ProgramMaster Logo
Conference Tools for MS&T24: Materials Science & Technology
Login
Register as a New User
Help
Submit An Abstract
Propose A Symposium
Presenter/Author Tools
Organizer/Editor Tools
About this Abstract
Meeting MS&T24: Materials Science & Technology
Symposium Ceramic Materials for Nuclear Energy Systems
Presentation Title Compositionally Complex Carbide Ceramics: A Perspective on Irradiation Damage
Author(s) Bai Cui, Lanh Trinh, Fei Wang, Kaustubh Bawane, Khalid Hatter, Zilong Hua, Linu Malakkal, Lingfeng He, Luke Wadle, Yongfeng Lu
On-Site Speaker (Planned) Bai Cui
Abstract Scope Extensive experimental and computational studies have demonstrated outstanding physical and chemical properties of the novel materials of compositionally complex carbides (CCC), enabling their promising applications in advanced fission and fusion energy systems. This presentation provides a comprehensive overview of radiation damage behavior reported in the literature to understand the fundamental mechanisms related to the impact of multi-principal metal components on phase stability, irradiation-induced defect clusters, irradiation hardening, and thermal conductivity of compositionally complex carbides. Several future research directions are recommended to critically evaluate the feasibility of designing and developing new ceramic materials for extreme environments using the transformative “multi-principal component” concept. Compared to the existing materials for nuclear applications including stainless steels, nickel alloys, ZrC, SiC, and potentially high-entropy alloys, as well as certain other compositionally complex ceramic families, CCC appear to be more resistant to amorphization, growth of irradiation defect clusters, and void swelling.

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

Advanced FCCI Barrier Coatings: Enhancing Fuel Cladding Performance Against Metallic Fuels at High Temperatures
Advanced Moderator Module (AMM) Manufacturing
Analysis of Crucible-Scale Corrosion Testing of Monofrax® K-3 Refractory in Contact with Glass Melts
Assessing High Uranium Density Ceramic Fuels for Implementation in Water Cooled Reactors
Atomic Scale Order in Swift Heavy Ion Irradiated MgAl2O4 Spinel Oxide
Atomistic Investigation of Defects and Defect-Phonon Scattering in ThO2
Beyond TRISO: Development of New Coated Particle Fuels
Challenge of Making Accurate Heat Capacity Measurements for Fluoride Salts
Characterization and Thermal Stability of (Li,Na,K)2O-Fe2O3-P2O5 Glasses for Waste Immobilization.
Chemical Durability of Cermet Waste Forms for Advanced Reactor Wastes
Chemical Thermodynamic Database Development and Applications for Molten Salt Reactors
Cluster Dynamics Simulations of Tritium and Helium Diffusion in Lithium Ceramics
Composition and Properties of Iron Phosphate Waste Forms for Radioactive Salt Waste Immobilization
Compositionally Complex Carbide Ceramics: A Perspective on Irradiation Damage
Corrosion Behavior of Nuclear Waste Glasses and Glass-Ceramics in Geological Repository Systems
Determining Waste Glass Corrosion via the EPA Method 1313, the Stirred Reactor Coupon Analysis, and Vapor Hydration Testing
Development of Kernel Fuels for High Temperature Gas Reactor and Space Systems
Developments in Producing Pyrolytic Carbon Coatings for Advanced Particle Fuel Forms
High-Burnup Structure Formation and Associated Fission Product Diffusion in UO2
High Temperature Ceramic Nuclear Fuels for Cross-Cutting Applications
Indirect Powder Bed Fusion of Ceramics for Neutron Radiation Shielding
Investigation of Technetium Management through Chalcogenides and Bimetallic Nanoparticles
Ion irradiation of UC and UN and their surrogates
Iron Phosphate Glasses for Waste Immobilization
Irradiation Study of TiN Inner-Wall Coating for Advanced Cladding to Suppress FCCI
Mechanisms Controlling Defect Evolution in Irradiated CeO2 Using In-Situ TEM Annealing
Microstructural Characterization and Thermal Oxidation of Zr Doped UO2
Molecular Dynamics Simulations of Displacement Cascades in LiAlO2 and LiAl5O8 Ceramics
MXene Hybrids as Promising Candidates for Iodine Gas Capture
Numerical Modeling of Graphite Oxidation In Water Vapor Ingress Accidental Conditions for High Temperature Gas-Cooled Reactors
Oxidation Behavior of the SiC Coating of TRISO Fuel Particles in Air or Water Vapor
Performance of CrAl/Al2O3 Multilayer H2 Permeation Barrier Designed for High Temperature Metal Hydride-Based Neutron Moderators
Phase Equilibria and Thermodynamics of Uranium Mononitride Fuel Undergoing Burn-Up in a Lead-Cooled Reactor
Phosphate-Based Dechlorination of Electrorefiner Salt Waste Using a Phosphoric Acid Precursor
Post-irradiation Examination of Irradiated Fuels with Pulsed Neutrons at LANSCE
Predicting the Durability of Nuclear Waste Immobilization Glasses Using Nonparametric Machine Learning
Progress on Modeling Refractory Corrosion of Waste Glass Melters
Protecting Structural Components in Molten Salt Nuclear Reactors with Functional Coatings.
Room Temperature Micro-Cold Spray of Ceramic Thick Films
Statistical Fracture Behavior of Doped UO2 Using a Ball-On-Ring Test Method
Structural Analysis of Swift Heavy Ion Irradiated β-Sc2Hf7O17 and γ-Sc2Hf5O13
Structural Manipulation of Ceramic Materials via Extreme Conditions
Thermal Diffusivity of UN Produced via Carbothermic Reduction Prior to Nitriding
Thermodynamic Modeling of the LiF-NaF-(La,Ce,Pu)F3 Systems for Molten Salt Reactor Applications
Time-Temperature-Transformation Diagram Development for a Coupled-Operation Glass Composition with SWPF

Questions about ProgramMaster? Contact programming@programmaster.org