In Situ and Ex Situ Observations of the Influence of Twin Boundaries on Heavy Ion Irradiation Damage Effects in 316L Austenitic Stainless Steels
Austenitic stainless steels are widely used in Light Water Reactors (LWR) in annealed or cold-worked conditions. Their in-service temperature ranges between 270 and 370 °C, and in some regions of the reactor they can receive radiation damage levels up to about 4 dpa/year. Radiation can induce defects such as micro-compositional segregation and formation of microstructural defects such as dislocation loops, stacking fault tetrahedra, precipitates, and voids, which may undermine corrosion and mechanical properties. One consequence of these damage effects is irradiation-assisted stress corrosion cracking (IASCC), which is one of the most important degradation mechanisms in austenitic stainless steels in LWR.
It is well-known that interfaces in materials can act as defect sinks for radiation induced defects and thus can reduce neutron-induced damage (or slow down damage accumulation with dose) . In addition to grain boundaries, polycrystalline austenitic stainless steels...