The journal of Nuclear Science and Technology - Volume 9, Number 3, September 2019
ISSN 1810-5408
Nuclear Science
and Technology
Volume 9, Number 3, September 2019
Published by
VIETNAM ATOMIC ENERGY SOCIETY
VIETNAM ATOMIC ENERGY INSTITUTE
NUCLEAR SCIENCE AND TECHNOLOGY
Volume 9, Number 3, September 2019
Editorial Board
Editor-in-chief
Tran Huu Phat (VINATOM)
Executive Editors
Vuong Huu Tan (VARANS)
Tran Chi Thanh (VINATOM)
Cao Dinh Thanh (VINATOM)
Hoang Anh Tuan (VAEA)
Editors
Nguyen Kien Cuong (VINATOM)
Nguyen Nhi Dien (VINATOM)
Nguyen Thi Kim Dung (VINATOM)
Ho Manh Dung (VINATOM)
Nguyen Nam Giang (VINATOM)
Trinh Van Giap (VINATOM)
Le Ngoc Ha (108 Military Central Hospital)
Phan Son Hai (VINATOM)
Le Hong Khiem (IOP)
Dao Tien Khoa (VINATOM)
Tran Hoai Nam (Duy Tan University)
Dang Duc Nhan (VINATOM)
Nguyen Hao Quang (VINATOM)
Nguyen Mong Sinh (VINATOM)
Tran Duc Thiep (IOP)
Dang Quang Thieu (VINATOM)
Le Ba Thuan (VINATOM)
Le Huy Ham (VAAS)
Nguyen Quoc Hien (VINATOM)
Le Van Hong (VINATOM)
Nguyen Trung Tinh (VARANS)
Tran Ngoc Toan (VINATOM)
Duong Thanh Tung (VARANS)
Nguyen Nu Hoai Vi (VARANS)
Nguyen Tuan Khai (VINATOM)
Pham Dinh Khang (VINATOM)
Science Secretary
Hoang Sy Than (VINATOM)
....................................................................................................................................................................................................................
Copyright: ©2008 by the Vietnam Atomic Energy Society (VAES), Vietnam Atomic Energy Institute
(VINATOM).
Pusblished by
Vietnam Atomic Energy Society, 59 Ly Thuong Kiet, Hanoi, Vietnam
Vietnam Atomic Energy Institute, 59 Ly Thuong Kiet, Hanoi, Vietnam
....................................................................................................................................................................................................................
Contents
Nuclear Transmutation of Long-Lived Fission Product I-129 in Radial Blanket of
Sodium-Cooled Fast Reactor
Peng Hong Liem, Donny Hartanto, Yoshihisa Tahara, Naoyuki Takaki…………………... 01
Calculation Results for Enhancing Ability of I-131 Radioisotope Production Using
Tellurium Dioxide Target on the Dalat Nuclear Research Reactor
Kien Cuong Nguyen, Ton Nghiem Huynh, Vinh Vinh Le, Ba Vien Luong, Quang Huy
Pham, Quoc Duong Tran, Van Cuong Bui …………………………………………..……..... 09
Modeling of the Dalat Nuclear Research Reactor (DNRR) with the Serpent 2 Monte
Carlo code
Nguyen Kien Cuong, Nguyen Thi Dung, Tran Viet Phu, Nguyen Huu Tiep, Pham Nhu
Viet Ha……………………………………………………………………………………………... 21
High-resolution fission studies with the planned GBS facility at ELI-NP
D. Choudhury, L. Csige, A.J. krasznahorkay, D.L. balabanski, P. Constantin, M.
Csatlós, J. Gulyás, A. Oberstedt………………………………………………………….......... 30
Applications of Electron Beam to Environmental Conservation
Koichi Hirota ………...…………………………………………………………………..……… 38
Temporal variation of stable isotopic values for dissolved nitrogen compounds in paddy
water environment
Makoto Saiki, Thu Nga Do, Thi Thuy Hai Cao, Takashi Nakamura, Thi Thao Ta, Kei
Nishida…………………………………………………………………………..…..…….. 42
Study on the isomeric decay of neutron-rich isotope 67Fe
L. X. Chung, P.-A. Söderström, A. Corsi, P. Doornenbal, A. Gillibert, P. D. Khue, B. D.
Linh, S. Nishimura, A. Obertelli, N. D. Ton..……………………………………….……..
48
Nuclear Science and Technology, Vol.9, No. 3 (2019), pp. 01-08
Nuclear Transmutation of Long-Lived Fission Product I-129 in
Radial Blanket of Sodium-Cooled Fast Reactor
Peng Hong Liem 1,2,*, Donny Hartanto3, Yoshihisa Tahara1 and Naoyuki Takaki1
1Nippon Advanced Information Service (NAIS Co., Inc.),
416 Muramatsu, Tokai-mura, Naka-gun Ibaraki, Japan
2Tokyo City University (TCU), 1-28-1 Tamazutsumi, Setagaya, Tokyo, Japan
3University of Sharjah, P.O. BOX 27272, Sharjah, United Arab Emirates
Emai: [email protected].jp
(Received 07 November 2019, accepted 25 November 2019)
Abstract: An investigation on the nuclear transmutation of elemental long-lived fission product
(LLFP) in a fast reactor is being conducted focusing on the I-129 LLFP (half-life 15.7 million
years) to reduce the environmental burden. The LLFP assembly is loaded into the radial blanket
region of a Japanese MONJU class sodium-cooled fast reactor (710 MWth, 148 days/cycle). The
iodine element containing I-129 LLFP (without isotope separation) is mixed with YD2 and/or YH2
moderator material to enhance the nuclear transmutation rate. We studied the optimal moderator
volume fraction to maximize the transmutation rate (TR, %/year) and the support factor (SF is
defined as the ratio of transmuted to produced LLFP). We also investigated the effect of LLFP
assembly loading position in the radial blanket and the severe power peak appeared at the fuel
assembly adjacent to the LLFP assembly.
Keywords: Nuclear transmutation, long-lived fission product, I-129, radial blanket, sodium-cooled
fast reactor.
first by elemental separation of MAs and
LLFPs from HLRW using chemical processes
(partitioning) and then followed by nuclear
reactions of the MAs and LLFPs into short-
lived or stable nuclides (transmutation). The
transmutation can be done in a nuclear reactor
or in an accelerator-driven transmutation
system (ADS).
I. INTRODUCTION
Transmutation of long-lived fission
products (LLFPs: Se-79, Zr-93, Tc-99, Pd-
107, I-129, and Cs-135) into short-lived or
stable nuclides by fast neutron spectrum
reactors is being revisited in Japan [1]. The
geological disposal of high-level radioactive
wastes (HLRW) as the nuclear fuel cycle by-
products raises public concern in Japan and
many other countries since even after
thousands of years, minor actinides (MAs)
and some LLFPs remain and become the main
contributors of the radioactive hazard. Under
the partitioning and transmutation (P&T)
strategy, research and development efforts are
being conducted to reduce the radioactive
waste to be stored and the above-mentioned
long-term hazard for future generations [2].
The P&T strategy in general is implemented
In this present study, we focus on the
transmutation of one LLFP, i.e. I-129 (half-life
15.7 million years) using a sodium-cooled fast
reactor, in particular in the radial blanket
region.
I-129 is selected in the present study
since
the
chemical
compound
and
manufacturing of this LLFP have been
previously investigated [3]. The investigation
concluded that BaI2 is the suitable chemical
compound for this particular LLFP from the
©2019 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute
NUCLEAR TRANSMUTATION OF LONG-LIVED FISSION PRODUCT I-129 IN …
points of view of chemical stability and easy
manufacturing process. As mentioned before,
the iodine from the partitioning process does
not experience isotope separation process that
is the LLFP iodine element consists of
approximately 76 at % of I-129 and the rest I-
127, if we assume that the LLFP iodine
originated from a spent mixed oxide (MOX)
fuel of a fast breeder reactor core at 80 GWd/t
(two years of irradiation). The I-129 LLFP
irradiated in a sodium-cooled fast reactor will
transmute into I-130 by neutron capture
reaction and then decay (half-life 15.5 second)
into stable Xe-130.
capability will be lower. However, for Japan,
the accumulated surplus of plutonium may be
one argument for not pursuing a high breeding
ratio. Although Japan has decided to close
MONJU fast reactor in December 2016,
scientific knowledge, design, engineering and
technical experiences as well as high quality
measurement data are well accumulated and
documented. Therefore, in this study we use
MONJU as the fast reactor system for the
LLFP transmutation (readers are referred to
JNC TN 4520 2004-001 document for detail
descriptions of MONJU reactor). Findings and
conclusions from this study should be also
valid in other similar fast reactor systems.
A sodium-cooled fast reactor is selected
in the present study because of its high excess
neutrons and high neutron flux available for
the transmutation. Radial blanket regions of a
fast reactor can be used for the proposed
LLFP transmutation since they provide large
space (volume) while the core characteristics
of the fast reactor are not significantly
affected. One important fast reactor
characteristic which is expected to be affected
is the reactor breeding capability. By
replacing radial blanket sub-assemblies (S/As)
containing depleted uranium with LLFP S/As
containing BaI2, the reactor breeding
II. TRANSMUTATION OF I-129 LLFP
The MONJU reactor core layout for the
proposed I-129 LLFP transmutation is shown
in Fig. 1. The reactor has a nominal power of
710 MWth and under this nominal power it is
typically operated in 148 days per core cycle.
From Fig. 1, one can observe that the reactor
core consists of inner and outer core regions,
and the core is surrounded by a radial blanket
region. The radial blanket region is occupied
by three layers of S/As.
Fig. 1. Japanese MONJU reactor configuration and dimension (left), and fuel and radial blanket S/A (right).
A radial blanket S/A may be replaced by the same dimension of a LLFP S/A.
2
PENG HONG LIEM et al.
It is obvious that the first layer of the
irradiation. For enhancing the I-129
transmutation, the volume fraction between
BaI2 and moderator material will be optimized.
The use of a moderator material in the LLFP
S/A is expected to increase the power peaking
factor (PPF) of fuel pins adjacent to the LLFP
S/A. Therefore, in the present study we also
estimate the PPF of the fuel pins of the most
outer fuel S/As in the outer core region.
radial blanket will have the highest neutron
flux and the hardest neutron spectrum.
However, since the mean free path of fast
neutrons is comparable to the radial blanket
S/A dimension, still moderately high neutron
flux can be expected in the second and the
last layers. In this study, the performance of
I-129 transmutation is investigated by
loading the I-129 LLFP S/As in the first,
second or third layers, as well as in several
combinations of layers.
100
10-1
10-2
10-3
10-4
10-5
10-6
10-7
10-8
104
103
102
101
100
10-1
10-2
10-3
10-4
Blanket layer 1
The effectiveness of LLFP (including
I-129) nuclides depends on both the neutron
flux level and LLFP capture cross sections,
namely the product of the two parameters. A
sodium-cooled fast reactor can provide a
high neutron flux level however in general
the LLFP neutron capture cross sections are
low in fast spectrum regimes as shown in
Fig. 2. For I-129, the neutron capture cross
section shows 1/v behavior in the thermal
energy region, followed by many resolved
resonances from several tens eV up to
several kilos eV. The neutron capture cross
section at 0.0253 eV and resonance integral
of I-129 are approximately 30 and 33 barns
(Shibata et al., 2011). On the other hand, in
the same figure, the neutron spectrum of the
blanket first layer is also shown. The neutron
spectrum shows higher contributions in keV
to MeV energy range. These data suggest
that softer neutron spectra than the one of a
typical sodium-cooled fast reactor will
enhance the neutron capture and increase the
transmutation rate of I-129.
I-129 (n,g) I-130
10-10 10-9 10-8 10-7 10-6 10-5 10-4 10-3 10-2 10-1 100 101 102
Energy (MeV)
Fig. 2. I-129 LLFP capture cross section (JENDL-4.0).
To evaluate quantitatively the I-129
transmutation effectiveness in a fast reactor,
the following performance indices are defined:
( )
( )
( )
N 0 − N T
(
)
Transmutation Rate TR ≡
T × N 0
≈ 휎퐶퐿퐿퐹푃
∅
(1)
푇푎푟푔푒푡
( )
N 0 − N(T)
γ × F × T
퐿퐿퐹푃
Support Factor (SF) ≡
퐿퐿퐹푃
푁
≈ (훾
퐹) (∅푇푎푟푔푒푡) (푁
)
퐹푖푠푠푖푙푒
퐶표푟푒
(2)
휎
푇푎푟푔푒푡
푐
퐿퐿퐹푃
휎
∅
퐶표푟푒
퐹
where N(0) and T are the number of
initial atoms of I-129 in a LLFP S/A and the
irradiation period, respectively, while and F
are the I-129 yield per fission and the total
fission rate of the core, respectively. From the
definition, it is clear that TR is the ratio of the
initially loaded I-129 to the amount of the
transmuted I-129 in the LLFP S/A per unit
In the present study, yttrium deuteride
(YD2) and yttrium hydride (YH2) moderator
materials are adopted for softening the neutron
spectrum in the LLFP S/A. Yttrium is superior
due to its low absorption cross section and does
not produce long-lived daughters during
3
NUCLEAR TRANSMUTATION OF LONG-LIVED FISSION PRODUCT I-129 IN …
time. On the other hand, SF indicates the ratio
of the amount of transmuted I-129 to the
amount of I-129 produced in the core over the
same period of time.
height of approximately 5 cm for burnup
calculation. The LLFP S/A was modeled as a
homogeneous mixture of BaI2 and moderator
material. For each predictor-corrector burnup
step, the effective and skip numbers of
generations (10,000 histories per generation)
were 100 and 20, respectively. The core
characteristics and transmutation performance
were evaluated at the beginning of cycle (BOC,
0 day), middle of cycle (MOC, 74 days) and
end of cycle (EOC, 148 days). The reactor is
assumed to be operated at its nominal power
(710 MWth) with all control rods fully
withdrawn.
As mentioned above, TR is strongly
dependent on the product of I-129 effective
microscopic capture cross section and the
neutron flux level. The performance index SR
is needed because during the I-129
transmutation in the blanket region, the reactor
core also produces I-129. If the condition of
SR>1.0 can be achieved, then the self-
produced I-129 can be transmuted (eliminated)
during electric power generation.
B. Analysis Results and Discussions
III. ANALYSIS AND RESULTS
The
performance
of
I-129
A. Analytical Tools, Library and Calculation
Model
transmutation in the radial blanket region of
MONJU reactor is summarized in Table I &
Fig. 3 and Table II Fig. 4 for YD2 and YH2
moderator material, respectively. The I-129
LLFP S/As are loaded into the first, second
and third layer of the radial blanket region,
or the combination of them. The BaI2 volume
fraction is varied from 100 % volume (no
moderator material) to 10 %.
The continuous energy Monte Carlo
SERPENT2 code [4] with JENDL-4.0 library
[5] was used for the neutronics and burnup
analysis. The MONJU reactor was modeled in
2-D R-Z geometry to reduce the overall
calculation time. In the axial (Z) direction, a
fuel S/A was divided into many zones with the
YD2 Moderator
YH2 Moderator
10
8
25
25
20
15
10
5
10
8
SF Lay. 1, 2, 3
SF Lay. 1, 2
SF Lay. 1
TR Lay. 1, 2, 3
TR Lay. 1, 2
TR Lay. 1
SF Lay. 1, 2, 3
SF Lay. 1, 2
SF Lay. 1
TR Lay. 1, 2, 3
TR Lay. 1, 2
TR Lay. 1
20
15
10
5
6
6
4
4
2
2
0
0
0
0
0
20
40
60
80
100
0
20
40
60
80
100
BaI2 Volume Fraction (%)
BaI2 Volume Fraction (%)
Fig.3. Performance of I-129 transmutation in LLFP
Fig. 4. Performance of I-129 transmutation in LLFP
S/A using YD2 moderator material.
S/A using YH2 moderator material.
4
PENG HONG LIEM et al.
Table I. Performance of I-129 transmutation in LLFP S/A using YD2 moderator material.
BaI2
[%v]
Mod
[%v]
Layer
1
Layer
2
Layer
3
TR
[%/y]
SF
PPF
100
70
0
YD2
YD2
YD2
YD2
YD2
YD2
YD2
YD2
YD2
YD2
YD2
YD2
YD2
0.74
1.17
1.77
3.37
1.60
1.04
0.68
1.36
0.91
2.32
1.47
0.97
2.06
1.35
3.80
2.38
1.56
3.83
2.43
12.6
14.0
12.1
5.8
6.0
4.4
2.9
10.7
7.6
5.0
3.5
2.3
9.3
6.5
2.0
1.4
0.9
4.3
2.9
0.89
0.93
1.13
1.50
0.94
30
60
90
30
40
10
70
YD2
YD2
YD2
YD2
YD2
YD2
YD2
0.94
1.10
40
60
YD2
YD2
YD2
YD2
YD2
YD2
YD2
1.13
1.37
10
90
YD2
YD2
YD2
YD2
YD2
YD2
1.50
First
the
I-129
transmutation
layer pairs (layer 1 & 2, layer 2 &3). The
largest TR value (3.83) is found for the 1 &2
layer pair with BaI2 volume fraction of 10 %,
while the largest SF value (14.0) is found for
the case of LLFP S/As loaded in all layers with
BaI2 volume fraction of 70 %. Concerning the
PPF, severe PPF values appear if LLFP S/As
are located in the first layer, in particular for
cases with a high volume fraction of moderator
material. For the case with the highest volume
fraction of moderator material (90 %), the PPF
is approximately 1.50.
performance of LLFP S/A with YD2 moderator
material is discussed. For cases where all three
layers are loaded with LLFP S/As, increasing
the moderator volume fraction (or equivalently
decreasing the BaI2 volume fraction) softens
the neutron spectrum and enhances the TR
values, but decreases the SF values. When
LLFP S/As loaded location is varied from a
single layer 1, 2 and 3, both the TR and SF
values decrease because of lower neutron flux
levels. Similar trends are observed for varied
5
NUCLEAR TRANSMUTATION OF LONG-LIVED FISSION PRODUCT I-129 IN …
Table II. Performance of I-129 transmutation in LLFP S/A using YH2 moderator material.
BaI2
[%v]
Mod
[%v]
Layer
1
Layer
2
Layer
3
TR
[%/y]
SF
PPF
100
70
0
YH2
YH2
YH2
YH2
YH2
YH2
YH2
YH2
YH2
YH2
YH2
YH2
YH2
0.74
12.6
0.89
1.41
2.03
2.74
1.55
30
60
90
30
1.46
2.26
3.58
2.44
1.55
0.96
1.90
1.20
4.38
2.68
1.57
3.16
1.89
7.44
4.35
2.44
5.12
2.92
1.37
17.4
15.4
6.1
40
10
70
9.2
YH2
6.5
YH2
YH2
4.1
YH2
YH2
YH2
YH2
15.0
10.0
9.5
1.43
2.18
40
10
70
60
90
30
YH2
6.5
YH2
YH2
3.8
YH2
YH2
YH2
YH2
14.2
9.1
2.04
2.84
4.0
YH2
2.7
YH2
1.5
YH2
YD2
YH2
YH2
YH2
5.8
2.74
0.96
YH2
YH2
3.5
16.3
Secondly the transmutation performance
of LLFP S/A with YH2 moderator material is
discussed. Compared to deuterium, hydrogen
mass is half therefore it is expected that with the
same nuclide density hydrogen moderates
significantly better than the deuterium does.
However, at the same time the use of hydrogen is
expected to increase the PPF value. From Table II
one can observe many similar trends appeared in
the YD2 cases related to moderator material
volume fraction, LLFP S/A location etc.
Nevertheless, severe PPF values already appear
at a low moderator material volume fraction of
30 %. The largest TR value (5.12) is found the 1
& 2 layer pair with volume fraction of 10 %,
while the largest SF value (17.4) is found for the
case of LLFP S/As loaded in all layers with BaI2
volume fraction of 70 %. These largest TR and
SF values of YH2 are higher than the ones of
YD2, with the penalty of severer PPF values.
6
PENG HONG LIEM et al.
One possible solution for decreasing the
I-129 Capture Reaction Rate
6x1014
5x1014
4x1014
3x1014
2x1014
1x1014
0
high PPF values is by using a combination of
YD2 and YH2 moderator materials as shown in
the last case of Table II. In this particular case
the moderator material volume fraction is fixed
to 30 % for all layers but the first layer uses
YD2 instead of YH2. The TR and SF values
decrease slightly from 1.46 to 1.37 and 17.4 to
16.3, respectively, but the PPF can be
improved from 1.41 to 0.96.
Layer 1
Layer 2
Layer 3
In order to have a deeper insight of the
Layer 1
Layer 2
Layer 3 Lay. 1 & 2 Lay. 2 & 3 Lay. 1, 2
& 3
LLFP Location
above-mentioned
results,
the
effective
microscopic cross sections of I-129, neutron
flux levels and I-129 neutron capture reaction
rates of LLFP S/A are processed and shown in
Figs. 5, 6 and 7, respectively. The figures are
for the case of YD2 moderator material with
30 % volume fraction (or equivalently BaI2
volume fraction of 70 %).
Fig. 7. I-129 neutron capture reaction rates for the
case of BaI2 volume fraction of 70 % (YD2
moderator material).
When LLFP S/As loaded location is
varied from single layer 1, 2 and 3, the
effective capture cross section of I-129
increases slightly (cf. Fig. 5) since the
neutron spectra become softer, while as
shown in Fig. 4 the neutron flux levels
decrease significantly. As a final result, the
I-129 capture rates (proportional to the TR
values) decrease. The same trends can be
observed for the pair of layers 1 & 2 and 2
& 3.
I-129 Capture XS
1.0
Layer 1
Layer 2
Layer 3
0.8
0.6
0.4
0.2
0.0
Layer 1
Layer 2
Layer 3 Lay. 1 & 2 Lay. 2 & 3 Lay. 1, 2
& 3
LLFP Location
The power peaking factors (PPF) for
the two moderator materials are shown in
Fig. 8. As expected the YH2 moderator
material shows higher PPF values than the
YD2. Under the highest moderator volume
fraction considered in the present work i.e.
90 % the PPF values are found to be 1.5
and 2.7 for YD2 and YH2, respectively. The
softest neutron spectrum amongst all cases
listed in Tables I and II is shown in Fig. 9,
namely the case of YH2 moderator material
Fig. 5. I-129 effective microscopic capture cross
section for the case of BaI2 volume fraction of 70 %
(YD2 moderator material).
Neutron Flux Level
1.0x1015
Layer 1
Layer 2
8.0x1014
6.0x1014
4.0x1014
2.0x1014
0.0
Layer 3
with volume fraction of 90 %.
A
Layer 1
Layer 2
Layer 3 Lay. 1 & 2 Lay. 2 & 3 Lay. 1, 2
& 3
LLFP Location
pronounced thermal peak can be observed
below 1 eV which is the reason for the high
PPF values.
Fig. 6. Total neutron flux levels for the case of BaI2
volume fraction of 70 % (YD2 moderator material).
7
NUCLEAR TRANSMUTATION OF LONG-LIVED FISSION PRODUCT I-129 IN …
(3.83 and 5.12) are found for a pair of LLFP
S/A loaded into the first and second layer with
BaI2 volume fraction of 10 %, while the largest
SF values (14.0 and 17.4) are found for the
case of 3 LLFP S/As with BaI2 volume fraction
of 70 % for YD2 and YH2 moderator material,
respectively. YD2 moderator material shows
better PPF values and should be used in a
3.0
2.5
2.0
1.5
1.0
0.5
0.0
YH2 Lay. 1, 2, 3
YH2 Lay. 1, 2
YH2 Lay. 1
YD2 Lay. 1, 2, 3
YD2 Lay. 1, 2
YD2 Lay. 1
LLFP S/A loaded in the first layer of the
blanket region.
0
20
40
60
80
100
BaI2 Volume Fraction (%)
ACKNOWLEDGMENTS
Fig. 8. The power peaking factors of the fuel pin in
the S/A adjacent to the LLFP S/A for YD2 and YH2
moderator materials.
This study comprises partly the results
of “Research and development of an
innovative transmutation system of LLFP by
fast reactors”, entrusted to the Tokyo
Institute of Technology by the Ministry of
Education, Culture, Sports, Science and
Technology of Japan (MEXT). Tables and
Figures were carefully prepared by Mrs.
Suzuko Ikegami.
BaI 10%, YH 90%
2
2
0.15
0.10
0.05
0.00
REFERENCES
[1]. Chiba S. et al., “Method to Reduce Long-lived
Fission Products by Nuclear Transmutations
with Fast Spectrum Reactors,” Sci. Rep., 7,
13961, 2017.
10-8 10-7 10-6 10-5 10-4 10-3 10-2 10-1 100 101 102
Energy (MeV)
Fig. 9. Neutron spectrum of the I-129 LLFP S/A
loaded in the first layer for the case of BaI2 volume
fraction of 10 % (YH2 moderator material).
[2]. International
Atomic
Energy
Agency,
Evaluation of Actinide Partitioning and
Transmutation, Technical Reports Series No.
214, IAEA, Vienna, 1982.
III. CONCLUDING REMARKS
The performance of elemental I-129
(chemical form BaI2) LLFP nuclear
transmutation in the radial blanket region of
the Japanese MONJU fast reactor was
investigated. Several important parameters
such as the LLFP S/A loading position,
moderator material type (YD2 or YH2) and its
volume fraction are taken into account for
optimizing the transmutation effectiveness
guided by using two performance indices, i.e.
the transmutation rate (TR, %/year) and
support factor (SF). The largest TR values
[3]. Osaka M. et al., “Research and Development
of Minor Actinide-Containing Fuel and Target
in a Future Integrated Closed Cycle System,”
Journal of Nuclear Science and Technology,
44(3), 309, 2007.
[4]. Leppänen J. et al. 2015. “The Serpent Monte
Carlo Code: Status, Development and
Applications in 2013,” Annals of Nuclear
Energy, 82, 142, 2015.
[5]. Shibata K. et al., “JENDL-4.0: A New Library for
Nuclear Science and Engineering,” Journal of
Nuclear Science and Technology, 48, 1, 2011.
8
Nuclear Science and Technology, Vol.9, No. 3 (2019), pp. 09-20
Calculation Results for Enhancing Ability of I-131
Radioisotope Production Using Tellurium Dioxide Target
on the Dalat Nuclear Research Reactor
Nguyen Kien Cuong, Huynh Ton Nghiem, Le Vinh Vinh, Luong Ba Vien,
Pham Quang Huy, Tran Quoc Duong, Bui Van Cuong
Dalat Nuclear Research Institute, Vietnam Atomic Energy Institute (VINATOM)
(Received 05 November 2019, accepted 25 November 2019)
Abstract: The paper presents the calculation results in re-design of neutron trap of the Dalat Nuclear
Research Reactor (DNRR) for I-131 radioisotope production using TeO2 target. The new design
permits for loading more TeO2 capsules from 9 to 12, 15 and 18 in the neutron trap. The enhancement
of radioisotope production was implemented by re-arrangement of the neutron trap without changing
the dimension or geometry of irradiation capsules. By using neutronics computer code as MCNP6, the
obtained calculation results of I-131 activity in 6 investigated cases showed that the new design by the
re-arrangement of the neutron trap can be used effectively for radioisotope production with thermal
neutron flux in average range from 5.3×1012 to 1×1013 n/cm2.s and the total activity of I-131 isotope
was increased from about 19.2% to 38.8% comparing with the original design using 9 capsules. The
negative reactivity insertion was from 0.60 βeff to 0.96 βeff when loading capsules that also met the
safety requirements of operational conditions of the DNRR.
Keyword: The DNRR, Radioisotope Production, TeO2, I-131, MCNP6 code.
production needs to be considered to satisfy the
domestic market [3]. Therefore, the design
calculation for enhancement of I-131
radioisotope production was carried out.
I. INTRODUCTION
Research reactor is a very useful facility
for radioisotope production by using neutron to
irradiate the target under normal operation. The
most convenience of this method is simple in
preparation of target capsules and in their
irradiation at fixed positions inside or outside
reactor core. I-131 isotope is an important
radioisotope that can be used for cancer
diagnostics and treatment and it is also
produced easily by using research reactor with
neutron irradiation on TeO2 target [1]. Even the
DNRR has low power of 500 kWt and average
thermal neutron flux is smaller than 5.0×1012
n/cm2.s, radioisotope production is still applied
as a main purpose to exploit the reactor during
more than 35 years from 1984 until now [2].
The demand of I-131 isotope utilization in
nuclear medicine is increasing so the effective
utilization of the DNRR for radioisotope
By modifying the neutron trap
structure, the activity of I-131 radioisotope
was increased by loading up to18 target
capsules but keeping their geometry and
dimension. In original design, 9 TeO2 capsules
have been irradiated at neutron trap of the
reactor core with 150 hours continuously, and
the average activity of I-131 isotope can be
reached to 35 Ci in each reactor operation
cycle. Basically, the method to produce I-131
isotope with half-life of about 8.05 days can
be expressed as follows [1]:
and the half-life of Te-
131 is only 25 minutes.
©2019 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute
CALCULATION RESULTS FOR ENHANCING ABILITY OF I-131 RADIOISOTOPE PRODUCTION…
Fig. 1. The micro cross section of Te-130 isotope [4]
The neutron capture cross section of Te-
130 isotope (see in Fig. 1) in the range from
0.01 eV to 1000 eV is quite high compared
with cross section at higher energy region. The
linear of Te-130 capture cross section is in
thermal energy range under 0.625 eV. So the
main calculation results for the I-131 isotope
production are in the thermal neutron energy
group. The average cross section of neutron
capture reaction of Te-130 is about 67 mili-
barns in this region.
The DNRR is a pool type reactor with
nominal power of 500 kWt and its main
purposes are radioisotope production,
neutron activation analysis and training.
Recently the demand of radioisotope market,
especially I-131 isotope is increasing so the
enhancement of I-131 production on the
DNRR has been considered in calculation by
using neutronics computer codes as MCNP6
[5] and ORIGEN2.1 [6]. The detailed
calculation about the neutron fluxes inside
TeO2 target and reaction rate as well as
neutron fluxes on the irradiation capsules
were done by MCNP6 code.
I-131 isotope can be chemically
separated from the TeO2 target after
irradiation by thermal neutron on research
reactor. By this method, high specific
activity of target can be obtained by
irradiation with high thermal neutron flux.
Of course, another method can be used on
the research reactor to produce I-131 as
fission product by irradiation of the U-235
HEU or LEU but the facility chain is very
complicated with solid waste treatment and
separation of I-131 through fission products
as well.
The number of loaded TeO2 capsules in
the neutron trap after modifying is about 12,
15 and 18 capsules. The rearrangement of
neutron trap with beryllium rods around can
be created more new irradiation channels.
The disadvantage of this rearrangement is
mainly in reducing the neutron flux and
adding more negative reactivity, however
the loaded mass of TeO2 increases and I-131
isotope activity increases from about 19.2%
to 38.8%.
10
NGUYEN KIEN CUONG et al.
II. CALCULATION MODEL AND METHOD
A. Calculation model
the total activity of I-131 after 150-hour reactor
operating continuously approximated 35 Ci.
Fig. 2 shows the loading positions of 9 TeO2
capsules at neutron trap.
Since 2014, the neutron trap of the DNRR
was modified for loading up to 9 capsules and
Fig. 2. The irradiation positions of 9 TeO2 capsules at neutron trap of the DNRR.
The TeO2 target material used for I-
131 isotope production is in powder with
99.9% purity and the mass of Te-130
isotope is about 27% in weight. The TeO2
target of 16 cm height is covered by
aluminum container with the dimension of
2.6 cm in diameter, 1.5 mm in thickness and
20 cm in height. The total volume of target
is about 65 cm3 that is equivalent with 200
to 230 gram of TeO2.
power days with the configuration of 92 LEU
fuel assemblies so the power peaking factor is
smaller than those at the beginning of
operation cycle, and the fuel cladding
temperature is lower than those of the fresh
core [7]. This means that the operational
conditions of the reactor meet safety
requirements.
After modifying the neutron trap, 6
available positions can be loaded by maximum
3 TeO2 capsules in each hole. At the present
work, the loading of 12 capsules with and
without graphite reflector on the top and
bottom, and the loading with 18 capsules was
considered in calculation models. The MCNP6
calculation models are shown in Fig. 3.
By removing the whole neutron trap
with beryllium blocks (about 6 half beryllium
blocks), the center of the neutron trap was
replaced by a beryllium rod to avoid the core
has too high power density. Until now, the
DNRR was operated about more than 300 full
Fig. 3. The arrangement of 12 and 18 TeO2 capsules at neutron trap of the DNRR
11
CALCULATION RESULTS FOR ENHANCING ABILITY OF I-131 RADIOISOTOPE PRODUCTION…
By rearrangement of the beryllium rods
around neutron trap, 2 new positions for
loading 3 capsules each hole can be created.
The total 15 capsules can be loaded in the new
neutron trap including 9 capsules are still in the
old positions. The calculation model by using
MCNP6 code is depicted in Fig. 4. The
rearrangement of the neutron trap satisfies the
safety condition as well as the thermal neutron
flux for radioisotope production.
5-2
5-1
5-3
5-4
5-5
Fig. 4. Loaded positions of 15 TeO2 target capsules in the neutron trap after its rearrangement
with 5 different options.
In the new design, it is very simple and
easy for carrying out experiments to test and to
confirm the calculation results. Under this
arrangement, the violation in thermal
hydraulics limitation is avoided as calculation
results were confirmed in the design of new
LEU core as the second candidate core for
reactor conversion from HEU to LEU fuel [7].
The calculations were carried out for 6
cases with different loading capsule numbers
from 9 to 18. The explanation of loading
scheme for detail calculations is described in
the Table I. It is convenient that all cases
studied were used old capsules and only adding
more capsules at neutron trap after its
rearrangement.
Table I. Calculation cases for enhancement of isotope production on the DNRR
Number of loaded
capsules
Detailed explanation
9
Original neutron trap using 9 capsules currently (Fig. 2) with 3 layers and in
each layer 3 capsules can be loaded
12
NGUYEN KIEN CUONG et al.
12 – Normal
12 – Graphite
All capsules located at the second ring with beryllium rod at center (Fig. 3).
Optimizing axial direction of thermal neutron flux.
Top and bottom of all capsules covered by light water reflector.
As the first new case above but top and bottom of all capsules covered by
graphite reflector.
15 – Long
(5-1, 5-3, 5-4)
Creating two more irradiation channels next to neutron trap following axial
direction in the reactor core (Fig. 4). It is an easy way for rearrangement of
neutron trap with 3 options to select.
15 – Wide
(5-2)
Creating two more irradiation channels next to neutron trap following radial
direction in the reactor core (Fig. 4). It is also very easy way for rearrangement
of neutron trap.
15 – Normal
(5-5)
Creating two more irradiation channels next to neutron trap at cell 5-6 and 9-6
(Fig. 4).
18
As the first new case above but each irradiation channel can be loaded with 3
capsules in axial direction.
Neutron fluxes have to be normalized to
get absolute value of neutron flux under
thermal power of reactor. The absolute neutron
flux can be obtained as follows [5]:
B. MCNP code for neutron fluxes and
reaction rate calculation
The MCNP6 code was used for neutron
flux and reaction rate calculation for TeO2
target in capsules. The calculation models of
the DNRR together with target capsules were
described as detail as possible in Fig. 2, Fig. 3
and Fig. 4. The ENDF/B7.1, JEFF3.3.1
libraries [8] were used for calculation. The
kcode option of MCNP6 code was chosen for
the calculation with positions of control rods
were set nearly critical status. The standard
deviation of keff in the calculation was lower
than 8.0×10-5 with 160 active cycles and 1
million particles in each calculation cycle. The
errors of obtained neutron flux as well as
reaction rates were smaller than 0.3%. The
burn-up calculations of the DNRR were mainly
at two points: after 300 full power days and
120 full power days.
Pv
1
(2)
m
m
1.60221013wf keff
v
where P is power of reactor (MW), is
the average number of neutrons released per
fission, wf is effective energy released per fission
(~ 193.7 MeV/fission to VVR-M2 LEU fuel
type) and keff is the effective multiplication factor.
Actually, getting data including neutron
flux, reaction rates and effective multiplication
factor from MCTAL file is easier than reading
from output file of MCNP6 code. Reaction
rates R are received in tally as:
Rmnx xn (E)m (E)dE
(3)
where Rmnx is reaction rate type x of nuclide
n in cell m, micro cross section has unit barns.
Neutron fluxes and reaction rates of Te-
130 reactions and multiplication factor were
calculated by using the MCNP6 code and three
group neutron fluxes in active cell are
provided by tally F4:
C. I-131 activity calculation after its
irradiation and cooling
The number density of I-131 can be
calculated by equation [9]:
(E)dE
(1)
m
m
13
CALCULATION RESULTS FOR ENHANCING ABILITY OF I-131 RADIOISOTOPE PRODUCTION…
+
(
(
)
(
)
)
( )
(
)(
)
(4)
( ) ( )
(9)
The number density of Te-130 can be
calculated by
(5)
( )
(10)
We can interfere that
With Avogadro
number;
( )
( )(
)
( )is mass of Tellurium in grams;
(6)
( ) is Atomic mass of Tellurium
With is micro cross section (barn); is
neutron flux (n/cm2.s); is decay constant (s-1)
and N is the number density of isotopes
(atoms/cm3).
in grams.
The activity A by Curie unit of I-131 can
be determined by formula:
The equation can be solved by using
numerical method or direct calculation after
integrating the equation (6). Then, the number
density of I-131 can be obtained as follows:
(11)
So the activity in each TeO2 target
capsule can be determined through the
calculation by applying equations from (4) to
(11).
I-131 activity can also be determined by
using ORIGEN2.1 code with input data
including irradiation time under specified
neutron flux, cooling time and substituted
capture cross section of Tellurium isotopes in
activation material library. The thermal neutron
flux and micro cross section of Tellurium
isotope were estimated by MCNP6 code.
(
)(
)
( )
(7)
And tirr and tdecay are irradiation and
decay time (hour), respectively.
III. CALCULATION RESULTS AND
DISCUSSIONS
In the equation (7), we can separate as
two parts with build-up process and decay time
as follows:
A. Thermal neutron flux
In the axial direction, the thermal
neutron flux distribution of 1 cm/node at the
original neutron trap without loading target
compares with modified for loading 9 capsules
(OLD) and the case having 6 irradiation
(8)
14
NGUYEN KIEN CUONG et al.
channels (NEW). The profiles of thermal
trap. However in case adding only 2 new
irradiation channels as in Fig. 4, the reducing
of thermal neutron flux at neutron trap is not
much and only under 5%. Fig. 5 depicts the
profile of thermal neutron trap without loading
capsules for radioisotope production.
neutron flux in three cases have different in
maximum point of 10% and in the average
between OLD and NEW type. It is a
disadvantage of thermal neutron flux when
creating new irradiation channels around neutron
Fig. 5. Thermal neutron flux distribution at original neutron trap and modified for loading 9 capsules (OLD)
and new irradiation positions (NEW)
Basing on the obtained calculation
results, the best distance in axial direction that
has the highest thermal neutron flux is from 8
to 48 cm at bottom to the top direction. The
height is equivalent to the length of 2 capsules
with light water or graphite reflector at the top
and the bottom.
just of 8.8×1012, 8.3×1012 and 8.0×1012 with
15, 18 and 12 capsules loaded, respectively.
The micro cross section in barn unit of Te-
130 isotope in thermal energy is about
1.53E-01 barns.
The thermal neutron flux inside all
capsules is high enough from 5.3×1012
n/cm2.s to 1.0×1013 n/cm2.s at top and
middle position at irradiation channel. The
detail axial distribution of each node of
0.78 cm/node inside 3 TeO2 target capsules
from bottom to the top is depicted in the
Fig. 7.
When loading TeO2 capsules at different
irradiation positions inside the reactor core
as in Table I, the average thermal neutron
flux inside TeO2 target capsules is described
in the Fig. 6. The maximum of average
thermal neutron flux in the 9 capsules loaded
is about 1.0×1013 n/cm2.s while in other cases
15
CALCULATION RESULTS FOR ENHANCING ABILITY OF I-131 RADIOISOTOPE PRODUCTION…
1.1E+13
1.0E+13
1.0E+13
9.8E+12
9.5E+12
15-Long
15-Wide
9-old
9.3E+12
9.0E+12
8.8E+12
8.5E+12
8.3E+12
8.0E+12
7.8E+12
7.5E+12
7.3E+12
7.0E+12
6.8E+12
6.5E+12
6.3E+12
6.0E+12
5.8E+12
5.5E+12
5.3E+12
5.0E+12
12-Nor
12-Graph
18-Nor
10-Cont
15-Nor
1
2
3
4
5
6
7
8
9
10 11 12 13 14 15 16 17 18
Number of container
Fig. 6. The average thermal neutron flux in each capsule of 6 calculated cases
Fig. 7. Relative thermal neutron flux of 3 capsules at each node in an irradiation channel
from bottom to the top
number of capsules is depicted on the Table
II with the difference of TeO2 mass in the
capsules. The maximum of the negative
reactivity insertion depends on the mass of
loaded targets and it will be increased when
increasing mass of Te-130.
B. Negative reactivity and safety
The negative reactivity insertion when
loading capsules was also estimated by
comparing with the core configuration
without
capsules.
Detailed
negative
reactivity insertion for different loading
16
NGUYEN KIEN CUONG et al.
Table III. Negative reactivity insertion with different mass of Te-130 from 54 to 62 gram in each
TeO2 target capsule
Number of capsule
Negative reactivity (cents)
(60.2 – 70.2)
9
12 normal
12 with graphite
15 long
(69.2 – 79.2)
(63.4 – 73.4)
(83.8 – 93.8)
15 wide
(81.4 – 91.4)
18
(86.5 – 96.5)
The range of negative reactivity is met
requirements for the reactor safety in normal
operation. In thermal hydraulics, fuel cladding
temperature is a parameter needed to be
evaluated to assure the nuclear safety. As
calculation results in the design of new LEU
cores of the DNRR, the maximum temperature
of the hottest channels in 2 core configurations
is lower than permission temperature 1030C.
So the rearrangement of neutron trap in 5
investigated cases will be also satisfied the
limit of fuel cladding temperature after
operating of 8 years.
irradiation time from 24 to 150 hours and the
same target mass of 62 gram Te-130,
calculation results are shown in Table II and
Fig. 8. The increasing of neutron flux is
proportional with I-131 activity in the scale 1
by 1. So the improvement of thermal neutron
flux is
a
way to enhancement for
radioisotope production. In case of the
DNRR, the method to increase neutron flux
is difficult because of its low power and
design especially using fuel with low power
density. If upgrading power of the DNRR
from 0.5 MW to 0.6 or 0.8 MW the expected
thermal neutron flux can be increased 16 to
35% respectively. Because the irradiation
positions in axial direction of 3 capsules in
each channel have different thermal neutron
flux, the archived activity of I-131 after 150-
hour irradiation of current 9 capsules has the
ratio as 1.00:1.20:1.54 as top, bottom and
middle positions of target.
C. Activity of I-131 isotope in 6 investigated
cases
The main effects of irradiation for
radioisotope production include thermal
neutron flux, mass of target and irradiation
time or decay time. With different neutron
flux range from 5.3E+12 to 1.0E+13 n/cm2.s,
Table II. Calculation results of I-131 activity (Ci) at different thermal neutron flux and the mass of Te-130 is
about 62 gram in a capsule
Activity of I-131 (Ci) with different thermal neutron flux (n/cm2.s)
Irradiation
F=5.3E+12 F=5.5E+12
F=6.8E+12
0.65
F=8.5E+12
0.81
time (hr.)
F=6.5E+12
0.62
F=7.8E+12
0.75
F=1.0E+13
0.96
24
48
0.51
0.99
1.42
1.83
2.20
2.54
0.53
1.02
1.48
1.90
2.28
2.63
1.21
1.26
1.45
1.58
1.86
72
1.75
1.83
2.10
2.28
2.69
96
2.24
2.34
2.69
2.93
3.45
120
144
2.69
2.82
3.23
3.52
4.14
3.11
3.25
3.73
4.07
4.78
17
Tải về để xem bản đầy đủ
59 trang
10560
Free
Bạn đang xem 20 trang mẫu của tài liệu "The journal of Nuclear Science and Technology - Volume 9, Number 3, September 2019", để tải tài liệu gốc về máy hãy click vào nút Download ở trên
File đính kèm:
- the_journal_of_nuclear_science_and_technology_volume_9_numbe.pdf
