SSRT facility for in-situ observation in high temperature water of irradiated materials

Nakano, Junichi; Koya, Toshio ; Endo, Shinya; Ugachi, Hirokazu ; Tsuji, Hirokazu; Tsukada, Takashi 

Irradiation assisted stress corrosion cracking (IASCC) is one of the key issues for the life management of light water reactor (LWR) core components. For understanding IASCC phenomenon, a slow stain rate testing (SSRT) facility with a capability for in-situ observation in high temperature water for irradiated materials was developed. The SSRT facility has an autoclave with a window for in-situ observation and has been designed for SSRT under boiling water reactor (BWR) condition. To simulate normal water chemistry (NWC) and hydrogen water chemistry (HWC) of BWR environment, dissolved oxygen and hydrogen concentrations (DO and DH) can be controlled within the range of 10 ppb to 32 ppm and 10 ppb to 2.8 ppm, respectively. Hydrogen peroxide can be injected into the autoclave to simulate the radiolysis of water in the reactor core. As a trial run, in-situ observation for an unirradiated material during tensile test in water at 561K was performed and it was confirmed that the load-elongation curve and images could be obtained successfully.