References
- USNRC, 1996, Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events, 10 CFR 50 50.61, US Nuclear Regulatory Commission, May
- USNRC, 1987, Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors, Regulatory Guide 1.154, US Nuclear Regulatory Commission, January
- Jhung, M. J., Park, Y. W. and Jang, C., 1999, 'Pressurized Thermal Shock Analyses of a Reactor Pressure Vessel Using Critical Crack Depth Diagrams,' The International Journal of Pressure Vessels and Piping, Vol. 76, No. 12, pp. 813-823 https://doi.org/10.1016/S0308-0161(99)00063-0
- Joo, J. H., Kang, K. J. and Jhung, M. J., 2002, 'Fracture Mechanics Analysis of Reactor Pressure Vessel Under Pressurized Thermal Shock - The Effect of Elastic-Plastic Behavior and Stainless Steel Cladding.,' Transactions of the Korean Society of Mechanical Engineers, A, Vol. 26, No. 1, pp. 39-47 https://doi.org/10.3795/KSME-A.2002.26.1.039
- Jhung, M. J., Kim, S. H., Lee, J. H. and Park, Y. W., 2001, 'Round Robin Analysis of Pressurized Thermal Shock for Reactor Vessel,' 16th International Conference on Structural Mechanics in Reactor Technology, Washington, USA, August
- Tada, H., P. C. Paris P. C. and Irwin, G. R., 2000, The Stress Analysis of Cracks Handbook, ASME Press
- Wu, X. R. and Carlsson, A. J., 1991, Weight Functions and Stress Intensity Factor Solutions, Pergamon Press, New York
- USNRC, 1982, 'NRC Staff Evaluation of Pressurized Thermal Shock,' SECY 82-465, US Nuclear Regulatory Commission
- Veseley, W. E., Lynn, E. K. and Goldberg, F. F., 1978, 'The OCT A VIA Computer Code : PWR Reactor Pressure Vessel Failure Probabilities Due to Operational Caused Pressure Transients,' NUREG-0258, US Nuclear Regulatory Commission
- EPRI, 1993, 'White Paper on Reactor Vessel Integrity Requirements for Level A and B Conditions,' TR-100251, Electric Power Research Institute, January
- USNRC, 1986, 'VISA-II, A Computer Code for Predicting the Probability of Reactor Vessel Failure,' NUREG/CR-4486, Battelle Pacific Northwest Laboratories, April
- ASME, 1998, ASME Boiler and Pressure Vessel Code Sec. XI, 'Proposed Code Case for Application of Master Curve Method,' Code Case N-629, The American Society of Mechanical Engineers
- Yoon, K. K., 2000, 'A direct fracture toughness model for irradiated reactor vessel weld material based on reference temperature,' Nuclear Engineering and Design, Vol. 198, pp. 253-259 https://doi.org/10.1016/S0029-5493(99)00343-X
- Sokolov, M. A., 1998, 'Statistical analysis of the ASME Kk database,' Journal of Pressure Vessel Technology, Vol. 120, pp. 24-28 https://doi.org/10.1115/1.2841880
Cited by
- Round robin analysis for probabilistic structural integrity of reactor pressure vessel under pressurized thermal shock vol.19, pp.2, 2005, https://doi.org/10.1007/BF02916185