Stress Corrosion Cracking of Alloy 800 in Secondary Side Crevice Environment

  • Maria-Lynn KomarEmail author
  • Guylaine Goszczynski
Conference paper
Part of the The Minerals, Metals & Materials Series book series (MMMS)


Alloy 800 nuclear grade (NG) is a material of choice for replacement steam generators (SG) due to its inherent resistance to primary water stress corrosion cracking (SCC). However, the long term performance of SGs depends on the performance of the material in upset conditions. Various degradation modes have been observed in Alloy 800NG under simulated secondary crevice environments (SCE) in C-ring and CERT experiments. Furthermore, the first incidences of SCE SCC have been observed in Alloy 800NG SG tubes in nuclear power plants and may be the sentinel events at the onset of more extensive cracking in the future. Understanding the parametric dependencies of SCC obtained under representative SCE and plausible transient conditions are keys to predicting future SG performance, validating mitigation strategies, and addressing life extension issues. The results of SCE crack growth rate (CGR) testing of Alloy 800NG in conditions representative of an acid-sulfate chemistry upset condition will be presented.


Alloy 800 Stress corrosion cracking Initiation Growth rate Acidic crevice environment 



The authors gratefully acknowledge Peter Andresen of GE R&D for his expert advice on CGR testing, and Sandy MacKay of Ontario Power Generation for performing the MULTEQ calculations of the standard and modified crevice chemistry pH.


  1. 1.
    B. Alexandreanu, O.K. Chopra, and W.J. Shack, Crack Growth Rates and Metallographic Examinations of Alloy 600 and Alloy 82/182 from Field Components and Laboratory Materials Tested in PWR Environments, NUREG/CR-6964 ANL-07/12 Argonne National Laboratory, May 2008Google Scholar
  2. 2.
    Status Review of Initiation of Environmentally Assisted Cracking and Short Crack Growth EPRI, Palo Alto, CA: December 2005. 1011788Google Scholar
  3. 3.
    Y.Z. Wang, R.W. Revie, R.N. Parkins, Mechanistic Aspects of Stress Corrosion Crack Initiation and Early Propagation, CORROSION 99, NACEGoogle Scholar
  4. 4.
    E. Richey, D. Morton, M. Schurman, SCC Initiation Testing of Nickel-Based Alloys Using In-Situ Monitored Uniaxial Tensile Specimens, LM-05K043, May 17, 2005Google Scholar
  5. 5.
    Materials Reliability Program Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 152 Welds (MRP-115), EPRI, Palo Alto, CA: 2004. 1006696Google Scholar
  6. 6.
    P. Andresen, Stress Corrosion Cracking: Mechanisms and Current Status Theme Meeting on Mechanisms, Occurrences, and Mitigation of Corrosion Degradation in NPP (Mumbai, India, 2011)Google Scholar
  7. 7.
    E. Serra, Stress Corrosion Cracking of Alloy 600, EPRI NP-2114-SR (Electric Power Research Institute, Palo Alto, CA, 1981)Google Scholar
  8. 8.
    R. Bandy, D. van Rooyen, Tests with Inconel 600 to Obtain Quantitative Stress-Corrosion Cracking Data for Evaluating Service Performance, BNL-NUREG-31814, (Upton, NY: Brookhaven National Laboratory, September 1982)Google Scholar
  9. 9.
    G. Was, K. Lian, The Role of Time-Dependent Deformation in Intergranular Crack Initiation of Alloy 600 Steam Generator Tubing Material NUREG/GR-0016, (Washington, DC: United States Nuclear Regulatory Commission, March 1998)Google Scholar
  10. 10.
    R.W. Staehle, Bases for Predicting the Earliest Penetrations Due to SCC for Alloy 600 on the Secondary Side of PWR Steam Generators, NUREG/CR-6737, (Washington, DC: United States Nuclear Regulatory Commission, September 2001)Google Scholar
  11. 11.
    Steam Generator Management Program: Alloy 800 Steam Generator Tubing Experience, (Palo Alto, CA: Electric Power Research Institute, 2012). 1024992Google Scholar
  12. 12.
    R. Kilian, R. Zimmer, R. Arenz, J. Beck, T. Schonherr, M. Widera, in Operating Experience with Alloy 800 SG Tubing in Europe. Proceedings 13th International Conference on Environmental Degradation in Nuclear Power Systems—Water Reactors, (Whistler, British Columbia, Canada: April 19–23, 2007)Google Scholar
  13. 13.
    D. Gómez-Briceño, M.S. García-Redondo, F. Hernández, in Update of Alloy 800 Behaviour in Secondary Side of PWR Steam Generators. Proceedings Fontevraud 7 Contribution of Materials Investigations to Improve the Safety and Performance of LWRs, (Avignon, Popes’ Palace, France: September 26–30, 2010)Google Scholar
  14. 14.
    M. Wright, in Establishing Threshold Conditions for Lead-Induced Cracking of Steam Generator Tube Alloys. Proceedings Water Chemistry of Nuclear Reactor Systems, BNES, 1996Google Scholar
  15. 15.
    Y. Lu, in Minimize Corrosion Degradation of Steam Generator Tube Materials—Updated ECP/pH Zone for Alloy 800 SG Tubing. 5th CNS International Steam Generator Conference, (Toronto, Ontario, Canada: November 26–29, 2006)Google Scholar
  16. 16.
    Y. Lu, in Define Optimum Conditions for Steam Generator tube Integrity and Extended Steam Generator Service Life. 15th International Conference on Nuclear Engineering, ICONE 15–10854 (Nagoya, Japan: April 22–26, 2007)Google Scholar
  17. 17.
    E. Pierson, J. Stubbe, P. Someville, in Resistance En Milieu Acide Des Materiaux Alternatifs Pour Tubes De GV (Alliages 690 Et 800), Fontevraud 3 International Symposium (Avignon, France: French Nuclear Energy Society, September 12–16, 1994) p. 556Google Scholar
  18. 18.
    D. Gómez-Briceño, M.S. García, F. Hernández, Effect of Secondary Cycle Sulphuric Acidic Injection on Steam Generator Tubes. Fontrevaud 3 International Symposium (Avignon, France: French Nuclear Energy Society, September 12–16, 1994) p. 565Google Scholar
  19. 19.
    J.M. Sarver, B.P. Migling, J.V. Monter, Constant Extension Rate (CERT) Testing of alloy 690 and 800 Nuclear Steam Generator Tubing. International Conference Fontevraud III, 1994Google Scholar
  20. 20.
    Tubetech NK38 TH 33111 02 Part 2 Technical Specification for DNGS Incoloy 800Google Scholar
  21. 21.
    M.L. Turi, G. Ogundele, G. Goszczynski, A.K. Jarvine, Stress Corrosion Cracking in Alloy 800 in Secondary Side Crevice Environment, in 16th International Symposium on Environmental Degradation of Materials in Nuclear Systems -Water Reactors, (The Minerals, Metal and Materials Society TMS, 2013), Doc Number: ED2013–3311Google Scholar
  22. 22.
    M. Mirzai, C. Marushka, S. Pagan, O. Lepik, G. Ogundele, M. Wright, G. Kharshafkjian, 1997 Stress Corrosion Cracking/Corrosion Fatigue/Fatigue in Alloy 600, in Proceedings Of 8th International Symposium on Environmental Degradation of Metals in Nuclear Power Systems—Water Reactors, vol 1. (La Grange Park, Illinois: American Nuclear Society Inc.) p. 11Google Scholar
  23. 23.
    M.D. Wright, M. Mirzai, in Lead-Induced SCC Propagation Rates in Alloy 600, ed. by F.P. Ford, S.M. Bruemmer, G.S. Was. 9th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors, (The Minerals, Metals and Materials Society, 1999) p. 657Google Scholar
  24. 24.
    Y. C. Lu, in Effect of Lead Contamination on Steam Generator Tube Degradation, ed by T.R. Allen, P.J. King, L. Nelson TMS. Proceedings of the 12th International Conference on Environmental Degradation of Materials in Nuclear Power System—Water Reactors (The Minerals, Metals & Materials Society, 2005)Google Scholar
  25. 25.
    S. MacKay, Ontario Power Generation, Private Communication, May 5, 2011Google Scholar
  26. 26.
    O.K. Chopra, W.K. Soppet, W.J. Shack, Effects of Alloy Chemistry, Cold Work, and Water Chemistry on Corrosion Fatigue and Stress Corrosion Cracking of Nickel Alloys and Welds, NUREG/CR-6721, ANL 01/07, Argonne National Lab, March 2001Google Scholar
  27. 27.
    Advanced Testing Techniques to Measure the PWSCC Resistance of Alloy 690 and its Weld Metals, EPRI, Palo Alto, CA, and U.S. Department of Energy, Washington, DC: 2004. 1011202Google Scholar
  28. 28.
    Materials Reliability Program, Mitigation of PWSCC in Nickel-Base Alloys by Optimizing Hydrogen in the Primary Water (MRP-213) (EPRI, Palo Alto, CA, 2007), p. 1015288Google Scholar
  29. 29.
    P.L. Andresen, F.P. Ford, K. Gott, R.L. Jones, P.M. Scott, T. Shoji, R.W. Staehle, R.L. Tapping, Expert Panel Report on Proactive Materials Degradation Assessment, BNL-NUREG-77111–2006, US NRC, February 2007Google Scholar
  30. 30.
    D.S. Morton, S.A. Attanasio, G.A. Young, in Primary Water SCC Understanding and Characterization Through Fundamental Testing in the Vicinity of the Nickel/Nickel Oxide Phase Transition, Proceedings 10th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors, NACE, 2001Google Scholar
  31. 31.
    D. Morton, G. Newsome, E. West, C. Ehlert, in SCC Growth Rate Testing of Cold Worked Stainless Steel in Hydrogen Deaerated Water, 17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors August 9–12, 2015, Ottawa, Ontario, CanadaGoogle Scholar
  32. 32.
    R.W. Staehle, J.A. Gorman, Quantitative assessment of submodes of stress corrosion cracking on the secondary side of steam generator tubing in pressurized water reactors: Part 1, critical review of corrosion science and engineering. Corrosion 59(11), 931 (2003)CrossRefGoogle Scholar

Copyright information

© The Minerals, Metals & Materials Society 2019

Authors and Affiliations

  1. 1.Kinectrics Inc.Toronto, OntarioCanada

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