Effect of Long-Term Thermal Aging on SCC Initiation Susceptibility in Low Carbon Austenitic Stainless Steels

  • So AokiEmail author
  • Keietsu Kondo
  • Yoshiyuki Kaji
  • Masahiro Yamamoto
Conference paper
Part of the The Minerals, Metals & Materials Series book series (MMMS)


The objective of this study was to clarify the effect of long-term thermal aging on SCC initiation susceptibility in low carbon austenitic stainless steels. Specimens used were Type 304L and 316L austenitic stainless steels. Both steels were cold worked to 20% thickness reduction (CW) followed by long-term thermal aging at 288 °C for 14,000 h (LTA). Creviced Bent Beam (CBB) testing was carried out to estimate the SCC initiation susceptibility under BWR simulated water condition at high temperature. The results of the CBB tests showed that Type 304L specimens with CW and LTA treatment exhibited no SCC susceptibility. In contrast, the SCC initiation susceptibility of Type 316L increased by the combination of cold work and long-term thermal aging. To understand these results, evaluations on the changes of microchemistry, microstructure and mechanical properties induced by the CW and LTA treatment have been performed, and their correlation with the SCC initiation susceptibility was discussed.


Low-carbon austenitic stainless steel Stress corrosion cracking Long-term thermal aging Cold work Creviced bent beam (CBB) test 


  1. 1.
    K.J. Leonard, T.M. Rosseel, and J.T. Busby, Light water reactor sustainability program materials aging and degradation pathway technical program plan. (Report ORNL/LTR-2012/327 Revision 5, Oak Ridge National Laboratory, 2016)Google Scholar
  2. 2.
    T. Tsukada et al., Report of examination of the samples from core shroud (2F3-H6a) at Fukushima Dai-ni nuclear power station unit-3. (Report JAERI-Tech 2004-044, Japan Atomic Energy Research Institute (currently, Japan Atomic Energy Agency), 2004)Google Scholar
  3. 3.
    Y. Okamura et al., Structural integrity evaluation for core shroud and PLR piping with SCC. J. High Press. Inst. Jpn. 43, 4–14 (2005)Google Scholar
  4. 4.
    T.M. Angeliu et al., in The IGSSCC behavior of L-grade stainless steel in 288 °C water. Proceeding of the 8th International Symposium on Environmental Degradation of Materials in Nuclear Power System Water Reactors, no. 8 (1997), pp. 649–662Google Scholar
  5. 5.
    J. Kuniya et al., Stress corrosion cracking susceptibility of various austenitic stainless steel pipe welds in high temperature oxygenated water. Boshoku-Gijutsu 31, 261–267 (1982)Google Scholar
  6. 6.
    R.M. Horn et al., Experience and assessment of stress corrosion cracking in L-grade stainless steel BWR internals. Nucl. Eng. Des. 174, 313–325 (1997)CrossRefGoogle Scholar
  7. 7.
    V. Kain et al., Effect of cold work on low-temperature sensitization behavior of austenitic stainless steels. J. Nucl. Mater. 334, 115–132 (2004)CrossRefGoogle Scholar
  8. 8.
    M. Tsubota, Y. Kanazawa, H. Inoue, in The Effect of Cold Work on SCC Susceptibility of Austenitic Stainless Steel. Proceeding of the 7th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors, no. 7 (1995), pp. 519–527Google Scholar
  9. 9.
    M. Akashi, T. Kawamoto, The effect of molybdenum addition on SCC susceptibility of stainless steel in oxygenated high temperature water. Boshoku-Gijutsu 27, 165–171 (1978)Google Scholar
  10. 10.
    M. Akashi et al., Metallurgical factors influencing the susceptibility of non-sensitized stainless steel to intergranular stress-corrosion cracking in high-temperature, high-purity water environments. Paper presented at corrosion 99, San Antonio, Texas, 25 April 1999, 451Google Scholar
  11. 11.
    H. Sahlaoui et al., Effect of ageing condition on the precipitates evolution, chromium depletion and intergranular corrosion susceptibility of AISI 316L: experimental and modeling results. Mater. Sci. Eng., A 372, 98–108 (2004)CrossRefGoogle Scholar
  12. 12.
    G. Mereic de bellefon, J.C. van Duysen, Tailoring plasticity of austenitic stainless steels for nuclear applications: review of mechanisms controlling plasticity of austenitic steels below 400 °C. J. Nucl. Mater. 475, 168–191 (2016)Google Scholar
  13. 13.
    M. Murayama et al., The combined effect of molybdenum and nitrogen on the fatigued microstructure of 316 type austenitic stainless steel. Scripta Mater. 41, 467–473 (1999)CrossRefGoogle Scholar

Copyright information

© The Minerals, Metals & Materials Society 2019

Authors and Affiliations

  • So Aoki
    • 1
    Email author
  • Keietsu Kondo
    • 1
  • Yoshiyuki Kaji
    • 1
  • Masahiro Yamamoto
    • 1
  1. 1.Japan Atomic Energy Agency, Nuclear Science and Engineering CenterNaka-GunJapan

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