Neutronic Analysis of Molten Salt Fast Reactor Utilizing Different Initial Fuel Loading

Molten salt reactors have the capability of operating in the thermal, epithermal, and fast neutron spectra and can also use different fuels to produce fission. These reactors utilize the thorium fuel cycle using molten fluoride or chloride salts as coolants. In this work molten salt fast reactor is simulated using MCNP6. Three initial fuels are studied, 233U, 235U, 239Pu. 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. Moreover, material evolution during the period of burnup is studied for the three types of fuel.