Calculation acceleration for fuel cycle simulation of molten salt reactor based on multi-group cross section method

Abstract

The time-consuming issue of transport calculations is prominent in the burnup calculation of nuclear reactor. Multi-Group Cross Section (MGXS) method is an acceleration technique developed based on the characteristics of Monte Carlo simulation, which can significantly reduce the computation time required to solve a single group cross section in transportation calculations. The effectiveness of the method has been verified in the test calculations of water reactor pins. However, liquid molten salt reactors (MSRs) exhibit significant differences from conventional water reactors in terms of neutron energy spectra and fuel cycle mode. The effectiveness of the MGXS method in MSR burnup simulations remains to be validated, and targeted adjustments are required during its application. In this study, OpenMC and ORIGEN2 are coupled to develop an accelerated calculation method for MSR burnup simulations based on the MGXS approach. The reasonable grouping structure of the MGXS method is explored, and the performance of different grouping structures is tested. Results show that the transport calculation can be accelerated by an average factor of 2.4 for a single burnup zone by using MGXS method and the acceleration effect is generally independent of the grouping structure adopted. The nuclide mass bias compared to the traditional direct solution can be reduced to approximately 1% when the fuel burnup is 250MWd/kg for the LEU loading scheme with the 10000 groups structure. For the TRU loading scheme, the mass bias compared to the traditional direct solution of important nuclides (such as U-233, U-235, Pu-239 and so on) can be controlled below 0.5% at a burnup of 230 MW d/kg. The results indicate that the grouping strategy proposed in this study can achieve the adaptation of MGXS to MSRs, and the 10000 groups structure adopted in the study exhibits good accuracy.