BWR Performance Monitoring

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Technical Report
BWR Performance Monitoring

1003600

Final Report, November 2002

EPRI Project Manager

P. Frattini

EPRI • 3412 Hillview Avenue, Palo Alto, California 94304 • PO Box 10412, Palo Alto, California 94303 • USA
800.313.3774 • 650.855.2121 • askepri@epri.com • www.epri.com
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REPORT SUMMARY

This report presents the status of chemistry control programs and issues impacting performance
of North American boiling water reactors (BWRs), as of September 2002. The review of actual
BWR plant performance, experiences and practices surveyed in the report will make it a valuable
source document for the upcoming revision of the BWR Water Chemistry Guidelines, scheduled
for publication in 2004.

Background
BWR water chemistry has evolved from essentially pure, relatively oxidizing water to the current
programs which include hydrogen addition for IGSCC mitigation; NMCA to minimize steam
line dose rates; and DZO addition to minimize shutdown dose rates. EPRI standardized these
applications through Guidance such as previous EPRI report 1003022, BWRVIP-92 BWR Vessel
and Internals Project NMCA Experience Report and Applications Guidelines.

Objectives
To compile and summarize the chemistry and radiation dose control data available for the 36
operating BWR plants in North America; and to create a database to serve as the source
document for revision of the BWR Water Chemistry Guidelines.

Approach
EPRI approached the 36 operating BWRs with a request for information regarding data on
chemistry, plant design, iron and copper data and control techniques, chemistry control methods,
and radiation dose. The project team also asked the plants if and to what extent they were
conforming to the 2000 Revision 2 of the BWR Water Chemistry Guidelines (TR-103515-R2).
EPRI issued an interim report for this study in 2001 (1003157), focusing on the material relating
to iron performance monitoring, (Iron Performance Monitoring: Database Maintenance and
Evaluation of BWR Field Experience with New Technology).

Results
Most BWR plants inject hydrogen into the feedwater to mitigate intergranular stress corrosion
cracking (IGSCC). Many of these plants also apply noble metal chemical addition (NMCA) to
minimize the concentration of hydrogen required for IGSCC mitigation, and to minimize
increases in main steam line dose rates. In addition, most plants add zinc depleted in the 64Zn
isotope (DZO addition) for shutdown radiation field minimization.
Iron and copper transport in the reactor coolant are key elements of BWR water chemistry.
Control of these elements in the feedwater is highly dependent on the type of condensate
polishing system in use. Consequently, this report examines the performance of deep-bed
demineralizers, precoat filter/demins, and prefilter/deep bed combinations. The latter type of

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condensate polishing system affords the best removal of both particulate and soluble materials,
combining the excellent solids removal of the prefilters with superior ion exchange performance
of the deep beds, eliminating the need to clean the deep bed resins. In view of the increasing use
of filtration media for BWR condensate polishing, a review of industry practice and experience
with precoated and non-precoated media is also included.
The chemistry and radiation buildup database presented in this report has significant value to
plant operators for planning process optimization and benchmarking. EPRI will also produce a
series of periodic electronic reports to update and disseminate the most current water chemistry
information to EPRI members. These reports include ‘BRAC Summary Report,’ incorporating
standardized shutdown dose rate data; ‘BWR Chemistry Sampling Frequency Report,’ surveying
which and with what frequency plants measure the various chemistry parameters; ‘BWR Resin
Usage Report,’ detailing the consumption of ion exchange media at the individual plants; and
‘Condensate, Feedwater, and Reactor Water Chemistry Report,’ providing a 12-month rolling
average of feedwater metals, hydrogen, and anions.

EPRI Perspective
Successful operation of a nuclear plant demands careful control of water chemistry, particularly
in BWRs, where control of iron and copper in the reactor coolant is essential. Optimum
operation of the condensate polishers is the key to controlling these and other metal species. This
document compiles recent BWR plant experience with various chemistries, and reviews the
comparative performance and properties of the three types of BWR condensate polishing
systems. The document forms the most comprehensive database available for BWR water
chemistry operations and radiation trends. As such, it will be the principal source document for
revision of the BWR Water Chemistry Guidelines, NMCA Guidelines, and for developing plant
operational strategies.

Keywords
Boiling water reactor
BWR
Water chemistry
Shutdown radiation
Condensate polishing

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ACKNOWLEDGMENTS

The authors would like to acknowledge the following BWR power plant and corporate
chemistry, engineering, and radiation protection staff members who contributed their time to
provide the performance and operating experience data documented in this report:
Company Plant Contributor
AmerGen Energy Clinton Art Daniels
AmerGen Energy Clinton Tom Gagnon
AmerGen Energy Clinton Kendal Kett
AmerGen Energy Clinton John Wilson
AmerGen Energy Oyster Creek Mike Ford
AmerGen Energy Oyster Creek A. J. Judson

Carolina Power & Light Brunswick Joan Bozeman

Carolina Power & Light Brunswick Jeff Ferguson

Comision Federal de Electricidad Laguna Verde Narciso Beyrut

Comision Federal de Electricidad Laguna Verde Alejandro Garcia M.
Comision Federal de Electricidad Laguna Verde Alfonso Ruiz

Constellation Nuclear Nine Mile Point Bill Aiken

Constellation Nuclear Nine Mile Point Jim Bach
Constellation Nuclear Nine Mile Point Dave Barcomb
Constellation Nuclear Nine Mile Point Mark Becker
Constellation Nuclear Nine Mile Point John Blasiak
Constellation Nuclear Nine Mile Point Jeff Gerber
Constellation Nuclear Nine Mile Point Mike Goldych
Constellation Nuclear Nine Mile Point Bruce Holloway
Constellation Nuclear Nine Mile Point Dave Kazyaka
Constellation Nuclear Nine Mile Point Larry McNeer
Constellation Nuclear Nine Mile Point Al Moisan
Constellation Nuclear Nine Mile Point John Richards
Constellation Nuclear Nine Mile Point Tony Salvagno
Constellation Nuclear Nine Mile Point Carl Senska
Constellation Nuclear Nine Mile Point Steve Sheahan
Constellation Nuclear Nine Mile Point Pete Thingvoll
Constellation Nuclear Nine Mile Point Mike Thornhill

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Detroit Edison Fermi 2 Dan Craine

Detroit Edison Fermi 2 Jim Czech
Detroit Edison Fermi 2 Greg Mulleavy
Detroit Edison Fermi 2 Bob Nearhoof
Detroit Edison Fermi 2 John Tokarski

Energy Northwest Columbia Dave Bennett

Energy Northwest Columbia Clay Madden
Energy Northwest Columbia Larry Mayne
Energy Northwest Columbia Larry Morrison

Entergy Nuclear Northeast Corporate Jeff Goldstein

Entergy Nuclear Northeast James A. FitzPatrick Bill Bock
Entergy Nuclear Northeast James A. FitzPatrick Crystal Boucher
Entergy Nuclear Northeast Pilgrim Larry Loomis
Entergy Nuclear Northeast Pilgrim Bill Mauro
Entergy Nuclear Northeast Pilgrim Paul McNulty
Entergy Nuclear Northeast Vermont Yankee Rick Gerdus
Entergy Nuclear Northeast Vermont Yankee Remi Morrissette

Entergy Operations, Inc. Grand Gulf Bruce Bryant

Entergy Operations, Inc. Grand Gulf Charlotte Freeman
Entergy Operations, Inc. Grand Gulf Mike Michalski
Entergy Operations, Inc. Grand Gulf Keith O’Neal
Entergy Operations, Inc. Grand Gulf Roger Tolbert
Entergy Operations, Inc. Grand Gulf E. G. Wright
Entergy Operations, Inc. River Bend Dean Burnett
Entergy Operations, Inc. River Bend Genia Woodcox

Exelon Corp. Corporate Harry Miller

Exelon Corp. Corporate Dan Malauskas
Exelon Corp. Dresden Tim Fisk
Exelon Corp. Dresden Lou Magers
Exelon Corp. LaSalle Mike Pearson
Exelon Corp. LaSalle Ed Wolfe
Exelon Corp. Limerick Dave Barron
Exelon Corp. Limerick Willy Harris
Exelon Corp. Limerick Marino Kaminski
Exelon Corp. Limerick Marjorie McLaughlin
Exelon Corp. Limerick Dave Ryan
Exelon Corp. Peach Bottom Art Arcilla
Exelon Corp. Peach Bottom Darrel Chase
Exelon Corp. Peach Bottom Dan Droddy
Exelon Corp. Peach Bottom Bob Scholz
Exelon Corp. Quad Cities Paul Behrens
Exelon Corp. Quad Cities Erryl Mendenhall

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Exelon Corp. Quad Cities Ken Ohr

Exelon Corp. Quad Cities Rod Wiebenga

FirstEnergy Corp. Perry Michael Doty

FirstEnergy Corp. Perry Don Forbush
FirstEnergy Corp. Perry John Grimm
FirstEnergy Corp. Perry Rob Leib
FirstEnergy Corp. Perry Ron Wolf

Kernkraftwerk Leibstadt Leibstadt Daniel Brack

Kernkraftwerk Leibstadt Leibstadt Wilfried Kaufmann

Nebraska Public Power Cooper Gary Bray

Nebraska Public Power Cooper Kimberly Perry
Nebraska Public Poer Cooper Joseph Strahan
Nebraska Public Power Cooper Cindy Weers

Nuclear Management Co. Duane Arnold Steve Funk

Nuclear Management Co. Duane Arnold Wendell Keith
Nuclear Management Co. Duane Arnold Louis Kriege
Nuclear Management Co. Duane Arnold Kent Levan
Nuclear Management Co. Monticello Joe Gitzen
Nuclear Management Co. Monticello Mark Holmes
Nuclear Management Co. Monticello Kevin Jepson
Nuclear Management Co. Monticello Glenn Mathiasen
Nuclear Management Co. Monticello John Peterson

PPL Susquehanna Corporate Dave Morgan

PPL Susquehanna Corporate John Pacer
PPL Susquehanna Susquehanna Tim Ball
PPL Susquehanna Susquehanna Ray Doebler
PPL Susquehanna Susquehanna Richard Doty
PPL Susquehanna Susquehanna Leonard Humpf
PPL Susquehanna Susquehanna Dan Miller
PPL Susquehanna Susquehanna Patrick Treier
PPL Susquehanna Susquehanna Bruce Rhoades

PSE&G Hope Creek Mike Fagan

PSE&G Hope Creek Mark Meltzer
PSE&G Hope Creek Andy Narayan

Southern Nuclear Corporate Dennis Rickertsen

Southern Nuclear Hatch Brian Arnold
Southern Nuclear Hatch William Conant
Southern Nuclear Hatch Randy Jones
Southern Nuclear Hatch Al Manning

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Tennessee Valley Authority Browns Ferry Mike Campbell

Tennessee Valley Authority Browns Ferry Rob Coleman
Tennessee Valley Authority Browns Ferry Jeff Fenton
Tennessee Valley Authority Browns Ferry Bert Huie
Tennessee Valley Authority Browns Ferry Keith Nesmith
Tennessee Valley Authority Browns Ferry Conrad Ottenfeld
Tennessee Valley Authority Browns Ferry Mike Scarboro

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CONTENTS

1 INTRODUCTION ....................................................................................................................1-1
EPRI BWR Chemistry Monitoring Database .........................................................................1-1
Electronic Reports .................................................................................................................1-3
BRAC Summary Report ...................................................................................................1-3
BWR Chemistry Sampling Frequency Report ..................................................................1-3
BWR Resin Usage Reports ..............................................................................................1-4
Condensate, Feedwater, and Reactor Water Chemistry Report ......................................1-4
Individual Plant and Utility Assistance...................................................................................1-4
Final Report...........................................................................................................................1-5
References ............................................................................................................................1-6

2 PLANT CHEMISTRY REGIME STATUS ...............................................................................2-1

Hydrogen Water Chemistry (HWC) .......................................................................................2-1
Noble Metals Chemical Application (NMCA).........................................................................2-1
Zinc Addition .........................................................................................................................2-2
References ............................................................................................................................2-2

3 IRON AND COPPER CONTROL ...........................................................................................3-1

Plant Design Factors Impacting Feedwater Iron and Copper Control...................................3-1
Feedwater Iron Control Status ..............................................................................................3-4
BWR Industry Iron Control Performance ..........................................................................3-4
Impact of Drains Path on Iron Control ..............................................................................3-8
Iron Injection ...................................................................................................................3-10
Hotwell Iron .........................................................................................................................3-12
Feedwater Copper Control Status.......................................................................................3-14
Seasonal Effects .................................................................................................................3-20
Low Iron and Localized Fuel Corrosion...............................................................................3-20
Transient Iron and Copper ..................................................................................................3-24

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Perry Jet Pump Fouling.......................................................................................................3-25

References .....................................................................................................................3-27

4 TRANSIENT CORROSION PRODUCTS ...............................................................................4-1

Introduction ...........................................................................................................................4-1
Approach ...............................................................................................................................4-2
Summary of EPRI Guidelines on Feedwater and Reactor Water Metals..............................4-2
Background and Literature Review .......................................................................................4-3
History of BWR Fuel Issues..............................................................................................4-3
River Bend Experience.....................................................................................................4-4
Duane Arnold (DAEC) Post-NMCA Fuel Monitoring ........................................................4-5
Industry Updates on Post-NMCA Fuel Monitoring............................................................4-6
Fuel Surveillance Results at Hatch prior to NMCA...........................................................4-7
Hatch Feedwater Chemistry Trends.................................................................................4-7
Hatch Unit 1 .................................................................................................................4-7
Hatch Unit 2 .................................................................................................................4-9
Hatch Fuel Inspection and Deposition Data ...................................................................4-10
Swedish Experience .......................................................................................................4-16
Correlated Causes of CILC-Related Fuel Failures .........................................................4-16
Summary and Mechanisms of Past Crud-Related Fuel Failures....................................4-17
Iron and Copper Data Evaluation ........................................................................................4-23
Steady State Loading Calculations.................................................................................4-25
Results and Observations ..............................................................................................4-26
Plant Startup Data ..........................................................................................................4-28
River Bend .................................................................................................................4-29
Columbia....................................................................................................................4-32
Conclusions from Transient Metals Evaluation ...................................................................4-35
Recommendations ..............................................................................................................4-37
References ..........................................................................................................................4-38

5 DRYWELL RADIATION DOSE RATES .................................................................................5-1

Monitoring of Drywell Radiation Dose Rates.........................................................................5-1
BRAC Monitoring Program and Practices ........................................................................5-1
BRAC Monitoring Results.................................................................................................5-2
Drywell Dose Rate Correlations ............................................................................................5-6

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Hot Spots ............................................................................................................................5-17

Impact of Feedwater Iron on Co-60.....................................................................................5-18
Impact of Feedwater Iron on Soluble Co-60...................................................................5-18
Impact of Feedwater Iron on Insoluble Co-60 ................................................................5-21
Current Iron Control Guidelines......................................................................................5-24
Hatch 1 Low Iron ............................................................................................................5-25
References ..........................................................................................................................5-31

6 PLANT CHEMISTRY IMPACT OF NMCA .............................................................................6-1

Noble Metal Loading .............................................................................................................6-1
NMCA Guidelines..................................................................................................................6-5
Summary of Plant Chemistry Responses..............................................................................6-6
Feedwater Dissolved Hydrogen .......................................................................................6-6
MSLRM Response .........................................................................................................6-10
Reactor Water Conductivity Response...........................................................................6-12
Offgas Sum of Six Activity Response .............................................................................6-15
Co-60 Response.............................................................................................................6-18
Iodine Response.............................................................................................................6-22
Plants Experiencing a Decrease in Drywell Shutdown Dose Rates after NMCA ................6-26
Plants Experiencing an Increase in Drywell Shutdown Dose Rates after NMCA................6-28
Summary of Recommendations with NMCA .......................................................................6-30
NMCA Impact on Nine Mile Point 2 RWCU Filter Demineralizer (F/D) Performance..........6-30
Issue Description ............................................................................................................6-30
Nine Mile Point 2 RWCU Conditions ..............................................................................6-31
Nine Mile Point 2 Data Evaluation (3) .................................................................................6-31
Conclusions on Nine Mile Point 2 RWCU F/D Performance ..........................................6-36
Similar Experience at a Japanese BWR.........................................................................6-42
References ..........................................................................................................................6-42

7 PLANT CHEMISTRY RESULTS VS. BWR WATER CHEMISTRY GUIDELINES

REVISIONS ...............................................................................................................................7-1
Feedwater Dissolved Oxygen Control...................................................................................7-1
Implementation of Revised Action Levels for HWC and HWC/NMCA ..................................7-3
Monitoring to Assure IGSCC Protection................................................................................7-4
“Startup/Hot Standby” versus “Power Operations” Action Levels .........................................7-7
Startup and Shutdown Iron Control .......................................................................................7-7

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Feedwater Flush Practices ...............................................................................................7-7

Impact of New Startup/Shutdown Iron Control Guidelines .............................................7-10
Shutdown Duration for Application of New Startup/Shutdown Iron Control
Guidelines.......................................................................................................................7-10
Startup/Shutdown Copper Monitoring ............................................................................7-10
Startup/Shutdown Iron Analysis Methods.......................................................................7-10
Startup/Shutdown Feedwater Flush Limits.....................................................................7-10
Deep Bed Only Plant Practices ......................................................................................7-11
Filter + Deep Bed Plant Practices...................................................................................7-12
Filter/Demineralizer Plant Practices ...............................................................................7-12
Startup/Shutdown Summary...........................................................................................7-12
References ..........................................................................................................................7-12

8 BWR CHEMISTRY MONITORING PRACTICES ...................................................................8-1

Sampling Frequencies...........................................................................................................8-1
Reactor Water Chloride and Sulfate.................................................................................8-1
Reactor Water Cobalt-60 & Zinc-65 .................................................................................8-2
Reactor Water Iron ...........................................................................................................8-4
Reactor Water Zinc...........................................................................................................8-4
Feedwater Iron and Copper..............................................................................................8-5
Feedwater Zinc.................................................................................................................8-5
Zinc Sampling and Analysis Methodologies..........................................................................8-7
Injection Method ...............................................................................................................8-7
Analysis Method ...............................................................................................................8-7
Other Sampling Issues ........................................................................................................8-11
Reactor Water Gamma Isotopics ...................................................................................8-11
Feedwater Sampling.......................................................................................................8-12
References ..........................................................................................................................8-17

9 OVERVIEW OF FIELD EXPERIENCE WITH CONDENSATE FILTERS...............................9-1

The Evolution and Design of Condensate Filters ..................................................................9-1
Application Challenge Severity Indexes ................................................................................9-4
RLI (Filter Run Length Index) ...........................................................................................9-5
IXI (Filter Precoat Ion Exchange Index)............................................................................9-5
DSI (Deep Bed Sulfate Index) ..........................................................................................9-6
Growth of Iron Removal Filter Septa Use..............................................................................9-7

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Precoat Applications............................................................................................................9-10
Non-Precoat Applications....................................................................................................9-14
Deep Bed Demineralizers – Potential for Condensate Filter Addition.................................9-17
Major Issues and Their Current Status................................................................................9-19
Full Scale Application Status ..........................................................................................9-19
Effluent Iron Concentrations ...........................................................................................9-20
Mechanical Integrity........................................................................................................9-20
Septa Useful Life ............................................................................................................9-20
Backwash Methods ........................................................................................................9-21
Ion Exchange Performance ............................................................................................9-23
Handling Backwash Waste Liquids ................................................................................9-24
Peach Bottom Trial of Melt-Blown Septa........................................................................9-24
Continuing Issues ...........................................................................................................9-25
References ..........................................................................................................................9-25

10 SUMMARY OF EXPERIENCE WITH DEEP BED CONDENSATE

DEMINERALIZERS .................................................................................................................10-1
Deep Bed Only Condensate Polishing ................................................................................10-1
Resin Selection...............................................................................................................10-2
Inlet Iron Concentration ..................................................................................................10-6
Effect of Condensate Temperature ................................................................................10-8
Resin Cleaning Frequency .............................................................................................10-8
Flow Transients ............................................................................................................10-10
Resin Cleaning Effectiveness.......................................................................................10-10
Resin Age .....................................................................................................................10-11
Cation/Anion Resin Ratio .............................................................................................10-12
Vessel Area Flow Rate .................................................................................................10-12
Filter + Deep Bed Condensate Polishing ..........................................................................10-13
Maintenance of Condensate Polishers and Support Systems ..........................................10-15
References ........................................................................................................................10-16

11 CONDENSATE POLISHING WITH ION EXCHANGER BYPASS.....................................11-1

Bypass Operation................................................................................................................11-1
Brunswick Experience ....................................................................................................11-2
Dresden Experience .......................................................................................................11-6
KKL Experience..............................................................................................................11-7

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Summary ...........................................................................................................................11-10
References ........................................................................................................................11-11

12 OTHER ISSUES FOR CONSIDERATION IN THE NEXT REVISION OF THE BWR

CHEMISTRY GUIDELINES.....................................................................................................12-1
References ........................................................................................................................12-11

13 CONCLUSIONS .................................................................................................................13-1
BWR Iron/Chemistry Monitoring Database .........................................................................13-1
Plant Chemistry Regime Status ..........................................................................................13-1
Iron and Copper Control......................................................................................................13-1
Transient Corrosion Products..............................................................................................13-2
Drywell Radiation Dose Rates.............................................................................................13-4
Plant Chemistry Impact of NMCA........................................................................................13-5
Plant Chemistry Results vs. BWR Chemistry Guidelines Revisions ...................................13-6
BWR Chemistry Monitoring Practices .................................................................................13-7
Experience with Condensate Filters....................................................................................13-7
Experience with Deep Bed Condensate Demineralizers.....................................................13-8
Condensate Polishing With Ion Exchanger Bypass ............................................................13-9
References ..........................................................................................................................13-9

A BROWNS FERRY 2.............................................................................................................. A-1

Browns Ferry 2 Milestones................................................................................................... A-1
Radiation Data ..................................................................................................................... A-3
Trend Data ........................................................................................................................... A-4
Feedwater Iron Control......................................................................................................... A-4
Reactor Water Sulfate Control ............................................................................................. A-4
Recirculation Piping Dose Rates.......................................................................................... A-7
Fuel Failures ........................................................................................................................ A-7

B BROWNS FERRY 3.............................................................................................................. B-1

Browns Ferry 3 Milestones................................................................................................... B-1
Radiation Data ..................................................................................................................... B-3
Trend Data ........................................................................................................................... B-3
Feedwater Iron Control......................................................................................................... B-4
Reactor Water Sulfate Control ............................................................................................. B-4

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Recirculation Piping Dose Rates.......................................................................................... B-7

C BRUNSWICK 1 ..................................................................................................................... C-1

Brunswick 1 Milestones........................................................................................................ C-1
Radiation Data ..................................................................................................................... C-3
Trend Data ........................................................................................................................... C-3
Feedwater Iron Control......................................................................................................... C-6
Recirculation Piping Dose Rates.......................................................................................... C-6
Recirculation Piping Gamma Scans..................................................................................... C-6
Fuel Failures ........................................................................................................................ C-7

D BRUNSWICK 2 ..................................................................................................................... D-1

Brunswick 2 Milestones........................................................................................................ D-1
Radiation Data ..................................................................................................................... D-1
Trend Data ........................................................................................................................... D-1
Feedwater Iron Control......................................................................................................... D-6
Recirculation Piping Dose Rates.......................................................................................... D-6
Recirculation Piping Gamma Scans..................................................................................... D-6
Stellite Reduction .............................................................................................................. D-7

E CLINTON............................................................................................................................... E-1
Clinton Milestones ................................................................................................................ E-1
Radiation Data ..................................................................................................................... E-1
Trend Data ........................................................................................................................... E-3
Feedwater Iron Control......................................................................................................... E-3
Recirculation Piping Dose Rates.......................................................................................... E-6
Recirculation Piping Gamma Scans..................................................................................... E-6

F COOPER NUCLEAR..............................................................................................................F-1
Cooper Milestones ................................................................................................................F-1
Radiation Data ......................................................................................................................F-2
Cooper – Recirculation System Dose Rates (mR/hr)............................................................F-3
Trend Data ............................................................................................................................F-3
Feedwater Iron Control..........................................................................................................F-5
Recirculation Piping Dose Rates...........................................................................................F-5
Recirculation Piping Gamma Scans......................................................................................F-5

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G DRESDEN 2.......................................................................................................................... G-1

Dresden 2 Milestones........................................................................................................... G-1
Radiation Data ..................................................................................................................... G-3
Trend Data ........................................................................................................................... G-3
Feedwater Iron Control......................................................................................................... G-6
Recirculation Piping Dose Rates.......................................................................................... G-6

H DRESDEN 3.......................................................................................................................... H-1

Dresden 3 Milestones........................................................................................................... H-1
Radiation Data ..................................................................................................................... H-2
Trend Data ........................................................................................................................... H-3
Feedwater Iron Control......................................................................................................... H-3
Recirculation Piping Dose Rates.......................................................................................... H-3

I DUANE ARNOLD.....................................................................................................................I-1
Duane Arnold Milestones .......................................................................................................I-1
Radiation Data .......................................................................................................................I-1
Trend Data .............................................................................................................................I-1
Feedwater Iron Control...........................................................................................................I-6
Recirculation Piping Dose Rates............................................................................................I-6
Recirculation Piping Gamma Scans.......................................................................................I-6

J FERMI 2..................................................................................................................................J-1
Fermi 2 Milestones ................................................................................................................ J-1
Radiation Data ...................................................................................................................... J-1
Trend Data ............................................................................................................................ J-3
Feedwater Iron Control.......................................................................................................... J-3
Recirculation Piping Dose Rates........................................................................................... J-6
Stellite™ Reduction............................................................................................................... J-6

K FITZPATRICK....................................................................................................................... K-1
FitzPatrick Milestones .......................................................................................................... K-1
Radiation Data ..................................................................................................................... K-2
Trend Data ........................................................................................................................... K-4
Feedwater Iron Control......................................................................................................... K-4
Recirculation Piping Dose Rates.......................................................................................... K-6

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Recirculation Piping Gamma Scan....................................................................................... K-6

Stellite™ Reduction.............................................................................................................. K-8

L GRAND GULF........................................................................................................................L-1
Grand Gulf Milestones...........................................................................................................L-1
Radiation Data ......................................................................................................................L-3
Trend Data ............................................................................................................................L-3
Feedwater Iron Control..........................................................................................................L-4
Recirculation Piping Dose Rates...........................................................................................L-6
Recirculation Piping Gamma Scans......................................................................................L-6

M HATCH 1 ..............................................................................................................................M-1
Hatch 1 Milestones...............................................................................................................M-1
Radiation Data .....................................................................................................................M-3
Trend Data ...........................................................................................................................M-4
Feedwater Iron Control.........................................................................................................M-4
Recirculation Piping Dose Rates..........................................................................................M-6
Recirculation Piping Gamma Scans.....................................................................................M-6
Stellite™ Reduction..............................................................................................................M-7

N HATCH 2............................................................................................................................... N-1

Hatch 2 Milestones............................................................................................................... N-1
Radiation Data ..................................................................................................................... N-3
Trend Data ........................................................................................................................... N-4
Feedwater Iron Control......................................................................................................... N-4
Recirculation Piping Dose Rates.......................................................................................... N-6
Recirculation Piping Gamma Scans..................................................................................... N-7
Stellite™ Reduction.............................................................................................................. N-7

O HOPE CREEK ...................................................................................................................... O-1

Hope Creek Milestones ........................................................................................................ O-1
Radiation Data ..................................................................................................................... O-2
Trend Data ........................................................................................................................... O-3
Feedwater Iron Control......................................................................................................... O-5
Recirculation Piping Dose Rates.......................................................................................... O-5
Recirculation Piping Gamma Scans..................................................................................... O-6

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Fuel Failures ........................................................................................................................ O-6

P LAGUNA VERDE 1............................................................................................................... P-1

Laguna Verde 1 Milestones.................................................................................................. P-1
Radiation Data ..................................................................................................................... P-2
Trend Data ........................................................................................................................... P-3
Feedwater Iron Control......................................................................................................... P-4
Recirculation Piping Dose Rates.......................................................................................... P-6

Q LAGUNA VERDE 2 .............................................................................................................. Q-1

Laguna Verde 2 Milestones.................................................................................................. Q-1
Radiation Data ..................................................................................................................... Q-3
Trend Data ........................................................................................................................... Q-3
Feedwater Iron Control......................................................................................................... Q-4
Recirculation Piping Dose Rates.......................................................................................... Q-6

R LASALLE 1........................................................................................................................... R-1

LaSalle 1 Milestones ............................................................................................................ R-1
Radiation Data ..................................................................................................................... R-3
Trend Data ........................................................................................................................... R-4
Feedwater Iron Control......................................................................................................... R-4
Recirculation Piping Dose Rates.......................................................................................... R-6

S LASALLE 2 ........................................................................................................................... S-1

LaSalle 2 Milestones ............................................................................................................ S-1
Radiation Data ..................................................................................................................... S-3
Trend Data ........................................................................................................................... S-3
Feedwater Iron Control......................................................................................................... S-3
Recirculation Piping Dose Rates.......................................................................................... S-5
Recirculation Piping Gamma Scans..................................................................................... S-6

T LIMERICK 1 ...........................................................................................................................T-1
Limerick 1 Milestones............................................................................................................T-1
Radiation Data ......................................................................................................................T-3
Trend Data ............................................................................................................................T-3
Feedwater Iron Control..........................................................................................................T-4

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Recirculation Piping Dose Rates...........................................................................................T-6

Recirculation Piping Gamma Scans......................................................................................T-6

U LIMERICK 2 .......................................................................................................................... U-1

Limerick 2 Milestones........................................................................................................... U-1
Radiation Data ..................................................................................................................... U-3
Trend Data ........................................................................................................................... U-3
Feedwater Iron Control......................................................................................................... U-4
Recirculation Piping Dose Rates.......................................................................................... U-6
Recirculation Piping Gamma Scans..................................................................................... U-6

V MONTICELLO....................................................................................................................... V-1
Monticello Milestones ........................................................................................................... V-1
Radiation Data ..................................................................................................................... V-2
Trend Data ........................................................................................................................... V-4
Feedwater Iron Control......................................................................................................... V-4
Reactor Water Sulfate Control ............................................................................................. V-5
Recirculation Piping Dose Rates.......................................................................................... V-7
Recirculation Piping Gamma Scans..................................................................................... V-7
Stellite™ Reduction.............................................................................................................. V-8

W NINE MILE POINT 1 ........................................................................................................... W-1

Nine Mile Point Milestones .................................................................................................. W-1
Radiation Data .................................................................................................................... W-2
Trend Data .......................................................................................................................... W-4
Feedwater Iron Control........................................................................................................ W-4
Reactor Water Sulfate Control ............................................................................................ W-4
Recirculation Piping Dose Rates......................................................................................... W-7
Recirculation Piping Gamma Scans Pipe............................................................................ W-7

X NINE MILE POINT 2 ............................................................................................................. X-1

Nine Mile Point 2 Milestones ................................................................................................ X-1
Radiation Data ..................................................................................................................... X-2
Trend Data ........................................................................................................................... X-3
Feedwater Iron Control......................................................................................................... X-4
Reactor Water Sulfate Control ............................................................................................. X-4

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Recirculation Piping Dose Rates.......................................................................................... X-7

Recirculation Piping Gamma Scans..................................................................................... X-7
Stellite™ Reduction.............................................................................................................. X-8

Y OYSTER CREEK .................................................................................................................. Y-1

Oyster Creek Milestones ...................................................................................................... Y-1
Radiation Data ..................................................................................................................... Y-2
Trend Data ........................................................................................................................... Y-4
Feedwater Iron Control......................................................................................................... Y-4
Recirculation Piping Dose Rates.......................................................................................... Y-7
Stellite™ Reduction.............................................................................................................. Y-7

Z PEACH BOTTOM 2................................................................................................................Z-1

Peach Bottom 2 Milestones...................................................................................................Z-1
Radiation Data ......................................................................................................................Z-2
Trend Data ............................................................................................................................Z-3
Feedwater Iron Control..........................................................................................................Z-5
Recirculation Piping Dose Rates...........................................................................................Z-7
Recirculation Piping Gamma Scans......................................................................................Z-8

AA PEACH BOTTOM 3..........................................................................................................AA-1

Peach Bottom 3 Milestones................................................................................................ AA-1
Radiation Data ................................................................................................................... AA-2
Trend Data ......................................................................................................................... AA-3
Feedwater Iron Control....................................................................................................... AA-4
Recirculation Piping Dose Rates........................................................................................ AA-6
Recirculation Piping Gamma Scans................................................................................... AA-6

BB PERRY .............................................................................................................................BB-1
Perry Milestones ................................................................................................................ BB-1
Radiation Data ................................................................................................................... BB-3
Trend Data ......................................................................................................................... BB-3
Feedwater Iron Control....................................................................................................... BB-4
Recirculation Piping Dose Rates........................................................................................ BB-6
Recirculation Piping Gamma Scans................................................................................... BB-6
Stellite™ Reduction............................................................................................................ BB-7

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CC PILGRIM...........................................................................................................................CC-1
Pilgrim Milestones ..............................................................................................................CC-1
Radiation Data ...................................................................................................................CC-2
Trend Data .........................................................................................................................CC-3
Feedwater Iron Control.......................................................................................................CC-3
Recirculation Piping Dose Rates........................................................................................CC-6
Recirculation Piping Gamma Scans...................................................................................CC-6
Stellite™ Reduction............................................................................................................CC-6

DD QUAD CITIES 1................................................................................................................DD-1

Quad Cities 1 Milestones ...................................................................................................DD-1
Radiation Data ...................................................................................................................DD-1
Trend Data .........................................................................................................................DD-4
Feedwater Iron Control.......................................................................................................DD-7
Recirculation Piping Dose Rates........................................................................................DD-7

EE QUAD CITIES 2 ................................................................................................................ EE-1

Quad Cities 2 Milestones ................................................................................................... EE-1
Radiation Data ................................................................................................................... EE-1
Trend Data ......................................................................................................................... EE-3
Feedwater Iron Control....................................................................................................... EE-6
Recirculation Piping Dose Rates........................................................................................ EE-6

FF RIVER BEND .................................................................................................................... FF-1

River Bend Milestones ....................................................................................................... FF-1
Radiation Data ................................................................................................................... FF-3
Trend Data ......................................................................................................................... FF-4
Feedwater Iron Control....................................................................................................... FF-4
Recirculation Piping Dose Rates........................................................................................ FF-6
Stellite™ Reduction............................................................................................................ FF-7
Fuel Failures ...................................................................................................................... FF-7

GG SUSQUEHANNA 1 ......................................................................................................... GG-1

Susquehanna 1 Milestones ............................................................................................... GG-1
Radiation Data .................................................................................................................. GG-3
Trend Data ........................................................................................................................ GG-3

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Feedwater Iron Control...................................................................................................... GG-4

Recirculation Piping Dose Rates....................................................................................... GG-6
Recirculation Piping Gamma Scans.................................................................................. GG-6
Stellite™ Reduction........................................................................................................... GG-8

HH SUSQUEHANNA 2...........................................................................................................HH-1
Susquehanna 2 Milestones ................................................................................................HH-1
Radiation Data ...................................................................................................................HH-2
Trend Data .........................................................................................................................HH-3
Feedwater Iron Control.......................................................................................................HH-4
Recirculation Piping Dose Rates........................................................................................HH-6
Recirculation Piping Gamma Scans...................................................................................HH-6
Stellite™ Reduction............................................................................................................HH-7

II VERMONT YANKEE ..............................................................................................................II-1

Vermont Yankee Milestones .................................................................................................II-1
Radiation Data ......................................................................................................................II-3
Trend Data ............................................................................................................................II-3
Feedwater Iron Control..........................................................................................................II-4
Recirculation Piping Dose Rates...........................................................................................II-6
Recirculation Piping Gamma Scans......................................................................................II-7
Stellite™ Reduction...............................................................................................................II-7
Fuel Failures .........................................................................................................................II-7

JJ COLUMBIA (FORMERLY WNP2)......................................................................................JJ-1

Columbia Milestones ........................................................................................................... JJ-1
Radiation Data .................................................................................................................... JJ-3
Trend Data .......................................................................................................................... JJ-3
Feedwater Iron Control........................................................................................................ JJ-4
Recirculation Piping Dose Rates......................................................................................... JJ-6
Recirculation Piping Gamma Scans.................................................................................... JJ-6

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LIST OF FIGURES

Figure 1-1 Structure of EPRI BWR Chemistry Monitoring Database .........................................1-2

Figure 3-1 Design Evolution of North American BWRs .............................................................3-3
Figure 3-2 1997-2001 Average Feedwater Iron for North American BWRs...............................3-5
Figure 3-3 2001 Average Feedwater Iron for North American BWRs........................................3-5
Figure 3-4 2001 Average Feedwater Iron vs. Condensate Polishing Type ...............................3-7
Figure 3-5 1997 – 2001 Trend in Average Feedwater Iron (North American BWRs) ................3-7
Figure 3-6 Iron Concentrations in Forward Pumped Drains ......................................................3-9
Figure 3-7 2001 Average Feedwater Iron vs. Drains Path.........................................................3-9
Figure 3-8 Susquehanna Iron Oxalate Injection Skid Schematic P&I Diagram .......................3-11
Figure 3-9 1998 – 2000 Average Hotwell Iron Concentration ..................................................3-13
Figure 3-10 2001 Average Hotwell Iron Concentration............................................................3-13
Figure 3-11 1998 - 2001 Annual Average Hotwell Iron Concentration ....................................3-14
Figure 3-12 2000 Average Feedwater Copper for BWRs with Copper Source in Main
Condenser........................................................................................................................3-15
Figure 3-13 2001 Average Feedwater Copper for BWRs Differentiated by Condensate
Polisher Type and Main Condenser Tube Material ..........................................................3-16
Figure 3-14 1999-2001 Average Feedwater Copper Differentiated by Condensate
Polisher Type (for plants with copper alloy condenser tubes)..........................................3-18
Figure 3-15 Average Feedwater Iron by Calendar Quarter .....................................................3-20
Figure 3-16 Reactor Water Soluble Metals Ratio ....................................................................3-21
Figure 3-17 Reactor Water Total Metals Ratio ........................................................................3-22
Figure 3-18 Feedwater Soluble Metals Ratio...........................................................................3-23
Figure 3-19 Feedwater Total Metals Ratio...............................................................................3-23
Figure 3-20 A Loop Drive Flow Difference From Expected Based On FCV Position...............3-25
Figure 3-21 B Loop Drive Flow Difference From Expected Based On FCV Position...............3-26
Figure 4-1 Duane Arnold Fuel Deposition Loading (Sum of Brushing and Scraping)................4-6
Figure 4-2 Hatch 1 Feedwater Metals Trend .............................................................................4-9
Figure 4-3 Hatch 2 Feedwater Metals Trend ...........................................................................4-10
Figure 4-4 Hatch 1 Single Cycle Exposure Fuel Deposition ....................................................4-13
Figure 4-5 Hatch 1 Multi-Cycle Exposure Fuel Deposition ......................................................4-14
Figure 4-6 Hatch 2 Single Cycle Exposure Fuel Deposition ....................................................4-15
Figure 4-7 Hatch 2 Multi-Cycle Exposure Fuel Deposition ......................................................4-15

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Figure 4-8 Monticello Fuel Deposition Data after 32,830 EFPH ..............................................4-21
Figure 4-9 Nine Mile Point 1 Fuel Deposition Data after 26,000 EFPH ...................................4-21
Figure 4-10 River Bend April 2000 Startup Metals Data..........................................................4-30
Figure 4-11 River Bend April 2000 Startup Iron Deposition Trends.........................................4-31
Figure 4-12 River Bend April 2000 Startup Copper Deposition Trends ...................................4-31
Figure 4-13 Columbia July 2001 Startup Metals Data .............................................................4-33
Figure 4-14 Columbia July 2001 Startup Iron Deposition Trends ............................................4-34
Figure 4-15 Columbia July 2001 Startup Copper Deposition Trends ......................................4-35
Figure 5-1 Most Recent BRAC Average Dose Rates for North American BWRs
Designated by Chemistry Regime at Time of Measurement..............................................5-5
Figure 5-2 Frequency Distribution of Most Recent BRAC Average Dose Rates for North
American BWRs.................................................................................................................5-5
Figure 5-3 BRAC and Average Soluble Co-60 ..........................................................................5-6
Figure 5-4 Most Recently Reported BRAC Dose Rate vs. Average Soluble Co-60 ..................5-7
Figure 5-5 Most Recently Reported BRAC Dose Rate vs. Average Soluble Co-60 by
Chemistry Regime..............................................................................................................5-8
Figure 5-6 Most Recently Reported BRAC Dose Rate vs. Average Soluble Co-60 for
HWC+Zn Plants .................................................................................................................5-9
Figure 5-7 Average BRAC Dose Rate by Chemistry Regime (based on most recently
reported dose rates)...........................................................................................................5-9
Figure 5-8 BRAC Dose Rate One Cycle After Chemical Decontamination by Chemistry
Regime.............................................................................................................................5-10
Figure 5-9 Average Impact on BRAC Dose Rates of Transitions Between Chemistry
Regimes ...........................................................................................................................5-11
Figure 5-10 BRAC Dose Rate vs. Soluble Co-60 (points distinguished by transitions
between chemistry regimes) ............................................................................................5-12
Figure 5-11 BRAC Dose Rate vs. Reactor Water Zinc (points distinguished by transitions
between chemistry regimes) ............................................................................................5-12
Figure 5-12 Reactor Water Soluble Co-60 vs. Feedwater Zinc ...............................................5-13
Figure 5-13 Reactor Water Soluble Co-60 vs. Reactor Water Zinc .........................................5-14
Figure 5-14 BRAC Dose Rates vs. Ratio of Reactor Water Soluble Co-60 to Reactor
Water Soluble Zinc...........................................................................................................5-15
Figure 5-15 BRAC Dose Rates vs. Ratio of Reactor Water Soluble Co-60 to Reactor
Water Soluble Zinc by Chemistry Regime In Effect Before BRAC Measurement............5-15
Figure 5-16 BRAC Dose Rates vs. Ratio of Reactor Water Soluble Zinc to Feedwater
Total Iron ..........................................................................................................................5-16
Figure 5-17 BRAC Dose Rates vs. Ratio of Reactor Water Soluble Zinc to Feedwater
Total Iron by Chemistry Regime In Effect Before BRAC Measurement ...........................5-17
Figure 5-18 BHD Dose Rate vs. Insoluble Co-60 ....................................................................5-18
Figure 5-19 Reactor Water Soluble Co-60 vs. Feedwater Iron ................................................5-19
Figure 5-20 Brunswick 1 and 2 Reactor Water Soluble Co-60 vs. Feedwater Iron..................5-20
Figure 5-21 Hatch 1 Reactor Water Soluble Co-60 vs. Feedwater Iron ..................................5-21

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Figure 5-22 Brunswick Units 1 and 2 Reactor Water Insoluble Co-60 Response to
Feedwater Iron .................................................................................................................5-22
Figure 5-23 Limerick 1 Reactor Water Insoluble Co-60 Response to Feedwater Iron.............5-23
Figure 5-24 Hatch 1 Reactor Water Insoluble Co-60 Response to Feedwater Iron.................5-24
Figure 5-25 Hatch 1 Feedwater Iron Trend..............................................................................5-25
Figure 5-26 Hatch 1 2001 Iron Data ........................................................................................5-26
Figure 5-27 Hatch 1 Reactor Water Co-60 Trend Data ...........................................................5-27
Figure 5-28 Plants in Low Iron Range, 2001 Averages ...........................................................5-28
Figure 5-29 Hatch 1 Monthly Average Feedwater Iron and Reactor Water Soluble Co-60
at Steady Power...............................................................................................................5-28
Figure 5-30 Hatch 1 Monthly Average Feedwater Iron and Reactor Water Insoluble Co-
60 at Steady Power..........................................................................................................5-29
Figure 5-31 Hatch 1 Monthly Average Soluble Co-60 vs. Feedwater Iron at Steady
Power After Post-NMCA Transient Conditions ................................................................5-30
Figure 5-32 Hatch 1 Monthly Average Insoluble Co-60 vs. Feedwater Iron at Steady
Power After Post-NMCA Transient Conditions ................................................................5-30
Figure 6-1 Noble Metal Loading vs Coupon Exposure Time .....................................................6-4
Figure 6-2 Duane Arnold Feedwater Hydrogen Trend Data ......................................................6-7
Figure 6-3 Hatch 1 Feedwater Hydrogen Trend Data................................................................6-7
Figure 6-4 FitzPatrick Feedwater Hydrogen Trend Data ...........................................................6-8
Figure 6-5 Peach Bottom 2 Feedwater Hydrogen Trend Data ..................................................6-8
Figure 6-6 Quad Cities 1 Feedwater Hydrogen Trend Data ......................................................6-9
Figure 6-7 Nine Mile 1 Feedwater Hydrogen Trend Data ..........................................................6-9
Figure 6-8 FitzPatrick Main Steam Line Radiation Monitor Trend Data...................................6-10
Figure 6-9 Peach Bottom 2 Main Steam Line Radiation Monitor Trend Data ..........................6-11
Figure 6-10 Nine Mile Point 1 Main Steam Line Radiation Monitor Trend Data ......................6-11
Figure 6-11 Duane Arnold Reactor Water Conductivity Trend Data ........................................6-12
Figure 6-12 Hatch 1 Reactor Water Conductivity Trend Data .................................................6-13
Figure 6-13 FitzPatrick Reactor Water Conductivity Trend Data .............................................6-13
Figure 6-14 Peach Bottom 2 Reactor Water Conductivity Trend Data ....................................6-14
Figure 6-15 Quad Cities 1 Reactor Water Conductivity Trend Data ........................................6-14
Figure 6-16 Nine Mile Point 1 Reactor Water Conductivity Trend Data...................................6-15
Figure 6-17 Duane Arnold Sum of Six Noble Gases Trend Data ............................................6-16
Figure 6-18 Hatch 1 Sum of Six Noble Gases Trend Data ......................................................6-16
Figure 6-19 FitzPatrick Sum of Six Noble Gases Trend Data..................................................6-17
Figure 6-20 Peach Bottom 2 Sum of Six Noble Gases Trend Data .........................................6-17
Figure 6-21 Nine Mile Point 1 Sum of Six Noble Gases Trend Data .......................................6-18
Figure 6-22 Duane Arnold Co-60 Trend Data..........................................................................6-19
Figure 6-23 Hatch 1 Co-60 Trend Data ...................................................................................6-19
Figure 6-24 FitzPatrick Co-60 Trend Data ...............................................................................6-20

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Figure 6-25 Peach Bottom 2 Co-60 Trend Data ......................................................................6-20

Figure 6-26 Quad Cities 1 Co-60 Trend Data ..........................................................................6-21
Figure 6-27 Nine Mile Point 1 Co-60 Trend Data.....................................................................6-21
Figure 6-28 Duane Arnold Reactor Water Iodine Isotopic Trend Data ....................................6-23
Figure 6-29 Hatch 1 Reactor Water Iodine Isotopic Trend Data ..............................................6-23
Figure 6-30 FitzPatrick Reactor Water Iodine Isotopic Trend Data .........................................6-24
Figure 6-31 Peach Bottom 2 Reactor Water Iodine Isotopic Trend Data.................................6-24
Figure 6-32 Quad Cities 1 Reactor Water Iodine Isotopic Trend Data ....................................6-25
Figure 6-33 Nine Mile Point 1 Reactor Water Iodine Isotopic Trend Data ...............................6-25
Figure 6-34 Duane Arnold BRAC Dose Rates and Milestones................................................6-26
Figure 6-35 Hatch 1 BRAC Dose Rates and Milestones .........................................................6-27
Figure 6-36 FitzPatrick BRAC Dose Rates and Milestones .....................................................6-27
Figure 6-37 Hatch 2 BRAC Dose Rates and Milestones .........................................................6-28
Figure 6-38 Peach Bottom 2 BRAC Dose Rates and Milestones ............................................6-29
Figure 6-39 Quad Cities 1 BRAC Dose Rates and Milestones ................................................6-29
Figure 6-40 Nine Mile Point 1 BRAC Dose Rates and Milestones ..........................................6-30
Figure 6-41 Nine Mile Point 2 Reactor Water Trend Data .......................................................6-32
Figure 6-42 Nine Mile Point 2 RWCU F/D “A” Performance Data............................................6-33
Figure 6-43 Nine Mile Point 2 RWCU F/D “B” Performance Data............................................6-34
Figure 6-44 Nine Mile Point 2 RWCU F/D “C” Performance Data ...........................................6-34
Figure 6-45 Nine Mile Point 2 RWCU F/D “D” Performance Data ...........................................6-35
Figure 6-46 Pourbaix Diagram For Fe-H2O at 25ºC (5) ...........................................................6-41
Figure 7-1 BWR Feedwater Dissolved Oxygen Results at Power Operating Conditions ..........7-2
Figure 7-2 Feedwater Dissolved Oxygen Results versus <30 ppb Action Level 1 ....................7-2
Figure 7-3 2001 Feedwater Dissolved Oxygen Results at Power Operating Conditions...........7-3
Figure 7-4 Implementation of Revised Action Levels for HWC and HWC+NMCA.....................7-4
Figure 7-5 Feedwater Flush Duration ........................................................................................7-8
Figure 7-6 Startup/Shutdown Iron Analysis Method ................................................................7-11
Figure 8-1 Reactor Water Chloride & Sulfate Sampling Frequency ..........................................8-2
Figure 8-2 Total Co-60 & Zn-65 Sampling Frequency ...............................................................8-3
Figure 8-3 Soluble & Insoluble Co-60 & Zn-65 Sampling Frequency ........................................8-3
Figure 8-4 Reactor Water Iron Sampling Frequency .................................................................8-4
Figure 8-5 Reactor Water Zinc Sampling Frequency.................................................................8-5
Figure 8-6 Feedwater Iron Sampling Frequency .......................................................................8-6
Figure 8-7 Feedwater Zinc Sampling Frequency.......................................................................8-6
Figure 8-8 Feedwater Zinc Variability vs. Injection Method .......................................................8-9
Figure 8-9 Reactor Water Zinc Variability vs. Injection Method .................................................8-9
Figure 8-10 Feedwater Zinc Variability vs. Analysis Method ...................................................8-10
Figure 8-11 Reactor Water Zinc Variability vs. Analysis Method .............................................8-10

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Figure 8-12 Reactor Water and Feedwater Zinc Variability by Plant Distinguished by
Analysis Method...............................................................................................................8-11
Figure 8-13 Feedwater Sample Line Velocity – Number of Plants in Range and
Cumulative Percentage of Plants At or Below Range ......................................................8-14
Figure 8-14 Feedwater Soluble Zinc Variability vs. Main Sample Line Velocity ......................8-15
Figure 8-15 Feedwater Insoluble Iron Variability vs. Main Sample Line Velocity ....................8-16
Figure 9-1 Upright Pleat Filter Septum – View of Expanded Pleats ..........................................9-2
Figure 9-2 Fold-over Pleat Filter Septum ...................................................................................9-2
Figure 9-3 Trend In Iron Removal Septa Use ............................................................................9-8
Figure 9-4 Current Distribution of Precoat Septa Types (% of Vessels) ....................................9-9
Figure 9-5 Tolerable Cooling Water Ingress at Action Level 1 Reactor Water Sulfate ............9-24
Figure 10-1 Annual Average Feedwater Iron for 1998 – 2001 for Plants with Deep Bed
Only Condensate Polishing..............................................................................................10-1
Figure 10-2 Oyster Creek Summer Reactor Water Sulfate for 1999 - 2001 (Normalized
to RWCU Flow = 400 gpm) ..............................................................................................10-6
Figure 10-3 1998 – 2001 Average CDI Iron for Deep Bed Only BWRs ...................................10-7
Figure 10-4 Condensate Polisher Service Time to Flow Resistance Increase
(Temperature = 95 °F)......................................................................................................10-9
Figure 11-1 Power Uprate Status ............................................................................................11-1
Figure 11-2 Brunswick 1 Coolant Sulfate and Chloride Trends during CDD Bypassing
(June 2001) ......................................................................................................................11-4
Figure 11-3 Brunswick 1 Coolant Sulfate and Chloride Trends during CDD Bypassing
(August-October 2001).....................................................................................................11-5
Figure 11-4 Brunswick 2 Sulfate Trend during CDD Bypassing (September 2001) ................11-5
Figure 11-5 Brunswick 2 Sulfate and Chloride Trends during CDD Bypassing (May-June
2002) ................................................................................................................................11-6
Figure 11-6 KKL Resin Consumption ......................................................................................11-9
Figure 11-7 KKL Reactor Coolant Anion Trends ...................................................................11-10
Figure 12-1 Reactor Water Sulfate and Estimated Cation Resin Quantity ..............................12-2
Figure 12-2 Pilgrim Reactor Water Conductivity and pH Responses ......................................12-3
Figure 12-3 Pilgrim Reactor Water Conductivity and Sulfate Responses................................12-3
Figure 12-4 Pilgrim Reactor Water Conductivity and Zinc Responses ....................................12-4
Figure 12-5 Pilgrim Reactor Water Dissolved Oxygen and ECP Responses ..........................12-4
Figure 12-6 Pilgrim Reactor Water, Feedwater and Condensate Conductivity Responses.....12-5
Figure 12-7 Pilgrim Main Steam Line Activity Response .........................................................12-5
Figure 12-8 Pilgrim Offgas Activity Response .........................................................................12-6
Figure 12-9 Pilgrim Reactor Thermal Power Response to Resin Intrusion..............................12-6
Figure 12-10 Pilgrim Reactor Water Total Iodine Response ...................................................12-7
Figure 12-11 Pilgrim Bottom Head Drain Temperature Response ..........................................12-7
Figure A-1 Power History, Browns Ferry 2 ............................................................................... A-4

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Figure A-2 Feedwater Iron, Browns Ferry 2.............................................................................. A-5

Figure A-3 Reactor Water Anions, Browns Ferry 2................................................................... A-5
Figure A-4 Reactor Water Cobalt-60, Browns Ferry 2 .............................................................. A-6
Figure A-5 BRAC History, Browns Ferry 2................................................................................ A-6
Figure A-6 Reactor Water Iodines, Browns Ferry 2 .................................................................. A-7
Figure A-7 Sum of 6 Noble Gases, Browns Ferry 2.................................................................. A-8
Figure B-1 Power History, Browns Ferry 3 ............................................................................... B-4
Figure B-2 Feedwater Iron, Browns Ferry 3.............................................................................. B-5
Figure B-3 Reactor Water Anions, Browns Ferry 3................................................................... B-5
Figure B-4 Reactor Water Cobalt-60, Browns Ferry 3 .............................................................. B-6
Figure B-5 BRAC History, Browns Ferry 3................................................................................ B-6
Figure C-1 Power History, Brunswick 1 .................................................................................... C-4
Figure C-2 Feedwater Iron, Brunswick 1 .................................................................................. C-4
Figure C-3 Reactor Water Cobalt-60, Brunswick 1................................................................... C-5
Figure C-4 BRAC History, Brunswick 1 .................................................................................... C-5
Figure D-1 Power History, Brunswick 2 .................................................................................... D-4
Figure D-2 Feedwater Iron, Brunswick 2 .................................................................................. D-4
Figure D-3 Reactor Water Cobalt-60, Brunswick 2................................................................... D-5
Figure D-4 BRAC History, Brunswick 2 .................................................................................... D-5
Figure E-1 Power History, Clinton............................................................................................. E-4
Figure E-2 Feedwater Iron, Clinton........................................................................................... E-4
Figure E-3 Reactor Water Cobalt-60, Clinton ........................................................................... E-5
Figure E-4 BRAC History, Clinton............................................................................................. E-5
Figure F-1 Power History, Cooper Nuclear Station....................................................................F-3
Figure F-2 Feedwater Iron, Cooper Nuclear Station ..................................................................F-4
Figure F-3 Reactor Water Cobalt-60, Cooper Nuclear Station ..................................................F-4
Figure F-4 BRAC History, Cooper Nuclear Station....................................................................F-5
Figure G-1 Power History, Dresden 2....................................................................................... G-4
Figure G-2 Feedwater Iron, Dresden 2 ..................................................................................... G-4
Figure G-3 Reactor Water Cobalt-60, Dresden 2 ..................................................................... G-5
Figure G-4 BRAC History, Dresden 2 ....................................................................................... G-5
Figure H-1 Power History, Dresden 3 ....................................................................................... H-4
Figure H-2 Feedwater Iron, Dresden 3 ..................................................................................... H-4
Figure H-3 Reactor Water Cobalt 60, Dresden 3...................................................................... H-5
Figure I-1 Power History, Duane Arnold .....................................................................................I-4
Figure I-2 Feedwater Iron, Duane Arnold ...................................................................................I-4
Figure I-3 Reactor Water Cobalt-60, Duane Arnold....................................................................I-5
Figure I-4 BRAC History, Duane Arnold .....................................................................................I-5
Figure J-1 Power History, Fermi 2 ............................................................................................. J-4

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Figure J-2 Feedwater Iron, Fermi 2 ........................................................................................... J-4

Figure J-3 Reactor Water Cobalt 60, Fermi 2 ............................................................................ J-5
Figure J-4 BRAC History, Fermi 2 ............................................................................................. J-5
Figure K-1 Power History, FitzPatrick ....................................................................................... K-4
Figure K-2 Feedwater Iron, FitzPatrick ..................................................................................... K-5
Figure K-3 Reactor Water Cobalt-60, FitzPatrick...................................................................... K-5
Figure K-4 BRAC History, FitzPatrick ....................................................................................... K-6
Figure L-1 Power History, Grand Gulf........................................................................................L-4
Figure L-2 Feedwater Iron, Grand Gulf......................................................................................L-5
Figure L-3 Reactor Water Cobalt-60, Grand Gulf ......................................................................L-5
Figure L-4 BRAC History, Grand Gulf........................................................................................L-6
Figure M-1 Power History, Hatch 1...........................................................................................M-4
Figure M-2 Feedwater Iron, Hatch 1 .........................................................................................M-5
Figure M-3 Reactor Water Cobalt-60, Hatch 1 .........................................................................M-5
Figure M-4 BRAC History, Hatch 1 ...........................................................................................M-6
Figure N-1 Power History, Hatch 2 ........................................................................................... N-4
Figure N-2 Feedwater Iron, Hatch 2 ......................................................................................... N-5
Figure N-3 Reactor Water Cobalt-60, Hatch 2.......................................................................... N-5
Figure N-4 BRAC History, Hatch 2 ........................................................................................... N-6
Figure O-1 Power History, Hope Creek .................................................................................... O-3
Figure O-2 Feedwater Iron, Hope Creek .................................................................................. O-4
Figure O-3 Reactor Water Cobalt-60, Hope Creek ................................................................... O-4
Figure O-4 BRAC History, Hope Creek .................................................................................... O-5
Figure O-5 Reactor Water Dose Equivalent Iodine, Hope Creek ............................................. O-7
Figure O-6 Sum of 6 Noble Gases, Hope Creek ...................................................................... O-7
Figure P-1 Power History, Laguna Verde 1 .............................................................................. P-4
Figure P-2 Feedwater Iron, Laguna Verde 1 ............................................................................ P-5
Figure P-3 Reactor Water Cobalt-60 (µCi/kg), Laguna Verde 1 ............................................... P-5
Figure P-4 BRAC History, Laguna Verde 1 .............................................................................. P-6
Figure Q-1 Power History, Laguna Verde 2.............................................................................. Q-4
Figure Q-2 Feedwater Iron, Laguna Verde 2 ............................................................................ Q-5
Figure Q-3 Reactor Water Cobalt-60 (µCi/kg), Laguna Verde 2............................................... Q-5
Figure Q-4 BRAC History, Laguna Verde 2 .............................................................................. Q-6
Figure R-1 Power History, LaSalle 1......................................................................................... R-4
Figure R-2 Feedwater Iron, LaSalle 1....................................................................................... R-5
Figure R-3 Reactor Water Cobalt-60, LaSalle 1 ....................................................................... R-5
Figure R-4 BRAC History, LaSalle 1......................................................................................... R-6
Figure S-1 Power History, LaSalle 2......................................................................................... S-4
Figure S-2 Feedwater Iron, LaSalle 2 ....................................................................................... S-4

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Figure S-3 Reactor Water Cobalt-60, LaSalle 2 ....................................................................... S-5

Figure S-4 BRAC History, LaSalle 2 ......................................................................................... S-6
Figure T-1 Power History, Limerick 1.........................................................................................T-4
Figure T-2 Feedwater Iron, Limerick 1.......................................................................................T-5
Figure T-3 Reactor Water Cobalt-60, Limerick 1 .......................................................................T-5
Figure T-4 BRAC History, Limerick 1.........................................................................................T-6
Figure U-1 Power History, Limerick 2 ....................................................................................... U-4
Figure U-2 Feedwater Iron, Limerick 2 ..................................................................................... U-5
Figure U-3 Reactor Water Cobalt-60, Limerick 2...................................................................... U-5
Figure U-4 BRAC History, Limerick 2 ....................................................................................... U-6
Figure V-1 Power History, Monticello........................................................................................ V-4
Figure V-2 Feedwater Iron, Monticello...................................................................................... V-5
Figure V-3 Reactor Water Anions, Monticello........................................................................... V-6
Figure V-4 Reactor Water Cobalt-60, Monticello ...................................................................... V-6
Figure V-5 BRAC History, Monticello........................................................................................ V-7
Figure W-1 Power History, Nine Mile Point 1........................................................................... W-5
Figure W-2 Feedwater Iron, Nine Mile Point 1......................................................................... W-5
Figure W-3 Reactor Water Anions, Nine Mile Point 1.............................................................. W-6
Figure W-4 Reactor Water Cobalt-60, Nine Mile Point 1 ......................................................... W-6
Figure W-5 BRAC History, Nine Mile Point 1........................................................................... W-7
Figure X-1 Power History, Nine Mile Point 2............................................................................. X-4
Figure X-2 Feedwater Iron, Nine Mile Point 2........................................................................... X-5
Figure X-3 Reactor Water Anions, Nine Mile Point 2................................................................ X-6
Figure X-4 Reactor Water Cobalt-60, Nine Mile Point 2 ........................................................... X-6
Figure X-5 BRAC History, Nine Mile Point 2............................................................................. X-7
Figure Y-1 Power History, Oyster Creek................................................................................... Y-4
Figure Y-2 Feedwater Iron, Oyster Creek................................................................................. Y-5
Figure Y-3 Reactor Water Cobalt-60, Oyster Creek ................................................................. Y-6
Figure Y-4 BRAC history, Oyster Creek ................................................................................... Y-6
Figure Z-1 Power History pre-NMCA, Peach Bottom 2 .............................................................Z-4
Figure Z-2 Power History post-NMCA, Peach Bottom 2 ............................................................Z-4
Figure Z-3 Feedwater Iron pre-NMCA, Peach Bottom 2............................................................Z-5
Figure Z-4 Feedwater Iron post-NMCA, Peach Bottom 2 ..........................................................Z-6
Figure Z-5 Reactor Water Cobalt-60 pre-NMCA, Peach Bottom 2 ............................................Z-6
Figure Z-6 Reactor Water Cobalt-60 post-NMCA, Peach Bottom 2 ..........................................Z-7
Figure Z-7 BRAC History, Peach Bottom 2................................................................................Z-8
Figure AA-1 Power History, Peach Bottom 3 .......................................................................... AA-3
Figure AA-2 Feedwater Iron, Peach Bottom 3 ........................................................................ AA-4
Figure AA-3 Reactor Water Cobalt-60, Peach Bottom 3 ........................................................ AA-5

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Figure AA-4 BRAC History, Peach Bottom 3 .......................................................................... AA-5

Figure BB-1 Power History, Perry........................................................................................... BB-4
Figure BB-2 Feedwater Iron, Perry ......................................................................................... BB-5
Figure BB-3 Reactor Water Cobalt-60, Perry ......................................................................... BB-5
Figure BB-4 BRAC History, Perry ........................................................................................... BB-6
Figure CC-1 Power History, Pilgrim ........................................................................................CC-4
Figure CC-2 Feedwater Iron, Pilgrim ......................................................................................CC-4
Figure CC-3 Reactor Water Cobalt-60, Pilgrim.......................................................................CC-5
Figure DD-1 Power History, Quad Cities 1 .............................................................................DD-5
Figure DD-2 Feedwater Iron, Quad Cities 1 ...........................................................................DD-5
Figure DD-3 Reactor Water Cobalt-60, Quad Cities 1............................................................DD-6
Figure DD-4 BRAC History, Quad Cities 1 .............................................................................DD-6
Figure EE-1 Power History, Quad Cities 2.............................................................................. EE-4
Figure EE-2 Feedwater Iron, Quad Cities 2............................................................................ EE-4
Figure EE-3 Reactor Water Cobalt-60, Quad Cities 2 ............................................................ EE-5
Figure EE-4 BRAC History, Quad Cities 2.............................................................................. EE-5
Figure FF-1 Power History, River Bend .................................................................................. FF-4
Figure FF-2 Feedwater Iron, River Bend ................................................................................ FF-5
Figure FF-3 Reactor Water Cobalt-60, River Bend................................................................. FF-6
Figure FF-4 BRAC History, River Bend .................................................................................. FF-7
Figure FF-5 Feedwater Copper, River Bend........................................................................... FF-8
Figure FF-6 Reactor Water Copper, River Bend .................................................................... FF-8
Figure GG-1 Power History, Susquehanna 1 ........................................................................ GG-4
Figure GG-2 Feedwater Iron, Susquehanna 1....................................................................... GG-5
Figure GG-3 Reactor Water Cobalt 60, Susquehanna 1 ....................................................... GG-5
Figure GG-4 BRAC History, Susquehanna 1......................................................................... GG-6
Figure HH-1 Power History, Susquehanna 2..........................................................................HH-4
Figure HH-2 Feedwater Iron, Susquehanna 2 ........................................................................HH-5
Figure HH-3 Reactor Water Cobalt-60, Susquehanna 2 ........................................................HH-5
Figure HH-4 BRAC History, Susquehanna 2 ..........................................................................HH-6
Figure II-1 Power History, Vermont Yankee ..............................................................................II-4
Figure II-2 Feedwater Iron, Vermont Yankee.............................................................................II-5
Figure II-3 Reactor Water Cobalt-60, Vermont Yankee .............................................................II-5
Figure II-4 BRAC History, Vermont Yankee...............................................................................II-6
Figure II-5 Reactor Water Iodines, Vermont Yankee .................................................................II-8
Figure II-6 Offgas Sum of 6 Noble Gases, Vermont Yankee .....................................................II-8
Figure JJ-1 Power History, Columbia ...................................................................................... JJ-4
Figure JJ-2 Feedwater Iron, Columbia..................................................................................... JJ-5
Figure JJ-3 Reactor Water Cobalt-60, Columbia ..................................................................... JJ-5

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Figure JJ-4 BRAC History, Columbia....................................................................................... JJ-6

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LIST OF TABLES

Table 2-1 Hydrogen, Zinc, and Noble Metals Status for North American BWRs .......................2-3
Table 3-1 Types of Condensate Polishing for North American BWRs.......................................3-2
Table 3-2 U.S. BWRs Currently Injecting Iron into the Feedwater...........................................3-10
Table 3-3 2000 Average Feedwater Copper (BWRs with copper source in main
condenser) .......................................................................................................................3-15
Table 3-4 2001 Average Feedwater Copper (BWRs with copper source in main
condenser) .......................................................................................................................3-17
Table 3-5 2000 Average Hotwell Copper (BWRs with copper source in main condenser)......3-19
Table 3-6 2001 Average Hotwell Copper (BWRs with copper source in main condenser)......3-19
Table 4-1 Crud-Related and Total BWR Fuel Failures, 1989 – 1999 (Number of Failed
Fuel Assemblies)................................................................................................................4-3
Table 4-2 Changes Influencing Primary Chemistry at Each Hatch Unit ....................................4-7
Table 4-3 Hatch 1 RALO Measurements.................................................................................4-11
Table 4-4 Hatch 2 RALO Measurements.................................................................................4-12
Table 4-5 Composition of Fuel Deposits at BWRs (from EPRI NP-522)..................................4-22
Table 4-6 Oskarshamn 1 Fuel Deposition (from EPRI NP-522) ..............................................4-23
Table 4-7 Design Features of Plants Selected for Mass Balance Calculations .......................4-24
Table 4-8 Chemistry Regimes During Data Evaluation Period at Plants Selected for
Mass Balance Calculations ..............................................................................................4-24
Table 4-9 Mass Balance Results for Iron Deposition on Fuel (normalized to 24-month
fuel cycle with 20 outage days) ........................................................................................4-27
Table 4-10 Mass Balance Results for Copper Deposition on Fuel (normalized to 24-
month fuel cycle with 20 outage days) .............................................................................4-27
Table 4-11 River Bend April 2000 Startup Metals Data ...........................................................4-29
Table 4-12 River Bend April 2000 Startup Mass Balance Results...........................................4-30
Table 4-13 River Bend 100% Power Average and Startup Maximum Iron and Copper
Deposition Rates..............................................................................................................4-32
Table 4-14 Columbia July 2001 Startup Metals Data .............................................................4-33
Table 4-15 Columbia July 2001 Startup Mass Balance Results ..............................................4-34
Table 5-1 BRAC Results Summary for North American BWRs .................................................5-4
Table 6-1 Noble Metal Application Loading ..............................................................................6-2
Table 6-2 Noble Metal Coupon Loading ...................................................................................6-3
Table 6-3 Nine Mile Point 1 Artifacts..........................................................................................6-5

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Table 6-4 NMCA Coupon Location and Temperature ...............................................................6-5

Table 6-5 Nine Mile Point 2 RWCU Influent and F/D Effluent Analysis on 04/11/01 ...............6-36
Table 6-6 Calculated Platinum ECP as a Function of Temperature, pH and Hydrogen
Concentration...................................................................................................................6-39
Table 7-1 Monitoring of Mitigation Conditions............................................................................7-5
Table 7-2 Reactor Coolant Dissolved Oxygen and H2/O2 Molar Ratio Control Limits................7-6
Table 7-3 Other Secondary Parameters Used to Confirm Mitigation........................................7-7
Table 7-4 Feedwater Flush Chemistry Analyses .......................................................................7-9
Table 7-5 Sample Locations During Feedwater Flush...............................................................7-9
Table 7-6 Feedwater Flush Criteria Target Values .................................................................7-11
Table 8-1 Chemistry Parameter Sampling Frequency...............................................................8-1
Table 8-2 Soluble Zinc Analysis Method & Injection Method .....................................................8-8
Table 8-3 Feedwater Sample Tubing.......................................................................................8-13
Table 9-1 Filter Septa Evaluated In BWR Condensate Polishing Applications..........................9-9
Table 9-2 BWR Precoated Iron Removal Septa ......................................................................9-11
Table 9-3 ACSI Values for BWR Precoated Iron Removal Septa............................................9-13
Table 9-4 Comparison of HydroGuard™ 10 µm Septa Performance ......................................9-14
Table 9-5 BWR Non-Precoat Iron Removal Septa ..................................................................9-15
Table 9-6 ACSI Values for BWR Non-Precoated Iron Removal Applications ..........................9-16
Table 9-7 Comparison of Pall 1 µm Non-Precoated Septa After One Year of Service............9-17
Table 9-8 BWR Condensate Deep Bed Only Plants................................................................9-18
Table 10-1 Resins Used with Deep Bed Only Condensate Polishing......................................10-3
Table 10-2 Summer Reactor Water Sulfate and RWCU Flow for 1999 and 2000
(Normalized to RWCU Flow = 400 gpm)..........................................................................10-5
Table 10-3 Resins Used with Filter + Deep Bed Condensate Polishing ................................10-14
Table A-1 Browns Ferry 2 Plant Design Parameters ................................................................ A-1
Table A-2 Browns Ferry 2 Milestones...................................................................................... A-2
Table A-3 Browns Ferry 2 Recirculation System Dose Rates .................................................. A-3
Table B-1 Browns Ferry 3 Plant Design Parameters ................................................................ B-1
Table B-2 Browns Ferry 3 Milestones....................................................................................... B-2
Table B-3 Browns Ferry 3 Recirculation System Dose Rates .................................................. B-3
Table C-1 Brunswick 1 Plant Design Parameters ..................................................................... C-1
Table C-2 Brunswick 1 Milestones............................................................................................ C-2
Table C-3 Brunswick 1 Recirculation System Dose Rates ....................................................... C-3
Table C-4 Brunswick 1 Recirculation Piping Gamma Scan Results ......................................... C-7
Table D-1 Brunswick 2 Plant Design Parameters ..................................................................... D-1
Table D-2 Brunswick 2 Milestones............................................................................................ D-2
Table D-3 Brunswick 2 Recirculation System Dose Rates ....................................................... D-3
Table D-4 Brunswick 2 Recirculation Piping Gamma Scan Results ......................................... D-7

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Table E-1 Clinton Plant Design Parameters ............................................................................. E-1

Table E-2 Clinton Milestones .................................................................................................... E-2
Table E-3 Clinton Recirculation System Dose Rates................................................................ E-3
Table E-4 Clinton Recirculation Piping Gamma Scan Results ................................................. E-6
Table F-1 Cooper Plant Design Parameters..............................................................................F-1
Table F-2 Cooper Milestones.....................................................................................................F-2
Table F-3 Cooper Recirculation Dose Rates .............................................................................F-3
Table F-4 Cooper Nuclear Station Recirculation Piping Gamma Scan Results ........................F-6
Table G-1 Dresden 2 Plant Design Parameters ....................................................................... G-1
Table G-2 Dresden 2 Milestones .............................................................................................. G-2
Table G-3 Dresden 2 Recirculation System Dose Rates .......................................................... G-3
Table H-1 Dresden 3 Plant Design Parameters........................................................................ H-1
Table H-2 Dresden 3 Milestones .............................................................................................. H-2
Table H-3 Dresden 3 Recirculation System Dose Rates .......................................................... H-3
Table I-1 Duane Arnold Plant Design Milestones .......................................................................I-1
Table I-2 Duane Arnold Milestones.............................................................................................I-2
Table I-3 Duane Arnold Recirculation System Dose Rates ........................................................I-3
Table I-4 Duane Arnold Recirculation Piping Gamma Scan Results ..........................................I-6
Table J-1 Fermi 2 Plant Design Parameters .............................................................................. J-1
Table J-2 Fermi 2 Milestones..................................................................................................... J-2
Table J-3 Fermi 2 Recirculation System Dose Rates ................................................................ J-3
Table K-1 FitzPatrick Plant Design Parameters........................................................................ K-1
Table K-2 FitzPatrick Milestones .............................................................................................. K-2
Table K-3 FitzPatrick Recirculation System Dose Rates .......................................................... K-3
Table K-4 FitzPatrick Recirculation Piping Gamma Scan Results ............................................ K-7
Table L-1 Grand Gulf Plant Design Parameters ........................................................................L-1
Table L-2 Grand Gulf Milestones ...............................................................................................L-2
Table L-3 Grand Gulf Recirculation System Dose Rates...........................................................L-3
Table L-4 Grand Gulf Recirculation Piping Gamma Scan Results ............................................L-7
Table M-1 Hatch 1 Plant Design Parameters ...........................................................................M-1
Table M-2 Hatch 1 Milestones ..................................................................................................M-2
Table M-3 Hatch 1 Recirculation System Dose Rates ..............................................................M-3
Table M-4 Hatch 1 Recirculation Piping Gamma Scan Results................................................M-7
Table N-1 Hatch 2 Plant Design Parameters ............................................................................ N-1
Table N-2 Hatch 2 Milestones................................................................................................... N-2
Table N-3 Hatch 2 Recirculation System Dose Rates .............................................................. N-3
Table N-4 Hatch 2 Recirculation Piping Gamma Scan Results ................................................ N-7
Table O-1 Hope Creek Plant Design Parameters ..................................................................... O-1
Table O-2 Hope Creek Milestones........................................................................................... O-2

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Table O-3 Hope Creek Recirculation System Dose Rates ...................................................... O-3
Table O-4 Hope Creek Recirculation Piping Gamma Scan Results ......................................... O-6
Table P-1 Laguna Verde 1 Plant Design Parameters............................................................... P-1
Table P-2 Laguna Verde 1 Milestones...................................................................................... P-2
Table P-3 Laguna Verde 1 Recirculation System Dose Rates ................................................. P-3
Table Q-1 Laguna Verde 2 Plant Design Parameters .............................................................. Q-1
Table Q-2 Laguna Verde 2 Milestones ..................................................................................... Q-2
Table Q-3 Laguna Verde 2 Recirculation System Dose Rates ................................................. Q-3
Table R-1 LaSalle Unit 1 Plant Design Parameters .................................................................. R-1
Table R-2 LaSalle 1 Milestones ................................................................................................ R-2
Table R-3 LaSalle 1 Recirculation System Dose Rates............................................................ R-3
Table S-1 LaSalle 2 Plant Design Parameters ......................................................................... S-1
Table S-2 LaSalle 2 Milestones ................................................................................................ S-2
Table S-3 LaSalle 2 Recirculation System Dose Rates............................................................ S-3
Table S-4 LaSalle 2 Recirculation Piping Gamma Scan Results.............................................. S-7
Table T-1 Limerick 1 Plant Design Parameters .........................................................................T-1
Table T-2 Limerick 1 Milestones ................................................................................................T-2
Table T-3 Limerick 1 Recirculation System Dose Rates............................................................T-3
Table T-4 Limerick 1 Recirculation Piping Gamma Scan Results .............................................T-7
Table U-1 Limerick 2 Plant Design Parameters........................................................................ U-1
Table U-2 Limerick 2 Milestones............................................................................................... U-2
Table U-3 Limerick 2 Recirculation System Dose Rates .......................................................... U-3
Table U-4 Limerick 2 Recirculation Piping Gamma Scan Results ............................................ U-7
Table V-1 Monticello Plant Design Parameters ........................................................................ V-1
Table V-2 Monticello Milestones ............................................................................................... V-2
Table V-3 Monticello Recirculation System Dose Rates........................................................... V-3
Table V-4 Monticello Recirculation Piping Gamma Scan Results ............................................ V-8
Table W-1 Nine Mile Point 1 Plant Design Parameters ........................................................... W-1
Table W-2 Nine Mile Point 1 Milestones .................................................................................. W-2
Table W-3 Nine Mile Point 1 Recirculation Piping Dose Rates................................................ W-3
Table W-4 Nine Mile Point 1 Recirculation Piping Gamma Scan Results............................... W-8
Table X-1 Nine Mile Point 2 Plant Design Parameters ............................................................. X-1
Table X-2 Nine Mile Point 2 Milestones .................................................................................... X-2
Table X-3 Nine Mile Point 2 Recirculation Piping Dose Rates.................................................. X-3
Table X-4 Nine Mile Point 2 Recirculation Piping Gamma Scan Results ................................. X-8
Table Y-1 Oyster Creek Plant Design Parameters ................................................................... Y-1
Table Y-2 Oyster Creek Milestones .......................................................................................... Y-2
Table Y-3 Oyster Creek Recirculation System Dose Rates...................................................... Y-3
Table Z-1 Peach Bottom 2 Plant Design Parameters ................................................................Z-1

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Table Z-2 Peach Bottom 2 Milestones......................................................................................Z-2

Table Z-3 Peach Bottom 2 Recirculation System Dose Rates ..................................................Z-3
Table Z-4 Peach Bottom 2 Recirculation Piping Gamma Scan Results ....................................Z-8
Table AA-1 Peach Bottom 3 Plant Design Parameters .......................................................... AA-1
Table AA-2 Peach Bottom 3 Milestones ................................................................................. AA-2
Table AA-3 Peach Bottom 3 Recirculation System Dose Rates ............................................. AA-3
Table AA-4 Peach Bottom 3 Recirculation Piping Gamma Scan Results............................... AA-6
Table BB-1 Perry Plant Design Parameters ........................................................................... BB-1
Table BB-2 Perry Milestones .................................................................................................. BB-2
Table BB-3 Perry Recirculation System Dose Rates .............................................................. BB-3
Table BB-4 Perry Recirculation Piping Gamma Scan Results............................................... BB-7
Table CC-1 Pilgrim Plant Design Parameters.........................................................................CC-1
Table CC-2 Pilgrim Milestones ...............................................................................................CC-2
Table CC-3 Pilgrim Milestones ..............................................................................................CC-3
Table CC-4 Pilgrim Recirculation Piping Gamma Scan Results .............................................CC-6
Table DD-1 Quad Cities Unit 1 Plant Design Parameters ......................................................DD-1
Table DD-2 Quad Cities 1 Milestones.....................................................................................DD-2
Table DD-3 Quad Cities 1 Recirculation Piping Dose Rates ..................................................DD-3
Table EE-1 Quad Cities 2 Plant Design Parameters .............................................................. EE-1
Table EE-2 Quad Cities 2 Milestones ..................................................................................... EE-2
Table EE-3 Quad Cities 2 Recirculation Piping Dose Rates................................................... EE-3
Table FF-1 River Bend Plant Design Parameters................................................................... FF-1
Table FF-2 River Bend Milestones ......................................................................................... FF-2
Table FF-3 River Bend Recirculation System Dose Rates ..................................................... FF-3
Table FF-4 River Bend Annual Average Feedwater and CDE Metals, 1997-2001................. FF-5
Table GG-1 Susquehanna 1 Plant Design Parameters ......................................................... GG-1
Table GG-2 Susquehanna 1 Milestones................................................................................ GG-2
Table GG-3 Susquehanna 1 Recirculation System Dose Rates ........................................... GG-3
Table GG-4 Susquehanna 1 Recirculation Piping Gamma Scan Results ............................. GG-7
Table HH-1 Susquehanna 2 Plant Design Parameters ..........................................................HH-1
Table HH-2 Susquehanna 2 Milestones .................................................................................HH-2
Table HH-3 Susquehanna 2 Recirculation System Dose Rates .............................................HH-3
Table HH-4 Susquehanna 2 Recirculation Piping Gamma Scan Results...............................HH-7
Table II-1 Vermont Yankee Plant Design Parameters ...............................................................II-1
Table II-2 Vermont Yankee Milestones......................................................................................II-2
Table II-3 Vermont Yankee Recirculation System Dose Rates .................................................II-3
Table JJ-1 Columbia Plant Design Parameters ....................................................................... JJ-1
Table JJ-2 Columbia Milestones.............................................................................................. JJ-2
Table JJ-3 Columbia Recirculation System Dose Rates ......................................................... JJ-3

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Table JJ-4 Columbia Recirculation Piping Gamma Scan Results ........................................... JJ-7

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1
INTRODUCTION

The purpose of this report is to compile and summarize available chemistry and radiation dose
control data for the thirty-six operating BWR (Boiling Water Reactor) plants in North America.
This work draws on plant data that are continually collected and maintained in the EPRI BWR
Chemistry Monitoring database. This database is maintained to support EPRI initiatives, such as:
• BWR Water Chemistry Guidelines revisions
• NMCA (Noble Metals Chemical Application) Guidelines development
• DZO (Depleted Zinc Oxide) Zinc Addition Optimization
• Plant demonstration results to achieve radiation field reduction
• BWRVIP (BWR Vessel and Internals Project) Radiolysis Model benchmarking
• BWR fuel issue evaluations
• BWR Condensate Filter Users Group
• Direct support on individual plant issues

EPRI BWR Chemistry Monitoring Database

The database encompasses condensate, feedwater and reactor water chemistry, radiation dose
data, plant design data, resins and filters in use, and other data on plant chemistry practices for
the thirty-six operating North American BWRs. The structure of the database is illustrated in
Figure 1-1.

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Introduction

Chemistry
ChemistryData
Data

Radiation
Radiation Plant
PlantDesign
Design
Dose
DATABASE
DoseData
Data Data
Data

Iron/Chem
Iron/ChemControl
Control
Technologies
Technologies
Figure 1-1
Structure of EPRI BWR Chemistry Monitoring Database

The database was started in 1997 to support BWR iron control initiatives. It was expanded to
support iron and radiation field correlations (1, 2). The data summaries and correlations
developed were applied as the technical bases for iron control guidelines in the EPRI BWR
Water Chemistry Guidelines – 2000 revision (3). The database scope was further expanded to
incorporate parameters in support of plant radiation control demonstration results (4) and NMCA
experience and application guidelines development (5). The database was also applied in EPRI
iron control tailored collaboration projects to requalify low-crosslinked ion exchange resins (6)
and to evaluate the performance of a deep bed condensate polishing system (7).

The objective of the ongoing monitoring activity is to maintain a single-source database that can
be used to support a wide range of EPRI studies and investigations related to chemistry and
radiation dose issues. One key objective of this effort is to reduce burden on plant personnel by
minimizing duplicate requests for data. In addition, the reliability of the data is maximized
through data quality verification and consistency checks.

The types of data included for each plant are described in more detail as follows:

Chemistry Data: Milestone dates and details of chemistry programs (NWC, HWC, NMCA,
NZO and DZO, Iron Addition) are included. The largest segment of the database is comprised of
chemistry trend data, including reactor power and the normally monitored chemistry parameters
for condensate, feedwater and forward-pumped drains, reactor water and RWCU (reactor water
cleanup). When available, ECP (electrochemical corrosion potential) data are also included. To
support the development of the NMCA guidelines, MSLRM (main steam line radiation monitor)
results, offgas release rate expressed as the sum of the six noble gases, and reactor water
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 EPRI Licensed Material

Introduction

radioiodines data were also included. In addition, chemistry data of auxiliary systems (e.g.,
radwaste, makeup, fuel pool, condensate storage tank) are available for many plants.

Plant Design Data: Major plant design data include BWR type, license start date/original end
date/extension date, power rating and amounts/dates of power uprates, circulating cooling water
source and chemical characterization, major system rated flows, reactor liquid mass inventory
and fuel surface area. Other data include materials of construction (condenser, heaters, reactor
recirculation piping, extraction steam piping, RWCU piping), condensate polisher system design
(filters, filter demineralizers and deep bed demineralizers) and RWCU system design.

Iron/Chemistry Control Technologies: This segment of the database includes details of the
materials, components and practices plants are employing to achieve chemistry control goals.
The technologies include deep bed condensate demineralizer bead ion exchange resins,
condensate F/D (filter demineralizer) precoat materials, RWCU F/D and RWCU deep bed resins,
filter septa (precoat, non-precoat), type of iron added and practices (e.g., cleaning/backwash
frequency, anion underlay, precoat relaxation).

Radiation Dose Data: Drywell piping shutdown radiation dose rate survey results and practices
employed in performing the surveys are collected along with piping gamma scan results. Dose
rates measured at hot spots are included along with details on cobalt materials (Stellite™)
replacement details, chemical decontamination, unit scram history, materials pre-conditioning
(e.g., electropolishing, chromium plating) and startup/shutdown strategies for dose control.

Electronic Reports

In an effort to report back to plants more frequently, the following electronically transmitted
reports have been developed. The reports are intended for use by the plant staff in benchmarking
and improvement initiatives, and are transmitted to all thirty-six operating North American
BWRs.

BRAC Summary Report

This electronically transmitted report to plants provides a summary of shutdown drywell

radiation dose rates, and thus fills an information gap. The first issue of the report was
transmitted in May 2001 and has been updated twice since then. The report will continue to be
updated twice per year, after spring and fall refueling outages when new drywell radiation survey
data are normally available.

BWR Chemistry Sampling Frequency Report

This electronic report provides sampling frequencies for the following parameters: reactor water
chloride and sulfate; reactor water iron and zinc; reactor water total, soluble, and insoluble
cobalt-60 and zinc-65; and feedwater iron, copper, and zinc. Frequencies were determined by

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Introduction

evaluation of chemistry data and/or plant surveys and confirmed by industry chemistry
personnel. The first issue of the report was transmitted in June 2002, and the report will be
issued annually. This report is intended for use as a benchmarking tool as corporate and plant
chemistry management strive to optimize resources and standardize practices.

BWR Resin Usage Reports

The BWR Resin Usage electronic reports contain information on condensate deep bed and filter
demineralizer resin usage. The information is issued in two separate reports to allow targeting to
the specific interests of generating company personnel based on the condensate polishing system
design of their specific plant(s). The BWR Condensate Demineralizer Deep Bed Resins report
includes a combination of charts and tables that give for each plant the total resin volume, cation
resin volume, anion underlay volume, resin supplier, product names and resin characteristics.
The BWR Condensate Filter Demineralizer Precoat Materials report gives for each plant the
supplier, product names, precoat dosage and precoat material properties. These reports were first
issued in September 2002, and updating is planned on an annual basis.

Condensate, Feedwater, and Reactor Water Chemistry Report

An additional electronically transmitted report, which is under development, will provide the
most recent 12-month average values for each plant for the following chemistry data parameters:
hotwell iron and copper; feedwater hydrogen, iron, copper, nickel, and zinc; and reactor water
chloride and sulfate. In addition, a number of calculated parameters will be given, such as
feedwater iron/zinc ratio and condensate polisher effluent chloride and sulfate concentrations.
The first issue of the report is scheduled for the fourth quarter of 2002. Semi-annual updates are
planned.

Individual Plant and Utility Assistance

The BWR Chemistry Monitoring Database is frequently applied in response to information

and/or data requests of utility personnel for benchmarking and assessments, and to support
individual plant improvement initiatives. Some examples are listed below:

Brunswick: Database support for evaluation of chemistry control impact of an extended power
uprate to 120% of original licensed thermal power

Cooper: List of users of Delaval Filter Demineralizers in Reactor Water Cleanup application

Cooper: Reactor water and feedwater chemistry and summary of chemistry regimes as part of an
optimum water chemistry benchmarking effort

Dresden: Database support for a chemistry assessment following implementation of an extended

power uprate

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Introduction

Entergy Nuclear Northeast: Historical reactor water zinc data for NMCA plants for use in fuel
failure root cause analysis

Exelon: Historical Co-60 data for use in a corporate evaluation

Exelon: Information on BRAC measurement (location, type of instrumentation used, etc.) for
development of a corporate BRAC measurement standard

FitzPatrick: Database support for a plant chemistry assessment

Hatch: Data evaluations to determine the need for feedwater iron addition

Hope Creek: Number of plants using anion underlays in condensate deep beds

Oyster Creek: Database support for assessment of feedwater iron control improvement options
with Deep Bed Only condensate polishing

Nine Mile Point 1: Technical information and contact personnel for recent chemical
decontaminations performed at Browns Ferry and Susquehanna

Nine Mile Point 1: Database support for an evaluation of long-term iron and copper control
options

Nine Mile Point 2: Database support for evaluation of elevated RWCU effluent conductivity
following NMCA

Pilgrim: Database support for an evaluation of long-term condensate polishing options

Quad Cities: Plant data summaries and evaluation to support a plant assessment

River Bend: Support assessment of elevated hotwell and feedwater iron levels when changing
from Normal Water Chemistry to moderate hydrogen water chemistry

Susquehanna: Trends in Fe-59 gamma scan activity

Vermont Yankee: Feedwater metals data for use in fuel failure evaluation

Final Report

This Final Report provides an update of the status of chemistry control programs and issues
impacting performance. The report documents results of surveys and evaluations to measure
actual BWR plant chemistry performance, experiences and practices versus the EPRI BWR
Water Chemistry Guidelines – 2000 Revision. It will be of value as a reference in preparation of
the next revision of the BWR Water Chemistry Guidelines. Recommendations are also included

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Introduction

on new areas to be addressed in the next revision of the EPRI BWR Water Chemistry
Guidelines.

The report includes an iron and radiation dose control summary for each of the thirty-six BWR
plants monitored. Focus areas include: updates to plant chemistry regimes; a summary of
industry HWC/NMCA experience; feedwater iron control status and industry progress in
implementing new iron control guidelines; industry progress in meeting feedwater dissolved
oxygen guidelines; updated correlations involving iron; cobalt-60 and radiation dose rates;
sampling and analysis practices; and other topics that assess industry practices vs. the EPRI
Water BWR Chemistry Guidelines. A separate appendix section for each monitored plant
summarizes chemistry and radiation dose data, and plant practices and milestone events that have
influenced the data.

References

1. “BWR Iron Control Monitoring,” TR-108737, Interim Report, December 1998.

2. “BWR Iron Control Monitoring,” TR-109565, Final Report, September 1999.

3. “EPRI BWR Water Chemistry Guidelines – 2000 Revision,” TR-103515-R2, Final Report,
February 2000.

4. “BWR Activity Control - Plant Demonstration Results,” Final Report, December 1999.

5. “BWRVIP-92: NMCA Experience Report and Application Guidelines,” TR-1003022, Final

Report, September 2001.

6. “Requalification of Low Crosslinked Resin,” TR-113368, Interim Report, August 1999.

7. “Condensate Demineralizer System Evaluation at Pilgrim,” TR-113369, Final Report, July

1999.

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2
PLANT CHEMISTRY REGIME STATUS

The status of hydrogen injection, zinc addition and application of noble metals at 36 operating
North American BWRs is shown in Table 2-1. This table updates the information given in Table
2-7 of the EPRI BWR Water Chemistry Guidelines – 2000 Revision (1) and adds entries for
NMCA.

Hydrogen Water Chemistry (HWC)

The oxidizing chemistry environment of the BWR primary system under normal water chemistry
(NWC) is a key factor for aggravating IGSCC (Intergranular Stress Corrosion Cracking).
Injecting hydrogen into the feedwater was shown in the early 1980s to reduce the ECP
(electrochemical corrosion potential) of the reactor coolant, and was thereby successful in
mitigating IGSCC of the reactor recirculation piping (1). Hydrogen addition suppresses the
formation of oxidizing radiolytic reaction products such as oxygen and hydrogen peroxide, thus
lowering the ECP driving force for IGSCC. Under NWC, the ECP of primary system materials
is typically in the range of 0 to -250 mV, referred to the standard hydrogen electrode (SHE).
Extensive laboratory and field testing has shown that IGSCC of BWR piping is mitigated by
lowering the ECP of sensitized stainless steel below -230 mV (SHE) through hydrogen injection,
and keeping the reactor coolant conductivity below 0.3 µS/cm at 25 oC. Low hydrogen addition
(HWC-L), defined as 0.4 to <1.0 ppm hydrogen in the final feedwater, is mainly targeted at
protecting the reactor recirculation piping. Moderate hydrogen addition (HWC-M), defined as
1.0 to <2.0 ppm hydrogen in the feedwater, is targeted at protecting reactor internals as well.

As of August 2002, thirty-one (31) plants were injecting hydrogen to protect vessel internals
and/or reactor recirculation system piping from IGSCC. Only five (5) plants were not injecting
hydrogen. Of the thirty-one plants injecting hydrogen:
• Twenty-two (22) plants applied noble metals as well (see NMCA, below). A 23rd plant also
performed NMCA, but did not initiate hydrogen injection after NMCA.
• Seven (7) plants were on moderate hydrogen (HWC-M).
• Four (4) plants were on low hydrogen (HWC-L), with one of these increasing flow into the
moderate range.

Noble Metals Chemical Application (NMCA)

A major drawback to HWC-M is the high operating dose rate in the main steam line, resulting
from increased levels of 16N. Noble Metal Chemical Addition (NMCA) has been developed to

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Plant Chemistry Regime Status

provide the IGSCC protection of HWC-M at lower hydrogen injection rates, thus reducing the
impact on operating radiation dose.

NMCA involves the deposition of small amounts of platinum and rhodium on the wetted
surfaces in contact with the reactor coolant. These noble metal deposits catalyze recombination
reactions of hydrogen with oxygen and hydrogen peroxide at these surfaces. The ECP response
of these treated surfaces is similar to that of a platinum surface in that protective ECPs are
achieved when the molar ratio of hydrogen to total oxidant in reactor water is as low as two.
This ratio is achieved at very low feedwater hydrogen concentrations (usually between 0.1 and
0.3 ppm). This is approximately an order of magnitude less than the hydrogen concentration
required by HWC-M; thus application of NMCA minimizes the increase in main steam line 16N
radiation because lower hydrogen addition levels can be used.

Twenty-three (23) plants had completed at least one application of noble metals prior to the fall
2002 refueling outages. Of these plants, Perry applied noble metals in 02/01 but did not begin
hydrogen addition until 7/02. Columbia applied noble metals in 05/01 but does not plan to start
injecting hydrogen until the fall of 2003. Cooper applied noble metals in 03/00 and planned to
start injecting hydrogen in 09/02. Vermont Yankee applied noble metals in 04/01 and plans to
start injecting hydrogen in 2003. An additional seven (7) plants have scheduled, or were
planning, NMCA. The remaining six (6) plants were still evaluating the NMCA process.

Zinc Addition

Addition of zinc into BWR reactor coolant can reduce Co-60 buildup on primary piping, thus
reducing radiation dose rates in the drywell and reducing personnel dose during outages.
Initially, NZO (natural zinc oxide) was added to the feedwater, but the increased 65Zn, from
neutron activation of the naturally occurring 64Zn isotope, added to the out of core radiation field.
Therefore, depleted zinc oxide (DZO) having <1% 64Zn is now used for BWR zinc injection.

As of August 2002, thirty-two plants were adding DZO to the feedwater for drywell radiation
field control. Four (4) plants were not adding zinc. None were adding NZO. Oyster Creek
(07/00) and Clinton (12/00) were the most recent plants to start adding DZO.

All thirty-two plants adding zinc target reactor water zinc concentrations at 5 ppb or higher.
Twenty-seven (27) have passive zinc addition systems and five (5) have active zinc injection
systems. One plant is planning to change its active system to a passive system in fall 2002.

The four plants that are currently not adding DZO are evaluating its use. At Nine Mile Point 1,
plant materials produce steady state average reactor water zinc concentrations of approximately
0.5 ppb. At Vermont Yankee, zinc levels were 8 ppb following NMCA.

References

1. “EPRI BWR Water Chemistry Guidelines – 2000 Revision,” TR-103515-R2, Final Report,
February 2000.

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Plant Chemistry Regime Status

Table 2-1
Hydrogen, Zinc, and Noble Metals Status for North American BWRs

FW H2 HWC FW H2 FW Zn Zn Inj. RW Zn Zn NMCA NMCA Noble Metal

BWR Injection Start (ppm) Injection Start Date (ppb) System Date Loading
Date (µg/cm2)

Browns Ferry 2 yes 12/99 0.3 yes 10/97 6.5 passive yes 3/01 0.95
Browns Ferry 3 yes 8/00 0.3 yes 12/95 8.1 passive yes 4/00 1.03
Brunswick 1 yes 6/90 1.1 yes 5/95 8-10 passive no NA
Brunswick 2 yes 1/89 1.1 yes 3/96 8-10 passive no NA
Clinton yes 6/02 NA yes 12/00 >5 passive yes 4/02 0.75
Columbia no 2003* NA yes 9/96 8 (1) passive yes 5/01 0.5
Cooper no 9/02* NA yes 12/99 5-10 passive yes 3/00 0.65
Dresden 2 yes 4/83 0.3 yes 12/96 6.9 passive yes 10/99 0.8
Dresden 3 yes 3/99 0.4 yes 7/98 6.5 passive yes 9/00 0.3
0.22, 0.8
10/96,
Duane Arnold yes 7/87 0.3 yes 12/94 8.0 passive yes (second
10/99
application)
Fermi 2 yes 9/97 1.1 yes 7/95 7.5 passive no NA
FitzPatrick yes 8/88 0.2 yes 1/89 5 passive yes 11/99 1.18
Grand Gulf yes 5/99 0.4 yes 2/98 5 passive no * NA
Hatch 1 yes 9/87 0.3 yes 8/90 6.6 passive yes 3/99 0.7
Hatch 2 yes 9/91 0.2 yes 8/90 6.6 passive yes 3/00 1.3
Hope Creek yes 1/93 0.72 yes 12/86 7.0 (2) active no 04/03* NA
Laguna Verde 1 no 10/02* NA yes 6/98 10.5 passive no 10/02* NA
Laguna Verde 2 no 4/03* NA yes 8/98 10.7 passive no 4/03* NA
LaSalle 1 yes 8/99 0.29 yes 7/94 5.5 passive yes 10/99 0.7
*Planned NA = not applicable (1) discontinued after 5/01 RFO (2) no zinc added 6/94 – 11/95

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Plant Chemistry Regime Status

FW H2 HWC FW H2 FW Zn Zn Inj. RW Zn Zn NMCA NMCA Noble Metal

BWR Injection Start (ppm) Injection Start Date (ppb) System Date Loading
2
Date (µg/cm )

LaSalle 2 yes 8/00 0.29 yes 6/95 6.3 passive yes 11/00 1.0
Limerick 1 yes 9/98 0.2 yes 1992 5.9 active yes 3/00 1.16
Limerick 2 yes 1/98 0.2 yes 1991 6.7 active yes 4/01 1.12
Monticello yes 2/89 1.4 yes 11/89 10.3 active no (1) NA
Nine Mile Point 1 yes 4/00 0.2 no Fall 2002* 0.4 NA yes 5/00 1.2
Nine Mile Point 2 yes 2/01 0.3 yes 4/88 7.7 passive yes 9/00 1.0
Oyster Creek yes 2/92 0.6 yes 7/00 7.0 passive no 9/02* NA
Peach Bottom 2 yes 5/97 0.2 yes 6/91 7.1 active yes 10/98 2.80
Peach Bottom 3 yes 3/97 0.2 yes 6/92 6.7 active yes 10/99 0.66
Perry no 8/02 NA yes 2/90 7.5 passive yes 2/01 0.9
Pilgrim yes 9/91 1.3 yes 12/96 7.7 (2) passive no * NA
Quad Cities 1 yes 10/90 0.4 yes 9/98 8.1 passive yes 4/99 0.6
Quad Cities 2 yes 10/90 0.3 yes 6/97 6.3 passive yes 1/00 1.0
River Bend no 12/01 0.3 (5) yes 6/97 (3) 5.6 (4) passive no * NA
Susquehanna 1 yes 1/99 1.8 no Spring NA no NA
2004*
Susquehanna 2 yes 8/99 1.8 no Spring NA no NA
2003*
Vermont Yankee no 1/03* NA no 7.8 NA yes 4/01 1.24

*Planned NA = not applicable (1) preliminary planning, no date (2) goal = 5-10 ppb (3) suspended 6/99 – 7/00 (4) increasing to 7 ppb
(5) increasing to 1.8 ppm

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3
IRON AND COPPER CONTROL

Plant Design Factors Impacting Feedwater Iron and Copper Control

The major factors influencing feedwater iron and copper control are:

1. Condensate polishing system design

2. Heater drains system design

3. Condenser tube materials of construction (for copper control)

The condensate polishing system design has the greatest influence on feedwater corrosion
products control at North American BWRs. There are three types of condensate polishing system
designs:
• Deep Bed Only (DB) – Deep bed demineralizers employing bead ion exchange resins. This
type of design has high capacity to remove ionic impurities but limited capability to remove
particulate impurities (such as iron and copper oxides).
• Filter + Deep Bed (F+DB) – Deep bed demineralizers downstream of high-efficiency
backwashable filters. These filters are commonly called pre-filters and they are used without
precoat materials. In North America, all pre-filter systems currently employ vertical
cylindrical filter septa with pleated filter media. This design provides both efficient removal
of insoluble crud particles and high capacity to remove ionic impurities.
• Filter Demineralizer (F/D) – Filter Demineralizers employ vertical cylindrical filter septa
that are precoated with filter aid materials made up totally or partially of powdered cation
and anion exchange resins. This design provides efficient removal of insoluble crud particles
but has only limited capacity to remove ionic impurities.

A summary of the current types of condensate polishing system designs for the thirty-six
operating North American BWRs is shown in Table 3-1. Of the twenty-two BWRs with deep bed
demineralizers, eight currently have only deep beds and fourteen have filters + deep beds. There
are fourteen plants designed with Filter Demineralizers.

Of the condensate Filter Demineralizer plants, all except Columbia and Vermont Yankee employ
precoatable pleated septa in at least some of the filter vessels. Browns Ferry and Monticello use
pleated septa in all of their filter vessels. Pleated septa provide more efficient iron removal and
allow lower precoat material usage than conventional porous cylindrical septa. By using a

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Iron and Copper Control

combination of pleated and conventional septa, most plants with Filter Demineralizers can
control the average feedwater iron concentration within the EPRI optimal range of 0.5 – 1.5 ppb.
Table 3-1
Types of Condensate Polishing for North American BWRs

Deep Bed Only Filter + Deep Bed Filter Demineralizer

(8 BWRs) (14 BWRs) (14 BWRs)

Dresden 3 (a) (d) (e) (j) Brunswick 1 Columbia

FitzPatrick (e) Brunswick 2 Browns Ferry 2 (m)
Grand Gulf (a) (e) Clinton (g)(k) Browns Ferry 3 (m)
Nine Mile Point 1 (d) (e) Dresden 2 (a) (d) (e) (g) (k) Cooper (i)
Nine Mile Point 2 (c) (d) Hope Creek (g) Duane Arnold (i)
Oyster Creek (b) (e) Laguna Verde 1 (g) Fermi 2 (i)

Pilgrim (f) (e) Laguna Verde 2 (g) Hatch 1 (i)

River Bend (d) (e) (j) LaSalle 1 (g) Hatch 2 (i)
LaSalle 2 (g) Monticello (m)

Limerick 1 (h) Peach Bottom 2 (i)

Limerick 2 (h) Peach Bottom 3 (i)
Perry Quad Cities 1 (i) (l)

Susquehanna 1 (g) Quad Cities 2 (i) (l)

Susquehanna 2 (g) Vermont Yankee
4. Notes:
a. EPRI-ARCS resin cleaner was added as retrofit
b. JRCS resin cleaning method was added as retrofit
c. EPRI TC – Low cross-linked resin requalification
d. Some low cross-linked resins in use
e. Some or all small size cation resin in use
f. EPRI TC – Iron control improvement
g. Pre-Filters were added as retrofit
h. Deep beds were added as retrofit
i. Some precoated pleated septa use
j. Pre-Filter retrofit in engineering or installation phase
k. Partial filtration
l. One F/D vessel has been added to accommodate extended power uprate
m. 100% precoated pleated septa use

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Iron and Copper Control

Over the past ten years, significant design changes have been implemented at BWRs with Deep
Bed Only condensate polishers. Four units including Dresden 2, Dresden 3, Grand Gulf and
Oyster Creek have retrofitted improved resin cleaning systems. The Advanced Resin Cleaning
System (ARCS) was retrofitted at Grand Gulf and Dresden (single ARCS shared by the two units).
The Japanese Resin Cleaning System (JRCS) was retrofitted at Oyster Creek. In all cases, more
effective resin cleaning, lower liquid waste generation and less operator labor were benefits
achieved. Improved iron control has also been realized by plants with only deep bed polishers
through a combination of improved resin cleaning design and/or practices and resin selection.
Low cross-linked cation resins and small size cation resins (gel resins with >10% cross-link)
have been demonstrated to reduce feedwater iron at some plants.

However, the most dramatic trend in condensate polisher system design in the past ten years has
been the conversion to the Filter + Deep Bed design. The evolution of condensate polisher
designs for North American BWRs since 1990 is shown in Figure 3-1.

20
18
16
14
Number of BWR Units

12
10
8
6
4
2
0
1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 Proj Proj
2002 2003

Deep Bed Only Filter + Deep Bed Filter Demineralizer

Figure 3-1
Design Evolution of North American BWRs

Figure 3-1 shows that in 1990, there were only three North American BWRs with Filters + Deep
Beds, while there were 17 Deep Bed Only and 16 Filter Demineralizer plants. By 1992, after the
two Limerick units were converted from Filter Demineralizers to Filter + Deep Beds through a
retrofit of deep beds designed to improve the removal of copper from the copper alloy condenser
tubes, the number of Filter Demineralizer plants was reduced to 14 and the number of Filter +
Deep Bed plants increased to 5. The number of Filter Demineralizer plants has remained
constant at 14 since 1992.

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Iron and Copper Control

Starting in 1997, the number of filter retrofits to deep bed plants has increased dramatically. By
the end of 2000 there were 13 Filter + Deep Bed plants, decreasing the number of Deep Bed
Only plants to 9. Based on the filter retrofits that are currently in the installation phase at
Dresden 3 and River Bend, it is projected that by the end of 2002 there will be 16 Filter + Deep
Bed plants and 6 Deep Bed Only plants. Two additional BWRs with deep beds have made the
technical decision to install filters by 2003 and are in the project authorization phase.

Of the existing Filter + Deep Bed plants, all of the filter systems process 100% of the condensate
flow except Clinton and Dresden. Clinton has nine deep bed demineralizer vessels and originally
installed filters directly upstream of three vessels (one filter dedicated to one demineralizer).
After an extended shutdown, Clinton committed to installing three additional filter vessels,
which went into service in 2000 and 2001. Clinton therefore remains a partial filtration plant,
with prefilters directly upstream of 6 of 9 demineralizer vessels.

The Dresden 2 partial Filter + Deep Bed system went online in December 2001, and the Dresden
3 system startup is planned for the fourth quarter of 2002. The Dresden filters were designed to
reduce the inlet iron to the condensate demineralizers to achieve iron control goals under
extended power uprate conditions. At each unit, the filter system is a headered design to treat
40% of the condensate flow at full power, thereby reducing the inlet iron to the condensate
demineralizers by approximately 40%.

Feedwater Iron Control Status

BWR Industry Iron Control Performance

Measures taken by the industry to improve iron control have been successful. This is evident in
the downward trend in the 1997-2001 average feedwater iron results for North American BWRs
shown in Figure 3-2.

The EPRI desired minimum and desired maximum feedwater iron thresholds of 0.5 ppb and 3.0
ppb are indicated in Figure 3-2 along with the Action Level 1 limit of 5 ppb. Results are shown
for all plants for which sufficient data were available to determine a reliable average.

Each of the average iron values shown in Figure 3-2 is the simple average of the total iron
(soluble and insoluble) reported by each plant at power operating conditions (>10% power). The
average may differ from the time-weighted calculation made by plants for chemistry indicator
tracking. For the purpose of this report, the data used in calculating the averages are screened.
When the concentration deviates from the normal plant-specific range, the plant is contacted to
determine whether the data point is valid; otherwise, no other screening is performed on the data
provided.

The annual average feedwater iron results for 2001 are shown in Figure 3-3. These results show
that the average feedwater iron values of all BWRs in 2001 were below the 3 ppb desired
maximum. Of the 36 plants represented, 34 had feedwater iron averages within the 0.5 – 3.0 ppb
desired range.

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10
9
8
Total Feedwater Fe (ppb)

7
6
5
4
3
2
1
0
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36

1997 1998
1999 2000
2001 DESIRED MINIMUM
DESIRED MAXIMUM ACTION LEVEL 1

Figure 3-2
1997-2001 Average Feedwater Iron for North American BWRs

6
Total Feedwater Fe (ppb)

0
ENF2

PER1

FIT1

NMP2

NMP1

DNPS3
DNPS2
CGS
HCNS

GGNS
HAT1

HAT2
LIM1

DUA1

PB2
LIM2

SUS2
MON1
SUS1
CNS1

BRF2
CLI1

LAS2
LAG2
LAG1

LAS1
PB3
PIL1
BRF3

QUA2
QUA1

OYC1
RIB1
VY
BRU2
BRU1

DB F/D F+DB

DESIRED MINIMUM DESIRED MAXIMUM ACTION LEVEL 1

Figure 3-3
2001 Average Feedwater Iron for North American BWRs

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Brunswick 1 and Brunswick 2 had feedwater iron averages of less than 0.5 ppb. Both of these
units have full flow Filter + Deep Bed condensate polishing designs and forward pumped heater
drains with minimal iron contribution from the drains. These plants have not implemented iron
addition because no detrimental impact on plant dose rates has been attributed to iron being too
low. Limerick 1 and 2, Susquehanna 1 and 2, Hope Creek, and Columbia add iron to maintain a
feedwater concentration in the range of 0.5 ppb – 1 ppb to avoid potential increases in radiation
fields. Additional details in iron injection and the impact on radiation dose rates are given later
in this section and in Section 5.

Of the Deep Bed Only plants, FitzPatrick and Nine Mile Point 2 had the lowest average
feedwater iron values in 2001. FitzPartick has low inlet iron, which enables an average feedwater
iron concentration of less than 1.5 ppb to be achieved using standard resins. Nine Mile Point 2
uses low cross-linked cation resins in a portion of the demineralizers to control iron at about 1.5
ppb. Both FitzPatrick and Nine Mile Point 2 clean with air scrub & backwash and flow-
optimized URC (Ultrasonic Resin Cleaner). Nine Mile Point 1 was the third Deep Bed Only
plant with average feedwater iron of less than 2 ppb, and also uses low cross-linked resins with
air scrub & backwash and URC for resin cleaning. In addition, at Grand Gulf, the low area flow
rates of the condensate demineralizers along with ARCS and resin cleaning frequency
optimization allowed feedwater iron to be controlled at less than 2 ppb using standard (not low
cross-linked) resin.

It is also of interest to note the relatively high feedwater iron of Laguna Verde 1, Laguna Verde
2, LaSalle 1 and LaSalle 2, all of which are Filter + Deep Bed plants with forward pumped
drains. These plants have full flow filters and deep beds, resulting in low (<0.1 ppb) iron in the
polisher effluent and they do not inject iron into the feedwater. The reason for the relatively high
iron at these units is the high iron source term in the forward pumped drains stream.

The 2001 average feedwater iron concentrations as a function of condensate polisher design type
are shown in Figure 3-4. The group with the highest average feedwater iron concentration has
Deep Bed Only condensate polishing. The partial Filter + Deep Bed category is only
representative of Clinton with six of nine demineralizer vessels having dedicated filters in 2001;
Dresden 2 is not included in this category because the filters only began partial operation in
December 2001. The Filter Demineralizer plants and Filter + Deep Bed plants have the lowest
average feedwater iron.

The industry average feedwater iron for North American BWRs has shown a decreasing trend
since 1997, as shown in Figure 3-5. The industry average iron has decreased by 1.18 ppb over
this period. The main contribution to the iron reduction has come from the conversion of Deep
Bed Only plants to Filter + Deep Bed plants. The average feedwater iron has also decreased by
0.42 ppb for Filter Demineralizer plants, largely due to the increased use of precoatable septa
designed for improved iron removal.

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2.5
2.17

2.0
Feedwater Iron (ppb)

1.5 1.34
1.13
1.03
1.0

0.5

0.0
Deep Bed Only Partial Filter + Deep Filter Demineralizer Full Filter + Deep
Bed Bed

Figure 3-4
2001 Average Feedwater Iron vs. Condensate Polishing Type

5.0

4.5

4.0
Total Feedwater Fe (ppb)

3.5

3.0

2.5

2.0

1.5

1.0

0.5

0.0
1997 1998 1999 2000 2001

All DB F+DB F/D

Figure 3-5
1997 – 2001 Trend in Average Feedwater Iron (North American BWRs)

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Iron and Copper Control

The average feedwater iron value has also decreased sharply, by 2.31 ppb since 1997, for Deep
Bed Only plants. This trend is attributed to a combination of factors. One factor is that high
feedwater iron plants were converted from the Deep Bed Only to Filter + Deep Bed design. The
remaining Deep Bed Only plants are ones that have demonstrated some success in reducing
feedwater iron through resin cleaning modifications and techniques, the use of special resins and
optimized resin management practices (e.g., resin cleaning frequency and resin replacement
frequency). Some of the remaining Deep Bed Only plants, such as FitzPatrick and Pilgrim, have
low hotwell iron, resulting in less challenge to reduce iron to acceptable feedwater levels.

Impact of Drains Path on Iron Control

Depending on materials of construction, plants with forward pumped drains can have a
significant iron input to the final feedwater from this source. The forward pumped drains stream
typically accounts for 30% of the final feedwater flow, with the remaining 70% from the
condensate polishing system effluent. In general, plants constructed predominantly of carbon
steel in the extraction stream, moisture separator drains and high pressure heater drain piping
have the highest contribution of iron from the forward pumped drains stream, as indicated in
Figure 3-6. The Laguna Verde and LaSalle units are examples of plants with mostly carbon steel
construction in the forward pumped drains flow path and, despite having full flow condensate
filters upstream of deep bed demineralizers, do not have to add iron to maintain the feedwater
concentration >0.5 ppb.

On average, the 2001 average feedwater iron concentration is higher for plants with forward
pumped drains than for plants with cascaded drains, as shown in Figure 3-7. However, materials
of construction must be taken into account to fully assess the impact of drains path on iron
control. For plants with all drains cascaded to the hotwell, the iron transported to the feedwater
system is equal to the iron contribution from the condensate polishing system effluent plus the
net iron transport from the feedwater piping. For plants with forward pumped high pressure
drains, the forward pumped drains may contribute significant iron to the final feedwater,
depending on the materials of construction of the piping in the flow path.

3-8
 EPRI Licensed Material

Iron and Copper Control

10

7.76
8
FPD Iron (ppb)

6
4.56

4
3.00 3.00
2.10
2
0.89 0.93
0.43 0.53
0.17 0.20
0
1 2 3 4 5 6 7 8 9 10 11
Plant Number

Mainly Chrome-Moly Some Chrome-Moly

Carbon Steel Mat'l Not Reported

Figure 3-6
Iron Concentrations in Forward Pumped Drains

2.5

1.99
2.0
1.78
Total Feedwater Iron (ppb)

1.57
1.5 1.38
1.16 1.19

1.0 0.90
0.72

0.5

0.0
Cascaded Forward Pumped
DB F/D F+DB All

Figure 3-7
2001 Average Feedwater Iron vs. Drains Path

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 EPRI Licensed Material

Iron and Copper Control

Iron Injection

Plants that are routinely injecting iron into the feedwater are shown in Table 3-2. Five of the six
units listed are Filter + Deep Bed plants, one is a Filter Demineralizer plant, and all have
cascaded drains. Three of these units inject iron in the insoluble hematite form, Fe2O3, and three
inject iron oxalate hexahydrate, Fe2(C2O4)3 • 6H2O, which is soluble in the feed tank. At River
Bend, full flow condensate filters are expected to go online in the fourth quarter of 2002, and
iron oxalate injection is planned (River Bend has <0.5 ppb iron in the forward pumped drains
stream). The iron oxalate hydrolyzes when it is diluted in the feedwater stream, going to the
insoluble iron oxide form. The oxalic acid byproduct produces a measurable increase in
feedwater specific conductance; for example, the feedwater specific conductance at Susquehanna
is typically approximately 0.07 µS/cm due to the formation of oxalic acid after injection.

Plants adding iron continue to report that more iron is injected into the feedwater system than is
measured in the final feedwater. Typically, 50% to 75% of the iron injected is accounted for by
final feedwater sampling and analysis. Plants adding iron oxalate report fewer maintenance
problems with the injection equipment than plants adding iron oxide.
Table 3-2
U.S. BWRs Currently Injecting Iron into the Feedwater

Feedwater Iron
Concentration (ppb)
Plant Condensate Type of Iron
Polishing Injected Without Fe After Fe
Type Addition Addition

Hope Creek F + DB Iron Oxide <0.1 0.5 – 1

Limerick 1 F + DB Iron Oxide ~0.2 0.5 – 1

Limerick 2 F + DB Iron Oxide ~0.2 0.5 – 1

Susquehanna 1 F + DB Iron Oxalate <0.1 0.5 – 1

Susquehanna 2 F + DB Iron Oxalate <0.1 0.5 – 1

Columbia F/D Iron Oxalate <0.5 0.5 – 1

Both Hatch 1 and Quad Cities 2 have experienced feedwater iron levels of <0.5 ppb in 2002.
Iron addition is being considered at Hatch.

A schematic diagram of the Susquehanna iron oxalate injection skid is presented in Figure 3-8.
The iron injection skid is located in the turbine building and consists of two small injection
pumps, a stainless steel tank with agitator, valves, instrumentation and tubing. The skid is
locally controlled and manually operated. The injection pumps discharge into 3/8″ stainless steel
tubing that includes a check valve to prevent condensate from flowing back into the iron
injection equipment. The injection pump discharge tubing connects to a 1″ nozzle in the
condensate demineralizer system effluent piping. Electrical power for the skid is provided from

3-10
 EPRI Licensed Material

Iron and Copper Control

a local lighting panel and demineralized water for filling and flushing is also provided. Curbing
is installed around the skid to contain any spillage or leakage.

Figure 3-8
Susquehanna Iron Oxalate Injection Skid Schematic P&I Diagram

A safety evaluation was prepared for the iron injection systems installed at Susquehanna. Major
areas addressed and conclusions of the evaluation are listed as follows:
• The condensate system into which iron is injected has no safety functions.
• The power is taken from a lighting panel that provides A.C. power for non-safety related
electrical loads.
• The demineralized water system, which provides water to the skid, has no safety functions.
• The turbine building structure, to which the iron injection skid is attached, has no safety
functions. The iron injection system is designed so that there is no adverse impact on the
turbine building structure.
• The iron injection tubing and components are designed for the maximum pressure and
temperature of the condensate system to assure that the condensate system pressure boundary
will be maintained.

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 EPRI Licensed Material

Iron and Copper Control

• The iron injection system modification has no impact on the performance or operation of the
condensate system.
• There are no potential effects on safety functions that could be caused by changes in reactor
water chemistry.
• The design prevents backflow from the condensate system to the iron injection skid. There
are no changes in flow paths nor are there any changes to radiological zones or ventilation
boundaries; therefore, the modification does not create the potential for cross-contamination
or an uncontrolled release of radioactive materials.
• The skid is located in an area that does not interact with main steam lines, which are analyzed
to show system integrity is maintained during a seismic event.
• There is no impact on Technical Specifications, license or operating requirements.

Hotwell Iron

A plot of 1998 – 2000 individual plant average hotwell iron concentrations is shown in Figure 3-
9. Average individual plant 2001 hotwell iron concentrations from those plants reporting data are
shown in Figure 3-10. Hotwell iron varies widely, from about 7 ppb to >30 ppb, among the
plants shown in Figure 3-10. The majority of plants with the lowest hotwell iron have moisture
separator reheaters and the majority with the highest hotwell iron have no reheat. When plants
designed with reheat have experienced a loss in the availability of the reheaters, hotwell iron has
increased dramatically. A summary of annual Hotwell iron concentrations for plants with and
without moisture separator reheaters is shown in Figure 3-11. The annual averages for plants
designed with reheat are about 14 ppb, compared with about 20 ppb for those plants which have
no reheat.

The magnitude of the hotwell iron concentration is also an indictor of the type of action that must
be taken by a specific plant to provide sufficient removal of iron from condensate. FitzPatrick,
having low hotwell iron, can control feedwater iron at approximately 1.5 – 2 ppb using only deep
beds and standard resins. Oyster Creek, with about 14 ppb hotwell iron, has reached the target
range with standard resins, but only after a resin aging period of about 6 months. Dresden (prior
to partial prefilters) and Nine Mile Point 2 have hotwell iron concentrations in the 20 ppb range
and find deep beds with standard resins alone are not sufficient for iron control.

Dresden has installed the ARCS, but improved resin cleaning with standard resins was not
sufficient to control feedwater iron with their hotwell iron and high area flow rate challenge.
Therefore, Dresden continued the use of some low cross-linked resins along with the ARCS.
After retrofitting partial (40%) condensate filtration, with the condensate demineralizer inlet iron
reduced to approximately 12 ppb, Dresden plans to control feedwater iron at acceptable levels
without the use of low cross-linked resins. In 2001, Nine Mile Point 2 had five of nine
demineralizer beds charged with low cross-linked resins and Nine Mile Point 1 had four of six
demineralizers using low cross-linked resins to achieve feedwater iron control targets.

3-12
 Total Hotwell Fe (ppb) Total Fe (ppb)

0
10
20
30
40
50
60
0
10
20
30
40
50

Figure 3-9
FIT1

Figure 3-10
HAT1 JAF
PILG1
PIL1
HATCH2
HAT2 HATCH1
VY PER1
BRU2 BRU1
PER1 VY
DUA1 DUA1

1998, Reheat
LIM1 BRU2
LGS1

1998, No Reheat
MON1
ENF2

1998-2000 Average
CGS LGS2
LIM2 LAS1
GGNS CGS

2001 Average Hotwell Iron Concentration

BRU1 LAG1
OYC MON1

Reheat
ENF2 CNS1
LAS2

1998 – 2000 Average Hotwell Iron Concentration

SUS1
CLI1
QUA1
1999, Reheat

SUS1
RIB1 NMP1
1999, No Reheat

No Reheat
QUA2 OCY1
EPRI Licensed Material

SUS2 RIB1
NMP1 BRF2
LAG2 SUS2
HCNS
NMP2
LAG2
BRF3 GGNS
LAG1 DNPS2
DNPS3 QUA1
HCNS BRF3
2000, Reheat

BRF2 DNPS3
2000, No Reheat

DNPS2 NMP2
QUA2
PB3
PB2
PB2 PB3

3-13
Iron and Copper Control
 EPRI Licensed Material

Iron and Copper Control

25

21.09
19.36 19.36 19.48
20
Hotwell Iron (ppb)

14.78 14.37 14.39 14.38

15

10

0
1998 1999 2000 2001

Reheat No Reheat

Figure 3-11
1998 - 2001 Annual Average Hotwell Iron Concentration

Feedwater Copper Control Status

Monitored North American BWR plants with a significant copper source in the main condenser
include Columbia (CGS), FitzPatrick (JAF), Laguna Verde 1 (LAG1), Laguna Verde 2 (LAG2),
Limerick 1 (LIM1), Limerick 2 (LIM2), Nine Mile Point 1 (NMP1), Nine Mile Point 2 (NMP2),
River Bend (RIB), and Vermont Yankee (VY). The ability to control feedwater copper at the
lowest concentrations is clearly a function of the condensate polishing system design at these
plants.

The 2000 average feedwater total copper concentration for each of these plants is plotted in
Figure 3-12. The EPRI Action Level 1 value of >0.2 ppb is shown on the plot. The insoluble,
soluble (non-filterable) and total values are also given in Table 3-3. The results show that the
lowest average values were achieved by plants having Filter + Deep Bed condensate polishing.
This performance is expected since the inlet copper is made up of significant soluble and
insoluble fractions and the Filter + Deep Bed design is capable of efficiently removing both
fractions. The Limerick stations were converted to the Filter + Deep Bed configuration from the
original Filter Demineralizers to improve copper control. Plants with Deep Bed Only condensate
polishing provide better copper control than Filter Demineralizer plants due to the much greater
ion exchange capacity available in the condensate polishing system. Laguna Verde 1 has higher
feedwater copper than the other Filter + Deep Bed plants.

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 EPRI Licensed Material

Iron and Copper Control

1
Feedwater Total Cu (ppb)

0.1

0.01

0.001
NMP2

NMP1
JAF

CGS
LIM2

LAG2

LAG1

VY
LIM1

RIB
DB F+DB F/D ACTION LEVEL 1

Figure 3-12
2000 Average Feedwater Copper for BWRs with Copper Source in Main Condenser

Table 3-3
2000 Average Feedwater Copper (BWRs with copper source in main condenser)

Copper (ppb)
Plant CP Type
Insoluble Soluble Total

Limerick 1 F+DB 0.002 0.004 0.006

Limerick 2 F+DB 0.003 0.007 0.010

Laguna Verde 2 F+DB 0.005 0.025 0.030

Nine Mile Point 2 DB 0.014 0.062 0.076

FitzPatrick DB 0.013 0.076 0.087

Laguna Verde 1 F+DB 0.010 0.108 0.118

River Bend DB 0.037 0.103 0.140

Nine Mile Point 1 DB 0.049 0.110 0.159

Columbia F/D 0.009 0.352 0.361

Vermont Yankee F/D 0.031 0.372 0.404

3-15
 EPRI Licensed Material

Iron and Copper Control

The 2001 average feedwater total copper concentration for all BWRs submitting data,
differentiated by condensate polisher type and main condenser tube material, is plotted in Figure
3-13. Note that that copper concentrations plotted for Dresden 2 and Dresden 3 appear to be
lower limit of detection values, and the actual copper concentrations are expected to be
significantly lower. The insoluble, soluble (non-filterable), and total values for those plants with
a copper source in the main condenser are also given in Table 3-4. The 2001 results also show
that the lowest average feedwater copper values for those plants with copper alloy main
condenser tube material were achieved by plants having Filter + Deep Bed condensate polishing,
with the Limerick units being the lowest of these. Nine Mile Point 2 and FitzPatrick, plants with
Deep Bed Only condensate polishing, achieved lower average feedwater copper concentrations
than Laguna Verde 1, which has filters upstream of deep beds. In 2001, the average feedwater
copper for River Bend was significantly higher than the other plants with Deep Bed Only
condensate polishing. The 2001 average feedwater copper for River Bend was 32% higher than
the 1999 average while the Laguna Verde 1, Nine Mile Point 1 and 2 and FitzPatrick
concentrations decreased over the same period. The Nine Mile Point 1 and 2 copper reduction is
attributed to an increased use of low cross-linked resins and the FitzPatrick copper reduction is
attributed to resin bed replacements. The reason for the Laguna Verde 1 decrease is not known at
this time but is suspected to be due to replacement of old resins having significant copper
loading.

1
Feedwater Total Cu (ppb)

0.1

0.01

0.001
DUA

MON
HCNS

CGS
LIM1

LAS1

LAS2
LIM2

QUA1

LAG2
QUA2

HAT1

HAT2
CLI
PER

OYC

VY
PIL

RIB
JAF
ENF2

NMP2

LAG1

NMP1
DNPS2
BRF3

BRF2
BRU1
BRU2

PB3

DNPS3
PB2

DB (Cu Alloy) DB (Titanium) DB (SS)

F+DB (Cu Alloy) F+DB (Titanium) F+DB (SS)
F/D (Cu Alloy) F/D (Titanium) F/D (SS)
Action Level 1

Figure 3-13
2001 Average Feedwater Copper for BWRs Differentiated by Condensate Polisher Type
and Main Condenser Tube Material

As indicated in Figure 3-12 and Figure 3-13, Filter + Deep Bed plants can readily control
feedwater copper below the EPRI Action Level 1 value. By implementing best practices, Deep
Bed Only plants are also capable of maintaining feedwater copper below the Action Level 1

3-16
 EPRI Licensed Material

Iron and Copper Control

value. The lower achievable feedwater copper control by Filter + Deep Bed plants compared to
plants with Deep Bed Only condensate polishing is attributed to the efficient removal of
insoluble copper by the filters and soluble copper by fixed bed ion exchange (the ion exchange
zone is normally not disturbed during the life of the resins). However, Filter Demineralizer
plants routinely operate with feedwater copper above Action Level 1. The EPRI BWR Water
Chemistry Guidelines (1) recognizes the challenge to Filter Demineralizer plants and states the
following: “An engineering evaluation should be performed before application of this value [the
0.2 ppb feedwater copper limit] at plants with copper alloy condenser tubes and powdered resin
filter/demineralizers, since it may not be achievable without costly plant modifications. In these
circumstances, a limit above 0.2 ppb may be justifiable based on previous performance and core
design considerations.”
Table 3-4
2001 Average Feedwater Copper (BWRs with copper source in main condenser)

Copper (ppb)
Plant CP Type
Insoluble Soluble Total

Limerick 1 F+DB 0.003 0.005 0.007

Limerick 2 F+DB 0.003 0.011 0.014

Laguna Verde 2 F+DB 0.004 0.021 0.024

Nine Mile Point 2 DB 0.013 0.053 0.066

FitzPatrick DB 0.009 0.061 0.071

Laguna Verde 1 F+DB 0.004 0.088 0.092

Nine Mile Point 1 DB 0.030 0.085 0.115

River Bend DB 0.049 0.137 0.186

Columbia F/D 0.012 0.408 0.412

Vermont Yankee F/D 0.035 0.388 0.424

For this same group of North American BWRs with copper alloy condenser tubes, the 1999-2001
average feedwater copper concentrations for the condensate polishing system design types are
shown in Figure 3-14. The results show the strong correlation of feedwater copper with
condensate polishing system design.

3-17
 EPRI Licensed Material

Iron and Copper Control

0.45
0.414 0.418

Average Total Feedwater Copper (ppb) 0.40 0.382

0.35

0.30

0.25

0.20
0.145
0.15
0.116 0.110
0.10
0.041 0.035
0.05 0.026

0.00
1999 2000 2001
Filter + Deep Bed Deep Bed Only Filter Demineralizer

Figure 3-14
1999-2001 Average Feedwater Copper Differentiated by Condensate Polisher Type (for
plants with copper alloy condenser tubes)

The 2000 and 2001 average hotwell copper concentrations are summarized in Table 3-5 and
Table 3-6, respectively. The hotwell total copper average ranges from about 1.7 ppb to >6 ppb
for these plants. The fraction reported as soluble (non-filterable through a 0.45 micron membrane
filter) accounts for as little as 41.6% and as much as 91.5% of the measured total copper. Laguna
Verde 2 has the highest hotwell copper and FitzPatrick has the lowest among the North
American BWRs.

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 EPRI Licensed Material

Iron and Copper Control

Table 3-5
2000 Average Hotwell Copper (BWRs with copper source in main condenser)

Copper (ppb)
Plant CP Type
Insoluble Soluble Total
FitzPatrick DB 0.22 1.52 1.74
Limerick 2 F+DB 0.49 1.90 2.39
Vermont Yankee F/D 0.82 1.61 2.43
Limerick 1 F+DB 0.36 2.35 2.70
Columbia F/D 1.05 2.09 3.14
Nine Mile Point 1 DB 0.31 3.07 3.38
Nine Mile Point 2 DB 0.90 3.70 4.59
Laguna Verde 1 F+DB 1.95 2.77 4.73
River Bend DB 1.92 2.87 4.79
Laguna Verde 2 F+DB 3.43 2.76 6.19

Table 3-6
2001 Average Hotwell Copper (BWRs with copper source in main condenser)

Copper (ppb)
Plant CP Type
Insoluble Soluble Total
FitzPatrick DB 0.19 1.08 1.27
Vermont Yankee F/D 0.81 1.62 2.44
Limerick 1 F+DB 0.41 2.09 2.50
Limerick 2 F+DB 0.51 2.13 2.64
Columbia F/D 1.18 2.09 3.28
Nine Mile Point 2 DB 0.42 3.20 3.62
Nine Mile Point 2 DB 0.74 3.02 3.77
River Bend DB 0.95 2.88 3.83
Laguna Verde 1 F+DB 2.48 2.75 5.24
Laguna Verde 2 F+DB 2.88 3.25 6.13

3-19
 EPRI Licensed Material

Iron and Copper Control

Seasonal Effects

BWR iron data for several years at reactor power greater than 90% were analyzed to determine
the presence of seasonal effects. The average feedwater iron by calendar quarter for all BWRs
by CP type is shown in Figure 3-15.

Deep Bed Only plants show the greatest variation by calendar quarter, with the second and third
quarter averages significantly lower than the first and fourth quarter averages. The first quarter
average feedwater iron for Deep Bed Only plants is 42% higher than the third quarter feedwater
iron. The third quarter is when condensate temperatures typically reach their peak seasonal
values. The seasonal variation is much less apparent for the Filter + Deep Bed plants (some of
which add iron) and Filter Demineralizer plants. There appears to be an indication that Filter
Demineralizer plants experience the lowest quarterly average feedwater iron in the first and
fourth quarters when condensate temperatures are the lowest.

4.0
Average Quarterly Feedwater Iron (ppb)

3.0

2.0

1.0

0.0
1QTR 2QTR 3QTR 4QTR

DB FILTER + DB F/D

Figure 3-15
Average Feedwater Iron by Calendar Quarter

Low Iron and Localized Fuel Corrosion

There have been indications of localized accelerated corrosion of Zircaloy fuel cladding in the
region of the spacers (1). Additional research is required to understand the root cause and factors
affecting the localized corrosion, including water chemistry. The BWR Water Chemistry
Guidelines – 2000 Revision states that one hypothesis for the localized corrosion is iron

3-20
 EPRI Licensed Material

Iron and Copper Control

starvation. It further states that “…it has been suggested that the enhanced localized corrosion
will occur if the concentration ratio of trivalent ions (Fe) to that of divalent ions (Zn, Ni, Cu) on
the fuel cladding surface is less than 2”. Two domestic BWRs have received recommendations
from their fuel supplier to try to maintain the Fe/Zn ratio at or greater than 2:1(2). One foreign
BWR has received a recommendation to maintain its feedwater iron concentration in accordance
with the following (3): 1.5 > [FW Fe] > 2*(0.2 + [FW Zn]), where the units for FW Fe and FW
Zn are µg/kg (ppb). These recommendations are to assure control of localized fuel clad
corrosion, rather than for radiation dose control. The iron starvation hypothesis focuses on the
metal ion concentration ratio at the fuel cladding surface, while the recommendations apply to
total metal concentration ratios in the feedwater. It is not possible to measure the ion
concentrations at the fuel cladding surface but they are measured in the feedwater and bulk
reactor water.

Using 2001 average data from the EPRI BWR Chemistry Monitoring Database, Fe/(Zn+Cu+Ni)
and Fe/(Zn+Ni) ratios for reactor water and feedwater were calculated for both soluble metals
and total metals. Figures 3-16 and 3-17 show the reactor water metals ratios for soluble and total
metals analysis results, respectively. Only one plant had a reactor coolant soluble metals ratio
above 2. Between 5 and 9 plants, depending on the ratio selected (with or without copper), had
reactor water total metals ratios greater than 2.

5.0
4.5
Reactor Water Soluble Metals Ratio

4.0
3.5
3.0
2.5
2.0
1.5
1.0
0.5
0.0
PIL

NMP2

NMP1
JAF

ENF
GGNS
MON

CGS
LAG2
LAG1

PB3

HAT1
PB2
LIM2

HAT2
BRF3

BRF2
VY

OYC
RIB

LIM1
BRU2

BRU1

Fe/(Zn+Cu+Ni) Fe/(Zn+Ni) Ratio = 2

Figure 3-16
Reactor Water Soluble Metals Ratio

3-21
 EPRI Licensed Material

Iron and Copper Control

25
Reactor Water Total Metals Ratio

20

15

10

0 PER

NMP2

PIL
NMP1
JAF

ENF
MON

GGNS

CGS
PB3
PB2

HAT1

LIM2

HAT2

LAG2

LAG1

BRF3
BRF2
OYC

VY
BRU2

LIM1

BRU1

RIB
Fe/(Zn+Cu+Ni) Fe/(Zn+Ni) Ratio = 2

Figure 3-17
Reactor Water Total Metals Ratio

Feedwater metals ratios for soluble and total metals analysis results are shown in Figures 3-18
and 3-19, respectively. Feedwater soluble metals ratios for all plants are less than 2. Between 18
and 22 plants, depending on the ratio selected (with or without copper), had feedwater water total
metals ratios greater than 2.

3-22
 Feedwater Total Metals Ratio Feedwater Soluble Metals Ratio

0
2
4
6
8
10
12
14
16
0.0
0.5
1.0
1.5
2.0
2.5

Figure 3-19
Figure 3-18
HAT1 LIM2
HAT2 HAT1
CGS QUA2
QUA2
HAT2
QUA1
LAG2
BRU2
PB2
LAS1
LIM1

Fe/(Zn+Cu+Ni)

Feedwater Total Metals Ratio

BRF3

Fe/(Zn+Cu+Ni)
BRF3

Feedwater Soluble Metals Ratio

LAS2
VY QUA1
DUA NMP2
NMP2 OYC
LAG2 MON
MON LAG1
BRU1 DNPS
DNPS3 GGNS
PB3 PB3
LAG1 Fe/(Zn+Ni) DUA
EPRI Licensed Material

BRF2

Fe/(Zn+Ni)
DNPS
LIM2 BRF2
LIM1 RIB
PB2
PIL
GGNS
PER
OYC
VY
ENF
PIL CLI
DNPS2 NMP1
RIB JAF
ENF

Ratio = 2
PER
Ratio = 2

JAF CGS
CLI BRU2
NMP1 BRU1

3-23
Iron and Copper Control
 EPRI Licensed Material

Iron and Copper Control

The industry data demonstrate that it is important to identify which ratio, if any, has an effect on
localized accelerated corrosion of fuel cladding. Assuming there is a correlation between a
metals ratio and localized accelerated corrosion, then it is important to define the appropriate
control parameter. If, for example, a plant is adding zinc to maintain a certain zinc concentration
in the reactor coolant to achieve drywell radiation dose goals, then if a minimum ratio of
Fe/(Zn+Ni) or Fe/(Zn+Ni+Cu) is required for fuel corrosion control, iron addition might be
needed.

Transient Iron and Copper

Crud deposition, in the form of metal oxides, on the Zircaloy-2 fuel cladding surface can affect
the corrosion of the cladding (1). In 1999, fuel failures occurred at River Bend (BWR 6). The
failures occurred in heavily crud-loaded areas 20 – 60 inches above the bottom of the fuel (1).
The cycle average feedwater iron and copper concentrations were 3.7 ppb and 0.2 ppb,
respectively, both well below the prevailing EPRI Action Level 1 limits for these corrosion
products. However, the mass and concentration of copper found on the fuel were in the same
ranges associated with CILC failures and the corrosion film thickness was high in the high
power, heavily crudded regions of the core. There was no evidence of nodular corrosion. The
unexpectedly high crud deposition on high-powered fuel bundles may have been due to an additional
high iron source during startup transients (when metals are not typically measured) or preferential
deposition.

It is well known that high (compared to cycle average) concentrations of iron can occur in the
condensate and feedwater systems during outages, startups and other transient conditions. This
represents a potentially high source of iron, and possibly copper, particularly during startup
conditions. Plants normally perform some type of feedwater flush to remove the bulk of the crud
from the feedwater system prior to startup; however, the practices and effectiveness of this flush
vary widely among the plants as discussed in Section 7.

The River Bend fuel failures emphasized the impact of feedwater iron and copper on fuel
performance. As a result, the following revisions were incorporated into the BWR Water
Chemistry Guidelines – 2000 Revision to reduce the power operation limit for feedwater copper
and to add diagnostic parameters for reactor water and feedwater iron indicative of the startup
transient. Since that time, additional fuel failures have occurred at other plants, and the root
cause evaluations of these failures is further complicated by the rapid conversion of plants to the
NMCA chemistry regime.

Available industry data from startup transient conditions that were provided for the EPRI BWR
Chemistry Monitoring Database were evaluated to assess the transient metals impact on metals
deposition on the fuel. These results are presented in Section 4 along with a summary of crud-
related fuel failures and surveillance results from published EPRI technical reports.
Recommendations are also given in Section 4 for revisions to transient iron monitoring for
consideration in the next revision of the EPRI BWR Water Chemistry Guidelines.

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 EPRI Licensed Material

Iron and Copper Control

Perry Jet Pump Fouling

Perry has reported on jet pump fouling experience which results in reactor recirculation loop
flow deviations from expected values based on flow control valve position (4). The A loop and
B loop drive flow differences are shown in Figures 3-20 and 3-21, respectively. These
differences have been attributed to metal oxide crud deposition on the jet pump nozzle surfaces.

The station reported that investigations into this issue indicate that for deposits to occur, certain
conditions must be satisfied. The “zeta potential” (a measure of the surface charge) of the
receptor surface must be the opposite of the zeta potential of the colloidal corrosion products in
the free fluid stream. The pH of the fluid influences this condition and the fluid temperature
must be high. Additionally, the fluid velocity must be high enough to reduce the thickness of the
boundary layer such that free charges that would normally cancel the electric field produced by
the surface are swept away, allowing the surface to attract colloidal products in the free stream.

Low conductivity conditions allow the surface charge to project further into the fluid stream. As
corrosion products begin to adhere to the surface, chemical changes in them may occur, whereby
their charge reverses, and they become the receptors for other particles in the stream.

4
Flow Difference from Expected (kgpm)

-2

-4

-6
12/8/91 12/7/93 12/7/95 12/6/97 12/6/99 12/5/01 12/5/03
A Loop
JP High Pressure Throat & Nozzle Cleaning
NMCA; 88 out of 100 JP Nozzles In-Core Cleaning

Figure 3-20
A Loop Drive Flow Difference From Expected Based On FCV Position

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4
Flow Difference from Expected (kgpm)
2

-2

-4

-6
12/8/91 12/7/93 12/7/95 12/6/97 12/6/99 12/5/01 12/5/03

B Loop
JP High Pressure Throat & Nozzle Cleaning
NMCA; 88 out of 100 JP Nozzles In-Core Cleaning

Figure 3-21
B Loop Drive Flow Difference From Expected Based On FCV Position

Samples taken from the loosely adherent layer on a jet pump nozzle contained 50 wt% Fe2O3
(hematite) and 50 wt % ZnCr2O4 (zinc chromite). Out of Core Cleaning using the High Pressure
Method was performed in Refuel Outage (RFO) 7 (March 1999). In-Core Cleaning of the
nozzles was performed in RFO8 (June 2000).

The station concluded that the higher fouling tendency at Perry than at other plants is attributed
to:
• The plant has had historically low conductivity.
• The jet pump nozzle design has a high flow velocity.
• Low iron values at Perry along with a long history of zinc addition may have created the
potential for higher concentrations of colloidal corrosion products in the reactor coolant.
• Feedwater chrome increased following modifications to a direct contact heater.

Figure 3-20 shows that the A Loop drive flow difference was fairly stable around zero difference
for an extended period following NMCA and in-core nozzle cleaning. Figure 3-21 shows the B
Loop drive flow difference after NMCA and in-core nozzle cleaning to have a similar trend to
the previous cycle trend. An evaluation of cycle data to determine whether NMCA helped
prevent fouling was to be performed, but this may not be definitive since there were several plant
scrams during the cycle. Perry has had no hydrogen injection either prior to or after NMCA.
The next evaluation will test whether hydrogen injection, which started in August 2002, has
helped reduce the rate of fouling.

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References

1. “EPRI BWR Water Chemistry Guidelines – 2000 Revision,” TR-103515-R2, Final Report,
February 2000.

2. “BWR Chemistry Committee Minutes”, Nashville, Tennessee., May 15-17, 2002.

3. Brack, Daniel, “13th BWR Chemistry Meeting”, Nashville, Tennessee, May 15-17, 2002.

4. Doty, Mike, “Jet Pump Fouling at Perry Power Plant”, BWR Chemistry Committee Meeting,
Nashville, Tennessee, May 15-17, 2002

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4
TRANSIENT CORROSION PRODUCTS

Introduction

The adequacy of transient metals control limits and actions as given in the EPRI BWR Water
Chemistry Guidelines – 2000 Revision (1) is examined here. Transient feedwater iron and
copper levels are generally higher than during steady operation, and the impact of these elevated
levels on BWR fuel deposits is evaluated in this section. The objectives of this evaluation are as
follows:
• Determine if the values for startup feedwater and reactor water metals reported by the
industry could result in excessive loading on BWR fuel and increase the potential for crud-
related fuel failures.
• Determine if the current startup diagnostic limits for feedwater and reactor water insoluble
iron are appropriate, if they should be revised, or if new diagnostic or control parameters are
required.
• Identify additional data required to assess the significance of transient metals transport.

This evaluation is driven by the reemergence of the potential role of crud transport on fuel
integrity. From 1992 - 1996, there were no reported crud-related fuel failures. This performance
was attributed to improved chemistry control along with material improvements made with
Zircaloy-2 cladding in the late 1980s and early 1990s. Significant chemistry improvements were
realized at several plants originally designed with filter/demineralizers for condensate polishing
and copper alloy main condenser tubes, when they either converted to titanium tubes or added
deep bed condensate polishers to reduce feedwater copper concentrations, thus reducing the
potential for CILC (crud induced localized corrosion) failures.

Following the River Bend single-cycle exposure fuel failures of 1998 – 1999, the potential for
crud-related fuel failures reemerged as an issue of major concern to the BWR industry. While all
details of the River Bend fuel evaluations and surveillances have not been released, the
occurrence of these failures has stimulated a renewed awareness of the potential impact of metals
deposition on fuel, particularly during transient (startup) conditions. This awareness is reflected
in the EPRI BWR Water Chemistry Guidelines – 2000 Revision, by revising the feedwater
copper Action Level 1 value at power operation (>10% power) from >0.5 ppb to >0.2 ppb and
adding diagnostic measurements for feedwater and reactor water metals during or just prior to
startup.

In 2001 and 2002, a key BWR industry focus area has been the impact of NMCA (Noble Metals
Chemical Applications) on fuel performance. Early results from Duane Arnold and extensive

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laboratory testing showed no significant impact on fuel performance, but recent inspections at
Duane Arnold, Peach Bottom, and Hatch show thicker deposits and evidence of spalling.
Evaluations are in progress to assess the impact and determine the likely causes of these
observations.

Approach

Key areas of evaluation involve the following;

• Review historical industry data to establish a baseline for fuel crud deposition.
• Review historical industry data on fuel failures, identifying those where crud loading was
determined to be a contributing cause (excluding instances of debris fretting failures).
• Perform mass balances for selected plants to compare calculated crud loadings to historical
baseline measured values at steady state and deposition rates of metals on fuel during
transient (startup) conditions.
• Assess the adequacy of the startup diagnostic limits in the BWR Water Chemistry Guidelines
– 2000 Revision (1) and subsequent BWR industry monitoring practices during startups.

Summary of EPRI Guidelines on Feedwater and Reactor Water Metals

The River Bend fuel failures emphasized the potential impact of feedwater iron and copper on
fuel performance. As a result, the following revisions were incorporated into the BWR Water
Chemistry Guidelines – 2000 Revision for iron and copper control:
• The feedwater copper Action Level 1 value at power operation (>10% power) was reduced
from >0.5 ppb to >0.2 ppb. This change was to address the apparent synergy between
elevated levels of feedwater copper and iron leading to fuel failures at River Bend. River
Bend data show that although feedwater iron and copper were below the prevailing Action
Level 1 limits, the levels were on the high side compared to industry averages. The basis for
the reduced copper limit is that the 0.5 ppb limit may be too high for plants operating with
feedwater iron >3 ppb (the average feedwater iron concentration at power operation at River
Bend leading up to the fuel failures was 3.7 ppb).
• A diagnostic parameter was added for startup/hot standby to measure reactor water insoluble
iron prior to initiation of significant feedwater flow to the reactor. This value can be used to
quantify the insoluble iron inventory that could deposit on the fuel during power ascension.
• A diagnostic parameter was added for startup/hot standby to measure feedwater/condensate
insoluble iron prior to initiation of significant feedwater flow to the reactor or at completion
of cleanup during long path recycle/flush. A target value of <100 ppb is suggested. A high
pressure heater drain target value of <100 ppb is also suggested prior to pumping the drains
forward at plants with forward pumped drains.

Many plants have adopted these guidelines and EPRI efforts to collect and evaluate data are
ongoing. For the reactor water and feedwater iron diagnostic parameters, the guidelines suggest
that analysis by a color comparator test is acceptable. However, the results from this test are

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normally not entered in the plant chemistry electronic database and are thus not provided in the
data routinely supplied to the EPRI BWR Chemistry Monitoring database. In addition, some
plants do not sample for feedwater metals during startups until approximately steady full power
is attained. While there are questions concerning the representativeness of samples at low
power, such results are important because they provide the data needed to determine the
significance of crud transport to the reactor and deposition on fuel surfaces during startup
transient conditions.

Background and Literature Review

EPRI studies on fuel surveillance programs conducted at a number of BWR stations over the past
30 years were reviewed. The early industry efforts focused on the causes of crud induced
localized corrosion (CILC), while more recent efforts are focused on the impact of the noble
metals chemical application (NMCA) process on fuel performance.

History of BWR Fuel Issues

A status summary of fuel failures occurring at U.S. BWRs since 1989 is given in TR-100978 (2).
A summary of the number of crud-related and non crud-related BWR fuel failures from 1989 -
1999 is given in Table 4-1. The results show a period of five years from 1992 – 1996 where no
CILC or crud related failures were reported. Furthermore, the CILC failures reported in 1997
and 1998 at Browns Ferry appear to be an anomaly, since the particular failed fuel had
previously been identified as susceptible to CILC failures; this fuel was manufactured in the
early 1980s and had been placed back in service in the late 1990s following a long shutdown.
Excluding the 1997 - 1998 failures, there were no crud-related failures from 1992-1998. The
seven failures attributed to crud and corrosion in 1999 occurred at River Bend. The totals given
in Table 4-1 include non crud-related failures, the attributed causes of which are debris fretting,
pellet clad interaction (PCI), manufacturing defects, or undetermined.
Table 4-1
Crud-Related and Total BWR Fuel Failures, 1989 – 1999 (Number of Failed Fuel
Assemblies)

Cause 1989 1990 1991 1992 1993 1994 1995 1996 1997 1998 1999

CILC 52 5 3 0 0 0 0 0 3 46 0

Crud &
0 0 0 0 0 0 0 0 0 0 7
Corrosion

Non Crud-
5 10 21 12 16 15 4 14 5 7 6
Related

Total 57 15 24 12 16 15 4 14 8 53 13

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River Bend Experience

Limited published data are available on the crud-related fuel failures that occurred at River Bend
in the late 1990s. A discussion is provided in the BWR Water Chemistry Guidelines - 2000
Revision (1) and is summarized below.

River Bend is a BWR-6 with deep bed condensate polishing, pumped-forward heater drains and
copper-alloy condenser tubes. In 1999, fuel failures were observed on high duty, single cycle
exposure fuel. Examinations showed heavily crud loaded regions between 20 and 60 inches
from the bottom of the fuel. The crud was found to contain 5 - 17 % zinc and 2-15% copper,
with the bulk of the remainder being iron crud. The highest copper concentrations were found at
the cladding-crud interface. The crud zinc concentration was reported to be typical for a zinc
injection plant, but the crud copper concentration was considered to be high for a plant with deep
bed condensate polishers and a copper-alloy condenser (1). Cycle average feedwater iron and
copper concentrations were within the EPRI guideline limits (3.7 ppb iron, 0.2 ppb copper). A
clear correlation was found between crud mass, measured cladding oxidation, and local power
level, since only the high power fuel rods experienced damage and there was no evidence of
corrosion nodules, which were identified as a precursor to CILC failures (1).

The mass of the crud deposited on the River Bend fuel was reported to be much greater than
expected for a plant with 3.7 ppb average feedwater iron. It was consequently reasoned that
either iron preferentially deposited on the high power fuel surfaces or another source of iron was
present. The source of copper is stated to be from the main condenser (1).

Additional information from River Bend, which was also documented by INPO (3), provides
some further insights. During the refueling outage just before the cycle where the fuel failures
were detected, a reactor coolant chemistry excursion occurred following the chemical
decontamination of an RHR (Residual Hear Removal) heat exchanger. Reactor coolant
conductivity reached a maximum value of 6.1 µS/cm when the RHR system was restored to
service and did not return to normal values until three days later. At this point in the outage, the
new fuel had been placed in the vessel. The RHR heat exchanger tubes at River Bend are
reported to be 70/30 copper/nickel. No analysis results of the reactor coolant were available to
determine the chemical constituents that caused the elevated conductivity. However, the
reference states that difficulties in the chemical decontamination process due to the heat
exchanger tube material resulted in the liberation of copper ions and oxalate precipitates during
the process. A second reactor coolant chemical excursion occurred during the subsequent plant
startup, with reactor water conductivity increasing to 0.946 µS/cm, which was attributed to
inputs from the pumped-forward feedwater heater drains system.

The River Bend outage and startup chemical excursions are identified as initiating events
ultimately leading to the fuel failures (3). The synergistic effect of the chemical excursions
along with somewhat elevated feedwater iron and copper concentrations led to the high crud
buildup on the fuel. The high local power in conjunction with the insulating properties of the
crud produced a high temperature at the cladding surface, causing accelerated oxidation of the
cladding. Copper in the crud layer deposited on the fuel lowered the thermal conductivity of the
deposits, thus resulting in high clad temperatures.

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Duane Arnold (DAEC) Post-NMCA Fuel Monitoring

The results through 2000 of an ongoing surveillance program at DAEC to examine the
effectiveness of NMCA and fuel effects have been published (4). DAEC is a condensate
filter/demineralizer plant with stainless steel main condenser tubes and has been injecting
depleted zinc oxide since 1994. As part of the surveillance program, fuel deposit samples were
obtained and analyzed.

Figure 4-1 is a plot of the reported total fuel deposition results for fuel that has been exposed
from one to three cycles (4). The total loading is the sum of crud obtained from fuel brushing
and fuel scraping samples. The metal present in the highest weight percent in the crud samples is
from iron, followed by zinc and nickel. The annual average feedwater iron at DAEC ranged
from 1.7 to 0.9 ppb from 1998-2000. Iron represents about 77% (by weight) of the total fuel
deposit with zinc accounting for about 18%. The brushing results, indicative of the loose outer
layer crud deposit, accounted for 62% of the total iron deposit for the single cycle exposure fuel
and increased to account for 75% on the three cycle exposure fuel. In contrast, brushing
removed only 34 - 49% of the zinc and 42 - 65% of the nickel. Similar to iron, the percentages
of zinc and nickel in the loose outer layer of crud increased with increasing cycles of exposure.
The data also show that most of the noble metals (66 - 71% for platinum and 73 - 96% for
rhodium) were in the loose outer layer. The peak noble metals and total crud loadings occurred
about 30 inches from the bottom of the fuel rods. The report concludes that the applications of
noble metals at DAEC have not altered the distribution of the brush/scrape deposits and the
concentration and composition of the deposits (4).

Visual observations given in an earlier report (5) on Duane Arnold that also includes fuel crud
liftoff measurements following three cycles of NMCA are summarized as follows:
• Three-cycle exposed fuel (the first cycle was completed before NMCA) appeared to have
heavier corrosion than baseline fuel (prior to initial NMCA), particularly at the lower
elevations (10 to 40 inches).
• Two-cycle exposed fuel (untreated rods exposed to noble metals reapplication) showed
similar results as the three-cycle exposed, plant treated fuel.
• Two-cycle exposed fuel (pre-treated rods that were also exposed to noble metals
reapplication) appeared to have heaviest corrosion of any of the previously inspected fuel
rods at DAEC. Corrosion was reported not to be excessive as would be evident by oxide
spalling.

Corrosion layer thickness is reported as Rod Average Lift-off (RALO) and Maximum Effective
Lift-off (MELO), which are measurements of the rod average and maximum corrosion,
respectively (5). The results are summarized below:
• RALO values after two-cycle exposure are reported to be within the band of historical data
obtained at DAEC and other BWRs.
• MELO values after two-cycle exposure (plant treated, first cycle without NMCA) were
reported to be within the historical BWR experience. However, the average MELO for these
rods was 0.5 mils higher than the baseline DAEC fuel measurements prior to NMCA.

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• MELO values after two-cycle exposure (NMCA for both cycles) were 0.7 mils higher than
the baseline DAEC measurements prior to NMCA.
• MELO values for the pretreated rods were 0.7 mils higher than the historical BWR levels.

1000

100
Deposition, mg/dm^2

10

0.1
Cycle 16 Cycle 15-16 Cycle 14-16

Fe Cu Zn Ni Pt Rh

Figure 4-1
Duane Arnold Fuel Deposition Loading (Sum of Brushing and Scraping)

The report (5) concludes that after two cycles of operation, the corrosion behavior of the fuel
rods is acceptable for continued operation. However, the lift-off data indicate that NMCA may
have a small effect on fuel rod corrosion, primarily on the lower portion of the rod. The visual
inspection results are reported to confirm the lift-off measurement trends.

Industry Updates on Post-NMCA Fuel Monitoring

At this time, no known fuel failures have been attributed to NMCA. However, initial evaluations
suggest that the fuel failures at Browns Ferry 2 and Vermont Yankee after NMCA may be
corrosion related.

New data from Duane Arnold following the first cycle of operation after noble metals
reapplication appear to show oxide spalling on some examined rods and thicker corrosion layers
than expected. At Peach Bottom 2, which received the second highest noble metal loading of
any BWR, spalling and a thicker than expected layer of corrosion products was observed on
some rods with three-cycle exposed fuel after one cycle under NMCA. At Hatch 1, oxide
spalling was observed at some fuel spacer locations. No firm conclusions were drawn on the
cause of the unexpected observations and measurements. Additional data are to be obtained via
hot cell examinations as well as additional poolside examinations at other NMCA plants. An

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interim recommendation is under consideration to limit the average noble metals loading during
application to no more than 30 µg/cm2 (based on fuel surface area) and to avoid reapplication to
the same fuel batch.

Fuel Surveillance Results at Hatch prior to NMCA

A multi-cycle fuel sampling program was conducted at the two Hatch units and the results were
documented in a 1994 report (6). Over 1000 fuel samples were collected and analyzed over the
course of the program. The program covered three cycles of operation at each unit, which
allowed the effects of significant chemistry changes that occurred during this period on fuel
performance to be examined. The chemistry changes consisted of starting hydrogen water
chemistry and GEZIP, and re-tubing the main condensers from brass to titanium.

Changes influencing primary water chemistry at Hatch 1 and Hatch 2, and the fuel cycle when
each change was implemented, are summarized in Table 4-2. Both Hatch units have filter
demineralizers for condensate polishing and cascaded heater drains. They are high power-
density plants (BWR 4s), and had past fuel failures attributed to CILC.
Table 4-2
Changes Influencing Primary Chemistry at Each Hatch Unit

Change Hatch 1 Hatch 2

HWC Cycle 12 (1) Cycle 10

GEZIP Cycle 13 Cycle 9

Condenser re-tube Cycle 13 Cycle 9

Table 4-2 Notes

1. Some testing of HWC was performed during Cycle 11.

Hatch Feedwater Chemistry Trends

Cycle average feedwater iron, copper and zinc concentrations for the three-cycle surveillance
program for each unit are plotted in Figures 4-2 and 4-3 (6).

Hatch Unit 1

The average feedwater iron trend over the surveillance period showed a steady increase from just
over 1 ppb to almost1.8 ppb. A steady increase in feedwater iron had been observed since cycle
9 (6). In cycle 12, feedwater iron began to increase shortly after the initiation of HWC.
Hydrogen was injected at 22 scfm, but there was frequent cycling of the hydrogen injection
system, and this may have contributed to the increases in feedwater iron seen during the first
cycle with HWC. It is also reported that the condensate filter demineralizer inlet iron steadily

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increased during the study period from an average value of about 17.4 ppb in cycle 12 to an
average value of about 24 ppb in cycle 14 (6).

The condenser was re-tubed in the refuel outage following cycle 12 (June 1990), and zinc
injection was initiated with NZO (August 1990) shortly after the start of cycle 13. NZO
injection began shortly after the cycle 13 startup to compensate for eliminating the natural zinc
source from the brass condenser tubes. Iron concentrations in cycle 13 continued to increase,
compared to the cycle 12 numbers. The report cites that unpublished studies show that BWRs
that have replaced brass condenser tubes with titanium experience increased levels of feedwater
iron during initial plant operation immediately following the tube replacement (6). HWC, which
had been in operation at 22 SCFM since 1987, was operated at the reduced flow of 17 SCFM in
cycle 13, and this change may have resulted in some higher transient iron levels observed in
cycle 13 (6). During cycle 14, hydrogen was injected at 17 SCFM with good availability but was
returned to 22 SCFM for the last six months of the cycle. Iron levels were higher in the last six
months of the cycle and this was attributed to the higher hydrogen levels (6).

During cycle 12, the last cycle with the brass condenser tubes, feedwater copper was reduced by
a factor of about two (from 0.7 down to 0.3 ppb) during the first few months following the start
of HWC (6). Copper then increased to about 0.5 ppb before leveling out at about 0.3 ppb for the
remainder of the fuel cycle. The decrease in copper concentration was coincident with the
implementation of HWC and was attributed to a reduction in the oxygen concentration of the
main steam, which resulted in a reduction in the corrosion rate of the brass condenser tubes (6).

Feedwater copper decreased significantly following re-tubing of the condenser. In the two
cycles following the condenser tube replacement, the feedwater copper concentration was
reported to be about 0.1 ppb, excluding values seen during power transients (6).

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2.0

1.8

1.6

1.4
Concentration, ppb

1.2

1.0

0.8

0.6

0.4

0.2

0.0
Cycle 12 Cycle 13 Cycle 14

Fw Fe FW Cu Fw Zn

Figure 4-2
Hatch 1 Feedwater Metals Trend

Hatch Unit 2

Feedwater iron at Hatch 2 increased over the surveillance program period, as shown in Figure 4-
3. The increase from cycle 8 to 9 was relatively small, while the increase from cycle 9 to 10 was
significantly higher. At Unit 2, both zinc injection with NZO and operation with titanium
condenser tubes commenced at the beginning of cycle 9, while HWC operation was started in
cycle 10. The increase in feedwater iron reported in cycle 10 is attributed to the effects of HWC.

Feedwater copper at Unit 2 was reported to be on a declining trend during Cycle 8, the operating
cycle before the condenser re-tube (6). By the end of the cycle, copper was about 0.3 ppb, with a
cycle average value of approximately 0.5 ppb. The condenser was re-tubed with titanium during
the subsequent refuel outage and copper values for Cycle 9 were significantly reduced. Copper
continued to trend down through Cycle 9 and into Cycle 10, with the average value
approximately 0.1 ppb in Cycle 10.

Natural zinc injection was introduced about 8 months after startup from Cycle 9. Although zinc
levels in the reactor water were observed to have increased following the start of NZO injection,
this increase was not reflected in feedwater zinc samples (6). The goal for GEZIP was to
maintain a reactor coolant zinc concentration of 6 ppb. Following the start of HWC in Cycle 10,
the coolant concentration goal was increased to 10 ppb. Again, an increase in zinc concentration
was measured in reactor water but not in feedwater. No explanation was available for this
anomaly.

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1.8

1.6

1.4
Concentration, ppb

1.2

1.0

0.8

0.6

0.4

0.2

0.0
Cycle 8 Cycle 9 Cycle 10

Fw Fe FW Cu Fw Zn

Figure 4-3
Hatch 2 Feedwater Metals Trend

Hatch Fuel Inspection and Deposition Data

As part of the surveillance program, visual inspections, thickness measurements, and crud
deposition measurements were made for selected fuel rods (6). Single and multi-cycle exposure
bundles were selected in an attempt to correlate chemistry changes to data from fuel
examinations.

Visual examinations were made with an underwater TV camera. Thickness measurements (or
ROXI for rod oxide thickness measurement system) were made using eddy-current type of
equipment. The results from ROXI are reported in mils, as rod average lift off (RALO). RALO
is the average of individual liftoff measurements along the length of a rod after brushing. These
measurements were made after brushing so the remaining tenacious crud layer will contribute to
high RALO results. Crud sampling was done using a nylon brush to remove the loosely adhered
layer and an aluminum stone to remove the tightly adhered layer.

The visual examinations of Unit 1 fuel are summarized as follows:

• Single cycle exposure fuel (Cycle 12) was observed to be in excellent condition with 0 - 20%
nodular oxide coverage. Most of the visible nodules were observed to have initiated along
surface scratches.
• Single cycle exposure fuel (Cycle 13) had significantly more nodule coverage (up to 80%)
than on the single cycle fuel exposed in Cycle 12.

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• Two-cycle exposure fuel (Cycle 13) was observed to have only a moderate increase in nodule
coverage compared to the single cycle exposure rods at the end of Cycle 12. The oxide
nodules appeared coarser and more diffuse than the single cycle exposure rods at the end of
cycle 12.
• One, two and three cycle exposure fuel (Cycle 14) was observed to have similar size and
distribution of nodules.

The visual examinations of Unit 2 fuel are summarized as follows:

• Single cycle exposure fuel (Cycle 8) was observed to be in excellent condition with little
nodule oxide coverage. Most of visible nodules were along surface scratches.
• Single cycle exposure fuel (Cycle 9) had significantly more nodule coverage (up to 40%)
than single cycle exposure fuel from Cycle 8.
• Two-cycle exposure fuel (Cycle 9) had fewer nodules than the single cycle exposure fuel.
Some of the nodules were surrounded by halos, which were identified as areas of
discoloration. Halos were more common above the 40-inch level of the rods. Rods from one
bundle that were inspected in Cycle 8 were also inspected in cycle 9. The predominant
feature observed was the halo effect.
• Two and three-cycle exposure fuel (Cycle 10) was observed to have similar nodule coverage
as the one and two cycle fuel after Cycle 9, with the three cycle exposure fuel having more
halo coverage. The same rods inspected at the end of the previous two cycles were also
inspected at the end of cycle 10. More halo coverage was observed as well after three cycles
than after two cycles.

RALO measurements for Unit 1 fuel are summarized in Table 4-3. The single cycle
measurements after the first HWC cycle was significantly lower than the single cycle
measurement after the first cycle with the condenser re-tube and the initiation of GEZIP. Also,
the second cycle measurement for fuel that had seen HWC, condenser re-tube and GEZIP was
higher than the second cycle measurement for fuel that had seen one cycle of HWC and one
cycle of HWC, GEZIP, and the condenser re-tube. The highest corrosion thickness generally
occurred within the first 60 inches of rod length, as measured from the bottom spacer. The
report also cites that the high liftoffs measured for the two and three cycle exposure fuel could be
due to the measurement bias (6).
Table 4-3
Hatch 1 RALO Measurements

EXPOSURE CYCLE LIFTOFF (mils)

1 Cycle 12 0 - 0.4
1 Cycle 13 0.51- 0.71
2 Cycle 13 0.47 – 0.78
2 Cycle 14 2.13 - 3.84
3 Cycle 14 1.69 - 2.94

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RALO measurements for Unit 2 fuel are summarized in Table 4-4. The largest increase was
seen from Cycle 9 to Cycle 10. HWC was implemented during cycle 10. The three cycle
exposure rods were found to have the peak thickness between 60 and 90 inches from the bottom
of the bundle.
Table 4-4
Hatch 2 RALO Measurements

EXPOSURE CYCLE LIFTOFF (mils)

1 Cycle 8 0 – 0.2
1 Cycle 9 0.25 – 0.64
2 Cycle 9 0.26 – 0.55
2 Cycle 10 1.07 - 1.76
3 Cycle 10 0.85 - 1.45

Hatch 1 fuel crud loading for single cycle and multi-cycle exposed fuel are shown in Figures 4-4
and 4-5, respectively (6). The data presented are the sum of the brush and scrape sample results
for each element. The results show that brushing removed between 51-74% of the iron, 21-37%
of the copper, and 28-49% of the zinc, indicating that the tightly adherent inner layer was more
concentrated in copper and zinc than the loose outer layer. This was true for both the single cycle
and multi-cycle exposure fuel.

The Hatch 1 copper deposit weight decreased significantly on the single cycle fuel from Cycle
14 compared to that of the previous cycles, as shown in Figure 4-4. The three cycle exposure
fuel shows only a modest increase in copper from the previous multi-cycle values. These effects
are attributed to the re-tubing of the condenser with titanium. The report (6) also cites early
Hatch data (no specific date was given, but is believed to be at the time of CILC failures) where
scraping results showed copper deposition as high as 60 mg/dm2. This is about a factor of 30
higher than the highest deposition rate for a single cycle exposure bundle shown in Figure 4-4.

Iron represented between 73 to 92% of the total crud (by weight) on the fuel. The highest
percentages were observed prior to GEZIP initiation, while the lowest percentages were
observed after GEZIP initiation. Copper represented between 0.3 and 2.7% of the total crud on
the fuel, with the highest percentage observed prior to the condenser retubing, and the lowest
percentage was observed on the single cycle exposure bundle examined at the end of cycle 14.

Mass balances were performed using the fuel deposition data to determine the fraction of
feedwater metals that deposits on the fuel. The report states that fuel deposit experience shows
that 70 - 90% of the feedwater crud deposits on fuel surfaces, the balance being removed by the
RWCU system or is deposited on non-fuel surfaces (6). For Unit 1, bundles sampled after cycle
12 and 13 show an average of 79% of the feedwater iron deposited on the fuel. Data from
bundles sampled after cycle 14 showed this number to be only slightly above 50%. One reason
given for the lower estimated deposition rate was possible inaccuracies in metals measurements
during cycle 14 (6). A similar finding was also reported for copper as well as zinc. The reports
states that the deposition of soluble species like copper and zinc is different than insoluble

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species like iron and thus, the 80-90% deposition rule of thumb on fuel surfaces is not expected.
The calculated percentages of feedwater copper and zinc depositing on the fuel surfaces, based
on the fuel crud samples, averaged about 3.1% for copper and 32% for zinc, with the highest
reported values being 6% for copper and 51% for zinc. Only a small portion of the feedwater
copper entering the reactor vessel could be accounted for in the crud deposit samples and the
RWCU system. The report cites the possibility that artificially high feedwater copper values
could have resulted from sample line contamination.

1000

100
Deposition, mg/dm^2

10

0.1
Cycle 12 Cycle 13 Cycle 14
Fe Cu Zn Ni Total

Figure 4-4
Hatch 1 Single Cycle Exposure Fuel Deposition

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1000

100
Deposition, mg/dm^2

10

0.1
Cycle 12,13 Cycle 13,14 Cycle 12,13,14

Fe Cu Zn Ni Total

Figure 4-5
Hatch 1 Multi-Cycle Exposure Fuel Deposition

Unit 2 fuel crud deposition results for single and multi-cycle exposures are shown in Figures 4-6
and 4-7. The Unit 2 trends are similar to those of Unit 1, except that the iron crud loading for the
single cycle exposure bundle in cycle 10 was higher than the single cycle exposure bundles in the
previous cycles. This may be due to higher feedwater iron in cycle 10 that in the previous two
cycles, as shown in Figure 4-3.
The Unit 2 crud composition was also similar to that of Unit 1. Iron accounted for 77 - 89% of
the crud, copper 0.4 - 3.4%, and zinc 4.8 - 15.8%. Reported mass balance results indicated that
about 70% of the feedwater iron deposited on the fuel, along with about 3% of the feedwater
copper and 28% of the feedwater zinc.

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1000

100
Deposition, mg/dm^2

10

0.1
Cycle 8 Cycle 9 Cycle 10

Fe Cu Zn Ni Total

Figure 4-6
Hatch 2 Single Cycle Exposure Fuel Deposition

1000
Deposition,mg/dm^2

100

10

1
Cycle 8,9 Cycle 9,10 Cycle 8,9,10

Fe Cu Zn Ni Total

Figure 4-7
Hatch 2 Multi-Cycle Exposure Fuel Deposition

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Swedish Experience

Results were reported from a program conducted in Sweden, where various Zircaloy-2 claddings
from different manufacturers were tested both in and out of reactor (7). The in-reactor testing
was conducted in all of the Swedish BWRs. Key results and conclusions are as follows:
• The copper fraction in fuel crud increases with oxide thickness. This trend was observed in
plants with high as well as low feedwater copper. The explanation is that the white oxide
formation begins as individual nodules that eventually spread out and cover the cladding
surface. Cracks in the white oxide layer allow the penetration of soluble ions like copper into
the depth of the oxide layer. Iron and other insoluble particles cannot penetrate through the
crack, and thus deposit on the oxide surface. As the oxide layer thickens, the amount of
soluble crud that penetrates the oxide layer increases, while the amount of insoluble crud that
deposits on the oxide surface remains about the same. Data cited from measurements made
on Ringhals 1 fuel in 1980 showed high crud loading (90 - 200 mg/dm2) with a copper
fraction of 50%. The crud was tightly adhered with heavy oxide spalling and an oxide
thickness of 10 - 150 microns. Ringhals 1 is a filter/demineralizer plant that initially had
brass main condenser tubes and was operating under NWC conditions.
• Higher crud deposition occurred on areas with thicker layers of white zirconium oxide.
• Frequent cycling of hydrogen water chemistry may increase the corrosion rate of fresh fuel
since the first protective oxide layer is not properly formed due to the frequent change in
oxidation potential. This conclusion was based on the measured oxide thickness of three
plants, all of which had very low feedwater copper (< 0.03 ppb). The oxide thickness of the
plant that was using HWC during its first cycle was reported to be a factor of five higher after
25,000 EFPH than two other plants that were not using HWC during their first cycle.

Correlated Causes of CILC-Related Fuel Failures

A study published in 1990 investigated the link between chemical inleakage and accelerated fuel
cladding corrosion based on the data of five BWRs that had experienced CILC failures (8). The
participating plants were not identified by name, but were those that had experienced CILC
related failures in the 1980s. Chemical inleakage refers to any type of excursion that adversely
affects reactor coolant chemistry. Examples of excursions cited are significant conductivity
increases, high condenser air inleakage, condenser tube leaks, organic intrusions from recycled
water and events that could result in abnormal corrosion product transport into the reactor vessel
from the feedwater system. The approach of this study was first to hypothesize the stages in the
corrosion process of the Zircaloy cladding, then model how chemical inleakage would affect the
specific corrosion stages. This was followed by the development of correlation coefficients and
testing of the correlations with data from another plant not used for developing the correlations.

A hypothesis for the Zircaloy corrosion stages was initially formulated as follows:
• White oxide nodule formation on a thin black protective film of zirconium dioxide. The rate
of formation is a function of the oxidation potential set forth by radiolysis, and thus is a
function of neutron flux, the extent of severe chemical intrusion events, and cladding
material condition.

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• Lateral spreading of the nodules to form a complete sheet of thicker white oxide. The rate of
spreading is a function of the number of nodules formed during the first stage, the oxidation
potential due to radiolysis, the cumulative supply of oxidants, and cladding material
condition.
• Thickening of the white oxide film to the point of cracking perpendicular to the cladding
surface. The rate of thickening is reported to be a function of oxidation potential from
radiolysis and material condition of the cladding.
• Deterioration of the thermal conductivity of the white oxide layer, the rate of which is a
function of copper deposition in cracks and power density. Copper is proposed to deposit in
the cracks through “wick boiling”, where coolant enters the crack, copper is deposited, and
the coolant leaves the crack as steam, thus creating countercurrent water/steam flow in the
crack.
• Through-wall corrosion to failure. It is postulated that as the thermal resistance to heat
transfer increases, the temperature at the crud/cladding interface increases significantly.
Once cracking of the oxide film occurs and copper deposits in the crack, it may take only a
few months to observe a through-wall failure.

The initial hypothesis of the corrosion stages assumed that copper played a role only in the
lowering of the thermal conductivity of the oxide layer once cracking occurred. However, the
report concludes that the role of copper in the latter stages of the corrosion process could not be
substantiated (8). The data and correlations suggest that the key role of copper may be in
accelerating the early life oxidation rate. The rapid drop in the thermal conductivity of the white
oxide observed in the latter stages that ultimately leads to through-wall corrosion may be the
result of steam blanketing in cracks and spalling of the zirconium oxide, and that the deposition
of copper in the white oxide structure would only be the result of cracking and steam blanketing
and not a contributing cause of the failure.

A significant finding of this study is a correlation between chemical transients during early core
life and cumulative oxidants with the growth rate of the white zirconium oxide layer, and thus
the time to reach a critical thickness for through-wall corrosion failure. Specifically, severe
chemical transients occurring within the first 2900 MWD/MT of operation (within the first few
months of an operating cycle) for a given batch of fuel correlated with peak oxidation rates and
incidents of localized corrosion failures observed in later core life. A severe conductivity
transient is defined as an excursion of >0.3 µS/cm, greater than a week in duration, and occurring
at high core power. A severe copper intrusion is not quantitatively defined, but it is
acknowledged that operating with a full power feedwater average concentration of >0.3 ppb
could accelerate the oxidation rate.

Summary and Mechanisms of Past Crud-Related Fuel Failures

A 1985 study reviews the experience of historical BWR and PWR crud-related fuel failures and
focuses on the effects of chemical and non-chemical parameters on fuel crud deposition and
build-up (9).

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Crud-induced fuel failures were reported in the mid 1970s at Japan’s Tsuruga BWR, which
began commercial operation in 1970 with only deep beds for condensate polishing. Failures
occurred in the fuel cladding primarily at and between the second and third spacers from the
bottom. The axial distribution of the crud layers was peaked toward the bottom of the
assemblies. Thick deposits were observed to have accumulated at the spacer-to-fuel rod
interfaces. The source of the crud was postulated to be from the precipitation of colloidal iron
oxides and hydroxides at the point where the feedwater mixes with reactor water in the vessel.
Precipitation of colloidal iron was attributed to very low feedwater oxygen (reported as
undetectable) in conjunction with very high feedwater iron, in the range from 10 to 50 ppb. The
proposed mechanism is that under very low oxygen conditions, feedwater iron is more likely to
be in the Fe3O4 form and when the iron comes in contact with oxygenated reactor coolant in the
mixing zone (approximately 200 ppb oxygen under NWC), the solubility of the iron decreases
resulting in the precipitating a gelatinous iron hydroxide or colloidal Fe2O3. The colloidal
material formed can then deposit on the lower tie plates and spacers of the fuel assemblies.

Fuel deposits were examined on an unfailed bundle that had 26,400 EFPH of operation (17.2
GWD/MTU). The average iron deposition on the fuel cladding was about 1400 mg/dm2, with a
reported maximum value of 5000 mg/dm2. The crud was composed of 98% iron, 0.5% nickel,
and 0.3% copper.

A review of operational chemistry data showed that during 1970 – 1972, the first two years of
commercial operation, feedwater iron levels were relatively low, averaging <5 ppb. During the
next three years, feedwater iron levels ranged from 10 to 50 ppb. In 1976, feedwater iron
returned to <5 ppb. The reactor coolant iron trend followed the feedwater iron trend over the
same time period, with reactor water levels ranging up to 100 ppb during the 1972 – 1975 period.
The initial increase in feedwater iron (1973) was attributed to operating with severe air
inleakage, resulting in an increase in feedwater system materials corrosion. The air inleakage
was subsequently repaired, resulting in no detectable feedwater dissolved oxygen. It was also
reported that in 1976 feedwater iron decreased to <5 ppb following the installation and operation
of a feedwater oxygen injection system.

The SGHWR (Steam Generating Heavy Water Reactor) in Dorset, England had between 20 and
30 failed fuel elements during the first 100 to 200 days of operation. Crud thickness was the
highest near the bottom of the assembly with a maximum reported crud thickness of 100 microns
on non-failed fuel. The crud was described as forming two layers, a dense, reflective, ferro-
magnetic layer adjacent to the cladding surface and a porous agglomerate on the surface. The
composition of the dense layer on the non-failed fuel was primarily CuO (40-65%), along with
Fe2O3 and Fe3O4. Prior to the failures, reactor coolant iron and copper levels were reported to be
approximately 200 ppb and 100 ppb, respectively.

Crud sampling in the early 1970s identified fuel crud deposits at the KRB-A reactor containing
high amounts of copper at a burn-up of 22,000 MWD/MTU. Crud samples from non-failed fuel
showed a peak thickness of 150 microns about 1 meter from the bottom of the fuel with an
adherent layer consisting of 45% copper, 11% iron, and 3 % nickel. An outer loose layer
consisted primarily of iron with little copper.

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The study addresses reasons why adherent crud tends to be detrimental to fuel integrity (9).
Dense, adherent crud layers tend to crack due to thermal expansion of the fuel. The cracks tend
to be perpendicular to the heat transfer surface and offer higher heat transfer resistance than the
crud layer itself. The increased heat transfer resistance leads to increased zircaloy oxidation.
Loose crud, although it may be thicker, has a significantly lower thermal resistance because of its
high porosity; the loose structure permits it to expand and contract without cracking.

The report also cites results of heat flux testing of artificial iron oxide crud on a stainless steel
tubing, some of which was impregnated with copper (9). The addition of copper to the crud was
reported to have consolidated the crud, resulting in a significant increase in the heat transfer
resistance. The measured tube wall temperatures were significantly higher with the crud
containing copper than with pure iron oxide crud.

Crud loading sample results were given for Brunswick 2, which has forward-pumped heater
drains (9). At EOC 1, the measured deposition was 400-700 mg/dm2 for a middle of the core
bundle and 1100-3600 mg/dm2 for a peripheral bundle. The source of the high iron loading was
attributed to corrosion of carbon steel piping from the high pressure heater drains system. Final
feedwater iron averaged about 13 ppb for the operating cycle. No fuel failures occurred during
this cycle as a result of the high crud loading.

Monticello and Nine Mile Point 1 data were evaluated to correlate fuel crud loading with plant
operating chemistry data (9). Monticello and Nine Mile Point 1 were chosen primarily because
of significant differences in their condensate treatment systems, with Monticello having
condensate filter/demineralizers and Nine Mile Point 1 having only deep bed condensate
demineralizers. Both plants had several operating cycles but differed significantly in fuel crud
loading, with Monticello considered a low crud input plant and Nine Mile Point 1 a high crud
input plant. Both plants also had a significant copper source at the time from their copper alloy
(brass) condenser tubes.

Reported fuel crud deposition data for Monticello and Nine Mile Point 1 are presented in Figures
4-8 and 4-9, respectively. The Monticello data are representative of 32,830 EFPH and the Nine
Mile Point 1 data are representative of 26,000 EFPH. In both plots, the x-axis gives the fuel rod
length, as measured from the bottom (inlet) to the fuel bundle. For Monticello, actual iron and
copper deposition results were provided while for Nine Mile Point 1 the copper results were
reported as a percentage of the total deposition after a given number of EFPH. For Nine Mile
Point 1, after 26,000 EFPH, the iron deposition represented 94% of the total mass deposited,
while copper represented 2%. These percentages were used to estimate the iron and copper
loading results given in Figure 4-9.

The following observations are reported based on the Monticello data (9):
• The rod appearance was “rather clean” with reddish-brown corrosion products.
• A tightly adherent layer was covered with a loosely adherent layer. Similar observations
were made in previous fuel cycle examinations.
• Monticello had low amounts of fuel deposits over the first five cycles, with peak levels
ranging from 20-200 mg/dm2 of iron.

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• The copper content of the deposits increased after the first and second fuel cycles, but the
reason for this is undetermined since no feedwater copper measurements were made during
the first three cycles of plant operation. The maximum value after 32,830 EFPH was a factor
of 5 higher than the maximum value after the first fuel cycle. Feedwater copper values were
reported to be generally <1 ppb during cycles 4 and 5, with some values as high as 1.5 ppb.
Reactor coolant copper data ranged from 10 - 20 ppb. The condensate filter/demineralizers
were reported to have a removal efficiency of about 50% for soluble copper in fuel cycle 5.
• Mass balances using fuel deposition data after 10,170 EFPH indicated that about 63% of
insoluble feedwater iron deposited on the fuel.

The following observations are based on the Nine Mile Point 1 results (9):
• Nine Mile Point 1 had higher levels of deposits on the fuel over the first three cycles than
Monticello, with peak levels ranging from 600 - 1000 mg/dm2 of iron.
• The high iron deposits on the fuel in the first three cycles are related to high feedwater iron,
which ranged from 10 to 100 ppb, with peak values >100 ppb during the first cycle of
operation. Following the first cycle of operation, feedwater iron was reported to be normally
<10 ppb.
• Feedwater copper was reported to be <0.1 ppb in the first three cycles of operation. Reactor
water was not analyzed for copper at that time. The report concludes that the low
concentration of copper in the fuel deposits indicates that reactor coolant copper levels were
under good control.
• Mass balances after 12,100 EFPH indicate about 84% of insoluble feedwater iron depositing
on the fuel.

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100
90
80
Deposition, mg/dm^2

70
60
50
40
30
20
10
0
0 20 40 60 80 100 120 140 160
Distance from Bottom, inches

Fe Cu

Figure 4-8
Monticello Fuel Deposition Data after 32,830 EFPH

1000
Deposition, mg/dm^2

100

10

1
0 20 40 60 80 100 120 140 160
Distance from Bottom, inches

Fe Cu

Figure 4-9
Nine Mile Point 1 Fuel Deposition Data after 26,000 EFPH

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An earlier study, published in 1979 (10) concluded the following regarding fuel crud deposition
from the data of a number of plants:
• Deposits on BWR fuel are primarily Fe2O3 with minor amounts of Fe3O4 (results under
normal water chemistry conditions).
• Deposits forming on fuel elements in deep bed condensate polishing plants are flocculent and
easy to remove, while deposits forming on filter/demineralizer plants are thinner and more
tenacious.
• The fuel deposit loading peaks some 50 to 100 cm above the bottom of the fuel bundle,
which has a total length of approximately 350 cm (140 inches). Nine Mile Point 1 results
showed deposition of >1000 mg/dm2 near the bottom to less than measurable at the bundle
outlet (top) for burn-ups ranging from 3.8 to 7.7 GWD/MTU.

Typical crud deposition and composition for the two types of condensate polishing designs are
presented in Table 4-5 (10). The results representative of deep bed plants are after 20,000 EFPH
operation while the data representing filter/demineralizer plants are after 16,000 EFPH. The
total deposition results reflect superior removal efficiency of corrosion products by
filter/demineralizers than by deep bed demineralizers. Insoluble iron represents the largest
contribution to the hotwell (condensate pump discharge) corrosion products. The higher
percentage of iron in the fuel deposit of deep bed plants reflects the poorer removal of insoluble
corrosion products with this type of condensate polishing compared to condensate
filter/demineralizers. The higher percentage of copper in the deposits of filter/demineralizer
plants is attributable to the poorer removal of inlet copper (approximately 50% - 70% soluble) by
the powdered resin precoats than by the deep bed demineralizer resins.
Table 4-5
Composition of Fuel Deposits at BWRs (from EPRI NP-522)

PARAMETER DEEP BED FILTER/DEMINERALIZER

Deposition (mg/dm2) 500 150

Iron (wt. %) 94 65
Copper (wt. %) 1 4

Cobalt (wt. %) 0.1 0.4

Zinc (wt. %) 0.3 14

Nickel (wt. %) 3 6

Others (wt. %) 2 10

Deposition data are also presented for Oskarshamn 1, a Swedish BWR with filter/demineralizers
in the condensate system and a brass condenser (10). The data, summarized in Table 4-6, show a
two-layer corrosion deposit consisting of a loosely held outer layer and a more adherent inner
2
layer. After 26,000 EFPH, the measured deposits ranged from 2 to 60 mg/dm with the peak
deposits occurring about 100 cm from the inlet end.

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Table 4-6
Oskarshamn 1 Fuel Deposition (from EPRI NP-522)

ELEMENT LOOSE LAYER (wt. %) ADHERENT LAYER (wt. %)

Iron 68 - 76 47 - 69
Copper 3.9 - 9.0 8.7 - 23

Zinc 8.9 - 11 12 - 13

Chrome 1.7 - 3.7 2.0 - 5.8

Nickel 7.1 - 8.3 6.2 - 8.4
Manganese 2.0 - 2.2 2.1 - 3.0

Iron and Copper Data Evaluation

Mass balance calculations at approximately full power were performed for ten plants using
chemistry data from the BWR Chemistry Monitoring Database to estimate deposition rates of
iron and copper on the fuel. One purpose of these calculations was to provide a baseline of
steady state full power deposition results against which transient startup deposition could be
compared. Plants were selected based on the following criteria:
• Include plants with copper alloy main condenser tubes.
• Ensure that all three types of condensate treatment systems are represented (deep bed only,
filter + deep bed, and filter/demineralizer).
• Susquehanna 1 was selected because of historically high feedwater iron prior to the
installation of condensate pre-filters. There have been no corrosion product related fuel
failures at Susquehanna. Only data prior to the installation of pre-filters were evaluated.
• River Bend was selected because of fuel failures related to crud.
• Peach Bottom 2 was selected because of the high noble metals loading.
• Other plants were selected based on having adequate feedwater and reactor water chemistry
data available to perform mass balance calculations.

Data were evaluated for the specific plants listed in Table 4-7. The table also gives some related
plant design characteristics. The chemistry regime for each plant during the period of data
evaluation is shown in Table 4-8.

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Table 4-7
Design Features of Plants Selected for Mass Balance Calculations

Condensate Condenser Tube Feedwater Drains

Plant
Treatment Design Materials System Design

Brass and Copper-

Columbia F/D Cascaded
Nickel
FitzPatrick DB Brass and Titanium Cascaded
Limerick 1 Filter + DB Brass Cascaded
Limerick 2 Filter + DB Brass Cascaded
Copper-Nickel and
Nine Mile Point 1 DB Cascaded
Stainless Steel
Brass and Copper-
Nine Mile Point 2 DB Pumped-Forward
Nickel
Peach Bottom F/D Titanium Cascaded
River Bend DB Brass Pumped-Forward
Susquehanna 1 DB Stainless Steel Cascaded
Brass and
Vermont Yankee F/D Cascaded
Stainless Steel

Table 4-8
Chemistry Regimes During Data Evaluation Period at Plants Selected for Mass Balance
Calculations

Hydrogen Water Noble Metals Zinc

Plant
Chemistry Chemistry Injection

Columbia No No Yes
FitzPatrick Yes Yes Yes
Limerick 1 Yes Yes Yes
Limerick 2 Yes Yes Yes
Nine Mile Point 1 Yes Yes No
Nine Mile Point 2 Yes Yes Yes
Peach Bottom Yes Yes Yes
River Bend Yes No Yes
Susquehanna 1 No No No
Vermont Yankee No Yes Natural

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Steady State Loading Calculations

The mass balance for metal “i” over specified time increment ∆t can be written as follows:

∆ Rx Mass (i) = FW(i) – RWCU(i) - D(i)

∆t

Where: ∆ Rx Mass (i) = the change in reactor coolant mass of metal i over a
∆t given time increment (lb/hr)
FW(i) = mass rate input of metal i entering the reactor from
the feedwater (lb/hr)
RWCU(i) = mass rate of a metal i removed by the RWCU system
(lb/hr)
D(i) = mass rate of a metal i removed by deposition on in-core and out
of core surfaces (lb/hr)

This mass balance can also be solved explicitly for the concentration of metal “i” in the reactor
coolant as follows:

CP(i) = CFW(i) * F
V(α+β)

Where: CP(i) = Reactor coolant concentration for a given element (i), ppb
CFW(i) = Feedwater concentration for a given element (i), ppb
F = Feedwater flow, lbs/hr
V = Coolant mass at operating temperature, lbs
α = total deposition rate, hr-1
β = RWCU removal rate, hr-1

Ample feedwater metals analyses were available for the mass balance calculations. Since these
are integrated samples, they represent a virtually continuous indication of feedwater metals
concentrations averaged over each 2 to 3 day period when metals sample collection filters are in
service. The main uncertainty in the overall quality of feedwater metals data is associated with
inadequate sample-line flow because the sampling systems at most plants operate with a sample
line velocity that is substantially lower than the recommended 6 ft/sec required for minimizing
sample-line effects.

Reactor water metals samples are considered to be less reliable than feedwater data. Sampling
frequencies for reactor water metals vary widely; as frequently as three times per week to as
infrequently as once per quarter. Reactor water samples are usually grab samples and sample
line velocities are generally less than 6 ft/sec.

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The calculations assume a 95% removal efficiency of the RWCU system for soluble and
insoluble metals from the reactor coolant. Although metals are normally not analyzed in the
RWCU effluent, removal efficiencies of at least 95% are expected. An actual removal efficiency
greater than 95% would not have a significant impact on the mass balance calculations.

The time period over which the data for each station were evaluated depended upon the
frequency of sampling reactor coolant metals and the duration of steady operation at or near full
power. The mass of each metal entering the reactor with the feedwater was determined for each
time increment. The mass removed by RWCU and the increase in inventory in the coolant over
the time increment were subtracted from the total entering with the feedwater, and the difference
gave the total mass deposited on primary surfaces. Ninety percent of the mass depositing on
primary system surfaces during the time increment were assumed to deposit on the surface of the
fuel.

To compare all results on the same basis, the total deposition for a two year fuel cycle with 20
outage days (17,040 full power hours) was calculated for each station, even though some stations
have yet to implement a two year cycle.

Results and Observations

The results of the iron mass balance calculations for the selected plants are summarized in Table
4-9. Similarly, the mass balance results for copper are given in Table 4-10. The iron deposition
results are given per unit of fuel surface area and the reactor coolant and feedwater iron values
are the average for the evaluation period. For Nine Mile Point 1 and 2 and River Bend the
calculations were performed using data from different fuel cycles. At both Nine Mile Point
units, significant improvements have been made in lowering feedwater iron with the use of low
crosslinked resin over the past few years. For River Bend, data from three fuel cycles were used,
including data from 1998 leading to the fuel failures encountered later in the operating cycle.

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Table 4-9
Mass Balance Results for Iron Deposition on Fuel (normalized to 24-month fuel cycle with
20 outage days)

Average Iron (ppb)

Iron Deposition,
Plant Reactor Feedwater
mg/dm2

Columbia 146 3.1 0.8

Fitzpatrick 176 8.4 1.6
Limerick 1 100 9.7 0.8
Limerick 2 108 4.2 0.7
Nine Mile Point 1 (2001) 103 46.0 1.8
Nine Mile Point 1 (1998) 214 15.5 2.4
Nine Mile Point 2 (2001) 113 15.6 1.0
Nine Mile Point 2 (2000) 222 5.7 2.1
Peach Bottom 2 158 3.1 1.3
River Bend (2001) 267 35.6 2.8
River Bend (2000) 400 42.9 4.1
River Bend (1997-98) 442 47.9 4.0
Susquehanna 1 1254 24 8.4
Vermont Yankee 129 7.7 1.1

Table 4-10
Mass Balance Results for Copper Deposition on Fuel (normalized to 24-month fuel cycle
with 20 outage days)

Copper (ppb)
Copper Deposition,
Plant mg/dm2
Reactor Feedwater

Columbia 48.5 12.0 0.40

Fitzpatrick -11.0 11.9 0.07
Limerick 1 0.3 0.7 0.07
Limerick 2 0.8 0.4 0.01
Nine Mile Point 1 (2001) 5.4 1.4 0.11
Nine Mile Point 1 (1998) 9.1 2.2 0.15
Nine Mile Point 2 (2001) 1.4 1.9 0.04
Nine Mile Point 2 (2000) 9.5 0.8 0.10
Peach Bottom 2 0.8 1.3 0.02
River Bend (2001) 19.8 4.6 0.25
River Bend (2000) 14.4 1.6 0.15
River Bend (1997-98) 22.4 2.5 0.20
Susquehanna 1 Very low <LLD <LLD
Vermont Yankee 38.9 7.8 0.4

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Observations from the steady state results are as follows:

• With the exception of Susquehanna 1, the calculated iron deposition values are within the
historical ranges reported in the literature review.
• The calculated copper deposition values are higher than the historical ranges as reported in
the literature review. This may be an artifact of the assumptions that 90% of the copper
deposition occurs on the fuel surfaces and the 24 month fuel cycle with a high capacity
factor. The maximum calculated copper deposition value (48.5 mg/dm2 for Columbia) is
roughly a factor of two higher than the maximum reported deposition value for Hatch 1 and 2
prior to condenser re-tubing (when equated to the same operating time).
• The plants with brass condenser tubes and filter/demineralizers (Vermont Yankee and
Columbia) have the highest copper deposition. These plants also have the highest feedwater
copper values.
• FitzPatrick shows a negative deposition for copper. This is attributed to a large variance in
reactor coolant copper concentrations over the data analysis period, even though only data for
approximately full power were used. Several full power reactor water copper values greater
than 50 ppb were included in the data analysis. The negative deposition may imply copper
release from the fuel surface, sample line contamination or problems with obtaining
representative samples.
• The highest steady state copper deposition at River Bend occurred during the cycle having
the crud related fuel failures. While copper deposition at River Bend was significantly lower
than at Columbia or Vermont Yankee, River Bend iron deposition was significantly higher.
• The use of low cross-linked resins at both Nine Mile Point plants has decreased iron and
copper deposition. Nine Mile Point 1 1998 and 2001 results show that both iron and copper
deposition were cut by about half after low cross-linked resins were implemented.

Plant Startup Data

The EPRI BWR Chemistry Monitoring Database indicates that few plants report startup metals
data. Plants normally do not report quantitative feedwater metals sampling and analysis results
from early startup and the ensuing power ascension period. The same is also true for reactor
water metals.

A sufficient number of startup metals measurements for meaningful mass balance calculations
during the power ascension were available from two plants, River Bend and Columbia. For
River Bend, the startup data included side-by-side collection of reactor and feedwater samples.
For Columbia, reactor water samples were collected during the power ascension but feedwater
sampling was only performed at 75% power. Mass balance calculations were performed as
previously described. For Columbia calculations, it was assumed that the reported 75% power
feedwater iron of 4.27 ppb and feedwater copper of 0.25 ppb were constant during the power
ascension. This assumption probably underestimates the deposition, since higher feedwater
concentrations are expected earlier in the power ascension.

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River Bend

Reactor water and feedwater metals concentrations for River Bend during the April 2000 startup
are presented in Table 4-11 and Figure 4-10. The data show that feedwater iron concentration
was less than 100 ppb prior to establishing significant feedwater flow. Reactor water iron was
much greater than 100 ppb during early startup (<10% power).

The iron and copper deposition rates and cumulative deposition on the fuel during the startup are
shown in Table 4-12 and in Figures 4-11 and 4-12. The calculations show a total iron deposition
of 7.3 mg/dm2 and total copper deposition of about 0.2 mg/dm2. This amount of iron is
equivalent to 1.8 - 2.7% of the estimated fuel cycle deposition at steady state for River Bend.
The transient copper deposition is equivalent to 0.9 - 1.4% of the deposition calculated for the
fuel cycle during steady full power operation.

The maximum startup deposition rates for iron and copper are compared to the steady power
operation values in Table 4-13. While the total mass deposited on the fuel surface during a
startup is small compared to the total deposition during a complete fuel cycle, the maximum
startup iron deposition rate is indicated to be a factor of two to three higher than the steady power
operation rate. The maximum startup copper deposition rate is 10 – 75% higher than the cycle
steady power deposition rate.

Negative deposition rates were calculated for iron at 2% and 7% power corresponding to the
initial 17 and 46 hours of startup, respectively. Reactor water iron values at these times were
high, indicating an apparent crud release from the fuel.
Table 4-11
River Bend April 2000 Startup Metals Data

Reactor Reactor
Power Cumulative Reactor Water Feedwater Feedwater
Water Iron
Time (hours) Copper (ppb) Iron (ppb) Copper (ppb)
(%) (ppb)

0 0 304 41 6.9 0.21

2 17 1374 11 14.7 0.18
7 42 524 13.8 19.8 0.37
17 65 56 7.4 16.5 0.55
45 89 103 4.2 16.5 0.55
67 109 107 5.9 3.8 0.31
75 113 76 3.9 13.6 0.31
100 200 80 3.3 7.5 0.14

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10000

1000
Concentration, ppb

100

10

0.1
0 10 20 30 40 50 60 70 80 90 100
Power, %

Rx Fe Rx Cu FW Fe FW Cu

Figure 4-10
River Bend April 2000 Startup Metals Data

Table 4-12
River Bend April 2000 Startup Mass Balance Results

Iron Cumulative Copper Cumulative

Deposition Iron Deposition Deposition Copper
Plant Cumulative
Rate Rate Deposition
Power Hours into
mg/dm2/hr mg/dm2 mg/dm2/hr mg/dm2
(%) Startup

0 0 0 0 0 0

2 17 -0.008 -0.143 0.0002 0.004

7 42 -0.0001 -0.148 0.0003 0.011

17 65 0.021 0.33 0.0004 0.020

45 89 0.037 1.22 0.0010 0.048

67 109 0.031 1.85 0.0014 0.077

75 113 0.042 2.02 0.0014 0.083

100 200 0.060 7.31 0.0012 0.197

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0.10 10
Iron Deposition Rate, mg/dm^2/hr 0.09 9

Cumulative Deposition, mg/dm^2

0.08 8
0.07 7
0.06 6
0.05 5
0.04 4
0.03 3
0.02 2
0.01 1
0.00 0
-0.01 -1
0 20 40 60 80 100
Power, %
Deposition Rate Cumulative Deposition

Figure 4-11
River Bend April 2000 Startup Iron Deposition Trends

0.0016 0.40

0.0014 0.35

Cumu;ative Deposition, mg/dm^2

Copper Deposition Rate,

0.0012 0.30

0.0010 0.25
mg/dm^2/hr

0.0008 0.20

0.0006 0.15

0.0004 0.10

0.0002 0.05

0.0000 0.00
0 20 40 60 80 100
Power, %

Deposition Rate Cumulative Deposition

Figure 4-12
River Bend April 2000 Startup Copper Deposition Trends

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Table 4-13
River Bend 100% Power Average and Startup Maximum Iron and Copper Deposition Rates

2001 2000 1998

Power Level / Metal 2 2
mg/dm /hr mg/dm /hr mg/dm2/hr

100% / Iron 0.017 0.023 0.026

Startup / Iron Not available 0.060 max. Not available

100% / Copper 0.0011 0.0008 0.0013

Startup / Copper Not available 0.0014 max. Not available

Columbia

The Columbia startup sampling and mass balance results are shown in Tables 4-14 and 4-15 and
in Figures 4-13, 4-14 and 4-15. The mass balance calculations for Columbia assumed that the
feedwater metals measured at 75% power represented the value over the entire startup, since this
single sample provided the only feedwater data for this startup. The highest measured reactor
coolant iron concentration was 850 ppb, which occurred at 75% power. Two reactor water
samples were taken 8 hours apart prior to startup. Iron was greater than 100 ppb in the first
sample and decreased to less than 100 ppb in the second sample. The highest measured reactor
coolant copper concentration was at 71 ppb, which was on a pre-startup sample.

The startup data evaluation indicates cumulative iron deposition of 3.3 mg/dm2 and cumulative
copper deposition of 0.115 mg/dm2. The highest iron deposition rate, 0.15 mg/dm2/hr, occurred
at 80% power while the highest copper deposition rate, 0.0047 mg/dm2/hr, occurred at 80 - 100%
power. A comparison of these results with the steady full power operating cycle results indicates
that the maximum startup iron deposition rate is a factor of 18 higher than the steady power rate
while the maximum startup copper deposition rate is a factor of approximately 1.6 higher than
the steady power value. The cumulative iron deposition for this startup is 2.2% of the deposition
that occurs during the steady power operation portion of the fuel cycle while the cumulative
startup copper deposition is 0.24% of the cumulative deposition during steady power operation
(Table 4-10).

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Table 4-14
Columbia July 2001 Startup Metals Data

Reactor Water Reactor Water

Plant Cumulative Time Iron Copper
Power (%) into Startup (hours) (ppb) (ppb)
0 0 225 14.2
0 8 88 71.7
15 16 54.6 34.2
17 24 13.7 16.1
17 32 7.5 13.3
22 40 13.8 11.8
24 48 7.29 8.4
49 56 50 19
64 64 66.8 12.7
75 72 850 7.8
79 80 73 22.5
80 112 29.7 20.1
100 120 94 32.2

1000
Concentration, ppb

100

10

1
0 10 20 30 40 50 60 70 80 90 100
Power, %

Rx Fe Rx Cu

Figure 4-13
Columbia July 2001 Startup Metals Data

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Table 4-15
Columbia July 2001 Startup Mass Balance Results

Cumulative Iron Cumulative Copper Cumulative

Time into Deposition Iron Deposition Copper
Plant Rate Deposition Rate, Deposition
Startup
Power (%) (hours) (mg/dm2/hr) (mg/dm2) (mg/dm2/hr) (mg/dm2)
0 0 0 0 0 0
0 8 0 0 0 0
15 16 0.014 0.11 0.0004 0.004
17 24 0.036 0.40 0.0010 0.018
17 32 0.037 0.70 0.0013 0.021
22 40 0.037 0.99 0.0014 0.032
24 48 0.037 1.3 0.0010 0.041
49 56 0.035 1.7 0.0008 0.048
64 64 0.033 1.6 0.0014 0.059
75 72 -0.005 1.83 0.0009 0.066
79 80 -0.001 1.79 0.0014 0.077
80 112 0.15 3.0 -0.00001 0.077
100 120 0.008 3.3 0.0047 0.115

0.16 4.5

0.14 4.0
Iron Deposition Rate, mg/dm^2/hr

Cumulative Deposition, mg/dm^2

0.12 3.5

0.10 3.0

0.08 2.5

0.06 2.0

0.04 1.5

0.02 1.0

0.00 0.5

-0.02 0.0
0 10 20 30 40 50 60 70 80 90 100
Power, %
Deposition rate Cumulative Deposition

Figure 4-14
Columbia July 2001 Startup Iron Deposition Trends

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0.005 0.12

Cumulative Deposition, mg/dm^2

0.004 0.10
Copper Deposition Rate,

0.003 0.08
mg/dm^2/hr

0.002 0.06

0.001 0.04

0.000 0.02

-0.001 0.00
0 20 40 60 80 100
Power, %

Deposition Rate Cumulative Deposition

Figure 4-15
Columbia July 2001 Startup Copper Deposition Trends

Conclusions from Transient Metals Evaluation

The adequacy of the startup metals limits and guidance in the BWR Water Chemistry Guidelines
– 2000 Revision was evaluated by reviewing published data on industry fuel failures, reviewing
fuel corrosion and crud deposition measurements and performing mass balances using actual
plant steady state and startup transient metals and operating data to estimate cumulative crud
deposition and deposition rates on fuel. The following conclusions can be drawn from this work:

1. The startup/hot standby diagnostic parameter for feedwater iron, normally taken at the
conclusion of the feedwater flush, is appropriate because it provides confirmation of the
removal of bulk loose crud from condensate and feedwater piping and equipment prior to
initiating feedwater flow to the reactor. This has the benefit of reducing the mass of crud that
would otherwise be transported to the reactor. However, differences in feedwater flush
practices make it difficult to compare the effectiveness from plant to plant. Many plants have
adopted the feedwater and reactor water diagnostic parameters for iron prior to the start of
significant feedwater flow during a plant startup. Plants normally meet the startup diagnostic
parameter of <100 ppb for feedwater iron based on sampling during feedwater flush
operations.

2. Feedwater/condensate metals samples taken during feedwater flush operations are not
indicative of feedwater metals concentrations during startup. During the transient conditions
of a plant startup, changes occur in condensate and feedwater system parameters that are not
experienced during feedwater flush operations and that have a large impact on metals

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concentrations. Examples are increasing feedwater velocities, rapid velocity fluctuations in

feedwater piping and equipment due to control responses to maintain reactor coolant level
and starting of pumps, increasing temperature and reduction of feedwater dissolved oxygen
from air-saturated levels (from several ppm to <100 ppb). In most plants, the first reactor
feed pump is started at <10% power, but the second feed pump is usually not started until
50% power. At some plants, the condensate pumps, condensate booster pumps and feed
pumps are slaved to the electric power grid and thus are not placed in service until the
generator is online. Starting these pumps creates flow transients that can cause significant
releases of corrosion products from feedwater system surfaces. Also, starting hydrogen
injection during power ascension may further affect dissolved oxygen levels in the
steam/condensate and thus affect corrosion product concentrations.

3. The diagnostic parameter added for the startup/hot standby condition to measure feedwater
and reactor water insoluble iron prior to initiation of significant feedwater flow provides only
an imperfect indication of the quantity of metals that could deposit on the fuel during power
ascension. Detailed feedwater and reactor water data taken during power ascension by two
plants indicate periods of metals release from the fuel and significant variations in deposition
rates.

4. Reactor coolant chemistry excursions during a refueling outage and/or during the subsequent
startup power ascension period are potential initiating events that can increase the
vulnerability to crud-induced fuel failures later in the fuel cycle. Historical evaluations of
fuel failures in the 1980s with the cause attributed to CILC found a strong correlation
between early cycle chemical excursions (occurring within the first 2900 MWD/MT of
operation) and the occurrence of fuel failures later in the operating cycle. Although the
mechanism of the 1998 - 1999 River Bend fuel failures appears to differ from that of
historical CILC failures because the presence of nodule oxide formation was not evident at
River Bend, the occurrence of a chemistry transient during refueling or early startup
operations is a common factor.

5. Measurements suggest that copper plays a significant role in fuel failures related to metals
transport. Copper may accelerate fuel cladding corrosion early in fuel life and/or may lower
the thermal conductivity of the deposit. Fuel examinations show two distinct crud layers, a
loose outer layer and an adherent inner layer. The outer layer is predominantly composed of
iron while the inner layer contains higher concentrations of zinc and copper than the outer
layer. Measurements at Swedish BWRs show that the copper fraction in the fuel crud
increases with oxide thickness whether feedwater copper is high or low, and that the copper
is found primarily in the inner layer.

6. Only the Tsuruga fuel failures of the 1970s were attributed strictly to high iron crud
deposition, which exceeded 1000 mg/dm2 and occurred under NWC conditions. However,
the iron deposition threshold above which fuel integrity can be affected under today’s HWC
and NMCA conditions, where sections of the fuel are exposed to a reducing chemistry
environment, and with higher-powered cores from power uprates, is not known. The
Tsuruga fuel failures of the 1970s were attributed to high iron crud deposition (>1000
mg/dm2) following operation with high feedwater iron in conjunction with very low
feedwater dissolved oxygen. Mass balance calculations performed for Susquehanna 1, prior

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to the installation of condensate filters and before starting moderate HWC, show an average
iron loading of 1250 mg/dm2 over a two-year operating cycle with no crud-related fuel
failures.

7. Evaluation of startup metals data for River Bend and Columbia indicate that, based on a 24-
month fuel cycle, total iron deposition for the startup transient is 1.7% – 2.8% of the steady
state deposition and transient copper deposition is 0.24% - 1.4% of the steady state copper
deposition. However, the Columbia analysis is based on only a single feedwater corrosion
product sample obtained during the startup. Neither of these plants was on HWC or NMCA
at the time.

8. Mass balance calculations on available detailed plant startup metals data indicate
significantly higher iron and copper deposition rates during plant startups than during steady
full power operating conditions. At River Bend and Columbia, which provided detailed
startup data, the peak iron deposition rate occurred between about 80% and 100% power and
the peak copper deposition rate occurred between 45% and 100% power. The peak transient
iron deposition rate ranged from 2 to 18 times the steady state rate, and the peak transient
copper deposition rate was 10% to 75% higher than the steady state rate.

9. Most plants normally do not perform reactor coolant and feedwater metals analyses during
power ascension. Reasons for not performing such sampling and analysis are:
– Samples taken during startup may not be representative of the process liquid due to
low sample line velocity and sample tubing conditioning uncertainties.
– Integrated samples taken with a corrosion products sampler may overload the
membrane filter discs and therefore not yield quantifiable results.
– Samples taken during power ascension at >10% power are not considered to be
representative of steady state power operating conditions and elevated results will
challenge station Chemistry Indicator goals.
– The extra sampling and sample preparation work adds extra burden, particularly for
stations that must use the labor intensive digestion process of corrosion product
samples for analysis by atomic absorption (AA) or inductively coupled plasma (ICP)
spectroscopy. For analyses performed by XRF, sample digestion is not required.

Recommendations

1. The diagnostic parameter for insoluble iron prior to the initiation of significant feedwater
flow given in Table 4-4 of the BWR Water Chemistry Guidelines – 2000 Revision at the
startup/hot standby condition should be revised to a control parameter with a limit of 100
ppb. For practical purposes, this parameter is measured at the conclusion of the feedwater
flush, which is performed during cold shutdown conditions. Consideration should therefore
be given to adding a table for feedwater/condensate – cold shutdown that would include this
and any additional parameters. The requirement for high pressure heater drain insoluble iron
to be <100 ppb prior to routing these drains forward for plants with pumped-forward drains
should remain in Table 4-4.

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2. The current diagnostic parameter for reactor water insoluble iron given in Table 4-4 of the
BWR Water Chemistry Guidelines – 2000 Revision at the startup/hot standby condition
should be maintained. Although available data indicate that this measurement alone is not
adequate to quantify the cumulative amount or rate of metals deposition on fuel during the
startup transient, such detailed startup data are needed to develop and support appropriate
revisions.

3. As a matter of practice, plants should be encouraged to perform concurrent reactor water and
feedwater metals sampling for soluble and insoluble metals during power ascension from
early startup through 100% power to further assess startup metals deposition and deposition
rates on fuel. Obtaining these samples should be made part of the chemistry startup plan.
Having these samples allows an evaluation to be performed, as needed, should fuel issues
that could be related to crud deposition emerge later in the fuel cycle. The startup sampling
should be planned and executed because it is not possible to know with certainty beforehand
whether any significant chemistry transients (which have been associated with past fuel
failures along with metals deposition) will occur during startup or early in the cycle. While it
is preferable that startup feedwater and reactor water metals samples be analyzed soon after
they are taken, plants could alternatively choose to preserve the samples for future analysis as
needed. The burden of analyzing startup samples should be less for plants that have
implemented XRF metals analysis technology.

4. With EPRI assistance, plants should develop an appropriate strategy for startup sampling
strategy and for evaluating the analyses to quantify startup metals deposition and the
influence of factors such as power ascension rate, hold points, startup chemistry excursions
and chemistry regime. The results of this effort would be used in developing further
guidance on transient metals monitoring and control. A sampling plan should be developed,
addressing when reactor water and feedwater sampling should start, volumes to be collected
for each sample station, and sample flow rates. Detailed startup data should be obtained
from a cross-section of plants representative of different condenser tube materials (copper
and non-copper), feedwater iron levels (low iron, high iron, iron addition), chemistry regimes
(HWC, HWC/NMCA), zinc levels and power densities. Reactor and feedwater metals
samples should be collected concurrently. Additional data should be recorded, such as
feedwater and reactor water dissolved oxygen, reactor thermal power, condensate flow,
feedwater flow and operating logs that show when equipment is started or placed in service.

References

1. “EPRI BWR Water Chemistry Guidelines – 2000 Revision,” TR-103515-R2, Final Report,
February 2000.

2. “BWR Failed Fuel Degradation, An EPRI Perspective,” TR-100978, December 2000.

3. “Water Chemistry Induced Fuel Leaks," INPO Significant Event Report SEN #204,
September 20, 1999.

4. “BWRVIP-89, BWR Vessel and Internals Project NMCA Materials Surveillance and Fuel
Deposit Sampling Program at Duane Arnold Energy Center (DAEC)”, December 2000.

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5. “Post NMCA Fuel Surveillance Program Duane Arnold Energy Center: RFO16 Fuel
Surveillance Program (BWRVIP-62)”, EPRI TR-1000162, July 2000.

6. “Fuel Surveillance through Condenser Change-out at Hatch 1 and Hatch 2 Reactors,” EPRI
TR-104702, December 1994.

7. “Corrosion Performance of Zircaloy-2 Cladding”, EPRI NP-6707-L, April 1990.

8. “Investigation of In-Leakage and Fuel Corrosion in Certain BWRs,” EPRI NP-6779-SL,

March 1990.

9. “Corrosion Product Buildup on LWR Fuel Rods”, EPRI NP-3789, April 1985.

10. “Survey of Corrosion Product Generation, Transport, and Deposition in Light Water
Reactors,”EPRI NP-522, March 1979.

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5
DRYWELL RADIATION DOSE RATES

Monitoring of Drywell Radiation Dose Rates

BRAC Monitoring Program and Practices

The BWR fixed point survey program, commonly referred to as BRAC (BWR Radiation Level
Assessment and Control), is discussed in General Electric document NEDC-12688, which was
issued in 1977 based on work sponsored jointly by General Electric and EPRI (1). The intent of
the BRAC program is to establish a consistent set of fixed survey points in order to monitor
radiation buildup, review plant operational and design factors for effect on dose rates, and to
provide reference data input to radiation buildup modeling. The BRAC program specifies
locations, frequency, timing, and instrumentation for periodic fixed-point radiation dose rate
surveys of BWR primary system components in order to provide consistent and comparable data.

Although all BWR units routinely collect some form of BRAC data during plant shutdowns,
there is considerable variability in the survey and data collection processes. The major
contributions to this variability are the survey method and equipment used, the time after
shutdown at which surveys are taken, and in-plant equipment configuration. There is a need in
the industry to improve the standardization of BRAC surveys to improve the reliability of the
data, thus improving the validity of data comparisons and trends.

Proper selection of survey instrumentation is important to provide accurate dose rate readings at
the intended measurement locations. Dose rate surveys should be taken using a detector with
directional bias response in order to accurately measure contact dose rates on each specific
component and to reduce the impact on the reading from extraneous background radiation and
contributions from nearby components. A specific recommended detector/shield housing
assembly is the Eberline HP220 A. This detector consists of a small Geiger-Mueller detector
inside a hemispherical tungsten shield, which provides a 7 to 1 attenuation front to back for
cobalt-60 gamma emitters. A digital readout ratemeter is preferred, but an analog model such as
the Eberline E530N is acceptable. The Eberline E520, a very common Geiger-Mueller
instrument, should not be used for directional BRAC surveys because the instrument switches to
a second, internal detector when on the highest scale.

The survey instrumentation used is not universally consistent among the plants, which can
contribute to difficulty in making comparisons. At a given plant, the use of a non-directional
instrument will generally result in dose rate readings higher than the actual contact
measurements of interest and will likely contribute to greater variability in the readings taken
during different outages. Data have variously been collected using TLD (Thermo-Luminescent

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Dosimeter), non-directional surveys meters (e.g., ion chambers such as the Eberline RO2), and
‘homemade’ directional detectors (GM detectors wrapped in lead).

Survey points are specified (1) throughout the primary system, and include the suction and
discharge piping of the recirculation pumps, suction and discharge piping of the reactor water
cleanup pumps, the main steam lines, the inlet and outlet of the regenerative and non-
regenerative heat exchangers, and points on the heat exchangers themselves. The BRAC average
values used throughout this report are the average of the recirculation suction and discharge
contact dose rate readings for a set of measurements from a given plant. Surveys should be
conducted with the component in its normal configuration; for example, with any insulation in
place and liquid-filled. The BRAC program does not specify a distance between the target
survey point and the system components (e.g., a valve in the pipe) nor does it recommend a
length for straight run of pipe. It only specifies that each unit should be consistent with its
selected point.

Surveys should be conducted during each refueling outage and during other outages that are long
enough to permit a meaningful survey. The surveys should be conducted between 7 and 14 days
after shutdown, with the 7-day minimum to allow short lived isotopes to decay. Variability in the
time at which surveys are taken may complicate the interpretation of the results; particularly in
the absence of piping gamma scan results.

Differences in plant design, access platforms, etc. contribute to inconsistency in the exact
location of the survey points. In addition, surveys may be conducted when systems are drained
or with insulation removed. Many units have changed insulation types over the years, which
would change the effective standoff distance or the radiation shielding value of the insulation.
Several plants have added permanent shielding to the BRAC components. Performing the
survey with systems drained will most likely result in higher readings than if the system was full.

Time after shutdown when the survey is taken can also vary significantly. Plants obtain BRAC
surveys for trending, even in short mid-cycle outages. The continued compression of outages
may make collecting consistent BRAC data in the specified 7 to 14 day range more difficult.
Shielding may be installed on components well before the 7-day minimum. Plants then must
either take the BRAC survey early or wait past the 14-day recommendation, when the system is
restored. In addition, if a plant experiences a fuel failure during the cycle, 7 to 14 days may
actually be insufficient to eliminate the contribution from the iodine-131 fission product.

BRAC Monitoring Results

A chart of the most recent reported BRAC point average dose rates for the thirty-six operating
North American BWRs is shown in Figure 5-1 (2). The BRAC value shown is normally the
average of the reactor recirculation suction and discharge contact dose rates measured with a
shielded directional probe in the vertical piping sections. Data points are differentiated by the
prevailing chemistry regime prior to the dose rate measurements. The pre-decon BRAC average
is also plotted if a chemical decontamination of the reactor recirculation loop was performed
during the outage. As shown in Figure 5-1, chemical decontaminations were performed within
the last two years at Browns Ferry 2, Laguna Verde 2, Susquehanna 1 and Susquehanna 2. These

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four units and Hatch 2 have the lowest BRAC dose rates; without performing a chemical
decontamination, the Hatch 2 BRAC dose rate decreased significantly one cycle after NMCA
was performed. In contrast, the Nine Mile 1 BRAC average dose rate is the highest shown in
Figure 5-1; the Nine Mile 1 BRAC average showed a major increase in the refueling outage one
cycle after NMCA was performed. The main reasons for the decreased post-NMCA BRAC dose
rates at Hatch 2 were well-established reducing conditions under HWC prior to NMCA and
sufficient zinc addition prior to and following NMCA. The main reasons for the increased post-
NMCA BRAC dose rates at Nine Mile 1 were that the plant had very little time with hydrogen
injection prior to NMCA and had no zinc addition either prior to or following NMCA.
Additional details are provided in TR-1003022 (3) and later in this report.

Chemical decontaminations of the recirculation piping were performed at both Susquehanna

units during their most recent refueling outages. If these chemical decontaminations had not been
performed, the Susquehanna units would had the highest and third highest BRAC dose rates, at
580 mR/hr and 1475 mR/hr for Unit 1 and Unit 2, respectively, among North American BWRs.
Both Susquehanna units implemented HWC in 1999 without zinc injection either prior to or after
HWC.

Figure 5-2 shows the frequency distribution of the most recent reported BRAC results. Based on
this data set, the industry mean and median BRAC dose rates are 207 mR/hr and 163 mR/hr,
respectively. The mean and median values are based on the post-decon dose rates for plants
where a chemical decontamination was performed.

The BRAC summary charts are updated as new data are received, typically following the fall and
spring refueling outages. The summary results are transmitted electronically twice per year to
each plant (2). The electronic report is intended to fill an information gap and can be used by
plant staff for benchmarking and improvement initiatives. Other objectives are to provide
additional incentive for plants to provide data in a timely manner and to work toward
standardizing BRAC survey programs.

In the electronic transmittal, only the plant or plants operated by the recipient are identified by
name. In addition, the plant-specific BRAC results include a historical trend for each plant,
showing milestones of when changes were made that may impact drywell dose rates. The
estimated dose rate contribution from Co-60 is also plotted if gamma scan data were provided.
BRAC/milestone trend plots for each plant are also provided in the Appendix sections of this
report.

The most recent and past survey results are given in Table 5-1. The table also includes survey
dates, date of the most recent refueling outage, and date of the last chemical decontamination (or
replacement) of the recirculation piping. The information in this Table, along with additional
information on chemistry regimes, including dates of implementation of NZO, DZO, HWC, and
NMCA, are provided to the participants in the electronically transmitted BRAC Summary
Report. The data in Table 5-1 and in Figures 5-1 and 5-2 are updated frequently as new data are
received; therefore the data presented here are those available at the time this report was
prepared.

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Table 5-1
BRAC Results Summary for North American BWRs

Plant Last Recirc Most Most Recent Previous

%
BRAC Plant Chem Recent BRAC BRAC
Date Date Change
Order Decon RFO (mR/hr) (mR/hr)
1 SUS2 2001 Apr-01 Mar-01 8 Mar-99 103 -93
2 BRF2 2001 Mar-01 Dec-01 25 May-10 4 525
3 HAT2 Sep-01 Sep-01 31 Mar-00 133 -77
4 LAG2 2000 Mar-00 Jun-00 35 Sep-97 276 -88
5 HAT1 1996 Mar-02 Mar-02 35 Oct-00 60 -42
6 LIM1 Mar-02 Mar-02 45 Apr-98 94 -53
pipe repl
7 DNPS2 Oct-01 Oct-01 65 Sep-99 60 8
1997
8 LIM2 Apr-01 Apr-01 80 May-99 103 -22
9 JAF 1994 Oct-00 Mar-01 81 Oct-00 78.75 3
pipe repl
10 VY Apr-01 May-02 92 May-01 76 21
1986
11 DNPS3 1994 Sep-00 Oct-00 100 May-98 120 -17
12 LAG1 1998 Aug-99 Jun-01 109 Aug-00 148.5 -27
13 HCNS Oct-01 Oct-01 114 Apr-00 124.5 -9
14 LAS2 1995 Nov-00 Sep-01 118 Nov-00 152.5 -23
15 ENF Oct-01 Nov-01 125 Apr-00 131 -5
16 CNS 1984 Nov-01 Nov-01 151 Mar-00 213 -29
17 BRF3 1995 Mar-02 Mar-02 160 Apr-00 138 16
18 QUA2 1997 Feb-02 Feb-02 205 Mar-01 295 -31
pipe repl
19 PB2 Sep-00 Sep-00 214 Sep-98 130.75 64
1985
20 NMP2 Mar-02 Mar-02 226 Jul-01 110 105
21 CLI 1993 Apr-02 Apr-02 232 Oct-00 237 -2
pipe repl
22 PB3 Sep-01 Sep-01 233 Oct-99 255 -9
1988
23 DUA 1995 Apr-01 May-01 245 Nov-99 288 -15
24 PER 1996 Feb-01 Feb-01 245 Apr-99 250 -2
25 LAS1 1996 Jan-02 Jan-02 257 Feb-01 158 63
26 MON Nov-01 Nov-01 278 Feb-01 325 -15
27 SUS1 2002 Mar-02 Mar-02 311 Mar-00 12 2492
28 CGS 1992 May-01 May-01 318 Sep-99 313 2
29 QUA1 1998 Oct-00 Apr-01 320 Oct-00 339 -6
30 BRU2 1996 Feb-01 Feb-01 331 Sep-00 316.7 5
31 GGNS 1995 Apr-01 May-01 340 Dec-99 308 10
32 PIL 1998 Apr-01 May-01 348 Aug-00 393 -11
33 BRU1 1995 Mar-02 Mar-02 363 Feb-00 450 -19
34 OYC 1991 Oct-00 May-01 390 Nov-00 432 -10
35 RIB 1992 Sep-01 Sep-01 473 Mar-00 290.5 63
36 NMP1 Mar-01 May-02 1000 Aug-01 905 10

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Drywell Radiation Dose Rates

1600
1400
Average Dose Rate (mR/hr)

1200
1000
800
600
400
200
0

PER
NMP2

NMP1
PIL
DNPS2

DNPS3

ENF
JAF

HCNS

CLI

CGS
DUA

MON
SUS2

LAG2
HAT1
LIM1

LIM2

LAG1

LAS2

PB2

PB3

LAS1

SUS1

GGNS
CNS
VY

OYC
QUA2

QUA1
BRF2
HAT2

BRF3

BRU1

RIB
BRU2
Plant BRAC Order

HWC HWC/NMCA NWC NWC/NMCA Pre Decon

Figure 5-1
Most Recent BRAC Average Dose Rates for North American BWRs Designated by Chemistry Regime
at Time of Measurement

10 100

Percent In or Less Than Range

9 90
8 80
Number of Plants

7 70
6 60
5 50
4 40
3 30
2 20
1 10
0 0
50-99
100-149
150-199
200-249
250-299
300-349
350-399
400-449

550-599
600-649
650-699
700-749
750-799
800-849
850-899
900-949
0-49

450-499
500-549

950-999
1000-1049

BRAC Dose Rate Range (mR/hr)

Number of Plants Percent In or Less Than Range

Figure 5-2
Frequency Distribution of Most Recent BRAC Average Dose Rates for North American BWRs

5-5
 EPRI Licensed Material

Drywell Radiation Dose Rates

Drywell Dose Rate Correlations

The most significant long-term source of BWR radiation fields is Co-60, which has a half-life of
5.27 years and accounts for 70% or more of the measured drywell dose rates. Co-60 is produced
from the neutron activation of naturally occurring Co-59. The main source of Co-59 in BWRs is
the Stellite™ hardfacing material originally used for its wear resistant properties in primary,
feedwater, and balance-of-plant system components. Stellite™ contains approximately 60%
cobalt and the original inventory is responsible for approximately 90% of the total cobalt input to
the reactor water (4). Other typical activated corrosion products that contribute to a lesser extent
to BWR radiation dose rates include Mn-54, Cr-51, Co-58, Fe-59, and Zn-65.

Previous studies based on data from the EPRI BWR Chemistry Monitoring database have found
that BRAC dose rates can be most clearly correlated with soluble Co-60 (5, 6). Direct
correlations between drywell dose rates and iron, total Co-60 or insoluble Co-60 were not
apparent. Therefore, correlations between BRAC dose rates and soluble Co-60 are examined
again here with updated data. A plot of BRAC dose rates, from lowest to highest, along with the
average soluble Co-60 is shown in Figure 5-3. Where a chemical decontamination was
performed, the pre-decontamination BRAC measurements are shown in Figure 5-3. The most
recent data do not demonstrate a strong correlation between soluble Co-60 and BRAC dose rates
for different plants as has been found previously. This is attributed in large part to recent changes
in chemistry regimes.

1800 9E-04
1600 8E-04
1400 7E-04
BRAC Dose Rate (mR/hr)

Soluble Co-60 (µCi/ml)

1200 6E-04
1000 5E-04
800 4E-04
600 3E-04
400 2E-04
200 1E-04
0 0E+00
PER

PIL
DNPS2

DNPS3

ENF
JAF
HCNS

CLI
CNS

GGNS
DUA

MON
CGS
NMP2

NMP1
HAT1
LIM1
LIM2

LAG1
LAS2
LAG2

BRF3
PB2

PB3

LAS1

SUS2
VY

OYC
QUA1
BRF2
HAT2

BRU2

BRU1
RIB

SUS1
QUA2

BRAC Soluble Co-60

Figure 5-3
BRAC and Average Soluble Co-60

5-6
 EPRI Licensed Material

Drywell Radiation Dose Rates

The most prevalent cause of the lack of correlation between BRAC dose rates and soluble Co-60
is the widespread implementation of NMCA. The change to operation under NMCA causes a
restructuring of crud in the primary system, the largest inventory of which is on the fuel. This
transition upsets the equilibrium between Co-60 on the fuel and in the coolant. Reactor water
zinc concentrations also tend to vary widely in this transition, since one of the countermeasures
to the increased Co-60 is to increase the rate of zinc injection. The high dose rates at the
Susquehanna units are another example of the chemistry transitions due to an abrupt chemistry
regime change; relatively low values of soluble Co-60 correspond with high BRAC dose rates as
a result of the transition from NWC to HWC without zinc injection.

A plot of recently reported BRAC average dose rates vs. soluble Co-60 is presented in Figure 5-
4. The soluble Co-60 concentration is the average for the operating cycle or period prior to the
BRAC survey. There is considerable variability, with a weaker industry trend of increasing dose
rates with increasing soluble Co-60 than has been previously observed. The correlation has
become weaker due to the recent transition to a more reducing chemistry environment at several
plants and as these plants have increased zinc injection to suppress Co-60. The data suggest an
increasing trend in drywell dose rates above approximately 1E-4 µCi/ml soluble Co-60.

1600

1400
Average BRAC Dose Rate(mR/hr)

1200

1000

800

600

400

200

0
1E-05 1E-04 1E-03
Soluble Co-60 (µCi/ml)

BRAC vs. Soluble Co-60 Expon. (BRAC vs. Soluble Co-60)

Figure 5-4
Most Recently Reported BRAC Dose Rate vs. Average Soluble Co-60

The most recently reported BRAC results are also plotted against soluble Co-60 in Figure 5-5,
but with the data symbols differentiated by chemistry regime. The chemistry regime indicated is
the one that was in effect leading up to the drywell survey. The data show some grouping among
plants with the same chemistry regime. In the data of Figure 5-5, no distinction is made between

5-7
 EPRI Licensed Material

Drywell Radiation Dose Rates

plants that are on HWC and ones that are on HWC after NMCA; in either case, the data point is
designated as HWC. Figure 5-5 clearly shows that zinc suppresses BRAC dose rates, as
evidenced by the similarity of the data for NWC+Zn and HWC+Zn and the widely divergent
results for the three data points for HWC without Zn.

A plot of the data for only those plants with the HWC+Zn chemistry regime (including NMCA
plants) is provided in Figure 5-6. A trend of increasing BRAC dose rates with increasing soluble
Co-60 is indicated. BRAC dose rates appear to increase rapidly at soluble Co-60 levels above
about 1E-4 µCi/ml. Again, the scatter in the data is attributed to recent transitions to more
reducing chemistry.

The most recent BRAC average for each chemistry regime was determined and the results are
plotted in Figure 5-7. However, only relative comparisons are valid because the amount of time
since conversion to these chemistry regimes varies. As expected, the average for HWC+Zn
plants is much lower than for HWC (no Zn injection). Another influencing factor here is the
number of chemical decontaminations performed over the plant life. HWC+Zn plants have
performed the most chemical decontaminations. It is noted that in Figures 5-5 and 5-7 the HWC
category, which does not inject zinc, includes only three plants now.

1600

1400
Average BRAC Dose Rate (mR/hr)

1200

1000

800

600

400

200

0
1E-05 1E-04 1E-03
Soluble Co-60 (µCi/ml)

HWC HWC+Zn NWC+Zn

Figure 5-5
Most Recently Reported BRAC Dose Rate vs. Average Soluble Co-60 by Chemistry Regime

5-8
 EPRI Licensed Material

Drywell Radiation Dose Rates

500
450
Average BRAC Dose Rate (mR/hr)

400
350
300
250
200
150
100
50
0
1E-05 1E-04 1E-03
Soluble Co-60 (µCi/ml)

HWC+Zn Expon. (HWC+Zn)

Figure 5-6
Most Recently Reported BRAC Dose Rate vs. Average Soluble Co-60 for HWC+Zn Plants

1200

1018
1000
Average BRAC Dose Rate (mR/hr)

800

600

400

220
191
200

0
HWC HWC+Zn NWC+Zn

Figure 5-7
Average BRAC Dose Rate by Chemistry Regime (based on most recently reported dose rates)

5-9
 EPRI Licensed Material

Drywell Radiation Dose Rates

Historical BRAC dose rate data were examined to compile the measurements taken at plants
during the next refueling outage after a chemical decontamination of the recirculation piping was
performed. Generally, the BRAC dose rates immediately after chemical decontamination are
very low, often <20 mR/hr. The results in the next outage were averaged for each chemistry
regime and are presented in Figure 5-8. These values are representative of the average one-cycle
recontamination dose rate. The results show that the lowest average recontamination rates are
achieved under the HWC+Zn chemistry regime, which has an even lower recontamination rate
than NWC+Zn. On average, HWC plants that do not inject zinc have the highest recontamination
rate. The NWC chemistry regime (no zinc injection) data point is presented for historical
comparison since there are no longer any plants operating under this regime.

600

517
500
BRAC Dose Rate (mR/hr)

400

300
232 220
200
128

100

0
NWC, No Zn HWC, No Zn NWC+Zn HWC+Zn

Figure 5-8
BRAC Dose Rate One Cycle After Chemical Decontamination by Chemistry Regime

The average impact of the sequencing of transitions between chemistry regimes was also
examined. The objective here is to determine, for example, whether BRAC dose rates are
significantly different for plants that applied HWC before zinc injection compared with plants
that established zinc injection on NWC and then started HWC. The results are presented in
Figure 5-9. Note that in Figure 5-9, “HWC-Zn” is the label for the group of plants that was on
HWC for a significant operating time period and then began zinc injection. “Zn-HWC” is the
label for the group of plants that had significant operating time with zinc injection before starting
HWC. The results show that the average dose rate is lower for Zn-HWC plants than for HWC-Zn
plants. The average dose rates for the Zn-HWC and HWC-Zn groups are substantially lower than
the “HWC(NMCA)” group, which does not inject zinc. The average dose rates for the “HWC-
Zn-NMCA” and “Zn-HWC-NMCA” groups are similar and also both significantly lower than
the average dose rate of the HWC(NMCA) group.

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 EPRI Licensed Material

Drywell Radiation Dose Rates

1200

1018
1000
BRAC Dose Rate (mR/hr)

800

600

400
304
233 239
202 178
200 125

0
HWC-Zn Zn-HWC HWC Zn-NMCA HWC-Zn- Zn-HWC- Zn
(NMCA) NMCA NMCA

Figure 5-9
Average Impact on BRAC Dose Rates of Transitions Between Chemistry Regimes

The impact of transitions between chemistry regimes can also be seen as a function of soluble
Co-60 in Figure 5-10. The data show that Zn-HWC plants (zinc added before the start of
hydrogen injection) have lower BRAC dose rates than HWC-Zn plants over the same Co-60
range. The high dose rate associated with HWC or HWC-NMCA without zinc is also apparent.

There appears to be a tendency toward increased BRAC dose rates at low reactor water zinc,
particularly at zinc concentrations less than 3 ppb, as shown in Figure 5-11. More significantly,
over the same range of reactor water zinc concentrations, the BRAC dose rates for HWC-Zn
plants are higher than those of Zn-HWC plants. This observation supports the benefits of
injecting zinc prior to starting HWC or immediately following a chemical decontamination.
Since the extent of metal oxide restructuring from ECP reduction in the primary system is greater
under HWC-NMCA than under HWC, by analogy the above findings also support EPRI
recommendations to add sufficient zinc preferably prior to beginning HWC-NMCA operations.

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Drywell Radiation Dose Rates

1600

1400

1200
BRAC (mR/hr)

1000

800

600

400

200

0
1E-05 1E-04 1E-03
Soluble Co-60 (µCi/ml)

HWC-Zn Zn-HWC HWC or HWC-NMCA Zn-NMCA

HWC-Zn-NMCA Zn-HWC-NMCA Zn

Figure 5-10
BRAC Dose Rate vs. Soluble Co-60 (points distinguished by transitions between chemistry regimes)

1600

1400

1200
BRAC (mR/hr)

1000

800

600

400

200

0
0 2 4 6 8 10 12 14
Reactor Water Zn (ppb)

HWC-Zn Zn-HWC HWC-NMCA Zn-NMCA HWC-Zn-NMCA Zn-HWC-NMCA Zn

Figure 5-11
BRAC Dose Rate vs. Reactor Water Zinc (points distinguished by transitions between chemistry
regimes)

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 EPRI Licensed Material

Drywell Radiation Dose Rates

A known effect of zinc addition is the suppression of reactor coolant Co-60 concentrations. The
amount of zinc that must be injected into the feedwater to suppress Co-60 varies widely among
the plants, as shown in Figure 5-12. The amount of zinc required depends on several variables,
including the feedwater iron concentration, crud inventory on the fuel, the extent of the crud
restructuring process particularly for NMCA plants and the RWCU flow rate. Some plants are
capable of achieving less than 1E-4 µCi/ml soluble Co-60 at feedwater zinc concentrations well
below 0.3 ppb while at other plants soluble Co-60 is greater than 1E-4 µCi/ml at average
feedwater zinc of about 1 ppb.

1E-03
Soluble Co-60 (µCi/ml)

1E-04

1E-05

1E-06
0.0 0.2 0.4 0.6 0.8 1.0 1.2
Average Feedwater Zinc (ppb)

Figure 5-12
Reactor Water Soluble Co-60 vs. Feedwater Zinc

The soluble Co-60 suppression effect of zinc in reactor water can be seen for the North American
BWRs in Figure 5-13. The general decreasing trend in Co-60 as reactor water Zn concentration
increases is still apparent despite the transitional chemistry effects of NMCA.

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 EPRI Licensed Material

Drywell Radiation Dose Rates

1E-03
Soluble Co-60 (µCi/ml)

1E-04

1E-05

1E-06
0 2 4 6 8 10 12 14
Reactor Water Zinc (ppb)

Figure 5-13
Reactor Water Soluble Co-60 vs. Reactor Water Zinc

Plots of reported BRAC dose rates versus the ratio of reactor water soluble Co-60 to reactor
water soluble zinc are presented in Figures 5-14 and 5-15. There is significant variability in the
data resulting in weak correlations for most of the trends. Use of all the data, regardless of the
operating chemistry regime, indicates a trend of increasing BRAC dose rates with higher ratios
as shown in Figure 5-14. A similar plot but with the plants distinguished by chemistry regime is
provided in Figure 5-15. The chemistry regime indicated is the one in effect prior to the drywell
survey. It is noted that the data of the single plant under the HWC NMCA chemistry regime is
not shown here but it is significantly different than for the other chemistry regimes. This plant
does not inject zinc, and the ratio of reactor water soluble Co-60 to reactor water soluble zinc is
orders of magnitude higher than for the other plants.

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 EPRI Licensed Material

Drywell Radiation Dose Rates

600

500
BRAC Dose Rate, mR/hr

400

300

200

100

0
1.00E-06 1.00E-05 1.00E-04
Co60s / RCI Zns

Figure 5-14
BRAC Dose Rates vs. Ratio of Reactor Water Soluble Co-60 to Reactor Water Soluble Zinc

600

500
BRAC Dose Rate, mR/hr

400

300

200

100

0
1.00E-06 1.00E-05 1.00E-04
Co60s/RCI Zns
DZO DZO HWC DZO HWC NMCA
DZO NMCA HWC HWC DZO HWC DZO NMCA
NMCA NZO Power (HWC DZO)

Figure 5-15
BRAC Dose Rates vs. Ratio of Reactor Water Soluble Co-60 to Reactor Water Soluble Zinc by
Chemistry Regime In Effect Before BRAC Measurement

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 EPRI Licensed Material

Drywell Radiation Dose Rates

A plot of BRAC dose rates versus the ratio of reactor water soluble zinc to feedwater total iron is
shown in Figure 5-16. This plot does not distinguish between the operating chemistry regimes.
A trend of decreasing dose rate with increasing ratio is indicated, although there is significant
scatter in the data. The reactor water zinc to feedwater iron ratio may have physical significance,
providing an indication that sufficient zinc is being fed to maintain a sufficient residual in the
reactor coolant to stabilize the incoming iron as it deposits primarily on the fuel. The same data
are plotted in Figure 5-17 but the points are distinguished by the chemistry regime in effect prior
to the BRAC measurement. These trends show less variability for the various operating
chemistry regimes than the BRAC dose rate versus the ratio of reactor water total Co-60 to
reactor water soluble zinc. As before, the data for the one plant operating with HWC NMCA
varies greatly from the remainder of the plants, and is not included in Figure 5-17. This plant
does not inject zinc and is a Deep Bed Only plant with relatively high feedwater iron, resulting in
a much lower reactor water soluble zinc/feedwater total iron ratio than that for the other plants
shown. The collection of more data over time with a diminishing number of chemistry regimes
(as more plants implement HWC and NMCA) should aid in developing improved correlations.

600

500
BRAC Dose Rate, mR/hr

400

300

200

100

0
0.1 1 10 100
RCI Zns / FW Fe Tot

Figure 5-16
BRAC Dose Rates vs. Ratio of Reactor Water Soluble Zinc to Feedwater Total Iron

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 EPRI Licensed Material

Drywell Radiation Dose Rates

500
BRAC Dose Rate, mR/hr
400

300

200

100

0
1 10 100
RCI Zns / FW Fe Tot

DZO DZO HWC DZO HWC NMCA

DZO NMCA HWC HWC DZO HWC DZO NMCA
NMCA NZO DZO NMCA
Power (DZO HWC NMCA) Power (DZO HWC) Power (HWC DZO)
Power (HWC DZO NMCA)

Figure 5-17
BRAC Dose Rates vs. Ratio of Reactor Water Soluble Zinc to Feedwater Total Iron by
Chemistry Regime In Effect Before BRAC Measurement

Hot Spots

The industry data show correlations between BRAC point dose rates and soluble Co-60. They
provide guidance on the threshold Co-60 activity concentrations above which BRAC dose rates
increase sharply for each chemistry regime. In addition, hot spot survey data were collected to
determine if there is any correlation with Co-60. Data collection efforts focused on obtaining hot
spot survey data along with the reactor water fractionation data. To simplify the data gathering
efforts by plant personnel, only contact dose rates from the reactor vessel bottom head drain line
were requested as this is a common hot spot location.

The reported bottom head drain (BHD) dose rates are plotted as a function of insoluble Co-60 in
Figure 5-18. In some cases, more than one data point is included from a single plant. The
insoluble Co-60 values used are the average values for the cycle prior to the surveys. The results
show an increasing trend in BHD dose rates as insoluble Co-60 increases. The available results
show that the lowest BHD dose rates correspond to average insoluble Co-60 concentrations of
less than 6E-5 µCi/ml. for several plants, insoluble Co-60 approaches and exceeds 1E-4 µCi/ml
as BHD dose rates exceed 10 R/hr. The highest BHD contact dose rate was 1000 mR/hr, which
corresponded to an insoluble Co-60 of 4.84E-4 µCi/ml.

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 EPRI Licensed Material

Drywell Radiation Dose Rates

10000
Bottom Head Drain Dose Rate (mR/hr)

1000

100

10

0.1
1E-06 1E-05 1E-04 1E-03 1E-02
Cycle Average Insoluble Co-60 (µCi/ml)
Ins Co-60 (cycle) Power (Ins Co-60 (cycle))

Figure 5-18
BHD Dose Rate vs. Insoluble Co-60

Impact of Feedwater Iron on Co-60

Industry correlations have shown that feedwater iron has an indirect effect on BWR dose rates:
• Iron affects soluble Co-60, which affects fixed drywell piping dose rates (mainly influenced
by incorporated Co-60).
• Iron affects insoluble Co-60, which affects hot spots.

Impact of Feedwater Iron on Soluble Co-60

Some plants, like the Susquehanna units, operated for years under normal water chemistry and
no zinc addition with high (5 – 10 ppb) feedwater iron. Under these conditions, soluble Co-60 is
suppressed. However, today no plants operate any longer with feedwater iron in this high range
and most plants have implemented or are planning HWC or HWC+NMCA.

Industry data in the low feedwater iron range show that as feedwater iron decreases, reactor
water soluble Co-60 continues to decrease. This type of trend is shown for Brunswick 2,
Limerick 1, and FitzPatrick in Figure 5-19.

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 EPRI Licensed Material

Drywell Radiation Dose Rates

1.E-03
Reactor Water Soluble Co-60 (µCi/ml)

1.E-04

1.E-05
0.1 1 10
Feedwater Total Fe (ppb)

FitzPatrick Limerick 1 Brunswick 2

Expon. (FitzPatrick) Expon. (Limerick 1) Expon. (Brunswick 2)

Figure 5-19
Reactor Water Soluble Co-60 vs. Feedwater Iron

Brunswick 1 and 2 operate with Filter + Deep Bed condensate polishing. These units routinely
operate with feedwater iron less than 0.5 ppb, and have HWC (no NMCA) and DZO addition.
The trend of decreasing reactor water soluble Co-60 with decreasing feedwater iron for each
Brunswick unit is shown in Figure 5-20. The decreasing trend in Co-60 with decreasing
feedwater iron is observed at the relatively high reactor water soluble Co-60 concentrations,
apparently due to a high cobalt source term at these units.

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 EPRI Licensed Material

Drywell Radiation Dose Rates

1.E-03
Reactor Water Soluble Co-60 (µCi/ml)

1.E-04
0.00 0.05 0.10 0.15 0.20 0.25 0.30 0.35 0.40 0.45 0.50
Feedwater Total Fe (ppb)

BRU1 BRU2 Expon. (BRU2) Expon. (BRU1)

Figure 5-20
Brunswick 1 and 2 Reactor Water Soluble Co-60 vs. Feedwater Iron

Monthly average feedwater iron and reactor water soluble Co-60 data were examined for Hatch
1. The data shown in Figure 5-21 are for stable plant operating conditions at approximately full
power under moderate HWC and with DZO addition.

While additional Hatch 1 data are needed at feedwater iron less than 0.5 ppb to fully establish a
correlation, it appears that soluble Co-60 has decreased or at least remained constant at
approximately 4E-5 µCi/ml as feedwater iron has decreased to and below 0.5 ppb. This is
consistent with industry experience. A further evaluation of Hatch data with low feedwater iron
and under HWC/NMCA chemistry is provided later in this section.

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 EPRI Licensed Material

Drywell Radiation Dose Rates

1.E-04
Reactor Water Soluble Co-60 (µCi/ml)

1.E-05
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9
Feedwater Total Fe (ppb)

HAT1

Figure 5-21
Hatch 1 Reactor Water Soluble Co-60 vs. Feedwater Iron

Impact of Feedwater Iron on Insoluble Co-60

For most plants, as feedwater iron increases, insoluble Co-60 appears to go through a minimum
in the 1 – 2 ppb iron range. However, specific plants show different responses. Data for
Brunswick Unit 1 show an overall decreasing trend in insoluble Co-60 with decreasing feedwater
iron while the Unit 2 data indicate the opposite effect, as shown in Figure 5-22. Brunswick 1 has
been on HWC since 6/90 and DZO since 5/95. Brunswick 2 has been on HWC since 1/89 and
DZO since 3/96. Brunswick does not report an increase in hot spots at either unit due to low iron.
The reasons for these different responses to feedwater iron are not known at this time.

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 EPRI Licensed Material

Drywell Radiation Dose Rates

Reactor Water Insoluble Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05
0.00 0.10 0.20 0.30 0.40 0.50
Feedwater Total Fe (ppb)

BRU1 BRU2 Power (BRU2) Power (BRU1)

Figure 5-22
Brunswick Units 1 and 2 Reactor Water Insoluble Co-60 Response to Feedwater Iron

Limerick 1 also reported an increase in insoluble Co-60 and Zn-65 as feedwater iron decreased
after installing pleated pre-filters upstream of their retrofitted deep bed Condensate
Demineralizers. Limerick 1 was on NWC during this period and was adding NZO at the time.
The results in Figure 5-23 show insoluble Co-60 frequently greater than 1E-4 µCi/ml. Hot spots
due to activated crud particles were detected, including the 1C RWCU pump internal casing dose
rate increase from a previous contact measurement of 5 R/hr to 40 R/hr, and a 100 R/hr hot spot
on the Unit 1 bottom head drain line. Consequently, Limerick implemented iron addition to
increase final feedwater iron to approximately 0.5 ppb, after which the insoluble Co-60 and Zn-
65 control improved.

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 EPRI Licensed Material

Drywell Radiation Dose Rates

Reactor Water Insoluble Co-60 (µCi/ml)

1.E-02

1.E-03

1.E-04

1.E-05
0.00 0.20 0.40 0.60 0.80 1.00 1.20 1.40
Feedwater Total Fe (ppb)

LIM1 Poly. (LIM1)

Figure 5-23
Limerick 1 Reactor Water Insoluble Co-60 Response to Feedwater Iron

Under moderate HWC and DZO, Hatch 1 monthly average data at approximately full power
appear to show an increasing trend in insoluble Co-60 as iron has decreased to 0.6 ppb and
lower. This trend is shown in Figure 5-24. The insoluble Co-60 results are still significantly less
than 1E-4 µCi/ml. The influence of feedwater iron on reactor water insoluble Co-60 after NMCA
is examined for Hatch 1 later in this section.

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Drywell Radiation Dose Rates

Reactor Water Insoluble Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
0 0.5 1 1.5 2

Feedwater Total Fe (ppb)

HAT1

Figure 5-24
Hatch 1 Reactor Water Insoluble Co-60 Response to Feedwater Iron

Current Iron Control Guidelines

In the EPRI BWR Water Chemistry Guidelines – 2000 Revision (7), the Action Level 1 value for
feedwater iron at >10% power is 5 ppb. All North American BWRs achieve average feedwater
iron concentrations below 5 ppb. The desired range for iron control is given as 0.5 ppb to 3.0
ppb. Modeling studies show that feedwater iron concentrations up to 3.0 ppb do not significantly
increase recirculation piping dose rates over a 20 year period (8). The basis for the 0.5 ppb lower
limit is that, based on U.S. plant experiences, operation at less than 0.5 ppb feedwater iron has a
risk of increasing reactor water insoluble activated corrosion products leading to increased hot
spots. If plants operate at <0.5 ppb feedwater iron, the effects on activated corrosion products
and dose rate trends should be closely monitored.

With the increased use of highly efficient iron removal septa in condensate polishing
applications, plants that had historically controlled feedwater iron above 0.5 ppb have observed
feedwater iron decreasing below 0.5 ppb. These plants are in a position where they must decide
whether the addition of iron is necessary. Hatch 1 is considering the need to add iron, their
evaluation is discussed below as an example.

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Drywell Radiation Dose Rates

Hatch 1 Low Iron

Hatch 1 iron decreased below the EPRI recommended desirable range lower threshold of 0.5 ppb
in 2001. The station consequently considered installing an iron addition system to increase iron
above 0.5 ppb.

Hatch 1 has F/D (filter demineralizers) for condensate polishing and operated for over a decade
on HWC, with NZO started in 1990 and DZO in 1994. Prior to NMCA in 3/99, Hatch 1 had
several years of moderate HWC. Drywell shutdown dose rates remained low after a fuel cycle
with NMCA/HWC. This indicates that the extent and effects of crud restructuring in the primary
system (mainly on the fuel) were less than at other plants that experienced significant increases
in shutdown dose rates.

The feedwater iron trend for Hatch 1 since 1997 is shown in Figure 5-25. Hatch 1 feedwater iron
data show a downward trend following the spring 1999 refueling outage when NMCA was
applied. Since mid-2001, feedwater iron has been in the range of approximately 0.2 – 0.5 ppb.

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001

0.0001
1/1/97 5/16/98 9/28/99 2/9/01 6/24/02 11/6/03

Insoluble Fe Soluble Fe NMCA

Figure 5-25
Hatch 1 Feedwater Iron Trend

A trend plot provided by Hatch showed that by May 2001, feedwater iron decreased to
approximately 0.4 ppb. This trend is shown in Figure 5-26. Prior to mid-April 2001, CDE
insoluble iron was normally 0.3 ppb higher than feedwater iron. From about the middle to the
end of April 2001, CDE insoluble iron was 0.6 to 0.8 ppb higher than feedwater insoluble iron.
From the EPRI database for Hatch 1, the 2000 averages for feedwater and CDE insoluble iron

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Drywell Radiation Dose Rates

were 0.93 ppb and 1.23 ppb, respectively. The 2000 averages show CDE insoluble iron at 0.3
ppb higher than feedwater iron, approximately the same as the difference indicated by the first
quarter 2001 trends.

Edwin I. Hatch Unit 1

2
14
1.8
1.6 12
CD E/FW Fe (ppb)

1.4 10

H W Fe (ppb)
1.2
1 8

0.8 6
0.6
4
0.4
2
0.2
0 0
01/01/01 01/21/01 02/10/01 03/02/01 03/22/01 04/11/01 05/01/01 05/21/01

FW Insol Fe CDE Insol Fe HW Insol Fe

Figure 5-26
Hatch 1 2001 Iron Data

The only F/D plant that is currently adding iron to the feedwater is Columbia Generating Station.
Iron oxalate is added to control the feedwater iron concentration at approximately 1 ppb to avoid
the potential for crud destabilization leading to insoluble cobalt-60 release and subsequent
increase in hot spots. Columbia experienced increased hot spot dose rates when iron decreased
below 0.5 ppb under NWC.

Hatch 1 reactor water soluble and insoluble Co-60 both increased during plant operation
following NMCA. The Co-60 trend data are shown in Figure 5-27. While insoluble Co-60
peaked in the 1E-3 µCi/ml range, soluble Co-60 never showed a sustained increase above 1E-4
µCi/ml. By the early 2001, both soluble and insoluble Co-60 had peaked following NMCA and
were trending down.

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Drywell Radiation Dose Rates

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
1/1/97 5/16/98 9/28/99 2/9/01 6/24/02 11/6/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure 5-27
Hatch 1 Reactor Water Co-60 Trend Data

Plants with 2001 average feedwater iron of 1.5 ppb or lower are shown in Figure 5-28. As
shown, the Brunswick units have average feedwater iron values in the <0.5 ppb range. At
Brunswick, Co-60 is lowest when feedwater iron is minimized and no impact on hot spot dose
rates is observed. This is in contrast with the experience of Limerick and Columbia, where
insoluble Co-60 increased when feedwater iron dropped significantly below 0.5 ppb, giving rise
to hot spots.

A trend plot of monthly average values of feedwater iron and reactor water soluble Co-60 for
Hatch 1 is shown in Figure 5-29. The data plotted include only stable operating periods. For most
periods, Co-60 appears to trend with iron. An exception is the period after NMCA when soluble
Co-60 increased above normal levels. A similar plot for insoluble Co-60 is shown in Figure 5-30.
For the initial months following NMCA, the data appear to show a response of increasing
insoluble Co-60 with decreasing iron. However, by May 2001, insoluble Co-60 decreased to
<1E-5 µCi/ml as feedwater iron decreased below 0.5 ppb. This result indicates that, after a
stabilization period following NMCA, the feedwater iron decrease to <0.5 ppb at Hatch 1 did not
cause soluble or insoluble Co-60 increases that would increase BRAC dose rates or hot spots.

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Drywell Radiation Dose Rates

6
Total Feedwater Fe (ppb)
5

0
ENF2

PER1

FIT1
HCNS

CGS
LIM1
HAT1

HAT2
DUA1
LIM2

MON1
SUS1
CNS1
BRF3

PB2
BRF2

CLI1

NMP2

LAS2
VY
BRU1

SUS2

QUA1
BRU2

QUA2
DB F/D
F+DB DESIRED MINIMUM
DESIRED MAXIMUM ACTION LEVEL 1

Figure 5-28
Plants in Low Iron Range, 2001 Averages

2.0 1.1E-04

1.8 1.0E-04

Reactor Water Soluble Co-60 (µCi/ml)

1.6 9.0E-05
Feedwater Total Iron (ppb)

1.4 8.0E-05

1.2 7.0E-05

1.0 6.0E-05

0.8 5.0E-05

0.6 4.0E-05

0.4 3.0E-05

0.2 2.0E-05

0.0 1.0E-05
Dec-96 Dec-97 Dec-98 Dec-99 Dec-00 Dec-01 Dec-02

FFW Fe Total NMCA RCI Co60 Soluble

Figure 5-29
Hatch 1 Monthly Average Feedwater Iron and Reactor Water Soluble Co-60 at Steady
Power

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Drywell Radiation Dose Rates

3.0 1E-03

Reactor Water Insoluble Co-60 (µCi/ml)

2.5
Feedwater Total Iron (ppb)

2.0 1E-04

1.5

1.0 1E-05

0.5

0.0 1E-06
Dec-96 Dec-97 Dec-98 Dec-99 Dec-00 Dec-01 Dec-02

FFW Fe total NMCA RCI Co60 Insoluble

Figure 5-30
Hatch 1 Monthly Average Feedwater Iron and Reactor Water Insoluble Co-60 at Steady
Power

To evaluate the influence of feedwater iron on reactor water Co-60 after NMCA, data were
compiled for the period of December 2000 through January 2002. During this period, reactor
water activated corrosion products appear to have stabilized from the effects of NMCA. A plot
of reactor water soluble Co-60 vs. feedwater iron is shown in Figure 5-31. The data show a mild
trend of increasing soluble Co-60 with decreasing feedwater iron concentration, but soluble Co-
60 remained well below 1E-4 µCi/ml. The soluble Co-60 levels are similar to those for Hatch 1
prior to NMCA (see Figure 5-21).

A plot of reactor water insoluble Co-60 vs. feedwater iron under stable chemistry after NMCA is
shown in Figure 5-32. The increasing trend in insoluble Co-60 with decreasing feedwater iron is
similar to the pre-NMCA trend (moderate HWC) shown in Figure 5-24. The post-NMCA trend
shows insoluble Co-60 going above 1E-4 µCi/ml during only one month as feedwater iron
decreased to 0.3 ppb, while other results in this low iron range show insoluble Co-60 controlled
to well below 1E-4 µCi/ml.

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Drywell Radiation Dose Rates

1E-04

9E-05
Reactor Water Soluble Co-60 (µCi/ml)

8E-05

7E-05

6E-05

5E-05

4E-05

3E-05

2E-05

1E-05
0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4
Feedwater Total Iron (ppb)

Figure 5-31
Hatch 1 Monthly Average Soluble Co-60 vs. Feedwater Iron at Steady Power After Post-
NMCA Transient Conditions

1.E-03
Reactor Water Insoluble Co-60 (µCi/ml)

1.E-04

1.E-05

1.E-06
0.00 0.20 0.40 0.60 0.80 1.00 1.20 1.40
Feedwater Total Iron (ppb)

Figure 5-32
Hatch 1 Monthly Average Insoluble Co-60 vs. Feedwater Iron at Steady Power After Post-
NMCA Transient Conditions

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Drywell Radiation Dose Rates

The EPRI BWR Water Chemistry Guidelines – 2000 Revision (7), Section 3.2.4.4 states the
following regarding feedwater iron: "Long term operation below 0.5 ppb may also be acceptable
in particular cases, but the current experience base is insufficient to eliminate concerns regarding
hot spots and increases in reactor water 60Co. Thus, plants operating consistently below 0.5 ppb
feedwater iron should monitor reactor water activity (soluble and insoluble) and plant radiation
buildup trends; and use these to evaluate the need for iron addition."

At Hatch 1, no significant increases in hot spot dose rates have been observed when iron
decreased below 0.5 ppb. There are several differences between current conditions at Hatch 1
and the conditions at other U.S. plants (Limerick 1 and Columbia) that experienced increased hot
spot dose rates following low feedwater iron operation. One major difference is that Hatch 1 has
operated since 1987 under HWC, and since July 1999 on HWC/NMCA, while both Limerick 1
and Columbia were on NWC. Hatch has also been injecting zinc since 1990 (NZO 8/90, DZO
2/94) while Columbia was just starting DZO addition and Limerick 1 was injecting NZO. It is
important to note that there is significant restructuring of the iron-based fuel deposit during
transition from NWC to HWC and in the transition from HWC to NMCA/HWC. This
restructuring of the iron-based fuel deposit may continue for many months of operation and
contribute to increases in reactor water Co-60 levels, which in turn impact dose rates. Once this
restructuring nears completion, and highly reducing conditions and sufficient DZO injection are
maintained, the fuel crud deposits appear to remain stable.

The available data indicate that, with well-established reducing conditions in the primary system
and low iron inventory on the fuel after several cycles of operation with feedwater iron of 0.5 -
1.5 ppb and DZO addition, further lowering the feedwater iron below 0.5 ppb does not increase
dose rates, compared to the increases observed when iron was lowered to <0.5 ppb under NWC
conditions. This conclusion is supported by the Hatch 1 experience, where BRAC dose rates
continue to decline and no increase in the occurrence of hot spots has been observed as feedwater
iron has decreased below 0.5 ppb. In general, more plants are able to reduce feedwater iron to
low levels through the implementation of new iron filtration technologies. With fully established
reducing conditions under NMCA/HWC, and stabilization of the iron oxide spinels by injecting
adequate DZO, it appears that operating with <0.5 ppb feedwater iron may provide the long-term
benefits of lower drywell radiation dose rates and lower zinc usage without increasing hot spots.
The feedwater iron, Co-60, and dose rate trends continue to be closely monitored at Hatch 1 and
other stations through the EPRI BWR Chemistry Monitoring Database activities.

References

1. G. F. Palino, “BWR Radiation Level Surveillance,” NEDC-12688, December 1977.

2. “EPRI BWR BRAC Summary” (Electronic Report), May 2002.

3. “BWRVIP-92: NMCA Experience Report and Application Guidelines,” TR-1003022, Final

Report, September 2001.

4. “BWR Cobalt Source Identification”, General Electric Company, February 1982. EPRI NP-
2263.

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Drywell Radiation Dose Rates

5. “BWR Iron Control Monitoring,” TR-108737, Interim Report, December 1998.

6. “BWR Iron Control Monitoring,” TR-109565, Final Report, September 1999.

7. “EPRI BWR Water Chemistry Guidelines – 2000 Revision,” TR-103515-R2, Final Report,
February 2000.

8. “Correlative Plant Data Study of Influence of Iron on BWR Activity Transport,” EPRI TR-
109566, May 1998.

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6
PLANT CHEMISTRY IMPACT OF NMCA

Noble Metal Chemical Application (NMCA) was developed to achieve the BWR
electrochemical corrosion potential (ECP) specification for mitigation of IGSCC (Intergranular
Stress Corrosion Cracking) (1) while avoiding the high operating dose rates and high hydrogen
gas usage associated with Moderate Hydrogen Water Chemistry. NMCA deposits very small
amounts of platinum and rhodium metal (approximately 1 µg/cm , a mass equivalent to a single
2

atomic layer) on the wetted surfaces in the reactor vessel and reactor coolant system. The ECP
response of these wetted surfaces is similar to that of a platinum surface, and a protective ECP
(<-230 mV SHE) is achieved when the molar ratio of hydrogen to total oxidant (oxygen plus
hydrogen peroxide) in the reactor water is two or more. With NMCA, protection is achieved at
feedwater hydrogen concentrations of 0.1 – 0.3 ppb; much lower than the 1 – 2 ppm
concentration required to achieve protection under HWC. The reduced hydrogen injection with
NMCA results in less N-16 production, thereby reducing main steam line radiation monitor
(MSLRM) dose rates and lowering overall operating dose.
NMCA has been accepted rapidly by the BWR industry. As of August 2002, twenty-three of the
thirty-six North American BWRs have applied noble metals, and one plant is planning NMCA in
October 2002. Seven additional North American plants are planning to apply noble metals after
2002, which would increase the number of NMCA plants to thirty.

NMCA plants are clearly benefiting from the IGSCC mitigation and are operating with dose
reduction benefits and hydrogen cost savings. However, the post-NMCA impact on shutdown
dose rates has differed among the plants. Some plants have seen decreases, or no significant
change, in shutdown dose rates following NMCA while others have seen increases. All plants
have observed chemistry transient effects following NMCA. These and other NMCA issues are
addressed in BWRVIP-92 (2).

Noble Metal Loading

Data on noble metal application loading reported by the plants to the EPRI BWR Iron/Chemistry
Monitoring database are summarized in Table 6-1. The target loading for the first application is
approximately 1 µg/cm2. The loading is normally based on NMCA coupon loading results but
may also be estimated by a material balance. The values give the combined mass of platinum and
rhodium deposited on surfaces, including the fuel. Note that Duane Arnold is the only plant that
has reapplied noble metals so far.

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Plant Chemistry Impact of NMCA

Table 6-1
Noble Metal Application Loading

Noble Metal
Plant Date Loading
(µg/cm2)

Browns Ferry 2 Mar-01 0.95

Browns Ferry 3 Apr-00 1.03

Clinton Apr-02 0.75

Columbia May-01 0.51

Cooper Mar-00 0.65

Dresden 2 Oct-99 0.8

Dresden 3 Sep-00 0.3

Duane Arnold (Initial) Oct-96 0.22

Duane Arnold (Reapplication) Oct-99 0.8

FitzPatrick Nov-99 1.18

Hatch 1 Mar-99 0.7

Hatch 2 Mar-00 1.3

LaSalle 1 Oct-99 0.7

LaSalle 2 Nov-00 1.0

Limerick 1 Mar-00 1.2

Limerick 2 Apr-01 1.1

Nine Mile Point 1 May-00 1.2

Nine Mile Point 2 Sep-00 1.0

Peach Bottom 2 Oct-98 2.80

Peach Bottom 3 Oct-99 0.66

Perry Feb-01 0.9

Quad Cities 1 Apr-99 0.6

Quad Cities 2 Jan-00 1.0

Vermont Yankee Apr-01 1.24

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Plant Chemistry Impact of NMCA

Over time, the amount of noble metals on the reactor components will decrease and reapplication
of noble metals will be required. To estimate when reapplication is required, durability
monitoring is performed using a Noble Metal Monitoring System (NMMS). The NMMS
contains a number of coupons, which are treated during the NMCA application and are exposed
to reactor coolant during plant operation. The coupons are removed at periodic intervals and
analyzed to quantify the amount of noble metals remaining on the surface. By trending the noble
metals retention, projections can be made to estimate when reapplication will be required.
Reapplication is recommended when “the total noble metal deposition on a plant’s deposition
2
monitoring coupons is projected to decrease into the range of 0.10 to 0.15 µg/cm ”; at that time
“the plant should prepare to reapply or should perform a more detailed evaluation of
reapplication timing”(2). Noble metal coupon loading results incorporated in the EPRI BWR
Chemistry Monitoring Database are summarized in Table 6-2.
Table 6-2
Noble Metal Coupon Loading

Coupon Loading (µg/cm2)

Coupon Exposure in Months

Plant 27- 33-

Initial 3-5 6-8 9-11 12-14 15-17 18-20 21-23 24-26
29 35

Duane Arnold (1st 0.16

0.219 0.126 0.106 0.149 0.123 0.136 0.133 0.126
application) 6

Duane Arnold
0.8
(Reapplication)

Hatch 2 1.38 1.17

Limerick 1 1.16 0.85

Nine Mile Point 1 3.4 1.9

Nine Mile Point 2 0.8 1.8

Peach Bottom 2 2.8 1.9 1.13

Peach Bottom 3 0.66 0.68

Quad Cities 1 1.2 0.8 0.7 0.6 0.2 0.6

Quad Cities 2 2.3 2.2 1.3 1.7

Vermont Yankee 1.24 1.86 1.23 1.55

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Plant Chemistry Impact of NMCA

Trends of the coupon loading data are shown in Figure 6-1. Five of the plants (Hatch 2,
Limerick 1, Peach Bottom 2, Quad Cities 1 and Quad Cities 2) exhibit a decreasing trend with
operating time. Duane Arnold (first application) and Peach Bottom 3, which had the lowest
initial loadings, exhibit an essentially flat trend. Nine Mile Point 1, which received the highest
initial loading, exhibits a steeply decreasing trend, while the Nine Mile Point 2 results suggest an
increase from the initial loading value.

4.0

3.5
Noble Metal Loading (µg/cm^2)

3.0

2.5

2.0

1.5

1.0

0.5

0.0
0 5 10 15 20 25 30 35 40
Months

DUA (1st) HAT2 LIM1 NMP1 NMP2 PB2 PB3 QUA1 VY QUA2

Figure 6-1
Noble Metal Loading vs Coupon Exposure Time

Nine Mile Point 1 has analyzed several surfaces since NMCA. The components removed and
noble metals deposition for each are listed in Table 6-3.

Reported coupon locations and temperature at the locations are listed in Table 6-4.

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Plant Chemistry Impact of NMCA

Table 6-3
Nine Mile Point 1 Artifacts

Temperature at Exposure Time Total Noble Metal

Description Location
Location (ºF) (months) Loading (µg/cm2)

Recirc
SS ECP Electrode 519-526 11 none detected
loop
40% Fe-Ni ECP Recirc
519-526 11 0.3
Electrode loop
Alloy 600 Shroud
Reactor 519-526 11 <0.1
Head Bolt
SS Indexing Collar Reactor 519-526 11 2.4
CS Cleanup Pipe RWCU 519-526 11 1.1

Table 6-4
NMCA Coupon Location and Temperature

Plant Coupon Location Temperature at Location (ºF)

Columbia RWCU
FitzPatrick CAVS sample line
Hatch 1 CAVS
Limerick 1 RWCU 513-530 (526 average)
Limerick 2 RWCU 498-531 (528 average)
Vermont Yankee RWCU 510-517

NMCA Guidelines

The impact of NMCA on shutdown dose rates and transient chemistry effects is addressed in
BWRVIP-92: NMCA Experience Report and Application Guidelines (2). In this section,
selected data from the EPRI BWR Iron/Chemistry Monitoring database are presented to illustrate
these effects.

NMCA drives the ECP down to low values (-500 mV SHE) at all wetted surfaces having a liquid
phase molar ratio of hydrogen to total oxidant of 2 or more. As a result the crud is restructured
from the hematite form (Fe2O3), which is stable under oxidizing conditions, to the spinel form
(Fe3O4), which is stable under reducing conditions. This is the same type of restructuring that has
occurred at most plants following the implementation of HWC. However, under NMCA with
hydrogen injection, the restructuring effects are greater. The corrosion films tend to restructure
on more of the surfaces as compared with a HWC environment. In addition, more of the primary
crud inventory on the fuel tends to restructure as well.
The restructuring of deposits at NMCA plants is evidences by increased insoluble and soluble
corrosion products (especially iron) in the reactor coolant. In addition, soluble and insoluble

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Plant Chemistry Impact of NMCA

forms of activation products such as Co-60, that are associated with the crud and corrosion films,
increase in the reactor coolant.

Increased drywell shutdown dose rates (measured at the BRAC points) that occur after initiating
HWC are mainly due to Co-60 incorporation into the corrosion film formed on the reactor
recirculation piping under reducing conditions. Addition of sufficient zinc into the feedwater to
maintain 5 – 10 ppb in reactor water has effectively countered the driving force for drywell dose
rates to increase under HWC. The benefits of zinc are greatest when zinc addition is initiated
and maintained at a sufficient level prior to the start of HWC. These same principles apply to
NMCA.

Summary of Plant Chemistry Responses

BWRVIP-92 (2) gives a summary and description of plant chemistry responses during the
operating period following NMCA. The objective of this section is to provide some actual plant
data showing these responses.

There are six plants for which post-NMCA operating and refueling outage data were evaluated in
preparing BWRVIP-92. These plants are Duane Arnold, Hatch 1, FitzPatrick, Peach Bottom 2,
Quad Cities 1 and Nine Mile Point 1. The first three of these plants experienced a reduction or
no significant change in drywell dose rates in the refueling outage after operating under NMCA
with hydrogen injection, while the latter three plants all experienced increased drywell shutdown
dose rates. Post-NMCA data for these plants on MSLRM response, reactor water conductivity,
offgas sum-of-six activity, Co-60, and reactor water iodines are presented. Feedwater dissolved
hydrogen data are also provided to enable assessment of its impact.

Feedwater Dissolved Hydrogen

The feedwater hydrogen concentration required at full power to achieve mitigation against
IGSCC is significantly reduced with NMCA. Of the six plants whose post-NMCA chemistry
response data are presented here, all except Nine Mile Point 1 had been injecting hydrogen for a
significant period prior to NMCA. Nine Mile Point 1 began injecting hydrogen to achieve 0.37
ppm in feedwater only about one month prior to NMCA. In May 2001, after the refueling outage,
feedwater hydrogen was reduced to 0.185 ppm to reduce the release rate of activated corrosion
products to the coolant while maintaining protection against IGSCC.

Feedwater hydrogen trend data for these plants are shown in Figures 6-2 through 6-7. Changes in
hydrogen injection rate are useful indicators for understanding some of the chemistry responses.

The Duane Arnold hydrogen trend data (Figure 6-2) begin after the first application of noble
metals; prior to NMCA, feedwater hydrogen averaged 0.65 ppm. The Peach Bottom 2 feedwater
hydrogen data (Figure 6-5) also begin after NMCA; the average concentration prior to NMCA
was approximately 0.3 ppm. Measured feedwater hydrogen data were provided by FitzPatrick
and Quad Cities 1; for the other four plants, the feedwater hydrogen concentrations are
calculated values based on the hydrogen injection rates and feedwater flow rates.

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Plant Chemistry Impact of NMCA

2.0
Feedwater Dissolved Hydrogen (ppm)

1.8
1.6
1.4
1.2
1.0
0.8
0.6
0.4
0.2
0.0
1/1/93 1/1/95 12/31/96 12/31/98 12/30/00 12/30/02
NMCA

Figure 6-2
Duane Arnold Feedwater Hydrogen Trend Data

2.0
1.8
1.6
Dissolved Hydrogen (ppm)

1.4
1.2
1.0
0.8
0.6
0.4
0.2
0.0
1/1/97 5/16/98 9/28/99 2/9/01 6/24/02 11/6/03

Feedwater NMCA

Figure 6-3
Hatch 1 Feedwater Hydrogen Trend Data

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Plant Chemistry Impact of NMCA

2.0
1.8
1.6
Dissolved Hydrogen (ppm)

1.4
1.2
1.0
0.8
0.6
0.4
0.2
0.0
12/1/96 12/1/97 12/1/98 12/1/99 11/30/00 11/30/01 11/30/02 11/30/03

Feedwater NMCA

Figure 6-4
FitzPatrick Feedwater Hydrogen Trend Data

2.0
1.8
1.6
Dissolved Hydrogen (ppm)

1.4
1.2
1.0
0.8
0.6
0.4
0.2
0.0
9/1/98 9/1/99 8/31/00 8/31/01 8/31/02 8/31/03

Figure 6-5
Peach Bottom 2 Feedwater Hydrogen Trend Data

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Plant Chemistry Impact of NMCA

2.0

1.5
Feedwater DH (ppm)

1.0

0.5

0.0
7/20/95 7/19/97 7/19/99 7/18/01 7/18/03
NMCA

Figure 6-6
Quad Cities 1 Feedwater Hydrogen Trend Data

1.0
0.9
0.8
Dissolved Hydrogen (ppm)

0.7
0.6
0.5
0.4
0.3
0.2
0.1
0.0
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Feedwater NMCA

Figure 6-7
Nine Mile 1 Feedwater Hydrogen Trend Data

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Plant Chemistry Impact of NMCA

MSLRM Response

A transient increase in MSLRM dose rate was observed at most plants during the initial
operating period after NMCA. These increases are attributed to increases in the volatile fraction
of N-16 caused by the reducing effect of surfaces that have noble metal. This effect gradually
subsides as noble metals are lost from those surfaces, as crud covers catalytically active sites,
and through removal by the reactor water cleanup system (2). MSLRM data provided by
FitzPatrick, Peach Bottom 2, and Nine Mile Point 1 are shown in Figures 6-8, 6-9, and 6-10,
respectively. As shown in Figure 6-8, the FitzPatrick post-NMCA MSLRM dose rate at 0.3 ppm
feedwater hydrogen peaked to lower values than the steady state dose rate with 0.5 ppm
hydrogen prior to NMCA. After the transient period, MSLRM dose rates were reduced further
after lowering the feedwater hydrogen concentration to 0.2 ppm.

Two initial post-NMCA MSLRM transients were observed at Peach Bottom 2, as shown in
Figure 6-9. The first MSLRM increase occurred during operation with minimum hydrogen
injection (hydrogen injected briefly) followed by a second transient after hydrogen was injected
to achieve a feedwater concentration of 0.15 ppm. After these initial transients, the MSLRM
reading trended proportionally with feedwater hydrogen concentration.

The transient increase in MSLRM dose rates after NMCA at Nine Mile Point 1 (Figure 6-10)
was higher than the pre-NMCA readings. This is because Nine Mile Point 1 had only a short
operating time of about 0.37 ppb feedwater hydrogen prior to NMCA. At this level of hydrogen
injection (prior to NMCA), MSLRM dose rates do not increase significantly.

3000

2500
Average MSLRM (mR/hr)

2000

1500

1000

500

0
12/1/96 12/1/97 12/1/98 12/1/99 11/30/00 11/30/01 11/30/02 11/30/03

NMCA

Figure 6-8
FitzPatrick Main Steam Line Radiation Monitor Trend Data

6-10
 EPRI Licensed Material

Plant Chemistry Impact of NMCA

Average Main Steam Line Radiation (mR/hr) 2000

1500

1000

500

0
9/1/98 9/1/99 8/31/00 8/31/01 8/31/02 8/31/03

Figure 6-9
Peach Bottom 2 Main Steam Line Radiation Monitor Trend Data

2000
Average Main Steam Line Radiation (mR/hr)

1800
1600
1400
1200
1000
800
600
400
200
0
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

NMCA

Figure 6-10
Nine Mile Point 1 Main Steam Line Radiation Monitor Trend Data

6-11
 EPRI Licensed Material

Plant Chemistry Impact of NMCA

Reactor Water Conductivity Response

The transient increase in reactor water conductivity observed after NMCA with hydrogen
injection is due to an increase in soluble metals, mainly iron, in the reactor coolant. The iron is
believed to be a product of the restructuring of primary system crud from the Fe2O3 form to the
Fe3O4 spinel form. The magnitude and duration of the transient is dependent on the extent of
restructuring; plants with significant prior operation on moderate HWC experienced less of an
impact than plants with low hydrogen or no prior operation with hydrogen injection.

Reactor water conductivity trends are shown in Figures 6-11, 6-12, 6-13, 6-14, 6-15 and 6-16 for
Duane Arnold, Hatch 1, FitzPatrick, Peach Bottom 2, Quad Cities 1 and Nine Mile Point 1,
respectively. Duane Arnold, the only plant to reapply noble metals, experienced a conductivity
transient following the second application, as shown in Figure 6-11.

1.0
0.9
0.8
0.7
RCI Conductivity

0.6
0.5
0.4
0.3
0.2
0.1
0.0
1/1/93 1/1/95 12/31/96 12/31/98 12/30/00 12/30/02

NMCA

Figure 6-11
Duane Arnold Reactor Water Conductivity Trend Data

6-12
 EPRI Licensed Material

Plant Chemistry Impact of NMCA

1.0

0.8
RCI Conductivity

0.6

0.4

0.2

0.0
1/1/97 5/16/98 9/28/99 2/9/01 6/24/02 11/6/03

RCI Cond NMCA

Figure 6-12
Hatch 1 Reactor Water Conductivity Trend Data

1.0
0.9
0.8
RCI Conductivity (µS/cm)

0.7
0.6
0.5
0.4
0.3
0.2
0.1
0.0
12/1/96 12/1/97 12/1/98 12/1/99 11/30/00 11/30/01 11/30/02 11/30/03

NMCA

Figure 6-13
FitzPatrick Reactor Water Conductivity Trend Data

6-13
 EPRI Licensed Material

Plant Chemistry Impact of NMCA

1.0
0.9
0.8
0.7
RCI Conductivity

0.6
0.5
0.4
0.3
0.2
0.1
0.0
9/1/98 9/1/99 8/31/00 8/31/01 8/31/02 8/31/03
NMCA

Figure 6-14
Peach Bottom 2 Reactor Water Conductivity Trend Data

1.0
0.9
0.8
0.7
RCI Conductivity

0.6
0.5
0.4
0.3
0.2
0.1
0.0
7/20/95 7/19/97 7/19/99 7/18/01 7/18/03

NMCA

Figure 6-15
Quad Cities 1 Reactor Water Conductivity Trend Data

6-14
 EPRI Licensed Material

Plant Chemistry Impact of NMCA

1.0
0.9
0.8
RCI Conductivity (µS/cm)

0.7
0.6
0.5
0.4
0.3
0.2
0.1
0.0
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

NMCA

Figure 6-16
Nine Mile Point 1 Reactor Water Conductivity Trend Data

Offgas Sum of Six Activity Response

Some plants have experienced increases in noble gas activity in the offgas after NMCA. This is
attributed to small amounts of uranium from the fuel that incorporated into crud deposits and is
released in the post-NMCA crud restructuring process (2).

Offgas trend data in the form of Sum of Six Activity (i.e., the sum of the activities of Kr-87, Kr-
88, Kr-85m, Xe-133, Xe-135, and Xe-138) are plotted in Figures 6-17, 6-18, 6-19, 6-20 and 6-21
for Duane Arnold, Hatch 1, FitzPatrick, Peach Bottom 2 and Nine Mile Point 1, respectively. As
shown in Figure 6-17, Duane Arnold experienced a larger increase in offgas noble gas activity
release rate following the initial noble metals application than after the first reapplication.
FitzPatrick (Figure 6-19), Peach Bottom 2 (Figure 6-20) and Nine Mile Point 1 (Figure 6-21)
also experienced offgas Sum of Six increases after NMCA, while the offgas Sum of Six release
rate decreased after NMCA at Hatch 1.
Nine Mile Point 1 is unique among these plants because hydrogen was started at about the same
time NMCA was performed. FRI (Fuel Reliability Index) calculations indicate that fuel integrity
has been maintained despite the increase in offgas activity. The increase is attributed to the crud
restructuring process, which was more extreme for Nine Mile Point 1 than for the other plants
that performed NMCA.

6-15
 EPRI Licensed Material

Plant Chemistry Impact of NMCA

300
275
250
Offgas Sum of 6 (µCi/sec)

225
200
175
150
125
100
75
50
25
0
1/1/93 1/1/95 12/31/96 12/31/98 12/30/00 12/30/02

Sum of 6 NMCA

Figure 6-17
Duane Arnold Sum of Six Noble Gases Trend Data

10000
9000
8000
Offgas Sum of 6 (µCi/sec)

7000
6000
5000
4000
3000
2000
1000
0
1/1/97 5/16/98 9/28/99 2/9/01 6/24/02 11/6/03

Sum of 6 NMCA

Figure 6-18
Hatch 1 Sum of Six Noble Gases Trend Data

6-16
 EPRI Licensed Material

Plant Chemistry Impact of NMCA

700

600
Offgas Sum of 6 (µCi/sec)

500

400

300

200

100

0
12/1/96 12/1/97 12/1/98 12/1/99 11/30/00 11/30/01 11/30/02 11/30/03

Sum of 6 NMCA

Figure 6-19
FitzPatrick Sum of Six Noble Gases Trend Data

700

600
Offgas Sum of 6 (µCi/sec)

500

400

300

200

100

0
9/1/98 9/1/99 8/31/00 8/31/01 8/31/02 8/31/03

Sum of 6 NMCA

Figure 6-20
Peach Bottom 2 Sum of Six Noble Gases Trend Data

6-17
 EPRI Licensed Material

Plant Chemistry Impact of NMCA

1500
1400
1300
1200
Offgas Sum of 6 (µCi/sec)

1100
1000
900
800
700
600
500
400
300
200
100
0
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Sum of 6 NMCA

Figure 6-21
Nine Mile Point 1 Sum of Six Noble Gases Trend Data

Co-60 Response

Soluble and insoluble Co-60 data are presented for Duane Arnold (Figure 6-22), Hatch 1 (Figure
6-23), FitzPatrick (Figure 6-24), Peach Bottom 2 (Figure 6-25), Quad Cities 1 (Figure 6-26), and
Nine Mile Point 1 (Figure 6-27). All plants that have applied noble metals have experienced
some increase in both soluble and insoluble Co-60. The post-NMCA data for Peach Bottom 2
(Figure 6-25) show increases above pre-NMCA levels of approximately 1E-4 µCi/ml soluble Co-
60 and 1E-5 µCi/ml insoluble Co-60. The impact of increased Co-60 on drywell shutdown dose
rates is largely dependent on whether sufficient zinc was maintained both before and after
operation with NMCA.

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 EPRI Licensed Material

Plant Chemistry Impact of NMCA

1.E-01
Reactor Water Co-60 (µCi/ml)

1.E-02

1.E-03

1.E-04

1.E-05

1.E-06
1/1/93 1/1/95 12/31/96 12/31/98 12/30/00 12/30/02

Insoluble Co-60 Soluble Co-60 NMCA

Figure 6-22
Duane Arnold Co-60 Trend Data

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
1/1/97 5/16/98 9/28/99 2/9/01 6/24/02 11/6/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure 6-23
Hatch 1 Co-60 Trend Data

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 EPRI Licensed Material

Plant Chemistry Impact of NMCA

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
12/1/96 12/1/97 12/1/98 12/1/99 11/30/00 11/30/01 11/30/02 11/30/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure 6-24
FitzPatrick Co-60 Trend Data

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06

1.E-07
9/1/98 9/1/99 8/31/00 8/31/01 8/31/02 8/31/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure 6-25
Peach Bottom 2 Co-60 Trend Data

6-20
 EPRI Licensed Material

Plant Chemistry Impact of NMCA

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
7/20/95 7/19/97 7/19/99 7/18/01 7/18/03

Insoluble Co-60 Soluble Co-60 Total Co-60 NMCA

Figure 6-26
Quad Cities 1 Co-60 Trend Data

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure 6-27
Nine Mile Point 1 Co-60 Trend Data

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 EPRI Licensed Material

Plant Chemistry Impact of NMCA

Iodine Response

Reactor water iodine trend plots are presented for Duane Arnold (Figure 6-28), Hatch 1 (Figure
6-29), FitzPatrick (Figure 6-30), Peach Bottom 2 (Figure 6-31), Quad Cities 1 (Figure 6-32) and
Nine Mile Point 1 (Figure 6-33). As shown in Figure 6-28, Duane Arnold did not experience a
significant change in iodine activity during the operating period following noble metals
reapplication. Hatch 1 and FitzPatrick experienced significant decreases in iodine activity after
NMCA, as shown in Figures 6-29 and 6-30.

BWRVIP-92 (2) provides guidance for interpreting reactor water iodine results following
NMCA: “During operation following NMCA, reactor water concentrations of short half-life
iodine isotopes such as I-134 can be suppressed relative to longer half-life isotopes such as I-131.
As a result, a standard fission product analysis using iodine data may incorrectly indicate a
diffusion isotopic distribution suggestive of fuel leakage. Post-NMCA noble gas data are not
affected in this manner. Therefore, noble gas data should be used in preference to iodine data in
order to assess whether a fuel leak is present.”

The Nine Mile Point 1 data do not exhibit a large decrease in iodines (Figure 6-33). I-133
increased after NMCA, coinciding with the start of hydrogen injection. I-131 remained about the
same, and I-132 and I-134 decreased slightly. While this may appear to suggest a possible fuel
leak, the offgas data clearly indicate that fuel integrity was not affected. This confirms the
BWRVIP-92 guidance that noble gas data are the preferable indicators of fuel integrity during
post-NMCA plant operation.

Quad Cities 1 did not experience a large decrease in iodines either (Figure 6-32). I-131
increased, and I-133 remained about the same. I-132, I-134, and I-135 decreased slightly.
Offgas data are not available for Quad Cities 1, but no fuel reliability problems have been
reported.

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Plant Chemistry Impact of NMCA

1.E-03
Reactor Water Iodines (µCi/ml)

1.E-04

1.E-05

1.E-06
1/1/93 1/1/95 12/31/96 12/31/98 12/30/00 12/30/02

I131 I132 I133 I134 I135 NMCA

Figure 6-28
Duane Arnold Reactor Water Iodine Isotopic Trend Data

1.E-01
Reactor Water Iodines (µCi/ml)

1.E-02

1.E-03

1.E-04

1.E-05

1.E-06
1/1/97 5/16/98 9/28/99 2/9/01 6/24/02 11/6/03

I131 I132 I133 I134 I135 NMCA

Figure 6-29
Hatch 1 Reactor Water Iodine Isotopic Trend Data

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Plant Chemistry Impact of NMCA

Reactor Water Iodines (µCi/ml) 1.E-02

1.E-03

1.E-04

1.E-05

1.E-06

1.E-07
12/1/96 12/1/97 12/1/98 12/1/99 11/30/00 11/30/01 11/30/02 11/30/03

I131 I132 I133 I134 I135 NMCA

Figure 6-30
FitzPatrick Reactor Water Iodine Isotopic Trend Data

1.E-02
Reactor Water Iodines (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
9/1/98 9/1/99 8/31/00 8/31/01 8/31/02 8/31/03

I131 I132 I133 I134 I135 NMCA

Figure 6-31
Peach Bottom 2 Reactor Water Iodine Isotopic Trend Data

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 EPRI Licensed Material

Plant Chemistry Impact of NMCA

Reactor Water Iodines (µCi/ml) 1.E-02

1.E-03

1.E-04

1.E-05

1.E-06
7/20/95 7/19/97 7/19/99 7/18/01 7/18/03

I-131 I-132 I-133 I-134 I135 NMCA

Figure 6-32
Quad Cities 1 Reactor Water Iodine Isotopic Trend Data

1.E-02
Reactor Water Iodines (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

I131 I132 I133 I134 I135 NMCA

Figure 6-33
Nine Mile Point 1 Reactor Water Iodine Isotopic Trend Data

6-25
 EPRI Licensed Material

Plant Chemistry Impact of NMCA

Plants Experiencing a Decrease in Drywell Shutdown Dose Rates after

NMCA

BRAC trend plots with pertinent milestone events indicated are presented for Duane Arnold,
Hatch 1, and FitzPatrick in Figures 6-34, 6-35 and 6-36, respectively. All three of these plants
had well-established HWC programs and injected DZO to maintain reactor water zinc in the 5 –
10 ppb range prior to and after NMCA. Duane Arnold and Hatch 1 applied noble metals at the
end of the fuel cycle while FitzPatrick performed a mid-cycle application.

2000
6 scfm 9 scfm 15 scfm 6 scfm
HWC
Pleat ed filters
1500
Dose Rate (mR/hr)

DZO

NMCA
1000
Power uprate

Chem decon

Chem decon

Chem decon

Chem decon
500

0
Jan-80 Sep-82 Jun-85 Mar-88 Dec-90 Sep-93 Jun-96 Mar-99 Nov-01

BRAC M ilestones Co-60 Dose

Figure 6-34
Duane Arnold BRAC Dose Rates and Milestones

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Plant Chemistry Impact of NMCA

400
HWC 22 scfm 30 45 50 45 5-6 8 scfm
350
Pleated Filters
300
NZO DZO
Dose Rate (mR/hr)

Chem decon, power uprate 1996

250

MSR retube, power uprate 1999

O2 Inj

CRB replacement complete

200 NMCA
Retube condenser 6/90
Begin replacing CRBs

150
Chem decon
100
50
0
Oct-86 Jun-89 Mar-92 Dec-94 Sep-97 Jun-00
BRAC (DPGSM) BRAC (RO2A) Milestones Co-60

Figure 6-35
Hatch 1 BRAC Dose Rates and Milestones

10.5 scfm
250
11 scfm 13.5 scfm 18.5 scfm 8 scfm
HWC
6 scfm
200
Begin cobalt material replacement 1985

Chem Decon, Retube Condenser 1994

NZO DZO
Dose Rate (mR/hr)

150
NMCA

100
Power uprate 1997
Chem Decon 1988

Chem Decon 1992

50

0
Jun-84 Feb-87 Nov-89 Aug-92 May-95 Feb-98 Nov-00
BRAC Milestones Co-60 Dose

Figure 6-36
FitzPatrick BRAC Dose Rates and Milestones

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Plant Chemistry Impact of NMCA

Hatch 2 completed its first refueling outage in September 2001 after NMCA was performed in
March 2000. The BRAC/milestone plot shown in Figure 6-37 shows the remarkable decrease in
the BRAC average dose rate to 31 mR/hr in September 2001.

400
17 scfm 30 36 40 - 45 50 - 65 8 scfm
HWC
350
NZO DZO
300
Dose Rate (mR/hr)

O2 Inj
250 Pleated Filters

Complete CRB replacement

MSR retube, power uprate
200
Begin CRB replacement
Recirc pipe replacement

NMCA
150
Retube condenser

Power uprate
100
50
0
Aug-83 Apr-86 Jan-89 Oct-91 Jul-94 Apr-97 Jan-00 Sep-02

RO2A DPGSM Milestones Co-60 Dose

Figure 6-37
Hatch 2 BRAC Dose Rates and Milestones

Plants Experiencing an Increase in Drywell Shutdown Dose Rates after

NMCA

BRAC trend plots with pertinent milestone events noted are presented for Peach Bottom 2, Quad
Cities 1, and Nine Mile Point 1 in Figures 6-38, 6-39 and 6-40, respectively. These plants
experienced significant increases in drywell shutdown dose rates in their refueling outages
following NMCA. Peach Bottom 2 and Quad Cities 1 maintained reactor water zinc at 2 – 3 ppb
after NMCA, and Peach Bottom 2 also maintained <5 ppb (about 4 ppb) reactor water zinc prior
to NMCA. Nine Mile Point 1, which experienced the largest dose rate increase, does not add
zinc and had virtually no time on HWC prior to NMCA. Peach Bottom 2 applied noble metals at
the end of the fuel cycle, while Quad Cities 1 and Nine Mile Point 1 performed mid-cycle
applications. Dose rate measurements taken during mid-cycle outages show that Quad Cities 1
BRAC dose rates have decreased since the last refuel outage, while Nine Mile Point 1 BRAC
dose rates increased.

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 EPRI Licensed Material

Plant Chemistry Impact of NMCA

250
NZO DZO

200 Pleated filters

Dose Rate (mR/hr)

HWC
Recirc pipe replacement 1985

150

Retube condenser 1991

`
100
NMCA

Power Uprate
Chem decon

50

0
Jun-84 Feb-87 Nov-89 Aug-92 May-95 Feb-98 Nov-00
BRAC Milestones Co-60

Figure 6-38
Peach Bottom 2 BRAC Dose Rates and Milestones

Measurement taken
4 days after
shutdown.
800

700
NMCA
600 47 scfm
HWC
Dose Rate (mR/hr)

500
Pleated Filters
400
DZO
300
Chem Decon 1987

Chem Decon 1989

Chem Decon 1986

Chem decon 1990

Chem decon 1992

Chem decon 1994

Chem decon 1998

Chem decon 1984

Chem decon 1996

200 Measurement taken

16 days after
100 shutdown.

0
Jan-84 Sep-86 Jun-89 Mar-92 Dec-94 Sep-97 Jun-00
BRAC Milestones BRAC - All

Figure 6-39
Quad Cities 1 BRAC Dose Rates and Milestones

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Plant Chemistry Impact of NMCA

T wo hrs
1500 aft er sd
Natural zinc from condenser

1250

Recirc pipe replacement 1982-83

Dose Rate (mR/hr)

Decon prior to pipe replacement

1000
HWC 8 scfm
750
NMCA

500
Partial decon

250

0
Jan-81 Sep-83 Jun-86 Mar-89 Dec-91 Sep-94 Jun-97 Mar-00 Nov-02

BRAC BRAC (all measurements) Milestones Co-60 Dose

Figure 6-40
Nine Mile Point 1 BRAC Dose Rates and Milestones

Summary of Recommendations with NMCA

Recommendations for controlling drywell dose rates with NMCA, which are detailed in
BWRVIP-92 (2), include:
• Maintaining 5 – 10 ppb reactor water zinc prior to and following NMCA
• Maintaining HWC at high availability prior to NMCA
• Implement end of cycle noble metal application rather than a mid-cycle application
• Maintaining 100% capacity RWCU operation after NMCA
• Increasing hydrogen injection in small steps after NMCA

NMCA Impact on Nine Mile Point 2 RWCU Filter Demineralizer (F/D)

Performance

Issue Description

After NMCA at Nine Mile Point 2, rapid increases in RWCU (Reactor Water Cleanup) F/D
effluent conductivity were measured with hydrogen injection in service. When hydrogen
injection was turned off, F/D effluent conductivity decreased to normal levels (<0.06 µS/cm).

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Plant Chemistry Impact of NMCA

The F/D effluent conductivity limit of 0.1 µS/cm was rapidly approached, causing the plant to
take actions to avoid alarm conditions. The RWCU system includes four F/D vessels, A through
D. The rate of effluent conductivity increase was highest for F/D "A" and next highest for F/D
"C". The trends from the conductivity monitor readings were confirmed by grab sample
measurements.

Other plants have reported short RWCU F/D run lengths following NMCA. Data from the EPRI
BWR Monitoring Database and Surveys indicated 5 of 15 plants experienced increased RWCU
F/D effluent conductivity after NMCA. The conductivity was reported to have returned to
normal in approximately 1.5 to 3 months. No other North American plants reported RWCU F/D
effluent conductivity greater than RWCU inlet conductivity. The experience of Nine Mile Point
2 is presented here to inform other plants of the impact and countermeasures considered.

Nine Mile Point 2 RWCU Conditions

1. The normal precoat is 4 containers of all powdered resin precoat with a 2:3 (dry weight)
cation/anion ratio for each F/D.

2. Septum media are diffusion bonded SS mesh, 64 microns absolute.

3. Individual F/D vessel flow = 170 – 175 gpm.

4. NMCA was applied 09/09 – 09/12/00.

5. Hydrogen injection was started on 1/26/01. The target hydrogen injection rate at full power is
15 scfm, which results in a final feedwater hydrogen concentration of 0.3 ppm.

6. NZO addition started in 04/88.

7. Switched to DZO addition 07/99.

8. The tubesheet to vessel seal, separating the influent and effluent sections of the F/D vessel, is
a metal-to-metal seal. This seal was not tight, as indicated by resin leakage to the effluent
strainer of the “B” F/D. This seal was improved for the “B” vessel, but not for the other
F/Ds.

9. The transition between precoat mode and hold mode is suspected to result in a momentary
low flow condition that could affect precoat retention.

Nine Mile Point 2 Data Evaluation (3)

Reactor water trend data are shown in Figure 6-41. Reactor water conductivity increased
following the startup of hydrogen injection on 01/26/01. The reactor water conductivity increase
coincided with an increase in soluble iron. Conductivity showed a decreasing trend to
approximately 0.1 µS/cm by 06/06/01. Conductivity returned to pre-hydrogen injection levels of
0.077 – 0.083 µS/cm in August 2001 following a sustained period of hydrogen injection. In

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Plant Chemistry Impact of NMCA

October 2001 conductivity again exceeded 0.1 µS/cm following return of hydrogen injection
after it was out of service for five days. Following the spring 2002 outage, hydrogen injection
has been in service consistently and conductivity is again at pre-hydrogen injection levels.

0.50 50
NMCA
0.45 45

0.40 HWC 40

0.35 35

Rx Sol Fe, Zn or Cu, ppb

Rx Conductivity, µS/cm

0.30 30

0.25 25

0.20 20

0.15 15

0.10 10

0.05 5

0.00 0
06/23/00 10/21/00 02/18/01 06/18/01 10/16/01 02/13/02 06/13/02

Rx Cond (Monit) Rx Fe Soluble ppb Rx Zn Soluble ppb Rx Cu Soluble ppb

Figure 6-41
Nine Mile Point 2 Reactor Water Trend Data

Performance data for RWCU F/Ds “A”, “B”, “C”, and “D” are shown in Figures 6-42, 6-43, 6-
44 and 6-45, respectively. The vertical lines on each figure indicate the date an F/D was placed
in service after backwash/precoat. The F/D “A” results in Figure 6-42 show a trend of increasing
effluent conductivity with time after a new precoat was placed in service. This behavior is
indicative of an increase in ion concentrations, such as would be experienced with ion exchange
breakthrough. Following the spring 2002 outage, conductivity again began to increase but then it
leveled off at less than 0.1 µS/cm. Even after a precoat change, the conductivity has remained
essentially constant.

RWCU F/D “C” and “D”, Figures 6-44 and 6-45 respectively, show a conductivity increase
behavior similar to that of “A” prior to the spring 2002 outage. Following the outage, F/D “C”
conductivity increased substantially. However, following a precoat change, the conductivity has
been very high, approximately 0.13 µS/cm, but fairly stable. F/D “D” conductivity following the
outage has been increasing with a smaller rate of increase following a precoat change.

In contrast, the F/D “B” results, shown in Figure 6-43, indicate sustained periods of low effluent
conductivity. The run starting on 03/23/01 showed a gradual effluent conductivity break over a

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Plant Chemistry Impact of NMCA

period of about one month as RWCU inlet conductivity increased from approximately 0.11
µS/cm to 0.14 µS/cm. From September 2001 up to the spring 2002 outage, F/D “B” conductivity
increased slowly and was typically less than 0.08 µS/cm. Following the outage the F/D “B”
conductivity has been reasonably steady at approximately 0.06 µS/cm.

0.20 200
0.18 180
0.16 160
Conductivity (µS/cm)

0.14 140

Flow (gpm)
0.12 120
0.10 100
0.08 80
0.06 60
0.04 40
0.02 20
0.00 0
06/23/00 10/21/00 02/18/01 06/18/01 10/16/01 02/13/02 06/13/02

Conductivity (Monit) Conductivity (grab) New Precoat Date

Rx Cond. (Monit) Filter Flowrate

Figure 6-42
Nine Mile Point 2 RWCU F/D “A” Performance Data

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Plant Chemistry Impact of NMCA

0.20 200
0.18 180
0.16 160
Conductivity (µS/cm)

0.14 140
0.12 120

Flow (gpm)
0.10 100
0.08 80
0.06 60
0.04 40
0.02 20
0.00 0
05/14/00 09/11/00 01/09/01 05/09/01 09/06/01 01/04/02 05/04/02 09/01/02
Conductivity (Monit) Conductivity (grab) Rx Cond (Monit)
New Precoat Date Filter Flowrate

Figure 6-43
Nine Mile Point 2 RWCU F/D “B” Performance Data

0.20 200
0.18 180
0.16 160
Conductivity (µS/cm)

0.14 140
0.12 120
Flow (gpm)

0.10 100
0.08 80
0.06 60
0.04 40
0.02 20
0.00 0
05/14/00 09/11/00 01/09/01 05/09/01 09/06/01 01/04/02 05/04/02 09/01/02

Conductivity (Monit) Conductivity (grab) Rx Cond (Monit)

New Precoat Date Filter Flowrate

Figure 6-44
Nine Mile Point 2 RWCU F/D “C” Performance Data

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Plant Chemistry Impact of NMCA

0.20 200
0.18 180
0.16 160
0.14 140
Conductivity (µS/cm)

0.12 120

Flow (gpm)
0.10 100
0.08 80
0.06 60
0.04 40
0.02 20
0.00 0
05/14/00 09/11/00 01/09/01 05/09/01 09/06/01 01/04/02 05/04/02 09/01/02
Conductivity (Monit) Conductivity (grab) Rx Cond (Monit)
New Precoat Date Filter Flowrate

Figure 6-45
Nine Mile Point 2 RWCU F/D “D” Performance Data

Chemistry results are summarized as follows:

1. Anions in RWCU F/D effluents were less than detectable. Effluent silica is not routinely
measured.

2. The reactor water conductivity initially increased with hydrogen injection in service, which
2+
corresponded to an increase in soluble iron (assumed to be Fe ).

3. A reactor water ion-conductivity balance was performed using 02/13/01 ion concentration
data. Concentrations of 11.7 ppb soluble Fe2+, 6.86 ppb soluble Zn2+, 0.419 ppb soluble Cu2+,
0.109 ppb soluble Ni2+ and 1.09 ppb SO42- impart a specific conductance of 0.169 µS/cm. This
is in agreement with the measured value of 0.170 µS/cm. On 4/11/01, an ion-conductivity
balance resulted in a calculated specific conductance of 0.114 µS/cm, while the measured
conductivity prior to the sample was 0.109 µS/cm and increasing. Therefore, the soluble
metals and anions measured in reactor water appear to account for the measured
conductivity.

4. RWCU influent and F/D effluent analysis results on samples taken on 04/11/01 are shown in
Table 6-5. Sulfate was the only detectable anion in reactor water and all anions were less
than detectable in the F/D effluents. Measured and calculated effluent conductivity values
increase as soluble iron increases. The reason that the calculated effluent conductivity values
are lower than the measured values may be that the recovery of soluble metals (especially

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Plant Chemistry Impact of NMCA

iron) by the ion cation exchange filter in the sample apparatus was significantly less than
100%.
Table 6-5
Nine Mile Point 2 RWCU Influent and F/D Effluent Analysis on 04/11/01

Fe Cu Ni Cr Zn Calc
Sample Cond SO4
Liters Sol. Sol. Sol. Sol. Sol. Cond
Point (µS/cm) (ppb)
(ppb) (ppb) (ppb) (ppb) (ppb) (µS/cm)
Influent 29.2 5.6 1.79 0.51 0.13 5.51 0.109 1.32 0.119

A Effluent 30.8 8.28 0 0 0 0.11 0.127 0.092

B Effluent 30.8 1.37 0.13 0.04 0 0.58 0.059 0.057

C Effluent 30.8 6.54 0.02 0.04 0.01 0.17 0.086 0.080

5. Based on cations in reactor water consisting mainly of 8 ppb soluble iron, 5.5 ppb soluble
zinc and 2 ppb soluble copper, the precoat material cation resin loading rate is 1.09 percent of
total sites per day.

6. Special gamma isotopic analyses on samples from 06/13/01 show about two times as much
Na-24 and Co-60 in the “A” effluent as in the “B” F/D effluent. However, the levels were
low in both samples.

7. A scrape sample from the RWCU filter effluent line was obtained to try to detect the
presence of Pt and Rh. The XRF analysis method was unable to detect the presence of noble
metals in the sample.

Conclusions on Nine Mile Point 2 RWCU F/D Performance

The rapid effluent conductivity increase in the “A”, “C” and “D” RWCU F/D effluents following
NMCA with hydrogen injection in service was attributed to an increase in soluble metals while
effluent anions remained below detectable levels. Consequently, this performance issue did not
represent a significant challenge to corrosion control of primary system components or materials.
The main impact was short RWCU F/D run times, causing excess cost and burial volumes, and
more frequent actions by the plant staff to avoid effluent conductivity alarms.

The “A” F/D performance initially appeared to be the poorest and may be associated with
frequent trips of the “A” Hold Pump. In contrast, the “B” RWCU F/D showed a much less rapid
effluent conductivity increase than the other F/Ds. The “B” F/D performance indicated that the
precoat mixture currently in use was capable of controlling soluble iron and other cationic and
anionic impurities.

The rapid conductivity breakthrough of the “A”, “C” and “D” F/Ds could be due to non-uniform
precoat distribution or to soluble metals going into solution downstream of the precoat. Soluble
iron removal by powdered resin precoat materials is not highly efficient, and non-ideal ion

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exchange precoat distribution would have a pronounced impact on soluble metals breakthrough.
Rapid soluble iron breakthrough can occur due to thinly precoated areas of the metal septa, most
likely due to partial fouling with crud and/or resins.

The early soluble iron breakthrough for the “A”, “C” and “D” F/Ds could be exacerbated by the
low cation to anion ratio of the precoat material. The cation/anion ratio is 2:3 (by dry weight), or
approximately 0.83:1 total cation to anion exchange sites. This powdered resin mixture is
normally appropriate when reactor water is neutral or even slightly basic due to zinc addition.
However, as the iron oxides in the primary system restructure due the change from oxidizing to
highly reducing conditions following the start of HWC/NMCA, the sum of the equivalent
concentrations of cationic impurities in reactor water is much greater than that of anions. The
impact on conductivity breakthrough of non-idealities in the distribution of the 2:3 cation/anion
precoat is amplified under these chemistry conditions.
However, the superior performance of the “B” F/D in comparison to the others may be due the
service history of this F/D. The “B” F/D had less service time than the other F/Ds and had minor
or no use while performing NMCA. This F/D was repetitively backwashed prior to disassembly
for maintenance in the summer of 2000. During this time, the “B” F/D was out of service for
approximately three months. After reassembly, F/D “B” was again backwashed, and then
precoated and returned to service on 09/21/00. The combination of less service time, repetitive
backwashing and work on the hold pump are probable contributing factors to the superior
performance of the “B” F/D over the others. In addition, if iron going into solution downstream
of the precoat when hydrogen is injected is the cause, then the “B” F/D should be less affected
since it would not have seen significant platinum and rhodium loading during NMCA. The
improved tubesheet to vessel seal of the “B” F/D was not a significant factor in its improved
soluble iron removal. In addition, contamination of sample lines with platinum and rhodium
does not appear likely since these lines were isolated during the performance of NMCA.
Some initial actions considered by Nine Mile Point 2 to either improve the F/D effluent
conductivity control or eliminate possible causes under NMCA/hydrogen injection chemistry
conditions are as follows:
1. Consider lowering the hydrogen injection rate to decrease the soluble iron source term while
maintaining the required hydrogen to total oxidant molar ratio. Lowering hydrogen injection
will expose less of the fuel surface area to highly reducing conditions, thus limiting the
amount of iron oxide undergoing restructuring and reducing the rate at which iron is imparted
to the reactor coolant. The target hydrogen injection rate with NMCA at full power at Nine
Mile Point 2 is 15 scfm, which results in a final feedwater hydrogen concentration of 0.30
ppm. Lowering the feedwater hydrogen concentration to approximately 0.20 - 0.25 ppm
(10.0 - 12.5 scfm at full power) may provide a significant reduction in soluble iron while
maintaining protection against IGSCC. As iron slowly decreases with time under
NMCA/HWC, then gradually increase hydrogen injection.
2. Take effort to reverse the possible fouling of the "A", "C" and "D" F/Ds. Such efforts as
double backwashes and practical means of drying the septa before backwashing should be
considered. Air drying tends to shrink the resins, facilitating backwashing them out of the
septa screen openings.

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3. Consider the application of a premixed powdered resin precoat material with an increased
fraction of cation resin compared with the present mixture. For example, changing from the
2:3 cation/anion (dry weight) mixture to a 1:1 cation/anion (dry weight mixture) increases the
quantity of cation resin available and increases the ratio of cation/anion exchange sites ratio
from 0.83 to 1.25. For maximum potential effectiveness, the increased cation resin precoat
can be applied in steps using the uniform precoat feed system as described below.
4. To improve the uniformity and coverage of the septa by the precoat material, apply the
precoat material in two steps, with half the precoat material applied in each step. The two-
step process and the more dilute precoat slurry promote improved precoat uniformity and
more complete coverage of all porous areas of the septa.
5. As a long-term improvement action, assure that adequate flow is maintained during the
transition from the precoat mode to the hold mode. Pump performance data indicate that
during this transition there could be a period of low flow, which could disturb the precoat.
The station performed actions 2 – 4 and found no significant improvement in RWCU F/D
effluent conductivity. The precoat-hold transition (action 5) is not a probable cause as only
soluble iron is present in the F/D effluent and its concentration is higher than the influent soluble
iron. Iron from sample lines also is not a probable cause since the RWCU inlet sample line had
greater exposure to noble metal chemicals during application than F/D effluent lines. Also both
the RWCU inlet (after the non-regenerative heat exchanger) and F/D effluent samples are cool at
the process tap. Comparisons of grab sample conductivities to on-line conductivity readings are
consistent, indicating that the conductivity cells are performing well.
Reactor water soluble iron became steady at around 5 ppb going into the spring 2002 outage and
has remained at this level for at least three months following the outage. This is significantly
higher than the pre-hydrogen injection levels of 1 ppb. RWCU F/D effluent iron is indicated to
be higher than inlet iron, even early in a F/D run with a fresh precoat, and the effluent
conductivity of all vessels is low when hydrogen is not in service.

The “B” F/D, which had little or no service during the noble metals application, continues to
perform well. Therefore, it appears that the soluble iron in the “A”, “C” and “D” RWCU F/D
effluents is caused by iron going into solution downstream of the precoats when hydrogen is in
service. It is hypothesized that noble metal deposition downstream of the precoats drives down
the ECP when sufficient hydrogen is present, causing iron oxide restructuring to the soluble
form.

At normal reactor coolant operating temperature the ECP under NWC is typically greater than
0.0 V (SHE) and under HWC the ECP is typically less than -0.300 V (SHE). Pourbaix diagrams
indicate a region of stable oxide at high and low temperatures under NWC. Under HWC, the
Pourbaix diagrams indicate regions of iron solubility and restructuring at low temperatures. At
operating temperature, when the ECP goes from greater than 0.0 V (SHE) to less than –0.300 V
(SHE) the Pourbaix diagram indicates iron restructures from Fe2O3 to Fe3O4. This causes soluble
iron to increase, peak and then gradually decrease.

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The region of solubility for Fe+2 becomes larger as the temperature decreases. In the RWCU
system the temperature of the water is reduced before it reaches the filter demineralizers.
Additionally, the ECP affects the form of iron at low temperatures. The ECP is also a function
of temperature. For platinum with HWC the Nernst Equation is (4):
1/2 +
E = E0-2.303(RT/nF)[log((PH2) )/H )]
or,
E = -(1.985E-4)(T(ºK))[pH+1/2log([H2]/(1000KH2))]
Where: E = Electrode Voltage
E0 = Standard Potential (zero for hydrogen)
2.303 R/F = Constant = 1.985E-4
n = Number of electrons
log[H+] = Log of hydrogen activity (expressed as pH)
PH2 = Partial pressure of hydrogen
[H2] = Dissolved hydrogen concentration (ppb) in excess of stoichiometric O2
KH2 = Hydrogen solubility coefficient (ppm/Atm)
pH = Assumed to be neutral

The ECP at the filter demineralizers will be different than that measured in the reactor coolant
system. Table 6-6 shows calculated results for ECP, using the Nernst Equation, for selected
temperatures, pH and hydrogen concentrations.
Table 6-6
Calculated Platinum ECP as a Function of Temperature, pH and Hydrogen Concentration

Temperature, ºF pH [H2], ppb ECP, V (SHE)

550 5.67 200 -0.543

550 5.67 20 -0.488

100 6.77 200 -0.362

100 6.77 20 -0.331

Assumed operating conditions for the filter demineralizers are temperature of 25ºC, pH in the
range of 6 to 8 and ECP in the range of –0.500 to +0.500 V (SHE). The Pourbaix diagram,
Figure 6-46, indicates that the region of Fe+2 solubility increases as the potential becomes more
negative.

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As stated earlier, filter demineralizers “A”, “C” and “D” were in service at high effluent
conductivity during NMCA or cleanup. Noble metals were available for incorporation into the
oxide layer on the surface of the filter demineralizers and downstream piping. When hydrogen is
injected it reduces the ECP at the oxide surface. The lower ECP yields conditions favorable for
solubilizing iron. Iron oxides on the F/D septa downstream of the precoat and on the
downstream piping are released increasing the F/D effluent conductivity. Since the “B” F/D had
the least exposure during NMCA and cleanup, there is insufficient noble metal present to reduce
the ECP, and the effluent conductivity from soluble iron is not affected.
A bounding estimate gives 6.24 lb/year of effluent iron per F/D. This corresponds to ~0.037 lb
Fe /ft2 of septa surface area or ~0.14 lb Fe /ft2 of vessel effluent chamber carbon steel surface.
Based on three F/Ds with this bounding case effluent iron, this source corresponds to
approximately 10.7% of the feedwater iron.

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Figure 6-46
Pourbaix Diagram For Fe-H2O at 25ºC (5)

Noble metals were not detected in a sample taken from a component in a F/D effluent. The
station has raised the effluent conductivity alarm point.

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Similar Experience at a Japanese BWR

A Japanese BWR applied noble metals during their 2001 outage. In January 2002 they began to
inject hydrogen, initially at a low rate to achieve a feedwater concentration of 0.05 ppm. In July
2002, the injection rate was increased to achieve a feedwater concentration of 0.1 ppm hydrogen.
In early August 2002, the plant observed that the RWCU effluent conductivity and iron
concentration exceeded those of the influent. Reported iron results were 4 ppb Fe in the RWCU
influent and 10 ppb Fe in the RWCU effluent. Another Japanese BWR that applied noble metals
in 2001 did not see this response. It was reported that a review of materials of construction was
performed. Both stations had filter elements and sample lines constructed of stainless steel.
However, the piping material in the plant that experienced the elevated RWCU effluent
conductivity and iron concentrations was carbon steel while the other plant’s piping was stainless
steel. The conclusion of the evaluation of this issue in Japan was that the source of high RWCU
effluent iron concentration was iron from the carbon steel piping going into solution in response
to the change to low dissolved oxygen concentrations after NMCA along with significant
hydrogen injection. (6)
Note that both Nine Mile Point 2 and the Japanese plant have carbon steel reactor water cleanup
system piping. The effluent compartment of the RWCU filter demineralizer at Nine Mile Point 2
is also constructed of carbon steel with no coating. Elevated corrosion rates of carbon steel occur
as the oxygen concentration decreases below about 25 ppb. With hydrogen injection, in the
presence of noble metals downstream of the F/Ds, the oxygen concentration at the surface of the
piping would be expected to be well below 25 ppb, resulting in elevated corrosion of the carbon
steel piping. Over a long enough period of time this may ultimately result in pipe wall thinning.

References

1. “EPRI BWR Water Chemistry Guidelines – 2000 Revision,” TR-103515-R2, Final Report,
February 2000.

2. “BWRVIP-92: NMCA Experience Report and Application Guidelines,” TR-1003022, Final

Report, September 2001.

3. Becker, Mark, Faivus, Mike and Giannelli, Joseph, “Elevated RWCU F/D Effluent
Conductivity After NMCA”, EPRI Workshop on Condensate Polishing, February 6-8, 2002

4. “Corrosion Potential (ECP) Measurement Sourcebook”, Final Report, January 1991

5. “Computer-Calculated Potential pH Diagrams to 300 C Degrees, Volume 2: Handbook of

Diagrams”, NP-3137-V2, June 1983

6. Personal Correspondence, Kuniie Oikawa (Organo Corporation) to Joseph F. Giannelli

(Finetech), 12 August 2002

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7
PLANT CHEMISTRY RESULTS VS. BWR WATER
CHEMISTRY GUIDELINES REVISIONS

Feedwater Dissolved Oxygen Control

One of the parameters affecting the rate of FAC (Flow Accelerated Corrosion) in the condensate
and feedwater system is dissolved oxygen. The oxygen concentration affects the form and
solubility of the iron oxide layer that forms on carbon steel piping. Data show that the rate of
FAC increases dramatically at dissolved oxygen concentrations less than approximately 25 ppb
(1). It follows that iron contributed from FAC of condensate and feedwater system materials can
be controlled by maintaining sufficient dissolved oxygen.

The EPRI BWR Water Chemistry Guidelines – 2000 Revision (1) addressed the effect of oxygen
on FAC by increasing the Action Level 1 minimum value for condensate and feedwater
dissolved oxygen concentration at power operating conditions (>10% power) from <15 ppb (2)
to <30 ppb. The Action Level 1 maximum dissolved oxygen value remained at >200 ppb.

Plant data show that the average and median concentrations of feedwater dissolved oxygen at
power operating conditions for the industry has progressively increased since 1997, as shown in
Figure 7-1. The percentage of plants with annual average dissolved oxygen concentrations
greater than 30 ppb has increased from 72% in 1997 to 97% in 2001, as shown in Figure 7-2.
The average final feedwater dissolved oxygen is higher for plants with forward-pumped drains
than cascaded drains, which is as expected since the dissolved oxygen concentration in the
forward-pumped drains stream is normally high. However, the average is skewed by the high
dissolved oxygen of a couple of forward-pumped drain plants.

The 2001 annual average feedwater dissolved oxygen at >10% power for each of the 32 plants
reporting sufficient data is presented in Figure 7-3. All plants had feedwater dissolved oxygen
averages below the >200 ppb maximum value. There was only one BWR with its annual average
lower than the <30 ppb action level value, compared to four BWRs with annual averages lower
than 30 ppb in 2000.

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Plant Chemistry Results vs. BWR Water Chemistry Guidelines Revisions

50

45

40

35
Feedwater DO (ppb)

30

25

20

15

10

0
1997 1998 1999 2000 2001

Median (All) Average (All) Average (Cascaded) Average (FPD)

Figure 7-1
BWR Feedwater Dissolved Oxygen Results at Power Operating Conditions

120

100 96.9
88.6
% of Plants with FFW DO >30 ppb

85.7
80.0
80 72.4

60

40

20

0
1997 1998 1999 2000 2001

Figure 7-2
Feedwater Dissolved Oxygen Results versus <30 ppb Action Level 1

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Plant Chemistry Results vs. BWR Water Chemistry Guidelines Revisions

90

80

70
Feedwater DO (ppb)

60

50

40

30

20

10

0
0 5 10 15 20 25 30 35
Rank

Cascaded FPD Action Level 1

Figure 7-3
2001 Feedwater Dissolved Oxygen Results at Power Operating Conditions

Implementation of Revised Action Levels for HWC and HWC/NMCA

In the EPRI BWR Water Chemistry Guidelines – 2000 Revision, Table 4-5b was added for
Reactor Water Action Level values applicable to plants with HWC or HWC+NMCA at power
operating conditions (1). Table 4-5b allows higher values associated with Action Level 2 and
Action Level 3 for chloride and sulfate when components are protected to –230 mV (SHE) by
HWC or HWC + NMCA. Under these chemistry regimes, an Action Level 2 value of 50 ppb
and an Action Level 3 value of 200 ppb may be allowable for both chloride and sulfate. NWC
values for chloride and sulfate are 20 ppb for Action Level 2 and 100 ppb for Action Level 3.

The North American BWRs were surveyed to determine if these new action levels have been
adopted and how plants assured that components were protected to –230 mV (SHE). To date,
twenty-six (26) of the thirty-one (31) units injecting hydrogen have responded. The results
presented in Figure 7-4 show that most of the plants have implemented the increased action
levels. One plant on HWC without NMCA that has adopted increased values reported that a
plant-specific evaluation showed that a component was not protected to –230 mV (SHE) at the
action levels allowed in Table 4-5b; this plant was able to justify Action Level 2 values above
the NWC maximums but below the HWC maximums. One of the HWC+NMCA plants reported
a plant-specific evaluation of a limiting component as the basis for allowing the increased Action
Levels.

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Plant Chemistry Results vs. BWR Water Chemistry Guidelines Revisions

14 13

12

10
Number of BWRs

6 5
4 4
4

0
Have Implemented New Action Levels Have Not Implemented New Action Levels

HWC HWC + NMCA

Figure 7-4
Implementation of Revised Action Levels for HWC and HWC+NMCA

Monitoring to Assure IGSCC Protection

Plants were also surveyed to determine the monitoring approach used to assure that IGSCC
mitigation (-230mV SHE) conditions are achieved and maintained. The results from the 26
responding plants are summarized in Table 7-1.

Twenty-two (22) of the 26 responding plants, regardless of whether the increased action levels
for reactor water chloride and sulfate were adopted, use secondary parameters either as the sole
measurements determining mitigating conditions or as means of confirming mitigating
conditions as indicated by ECP measurements. The most common secondary parameter, used by
16 of the 26 responding plants, is reactor coolant dissolved oxygen. The sample point for
dissolved oxygen measurements is either the reactor water cleanup (RWCU) influent or a reactor
recirculation line. The second most common secondary parameter used to confirm mitigation is
H2/O2 molar ratio, used by 12 of the responding plants. The reactor coolant dissolved oxygen
control limits and H2/O2 molar ratio control limits for the responding plants are listed in Table
7-2.

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Plant Chemistry Results vs. BWR Water Chemistry Guidelines Revisions

Table 7-1
Monitoring of Mitigation Conditions

Mitigation Method(s)
Secondary Validation Method for
Plant Secondary
ECP/Location Parameters Used Secondary Parameters
Parameters

Rx water DH, H2/O2

Radiolysis modeling with
Browns Ferry 2 & 3 No Yes ratio, NM coupon
no benchmark data
loading
MSLRM, H2 flow, Plant specific ECP data
Brunswick 1 & 2 No Yes FW DH, main to match sister plant
steam O2 radiolysis modeling
Plant specific ECP data
Rx water DO, H2/O2
Dresden 2 & 3 Yes/RWCU Yes with plant specific
ratio, FW DH
radiolysis modeling

Yes/CAV
Vessel, Recirc
Duane Arnold No
Pipe Flange, In
Core Probes

Fermi 2 No Yes Rx water DO

MSLRM, Rx water Radiolysis modeling with
FitzPatrick No Yes
DO, H2/O2 ratio sister plant data
Plant specific ECP data
MLSRM, H2/O2
Hatch 1 Yes/BHDL Yes with plant specific
Ratio, H2 flow
radiolysis modeling

Plant specific ECP data

Yes/BHDL, MLSRM, H2/O2
Hatch 2 Yes with plant specific
NMMS, LPRM Ratio, H2 flow
radiolysis modeling

Hope Creek No Yes Rx water DO

Rx water DO, H2/O2 Radiolysis modeling with
LaSalle 1 & 2 Yes/RWCU Yes
ratio, FW DH no benchmark data
Yes/NMMS
Limerick 1 & 2 Yes Rx water DO Plant specific ECP data
skid off RWCU
Plant specific ECP data
H2 flow, Rx water
Monticello No Yes with plant specific
DO
radiolysis modeling

Yes/Recirc
Loop Flange,
Nine Mile Point 1 No
GE Monitoring
Loop off RWCU

Rx water DO, Rx
Radiolysis modeling with
Nine Mile Point 2 No Yes water DH, H2/O2
no benchmark data
ratio

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Plant Chemistry Results vs. BWR Water Chemistry Guidelines Revisions

Table 7-1 (continued)

Monitoring of Mitigation Conditions

Mitigation Method(s)
Secondary Validation Method for
Plant Secondary
ECP/Location Parameters Used Secondary Parameters
Parameters

Rx water DO, H2 Radiolysis modeling with

Oyster Creek No Yes
flow no benchmark data

Rx water DO, Rx
Pilgrim Yes/Recirc Yes water DH, MSLRM, Plant specific ECP data
H2 flow
Plant specific ECP data
Rx water DO, H2/O2
Quad Cities 1 & 2 Yes/RWCU Yes with plant specific
ratio, FW DH
radiolysis modeling
Rx water DO, Radiolysis modeling with
River Bend No Yes
MSLRM, FW DH no benchmark data

Yes/LPRM
Susquehanna 1 & 2 String Lower No
Vessel Head

Table 7-2
Reactor Coolant Dissolved Oxygen and H2/O2 Molar Ratio Control Limits

Dissolved Oxygen H2/O2 Molar Ratio (NMCA Plants)

Number of Number of
Limit (ppb) Limit
Plants Plants

10 1 >4 1

<5 1 >3 3

<2 4 >2 6

<1 9
Calculated from
2
model results
None detected 1

A summary of the other secondary parameters in use and the number of plants using each is
found in Table 7-3. Most plants use multiple parameters, as noted in Table 7-1.

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Table 7-3
Other Secondary Parameters Used to Confirm Mitigation

Number of
Secondary Parameter Plants

Main Steam Line Radiation 7

Feedwater Dissolved Hydrogen Concentration 9

Hydrogen Injection Rate 6

Reactor Water Dissolved Hydrogen 4

Noble Metal Coupon Loading 2

Nineteen plants reported that radiolysis modeling was performed, and 12 of these reported that
the modeling was benchmarked using plant-specific data or data from a sister plant. Seven
plants reported performing modeling with no benchmarked data. A BWRVIP benchmarking task
to validate the existing EPRI radiolysis model code is in progress.

“Startup/Hot Standby” versus “Power Operations” Action Levels

Eleven of 12 responding plants report that the typical duration of “Startup/Hot Standby”
condition is 24 hours or less. Four plants state that Action Levels are changed to “Power
Operations” values after 24 hours; seven plants report that Action Level values are not changed.
Of the four plants that change Action Levels, two plants state that meeting conductivity values
for power operation might be difficult.

Startup and Shutdown Iron Control

In the EPRI BWR Water Chemistry Guidelines – 2000 Revision, insoluble iron was added as a
diagnostic parameter for both feedwater and reactor water during the Startup/Hot Standby
condition. Specifically, it is suggested that feedwater iron be <100 ppb prior to initiation of
significant feedwater flow to the reactor or at the completion of the feedwater flush. In practice,
the feedwater flush sample analyses are used for this purpose. These cold shutdown or early
startup results are not normally included in the plant chemistry databases, so BWRs were
surveyed to assess how the new guidelines were addressed and the sampling results obtained.
Responses were received from 16 plants; every plant did not provide a complete response to
every question.

Feedwater Flush Practices

Eleven of fourteen plants perform a feedwater flush during every refueling outage. Three other
plants use chemistry parameters to determine whether or not a flush is required. Six out of
eleven plants responded that flushes would not be cancelled due to schedule constraints, but two

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Plant Chemistry Results vs. BWR Water Chemistry Guidelines Revisions

of those six said that the flush may be shortened. There is wide variability in the duration of
feedwater flushes, from <4 hours to approximately 1 week, as shown in Figure 7-5.

4
4

3 3 3
Number of BWRs

2
2

0
< 4 hours 4-8 hours 8-12 hours 24-48 hours 48 hours to 1
week
Duration of Feedwater Flush

Figure 7-5
Feedwater Flush Duration

Nine of fourteen plants perform only a long path flush, while five of fourteen perform a
combination of long and short path flushes. Ten out of thirteen plants specifically reported that
each feedwater heater train is flushed separately.

Twelve out of fifteen plants use chemistry results as the feedwater flush termination criterion.
One uses time only and two use a combination of time and chemistry results. The chemistry
analyses and sample locations are given in Tables 7-4 and 7-5, respectively. Note that most
plants perform more than one analysis at more than one location.

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Table 7-4
Feedwater Flush Chemistry Analyses

Number of Plants
Analysis (14 plants responding)

Conductivity 9

Iron by Color Comparison 8

Turbidity 5

TOC 2

Dissolved Oxygen 2

Total (post-UV) Anions 1

Anions 1

Filterable Solids 1

Visual Crud 1

Metals 1

Table 7-5
Sample Locations During Feedwater Flush

Number of Plants
Location (14 plants responding)

Individual or Common Condensate Polisher Effluent 8

Feedwater 6

CDI or Hotwell 5

LP Heater Effluent 1

5th Heater Effluent 2

Reactor Feed Pump Suction 2

6th Heater Effluent 1

Drywell Sump 1

Heater Drains (FPD Plants) 1

Reactor Water 1

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Plant Chemistry Results vs. BWR Water Chemistry Guidelines Revisions

Impact of New Startup/Shutdown Iron Control Guidelines

Twelve out of thirteen plants responded that sampling reactor water insoluble iron prior to
initiating significant feedwater flow to the reactor is a new task; three of the twelve do not intend
to follow this guideline.

Four out of thirteen plants responded that sampling of feedwater/condensate insoluble iron at this
condition is a new task. Two (both with forward pumped drains) of those four state that the 100
ppb target is not achievable, and the other two have no data with which to assess whether the
target value is achievable. One FPD plant says that the 100 ppb target is achievable prior to
pumping forward.

Shutdown Duration for Application of New Startup/Shutdown Iron Control

Guidelines

One plant of six will implement feedwater flush and sampling only after a refueling outage. (In
the past, most of this plant’s non-refuel outages have lasted only a few days.) Five out of nine
plants responded that there is a minimum time after which feedwater flush and sampling will be
performed: two after a three- to four-day outage; two after an outage of four weeks; and one
after an outage of greater than two months.

Startup/Shutdown Copper Monitoring

Three out of four responding plants with a significant condenser copper source will monitor
copper as a diagnostic parameter for reactor water and feedwater. The fourth plant states that
copper has not been a problem in the past.

Startup/Shutdown Iron Analysis Methods

Feedwater/condensate insoluble iron analysis methods used to meet the new startup/shutdown
iron control guidelines are summarized in Figure 7-6. Eleven (11) of 11 plants analyze by color
comparison, and most of those plants also analyze the filter disk by XRF.

Startup/Shutdown Feedwater Flush Limits

Ten of fourteen responding plants set target values for insoluble iron during feedwater flush.
One plant uses total iron; one uses total suspended solids; and two plants use turbidity as
feedwater flush criteria as shown in Table 7-6.

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Table 7-6
Feedwater Flush Criteria Target Values

Parameter Target Value Number of Plants

100 ppb 6

Insoluble Iron 10 ppb 2

Unknown 2

Total Iron 100 ppb 1

Turbidity 0.2 NTU 2

Total Suspended Solids 15 ppb 1

5 5
5

4
Number of BWRs

1
1

0
Color Comparison Color Comp + XRF Color Comp + ICP
Analysis Method

Figure 7-6
Startup/Shutdown Iron Analysis Method

Deep Bed Only Plant Practices

Two plants with only deep beds for condensate polishing report using the oldest bed(s) for
feedwater flush. One removes the bed from use prior to startup and the other prior to reaching
150 psig reactor pressure. Both have had radwaste processing problems in the past related to

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Plant Chemistry Results vs. BWR Water Chemistry Guidelines Revisions

resin cleaning waste liquids after feedwater flushes. Radwaste problems are due to the high iron
loading during the feedwater flush and, in some cases, changes in the form of iron that reduces
the effectiveness and/or throughput of radwaste liquid processing equipment.

Filter + Deep Bed Plant Practices

Two responding plants with Filter + Deep Bed condensate polishing backwash the filters prior to
the feedwater flush and usually remove filters from service due to high differential pressure
during the flush. Neither plant reported experiencing radwaste problems related in particular to
feedwater flushes.

Filter/Demineralizer Plant Practices

Seven of nine responding plants with filter/demineralizers use 3 F/Ds for flushing; one uses two
and one uses two or three. Three plants select vessels with the oldest septa for use; two base the
decision on differential pressure; and two use septa that have previously been used in outage
service. One out of nine plants use a different precoat for optimum iron removal, and one uses
an anion overlay for flushing operations. Two of nine plants usually have to remove F/Ds from
service due to high differential pressure during flushes; three of nine sometimes experience high
differential pressure. Five out of nine plants backwash and precoat prior to use during or after
startup; the other four sometimes backwash and precoat based on differential pressure
experienced during flushing. None of the nine plants reported experiencing radwaste-processing
problems attributable to feedwater flushes.

Startup/Shutdown Summary

Ten out of thirteen plants (77%) will sample reactor water insoluble iron as recommended. Nine
out of thirteen plants already sample feedwater/condensate iron prior to initiation of significant
feedwater flow and four others will add this as a new task. Only two plants (Laguna Verde 1
and 2), which have a high iron source in the high pressure heater drains, had concerns about
meeting the 100 ppb target.

Assuming that these thirteen plants are representative of all BWRs, it appears that a significant
number of plants have implemented or are prepared to implement the use of reactor water and
feedwater insoluble iron as a diagnostic parameter during Startup/Hot Standby conditions.

References

1. “EPRI BWR Water Chemistry Guidelines – 2000 Revision,” TR-103515-R2, Final Report,
February 2000.

2. “EPRI BWR Water Chemistry Guidelines – 1996 Revision,” TR-103515-R1, December

1996.

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8
BWR CHEMISTRY MONITORING PRACTICES

Sampling Frequencies

The EPRI BWR database was reviewed to compare the sampling frequencies recommended in
the EPRI BWR Water Chemistry Guidelines – 2000 Revision (1) to actual plant practice. The
most recent data available for each plant were examined to determine current sampling practice.
The results of this review were the basis for the EPRI BWR Sampling Frequency Report (2),
which was initially issued in June 2002. The report was reissued in July 2002 with plant updates
included. A summary of the report’s findings is shown in Table 8-1.
Table 8-1
Chemistry Parameter Sampling Frequency

Guidelines Percent Meeting or

Parameter
Recommendation Exceeding Recommendation

Reactor Water Chloride &

Daily 75
Sulfate

Reactor Water Total Co-60

(and other total gamma Weekly 100
isotopics)

Reactor Water Soluble &

Insoluble Co-60 (and other Twice/Month 94
sol/insol gamma isotopics)

Reactor Water Iron Monthly 92

86% weekly or more

Reactor Water Zinc As needed
frequently

Feedwater Metals Weekly - Integrated 100

Reactor Water Chloride and Sulfate

Reactor water chloride and sulfate sampling frequencies are presented in Figure 8-1. The EPRI
BWR Water Chemistry Guidelines – 2000 Revision recommends daily sampling, but allows the
sampling frequency to be relaxed if the plant can use conductivity or chemistry trends to ensure
that the Action Level 1 values are not exceeded. Twenty-seven (27) out of 36 plants meet the

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recommended daily frequency (with one plant sampling twice a day). All remaining plants
sample at least 3 times per week.

Reactor Water Cobalt-60 & Zinc-65

The sampling frequencies for total Co-60 and Zn-65 are shown in Figure 8-2. Thirty-six (36)
units meet the Guidelines recommendation of weekly sampling, with 13 units sampling more
frequently. The same sampling frequency also applies to the “total” values for other gamma
isotopic parameters.

The EPRI BWR Water Chemistry Guidelines – 2000 Revision recommends twice/month
sampling and analysis of soluble and insoluble Co-60 and Zn-65 fractions in reactor water.
Other gamma isotopic parameters are measured at the same time. Thirty-four (34) of the 36
BWRs meet or exceed this recommendation. Two plants sample monthly. These results are
summarized in Figure 8-3.

2.8%
13.9%
Twice/Day
3x/Week
5.6%
2x/Week
5.6%
Every 2 Days

72.2%
Daily

Figure 8-1
Reactor Water Chloride & Sulfate Sampling Frequency

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2.8% 5.6%
Daily Every 2 Days

13.9%
Twice/Week

13.9%
63.9% 3x/Week
Weekly

Figure 8-2
Total Co-60 & Zn-65 Sampling Frequency

5.6%
5.6% 13.9%
Monthly
Twice/Month Twice/Week

13.9%
3x/Week

61.1%
Weekly

Figure 8-3
Soluble & Insoluble Co-60 & Zn-65 Sampling Frequency

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Reactor Water Iron

Reactor water iron sampling frequencies are shown in Figure 8-4. Thirty-three (33) units meet
the Guidelines recommendation of monthly sampling, with 16 units sampling more frequently.
Two (2) units sample quarterly and one (1) unit does not sample.

2.8%
5.6%
Not Sampled 19.4%
Quarterly
Twice/Week

5.6%
3x/Week

47.2%
Monthly 19.4%
Weekly

Figure 8-4
Reactor Water Iron Sampling Frequency

Reactor Water Zinc

Thirty-two BWRs are currently injecting DZO. Two units have a significant “natural” source of
zinc in plant materials of construction as the only zinc source. Reactor water zinc sampling
frequencies are summarized in Figure 8-5. The EPRI BWR Water Chemistry Guidelines – 2000
Revision recommends that reactor water zinc should be sampled as needed to support zinc
injection programs. The majority of plants with zinc injection sample for reactor water zinc at
least weekly, with only 3 units sampling less frequently. All plants sample at least monthly,
except for 2 plants that have no set frequency for zinc analysis; these plants do not inject zinc
and have no zinc source in plant materials of construction. Note that the category in Figure 8-5
labeled “1-5/wk” applies to two units where the reactor water zinc sampling frequency can vary
from once per week to five times per week.

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6%
3% No Set Frequency
Monthly 6% 6%
6% Daily 1-5x/Week
Twice/Month 22%
Twice/Week

27%
Daily
24%
3x/Week

Figure 8-5
Reactor Water Zinc Sampling Frequency

Feedwater Iron and Copper

Feedwater iron and copper sampling frequencies are summarized in Figure 8-6. Most plants
analyze all metals samples for both iron and copper. All thirty-six plants monitor feedwater iron
via at least weekly integrated samples as recommended by the Guidelines. The same is true for
copper with the exception of one plant, which has no significant copper source and analyzes for
iron daily and copper monthly. The most prevalent practice (78%) is to change out the filter discs
in the corrosion products sampler three times per week, providing three separate integrated
samples during each week at power operating conditions.

Feedwater Zinc

There is no EPRI-recommended sampling frequency for feedwater zinc. For the 32 of 36 BWRs
adding zinc (DZO), this parameter must be sampled as required to meet the plant’s feedwater
zinc target concentration and to ensure that the fuel warranty limit is not exceeded. All plants,
except two which do not add DZO and have no specified frequency for zinc analysis, are
reporting feedwater zinc results at least weekly with the predominant frequency being three
times per week, as indicated in Figure 8-7.

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3%
19% Daily
Weekly

78%
3x/Week

Figure 8-6
Feedwater Iron Sampling Frequency

6%
11% No Set Frequency
Weekly

83%
3x/Week

Figure 8-7
Feedwater Zinc Sampling Frequency

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BWR Chemistry Monitoring Practices

Zinc Sampling and Analysis Methodologies

During the development of the BWRVIP-92 document (3), the considerable variability in the
measured feedwater and reactor water zinc concentrations was noted. The plants’ sampling and
analysis methodologies were surveyed to help develop a better understanding of reasons for this
variability.

All thirty-six BWRs were surveyed, and thirty-two responses were received. Responses were not
provided by all plants to all questions. Table 8-2 identifies the reactor water and feedwater zinc
analysis method and injection method for each plant. Reactor water and feedwater zinc data at
power greater than 10% were used to determine variability in soluble zinc concentrations for
2001. (For FitzPatrick, 2000 data were used for reactor water because of sampling problems in
2001.) Variability was calculated by dividing the annual standard deviation by the annual
average concentration.

Injection Method

Among the plants responding to the survey, 5 plants use an active zinc addition system and a
passive zinc addition system is used in 23 plants. Two (2) plants (Nine Mile Point 1 and
Vermont Yankee) have a natural zinc source as the result of condenser materials of construction.
Feedwater zinc variability and reactor zinc variability by plant and zinc injection method are
shown in Figures 8-8 and 8-9, respectively. The zinc variabilites for both active and passive
addition systems are distributed over the entire range of values. The variability is at the higher
end of the range for the plants where the zinc source is condenser materials of construction.

Analysis Method

Feedwater zinc is analyzed using ion chromatography (IC) at 2 plants, ion coupled plasma (ICP)
at 16 plants, x-ray fluorescence (XRF) at 12 plants and atomic absorption (AA) at 3 plants.
Reactor water zinc analysis is performed using IC at 10 plants, ICP at 19 plants XRF at 3 plants,
and AA at one plant. (River Bend uses either XRF or AA for both feedwater and reactor water
zinc analysis.) Feedwater zinc variability and reactor zinc variability by plant and zinc analysis
method are shown in Figures 8-10 and 8-11, respectively. The feedwater zinc variabilites for
ICP and XRF are distributed over the entire range of values, while AA variability (2 plants) is in
the lower half of the range and IC variability (2 plants) is at the upper end of the range. The
reactor water zinc variability for IC and ICP is distributed over the range of values, while XRF
variability (2 plants) is at the upper end of the range.

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Table 8-2
Soluble Zinc Analysis Method & Injection Method

Feedwater Zinc Reactor Water Zinc Zinc Injection Method

Plant
Analysis Method Analysis Method

Browns Ferry 2 & 3 ICP ICP Passive

Brunswick 1 & 2 ICP ICP Passive
Clinton XRF IC Passive
Columbia ICP ICP Passive
Cooper ICP ICP Passive
Dresden 2 & 3 IC IC Passive
Duane Arnold ICP IC Passive
Fermi 2 ICP ICP Passive
Grand Gulf XRF ICP Passive
Hatch 1 & 2 XRF IC Passive
Hope Creek ICP ICP Active
FitzPatrick ICP ICP Passive
LaSalle 1 & 2 XRF IC Passive
Limerick 1 & 2 ICP ICP Active
Monticello ICP ICP Active
Nine Mile Point 1 XRF XRF Condenser (2)
Nine Mile Point 2 XRF XRF Passive
Oyster Creek XRF ICP Passive
Peach Bottom 2 XRF ICP Active (1)
Peach Bottom 3 XRF ICP Passive
Perry ICP ICP Passive
Pilgrim ICP ICP Passive
Quad Cities 1 & 2 AA IC Passive
River Bend XRF or AA XRF or AA Passive
Vermont Yankee ICP ICP Condenser (2)
Table 8-2 Notes
1. Will change to passive system in late 2002.
2. No zinc injection; condenser materials provide zinc source.

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1.2

1.0

0.8
Variability

0.6

0.4

0.2

0.0
PER

DNPS
PIL

NMP2

NMP1
ENF

JAF
MON

HCNS

CGS
DUA

CLI

DNPS
HAT1

HAT2

LAS2

BRF2
BRF3
PB3
LIM1

PB2

LIM2
VY
QUA1

BRU1
BRU2
QUA2

LAS1

RIB
OYC

Active Passive No DZO

Figure 8-8
Feedwater Zinc Variability vs. Injection Method

1.6

1.4

1.2

1.0
Variability

0.8

0.6

0.4

0.2

0.0
PER

DNPS2

DNPS3

NMP1
ENF

JAF
MON

CGS

HCNS
DUA

CLI

LIM1
BRF3

LIM2
PB2
LAS2

HAT2
BRF2

PB3

NMP2
PIL
LAS1
VY
QUA1
BRU1

BRU2

HAT1

RIB
OYC

QUA2

Active Passive No DZO

Figure 8-9
Reactor Water Zinc Variability vs. Injection Method

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1.2

1.0

0.8
Variability

0.6

0.4

0.2

0.0
PER

DNPS
PIL

NMP2

NMP1
ENF

JAF
HCNS

CGS

CLI
MON

DNPS
DUA
HAT1

HAT2

LAS2

BRF2
BRF3
PB3
LIM1

PB2

LIM2
VY
QUA1

QUA2
BRU1
BRU2

LAS1

RIB
OYC

IC ICP XRF AA

Figure 8-10
Feedwater Zinc Variability vs. Analysis Method

1.6

1.4

1.2

1.0
Variability

0.8

0.6

0.4

0.2

0.0
PER

NMP1
DNPS2

DNPS3
ENF

JAF
CGS

HCNS
MON

CLI

LIM1
BRF3

LIM2
HAT1

PB2
LAS2

BRF2
HAT2

PB3

NMP2

LAS1
PIL
OYC
DUA

VY
RIB
QUA1
BRU1

BRU2

QUA2

IC ICP XRF XRF or AA

Figure 8-11
Reactor Water Zinc Variability vs. Analysis Method

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BWR Chemistry Monitoring Practices

Variability of both reactor water and feedwater soluble zinc, distinguished by analysis type, is
plotted in Figure 8-12. It can be seen that, for most plants, the reactor water variability remains
in the range of 0.2 – 0.4, while the feedwater variability ranges between 0.2 and 0.6. Even
among stations using the same analysis method for both reactor water and feedwater zinc (e.g.,
Dresden, Limerick, and Browns Ferry), there is a significantly higher variation in feedwater zinc
than in reactor water zinc. As shown earlier, these variations do not seem to be related to either
zinc injection method or analysis method. The differences in variability may be related to actual
process fluctuations or other factors such as specific sampling and laboratory techniques (e.g.,
number of cation papers used, sample line velocity, digestion process), as well as to the fact that
reactor water concentrations are higher and easier to quantify.

1.4

1.2

1.0
Variabiliy

0.8

0.6

0.4

0.2

0.0
PIL
ENF

JAF
CLI
HCNS

CGS
PER

DUA

VY
HAT1

HAT2

LAS2

DNPS2

DNPS3
RIB
OYC

MON

QUA1

BRU1
BRU2
QUA2

LAS1

BRF2
BRF3
PB3
LIM1

NMP1
LIM2
NMP2
PB2
FW - IC FW - ICP FW - XRF
FW - AA FW - XRF or AA RW - IC
RW - ICP RW - XRF RW - XRF or AA

Figure 8-12
Reactor Water and Feedwater Zinc Variability by Plant Distinguished by Analysis Method

Other Sampling Issues

Reactor Water Gamma Isotopics

An apparent low bias in reactor water gamma isotopic results was discovered by FitzPatrick in
May 2001, after the station learned of a low bias at a PWR station, where a reactor coolant
sample was filtered through a 0.45 micron membrane filter using a laboratory filtration
apparatus. Gamma spectroscopy of the raw sample, the membrane filter, and the filtrate showed
that the sum of the (filter + filtrate) activity was significantly lower than that of the raw sample.
At FitzPatrick, reactor coolant is normally sampled every other day and filtered through a

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BWR Chemistry Monitoring Practices

membrane filter, followed by counting the membrane filter and filtrate. These are the data
normally reported to the EPRI BWR database. A monthly analysis is also performed using a
membrane filter and cation exchange filter stack to measure filterable and non-filterable species.

Testing at FitzPatrick showed that the sum of the (filter + filtrate) activity was significantly
lower than that of the raw sample. Further testing showed that species passing through the 0.45
micron membrane filter were apparently removed or held up on the fritted glass support that was
being used in the filtration apparatus. After changing from the fritted glass support to a stainless
steel mesh support the sum of the (filter + filtrate) results were in close agreement with the
results for the raw (unfiltered) coolant sample. FitzPatrick now only uses the stainless steel mesh
support in the laboratory filtration apparatus. Plants using the filter and filtrate method of
analyzing for filterable and non-filterable activity should confirm that the filter apparatus is not
biasing the results. The use of the membrane filter + ion exchange filter stack approach is
preferred.

Feedwater Sampling

Feedwater sample system design data were compiled for North American BWRs. Data received
from 22 plants are presented in Table 8-3. The sample tubing shown in Table 8-3 is the main
tubing run from the sample point to the sample station. Tubing sizes typically change at the
sample tap (e.g., there may be an isokinetic probe in the process piping) and at the sample
station. The sample flow rate is the flow through the corrosion products sampler, and it varies
from 100 – 550 cc/min. The bypass flow is used to increase the velocity in the sample line.

The results show a wide variation in main sample tubing length. The length varies from 25-50
feet to greater than 400 feet. The sample tubing length is a function of plant design and
equipment layout and is therefore not a controllable operating variable.

The liquid velocity in the sample tubing was calculated from the measured volumetric flow rate.
The sample and bypass flow rates are measured after the feedwater sample is cooled.
Consequently, the actual velocity at the temperature of the feedwater sample can be roughly 15
to 20% greater than the values shown in Table 8-3.

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Table 8-3
Feedwater Sample Tubing

Sample Tubing (Main)

Sample Bypass Velocity in Main
Flow Rate Flow Rate Sample Tubing
Length Diameter
Plant Wall (in) (cc/min) (cc/min) (ft/sec)
(ft) (in)

Browns Ferry 2 335 0.25 0.049 100-150 800-1000 4.2 - 5.4

Browns Ferry 3 335 0.25 0.049 100-150 800-1000 4.2 - 5.4

Columbia 318 0.5 0.065 150 1000 0.91

FitzPatrick >400 0.375 0.065 100 600 1.30

Grand Gulf 25-50 0.25 0.0625 350 200-250 3.8-4.1

Hatch 1 >100 0.375 0.065 100-200 700-800 1.44 - 1.80

Hatch 2 >100 0.375 0.065 100-200 700-800 1.44 - 1.80

Hope Creek 200 0.375 0.065 500 1500 3.6

Limerick 1 >300 0.375 Note 1 100-200 800 1.6 - 1.8

Limerick 2 >300 0.5 Note 1 100-200 800 0.71 - 0.79

Monticello 142 0.375 0.065 1200 2100 5.93

Nine Mile Point 1 ~50-75 0.375 0.0625 100-250 700-1000 1.44 - 2.25

Nine Mile Point 2 219 0.25 0.035 100-150 800-1000 3.0 - 3.8

Oyster Creek ~50 0.25 0.035 100 650 2.5

Peach Bottom 2 200-300 0.375 0.049 450-550 1000-1500 2.0 - 2.9

Peach Bottom 3 200-300 0.375 0.049 450-550 1000-1500 2.0 - 2.9

Pilgrim 210 0.25 0.035 100 1200 4.3

Quad Cities 1 160 0.375 0.1125 150 850 4.8

Quad Cities 2 250 0.375 0.1125 150 850 4.8

Susquehanna 1 385 0.25 0.054 400 800 6.4

Susquehanna 2 385 0.25 0.054 400 800 6.4

Vermont Yankee 0.25 0.049 100 500 2.8

Table 8-3 Notes

1. Assume 0.065” wall thickness.

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ASTM D 5540-94a (4) specifies flow control and temperature control for on-line water sampling
and analysis. This reference states, “the loss of ionic and particulate components is minimized
by maintaining the liquid velocity at 1.8 M/sec in the sample tubing transporting the sample.
The turbulent flow at 1.8 M/sec (6 ft/sec) presents a stable condition for the deposition and
removal.” The basis for this recommendation is research sponsored by EPRI (4 - 10), which
shows that deposition reaches a minimum at a fluid velocity of approximately 6 ft/sec, and that
deposit weights increase exponentially at lower velocities. Linear velocity rather than Reynolds
Number controls the deposition of particles in sample lines. Particle deposition is also minimized
at velocities above 20 ft/sec, but a velocity this high would not be practical due to pressure drop
and sample waste liquid generation limitations. A constant 6 ft/sec water velocity is necessary
for representative sampling of insoluble particulates to prevent excess deposition in the sample
line and of dissolved species in high purity water (conductivity <1 µS/cm) to prevent excessive
response delays (3 – 10 hours) by process chemistry monitors.

Therefore, to assure a representative final feedwater sample, which for most plants contains
insoluble iron oxide as the major impurity, a sample velocity of 6 ft/sec is recommended. As
indicated in Table 8-3, the main feedwater sample line velocities vary from <1 ft/sec to 6.4
ft/sec. Only two plants achieve the recommended sample velocity of 6 ft/sec. Figure 8-13
displays the range of sampling velocities and the cumulative percentage of plants at or below the
corresponding sample line velocity range for the 22 plants that responded to the survey.

6 120

5 100

4 80
Number

Percent
3 60

2 40

1 20

0 0
0-1 1-2 2-3 3-4 4-5 5-6 6-7
Velocity (ft/sec)

Number in Range % less than or equal to range

Figure 8-13
Feedwater Sample Line Velocity – Number of Plants in Range and Cumulative Percentage
of Plants At or Below Range

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Plots of feedwater soluble zinc variability and feedwater insoluble iron variability versus main
sample line velocity are shown in Figures 8-14 and 8-15 respectively. Variability was calculated
by dividing the annual standard deviation by the annual average for each parameter. For soluble
zinc, the correlation between variability and sample velocity was not apparent at velocities less
than 2.8 ft/sec. The insoluble iron correlation, although weaker, seems to apply to the entire
range of velocities. For insoluble iron, the four plants with the highest insoluble feedwater iron
variability were not included in the analysis. One of these four was Susquehanna 1, which
injects iron oxalate into the feedwater.

0.9

0.8 Velocity greater than

2.8 ft/sec
0.7
Feedwater Zns Variability

0.6

0.5
-0.3445x
y = 1.8225e
0.4 2
R = 0.5914
0.3

0.2

0.1

0
2.0 2.5 3.0 3.5 4.0 4.5 5.0 5.5 6.0 6.5
Main Sample Line Velocity (ft/sec)

Figure 8-14
Feedwater Soluble Zinc Variability vs. Main Sample Line Velocity

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0.9

0.8 Data points for the 4

plants with the highest
0.7 variability were not
Feedwater Fei Variability

used.
0.6

0.5

0.4

0.3
y = -0.1225Ln(x) + 0.6137
2
0.2 R = 0.256

0.1

0
0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0
Main Sample Line Velocity (ft/sec)

Figure 8-15
Feedwater Insoluble Iron Variability vs. Main Sample Line Velocity

For plants with low velocity, the main options for increasing the sample velocity are to decrease
the sample tubing inside diameter (an expensive modification), to increase the bypass flow
and/or perhaps maintain a diversion flow (e.g., grab sample flow after the primary sample cooler
to the sink). A flow increase may be limited by the available sample cooling capacity; the
temperature of the sample at the sample station may be limited by the sampling apparatus design
or personnel safety concerns. Increasing sample flow to achieve the recommended velocity
could require increasing the sample cooling capacity.

A feedwater sample flow assessment was performed for Oyster Creek under EPRI BWR
Chemistry Monitoring individual plant support. The main sample tubing run from the sample
point to the feedwater sample sink is 1/4" tubing with a wall thickness of 0.035". The tubing in
the feedwater sample line segment to the corrosion products sampler is 1/16" tubing with a wall
thickness of 0.010". The total indicated sample flow at the sink for corrosion products sampling
was found to be 600 – 720 cc/min. The corresponding velocity in the main run of tubing at the
feedwater process temperature of 315 oF, from the sample point to the sink, was 2.19 – 2.63
ft/sec. The flow rate through the 1/16" tubing was regulated by procedure to 50 - 70 cc/min.
With this sample flow rate the liquid velocity in this tubing segment ranged from 3.0 - 4.2 ft/sec.

The assessment concluded that the 1/16" sample tubing is the appropriate size for the sample line
segment to the corrosion products sampler to achieve sufficient velocity to assure a
representative sample based on minimizing deposition in the sample line. To assure sample
representativeness, it would be preferable to maintain a sample flow of 100 cc/min through this
tubing to meet the 6 ft/sec velocity criterion. The station reported that the sample flow was

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BWR Chemistry Monitoring Practices

previously set at 100 cc/min, but when the practice of running each corrosion products sample
filter stack for a full week was implemented, it was not possible to maintain the flow at 100
cc/min. This was because the membrane filter became overloaded, causing the sample flow to
drop off. The basis for selecting weekly integrated sampling at that time was to limit the number
of samples that had to undergo the time consuming and labor intensive sample preparation
process for analysis by ICP.

After implementating zinc injection, feedwater metals sampling was changed back to 3 times per
week to allow timely tracking of zinc, but the sample flow rate was not changed. With 3 samples
per week, a constant sample flow rate of 100 ml/min should be sustainable over the sample
period.

The assessment also concluded that the flow through the main sample line should be set to
maintain 6 ft/sec. Based on 1/4" sample tubing with 0.035" wall and a process temperature of
315 oF, the required feedwater total sample flow is 1645 cc/min (measured after cooling at the
sink). The allowable bypass flow does not appear to be limited by the available sample cooling
capacity of the system.

References

1. “EPRI BWR Water Chemistry Guidelines – 2000 Revision,” TR-103515-R2, Final Report,
February 2000.

2. “EPRI BWR Sampling Frequency Report” (electronic report), July 2002.

3. “BWRVIP-92: NMCA Experience Report and Application Guidelines,” TR-1003022, Final

Report, September 2001.

4. ASTM D 5540-94a (Reapproved 1999)

5. Eater, Lloyd, “Make sure water-chemistry samples are representative,” Power, July 1999.

6. L.J. Bird, “Requirements for crud-sampling systems for PWR primary-coolant circuits,”
EPRI Report NP-3402-SR, March 1984.

7. B. Emory, “Crud sample-system design criteria,” EPRI Report NP-3402-SR, March 1984.

8. L.L. Sundberg, “Sampling of metallic impurities in BWRs,” EPRI Report NP-3402-SR,

March 1984.

9. “Survey of Corrosion-Product Generation, Transport, and Deposition in Light-Water Nuclear

Reactors,” EPRI Report NP-522, March 1979.

10. “Water Chemistry and Radiation Build-up at the Commonwealth Edison Company LaSalle 1
BWR,” EPRI Report NP-4823, September 1996.

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9
OVERVIEW OF FIELD EXPERIENCE WITH
CONDENSATE FILTERS

This section provides an overview of the use of iron removal filter septa in Boiling Water
Reactor (BWR) condensate applications. Iron removal filter septa are those septa capable of
achieving 1.5 ppb or less effluent iron without precoats for non-precoated septa or with thin
precoats (0.05 dry pounds/10″ filter septum length) for precoatable septa.

More detailed analysis is available to members of the EPRI BWR Condensate Filter Users
Group. The group was formed to facilitate the improvement of high efficiency iron removal
condensate filter technology through evaluation of septa performance and communication of
information concerning filter septa design, performance, modification, and operating experience.

The Evolution and Design of Condensate Filters

Pleated media filter septa were the first condensate iron removal filter septa to evolve from the
EPRI BWR Feedwater Iron Reduction Program. Under the program, iron removal filter septa
and special iron removal cation exchange resins were evaluated as alternative means of reducing
feedwater iron concentrations. Although the special resins could achieve low effluent iron
concentrations, in most cases they produced unacceptable reactor water sulfate levels and/or high
operating costs due to the need for frequent resin replacement. Therefore, iron removal filter
septa currently are the means most often employed to reduce feedwater iron levels to the range of
0.5 to 1.5 ppb.

Since non-precoated iron removal septa were able to provide effluent iron concentrations less
than 1.5 ppb, many users of precoated yarn septa have adopted the use of iron removal septa with
reduced precoat dosages. As a consequence, iron removal filter septa are now used in both non-
precoat and precoat applications. Initially, only pleated filter septa were available from US
manufacturers for enhanced iron removal.

Pleated filter media include two basic types: upright pleats (offered by Graver Technologies and
Pall Corporation) and fold-over pleats (offered by Pall Corporation). In both designs, the flow
path is radial, from outside to inside. As the name implies, upright-pleated septa have the pleats
standing upright around an inner support core as shown in Figure 9-1. In fold-over septa shown
in Figure 9-2, the filter medium is folded back and forth in layers upon itself in the annular space
between an outer protective layer and a center core.

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Figure 9-1
Upright Pleat Filter Septum – View of Expanded Pleats

Figure 9-2
Fold-over Pleat Filter Septum

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Overview of Field Experience with Condensate Filters

Precoated and non-precoated iron removal septa have evolved along different paths. From the
start, non-precoated septa have been pleated septa, in either upright or fold-over pleat
configurations. Only the material of construction has changed. Polysulfone, polyaramid,
polyethylene, and polyester were tried before settling on the presently used polypropylene.

Graver Technologies and Pall Corporation both offer non-precoated iron removal septa. Both
Graver and Pall HydroGuard™ septa are upright-pleated, while Pall BPF-4 and BPF-5 septa are
comprised of fold-over pleats. The newer Pall BPF-5 design differs from the older BPF-4 in
effluent side support layer material, which Pall believes is less susceptible to plugging. The Pall
HydroGuardTM septa were formerly provided by US Filter-Filterite Power Generation (FPG).

The evolution of precoated septa is more involved. The evolution of the material of construction
was the same as for non-precoated septa. However, because of ion exchange performance
deficiencies and occurrence of media splitting, several alternative configurations have evolved.

Graver Technologies introduced pleats surrounded by wound yarn or a layer of melt-blown

polypropylene sheet to improve ion exchange performance with a more even precoat layer, while
maintaining low effluent iron concentrations. Graver DualGuard™ filters have polypropylene
pleats surrounded by a polypropylene mesh support layer covered with wound yarn that serves as
the precoat retention media. DualGuard™ II filters have a melt-blown polypropylene precoat
retention layer over polypropylene pleats; the protective cage is on the outside of the precoat
retention layer.

To address the problem of media splitting, FPG laminated a different type of support layer to the
underside of the pleats of its HydroGuard™ septa. Prior to this, the support layer was not
attached to the media, and was a simple square open mesh. The new layer is a melt-blown
gauze-like material. Thus far, the new downstream support layer has eliminated or appreciably
reduced the resin bleed-through which results from filter media splitting. Pall HydroGuard™
iron removal septa currently used in both precoat and non-precoat BWR condensate filter
applications are all upright-pleated polypropylene septa. There are slight differences between
the non-precoat and precoatable septa. The pleat height on the precoatable septa is slightly
shorter than on the non-precoat septa to accommodate precoat materials in the annular volume
between the surrounding open protective cage and the pleated media.

Pall's answer to the ion exchange performance deficiencies and media splitting occurrences was
to offer cylindrical melt-blown polypropylene septa with enhanced iron removal capability.
These septa have been free of media splitting problems, and ion exchange performance has been
superior to that obtained from precoats on pleated media. Melt-blown septa with modified
surface chemistry for improved iron removal have been used at both Peach Bottom and Quad
Cities. However, plant experience has demonstrated that acceptable iron removal with these
septa requires a precoat dosage exceeding the 0.05 dry pounds/10″ septum length stated in the
definition of iron removal septa. For this reason, the Pall melt-blown septa are no longer
included in the category of iron removal septa.

Domestic suppliers are continuing their development efforts related to BWR condensate filter
applications. Pall and Exelon have performed small-scale test work at Peach Bottom to develop

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optimal cylindrical melt-blown polypropylene septa. FPG and Browns Ferry developed surge
chamber septa which have demonstrated improved backwashability in laboratory testing. The
septa have not yet been installed at Browns Ferry, so no full scale results are available at this
time.

In addition, Organo and Toshiba hollow fiber filters (HFFs) have emerged as a practical
alternative to the non-precoat iron removal filter types mentioned above. Useful lives greater
than 10 years have been demonstrated in Japan. HFFs were included in the initial exploratory
trials at Hope Creek. Although the HFFs performed well, at the time they were more expensive
compared to pleated media septa based on 4 to 6 years anticipated lives for the pleated media
septa. More realistic projected lives of 2 to 4 years for the pleated media septa and with HFF
improvements, HFF was found to be competitive with pleated media septa in at least one recent
cost study.

Application Challenge Severity Indexes

Although low cross-linked cation exchange resins and iron removal filter septa have clearly
demonstrated benefits in reducing feedwater iron concentrations, both have potential liabilities.
The significance of these liabilities depends on plant-specific factors. The severity of the
challenge to the iron removal technologies can vary from plant to plant and can be defined in
terms of plant specific conditions. Low cross-linked cation resins present the risk of increased
sulfur release, which challenges a plant’s ability to control reactor water sulfate (in addition to
other ionic impurities). One potential liability for iron removal septa, used with or without
precoats, is a rapid decline in run lengths based on differential pressure, and consequently short
useful lives. For precoated iron removal septa, another potential liability is unsatisfactory ion
exchange performance during cooling water ingress periods.

Since both technologies for enhanced iron removal are relatively new, potential and current users
look to experience at other plants to assist in evaluating the alternative technologies. For this purpose,
the “Application Challenge Severity Index” (ACSI) concept (1, 2) was developed as a means of
ranking plants according to their vulnerability to the potential liabilities associated with the
technology of interest. So far, three different index values have been developed:
• RLI: Filter Run Length Index
• IXI: Filter Precoat Ion Exchange Index
• DSI: Deep Bed Sulfate Index

In each case, increasing index values indicate increasing challenge severity. Thus far the major
use of the Deep Bed Sulfate Index (DSI) has been for identifying Deep Bed Only condensate
polishing plants that are potential candidates for condensate filter retrofits.

Because of variations in physical and operating characteristics of the filters that are not included
in the indexes, simple correlations of life or run length to ACSI are not always possible.
However, the indexes may be useful in identifying plants with similar values in order to evaluate,
by inference, the effects of physical and operating characteristics on filter performance.

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RLI (Filter Run Length Index)

The RLI provides an indication of the impact of plant specific conditions on filter run lengths
and useful lives, and is defined as follows:

RLI = 100 * (Ff / (NS * LS))2 * Fei

where

Ff = Flow per Condensate Filter (gpm)

NS = Number of Filter Septa per Vessel
LS = Septum Length (inches)
Fei = Inlet Insoluble Iron (ppb)

The basis for the RLI is that the pressure drop across a filter septum is a function of the quantity
of filter cake deposited and the flow rate through the filter septum. For constant rate filtration,
the rate of cake deposition is directly proportional to flow rate. For laminar flow, pressure drop
is also directly proportional to flow rate. Therefore, under these conditions, run time to a
specified pressure drop is proportional to the flow rate squared. An implicit assumption in the
form of the equation is that the flow through the filter media and filter cake is laminar.

It is important to recognize that the index does not take into account the characteristics of the
filter septa, the cleaning method and frequency, the use of precoats and/or precoat relaxation, run
termination criteria or temperature. All of these factors may affect the performance of the filters.
In addition, with compressible cakes, the pressure drop across the filter may become a power
function of flow rate as pressure drop increases.

IXI (Filter Precoat Ion Exchange Index)

The IXI is applicable to precoats applied directly to upright pleats. It provides an indication of
the potential impact of unsatisfactory ion exchange performance on maintaining reactor water
sulfate concentrations at or below 5 ppb.

IXI = 2288.1 * At / (N * LS)/(ln[50 * (Fr /Ax - 1)]-2.34)

where
At = Cooling Water Total Anions (meq/l)
Ax = Cooling Water Sulfate (ppm)
Fr = RWCU Flow (gpm)
N = Number of Condensate Filter Septa In Service
LS = Filter Septum Length (inches)

The numerical value of the IXI is the precoat dosage (dry pounds precoat material/10″ of septum
length) required to prevent reactor water sulfate from exceeding 5 ppb during a 30 day filter run
length with a 0.1 gpm condenser leak rate. It is noted that the required precoat dosage (IXI) is
proportional to the required run length and is a non-linear function of the cooling water ingress rate.

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A number of assumptions are inherent in the form of the IXI:

• All of the Condensate Filter Demineralizers (CF/Ds) use upright-pleated filter septa
• The total condensate flow is treated via filter/demineralizers only
• RWCU precoat sulfate removal efficiency = 100%
• Non-bypassed CF/D precoat anion removal efficiency = 95% @ CDI conductivity less than
0.1 µS/cm
• The CF/Ds operate as a unit (begin and end runs together)
• CF/D precoat material consists of all resin with cation/anion (dry wt.) = 4/5
• Anion resin ultimate capacity = 3.74 meq/dry gram

It should also be noted that the IXI algorithm is based on limited test data for a single type of
septum with upright pleats.

DSI (Deep Bed Sulfate Index)

This index was developed to assist in identifying plants that would benefit more from using
prefilters rather than crud resins to reduce feedwater iron concentrations. The major
disadvantage of crud resins is their production of organosulfur compounds, which contribute to
reactor water sulfates. Thus, plants with an already high reactor water sulfate potential would be
better candidates for prefilters than those with a lower potential.

In physical terms, the DSI provides an indication of the steady state reactor water sulfate
concentration (ppb) resulting from the release of 0.033 mg/hr of sulfur (equivalent to 0.10 mg/hr
of sulfate) per cubic foot of condensate demineralizer resin (cation and anion). For calculation
purposes, RWCU sulfate removal efficiency is assumed to be 100%. The DSI is defined as
follows:

DSI = 0.44 * (Nb x V)/Fr

where

Nb = Total Number of Online Condensate Demineralizer Beds

V = Total Resin Volume per Bed (ft3)
Fr = Normal RWCU Flow (gpm)

Organosulfur production rate is proportional to cation resin volume, therefore the total
condensate demineralizer resin volume is found in the numerator. DSI does not account for
differences due to cation and anion resin characteristics, cation/anion ratio, cleaning method and
frequency, cooling water ingress, or temperature. Recently, it has become apparent that the
combination of cooling water ingress and bed cleaning frequency has a strong influence on
reactor water sulfate concentrations. In addition, other plant specific factors such as deep bed
system hardware, operating practices, resin types and replacement policies, and heater drain

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Overview of Field Experience with Condensate Filters

routing make it difficult to draw comparisons from DSI. As a result, DSI is no longer perceived
as a sole accurate identifier of plants likely to benefit from condensate filter retrofits. That is,
plants with low DSI may have undesirable high average reactor water sulfate concentrations that
could be lowered by the reduced frequency or complete elimination of resin cleaning operations
made possible by the addition of prefilters.

Growth of Iron Removal Filter Septa Use

The initial use of pleated filter septa in a domestic BWR condensate non-precoat application was
at Perry in August 1991, using Graver’s polyester upright-pleated septa that have since been
replaced with other iron removal septa. Pleated filter septa were initially used in a domestic
BWR precoat application at Hatch in January 1995, using Pall polyaramid fold-over pleated
septa that have also been replaced with other iron removal septa. The dramatic increase in the
use of iron removal septa in BWR condensate polishing systems is illustrated in Figure 9-3. The
increase in non-precoated applications is a function of the conversion of Deep Bed Only plants to
Filter + Deep Bed plants. The initial early increase in iron removal septa usage in precoat
applications has leveled off as plants have balanced the advantages of iron removal septa (good
iron removal at lower precoat dosage) versus the disadvantages (vulnerability to resin leakage,
diminished run times, shorter life, higher replacement cost).

Quad Cities and Peach Bottom temporarily decreased the fraction of filters using pleated
precoated septa because of media splitting concerns. Pall HydroGuard™ septa with laminated
supports have thus far lessened splitting concerns, which has been demonstrated by Monticello
and Browns Ferry using pleated septa in all vessels. However, recent analysis of septa age and
sulfate data has indicated a potential relationship between pleated septa age and increased reactor
water sulfate at Filter Demineralizer plants, with the effect being greater for those plants with a
higher fraction of pleated septa. In addition, several plants have experienced increased effluent
iron as pleated septa age increases. Both of these phenomena suggest the possibility of age-
related mechanical failure, although other age-related effects may be possible. Further
investigation into these effects will be conducted under the auspices of the BWR Condensate
Filter Users Group.

Pleated septa have been and remain the exclusive iron reduction media in non-precoat
applications. The septa were either retrofits to existing vessels or supplied as original equipment
in new filter vessels added to condensate polishing systems for the specific purpose of reducing
feedwater iron. Of the 84 vessels with iron removal septa in non-precoat applications, 52 are
new vessels.

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160

Number of Vessels with Iron Removal Septa

140

120

84 84
100
65 66
80 52
37
60

40
16
60 62
51 54 54
20 45
27

0
1995 1996 1997 1998 1999 2000 2001 2002

Precoat Non-Precoat

Figure 9-3
Trend In Iron Removal Septa Use

For the precoat applications, the majority of septa are also pleated, with Columbia and Vermont
Yankee the only stations with CF/Ds using no pleated septa. Current distribution of precoat filter
septa types is shown in Figure 9-4.

All of the precoat applications of iron removal septa have been in vessels originally designed for
cylindrical non-pleated septa. For vessels with top tubesheets, new tubesheets were required
since the pleated septa have larger diameters than the cylindrical septa they replaced. New
tubesheets were not required for bottom tubesheet vessels.

The type and characteristics of the iron removal filter septa used thus far in BWR condensate
applications are listed in Table 9-1. Several of these have been discontinued by the manufacturer. All
of the septa listed use pleated filter media, except for the Pall non-pleated meltblown septa.

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36% Yarn Wound

52% Pleated

7%
Pleated/Meltblown
5% Meltblown

Figure 9-4
Current Distribution of Precoat Septa Types (% of Vessels)

Table 9-1
Filter Septa Evaluated In BWR Condensate Polishing Applications

Rating Manufacturing
Mfg. Type Media Material Application Initial Use
(µm) Status

Precoat Retain
Falban Polyolefin 4-5 Precoat Jan-98 Discontinued
Layer + Pleats

Graver Upright Pleats Polyester 0.6 Non-Precoat Aug-91 Continued

Graver Upright Pleats Polyester 1 Non-Precoat Dec-95 Continued

Graver Upright Pleats Polypropylene 3 Non-precoat Jun-01 Continued

Wound Yarn + Polypropylene +
Graver 0.6 Precoat Apr-97 Discontinued
Pleats Polyester
Wound Yarn +
Graver Polypropylene 1 Precoat Apr-97 Discontinued
Pleats
Wound Yarn +
Graver Polypropylene 5 Precoat Apr-98 Discontinued
Pleats
Precoat Retain
Graver Polypropylene 10 Precoat May-99 Continued
Layer + Pleats
Precoat Retain
Graver Polypropylene 1 Precoat Sep-00 Continued
Layer + Pleats
Pall Fold-Over Pleats Polyaramid 1.4 Non-Precoat Oct-94 Discontinued

Pall Fold-Over Pleats Polypropylene 1 Non-Precoat Dec-95 Continued

Fold-Over Pleats
Pall New Type Support Polypropylene 1 Non-Precoat Jul-99 Continued
Layer

Pall Fold-Over Pleats Polyaramid 1.4 Precoat Jan-95 Discontinued

Pall Fold-Over Pleats Polypropylene 1 Precoat Jan-97 Discontinued

Melt-Blown
Pall Non-Pleated 5 Precoat Jun-97 Continued
Polypropylene

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Table 9-1 (continued)

Filter Septa Evaluated In BWR Condensate Polishing Applications

Rating Manufacturing
Mfg. Type Media Material Application Initial Use
(µm) Status

Melt-Blown
Pall Non-Pleated Polypropylene 5 Precoat 2001 Continued
(improved design)

FPG Upright Pleats Polysulfone 0.5 Non-Precoat Sep-95 Discontinued

FPG Upright Pleats Polypropylene 2 Non-Precoat May-95 Continued

FPG Upright Pleats Polypropylene 4 Non-Precoat May-95 Continued

FPG Upright Pleats Polypropylene 10 Non-Precoat Jan-00 Continued

FPG Upright Pleats Polypropylene 2 Precoat May-95 Continued

FPG Upright Pleats Polypropylene 4 Precoat May-95 Continued

FPG Upright Pleats Polypropylene 10 Precoat Mar-96 Continued

Polypropylene
FPG Upright Pleats Laminated Support 1 Precoat Apr-00 Continued
Layer

Polypropylene
FPG Upright Pleats Laminated Support 10 Precoat May-99 Continued
Layer

Polypropylene
FPG Upright Pleats Laminated Support 18 Precoat Jan-00 Continued
Layer

Upright Pleats
Double Media
FPG Polypropylene 4 Precoat Sep-98 Discontinued
Layers, Bottom 2
Cartridges

Precoat Applications

A list of the domestic BWR plants using precoated iron removal septa is provided in Table 9-2.
Most plants use all powdered ion exchange resin precoat materials on their pleated septa.
Resin/fiber mixtures are currently used at Duane Arnold, Hatch, Peach Bottom, and occasionally
at Quad Cities. The all-resin precoat materials are premixed formulations, except at Monticello,
where the cation and anion resins are mixed in the precoat tank prior to application.

All of the plants listed in Table 9-2, except for Duane Arnold, use Pall HydroGuard™ iron
removal septa in one or more condensate filter vessels. Except for Fermi 2 and Duane Arnold,
the Pall HydroGuard™ septa are installed in bottom tubesheet vessels. Quad Cities has installed
an eighth vessel in Unit 2 and is in the process of installing an eighth vessel in Unit 1 as part of a
planned power uprate.

The relative performance of the Pall HydroGuard™ septa among the plants with bottom
tubesheet vessels has generally been consistent with the RLI concept. RLI and IXI values for
BWR precoat septa applications are shown in Table 9-3.

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Table 9-2
BWR Precoated Iron Removal Septa

Number
Flow per
of Precoat
Current Rating Vessel Type Flow Tubesheet
Plant Septa Type Vessels Mat’l 0
Septa (µm) @60 F Of BW Balancing Location
With Type
(gpm)
Septa

Browns Ferry Polypropylene 10 3 of 9

Pall (1) All Resin 3200 Surge No Bottom
2 Pleated 18 6 of 9

10 4 of 9
Browns Ferry Polypropylene
Pall (1) All Resin 3200 Surge No Bottom
3 Pleated 18 5 of 9

Polypropylene Steady
Cooper Pall (1) 10 3 of 7 All Resin 2718 No Bottom
Pleated State

Melt-blown
Graver
Polypropylene+ Fiber/ Air
Duane Arnold DualGuard™ 5 1 of 5 2800 No Top
Polypropylene Resin Bump
II
Pleats

Polypropylene Air
Fermi 2 Pall (1) 4 2 of 8 All Resin 2423 No Top
Pleated Bump

Polypropylene
Pall (1) 10 2 of 7
Pleated

Melt-blown
Graver Fiber/
Hatch 1 Polypropylene + 3486 Surge Yes Bottom
DualGuard™ 5 1 of 7 Resin
Polypropylene
II
Pleats

Pall APF-2 Non-pleated 5 1 of 7

Polypropylene
Pall (1) 1 1 of 7
Pleated

Polypropylene
Pall (1) 5 1 of 7
Pleated

Polypropylene Fiber/
Hatch 2 Pall (1) 10 1 of 7 3700 Surge Yes Bottom
Pleated Resin

Melt-blown
Graver
Polypropylene +
DualGuard™ 5 1 of 7
Polypropylene
II
Pleats
Polypropylene
Monticello Pall (1) 10 5 of 5 All Resin 2911 Surge Yes Bottom
Pleated
Peach Bottom Polypropylene
Pall (1) 10 8 of 10 All Resin 2880 Surge Yes Bottom
2 Pleated
Polypropylene
Pall (1) 10 8 of 10 All Resin
Peach Bottom Pleated
2880 Surge Yes Bottom
3 Fiber/
Pall IFR-1 (2) Non-pleated (3) 5 1 of 10
Resin

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Table 9-2 (continued)

BWR Precoated Iron Removal Septa

Number
Flow per
of Precoat
Current Rating Vessel Type Flow Tubesheet
Plant Septa Type Vessels Mat’l 0
Septa (µm) @60 F Of BW Balancing Location
With Type
(gpm)
Septa

Polypropylene
4 2 of 7
Pleated
Pall (1)
Polypropylene
10 1 of 7
Pleated

Pall
Quad Cities 1 Non-pleated 5 1 of 7 All resin 2800 Surge Yes (Manual) Bottom
APF-2

Pall IFR-1
Non-pleated 5 1 of 7
(2)
Graver Melt-blown
DualGuard™ Polypropylene + 1 1 of 7
II Polypropylene
Polypropylene
Pall (1) 4 1 of 8
Pleated

Polypropylene
Pall (1) 10 1 of 8
Pleated

Quad Cities 2 Melt-blown All resin 2800 Surge Yes (Manual) Bottom
Graver
Polypropylene +
DualGuard™ 1 3 of 8
Polypropylene
II
Pleats

Pall
Nonpleated 5 1 of 8
APF-2

Table 9-2 Notes

1. HydroGuard™, formerly by FPG.
2. Melt-blown polypropylene with modified surface chemistry.

The stations with the three highest RLI values are Quad Cities, Browns Ferry and Peach Bottom.
The highest RLI value based on 2001data is that for Peach Bottom 2. Peach Bottom 2 and 3 CDI
iron concentrations are among the highest for all North American BWRs, as illustrated in Figures 3-9
and 3-10. Average CDI iron for 2001 (based on monthly samples) was 56.2 ppb at Peach Bottom 2.
This is higher than has been seen in the past; the annual average was 30.1 ppb in 2000 and 21.5 in
1999. Three values in July and August 2001 were above 100 ppb; contributing to the high annual
average. Without these unusually high values, the annual average CDI iron for Peach Bottom 2 in
2001 was 42 ppb, about the same as that for Peach Bottom 3. The stations with the three lowest RLI
values for 2001 are Hatch 2, Hatch 1, and Duane Arnold; CDI iron concentrations at these plants are in
the low range for North American BWRs.

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Table 9-3
ACSI Values for BWR Precoated Iron Removal Septa

2001 Average Insoluble Fe 2001

Plant IXI
FFW (ppb) CDI (ppb) RLI (1)

Hatch 1 0.64 8.9 18.6 0.0073

Hatch 2 0.73 9.3 21.7 0.0076

Cooper 2 (2) 1.18 14.8 24.5 0.3013

Fermi 2 0.67 15.2 25.6 0.0263

Duane Arnold 0.75 12.2 29.6 ?

Monticello 1.08 12.7 32.8 0.0226

Quad Cities 1 1.45 16.7 39.9 0.0238

Quad Cities 2 (3) 1.12 17.0 41.3 0.0208

Browns Ferry 3 1.22 20.2 63.0 0.0056

Browns Ferry 2 1.24 22.9 71.4 0.0056

Peach Bottom 3 1.84 43.0 79.8 0.0058

Peach Bottom 2 1.27 56.6 105.0 0.0058

Table 9-3 Notes:
1. Accuracy is +/- 2 or 3 due to infrequency of CDI iron analysis.
2. FFW Fe is total; CDI Fe is 1998 average.
3. IXI reflects addition of new vessel. With planned increase in RWCU flow to 520 gpm, IXI = 0.0166.

A comparison of HydroGuard™ 10 micron septa performance in 2000 at Browns Ferry 3 and

Peach Bottom 2 is shown in Table 9-4. As the table indicates, run length is considerably shorter
and dP rise rate considerably higher at Browns Ferry, whose RLI indicates a more severe
challenge to the filter. Because of these factors, useful life of the septa is expected to be shorter
at Browns Ferry than at Peach Bottom. Browns Ferry uses precoat relaxation, which appears to
add an average of 2-3 days to each run. Thus far it is not known whether the relaxation will
ultimately have a beneficial or adverse effect on septa life.

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Table 9-4
Comparison of HydroGuard™ 10 µm Septa Performance

Browns Ferry Peach Bottom

Vessel 3C Vessel 2E

RLI (2000) 71.1 56.1

Installation Date 5/99 7/99

Days in Service (for data below) 568 574

Average Run Length (days) 20.7 41.5

Average dP Rise Rate (psi/day) 0.71 0.12

Current dP Rise Rate (psi/day) 0.79 0.20

Current Ending dP (psi) 18 8

Non-Precoat Applications

North American BWRs using iron removal septa without precoats are listed in Table 9-5. All of
the plants, except Brunswick and Limerick, have top tubesheet filter vessels. This is in direct
contrast to the situation for precoated septa where only 2 of the 12 plants have top tubesheet
filter vessels.

Additional information on North American non-precoat condensate applications of iron removal

septa is given in Table 9-6. In particular, challenge severity index values for RLI and DSI (the
potential impact on reactor water chemistry of organosulfur releases from condensate
demineralizer resin beds) are provided. The DSI values are noteworthy in being high for the
BWR plants that have added filters to their condensate polishing systems. That is, the plants that
elected to use iron removal septa rather than low cross-linked cation resins to reduce feedwater
iron appear to have made the better technical choice between the two alternative methods.

Of the iron removal septa applications listed in Table 9-6, the most recent installation is at
Dresden 2, which installed two filter vessels handling 40% of the total condensate flow in
December 2001. River Bend is in the process of installing a headered full-flow condensate
filtration system, expected to be operational by the end of 2002. The current applications cover a
wide range of RLI values, from 118 at Hope Creek to 11 at Perry. The applications encompass
septa uses in top and bottom tubesheet filter vessels.

Currently, septa from two suppliers, Graver Technologies and Pall, are used in the non-precoat
applications of iron removal septa. Both suppliers’ septa are installed in both top and bottom
tubesheet vessels.

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Table 9-5
BWR Non-Precoat Iron Removal Septa

Number
Flow per
of Slaved or
Current Rating Vessel Type Of Flow Tubesheet
Plant Septa Type Vessels 0 Headered to
Septa (µm) @60 F BW Balancing Location
With Demin Vessels
(gpm)
Septa

Polypropylene
Pall 1 2
Pleated
Pall Polypropylene Steady
Brunswick 1 4 1 3500 Headered Yes Bottom
(1) Pleated State
Polypropylene
Graver 3 1
Pleated
Polypropylene
Pall 1 4
Pleated
Pall Polypropylene Steady
Brunswick 2 4 1 3500 Headered Yes Bottom
(1) Pleated State
Polypropylene
Graver 3 1
Pleated
Pall Polypropylene
4 1 Air Bump
(1) Pleated No
Clinton 2889 Slaved Top
Polypropylene (2)
Pall 1 5 Surge
Pleated
Polypropylene
Hope Creek Pall 1 4 6995 Headered Surge No Top
Pleated
Pall Polypropylene
Dresden 2 2 3920 Headered Air Bump Top
(1) Pleated
Pall
Polypropylene 7
Laguna Verde 1 (1) or (4) 1760 Slaved Air Bump No Top
Pleated (3)
Graver
Pall
Polypropylene 7
Laguna Verde 2 (1) or (4) 1760 Slaved Air Bump No Top
Pleated (3)
Graver
Pall Polypropylene
LaSalle 1 4 7 2850 Slaved Air Bump No Top
(1) Pleated
Pall Polypropylene
LaSalle 2 4 7 2850 Slaved Air Bump No Top
(1) Pleated
Polypropylene
Limerick 1 Pall 1 8 3673 Headered Surge Yes (manual) Bottom
Pleated
Polypropylene
Limerick 2 Pall 1 8 3673 Headered Surge Yes (manual) Bottom
Pleated
Pall Polypropylene
Perry 10 8 2495 Headered Air Bump Yes Top
(1) Pleated
Pall Polypropylene
Susquehanna 1 10 6 4990 Headered Air Bump Yes Top
(1) Pleated
Pall Polypropylene
Susquehanna 2 4 6 4990 Headered Air Bump Yes Top
(1) Pleated

Table 9-5 Notes

1. HydroGuard™, formerly by FPG.
2. Partial filter installation. Six filter vessels are currently slaved to six demineralizers. Design flow is based on 1/9 of total
condensate flow. Flow balancing is not used, but flows are controlled for each septa type.
3. 6 vessels normally in service.
4. Pall septa are 4 micron; Graver septa are 1 micron.

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Table 9-6
ACSI Values for BWR Non-Precoated Iron Removal Applications

2001 Average Insoluble Fe (ppb)

Plant RLI DSI
FFW CDI

Brunswick 1 0.39 14.6 20.7 1.68-2.16

Brunswick 2 0.35 10.7 15.2 1.68-2.16

69.1 (2)
Clinton 1.34 18.5 (1) 2.64-2.97
104.3 (3)

Dresden 2 2.57 24.1 45.7 0.83

Hope Creek 0.70 22.7 118.2 3.35-3.66

Laguna Verde 1 1.60 22.1 63.8 3.17

Laguna Verde 2 1.62 17.8 51.4 3.17

LaSalle 1 1.82 14.4 (4) 39.4 2.01

LaSalle 2 1.49 16.5 (5) 45.2 1.98

Limerick 1 0.57 12.1 57.8 3.19

Limerick 2 0.80 14.4 68.8 3.19

Perry 0.78 12.0 11.0 2.23

Susquehanna 1
1.11 15.7 60.8 2.94
(1)

Susquehanna 2
0.89 17.0 65.9 2.94
(1)

Table 9-6 Notes

1. CDI Fe is 1999 average.
2. Value for vessel with Pall HydroGuard™ septa..
3. Value for 5 vessels with Pall septa.
4. 1999 average.
5. 2000 average.

A comparison of Pall 1 µm septa is shown in Table 9-7. Although the RLI value for
Susquehanna in 2000 was lower than that of Hope Creek, septa performance at Hope Creek was
superior. It is likely that the Hope Creek septa outperform those of Susquehanna because of a
more effective backwash method at Hope Creek.

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Table 9-7
Comparison of Pall 1 µm Non-Precoated Septa After One Year of Service

Hope Creek Susquehanna 1

2000 RLI 89.0 58.5
Model BPF-4 BPF-5
Average Run Length (days) 17 (1) 21

Average dP Rise Rate (psi/day) 0.06 0.19

Last dP Rise Rate (psi/day) (2) 0.08 (649 days) 0.54 (330 days)
Table 9-7 Notes
1. Initial run lengths were 14 days, then 21 days, then back to 14 days.
2. At ~570 days age, Susquehanna 1 filter run length was much shorter than 21 days to 15 psi dP.

A similar comparison for bottom tubesheet vessels may be made between Brunswick and
Limerick. The Brunswick RLI of about 20 compared to that of about 60 for Limerick would
suggest that septa at Brunswick should last significantly longer than at Limerick. However,
experience has shown that the lives are similar; septum life at both stations is approximately 3
years. It may therefore be deduced that more effective backwashing at Limerick contributes to
improved run length despite a fairly high challenge based on RLI.

Deep Bed Demineralizers – Potential for Condensate Filter Addition

Of the twenty-two North American BWR plants with condensate deep bed demineralizers, only
the six units listed in Table 9-8 do not have or are not in the process of implementing prefilters
upstream of the deep beds. As shown, in 2001, Grand Gulf, Nine Mile Point 1, Oyster Creek and
Pilgrim had feedwater iron concentrations above the upper limit of the 0.5 to 1.5 ppb range
within which dose rates are minimized, and these plants could benefit from the improved iron
removal offered by prefilters. FitzPatrick and Nine Mile Point 2 maintained an annual average
feedwater iron of less than 1.5 ppb.

Reducing reactor water sulfate concentrations is another potential benefit of prefilters. The bed
and flow disturbances inherent to bed cleaning invariably result in releases of soluble and
insoluble sulfate-bearing substances to the reactor which produce sulfate spikes and elevate
average sulfate concentrations. Prefilters normally allow undisturbed use of demineralizer resin
beds for one or more fuel cycles. FitzPatrick, Grand Gulf, and Pilgrim all had 2000 average
reactor water sulfate concentrations exceeding 2 ppb. FitzPatrick’s average sulfate in 2002 has
been reduced to less than 2 ppb. This improvement is attributable to the use of high cross-linked
resins in the condensate demineralizers in combination with larger anion resin underlays. Grand
Gulf’s year to date sulfate average is also less than 2 ppb.

High condensate temperature is also a factor in sulfate control, since the higher temperatures
increase the release of sulfate ions and organosulfur compounds from deep bed demineralizer
cation resins. Grand Gulf has the highest condensate temperature of the six remaining Deep Bed
Only plants.

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Table 9-8
BWR Condensate Deep Bed Only Plants

2001 Avg.
Condensate
Insol. Fe
Temperature
Comm. (ppb) 2001 Avg.
Drains Potential Benefit(s)
Plant Operation Reactor Water
Path from Filter Retrofit
Start Avg Max Sulfate (ppb)
FFW CDI
(0F) (0F)

James A. 2.59 (through

Jul-75 110 125 1.43 7.52 Cascaded Lower rx sulfate
FitzPatrick September)

Forward Lower rx sulfate and

Grand Gulf Jul-85 124 136 1.65 13.5 2.03
Pumped feedwater iron

Reduce use of low

Nine Mile cross-linked resins
Dec-69 105 130 1.61 17.2 Cascaded 1.64
Point 1 (1) and lower feedwater
iron

Reduce use of low

Nine Mile Forward cross-linked resins;
Apr-88 115 120 1.41 19.3 1.94
Point 2 (2) Pumped minimize sulfate
excursions

Dec-69
Oyster Lower rx sulfate and
88 118 2.79 14.5 Cascaded 1.57
Creek feedwater iron
(3)

Lower rx sulfate and

Dec-72
feedwater iron;
Pilgrim 98 121 2.41 8.81 Cascaded 2.27
improved condensate
(4)
system performance

Table 9-8 Notes

1. Using low cross-linked resins in 4 of 6 demineralizer vessels and maximum RWCU flow (6%).
2. Sulfate excursions still occur even while using low cross-linked resins in 5 of 9 demineralizer vessels.
3. Evaluating life extension.
4. Seeking life extension

Although prefilters offer significant performance and cost benefits, a prefilter retrofit requires a
significant capital investment. For plants with limited years of licensed operation remaining, the
economics of prefilter addition may be difficult to justify.

Thus far, the oldest plant to justify a condensate filter retrofit is Dresden 2, with a commercial
operation start in June 1970. The Dresden retrofit is for partial flow filters that were needed to
maintain iron and chemistry control associated with a power uprate. Dresden 3 prefilters are
expected to begin service late in 2002. The oldest plant to justify a full-flow condensate prefilter

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retrofit is Susquehanna 2, with a commercial start in June 1983. As shown in Table 9-8, only
Nine Mile Point 1 and Oyster Creek have commercial operation start dates prior to June 1970.
Life extension is under evaluation at Oyster Creek.

Of the six plants listed in Table 9-8, only FitzPatrick, Nine Mile Point 2 and Grand Gulf are not
either committed to, or actively evaluating, a retrofit of condensate prefilters. Nine Mile Point 1
intends to add prefilters. Oyster Creek has made the decision to add prefilters based on technical
considerations and is seeking funding. Pilgrim is evaluating a condensate filter retrofit as an
alternative to other modifications to upgrade the condensate demineralizers, and may prove to be
an important contributor to life extension efforts.

In addition to the factors mentioned above, other reasons for consideration of prefilter
installation include the following: meeting CPI goals; reducing long-term plant labor
requirements; reducing the quantity of radioactive waste that must be buried or stored;
optimizing iron to meet radiation dose goals; reducing DZO costs; and reducing RWCU flow to
allow more electricity production. With only two plants currently using low cross-linked resins
(Nine Mile Point 1 and Nine Mile Point 2), the long-term availability of the resins is uncertain.
The industry trend, demonstrated by the dwindling number of Deep Bed Only plants, is toward
the Filter + Deep Bed design for previously Deep Bed Only plants.

Major Issues and Their Current Status

Full Scale Application Status

Iron removal filter septa are now in use at twenty-seven (75%) of the North American BWR
stations in precoat and non-precoat applications. Most septa currently in service have been in
place long enough for effective performance evaluation. Retrofits of condensate filters for iron
removal are in the design/procurement phase at River Bend and in the installation phase at
Dresden 3.

Pall BPF-4 septa, without precoats, are in service at Clinton, Hope Creek, and Brunswick. Non-
precoat BPF-5 septa are currently in service at Limerick.

Pall HydroGuard™ upright-pleated septa, without precoats, are in service at Brunswick, Clinton,
Dresden, LaSalle, Laguna Verde, Perry and Susquehanna. Pall HydroGuard™ precoatable
pleated septa are in service at Browns Ferry, Cooper, Fermi, Hatch, Monticello, Peach Bottom,
and Quad Cities.

Precoated Graver septa with precoat retention layers over upright pleats are in service at Duane
Arnold, Hatch, and Quad Cities.

Pall precoated melt-blown polypropylene cylindrical septa are in use at Hatch and Quad Cities,
and septa with improved surface chemistry were recently installed at Peach Bottom and Quad
Cities.

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Effluent Iron Concentrations

In the absence of mechanical failures, most of the pleated iron removal septa have achieved
satisfactory effluent iron concentrations. Obtaining satisfactory effluent concentrations with
practical run lengths while using Pall melt-blown cylindrical septa continues to be problematic
compared to pleated septa, requiring use of specific precoat materials to obtain the desired
effluent iron concentration. Low iron concentrations for non-precoated septa have been
successfully addressed with iron injection. Some precoat users have kept yarn-wound septa in
service to maintain feedwater iron above the recommended lower limit. Browns Ferry and
Monticello use pleated septa in all vessels while maintaining feedwater iron concentration
greater than 0.5 ppb. Browns Ferry is the only station using 18 micron pleated septa.

Mechanical Integrity

In the early stages of use, iron removal septa from all domestic suppliers suffered isolated
mechanical failures. These early failures occurred in joints, end fittings, protective cages, and
septum-to-tubesheet seals. Modifications to the septa or to manufacturing procedures have
eliminated these types of failures.

More recently, defects in or breeches of the filter media itself became a serious problem resulting
in the plugging of effluent resin traps and/or reactor water sulfate perturbations one to two years
after installation. Modifications to septa construction and manufacturing techniques have been
made to eliminate or appreciably reduce these premature failures. In particular, the laminated
support layer design developed by FPG has made mechanical integrity much less of a concern.

Septa Useful Life

Septa useful life, although still the number one concern, has improved with experience.
Although the 5 to 6 year lives projected in the early 90s are not being achieved, 3 to 4 year lives
have been achieved at many stations.

Demonstrated useful lives at units with moderate to high RLI values are particularly noteworthy.
Among these are the practical lives demonstrated for non-precoated Pall BPF-4 septa at Limerick
and at Hope Creek, and for precoated HydroGuard pleated septa at Peach Bottom, the incidents
of filter media splitting notwithstanding.

Among non-precoat pleated septa users, Susquehanna has the shortest useful lives
(approximately 18 months) due to relatively high inlet iron and flow rates and an ineffective
backwash method. Quad Cities' pleated septa lives are also relatively short because of high inlet
iron and flow.

Based on results at Limerick, Pall BPF-5 fold-over pleated septa seem slightly better than BPF-
4s with respect to useful life. The Pall HydroGuard™ septa with laminated effluent side support

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layer are a demonstrated improvement over earlier designs. Graver’s DualGuard™ II septa have
performed better than the original DualGuard™ septa with respect to run length and useful life.

Where useful lives have been judged impractical or unacceptable, modifications to backwash
and/or operating techniques are being investigated. Operating techniques encompass run
termination criteria, ∆P or run time, and, in the case of precoat septa, precoat material and
dosage. Results of these investigations continue to be evaluated.

Backwash Methods

Material collected on the filter during the service run is removed by backwashing the septa. As
noted above, the effectiveness of the backwash can have a major impact on filter performance
and service life of the septa. There are two basic backwash methods commonly used for
cleaning septa in BWR condensate filters: steady state and non-steady state. Graver’s “Mod III”
backwash method is the only steady state method in use in BWR applications, and is used only
with bottom tubesheet filter vessels. Three variations of the non-steady state method are in use:
“Air Surge”; “Air Bump”; and “Slam Dunk”. Backwash procedures vary from manufacturer to
manufacturer and from plant to plant. Representative examples are given below.

The “Mod III” method is currently used at two BWR stations using iron removal septa:
Brunswick and Cooper. It consists of the following timed steps:
• The vent and drain valves are opened while simultaneous backflush flows of air and water
remove the bulk of the material from the septa.
• The drain valve is then closed while air and water flows continue, resulting in the vessel
refilling.
• Backwash water and air to the vessel are secured and a drain valve is opened, allowing the
vessel to drain.
• The drain valve is closed and addition of air and water at a higher flow rate than the earlier
step refills the vessel.
• The drain valve is then opened with the air and water still being supplied to the vessel.
• After a preset time, the water to the vessel is secured.
• With air still being supplied to the vessel, the vessel is drained.
• The air and drain valves are closed and the vessel is filled with water, completing the
backwash cycle.

The non-steady state method is the most common backwash method used in BWR condensate
filter applications. The “Air Bump” and “Slam Dunk” variations are used in top tubesheet
systems, and the “Air Surge” variation is used in bottom tubesheet systems.

An example of an “Air Surge” method is shown below. The vessel is initially full of water, and is
backwashed using timed steps:

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• The vent and drain valves are opened to begin draining the vessel.
• The backwash water valve is opened to begin backwashing the septa.
• The air header to the vessel is pressurized followed by cycling the air valve to the vessel.
This short burst of air causes water in the vessel lower plenum to surge upward through the
filter media, removing material from the media. This process is repeated several times as the
level in the vessel is lowered. This is followed by more backwash surge steps as the vessel is
refilled, and then again as it is drained.
• The vessel is refilled with water completing the backwash cycle.

An example of an “Air Bump” method consists of the following steps:

• After the vessel is isolated, the vent is opened to depressurize the vessel.
• The vessel is emptied by opening the air and drain valves until a preset time or preset level is
reached, at which time these valves and the vent are closed.
• The vessel is then partially drained (or filled in repeat steps) to provide an air space above the
liquid level in the top dome above the tubesheet. The air valve is opened to pressurize the
vessel, including its top dome. The quick opening drain valve is opened, for a preset time.
The expanding air in the top dome forces water downward through the filter septa media,
removing solids from the media surface.
• The vessel is vented and then refilled. A number of air bump steps can be performed each
time a filter is backwashed; the normal number of repeats varies for different plants.
Following the completion of air bump steps the vessel is filled, completing the backwash
cycle.

The “Slam Dunk” method is currently used at only two stations, Clinton and Hope Creek. This
method is only used with top tubesheet filters and requires a large diameter air supply line
attached to the top plenum and a vent below the tubesheet.
• The vessel is filled with water above and below the tubesheet at the start of the backwash.
• The drain and vent below the tubesheet are opened, draining water from below the tubesheet.
Water above the tubesheet does not drain.
• After a preset time or preset level drop, a quick acting valve opens on the air supply line to
the top of the plenum. This pushes the water above the tubesheet through the septa, forcing
the material on the outside of the septa into the evacuated region below the tubesheet.

The basic differences between the two alternative methods with top tubesheets are important.

In the “Air Bump” method, before the bump, the entire vessel is liquid filled except for the
pressurized air dome in the top plenum. Backwash starts when the main drain valve is opened.
The volume of liquid in the top plenum determines the amount of backwash water delivered per
bump. Backwash velocity is determined by the maximum achievable drain rate. The liquid
waste volume/bump approximates the vessel volume.

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In the “Slam Dunk” method, before bump, some water is drained from below the tubesheet
(main drain and vent below tubesheet are opened). The backwash starts when the air supply
valve is opened. The entire top plenum volume determines the amount of backwash water
delivered per bump. Backwash velocity is determined by the maximum achievable air delivery
rate. The liquid waste volume/bump equals the vessel volume.

The EPRI BWR Condensate Filter Users Group has made efforts to model top tubesheet
backwashing for optimization of design and operating parameters. Some of the parameters
considered include: air and water volumes above the tubesheet; air pressure; air delivery rate;
drain valve type, size and opening speed; and drain piping diameter, length, fittings and routing.
The model is customized for the specific plant design and operating conditions. The effects of
these variables on backwash water volume and velocity can be evaluated for initial designs or for
effectiveness of proposed modifications.

Ion Exchange Performance

Although pilot plant tests clearly indicated that the performance of precoats applied to pleated
media was significantly inferior to that of precoats applied to yarn-wound septa, this has not been
a major concern except at stations with medium to high cooling water solids. The consequence
of poor ion exchange capabilities is site specific, and depends on the frequency and size of
condenser leaks, cooling water quality and the Reactor Water Cleanup flow rate. Where precoat
performance is an issue, precoat materials may be altered when condenser leaks are encountered,
or runs may be terminated early to minimize effects on reactor water chemistry. At Quad Cities,
for example, both precoat material selection and early run termination at the request of chemistry
are used to minimize the impact of condenser leaks on reactor water chemistry.

The value of precoats is evident from a recent evaluation for Browns Ferry concerning use of
upright pleated septa in all vessels. The maximum tolerable cooling water ingress rate for Action
Level 1 reactor sulfate was calculated, assuming that all vessels contain upright pleated septa
(currently only true at Browns Ferry and Monticello), and that the septa are precoated on a
uniformly staggered scheduled. The results, shown in Figure 9-5, indicate that for all plants
investigated, the maximum tolerable ingress rate is about 8 to 10 times higher with precoats than
without precoats. In general, the maximum ingress rate is higher for those plants with lower IXI.

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1.0 10
0.90
0.9
0.8 1

IXI (dry lbs/10" septum length)

Max. CW Ingress (gpm)

0.7
0.60
0.6 0.55 0.1
0.5
0.40 0.39
0.4 0.01
0.30
0.3
0.2 0.001
0.08 0.10
0.1 0.05 0.07
0.03 0.04 0.03 0.01
0.0 0.0001
Browns Peach Hatch Quad Cities Monticello Fermi 2 Cooper
Ferry Bottom

With Precoat Without Precoat IXI

Figure 9-5
Tolerable Cooling Water Ingress at Action Level 1 Reactor Water Sulfate

Handling Backwash Waste Liquids

Because of longer run lengths, and lower precoat dosages where the septa must be precoated,
backwash waste liquids have had an adverse effect on radwaste filter throughputs at some
stations. The addition of small amounts of polyelectrolytes to the backwash waste liquids has
improved liquid waste processing at many stations. This simple procedure also been
successfully applied in eliminating process problems without increasing radwaste liquid TOC
(total organic carbon) concentrations for non-precoat applications.

Peach Bottom Trial of Melt-Blown Septa

Peach Bottom 3 is testing Pall melt-blown polypropylene septa that are outgrowths of the APF-2
design with modified surface chemistry. The 5-micron septa are currently installed in two filter
vessels, and various precoats, including resin only and resin/fiber mixtures, have been tried.
During the initial runs, effluent iron did not change from run to run, implying that the septa were
cleaning better than the pleated septa currently in use at Peach Bottom. (In the case of the
pleated septa, the effluent iron improves from run to run, presumably an indication that the septa
are fouling.) Effluent iron was higher than desired, at about 3-4 ppb. Modifications to improve
effluent iron without detracting from cleaning ability are currently being investigated.

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Continuing Issues

The following issues are reported to be of major concern to users of BWR condensate filters:
• Cost
• Effluent Quality
• Ion Exchange Performance (precoat applications)
• Run Lengths
• Useful Life
• Radwaste Reduction
• Equipment Issues

With an ever-increasing number of plant life extensions being sought along with increased
challenges due to power uprates, the improvement of condensate filter performance remains an
important issue. The EPRI BWR Condensate Filter Users Group is involved in a continuing
effort to evaluate septa performance and disseminate information in order to address industry
issues and drive the optimization of this technology.

References

1. “BWR Iron Control Monitoring,” TR-108737, Interim Report, December 1998.

2. “BWR Iron Control Monitoring,” TR-109565, Final Report, September 1999.

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10
SUMMARY OF EXPERIENCE WITH DEEP BED
CONDENSATE DEMINERALIZERS

Deep Bed Only Condensate Polishing

All of the nine U.S. BWRs that operated with Deep Bed Only condensate polishing for most of
2001 have demonstrated the capability to achieve feedwater iron in the upper portion of the
desired range of 0.5 ppb to 3 ppb. However, consistent iron control within the desired range has
proven difficult, as shown by the 1998 – 2001 annual average feedwater iron concentrations for
Deep Bed Only plants in Figure 10-1. The inconsistency in iron control at Deep Bed Only plants
is attributed to their greater vulnerability to upsets and other changes in operating conditions
compared to Filter + Deep Bed or Filter Demineralizer systems.

5
Feedwater Iron, ppb

0
JAF DNPS3 NMP2 DNPS2 OYC PIL RIB NMP2 GGNS

1998 1999 2000 2001

Figure 10-1
Annual Average Feedwater Iron for 1998 – 2001 for Plants with Deep Bed Only Condensate
Polishing

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Summary of Experience with Deep Bed Condensate Demineralizers

The challenge to maintain reliable iron control along with maintaining reactor water sulfate
control and minimizing operating costs have driven several plants with only deep beds to retrofit
pre-filters. As shown in Section 3 (Figure 3-1), the 2002 projected number of North American
BWRs with Deep Bed Only condensate polishing is six, based on pre-filter retrofits that are in
operation, or in the design or installation phase.

Several factors must be optimized to realize the best achievable iron control performance with
Deep Bed Only condensate polishers. These variables are discussed below:

Resin Selection

Resin selection is one important variable influencing iron control for Deep Bed Only plants. A
listing of resins in use in these plants is given in Table 10-1. The information in Table 10-1,
along with additional details, is periodically updated and distributed to the industry in the EPRI
BWR Deep Bed Resins electronic report (1).

The 2001 average feedwater iron results, presented in Section 3 (Figure 3-3), show that
FitzPatrick and Nine Mile Point 2 achieved the lowest values of the Deep Bed Only plants. Nine
Mile Point 2 applied Dow Guardian CR-1 low cross-linked cation resin to maximize iron
removal efficiency. Nine Mile Point 2 uses optimized air scrub & backwash along with URC
(Ultrasonic Resin Cleaner) to perform periodic external cleaning of the resins. FitzPatrick,
which has low hotwell iron and applies optimized URC for resin cleaning, currently uses 16%
cross-linked cation resin in all vessels.

In addition to Nine Mile Point 2, other plants currently applying low cross-linked resins to meet
iron control objectives are Nine Mile Point 1 and River Bend. Both Nine Mile Point 1 and River
Bend have experienced improved iron control through the use of low cross-linked resins and
optimizing resin cleaning practices with existing equipment. River Bend plans to discontinue the
use of low cross-linked resins by the end of 2002, when full-flow filtration is to be implemented.

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Summary of Experience with Deep Bed Condensate Demineralizers

Table 10-1
Resins Used with Deep Bed Only Condensate Polishing

Number of
Anion
Plant Supplier Cation Resin Demineralizer
Resin
Vessels

Dow 575C SBR-C 2

Dresden 3
Rohm &
IRN-97 IRN-78 5
Haas

Rohm &
FitzPatrick IRN-99 IRN-78 8
Haas

Grand Gulf Dow HGR-W2 SBR-C 8

Guardian CR-1 SBR-C 4

Nine Mile Point 1 Dow
575C SBR-C 2

Guardian CR-1 550A 5

Nine Mile Point 2 Dow
HGR-W2 SBR-C 4

575C SBR-C 5
Oyster Creek Dow
HGR-W2 SBR-C 2

Pilgrim Dow 575C SBR-C 7

Purecat C-550 LS A-284-C

1
(1) (2)
US Filter
River Bend
(2) C-361 (3) A-284-C 8

Guardian CR-1 A-284-C 1

Table 10-1 Notes
1. Purecat C-550 LS is treated Dow 575C.
2. A-284-C is treated Dow SBR-C.
3. C-361 is treated Dow HGR-W2.

The use of low cross-linked resins has been observed to increase reactor water sulfate or impair
ion exchange kinetics based on laboratory measurements of the sulfate mass transfer coefficient
of in-service resin samples. Plants using these resins have learned that the useful life of low
cross-linked resins must be limited to approximately 6 months to 1 year to maintain sulfate
control. This short media life results in significant increases in operating costs and radwaste
burial volumes. Some of the plants have therefore applied these resins as an interim operating
strategy to control iron, accepting the higher operating costs required for control of sulfate, until
condensate filters can be retrofitted. Most plants using these resins also exclude the use of low

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Summary of Experience with Deep Bed Condensate Demineralizers

cross-linked resins in outage service because the elevated dissolved oxygen levels appear to
increase the cation resin decomposition rate, leading to sulfate control problems. Similarly, low
cross-linked resins from condensate demineralizers are normally restricted from use in radwaste
demineralizers.

With the declining number of BWR plants projected to continue operating with Deep Bed Only
condensate polishing, the U.S. market for low cross-linked resin is shrinking. Based on current
projections, only two North American BWRs will be using low cross-linked resins in some of
their condensate demineralizer vessels in 2003. Therefore, resin manufacturers have little
incentive to invest in research and development to improve the stability of low cross-linked
resins for this market. If development efforts directed at foreign markets are successful, these
may benefit users in the U.S.

Some plants in the U.S. and in Japan are evaluating the use of weak acid cation exchange resins.
These resins offer potential for improved iron removal compared to standard resins and they
have no sulfur in the functional groups; the cation exchange functionality is based on carboxylic
acid instead of sulfonic acid. One of the concerns with using weak acid resins is their limited
ability to provide acceptable control of reactor water chemistry in the event of a small chronic or
sudden large condenser leak. The suitability of weak acid resins in BWR condensate polishing is
highly dependent on the main condenser circulating water concentrations, composition and
design leak rate at which acceptable reactor water chemistry must be maintained.

FitzPatrick and Pilgrim have been successful in controlling average feedwater iron below
approximately 2 ppb without using low cross-linked resins. These plants have inlet iron
concentrations in the 6 – 9 ppb range; among the lowest of the North American BWRs. The low
inlet iron concentrations decrease the challenge to control feedwater iron with Deep Bed Only
condensate polishing. FitzPatrick has demonstrated feedwater iron control in the 1.5 - 2 ppb
range for several years. This level of control could be the result of using a relatively high
cation/anion resin ratio (recently reduced from 1.2 to 0.86 by volume) along with optimized resin
cleaning and resin management practices. Pilgrim has been able to reduce feedwater iron by
increasing the use of smaller than standard particle size cation resins.

Several other plants are using smaller than standard cation resins; indicated in Table 10-1 by the
plants using IRN-97, 575C, and IRN-99. The cation resin harmonic mean diameter ranges from
approximately 500 microns to 650 microns for these resins, which is significantly smaller than
the conventional cation resins (based on Dow HGR-W2) that have a harmonic mean diameter of
approximately 790 microns. With the smaller resin size, the additional cation resin surface area
appears to provide some benefit in iron removal capability, particularly after the resins age for 4
– 6 months. These cation resins are also often applied with anion resins having similar hydraulic
properties, thus the cation and anion components have a lower tendency to hydraulically
separate. The tendency to separate could result in a cation resin heel in the bottom of the
demineralizer vessel, contributing higher sulfate to the reactor water than if a mixture of cation
and anion resins was present.

Two resin strategies are gaining wide acceptance to improve control of reactor water sulfate. One
is a trend toward higher cross-linked cation resins. For a gel cation resin, the percent cross-

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Summary of Experience with Deep Bed Condensate Demineralizers

linkage refers to the percentage of divinylbenzene in the styrene-divinylbenzene copolymer.

Some data from resin manufacturers indicate that higher cross-linked cation resins impart less
extractable organic sulfur species to the bulk liquid. Dow was the first to offer a higher cross-
linked gel cation resin, 575C, to the U.S. market. This resin has a nominal cross-linking of 12%,
compared to 10% for conventional gels used in BWR condensate polishing. Rohm & Haas also
offers the IRN-99 16% cross-linked gel resin. Oyster Creek has experienced improved reactor
water sulfate control and improved iron control with the Dow 575C cation resin; however, other
improvements were made at the same time (as discussed below), making it difficult to separate
the factors contributing to lower sulfate. FitzPatrick has been able to maintain <2 ppb feedwater
iron while improving reactor water sulfate control in the summer months by using Rohm & Haas
IRN-99 cation resin with anion resin underlays (discussed below).

A second resin strategy that has gained rapid acceptance and is currently used by all BWR Deep
Bed Only plants is to routinely operate with a portion of the anion resin applied as an “anion
underlay” in the bottom of the condensate demineralizer vessel. When properly implemented,
this strategy assures that the last resin contacted by the condensate in the downflow
demineralizer is anion resin, which has a final opportunity to remove sulfur impurities from the
condensate prior to entering the feedwater to the reactor. The quantity of anion resin used as the
underlay has varied from as little as 20 ft3 to as much as 75 ft3. At Oyster Creek, approximately
55 to 60 ft3 of the 90 ft3 of anion resin in each bed is routinely applied as an underlay. The
improvement in reactor water sulfate between the summer of 2000, when the underlays were
used, and the summer of 1999, when no anion underlays were used, is shown in Table 10-2. The
normalized reactor water sulfate was reduced by as much as 61% in summer 2000. Note that
actual reactor water sulfate concentrations in 1999 were lower than those shown in Table 10-2
because reactor water cleanup flow was higher. The sulfate levels at Oyster Creek continued to
be lowered through the summer of 2001, as shown in Figure 10-2, during the time that the
operational consistency of the anion underlay process was further improved.
Table 10-2
Summer Reactor Water Sulfate and RWCU Flow for 1999 and 2000 (Normalized to RWCU
Flow = 400 gpm)

Normalized Normalized 2000

2000 RWCU
1999 Sulfate 1999 RWCU Sulfate
Flow (gpm)
(ppb) Flow (gpm) (ppb)

June 6.24 400 2.44 400

July 5.16 400 2.76 400

August 5.39 400 4.42 (1) 400

September 5.36 400 2.73 400

Table 10-2 Notes

1. Value for August includes transient following plant scram.

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6
Reactor Water Sulfate (ppb)

5 SCRAM
Peak summer temperature
approximately 125 F.
4 Values normalized to constant
RWCU flow rate.
3

0
June July August September

1999 - No anion underlay 2000 - First summer anion underlay

2001 - Improved underlay procedure

Figure 10-2
Oyster Creek Summer Reactor Water Sulfate for 1999 - 2001 (Normalized to RWCU Flow =
400 gpm)

Inlet Iron Concentration

The average CDI iron concentration varies from about 6 ppb to more than 20 ppb among plants
with Deep Bed Only condensate polishing. The annual average CDI iron concentrations from
1998 – 2001 for Deep Bed Only plants are shown in Figure 10-3. The higher the inlet iron
concentration, the greater the challenge to achieve and maintain acceptably low feedwater iron
concentrations. This is clearly apparent in the relatively low feedwater iron achieved at
FitzPatrick and Pilgrim where there is no use of low cross-linked resins. Improving the current
level of understanding of the plant design and operating variables that influence CDI iron and
how those variables can be practically managed to reduce CDI iron would have large economic
benefits to plants.

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Summary of Experience with Deep Bed Condensate Demineralizers

35

30

25
CDI Iron, ppb

20

15

10

0
JAF PIL OYC NMP1 RIB DNPS3 GGNS NMP2 DNPS3

1998 1999 2000 2001

Figure 10-3
1998 – 2001 Average CDI Iron for Deep Bed Only BWRs

Other issues involving CDI iron are measurement frequency and accuracy. CDI iron is typically
measured once per week using a sampling period of 1 – 3 days. Resin beds are typically cleaned
on a service time or throughput basis to approximate an iron loading basis, assuming CDI iron is
constant. However, the analysis results typically show significant variability, say from 11 to
more than 20 ppb for a plant averaging about 15 ppb. If accurate and timely results were
available, resin cleaning could be performed on a crud loading basis. XRF (x-ray fluorescence)
for iron analysis requires significantly less sample preparation time than ICP (ion coupled
plasma) or AA (atomic absorption) spectroscopy methods, and is therefore in increasing use.
With this method, it has become more practical to actually track CDI iron and take action on
resin cleaning using information based on crud loading. Dresden has implemented this method
of tracking iron loading as the basis for removing a demineralizer from service for resin cleaning.

The accuracy of the CDI iron results also appears to be more questionable than the accuracy of
feedwater integrated samples at many plants. The CDI iron corrosion products sample may not
always be flowing and this can result in sampling inaccuracies. In addition, if the sample filter
becomes overloaded, the results will be uncertain. In addition to the impact on condensate
demineralizers, it is also important to assure the accuracy of CDI or hotwell corrosion products
determinations because a significant increase in concentration could be indicative of changes in
the plant. Two examples of plant changes that have resulted in increased CDI or hotwell iron are
the start of hydrogen water chemistry and degraded performance of moisture separator reheaters.

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Summary of Experience with Deep Bed Condensate Demineralizers

Effect of Condensate Temperature

The North American BWRs, as a group, show seasonal feedwater iron trend for plants with Deep
Bed Only condensate polishing. This is shown in Section 3 (Figure 3-15). For Deep Bed Only
plants, the average feedwater iron is highest in the first and fourth quarters, corresponding to the
calendar quarters with the lowest condensate temperatures. The Filter Demineralizer and Filter +
Deep Bed condensate polishing plants show much less variation by calendar quarter than Deep
Bed Only plants.

The variation in feedwater iron control for Deep Bed Only plants is thought to be a function of
condensate temperature. Calendar quarter averages are used in Figure 3-15 because condensate
temperature data are generally not provided at this time to the EPRI BWR Chemistry Monitoring
Database. Recognizing that temperature is the important process variable helps explain why
some plants experience more seasonal variation in iron control than others. For example, the
seasonal variations are less extreme at some plants like Dresden, where once-through main
condenser cooling is used in the summer and recirculating cooling is used in the winter.

Resin Cleaning Frequency

A variety of approaches have been taken by Deep Bed Only plants to establish a resin cleaning
frequency. A rational basis, which has been successfully used as a guide by several plants, is to
establish a cleaning frequency that normally limits the crud accumulation on each bed to a level
at which the increase in flow-resistance is still small. This approach has also been found to
provide acceptable control of effluent iron under normal conditions and to limit the impact of
flow transients on feedwater iron perturbations.

When a condensate polisher bed is placed in service after cleaning or with new resin, it begins to
accumulate insoluble crud, predominantly iron oxide, which is filtered out of the feed stream. In
the early part of the service run, the mass of iron crud removed by the bed is not sufficient to
affect the resistance to flow through the bed. As the crud inventory increases, the resistance to
flow begins to increase, first at a slow rate and eventually accelerating until a steady increase rate
is attained.

If the bed is maintained in service for extended periods after the steady rate of increase is
attained, the flows among beds become unbalanced. The most recently cleaned bed will tend to
have the highest flow (lowest resistance), and the most highly loaded bed will tend to have the
lowest flow (highest resistance). The flow balance can be restored by throttling the individual
vessel valves to compensate for the pressure drop caused by crud loading, but this action
increases the differential pressure across the system and can obscure other causes of
demineralizer flow restrictions.

The majority of the crud is removed in the upper portion of the bed. As crud continues to
accumulate, portions of the crud layer may eventually crack or collapse. Such cracking or
collapse of the crud layer can result in performance problems such as flow channeling in the bed
(also affecting ion exchange) and sudden crud breakthrough to the effluent. It is therefore
important to avoid large accumulations of crud on the beds.

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Summary of Experience with Deep Bed Condensate Demineralizers

The crud accumulation on each bed is removed by cleaning the resins. A practical and
reasonable choice for cleaning frequency is to clean often enough to maintain approximately
balanced flows among the vessels in parallel service without the need to throttle the effluent
valves. This is accomplished by limiting the crud inventory to an amount that does not
significantly increase the resistance to flow.

A correlation was previously presented (2) that gives the service time to the onset of increased
resistance due to crud accumulation as a function of the inlet iron concentration. Sample results
of this correlation are shown graphically in Figure 10-4.

260
Time to Increased Crud Resistance (days)

240
220
200
180
160
140
120
100
80
60
40
20
0
5 10 15 20 25
Condensate Polisher Inlet Crud (ppb)

Plant data at 50 gpm/ft2 CORRELATION

45 gpm/ft2 40 gpm/ft2
35 gpm/ft2 30 gpm/ft2

Figure 10-4
Condensate Polisher Service Time to Flow Resistance Increase (Temperature = 95 °F)

As expected, the lower the area flow rate, the longer the service time to the crud resistance
increase point at a given CDI iron concentration. It is important to note that the curves shown in
Figure 10-4 give the points where the resistance to flow due to crud accumulation would just
begin to increase. The bed does not necessarily have to be removed from service and cleaned at
precisely that service time to maintain approximately balanced flows or acceptably low crud
inventory. At a given iron concentration, an increase in the range of about 10% to 20% in the
indicated service time normally would not cause a major increase in resistance to flow.
However, the implicit assumption in the correlation is that the average iron concentration used is
a representative value for the entire service period.

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Summary of Experience with Deep Bed Condensate Demineralizers

Flow Transients

Plants with Deep Bed Only condensate polishing generally cannot reliably maintain feedwater
iron in the desired range during and immediately following condensate flow transients.
Recovery of iron control following a flow transient may not occur until all of the resin beds have
been effectively cleaned at least one time.

Significant condensate flow transients can and do occur during conditions such as startups,
scrams, power reductions or increases, placing condensate polisher vessels in service or
removing them from service, and feedwater flow changes to maintain reactor level. Sudden flow
increases can occur even during a power reduction as condensate flow drops below the point
where condensate pump and/or condensate booster pump minimum flow valves suddenly open,
causing a surge in flow through the resin beds.

Action should be taken to minimize the severity of flow transients. This can be done by
developing an understanding of operational practices and system responses at various plant
conditions. For example, it may be possible to plan a power reduction to avoid a minimum flow
valve opening. It may also be possible to limit the power ascension rate to limit the rate of flow
increase through the Condensate Polisher beds. Controlling the rate at which vessels are valved
in and out of service can also minimize flow surges during these evolutions.

The effects of flow transients on iron transport can be minimized by maintaining a low crud
inventory on each demineralizer bed. This can be accomplished by establishing an appropriate
resin cleaning frequency using the curves given in Figure 10-2 as guidance along with
sufficiently frequent and representative sampling of the CDI iron.

Resin Cleaning Effectiveness

In addition to cleaning resins at an appropriate frequency, the process used must be effective in
removing and separating the crud and resin fines from the resins. If the external resin cleaning
process is not performing effectively, the resin transfer operations could actually exacerbate
water quality problems.

Several Deep Bed Only plants have demonstrated that feedwater iron can be reduced by
improving the effectiveness of the external resin cleaning process. Most of these plants were
originally designed with the URC process, which was first set up for use with the original resins
and cation/anion ratio in conjunction with chemical regeneration. Over the past 10 years or so,
several plants have optimized the operation of the URC to remove crud and resin fines; realizing
benefits of improved iron control and reduced sulfate spiking. Optimizing the URC performance
avoids the capital cost of installing a new resin cleaning system, but requires higher O&M costs
to maintain effective cleaning.

Improved resin cleaning processes have also been developed and implemented, including the
ARCS (Grand Gulf, Dresden) and the JRC method (Oyster Creek). These plants have all
realized the benefits of iron reduction, reduced sulfate spiking through improved resin fines

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Summary of Experience with Deep Bed Condensate Demineralizers

removal, lower pressure drop, lower liquid waste volumes, reduced operator labor, and reduced
radiation exposure.

Resin Age

Iron removal performance at Oyster Creek, River Bend, Pilgrim, Nine Mile 1 and Dresden
improves as the resins age for about 6 months. This is believed to be due to some decrosslinking
of the cation resin at the outer surface, causing it to have iron removal properties analogous to
but less pronounced than low crosslinked resins. In contrast, FitzPatrick experiences the best iron
removal from new resin beds. The reason for this difference is not known.

Pilgrim and FitzPatrick, which do not use low crosslinked resins, have typically maintained beds
in condensate service for about two years before disposal. The stresses on the resins at these
plants are low compared to most plants because maximum condensate temperatures are not
extreme, temperatures over 120 oF occur no more than four months per year, and resins need
cleaning only about every 100 days due to the low source term. The lower cleaning frequency
decreases the stresses on the resins from the transfers and cleaning process and results in less
exposure time to oxygenated water. Grand Gulf also uses standard or high crosslinked resins,
but reports resin bed life of no more than 18 months to maintain reactor water sulfate control.
Grand Gulf’s condensate temperature, the highest among North American BWRs, sometimes
approaches or exceeds 140 oF, accelerating the decomposition of the cation resin and the anion
resin.

Some Deep Bed Only plants have replaced the anion resin portion of a bed while continuing to
use the cation resin in an effort to maintain the iron removal capability while improving reactor
water anions control (mainly sulfate). Nine Mile 1 and Nine Mile 2 have done this when the
anion resins show indications of fouling. Replacing the anion resin at these plants has had the
immediate impact of restoring ion exchange performance, but the time for the new anion resin to
become fouled is typically significantly shorter than for the original anion resin. This is thought
to be due to decrosslinking of the cation resin, which allows more and higher molecular weight
organic extractables to diffuse out of the cation resin pores, causing the new anion resin to foul
faster in the presence of the aged cation resin. In another example, the anion resin was replaced
in the 2G demineralizer at Dresden 2 in 12/01 but the cation resin, which was Dow 575C, was
kept in service.

The impact of condenser leaks is another factor to consider in deciding whether to extend the
service life of the cation resin while replacing the anion resin in a Deep Bed Only demineralizer.
If the cation resin has high ionic loading, particularly sodium, from a condenser leak (10% of ion
exchange sites or higher), the equilibrium leakage from the cation resin appears to coincide with
elevated reactor water anions as well as cations. This was the case at Oyster Creek following a
salt water condenser leak. When the beds were disturbed by resin cleaning, the reactor water
sulfate increased (3). The sulfate source was higher organic (post-UV) sulfate, presumably
organo-sulfur decomposition products that were not being effectively removed by the anion resin
despite excellent measured resin kinetics.

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Cation/Anion Resin Ratio

Nine Mile 2 and FitzPatrick are the only two Deep Bed Only BWR plants currently using a C/A
(cation/anion) resin ratio approaching or equal to 1.0 by volume. This ratio provides an excess
of cation resin sites over anion resin sites based on the total resin in the bed (including the anion
underlay). These plants are. The remaining plants use a C/A volume resin ratio between about
0.56 and 0.66, indicative of an approximately equivalent mix based on total cation and anion
exchange sites.

EPRI studies have shown that insoluble iron oxide particles in BWR condensate iron are
attracted to the cation resin (4). Analysis of resin samples show that iron crud adheres tightly to
the surface of the cation resin but only loosely to anion resins, from which it can be easily
removed by rinsing or minor mechanical agitation.

The iron control benefit of an increased cation resin ratio was demonstrated in full-scale plant
applications at Susquehanna. Prior to installing pre-filters, different resin ratios of standard
resins were applied at Susquehanna, and it was found that 2/1 C/A (by volume) beds provided
better iron removal than 1/1 C/A (by volume) beds, which provided better iron removal
performance than equivalent mix beds. The increased cation resin surface area per bed with the
higher C/A appeared to be responsible for the improved iron removal. The down side of an
increased C/A ratio is that more cation resin is available to contribute ionic sulfate (from thermal
desulfonation) and organic sulfur (from resin oxidation), and this may reduce bed useful life
based on reactor water chemistry (particularly sulfate). The net benefit has to be determined by
experience. It is noted that the original resin ratio used at most BWRs designed with deep beds
and chemical regeneration was 2/1 cation/anion by volume.

As previously noted, routine anion resin underlays are now being implemented by all Deep Bed
Only plants. This is accomplished by hydraulically classifying the cation and anion resins and
transferring a portion of the anion resin to the demineralizer vessel for use as the underlay. The
main purpose of the underlays is to reduce sulfate input to reactor water by assuring that there is
anion resin at the bottom of the demineralizer bed to remove sulfur impurities from the
condensate. The anion underlay application also results in a resin layer rich in cation resin on
top of the bed. Some plants report improved iron removal when anion underlays are applied.

Vessel Area Flow Rate

The lower the demineralizer vessel area flow rate (superficial velocity) in a Deep Bed Only
plant, the lower the challenge to achieve effective iron control. Among the North American
BWRs with Deep Bed Only condensate polishing, Grand Gulf’s area flow rate of about 30
gpm/ft2 is the lowest. Other deep bed plants have area flow rates in the range of about 40 gpm/ft2
to almost 60 gpm/ft2.

Area flow rate is not a variable that can be effectively controlled with existing equipment. With
full-flow condensate polishing, the flow rate is fixed based on the full-power condensate flow
and the number and size of demineralizer vessels. The area flow rate can be temporarily lowered
by operating with all beds in service (no spare), but the plant would have to operate with fewer

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beds in service when a vessel must be removed for cleaning. The flow transients through iron-
loaded beds due to these types of manipulations must be carefully controlled because they may
disturb the crud layers and contribute to higher feedwater iron. Another consideration is that
operating with all vessels in service puts more cation resin in line with the condensate flow and
could increase reactor water sulfate. The alternative of bypassing a portion of the condensate
flow may not result in a net reduction in feedwater iron with Deep Bed Only condensate
polishing.

Filter + Deep Bed Condensate Polishing

Plants with Filter + Deep Bed condensate polishing rely on the upstream filters to remove
insoluble metal oxides and other particulate impurities from the condensate. For plants with full-
flow filtration, there is no need to consider the use of low cross-linked resins or other special
resins designed to enhance crud removal. With upstream full-flow filtration, there is normally no
need to externally clean the deep bed demineralizer resins. This allows the resin bed to remain
fixed throughout its useful life, so the resin ion exchange capacity can be more fully utilized
while maintaining low effluent impurities compared with beds that are mixed by periodic
external resin cleaning.

Two U.S. plants currently have partial-flow condensate filtration. These are Dresden 2 and
Clinton. Dresden 2 has a headered partial-flow filtration system that treats 40% of the
condensate flow at full power. This means that each demineralizer bed is still challenged with
approximately 60% of the hotwell iron. Therefore, the condensate demineralizers at Dresden
still required periodic resin cleaning (via ARCS) and the benefits of undisturbed ion exchange
resin beds cannot be achieved. In the selection of ion exchange resins for Dresden, the crud
removal capability of the resins must still be considered. Implementation of 40% headered
condensate filtration at Dresden 3 is planned by the end of 2002.

At Clinton, six of the nine demineralizer vessels have dedicated filters upstream. For those
demineralizers with filters, resin cleaning should not be required. For the three demineralizer
vessels without filters upstream, the resins must continue to be periodically cleaned (by URC).

The ion exchange resins used in deep bed demineralizers with upstream filters are listed in Table
10-3. The cation resins are all gel types having cross-linkage of 10% or higher with the exception
of Clinton, which uses four beds having some weak acid resin. Hope Creek has loaded Rohm and
Haas IRN-99 16% cross-linked cation resin in all beds to benefit from its greater stability and
lower extractable sulfur impurities. Some of the cation and anion resins are used together to
provide a mixture that is less prone to hydraulic separation than standard resins. The Dow
575C/SBR-C and Rohm and Haas IRN-97/IRN-78 are examples of such “less separable”
mixtures. The objective of less separable resins in this application are to avoid hydraulic
classification during resin transfer into the demineralizer or during hydraulic disturbances that
would tend to result in cation resin settling to the bottom of the bed.

Eleven of fourteen plants with Filter + Deep Bed condensate polishing add some of the anion
resin as an anion underlay when a bed is replaced. This tends to provide better sulfate control
during operating periods of peak condensate temperature and to extend life of the resin bed.

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Plants with Filter + Deep Bed condensate polishing, in particular those with full-flow filtration or
those demineralizers that have a dedicated filter directly upstream, normally target resin useful
lives of 3 to 4 years or more. This results in significantly lower operating costs for resin
purchase and disposal compared to Deep Bed Only plants, which typically replace low cross-
linked resins after only 6 months to 1 year of service and replace other resins after 1 – 2 years of
service.
Table 10-3
Resins Used with Filter + Deep Bed Condensate Polishing

# of
Plant Supplier Cation Resin Anion Resin
Vessels

Standard resin Non-porous standard resin

Epicor 3
(premixed) (premixed)
Brunswick 1
Rohm &
IRN-170 (1) IRN-170 3
Haas

Standard resin Non-porous standard resin

Epicor 5
(premixed) (premixed)
Brunswick 2
Rohm &
IRN-170 (1) IRN-170 1
Haas

Dow HGR-W2 SBR-C 5

Clinton
Rohm &
IRN-99/IRC-86RF IRN-78 4
Haas

Dow 575C SBR-C 5

Dresden 2
Rohm &
IRN-97 IRN-78 2
Haas

Rohm &
Hope Creek IRN-99 IRN-78 7
Haas

Laguna Verde 1 Dow Monosphere 650C Monosphere 550A 7

Laguna Verde 2 Dow Monosphere 650C Monosphere 550A 7

Rohm &
LaSalle 1 IRN-97 IRN-78 7
Haas

Rohm &
IRN-99 IRN-78 4
Haas
LaSalle 2
Rohm &
IRN-97 IRN-78 3
Haas

Rohm &
Limerick 1 IRN-160 (2) IRN-160 8
Haas

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Table 10-3 (continued)

Resins Used with Filter + Deep Bed Condensate Polishing

# of
Plant Supplier Cation Resin Anion Resin
Vessels

Rohm &
Limerick 2 IRN-160 IRN-160 8
Haas

Rohm &
Perry IRN-97 IRN-78 6
Haas

Rohm &
IRN-97 IRN-78 6
Haas
Susquehanna 1
US Filter
Purecat C-550 LS A-284-C 1
(3)

Rohm &
Susquehanna 2 IRN-97 IRN-78 7
Haas
Table 10-3 Notes
1. Premixed IRN-99 and IRN-78
2. Premixed IRN-97 and IRN-78
3. Cation resin is treated Dow 575C; anion resin is treated Dow SBR-C.

Maintenance of Condensate Polishers and Support Systems

An effective maintenance program is essential for Deep Bed Only plants for long term control of
feedwater iron and reactor water quality. Although the condensate polishing and support
systems are typically categorized as “balance of plant” equipment, it is important to adequately
maintain them to assure performance. Several plants have taken an aggressive system
maintenance approach.

An obvious area that directly affects iron control for Deep Bed Only plants is the maintenance of
the resin cleaning system. The key focus should be on process monitoring to identify
performance problems to trigger corrective maintenance as required. Preventive maintenance
programs should be established to avoid problems that will eventually affect performance. This
approach to maintenance is important no matter what resin cleaning system is used.

For all deep bed plants, the polisher service vessels should be periodically opened and internally
inspected to assure that resin transfers are complete and that the condition of the internal lining,
inlet distributor, resin distributor, and underdrain system are satisfactory. Problems in any of
these areas can affect resin life and flow distribution, which can affect iron removal and ion
exchange performance. New approaches to the internal lining of vessels show promise,
particularly the use of a coating instead of replacement with new rubber lining; the coating
requires less down time to apply and appears to be a lower source of sulfur. A coating has been
applied to condensate polisher service vessels at Clinton and Brunswick.

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Summary of Experience with Deep Bed Condensate Demineralizers

To perform a proper internal inspection of the vessel, isolation valves should be maintained to
avoid significant leaks that would impede inspections or create a contamination concern. With
shorter refueling outage durations, means of vessel isolation should be available to allow on-line
inspections and maintenance. Condensate polisher vessel and support system areas should be
maintained as clean as practical, including maintaining low contamination levels and low
radiation dose rates.

The resin transfer equipment and sequences should provide essentially complete resin transfers
while minimizing resin attrition. This typically means that the transfer, backwashing, mixing,
and rinsing capabilities of the regeneration system must function properly. It is important to
maintain the integrity of each bed to keep older resins separate from newer resins and to avoid
co-mingling of different types of resins (standard, low cross-linked, less-separable, etc.).

It is also important to maintain an effective means of measuring the resin inventory of each bed.
When a bed is short, the iron removal and ion exchange efficiency often measurably decrease.
While most plants have some means of measuring the total bed volume (usually via a “top of
bed” sight window on a regeneration system vessel or resin cleaning holding tank), the majority
of plants do not have the means to determine the volumes of the cation resin and anion resin
components in each bed. So, when a bed is short, it may be difficult to determine how much
cation resin and anion resin to add. The best approach is to maintain the resin transfer and
cleaning systems as required to avoid conditions that cause resin losses.

The capability to obtain representative samples of the ion exchange resins during their service
life should be provided. In-service resin samples are typically analyzed to confirm calculations
of ionic loading, to measure ion exchange kinetics and to assess the effectiveness of resin
cleaning operations. For resin beds that are periodically transferred out of the demineralizer
vessel for external cleaning and then returned for another service cycle, resin sampling during
the transfer(s) is appropriate. Alternatively, resins from such beds may be sampled from tanks in
the regeneration or resin cleaning system. For beds that treat fully filtered condensate and
therefore remain undisturbed in the service vessel throughout their useful lives, means to obtain a
resin core sample from the demineralizer vessel is needed.

References

1. “EPRI BWR Deep Bed Resins Report” (Electronic Report Draft), September 2002.

2. “BWR Iron Control Monitoring,” TR-109565, Final Report, September 1999.

3. Hillman, Robert J. and Joseph F. Giannelli, “Sulfate from a Salt Water Condenser Leak At
Oyster Creek Nuclear Generating Station,” 2000 EPRI Workshop on Condensate Polishing,
June 26 – 28, 2000, Annapolis, Maryland.

4. “BWR Iron Control – Volume 1: Deep Beds, TR-107297-V1, December 1996.

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11
CONDENSATE POLISHING WITH ION EXCHANGER
BYPASS

Bypass Operation

The condensate polishing systems in most BWRs were designed to accommodate flows
associated with the original full power heat balance of the plant, with little excess capacity. In
general, the systems were designed in one of two ways: 1) with a spare vessel that would be
placed in service when another vessel was removed from service for cleaning or vessel
maintenance, or 2) with vessels sized such that when one vessel was taken out of service the
vessels remaining in service could accommodate the additional flow.

At several plants, condensate flows have increased significantly above the original design flows
due to significant power uprates. Power uprates or extended power uprates (EPU) have been
made, are in progress and are being planned for future implementation at many utilities. Figure
11-1 shows the status of power uprates.

8 40

7 35

6 30
Cumulative No. of Uprates
5 25
No. of Uprates

4 20

3 15

2 10

1 5

0 0
1985 1992 1994 1995 1996 1997 1998 1999 2000 2001 2002
No. Implemented No. Pending
No. Approved No. Planned for Future Submittal
No. Under Evaluation Cumulative No. of Uprates

Figure 11-1
Power Uprate Status

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Condensate Polishing with Ion Exchanger Bypass

Thirty-five power uprates have either been implemented or are in some stage of evaluation,
approval or implementation through 2002. The impact of the increased condensate flow
associated with a power uprate on the condensate polishing system must be evaluated.
Alternatives need to be considered if the existing design is not adequate for continuous operation
at the increased flow.

One alternative is to install additional vessels (filter demineralizers, deep bed demineralizers or
pre-filters) to accommodate the increased flow. At some plants this is impractical or very costly
due to insufficient space for additional vessels or other limitations.

Plants that have a spare vessel may be able to place this vessel in service for the additional flow.
However, when a vessel is then taken out of service for cleaning or maintenance, if the
remaining vessels are not capable of processing the full flow, the condensate flow would have to
be reduced by reducing plant power, or a portion of the condensate may need to be routed to
bypass the condensate polishing system.

The chemistry impact of the alternatives must be part of the evaluation process. Conductivity,
sulfate concentration, chloride concentration and iron transport are among the parameters that
can be significantly affected, particularly if a portion of flow bypasses the condensate polishing
system or if deep bed demineralizers have significant and frequent flow changes. The response
to condenser tube leaks will need to be addressed and improved monitoring capability may be
required.

The experiences of three plants that have made changes to condensate polishing operations as a
result of power uprates are discussed below.

Brunswick Experience

Each Brunswick plant has a GE - BWR 4/Type 5g (Mark II) reactor. Unit 2 started commercial
operation in November 1975 and Unit 1 began commercial operation in March 1977. The OLTP
(original licensed thermal power) for each unit was 821 MWe and 2436 MWth. A stretch power
uprate to 105% OLTP was implemented at Unit 1 in 1996 and Unit 2 in 1997, increasing the
power of each unit to 895 MWe and 2558 MWth. Power was increased at Unit 1 in June 2002 to
113% OLTP (2758 MWth). A power uprate to 120% OLTP is being planned. The plant design
includes forward-pumped high pressure drains (FPD) and moisture separator reheaters. The
original condenser tubes, which were copper/nickel alloy, were replaced with titanium. The
main condenser cooling water source is seawater.

The original condensate polishing system design consisted of four Condensate Filter
Demineralizers (CFDs), which were operated with precoat materials applied, and six Deep Bed
Condensate Demineralizers (CDDs), with chemical regeneration of the resins. The CFDs are
now operated as non-precoat filters and the CDDs are now operated in a resin replacement mode;
i.e., chemical regeneration of the resins has been abandoned. The system is configured such that
it can be operated either with the CFDs upstream of the CDDs (pre-filter mode) or with the
CFDs downstream of the CDDs (post-filter mode). The system is always operated in the pre-

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Condensate Polishing with Ion Exchanger Bypass

filter mode. The system also has the capability to partially bypass some flow around the CFDs
or the CDDs, or both.

The power uprate to 120% OLTP will require bypassing a portion of the flow around the deep
bed demineralizers when one of the six beds must be removed from service for resin
replacement. Both units have operated for short periods of time with one deep bed demineralizer
bypassed. Data for operating in this mode are presented in Figures 11-2 and 11-3 for Unit 1 and
Figures 11-4 and 11-5 for Unit 2.

Figure 11-2 shows the data for a two-day period in June 2001 when the plant operated with one
deep bed demineralizer out of service. The data show a significant decrease of approximately
0.5 ppb in reactor coolant sulfate. Chloride levels remained below 0.5 ppb during this period.

From August to October 2001, Unit 1 operated with one deep bed demineralizer out of service
for underdrain strainer and lining upgrades. The data in Figure 11-3 also show a reduction in
coolant sulfate concentration between 0.5 and 1 ppb.

At Unit 2, a deep bed demineralizer was bypassed for a four-day period in September 2001. Unit
2 showed a reduction in reactor coolant sulfate levels of 0.2 - 0.4 ppb during this period as shown
in Figure 11-4.

Data for Unit 2, which operated with one demineralizer out of service for various periods
between May 1, 2002 and June 15, 2002, are shown in Figure 11-5. The data show a cyclical
trend where one demineralizer was removed from service for a few days, followed by a period of
all demineralizers in service, and then back again to one out–of-service demineralizer. This
cyclical pattern lasted for the first three weeks in May 2002, until a condenser tube leak occurred
during the last week of the month. The data in the first three weeks of May appear to show some
lower sulfate values with one demineralizer out of service but the data are not always consistent.

Reactor coolant chloride levels initially remained at 0.5 ppb, but just after 5/14/02, the data show
a slight increase in chloride levels when a demineralizer was bypassed. This may have been the
first indication of a change in condenser salt water leakage. The data just before 5/24/02 clearly
show a significant increase in chloride concentrations when a demineralizer was bypassed.
Between 5/19/02 and 5/21/02, chloride increased from 0.7 to 1.2 ppb following the bypassing of
one demineralizer. On 5/23/02, chloride increased from 0.56 to 2.53 ppb, within 14 hours of
bypassing one demineralizer.

From 5/24/02 through 6/3/02, all demineralizers were in service while the station addressed the
condenser tube leak. CPD chloride reached a maximum of 10 ppb, while reactor coolant
chloride levels were about 1 ppb (all demineralizers in service). The tube leak was repaired on
6/1/02, as CPD chloride levels returned to < 0.5 ppb. With reactor water sulfate levels near 3
ppb on 6/3/02, one demineralizer was bypassed. Sulfate decreased to about 2.7 ppb, while
chloride increased slightly (to 0.77 - 0.92 ppb). On 6/6/02, another condenser tube leak
occurred, resulting in a significant increase in reactor water chloride concentration to about 3
ppb, with one demineralizer still in bypass. Reactor water sulfate at this time was 3.2 ppb. The
bypassed demineralizer was placed back in service. Reactor water sulfate continued to increase

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Condensate Polishing with Ion Exchanger Bypass

and reached a maximum value of about 13 ppb on 6/7. The station data noted resin fines present
in the demineralizer effluent (“C” Demineralizer) that was placed back in service. It was also
noted that following the placement of the bypassed demineralizer back in service, the RWCU
system was shutdown for a period of about 11 hours. Reactor coolant chloride reached a
maximum value of about 4 ppb while RWCU was out of service. Once RWCU was restored,
reactor coolant chloride levels decreased to between 0.5 and 0.7 ppb. CPD chloride levels
reached a maximum of 9 ppb prior to the repair of the tube leak.

The bypass data show that while sulfate levels could decrease due to the reduction in sulfate
source term (the amount of deep bed resin in contact with the condensate), the occurrence of a
condenser tube leak with a demineralizer bypassed will quickly elevate coolant chloride levels
above 1 ppb (median value used in the INPO/WANO Chemistry Indicator formula).

5.0 150

4.5

4.0 120
Reactor Coolant Anions, ppb

3.5

Temperature (F)
3.0 90

2.5 B Demin NIS

2.0 60

1.5

1.0 30

0.5

0.0 0
6/26/01 6/27/01 6/28/01 6/29/01 6/30/01 7/1/01 7/2/01 7/3/01 7/4/01 7/5/01

Sulfate Chloride Temperature

Figure 11-2
Brunswick 1 Coolant Sulfate and Chloride Trends during CDD Bypassing (June 2001)

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Condensate Polishing with Ion Exchanger Bypass

14 140

12 120
Reactor Coolant Anions, ppb

10 100

Temperature, (F)
8 80
B Demin NIS

6 60

4 40

2 20

0 0
7/28/01 8/12/01 8/27/01 9/11/01 9/26/01 10/11/01 10/26/01 11/10/01

Sulfate Chloride Temperature

Figure 11-3
Brunswick 1 Coolant Sulfate and Chloride Trends during CDD Bypassing (August-October
2001)

5 150

4 120
Reactor Coolant Anions, ppb

Temperature, (F)
3 90

A Demin NIS
2 60

1 30

0 0
8/31/01 9/2/01 9/4/01 9/6/01 9/8/01 9/10/01 9/12/01 9/14/01 9/16/01

Sulfate Chloride Temperature

Figure 11-4
Brunswick 2 Sulfate Trend during CDD Bypassing (September 2001)

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Condensate Polishing with Ion Exchanger Bypass

10

9
Anions (ppb) or # of CDDs in Service

0
4/24/02 5/4/02 5/14/02 5/24/02 6/3/02 6/13/02 6/23/02

SO4 Rx Cl CPD Cl # of CDDs I/S

Figure 11-5
Brunswick 2 Sulfate and Chloride Trends during CDD Bypassing (May-June 2002)

At Brunswick a condenser leak of less than 0.1 gpd with 1/6 of the condensate flow bypassed
will cause reactor water chloride to exceed 1 ppb. A maximum of approximately 0.05 ppb
chloride is allowed in the condensate pump discharge at 120% OLTP to maintain reactor water
chloride concentrations below 1.0 ppb with 15 % condensate bypass. The most accurate
indication of a condenser leak with condensate bypass could be obtained from the reactor water
chloride concentration using an in line ion chromatograph (IC). Any indication of increased
chlorides would require the plant to discontinue bypass until the condenser leak was plugged.

Dresden Experience

Dresden Station has two operating BWR-3 reactors. Commercial operation began for Dresden 2
in June 1970 and for Dresden 3 in November 1971. Dresden 2 is rated for 912 MWe and 2957
MWth following the power uprate that was fully implemented early in 2002. Dresden 3 is
currently rated for 833 MWe and 2527 MWth, and will be modified for power uprate as at
Dresden 2 following the fall 2002 refueling outage. The plant design includes cascaded heater
drains, no moisture separator reheaters and a reactor water cleanup system with a maximum
capacity of approximately 700 gpm (normally operated at about 600 gpm). Condensate flow
from the main condenser hotwells is processed through the deep bed condensate demineralizers
and represents 100% of the final feedwater flow. The condenser tubes are constructed of 304
stainless steel and the main condenser circulating water is river water (once-through cooling in
the summer and recirculation cooling in the winter).

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Condensate Polishing with Ion Exchanger Bypass

The condensate demineralizer system includes seven (7) deep bed demineralizer vessels. The
original equipment manufacturer is Graver. The system was originally designed for chemical
regeneration of the ion exchange resins. However, since the 1980s, condensate demineralizer
resins have not been regenerated but are replaced when no longer suitable for condensate
polishing service based on ionic loading, anion resin kinetics or chemistry performance limits.

At 100% power, the station normally runs with six (6) demineralizer vessels in service with the
seventh vessel used as a spare. Based on the 15% extended power uprate, the full power
condensate flow was increased from 19,600 gpm to 22,540 gpm. The original flow design for
the demineralizer vessels was 3267 gpm normal and 3561 gpm maximum. At extended power
uprate conditions, the normal flow per vessel with 6 beds is 3757 gpm and with 7 beds is 3220
gpm.

As part of the Dresden 2 EPU modifications, a headered partial flow condensate filtration system
was installed upstream of the condensate demineralizers. The filter system processes
approximately 40% of the condensate flow at full power. A full-flow filtration system was not
installed due to space limitations. Removal of iron crud particles by the filters approaches 100%,
thus reducing the inlet iron to the condensate demineralizers by approximately 40%. The filter
system consists of two 66” diameter by 12’-4” straight side top tubesheet filter vessels
employing pleated polypropylene filter septa. A similar system is being installed for Unit 3 and
is planned to be placed in service in conjunction with EPU following the fall 2002 refueling
outage.

In Dresden Unit 2, when going from 7 condensate demineralizer vessels in service to six vessels
for the purpose of removing one vessel for resin cleaning, the average flow through the vessels is
increased by 16.7% . To accomplish this evolution, first the condensate demineralizer bypass
valve is opened, decreasing the flow through each vessel, and then the bed to be cleaned is
removed from service. This procedure avoids a sudden increase in flow through the in-service
vessels to avoid disturbance of the crud layer on the bed that can lead to iron breakthrough.
After the vessel is removed from service, the flow through each in-service vessel is about equal
to the flow with 7 vessels in service and no bypass. Finally, the condensate demineralizer bypass
valve is gradually closed, raising the flow per vessel with the full condensate flow going through
6 vessels.

So far, Dresden 2 has only remained in the condensate demineralizer bypass condition for the
short time (an hour or so) that it takes for the valve manipulations. No adverse impacts on
reactor water chemistry have been reported.

KKL Experience

The Leibstadt Nuclear Power Plant (KKL) is a GE Boiling Water Reactor (BWR-6) located in
northern Switzerland that started commercial operation in 1984. It had an original net electrical
power output of 942 MWe (3012 MWth) and through a series of power uprates currently
produces 1230 MWe (3600 MWth), which equates to a 19.5% increase in thermal power.

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Condensate Polishing with Ion Exchanger Bypass

The plant has forward pumped heater drains and operates with normal water chemistry (NWC)
and depleted zinc oxide injection (DZO). Condenser tubes are titanium. The cooling water
source is taken from the confluence of the Rhine and Aare rivers. The current Reactor Water
Clean-up (RWCU) system flow is approximately 1% of the 3600 MWth power uprate total
feedwater flow.

The condensate polishing system includes five (5) bottom tubesheet powdered resin condensate
filter demineralizers (CFDs). There are 378 elements in each vessel. All five CFDs were
modified in 1988/1989 with draft tubes. The original design was to run with four (4) CFDs in
service and one (1) CFD in standby. Shortly after commissioning, all five (5) filters were
operated in order to achieve reasonable filter run lengths.

As part of a power uprate and to reduce resin consumption, a study was performed to evaluate
non-precoat filter elements in one CFD (1, 2). During the refuel outage in 1997, Memtec
HGPPB-1 (1 micron) non-precoated filter elements were installed in one CFD vessel.
Equipment modifications and procedure revisions were also made for the backwash of the non-
precoated filter only.

These elements had to be replaced after two years of operation because of cracks in the tips of
the pleats. They were replaced with an improved version of the same 1-micron rating elements.

In the 2001 refuel outage, the non-precoated pleated Memtec HGPPB-1 elements were again
replaced with the same type element but with a 4-micron rating. Precoated yarn wound elements
in another vessel were replaced with non-precoated pleated Pall BPF-5 elements with a 1-micron
rating.

Resin consumption for the cycle following installation of the non-precoated elements was
approximately 27% (3080 lbs) less than the previous operating cycle, as shown in Figure 11-6
(compare data point 96/97 to 97/98). Over the next three years, resin consumption increased,
with the 2000/2001 usage just slightly under the value for the last operating cycle with all
precoat filters. The increase is attributed to the 12% power uprate over the period. Resins
consumption has since declined with the operation of the second non-precoat filter.

Improvements in reactor water quality were also observed, as shown in Figure 11-7. The
average reactor coolant sulfate value for the two years prior to installation of the first non-
precoat filter was 1.68 ppb. For the four-year period with one operating non-precoat filter,
reactor coolant sulfate averaged 1.41 ppb. Reactor coolant sulfate has averaged 1.33 ppb since
the initial operation of the second non-precoat filter. As shown in Figure 11-7, 50% of the total
condensate flow is processed through the non-precoat filters.

Because 50% of the total condensate flow not processed through ion exchange media, KKL has
implemented improved condenser leak monitoring capabilities. These capabilities include
hotwell conductivity instrumentation and CFD inlet sodium monitoring. The station installed two
new sodium monitors (Swan, Inc. Soditrace) in the CFD inlet. The stated range of detection for
the new sodium monitors is 0.001–100 ppb. Prior to operation of the first non-precoat filter,
KKL lowered condenser leakage action levels. The most significant change was the lowering of

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Condensate Polishing with Ion Exchanger Bypass

the CFD inlet action level conductivity value for a large leak from 5 µS/cm to 0.2 µS/cm with
one non-precoat filter and to 0.15 µS/cm with two non-precoat filters.

40,000 5.0

and No. of Non-Precoat Filters in Service

36,000 4.5
32,000 4.0
Precoat Material Usage
(dry lbs resin/year)

28,000 3.5
24,000 3.0

Iron (ppb)
20,000 2.5
16,000 2.0
12,000 1.5
8,000 1.0
4,000 0.5
0 0.0
84
84/85
85/86
86/87
87/88
88/89
89/90
90/91

92/93
93/94
94/95
95/96
96/97
97/98
98/99
99/00
00/01
91/92

01/02
Year

Precoat Material CDE Iron Feedwater Iron Non-Precoat CFDs

Figure 11-6
KKL Resin Consumption

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Condensate Polishing with Ion Exchanger Bypass

8.0 60

Percent of Total Condensate Flow through Non-

Reactor Coolant Chloride and Sulfate (ppb)

6.0 45

Precoat Filters
4.0 30

2.0 15

0.0 0
3/7/1994 3/6/1996 3/6/1998 3/5/2000 3/5/2002

Chloride Sulfate Flow

Figure 11-7
KKL Reactor Coolant Anion Trends

Summary

Based on the experiences of the three plants, the practicality of a BWR bypassing or not applying
ion exchange media to a portion of the condensate flow has been demonstrated. KKL is the only
plant of these three that normally operates with continuous, sustained bypassing. Dresden 2
bypasses for a short time when a vessel is removed from service at full power, and the same
practice is planned for Dresden 3. Brunswick has bypassed the deep bed demineralizers for
some extended periods, and reactor water ionic impurities have either increased or decreased
depending on the condenser leak rate. There have not been any significant detrimental effects to
date from partially bypassing or not applying ion exchange resins at all three plants. Benefits
from partially bypassing the ion exchange media include reduced sulfates and reduced resin
usage and disposal costs.

Prior to bypassing ion exchange media, a detailed evaluation of the required operational and
monitoring changes should be performed. It may be worthwhile to collect test data to ensure that
the response of the plant is as expected. Greater attention to monitoring plant performance,
particularly condenser inleakage, will be required. For plants with full-flow condensate filtration
upstream of deep beds (such as Brunswick) or filter demineralizers with a portion of the vessels
equipped with non-precoat filter septa (like KKL), the feedwater and reactor water impact is
strictly related to the soluble species in the condensate pump discharge stream. For a plant with
partial headered condensate filtration (such as Dresden) or with Deep Bed Only condensate
polishing, flow changes through operating deep bed demineralizers need to be performed

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Condensate Polishing with Ion Exchanger Bypass

gradually to avoid disturbing the crud layer that has built up on the beds. This may require
equipment, control system and procedural changes. Finally, plant personnel should be trained in
the changes to be made, anticipated results of the changes and reasons for the changes.

Plants with condensate filter demineralizers may be the best candidates to consider partial
bypassing, similar to KKL, because they must normally maintain low condenser cooling water
inleakage due to the ion exchange performance limitations of the powdered resin precoats.
However, plants with copper alloy condensers and those with a significant cobalt source term
upstream of the condensate polishing system should carefully evaluate the impact of bypassing.
At Filter + Deep Bed plants such as Brunswick, the full-flow filters maintain low feedwater
corrosion products when a deep bed demineralizer is bypassed. Seawater cooled plants must
maintain very low condenser inleakage rates, below 1 gpd, to realistically consider ion exchange
bypassing. Although the technology exists for locating condenser tube leaks in the range of 1
gpd, it has not been widely applied at such low leak rates. Chemistry monitoring instrumentation
for the early detection of a small condenser leak is available.

References

1. Wilfried Kaufmann and Daniel Brack, “Condensate Cleaning in a BWR With Non-
precoated Filter Septa Without any Ion Exchange Capacity”, Proceedings: 1999
Workshop on Condensate Polishing, EPRI, TR-113281.

2. Daniel Brack , “13th BWR Chemistry Committee Meeting”, Nashville, TN, May 15-
17,2002

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12
OTHER ISSUES FOR CONSIDERATION IN THE NEXT
REVISION OF THE BWR CHEMISTRY GUIDELINES

During the ongoing BWR data gathering effort and the many contacts with station personnel in
preparation of this report, a number of issues were suggested for consideration in the next update
of the EPRI BWR Water Chemistry Guidelines. These issues are summarized below:

1. Provide the additional guidance, correlations or tools needed for a site-specific determination
of when a plant can continue operating or when it must shut down during a transient, such as
a resin intrusion. This topic is addressed in Appendix C of the EPRI BWR Water Chemistry
Guidelines (1), which states, “…an action plan in the event of a chemical intrusion should be
developed and approved by all responsible chemistry and fuels staff.” It goes on to say, “A
plant specific plan will need to include reaction times and action levels for response to
specific transients if they occur. It is important that the potential effect of a transient on all
plant components, including the fuel, be considered.” When a resin intrusion occurred at
Pilgrim in December of 2000, an evaluation had not been performed in advance. It was
found that the available BWRVIP crack growth rate model was inadequate at the high
concentrations present during the intrusion. Per input from chemistry managers, this
guidance should include a list of questions that should be asked and several examples that
stations have developed, to address some of the more common chemistry excursions. Some
of the considerations include guidance on whether to stay on HWC, to reduce power (effect
on crack growth rate), to shut down, change out CP resin, change out RWCU resin, secure Zn
addition, etc.

2. Provide guidance on expected plant responses to a resin intrusion under the various
chemistry regimes, including NWC, HWC, NMCA with hydrogen injection and zinc
injection. This will help plants confirm or eliminate a resin intrusion as the cause of a
transient. A technical explanation for each response should be included. While some of the
chemistry responses to a resin intrusion are well known, other chemistry and system
responses are not obvious. For example, the December 2000 resin intrusion at Pilgrim, while
on moderate HWC with DZO injection at approximately 100% power, produced the
following responses (2):
• Feedwater conductivity increased.
• Reactor water conductivity increased.
• Reactor water pH decreased.
• Reactor water sulfate increased.
• ECP increased.

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Other Issues for Consideration in the Next Revision of the BWR Chemistry Guidelines

• Reactor water zinc increased by approximately 100 ppb.

• Reactor water dissolved oxygen increased.
• Reactor water total iodine decreased by a factor of 10.
• Hotwell raw and cation conductivity increased.
• Main steam line radiation levels decreased.
• Offgas radiation monitors spiked high.
• Reactor power increased by approximately 8 MWth.

For illustration, data from the Pilgrim 12/00 resin intrusion are shown in Figures 12-1 through
12-11. Note that it is not known whether the indicated bottom head drain temperature increase
shown in Figure 12-11 was due to a flow change or cleaning of the RTD thermowell by the low
pH from the resin intrusion.

PNPS 12/2/00 Resin Intrusion

1200 1.60

1.40
1000

1.20

Cation Resin Quantity (liters)

Reactor Water Sulfate (ppb)

800
1.00

600 0.80

0.60
400

0.40

200
0.20

0 0.00
12/01/00 12/02/00 12/03/00 12/04/00 12/05/00 12/06/00 12/07/00 12/08/00

S ulfate Calc Rx SO4 Cation Resin

Figure 12-1
Reactor Water Sulfate and Estimated Cation Resin Quantity

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Other Issues for Consideration in the Next Revision of the BWR Chemistry Guidelines

PNPS 12/2/00 Resin Intrusion

6.000 8

7
5.000

4.000
5
uS/cm

pH
3.000 4

3
2.000

1.000
1

0.000 0
12/2/2000 0:00 12/2/2000 12:00 12/3/2000 0:00 12/3/2000 12:00 12/4/2000 0:00 12/4/2000 12:00

Rx Conductivity Rx pH

Figure 12-2
Pilgrim Reactor Water Conductivity and pH Responses

PNPS 12/2/00 Resin Intrusion (Rx Water Sulfates)

6.000 700.0
640

600.0
5.000
520
488
500.0
4.000

375 400.0
uS/cm

ppb

3.000

300.0
260

2.000
175 200.0
160
140
116
1.000 90
69 100.0

1.6
0.000 0.0
12/2/00 12/2/00 12/2/00 12/2/00 12/2/00 12/2/00 12/2/00 12/2/00 12/3/00 12/3/00 12/3/00
4:48 7:12 9:36 12:00 14:24 16:48 19:12 21:36 0:00 2:24 4:48
Rx Conductivity Rx Sulfates

Figure 12-3
Pilgrim Reactor Water Conductivity and Sulfate Responses

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Other Issues for Consideration in the Next Revision of the BWR Chemistry Guidelines

PNPS 12/2/00 Resin Intrusion (Reactor Water Zinc)

6.000 120

108

5.000 100 100

4.000 80

65
uS/cm

ppb
3.000 60

42
2.000 40

22
1.000 20
16
12 12.5
8.9
6.2
0.000 0
12/2/00 12/2/00 12/2/00 12/2/00 12/2/00 12/2/00 12/2/00 12/2/00 12/3/00 12/3/00 12/3/00
4:48 7:12 9:36 12:00 14:24 16:48 19:12 21:36 0:00 2:24 4:48
Rx Conductivity Rx Zinc

Figure 12-4
Pilgrim Reactor Water Conductivity and Zinc Responses

PNPS 12/2/00 Resin Intrusion

0 10.000

-100 8.000

-200 6.000
mVSHE

ppb

-300 4.000

-400 2.000

-500 0.000

-600 -2.000
11/30/2000 12/1/2000 12/2/2000 12/3/2000 12/4/2000 12/5/2000 12/6/2000 12/7/2000 12/8/2000 12/9/2000 12/10/2000
0:00 0:00 0:00 0:00 0:00 0:00 0:00 0:00 0:00 0:00 0:00

ECP Dissolved O2

Figure 12-5
Pilgrim Reactor Water Dissolved Oxygen and ECP Responses

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Other Issues for Consideration in the Next Revision of the BWR Chemistry Guidelines

PNPS 12/2/00 Resin Intrusion Feedwater/Condensate Conductivity

0.4
0.379

0.35

0.313
0.3

0.25

Hotwell
uS/cm

0.2 Feedwater
Hotwell Cation

0.15 0.151
08:50
Start Resin
Intrusion
0.1

0.057
0.05

0
8:02:39
8:10:23
8:18:07
8:25:51
8:33:35
8:41:19
8:49:03
8:56:47
9:04:31
9:12:15
9:19:59
9:27:43
9:35:27
9:43:11
9:50:55
9:58:39
10:06:23
10:14:07
10:21:51
10:29:35
10:37:19
10:45:03
10:52:47
11:00:31
11:08:15
11:15:59
11:23:43
Time

Figure 12-6
Pilgrim Reactor Water, Feedwater and Condensate Conductivity Responses

PNPS 12/2/00 Resin Intrusion Main Steam Activity

4500

4000

3500
12/2 Resin Intrusion

3000
Average mR/hr

2500

2000

1500
H2 Reduction

1000

500

0
20-Nov-00 25-Nov-00 30-Nov-00 05-Dec-00 10-Dec-00 15-Dec-00 20-Dec-00 25-Dec-00
00:00:00 00:00:00 00:00:00 00:00:00 00:00:00 00:00:00 00:00:00 00:00:00

Figure 12-7
Pilgrim Main Steam Line Activity Response

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Other Issues for Consideration in the Next Revision of the BWR Chemistry Guidelines

PNPS 12/2/00 Resin Intrusion Offgas Rad Monitor Activity

90 1800

80 1600

70 1400

60 1200

50 1000
mr/hr

cps
40 800
Resin Intrusion

30 600
Offgas Sample

20 400

10 200

0 0
12/2/2000 6:08 12/2/2000 7:48 12/2/2000 9:18 12/2/2000 10:58 12/2/2000 12:38 12/2/2000 14:18 12/2/2000 15:58 12/2/2000 17:38

Offgas-A Offgas -B PTRM-A

Figure 12-8
Pilgrim Offgas Activity Response

PNPS 12/2/00 Resin Intrusion

1996

1994

1992

1990
Resin Intrusion
Reactor Power (MWT)

1988

1986

1984

1982

1980

1978
12/1/00 19:12 12/2/00 0:00 12/2/00 4:48 12/2/00 9:36 12/2/00 14:24 12/2/00 19:12 12/3/00 0:00 12/3/00 4:48

Figure 12-9
Pilgrim Reactor Thermal Power Response to Resin Intrusion

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Other Issues for Consideration in the Next Revision of the BWR Chemistry Guidelines

PNPS 12/2/00 Resin Intrusion Reactor Water Total Iodine)

6.00E-03

5.00E-03

4.00E-03

Downpower
uCi/ml

3.00E-03

2.00E-03

Resin Intrusion

1.00E-03
Downpower
Forced outage

0.00E+00
10/16/00

10/23/00

11/13/00

11/20/00

12/11/00

12/18/00
9/4/00

9/11/00

9/18/00

9/25/00

10/2/00

10/9/00

11/6/00

12/4/00
10/30/00

11/27/00

12/25/00

1/1/01
Figure 12-10
Pilgrim Reactor Water Total Iodine Response

PNPS 12/2/00 Resin Intrusion Bottom Head Drain Temperature

600

500

400
Deg F

300

200

100

0
01-Dec-00 02-Dec-00 02-Dec-00 02-Dec-00 02-Dec-00 02-Dec-00 03-Dec-00 03-Dec-00
19:12:00 00:00:00 04:48:00 09:36:00 14:24:00 19:12:00 00:00:00 04:48:00

Figure 12-11
Pilgrim Bottom Head Drain Temperature Response

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Other Issues for Consideration in the Next Revision of the BWR Chemistry Guidelines

3. A recurring jet pump fouling phenomenon was reported by Perry to be the cause of
inadequate jet pump performance. The occurrence and causes of this issue should be
captured in the EPRI BWR Water Chemistry Guidelines so that chemistry personnel can
recognize and inform plant management of conditions conducive to jet pump fouling.
During some operating periods starting in 10/00, Perry was not able to produce maximum
power due to an inability to achieve maximum core flow. This is an economic issue, not a
safety issue. Contributing causes to jet pump fouling at Perry were concluded to be low
reactor water conductivity and an operating philosophy that results in more time at high core
flow; these conditions promote an attraction (attributed to zeta potential) of crud particles to
the jet pump surface. Perry is a “filter + deep bed” plant with forward pumped drains and no
iron addition; the average feedwater iron concentration in 2000 was approximately 0.4 ppb.
A low iron-to-zinc ratio in the reactor water was also suspected as a contributing cause that
could promote a loosely adherent corrosion film to develop on the fuel. Portions of such a
loosely adherent film can break off and be transported to jet pump nozzle and throat surfaces.
Calculations indicated that a deposit about 40 mils thick would be required to produce the
flow shortfall measured, and a jet pump inspection in RFO-07 showed almost twice this
amount. Perry experienced the flow degradation with zinc injection but no HWC or NMCA.
Perry has since applied NMCA in February 2001 and began injecting hydrogen in August
2002. NMCA with hydrogen injection was recommended by General Electric and
independent studies as a means of controlling jet pump fouling by forming a tighter and
thinner corrosion film that can incorporate more transition metals (zinc) at reduced ECP.

4. For plants adding iron to the feedwater to maintain >0.5 ppb, the significance of more iron
being added than is measured in the final feedwater should be addressed. It is currently not
known whether the difference is sampling error or an actual process phenomenon. If it is a
process phenomenon, when should it become a concern? For example, if it is a real process
phenomenon and an inventory of iron is being accumulated in the feedwater train, is there a
point where the potential release during a transient (such as a plant startup) becomes
excessive? Susquehanna, Unit 1 and Unit 2, continues to estimate that only 50% to 75% of
the iron being injected (as iron oxalate) is being measured in the final feedwater. Limerick
also reports less of an increase in final feedwater iron than expected based on the quantity of
iron oxide being injected.

5. The significance of increased H2O2 levels measured in reactor water and condensate under
cold shutdown conditions should be addressed. Hydrogen peroxide in reactor water and
RWCU effluent during cold shutdown have been measured in the 2 ppm range at Oyster
Creek and Pilgrim. Condensate (hotwell) levels have increased to the hundreds of ppb levels
in cold shutdown. BWRs, especially those with deep bed condensate demineralizers, have
experienced rapid increases in organic sulfur species in condensate, thereby increasing
sulfate concentrations in reactor water and in control rod drive inputs during cold shutdown.
Resins used in reactor water cleanup systems are also affected. The impact of hydrogen
peroxide under cold shutdown conditions, on resin degradation and sulfur release, and on
potential countermeasures to suppress or avoid its potential negative impacts on the system
should be considered. Sampling locations and analysis methods should also be addressed.

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Other Issues for Consideration in the Next Revision of the BWR Chemistry Guidelines

6. Improved guidance on the proper control parameters for DZO addition is needed. Zinc control
strategies vary among plants, and it is not well defined if and when feedwater zinc, reactor water
zinc or reactor water Co-60 is the correct control parameter. For plants trying to control
reactor water zinc, DZO usage during different operating periods can vary by a factor of
about 10, which has a significant impact on DZO operating cost. Other plants try to add just
enough zinc to achieve a minimum Co-60 concentration in the reactor coolant. The reactor
water Zn/Co-60 ratio has also been suggested as a parameter for dose control after NMCA.
The industry BRAC dose rate data also appear to correlate with the ratio of reactor water
Zn/feedwater iron.

7. In Appendix C of the EPRI BWR Water Chemistry Guidelines – 2000 Revision, the results
of a startup study illustrate the potential for high crack growth rates during startup/hot
standby conditions. Additional guidance is needed on chemistry control during shutdown
and startup conditions for minimizing crack growth rates during startups.

8. INPO is assessing each plant’s program for implementing the EPRI BWR Water Chemistry
Guidelines as part of the BWRVIP program. Compliance is an issue and a formal method for
evaluating and documenting deviations from the guidelines has been published in BWRVIP-
94 and approved by operating company executives. This should be captured in the Chemistry
Guidelines, noting the sections or parameters that will be evaluated for compliance.

9. The wording and format of the EPRI BWR Water Chemistry Guidelines in Section 5
regarding Table 5-1, Recommended BWR Chemistry Database Parameters, should be more
consistent with the wording of Section 7 of the PWR Secondary Guidelines. Regarding on-
line instrumentation, the PWR document states that, "in these tables a distinction is made
between the instrumentation array consistent with the chemistry control parameters of
Sections 5 and 6 and the optional instrumentation array recommended for problem
diagnosis". In the PWR Secondary Guidelines, Tables 7-1 and 7-2 distinguish between
continuous instrumentation recommended for control (C) vs. diagnostic (D) parameters.
Also Section 7.4 (Optimization of Data Collection), Table 7-4, of the PWR Secondary
Chemistry Guidelines differentiates between uses of data by mandatory requirements, short
term monitoring, and long term monitoring. Tables 7-4, 7-5 and 7-6 provide a framework for
a PWR to put together a plant specific optimization plan, which is now also required by
VIP/INPO for the BWRs. Therefore, the next revision of the BWR Chemistry Guidelines
should demonstrate how a BWR can develop a plant specific optimization plan and discuss
the value and use of each parameter in Table 5-1.

10. Table B-3 of the “EPRI BWR Water Chemistry Guidelines – 2000 Revision” addresses
diagnostic parameters for torus/suppression pool. The recommended conductivity limit is
<5.0 µS/cm. At FitzPatrick, for instance, the torus has a coating that contains zinc, some of
which goes into solution and causes an increase in conductivity. FitzPatrick measures torus
water concentrations of cationic and anionic species and calculates a corrected conductivity
by excluding zinc. The corrected conductivity is compared to the recommended limit. If this
practice is acceptable, then a note should be included under Table B-3.

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Other Issues for Consideration in the Next Revision of the BWR Chemistry Guidelines

11. It was suggested that a section should be considered on ensuring cleanliness of newly
installed equipment. For example, at Brunswick even though the new MSR chevrons met
supplier requirements, including steam cleaning, the station still had a small reactor water
anion transient on startup. A transient was also reported after installing new underdrain
strainers in a condensate demineralizer vessel. At FitzPatrick and Susquehanna, increased
TOC in condensate was measured following replacement of feedwater heaters; the TOC
source was residual oil on the surfaces of the new heaters.

12. The NaCl and Na2SO4 plots in Figure 5-1 of the EPRI BWR Water Chemistry Guidelines
should be corrected. The x-axis is labeled anion concentration, but it appears that the salt
concentration from Table 5-2 was used. The plot should be revised to use the anion
concentration values.

13. A concise statement of the basis for each of the chemistry control parameter action level
value would be beneficial. Such statements can be effectively used to inform management
why actions are required.

14. Additional guidance and discussion should be given on the impact of low feedwater iron and
the need for iron addition. The dose impact of operating with feedwater iron <0.5 ppb should
be addressed for plants that have had several cycles with DZO injection and moderate HWC
or HWC/NMCA. In addition, the need to maintain a minimum feedwater ratio of Fe/Zn,
and/or Fe/(Zn + Ni) and/or Fe/(Zn + Ni + Cu) to minimize localized corrosion at fuel spacer
contacts should be addressed. A discussion of this issue is included in Section 3.3.5 if the
BWR Water Chemistry Guidelines – 2000 Revision, with ratios of >2 to as low as 1
recommended depending on the fuel supplier. Based on EPRI BWR Chemistry Monitoring,
a number of plants have ratios below 2.

15. Updates of fuel surveillance data for plants that have performed NMCA should be included.
Specifically, Section 3 of the BWR Water Chemistry Guidelines – 2000 Revision should
include a discussion on the effects of NMCA on fuel and its potential impact on fuel
corrosion with and without hydrogen injection.

16. Additional guidance should be included on the use of post-UV IC analyses. Either
recommended limits or the methodology for developing site-specific limits should be
addressed. This would be particularly applicable to radwaste sample tank liquids that must
be returned to the condensate system, demineralized makeup water, condensate storage tanks
and CRD source water during plant shutdowns.

17. Additional guidance should be included on boron sources, the impact of boron on reactor
water chemistry (ion-conductivity balance) and effluents, and the appropriate boron control
range. This could be addressed in Section 5 of the BWR Water Chemistry Guidelines – 2000
Revision.

18. The EPRI BWR Water Chemistry Guidelines – 2000 Revision requires continuous feedwater
metals samples when power is >10%. At some plants, Vermont Yankee for example, steam
dumping is still occurring at this power level and metals are very difficult to measure
accurately for many reasons. The plant is then in a transient condition where feedwater and

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Other Issues for Consideration in the Next Revision of the BWR Chemistry Guidelines

condensate pumps are being started and placed in service. Crud bursts are likely. Integrated
metals samples are not practical. Grab samples may be the only way to monitor these
elements. The time between 10% power and 50% power is brief. Guidance should be
included that would allow grab sampling below 50% power and continuous corrosion
product sampling starting when plant power is above this level.

19. Additional guidance is needed on when a plant can use the higher action level limits for
chloride and sulfate. Should the Prior to Startup and Power Operation values be considered
hold points? During a recent review, Exelon elected too use the values Prior to Startup as
hold points and the Prior to Power Operation values as goals. Consideration should also be
given to whether the basis for action levels should be changed to reactor coolant temperature
or plant modes; i.e., differentiated according to >200 oF and < 200 oF or according to
Startup, Power Operation and Cold Shutdown.

20. Section 5.4.2, Secondary Parameters, of the BWR Water Chemistry Guidelines – 2000
Revision needs to be updated to reflect the information in the BWRVIP 92 experience report
on NMCA in the area of monitoring dissolved hydrogen and oxygen molar ratio when the
sample line is coated with noble metals and the dissolved oxygen reads zero with hydrogen
injection.

21. Table 2-3, NMCA Process Parameters, of the BWR Water Chemistry Guidelines – 2000
Revision needs to be updated to be consistent with best industry experience.

22. Sections 2 and 5 need to be updated in general for consistency with BWRVIP 62. In
particular, the validity of the EPRI empirical model and factors of improvements should be
reviewed and revised as appropriate.

References

1. “EPRI BWR Water Chemistry Guidelines – 2000 Revision,” TR-103515-R2, Final Report,
February 2000.

2. Loomis, Larry, “Resin Intrusion Response at Pilgrim Nuclear Power Station,” EPRI
PWR/BWR Plant Chemistry Meeting, San Antonio, Texas, January 10-12, 2001.

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13
CONCLUSIONS

BWR Iron/Chemistry Monitoring Database

The wide range of EPRI chemistry and radiation dose evaluations that draw on the EPRI BWR
Chemistry Monitoring Database confirms the usefulness and value of its monitoring efforts to
the BWR industry. The database is also frequently used to address individual plant requests on
specific issues, and for use in benchmarking efforts and assessments.

In addition, periodic electronic reports have been introduced to provide regular feedback to the
plants of the monitoring effort. The electronic reports are structured to disseminate information
in a concise format to assist in plant benchmarking efforts. The EPRI BWR BRAC Summary
Report, the first of these periodic electronic reports, provides feedback to the industry on a
regular biannual basis. This report has disseminated useful information for benchmarking and
has contributed to the standardization of drywell dose rate measurements. The EPRI BWR
Chemistry Sampling Frequency Report, which is updated and electronically transmitted
annually, has been put to use by corporate and plant chemistry managers for optimizing
resources and standardizing practices among plants. The EPRI BWR Condensate Demineralizer
Deep Bed Resins Report and the EPRI BWR Condensate Filter Demineralizer Precoat Materials
Report, which is also updated annually and transmitted electronically, is used by chemists,
systems engineers and operations personnel that have a stake in optimizing condensate polisher
operations. A fourth electronic report, targeted for the fourth quarter of 2002, will provide
concise performance summaries of condensate, feedwater and reactor water chemistry results.

Plant Chemistry Regime Status

As of August 2002, 31 of the 36 operating North American BWRs were injecting hydrogen into
the feedwater to mitigate IGSCC of the reactor recirculation piping and/or reactor vessel
internals. Twenty-three plants had performed NMCA and 22 of these were injecting hydrogen.
A total of 32 plants were adding DZO to the feedwater for drywell radiation field control.

Iron and Copper Control

Feedwater iron and copper control is highly dependent on the type of condensate polishing
system. Many plants that were originally designed with only deep beds for condensate polishing
have added or are adding non-precoat pre-filters upstream of the deep beds. Based on retrofits in
progress, there will be 16 plants by the end of 2002 with Filter + Deep Bed condensate polishing

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Conclusions

and 6 with Deep Bed Only condensate polishing. Most plants with Filter Demineralizers are
using a combination of precoatable pleated and conventional septa to control iron, although two
plants are using all pleated septa and two others (with copper alloy condenser tubes) use no
pleated septa. Plants that continue to operate with Deep Bed Only condensate polishers are using
a combination of resin selection, improved resin cleaning systems or methods, and resin
management practices to realize the best achievable iron control within the limitations of this
design.

North American BWRs have improved feedwater iron control from an industry average of 2.54
ppb in 1997 to 1.38 ppb in 2001. The 2001 results show that the average feedwater iron values
of all plants were below the 5 ppb EPRI Action Level 1 value. Of the 33 plants for which 2001
data were available, 32 had feedwater iron averages within the 0.5 – 3.0 ppb desired range, no
plant had averages greater than 3 ppb and 2 had averages less than 0.5 ppb. The two plants with
less than 0.5 ppb iron had Filters + Deep Bed condensate polishing with no iron addition. Iron is
being injected into the feedwater at six BWR units, either in the oxide or the oxalate form, to
maintain feedwater iron above the 0.5 ppb threshold concentration. Below this threshold, some
plants have experienced an increase in dose rates at hot spots.

These evaluations have identified condensate temperature as having a significant impact on

feedwater iron control for Deep Bed Only plants. Calendar quarter averages for Deep Bed Only
plants at >90% power show that feedwater iron in the second and third quarter is significantly
lower than the first and fourth quarter. The first quarter average feedwater iron for Deep Bed
Only plants is 42% higher than the third quarter average feedwater iron.

Hotwell iron varies widely among plants, from about 6 ppb to >30 ppb. The use of moisture
separator reheaters appears to influence hotwell iron. The average hotwell iron for plants
designed with reheat is about 14 ppb, compared with about 20 ppb for those plants which have
no reheat.

For plants with copper alloy condenser tubes, average feedwater copper concentrations in the
range of 0.01 ppb are achievable by plants with Filter + Deep Bed condensate polishing. Deep
Bed Only plants with copper alloy condenser tubes have demonstrated feedwater copper control
in the 0.1 – 0.2 ppb range. Plants with copper alloy condenser tubes and Filter Demineralizer
condensate polishing have been unable to routinely control feedwater copper below the EPRI
Action Level 1 value of 0.2 ppb.

Two U.S. BWRs and one foreign BWR are controlling the feedwater Fe/Zn ratio above 2 to
avoid the potential for localized corrosion of the Zircaloy fuel cladding in the region of the
spacers. Several plants have feedwater Fe/Zn, Fe/(Zn+Ni) and/or Fe/(Zn+Ni+Cu) ratios that are
lower than 2. Additional guidance is needed on the importance of maintaining these ratios for
minimizing the probability of fuel corrosion.

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Conclusions

Transient Corrosion Products

Many plants have adopted the startup/hot standby feedwater and reactor water diagnostic
parameters for iron prior to initiating significant feedwater flow. Plants normally meet the
startup suggested limit of <100 ppb for feedwater iron when sampling at the conclusion of
feedwater flush operations prior to startup. This measurement confirms the removal of bulk
loose crud from condensate and feedwater piping and equipment, reducing the mass of crud that
would otherwise be transported to the reactor. Consequently, the diagnostic parameter for
insoluble iron prior to the initiation of significant feedwater flow given in Table 4-4 of the BWR
Water Chemistry Guidelines – 2000 Revision at the startup/hot standby condition should be
revised to a control parameter with a limit of 100 ppb. At the same time, feedwater flush
practices may need to be optimized to maximize their effectiveness. The current diagnostic
parameter for reactor water insoluble iron should remain a diagnostic parameter, although
available data indicate that this measurement alone is not adequate to quantify the cumulative
amount or rate of metals deposition on fuel during the startup transient. Additional detailed
startup data are needed to develop and support appropriate revisions.

Feedwater/condensate metals samples taken during feedwater flush operations are not indicative
of feedwater metals concentrations during the startup transient. During startup, velocities in the
condensate and feedwater systems are changing due to the flow demand from power ascension,
feedwater flow changes to maintain coolant level in the reactor, and the starting of pumps. Also,
changes in temperature and in the dissolved oxygen concentration influence the transport of
corrosion products. These changes can cause significant releases of corrosion products from
feedwater system surfaces. This is confirmed by detailed feedwater and reactor water data taken
during power ascension by two plants that show significant variations in deposition rates.
Evaluation of available detailed plant startup metals data indicate significantly higher deposition
rates for iron (2 – 18 times) and somewhat higher for copper (1.1 – 1.75 times) than during
steady full power operating conditions. The limited data show that the peak iron deposition rate
occurred between about 80% and 100% power and the peak copper deposition rate occurred
between 45% and 100% power.

Reactor coolant chemistry excursions during a refueling outage, the subsequent startup power
ascension or early in the fuel cycle (within the first 2900 MWD/MT) are potential initiating
events that can increase the vulnerability to crud-induced fuel failures later in the fuel cycle.
Copper appears to play a significant role in fuel failures related to metals transport, potentially
accelerating fuel cladding corrosion early in fuel life, possibly by lowering the thermal
conductivity of the deposit. While past fuel failures attributed strictly to high iron crud
deposition are limited to Tsuruga in the 1970s, and these had high crud deposition (exceeding
2
1000 mg/dm ), these failures occurred under NWC conditions. The iron deposition threshold
above which fuel integrity can be affected under today’s HWC and NMCA conditions, along
with zinc injection and with higher powered cores, is not known.

Most plants do not perform reactor coolant and feedwater metals analyses during power
ascension for several reasons. These include questions on whether the sample is representative
and the results are quantifiable, the impact of power ascension samples taken at >10% power on

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Conclusions

the ability to meet station Chemistry Indicator goals and the chemistry technician burden of extra
sampling and sample preparation. The latter is particularly true for stations that do not use XRF
but must digest corrosion products samples for analysis by an AA or ICP analytical technique.
Despite these reasons, plants should be encouraged to perform concurrent reactor water and
feedwater sampling for soluble and insoluble metals during power ascension from early startup
through 100% power as part of the chemistry startup plan. Having these samples allows an
evaluation to be performed later as needed, should fuel issues that could be related to chemistry
transients and crud deposition emerge during the fuel cycle. While it is preferable that startup
feedwater and reactor water metals samples be analyzed soon after they are obtained, plants
could alternatively choose to preserve the samples for future analysis as needed.

Plants should develop an appropriate startup sampling strategy with EPRI assistance. The results
of startup sampling should be evaluated to quantify startup metals deposition and the influence of
factors such as power ascension rate, hold points, startup chemistry excursions and chemistry
regime. The results, based on data from a cross-section of plants having different condenser tube
materials (copper and non-copper), feedwater iron levels (low iron, high iron, iron addition),
chemistry regimes (HWC, HWC/NMCA), zinc levels and power densities, can be used to
develop further guidance on transient metals monitoring and control.

Drywell Radiation Dose Rates

The most recent mean and median BRAC dose rates for the BWR industry are 207 mR/hr and
163 mR/hr, respectively. The mean and median values are based on the post-decon dose rates
when a chemical decontamination was performed. The BRAC point dose rates are mainly
influenced by the amount of Co-60 incorporated in the piping corrosion film.

Previous industry data evaluations showed that the BRAC point dose rates correlate with the
level of soluble Co-60 in the reactor coolant (3). BRAC point dose rates were shown to increase
exponentially at soluble Co-60 concentrations above approximately 7.5E-5 to 1E-4 µCi/ml.
However, with the widespread changes in chemistry regimes, particularly the conversion to the
highly reducing chemistry environment with NMCA that causes a restructuring of primary
system crud (particularly the large inventory on the fuel), this correlation is less clear. The start
of hydrogen injection after NMCA coincides with chemistry transitions that tend to increase Co-
60 in the reactor coolant. Zinc injection is increased as a countermeasure to suppress the dose
impact, and the industry data show a decreasing trend in soluble Co-60 with increasing reactor
water zinc concentration. BRAC dose rates tend to decrease as the ratio of reactor water soluble
Co-60/reactor water zinc decreases. In addition, the industry data show a general trend of
decreasing BRAC dose rates as the ratio of reactor water zinc/feedwater iron increases.

The results also clearly show the influence of chemistry regime and zinc injection. The average
dose rate for HWC+Zn plants is much lower than for plants with HWC without zinc injection.
Similarly, NWC+Zn plants also have a lower average BRAC dose rate than HWC plants without
zinc injection. The average recontamination rate following a chemical decontamination of the
recirculation piping is lowest under HWC+Zn chemistry and the highest under HWC without

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Conclusions

zinc injection. Also, the sequence of implementation is important. Plants that injected zinc for a
significant period prior to starting HWC achieve lower dose rates than plants starting HWC
before injecting zinc. This observation also holds when comparing results from plants over the
same range of reactor water zinc concentrations. Since the extent of metal oxide restructuring
from ECP reduction in the primary system is greater under HWC-NMCA than under HWC, the
above findings also support, by analogy, EPRI recommendations to add sufficient zinc prior to
beginning HWC-NMCA operations.

An expanded set of BHD (reactor vessel bottom head drain) data show that on an overall
industry basis dose rates from this “hot spot” location correlate with the activity of insoluble Co-
60 in the reactor coolant. Generally, insoluble Co-60 concentrations of less than about 6E-5
µCi/ml correspond to BHD dose rates of 1R/hr or lower. As insoluble Co-60 approaches and
exceeds 1E-4 µCi/ml, BHD dose rates tend to exceed 10 R/hr.

The industry data still support the previous general finding that lower feedwater iron
concentrations correlate with lower soluble Co-60 levels in the reactor coolant. However,
insoluble Co-60 appears to go through a minimum at some plants, increasing again as feedwater
iron is reduced below about 0.5 ppb. Recent data from Hatch 1, at which feedwater iron dropped
below 0.5 ppb during extended operating periods, confirm that as iron decreased, soluble Co-60
decreased or remained approximately constant while insoluble Co-60 increased. As feedwater
iron decreased, reactor water soluble and insoluble Co-60 remained well below 1E-4 µCi/ml.
BRAC dose rates have continued to decrease at Hatch 1 and no increases in hot spot dose rates
have been detected. The plant is evaluating the need for iron addition.

The Hatch 1 experience, along with other industry data, suggests that under fully established
reducing conditions with NMCA/HWC, stabilized iron oxide spinels with adequate DZO
injection and low fuel crud inventory after several (at least three) cycles with relatively low (<1.5
ppb) feedwater iron, operating with <0.5 ppb feedwater iron may provide the benefits of lower
drywell radiation dose rates and lower zinc usage without increases in hot spots.

Plant Chemistry Impact of NMCA

The transient chemistry effects observed following the application of noble metals are mainly
due to the restructuring of primary system crud from the Fe2O3 form, the stable form under
oxidizing conditions, to the spinel form (Fe3O4), which is stable under reducing conditions. This
restructuring is similar to that which occurs in the transition from the oxidizing conditions under
NWC to the reducing conditions under HWC, except that conditions under NMCA with
hydrogen injection are more reducing (lower ECP) and can affect more of the fuel surface area
on which crud is deposited.
2
The initial noble metal loading immediately after application varied from 0.22 to 2.8 µg/cm . In
the first reapplication at Duane Arnold, the first plant to reapply NMCA, the noble metal loading
2
was 0.8 µg/cm . Sample coupons have been analyzed to track the decline with time of noble
metals loading on primary surfaces.

13-5
 EPRI Licensed Material

Conclusions

All plants showed an initial increase in MSLRM dose rates after NMCA with hydrogen
injection. This increase is attributed to an increase in N-16 in the steam caused by the reducing
effect of surfaces that have noble metal. These dose rates are lower than under moderate HWC
and show a decreasing trend following the initial peak.

Increases in offgas noble gas activity after NMCA are attributed to small amounts of uranium
from fuel that have been incorporated into crud deposits and are released in the post-NMCA crud
restructuring process. Duane Arnold experienced a larger increase in the offgas noble gas
activity release rate following the initial noble metals application than after reapplication.
Reactor water iodines tend to be suppressed following NMCA. Noble gas data should be used in
preference to iodine data for monitoring the presence of a fuel leak under post-NMCA operating
conditions.

The reactor water conductivity transient after NMCA is due to soluble metals, mainly iron, from
the crud restructuring process. Duane Arnold experienced a similar spike following the first
reapplication. The crud restructuring process is also the cause of the increase in both soluble and
insoluble Co-60 experienced by all plants that have transitioned to NMCA with hydrogen
injection. The impact of increased Co-60 on drywell shutdown dose rates is largely dependent
on whether sufficient zinc was maintained both before and after operation with NMCA.

The impact on drywell piping shutdown dose rates varied from plant to plant after operating for
the first cycle under NMCA with hydrogen injection. Duane Arnold, Hatch 1, Hatch 2, and
FitzPatrick measured decreases in BRAC average dose rates. Peach Bottom 2, Quad Cities 1,
and Nine Mile Point 1 experienced increases in BRAC average dose rates. The plants that
experienced lower BRAC point dose rates after a cycle under NMCA/HWC had well-established
HWC programs prior to NMCA and sufficient zinc injected prior to and following NMCA.
Recommendations for controlling drywell dose rates with NMCA are detailed in BWRVIP-92
(1); these recommendations emphasize the need to maintain 5 – 10 ppb reactor water zinc prior
to and following NMCA and to maintain HWC at high availability prior to NMCA.

Nine Mile Point 2 continues to have elevated specific conductance in the RWCU F/D effluent
after NMCA with hydrogen injection. This appears to be due to corrosion of carbon steel
materials downstream of the ion exchange precoat. An evaluation showed that noble metals
depositing on the carbon steel piping with low dissolved oxygen levels and hydrogen injection
causes iron to go into solution. A similar experience at a plant in Japan was likewise determined
to be corrosion of the carbon steel effluent piping under conditions of NMCA and hydrogen
injection.

Plant Chemistry Results vs. BWR Chemistry Guidelines Revisions

The percentage of plants with annual average feedwater dissolved oxygen concentrations greater
than the Action Level 1 minimum value of <30 ppb increased from 72% in 1997 to 97% in 2001.

13-6
 EPRI Licensed Material

Conclusions

In the EPRI BWR Water Chemistry Guidelines – 2000 Revision, Table 4-5b was added for
reactor water chloride and sulfate limits applicable to plants with HWC or HWC+NMCA at
power operating conditions (2). Of the 26 plants injecting hydrogen during the survey period, 17
reported that they have implemented the new Action Level values, while 9 plants reported that
they have not adopted the new values.

ECP is used to monitor IGSCC mitigation conditions (-230mV SHE) at 15 of the 26 units
reporting this information. All 26 of these units use one or more secondary parameters either
alone or in combination with ECP. The most commonly used secondary parameter is reactor
water dissolved oxygen. Other secondary parameters include main steam line radiation levels,
feedwater dissolved oxygen concentration, hydrogen injection rate, reactor water dissolved
hydrogen and H2/O2 molar ratio. Nineteen of these 26 units also reported using a radiolysis
model, with the model being benchmarked at 12 of these plants using plant specific data or sister
plant data.

Most plants responding have or are prepared to measure reactor water and feedwater insoluble
iron as diagnostic parameters during Startup/Hot Standby conditions. Most plants are
performing some form of feedwater flush prior to startup from a refueling outage. This serves as
the action to meet the EPRI recommendation that feedwater iron should be reduced to the
suggested target of <100 ppb prior to initiation of significant feedwater flow to the reactor or at
the completion of the feedwater flush. Eleven out of 14 plants sample feedwater/condensate iron
prior to initiation of significant feedwater flow to the reactor. Only two plants (Laguna Verde 1
and 2), which have a high iron source in the high pressure heater drains, had concerns about
meeting the 100 ppb target. The decision to perform a feedwater flush prior to startup from an
unplanned outage is dependent on outage duration. Ten out of thirteen plants (77%) sample
reactor water insoluble iron as recommended.

BWR Chemistry Monitoring Practices

Seventy-five percent of plants sample reactor water at least daily to determine anion
concentrations and 75% at least weekly for soluble and insoluble gamma isotopics. Ninety-four
percent (94%) of plants report feedwater zinc results at least weekly, with the predominant
frequency (83%) being three times per week. For most plants, the variability in reactor water
zinc measurements is lower than feedwater zinc variability. This is true even among stations
using the same analysis method for both reactor water and feedwater zinc. These variations do
not appear to be related to either the method of zinc injection or the analysis method. The
differences in variability may be related to actual process fluctuations or other factors such as
specific sampling and laboratory techniques (e.g., number of cation papers used, sample line
velocity, sample tubing length, digestion process, etc.), as well as to the fact that reactor water
concentrations are higher and easier to quantify.

Most plants to not meet the velocity criterion of 6 ft/sec in the main sample tubing run from the
feedwater sample tap to the feedwater sample station. An evaluation for one plant where the
velocity was low showed that the 6 ft/sec criterion was achievable by increasing the corrosion

13-7
 EPRI Licensed Material

Conclusions

products sampler bypass flow. An evaluation of industry data confirms that the variability in
feedwater soluble zinc and insoluble iron is reduced as the sample line velocity approaches 6
ft/sec.

Experience with Condensate Filters

The first use of pleated filter septa in a domestic BWR condensate non-precoat application was
at Perry in August 1991. Pleated filter septa were first used in a domestic BWR precoat
application at Hatch in January 1995. By 2002, the use of iron removal septa in North American
BWRs had grown to 62 vessels in precoat service and 84 vessels in non-precoat service. Iron
removal filter septa are now in use at twenty-six North American BWR stations in precoat and
non-precoat applications.

Most filter demineralizer plants have kept some yarn-wound septa in service to maintain
feedwater iron above the recommended lower limit of 0.5 ppb, while two plants use pleated septa
in all vessels and are still able to maintain the feedwater iron concentration above 0.5 ppb.

The mechanical integrity of the septa has improved. The problems with past failures at joints,
end fittings, protective cages and septum-to-tubesheet seals have been resolved. The occurrence
of some filter media failures has prompted changes in media construction to address this issue.
With the evolution of septum design and quality, septa useful lives of 3 to 4 years have been
achieved at many stations. At some plants, such as Susquehanna, an ineffective backwash has
been identified as the cause of shorter than average septa useful life.

The EPRI BWR Condensate Filter Users Group is driving the optimization of this technology.
The group is involved in a continuing effort to evaluate septum performance and to disseminate
information addressing industry issues.

Experience with Deep Bed Condensate Demineralizers

Plants with Deep Bed Only condensate polishing face a greater challenge to control feedwater
iron than plants with either Filter + Deep Bed or Filter Demineralizer systems. All of the U.S.
BWRs that operated with Deep Bed Only condensate polishing in 2001 have demonstrated the
capability to achieve feedwater iron in the upper portion of the desired range of 0.5 ppb to 3 ppb,
although consistent iron control within this range has not been achieved in all cases.

The number of Deep Bed Only BWR plants continues to decline, with only six plants projected
to have this condensate polishing design by the end of 2002. In addition, the two Dresden units
will be the only plants with partial (40%) condensate filtration in a headered configuration, and
will thus continue to rely on deep beds for iron filtration. Clinton also has three deep bed
condensate demineralizers that do not have slaved filters upstream, so these three resin beds
continue to require periodic resin cleaning.

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 EPRI Licensed Material

Conclusions

Only two Deep Bed Only plants now use low cross-linked resins to meet iron control objectives.
The low cross-linked resins are typically being replaced after about 12 months of service to
avoid sulfate control problems. The other Deep Bed Only plants have been able to meet plant-
specific objectives for feedwater iron control with cation resins having 10% or higher cross-
linkage along with optimized resin cleaning and resin management practices.

A significant issue for plants having deep bed condensate demineralizers has been the limited
ability to control reactor water sulfate, particularly in the summer months when condensate
temperatures reach peak values. While the EPRI Action Level 1 value of 5 ppb for reactor water
sulfate has typically not been exceeded, the 2 ppb threshold concentration used in the WANO
CPI (Chemistry Performance Indicator) calculation is being exceeded. To address this problem,
plants are applying higher cross-linked cation resin (more stable against oxidative
decomposition) and anion resin underlays. In addition, these plants tend to replace resin beds
sooner to control reactor water chemistry. One plant is using carboxylic cation resin in some
demineralizers to reduce the sulfate source.

Condensate Polishing With Ion Exchanger Bypass

Because of the increased condensate flows from BWR plant power uprates, two U.S. plants and
one foreign plant bypass the condensate polishing system with a portion of the condensate flow.
For these plants, retrofitting additional condensate polishing vessels was not feasible or cost
prohibitive.

At Brunswick, when seawater inleakage into the main condenser is very low (<1 gpd), filtering
100% of the condensate while bypassing about 15% around the deep bed demineralizers resulted
in decreased reactor water sulfate while reactor water chloride was maintained below 1 ppb. At
Dresden 2, using 40% condensate filtration and bypassing flow around the condensate
demineralizers for short durations to avoid the flow surge that could increase feedwater iron
when a bed is being removed from service for resin cleaning, no detrimental impact on reactor
water chemistry has been reported. At KKL with 2 of 5 condensate filter demineralizer vessels
equipped with non-precoat septa and the rest with precoated septa, reactor water chemistry has
improved while the consumption of precoat materials has been reduced.

In general, the feasibility of partial bypassing of the condensate polisher is highly dependent on
the plant-specific capability to maintain sufficiently low cooling water inleakage to the main
condenser to maintain acceptable reactor water quality. Plants with copper alloy condensers and
those with a significant cobalt source term upstream if the condensate polishing system should
carefully evaluate the impact of bypassing.

References

1. “BWRVIP-92: NMCA Experience Report and Application Guidelines,” TR-1003022, Final

Report, September 2001.

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 EPRI Licensed Material

Conclusions

2. “EPRI BWR Water Chemistry Guidelines – 2000 Revision,” TR-103515-R2, Final Report,
February 2000.

3. “Iron Performance Monitoring – Database Maintenance and Evaluation of BWR Field

Experience with New Technology,” TR-1003157, Interim Report, December 2001.

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 EPRI Licensed Material

A
BROWNS FERRY 2

Table A-1
Browns Ferry 2 Plant Design Parameters

Parameter Value
Commercial Operation Date 3/75
Capacity (MWe) 1152
BWR Type 4
Drains Path Cascaded
Condensate Polishing Filter Demineralizer
RWCU Capacity (% Feedwater Flow), Normal/Maximum 0.9/0.9

Browns Ferry 2 Milestones

Milestone events for Browns Ferry 2 are given in Table A-2.

Browns Ferry 2 was in an extended shutdown from the mid-1980s to the early 1990s. The
condenser was re-tubed and a chemical decontamination of the recirculation piping was
performed in 1991. The recirculation piping safe ends and risers were also replaced. A second
decontamination was performed in 1993. DZO addition began in 10/97 and hydrogen injection
started in 12/99. A 5% power uprate was implemented in 1999. Browns Ferry 2 performed
another chemical decontamination of the recirculation system piping in 3/01 which removed 78
curies. NMCA was implemented in 3/01. Hydrogen flow was lowered from 80 scfm to 12 scfm
following NMCA implementation.

Use of pleated filters began in 1997 and was extended to all nine F/D vessels in 2000

A-1
 EPRI Licensed Material

Browns Ferry 2

Table A-2
Browns Ferry 2 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 4/99

Condenser Retube 1991

Recirc. Pipe
1991
Replacement

RWCU Pipe
X
Replacement

Extraction Steam
Pipe Replacement

Chem. Decon. 1991 X 3/01

→ → →
HWC (scfm) 12/99
(80) (12) (12)

NMCA 3/01

NZO

DZO 10/97 → → → → →

Iron Injection

Crud Resins

Pleated Filters 3/97 → → → → →

A-2
 EPRI Licensed Material

Browns Ferry 2

Radiation Data

Recirculation System dose rates are summarized in Table A-3.

Table A-3
Browns Ferry 2 Recirculation System Dose Rates

Browns Ferry 2 – Recirculati on System Dose Rates (mR/hr)

May-91 Oct-91 Feb-92 Sep-92 Jan-93 May-93 Oct-94 Mar-96 Sep-97

(1) (2)

EFPY

BRAC 55 300 349 34 360 296 425

A Suction 50 120 220 300 400 40 400 350 500

B Suction 50 300 350 90 500 400 400

A Discharge 25 300 325 4 250 225 400

B Discharge 95 300 320 3 290 210 400

Avg Risers

Browns Ferry 2 – Recirculati on System Dose Rates (mR/hr)

Apr-99 Mar-01 Mar-01

(1) (2)

EFPY

BRAC 315 375 4

A Suction 360 450 5

B Suction 380 400 9

A Discharge 270 350 1

B Discharge 250 300 1

Avg Risers
Table A-3 Notes
1. Pre decon
2. Post decon

A-3
 EPRI Licensed Material

Browns Ferry 2

Trend Data

Power, feedwater iron, reactor water anions, reactor water cobalt-60 and BRAC history trend
plots for Browns Ferry 2 are presented in Figures A-1, A-2, A-3, A-4, and A-5 respectively.

100

80
Power (%)

60

40

20

0
5/2/95 6/5/96 7/10/97 8/14/98 9/18/99 10/22/00 11/26/01 12/31/02

Power NMCA

Figure A-1
Power History, Browns Ferry 2

Feedwater Iron Control

Feedwater insoluble iron at Browns Ferry 2 has averaged between 1.0 and 1.5 ppb since 1998.
The feedwater average insoluble iron was 1.2 ppb for 2001. Pleated filter septa were first
installed in 1997, and were installed in all F/D vessels as of 12/00.

Reactor Water Sulfate Control

Browns Ferry 2 experienced increased average reactor water sulfate in 2001, as shown in Figure
A-3. The annual average increased from 1.0 ppb in 2000 to 2.2 ppb in 2001. The 2002 average
through mid-July was over 3 ppb. Data evaluation by the EPRI BWR Condensate Filter Users
Group indicated that a possible cause of the increased reactor water sulfate could be effects
related to pleated filter septa age. Average pleated septa age at Browns Ferry 2 was 1.0 years in
December 2000 and 1.7 years in December 2001.

A-4
 EPRI Licensed Material

Browns Ferry 2

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001
5/2/95 6/5/96 7/10/97 8/14/98 9/18/99 10/22/00 11/26/01 12/31/02

Insoluble Fe Soluble Fe NMCA

Figure A-2
Feedwater Iron, Browns Ferry 2

5.0 100

4.5 90

4.0 80
Reactor Water Anions (ppb)

3.5 70

3.0 60 Power (%)

2.5 50

2.0 40

1.5 30

1.0 20

0.5 10

0.0 0
1/1/00 7/19/00 2/4/01 8/23/01 3/11/02 9/27/02

Cl NO3 SO4 Power

Figure A-3
Reactor Water Anions, Browns Ferry 2

A-5
 EPRI Licensed Material

Browns Ferry 2

Reactor Water Co-60 (µCi/ml) 1.E-02

1.E-03

1.E-04

1.E-05

1.E-06
5/2/95 6/5/96 7/10/97 8/14/98 9/18/99 10/22/00 11/26/01 12/31/02

Insoluble Co-60 Soluble Co-60 NMCA

Figure A-4
Reactor Water Cobalt-60, Browns Ferry 2

500
NMCA
Retube Condenser, replace RWCU pipe,
replace Recirc pipe safe ends & risers,

Replace CRB pins/rollers, chem decon

400
Dose Rate (mR/hr)

HWC
Chem Decon (recirc piping)

300
Pleated Filters

200
DZO
Power Uprate

Chem Decon

100

0
Jan-90 Sep-92 Jun-95 Mar-98 Dec-00

BRAC Milestones

Figure A-5
BRAC History, Browns Ferry 2

A-6
 EPRI Licensed Material

Browns Ferry 2

Recirculation Piping Dose Rates

The historical BRAC data for Browns Ferry 2 show dose rates in the high range among U.S.
BWRs, averaging over 300 mR/hr from 1995 through 3/01. These dose rates are also high in
comparison to those of Browns Ferry 3, where the last reported value was 160 mR/hr in April
2002. The major difference between the two units is that Unit 3 began injection of depleted zinc
oxide in 1995, while Unit 2 implemented DZO in 10/97.

A chemical decontamination performed prior to NMCA reduced the BRAC dose rate from 375
to 4 mR/hr.

Reactor water soluble Co-60 has increased from a low of 9.0E-5 µCi/ml in 1998 (the year after
zinc injection was started) to 1.30 E-4 µCi/ml in 2001 under HWC. Insoluble Co-60 was 1.68E-
4 µCi/ml in 2001, an increase from the 2000 average of 2.0E-5 µCi/ml prior to NMCA.

Fuel Failures

Browns Ferry 2 experienced fuel failures in early 2002. Investigation into the cause of the
failures is ongoing. Reactor water iodine and offgas sum of six noble gases trend data are
presented in Figures A-6 and A-7, respectively.

1.E-02
Reactor Water Iodines (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
5/2/95 6/5/96 7/10/97 8/14/98 9/18/99 10/22/00 11/26/01 12/31/02

I131 I132 I133 I134 I135 NMCA

Figure A-6
Reactor Water Iodines, Browns Ferry 2

A-7
 EPRI Licensed Material

Browns Ferry 2

12000

10000
Offgas Sum of 6 (µCi/sec)

8000

6000

4000

2000

0
5/2/95 6/5/96 7/10/97 8/14/98 9/18/99 10/22/00 11/26/01 12/31/02

Sum of 6 NMCA

Figure A-7
Sum of 6 Noble Gases, Browns Ferry 2

A-8
 EPRI Licensed Material

B
BROWNS FERRY 3

Table B-1
Browns Ferry 3 Plant Design Parameters

Parameter Value

Commercial Operation Date 3/77

Capacity (MWe) 1152

BWR Type 4

Drains Path Cascaded

Condensate Polishing Filter Demineralizer

RWCU Capacity (% Feedwater Flow), Normal/Maximum 0.9/0.9

Browns Ferry 3 Milestones

Milestone events for Browns Ferry 3 are given in Table B-2.

The station was shutdown for an extended period from the mid-1980s to the mid-1990s. The
condenser was retubed, a chemical decontamination of the recirculation system was performed
and the recirculation piping ring header, safe ends, and risers were replaced in 1995 prior to
startup. DZO addition was started in 12/95. NMCA was performed in 4/00 and HWC was
started in 8/00.

Pleated filter use began in 1996 and was extended to all nine F/D vessels in 2001.

B-1
 EPRI Licensed Material

Browns Ferry 3

Table B-2
Browns Ferry 3 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1998 2000 2001 2002
1993

Power Uprate 10/98

Condenser
X
Retube

Recirc. Pipe
X
Replacement

RWCU Pipe
X
Replacement

Extraction Steam
Pipe Replacement

Chem. Decon. 1/92 X

8/00 → →
HWC (scfm)
(12) (12) (12)

NMCA 4/00

NZO

DZO 12/95 → → → → → → →

Iron Injection

Crud Resins

Pleated Filters 11/96 → → → → → →

B-2
 EPRI Licensed Material

Browns Ferry 3

Radiation Data

Recirculation System dose rates are summarized in Table B-3:

Table B-3
Browns Ferry 3 Recirculation System Dose Rates

May-96 Sep-96 Feb-97 Sep-98 Apr-00 Mar-02

EFPY

BRAC 29 41 70 98 (1) 138 (1) 160

A Suction 40 50 90 140 170 170

B Suction 25 60 80 110 140 170

A Discharge 50 30 60 700 (2) 700 (2) 180

B Discharge 30 25 50 70 120 120

Avg Risers
Table B-3 Notes
1. B discharge used twice to calculate average.
2. External contamination on pipe.

Trend Data

Power, feedwater iron, reactor water anions, reactor water cobalt-60 and BRAC history trend
plots for Browns Ferry 3 are presented in Figures B-1, B-2, B-3, B-4, and B-5 respectively.

B-3
 EPRI Licensed Material

Browns Ferry 3

100

80
Power (%)

60

40

20

0
3/28/95 5/1/96 6/5/97 7/10/98 8/14/99 9/17/00 10/22/01 11/26/02 12/31/03

% Power NMCA

Figure B-1
Power History, Browns Ferry 3

Feedwater Iron Control

Feedwater insoluble iron at Browns Ferry 3 averaged 1.04 ppb in 1999, 1.1 ppb in 2000, and 1.2
ppb in 2001.

Reactor Water Sulfate Control

Browns Ferry 3 experienced increased average reactor water sulfate in from the end of 2000
through July 2002 as shown in Figure A-3. The average increased from 0.74 ppb in December
2000 to 1.2 ppb in December 2001. The 2002 average through mid-July is over 2 ppb. Data
evaluation by the EPRI BWR Condensate Filter Users Group indicated that a possible cause of
the increased reactor water sulfate could be effects related to pleated filter septa age. Average
pleated septa age at Browns Ferry 3 was 0.9 years in December 2000 and 1.9 years in December
2001.

B-4
 EPRI Licensed Material

Browns Ferry 3

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001
3/28/95 5/1/96 6/5/97 7/10/98 8/14/99 9/17/00 10/22/01 11/26/02 12/31/03

Insoluble Fe Soluble Fe NMCA

Figure B-2
Feedwater Iron, Browns Ferry 3

5.0 100

4.5 90

4.0 80
Reactor Water Anions (ppb)

3.5 70

3.0 60 Power (%)

2.5 50

2.0 40

1.5 30

1.0 20

0.5 10

0.0 0
1/1/00 7/19/00 2/4/01 8/23/01 3/11/02 9/27/02

Cl NO3 SO4 Power

Figure B-3
Reactor Water Anions, Browns Ferry 3

B-5
 EPRI Licensed Material

Browns Ferry 3

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
3/28/95 5/1/96 6/5/97 7/10/98 8/14/99 9/17/00 10/22/01 11/26/02 12/31/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure B-4
Reactor Water Cobalt-60, Browns Ferry 3

400
DZO
Pleated Filters
Condenser retube; partial recirc pipe repl;

300
RWCU pipe repl; chem decon
Dose Rate (mR/hr)

NMCA
HWC

200
Power Uprate
Chem decon

100

0
Jun-90 Feb-93 Nov-95 Aug-98 May-01

BRAC Milestones

Figure B-5
BRAC History, Browns Ferry 3

B-6
 EPRI Licensed Material

Browns Ferry 3

Recirculation Piping Dose Rates

Although the BRAC average dose rate Browns Ferry 3 increased to138 mR/hr in 4/00, it remains
significantly lower than the pre-decon dose rate at Unit 2. The most significant difference
between the two units is the earlier start of DZO injection at Unit 3 in 1995; zinc injection was
started in 1997 at Unit 2.

Average reactor water soluble Co-60 activity was 1.61E-4 µCi/ml in 2000 and 1.30E-4 µCi/ml in
2001. Average reactor water insoluble Co-60 was 1.89E-4 µCi/ml in 2000 and 3.29E-4 µCi/ml
in 2001.

B-7
 EPRI Licensed Material

C
BRUNSWICK 1

Table C-1
Brunswick 1 Plant Design Parameters

Parameter Value

Commercial Operation Date 1976

Capacity (MWe) 895
BWR Type 4
Drains Path Forward Pumped
Condensate Polishing Filter + Deep Bed
RWCU Capacity (% Feedwater Flow), Normal/Maximum 0.88/0.88

Brunswick 1 Milestones

Milestone events for Brunswick 1 are given in Table C-2.

The original copper/nickel condenser tubes were replaced with titanium in 1983. Partial
replacements of extraction steam piping with Chrome/Moly occurred in 1983 and 1987 and are
ongoing. The recirculating pipe safe ends and external risers were replaced in 1991. An
extended shutdown began in April 1992; the plant was restarted in January 1994. A chemical
decontamination of the recirculation piping in 1993 removed 31.3 curies; another chemical
decontamination in 1995 removed 186.6 curies. DZO injection was started in May of 1995. A
5% power uprate was implemented in 1996. A 15% power uprate to 958 MWe is currently being
implemented.

C-1
 EPRI Licensed Material

Brunswick 1

Table C-2
Brunswick 1 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power X X
Uprate
Condenser 1983
Retube

Recirc. Pipe 1991

Replacement (1)
RWCU Pipe
Replacement
Extraction
1983,
Steam Pipe
1987
Replacement

Chem. 1987, X X
Decon. 1990
1990 → →
HWC (scfm) (10- → → → → → → → →
(35) (39.6)
18)
NMCA

NZO

DZO 5/95 → → → → → → →

Iron Injection

Crud Resins

Pleated 8/94 → → → → → → → →
Filters
Table C-2 Notes
1. Safe ends and risers

C-2
 EPRI Licensed Material

Brunswick 1

Radiation Data

Recirculation System dose rates are summarized in Table C-3:

Table C-3
Brunswick 1 Recirculation System Dose Rates

Feb- Mar- Oct- Nov- May- Apr- Apr-93 Apr- Apr- Jan-
87 (1) 87 (2) 90(1) 90 (2) 92 93 (1) (2) 95 (1) 95 (2) 96

EFPY

BRAC 212 61.7 343.8 82.5 216.3 177.5 23.8 1525 693.8 292.5

A Suction 300 85 250 200 20 1500 350 300

B Suction 160 60 400 120 225 150 10 1900 1200 700

A Discharge 225 80 300 45 195 110 15 1400 575 70

B Discharge 250 45 375 80 195 250 50 1300 650 100

Avg Risers 185 22.2 687.5 40.5 90 69 4.8 1112.5 80 142.5

Mar- Oct- Nov- May- Feb- Mar- Mar-

Sep-02
96 96 97 98 00 00 02

EFPY

BRAC 295 417.5 550 556 450 475 362.5 331.25

A Suction 280 220 300 450 500 400 350 325

B Suction 700 750 800 950 450 700 500 300

A Discharge 80 400 350 400 400 400 300 400

B Discharge 120 300 750 425 450 400 300 300

Avg Risers 170 218.75 237.5 325 320 355

Table C-3 Notes
1. Pre-decon
2. Post-decon

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Brunswick 1
are presented in Figures C-1, C-2, C-3 and C-4, respectively.

C-3
 EPRI Licensed Material

Brunswick 1

100

80
Power (%)

60

40

20

0
9/30/94 2/12/96 6/26/97 11/8/98 3/22/00 8/4/01 12/17/02

Figure C-1
Power History, Brunswick 1

10
Feedwater Fe (ppb)

0.1

0.01
9/30/94 2/12/96 6/26/97 11/8/98 3/22/00 8/4/01 12/17/02
Insoluble Fe Soluble Fe

Figure C-2
Feedwater Iron, Brunswick 1

C-4
 EPRI Licensed Material

Brunswick 1

1.E-03
Reactor Water Co-60 (µCi/ml)

1.E-04

1.E-05

1.E-06
9/30/94 2/12/96 6/26/97 11/8/98 3/22/00 8/4/01 12/17/02

Insoluble Co-60 Soluble Co-60

Figure C-3
Reactor Water Cobalt-60, Brunswick 1

2500 HWC 10-18 scfm 35 scfm 39.6 scfm

2000 Extended Pleated Filters

Shutdown DZO
Dose Rate (mR/hr)

Chem decon; partial extraction

extraction steam pipe replacement

steam pipe replacement

1500
Partial recirc pipe replacement
Retube condenser; partial

(safe ends and risers)

1000
Chem decon

Chem decon

Power Uprate
Chem decon

500

0
Jun-82 Feb-85 Nov-87 Aug-90 May-93 Feb-96 Nov-98 Jul-01

BRAC Milestones Co-60 Dose

Figure C-4
BRAC History, Brunswick 1

C-5
 EPRI Licensed Material

Brunswick 1

Feedwater Iron Control

Brunswick’s original plant design included filter demineralizers followed by deep beds for
condensate polishing. The filter demineralizers were retrofitted with non-precoat pleated filter
septa in 1994. Feedwater iron concentrations at Unit 1 are routinely <0.5 ppb, which are among
the lowest in the industry. Unit 1 had an average soluble feedwater iron concentration in 2001 of
about 0.106 ppb, which continues to be among the higher values reported by BWRs. The
average insoluble feedwater iron in 2001 was 0.389 ppb. The ratio of soluble to insoluble
feedwater iron is also high compared to other plants.

Recirculation Piping Dose Rates

The most significant increases in drywell dose rates were clearly related to hydrogen injection,
and the largest increase followed operation at moderate HWC conditions of 1 ppm H2 in the
feedwater. Frequent cycling of hydrogen injection may also have been a contributing factor.

Dose rates tended to stabilize at approximately 200 mR/hr without hydrogen injection. Addition
of depleted zinc oxide was implemented after a chemical decontamination in 1995. The unit also
had a power uprate and increased the hydrogen injection rate since 1996. The most recent
surveys indicate a slowly decreasing trend in BRAC dose rates.

Reactor water soluble Co-60 typically exceeds 1.0E-4 µCi/ml, and the average for 2001 was
1.83E-4 µCi/ml, which is in the upper range reported by North American BWRs. Soluble Co-60
appears to trend in cycles, increasing as the feedwater zinc concentration decreases from about
0.3 ppb to 0.1 ppb. With increased reactor water and feedwater zinc concentrations after the
2002 refuel outage, the average reactor water soluble Co-60 concentration has decreased to
8.32E-5 µCi/ml.

Recirculation Piping Gamma Scans

Gamma scan data for Brunswick 1 are summarized in Table C-4.

A gamma scan of the recirculation pipe in 1990, before the chemical decontamination was
performed, indicated the corrosion film had approximately 18 µCi/cm2 of total activity with
about 80% of the activity due to Co-60. The March 2000 gamma scan had 47.4 µCi/cm2 total
activity with almost 43% due to Cr-51incorporation in the corrosion film due to HWC
conditions.

C-6
 EPRI Licensed Material

Brunswick 1

Table C-4
Brunswick 1 Recirculation Piping Gamma Scan Results

Oct-90 Nov-90
Mar-00
(1) (2)

Total Activity
18 4.85 47.4
(µCi/cm2)

% Co-60 78.8 81.2 52.5

% Co-58 15.4 11.3 3.5

% Mn-54 5.8 7.6

% Zn-65 1.2

%Cr-51 42.8
Table C-4 Notes
1. Pre-decon
2. Post-decon

Fuel Failures

The station entered a forced shutdown in 9/02 to perform fuel sipping to identify leaking fuel
assemblies. Three leaking fuel bundles were identified, all of which were initially installed in
the 3/02 refuel outage. The station had also experienced fuel failures in 1999, as the offgas
release rate (sum of six) prior to the 3/00 refuel outage was about 15,000 µCi/sec.

C-7
 EPRI Licensed Material

D
BRUNSWICK 2

Table D-1
Brunswick 2 Plant Design Parameters

Parameter Value

Commercial Operation Date 1975

Capacity (MWe) 895
BWR Type 4
Drains Path Forward Pumped
Condensate Polishing Filter + Deep Bed
RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1

Brunswick 2 Milestones

Milestone events for Brunswick 2 are given in Table D-2.

The original copper/nickel condenser tubes were replaced with titanium in 1984. The
recirculation piping external risers and safe ends were replaced in 1989. Partial extraction steam
piping replacement with chrome/moly occurred in 1984 and 1988 and continues on an ongoing
basis. An extended shutdown began in April 1992 and continued until restart in May 1993.

A chemical decontamination of the recirculation piping in 1991 removed 76 curies; the 1994
chemical decontamination removed 25.3 curies; the decon in 1996 removed 281 curies. DZO
injection was started in March of 1996. Pleated filters were installed beginning in January 1995.
A 5% power uprate was implemented in 1997. Implementation of a 15% power uprate to 951
MWe is planned for 2003.

Radiation Data

Recirculation System dose rates are summarized in Table D-3.

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Brunswick 2
are presented in Figures D-1, D-2, D-3 and D-4, respectively.

D-1
 EPRI Licensed Material

Brunswick 2

Table D-2
Brunswick 2 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate X

Condenser
1984
Retube

Recirc. Pipe 1988

Replacement (1)

RWCU Pipe
Replacement

Extraction Steam
1984,
Pipe
1988
Replacement

1984,
1985,
Chem. Decon. X 2/96
1989,
1991

1989 → → →
HWC (scfm) → → (39.6) → → → → →
(10-15) (20) (35)

NMCA

NZO

DZO 3/96 → → → → → →

Iron Injection

Crud Resins

Pleated Filters 1/95 → → → → → → →

Table D-2 Notes
1. Safe ends and risers

D-2
 EPRI Licensed Material

Brunswick 2

Table D-3
Brunswick 2 Recirculation System Dose Rates

Brunswick 2 – Recirculation System Dose Rates (mR/hr)

Dec-85 (1) Jan-86 (2) Apr-87 Apr-88 Sep-89 Oct-89 (2) Aug-90 Jan-91 Apr-91
(1)
(1)

EFPY

BRAC 350 82.5 172.5 142.8 247.5 58.8 468.8 243.8 906.3

A Suction 190 75 450 275 700

B Suction 150 45 425 200 700

A Discharge 325 95 155 150 150 60 500 300 825

B Discharge 375 70 190 135 500 55 500 200 1400

Avg Risers 295.5 30.9 170 144 201 7.1 942.5 990 2020

Brunswick 2 – Recirculation System Dose Rates (mR/hr)

Sep-91 (2) Oct-91 May-92 Mar-94 (1) Mar-94 (2) Feb-96 (1) Feb-96 (2) Mar-96 Sep-97

EFPY

BRAC 275 14.5 60 981.3 612.5 2175 38.8 10.3 325

A Suction 300 20 90 975 600 2200 40 17 350

B Suction 350 18 80 1100 600 2500 65 6 350

A Discharge 250 10 35 950 675 2000 20 10 300

B Discharge 200 10 35 900 575 2000 30 8 300

Avg Risers 740 625.7 535.5 1652.5 1912.5 1283 31 19.8 350

Brunswick 2 – Recirculation System Dose Rates (mR/hr)

Jun-98 Apr-99 Apr-99 May-99 Sep-00 Feb-01 Apr-02

EFPY

BRAC 362.5 437.5 400 375 317 331 375

A Suction 450 450 450 450 350 350 400

B Suction 350 350 425 350 375 350

A Discharge 300 300 375 350 300 300 400

B Discharge 350 350 350 350 300 300 350

Avg Risers 350 410 379.2 325 328 385

Table D-3 Notes
1. Pre-decon
2. Post-decon

D-3
 EPRI Licensed Material

Brunswick 2

100

80
Power (%)

60

40

20

0
9/30/94 2/12/96 6/26/97 11/8/98 3/22/00 8/4/01 12/17/02

Figure D-1
Power History, Brunswick 2

10
Feedwater Fe (ppb)

0.1

0.01
9/30/94 2/12/96 6/26/97 11/8/98 3/22/00 8/4/01 12/17/02

Insoluble Fe Soluble Fe

Figure D-2
Feedwater Iron, Brunswick 2

D-4
 EPRI Licensed Material

Brunswick 2

Reactor Water Co-60 (µCi/ml) 1.E-02

1.E-03

1.E-04

1.E-05

1.E-06
9/30/94 2/12/96 6/26/97 11/8/98 3/22/00 8/4/01 12/17/02

Insoluble Co-60 Soluble Co-60

Figure D-3
Reactor Water Cobalt-60, Brunswick 2

2500 HWC 10 scfm 20 scfm 35 scfm 39.6 scfm

12-15 scfm
Chem decon, retube condenser & partial

2000
Partial extraction steam pipe repl.
extraction steam pipe replacement

Extended SD
Dose Rate(mR/hr)

Pleated filters
1500
recirc pipe replacement
Chem decon, partial

DZO
Chem Decon
Chem decon

Power uprate

1000
Chem Decon
Chem Decon

500

0
Mar-83 Nov-85 Aug-88 May-91 Feb-94 Nov-96 Aug-99 Apr-02
BRAC Milestones Co-60 Dose

Figure D-4
BRAC History, Brunswick 2

D-5
 EPRI Licensed Material

Brunswick 2

Feedwater Iron Control

Feedwater insoluble iron concentrations at Unit 2 are typically less than those in Unit 1. The
average total feedwater iron concentration in 2001 was 0.443 ppb, which is the lowest among
North American BWRs reporting data. Soluble iron contributes about one third of the total
feedwater iron. The Brunswick 2 feedwater iron represents the achievable minimum possible
with pleated prefilters plus deep beds in the condensate polishing system and a low iron
contribution from the forward pumped drains.

Recirculation Piping Dose Rates

Brunswick 2 continues to have high average reactor water soluble Co-60 concentrations,
averaging 2.38E-4 µCi/ml in 2000 and 2.31E-4 µCi/ml in 2001. The industry data indicate that
those plants with higher reactor water soluble Co-60 tend to have higher recirculation pipe dose
rates. Soluble Co-60 appears to trend in cycles, increasing as the feedwater zinc concentration
decreases from about 0.4 ppb to 0.1 ppb. Since increasing feedwater and reactor water zinc
concentrations, the average soluble Co-60 concentration in 2002 through mid-September has
decreased to 1.63E-4 µCi/ml.

Prior to hydrogen injection, dose rates tended to stabilize at less than 200 mR/hr. The most
significant increases in drywell dose rates were clearly related to hydrogen injection, and the
largest increase followed operation at moderate HWC conditions of 1 ppm H2 in the feedwater.
Frequent cycling of hydrogen injection may also have been a contributing factor.

The plant started depleted zinc oxide addition after a chemical decontamination in 1996. The
next dose rate survey, after operation with zinc addition and hydrogen injection, indicated a
BRAC value of approximately 300 mR/hr. Although this was higher than dose rates measured at
Brunswick 1 after operating with both zinc addition and hydrogen injection, dose rates at Unit 2
appear to have stabilized at current conditions at 300 to 400 mR/hr.

Recirculation Piping Gamma Scans

Gamma scan data for Brunswick 2 are shown in Table D-4.

A gamma scan of the recirculation pipe before the chemical decontamination in 1991 indicated
that the corrosion film had approximately 23 µCi/cm2 total activity on the suction piping and 11
µCi/cm2 total activity on the discharge piping. Co-60 accounted for between 65 and 80% of the
total activity while Co-58 accounted for an additional 20% of the activity. The 2001 gamma
scan taken after years of HWC operation shows that over 30% of the activity is due to Cr-51.
The total activity increased to 28.3 µCi/cm2, which is similar to the Unit 1 data, although the
Unit 2 increase is not as pronounced as that in Unit 1.

D-6
 EPRI Licensed Material

Brunswick 2

Table D-4
Brunswick 2 Recirculation Piping Gamma Scan Results

1991 1991
Date 2001
(1) (2)

Total Activity
17.6 0.70 28.3
(µCi/cm2)

% Co-60 70.1 84 58.6

% Co-58 19.8 20 7.6

% Mn-54 7.4

% Zn-65 1.5

% Cr-51 32.2
Table D-4 Notes
1. Pre-decon
2. Post-decon

Stellite Reduction

Brunswick 2 has replaced some Stellite-containing valves and control rod blades.
Approximately 4.2% of the initial Stellite surface area of 330 square feet has been replaced
with alternative materials.

D-7
 EPRI Licensed Material

E
CLINTON

Table E-1
Clinton Plant Design Parameters

Parameter Value
Commercial Operation Date 4/87
Capacity (MWe) 980
BWR Type 6
Drains Path Cascaded
Condensate Polishing Partial Filter + Deep Bed
RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1

Clinton Milestones

Milestone events for Clinton are given in Table E-2.

The plant performed a chemical decontamination of the recirculation and reactor water cleanup
piping in 10/93. Pleated septa were installed in three filter vessels slaved to three of the nine
demineralizer vessels in 7/95; between 6/00 and 5/01, three more filter vessels were added for a
total of six. DZO injection was started 12/00, and HWC in 6/02. NMCA was performed in 4/02.

Radiation Data

Recirculation System dose rates are summarized in Table E-3.

E-1
 EPRI Licensed Material

Clinton

Table E-2
Clinton Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate

Condenser
Retube

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction
Steam Pipe
Replacement

Chem. Decon. 10/93

HWC (scfm) 6/02

NMCA 4/02

NZO

DZO 12/00 → →

Iron Injection

Crud Resins

Pleated Filters 7/95 → → → → → → →

E-2
 EPRI Licensed Material

Clinton

Table E-3
Clinton Recirculation System Dose Rates

Clinton – Recirculation S ystem Dose Rates (mR/hr)

Oct-88 Apr-90 Oct-91 Oct-93 Oct-94 Oct-96 Sep-98 Oct-00 Apr-02

EFPY

BRAC 243 478 370 387 304 330 207 237 232

A Suction 406 323 333 243 255 272

B Suction 496 393 407 257 209 271

A Discharge 316 256 325 164 285 190

B Discharge 331 242 256 165 197 193

Avg Risers

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Clinton are
presented in Figures E-1, E-2, E-3 and E-4, respectively.

Feedwater Iron Control

Feedwater iron is controlled using non-precoated pleated condensate pre-filters piped directly to
six of nine deep bed condensate demineralizers. The station currently does not have plans to
install filters upstream of the remaining three demineralizers, so it is expected to remain a Partial
Filter + Deep Bed plant. With the six filter vessels in service, and with one of the 3 remaining
deep beds out of service, the station will have a system with 75% Filter + Deep Bed and 25%
Deep Bed Only condensate polishing.

Data after 4/99 show that feedwater iron was in the 1 – 2 ppb range after startup from an
extended outage. Insoluble iron values then increased to between 2 and 4 ppb and finally
decreased to 1 – 2 ppb after 4/00. The insoluble feedwater iron annual average concentration
was 2.52 ppb in 2000 and decreased to 1.34 in 2001 after addition of three additional filter
vessels between 7/00 and 5/01.

E-3
 EPRI Licensed Material

Clinton

100

80
Power (%)

60

40

20

0
4/1/99 10/18/99 5/5/00 11/21/00 6/9/01 12/26/01 7/14/02

Figure E-1
Power History, Clinton

100

10
Feedwater Fe (ppb)

0.1

0.01

0.001
4/1/99 10/18/99 5/5/00 11/21/00 6/9/01 12/26/01 7/14/02
Insoluble Fe Soluble Fe

Figure E-2
Feedwater Iron, Clinton

E-4
 EPRI Licensed Material

Clinton

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
4/1/99 10/18/99 5/5/00 11/21/00 6/9/01 12/26/01 7/14/02

Insoluble Co-60 Soluble Co-60

Figure E-3
Reactor Water Cobalt-60, Clinton

500
Pleated filters

400
Extended shutdown
Dose Rate (mR/hr)

300
DZO

200
NMCA
Chem decon

100
HWC

0
Oct-88 Oct-90 Sep-92 Sep-94 Sep-96 Sep-98 Sep-00 Sep-02

BRAC Milestones Co-60 Dose

Figure E-4
BRAC History, Clinton

E-5
 EPRI Licensed Material

Clinton

Recirculation Piping Dose Rates

The BRAC dose rate of 207 mR/hr reported in 9/98 was taken during the extended plant
shutdown and therefore includes some decrease due to isotopic decay. In 10/00, the BRAC dose
rate was 236 mR/hr, representative of operation prior to initiating DZO in 12/00. Prior to
NMCA in 4/02, the BRAC dose rate had remained constant at 232 mR/hr.

The average insoluble Co-60 was 3.47E-5 µCi/ml in 1999 and 6.50E-5 µCi/ml in 2000. The
average soluble Co-60 was 7.02E-5 µCi/ml in 1999 and 7.97E-5 µCi/ml in 2000. Limited data
available for 2001 indicates a soluble Co-60 average of 3.68E-5 µCi/ml and an insoluble Co-60
average of 8.91E-6 µCi/ml. These values are in the low range among North American BWRs.

Recirculation Piping Gamma Scans

Gamma scan data for Clinton are summarized in table E-4.

The plant has reported data from two pipe gamma scans. The results show a small increase in
total pipe activity, with Co-60 being the predominant isotope.
Table E-4
Clinton Recirculation Piping Gamma Scan Results

Date RF-5 Oct-96

Total Activity
19.4 21.0
(µCi/cm2)

% Co-60 51 51

% Co-58 9 9

% Mn-54 35 35

% Fe-59 5 5

E-6
 EPRI Licensed Material

F
COOPER NUCLEAR

Table F-1
Cooper Plant Design Parameters

Parameter Value
Commercial Operation Date 7/74
Capacity (MWe) 801
BWR Type 4
Drains Path Cascaded
Condensate Polishing Filter Demineralizer
RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1

Cooper Milestones

Milestone events for Cooper are given in Table F-2.

The recirculation pipe was replaced during 1984 – 1985. The original material was 304 stainless
steel, while the new material is seamless, electropolished and passivated 316L. DZO addition
began 12/99 and NMCA was performed 3/00. HWC is planned for 2002.

F-1
 EPRI Licensed Material

Cooper Nuclear

Table F-2
Cooper Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate

Condenser Retube

Recirc. Pipe 1984-

Replacement 1985

RWCU Pipe
1/87
Replacement

Extraction Steam
1/88
Pipe Replacement

11/84
Chem. Decon.
3/90

HWC (scfm)

NMCA 3/00 → →

NZO

DZO 12/99 → → →

Iron Injection

Crud Resins

Pleated Filters 12/97 → → → → →

Radiation Data

Recirculation System dose rates are summarized in Table F-3:

F-2
 EPRI Licensed Material

Cooper Nuclear

Table F-3
Cooper Recirculation Dose Rates

Cooper – Recirculation S ystem Dose Rates (mR/hr)

Oct-85 Oct-86 Mar-88 Apr-89 Oct-91 Mar-93 Mar-97 Dec-98 Mar-00 Nov-01

EFPY 0.13 0.98 2.04 2.91 4.81 5.97 8.02 9.35

BRAC 9 70 95 95 160 175 230 171 213 151

A Suction 8 60 80 70 155 180 225 195 180 165

B Suction 6 60 90 95 165 170 200 165 220 155

A Discharge 17 85 90 110 160 200 300 190 225 155

B Discharge 9 110 105 160 155 200 135 220 130

Avg Risers 200 290 178 305 350 340 341

Trend Data
Power, feedwater iron, reactor water cobalt-60, and BRAC history trend plots for Cooper are
presented in Figures F-1, F-2, F-3 and F-4, respectively.

100

80
Power (%)

60

40

20

0
10/28/95 3/11/97 7/24/98 12/6/99 4/19/01 9/1/02

Power NMCA

Figure F-1
Power History, Cooper Nuclear Station

F-3
 EPRI Licensed Material

Cooper Nuclear

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001
10/28/95 3/11/97 7/24/98 12/6/99 4/19/01 9/1/02

Insoluble Fe Soluble Fe NMCA

Figure F-2
Feedwater Iron, Cooper Nuclear Station

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
10/28/95 3/11/97 7/24/98 12/6/99 4/19/01 9/1/02

Insoluble Co-60 Soluble Co-60 NMCA

Figure F-3
Reactor Water Cobalt-60, Cooper Nuclear Station

F-4
 EPRI Licensed Material

Cooper Nuclear

300

250 Oxygen inj to condensate

Extraction steam pipe replacement

Recirc pipe replacement 1984-85
Dose Rate (mR/hr)

RWCU pipe replacement 1987

200

(ongoing) - Late 80s

150
Pleated filters

100
Chem decon

Chem decon
DZO

50 NMCA

0
Jan-84 Sep-86 Jun-89 Mar-92 Dec-94 Sep-97 Jun-00 Mar-03

BRAC Milestones Co-60 Dose

Figure F-4
BRAC History, Cooper Nuclear Station

Feedwater Iron Control

The feedwater iron during 1996 - 1998 was in the range of approximately 0.7 to 2.5 ppb. Feedwater
iron averaged 1.3 ppb in 1998, down from 2.2 ppb in 1997, after installing pleated filter septa in
the condensate filter demineralizer system. Feedwater iron remains low, averaging 1.18 in 2001.

Recirculation Piping Dose Rates

Since 1984 when the 304SS recirculation piping was replaced with 316L SS, the piping dose
rates steadily increased to a peak of 230 mR/hr in 3/97. The 12/98 survey showed a decrease in
the BRAC dose rate to 171 mR/hr. However the BRAC average dose rate increased to 213
mR/hr in 3/00 after initiating DZO in 12/99. NMCA was performed in 3/00, and the BRAC dose
rate subsequently decreased to 151 mR/hr in 11/01.

Recirculation Piping Gamma Scans

Gamma scan data for the recirculation piping are summarized in Table F-4.

Gamma scans reported since 1993 show similar total activity and isotopic mixes with both Co-60
and Mn-54 being the major constituents.

F-5
 EPRI Licensed Material

Cooper Nuclear

Table F-4
Cooper Nuclear Station Recirculation Piping Gamma Scan Results

Oct- Oct- Mar- Apr- Oct- Mar- Mar- Oct- Mar- Nov-
85 86 88 89 91 93 97 98 00 01

Total Activity
0.43 7.4 8.4 9.0 12.7 13.6 13.8 13.4 11.5 10.2
(µCi/cm2)

% Co-60 71 48 51 52 44 46 53 54 60 62

% Co-58 15 28 14 13 9 6 4 4 3 3

% Mn-54 15 24 35 35 44 49 38 37 30 30

% Fe-59 5 4 8 6

F-6
 EPRI Licensed Material

G
DRESDEN 2

Table G-1
Dresden 2 Plant Design Parameters

Parameter Value
Commercial Operation Date 6/70
Capacity (MWe) 912
BWR Type 3
Drains Path Cascaded
Condensate Polishing Partial Filter + Deep Beds
RWCU Capacity (% Feedwater Flow), Normal/Maximum 3/3

Dresden 2 Milestones

Milestone events for Dresden 2 are given in Table G-2. The RWCU piping was replaced in 1997
with 316L stainless steel; the original material was 304 stainless steel. The extraction steam
piping was replaced with chrome/moly. DZO injection was started 12/96. An ARCS (Advanced
Resin Cleaning System) was installed in 12/98. NMCA was performed in 10/99. Retrofit of a
partial filtration system (40% of full power flow) upstream of the condensate demineralizers was
completed in December 2001. A 17% power uprate was implemented in 12/01.

G-1
 EPRI Licensed Material

Dresden 2

Table G-2
Dresden 2 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 12/01

Condenser
Retube

Recirc. Pipe
Replacement

RWCU Pipe
5/97
Replacement

Extraction Steam
Pipe Yes
Replacement

84,86,
Chem. Decon. 2/93 7/95
88,90

→
→ → (8- →
HWC (scfm) 4/83 → → → → → → (10-
(40) 10) (13)
12)

NMCA 10/99

NZO

DZO 12/96 → → → → → →

Iron Injection

Crud Resins 12/95 → → → → → → →

Pleated Filters 12/01 →

Oxygen Injection 10/96 → → → → → →

G-2
 EPRI Licensed Material

Dresden 2

Radiation Data

Recirculation System dose rates are summarized in Table G-3.

Table G-3
Dresden 2 Recirculation System Dose Rates

Dresden 2 – Recirculation System Dose Rates (mR/hr)

Oct-84 Apr-85 Nov-86 May-87 Oct-88 Feb-89 Sep-90 Feb-91

Nov-74 Jan-83
(1) (2) (1) (2) (1) (2) (1) (2)

EFPY
BRAC 131 251 256 57 193 34 169 26 328 16.5

A Suction
B Suction
A Discharge
B Discharge
Avg Risers

Dresden 2 – Recirculation System Dose Rates (mR/hr)

Jan-93 Feb-93
Apr-98 Sep-99 Oct-01
(1) (2)

EFPY
BRAC 350 105 70 60 65

A Suction 60
B Suction 50
A Discharge 70
B Discharge 80
Avg Risers
Table G-3 Notes
1. Pre-decon
2. Post-decon

Trend Data

Power history, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Dresden
2 are presented in Figures G-1, G-2, G-3 and G-4, respectively.

G-3
 EPRI Licensed Material

Dresden 2

110
100
90
80
70
Power (%)

60
50
40
30
20
10
0
10/28/95 3/11/97 7/24/98 12/6/99 4/19/01 9/1/02

NMCA

Figure G-1
Power History, Dresden 2

100

10
Feedwater Fe (ppb)

0.1

0.01
10/28/95 3/11/97 7/24/98 12/6/99 4/19/01 9/1/02

Insoluble Fe Soluble Fe NMCA

Figure G-2
Feedwater Iron, Dresden 2

G-4
 EPRI Licensed Material

Dresden 2

1.E+00

1.E-01
Reactor Water Co-60 (µCi/ml)

1.E-02

1.E-03

1.E-04

1.E-05

1.E-06
10/28/95 3/11/97 7/24/98 12/6/99 4/19/01 9/1/02

Insoluble Co-60 Soluble Co-60 Total Co-60 NMCA

Figure G-3
Reactor Water Cobalt-60, Dresden 2

450 HWC
400 Crud resins Pleated Filters
350
Oxygen inj.
Dose Rate (mR/hr)

300
DZO
250
RWCU pipe replacement

ARCS
200 NMCA
150
Power Uprate
Chem decon

Chem decon

Chem decon

Chem decon

Chem decon

Chem decon

100
50
0
11/1/74 10/31/78 10/30/82 10/29/86 10/28/90 10/27/94 10/26/98 10/25/02

BRAC Milestones Contact Dose Rate

Figure G-4
BRAC History, Dresden 2

G-5
 EPRI Licensed Material

Dresden 2

Feedwater Iron Control

Dresden 2 has discontinued the use of low crosslinked resins. Average insoluble iron was 2.83
ppb in 2001, but has decreased to 1.79 for the first half of 2002, after the installation of partial
condensate prefilters.

Recirculation Piping Dose Rates

Dresden 2 is currently injecting both hydrogen and depleted zinc oxide. Injection of DZO started
in December 1996. Six chemical decontaminations of recirculation piping were performed from
1984 through July 1995. The BRAC dose rate was 60 mR/hr in September 1999, about 3 years
after DZO injection was started and just before NMCA in October 1999. The BRAC dose rate in
October 2001, the first refuel outage after NMCA, increased slightly to 65 mR/hr. Data from
1997 and 1998 indicated average soluble Co-60 at Dresden 2 was high compared to the rest of
the industry at about 3E-4 µCi/ml. Recent results are lower, with an average soluble Co-60 of
2.00E-4 µCi/ml in 2001. Average insoluble Co-60 was also high in 1997 at 7.3E-4 µCi/ml and
6.78E-4 µCi/ml in 2001. Both soluble and insoluble Co-60 increased following NMCA in
October 1999, but recent data indicate a downward trend in soluble Co-60.

G-6
 EPRI Licensed Material

H
DRESDEN 3

Table H-1
Dresden 3 Plant Design Parameters

Parameter Value
Commercial Operation Date 11/71
Capacity (MWe) 833
BWR Type 3
Drains Path Cascaded
Condensate Polishing Deep Beds
RWCU Capacity (% Feedwater Flow), Normal/Maximum 3/3

Dresden 3 Milestones

Milestone events for Dresden 3 are given in Table H-2.

The recirculation piping was replaced in 1985 with 316NG stainless steel; the original material
was 304 stainless steel. The RWCU piping was replaced in 1997 with 316L stainless steel; the
original material was 304 stainless steel. The extraction steam piping was replaced with
chrome/moly. DZO injection was started in 7/98. An ARCS (Advanced Resin Cleaning
System) was installed in 12/98 and NMCA was performed in 9/00. Retrofit of a partial filtration
system (40% of full power flow) upstream of the condensate demineralizers was in progress in
2002 to support increased condensate flow due to a planned power uprate.

H-1
 EPRI Licensed Material

Dresden 3

Table H-2
Dresden 3 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate

Condenser Retube

Recirc. Pipe 1985

Replacement

RWCU Pipe 6/97

Replacement

Extraction Steam X
Pipe Replacement Date?

Chem. Decon. 1983 4/94

HWC (scfm) 3/99 → → →

(45) (45) (13) (13)

NMCA 9/00

NZO

DZO 7/98 → → → →

Iron Injection

Crud Resins 6/97 → → → →

Pleated Filters

Oxygen Injection 10/96 → → → → → →

Radiation Data

Recirculation system dose rates are summarized in Table H-3.

H-2
 EPRI Licensed Material

Dresden 3

Table H-3
Dresden 3 Recirculation System Dose Rates

Dresden 3 – Recirculation System Dose Rates (mR/hr)

Sep-83 Mar-84 Oct-85 Sep-86 Mar-94 Apr-94

Mar-88 Jan-91 May-98 Oct-00
(1) (2) (1) (2) (1) (2)

EFPY

BRAC 313 14 329 0 78 90 146 22 120 100

A Suction

B Suction

A Discharge

B Discharge

Avg Risers
Table H-3 Notes
1. Pre-decon
2. Post-decon

Trend Data
Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Dresden 3 are
presented in Figures H-1, H-2, H-3 and H-4, respectively.

Feedwater Iron Control

Some low cross-linked resin beds were installed in 1997 at Dresden 3 to control feedwater iron,
but have been discontinued as of late 2001. Average insoluble iron was 1.80 ppb in 2000 and
2.52 ppb in 2001. Dresden 3 is in the process of installing condensate pre-filters which will
process 40% of the total condensate flow.

Recirculation Piping Dose Rates

Dresden 3 started DZO injection in 7/98 and hydrogen injection in 3/99. A difference between
the two Dresden units is that the recirculation piping at Unit 3 is 316 SS while that at Unit 2 is
304 SS. One chemical decontamination has been performed since replacing the recirculation
system piping in 1985. A dose rate of 120 mR/hr was reported in May 1998, before the start of
DZO or HWC. NMCA was performed in September 2000 and the BRAC measurement in
October 2000 was 100 mR/hr. Average reactor water soluble Co-60 was 9.20E-5 µCi/ml in 2000
and 1.24E-4 µCi/ml in 2001. The average insoluble Co-60 was 1.74E-4 µCi/ml in 2000 and
1.24E-4 µCi/ml in 2001.

H-3
 EPRI Licensed Material

Dresden 3

110
100
90
80
70
Power (%)

60
50
40
30
20
10
0
10/28/95 3/11/97 7/24/98 12/6/99 4/19/01 9/1/02

NMCA

Figure H-1
Power History, Dresden 3

100

10
Feedwater Fe (ppb)

0.1

0.01
10/28/95 3/11/97 7/24/98 12/6/99 4/19/01 9/1/02

Insoluble Fe Soluble Fe NMCA

Figure H-2
Feedwater Iron, Dresden 3

H-4
 EPRI Licensed Material

Dresden 3

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
10/28/95 3/11/97 7/24/98 12/6/99 4/19/01 9/1/02

Insoluble Co-60 Soluble Co-60 Total Co-60 NMCA

Figure H-3
Reactor Water Cobalt 60, Dresden 3

450
400
Oxygen inj
350
Recirc pipe replacement; chem decon

Crud resins
Dose Rate (mR/hr)

300 DZO
Chem decon

250
RWCU pipe replacement

ARCS
200
1985

HWC
150 NMCA
Chem decon

100
50
0
Jun-83 Feb-86 Nov-88 Aug-91 May-94 Feb-97 Nov-99 Jul-02
BRAC Milestones

Figure H-4
BRAC History, Dresden 3

H-5
 EPRI Licensed Material

I
DUANE ARNOLD

Table I-1
Duane Arnold Plant Design Milestones

Parameter Value
Commercial Operation Date 2/75
Capacity (MWe) 545
BWR Type 4
Drains Path Cascaded
Condensate Polishing Filter Demineralizer
RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1

Duane Arnold Milestones

Milestone events for Duane Arnold are given in Table I-2.

The plant performed a number of chemical decontaminations of the recirculation system piping.
The last reported chemical decontamination was in 1995. The plant started injecting DZO 4/94.
In 10/96, Duane Arnold was the first BWR to perform NMCA, and the only BWR to repeat
application, in 10/99. Duane Arnold began use of pleated septa in one of five filter demineralizer
vessels in 1995, discontinued pleated septa use in 11/97, and resumed pleated septa use in one of
five vessels in 3/99.

Radiation Data

Recirculation System dose rates are summarized in Table I-3.

Trend Data

Power, feedwater iron, reactor water cobalt-60, and BRAC history trend plots for Duane Arnold
are presented in Figures I-1, I-2, I-3 and I-4, respectively.

I-1
 EPRI Licensed Material

Duane Arnold

Table I-2
Duane Arnold Milestones

Duane Arnold

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 8/85

Condenser Retube

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction Steam
Pipe Replacement

7/90
Chem. Decon. 7/93 3/95
3/92

1987 → → →
HWC (scfm) → → → → → → →
(6) (9) (15) (6)

NMCA 10/96 10/99

NZO

DZO 12/94 → → → → → → → →

Iron Injection

Crud Resins

Pleated Filters 3/95 → → 3/99 → → →

I-2
 EPRI Licensed Material

Duane Arnold

Table I-3
Duane Arnold Recirculation System Dose Rates

Duane Arnold – Recirculation System Dose Rates (mR/hr)

Jun-80 Dec-82 Jan-85 Apr-87 Aug-88 Jun-90 Sep-90 Feb-92
(1) (2) (1)
EFPY 3.87 4.79 5.92 7.08 8.17 9.29 9.29 10.57
BRAC 563 506 681 431 530 613 293 1925
A Suction 800 475 1600
B Suction 400 90 1500
A Discharge 500 300 1600
B Discharge 1100 650 3000
Avg Risers 450 130

Duane Arnold – Recirculation System Dose Rates (mR/hr)

Apr-92 Jul-93 Oct-93 Mar-95 Apr-95 Nov-96 May-98 Nov-99
(2) (1) (2) (1) (2)
EFPY 10.57 11.71 11.71 12.95 12.95 14.13 15.62 16.92
BRAC 938 1438 838 1125 600 400 248 288
A Suction 450 1250 900 1000 500 180 130 130
B Suction 800 900 150 500 100 70 60 70
A Discharge 1600 2500 1800 1500 1000 700 300 500
B Discharge 900 1100 500 1500 800 350 500 450
Avg Risers

Duane Arnold – Recirculation System Dose Rates (mR/hr)

May-01
EFPY
BRAC 245
A Suction 60
B Suction 120
A Discharge 400
B Discharge 400
Avg Risers

Table I-3 Notes:

1. Pre decon
2. Post decon

I-3
 EPRI Licensed Material

Duane Arnold

110
100
90
80
70
Power (%)

60
50
40
30
20
10
0
1/1/93 5/16/94 9/28/95 2/9/97 6/24/98 11/6/99 3/20/01 8/2/02 12/15/03

NMCA

Figure I-1
Power History, Duane Arnold

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001
1/1/93 5/16/94 9/28/95 2/9/97 6/24/98 11/6/99 3/20/01 8/2/02 12/15/03

Insoluble Fe Soluble Fe NMCA

Figure I-2
Feedwater Iron, Duane Arnold

I-4
 EPRI Licensed Material

Duane Arnold

1.E-01
Reactor Water Co-60 (µCi/ml)

1.E-02

1.E-03

1.E-04

1.E-05

1.E-06
1/1/93 1/1/95 12/31/96 12/31/98 12/30/00 12/30/02

Insoluble Co-60 Soluble Co-60 NMCA

Figure I-3
Reactor Water Cobalt-60, Duane Arnold

2000
6 scfm 9 scfm 15 scfm 6 scfm
HWC
Pleated filters
1500
Dose Rate (mR/hr)

DZO

NMCA
1000
Power uprate

500
Chem decon

Chem decon

Chem decon
Chem decon

0
Jan-80 Sep-82 Jun-85 Mar-88 Dec-90 Sep-93 Jun-96 Mar-99 Nov-01

BRAC Milestones Co-60 Dose

Figure I-4
BRAC History, Duane Arnold

I-5
 EPRI Licensed Material

Duane Arnold

Feedwater Iron Control

With pleated filter septa installed in one of five filter demineralizer vessels since 3/99, feedwater
iron has remained within the EPRI recommended range. Feedwater average insoluble iron was
1.34 ppb in 2000 and 0.75 ppb in 2001.

Recirculation Piping Dose Rates

Prior to starting HWC in 1987, dose rates were relatively stable in the range of 500 to 600
mR/hr. Dose rates increased dramatically to 1900 mR/hr in 1992, the outage after a chemical
decontamination. The plant performed four more chemical decontaminations, the last of which
occurred in 3/95. This decontamination was performed after two months of DZO injection.
Since then, with continued DZO injection and after NMCA, dose rates appear to have decreased
and have stabilized between about 250 and 300 mR/hr.

Soluble and insoluble Co-60 at Duane Arnold each averaged less than 5E-5 µCi/ml for 1998,
1999 and 2000. Soluble Co-60 averaged 2.59E-5 µCi/ml in 2001, while the average for
insoluble Co-60 was 1.02E-4 µCi/ml.

Recirculation Piping Gamma Scans

Gamma scan data for Duane Arnold are summarized in Table I-4.

Piping gamma scans parallel the dose rate data and show decreasing total activity. Cr-51 is the
most abundant isotope at Duane Arnold and, except for the 1992 data, has been a major
contributor based on the reported data. Co-60 also remains a major constituent in the isotopic
mix.
Table I-4
Duane Arnold Recirculation Piping Gamma Scan Results

Jul-90 Jul-90 Mar-92 Apr-92 Oct-96 Apr-98

(1) (2) (1) (2)
Total Activity
55.8 42.5 53.1 25.3 30.6 6.8
(µCi/cm2)
% Co-60 40 41 87 94 9 32
% Co-58 12 13 11 5 2 3
% Mn-54 2 0 2 1 0 3
% Zn-65 2 2 3 13
% Cr-51 44 43 86 48
Table I-4 Notes
1. Pre decon
2. Post decon

I-6
 EPRI Licensed Material

J
FERMI 2

Table J-1
Fermi 2 Plant Design Parameters

Parameter Value

Commercial Operation Date 1/88

Capacity (MWe) 1150

BWR Type 4

Drains Path Forward Pumped

Condensate Polishing Filter Demineralizer

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1.1

Fermi 2 Milestones

Milestone events for Fermi 2 are given in Table J-2.

The plant electrical capacity was increased in January 1992 from 1093 MWe to 1139 MWe, and
again in January 1996 to 1150 MWe. There was an extended shutdown following the turbine
blade failure on 12/93 which resulted in a large ingress of main condenser cooling water into
plant systems.

The condenser was retubed with titanium in April 1991; the original material was admiralty
brass. Approximately 150 ft of extraction steam piping to the #5 heater was replaced in 10/92.
The original extraction steam piping material was carbon steel, and the new material is chrome-
moly.

DZO injection began in July 1995 and HWC was started in September 1997. Pleated filter septa
were first installed in January 1999.

Radiation Data

Recirculation System dose rates are summarized in Table J-3.

J-1
 EPRI Licensed Material

Fermi 2

Table J-2
Fermi 2 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 9/92 11/96

Condenser Retube 1991

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction Steam
10/92
Pipe Replacement

Chem. Decon.

HWC (scfm) 9/97 → → → → →

NMCA

NZO

DZO 7/95 → → → → → → →

Iron Injection

Crud Resins

Pleated Filters 1/99 → → →

J-2
 EPRI Licensed Material

Fermi 2

Table J-3
Fermi 2 Recirculation System Dose Rates

Fermi 2 – Recirculation S ystem Dose Rates (mR/hr)

Nov-88 Apr-91 Oct-92 Mar-94 Nov-96 Oct-97 Oct-98 Apr-00 Nov-01

EFPY

BRAC 43 91 133 163 128 125 125 131 125

A Suction 40 90 160 190 140 120 110 140 120

B Suction 25 95 50 60 40 80 100 65 90

A Discharge 46 90 120 200 150 150 150 160 150

B Discharge 60 90 200 200 180 150 140 160 140

Avg Risers 119 164 339 424 296 223

Trend Data

Power, feedwater iron, reactor water cobalt-60, and BRAC history trend plots for Fermi 2 are
presented in Figures J-1, J-2, J-3 and J-4, respectively.

Feedwater Iron Control

Fermi uses pleated septa in two condensate filter demineralizer vessels, with the main incentive
being reduced radwaste generation due to the thinner precoats required.. Average feedwater
insoluble iron was 0.68 ppb in 2000 and 0.67 ppb in 2001. Effluent iron from the condensate
filter demineralizers was 1.45 ppb in 2000 and 1.51 ppb in 2001 and is diluted in the final
feedwater by the low iron contribution, approximately 0.1 ppb, from the forward pumped drains.

J-3
 EPRI Licensed Material

Fermi 2

100

80
Power (%)

60

40

20

0
3/2/97 7/15/98 11/27/99 4/10/01 8/23/02

Figure J-1
Power History, Fermi 2

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001
3/2/97 7/15/98 11/27/99 4/10/01 8/23/02

Insoluble Fe Soluble Fe

Figure J-2
Feedwater Iron, Fermi 2

J-4
 EPRI Licensed Material

Fermi 2

Reactor Water Co-60 (µCi/ml) 1.E-03

1.E-04

1.E-05

1.E-06
3/2/97 7/15/98 11/27/99 4/10/01 8/23/02
Insoluble Co-60 Soluble Co-60

Figure J-3
Reactor Water Cobalt 60, Fermi 2

250
Extended DZO
shutdown

200
Replace LP turbine rotors; begin valve

HWC
Retube condenser; replace remaining

Extraction steam pipe replacement

component replacement 1994

Dose Rate (mR/hr)

150
Replace 20 CRBs

(partial) 1992
CRBs

100
Power uprate 1992

Power uprate 1996

Pleated filters

50

0
Jun-84 Feb-87 Nov-89 Aug-92 May-95 Feb-98 Nov-00
BRAC Milestones

Figure J-4
BRAC History, Fermi 2

J-5
 EPRI Licensed Material

Fermi 2

Recirculation Piping Dose Rates

Recirculation piping dose rates showed a gradual increasing trend from the start of plant
operation until 1994, when the dose rates peaked 162 mR/hr. Surveys taken after 1994 indicate
the BRAC dose rates have stabilized at approximately 125 mR/hr. Fermi started adding DZO to
the feedwater in 1995 and moderate hydrogen injection was implemented in late 1997.

Soluble Co-60 activity has remained <1E-4 µCi/ml, with averages of 5.51E-5 µCi/ml in 2000
and 9.05E-5 µCi/ml in 2001. Insoluble Co-60 activity is also low with an average of 4.41E-5
µCi/ml in 2000 and 3.29E-5 µCi/ml in 2001.

Stellite™ Reduction

Fermi 2 has made significant progress in cobalt materials reduction. Since 1988, all control rod
blades have been replaced with low-cobalt materials. The low pressure turbine rotors were
replaced in 1994. A program to replace Stellite™ valve internals with 400-series stainless steel
was begun in 1994, continuing through several outages.

J-6
 EPRI Licensed Material

K
FITZPATRICK

Table K-1
FitzPatrick Plant Design Parameters

Parameter Value

Commercial Operation Date 7/75

Capacity (MWe) 860

BWR Type 4

Drains Path Cascaded

Condensate Polishing Deep Bed

RWCU Capacity (% Feedwater Flow), Normal/Maximum 0.91/1.1

FitzPatrick Milestones

Milestone events for James A. FitzPatrick Nuclear Power Plant are given in Table K-2.

The main condenser was retubed in 1994 with admiralty brass (same as original material). The
impingement tubes and tubes adjacent to the air removal piping are titanium. These tubes were
initially 304 stainless steel but were replaced with titanium in early plant operation. Chemical
decontaminations of the RWCU and recirculation piping removed 63 curies in 1988, 49 curies in
1992, and 37 curies in 1994. A power uprate to 2536 MWth (860 MWe) was implemented in
01/97. NZO was started in 01/89, with the station switching to passive DZO in 04/96. NMCA
was performed in 11/99. Hydrogen injection was lowered from 18.5 scfm to 10.5 scfm in 12/99,
reduced again to 6 scfm in 04/00, but raised to 8 scfm in 07/01.

K-1
 EPRI Licensed Material

Fitzpatrick

Table K-2
FitzPatrick Milestones

FitzPatrick

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 1/97

Condenser Retube 1/94

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction Steam
Pipe Replacement

9/88
Chem. Decon. 12/94
5/92

8/88
(11)
HWC (scfm) → → 18.5 → → → 10.5 6.0 8.0 →
7/90
(13.5)

NMCA 11/99

NZO 1/89 → → →

DZO 4/96 → → → → → →

Iron Injection

Crud Resins

Pleated Filters

Radiation Data

Recirculation System dose rates are summarized in Table K-3.

K-2
 EPRI Licensed Material

Fitzpatrick

Table K-3
FitzPatrick Recirculation System Dose Rates

FitzPatrick – Recirculation System Dose Rates (mR/hr)

Sep- Oct- Sep- Apr- Mar- Jan- May- Mar- Nov- Apr-
88 (1) 88 (2) 89 90 91 92 (1) 92 (2) 93 93 94
EFPY 8.5 8.5 9.3 9.7 10.3 10.6 10.6 10.7 11.2 11.5
BRAC 168 43 116 116 120 113 10 34 90 113
A Suction 180 15 166 180 153 147 11 50 100 140
B Suction 150 75 89 95 114 108 13 30 90 100
A Discharge 190 25 94 89 106 96 8 30 70 90
B Discharge 150 55 101 108 102 7 25 100 120
Avg Risers 525 53 279 275 295 266 6 50 112 154

FitzPatrick – Recirculation System Dose Rates (mR/hr)

Jan- May- Feb- Sep- Oct- Nov- May- Dec- May- Aug-
95 95 96 96 96 96 97 97 98 98
EFPY 12.5 13 13.5 13.6 14 14.5 14.8
BRAC 15.5 31 65 96 85 78 68 92.5 87.5 108
A Suction A9 15 70 100 70 80 70 70
B Suction 17 20 90 100 85 110 90 140
A Discharge 14 45 0 85 70 90 80 110
B Discharge 12 45 100 100 85 90 110 110
Avg Risers 12 65 140 145 164 265 125 165 170

FitzPatrick – Recirculati on System Dose Rates (mR/hr)

Oct- Jul- Oct- Apr- Oct- Mar- Oct-
98 99 99 00 00 01 02
EFPY 15.2
BRAC 80 113 104 61 79 81 97.5
A Suction 70 70 85 40 80 80 90
B Suction 90 100 100 50 85 80 80
A Discharge 70 110 110 75 75 90 100
B Discharge 90 170 120 80 75 75 120
Avg Risers 172 155 90 164 190 225
Table K-3 Notes
1. Pre-decon
2. Post-decon

K-3
 EPRI Licensed Material

Fitzpatrick

Trend Data

Power, feedwater iron, reactor water Co-60 and BRAC history trend plots for FitzPatrick are
presented in Figures K-1, K-2, K-3 and K-4, respectively.

100

80
Percent Power

60

40

20

0
12/1/96 12/1/97 12/1/98 12/1/99 11/30/00 11/30/01 11/30/02 11/30/03

% Power NMCA

Figure K-1
Power History, FitzPatrick

Feedwater Iron Control

FitzPatrick remains one of the more successful plants with Deep Bed Only condensate polishing
in controlling feedwater iron using standard resins and URC for resin cleaning. The 2001
average feedwater insoluble iron concentration was 1.43 ppb, while the average for 2000 was
2.01 ppb. It is noted that, starting in 1999, each iron data point is the average from the two
feedwater trains.

In 2002, Fitzpatrick began the use of 16% cross-linked gel cation resin in all condensate
demineralizers to improve sulfate control.

K-4
 EPRI Licensed Material

Fitzpatrick

JAF

1000

100
Feedwater Fe (ppb)

10

0.1

0.01

0.001

0.0001
12/1/96 12/1/97 12/1/98 12/1/99 11/30/00 11/30/01 11/30/02 11/30/03

Insoluble Fe Soluble Fe NMCA

Figure K-2
Feedwater Iron, FitzPatrick

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
12/1/96 12/1/97 12/1/98 12/1/99 11/30/00 11/30/01 11/30/02 11/30/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure K-3
Reactor Water Cobalt-60, FitzPatrick

K-5
 EPRI Licensed Material

Fitzpatrick

250 10.5 scfm

11 scfm 13.5 scfm 18.5 scfm 8 scfm
HWC
6 scfm
200
Begin cobalt material replacement 1985
NZO DZO

Chem Decon, Retube Condenser 1994

Dose Rate (mR/hr)

150
NMCA

100

Power uprate 1997

Chem Decon 1988

Chem Decon 1992

50

0
Jun-84 Feb-87 Nov-89 Aug-92 May-95 Feb-98 Nov-00
BRAC Milestones Co-60 Dose

Figure K-4
BRAC History, FitzPatrick

Recirculation Piping Dose Rates

Prior to implementation of NZO in 1989, the BRAC dose rate was 168 mr/hr. Two chemical
decontaminations were performed on the recirculation piping after starting NZO injection. Since
the last chemical decontamination in 1994, BRAC dose rates have typically remained less than
100 mR/hr.

The BRAC average dose rate in the first refuel outage (10/00) following the 1999 mid-cycle
NMCA outage was 78 mR/hr. A similar result was obtained in 3/01. The BRAC average dose
rate has risen to 97.5 as of the latest measurements taken in October 2002

Soluble Co-60, which had averaged approximately 3.0E-5 µCi/ml prior to NMCA, had almost
doubled in 2000 with an average of 5.39E-5 µCi/ml. The 2001 average of 4.33E-5 µCi/ml is
slightly lower than the 2000 average. Average insoluble Co-60 had also increased since NMCA
from 3.68E-5 µCi/ml in 1999 to 1.66E-4 µCi/ml in 2000. The 2001 average of 5.6E-5 µCi/ml is
also lower than the 2000 average.

Recirculation Piping Gamma Scan

Gamma scan data for FitzPatrick are summarized in Table K-4.

K-6
 EPRI Licensed Material

Fitzpatrick

Pre-chemical decontamination gamma scan results from 12/94 (five years after the
implementation of NZO and HWC) indicated the pipe corrosion film had about 11 µCi/cm2 total
activity with about 35% due to Co-60, about 25 % due to Mn-54 and 27% due to Zn-65. A
gamma scan performed in 11/96, six months after the start of DZO, showed a total activity of
about 4 µCi/cm2, with 39% of the activity due to Co-60, 28% from Mn-54, and 20% from Zn-65.
The contribution to total activity from Cr-51 became significant in 1998; in previous
measurements, Cr-51 was not detected. The Cr-51 activity contribution increased to 51 percent
of the total activity in 10/00. The activity of Co-60 in the piping film was essentially unchanged
in the 10/98 and 10/00 measurements as the Cr-51 activity increased. Preliminary results from
the 10/02 scan indicate that Cr-51 still remains the predominant contributor to activity at 55% of
the total. Co-60 activity has increased from 2.5 µCi/cm2 in 10/00 to 3.9 µCi/cm2 in 10/02, with
21% of the total activity attributed to Co-60.
Table K-4
FitzPatrick Recirculation Piping Gamma Scan Results

Sep- May- Jan- May- Dec- Jan- Nov-

89 90 92 (1) 92 (2) 94 (1) 95 (2) 96

Total Activity
22.4 17.8 11.1 0.5 10.8 0.73 4.11
(µCi/cm2)

% Co-60 18 21 49 40 35 55 39

% Co-58 9 8 2 6 7

% Mn-54 6 7 14 20 24 20 27

% Zn-65 64 63 35 40 27 24 20

%Cr-51 nd nd nd nd nd nd nd

Date 10/02
10/98 10/00
(3)

Total Activity
8.86 15.8 18.45
(µCi/cm2)

% Co-60 30 16 21.1

% Co-58 10 6 6.5

% Mn-54 30 8 4.6

% Zn-65 8 17 11.4

% Cr-51 14 51 54.7
Table K-4 Notes
1. Pre-decon
2. Post-decon
3. Preliminary results

K-7
 EPRI Licensed Material

Fitzpatrick

Stellite™ Reduction

As of 10/00, FitzPatrick has replaced 100 of 137 control rod blades with non-Stellite™ bearing
materials. The station has also has an ongoing program to evaluate valve replacements with non-
Stellite™ materials.

K-8
 EPRI Licensed Material

L
GRAND GULF

Table L-1
Grand Gulf Plant Design Parameters

Parameter Value

Commercial Operation Date 7/85

Capacity (MWe) 1250

BWR Type 6

Drains Path Forward Pumped

Condensate Polishing Deep Beds*

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1%

*Filters used only for plant startups

Grand Gulf Milestones

Milestone events for Grand Gulf are given in Table L-2.

Chemical decontaminations of the recirculation system piping were performed in 1992 and 1995.
The 1995 chemical decontamination removed 24.1 curies. A new resin cleaning process, the
ARCS (Advanced Resin Cleaning System), was installed in 1995 as a replacement for the
original URC (Ultrasonic Resin Cleaner). The ARCS was initially in use from 12/95 to 08/96, at
which time it was shut down for modification. ARCS operations resumed on 10/96. DZO
injection was initiated in 02/98 and HWC in 05/98.

L-1
 EPRI Licensed Material

Grand Gulf

Table L-2
Grand Gulf Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power
Uprate

Condenser
Retube

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction
Steam Pipe
Replacement

Chem.
5/92 4/95
Decon.

→
HWC (scfm) 5/99 → →
(55)

NMCA

NZO

DZO 2/98 → → → →

Iron Injection

Crud Resins 11/93 →

Pleated
Filters

L-2
 EPRI Licensed Material

Grand Gulf

Radiation Data

Recirculation System dose rates are summarized in Table L-3.

Table L-3
Grand Gulf Recirculation System Dose Rates

Grand Gulf – Recirculation System Dose Rates (mR/hr)

Oct- Dec- Apr- Nov- May- Nov- Apr- Nov- May- Dec- May- Sep-
86 87 89 90 92 93 95 96 98 99 01 02

EFPY

BRAC 145 175 235 220 310 425 418 380 308 340 293

A Suction 143 170 230 220 320 400 450 380 340 (1) 310

B Suction 154 180 260 240 320 500 420 440 340 360 340

A Discharge 153 180 230 200 300 300 400 330 280 330 250

B Discharge 129 170 220 220 300 500 400 370 270 310 270

Avg Risers 173 196 293 296 308 408 408 446 499 347 380 314
Table L-3 Notes
1. A suction not measured; B suction used twice in average.

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Grand Gulf are
presented in Figures L-1, L-2, L-3 and L-4, respectively.

L-3
 EPRI Licensed Material

Grand Gulf

100

80
Power (%)

60

40

20

0
1/1/94 1/1/96 12/31/97 12/31/99 12/30/01 12/30/03

Figure L-1
Power History, Grand Gulf

Feedwater Iron Control

Feedwater insoluble iron concentrations averaged 3.52 ppb in 2000 and 1.65 ppb in 2001.
Although Grand Gulf has precoatable condensate filters upstream of the deep bed condensate
polishers. These filters are used only during startup and outages to minimize crud transport; the
filters are not rated for full power operation. The precoat filters are used to reduce the amount of
crud loading on the deep bed resins during startups. Grand Gulf tried using low cross-linked
resins for enhanced crud removal in 1993 and 1994, but these were removed due to unacceptable
reactor water chemistry. The plant now relies on standard resins and the new resin cleaning
system to control feedwater iron. Iron control improved in 2001 through efforts to optimize the
resin cleaning frequency.

L-4
 EPRI Licensed Material

Grand Gulf

100

10
Feedwater Fe (ppb)

0.1

0.01

0.001
1/1/94 1/1/96 12/31/97 12/31/99 12/30/01 12/30/03

Insoluble Soluble

Figure L-2
Feedwater Iron, Grand Gulf

1.E-01
Reactor Water Co-60 (µCi/ml)

1.E-02

1.E-03

1.E-04

1.E-05

1.E-06
1/1/94 1/1/96 12/31/97 12/31/99 12/30/01 12/30/03

Insoluble Co-60 Soluble Co-60

Figure L-3
Reactor Water Cobalt-60, Grand Gulf

L-5
 EPRI Licensed Material

Grand Gulf

450
400
350
Dose Rate (mR/hr)

300
250
200
Crud resins
150
ARCS
Chem decon

Chem decon
100
DZO
50
HWC
0
Oct-86 Jun-89 Mar-92 Dec-94 Sep-97 Jun-00

BRAC Milestones Co-60 Dose

Figure L-4
BRAC History, Grand Gulf

Recirculation Piping Dose Rates

BRAC dose rates following the 1995 recirculation piping chemical decontamination were on a
declining trend but increased after the first full fuel cycle of HWC operation. The most recent
measurement of 292.5 mR/hr in 9/02 is the lowest since the 1995 chemical decontamination, and
has decreased since the 12/99 value of 308 mR/hr (7 months after initiating HWC). Soluble
reactor coolant Co-60 averaged 6.7E-5 µCi/ml in 2000 and 6.04E-5 µCi/ml in 2001. Insoluble
reactor coolant Co-60 averaged 1.7E-5 µCi/ml in 2000 and 1.4E-5 µCi/ml in 2001. Soluble Co-
60 shows a decrease following the start of DZO injection.

Recirculation Piping Gamma Scans

Gamma scan data for Grand Gulf are summarized in Table L-4.

The gamma scan data indicate that the total activity of the corrosion film in 5/92 was
approximately 19 µCi/cm2, with 66% of the activity due to Co-60 and 29% due to Mn-54. The
total activity was reduced to about 5 µCi/cm2 by chemical decontamination, with about 96% of
the activity attributable to Co-60.

L-6
 EPRI Licensed Material

Grand Gulf

Table L-4
Grand Gulf Recirculation Piping Gamma Scan Results

Date May- Jun-92

92 (1) (2)

Total Activity
18.75 5.1
(µCi/cm2)

% Co-60 65.9 96.1

% Co-58 5.6

% Mn-54 28.5 3.9

% Zn-65

%Cr-51

Table L-4 Notes

1. Pre decon
2. Post decon (B discharge only)

L-7
 EPRI Licensed Material

M
HATCH 1

Table M-1
Hatch 1 Plant Design Parameters

Parameter Value

Commercial Operation Date 12/75

Capacity (MWe) 910

BWR Type 4

Drains Path Cascaded

Condensate Polishing Filter Demineralizer

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1

Hatch 1 Milestones

Milestone events for Hatch 1 are given in Table M-2.

HWC was implemented in 1987. The main condenser admiralty brass condenser tubes were
replaced with titanium in 6/90, followed by NZO implementation in 8/90 to compensate for the
loss of the natural zinc source. The station switched to DZO in 2/94. Power uprates were
implemented in 1996 and 1999. The current station electrical power output is 112% of the
original design. NMCA was implemented in 3/99 upon which the station lowered hydrogen flow
from 45 scfm (moderate hydrogen flow plant) to 6 scfm. Hydrogen flow was increased to 8 scfm
in 2000. The current hydrogen injection rate is 10 scfm. The NRC has approved operating
license extension to a new expiration date of August 2034.

The station has performed chemical decontaminations of the reactor recirculation piping and the
RWCU piping. The 10/91 decontamination of the RWCU piping removed 22.5 curies while the
4/96 decontamination of the RWCU piping removed 113 curies. The 4/96 decontamination of
the recirculation suction piping removed 72.9 curies.

M-1
 EPRI Licensed Material

Hatch 1

Table M-2
Hatch 1 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 4/96 4/99

Condenser Retube 6/90

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction Steam
Pipe Replacement

Chem. Decon. 10/91 3/96

HWC (scfm) 9/87 → → → → → → →

→ → →
(22) (30) (45) (50) (45) (6) (8) (10)

NMCA 3/99

NZO 8/90 →

DZO 2/94 → → → → → → → →

Iron Injection

Crud Resins

Pleated Filters 1/95 → → → → → → →

M-2
 EPRI Licensed Material

Hatch 1

Radiation Data

Recirculation System dose rates are summarized in Table M-3.

Table M-3
Hatch 1 Recirculation System Dose Rates

Hatch 1 – Recirculation System Dose Rates (mR/hr)

Nov- Mar- Sep- Feb- Feb- Sep- Nov- Apr- Mar-

86 87 88 90 91 91 (1) 91 (2) 92 (3) 93

EFPY

BRAC 113 118 135 161 184 320 38 118 132

A Suction 120 100 150 150 200 280 48 130 166

B Suction 150 150 120 250 180 250 50 180 196

A Discharge 80 100 120 125 175 250 25 80 77

B Discharge 100 120 150 120 180 500 30 80 87

Avg Risers

Hatch 1 – Recirculation System Dose Rates (mR/hr)

Dec- Sep- Mar- Apr- Oct- Mar- Jan- Oct- Mar-

93 94 96 (1) 96 (2) 97 99 00 00 02

EFPY

BRAC 104 165 228 113 169 155 54 60 35

A Suction 122 187 270 236 216 193 50 65 37

B Suction 147 196 221 206 190 153 81 67 46

A Discharge 73 77 226 3 108 143 40 56 29

B Discharge 74 87 196 7 105 131 43 53 28

Avg Risers
Table M-3 Notes
1. Pre decon
2. Post decon
3. Values prior to Mar-93 were obtained with unshielded probe.

M-3
 EPRI Licensed Material

Hatch 1

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Hatch 1 are
presented in Figures M-1, M-2, M-3 and M-4, respectively.

100

80
Power (%)

60

40

20

0
1/1/97 5/16/98 9/28/99 2/9/01 6/24/02 11/6/03

Power NMCA

Figure M-1
Power History, Hatch 1

Feedwater Iron Control

Average feedwater total iron was 0.92 ppb in 1999, 1.03 ppb in 2000, and 0.65 ppb in 2001.
There was no significant change to feedwater iron following the March 1999 application of
NMCA. The data from the latter half of 2001 and into 2002 show feedwater iron at times less
than 0.5 ppb, the threshold below which the impact on insoluble Co-60 should be monitored and
the need for iron addition considered.

M-4
 EPRI Licensed Material

Hatch 1

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001

0.0001
1/1/97 5/16/98 9/28/99 2/9/01 6/24/02 11/6/03

Insoluble Fe Soluble Fe NMCA

Figure M-2
Feedwater Iron, Hatch 1

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
1/1/97 5/16/98 9/28/99 2/9/01 6/24/02 11/6/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure M-3
Reactor Water Cobalt-60, Hatch 1

M-5
 EPRI Licensed Material

Hatch 1

400
HWC 22 scfm 30 45 50 45 5-6 8 scfm
350
Pleated Filters
300
NZO DZO
Dose Rate (mR/hr)

Chem decon, power uprate 1996

250

MSR retube, power uprate 1999

O2 Inj

CRB replacement complete

200 NMCA
Retube condenser 6/90
Begin replacing CRBs

150
Chem decon
100
50
0
Oct-86 Jun-89 Mar-92 Dec-94 Sep-97 Jun-00
BRAC (DPGSM) BRAC (RO2A) Milestones Co-60

Figure M-4
BRAC History, Hatch 1

Recirculation Piping Dose Rates

Hatch 1 started natural zinc injection in 8/90 and switched to DZO in 2/94. Since the start of
zinc addition, two chemical decontaminations have been performed to reduce recirculation
piping dose rates. Prior to the 1991 decontamination, the BRAC average dose rate had increased
to 320 mR/hr. Prior to the 1996 decontamination, the BRAC dose rate had increased 268 mR/hr.

Since 1997, BRAC dose rates have decreased; the most significant drop occurred following
NMCA implementation and the lowering of the hydrogen flow from 45 scfm to the 6-8 scfm
range. The reported value from the 3/02 outage was 35 mR/hr.

Reactor water average soluble Co-60 activity was 6.81E-5 µCi/ml in 2000 and 3.93E-5 µCi/ml in
2001. The average reactor water insoluble Co-60 was 1.47E-4 µCi/ml in 2000 and 4.91E-5
µCi/ml in 2001.

Recirculation Piping Gamma Scans

Gamma scan data for Hatch 1 are summarized in Table M-4.

Pipe gamma scans show that corrosion film activity reached its highest value in the 3/99 outage
when NMCA was performed. The gamma scan in 2002 corresponding to the lowest BRAC dose
rate of 35 mR/hr is the lowest reported piping activity at Hatch 1 without prior chemical

M-6
 EPRI Licensed Material

Hatch 1

rate of 35 mR/hr is the lowest reported piping activity at Hatch 1 without prior chemical
decontamination. The fraction of Zn-65 in the film has decreased since DZO initiation in 1994.
Cr-51 contributes the majority of the total deposited activity; however, about 50% of the dose
rate is attributed to Co-60.
Table M-4
Hatch 1 Recirculation Piping Gamma Scan Results

Feb- Sep- Nov- Mar- Oct- Mar- Apr- Oct- Mar- Oct- Mar-
90 91(1) 91(2) 94 94 96(1) 96(2) 97 99 00 02

Total Activity
14.4 16.6 1.8 16.7 19 47 13.9 27 41 11.5 9
(µCi/cm2)

% Co-60 51.4 48 55 25 30 18 34 22 15 15 12

% Co-58 6.9 7 11 2.4 4 4.6 7.2 1.1 5 6 3

% Mn-54 13.8 14 5 10 11 1.7 2.9 3 3 8 6.3

% Zn-65 25.7 27 27 60 48 33 57 19 7.7 11 9.6

% Cr-51 0 0 0 0 5.8 41 56 57 67 61 65
Table M-4 Notes
1. Pre-decon
2. Post -decon

Stellite™ Reduction

Hatch 1 reported replacement of all original control rod blades from 1990 through 1996.

M-7
 EPRI Licensed Material

N
HATCH 2

Table N-1
Hatch 2 Plant Design Parameters

Parameter Value

Commercial Operation Date 9/79

Capacity (MWe) 925

BWR Type 4

Drains Path Cascaded

Condensate Polishing Filter Demineralizer

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1

Hatch 2 Milestones

Milestone events for Hatch 2 are given in Table N-2.

The original 304 stainless steel recirculation piping was replaced with 316 stainless steel in 1984.
The main condenser admiralty brass tubes were replaced with titanium in 1989. NZO was
initiated in 1990, while DZO was implemented in 1993. HWC was initiated in 1991. The
station operated under moderate hydrogen water chemistry conditions as the hydrogen flow was
raised from 40 to 65 scfm between 1995 and 1998 to provide protection for lower vessel
internals. The station implemented NMCA in 3/00 and subsequently reduced hydrogen injection
flow to 8 scfm. The current hydrogen injection rate is 10 scfm.

Power uprates were implemented in 1995 and 1998. The current electrical power output is 114%
of the original design value. NRC approval was granted for a 20-year extension of the operating
license to June 2038.

N-1
 EPRI Licensed Material

Hatch 2

Table N-2
Hatch 2 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 1/95 11/98

Condenser
12/89
Retube

Recirc. Pipe
8/84
Replacement

RWCU Pipe
Replacement

Extraction Steam
Pipe
Replacement

Chem. Decon.

→ → →
9/91 → → → → →
HWC (scfm) (40- → → (50- (45-
(17) (30) (36) (8) (10) (10)
45) 65) 50)

NMCA 3/00

NZO 8/90 →

DZO 12/93 → → → → → → → → →

Iron Injection

Crud Resins

Pleated Filters 1/97 → → → → →

N-2
 EPRI Licensed Material

Hatch 2

Radiation Data

Recirculation System dose rates are summarized in Table N-3.

Table N-3
Hatch 2 Recirculation System Dose Rates

Hatch 2 – Recirculation System Dose Rates (mR/hr)

Aug- Aug- Apr- Sep- Feb- Sep- Feb- Jan- Sep- Oct-
83 84 85 86 88 89 91 92 92 92

EFPY

BRAC 128 3 41 65 80 86 158 140 213 194

A Suction 150 2 36 60 70 85 140 140 250 200

B Suction 150 2 34 38 45 80 150 140 200 200

A Discharge 150 4 32 60 80 95 160 160 200 175

B Discharge 60 3 60 100 125 85 180 120 200 200

Avg Risers

Hatch 2 – Recirculation System Dose Rates (mR/hr)

Nov- Mar- Nov- Apr- Sep- Mar- Sep- Sep- Mar- Sep-
92 93 93 94 95 97 97 98 00 01

EFPY

BRAC 160 295 281 250 209 193 133 218 133 31

A Suction 160 260 250 300 195 200 130 210 122 25

B Suction 140 260 300 200 240 180 120 260 180 40

A Discharge 160 300 275 200 180 195 130 200 110 30

B Discharge 180 360 300 300 220 195 150 200 120 30

Avg Risers 375

N-3
 EPRI Licensed Material

Hatch 2

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Hatch 2 are
presented in Figures N-1, N-2, N-3 and N-4, respectively.

100

80
Power (%)

60

40

20

0
1/1/97 5/16/98 9/28/99 2/9/01 6/24/02 11/6/03

Power NMCA

Figure N-1
Power History, Hatch 2

Feedwater Iron Control

Average feedwater total iron was 1.11 ppb in 1999, 0.92 ppb in 2000, and 0.74 ppb in 2001.
Unlike Hatch 1, feedwater iron has generally been maintained greater than the recommended
lower limit of 0.5 ppb.

N-4
 EPRI Licensed Material

Hatch 2

100

10
Feedwater Fe (ppb)

0.1

0.01

0.001

0.0001
1/1/97 5/16/98 9/28/99 2/9/01 6/24/02 11/6/03

Insoluble Fe Soluble Fe NMCA

Figure N-2
Feedwater Iron, Hatch 2

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
1/1/97 5/16/98 9/28/99 2/9/01 6/24/02 11/6/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure N-3
Reactor Water Cobalt-60, Hatch 2

N-5
 EPRI Licensed Material

Hatch 2

400
17 scfm 30 36 40 - 45 50 - 65 8 scfm
HWC
350
NZO DZO
300
Dose Rate (mR/hr)

O2 Inj
250 Pleated Filters

Complete CRB replacement

MSR retube, power uprate
200

Begin CRB replacement

Recirc pipe replacement

NMCA
150
Retube condenser

Power uprate
100
50
0
Aug-83 Apr-86 Jan-89 Oct-91 Jul-94 Apr-97 Jan-00 Sep-02

RO2A DPGSM Milestones Co-60 Dose

Figure N-4
BRAC History, Hatch 2

Recirculation Piping Dose Rates

Hatch 2 initiated NZO addition in 1990, and switched to DZO addition in 12/93. BRAC dose
rates showed a steadily increasing trend through March 1993, when the average dose rate peaked
at 295 mR/hr. The next five BRAC readings showed a declining trend, with a value of 133
mR/hr in the 9/97 outage. The next reported value was 218 mR/hr, in the 9/98 outage. The
increase in dose rate could be attributed to the increase in hydrogen flow that occurred between
the 9/97 and 9/98 measurements. During this period, hydrogen flow was increased by as much
as 20 scfm. The dose rate during the 3/00 outage when NMCA was applied showed a return to
approximately the 9/97 average. In 9/01, the BRAC dose rate was 31 mR/hr. This is the lowest
reported value since the 1984 outage when the recirculation system piping was replaced.

The average soluble reactor water Co-60 was 2.74E-5 µCi/ml in 1999, 4.26E-5 µCi/ml in 2000
and 4.93E-5 µCi/ml in 2001. Average insoluble Co-60 was 5.08E-5 µCi/ml in 1999, 5.81E-5
µCi/ml in 2000 and 1.71E-5 µCi/ml in 2001.

N-6
 EPRI Licensed Material

Hatch 2

Recirculation Piping Gamma Scans

Gamma scan data for Hatch 2 are summarized in Table N-4.

Through 3/00 the fraction of Co-60 present in the corrosion film decreased over the years, even
though the 9/98 and 3/00 scans show higher total activity than the previous data reported. The
9/01 gamma scan data correlates with the low BRAC measurement reported. Co-60 contributes
approximately 50% to the observed dose rate. Since 9/97, Cr-51 is the predominant contributor
to the total piping surface activity.
Table N-4
Hatch 2 Recirculation Piping Gamma Scan Results

Mar- Oct- Apr- Apr- Oct- Mar- Sep- Mar- Sep-

91 92 94 95 95 97 98 00 01

Total Activity
11.8 15 17.4 18.4 15.8 29.8 31.2 35.5 8.7
(µCi/cm2)

% Co-60 50 38 42 45 50 23.4 10.6 12 16

% Mn-54 18 17 6.3 7.4 6.9 6.4 3 6.5 5.7

% Zn-65 19 37 46 36 32 10 8 7 14

% Co-58 9.3 3 3 4 4 4 5.4 5 9.2

% Cr-51 50 70 65 54

Stellite™ Reduction

Hatch 2 replaced 137 control rod blades between 1992 and 1997.

N-7
 EPRI Licensed Material

O
HOPE CREEK

Table O-1
Hope Creek Plant Design Parameters

Parameter Value

Commercial Operation Date 12/86

Capacity (MWe) 1136

BWR Type 4

Drains Path Cascaded

Condensate Polishing Filter + Deep Bed

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1

Hope Creek Milestones

Milestone events for Hope Creek are given in Table O-2.

NZO was implemented in 12/86. The station subsequently experienced significant increases in
Zn-65 activity in a number of plant systems while attempting to achieve the desired reactor water
zinc concentration. The inability to achieve the desired reactor water soluble zinc concentration
was the direct result of high feedwater iron. Hope Creek applied various types of crud resins
from 1991 through 1995 in the condensate polisher deep beds to control feedwater iron. In 2/93,
the station switched to DZO addition and initiated HWC. In June 1999, Hope Creek started the
operation of full-flow condensate pre-filters and iron addition. A 1.4% power uprate was
implemented in 8/01.

O-1
 EPRI Licensed Material

Hope Creek

Table O-2
Hope Creek Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate X

Condenser
Retube

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction Steam
#4
Pipe #3 Htr
Htr
Replacement

Chem. Decon.

→
2/93
HWC (scfm) → → → → → → → →
(21.5) (35,
7/99)

NMCA

NZO 12/86

→
DZO 2/93 (Note → → → → → → → →
1)

Iron Injection 6/99 → → →

Crud Resins 1992 → → X

Pleated Filters 6/99 → → →

Table O-2 Notes

1. DZO suspended for Cycle 6, 4/94 – 11/95.

Radiation Data

Recirculation System dose rates are summarized in Table O-3.

O-2
 EPRI Licensed Material

Hope Creek

Table O-3
Hope Creek Recirculation System Dose Rates

Hope Creek - Recirculation System Dose Rates (mR/hr)

Sep- Feb- Mar- Sep- Feb- Mar- Nov- Oct- Apr- Oct-
87 88 89 89 91 94 95 97 00 01

EFPY .75 1.04 1.82 2.29 3.29 6.02 7.40 8.78 10.91

BRAC 57 72 111 112.5 93 100 138.2 112.5 124.5 114

A Suction 64 88 124 128 107 118 118 130 135 135

B Suction 70 82 127 136 110 110 180 130 114 125

A Discharge 59 65 102 96 85 88 160 105 105

B Discharge 36 52 92 90 70 84 95 85 90

Avg Risers 96 127 203 209 171 183.6 227.9 206 182 196

Trend Data

Power, feedwater iron, reactor water cobalt-60, and BRAC history trend plots for Hope Creek
are presented as Figures O-1, O-2, O-3 and O-4, respectively.

100

80
Power (%)

60

40

20

0
1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Figure O-1
Power History, Hope Creek

O-3
 EPRI Licensed Material

Hope Creek

100

10
Feedwater Fe (ppb)

0.1

0.01
1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Total Fe

Figure O-2
Feedwater Iron, Hope Creek

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Insoluble Co-60 Soluble Co-60

Figure O-3
Reactor Water Cobalt-60, Hope Creek

O-4
 EPRI Licensed Material

Hope Creek

Feedwater Iron Control

Hope Creek applied low cross-linked resins for enhanced crud removal for 3 years before
returning to conventional resins. The use of crud resins was discontinued due to increased
reactor water sulfate concentration. The plant has installed full-flow condensate pre-filters
upstream of the original deep bed polishers to reduce feedwater iron. Operation of the filters
began in 6/99 along with iron addition. Average feedwater iron for 1999 was 5.50 ppb before
pre-filter operation, and 1.01 ppb after pre-filters and iron injection were in operation. The
average feedwater iron concentration was 0.87 ppb in 2000 and 0.70 ppb in 2001.

200
NZO DZO DZO

HWC
150
Dose Rate (mR/hr)

100
Pleated filters

Power Uprate
50

0
Jun-84 Feb-87 Nov-89 Aug-92 May-95 Feb-98 Nov-00

BRAC Milestones Co-60 Dose

Figure O-4
BRAC History, Hope Creek

Recirculation Piping Dose Rates

NZO was initiated in 12/86 when the station commenced commercial operation. Dose rates
increased to just over 100 mR/hr after the first few years of operation and remained at this level
until the mid 1990s. HWC and DZO were implemented in 1/93. The BRAC dose rate following
the first full cycle of HWC operation remained at the 100 mR/hr level but increased by about
40% after the second cycle of HWC operation. Dose rates for the most recent measurements
have been in the 110-120 mR/hr range.

Reactor water soluble Co-60 activity averaged 1.31E-4 µCi/ml in 2000 and 5.29E-5µCi/ml in
2001. Reactor water insoluble Co-60 activity averaged 1.17E-4 µCi/ml in 2000 and 1.44E-4
µCi/ml in 2001.

O-5
 EPRI Licensed Material

Hope Creek

Recirculation Piping Gamma Scans

Gamma scan data for Hope Creek are summarized in Table O-4.

Since 10/2, the piping total activity has been stable in the range of 14 – 18 µCi/cm2. The gamma
scan results show a decreasing trend in Zn-65 activity after switching from NZO to DZO in
1993. By 4/00, the contribution to total activity from Co-60 increased to 26%, at which time the
Co-60 contribution to the total contact dose rate was 56%.
Table O-4
Hope Creek Recirculation Piping Gamma Scan Results

Sep- Feb- Mar- Sep- Feb- Oct- Mar- Nov- Sep- Apr- Oct-
87 88 89 89 91 92 94 95 97 00 01

Total Activity
10.9 14.9 22.0 22.1 19.2 17.0 18.0 14.5 18.1 14.0 11.3
(µCi/cm2)

% Co-60 10 8 11 12 13 14 15 22 22 26 27

% Mn-54 10 10 13 19 21 29 43 53 57 60 53

% Zn-65 54 68 64 58 55 45 26 9 4 0 1

% Co-58 21 11 9 6 5 4 4 5 3 4 9.4

% Cr-51 0 0 0 0 0 0 0 0 0 0 0

% Fe-59 8 12 11 14 11 9.6

Fuel Failures

The station experienced a fuel failure in March 2002 with the offgas activity (sum of six noble
gases) increasing from about 50 µCi/sec to over 1500 µCi/sec (the station reports that a
maximum value of 3800 µCi/sec was reached). The station performed suppression testing and
has inserted two control rods to suppress the leak, reporting that the offgas activity has remained
about 1100 µCi/sec since leak suppression. Reactor water dose equivalent iodine and offgas sum
of 6 noble gases trend data are presented in Figures O-5 and O-6, respectively.

O-6
 EPRI Licensed Material

Hope Creek

Dose Equivalent Iodine (µCi/ml) 1.E-03

1.E-04

1.E-05

1.E-06
1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

DEI

Figure O-5
Reactor Water Dose Equivalent Iodine, Hope Creek

2000

1750

1500
Sum of 6 (µCi/sec)

1250

1000

750

500

250

0
1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Figure O-6
Sum of 6 Noble Gases, Hope Creek

O-7
 EPRI Licensed Material

P
LAGUNA VERDE 1

Table P-1
Laguna Verde 1 Plant Design Parameters

Parameter Value

Commercial Operation Date 7/90

Capacity (MWe) 686 (est.)

BWR Type 5

Drains Path Forward Pumped

Condensate Polishing Filter + Deep Bed

RWCU Capacity (% Feedwater Flow), Normal/Maximum 0.85/0.85

Laguna Verde 1 Milestones

Milestone events for Laguna Verde 1 are given in Table P-2.

A Reactor Water Clean-Up System chemical decontamination was performed in 1994, during the
third refueling outage. The recirculation system piping was chemically decontaminated in 3/98.
Condensate pre-filters were installed and phased into service over several months starting in
3/97. DZO addition was implemented following the 1998 recirculation system piping chemical.
A 5% power uprate was implemented in 6/99. HWC and NMCA are planned for the fall of
2002.

P-1
 EPRI Licensed Material

Laguna Verde 1

Table P-2
Laguna Verde 1 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 6/99

Condenser Retube

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction
Steam Pipe
Replacement

Chem. Decon. (RWCU) 3/98

HWC (scfm)

NMCA

NZO

DZO 6/98 → → → →

Iron Injection

Crud Resins

Pleated Filters 3/97 → → → → →

Radiation Data

Recirculation System dose rates are summarized in Table P-3.

P-2
 EPRI Licensed Material

Laguna Verde 1

Table P-3
Laguna Verde 1 Recirculation System Dose Rates

Laguna Verde 1 – Recirculati on System Dose Rates (mR/hr)

May- Sep- Dec- Apr- Jul- Oct- May- Aug- Aug- Jun-
90 91 92 94 95 96 98 (1) 99 00 01

EFPY 0.21 1.26 1.93 3.00 3.96

BRAC 7.9 90.6 105 205 336 215 46 264 148.5 109

A Suction 4.5 88 126 342 209 400 35 212 98 125

B Suction 6.7 83.3 76 211 861 220 83 471 86 88

A Discharge 10.5 88.9 100 122 134 185 42 154 350 128

B Discharge 9.9 102 120 146 142 57 25 217 60 95

Avg Risers
Table P-3 Notes
1. Post-decon

Trend Data

Power, feedwater iron, reactor water cobalt-60, and BRAC history trend plots for Laguna Verde
1 are presented in Figures P-1, P-2, P-3 and P-4, respectively. Note that the Co-60 data are
reported in units of µCi/kg, rather than µCi/ml as reported by U.S. plants.

P-3
 EPRI Licensed Material

Laguna Verde 1

100

80
Power (%)

60

40

20

0
12/1/96 4/15/98 8/28/99 1/9/01 5/24/02 10/6/03

Figure P-1
Power History, Laguna Verde 1

Feedwater Iron Control

Laguna Verde 1 began phasing the condensate pre-filters into service in March 1997. The
average feedwater insoluble iron was 2.00 ppb in 1999, 2.69 ppb in 2000, and 1.59 ppb in 2001.
Prior to pre-filter implementation, feedwater iron typically ranged between 3 to 4 ppb.

The condensate polisher system average effluent iron was 0.47 ppb in 1999, 0.44 ppb in 2000,
and 0.28 ppb in 2001. The average insoluble iron in the forward pumped heater drains was 4.42
ppb in 1999, 2.43 ppb in 2000, and 1.7 ppb in 2001. The forward pumped heater drains
contribute significantly to the total feedwater iron.

P-4
 EPRI Licensed Material

Laguna Verde 1

100

10
Feedwater Fe (ppb)

0.1

0.01

0.001

0.0001
12/1/96 4/15/98 8/28/99 1/9/01 5/24/02 10/6/03

Insoluble Fe Soluble Fe

Figure P-2
Feedwater Iron, Laguna Verde 1

1.E+01

1.E+00
Reactor Water Co-60 (µCi/kg)

1.E-01

1.E-02

1.E-03

1.E-04

1.E-05
12/1/96 4/15/98 8/28/99 1/9/01 5/24/02 10/6/03

Insoluble Co-60 Soluble Co-60

Figure P-3
Reactor Water Cobalt-60 (µCi/kg), Laguna Verde 1

P-5
 EPRI Licensed Material

Laguna Verde 1

400

350

300
Dose Rate (mR/hr)

250
Pleated filters
200
DZO
150

Power Uprate
Chem decon
100

50

0
May-90 Sep-91 Jan-93 Jun-94 Oct-95 Mar-97 Jul-98 Nov-99 Apr-01

BRAC Milestones

Figure P-4
BRAC History, Laguna Verde 1

Recirculation Piping Dose Rates

After peaking at 336 mR/hr in 5/95 and reaching an operating minimum of 46 mR/hr in 4/98
following a chemical decontamination, the recirculation piping BRAC average dose rate peaked
again in 8/99 at 264 mR/hr. This peak dose rate occurred approximately one year after DZO
initiation and two months after a power uprate. DZO was not implemented until about a month
after the unit restarted following the chemical decontamination outage in 3/98. Since the 8/99
data, with continued DZO injection and lower feedwater iron, dose rates have shown a steady
decrease to109 mR/hr in June 2001.

Soluble reactor coolant Co-60 averaged about 6.35E-5 µCi/ml (6.35E-2 µCi/kg) in 2000 and
4.0E-5 µCi/ml (4.0E-2 µCi/kg) in 2001. Insoluble reactor coolant Co-60 averaged 6.9E-5
µCi/ml (6.9E-2 µCi/kg) in 2000 and 1.86E-5 µCi/ml (1.86E-2 µCi/kg) in 2001. Steady state
soluble Co-60 values have trended down since starting DZO injection; however, the spikes
reported from transients, particularly from startups, have a significant impact on the averages.

P-6
 EPRI Licensed Material

Q
LAGUNA VERDE 2

Table Q-1
Laguna Verde 2 Plant Design Parameters

Parameter Value
Commercial Operation Date 4/95
Capacity (MWe) 686 (est.)
BWR Type 5
Drains Path Forward Pumped
Condensate Polishing Filters + Deep Beds
RWCU Capacity (% Feedwater Flow), Normal/Maximum 0.85/0.85

Laguna Verde 2 Milestones

Milestone events for Laguna Verde 2 are given in Table Q-2.

Condensate pre-filters were added as a retrofit, and the first filter vessel was placed into service
in 3/98. The remaining filters were phased into operation during 1998. DZO addition was
implemented in 8/98, four months prior to a recirculation piping chemical decontamination. A
second chemical decontamination of the recirculation piping was performed in 4/00. A 5%
power uprate was implemented in 7/99. HWC and NMCA are planned in 2003.

Q-1
 EPRI Licensed Material

Laguna Verde 2

Table Q-2
Laguna Verde 2 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 7/99

Condenser Retube

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction Steam
Pipe Replacement

Chem. Decon. 12/98 4/00

HWC (scfm)

NMCA

NZO

DZO 8/98 → → → →

Iron Injection

Crud Resins

Pleated Filters 3/98 → → → →

Q-2
 EPRI Licensed Material

Laguna Verde 2

Radiation Data

Recirculation System dose rates are summarized in Table Q-3.

Table Q-3
Laguna Verde 2 Recirculation System Dose Rates

Laguna Verde 2 – Recirculati on System Dose Rates (mR/hr)

Mar- Aug- Sep- Mar- Jun-

95 96 97 00 (1) 00 (2)

EFPY

BRAC 3.9 102 276 141 34

A Suction 4.3 205 250 240 22

B Suction 4.5 77 98 75 29

A Discharge 3.7 37.5 410 181 32

B Discharge 3.0 88 345 69 55

Avg Risers

Table Q-3Notes
1. Pre-decon
2. Post-decon

Trend Data

Power, feedwater iron, reactor water cobalt-60, and BRAC history trend plots for Laguna Verde
2 are presented in Figures Q-1, Q-2, Q-3 and Q-4, respectively. Note that the cobalt-60 data are
reported in units of µCi/kg, rather µCi/ml as reported by U.S. plants.

Q-3
 EPRI Licensed Material

Laguna Verde 2

100

80
Power (%)

60

40

20

0
12/1/96 12/1/98 11/30/00 11/30/02

Figure Q-1
Power History, Laguna Verde 2

Feedwater Iron Control

Condensate pre-filters were phased into service during 1998 at Laguna Verde 2. Insoluble
feedwater iron has decreased from an average of 9.05 ppb in 1997 to an average of 2.69 ppb in
2000 and 1.62 ppb in 2001.

There is a significant iron contribution to the final feedwater from the forward pumped heater
drains. The average insoluble iron from this source was 4.79 ppb in 1999, 2.4 ppb in 2000, and
2.3 ppb in 2001. By comparison, the average condensate polishing system effluent iron was 0.15
ppb in 1999, 0.27 ppb in 2000, and 0.09 ppb in 2001.

Q-4
 EPRI Licensed Material

Laguna Verde 2

1000

100
Feedwater Fe (ppb)

10

0.1

0.01

0.001
12/1/96 12/1/98 11/30/00 11/30/02

Insoluble Fe Soluble Fe

Figure Q-2
Feedwater Iron, Laguna Verde 2

1.E+02

1.E+01
Reactor Water Co-60 (µCi/kg)

1.E+00

1.E-01

1.E-02

1.E-03

1.E-04

1.E-05
12/1/96 12/1/98 11/30/00 11/30/02

Insoluble Co-60 Soluble Co-60

Figure Q-3
Reactor Water Cobalt-60 (µCi/kg), Laguna Verde 2

Q-5
 EPRI Licensed Material

Laguna Verde 2

400

350

300
Dose Rate (mR/hr)

250

200
DZO
150 Pleated Filters

Power Uprate
Chem Decon

Chem Decon
100

50

0
Mar-95 Apr-96 May-97 Jun-98 Jul-99 Aug-00 Sep-01

BRAC Milestones

Figure Q-4
BRAC History, Laguna Verde 2

Recirculation Piping Dose Rates

Prior to DZO, the highest reported BRAC dose rate was 276 mR/hr in 9/97. Soluble Co-60 has
decreased since DZO initiation, and two chemical decontaminations have been performed. The
most recent measurement (post-decontamination) was 34 mR/hr.

Reactor soluble and insoluble Co-60 activity has been decreasing since DZO implementation.
Soluble reactor coolant Co-60 averaged about 4.42E-3 µCi/ml (4.42 µCi/kg) in 2000 and 2.5E-5
µCi/ml (2.5E-2 µCi/kg) in 2001. Insoluble reactor coolant Co-60 averaged 1.5E-4 µCi/ml (1.5E-
1 µCi/kg) in 2000 and 1.49E-5 µCi/ml (1.49E-2 µCi/kg) in 2001. Spikes reported from
transients, particularly from startups, have a significant impact on the averages.

Q-6
 EPRI Licensed Material

R
LASALLE 1

Table R-1
LaSalle Unit 1 Plant Design Parameters

Parameter Value

Commercial Operation Date 1/84

Capacity (MWe) 1200

BWR Type 5

Drains Path Forward Pumped

Condensate Polishing Filters + Deep Beds

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1

LaSalle 1 Milestones

Milestone events for LaSalle 1 are given in Table R-2.

Chemical decontaminations of the recirculation system piping were performed in 1988, 1989,
and 1994, with 86 curies removed in 1994, 80 curies removed in 1989, and 50 curies removed in
1988. A partial chemical decontamination (one recirculation pump and RWCU piping) was
performed in 1996. LaSalle 1 was shutdown from 9/96 to 8/98.

DZO injection was initiated in 07/94, while HWC was initiated in 08/99. NMCA was performed
at LaSalle Unit 1 in 10/99. Startup of the condensate pre-filters was implemented in 02/01,
followed by a power uprate from 1153 MWe to 1200 MWe in 03/01.

R-1
 EPRI Licensed Material

LaSalle 1

Table R-2
LaSalle 1 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 3/01 →

Condenser Retube

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction Steam
Pipe Replacement

3/88,
Chem. Decon. 4/94 2/96
10/89

8/99
HWC (scfm) → → →
(8-10)

NMCA 10/99

NZO

DZO 7/94 → → → → → → → →

Iron Injection

Crud Resins 1/94 12/95 9/99 11/00

Pleated Filters 02/01 →

R-2
 EPRI Licensed Material

LaSalle 1

Radiation Data

Recirculation System dose rates are summarized in Table R-3.

Table R-3
LaSalle 1 Recirculation System Dose Rates

LaSalle 1 – Recirculation System Dose Rates (mR/hr)

May-84 Oct-85 Mar-88 Jun-88 Jan-90 Feb-91 Oct-92 Jun-94 Jul-94

(1) (2) (2) (1) (2)

EFPY

BRAC 100 205 288 50 50 223 420 460 65

A Suction

B Suction

A Discharge

B Discharge

Avg Risers

LaSalle 1 – Recirculation System Dose Rates (mR/hr)

Feb-96 Oct-99 Feb-01 Jan-02

EFPY

BRAC 82 181 157.5 256

A Suction 42.5 200 320

B Suction 67.5 180 260

A Discharge 65.5 140 290

B Discharge 152.5 110 156

Avg Risers
Table R-3 Notes
1. Pre-decon
2. Post decon

R-3
 EPRI Licensed Material

LaSalle 1

Trend Data

Power, feedwater iron, reactor water cobalt-60, and BRAC history trend plots for LaSalle 1 are
presented in Figures R-1, R-2, R-3 and R-4, respectively

Feedwater Iron Control

Feedwater iron was significantly reduced following implementation of condensate filtration in

2001. The average feedwater iron in 2001 was 1.82 ppb compared to an average of 3.56 ppb for
2000.

100

80
Power (%)

60

40

20

0
1/1/96 2/4/97 3/11/98 4/15/99 5/19/00 6/23/01 7/28/02 9/1/03

% Power NMCA

Figure R-1
Power History, LaSalle 1

R-4
 EPRI Licensed Material

LaSalle 1

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001
1/1/96 2/4/97 3/11/98 4/15/99 5/19/00 6/23/01 7/28/02 9/1/03

Insoluble Fe Soluble Fe NMCA

Figure R-2
Feedwater Iron, LaSalle 1

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
1/1/96 2/4/97 3/11/98 4/15/99 5/19/00 6/23/01 7/28/02 9/1/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure R-3
Reactor Water Cobalt-60, LaSalle 1

R-5
 EPRI Licensed Material

LaSalle 1

Recirculation Piping Dose Rates

Dose rates steadily increased after the 1989 chemical decontamination and reached a peak value
of 460 mR/hr in 6/94, prior to performing the third chemical decontamination. DZO was
initiated following the 1994 chemical decontamination. The next reported drywell dose rate
measurement was 82 mR/hr in 2/96, approximately 19 months after the initiation of DZO. The
BRAC dose rate after 14 months of operation under NMCA and HWC was 256 mR/hr.

Isotopic data indicate an average reactor coolant soluble Co-60 activity of 1.67E-4 µCi/ml in
2001 and 1.83E-4 µCi/ml in 2000. The average reactor coolant insoluble Co-60 activity was
5.47E-4µCi/ml in 2001and 3.56E-4µCi/ml in 2000. Soluble Co-60 activity increased following
NMCA.

Crud Resins Crud Resins

600
DZO

500 Extended Shutdown Pleated Filters

Dose Rate (mR/hr)

400 HWC

300
Chem decon

NMCA

200
Power Uprate
Chem decon

Chem decon

Chem decon

100

0
May-84 Jan-87 Oct-89 Jul-92 Apr-95 Jan-98 Oct-00 Jul-03

BRAC Milestones

Figure R-4
BRAC History, LaSalle 1

R-6
 EPRI Licensed Material

S
LASALLE 2

Table S-1
LaSalle 2 Plant Design Parameters

Parameter Value

Commercial Operation Date 10/84

Capacity (MWe) 1200

BWR Type 5

Drains Path Forward Pumped

Condensate Polishing Filters + Deep Beds

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1

LaSalle 2 Milestones

Milestone events for LaSalle 2 are given in Table S-2.

Chemical decontaminations of the recirculation system piping were performed in 1989 and 1995.
The 1995 chemical decontamination of the recirculation piping removed 88.5 curies. The plant
was in an extended shutdown from 9/96 to 4/99.

DZO injection began in 6/95. HWC was initiated in 8/00 and NMCA was performed in11/00.
Startup of the condensate pre-filters began in 2/01. A power uprate from 1153 MWe to 1200
MWe was implemented in 3/01.

S-1
 EPRI Licensed Material

LaSalle 2

Table S-2
LaSalle 2 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power 3/01
Uprate

Condenser
Retube

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction
Steam Pipe
Replacement

Chem. 11/88 3/95

Decon.

8/00 → →
HWC (8-10) (8-10) (8-10)

11/00
NMCA

NZO

DZO 6/95 → → → → → → →

Iron Injection

Crud Resins 1/94→ 12/95 9/99→ 11/00

Pleated
02/01 →
Filters

S-2
 EPRI Licensed Material

LaSalle 2

Radiation Data

Recirculation System dose rates are summarized in Table S-3:

Table S-3
LaSalle 2 Recirculation System Dose Rates

LaSalle 2 - Recirculation System Dose Rates (mR/hr)

Apr-85 Feb-89 Jan-92 Mar-95 Apr-95 Nov-96 Nov-00 Sept-01 Nov-00 Sep-01
(1) (2)

EFPY

BRAC 53 25 425 618 34 110 153 117 153 118

A Suction 180

B Suction 50

A Discharge 110

B Discharge 100

Avg Risers
Table S-3 Notes
1. Pre-decon
2. Post-decon

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots are presented in
Figures S-1, S-2, S-3 and S-4 respectively.

Feedwater Iron Control

Feedwater iron was significantly reduced following implementation of condensate filtration in

2001. The average feedwater iron in 2001 was 1.49 ppb compared to an average of 3.53 ppb for
2000.

S-3
 EPRI Licensed Material

LaSalle 2

100

80
Power (%)

60

40

20

0
3/15/95 3/14/97 3/14/99 3/13/01 3/13/03

NMCA

Figure S-1
Power History, LaSalle 2

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001
3/15/95 3/14/97 3/14/99 3/13/01 3/13/03

Insoluble Fe Soluble Fe NMCA

Figure S-2
Feedwater Iron, LaSalle 2

S-4
 EPRI Licensed Material

LaSalle 2

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
3/15/95 3/14/97 3/14/99 3/13/01 3/13/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure S-3
Reactor Water Cobalt-60, LaSalle 2

Recirculation Piping Dose Rates

Dose rates increased from 25 mR/hr after the 1989 chemical decontamination to 425 mR/hr just
prior to the 1995 chemical decontamination. LaSalle 2 started DZO injection after the 1995
chemical decontamination, and the BRAC average dose rate measured in 9/99 was 153 mR/hr,
indicating that DZO was effective in suppressing piping recontamination. The BRAC average
dose rate in September 2001, after about 10 of operation months under NMCA with HWC, had
decreased to 117 mR/hr. This is approximately equal to the first BRAC reading after starting
DZO, following the 1995 chemical decontamination. Average soluble reactor water Co-60 was
9.59E-5 µCi/ml in 2000 and 1.69E-4 µCi/ml in 2001. The average insoluble Co-60 was 1.28E-4
µCi/ml in 2000 and 2.99E-4 µCi/ml in 2001. Both soluble and insoluble Co-60 activity have
increased following NMCA.

S-5
 EPRI Licensed Material

LaSalle 2

Pleated Filters
600
NMCA
500 DZO
Dose Rate (mR/hr)

400
Extended Shutdown HWC

300 Crud Resins Crud Resins

200
Chem decon

Chem decon

Power Uprate
100

0
Jun-84 Feb-87 Nov-89 Aug-92 May-95 Feb-98 Nov-00

BRAC Milestones Co-60 Dose

Figure S-4
BRAC History, LaSalle 2

Recirculation Piping Gamma Scans

Gamma scan data are shown in Table S-4.

Measurements made during 11/96 showed a total activity of 8.3 µCi/cm2, with 47% of the
activity from Co-60 and 42% from Mn-54. Since these measurements, the plant has started both
DZO and hydrogen injection, which should increase the contribution of Zn-65 and Cr-51 in the
pipe corrosion film.

S-6
 EPRI Licensed Material

LaSalle 2

Table S-4
LaSalle 2 Recirculation Piping Gamma Scan Results

Date 11/96

Total Activity
8.3
(µCi/cm2)

% Co-60 46.7

% Co-58 4.5

% Mn-54 41.6

% Zn-65 2.5

%Cr-51 nd

S-7
 EPRI Licensed Material

T
LIMERICK 1

Table T-1
Limerick 1 Plant Design Parameters

Parameter Value

Commercial Operation Date 2/86

Capacity (MWe) 1200

BWR Type 4

Drains Path Cascaded

Condensate Polishing Filter + Deep Bed

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1.1/1.1

Limerick 1 Milestones

Milestone events for Limerick 1 are given in Table T-1.

Power uprates were implemented in 1995 and again in 1998. Current electrical power output is
approximately 114% of original design. Full flow deep bed demineralizers were installed
downstream of the original condensate filter/demineralizers in 1992 to control copper from the
condenser tubes. Non-precoat filter septa were subsequently installed in the original
filter/demineralizer vessels; pleated filters septa were first used in 1994. NZO was started in
1992 and the plant switched to DZO in 5/97. Feedwater iron injection was implemented in 2/97.
HWC was initiated 9/98, and NMCA was performed in 3/00.

T-1
 EPRI Licensed Material

Limerick 1

Table T-2
Limerick 1 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate X X

Condenser Retube

Recirc. Pipe
Replacement

RWCU Pipe 4/00

Replacement (1)

Extraction Steam
Pipe Replacement

Chem. Decon.

→ →
HWC (scfm) 9/98 → →
(6-10) (8-10)

NMCA 3/00

NZO 9/92 → → → →

DZO 5/97 → → → → →

Iron Injection 2/97 → → → → →

Crud Resins

Pleated Filters 10/94 → → → → → → → →

Table T-2 Notes
1. One of 3 50% pumps and nearby piping replaced with 100% pump.

T-2
 EPRI Licensed Material

Limerick 1

Radiation Data

Recirculation System dose rates are summarized in Table T-3:

Table T-3
Limerick 1 Recirculation System Dose Rates

Limerick 1 – Recirculation System Dose Rates (mR/hr)

Jan-94 Feb-96 Apr-98 Mar-00 Mar-02

EFPY

BRAC 137.5 109 93.5 168 44.5

A Suction 140 118 100 39

B Suction 160 126 110 50

A Discharge 130 103 80

B Discharge 120 90 84

Avg Risers 374 223 217 63

Trend Data

Power history, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for
Limerick 1 are provided in Figures T-1, T-2, T-3 and T-4, respectively.

T-3
 EPRI Licensed Material

Limerick 1

100

80
Power (%)

60

40

20

0
10/28/95 10/27/97 10/27/99 10/26/01 10/26/03

NMCA

Figure T-1
Power History, Limerick 1

Feedwater Iron Control

The annual average total feedwater iron was 0.64 ppb in 1999, 0.65 ppb in 2000, and 0.58 ppb in
2001. Following the implementation of the deep bed demineralizers in 1992, the
filter/demineralizer septa were changed to non-precoat pleated septa in 1994. The station
implemented feedwater iron injection in 1997 because of low CDE effluent iron.

T-4
 EPRI Licensed Material

Limerick 1

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001
10/28/95 10/27/97 10/27/99 10/26/01 10/26/03

Insoluble Fe Soluble Fe NMCA

Figure T-2
Feedwater Iron, Limerick 1

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
10/28/95 10/27/97 10/27/99 10/26/01 10/26/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure T-3
Reactor Water Cobalt-60, Limerick 1

T-5
 EPRI Licensed Material

Limerick 1

300
NZO DZO

250 Pleated Filters

Fe Inj
Dose Rate (mR/hr)

200 HWC
NMCA

RWCU Pipe Repl Begins

150

Deep Beds Installed

100

Power Uprate
50

0
6/1/84 2/26/87 11/22/89 8/18/92 5/15/95 2/8/98 11/4/00 8/1/03

BRAC Co-60 Milestones

Figure T-4
BRAC History, Limerick 1

Recirculation Piping Dose Rates

BRAC dose rates decreased from 137 mR/hr in 1994 to 109 mR/hr in 1996 and then to 93.5
mR/hr in 1998. NZO/DZO, pleated filters and iron injection were implemented in this time
frame. The BRAC dose rate increased to 168 mR/hr in 3/00 under the first cycle with HWC.
The BRAC dose rate decreased significantly after the first full cycle with NMCA and HWC to
44 mR/hr in 3/02.

The average reactor water soluble Co-60 activity was 1.82E-3 µCi/ml in 1999, 9.31E-5 µCi/ml
in 2000, and 5.32E-5 µCi/ml in 2001. The average reactor water insoluble Co-60 activity was
8.04E-5 µCi/ml in 1999, 2.82E-4 µCi/ml in 2000, and 1.60E-4 µCi/ml in 2001. The significant
decline in BRAC dose rate in 2002 corresponds with the decline in reactor water soluble Co-60
concentration.

Recirculation Piping Gamma Scans

Limerick 1 gamma scan data are summarized in Table T-4.

T-6
 EPRI Licensed Material

Limerick 1

Table T-4
Limerick 1 Recirculation Piping Gamma Scan Results

May-87 Jan-89 Sep-90 Mar-92 Feb-94 Feb-96 Apr-98

Total Activity (µCi/cm2) 7.8 6.4 6.6 10.7 12 13.5 11

% Co-60 22 31 41 41 40 37 43

% Co-58 33 16 16 15 7.5 4.4 7

% Mn-54 2.5 8.5 29 33 43 32 19

% Zn-65 41 40 11 7 6.6 20 27

The total activity shows a gradual increase from 1987 to 1996, but then declines in 1998
following DZO implementation in 5/97. Co-60 remains the major contributor to the total
activity. The contribution from Mn-54 reached a peak in 1994, but has declined since with the
1998 value approximately half of the 1994 level. In 1998, Zn-65 was still a major contributor to
the total surface activity.

T-7
 EPRI Licensed Material

U
LIMERICK 2

Table U-1
Limerick 2 Plant Design Parameters

Parameter Value

Commercial Operation Date 1/90

Capacity (MWe) 1200

BWR Type 4

Drains Path Cascaded

Condensate Polishing Filter + Deep Bed

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1.1/1.1

Limerick 2 Milestones

Milestone events for Limerick 2 are given in Table U-2.

Power uprates were implemented in 1995 and again in 1999. Current electrical power output is
approximately 114% of original design. Limerick 2 installed full flow deep beds downstream of
the original condensate filter/demineralizers in 1993 to control copper from the condenser tubes.
Non-precoat filter septa were subsequently installed in the original filter/demineralizer vessels;
pleated septa use began in 1995. Natural zinc oxide injection began in 1991 and the station
switched to DZO in 5/97. Feedwater iron injection was started in 3/97. HWC was initiated in
1/98 and NMCA was performed in 4/01.

U-1
 EPRI Licensed Material

Limerick 2

Table U-2
Limerick 2 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power
X X
Uprate

Condenser
Retube

Recirc. Pipe
Replacement

RWCU Pipe 4/01

Replacement (1)

Extraction
Steam Pipe
Replacement

Chem.
Decon.

HWC (scfm) → →
1/98 → →
(6-10) (8-10)

NMCA 4/01

NZO 1991 → → → →

DZO 5/97 → → → → →

Iron Injection 3/97 → → → → →

Crud Resins

Pleated
4/95 → → → → → → →
Filters
Table U-2 Notes
1. One of 3 50% pumps and nearby piping replaced with 100% pump.

U-2
 EPRI Licensed Material

Limerick 2

Radiation Data

Recirculation System dose rates are summarized in Table U-3.

Table U-3
Limerick 2 Recirculation System Dose Rates

Limerick 2 – Recirculation System Dose Rates (mR/hr)

Jan-95 Jan-97 May-99 Apr-01

EFPY

BRAC 164 107.5 102.5 80

A Suction 160 110 105 90

B Suction 180 120 120 80

A Discharge 160 100 90 80

B Discharge 155 100 95 70

Avg Risers 241 147 182 166.5

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Limerick 2 are
presented in Figures U-1, U-2, U-3 and U-4, respectively.

U-3
 EPRI Licensed Material

Limerick 2

100

80
Power (%)

60

40

20

0
10/28/95 10/27/97 10/27/99 10/26/01 10/26/03

NMCA

Figure U-1
Power History, Limerick 2

Feedwater Iron Control

The annual average total feedwater iron was 0.69 ppb in 1999, 0.74 ppb in 2000, and 0.80 ppb in
2001. Following the implementation of the deep bed demineralizers in 1993, the
filter/demineralizer septa were changed to non-precoat pleated septa in 1995. The station
implemented feedwater iron injection in 1997 because of low CDE effluent iron.

U-4
 EPRI Licensed Material

Limerick 2

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001
10/28/95 10/27/97 10/27/99 10/26/01 10/26/03

Insoluble Fe Soluble Fe NMCA

Figure U-2
Feedwater Iron, Limerick 2

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
10/28/95 10/27/97 10/27/99 10/26/01 10/26/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure U-3
Reactor Water Cobalt-60, Limerick 2

U-5
 EPRI Licensed Material

Limerick 2

300
NZO DZO

250
Pleated filters
Dose Rate (mR/hr)

200
Iron inj

150

Begin RWCU Pipe Repl

HWC

Deep Beds Installed

100

Power Uprate
50
NMCA

0
Jun-84 Feb-87 Nov-89 Aug-92 May-95 Feb-98 Nov-00 Aug-03

BRAC Co-60 Dose Milestones

Figure U-4
BRAC History, Limerick 2

Recirculation Piping Dose Rates

The BRAC reported dose rates data show a decreasing trend between 1995 and 2000. Unlike
Limerick 1, Limerick 2 did not see an increase in BRAC dose rate following the first full cycle
of operation under HWC.

Average reactor water soluble Co-60 activity was 4.57E-5 µCi/ml in 1999, 3.94E-5 µCi/ml in
2000, and 5E-5 µCi/ml in 2001. The average reactor water insoluble Co-60 activity was 1.04E-4
µCi/ml during 1999, 5.43E-5 µCi/ml in 2000, and 1.20E-3 µCi/ml in 2001.

Recirculation Piping Gamma Scans

Limerick 2 gamma scan results are summarized in Table U-4.

U-6
 EPRI Licensed Material

Limerick 2

Table U-4
Limerick 2 Recirculation Piping Gamma Scan Results

Apr-91 Feb-93 Feb-95 Jan-97 May-99

Total Activity (µCi/cm2) 1.9 9.3 12.9 11.3 10.35

% Co-60 16 38 38 40 42

% Co-58 50 14 5 5 11

% Mn-54 15 23 40 27 24

% Zn-65 13 22 14 22 16

The total specific activity showed an increasing trend from 4/91 – 2/95. The 1/97 and 5/99
results suggest that the activity has stabilized. Co-60 is the major contributor to the total activity.

U-7
 EPRI Licensed Material

V
MONTICELLO

Table V-1
Monticello Plant Design Parameters

Parameter Value

Commercial Operation Date 6/71

Capacity (MWe) 616

BWR Type 3

Drains Path Cascaded

Condensate Polishing Filter Demineralizer

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1

Monticello Milestones

Milestone events for Monticello are given in Table V-2.

A 1989 chemical decontamination was performed on the recirculation piping and on the RHR
(Residual Heat Removal) system piping. This chemical decon, which was performed prior to the
start of zinc injection, resulted in the removal of 97.3 curies. HWC was initiated in 2/89,
followed by NZO in 11/89. The 1991 chemical decontamination was performed on the
recirculation, RHR, and RWCU systems, and resulted in the removal of 114 curies, including 36
curies of Zn-65. The 1993 chemical decontamination was performed on the recirculation and
RHR piping, and 228 curies were removed, including 173 curies of Zn-65. DZO began in 4/93.
Pleated septa were installed in one of five F/D vessels in 1996, extended to three of five in 1997
and to all vessels in 1999. A power uprate was implemented in 10/98. The power uprate
provided an increase from 1670 MWth to 1775 MWth (576 MWe to 616 MWe).

V-1
 EPRI Licensed Material

Monticello

Table V-2
Monticello Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 10/ 98

Condenser
1984
Retube

Recirc. Pipe
1984
Replacement

RWCU Pipe
Replacement

Extraction
Steam Pipe
Replacement

8/87
Chem.
9/89 2/93
Decon.
4/91

2/89 → → → → → →
HWC (scfm) (15- → → → →
40) (40) (40) (40) (40) (35) (35)

NMCA

NZO 11/89 →

DZO 4/93 → → → → → → → → →

Iron Injection

Crud Resins

Pleated
2/96 → → → → → →
Filters

Radiation Data

Recirculation System dose rates are summarized in Table V-3.

V-2
 EPRI Licensed Material

Monticello

Table V-3
Monticello Recirculation System Dose Rates

Monticello – Recirculation System Dose Rates (mR/hr)

May-86 Nov-87 Apr-89 Sep-89 Sep-89 Jul-90 Sep-90 Apr-91 Apr-91
(1) (2) (1) (2)
EFPY 1.16 2.34 3.65 4.00 4.00 4.65 4.83 5.35 5.35
BRAC 199 356 613 758 21 258 300 613 21
A Suction
B Suction
A Discharge
B Discharge
Avg Risers 1050 157 840 50

Monticello – Recirculation System Dose Rates (mR/hr)

Jan-92 Apr-92 Feb-93 Feb-93 Apr-94 Jun-94 Sep-94 Apr-96 May-97
EFPY 5.95 6.24 7.02 7.02 8.06 9.26 10.72 12.19
BRAC 273 478 467 (3) 38 (4) 300 250 231 231 215
A Suction 300
B Suction 200
A Discharge 200
B Discharge 225
Avg Risers 700 960 1400 475 555 475 345 380

Monticello – Recirculation System Dose Rates (mR/hr)

Mar-98 Apr-99 Jan-00 Feb-01 Nov-01
EFPY
BRAC 216 288 338 325 278
A Suction 750 650
B Suction 150 100
A Discharge 200 160
B Discharge 250 200
Avg Risers 375 250 443 354
Table V-3 Notes
1. Pre-decon
2. Post-decon
3. 1350 if B discharge hot spot is included.
4. 114 if B discharge hot spot is included.

V-3
 EPRI Licensed Material

Monticello

Trend Data

Power, feedwater iron, reactor water anions, reactor water cobalt-60, and BRAC history trend
plots for Monticello are presented in Figures V-1, V-2, V-3, V-4, and V-5 respectively.

100

80
Power (%)

60

40

20

0
12/1/96 12/1/97 12/1/98 12/1/99 11/30/00 11/30/01 11/30/02 11/30/03

Figure V-1
Power History, Monticello

Feedwater Iron Control

Following a gradual increasing trend in feedwater iron from 1997 to 1999, feedwater iron at
Monticello decreased significantly in 2000 after extending the use of pleated precoatable septa to
all filter/demineralizer vessels. Average insoluble iron peaked in 1999 at 1.87 ppb, decreased to
0.88 ppb in 2000 and was 1.1 ppb in 2001. Both insoluble nickel and chromium also peaked in
1999.

V-4
 EPRI Licensed Material

Monticello

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001
12/1/96 12/1/97 12/1/98 12/1/99 11/30/00 11/30/01 11/30/02 11/30/03

Insoluble Fe Soluble Fe

Figure V-2
Feedwater Iron, Monticello

Reactor Water Sulfate Control

Reactor water sulfate at Monticello increased from 1.37 ppb in 1999 and 1.49 ppb in 2000 to
3.25 ppb in 2001 and 2.38 ppb for 2002 (through mid-July). The increase in early 2001 is
attributed to pump bearing oil ingress. After replacement of oil contaminated septa, sulfate
values returned to about 1.7 ppb. There is a possibility that the increase in 2000 as shown in
Figure V-3 is related to aging effects of pleated septa. Monticello is one of only 3 Filter
Demineralizer plants using pleated septa in all vessels.

V-5
 EPRI Licensed Material

Monticello

10 100

9 90

8 80
Reactor Water Anions (ppb)

7 70

6 60

Power (%)
5 50

4 40

3 30

2 20

1 10

0 0
1/1/00 7/19/00 2/4/01 8/23/01 3/11/02 9/27/02

Cl NO3 SO4 Power

Figure V-3
Reactor Water Anions, Monticello

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
12/1/96 12/1/97 12/1/98 12/1/99 11/30/00 11/30/01 11/30/02 11/30/03

Insoluble Co-60 Soluble Co-60

Figure V-4
Reactor Water Cobalt-60, Monticello

V-6
 EPRI Licensed Material

Monticello

1400 15-30 scfm 45 scfm

HWC
1200 NZO DZO

1000
Dose Rate (mR/hr)

Pleated filters
Retube condenser, recirc pipe repl.

CRB replacement mostly complete

800

600
Replace fw reg valves

400

Power uprate
Chem decon

Chem decon

Chem decon

Chem decon
200

0
Jun- Feb- Nov- Aug- May- Feb- Nov-
BRAC Milestones Co-60 Dose B disch hot spot included

Figure V-5
BRAC History, Monticello

Recirculation Piping Dose Rates

Monticello initiated both natural zinc injection and hydrogen water chemistry in 1989. The plant
switched to DZO in April 1993. Recirculation piping dose rates were high prior to the start of
HWC and zinc injection, and then continued to increase, with the BRAC average reaching 467
mR/hr prior to the chemical decontamination in 1993. Following the 1993 decon and the start of
DZO, dose rates peaked at 300 mR/hr in 4/94, after which they decreased and stabilized at
approximately 230 mR/hr for a number of years. BRAC dose rates increased to 338 mR/hr
following the 10/98 power uprate, but have since decreased to 278 mR/hr.

Average soluble Co-60 was 6.35E-4 µCi/ml in 1998, and then decreased to 1.15E-4 µCi/ml in
1999 and to 6.85 E-5 µCi/ml in 2000. The 2001 average soluble Co-60 was 5.78E-5 µCi/ml.
The 1998 average concentration is one of the higher soluble Co-60 values reported by BWRs.

Recirculation Piping Gamma Scans

Recirculation piping gamma scan results are summarized in Table V-4. While total activity
reached almost 70 µCi/cm2 in 2/93, a major contribution was from Zn-65. Since switching from
NZO to DZO in 4/93, both the total activity and the contribution from Zn-65 have decreased,
while the contribution from Co-60 has increased.

V-7
 EPRI Licensed Material

Monticello

Table V-4
Monticello Recirculation Piping Gamma Scan Results

5/86 11/87 9/89 7/90 9/90 4/91 1/92 4/92

Total Activity
9.7 15.5 27.3 19.6 13.4 23.5 26.1 56.9
(µCi/cm2)

% Co-60 63 73 84 41 63 55 11 9

% Co-58 10 7 3 7 8 5 3 3

% Mn-54 13 12 9 3 5 2 1

% Zn-65 13 8 4 17 23 38 38 45

% Cr-51 32 48 43

2/93 6/94 10/94 4/96

Total Activity
67.4 30.2 25.1 23.6
(µCi/cm2)

% Co-60 10 17 24 36

% Co-58 2 3 3 3

% Mn-54 <1 <1 <1 2

% Zn-65 81 66 73 42

% Cr-51 7 14 17

Stellite™ Reduction

Stellite™ feedwater regulating valves were replaced in 1987. As of December, 1997, most
Stellite™ control rod blades were reported to have been replaced.

V-8
 EPRI Licensed Material

W
NINE MILE POINT 1

Table W-1
Nine Mile Point 1 Plant Design Parameters

Parameter Value

Commercial Operation Date 4/69

Capacity (MWe) 610

BWR Type 2 (Non-jet pump)

Drains Path Cascaded

Condensate Polishing Deep Bed

RWCU Capacity (% Feedwater Flow), Normal/Maximum 2.74/5.14

Nine Mile Point Milestones

Milestone events for Nine Mile Point 1 are given in Table W-2.

NMCA was performed in 05/00 in a mid cycle outage and hydrogen injection was started in
06/00. For the first year of operation under NMCA and HWC, the hydrogen injection rate was 8
scfm. Following the spring 2001 refuel outage, the hydrogen injection rate was reduced to 4
scfm. This was done in an effort to minimize further increases in drywell dose rates following
HWC initiation with NMCA. DZO implementation is scheduled to be completed before the
spring 2003 refuel outage.

W-1
 EPRI Licensed Material

Nine Mile Point 1

Table W-2
Nine Mile Point 1 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power
Uprate

Condenser
Retube

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction
Steam Pipe
Replacement

Chem.
Decon.

HWC (scfm) 6/00 5/01

→
(8) (4)

NMCA 5/00

NZO

DZO

Iron Injection

Crud Resins X X → →

Pleated
Filters

Radiation Data

Recirculation System dose rates are summarized in Table W-3.

W-2
 EPRI Licensed Material

Nine Mile Point 1

Table W-3
Nine Mile Point 1 Recirculation Piping Dose Rates

Nine Mile Point 1 – Recirculation System Dose Rates (mR/hr)

Jan - Jun- Nov- Mar- Apr- Mar- Apr- Apr- Sep-
84 86 88 93 94 95 97 99 00
EFPY
BRAC 250 312 337 276 299 244 250 235 175
A Suction 300 344 450 300 325 280 280 250 130
A Discharge 130 226 300 200 240 180 200 220 115
B Suction 290 358 370 300 350 300 280 250 275*
B Discharge 310 383 377 360 350 280 300 220 175
C Suction 310 372 362 300 325 280 300 250 175
C Discharge 310 388 379 300 350 300 280 300 160
D Suction 120 174 200 200 200 200 180 220 275*
D Discharge 325 398 400 300 300 240 280 240 150
E Suction 275 290 316 300 300 200 200 200 150
E Discharge 130 190 212 200 250 180 200 200 145
*may be related to shutdown cooling loop modifications

Nine Mile Point 1 – Recirculation System Dose Rates (mR/hr)

Mar Mar Apr Apr Apr Apr Aug May May
18-01 28-01 06-01 12-01 23-01 30-01 24-01 15-02 17-02
EFPY
BRAC 710 698 675 630 630 638 905 1055 1000
A Suction 700 600 650 600 600 600 1000 1150 1100
A Discharge 550 550 550 600 500 500 800 1000 850
B Suction 800 750 700 600 700 700 850 900 900
B Discharge 850 800 800 700 800 800 1100 1200 1100
C Suction 700 700 700 700 600 650 900 1000 1000
C Discharge 800 800 750 600 700 750 1050 1200 1200
D Suction 800 875 750 700 650 650 850 1000 1000
D Discharge 750 700 750 700 700 700 950 1200 1100
E Suction 600 650 600 600 550 550 800 1000 950
E Discharge 550 550 500 500 500 475 750 900 800

W-3
 EPRI Licensed Material

Nine Mile Point 1

Trend Data

Power, feedwater iron, reactor water anions, reactor water cobalt-60 and BRAC history trend
plots for Nine Mile Point 1 are presented in Figures W-1, W-2, W-3, W-4, and W-5 respectively.

Feedwater Iron Control

The use of crud removal resins for improved feedwater iron control was first implemented at the
station in 1995 for a short period, followed by re-implementation in 2000. Since re-
implementation, the station has used a combination of crud removal resin beds and standard resin
or high cross-linked resin (12% cross-linkage) beds. The typical configuration consists of four
beds of crud removal resins and two beds of standard resins or high cross-linked resin beds.

Insoluble feedwater iron has shown a declining trend since re-implementation of crud removal
resins. Insoluble feedwater iron averaged 5.24 ppb for 1999, 3.15 ppb for 2000, and 1.61 ppb for
2001. The station reports recent spiking of feedwater iron following condensate temperature
variations that are caused by upwelling of the cooling water supply (Lake Ontario).

Reactor Water Sulfate Control

The station has observed increases in reactor coolant sulfate over time due to the use of the crud
removal resins. An increase in sulfate to levels slightly above 2 ppb was observed in the winter
of 2001-2002 that resulted in the station replacing the crud removal beds after about one year of
service prior to the spring of 2002. The sulfate increase was observed even with operation of the
RWCU system near rated capacity (5.14% of total feedwater flow).

W-4
 EPRI Licensed Material

Nine Mile Point 1

110
100
90
80
70
Power (%)

60
50
40
30
20
10
0
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

NMCA

Figure W-1
Power History, Nine Mile Point 1

100

10
Feedwater Fe (ppb)

0.1

0.01

0.001
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Insoluble Fe Soluble Fe NMCA

Figure W-2
Feedwater Iron, Nine Mile Point 1

W-5
 EPRI Licensed Material

Nine Mile Point 1

5.0 100
4.5 90
4.0 80
Reactor Water Anions (ppb)

3.5 70
3.0 60

Power (%)
2.5 50
2.0 40
1.5 30
1.0 20
0.5 10
0.0 0
1/1/00 7/19/00 2/4/01 8/23/01 3/11/02 9/27/02
Cl NO3 SO4 Power

Figure W-3
Reactor Water Anions, Nine Mile Point 1

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure W-4
Reactor Water Cobalt-60, Nine Mile Point 1

W-6
 EPRI Licensed Material

Nine Mile Point 1

1500
Natural zinc from condenser
Two hrs
after sd
1250
Dose Rate (mR/hr)

1000
Recirc pipe replacement 1982-83
Decon prior to pipe replacement

HWC 8 scfm 4 scfm

750
NMCA
500
Partial decon

250

0
1/1/81 4/15/84 7/29/87 11/10/90 2/22/94 6/6/97 9/18/00

BRAC BRAC (all measurements) Milestones Co-60 Dose

Figure W-5
BRAC History, Nine Mile Point 1

Recirculation Piping Dose Rates

Pipe dose rates were relatively stable, ranging from 200 to 300 mR/hr from 1984 through 1999.
Dose rates were at their lowest (175 mR/hr) in 09/00 immediately following NMCA. However,
in 03/01 when Nine Mile Point 1 shut down for a refueling outage, the BRAC dose rates
increased to of 710 mR/hr, decaying to 638 mR/hr six weeks later. A further increase to 905
mR/hr was measured 57 hours after an unplanned shutdown in 08/01. The measured dose rate
15 hours after a scheduled short outage in 05/02 was 1055 mR/hr. Average reactor water soluble
Co-60 activity was 1.89E-4 µCi/ml in 1999, 1.60E-4 µCi/ml in 2000, and 6.3E-4 µCi/ml in 2001.
Average insoluble Co-60 activity was 3.31E-5 µCi/ml in 1999, 7.97E-5 µCi/ml in 2000, and
3.78E-3 µCi/ml in 2001. The dose rate increase coincides with the soluble and insoluble
increases in reactor coolant Co-60.

Recirculation Piping Gamma Scans Pipe

Gamma scan data for Nine Mile Point 1 are summarized in Table W-4.

Pipe gamma scan data show that the total activity in the corrosion film did not change
significantly between 11/88 and 09/00. During that time the Co-60 contribution decreased
slightly. The gamma scan data from 03/01 showed a significant increase in the total activity on
the piping. Although the activity due to Co-60 increased, the percentage of Co-60 has decreased

W-7
 EPRI Licensed Material

Nine Mile Point 1

corresponding with increases in Co-58 and Fe-59 in the corrosion film. The BRAC dose rate
measurement in 03/01 was about 700 mR/hr. The dose from Co-60 was about 60% of this value.

Table W-4
Nine Mile Point 1 Recirculation Piping Gamma Scan Results

Date 11/88 9/00 3/01

Total Activity
14.4 14.2 64.7
(µCi/cm2)

% Co-60 87 64 37

% Co-58 10 34

% Mn-54 12 26 13

% Fe-59 15

W-8
 EPRI Licensed Material

X
NINE MILE POINT 2

Table X-1
Nine Mile Point 2 Plant Design Parameters

Parameter Value

Commercial Operation Date 4/88

Capacity (MWe) 1207

BWR Type 5

Drains Path Forward Pumped

Condensate Polishing Deep Bed

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1.75/2

Nine Mile Point 2 Milestones

Milestone events for Nine Mile Point 2 are given in Table X-2.

Nine Mile Point 2 started NZO injection in 1988, when the plant began commercial operation,
and switched to DZO in 7/99. A 5% power uprate to 3467 MWth was implemented in 9/94. A
failure of the moisture separator reheater in 6/97 caused hotwell iron concentrations to increase,
leading to increased feedwater iron. Controlled trials to requalify one type of crud removal resin
began in 1/98 through an EPRI Tailored Collaboration project. NMCA was performed in 9/00
and hydrogen injection was started in 2/01.

X-1
 EPRI Licensed Material

Nine Mile Point 2

Table X-2
Nine Mile Point 2 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 9/94

Condenser
Retube

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction
Steam Pipe
Replacement

Chem. Decon.

HWC 2/01 →
(10) (15)

NMCA 9/00

NZO 4/88 → → → → → → →

DZO 7/99 → → →

Iron Injection

Crud Resins 8/93 X 1/98 → → → →

Pleated Filters

Radiation Data

Recirculation System dose rates are summarized in Table X-3.

X-2
 EPRI Licensed Material

Nine Mile Point 2

Table X-3
Nine Mile Point 2 Recirculation Piping Dose Rates

Nine Mile Point 2 – Recirculation System Dose Rates (mR/hr)

Nov- Nov- Mar- Oct- Apr- Sep- May- Nov- Apr-

88 90 92 93 95 96 98 98 99

EFPY 0.45 1.4 2.25 3.4 4.6 5.75 6.9 7.1 7.1

BRAC 24 118 164 186 222 230 243

A Suction 30 135 182 205 275 250 270

B Suction 20 140 152 206 225 250 250 225 180

A Discharge 18 98 183 177 200 200 250

B Discharge 27 99 138 158 190 220 200 175 150

Avg Risers 36 199 283 331 379 353 364 350 305

Nine Mile Point 2 – Recirculation System Dose Rates (mR/hr)

Jul- Mar- Sep- Nov- Feb- May- Jul- Mar-

99 00 00 00 01 01 01 02

EFPY 7.5 8.1 8.4 8.6 8.9 9.0 9.2 9.7

BRAC 179 198 163 129 68 110 226

A Suction 200 190 190 155 125 145 275

B Suction 190 190 200 170 140 100 120 275

A Discharge 175 175 140 100 20 85 175

B Discharge 150 150 225 150 120 25 90 180

Avg Risers 305 299 235

Trend Data

Power, feedwater iron, reactor water anions, reactor water cobalt-60, and BRAC history trend
plots for Nine Mile Point 2 are presented in Figures X-1, X-2, X-3, X-4, and X-5 respectively.

X-3
 EPRI Licensed Material

Nine Mile Point 2

120
110
100
90
80
Power (%)

70
60
50
40
30
20
10
0
7/1/98 7/1/99 6/30/00 6/30/01 6/30/02 6/30/03

NMCA

Figure X-1
Power History, Nine Mile Point 2

Feedwater Iron Control

The use of low cross-linked crud removal resins, implemented under an EPRI Tailored
Collaboration project, resumed in January 1998. Nine demineralizers are required for 100%
power operation. The station typically operates with crud removal resins in five demineralizers
and standard resins in four demineralizers. The use of the crud removal resins in conjunction
with improvements in resin cleaning (URC) has resulted in a decline in feedwater iron. The
average feedwater insoluble iron concentration was 2.17 ppb in 1999, 1.61 ppb in 2000, and 1.41
ppb in 2001. Condensate demineralizer effluent concentrations averaged 5.95 ppb in 1999, 3.12
ppb in 2000, and 1.60 ppb in 2001. The forward pumped heater drains averaged 1.14 ppb in
1999, 0.71 ppb in 2000, and 0.70 ppb in 2001. The forward pump heater drains act to dilute the
iron content of the demineralizer effluent.

A seasonal effect in feedwater iron is apparent as iron tends to be lower in the summer months
and higher in the winter months.

Reactor Water Sulfate Control

The station has observed frequent sulfate spikes over the past few years, many of which appear
to be as a result of power transients. Sulfate spikes have also been observed when beds are
placed in service. The station identified a deficiency in a piping design for an air/sluice line that

X-4
 EPRI Licensed Material

Nine Mile Point 2

allows the accumulation of a small amount of resin in a dead leg as a source of resin when beds
are placed in service.

100

10
Feedwater Fe (ppb)

0.1

0.01

0.001
7/1/98 7/1/99 6/30/00 6/30/01 6/30/02 6/30/03
Insoluble Fe Soluble Fe NMCA

Figure X-2
Feedwater Iron, Nine Mile Point 2

X-5
 EPRI Licensed Material

Nine Mile Point 2

5.0 100
4.5 90
4.0 80
Reactor Water Anions (ppb)

3.5 70
3.0 60

Power (%)
2.5 50
2.0 40
1.5 30
1.0 20
0.5 10
0.0 0
1/1/00 7/19/00 2/4/01 8/23/01 3/11/02 9/27/02
Cl NO3 SO4 Power

Figure X-3
Reactor Water Anions, Nine Mile Point 2

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
7/1/98 7/1/99 6/30/00 6/30/01 6/30/02 6/30/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure X-4
Reactor Water Cobalt-60, Nine Mile Point 2

X-6
 EPRI Licensed Material

Nine Mile Point 2

300 NZO DZO

250
Crud resins Crud resins
Dose Rate (mR/hr)

Moisture separator reheater failure

200

150

Replace 12 CRBs

Replace 12 CRBs
100

Power uprate
NMCA

50
HWC

0
Jun-84 Feb-87 Nov-89 Aug-92 May-95 Feb-98 Nov-00 Aug-03

BRAC Milestones Co-60 Dose

Figure X-5
BRAC History, Nine Mile Point 2

Recirculation Piping Dose Rates

Historical BRAC data showed steadily increasing dose rates from the initial plant operation to
242 mR/hr in 1998. Nine Mile Point 2 had been adding natural zinc oxide since the initial
startup and switched to DZO in July 1999. NMCA was implemented in 9/00, but HWC was not
started until 2/01, a month before a refuel outage. With limited time on NMCA and HWC, the
BRAC dose rate average in the 3/01 refuel outage was 68 mR/hr. The BRAC dose rate in 3/02,
following a complete cycle of NMCA with HWC, was 226 mR/hr. Both soluble and insoluble
Co-60 activity increased during the first full cycle with NMCA and HWC. The average reactor
coolant soluble Co-60 activity was 7.12E-5 µCi/ml in 2000 and 2.11E-4 µCi/ml in 2001. The
average reactor coolant insoluble Co-60 activity was 6.13E-5 µCi/ml in 2000 and 2.38E-4
µCi/ml in 2001.

Recirculation Piping Gamma Scans

Nine Mile Point 2 gamma scan results are summarized in Table X-4. There are no gamma scan
data reported after 1993.

X-7
 EPRI Licensed Material

Nine Mile Point 2

Table X-4
Nine Mile Point 2 Recirculation Piping Gamma Scan Results

Nov-88 Nov-90 Mar-92 Jan-93

Total Activity (µCi/cm2)

3.1 11.9 13.7 16.8

% Co-60 19 34 40 38

% Co-58 29 11 8 4

% Mn-54 13 21 17 25

% Zn-65 32 30 32 29

The piping total activity was still on an increasing trend at the time of the 1/93 gamma scan; this
trend is consistent with the BRAC dose rate trend. The gamma scan data also indicate that
approximately one-third of the recirculation piping activity during this period when NZO was
being injected was from Zn-65. The Zn-65 percentage is relatively constant while the fraction of
Co-60 in the piping corrosion film increased from 19% to 38%.

Stellite™ Reduction

Nine Mile Point 2 has reported replacement of 24 control rod blades. Twelve blades were
replaced in 1995 and an additional 12 were replaced in 1996.

X-8
 EPRI Licensed Material

Y
OYSTER CREEK

Table Y-1
Oyster Creek Plant Design Parameters

Parameter Value

Commercial Operation Date 12/69

Capacity (MWe) 640

BWR Type 2 (Non-Jet Pump)

Drains Path Cascaded

Condensate Polishing Deep Bed

RWCU Capacity (% Feedwater Flow), Normal/Maximum 2.6/4.6%

Oyster Creek Milestones

Milestone events for Oyster Creek are given in Table Y-2.

The original aluminum-bronze condenser tubes were replaced with titanium tubes in 1975. The
1986 chemical decontamination of the recirculation piping removed approximately 55.3 curies.
The 1991 decon of the recirculation piping removed 38.3 curies, and a concurrent
decontamination of the RWCU piping removed 8.3 curies. Hydrogen injection was started in
1992. New Japanese resin cleaning technology was implemented in 1/93. DZO was
implemented in 7/00, and NMCA was performed in 10/02.

Y-1
 EPRI Licensed Material

Oyster Creek

Table Y-2
Oyster Creek Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power
Uprate

Condenser
1975
Retube

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction
Steam Pipe
Replacement

Chem. 5/86
Decon. 4/91

2/92 → → → → → →
HWC (scfm) (12) → → → (8)
(5) (10.6) (10) (8-10) (10) (10) (11.5)

NMCA 10/02

NZO

DZO 7/00 → →

Iron Injection

Crud Resins 12/93 → 12/95

Pleated
Filters

Radiation Data

Recirculation System dose rates are summarized in Table Y-3.

Y-2
 EPRI Licensed Material

Oyster Creek

Table Y-3
Oyster Creek Recirculation System Dose Rates

Oyster Creek – Recirculation System Dose Rates (mR/hr)

Mar- May- Apr- Feb- May- Jun- Nov- Sep- Sep- Nov-
86 86 90 91 91 92 92 94 96 98
EFPY 9.0 9.0 10.9 11.6 11.6 12.5 12.9 14.4 16.1 18.1
BRAC 628 353 282 256 40.7 188 188 294 389 391
A Suction 300 300 10 200 200 300 400 430
A Discharge 300 320 20 180 200 300 400 440
B Suction 280 200 12 200 180 300 400 420
B Discharge 320 200 8 160 180 200 400 220
C Suction 300 280 240 220 180 320 380 420
C Discharge 260 260 24 180 180 280 350 410
D Suction 260 280 15 160 180 320 380 400
D Discharge 240 260 18 180 200 300 400 360
E Suction 300 240 20 220 200 300 400 400
E Discharge 260 220 40 180 180 320 380 410

Oyster Creek – Recirculation System Dose Rates (mR/hr) (continued)

Jan- Aug- Nov- May-
00 00 00 01
EFPY 19.3 19.9 20.1
BRAC 390 404 432 390
A Suction 420 420 420 340
A Discharge 400 400 440 360
B Suction 420 400 460 440
B Discharge 200 360 460 420
C Suction 400 360 360 360
C Discharge 410 400 440 360
D Suction 410 420 440 440
D Discharge 430 440 480 420
E Suction 410 440 410 380
E Discharge 400 400 410 380

Y-3
 EPRI Licensed Material

Oyster Creek

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots are presented for
Oyster Creek in Figures Y-1, Y-2, Y-3 and Y-4, respectively.

100

80
Power (%)

60

40

20

0
12/1/98 12/1/99 11/30/00 11/30/01 11/30/02 11/30/03

Figure Y-1
Power History, Oyster Creek

Feedwater Iron Control

In 1993, Oyster Creek installed resin cleaning technology from Japan to improve removal of
crud and resin fines. In addition, the plant applied low cross-linked resins for enhanced crud
removal in two of the seven condensate demineralizer service vessels for a 2-year period from
December 1993 through December 1995, until the increase in reactor water sulfate concentration
became unacceptable. The plant returned to using conventional resins, with the exception of one
bed of less separable resins (Dow C500ES/SBR-C), which also had a lower void fraction and
greater cation resin surface area than the conventional resins. During the October 1998 refuel
outage Oyster Creek replaced all seven of the condensate beds with the Dow C500ES/SBR-C
less separable resins. The current resin mix at Oyster Creek includes 5 beds of Dow 575C/SBR-
C and 2 beds of Dow HGR-W2/SBR-C. The two beds with HGR-W2 will be replaced with
Dow 575C at their next scheduled replacement. The Dow 575C cation resin is a 12%
crosslinked gel.
The average feedwater total iron concentration was 2.55 ppb in 1999, 3.29 ppb in 2000, and 2.83
ppb in 2001. The 2000 average insoluble iron was higher than 2001 and in previous years,

Y-4
 EPRI Licensed Material

Oyster Creek

influenced by the four plant startups and the effects of resin changeouts due to condenser
inleakage (aged resin provide much better iron removal than new resins at Oyster Creek).

100

10
Feedwater Fe (ppb)

0.1

0.01

0.001
12/1/98 12/1/99 11/30/00 11/30/01 11/30/02 11/30/03

Insoluble Fe Soluble Fe

Figure Y-2
Feedwater Iron, Oyster Creek

Y-5
 EPRI Licensed Material

Oyster Creek

Reactor Water Co-60 (µCi/ml) 1.E-02

1.E-03

1.E-04

1.E-05

1.E-06
12/1/98 12/1/99 11/30/00 11/30/01 11/30/02 11/30/03

Insoluble Co-60 Soluble Co-60 Total Co-60

Figure Y-3
Reactor Water Cobalt-60, Oyster Creek

600
5 scfm 12 scfm 10.6 scfm
HWC
500
Replace CRBs, LP turbine blades 11/92
Chem decon; replace CRBs, LP turbine

Crud resins DZO

new resin cleaning technology 1/93
Dose Rate (mR/hr)

400
blades, valves

300
Retube condenser 1975

Replace CRBs, valves

Replace valve plugs

200
Replace CRBs

Replace CRBs

Replace CRBs
Chem decon

100

0
Jun-84 Feb-87 Nov-89 Aug-92 May-95 Feb-98 Nov-00

BRAC Milestones

Figure Y-4
BRAC history, Oyster Creek

Y-6
 EPRI Licensed Material

Oyster Creek

Recirculation Piping Dose Rates

Piping dose rates at Oyster Creek are typically greater than 250 mR/hr. The last chemical
decontamination in 1991 reduced the dose rates to 40 mR/hr, but the piping recontaminated with
dose rates increasing to 390 mR/hr by September 1996. The BRAC dose rate remained near this
level until 11/00 when the dose rate increased to 432 mR/hr. The possible causes of this increase
are the multiple plant shutdowns and restarts in 2000 and a fuel failure that occurred in July
2000. Oyster Creek also increased hydrogen injection in response to the fuel failure. The 5/01
measurement decreased to 390 mR/hr.

Average reactor water soluble Co-60 activity was 8.38E-5 µCi/ml in 1999, 1.30E-4 µCi/ml in
2000, and 5.79E-5 µCi/ml in 2001. Average reactor water insoluble Co-60 activity was 8.66E-6
µCi/ml in 1999, 5.45E-5 µCi/ml in 2000, and 7.49E-6 µCi/ml in 2001. Both soluble and
insoluble reactor water Co-60 increased in 2000 and were affected by the four plant outages
during the year.

Stellite™ Reduction

The Stellite™ content of primary, steam, condensate, and feedwater system materials at Oyster
Creek (BWR 2) is approximately 40 % higher than for a typical BWR 4. Between 1987 and
1997, the Stellite™ surface area at Oyster Creek was reduced by approximately 50%, from 177
ft2 to 91 ft2, through replacement of control rod blade components. Total Co-59 input to the
reactor from the feedwater has been reduced by greater than 50% through replacement of low
pressure rotor blades, valve replacements and improvements in iron control.

Y-7
 EPRI Licensed Material

Z
PEACH BOTTOM 2

Table Z-1
Peach Bottom 2 Plant Design Parameters

Parameter Value

Commercial Operation Date 7/74

Capacity (MWe) 1159

BWR Type 4

Drains Path Cascaded

Condensate Polishing Filter Demineralizer

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1.2

Peach Bottom 2 Milestones

Milestone events for Peach Bottom 2 are given in Table Z-2.

The original 304 stainless steel recirculation piping was replaced in 1985 with 316NG stainless
steel. The plant performed a chemical decontamination of the original recirculation piping prior
to the replacement. There have been no chemical decontaminations performed on the new
recirculation piping. The original admiralty brass condenser tubes were replaced in 1991 with
titanium tubes. NZO was initiated in 6/91 and the station switched to DZO in 10/96. A 7%
power uprate was implemented in 10/96. HWC was initiated in 5/97. NMCA was performed in
10/98. The total mass of noble metal added at Peach Bottom 2 was 4.53 kg, the largest of any
application to date. The calculated fuel deposition was 61.5 µg/cm2.

Z-1
 EPRI Licensed Material

Peach Bottom 2

Table Z-2
Peach Bottom 2 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 10/96

Condenser
1991
Retube

Recirc. Pipe
1985
Replacement

RWCU Pipe
Replacement

Extraction Steam
Pipe Replacement

Chem. Decon. 8/84

→
5/97 → →
HWC (scfm) (10- → →
(13) (6) (10)
13)

NMCA 10/98

NZO 6/91 → → →

DZO 10/96 → → → → → →

Iron Injection

Crud Resins

Pleated Filters 5/96 → → → → → →

Radiation Data

Recirculation System dose rates are summarized in Table Z-3.

Z-2
 EPRI Licensed Material

Peach Bottom 2

Table Z-3
Peach Bottom 2 Recirculation System Dose Rates

Peach Bottom 2 – Recirculati on System Dose Rates (mR/hr)

Mar-91 Nov-92 Oct-94 Sep-96 Sep-98 Sep-00

EFPY 2.7 3.5 5.0 6.8 8.6 10.4

BRAC 109 114 128 111 131 214

A Suction 90 130 135 117 150 270

B Suction 110 105 112 107 104 215

A Discharge 115 90 117 124 187

B Discharge 120 130 137 105 145 185

Avg Risers 173 221 279 190 230 348

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Peach Bottom 2
are presented in Figures Z-1 through Z-7.

Z-3
 EPRI Licensed Material

Peach Bottom 2

110
100
90
80
70
Power (%)

60
50
40
30
20
10
0
9/23/94 4/11/95 10/28/95 5/15/96 12/1/96 6/19/97 1/5/98 7/24/98
Power NMCA

Figure Z-1
Power History pre-NMCA, Peach Bottom 2

110
100
90
80
70
Power (%)

60
50
40
30
20
10
0
9/1/98 9/1/99 8/31/00 8/31/01 8/31/02 8/31/03
Power NMCA

Figure Z-2
Power History post-NMCA, Peach Bottom 2

Z-4
 EPRI Licensed Material

Peach Bottom 2

Feedwater Iron Control

Average feedwater insoluble iron was 1.70 ppb in 2000 and 1.27 ppb in 2001. The station began
the implementation of pleated septa in 1996. The majority of the filter/demineralizers now
contain pleated septa.

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001
9/23/94 4/11/95 10/28/95 5/15/96 12/1/96 6/19/97 1/5/98 7/24/98
Insoluble Fe Soluble Fe NMCA

Figure Z-3
Feedwater Iron pre-NMCA, Peach Bottom 2

Z-5
 EPRI Licensed Material

Peach Bottom 2

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001
9/1/98 9/1/99 8/31/00 8/31/01 8/31/02 8/31/03

Insoluble Fe Soluble Fe NMCA

Figure Z-4
Feedwater Iron post-NMCA, Peach Bottom 2

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06

1.E-07
9/23/94 4/11/95 10/28/95 5/15/96 12/1/96 6/19/97 1/5/98 7/24/98

Insoluble Co-60 Soluble Co-60 NMCA

Figure Z-5
Reactor Water Cobalt-60 pre-NMCA, Peach Bottom 2

Z-6
 EPRI Licensed Material

Peach Bottom 2

Reactor Water Co-60 (µCi/ml)

1.E-02

1.E-03

1.E-04

1.E-05

1.E-06

1.E-07
9/1/98 9/1/99 8/31/00 8/31/01 8/31/02 8/31/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure Z-6
Reactor Water Cobalt-60 post-NMCA, Peach Bottom 2

Recirculation Piping Dose Rates

Pipe dose rates were relatively constant between 109 and 131 mR/hr from 1991 to 1998. NMCA
was performed in October 1998, and the BRAC average dose rate increased to 214 mR/hr in 9/00
after the first cycle of HWC and NMCA.

The average reactor water soluble Co-60 activity was 1.69E-4 µCi/ml in 2000 and 5.96E-5
µCi/ml in 2001. The average reactor water insoluble Co-60 activity was 8.83E-5 µCi/ml in 2000
and 6.56E-6 µCi/ml in 2001.

Z-7
 EPRI Licensed Material

Peach Bottom 2

250
NZO DZO

200 Pleated filters

Dose Rate (mR/hr)

HWC
Recirc pipe replacement 1985

150

Retube condenser 1991

`
100
NMCA

Power Uprate
Chem decon

50

0
Jun-84 Feb-87 Nov-89 Aug-92 May-95 Feb-98 Nov-00
BRAC Milestones Co-60

Figure Z-7
BRAC History, Peach Bottom 2

Recirculation Piping Gamma Scans

Gamma scan data for Peach Bottom 2 are summarized in Table Z-4.

The gamma scan following the first cycle of HWC and NMCA (9/00) shows an increase in total
activity after remaining relatively constant since 1991. Co-60 is the dominant isotope in the 9/00
scan. However, Zn-65 activity showed an increase even though DZO had been initiated in
10/96.
Table Z-4
Peach Bottom 2 Recirculation Piping Gamma Scan Results

3/91 10/94 9/96 10/98 9/00

Total Activity (µCi/cm2) 9.37 11.42 9.2 11.5 15.1

% Co-60 64 48 43 47 62

% Co-58 19 5 3 5 6

% Mn-54 9 34 29 35 9

% Zn-65 9 10 17 9 20

Z-8
 EPRI Licensed Material

AA
PEACH BOTTOM 3

Table AA-1
Peach Bottom 3 Plant Design Parameters

Parameter Value

Commercial Operation Date 12/74

Capacity (MWe) 1159

BWR Type 4

Drains Path Cascaded

Condensate Polishing Filter Demineralizer

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1.2

Peach Bottom 3 Milestones

Milestone events for Peach Bottom 3 are given in Table AA-2.

The original 304 stainless steel recirculation piping was replaced in 1988 with 316NG stainless
steel. The plant performed a chemical decontamination of the original recirculation piping prior
to the replacement. There have been no chemical decontaminations performed on the new
recirculation piping. The original admiralty brass condenser tubes were replaced in 1991 with
titanium tubes. NZO was initiated in 6/92 and the station switched to DZO in 10/96. A 7%
power uprate was implemented in 10/95. HWC was initiated in 3/97, and NMCA was performed
in 10/99.

AA-1
 EPRI Licensed Material

Peach Bottom 3

Table AA-2
Peach Bottom 3 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power
10/95
Uprate

Condenser
12/91
Retube

Recirc. Pipe
12/88
Replacement

RWCU Pipe
Replacement

Extraction
Steam Pipe
Replacement

Chem.
Decon.

3/97 → → → → →
HWC (scfm)
(13) (12) (12) (6-8) (8-11) (11)

NMCA 10/99

NZO 6/92 → → →

DZO 10/96 → → → → → →

Iron Injection

Crud Resins

Pleated
5/95 → → → → → → →
Filters

Radiation Data

Recirculation System dose rates are summarized in Table AA-3.

AA-2
 EPRI Licensed Material

Peach Bottom 3

Table AA-3
Peach Bottom 3 Recirculation System Dose Rates

Peach Bottom 3 - Recirculation System Dose Rates (mR/hr)

Nov-91 Nov-93 Oct-95 Oct-97 Oct-99 Sep-01

EFPY

BRAC 58 164 179 182.5 255 232

A Suction 175 240 200

B Suction 165 220 180

A Discharge 220 220 300

B Discharge 170 240 250

Avg Risers 282

Trend Data

Power, feedwater iron, reactor water cobalt-60, and BRAC history trend plots for Peach Bottom
3 are presented in Figures AA-1, AA-2, AA-3 and AA-4, respectively.

100

80
Power (%)

60

40

20

0
1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Power NMCA

Figure AA-1
Power History, Peach Bottom 3

AA-3
 EPRI Licensed Material

Peach Bottom 3

Feedwater Iron Control

Average feedwater insoluble iron was 1.98 ppb in 2000 and 1.91 ppb in 2001. The station began
the use of pleated septa in 1995. The majority of the filter/demineralizers now contain pleated
septa. Hotwell iron averaged 43 ppb in 1999, 29 ppb in 2000, and 43 ppb in 2001. The hotwell
iron average at Peach Bottom 3 is the highest reported among North American BWRs.

10

1
Feedwater Fe (ppb)

0.1
Y

0.01

0.001
1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Insoluble Fe Soluble Fe NMCA

Figure AA-2
Feedwater Iron, Peach Bottom 3

AA-4
 EPRI Licensed Material

Peach Bottom 3

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure AA-3
Reactor Water Cobalt-60, Peach Bottom 3

300

250
NZO DZO
Dose Rate (mR/hr)

200
Recirc pipe replacement 12/88

150
Retube condenser 12/91

Pleated filters

100
Power uprate

HWC

50 NMCA

0
Jun-84 Feb-87 Nov-89 Aug-92 May-95 Feb-98 Nov-00

BRAC Milestones Co-60 Dose

Figure AA-4
BRAC History, Peach Bottom 3

AA-5
 EPRI Licensed Material

Peach Bottom 3

Recirculation Piping Dose Rates

BRAC dose rates increased from the first reported value of 58 mR/hr in 11/91 to 182 mR/hr in
10/97. The 10/97 value of 182 mR/hr occurred seven months after HWC was initiated. The
BRAC dose rate increased to 255 mR/hr in 10/99 following the first full cycle of HWC. The
dose rate decreased to 232 mR/hr following the first full cycle of HWC and NMCA.

Average reactor water soluble Co-60 activity was 1.68E-4 µCi/ml in 2000 and 1.11E-4 µCi/ml in
2001. Average reactor water insoluble Co-60 activity was 4.51E-5 µCi/ml in 2000 and 4.33E-5
µCi/ml in 2001.

Recirculation Piping Gamma Scans

Peach Bottom 3 gamma scan results are summarized in Table AA-4.

Table AA-4
Peach Bottom 3 Recirculation Piping Gamma Scan Results

Date 11/91 11/93 10/95

Total Activity 5.2 11.1 17.1

(µCi/cm2)

% Co-60 54 54 56

% Co-58 15 8 4

% Mn-54 12 19 20

% Zn-65 17 17 18

The data show that Co-60 was the dominant isotope contributing to the surface activity in the
mid 1990s (no data after 10/95 is available). Zn-65 was also a major contributor since DZO had
yet to be implemented.

AA-6
 EPRI Licensed Material

BB
PERRY

Table BB-1
Perry Plant Design Parameters

Parameter Value
Commercial Operation Date 11/87
Capacity (MWe) 1306
BWR Type 6
Drains Path Forward Pumped
Condensate Polishing Filter + Deep Bed
RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1

Perry Milestones

Milestone events for Perry are given in Table BB-2.

NZO injection began in 1990 and the plant switched to DZO in 1997. From 1991 through 1994,
all condensate filters were upgraded with pleated septa. A chemical decontamination of the
recirculation piping, RWCU piping, and fuel pool cooling piping removed a total of 100 curies in
1996. A 5% power uprate was implemented in 6/00. NMCA was performed in 2/01, but
hydrogen injection was not initiated until 8/02.

BB-1
 EPRI Licensed Material

Perry

Table BB-2
Perry Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 6/00

Condenser Retube

Recirc. Pipe
Replacement

RWCU Pipe 8/92

Replacement partial

Extraction Steam
Pipe Replacement

Chem. Decon. 8/91 2/94 1/96

(partial)

HWC (scfm) 8/02

(10)

NMCA 2/01

NZO 2/90 → → → →

DZO → → → → → →

Iron Injection

Crud Resins

Pleated Filters 8/91 → → → → → → → → → →

BB-2
 EPRI Licensed Material

Perry

Radiation Data

Recirculation System dose rates are summarized in Table BB-3.

Table BB-3
Perry Recirculation System Dose Rates

Perry - Recirculation System Dose Rates (mR/hr)

Apr- Dec- May- Feb- Feb- 96 Mar- 96 Sep- Apr- Jun- Feb-
89 90 92 94 (1) (2) 97 99 00 01

EFPY 8.2 9.2

BRAC 100 149 173 231 178* 92** 248 250 205 245

A Suction 114 275 210 280

B Suction 132 300 250 300

A
78 225 180 200
Discharge

B
75 200 180 200
Discharge

Avg Risers 141 250 205

Table BB-3 Notes
1.Pre decon
2.Post-decon

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Perry are
presented in Figures BB-1, BB-2, BB-3 and BB-4, respectively.

BB-3
 EPRI Licensed Material

Perry

110
100
90
80
70
Power (%)

60
50
40
30
20
10
0
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Power NMCA

Figure BB-1
Power History, Perry

Feedwater Iron Control

Perry uses non-precoated pleated filters upstream of the deep bed condensate polishers. Total
feedwater iron averaged 0.34 ppb in 1999, 0.42 ppb in 2000, and 0.8 ppb in 2001. Condensate
polisher system effluent data show that the iron concentration in this stream is very low (<0.1
ppb).

BB-4
 EPRI Licensed Material

Perry

100

10
FW Fe (ppb)

0.1

0.01

0.001
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Insoluble Fe Soluble Fe NMCA

Figure BB-2
Feedwater Iron, Perry

1.E-01

1.E-02
Reactor Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure BB-3
Reactor Water Cobalt-60, Perry

BB-5
 EPRI Licensed Material

Perry

400
NZO DZO

350 Pleated filters

300
NMCA
Dose Rate (mR/hr)

RWCU pipe replacement (partial)

250

Begin CRB replacement 1995

Chem decon (partial) 8/91
200

150

Chem decon 1996

Power Uprate
100 Chem decon

50

0
Jan-89 Sep-91 Jun-94 Mar-97 Dec-99 Sep-02

BRAC Milestones Co-60 Dose Ion Chamber

Figure BB-4
BRAC History, Perry

Recirculation Piping Dose Rates

BRAC dose rates steadily increased during the first seven years of operation to 231 mR/hr in
2/94. The next BRAC measurement in 1996, prior to performing the chemical decontamination,
was 178 mR/hr. The dose rate increased to 250 mR/hr at the end of the first cycle following the
1996 chemical decontamination and has remained near this value over the next two
measurements.

The average reactor water soluble Co-60 activity was 7.76E-5 µCi/ml in 2000 and 2.64E-4
µCi/ml in 2001. The average reactor water insoluble Co-60 activity was 5.78E-5 µCi/ml in 2000
and 8.87E-4 µCi/ml in 2001.

Recirculation Piping Gamma Scans

Gamma scan data for Perry are summarized in Table BB-4.

The most recent reported measurements show that the total activity and Co-60 contribution have
essentially remained unchanged.

BB-6
 EPRI Licensed Material

Perry

Table BB-4
Perry Recirculation Piping Gamma Scan Results

Apr-89 Dec-90 May-92 Apr-99 Feb-01

Total Activity
5.45 6.48 13.2 11.0 11.0
(µCi/cm2)

% Co-60 53 56 59 69 66

% Co-58 26 11 11 7 16

% Mn-54 11 6 9 10 9.2

% Zn-65 5 26 21 11 5.1

Stellite™ Reduction

The station began replacing control rod blades with non-Stellite™ pins and rollers in 1995.

BB-7
 EPRI Licensed Material

CC
PILGRIM

Table CC-1
Pilgrim Plant Design Parameters

Parameter Value

Commercial Operation Date 12/72

Capacity (MWe) 687

BWR Type 3

Drains Path Cascaded

Condensate Polishing Deep Bed

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1.5

Pilgrim Milestones

Milestone events for Pilgrim are given in Table CC-2.

The original condenser tubes have been replaced with titanium tubes. The recirculation system
piping was replaced in 1984. The extraction steam piping was replaced with chrome-moly in
1985. Hydrogen injection began in September 1991. Portions of the RWCU piping were
replaced in 1993 and 1995. Pilgrim performed a partial chemical decontamination of the
recirculation system piping (discharge piping) in March 1997 after starting DZO injection in
December 1996.

CC-1
 EPRI Licensed Material

Pilgrim

Table CC-2
Pilgrim Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate

Condenser Retube

Recirc. Pipe
1/84
Replacement

RWCU Pipe
1/93 1/95
Replacement

Extraction
Steam Pipe 6/85
Replacement

Chem. Decon. 3/97

HWC (scfm) 9/91

→ → → → → → → → → →
(32)

NMCA

NZO

DZO 12/96 → → → → → →

Iron Injection

Crud Resins

Pleated Filters

Radiation Data

Recirculation System dose rates are summarized in Table CC-3.

CC-2
 EPRI Licensed Material

Pilgrim

Table CC-3
Pilgrim Milestones

Pilgrim - Recirculation System Dose Rates (mR/hr)

Jul-91 May-93 May-95 Mar-97 Dec-97 May-99 Aug-00 May-01

(1)

EFPY

BRAC 126 182 217 340 408 385 392 348

A Suction 140 200 250 500 1200 1000 900 700

B Suction 125 200 250 800 350 400 450 450

A Discharge 130 150 170 20 38 80 110 120

B Discharge 110 180 200 40 46 60 110 120

Avg Risers 272 345 359 40 80 190 275

Table CC-3 Notes:
1. Post Decon.

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Pilgrim are
presented in Figures CC-1, CC-2, CC-3 and CC-4, respectively.

Feedwater Iron Control

Average feedwater iron concentrations have decreased significantly since 1997. The average
feedwater iron was 3.88 ppb in 1998, 2.63 ppb in 1999, 2.24 ppb in 2000 and 2.27 ppb in 2001.
Pilgrim is a deep bed condensate polishing plant that does not employ low crosslinked resins for
iron control. Feedwater iron has decreased due to improvements in the URC process and in resin
management. Hotwell iron at Pilgrim is in the 6– 8 ppb range, which is among the lowest in the
industry.

CC-3
 EPRI Licensed Material

Pilgrim

100

80
Power (%)

60

40

20

0
8/9/96 12/22/97 5/6/99 9/17/00 1/30/02 6/14/03

Figure CC-1
Power History, Pilgrim

100

10
Feedwater Fe (ppb)

0.1

0.01

0.001

8/9/96 12/22/97 5/6/99 9/17/00 1/30/02 6/14/03

Insoluble Fe Soluble Fe Total Fe

Figure CC-2
Feedwater Iron, Pilgrim

CC-4
 EPRI Licensed Material

Pilgrim

1.E-02

Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06

8/9/96 12/22/97 5/6/99 9/17/00 1/30/02 6/14/03

Insoluble Co-60 Soluble Co-60

Figure CC-3
Reactor Water Cobalt-60, Pilgrim

500
HWC; Oxygen Injection

400
Extraction steam pipe replacement (date
Dose Rate (mR/hr)

Retube condenser (date unknown)

300
RWCU pipe replacement 1995
RWCU pipe replacement 1993
Recirc pipe replacement 1984

DZO
unknown)

40 CRBs replaced 1991

200
Chem decon

100

0
Jan-83 Sep-85 Jun-88 Mar-91 Dec-93 Sep-96 Jun-99

BRAC Co-60 Dose Milestones

Figure CC-4
BRAC History, Pilgrim

CC-5
 EPRI Licensed Material

Pilgrim

Recirculation Piping Dose Rates

Since starting DZO, dose rates have peaked at about 408 mR/hr and show a decreasing trend
from 12/97 through 5/01. Reactor coolant Co-60 activity is relatively low at Pilgrim. The
average soluble Co-60 was 5.02E-5 µCi/ml in 2000 and 3.84E-5 µCi/ml in 2001. The average
insoluble Co-60 activity was 6.11E-5 µCi/ml in 2000 and 3.25E-5 µCi/ml in 2001.

Recirculation Piping Gamma Scans

Gamma scan data for Pilgrim are summarized in Table CC-4.

The result from the 5/01 outage shows that about 53% of the total deposited activity was due to
Co-60.
Table CC-4
Pilgrim Recirculation Piping Gamma Scan Results

Date May-01

Total Activity
24
(µCi/cm2)

% Co-60 53.3

% Mn-54 2.5

% Zn-65 5.4

% Co-58 1.3

% Cr-51 37.9

Stellite™ Reduction

The station replaced 40 CRBs in 1991 with non-Stellite™ pins and rollers. The station has also
replaced both low pressure turbine rotors with non-Stellite™ materials.

CC-6
 EPRI Licensed Material

DD
QUAD CITIES 1

Table DD-1
Quad Cities Unit 1 Plant Design Parameters

Parameter Value

Commercial Operation Date 2/73

Capacity (MWe) 833

BWR Type 3

Drains Path Cascaded

Condensate Polishing Filter Demineralizer

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1.22/2

Quad Cities 1 Milestones

Milestone events for Quad Cities 1 are given in Table DD-2.

Quad Cities 1 has had a number of recirculation piping chemical decontaminations. HWC was
initiated in 10/90. Pleated filter septa were installed in some condensate filter demineralizers
beginning in 6/95. DZO injection started in 9/98 and NMCA was performed in 4/99.

Radiation Data

Recirculation System dose rates are summarized in Table DD-3.

DD-1
 EPRI Licensed Material

Quad Cities 1

Table DD-2
Quad Cities 1 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power
Uprate

Condenser
Retube

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction
Steam Pipe
Replacement

84,86,
Chem. 87,89,
3/94 4/96 11/98
Decon 90,
10/92

10/90 →
HWC (scfm) → → → → → → → → →
(47) (8-10)

NMCA 4/99

NZO

DZO 9/98 → → → →

Iron Injection

Crud Resins

Pleated
6/95 → → → → → → →
Filters

DD-2
 EPRI Licensed Material

Quad Cities 1

Table DD-3
Quad Cities 1 Recirculation Piping Dose Rates

Quad Cities 1 – Recirculation System Dose Rates (mR/hr)

Jan- Jan- Sep- Sep- Sep- Sep- Nov- Nov- Sep- Sep-
86 (1) 86 (2) 87 (1) 87 (2) 89 (1) 89 (2) 90 (1) 90 (2) 92 (1) 92 (2)

EFPY

BRAC 347 62.5 345 56 343 36 205 52 270 53

A Suction 294 51 293 46 397 40 180 51 320 59

B Suction 400 74 397 65 289 31 50 53 120 47

A
500 480
Discharge

B
90 160
Discharge

Avg Risers

Quad Cities 1 – Recirculation System Dose Rates (mR/hr)

Nov- Mar-94 Mar-94 Feb- Jan- Nov-98 Nov- 98 Apr- Oct-00 Oct-
93 (1) (2) 96 97 (1) (2) 99 (3) 00

EFPY

BRAC 252 268 78 210 473 553 176 143 781 339

A Suction 320 330 80 180 520 580 22 120 660 225

B Suction 240 120 77 280 470 440 220 180 750 394

A
430 420 260 680 950 400 180 934 501
Discharge

B
17 200 120 220 240 60 90 234
Discharge

Avg Risers 513 112 176 666

DD-3
 EPRI Licensed Material

Quad Cities 1

Table DD-3 (continued)

Quad Cities 1 Recirculation Piping Dose Rates

Quad Cities 1 – Recirculation System Dose Rates (mR/hr)

4/01

EFPY

BRAC 320

A Suction 420

B Suction 140

A Discharge 400

B Discharge

Avg Risers
Table DD-3 Notes
1. Pre decon
2. Post decon
3. Measured early in outage

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Quad Cities 1
are presented in Figures DD-1, DD-2, DD-3 and DD-4, respectively.

DD-4
 EPRI Licensed Material

Quad Cities 1

110
100
90
80
70
Power (%)

60
50
40
30
20
10
0
07/20/95 07/19/97 07/19/99 07/18/01 07/18/03

NMCA

Figure DD-1
Power History, Quad Cities 1

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001
07/20/95 07/19/97 07/19/99 07/18/01 07/18/03

Insoluble Fe Soluble Fe Total Fe NMCA

Figure DD-2
Feedwater Iron, Quad Cities 1

DD-5
 EPRI Licensed Material

Quad Cities 1

Reactor Water Co-60 (µCi/ml) 1.E-02

1.E-03

1.E-04

1.E-05

1.E-06
7/20/95 7/19/97 7/19/99 7/18/01 7/18/03

Insoluble Co-60 Soluble Co-60 Total Co-60 NMCA

Figure DD-3
Reactor Water Cobalt-60, Quad Cities 1

Measurement taken
4 days after
shutdown.
800

700
NMCA
600 47 scfm
HWC
Dose Rate (mR/hr)

500
Pleated Filters
400
DZO
300
Chem Decon 1987

Chem Decon 1989

Chem Decon 1986

Chem decon 1990

Chem decon 1992

Chem decon 1994

Chem decon 1998

Chem decon 1984

Chem decon 1996

200 Measurement taken

16 days after
100 shutdown.

0
Jan-84 Sep-86 Jun-89 Mar-92 Dec-94 Sep-97 Jun-00
BRAC Milestones BRAC - All

Figure DD-4
BRAC History, Quad Cities 1

DD-6
 EPRI Licensed Material

Quad Cities 1

Feedwater Iron Control

The average total feedwater iron was 1.00 ppb for 1998 and 1.64 ppb in 1999 after NMCA.
Feedwater iron has significantly decreased as the use of pleated filter septa has been extended to
additional vessels; the average for 2001 was 1.46 ppb.

Recirculation Piping Dose Rates

Quad Cities 1 performed nine chemical decontaminations from 1984 through 1998, lowering
outage BRAC average dose rates to approximately 60 mR/hr. Pre-decon dose rates were usually
between 250 and 350 mR/hr. DZO injection was started just before the last decon in November
1998. The BRAC result from the October 2000 refueling outage was 339 mR/hr, indicating a
significant increase from 143 mR/hr measured in the April 1999 outage when NMCA was
performed. A measurement in April 2001 shows a slight decrease to 320 mR/hr.

The average insoluble Co-60 for 1999 was 1.51E-3 µCi/ml, strongly influenced by the high
levels occurring after NMCA. Average soluble Co-60 was 2.57E-4 µCi/ml in 1999, also
influenced by increases after NMCA. In 2001, average insoluble Co-60 was 4.09E-4 µCi/ml,
and average soluble Co-60 was 2.33E-4 µCi/ml.

DD-7
 EPRI Licensed Material

EE
QUAD CITIES 2

Table EE-1
Quad Cities 2 Plant Design Parameters

Parameter Value

Commercial Operation Date 3/73

Capacity (MWe) 912

BWR Type 3

Drains Path Cascaded

Condensate Polishing Filter Demineralizer

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1.2/1.2

Quad Cities 2 Milestones

Milestone events for Quad Cities 2 are given in Table EE-2.

Quad Cities 2 has had a number of recirculation piping chemical decontaminations. HWC has
been applied since 1990. Pleated filter septa were first used in the condensate filter
demineralizers in 10/96. DZO injection was started in 6/97 and NMCA was performed in 1/00.
A 17.8 percent power uprate was implemented in 3/02.

Radiation Data
Recirculation System dose rates are summarized in Table EE-3.

EE-1
 EPRI Licensed Material

Quad Cities 2

Table EE-2
Quad Cities 2 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 3/02

Condenser
Retube

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction
Steam Pipe
Replacement

1983
1984
Chem. 1986
3/93 3/95 3/97
Decon. 1988
1990
1992

10/90 → → →
HWC (scfm) → → → → → → →
(47) (8-10) (8-10) (8-10)

NMCA 1/00

NZO

DZO 6/97 → → → → →

Iron Injection

Crud Resins

Pleated
10/96 → → → → → →
Filters

EE-2
 EPRI Licensed Material

Quad Cities 2

Table EE-3
Quad Cities 2 Recirculation Piping Dose Rates

Quad Cities 2 – Recirculation System Dose Rates (mR/hr)

Mar- Mar-
Apr- Apr- Feb- Apr- Jan- Jan-92 Mar- Mar-
93 95
88 88 90 (1) 90 (2) 92 (1) (2) 93 (2) 95 (1)
(1) (2)

EFPY

BRAC 235* 40** 295 17 250 58 223 35 245 33

A Suction 300 18 260 15 250 18 260 340 20

B Suction 220 28 280 15 250 20 180 180 60

A Discharge 280 94 380 20 320 175 310 55 320 40

B Discharge 140 18 260 18 180 20 140 15 140 10

Avg Risers

Quad Cities 2 – Recirculation System Dose Rates (mR/hr)

May- Mar- Mar- Oct- Oct- Feb- Jan- Mar- Aug- Feb-
96 (1) 97 (1) 97 (2) 97 98 99 00 01 01 02

EFPY

BRAC 150 203 31 79 99 149 338 295 116 205

A Suction 260 400 46 110 160 220 300 300 99 315

B Suction 100 90 32 55 44 56 400 600 28

A Discharge 160 200 25 120 140 220 250 90 243 95

B Discharge 80 120 20 32 52 100 300 190 94

Avg Risers 152 224 375

Table EE-3 Notes
1. Pre-decon
2. Post-decon

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Quad Cities 2
are presented in Figures EE-1, EE-2, EE-3 and EE-4, respectively.

EE-3
 EPRI Licensed Material

Quad Cities 2

100

80
Power (%)

60

40

20

0
6/1/94 10/14/95 2/25/97 7/10/98 11/22/99 4/5/01 8/18/02 12/31/03

% Power NMCA

Figure EE-1
Power History, Quad Cities 2

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001
6/1/94 10/14/95 2/25/97 7/10/98 11/22/99 4/5/01 8/18/02 12/31/03

Insoluble Fe Soluble Fe Total Fe NMCA

Figure EE-2
Feedwater Iron, Quad Cities 2

EE-4
 EPRI Licensed Material

Quad Cities 2

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
6/1/94 10/14/95 2/25/97 7/10/98 11/22/99 4/5/01 8/18/02 12/31/03

Insoluble Co-60 Soluble Co-60 NMCA

Figure EE-3
Reactor Water Cobalt-60, Quad Cities 2

600
47 scfm
HWC

500
Pleated Filters
Dose Rate (mR/hr)

400
DZO

300
NMCA

200
Chem Decon

Chem Decon

Chem Decon

Chem Decon

Chem Decon
Chem Decon

Chem Decon

Chem Decon

100

0
Jul-83 Mar-86 Dec-88 Sep-91 Jun-94 Mar-97 Dec-99 Aug-02

BRAC Milestones

Figure EE-4
BRAC History, Quad Cities 2

EE-5
 EPRI Licensed Material

Quad Cities 2

Feedwater Iron Control

Quad Cities 2 began the use of pleated septa in condensate demineralizer vessels in 10/96.
Feedwater iron showed a decreasing trend starting in 9/97. Average feedwater total iron in 1999
was 2.02 ppb. Total iron decreased to consistently less than 2 ppb in November 1999. The
average for 2001 was 1.38 ppb.

Recirculation Piping Dose Rates

Quad Cities 2 controlled outage BRAC dose rates at less than 100 mR/hr by performing nine
chemical decontaminations from 1983 through 1997. Pre-decon dose rates were usually between
200 and 300 mR/hr. No additional chemical decontaminations have been reported since 1997.
The plant started DZO injection in 6/97 after surveys showed a BRAC average dose rate of 234
mR/hr. A BRAC value of 338 mR/hr was reported immediately after performing NMCA, but
has decreased to 205 mR/hr in 2/02. Average reactor water insoluble Co-60 was 3.74E-4 µCi/ml
in 1999, and 1.13E-3 µCi/ml in 2001. Average soluble Co-60 was 1.51E-4 µCi/ml in 1999 and
3.30E-4 µCi/ml in 2001.

EE-6
 EPRI Licensed Material

FF
RIVER BEND

Table FF-1
River Bend Plant Design Parameters

Parameter Value

Commercial Operation Date 6/86

Capacity (MWe) 1086

BWR Type 6

Drains Path Forward Pumped

Condensate Polishing Deep Bed

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1

River Bend Milestones

Milestone events for River Bend are given in Table FF-2.

The recirculation and RWCU piping were chemically decontaminated in 1992. The activity
removed included 48.9 curies from the recirculation piping and 15.7 curies from the RWCU
piping. The use of crud removal resins began in the mid-1990s. Their use was stopped in 1997
due to reactor coolant sulfate concerns, but was resumed in 1999. DZO was initiated in 7/97.
The station experienced fuel failures in 1998 that were attributed to excess crud loading on new
fuel. DZO was secured in 6/99 as part of the fuel failure recovery plan, and resumed again in
7/00. A 5% power uprate was implemented in October 2000. HWC was initiated in 12/01 but
was discontinued until condensate filtration is available (scheduled by the end of 2002).

FF-1
 EPRI Licensed Material

River Bend

Table FF-2
River Bend Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 10/00

Condenser Retube

Recirc. Pipe
Replacement

RWCU Pipe
3/92
Replacement

Extraction Steam
Pipe Replacement

Chem. Decon. X

HWC (scfm)

NMCA

NZO

DZO 6/97 → 6/99 7/00 → →

Iron Injection

Crud Resins 2/94 → → 3/97 7/99 → → →

Pleated Filters

FF-2
 EPRI Licensed Material

River Bend

Radiation Data

Recirculation System dose rates are summarized in Table FF-3.

Table FF-3
River Bend Recirculation System Dose Rates

River Bend – Recirculation System Dose Rates (mR/hr)

1992 1992
1987 1989 1990 1993 1994 1996 9/97 4/99
(1) (2)

EFPY 1.00 2.03

BRAC 215 355 322.5 315 37.5 231 300 312 318 485

A Suction 280 350 160 310 15 200 nm 320 310 650

B Suction 200 320 460 400 50 300 300 400 380 370

A Discharge 180 350 350 280 45 225 400 260 300 440

B Discharge 200 400 320 270 40 200 200 270 280 480

Avg Risers 310 650 520 63 490 764 600 463

River Bend – Recirculation System Dose Rates (mR/hr) continued

3/00 9/01

EFPY

BRAC 291 473

A Suction 311 555

B Suction 355 500

A Discharge 260 411

B Discharge 236 324

Avg Risers 559 280

Table FF-3 Notes
1. Pre-decon
2. Post-decon.

FF-3
 EPRI Licensed Material

River Bend

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for River Bend are
presented in Figures FF-1, FF-2, FF-3 and FF-4, respectively. Only total Co-60 values were
provided until late 1999, after which soluble and insoluble fractions have been reported.

100

80
Power (%)

60

40

20

0
3/15/97 9/16/98 3/19/00 9/20/01 3/24/03

Figure FF-1
Power History, River Bend

Feedwater Iron Control

The concentration of total feedwater iron has averaged between 2.88 and 3.88 ppb since 1997,
while total feedwater copper averaged between 0.14 and 0.19 ppb and total CDE iron averaged
between 3.67 and 5.54 ppb over the same time period (see Table FF-4). During a non-refueling
outage in April 1998, it was found that two of the three ultrasonic transducer banks on the URC
system were not operating, which is likely the reason for the increased feedwater iron starting in
the latter part of 1997. Iron data from the forward pump heater drains are not available. Based
on the fact that the CDE data is higher than final feedwater, the iron content of the forward pump
drains is expected to be less than CDE (or a significant amount of the CDE iron is depositing in
the feedwater heaters and piping). It is noted that the difference between final feedwater and
CDE iron was about 3 ppb in 1998 and 1999 but less than 1 ppb in 2000 and 2001, indicating
improvements in resin cleaning.

Feedwater dissolved oxygen concentration was maintained below the 1996 BWR Water
Chemistry Guidelines lower limit of 20 ppb in 1998 with a reported annual average value of 17.7

FF-4
 EPRI Licensed Material

River Bend

ppb. The average was increased to 25.9 ppb in 1999 and 28.1 ppb in 2000. In 2001, the station
annual feedwater dissolved oxygen average was 41.7 ppb, which complies with the revised lower
limit of 30 ppb from the BWR Water Chemistry Guidelines, 2000 Revision.

1000

100
Feedwater Fe (ppb)

10

0.1

0.01

0.001
3/15/97 9/16/98 3/19/00 9/20/01 3/24/03

Insoluble Fe Soluble Fe

Figure FF-2
Feedwater Iron, River Bend

Table FF-4
River Bend Annual Average Feedwater and CDE Metals, 1997-2001

1997 1998 1999 2000 2001

Feedwater Iron
3.88 3.26 3.04 3.40 2.88
(ppb)

Feedwater
0.17 0.19 0.16 0.14 0.19
Copper (ppb)

CDE Iron (ppb) 4.85 7.54 6.08 4.26 3.67

FF-5
 EPRI Licensed Material

River Bend

Reactor Water Co-60 (µCi/ml) 1.E-02

1.E-03

1.E-04

1.E-05

1.E-06
3/15/97 9/16/98 3/19/00 9/20/01 3/24/03

Total Co-60 Insoluble Co-60 Soluble Co-60

Figure FF-3
Reactor Water Cobalt-60, River Bend

Recirculation Piping Dose Rates

The highest BRAC average dose rate reported at River Bend was 485 mR/hr in 4/99. The BRAC
average dose rate increased by about 160 mR/hr after almost two years of DZO injection. The
plant data show higher crud levels in the cycle leading up to the 4/99 outage.

Average soluble reactor water Co-60 activity was 1.07E-4 µCi/ml in 1999, 1.26E-4 µCi/ml in
2000, and 5.65E-5 µCi/ml in 2001. The average insoluble Co-60 was 1.33E-3 µCi/ml in 1999,
7.98E-5 µCi/ml in 2000, and 3.48E-4 µCi/ml in 2001. Early in 2000, the soluble and insoluble
Co-60 activity levels were approximately the same, with an average soluble to insoluble Co-60
ratio of 1.38. After September 2000, the soluble Co-60 showed a greater increase than the
insoluble fraction, with an average ratio of 3.48. The ratio is 2002 appears to be returning to the
early 2000 value.

FF-6
 EPRI Licensed Material

River Bend

500

400
Crud Resins Crud Resins
Dose Rate(mR/hr)

replacement; begin cobalt material

Chem decon; partial RWCU pipe
300
DZO DZO

replacement
200
HWC

Power Uprate
100

0
Jun-84 Feb-87 Nov-89 Aug-92 May-95 Feb-98 Nov-00 Aug-03

BRAC Milestones

Figure FF-4
BRAC History, River Bend

Stellite™ Reduction

The station reports the replacement of 29 of 145 CRBs with non-Stellite™ pins and rollers along
with the replacement of high and low pressure turbine rotors with non-Stellite™ materials.

Fuel Failures

Section 4 of this report (Transient Corrosion Products) discusses the crud-related fuel failures at
River Bend that occurred in 1998-1999. The failed fuel was reported to be high exposure single
cycle fuel with crud deposits containing a higher percentage of copper (2-15%) than what would
be expected for a deep bed condensate polishing plant.

Feedwater and reactor water copper trends are shown in Figures FF-5 and FF-6, respectively.
During the time of the fuel failures, feedwater copper values typically ranged from 0.1 to 0.2
ppb, while reactor water copper ranged from 1.5 to 3 ppb. These values are on the high end of
what is typically seen at deep bed condensate polishing plants (Reference Figure 3-13).

Following the fuel failures, efforts were made to reduce copper and iron. This included the use
of crud removal resins, improved feedwater oxygen control, and optimization of URC. Data for
much of 2001 and 2002 shows both feedwater and reactor water copper values on an increasing
trend. Reactor coolant copper in 2002 is two to three times higher than the levels leading up to
the 1998 fuel failures; there is currently no evidence of fuel failure.

FF-7
 EPRI Licensed Material

River Bend

0.50
0.45
0.40
0.35
Feedwater Cu (ppb)

0.30
0.25
0.20
0.15
0.10
0.05
0.00
3/15/97 9/16/98 3/19/00 9/20/01 3/24/03

Insoluble Cu Soluble Cu

Figure FF-5
Feedwater Copper, River Bend

10
9
8
Reactor Water Cu (ppb)

7
6
5
4
3
2
1
0
3/15/97 9/16/98 3/19/00 9/20/01 3/24/03

Insoluble Cu Soluble Cu

Figure FF-6
Reactor Water Copper, River Bend

FF-8
 EPRI Licensed Material

GG
SUSQUEHANNA 1

Table GG-1
Susquehanna 1 Plant Design Parameters

Parameter Value

Commercial Operation Date 6/83

Capacity (MWe) 1131

BWR Type 4 Mark II

Drains Path Cascaded

Condensate Polishing Filter + Deep Bed

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1.03/1.03

Susquehanna 1 Milestones

Milestone events for Susquehanna 1 are given in Table GG-2.

A 7% power uprate was implemented in 1995. The use of condensate pre-filters was started in
5/98 and feedwater iron injection began in 8/98. HWC was initiated in 1/99. A chemical
decontamination of the recirculation system piping was performed in March 2000 and a chemical
decontamination of the recirculation system piping and the RWCU drywell piping was
performed in 3/02. The 3/00 chemical decontamination removed 49 curies while the 3/02
chemical decontamination removed about 203 curies. While more activity was removed in the
3/02 decontamination, the final dose rates were higher than anticipated. The decontamination
process involved sloshing of decontamination chemicals (NP-LOMI) behind plugs in the jet
pump nozzles and suction nozzles. The station encountered problems during the LOMI step with
inleakage and inventory which reduced the effectiveness of the sloshing steps.

GG-1
 EPRI Licensed Material

Susquehanna 1

Table GG-2
Susquehanna 1 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 4/95

Condenser Retube

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction Steam
1/93 → → 12/96
Pipe Replacement

Chem. Decon. 3/00 3/02

HWC (scfm) 1/99 → → →

NMCA

NZO

DZO

Iron Injection 8/98 → → → →

Crud Resins 2/93 → 4/95

Pleated Filters 5/98 → → → →

GG-2
 EPRI Licensed Material

Susquehanna 1

Radiation Data

Recirculation System dose rates are summarized in Table GG-3.

Table GG-3
Susquehanna 1 Recirculation System Dose Rates

Susquehanna 1 – Recirculati on System Dose Rates (mR/hr)

Feb-85 Feb-86 Sep-87 Mar-89 Sep-90 Mar-92 Sep-93 Mar-95 Sep-96

EFPY 1.36 2.06 3.21 4.37 5.53 6.73 7.82 8.97 10.14

BRAC 153 145 130 175 143 135 140 112 140

A Suction

B Suction

A Discharge

B Discharge

Avg Risers 217 203 160 300 135 135 107 120 130

Susquehanna 1 – Recirculation Syst em Dose Rates (mR/hr) (continued)

Mar-00 Mar-00 Mar-02 Mar-02
Apr-98 May-99
(1) (2) (1) (2)

EFPY 11.51 12.42 13.24 14.92 14.92 14.92

BRAC 165 105 203 12 1475 311
A Suction 150 100 180 10 700 4
B Suction 220 100 210 22 1500 160
A Discharge 140 120 200 6 2100 740
B Discharge 150 100 220 10 1600 340
Avg Risers 130 130 340 43 1450 351
Table GG-3 Notes
1. Pre-decon.
2. Post-decon.

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Susquehanna 1
are presented in Figures GG-1, GG-2, GG-3 and GG-4, respectively.

GG-3
 EPRI Licensed Material

Susquehanna 1

100

80
Power (%)

60

40

20

0
1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Figure GG-1
Power History, Susquehanna 1

Feedwater Iron Control

Feedwater iron concentrations at Susquehanna 1 had been historically high compared to other
U.S. BWRs prior to the initiation of condensate filtration. Feedwater iron levels in 1997 and in
1998 (prior to filter operation) averaged about 9.5 ppb. Three months after filter operation
commenced, the station initiated feedwater iron injection because of low CDE iron concentration
(0.1 – 0.3 ppb). The average feedwater iron for 1999 was 1.11 ppb, 0.87 ppb in 2000, and 1.11
ppb in 2001.

GG-4
 EPRI Licensed Material

Susquehanna 1

100

10
Feedwater Fe (ppb)

0.1

0.01

0.001
1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Insoluble Fe Soluble Fe

Figure GG-2
Feedwater Iron, Susquehanna 1

1.E-02
Reactor Water Co-60 (uCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Insoluble Co-60 Soluble Co-60

Figure GG-3
Reactor Water Cobalt 60, Susquehanna 1

GG-5
 EPRI Licensed Material

Susquehanna 1

1500

Crud resins Fe injection

1200

Ext. steam piping replacement completed

Start replacing extraction steam piping
Dose Rate (mR/hr)

Cond. Prefilters
900

Start replacing pins/rollers

HWC
600

Power Uprate

Chem Decon
Chem decon
300

0
Jun-84 Jun-87 May-90 May-93 May-96 May-99 May-02

BRAC Milestones Co-60 Dose

Figure GG-4
BRAC History, Susquehanna 1

Recirculation Piping Dose Rates

Prior to 1998, recirculation piping dose rates for Susquehanna 1 had been relatively stable,
ranging from 112 mR/hr to 175 mR/hr. The dose rates had reached a minimum in 1995 at 112
mR/hr, while low cross-linked resins for enhanced crud removal were in use. During this time
the feedwater iron concentrations were also reduced. After a power uprate and the removal of
the low cross-linked resins from service, dose rates gradually trended higher until a forced
outage in 1999 when dose rates reached a new plant minimum of 102 mR/hr. This low dose rate
was recorded after the plant completed installation of condensate pre-filters, implemented
feedwater iron injection and began HWC. However, the dose rate increased to 203 mR/hr in
3/00 and the first chemical decontamination of the unit was performed. The dose rate
significantly increased during the next cycle as the reported BRAC dose rate in 3/02 was 1475
mR/hr, about a factor of 7.25 higher than the 3/00 reported value.

The average reactor water soluble Co-60 activity was 3.65E-5 µCi/ml in 1999, 1.51E-4 µCi/ml in
2000, and 3.93E-5 µCi/ml in 2001. Reactor water insoluble Co-60 activity averaged 3.07E-5
µCi/ml in 1999, 1.78E-4 µCi/ml in 2000, and 3.70E-4 µCi/ml in 2001.

Recirculation Piping Gamma Scans

Gamma scan data for Susquehanna 1 are summarized in Table GG-4.

GG-6
 EPRI Licensed Material

Susquehanna 1

The gamma scan data indicate that in 1995, when the BRAC dose rate reached a relative
minimum of 112 mR/hr, the piping total activity and Co-60 activity peaked at 16.2 µCi/cm2 and
7.1 µCi/cm2, respectively. The reason for this apparent discrepancy has not been determined.
The data also indicated a change in the isotopic mix. In May 1999 (first scan after HWC
initiation), Cr-51 represented a major fraction of the total activity. The 3/00 gamma scan results
show no Cr-51 and indicate that an effective chemical decontamination was performed. Gamma
scan data associated with the highest BRAC dose rate of 1475 mR/hr in 3/02 are not available.
Table GG-4
Susquehanna 1 Recirculation Piping Gamma Scan Results

Date Feb-85 Feb-86 Sep-87 Mar-89 Sep-90 Mar-92 Sep-93 Mar-95 Sep-96

Total Activity
13.8 12.1 12.3 10.4 10.8 12.5 11 16.2 10
(µCi/cm2)

% Co-60 39 45 48 49 46.3 39.2 50 39.5 32

% Co-58 23 15 8.1 2.9 3.7 2.4 nd 4.3 3

% Mn-54 29 28 27 30 31.5 35.2 36.3 39.5 51

% Zn-65 3.6 6.6 11.3 13.5 13.9 14.4 11 4.9 6

% Fe-59 5.0 5.8 5.7 4.8 4.6 8.8 2.7 7.4 8

Mar-00 Mar-00 Mar-02

Date Apr-98 May-99
(1) (2) (3)

Total Activity
7.41 7.20 14.7 0.22 2.2
(µCi/cm2)

% Co-60 40.5 46 36.2 45 66

% Co-58 4.2 3.3 21.2 0 16.1

% Mn-54 47.5 5.1 7.3 45 0

% Zn-65 3.6 1.9 14.7 0 8.9

% Fe-59 7.8 2.9 0 10 0.7

% Cr-51 0 41 0 0 0
Table GG-4 Notes
1. Pre-decon
2. Post-decon
3. Post-decon (risers only)

GG-7
 EPRI Licensed Material

Susquehanna 1

Stellite™ Reduction

Susquehanna 1 has reduced the Stellite™ surface area by 15 % since 1990, from an initial
surface area of 52 ft2 to 44 ft2. The station initiated replacement of CRBs with non-Stellite™
pins and rollers in 1990.

GG-8
 EPRI Licensed Material

HH
SUSQUEHANNA 2

Table HH-1
Susquehanna 2 Plant Design Parameters

Parameter Value

Commercial Operation Date 2/85

Capacity (MWe) 1131

BWR Type 4 Mark II

Drains Path Cascaded

Condensate Polishing Filter + Deep Bed

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1.03/1.03

Susquehanna 2 Milestones

Milestone events for Susquehanna 2 are given in Table HH-2.

A 7% power uprate was implemented in 6/94. Replacement of the extraction steam piping from
carbon steel to chrome/moly was started in 1993 and completed in 1997. The project
encompassed upgrades to the turbine shell to feedwater heater shell piping, and to the No. 2, No.
3 and No. 4 extraction steam piping. Low cross-linked resins for enhanced crud removal were in
use from 2/93 – 6/96. The use of condensate pre-filters began in 6/99 and feedwater iron
injection began a month later. HWC was initiated in 8/99. A chemical decontamination of the
recirculation system piping in 3/01 removed 116 curies.

HH-1
 EPRI Licensed Material

Susquehanna 2

Table HH-2
Susquehanna 2 Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 6/94

Condenser Retube

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction Steam
Start → → → End
Pipe Replacement

Chem. Decon.

HWC (scfm) 8/99 → → →

NMCA

NZO

DZO

Iron Injection 7/99 → → →

Crud Resins 2/93 → → 6/96

Pleated Filters 6/99 → → →

Radiation Data

Recirculation System dose rates are summarized in Table HH-3.

HH-2
 EPRI Licensed Material

Susquehanna 2

Table HH-3
Susquehanna 2 Recirculation System Dose Rates

Susquehanna 2 – Recirculati on System Dose Rates (mR/hr)

Jul-86 Feb-88 Sep-89 Feb-91 Sep-92 Mar-94 Sep-95 Mar-97 Mar-99

EFPY 1.41 2.66 3.79 4.96 6.21 7.41 8.63 9.79 11.5

BRAC 124 111 180 195 120 140 140 133 103

A Suction 100

B Suction 100

A Discharge

B Discharge 110

Avg Risers 188 300 320 375 155 190 190 163 135

Susquehanna 2 – Recirculation System Dose Rates (mR/hr)

Mar-01 Mar-01
(1) (2)

EFPY 13.14 13.14

BRAC 580 7.5

A Suction 540 4

B Suction 580 8

A Discharge 600 10

B Discharge 600 8

Avg Risers 1250 123

Table HH-3 Notes

1. Pre-decon
2. Post decon

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Susquehanna 2
are presented in Figures HH-1, HH-2, HH-3 and HH-4, respectively.

HH-3
 EPRI Licensed Material

Susquehanna 2

Feedwater Iron Control

Prior to initiation of condensate filtration in 5/99, feedwater iron concentrations at Susquehanna

2 were historically high among U.S. BWRs, similar to those at Susquehanna 1.

The average feedwater iron prior to pre-filter operation was 8.12 ppb (1998-1999). Feedwater
iron averaged 0.83 ppb in 1999 following pre-filter operation, 0.92 ppb in 2000, and 0.89 ppb in
2001. Feedwater iron injection is necessary since CDE iron is typically low (0.03-0.2 ppb).

100

80
Power (%)

60

40

20

0
1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Figure HH-1
Power History, Susquehanna 2

HH-4
 EPRI Licensed Material

Susquehanna 2

100

10
Feedwater Fe (ppb)

0.1

0.01

0.001
1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Insoluble Fe Soluble Fe

Figure HH-2
Feedwater Iron, Susquehanna 2

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02 12/31/03

Insoluble Co-60 Soluble Co-60

Figure HH-3
Reactor Water Cobalt-60, Susquehanna 2

HH-5
 EPRI Licensed Material

Susquehanna 2

600 HWC
Crud Resins

500

Start replacing extraction steam piping

40% extraction steam piping replaced

Pleated Filters
Dose Rate (mR/hr)

Start replacing pins/rollers 1991

400

300

1993

1994
Power Uprate 1994
Iron Inj
200

Chem Decon
100

0
Jun-84 Feb-87 Nov-89 Aug-92 May-95 Feb-98 Nov-00

BRAC Milestones Co-60 Dose

Figure HH-4
BRAC History, Susquehanna 2

Recirculation Piping Dose Rates

Prior to HWC, reactor BRAC dose rates at Susquehanna 2 were relatively stable, ranging
fromm100-200 mR/hr. The last reported BRAC average prior to HWC was 103 mR/hr in March
1999.

HWC, condensate filtration and iron injection were started in the summer of 1999. When the
plant was shut down for the next refueling outage in March 2001, the BRAC average dose rate
had increased to 580 mR/hr. A successful chemical decontamination of the recirculation piping
reduced the dose rate to 7.5 mR/hr.

The average reactor water soluble Co-60 activity was 3.46E-5 µCi/ml in 1999, 3.64E-5 µCi/ml
in 2000, and 2.46E-5 µCi/ml in 2001. The average reactor water insoluble Co-60 was 3.15E-5
µCi/ml n 1999, 5.79E-5 µCi/ml in 2000, and 9.92E-5 µCi/ml in 2001

Recirculation Piping Gamma Scans

Gamma scan results are summarized in Table HH-4.

The piping gamma scan results show a small variation in the total activity through 3/99. The
next measurement, in 3/01, shows an increase in the total surface activity by a factor of about 7

HH-6
 EPRI Licensed Material

Susquehanna 2

following operation under HWC with no DZO injection. The pre-decontamination measurement
in 3/01 shows that 60% of the total activity is from Co-60 compared to 31% in 3/99.
Table HH-4
Susquehanna 2 Recirculation Piping Gamma Scan Results

Jul-86 Sep-89 Feb-91 Sep-92 Mar-94 Sep-95 Mar-97 Mar-99

Total Activity (µCi/cm2) 12.1 13.9 13.6 7.9 8.6 12.5 11.4 8.4

% Co-60 34 38 48 49 48 47 38 31

% Co-58 23 9 7 nd nd 3 4 5

% Mn-54 33 44 34 38 47 38 47 55

% Zn-65 5.7 5 6.9 2.5 3.5 4.8 5.3 3.6

% Fe-59 4.1 4.6 4.4 10.1 2.3 6.4 6.1 5.9

Mar- Mar-
01 (1) 01 (2)

Total Activity
57.7 0.21
(µCi/cm2)

% Co-60 60 100

% Co-58 22 0

% Mn-54 6.9 0

% Zn-65 11 0

% Fe-59 0 0
Table HH-4 Notes
1. Pre-decon (B Suction)
2. Post-decon (A & B Suction, B Discharge)

Stellite™ Reduction

The Susquehanna 2 Stellite™ surface area has been reduced by 14 % since 1990, from an initial
surface area of 52 ft2 to 44.8 ft2. The station initiated replacement of CRBs with non-Stellite™
pins and rollers in 1991.

HH-7
 EPRI Licensed Material

II
VERMONT YANKEE

Table II-1
Vermont Yankee Plant Design Parameters

Parameter Value

Commercial Operation Date 11/72

Capacity (MWe) 550

BWR Type 4

Drains Path Cascaded

Condensate Polishing Filter Demineralizer

RWCU Capacity (% Feedwater Flow), Normal/Max 1/1

Vermont Yankee Milestones

Milestone events for Vermont Yankee are given in Table II-2.

The recirculation piping was replaced in 1986; the original material was 304 stainless steel and
the replacement material was 316 Hitachi stainless steel. NMCA was performed in 04/01 after
which hydrogen injection was tested. HWC has not yet been implemented due to high dose rates
encountered during testing.

II-1
 EPRI Licensed Material

Vermont Yankee

Table II-2
Vermont Yankee Milestones

Vermont Yankee

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate

Condenser
Retube

Recirc. Pipe
1986
Replacement

RWCU Pipe
1986
Replacement

Extraction
Steam Pipe
Replacement

Chem. Decon. 1985

HWC (scfm)

NMCA 04/01

NZO (1)

DZO

Iron Injection

Crud Resins

Pleated Filters
Table II-2 Notes
1. The source of feedwater zinc is the main condenser tube material.

II-2
 EPRI Licensed Material

Vermont Yankee

Radiation Data

Recirculation System dose rates are summarized in Table II-3. Early measurements prior to
recirculation pipe replacement are not shown.
Table II-3
Vermont Yankee Recirculation System Dose Rates

Vermont Yankee – Recirculation System Dose Rates (mR/hr)

8/87 7/88 2/89 3/90 9/90 3/91 5/91 9/91 3/92

EFPY 10.5 11.2 11.7 12.6 13 13.5 13.7 13.9 14.4

BRAC

A Suction 80 75 75 85 85 85 100 100 100

B Suction

A Discharge

B Discharge

Avg Risers

Vermont Yankee – Recirculation System Dose Rates (mR/hr)

4/93 8/93 3/95 10/96 4/98 10/99 05/01 5/02

EFPY 15.3 15.6 16.8 18.2

BRAC 76 92

A Suction 95 100 80 65 66 93 80

B Suction 62 269 75

A Discharge 193 188 80

B Discharge 188 188 70

Avg Risers

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Vermont
Yankee are presented in Figures II-1, II-2, II-3 and II-4, respectively.

II-3
 EPRI Licensed Material

Vermont Yankee

100

80
Power (%)

60

40

20

0
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02

% Power NMCA

Figure II-1
Power History, Vermont Yankee

Feedwater Iron Control

Vermont Yankee feedwater iron annual average was 1.4 in 1999, 1.1 in 2000, and 1.2 in 2001.
The plant uses non-pleated precoated septa in its condensate filter demineralizers. Seasonal
variation is apparent in Figure II-2. The data also shows that the measured Condensate
Demineralizer Effluent iron concentrations average approximately 0.7 ppb higher than feedwater
iron concentrations.

II-4
 EPRI Licensed Material

Vermont Yankee

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02

Insoluble Fe Soluble Fe NMCA

Figure II-2
Feedwater Iron, Vermont Yankee

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02

Insoluble Co-60 Soluble Co-60 NM CA

Figure II-3
Reactor Water Cobalt-60, Vermont Yankee

II-5
 EPRI Licensed Material

Vermont Yankee

1200 Natural zinc from

1000

Begin CRB pins/rollers replacement

Chem decon; recirc & RWCU pipe

Dose Rate (mR/hr)

800
NMCA

600

replacement
Replace FW reg valves
400
Chem decon
200

0
Jan-74Dec-76Dec-79Dec-82Dec-85Dec-88Dec-91Dec-94Dec-97Dec-00

BRAC A Suction Only Milestones Co-60 Dose

Figure II-4
BRAC History, Vermont Yankee

Recirculation Piping Dose Rates

Vermont Yankee has had consistently low BRAC dose rates since replacing the recirculation
piping in 1985, as shown in Figure II-4. The BRAC data reported prior to 2001 were A Suction
values only. Vermont Yankee continued to have recirculation system piping dose rates of less
than 100 mR/hr through the 5/01 outage when NMCA was applied. One year after NMCA, the
BRAC dose rate had increased from 76 to 92 mR/hr. Vermont Yankee has a natural zinc source
in the brass condenser tubes, which contributes between 0.15 to 0.30 ppb zinc to the feedwater.

Reactor water soluble Co-60 activity remains low at Vermont Yankee, averaging 6.0 E-5 µCi/ml
in 2000 and 1.9E-4µCi/ml in 2001. The data show that the soluble Co-60 tends to increase at the
end of the fuel cycle.

It is significant to note that Vermont Yankee was not always a low dose rate plant, although the
zinc source from the condenser has been present since the initial plant startup. Prior to replacing
their recirculation piping, the contact dose rates approached 1200 mR/hr, suggesting that plant
conditions and/or initiatives may account for the subsequent low dose rates. The replacement
recirculation pipe is electropolished Hitachi 316L stainless steel. Apparently, the combination of
piping replacement and choice of material (electropolished stainless steel), the cobalt materials
replacement program, zinc from the condenser and the fact that the plant has been successful in
avoiding preventable transients (scrams) all contribute to maintaining low dose rates.

II-6
 EPRI Licensed Material

Vermont Yankee

Vermont Yankee has also replaced the low pressure turbine, which was reported to have resulted
in a 25 % decrease in the hotwell iron, which averages around 10-12 ppb. At the same time, zinc
and copper in the hotwell also decreased and changed from predominantly insoluble (filterable)
species to soluble species. In depth chemistry data from previous cycles were not available for
review.

Recirculation Piping Gamma Scans

One set gamma scan data from 1993 shows a total activity of 3.5 µCi/cm2 with about 60%
attributable to Co-60. The fraction of Zn-65 in the recirculation pipe deposit was low at less than
5% of the total activity reported.

Stellite™ Reduction

Vermont Yankee has replaced the Stellite™ control rod pins and rollers and the feedwater
regulating valves. The feedwater regulating valves were changed prior to replacing the
recirculation piping, thus eliminating a major source of cobalt input to the reactor.

Fuel Failures

Vermont Yankee first identified fuel leakage in December 2001; additional leaking fuel was
identified in March 2002. The plant was shut down for a mid-cycle outage in May 2002 and
forty second-cycle bundles were discharged, of which 4 had leakers, with two failures occurring
in one bundle. The removed bundles were replaced with 16 new fuel assemblies and 24 reinserts
from the spent fuel pool which were discharged the previous cycle. Investigation into the cause
of the fuel failures is not yet complete. Reactor water iodine and offgas sum of 6 noble gases
trend data are presented in Figures II-5 and II-6, respectively.

II-7
 EPRI Licensed Material

Vermont Yankee

Reactor Water Iodines (µCi/ml) 1.E-02

1.E-03

1.E-04

1.E-05

1.E-06
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02

I131 I132 I133 I134 I135 NMCA

Figure II-5
Reactor Water Iodines, Vermont Yankee

10000
9000
8000
Offgas Sum of 6 (µCi/sec)

7000
6000
5000
4000
3000
2000
1000
0
1/1/97 1/1/98 1/1/99 1/1/00 12/31/00 12/31/01 12/31/02

Sum of 6 NMCA

Figure II-6
Offgas Sum of 6 Noble Gases, Vermont Yankee

II-8
 EPRI Licensed Material

JJ
COLUMBIA (FORMERLY WNP2)

Table JJ-1
Columbia Plant Design Parameters

Parameter Value

Commercial Operation Date 12/84

Capacity (MWe) 1180

BWR Type 5/6

Drains Path Cascaded

Condensate Polishing Filter Demineralizer

RWCU Capacity (% Feedwater Flow), Normal/Maximum 1/1

Columbia Milestones

Milestone events for Columbia are given in Table JJ-2.

A recirculation pipe chemical decontamination performed in 1992 removed 44.5 curies. Iron
injection was initiated 6/96 using iron oxalate. DZO injection was started 9/96. NMCA was
performed 5/01 and hydrogen injection is planned for fall 2003.

JJ-1
 EPRI Licensed Material

Columbia (formerly WNP2)

Table JJ-2
Columbia Milestones

Pre-
Milestone 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002
1993

Power Uprate 6/94

Condenser
Retube

Recirc. Pipe
Replacement

RWCU Pipe
Replacement

Extraction
Steam Pipe
Replacement

Chem.
4/92 5/95
Decon.

HWC (scfm)

NMCA 5/01

NZO

DZO 9/96 → → → → → →

Iron Injection 6/96 → → → → → →

Crud Resins

Pleated
Filters

JJ-2
 EPRI Licensed Material

Columbia (formerly WNP2)

Radiation Data

Recirculation System dose rates are summarized in Table JJ-3.

Table JJ-3
Columbia Recirculation System Dose Rates

Columbia – Recirculation System Dose Rates (mR/hr)

Jun-87 May-88 May-89 May-90 Jun-91 Jun-93 Jul-94 Apr-96

EFPY 1.3 2.1 2.6 3.3 4.0 5.4 5.9

BRAC 49 117 100 410 316 165 320 400

A Suction 55 107 58 598 176 280 260 400

B Suction 68 111 300 244 438 200 280

A Discharge 66 108 18 443 295 30 300 460

B Discharge 8 140 22 356 354 150 440 340

Avg Risers

Columbia – Recirculation System Dose Rates (mR/hr)

Apr-97 May-98 Sep-99 May-01

EFPY 8.3 11

BRAC 377 330 313 318

A Suction 370 340 380 360

B Suction 300 280

A Discharge 460 400 320 350

B Discharge 300 280 240 280

Avg Risers 600

Trend Data

Power, feedwater iron, reactor water cobalt-60 and BRAC history trend plots for Columbia are
presented in Figures JJ-1, JJ-2, JJ-3, and JJ-4, respectively.

JJ-3
 EPRI Licensed Material

Columbia (formerly WNP2)

100

80
Power (%)

60

40

20

0
1/1/96 5/15/97 9/27/98 2/9/00 6/23/01 11/5/02

Power NMCA

Figure JJ-1
Power History, Columbia

Feedwater Iron Control

Columbia initiated feedwater iron injection in 6/96 to maintain iron concentrations at or above
0.5 ppb to control the transport of activated corrosion products. Iron is added as iron oxalate,
which results in an increase in feedwater conductivity. The average feedwater conductivity for
the cycle starting 6/96 was 0.067 µS/cm. Average insoluble feedwater iron for was 0.76 ppb in
2000 and 0.68 ppb in 2001. Columbia does not use pleated filter septa in its condensate F/Ds.

JJ-4
 EPRI Licensed Material

Columbia (formerly WNP2)

10

1
Feedwater Fe (ppb)

0.1

0.01

0.001
1/1/96 5/15/97 9/27/98 2/9/00 6/23/01 11/5/02
Insoluble Fe Soluble Fe NMCA

Figure JJ-2
Feedwater Iron, Columbia

1.E-02
Reactor Water Co-60 (µCi/ml)

1.E-03

1.E-04

1.E-05

1.E-06
1/1/96 5/15/97 9/27/98 2/9/00 6/23/01 11/5/02

Insoluble Co-60 Soluble Co-60 NMCA

Figure JJ-3
Reactor Water Cobalt-60, Columbia

JJ-5
 EPRI Licensed Material

Columbia (formerly WNP2)

500

400
Dose Rate (mR/hr)

300
Iron Injection

200
DZO

Power uprate
Chem decon

100
NMCA

0
Jun-87 Jan-89 Sep-90 May-92 Dec-93 Aug-95 Apr-97 Nov-98 Jul-00 Mar-02

BRAC Milestones Co-60 Dose

Figure JJ-4
BRAC History, Columbia

Recirculation Piping Dose Rates

Recirculation piping dose rates were reduced in 1992 by chemical decontamination of the
discharge piping. Pipe dose rates continue to be in the high range for BWRs at 300 to 400
mR/hr. DZO injection was started in 9/96 after the start of iron injection and dose rates slowly
decreased to 313 mR/hr in 9/99, and remained relatively constant at 318 mR/hr in 5/01, prior to
NMCA. In 1999, average reactor water insoluble Co-60 and soluble Co-60 were 1.16E-4 µCi/ml
and 1.06E-4 µCi/ml, respectively. Average insoluble Co-60 decreased to 9.62E-5 µCi/ml in
2000, and increased to 4.04E-4 µCi/ml following NMCA in 5/01. Soluble Co-60 also showed an
increase after NMCA; the 2000 average was 8.48E-5 µCi/ml and the 2001 average was 1.52E-4
µCi/ml.

Recirculation Piping Gamma Scans

Gamma scan data for Columbia are summarized in Table JJ-4.

The gamma scan results from 1994 through 1998 indicate a decreasing trend in total surface
activity. The Co-60 contribution remained essentially the same throughout this period. The 5/01
result shows an increase in the total surface activity but a decrease in the Co-60 contribution.

JJ-6
 EPRI Licensed Material

Columbia (formerly WNP2)

Table JJ-4
Columbia Recirculation Piping Gamma Scan Results

Apr-94 Apr-95 Apr-96 Apr-97 Apr-98 May-01

Total Activity
38.2 19.9 19.4 16.2 14.4 18.1
(µCi/cm2)

% Co-60 84 88 88 90 79 61

% Co-58 2 4 4 3 4 9

% Mn-54 2 2 2 4 10 20

% Zn-65 12 6 6 2 3 5

JJ-7
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