Research on fission and fusion neutron shielding performance of various materials

Abstract

One of the most challenges in neutron shielding analyses is the selection of materials that protect the components of nuclear power plants and workers against radiation effects, including neutron and gamma fluxes. The purpose of this study is to examine the neutron reduction effect of various single and multi-element materials. To achieve that MCNP neutron transport and an analytical approach have been proposed to model experiments setup with complex geometry. The materials are inserted between neutron source and the water tank to reduce the neutron effect in the model. Firstly, fast neutron removal cross-section of each material has been calculated. Then, the surface flux tally has been performed to calculate neutron flux as a function of distance and energy for thermal and fast neutrons. The neutron fluxes are decreased depending on the distance and absorbing properties of each candidate. Regarding neutron attenuation performance, the materials have different neutron reduction abilities for each neutron range. Both MCNP and analytical results have been compared with experimental data. The obtained data for thermal and fast neutrons can be useful for the design and development of component and fission-fusion reactors.