Multilayer shielding simulation for a cylindrical 241Am-Be source in order to further reduce neutron equivalent dose using MCNP5 code

In order to simulate neutron shields, MCNP5 calculation code was used and three types of homogeneous and separated shield multilayer arrangement, irradiated with 241Am-Be neutron sources were investigated. In these shields, the polyethylene (C2H4) and polystyrene (C8H8) were used as moderator material, and the boron carbide (B4C), as a thermal neutron absorber material and stainless steel as a absorber gamma rays materials. The ultimate goal of this research, was obtaining the best type of arrangement for the source to achieve the lowest fluence and dose produced outside to ensure their allowable level. To verify the simulation, the results were compared with an experimental work.