Nuclear Engineering and Design 187 (1999) 79 – 119

EPR core design

G. Sengler a,*, F. Forêt b, G. Schlosser c, R. Lisdat d, S. Stelletta e
a
Nuclear Power International, 6 Cours Michelet, 92064, Paris, La Defense Cedex, France
b
Framatome, Tour Framatome, 92084 Paris La Defense Cedex, France
c
Siemens, Bunsenstraße 43, 91050 Erlangen, Germany
d
Preuben Elektra–Kernkraft, Tresckows Str. 5, 30457 Hanno6er, Germany
e
EDF SEPTEN, 12 – 14 A6enue Dutrié6oz, 69628 Villeuzbanne Cedex, France

Abstract

The nuclear industry has to face the increasing impact of deregulation, competition and new products like the EPR
have to rely on all possible means to reduce the generation costs for compensating the high initial investment. As far
as the core is concerned this reduction of generation costs is obtained mainly under given power level boundary
conditions by increasing the burn-up of the fuel and by providing the margins needed to the operator to adopt all
types of fuel managements which will allow to maximize the availability of the plant. Essentials of the EPR core
design basis and some representative results of basic design neutronic and thermal hydraulic studies are described in
the present paper for illustrating the potentials of the EPR under the boundary conditions prevailing at the end of
the Basic Design Phase. © 1999 Published by Elsevier Science S.A. All rights reserved.

1. Introduction find out, under given boundary conditions, which

are those providing the most favourable cost/
The EPR is a product for the beginning of the benefit ratio.
next century. It has to fulfil the highest require- At the end of the Basic Design Phase we were
ments with respect to its public acceptance and its just closing the first loop of iterations allowing a
competitiveness compared to alternative power comprehensive and consistent presentation of the
production systems. The EPR has a good chance product. The following description corresponding
to meet these ambitious targets as its development to this stage of the design work is providing a
is performed in very narrow collaboration with typical design basis, mainly used for calculation
major european utilities and meets the EUR (Eu- purposes. Later refinements are not at all ex-
ropean Utilities’ Requirements). As far as the cluded for example when entering in the detailed
essential core design features are concerned, there design or even at later stages.
have been long discussions and assessments to
The core design basis and major results of
design analyses which are presented in this paper
are giving a good idea about the already proven
capabilities and the potentials for evolution of the
* Corresponding author. EPR core.

0029-5493/99/$ - see front matter © 1999 Published by Elsevier Science S.A. All rights reserved.
PII: S 0 0 2 9 - 5 4 9 3 ( 9 8 ) 0 0 2 5 9 - 3
80 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

2. EPR-core design-summary description Gadolinium in the form of Gd2O3, the Gd enrich-

ment being in the range of 8–10 wt%. These
The core of the EPR with a thermal power of poisons are added in the present design to reach
4250 MW were designed mainly for being oper- negative moderator temperature coefficients inclu-
ated with UO2 fuel but having however a capacity sively at unrodded HZP-conditions and are dis-
for MOX recycling of 50%. Its main features and tributed over the fuel so as to reduce the radial
operating conditions were selected for reaching on peaking factors at the beginning of the cycle.
the one hand: Depending on the quantity of burnable poisons to
“ a high thermal efficiency of the plant and low be added for meeting the design or safety targets,
fuel cycle costs, and on the other hand, the number of Gd-rods per fuel assembly can be
“ an extended flexibility with respect to the fuel
varied from 8 to 20 as shown in Fig. 2 depending
cycle length and to the plant manoeuvrability.
on the type of fuel management.
The reactor core consists of an array of 241 fuel
The fuel rods are mechanically maintained ax-
assemblies, identical from the point of view of
ially and radially in the fuel assembly structure by
their mechanical design and directly derived from
at least ten Zircaloy made spacer grids. The grids
the product line of the fuel implemented in a large
number of PWRs when the first EPR is supposed located in the active part of the fuel have integral
to be loaded. These assemblies are of the type mixing vanes which are promoting the mixing of
17× 17 and are for both types of fuel (UO2 or the coolant and improving the thermal exchange
MOX) containing 264 fuel rods. This type of fuel between the cladding and the coolant.
offers the capacity to be loaded with UO2 fuel
with enrichments of up to 5 wt% of U235, single
fuel assemblies of this type remaining subcritical
when placed in pure water up to this enrichment
limit. This is one of the preconditions for reaching
a high fuel burn-up.
The fuel rods are made of Zircaloy tubings
containing originally:
“ either uranium dioxide ceramic pellets, of
which the U235 initial enrichment is as said
before below or equal 5.0 wt%. Some of the
fuel rods may also contain burnable-poisons
bearing fuel pellets which are in this case not or
much less enriched than pure UO2 pellets,
“ or uranium dioxide ceramic pellets made of
depleted uranium where Pu is added in the
form of PuO2 as fissile material. The maximum
envisaged fissile Pu-enrichment in one fuel rod
is limited due to factory constraints to 7.44
wt% (12 wt% for total Pu content). For neu-
tron physical reasons, this fissile Pu enrichment
is limited to 7.0 wt% in the average in every
MOX-assembly. MOX-assemblies are therefore
containing different zones with different Pu
enrichments (see Fig. 1)
The nature of the burnable poisons, comixed to
the UO2 support in order not to generate addi-
tional volumes of wastes is assumed to be Fig. 1. Radial description of a MOX assembly.
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 81

same type, consisting of 24 individual and identi-

cal absorber rods, fastened together to a common
spider assembly. These rods are all containing
neutrons absorbing materials over a length which
is aimed to cover nearly the complete active
length of the fuel. At the present stage, hybrid
rods (AIC-B4C) are considered for the design,
corresponding to an existing and proven technol-
ogy:
AIC, being implemented in the lower part over
a length of 1.0 m.
B4C at the upper part, over a length of 3.11 m.
However, it is known that new technical solu-
tions improving the long term mechanical be-
haviour under irradiation or the physico-chemical
behaviour of the rods under extreme conditions
and simultaneously improving their neutron phys-
ical efficiency are under development. But as these
new solutions, not yet precisely defined, will be
necessarily compatible with previous techniques
and additionally offer improved performances,
the design based on present techniques is consid-
Fig. 2. Burnable poison rods arrangement within an assembly. ered being conservative.
The core is cooled and moderated by light
A total of 25 positions in the 17 × 17 array are water at a pressure of 155 bar. The moderator
equipped with guide thimbles which are joined to coolant contains boron (possibly 10B enriched) as
the grids and to the top and bottom nozzles. neutron absorber. The boron concentration in the
These guide thimbles are used as locations for rod coolant is varied for controlling slow reactivity
cluster control assemblies (RCCA) and possibly changes necessary for compensating Xe-poisoning
for the fixed or moveable incore instrumentation or burn-up effects during operation at power and
fingers or for the neutron source assemblies. In for the compensation of large reactivity changes
the present design the central guide thimble is not associated to large temperature variations during
used and could be, in the future replaced by a fuel cool-down or heat-up phases.
rod.
The bottom nozzle of the fuel assembly which
serves as the bottom structural element is shaped 3. Nuclear and thermalhydraulic designs
for directing and to a certain extent equalizing the
flow distribution and also for filtering out small 3.1. Design basis
debris.
The top nozzle of the fuel assembly is the upper The design basis for nuclear and thermalhy-
structural element of the fuel assembly. This top draulic design is the sum of operational require-
nozzle also supports the hold down springs of the ments and safety requirements.
fuel assembly and offers an appropriate housing The major EPR core features like the type of
for the upper part of the RCCA. The fuel element fuel assembly, the size of the core, the introduc-
has a rotation symetry of y/2 which provides tion of a heavy reflector and the main operating
maximum flexibility for reloadings. parameters were completely driven by the objec-
The core will have a fast shutdown system tive to maximize the plant efficiency and to imple-
made of up to 89 RCCAs. All RCCAs are of the ment margins allowing the maximization of the
82 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

core flexibility with respect to the duration of the The nuclear design for the moment concen-
fuel cycles, the capacity to fulfil European grid trates on cores featuring some of the most de-
requirements, the recycling of significant quanti- manding types of fuel managements. It is
ties of Pu and simultaneously the objective to postulated that this approach gives sufficient free-
minimize as much as reasonably achievable the dom to accomodate later a large spectrum of
fuel cycle costs mainly by improving the neutron other types of less demanding managements. The
ecomony and increasing the fuel burn-up. neutronic and thermalhydraulic designs consist of
The more detailed EPR core data today avail- defining the major core parameters and associated
able are the results of first nuclear and thermalhy- systems characteristics which allow to meet under
draulic design analyses performed with the the above mentioned boundary conditions the
following orientations and assumptions: following safety or decoupling criteria:
1. The steam pressure in the steam generator For normal operating conditions (PCC1)2, con-
shall be of 72.5 bar under best estimate or trols, surveillance and limitation systems are
thermal hydraulic (conservative) assumptions maintaining automatically the plant within the
concerning the coolant flowrate. LCO3 limits postulated for accident analyses and
2. The core is designed for reaching a batch- thus far from the integrity limits of the fuel clad-
burn-up of at least 55 and possibly ] 60 ding. These systems are relying on efficient, accu-
GWd/tHM. This objective is consistent with rate and reliable instrumentation concepts.
the expected improvements of fuel perfor- In case of PCC2 events (anticipated operational
mances in progress on existing plants. occurences), in general, automatic countermea-
3. Margins are available to provide large flexibil- sures (limitations) will be actuated with the aim to
ity for defining the loading schemes. OUT/IN terminate the abnormal transients at an early
and IN/OUT types of loadings are possible. stage and to return as far as possible the plant in
4. The cycle length of 18 months is taken as PCC1 conditions without tripping the plant. The
design basis for optimization. trip function (protection) relying on the accurate
5. Two years cycles are feasible. monitoring of essential core parameters is actu-
6. Extended Pu recycling is possible up to a 50% ated only when there is no other solution or when
MOX-core, on the basis of the 18 months-cy- the first automatic actions did not succeed to
cle and the two types of loadings. terminate the transient.
7. A systematic stretch-out operation of up to 70 For both PCC1 conditions and PCC2 events,
EFPD is allowed. there shall be no loss of integrity of the fuel. For
8. The core control is defined for meeting the PCC2 events and events of lower probability of
manoeuvrability requirements expressed in the occurrence which are resulting in a plant shut-
EUR grid requirements: capacity for sched- down, the target as far as the shutdown capabili-
uled or unscheduled large and fast power vari- ties are concerned is to bring and maintain the
ations over the total range of 20 – 100% RP, plant in subcritical conditions up to and during
capacity for permanent primary and secondary the safe shutdown state relying on safety grade
control in a range of 9 12.5% between 50 and means in the frame of the safety demonstration.
100% RP (Lisdat, Miossec1, Sengler, POWER-
GEN 96: Flexibility of Power generation with 3.2. Neutronic design
the EPR based on the French and German
experience). It is optimized for minimizing the 3.2.1. Number and type of fuel assemblies
mechanical and thermal stresses and the excess The large size of the core characterized by:
of core margins needed to cover the complete fuel assemblies of the type 17× 17 and,
spectrum of corresponding types of operation.

2
1
Claude Miossec: EDF–EPN Site Cap–Ampère, 1 place PCC stands for plant condition category.
3
Pleyel 93282 St Denis Cedex, France. LCO stands for limiting conditions of operation
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 83

an active length of 420 cm, tions category 1 and 2 events including the
was selected mainly for economical reasons maximum overpower condition,
(Forêt, Schlosser, Sengler, TOPNUX 96: EPR 4. the fuel management will be such that the
Fuel Cycle cost reduction). It significantly con- resulting power and burn-up distributions will
tributes to improve the neutron economy, by of- be consistent with the assumptions taken for
fering a low surface to volume ratio and lowering demonstrating the mechanical integrity of the
the power density which results in a lower xenon fuel rods.
poisoning and a reduced Doppler feedback. The Based on past experiences, as a first de-coupling
implemention of a heavy reflector with an average assumption for starting the neutronic and ther-
thickness of 19.4 cm provides an additional neu- malhydraulic design, the target fixed for defining
tron economy corresponding to an equivalent sav- the loading patterns was not to exceed a best
ing of 0.15 and 0.06 wt% of U5 enrichment, estimate value of 1.6 for the F N DH (integrated

respectively for the first and later equilibrium relative fuel pin power after reconstruction of the
cycles. The low linear heat generation rate of 3-D pin by pin power map) for the most demand-
154.9 W cm − 1 additionaly provides margins ing types of fuel managements. The performed
which allow to increase the burn-up by increasing neutronic calculations for all relevant types of
the U5 enrichment up to 5.0 wt% and conse- loadings, including the transition cycles have
quently the reciprocal reload fraction. shown that this target can be met (as illustrated in
The requested cycle lengths of 12, 18 and 24 Table 3, column 10).
months are reached approximately with reciprocal In fact the justification of the design is based on
the calculation of extreme power shapes which
reload fractions of, respectively, 6, 4 and 3. This
may affect fuel design limits. These calculations
means that even with very long fuel cycles one can
are performed with proven methods verified fre-
meet the requested fuel burn-up target.
quently with measurements from operating reac-
It can be seen also in Table 3 that for the
tors. The conditions under which limiting power
envisaged types of loadings characterized in
shapes are assumed to occur are chosen conserva-
columns 1–4, the operational and economical
tively with regard to any permissible operating
targets with respect to cycle lengths and fuel
state. In addition, uncertainties are considered for
burn-up are reached (respectively, columns 8 and covering calculation uncertainties and allowances
9). for manufacturing tolerances.
The calculation of the power distributions is
3.2.2. Fuel managements and power distributions relying on a two-energy group 3-D model diffu-
The margins mentioned above are sufficient sion code incorporating the most recent nodal
also to cover more demanding 50% MOX-cores technologies. In this modelization, the fuel assem-
and low leakage loadings which are improving the bly is represented radially by four nodes and by
fuel cycle economy in the average by at least 2.5% 20 nodes in axial direction, from which 18 for the
compared to OUT/IN loadings (see examples of active part. The local reconstruction of the flux,
different types of studied loading patterns in Figs. power, burn-up and reaction rates is based on a
3 – 9) with respect to the power distribution, the combination of homogeneous intranodal fluxes
nuclear design basis is that, with at least a 95% computed at each step and tabulated power form
confidence level: factors. The homogeneous intranodal flux is re-
1. the fuel power density remains below 450 W constructed using surface currents, surfaces fluxes,
cm − 1 under normal conditions (Fq 52.90), corner point fluxes and the nodal average flux.
2. the fuel peak power under abnormal condi- The power form factors are the results of lattice
tions including the maximum overpower con- computations performed by a 2-D code solving
dition will not cause local fuel melting, the Boltzmann transport equation using a multi-
3. the fuel will not operate with a power distribu- group formalism and a collision probability
tion that violates the departure from nucleate method for various 2-D geometries and burn-up
boiling (DNB) design basis under plant condi- conditions.
84 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

Table 1
Essential core and primary system parametersa,b

BE-conditions TH-conditions

Nuclear power 4250 MW

Number of fuel assemblies 241
Number of fuel rods/fuel assembly 264
Fuel assembly length (cold) 485 cm
Active length (cold) 420 cm
Average linear power (at rated power) 154.9 W cm−1
Power density in hot conditions 89.3 kw dm−3
Radial peaking factor (F 2-D
xy , best estimate value at rated power, w/o control rods, :1.6
obtained by 2-D-calculations)
Diameter of fuel rods 0.95 cm
Fuel rod pitch 1.26 cm
Total thermal design flowrate 21 895 kg s−1 21 035 kg s−1
Core by-pass flowrate 5.19% 7%
Coolant nominal inlet temperature 291.8°C 291.5°C
Mean coolant nominal core outlet temperature 327.1°C 328.8°C
Mean coolant nominal vessel outlet temperature 325.5°C 326.5 3°C
Steam pressure in steam generator 72.5 bar 72.5 bar
Acti6e core
Equivalent diameter (mm) 3767
Core average active fuel height, first core (mm, cold dimensions) 4200
Height-to-diameter ratio 1.115
Total cross-section area (cm2) 111450
Hea6y, radial reflector thickness and composition
Water plus steel (mm) Between 77 and
297 (average
194)
Fuel rods
Number 63624
Outside diameter (mm) 9.50
Diametral gap (mm) 0.15
Clad thickness (mm) 0.625
Clad material Zircaloy
Pu 6ector for MOX fuel assemblies (% wt)
Pu238 4.0
Pu239 50.0
Pu240 53.0
Pu241 12.0
Pu242 9.5
Am241 1.5
Co-mixed burnable poison
Material Gd2O3
Gadolinium enrichment (% wt) 8
UO2 carrier enrichment for all types of fuel management except 0.25
UO2-INOUT-24 months (% wt)
UO2 carrier enrichment for UO2-INOUT-24 months (% wt) 2.5
Absorber rod cluster control assemblies
Absorber (2)
1) AIC part (lower part) composition
Ag 80.0%
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 85

Table 1 (Continued)

BE-conditions TH-conditions

In 15.0%
Cd 5.0%
Density (g cm−3) 10.17
Absorber outer diameter (mm) 7.65
Length (mm) 1000
2) B4C part (upper part) composition: natural boron with 19.9 atoms of B10 percent
Density (g cm−3) 1.79
Absorber diameter (mm) 7.47
Length (mm) 3110

Pu239+Pu241
a
The fissile Pu enrichment is defined as:e = .
(U+Pu+Am)
b
It is important to note that hybrid AIC-B4C control rods are considered only as a calculation reference absorber in term of
integral worth. This type of rods may be replaced later on by new types resulting from R&D work and having at least the same
efficiency.

The target of the core designer is always to find in a way to avoid axial power peaks.
appropriate solutions for minimizing the local As far as radial power distributions are con-
power peaks. In this attempt, the classical way is cerned, the shape in horizontal section is a func-
to consider separately the means to reduce radial tion of:
peaking factors and the means to operate the core “ the loading pattern of fuel assemblies,
“ the location of poisoned rods,
“ the insertion of rod cluster control assemblies
“ the core burn-up,
“ the power level and the moderator density,
“ the concentration and the distribution of xenon
and samarium.
On the contrary, the effect of non-uniform flow
distribution is negligible.
With respect to radial power distributions the
designer defines the fuel management (loading
pattern, introduction of Gd-rods...) and the loca-
tion of the control rods which at the same time
correspond to the best compromise between eco-
nomical optimization and minimization of radial
peaks.
For illustration, the figures Figs. 10–23 are
showing the radial power distributions per assem-
bly for one eighth of the core for different burn-
up steps of the cycle 1 and of the equilibrium
cycle of the different types of fuel managements.
These distributions are obtained by integrating
the nuclear power over the core height.
As already mentioned above, the power of the
hot channel in the core results from the superim-
position of the macroscopic power distribution in
Fig. 3. First core loading pattern. the core and the pin by pin distribution in the
86 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

Fig. 4. Reloading pattern for UO2-OUTIN-18 months equilibrium cycle.

assembly. For the purpose of illustration, assembly are shown in Figs. 28 and 29, corresponding to full
pin by pin power distributions at beginning of life power states without rods inserted.
and end of cycle 1 are given for the same assembly The means to limit the axial power distribution
in Figs. 24 and 25, respectively. Other examples are mainly linked to the core control principles.
Figs. 26 and 27 are, respectively given for an Signals are available from neutron flux incore or
assembly with Gd pins and for a Mox assembly. excore instrumentation. These signals are used for
With respect to the radial power distribution used the core control during normal operation to deter-
for DNB calculations, a single reference radial mine the average axial power distribution of the
design distribution is selected for reasons of sim- core which is characterized either by the axial offset
plification. This design power distribution is chosen (AO) or the axial power difference DP:
to be conservative, enveloping all types of calcu-
PT − PB
lated power distributions and minimizing the AO=
PT + PB
benefits or flow redistributions.
On the other hand, the shape of the power where PT and PB are the power fraction in the top
distribution in the axial direction depends mainly and bottom halves of the core.
on:
DP= PT − PB = AO× Pr
“ the insertion of control rods,
“ the power level, where Pr is the relative power level in comparison
“ the axial xenon distribution, with the nominal power level.
“ the feedback effects resulting from Doppler For minimizing axial power distributions, the
effect and moderator density, core control has to be designed appropriately
“ the fuel burn-up. (Section 3.4). This is the case for the EPR where
For the purpose of illustration representative control rods are nearly totally withdrawn at full
axial power shapes for cycle 1 at different burn-ups power. In addition, axial oscillations due to Xe
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 87

Fig. 5. Reloading pattern for UO2-INOUT-18 months equilibrium cycle.

are prevented by maintaining automatically the 3.2.3. Boron concentration, reacti6ity coefficients,
axial offset within a narrow operating band, keep- shutdown efficiency
ing the axial Xe-distribution in phase with the Following the defense in depth concept, the core
actual power distribution. is designed to have stabilizing reactivity coefficients
Finally, the worst or limiting power distributions and to have an efficiency of the shutdown system
which can occur during normal operation (plant which allows to meet the safety objectives.
condition category 1 events) have to be considered Typical values for boron concentrations (ex-
as the starting point for analysis of plant condition pressed considering natural boron) and moderator
category 2, 3 and 4 events. These limiting power temperature coefficients are shown for illustration
distributions are at the present stage of design in Table 3, respectively in the columns 5–7 for the
generated in a conservative way; later on, on the different types of fuel managements. The number
real plant, the limitations of the maximum linear of burnable poisons indicated in column 3 is
power density Q(z) and of the DNBR, based adjusted at present partly to limit the critical boron
mainly on fixed incore instrumentation, will ensure concentration at HZP, all rods out at BOL, in such
that the real power distributions measured online a way that the moderator temperature coefficient
are remaining less penalizing than those generated temperature under these conditions remains nega-
for the design (Section 3.5.3.2). In practice, due to tive, including consideration of uncertainties. This
the the conservative approach on power distribu- is at present an overconservative constraint which
tions, there are margins available which may be could be relaxed by taking appropriate counter-
used in an appropriate way in a next iteration step. measures, e.g. by performing the accident analyses
The major core characteristics, important for the with more positive moderator temperature coeffi-
nuclear design are gathered in Table 1. cients. All the other feed back coefficients (fuel
88 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

Fig. 6. Reloading pattern for UO2-INOUT-12 months equilibrium cycle.

temperature coefficient, power coefficient, void assumption based in a first stage on past expe-
coefficient) are negative. rience with respect to the analysis of subcool-
The requirements for the design of the shut- ing accidents of plants of the previous
down system are the following: generation and confirmed later by real accident
“ Maintain subcriticality assuming one stuck rod analyses for the EPR). This subcriticality in-
after actuation of the reactor trip until condi- cludes, as before, a 500 pcm margin for provid-
tions are reached at which the boration with ing additional flexibility for the reloadings and
safety classified systems is effective. This crite- transition cycles.
rion consists in reaching a subcriticality of 500 “ Maintain in the long term the plant succritical
pcm after reactor trip and cooldown of the at HZP after a reactor trip from full power.
primary circuit to 260°C (conservative decou- This means that with all rods in, the subcriti-
pling assumption) for all types of fuel manage- cality after trip has to be sufficient to cover the
ments with consideration of an additional 500 long term depletion of the xenon in order to be
pcm margin for covering the variability of the able to cope with a loss of the operational
reshuffling in a real plant (in order to minimize boration function.
possible constraints on reshufflings). The sizing of the shutdown system is taking
“ Compliance with PPC 4 design criteria after a into account as well allowances for covering ini-
main steam line break, assuming one stuck rod. tial conditions of accidents in accordance with the
In this case, a temporary return to criticality is defined limiting conditions of operation as uncer-
considered being acceptable. In terms of reac- tainties of the design tools and of measurement
tivity, this criterion corresponds to request a systems. To the limiting conditions of operation
subcriticaity of 2500 pcm at HZP after reactor considered for this purpose are belonging parame-
trip from full power (conservative decoupling ters like:
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 89

Fig. 7. Reloading pattern for UO2-INOUT-24 months equilibrium cycle.

“ the reactor power, long fuel cycles on the other, high boron con-
“ the initial insertion of control rods (maximum centrations are needed at the beginning of cy-
needed negative reactivity inserted for control cles. These high boron concentrations might not
purposes), be suitable from the point of view of the clad-
“ the initial axial power and Xe distributions. ding corrosion. It is therefore envisaged under
A RCCA pattern which inclusively covers the these conditions to operate the plant with B10
needs for MOX-cores is shown in Fig. 30. Mini- enriched boron.
mum shutdown margins calculated at bounding At present, the proposal would be to operate
end of cycle conditions for the relevant fuel the plant with at least 28.50 wt% of B10 for UO2
managements are shown in Table 3, column 12. cores with the highest U5 enrichment and 36
It can be seen that the requirements are over wt% of B10 in case of 50% Mox cores. These
fulfilled for uranium cores for which less RC- boron enrichments would allow to operate the
CAs would in fact be needed. plants at boron concentrations B 1400 ppm at
The shutdown and the boration systems are BOL without Xe.
designed also to satisfy long term requirements For covering all the needs, the theoretical
after reactor trip, corresponding to subcriticality boron concentration of the IRWST has to be in
requirements for all types of shut down condi- the range of 3200 ppm when considering natural
tions covering operational and accidental scenar- boron, 2260 ppm when considering enriched
ios.Due to the relatively low efficiency of the boron. The real IRWST boron-concentration
boron observed for high enriched uranium cores will be somewhat higher to account for uncer-
and its even lower efficiency in case of MOX- tainties and partial dilutions which may inter-
cores on the one hand and the consideration of vene in some accidental scenarios.
90 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

Fig. 8. Reloading pattern for MOX-OUTIN-18 months equilibrium cycle.

3.3. Thermal hydraulic design and core parameters The major parameter from thermal hydraulic
point of view is the DNB (departure from nucleate
The purpose of the thermal hydraulic design is boiling), depending on main parameters such as
to provide an adequate heat transfer, coherent with the mean heat generation, the power distribution,
the distribution of the heat generation such that the coolant temperatures and pressures and the
the heat removal by the reactor coolant system or core flowrate.
by the safety injection system allow to fulfill: The DNB also depends on the detailed geomet-
“ on the one hand operational targets in terms of rical characteristics of the real fuel, location of
available margins to minimize the constraints mixing grids and efficiency of mixing vanes etc.
on core loadings and allow easy load follow and Empirical DNB correlations are generally estab-
lished by every fuel supplier for a given fuel, based
frequency control operation,
on extensive DNB tests performed on thermal
“ and on the other hand to meet the safety targets.
hydraulic test loops.
The thermal hydraulic design consists in princi-
For the EPR the main fuel characteristics are
ple of defining the main thermal hydraulic parame- known, but not the details which are specific for a
ters and then of sizing the setpoints of limitation given supplier. Thus the thermal hydraulic design
and protection functions in an iterative optimiza- of the EPR is not based on a specific DNB
tion process in order to meet these design targets. prediction correlation.
In fact, for the EPR with its large core and low The way selected to overcome this situation was
linear heat rate, there is no major problems to meet to make use of CHF tables of the open literature
all these targets, so that there was no need for and to apply them to reactor design calculations
iterative work since the main primary components after having performed appropriate validation
were selected to rely on already existing designs. tests and set an appropriate criterion in terms of
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 91

Fig. 9. Reloading pattern for MOX-INOUT-18 months equilibrium cycle.

DNBR. For doing that, the CHF values obtained By preventing DNB, an adequate heat transfer
by the tables were compared to the results of tests between the fuel clad and the reactor coolant is
performed on real rod bundles of different sets of ensured. The prevention of DNB is relying on
presently known advanced fuels. The resulting appropriately sized limitation and protection
DNB predictor is enveloping all considered fuels functions based on on-line DNBR calculations.
and can be considered conservative at the present These I and C functions are using fixed incore flux
stage of the design, providing margins for future measurements to reconstruct the local thermal
actual fuel applications. hydraulic conditions and to calculate the CHF.
The DNBR values characterizing the operation The reconstruction uncertainties as well as un-
of the EPR and preliminary setpoints of limita- certainties related to the fuel geometry and to the
tion and protection functions where obtained via thermal hydraulic modelization are considered for
this particular methodology and cannot be com- sizing the setpoints, the principle of sizing being
pared directly with values obtained for other that there is a 95% probability at a 95% confi-
plants with DNBR correlations applying for spe- dence level that the DNB will not occur when the
cific fuels. on-line calculated DNBR threshold is reached or
when other protective functions have been actu-
3.3.1. Thermal hydraulic design with respect to the ated.
DNBR The methodology used for defining the
The design basis on DNB is to ensure with a thresholds of the I and C functions used to ensure
probability of 95% that DNB will not occur on the thermal hydraulic design criteria is based
the limiting fuel rods during normal operation mainly on the analyses of events of PCC 2 for
and anticipated transients (PCC 1 +2) at a confi- which the integrity of the first barrier is manda-
dence level of 95%. tory. But there are also some other particular
92 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

Fig. 10. Normalized power density distribution near beginning of life unrodded core, hot full power, equilibrium xenon for cycle 1.
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 93

Fig. 11. Normalized power density distribution near end of life unrodded core, hot full power, equilibrium xenon for cycle 1.
94 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

Fig. 12. Normalized power density distribution near beginning of life unrodded core, hot full power, equilibrium xenon for
UO2-OUTIN-18 months.
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 95

Fig. 13. Normalized power density distribution near end of life unrodded core, hot full power, equilibrium xenon for UO2-OUTIN-
18 months.
96 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

Fig. 14. Normalized power density distribution near beginning of life unrodded core, hot full power, equilibrium xenon for
UO2-INOUT-18 months.
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 97

Fig. 15. Normalized power density distribution near end of life unrodded core, hot full power, equilibrium xenon for UO2-INOUT-
18 months.
98 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

Fig. 16. Normalized power density distribution near beginning of life unrodded core, hot full power, equilibrium xenon for
UO2-INOUT-12 months.
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 99

Fig. 17. Normalized power density distribution near end of life unrodded core, hot full power, equilibrium xenon for UO2-INOUT-
12 months.
100 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

Fig. 18. Normalized power density distribution near beginning of life unrodded core, hot full power, equilibrium xenon for
UO2-INOUT-24 months.
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 101

Fig. 19. Normalized power density distribution near end of life unrodded core, hot full power, equilibrium xenon for UO2-INOUT-
24 months.
102 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

Fig. 20. Normalized power density distribution near beginning of life unrodded core, hot full power, equilibrium xenon for
MOX-OUTIN-18 months.
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 103

Fig. 21. Normalized power density distribution near end of life unrodded core, hot full power, equilibrium xenon for MOX-OUTIN-
18 months.
104 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

Fig. 22. Normalized power density distribution near beginning of life unrodded core, hot full power, equilibrium xenon for
MOX-INOUT-18 months.
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 105

Fig. 23. Normalized power density distribution near end of life unrodded core, hot full power, equilibrium xenon for MOX-INOUT-
18 months.
106 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

Fig. 24. Rodwise power distribution in typical assembly near beginning of life, hot full power, equilibrium xenon for cycle 1.

events of PCC 3 and 4 for which this integrity is protection channel used to protect the core. Three
requested too for reasons of simplification of the cases are considered:
safety demonstration (decoupling assumption) in “ Transients of type 1 for which the DNBR
particular in case of a failure of another barrier. protection is efficient. These transients occur at
The methodology used for sizing with respect to power and are relatively slow. For these tran-
DNB in fact depends on the types of reactor sients, the DNB is avoided by setting the
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 107

Fig. 25. Rodwise power distribution in typical assembly near end of life, hot full power, equilibrium xenon for cycle 1.

DNBR threshold of the DNBR protection pared to the response time of this protection:
channel at a limit which guarantees non occur- For these transients, the protection is based on
rence of DNB, under consideration of all types specific protections detecting the event (like
of uncertainties. ‘low pump speed’, for detection of loss of RCS
“ Transients of type 2 which occur at power but flow). In this case, one has to consider that
for which the DNBR protection channel is not once the setpoints for these specific protections
efficient. These transients are too fast com- have been fixed, the minimum DNB during the
108 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

Fig. 26. Rodwise power distribution in typical assembly with 20 GD pins at HFP, BOL.

corresponding transients only depends on the DNBR is belonging to these parameters which
initial conditions at which the event occurs. In are submitted to operational constraints. A
order not to overshoot the DNB-limits during surveillance/limitation function is in charge to
the transient, the limiting conditions of opera- ensure that the actual DNBR always exceeds
tion which are defining the worst initial condi- the DNBR threshold fixed for initial conditions
tions for these events have to be defined by of accidents also called DNBRLCO (mainly with
appropriate accident analyses. The actual regard to the loss of flow event). This setpoint
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 109

Fig. 27. Rodwise power distribution in typical MOX assembly at HFP, BOL.

is defined taking into account all the conditions. For these events, the methods
uncertainties linked to the fuel geometry, and defined for the previous types cannot apply.
those related to the surveillance/limitation Here, specific protection functions and safety
functions. systems must intervene. The aim of the
“ Transients of type 3 occuring at very low corresponding accident analyses will be the
power level or at subcritical conditions or sizing of these protections and systems for
leading to recriticality at low temperature ensuring that the minimum DNBR limits
110 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

Table 2
Thermal hydraulic design data

Total core heat output (MWth) 4250

Number of loops 4
Nominal system pressure (absolute) (MPa) 15.5
Coolant flow
Core flow area (m2) 5.9
Core average coolant velocity (m s−1) 5.0
Core average mass velocity (g cm−2 s−1) 331.5
Total mass flow rate/loop (kg s−1) 21035
Effective mass flow rate (t h−1) 78250
Best estimate flow/loop (m3 h−1) 25460
Thermal design flow/loop (m3 h−1) 26520
Mechanical design flow/loop (m3) 27581
Coolant temperature (°C)
Nominal inlet 291.5
Average rise in vessel 35
Average rise in core 37.3
Fig. 28. Typical axial power shape equilibrium xenon, occur- Average in core 310.15
ing at beginning of life cycle 1. Average in vessel 309
Heat transfer
guarantying the integrity of the fuel will not be Heat transfer surface area (m2) 7975
violated during the accident. Average core heat flux (W cm−2) 51.9
This approach and the results of these sizing Maximum core heat flux (nominal 146.9
analyses are presented in Fig. 31 operation) (W cm−2)
Average linear power density (W cm−1) 154.9
Peak linear power for normal operating 450
3.3.2. Core flow design basis and thermal conditions (FQ = 2.93), (W cm−1)
hydraulic main characteristics Hot spot pellet center temperaturea (°C) 1770
The core cooling is ensured by the primary Peak linear power protection threshold 590
coolant flow. The flowrate and the primary (W/cm)
Hot spot pellet center temperaturea (°C) 2200
DNB ratio: minimum DNBR under nominal
operating conditions with
FDH= 1.75−cos 1.45 2.33
FDH =1.60−cos 1.45 2.77
Fuel assembly
Number of fuel assemblies 241
Fuel assembly pitch (cm) 21.504
Active fuel height (cm) 420
Lattice pitch (cm) 1.26
Number of fuel rods per assembly 264
Number of control rod assembly or 25
instrumentation guide thimbles per
assembly
Outside fuel rod diameter (cm) 0.95
Guide thimble diameter (cm) 1.225
Core performance characteristics
Power density in hot conditions [KW (core 89.3
litre)−1]
Power density in hot conditions (KW (fuel 99.3
litre)−1]
Fig. 29. Typical axial power shape equilibrium xenon, occur-
ing at end of life cycle 1. a
Values for UO2 unirradiated fuel (dependent on burn-up).
Table 3
Neutronic design/major characteristics of typical fuel managements

Batch size No. of FA per No. of Gd Enrichments amod BOL- CB at BOL CB at BXL Cycle length Average dis- Max Fxy over Max Fq over Subcriticality
type pins per FA U5, Pu in HZP-ARO (ppm) assum- (ppm) assum- (MWd charge BU cycle, BE, cycle, BE, (pcm) at
wt% (pcm °C−1) ing natural ing natural tHM−1)/ (MWd ARO ARO HZP/ at
boron boron EFPD tHM−1) 260°C after
trip from
HFP (N-1)

Column No. 1 2 3 4 5 6 7 8 9 10 11 12
UO2-core, 1st 109 0 1.9 U5 −6.0 960 688 16060/468 16880 1.505 2.246 6007/3956
cycle 84 20 3.0 U5

G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

28 12 3.6 U5
20 0 3.6 U5
UO2-core, 40 24 8 4.9 U5 −15.8 1474 1093 10691/312 64660 1.581 1.879 5025/2613
equilibrium 16 0 4.9 U5
12 months-
cycle IN/
OUT
loading
UO2-Core, 18 60 36 8 4.9 U5 −6.2 1983 1582 15112/441 60142 1.539 1.864 4657/2253
months 24 0 4.9 U5
equilibrium-
cycle IN/
OUT
loading
UO2-core, 18 60 16 12 4.9 U5 −8.8 1851 1465 14710/429 59766 1.523 1.783 3515/1013
months- 44 0 4.9 U5
equilibrium
cycle OUT/
IN loading
UO2-core, 24 80 44 16 4.9 U5 −4.9 1992 1585 18888/549 56540 1.593 1.957 4240/1888
months 36 12 4.9 U5
equilibrium-
cycle IN/
OUT
loading
50% MOX- 68 28 16 4.9 U5 −17.0 2294 1832 16576/483 UO2: 59240 1.566 1.847 2821/602
core, 18 40 7.0 Pu MOX: 58620
months-
equilibrium
cycle, IN/
OUT load
ing
50% MOX- 68 16 8 4.9 U5 −17.2 2220 1790 15813/462 UO2: 63048 1.570 1.821 3524/1242
core, 18 12 0 4.9 U5 MOX: 51412
months- 40 7.0 Pu
equilibrium
cycle, OUT/
IN loading

111
112 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

coolant temperature together have to be uncertainties and allowances in the direction

optimized for ensuring the heat transfer to the which maximizes the efforts on the
secondary side in an optimized way for reaching components.
on the one side high secondary steam pressure As far as the thermal hydraulic core design is
with acceptable heat exchange surfaces in the concerned, it must be known that not the total
steam generators and on the other hand flowrate generated by the reactor coolant pumps
ensuring acceptable thermal hydraulic conditions will enter the fuel region of the core. A part of
in the core. the flow is bypassing the fuel rods through the
In practice, three different flowrates are thimble tubes of the fuel or is used to cool the
considered for the design, depending on the heavy reflector and to maintain as far as
purpose: possible cold leg temperature conditions in the
“ A ‘thermal’ flowrate minimizing the flow upper dome of the RPV. This last feature was
entering the core used for the thermal design introduced in an early phase of the development
of the core. This flowrate considers all the and might be dropped in a later design phase.
uncertainties or allowances in a penalizing For the evaluation of the ‘thermal’ flow, a total
way for the core cooling. bypass of 7% is considered up to now.
“ A ‘best estimate’ flowrate which is used to The reactor coolant system temperature is a
predict the secondary side pressure under best function of the reactor power. The relationship
estimate assumptions and to design the between the primary coolant temperatures and
reactor coolant pumps. the power level are illustrated in the part load
“ A ‘mechanical’ flowrate maximizing the flow diagrams shown in Figs. 32 and 33, respectively
entering the core which is used for the for thermal hydraulic conditions and best
mechanical design of the components. This estimate conditions, the principle being that at
mechanical flowrate considers the nominal power, the secondary steam pressure be
always 72.5 b.
The shape of the part load diagram was de-
fined for having approximately a constant
primary temperature in the range of 50–100%
RP. The reason for this choice was the
minimization of the thermal loads and
mechanical stresses and wear on primary
components during load follow operation, in the
power range where the plant is believed to be
operated most of the time.
The main thermal and hydraulic chara-
cteristics of the EPR are gathered in Table 2.

3.4. Core control principles/control rods/control

rod manoeu6ring

The reactivity of the core is controlled at

power by the means of changes of the boron
concentration and the insertion of RCCAs.
As a general rule, slow reactivity variations
resulting either from changes of the xenon
concentration (e.g. following daily load
Fig. 30. Possible RCCA pattern covering enveloping require- variations) or due to the evolution of the
ments. burn-up are compensated by adjusting the boron
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 113

Fig. 31. Transient analysis method.

concentration. Fast reactivity changes necessary “ The control of RCCA-position with respect to
for adapting the power level are obtained by shutdown efficiency.
modifying the RCCAs insertions. The essential features of the core control
The core control relies on the three main closed systems are the following:
loop controls: “ All the RCCAs are ‘black’ rods.
“ The reactor coolant temperature control. “ At rated power, no control rods shall be deeply
“ The axial power distribution control. inserted.
114 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

Fig. 32. Part load diagram at thermalhydraulic flowrate.

“ The power defect reactivity due to power level plant operation by adjusting appropriately the
variations is essentially covered by movement setpoints.
of control rod groups (temperature control). The part load temperature diagram is defined
“ During load follow operation long term to favour load variations between 50 and 100%
reactivity changes (BU, Xe) are mainly of rated power. Over this range, the mean
compensated by modifications of the boron coolant temperature is maintained approximately
concentration. At partial load the RCCAs are constant at : 308.7°C. This reduces the thermal
remaining inserted as much as necessary for and thermomechanical stresses in systems and
meeting the spinning reserve required by the components (less volume contractions, less
dispatching. actuations of fluid systems, less steps of control
“ The axial power distribution is mainly rods for compensation of reactivity variations).
controlled at high power levels by moving In this range the plant has its highest capacity
slightly inserted control groups and at lower to perform fast load variations with high
power levels by adjusting the overlap of the amplitudes or numerous load variations of
already inserted control groups. smaller amplitudes associated to primary and
“ In order to ensure sufficient shutdown margins, secondary frequency control.
the amount of negative reactivity inserted by Over the range of 20–50% of rated power, the
RCCAs is automatically adjusted by modifying secondary pressure is maintained nearly
the boron concentration in the coolant water. constant, whereas the average primary
“ Rod dropping is used for fast power reductions temperature is reduced from : 308.7°C at 50%,
in case of perturbed operating conditions. to : 303°C at 20% of rated power (for a
“ The core control is fully automated. temperature at HZP of 299.1°C). In this range
“ The operator is in charge of optimizing the where the primary temperature is moving with
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 115

Fig. 33. Part load diagram at best estimate flowrate.

the reactor power, the capacity for load The capacity to predict and to ‘measure’ the
variations is somewhat reduced compared to the 3-dimensional power distribution relies, on the
power range between 100 and 50% of rated other hand, essentially on two types of incore
power for reducing the mechanical and instrumentations:
thermomechanical loads (Lisdat, Miossec, 1. The ‘movable’ instrumentation which is also
Sengler, POWERGEN 96: Flexibility of Power called the reference instrumentation, which is
generation with the EPR based on the French principally used for validating the core
and German experience). models used for the core design and for
calibrating the other sensors used for core
3.5. Core instrumentation/core sur6eillance/core surveillance and core protection purposes.
protection 2. The fixed incore instrumentation used for
delivering the necessary online information to
3.5.1. General the different core surveillance and core
As far as the protection of the core is protection systems which are in charge to
concerned, the safety approach relies partly on actuate the appropriate actions or
the capacity to predict and to ‘measure’ as well countermeasures when anomalies are detected
the nuclear power level (or level of neutron or when predefined limits are exceeded.
fluxes) as the three dimensional power In fact, the strict differentiation of functions
distribution. between excore and incore instrumentations does
The measurement of the nuclear power level not totally apply: excore instrumentation can be
(or neutron flux level) is essentially performed partially used for providing information about
by the so called excore instrumentation which is the ‘macroscopic’ power distribution (axial and
based on the temperature measurements in the azimuthal tilts) and the fixed incore
loops and wide range excore flux-measurements instrumentation can as well be used for meas-
as it is classically done on PWRs. uring the power level.
116 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

Fig. 34. Main features of the aeroball measurement system.

 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 117

3.5.3. The ‘incore’ instrumentation

The EPR incore instrumentation is ‘top-
mounted’ and consists of:
“ An aeroball measurement system as the
movable reference core instrumentation.
“ A certain number of fixed incore-detector
fingers containing axially distributed self
powered neutron detectors (SPND). These
fixed incore detectors are used for the
elaboration in 4-fold redundant protection
channels of the state parameters which are
necessary for core surveillance and core
protection purposes under power operation.
“ A certain number of core outlet
thermocouples (COTC) having the same
radial locations as the PDDs and used mainly
for measuring the margin to saturation in
post accident or degraded thermal hydraulic
conditions.

3.5.3.1. The Aeroball system. The reference system

for the power distribution assessment is an
Aeroball system. This system is simple and reliable.
The guide tubes for the balls have a small inner
Fig. 35. Possible arrangement of incore instrumentation
diameter (2.0 mm), their bend radii can be very
fingers.
small (only a few cm) and there are no major
constraints for locating the measurement room and
for routing the tubings (see main features in Fig.
3.5.2. Excore instrumentation 34). In addition, the time necessary for a flux
During load operation, the nuclear power measurement is very short: 3 min for activation, 5
level is principally measured by the means of a min for activity measurements. This system
4-fold redundant primary heat balance, relying therefore allows flux mapping measurements in
on temperature measurements in the cold and time intervals of 10–15 min.
hot legs of the primary loops. These Number of probes. The aeroball probes are
measurements are combined in such a way that carried and distributed over the core by the means
a power measurement remains possible also in of 12 incore lances. Each incore lance can on
the case of assymmetric or (N-1) loop operation. principle bear four aeroball fingers and one PDD
This primary heat balance is speeded up by finger. The mechanical solution relies on the proven
using in addition excore neutron-flux- technique applied for German PWRs which leads
measurements (Power range) which have very to a proposal for the radial distribution of some 40
short response times. probes as indicated in Fig. 35.
As the core must be supervised and protected This ‘density’ of radial measurements and type
also when being operated at very low power of distribution for the probes was shown being
levels or in subcritical conditions, the convenient in Germany for the validation of the
appropriate surveillance and protective functions nuclear core design.
are relying on redundant excore neutron flux Axial resol6ing power. After activation, the
channels covering : 9 – 10 decades of the total activity of the balls-columns is measured in a
neutron flux range below the nominal power. measuring table with the means of some 30
118 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119

‘surface-barrier’ semi conductors per column. information necessary for characterizing the
These detectors are equally distributed over the power distribution and the thermal hydraulic
active length and are integrating the activity conditions is distributed to all the surveillance and
over a length of :6 cm protective channels simultaneously. This solution
Operational power range of the aeroball system. is acceptable due to the intrinsic system
The electronic part of the measuring system redundancy, the overlapping of information, the
consists in pulse counters. This technique diversity of sensors and the independence of their
combined with the short period of the V52 calibration procedures. This will allow to detect
isotope (3.7 min) which is serving as indicator, degraded operating conditions and failures of
restricts the range of power at which accurate sensors and to differentiate these two types of
events.
3-D-flux mapping is possible. In practice,
The fixed incore instrumentation which is a part
acceptable 2-D-flux maps can be obtained at
of these systems consists of SPNDs and COTCs.
: 5% RP and the necessary accuracy for
The SPNDs have a fast response time. At every
3-D-flux maps is reached at some 30% RP.
radial location at least six SPNDs are placed in a
PDD-finger where they cover appropriately the
3.5.3.2. The fixed incore instrumentation. The active core height. Each of the yokes of the
specific core surveillance and core protection aeroball system can contain one PDD-finger,
systems are based on digital equipment which which is replaceable in case of detector defects. As
are aimed at elaborating the relevant limiting shown in Fig. 35, 12 of those fingers are used in
core parameters. These systems are relying on total for ‘covering’ the core.
more or less sophisticated models or data The COTCs are axially located in the top
processing (e.g. 3-D-core model for core nozzle of the instrumented fuel assemblies and are
surveillance and simplified algorithms for core on principle radially distributed over the core in
protection) for elaborating the safety relevant the same way than PDD fingers. This is an
state parameters of the core like peak power, already proven mechanical solution but the
DNBR, etc. pertinence of this radial distribution with respect
These systems are operating properly in case to the functions of these sensors has still to be
of relatively slow core related PCC2 events and confirmed. At every location, there is space for
were introduced in order not to penalize more three thermocouples. These sensors are mainly
than necessary the operation of the plant. As a used for post accident measurement purposes but
consequence, they are able to operate without they may be also used for getting additional radial
significant loss of accuracy, under conditions like information on the radial power distribution and
dropped control rods, RCCA-misalignments, loss on the local thermalhydraulic conditions.
of one RCP, and in case of single failures.
In addition, the fixed core instrumentation
provides the core surveillance and core
4. Conclusions/outlook
protective systems with sufficient redundant and
diverse information:
The major features of the EPR-core as resulting
“ on RCCA positions,
from a first loop of iterations were presented in
“ on temperatures, flowrates, power level,
this paper.
“ on axial power distributions, They are corresponding to the status reached
“ on radial power distributions, mid 1997 and are pointing out the large flexibility
for allowing these systems to cover PCC1 and with respect to the selection of the cycle lengths,
PCC2 events, even in case of partial unavailability of different types of fuel managements including
of sensors. significant MOX recycling etc. This core contains
For reaching these targets, the available even margins in excess which could be used under
 G. Sengler et al. / Nuclear Engineering and Design 187 (1999) 79–119 119

different boundary conditions to improve potential evolution is studied in a later design

the cost/benefit ratio by increasing the power phase called BDOP (Basic Design Optimization
level up to the limit imposed by major Phase) which is expected to be completed at the
components on the secondary side. This end of 1998.