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Shibata, Akira; Nakamura, Natsuki; Futakawa, Masatoshi; Maekawa, Katsuhiro*
Zairyo Shiken Gijutsu, 62(1), p.35 - 40, 2017/01
Welding between irradiated and un-irradiated type 316 stainless steel is demanded for in-pile IASCC tests in Japan Materials Testing Reactor. Specimens were made from type 316 stainless steel which was irradiated up to 610
n/m
(
1 MeV) in Reactor and welding tests of irradiated and un-irradiated type 316 stainless steels were performed to confirm integrity of the welding. Relationship between heat input and amount of Helium bubbles in welding bead was evaluated and hardness test was performed. It appeared that there is threshold value of heat input for welding without problem.
Shibata, Akira; Wakui, Takashi; Nakamura, Natsuki; Futakawa, Masatoshi; Maekawa, Katsuhiro*; Naka, Michihiro
Zairyo Shiken Gijutsu, 62(1), p.41 - 47, 2017/01
Nuclear reactor fuel cladding material has been gradually replaced from Zircaloy-4 to Zr-Nb alloys such as M5. To reveal the origin of good corrosion resistance of Zr-Nb alloys, evaluation on oxide layer of fuel cladding tube Zircaloy-4 and M5 was performed. Nano-indentation with spherical indenter was performed on oxide layer of those materials. Test results of the indentation were evaluated by the inverse analysis using Kalman filter and Finite Element Method. The results analysis shows that the oxide layer of M5 has more ductility compare from that of Zircaloy 4. Thus, oxide layer of Zircaloy4 could be injured by smaller strain compare from that of M5.