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Meeting 2024 TMS Annual Meeting & Exhibition
Symposium Ceramics and Ceramic-based Composites for Nuclear Fission Applications
Presentation Title Advanced Fuel for Integrating Nuclear SMRs with Renewables
Author(s) James Edwin Portwin, Patrick Burr, Jessica Veliscek Carolan, Edward Obbard, Gordon Thorogood
On-Site Speaker (Planned) James Edwin Portwin
Abstract Scope Iodine-induced stress corrosion cracking of zirconium cladding is likely to become a limiting factor on the capabilities of load-following in commercial reactors. We propose a strategy to prevent PCCI by locking iodine within a pellet matrix to reduce/retard cladding interaction, sequestering it into secondary phases with high chemical affinities to iodine. Iodine trapping additives/dopants will be mixed homogeneously in UO2 through a sol-gel synthesis method before a source of iodine (CsI) is powdered, incorporated, and a pellet is pressed and sintered. Iodine release can be measured via TGA through volatilisation of CsI, allowing us to determine the dopants’ potential to lock iodine in a UO2 pellet matrix. Preliminary results indicate challenges associated with methods of iodine implantation into pellets; sintering of CsI volatilises iodine prematurely. A complimentary approach of implanting iodine through ion accelerators avoids the issues of iodine volatilisation during sintering; however, limits the interaction region around 130 nm.
Proceedings Inclusion? Planned:
Keywords Nuclear Materials, Ceramics, Characterization

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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