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Meeting 2023 TMS Annual Meeting & Exhibition
Symposium Microstructural, Mechanical and Chemical Behavior of Solid Nuclear Fuel and Fuel-cladding Interface
Presentation Title
Author(s)
On-Site Speaker (Planned)
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Proceedings Inclusion? Planned:
Keywords Nuclear Materials, Ceramics, Process Technology

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

Accelerating the Qualification of Nuclear Fuels Through Advanced Characterization and Multiscale Modeling
Advanced Characterization of Fuel-cladding Chemical Interaction in HT9 Clad U-Mo-Ti-Zr Metallic Fuel Irradiated in Advanced Test Reactor
Analysis of Secondary Phase Formation at U-10Mo Fuel/Cladding Interfaces During Manufacturing
Assessing the Influence of Microstructure on Uranium Hydride Size Distributions via Small Angle Neutron Scattering
Assessment of High-density Fuels During Hydrogen Interaction
Atomistic Simulations of Silicon Carbide Layer in Tristructural Isotropic Fuel Particles
Atomistically-informed Cluster Dynamics Modelling of Defect Evolution in Irradiated ThO2
Characterization of Crystal Structure Evolution in U-2wt.%Zr Using Neutron Diffraction with Particular Focus on the Beta-Uranium Phase
Characterization of the Radial Microstructural Evolution in Commercial LWR UO2 with Different Power Histories
Characterization of U-10Mo Fuel Exposed to Intermediate Temperature Irradiation Conditions at the High Flux Isotope Reactor
Chemical Interaction and Compatibility of Uranium Nitride and Alumina Forming Austenitic Alloys
Chemical Structures and Thermodynamics of Uranium Nitride and Uranium Carbide
Comparing the Impact of Thermal Stresses and Bubble Pressure on Intergranular Fracture in UO2 Using 2D Phase Field Fracture Simulations
Cr-doped UO2 Studied Using XAS and Neutron Scattering
Creep Testing of 70% Theoretical Density U10Zr
Determination of the Hydrogen Heat of Transport in Zircaloy-4
Diffusion Coefficients of Zr- and Cr-based Binary Systems for Simulation of Cr-coated Zircaloy Nuclear Fuel Cladding
Extended Defect Coalescence in Kr Irradiated UO2 During High Temperature Annealing
Fabrication and Characterization of Uranium Carbide
Fracturing and Fragmentation of Cr2O3-doped UO2 Pellets with Controlled Microstructure Under Prototypic LOCA and RIA Thermal Transients
Fuel Performance Analysis of an Annular Type Metallic U-10Zr Fuel
High Resolution Microscopic Studies on HT-9 Cladding from U-10Zr Fuel Irradiated at Fast Flux Test Facility
High Temperature Steam Oxidation Performance of Alloyed, High Density Fuel Composite: U3Si2 + 50wt% UB2
Improved Model of Microcracking Behavior in High Burnup UO2 Fuel
In Situ EBSD Studies of Blocky Grain Growth in Welded Zircaloy-4
Interconnectivity Quantification and Corrosion Mechanisms in Zr Alloys
Interfacial Microstructure Evolution in Al6061-Al6061 HIP Bonded Samples for Cladding Applications on U-10Mo Monolithic Fuel
Irradiation Performance of Densely Packed UN TRISO Fuel in a 3D-Printed SiC Matrix
Lower Length Scale Fuel Performance Modeling of U-Mo Fuel
Magnetism and Finite Temperature Effects in δ-UZr2: A Density Functional Theory Analysis
Microstructural Characterization of the SiO2-SiC Interface of Oxidized TRISO Particles
Microstructure and Phase Evolutions of U-Zr System in Thermal Cycling Neutron Diffraction Experiments
Modeling Fission Gas Release Behavior from Microcracking and Thermal Diffusion at High Burnup in UO2 Fuel in BISON
Modeling Low-temperature Hydrided Zircaloy Cladding Failure Under a Reactivity-initiated Accident
Modeling of Fission Gas Behavior in Uranium Nitride Fuel
Modeling Stoichiometry Controlled Defect Dependent Densification in UO2±X
Molecular Dynamics Based Microstructural Evaluation of the Surviving Defects in α-U Induced by a Single Displacement Cascade
Molecular Dynamics Study of the Anisotropic Elastic Response of Defects in Alpha-Uranium
Multiphysics Modeling of Nuclear Fuels at the Mesoscale
Nuclear Fuels and Interfaces for Advanced Specialty Microreactors
Numerical Modeling of AA6061 Cladding Diffusion Bonding for the U-10Mo Monolithic Fuel
O-14: Calculation of Grain Boundary Diffusion Coefficients in Gamma U-Mo Using Atomistic Simulations
O-15: Experimental Methods for Comprehensive PIE of Test Fuel Rods
Observations from Microscopy on High Burnup Light Water Reactor Fuel Before and After LOCA Testing
Performance of FeCrAl Alloys Under Long-term Graphite Exposure
Scaling laws in nanoindentation investigation of metallic uranium alloys
Small-scale Mechanics Quantification of UO2 Fracture Toughness
Spark Plasma Sintering – Innovative Approaches for High Temperature Creep Testing and Transient Behavior of Nuclear Fuels
The Evolution of the Microstructure of Low-enriched Uranium Fuels During Irradiation in the Advanced Test Reactor
The Fabrication, Advanced Characterization, Advanced Test Reactor Irradiation, Post Irradiation Examination, and Materials Informatics for Annular U-10Zr Metallic Fuel
Thermochemical Investigation of Advanced Reactor Fuels and Fuel-clad Chemical Interaction
Uncertainty Quantification of Thermal Performance of UO2 Fuel Pellets
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