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Description
Oxide dispersion strengthened (ODS) alloys are composed of a metallic matrix (typically bcc or fcc) containing a uniform distribution of nanoscale oxide particles. These alloys generally exhibit strong mechanical performance, retaining high strength even at elevated temperatures reaching 800°C. In addition, the nanoscale precipitates function as trapping sites for defects generated by irradiation, considerably enhancing the material's resistance to radiation damage. Owing to these characteristics, ODS alloys are considered attractive candidates for use in various components of Generation IV nuclear reactors, such as high-temperature gas-cooled reactors.
In this work, FeCrAl-Y2O3 ODS alloys with minor additions (0.5–1.0%) of Ti and V were investigated to assess how chemical composition and processing parameters influence microstructure and properties, including behavior under irradiation. The alloys were produced via mechanical alloying followed by spark plasma sintering under optimized conditions. Microstructural characterization using SEM and EBSD showed a fine, uniform grain size of around 1 µm across all samples, with vanadium addition appearing to produce a slight further refinement in grain size. TEM analysis revealed a dense and homogeneous distribution of nanoscale oxide particles (10–40 nm in size), along with the presence of Ti-rich nanometric carbides. Room-temperature tensile testing showed yield strength up to 900 MPa and ultimate tensile strength up to 1050 MPa, while in-situ high-temperature testing up to 800°C demonstrated encouraging mechanical performance under elevated temperatures. Irradiation resistance was assessed through Fe2+ ion implantation at room temperature and 300°C, up to a dose of 5 dpa. Nanoindentation measurements indicated only limited irradiation-induced hardening, while grazing incidence XRD showed increases in both crystallite size and lattice strain following irradiation. Detailed TEM investigations were carried out to characterize irradiation-induced defects and assess the stability of the oxide precipitates. Overall, these findings support the potential of FeCrAl ODS alloys as candidate materials for Generation IV nuclear reactor applications.
