Simulation of absorbed doses distribution in a polyethylene phantom for BNCT application at the Dalat research reactor

Nội dung text: Simulation of absorbed doses distribution in a polyethylene phantom for BNCT application at the Dalat research reactor

1. Nuclear Science and Technology, Vol.10, No. 4 (2020), pp. 01-07 Simulation of absorbed doses distribution in a polyethylene phantom for BNCT application at the Dalat research reactor Pham Dang Quyet1*, Pham Ngoc Son2, Nguyen Nhi Dien2, Nguyen An Son1, Trinh Thi Tu Anh1, Cao Dong Vu2 1The University of Dalat, 01 Phu Dong Thien Vuong Street, Dalat, Lamdong 2Nuclear Research Institute, 01 Nguyen Tu Luc Street, Dalat, Lamdong *Email: quyetpd@dlu.edu.vn (Received 26 November 2020, accepted 28 December 2020) Abstract: In this paper, the distribution of absorbed dose components in a polyethylene phantom for BNCT application at Dalat Nuclear Research Reactor (DNRR) were calculated using the MCNP5 code. The configuration of horizontal neutron channel No.2 of the DNRR, which contains a cylindrical collimator with neutron filters of 20-cm Si and 3-cm Bi, was simulated. The results show that the gamma dose along the central axis of the phantom has the maximum value of 1.82×10-6 Gy at the 0.5-cm depth, and reduces to 9.05×10-7 Gy at the 3-cm depth. The main contribution to gamma dose is due to the interaction of thermal neutron with hydrogen in the phantom via the 1H(n,γ)2H reaction, and its value is much smaller than thermal neutron dose. The total absorbed dose along the central axis of the phantom has the maximum value of 7.87×10-5 Gy at the 0.5-cm depth, and decreases rapidly to 1.52×10-5 Gy at the 3-cm position, and mainly depends on the boron and thermal neutron doses caused by the 10B(n, α)7Li and 14N(n, p)14C reactions, respectively. Keywords: Dose distribution, BNCT, polyethylene phantom, thermal neutron flux. I. INTRODUCTION diameter ( 10 m) and the rest i.e. 0.48 MeV is deposited by the emitted gamma ray Boron neutron capture therapy (BNCT) throughout the medium. Because the high LET is an effective cancer treatment method for of two particles (4He  150 keV/m and 7Li  several types of brain tumors such as 175 keV/m) have limited path lengths in Glioblastoma Multiforme (GBM), tissue (5  9 m) [1, 3], therefore, it can Astrocytomas, etc. This method is based on the selectively destroy tumor cells and spare nuclear reaction that occurs when a stable normal cells [4]. isotope, 10B, is irradiated with thermal neutrons (with the energy < 0.05 eV) to produce high Up to now, the best neutron source for linear transfer (LET) 4He nuclei (α-particles) BNCT is a thermal nuclear research reactor [3]. and 7Li recoiling nuclei [1, 2]. In some reactors, such as HANARO in the Republic of Korea [5] and MURR in the USA 4 7 He + Li + 2.79 MeV (6%) [6], the single crystal filters of Si and Bi are 10 1 11 B + n B used to produce thermal neutrons for BNCT 4He + 7Li* + 2.31 MeV (94%) research. The in-phantom depth absorbed dose distribution for BNCT application has been 7Li + g + 0.48 MeV computed using the MCNP5 code [7-9]. In As it is seen, in 94% of the events the case of the TRR and MNSR reactors in Iran, released energy of 2.31 MeV of two above the absorbed doses distribution in phantoms particles is deposited within about a cell were computed based on the thermal column of ©2020 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute
2. SIMULATION OF ABSORBED DOSES DISTRIBUTION IN A POLYETHYLENE PHANTOM the reactor, however gamma-rays dose that II. MATERIALS AND METHODS include gamma rays from reactor and captured gamma rays in material of the phantom were The DNRR is a 500-kW pool-type not estimated [8-9]. Two types of materials, research reactor, which was reconstructed and which are usually used for designing a upgraded from the former 250-kW TRIGA Mark II reactor. Light water is used as both moderator phantom for BNCT research, are water and and coolant, and the beryllium and graphite polyethylene because their densities are similar materials surrounding the reactor core are used as to that of a tissue. Polyethylene phantoms have a neutron reflector. The operation of the DNRR been used for experimental trial at the LVR-15 is mainly for the purposes of radioisotope reactor in Czech Republic, the KUR reactor in production, neutron activation analysis, basic and Japan, and the BMRR reactor in USA [10]. applied research, and training [15]. Water phantoms have been used experimentally at the JRR No. 4 in Japan [11], The reactor has four horizontal neutron the TRIGA Puspati reactor in Malaysia [12] channels that penetrate the concrete shield and and the TRR reactor in Iran [13]. The the aluminum tank and pass through the pool prototype water phantom was also developed water to the graphite reflector. Three of the for the design simulation of thermal neutron neutron channels (No.1, No.2 and No.4) are collimators for neutron capture studies at the oriented radially with respect to the center of DNRR [14]. The purpose of this study is to the reactor core, and the channel No.3 is simulate the distribution of absorbed dose tangential to the outer edge of the reactor core. components in the polyethylene phantom for The radial channel No.2 terminates at the outer BNCT research. The recent configuration of edge of the graphite reflector that can produce the horizontal neutron channel No.2 of the a relatively high thermal neutron flux. This is DNRR were simulated using the MCNP5 code. the reason that the neutron channel No.2 was The absorbed doses distribution in the chosen in this study. The configuration of the polyethylene phantom at the horizontal channel horizontal channel No.2 of the DNRR is No.2 was then calculated. presented in Fig. 1. Beam stop Water Fe Bi single crystal Si single crystal Pb SWX-277 Pb Air Al sheet Sample irradiation Graphite position D1= 9 cm D2= 3 cm Beam port No.2 Reactor core 99.6 cm 50.7 cm 90 cm 18 cm 40 cm Concrete Stainless steel Fig. 1. The configuration of the horizontal channel No.2 of the DNRR As seen in Fig. 1, after traversing the produced for different applications, including filters including Si and Bi single crystal nuclear data measurement, prompt gamma layers, with the thicknesses of 20 and 3 cm, neutron activation analysis (PGNNA) and respectively, thermal neutron beam will be BNCT study [16]. The simulated and 2
3. PHAM DANG QUYET et al. measured thermal neutron fluxes and gamma normalized according to the experimental dose rates at 1-cm depth in the water phantom values to calculate thermal neutron flux and at the channel No.2 are shown in Table I [14]. gamma dose rate distributions in the Then, the simulated results have been polyethylene phantom. Table I. Simulated and measured results of the thermal neutron flux and gamma dose rate in the water phantom at the channel No.2 Parameters Simulation Experimental Thermal neutron flux 2.27 107 (2.13 0.04) 107 (n/cm2/s) Gamma dose rate 3.33 10-3 (3.01 0.90) 10-3 (Gy/h) In this study, the polyethylene respectively, was used and located at 40 cm phantom ( = 0,92 g/cm3) with its length, far from the beam port of the channel No.2 width and depth of 25, 16 and 16 cm, (Fig. 2). Beam stop Air Water Fe Bi single crystal Si single crystal Pb SWX-277 Pb Al sheet Entrance of phantom Graphite Polyethylene D = 15.2 cm D = 9 cm 1 3 phantom Reactor Entrance of channel No.2 D4= 3 cm core 99.6 cm 50.7 cm 90 cm 18 cm 40 cm Concrete Stainless steel Fig. 2. The real phantom position at the channel No.2 of the DNRR In BNCT, there are four components of to interaction of thermal neutron with the interested absorbed dose: the boron dose, the hydrogen of phantom by the 1H(n,γ)2H thermal neutron dose, the gamma dose, and the reaction. The DB and DN doses are determined fast neutron dose. However, for the current as [17]: configuration of channel No.2, the fast neutron 14 DB 7.43CB 10 th (2) dose is very low and can be eliminated [14], so and the total absorbed dose in BNCT is calculated 14 by equation [8]: DN 6.78CN 10 th (3) 10 D DB DN Dg (1) In which, CB is the B concentration (the value in this calculation was chosen of 30 Where, D is the total absorbed dose, D 14 B ppm [1, 7]), CN is the N concentration (it is the boron dose, DN is the thermal neutron was chosen of 2 % [17]), and th is the dose, and Dγ is the gamma dose that includes thermal neutron fluence (n/cm2). In BNCT, gamma from the reactor core and gamma due when calculating the dose, the concept of 3
4. SIMULATION OF ABSORBED DOSES DISTRIBUTION IN A POLYETHYLENE PHANTOM thermal neutron fluence (th) is used instead E2  (E)dE of the thermal neutron flux (th). The th n (5) relationship between these two quantities is E1 calculated as follows: E2 D (E)R(E)dE (6) g n E1 th th t (4) Where, th is the thermal neutron flux, Where, th is the thermal neutron flux n(E) is the neutron flux, E1 and E2 are the 2 (n/cm /s), and t is time (s). lower and upper limit of the energy range,  The thermal neutron flux and gamma respectively, Dg is the gamma dose rate and dose rate in the polyethylene phantom were R(E) is the flux to dose conversion factor. calculated using tally F4 with DE4/DF4 cards of MCNP5, the gamma dose was determined III. RESULTS AND DISCUSSION using the fluence-to-KERMA conversion The thermal neutron flux and gamma factors reported in ICRU 63 [18]. With these dose distributions along the central axis of the MCNP cards we can calculate the following polyethylene phantom at the neutron horizontal values of thermal neutron flux and the gamma channel No.2 were calculated and the results dose rate [8, 9]: are shown in Tables II and III, respectively. Table II. The thermal neutron flux along central axis of the polyethylene phantom 2 2 Position (cm) th (n/cm /s) Position (cm) th (n/cm /s) No. No. z Mean Err. (%) z Mean Err. (%) 1 0 2.91×107 1.0 7 5 1.89×106 4.3 2 0.5 3.33×107 1.2 8 6 1.18×106 5.1 3 1 2.43×107 1.4 9 7 7.12×105 7.2 4 2 1.24×107 2.0 10 8 4.65×105 8.0 5 3 6.41×106 2.6 11 9 2.88×105 8.5 6 4 3.60×106 3.4 12 10 1.78×105 12.5 Table III. The gamma dose along central axis of the polyethylene phantom Dg (Gy) Dg (Gy) Position (cm) No. Mean Err. (%) Mean Err. (%) z from reactor with phantom from reactor without phantom -6 1 0 1.18×10 4.5 4.06×10-9 17.2 -6 2 0.5 1.82×10 4.6 5.28×10-9 25.7 -6 3 1 1.69×10 5.0 4.24×10-9 17.1 -6 4 2 1.18×10 5.3 4.16×10-9 17.1 4
5. PHAM DANG QUYET et al. Dg (Gy) Dg (Gy) Position (cm) No. Mean Err. (%) Mean Err. (%) z from reactor with phantom from reactor without phantom -7 5 3 9.05×10 6.0 4.94×10-9 18.1 6 4 5.61×10-7 7.1 4.14×10-9 17.4 -7 7 5 3.87×10 7.2 4.82×10-9 17.8 -7 8 6 2.93×10 9.0 5.29×10-9 18.9 9 7 2.26×10-7 7.9 4.62×10-9 17.6 10 8 2.15×10-7 20.0 4.36×10-9 20.2 11 9 1.43×10-7 9.3 3.87×10-9 17.8 12 10 9.52×10-8 11.0 3.33×10-9 18.1 Table II and Fig. 3 show that the thermal is dominated by gamma rays produced by the neutron flux has the maximum value of thermal neutron capture reactions. At the 3.33×107 n/cm2/s at 0.5-cm depth and then surface of the phantom, gamma dose increases decreases to 6.41×106 n/cm2/s at 3-cm depth in from 4.06×10-9 Gy (without the phantom) to the polyethylene phantom. 1.18×10-6 Gy (with the phantom). The gamma As can be seen in Table III and Fig. 4, dose in the phantom arises due to the thermal the gamma dose in the polyethylene phantom neutron capture reaction as 1 H(n,g)2 H . 7 3.5x10 -6 1.8x10 Gamma dose from reactor with phantom -6 7 1.6x10 Gamma dose from reactor without phantom ) 3.0x10 -6 /s 1.4x10 2 -6 2.5x107 1.2x10 n/cm -6 ( 1.0x10 -7 2.0x107 8.0x10 6.0x10-7 1.5x107 4.0x10-7 Dose (Gy) Dose 2.0x10-7 1.0x107 6.0x10-9 5.0x10-9 Thermal neutron Thermal flux 6 5.0x10 4.0x10-9 3.0x10-9 0.0 2.0x10-9 0 2 4 6 8 10 0 2 4 6 8 10 Depth in phantom (cm) Depth in phantom (cm) Fig. 3. Thermal neutron flux distribution according Fig. 4. Gamma doses distribution according to the to the phantom depth phantom depth Figs. 3 and 4 show that the thermal thermal neutron capture reaction as neutron flux and the gamma dose curves have 1H(n,g)2 H . the same shape. The reason is that the gamma dose mixed in the incident neutron beam from MCNP5 calculations have shown that the reactor core is relatively low. It means that the thermal neutron dose is much higher the gamma dose is mainly dominated by the than the gamma dose (Fig. 5), which is the 5
6. SIMULATION OF ABSORBED DOSES DISTRIBUTION IN A POLYETHYLENE PHANTOM difference of BNCT applications that are total absorbed dose mainly due to the based on the horizontal neutron channel in contribution of the boron and thermal comparison to the thermal column of a neutron doses, and it has the maximum nuclear reactor. In Fig. 6, the absorbed value of 7.87×10-5 Gy at the 0.5-cm depth doses along the central axis of the and decreases rapidly to 1.52×10-5 Gy at the polyethylene phantom are presented. The 3-cm depth in the phantom. 5.0x10-6 8.0x10-5 Thermal neutron dose Total dose Gamma dose from reactor with phantom 7.0x10-5 Boron dose 4.0x10-6 Thermal neutron dose 6.0x10-5 Gamma dose from reactor with phantom -5 3.0x10-6 5.0x10 4.0x10-5 -6 Dose (Gy) Dose 2.0x10 Dose (Gy) Dose 3.0x10-5 2.0x10-5 1.0x10-6 C = 2 % N 1.0x10-5 CB = 30 ppm, CN = 2 % 0.0 0.0 0 2 4 6 8 10 0 2 4 6 8 10 Depth in phantom (cm) Depth in phantom (cm) Fig. 5. Gamma and thermal neutron dose Fig. 6. Distribution of all absorbed doses according to distribution according to the phantom depth phantom depth IV. CONCLUSIONS head and neck cancer”, Radiation Oncology, 7, pp. 146, 2012. Calculations of thermal neutron flux and [3]. Matsumoto T., “Design of neutron beams for gamma dose rate in the polyethylene phantom boron neutron capture therapy for Triga were performed to evaluate the absorbed dose reactor”, Journal of Nuclear Science and components in BNCT application. Calculated Technology, 33 (2), pp.171-178, 1996. absorbed dose distribution along the central [4]. Barth R. F., Coderre J. A., Vicente M. G. H. axis of the polyethylene phantom shows that and Blue T. E., Boron Neutron Capture the total absorbed dose has the maximum value Therapy, Current Status and Future Prospects, of 7.87×10-5 Gy at the 0.5-cm depth in the pp. 431-459, 2005. phantom, and is mainly dependent on the boron [5]. Myong-Seop K., Byung-Chul L., Sung-Yul H. and thermal neutron doses with the dominant et al., “Development and characteristics of the component of boron dose. Maximum thermal HANARO neutron irradiation facility for neutron flux at the 0.5-cm depth in the applications in the boron neutron capture phantom is 3.33×107 n/cm2/s1. therapy field”, Phys. Med. Biol., 52, pp. 2553- 2566, 2007. REFERENCES [6]. Brockman J., Nigg D. W., Hawthorne M. F. and McKibben C., “Spectral performance of a [1]. Barth R. F., Soloway A. H. and Fairchild R. G., composite single-crystal filtered thermal “Boron Neutron Capture Therapy of Cancer”, neutron beam for BNCT research at the Cancer Research, 50, pp. 1061-1070, 1990. University of Missouri”, Applied Radiation and Isotopes, 67, pp. 223-225, 2009. [2]. Barth R. F., Vicente M. G. H., Harling O. K. et al., “Current status of boron neutron capture [7]. Nigg D. W. and Eng D., “Methods for radiation therapy of high grade gliomas and recurrent dose distribution analysis and treatment 6
7. PHAM DANG QUYET et al. planning in boron neutron capture therapy”, neutron flux and gamma dose in Tehran Int. J. Radiat. Oncol. Biol. Phys., 28(5), pp. research reactor boron neutron capture therapy 1121-1134, 1994. beam line”, J. Canc. Res. Therapy, 12(2), 2016. [8]. Kasesaz Y., Bavarnegin E., Golshanian M. et [14]. Pham Dang Quyet, Pham Ngoc Son, Nguyen al., “BNCT project at Tehran Research Nhi Dien, Trinh Thi Tu Anh and Cao Dong Reactor: current and prospective plans”, Vu, “Simulation design of thermal neutron Progress in Nuclear Energy, 91, pp. 107, 2016. collimators for neutron capture studies at the Dalat Research Reactor”, Asian Journal of [9]. Monshizadeh M., Kasesaz Y., Khalafi H. and Hamidi S., “MCNP design of thermal and Scientific Research, 13(3), pp. 214-218, 2020. epithermal neutron beam for BNCT at the [15]. Nguyen Nhi Dien et al., Utilisation of the Isfahan MNSR”, Progress in Nuclear Energy, Dalat Research Reactor after its core 83, pp. 427-432, 2015. conversion, Joint IGORR 2014/ IAEA [10]. Allen B. J, Moore D. E. and Harrington B. V., Technical Meeting, 17–21 November, Progress in neutron capture therapy for cancer, Bariloche, Argentina, 2014. New York, 1992. [16]. Pham Ngoc Son, Vuong Huu Tan, Nguyen Nhi [11]. Yamamoto T., Matsumura A., Yamamoto K., Dien, Nguyen Xuan Hai, Tran Tuan Anh, Ho Kumada H., Shibata Y. and Nose T., “In- Huu Thang and Cao Dong Vu, “Development phantom two-dimensional thermal neutron of thermal filtered neutron beam based on the distribution for intraoperative boron neutron radial channel No. 2 of Dalat research reactor”, capture therapy of brain tumours”, Phys. Med. The Annual Report for 2010, Science and Biol. 47, pp. 2387–2396, 2002. Technics Publishing House, pp. 21-27, 2010. [12]. Solleh M. R. M., Mohamed A. A., Tajuddin [17]. Nakagawa Y., Pooh K., Kobayashi T. et al., A.A., Rabir M. H., Zin M. R. M., Yazid H., “Clinical review of the Japanese experience with Azman A., Yoshiaki K. and Hiraga F., Neutron boron neutron capture therapy and a proposed and gamma measurement with water phantom strategy using epithermal neutron beams”, J. for boron neutron capture therapy (BNCT) Neuro-Oncol., 62, pp. 87-99, 2003. reactor Triga Puspati, 2014. [18]. Masouli S. F., “Simulation of the BNCT of brain [13]. Bavarnegin E., Sadremomtaz A., Khalafi H. and tumors using MCNP code: beam designing and Kasesaz Y., “Measurement of in-phantom dose”, Iran J. Med. Phys., 9 (3), 2012. 7