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Meeting MS&T24: Materials Science & Technology
Symposium Ceramic Materials for Nuclear Energy Systems
Presentation Title Indirect Powder Bed Fusion of Ceramics for Neutron Radiation Shielding
Author(s) Arturo Hernandez-Barreto, Julian R. Gawel, Desiderio Kovar, Sheldon Landsberger, Jason Benkoski
On-Site Speaker (Planned) Arturo Hernandez-Barreto
Abstract Scope Additive manufacturing of ceramic-polymer composites for radiation shielding is conventionally done via material extrusion methods. Filament extrusion is constrained to relatively low ceramic solids loadings, which limits shielding capabilities. A powder bed fusion approach allows larger solids loadings of ceramics. In this study, zirconium diboride and nylon 11 powders are mixed and printed via selective laser sintering and evaluated as neutron radiation shields.

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

Advanced FCCI Barrier Coatings: Enhancing Fuel Cladding Performance Against Metallic Fuels at High Temperatures
Advanced Moderator Module (AMM) Manufacturing
Assessing High Uranium Density Ceramic Fuels for Implementation in Water Cooled Reactors
Atomic Scale Order in Swift Heavy Ion Irradiated MgAl2O4 Spinel Oxide
Atomistic Investigation of Defects and Defect-Phonon Scattering in ThO2
Beyond TRISO: Development of New Coated Particle Fuels
Challenge of Making Accurate Heat Capacity Measurements for Fluoride Salts
Chemical Thermodynamic Database Development and Applications for Molten Salt Reactors
Cluster Dynamics Simulations of Tritium and Helium Diffusion in Lithium Ceramics
Compositionally Complex Carbide Ceramics: A Perspective on Irradiation Damage
Corrosion Behavior of Nuclear Waste Glasses and Glass-Ceramics in Geological Repository Systems
Determining Waste Glass Corrosion via the EPA Method 1313, the Stirred Reactor Coupon Analysis, and Vapor Hydration Testing
Development of Kernel Fuels for High Temperature Gas Reactor and Space Systems
Developments in Producing Pyrolytic Carbon Coatings for Advanced Particle Fuel Forms
FLiNaZr salt spill splatter and aerosol formation in molten salt reactor accident scenario
High-Burnup Structure Formation and Associated Fission Product Diffusion in UO2
High Temperature Ceramic Nuclear Fuels for Cross-Cutting Applications
Indirect Powder Bed Fusion of Ceramics for Neutron Radiation Shielding
Ion Irradiation of UC and UN and Their Surrogates
Iron Phosphate Glasses for Waste Immobilization
Irradiation Study of TiN Inner-Wall Coating for Advanced Cladding to Suppress FCCI
L-3: Characterization and Thermal Stability of (Li,Na,K)2O-Fe2O3-P2O5 Glasses for Waste Immobilization.
L-4: MXene Hybrids as Promising Candidates for Iodine Gas Capture
L-5: Protecting Structural Components in Molten Salt Nuclear Reactors with Functional Coatings
L-6: Thermal Diffusivity of UN Produced via Carbothermic Reduction Prior to Nitriding
Mechanisms Controlling Defect Evolution in Irradiated CeO2 Using In-Situ TEM Annealing
Molecular Dynamics Simulations of Displacement Cascades in LiAlO2 and LiAl5O8 Ceramics
Numerical Modeling of Graphite Oxidation In Water Vapor Ingress Accidental Conditions for High Temperature Gas-Cooled Reactors
Oxidation Behavior of the SiC Coating of TRISO Fuel Particles in Air or Water Vapor
Performance of CrAl/Al2O3 Multilayer H2 Permeation Barrier Designed for High Temperature Metal Hydride-Based Neutron Moderators
Phase Equilibria and Thermodynamics of Uranium Mononitride Fuel Undergoing Burn-Up in a Lead-Cooled Reactor
Phosphate-Based Dechlorination of Electrorefiner Salt Waste Using a Phosphoric Acid Precursor
Post-irradiation Examination of Irradiated Fuels with Pulsed Neutrons at LANSCE
Predicting the Durability of Nuclear Waste Immobilization Glasses Using Nonparametric Machine Learning
Progress on Modeling Refractory Corrosion of Waste Glass Melters
Room Temperature Micro-Cold Spray of Ceramic Thick Films
Statistical Fracture Behavior of Doped UO2 Using a Ball-On-Ring Test Method
Structural Analysis of Swift Heavy Ion Irradiated β-Sc2Hf7O17 and γ-Sc2Hf5O13
Structural Manipulation of Ceramic Materials via Extreme Conditions
Thermodynamic Modeling of the LiF-NaF-(La,Ce,Pu)F3 Systems for Molten Salt Reactor Applications
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