ProgramMaster Logo
Conference Tools for Materials Science & Technology 2020
Login
Register as a New User
Help
Submit An Abstract
Propose A Symposium
Presenter/Author Tools
Organizer/Editor Tools
About this Abstract
Meeting Materials Science & Technology 2020
Symposium Advanced Characterization of Materials for Nuclear, Radiation, and Extreme Environments
Presentation Title Investigation of Uranium Silicide Fuel Form Additions through Rietveld Refinement and Internal Standard P-XRD
Author(s) Zachary Acosta, Cole Moczygemba, Elizabeth Sooby Wood
On-Site Speaker (Planned) Zachary Acosta
Abstract Scope Uranium oxides have historically been utilized to power nuclear reactors due to their resistance to further oxidation and ease of large scale fabrication. However, other fuel forms may make more ideal candidates. U3Si2 has a higher thermal conductivity in addition to having a higher uranium density, making it more economical and less prone to cracking during thermal cycling. Interstitial additions may be introduced into the uranium silicide fuel form to boost lacking properties. Zirconium has been hypothesized as an ideal interstitial material due to its low neutron cross section and resistance to corrosion. Zirconium additions to U3Si2 were fabricated at 2%, 7%, and 14% by volume percent. These additions were characterized by powder-XRD, pre and post thermal solutionization, using the internal standard method and Rietveld refinement. The selected method allows for structural information to be derived about the atomic lattice, identification of secondary phases, and the quantification of interstitial mixing.
Proceedings Inclusion? Planned: Publication outside of MS&T

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

In Situ Observation of Short- and Long-Timescale Material Property Evolution Under Extreme Conditions
Analysis of Heavy Ion Irradiation Damage in Commercially Pure Titanium and Titanium Alloys
Benefits of Using High Energy Ions in Ion Irradiation Experiments to Evaluate Void Swelling
Characterization of Microstructure Evolution in Ceramic Materials Using Acoustic and Thermal Transport Measurements
Characterization of Stress and Environment Dependent Fracture Mechanisms of SiC/SiC CMCs
Controlling Helium Morphology in Pure Metals: Effects of Helium Defects on Deformation and Strength
Corrosion Control of Austenitic Stainless Steel and Nickel-Based Alloys in Molten Chloride Salt Environments
Design of a Hot Hydrogen Test Loop for Testing of Nuclear Thermal Rocket Elements
Development of a Combined Thermal Hydraulic and Materials Corrosion Liquid-Sodium Experimental Facility
Development of an In-Situ Mechanical Test System for Advanced Reactor Coolants
Effects of Post-Processing Variability on Radiation Response of Additively-Manufactured HT9
Enabling In-situ Crack Growth Testing and Monitoring in VTR Cartridge Loop Environments
Explaining the Corrosion Morphology of Structural Materials in Molten Fluoride Salts With/Without Radiation
Fundamental In-situ Experiments Coupled to High-throughput Approaches to Understand Radiation Damage in FCC and BCC Compositionally Complex Alloys
In situ Crack Loading and Measurement Techniques for Gen IV Reactor Coolant Media
In Situ Observation of Irradiation Damage in Polycrystalline Nuclear Graphite
Investigation of Uranium Silicide Fuel Form Additions through Rietveld Refinement and Internal Standard P-XRD
Mobility of Hydrogen in YH2 Probed by Nuclear Magnetic Resonance
Modified stereo TEM for 3D analysis of defects
Nanomechanical Change of Tungsten in ELM Conditions
Nonlinear Ultrasound for Nondestructive Evaluation of Microstructural Defects
Response of an Additively Manufactured 316 Stainless Steel Subjected to High Temperature Heavy Ion Irradiations
Selective Irradiation Behavior in Dual Phase 308L Filler of SA508-304L Dissimilar Metal Weldment after Proton Irradiation
STEM Characterization of Dislocation Loops for an Irradiated Model FCC Alloy
Study on Hydrogen Isotopes Solubility and Diffusivity in Y- and Co-doped Barium-zirconates Using Tritium Imaging Plate Technique
Swelling of Nuclear Reactor Steels: Modeling, Theory, and Accelerated Testing
Unveiling High Temperature Damage Mechanisms via In-situ Digital Image Correlation of Chromium-coated Zirconium-based Fuel Claddings

Questions about ProgramMaster? Contact programming@programmaster.org