Co-influence of Irradiation and Helium gas on swelling of pressurized-water nuclear reactors materials components: experimental and numerical Simulations – CoIrrHeSim

Extending nuclear power plants lifetime requires a very good knowledge and a good control of the mechanisms of ageing of all their components. The mechanisms of ageing of the components of a nuclear reactor are very varied. Some of them, such as material fatigue, stress corrosion, corrosion-erosion, wear by friction are quite conventional and are found in many other plants or industrial objects. Other mechanisms are more specific to nuclear power plants, particularly the embrittlement and possible swelling under irradiation, and corrosion under radiation. These mechanisms do not act individually: it is their joint action, "synergistic", that helps to accelerate the ageing of components of a nuclear power plant, and that need to be controlled. This study deals with nuclear reactors material components of Pressurized-Water Reactors (PWR), such as Reactor Pressure Vessel Internals. These components contribute to the cooling of the core by channelling the primary coolant, to its support and to maintain the alignment between fuel assemblies, control rods and instrumentation of the core. This equipment, made of austenitic stainless steels, is subjected to intense irradiation: dose levels up to 120 dpa will be reached at certain locations after 60 years operation at a temperature between 280°C and 380°C. These extreme conditions alter significantly their properties: increase of yield strength, reduction in ductility, etc. For the hottest and most irradiated parts of the baffle assembly a phenomenon of swelling due to irradiation may be possible under PWR conditions. The sensitivity of materials to swelling when irradiated has been studied mainly in the context of work on fast breeder reactors (FBR) above 400°C. It is not therefore possible to extrapolate the FBR data to the operating conditions in PWRs without considering additional factors (in particular the effect of gas produced by the “thermal” neutrons in the PWR spectrum). The data available for the internal components of PWR vessels come from investigations on decommissioned reactors. In most cases, the irradiation temperature and the dose are too low for observable swelling. However, the “hottest” parts of some baffle bolts contain a few cavities (or bubbles): not many, but enough to raise questions about the cavities’ stability, in particular with the aim of reactors lifetime extension. Additional studies suggest that the phenomenon’s temperature dependency is less pronounced than it is for FBRs, making it even more difficult to extrapolate.
The main objective of this project is to define, at the nanoscale, mechanisms of swelling occurring in austenitic stainless steels under long-term neutron irradiation. By coupling experimental and numerical modelling, the ambition of our project is to determine the kinetic of these mechanisms. This will allow anticipating any swelling-related ageing of PWR vessels internal components. The accuracy of this prediction is essential for nuclear power plant safety. Our project brings together public and industrial research laboratories, relying on fundamental studies of materials with considerable industrial issues. It is part of an international context where complementary research programmes exist. By combining numerical simulations and experiments, it will take in consideration synergistic effects of irradiation and helium gas on austenitic stainless steels. This line of research is a novelty compared to in-pile experiments previously performed.