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Meeting MS&T22: Materials Science & Technology
Symposium Ceramics for a New Generation of Nuclear Energy Systems and Applications
Presentation Title Comparison of ZrC-TZM Mechanical and Structural Properties Before and After Extended Carbon Exposure
Author(s) Peyton McGuire, Erofili Kardoulaki, Ming Tang
On-Site Speaker (Planned) Peyton McGuire
Abstract Scope Mo alloy TZM has demonstrated favorable mechanical properties for use in high-temperature environments. However, carbon exposure in-service can carburize TZM and alter its mechanical properties. Ceramic coatings for TZM are considered to help strengthen the overall system, with particular emphasis being placed upon zirconium carbide (ZrC) due to its high-temperature stability and favorable mechanical properties. In this work, ZrC was coated on a TZM substrate through chemical vapor deposition. The sample was exposed to graphite for 40 days at 1000°C to test the carburization resistance the coating can offer. Samples were analyzed using nanoindentation and TEM/STEM techniques. The hardness and elastic moduli of both exposed and non-exposed coatings/substrates were analyzed and compared against literature to see whether or not the exposure to carbon had significantly affected the structural and/or mechanical properties, and if such a difference would affect the use of ZrC as a protective coating of TZM claddings.

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

A Physics-Based Cluster Dynamics Model of Radiation-Enhanced Growth of Oxides
Additive Manufacturing of Ceramics for Nuclear Applications
Bismuth Loaded Carbon Foam as an Effective Radio Iodine Sorbent
Characterization of Radiation Effects in Ceramics with Spallation Neutron Probes
Characterizing Effects of Aging Bismuth Laden Sorbents in NOx Atmosphere for Radioiodine Capture
Cluster Dynamics Simulations of Point Defects and Fission Gas Evolution in Irradiated Ceramic Nuclear Fuels
Comparison of ZrC-TZM Mechanical and Structural Properties Before and After Extended Carbon Exposure
Corrosion of SiC in Molten Salt and Liquid Lead
Development of Novel TRU-containing Ceramics for Nuclear Waste Immobilization
Environmental Degradation of Ceramic Materials in Nuclear Energy Systems
Fabrication and Properties of Sintered Yttrium Hydride
Integration of Nuclear Fuel and Embedded Sensors within Additively Manufactured SiC Components
J-1: Development and Characterization of Ga/Ta Doped Li7La3Zr2O12 for Direct LiT Electrolysis
J-2: Evaluation of In-Flow Mechanical Robustness of Metal-Functionalized Porous Silica Materials
Microstructural Evolution in Ceramic Nuclear Fuels and their Surrogates under Irradiation
Modeling Vibrational Modes in Raman Spectra of ThO2
Phonon Broadening in High Entropy Ceramic Carbide
Radiation Damage of Ion-irradiated High Entropy Ceramics
Radiation Effects in Single-crystal High-entropy Oxides
SiC Oxidation and Irradiation Resistance in Advanced Nuclear Reactor TRISO Fuel
Single Component Variations in Glass Ceramic Waste Forms
Sulfur Retention of Low Activity Waste Glasses
Synthesis and Characterization of Super Occluded LiCl-KCl in Zeolite-4A as a Chloride Salt Waste Form Intermediate

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