Conference Logo ProgramMaster Logo
Conference Tools for 2026 TMS Annual Meeting & Exhibition
Login
Register as a New User
Help
Submit An Abstract
Propose A Symposium
Presenter/Author Tools
Organizer/Editor Tools

About this Abstract

Meeting 2026 TMS Annual Meeting & Exhibition
Symposium Metallic Fuels - Design, Fabrication, and Characterization
Presentation Title Investigating Phonon Transport in Uranium Nitride Using Neutron scattering
Author(s) Shaofei Wang, Adam Aczel, Barry Winn, Chris A Marianetti, Michael E. Manley
On-Site Speaker (Planned) Shaofei Wang
Abstract Scope Uranium nitride (UN) has a high thermal conductivity (13.92 W/m·K at 300 K), making it a promising candidate fuel for future fast neutron reactors. Understanding its thermal transport behavior is essential for improving reactor performance and safety. From 300 K to 2000 K, thermal conductivity in UN is predominantly governed by lattice vibrations. Therefore, characterizing the phonons is critical to understanding its thermal conductivity. To gain direct insight into the microscopic mechanisms of heat transport in UN, we performed inelastic neutron scattering to measure phonon energies, group velocities, and lifetimes. These results offer direct insight into the microscopic mechanisms of heat conduction and provide a benchmark for testing modern first-principles models, especially for materials with strong 5f-electron correlations. This neutron work was supported by the U.S.DOE Office of Science, EFRC, Center for Thermal Energy Transport under Irradiation, and used resources at SNS and HFIR.
Proceedings Inclusion? Planned:
Keywords Nuclear Materials, Characterization, High-Temperature 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

Questions about ProgramMaster? Contact programming@programmaster.org | TMS Privacy Policy | Accessibility Statement