Development of Level 2 PSA Methodology for Sodium-Cooled Fast Reactors (3) Development of Technical Basis in the Transition Phase of UnprotectedEvents
Level 2 PSA methodology was developed for the risk evaluation of sodium-cooled fast reactors. For this purpose, a phenomenological event tree was developed as well as a technical basis to quantify the probability of event sequences in the Level 2 PSA, focusing on the transition phase in an unprotected loss of flow accident in this study. In addition, dominant factors were also identified through parametric analyses using the SIMMER-III code. In the Japan sodium-cooled fast reactor, an inner duct is introduced into a fuel subassembly for enhancing molten fuel discharge from disrupted core in the transition phase. The parametric study showed that the analytical case without the fuel discharge through the inner duct resulted in an occurrence of recriticality regardless of the fuel discharge through control rod guide tubes. This suggests that the fuel discharge through the inner duct is essential to avoid severe recriticality in the transition phase. It could be concluded that the fuel discharges from the inner duct were the most dominant factor in the transition phase. On the other hand, the effects of fuel penetration into the blanket region, the falling-down of blanket fuels, etc. Were less significant. The experimental findings on the fuel discharge behavior and its driving force formation were summarized from the CABRI and EAGLE experiments. Using past experimental evidences, furthermore, the technical database was developed in order to quantify the probability of the Level 2 PSA.
Level-2 PSA unprotected loss of flow accident transition phase SIMMER-III event tree
Hidemasa Yamano Yoshiharu Tobita Ikken Sato
Japan Atomic Energy Agency 4002 Narita-cho, O-arai, Ibaraki, 311-1393, Japan
国际会议
上海
英文
607-619
2010-10-10（万方平台首次上网日期，不代表论文的发表时间）