A High Density Uranium Zirconium Carbonitride LEU Fuel for Application in Fast Reactors

U-Zr-C-N fuel is a high density, high temperature fuel that has potential for application in different types of reactors, including fast reactors. In the past, reactor tests using U-Zr-C-N fuel have been performed to low burnup. However, reactor-testing data is still needed at high burnup to confirm the optimal performance of this fuel type. The SM-3 reactor, which is a high-flux reactor located in Dmitrovgrad, Russia, will be used to test a U-Zr-C-N (U0.9Zr0.1C0.5N0.5) fuel to ~40% burnup. The fuel will then be examined to determine its performance during irradiation. The fuel that will be tested has a density of 12.5 g/cm3 and an enrichment of 19.73% (uranium-235), and the uranium density of this fuel material is 11.3 g/cm3. Over 1200 effective days of irradiation will be required to achieve the targeted burnup. This paper desribes the design of the experiment that will be performed in the SM-3 reactor and discusses the results of different calculations that have been performed to show that the experiment design will meet all objectives. (author)