Nghiên cứu đánh giá các thành phần liều phục vụ nghiên cứu BNCT trên kênh ngang của lò phản ứng hạt nhân đà lạt tt tiếng anh

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 neutr

MINISTRY OF SCIENCE AND

TECHNOLOGY

MINISTRY OF EDUCATION AND

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

Người hướng dẫn khoa học 1: PGS.TS Nguyễn Nhị Điền

Người hướng dẫn khoa học 2: TS Trịnh Thị Tú Anh

Phản biện 1:

Phản biện 2:

Phản biện 3:

Luận án sẽ được bảo vệ trước Hội đồng cấp Viện chấm luận án tiến sĩ họp tại vào hồi giờ ngày tháng năm 20

Có thể tìm hiểu luận án tại:

- Thư viện Quốc gia Việt Nam

- Thư viện Trung tâm Đào tạo hạt nhân

1

INTRODUCTION

Radiation therapy is a method applied to treat tumors through the interaction of radiation 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

1998 Byung-Jin et al used MCNP to design a thermal neutron beam on the horizontal channel

of HANARO reactor with a power of 30 MW using Si and Bi filters The thermal neutron flux and the ratio of gamma dose rate to thermal neutron flux at the sample irradiation position were 2.6×109 n.cm-2.s-1 và 1.2×10-13 Gy.cm2.n-1, respectively

In Vietnam, Dalat reactor has been reconstructed from the original TRIGA Mark II reactor and officially put into operation with a nominal power of 500 kW on March 20, 1984 Filtered neutron beams from horizontal channels No.3 and No.4 have been in use since the 1990s for basic and applied researches Since 2011, the Channel No.2 of Dalat Reactor (CN2DR) was put into use with a number of good quality neutron beams such as pure thermal neutron and mono-energetic epithermal neutron beam 2 keV, created by filtering techniques with neutron flux of 1.5×106 n.cm-2.s-1 These neutron beams have been used mainly for the study on nuclear data, nuclear structure, etc

Although, in the world, the BNCT method has been applied since the 1960s of the 20th

century for clinical or preclinical research in many countries such as Japan, the USA, Korea Republic of, Iran, Italy, Czech Republic, Finland, Netherlands, v.v…, meanwhile, up to now,

in Vietnam there is no any BNCT facility as well as detail studies on dose calculation of neutron reaction with10B in BNCT method

In order to achieve the above objectives, the contents of the thesis need to be done include: (i) studying and calculating the absorbed dose of reaction 10B(n,α)7Liin 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 phantom with a ThermoLuminescence Dosimeter (TLD), evaluating and discussion of data between experiment and simulation Chapter 3 presents the results of simulating the design of some new configurations for neutron beams, thereby proposing the optimal configuration for BNCT at CN2DR

Chapter 1: OVERVIEW

The objective of this chapter is to present the principles and components of absorbed doses that appear in BNCT, evaluation and determination of the major dose components in the

Neutron source (Reactor)

Fig 1.2 Technical illustration of BNCT using thermal neutron beam to treat brain tumors

1.2 Treatment of brain tumors by BNCT in the world

This section presents an overview of the clinical trial situation of BNCT in the world during the 1968-1999 period

1.3 Neutron KERMA factor in tissue

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:

is the number of nuclei of the interest isotope

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

(%)

KERMA factor (Gy.cm2)

Ratio (%)

Thus, the rate of neutron KERMA contribution of hydrogen and nitrogen elements is mainly

in tissue (accounting for 98.2%) Therefore, when calculating neutron KERMA in tissue, we can simply calculate neutron KERMA of nitrogen or add neutron KERMA of hydrogen

1.4 Theoretical calculation of absorbed dose in BNCT

1.4.1 Absorbed dose and dose measurement unit

This section presents the theory of concepts and measurement units of absorbed dose, equivalent dose and radiation weighting factor of radiation types

4

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 10B(n,α)7Li is expressed as follows:

(6%) (MeV)2.79LiHe

(94%) (MeV)2.31MeV)(0.478Li

HeBeV)(0.025n

7 4

* 11 1

10

++

+γ

++

〈

→+

The boron dose is calculated by the formula:

th B

B 13

B 1.6 10 C σ Q

Here, the symbols D, Cx, σx and Q are used as follows: D is the absorbed dose, C is the concentration of the element, x is the symbol of the element, σ is the reaction cross-section of the element to the thermal neutron, Q is the kinetic energy of product particles (MeV), và Φth

is the thermal neutron fluence (n.cm-2)

In BNCT, when calculating the dose, the concept of thermal neutron fluence (Φth) is used instead of the thermal neutron flux (φth) The relationship between these two quantities

th B 14

B 7.43 10 C

In which, the value 7.43×10-14 is the KERMA factor of boron for thermal neutron, CB (ppm) (ii) The thermal neutron dose (DN) generated from the reaction 14N(n,p)14C is calculated by the formula:

th N

N 13

D = × − × × × ×Φ (1.11) Combined with the neutron KERMA factor for the 14N nucleus, the reaction cross-section, and the nitrogen concentration in the tissue, equation (1.11) is rewritten as follows:

th N 14

N 6.78 10 C

In which, the value 6.78×10-14 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 reaction1H(n,n')1H The fast neutron dose is calculated by the formula:

fE

σC101.6

where σsH is the elastic scattering cross-section between fast neutron and hydrogen (cm2), Ef

is the released energy of the reaction (MeV), Φf is the fast neutron fluence (n.cm-2), vàf =0.5

is the absorption coefficient for fast neutron in tissue

(iv) The gamma dose (Dγ) generated by gamma rays formed in the reaction 1H(n,γ)2H 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 reaction1H(n,γ)2Hand calculated by the formula:

) 22 2 ( th ) 22 2 ( H H 13

in which fγ(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:

th 14 (2.22) 1.0 10

) 478 0 ( th ) 478 0 ( B B 13

in which, the value 1.0×10-16 is the KERMA of the 0.478 MeV gamma-ray for the whole body, calculated for 1 ppm unit of boron

1.4.3 Total absorbed dose in BNCT

From equations 1.10, 1.12, 1.15 and 1.17 can see that the multipliers in equations 1.15 and 1.17 are about 10 and 100 times smaller than those in equations 1.10 and 1.12 On the other hand, the radiation weighting factor of gamma is about 20 times smaller than that of heavy charged particles, so the absorbed dose in the BNCT method is usually only interested

in the two dose-components caused by the thermal neutron reaction by 10B và 14N in tissue

th 14 N

C43.7(

where D (Gy) is the total absorbed dose in BNCT

1.5 Components of the research model of BNCT in the world

1.5.1 Filtered neutron beam

This section presents the neutron beam used for BNCT, basic principle of thermal neutron filter technique, and some filtered neutron beams that have been used for BNCT research around the world

Table 1.7 Some reactors generate a thermal neutron beam with Si and Bi

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:

0

t t

t

e e

e N

I

f C

λ λ

λ

σε

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):

( ) ( )2 2

δεδ

δφφ

6

Here, δφ , δC,and δε is the relative error of thermal neutron flux, of the number of counts in gamma peak, and of the gamma full-peak efficiency at the energy of interest; and ∆φ is the absolute error of thermal neutron flux, respectively

1.5.4 Determination of boron concentration by PGNAA

This section presents an overview of the use of the PGNAA technique to determine boron concentration in BNCT research

In Japan, measurements of 10B concentration in tumors, tissues, blood, and cell cultures

at Musashi Institute of Technology by Matsumoto and Aizawa, the detection limit for 10B of PGNAA systems in these experiments to 2.5 ppm for 1 ml volume samples and 10 ppm for 0.3 ml volume samples

In Vietnam, the PGNAA system has been installed to use at the tangential horizontal

channel No.3 of Dalat reactor from 1988, at the radial channel No.2 since 2011, to serve the research directions of PGNAA

1.5.5 Determination of gamma dose by TLD

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

Table 1.10 Errors of some types of TLDs

1.6.2 Input file structure and tally types

This section introduces the MCNP5 program, input file structure and tally types in MCNP5

1.6.3 The conversion for neutron and gamma flux to dose rate

This section presents how to convert for neutron and gamma flux to the absorbed dose rate

1.6.4 Evaluation of errors

This section presents the relative error and the meaning of the relative error value of the results of running the MCNP5 code

Table 1.14 The meaning of the relative error value R in MCNP5

In order to track the retrieval results, MCNP5 also issues FOM (Figure Of Merit) 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

7

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

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

1.7.2 Filtered neutron beam at channel No.2 Dalat reactor

By using a combination of the 80cm Si and 4cm Bi single-crystal filters, the quality thermal neutron beam was generated at CN2DR, and the physical parameters of this neutron beam are listed in Table 1.16

high-Table 1.16 Physical parameters of thermal neutron beam at the exit of CN2DR

Thermal neutron flux

φth (n.cm-2.s-1)

Cadmi ratio

RCd(Au)

Filter length (cm) The diameter of

the neutron beam (cm)

is controlled by the PGNAA technique and thermal neutron flux is determined by the NAA technique

8

Chapter 2: SIMULATION AND EXPERIMENTAL

The purpose of this chapter is to simulate (by MCNP5) and experimentally set-up the horizontal CN2DR with the recent configuration to: determine, compare and evaluate the values of thermal neutron flux and gamma dose rate in water phantom At the same time, a boron concentration curve has been also built using the PGNAA technique to test the ability

of determination of boron concentration in liquid samples

2.1 Simulation for BNCT research model at Dalat reactor

2.1.1 Filtered neutron beam channel No.2

The neutron beam guide system at CN2DR, according to the cylindrical design, has a total length of 240.3 cm divided into two parts: (i) the neutron beam guide; and (ii) the neutron beam collimation

Figure 2.3 shows the structure of the neutron beam guide system at CN2DR, with cylindrical collimation of 20 cm Si and 3 cm Bi filters (called for short - recent configuration)

1E-09 1E-08 1E-07 1E-06 1E-05 1E-04 1E-03 1E-02 1E-01 1E+00 1E+00

1E+01 1E+02 1E+03 1E+04 1E+05 1E+06 1E+07 1E+08 1E+09 1E+10 1E+11

1E-9 1E-8 1E-7 1E-6 1E+03

1E+04 1E+05 1E+06 1E+07 1E+08

Energy (MeV)Fig 2.5 The spectrum shape at the sample irradiation position at CN2DR

9

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)

2.1.2 Water phantom

In the thesis, a water phantom (hereinafter referred to as phantom) of rectangular parallelepiped with dimensions of 25 cm × 16 cm × 16 cm (see Fig 2.11) has been designed and manufactured, the shell of the phantom is made from organic glass plates with 2 mm in thickness Figure 2.6 shows the phantom model used at CN2DR simulated by MCNP5

Cell tính thông lượng nơtrôn

Đường trung tâm

Cell tính suất liều gamma

Phantom nướcWater phantom Cell for neutron flux calculation

Cell for gamma dose rate calculation Center line

Fig 2.6 Phantom model used at CN2DR was simulated by MCNP5 Figure 2.8 shows the detailed structure of CN2DR and the phantom position used in the MCNP5 simulation

Water phantom

Table 2.1 The thermal neutron flux in the phantom is simulated by MCNP5 with

the recent configuration

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

Table 2.4 The evaluation result of simulation parameters of gamma dose rate in

the phantom with the recent configuration

No Coordinate (cm) No of the

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 research at CN2DR, experiments on the recent configuration (introduced and simulated in Section 2.1) have been preformed

Experimental set-up steps include detector calibration, water phantom and activation foils preparation, irradiation - measuring - spectra processing from activated foils with thermal neutron beam at CN2DR

11

2.2.1 Detector calibration

In this thesis, the gamma spectrometer system using a high-purity germanium detector (model: GR7023) of Canberra firm (Fig 2.9) has been used, and its parameters are listed in Table 2.5

Fig 2.9 Gamma spectrometer system uses HPGe detector type GR7023 at CN2DR Table 2.5 Characteristics of gamma spectrometer used at CN2DR

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

-0.3 -0.2 -0.1 0.0 0.1 0.2 0.3 0.4 0.5

Fig 2.10 The absolute full-peak efficiency curve for gamma rays of the HPGe detector for

standard samples is 5 cm from end-cap of the detector Based on the relationship between the absolute full-peak efficiency of detector and gamma-ray energy, we extrapolated the absolute full-peak efficiency of detector for 1434 keV gamma-ray of 52V emitted from the 51V foil, this efficiency used to calculate the thermal neutron flux (in section 2.2.2.2) Using the least-square method, the results of calculating the absolute full-peak efficiency and relative error of the HPGe detector for 1434 keV gamma-ray are 0.6644% and 1.5%, respectively