Development of high irradiation resistant and corrosion resistant oxide dispersion strengthened austenitic stainless steels

Takahiro Ishizaki, Yusaku Maruno, Kiyohiro Yabuuchi, Sosuke Kondo, Akihiko Kimura

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

Abstract

The next generation of light water reactors, resource renewable BWR (RBWR), which can be burned trans uranium (TRU) is currently under development at Hitachi. The RBWR requires a high flux of fast neutron for efficient burning of the TRU, which is four times as large as that of the ABWR. Therefore, structural materials require both a high resistance to corrosion and to irradiation. In this study, oxide dispersion strengthened austenitic stainless steels (ODS-ASUS) with high corrosion resistance have been developed. The objective of this research is to evaluate irradiation resistance and SCC susceptibility in a simulated reactor water environment for the ODS-ASUS. The materials were irradiated with 6.4, MeV Fe3+ at 673, K up to 8.0, dpa using the DuET facility at Kyoto University. The creviced bent beam (CBB) test is conducted to assess the SCC susceptibility in the hot water, 288, °C, 8, MPa with a dissolved oxygen of 8, ppm.

Original languageEnglish
Title of host publicationProceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors
EditorsDenise Paraventi, Michael Wright, John H. Jackson
PublisherSpringer International Publishing
Pages605-615
Number of pages11
ISBN (Print)9783319672434
DOIs
Publication statusPublished - 2018 Jan 1
Externally publishedYes
Event18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, 2017 - Portland, United States
Duration: 2017 Aug 132017 Aug 17

Publication series

NameMinerals, Metals and Materials Series
VolumePart F9
ISSN (Print)2367-1181
ISSN (Electronic)2367-1696

Other

Other18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, 2017
CountryUnited States
CityPortland
Period17/8/1317/8/17

Keywords

  • IASCC
  • Ion irradiation test
  • ODS
  • RIS
  • SCC
  • Stainless steel

ASJC Scopus subject areas

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

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  • Cite this

    Ishizaki, T., Maruno, Y., Yabuuchi, K., Kondo, S., & Kimura, A. (2018). Development of high irradiation resistant and corrosion resistant oxide dispersion strengthened austenitic stainless steels. In D. Paraventi, M. Wright, & J. H. Jackson (Eds.), Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors (pp. 605-615). (Minerals, Metals and Materials Series; Vol. Part F9). Springer International Publishing. https://doi.org/10.1007/978-3-319-67244-1_39