Lifetime estimation of a BWR core shroud in terms of IGSCC

Stress Corrosion Cracking in Light Water Reactors: Good Practices and Lessons Learned. — Vienna: International Atomic Energy Agency, 2011. BWRVIP Vessel and Internals Project Update, Drew Odell, Exelon, BWRVIP Integration Committee Technical Chairman EPRI-NRC Technical Exchange Meetings, May 22-25, 2017. Status Report: Intergranular Stress Corrosion Cracking of BWR Core Shrouds and Other Internal Components, NUREG-1544, 1996. J.K. McKinley, W. H. Bamford, A. M. Ruminski, T. L. Meikle, J. L. McFadden, “Comparison of Boiling Water Reactor and Pressurized Water Reactor Experience with Cracking of Austenitic Stainless Steel”, Enclosure 3 to MRP 2018-011, Westinghouse Electric Company, 2018. A. Roth, “Review of Intergranular Cracking in Austenitic Stainless Steel Components of BWR RPV-Internals,” Fontevraud 8, Avignon, France, September 15-18, 2014. Nuclear Regulatory Commission Document, NUREG-1544, “Status Report: Intergranular Stress Corrosion Cracking of BWR Core Shrouds and Other Internal Components,” March 1996. Nuclear Regulatory Commission Generic Letter 94-03, “Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors,” July 25, 1994. W. Lunceford, N. Palm, and D. Sommerville, “BWR Core Shroud Off-Axis Cracking Inspection Experience,” International Light Water Reactor Materials Reliability Conference and Exhibition, EPRI, Chicago, Illinois, August 1-4, 2016. BWR Vessel and Internals Project, BWR Core Shroud Repair Design Criteria, Rev. 2 (BWRVIP-02), EPRI, Palo Alto, CA: 1999. 112642. BWR Vessel and Internals Project: BWR Core Shroud Inspection and Flaw Evaluation Guidelines (BWRVIP-76, Revision 1-A), EPRI, Palo Alto, CA: 2015. 3002005566. Nuclear Regulatory Commission Final Safety Evaluation, “Final Safety Evaluations of the Boiling Water Reactor Vessel and Internals Project 76, Rev. 1 Topical Report, 'Boiling Water Reactor Core Shroud Inspection and Flaw Evaluation Guidelines, '” November 12, 2014 (ML14266A227). BWRVIP-311: BWR Vessel and Internals Project: Core Shroud Weld Off-Axis Crack Growth Assessment. EPRI, Palo Alto, CA: 2017. 3002010682. Ford, 1980, Corrosion, 36, 597, 10.5006/0010-9312-36.11.597 Ford, 1982 F.P. Ford, P.L. Andresen, Proc. 3rd Int. Symp. Environ. Degrad. Mat. Nucl. Power Syst.--Water Reactors, p. 2789. Metallurgical Soc., Warrendale, PA (1988). F.P. Ford, P.L. Andresen, The Theoretical Prediction of the Effect of System Variables on the Cracking of Stainless Steels and Its Use in Design, Paper No. 83, Corrosion/87, NACE, Houston, TX (1987). Ford, 1987 F. P. Ford, D. F. Taylor, P. L. Andresen And R. G. Ballinger, Corrosion-Assisted Cracking of Stainless and Low-Alloy Steels in LWR Environments, NP-5064M, Final Report, Electric Power Research Institute, Palo Alto, CA (1987). Andresen, 1981, Corrosion, 38, 531, 10.5006/1.3593856 A, Molanderand, B. Rosberg,Proc. 3rd Int. Symp. Environ. Degrad. Mat. Nucl. Power Systems--Water Reactors, P. 333. The Metallurgical Soc., Warrendale, PA (1988). Ruther, 1984, Corrosion, 40, 518, 10.5006/1.3593887 Ljungberg, 1985, Corrosion, 41, 290, 10.5006/1.3582006 Ljungberg, 1988, Corrosion, 44, 66, 10.5006/1.3583909 L. G. Ljungberg, C. Jannson and D. Cubicciotti, Proc. 3rdlnt. Syrup. Environ. Degrad. Mat. Nucl. Power Syst.--Water Reactors, P. 315. The Metallurgical Soc., Warrendale, PA (1988). Ljungberg, 1984 Park, 1986, Trans. ASME, 108, 20 Eghan, 1985, Corrosion, 41, 474, 10.5006/1.3583830 Macdonald, D.D. et al., “Estimation of Corrosion Potentials in the Heat Transport Circuit of LWRs,” Proc. Int. Conf. Chemistry in Water Reactors: Operating Experience & New Developments, Nice, France, April 24-27, 1994, French Nuclear Energy Society. C. P. Ruiz, et al., Modeling Hydrogen Water Chemistry for BWR Applications, EPRI NP-6386, Electric Power Research Institute, June 1989. Macdonald, 1990, Corrosion, 46, 380, 10.5006/1.3585122 Macdonald, 1988, Corrosion, 44, 186, 10.5006/1.3583923 Macdonald, 1992, J. Electrochem. Soc, 139,12,3434 Macdonald, 1992, Corrosion, 48, 194, 10.5006/1.3315925 D. D. Macdonald, A. C. Scott, and P. R. Wentrcek. J. Electrochem. Soc., 126, 908-911 (1979); 128, 250 (1981). Macdonald, 1991, Corr. Sci., 32, 51, 10.1016/0010-938X(91)90063-U I. Balachov, D. D. Macdonald, B. Stellwag, N. Henzel, and R. Kilian, “Prediction of Materials Damage History from Stress Corrosion Cracking in Boiling Water Reactors”. Proc. ASMEIJSME Joint Press. Vess. Piping Conf., San Diego, CA, 376, 101-109 (July 26-30, 1998). Fekete, 2018, An Advanced Coupled Environment Fracture Model for Hydrogen-induced Cracking in Alloy 600 in PWR Primary Heat Transport Environment, Theoretical and Applied Fracture Mechanics, 95, 233, 10.1016/j.tafmec.2018.02.013 B. Fekete, I. Balachov, D. D. Macdonald, “Deterministic Modeling of Stress Corrosion Cracking in Light Water Nuclear Reactors.” Proc. 21st Int. Conf. on Water Chem. in Nucl. React. Syst., San Francisco, CA (2018). I. Balachov, B. Fekete, D. D. Macdonald, “Parameter Sensitivity Analysis on Deterministic Models of Stress Corrosion Cracking in Light Water Nuclear Reactors.” Proc. 21st Int. Conf. on Water Chem. in Nucl. React. Syst., San Francisco, CA (2018). ABAQUS User Manual Version 6.9. Pennsylvania, PA: Karlsson & Sorensen, Incorporated; 2009. Spencer, 2015 Spencer, 2019, Modular system for probabilistic fracture mechanics analysis of embrittled reactor pressure vessels in the grizzly code, Nucl. Eng. Des., 341, 25, 10.1016/j.nucengdes.2018.10.015 Li, 2015, Remaining Life Prediction of the Core Shroud Due to Stress Corrosion Cracking Failure in BWRs Using Numerical Simulations, J. Nucl. Sci. Technol., 52, 96, 10.1080/00223131.2014.930703 Nuclear and Industrial Safety Agency, 2004 Shi, 2014, Prediction of Crack Growth Rate in Type 304 Stainless Steel Using Artificial Neural Networks and the Coupled Environment Fracture Model, Corr. Sci., 89, 69, 10.1016/j.corsci.2014.08.011