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Meeting 2017 TMS Annual Meeting & Exhibition
Symposium Ceramic Materials for Nuclear Energy Research and Applications
Presentation Title Anisotropic Thermal Conductivity and Interface Resistance in Pyrolytic Carbon Coated Zirconia Particles
Author(s) Yuzhou Wang, David Hurley, Erik Luther, Miles Beaux, Venkateswara Rao, Igor Usov, Marat Khafizov
On-Site Speaker (Planned) Yuzhou Wang
Abstract Scope Pyrolytic carbon (PyC) is used in composite materials. Its primary function is to act as an interphase layer that arrests crack propagation offering greater mechanical integrity to the composite. Its anisotropic properties due to layered structure of graphene sheets can be a limiting factor in nuclear applications. Here we investigate thermal transport in a composite consisting of zirconia (ZrO2) spherical kernels coated with a thin PyC layer. We implement laser-based thermal wave microscopy to measure anisotropic thermal conductivity in PyC and PyC/ZrO2 interface thermal resistance. The conductivity in circumferential direction is about 15 W/mĚK whereas radial conductivity is less than 1 W/mĚK. The interface resistance is on the order of 10-8 m2ĚK/W. The results of this work have implications for development of accident tolerant fuels where precise knowledge of conductivity is a critical design parameter and measurement of anisotropy in PyC layers of TRISO fuels to assess their mechanical integrity.
Proceedings Inclusion? Planned: Supplemental Proceedings volume


A TEM Study of Microstructure of Hi-Nicalon Type S SiC Composite beyond Ultimate Shear Strength
Alpha-damage Formation in Mixed Americium-uranium Compounds
Anisotropic Thermal Conductivity and Interface Resistance in Pyrolytic Carbon Coated Zirconia Particles
Atomistic Simulation of Swift Heavy Ion Irradiation Effects in UO2 and CeO2
Ceramic Materials for Nuclear Energy Research and Applications
Correlation Between Particle Size and Grain Size Distributions in Single/Multiphase Ceramic Oxide Surrogate Materials
Effect of Burn-up on the Thermal Conductivity of Fast Reactor MOX Fuel
Evaluation of Creep Behavior of UO2 at Sub-grain Length Scales
Evolution of Irradiation Defects in Ti2AlC Ceramics During Heavy Ion Irradiation
Five-dimensional Representation of Grain Boundary Energies in UO2
High Burn-up Nuclear Fuel, Impact of Fission Gases
Highlights of Ceramic Nuclear Fuel Research within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program
In-Situ Tritium Measurements from γ-LiAlO2 Pellets Irradiated in TMIST-3A
In Situ Synchrotron Characterization of the Field Assisted Sintering of UO2
Ionization-Induced Damage Annihilation in Silicon Carbide
Irradiation-induced Recrystallization in UO2: A Phase Field Study
Irradiation Dependent Deformation and Thermal Properties of SiC and SiO2 Measured by Using Nanomechanical Raman Spectroscopy
Irradiation Effects on Electrochemical Performance of TiO2 Anode
Micro-Mechancial Interphase Property Evaluation for SiC-SiC Composites
Microstructural Characterization of the Processes, Stability, and End-of-Range Effects in Heavily Irradiated Pyrochlores
Modeling the Effect of Percolation on Fission Gas Release in UO2 Nuclear Fuels
Molecular Dynamics Simulations of Thermal Transport in Uranium Dioxide with Intrinsic Defects and Fission Products
Multi-scale Coupled Radiation Damage and Heat Transport Modeling for Dispersed Nuclear Fuels
Multi-scale Modeling of Fracture Behavior in SiC with a Phase Field Fracture Model
Neutron Irradiated SiC Advanced Analysis to Understand Fission Product Transport: Safety Tested TRISO Coated Particles
One-Dimensional String-like Relaxation in Actinide Oxides
Phase Field Modeling of Uranium Dioxide Sintering and Densification
Probing Oxygen Defects in Ion Irradiated Actinide and Analogue Oxides Using Neutron Total Scattering
Processing Routes for Improving Purity and Theoretical Density of UN Microspheres
Progress in Development of Non-Oxide Ceramic Nuclear Fuels
Radiation-Stability of Zirconium Carbide and Nitride Ceramics for Advanced Fuel Cycles
Radiation Damage on UO2 and UN
Role of Ion Species in Radiation Effects of Lu2Ti2O7
Sensitivity Analysis and Uncertainty Quantification of the MARMOT Mesoscale Fuel Performance Code
Study of Oxide Dispersion Strengthened 316L Austenitic Steel by Mechanical Milling
Study of Point and Extended Defects in Fluorite UO2 with Variable Charges Empirical Potentials
The Roles of Surfaces, Chemical Interfaces, and Disorder on Plutonium Incorporation in Pyrochlores
Theoretical and Experimental Investigation of the Interrelationship Between Radiation Damage and Ionic Transport in Pyrochlore
Thermal-Mechanical Properties of Sintered UO2
Thermal Transport Properties of Uranium Dioxide from Atomistic Simulations
Thermoelectric Properties of Doped and Pure UO2 at High Temperatures

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