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Meeting 2026 TMS Annual Meeting & Exhibition
Symposium Ceramics and Ceramic-Based Composites for Nuclear Applications III
Presentation Title Design of High-Temperature Composite Radiation Shields Using Bayesian Optimization
Author(s) Byron Millet, Jim Steppan, Taylor Sparks, Tom Meaders, Lee Sorensen, Matt Coventry
On-Site Speaker (Planned) Byron Millet
Abstract Scope Radiation shielding (RS) that is stable at operating temperatures up to 650 °C is essential for future nuclear systems, including electromagnetic pumps for liquid sodium-cooled fast reactors and advanced microreactors. We demonstrate the application of Bayesian optimization to the RS design configuration and material selection for multi-layer composite radiation shields produced via filament winding with in situ potting. Composite radiation shields were fabricated using several radiation-shielding filler materials. An open-source, physics-based radiation simulation developed using GEANT4 for the evaluation of multi-layer RS configurations was coupled with Python-based machine learning algorithms to optimize composite RS materials and designs. Proof-of-concept composite shields, stable at 650 °C, showed improved neutron and gamma attenuation compared to an equivalent thickness of borated HDPE. The simulation was enhanced by incorporating CAD-based geometry to enable application-specific radiation shield optimization and streamline development by reducing the need for many expensive and complex experimental radiation exposure tests.
Proceedings Inclusion? Planned:
Keywords Computational Materials Science & Engineering, Machine Learning, Modeling and Simulation

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Characterization of thermal physical properties and defects in Ln-doped UO2
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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
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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
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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
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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
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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
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Quantification of Porosity in Graphite
Separate effect studies on fission gas release behavior of UO2 and Cr2O3-doped UO2 of various microstructures
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