Calculation results for enhancing ability of I-131 radioisotope production using tellurium dioxide target on the Dalat nuclear research reactor
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
CALCULATION RESULTS FOR ENHANCING ABILITY OF I-131 RADIOISOTOPE PRODUCTION…
150
160
170
180
2.62
2.75
2.87
2.99
2.71
2.85
2.98
3.10
3.21
3.37
3.52
3.67
3.36
3.52
3.68
3.84
3.85
4.04
4.23
4.40
4.19
4.40
4.60
4.80
4.93
5.18
5.42
5.65
Fig. 8. Activity of I-131 using about 62 gram TeO2 target capsule with different thermal neutron flux
The calculation results of I-131
activity with different target mass in 6
investigated cases are described in Table III
and Fig. 9. The irradiation condition includes
150-hour irradiation time, mass of target
from 54 to 62 gram of Te-130. The obtained
results showed that the total activity of I-131
in new design comparing with old design
using 9 capsules is increased from 19.24 to
38.77%. It is satisfied the requirements for
enhancement of radioisotope production on
the DNRR.
Table III. Calculation results of I-131 activity of 6 cases with different mass of TeO2 target
9
12-normal
12-graphite
Mass of
TeO2 target
(gram)
After 150
hrs
Decay 18
hrs
After 150
Decay 18
hrs
After 150
Decay 18
hrs.
hrs
hrs
200
210
220
230
31.23
32.79
34.35
35.91
29.27
30.74
32.20
33.66
38.67
40.60
42.53
44.47
36.25
38.06
39.87
41.69
39.47
41.45
43.42
45.39
37.00
38.85
40.70
42.56
Increasing %
(comparing
with 9
19.24
20.89
capsules)
15-Long (5-1, 5-3, 5-4)
After 150 Decay 18
15-Wide (5-2)
After 150 Decay 18
18
Mass of
TeO2 target
After 150
Decay 18
18
NGUYEN KIEN CUONG et al.
(gram)
200
hrs.
hrs.
hrs.
hrs.
hrs.
hrs.
44.95
47.20
49.44
51.69
42.14
44.24
46.35
48.46
43.86
46.05
48.25
50.44
41.12
43.17
45.23
47.29
51.00
53.55
56.10
58.65
47.81
50.20
52.59
54.98
210
220
230
Increasing %
(comparing
with 9
30.53
28.81
38.77
capsules)
10 (5-5)
13 (5-5)
15 (5-5)
Mass of
TeO2 target
(gram)
After 150
hrs.
Decay 18
hrs.
After 150
hrs.
Decay 18
hrs.
After 150
hrs.
Decay 18
hrs.
200
210
220
230
37,18
34,85
41,26
38,68
45,32
42,48
39,03
40,89
42,75
36,59
38,33
40,08
43,32
45,39
47,45
40,61
42,55
44,48
47,58
49,85
52,11
44,61
46,73
48,85
Increasing%
(comparing
with 9
16.00
24.32
31.10
capsules)
Fig. 9. Total activity of I-131 with different TeO2 mass and under 150-hour irradiation
In 6 new cases considered, the simplest
way is rearrangement of neutron trap with an
option to create 2 irradiation channels by
replacing the beryllium rods at cell 5-6 and 9-
6 to load 6 capsules more. The new core
configuration still keeps the neutron trap, 92
LEU fuel assemblies and 10 beryllium rods.
And this core configuration was confirmed the
safety by evaluating the maximum fuel
cladding temperature by PLTEMP4.2 code
[10] at hottest fuel assembly located at cell 4-
5. Following the calculation results, under
27oC of water at core inlet, the maximum
temperature is still lower than 88oC and meets
requirements for operating limit condition of
the VVR-M2 LEU fuel type of the DNRR.
19
CALCULATION RESULTS FOR ENHANCING ABILITY OF I-131 RADIOISOTOPE PRODUCTION…
production on the Dalat Nuclear Research
Reactor” with grant number ĐTCB.10/191
VNCHN.
IV. CONCLUSIONS AND REMARKS
The calculation about the enhancement
radioisotope production for I-131 by
rearrangement of neutron trap to loading more
target capsules using MCNP6 code were
conducted. In general, 6 cases for enhancement
of I-131 radioisotope production on the DNRR
can be carried out easily and meet the safety
requirements in reactor operation. Comparing
with old design using 9 containers, the total
activity of I-131 isotope was increased from
about 19.2% to 38.8% with loading TeO2 target
mass increasing from 25% to 50%. The
rearrangement of beryllium rods around
neutron trap at cell 5-6 and 9-6 to create 2 new
irradiation channels is highly considered
because of safety, easy implementation and
effectiveness in production of I-131.
REFERENCES
[1]. International Atomic Energy Agency, Manual
for reactor produced radioisotopes, IAEA
TECDOC-1340. IAEA, Vienna, Austria; 2003.
p. 121-4.
[2]. Nuclear Reesearch Institute, Safety Analysis
Report for the DNRR, version 2012.
[3]. Nguyen Kien Cuong et al., “Some
enhancement plans for radioisotope production
at neutron trap of the reactor core using LEU
fuel”, the 10th National Conference on Nuclear
Science and Technology, Vung-tau City, 2013.
[5]. X-5 Monte Carlo Team, MCNP – A General
Monte Carlo N-Particle Transport Code,
Version 5, LAUR-03-1987, Los Alamos
National Laboratory, Los Alamos, New
Mexico (2003).
The design calculation is a very
important step before carrying out testing a
reactor operating cycle with new design for
changing neutron trap to satisfy safety of the
normal operation condition of the DNRR. The
increasing I-131 product and operation time as
well as using economical fuel are the main
purposes in effective utilizations and
applications of the DNRR. The calculation
results can also be used as reference for the
new research reactor in design of radioisotope
production especially of I-131 isotope.
[6]. A.G. Croff, A User Manual for the ORIGEN2
Computer Code, Oak Ridge NL, 1980.
[7]. LUONG Ba Vien, LE VinhVinh, HUYNH
Ton Nghiem, NGUYEN Kien Cuong, “Design
Analyses for full core conversion of the Dalat
Nuclear Research Reactor”, Nuclear Science
and Technology,Vol. 4. No. 1 (2014), pp. 10-
25, Vietnam Atomic Energy Association,
August, 2014.
ACKNOWLEDGEMENT
The authors would like to express
special thanks for the helpful discussions and
willing carrying out experiments on the DNRR
of the staff in Reactor Physics and Engineering
Department, Reactor Center. We also
appreciate the financial support, valuable
helping of Ministry of Science and Technology
and VINATOM for implementation of the
research project: “Study on effective, economic
in using fuel and enhancement of radioisotope
[9]. Abdessamad Didi, Ahmed Dadouch, Hassane
El Bekkouki, “Feasibility study for Production
of Iodine-131 using dioxide of Tellurium-
130”, International Journal of Pharmaceutical
Sciences 2016, Vol. 8, Issue 11, 327-331.
[10]. P. Olson and M. Kalimullah, “A User’s Guide
to the PLTEMP/ANL Code, Version 4.2”,
Nuclear Engineering Division, Argonne
National Laboratory, USA, July 7, 2015.
20
12 trang
10560
Free
Bạn đang xem tài liệu "Calculation results for enhancing ability of I-131 radioisotope production using tellurium dioxide target on the Dalat nuclear research reactor", để tải tài liệu gốc về máy hãy click vào nút Download ở trên
File đính kèm:
- calculation_results_for_enhancing_ability_of_i_131_radioisot.pdf
