|About this Abstract
||2023 TMS Annual Meeting & Exhibition
||Microstructural, Mechanical and Chemical Behavior of Solid Nuclear Fuel and Fuel-cladding Interface
||Chemical Structures and Thermodynamics of Uranium Nitride and Uranium Carbide
||Xiaofeng Guo, Vitaliy Goncharov, Juejing Liu, Arjen van Veelen, Joshua White, Hongwu Xu
|On-Site Speaker (Planned)
In the U.S. and many other countries, carbide and nitride matrices have received considerable attention as advanced nuclear fuel types. Compared to UO2, both of UC and UN have the advantages of high thermal conductivity, high melting point, and high fissile density. A fundamental understanding of the material chemistry and thermodynamic properties of UC and UN is critical for predicting their behavior under reactor or extreme conditions. In this work, we investigated (1) the local structures by X-ray diffraction and X-ray absorption fine structure, (2) bulk thermal oxidations by TGA – DSC, and thermochemical reactions, including the enthalpy of oxidation and standard enthalpy of formation of UN and UC by high temperature drop solution calorimetry. These updated understanding of UN and UC have two implications: (1) enable thermodynamic modeling and DFT computation for U-C, U-N, and U-C-N; and (2) a foundation for future studies on UC-, and UN-derived waste forms.
||Ceramics, Nuclear Materials, High-Temperature Materials