1 |
H.W. Chou, C.C. Huang, Effects of fracture toughness curves of ASME Section XI-Appendix G on a reactor pressure vessel under pressure-temperature limit operation, Nucl. Eng. Design 280 (2014) 404-412.
DOI
|
2 |
M.J. Jhung, Reactor probabilistic integrity evaluation (R-PIE) code, KINS/RR-545, Korea Institute of Nuclear Safety, Daejeon (Korea), 2008.
|
3 |
U.S. Nuclear Regulatory Commission, Radiation embrittlement of reactor vessel materials, Regulatory Guide 1.99, Rev. 2, USNRC, Washington (DC), 1988.
|
4 |
F.A. Simonen, K.I. Johnson, A.M. Liebetrau, D.W. Engel, E.P. Simonen, VISA-II: a computer code for predicting the probability of reactor vessel failure, NUREG/CR-4486 (PNL-5775), Pacific Northwest Laboratory, Richland (WA), 1986.
|
5 |
H.W. Chou, C.C. Huang, B.Y. Chen, H.C. Lin, R.F. Liu, Failure probability assessment for a boiling water reactor pressure vessel under low temperature over-pressure event, PVP2012-78244, Proceedings of the ASME 2012 Pressure Vessels & Piping Conference, Toronto (Canada), 2012.
|
6 |
American Society of Mechanical Engineers, Fracture toughness criteria for protection against failure, ASME Boiler and Pressure Vessel Code, Section XI, Appendix G, ASME, New York (NY), 2011.
|
7 |
R.D. Cheverton, D.G. Ball, The role of crack arrest in the evaluation of PWR pressure vessel integrity during PTS transients, Oak Ridge National Laboratory, Oak Ridge (TN), 1984.
|
8 |
I.S. Raju, J.C. Newman Jr., Stress intensity factors for internal and external surface cracks in cylindrical vessels, J. Pressure Vessel Technol. 104 (1982) 293-298.
DOI
|
9 |
D.A. Curry, A model for predicting the influence of warm prestressing and strain aging on the cleavage fracture toughness of ferritic steels, Int. J. Fracture 22 (1983) 149-159.
|
10 |
U.S. Nuclear Regulatory Commission, Fracture toughness requirements for protection against pressurized thermal shock events, Title 10, Section 50.61, USNRC, Washington (DC), 1984.
|
11 |
U.S. Nuclear Regulatory Commission, Alternative fracture toughness requirements for protection against pressurized thermal shock events, Title 10, Section 50.61a, USNRC, Washington (DC), 2010.
|
12 |
Y. Kanto, M.J. Jhung, K. Ting, Y.B. He, K. Onizawa, S. Yoshimura, Summary of international PFM round robin analyses among Asian countries on reactor pressure vessel integrity during pressurized thermal shock, Int. J. Pressure Vessels Piping 90 (2012) 46-55.
|
13 |
M.J. Jhung, S.H. Kim, Y.H. Choi, Y.S. Chang, W.Y. Xu, J.M. Kim, J.W. Kim, C. Jang, Probabilistic fracture mechanics round robin analysis of reactor pressure vessels during pressurized thermal shock, J. Nucl. Sci. Technol. 47 (2010) 1131-1139.
DOI
|
14 |
G. Qian, M. Niffenegger, Procedures, methods and computer codes for the probabilistic assessment of reactor pressure vessels subjected to pressurized thermal shocks, Nucl. Eng. Design 258 (2013) 35-50.
DOI
|
15 |
G. Qian, V.F. Gonzalez-Albuixech, M. Niffenegger, Probabilistic assessment of a reactor pressure vessel subjected to pressurized thermal shocks by using crack distributions, Nucl. Eng. Design 270 (2014) 312-324.
DOI
|
16 |
G. Qian, M. Niffenegger, Deterministic and probabilistic analysis of a reactor pressure vessel subjected to pressurized thermal shocks, Nucl. Eng. Design 273 (2014) 381-395.
DOI
|
17 |
C.C. Huang, H.W. Chou, B.Y. Chen, R.F. Liu, H.C. Lin, Probabilistic fracture analysis for boiling water reactor pressure vessels subjected to low temperature over-pressure event, Ann. Nucl. Energy 43 (2012) 61-67.
DOI
|