NSUF 26-5670: Effect of Extrusion Temperature on Mechanical Property Evolution for FeCrAl-ODS Alloys Following Neutron Irradiation at Liquid Metal Fast Reactor Operating Temperatures
This RTE aims to investigate the difference in mechanical properties of FeCrAl-ODS alloys following high-dose neutron irradiation as a function of extrusion temperature. If successful, this experiment will help quantify how differences in nanoscale precipitate dispersions generated during alloy consolidation affect the retained ductility and mechanical strength of low-Cr FeCrAl-ODS alloy candidates within the operating temperature regime for liquid metal lead-bismuth eutectic (LBE) fast reactors. This RTE will generate data to demonstrate the viability of low-Cr FeCrAl-ODS alloys as viable candidates for fast reactor operating conditions after high-dose neutron irradiation. Post-irradiation examination will be completed on two FeCrAl alloys following neutron irradiation to 8, 16, and 50 dpa at target irradiation temperatures of 385°C and 525°C within Oak Ridge National Laboratory’s High Flux Isotope Reactor. A total of 36 tensile tests will be conducted at the irradiated materials examination facility (IMET) using the vacuum mechanical test frame available within Hot Cell 1, comprising 12 tensile tests at 23°C and 24 tests at the target irradiation temperature, totaling a total of 6 days of required testing time within the IMET facility. The alloys of interest, alloy 106ZY10C and 106ZYA15C, are of the same nominal composition, but differ in the extrusion temperature used for consolidation into ODS alloy bars. As such, the initial ductility is higher for the latter alloy due to the use of a higher extrusion temperature, but the precipitate and grain boundary sink strength may be lower, which may cause differences in the evolution of strength and ductility as a function of increasing dose. Over the course of 8 months, we expect to collect all data on these specimens, resulting in a new understanding of nanoscale defect evolution wrought and ODS FeCrAl alloys as potential fuel cladding.
Additional Info
| Field | Value |
|---|---|
| Awarded Institution | First American Nuclear Company |
| DOI | 10.46936/NSUF/60017655 |
| Embargo End Date | 2028-04-22 |
| Facility Tech Lead | Kory Linton |
| NSUF Call | FY 2026 RTE 1st Call |
| PI | Andrew Caldwell |
| PIE Facilities | Irradiated Materials Examination and Testing Facility |
| Project Member | Dr. Stuart Maloy, Senior Nuclear Materials Adviser - Pacific Northwest National Laboratory (https://orcid.org/0000-0001-8037-1319) |
| Project Member | Dr. Caleb Massey, R&D Staff Member - Oak Ridge National Laboratory (https://orcid.org/0000-0003-1093-3958) |
| Project Member | Mr Philip Jensen, VP Nuclear Fuels - First American Nuclear (https://orcid.org/0000-0003-4286-1039) |
| Project Member | Dr. Andrew Caldwell, Materials Science Engineer - First American Nuclear (https://orcid.org/0009-0008-4042-738X) |
| Project Type | RTE |