Design of a portable shield for neutron sources using MCNP computer code

Abstract EN

In this work, a design of a portable shield for neutron sources is proposed using MCNP5 code for neutron and gamma total dose rate calculation.

The shield is composed of three layers, as follows: 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 is contained in a stainless steel container of 0.5 mm thick.

The source design verification has been carried out by experimental measurements at different source- to detector distance.

The energy spectrum of an241Am-Be source is considered in order to study the neutron moderation due to successive elastic scattering with PMMA layer.

The neutron and photon total dose rates have been calculated at the surface and one meter from the shield.

The results of MCNP5 model are 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 controlling external radiation level during neutron sources transportation, ensuring the safety transport conditions and protection of radiation workers in different applications.