New developments are under way to reduce the weight and volume of neutron shielding structures using multi-layered materials. The present study aimed at designing and simulating an appropriate neutron shielding material based on a ۲۵۲Cf source using MCNPX code. The proposed design is composed of concentric cylinders and sphere layers with a source. The shielding matter consists of paraffin and paraffin +۱۰% graphite as a moderator, beryllium as a reflector and multiplier and boron carbide and lead tungsten as a thermal neutron and gamma absorber, respectively. The results indicate that, compared to previously reported shielding assemblies, the volume and the weight of the proposed design could be significantly reduced by about ۹۷% and ۷۵%, respectively. Thermal and fast neutron fluxes in the irradiation channel were optimized to achieve maximum values for NAA, PGNAA and other applications.