Neutron transport and activation comparison between OpenMC and FISPACT-II in ARC-class reactor

Abstract

In a fusion reactor, high-energy neutron fluxes strike the materials causing radiation damage and triggering nuclear reactions that alter the chemical composition of the materials through transmutation. This investigation employs the Monte Carlo code OpenMC with Direct Accelerated Geometry Monte Carlo (DAGMC), a software package that allows users to perform Monte Carlo radiation transport directly on CAD models. The analysis was conducted on an Affordable Robust Compact (ARC) class reactor using a 3D CAD geometry. OpenMC supports depletion calculations allowing for the time evolution of the radioactive inventories evaluation. This study focuses on the machine’s nuclear performance, analyzing tritium production, neutron fluxes, and power deposition to assess the reactor’s behavior. It also explores the primary aspects of neutron irradiation on solid materials in the ARC class reactor, with particular emphasis on neutron-induced activation and displacements per atom (DPA). The neutronic results indicate that the predicted tritium breeding ratio and power density are consistent with both the reactor design specifications and the findings in existing literature, while the DPA and activation calculations reveal key areas of material degradation due to neutron irradiation. To validate the accuracy of the results, a comparison was made with corresponding results obtained using the FISPACT-II code. The agreement between the two codes serves as a benchmark for the reliability of OpenMC in predicting nuclear activation phenomena in fusion reactors.