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 Fe 3+ 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.
CITATION STYLE
Ishizaki, T., Maruno, Y., Yabuuchi, K., Kondo, S., & Kimura, A. (2019). Development of high irradiation resistant and corrosion resistant oxide dispersion strengthened austenitic stainless steels. In Minerals, Metals and Materials Series (pp. 605–615). Springer International Publishing. https://doi.org/10.1007/978-3-030-04639-2_39
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