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Meeting 2024 TMS Annual Meeting & Exhibition
Symposium Irradiation Testing: Facilities, Capabilities, and Experimental Designs
Presentation Title Post-irradiation Examination of AGR-5/6/7 TRISO Fuel with Micro X-ray Computed Tomography
Author(s) William Chuirazzi, Rahul Reddy Kancharla, John Stempien, Swapnil Kishor Morankar
On-Site Speaker (Planned) Swapnil Kishor Morankar
Abstract Scope Tristructral isotropic (TRISO) coated fuel particles are a fuel form under study for commercial qualification to power certain next-generation nuclear reactor designs. Nondestructive micro X-ray Computed Tomography (XCT) examinations have recently been used to complement traditional destructive analyses. While XCT results of AGR-2, AGR-3/4, and initial results on a small sample of AGR-5/6/7 particles have previously been reported, this work focuses on a more comprehensive examination of a larger sample of irradiated AGR-5/6/7 particles at Idaho National Laboratory’s Irradiated Materials Characterization Laboratory (IMCL). These results enable analysis of post-irradiation TRISO particle properties such as layer thicknesses as well as observations of defects and buffer layer delaminations. This presentation also includes application of lessons learned from previous XCT examinations to improve and expedite data collection. Finally, the impact of post-irradiation examination XCT on destructive analysis as well as XCT data's usefulness in observing fuel performance under irradiation conditions is discussed.
Proceedings Inclusion? Planned:
Keywords Characterization, Nuclear Materials,

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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