A Report on the Microstructure of As-Fabricated, Heat Treated and Irradiated ZrC Coated Surrogate TRISO Particles

Zirconium carbide is a candidate to either replace or supplement silicon carbide as a coating material in fuel particles for high temperature gas-cooled reactor fuels. Desirable characteristics of ZrC as a fuel coating include high melting point, adequate fission product retention capability, appropriate neutronic characteristics, resistance to fission product palladium corrosion, and reasonable thermal conductivity. However, there is not sufficient data to demonstrate the suitability of ZrC for nuclear fuel applications. The US and Japan have initiated a collaborative Study to evaluate the feasibility of using ZrC as a fuel coating material, in which microstructural, thermophysical, and thermomechanical properties of developmental ZrC coatings are being evaluated for both unirradiated and irradiated conditions. This paper presents the results of a joint endeavor to study the microstructural evolution, including the effects of C/Zr ratio, in as-fabricated and heat treated samples of nominally stoichiometric and carbon rich ZrC coated surrogate microspheres irradiated to fluence of 2 and 6 dpa at 800 and 1250 degrees C. Grain growth was the primary irradiation effect observed in all the samples. Microstructural examination revealed greater grain growth in stoichiometric ZrC coatings predicting mechanical unsuitability of currently proposed ZrC for nuclear fuel coating applications.