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Meeting 2024 TMS Annual Meeting & Exhibition
Symposium Irradiation Testing: Facilities, Capabilities, and Experimental Designs
Sponsorship TMS Structural Materials Division
TMS: Advanced Characterization, Testing, and Simulation Committee
TMS: Mechanical Behavior of Materials Committee
TMS: Nuclear Materials Committee
Organizer(s) Walter G. Luscher, Pacific Northwest National Laboratory
Peter Hosemann, University of California, Berkeley
Andrew K. Hoffman, Catalyst Science Solutions
Joris Van den Bosch, SCK CEN
Brenden Heidrich, Nuclear Science User Facilities
Scope Irradiation testing is integral to the development and acceptance of materials and components intended for radiation environments. Irradiation testing addresses a broad array of concerns ranging from the validation of models describing irradiated material behavior to providing proof-of-concept information to justify further development by industry or acceptance by regulatory authorities. Nuclear energy production frequently drives the need for developing materials with enhanced irradiation performance. Current material development efforts include those intended for established reactor designs as well as those being considered for use in either fusion or advanced reactor concepts. Outside the reactor, additional structures, such as those in spent fuel pools and storage casks, must also withstand the challenges posed by long-term exposure to radiation during subsequent spent fuel handling, storage, and disposal. Beyond energy production, irradiation testing can also help develop and refine isotope production processes as well as shielding requirements. These efforts support a wide array of applications ranging from medical diagnostics and scientific research to enhancing worker safety and prolonging space exploration missions, respectively. Irradiation testing is clearly a critical aspect of material development and a wide array of test capabilities are required. The aim of this symposium is to highlight facilities with irradiation testing capabilities that enable data collection from materials exposed to neutron, proton, ion, or gamma irradiation. Topics of interest for this symposium include irradiation vehicle design, in-situ monitoring and control, irradiation facility capabilities, experimental design, and post-irradiation examination capabilities. Test vehicle designs used to support drop-in or instrumented lead experiments in materials research reactors are of interest as well as the experimental configurations used to facilitate beamline irradiations. Active and passive methods of monitoring and controlling key parameters, such as temperature and flux, are also interest. Finally, methods of obtaining data from experiments either during irradiation (e.g., in-situ data collection) or from post-irradiation examination are also of interest. This symposium is intended to bring together national laboratory, university, and nuclear industry researchers from around the world to discuss the current capabilities and challenges associated with the design and execution of irradiation experiments.
Abstracts Due 07/15/2023
Proceedings Plan Planned:

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 410C
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 275C and 400C 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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