Stress corrosion cracking of cold worked austenitic stainless steel pipes in BWR reactor water

Seppo Tähtinen, Hannu Hänninen, Margareta Trolle

Research output: Chapter in Book/Report/Conference proceedingConference article in proceedingsScientificpeer-review

5 Citations (Scopus)

Abstract

The results of the failure analysis, material characterization and stress corrosion cracking (SCC) tests of cold bent, seamless austenitic stainless steel pipes removed from scram system 354 of Ringhals 1 BWR plant after 11 years operating time at 240 °C are presented. The material characterization and also the stress corrosion cracking susceptibility study of the cold bent pipes were performed for materials taken from different depths of the pipe wall thickness. Stress corrosion tests were carried out in simulated BWR water environments at 288°C with varying oxygen contents. The results indicated that highly cold worked material is susceptible to IGSCC without any marked sensitization under constant loading, while dynamic loading results in TGSCC.
Original languageEnglish
Title of host publicationProceedings of the Sixth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems
Subtitle of host publicationWater Reactors
EditorsRobert E. Gold, Edward P. Simonen
Place of PublicationWarrendale
PublisherMinerals, Metals and Materials Society, TMS
Pages265-275
ISBN (Print)978-0-87339-258-7
Publication statusPublished - 1993
MoE publication typeA4 Article in a conference publication
Event6th International symposium on Environmental Degradation in Nuclear Power Systems - San Diego, United States
Duration: 1 Aug 19935 Aug 1993

Conference

Conference6th International symposium on Environmental Degradation in Nuclear Power Systems
CountryUnited States
CitySan Diego
Period1/08/935/08/93

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  • Cite this

    Tähtinen, S., Hänninen, H., & Trolle, M. (1993). Stress corrosion cracking of cold worked austenitic stainless steel pipes in BWR reactor water. In R. E. Gold, & E. P. Simonen (Eds.), Proceedings of the Sixth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems: Water Reactors (pp. 265-275). Minerals, Metals and Materials Society, TMS.