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Meeting 2024 TMS Annual Meeting & Exhibition
Symposium Irradiation Testing: Facilities, Capabilities, and Experimental Designs
Presentation Title Measurement of Hydrogen Vapor Pressure Over Two-phase Zirconium/Zirconium Hydride Material between 275°C and 400°C Under the Effects of Neutron Irradiation
Author(s) Kenneth J. Geelhood, Samuel Goodrich, Travis Zipperer, Eric Choi, Walter Luscher, Corey Hines, Hillary Bennett
On-Site Speaker (Planned) Kenneth J. Geelhood
Abstract Scope The epithermal beamline at Washington State University’s Dodgen Research Facility was utilized to interrogate the influence of neutron irradiation on vapor pressure over hydrided getters. Neutrons produced in WSU’s TRIGA Reactor were collimated and reduced to a 2-inch beam prior to interacting with the hydrided getter sample inside the measurement apparatus. Samples evaluated in this study consisted of nickel-plated zirconium alloy tubes hydrided to an H/Zr ratio of 0.4. Previously established pressure-composition-temperature diagrams provide the temperature dependence of the vapor pressure exerted by hydrided zirconium at relatively high temperatures (> 500°C). Low-temperature data are more challenging to obtain due to the corresponding lower pressures and the influence of neutron irradiation is unknown. Although it was hypothesized that neutron irradiation may enhance vapor pressure, data measured in this study show indicate vapor pressures are not increased by an epithermal neutron flux of 1x109 n/cm2-s in the temperature range of 275°C -400°C.
Proceedings Inclusion? Planned:
Keywords 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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