| About this Abstract |
| Meeting |
2025 TMS Annual Meeting & Exhibition
|
| Symposium
|
Mechanical Behavior of Nuclear Reactor Materials and Components IV
|
| Presentation Title |
Irradiation-creep and irradiation-creep-fatigue of austenitic and ferritic-martensitic alloys for advanced nuclear reactors |
| Author(s) |
Charles A. Hirst, Mackenzie Warwick, Wyatt Peterson, Kevin Field |
| On-Site Speaker (Planned) |
Charles A. Hirst |
| Abstract Scope |
The next generation of advanced reactors will subject structural materials to extreme combinations of temperature, irradiation, and stress. As a result, it is critical to evaluate candidate material properties under these coupled environmental factors. This talk will focus on the recent development of irradiation-creep and irradiation-creep-fatigue facilities at the University of Wisconsin Ion Beam Laboratory and Michigan Ion Beam Laboratory. 3 MeV H+ ions are accelerated through 30μm-thin foils to achieve displacement damage without significant implantation. Specimens are mechanically loaded and heated through a combination of external input and internal beam heating. Preliminary results of thermal- and irradiation-creep of austenitic alloys 316L and A709, and ferritic-martensitic alloy Grade 91 will be shown. In addition, progress towards the application of time-dependent loads, enabling irradiation-tensile and irradiation-fatigue testing, will be reported. The development of these in situ ion irradiation testing facilities will yield crucial insight into the mechanical behavior of nuclear materials. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Nuclear Materials, Mechanical Properties, Characterization |