Understanding how irradiation enables swelling-induced embrittlement and defects self-organization
Abstract: Predicting the behaviors of materials in nuclear environments is a grand challenge. Irradiation damage in microstructure and its effect on mechanical properties of materials determine the performance of nuclear core components. Austenitic 316 stainless steels have been proposed as one of the candidate materials for cladding and structural components in advanced nuclear reactors, while irradiation damage can cause significant degradation, including volumetric changes, hardening, embrittlement, etc. The first part of this presentation will discuss irradiation-induced voids and void swelling-induced embrittlement in 316 stainless steels. In nuclear fuels, gas-driven swelling and fission gas release are life limiting. Self-organization of fission gas bubbles leads to the formation of fission gas bubble superlattices, a highly efficient mechanism for fission gas storage at high pressure, persisting to a very high fission density. Fundamental understanding of the underlying physics that govern the self-organization of gas bubbles is essential to predicting the lifetime of nuclear fuel elements. In the second part of this presentation, the role of materials’ anisotropic properties on the formation of gas bubble superlattices will be discussed.
Bio: Dr. Cheng Sun is a senior scientist working in the Materials and Fuels Complex at Idaho National Laboratory (INL). He received his Ph.D. from Texas A&M University in 2013 and has been working in the field of nuclear materials for more than 10 years. He currently leads the Reactor Structural Materials Group in the Division of Advanced Characterization and Post-irradiation Examination at INL. He received INL’s 2023 Exceptional Scientific Achievement Award, 2022 Rising Star Award from Journal of Nuclear Materials, and INL’s 2019 Early Career Exceptional Achievement Award