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Tiểu ban B: Vật lý hạt nhân, Số liệu hạt nhân, Phân tích hạt nhân Máy gia tốc Section B: Nuclear physics, Nuclear data, Nuclear analysis and Accelerator NGHIÊN CỨU TÍNH TỐN VÀ MƠ PHỎNG Q TRÌNH SẢN XUẤT 99mTc BẰNG PHƯƠNG PHÁP CHIẾU XẠ NEUTRON DÂY MoO3 SỬ DỤNG MCNPX THEORETICAL CALCULATION AND SIMULATION OF 99MTC PRODUCTION PROCESS BY NEUTRON IRRADIATION OF MOO3 WIRE USING MCNPX CODE N.M QUACH1, V.T NGUYEN1,*, M.D DO2, H SUEMATSU2 Department of Nuclear Engineering and Environmental Physics, School of Engineering Physics,Hanoi University of Science and Technology (HUST) Department of Nuclear System Safety Engineering, Extreme Energy-Density Research Institute, Nagaoka University of Technology * E-mail: thai.nguyenvan@hust.edu.vn Tóm tắt: Đồng vị phóng xạ Technitium-99m (99mTc), sản phẩm phân rã β 99Mo sử dụng phổ biến y tế, tạo lò phản ứng hạt nhân hai phương pháp bản: phản ứng phân hạch 235U làm giàu cao phản ứng bắt nơtron 98Mo Mặc dù phương pháp phân hạch tạo đồng vị 99Mo có hoạt độ cao lại gây lo ngại an ninh hạt nhân Phương pháp (n, γ) có quy trình đơn giản với nhiều ưu điểm, nhiên, có số hạn chế suất thấp hoạt độ hàm lượng trình chiết xuất phức tạp Rất nhiều giải pháp đề xuất để đơn giản hóa kỹ thuật chiết xuất nâng cao hiệu thu hồi 99Mo Bia MoO3 cấu trúc xốp với diện tích tiếp xúc bề mặt lớn sau chiếu xạ tạo 99Mo với khả hịa tan cao thu hồi lại nước cách đơn giản Bài báo trình bày kết nghiên cứu điều kiện chiếu xạ thích hợp dây MoO3 có cấu trúc xốp điều kiện Lò phản ứng nghiên cứu hạt nhân Đà Lạt Tính tốn lý thuyết thời gian chiếu xạ thực dựa phân bố thông lượng neutron vị trí khác kênh Ngồi ra, q trình kích hoạt neutron 98Mo dây MoO3 xốp mô phần mềm MCNPX để kiểm tra tương tác hạt Kết tính tốn, phân tích thảo luận báo sử dụng trình thiết kế thực thí nghiệm Lị phản ứng nghiên cứu hạt nhân Đà Lạt Từ khố: 99Mo/99mTc, chiếu xạ, tính tốn lý thuyết, mơ MCNPX Abstract: The most widely used medical radioisotope, Technitium-99m (99mTc), is a daughter product of β the decay of 99 Mo which can be produced in nuclear reactors by two basic methods: fission reaction of highly enriched 235U and neutron capture reaction of 98Mo Although the first method can produce very high specific activity obtainable of 99Mo, it causes nuclear security concerns The (n, γ) method uses a simple procedure with many advantages, however, there are some limitations such as lower productivity at both activity and quantity due to the complex extraction process A lot of efforts have been carried out to simplify the extraction techniques as well as improve the recovery efficiency of 99Mo after the irradiation process Porous MoO3 target structure with a large surface area can provide high solubility of 99Mo/99mTc after being irradiated and it is possible to recover 99Mo/99mTc simply by pouring water This paper presents an investigation of appropriate irradiation conditions for MoO3 wire with porous structure in Dalat Nuclear Research Reactor Theoretical calculation of irradiation time was obtained based on neutron flux distribution at different channels and locations Also, the neutron activation process of 98Mo in porous MoO3 wire was simulated with MCNPX code to examine the interactions between the particles Results were analyzed, discussed, and will be further applied in experiments at Dalat Nuclear Research Reactor Keywords: 99Mo/99mTc, irradiation, theoretical calculation, MCNPX simulation INTRODUCTION 99Mo can be produced in nuclear reactors by two basic methods: fission reaction of highly enriched U and neutron capture reaction of 98Mo The fission method causes nuclear security concerns such as generating large quantities of radioactive waste and reprocessing of the unused uranium targets due to proliferation concerns [1] Although the simple method with neutron capture reaction (n, γ) is being developed to replace, there are some limitations of low yields in both activity and quantity [2] Solutions were proposed to overcome these burdens, such as using a high density of MoO3 pellet or using solutions with extremely adsorptive However they are limited at laboratory scale for 99Mo production due to the complicated extraction process [3] With the purpose to improve the recovery 99mTc from 99Mo, Seki [4] used α-MoO3 powder which was irradiated in an experimental reactor and then dispersed in water After the separation and precipitation process, the concentration of 99Mo in water is higher than the solubility limit It is probable that 99Mo isotopes can behave as “hot atoms”, and water can be used to recover 99Mo 235 153 Tuyển tập báo cáo Hội nghị Khoa học Cơng nghệ hạt nhân tồn quốc lần thứ 14 Proceedings of Vietnam conference on nuclear science and technology VINANST-14 from the irradiated MoO3 target [4] From this finding, we proposed a new target in porous form for the irradiation experiment Porous target containing many pores (voids) with large surface area Moreover, when using wire instead of powder to create a porous structure, the unreacted components are fixed avoiding contamination of the recovery solution Therefore, porous 98MoO3 target can provide high solubility of 99Mo/99mTc after being irradiated and it is possible to recover 99Mo/99mTc simply by pouring water 98MoO3 wire will be heated in the electric furnace to form a porous structure and then irradiated in DNRR to produce 99Mo This paper presents preliminary calculations by theory and MCNPX simulation to investigate appropriate irradiation conditions for MoO3 wire with porous structure in DNRR METHODOLOGY 2.1 Theoretical background decays into a meta-stable isomer of 99Tc through the emission of a beta particle with a half-life time of 66 hours Technetium-99m then (a half-life time of hours) decays into 99Tc, which is also unstable but has a long half-life time (2.13x105 years) Fig shown the decay scheme of 99Mo and 99mTc The nuclear properties of 99mTc are ideal for medical imaging as it emits readily available photon energy (~140 keV) that is sufficient to determine the exact molecular structure of the coordination compounds by scintillation instruments such as gamma cameras [5] The data collected by the gamma camera are analyzed to produce detailed structural and functional images of certain human organs that are otherwise difficult or impossible to image In the neutron capture process, 99Mo is produced by irradiation of molybdenum (98Mo) in thermal neutron flux, {98Mo (n,𝛾)99Mo} 99Mo Figure Decay scheme of 99Mo and 99mTc The thermal (~ 0.025 eV) or epithermal (0.025–1.0 eV) neutrons produced by the fission of uranium in a nuclear reactor can be used to attack stable 98Mo target material installed in a nuclear reactor In this process, the 98Mo nuclei capture a neutron and transform into a 99Mo isotope as shown in Figure Figure Neutron activation of 98Mo to generate 99Mo Reaction cross-sections depict how the probability of a particular reaction occurrence changes with respect to projectile energy (in the case of neutron energy, see Figure 3) It can be divided into a number of regions, with the most important regions as thermal, resonance, and fast neutron The major interest is in the thermal energy region where both the cross-section and the neutron flux are high In the thermal region, the microscopic cross-section of 98Mo is approximately 0.13 b (barns) 154 Tiểu ban B: Vật lý hạt nhân, Số liệu hạt nhân, Phân tích hạt nhân Máy gia tốc Section B: Nuclear physics, Nuclear data, Nuclear analysis and Accelerator Figure Cross-sections of the 98Mo (n, γ) 99Mo reaction As with the thermal fission case, the amount of 99Mo produced from this reaction is dependent on the thermal neutron flux 𝜑[n.cm-2.s-1], the number of 98Mo atoms initial condition 𝑁𝑀𝑜#$%,0 [𝑛], and the irradiation time t [s] Since some amount of 98Mo will be consumed by the capture reaction, it is necessary to establish the variation of 98Mo concentration Since 98Mo is stable and is not being produced by other reactions, the change in its population is due entirely to thermal neutron capture Therefore, the amount of 98 Mo presentin the system at time t is determined as: (1) 𝑁𝑀𝑜−98 = 𝑁𝑀𝑜−98,0 𝑒(−𝜑𝜎𝑀𝑜−98 )𝑡 99Mo is produced only through the neutron capture of decay or further neutron capture: 𝑁𝑀𝑜−99 = 𝜑𝜎𝑀𝑜−98 𝑁𝑀𝑜−98,0 𝜆𝑀𝑜−99 + 𝜑(𝜎𝑀𝑜−99 − 𝜎𝑀𝑜−98 ) 98 Mo but also lost through radioactive [𝑒(−𝜑𝜎𝑀𝑜−98 )𝑡 − 𝑒−(𝜆𝑀𝑜−99 +𝜑𝜎𝑀𝑜−99 )𝑡 ] 𝑙𝑛2 where the decay constant for 99Mo is determined by 𝜆 = 𝑇 1/2 (2) [s-1] If secondary absorptions are neglected, the solution gives the irradiation period t as follows: 99 Mo activity (Bq) produced after 𝐴𝑡ℎ [𝐵𝑞/𝑔] = 𝜆𝑀𝑜−99 𝑁𝑀𝑜−99 = 𝜑𝜎𝑀𝑜−98 𝑁𝑀𝑜−98,0 [1 − 𝑒 −𝜆𝑀𝑜−99 𝑡 ] (3) 2.2 MCNPX code simulation Dalat reactor and its parameters: The DNRR was rebuilt and upgraded from the USA TRIGA MARK II reactor, which was 500 kW pool - typed, light water - cooled, and moderated Radioisotope manufacturing, neutron activation analysis, fundamental and applied research, and workforce training are all now possible with the reactor in operation In the working configuration of DNRR, the neutron trap was designed to have maximum thermal neutron flux in the reactor core Neutron trap is a water cylinder surrounded by Beryllium blocks located in the center of the core which has 6.5cm in diameter and about 2050cm3 in volume MCNPX code was used to determine the neutron flux distribution at the neutron trap position and reaction rate of neutron activation The MCNPX has been being officially used for core management of DNRR with ENDF/B 7.0 library [7] The calculation model for DNRR using MCNPX computer code is shown in Figure The parameters of 98Mo wire is listed in Table 155 Tuyển tập báo cáo Hội nghị Khoa học Công nghệ hạt nhân toàn quốc lần thứ 14 Proceedings of Vietnam conference on nuclear science and technology VINANST-14 Figure Calculation model of DNRR using MCNP5 computer code Table Summary description of the sample Aluminum tube MoO3 wire Diameter Outer diameter: 1.5cm Inner diameter: 1.25 cm Diameter: 0.5 cm Length 10 cm 10 cm Component 100% Aluminum 75% of mass 98Mo, 25% of mass 16O Estimates of 99Mo yield from neutron capture on 98Mo may be produced by integrating the total capture cross-section and neutron flux at the energy ranging from 10-11 to 20-1 MeV The thermal neutron flux and neutron absorption cross-section is divided into corresponding bins according to the energy For small quantities of 98Mo (N98), where self-shielding and target burn-out effects are negligible, the 99Mo production rate can be approximated by (1) 𝑑𝑁99 = −𝜆𝑁99 + ∑ 𝜎̅𝑗 𝜙̅𝑗 𝑁98 𝑑𝑡 𝑗 where t is irradiation time, 𝜆 is the 99Mo decay constant (𝜆 = 𝑙𝑛2/ 𝑇1/2 ), 𝜎̅𝑗 is the energy group javeraged (n,) cross – section, and 𝜙̅𝑗 the energy group-averaged neutron flux which calculated by MCNPX The transmutation of the 99Mo product has been neglected—its inclusion would simply introduce a modification to 𝜆 in Equation that is of the order of the uncertainty of itself The solution of Equation gives the 99Mo activity (Bq) produced after irradiation period t according to: 𝜆𝑁99 (𝑡) = (1 − 𝑒 −𝜆𝑡 ) 𝑁98 ∑ 𝜎̅𝑗 𝜙̅𝑗 (2) 𝑗 where the 98Mo in equation (5) is the number of 98Mo atoms of course calculated from the mass of target molybdenum oxide MoO3 𝑁98 = 𝑁𝑀𝑜−98,0 = 𝑚𝑀𝑜𝑂3 × 75% × 𝑁𝐴 97.9054 𝑔/𝑚𝑜𝑙 − 𝑀𝑜98 (3) where 𝑁𝐴 is the Avogadro constant The equation activity increases linearly with t for irradiation periods much less than the 99Mo decay half-life (T1/2) of 66h In this way, the short-time production rate within the approximation given is calculated 156 Tiểu ban B: Vật lý hạt nhân, Số liệu hạt nhân, Phân tích hạt nhân Máy gia tốc Section B: Nuclear physics, Nuclear data, Nuclear analysis and Accelerator On the other hand, using the FM card in MCNPX code, the tally printout will indicate the number of 99Mo (atoms/ cm3 ) produced as a result of (n, 𝛾) capture with 98Mo The number of reactions was calculated by equation (7) below: (4) 𝑄 = 𝐶 ∫ 𝜙(𝐸) 𝜎(𝐸) 𝑑𝐸 where 𝜙(𝐸): Neutron flux (neutrons/cm2.s), 𝜎(𝐸): Microscopic cross section (barn), C: Normalization constant (atoms/barn.cm) 𝐶𝑀𝑜𝑂3 = (5) (𝜌𝑀𝑜𝑂3 )(0.6022)(𝜀𝑀𝑜𝑂3 ) 𝑀𝑀𝑜𝑂3 where 𝜌𝑥: Density (g/cm3), 0.6022: Result of product of the Avogadro’s Number and equivalence between barn unit and cm-2 (atoms/mol), 𝜀𝑀𝑜7): MoO3 in the material, 𝑀𝑀𝑜7): Molecular weight of the MoO3 (g/mol) Fig Input card in MCNPX code In this way, the 99Mo activity (Bq) produced after irradiation was calculated according to: 𝑄[ 𝐴𝑟𝑒𝑎𝑐𝑡𝑖𝑜𝑟𝑎𝑡𝑒 [𝐵𝑞/𝑔] = 𝑎𝑡𝑜𝑚𝑠 ] × 𝑉𝑐𝑒𝑙𝑙 [𝑐𝑚3 ] × 𝜆 [ 𝑠 ] 𝑐𝑚3 𝑔 𝜌 [ ] × 𝑉𝑐𝑒𝑙𝑙 [𝑐𝑚3 ] 𝑐𝑚 (6) RESULTS AND DISCUSSION Neutron spectra were divided into energy groups Neutron spectra obtained from calculated at the neutron trap are shown in Table This result of input has not updated the burnof the fuel Table Neutron flux of the neutron trap using MCNPX Flux Calculation Range Energy (n/cm s) Relative Error (%) Thermal neutrons ~ 0.025 eV 1.48.1011 0.15 Resonance neutrons 0.025 eV – 0.4 eV 1.23.1011 0.15 Fast neutrons MeV – 20 MeV 5.65 1010 2.26 The Fig shows that the maximum thermal neutron flux value is located 20cm away from the core bottom The sample was inserted into the neutron trap at the position where have maximum thermal neutron flux The shift of maximum flux is mainly affected by the control rod position 157 Tuyển tập báo cáo Hội nghị Khoa học Cơng nghệ hạt nhân tồn quốc lần thứ 14 Proceedings of Vietnam conference on nuclear science and technology VINANST-14 1.00 1.0 0.98 0.96 Relative unit Relative unit 0.8 0.94 0.6 0.4 0.92 0.2 0.90 -3 -2 -1 0.0 10 20 Position[cm] 30 40 50 60 Position[cm] Figure Thermal neutron distribution at neutron capture a) Thermal neutron distribution in radialdirection b)Thermal neutron distribution in axial direction The decay constant for 99Mo (66h half-life) is determined by 𝜆 = 2.9167 × 10-6 𝑠-1.The thermal neutron absorb cross-section for 98Mo(n, 𝛾)99Mo is approximately 0.13barn so that 𝜎 = 0.13𝑏 = 1.3 × 10-25 𝑐𝑚2, thermal neutron flux of DNRR take from the result simulated by MCNPX code is 1.48x1011 n/cm2.s So the result theoretical calculation of 99Mo activity (Bq) produced after irradiation is 8922 [Bq/g.s] With the MCNPX code, the cross-section and thermal neutron flux were divided into eighty-eight bins based on energy see (Fig 7) Substituted back into Equation (5), the short-time production rate which calculated using the average group is 8811[Bq/g.s] By using the FM card in tally F4, the number of 99Mo produced as a result of (n, 𝛾) capture with 98Mo is 1.295x1010 atoms/cm3 Substituted back into Equation (9), 99Mo activity (Bq) produced after irradiation is 8056 [Bq/g.s] To begin with compare between the theoretical calculation and the calculation using the average group In the theoretical calculation using the average thermal neutron absorption cross-section of 98Mo multiplied directly by the thermal neutron flux of DNRR which take from the result of MCNPX On the other hand, in the calculation using the average group, the cross- section of 98Mo and neutron flux are subdivided according to the energy, calculating each group's average value After that multiply the cross-section with the corresponding neutron and the calculated sum of them This is the reason for the small difference between the two values above (1.24%) Secondly, compare between the theoretical calculation and the calculation using the FM card of tally F4 In MCNPX the calculation considers additional factors such as self- shielding, target burned effects, or scattering between materials which theoretical calculations are negligible This is the reason for the difference between the theoretical calculation and the calculation using the FM card (9.71%) 1E10 10000 1E9 1E8 1000 1000000 100 100000 10000 10 1000 100 10 0.1 Cross Section (B) Neutron flux (n.cm-2.s-1) 1E7 0.1 0.01 0.001 1E-4 1E-5 0.01 Neutron Flux Cross section Group AVG 0.001 1E-6 1E-11 1E-10 1E-4 1E-9 1E-8 1E-7 1E-6 1E-5 1E-4 1E-3 0.01 0.1 10 E(Mev) Figure Log-log plot superimposing the Neutron Flux of Position neutron trap with the Cross sectionfor the reaction 98 Mo(n,𝛾)99Mo The group average shows the integrated flux over the discrete 88 energy bin 158 Tiểu ban B: Vật lý hạt nhân, Số liệu hạt nhân, Phân tích hạt nhân Máy gia tốc Section B: Nuclear physics, Nuclear data, Nuclear analysis and Accelerator 1.2E+15 The number 99 Mo present (atoms) 1.0E+15 8.0E+14 6.0E+14 4.0E+14 Theoretical Average Group MCNPX code 2.0E+14 0.0E+00 0E+00 1E+06 2E+06 3E+06 4E+06 5E+06 6E+06 7E+06 Time(s) Figure The number 99Mo present at coressponding time The 99Mo activity was calculated as a function of time Based on Fig 8, the time to the number of Mo became saturated can be determined At that time, the number of 99Mo produced by neutron capture of 98Mo was equal to the number of 99Mo lost by decay With this result, after synthesizing the porous sample, measure the parameter of the sample, and determine the specific activity at least essential of 99Mo, the position insert sample and time irradiation can be calculated In the DNRR, besides neutron traps, the configuration has two other positions where can insert the sample such as wet channels 1-4 and dry channels 13-2, 1- Therefore, in the future, the research will continue to be studied at the two above positions 99 CONCLUSION The growing demand for 99mTc and the inherent problem associated with 99Mo production from fission products has increased the interest for economically feasible alternative sources of 99Mo Production of 99Mo via the neutron capture method can be a feasible alternative to fission- derived 99Mo This report proposes the porous MoO3 target made from Mo wire together with the investigation of specific activity also reaction rate of the irradiation process by MCNPX with the conditions of the Da Lat Nuclear Research Reactor Therefore, simplifies the separation of 99Mo/99mTc as well as improves the performance of 99Mo ACKNOWLEDGMENTS This research was funded by the scholarship from the Global Academic-Industry Consortium for Collaborative Education Program (GAICCE), the AUN/SEED-Net project that is supported by the Japanese Government through JICA for the Double Degree Program at Hanoi University of Science and Technology (HUST), Vietnam and Nagaoka University of Technology, Japan REFERENCES [1] Lyra M, et al., “Alternative production methods to face global molybdenum-99 supplyshortage,” National Library of Medicine, 49-55, 2011 [2] RE Boyd, “Molybdenum-99: Technetium-99m generator” , National Center forBiotechnology Information, 123145, 1982 [3] A.Kimura, et al., “Development of 99mTc Extraction Techniques from 99Mo by (n, 𝛾) Reaction”, JEAE Review, 2010-053, 2010 [4] M Seki, (2017), “Development of water nuclide separation method for medical 99Mo 99mTc generator development”, Nagaoka University of Technology [5] CG Whipple, “Medical isotope production without highly enriched uranium,”ScienceEngineering Medicine, 2009 [6] 2012, “Báo cáo phân tích an tồn lị phản ứng hạt nhân Đà Lạt,” Viện Nghiên cứu hạtnhân Đà Lạt [7] MB Chadwick et al., “ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology,” Nuclear DataSheets, 2006 159
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