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In order to provide an appropriate neutron beam for Boron Neutron Capture Therapy (BNCT), a special Beam Shaping Assembly (BSA) must be designed based on the neutron source specifications. A typical BSA includes moderator, reflector, collimator, thermal neutron filter, and gamma filter. In common BSA, the reflector is considered as a layer which covers the sides of the moderator materials. In this paper, new reflector/moderator geometries including multi-layer and hexagonal lattice have been suggested and the effect of them has been investigated by MCNP4C Monte Carlo code. It was found that the proposed configurations have a significant effect to improve the thermal to epithermal neutron flux ratio which is an important neutron beam parameter. [ABSTRACT FROM AUTHOR]

The feasibility of using a laser-accelerated proton beam to produce a neutron source, via (p,n) reaction, for Boron Neutron Capture Therapy (BNCT) applications has been studied by MCNPX Monte Carlo code. After optimization of the target material and its thickness, a Beam Shaping Assembly (BSA) has been designed and optimized to provide appropriate neutron beam according to the recommended criteria by International Atomic Energy Agency. It was found that the considered laser-accelerated proton beam can provide epithermal neutron flux of ∼2×10 6 n /cm 2 shot. To achieve an appropriate epithermal neutron flux for BNCT treatment, the laser must operate at repetition rates of 1 kHz, which is rather ambitious at this moment. But it can be used in some BNCT researches field such as biological research. [ABSTRACT FROM AUTHOR]

Abstract: Optimization of the Beam Shaping Assembly (BSA) has been performed using the MCNP4C Monte Carlo code to shape the 2.45MeV neutrons that are produced in the D–D neutron generator. Optimal design of the BSA has been chosen by considering in-air figures of merit (FOM) which consists of 70cm Fluental as a moderator, 30cm Pb as a reflector, 2mm 6Li as a thermal neutron filter and 2mm Pb as a gamma filter. The neutron beam can be evaluated by in-phantom parameters, from which therapeutic gain can be derived. Direct evaluation of both set of FOMs (in-air and in-phantom) is very time consuming. In this paper a Response Matrix (RM) method has been suggested to reduce the computing time. This method is based on considering the neutron spectrum at the beam exit and calculating contribution of various dose components in phantom to calculate the Response Matrix. Results show good agreement between direct calculation and the RM method. [Copyright &y& Elsevier]

Abstract: A 3-D model for 1MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k0-INAA and absolute method. The average values of and by MCNP code were (2.19±0.03)×1012 cm−2 s−1, (1.26±0.02)×1011 cm−2 s−1 and (3.33±0.02)×1010 cm−2 s−1, respectively. These average values were consistent with the experimental results obtained by k 0-INAA. The findings show a good agreement between MCNP code results and experimental results. [Copyright &y& Elsevier]