Abstract
After the Japanese Fukushima-Daiichi accident, the extreme event beyond the design basis accident is realized to be possible in the combined disasters. The current mitigation strategy of ECCS could fail because the low pressure injection system with electrical pumps will fail in a station blackout accident. To utilize the best of residual steam by the turbine driven pump is a possible alternate mitigation strategy and is analyzed in the paper. An advanced safety analysis tool with fast, accurate and integrated man-machine interface is necessary to analyze more different cases in the extreme accident and provide more safety precautions and more operation strategies for the plant owners. The TRACE code, the latest and advanced best-estimate simulation code, incorporates the four important codes, TRAC-P、TRAC-B、RELAP5 and RAMONA, and a graphic user interface, Symbolic Nuclear Analysis Package (SNAP), to provide a modern thermal-hydraulic analysis tool with fast and integrated inputs, and will become the NRC’s flagship thermal-hydraulic analysis tool in the near future. The TRACE model of Chinshan nuclear power plant with the same BWR/4 reactor of Fukushima-Daiichi NPP is developed, (1) based on the plant design data; (2) consists of different modules to simulate the reactor systems; and (3) analyzes the 3D thermal-hydraulic phenomena through the 3D VESSEL component and more practical thermal-hydraulic phenomena can be analyzed in the downcomer, fuel-assembly reactor core, core bypass, and upper and lower plenum. The Chinshan TRACE model, which has been benchmarked through several transient cases with the Chinshan FSAR report, the start-up data and the transient results of RETRAN data, can be adopted for analyzing both hypothetical transient scenarios and loss-of-coolant accidents, and further more for the alternate mitigation strategies of the extreme accidents of Fukushima-Daiichi type. In this paper, a double-ended guillotine (DEG) break on the recirculation loop is analysis. The Fukushima-Daiichi type accidents, the extended station blackout (SBO) accidents, are evaluated with several scenarios like no break, one SRV stuck open and the various break areas with the 1%, 10%, 100% cross areas of recirculation loop. The current RCIC injection flow rate is not sufficient in a very small break like 1% break area of a recirculation loop and even in the stuck open of a safety/ relieve valve (SRV). The reactor water level will sharply reduce when the reactor pressure is released and result in a fast increase of the fuel temperature. In this situation, the reactor pressure will increase once the coolant being injected that will reduce the effect of the external low pressure injecting system. Thus, the turbine driven pumps, the RCIC pump and the HPCI pump, are one of the important alternate mitigation strategies in the extended SBO. Through this paper, the more advanced analysis on the combined accidents could be performed for the improvement of nuclear safety.
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