References
- S. KW. YU and K. R. HEDGES, “The Next Generation CANDU Design,” Proc. 13th Pacific Basin Nuclear Conf., Shenzhen, Guangdong, China, October 22–24, 2002, Chinese Nuclear Society (2002).
- “Quality Assurance of Analytic, Scientific, and Design Programs for Nuclear Power Plants,” CSA N286.7-9, Canadian Standards Association (1999).
- N. K. POPOV, D. J. WREN, V. G. SNELL, A. WHITE, and P. BOCZAR, “Development and Qualification of AECL Computer Codes for ACR Safety Analysis Applications,” Proc. 24th Annual CNS Conf., Toronto, Ontario, Canada, June 8–11, 2003, Canadian Nuclear Society (2003).
- D. J. WREN, N. K. POPOV, V. LANGMAN, and V. G. SNELL, “Computer Code Applicability Assessment for the Advanced CANDU Reactor,” Proc. 12th Int. Conf. Nuclear Engineering (ICONE 12), Arlington, Virginia, April 25–29, 2004, American Society of Mechanical Engineers (2004).
- A. ABDUL-RAZZAK, A.C.D. WRIGHT, N. K. POPOV, and D. J. WREN, “Classification and Ranking of Fundamental Phenomena for Validation of ACR Computer Programs,” Proc. 24th Annual CNS Conf., Toronto, Ontario, Canada, June 8–11, 2003, Canadian Nuclear Society (2003).
- A. C. D. WRIGHT, A. ABDUL-RAZZAK, N. K. POPOV, D. WREN, H. CHAO, and S. GIRGIS, “Validation Matrices for the Computer Codes Used for ACR Safety Analysis,” Proc. 24th Annual CNS Conf., Toronto, Ontario, Canada, June 8–11, 2003, Canadian Nuclear Society (2003).
- E. O. MOECK, J. C. LUXAT, L. A. SIMPSON, M. A. PETRILLI, and P. D. THOMPSON, “Validation of Computer Codes Used in Safety Analyses of CANDU Power Plants,” Proc. IAEA Technical Committee Mtg. Advances in Heavy Water Reactor Technology, Bombay, India, January 29-February 1, 1996, IAEA-TECDOC-984, International Atomic Energy Agency (1997).
- B. E. BOYACK, I. CATTON, R. B. DUFFEY, P. GRIFFITH, K. R. KATSMA, G. S. LELLOUCHE, S. LEVY, U. S. ROHATGI, R. A. SHAW, G. E. WILSON, W. WULFF, and N. ZUBER, “Quantifying Reactor Safety Margins, Parts 1-6,” Nucl. Eng Des., 119, 1 (1990).
- “Draft Regulatory Guide DG-1120: Transient and Accident Analysis Methods,” United States Nuclear Regulatory Commission (2002).
- N. K. POPOV, H. SILLS, V. G. SNELL, B. BOYACK, and V. LANGMAN, “Large-Break Loss-of-Coolant Accident Phenomena Identification and Ranking Table (PIRT) for the Advanced CANDU Reactor,” Proc. 12th Int. Conf. Nuclear Engineering (ICONE 12), Arlington, Virginia, April 25–29, 2004, American Society of Mechanical Engineers (2004).
- N. K. POPOV, A. ABDUL-RAZZAK, H. SILLS, V. G. SNELL, and V. LANGMAN, “Single Channel Flow Blockage Accident Phenomena Identification and Ranking Table (PIRT) for the Advanced CANDU Reactor,” Proc. Int. Congress Advances in Nuclear Power Plants, Pittsburgh, Pennsylvania, June 13–17, 2004, American Nuclear Society (2004).