Estimation of temperature-induced reactor coolant system and steam generator tube creep rupture probability under high-pressure severe accident conditions

References

• Bayless , P.D. 1995 . Severe Accident Natural Circulation Studies at the INEL , Idaho National Engineering Laboratory . NUREG/CR-6285 (INEL-94/0016)
   Google Scholar
• Probabilistic Safety Assessment for Ulchin Units 3&4 [Containment Performance Analysis], Korea Hydro & nuclear power Co., Ltd, 411 Yeongdongdae-ro Seoul, Korea, 2004 [in Korean]
   Google Scholar
• Breeding , R.J. 1990 . Evaluation of Severe Accident Risks: Surry Unit 1 , Albuquerque, NM 87185 : Sandia National Laboratories . NUREG/CR-4551
   Google Scholar
• NRC . 1990 . Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants , Washington, DC 20555 : Nuclear Regulatory Commission . NUREG-1150, Final Summary Report
   Google Scholar
• Gauntt , R.O. 2005 . MELCOR Computer Code Manuals, vol.1: Primer and User's Guide, Version 1.8.6 , Albuquerque, NM 87185-0739 : SAND 2005-5713, Sandia National Laboratories . NUREG/CR-6119
   Google Scholar
• McKay , M.D. , Conover , W.J. and Beckman , R.J. 1979 . A comparison of three methods for selecting values of input variables in the analysis of output from a computer code . Technometrics , 221 : 239 – 245 .
   Web of Science ®Google Scholar
• Lee , B. 2009 [in Korean] . Development of Probability Evaluation Methodology for High Pressure/Temperature Gas Induced RCS Boundary Failure and SG Creep Rupture , Daedeok-daero 989-111 Daejeon, Korea : Energy Research Institute . KAERI/TR-3756/2009, Korea Atomic
   Google Scholar
• NRC, SGTR Severe Accident Working Group . 1998 . Risk Assessment of Severe Accident-Induced Steam Generator Tube Rupture , Washington, DC 20555-0001 : Nuclear Regulatory Commission . NUREG-1570, U.S
   Google Scholar
• Gauntt , R.O. Uncertainty Analyses using the MELCOR Severe Accident Analysis Code . Workshop on Evaluation of Uncertainties in Relation to Severe Accidents and Level 2 Probabilistic Safety Analysis .
   Google Scholar
• Liao , Y. and Vierow , K. 2005 . MELCOR analysis of steam generator tube creep rupture in station blackout severe accident . Nucl. Technol , 152 : 302 – 313 .
   Web of Science ®Google Scholar
• Stewart , W.A. , Pieczynski , A.T. and Srinivas , V. 1990 . Experiments on Natural Circulation in a Pressurized Water Reactor Model for Degraded Core Accidents Palo Alto , CA EPRI Project No.RP2177-5 Final Report, EPRI
   Google Scholar
• Stewart , W.A. , Pieczynski , A.T. and Srinivas , V. 1990 . Experiments on Natural Circulation in a Pressurized Water Reactor Model with High Pressure SF6 Palo, CA , , Alto Draft EPRI Report, EPRI
   Google Scholar
• Boyd , C.F. 2003 . CFD Analysis of 1/7th Scale Steam Generator Inlet Plenum Mixing During a PWR Severe Accident , Washington, DC 20555-0001 : Commission . NUREG-1781, U.S. Nuclear Regulatory
   Google Scholar
• Boyd , C.F. 2004 . CFD Analysis of Full-Scale Steam Generator Inlet Plenum Mixing During a PWR Severe Accident , Washington, DC 20555-0001 : Commission . NUREG-1788, U.S. Nuclear Regulatory
   Google Scholar
• Martinez , G.M. 1994 . Independent Review of SCDAP/RELAP5 Natural Circulation Calculations , Albuquerque, New Mexico 87185 and Llvermore, California 94550 : National Laboratories . SAND91-2089, Sandia
   Google Scholar
• IAEA . 2005 . Natural Circulation in Water Cooled Nuclear Power Plants: Phenomena, Models, and Methodology for System Reliability Assessments , Wagramer Strasse 5, P.O. Box 100 A-1400 Vienna, Austria : International Atomic Energy Agency . IAEA-TECDOC-1474
   Google Scholar
• Park , J. and Lee , J. 1997 [in Korean] . MELCOR Natural Circulation Model for SGTR Sequence During the Severe Accident in a 3 Loop Westinghouse PWR , 62 Gwahak-ro Daejeon, Korea : Safety . KINS/AR-482, Korea Institute of Nuclear
   Google Scholar
• Nutt , W.T. and Wallis , G.B. 2004 . Evaluation of nuclear safety from the outputs of computer codes in the presence of uncertainties . Reliab. Eng. Syst. Safe , 93 : 57 – 77 .
   Google Scholar