ProgramMaster Logo
Conference Tools for 2023 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 2023 TMS Annual Meeting & Exhibition
Symposium Ceramic Materials for Nuclear Energy Research and Applications
Presentation Title Comprehensive Characterization of Damage in Ion Irradiated Ceramics for Validation of Atomistic Models
Author(s) Marat Khafizov, Joshua Ferrigno, Erika Nosal, Saqeeb Adnan, Kaustubh K. Bawane, Amey Khanolkar, Miaomiao Jin, Linu Malakkal, Chao Jiang, Lingfeng He, David Hurley
On-Site Speaker (Planned) Marat Khafizov
Abstract Scope Validation of first principles models for microstructure evolution under irradiation has been hindered by availability of table-top experimental methods suitable for defect characterization at the atomic level. We discuss extensive characterization of defects in ion irradiated fluorite oxides including urania, thoria and ceria. Irradiations over a range of temperatures to a few dpa dose provide optimal conditions for producing both point and extended defects. Electron microscopy is used to characterize extended defects, while optical and Raman spectroscopies are used to characterize point defects. Quantitative analysis of dislocation loops evolution using rate theory model provides means for measuring mobility of defects. Raman and optical spectroscopies probe vibrational and electronic properties of defects all tightly coupled to atomic structure of the defect. Finally, thermal conductivity measurements are used as ultimate test. Presented methods offer an attractive solution for validation of atomistic models critical for development of material performance codes with predictive capability.
Proceedings Inclusion? Planned:
Keywords Nuclear Materials, Ceramics, Characterization

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

Advanced Characterization and Modeling of Nanoprecipitates in Spent Nuclear Fuel
Atomic Scale Simulation of Amorphous Intergranular Films in Nuclear Fuel Materials
Atomistic-scale Simulations used to Simulate Creep in Oxide Fuel
Atomistic and Mesoscale Modeling of Fission Gas and Fission Products Diffusivity in TRISO Fuel Kernels
Atomistic Investigation of Radiation-induced Defects in ThO2
Cluster Dynamics Modeling of Defects and Fission Gas in Gd Doped UO2 under Irradiation
Comprehensive Characterization of Damage in Ion Irradiated Ceramics for Validation of Atomistic Models
Defect Chemistry and Radiation Stability of (Gd & Zr) Co-doped UO2
Development of UC/UO2 Composite Fuels for Light Water Reactors
Diffusion Properties in Uranium-plutonium Mixed Oxides: Atomic Scale Investigation of the Effect of Composition and Chemical Disorder
Emulation of Microstructures and Tritium Behavior in Lithium Aluminate by Ion Irradiation
Exploring Irradiation-induced Phase Evolution in WC
Hidden Defect Evolution Mechanism in ThO2 Revealed by Atomistic Modeling
High-entropy Carbide Ceramics: New Materials for Extreme Environments in Nuclear Energy Applications
Impact of Resonance Scattering on the Thermal Conductivity of ThO2
Improving Uranium Mononitride Behaviour using Uranium Diboride Addition
Irradiation- and Dopant-induced Structural Changes in Ceramic Nuclear Fuels Probed via Elastic and Optical Properties
Low-temperature Fabrication of Ceramic Tritium Breeder Materials, for Enhanced Control of Microstructure and Phase Formation
Microstructural Characterization of Neutron Irradiated Concrete Minerals
Microstructural, Mechanical and Thermal Characterization of High Entropy Carbide Ceramics
Modelling the Melting Temperature of CrUO4 to Assess its Behaviour during the Sintering of Cr-doped UO2
Multiphysics Modeling of High Burnup UO2 at Mesoscale
Multiscale Modeling for High-burnup Structure Formation in UO2
O-1: Uranium Silicide Processing for Advanced Reactors
O-31: Radiation Damage in Lithium Oxide, a Surrogate for Beryllium Carbide
Oxidation Behavior and Mechanisms of the SiC Coating in TRISO Fuel Particles
Phase Equilibria and Thermodynamics of Tri-carbide Fuels for Nuclear Thermal Propulsion
Predicting Mechanical Behavior of Uranium Oxide Fuel Pellets Using Full-field Defect Diffusion Modeling in a Crystal Plasticity Framework
Quantifying the Impact of Fast Interface Diffusion and Free Surface Evolution on Fission Gas Release in UO2 Using a Phase-field Model
Radiation Shielding Ceramics with Enhanced Performance and Scalability
Radiation Studies on the TiZrNbHfTa High Entropy Alloy and Its Hydrides
Relating Microstructural Evolution and Stoichiometry to Tritium Release from Ternary Lithium Ceramics
Revealing The Microstructure and Irradiation Effects on UO2 Fracture via Coupled Phase-field and MD Simulations Approach
Scanning Transmission Electron Microscopy of Nanoprecipitates in Spent UO2 Nuclear Fuel
Silica Formation on SiC Following Steam Attack
Simulation of Irradiation-induced Bubble Over-pressurization and Application in Fuel Performance
Soft X-ray Synchrotron Radiation Spectromicroscopy of Spent Nuclear Fuel Focused Ion Beam Sections
Surface Modification Strategies for Hydrogen Retention in Hydride Moderators
Susceptibility of Nuclear Fuel Ceramics to Oxidation and Hydridization during Off Normal Events
Thermomechanical Characterization of Advanced Reactor Alloys and Composites Exposed to High-temperature Gas Environments
Zirconia-coated Uranic Fuel Particles Processing and In Situ Sintering Characterisation

Questions about ProgramMaster? Contact programming@programmaster.org