Physics Evaluation of Alternative Uranium-Based Oxy-Carbide Annular Fuel Concepts for Potential Use in Compact High-Temperature Gas-Cooled Reactors

Abstract

Abstract Lattice physics calculations have been carried out to evaluate the performance and safety characteristics of a modified high temperature gas-cooled reactor (HTGR) prismatic fuel block concept, based on the MHTGR-350 benchmark problem. Key changes were to replace the conventional tri-structural isotropic (TRISO)-filled fuel compacts with heterogeneous, multilayer annular fuel pellets made with UCO, ThCO, or (U,Th)CO. These fuel pellets have multiple protective cladding layers of pyrolytic carbon and silicon carbide, which will give it robust qualities. With the increased loading of U-235 in the fuel block, it was necessary to replace up to 78 fuel holes and 42 coolant holes with a hydrogen-based moderator (7LiH), in order to ensure a thermal neutron energy spectrum in the lattice. Calculation results demonstrate that the modified fuel concept has several advantages and some challenges relative to the conventional MHTGR-350 design concept. With the increased uranium loading and the reduced neutron leakage due the use of 7LiH moderator rods, higher burnup levels and lower natural uranium consumption levels can be achieved with the same level of uranium enrichment. In addition, the expected fuel residence time increased by a factor of 20 or more, making such a concept very attractive for use in small, modular, “nuclear battery” HTGRs that would only need to be fueled once. Calculation results for the current concept indicate positive graphite and hydrogen moderator temperature coefficients, and further modifications will be required to ensure a negative power coefficient of reactivity.