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Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors pp 457–483Cite as

Grain Boundary Damage Evolution and SCC Initiation of Cold-Worked Alloy 690 in Simulated PWR Primary Water

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Part of the book series: The Minerals, Metals & Materials Series ((MMMS))

Abstract

Long-term grain boundary (GB) damage evolution and stress corrosion crack initiation in alloy 690 are being investigated by constant load tensile testing in high-temperature, simulated PWR primary water. Six commercial alloy 690 heats are being tested in various cold work conditions loaded at their yield stress. This paper reviews the basic test approach and detailed characterizations performed on selected specimens after an exposure time of ~1 year. Intergranular crack nucleation was observed under constant stress in certain highly cold-worked (CW) alloy 690 heats and was found to be associated with the formation of GB cavities. Somewhat surprisingly, the heats most susceptible to cavity formation and crack nucleation were thermally treated materials with most uniform coverage of small GB carbides. Microstructure, % cold work and applied stress comparisons are made among the alloy 690 heats to better understand the factors influencing GB cavity formation and crack initiation.

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References

  1. D.J. Paraventi, W.C. Moshier, Alloy 690 SCC growth rate testing, in Workshop on Cold Work in Iron- and Nickel-Base Alloys (EPRI, 2007)

    Google Scholar 

  2. P.L. Andresen, M.M. Morra, J. Hickling, A. Ahluwalia, J. Wilson, Effect of deformation and orientation on SCC of alloy 690, in 14th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors (American Nuclear Society, 2009), p. 846

    Google Scholar 

  3. D.R. Tice, S.L. Medway, N. Platts, J.W. Startmand, Crack growth testing on cold worked alloy 690 in primary water environment, in 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors (The Minerals, Metals & Materials Society, 2011), p. 71

    Google Scholar 

  4. S.M. Bruemmer, M.J. Olszta, N.R. Overman, M.B. Toloczko, Cold work effects on stress corrosion crack growth in alloy 690 tubing and plate materials, in 17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors (Canadian Nuclear Society, 2015)

    Google Scholar 

  5. M.B. Toloczko, S.M. Bruemmer, Crack growth response of alloy 690 in simulated PWR primary water, in 14th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors (American Nuclear Society, 2009), p. 706

    Google Scholar 

  6. M.B. Toloczko, S.M. Bruemmer, Cold rolling effects on stress corrosion crack growth in alloy 690 tubing and plate materials, in 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors (The Minerals, Metals & Materials Society, 2011), p. 91

    Google Scholar 

  7. R.H. Jones, S. Breummer, Environment-induced crack growth processes in nickel-base alloys, in 1st International Conference on Environment-Induced Cracking of Metals (1988), p. 287

    Google Scholar 

  8. G.S. Was, Grain-boundary chemistry and intergranular fracture in austenitic nickel-base alloys—A review. Corrosion (Houston) 46, 319–330 (1990)

    Article  CAS  Google Scholar 

  9. P. Andresen, M.M. Morra, A. Ahluwalia, Effect of deformation temperature, orientation and carbides on SCC of alloy 690, in 16th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors (NACE International, 2013)

    Google Scholar 

  10. K. Arioka, T. Yamada, T. Terachi, G. Chiba, Influence of carbide precipitation and rolling direction on intergranular stress corrosion cracking of austenitic stainless steels in hydrogenated high-temperature water. Corrosion (Houston) 62, 568–575 (2006)

    Article  CAS  Google Scholar 

  11. S.M. Bruemmer, M.J. Olszta, N.R. Overman, M.B. Toloczko, Microstructural effects on stress corrosion cracking of cold-worked alloy 690 tubing and plate materials, in 16th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors (NACE International, 2013)

    Google Scholar 

  12. K. Arioka, R.W. Staehle, T. Yamada, T. Miyamoto, T. Terachi, Degradation of alloy 690 after relatively short times. Corrosion (Houston) 72, 1252–1268 (2016)

    Article  CAS  Google Scholar 

  13. Z. Zhai, M.B. Toloczko, K. Kruska, S. Bruemmer, Precursor evolution and SCC initiation of cold-worked alloy 690 in simulated PWR primary water. Corrosion (Houston), (2017) (under review)

    Google Scholar 

  14. K. Arioka, Whitney award lecture: Change in bonding strength at grain boundaries before long term SCC initiation. Corrosion (Houston) 71(2015), 403–419 (2014)

    Google Scholar 

  15. M.B. Toloczko, N.R. Overman, M.J. Olszta, S.M. Bruemmer, Pacific Northwest National Laboratory investigation of stress corrosion cracking in nickel-base alloys, in Stress Corrosion Cracking of Cold-Worked Alloy 690, NUREG/CR-7103 vol. 3 (Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, 2015)

    Google Scholar 

  16. Z. Zhai, M.B. Toloczko, K. Kruska, D.K. Schreiber, M.J. Olszta, N.R. Overman, S. Bruemmer, Precursor damage evolution and stress corrosion crack initiation of cold-worked alloy 690 in PWR primary water. Pacific Northwest National Laboratory: Technical Milestone Report M2LW-16OR0402034, Light Water Reactor Sustainability Program, DOE Office of Nuclear Energy, Sept 2016

    Google Scholar 

  17. S.M. Bruemmer, M.J. Olszta, D.K. Schreiber, M.B. Toloczko, Corrosion and stress corrosion crack initiation of cold worked alloy 600 and alloy 690 in PWR primary water environments. Pacific Northwest National Laboratory: Technical Milestone Report M2LW-13OR0402035, Light Water Reactor Sustainability Program, DOE Office of Nuclear Energy, Sept 2014

    Google Scholar 

  18. Z. Zhai, M.J. Olszta, M.B. Toloczko, S.M. Bruemmer, Precursor corrosion damage and stress corrosion crack initiation in alloy 600 during exposure to PWR primary water, in 17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors (Canadian Nuclear Society, 2015)

    Google Scholar 

  19. K. Kruska, Z. Zhai, M.B. Toloczko, S. Bruemmer, Characterization of SCC initiation precursors in cold-worked alloy 690, in CORROSION 2017, NACE (2017)

    Google Scholar 

  20. K. Arioka, T. Yamada, T. Miyamoto, T. Terachi, Dependence of stress corrosion cracking of alloy 690 on temperature, cold work, and carbide precipitation—role of diffusion of vacancies at crack tips. Corrosion (Houston) 67, 035006-035001–035006-035018 (2011)

    Article  Google Scholar 

  21. H.G. Van Bueren, Theory of the formation of lattice defects during plastic strain. Acta Metall. 3, 519–524 (1955)

    Article  Google Scholar 

  22. J. Cadek, Creep in Metallic Materials (Elsevier, 1988)

    Google Scholar 

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Acknowledgements

The authors gratefully acknowledge the financial support from the Office of Nuclear Energy, U.S. Department of Energy through the Light Water Reactor Sustainability Program. In addition, support is recognized from the U.S. Nuclear Regulatory Commission for SCC crack growth rate testing and pre-test microstructural characterizations of the CW materials and from the Office of Basic Energy Sciences, U.S. Department of Energy for high-resolution grain boundary examinations. Key experimental support was provided by Dr. John Deibler at PNNL for conducting the finite element modeling. Key technical assistance from Robert Seffens, Clyde Chamberlin, Anthony Guzman and Ryan Bouffioux is acknowledged for SCC initiation testing and materials preparation activities.

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Authors and Affiliations

  1. Pacific Northwest National Laboratory, 622 Horn Rapids Rd, 999, Richland, WA, 99352, USA

    Ziqing Zhai, Mychailo Toloczko, Karen Kruska, Daniel Schreiber & Stephen Bruemmer

Authors
  1. Ziqing Zhai
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  2. Mychailo Toloczko
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  3. Karen Kruska
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  4. Daniel Schreiber
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  5. Stephen Bruemmer
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Corresponding author

Correspondence to Ziqing Zhai .

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Editors and Affiliations

  1. Idaho National Laboratory, Idaho Falls, ID, USA

    John H. Jackson

  2. Naval Nuclear Laboratory, Pittsburgh, PA, USA

    Denise Paraventi

  3. Chalk River Laboratories, Canadian Nuclear Laboratories, Chalk River, ON, Canada

    Michael Wright

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© 2019 The Minerals, Metals & Materials Society

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Zhai, Z., Toloczko, M., Kruska, K., Schreiber, D., Bruemmer, S. (2019). Grain Boundary Damage Evolution and SCC Initiation of Cold-Worked Alloy 690 in Simulated PWR Primary Water. In: Jackson, J., Paraventi, D., Wright, M. (eds) Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors. The Minerals, Metals & Materials Series. Springer, Cham. https://doi.org/10.1007/978-3-030-04639-2_29

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