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PWSCC Initiation of Alloy 600: Effect of Long-Term Thermal Aging and Triaxial Stress

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Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors

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

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Abstract

Thermally aged nickel based Alloy 600 was investigated to evaluate the effects of long-term thermal aging and triaxial stress on primary water stress corrosion crack initiation behavior. Long-term thermal aging was simulated by heat treatment at 400 °C, a temperature that does not cause excessive formation of second phases that cannot form in nuclear power plant service conditions. Triaxial stress was applied by a round notch in the gauge length of some test specimen; other specimens were smooth. Slow strain rate tests (SSRT) monitored by the direct current potential drop method were conducted to evaluate stress corrosion crack initiation susceptibility of the thermally aged specimens in the primary water environment. For smooth specimens (which experience uniaxial stress), the susceptibility of those thermally aged for the equivalent of 10-years was the highest, while the susceptibility of the as-received specimens was the lowest. However, for the notched specimens (which experience triaxial stress), the specimens thermally aged for the equivalent of 20-years showed the highest susceptibility, while the as-received specimens showed the lowest.

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Acknowledgements

This work was financially supported by the Nuclear Safety Research Program through the Korea Foundation of Nuclear Safety (KOFONS), granted financial resource from the Nuclear Safety and Security Commission (NSSC), Republic of Korea (No. 1403006) and by the Human Resources Development of the Korea Institute of Energy Technology Evaluation and Planning (KETEP) grant funded by the Korea Government Ministry of Trade Industry and Energy (MOTIE) (No. 20174030201430).

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Authors and Affiliations

  1. Department of Nuclear Engineering, School of Mechanical, Aerospace and Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST), 50 UNIST-Gil, Eonyang-Eup, Ulju-Gun, Ulsan, 44919, Republic of Korea

    Seung Chang Yoo, Kyoung Joon Choi, Seunghyun Kim & Ji Hyun Kim

  2. Department of Mechanical Engineering, Korea University, 50 UNIST-Gil, Eonyang-Eup, Ulju-Gun, Ulsan, 689-798, Republic of Korea

    Ji-Soo Kim, Byoung Ho Choi & Yun-Jae Kim

  3. Department of Nuclear Engineering, Sejong University, 50 UNIST-Gil, Eonyang-Eup, Ulju-Gun, Ulsan, 689-798, Republic of Korea

    Jong-Sung Kim

Authors
  1. Seung Chang Yoo
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  2. Kyoung Joon Choi
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  3. Seunghyun Kim
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  4. Ji-Soo Kim
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  5. Byoung Ho Choi
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  6. Yun-Jae Kim
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  7. Jong-Sung Kim
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  8. Ji Hyun Kim
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Corresponding author

Correspondence to Seung Chang Yoo .

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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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Yoo, S.C. et al. (2019). PWSCC Initiation of Alloy 600: Effect of Long-Term Thermal Aging and Triaxial Stress. 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_18

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