Preliminary analysis of the (U, Pu)Be fuel in an intermediate spectrum sodium-cooled reactor with a lower linear heat rating

中間スペクトルを有する低線出力型ナトリウム冷却炉における(U, Pu)Be燃料の予備解析

桑垣 一紀  ; 近澤 佳隆  ; Yan, X. 

Kuwagaki, Kazuki; Chikazawa, Yoshitaka; Yan, X.

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)Be as a fuel, containing the moderator. The purpose of this study is to confirm the feasibility of the (U, Pu)Be fuel and the possibility of improving its core safety. The (U, Pu)Be 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)Be-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)Be-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)Be fuel than that of the MOX fuel. These results indicated that the (U, Pu)Be fuel has the potential to design an intermediate spectrum sodium-cooled reactor with improved safety performance.