Irradiation Assisted Stress Corrosion Cracking And Impact On Life Extension

ABSTRACT The paper aims to give a present status of knowledge on Irradiation Assisted SCC (IASCC) from the point of view of nuclear power plant life extension. Field experience as well as laboratory test results are considered to illustrate how IASCC already impacts plants operation. IASCC has been differentiated with difficulty from standard IGSCC, but above a threshold radiation dose an enhanced portion of intergranular fracture are convincing evidences for IASCC. IASCC cracking in BWR core shrouds started wide studies of materials of irradiated reactor internals. The last cases of IASCC have been found in PWR/WWER baffle bolts. Laboratory slow strain rate tests as well as constant load and crack growth rate tests in simulated BWR and PWR environments resulted in the discovery of the IASCC threshold dose, the threshold stress and that the cracking kinetics increase with neutron exposure. The present understanding of the mechanism of IASCC in BWR / PWR systems is given assuming no fundamental differences between the two environments. Localized deformation on grain boundaries affected by segregation and environmental effects is the most likely mechanism. Finally, plant life extensions likely will bring an increase of IASCC risk due to higher accumulated doses. The impact on BWR and PWR internal components is estimated and discussed. It is concluded that further critical experiments and complex data analyses are urgently needed. INTRODUCTION The paper aims to give a present picture regarding what we know and do not yet know about Irradiation Assisted Stress Corrosion Cracking (IASCC). The IASCC is a term used for stress corrosion cracking of structural materials appearing after irradiation in water environments. The IASCC phenomenon concerns mainly irradiated austenite stainless steels (ASS) and Ni based alloys of BWR and PWR reactor vessel internals (RVI). In the paper only ASS of 300 series will be considered.