Nuclear Engineering and Technology 55 (2023) 1775e1782

Contents lists available at ScienceDirect

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

Evaluation of neutronics parameters during RSG-GAS commissioning
by using Monte Carlo code
Surian Pinem a, *, Wahid Luthﬁ a, Peng Hong Liem b, c, Donny Hartanto d, 1aResearch Center for Nuclear Reactor Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency (BRIN), 80th
Building of PUSPIPTEK Serpong, South Tangerang, Banten, Indonesiab
Cooperative Major in Nuclear Energy, Graduate School of Engineering, Tokyo City University (TCU), 1-28-1, Tamazutsumi, Setagaya, Tokyo, Japanc
Scientiﬁc Computational Division, Nippon Advanced Information Service, (NAIS Co., Inc.), 416 Muramatsu, Tokaimura, Ibaraki, Japand
Oak Ridge National Laboratory One Bethel Valley Road, Oak Ridge, TN, 37830, USAa r t i c l e i n f o

a b s t r a c t

Article history:
Received 30 September 2022
Received in revised form
19 December 2022
Accepted 8 January 2023
Available online 16 January 2023

Several reactor physics commissioning experiments were conducted to obtain the neutronic parametersat the beginning of the G.A. Siwabessy Multi-purpose Reactor (RSG-GAS) operation. These parameters are
essential for the reactor to safety operate. Leveraging the experimental data, this study evaluated thecalculated core reactivity, control rod reactivity worth, integral control rod reactivity curve, and fuelreactivity. Calculations were carried out with Serpent 2 code using the latest neutron cross-section data
ENDF/B-VIII.0. The criticality calculations were carried out for the RSG-GAS ﬁrst core up to the third core
conﬁguration, which has been done experimentally during these commissioning periods. The excess
reactivity for the second and third cores showed a difference of 510.97 pcm and 253.23 pcm to theexperiment data. The calculated integral reactivity of the control rod has an error of less than 1.0%
compared to the experimental data. The calculated fuel reactivity value is consistent with the measured
data, with a maximum error of 2.12%. Therefore, it can be concluded that the RSG-GAS reactor core modelis in good agreement to reproduce excess reactivity, control rod worth, and fuel element reactivity.
© 2023 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:
Neutronic parameter
Control rod worth
Fuel element reactivity
RSG-GAS
Serpent 2

1. Introduction
The G. A. Siwabessy Multi-purpose Reactor (RSG-GAS) is an open
pool, water-cooled, and water-moderated reactor with a nominal
power of 30 MW, which reached its ﬁrst criticality in July 1987. The
reactor core uses plate-type U3O8eAl fuel, which is later converted toU3Si2eAl fuel with the same uranium density of 2.96 g/cc andenriched at 19.75%. For a better neutron economy, beryllium is used
as a reﬂector. The RSG-GAS typical working core conﬁguration (TWC)
consists of 40 fuel elements (FE), 8 control elements (CE), and 30
beryllium reﬂector elements. The TWC core was achieved through 5transition cores with different amounts of fuel loading. The averagethermal neutron ﬂux is 2.0  1014 n/cm2s, and its maximum neutron
ﬂux is at the center irradiation position (CIP), up to 5.38  1014 n/
cm2s. The RSG-GAS is equipped with several test facilities such as a
CIP, 4 small irradiation positions (IP) in the reactor core, beam tubes
for radioisotope production and basic science experiments, and power reactors fuel development such as power ramp test, fuel irradiation facilities, and others.
During the RSG-GAS commissioning period, experiments related* Corresponding author.
E-mail address: suri001@brin.go.id (S. Pinem).
1
Work completed while employed at the University of Sharjah.

to the reactor physics parameters have been carried out. These parameters are essential for the safety of reactor operation. With theavailability of the commissioning experiment data, as well as therecent rapid development of high-ﬁdelity reactor physics codes andnuclear data libraries, it is now possible to evaluate the reactor physics
parameters with almost exact modeling of the experimented reactor
conditions. It is worthily noted here that not only static experiments
but whole core burnup analyses based on the detailed operation
history (including control rod positions) are becoming realistic.
Reactivity is among these parameters, described as the departurefrom reactor criticality. Variations in absolute value, initiating event,and reactivity duration directly inﬂuence reactor safety operation,
making reactivity one of the most important parameters related toreactor kinetics. The reactivity of the control rod depends on theposition of the rod inside the core, the displacement of the control rod
position, and its material composition [1,2]. For this reason, themeasurement of control rod reactivity during the commissioning
period of RSG-GAS ﬁrst core was carried out by applying several
methods such as rod drop, reactor period method, reactivity meter,
and bank rod versus single rod compensation. All methods have theiradvantages and disadvantages. Based on the previous evaluation
measured control rod reactivity in RSG-GAS ﬁrst core, the compensation method is chosen to measure control rod reactivity [3].

https://doi.org/10.1016/j.net.2023.01.008
1738-5733/© 2023 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/).

S. Pinem, W. Luthﬁ, P.H. Liem et al.

Nuclear Engineering and Technology 55 (2023) 1775e1782

The commissioning of the RSG-GAS provides a comprehensive
range of experimental or measurement data for neutronics parameters with various core conﬁgurations and their corresponding
control rod positions. These data can be used as benchmarking data
to verify and validate the model of the reactor core. The criticality
calculation and kinetic parameters of RSG-GAS ﬁrst core have been
carried out using various methods [4e10]. In this paper, we evaluate the excess reactivity, the control rod worth, the integral
reactivity curve, and the reactivity of the fuel element. The resultsof this calculation could become good benchmark data for validating the calculation technique related to neutronics parametersin research reactors. The evaluation of neutronics parameters was
carried out using the RSG-GAS second (transition) core because, inthat core, most of the fuel elements from the ﬁrst core were also
being used besides several new fuel elements. The measurement
data for neutronics parameters, i.e., the criticality, control rod
worth, control rod reactivity curve, and fuel element reactivity
value, are also available at the beginning of the cycle (BOC) and endof the cycle (EOC) conditions. The reactor was operated in one
operating cycle from BOC to EOC for about 75.6 days (322,609
MWD). Calculations were performed using a continuous energyMonte Carlo Serpent 2 code [11,12] with the latest nuclear data library ENDF/B-VIII.0 [13]. The analyses are carried out according tothe operation history of the particular core, and the calculated results are compared with the experimental data.

Table 1Main design data of RSG-GAS [14].
Parameters

First core

Nominal power (MWth)No. of fuel elements
No. of control elements
No. of beryllium blocks
Fuel/control element dimension (mm)
Fuel plate thickness (mm)
Coolant channel width (mm)Number of plates per fuel element
Number of plates per control element
Fuel plate clad material
Fuel plate clad thickness (mm)
Fuel plate dimension (mm)
Fuel meat material
U-235 enrichment (w/o)
Uranium density in meat (g/cc)
U-235 loading per fuel element (g)U-235 loading per control element (g)Absorber meat material
Absorber thickness (mm)Absorber clad material

10.6
13.60
12
16
6
6
42
58
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-321

Second core

Third core
18.19
22
8
29

aluminum matrix, with 19.75% enriched uranium and 2.96 gU/cc
uranium density. The active length of the fuel element is 60 cm,
with nominal U-235 loading per standard fuel element being 250 g.
The control fuel element is a fuel element designed to facilitate afork-type neutron absorber containing an AgeIneCd, as shown in
Fig. 1 (b). The control fuel element has 15 fuel plates to provide
space for the control rod. A total of three fuel element plates weretaken at each side of the fuel element, which made the nominal U235 loading for the control fuel element 178.57 g. The primary data
for the ﬁrst to the third core for RSG-GAS are shown in Table 1, andthe second core conﬁguration is shown in Fig. 2.
Beryllium is used as a reﬂector material surrounding the RSG-GAS
core. Two rows of beryllium elements enclose the two sides of thereactor core to provide ﬂexibility in adding desired irradiation kit. The
beryllium elements consist of a lower end ﬁtting, a rectangularberyllium rod with a height of 683 mm, and an outer cross-sectional
dimension of 79 mm  75 mm. Several beryllium elements are provided with one oriﬁce with a diameter of 50 mm for the irradiatedsamples. The other two sides of the reactor core are connected to aberyllium reﬂector block specially designed to provide neutron beam

1.1. Transition core of RSG-GAS
The TWC core of the RSG-GAS is achieved after operating thereactor for 6 transition cores with a gradual increase in both thenumber of fuel elements and the reactor power [14]. The ﬁrst core
conﬁguration consists of 12 fuel and 6 control rods with amaximum operating power of 10.6 MW. The second core conﬁguration consists of 16 standard fuel elements, 6 control rods, and 58
beryllium blocks and achieves maximum power at 13.6 MW. The
third RSG-GAS core consists of 22 fuel elements and 8 control rods
with a maximum power of 18.19 MW.
The RSG-GAS fuel element is a material testing reactor (MTR)
type fuel element with a standard fuel element consisting of 21 fuelplates, as shown in Fig. 1(a). Each fuel plate consists of an AlMg2
frame and two cladding sheets with the same material enclosing
U3O8eAl fuel meat. U3O8eAl is a uranium oxide fuel dispersed in an

Fig. 1. Standard fuel element (a) and control element (b) layout (mm) [6].
1776

S. Pinem, W. Luthﬁ, P.H. Liem et al.

Nuclear Engineering and Technology 55 (2023) 1775e1782

tubes. A conﬁguration of 6 neutron beam tubes (S-1, S-2, S-3, S-4, S-5,
and S-6) mounted on a beryllium block reﬂector is shown in Fig. 3.
2. Experiment method
2.1. Excess reactivity
Excess reactivity was measured after fuel elements were loaded
and the reactor achieved criticality. This critical condition was thebasis for determining excess reactivity caused by loading fuel andreﬂector elements. The RSG-GAS second core was critical with 11
fuel elements and 6 control rods, while the control rod bank was
600 mm and the regulating rod was 462 mm. After reaching thecritical condition, excess reactivity measurement was carried out
using the compensation method with a reactivity meter and control rod bank. Loading additional fuel elements for excess reactivity
were added one by one with a total of 4 new fuel elements and 6
reﬂector elements at once. Reactivity measurement was conductedat each fuel loading and all six reﬂectors by measuring the control
rod displacement using the compensation method. Core conﬁguration after fuel loading for excess reactivity to RSG-GAS second
core is shown in Fig. 2.

Fig. 2. RSG-GAS second core conﬁguration [14].

2.2. Control rod worth
The control rod worth was determined using a reactivity meter.
Based on the point kinetic equation, the time-dependent reactivityrðtÞ is deﬁned as follows,rðtÞ ¼ b þ L

d½lnPðtÞ   b
dt

∞
ð

DðtÞ

Pðt  tÞ
dt
PðtÞ

0

with

DðtÞ ¼

Xbj ljjbelj twhere PðtÞ is time-dependent reactor power, b is the effective
delayed neutron fraction, bj is the jth group of delayed neutron
fraction, lj is a decay constant of the jth group of delayed neutron
precursors, and L is the effective neutron generation time. All
quantities b, bj , lj , and L were input data that have been determined from reactor vendor design data [3], while PðtÞ was counted
as the signal amplitude taken from the compensated ionizationchamber (CIC) neutron detector (JKT04). The JKT04 is a detector
connected to a reactivity meter.
Initially, the reactor was operated at source-free low power
conditions so that it was free from neutron source effect and thereactivity feedbacks were negligible. The core temperature whenmeasuring the reactivity of the control rods was 28  C. The positionof the control rods characterized source-free conditions thatremain unchanged (the critical state) at different power levels.

Fig. 3. Beryllium block reﬂector and beam tube of the RSG-GAS reactor (unit in mm)[14].

Table 2Control rod calibration position data of second core at BOC.
Calibrated control rod

JDA-01/E-9
JDA-03/F-8
JDA-04/F-5
JDA-05/C-5
JDA-06/C-8
JDA-07/D-4

Control rod position (mm)Number of steps

JDA01
/E-9

JDA03
/F-8

JDA04
/F-5

JDA05
/C-5

JDA06
/C-8

JDA07
/D-4

0e600
289e206
290e184279e221323e239277e223

320e5250e600
290e184279e221323e239277e223

320e525289e2060e600
279e221323e239277e223

320e525289e206
290e1840e600
323e239277e223

320e525289e206
290e184279e2210e600
277e223

320e525289e206
290e184279e221323e239
0e600

1777

20
24
28
19
23
18

Table 3
Control rod calibration position data of second core at EOC.
Calibrated control rod

Control rod position (mm)JDA-01/E-9
JDA-03/F-8
JDA-04/F-5
JDA-05/C-5
JDA-06/C-8
JDA-07/D-4

Number of steps

JDA01/E-9

JDA03/F-8

JDA04/F-5

JDA05/C-5

JDA06/C-8

JDA07/D-4

0e600
378e290378e279362e298386e285358e299366e2940e600
378e279362e298386e285358e299366e294378e2900e600
362e298386e285358e299366e294378e290378e2790e600
386e285358e299366e294378e290378e279362e2980e600
358e299366e294378e290378e279362e298386e2850e600

measurement was carried out by removing each fuel from the core.
The control rod was then positioned to bring the reactor back to
critical condition; this position was then recorded. After the measurement is complete, the corresponding fuel is returned to its
original position, so the measured fuel element is the only fueloutside the core while the measurement incur. The difference incontrol rod position determined the fuel element reactivity worth
when the measured fuel was in and outside the core. The control
rod position corresponded to fuel reactivity measurement at the
BOC and EOC of the second core conﬁguration, shown in Tables 4
and 5. The reactivity of each fuel element was determined byﬁtting the control rod position to the previously measured integral
reactivity curve of the control rod.

Table 4Control rod position when measuring fuel element reactivity at BOC.
Fuel element

Core position

Full core
RI-13-out
RI-14-out
RI-15-out
RI-18-out
RI-19-out
RI-24-out
RI-25-out
RI-27-out

e
C-9
F-9
D-8
C-4
E6
D-9
D-7
F-4

Control rod position (mm)Bank

Regulating rod

253
286
286
286
286
347
286
345
286

253
237
286
543
242
347
321
345
278

20
21
25
18
21
16

During the control rod calibration, the control rod banks were in
critical positions while the single control rod to be calibrated waswithdrawn. After waiting for about 10 s to ensure the transient
event caused by the calibrated control rod insertion was done, thebank rods were changed to compensate for the positive reactivity
and brought the reactor to a critical state again. Each step of control
rod calibration produced a reactivity value of about 10e20 cents.
This procedure was repeated until the whole part of the control rod
was calibrated. The accumulation of measured reactivity for eachstep of control rod compensation is the control rod's worth. Control
rod positions and their corresponding bank rod positions areshown in Tables 2 and 3 for the second core conﬁguration at the
beginning of cycle (BOC) and end of cycle (EOC), respectively.

3. Calculation method
The ﬁrst core of the RSG-GAS model from the previous study
was used [10]. The depletion calculation by Serpent 2 code with
ENDF/B-VIII.0 nuclear data was carried out using 50,000 neutron
histories per cycle, and the total number of cycles was 200 with 50
inactive cycles, providing the standard deviation of the effective
multiplication factor (keff) about 30e40 pcm. In addition, for thecontrol rod worth calculations, the neutron histories per cycle wereincreased to 400,000 neutrons, and the total number of cycles is
500 with 100 inactive cycles to reduce the statistical uncertainty.
The thermal scattering libraries S(a,b) for hydrogen in the lightwater and beryllium as metal were included in the calculations.
Each fuel element was divided into ﬁve axial zones, and the material for each zone was deﬁned separately. In total, 90 depletion
zones were considered. The Serpent core model shown in Fig. 4.

2.3. Fuel element reactivity
Measuring reactivity of fuel element was also done while thereactor was operated at the source-free low power. The reactivity
Table 5Control rod position when measuring fuel element reactivity at EOC.
Fuel element

Full core
RI-12-out
RI-13-out
RI-14-out
RI-15-out
RI-16-out
RI-17-out
RI-18-out
RI-19-out
RI-20-out
RI-21-out
RI-22-out
RI-23-out
RI-24-out
RI-25-out
RI-26-out
RI-27-outCore position

e
D-5
C-9
F-9
D-8
C-7
C-6
C-4
E6
F-7
E5
F-6
E4
D-9
D-7
E8
F-4

Control rod position (mm)Bank

Regulating rod

321
391
365
362
385
380
385
353
433
409
385
396
362
365
430
400
360

323
391
363
366
385
390
395
352
429
406
392
394
360
361
418
401
357

Fig. 4. Second working core conﬁguration of RSG-GAS.
1778

S. Pinem, W. Luthﬁ, P.H. Liem et al.

Nuclear Engineering and Technology 55 (2023) 1775e1782

Table 6
Measured and calculated excess reactivity and control rod worth.
Parameters

2nd Core

3rd core

Experiment

Serpent 2Experiment

Serpent 2

9.48 ± 0.02
10.03 ± 0.001
9.31 ± 0.01
9.57 ± 0.01
Excess
(5.8%)(2.7%)
reactivity(%)
Control rod 18.12 ± 0.043 17.77 ± 0.009 20.10 ± 0.040 19.16 ± 0.01
worth (%)
(1.9%)(4.7%)

Table 7
Control rod reactivity worth at BOC.
Control Rod

Reactivity control rod at BOC condition ($)
Experiment

Serpent 2

JDA01
JDA03
JDA04
JDA05
JDA06
JDA07

3.660 ± 0.29
4.475 ± 0.32
5.328 ± 0.34
3.345 ± 0.29
4.315 ± 0.31
3.170 ± 0.28

3.665 ± 0.14 (0.13%)
4.466 ± 0.21 (0.21%)
5.357 ± 0.55 (0.55%)
3.374 ± 0.86 (0.87%)
4.358 ± 0.99 (0.99%)
3.171 ± 0.02 (0.02%)
Fig. 6. JDA03 integral control rod worth curve at BOC.

4. Result and discussion

found for the third core, i.e., around 4.7%. Considering the measurement errors and the rod interaction effects, the Serpent 2
values agreed with the experiment data.

4.1. Core excess reactivity
Excess reactivity at BOC is evaluated by calculating the keff. The
calculated result of the excess reactivity and control rods worth of
the second core and third core are shown in Table 6. The deviation
between Serpent 2 and experiment data is formulated as follows,

4.2. Control rod worth
The measured and calculated results of the control rod worth forthe second core at BOC are shown in Table 7. The difference between a calculation and measurement peaked at 0.99% for the
JDA06 rod. These results indicate that the control rod worth basedon Serpent 2 calculations is within good agreement with theexperiment.
The calculated and measured integral control rod worth curve (Scurve) at BOC for the second core is shown in Figs. 5e10. The
compensation method explained before determined the integral
control rod curve by calculating the reactivity derived from the keffvalues corresponding to the control rod positions. The Serpent 2

Serpent 2
 1  100%Experiment

The deviation in excess reactivity between calculated andexperiment values for the second core was 5.8% (510.97 pcm), andthe third core was 2.7% (253.23 pcm). The calculated values are
considered in good agreement with the experiment data since thecalculations also involved the whole core burnup calculations of theﬁrst and second cores. For the control rod worth, the Serpent 2
values show an underestimation slightly. The highest deviation wasFig. 5. JDA01 integral control rod worth curve at BOC.

Fig. 7. JDA04 integral control rod worth curve at BOC.
1779

S. Pinem, W. Luthﬁ, P.H. Liem et al.

Nuclear Engineering and Technology 55 (2023) 1775e1782

Fig. 8. JDA05 integral control rod worth curve at BOC.

Fig. 10. JDA07 integral control rod worth curve at BOC.

calculated results of integral control rod worth were again within
the range of measured data.
At the EOC of the second core, the control rod calibration was
also carried out. The measured and calculated control rod worth at
EOC are shown in Table 8. The maximum difference between thecalculated and measured data at EOC was 0.78%, comparable to the
BOC. The integral control rod worth of each control rod was also
calculated and shown in Figs. 11e16 with its corresponding
measured data. The most signiﬁcant difference between calculations and measured data occurred in the JDA07 rod but was still
within the measurement uncertainty range.

Table 8Control rod reactivity worth at EOC.
Control Rod

JDA01
JDA03
JDA04
JDA05
JDA06
JDA07

Reactivity control rod at EOC condition ($)
Experiment

Serpent 2

3.390 ± 0.29
4.050 ± 0.31
4.710 ± 0.32
3.270 ± 0.29
4.027 ± 0.31
2.920 ± 0.26

3.405 ± 0.44 (0.44%)
4.052 ± 0.04 (0.04%)
4.681 ± 0.61 (0.61%)
3.244 ± 0.78 (0.78%)
4.051 ± 0.58 (0.58%)
2.938 ± 0.60 (0.60%)

control rod curve (S curve). The control rod S curve was obtained
before the fuel reactivity measurement was carried out during theexperiment. The calculation of fuel element reactivity with Serpent
2 was carried out by simulating the control rod position as theexperiment occurred. Fuel element reactivity worth is then calculated according to the measured core reactivity (experimental) or

4.3. Fuel element reactivity
The fuel element reactivity worth is determined based on thecontrol rod position and compared to the control rod's integral

Fig. 9. JDA06 integral control rod worth curve at BOC.

Fig. 11. JDA01 integral control rod worth curve at EOC.
1780

S. Pinem, W. Luthﬁ, P.H. Liem et al.

Nuclear Engineering and Technology 55 (2023) 1775e1782

Fig. 12. JDA03 integral control rod worth curve at EOC.

Fig. 14. JDA05 integral control rod worth curve at EOC.

Fig. 13. JDA04 integral control rod worth curve at EOC.

Fig. 15. JDA06 integral control rod worth curve at EOC.

the keff (calculation) when the fuel element is inside and outside
the core.
The calculated results for fuel element reactivity at BOC and EOC
are shown in Tables 9 and 10. For the second core BOC, Serpent 2
results were very close to the measured data with a maximum
deviation of 1.82%. Meanwhile, the maximum deviation for thesecond core EOC was around 2.12%. The primary source of errorsthat might contribute to the uncertainty of the fuel reactivity
measurement was related to the control rod position uncertainty,
which was estimated to be ±1 mm. From the calibrated S curve of
all 6 control rods, a 1 mm change of control rod position equals areactivity value of about 4.09 cents.

control rod worth, and fuel element reactivity have been carriedout using the continuous energy Monte Carlo Serpent 2 code with
the ENDF/B-VIII.0 nuclear data. The difference between the calculation and measurement of the excess reactivity for the RSG-GAS
second and third cores was around 672 pcm and 511 pcm,
respectively. The calculated integral control rod reactivity curves
obtained by the bank rod compensation method were within theuncertainty range of measured data. The maximum difference incontrol rod worth at the RSG-GAS second core BOC and EOC is lessthan 1%. The calculated fuel element reactivity showed a maximum
deviation of around 2% compared to the measured data. Overall, it
can be concluded that the calculated neutronics parameters agree
well with the measured commissioning data accumulated for the
RSG-GAS second core.

5. Conclusions
Neutronics parameter analysis is essential to ensure a safe
reactor operation. The analyses related to core excess reactivity,
1781

S. Pinem, W. Luthﬁ, P.H. Liem et al.

Nuclear Engineering and Technology 55 (2023) 1775e1782

Curation, Visualization, Writing Original Draft, Review & Editing.
Peng Hong Liem: Conceptualization, Software, Writing Original
Draft, Review & Editing. Donny Hartanto: Data Curation, Methodology, Review & Editing.
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.
Acknowledgments
The authors are very grateful to the RSG-GAS multipurpose
reactor staff members for conducting the experiments. This
research was supported by the Indonesian government through
DIPA for the 2021 ﬁscal year funding number SP DIPA080.01.1.450310/2021.
References[1] Q.B. Do, G.T. Phan, K.C. Nguyen, Q.H. Ngo, H.N. Tran, Criticality and rod worth
analysis of the DNRR research reactor using the SRAC and MCNP5 codes, Nucl.
Eng.
Des.
343
(2019)
197e209,https://doi.org/10.1016/
j.nucengdes.2019.01.011.
[2] P. Kaviani, J. Mokhtari, S.M. Mirvakili, Integral form of the control rod calibration curve in the new core conﬁguration of HWZPR using rod insertion
method, Prog. Nucl. Energy 125 (2020), 103375, https://doi.org/10.1016/
j.pnucene.2020.103375.
[3] PRSG, Report of the Operation of RSG-GAS Reactor First Core, vol. 60, National
Nuclear Energy Agency of Indonesia, Serpong, Indonesia, 1988.
[4] L.P. Hong, Monte Carlo calculations on the ﬁrst criticality of the multipurpose
reactor GA Siwabessy, At. Indones. 24 (2) (1998).
[5] T.M. Sembiring, S. Pinem, Validation of the Monte Carlo code MVP on the ﬁrst
criticality of Indonesian multipurpose reactor, in: Proceedings of International
Conference on Research Reactors, Safe management and effective utilization,
Sydney, Australia, 2007, pp. 5e9 (Nov).
[6] P.H. Liem, T.M. Sembiring, Benchmarking the new JENDL-4.0 library on criticality experiments of a research reactor with oxide LEU (20 w/o) fuel, lightwater moderator and beryllium reﬂectors, Ann. Nucl. Energy 44 (2012)
58e64, https://doi.org/10.1016/j.anucene.2012.01.017.
[7] T.M. Sembiring, P.H. Liem, Accuracy of the ENDF/B-VII. 0 nuclear data libraryon the ﬁrst criticality experiments of the Indonesian multipurpose reactor RSG
GAS, in: Proceedings of the 2013 International Conference on Nuclear Data for
Science and Technology, 2013, pp. 4e8. ND2013.
[8] P.H. Liem, T. Surbakti, D. Hartanto, Kinetics parameters evaluation on the ﬁrst
core of the RSG GAS (MPR-30) using continuous energy Monte Carlo method,
Prog. Nucl. Energy 109 (2018) 196e203, https://doi.org/10.1016/
j.pnucene.2018.08.014.
[9] P.H. Liem, D. Hartanto, Sensitivity and uncertainty analysis on the ﬁrst core
criticality of the RSG GAS multipurpose research reactor, Prog. Nucl. Energy
114 (2019) 46e60, https://doi.org/10.1016/j.pnucene.2019.03.001.
[10] D. Hartanto, P.H. Liem, Analysis of the ﬁrst core of the Indonesian multipurpose research reactor RSG-GAS using the Serpent Monte Carlo code and the
ENDF/B-VIII. 0 nuclear data library, Nucl. Eng. Technol. 52 (12) (2020)
2725e2732, https://doi.org/10.1016/j.net.2020.05.027.
€nen, M. Pusa, T. Viitanen, V. Valtavirta, T. Kaltiaisenaho, The Serpent[11] J. Leppa
Monte Carlo code: status, development and applications in 2013, Ann. Nucl.
Energy 82 (2015) 142e150, https://doi.org/10.1016/j.anucene.2014.08.024.
€nen, S. Sikorin,[12] Alberto Talamo, Yan Cao, Yousry Gohar, V. Valtavirta, J. Leppa
S. Mandzik, S. Polazau, T. Hryharovich, Serpent transient analyses of GIACINT
geometrical change experiments, Ann. Nucl. Energy 64 (2021) 1e30, https://doi.org/10.1016/j.anucene.2021.108601.
[13] D.A. Brown, M.B. Chadwick, R. Capote, A.C. Kahler, A. Trkov, M.W. Herman, etal., ENDF/B-VIII. 0: the 8th major release of the nuclear reaction data librarywith CIELO-project cross sections, new standards and thermal scattering data,
Nucl.
Data
Sheets148(2018)
1e142,
https://doi.org/10.1016/
j.nds.2018.02.001.
[14] U. Jujuratisbela, B. Arbie, S. Pinem, L. Suparlina, O.P. Singh, Kinetics parametermeasurements on RSG-GAS, a low-enriched fuel reactor, in: Proceedings of
XIV International Conference on Reduced Enrichment for Research and Test
Reactor, 1991. Jakarta, Indonesia, 4-7 Nov, https://inis.iaea.org/collection/
NCLCollectionStore/_Public/35/047/35047142.pdf.

Fig. 16. JDA07 integral control rod worth curve at EOC.

Table 9Fuel element reactivity worth at BOC.
Fuel element

Core position

Reactivity (cent)Experiment

Serpent 2

RI-13
RI-14
RI-15
RI-18
RI-19
RI-24
RI-25
RI-27

C-9
F-9
D-8
C-4
E-6
D-9
D-7
F-4

161.85219.50429.10167.67579.19248.26568.36207.35160.35 ± 2.27
218.90 ± 1.95
421.30 ± 2.01
166.20 ± 1.60
573.81 ± 2.05
249.09 ± 2.20
563.20 ± 1.61
206.62 ± 1.72

Different (%)

0.92
0.27
1.82
0.88
0.93
0.33
0.91
0.35

Table 10
Fuel element reactivity worth at EOC.
Fuel element

RI-12
RI-13
RI-14
RI-15
RI-16
RI-17
RI-18
RI-19
RI-20
RI-21
RI-22
RI-23
RI-24
RI-25
RI-26
RI-27

Core position

D-5
C-9
F-9
D-8
C-7
C-6
C-4
E-6
F-7
E-5
F-6
E-4
D-9
D-7
E-8
F-4

Reactivity (cent)Experiment

Serpent 2

363.11
175.24226.89335.55320.89344.26194.49532.04438.82341.68383.79221.16234.54514.84403.86209.90357.14 ± 1.87
173.08 ± 1.73
223.98 ± 1.98
330.26 ± 1.65
316.23 ± 1.89
339.05 ± 1.99
192.23 ± 1.69
520.76 ± 2.02
430.65 ± 2.21
336.45 ± 2.20
377.21 ± 1.60
218.21 ± 1.95
231.29 ± 1.60
503.92 ± 2.22
396.84 ± 1.69
207.12 ± 1.82

Different (%)

1.64
1.23
1.28
1.58
1.45
1.51
1.16
2.12
1.86
1.53
1.71
1.34
1.39
2.12
1.74
1.32

Credit authorship contribution statement
Surian Pinem: Conceptualization, Validation, Formal Analysis,
Writing Original Draft, Review & Editing. Wahid Luthﬁ: Data

1782