About this Abstract |
Meeting |
2022 TMS Annual Meeting & Exhibition
|
Symposium
|
Materials Systems for the Future of Fusion Energy
|
Presentation Title |
NOW ON-DEMAND ONLY - Post-irradiation Examination of High-dose Ion Irradiated MA956 ODS Alloy |
Author(s) |
Yu Lu, Yaqiao Wu, Ramprashad Prabhakaran, Megha Dubey, Lin Shao, Jing Wang, Dalong Zhang |
On-Site Speaker (Planned) |
Yu Lu |
Abstract Scope |
Oxide dispersion strengthened (ODS) alloys are candidate cladding materials for the next generation advanced nuclear reactors due to demonstrated excellent resistance to irradiation damage, high-temperature creep and superior mechanical properties under high temperature and radiation environment. MA956 ODS alloys have been being studied, however, only very limited data was reported regarding to its microstructural evolution under irradiation to date, which makes it difficult to understand its performance fundamentally. In this study, the MA956 ODS alloys are ion irradiated to different doses (2.5, 50 and 100 dpa) under different temperatures (190°C and 320°C). Post-irradiation characterizations are performed by using scanning transmission electron microscopy (STEM), atom probe tomography (APT) and nanoindentation techniques. The irradiation induced defects and the effects of irradiation dose and temperature on the evolution of the dispersoid are illustrated and discussed in details. |
Proceedings Inclusion? |
Planned: |
Keywords |
Nuclear Materials, Iron and Steel, |