Fig. 9-5 Neutron and gamma radiations in a reactor core result in various effects on the coolant (high temperature water) of a LWR and on structural materials (stainless steel). As a result of the synergy of these effects, IASCC susceptibility (evaluated by the fracture surface in specimens after post-irradiation IASCC tests) will be manifested.

Fig. 9-6 shows the IASCC susceptibility of 316 stainless steels irradiated in a research reactor under simulated fusion reactor conditions, and in a FBR. Significant IASCC susceptibility was observed in the specimens irradiated at higher temperatures.

Stress corrosion cracking (SCC) in high temperature water is one of the main causes of failure of reactor structures made of stainless steels. In the structure, damage of the piping in a reactor system caused by SCC was overcome as a result of a number of studies. Now, the irradiation induced stress corrosion cracking (IASCC) of austenitic stainless steel for core internals becomes an important research subject in relation to the life evaluation of an LWR. Radiation affects the SCC sensitivity of materials and the corrosiveness of the environment through various tests as illustrated in Fig. 9-5. Researchers in JAERI are studying the controlling factors of IASCC and its mechanism. In order to simulate the in-reactor environment, IASCC tests of irradiated materials at high temperature and in high pressure water are performed in a hot cell. Electrochemical corrosion tests and microstructural observations with an optical microscope and electron microscope are also performed. The effects of irradiation temperature and alloy composition on the IASCC susceptibility were investigated by post-irradiation examinations. These results contribute to the design of a fusion reactor, because fusion reactors also have the IASCC problem. Fig. 9-6 shows the IASCC susceptibility of 316 stainless steels irradiated in a research reactor under simulated fusion reactor conditions, and in a FBR. Significant IASCC susceptibility was observed in the specimens irradiated at higher temperatures.

Reference T.Tsukada et al., Slow Strain Rate Tensile Tests in High Temperature Water of Spectrally Tailored Irradiated Type 316 Materials for Fusion Reactor Applications, CORROSION/92 (NACE,1992), Paper No.104. T.Tsukada et al., Evaluation of Irradiation Assisted Stress Corrosion Cracking (IASCC) of Type 316 Stainless Steel Irradiated in FBR, J.Nucl. Mater., 207(1993)159.

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