Assessment of the Absorbed Dose Components of BNCT Method at the Dalat Research Reactor

International Journal of Scientific Research and Engineering Development-– Volume 7 Issue 3, May-June 2024
Available at www.ijsred.com
ISSN : 2581-7175 ©IJSRED:All Rights are Reserved Page 2527
Assessment of the Absorbed Dose Components of BNCT Method at the
Dalat Research Reactor
Dang Quyet Pham*
*Nuclear Research Institute, 01 Nguyen Tu Luc, Dalat, Vietnam
Email: quyetpd@gmail.com
Thi Tu Anh Trinh**
**Office of National Assembly Delegations and People Councils, 02 Tran Hung Dao, Dalat, Vietnam
Email: anhttt@dlu.edu.vn
Xuan Hai Pham*
*Nuclear Research Institute, 01 Nguyen Tu Luc, Dalat, Vietnam
Email: phamxuanhai2007@yahoo.com
-------------------------------------------------************************---------------------------------------------------
Abstract:
At the Dalat Nuclear Research Reactor (DNRR) in Vietnam, some calculations and experiments of the
Boron Neutron Capture Therapy (BNCT) method have been performed at horizontal channel No. 2 of the
DNRR using a phantom. This research used the Monte Carlo N-Particle version 5 (MCNP5) code to
simulate and calculate the distribution of absorbed dose components of the BNCT method. The collimator
of horizontal neutron channel No. 2 of the DNRR was changed from cylindrical to conical to increase the
flux of the neutron beam. Simultaneously, neutron crystal filters corresponding to 20 cm Si and 3 cm Bi
are also employed to produce high-purity thermal neutron beams. The thermal neutron flux and absorbed
dose components have been computed in a water phantom. The gamma dose from the reactor core of the
DNRR can be omitted when calculating the total absorbed dose in the BNCT method.
Keywords —BNCT, water phantom, absorbed dose,MCNP, collimator
-------------------------------------------------************************--------------------------------------------------
1. INTRODUCTION
Boron Neutron Capture Therapy (BNCT) will
selectively damage cancer cells that are difficult to
achieve with other treatments. Therefore, BNCT
was suggested as a possibility to treat brain tumors
in 1951 [1-3]. So far, the neutron sources for BNCT
are a thermal nuclear research reactor or an
accelerator [4–5]. For example, the HANARO
reactor in Korea [3] used a combination of single-
crystal Si and Bi filters to generate a thermal
neutron beam for BNCT research because single-
crystal Si and Bi have a relatively small total cross-
section for thermal neutrons. Furthermore, single-
crystal Bi reduces gamma rays mixed in the neutron
beam from the reactor core as well as secondary
gamma rays created by the single-crystal Si filter.
As a result, they are often used to generate pure
thermal neutron beams.Before conducting clinical
trials, preclinical studies are often simulated and
tested on models (phantoms). Two types of
materials commonly used to design phantoms for
BNCT research are water and polyethylene because
the densities of these two materials are almost
similar to that of tissue. Water phantom has been
used experimentally at the Tehran research reactor
(TRR) in Iran[6]. For the last ten years, simulation
calculations and experiments related to the
absorbed dose of the BNCT method have been
performed at horizontal channel No. 2 of the DNRR
in Vietnam. However, the gamma dose (including
the dose of gamma rays from the reactor and
gamma rays produced by the reaction of the
phantom material with neutrons) has not been
calculated in detail. This study provides
information about assessing the absorbed dose
components of the BNCT method using the
MCNP5 code. These include improvements in the
shape of the collimator to increase the thermal
neutron flux and detailed calculations of component
gamma doses.
RESEARCH ARTICLE OPEN ACCESS

ISSN : 2581-7175 ©IJSRED: All Rights are Reserved Page 2528
2. MATERIALS AND METHODS
2.1. Confirm the simulated values for cylindrical
collimator by experiments
The DNRR is a 500 kW pool-type research reactor
with four horizontal neutron channels. Three of the
neutron channels (Nos. 1, 2, and 4) are oriented
radially toward the reactor core's center, whereas
channel No. 3 is tangential to the reactor core's
outer edge [7-8].
Currently, the horizontal channel No.2 design
includes a cylindrical collimator with a total length
of 240.3 cm made upof two parts. The first part is
150.3 cm long and 9 cm in diameter. It installs
neutron filters made of Si and Bi crystals with
thicknesses of 20 cm and 3 cm, respectively. The
second part is 90 cm long with an outer diameter of
20.1 cm and a inner diameter of 3cm. The current
configuration of horizontal channel No. 2 is seen in
Figure 1.
An aluminum plate is used to stop water leaks,
while the lining layers around the collimators are
composed of Pb and WWX-277 (a neutron
shielding material from the Shieldwerx company)
as gamma and neutron absorption materials,
respectively [9].
90 cm 18 cm
D2= 3 cm
Concrete
50.7 cm
Beam port No.2
40 cm
Sample irradiation
position
D1= 9 cm
Water
Graphite
Reactor
core
Si single crystal SWX-277
Air
Pb
Bi single crystal
Al sheet
Stainless steel
Pb
Beam stop
Fe
Water
Graphite
Reactor
core
99.6 cm
Entrance No.2
Fig.1. Structure of horizontal channel No.2 with the cylindrical collimator
As shown in Figure 1, a thermal neutron beam will
be generated after passing through a single-crystal
filter assembly made up of 20 cm Si and 3 cm Bi,
which will be used for a variety of applications
such as nuclear data measurement, neutron
activation analysis, and BNCT research.
First, the MCNP5 code is used to simulate the
structure of horizontal channel No. 2 in Figure 1,
and then the thermal neutron flux and gamma dose
rate in the water phantom were calculated using
tally F4 with the DE4/DF4 cards of MCNP5. The
following equations [10] are used to compute these
values:
∫
=
ϕ
=
φ
eV
5
.
0
E
meV
1
n
th
Cd
dE
)
E
( (1)
∫
=
2
1
)
(
)
(
E
E
n dE
E
R
E
D ϕ
γ
&
(2)
where φthis the thermal neutron flux, and ϕn(E)is the
neutron flux. In this paper, the energy range of
thermal neutrons (0.025 eV) is calculated from
1.0×10-3
eV to 0.5 eV, γ
D
&
is the gamma dose rate
and R(E) is the conversion factor of neutron flux to
gamma dose rate.
Second, we used the activation foils and
Thermoluminescent Dosimeters to measure thermal
neutron flux and gamma, respectively, to validate
the above-simulated and computed values. The
results showed that there was a good agreement
between experimental and simulation results for the
thermal neutron flux and gamma dose rate, as
reported by us [11].
2.2. Change the shape of the collimator to increase
the neutron flux
Finally, we utilized the MCNP5 code to simulate
modifying the form of the neutron collimator to
increase the flux of the neutron beam, which also
increases the total absorbed dose of the BNCT
method, as shown in Eq. (3). The conical collimator
design of horizontal channel No. 2, which was
simulated to increase the neutron flux and calculate
the total absorbed dose for the BNCT application, is
seen in Figure 2. The simulated result of thermal
neutron flux at 2 cm depth in the phantom supplied
by the conical collimator may be enhanced by a
factor of 8.35, as shown in Table 1.

90 cm 18 cm
Concrete
50.7 cm
Bi single crystal
SWX-277
Air
Pb
Si single crystal
Al sheet
Stainless steel
Pb
Fe
D1= 14.6 cm
Water
Graphite
Reactor
core
Beam stop
D2= 5.3 cm
99.6 cm
Fig.2. Structure of horizontal channel No.2 with the new collimator (conical shape)
Table 1. Simulated results of the thermal neutron flux and
gamma dose rate forthe cylindrical and conical collimators
Type of
collimator
Thermal
neutron flux φth
(n.cm2
.s-1
)
Gamma dose
rate γ
D
&
(Gy.h-1
)
Filters
(cm)
Si Bi
Cylinderical
collimator (A)
1.12×107
1.89×10-3
20 3
Conical
collimator (B)
9.35×107
1.90×10-2
20 3
Ratio B/A 8.35 10.05
2.3. Calculation of absorbed dose components
There are four absorbed dose components that are
often studied in BNCT: (i) boron dose, (ii) thermal
neutron dose, (iii) gamma dose, and (iv) fast
neutron dose [12-15]. However, with the current
configuration of horizontal channel No. 2, the fast
neutron dose is extremely low and may be
negligible. Consequently, the following estimate of
the total absorbed dose in BNCT is [13-14]:
γ
+
Φ
×
×
−
×
+
×
−
×
=
γ
+
+
=
D
th
)
N
C
14
10
78
.
6
B
C
14
10
43
.
7
(
D
N
D
B
D
D
(3)
where D is the total absorbed dose; DB is the boron
dose; DN is the thermal neutron dose; Dγ is the
gamma dose, including gamma from the reactor
coreand gamma generated by the interaction of
thermal neutrons with the hydrogen of the water
phantom; CB is the concentration of 10
B (the value
chosen in this calculation is 30 ppm); CN is the
concentration of 14
N (chosen to be 2%)[14], and Φth
is the thermal neutron fluence (n.cm-2
). When
calculating dose in BNCT, the concept of thermal
neutron fluence (Φth) is often used instead of the
concept of thermal neutron flux (φth). The
relationship between these two quantities is
calculated as follow:
t
th
th
×
φ
=
Φ (4)
where t is the time in seconds.
3. Results and Discussion
3.1. Neutron fluxes and gamma doses for the
conical collimator
According to Table 1, when both the shape and the
solid angle of the collimator are changed, the
thermal neutron flux and gamma dose rate increase
approximately 8.35 times and 10.05 times,
respectively. Simulation results for the distribution
of thermal neutron flux and gamma dose in the
water phantom are presented in Tables 2 and 3,
respectively.
Table 2. Thermal simulated neutron flux along the central axis
of the phantom by MCNP5
No.
Position
(cm)
φth (n.cm-2
.s-1
)
Mean Error (%)
1 0 1.46×108
1
2 0.5 1.78×108
1
3 1 1.49×108
1
4 2 9.35×107
1
5 3 5.69×107
1
6 4 3.53×107
1
7 5 2.20×107
1
8 6 1.42×107
1
9 7 9.32×106
1
10 8 6.15×106
2
11 9 4.05×106
2
12 10 2.70×106
2
Table 3. Gamma dose along the central axis of the phantom
simulated by MCNP5

No.
Positi
on
(cm)
Dγ (Gy) Dγ (Gy)
Mean
Error
(%)
Mean
Error
(%)
from reactor, with
phantom
from reactor, without
phantom
1 0 5.99×10-6
1 4.56×10-8
7
2 0.5 8.11×10-6
1 4.64×10-8
7
3 1 8.02×10-6
1 4.93×10-8
7
4 2 6.88×10-6
1 4.47×10-8
7
5 3 5.27×10-6
1 4.36×10-8
7
6 4 4.00×10-6
1 4.47×10-8
7
No.
Positi
on
(cm)
Dγ (Gy) Dγ (Gy)
Mean
Error
(%)
Mean
Error
(%)
from reactor, with
phantom
from reactor, without
phantom
7 5 3.06×10-6
1 4.43×10-8
7
8 6 2.39×10-6
2 4.21×10-8
7
9 7 1.79×10-6
2 4.41×10-8
8
10 8 1.38×10-6
2 4.23×10-8
8
11 9 1.14×10-6
2 3.98×10-8
8
12 10 8.74×10-7
2 4.22×10-8
8
Table 2 and Figure 3 show that the maximum value
of thermal neutron flux is 1.78×108
n.cm-2
.s-1
at 0.5
cm depth and then decreases to 5.69×107
n.cm-2
.s-1
at 3 cm depth in the water phantom.
0 2 4 6 8 10
0.0
2.0x107
4.0x107
6.0x107
8.0x107
1.0x108
1.2x108
1.4x108
1.6x108
1.8x108
2.0x108
Thermal
neutron
flux
(n.cm
-2
.s
-1
)
Depth in phantom (cm)
Fig. 3. Thermal neutron flux distribution in the phantom
Figure 4 displays the two-dimensional distribution
of the thermal neutron flux in the phantom. The
neutron flux area accounts for approximately 87%
(with a value of 1.25×108
n.cm-2
.s-1
) of the total
neutron flux, which is dispersed mostly from the
phantom surface to a depth of about 3 cm and
drastically decreases at depths higher than 5 cm in
the phantom.
0 2 4 6
0
2
4
6
Radial distance from center of beam port (cm)
Depth
in
phantom
(cm)
2.50E+06
2.45E+07
4.65E+07
6.85E+07
9.05E+07
1.13E+08
1.35E+08
1.57E+08
1.79E+08
Thermal neutron flux
(n.cm-2.s-1)
Fig. 4. 2-D thermal neutron flux distribution in the phantom
3.2. Absorbed dose components of BNCT method
After obtaining the simulated values of thermal
neutron flux and gamma dose, as indicated in
Tables 2 and 3, respectively, the total absorbed dose
of the BNCT method has been calculated by
applying Eqs. (3) and (4). The results of these
calculations are reported in Table 4.
Table 4. Calculated results of the total absorbed dose in the water phantom for the conical collimatior
No.
Position
(cm)
Dose (Gy)
Boron dose
DB
Thermal neutron
dose
DN
Gamma dose
Dγ (with
phantom)
Gamma dose
Dγ (without
phantom)
Total absorbed
dose
D
1 0 3.25×10-4
1.98×10-5
5.99×10-6
4.56×10-8
3.51×10-4
2 0.5 3.97×10-4
2.41×10-5
8.11×10-6
4.64×10-8
4.29×10-4
3 1 3.32×10-4
2.02×10-5
8.02×10-6
4.93×10-8
3.60×10-4
4 2 2.08×10-4
1.27×10-5
6.88×10-6
4.47×10-8
2.28×10-4

No. Position
(cm)
Dose (Gy)
Boron dose
DB
Thermal neutron
dose
DN
Gamma dose
Dγ (with
phantom)
Gamma dose
Dγ (without
phantom)
Total absorbed
dose
D
5 3 1.27×10-4
7.72×10-6
5.27×10-6
4.36×10-8
1.40×10-4
6 4 7.87×10-5
4.79×10-6
4.00×10-6
4.47×10-8
8.75×10-5
7 5 4.90×10-5
2.98×10-6
3.06×10-6
4.43×10-8
5.51×10-5
8 6 3.17×10-5
1.93×10-6
2.39×10-6
4.21×10-8
3.60×10-5
9 7 2.08×10-5
1.26×10-6
1.79×10-6
4.41×10-8
2.39×10-5
10 8 1.37×10-5
8.34×10-7
1.38×10-6
4.23×10-8
1.60×10-5
11 9 9.03×10-6
5.49×10-7
1.14×10-6
3.98×10-8
1.08×10-5
12 10 6.02×10-6
3.66×10-7
8.74×10-7
4.22×10-8
7.30×10-6
Note: The concentrations of boron (CB) and nitrogen (CN) used are 30 ppm and 2%, respectively.
The reactions captured thermal neutrons from
hydrogen in the water phantom H
γ)
H(n, 2
1
, which
mainly contributed to the gamma dose in the water
phantom, as shown in Table 3 and Figure 5.
0 2 4 6 8 10
4x10-8
5x10-8
1.0x10-6
2.0x10-6
3.0x10-6
4.0x10-6
5.0x10-6
6.0x10-6
7.0x10-6
8.0x10-6
9.0x10-6
Dose
(Gy)
Gamma dose from reactor, with phantom
Gamma dose from reactor, without phantom
Fig. 5. Gamma dose distribution by in the phantom
The gamma radiation at the phantom's surface is
enhanced approximately 130 times compared to
that without the phantom (from 4.56×10-8
Gy to
5.99×10-6
Gy) but about three times lower than the
thermal neutron dose, as presented in Figure
6.Furthermore, Figures 3 and 5 indicate that the
curves of thermal neutron flux and gamma dose
have the same shape, which means the thermal
neutron capture reaction of hydrogen mainly
contributed to the gamma dose.The MCNP5
calculation also revealed that the thermal neutron
dose was approximately three times greater than the
gamma dose (Figure 6). Our results also indicate
that the gamma dose is the most significant
difference when using the horizontal channel or the
thermal column of the reactor for BNCT research.
It is smaller than the thermal neutron dose for the
horizontal channel at DNRR in Figure 7(a), but not
for the thermal column of a reactor at TRR in
Figure 7(b).
0 2 4 6 8 10
0.0
2.0x10-6
4.0x10-6
6.0x10-6
8.0x10-6
1.0x10-5
1.2x10-5
1.4x10-5
1.6x10-5
1.8x10-5
2.0x10-5
2.2x10-5
2.4x10-5
2.6x10-5
CN = 2 %
Dose
(Gy)
Thermal neutron dose
Gamma dose from reactor, with phantom
Fig. 6. Gamma and thermal neutron doses distribution in
phantom
The absorbed dose components along the central
axis of the water phantom at the DNRR are shown
in Figure 7(a). The total absorbed dose, reaching a
maximum of 4.24×10-4
Gy at 0.5 cm depth and
rapidly decreasing to 1.36×10-4
Gy at 3 cm depth in
the phantom, is mainly due to the boron and
thermal neutron doses. Our results also show a
relative agreement between the shapes of absorbed
dose components in the phantom simulated at the
DNRR and the results measured in the phantom at
the TRR in Iran in Figure 7(b).

0 2 4 6 8 10
0.0
5.0x10-5
1.0x10-4
1.5x10-4
2.0x10-4
2.5x10-4
3.0x10-4
3.5x10-4
4.0x10-4
4.5x10-4
(a)
CB = 30 ppm, CN = 2 %
Dose
(Gy)
Total absorbed dose
Boron dose
Thermal neutron dose
Gamma dose
Fig. 7. Distribution of absorbed dose components in the phantom: (a) at DNRR; (b) at TRR
4. Conclusion
The MCNP5 code was used to assess the
contribution of absorbed dose components in a
water phantom of BNCT research at the DNRR.
The total absorbed dose depends mainly on the
boron and thermal neutron doses. The gamma dose
from the reactor core of the DNRR contributes very
little compared to the gamma dose generated from
the phantom material, and it can be omitted when
calculating the total absorbed dose in the BNCT
method.
Acknowledgements
The authors would like to thank Dr. Pham Ngoc
Son of Dalat Nuclear Research Institute for his
invaluable assistance regarding experimental
measurements on horizontal channel No. 2 of the
reactor to determine neutron flux and gamma dose
rate.
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Competing Interest: The authors have declared
that no competing interest exists.
Ethical approval: This study does not contain any
studies with human or animal subjects performed
by any of the authors.
Author Contributions: The first draft of the
manuscript was written by Pham Dang Quyet and
all authors commented on previous versions of the
manuscript. All authors read and approved the final
manuscript.

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