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

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 hồn 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 bảo vệ trước Hội đồng cấp Viện chấm luận án tiến sĩ họp vào hồi ngày tháng Có thể tìm hiểu luận án tại: - Thư viện Quốc gia Việt Nam - Thư viện Trung tâm Đào tạo hạt nhân năm 20 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 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 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 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 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 Hence, a study on dose components evaluation for BNCT research on the horizontal channel No of the Dalat reactor is the subject proposed to carry out in this thesis Objectives The main objectives of the thesis are to simulate the optimal design of thermal neutron beam at the exit of the CN2DR for BNCT research by MCNP5 code; simulation, calculation and experimental measurement of typical parameters of thermal neutron flux distribution and radiation dose components in a water phantom model; development of methods for analyzing boron concentration in water samples to apply in BNCT experiments and other related applications 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 10 B(n, α) 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 chapters Chapter 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 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 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 total absorbed dose as well as the method of calculation and determination of these dose components 1.1 The principle of BNCT This section presents the principle of BNCT, concentration, and cross-section of the reaction of elements in the tissue with thermal neutrons Figure 1.2 illustrates the BNCT technique using a thermal neutron beam to treat brain tumors Tumor Thermal neutron 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: 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) (%) H 10.7 4.49E-15 2.49 C 14.5 3.49E-18 0.00 - N - 2.2 - 1.73E-13 - 95.67 - 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 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: 10 B+1 n (0.025 eV) →11 B* 〈 10 B(n, α) Li is expressed as He+ Li + γ (0.478 MeV) + 2.31 (MeV) He + Li + 2.79 (MeV) (94%) (6%) The boron dose is calculated by the formula: D B = 1.6 × 10−13 × C B × σ B × Q × Φ th (1.8) 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), Φ th is the thermal neutron fluence (n.cm-2) In BNCT, when calculating the dose, the concept of thermal neutron fluence ( Φ th ) is used instead of the thermal neutron flux ( φ th ) The relationship between these two quantities is calculated as follows: Φ th = φ th × t (1.9) -2 -1 In which φth is the thermal neutron flux (n.cm s ), and t is time (s) Combined with the neutron KERMA factor for the 10B nucleus and the 10B concentration in the tumor (used at the ppm concentration), equation (1.8) is rewritten as follows: D B = 7.43 × 10−14 × C B × Φ th (1.10) 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 calculated by the formula: 14 N(n, p)14 C is D N = 1.6 × 10 −13 × C N × σ N × Q × Φ th (1.11) nucleus, the reaction crossCombined with the neutron KERMA factor for the section, and the nitrogen concentration in the tissue, equation (1.11) is rewritten as follows: 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 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), 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 H(n, γ) 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 H(n, γ) 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) where the value 1.0×10-14 is the KERMA factor of 2.22 MeV gamma-ray for the volume that is irradiated • The absorbed dose caused by 478 keV gamma-ray when the nucleus Li * dissolves the excitation energy, calculated by the formula: D γ (0.478) = 1.6 × 10 −13 × C B × σ B × E γ ( 0.478) × Φ th × f γ ( 0.478) (1.16) In which f γ ( 0.478) is the absorption coefficient in the volume that is irradiated by 0.478 MeV gamma-ray Similar to equation (1.14), equation (1.16) is rewritten as follows: D γ (0.478) = 1.0 × 10 −16 × C B × Φ th (1.17) -16 in which, the value 1.0×10 is the KERMA of the 0.478 MeV gamma-ray for the whole body, calculated for 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 14N in tissue (1.18) D = (7.43 × C B + 6.78 × C N ) × 10 −14 × Φ th 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 singlecrystal filters Length of filter (cm) Power Reactor (MW) Si Bi MURR 10 50 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 × (1 − e − λt1 )× e − λt × − e − λt 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): ( δφ = (δC )2 + (δε )2 ∆φ = φ × δφ ) (1.27) (1.28) 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 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 No Table 1.10 Errors of some types of TLDs Dosimeter Material TLD-300 CaF2:Tm TLD-600 LiF:Mg,Ti TLD-700 LiF:Mg,Ti TLD-900 CaSO4:Dy Error (%) 30 5.1 5.1 1.6 Use the MCNP5 code in BNCT 1.6.1 Introduction MCNP5 is a code that uses the Monte Carlo method to simulate nuclear physics processes for neutron, photon, and electron This is a very powerful computational tool that can simulate the transport of neutron, photon, and electron, and solve three-dimensional radiation transport problems used in many fields from reactor design to radiation safety and medical physics 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 Range of R > 0.5 0.2 – 0.5 0.1 – 0.2 < 0.1 < 0.05 Quality of the Tally Not meaningful The factor of a few Questionable Generally reliable Generally reliable for point detectors 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 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 10 12 120 At 7cm depth 100 80 60 40 20 -6 Depth in phantom (cm) -4 -2 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) IAEA >1 1×108 n.cm2.s-1 and the ratio of gamma dose rate to thermal neutron flux < ×10-13 Gy.cm2.n-1 3.5 Summary of chapter As described above, from the results compared between simulation and experiment for the thermal neutron flux distribution and gamma dose rate in the phantom of the recent configuration, the author has improved the design and proposed a new configuration for BNCT research at CN2DR The thermal neutron flux of the new configuration has increased by about 12 times compared to the recent configuration and ensures the permitted safety for the gamma dose rate However, with this new design result focusing primarily on the goal of neutron flux achieved in the phantom, the radiation shielding structures outside the channel need to be additionally designed when the new collimation model is applied 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 method can be used to design and improve the CN2DR to meet the requirements for BNCT research - Develop a method for analyzing boron concentration in water samples by the PGNAA technique The results show that the PGNAA facility at CN2DR perfectly meets the requirements for BNCT research In addition, this result can be apply extended for the quantitative analysis of boron in biological, pharmacological and environmental samples - Proposing the optimal configuration design for the BNCT system at CN2DR by MCNP5 code The thermal neutron flux of the new configuration increased by 12 times compared to the recent configuration and still ensures permissible safety for the gamma dose rate - The research results of experimental and simulation allow concluding that CN2DR completely meets the requirements of neutron beam technique and radiation safety to carry out physical research and training on the BNCT method SUGGESTIONS FOR FURTHER WORK From the contents and conclusions in the thesis, the author has some suggestions for further research directions as follows: (1) Design, manufacture and install the conical collimator tube and cm Si +1 cm Bi filter combination as proposed (2) Experimental measurement of neutron flux and gamma dose rate before and inside the water phantom to verify and evaluate with simulation results presented in section 3.4.4 of this thesis (3) Proposing to combine the research with the Center for Research and Production of Radioisotopes to test the BNCT technique on animals (some mice with tumors in the legs) LIST OF PUBLICATIONS [1] C.D Vu, T.Q Thien, H.V Doanh, P.D Quyet, T.T.T Anh, and N.N Dien (2014), “Characterization of neutron spectrum parameters at irradiation channels for neutron activation analysis after full conversion of the Dalat nuclear research reactor to low enriched uranium fuel”, Nucl Sci Technol., (Vietnam), Vol 4, No 1, pp 70-75 [2] Trinh Thi Tu Anh, Nguyen Danh Hung, Pham Dang Quyet, Pham Ngoc Son (2018), “Dose Calculation and Measurement from B10(n, α)Li7 Reaction Using Filtered Neutron Beam at Nuclear Research Institute”, Nucl Sci Technol., (Vietnam), Vol.8, No 1, pp 2935 [3] Pham Dang Quyet, Pham Ngoc Son and Trinh Thi Tu Anh (2018), “Measurement of inphantom thermal neutron flux distribution in Dalat Research Reactor boron neutron capture therapy beam line”, Proceedings of 5th Academic conference on Natural Science for Young Scientists, Master and Ph.D Students from ASEAN Countries 4-7 October 2017, Da Lat, Viet Nam, Publishing house for Sci & Tech ISBN: 978-604-913-714-3, pp 329-335 [4] Trinh Thi Tu Anh, Pham Dang Quyet, Mai Nguyen Trong Nhan & Pham Ngoc Son (2019), “Measurement of Neutron Flux and Gamma Dose Rate Distribution Inside a Water Phantom for BNCT Study at Dalat Research Reactor”, SAINS Malaysiana, 48(1), pp 191-197 [5] Pham Dang Quyet, Pham Ngoc Son, Trinh Thi Tu Anh, Nguyen Nhi Dien, and Cao Dong Vu, “Simulation Design of Thermal Neutron collimators for Neutron Capture Studies at the Dalat Research Reactor” The article has been accepted for publication by Asian Journal of Scientific Research 24 ... trình hồn 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:... ngày tháng Có thể tìm hiểu luận án tại: - Thư viện Quốc gia Việt Nam - Thư viện Trung tâm Đào tạo hạt nhân năm 20 INTRODUCTION Radiation therapy is a method applied to treat tumors through the... Vietnam there is no any BNCT facility as well as detail studies on dose calculation of neutron reaction with10B in BNCT method Hence, a study on dose components evaluation for BNCT research on the