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Meeting 2020 TMS Annual Meeting & Exhibition
Symposium Accelerated Materials Evaluation for Nuclear Applications Utilizing Irradiation and Integrated Modeling
Presentation Title Microstructure of HT-9 Cladding After fuel-cladding Chemical Interaction with an Annular U-10Zr Fuel Irradiated to 3.3% FIMA
Author(s) Xiang Liu, Luca Capriotti, Tiankai Yao, Jason M Harp, Lingfeng He
On-Site Speaker (Planned) Xiang Liu
Abstract Scope U-Zr fuel and ferritic/martensitic HT-9 cladding is the primary fuel system for fast reactors. Although a low smear density (~75%) was often employed to avoid the premature mechanical failure of the cladding, fuel swelling eventually brings the fuel and cladding into contact and leads to fuel-cladding chemical interaction (FCCI). FCCI is a limiting factor that restricts the fuel performance at high burnups. Annual fuel forms are being investigated due to their advantages in back end fuel cycle. Here, we investigated the FCCI layer and nearby cladding regions of an annular U-10Zr fuel with HT-9 cladding irradiated to 3.3% FIMA. Significant amount of fission products diffused into the cladding and formed intergranular precipitates, whereas the outgoing C diffusion led to decarbonization of the cladding. In the FCCI layer, a fine-grained U,Fe-rich phase (wastage), with a high number density of ~15 nm intragranular voids and ~50 nm intergranular voids was found.
Proceedings Inclusion? Undecided


A Comparative Study of Two Nanoindentation Approaches for Assessing Mechanical Properties of Ion-irradiated Stainless Steel 316
A Comparison of Ring Pull and Axial Tensile Tests of HT-9 and 14YWT Thin-walled Tubes
A Virtual Experiment Approach to Positron Annihilation Spectroscopy
Alpha Self-Irradiation of Archive and Irradiated Fast Reactor Fuels
Analysis of the Oxide Nanoparticles Trapping Behavior in an ODS Eurofer Steel by Means of Positron Annihilation Spectroscopy
Behaviors of Implanted Xe and Kr Gas Bubbles in CeO2 After Annealing and Rapid Heating Test
Changes in the Starting Microstructures of U-Mo Fuels due to the Effects of Neutron Irradiation
Characterization of Helium Implanted Single Crystal Titanium
Comparison of Radial Microstructural Changes in Fast Reactor MOX Fuels Across Varying Burnup Profiles
Controlling Helium Morphology in Pure Metals: Toward Uniform Samples for the Accelerated Measurement of Bulk Irradiated Properties
Damage Mitigation Strategies for Femtosecond Laser Machining of Micro-tensile Bars
Defect Evolution and Radiation Resistance of Advanced Fusion Materials Under Heavy Ion and Low Energy Helium Irradiation
Development of Advanced Low N-12Cr (wt.%) Ferritic/Martensitic Steel for Reactor Applications
Diffusion Analysis of Metallic Fission Products in Tristructural-isotropic Coated Fuel Using Representative Diffusion Couples
Diffusion of Fission Products in Virgin Nuclear Graphite
Direct Compaction of Dispersion Fuels Using a Matrix Deposition on the Fuel Particles
Direct Measurement of Radiation Damage Through the Energy Stored in Defects: Simulations and Experiments
Effective Defect Sinks in Metallic Composite with Nanodispersoids: In situ Ion Radiation Transmission Electron Microscopy and Position Annihilation Lifetime Spectroscopy
Effects of Helium Ion Irradiation on Single Crystal Vanadium
Fabrication and Characterization of Massive Crack-free Delta Phase-zirconium Hydride for High-performance Moderator Application
Fabrication of Low-enriched Uranium Dispersion Targets with a High Uranium Density for Mo-99 Production
High-Throughput Synthesis and Ion Irradiation of High-Entropy Alloys using Additive Manufacturing
Impact of Ionization Effects and Defect Trapping on Microstructure Evolution in Light Ion Irradiated Uranium Dioxide
In-situ Heavy Ion Irradiation of FCC and BCC High-entropy Alloys at Cryogenic and High Temperatures
In-situ Neutron Characterization of Advanced Nuclear Fuels - The Road to a New Neutron Irradiation Testing Capability
In-situ Observation of Radiation-induced Phase Transformation in U-Mo
In-situ Studies on the Mechanical Properties of He Ion Irradiated Nanotwinned Ag
Interphase distribution behavior of oxide nanoparticles triggered by isothermal ferrite transformation in 9Cr ODS steels
Irradiation Behavior of Mechanically Processed Zr-Nb Multilayers at Very High Doses
Kinetic Study on the Evolution of Nano-ceramic Coatings Under Heavy Ions Irradiation
Linking Defect Structure and Property Evolution in Ion-irradiated Tungsten: A Multi-facetted View
Mechanical Properties of Ion Irradiated and Helium Implanted HT9 Micropillars
Microstructural and Micro-chemical Characterization of Safety Tested TRISO UCO Fuel Kernels Irradiated in the Advanced Test Reactor
Microstructural Changes and Corrosion of Proton-pre-irradiated Hastelloy N in FLiNaK Molten Salt
Microstructure and Mechanical Behavior of Directed Energy Deposition Laser Additively Manufactured T-91
Microstructure of HT-9 Cladding After fuel-cladding Chemical Interaction with an Annular U-10Zr Fuel Irradiated to 3.3% FIMA
Multiple Scale Mechanical Testing of Neutron Irradiated FeCrAl Alloys
Neutron Irradiation Damage and Fission Product Transport in the SiC Layer of TRISO Fuel Particles
On a Theory Based Accelerated Testing Methodology for Swelling.
On the Role of Heterogeneity in Concentrated Solid‒solution Alloys in Enhancing their Irradiation Resistance
Promotion and Suppression of the G-phase in Steels
Radiation Response of HT9 Ferritic/Martensitic Alloys as a Function of Interstitial Content
Rapid Investigation of Irradiation Temperature Sensitivity Using Charged Particles
Recent Applications of Ex-situ Transient Grating Spectroscopy to the Study of Radiation-induced Degradation of Nuclear Materials
Small Scale Mechanical Testing of Ceramic Interfaces in Nuclear Materials: Characterizing the Impact of Elastic Mismatch on Stress Intensity and Property Extraction
Synthesis of Intermetallic UZr2+x and its Phase Transformation
Temperature Shift Evaluation for G-phase Clustering in Ferritic-martensitic Alloys
Testing of Nuclear Fuels and Materials in the Advanced Fuels Campaign
Three-dimensional Analysis of the IPyC/SiC Interface in Irradiated TRISO Fuel Particles

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