Neutronic Assessment of Fluoride and Chloride Fuels for Safety in Molten Salt Fast Reactors

Fast breeder reactors have been widely recognized as a promising solution for sustainable energy generation, with Molten Salt Reactors (MSRs) offering significant advantages in fuel cycle performance and resource utilization. Selecting the appropriate fuel composition is crucial for MSRs, as it directly influences neutronic behavior, safety parameters, and fuel cycle performance through changes in the neutron spectrum. Given the wide range of possible fuel salts, carrier salts, and fissile-fertile isotope combinations, this study systematically compares the performance of fluoride and chloride fuel salts in a Molten Salt Fast Reactor (MSFR) operating on the Th-U cycle. For this purpose, four fuel compositions are modeled using MCNP6, with start-up fissile isotopes of U-233 and Transuranic (TRU) isotopes dissolved in fluoride and chloride salts. The model is used to evaluate the flux distribution in the core and blanket, as well as safety parameters namely Doppler and density coefficients. The initial breeding ratio is also estimated. Burnup is performed for a period of six month. During the Burnup, the variation in effective multiplication factor is estimated.