Analysis of PWR RPV lower head SBLOCA scenarios with the failure of high-pressure injection system using MAAP5

摘要

Through the use of MAAP5 code we carried out an analysis of the severe accident scenarios initiated by an unusual small-break loss-of-coolant accident (SBLOCA) located at the reactor pressure vessel (RPV) lower head with the failure of high-pressure injection system (HPIS) in a PWR power plant. The cases of 0.4-inch-break LOCA with and without manual depressurization of the reactor coolant system (RCS), and 1.0-inch-break LOCA were studied. The results showed that without manually depressurizing the RCS, reactor core underwent slow heatup and completely melted and eventually failed the RPV lower head at the primary system pressure of 6.87 MPa for 0.4-inch-break LOCA and 2.71 MPa for 1.0-inch-break LOCA. There was a strong interaction between the discharged melt and collected pool water in the reactor cavity at the time of RPV lower head failure by creep. A 0.472 MPa containment peak pressure was caused by hydrogen combustion for 0.4-inch-break LOCA. For 1.0-inch-break LOCA the pressure peaked twice, the first one (0.396 MPa) by intense steam generation and the second one (0.331 MPa) by hydrogen combustion. These pressure spikes were lower than the containment design pressure of 0.52 MPa. It was not likely to stop the cavity from being eroded by molten corium through continuously pouring water down into the cavity. Thus there were constant reactor cavity ablation and hydrogen generation, leading to dangerous erosion depths and elevated hydrogen volume fraction in the presence of containment spray system. On the other hand, timely manually opening the pressurizer (PZR) safety valves (SVs) was an effective mitigation measure to recover the core coolabilty by cold-leg accumulator injection system (AIS) and low-pressure injection system (LPIS). Besides, reasonably manually opening the steam generators (SGs) SVs while keeping the auxiliary feedwater on also helped to depressurize the RCS and prevent the severe accident. Both of the two mitigation measures successfully prevented the core from complete melt, but the latter one is preferable to the former one provided no steam generator tube rupture takes place.