Simulation of Neutron Transport in Shielding Materials Using Monte Carlo Methods for Enhanced Nuclear Reactor Safety and Space Mission Shielding Design

Abstract

The accurate simulation of neutron transport in shielding materials is a cornerstone of radiation protection in nuclear engineering and aerospace applications. Monte Carlo methods provide a statistically robust framework for modeling the stochastic behavior of neutrons as they interact with matter. This paper explores the implementation of Monte Carlo techniques in simulating neutron transport through various shielding materials, emphasizing their application in nuclear reactor containment and spacecraft radiation shielding. Using high-fidelity computational models, the study investigates neutron flux attenuation, energy spectrum shifts, and dose rate reductions in commonly used shielding substances like borated polyethylene, lead, and water. It also evaluates material performance under diverse neutron energy ranges and geometrical configurations. The insights generated aid in optimizing shielding material selection and thickness for both terrestrial and extraterrestrial environments.

Keywords: Neutron transport, Monte Carlo simulation, Radiation shielding, Nuclear reactor safety, Space mission design, Shielding materials, Neutron interaction