Design of an irradiation rig using screen method for silicon transmutation doping at the Dalat research reactor

fore, the design 
and testing of an irradiation rig using flux 
screen method for NTD-Si at the DNRR are 
necessary to have a better understanding and 
experience in NTD-Si application. The new 
irradiation rig using various screen materials 
(stainless steel, aluminum and light water) was 
designed, installed and tested in the core of 
DNRR. The results would provide good 
experiences in the application of NTD-Si on 
the new research reactor, which has planned to 
be built in Vietnam. 
DESIGN OF AN IRRADIATION RIG USING SCREEN METHOD FOR  
2 
II. CALCULATION MODEL AND 
EXPERIMENTAL METHOD 
A. Dalat research reactor and its 
parameters 
The DNRR, 500-kW pool-typed, light 
water cooled and moderated, was reconstructed 
and upgraded from the USA 250 kW TRIGA 
MARK II reactor. The reactor has been 
officially put into operation for the purposes of 
radioisotope production, neutron activation 
analysis, fundamental and applied research, 
and manpower training. The summary 
description of the DNRR is shown in Table I 
and Figure 1 [3]. 
Table I. Summary description of the DNRR [3] 
Parameter Description 
Nominal power 500 kW 
Neutron flux (thermal, max) 2 1013 n/cm2.s 
Fuel VVR-M2, mixed UO2-Al, 19,75% enrichment 
Moderator and coolant Light water 
Reflector Graphite, beryllium and light water 
Core cooling Natural convection 
Heat rejection Two-loop cooling system 
Control rods 2 safety, 4 shim (B4C) and 1 regulating (stainless steel) 
Fig. 1. Present working configuration of DNRR. 
Since thermal neutrons are mainly used in 
NTD-Si, higher thermal neutron flux would 
result in a shorter irradiation time. The neutron 
trap, which has maximum thermal neutron flux in 
the core of DNRR, is dedicated for NTD-Si 
testing. Neutron trap is a water cylinder 
surrounded by Beryllium blocks located in center 
of the core. The neutron trap has 6.5 cm in 
diameter and about 2050 cm
3
 in volume [3]. An 
aluminum tube installed in the neutron trap to 
PHAM QUANG HUY et al. 
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load irradiation samples which have maximum of 
4.2 cm in diameter as shown in Figure 1. 
B. Determination of neutron spectrum and 
flux distribution 
Determination of neutron spectrum and 
flux distribution at the neutron trap was 
obtained by using foil activation method. Bare 
gold foils and Cadmium covered gold foils 
were irradiated to obtain absolute neutron flux 
[4]. The following equation can be used to 
obtain thermal neutron flux: 
00,
( , ) ( , )2. .
(1 )
n b Cd
th T
b CdA act th
T A T A TAe
T m mN G e


 
 
 
(1) 
Where, bm - mass of bare gold foil 
(g); Cdm - mass of Cadmium covered gold 
foils (g); T - Irradiation duration (s);  - 
cooling duration after irradiation (s); mt - 
measurement in real time (s); 
,m efft - effective 
time measurement (s);  - decay constant of 
nuclide compound (s
-1
);  - counting 
efficiency of detector;  - gamma abundance 
factor; m - mass of foil (g); - isotope 
enrichment; thG - thermal neutron self-
shielding factor; AN - Avogadro constant; 
A - Atomic number of isotope; G – ratio of 
isotope in foil; Tn neutron temperature (K); T0 
room temperature (293 K). 
To determine the thermal neutron flux 
distribution at the neutron trap, Lu-176 foils 
which have large thermal neutron absorption 
cross-section were used. Thermal neutron 
flux distribution was scaled relatively. The 
relative thermal neutron distribution was 
obtained by comparing of ratios of 
corresponding Lu-176 activities at various 
positions in the neutron trap. 
The method used for neutron spectrum 
measurement is based on activation of a set of 
foils and determination of reaction rates. 
SANDBP program is used to obtain the 
neutron energy spectra after irradiation of the 
foils. The SANDBP program is designed to 
obtain a ’best fit’ neutron spectrum for a given 
input set of infinitely dilute foil activities [4]. 
MCNP5 computer code was also used 
for calculation of neutron flux distribution and 
neutron spectrum. This computer code is 
developed at the Los Alamos National 
Laboratory, USA [5]. The MCNP5 has been 
being officially used for core management of 
DNRR with ENDF/B7.0 library [6]. The 
calculation model for DNRR using MCNP5 
computer code is shown in Figure 2. 
Fig. 2. Calculation model of DNRR using MCNP5 computer code. 
Neutron spectrum divided into 3 
energy groups calculated using MCNP5 
computer code and neutron spectrum obtained 
from experiments at the neutron trap are shown 
in Table II. The reliability of simulation by 
MCNP5 computer code was confirmed through 
DESIGN OF AN IRRADIATION RIG USING SCREEN METHOD FOR  
4 
the good agreement of experimental and 
computational result. The discrepancies are 
about 2% in thermal neutron flux and more 
than 4% in epithermal neutron flux in 
comparison between the experimental and 
computational results. 
Table II. Comparison of neutron flux of the neutron trap between calculation and experiment. 
Flux (n/cm
2
.s) Calculation 
Experiment 
(error ±5-6%) 
Thermal 2.24 x 10
13
 2.29 x 10
13
epithermal 6.52 x 10
12
 6.22 x 10
12
Fast 2.56 x 10
12
 2.64 x 10
12
Fig. 3. Thermal neutron distributions in radial 
direction at the neutron trap (zero position is the 
neutron trap center) 
Fig. 4. Thermal neutron distributions in axial 
direction at the neutron trap 
The experimental and computational 
results in Figure 3 show that the deviation of 
thermal neutron flux distribution in the radial 
direction at the neutron trap ranges within 5-
7%. The experimental result is consistent with 
the computational result for thermal neutron 
flux distribution in the axial direction as shown 
in Figure 4. The maximum thermal neutron 
flux value is located at 20 cm away from the 
core bottom. The shift of maximum flux to the 
bottom of the neutron trap is mainly affected 
by control rod positions. With a large 
discrepancy of the thermal neutron distribution 
in both radial and axial directions, the current 
neutron trap needs to be redesigned to meet the 
requirements of NTD-Si test. 
2.3. Design and testing of irradiation rig 
using flux screen. 
The flux screen method has been 
selected for the purpose of NDT-Si testing in 
DNRR because of its consistent with 
characteristics design of DNRR core 
configuration. The principle of flux screen 
method to flatten the neutron flux distribution 
is shown in Figure 5. A uniformity of the 
irradiation flux is achieved by using screens 
from different materials to absorb neutron or 
change the appropriate thickness of neutron 
absorbers. The screens are made of strong 
neutron absorbers in high neutron flux region 
(stainless steel) and weak absorbers in low flux 
region (aluminum). 
1.8
1.9
2
2.1
2.2
2.3
-2.5 -2 -1.5 -1 -0.5 0 0.5 1 1.5 2 2.5
Th
er
m
al
 n
eu
tr
o
n
 f
lu
x 
 (
x1
0
^3
 n
/c
m
2
.s
) 
 Possition (cm) 
Calculation
(MCNP)
Experiment
0
0.2
0.4
0.6
0.8
1
1.2
0 10 20 30 40 50 60 70
R
el
at
iv
e 
u
n
it
Position (cm) 
Calculation
(MCNP)
Experiment
PHAM QUANG HUY et al. 
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Fig. 5. Flux screen method [1]. 
Dimensions of the irradiation rig and 
screen materials used to flatten flux 
distribution has been calculated and modified 
repeatedly by comparing calculation results 
using MCNP computer code and experiment 
results. Figure 6 shows details of the design 
and materials of irradiation rig. 
Fig. 6. The design of the irradiation rig for NTD-Si testing using flux screen method 
Fig. 7. Silicon ingot (left hand side) and 
aluminum ingot (right hand side) 
Fig. 8. Aluminum ingots with Lu foils 
The important parameters related to 
nuclear safety and radiation safety issues 
such as reactivity and radiation dose rate 
have been calculated before installing the 
irradiation rig in the neutron trap [7], [8]. 
Then, the experiment using aluminum ingots 
were performed. The aluminum ingots have 
similar neutronic characteristics and 
dimension of testing silicon ingots were 
inserted in the rig. Each ingot has a diameter 
of 4 cm and 2 cm in length. Lu-176 foils 
were attached on aluminum ingots and 
irradiated for the determination of thermal 
neutron flux distribution in the rig as shown 
in Figures 7 and 8. 
III. RESULTS AND DISCUSSIONS 
Figure 9 shows calculation results of 
neutron spectra when using water, aluminum 
or silicon in the irradiation rig. The fast 
neutron flux is higher than that in case of 
replacing of water volume in the rig by 
aluminum or silicon ingots. 
Experimental results and calculations 
presented in Table III show that the deviation of 
thermal flux in axial direction at the irradiation 
region is within 5%. The distribution of thermal 
neutron flux in experiment is consistent with the 
calculation results. The thermal neutron flux 
peaks at 10 cm away from the bottom of the 
DESIGN OF AN IRRADIATION RIG USING SCREEN METHOD FOR  
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irradiation rig in the experiment and about 12 cm 
in the calculation. The flux distribution in the top 
half from the 30-34 cm of the irradiation rig tends 
to increase but still below the desired value of 5% 
in both calculation and experiment. These results 
show that the irradiation region of 10 cm to 36 
cm in axial direction of the irradiation rig is 
suitable for NTD-Si test. 
Fig. 9. Calculation of neutron spectra with water, aluminum or silicon in the irradiation rig 
Table III. The relative distribution of thermal neutron flux in axial direction at the irradiation rig 
using flux screen method 
Position (cm) 
Experiment 
(error ±5-6%) Calculation 
10 1.000 0.991 
12 0.992 1.000 
14 0.999 0.995 
16 0.991 0.987 
18 0.960 0.972 
20 0.980 0.974 
22 0.967 0.969 
24 0.968 0.955 
26 0.979 0.962 
28 0.961 0.951 
30 0.958 0.953 
32 0.955 0.968 
34 0.960 0.971 
36 0.948 0.954 
Table IV and Table V present the 
deviation of thermal neutron flux in radial 
direction at the irradiation rig in calculation 
and experiment. The deviation is about 2% in 
experiment and 1% in the calculation for 
each position in the rig. The distribution of 
thermal neutron flux in radial direction in 
experiment is consistent with calculation 
result. The maximum thermal neutron flux 
value drops to 1.31×10
13
 n/cm
2
.s in the 
irradiation rig due to combination of the 
absorption effects from the screen layers and 
the replacement of moderator by aluminum 
ingots in the irradiation rig. These results 
confirm that the irradiation rig using flux 
screen is capable of NTD-Si testing with the 
neutron flux uniformity in both radial and 
axial direction are less than 5%. 
0.0
0.2
0.4
0.6
0.8
1.0
1.2
1.E-10 1.E-08 1.E-06 1.E-04 1.E-02 1.E+00
R
el
a
ti
v
e 
u
n
it
Energy (MeV) 
PHAM QUANG HUY et al. 
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Table IV. Calculation result of thermal neutron flux distribution in radial direction 
at the irradiation rig using flux screen 
Flux ×10
13
 n/cm
2
.s 
Position (cm) Left side (-1.9 cm) Center (0 cm) Right side (+1.9 cm) 
10 1.30 1.30 1.31 
20 1.26 1.28 1.27 
30 1.25 1.26 1.25 
Table V. Experiment result of thermal neutron flux distribution in radial direction 
at the irradiation rig using flux screen 
Relative unit 
Position (cm) Left side (-1.9 cm) Center (0 cm) Right side (+1.9 cm) 
10 0.98 1.00 0.99 
20 0.96 0.98 0.97 
30 0.95 0.96 0.96 
Since fast neutron could cause defects in 
the Silicon crystal, the thermal/fast flux ratio of the 
irradiation channels using for NTD-Si should be at 
least 7/1 [9]. The calculation of the thermal/fast 
flux ratio of the irradiation rig using flux screen, 
however, is 4.38/1. The target resistivity of testing 
Silicon ingots would be affected by this low 
thermal/fast flux ratio after irradiation. In order to 
archive an acceptable uniformity, the silicon ingots 
in the irradiation area of low thermal/fast flux ratio 
of the rig should be replaced by dummy ingots 
(aluminum). 
During the irradiation time, the change 
of power level and control rod positions of the 
reactor would affect the silicon irradiation 
condition. According to the operational 
workbook and lookup table of excess reactivity 
of the reactor, the power level and changing in 
control rod positions would cause an error less 
than 3% of the irradiation condition [7], [10]. 
IV. CONCLUSIONS 
The irradiation rig using various screen 
materials has been designed and installed for 
NTD-Si testing at DNRR. The test with 
aluminum ingots, which have similar neutronic 
characteristics with silicon ingots, was carried 
out to confirm the appropriate design of the 
irradiation rig. The calculated and experimental 
results show that the uniformity of thermal 
neutron flux in the irradiation rig is less than 
5% in height and 2% in radius, respectively. 
However, the low thermal/fast flux ratio of the 
irradiation rig would affect target resistivity of 
testing Silicon ingots after irradiation. The 
processes of designing, installing and testing of 
the irradiation rig would provide good 
experiences in the application of NTD-Si on 
the new research reactor, which has planned to 
be built in Vietnam. 
REFERENCES 
[1]. International Atomic Energy Agency - 
TECDOC-1681, “Neutron Transmutation 
Doping of Silicon at Research Reactors”, 
IAEA, Vienna, 2012. 
[2]. Phan An et al., “Research on using neutron of the 
Dalat Nuclear Research Reactor for doping single 
silicon crystals used in semiconductor 
components”, Report in Vietnamese, Z181, 1988. 
DESIGN OF AN IRRADIATION RIG USING SCREEN METHOD FOR  
8 
[3]. Dalat Nuclear Research Institute, “Safety 
Analysis Report”, 2012. 
[4]. International Atomic Energy Agency, 
“Regional Training Course on calculation and 
measurement of neutron flux spectrum for 
research reactors”, Serpong, Indonesia, 1993. 
[5]. X-5 Monte Carlo Team, “MCNP, A General 
Monte Carlo N-Particle Transport Code, 
Version 5", Los Alamos National Laboratory 
report LA-UR-03-1987, April 2003. 
[6]. M. B. Chadwick et al., “ENDF/B-VII.0: Next 
Generation Evaluated Nuclear Data Library for 
Nuclear Science and Technology", Nuclear 
Data Sheets, 107, 2006. 
[7]. Dalat Nuclear Research Institute, “Lookup 
table of excess reactivity of the Dalat Nuclear 
Research Reactor”, Internal Report in 
Vietnamese, 2018. 
[8]. Dalat Nuclear Research Institute, “Experiment 
and calculation results on the designed 
irradiation rigs for testing NTD-Si”, Internal 
Report in Vietnamese, 2018. 
[9]. International Atomic Energy Agency, “Silicon 
Transmutation Doping Techniques and 
Practices”, Proc. IAEA Consultants Mtg, 
Otwock-Świerk, Poland, 1985. 
[10]. Dalat Nuclear Research Institute, “Operational 
Workbook of the Dalat Research Reactor”, 
Workbook in Vietnamese, 2018.