Date of Award


Document Type

Campus Access Thesis


Nuclear Engineering

First Advisor

Travis W. Knight


The characterization of an advanced nuclear fuel for Gas Fast reactors has been studied. Portions of a dispersion/composite fuel involving Uranium Carbide (UC) and Zirconium Carbide (ZrC) have been characterized and studied. Uranium carbide (UC) microspheres produced in the USC-Nuclear Materials Laboratory were subjected to metallographic techniques, and then characterized by analytical methods. A method for separation of spherical and non-spherical microspheres was developed involving an inclined plane. Metallography was done using the LECO SS-1000 grinder/polisher system. Quantitative analysis and imagery were gathered using scanning electron microscopy (SEM), back scattering electron microscopy (BSE), electron microprobe for quantitative analysis (EPMA), and x-ray photoelectron spectroscopy (XPS). X-ray diffraction (XRD) was also used to find the crystal structure of the microspheres. The UC microspheres were further investigated by annealing. The annealing process was completed using a CM Furnace in an inert argon gas. The results of the experiment were analyzed using the same methods mentioned above. Uranium diffusion was found in the ZrC matrix and was confirmed to be possible through diffusion calculations.