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Meeting 2023 TMS Annual Meeting & Exhibition
Symposium Microstructural, Mechanical and Chemical Behavior of Solid Nuclear Fuel and Fuel-cladding Interface
Sponsorship TMS Structural Materials Division
TMS: Advanced Characterization, Testing, and Simulation Committee
TMS: Nuclear Materials Committee
Organizer(s) Yi Xie, Purdue University
Miaomiao Jin, Pennsylvania State University
Jason M. Harp, Oak Ridge National Laboratory
Fabiola Cappia, Idaho National Laboratory
Jennifer K. Watkins, Idaho National Laboratory
Michael R. Tonks, University of Florida
Scope Fuel and fuel-cladding interaction constitute the key to understanding fuel performance. The combined effect of microstructural evolution and chemical change cause loss of performance in various forms such as embrittlement, deformation, phase instability, etc, which need to be well understood to enable materials evaluation and prediction in normal and accident scenarios. The designs of current and next-generation reactors are varied. The types of fuels include ceramic, metal, and composite fuels, including UO2-, UN-, U3Si2-, U-Zr-, U-Mo-based, and TRISO. There are also multiple types of cladding materials in use/consideration, e.g., zircoloy, stainless steels, SiC/SiC composite, ODS, HEA concepts and coating, made by various advanced manufacturing methods. In particular, the confounding factors from chemically active fission products (lanthanides, tellurium, etc.) and impurities (oxygen, carbon, etc.) can complicate the fuel performance analysis, due to the changes in fuel and fuel-cladding interaction. This symposium aims to bring together experimental and computational investigations that assist in understanding the microstructural, mechanical, and chemical changes in these solid fuels and fuel-cladding interfaces. Both the synergistic and separate effects of involved physical processes, with fresh or certain burn-up or surrogate fuels, are of interest. Analysis of advanced fuel types and cladding concepts are strongly encouraged. This symposium also calls for multi-scale modeling and simulations and fuel performance modeling.

The topics of interest include experimental and modeling efforts in the following aspects, but not limited to:

• Evolution of defects, microstructure, and phase in fuels or fuel surrogates
• Impact of impurities on microstructure and phase transformation of fuels or fuel surrogates
• Behaviors of fission products in fuels and/or cladding
• Fuel-cladding mechanical and chemical interactions
• Advanced manufacturing and characterization technologies for nuclear fuels

Abstracts Due 07/17/2022
Proceedings Plan Planned:

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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