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Meeting 2022 TMS Annual Meeting & Exhibition
Symposium Mechanical Behavior and Degradation of Advanced Nuclear Fuel and Structural Materials
Presentation Title Ion Irradiation Effects on Microstructure Evolution and Mechanical Properties of Silicon Oxycarbide
Author(s) Kathy Lu, Sanjay Kumar Singh, Kaustubh Bawane
On-Site Speaker (Planned) Kathy Lu
Abstract Scope Polymer derived ceramic is a promising nuclear material due to its irradiation resistant and high temperature stable nature. In this study, silicon oxycarbide (SiOC) was fabricated by pyrolysis of a polysiloxane precursor polymer at 1000C and 1500C in an Ar atmosphere. The 1000C pyrolyzed SiOC is fully amorphous while the 1500C sample involves crystalline β-SiC nanodomains, turbostratic carbon network, and amorphous SiOC. After 100 keV He ion irradiation, there is no detectable microstructural changes for the 1000C pyrolyzed SiOC. However, the 1500C pyrolyzed SiOC shows amorphization. Neither sample has He bubbles, elemental segregation or voids as a result of irradiation. Raman studies showed effect of irradiation on Si-O-C bonding. Irradiation temperature has no perceivable effect on microstructure and mechanical properties of SiOC. This work offers important guidance for development of radiation-tolerant polymer derived SiOC materials with applications in advanced nuclear reactors.
Proceedings Inclusion? Planned:
Keywords Ceramics, Characterization, Nuclear Materials

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

A Case Study on Radiation-induced Degradation of High-entropy Alloys
A Comprehensive Study of Responses of SiC Materials to Neutron Irradiation for ATF Cladding Application
Accelerating Advanced Fuel Development and Analysis by Combining Modelling and Experiment
Advanced Characterization and Multiscale Testing for SiC Ceramic Matrix Composite Cladding as Accident Tolerant Fuel Candidate Materials for LWR Applications
Advances in SiGA Development for Nuclear Applications
Atomistic Simulation Study of the Effect of Hydride Morphology on the Ductility of Polycrystalline Zirconium
Behavior of High Entropy Alloy in Molten Salt Environments Under Biaxial Stresses
Characterizing and Testing High Dose Neutron Irradiated Materials for Cladding Applications
Chemical Redistribution of Alloying Elements through Oxide/Metal Interface of Irradiated ZrNbFe Alloys and Its Implication on Corrosion Behavior
Comparison of Neutron Irradiation Effects in PM-HIP and Cast Grade 91 Steel
Compositionally Graded Specimen: A High-throughput Approach for Nuclear Material Development
Creep Performance of IronClad Accident Tolerant Fuel Cladding
Deconvoluting Properties of Additively Manufactured Alloy 718 Utilizing Coupled Microscopy and Machine Learning
Degradation Modes of Core Materials under Multiple Components of the Reactor Environment
Development of Advanced Nuclear Fuels for Current and Next Generation Reactors
Development of PVD Cr Coatings for Hydrothermal Corrosion Resistance of SiC-SiCf Fuel Cladding in LWRs
Diffusion in Doped and Undoped Amorphous Zirconia
Effect of Free Surface on Displacement-cascade Damage in Neutron Irradiated Nickel
Effect of Residual Strain on Short Time Oxidation Behavior of Machined 304 Stainless Steel in High Temperature, High Pressure Deaerated Water
Effective Bias of Cavities in BCC Fe: As Revealed by Atomistic Calculations
Effects of Low-temperature Neutron Irradiation, Hydrogen Charging, and Post-weld Heat Treatment on Tensile Properties of Welded Zircaloy-4
Effects of Steel Composition and Grain Size on Diffusion with Neodymium
Fabrication and Thermophysical Properties of (U,Zr)C; A High Uranium Density Fuel Candidate for Nuclear Thermal Propulsion Reactors
Fiber/Matrix Debonding of SiC/SiC Composites Evaluated Using the Micropillar Compression
Formation of UN in U-Mo Systems by Mechanical Alloying
Helium Implantation Responses of Co-deposited Copper-tungsten Nanocomposites
High Temperature Mechanical Testing of Uranium Fuel Pellets
Hydrogen Accommodation in the TiZrNbHfTa High Entropy Alloy
Impact Fretting Wear Behavior of Cr-alloy Coating Layer for Accident-tolerant Fuel cladding
Impact of Lithium Accommodation on Defect Chemistry in ZrO2
In Situ Microstructural Evolution in Face-centered Cubic Compositionally Complex Alloys under Dual-beam Heavy-ion Irradiation
In Situ Study of High Temperature Mechanical Behavior of Irradiated FeCrAl Alloys
Interface Characterization of an Explosion Welded Stainless Steel-clad Plate for Neutron Irradiation Studies
Investigating the Thermophysical Properties and Key Contributions to the Thermal Conductivity of Different Nitride Systems
Investigation of Degradation Mechanism of Accident Tolerant Fuel (ATF) Coated Cladding Concepts during Interim Storage and Transportation of Used Nuclear Fuels
Investigation of Elemental Segregation and Precipitation in Ion-irradiated Advance Austenitic Alloy A709 Using Advance Techniques
Ion Irradiation Effects on Microstructure Evolution and Mechanical Properties of Silicon Oxycarbide
Manufacturing of Oxide Dispersion Strengthened (ODS) Steel Fuel Cladding Tubes Using Cold Spray Technology
Mesoscale Modeling of the Relationships between Microstructure and Mechanical Properties in the Porous Pyrocarbon Buffer Layer for TRISO Particle Fuel
Micromechanical Testing of Femtosecond Laser Machined Tensile Samples of Varied Geometries
Microscale Thermal Conductivity and Residual Stress Measurements in TRISO Particle Coatings
Microstructural and Mechanical Properties of Hot Deformation Behavior of Zr-4 Alloy
Microstructure and Surface Chemistry of FeCrAl Alloys Accident Tolerant Fuel Cladding Subjected to Fast Heating Rate in Aqueous Environment
Microstructure Characterization and Micro-mechanical Properties of 14YWT Tubing after Proton Irradiation
Microstructure, Mechanical Properties, and Irradiation Response of Fe-Cr-Ni-based Multi-principal Element Alloys
N-32: Dynamics of Helium Bubbles during Thermal Annealing: A Data-driven Approach
N-34: Influence of the Bulk Chemical Composition on the Microstructure Evolution of Irradiated Chemically-tailored Nuclear RPV Steels
N-35: Investigation of Ion Irradiation Effects on Mineral Analogues of Concrete Aggregates
N-36: Stress Distribution of Disk Geometry Under Three-point Bending Tests to Evaluate Mechanical Properties of Neutron-irradiated Tungsten for Future Fusion Devices
N-37: Study of Microstructure, Hydrogen Solubility and Corrosion of Ta-modified Zr-1Nb Alloys for Nuclear Applications
Neutron Irradiation Effects on Mechanical Anisotropy in Alloy 625
NOW ON-DEMAND ONLY - Assessment of Local Deformation Behavior in Mesoscale Tensile Specimens via Digital-image Correlation
NOW ON-DEMAND ONLY - Deformation Behavior of Helium Irradiated Nano-pillars Containing a Helium Gas Bubble Superlattice
NOW ON-DEMAND ONLY - Development of Direct and Indirect Ab Initio Radiation Damage Models
NOW ON-DEMAND ONLY - Effect of Microstructure and Rolling Treatment on Static Recrystallization Behavior in Monolithic U-10Mo Fuel Foils
NOW ON-DEMAND ONLY - Material Degradation Analysis through Machine Learning-based Information Extraction from Legacy SFR Metallic Fuel Performance Data
NOW ON-DEMAND ONLY – On the Diffusion of Hydrogen Atoms towards Notch Tips in Zirconium Polycrystals: A CPFE Analysis
Phase-field Simulations of Fission Gas Bubbles in High Burnup UO2 to Inform Engineering-scale Fuel Performance Modeling
Phase Field Fracture Study of the Effect of Gas Bubble on Fracture at U-Mo/Zr Interface
Phase Stability in FeCrAl Alloys: Mapping the Miscibility Gap and Understanding the Impact of Alpha Prime Precipitation on Material Properties
Progress on Experimental Investigation of Degradation Mechanisms of ATF Coated Cladding under Transient Conditions
Radiation-Induced Segregation in Binary Alloy Systems Examined Via Phase Field Simulations
Role of Powder Microstructure and Mechanical Properties on Deposition and Properties of Cold Spray Cr Coatings
Simulation of Shearing-induced Edge and Interfacial Fractures in U-10Mo Monolithic Fuel Plates
Size-dependent Radiation Damage Mechanisms in Nanowires and Nanoporous Structures
Small Scale Mechanical Testing of Irradiated Cladding and Fuel for Nuclear Reactors
Strain Rate Sensitivity Studies of Commercial FeCrAl Alloy
Study of the Irradiation Induced Microstructure and Mechanical Properties in Low Alloyed Ferritic Steels
Temperature Sensitive Dislocation Dynamics Modeling of Hardening and Embrittlement
The Interaction between an Extended Edge Dislocation and a Helium Bubble in Copper
Through-thickness Microstructure Characterization in a Centrifugally Cast Austenitic Stainless Steel Nuclear Reactor Primary Loop Pipe Using Time-of-flight Neutron Diffraction
Unraveling the Early Stage Ordering of Krypton Solid Bubbles in Molybdenum: A Multi-modal Study
Uranium Nitride/Uranium Boride Composite Materials

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