The original internal flooding probabilistic safety analysis （PSA） study of Krsko Nuclear Power Plant （two-loop Pressurized Water Reactor （PWR） plant of Westinghouse design） was performed in mid nineties and lim...The original internal flooding probabilistic safety analysis （PSA） study of Krsko Nuclear Power Plant （two-loop Pressurized Water Reactor （PWR） plant of Westinghouse design） was performed in mid nineties and limited to reactor core damage risk （Level 1 PSA）. In 2003, it was, together with other safety and hazard analyses, subject to the Periodic Safety Review （PSR）. In the PSR, it was stated that methodological PSA approaches and guidelines have evoluted during the past decade and several observations were provided, concerning the area screening process, residual risk and treatment of plant damage states and risk from radioactivity releases （i.e., Level 2 PSA）. In order to address the PSR observations, upgrade ofKrsko NPP internal flooding PSA was undertaken. The area screening process was revisited in order to cover the areas without automatic reactor trip equipment. The model was extended to Level 2. Residual risk was estimated at both Level 1 and Level 2, in terms of core damage frequency （CDF） and large early release frequency （LERF）, respectively.展开更多
文摘The original internal flooding probabilistic safety analysis （PSA） study of Krsko Nuclear Power Plant （two-loop Pressurized Water Reactor （PWR） plant of Westinghouse design） was performed in mid nineties and limited to reactor core damage risk （Level 1 PSA）. In 2003, it was, together with other safety and hazard analyses, subject to the Periodic Safety Review （PSR）. In the PSR, it was stated that methodological PSA approaches and guidelines have evoluted during the past decade and several observations were provided, concerning the area screening process, residual risk and treatment of plant damage states and risk from radioactivity releases （i.e., Level 2 PSA）. In order to address the PSR observations, upgrade ofKrsko NPP internal flooding PSA was undertaken. The area screening process was revisited in order to cover the areas without automatic reactor trip equipment. The model was extended to Level 2. Residual risk was estimated at both Level 1 and Level 2, in terms of core damage frequency （CDF） and large early release frequency （LERF）, respectively.
