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Meeting 2022 TMS Annual Meeting & Exhibition
Symposium Advanced Characterization and Modeling of Nuclear Fuels: Microstructure, Thermo-physical Properties
Presentation Title Modeling Irradiation-enhanced Diffusion in Advanced Ceramic Nuclear Fuels
Author(s) Michael Cooper, Christopher Matthews, Vancho Kocevski, Christopher Stanek, David Andersson
On-Site Speaker (Planned) Michael Cooper
Abstract Scope Nuclear fuel performance and the degradation of fuel properties are governed, in many respects, by the formation and diffusion of point defects and clusters in the lattice. For example, the diffusion of fission gas and vacancies through the lattice controls fission gas swelling and release, which are key performance metrics that also impact thermal transport. During reactor operation, the concentration and diffusion of defects can be enhanced through irradiation processes. In this work, we represent atomic scale calculations of the diffusion mechanisms of host and impurity (Xe) defects in ceramic nuclear fuels, such as doped UO2, UN, and UC. The atomic scale predictions of the stability and diffusivity of point defect and clusters in these systems have then been implemented in cluster dynamics simulations to predict irradiation-enhanced defect concentration and diffusivities. The importance of the in-reactor conditions and of including various defects will be discussed.
Proceedings Inclusion? Planned:
Keywords Nuclear Materials, Modeling and Simulation, Ceramics

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

A Combined Molecular and Cluster Dynamics Approach to Determine Radiation Enhanced Diffusion in UMo Alloys
A Monte-Carlo Solver for Coupled Electron-phonon Boltzmann Transport Equation in Metallic α-U
A Predictive Approach to Model Thermal Conductivity Degradation for In-pile UO2
Accelerating Nuclear Fuel Qualification through Multiscale Models
Adding Irradiation-assisted Grain Growth to the MARMOT Tool for UO2 Nuclear Fuel
Advanced Characterization of Oxidation Behavior of TRISO Fuel SiC Coating
An Atomistic Study of the Anisotropic Elastic Response of Defects in Alpha Uranium
An Atomistically-informed Cluster Dynamics Approach for Defect Evolution in ThO2 under Irradiation
An Integrated Approach for Coupling Experimental Data, Physics-based Models, and Machine Learning Algorithms for Predicting the Effective Thermal Conductivity of U-based Fuels
Bulk Thermal Conductivity Measurement of Fuels and Surrogates
Centipede: A New Tool for Calculating Irradiation Enhanced Transport of Defects in Nuclear Fuel
Comparison of Observations from the Microstructure of Two High Burnup Fuel Samples Operated at Different Linear Heat Generation Rates
Correlating Atomic Scale Microstructure with Mechanical Properties in Low-density Pyrocarbon Used in TRISO Particle Fuel Buffer Layer
Correlative APT-TEM Investigation of Defects’ Influence on Thermal Diffusivity in ThO2 Nuclear Fuel
Diffusion Coefficients of Zr- and Cr-based Binary Systems for Simulation of Cr-coated Zircaloy Nuclear Fuel Cladding
Evolution of the Internal Layer Structure in Irradiated TRISO Fuel
Experimentally Validated Model for Investigating High-burnup Structure Formation in U-Mo Fuels
High-resolution Thermal Conductivity and Thermal Boundary Resistance Mapping in TRISO
High-throughput Viscosity Measurements of Molten Salts for Molten Salt Reactors
Investigation of Damage Structure Evolution on Proton Irradiated Zr-alloys of Various Compositions Using Synchrotron X-ray Diffraction and TEM
Investigation of Hot-cell Capable Thermal Conductivity Measurements for Ceramic Fuels
Mesoscale Hybrid Model for Fission Gas Behavior in UO2: Coupling the Phase Field Method to Spatially Resolved Cluster Dynamics
Mesoscale Model of Gas Bubble Evolution and Creep in Monolithic UMo Fuels
Mesoscale Modeling of Effective Thermal Conductivity in U-Zr Fuels with Heterogeneous Phases
Micromechanical Behavior of Thermally Loaded Monoclinic U-6Nb
Microstructural Characterization of the Porous Pyrocarbon Buffer Layer in TRISO Fuel Particles
Modeling Irradiation-enhanced Diffusion in Advanced Ceramic Nuclear Fuels
Multiphysics Modeling of Fracture in Sintered Uranium Dioxide Pellets
N-1: 3D Reconstruction and Quantification of Oxide Nano-porosity in Zirconium Alloys
N-2: An Experimentally Validated Mesoscale Model for the Effective Thermal Conductivity of UZr Fuels
N-4: Atomistic Modeling of Transport Properties and Interaction with Point Defects of α-U Tilt Grain Boundaries
N-5: Characterization of Additively Manufactured UO2 Fuel Pellets with Pulsed Neutron Techniques and 450 keV X-ray CT
N-6: Characterization of Nuclear Materials from the Millimeter to the Nanometer
N-7: Investigation of the Impact the 3D Fission Product Structure has on the Local Thermal Conductivity in FBR MOX Fuel
On the Phases Observed in Irradiated U-19Pu-14Zr Fuels
Perspectives on Synchrotron Micro-computed Tomography and Serial Sectioning Applied to Metallic Nuclear Fuels
Predicting Thermophysical Properties of Actinide Oxides Using Atomic Scale Simulation
Propose Advanced Nuclear Fuels with High Thermal Conductivity Using Machine Learning
Pulsed Neutron Characterization of Irradiated Fuels at LANSCE
Structural Analysis of the IPyC/SiC Interface of AGR-2 Irradiated and Safety Tested TRISO Fuel
Synergistic Electron/Thermal Microscope for High-throughput Screening of Emerging Nuclear Materials
The Effect of the Proton Irradiation Dose Rate on the Evolution of Microstructure in Zr Alloys: A Synchrotron Micro-beam X-ray and TEM Study
The Influence of Radiation-induced Microstructural Defects on the Optical and Elastic Properties of Ceramic Nuclear Fuels
Thermal Conductivity Degradation by Solid Fission Products: Machine Learning Coupled with First Principles Model
Thermal Conductivity Measurement of Microstructures in Irradiated Nuclear Fuels
Thermal Energy Transport in Defect-bearing and Uranium-doped Single Crystal Thorium Dioxide
Thermal Stability of Uranium Compounds and Advanced Nuclear Materials under Extreme Conditions
Thermal Transport Behavior of Pristine and Zirconium-doped Alpha-Uranium
Thermo-physical Properties of the Ternary (U2Cr)N3 Phase
Thermophysical Properties of Liquid Chlorides from 600-1600 K

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