Monte Carlo simulation of dose and residual activation in a polyethylene shuttle of the TRIGA Mark II reactor

Abstract

This study focuses on the use of polyethylene as the primary material for the fabrication of an irradiation shuttle intended for the TRIGA Mark II reactor. Designed to accommodate multiple sample capsules, the shuttle is subjected to prolonged neutron irradiation under real operating conditions. The objective is to analyze the interactions between thermal neutrons and polymeric materials in order to better understand their structural and radiological behavior under irradiation. The dose distribution was numerically modeled using the PHITS code, enabling precise identification of the regions of maximum intensity within the shuttle. The results provide a basis for optimizing the materials employed in nuclear environments, integrating radiological safety requirements with durability and mechanical performance. • Monte Carlo modeling of the radiological behavior of polyethylene under TRIGA neutron irradiation. • Identification of maximum dose regions in the shuttle through precise PHITS simulations. • Analysis of residual activation and its impact on safety and management of irradiated materials. • Optimization of materials for durability and safety in experimental nuclear environments.