Helium Effects on IASCC
The formation and growth of cracks by irradiation-assisted stress corrosion cracking (IASCC) of stainless steel reactor internals is a critical issue for the safe long-term operation of LWRs and such cracking instances already occurred in Swiss reactors (core shroud in KKM, baffle bolts in KKB). Recent studies show that a huge amount of helium (up to 1000 atomic parts per million) can be accumulated in some reactor internal components of pressurized water reactors (PWR) after long-term operation (above 40 years) that could significantly increase (or even dominate) the IASCC susceptibility at high doses in addition to conventional mechanism like irradiation-induced segregation and hardening etc. and that material test reactor data thus potentially underestimate the real IASCC concern for PWR long-term operation.
The main scientific goal of this new PhD project, which is a joint activity of ANM and INTEGER and is supported by swissnuclear, is to make a contribution on the clarification of the unexplored role of He in IASCC of austenitic stainless steels and to determine thresholds concentration for He effects on mechanical properties and IASCC in air and high-temperature water, respectively. The project will thus provide an important contribution to the assurance of the safe long-term operation of Swiss PWRs (Beznau-I is the oldest PWR world-wide) and directly contributes to the maintenance of expertise and education of young specialists in the field of material ageing of reactor internals.
The main scientific goal of this new PhD project, which is a joint activity of ANM and INTEGER and is supported by swissnuclear, is to make a contribution on the clarification of the unexplored role of He in IASCC of austenitic stainless steels and to determine thresholds concentration for He effects on mechanical properties and IASCC in air and high-temperature water, respectively. The project will thus provide an important contribution to the assurance of the safe long-term operation of Swiss PWRs (Beznau-I is the oldest PWR world-wide) and directly contributes to the maintenance of expertise and education of young specialists in the field of material ageing of reactor internals.