Preliminary Analysis of the (U, Pu)Be₁₃ Fuel in an Intermediate Spectrum Sodium-Cooled Reactor with a Lower Linear Heat Rating

Various methods have been explored to improve the safety characteristics of sodium-cooled fast reactors (SFRs), and one of the primary approaches is to load moderators, such as zirconia or beryllium (Be), into the reactor core. A moderator softens the neutron spectrum in the core; hence, safety-enhancing effects, such as reduced sodium-void reactivity, can be expected. This study uses (U, Pu)Be₁₃ as a fuel, containing the moderator. The purpose of this study is to confirm the feasibility of the (U, Pu)Be₁₃ fuel and the possibility of improving its core safety. The (U, Pu)Be₁₃ fuel is loaded to replace the mixed-oxide (MOX) fuel in a reference SFR core with a lower linear heat rating. Characteristics of the (U, Pu)Be₁₃–fueled core are evaluated by comparing with the reference MOX-fueled core by neutronics analyses.

The results show the feasibility of designing a (U, Pu)Be₁₃–fueled core with an intermediate neutron spectrum, where the sodium-void reactivity can be reduced by 3.8 $ from that of the reference MOX-fueled core, showing a negative sodium-void reactivity is achievable. In addition, it was found that the maximum fuel temperature for an unprotected loss-of-flow accident can be mitigated by two factors: (1) the temperature rise at the unprotected loss-of-flow accident can be reduced by appropriately adjusting the Be weight fraction in the fuel, and (2) the higher thermal conductivity of the (U, Pu)Be₁₃ fuel than that of the MOX fuel. These results indicated that the (U, Pu)Be₁₃ fuel has the potential to design an intermediate spectrum sodium-cooled reactor with improved safety performance.