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Meeting MS&T22: Materials Science & Technology
Symposium Ceramics for a New Generation of Nuclear Energy Systems and Applications
Presentation Title J-1: Development and Characterization of Ga/Ta Doped Li7La3Zr2O12 for Direct LiT Electrolysis
Author(s) Rahul Rajeev, Shraddha Jadhava, Brenda Garcia-Diaz, Christopher Dandeneau, Dale Hitchcock, Kyle Brinkman
On-Site Speaker (Planned) Rahul Rajeev
Abstract Scope Li7La3Zr2O12(LLZO) has proven to be an excellent material for all-solid-state battery electrolytes with good stability and ionic conductivity. In this work, LLZO is being developed for use in a direct lithium tritide (LiT) electrolysis process for hydrogen isotope extraction from Lead Lithium (Li-Pb) blanket in a nuclear fusion reactor. Three different routes were used to synthesize phase pure LLZO with the addition of aliovalent dopants Gallium (Ga) and Tantalum (Ta) used to stabilize the high conducting cubic LLZO phase. The stability was studied by immersing LLZO powders and sintered pellets in the Li-Pb blanket material at 450oC in an inert environment for several hours. Post exposure characterization tests were performed including powder X-ray diffraction(PXRD) Rietveld analysis, transmission electron microscopy (TEM), and energy dispersive X-Ray (EDX) analysis to detect potential phase degradation.

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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