Nuclear Engineering and Design 418 (2024) 112899

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Nuclear Engineering and Design
journal homepage: www.elsevier.com/locate/nucengdes

JENDL-5.0 nuclear data sensitivity, uncertainty, and similarity analyses on
the criticality of RSG GAS multipurpose research reactor
Peng Hong Liem a, b, *, Donny Hartanto caCooperative Major in Nuclear Energy, Graduate School of Engineering, Tokyo City University (TCU), 1-28-1 Tamazutsumi, Setagaya, Tokyo, Japan
Scientific Computational Division, Nippon Advanced Information Service (NAIS Co., Inc.), 416 Muramatsu, Tokaimura, Ibaraki, Japanc
Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37830, USAb

A R T I C L E I N F O

A B S T R A C T

Keywords:
JENDL-5.0
S/U & similarity analysis
RSG GAS multipurpose reactor
Covariance matrices
AMPX-6The criticality, sensitivity, uncertainty and similarity analysis of the clean, first core of the G. A. Siwabessy
Multipurpose Reactor (RSG GAS), using the newly released Japanese Nuclear Data Library version 5 (JENDL-5)
was conducted to contribute the validation effort of the JENDL-5, especially for the application on beryllium
reflected, light-water moderated, low-enriched uranium (19.75 wt%) fueled material testing reactors. The
JENDL-5 sensitivity coefficients and covariance matrices were prepared by MCNP6 and AMPX-6, respectively.
The keff uncertainty was then evaluated with TSUNAMI-IP module of SCALE-6. The same module was used forsimilarity analysis by comparing the sensitivity data of 3,690 criticality experiments. The criticality (keff)analysis results ([C/E-1]) showed maximum overestimation of 801 pcm, 477 pcm and 547 pcm, for JENDL-5,
JENDL-4.0 and ENDF/B-VIII.0, respectively. The keff uncertainty originated from the nuclear data was esti­
mated to be 620 pcm, 644 pcm and 637 pcm for JENDL-5 only, JENDL-5 & ENDF/B-VIII.0, and ENDF/B-VIII.0
covariance data, respectively, which are comparable with the keff ([C/E-1]) values. About 125 experiment cases
were found to have correlation factors of > 0.8, however, no experiment showed a strong correlation factor(>0.9), therefore, in the future criticality experiments similar to the RSG GAS should be conducted.

1. Introduction
Almost twelve years after the Japanese Evaluated Nuclear Data
version 4 (JENDL-4.0) was released (2012) (Shibata et al., 2011), thenew version, i.e., the JENDL-5, was released in December 2021 (Iwa­
moto et al., 2023)). The new release was mainly aimed to contribute to
the existing important issues on nuclear energy development in Japan(Iwamoto et al., 2020). The issues include the treatment of nuclear waste
and the enhancement of nuclear safety. Related to the present work, it isworthily noted that the new JENDL-5 provides a neutron-induced re­action data sub-library from H (Z = 1) to Fm (Z = 100) for 795 nuclides
and a thermal-scattering law (TSL) sub-library for 37 materials(covering 62 elements). To appreciate the significance of the newrelease, for comparison, the previous version, JENDL-4.0, provides aneutron-induced reaction data sub-library for only 406 nuclides and a
TSL sub-library only for 15 materials. For nuclear data uncertainty
evaluation, reliable covariance data are essential. The newly released
JENDL-5 provides more covariance data than the previous version(JENDL-4.0); however, as will be reported later, the provided covariancedata of some important nuclides are still missing for the present
application.
In our previous works, the criticality experiments of the clean, firstcore of the G. A. Siwabessy Multipurpose Reactor (RSG-GAS) were used
to validate the JENDL and other worldwide used nuclear data libraries(ENDF and JEFF). The RSG GAS is a light water-moderated and cooledmaterial testing reactor that uses beryllium as its neutron reflector. It is
designed with a maximum thermal power of 30 MW. The first criticalityof this reactor was achieved in July 1987, and it reached 30 MW ofpower after going through 6 transition cores. Initially, this reactor used amaterial testing reactor (MTR type) oxide fuel (U3O8-Al) with lowenrichment, 19.75 w/o U-235. To improve the performance of thereactor, the RSG-GAS core was then converted to silicide fuel (U3Si2-Al)
first with the same uranium density, 2.96 g/cm3, and in the future withhigher uranium densities. The new equilibrium silicide core was pro­
posed and designed by Liem et al. (1998). The transition cores fromoxide to silicide have been also investigated, and the transition strategies
were established by Liem and Sembiring (2010). Based on the proposed
strategies, the conversion work was started by operating a series of* Corresponding author at: Scientific Computational Division, Nippon Advanced Information Service (NAIS Co., Inc.), 416 Muramatsu, Tokaimura, Ibaraki, Japan.
E-mail address: liemph@nais.ne.jp (P.H. Liem).
https://doi.org/10.1016/j.nucengdes.2024.112899
Received 13 September 2023; Received in revised form 19 November 2023; Accepted 1 January 20240029-5493/© 2024 Elsevier B.V. All rights reserved.

P.H. Liem and D. Hartanto

Nuclear Engineering and Design 418 (2024) 112899

release; therefore, we utilized all the available JENDL-5.0 covariancedata for the present U/A. Covariance data that were not provided by the
JENDL-5.0 release were prepared from the ENDF/B-VIII.0 nuclear data
(Brown et al., 2018). The justification is as follows. First, we consideredthat presently the JENDL-5.0 and the ENDF/B-VIII.0 are in a similarstate-of-the-art, and the covariance data are expected to be consistent
with each other. Second, before the present work, we conducted vali­
dation of the ENDF/B-VIII.0 nuclear data against the same criticality
experiments (Hartanto et al., 2019; Hartanto and Liem, 2020). The
validation results (keff) showed that the ENDF/B-VIII.0 had the sameorder of accuracy compared to the then-available JENDL-4.0 and ENDF/
B-VII.1. To improve the accuracy of the evaluated uncertainty, thesensitivity profiles and the covariance data were prepared into 252
energy-group and 56 energy-group, respectively. We also generate a
special JENDL-5 covariance matrix with additional ENDF/B-VIII.0
covariance data for JENDL-5 nuclides which have no covariance data.
The similarity analysis is never conducted for the RSG GAS although
up to the present thousands of high-fidelity critical experiments havebeen compiled into databases available for the analysis, such as ICSBEP
Handbook (OECD/NEA, 2016) and IRPhEP Handbook (OECD/NEA,
2017).
Through the present work, we expect that one can (1) estimate better
the JENDL-5.0 dominant nuclides and their reactions which have sig­
nificant sensitivity against the keff, (2) estimate the uncertainty of
JENDL-5.0 against keff and its dependency on the used covariance data,(3) some helpful information as the feedback to the JENDL evaluatorsand experts for further improvement in the future, (4) whether criti­
cality experiments similar to the RSG GAS are adequately available orsimilar criticality experiments are needed in the future. Of course, one
should be reminded that although the above-mentioned expectation ismainly valid for this particular type of reactor, some educated extrap­
olation of the results could be useful for reactors with similar neutron
spectra and/or reactors with similar fuel compositions.
The manuscript is organized as follows. First, the RSG GAS first core
criticality experiments are briefly discussed, followed by the method­
ology used in the present S/U and similarity analysis. After the criti­
cality, S/U, and similarity analysis results are presented and discussed,
the concluding remarks are given.

Fig. 1. First criticality core configuration of the RSG GAS first core (modified
from Reference (Liem and Sembiring, 2012)).

mixed oxide-silicide cores (Sembiring et al., 2001). A full silicide corehas been achieved at Core No. 45 after going through 10 cycles oftransition (mixed) oxide-silicide core operation as planned.
The accuracy of the JENDL-3.3 and JENDL-4.0 were validated
against the clean, first core of RSG GAS by using the continuous energyMonte Carlo MVP-II code (Nagaya et al., 2005), and the results werereported by Liem (1995), Liem (1998) and Liem and Sembiring (2012),
respectively. In these particular works, we only validated the nuclear
data against the criticality (multiplication factor, keff) experiments, andno sensitivity analysis (S/A) nor uncertainty analysis (U/A) was con­ducted. The JENDL-4.0 showed an excellent agreement with the keffoverestimation of only ~ 0.2 % and was better than the older JENDL-3.3(~0.3 %). The S/A of JENDL-4.0 on the first core of RSG GAS criticality
was for the first time conducted by using the continuous energy Monte
Carlo MCNP-6.2 (Werner et al., 2017) and reported by Liem et al.
(2019a), Liem et al. (2019b). To cope with the smaller number of the
JENDL-4.0 covariance data, we conducted the U/A using the thenavailable ENDF/B-VII.1 covariance data (Chadwick et al., 2011). It
should be noted that covariance data were compiled into 44 energy
groups, and not all of them were of high quality. The S/A for both
JENDL-4.0 and ENDF/B-VII.1 showed consistent results in terms ofdominant nuclides and their sensitivity values. The large negative sen­sitivities were found in the (n, gamma) reaction of H-1, U-235, Al-27, U238, and Be-9. In contrast, large positive sensitivities were found in U235 (total nu and fission), H-1 (elastic), Be-9 (free gas, elastic), and H-1(lwtr.20 t, inelastic). The U/A result of ENDF/B-VII.1 showed that theuncertainty of the nuclear data against the keff was approximately 0.6 %which was comparable with the [C/E – 1.0] values of keff. With thesimilar state of the art of JENDL-4.0 and ENDF/B-VII.1, we expected thatthe uncertainty of JENDL-4.0 would be in the same order of magnitude.
As a continuation of our previous work, in the present work, we
conducted the same S/A and U/A for the newly released JENDL-5.0. As
mentioned before, compared to the previous JENDL-4.0, a significantly
increased number of high-quality covariance data are available for the

2. Criticality experiments
The RSG GAS is a light-water moderated and cooled, pool-type MTR
that uses beryllium (Be) as its neutron reflector. It is designed with amaximum thermal power of 30 MW. The first criticality of this reactor
was achieved in July 1987. The first core used an MTR plate-type oxide
fuel (U3O8-Al) with low enrichment, 19.75 wt% U-235 (at present thecore has been converted to silicide fuel). Figs. 1 and 2 show the firstcriticality and full core configurations of the first core of RSG GAS,
respectively. The main reactor data for the first core is shown in Table 1
and the criticality configurations are summarized in Table 2. The stan­
dard fuel element (FE) and the control fuel element (CE) are shown in
Figs. 3 and 4, respectively. Other main specifications of the reactors, aswell as the details of the criticality experiments, can be found elsewhere(Liem and Sembiring, 2012; BATAN, 1980).
There were two groups of criticality experiments conducted for the
first core of RSG GAS, i.e. (1) the criticality approach experiment forachieving the first criticality and the criticality experiment for coreexcess reactivity loading, and (2) control rod calibrations under criticalfull core conditions. The first criticality was achieved at the 9-th loadingof standard fuel element (Liem and Sembiring, 2012), i.e. the regulatingrod (RR) was pulled out slowly (reactor periods were greater than 50 s)
while the other 5 shim rods were fully withdrawn. The insertion positionof the RR at the first criticality was 475 mm. This configuration, shownin Fig. 1, was chosen as one of the most appropriate for this study.
Following the first criticality, the loading of additional fuel elements andreflector elements was conducted to achieve a full core configuration
2

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Nuclear Engineering and Design 418 (2024) 112899

Fig. 2. Full core configuration of the RSG GAS first core (modified from Reference (Liem and Sembiring, 2012)).

(Fig. 2) with sufficient excess reactivity for one core cycle. The reactivity
gains were measured at each loading step by calibrating the difference ofthe regulating rod position with a reactivity meter (compensation
method with the other 5 shim rods in bank configuration). The mea­
surement of the accumulated excess reactivity by this method should beconsidered as an uncorrected value. The excess reactivity value was thencorrected by a method described elsewhere (Liem and Sembiring, 2012).
In the second group experiments, the control rod calibrations wereconducted following the excess reactivity loading for the six control rods
with various methods. Many combinations of control rod positions canbe found during the calibrations which gave a critical core condition. In
this work, the combinations of control rod positions that occurred dur­
ing control rod calibration with the bank compensation method wereselected for the benchmark calculations. Case 2-1 of the second group isselected for the S/U analysis.

Hereafter, the explanation is focused on the JENDL-5 evaluated nuclear
data library.
The sensitivity data, i.e. the sensitivity coefficients (Sk,α ) in 252
energy-group are prepared using the MCNP-6.2 (Werner et al., 2017)
code with the KSEN option. The KSEN option is based on the linear
perturbation theory with an adjoint weighting function (Kiedrowski and
Brown, 2013):()
1〈φ+ , Σα − Sα − keff
Fα , φ〉
α ∂keff
Sk,α == −,1
keff ∂αFα φ〉〈φ+ , keff
where φ and φ+ are the forward and adjoint neutron flux; Σα is the α-type
macroscopic cross-section; Sα and Fα are the scattering and fission op­
erators, respectively; keff is the eigenvalue (effective neutron multipli­
cation factor). The 〈〉 operator is the integration of space and energy
domains. Considering the reaction types (α), in the present version of
MCNP-6.2, the KSEN option can provide reaction types as shown in
Table 3. MCNP-6.2 does not solve the adjoint flux directly, instead, it
uses the Iterated Fission Probability method to compute the adjointweighted integrals for the sensitivity coefficients (cf. Eq. (1)). Forinter-library comparison, the sensitivity profiles were prepared for
JENDL-5, JENDL-4.0, and ENDF/B-VIII.0.
The covariance data, i.e. the covariance matrices (C), in COVERX 56

3. Methodology
3.1. Sensitivity, uncertainty, and similarity analysis
The calculation flow of the S/U and similarity analysis is shown in
Fig. 5. The calculation flow is not specific to JENDL-5 but is also applied
to JENDL-4.0 and ENDF/B-VIII.0 evaluated nuclear data libraries.
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Nuclear Engineering and Design 418 (2024) 112899

number of JENDL-5 nuclides with covariance data is significantly lessthan the one of ENDF/B-VIII.0, in the present work we also compiled a
special covariance matrix of JENDL-5 with the addition of ENDF/BVIII.0 covariance data for JENDL-5 nuclides which have no covariancedata. Using this special JENDL-5 covariance matrix, the uncertainty of
JENDL-5 can also be compared to the one of ENDF/B-VIII.0.
Based on the sensitivity coefficients (Eq. (1)) and covariancematrices (Eq. (2)), the keff uncertainty matrix, Ukk , can be estimated as

Table 1Main design data of RSG GAS (Liem and Sembiring, 2012).
Reactor Type

Pool Type

Fuel Element Type
Moderator/Coolant
Reflector
Nominal Power (MWth)No. of Fuel Elements
No. of Control Elements
No. of Fork Type Absorber (pairs)
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
Fuel Plate Clad Thickness (mm)
Fuel Meat Dimension (mm)
Fuel Meat Material
U-235 Enrichment (wt. %)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

LEU Oxide MTR
H2O
Be & H2O
10.7
12
6
6
77.1 × 81 × 600
1.3
2.55
21
15
AlMg2
0.38
0.54 × 62.75 × 600
U3O8-Al
19.75
2.96
250
178.57
Ag-In-Cd
3.38
SUS-3210.85

Ukk = Sk Cαα STk
where T indicates a transpose. The diagonal elements of the Ukk consistsof relative variance values, σ 2 , and by taking the squared root of thesevalues, i.e., σ, the keff uncertainty for each nuclide-reaction can be ob­tained. For the calculations of Ukk the TSUNAMI-IP module (Rearden
and Jessee, 2009) of SCALE-6.3 (Wieselquist et al., 2021) is used. As
mentioned above, the uncertainty matrices were calculated for JENDL-5
only, JENDL-5 + ENDF/B-VIII.0, while the ENDF/B-VIII.0 covariancematrices were taken for the SCALE 6.3.
The similarity analysis uses the same TSUNAMI-IP module. The
number of criticality experiments taken into account for the present
similarity analysis is 3,690 cases, obtained from ICSBEP Handbook(OECD/NEA, 2016). The similarity evaluation is based on the sensitivity
coefficients, either alone or in combination with cross-section uncer­
tainty information. The similarity of the RSG GAS core with these crit­
icality experiments is expressed with an index, i.e. correlation factor (ck)which has a range from 0 to 1.0. ck close to 1.0 indicates a strong cor­
relation, that is, a strong similarity exists between the RSG GAS core andthe corresponding criticality experiment. The complete derivation of theck can be found in the SCALE/TSUNAMI-IP module manual (Rearden
and Jessee, 2009).

Table 2
Criticality and S/U analysis cases.
Case

First criticality and excess reactivity loading (FirstGroup)

ReactorCondition

1-1
1-2
1-3
Case

First 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/other rod bank position)(Second Group, 12 FEs, 6 CEs)
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)

CriticalSupercritical
Subcritical
ReactorCondition

2-1
2-2
2-3
2-4
2-5
2-6
○

3.2. Monte Carlo modeling
The active part (7.71 × 8.1 × 60 cm3) of both FE and CE weremodeled as their exact geometry and dimensions while the top and endfitting of the elements were modeled approximately since their geome­try is very complicated, that is, the structure materials were homoge­
nized with water by volume 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 ge­
ometry, the core grid and bottom support were also treatedapproximately as for the top or end-fitting of fuel elements. Thisapproximation did not deteriorate the accuracy of calculation results
since it was applied in the non-active parts of the core. The movable
control rods (absorber blades) were modeled according to their exact
geometry and dimensions. Consequently, a 60 cm water layer above thecore had to be included in the calculation to provide enough space forthe absorber blades when a control rod was fully withdrawn. Approxi­
mately 30 cm water layers were included below the core bottom sup­port, and around the beryllium block and element reflectors. Vacuum
boundary conditions were imposed on the outer boundary of the reactor
system.
All calculations in the present benchmark were conducted with the
JENDL-5 library for a room temperature of 300 K. The measured criticaleffective multiplication factors were corrected when the core 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. Initial 100
batches were skipped to guarantee the fundamental mode had beenachieved before statistical evaluation of keff and other tallies wereconducted. 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 comparedwith the experimental/measured ones. The inter-library comparison
results also provide valuable information, differences in the critical keff,

Critical ○CriticalCriticalCriticalCriticalCritical: Selected case for S/U analysis.

energy-group format are prepared by employing the AMPX-6 code
(Bostelmann et al., 2021; Wiarda et al., 2016), particularly the PUFF-IV
module (Wiarda and Dunn, 2008) is utilized. The 56 energy-groupstructure is identical to the one used in the SCALE-6.2 (Rearden andJessee, 2016) 56 energy-group covariance data. The nuclear data pa­
rameters (i.e., group-wise nuclide-reaction specific cross sections) arerepresented by the vector α : = (αm ), m = 1, 2, ⋯, M, where M is thenumber of nuclide-reaction pairs × the number of energy groups. The
corresponding covariance matrix containing the relative variances andcovariances in the nuclear data are[]
COV(αm , αp ), m = 1, 2, ⋯, M; p = 1, 2, ⋯, M,
Cαα :=
αm , αpwhere()
COV αm , αp = 〈δαm , δαp 〉.
The δαm and δαp represent the difference between the values andexpectation values of the nuclear data parameters.
For inter-library comparison, the covariance matrices were evalu­
ated for JENDL-5 and ENDF/B-VIII.0. Amongst the 795 nuclides of
JENDL-5, only 105 nuclides have covariance data (see Table 4) while the
ENDF/B-VIII.0 has 251 nuclides with covariance data. Hence, since the4

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Nuclear Engineering and Design 418 (2024) 112899

Fig. 3. Standard fuel element (FE) of RSG GAS reactor (unit mm).

excess reactivity, and control rod worth amongst libraries will also bepresented and discussed.

criticality (9 FEs, 6 CEs, RR = 475 mm)). Other cases, i.e. Second group,
showed relative differences of less than 50 pcm. The keff values of
JENDL-5 differed considerably from the ones of JENDL-4.0 and ENDF/BVIII.0, where the maximum relative differences were 321 pcm and 252
pcm, respectively (both for Case 1-1, first criticality (9 FEs, 6 CEs, RR =475 mm)). Other cases, i.e. Second group, showed relative differences ofless than 220 pcm and 197 pcm, respectively. The keff differences be­
tween the JENDL-5 and previous JENDL-4.0 showed the same order andtrend as reported in Table 4 of Tada et al. (2023). The table shows thereactivity differences of uranium-fueled light-water-moderated systems
when each nuclide is changed from JENDL-5 to JENDL-4.0. The lightwater-moderated systems refer to LCT-02-01, LCT02-05, LCT26-01,
LCT48-03, LCT48-04, and LCT79-02, where LCT stands for LEUCOMP-THERM (OECD/NEA, 2016). They concluded that the JENDL-5(C/E) values were within the uncertainty of the (C/E) values for manycases. The difference between JENDL-4.0 and JENDL-5 is from − 407
pcm to − 9 pcm. They also conducted a ‘crude’ sensitivity analysis wherea single JENDL-5 nuclide was replaced by JENDL-4.0 and the keff dif­ference was evaluated. It was found that the dominant contributors were
H in H2O (TSL), O-16, U-235, and U-238.
In our recent work (Liem and Hartanto, 2023), the above-mentioned
crude sensitivity analysis was conducted for the RSG GAS first criticality
by using MVP3 code (Nagaya et al., 2016). We found that the main
contributors were the nuclides composing the fuel region, i.e. U-235, U238, and H2O. Relatively large reactivity changes are observed for U235, U-238, and water, namely around − 50 to − 60 pcm. It should be
noted that one can not separate the bounded hydrogen S(α,β) thermal
library from the free gas (fast) library of MVP libraries, therefore, the4. Analysis results and discussion
4.1. Criticality analysis
The MCNP criticality analysis results for JENDL-5, JENDL-4.0, and
ENDF/B-VIII.0 are shown in Table 5, divided into two groups in accor­
dance with Table 2. The experiment/measured keff is 1.0, hence thecalculated to experiment (C/E) values can be readily estimated from thetable (excluding non-critical cases, 1-2 and 1-3). In general, the JENDL4.0 results show the closest agreement with the experiment results,
followed by ENDF/B-VIII.0 and JENDL-5. Of the first group, the firstcriticality (Case 1-1) consisted of 9 FEs and 6 CEs, and the core had asmall excess reactivity equivalent to the insertion of the regulation rod(RR, position 475 mm). For this first criticality condition, the JENDL-5,
JENDL-4.0, and ENDF/B-VIII.0 results (keff) show an overestimation of
801 pcm, 477 pcm, and 547 pcm, respectively. For the second group, 6
cases of critical full core during control rod calibration (Case 2-1 to 2-6)
consisted of 12 FEs and 6 CEs, and all control rods were inserted
partially, except the measured control rod. For the second group, themaximum overestimation of keff values of JENDL-5, JENDL-4.0, and
ENDF/B-VIII.0 were 650 pcm, 431 pcm, and 452 pcm, respectively.
Concerning the inter-library comparison, the following results wereobserved. The keff values of JENDL-4.0 were considerably close to theones of the ENDF/B-VIII.0, where the maximum relative difference be­
tween the two libraries was only 70 pcm (First group, Case 1-1, first5P.H. Liem and D. Hartanto

Nuclear Engineering and Design 418 (2024) 112899

Fig. 4. Control fuel element (CE) of RSG GAS reactor (unit mm).

6

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Nuclear Engineering and Design 418 (2024) 112899

Fig. 5. Calculation flow of the sensitivity, uncertainty, and similarity analysis.

differences compared to JENDL-5 of more than 5 % were observed for
JENDL-4.0O-16 (n, alpha) reaction, and ENDF/B-VIII.0B-9 (n, gamma)
reaction. One can expect that these differences originated from thecorresponding cross section differences amongst the libraries.
From Table 7, one can observe that the sensitivities in the increasing
direction were dominated (absolute values > 1.0 %) by U-235 (total nu)& (fission), H-1 (free gas, elastic) & (TSL, inelastic), Be-9 (elastic) & (n,2n), O-16 (elastic), and Al-27 (elastic). As for the inter-library compar­
ison, sensitivity coefficients differences compared to JENDL-5 of morethan 5 % were observed for JENDL-4.0H-1 (free gas, elastic) & (TSL,
inelastic), Be-9 (TSL, inelastic) & (TSL, elastic), ENDF/B-VIII.0 Al-27(inelastic), U-238 (elastic) and Fe-56 (elastic).
The sensitivity analysis results of JENDL-5 showed similar trends as
in our previous work of JENDL-4.0 (Liem et al., 2019a). The energydependent sensitivity profiles for some JENDL-5 dominant nuclidereaction pairs are shown in Figs. 6 to 13. It can be observed that theuse of 252 energy groups can catch the dominant resonance structures of
the cross sections.
The U-235 sensitivity (Fig. 6) shows positive profiles for average νand fission reactions, and a negative profile for (n,gamma) or capture
reaction, in the thermal energy region (<1.0 eV). As for the U-238
sensitivity (Fig. 7), the negative profile of (n,gamma) reaction appears
not only in the thermal energy region but at many resonance energies in
the range of 1 eV to 1 keV. In the opposite direction of the resonances,the positive profile of elastic scattering can be observed. In the fast en­ergy region of more than 1 MeV, several positive profiles fission related
reactions, and inelastic reaction can be observed. The sensitivity of thefree gas H-1 and bounded H-1 (Fig. 8) shows positive profiles in the freegas H-1 elastic scattering reaction (greater than 10 eV). In the thermalenergy region, the bounded H-1 (TSL) inelastic scattering profile re­sembles the JENDL-4.0 where a sharp peak appears in the U-238 low
energy resonance (6.67 eV). The free gas and bounded Be-9 sensitivity(Fig. 9) shows the dominant positive profile in the elastic scattering
reaction, especially in the fast energy region. Fig. 10 indicates the mostly
positive profile of O-16 elastic scattering reaction especially in the fast
energy region. A relatively small negative profile of (n, alpha) reaction
appears in the energy region of more than several MeV. The Al-27
sensitivity (Fig. 11) shows a large negative profile of capture reactionin the thermal energy region (below 1 eV), while relatively smaller
positive profiles for elastic and inelastic scattering reactions appear in
the fast energy region. The Fe-56 sensitivity (Fig. 12) shows two domi­
nant profiles, namely, a positive elastic scattering profile and a negative

Table 3
Considered reaction types for the sensitivity coefficients (KSEN option).
No.

MCNP6.2 MT No.

Reaction Type

1
2
3
4
5
6
7
8
9
10

2
4
–6
16
102
103
104
107
–7
–1018

Elastic scattering
Inelastic scattering
Fission(n,2n) scattering
Capture (n,γ)(n,p) proton production(n,d) deuterium production(n,α) alpha particle production
Average number of neutrons produced per fission (total)
Fission spectrum

reactivity change from replacing the library is the summation of both S(α,β) thermal and fast library of hydrogen. Hence, in the study, thereactivity change of water is further broken down into H-1 and O-16
contributions. The contribution of H-1 is − 162 pcm while the one of O16 is + 96 pcm. Al-27 and Be-9 show negligible reactivity changes.
Relatively strong absorber materials, i.e. Ag, In, and Cd, also do notshow large reactivity changes. These results are also consistent with theones obtained by Tada et al. (2023) mentioned above.
4.2. Sensitivity analysis
The sensitivity analysis has been conducted with the KSEN option of
MCNP using JENDL-5, JENDL-4.0, and ENDF/B-VIII.0 libraries, for theselected case, i.e., Case 2-1 (Second group, 12 FEs, 6 CEs) where the fullcore was critical during C-8 regulation rod calibration. The sensitivity
coefficients for each nuclide and nuclear reaction were obtained andsorted according to their absolute values. Furthermore, the sensitivity
coefficients were grouped according to their signs into two groups as inour previous work (Liem et al. 2019a), one for sensitivity coefficientsthat tend to decrease keff (Table 6), and another one that tends to in­crease keff (Table 7). In the two tables, coefficients’ absolute values
smaller than 0.1 % were not shown, namely, only dominant coefficientswere shown and discussed.
From Table 6, one can observe that the keff sensitivities in thedecreasing direction were dominated by (n, gamma) reactions of H-1, U235, Al-27, U-238, Be-9, Mn-55, Ag-109, In-115, Ag-107, Cd-113, andFe-56. The (n, alpha) reactions of Be-9 and O-16, also contributed more
weakly. As for the inter-library comparison, sensitivity coefficients7

P.H. Liem and D. Hartanto

Nuclear Engineering and Design 418 (2024) 112899

Table 4JENDL-5 nuclides with covariance data processed by AMPX code.
No.

Nuclide ID

ZAIDFile Name

No.

Nuclide ID

ZAIDFile Name

1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24
25
26
27
28
29
30
31
32
33
34
35
36
37
38
39
40
41
42
43
44
45
46
47
48
49
50
51
52
53
54
55

125
525
528
628
725
728
825
1125
1725
2231
2425
2431
2434
2437
2525
2631
2637
2725
2825
2831
4025
8225
8231
8234
8237
8325
8925
8928
8931
9025
9028
9031
9034
9037
9040
9043
9046
9125
9128
9131
9134
9137
9213
9216
9219
9222
9225
9228
9231
9234
9237
9337
9340
9343
9346

1001
5010
5011
6013
7014
7015
8016
11,023
17,03522,04824,05024,052
24,05324,05425,05526,056
26,05827,05928,05828,06040,09082,20482,20682,207
82,20883,20989,22589,22689,22790,22790,22890,22990,23090,231
90,232
90,23390,234
91,22991,23091,231
91,232
91,23392,23092,231
92,232
92,23392,234
92,23592,23692,237
92,23893,234
93,23593,236
93,237n_001-H-001.dat
n_005-B-010.dat
n_005-B-011.dat
n_006-C-013.dat
n_007-N-014.dat
n_007-N-015.dat
n_008-O-016.dat
n_011-Na-023.dat
n_017-Cl-035.dat
n_022-Ti-048.dat
n_024-Cr-050.dat
n_024-Cr-052.dat
n_024-Cr-053.dat
n_024-Cr-054.dat
n_025-Mn-055.dat
n_026-Fe-056.dat
n_026-Fe-058.dat
n_027-Co-059.dat
n_028-Ni-058.dat
n_028-Ni-060.dat
n_040-Zr-090.dat
n_082-Pb-204.dat
n_082-Pb-206.dat
n_082-Pb-207.dat
n_082-Pb-208.dat
n_083-Bi-209.dat
n_089-Ac-225.dat
n_089-Ac-226.dat
n_089-Ac-227.dat
n_090-Th-227.dat
n_090-Th-228.dat
n_090-Th-229.dat
n_090-Th-230.dat
n_090-Th-231.dat
n_090-Th-232.dat
n_090-Th-233.dat
n_090-Th-234.dat
n_091-Pa-229.dat
n_091-Pa-230.dat
n_091-Pa-231.dat
n_091-Pa-232.dat
n_091-Pa-233.dat
n_092-U-230.dat
n_092-U-231.dat
n_092-U-232.dat
n_092-U-233.dat
n_092-U-234.dat
n_092-U-235.dat
n_092-U-236.dat
n_092-U-237.dat
n_092-U-238.dat
n_093-Np-234.dat
n_093-Np-235.dat
n_093-Np-236.dat
n_093-Np-237.dat

56
57
58
59
60
61
62
63
64
65
66
67
68
69
70
71
72
73
74
75
76
77
78
79
80
81
82
83
84
85
86
87
88
89
90
91
92
93
94
95
96
97
98
99
100
101
102
103
104
105

9349
9352
9428
9431
9434
9437
9440
9443
9446
9452
9458
9540
9543
9546
9547
9549
9552
9553
9625
9628
9631
9634
9637
9640
9643
9646
9649
9652
9655
9740
9743
9746
9749
9752
9755
9843
9849
9852
9855
9858
9861
9864
9867
9911
9912
9913
9914
9915
9916
9936

93,23893,23994,236
94,237
94,23894,23994,24094,241
94,242
94,244
94,246
95,24095,241
95,242
95,242
95,243
95,244
95,24496,24096,241
96,242
96,24396,24496,24596,246
96,247
96,24896,24996,25097,24597,246
97,247
97,24897,24997,25098,246
98,24898,24998,25098,251
98,252
98,253
98,25499,251
99,252
99,25399,25499,25499,255100,255

n_093-Np-238.dat
n_093-Np-239.dat
n_094-Pu-236.dat
n_094-Pu-237.dat
n_094-Pu-238.dat
n_094-Pu-239.dat
n_094-Pu-240.dat
n_094-Pu-241.dat
n_094-Pu-242.dat
n_094-Pu-244.dat
n_094-Pu-246.dat
n_095-Am-240.dat
n_095-Am-241.dat
n_095-Am-242.dat
n_095-Am-242 m1.dat
n_095-Am-243.dat
n_095-Am-244.dat
n_095-Am-244 m1.dat
n_096-Cm-240.dat
n_096-Cm-241.dat
n_096-Cm-242.dat
n_096-Cm-243.dat
n_096-Cm-244.dat
n_096-Cm-245.dat
n_096-Cm-246.dat
n_096-Cm-247.dat
n_096-Cm-248.dat
n_096-Cm-249.dat
n_096-Cm-250.dat
n_097-Bk-245.dat
n_097-Bk-246.dat
n_097-Bk-247.dat
n_097-Bk-248.dat
n_097-Bk-249.dat
n_097-Bk-250.dat
n_098-Cf-246.dat
n_098-Cf-248.dat
n_098-Cf-249.dat
n_098-Cf-250.dat
n_098-Cf-251.dat
n_098-Cf-252.dat
n_098-Cf-253.dat
n_098-Cf-254.dat
n_099-Es-251.dat
n_099-Es-252.dat
n_099-Es-253.dat
n_099-Es-254.dat
n_099-Es-254 m1.dat
n_099-Es-255.dat
n_100-Fm-255.dat

Table 5
Criticality analysis results (keff).
Case

First criticality and excess reactivity loading(First Group)

JENDL-5

1-1
1-2
1-3

First 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)

1.00801
1.101670.92088

Case

Calibrated rod/grid position(calibrated rod position/other rod bank position)(Second Group, 12 FEs, 6 CEs)
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)

JENDL-5

2-1
2-2
2-3
2-4
2-5
2-6

1.00465
1.004781.00599
1.00584
1.00650
1.00614

8

JENDL-4.0±±±

0.00008
0.00008
0.00008

1.00477
1.09949
0.91957

ENDF/B-VIII.0±±±

JENDL-4.0
±
±
±
±
±
±

0.00008
0.00008
0.00008
0.00008
0.00008
0.00008

1.00282
1.00257
1.00379
1.00400
1.004311.004220.00008
0.00008
0.00008

1.00547
1.099120.91974±±±0.00008
0.000070.00008

±
±
±
±
±
±

0.00008
0.00008
0.00008
0.00008
0.00008
0.00008

ENDF/B-VIII.0
±
±
±
±
±
±

0.00008
0.00008
0.00008
0.00008
0.00008
0.00008

1.00302
1.00301
1.00406
1.00415
1.00452
1.00426

P.H. Liem and D. Hartanto

Nuclear Engineering and Design 418 (2024) 112899

Table 6
Sensitivities of keff in the decreasing direction (abs. value > 0.1 %).

(shaded figures show a discrepancy of more than 5 %).

capture profile. The elastic scattering profile has a large profile in thethermal energy region and complicated profiles stem from the reso­
nances. The capture profile is much simpler which shows a large nega­tive profile in the thermal energy region (below 1 eV). The less dominant(n,gamma) or capture sensitivity profiles of Mn, Ag, In, and Cd isotopes
are shown in Fig. 13. Among the isotopes, the Mn-55 shows a notably
large negative profile in the thermal energy region (below 1 eV).

Case 2-1 (Second group, 12 FEs, 6 CEs) where the full core was criticalduring C-8 regulation rod calibration. The uncertainty analysis results
are shown in Table 8 while the uncertainty main contributors (forJENDL-5) are shown in Table 9.
From Table 8, the keff uncertainties due to nuclear data are pre­sented for three combinations of sensitivity coefficients and covariancematrices, i.e. JENDL-5 only, JENDL-5 & ENDF/B-VIII.0, and ENDF/BVIII.0. If only 105 covariance matrices of JENDL-5 was used the un­certainty was around 620 pcm, while if the additional covariancematrices of ENDF/B-VIII.0 was also used, the uncertainty increased to
644 pcm, hence, for the present system, the difference was small, i.e.,around 24 pcm. The uncertainty of ENDF/B-VIII.0 was evaluated to be

4.3. Uncertainty analysis
The uncertainty analysis has been conducted with TSUNAMI-IP
using JENDL-5 and ENDF/B-VIII.0 libraries, for the selected case, i.e.,
Table 7
Sensitivities of keff in the increasing direction (abs. value > 0.1 %).

TSL: Thermal Scattering Law; (shaded figures show a discrepancy of more than 5 %).
9

P.H. Liem and D. Hartanto

Nuclear Engineering and Design 418 (2024) 112899

Fig. 6. Energy-dependent sensitivities of k to U-235 cross-sections (JENDL-5).

Fig. 7. Energy-dependent sensitivities of k to U-238 cross-sections (JENDL-5).

10

P.H. Liem and D. Hartanto

Nuclear Engineering and Design 418 (2024) 112899

Fig. 8. Energy-dependent sensitivities of k to H-1 free-gas and S(α,β) cross-sections (JENDL-5).

Fig. 9. Energy-dependent sensitivities of k to Be-9 free-gas and S(α,β) cross-sections (JENDL-5).

11

P.H. Liem and D. Hartanto

Nuclear Engineering and Design 418 (2024) 112899

Fig. 10. Energy-dependent sensitivities of k to O-16 cross-sections (JENDL-5).

Fig. 11. Energy-dependent sensitivities of k to Al-27 cross-sections (JENDL-5).

12

P.H. Liem and D. Hartanto

Nuclear Engineering and Design 418 (2024) 112899

Fig. 12. Energy-dependent sensitivities of k to Fe-56 cross-sections (JENDL-5).

Fig. 13. Energy-dependent sensitivities of k to Mn, Ag, In and Cd isotopes cross-sections (JENDL-5).

13

P.H. Liem and D. Hartanto

Nuclear Engineering and Design 418 (2024) 112899

Table 8
Uncertainty analysis results (pcm = 10− 5).

Table 10Similarity analysis results with ck > 0.8.

Sensitivity Coefficients

Covariance Matrices

Uncertainty (pcm)No.

Experiment Series

Cases

JENDL-5(MCNP6.2, KSEN option)
JENDL-5(MCNP6.2, KSEN option)
ENDF/B-VIII.0(MCNP6.2, KSEN option)

JENDL-5 only
(AMPX-6)
JENDL-5 & ENDF/B-VIII.0 *(AMPX-6)
ENDF/B-VIII.0(SCALE 6.3.0)

620

1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16

HEU-COMP-THERM-010
HEU-COMP-THERM-021
HEU-MET-THERM-006
HEU-MISC-THERM-002
HEU-MISC-THERM-001
HEU-MISC-THERM-002
HEU-SOL-THERM-005
HEU-SOL-THERM-006
HEU-SOL-THERM-007
HEU-SOL-THERM-021
HEU-SOL-THERM-031
HEU-SOL-THERM-036
HEU-SOL-THERM-044
HEU-SOL-THERM-048
LEU-SOL-THERM-006
LEU-SOL-THERM-016
Total number of cases

1
41
12
1
5
1
11
21
7
1
2
3
13
1
4
1
125

644
637

*: For JENDL-5 nuclides which have no covariance data.

637 pcm, which was comparable with the ones of the JENDL-5. In otherwords, the nuclear data uncertainties of JENDL-5 and ENDF/B-VIII.0 are
in the same order.
In Table 9, the JENDL-5 uncertainty contributors are sorted andlisted when the JENDL-5 & ENDF/B-VIII.0 covariance matrices wereused. The dominant contributors of the table, i.e. > 0.1 %, are U-235(chi), (nubar) & (fission), H-1 (n, gamma), (elastic) & (TSL, inelastic), O16 (elastic), and Al-27 (n, gamma). In the RSG GAS, the LEU (19.75 wt%
U-235 enrichment) is used and consequently, the U-235 fission-relatedparameters contribute to the keff uncertainty significantly. Water isused as a moderator and coolant material so that it exists in the core,reflector, and other structural regions in large quantities. The H-1
component of the water contributes significantly both as free gas(elastic) and when bounded by O-16 (TSL, inelastic).
Recalling back the JENDL-5 [C/E-1] of Case 2-1 shown in Table 5(~465 pcm), the evaluated keff uncertainty of JENDL-5 of 644 pcm wascomparable. However, it is premature to conclude that the deviation of[C/E] from 1.0 originated only from the nuclear data, since other un­
certainties were not taken into account, such as uncertainties in thecompositions, dimensions, criticality measurements, etc.

show ck > 0.8, and their total cases are 125 experiments. Surprisingly,
amongst the 16 series, 14 of them related to high-enriched uranium(HEU) although the RSG GAS core uses a relatively low 19.75 wt%
enrichment (LEU). The other 2 series are under the LEU group. Amongst
these 125 experiments, the top 20 cases are listed in Table 11. The
largest ck (0.857) was found for HEU-MISC-THERM-001-023. HEUMISC-THERM-001 is a criticality experiment series conducted by ORNL
with the SPERT-D experimental reactor (Crawford and Palmer, 1992). In
the SPERT-D experiment series, aluminum-clad plate-type fuel in water,
dilute uranyl nitrate, or borated uranyl nitrate was used. Except that
SPERT-D used HEU (93 wt% enrichment), the fuel elements are indeed
very similar to the ones of the RSG GAS.
From Table 11, it can be concluded that at present there is no single
criticality experiment that has a strong similarity (in this case ck > 0.9)
with the RSG GAS core. Therefore, it is suggested that more criticality
benchmark experiments similar to RSG GAS should be conducted in the
future. Furthermore, the RSG GAS first criticality experiments should beconsidered to be included in the benchmark databases.

4.4. Similarity analysis
The similarity analysis results are shown in Tables 10 and 11. In
Table 10, the criticality experiment series whose member cases have ck> 0.8 are tabulated. We found that 16 criticality experiment series under
the thermal spectrum category with various numbers of member cases
Table 9Uncertainty contributors (>0.001 %).
NuclideReactionwith

NuclideReaction% dR/R due to this matrix

u-235
hfreegas
u-235
h-1
u-235
hfreegas
hfreegas
al-27
al-27
al-27
be-9
be-9
u-238
u-235
u-235
be-9
u-238
bebound
be-9
be-9
fe-56
o-16
u-238
ag-109u-238

chin,gammanubar
n,n’
fissionelastic
elastic
n,gammaelastic
n,n’
elasticn,2nn,gamman,gammafissionn,gamma
n,n’
n,n’
n,2n
n,alphan,gammaelastic
elastic
n,gammaelastic

with
with
with
with
with
with
with
with
with
with
with
with
with
with
with
with
with
with
with
with
with
with
with
with
with

u-235
hfreegas
u-235
h-1
u-235
hfreegas
hfreegas
al-27
al-27
al-27
be-9
be-9
u-238
u-235
u-235
be-9
u-238
bebound
be-9
be-9
fe-56
o-16
u-238
ag-109u-238

chin,gammanubar
n,n’
fissionelastic
n,gamman,gammaelastic
n,n’
elasticn,2nn,gamman,gamman,gamman,gamman,n’n,n’elasticn,alphan,gammaelastic
elastic
n,gamman,gamma

3.76E-01
3.48E-01
3.00E-01
2.41E-01
1.62E-01
1.27E-01− 1.14E-01
1.07E-01
8.63E-02
7.15E-02
6.29E-02
3.56E-02
3.46E-02
3.19E-02− 2.80E-02
2.37E-02
2.03E-02
1.86E-02− 1.71E-02
1.43E-02
1.37E-02
1.28E-02
1.26E-02
1.20E-02
1.16E-02

14

+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−
+/−

1.78E-04
9.93E-06
1.19E-05
6.47E-04
1.08E-05
2.85E-05
1.92E-05
8.73E-07
7.24E-05
2.28E-05
1.17E-05
8.63E-06
5.28E-07
1.48E-07
4.38E-07
2.01E-07
6.28E-06
1.27E-05
1.82E-06
1.77E-07
1.17E-07
7.51E-07
2.76E-06
1.81E-07
2.00E-06

P.H. Liem and D. Hartanto

Nuclear Engineering and Design 418 (2024) 112899

Table 11Top-20 experiments with ck > 0.8.
No

Experiment Casekeffkeff (SD)

Uncertainty (%)

ck

1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20

HEU-MISC-THERM-001-023
HEU-MISC-THERM-001-022
HEU-MISC-THERM-001-021
HEU-MET-THERM-006-018
HEU-MISC-THERM-001-020
HEU-MET-THERM-006-017
HEU-MET-THERM-006-013
HEU-MET-THERM-006-016
HEU-SOL-THERM-006-015
HEU-SOL-THERM-006-024
HEU-SOL-THERM-006-019
HEU-SOL-THERM-006-005
HEU-MET-THERM-006-010
HEU-SOL-THERM-005-004
HEU-MET-THERM-006-011
HEU-SOL-THERM-006-006
HEU-MET-THERM-006-002
HEU-SOL-THERM-006-016
HEU-MET-THERM-006-003
HEU-SOL-THERM-044-024

1.00371.00111.0014
0.99830.9984
0.99781.02061.0017
1.00701.00811.0072
1.0082
1.00761.00930.99940.99940.99930.99891.00411.02000.00030.00030.00030.00030.00030.00030.00030.00030.00020.00020.00030.00020.00030.00030.00030.00020.00030.00020.00030.00050.70850.73230.77530.77560.81730.83260.91420.8206
0.61920.61550.61170.6242
0.95450.67640.90230.57990.9626
0.57550.91980.6459

0.857
0.856
0.849
0.848
0.843
0.839
0.830
0.829
0.828
0.828
0.828
0.827
0.826
0.825
0.825
0.824
0.824
0.824
0.824
0.823

SD: Standard Deviation

5. Concluding remarks

Acknowledgments

The JENDL-5 criticality, sensitivity, uncertainty, and similarity an­alyses were conducted for the clean first core of the RSG GAS Multi­
purpose Reactor, which is a beryllium-reflected, light-water moderated,
19.75 wt% enriched LEU-fueled MTR. The sensitivity coefficientsneeded were prepared with the KSEN option of MCNP6.2 for JENDL-5,
JENDL-4.0, and ENDF/B-VIII.0. New covariance matrices were preparedwith AMPX-6 for JENDL-5 only and the combined JENDL-5 & ENDF/BVIII.0 for JENDL-5 nuclides with missing covariance data. The nuclear
data-originated uncertainty was finally estimated with the TSUNAMI-IP
module of SCALE-6.2.
The criticality (keff) analysis results ([C/E-1]) showed a maximum
overestimation of 801 pcm, 477 pcm, and 547 pcm, for JENDL-5,
JENDL-4.0 and ENDF/B-VIII.0, respectively. Dominant nuclidereaction sensitivities and several large differences between libraries
were identified. The keff uncertainty originated from the nuclear data
was estimated to be 620 pcm, 644 pcm, and 637 pcm for JENDL-5 only,
JENDL-5 & ENDF/B-VIII.0, and ENDF/B-VIII.0 covariance data,
respectively, which are comparable with the keff ([C/E-1]) values.
The similarity analysis results showed that at present there is no
single criticality experiment that has a strong similarity (ck > 0.9) with
the RSG GAS core.

We express our gratitude to all members of the commissioning groupof RSG GAS for their efforts in compiling the criticality experiment data.
We are also grateful to Mrs. Suzuko Ikegami for the preparation of thetables and graphs.
References
BATAN, 1980. Multipurpose Reactor G. A. Siwabessy Safety Analysis Report - Rev. 7,
Jakarta.
Bostelmann, F., et al., 2021. On the creation of the new ENDF/B-VIII.0 Covariance
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CRediT authorship contribution statement
Peng Hong Liem: Conceptualization, Methodology, Investigation,
Formal analysis, Data curation, Visualization, Writing – original draft,
Writing – review & editing, Supervision. Donny Hartanto: Data cura­tion, Formal analysis, Methodology, Software, Visualization, Writing –original draft, Writing – review & editing.
Declaration of competing interest
The authors declare that they have no known competing financial
interests or personal relationships that could have appeared to influence
the work reported in this paper.
Data availability
The authors do not have permission to share data.

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Nuclear Engineering and Design 418 (2024) 112899

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