About this Abstract |
| Meeting |
2026 TMS Annual Meeting & Exhibition
|
| Symposium
|
Developments in Advanced Nuclear Structural Materials
|
| Presentation Title |
Mechanical Property Measurements of Neutron-Irradiated 14YWT, 9YWTV and OFRAC ODS Ferritic Alloys |
| Author(s) |
David T. Hoelzer, Jess Werden, TS Byun, Annabelle Le Coq, Caleb Massey |
| On-Site Speaker (Planned) |
David T. Hoelzer |
| Abstract Scope |
Oxide dispersion strengthened (ODS) ferritic alloys are widely considered for advanced cladding concept to be deployed as fast-reactor cladding because of their beneficial high-temperature mechanical properties and irradiation resistance. The deployment of ODS ferritic cladding in advanced reactors requires the collection of relevant performance data, such as the response of mechanical properties to neutron irradiation at relevant doses and temperatures of advanced reactors. To assess the neutron irradiation tolerance of ODS ferritic alloys, post irradiation examination (PIE) of the tensile properties of three ODS ferritic alloys, 14YWT, 9YWTV and OFRAC was performed following neutron irradiations at the High Flux Isotope Reactor. Sub-sized SS-J2 tensile specimens were exposed to target temperatures of 300°C, 385°C or 525°C and doses of 8, 16 and 50 dpa. This presentation will correlate the processing history of the three ODS ferritic alloys to their neutron irradiation hardening behavior obtained from the tensile tests. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Nuclear Materials, High-Temperature Materials, Mechanical Properties |