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Science and Technology of Nuclear Installations
Volume 2008, Article ID 745178, 10 pages
Research Article

RELAP5/MOD3.3 Code Validation with Plant Abnormal Event

Reactor Engineering Division, Jožef Stefan Institute, Jamova cesta 39, 1000 Ljubljana, Slovenia

Received 27 February 2008; Accepted 3 June 2008

Academic Editor: Martina Adorni

Copyright © 2008 Andrej Prošek and Borut Mavko. 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. A. Prošek, F. D'Auria, and B. Mavko, “Review of quantitative accuracy assessments with fast Fourier transform based method (FFTBM),” Nuclear Engineering and Design, vol. 217, no. 1-2, pp. 179–206, 2002. View at Publisher · View at Google Scholar
  2. A. Prošek, F. D'Auria, D. J. Richards, and B. Mavko, “Quantitative assessment of thermal-hydraulic codes used for heavy water reactor calculations,” Nuclear Engineering and Design, vol. 236, no. 3, pp. 295–308, 2006. View at Publisher · View at Google Scholar
  3. F. D'Auria, “The role of experimental database in the validation of thermalhydraulic system codes,” International Journal of Heat and Technology, vol. 21, no. 1, pp. 21–30, 2003. View at Google Scholar
  4. A. Petruzzi and F. D'Auria, “Thermal-hydraulic system codes in nulcear reactor safety and qualification procedures,” Science and Technology of Nuclear Installations, vol. 2008, Article ID 460795, 16 pages, 2008. View at Publisher · View at Google Scholar
  5. IAEA-TECDOC-1550, Proceedings of a Technical Meeting on Deterministic Analysis of Operational Events in Nuclear Power Plants, Dubrovnik, Croatia, May 2005.
  6. P. Groudev and A. Stefanova, “Validation of RELAP5/MOD3.2 model on trip off one main coolant pump for VVER 440/V230,” Nuclear Engineering and Design, vol. 236, no. 12, pp. 1275–1281, 2006. View at Publisher · View at Google Scholar
  7. A. Prošek, B. Kvizda, B. Mavko, and T. Kliment, “Quantitative assessment of MCP trip transient in a VVER,” Nuclear Engineering and Design, vol. 227, no. 1, pp. 85–96, 2004. View at Publisher · View at Google Scholar
  8. I. Parzer, “RELAP5/MOD3.3 simulation of MSIV closure events in a two-loop PWR,” Transactions of the American Nuclear Society, vol. 87, pp. 202–203, 2002. View at Google Scholar
  9. Information Systems Laboratories, Inc., Nuclear Safety Analysis Division, “RELAP5/MOD3.3 Code Manual, NUREG/CR-5535 Rev. P3,” Rockville, Maryland, Idaho Falls, Idaho, USA, March 2003.
  10. I. Parzar, B. Mavko, and B. Krajnc, “Simulation of a hypothetical loss-of-feedwater accident in a modernized nuclear power plant,” Journal of Mechanical Engineering, vol. 49, no. 9, pp. 430–444, 2003. View at Google Scholar
  11. A. Prošek, I. Parzer, and B. Krajnc, “Simulation of hypothetical small-break loss-of-coolant accident in modernized nuclear power plant,” Electrotechnical Review, vol. 71, no. 4, pp. 199–204, 2004. View at Google Scholar
  12. I. Parzer, “Break model comparison in different RELAP5 versions,” in Proceedings of the International Conference Nuclear Energy for New Europe, pp. 217.1–217.8, Nuclear Society of Slovenia, Portorož, Slovenia, September 2003.
  13. F. D'Auria, M. Leonardi, and R. Pochard, “Methodology for the evaluation of thermal-hydraulic codes accuracy,” in Proceedings of International Conference on New Trends in Nuclear System Thermalhydraulics, Pisa, Italy, May-June 1994.
  14. A. Prošek and M. Leskovar, “Improved FFTBM by signal mirroring as a tool for code assessment,” in Proceedings of International Congress on Advances in Nuclear Power Plants (ICAPP '07), pp. 7121-1–7121-9, Nice, France, May 2007.