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Meeting Materials Science & Technology 2019
Symposium Materials for Nuclear Applications
Presentation Title Development and Performance of High Temperature Irradiation Resistant Thermocouples
Author(s) Brian J. Jaques, Richard Skifton, Lan Li, Beck Perrine, Scott Riley, Ember Sikorski
On-Site Speaker (Planned) Brian J. Jaques
Abstract Scope In order to decrease the nuclear innovation cycle time, the future of nuclear technology is dependent on the acquisition of reliable, real-time in-pile data. Specifically, the development of high temperature irradiation resistant thermocouples (HTIR-TCs) for continuous, accurate, and stable temperature monitoring in reactors exceeding 1250 C is paramount for process/experiment monitoring, simulation validation, and code development. Prior to deployment, HTIR-TCs are thermally conditioned at temperatures above the maximum service temperature to reduce drift with time at exposure. In this work, effects of thermal stabilization treatments on the properties of thermocouples is predicted and characterized. Modeling and simulation (density functional theory-based) were coupled with experiment to optimize such treatments and predict their behavior at temperature over time. HTIR-TCs were examined before and after the stabilization heat treatment, as well as extended (1000 hour) exposures, for mechanical properties, Seebeck, and chemical behavior.
Proceedings Inclusion? Definite: At-meeting proceedings

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

A Metallic Multilayer Composite for use in Fluoride Molten Salt Reactors
Aging Behavior and Microstructure Evolution in Ni-Cr-Mo-W (Haynes 244) Alloy After Surface Treatment by Laser Shock Peening (LSP)
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
Characterization of the Impact of Fission Product Inclusion on Phase Development in U3Si2 Fuel
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
Enhancing the Properties of a Cast FeNiMnCr10 Co-free High-entropy Alloy Through Hot Rolling
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
High-Entropy Carbide Ceramics for Extreme Environments
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
Materials for Capture of Uranium for Nuclear Fuel from Fertilizer
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
Predicting Concrete’s Response to Irradiation
PVD Coating of Surrogate Fuels for Deep Space Nuclear Thermal Propulsion
Rapid Multiscale Simulation of Cladding Performance: Application to HT-9
Spectral Thermal Conductivity Predictions in UO2 with Xe Inclusions
Steam Oxidation and Microstructural Characterization of U3Si2 alloyed with Al, Cr, Nb, Y, and Zr
Temperature Impacts on Damage Response in Mixed Carbides
The Effects of Ultrasonic Nanocrystal Surface Modification at Room Temperature and Elevated Temperatures on Residual Stress, Microstructure and Mechanical Properties of Nuclear Alloys IN600 and IN690.
Thermophysical Properties of Binary Cl & F Compositions for Next Generation Molten Salt Reactors
Transmutation-induced Precipitation in Tungsten Irradiated with a Mixed Energy Neutron Spectrum
Unveiling SiC/SiC CMC Cladding Failure Mechanisms and Hermetic Performance with In-situ 3D-Digital Image Correlation
Uranium Nitride and High Temperature Irradiation Resistant Thermocouples towards Accident Tolerant Nuclear Fuel
Using ACRT with MVB Furnace to Achieve Low-cost CZT

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