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Meeting MS&T26: Materials Science & Technology
Symposium Ceramics and Glasses for Nuclear Energy Applications
Presentation Title K-1: Size Effects on Oxidation Behaviors of IG-110 Nuclear Graphite During Water Vapor Ingress Accidents for High Temperature Gas-Cooled Reactors
Author(s) Seula Lee, Yi Je Cho
On-Site Speaker (Planned) Seula Lee
Abstract Scope Nuclear graphite is used in high temperature gas-cooled reactors owing to its thermal stability, high thermal conductivity, low thermal expansion, and neutron-moderating ability. During water vapor ingress accidents, however, graphite oxidation can induce mass loss, pore growth, and structural degradation. Oxidation rate and depth are governed by oxidant diffusion length and pore structure, which determine gas transport and reaction distribution. Sample size and geometry can also affect oxidation behaviors, but these effects are often neglected in kinetic models. This limits the accurate oxidation prediction of graphite components in different geometries. In this study, IG-110 nuclear graphite specimens with different sizes and geometries were oxidized in He/H2O atmospheres up to 1100 °C. Mass loss was used to calculate oxidation rates and activation energies, while cross-sectional observations evaluated pore evolution and oxidation depth. A kinetic model incorporating sample size correction parameters effectively described the observed oxidation discrepancies.

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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