Abstract
Recent studies related to Monte Carlo based homogenization have shown that we can use a one-step method to generate the macroscopic cross section database (MCSD) of each grid element of the reactor core. This paper presents a systematic study on the influence of energy groups and fuel rod groups on MCSD. The Monte Carlo method for generating multi-group MCSD is studied based on the Monte Carlo program OpenMC with ENDF-B-VII.1 continuous energy cross section library. Through Monte Carlo simulation of the whole model, a fine 69 groups MCSD of lattice level can be directly generated by a one-step method, without leakage correction, and compared with 2 and 8 groups MCSDs. Furthermore, a method is proposed to group the fuel rods in the model according to the symmetry and neutron energy spectrum similarity. OpenMOC was used to test the generated MCSDs. The effective multiplication factors and the power distribution were in good agreement between OpenMC and OpenMOC results. It demonstrates that the Monte Carlo method is feasible to generate the highly accurate multi-group MCSD by the one-step method of full model calculation and the fuel grouping method.
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