Annals of Nuclear Energy 154 (2021) 108107

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Annals of Nuclear Energy
journal homepage: www.elsevier.com/locate/anucene

Analysis of the excess reactivity and control rod worth of RSG-GAS
equilibrium silicide core using Continuous-Energy Monte Carlo Serpent2
code
Tagor Malem Sembiring a,⇑, Surian Pinem b, Donny Hartanto c,d, Peng Hong Liem e,fa

Center for Nuclear Energy Assessment, National Nuclear Energy Agency of Indonesia, BATAN, Jl. Kuningan Barat, Jakarta, Indonesia
Center for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency of Indonesia, BATAN, Building No. 80, PUSPIPTEK, Tangerang Selatan, Banten, Indonesia
Department of Mechanical and Nuclear Engineering, University of Sharjah, P.O. BOX 27272, Sharjah, United Arab EmiratesdNuclear Energy System Simulation and Safety Research Group, Research Institute of Sciences and Engineering, University of Sharjah, P.O. BOX 27272, Sharjah, United Arab Emiratese
Cooperative Major in Nuclear Energy, Graduate School of Engineering, Tokyo City University (TCU), 1-28-1, Tamazutsumi, Setagaya, Tokyo, Japanf
Scientific Computational Division, Nippon Advanced Information Service, (NAIS Co., Inc.), 416 Muramatsu, Tokaimura, Ibaraki, Japanbca r t i c l e

i n f oArticle history:
Received 19 October 2020
Received in revised form 17 December 2020
Accepted 24 December 2020
Available online xxxx
Keywords:
RSG-GAS
Serpent2
Equilibrium core
Excess reactivity
Control rod worth
Beryllium reflectora b s t r a c t
This present work is a continuation of our previous work to verify the RSG-GAS equilibrium silicide coreparameters. Earlier, the burnup of 22 fuel elements in the 88th equilibrium core has been measured, andit showed a good agreement between the experimental and calculation results. In this paper, several
important neutronics parameters documented from the experiment at the same core, such as the excess
reactivity, integral control rod worth, and total control rod worth, are confirmed using the full core calculation results of Monte Carlo Serpent2 code in conjunction with ENDF/B-VII.1 and the most recent
ENDF/B-VIII.0 nuclear data libraries. The excess reactivity and integral control rod worth were measured
by the positive and negative reactivity compensation method. It is worth mentioning that the measured
integral control rod worth hasn’t been verified because the in-core fuel management code of RSG-GAS(BATAN-FUEL) solves only for a 2-dimensional configuration. The fuel element burnup at the end of cycle
calculated by BATAN-FUEL and Serpent2 is also compared and discussed. The 3-D Serpent2 model of RSGGAS first core is modified, and several improvements in the models are proposed, including the explicit
modeling of 6 neutron beam tubes in the beryllium block reflector. It is found that the calculated integral
control rod worth has a good agreement with the experiment, while Serpent2 overestimates the excess
reactivity by a maximum of 661 pcm. The impact of nuclear data is insignificant for the selected coreparameters.
Ó 2020 Elsevier Ltd. All rights reserved.

1. Introduction
RSG-GAS (Reaktor Serba Guna – G.A. Siwabessy) reactor, formerly named MPR-30, is an open pool material testing reactor witha nominal thermal power of 30 MW. It is located at the Puspiptek
Complex, Setu, Tangerang Selatan, Indonesia. Since the first criticality in July 1987, the reactor has been operated through 5 transition cores and 95 full core configurations. The first fullconfiguration was reached at the 6th core configuration using 40
standard fuel elements (FE) and 8 control fuel elements (CE). The
FEs and CEs are arranged in the 10  10 core grid positions. Thereare 21 fuel plates in a FE with low enriched uranium fuel meat, and⇑ Corresponding author.
E-mail addresses: tagorms@batan.go.id (T.M. Sembiring), pinem@batan.go.id (S.
Pinem), dhartanto@sharjah.ac.ae (D. Hartanto), liemph@nais.ne.jp (P.H. Liem).
https://doi.org/10.1016/j.anucene.2020.108107
0306-4549/Ó 2020 Elsevier Ltd. All rights reserved.

the total mass of U-235 is 250 g per element. The 15 fuel plates and2 Ag-In-Cd blades (one on each side) are assembled in a CE, and thetotal mass of U-235 is 178.6 g. The detailed geometry and data
specification of the FE and CE can be found in several references(BATAN, 1997; Liem et al., 1998; Pinem et al., 2016).
The reactor reached a typical working core that can be operated
at the nominal power of 30 MWth at the 6th core configuration. For
the typical working core, there is no fixed number of fresh and discharged FEs/CEs as well as the reshuffling pattern. The number offresh FEs/CEs in the RSG-GAS’ typical working core could be 6/1
or 6/2, and their locations in the core grid could be different
between each typical working core. Later, during the core conversion from oxide fuel (U3O8-Al) to the silicide fuel (U3Si2-Al), a new
in-core fuel management strategy was introduced to achieve theequilibrium core. The research on applying the new in-core fuelmanagement strategy has been carried out for the mixed core of

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Annals of Nuclear Energy 154 (2021) 108107

the oxide and silicide fuels to obtain an equilibrium silicide core by
the neutron diffusion method code (Liem et al., 1998; Sembiring
et al., 2001).
The equilibrium core means that all fresh, discharged, andreshuffled FEs and CEs have a predetermined loading, discharging,
and reshuffling pattern (Liem et al., 1998). It is usually establishedin the core fuel management of a power reactor because the number of fuel assemblies being handled can be hundreds. A research
reactor core is usually operated with a typical working core rather
than under the equilibrium core condition because it operates atlower power and has fewer fuel elements than the power reactor.
However, the in-core fuel management strategy for an equilibrium
core minimizes the human errors in the loading, unloading, andshuffling of the FEs and CEs since the pattern is fixed. Moreover,
the refueling time between each cycle is constant, and hence thebeginning of the next cycle can be determined more precisely.
In this present work, the important core parameters measured
in the equilibrium core of RSG-GAS, such as excess reactivity andcontrol rod worth, are evaluated using the analytical tools. Thereis a strong motivation to perform this work because the measured
core parameters of the equilibrium core haven’t been verified adequately analytically. Previous researches focused on the neutronics
parameters evaluation of RSG-GAS first-core which used only fresh
fuels (Liem and Sembiring, 2012; Liem et al., 2018, 2019, 2020;
2020;; Hartanto et al., 2020). The RSG-GAS equilibrium coreselected in this work is the 88th core in which the experiment
for excess reactivity and each control rod worth was also performed. The fuel elements’ burnup of the 88th core have been measured and compared with the analytical results using a linear
relationship between reactivity and burnup (Pinem et al., 2016).
The measurements were carried out for 22 irradiated FEs with aburnup level in the range of 20.85%-46.55% (loss of U-235). The
evaluation showed a good agreement between the measurement
and calculated burnup with the maximum error of 8%. Based onthese results, the neutronic parameters of the 88th core of RSGGAS are verified in the present work by the 3-dimension (3-D)
continuous-energy Monte Carlo Method Serpent2 code
(Leppänen et al., 2015).
This work also provides a meaningful discussion on a research
reactor using beryllium (Be) as a reflector element. RSG-GAS has
37 beryllium reflector elements in addition to the L-shaped beryllium block reflectors at the two sides of the core periphery. The
core also has 8 thermal neutron flux traps called irradiation position (IP): 4 IPs (2  2 grid) are located at the center of the core (central irradiation position, CIP), and the other 4 IPs (1  1 grid) are inamong FEs and CEs. Therefore, the core has a high heterogeneity or
various gradient of neutron flux. Hence, the Monte Carlo method
code, such as Serpent2, is used for this work.
Additionally, the detailed geometry model of RSG-GAS core isrevised in this work since the beam tubes were homogenized with
the Be block reflector previously (Liem and Sembiring, 2012; Liem
et al., 2018, 2019; 2020;; Hartanto et al., 2020). The effect of the
impurity of Be reflector is also considered. The experimental coreparameters, such as excess reactivity and control rod worth, arecompared with the calculated Serpent2 code. The calculated burnup swing, BOC (beginning of cycle) to EOC (end of cycle), of each
FE and CE is validated by the declared burnup. The effect of therecent ENDF/B nuclear data, ENDF/B-VIII.0, is also evaluated(Brown et al., 2018).
The rest of this paper is organized as follows: Section 2 discusses the description of the RSG-GAS 88th equilibrium core. Section 3 deals with the methodology adopted in the calculations.
Section 4 discusses the results, and finally, Section 5 presents theconclusions and recommendations for future works.

2. Equilibrium core (Core 88th) of RSG-GAS reactor
Fig. 1 shows the 88th equilibrium core configuration of the RSGGAS. Each FE and CE has a unique identification number, e.g., RI570 is fresh FE at position H-9 of the core grid. Table 1 shows
the detailed information of the declared burnup (% loss of U-235)
for all FEs and CEs taken from the RSG-GAS in-core fuel management code’s calculation results. Among them, there are 5 fresh
FEs and 1 fresh CE. As mentioned previously, there are 4 locations(at grid D-6, D-7, E-6, and E-7) for CIP, as well as 4 locations (at grid
B-6, D-9, E-4, and G-7) for the IPs (irradiation position). The CIP and
IPs have been utilized for the radioisotope (RI) production which
requires a high thermal neutron flux. At the nominal power of 30
MWth, the average thermal neutron flux at CIP and IPs are about
2  1014 neutron cm 2 s 1.
The CEs are divided into 8 burnup classes at grid locations of B7, C-5, C-8, D-4, E-9, F-5, F-8, and G-6. Therefore, the control rod
worth depends on its burnup class and the burnup distributionof FEs around the CE. The control rod worth of each position byexperiment and calculations is presented in this paper.
RSG-GAS core has two types of reflector element: the beryllium
reflector elements (B, at 29 grid locations) and the beryllium reflector elements with stopper (BS, at 7 grid locations). The BS can beused for target irradiation since it has a hole with a diameter andlength of 5 cm and 60 cm, respectively. The hole is plugged witha stopper if the BS is not utilized.
Fig. 2 shows the configuration of 6 neutron beam tubes (S-1, S2, S-3, S-4, S-5, and S-6) installed at the beryllium block reflector,
and Table 2 shows the important data of the beam tubes. The neutron flux in the core, especially near the beryllium block, is influenced by the condition of beam tubes, e.g., filled with water orair. This condition also affects the burnup of FEs next to the beryllium block. Therefore, all beam tubes are modeled in detail basedon the design data (BATAN, 1997).

3. Methodology
3.1. RSG-GAS 3-D core model
The Serpent core model of RSG-GAS from the previous study(Hartanto and Liem, 2020) was modified to accommodate the fuelcompositions at the BOC of the 88th core, calculated by the in-corefuel management code and documented in the operation report(PRSG-BATAN, 2015). The burnup of 22 FEs at BOC is also adjusted
by considering the calculation to the measurement ratio reported
in the previous work (Pinem et al., 2016).
It is found at the beginning of research work that the calculated
excess reactivity at BOC is higher than experimental value using
previous Serpent core Model (Hartanto and Liem, 2020). Therefore,
to improve the calculation of Serpent2 code, several importantrefinements related to beryllium reflector element and beryllium
block reflector are implemented into the core model, such as:
1. The six beam tubes, which are filled with water, are included
in detail in the reflector block region to improve the calculation
accuracy. In the previous work, the reflector block consisted of a
mixture of homogenized water and beryllium. It is expected thatthe composition of the reflector block affects the multiplication
factor and the neutron flux nearby the reflector block.
2. The chemical impurity is added to the material compositionof the beryllium reflector. Based on the fabricant data, the maximum impurity of the beryllium reflector element and beryllium
block reflector is around 10.113 ppm boron equivalent (BATAN,
1997).

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Tagor Malem Sembiring, S. Pinem, D. Hartanto et al.

Fig. 1. The 88th core configuration of the RSG-GAS reactor.

Table 1
Declared burnup values of the irradiated FEs and CEs for the 88th equilibrium core of RSG-GAS reactor.
No

IdentificationNumber

Grid location

Declared burnup (% loss of U-235)

No

Identification Number

Grid location

Declared burnup (% loss of U-235)

1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21

RI-523
RI-524
RI-525
RI-526
RI-527
RI-528
RI-529
RI-530
RI-531
RI-532
RI-533
RI-534
RI-535
RI-536
RI-537
RI-538
RI-539
RI-540
RI-541
RI-542
RI-543

B-8
D-8
B-5
F-6
G-8
C-7
C-9
A-8
F-4
G-5
G-4
B-9
E-5
F-7
B-7
G-6
E-9
D-4
C-6
D-10
A-6

45.72
44.42
46.55
46.56
47.85
39.34
38.58
41.77
40.64
42.30
33.51
32.91
34.43
33.39
51.50
45.34
38.64
31.62
35.32
27.13
26.64

22
23
24
25
26
27
28
29
30
31
32
33
34
35
36
37
38
39
40
41
42

RI-544
RI-545
RI-546
RI-547
RI-548
RI-549
RI-550
RI-551
RI-552
RI-553
RI-554
RI-555
RI-556
RI-557
RI-558
RI-559
RI-560
RI-561
RI-562
RI-563
RI-565

H-5
H-7
D-3
H-6
B-4
D-5
A-7
E-8
A-5
E-10
C-4
E-3
G-9
F-9
A-4
H-8
C-10
C-5
F-8
F-5
F-10

28.97
27.36
29.38
20.85
20.01
21.18
21.00
21.49
14.43
12.84
13.78
13.98
14.37
6.70
6.42
7.38
7.15
24.21
16.21
8.31
6.92

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Annals of Nuclear Energy 154 (2021) 108107

Fig. 2. The beam tubes configuration of RSG-GAS reactor (unit in mm).

Table 2
Utilization and position of the beam tubes.
Beam
Tube

Utilization

Position, cm (frombottom of active core)

Innerdiameter,
cm

S-1
S-2
S-3
S-4

I-125 loop
Neutron radiography–
Triple Axis Spectrometry(TAS)- Four Cycle Diffractometer (FCD)- High Resolution Powder
Diffractometer (HRPD)- Small Angle Neutron
Spectrometer (SANS)- High Resolution Small
Angle- Neutron
Spectrometer(HRSANS)
Powder Diffractometer (PD)

20
40
20
20

16
16
16
16

20

16

40

16

S-5

S-6

3. The depletion of beryllium in the reflector is considered. The
COUPLE/ORIGEN code in the SCALE6.2 package (Rearden andJessee, 2018) is used for the calculation of the beryllium composition after operating for more than 30 years (from the 1st to the87th core). In this calculation, 252 energy-group neutron spectrain the reflectors were tallied using Serpent2, and they were usedin the depletion calculation.

Fig. 3. Cross section view of the RSG-GAS Core at 20 cm from the bottom of active
core.

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3. The positive and negative compensation CRs are withdrawn andinserted, step by step. The maximum reactivity per step is ± 20
cents.
4. The positive compensation is carried out first, and then the negative compensation. This step is carried out continuously until
the positive compensation CR at 600 mm (fully up) and the negative compensation CR at 0 mm (fully down).
5. The accumulation of all reactivity steps from each compensated
CR is the control rod worth. Based on the reactivity meter, it isnoted that the error of reactivity is ± 1 cent/step.
The control rod worth evaluation was carried for all step positions of experiments, as shown in Table 4. The experimental
results, such as S-curve (integral rod worth curve) and the total
worth of each control rod, are compared with the calculation
results by Serpent2 with ENDF/B-VII.1 (Chadwick et al., 2011)and ENDF/B-VIII.0 nuclear data libraries. The ENDF/B-VIII.0 was
demonstrated to have a better agreement with experimental
results for the CR worth in the OPAL research reactor (Maul,
2018). In addition, the calculated total neutron flux distribution
in the core at the initial and final positions of each calibration caseis plotted to analyze its spatial effect on the CR worth.
The experimental excess reactivity can be determined by the Scurve of 8 CRs reactivity worth. Based on the criticality position of
all CRs bank position, the total positive and negative reactivities
can be determined from the S-curve. The experimental excess reactivity is the total positive reactivity which will be compared withthe calculation results in this study.

Fig. 4. Cross section view of the RSG-GAS Core at 40 cm from the bottom of active
core.

Figs. 3 and 4 depict the RSG GAS 3-D geometry models plottedby Serpent2 in the present work at 2 elevations to clearly illustrate
the addition of beam tubes into the model.

3.3. Burnup calculation
The burnup calculation to reach the EOC of the 88th core has
also been carried out in this work by following the one cycle operation history, such as control rod position, operation time, shutdown time, and reactor power. These data are collected from thelog-book of the reactor operation. The reactor operation datashows 69 different positions of control rod while operating from
BOC to EOC. The total energy produced in this cycle is about 630
MWd. In the burnup calculation by Serpent2, each FE and CE isdivided into 5 axial regions (12 cm height for each region). The
burnup of all axial regions in each FE/CE is later averaged and compared to the results calculated by the in-core fuel management
code of RSG-GAS, BATAN-FUEL code (Liem, 1996). The calculations
by Serpent2 are carried out using ENDF/B-VII.1 and the ENDF/BVIII.0 nuclear data libraries.

3.2. Control rod worth experiments and calculations
The control rod (CR) worth of the 88th core was measured by
the control rod calibration using the positive to negative reactivity
compensation method. In this method, one CR is chosen as a positive compensation rod, and its initial position is fully inserted into
the core (at 0 mm). Another CR is selected as a negative compensation rod, and it is fully withdrawn from the core initially (at600 mm). Meanwhile, the other 6 CRs are set at the bank position,
so the core is in the critical condition at low power. In each experiment, the worth of two CRs can be measured. Table 3 shows theexperimental condition for 4 cases of CR calibration in the 88th
core. The CR is identified by the code number, for example, JDA01 is the CR located at grid E-9. The detailed procedure of the positive to negative reactivity compensation method is explained by
the steps below:

4. Results and discussions

1. The core is set-up at low power with the criticality condition byadjusting one CR at 0 mm (fully down), one CR at 600 mm (fullyup), and 6 CRs at bank position. The CRs at 0 mm and 600 mm
are called positive and negative compensation CRs, respectively.
2. The compensated ionization chamber (CIC) detector (JKT04) isused to measure the positive and negative reactivities. The current of the JKT04 detector is 2.5  10-8 A or equal to 230 kW(low power). The JKT04 is connected to a reactivity meter.

The first results discussed is the impact of the refined modification into the neutron multiplication factor k of the core, including
the inclusion of the beam tubes into the Be block reflector, theaddition of chemical impurity in the Be reflector, and the depletionof the Be reflector after operating for 87 cycles. In Serpent2 simulation, all materials are at room temperature. The thermal scattering libraries of Be and H2O at room temperature are also used aswell as the Doppler broadening rejection correction (DBRC)

Table 3Control rod calibration cases in the 88th equilibrium core.
Case

1
2
3
4

Position of control rod, mm
JDA-01/ E-9

JDA-02/ G-6

JDA-03/ F-8

JDA-04/ F-5

JDA-05/ C-5

JDA-06/ C-8

JDA-07/ D-4

JDA-08/ B-7

0
274
260
277

262
0
260
277

262
274
600
277

262
274
260
0

600
274
260
277

262
600
260
277

262
274
0
277

262
274
260
600

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Annals of Nuclear Energy 154 (2021) 108107

Table 4
Step positions for the 4 CR worth experiment cases.
Case

Number of Step Position
Positive compensation

Negative compensation

1
2
3
4

13
14
14
14

14
11
14
10

Table 5
Effect of the refined modification to the core reactivity.
Modifications

ENDF/B-VII.1,
pcm

Inclusion of beam tubes
Addition of chemical impurity inBeDepletion of Be reflector
Total

ENDF/B-VIII.0,
pcm

155.04 ± 15.49
465.23 ± 15.44

146.61 ± 16.10
459.33 ± 15.44

192.00 ± 15.55
812.27 ± 26.84

158.25 ± 15.54
764.19 ± 27.19

Fig. 6. Calculated and experimental integral CR worth of JDA-02.
Table 6Comparison of excess reactivity between experiment and Serpent2 calculation.

Excess reactivity(%Dk/k)(C-E)/E*

ExperimentCalculation w/
ENDF/B-VII.1

Calculation w/
ENDF/B-VIII.0

7.305

7.937 ± 0.00011

7.966 ± 0.00011

8.7%

9.0%

* C = calculation result and E = experimental result.

method. The number of neutron histories per cycle is 100,000, andthe total number of cycles is 500 with 100 inactive cycles, resulting
in a k value with a standard deviation of less than 13 pcm.
As shown in Table 5, the highest impact on the reactivity is dueto the Be reflector’s chemical impurity, which provided negative
reactivity of 465.23 pcm. The explicit modeling of beam tubes
reduces the reactivity by about 155.04 pcm. A similar impact isalso noticed by taking into account the depletion of Be. Overall,these modifications decrease the total core reactivity by a maximum of 812.27 pcm, and hence it should be considered in the calculation. Meanwhile, it is shown that both ENDF/B-VII.1 and ENDF/
B-VIII.0 have a similar trend in the reactivity decrement.
Fig. 7. Calculated and experimental integral CR worth of JDA-03.

Fig. 5. Calculated and experimental integral CR worth of JDA-01.

Fig. 8. Calculated and experimental integral CR worth of JDA-04.
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Fig. 9. Calculated and experimental integral CR worth of JDA-05.

Fig. 11. Calculated and experimental integral CR worth of JDA-07.

Fig. 10. Calculated and experimental integral CR worth of JDA-06.

Fig. 12. Calculated and experimental integral CR worth of JDA-08.

Table 6 compares the excess reactivity between the experimental and calculated results for the 88th core of RSG-GAS at BOC (coldand Xe free condition). The corrections from Table 5 and the burnup (Pinem et al., 2016) are included in the calculations. The relative difference of the ENDF/B-VII.1 and ENDF/B-VIII.0 with theexperiment is about 8.7% and 9.0%, respectively. The differences
are equal to 632 pcm and 661 pcm of reactivity. It is consideredhigh compared to the maximum error of ½b, which is about 350
pcm. It is due to a lack of burnup distribution data at BOC of thiscore. Only 22 out of 35 irradiated fuel elements have the measured
burnup fraction data, while the remaining fuel elements are using
the burnup fraction data from the BATAN-FUEL code. It is expectedthat the burnup distribution may slightly differ depending on theoperation history, including the control rod insertion position.
Therefore, the multicycles core burnup analysis is required to geta more accurate burnup distribution.
The integral reactivity worth of each CR in the 88th core areshown in Figs. 5–12 and summarized in Table 7. For the control
rod worth calculation, the number of neutron histories per cycle

Table 7
Experimental and calculated total CR worth.
Control
Rod

Experiment(Cents)

ENDF/B-VII.1(Cents)

ENDF/B-VIII.0(Cents)JDA01

204.25 ± 12

JDA02224.75 ± 13

JDA03243.75 ± 14

JDA04241.10 ± 13

JDA05

234.50 ± 13

JDA06

182.75 ± 10

JDA07

242.95 ± 13

JDA08

181.30 ± 10

222.79 ± 6.63
(9.1%)*
237.00 ± 6.98
(5.4%)
243.01 ± 6.86
(0.3%)
240.61 ± 6.89
(0.2%)
241.99 ± 6.91
(3.2%)
203.37 ± 6.13
(11.3%)
249.33 ± 6.86
(2.6%)
196.53 ± 6.06
(8.4%)

211.16 ± 6.60
(3.4%)
239.57 ± 6.88
(6.6%)
240.68 ± 6.89
(1.3%)
254.84 ± 6.91
(5.7%)
233.07 ± 6.88
(0.6%)
195.51 ± 6.01
(7.0%)
244.73 ± 6.87
(0.7%)
205.16 ± 6.24
(13.2%)*(C-E)/E*100%.
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Annals of Nuclear Energy 154 (2021) 108107

Fig. 13. Total neutron flux distribution at the initial positions of JDA-02 (0 mm) and JDA-06 (600 mm).

Fig. 14. Total neutron flux distribution at the final positions of JDA-02 (600 mm) and JDA-06 (0 mm).

is increased to 200,000, and the total number of cycles is 600 with
100 inactive cycles, resulting in a k value with a standard deviationof less than 9 pcm. The smallest difference between calculations
and experiments for the integral CR worth is for JDA-03 (Fig. 7), followed by JDA-07 and JDA-05. It is also shown that the different
nuclear data have less impact on these 3 control rods. However,the largest difference between calculations and experiments forthe integral CR worth is shown in JDA-06 and JDA-08 (Figs. 10
and 12). Based on the Figs. 5–12, it is clear that the nuclear dataof ENDV/B-VII.1 and ENDV/B-VIII.0 give the difference result on
the integral CR worth of JDA-01, JDA-04, JDA-06 and JDA-08. In thispaper, the impacts of these libraries are expressed by the difference of the relative error of these libraries. Based on Table 7, therelatively higher impacts are in JDA-01, JDA-04, JDA-06 and JDA08 with the difference of 6%, 5.5%, 4.3% and 4.8%, respectively.
The rest are within 1% 2,6%. However, none of these nuclear data
can give the consistently relative error closer with experimental

results. For example, the calculated worth of JDA-04 using ENDF/
B-VII.1 is closer to experimental worth than the worth of ENDF/
B-VIII. On the other hand, the calculated worth of JDA-01 using
ENDF/B-VIII.0 is closer to the experimental worth. It should be
noted that the rod worth in dollars or cents is calculated by using
the adjoint-weighted total effective delayed neutron fractions calculated by Serpent2.
The total control rod worth of each CR is summarized in Table 7.
Compared to the experimental worth, the calculated results using
ENDF/B-VII.1 nuclear data provide an average relative error of 5.1%,
while a smaller difference is provided using ENDF/B-VIII.0 with theaverage relative error of 4.8%. It is consistent with the OPAL
research reactor results in which ENDF/B-VIII.0 was shown to havea better agreement (Maul, 2018).
Figs. 13–16 illustrate the total neutron flux distributions at
the initial and final conditions in each core grid during therod worth calibration, such as JDA-02 vs. JDA-06 (Figs. 13 and8

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Fig. 15. Total neutron flux distribution at the initial positions of JDA-03 (0 mm) and JDA-07 (600 mm).

Fig. 16. Total neutron flux distribution at the final positions of JDA-03 (600 mm) and JDA-07 (0 mm).

14) and pair of JDA-03 vs. JDA-07 (Figs. 15 and 16). For the pairof JDA-02 and JDA-06, the neutron flux is high at the bottom
part of the core at the initial condition (Fig. 13). However, it
shifts to the upper part of the core in the final condition(Fig. 14). Meanwhile, the high neutron flux in the pair of JDA03 and JDA-07 is at the right side of the core initially, but it
shifts only to the upper left part of the core in the final condition. It confirms that the control rod interaction in the RSG-GAS
is high, as analyzed in the previous work (Liem et al., 2002).
Additionally, it is found that the maximum absolute differenceof total neutron flux in each mesh grid between the twonuclear data is about 5.6%.
The burnup distribution at EOC is depicted in Fig. 17. Three calculated values for each FEs and CEs are shown; the first, second,

and third rows are the calculation results by Serpent2 with
ENDF/B-VII.1, Serpent2 with ENDF/B-VIII.0 and BATAN-FUEL code,
respectively. BATAN-FUEL results are similar to the Serpent2 with
the ENDF/B-VII.1 for the 2nd – 8th burnup classes. However, for the
1st burnup class of FEs at A-9, C-3F-3, H-4, and H-9 grid positions,the calculated burnup by Serpent2 code is higher than the BATANFUEL calculation results. This is due to the higher fission reaction
rates (cf. Fig. 18) by the Serpent2 at those grid positions higher
than the BATAN-FUEL calculation results, as shown by the higher
power peaking factor. However, for the 2nd – 8th burnup classesof FEs, the burnup rate in the Serpent2 is lower than in the
BATAN-FUEL code since the burnup change is not high comparedto the 1st burnup class, although the power peaking factor isslightly higher.
9Tagor Malem Sembiring, S. Pinem, D. Hartanto et al.

Annals of Nuclear Energy 154 (2021) 108107

Fig. 18. Calculated power peaking factor by Serpent2 code and BATAN-FUEL code.

Fig. 17. Calculated burnup distribution (% loss of U-235) by Serpent2 code and
BATAN-FUEL code at EOC.

analysis, Writing - original draft, Writing - review & editing. DonnyHartanto: Methodology, Software, Formal analysis, Data curation,
Visualization, Writing - review & editing. Peng Hong Liem: Conceptualization, Methodology, Writing - review & editing,
Supervision.

5. Conclusions
The experimental/measured core parameters of the RSG-GAS
88th equilibrium core, including the excess reactivity, integral control rod worth, and total control rod worth, have been confirmed
by using continuous energy Monte Carlo Serpent2 code with 2 different nuclear data libraries: ENDF/B-VII.1 and ENDF/B-VIII.0. A
more detailed configuration and accurate composition were usedin Serpent2, such as the explicit modeling of the neutron beamtubes, impurity in the Be reflector, and the Be reflector’s depletion.
It is found that the relative difference between the calculated andexperimental excess reactivity is about 9% or equivalent to 661
pcm, which is considered to be high. However, the calculated integral control rod worth provides a good agreement with the experimental results. The calculated total control rod worth also concurs
with the experimental results. The maximum relative difference
for the control rod worth is about 11.3% for JDA-06 by ENDF/BVII.1 and about 13% for JDA-08 by ENDF/B-VIII.0. Meanwhile, thecalculated burnup at EOC between BATAN-FUEL and Serpent2
agrees well, except for the 1st class burnup of FEs in which Serpent2 overestimates the burnup fraction by about 13%. This researchalso shows no significant impact of different nuclear data libraries,
ENDF/B-VII.1 and ENDF/B-VIII, on the selected equilibrium coreparameters such as the reactivity and the EOC burnup fraction.
However, the significant discrepancy of JDA-06 and JDA-08 control
rod worth by different nuclear data libraries deserves further
investigation. Sensitivity analysis will be performed for these twocases. Moreover, to further improve the accuracy, multicycles coreburnup analysis will be performed by considering the operation
history, including the control rod insertion position.

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.
Acknowledgments
This study has been partially supported by the Finance Ministryof Indonesia by fiscal year of 2020, DIPA PKSEN-BATAN 2020. The
authors are very grateful to staff members of the RSG GAS multipurpose reactor for conducting the experiments. The authors also
acknowledge the use of the HPC Facility at the University of Sharjah for part of the work reported in this paper.
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CRediT authorship contribution statement
Tagor Malem Sembiring: Conceptualization, Methodology, Validation, Formal analysis, Investigation, Writing - original draft. Surian Pinem: Conceptualization, Methodology, Validation, Formal
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