Neutronics analysis for MSR cell with different fuel salt channel geometries

Abstract

The neutronic properties of molten salt reactors (MSRs) differ from those of traditional solid fuel reactors owing to their nuclear fuel particularity. Based on the Monte-Carlo N particle transport code, the effects of the size and shape of the fuel salt channel on the neutron physics of an MSR cell are investigated systematically in this study. The results show that the infinite multiplication factor ( k ∞ ) first increases and then decreases with the change in the graphite cell size under certain fuel volume fraction ( FVF ) conditions. For the same FVF and average chord length, when the average chord length is relatively small, the k ∞ values for different fuel salt channel shapes agree well. When the average chord length is relatively large, the k ∞ values for different fuel salt channel shapes differ significantly. In addition, some examples of practical applications of this study are presented, including cell selection for the core and thermal expansion displacement analysis of the cell.