Comparison of radiation induced degradation in several austenitic stainless steels used for core internals in LWR

T. Aoki, T. Fukuda, Y. Isobe, A. Hasegawa, M. Sato, K. Abe, K. Matsueda, Y. Nishida

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

4 Citations (Scopus)

Abstract

In this study, existing (SUS304 and SUS347) and alternative candidate (SUS310+Nb and XM-19) stainless steels for core internals in light water reactors (LWRs) were irradiated with light ions. Post-irradiation analysis of alloying elements near grain boundaries and observation of microstructural changes were performed by field emission gun transmission electron microscope (FE-TEM) and X-ray energy dispersive spectroscopy (EDS). The light ion irradiation was carried out through two steps using two ion species: 3 MeV helium ions were implanted at room temperature up to 2000 appm to simulate Ni(n,α) reaction in core internals steels followed by 2 MeV H2+ irradiation at 300°C to a dose of 1 dpa. The irradiation tests were carried out using Dynamitron accelerator at Tohoku University. The dose rate of the proton irradiation was approximately 4×10-5 dpa/sec. Following irradiation, depleted Cr and enriched Ni areas were observed near grain boundaries. The magnitude of segregation for Cr and Ni strongly depended on chemical compositions of the specimens. Segregation of some impurities was also found near grain boundaries. Besides segregation of alloying elements and impurities, small cavities and an increase in dislocation density were observed for nearly all the stainless steels irradiated in this study. The size and density of the cavities depended on both chemical compositions of the specimens and He contents. The microchemical and microstructural changes were discussed as a function of Ni and Cr equivalent in the specimen. Besides, because hydrogen as well as helium accumulated in the sample may play an important roll on radiation induced degradation, the effect of pre-injected helium existence in material on hydrogen diffusion was evaluated.

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
PublisherMinerals, Metals and Materials Society
Pages1099-1106
Number of pages8
ISBN (Print)0873394755, 9780873394758
DOIs
Publication statusPublished - 1999
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 -
CountryUnited States
CityNew Portbeach, CA
Period99/8/199/8/5

ASJC Scopus subject areas

  • Engineering(all)

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