ProgramMaster Logo
Conference Tools for 2018 TMS Annual Meeting & Exhibition
Register as a New User
Submit An Abstract
Propose A Symposium
Presenter/Author Tools
Organizer/Editor Tools
About this Abstract
Meeting 2018 TMS Annual Meeting & Exhibition
Symposium Materials and Fuels for the Current and Advanced Nuclear Reactors VII
Presentation Title Importance of the Amount of Lubricant on the Die Wall during UO2 Powder Cold Compaction
Author(s) Ousse´ni Marou Alzouma, Anne-Charlotte Robisson
On-Site Speaker (Planned) Ousse´ni Marou Alzouma
Abstract Scope External lubrication is often used to complete compaction process of powder materials. The main goal of this method is to reduce the amount of admixed internal lubricant (Zinc stearate spray) within the raw material. The application of external lubricants enhances the density uniformity and the mechanical strength of the resulting green pellets. This study investigates the effects of the external lubricant amount for UO2 powder compaction and the properties of the corresponding green pellets. Results show that there is a quantity from which, the external lubricant on the die wall becomes detrimental to the friction index, and the ejection force measured during the pressing cycle. The quality of the green pellets (surface defects, mechanical resistance) can also be affected by a significant amount of lubricant. Thus the quantity of the die wall lubricant must be optimized in order to assure a mixed lubrication mode.
Proceedings Inclusion? Planned: Supplemental Proceedings volume


3D Characterization of High Fluence Irradiated UZr and UMo Fuels
A New Physically Based, Quantitatively Predictive Low Flux-high Fluence Model of RPV Embrittlement
A Study on the High Energy Ball Milling and Spark Plasma Sintering of Fe-Cr Based Alloys
A Unified Model for Irradiation Creep and Stress-free Growth in Zirconium Alloys
Additive Stainless Steel for Nuclear: From Material Aspects to Quality Part
Advanced Characterization of Irradiated UO2 Fuel
Advanced Manufacturing of HT9 Steel for Extreme Environments
Ar Bubble Effects on Precipitation Reactions in Solubilized AISI 316L Steel Irradiated with Heavy Ions
Asymptotic Expansion Homogenization of the Stiffness Tensor and Thermal Conductivity of a 2D Exemplar-guided Digital Reconstruction of an Al3Hf-Al Microstructure with Comparison to Experiment
Atom Probe Examinations of Zircaloy Irradiated at 358-410C
Atom Probe Tomography Study of Microstructural Evolution of Cast Duplex Stainless Steels after 10,000 Hour Thermal Aging
Characterization of Intragranular Creep Deformation in Uranium Dioxide Using Electron Backscatter Diffraction and Electron Channeling Contrast Imaging
Characterization of Stress and Microstructure of Zr-4 alloy Processed by Pulsed Laser
Characterization of the Microstructure and Grain Boundary Character of 14-YWT Nanostructured Ferritic Alloys Following Different Deformation Processing Paths
Characterization of U-Zr-RE Metallic Fuel Fabricated by Injection Casting
Characterizing and Modelling Precipitation in Zirconium Alloys
Chemical and Microstructural Analysis of Irradiated Mixed Oxide Fuels
Chemical Compatibility of Refractory Carbides with Hydrogen at Very High Temperatures Relevant for Nuclear Thermal Propulsion Applications
Corrosion Assessment of an Alloy/Oxide Composite Using Electrochemical Techniques
Corrosion of SiC with Cr, CrN, and TiN Coatings in High Temperature Water
Creep Related Microstructural Evolution of Alloy 617-based ODS Alloy
Development of an Alternative Manufacturing Process for U3Si2 Fuel by a Novel Additive Manufacturing Process
Development of Laves and B2 Manipulated Advanced Ferritic Alloys
Development of YSZ Environmental Barrier Coatings for the Molten Salt Fast Reactor
Dislocation Dynamics of Alloys for High Temperature Nuclear Reactors
Dual Ion Beam Irradiation of Commercial-grade Austenitic Alloys Relevant to LWR Core Components at High Dose
Dynamic Strain Aging in Alloy 709 (Fe-25Ni-20Cr)
Effect of 0.25 and 2.0 MeV He-ion Irradiation on Cr Atoms Distribution in Model Fe-Cr Alloys
Effect of C and Si Impurities in U10Mo Alloy: Discovery of New Quaternary Si-rich Phase and its Influence on Transformation Kinetics
Effect of Cold Working on the Corrosion and Carburization Behavior of Alloy 800HT in High Temperature CO2 Environment
Effect of Grain Elastic Anisotropy on Stress Intensification at Intergranular Stress Corrosion Cracking Iinitiation Sites in Austenitic Stainless Steels and Nickel-based Alloys in Light Water Reactor Environment.
Effect of Neutron Irradiation on the Mechanical Properties and Microstructure of Friction Stir Processed ODS Alloys
Effect of Pd on the Grain Boundary Character Distribution in SiC
Effect of Thermomechanical Processing on the Microstructure of U-9Zr-3Nb and U-9Zr-3Mo Alloys
Effects of Proton Irradiation on Microstructure in Additively Manufactured 316L Stainless Steel Made by Laser Powder Bed Fusion
Electron Microscopy Analysis of TRISO Fuel Particles with Failed SiC Layers from the AGR-2 Irradiation
Examining the Effects of Neutron Irradiation on Zirconium-alloy Oxide Film Microstructure Using Focused Ion Beam Techniques
Experimental Studies on Microstructure and Mechanical Properties of High Burnup Urania
Fabrication of Lumped Gd2O3 Inserted Oxide Pellets ‎for Burnable Absorber Fuel
Fabrication of ODS FeCrAl Tube for Accident Tolerant Fuel Cladding Applications
Fabrication of PyC/SiC Diffusion Couples Using Fluidized Bed CVD Techniques for Radiation Enhanced Diffusion Testing
Grain Boundary Engineering for Improved Resistance to Corrosion and Stress Corrosion Cracking Resistance of Nuclear Alloys
High-temperature Mechanical Properties of Zirconium Hydrides Studied with Nanoindentation
High Temperature Strength Characterization of Alloy 709
Impact of Low Dose Ion Irradiation on Raman Spectra and Thermal Conductivity in 3C-SiC
Implementation and Validation of a Physically-based Fuel Cladding Oxidation Model in BISON Nuclear Fuel Performance Code
Implementation of Viscoplastic Model to Predict the Failure in Zircaloy-4 due to Pellet Cladding Mechanical Interaction (PCMI)
Importance of the Amount of Lubricant on the Die Wall during UO2 Powder Cold Compaction
Improvements and Applications of the FAST Fuel Model to Thorium-based and Mixed Oxide Fuels
In-situ Characterization of Dispersoid Evolution during Annealing of ODS FeCrAl Mechanical Alloyed Powders
In-situ Elevated Temperature Micro-cantilever Testing of UO2
In-situ TEM Observation and MD Simulation of the Radiation Defects near Carbon Nanotube in Aluminum
In-situ Testing of Fouling-resistant Coatings for PWR Fuel Cladding
In Situ EBSD Analysis of Deformation Mechanisms in Highly Irradiated Austenitic Steels
In Situ Neutron Diffraction Analysis of Strain-induced Processes in 10.7-dpa Irradiated AISI 304L Steel
Investigation of the Role of Cr and Cr Carbides at Grain Boundaries in Alloy 600 for Stress Corrosion Cracking
Investigation of Tin as a Fuel Additive to Control FCCI
Investigation on the Damage Mechanism of Plasma-materials Interface by Multi-scale Electron Microscopy Methods
Ion Irradiation Effects on the Structure and Thermal Properties of Zirconium Diboride
Irradiation Effects on Fe-9%Cr Grain Boundary Strength Measured via In-situ TEM Testing
Irradiation of Additively Manufactured Grade 91 Ferritic/Martensitic Steel
Isothermal Transformation Kinetics ofáγ phase fromáα+γ'áPhase Mixture in U-10wt.%Mo Alloys
Kinetic Evolution of Transmutation Helium Accumulation at Y-Ti-O Oxides in Nanostructured Ferritic Alloys under Irradiation
Low Cycle Fatigue Resistance of Zircaloy-4 under Uniaxial and Torsion Loading
Mechanical Property Measurements of Zicaloy Hydride Structure by Using Nanoindentation and Nano Mechanical Raman Spectroscopy
Microstructural and Mechanical Integrity of Laser Weldment of Neutron Irradiated AISI 304 SS
Microstructural Characterization of Plutonium Based Fuels
Microstructural Characterization of U-Mo Fuel Plates Irradiated in the Advanced Test Reactor: Recent Observations
Microstructure Based Hardening Models for Alloys Irradiated with Charged Particles an in the ATR and BOR60 Reactors
Microstructure Evolution in Neutron Irradiated and Ion Irradiated Alloy T91
Multiscale Irradiation Effects of Tungsten Based Materials for Nuclear Power
Neutron irradiation Induced Microstructures in Ferritic/Martensitic Steel HT9
New Insights on Denuded Zone Formation in Polycrystalline Materials
Non-destructive 3D Neutron Imaging of Composition in Nuclear Fuels
Phase Transformation Kinetics in Rolled U-10 wt. % Mo Foil: Effect of Post-rolling Heat Treatment and Prior γ-UMo Grain Size
Phase Verification and Thermophysical Properties of Pu-Zr Alloys
Post-irradiation Examination (PIE) of Irradiated Hafnium
Probing Local Disorder in Ln-UO2 (Ln = Y, Nd, La) and UO2+x Systems
Progress in Developing High Dose Radiation Tolerant Ferritic Steels for Nuclear Applications
Radiation Effect on Nanomaterials at High Temperature -New Type of Radiation Detector for TREAT Nuclear Reactor-
Radiation Effects on HT9 Tempered Martensitic Steels as a Function of Initial Dislocation Density
Radiation Resistant Elemental Combination High Entropy Complex Concentrated Alloys for Nuclear Applications
Radiation Response of Nanoporous Metals
Radiation Tolerance of Equiatomic Multicomponent Single Phase Alloys Subjected to Ion Irradiation at 16 K
Recrystallization of a Nanostructured Ferritic Alloy after Cold Work
Shear Punch Measurement of the Mechanical Properties of Irradiated Cladding Material from ATR Irradiations
The ATR-2 RPV Steel Irradiation Hardening Data Base: An Overview and Some Major Findings
The Effect of Dpa and Dpa Rate on the Strength and Precipitates Stability in Ion-irradiated Inconel 718
The Elastic Constants of γ-phase U – 8 wt% Mo between 25-650░C via Resonant Ultrasound Spectroscopy
The Role of Grain Boundaries and Second Phase Particles in Predicting Nucleation Sites for Hydride Precipitation in Zirconium Alloys
Ti effect on Microstructrue Stability and Mechanical Properties in Reduced Activation Ferritic-martensitic Steel
Understanding Micromechanical Deformation in Hard-facing Alloys for Improving Galling Resistance
Using Synchrotron X-ray Diffraction and Transmission Electron Microscopy to Study the Dislocation Structures Found in Proton Irradiated Zr-Nb Alloys
Very High Temperature Steam Oxidation of LWR FeCrAl Fuel Cladding

Questions about ProgramMaster? Contact