Griffin's Depletion Calculation of Molten Salt Reactor Experiment

This study demonstrates the Molten Salt Reactor Experiment depletion calculation using Griffin. For this purpose, the multigroup macroscopic and one-group microscopic cross sections were generated using Shift, and the decay transmutation library from ORIGEN was converted to an ISOXML file. The Monte Carlo code Serpent2 was also used to benchmark and supplement multigroup cross sections generated by Shift. The 8-group structure was used and was quite when comparing Griffin results to continuous-energy (CE) Monte Carlo results. Additionally, accurate Griffin constant-flux depletion calculations were compared against ORIGEN calculations, and the calculated nuclide densities generally agreed with those of ORIGEN. Moreover, for time-dependent depletion calculations, the average difference in keff between Griffin and CE Monte Carlo results was about 155 pcm. Good agreement in the nuclide density of ²³⁵U and ¹³⁵Xe was also observed, with small differences up to ~0.5%.