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In CSR1000 (Chinese Supercritical Water Cooled Reactor 1000), based on the existing SCAC-CSR1000 code, by incorporating break flow model and passive safety system, CSR1000-DP02 code for LOCA analysis is developed. 50% hot leg break loss of coolant accident (LOCA) is analyzed. The results show that, when LOCA happens, the reactor scrams immediately. The main feed water mass flow, and coolant mass flow...
CHF (Critical Heat Flux) is an important nuclear thermal–hydraulic parameter, and it is closely related to the reactor safety. Based on Rough Set Decision Model, this article proposed a new method to seek for the main factors which may affect CHF at low pressure, low flow conditions in narrow channels. Nine condition attributes are selected to build the CHF fault diagnosis model in narrow rectangular...
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