SUMMARY

A series of accident simulations were performed using INL’s thermal hydraulics code RELAP5-3D to analyze steady-state and transient behavior of a sodium cooled fast reactor. The reactor chosen for this study was General Electric’s S-PRISM, which is a 1,000 MWt pool-type sodium-cooled fast reactor, designed for either an Oxide or Metal fueled core. CEA’s reactor physics code ERANOS 2.0 was used to calculate key core characteristics including power profiles, reactivity feedback coefficients and delayed neutron parameters. S-PRISM was then redesigned for a Nitride fueled core to take advantage of the Nitride fuel’s high thermal conductivity and melting temperature. Loss of flow, loss of heat sink, loss of power and inadvertent control rod withdrawal accidents were simulated for each core at beginning, middle and end of cycle to determine if one fuel type provides significant safety advantages over the others.