Optimized chemical composition and heat treatment conditions of 316 CW and high-chromium austenitic stainless steels for PWR baffle former bolts

T. Yonezawa, T. Iwamura, K. Fujimoto, K. Ajiki

Research output: Chapter in Book/Report/Conference proceedingConference contribution

11 Citations (Scopus)

Abstract

To develop irradiation assisted stress corrosion cracking (IASCC) resistant austenitic stainless steels that are resistant to stress corrosion cracking under irradiation in PWR, the authors investigated the optimized chemical compositions and heat treatment conditions of 316 cold working (CW) and high-chromium austenitic stainless steels for PWR baffle former bolts. IASCC susceptibility was evaluated using SSRT tests in simulated PWR primary water for simulated austenitic stainless steels whose chemical compositions are simulated to the grain boundary chemical composition of 316 stainless steel after irradiation of 2 × 1022 n/cm2 (E>0.1MeV). The optimized chemical composition of 316 CW stainless steel consists of ultra-low-impurity levels and high-chromium content within the requirements of the ASTM A193 B8M. About 20% cold working before aging and after solution treatment has also been recommended for this steel to recover sensitization and to induce the grain boundary precipitation of coherent M23C6 carbides. Heating at 700 to 725°C for 20 to 50 hours was selected as a suitable aging condition. Cold working of 5 to 10% for this steel after aging was selected to meet the requirements of mechanical properties for PWR baffle former bolts. The optimized chemical composition of the high-chromium austenitic stainless steel consists of ultra-low-impurity levels, 30% chromium, and 30% nickel contents in order to increase the primly water stress corrosion cracking (PWSCC) resistance and bring the thermal expansion coefficient of this material close to that of a 304 stainless steel for baffle plates. Also, heating at 700 to 725°C for longer than 40 hours was selected as the suitable aging condition. Cold working of 10 to 15% after aging was selected to meet the requirements of mechanical properties for PWR baffle former bolts.

Original languageEnglish
Title of host publicationProceedings of the Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors -
EditorsS. Bruemmer, P. Ford, G. Was, S. Breummer, P. Ford, G. Was
PublisherMinerals, Metals and Materials Society
Pages1015-1026
Number of pages12
ISBN (Print)0873394755, 9780873394758
DOIs
Publication statusPublished - 1999
Externally publishedYes
EventProceedings of the Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors - - New Portbeach, CA, United States
Duration: 1999 Aug 11999 Aug 5

Publication series

NameProceedings of the Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors -

Other

OtherProceedings of the Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors -
Country/TerritoryUnited States
CityNew Portbeach, CA
Period99/8/199/8/5

ASJC Scopus subject areas

  • Engineering(all)

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