Miao Song

Microstructural Study on the Stress Corrosion Cracking of Alloy 690 in Simulated Pressurized Water Reactor Primary Environment

This study was aimed at investigating the intergranular attack near a stress corrosion crack (SCC) of alloy 690 in simulated pressurized water reactor (PWR) primary water environment. Solution annealed alloy 690 was evaluated for its SCC initiation susceptibility in 360 °C hydrogenated pure water using slow strain rate tensile technique. After the test, a grain boundary showing SCC initiation was sampled with Focused Ion Beam (FIB) milling. The microstructure and elemental distribution near the crack tip were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM). The results show that intergranular oxidation occurs ahead of the crack tip and is preceded by diffusion induced grain boundary migration. The oxides at the crack tip are mainly composed of NiO and Cr2O3 which maintain rigid orientations with the neighboring grains. The adjacent migration zone is free of oxidization as a compact layer of Cr2O3 dominates at the oxide/substrate interfaces and the very tip region.

Irradiation assisted stress corrosion cracking (IASCC) of nickel-base alloys in light water reactors environments part II: Stress corrosion cracking

Radiation-induced microstructural changes control the Irradiation Assisted Stress Corrosion Cracking (IASCC) of core materials, which is a key factor in the extension of the operating lifetime of Light Water Reactors (LWRs). Nickel-base alloys are considered as potential structural materials to replace highly IASCC susceptible austenitic stainless steels. Constant extension rate tensile (CERT) tests were conducted on proton irradiated high strength nickel-base alloy 625 with two different heat treatment conditions (625Plus and 625DA) in both simulated BWR NWC and PWR primary water. Crack length per unit area and fraction of grain boundaries that cracked were used to assess the IASCC susceptibility. Both 625Plus and 625DA showed a very high IASCC susceptibility. 625DA also exhibited greater changes in all microstructure features than 625Plus.

Irradiation assisted stress corrosion cracking (IASCC) of nickel-base alloys in light water reactors environments—Part I: Microstructure characterization

Nickel-base alloys, 625DA and 625Plus have received renewed interest as potential structural materials in nuclear reactors to replace the austenitic stainless steels, which show high susceptibility of irradiation-assisted stress corrosion cracking (IASCC). We investigated the microstructural response of both alloys after 2 MeV protons irradiated to 5dpa at 360 °C in the Michigan Ion Beam Laboratory (MIBL). Transmission electron microscopy was performed on plan-viewed samples with a depth range 9–12 μm prepared by jet-polishing. Detailed analysis included changes in phases, dislocation loops, voids swelling, and radiation induced segregation (RIS). Nano-scaled irradiation-induced precipitates and dislocation loops were pervasive. Voids were absent in these alloys. RIS occurred at random high angle grain boundaries examined. A complete characterization of the irradiated microstructure is required to understand their mechanical and IASCC behavior.