Monte Carlo analysis of HDPE using PHITS and MCNP for neutron shielding applications.

The nuclear radiation shielding materials are having a great importance in reactor shielding, nuclear laboratories, radiation sources storage, as well as in space radiation shielding. The experimental methods are time consuming, facility dependent, as well as costly affair in order to manufacture the shielding materials. The alternative approach is to use nuclear transport codes such as GEANT, MCNP, PHITS, FLUKA for nuclear radiation shielding application. The present work is planned to study the prediction capabilities of transport code for calculation of shielding parameters using two different radiation transport codes and comparison with experimental results. The simulation measurements were carried out by Monte Carlo analysis, 241Am–Be neutron source & neutron dosimeter, and High Density Polyethylene (HDPE) were modeled using MCNP 6 and PHITS 3.22 codes. Subsequently, experimental neutron dose rates were measured after different thickness of HDPE samples using live 241Am–Be neutron source and NRF31 dosimeter. The study shows a good agreement between the experimental and simulated results which suggest that the simulation method can be used for the optimization of new shielding composites. [ABSTRACT FROM AUTHOR]