Effects of Neutron Irradiation from a 241 Am-Be Source on the Microstructure, Electrochemical, and Mechanical Properties of ODS Ferritic Alloys

The availability of structural materials resistant to neutron irradiation is a key aspect in the design of fusion reactors currently under development. Among the alloys that are being investigated for this type of application, there are the so-called oxide dispersion strengthened (ODS) ferritic alloys, which exhibit enhanced mechanical properties and improved corrosion and irradiation resistance. The purpose of this work was to study the effects of neutron irradiation on the microstructure, electrochemical behavior, and mechanical properties of Fe-Cr-W-Ti ferritic alloys with dispersion of Y2O3 particles in their matrix, in concentrations of 0, 0.4, and 0.8 wt%. The fabrication process of these alloys, which involved high-energy milling followed by spark plasma sintering, has been detailed in a previous work. A 241Am-Be neutron source was used to irradiate the ODS alloys. The irradiated samples were characterized for particle size distribution, phase composition by X-ray diffraction, microstructure by scanning electron microscopy/energy-dispersive X-ray spectroscopy and electron backscatter diffraction, Vickers microhardness, and corrosion rate by potentiodynamic polarization tests. It was observed that incorporating Y2O3 into the ferritic alloys changed their properties, refining their grain size and improving their resistance to neutron irradiation. In general, the results showed that the ODS ferritic alloys studied maintained good mechanical properties and microstructural stability after neutron irradiation, attributed to a reduced dislocation density, lower porosity, and grain size refinement. Among the three ODS ferritic alloys analyzed, the 0.8 wt% ODS alloy exhibited the best properties after neutron irradiation, making it a promising candidate to further study as a structural material for potential use in nuclear fusion reactors.