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Science and Technology of Nuclear Installations
Volume 2012, Article ID 957285, 15 pages
http://dx.doi.org/10.1155/2012/957285
Research Article

RELAP5 Analysis of OECD/NEA ROSA Project Experiment Simulating a PWR Loss-of-Feedwater Transient with High-Power Natural Circulation

1Nuclear Safety Research Center, Japan Atomic Energy Agency (JAEA), Tokai-mura, Ibaraki-ken 319-1195, Japan
2Nuclear Energy System Safety Division, Japan Nuclear Energy Safety Organization (JNES), Minato-ku, Tokyo 105-0001, Japan

Received 6 August 2011; Accepted 12 November 2011

Academic Editor: Klaus Umminger

Copyright © 2012 Takeshi Takeda et al. This is an open access article distributed under the Creative Commons Attribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.

Linked References

  1. C. L. Nalezny, “Summary of nuclear regulatory commission's LOFT program experiment,” Tech. Rep. NUREG/CR-3214, EGG-2248, 1983. View at Google Scholar
  2. W. Ambrosini, F. D'Auria, and G. M. Galassi, “Lesson learned from the application to LOBI tests of CATHARE and RELAP5 codes,” in Proceedings of the 1st Meeting of the Nuclear Society of Slovenia, Bovec, Slovenia, 1992.
  3. United States Nuclear Regulatory Commission, “Reactor safety study—an assessment of risks in U. S. commercial nuclear power plants,” Tech. Rep. WASH-1400 (NUREG-075/14), 1975. View at Google Scholar
  4. The ROSA-V Group, “ROSA-V large scale test facility (LSTF) system description for the third and fourth simulated fuel assemblies,” JAERI-Tech 2003-037, Japan Atomic Energy Research Institute, Ibaraki, Japan, 2003. View at Google Scholar
  5. Y. Kukita, Y. Anoda, and K. Tasaka, “Summary of ROSA-IV LSTF first-phase test program—integral simulation of PWR small-break LOCAs and transients,” Nuclear Engineering and Design, vol. 131, no. 1, pp. 101–111, 1991. View at Google Scholar · View at Scopus
  6. K. E. Carlson, “RELAP5/MOD3 code manual (draft),” Tech. Rep. NUREG/CR-5535, EGG-2596, 1990. View at Google Scholar
  7. T. Takeda, H. Asaka, and H. Nakamura, “Analysis of the OECD/NEA ROSA project experiment simulating a PWR small break LOCA with high-power natural circulation,” Annals of Nuclear Energy, vol. 36, no. 3, pp. 386–392, 2009. View at Publisher · View at Google Scholar · View at Scopus
  8. H. Nakamura, T. Watanabe, T. Takeda et al., “RELAP5/MOD3 code verification through PWR pressure vessel small break loca tests in OECD/NEA rosa project,” in Proceedings of the 16th International Conference on Nuclear Engineering (ICONE-16 '08), pp. 659–668, Orlando, Fla, USA, May 2008. View at Publisher · View at Google Scholar · View at Scopus
  9. N. Zuber, “Problems in modeling small break LOCA,” USNRC Report NUREG-0724, 1980. View at Google Scholar
  10. H. Kumamaru and K. Tasaka, “Recalculation of simulated post-scram core power decay curve for use in ROSA-IV/LSTF experiments on PWR small-break LOCAs and transients,” Tech. Rep. JAERI-M 90-142, Japan Atomic Energy Research Institute, Ibaraki, Japan, 1990. View at Google Scholar
  11. ROSA Project Team, “Final integration report of OECD/NEA ROSA project,” JAEA-Research 2010-9002, Japan Atomic Energy Agency, Ibaraki, Japan, 2010. View at Google Scholar
  12. Y. Kukita, H. Nakamura, K. Tasaka, and C. Chauliac, “Nonuniform steam generator U-tube flow distribution during natural circulation tests in ROSA-IV large scale test facility,” Nuclear Science and Engineering, vol. 99, no. 4, pp. 289–298, 1988. View at Google Scholar · View at Scopus
  13. H. Asaka, Y. Kukita, T. Yonomoto, Y. Koizumi, and K. Tasaka, “Results of 0.5% cold-leg small-break LOCA experiments at ROSA-IV/LSTF: effect of break orientation,” Experimental Thermal and Fluid Science, vol. 3, no. 6, pp. 588–596, 1990. View at Google Scholar · View at Scopus
  14. K. H. Ardron and R. A. Furness, “A study of the critical flow models used in reactor blowdown analysis,” Nuclear Engineering and Design, vol. 39, no. 2-3, pp. 257–266, 1976. View at Google Scholar · View at Scopus
  15. D. W. Sallet, “Thermal hydraulics of valves for nuclear applications,” Nuclear Science and Engineering, vol. 88, no. 3, pp. 220–244, 1984. View at Google Scholar · View at Scopus
  16. Susyadi and T. Yonomoto, “Analysis on non uniform flow in steam generator during steady state natural circulation cooling,” JAERI-Research 2005-011, Japan Atomic Energy Research Institute, Ibaraki, Japan, 2005. View at Google Scholar
  17. G. B. Wallis, One-Dimensional Two-Phase Flow, McGraw-Hill Book, New York, NY, USA, 1969.
  18. T. Yonomoto, Y. Anoda, Y. Kukita, and Y. Peng, “CCFL characteristics of PWR steam generator U-tubes,” in Proceedings of the ANS International Topical Meeting on Safety of Thermal Reactors, American Nuclear Society, Portland, Ore, USA, 1991.