Evaluation of Applicability of Geant4 and OpenMC in Tritium Breeding Calculation for Fusion Blanket

Abstract

Tritium breeding calculation is a crucial aspect of fusion blanket design. Currently, the typical particle transport code MCNP is extensively utilized in tritium breeding research, though the distribution of its new version is strictly restricted. The open-source Monte Carlo particle transport code Geant4 and OpenMC are regarded as possible substitutes, but their accuracy in tritium breeding calculation for fusion blanket has not been verified sufficiently. In this paper, we evaluated the suitability of Geant4 and OpenMC for tritium breeding calculation based on experimental data and the MCNP results, focusing on two types of blanket mock-ups and a novel blanket made of spent nuclear fuel. The results indicate that both Geant4 and OpenMC are suitable for tritium breeding calculations, though the overall deviation in Geant4 is slightly larger. OpenMC has better pertinence and ease of use for neutron transport problems and smaller TBR deviation. However, discrepancies are observed in the calculation of fission nuclide reaction rates for the spent nuclear fuel blanket.