Study on method to achieve high transmutation of LLFP using fast reactor

The purpose of this study is to clarify the method to achieve high transmutation rates of four long-lived fission products (79Se, 99Tc, 107Pd, and 129I) using a fast reactor. New LLFP target assemblies were invented in consideration of the suppression of thermal spikes in adjacent fuel assemblies by combining YD2 and YH2 moderators or using a thermal neutron filter material. It was clarified that the high transmutation rate of about 8%/year was achieved, if the new LLFP target assemblies of 4 nuclides were loaded in the blanket region of the sodium cooled, MOX fueled fast reactor. The feasibility of the LLFP transmutation target was clarified through experiments on material properties and fabrication of the LLFP target, YH2 and YD2 moderators.

Optimization of double layered beam shaping assembly using genetic algorithm

The genetic algorithm method is a new method used to obtain radiation beams that meet the IAEA requirements. This method is used in optimization of configurations and compositions of materials that compose double layered Beam Shaping Assembly (BSA). The double layered BSA is modeled as having two layers of material for each of the components, which are the moderator, reflector, collimator, and filter. Up to 21st generation, the optimization results in four (4) individuals having the capacity to generate the most optimum radiation beams. The best configuration, producing the most optimum radiation beams, is attained by using combinations of materials, that is by combining Al with either one of CaF2 and PbF2for moderator; combining Pb material with either Ni or Pb for reflector; combining Ni and either FeC or C for collimator, and FeC+LiF and Cd for fast and thermal neutron filter. The parameters of radiation resulted from the four configurations of double layer BSA adequately satisfy the standard of the IAEA.

Double Layer Collimator for BNCT Neutron Source Based on 30 MeV Cyclotron

<span>A research of design of double layer collimator using </span>⁹<span>Be(p,n) neutron source has been conducted. The research objective is to design a double layer collimator to obtain neutron sources that are compliant with the IAEA standards. The approach to the design of double layer collimator used the MCNPX code. From the research, it was found that the optimum dimensions of a beryllium target are 0.01 mm in length and 9.5 cm in radius. Collimator consists of a D</span>₂<span>O and Al moderator, Pb and Ni as a reflector, and Cd and Fe as a thermal and fast neutron filter. The gamma filter used Bi and Pb. The quality neutron beams emitted from the double layer collimator is specified by five parameters: epithermal neutron flux 1 ×10</span>⁹<span> n/cm</span>²<span>s; fast neutron dose per epithermal neutron flux 5 ×10</span>¹³<span> Gy cm</span>²<span>s; gamma dose per epithermal neutron flux 1×10</span>¹³<span> Gy cm</span>²<span>s; ratio of the thermal neutron flux of epithermal neutron flux 0; and the ratio of epithermal neutron current to total epithermal neutron 0.54.</span>

OPTIMIZATION OF A NEUTRON BEAM SHAPING ASSEMBLY DESIGN FOR BNCT AND ITS DOSIMETRY SIMULATION BASED ON MCNPX

This article involves two main objectives of BNCT system. The first goal includes optimization of 30 MeV Cyclotron-based Boron Neutron Capture Therapy (BNCT) beam shaping assembly. The second goal is to calculate the neutron flux and dosimetry system of BNCT in the head and neck soft tissue sarcoma. A series of simulations has been carried out using a Monte Carlo N Particle X program to find out the final composition and configuration of a beam shaping assembly design to moderate the fast neutron flux, which is generated from the thick beryllium target. The final configuration of the beam shaping assembly design includes a 39 cm aluminum moderator, 8.2 cm of lithium fluoride as a fast neutron filter and a 0.5 cm boron carbide as a thermal neutron filter. Bismuth, lead fluoride, and lead were chosen as the aperture, reflector, and gamma shielding, respectively. Epithermal neutron fluxes in the suggested design were 2.83 x 109 n/s cm-2, while other IAEA parameters for BNCT beam shaping assembly design have been satisfied. In the next step, its dosimetry for head and neck soft tissue sarcoma is simulated by varying the concentration of boron compounds in ORNL neck phantom model to obtain the optimal dosimetry results. MCNPX calculation showed that the optimal depth for thermal neutrons was 4.8 cm in tissue phantom with the maximum dose rate found in the GTV on each boron concentration variation. The irradiation time needed for this therapy were less than an hour for each level of boron concentration.Keywords: Optimization, Beam Shaping Assembly, BNCT, Dosimetry, 30 MeV Cyclotron, MCNPX. OPTIMASI DESAIN KOLIMATOR NEUTRON UNTUK SISTEM BNCT DAN UJI DOSIMETRINYA MENGGUNAKAN PROGRAM MCNPX. Telah dilakukan penelitian tentang sistem BNCT yang meliputi dua tahapan simulasi dengan menggunakan program MCNPX yaitu uji simulasi untuk optimasi desain kolimator neutron untuk sistem BNCT berbasis Siklotron 30 MeV dan uji simulasi untuk menghitung fluks neutron dan dosimetri radiasi pada kanker sarkoma jaringan lunak pada leher dan kepala. Tujuan simulasi untuk mendapatkan desain kolimator yang paling optimal dalam memoderasi fluks neutron cepat yang dihasilkan dari sistem target berilium sehingga dapat dihasilkan fluks neutron yang sesuai untuk sistem BNCT. Uji optimasi dilakukan dengan cara memvariasikan bahan dan ketebalan masing-masing komponen dalam kolimator seperi reflektor, moderator, filter neutron cepat, filter neutron thermal, filter radiasi gamma dan lubang keluaran. Desain kolimator yang diperoleh dari hasil optimasi tersusun atas moderator berbahan Al dengan ketebalan 39 cm, filter neutron cepat berbahan LiF2 setebal 8,2 cm, dan filter neutron thermal berbahan B4C setebal 0,5 cm. Untuk reflektor, filter radiasi gamma dan lubang keluaran masing-masing menggunakan bahan PbF2, Pb dan Bi. Fluks neutron epithermal yang dihasilkan dari kolimator yang didesain adalah sebesar 2,83 x 109 n/s cm-2 dan telah memenuhi seluruh parameter fluks neutron yang sesuai untuk sistem BNCT. Selanjutnya uji simulasi dosimetri pada kanker sarkoma jaringan lunak pada leher dan kepala dilakukan dengan cara memvariasikan konsentrasi senyawa boron pada model phantom leher manusia (ORNL). Selanjutnya model phantom tersebut diiradiasi dengan fluks neutron yang berasal dari kolimator yang telah didesain sebelumnya. Hasilnya, fluks neutron thermal mencapai nilai tertinggi pada kedalaman 4,8 cm di dalam model phantom leher ORNL dengan laju dosis tertinggi terletak pada area jaringan kanker. Untuk masing-masing variasi konsentrasi senyawa boron pada model phantom leher ORNL supaya dapat mematikan jaringan kanker, membutukan waktu iradiasi neutron kurang dari satu jam.Kata kunci: Optimasi, Kolimator, BNCT, Dosimetri, Siklotron 30 MeV, MCNPX

A Conceptual Design Optimization of Collimator With 181Ta as Neutron Source for Boron Neutron Capture Therapy Based Cyclotron Using Computer Simulation Program Monte Carlo N Particle Extended

The optimization of collimator with 30 MeV cyclotron as neutron source and ¹⁸¹Ta as its proton target. cyclotron assumed work at 30 MeV power with 1 mA and 30 kW operation condition. Criteria of design based on IAEA’s recommendation. Using MCNPX as simulator, the result indicated that with using ¹⁸¹Ta as target material with 0.55 cm thickness and 19 cm diameter, 25 cm and 45 cm PbF₂ as reflector and back reflector, 30 cm ³²S as a moderator, 20 cm ⁶⁰Ni as fast neutron filter, 2 cm ²⁰⁹Bi as gamma filter, 1 cm ⁶Li₂CO_3- polyethylenes as thermal neutron filter, and 23 cm diameter of aperture, an epithermal neutron beam with intensity 4.37 x 10⁹ n.cm^-2.s^-1, fast neutron and gamma doses per epithermal neutron of 1.86 x 10^-16Gy.cm².n^-1and 1.93 x 10^-13Gy.cm².n^-1, minimum thermal neutron per epithermal neutron ratio of 0.003, and maximum directionality 0,728, respectively could be produced. The results have passed all the IAEA’s criteria.

Neutron filter efficiency of beryllium and magnesium fluorides

The neutron filter efficiency of MgF2 and BeF2 has been investigated as a function of neutron incident energy at different temperatures, starting from the phonon density of states (PDOS) calculated using first-principles techniques, and the results are compared with those of MgO and BeO. Recently, MgF2 has been suggested as a neutron filter and neutron transmission through it has been experimentally studied. For MgF2, excellent agreement between calculated and available experimental data has been achieved for the phonon dispersion relations, constant-volume specific heat, inelastic scattering cross sections and neutron transmission. The PDOSs of MgF2 and BeF2 are found to differ significantly owing to the crystal structure and the cations' mass difference. The inelastic scattering cross sections and filter efficiencies of MgF2 and BeF2 show different behaviours, which can be understood to originate from the above PDOS incongruity and the large difference in absorption cross sections of Be and Mg nuclei. BeF2 is predicted to be a better neutron filter than MgF2 and MgO, over the temperature range of interest, while it has less ability than BeO to transmit low-energy neutrons.

Optimization of a Beam Shaping Assembly Design for Boron Neutron Capture Cancer Therapy Facility Based on 30 MeV Cyclotron

A series of simulations has been carried out using a Monte Carlo N Particle X code to find out the final composition and configuration of a neutron Beam Shaping Assembly (BSA) to moderate the fast neutron flux which is generated from the thick disk-type beryllium target. The final configuration for neutron BSA design included 35 cm lead as reflector, 39 cm alumina as moderator, 8.2 cm lithium fluoride as fast neutron filter and 0.5 cm boron carbide as thermal neutron filter. Bismuth, lead fluoride, and lead were chosen as the aperture, reflector, and gamma shielding, respectively. The disk-type of beryllium target is 19 cm in diameter with 0.5 cm thickness which is covered by copper plate to hold the water pressured coolant. A higher yield of neutron production requires a higher intensity of proton beams, which generate much heats and causes the target material to melt. Therefore, it is useful to consider the temperature distribution on the target material with flowing water coolant by means of computer modeling while designing the target. ANSYS-Fluent code will be used to estimate the thermal transfer and heat calculation in a solid target during beam irradiation. Epithermal neutron flux in the suggested design were 1,03x10⁹ n/cm² s, with almost all IAEA parameters for BNCT BSA design has been satisfied.