참고문헌
- Materials Reliability Program: Review of Stress Corrosion Cracking of Alloys 182 and 82 in PWR Primary Water Service (MRP-220). EPRI, Palo Alto, CA: 2007. 1015427.
- K. Dozaki et al, "Effects of Dissolved Hydrogen Content in PWR Primary Water on PWSCC Initiation Property," E-Journal of Advanced Maintenance, vol.2, pp. 65-76 (2010).
- E. S. Hunt and D. J. Gross, "PWSCC of Alloy 600 Materials in PWR Primary System Penetrations," EPRI TR-103696, EPRI (1994).
- D. J. Seman, G. L. Webb and R. J. Parrington, "Primary Water Stress Corrosion Cracking of Alloy 600 - Effects of Processing Parameters," Proceedings: 1991 EPRI Workshop on PWSCC of Alloy 600 in PWRs, EPRI TR-100852, EPRI (1992).
-
Materials Reliability Program: Effects of Hydrogen, pH, Lithium and Boron on Primary Water Stress Corrosion Crack Initiation in Alloy 600 for Temperatures in the Range 320 - 330
${^{\circ}C}$ (MRP-147). EPRI, Palo Alto, CA: 2005. 1012145. - Materials Reliability Program (MRP) Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Materials (MRP-55) Revision 1, EPRI, Palo Alto, CA: 2002. 1006695.
- Materials Reliability Program Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115), EPRI, Palo Alto, CA: 2004. 1006696.
- S.A. Attanasio and D.S. Morton, "Measurement of the Ni/ NiO Transition in Ni-Cr-Fe Alloys and Updated Data and Correlations to Quantify the Effect of Aqueous Hydrogen and Primary Water SCC," Proc. 11th Int. Conf. on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Stevenson, USA, August 11-14, 2003.
- S.E. Cumblidge, S.R. Doctor, P.G. Heasler, and T.T. Taylor, "Results of the Program for the Inspection of Nickel Alloy Components," NUREG/CR-7019, U.S. Nuclear Regulatory Committee (2010).
- Materials Reliability Program: Welding Residual and Operating Stresses in PWR Alloy 182 Butt Welds (MRP-106), EPRI, Palo Alto, CA: 2004. 1009378.
- F. A. Simonen et al, "Probabilistic Fracture Mechanics Evaluation of Selected Passive Components - Technical Letter Report," PNNL-16625, Pacific Northwest Laboratory (2007).
- D. Rudland et al, "Development of Computational Framework and Architecture for Extremely Low Probability of Rupture(xLPR) Code," Proceedings of the ASME Pressure Vessels & Piping Division / K-PVP Conference, Bellevue, USA, July 18-20, 2010.
- D. Datta and C. Jang, "Development of an advanced PFM code for integrity evaluation of nuclear piping system", submitted to International Journal of Pressure Vessels and Piping (2010).
- D. Datta and C. Jang, "Integrity Evaluation of Nuclear Piping System under Combined Aging Mechanisms - Development of PINTIN-CAM Code," submitted to Journal of Pressure Vessel Technology (2011).
- D. Datta, "Development of an advanced PFM code for the integrity evaluation of nuclear piping system under combined aging mechanisms," Ph.D. Thesis, Department of Nuclear and Quantum Engineering (KAIST), 2010.
- J. D. Hong and C. Jang, "Probabilistic Fracture Mechanics Application for Alloy 600 components in PWRs," Proc. of 2010 International Congress on Advanced in Nuclear Power Plants (ICAPP'10), San Diego, USA , June 13-17, 2010.
- J. D. Hong and C. Jang, "Probabilistic Fracture Mechanics Application for Alloy 82/182 Welds in PWRs," Proc. of the ASME 2010 Pressure Vessels & Piping Division / KPVP Conference, Bellevue, USA, July 18-20, 2010.
- M. A. Khaleel and F. A. Simonen, "Evaluations of Structural Failure Probabilities and Candidate Inservice Inspection Programs," NUREG/CR-6986, U.S. Nuclear Regulatory Committee (2009).
- L. J. Bond, T. T. Taylor, S. R. Doctor, A. B. Hull and S. N. Malik, "Expert Panel Report on Proactive Materials Degradation Assessment," NUREG/CR-6923, U.S. Nuclear Regulatory Committee (2007).
- B. Grimmel, "U. S. Plant Experience with Alloy 600 Cracking and Boric Acid Corrosion of Light-Water Reactor Pressure Vessel Materials," NUREG-1823, U.S. Nuclear Regulatory Committee (2005).
- A. Zahoor, "Ductile Fracture Handbook", EPRI NP-6301-D, EPRI (1989).
- S. R. Gosselin, F. A. Simonen, P. G. Heasler, S. R. Doctor, D. A. Jackson and W. E. Norris, "Fatigue Crack Flaw Tolerance in Nuclear Power Plant Piping," NUREG/CR-6934, U.S. Nuclear Regulatory Committee (2007).
- W. J. Mills and C. M. Brown, "Fracture Toughness of Alloy 600 and EN82H Weld in Air and Water," Bettis Atomic Power Laboratory Report No.B-T-3264, Bettis Atomic Power Laboratory (2001).
- D. O. Harris, D. D. Dedhia and S. C. Lu, "Theoretical and User's Manual for pc-PRAISE, A Probabilistic Fracture Mechanics Computer Code for Piping Reliability Analysis," NUREG/CR-5864, U.S. Nuclear Regulatory Committee (1992).
- R. Tregoning, L. Abramson and P. Scott, "Estimating Lossof- Coolant Accident (LOCA) Frequencies Through the Elicitation Process," NUREG-1829, U.S. Nuclear Regulatory Committee (2005).
- S. E. Cumblidge et al, "Nondestructive and Destructive Examination Studies on Removed from-Service Control Rod Drive Mechanism Penetrations," PNNL-16628, Pacific Northwest Laboratory (2007).
피인용 문헌
- Effects of Cracking Test Conditions on Estimation Uncertainty for Weibull Parameters Considering Time-Dependent Censoring Interval vol.10, pp.1, 2016, https://doi.org/10.3390/ma10010003
- Uncertainty Evaluation of Weibull Estimators through Monte Carlo Simulation: Applications for Crack Initiation Testing vol.9, pp.7, 2016, https://doi.org/10.3390/ma9070521