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Meeting 2026 TMS Annual Meeting & Exhibition
Symposium Metallic Fuels - Design, Fabrication, and Characterization
Presentation Title First-Principles Evaluation of Heat Capacity in Disordered Uranium Oxycarbide Phases for TRISO Fuel Application
Author(s) Asmabi Thottathil, Jerzy Szpunar, Barbara Szpunar
On-Site Speaker (Planned) Asmabi Thottathil
Abstract Scope Accurate prediction of thermal behaviour in TRISO fuel is essential for ensuring the safety and performance of next-generation high-temperature reactors. A key challenge lies in understanding the thermophysical properties of uranium oxycarbide (UCO) fuel, which undergoes significant compositional and structural evolution under irradiation. Among the irradiation-induced phases, rocksalt (UC₁₋ₓOₓ) and fluorite (UCₓO₂₋ₓ) structures emerge within the fuel kernel. This study employs first-principles methods to evaluate the isochoric (Cᵥ) and isobaric (Cₚ) specific heat capacities of these disordered UCO phases. By integrating the quasi-harmonic approximation (QHA) with finite-temperature ab initio molecular dynamics (AIMD), we account for both vibrational and anharmonic contributions to heat capacity in microstructures representative of irradiation-driven disorder. The findings will enhance understanding of how carbon–oxygen distribution influences the thermal storage behaviour of UCO fuel. The resulting data serve as input for fuel performance codes such as BISON, improving thermal transport predictions of the TRISO under irradiation conditions.
Proceedings Inclusion? Planned:
Keywords Modeling and Simulation, High-Temperature Materials, Nuclear Materials

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

Advancements in Metallic Fuel Development and Qualification: Materials Challenges
Advances in the Understanding of Fuel-Cladding Chemical Interaction from FFTF MFF HT9 Clad U-Zr Metal Fuels
An Innovative Technique for Rapid Elemental and Isotopic Characterization of Nuclear Fuels
Analysis of Higher Burnup Sodium Free Annular U-10Zr Fuel
Assessing Thermal Conductivity in U-Zr Fuels: A Comparison of Modeling and Experimental Results
Atomic Mobility Assessments in the U-C-X Ternary Systems
Capturing fission products in metallic fuels using nanostructured interfaces
Comparison of FAST and Historical Irradiation Testing of U-Zr Fuels
Component Assembly and Fuel Fabrication for the IMPACT Advanced Test Reactor Irradiation Experiment
Compositional Effects on Phonon Dispersion and Lifetimes of UxTh1-xO2 using Inelastic X-Ray Scattering
Database-Driven Validation of BISON Metallic Fuel Performance Models for Sodium-Cooled Fast Reactors
Effects of ZrN coating and heat treatment on U-Mo dispersion fuel systems under irradiation
Fabrication and Properties of Metal Fuels with Controlled Porosities by Spark Plasma Sintering
Faulted and Perfect Loop Evolution with Dose and Temperature in Proton Irradiated Uranium Mononitride (UN)
First-Principles Assessment of Thermal Conductivity, in UCO TRISO Fuel with Varying Carbon Stoichiometry
First-Principles Evaluation of Heat Capacity in Disordered Uranium Oxycarbide Phases for TRISO Fuel Application
High-Resolution Characterization of U-Zr Metallic Fuel for Elucidating Precipitation and Fission Products Distribution Using Atom Probe Tomography
Influence of Heterogeneous Microstructures and Large Deformation on Blister Formation in Monolithic U–Mo Nuclear Fuels
Interactions between U-10Mo fuel with a burnable matrix
Investigating Phonon Transport in Uranium Nitride Using Neutron scattering
Investigating the structural and chemical properties in the oxidation of U-Th MOx fuel
Metal Fuels Opportunities Beyond Sodium Fast Reactors
Microalloying Metallic Fuels for Tracking and Traceability
Microstructural Investigation into Two Variants of Fuel-Cladding Chemical Interaction Observed in an Irradiated HT-9 Clad U-10Zr Metallic Fuel
Microstructure Analysis of Uranium-Molybdenum Fuel Alloy from Accelerated Irradiation at Various Temperatures
Novel Phase Identification and Characterization: Experimental insights in The U-Tc Binary System for Metallic Fuel Modeling
Scaling Irradiation Behaviors: Examining Steady-State Swelling, Redistribution, and Fission Gas Release in Larger Diameter U-Zr Fuel Pins
Site-specific porosity and high temperature behavior of U-10Zr fuel
Smaller and faster: conventional vs. nanocalorimetry techniques for determining thermophysical properties of nuclear fuels
Thermal Expansion and Neutron Cross-section of U-Mo fuel to 1000C
Thermal transport of uranium nitride (UN) after irradiation
Thermodynamic and Kinetic Pathways of Impurity-Induced Degradation in U-Ti-Nb-Mo-C Alloys
U-Zr Alloy Properties Review and Applicability to Lightbridge Corporation Fuel Performance Activities

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