Status of FeCrAl ODS Irradiations in the High Flux Isotope Reactor
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD)
FeCrAl oxide-dispersion strengthened (ODS) alloys are an attractive sub-set alloy class of the more global FeCrAl material class for nuclear applications due to their high-temperature steam oxidation resistance and hypothesized enhanced radiation tolerance. A need currently exists to determine the radiation tolerance of these newly developed alloys. To address this need, a preliminary study was conducted using the High Flux Isotope Reactor (HFIR) to irradiate an early generation FeCrAl ODS alloy, 125YF. Preliminary post-irradiation examination (PIE) on these irradiated specimens have shown good radiation tolerance at elevated temperatures (≥330°C) but possible radiation-induced hardening and embrittlement at irradiations of 200°C to a damage level of 1.9 displacement per atom (dpa). Building on this experience, a new series of irradiations are currently being conceptualized. This irradiation series called the FCAD irradiation program will irradiate the latest generation FeCrAl ODS and FeCr ODS alloys to significantly higher doses. These experiments will provide the necessary information to determine the mechanical performance of irradiated FeCrAl ODS alloys at light water reactor and fast reactor conditions.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE); USDOE Office of Science (SC), Basic Energy Sciences (BES)
- DOE Contract Number:
- AC05-00OR22725
- OSTI ID:
- 1319204
- Report Number(s):
- ORNL/TM-2016/394; AF5810000; NEAF278; TRN: US1601885
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ATOMIC DISPLACEMENTS
IRRADIATION
OXIDES
IRON BASE ALLOYS
CHROMIUM ALLOYS
ALUMINIUM ALLOYS
RADIATION HARDENING
TEMPERATURE RANGE 0400-1000 K
POST-IRRADIATION EXAMINATION
DISPERSION HARDENING
EMBRITTLEMENT
TERNARY ALLOY SYSTEMS
REACTOR MATERIALS