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Liu L.,North China Electrical Power University | Liu L.,Beijing Passive Technology Key Laboratory | Zhou T.,North China Electrical Power University | Zhou T.,Beijing Passive Technology Key Laboratory | And 7 more authors.
Progress in Nuclear Energy | Year: 2015

CSR1000 was selected as the research object. A code named SCAC-CSR1000 has been developed based on the SCAC code. The reliability of the code was verified by comparing the result of SCAC-CSR1000 and SCTRAN. Then the safety analysis was carried out. Five events were selected, that are partial loss of reactor coolant flow, isolation of main steam line, uncontrolled CR withdrawal, reactor coolant pump seizure and loss of feed water heating. By the numerical analyses, it was found that the MCST does not exceed 1260 °C, meets the design safety requirements. The 2nd MCST are higher than 1st MCST. The isolation of main steam line has the least safety margin. © 2015 Elsevier Ltd. All rights reserved. Source


Liu L.,North China Electrical Power University | Liu L.,Beijing Passive Technology Key Laboratory | Zhou T.,North China Electrical Power University | Zhou T.,Beijing Passive Technology Key Laboratory | And 7 more authors.
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2016

CSR1000 was selected as the research object. A code named SCAC-CSR1000 has been developed based on the SCAC code. The reliability of the code was verified by comparing the results of SCAC-CSR1000 and SCTRAN. Then the safety analysis was carried out. Five events were selected, that are partial loss of reactor coolant flow, isolation of main steam line, uncontrolled CR withdrawal, reactor coolant pump seizure and loss of feed water heating. By the numerical analysis, it was found that the MCST does not exceed 1260℃ and meets the design safety requirements. The 2nd MCST are higher than the 1st MCST. The isolation of main steam line has the less safety margin. © 2016, Yuan Zi Neng Chuban She. All right reserved. Source


Zhou T.,North China Electrical Power University | Zhou T.,Beijing Passive Technology Key Laboratory | Yang X.,North China Electrical Power University | Yang X.,Beijing Passive Technology Key Laboratory | And 7 more authors.
Annals of Nuclear Energy | Year: 2015

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 channels, and the rules of CHF occurring are obtained. These attributes include pipe diameter, circulating mode and flow stability, etc. The results show that flow instability could cause the CHF reach a minimum and CHF could be improved with the increase of pipe size. Diagnostic results in this method could predict the main influencing factors for CHF very well, and provide a new theoretical approach for the safety analysis of nuclear plants. © 2014 Elsevier B.V. All rights reserved. Source

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