Pwscc initiation of alloy 600: effect of long-term thermal aging and triaxial stress

Seung Chang Yoo; Kyoung Joon Choi; Seunghyun Kim; Ji Soo Kim; Byoung-Ho Choi; Yun-Jae Kim; Jong Sung Kim; Ji Hyun Kim

doi:10.1007/978-3-030-04639-2_18

Pwscc initiation of alloy 600: effect of long-term thermal aging and triaxial stress

Seung Chang Yoo, Kyoung Joon Choi, Seunghyun Kim, Ji Soo Kim, Byoung-Ho Choi, Yun-Jae Kim, Jong Sung Kim, Ji Hyun Kim

School of Mechanical Engineering

Research output: Chapter in Book/Report/Conference proceeding › Conference contribution

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.

Original language	English
Title of host publication	Minerals, Metals and Materials Series
Publisher	Springer International Publishing
Pages	281-292
Number of pages	12
ISBN (Print)	9783030046385, 9783030046392, 9783319515403, 9783319651354, 9783319728520, 9783319950211
DOIs	https://doi.org/10.1007/978-3-030-04639-2_18
Publication status	Published - 2019 Jan 1
Event	18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors 2019 - Boston, United States Duration: 2019 Aug 18 → 2019 Aug 22

Publication series

Name	Minerals, Metals and Materials Series
ISSN (Print)	2367-1181
ISSN (Electronic)	2367-1696

Conference

Conference	18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors 2019
Country	United States
City	Boston
Period	19/8/18 → 19/8/22

Keywords

Nickel based alloy
Primary water stress corrosion cracking
Thermal aging
Triaxial stress

ASJC Scopus subject areas

Electronic, Optical and Magnetic Materials
Energy Engineering and Power Technology
Mechanics of Materials
Metals and Alloys
Materials Chemistry

Access to Document

10.1007/978-3-030-04639-2_18

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Cite this

Yoo, S. C., Choi, K. J., Kim, S., Kim, J. S., Choi, B-H., Kim, Y-J., Kim, J. S., & Kim, J. H. (2019). Pwscc initiation of alloy 600: effect of long-term thermal aging and triaxial stress. In Minerals, Metals and Materials Series (pp. 281-292). (Minerals, Metals and Materials Series). Springer International Publishing. https://doi.org/10.1007/978-3-030-04639-2_18

Pwscc initiation of alloy 600 : effect of long-term thermal aging and triaxial stress. / Yoo, Seung Chang; Choi, Kyoung Joon; Kim, Seunghyun; Kim, Ji Soo; Choi, Byoung-Ho ; Kim, Yun-Jae; Kim, Jong Sung; Kim, Ji Hyun.

Minerals, Metals and Materials Series. Springer International Publishing, 2019. p. 281-292 (Minerals, Metals and Materials Series).

Research output: Chapter in Book/Report/Conference proceeding › Conference contribution

Yoo, SC, Choi, KJ, Kim, S, Kim, JS, Choi, B-H , Kim, Y-J, Kim, JS & Kim, JH 2019, Pwscc initiation of alloy 600: effect of long-term thermal aging and triaxial stress. in Minerals, Metals and Materials Series. Minerals, Metals and Materials Series, Springer International Publishing, pp. 281-292, 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors 2019, Boston, United States, 19/8/18. https://doi.org/10.1007/978-3-030-04639-2_18

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