ICNC 2023 - The 12th International Conference on Nuclear Criticality Safety
October 1st – 6th, 2023 in Sendai, JapanThe First Core Criticality Analysis of the RSG GAS Multipurpose Research
Reactor using the Newly Released JENDL-5 Nuclear Data Library
Peng Hong Liem (1, 2) *and Donny Hartanto (3)(1)

Scientific Computational Division, Nippon Advanced Information Service (NAIS Co., Inc.)
416 Muramatsu, Tokaimura, Ibaraki, 319-1112, Japan(2)
Cooperative Major in Nuclear Energy, Graduate School of Engineering, Tokyo City University(TCU), 1-28-1 Tamazutsumi, Setagaya, Tokyo, Japan(3)
Oak Ridge National Laboratory (ORNL),One Bethel Valley Road, Oak Ridge, TN 37830, United States of America* liemph@nais.ne.jp

ABSTRACTIn the present work, the criticality experiments of the clean first core of the G. A.
Siwabessy Multipurpose Reactor (RSG GAS) were analyzed by using the newly released
Japanese Nuclear Data Library version 5 (JENDL-5) to contribute to the validation effortof the JENDL-5, especially for the application on the beryllium reflected, light-watermoderated, low-enriched uranium (LEU, 19.75 % enrichment) fueled material testing
reactors (MTRs). Compared to the other types of reactors, the number of testing/research
reactors are very limited, and it is expected that the results would be beneficial not only
to the research reactor communities but also to the criticality safety communities. In thepast, we have already conducted similar criticality analyses using the then-available
nuclear data libraries such as JENDL-3.2, 3.3, 4.0, ENDF/B-VII.1, VIII.0, and JEFF. In
the present work, a continuous energy Monte Carlo MVP3 code is used in conjunction
with the newly released JENDL-5 nuclear data library. The JENDL-5 keff C/E values arearound 1.005 which shows the high fidelity of the library. Sensitivity analysis results
reveal relatively large reactivity changes (-50 to -60 pcm) when U-235, U-238, or water
JENDL-5 library is changed by the one of JENDL-4.0. Furthermore, the H-1 and O-16
contributions to the water reactivity change are found to be -162 pcm and +96 pcm, which
are considerably large. The JENDL-5 average values of effective delayed neutron
fraction (βeff), generation time (Ʌ), and Rossi-Alpha parameter (α) are found to be 738
pcm, 73 μsec, and 101 sec-1, respectively. These average values are very close to the onesof JENDL-4.0, 738 pcm, 74 μsec, and 100 sec-1, respectively.
KEYWORDSJENDL-5, Criticality benchmark, RSG GAS first core criticality, LEU oxide fuel, Be reflector

1. INTRODUCTION
Almost twelve years after the Japanese Evaluated Nuclear Data version 4 (JENDL-4.0) was released(2012) [1], the new version, i.e., the JENDL-5, was released in December 2021 [2]. In our previous
work [3, 4, 5, 6], the criticality experiments of the clean first core of the G. A. Siwabessy Multipurpose
Reactor (RSG GAS) [7] were used to validate the JENDL and other worldwide used nuclear datalibraries (ENDF and JEFF). In the present work, the same criticality experiments were analyzed by using
the new JENDL-5 to contribute to the validation effort of the JENDL-5, especially for the applicationon the beryllium reflected, light-water moderated, low-enriched uranium (LEU, 19.75 % enrichment)
fueled material testing reactors (MTRs). The remainder of this paper is organized as follows. Section 2

presents a brief description of the RSG GAS reactor. Section 3 explained the methodology used in thecriticality calculations. In Section 4, the results of the criticality analysis, control rod worth, core excess
reactivity, and kinetic parameters are discussed. The conclusions and the future work are given in
Section 5.

2. BRIEF DESCRIPTION OF RSG GASThe RSG GAS is a light-water moderated and cooled, pool-type MTR that uses beryllium (Be) as its
neutron reflector. It is designed with a maximum thermal power of 30 MW. The first criticality of thisreactor was achieved in July 1987. The first core used an MTR plate-type oxide fuel (U3O8-Al) with lowenrichment, 19.75 wt % U-235 (at present the core has been converted to silicide fuel). Fig. 1 shows thefirst criticality and full core configurations of the first core of RSG GAS. The main reactor data for the
first core is shown in Table I. The standard fuel element (FE) and the control fuel element (CE) are
shown in Fig. 2. Other main specifications of the reactors, as well as the detail of the criticality
experiments, can be found elsewhere [5, 7].

Figure 1. First criticality (left) and full core (right) configurations of the RSG GAS first core (plain
numbers in the core show loading step of fuel, Be reflector and dummy elements after the firstcriticality; empty lattices contain water)

3. METHODOLOGY
3.1. Selection of Criticality Benchmark Cases
There were two sequences of criticality experiments conducted for the first core of RSG GAS, i.e. (1)the criticality approach experiment for achieving the first criticality, and (2) the criticality experimentfor core excess reactivity loading followed by control rod worth calibrations. At the end of the first
sequence of the criticality experiment, the first criticality was achieved at the 9-th loading of standard
fuel element [5], i.e. the regulating rod (RR) was pulled out slowly (reactor periods were greater than2

50 s) while the other 5 shim rods were fully withdrawn. The insertion position of the RR at the firstcriticality was 475 mm. This configuration, shown on the left side of Fig. 1, was chosen as one of themost appropriate for this study.
In the second sequence of the criticality experiment, the loading of additional fuel elements and reflector
elements was conducted to achieve a full core configuration with sufficient excess reactivity for one
core cycle. The reactivity gains were measured at each loading step by calibrating the difference of theregulating rod position with a reactivity meter (compensation method with the other 5 shim rods in bankconfiguration). The measurement of the accumulated excess reactivity by this method should beconsidered as an uncorrected value. The excess reactivity value was then corrected by a method
described elsewhere [5]. After that, the control rod calibrations were conducted following the excess
reactivity loading for the six control rods with various methods. Many combinations of control rod
positions can be found during the calibrations which gave a critical core condition. In this work, thecombinations of control rod positions that occurred during control rod calibration with the bankcompensation method were selected for the benchmark calculations.

Table I. Main reactor data of RSG GAS first core
Reactor Type
Pool Type
Fuel Element TypeLEU Oxide MTR
Moderator/Coolant
H2O
Reflector
Be & H2O
Nominal Power (MWth)
10.7
No. of Fuel Elements
12
No. of Control Elements
6
No. of Fork Type Absorber (pairs)
6
Fuel/Control Element Dimension (mm)77.1 x 81 x 600
Fuel Plate Thickness (mm)1.3
Coolant Channel Width (mm)2.55
No. of Plate per Fuel Element21
No. of Plate per Control Element
15
Fuel Plate Clad Material
AlMg2
Fuel Plate Clad Thickness (mm)0.38
Fuel Meat Dimension (mm)0.54 x 62.75 x 600
Fuel Meat Material
U3O8-AlU-235 Enrichment (wt. %)19.75
Uranium Density in Meat (g/cm3)
2.96
U-235 Loading per Fuel Element (g)250
U-235 Loading per Control Element (g)178.57
Absorber Meat Material
Ag-In-Cd
Absorber Thickness (mm)3.38
Absorber Clad Material
SUS-321
Absorber Clad Thickness
0.85

3.2. Criticality Calculation Model and Conditions
The active part (7.71 x 8.1 x 60 cm3) of both FE and CE were modeled as their exact geometry anddimensions while the top and end-fitting of the elements were modeled in an approximate manner since
their geometry is very complicated, that is, the structure materials were homogenized with water byvolume weighting. An exact modeling approach was also taken for the active parts of the Be reflector
elements, Be block elements, and irradiation positions. Considering their complicated geometry, thecore grid and bottom support were also treated approximately as for the top or end-fitting of fuel3

elements. This approximation did not deteriorate the accuracy of the MVP3 [8] calculation results since
it was applied in the non-active parts of the core. The movable control rods (absorber blades) weremodeled as their exact geometry and dimensions. Consequently, a 60 cm water layer above the core had
to be included in the calculation to provide enough space for the absorber blades when a control rod was
fully withdrawn. Approximately 30 cm water layers were included below the core bottom support, andaround the beryllium block and element reflectors. Vacuum boundary conditions were imposed on theouter boundary of the reactor system.

62.75

1.62
0.38

2.2

1.30

2.3

62.75

1.1

2.3
4.5

2.55

0.38

1.30

81.0
80.50

81.0

80.50

2.55

3.38

5.08

65.0

4.5

70.75

70.75

76.1

76.1
77.1

77.1

Figure 2. Standard fuel element (left) and control fuel element (right) of the RSG GAS (unit mm)

All MVP3 calculations in the present benchmark were conducted with the JENDL-5 library for a room
temperature of 300 K. The measured critical effective multiplication factors were corrected when thecore isothermal temperature was not identical to 300 K. The total number of batches (generations) was
10,000 where each batch consists of 10,000 histories, i.e. the total number of effective histories was 100
million. Additional initial 100 batches were skipped to guarantee the fundamental mode had beenachieved before statistical evaluation of keff and other tallies were conducted. Under these calculation
conditions, the fractional standard deviation (FSD) for keff was less than 0.01 % for all cases. The main
benchmark results are the keff values which can be directly compared with the experimental/measured
ones. The inter-library comparison results also provide valuable information, differences on the criticalkeff, excess reactivity, control rod worth, and kinetics parameters amongst libraries will also bepresented and discussed. As for the calculations of the kinetics parameters, the MVP3 differential
operator sampling with perturbed source effect and default number of propagation batches (10) was
adopted as in our previous work [9].

4. ANALYSIS RESULTS
The analysis results are shown in Tables II and III for criticality-related values and kinetic parameters,
respectively. In the tables, the results for the JENDL-4.0 of our previous work are also shown for interlibrary comparison. As mentioned in Section 3, since the benchmark cases are divided into two groups
according to the two sequences of critical experiments, the tables are also divided accordingly.

4In general, for the first group of Table II, one can observe that although the keff C/E values of JENDL5 are very close to 1.0, the values are slightly larger than the ones of JENDL-4.0. The maximum
difference between the two libraries is found for the full core configuration with all control rods inserted,
i.e. around 262 pcm. In the same table, the experiment data (keff or the excess reactivity) for the fullcore is included but needs several remarks since it is not purely experiment data. The measurement detailof core excess reactivity and the correction method is given in [5]. It should be noted here that themeasured excess reactivity must be converted from $ unit to absolute value to get the (absolute unit)
keff, and therefore, the accuracy depended on the accuracy of the calculated effective delayed neutron
fraction, βeff (=765 pcm). No measured data were then available for βeff so further assessment of theaccuracy of the calculated βeff was not possible. This factor also contributed to the difference between
the experiment data and Monte Carlo results. The sub-criticality of the full core when all control rods
were inserted was not measured and only calculation results are available. Combined with the calculated
excess reactivity values, the calculated total worth of all control rods can be estimated. The experimental
data for the total control rod worth was obtained by a simple arithmetic summation of the single control
rod worth, while the single control rod worth was measured by a reactivity meter with the shim rod bankcompensation method. It is well known that the interference between control rods, to some extent, may
deteriorate the accuracy of the total control rod worth obtained by the summation of the single control
rod worth [10]. Combined with the uncertainty of the calculated βeff value, the MVP3 results for
JENDL-5 and JENDL-4.0 can be judged to be well agree with the experiment data.
The second group of Table II shows the comparison between MVP3 results with several criticalconditions of the RSG GAS first core that occurred during control rod calibrations. These criticalconditions were achieved when a calibrated rod was fully inserted and the other rods were in a certainbank position. The measured critical conditions shown in the table were already corrected for the Jouleeffect (which turned out to be negligible) of the primary cooling pumps. Unfortunately, themeasurements for the critical conditions were not repeated so no reliable measurement error prediction
can be deducted. Similar trends of keff C/E values can be observed for the two JENDL libraries.
A sensitivity study was conducted to investigate the reactivity change when a single nuclide or material
library of JENDL-5 is replaced by one of the JENDL-.40, and the result is summarized in Fig. 3. The
main nuclides composing the fuel region are U-235, U-238, O-16 and Al-27 (meat and clad) and H-1
and O-16 (water moderator/coolant), while the reflector region mainly consists of Be-9. In other reactor
structural regions, Al-27, water (H-1 and O-16) are dominant. It should be noted that one can notseparate the bounded hydrogen S(α,β) thermal library from the free gas (fast) library of MVP libraries,
therefore, the reactivity change from replacing the library is the summation of both S(α,β) thermal andfast library of hydrogen. In the study, the reactivity change of water is further broken down into H-1 andO-16 contributions. Relatively large reactivity changes are observed for U-235, U-238, and water,
namely around -50 to -60 pcm. Furthermore, the contribution of H-1 is -162 pcm while the one of O-16
is +96 pcm. Al-27 and Be-9 show negligible reactivity changes. Relatively strong absorber materials,
i.e. Ag, In, and Cd, also do not show large reactivity changes.
Table III shows the kinetic parameters calculated by using the JENDL-5 library, and the results are also
compared to the ones using JENDL-4.0 [9]. The JENDL-5 average values of βeff, generation time (Ʌ),
and Rossi-Alpha parameter (α) for the second group benchmark cases are found to be 738 pcm, 73 μsec,
and 101 sec-1, respectively. These average values are very close to the ones of JENDL-4.0, 738 pcm, 74
μsec, and 100 sec-1, respectively. It is assumed that there is no significant revision of delayed neutron
data of U-235 and U-238 from JENDL-4.0 to JENDL-5. The calculated βeff values are slightly smaller
than the βeff value (765 pcm) provided by the reactor vendor, which was used during the commissioningof the reactor. The non-critical core kinetic parameters (from the first group of Table III) are listed forreference only.

5

Figure 1. Sensitivity analysis results by changing a single JENDL-5 library with JENDL-4.0

5. CONCLUSIONS AND FUTURE WORK
Criticality analysis of the newly released JENDL-5 nuclear data library against the criticality
experiments of the first core of RSG GAS has been conducted using the continuous energy Monte Carlo
MVP3 code. In general, the JENDL-5 keff C/E values are around 1.005 which shows the high fidelityof the library. Sensitivity analysis results reveal relatively large reactivity changes (-50 to -60 pcm) whenU-235, U-238, or water JENDL-5 library is changed by the one of JENDL-4.0. Furthermore, the H-1
and O-16 contributions to the water reactivity change are found to be -162 pcm and +96 pcm, which areconsiderably large. The reactivity changes obtained in this sensitivity analysis show the same order andtrend as reported in Table 4 of Ref. [11] (The dk/k values of uranium-fueled light-water moderated
system when each nuclide is changed from JENDL-5). As for future work, we plan to conduct sensitivity
and uncertainty of keff of JENDL-5 similar to our previous work for JENDL-4.0 and ENDF/B-VII.1
[12].

6

Table II. Criticality analysis results for JENDL-5 (present work) and JENDL-4.0 [5]Experiment
data

Core configuration
keffFirst criticality (9 FEs, 6 CEs, RR=475 mm)

Δρ(J5-J4)

JENDL-4.0

JENDL-5

1.002431.00496

251

1.002

1.005

-

1.09854

1.10054

165

1.006

1.007

-

0.918800.92102

262

-

-

-

17.81

17.71

-97

1.000

0.995

-

JENDL-4.0

JENDL-5

1.003421.00536

192

1.003

1.005

-

1.00261.00470

208

1.003

1.005

-

1.00277

1.00516

237

1.003

1.005

-

1.002311.00455

222

1.002

1.004

-

1.003

1.00519

217

1.003

1.005

-

1.002811.00479

197

1.003

1.005

-

(pcm) d)

1.0
C/E
keff

Full core (9 FEs, 6 CEs, CRs all up)
C/E

1.09242 a)keff
Full core (9 FEs, 6 CEs, CRs all down)
C/En.c. b)Δρ(%)
Control rods worth
C/E
Calibrated rod / grid position(calibrated rod position / other rod bank position)

17.80 c)Experiment
data

JDA05 / C-5(600 mm / 288 mm)

keff

JDA06 / C-8(600 mm / 290 mm)

keff

JDA07 / D-4(600 mm / 282 mm)

keff

JDA01 / E-9(600 mm / 284 mm)

keff

JDA04 / F-5(600 mm / 290 mm)

keff

JDA03 / F-8(600 mm / 293 mm)

keff

C/E

1.00008 a)Δρ(J5-J4)(pcm) d)1.00008
C/E
1.00008
C/E
1.00008
C/E
1.00008
C/E
1.00008C/Ea) β=0.00765, excess reactivity was already corrected (see text)b) n.c. = not conductedc) Shim rod bank compensation method, measured by a reactivity meter, summationof single control rod worthd) FSDs for all keff are less than 0.01% (J4: JENDL-4.0, J5: JENDL-5)

7

Table III. Kinetic parameters analysis results for JENDL-5 (present work) and JENDL-4.0 [9]First criticality andexcess reactivity loadingFirst criticality(9 FEs, 6 CEs, RR=475 mm)

Full core(12 FEs, 6 CEs, CRs all up)

Full core(12 FEs, 6 CEs, CRs all down)

Calibrated rod / grid position(calibrated rod position / otherrod bank position)
JDA06 / C-8(600 mm / 290 mm)

JDA01 / E-9(600 mm / 284 mm)

JDA03 / F-8(600 mm / 293 mm)

JDA05 / C-5(600 mm / 288 mm)

JDA04 / F-5(600 mm / 290 mm)

JDA07 / D-4(600 mm / 282 mm)

Kinetics
Parameters

JENDL-4.0

JENDL-5

JENDL-4.0/
JENDL-5

β eff (pcm)

751.4 ± 3.4

749.5 ± 3.4

1.003

Ʌ (msec)

82.79 ± 0.08

81.82 ± 0.07

1.012

α (1/sec)

90.8 ± 0.4

91.6 ± 0.40.991

β eff (pcm)

727.7 ± 3.2

728.9 ± 3.1

0.998

Ʌ (msec)

76.51 ± 0.07

75.56 ± 0.06

1.013

α (1/sec)

95.1 ± 0.4

96.5 ± 0.40.986

β eff (pcm)

742.0 ± 3.5

747.2 ± 3.5

0.993

Ʌ (msec)

69.17 ± 0.06

68.20 ± 0.06

1.014

α (1/sec)

107.3 ± 0.5

109.6 ± 0.5

0.979

Kinetics
Parameters

JENDL-4.0

JENDL-5

JENDL-4.0/
JENDL-5

β eff (pcm)

738.6 ± 3.3

737.8 ± 3.3

1.001

Ʌ (msec)

73.60 ± 0.06

72.72 ± 0.06

1.012

α (1/sec)

100.4 ± 0.5

101.5 ± 0.50.989

β eff (pcm)

734.6 ± 3.3

743.3 ± 3.3

0.988

Ʌ (msec)

75.32 ± 0.06

71.90 ± 0.06

1.048

α (1/sec)

100.9 ± 0.5

103.4 ± 0.50.976

β eff (pcm)

742.3 ± 3.4

733.0 ± 3.4

1.013

Ʌ (msec)

73.42 ± 0.06

72.31 ± 0.06

1.015

α (1/sec)

101.1 ± 0.4

101.4 ± 0.50.998

β eff (pcm)

741.9 ± 3.4

735.2 ± 3.3

1.009

Ʌ (msec)

73.67 ± 0.07

72.73 ± 0.06

1.013

α (1/sec)

100.7 ± 0.5

101.1 ± 0.50.996

β eff (pcm)

737.1 ± 3.4

739.2 ± 3.3

0.997

Ʌ (msec)

75.32 ± 0.07

74.30 ± 0.07

1.014

α (1/sec)

97.9 ± 0.5

99.5 ± 0.50.984

β eff (pcm)

735.0 ± 3.4

741.8 ± 3.3

0.991

Ʌ (msec)

73.93 ± 0.07

73.09 ± 0.06

1.012

α (1/sec)

99.4 ± 0.5

101.5 ± 0.5

0.980

8

ACKNOWLEDGMENTS
We express our gratitude to all members of the commissioning group of RSG GAS for their efforts ofcompiling the criticality experiment data. We are also grateful to Mrs. Suzuko Ikegami for the
preparation of tables and graphs.

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