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Analysis of the First Core of the Indonesian Multipurpose Research Reactor
RSG-GAS using the Monte Carlo Serpent Code and the ENDF/B-VIII.0 Nuclear
Data Library
Article in Nuclear Engineering and Technology · May 2020
DOI: 10.1016/j.net.2020.05.027

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Nuclear Engineering and Technology xxx (xxxx) xxx

Contents lists available at ScienceDirect

Nuclear Engineering and Technology
journal homepage: www.elsevier.com/locate/net

Analysis of the ﬁrst core of the Indonesian multipurpose research
reactor RSG-GAS using the Serpent Monte Carlo code and the ENDF/BVIII.0 nuclear data library
Donny Hartanto a, *, Peng Hong Liem b, cabc

Department of Mechanical and Nuclear Engineering, University of Sharjah, P.O. BOX 27272, Sharjah, United Arab Emirates
Cooperative 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, Japana r t i c l e i n f o

a b s t r a c t

Article history:
Received 16 January 2020
Received in revised form
8 May 2020
Accepted 26 May 2020
Available online xxx

This paper presents the neutronics benchmark analysis of the ﬁrst core of the Indonesian multipurpose
research reactor RSG-GAS (Reaktor Serba Guna G.A. Siwabessy) calculated by the Serpent Monte Carlocode and the newly released ENDF/B-VIII.0 nuclear data library. RSG-GAS is a 30 MWth pool-type material testing research reactor loaded with plate-type low-enriched uranium fuel using light water as acoolant and moderator and beryllium as a reﬂector. Two groups of critical benchmark problems arederived on the basis of the criticality and control rod calibration experiments of the ﬁrst core of RSG-GAS.
The calculated results, such as the neutron effective multiplication factor (k) value and the control rod
worth are compared with the experimental data. Moreover, additional calculated results, including theneutron spectra in the core, ﬁssion rate distribution, burnup calculation, sensitivity coefﬁcients, andkinetics parameters of the ﬁrst core will be compared with the previous nuclear data libraries (interlibrary comparison) such as ENDF/B-VII.1 and JENDL-4.0. The C/E values of ENDF/B-VIII.0 tend to beslightly higher compared with other nuclear data libraries. Furthermore, the neutron reaction crosssections of 16O, 9Be, 235U, 238U, and S(a,b) of 1H in H2O from ENDF/B-VIII.0 have substantial updates;
hence, the k sensitivities against these cross-section changes are relatively higher than other isotopes in
RSG-GAS. Other important neutronics parameters such as kinetics parameters, control rod worth, andﬁssion rate distribution are similar and consistent among the nuclear data libraries.
© 2020 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the
CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/).

Keywords:
RSG-GAS First core
SERPENT
ENDF/B-VIII.0
Critical benchmark
Sensitivity analysis

1. Introduction
Recently, the Cross Section Evaluation Working Group, in
collaboration with several international organizations has released
a new ENDF/B-VIII.0 nuclear data library [1]. The new nuclear data
cover neutron reaction libraries for 557 isotopes and thermal
scattering libraries for 34 materials. The release is a signiﬁcant
improvement compared with its predecessor ENDF/B-VII.1, which
has neutron reaction libraries for 423 isotopes and thermal scattering libraries for 21 materials [2]. Additionally, the new nuclear
data adopts substantial updates for neutron reaction libraries suchas 1H, 9Be, 56Fe, 235U, and 238U [1] that affect the criticality calculation [3e5] including the ﬁrst core of the Indonesian multipurpose* Corresponding author.
E-mail address: dhartanto@sharjah.ac.ae (D. Hartanto).

research reactor RSG-GAS.
The ﬁrst core of RSG-GAS is a neutronics benchmark problem
deﬁned as a research reactor with 20% enriched U3O8 fuel dispersedin the Al matrix using light water a moderator/coolant and beryllium reﬂectors [6]. The benchmark was derived from the reactor
physics experiments during the ﬁrst criticality of the reactor back inthe mid of 1987. Only a few well-documented critical benchmark
problems are available and open to the public based on a research
reactor that uses 20% enriched uranium fuels. Therefore, the ﬁrst
core of the RSG-GAS benchmark problem is considered important,and it can be used to verify and validate the neutronics calculation
tool and the nuclear data library to support the research reactor
community as suggested by Komuro [7].
The benchmark problem consists of two different core conﬁgurations, one is the ﬁrst criticality core that has nine fresh standard
fuel elements and six control fuel elements, and another is the ﬁrst
working core of RSG-GAS that is made up of 12 fresh standard fuelhttps://doi.org/10.1016/j.net.2020.05.027
1738-5733/© 2020 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/
licenses/by-nc-nd/4.0/).

Please cite this article as: D. Hartanto, P.H. Liem, Analysis of the ﬁrst core of the Indonesian multipurpose research reactor RSG-GAS using theSerpent Monte Carlo code and the ENDF/B-VIII.0 nuclear data library, Nuclear Engineering and Technology, https://doi.org/10.1016/
j.net.2020.05.027

2

D. Hartanto, P.H. Liem / Nuclear Engineering and Technology xxx (xxxx) xxx

Fig. 1. Standard (left) and control (right) fuel element layout.

Table 1
Major parameters of standard and control fuel elements [6].
Parameter

Value

Dimension (mm)
Fuel plate thickness (mm)
Coolant channel width (mm)No. of plate per FE
No. of plate per CE
Fuel plate clad material
Fuel plate clad thickness (mm)
Fuel meat dimension (mm)
Fuel meat material
U-235 enrichment (wt%)
U density in meat (g/cc)
U-235 loading per FE (g)U-235 loading per CE (g)Absorber meat material
Absorber thickness (mm)Absorber clad material
Absorber clad thickness (mm)77.1  81  600
1.3
2.55
21
15
AlMg2
0.38
0.54  62.75  600
U3O8eAl
19.75
2.96
250
178.6
AgeIneCd
3.38
SS-3210.85

elements and six control fuel elements. Liem [6] described thedetail of the ﬁrst core criticality experiment. Since the benchmark
problem was established, extensive works such as criticality calculations using the Monte Carlo codes [6,8e12], criticality calculation using deterministic codes [13,14], and control rod worth
calculation using the Monte Carlo and deterministic codes [15]have been done.
In this study, the new ENDF/B-VIII.0 is applied to these benchmark problems to investigate the impact of the newly released library on the RSG-GAS. The benchmark calculation work covers theeffective neutron multiplication factor and control rod worth,which are compared against the experimental data. Furthermore,neutron spectra in the core, ﬁssion rate distributions, kinetics parameters, burnup calculation, and the sensitivity coefﬁcients of theﬁrst core calculated using ENDF/B-VIII.0 are also compared with theones with the previous version of nuclear data libraries such as
ENDF/B-VII.1 [2] and JENDL-4.0 [16]. The calculations have been
performed by using the 3-D continuous-energy Serpent Monte
Carlo code [17]. The Monte Carlo method is recognized for its accuracy because it solves the neutron transport equation without
any approximations, such as no simpliﬁcation in the geometry andno discretization in the neutron energy. Conversely, the computing

time of the Monte Carlo method is expected to be longer. Nonetheless, the issue can be managed with the recent advancement incomputing technology. The Serpent Monte Carlo code is chosen inthis study because it is widely used and well veriﬁed and validated,
including its application for the research reactor. Moreover, thecomputing speed of Serpent is expected to be faster because it
adopts the unionized energy grid format and the Woodcock deltatracking method [17].
The rest of this paper is organized as follows: Section 2 discusses
the description of the RSG-GAS ﬁrst core, Section 3 analyzes anddeals with the results, and Section 4 presents the conclusions andrecommendations for future works.

2. First core of RSG-GAS description
The RSG-GAS multipurpose research reactor is a pool-typematerial testing research reactor that is owned and operated by
the Indonesian National Nuclear Energy Agency (BATAN). The
location is in Serpong, West Java, and it can operate at a maximum
power of 30 MWth for approximately 30 full power days andachieve the maximum discharge burnup up to 53.7%. The reactor
uses low-enriched uranium (19.75% U-235) fuel using light water asa moderator and coolant, and beryllium reﬂectors. Initially, the fuelwas U3O8 dispersed in the Al matrix; however, it was converted toU3Si2 fuel dispersed in the Al matrix. The typical working core was
loaded with 40 standard and control fuel elements. Since it reachedthe ﬁrst criticality in July 1987, RSG-GAS has been used for many
applications, such as the production of medical and industrial radioisotopes, the irradiation of gemstones, and the testing ofmaterials.
Fig. 1 and Table 1 show the standard and control fuel element
conﬁguration and parameters. As depicted, the standard fuelelement (FE) consists of 21 fuel plates, while the control fuelelement (CE) has 15 fuel plates to provide a space for the control
rods. Each fuel plate contains 19.75% enriched U3O8 meat dispersed
in an aluminum matrix and surrounded by aluminum cladding on
both sides of the fuel meat. The fuel plates are arranged and held by
two Al side plates. Meanwhile, the control rod (CR) is a fork-typecontaining a silver-indium-cadmium absorber.
As mentioned above, two core conﬁgurations were documented
during the experiment to achieve the ﬁrst working core. Fig. 2

Please cite this article as: D. Hartanto, P.H. Liem, Analysis of the ﬁrst core of the Indonesian multipurpose research reactor RSG-GAS using theSerpent Monte Carlo code and the ENDF/B-VIII.0 nuclear data library, Nuclear Engineering and Technology, https://doi.org/10.1016/
j.net.2020.05.027

D. Hartanto, P.H. Liem / Nuclear Engineering and Technology xxx (xxxx) xxx

Fig. 2. First criticality core conﬁguration of RSG-GAS.

Fig. 3. First working core conﬁguration of RSG-GAS.

Please cite this article as: D. Hartanto, P.H. Liem, Analysis of the ﬁrst core of the Indonesian multipurpose research reactor RSG-GAS using theSerpent Monte Carlo code and the ENDF/B-VIII.0 nuclear data library, Nuclear Engineering and Technology, https://doi.org/10.1016/
j.net.2020.05.027

3

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D. Hartanto, P.H. Liem / Nuclear Engineering and Technology xxx (xxxx) xxx
Table 2
Calculation cases for the ﬁrst core of RSG-GAS.
First group (First criticality and excess reactivity loading)Remarks

First criticality (nine FEs, six CEs, RR inserted up to 475 mm)
Full core (12 FEs, six CEs, CRs all up)
Full core (12 FEs, six CEs, CRs all down)

CriticalSupercritical, k ¼ 1.09242 [10]Subcritical,
CR worth ¼ 17.18% [10]Second group (Control rod calibrations, full core/12 FEs, six CEs)

Remarks

CR ID: JDA06
C-8 fully inserted, other rods inserted up to 290 mm
CR ID: JDA01
E9 fully inserted, other rods inserted up to 284 mm
CR ID: JDA03
F-8 fully inserted, other rods inserted up to 293 mm
CR ID: JDA05
C-5 fully inserted, other rods inserted up to 288 mm
CR ID: JDA04
F-5 fully inserted, other rods inserted up to 290 mm
CR ID: JDA07
D-4 fully inserted, other rods inserted up to 282 mm

CriticalTable 3C/E of k for different nuclear data libraries.
First group

ENDF/B-VIII.0

ENDF/B-VII.1

JENDL-4.0

First criticality
Full core (CRs all up)
Full core (CRs all down)

1.00581.0065
N/A

1.00511.0064N/A

1.00471.0062
N/A

1.00311.00311.00431.00431.00481.00451.00301.00281.00411.00411.00451.0042

1.00261.00251.00371.00381.0042
1.0038

Second group
CR ID: JDA06
CR ID: JDA01
CR ID: JDA03
CR ID: JDA05
CR ID: JDA04
CR ID: JDA07

CriticalCriticalCriticalCriticalCritical

depicts the smallest core arrangement that achieved the ﬁrst criticality. Nine FEs are located at grids C-6, C-7, D-5, D-7, D-8, D-9, E6,E8, and F-7, while the six CEs are located at grids C-5, C-8, D-4,
E9, F-5, and F-8. The regulation rod (RR, positioned at grid C-8)
was inserted slightly to make the core critical. By contrast, the ﬁrst
full working core conﬁguration, illustrated in Fig. 3, consists of 12
FEs and six CEs. The locations of the six CEs are identical to theprevious conﬁguration. However, in addition to the nine FEs in theﬁrst criticality core, three more were added at grids E4, E5, andF-6. Moreover, the number of beryllium reﬂector surrounding thecore was increased in the latter conﬁguration to reduce the neutron
leakage. This core conﬁguration was made critical by inserting
partial control rods into the core.
Table 2 summarizes the benchmark calculation cases of the ﬁrst

Fig. 4. Neutron spectra for different nuclear data libraries.

Please cite this article as: D. Hartanto, P.H. Liem, Analysis of the ﬁrst core of the Indonesian multipurpose research reactor RSG-GAS using theSerpent Monte Carlo code and the ENDF/B-VIII.0 nuclear data library, Nuclear Engineering and Technology, https://doi.org/10.1016/
j.net.2020.05.027

D. Hartanto, P.H. Liem / Nuclear Engineering and Technology xxx (xxxx) xxx

5

Table 4
Effect of speciﬁc isotope neutron reaction library from ENDF/B-VIII.0 on k of othernuclear data libraries.
Isotopes from ENDF/B-VIII.0

ENDF/B-VII.1

JENDL-4.0

27

7.00 ± 4.67
4.00 ± 4.67
28.00 ± 4.67
56.00 ± 4.67
12.00 ± 4.60
93.00 ± 4.67
59.00 ± 4.60
28.00 ± 4.60
29.00 ± 4.67
8.00 ± 4.67

2.00 ± 4.67
29.00 ± 4.60
165.00 ± 4.60
51.00 ± 4.60
0.00 ± 4.60
148.00 ± 4.67
72.00 ± 4.60
47.00 ± 4.60
180.00 ± 4.60
4.00 ± 4.67

Al in fuel27
Al in structural
9
Be
1
H
16
O in fuel16
O in coolant
235
U
238
U
S(a,b) of 1H in H2O
S(a,b) of 9Be

core of RSG-GAS performed in this work and categorizes them into
the ﬁrst and second groups. The ﬁrst group has three cases, but only
the ﬁrst criticality case was measured directly during the criticality
approach experiment. The other two cases were derived on thebasis of the incorporation of the experimental result and additional
analytical calculation [8,10]. Meanwhile, the second group has six
critical cases, and they were evaluated directly from the control rod
calibration experiment. Each case in the second group is identiﬁedby its corresponding control rod identiﬁcation number such as
JDA01 for the control rod at grid E9.

3. Analysis results and discussion
Before dealing with the results, the 3-D calculation model andcondition are discussed brieﬂy, and the geometry of the active region of the ﬁrst core RSG-GAS in this work is modeled explicitly.
However, homogenized mixture models by conserving the volumeof each region are used for the top and end-ﬁtting regions of theelements, core grid, and bottom supports. This approach is expected to have a negligible impact on the criticality calculation
because they are located outside of the active core region. Allcomponents in the core have a temperature of 300 K. The Serpent

Fig. 6. k as a function of burnup for different nuclear data libraries.

Monte Carlo code has been used with 400,000 neutron histories
per cycle, and the total number of cycles is 2200 with 200 inactive
cycles. Using this calculation conditions, the standard deviation of kis less than 4 pcm. Moreover, the thermal scattering libraries S(a,b)
for hydrogen in the light water and beryllium as metal are includedin the calculation.
Table 3 compares the ratio of the k values of Serpent to theexperimental values for different nuclear data libraries and for core
conﬁgurations. Overall, the calculation results are very close to theexperimental results taking into account the uncertainties due tonuclear data is about 0.006 [11], and the new ENDF/B-VIII.0 provides slightly higher k values than other libraries. The highest
overestimation was observed for the full core conﬁguration without
any insertion of control rods (1st group, 2nd case), which is nearly
0.646%. It was expected as the reference k value was not purely
experimental data. During the experiment, the excess reactivity

Fig. 5. Relative radial ﬁssion power distribution for different nuclear data libraries.

Please cite this article as: D. Hartanto, P.H. Liem, Analysis of the ﬁrst core of the Indonesian multipurpose research reactor RSG-GAS using theSerpent Monte Carlo code and the ENDF/B-VIII.0 nuclear data library, Nuclear Engineering and Technology, https://doi.org/10.1016/
j.net.2020.05.027

6

D. Hartanto, P.H. Liem / Nuclear Engineering and Technology xxx (xxxx) xxx

Table 5
Positive sensitivities of k.
Isotope

Reaction

ENDF/B-VIII.0

Ratio of ENDF/B-VII.1 to ENDF/B-VIII.0

Ratio of JENDL-4.0 to ENDF/B-VIII.0

235

n prompt

9.878E-01
3.503E-01
2.922E-01
5.967E-02
4.275E-02
2.807E-02
2.767E-02
1.100E-02
1.082E-02
7.394E-03
7.281E-03
4.806E-03
4.675E-03
3.373E-03
2.032E-03
1.752E-03

1.00
1.00
1.01
1.02
1.02
0.92
1.00
1.03
1.02
0.91
1.01
1.01
0.97
1.01
1.04
0.98

1.00
1.00
1.01
1.01
1.00
1.07
1.05
1.05
0.93
0.64
1.01
1.03
1.03
1.01
1.13
1.04

U
235
U
1
H
9
Be
16
O
1
H
27
Al
9
Be
27
Al
9
Be
235
U
238
U
238
U
238
U
238
U
56
Fe

(n,f)
(n,el)(n,el)(n,el)
S(a,b)(n,el)(n,xn)(n,inl)S(a,b)n delayedn prompt(n,el)(n,f)(n,inl)(n,el)

Table 6
Negative sensitivities of k.
Isotope

Reaction

ENDF/B-VIII.0

Ratio of ENDF/B-VII.1 to ENDF/B-VIII.0

Ratio of JENDL-4.0 to ENDF/B-VIII.0

1

(n,g)
(n,g)
(n,g)
(n,g)
(n,g)
(n,g)
(n,g)
(n,g)
(n,g)
(n,g)
(n,g)
(n,g)

1.679E-01
1.237E-01
5.061E-02
3.758E-02
8.391E-03
4.346E-03
3.388E-03
2.450E-03
1.777E-03
1.759E-03
1.351E-03
1.217E-03

1.00
0.99
1.01
1.01
1.01
1.01
1.01
0.97
1.02
0.76
1.03
1.02

1.01
0.99
0.99
1.02
0.93
1.02
1.02
0.98
1.04
0.75
0.96
1.01

H

235

U
27
Al
238
U
9
Be
55
Mn
109
Ag
115
In
107
Ag
16
O
113
Cd
56
Fe

Fig. 7. Control rod worth for different nuclear data libraries.

was evaluated by the summation of the individually measured
control rod reactivity, and the interference effect of the control rod
was neglected. Moreover, the excess reactivity was initially in a unitof $, and it was converted later using the beff (total effective delayed
neutron fraction) value calculated by the Monte Carlo method as
the measured beff value was unavailable [10].
For the achievement of prime insight into the new ENDF/B-VIII.0
nuclear data library, the neutron spectra in the fuel meat arecompared in Fig. 4, the effect of the neutron cross-section of thespeciﬁc isotope from ENDF/B-VIII.0 is investigated and presented inTable 4, and the radial power distribution is plotted in Fig. 5. The
calculation was performed for the full core conﬁguration without
any control rods (1st group, 2nd case). Fig. 4 depicts that a sharp
increase arises in the relative difference of the spectra at the highneutron energy region, which is regarded as likely as the neutron
ﬂux at that energy is fairly low and thus the uncertainty of ﬂux forthat energy region calculated by Serpent is high. However, theneutron spectrum by JENDL-4.0 has a higher relative difference to
ENDF/B-VIII.0 compared with ENDF/B-VII.1.
Table 4 shows the effect of the neutron cross-section library ofseveral main isotopes in RSG-GAS from ENDF/B-VIII.0 on the k valueof other nuclear data libraries. For ENDF/B-VII.1, the maximum
decrease of k is given by 16O (in coolant) roughly 93 pcm, followedby 1H (in coolant) approximately 56 pcm, S(a,b) of H in H2O nearly
29 pcm, and 9Be roughly 28 pcm from ENDF/B-VIII.0, and k increases approximately 59 pcm when 235U from ENDF/B-VIII.0 isused. For JENDL-4.0, replacing 9Be and 16O (in coolant) with ENDF/
B-VIII.0 decreases k by roughly 165 pcm and 148 pcm, respectively,
whereas replacing S(a,b) of H in H2O, 235U, and 238U with ENDF/BVIII.0 increases k by nearly 180 pcm, 72 pcm, and 47 pcm, respectively. It is worth mentioning that 1H, 16O, 235U, and 238U are theisotopes that have major updates in ENDF/B-VIII.0 from CIELO(Collaborative International Evaluated Library Organization) evaluations [18]. Isotope Be-9 and S(a,b) for H in H2O were also updated
in ENDF/B-VIII.0 [1].
The relative ﬁssion rate distribution is calculated by scoring theﬁssion reaction rate using the reaction multiplier in Serpent. The
result illustrated in Fig. 5 shows the peak-to-average radial ﬁssion

Please cite this article as: D. Hartanto, P.H. Liem, Analysis of the ﬁrst core of the Indonesian multipurpose research reactor RSG-GAS using theSerpent Monte Carlo code and the ENDF/B-VIII.0 nuclear data library, Nuclear Engineering and Technology, https://doi.org/10.1016/
j.net.2020.05.027

D. Hartanto, P.H. Liem / Nuclear Engineering and Technology xxx (xxxx) xxx

7

Table 7
Kinetics parameters for different nuclear data libraries.
First group

Kinetics Parameters

ENDF/B-VIII.0

ENDF/B-VII.1

JENDL-4.0

Abs. RMSE

First criticality

beff (pcm)
L (ms)a (1/s)beff (pcm)
L (ms)a (1/s)beff (pcm)
L (ms)a (1/s)747.16 ± 1.6
84.06 ± 0.04
88.84 ± 0.20
730.01 ± 1.4
73.91 ± 0.03
98.82 ± 0.20
742.94 ± 1.7
67.70 ± 0.03
109.73 ± 0.26

747.44 ± 1.6
84.95 ± 0.04
87.94 ± 0.19
730.08 ± 1.4
74.69 ± 0.03
97.83 ± 0.19
743.42 ± 1.7
68.32 ± 0.03
108.79 ± 0.26

749.30 ± 1.6
86.81 ± 0.04
86.27 ± 0.19
730.30 ± 1.5
76.20 ± 0.03
95.61 ± 0.19
744.74 ± 1.7
69.38 ± 0.03
107.31 ± 0.25

0.95
1.14
1.07
0.12
0.95
1.34
0.76
0.70
0.99

Full core (CRs all up)

Full core (CRs all down)

power distribution in each fuel elements. It is shown that there is agood agreement among the three libraries with a relative differenceof less than 0.8%.
A hypothetical depletion calculation was also performed for thesame core conﬁguration with all control rods withdrawn. The
thermal power was set to 10 MWth in the calculation to accommodate a small core size, and it was depleted up to 25 days. The
Chebyshev Rational Approximation Method was selected as thedepletion algorithm, while linear extrapolation and linear interpolation are adopted for the predictor and corrector methods,
respectively, as recommended [19]. For each library, its corresponding version of radioactive decay data and neutron ﬁssion
yield data was used. Fig. 6 depicts that the k values as the functionof burnup among the three libraries are consistent, and the fraction
losses of 235U after 25 days for ENDF/B-VIII.0, ENDF/B-VII.1, andJENDL-4.0 are 7.617%, 7.605%, and 7.602%, respectively, whichshows excellent agreement among the nuclear data libraries.
However, the impact of the updated decay data library can beconsidered small in this hypothetical burnup calculation as thepower used is small and the total burn days are 25 days.
The k sensitivity analysis was also performed for the 1st case of
the 2nd group (CR ID: JDA06) where the control rod was calibrated;
thus, the sensitivity of the isotopes and reactions in the control
absorber can be analyzed. The k sensitivity is deﬁned as the ratio
between the relative change in k to the relative change in parameters such as cross section and n (average number of neutrons per
ﬁssion). In Serpent, the sensitivity is calculated by the collisionhistory method [20]. The results are presented in Tables 5 and 6 forpositive and negative sensitivities to k, respectively. Only the sensitivities that have an absolute value larger than 0.1% are shown andthe ratio of calculated sensitivity using ENDF/B-VII.1 or JENDL-4.0 to
ENDF/B-VIII.0 that has an absolute difference larger than 5% arehighlighted in grey. Table 5 exhibits that the elastic scattering is thedominant reaction contributing to the positive sensitivities, followed by others such as n, ﬁssion, S(a,b), inelastic scattering, andneutron production (n,xn). The top ﬁve contributors in Table 5 are nprompt and (n,f) of 235U, (n,el) of 1H, 9Be, and 16O, and S(a,b) of 1H in
H2O. Meanwhile, in Table 6, the dominant reaction of the negative
sensitivities is neutron capture, and the top 5 contributors are 1H,
235
U, 27Al, 238U, and 9Be. These results are also consistent with theprevious sensitivities produced by Whisper-1.1 and ENDF/B-VII.1[11]. With regard to the inter-library comparison, JENDL-4.0 produces signiﬁcant differences compared with ENDF/B-VIII.0, especially for S(a,b) of 9Be, (n,g) of 16O, (n,inl) of 238U, S(a,b) of 1H in H2O,
(n,inl) of 27Al, and (n,g) of 9Be, while ENDF/B-VII.1 provides significant differences for (n,g) of 16O and S(a,b) of 9Be and 1H in H2O.
These results are clearly reﬂected in Table 4 when the speciﬁc
isotope neutron reaction library from ENDF/B-VIII.0 is used.
The total control rod worth is compared in Fig. 7, and it isdeﬁned as the reactivity difference when all control rods arewithdrawn and inserted. During the experiment, the worth of eachcontrol rod was measured by the shim rod bank compensation
method, and the arithmetic summation of each control rod worth
was done to obtain the total control rod worth, without considering
the rod shadowing effect. Accordingly, the difference is noticed
between the experimental and calculated results. However, thetotal control rod worth of the three nuclear data libraries agrees
very well with an absolute root mean square error (RMSE) of 0.03.
The calculated kinetics parameters, beff (total effective delayed
neutron fraction) L (prompt neutron generation lifetime), and a(Rossi-alpha value) are summarized in Table 7 for the core conﬁgurations in the 1st group. The kinetics parameters were calculatedusing the adjoint method available in Serpent, speciﬁcally by theiterated ﬁssion probability method. Generally, the value of the kinetics parameters for each core is consistent among the three nuclear data libraries with the maximum absolute RMSEs of 0.95, 1.14,
1.34 for beff, L, and a respectively.
4. Conclusions
The newly released ENDF/B-VIII.0 nuclear data library has been
adopted to characterize several important neutronics parameters of
the ﬁrst core of the Indonesian multipurpose research reactor RSGGAS benchmark problem. The Serpent Monte Carlo code has beenused as the calculation tool. Moreover, the C/E of the k values by
ENDF/B-VIII.0 tends to be slightly higher than other nuclear datalibraries, and the maximum value is approximately 1.00646 for thefull working core conﬁguration when all the control rods arewithdrawn. The 16O neutron cross from ENDF/B-VIII.0 provides thelargest reduction to the k value of ENDF/B-VII.1 and using 9Be and16
O neutron cross sections from ENDF/B-VIII.0 contributes largely tothe decrease of the k value of JENDL-4.0. On the one hand, replacing
S(a,b) of 1H in H2O, 235U and 238U neutron cross sections from
ENDF/B-VIII.0 cause the increase of the k value of JENDL-4.0. These
isotopes also produce relatively high sensitivity coefﬁcients of k. On
the other hand, the kinetics parameters, control rod worth, ﬁssion
rate distribution, and average burnup in %loss of 235U are consistentamong the nuclear data libraries. In the future, the uncertainty
analysis of k using ENDF/B-VIII.0 is planned to be conducted.
Declaration of competing interest
The authors declare that they have no known competing
ﬁnancial interests or personal relationships that could haveappeared to inﬂuence the work reported in this paper.
Appendix A. Supplementary data
Supplementary data to this article can be found online athttps://doi.org/10.1016/j.net.2020.05.027.

Please cite this article as: D. Hartanto, P.H. Liem, Analysis of the ﬁrst core of the Indonesian multipurpose research reactor RSG-GAS using theSerpent Monte Carlo code and the ENDF/B-VIII.0 nuclear data library, Nuclear Engineering and Technology, https://doi.org/10.1016/
j.net.2020.05.027

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Please cite this article as: D. Hartanto, P.H. Liem, Analysis of the ﬁrst core of the Indonesian multipurpose research reactor RSG-GAS using theSerpent Monte Carlo code and the ENDF/B-VIII.0 nuclear data library, Nuclear Engineering and Technology, https://doi.org/10.1016/
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