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Irradiation-assisted stress corrosion cracking (IASCC) is a term used to describe cracking of austenitic materials in components subjected to a relatively high fast neutron flux. Like intergranular stress corrosion Paper No.18360 cracking (IGSCC), IASCC appears as intergranular cracks, but thermal sensitization of the grain boundaries is not required for the material to become susceptible to cracking. Service failures caused by IASCC have occurred in components such as core shrouds and top guides in boiling water reactors (BWRs) and baffle bolts in pressurized water reactors (PWRs).
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The austenitic Ni-base Alloy 600 has been extensively used as structural material in primary water reactors (PWR). Despite good resistance against general corrosion in water-cooled nuclear power reactors, the material has been susceptible to stress corrosion cracking (SCC). These observations have led to ongoing discussions of the underlying embrittlement mechanism(s). Internal oxidation of the grain boundary (GB) at typical operating temperatures is one such mechanism, although debate continues on the exact mechanisms at play.