2023 Annual Meeting

Presentation information

Oral presentation

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

[3F05-09] Light Water Reactor Materials

Wed. Mar 15, 2023 10:40 AM - 11:55 AM Room F (12 Bildg.3F 1232)

Chair:Kazunori Morishita(Kyoto University)

11:40 AM - 11:55 AM

[3F09] Correlation between crack growth and corrosion behavior of cobalt-based alloy in simulated PWR primary water

*Takuyo Yamada1, Takahiro Sasaoka2, Takumi Terachi2, Yoshiari Kaneshima1, Kohei Kokutani1, Koji Arioka1 (1. INSS, 2. KEPCO)

Keywords:Co-based alloy, Simulated PWR primary water, Crack growth behavior, Corrosion behavior

Correlation between crack growth and corrosion behaviors of Co-based alloy in simulated PWR primary water were investigated at the temperature ranging between 250-320℃ using forged and welded materials. Preferential oxidation near the carbide/Co-matrix boundary was observed in both alloys after the tests. Such as preferential oxidation might be affected to the crack growth behavior of Co-based alloys.