As sodium cooled fast reactor (SFR) uses liquid sodium as coolant, the risk of sodium fire is brought to reactor safety, which is different from conventional fire. Sodium is very chemically active, and violent chemical reactions can happen when sodium is exposed to air or water. Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is probably one of the main contributors to the total reactor risk. In this paper, the methodology of fast reactor sodium fire risk assessment is studied, and the principles and procedure of sodium fire probabilistic safety assessment (PSA) are given. The application of this technology in China Experimental Fast Reactor (CEFR) is explored, and several key problems which need more research in the future during the process of sodium fire probabilistic safety assessment are discussed.
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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
Discussion on the Application of Fire Probability Safety Assessment to Sodium Cooled Fast Reactor
Wei Song,
Wei Song
Nuclear and Radiation Safety Center, Beijing, China
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Hongyi Yang,
Hongyi Yang
China Institute of Atomic Energy, Beijing, China
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Chunming Zhang,
Chunming Zhang
Nuclear and Radiation Safety Center, Beijing, China
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Jiaxu Zuo
Jiaxu Zuo
Nuclear and Radiation Safety Center, Beijing, China
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Wei Song
Nuclear and Radiation Safety Center, Beijing, China
Hongyi Yang
China Institute of Atomic Energy, Beijing, China
Chunming Zhang
Nuclear and Radiation Safety Center, Beijing, China
Jiaxu Zuo
Nuclear and Radiation Safety Center, Beijing, China
Paper No:
ICONE21-15780, V002T05A020; 8 pages
Published Online:
February 7, 2014
Citation
Song, W, Yang, H, Zhang, C, & Zuo, J. "Discussion on the Application of Fire Probability Safety Assessment to Sodium Cooled Fast Reactor." 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. V002T05A020. ASME. https://doi.org/10.1115/ICONE21-15780
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