|About this Abstract
||2018 TMS Annual Meeting & Exhibition
||Environmentally Assisted Cracking: Theory and Practice
||IASCC Behavior of Additively Manufactured 316L Stainless Steel in Light Water Reactor Environments
||Mi Wang, Miao Song, Xiaoyuan Lou, Raul Rebak, Gary Was
|On-Site Speaker (Planned)
316L stainless steel is widely used in nuclear reactor internal components because of its reliable mechanical properties, good corrosion resistance, weldability, and reasonable cost. Additive manufacturing (AM) is being explored as a processing route for these components, however no data exists on their susceptibility of irradiation assisted stress corrosion cracking (IASCC) in reactor environments. IASCC behavior was studied using constant extension rate tensile (CERT) tests on two heat treatment conditions (stress-relieved and post-printing hot-isostatically-pressed (HIPed)) of AM 316L that were irradiated with 2 MeV protons to doses of 5 dpa at 360°C at a strain rate of 1 x 10-7 s-1 in both PWR primary water and BWR NWC environments to a fixed strain. Better IASCC resistance was found in HIPed AM 316L. The cracking behavior is interpreted in the context of the irradiated microstructure (dislocation loops, radiation induced precipitates), grain boundary properties, and testing environments.
||Planned: Supplemental Proceedings volume