Design of an irradiation rig using screen method for silicon transmutation doping at the Dalat research reactor
Nuclear Science and Technology, Vol.9, No. 1 (2019), pp. 01-08
Design of an irradiation rig using screen method for silicon
transmutation doping at the Dalat research reactor
Pham Quang Huy, Nguyen Nhi Dien, Nguyen Kien Cuong,
Tran Quoc Duong, Vo Doan Hai Dang, Trang The Dat
Dalat Nuclear Research Institute, Vietnam Atomic Energy Institute
01 Nguyen Tu Luc, Dalat city, Vietnam
Email: [email protected]
(Received 06 May 2019, accepted 19 June 2019)
Abstract: The neutron transmutation doping of silicon (NTD-Si) at research reactors has been
successfully implemented in many countries to produce high-quality semiconductors. In the late
1980s, NTD-Si has been tested at the Dalat Nuclear Research Reactor (DNRR) but the results have
been limited. Therefore, the design and testing of an irradiation rig for NTD-Si at the DNRR are
necessary to have a better understanding in order to apply the NTD-Si in a new research reactor of the
Research Centre for Nuclear Science and Technology (RCNEST), which has planned to be built in
Viet Nam. This paper presents the design and testing of a new irradiation rig using screen method for
testing NTD-Si at the DNRR. The important parameters in the rig such as neutron spectrum and
thermal neutron flux distribution were determined by both calculation using MCNP5 computer code
and experiment. The aluminum ingots, which have similar neutronic characteristics with silicon
ingots, were irradiated in the rig to verify the appropriate design. The uniformity of thermal neutron
flux in the rig is less than 5% in axial and 2% in radial directions, respectively. However, the
thermal/fast flux ratio of the irradiation rig is 4.38/1 would affect target resistivity of testing Silicon
ingots after irradiation.
Keywords: NTD-Si, MCNP5, Dalat Nuclear Research Reactor (DNRR), screen method, foil
activation, irradiation rig.
Depending on the design characteristics and
core configuration of the reactors, the method
applied for neutron flux uniformity should be
chosen. In the late 1980s, the inversion method
has been tested at the DNRR but the results
have been limited [2]. Therefore, 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.
I. INTRODUCTION
One of the most important requirements of
NTD-Si is to maintain the uniformity of radial
and axial thermal neutron flux in irradiation
channels. At present, the uniformity of irradiation
from 5 ~ 6% would meet commercial request
depend on ingot dimension [1]. In a research
reactor, normally, the axial and radial
distributions of thermal neutron in irradiation
channels are not uniformed. In order to meet
customer’s requirements,
a
change in
configuration of irradiation channels is needed to
uniform thermal neutron flux distribution and
maximize usage of the channel. Three major
methods could be applied to axially uniform
thermal neutron flux distribution are inversion,
reciprocating motion and flux screen [1].
©2019 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute
DESIGN OF AN IRRADIATION RIG USING SCREEN METHOD FOR …
MARK II reactor. The reactor has been
II. CALCULATION MODEL AND
EXPERIMENTAL METHOD
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].
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
Table I. Summary description of the DNRR [3]
Parameter
Description
Nominal power
500 kW
21013 n/cm2.s
Neutron flux (thermal, max)
Fuel
VVR-M2, mixed UO2-Al, 19,75% enrichment
Light water
Moderator and coolant
Reflector
Graphite, beryllium and light water
Natural convection
Core cooling
Heat rejection
Control rods
Two-loop cooling system
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 cm3 in volume [3]. An
aluminum tube installed in the neutron trap to
2
PHAM QUANG HUY et al.
load irradiation samples which have maximum of
4.2 cm in diameter as shown in Figure 1.
A
- Atomic number of isotope; G – ratio of
isotope in foil; Tn neutron temperature (K); T0
room temperature (293 K).
B. Determination of neutron spectrum and
flux distribution
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.
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:
2.A.e
Tn A (T,) ACd (T,)
b
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].
th
NA0,actG (1eT
)
T0
mb
mCd
th
(1)
mb
Where,
- mass of bare gold foil
mCd
(g);
- mass of Cadmium covered gold
foils (g);
T
- Irradiation duration (s);
tm
-
-
cooling duration after irradiation (s);
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.
measurement in real time (s); tm,eff - effective
time measurement (s);
- decay constant of
nuclide compound (s-1);
- counting
efficiency of detector; - gamma abundance
factor; - mass of foil (g); - isotope
enrichment; - thermal neutron self-
shielding factor;
m
Gth
NA
- Avogadro constant;
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
3
DESIGN OF AN IRRADIATION RIG USING SCREEN METHOD FOR …
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.
Experiment
Flux (n/cm2.s)
Calculation
(error ±5-6%)
2.29 x 1013
6.22 x 1012
2.64 x 1012
Thermal
epithermal
Fast
2.24 x 1013
6.52 x 1012
2.56 x 1012
2.3
2.2
2.1
2
1.2
Calculation
(MCNP)
Experiment
1
0.8
0.6
0.4
0.2
0
Calculation
(MCNP)
Experiment
1.9
1.8
-2.5 -2 -1.5 -1 -0.5 0 0.5 1 1.5 2 2.5
0
10 20 30 40 50 60 70
Position (cm)
Possition (cm)
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).
4
PHAM QUANG HUY et al.
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. 5. Flux screen method [1].
Fig. 6. The design of the irradiation rig for NTD-Si testing using flux screen method
Fig. 7. Silicon ingot (left hand side) and
Fig. 8. Aluminum ingots with Lu foils
aluminum ingot (right hand side)
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
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.
similar
neutronic
characteristics
and
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
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.
5
DESIGN OF AN IRRADIATION RIG USING SCREEN METHOD FOR …
irradiation rig in the experiment and about 12 cm
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.
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%
1.2
1.0
0.8
0.6
0.4
0.2
0.0
1.E-10 1.E-08 1.E-06 1.E-04 1.E-02 1.E+00
Energy (MeV)
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
Experiment
Position (cm)
(error ±5-6%)
1.000
0.992
0.999
0.991
0.960
0.980
0.967
0.968
0.979
0.961
0.958
0.955
0.960
0.948
Calculation
0.991
1.000
0.995
0.987
0.972
0.974
0.969
0.955
0.962
0.951
0.953
0.968
0.971
0.954
10
12
14
16
18
20
22
24
26
28
30
32
34
36
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 1.31×1013 n/cm2.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%.
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PHAM QUANG HUY et al.
Table IV. Calculation result of thermal neutron flux distribution in radial direction
at the irradiation rig using flux screen
Flux ×1013 n/cm2.s
Position (cm)
10
Left side (-1.9 cm)
Center (0 cm)
1.30
Right side (+1.9 cm)
1.30
1.26
1.25
1.31
1.27
1.25
1.28
20
30
1.26
Table V. Experiment result of thermal neutron flux distribution in radial direction
at the irradiation rig using flux screen
Relative unit
Position (cm)
10
Left side (-1.9 cm)
Center (0 cm)
1.00
Right side (+1.9 cm)
0.98
0.96
0.95
0.99
0.97
0.96
0.98
20
30
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).
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.
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].
REFERENCES
[1]. International Atomic Energy Agency
-
TECDOC-1681, “Neutron Transmutation
Doping of Silicon at Research Reactors”,
IAEA, Vienna, 2012.
IV. CONCLUSIONS
[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.
The irradiation rig using various screen
materials has been designed and installed for
NTD-Si testing at DNRR. The test with
7
DESIGN OF AN IRRADIATION RIG USING SCREEN METHOD FOR …
[3]. Dalat Nuclear Research Institute, “Safety
Analysis Report”, 2012.
[7]. Dalat Nuclear Research Institute, “Lookup
table of excess reactivity of the Dalat Nuclear
Research Reactor”, Internal Report in
Vietnamese, 2018.
[4]. International
Atomic
Energy
Agency,
“Regional Training Course on calculation and
measurement of neutron flux spectrum for
research reactors”, Serpong, Indonesia, 1993.
[8]. Dalat Nuclear Research Institute, “Experiment
and calculation results on the designed
irradiation rigs for testing NTD-Si”, Internal
Report in Vietnamese, 2018.
[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.
[9]. International Atomic Energy Agency, “Silicon
Transmutation Doping Techniques and
Practices”, Proc. IAEA Consultants Mtg,
Otwock-Świerk, Poland, 1985.
[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.
[10].Dalat Nuclear Research Institute, “Operational
Workbook of the Dalat Research Reactor”,
Workbook in Vietnamese, 2018.
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