| About this Abstract |
| Meeting |
2024 TMS Annual Meeting & Exhibition
|
| Symposium
|
Materials Corrosion Behavior in Advanced Nuclear Reactor Environments
|
| Presentation Title |
Impact of Pre-irradiation and Water Chemistry on In-situ Irradiation-corrosion Behavior of Zircaloy-4 |
| Author(s) |
Peng Wang |
| On-Site Speaker (Planned) |
Peng Wang |
| Abstract Scope |
This study explores the in-reactor accelerated corrosion of zirconium alloys, which arises as a synergistic effect of several factors, including changes induced by irradiation in the microstructure and microchemistry of the base metal, modifications in water chemistry resulting from radiolysis, and the impact of irradiation on the oxide layer. The focus is on the in-situ irradiation-corrosion behavior of Zircaloy-4 in various water environments at 320°C, considering the influence of proton irradiation and examining the separate effects of different mechanisms on corrosion behavior. Using STEM and EDS characterization on FIB lift-out samples, the differences in oxide microstructure and composition are evaluated. This research enhances our understanding of the impact of irradiation damage and radiolysis on corrosion in Zr alloys, contributing to the improvement of safety and performance in nuclear reactor systems. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Characterization, Nuclear Materials, |