Microscopy and Microanalysis | 2021
Post-Irradiation Analysis of Additively Manufactured Stainless Steel 316L Specimens
Abstract
The nuclear industry is exploring additive manufacturing technologies for reactor internal applications in order to enhance performance, reduce economic costs, and shorten production cycles (Bergeron and Crigger, 2018; Song et al., 2019). Nuclear reactor materials are known to be susceptible to hardening and embrittlement, irradiation-assisted stress corrosion cracking, and void swelling when exposed to high temperature radiation fields for long periods of time (Was and Andresen, 2012). Austenitic stainless steels are a leading candidate to demonstrate AM in nuclear applications given their weldability, good mechanical properties, and corrosion resistance at high temperatures (Şahin and Übeyli, 2008). Stainless steel 316L (SS-316L) has the added benefit of an extensive history of use in a variety of reactor applications as well as a wide availability of alloy powders in the particle size specifications necessary for compatibility with the laser-based additive manufacturing processes.