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1
Simulation of the OECD NEA Kalinin-3 benchmark (phase 2) by
the
ATHLET-KIKO3D code system
A. Keresztúri, I. Trosztel, Zs. Elter, Gy. Hegyi Hungarian
Academy of Sciences
Centre for Energy Research
1. INTRODUCTION
As a logical continuation of the former benchmark activity of
AER devoted to the VVER reactor physics [1-3], now there is a
rather large interest of the Kalinin-3 benchmark recently completed
and published by the Nuclear Energy Agency (NEA) of the
Organization for Economic Cooperation and Development (OECD) [4].
The goals of these exercises were the evaluation of the prediction
capability of advanced code systems by means of a code to code
comparison. Meanwhile the last task, analyzing a real plant
transient, is good to verify the prediction capability of best
estimate coupled codes. These code systems are characterized by
direct coupling between the one dimensional thermal hydraulics
plant models with three dimensional neuton kinetics models. Among
others ATHLET-QUABOX/CUBBOX [5] ATHLET-DYN3D [6] TRAC-PARCS [7]
CATHARE-CRONOS-FLICA4 [8] and the Hungarian ATHLET-KIKO3D [9]
belong to this group. The new, KALININ-3 transient „Switching-off
of one Main Circulation Pump (MCP)” starts from the nominal power
at a MOC burnup condition. During the transient the coolant flow is
rearranged in the reactor pressure vessel and a spatially dependent
coolant temperature change can be observed, meanwhile the limiting
controller starts to decrease the reactor power by insertion of
control group number 10 and 9. At the beginning of the transient an
asymmetric core condition can be observed which eventually leads to
symmetric core behavior at lower power level. Broad ranges of the
parameter changes can be observed. The experiment is very well
documented, their uncertainties are known and the measurements were
performed with a quite high frequency [4]. These facts allow
applying the studied transient not only for validation purposes but
also for uncertainty analysis as a part of the NEA/OECD LWR
Uncertainty Analysis. The benchmark consists of four separate
exercises as follows:
1. Exercise: The system thermal hydraulic characteristics and
modeling of the plant regulation were investigated without using
the complicated 3D reactor physics model in order to clarify the
capability of thermal hydraulics. In our former paper, the ATHLET
model using point reactor kinetics was used for the solution of
this exercise. The results were compared to documented experimental
data. Rather good agreement was observed [10].
2. Exercise: The spatial reactor kinetics simulation with given
time dependent core boundary conditions, such as insertion of
control groups inlet temperature and flow as a time dependent
function are used. The appropriate data was distributed by GRS.
3. Exercise: full simulation of the reactor with coupled best
estimate code using 3D reactor physics model (analysis of the
transient in its entirety).
4. Exercise: The different types of uncertainties of the input
parameters will be investigated.
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2
In the paper the solution of the second exercise is outlined
using the coupled ATHLET/ KIKO3D code complex and the prescribed
time dependent- boundary conditions (coolant inlet of assemblies
and control rod positions as a function of time). New ATHLET input
model and some code development concerning the KIKO3D code will be
presented. The results are compared to some other solutions and the
documented experimental data. 2. THE CORE LOADING AND THE TRANSIENT
The transient was measured during the first cycle of the Unit 3,
NPP Kalinin. The original fuel loading of the core consists of five
types of assemblies:
1. 48 FA with U235 enrichment of 1.3 % (all pins has the same
enrichment); 2. 42 FA with U235 enrichment of 2.2 % (all pins has
the same enrichment); 3. 37 FA with average U235-enrichment of 2.98
% (303 pins with 3 % and
9 gadolinium fuel pins with 2.4 %-enrichment); 4. 24 radial
profiled FA with average U235-enrichment of 3.9 % (243 pins with 4
%, 60
pins with 3.6 % and 9 gadolinium pins with 3.3 %-enrichment); 5.
12 radial profiled FA with average U235-enrichment of 3.9 % (240
fuel pins with 4 %,
66 fuel rods with 3.6 % and 6 gadolinium fuel rods with 3.3
%-enrichment). A sixty degree symmetrical loading was developed
from these assemblies, but at the 96th effective day a defected
assembly with coordinates 07-32 had to be replaced by a “fresh”
standard FA with U235-enrichment of 1.6 %. The spacers and the
leading tubes of these FAs were made of stainless steel. The fuel
loading map in the reactor core of Unit 3 NPP Kalinin after the
replacement of the defected assembly by a standard one is
demonstrated in Figure 1. The data of the power load (necessary to
calculate the fuel burnup) for Unit 3 NPP Kalinin from the
beginning of the first fuel cycle up to the day when the experiment
with the switching off of one MCP took place, was provided
separately. Main scenario sequences recovered from the measured
data histories can be systematized so [4]:
Manually switching-off MCP number 1 at t=0s. The signal ‘one
pump out of operation’ is generated at 1.41 s. The following
sequence of actuations for reactor limiting controller and
automatic
reactor power controller is recorded: - At t=1.41 s the reactor
limiting controller starts to decrease the reactor power.
Control rod bank (CRB) #10 starts to move downwards. When the
CRB #10 reaches 50 % insertion depth (at about 60 s) the CRB #9
also starts to enter the active core according to the control rod
movement algorithm.
- Protection system level #1 of the automatic reactor power
controller switches from option ‘T’ (keeping the secondary loops’
parameter constant) to option ‘H’ (keeping neutron power
constant)
- Control rod controller decouples from automatic reactor power
controller. At t=71 s the reactor power load-off procedure is
finished and power reaches a level of
67.2 % Pnom. At this moment the position of the CRB #10 is at
43.4 % and remains there till the end of the transient. CRB #9 is
inserted into the core and reaches at 71 s the position of 93.1 %
and stays there till 180 s. After that, it returns back to 100 %.
The automatic reactor power controller is again switched on to the
control rod controller with option ‘H’ and it starts to keep the
power level in the range of 66.2 - 67.3 % Pnom.
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In connection of Control Bank, the further information is
important: At the beginning of the transient the control group 9 is
fully withdrawn from the core,
its position is 362 cm (104% by the measurements), as the upper
control rods’ end- switches are located at 14 cm above the active
core. It means that the real insertion of the CR 9 started by a
delay of 7.29 s [11]
According to the measurement system established at the NPP, the
positions of Control Bank are given with respect to the position of
the lower end switches. They are located 17.25 cm higher than the
bottom of the reactor core. The length of the reactor core is 355
cm and the distance between the lower and the upper end switches of
352 cm. It corresponds to the 100% insertion of control rod.
The Exercise 2 focuses on the 3D core behavior and -for the
easier comparison- the boundary conditions for the core are
provided by the benchmark team. 3. THE ATHLET /KIKO3D COUPLED CODE
SYSTEM The coupling of ATHLET with the spatial kinetics code KIKO3D
extends its application to a wide range of VVER plant transients.
This coupled code complex has been successfully used for more than
12 years to examine security issues of VVER, to perform
best-estimate analysis covering core thermal hydraulics, reactor
physics and plant dynamics [9], [10]. In this exercise the inlet
condition of core bottom and some outlet condition were prescribed
on the basis of a detailed calculation made by ATHLET-BIPR-VVER
best estimate code. That is why the detailed core model used in the
first exercise was converted to a simplified one. It consists of
163 separate thermo-hydraulic channels with the upper plenum (see
in Figure 3). Each channel represents the flow condition of a given
assembly wherever located in the core. No coolant mixing between
these channels are considered as the mass flow rate rather high
during the transient progression. The assemblies are divided in ten
axial nodes of the same height. The fuel rod gap conductance as
well as the material properties was taken from the final
specification. In the benchmark specification the characteristics
of the fuel assemblies for the KALININ-3 core are extensively
described. Concerning the original sixty degree turning symmetry of
the core, it consists of 29 fuel assembly types, each one with
unique axial material composition. Axially the fuel parts of the
core divided 10 layers. There are reflector nodes in radial
direction in all elevation and below and above all the assembly.
There are 283 unrodded compositions. Take into account the two
level structure of the control rod: one contains
Dysprosium-Titanium and the other mainly B4C absorber, there are
further 2 times 110 rodded compositions. For the asymmetric sector
a detached data set was developed, too. A complete set of diffusion
coefficients macroscopic cross section for scattering, absorption
and fission as a function of moderator temperature and density and
fuel temperature were calculated by HELIOS 1.9 code for each
composition. The kappas, the group inverse neutron velocities,
decay constants and betas are also provided for each composition.
Originally the KIKO3D reactor kinetic code calculation is based on
response matrix library which had to be modified for the above
depicted table given for the benchmark. Based on the actual reactor
conditions the appropriate total cross sections are obtained from
the look-up tables using a linear interpolation scheme. To predict
the actual cross section for
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the bottom reflector a fuel temperature equal to the inlet
coolant temperature and a coolant density equal to the inlet
coolant density are used. The same procedure used for the reflector
on the top and for radial reflector nodes.
4. THE STADY-STATE CALCULATIONS Our modified calculation tool
was verified by comparison our results to other one. The benchmark
team recommended the hot zero power state for that purpose. Steady
state calculations were performed using the following
parameters:
0.1% of nominal power Moderator and fuel temperature: 552.15 K
Moderator density: 767.1 kg/m3 Boron concentration Cb=660 ppm or
3.6 g/kgH20 Control rod group #10 is 82.95 % (302 cm according to
the measurement) inserted
from below the PSU/REL library was used
Calculations were done for the symmetrical case, where the newly
loaded assembly was not taken into account, and the asymmetrical
case, too. In the second column of Table 1 it is sign as “sym” or
“asym”. The results presented here are preliminary result from the
working materials of the 4th Workshop of OECD Kalinin-3 Benchmark,
held in Karlsruhe, Germany in May [16-18].
Table 1 Comparison of the eigenvalue of different calculations
Code References
Core Keff [-] SCRM [%]
CR No 9 [pcm]
CR No 10 [pcm]
CORSAR [16] Sym 1.010978 9.66 - PARCS [18] Sym 1.011190 11.52 -
1013.0 DYN3D [18] Sym 1.011311 11.25 - DYN3D [17] Sym 1.010832
11.25 985 746.9 ATHLET-KIKO3D
Sym 1.01408 11.10 984 809
ATHLET-KIKO3D
Asym 1.01356 11.07 1002 789.
From Table 1 one can see that rather similar results were given
by the different codes in hot zero power cases. In that state the
symmetrical case slightly differs from the asymmetrical one. The
frozen xenon concentration corresponds to the full power case and
the boron concentration which is the critical value of the full
power case give explanation for the supercritical eigenvalue. It is
slightly smaller for the asymmetrical core, as the enrichment of
the fresh fuel is smaller and the burnup level is not too high of
the core. Comparison of the radial power distributions calculated
by different codes for the symmetrical case can be seen on Figure
4.The agreement is acceptable. Figure 5 shows, how much
perturbation caused in the power distribution the insertion of a
fresh low enriched fuel assembly into the core. The control group
worth as a function of insertion can be seen on figure 6 (group 10)
and 7 (group 9). Good agreement can be seen between the PARCS and
KIKO3D calculation. It is
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worth to see, that the control rod behavior practically does not
depend from the core symmetry. One dimensional axial power
distribution for different state (H10= 302 / 352 cm) can be seen on
Figure 9 and 10. The agreement is similar the above mentioned
cases. The initial state of the transient was investigated, too. In
the Hot Power State only the asymmetric core loading was
calculated. The efficiency of control rod group number 9 and 10 can
be seen on Figure 11 and 12. The calculations made by KIKO3D and
PARS show good agreement. The integral worth of rod groups is
summarized in Figure 13. 5. SIMULATION OF THE TRANSIENT The initial
steady state calculation can be characterized by the following
parameters:
The thermal power: 2962 Mw Average fuel temperature: 609.1 C
Average moderator temperature: 305.4 C Average moderator density:
714.2 kg/m3 Boron concentration Cb=660 ppm Height of control group
number 10 is 302 cm, according to the measurement the PSU/REL
library was used Keff for steady state case: 0.99736
The radial power distribution at the beginning of the transient
and the one dimension axial power distribution in some different
time points can be seen on Figure 15 and 14. It has to be mention
that all participants using the PSU/REL library slightly
underestimate the steady state eigenvalue. The time dependent
boundary conditions are shown on Figure 16 and 17. Instead of
presenting the all the time dependent flow data, nine of them
signed by gray hexagon on Figure 15 were chosen to represent the
transient. The measured and calculated time dependent power can be
seen on Figure 18. One can see that the final power state slightly
underestimated. In Figure 19 a measured and calculated pressure is
presented. The reason of the difference between the two curves is
the fact, that the position of the simulation was not perfect.
Figure 20 -22 show the average parameters of the core, during the
transient. These figures well characterize the slight perturbation
of the core and are good for further inter-comparisons. The power
and temperature outlet for the chosen assemblies can be seen on
Figure 23-24. One can see the effect of the temperature
perturbation one side and the power decrease due to control group
insertion on the other side. Some of the assemblies (number 26, 36,
37 and 52) are the vicinity of the MCP 1. Two assemblies number 52
112 belong to the control group 10. The smallest outlet temperature
belongs to the fresh assembly (number 97). Concerning the measured
and calculated outlet temperature a rather large discrepancy can be
seen on Figure 25. Its reason was discussed in [11-13] a very
detailed manner. That is why the Kq were evaluated from the
calculated temperature rises. The measured and evaluated Kq as a
function of time show rather good agreement (see Figure 26 and
27).
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6. CONCLUSIONS The second phase of the KALININ-3 benchmark was
analyzed with the ATHLET-KIKO3D code complex. Appropriate ATHLET
nodalization was developed for this task. In case of spatial
reactor kinetic code, the PSU cross section library was implemented
into the KIKO3D code. The code to code comparison revealed that the
results prepared by the ATHLET-KIKO3D results are in good agreement
with most of the trends predicted by similar codes. The reactivity
values of different control rod groups and power distribution at
HZP and HFP were used to prove that the given cross section library
works properly. Some comparison of the result obtained for the
KALININ 3 benchmark transient with the appropriate measurements
clearly shown the efficiency of the coupled code. REFERENCES [ 1]
http://aerbench.kfki.hu [ 2] S.Kliem, S.Danilin, A.Hämäläinen,
J.Hadek, A.Keresztúri, S.Siltanen; Qualification of
Coupled 3-D neutron-Kinetic/Thermal-Hydraulic code systems by
the calculation of MSLB Benchmark in NPP with VVER-440 reactor,
Nuclear Science and Engineering, 157, 280-298 (2007)
[ 3] A.Kotsarev, M.Lizorkin, R.Petrin, Definition of 7th Dynamic
AER Benchmark VVER -440 pressure vessel coolant mixing by
re-connection of an isolated loop, RRC, Kurchatov Institute,
Institute of Nuclear Reactors. 2010
[ 4] A. Tereshonok, et al..; DESCRIPTION OF A TRANSIENT CAUSED
BY THE SWITCHING-OFF OF ONE OF THE FOUR OPERATING MCP AT NOMINAL
REACTOR POWER AT NPP KALININ UNIT 3, Final Draft
NEA/NSC/DOC(2009)6.
[5] S. Langenbuch, K.-D. Schmiedt, K. Velkov; The coupled Code
System ATHLET-QUABOX/CUBBOX, Model Features, and Results for the
Core Transients of the OECD PWRMSLB Benchmark, Int. Conf. on
Mathematics and Computation, Reactor Physics and Environmental
Analysis in Nuclear Applications, Vol. 1. pp 351-358. Sept 27-30.
1999, Senda Editorial. Madrid
[ 6] U. Grundmann, et. al.; Coupling of the Thermohydraulic Code
ATHLET with the Neutron Kinetics Core Model DYN3D, Mathematics and
Computation, Reactor Physics and Environmental Analysis, Int. Conf.
Apr. 30 - May 4. 1995. American Nuclear Society, La Grange Park,
III., USA, Vol. 1.P 257.
[ 7] R.M. Miller et. all.; Completed Report for the Coupled
TRAC-M/PARCS Code PU/NE99-20.Purdue University 1999
[ 8] E. Royer, I. Toumi; CATHARE - CRONOS - FLICA Coupling with
ISAS: A Powerful tool for Nuclear Studies. CEA-Saclay. 6th Int.
Conf. of Nuclear Engeneering. ICONE-6472. May 10-14.1998.
[ 9] Hegyi, Gy., Kereszturi, A., Trosztel, I., Langenbuch, S.,
Horche, W., Velkov, K., Improvement of Plant Transient Analysis for
VVER by Coupling KIKO3D with ATHLET. Proceedings of ICONE-6, San
Diego CA, USA.
[10]. István Trosztel, György Hegyi, András Keresztúri, Sergey
Nikonov; SOLUTION OF THE OECD NEA KALININ-3 COOLANT TRANSIENT
BENCHMARK PHASE 1 PROBLEM BY USING THE ATHLET CODE, 18th Symposium
of AER, Varna, Bulgaria, Sept. 21-28, 2009.
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[11] S. P. Nikonov, K. Velkov, A. Pautz; COMPARISON OF IN-CORE
THERMOCOUPLE AND SPND MEASURED DATA WITH THE ATHLET-BIPR-VVER
PREDICTION 20th Symposium of AER, Espoo, Finland, Sept. 20-24,
2010.
[12] S. P. Nikonov, K. Velkov, A. Pautz-PREDICTION OF MEASURED
SPND READINGS WITH THE COUPLEDCODE SYSTEM ATHLET-BIPR-VVER
[13] S. Nikonov, M. Lizorkin, S. Langenbuch, K. Velkov-MODELING
OF THERMOCOUP-LES’ MEASUREMENTS AT CORE EXIT OF REACTOR VVER-1000
WITH THE COUPLED CODE ATHLET/BIBR-VVER, 17th Symposium of AER,
Yalta, Crimea, Ukraine, Sept. 24-29, 2007.
[14] S. Nikonov, K. Velkov, ACCOUNTING FOR THE INERTIA OF THE
THERMO- COUPLES' MEASUREMENTS BY MODELLING OF A NPP KALININ-3
TRANSIENT WITH THE COUPLED SYSTEM CODE ATHLET-BIPR-VVER, 18th
Symposium of AER, Eger, Hungary, Oct. 6-10, 2008.
[15] V. A. Tereshonok, S.P. Nikonov, M.P.Lizorkin, K.Velkov, A.
Pautz, K.Ivanov, INTER-NATIONAL BENCHMARK FOR COUPLED CODES AND
UNCERTAINTY ANALYSIS IN MODELLING: SWITCHING-OFF OF ONE OF THE FOUR
OPERATING MAIN CIRCULATION PUMPS AT NOMINAL POWER AT NPP KALININ
UNIT 3, 18th Symposium of AER on VVER Reactor Physics and Reactor
Safety, Hungary, Eger, Oct. 6-10,2008
[16] Yann Perin, OECD KALININ-3 BENCHMARK: NEUTRONIC
CALCULATIONS OF EXERCISES 3A AND 3B WITH PARCS [17] Jan Hádek, R.
Meca, Progress in Solution of Kalinin-3 Benchmark Exercise 1 and 2
[18] Emiliya Georgieva, K. Ivanov, Status of PSU/REL Benchmark
Cross Section Library
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Figure 1 Final Core loading with the assembly (signed by letter
N) which unmake the 60
degree symmetry of the loading
Figure 2 Arrangement of the control rod groups within the
core
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Figure 3 ATHLET nodalisation for exercise 2
Figure 4 Radial power distribution at HZP state for the
symmetrical core calculated by ATHLET-KIKO3D (bellow) with
comparison to PARCS results (above)
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Figure 5 Assembly wise normalised power distribution at HZP
calculated byATHLET-KIKO3D. The fresh assembly and its 60 degree
symmetrical counterparts are signed by grey
HZP Control Rods Worth (CR09)
0
200
400
600
800
1000
1200
0 60 120 180 240 300 360Inserted length [cm]
Rod
wor
th [
pcm
]
Rodworth(pcm)-HZP-ARodworth(pcm)-HZP-SRodworth(pcm)-HZP-PARCS
Figure 6 Control rod worth calculated by ATHLET-KIKO3D and PARCS
for group 9 at
HZP state
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Figure 7 Control rod worth calculated by ATHLET-KIKO3D and PARCS
for group 10
at HZP state
Figure 8 Integral control rod worth calculated by ATHLET-KIKO3D
for all groups at HZP case
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Figure 9 Axial power distribution calculated by ATHLET-KIKO3D,
PARCS and some
coupled codes for HZP case H10=302 cm
Figure 10 Axial power distribution calculated by ATHLET-KIKO3D,
ATHLET-DYN3D for
HZP case H10=352 cm
Axial power distribution (HZP case ARO)
0.0
0.2
0.4
0.6
0.8
1.0
1.2
1.4
1.6
1 2 3 4 5 6 7 8 9 10
Axal layer
Rel
ativ
e Po
wer
KIKO3D DYN3D
Axial power distribution (HZP case H10=302 cm)
0.0
0.2
0.4
0.6
0.8
1.0
1.2
1.4
1.6
1 2 3 4 5 6 7 8 9 10
Axal layer
Rel
ativ
e Po
wer
KIKO3D DYN3D PARCS NRI
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Figure 11 -Control rod worth calculated by ATHLET-KIKO3D and
PARCS for group 10 at
HFP state
Figure 12 Control rod worth calculated by ATHLET-KIKO3D and
PARCS for group 10
at HFP state
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Figure 13 Integral worth of Control rod groups calculated by
ATHLET-KIKO3D at HFP
state
Figure 14 One dimension axial power distribution during the
transient
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Fig. 15 Assembly wise normalised power distribution at HFP
calculated byATHLET-KIKO3D. The fresh assembly and those which were
chosen to present the transient signed by
grey color
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Figure 16 Position of the control rod group groups during the
transient
Figure 17 Inlet temperatures of the chosen assemblies as a
function of time
280
282
284
286
288
290
292
294
0 50 100 150 200 250 300
Tem
pera
ture
(C)
Time (s)
Core inlet temperature
calc. of FA 26.calc. of FA 37.calc. of FA 38.calc. of FA
52.calc. of FA 67.calc. of FA 82.calc. of FA 97.
calc. of FA 112.calc. of FA 126.
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Figure 18 Comparison of the measured and calculated core
power
Figure 19 Comparison of the measured and calculated pressure
above the core
15
15.2
15.4
15.6
15.8
16
0 50 100 150 200 250 300
Pres
sure
(MPa
)
Time (s)
Pressure at the upper core
calc. upper core30YCR10FP901XQ01
1800
2000
2200
2400
2600
2800
3000
0 50 100 150 200 250 300
Pow
er (M
W)
Time (s)
Core power
calc.30YCR00FX001XQ01
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Figure 20 The core average density during the transient
Figure 21 The core average fuel temperature during the
transient
400
450
500
550
600
650
700
0 50 100 150 200 250 300
Tem
pera
ture
(C)
Time (s)
Average fuel temperature in the core
700
710
720
730
740
750
0 50 100 150 200 250 300
Den
sity
(kg/
m3)
Time (s)
Core average density
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Figure 22 The core average coolant temperature during the
transient
Figure 23 The calculated relative power of the chosen during the
transient
0.4
0.5
0.6
0.7
0.8
0.9
1
0 50 100 150 200 250 300
Rel
ativ
e po
wer
(-)
Time (s)
Assemblies relative power
calc. of FA 26.calc. of FA 37.calc. of FA 38.calc. of FA
52.calc. of FA 67.calc. of FA 82.calc. of FA 97.
calc. of FA 112.calc. of FA 126.
280
285
290
295
300
305
310
315
320
0 50 100 150 200 250 300
Tem
pera
ture
(C)
Time (s)
Average coolant temperature in the core
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20
Figure 24 The calculated outlet temperature of the chosen during
the transient
Figure 25 Comparison of the measured and calculated temperature
outlet for a given assembly
305
310
315
320
325
330
0 50 100 150 200 250 300
Tem
pera
ture
(C)
Time (s)
FA 11-18 outlet temperature
calc. of FA 37.30YQR00CT034XQ02
305
310
315
320
325
330
0 50 100 150 200 250 300
Tem
pera
ture
(C)
Time (s)
Calculated FA outlet temperature
calc. of FA 26.calc. of FA 37.calc. of FA 38.calc. of FA
52.calc. of FA 67.calc. of FA 82.calc. of FA 97.
calc. of FA 109.calc. of FA 112.calc. of FA 126.
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Figure 26 Comparison of the measured and calculated Kq for one
of the chosen assembly
during the transient
Figure 27 Comparison of the measured and calculated Kq for one
of the chosen assembly
during the transient
0.9
0.95
1
1.05
1.1
1.15
1.2
0 50 100 150 200 250 300
kQ (-
)
Time (s)
kQ in FA 10-23
calc. of FA 52.30YQR02FX052XQ01
0.9
0.95
1
1.05
1.1
1.15
1.2
0 50 100 150 200 250 300
kQ (-
)
Time (s)
kQ in FA 11-20
calc. of FA 38.30YQR02FX038XQ01