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About this Symposium

Meeting MS&T26: Materials Science & Technology
Symposium Ceramics and Glasses for Nuclear Energy Applications
Sponsorship ACerS Energy Materials and Systems Division
Organizer(s) Charmayne E. Lonergan, Missouri University of Science and Technology
Lingfeng He, Texas A&M University
Jake W. Amoroso, Savannah River National Laboratory
Krista Carlson, University of Nevada, Reno
James Neeway, Pacific Northwest National Laboratory
Brian J. Riley, Pacific Northwest National Laboratory
Scope Advanced ceramics and glasses are attractive for next generation nuclear energy as they expand the potential of the associated systems. Ceramics increase the operating temperature window to allow higher thermal efficiency, and exposure to variable environments. Glasses offer opportunities for waste treatment solutions, in-situ monitoring via newly designed windows or sensors, and coatings.

Although ceramic nuclear fuels have been widely used in current nuclear reactors, advanced energy systems call for improved materials with optimized properties. As an example, ceramic coatings are under active research to improve accident tolerance of nuclear fuels. Ceramic and glass materials can also contribute to enhanced safety of nuclear systems such as neutron moderation and shielding materials.

High entropy ceramics and advanced composites are promising options for advanced nuclear applications. Moreover, glass, ceramic, cement, and/or geopolymer compounds have been applied in long-term waste disposal. Processing, properties, testing and accelerated qualification of these materials should be expanded to meet the future needs of the nuclear industry. This symposium focuses on experimental and computational studies of ceramics and glasses for nuclear energy research and applications.


Abstracts are solicited on, but not limited to, the following topics:
• Materials Design, Synthesis/Processing, Additive Manufacturing, Accelerated Materials Qualification (including ceramic fuels, ceramic coatings, glass/ceramic waste forms and inert matrix fuels)
• Vitrification Technologies and Radioactive Waste Immobilization
• Radiation Damage Effects, Chemical Compatibility and Corrosion, Thermophysical Properties
• High Entropy Ceramics and Advanced Composites for Next-Generation Nuclear Applications
• Materials for Molten Salt Reactor Application
• Neutron Moderation and Shielding Materials
• Synergistic Multi-scale Modeling and Experimental Study on Microstructure Evolution, Mechanical and Physical Properties of Ceramics in Nuclear Energy Environments
•High Temperature Ceramics for Fusion Applications

Abstracts Due 05/19/2026

PRESENTATIONS APPROVED FOR THIS SYMPOSIUM INCLUDE


Addressing Integration and Compatibility Challenges in SiC-Based Materials for Fusion Applications
Analysis of Crucible-Scale Corrosion Testing of Monofrax® K-3 Refractory in Contact with Glass Melts
Decoding the Structure of Transition Metal Cations Containing Aluminoborosilicate-Based Nuclear Waste Glasses
Densification Mechanisms of Oxide Cermets Sintered by Spark Plasma Sintering for Nuclear Waste Form Application
Design Defects in Nuclear Fuels Using Boron Neutron Capture Method
Effects of Swift Heavy Ion Irradiation on the Oxidation Behavior of Zirconium Carbide
From Irradiation Defects to Thermal Conductivity: A Defect-Informed Model for He Ion-Irradiated Ceria
Impact of MgO on Nepheline Crystallization in Sodium Aluminoborosilicate Glasses
Impact of Potassium on Structure, Chemical Durability, and Nepheline Crystallization of Sodium Aluminoborosilicate Glasses
Impact of Radiation Damage in Uranium Dioxide: Integrating Rate Theory Defect Evolution with Irradiated Annealed Sample
Impact of Vanadium on Structure and Properties of Sodium Aluminoborosilicate Glasses
In-Situ TEM Observation of Irradiation-Induced Basal Plane Formation and Ordering in Nuclear Graphite
Influence of Compositional Complexity on the Structural Integrity of Residual (SIR) and Chemical Leaching Behavior in Nuclear Waste Glasses
International Advances in Glass Waste Forms for Low- and Intermediate-Level Radioactive Waste
Investigating the Impact of BiPO4 on Sulfur Solubility in Sodium Aluminoborosilicate Glasses
Key Synergies in Zirconium Carbide (ZrC) and its Alloying with Uranium (U,Zr)C for Nuclear Fuel Applications
Microstructural Characterization and Melt Composition Changes During Corrosion of Monofrax® K-3 in HLW Glasses
Processing, Properties, and Irradiation Response of ε-W2B5 Based Ceramics for Compact Fusion Reactor Shielding
Pyrolysis of Polymer-Derived SiC Coatings on Surrogate TRISO Fuel Particles
Revisiting Radioiodine Sorbent Regeneration
Sintered SiC and its Potential in Nuclear Energy Applications
Size Effects on Oxidation Behaviors of IG-110 Nuclear Graphite During Water Vapor Ingress Accidents for High Temperature Gas-Cooled Reactors
Sorbent Characterization and Method Development for Mercury Capture for the Hanford WTP
Structure of MoO3 Containing Alkali Aluminoborophosphate Glasses


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