Calculation results for enhancing ability of I-131 radioisotope production using tellurium dioxide target on the dalat nuclear research reactor

. The absolute neutron 
flux can be obtained as follows [5]: 
m
efff
m
kw
vP

1
106022.1 13 
  (2) 
where P is power of reactor (MW), v 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: 
 dEEER m
n
xmnx )()(  (3) 
where Rmnx is reaction rate type x of nuclide 
n in cell m, micro cross section has unit barns. 
C. I-131 activity calculation after its 
irradiation and cooling 
The number density of I-131 can be 
calculated by equation [9]: 
CALCULATION RESULTS FOR ENHANCING ABILITY OF I-131 RADIOISOTOPE PRODUCTION 
14 
 ( 
 ) 
(4) 
(5) 
We can interfere that 
 ( 
 )( ( 
 ) ) 
 (6) 
With  is micro cross section (barn); is 
neutron flux (n/cm
2
.s);  is decay constant (s-1) 
and N is the number density of isotopes 
(atoms/cm
3
). 
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: 
( 
 )( 
 )
 ( 
 ) 
 (7) 
And tirr and tdecay are irradiation and 
decay time (hour), respectively. 
In the equation (7), we can separate as 
two parts with build-up process and decay time 
as follows: 
 (8) 
 + 
( 
 ( 
 ) )( 
 ( 
 ) 
 ( 
 ) )
 ( 
 ) ( 
 )
(9) 
The number density of Te-130 can be 
calculated by 
 ( 
 ) 
 (10) 
With 
 Avogadro 
number; 
 ( 
 )is mass of Tellurium in grams; 
 ( 
 ) is Atomic mass of Tellurium 
in grams. 
The activity A by Curie unit of I-131 can 
be determined by formula: 
 (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. 
III. CALCULATION RESULTS AND 
DISCUSSIONS 
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 
NGUYEN KIEN CUONG et al. 
15 
channels (NEW). The profiles of thermal 
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 
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. 
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. 
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 1.0×10
13
 n/cm
2
.s while in other cases 
just of 8.8×10
12
, 8.3×10
12
 and 8.0×10
12
 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 5.3×10
12
n/cm
2
.s to 1.0×10
13
 n/cm
2
.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. 
CALCULATION RESULTS FOR ENHANCING ABILITY OF I-131 RADIOISOTOPE PRODUCTION 
16 
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 
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 
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. 
5.0E+12
5.3E+12
5.5E+12
5.8E+12
6.0E+12
6.3E+12
6.5E+12
6.8E+12
7.0E+12
7.3E+12
7.5E+12
7.8E+12
8.0E+12
8.3E+12
8.5E+12
8.8E+12
9.0E+12
9.3E+12
9.5E+12
9.8E+12
1.0E+13
1.0E+13
1.1E+13
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18
T
h
e
rm
al
 n
e
u
tr
o
n
 f
lu
x 
(n
/c
m
2
.s
)
Number of container
15-Long
15-Wide
9-old
12-Nor
12-Graph
18-Nor
10-Cont
15-Nor
NGUYEN KIEN CUONG et al. 
17 
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) 
9 (60.2 – 70.2) 
12 normal (69.2 – 79.2) 
12 with graphite (63.4 – 73.4) 
15 long (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 103
0
C. 
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. 
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/cm
2
.s, 
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. 
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 
Irradiation 
time (hr.) 
Activity of I-131 (Ci) with different thermal neutron flux (n/cm
2
.s) 
F=5.3E+12 F=5.5E+12 F=6.5E+12 F=6.8E+12 F=7.8E+12 F=8.5E+12 F=1.0E+13 
24 0.51 0.53 0.62 0.65 0.75 0.81 0.96 
48 0.99 1.02 1.21 1.26 1.45 1.58 1.86 
72 1.42 1.48 1.75 1.83 2.10 2.28 2.69 
96 1.83 1.90 2.24 2.34 2.69 2.93 3.45 
120 2.20 2.28 2.69 2.82 3.23 3.52 4.14 
144 2.54 2.63 3.11 3.25 3.73 4.07 4.78 
CALCULATION RESULTS FOR ENHANCING ABILITY OF I-131 RADIOISOTOPE PRODUCTION 
18 
150 2.62 2.71 3.21 3.36 3.85 4.19 4.93 
160 2.75 2.85 3.37 3.52 4.04 4.40 5.18 
170 2.87 2.98 3.52 3.68 4.23 4.60 5.42 
180 2.99 3.10 3.67 3.84 4.40 4.80 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 
Mass of 
TeO2 target 
(gram) 
9 12-normal 12-graphite 
After 150 
hrs 
Decay 18 
hrs 
After 150 
hrs 
Decay 18 
hrs 
After 150 
hrs 
Decay 18 
hrs. 
200 31.23 29.27 38.67 36.25 39.47 37.00 
210 32.79 30.74 40.60 38.06 41.45 38.85 
220 34.35 32.20 42.53 39.87 43.42 40.70 
230 35.91 33.66 44.47 41.69 45.39 42.56 
Increasing % 
(comparing 
with 9 
capsules) 
19.24 20.89 
Mass of 
TeO2 target 
15-Long (5-1, 5-3, 5-4) 15-Wide (5-2) 18 
After 150 Decay 18 After 150 Decay 18 After 150 Decay 18 
NGUYEN KIEN CUONG et al. 
19 
(gram) hrs. hrs. hrs. hrs. hrs. hrs. 
200 44.95 42.14 43.86 41.12 51.00 47.81 
210 47.20 44.24 46.05 43.17 53.55 50.20 
220 49.44 46.35 48.25 45.23 56.10 52.59 
230 51.69 48.46 50.44 47.29 58.65 54.98 
Increasing % 
(comparing 
with 9 
capsules) 
30.53 28.81 38.77 
Mass of 
TeO2 target 
(gram) 
10 (5-5) 13 (5-5) 15 (5-5) 
After 150 
hrs. 
Decay 18 
hrs. 
After 150 
hrs. 
Decay 18 
hrs. 
After 150 
hrs. 
Decay 18 
hrs. 
200 37,18 34,85 41,26 38,68 45,32 42,48 
210 39,03 36,59 43,32 40,61 47,58 44,61 
220 40,89 38,33 45,39 42,55 49,85 46,73 
230 42,75 40,08 47,45 44,48 52,11 48,85 
Increasing% 
(comparing 
with 9 
capsules) 
16.00 24.32 31.10 
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 
27
o
C of water at core inlet, the maximum 
temperature is still lower than 88
o
C and meets 
requirements for operating limit condition of 
the VVR-M2 LEU fuel type of the DNRR. 
CALCULATION RESULTS FOR ENHANCING ABILITY OF I-131 RADIOISOTOPE PRODUCTION 
20 
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. 
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. 
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 
production on the Dalat Nuclear Research 
Reactor” with grant number ĐTCB.10/191 
VNCHN. 
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. 
[4]. https://wwwndc.jaea.go.jp/j40fig/jpeg/te130_f
1.jpg 
[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). 
[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. 
[8]. https://www.oecd-
nea.org/dbdata/jeff/jeff33/#_jeff-
33_downloads 
[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.