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Meeting Materials Science & Technology 2020
Symposium Advanced Characterization of Materials for Nuclear, Radiation, and Extreme Environments
Presentation Title Development of a Combined Thermal Hydraulic and Materials Corrosion Liquid-Sodium Experimental Facility
Author(s) Dustin Mangus, Juwan Johnson, Brett Leitherer, Peter Beck, Seth Walton, Guillaume Mignot, Wade Marcum, Julie Tucker, Samuel Briggs
On-Site Speaker (Planned) Dustin Mangus
Abstract Scope The Versatile Test Reactor (VTR) program seeks to design and construct a fast-spectrum research reactor to assist in addressing the capability gap in the testing of advanced core materials, fuels, and instrumentation in prototypical next-generation nuclear reactor environments. Current design concepts utilize the proven experience of pool-type sodium-cooled reactors. In support of this program, Oregon State University has developed the Glovebox for Experimental Liquid Sodium (GELS) facility to support the development of test rigs and instrumentation enabling environmentally assisted cracking experiments with in-situ monitoring capabilities in proposed cartridge loop environments. This facility allows for chemistry-controlled flowing or static liquid sodium test capabilities, facilitating both thermal-hydraulic and materials corrosion experimental needs. This is done using a secondary diagnostics loop with oxygen cold-trapping capabilities and a combination of conduction and moving-magnet pumps. An overview of facility design specifications and preliminary experimental results will be presented.
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

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