Modeling of the Dalat Nuclear Research Reactor (DNRR) with the Serpent 2 Monte Carlo code
Nuclear Science and Technology, Vol.9, No. 3 (2019), pp. 21-29
Modeling of the Dalat Nuclear Research Reactor (DNRR) with
the Serpent 2 Monte Carlo code
Nguyen Kien Cuong1*, Nguyen Thi Dung2, Tran Viet Phu2,
Nguyen Huu Tiep2, Pham Nhu Viet Ha2**
1Nuclear Research Institute (NRI), Vietnam Atomic Energy Institute (VINATOM),
01 Nguyen Tu Luc, Dalat, Lamdong 670000, Viet Nam
2Institute for Nuclear Science and Technology (INST), Vietnam Atomic Energy Institute (VINATOM),
179 Hoang Quoc Viet str., Cau Giay dist., Hanoi 100000, Viet Nam
(Received 17 September 2019, accepted 19 November 2019)
Abstract: This paper presents a model development of the Dalat Nuclear Research Reactor (DNRR)
loaded with low enriched uranium (LEU) fuel using the Serpent 2 Monte Carlo code. The purpose is
to prepare the DNRR Serpent 2 model for performing fuel burnup calculations of the DNRR as well
as for generating multi-group neutron cross sections to be further used in the kinetics calculations of
the DNRR with a 3D reactor kinetics code. The DNRR Serpent 2 model was verified through
comparing with the MCNP6 criticality calculations under different reactor conditions. The parameters
to be compared include the effective neutron multiplication factor, radial and axial power
distributions, and thermal neutron flux distributions. The comparative results generally show a good
agreement between Serpent 2 and MCNP6 and thus indicate that the DNRR Serpent 2 model can be
used for further calculations of the DNRR.
Keywords: DNRR, neutronics, Monte Carlo simulation, Serpent 2, MCNP6.
36%. In the framework of the program on
Russian Research Reactor Fuel Return
(RRRFR) and the program on Reduced
Enrichment for Research and Test Reactor
(RERTR), the DNRR core was partly
converted to low enriched uranium (LEU) fuel
with 235U enrichment of 19.75% in September
2007 and the full core conversion to LEU fuel
was performed during the period from
November 24th, 2011 to January 13th, 2012 [2].
Presently, the operation of DNRR is mostly
I. INTRODUCTION
The 250 kW TRIGA Mark-II reactor, a
pool-type multipurpose research reactor
designed and manufactured by General
Atomics [1], was installed in the Dalat Nuclear
Research Institute, Vietnam in early 1960s. It
was upgraded to 500 kW in early 1980s and
named the Dalat Nuclear Research Reactor
(DNRR). In this upgrade, the main structures
of the TRIGA Mark-II reactor were remained
whereas the reactor core was loaded with the
Russian VVR-M2 fuel type and cooled by
natural convection. The first criticality of the
DNRR was achieved on November 1st, 1983
and the operation with full power was achieved
in March 1984. During the period from 1984 to
2007, the core was loaded with highly enriched
uranium (HEU) fuel with 235U enrichment of
for
radioisotope
production,
neutron
activation analyses, education and researches.
The DNRR is operated mainly in a continuous
run of about 100-130 hrs in a period of 3-4
weeks and then the reactor is shut down for
maintenance
activities
during
which
experimental physics researches can also be
conducted [3].
©2019 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute
MODELING OF THE DALAT NUCLEAR RESEARCH REACTOR (DNRR) WITH…
The DNRR, i.e., a TRIGA Mark-II
reactor loaded with the Russian VVR-M2 fuel,
consists of various components such as a
neutron trap, water gaps, irradiation channels,
horizontal beam ports, beryllium blocks,
control rods, rotary specimen rack and graphite
reflector, which make the reactor geometry
become complicated and difficult to be
simulated with computer codes. In certain
applications such as in the field of in-core fuel
and MCNP6 in the cases corresponding to the
complete withdrawal and the full insertion of
the four shim rods and the automatic regulating
rod as well as under the criticality condition of
the core. In addition, power and thermal
neutron flux distributions were also examined
throughout the radial core and along several
typical axial channels.
II. CALCULATION METHODOLOGY
management,
deterministic
codes
like
The horizontal cross section view and
main specifications of the DNRR are shown in
Fig. 1 and Table I, respectively [2]. The reactor
core consists of hexagonal cells including fuel
bundles, control rods, in-core irradiation
channels, beryllium blocks, and a neutron trap.
The fuel bundle is the Russian VVR-M2 type
made of UO2-Al dispersion cladded in
aluminum. The total 235U mass in each LEU
fuel bundle is about 49.7 g distributed in three
coaxial fuel tubes as illustrated in Fig. 2. The
outer tube has a hexagonal shape and the two
inner tubes are cylindrical. The core is
controlled by seven control rods: two safety
rods, four shim rods and one automatic
regulating rod. The safety and shim rods are
made of boron carbide (B4C), while the
automatic regulating rod is made of stainless
steel. The absorption length of each control rod
is 65 cm, ensuring to completely cover the 60
cm long active core. The dry and wet channels
are covered by aluminum cylinders with the
thickness of 0.5 mm. The neutron trap located
at the core center is a water cylinder with 6.5
cm in diameter and 60 cm in length surrounded
by the beryllium blocks. The beryllium blocks
have the same outer shape and dimension with
the fuel bundle. At several peripheral cells, if
no fuel bundle is loaded, the beryllium blocks
are loaded for providing supplementary
neutron reflection. A ring of serrated beryllium
blocks is also located between the active core
and the graphite reflector to act as an additional
WIMSD/CITATION and SRAC can be used
for modeling and analysis of the DNRR
[4][5][6]. Nevertheless, Monte Carlo codes like
MCNP [5][6][7], Serpent 2 [8] are often
preferred to simulating complicated reactor
geometries like the DNRR. It is because an
arbitrary reactor geometry can be only
simulated without any approximation by using
a Monte Carlo code, which also allows using
continuous neutron energy cross sections,
eliminating the need for energy groups
collapsing as required with the deterministic
ones. In particular, spatial homogenization and
fuel burnup calculation can also be performed
with the Serpent 2 Monte Carlo code.
In the present study, a model of the
DNRR loaded with LEU fuel was developed
and analyzed using the Serpent 2 Monte Carlo
code. The goal is to prepare the DNRR model
with Serpent 2 for performing fuel burnup
calculations of the DNRR and for generating
multi-group neutron cross sections to be further
used in the kinetics analysis of the DNRR with
a 3D reactor kinetics code. The DNRR core
loaded with 92 LEU fuel bundles and its
surrounding structures were simulated with
high detail using Serpent 2. The DNRR
Serpent 2 model was then verified through
comparing with the MCNP6 criticality
calculations under different reactor conditions.
The effective neutron multiplication factor, keff,
was calculated and analyzed using Serpent 2
22
NGUYEN KIEN CUONG, PHAM NHU VIET HA et al.
reflector. This beryllium ring and the core are
The DNRR model with MCNP6 was
developed similarly to that reported with
MCNP5 in Ref. [5]. It is noted that the
DNRR model with MCNP had been well
validated against experiments and other
calculation results obtained with different
codes [2][5]. By using MCNP, the complex
geometry of the DNRR including the fuel
bundles, control rods, in-core irradiation
channels, beryllium rods, horizontal beam
tubes, graphite reflector, rotary specimen
rack, thermal column and thermalizing
column was simulated with high accuracy
and detail. This calculation model covers
from the active core to reactor tank following
the radial direction with 198.72 cm in
diameter and the axial direction with 187 cm
in height. Tallies 4 and 7 were used for
calculating the neutron flux and power
distributions, respectively. Then a DNRR
model was developed with Serpent 2 using
the same input data for the reactor geometry
and materials with the MCNP6 model as
described below.
placed in a cylindrical aluminum shell, which
is the lower section of the supporting structure.
The thickness of the graphite reflector is 30.5
cm. The core and the graphite reflector are
placed in the reactor pool.
In this work, the DNRR configuration
with 92 LEU fuel bundles (Fig. 2) was simulated
using the Serpent 2 Monte Carlo code [8]. The
DNRR model with Serpent 2 was then verified
through comparing with the MCNP6 criticality
calculations under different reactor conditions
including the critical condition and the cases of
complete withdrawal and full insertion of the four
shim rods and the regulating rod. The radial
layout of the DNRR core modeled in MCNP6
and Serpent 2 was shown in Fig. 3. The
ENDF/B-VII.0 data library [9] was used in both
the Serpent 2 and MCNP6 calculations. The
parameters to be calculated and compared
include the keff; radial power distributions; axial
power distributions in cell 4-5 and cell 7-10; and
axial thermal neutron flux distributions in cell 4-5,
cell 7-10 and neutron trap.
Fig. 1. Horizontal cross section view of the DNRR.
23
MODELING OF THE DALAT NUCLEAR RESEARCH REACTOR (DNRR) WITH…
Serpent
is
a
three-dimensional
Therefore, Serpent 2 has been widely used in
the calculations of nuclear reactors in general
and of TRIGA reactors in particular [10][11].
To simulate the DNRR geometry with high
detail in the Serpent 2 model, we adopted all
the data that were input to the DNRR MCNP6
model as above mentioned. The only
simplification is that the rounded corners of
the outer hexagonal tubes of the VVR-M2
fuel bundles were not modeled in the Serpent
2 simulation. It is worth noting that the delta
tracking method was used in the Serpent 2
calculations as compared to the track length
method used in the MCNP6 calculations.
continuous-energy Monte Carlo reactor
physics burnup calculation code developed at
the VTT Technical Research Centre of
Finland [8]. Similar to MCNP, the code
allows easily modeling of complicated reactor
geometries for criticality calculations, fuel
cycle studies, etc. Furthermore, it has also
various powerful capabilities such as
automated burnup sequence for spatial
homogenization,
calculations, transient simulations, sensitivity
calculations, reactor geometry pre-
implementation and fast running time.
coupled
multi-physics
Fig. 2. Radial layout of the VVR-M2 fuel bundle (left) and configuration of the DNRR core with 92 LEU bundles
(right). SR: safety rod, ShR: shim rod, and AR: automatic regulating rod.
Fig. 3. The DNRR core with 92 LEU bundles modeled in MCNP6 (left) and Serpent 2 (right).
24
NGUYEN KIEN CUONG, PHAM NHU VIET HA et al.
Table I. Main specifications of the DNRR.
Reactor type
Pool type
Nominal thermal power
Coolant and moderator
Core cooling mechanism
Reflector
500 kW
Light water
Natural convection
Graphite, beryllium and light water
Active core height
Core equivalent diameter
Fuel pitch
60 cm
44.2 cm
3.5 cm
VVR-M2 type, dispersed UO2-Al with 19.75% enrichment,
aluminium cladding
Fuel type
Number of control rods
7 (2 safety rods, 4 shim rods, 1 regulating rod)
B4C
Material of safety and shim rods
Material of automatic regulating rod
Vertical irradiation channels
Horizontal beam ports
Stainless steel
4 (1 neutron trap, 1 wet channel, 2 dry channels)
4 (1 tangential, 3 penetrant)
position of the regulating rod was adjusted
correspondingly to search for the criticality of
the core. It can be seen that there was a good
III. RESULTS AND DISCUSSION
A. Effective neutron multiplication factor
agreement within 71 pcm in the keff values
calculated using Serpent 2 and MCNP6. Also,
the statistic errors of the keff values in the
Serpent 2 and MCNP6 calculations were
almost identical to each other with the same
number of neutron history used. The difference
in the keff values of at most 71 pcm might be
mainly related to the two reasons as follows:
(1) the rounded corners of the outer hexagonal
tubes of the VVR-M2 fuel bundles were not
included in the Serpent 2 model, and (2)
different neutron tracking methods used in the
Serpent 2 (delta tracking method) and MCNP6
(track length method) calculations.
The keff values obtained by Serpent 2
and MCNP6 in the cases of (a) criticality
condition (four shim rods were inserted 40.5
cm in the core and the regulating rod was
inserted 40 cm in the core), (b) complete
withdrawal of the four shim rods and the
regulating rod from the core, and (c) full
insertion of the four shim rods and the
regulating rod in the core were shown in Table
II. In all the calculations, the two safety rods
were assumed completely withdrawn from the
core as they are used only for emergency
shutdown. In this investigation, the four shim
rods were inserted in the core and then the
Table II. Comparison of keff calculated using Serpent 2 and MCNP6.
Position of control rods (cm)
Four shim rods Regulating rod
keff
Serpent 2
MCNP6
Deviation
(pcm)
25
MODELING OF THE DALAT NUCLEAR RESEARCH REACTOR (DNRR) WITH…
Full insertion
Complete withdrawal Complete withdrawal
40.5 40
Full insertion
-37
71
0.97103 0.00007
1.07864 0.00007
1.00073 0.00007
0.97140 0.00006
1.07793 0.00005
1.00018 0.00006
55
radial power distributions calculated by
Serpent 2 and MCNP6 in the cases was 6.2%
(critical condition) and -5.7% (complete
withdrawal of control rods), which appeared
at the fuel bundles located in the periphery
of the core. The maximum value of the
relative radial power distribution was found
at cell 4-5 in both cases. This value was
1.424 under the critical condition and 1.374
in case of complete withdrawal of control
rods. Hence, the insertion of control rods
under the critical condition increased the
maximum value of the radial power
distribution 3.6%.
B. Power and thermal neutron flux
distributions
In this Section, power and thermal
neutron flux distributions were calculated
and analyzed using Serpent 2 in comparison
with the MCNP6 calculations. The two
cases including the criticality of the core
and complete withdrawal of the shim rods
and the regulating rod as shown in Table II
were considered. Fig. 4 displays the relative
radial power distributions calculated by
Serpent 2 and MCNP6. It can be seen that
there was a good agreement between the
two codes. The maximum difference in the
Fig. 4. Relative radial power distribution in the core under the critical condition (left) and in the case of
complete withdrawal of control rods (right). The upper and lower values were obtained with Serpent 2 and
MCNP6, respectively.
The axial power and thermal neutron
flux distributions in cell 4-5 and cell 7-10 were
shown in Figs. 5-6. Cell 4-5 is the position at
which the relative radial power was highest
and cell 7-10 is the position next to the
regulating rod, which can show clearly the
effect of control rod insertion. Fig. 5 shows a
good agreement between the axial power
distributions obtained with Serpent 2 and
MCNP6 in these two cells. The respective
maximum difference between the codes was
1.5% (cell 4-5) and 4.9% (cell 7-10). Similarly,
the axial thermal neutron flux distributions in
cell 4-5 and cell 7-10 calculated by Serpent 2
compared well with those calculated by
MCNP6 as shown in Fig. 6. The respective
maximum difference between the codes in
this case was 3.3% (cell 4-5) and -4.2% (cell
26
NGUYEN KIEN CUONG, PHAM NHU VIET HA et al.
7-10). Additionally, the axial thermal
in Fig. 7. Also, the maximum thermal
neutron flux in the neutron trap was found
consistent with that reported in the Safety
Analysis Report [2].
neutron flux distribution in the neutron trap
calculated by Serpent 2 agreed well within
2.8% with the MCNP6 calculations as shown
ShR = 40.5 cm, AR = 40 cm
2.0
1.8
1.6
1.4
1.2
1.0
0.8
0.6
0.4
0.2
0.0
Cell 4-5
Cell 7-10
-30
-20
-10
Core height (bottom to top) (cm)
MCNP6
0
10
20
30
SERPENT 2
Control rods out
2.0
1.8
1.6
1.4
1.2
1.0
0.8
0.6
0.4
0.2
0.0
Cell 4-5
Cell 7-10
-30
-20
-10
Core height (bottom to top) (cm)
MCNP6
0
10
20
30
SERPENT 2
Fig. 5. Axial power distribution along the fuel bundles 4-5 and 7-10 under the critical core condition (left)
and in the case of complete withdrawal of control rods (right).
ShR = 40.5 cm, AR = 40 cm
8.0E+12
Cell 4-5
7.0E+12
6.0E+12
Cell 7-10
5.0E+12
4.0E+12
3.0E+12
2.0E+12
1.0E+12
0.0E+00
-40
-30
-20
-10
Core height (bottom to top) (cm)
MCNP6
0
10
20
30
40
SERPENT 2
27
MODELING OF THE DALAT NUCLEAR RESEARCH REACTOR (DNRR) WITH…
Control rods out
8.0E+12
Cell 4-5
7.0E+12
6.0E+12
Cell 7-10
5.0E+12
4.0E+12
3.0E+12
2.0E+12
1.0E+12
0.0E+00
-40
-30
-20
-10
Core height (bottom to top) (cm)
MCNP6
0
10
20
30
40
SERPENT 2
Fig. 6. Thermal neutron flux distribution along the fuel bundles 4-5 and 7-10 under the critical core condition
(left) and in the case of complete withdrawal of control rods (right).
2.2E+13
2.0E+13
1.8E+13
1.6E+13
1.4E+13
1.2E+13
1.0E+13
8.0E+12
6.0E+12
4.0E+12
2.0E+12
0.0E+00
-30
-20
-10
0
10
20
30
Core height (bottom to top) (cm)
Rods in, MCNP6
Rods in, SERPENT 2
Rods out, SERPENT 2
Rods out, MCNP6
Fig. 7. Thermal neutron flux distribution along the neutron trap under the critical core condition (Rods in)
and in the case of complete withdrawal of control rods (Rods out).
IV. CONCLUSIONS
in the keff values obtained with Serpent 2 and
MCNP6. In addition, the comparison of power
and thermal neutron flux distributions in the
core generally shows a good agreement
between the two codes.
In this study, the Serpent 2 Monte Carlo
code was applied to model the DNRR. The
DNRR Serpent 2 model was verified through
comparing with the MCNP6 criticality
calculations under different reactor conditions.
These conditions include the critical condition
of the core, full insertion of the four shim rods
and the regulating rod in the core, and
complete withdrawal of the four shimd rods
and the regulating rod from the core. It is found
that there was a good agreement within 71 pcm
Consequently, the results indicate that
the DNRR model developed herein with
Serpent 2 is reliable and can be used for further
analyses of the DNRR. The DNRR Serpent 2
model is now being used for generating multi-
group neutron cross sections to be used in the
kinetics analysis of the DNRR with a 3D
28
NGUYEN KIEN CUONG, PHAM NHU VIET HA et al.
nuclear research reactor with HEU fuel using
reactor kinetics code. Also, this model with
Serpent 2 is being planned for performing fuel
burnup calculations of the DNRR.
SRAC and MCNP5,” Science and Technology
of Nuclear Installations, Article ID 2615409,
Volume 2017.
ACKNOWLEDGMENTS
[6]. Q. B. Do, G. T. 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,” Nuclear
Engineering and Design 343, 197-209, 2019.
This research is funded by Ministry of
Science and Technology of Vietnam under
grant number DTCB.06/18/VKHKTHN. The
visit of Pham Nhu Viet Ha in North Carolina
State University during this work was
supported by the IAEA TC Project VIE9016.
[7]. D. B. Pelowitz et al., MCNP6TM User’s
Manual, Version 1.0, LA-CP-13-00634, Rev.
0, 2013.
[8]. J. Leppänen, M. Pusa, T. Viitanen, V.
Valtavirta, and T. Kaltiaisenaho. “The Serpent
Monte Carlo code: Status, development and
applications in 2013,” Annals of Nuclear
Energy 82, 142-150, 2015.
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