Development and verification of a model for generation of MSFR few-group homogenized cross-sections based on a Monte Carlo code OpenMC

A concept of molten salt fast reactor (MSFR) was proposed in EVOL to burn transuranium element discharged from pressurized water reactors. MSFR is featured by fast spectrum and using liquid fuel salt containing UF4 or ThF4. Some issues are presented, for instance, global material arrangement affects the local neutron spectrum due to long neutron free path, and fluoride salt (LiF-BeF2) has nonnegligible thermal neutron scattering effect. Thus, lattice code prepared for thermal-spectrum reactor is not suitable for MSFR calculation. In this study, “two-step” calculation scheme combining Monte Carlo method and deterministic method was prepared for MSFR calculation. A tool named TRANS was developed to transfer tally data from an open source Monte Carlo code OpenMC into few-group homogenized cross-sections, and one benchmark based on pressurized water reactor and two types of model based on MSFR were used for verification. Besides, the applicability of few-group parameters generated by different model to MSFR whole-core calculation was analyzed. Finally, MSFR neutronics characteristics at steady-state were calculated using MOREL. The results show that the few-group parameters generated by one-dimension (1D) and two-dimension (2D) model are correct, and it is feasible to use OpenMC to generate few-group parameters. In case of 1D homogenization model, few-group parameters by 1D model (b) can give more accurate results both for eigenvalue and flux distribution. In MSFR whole-core calculation, using few-group cross-sections generated by 2D model has better accuracy in flux distribution, however, using few-group cross-sections generated by 1D model has better accuracy in keff calculation. Moreover, the neutronics parameters of MSFR calculated by MOREL code agree well with that by other institutes.