2018 Fall Meeting

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Oral presentation

V. Nuclear Fuel Cycle and Nuclear Materials » 502-1 Nuclear Materials, Degradation, Radiation Effects, and Related Technology

[2C04-07] Fuel Cladding Material

Thu. Sep 6, 2018 10:50 AM - 11:55 AM Room C (B21 -B Building)

Chair:Sho Kano(Univ. of Tokyo)

11:35 AM - 11:50 AM

[2C07] Permeation behavior of tritium through FeCrAl-ODS ferritic steel

*Yudai Urabe1, Kenichi Hashizume1, Teppei Otsuka2, Kan Sakamoto3 (1. Kyushu Univ., 2. Kindai Univ., 3. NFD)

Keywords:fuel cladding, FeCrAl, ODS

A FeCrAl-ODS ferritic steel has been examined as a candidate Fe-Cr-Al alloy for the fuel cladding in BWRs. The property examined in this study is tritium permeability. In this work, two specimens : FeCrAl-ODS steel and oxidized FeCrAl-ODS steel were tested by means of hydrogen transmission experiment using tritium tracer technique.