Silicon carbide (SiC) and SiC matrix composites (SiCf/SiC) are being investigated as potential fuel cladding materials for advanced PWRs in order to improve the safety of nuclear power plants. The conceptual design of multi-layered SiC cladding (consisting of a monolithic SiC layer, SiCf/SiC composite layer and a monolithic SiC coating layer) has been investigated to meet the fuel requirements of both the strength and impermeability. A stress distribution model of the triple-layered SiC is developed on the basis of the theory of thermo-elasticity mechanics, taking radial temperature gradient and swelling effects into account as well. The heat transferring behavior of the cladding is investigated by analyzing the temperature distribution under steady conditions. Finite Element Analysis (FEA) code ANSYS is used to obtain the stress and temperature nephogram of multi-layered SiC fuel cladding under simulated steady conditions. Compared with the results of ANSYS, the stress distribution model and temperature distribution is validated.
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2017 25th International Conference on Nuclear Engineering
July 2–6, 2017
Shanghai, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5781-6
PROCEEDINGS PAPER
Stress Analysis of Three-Layer SiC Cladding for PWRs
Weiping Lan,
Weiping Lan
North China Electric Power University, Beijing, China
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Dachao Lin,
Dachao Lin
North China Electric Power University, Beijing, China
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Zulong Hao,
Zulong Hao
North China Electric Power University, Beijing, China
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Fenglei Niu
Fenglei Niu
North China Electric Power University, Beijing, China
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Weiping Lan
North China Electric Power University, Beijing, China
Dachao Lin
North China Electric Power University, Beijing, China
Zulong Hao
North China Electric Power University, Beijing, China
Fenglei Niu
North China Electric Power University, Beijing, China
Paper No:
ICONE25-66814, V003T02A032; 6 pages
Published Online:
October 17, 2017
Citation
Lan, W, Lin, D, Hao, Z, & Niu, F. "Stress Analysis of Three-Layer SiC Cladding for PWRs." Proceedings of the 2017 25th International Conference on Nuclear Engineering. Volume 3: Nuclear Fuel and Material, Reactor Physics and Transport Theory; Innovative Nuclear Power Plant Design and New Technology Application. Shanghai, China. July 2–6, 2017. V003T02A032. ASME. https://doi.org/10.1115/ICONE25-66814
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