《Nuclear Power Safety》

Severe Accident Phenomena

Weimin Ma
Division of Nuclear Power Safety
Royal Institute of Technology (KTH)
Stockholm, Sweden

Contents
 A severe accident scenario
 In-vessel phenomena
o Core degradation (early phase)
o Core reflooding
o Corium relocation (late phase)
o Vessel failure
 Ex-vessel phenomena
o Fuel coolant interactions (steam explosion)
o Debris bed coolability
o Corium spreading
o Molten core concrete interactions (MCCI)
o Hydrogen behavior
 Fission product behavior (Source term)

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 Illustration of a severe accident scenario

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A severe accident scenario

 Loss of coolant for a long period of time, leading to core uncovery.
 Core heat-up due to decay of Fission Products (FP)
 Core material oxidation (by steam)
 Liquefaction and melting of core materials
 Release and transport of Fission Products (FP)
 Loss of core geometry
 Relocation of core materials into the lower plenum
 In-vessel fuel coolant interactions (FCI)
 Evolution of debris bed/molten pool in the lower head
 Creep failure of the lower head
 Corium slump
 Direct containment heating (DCH)
 Ex-vessel fuel coolant interactions (FCI) if the reactor cavity is wet
 Melt spreading if the reactor cavity is dry
 Molten core concrete interactions (MCCI) if the reactor cavity is dry
China syndrome
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 Severe accident phenomena in brief

Core heatup and

degradation / Reflood
H2
Atmosphere mixing In-Vessel combustion
Coolability In-Vessel
FCI

FP production

DCH Melt Pool

Convection
FP behavior
(Source term)
Vessel Ex-Vessel
Creep and Failure FCI (wet cavity)

Hole Melt
Ablation Spreading (dry)

Ex-Vessel debris coolability

(wet) /
MCCI (dry)

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Another severe accident scenario

 Loss of coolant for a long period of time, leading to core uncovery.

 Core heat-up due to decay of Fission Products (FP)
 Core material oxidation (by steam)
 Liquefaction and melting of core materials
 Release and transport of Fission Products (FP)
 Loss of core geometry
 Relocation of core materials into the lower plenum
 Evolution of debris bed/molten pool in the lower head
 Core reflooding (as in TMI-2)

Accident was stabilized and terminated in the RPV

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 Core degradation

T (K)

Late Phase
Core collapse
and relocation
Molten pool formation

Partial fuel melting

core blockages
(Ceramic melts)
Fission
product
release

Early Phase
Local core damages
(metallic melts)
@1500 K
Start of H2
significant
production
Ag-In-Cd melting

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Oxidation
 Potential H2 generation

CR CR
structure absorber

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 Reflooding of a damaged core
 Reflooding effects
 Enhanced oxidation
 Temperature escalation and large H2 peak
production
 Additional core damage
 Debris bed formation due to thermal shock and
collapse of upper fuel rods
 Large steam production and heat transfer to upper
structures
 Large FP release due to fuel shattering
 Oxidation escalation is due to the oxidation of
molten or frozen U-O-Zr mixtures
 Data recently produced on oxidation of solid U-O-
Zr mixtures but no data for liquid mixtures

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In-core molten corium pool

 Rapid transition from solid-liquid

debris to a molten pool due to bad
cooling resulting from steam
diversion by crust
 Molten pool growth because the
surface heat transfer could not
compensate the internal FP decay
heat
 TMI-2: molten pool supported by
a lower bowl-shaped crust
(thickness ~ 10 cm)
Steam flow diversion

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 Corium relocation into the lower plenum

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Corium fragmentation in the lower plenum

 Molten corium interacts with

residual water in the lower
plenum (fuel coolant
interactions - FCI)
 Fragments (debris) may be
formed if small jet
 Steam explosion risk
 Pressure peak
 Oxidation of the remaining Zr
 Molten corium underwater
spreading may be observed if
large jet

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 Oxidation of Zr droplet

 MISTEE-HT experiment at KTH-NPS

 A cyclic process of bubble growth and
detachment. The bubbles do not seem to
condense even at high subcooling
 Hydrogen/vapor bubble size increases and
bubble detaches from the droplet rear upon
reaching a critical size
 After bubble detachment the oxidation rate
reaches a peak value and then decreases following
bubble growth.
 No spontaneous steam explosions probably
because of significant void.

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Corium evolution in the lower plenum

 Option 1: Internal cooling (TMI-2)

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 Corium evolution in the lower plenum (cont’d)

 Option 2: External cooling (IVR: In-vessel melt retention)

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In-vessel melt retention (IVR)

 Melt pool heat transfer vs. External cooling
Heat flux

Melt
Angular position

Coolant inflow

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 Melt pool convection and heat transfer

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Cooling capacity of external cooling

 CHF of external cooling

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 Vessel failure (no cooling)
 Reactor pressure vessel (RPV) is designed with sufficient strength
(ample margin to yield and ultimate stress) during normal operation.
 However, the strength degrades greatly the vessel steel is at high
temperature (>800°C).
 At such high temperatures even a modest pressure (20-25 bars) loading
can induce creep in the vessel steels.
 Continuation of the thermal and pressure loads on the RPV lower head can
lead to creep displacements sufficient to cause rupture.
 Parameters/factors of interest
 Failure time
 Failure location
 Melt discharge
 Effect of penetrations.

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Vessel failure (contd.)

 Option 3: No cooling  Vessel failure

 Vessel creep failure under the attack of high-temperature melt

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 Corium slump

 Option 1: high-pressure ejection of melt (dry cavity)

 Direct Containment Heating (DCH)

A set of induced phenomena:

• Corium dispersal into the cavity,
debris entrainment to containment;
• Heat exchange between debris and
gases;
• Oxidation of metallic part of
corium, H2 production and
possible
• Subsequent combustion of
prexisting/produced H2

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Corium slump
 Option 2: low-pressure ejection of melt (ADS,
wet cavity)
 Fuel coolant interactions (FCI)
 Melt jet fragmentation (deep water pool) 
forming debris bed
 Oxidation of metal components (Zr, Fe)
 Steam explosion
 Melt underwater spreading (shallow water pool) 
melt layer

Shallow pool
Deep pool

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 Possible locations of forming debris beds

(a) in the core (b) in the lower plenum (c) in the reactor cavity

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DEFOR experiment at KTH

 Fuel coolant interactions in a deep water pool

DEFOR test (KTH)

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 DEFOR experiment at KTH (cont’d)
 Debris characteristics

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Debris bed coolability

 Cooling schemes: Two-phase flow in porous media

(a) Top flooding (b) Bottom-fed

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 Debris bed coolability

 Quench of an ex-vessel debris bed

 Some results

Zheng Huang, Weimin Ma, Numerical investigation on quench of an ex-vessel debris bed at prototypical
scale, Annals of Nuclear Energy 122: 47–61, 2018.

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Melt underwater spreading

 PULiMS experiment at KTH-NPS

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 Corium spreading

 EPR: Spreading chamber designed to minimize corium height and

thus thermal loading.

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Steam explosion
 (Molten) Fuel- or Corium- Coolant Interaction (FCI)
 A process by which a certain quantity of molten core materials is put in close
contact with water coolant, thus transferring energy to it more or less rapidly
with subsequent vaporization of the water
 Steam or vapor or thermal explosion
 FCI process by which energy of all or part of the corium is transferred to the
water in a time-scale smaller than the time-scale for system pressure relief and
induces dynamic loading of surrounding structures
 Thermal detonation
 Self-sustained steam explosion process first introduced by S.J. Board et al. in
1974, which assimilates the structure of “extremely violent thermal
explosions” to that of chemical detonations
 Efficiency
 Conversion Ratio (CR) of mechanical energy output to thermal energy of the
corium (in general, the corium present in the water at the time of the SE).

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 Hydrogen explosion/Steam explosion

Explosion Driving force Acceleration Detonation

Hydrogen Combustion/ Flame Chemical
explosion Chemical energy detonation
Steam Boiling/ Vapor Thermal
explosion Thermal energy detonation

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Steam explosion issues

 Outcomes
 Strong shock waves
 Hydrodynamic loading to the surrounding system
 In-vessel steam explosion
 Alpha-mode containment failure
 Considered resolved from risk perspective
 Reactor pressure vessel failure
 E.g., failure of in-vessel retention strategies.
 Ex-vessel steam explosion
 Damage to cavity
 Challenge containment integrity
 Severe loads to primary system piping

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 Steam (vapor) explosion

Alpha-mode containment failure

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Conceptual picture of SE development

 Phases of steam explosion

Mixing Triggering Propagation Expansion

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 Examples of mixing
 KROTOS experiment

Fig. 1.Test KT-2 with

Corium melt corium
mixing melt
in KT-2 Fig. 2. Test K-57melt
Alumina withmixing
alumina in melt
K-57
(CANON). Viewing area 10 by 20 cm (CANON). Viewing area 10 by 20 cm

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Steam explosion

 MISTEE experiment at KTH-NPS

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 Stratified steam explosion

 SES experiment at KTH-NPS

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Molten Core Concrete Interactions (MCCI)

 MCCI occurs in a dry cavity or wet cavity with uncoolable corium

 Characterized by the interaction between the corium melt and the
concrete basemat
 Most existing NPPs: MCCI is unconfined and thus may proceed
until the basemat has been completely penetrated (China
syndrome).
 Some new NPPs (e.g. EPR, VVER-1200/AES-91) have dedicated
design features, i.e. core catchers, to avoid melt attack on the
structural concrete.

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 Composition of concrete
 Cement + Aggregates+ H2O ⇒ Hardened Concrete + Heat
 Cement is a mixture of processed limestone, shales, and clays which contain
the following compounds: CaO (lime), Al2O3 (Alumina), SiO2 (silica) and
iron oxides.

sand gravels stones

0-2mm 2-31.5 mm 20-80 mm
+ adjuvants ( plastifiers, setting accelerators,…)
+ rebar steel, presstressing cables….
 Hardened cement contains hydrated phases (chemically bound water)

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Types of concrete in NPPs

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 Decomposition of concrete at high temp.

 100°C: Evaporation of free

and physically bound water
 ~400°C: Dissociation of
calcium hydroxyde
 600°C:  quartz
expansive transformation
 ~700°C: Dissociation of
calcium carbonate
 800°C: Total loss of
hydratation water
 1200°C: Melting starts

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Ablation and decomposition products

 Decomposition enthalpy 2.12.3 MJ/kg, Tliquidus~ 1573 K
 Ablation of concrete
 Bubbling of the concrete volatiles (H2O, CO2)
 Modification of convection regime (induced flow)
 Aerosol generation
 Mixing of concrete decomposition products with corium
 Chemical reaction

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 MCCI test

 MCCI with UO2 based melt

(Framatome)

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Hydrogen behavior and control

 TMI-2 accident triggered extensive research work on hydrogen

behavior in severe accidents

 Control H2: Installations of igniters and PARs

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 Siemens spark igniter
 Mitigation of hydrogen combustion
by early ignition
 Can prevent accumulation of large
H2 masses during the course of a
severe accident
 Early ignition generally leads to
standing diffusion flames at H2
release location.
 Safe implementation without
detonation transition is in principle
possible if average distance small
enough.

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Hydrogen behavior and control (contd.)

 A hydrogen explosion in Fukushima accidents

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 Source term

 Generic definition: The amount of radioactive or hazardous

material released to the environment from a facility following an
accident.
 Specific definition for NPP: The quantity, time history, and
chemical and physical form of radionuclides (radioactive fission
products) released to the environment, or present in the
containment atmosphere, during the course of a severe accident
 Assessment of the source term is important, since it is the essential
first step in assessing any exposures (to workers, or public)

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Fission products

 2000 kg of fission products (FP) in a PWR-900 at equilibrium

 Xe : 300 kg Kr : 22 kg
 Cs : 160 kg I : 13 kg
 Mo : 180 kg Ru : 140 kg
 Zr : 200 kg Ba : 80 kg
 Radioactive period (half-life):
 133Xe : 5 days 85Kr : 10 years
 131I : 8 days 137Cs : 30 years

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 Fission products (cont’d)

 Less dependent on accident scenarios but tightly connected with

fuel type (composition)
 Main task of source term assessment: Evaluation of FP passing to
the containment (timing, quantity, species)
 Pathway followed by the FP:
 FP release from the fuel rods
 Retention in the reactor coolant system (RCS)
 Behavior in the containment: Distribution, trapping for aerosols and
iodine, chemical forms

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Source term pathway

FPs and materials release from fuel

RCS: transport, physical processes,

chemical processes

Containment: transport, physical Containment bypass

processes, chemical processes,
engineered mitigation processes

Containment failure
Leaks through cracks

Release to environment

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 FP transformation

 In its pathway the Source Term is transformed.

 The transformation strongly depends on physical, chemical and
engineering processes.
 Physical processes:
 Aerosol physics and dynamics
 Chemical processes:
 Thermochemistry under radiation
 Iodine chemistry
 Engineering processes:
 Mitigation measures (spays, filters etc.)

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Aerosol physics and dynamics

 Basic processes

(Source: Multiphase Flow Handbook, Taylor & Francis -CRC Press, 2005)

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 Possible mitigation measures

 Various measures can be taken to stop SA progression

 Core reflooding
 If water available (pumps available)
 But risk of a large H2 generation and consequently risk of H2 explosion
 But risk of in-vessel steam explosion if corium pool formed
 Flooding the reactor pit
 If water available and reactor design previously adapted
 But risk of ex-vessel steam explosion
 Installation of a core-catcher (in or ex-vessel)
 Possible only at the reactor design stage

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SA issues identified in EURSAFE

D. Magallon, et al., 2005. European expert network for the reduction of uncertainties in 《Nuclear Power Safety》 61
severe accident safety issues (EURSAFE). Nuclear Engineering and Design 235: 309–346
 Severe accident research priorities (SARP)

W. Klein-Heßling, et al., 2014. Conclusions on severe accident research priorities. Annals of 《Nuclear Power Safety》 62
Nuclear Energy 74: 4–11–346

SARP (contd.)

W. Klein-Heßling, et al., 2014. Conclusions on severe accident research priorities. Annals of 《Nuclear Power Safety》 63
Nuclear Energy 74: 4–11–346
 SARP (contd.)

W. Klein-Heßling, et al., 2014. Conclusions on severe accident research priorities. Annals of 《Nuclear Power Safety》 64
Nuclear Energy 74: 4–11–346

Reading assignment

 Read Pages 549-581: In-vessel and Ex-vessel corium retention

strategies, Nuclear Safety in Light Water Reactors by Sehgal

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