MINISTRY OF SCIENCE AND

MINISTRY OF EDUCATION AND

TECHNOLOGY

TRAINING

VIETNAM ATOMIC ENERGY INSTITUTE

Summary of Doctoral Thesis

STUDY FOR EVALUATION OF DOSE COMPONENTS
FOR BNCT RESEARCH ON THE HORIZONTAL
CHANNEL OF THE DALAT REACTOR

Author:
Pham Dang Quyet

Supervisor:
Ass. Prof. Dr. Nguyen Nhi Dien
Dr. Trinh Thi Tu Anh

A thesis submitted in fulfillment of the requirements
for the degree of Doctor of Philosophy

Hanoi – 2020


Công trình được hoàn thành tại: Viện Nghiên cứu hạt nhân-Viện Năng lượng
nguyên tử Việt Nam

with the cells. In which, Neutron Capture Therapy (NCT) is a technique that was designed to
destroy tumors at the cellular level. Some elements such as 10B, 6Li, 157Gd, and 235U can be
used in NCT. However, 10B can be used for treatment of brain cancer with a concentration of
about 30-60 ppm.
After Goldhaber discovered the unusually large thermal neutron capture cross-section of
the naturally occurring isotope 10B in 1934. In 1936 Locher proposed the idea of the Boron
Neutron Capture Therapy (BNCT) method (Fig. 1.2) and that was suggested to the treatment
of brain tumors in 1951.
Important physical problems in the BNCT method that need to be considered, include (i)
designing neutron channels; (ii) calculation, simulation and experimental measurement to
identify neutron flux distribution characteristics, neutron dose, and gamma dose in a phantom
model; (iii) calculation of dose components from neutron capture reaction in BNCT based on
information of neutron energy spectrum; (iv) development of analytical techniques to quantify
boron concentration during radiotherapy.
In BNCT, the absorbed dose will consist of 4 dose components of interest, namely: (i)
the boron dose; (ii) the thermal neutron dose; (iii) the fast neutron dose; and (iv) the gamma
dose. However, the absorbed dose mainly consists of the first 2 dose components and can only
be determined indirectly through neutron flux and concentration of 10B nuclide. In which, the
thermal neutron flux is usually determined by the method of Neutron Activation Analysis
(NAA), and the concentration of 10B is determined by the method of Prompt Gamma Neutron
Activation Analysis (PGNAA).
Historically, the best neutron sources with the energy and flux levels required for BNCT
were extracted from nuclear research reactors by the following methods: (i) neutron spectrum
shifting by re-arranging shielding materials or (ii) using filters, which is commonly used to
create a mono-energetic neutron beam not only for BNCT but also for many other research
purposes.
The improvement of the design of horizontal beam ports or thermal columns of the
research reactor to extract thermal neutron beam for BNCT research was often calculated and
simulated by some typical codes, such as DORT, MacNCTPLAN, SERA, MCNP (Monte
Carlo N–Particle), etc. However, MCNP is still the most commonly used code. In Korea, in

include: (i) studying and calculating the absorbed dose of reaction 10 B(n, α) 7 Li in BNCT; (ii)
simulating the absorbed dose distribution of BNCT in a water phantom at CN2DR using
MCNP5 code; (iii) determining the absorbed dose distribution in BNCT with water phantom
using CN2DR, and (iv) proposal of an optimal configuration design for the BNCT facility at
CN2DR with thermal neutron flux at the entrance of phantom more than 1×108 n.cm-2.s-1 and
the ratio of gamma dose rate to thermal neutron flux less than 3×10-13 Gy.cm2.n-1.
Scientific and practical significance
The results of the thesis are of scientific significance for the first time approaching and
researching physics in the BNCT method in Vietnam using the neutron channel of Dalat
reactor, providing new information about design improvement for neutron throughput at the
sample irradiation position, simulation and experiment results of neutron flux distribution and
dose components in phantom, making a meaningful contribution to knowledge and premise
development research for the application of BNCT in Vietnam in the future.
The practical significance of the thesis is that the research results of neutron beam
design improvement have proven to increase neutron flux at the experimental position of
CN2DR to 12 times, thereby contributing to increasing effective exploitation of horizontal
channels of the Dalat reactor. In addition, the results of the thesis also contributed
significantly to improving the capacity of simulation research and experimental measurement
in the field of neutron physics and related applications on neutron beams from the reactor.
The results of the thesis also have practical implications when serving the training and
development of nuclear human resources.
Thesis structure
The structure of the thesis consists of 3 chapters. Chapter 1 presents an overview of the
method for calculating the absorbed dose in BNCT, including the principle of BNCT, the dose
components generated in BNCT, neutron KERMA factor for the elements in the tissue, the
NAA method to determine thermal neutron flux, and phantom for BNCT. Chapter 2 presents
the simulation of the thermal neutron flux distribution in a water phantom using MCNP5
code, experiments at CN2DR, including design of the water phantom, set-up the measuring
system, determining the distribution of thermal neutron flux in the water phantom, building a
standard curve of boron concentration in solution samples, measuring gamma dose rate in

1.3.1. Interaction cross-section of neutron
This section presents concepts and types of interaction cross-sections of the neutron.
1.3.2. Neutron KERMA factor in tissue
The neutron field is often described in the term flux φ(E), when a mono-energetic
neutron beam interacts with a nucleus in the tissue, the released energy of the reaction per unit
of mass (Kinetic Energy Released per unit Mass – KERMA), defined by the expression:
N 
(1.5)
K = σ × t × E ×Φ
 m
N 
In which σ is the neutron cross-section,  t  is the number of nuclei of the interest isotope
m

in mass m, and E is the average energy transfer in the nuclear reaction.
Table 1.3 lists the KERMA factors for thermal neutrons (neutron KERMA) of the
elements in the tissue.
Table 1.3. KERMA factors for the thermal neutrons of the elements in tissue
Concentration
KERMA factor
Ratio
No.
Element
(%)
(Gy.cm2)
(%)
1
H
10.7
4.49E-15

3


1.4.2. The dose components in BNCT
In BNCT, there are four components of interested absorbed dose: (i) the boron dose; (ii)
the thermal neutron dose; (iii) the fast neutron dose; and (iv) the gamma dose.
(i)

The boron dose (DB) generated from the reaction
follows:
4
10

B+1 n (0.025 eV) →11 B* 〈

10

B(n, α) 7 Li is expressed as

He+ 7 Li + γ (0.478 MeV) + 2.31 (MeV)
He + 7 Li + 2.79 (MeV)

4

(94%)
(6%)

The boron dose is calculated by the formula:
D B = 1.6 × 10−13 × C B × σ B × Q × Φ th


D N = 6.78 × 10−14 × C N × Φ th
(1.12)
-14
In which, the value 6.78×10 is the KERMA factor of nitrogen for thermal neutron, CN (%).
(iii) The fast neutron dose (Df) generated from recoil protons released during elastic
scattering, which occurs when the fast neutrons interact with hydrogen in reaction 1 H(n, n ' )1 H .
The fast neutron dose is calculated by the formula:
(1.13)
D f = 1.6 × 10−13 × C H × σ sH × E f × Φ f × f
where σsH is the elastic scattering cross-section between fast neutron and hydrogen (cm2), Ef
is the released energy of the reaction (MeV), Φ f is the fast neutron fluence (n.cm-2), và f = 0.5
is the absorption coefficient for fast neutron in tissue.
(iv) The gamma dose ( D γ ) generated by gamma rays formed in the reaction 1 H(n, γ) 2 H and
gamma rays mixed in the incoming neutron beam (through interactions of the neutron beam
or from the reactor core). The gamma dose in tissue generated as primarily results when the
hydrogen in the tissue absorbs thermal neutron in reaction 1 H(n, γ) 2 H and calculated by the
formula:
D γ (2.22) = 1.6 × 10 −13 × C H × σ H × E γ ( 2.22) × Φ th × f γ ( 2.22)
(1.14)
14N

in which f γ ( 2.22) = 0.278 is the whole body's absorption coefficient for 2.22 MeV gamma-ray.
Combined with the concentration of 1H, the reaction cross-section, and the gamma
absorption coefficient in tissue, equation (1.14) is rewritten as follows:
D γ (2.22) = 1.0 × 10 −14 × Φ th
(1.15)
4


where the value 1.0×10-14 is the KERMA factor of 2.22 MeV gamma-ray for the volume that

Reactor
(MW)
Si
Bi
MURR
10
50
8
HANARO
30
40
15

1.5.2. Phantom
This section presents the selection of materials for phantom in BNCT research. The two
materials commonly used to make phantom are water and polyethylene because the density of
these two materials is almost similar to tissue.
1.5.3. Determination of thermal neutron flux by NAA technique
This section presents the theoretical basis of the neutron activation method and formula
to calculate thermal neutron flux by the NAA technique. The thermal neutron flux of an
activated foil can be determined by the equation:
C× f ×λ
(1.26)
φ=
ε × I × N × σ 0 × (1 − e − λt1 )× e − λt 2 × 1 − e − λt 3
The main source of error when calculating thermal neutron flux is caused by the number
of counts in gamma peak of interest (C) and gamma full-peak efficiency (ε). Therefore, the
relative error and the absolute error of thermal neutron flux are calculated by the formulas
(1.27) and (1.28):


This section presents an overview of the literature on the use of a TLD to determine the
gamma dose in the BNCT research. In which, the TLD-900 (CaSO4: Dy) will be a good
choice because this dosimeter has a very high sensitivity to gamma rays. Errors for some
types of TLDs are listed in Table 1.10.
No.
1
2
3
4

Table 1.10. Errors of some types of TLDs
Dosimeter
Material
TLD-300
CaF2:Tm
6
TLD-600
LiF:Mg,Ti
7
TLD-700
LiF:Mg,Ti
TLD-900
CaSO4:Dy

Error (%)
30
5.1
5.1
6


criteria after each retrieval of results. In addition, to evaluate the accuracy of R, one uses the
Variance of Variance (VOV), the value of VOV must be less than 0.1 for all types of Tally.
6


1.6.5. Simulation and calculation of absorbed dose in BNCT
In the world, most research centers including BNCT research have proposed a treatment
plan based on the Monte Carlo method to calculate the dose distribution in BNCT. Some
results of the comparison between simulation and experiment for neutron flux in a water
phantom at HFR reactor are shown in Fig. 1.22.

Relative neutron flux (%)

Relative neutron flux (%)

120

Exp.
100

Simulation

80

Center line
60
40
20
0


-4

-2

0

2

4

6

Distance from beam axis (cm)

Fig. 1.22. The neutron flux distribution in water phantom by simulation and experiment at
HFR reactor (Netherlands)

1.6.6. Design of a neutron beam for BNCT
This section presents the use of MCNP to design neutron beams for BNCT by
Matsumoto in 1996 (Japan), by Monshizadeh in 2015 (Iran), and two basic parameters in
thermal neutron beam design for BNCT research.
Table 1.15. Basic parameters in the design of thermal neutron beam for BNCT research
Dgamma/φth
φth
Power
No.
Reactor
(MW)
(×109 n.cm-2.s-1)
(×10-13 Gy.cm2.n-1)


φth (n.cm .s )

Cadmi ratio
RCd(Au)

1.6×106

420

-2

-1

Filter length (cm)
Si

Bi

80

4

The diameter of
the neutron beam
(cm)
3

1.8. Summary of chapter 1
Although there are 4 dose components generated during radiotherapy by BNCT, only 2

Graphite

Air

SWX-277
Al sheet

Pb

D1= 9 cm
Reactor
core

Sample irradiation
position
D2= 3 cm

Beam port No.2
99.6 cm

50.7 cm

90 cm

Concrete

18 cm

Stainless steel



-2 -1

Neutron flux (n.cm .s )

1E+08
1E+07
1E+06
1E+05
1E+04
1E+03
1E-9

1E-8

1E-7

1E-6

Energy (MeV)

Fig. 2.5. The spectrum shape at the sample irradiation position at CN2DR
8


It can easily realize that after passing through a filter combination of 20 cm Si and 3 cm
Bi single-crystal, an obtained thermal neutron spectrum (En < 0.414 eV) has a high-purity,
and reaches a maximum value of about 2.5×107 n.cm-2.s-1. However, this maximum value has
been reduced by about 5×103 times compared to the maximum value at the channel's entrance,
close to the reactor core (about 1.4×1011 n.cm-2.s-1).

MCNP5 simulation.
Water

Bi single crystal

Si single crystal Pb

Graphite
D1= 15.2 cm

SWX-277
Al sheet

Pb
Entrance of phantom
Water
phantom

D3= 9 cm

Entrance of channel No.2

Reactor
core

Beam stop

Air

Fe


x

z

Mean

Err. (%)

1

0

0

2.23E+07

0.39

2

0

0.5

2.79E+07

3

0


2.97

0.52

92

5

6

4.36E+05

3.32

2.13E+07

0.6

93

5

7

3.21E+05

3.86

-----

1
0
3.99
Yes
0
5.63E-03
2
0
2.71
Yes
-------------------------0.5
3.14E-04
7.00
No
29
0
0
4.49E-04
30
0
5.93
No
-------------------------

2.1.4. Evaluation of simulation errors
Parameters for evaluation of simulating thermal neutron flux and gamma dose rate in
the phantom are shown in Tables 2.3 and 2.4, respectively.
Table 2.3. The evaluation result of simulation parameters of thermal neutron flux in
the phantom with the recent configuration
No.


0.0000

9

0.0042
-----

16
-----

0.0001
-----

9

1×10
-----

1
-----

Table 2.4. The evaluation result of simulation parameters of gamma dose rate in
the phantom with the recent configuration
No.
1

Coordinate (cm)
x
z

0.15

0.0005

3

0

≠0

7×109

0.0161

0.15

0.0004

Phantom
Yes
No

Using the values presented in Tables 2.3 and 2.4 for comparing with the required
parameters in Table 1.14, it can be confirmed that the selection of the number of the history
(NOH) for the above cases satisfies the mandatory requirement of the MCNP5 program. That
means the data of simulation results are reliable.
To verify the above-simulated data, experimental measurements at CN2DR were
conducted to compare and evaluate the results between experiments and simulations.
2.2. Experiments on the recent configuration for BNCT research at Dalat reactor
In order to have a basis for changing and proposing a new configuration for BNCT

To determine the dependence of the detector's detection efficiency on energy, eight
standard sources of 133Ba, 109Cd, 57Co, 22Na, 137Cs, 54Mn, 65Zn, and 60Co have been used. The
above sources have 14 defined energy peaks that range from 81 keV to 1332.5 keV.
The full-peak efficiency curve for gamma rays of the HPGe detector used in the thesis is
shown in Fig. 2.10.
0.5
0.4
0.3

Log(Eff. %)

0.2
0.1
0.0
-0.1
-0.2

y = 1.7003x5 - 23.1623x4 + 125.5631x3 - 338.6775x2 + 453.9850x - 241.1282

-0.3
1.8

2.0

2.2

2.4

2.6


2.2.2.1.
Preparation of phantom and activation foils, irradiation - measuring - spectra
processing
(i) Preparation of phantom: a rectangular parallelepiped with a length of 25 cm, a
width of 16 cm and a height of 16 cm have been designed and manufactured, the outer shell of
the phantom is made from organic glass plates of 2 mm thickness (Fig. 2.11), and is filled
with water (called in short - phantom).

Fig. 2.11. Phantom used at CN2DR

(ii) Preparation of activation foils: Each 51V activation foil has a diameter of 1.27 cm
and a purity of 99.88%. These activation foils are placed at the position necessary to
determine thermal neutron flux in the phantom and are analyzed with the parameters listed in
Table 2.8.
Table 2.8. The decay properties of the nucleus in the activated foil
I
T1/2
Eγ
Activated product
(minute)
(%)
(keV)
52
V
3.75
1434.08
100

(iii) The irradiation of activation foils: Configuration of irradiation samples with a
thermal neutron beam at CN2DR with the recent configuration is shown in Fig. 2.13.

1
2
-----

Coordinate (cm)
x
0
0
-----

z
1
2
-----

Peak area
(number count)
832760
947480
-----

Thermal neutron flux
Error (%)
0.55
0.47
-----

(n.cm-2.s-1)
2.13E+07
1.23E+07

cps/ml
cps/ppm/ml
(ppm)
(ml)
count
(%)
time (s)
10
2
37526
6.1
81107
0.46
0.23
0.023
25
1.8
53846
2.7
61721
0.87
0.48
0.019
100
0.65
14449
2.3
10102
1.43
2.20

time
(h)
2

2

M2

0

2

1

3.01

3.01E-03

6

3

M3

0

4

2


Dose rate

Dose
(Sv)

Gy.h-1

Error (%)

11.24

5.62E-03

6

2.3. Summary of chapter 2
As above mentioned, the MCNP5 program has been used to simulate and calculate
thermal neutron flux in the phantom; method of neutron activation analysis was used to
determine thermal neutron flux in the phantom andTLD-900 was used to measure gamma
dose rate in the phantom. In addition, PGNAA technique is also used to create the calibration
curve for boron concentration in solution.

Chapter 3: RESULTS AND DISCUSSIONS

The main objectives of this chapter are: (i) to evaluate the results between simulation
and experiment, which were presented in Chapter 2; and (ii) to simulate a number of
configurations for CN2DR, thereby suggesting the optimal configuration for BNCT research
on the channel No.2 of Dalat reactor.
3.1. Assessing the results between simulation and experiment with the recent
configuration on CN2DR


Fig. 3.1. Thermal neutron flux distribution along the phantom center axis by
experiment and simulation
14


Figure 3.1 shows that the thermal neutron flux distribution along the phantom axis is a
function of the phantom depth. The thermal neutron flux has a maximum value of about
2.3×107n.cm-2.s-1 at a depth of 1 cm in the phantom and decreases rapidly to 4.0×106 n.cm-2.s1 at 4 cm deep. At the same time, this figure also shows a good agreement between
experimental data and simulation results.
Figure 3.2 shows the result of comparing the thermal neutron flux according to the
radius of the neutron beam, at the position z = 1 cm in the phantom between simulation by
MCNP5 and experimental measurement by foil activation technique (This data is extracted
from Tables 2.1 and 2.10).
MCNP5
Exp.
Fitting line MCNP5

Thermal neutron flux (n.cm-2.s-1)

2.50E+07 At z = 1 cm in phantom
2.00E+07
1.50E+07
1.00E+07
5.00E+06
0.00E+00

R2 = 0.983
-6



(n.cm-2.s-1)

•

Dγ (Gy.h-1)

Dγ / φth (Gy.cm2.n-1)

Yes (A)

3.18E+07

3.06E-03

2.67E-14

No (B)

1.69E+07

2.29E-05

3.76E-16

A/B

1.88

133.62

1.00E-03
R2 = 0.989

0.00E+00
0

2

4

6

8

10 12 14 16 18 20 22 24

Depth in phantom (cm)

Fig. 3.3. The gamma dose rate distribution along the center axis of the phantom with
MCNP5 and experiment

From the curve shown in Fig. 3.3, the gamma dose rate has a maximum value of about
5.5×10-3 (Gy.h-1) at the phantom entrance and decreases rapidly to about 1.8×10-3 (Gy.h-1) at 4
cm position in the phantom. In addition, these results also show a good agreement between
simulated and measured data for gamma dose rate along the center axis of the phantom.
The results of gamma dose rate according to the neutron beam radius, at position z = 3
cm in phantom by MCNP5 and experiment, are shown in Fig. 3.4 (This data is extracted from
Tables 2.2 and 2.12).

2.00E-03 At z = 3 cm in phantom

6

8

Radial distance from center of beam port (cm)

Fig. 3.4. The gamma dose rate distribution along the radius of the neutron beam at z = 3 cm
in the phantom using MCNP5 and TLD

It can be seen the curve shown in Fig. 3.4, the gamma dose rate in radial from the center
of the neutron beam at z = 3 cm in the phantom has good symmetry within a radius of about 5
cm. At the same time, this figure also shows a good agreement between MCNP5 simulation
and experimental measurement with TLD-900.
From the data shown in Figures 3.1 to 3.4 it can affirm that there is a good agreement
between MCNP5 simulation and experimental values of thermal neutron flux and gamma
dose rate in the phantom. This is the basis to use the MCNP5 program to simulate different
CN2DR configuration to propose the most optimal configuration for BNCT research.
16


Besides, Figures 3.1 to 3.4 also show that the curve shapes of thermal neutron flux
distribution, as well as of gamma dose rate distribution in the phantom, are similar to each
other. That is, there is a relationship between the gamma dose rate and thermal neutron flux in
the phantom. To verify the above comments, the simulated data of the gamma dose rate
distribution along the center axis in the phantom with and without phantom have been
compared.
The simulation results of the gamma dose rate distribution along the center axis of the
phantom in case with and without phantom is shown in Fig. 3.5.

Gamma dose rate (Gy.h-1)

cm, in the case with the phantom is much greater than that in case without phantom
The above result can be explained that the gamma dose rate measured in the phantom is
mainly caused by 2.22 MeV gamma-ray. Because the dose rate of the 2.22 MeV gamma-ray
in phantom is related to thermal neutron flux, and this flux also has a rapid decline similar to
that of a gamma dose rate in the range of 0-4 cm in the phantom (Fig. 3.1). Thus, from the
results of the above gamma dose rate simulation (Fig. 3.5), it also allows assertion that the
contribution of the gamma dose rate mixed in the neutron beam of CN2DR to the phantom is
very low.
3.2. Determination of the absorbed dose of BNCT in phantom
From the results of comparison and evaluation of thermal neutron flux and gamma dose
rate in the phantom (presented in section 3.1), we apply equations (1.10), (1.12) and (1.18) to
calculate the componential absorbed doses of DB, and DN, and the total absorbed dose of D in
BNCT research for experimental data, at each position in the phantom, with nitrogen
concentration of 2% and assumed 10B concentration of 30 ppm. Figure 3.6 shows the
relationship between the total absorbed dose and the two dose-components of DB and
DNcomponents, along the central axis in the phantom. In particular, the contribution of the DN
dose is small when the concentration of 10B used is 30 ppm.
Boron dose
Nitrogen dose
Total absorbed dose

Absorbed dose (Gy)

5.00E-05
4.00E-05
3.00E-05
2.00E-05
1.00E-05

0.00E+00


2.13E+07

Depth in phantom (cm)

1.86E+07

5
1.60E+07

2.663E+06

4

1.33E+07
1.07E+07

5.325E+06

3

7.99E+06

7.988E+06

2

1.065E+07

5.33E+06

10

Cps/ml

8
6
4
2
R2 = 0.998

0
0

100

200

300

400

500

Boron concentration (ppm)

Fig. 3.8. Calibration curve of boron concentration in solution is conducted at CN2DR

The result in Fig. 3.8 shows that the peak counting rate of 478 keV gamma of 10B is a
first-order function according to the boron concentration in the sample. In BNCT studies
around the world, the boron concentration in tumors ranges from 30 to 100 ppm and in normal

D4= 5.3 cm
Reactor
core

99.6 cm

50.7 cm

90 cm

Concrete

18 cm

Stainless steel

Fig. 3.9. Design drawing of the conical collimator tube of CN2DR

Simulating of thermal neutron flux and gamma dose rate at the phantom entrance has
been performed with a combination configuration of 20 cm Si + 3 cm Bi filters and conical
collimator (new configuration) as in Fig. 3.9. These simulation results are listed in Table 3.8.
Table 3.8. The neutron flux at the entrance of the phantom with the cylindrical
collimator and the conical collimator
Collimator tube

•

•

φ th


1.25

Based on the data presented in Table 3.8 can see that when changing the cylindrical
collimator configuration to the total conical collimator configuration with the 20 cm Si + 3 cm
Bi filters combination, the thermal neutron flux increased about 5.5 times and the gamma
dose rate increased about 6.8 times. The above increase in the thermal neutron flux and
gamma dose rate can be explained by the solid angle of the phantom entrance to the neutron
field in the reactor core increased.
3.4.2. Optimizing the collimator tube length
To serve the optimal length of the collimator tube, simulating a conical collimator tube
without filters, corresponding to the lengths (L): 40, 90, 140 and 240 cm has been conducted.
Parameters of the collimator tube are listed in Table 3.9.
Table 3.9. Parameters of collimator tube used in MCNP5 simulation
L (cm)
240
140
90
40

D5 (cm)
14.6
19.4
19.4
19.4

D4 (cm)
5.3
7.8
9.4


D5= 19.4 cm

99.6 cm

Pb

D4= 7.8 cm

50.7 cm

90 cm

Concrete

18 cm

Stainless steel

Fig. 3.10b. Configuration of CN2DR with the conical collimator tube length of 140 cm

The results of simulating thermal thermal neutron flux and gamma dose rate at the
sample irradiation position (Fig. 2.3) according to the different lengths of the conical
collimator tubes at CN2DR when the filters are not used are presented in Table 3.10.
Table 3.10. Thermal neutron flux and gamma dose rate at the sample irradiation
position along the length of the conical collimator tube.
Collimator tube
length (cm)
240
140

The results in Table 3.10 show that, as the length of the collimator tube decreases, the
neuron flux increases. However, the increase in neutron flux is slower than the increase in the
gamma dose rate. When the collimator tube length is reduced from 240 cm to 40 cm, the
neuron flux increases only 1.30 times; meanwhile gamma dose rate increased to 2.04 times.
Furthermore, the opening of the neutron beam increases from 5.3 cm to 12.47 cm in diameter,
which also means an increase in gamma dose rate to areas that do not need irradiation.
Therefore, the optimal collimator tube to choose is a cone with a length of 240 cm.
3.4.3. Optimizing filter length
To find the optimal length of the filter combination that satisfies the maximum thermal
neutron flux and ensures that the ratio of the gamma dose to thermal flux meets the
recommended requirement (< 3×10-13 Gy.cm2.n-1) of IAEA. Simulations of Si and Bi filter
combinations with lengths from 5 cm to 20 cm for Si and 1 cm to 3 cm for Bi have been
performed. The results of these simulations are shown in Table 3.11.

No.
1
2
3
-----

Table 3.11. The thermal neutron flux and gamma dose rate at the phantom entrance
correspond to the length of the different filter combinations
•
•
Filter (cm)
•
φth
Dγ / φth
Dγ / φth
-1


The simulation results presented in Table 3.11 show that some of the suitable filter
combinations used for BNCT are: 5 cm Si + 1 cm Bi; 10 cm Si + 1 cm Bi; 15 cm Si + 1 cm
Bi; 5 cm Si + 2 cm Bi; 10 cm Si + 2 cm Bi; and 5 cm Si + 3 cm Bi. In particular, a
combination of 5 cm Si + 1 cm Bi filter is the best choice in these cases. The reason for the
choice is that the neutron flux is the highest, the irradiation time to reach the recommended
neutron flux limit of 1×109 n.cm-2.s-1 is the shortest and still ensures the ratio of gamma dose
rate to thermal neutron flux is less than 3×10-13 Gy.cm2.n-1. The neutron spectrum at the
phantom entrance of the new and the recent configuration is shown in Fig. 3.11.
New configuration
Recent configuration

1E+09

Neutron flux (n.cm-2.s-1)

1E+08
1E+07
1E+06
1E+05
1E+04
1E+03
1E-9

1E-8

1E-7

1E-6



Pb

D5= 14.6 cm
D4= 5.3 cm
Reactor
core

99.6 cm

50.7 cm

90 cm

Concrete

18 cm

Stainless steel

Fig. 3.12. General design drawing of the new configuration for BNCT research at CN2DR

This new configuration has a beam line tube length (also neutron collimator tube) of 240
cm, with a diameter of 15.2 cm close to the reactor's reflector and the diameter of 5.3 cm
toward the phantom.
21


Simulation results of thermal neutron flux distribution in the phantom with the new
configuration proposed for BNCT research at CN2DR (the conical collimator tube with a


4.88E+08

4.23E+08

2.77E+08

1.18E+08

0.4

4.97E+08

4.82E+08

4.16E+08

2.77E+08

1.28E+08

0.9

4.28E+08

4.10E+08

3.51E+08

2.36E+08

(cm)
history
0.0013
256
0.0000
0
5×108
0.0013
252
0.0000
1
5×108
---------------------

the parameters presented in Tables 2.3 and 3.13 shows that with the new configuration:
number of the history is halved but the FOM parameter increases and the R parameter
decreases. So just using a small amount of time but we still get accurate simulation results.
This proves that the new design configuration is more optimal than the recent configuration.
Figure 3.14 represents an improvement of the thermal neutron beam in the phantom of
the new configuration compared to the recent configuration.

Thermal neutron flux (n.cm-2.s-1)

5.5E+08
New configuration
Recent configuration

5.0E+08
4.5E+08
4.0E+08

neutron beams used for BNCT research in the world.
Table 3.14. Some designs of thermal neutron beams based on the reactors of
countries have been implemented by MCNP5
•

Power
(MW)

φ th

Dγ / φth

×109 (n.cm-2.s-1)

TRIGA Mark II

0.1

1.5

×10-13 (Gy.cm2.n-1)
1.7

MURR

10

0.96

3.99

In addition, from the creating of the calibration curve of boron concentration for the
solution sample, it can confirm that the PGNAA facility at CN2DR completely meets the
process of controlling boron concentration in the BNCT research. In addition, this result also
shows the high applicability of the PGNAA facility at the Dalat reactor for the quantitative
analysis of boron in the biological, medical, pharmacological, and environmental samples.
CONCLUSIONS
From the obtained results can conclude that the author's thesis has achieved the set of
objectives, which is to approach and start up a new research direction of neutron beam
application from Dalat reactor to study and determine the basic parameters of the BNCT
method.
In order to meet the above objectives, the scientific and practical results of the thesis
have been achieved including:
- Study and determine the dose components in BNCT. The obtained results can conclude
that the absorbed dose in BNCT depends mainly on the thermal neutron flux and the
concentration of boron in the tumors.
- Simulation and determination of absorbed dose distribution of BNCT in the water
phantom at CN2DR by MCNP5 code and NAA method, respectively. The obtained results
show that there is a good agreement between experimental data and simulation results.
Therefore, it can confirm that the experimental arrangement is satisfactory and the simulation
23