NE Seminar: “Microstructural Variability and Irradiation Assisted Stress Corrosion Cracking: Observations and Implications”

Date/Time
Date(s) - 11/14/2024
1:55 pm - 2:55 pm

Location
Rhines 125

Categories


Abstract

Three pairs of specimens with statistically significant differences in irradiation-assisted stress corrosion cracking (IASCC) response that cannot be explained by existing crack growth rate (CGR) models have been characterized in detail by optical microscopy, microhardness, ferrite scope measurements, scanning electron microscopy (SEM), electron backscatter diffraction (EBSD), atom probe tomography (APT), and nanoindentation, with the objective of identifying key microstructural variables responsible for IASCC response in low electrochemical corrosion potential (ECP) environments of Pressurized Water Reactors (PWRs).

The materials examined covered Type 316 and Type 316Ti baffle bolt material at 25 dpa, Type 304 from a core shroud at ~10 dpa, and two heats of Type 304L control rod material at ~10 and 3.5 dpa. Two independent factors were identified that exacerbate IASCC susceptibility, although no single factor or sole microstructural variable explains the difference in CGR among the pairs of specimens. The potential implications of these observations, with respect to advanced manufacturing, will be discussed.

Bio

Peter Chou, Ph.D.

Principal Technical Leader
Electric Power Research Institute

Dr. Peter Chou is a Principal Technical Leader in the Nuclear Sector of the Electric Power Research Institute (EPRI), where he has worked for 20 years. He received his PhD in Materials Science and Engineering at the University of California, Berkeley. He manages research on environmentally assisted cracking, non-irradiated and irradiated, in the primary environments of light-water reactors, focusing on the link between material microstructure, material reliability, and on the application of advanced characterization and micromechanical testing for industrial research. More recently, he supports EPRI’s effort to assess potential materials issues related to the sodium fast reactor and the gas-cooled thermal reactor, from an operational perspective.