Abstract
Austenitic stainless steels are used for core internal structures in sodium-cooled fast reactors (SFRs) and light-water reactors (LWRs) because of their high strength and retained toughness after irradiation (up to 80 dpa in LWRs), unlike ferritic steels that are embrittled at low doses (<1 dpa). For fast reactors, operating temperatures vary from 400 to 550 °C for the internal structures and up to 650 °C for the fuel cladding. The internal structures of the LWRs operate at temperatures between approximately 270 and 320 °C although some parts can be hotter (more than 400 °C) because of localised nuclear heating. The ongoing operability relies on being able to understand and predict how the mechanical properties and dimensional stability change over extended periods of operation. Test reactor irradiations and power reactor operating experience over more than 50 years has resulted in the accumulation of a large amount of data from which one can assess the effects of irradiation on the properties of austenitic stainless steels. The effect of irradiation on the intrinsic mechanical properties (strength, ductility, toughness, etc.) and dimensional stability derived from in- and out-reactor (post-irradiation) measurements and tests will be described and discussed. The main observations will be assessed using radiation damage and gas production models. Rate theory models will be used to show how the microstructural changes during irradiation affect mechanical properties and dimensional stability.
Highlights
Austenitic stainless steels (SS) are widely used within the cores of sodium fast reactors (SFRs) and in light-water reactors (LWRs)
Even though the load drops often observed for irradiated 316 reactors because the high neutron energy (n, α) reaction cross-sections for naturally ocaustenitic stainless steels are coincident with reduced ductility, the fracture surfaces show curring Ni results in an order of magnitude higher He production per atom compared ductile failure at doses ~9 dpa [20]
Models based on point defect diffusion, including that of gaseous atoms that assist with stabilisation of large three-dimensional vacancy clusters, depend on knowing the gas production rate and the percentage of displaced atoms and associated vacancies that are mobile, the so-called freely-migrating point defects (FMDs)
Summary
Austenitic stainless steels (SS) are widely used within the cores of sodium fast reactors (SFRs) and in light-water reactors (LWRs). Atucha reactor uses a mix of natural and enriched uranium (0.85%), with heavy water for moderation and cooling It is like an LWR in having a single pressure vessel made of low-alloy ferritic steel. Austenitic stainless steels are used in reactor cores because of their high strength and toughness They retain sufficient toughness for operability after irradiation in LWRs to doses of approximately 80 dpa [1], unlike ferritic and ferritic-martensitic steels that are embrittled by irradiation at very low doses (
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