| About this Abstract |
| Meeting |
2022 TMS Annual Meeting & Exhibition
|
| Symposium
|
Mechanical Behavior and Degradation of Advanced Nuclear Fuel and Structural Materials
|
| Presentation Title |
Characterizing and Testing High Dose Neutron Irradiated Materials for Cladding Applications |
| Author(s) |
Stuart Andrew Maloy, Ben Eftink, Tarik Saleh, Mychailo Toloczko, Dave Hoelzer, T. S Byun |
| On-Site Speaker (Planned) |
Stuart Andrew Maloy |
| Abstract Scope |
The Nuclear Technology R&D program has significant experience at qualifying metallic fuels for fast reactor applications. In this process, new fuels and cladding materials must be developed and tested to high burnup levels (e.g. >20%) requiring cladding to withstand very high doses (greater than 200 dpa) while in contact with the coolant and the fuel. New ferritic/martensitic and ferritic Oxide Dispersion Strengthened (ODS) alloys have been developed and tested with improved radiation tolerance after high dose neutron exposures. Research includes developing ferritic/martensitic and ferritic ODS alloys in plate and tube form for future nuclear applications. Recent progress in high dose irradiated materials testing and materials development will be presented including recent plans for testing high dose irradiated materials. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Nuclear Materials, Mechanical Properties, Characterization |