Design of portable shield for neutron sources using MCNP Computer Code

In this work a proposed design of portable shield for neutron sources was performed theoretically using MCNP5 code. The shield is composed of three layers; lead layer for gamma attenuation, PMMA (Poly-methyl methacrylate) layer for fast neutron thermalization and a layer of B2O3 for thermal neutron absorption. The shield materials are combined in stainless steel container of 0.5 mm thick.
The source design verification was carried out by experimental measurements at different source-center distance. The energy spectrum of 241Am-Be source is considered in order to study the neutron moderation due to successive elastic scattering with PMMA composition. The neutron and photon total dose rates was calculated at the surface and one meter from the shield. The results of MCNP5 model were compared with the international transport regulation to ensure the suitability of neutron shield proposed design. The results show that the proposed design of the portable shield satisfies the dose rate limits and can be used for control of external radiation level during neutron source transportation, ensure the safety transport conditions and protection of radiation workers in different applications.