An evaluation of tri-valent oxide (Cr2O3) as a grain enlarging dopant for UO2 nuclear fuels fabricated under reducing environment

A study was performed to evaluate the microstructure and crystallography of nominally 500–2000 Cr2O3-doped UO2 fabricated in a temperature range of 1150–1750°C under reducing experimental conditions. An increase in grain size of the samples was observed with the increase in heat treating temperature as expected. For a given sintering temperature (1700–1750°C), an increase in the grain size was also observed with the increase in Cr2O3 concentration up to a value of ~1000–1200 wppm. A decrease in fission gas release as a function of grain size was estimated for the Cr2O3-doped UO2 samples assuming specified post-irradiation annealing conditions. A nearly linear decrease was obtained in the lattice parameter of the Cr2O3-doped UO2 fcc phase with the increase in Cr2O3 concentration, especially up to a nominal value of 1000 wppm. The lattice parameter decrease was also persistent with the increase in the average grain size as a result of addition of Cr2O3 into the UO2 lattice. An increase in the crystallite size and a decrease in the microstrain of the fcc phase were observed with the increase in the average grain size of the samples, indicating a higher crystallinity of the Cr2O3-doped samples than that of the undoped UO2 sample.