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Meeting MS&T23: Materials Science & Technology
Symposium Ceramics for New Generation Nuclear Energy System Application
Presentation Title Phase Equilibria and Thermodynamics of Uranium Mononitride Fuel Undergoing Burn-Up in a Lead-cooled Reactor
Author(s) Ronald Booth, E. Reece McManus, Antoine Claisse, Theodore M Besmann
On-Site Speaker (Planned) Ronald Booth
Abstract Scope Uranium nitride has excellent thermal conductivity and high uranium volume fraction, making it of interest for modern lead-cooled nuclear reactor designs. During the operational life of a UN-fueled reactor many different fission products will be generated, including rare earth and alkaline earth elements. Understanding how the generated fission products interact with the UN fuel is important for predicting their long-term stability. In this work, a thermodynamic database is being developed to analyze the interaction of UN with the developed fission products. The database will be used in thermodynamic equilibrium calculations to understand the phase equilibria and thermodynamic behavior expected, particularly at high UN fuel.


A New Method for Measuring Refractory Corrosion of Ceramics in Glass
Advanced Characterization of Nuclear Fuels to Support Qualification of Nuclear Fuels
AI/ML-assisted Design of Phosphate Nuclear Waste Forms
Analysis of Radially Resolved Thermal Conductivity in High Burnup Mixed Oxide Fuel
Atomistic Understanding of Thermal Conductivity Degradation in Irradiated Oxide Fuels
Creep Predictions in UO2 by Atomistic to Meso-scale Simulations
Crucible-scale Corrosion Testing of Monofrax® K-3 Refractory in Contact with Glass Melts
Crystal Growth of Actinide Materials as Potential Nuclear Waste Forms
Decoding the Structural Descriptors Controlling Nepheline Crystallization in Borosilicate-based Nuclear Waste Glasses
Development of Radiation Attenuating Geopolymer-particulate Composites
Development Strategy for SiC/SiC Composite Accident Tolerant Fuel Cladding
Disordered Enthalpy-entropy Descriptor for High-entropy Ceramics Discovery
Engineered Ceramic Composites for Neutron Moderation and Shielding in Advanced Reactors
Evolution of the Chemical State in Molten Salt Reactors during Operation and Implications for Materials Behavior
Impact of Phonon Resonant Scattering on Thermal Conductivity of Uranium-doped ThO2
Interface Effect on the Distributions of Radiation Induced Defects
Investigating the Effects of Irradiation on Microstructure, Micromechanical and Thermal Properties of High Entropy Carbide Ceramics
Investigating the Radiation Response of Oxide Materials with Neutron Scattering
Iron-phosphate Glasses for the Immobilization of Dehalogenated Chloride-based Waste Salt
M-2: Synthesis of Uranium Nitride Nuclear Fuel
Metal-halide Perovskites as Innovative and Cost-effective Salt Waste Form
Metal Hydride Moderators: A Historical Perspective of Their Design and Implementation
Microstructure and Mechanical Properties of Ceramics in Y-Ti-O System
Modeling of Pressure-driven Inter-granular Fracture in High Burnup Structure UO2 during LOCA Using a Phase-field Approach
Modeling the Effect of Point Defect Scattering on the Thermal Conductivity of ThO2
Non-Equilibrium Ionic Transport in Oxides
Oxygen Vacancy Formation Energetics in MgO-based High Entropy Oxides from DFT and Experimental Validation
Phase Equilibria and Thermodynamics of Uranium Mononitride Fuel Undergoing Burn-Up in a Lead-cooled Reactor
Phonon Modal Analysis of Thermal Transport in ThO2 with Defects
Processing of High Entropy Metal Carbides: A New Class of Ultrahigh Temperature, Irradiation Resistant Ceramics
Proton Irradiation and Characterization of ThO2, UxTh1-xO2, CeO2, UO2 and Zr:UO2 Single Crystals
Scalable Manufacturing of Garnet Structured LLZO Ceramic Tubes With Applications in Next Generation Fusion Systems
Stability of Radiation−Induced Bixbyite Phase in δ−Sc4Hf3O12
Status and Outlook of Tristructural Isotropic Coated Particle Fuel Technology
Thermal Oxidation and Thermodynamics of Uranium Nitride and Uranium Carbide
Thermal Property Mapping of Surrogate TRISO Particles
Thermodynamic Assessment of Chromium and Nickel Corrosion in Molten Fluoride Salts
Uncertainty Quantification and Propagation of NaCl-KCl-MgCl2 Thermodynamic Functions for Molten Salt Applications
Welding Development of Cladding Materials for Ceramic Fuels

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