The influence of corrosion potential on stress corrosion cracking of stainless steels in pressurized water reactor primary coolant environment

Due to ageing of the nuclear power plants, stainless steel core components are subjected to increasing radiation doses. This exposure enhances their susceptibility to a special type of stress corrosion cracking (SCC), known as irradiation-assisted SCC (IASCC), which is infl uenced by high temperature water environment, mechanical stresses and the presence of irradiation. One way to assess IASCC is to measure the corrosion potential, which estimates whether it is likely that IASCC occurs (high value of potential) or not (low value). The main task of this research project is to try to fi nd a direct link between the value of the corrosion potential and SCC present in stainless steels exposed to pressurized water reactor (PWR) conditions. Different values of the corrosion potential have been achieved by changing the water chemistry environment by adding oxygen or nitrogen into the autoclave.