About this Abstract |
| Meeting |
2026 TMS Annual Meeting & Exhibition
|
| Symposium
|
Accelerated Qualification Methods for Nuclear Reactor Structural Materials
|
| Presentation Title |
Fatigue testing of miniature Alloy 709 specimens |
| Author(s) |
Abhishek Kc, Steven Frankowski, Wyatt Peterson, Caleb Massey, Stephen Taller, Kevin Field, Khalid Hattar, Miguel Crespillo, Eric Lang, Charles Hirst |
| On-Site Speaker (Planned) |
Abhishek Kc |
| Abstract Scope |
Alloy 709 is an advanced austenitic steel being considered for sodium-cooled fast reactors due to its excellent creep resistance as well as compatibility with liquid sodium (Na). We examined the performance of SSJ geometry tensile specimens under fatigue loading, and ion-irradiated micropillars. The tests were carried out under stress-controlled cyclic loading with loading rate comparable with 1x10-3 s-1 strain rate so that the results can be compared with bulk strain-controlled experiments. Analysis of slip and dislocation activity around precipitates and irradiation-induced defects will be presented. S/TEM techniques will be leveraged to analyze changes in microstructure with particular focus on precipitate stability of Alloy 709. Changes in microstructure as well as physical properties will be compared with ongoing studies on neutron-irradiated Alloy 709 specimens. The results of this study will further validate the feasibility of using miniature specimens for use in accelerated testing. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Nuclear Materials, Iron and Steel, Characterization |