Abstract
The Maanshan nuclear power plant operated by Taiwan Power Company is a Westinghouse PWR. The rated core thermal power is 2775 MW. The Loss of coolant accident (LOCA) is one of the design basis accidents (DBAs) in light water reactors. Although a LOCA is considered very unlikely to happen, the safety systems must be designed to secure adequate cooling of the reactor core during the accident to prevent the meltdown of the core. By the separate effects tests, several important phenomena of thermal-hydraulic, such as critical flow, CCFL, and ECC bypass, are discussed and verified. Then each separate models are integrated to the Maanshan TRACE model. The Maanshan TRACE model is also verified by referring to the start-up test data and FSAR data. Furthermore, IIST facility had been established for safety studies of the Westinghouse three- loops PWR, especially for Maanshan nuclear power plant. The IIST TRACE model also has been established and verified its accuracy by referring ot the test data of IIST facility LOCA experiments. The experiences of IIST LOCA analysis are used in the Maanshan LOCA analysis. After verifications, this study uses the Maanshan TRACE model to analysis the two types of LOCA, LBLOCA and SBLOCA. Several important thermal-hydraulic parameters, such as core water level, break flow rate, system pressure, and peak cladding temperature, are calculated and discussed in this study. LOCA analytical results indicate that the Maanshan TRACE model predicts not only the behaviors of important plant parameters in consistent trends with the Maashan FSAR data, but also provides a greater margin for the PCT evaluation.