Oxide-dispersion-strengthened (ODS) materials are under development for use in applications in both nuclear fission and nuclear fusion power generation. The materials offer high-temperature stability and creep strength, coupled with irradiation damage resistance. The oxide dispersion is typically based on a small fraction (<0.5% by weight) of yttria. The oxide particles act to pin the grain boundaries and the particle/matrix interface acts as a sink for He under irradiation in reactor environments. In this experiment we will investigate the effects of the material processing parameters on the dissolution and internal strain of the dispersoid particles.