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Meeting MS&T26: Materials Science & Technology
Symposium Ceramics and Glasses for Nuclear Energy Applications
Presentation Title Pyrolysis of Polymer-Derived SiC Coatings on Surrogate TRISO Fuel Particles
Author(s) Kishore Behera, Kathy Lu
On-Site Speaker (Planned) Kishore Behera
Abstract Scope Polymer-derived silicon carbide has strong potential for nuclear and high-temperature applications because SiC provides excellent thermal stability, oxidation resistance, mechanical stability, and fission-product retention capability. In TRISO fuel particles, the SiC layer protects the fuel kernel and retains metallic and gaseous fission products under reactor operating conditions. In this study, surrogate TRISO fuel particles were coated using an SMP-10 polycarbosilane-based precursor and added with ball-milled nano-SiC filler to reduce shrinkage during pyrolysis. The coated particles were pyrolyzed in argon atmosphere at 1000, 1300, and 1600 °C to study densification, phase evolution, and microstructure changes. Uniform SiC-based coatings with an average thickness of 82 ± 5 μm were obtained. With increasing pyrolysis temperature, the amorphous polymer derived SiC network progressively transformed into a more ordered and crystalline SiC structure. The 1600 °C coating showed nano SiC crystallization, better densification and nano hardness of 18 ± 1 GPa compared with lower-temperature coatings.

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

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
K-1: Size Effects on Oxidation Behaviors of IG-110 Nuclear Graphite During Water Vapor Ingress Accidents for High Temperature Gas-Cooled Reactors
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
Sorbent Characterization and Method Development for Mercury Capture for the Hanford WTP
Structure of MoO3 Containing Alkali Aluminoborophosphate Glasses

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