This paper describes the preliminary safety analysis of a thermal-spectrum SCWR concept (CSR1000), which was proposed by Nuclear Power Institute of China (NPIC). The passive safety system and the design of the two-pass core concept characterize the safety performance of CSR1000. With code SCTRAN (a one-dimensional safety analysis code for SCWRs), loss of coolant flow accidents (LOFA) and loss of coolant accident (LOCA) as well as some other typical transients and accidents were analysed. The maximum cladding surface temperature (MCST) was regarded as an important criterion. The sensitivity analyses of some crucial parameters are helpful for the safety evaluation. Thus some parameters about the safety system and the actuation conditions, such as the delay time of the ADS actuation, the break area in LOCA analysis, were also involved in this paper. The analyses have shown that the proposed passive safety system is capable to mitigate the consequence of the selected abnormalities. The results will be a useful reference for the future development of CSR1000.
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2013 21st International Conference on Nuclear Engineering
July 29–August 2, 2013
Chengdu, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5579-9
PROCEEDINGS PAPER
Preliminary Safety Analysis of CSR1000
Pan Wu,
Pan Wu
Xi’an Jiaotong University, Xi’an, Shaanxi, China
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Junli Gou,
Junli Gou
Xi’an Jiaotong University, Xi’an, Shaanxi, China
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Jianqiang Shan,
Jianqiang Shan
Xi’an Jiaotong University, Xi’an, Shaanxi, China
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Bo Zhang,
Bo Zhang
Xi’an Jiaotong University, Xi’an, Shaanxi, China
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Xiang Li
Xiang Li
Nuclear Power Institute of China, Chengdu, Sichuan, China
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Pan Wu
Xi’an Jiaotong University, Xi’an, Shaanxi, China
Junli Gou
Xi’an Jiaotong University, Xi’an, Shaanxi, China
Jianqiang Shan
Xi’an Jiaotong University, Xi’an, Shaanxi, China
Bo Zhang
Xi’an Jiaotong University, Xi’an, Shaanxi, China
Xiang Li
Nuclear Power Institute of China, Chengdu, Sichuan, China
Paper No:
ICONE21-16502, V002T05A055; 10 pages
Published Online:
February 7, 2014
Citation
Wu, P, Gou, J, Shan, J, Zhang, B, & Li, X. "Preliminary Safety Analysis of CSR1000." Proceedings of the 2013 21st International Conference on Nuclear Engineering. Volume 2: Plant Systems, Construction, Structures and Components; Next Generation Reactors and Advanced Reactors. Chengdu, China. July 29–August 2, 2013. V002T05A055. ASME. https://doi.org/10.1115/ICONE21-16502
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