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Meeting 2021 TMS Annual Meeting & Exhibition
Symposium Ceramic Materials for Nuclear Energy Research and Applications
Presentation Title Hydrothermal Corrosion Study of Additive Manufactured SiC Fibers
Author(s) Arunkumar Seshadri, Akshay Dave, Bren Phillips, Koroush Shirvan, Shay Harrison, Joseph Pegna
On-Site Speaker (Planned) Arunkumar Seshadri
Abstract Scope SiC/SiC fiber composites are pursued as fuel cladding materials to improve the accident tolerance of light water nuclear reactors (LWR). With the advent of advanced additive manufacturing of ceramic Laser-Induced Chemical Vapor Deposition (LCVD) developed by Free Form Fiber, high purity fibers are developed with the required stoichiometric and dimensional precision. The ability of these fibers to be used in practical applications requires the knowledge of its corrosion performance in extreme environment of a LWR. In the present work, hydrothermal corrosion of commercially available carbon-rich Hi-Nicalon fibers is compared to the stoichiometric and silicon-rich fibers manufactured using LCVD. Autoclave testing was carried out at 310 C and 14 MPa. Relative measurements based on the hydrothermal corrosion experiments reveal that LCVD fibers show good potential in terms of corrosion resistance compared to commercial fibers. The results also indicate that impact of stoichiometry is likely dominant compared to a particular manufacturing technique.
Proceedings Inclusion? Planned:
Keywords Additive Manufacturing, Nuclear Materials, Ceramics

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

A Model of Grain Boundary Energy Anisotropy in Uranium Dioxide Nuclear Fuel
A Thermo-mechanical Coupled Phase Field Dynamic Fracture Model and Its Application in UO2
Comprehensive Treatment of Thermal Transport Under Irradiation in ThO2
Development of Hydrothermal Corrosion Barrier Coatings for High-density Nuclear Fuels
Development of Yttrium Hydride for High Temperature Moderator Application
Effect of UV and Gamma Irradiation on the Hydrothermal Corrosion of Ion-irradiated SiC
Electron Microscopy Characterization of the Fuel-cladding Interaction in Annular Fast Reactor MOX
Evolution of Ion Irradiated Nitride Ceramics Properties for Coated Particle Fuel Systems
Exotic Magneto-elastic Properties in Uranium Dioxide
Hydrothermal Corrosion of Silicon Carbide
Hydrothermal Corrosion Study of Additive Manufactured SiC Fibers
Impact of Dislocation Loops on Thermal Conductivity of CeO2
In-situ Measurement of Tritium Release from Lithium Aluminate Under Neutron Irradiation
Influence of Dose Rate and Temperature on Mass Transport in Hematite
Ionization Effects on Damage Accumulation Behavior in SiC
Irradiation Damage in High-entropy Carbide Ceramics
Irradiation Effects on Zirconium Alloy Oxides and Their Impacts on In-reactor Corrosion Rates
Microstructural Analysis and Micro-mechanical Testing on Xenon-irradiated Uranium Dioxide
Microstructural and Fission Products Analysis from Irradiated UO2 Fuel Using Atom Probe Tomography
Microstructural Characterization of Radiation Effects in 3D printed SiC
Microstructure and Chemical States of Fission Products in Irradiated AGR-1 and AGR-2 TRISO Particle UCO Fuel Kernels
Modeling of Pressure-driven Inter-granular Fracture in High Burnup Structure UO2 during LOCA Using A Phase-field Approach
Molecular Dynamics Investigations of AlN-based Piezoelectric Ceramics under Irradiation
Multiscale Modeling of High Burn-up Structure (HBS) Formation and Evolution in UO2
On the Role of Neutron Irradiation Damages on Fission Products Transport in the SiC Layer of TRISO Fuel Particles
Oxidation Behavior of TRISO Fuel Materials
Phase-field Modeling of Bubble Growth During High Burn-up Structure Formation in UO2
Radiation Tolerance of Nanoporous Gadolinium Titanate
Radiolytic Damage and Hydrogen Generation at Carbide – Water Interfaces
TEM Characterization of Dislocation Loops in Ion-irradiated Single Crystal ThO2
TMIST-3A Post-irradiation Examination
Towards a Model of Coupled Irradiation and Corrosion

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