An Experimental study on effective depressurization actions for PWR vessel bottom small break LOCA with HPI failure and gas inflow, ROSA-V test SB-PV-04

高圧注入系不作動とガス流入を想定したPWR容器底部小破断LOCAに及ぼす効果的減圧操作に関する実験研究,ROSA-V実験SB-PV-04

鈴木 光弘; 竹田 武司 ; 浅香 英明; 中村 秀夫  

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

大型非定常試験装置を使用したROSA-V計画において、加圧水型原子炉(PWR)の小破断冷却材喪失事故模擬実験を実施し、全高圧注入系不作動時に重要なアクシデントマネジメント(AM)策の炉心冷却効果を調べた。原子炉底部計装管10本破断を模擬した実験(SB-PV-04)で、蒸気発生器逃がし弁全開操作と補助給水作動によるAM策は、蓄圧注入系からの非凝縮性ガス流入による著しい減圧阻害にもかかわらず、低圧注入系が作動して炉心露出防止に効果的であることを示した。AM策として1次系冷却速度-55K/hの減圧操作を実施した前実験では炉心露出に至ったことに比較し、急減圧操作は1次系保有水量を多く保存する効果があり、炉心冷却上有用であることを明らかにした。

A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection system during an SBLOCA at a pressurized water reactor (PWR). The experiment (SB-PV-04) simulating a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes clarified that rapid depressurization action by full-opening of relief valves and supplying auxiliary feedwater were effective to avoid core uncovery through actuation of low pressure injection system irrespective of significantly degraded depressurization by non-condensable gas inflow from the accumulator tanks. It is clarified that the effective core cooling was established by the rapid primary cooling which contributed to preserve larger primary coolant mass than in the previous experiment (SB-PV-03) which was conducted with smaller primary cooling rate of -55 K/h as another AM action and resulted in core heatup.