References
- ASME, 1998, ASME Boiler and Pressure Vessel Code, Section XI Rules for Inspection of Nuclear Power Plant Components, Appendix A Analysis of Flaws, The American Society of Mechanical Engineers
- Bass, B. R., Pugh, C. E., Keeney, J., Schulz, H. and Sievers, J., 1996, 'CSNI Project for Fracture Analysis of Large-Scale International Reference Experiments (FALSIRE II),' NEA/CSNI/R(96)1, OECD/NEA, November
- Dickson, T. L. and Williams, P. T., 2003, Fracture Analysis of Vessels - Oak Ridge FAVOR, v02.4, Computer Code: Theory and Implementation of Algorithms, Methods and Correlations, draft NUREG, US Nuclear Regulatory Commission
- Faidy, C., 2003, PROSIR - Probabilistic Structural Integrity of a PWR Reactor Pressure Vessel, Proposed to OECD/NEA PWG3 - Metal Group, EDF - SEPTEN, July
- Jang, C., Kang, S. C., Moonn, H. R., Jeong, I. S. and Kim, T. R., 2003, 'The Effects of the Stainless Steel Cladding in Pressurized Thermal Shock Evaluation,' Nuclear Engineering and Design, Vol. 226, pp.127-140 https://doi.org/10.1016/S0029-5493(03)00190-0
- Jang, C., Jhung, M. J., Kang, S. C. and Choi, Y. H., 2004, 'The Effect of Analysis Variables on the Failure Probability of the Reactor Pressure Vessel by Pressurized Thermal Shock,' Transactions of the Korean Society of Mechanical Engineers (A), Vol. 28, No.6, pp.693-700 https://doi.org/10.3795/KSME-A.2004.28.6.693
- Jhung, M. J., Kim, S. H., Lee, J. H. and Park, Y. W., 2003, 'Round Robin Analysis of Pressurized Thermal Shock for Reactor Pressure Vessel,' Nuclear Engineering and Design, Vol. 226, pp. 141-154 https://doi.org/10.1016/S0029-5493(03)00191-2
- Jhung, M. J. and Park, Y. W., 1999, 'Deterministic Structural and Fracture Mechanics Analyses of Reactor Pressure Vessel for Pressurized Thermal Shock,' Structural Engineering and Mechanics, Vol. 8, No. 1, pp. 103-118 https://doi.org/10.12989/sem.1999.8.1.103
- Jung, S. G., Jin, T. E., Jhung, M. J. and Choi, Y. H., 2003, 'Probabilistic Fracture Mechanics Analysis of Reactor Vessel for Pressurized Thermal Shock - The Effect of Residual Stress and Fracture Toughness -,' Transactions of the Korean Society of Mechanical Engineers (A), Vol. 27, No.6, pp. 987-996 https://doi.org/10.3795/KSME-A.2003.27.6.987
- Sievers, J. and Schulz, H., 1999, 'Final Report on the International Comparative Assessment Study of Pressurized - Thermal - Shock in Reactor Pressure Vessels (RPV PTS ICAS),' NEA/CSNI/R(99)3, OECD/NEA, May
- USNRC, 1988, Radiation embrittlement of reactor vessel materials, Regulatory Guide 1.99, Revision 2, U.S. Nuclear Regulatory Commission
- USNRC, 1998, Technical Basis for an ASTM Standard on Determining the Reference Temperature, To, for Ferritic Steels in Transition Range, NUREG/CR-5504, US Nuclear Regulatory Commission