Development of Linkage Program Code OpenMC and ORIGEN2.2 for Neutronic Analysis and Burnup Nuclear Reactor Program

Abstract

Abstract This paper discusses the development of program code that coupling the ORIGEN2.2 for burn-up analysis with the Monte Carlo program OpenMC for neutron analysis program called OpenMC-ORIGEN (Op-OR). The results of the program have been compared with benchmark results from previously published results and MCNP6 program. The acquired results show a good agreement with benchmark results. This linkage program codes performed well for designing a new advanced reactor and analysing the neutronics parameter and burnup/depletion calculation.