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Meeting 2024 TMS Annual Meeting & Exhibition
Symposium Irradiation Testing: Facilities, Capabilities, and Experimental Designs
Presentation Title Ultrafast-Electron-diffraction Studies of Radiation-damaged Materials: An Example on the Melting Behavior of He-implanted W
Author(s) Mianzhen Mo, Ling Wang, Thies Albert, Alfredo Correa, Zhijiang Chen, Leora Dresselhaus-Marais, Mungo Frost, Nicholas Hartley, Laurenz Kremeyer, Matthias Kling, Emma McBride, Samuel Murphy, Benjamin Ofori-Okai, Alexander Hume Reid, Adam Summers, Klaus Sokolowski-Titen, Xiaozhe Shen, Artur Tamm, Yongqiang Wang, Xueli Zheng, Siegfried Glenzer
On-Site Speaker (Planned) Mianzhen Mo
Abstract Scope The development of theories for predicting radiation damage across multiple length and time scales plays a vital role in materials design for future fusion reactors. This is a scientific area with an urgent need for experimental data that can validate modelling assumptions. However, existing techniques for irradiation testing are constrained by temporal resolutions and cannot reach ultrafast time scales required to understand the atomic dynamics in radiation-damaged materials. To address this challenge, we apply the technique of ultrafast electron diffraction (UED) to visualize the structural dynamics of radiation-damaged materials on atomic time and length scales, and understand how radiation defects change the material properties and performance under fusion reactor conditions. We conduct the experimental studies in the MeV-UED instrument of LCLS at SLAC. As an example, we will showcase the results of studying the melting behavior of W with He implantation. Molecular dynamics simulation results will be also presented.
Proceedings Inclusion? Planned:
Keywords Characterization, Nuclear Materials, Phase Transformations

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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