Development of technical database in the unprotected events for level 2 PSA of sodium-cooled fast reactors

ナトリウム冷却高速炉炉停止失敗事象のレベル2PSA技術データベースの整備

山野 秀将   ; 飛田 吉春; 佐藤 一憲 

Yamano, Hidemasa; Tobita, Yoshiharu; Sato, Ikken

As part of the development of a Level 2 probabilistic safety assessment methodology for the risk evaluation of sodium-cooled fast reactors, technical database was developed to quantify the probability of event sequences, focusing on the transition and post-disassembly expansion phases in an unprotected loss of flow accident in this study. Dominant factors were also identified through parametric analyses using the SIMMER-III code. As for the transition phase, in Japan sodium-cooled fast reactor, an inner duct is introduced into a fuel assembly for enhancing molten fuel discharge from disrupted core. In the post-disassembly expansion phase, important headings in developing the event tree included pressurization in the core, energy dissipation effect of internal structures, bubble growth in the upper sodium plenum, and in-vessel structure response. The parametric analyses showed that the energy dissipation effect of internal structures was significant.