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Meeting Materials Science & Technology 2019
Symposium Materials for Nuclear Applications
Presentation Title P3-67: Investigation of Recrystallization and Grain Growth in Uranium 10 wt% Molybdenum
Author(s) Jacqueline I. Reeve, Vineet Joshi, David Field
On-Site Speaker (Planned) Jacqueline I. Reeve
Abstract Scope Uranium 10 wt% Molybdenum (U-10Mo) is being explored as a replacement for highly enriched uranium (HEU) fuels used in advanced research and test reactors. In this study, the texture and grain growth kinetics of depleted U-10Mo are analyzed. The U-10Mo microstructure and transitions that occur during recrystallization and grain growth are important due to their influence on the fuels performance and swelling kinetics in reactor operation. U-10Mo samples as hot rolled (80% total reduction) and as cold rolled (10%, 30%, and 50% total reduction) were annealed at 600˚C and 700˚C for time periods ranging from 1-8 hours. Electron Backscattered Diffraction (EBSD) was employed to observe and quantify recrystallization and grain growth in the samples. The effects of plastic deformation and annealing temperature control the nucleation and growth rates for recrystallization. Optimum conditions are identified to obtain specific grain size distributions.


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Anisotropic Thermal Transport in Uranium Dioxide Induced by Dislocation
Anisotropy in Thermal Creep and Creep Life Prediction of Zr-2.5%Nb Pressure Tube Alloy
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Combined Use of In-situ and Ex-situ TEM to Characterize Irradiation Induced Dislocation Loops in F/M Steels for Nuclear Applications
Comparison of In-situ Micro- and Ex-situ Meso-scale Tensile Testing for the Evaluation of Mechanical Properties of Stainless Steels
Compatibility of U3Si2 fuel with FeCrAl and SiC/SiC Based Cladding
Computational Studies of Environmental Degradation of Silicon Carbide
Design of Alloy Chemistry to Mitigate Fuel-Cladding Chemical Interactions in Uranium-based Metallic Fuels
Development and Performance of High Temperature Irradiation Resistant Thermocouples
Effect of Ultrasonic Nanocrystalline Surface Modification (UNSM) on the Oxidation Behavior of Alloy 800HT in a Supercritical Carbon Dioxide (SCO2) Environment
Effects of Ti and Al Additions on Irradiation Behavior of FeMnNiCr Based High-Entropy Alloys
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Evolution of Microstructure, Deformation Mechanisms, and Internal Damage During Creep-Fatigue Testing of Alloy 709 (Fe-20Cr-25Ni)
Exploring TRISO Layer Properties and Performance for Multiple Reactor Concepts
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In-situ Characterization of Zirconium Alloy Degradation to Support Nuclear Sensing Applications
In-situ Ion Irradiation Study of Silicon Carbide-Carbon Coated Nanostructured Ferritic Alloy
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Multi-dimensional, In-situ Mechanical Testing to Evaluate Damage and Fracture of Chromium-Coated Zirconium-based Fuel Claddings
P3-62: Characterization and Oxidation of Graphite and Silicon Carbide in TRISO Nuclear Fuel
P3-63: Corrosion Behavior of Metallic Alloys in a Molten Chloride Eutectic Salt for Nuclear Applications
P3-64: Corrosion Behavior of Nanostructured Stainless Steels and High Entropy Alloys
P3-65: Effects of Deposition Conditions on the Production of ZrO2 Coatings Produced by PE-CVD as Environmental Barrier Coatings for the Molten Salt Reactor
P3-67: Investigation of Recrystallization and Grain Growth in Uranium 10 wt% Molybdenum
P3-68: Oxidation of TRISO particles and Matrix Graphite in Mixed Gas Atmospheres
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Transmutation-induced Precipitation in Tungsten Irradiated with a Mixed Energy Neutron Spectrum
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