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Meeting 2026 TMS Annual Meeting & Exhibition
Symposium Ceramics and Ceramic-Based Composites for Nuclear Applications III
Presentation Title An Introduction to Idaho National Laboratories Sample Preparation Laboratory
Author(s) Joshua J. Kane
On-Site Speaker (Planned) Joshua J. Kane
Abstract Scope The Sample Preparation Laboratory (SPL) at Idaho National Laboratory (INL) is a cutting-edge hazard category 3 facility designed to advance irradiated nuclear materials research. This new facility offers high-throughput sample preparation, mechanical testing, surface science, and microstructural analysis, addressing critical gaps in the United States’ infrastructure for testing high-activity irradiated materials. Utilizing automated operations in shielded cells and along with INL’s existing advanced modeling and simulation capabilities, the facility hopes to significantly reduce the time required for developing, testing, and ultimately qualifying new radiation-resistant materials. The Sample Preparation Laboratory and facility staff look forward to collaborating with industry, university researchers and students, as well as further supporting long-standing efforts across the Department of Energy Laboratory complex and internationally. This presentation will provide a brief overview of SPL’s capabilities, its current mission, and anticipated future role in nuclear materials research.
Proceedings Inclusion? Planned:

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

AI/ML-assisted Design of Phosphate Nuclear Waste Forms
An Introduction to Idaho National Laboratories Sample Preparation Laboratory
Anisotropic Thermal Response of SiC-SiC Composite Cladding Via IR Thermography
Atomistic mechanism of faulted loop formation in fluorite oxide under irradiation
Build anisotropy in binder jet printed ZrC and its implications
Can studies on specific fuels from past irradiations be valorised for future SMR/AMR
Capturing the in-situ evolution of thermal diffusivity of SiC during ion irradiation
Ceramic-Metal Joint Technology for Fusion Reactor Functional Coating
Characterization of thermal physical properties and defects in Ln-doped UO2
Characterizing Thermal Conductivity in Anisotropic Textured Graphite Composites Using Frequency Domain Thermoreflectance
Compatibility of SiC materials with liquid metal/molten salt under simultaneous irradiation and corrosion at higher temperatures
Computational study of the impact of dopants on UO2 creep rates
Data-driven prediction of graphite oxidation behaviors in accidental conditions of high temperature gas-cooled reactors: graphite size and shape effects
Design of High-Temperature Composite Radiation Shields Using Bayesian Optimization
Development of Coated Particle Fuels with New Architectures for an Expanded Service Envelope (Invited Talk)
Development of Ultra-High Temperature Instruments for Measuring Thermophysical Properties in Nuclear Applications
Developments in zirconium carbide-based materials from the recent Nuclear Thermal Propulsion programs
Effective Thermal Conductivity Modeling of Synthetic TRISO Fuel Compacts
Effects of Compositional Complexity on the Ion Irradiation Response of Pyrochlore Oxides
Entrained Hydride Ceramic Composite Moderators for Transforming Reactor Economics
Evolution of Fission Products and Nanograins in UO2 under In-Situ Ion Irradiation
Fabrication of Engineered Oxide Fuels Mimicking High Burnup Fuels and their Fragmentation Mechanisms Under Simulated LOCA and RIA Thermal Transient
Fracture in low-textured pyrolytic carbons: transitioning from homogeneous to heterogeneous bond breaking morphology with increasing microstructural order
Graphite for Accelerator Beam Intercepting Devices
High Temperature X-ray and Neutron Studies of Uranium Carbide Systems
Image-Based Thermal Modeling of SiC/SiC Composites Using Synchrotron XCT
Impacts of Fission Product Speciation on Mechanical Properties of Advanced Ceramic Nuclear Fuels and Their Surrogates
Improved Radiation Shielding in SPS-Processed B4C Ceramics via High-Entropy Alloy Additions
In-situ irradiation of zirconium carbide and zirconium nitride above 800°C
Inferring the Local, Temperature-Dependent, Anisotropic Thermal Conductivity of TRISO fuel Constituents from Modulated Photothermal Phase Data
Investigation of FLiBe salt infiltration into graphite for molten salt reactors
Ion Beam Irradiation and Analysis of Ultra-High Temperature Ceramics
Ion irradiation and mechanical properties of additively manufactured tungsten carbide for nuclear power applications in molten salt reactors
Irradiation Experiment on the Performance of Carbon Composites at High Temperatures
Measured Radiation-Induced Bowing of SiC/SiC Composite Components under Neutron Flux Gradients
Mesoscale modeling of restructuring in high burnup UO2 fuel
Mitigation of Room Temperature Oxidation of Uranium Mononitride Through the Use of Powder Injection Molding
Modelling Failure Probability Distribution in Nuclear Graphite
Multi-Phase Modeling of Defect Accumulation and Thermal Conductivity Degradation for Irradiated UO2 Fuel
Multiscale Modeling of Fracture in Nuclear Fuels
Nanoscale Lithium Redistribution and Void Formation in Neutron Irradiated LiAlO₂ Ceramics Studied by EFTEM and Electron Diffraction
Neutron Irradiation Effects on Thermal Conductivity and Dimensional Stability of TiC, TiB₂, and ZrB₂ Ultra-High-Temperature Ceramics
Oxygen diffusion mediated by multi-dimensional defects in ThO2
Preliminary Results from Post Irradiation Examination of AGR-5/6/7 TRISO Fuel
Preliminary Results on Non-destructive TRISO Defect Identification via X-ray Imaging
Progression of SiGAŽ Cladding Technology To Support Nuclear Power Generation
Properties and behaviour of oxide nuclear fuels: Insights from atomic scale modelling
Quantification of Porosity in Graphite
Separate effect studies on fission gas release behavior of UO2 and Cr2O3-doped UO2 of various microstructures
Some thoughts on the ‘linear inelastic’ behaviour of nuclear graphite
The In Situ Mechanical Response of Surrogate TRISO Using Synchrotron Micro-Computed Tomography
The thermodynamic stability of chernobylites
Understanding tritium trapping in permeation barrier coatings for fusion breeder blanket applications.

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