ProgramMaster Logo
Conference Tools for 2016 TMS Annual Meeting & Exhibition
Login
Register as a New User
Help
Submit An Abstract
Propose A Symposium
Presenter/Author Tools
Organizer/Editor Tools
About this Abstract
Meeting 2016 TMS Annual Meeting & Exhibition
Symposium Materials and Fuels for the Current and Advanced Nuclear Reactors V
Presentation Title Additive Manufacturing of Uranium-6 Wt. Pct. Niobium
Author(s) Amanda Wu, Gilbert F. Gallegos, Matthew W. Wraith, Stephen C. Burke, Donald W Brown
On-Site Speaker (Planned) Amanda Wu
Abstract Scope Uranium-niobium alloys require specific melting and forging processes to achieve reasonable homogeneity of the alloying elements and to overcome shape memory effects and complex phase transformations, respectively. The challenges and expenses associated with these materials processes have driven our exploration of alternative manufacturing methods capable of producing geometrically complex parts while minimizing processing time and waste of this limited material. Powder bed fusion additive manufacturing introduces significant processing and materials challenges toward achieving structural integrity. Broadly addressed aspects affecting powder bed fusion structures include surface roughness, residual stresses, and porosity. Here, we focus on materials-specific challenges—namely, the influence of multiple pass reheating during processing and post-processing homogenization on phase transformations and the role of impurities and oxides on grain morphology and mechanical behavior—with a view of adopting powder bed fusion processing for the shape memory alloy, uranium-6 wt% niobium.
Proceedings Inclusion? Planned: A print-only volume

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

3D Microstructural Characterization of UO2+x Using High-energy X-rays
AA-10: Effects of β-decay on Ceramic Nuclear Waste Forms
AA-11: Low Temperature Friction Stir Welding (FSW) of Cr-Mo Steels
AA-12: Effects of Irradiation on the Interfacial Reaction between SiC and ODS Steels
AA-1: Influence of Zirconium Hydride on the Biaxial Thermal Creep Behavior of Zircaloy-4 Cladding at 573 K and 773 K
AA-2: Fractography of Neutron-irradiated Alloy 690
AA-3: Fabrication of Interconnected SiC Reinforced ZrO2 Composites by the Coat-mix Process and Spark Plasma Sintering
AA-4: Formation of Silicide Coatings on Refractory Alloy Substrates for Accident Resistant Nuclear Fuel Cladding
AA-5: Thermal and Mechanical Properties of Bulk Fe2B
AA-6: Thermodynamic Assessment of U-Eu-O System
AA-7: Thermophysical Properties of Molten Zr-Ni Alloys Measured by Electrostatic Levitation
AA-8: A Study on the Diffusion of Volatile Fission Products in the Graphite Matrix of HTGR
Additive Manufacturing of Uranium-6 Wt. Pct. Niobium
Advanced Fuels by Field Assisted Sintering Technology – Fuel Properties Characterization and Accident Tolerance
Advanced Nuclear Fuels and Materials Development and Philosophy of the DOE Advanced Fuels Campaign
Al-Ti-Cr Coating on Zr Alloys for Enhancing Accident Tolerance of Fuel Claddings
Atom Probe Examinations of Zircaloy Irradiated at Nominally 358C
Atomic-level Characterization of the Metal-oxide Interface of a Zircaloy-4 Cladding from Commercial LWR Irradiated Fuel
BISON Fuel Performance Code Examination of Coating/Clad Interfaces for Accident Tolerant Fuels Irradiation Testing
Bulk Extraction and XAS Characterization of Oxides in Nanostructured Ferritic Alloy MA957
Characterization of High Burnup Structure in LWR Irradiated Urania
Characterization of Thermal Aging Embrittlement of Cast Duplex Stainless Steels by Mechanical Testing and FEM Modeling
Characterization via Transmission Electron Microscopy of the Diffusional Interactions between U-10Mo and AA6061 Alloys at 600C
Chemical Dependence of the Amorphization Behavior of the UMo-Al Interaction Layer in Dispersion Fuels
Comparison of Zirconium Oxidation Behavior under Oxygen-rich Gaseous and High Humidity Environments via In-situ TEM
Comprehensive EBSD Analysis of the SiC Layer from AGR-1 and AGR-2 Constituent TRISO Fuel Batches
Correlation of Fission Product Transport to Grain Boundary Character in Neutron Irradiated Tristructural Isotropic Coated Nuclear Fuel Particles
Corrosion Studies on U-10Mo Fuel for Research Reactor Applications
CRUD Mitigation And Growth
Deposition of Compatibility Films on SiC for Environmental Barrier Coatings
Design of Radiation Tolerant Materials via Interface Engineering
Development of Engineering Parameters for Low Pressure Diffusion Bonds of 316 SS Tube-to-Tube Sheet Joints for FHR Heat Exchangers
Development of Fe-12Cr-5.6Al ODS Alloys for Nuclear Applications
Development of ODS FeCrAl Alloys for Accident-tolerant Fuel Cladding
Down Selection of Clad Material for LEU Fuel Elements for the TREAT Reactor
Effect of Cold Rolling on the Integrity and SCC Susceptibility of Twin Boundaries of Alloy 690
Effect of Heat Treatment and Chemical Composition on the Precipitation Behavior in Commercialized Age Hardening Nickel Based Alloys
Fabrication and Qualification of Small Scale Irradiation Experiments in Support of the Accident Tolerant Fuels Program
Fabrication of Graphite Composite Fuel with Controlled Thermal Transport Properties
Fabrication of Mock Up LEU Fuel Elements for the TREAT Reactor
Ferritic Steels Cladding for Accident Tolerant Fuel in Light Water Power Reactors
Fracture Criteria for Liquid Sodium Embrittlement in T91 Martensitic Steel
Fuel and Materials Development, Testing and Qualification for the Traveling Wave Reactor
Grain Orientation Factor and Stress Corrosion Crack Initiation in Neutron-irradiated Austenitic Stainless Steels
Helium Behavior after Thermal Treatment in V and Fe-based Systems
High Resolution Electron Microscopy Examination of Fission Product Precipitates in Triso Coated Particles
In-pile Creep of High Purity SiC and Selected FeCrAl Alloys
Interdiffusion and Reaction between Al vs. X (X = Zr, Mo, U) Diffusion Couples
Investigation of Thermal Conductivity in Ion Irradiated Samples Using Laser Based Thermoreflectance Methods
Irradiation-induced Microstructure of Precipitate Hardened Nickel Based Alloy
Irradiation Memory Effects in Zirconium Alloy Corrosion
Laser Shock Peening of Oxide-Dispersion-Strengthened Austenitic Stainless Steels
Mechanical and Microstructural Characterization of Some High Fluence Intermediate Flux Neutron Irradiated Reactor Pressure Vessel Steels
Mechanical Properties of Materials and Phases Relevant to Monolithic U-Mo Fuel System
Mechanical Testing of UO2 Fuel at Elevated Temperatures
Microstructural Characterization of Creep-Fatigue Interactions in 9Cr-1MoV Steel and Welds
Microstructural Development and Phase Transformations in Hot Isostatic Pressed Monolithic U-Mo Fuel Plates in AA6061 Cladding with Zr Diffusion Barrier
Microstructural Investigation of TREAT Graphite Fuel Blocks
Microstructure-based Finite Element Analysis of the Effect of Homogenization on the U-10Mo/Zr Interface
Microstructure and Phase Stability of Oxide Dispersion Strengthened Steels
Microstructure Characterization of P91 and P92 Steels and Weld Metals
Microstructure Characterization of TRISO Fuels by Atom Probe Tomography
Migration of Lanthanides in U-Zr Alloy Fuel under a Thermal Gradient
Miniature Bulge Test for Measuring HIPed Aluminum/Aluminum and Aluminum/Uranium Interfacial Fracture Toughness
Modeling Solute Segregation during Solidification of U-Mo Alloys
Nanostructured Vanadium Carbide Coating on the F/M Stainless Steel for Mitigating Fuel Cladding Chemical Interaction
Oxidation Behavior of Accident-Tolerant FeCrAl Cladding Alloys
Phase-Specific Nanoindentation of Wear-Resistant Alloys for Nuclear Power Plant Applications
Precipitation in 316 Stainless Steels under Irradiation in Light Water Reactor Conditions
Processbility Assessment of Accident-Tolerant FeCrAl Cladding Alloys
Recent Results of Microstructural Characterization of U-10Mo Monolithic Fuel Plates Irradiated in the Advanced Test Reactor
Recrystallization and Texture Development in Rolled U-10 wt% Mo Alloys
SiC/SiC Composites for Current and Advanced Reactors
Simulation of Hafnium-Aluminum Thermal Neutron Absorber Material
Solid-state Diffusion Bonding of Ni-base Hastelloy-X
Study of Microstructural Evaluation and Thermal Creep Behavior of Heat-Treated Zr-Excel Pressure Tube Materials
Swift Heavy Ion Irradiation Induced Interactions in the UMo/X/Al Trilayer System
Synchrotron Characterization of Fission Products in the SiC Containment Layer in High Burnup TRISO Fuel Particles
Synchrotron Characterization of Oxidation in Nuclear Claddings for LWR Applications
Synthesis and Characterization of Magnetron Sputtered Cr2AlC Coatings to Improve Oxidation Resistance of Zirconium Alloys
TEM Investigation of Phases Formed in Ternary U-Pu-Zr Systems
TEM Study of Damaged Archive and Irradiated SUPERFACT Fuels
Temperature Effect of Microstructural Evolution in Advanced Nanostructured Alloys by in-situ Synchrotron X-ray Diffraction
Texturing, Microcracking and Delamination in 14YWT Nanostructured Ferritic Alloys
The Effect of Grain Size on the Homogenization Kinetics and Eutectoid Decomposition in U-10 wt% Mo Alloys
The Status of a Quantitative Multiscale Master Model of Helium-Displacement Damage Interaction Effects on Cavity Evolution in Fusion Structural Alloys
The Thermal Properties of Fresh and Spent U-Mo Fuels: An Overview
Thermal Conductivity of High Plutonium Content MOX Fuels
Thermal Desorption Spectroscopy of High Fluence Irradiated Ultrafine and Nanocrystalline Tungsten: Helium Trapping and Desorption Correlated with Morphology
Thermal Expansion of a 3-phase Ceramic Composite: An In-situ High Temperature X-ray Diffraction Study
Thermal Stability of Nanoscale Hardening Features in Irradiated Reactor Pressure Vessel Steels
Thermomechanical Modeling of Triso Fuel Particles Silicon Carbide Matrix
Thermomechanical Processing and Microstructural Evolution of Alloy 690, and Its Effects on Stress Corrosion Cracking
TRISO Coating Development for Uranium Nitride Kernels

Questions about ProgramMaster? Contact programming@programmaster.org