VALIDATION OF BATAN'S STANDARD 3-D DIFFUSION CODE,
BAT AN-3DIFF, ON THE FIRST CORE OF RSG GAS
Liem Peng Hong"
ABSTRACT
VALIDATION OF BATAN'S STANDARD 3-D DIFFUSION CODE, BATAN3DIFF, ON THE FIRST CORE OF RSG GAS. Batan-3DIFF code validation efforts have
been conducted on the experiment data compiled during the commissioning phase of Reaktor
Serba GUlla G.A. Siwabessy (RSG GAS) first core. The validation work covers the effective
multiplication factor, excess reactivity and the control rod worth. One set of 4-group
macroscopic cross section library prepared by WIMS/D4 cell calculation code and another set
derived from lAFUEL library were used in the present analysis. In order to simulate the
existence of the control rod absorber material in the core, the extrapolation lengths for each
group are calculated based on the dirty blackness coefficients. The ratio of the calculated and
experiment (C/E) values for Batan-3DIFF code with WIMS/D4 and IAFUEL libraries for the
first criticality condition are 0.988 and 0.994, respectively, while for the excess reactivity of
the full core are 0.993 and 1.000, respectively. In general, the WIMS/D4 library gives slightly
lower effective multiplication factors than the IAFUEL library does. The C/E values for
Batan-3DIFF code with WIMS/D4 and lAFUEL libraries for the total control rod worth are
worse, that is, 0.893 and 0.881, respectively.
INTRODUCTION
Several Batan's standard neutron diffusion codes [1,2,3] as well as a
dedicated in-core fuel management code for research reactors [4] have been
developed, verified and validated to meet one objective of the research and
development activities conducted by the Center for Multipurpose Reactor,
National Atomic Energy Agency (Batan). Since the first version of the codes
were launched several important improvements, modifications as well as
addition of new options on the codes have been continuously conducted to
response to the new requirements, users' comments and feedback. Important
options developed and verified are the capability of the codes to calculate the
adjoint neutron flux (that is solving the adjoint neutron eigenvalue problems),
external neutron source (that is solving the fixed source problems), integral
kinetic parameters, reactivity changes based on the perturbation theory etc.
The options recently developed, which are closely related to the
present work, were the two options for treating the existence of strong
absorber materials in the reactor [5]. Particularly, from the point of view of
reactor safety these two options are of great importance. The first option for
treating strong absorber uses a pair of blackness coefficients which are then
converted to effective diffusion parameters (effective diffusion constant and
absorption cross section of the absorber materials), while the second one
Center for Multipurpose Reactor, National AtomicEnergy Agency
47
adopts the logarithmic derivative constant applied as an internal boundary
condition for the diffusion calculations.
In one hand, with all of these options, the Batan's standard codes have
to be systematically verified and validated on the available benchmark
problems and experimental data. On the other hand, plentiful sources of
experimental data are available from the long eleven years of RSG GAS
operation; especially for the ones obtained during the commissioning works
which have not been utilized for validation works. In the present work,
validation results of Batan-3DIFF code (for 3-D multigroup neutron diffusion
calculations) on the first core of RSG GAS experiment data will be reported.
The first core was chosen because of, first, the completeness of the
experiment data, and secondly, the fuel elements, absorber and beryllium
reflectors were still fresh so that uncertainties raised from the fuel burnup and
absorber depletion calculations, and from the calculation of lithium poisoning
in the beryllium reflectors can be eliminated.
The organization of the paper is as follows. In the next section, the
neutronic experiments of the RSG GAS first core are presented in detail.
The third section discusses the neutronic modeling and calculation procedure
for simulating the experiments. In the fourth section, the Batan-3DIFF
calculation results are discussed and compared with the experiment results.
The last section gives the concluding remarks of the present work.
FIRST CORE NEUTRONIC
EXPERIMENTS
In this section, some neutronic experiments closely related to the
present validation works are briefly reviewed. These include the criticality
experiment and control rod calibration.
The establishment of the first full core of RSG GAS (12 standard fuel
elements and 6 control fuel elements) consists of two steps. The first step was
to achieve a critical condition of the core with the smallest number of fuel
elements, i.e. achieving the first criticality of the first core. Following the first
criticality, in the second step, a certain number of fuel elements were then
added into the core to provide sufficient excess reactivity to the core so that
the reactor can be operated for a certain period without refueling, i.e. the
excess reactivity loading. The configuration for the first criticality needs only
fresh 9 standard and 6 control fuel elements (Fig. 1) while the full
configuration of the first core established in the second step consists of 12
fresh standard fuel elements and 6 control fuel elements (Fig. 2).
In the second step, at each fuel element loading step, the reactivity
gains were measured by calibrating the difference of the regulating rod
position with a reactivity meter (single rod vs. bank rods compensation
method). The accumulated excess reactivity measured by this method should
be considered as an uncorrected excess reactivity value. The excess reactivity
value was then corrected by a method described in Ref. 6 using
48
2-D multigroup neutron diffusion code, Batan-2DIFF [2] in XY 2-D reactor
geometry . The measured uncorrected excess reactivity value was
8.14 % 11k / k (10.63 $) while after correction on the measured excess
reactivity was done the corrected excess reactivity of the first core was found
to be 8.46 % 11k / k (11.054 $). This excess reactivity was converted to
effective multiplication factor of 1.09242.
Following the excess reactivity loading, control rod worth calibrations
were conducted for the six control rods with various methods. During the
calibrations one can find many combinations of control rods' position which
gave a critical core condition suitable for the present validation work. Since
the control rods were inserted in the core the validity of the code to treat the
strong absorber can be checked. Furthermore, during the control rod
calibration experiment the primary cooling pumps worked producing coolant
flow of 2500 m3lhour which increased slightly the core temperature
(Joule effect) isothermally. The reactivity effect of the Joule effect from the
primary cooling system was measured to correct the experiment data.
CALCULA TION PROCEDURE AND MODELING
The calculation procedure of the validation works is shown in Fig. 3,
which consists of two main parts, i.e. the 4 group macroscopic cross sections
library generation and the whole core neutron diffusion calculations using
Batan's standard neutron diffusion code, Batan-3DIFF in XYZ 3-D reactor
geometry.
The procedure and modeling used in the cross sections generation for
fuel element using WIMS/D4 code have been validated and extensively
discussed in Ref. [5], therefore, it will only be briefly reviewed here. For
standard fuel elements, the multiplate option is chosen where the unit cell
consists of eleven and a half identical fuel regions and one extra region with
each fuel region is divided further into fuel meat, aluminum cladding, water
moderator regions. The extra region is used to include regions that are not
covered by meat-cladding-moderator fuel plate configurations. Those regions
are side plates, small parts of aluminum cladding adjacent to side plates and
water gaps among fuel elements. Besides the cross section library obtained
from the WIMS/D4 code, the cross section library derived from IAFUEL
code [6] to date used by the Center for the RSG GAS in-core fuel
management will also be used for comparison.
In addition to the cross section generation, in order to treat the strong
absorber in the core, the extrapolation length of the RSG GAS control rod has
to be determined. The procedure to determine the extrapolation length of the
RSG GAS control rod follows the one already validated and discussed in
Ref. [5]. Batan's standard diffusion codes can treat strong absorbers as nondiffusing regions with an appropriate set of internal boundary conditions that
usually derived from the blackness coefficient a.
49
(1)
where
and d are neutron flux and extrapolation distance. In this case, the
neutron current and flux at both sides of the absorber surface are assumed to
be identical. The equation can be rewritten by using the neutron current
expressIOn,
<I>
(2)
where D and J denote the neutron diffusion coefficient and neutron current as
function of neutron energy. From the definition of a blackness coefficient,
the above boundary condition C=a, that is the surface current-to-flux ratio.
For example, if the dirty blackness coefficient, a(DB), is used to compute the
extrapolation distance then one can write,
o
d=
a
(DB)
(3)
o
11+3E.(L.<)]
= 0.7104),{1_
2E3 (La")
where Au-, La and "t represent the transport mean free path, absorption cross
section and absorber material tickness, respectively.
In the codes, this latter expression for the boundary condition is used.
The energy dependent d is supplied by code users to simulate the existence of
the strong absorber. Use of Eq.(3) in Batan-3DIFF code requires a fast and
efficient calculation way of the exponential integral functions, En (n=3,4).
The rational approximation of these integral functions as reported by Makino
[8] is used where the maximum relative error of the approximation are in the
order of 10-6.
The whole core diffusion calculations in 3-D reactor geometry with
Batan-3DIFF code can simulate precisely the experiment conditions,
especially, the position of the control rods within the core.
50
RESULTS AND DISCUSSION
In Table 2, the comparison between Batan-3DIFF calculation results
with experiment data for the first criticality and excess reactivity of
RSG GAS first core is shown. In the table, pure experiment data
(no correction) are for the critical condition under first criticality core
configuration (9 fuel elements - FEs, 6 control elements - CEs, with the
regulation rod at the position of 475 mm). For this critical condition Batan3DIFF code with WIMS/D4 and IAFUEL libraries gives ratio of calculated
to experiment (C/E) values of 0.988 and 0.994, respectively. The IAFUEL
library gives a closer calculation results than the one with WIMS/D4 library;
however, both libraries underestimate the critical condition. Under the first
criticality core configuration, the regulation rod is only inserted 125 mm from
the core top position; however, the reactivity on inserted part of the
regulation rod is too small to determine the accuracy of the extrapolation
length calculation.
As already discussed in the previous subsection, the other two
experiment data, that is, the excess reactivity under the full core configuration
(12 FEs, 6 CEs, and all six control rods fully withdrawn) and the total control
rod worth are not purely experiment data. The measured excess reactivity has
been corrected using Batan-2DIFF code and then converted from $ unit to
absolute value in order to obtain the effective multiplication factor, keff, and
consequently, the accuracy depended on the accuracy of the calculated /3eff
(==0.00765). No measured data are available for /3effso that further assessment
on the accuracy of the calculated /3effis not possible. For the excess reactivity,
the C/E values for Batan-3DIFF with WIMS/D4 and IAFUEL libraries are
0.993 and 1.000, respectively. These C/E values are slightly improved, and
again the WIMS/D4 library gives a slightly lower keff value than the IAFUEL
library does (around 1 $).
Furthermore, the experiment value of the total control rod worth is
obtained simply by an arithmetic summation of the individual control rod
worth, or in other words, it is assumed that no significant control rod
interaction occurs. For this experiment value, the C/E values of the Batan3DIFF with WIMS/D4 and IAFUEL libraries are 0.893 and 0.881,
respectively. Considerably large deviations between experiment and
calculated values can be observed. This will be discussed later.
Table 3 shows the comparison between Batan-3DIFF calculation
results with several critical conditions of the RSG GAS first core established
during control rod calibrations. These critical conditions were achieved when
the calibrated rod was fully inserted and the other rods were in a certain bank
position. The measured critical conditions shown in the table have been
corrected for the Joule effect (which turned out to be negligible) of the
primary cooling pumps.
51
This experiment is very important in that large portions of control rods
were inserted into the core so that the accuracy of the extrapolation length
calculation for simulating the existence of strong absorber in the core can be
evaluated accurately. For this experiment, the cm values for Batan-3DIFF
with WIMS/D4 and IAFUEL libraries are closer than those for the first
criticality condition. Since both conditions were critical so that no correction
(neither for excess reactivity calculation nor from the calculated /3elfl are
needed then the following· conclusion can be drawn. The results strongly
suggest that the calculated extrapolation
lengths produce slightly
underestimated reactivity values for the control rods under consideration. In
other words, if the extrapolation lenghts were correct then the cm values for
the control rod calibration would produce the same order of, but not better,
accuracy with the cm value for the first criticality experiment. This result is
consistent with the underestimated calculation results of the total control rod
worth shown in Table 2. However, the deviation from the experiment value
of total control rod worth is still considered large and may not just because of
the error from the extrapolation length estimation. Thus, the assumption and
consequently the procedure to obtain the experimental value of the total
control rod worth need to be reevaluated.
The next discussion will focus on the differences of the nuclear data
libraries used for the validation which were calculated by WIMS/D4 code
and derived from IAFUEL library. The neutron reaction rates as well as the
neutron balance over the reactor are shown in Tables 4 and 5, respectively,
for the first criticality condition. In the lowest parts of the tables, the relative
differences between the twos are given. From Table 4, it can be observed
that, in one hand, the total fission or production rates for WIMS/D4 library
are about 0.5 % greater than the ones for IAFUEL library, on the other hand,
the absorption rate are about 1.1 % greater than the one for IAFUEL library.
This is one reason for the slightly lower keffofWIMS/D4Iibrary.
From the neutron balance (Table 5) this can also be observed more
clearly. While the neutron production was 100 % (normalized) the neutron
absorption were 100.3 % and 99.8 % for WIMS/D4 and IAFUEL libraries,
respectively. Besides neutron absorption, neutron loss may arise from
escaping from the reactor system (total leakage) and from control rod
absorption. The control rod absorption for the two libraries was essentially
the same, but the total leakage for WIMS/D4 library was slightly larger than
the one for IAFUEL library. The latter trend made the values of keff for
WIMS/D4 library smaller. To elaborate more on these differences, the
magnitude of the fuel element macroscopic cross sections for the two
libraries have been investigated. It turns out that the eta values (1]) for
WIMS/D4 library in the thermal and epithermal energy groups are slightly
lower than, while the diffusion coefficients for the thermal energy group are
higher than, that for IAFUEL library. The former factor gives relatively lower
52
infinity multiplication factor while the latter one enhances the neutron
leakage. They combine to produce lower kejffor WIMSlD41ibrary.
Fig. 4 shows the fuel element channel factor (FECF) distribution or
commonly known as the radial power peaking factor distribution for the two
libraries. Essentially there is no significant difference between the FECF
distributions of the two libraries. Fig. 5 shows the maximum values of the
fuel element axial power distribution, commonly known as the axial power
peaking factor, for the two libraries. Again no significant difference between
the axial power peaking factors of the two libraries is observed. The last
figure, Fig. 6, shows the axial power profiles for the control element at core
grid C-8 (where the regulation rod is inserted 12.5 cm) and the fuel element
at core grid E-6 under first criticality of the RSG GAS first core calculated by
Batan-3DIFF code. No significant difference is observed between the
two libraries.
CONCLUDING REMARKS
Batan-3DIFF code validation efforts have been conducted on the
experiment data compiled during the commissioning phase of RSG GAS first
core. The validation work covers the effective multiplication factor, excess
reactivity and the control rod worth. Two sets of macroscopic cross section
library were used, i.e. one set was prepared by WIMSID4 cell calculation
code and another set was derived from IAFUEL library. The C/E values for
Batan-3DIFF code with WIMSID4 and IAFUEL libraries for the first
criticality condition are 0.988 and 0.994, respectively, while for the excess
reactivity of the full core are 0.993 and 1.000, respectively. In general, the
WIMSID4 library gives slightly lower effective multiplication factors than
the IAFUEL library does. The C/E values for Batan-3DIFF code with
WIMSID4 and IAFUEL libraries for the total control rod worth are worse,
that is, 0.893 and 0.881, respectively. This indicates that the assumption that
no significant control rods interference occurs, from which the experiment
value of the total control rod worth may be obtained by a simple arithmetic
summation of the individual control rod worth, have to be reevaluated.
ACKNOWLEDGMENTS
The author express their gratitude to the commissioning group of RSG
GAS for providing the criticality experiment data.
53
REFERENCES
1. P.H. LIEM, "Introduction to Diffusion Code Programming", IAEA
Regional Training Course on Calculation and Measurement of Neutron
Flux Spectrum for Research Reactors, Lecture Notes, Jakarta (1993)
2.
P.H. LIEM, "Development and Verification of Batan's Standard, TwoDimensional Multigroup Neutron Diffusion Code (Batan-2DIFF)", Atom
Indonesia 20 (1) (1994)
3.
P.H. LIEM, "Pengembangan Program Komputer Standar Batan Difusi
Neutron Banyak Kelompok 3-D (Batan-3DIFF)", Risalah Komputasi
dalam Sains dan Tekonologi Nuklir V, PPI-Batan, Jakarta (1995)
4.
P.H. LIEM, "Batan-FUEL: A General Fuel Management Code", Atom
Indonesia 22 (2) (1996)
5.
P.H. LIEM and T.M. SEMBIRING, "Validation of Batan's Standard
Neutron Diffusion Codes for Control Rod Worth Analysis", Atom
Indonesia 23 (2) (1997)
6.
M. WICKERT, "Concept and Methods of the Program MAIN:
Controlling the IAFUEL Program Cycle for Neutronic Calculations
Regarding Research Reactor", Interatom Bericht/Report Ident-No:
54.07100.4 (1986)
7.
T.M. SEMBIRING and P.H. LIEM, "Validation of Batan's Standard
Diffusion Codes on IAEA Benchmark Static Calculations", Atom
Indonesia 23 (2) (1997)
8. Y. MAKINO, "Some Rational Approximations for the Exponential
Integral Functions En (n==I,2,3,4)", Nukleonik 9 (1967).
54
Table 1. Reactor main design data of RSG GAS.
General
Reactor Type
Fuel Element Type
Cooling System
Pool Type
LEU Oxide MTR
Forced Convection
Down Flow
Moderator/Coolant
Reflector
H20
Nominal Power (MWt)
Be
30
& H20
FueUControl Elements
Fuel/Control Element Dimension (mm)
Fuel Plate Thickness (mm)
Coolant Channel Width (mm)
No. of Plate per Fuel Element
No. of Plate per Control Element
Fuel Plate Clad Material
77.1x81x600
1.3
2.55
21
15
Fuel Plate Clad Thickness (mm)
Fuel Meat Dimension (mm)
Fuel Meat Material
0.38
0.54x62.75x600
U30gAl
19.75
2.96
250
178.6
Ag-In-Cd
3.38
SS-321
0.85
U-235 Enrichment (w/o)
Uranium Density in Meat (g/cm3)
U-235 Loading per Fuel Element (g)
U-235 Loading per Control Element (g)
Absorber Meat Material
Absorber Thickness (mm)
Absorber Clad Material
Absorber Clad Thickness
AIMg2
55
ta
VI
0'\
a) p = 0.00765, excess reactivity was already corrected (see text)
b) n.c. = not conducted
c) Shim rod bank compensation method, measured by reactivity meter,
summation of single control rod worth
Note:
0.932235
0.994
1.084655
Batan-3DIFF
0.993849
1.091962
0.881
15.69
1.000
0.988
diffusion
calculations
15.90
0.893
1.0
0.993
C/E
WIMS/D4
Library
IAFUEL
Library
Experiment
1.09242a)
n.c.
b)0.988156
i1P(%)
17.80
c)0.925077
Core
configuration
Keff
-
Table 2. Comparison of Batan-3DIFF calculation results with experiment data for
first criticality and excess reactivity of RSG GAS first core (300 K).
-.J
\J1
1.004959
1.003670
1.004
1.004609
1.002923
1.005
1.00008
0.995535
0.997942
0.996605
1.002605
1.003
1.003139
Batan-3DIFF
diffusion
data calculations
0.995881
0.996062
0.996
0.995
0.997
0.998
IAFUELC/E
C/E
Library
Experiment
1.00008
WIMS/D4
Library
Keff
Calibrated
roda)/ 0.997672
grid
position
Ketf
Ken
a) p = 0.00765
Note:
JDA06 / C-8
Table 3. Comparison of Batan-3DIFF calculation results with experiment data obtained during
control rod calibration experiments ofRSG GAS first core (300 K).
Table 4. Neutron reaction rates over the reactor calculated by Batan-3DIFF
code for the first criticality of the RSG GAS first core.
Group
1
2
3
4
TOTAL
Production
6.07£+17
2.04£+17
1.53£+14
O.OOE+OO 8.12£+17
Fission
5.73£+15
3.39£+15
4.12£+16
7.61£+17
8.12£+17
Absorption
_
Top Leakage
6.19£+08
Bottom Leakage
1.27£+09
Left Leakage
8.80£+09
Right Leakage
5.15£+08
Forward Leakage
1.03£+14
Backward Leakage
8.74£+13
Total Leakage
1.90£+14
CRAbsorption
1.02£+13
3.08£+15
5.50£+08
1.15£+09
8.47£+09
4.59£+08
9.46£+13
8.38£+13
1.78£+14
7.65£+13
5.37£+16
5.97£+08
1.06£+09
7.90£+09
4.98£+08
7.89£+13
7.31£+13
1.52£+14
1.59£+15
7.64£+17
2.58£+09
8.20£+09
7.01£+ 10
2.01E+09
5.78£+14
5.26£+14
1.10£+15
3.16£+15
8.14£+17
4.35£+09
1.17£+ 10
9.53£+10
3.48£+09
8.54£+14
7.70£+14
1.62£+15
4.83£+15
6.04£+17
5.46£+15
2.03£+17
3.05£+15
1.52£+14
4.30£+16
0.00£+00
7.56£+17
8.08£+17
8.08£+17
__
5.90£+08
1.22£+09
2.95£+15
5.24£+08
1.11£+09
5.33£+16
5.68£+08
1.02£+09
7.56£+17
2.47£+09
7.87£+09
8.06£+17
4.15£+09
1.12£+ 10
8.50£+09
4.86£+08
9.72£+13
8.29£+13
1.80£+14
1.00£+13
8.18£+09
4.32£+08
8.97£+13
7.96E+13
1.69£+14
7.65£+13
7.63£+09
4.69£+08
7.50£+13
6.97£+13
1.45£+14
1.59£+15
6.83£+ 10
1.90£+09
5.58£+14
5.08£+14
1.07£+15
3.16£+15
9.26£+ 1O.
3.28£+09
8.19£+14
7.40£+14
1.56£+15
4.84£+15
0.498
0.000
0.495
0.805
5.095
0.680
1.078
3.518
0.494
1.075
4.109
WIMSID4 library
IAFUEL Library
Production
Fission
Absorption
Top Leakage
Bottom Leakage
Left Leakage
Right Leakage
Forward Leakage
Backward Leakage
Total Leakage
CRAbsorption
(WIMS-IAFUEL)/IAFUEL*lOO
%
Production
0.495
0.497
Fission
Absorption
Total Leakage
4.935
0.740
5.643
11.269
4.502
5.376
CRAbsorption
2.237
58
_
_
0.025
__
-
6.622
6.602
0.31
4.00
0.019
0.022
0.196
0.009
0.018
0.021
0.010
94.115
0.392
2.953.63
0.136
24.76
5.081
5.324
0.378
93.797
0.019
40.389
100.000
100.002
1 74.834
0.706
0.418
25.147
0.00
4.47
0.197
93.624
93.571
0.580.58
0.000
0.000.00
0.132
100.348
0.00
909.772
.200
.596
.193balances
.599
0.676
0.023
0.001
0.022
4.41
0.00
0.25
4.93
8.33 the reactor calculated by Batan-3DIFF code
3
0.380
0.365
0.18
Table 5.TOTAL
Neutron
over
10.73_
for the first criticality ofthe RSG GAS first core.
Production
Group
59
F
C
BE
PNRA
BF3
: Fuel Element
: Control Element
: Be Reflector Element
: Pneu. Rabbit System
: BF3 Neutron Detector
BSBS+
DE
HYRA
*
: Be Refl. El. without Stopper
: Be Refl. El. with Stopper
: Dummy Element
: Hydraulic Rabbit System
: Neutron Source (Cf-252)
Figure 1. First criticality core configuration of the RSG GAS first core.
60
F
C
BE
: Fuel Element
: Control Element
: Be Reflector Element
PNRA : Pneu. Rabbit System
BFJ
: Neutron Detector
BSBS+
DE
HYRA
*
: Be Refl. EI. without Stopper
: Be Refl. EI. with Stopper
: Dummy Element
: Hydraulic Rabbit System
: Neutron Source (Cf-252)
Figure 2. Full core configuration of the RSG GAS first core.
61
Absorber
geometry &
composition
FE/CE
geometry &
composition
First core
config.
First core
criticality
experiments
Experimental
keff
Figure 3.
Batan-3DIFF
RSG GAS.
62
code validation
procedure on the first core of
0.996
1.304
1.303
1.000
1.139
8 DI
1.006
D 0.930
1.316
1.152
0.998
0.996
1.341
1.078
0.991
0.993
1.018
1.305
0.483
1.315
1.006
1.114
1.112
1.002
1.351
1.327
1.308
1.322
1.304
1.3460.807
1.312
1.307
581.316
71.011
61.214
0.996
0.768
1.001
1.309
1.313
1.299
1.313
0.909
1.002
0.503
1.021
1.148
0.475
0.803
0.994
40.777
1.306
1.302
tI.1f410;k
1.310
0.996
1.309
1.317
1.311
1.336
9
0.513
1.138
1.001
1.206
0.911
I:m~1$,1~
0.936
1.075
Fuel
element
channels factors (FECF) under first criticality
FECFs
calculated
with
WIMS/D4
andand
IAFUEL
libraries
while
the
12.5 em
from
theregulation
corethetop).
third
calculated
figures
bythe
denote
Batan-3DIFF
ratiorod
(the
of which
first
WIMS
second
IAFUEL;
figures
represents
shaded
grid
marks
isover
inserted
at thethe
depth
of
8
Figure 5. Maximum values of the fuel element axial power distribution under
first criticality calculated by Batan-3DIFF (the first and second
figures represents FECFs calculated with WIMS/D4 and IAFUEL
l,ibraries while the third figures denote the ratio of WIMS over
IAFUEL; the shaded grid marks the regulation rod which is inserted
at the depth of 12.5 em from the core top).
63
1.1
1.1
='c
1.1
.e
1.1
::J
~
CI>
IE
e
a.
1.1
J
0.1
-+-1
~
~
-4
(u/
_I -4
-6-1 ~
(10
0.1
-tf-C ~
IjII
! I~I)
fa c I)
! /4
(Ld
IjII
(10
(0 1 I)
I)
I
I
8./
II
~
11
\8
II
Axial Direction (cm)
11
II
II
I
---.J
Figure 6. The axial power profiles for the control element at core grid C-8
(where the regulation rod is inserted 12.5 cm) and the fuel element
at core grid E-6 under first criticality of the RSG GAS first core
calculated by Batan-3DIFF code.
64