A computer code, named COMPASS, is being developed employing the Moving Particle Semi-implicit (MPS) method for various complex phenomena of core disruptive accidents (CDAs) in the sodium-cooled fast reactors (SFRs). The COMPASS is designed to analyze multi-physics problems involving thermal hydraulics, structure and phase change, in a unified framework of MPS method. In FYs2006 and 2007 (Japanese Fiscal Year, hereafter), the development of basic functions of COMPASS was completed and fundamental verification calculations were carried out. In FY2007, the integrated verification program using available experimental data for key phenomena in CDAs was also started. In this paper, we show the basic verification calculations for the phase change model of COMPASS and the results of experimental analyses, together with the outline of the formulation of MPS method and the conceptual design of the COMPASS code.
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17th International Conference on Nuclear Engineering
July 12–16, 2009
Brussels, Belgium
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-4355-0
PROCEEDINGS PAPER
Next Generation Safety Analysis Methods for SFRs—(3) Thermal Hydraulics Models of COMPASS Code and Experimental Analyses
Yuichi Yamamoto,
Yuichi Yamamoto
Japan Systems Corporation, Kawasaki, Kanagawa, Japan
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Etsujo Hirano,
Etsujo Hirano
Japan Systems Corporation, Kawasaki, Kanagawa, Japan
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Masaya Oue,
Masaya Oue
Japan Systems Corporation, Kawasaki, Kanagawa, Japan
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Sensuke Shimizu,
Sensuke Shimizu
Japan Systems Corporation, Kawasaki, Kanagawa, Japan
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Noriyuki Shirakawa,
Noriyuki Shirakawa
The Institute of Applied Energy, Japan
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Seiichi Koshizuka,
Seiichi Koshizuka
The University of Tokyo, Japan
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Hidemasa Yamano,
Hidemasa Yamano
Japan Atomic Energy Agency, Japan
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Yoshiharu Tobita
Yoshiharu Tobita
Japan Atomic Energy Agency, Japan
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Yuichi Yamamoto
Japan Systems Corporation, Kawasaki, Kanagawa, Japan
Etsujo Hirano
Japan Systems Corporation, Kawasaki, Kanagawa, Japan
Masaya Oue
Japan Systems Corporation, Kawasaki, Kanagawa, Japan
Sensuke Shimizu
Japan Systems Corporation, Kawasaki, Kanagawa, Japan
Noriyuki Shirakawa
The Institute of Applied Energy, Japan
Seiichi Koshizuka
The University of Tokyo, Japan
Koji Morita
Kyushu University, Japan
Hidemasa Yamano
Japan Atomic Energy Agency, Japan
Yoshiharu Tobita
Japan Atomic Energy Agency, Japan
Paper No:
ICONE17-75521, pp. 409-418; 10 pages
Published Online:
February 25, 2010
Citation
Yamamoto, Y, Hirano, E, Oue, M, Shimizu, S, Shirakawa, N, Koshizuka, S, Morita, K, Yamano, H, & Tobita, Y. "Next Generation Safety Analysis Methods for SFRs—(3) Thermal Hydraulics Models of COMPASS Code and Experimental Analyses." Proceedings of the 17th International Conference on Nuclear Engineering. Volume 5: Fuel Cycle and High and Low Level Waste Management and Decommissioning; Computational Fluid Dynamics (CFD), Neutronics Methods and Coupled Codes; Instrumentation and Control. Brussels, Belgium. July 12–16, 2009. pp. 409-418. ASME. https://doi.org/10.1115/ICONE17-75521
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