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Effect of Specimen Size on the Crack Growth Rate Behavior of Irradiated Type 304 Stainless Steel

  • A. JenssenEmail author
  • P. Chou
  • C. Tobpasi
Conference paper
Part of the The Minerals, Metals & Materials Series book series (MMMS)

Abstract

Crack growth rate (CGR) testing in BWR normal water chemistry was performed on compact tension (CT) specimens of two different sizes (B = 8 mm and B = 19 mm), machined from a Type 304 SS core shroud at a dose of ~1 dpa. The objectives were to study the effect of specimen size on the CGR, and to determine the K validity limit for a CT specimen dimension used in previous studies. The results show that for materials with significant strain hardening capacity remaining, there is no effect of specimen size on the CGR when testing is conducted at stress intensity factors valid according to ASTM E399 using the flow strength. For materials at higher dose in which the strain hardening capacity is lost or greatly reduced, a different K validity criterion might be applicable.

Keywords

Stress corrosion cracking Crack growth rate Specimen size K validity Irradiated stainless steel 

Notes

Acknowledgements

The work presented in this paper was funded by EPRI. Material from the Barsebäck 2 core shroud was procured with the assistance of Pål Efsing and Björn Forssgren, Ringhals AB. Testing and post-test examinations were performed by many colleagues at Studsvik. All these contributions are gratefully acknowledged.

References

  1. 1.
    Standard Test Method for Linear-Elastic Plane-Strain Fracture Toughness KIc of Metallic Materials, E399, Annual Book of ASTM Standards, Volume 03.01 (ASTM International, West Conshohocken, PA, 2006)Google Scholar
  2. 2.
    P. L. Andresen, K/Size Effects on SCC in Irradiated, Cold Worked and Unirradiated Stainless Steel, Paper presented at the 11th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors. Stevenson, USA (ANS, August 2003), pp. 870–886Google Scholar
  3. 3.
    G.S. Was, Recent Developments in Understanding Irradiation Assisted Stress Corrosion Cracking, (Paper presented at the 11th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors. Stevenson, USA. (ANS, August 2003), pp. 965–985Google Scholar
  4. 4.
    A. Jenssen, J. Stjärnsäter, R. Pathania. Crack Growth Rates of Irradiated Commercial Stainless Steels in BWR and PWR Environments, (Paper presented at the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors, Colorado Springs, CO, August 2011)Google Scholar
  5. 5.
    A. Jenssen, P. Efsing, K. Gott and P-O Andersson, Crack Growth behavior of Irradiated Type 304L Stainless Steel in Simulated BWR Environment, Paper presented at the 11th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors, Stevenson, WA. (ANS, August 2003), p. 1015)Google Scholar
  6. 6.
    A. Jenssen, J. Stjärnsäter, R. Pathania, Crack Growth Rate Testing of Fast Reactor Irradiated Type 304L and 316 SS in BWR and PWR Environments, Paper presented at the 14th International Conference on Environmental Degradation of Materials in Nuclear Power Systems, (Virginia Beach, VA, August 2009)Google Scholar
  7. 7.
    A. Jenssen, J. Stjärnsäter, R. Pathania, R. Carter, Crack Growth Rate Testing of Irradiated Type 304L and Type 316L in BWR Environments, Paper presented at the International Boiling Water Reactor and Pressurized Water Reactor Materials Reliability Conference and Exhibit Show, National Harbor, MD, (July 2012)Google Scholar
  8. 8.
    A. Jenssen, R. Pathania, R. Carter, Crack Growth in Irradiated Austenitic Stainless Steels in BWR Environments, Paper presented at Fontevraud 8 - Contribution of Materials Investigations and Operating Experience to LWRs’ Safety, Performance and Reliability, Avignon, France, (September 2014)Google Scholar
  9. 9.
    Effect of Specimen Size on the Crack Growth Rate Behavior of Irradiated Type 304 Stainless Steel: Phase 1—Testing of Type 304 at 1 dpa. EPRI, Palo Alto, CA: 2012. 1023962Google Scholar
  10. 10.
    L.W. Niedrach, W.H. Stoddard, Monitoring pH and Corrosion Potentials in High Temperature Aqueous Environments. Corrosion 41, 45–51 (1985)CrossRefGoogle Scholar
  11. 11.
    BWRVIP-265: BWR Vessels and Internals Project, Crack Growth in High Fluence BWR Materials-Phase 2. EPRI, Palo Alto, CA: 2012. 1026508Google Scholar
  12. 12.
    E. D. Eason, R. Pathania, Irradiation- Assisted Stress Corrosion Crack Growth Rates of Austenitic Stainless Steels in Light Water Reactor Environments, Paper presented at the 17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors, Ottawa, Ontario, Canada, (August 2015)Google Scholar
  13. 13.
    A. Jenssen, C. Jansson and J. Sundberg, The Effect of Hold Time on the Crack Growth Rate of Sensitized Stainless Steel in High Temperature Water, (Paper presented at the 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems–Water Reactors, Snowbird, UT, August 2005)Google Scholar
  14. 14.
    A. Jenssen, P. Efsing and J. Sundberg, Influence of Heat Treatment, Aging and Neutron Irradiation on the Fracture Toughness and Crack Growth Rate in BWR Environments of Alloy X-750, Paper presented at the 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors, Snowbird, UT, (August 2005)Google Scholar
  15. 15.
    P. L. Andresen, Similarity of Cold Work and radiation Hardening in Enhancing Yield Strength and SCC Growth of Stainless Steel in Hot Water, Paper presented at Corrosion 2002, NACE 2002, Paper #02509)Google Scholar
  16. 16.
    T. Sato and T. Shoji, Effects of Specimen Size and Thickness on CGR in High Temperature Waters, Paper presented at the 11th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors. Stevenson, USA. ANS, (August 2003) p. 862Google Scholar
  17. 17.
    M. Itow, M. Kikuchi and S. Suzuki, SCC Crack Growth Rates of Type 304 Stainless Steel at High K Region in Simulated BWR Environment, Paper presented at Corrosion 2000, Orlando, FL, NACE, 2000, Paper #00221Google Scholar
  18. 18.
    Morton, D. S., et al., In search of the true temperature and stress intensity factor dependencies for PWSCC. Paper presented at the 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors, Snowbird, UT, (August 2005), pp. 977–988Google Scholar
  19. 19.
    Richey, E., Morton, D. S., Moshier, W. C., Influence of Specimen Size on the SCC Growth Rate of Ni-Alloys Exposed to High Temperature Water. Paper presented at Corrosion 2006. Houston, TX, NACE, 2006. Paper #06513Google Scholar
  20. 20.
    E. Richey, D. Morton and W. Moshier, Influence of Specimen Size on the SCC Growth Rate of Ni-Alloys Exposed to High Temperature Water, Report LM-05K151, Lookheed Martin, (October 2005)Google Scholar

Copyright information

© The Minerals, Metals & Materials Society 2019

Authors and Affiliations

  1. 1.Studsvik Nuclear ABNyköpingSweden
  2. 2.Electric Power Research InstitutePalo AltoUSA

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