| Abstract Scope |
Laser-based additive manufacturing (AM) has been revolutionizing space and aerospace industries, demonstrating its capabilities to redesign rocket engines, turbine components, and even a full-size rocket nozzle and body for performance enhancement and cost/time-saving. For the nuclear industry, one inevitable barrier to designing next-generation reactors by AM is the limited understanding of material performance in nuclear environments, especially the integrated phenomena involving stress, corrosion, radiation, and high temperature. This talk highlights our continuing efforts and reviews a large body of literatures in the past 10 years to provide the most updated information regarding the material behavior of AM 316L stainless steel (SS) in nuclear environments, with the primary focus on radiation response, corrosion and corrosion cracking, and irradiation assisted stress corrosion cracking (IASCC). In 2020, EPRI/Westinghouse submitted the ASME code case application for ASME Section III Division I (Record 20-254) for AM 316L SS. While the first ASME code of AM is close to be established for LWRs, ASME code does not include corrosion and radiation related properties. This talk intends to update the community with the most recent data development regarding AM 316L SS in light water reactor environments and our perspectives on the adoption of AM for nuclear service. |