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Meeting 2024 TMS Annual Meeting & Exhibition
Symposium Irradiation Testing: Facilities, Capabilities, and Experimental Designs
Presentation Title Electron Energy Loss Spectroscopy (EELS) Characterization of Fuel Cladding Chemical Interaction (FCCI) Region in U-Zr Metallic Fuel Cladded with HT-9
Author(s) Arnold Pradhan, Daniele Salvato, Fei Xu, Tiankai Yao
On-Site Speaker (Planned) Arnold Pradhan
Abstract Scope Fuel cladding chemical interaction (FCCI) is one of the main factors that could limit the fuel performance of metallic fuels at high burnups. In this talk, we will focus on how the electron energy loss spectrum helps to reveal chemical and electronic structure information of light elements as well as heavy elements to improve mechanistic understanding of FCCI. EELS data was collected from FCCI region of U-10Zr solid fuel pin (with HT9 cladding) irradiated to burnup of 13 at% and analyzed using Hyperspy, an open-source python library. Elements namely C, O, Fe, Zr and lanthanides were mapped using the integration method. Decomposition was used to denoise the data while power law background subtraction was performed following deconvolution. The results show formation of different crystallographic phases with distribution of lanthanides, especially Ce and Nd, throughout the FCCI region. A C rich Zr-C rind was observed in the middle region of FCCI.
Proceedings Inclusion? Planned:
Keywords Nuclear Materials, Characterization, Other

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

Accelerated Irradiation Creep Testing of Structural Materials for Advanced Reactors
Accelerating the Pace of Radiation Damage Experiments through Novel Sample Geometries, Beam Line Architecture, and Machine Learning Analysis
Advancing Post-irradiation Examination of Structural Materials in INL Facilities
Advancing Thermo-physical Property Characterization Techniques and Methods for Irradiated Nuclear Fuels
Atom Probe Tomography Examinations of Bulk Zircaloy Irradiated at Nominally 410°C
Atom Probe Tomography (APT) Characterization of Annular U-Zr Metallic Fuel Cladded with HT-9
Challenges and Solutions for Fast Neutron Irradiation of Bulk Material Specimens
Comparison between Ion and Neutron Irradiated Tungsten to Simulate Damage in Commercial Nuclear Fusion Reactors
Deployment and Testing of a Fiber-based Instrument for In-reactor Thermal Property Measurements at MIT Research Reactor
Developing Irradiation Experiments to Enable Characterization and Qualification of Advanced Nuclear Materials
E-21: High-Throughput Study of Temperature Effects on Void Swelling in Ion Irradiated SS304
Electron Energy Loss Spectroscopy (EELS) Characterization of Fuel Cladding Chemical Interaction (FCCI) Region in U-Zr Metallic Fuel Cladded with HT-9
Harnessing HFIR Neutron Irradiations: Innovative Experiments and Standardized Capabilities
Increasing Ion Irradiation Sample Throughput with Gas Implantation Gradients
INL’s Holistic Approach to Post-irradiation Examination of Nuclear Fuel Systems
INL’s Mission Incorporating Neutrons in Post-irradiation Examination of Nuclear Materials
Investigating Water Ice Under Ion Irradiation for Future Exploration of Europa
Irradiation Testing of 316H Stainless Steel at Oak Ridge National Laboratory
Irradiation Vehicles for Materials Separate Effects Experiments Supporting the Tritium Modernization Program
Measurement of Hydrogen Vapor Pressure Over Two-phase Zirconium/Zirconium Hydride Material between 275°C and 400°C Under the Effects of Neutron Irradiation
Neutron Irradiation as a Function of Temperature – Experiment (NIFT-E)
Nuclear Fuel Salt Irradiation and Post-irradiation Processing Capabilities at The Ohio State University Research Reactor
Post-irradiation Examination of AGR-5/6/7 TRISO Fuel with Micro X-ray Computed Tomography
The Role of Nuclear Science User Facilities in Nuclear Energy Materials Research and Development
Ultrafast-Electron-diffraction Studies of Radiation-damaged Materials: An Example on the Melting Behavior of He-implanted W
Westinghouse Hot Cell Facility and Laboratories

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