抄録
To evaluate factors affecting irradiation assisted stress corrosion cracking (IASCC) behavior, the OECD Halden Reactor Project (HRP) is performing in-core crack growth and crack initiation studies in the Halden boiling water reactor (HBWR). In the crack growth studies, compact tension (CT) specimens are prepared from irradiated stainless steels (SSs), types 316NG, 347, 304L and 304 with neutron fluences in the range of 9.0x10^<20> - 1.2x10^<22> n/cm^2 (E > 1 MeV). The crack growth measurements are conducted in simulated boiling water reactor (BWR) and pressurized water reactor (PWR) environments. In a crack initiation study, miniature tensile specimens are machined from a type 304L SS irradiated to 8.0x10^<21> n/cm^2 (E > 1 MeV). Two specimens are installed in each test unit which has pressurized bellows for loading and a linear variable differential transformer (LVDT) for failure detection. Constant load is applied to the specimens in the 15 test units which are exposed to a simulated BWR environment. 5 signal changes indicating specimen failure have been detected since May 2002, when the test began.
