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Meeting Materials Science & Technology 2020
Symposium Advanced Characterization of Materials for Nuclear, Radiation, and Extreme Environments
Presentation Title In Situ Observation of Irradiation Damage in Polycrystalline Nuclear Graphite
Author(s) Dong Liu, David Cherns, Joshua Kane, Weiying Chen, Steve Jones, Karthik Chinnathambi, William Windes
On-Site Speaker (Planned) Dong Liu
Abstract Scope Understanding irradiation induced damage in nuclear graphite is of particular relevance to operating AGRs and some GenIV designs. Ex situ C+ implantation (Surrey Ion Beam Centre, UK) and in situ Ar+ irradiation (IVEM, ANL, USA) were used to simulate neutron damage over a range of temperatures up to 1000C. The surface morphological changes as well as residual stresses introduced by carbon ions were characterized and the decrease in crystallite size was compared to samples with the same neutron doses. It was found that large amount of porosity was introduced due to irradiation which has subsequently contributed to the volume change. The Ar+ irradiation carried out in situ within a TEM showed that basal plane dislocation movement was the primary cause of the macro-scale behavior. A novel mechanistic model was proposed based on these observations and it is considered to be applicable to all materials with nano-scale 2D graphite-like polycrystalline structures.
Proceedings Inclusion? Planned: Publication outside of MS&T


In Situ Observation of Short- and Long-Timescale Material Property Evolution Under Extreme Conditions
Analysis of Heavy Ion Irradiation Damage in Commercially Pure Titanium and Titanium Alloys
Benefits of Using High Energy Ions in Ion Irradiation Experiments to Evaluate Void Swelling
Characterization of Microstructure Evolution in Ceramic Materials Using Acoustic and Thermal Transport Measurements
Characterization of Stress and Environment Dependent Fracture Mechanisms of SiC/SiC CMCs
Controlling Helium Morphology in Pure Metals: Effects of Helium Defects on Deformation and Strength
Corrosion Control of Austenitic Stainless Steel and Nickel-Based Alloys in Molten Chloride Salt Environments
Design of a Hot Hydrogen Test Loop for Testing of Nuclear Thermal Rocket Elements
Development of a Combined Thermal Hydraulic and Materials Corrosion Liquid-Sodium Experimental Facility
Development of an In-Situ Mechanical Test System for Advanced Reactor Coolants
Effects of Post-Processing Variability on Radiation Response of Additively-Manufactured HT9
Enabling In-situ Crack Growth Testing and Monitoring in VTR Cartridge Loop Environments
Explaining the Corrosion Morphology of Structural Materials in Molten Fluoride Salts With/Without Radiation
Fundamental In-situ Experiments Coupled to High-throughput Approaches to Understand Radiation Damage in FCC and BCC Compositionally Complex Alloys
In situ Crack Loading and Measurement Techniques for Gen IV Reactor Coolant Media
In Situ Observation of Irradiation Damage in Polycrystalline Nuclear Graphite
Investigation of Uranium Silicide Fuel Form Additions through Rietveld Refinement and Internal Standard P-XRD
Mobility of Hydrogen in YH2 Probed by Nuclear Magnetic Resonance
Modified stereo TEM for 3D analysis of defects
Nanomechanical Change of Tungsten in ELM Conditions
Nonlinear Ultrasound for Nondestructive Evaluation of Microstructural Defects
Response of an Additively Manufactured 316 Stainless Steel Subjected to High Temperature Heavy Ion Irradiations
Selective Irradiation Behavior in Dual Phase 308L Filler of SA508-304L Dissimilar Metal Weldment after Proton Irradiation
STEM Characterization of Dislocation Loops for an Irradiated Model FCC Alloy
Study on Hydrogen Isotopes Solubility and Diffusivity in Y- and Co-doped Barium-zirconates Using Tritium Imaging Plate Technique
Swelling of Nuclear Reactor Steels: Modeling, Theory, and Accelerated Testing
Unveiling High Temperature Damage Mechanisms via In-situ Digital Image Correlation of Chromium-coated Zirconium-based Fuel Claddings

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