Abstarct: Neutron sources are suitable for some applications such as medical, industrial, agricultural etc. The main aim of this thesis is devoted to the design and simulation of appropriate neutron shielding materials baxsed on a 252Cf source, using MCNPX code to reduce weight and volume of neutron shielding structures baxsed on multi-laxyered materials. Another purpose of this study is to compare neutron flux and dose of 252Cf source with that of 241Am-Be neutron source, using MCNPX code. The proposed design is composed of three-concentric cylinder laxyers with source. It consists of three parts including paraffin and paraffin with 10% graphite as moderator, beryllium as reflector and borated polyethylene and lead Tungstate as thermal neutron and gamma absorber. The results show that compared with the traditional shielding structures, the volume and the weight of the proposed design is significantly decreased by about 89% and 58%, respectively. Finally, in addition to designing shield for 252Cf neutron source, we consider some radiation sites that are used in applications such as NAA, PGNAA etc.