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Meeting 2024 TMS Annual Meeting & Exhibition
Symposium Ceramics and Ceramic-based Composites for Nuclear Fission Applications
Presentation Title Using In Situ Neutron Powder Diffraction to Study the Thermal Expansion of Fission Product Doped UN
Author(s) Melody Ranger, Jennifer H Stansby, Faris B Sweidan, James R Hester, Patrick A Burr, Denise Adorno Lopes, Vanessa K Peterson, Pär Olsson, Edward G Obbard
On-Site Speaker (Planned) Melody Ranger
Abstract Scope UN fuel is being qualified for use in advanced reactors, such as Sweden's SUNRISE lead-cooled reactor. UN doped with Zr, Nb, Mo and Ru at 1, 1.5 and 2 at% each were used to research the effect of fission product solutes and second phases on the thermal expansion of UN at high temperature. Doping with a single solute at 10 at% was also explored to isolate separate effects of each fission product on the thermal expansion. Samples were heated to 1000°C with in situ neutron powder diffraction to determine real time physical and chemical evolution. 10 at% Zr is fully soluble, while Mo, Nb and Ru second phases are observed in UN. Adding 10 at% Mo in UN decreases lattice expansion over the 950°C temperature range by 0.003363 ± 0.001606 Å. The thermal expansion coefficients, phase fractions, and UN occupancies are reported for each sample over the studied temperature range.
Proceedings Inclusion? Planned:
Keywords Characterization, Nuclear Materials,

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

Advanced Fuel for Integrating Nuclear SMRs with Renewables
Analyzing ZrN and LiF-doped ZrN as a Shield Material for Fusion and Space Reactors
Aspects of Graphite Performance in Molten Salt
Beryllium Carbide Tolerance to Radiation Damage for Advance Reactor Moderators
Correlating Atomic Structure with Elastic Properties in Non-textured Pyrocarbon
Elucidating the Effect of Radiation-induced Defect Accumulation on Swelling in UN Using In-situ TEM Irradiation
Evaluating the Strength of TRISO Pebbles via Drop Tests and Nondestructive Techniques
Exploring Hydrogen Absorption with High-density Fuels
From First-principles, Modeling the Effect of Point-defect Phonon Scattering on the Thermal Conductivity of Oxide Fuels
Impact of Grain Boundary and Surface Diffusion on Predicted Fission Gas Bubble Microstructural Evolution Behavior and Release in UO2 Nuclear Fuel
In Situ High-temperature 3D Imaging of the Damage Evolution in a SiC Nuclear Fuel Cladding Material
Mechanical and Thermophysical Properties of ZrC, NbC, and TaC Binary Carbide Surrogate Fuels for Nuclear Thermal Propulsion Systems
Microstructure, Mechanical Properties, and Residual Stresses of Cr-Coated SiC Fuel Cladding for Light Water Reactors
Multiphysics and Multiscale Modeling of the Mechanical Properties of the Porous Pyrocarbon Buffer Layer in TRISO Particle Fuel
Multiphysics and Multiscale Modeling of Micro- and Macro- cracking in High Burnup UO2 Fuels
Next Steps in TRISO Fuel Technology Development
Radially Resolved Thermo-physical Modelling in High Burnup Oxide Nuclear Fuel
Radiation Effects and Corrosion of Silicon Carbide in Nuclear Reactor Environments
Radiation Effects in Ceramics for Immobilization of Actinide-containing Nuclear Waste
Reduced Processing Temperature of Advanced Ceramic Composites
Response of the U3Si2 + 50 wt%UB2 Composite Alloyed with Al, Al2O3, Y and Y2O3 in High-Temperature Oxidizing Atmospheres
Steam Corrosion of Cr- and Zr-containing Uranium Nitride Fuels: Mechanistic Insights from In Situ Neutron Diffraction
Thermal Conductivity Suppression in Uranium-doped Thorium Dioxide Due to Phonon Resonant Scattering
Thermo-mechanical Characterization of a Novel Alumina-Yttria Ceramic Coating for Lead Fast Reactors
Thermodynamics of Complex Carbides for Nuclear Applications in Extreme Environments
Thermophysical Properties of Solid Solution Carbide Fuels for Nuclear Thermal Propulsion
Thermophysical Properties of Uranium Nitride-metal Composite Fuels
Unraveling the Influence of Charge Effect on Defect Recombination in ThO2
Uranium Enrichment Homogeneity Study on HALEU/LEU UO2 Fuel Pellets
Using In Situ Neutron Powder Diffraction to Study the Thermal Expansion of Fission Product Doped UN

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