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Meeting 2016 TMS Annual Meeting & Exhibition
Symposium Accelerated Materials Evaluation for Nuclear Application Utilizing Test Reactors, Ion Beam Facilities and Integrated Modeling
Presentation Title Initial Post Irradiation Examination Results of a Novel Fuel Concept with Enhanced Thermal Properties
Author(s) Andrew M Casella, David J Senor, Edgar Buck, Mehdi Balooch, Peter Hosemann
On-Site Speaker (Planned) Andrew M Casella
Abstract Scope A new nuclear fuel concept, designed for improved thermal performance, has been fabricated at the University of California, Berkeley, irradiated in the MIT research reactor, and examined at the Pacific Northwest National Laboratory. The fuel consists of uranium zirconium hydride. The zircaloy-2 cladding has been slightly oxidized to prevent hydrogen migration from the fuel to the cladding. Additionally, the traditional helium backfill has been replaced by lead-bismuth eutectic (LBE) to improve gap thermal conductivity. In-situ temperature measurements during irradiation suggest a continuous decrease in thermal conductivity with increasing burnup. Post-irradiation examination of this novel fuel form suggests reduction of the cladding wall thickness due to a cladding-LBE reaction. The fuel pellets appear intact as evaluated by SEM/EDS, but uranium and zirconium presence in the eutectic indicates some surface erosion of the fuel has occurred.
Proceedings Inclusion? Planned: A print-only volume

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

A Synchrotron Peak Broadening and Modelling Study of Proton-Irradiated Zircaloy-2
Accelerated Irradiation for Emulation of Radiation Damage in Reactor
Accelerating Post-irradiation Examination with Latest-generation Electron Microscopy Hardware and Software
An Experimental Study to Elucidate Stage IV Recovery Mechanism of Heavy Ion Irradiated High Purity Molybdenum
Atom Probe Tomography Investigations of Reactor Pressure Vessel Steels Using High Dose Charged Particle Irradiations
Atomic Scale Characterisation of Radiation Damage in Superconducting Perovskites for Nuclear Applications
Change of Slip Anisotropy in Zr Alloys Due to Irradiation
Characterisation of Reactor Core Materials Performance Using Materials Test Reactors - A Canadian Perspective
Cluster Dynamics Modelling of Void Nucleation and Growth in Ferritic Steels
Comparison of Neutron, Proton, and Self-ion Irradiation of Fe-9%Cr ODS at 3 dpa, 500C
Correlative and Dynamic S/TEM Characterization of Heavily Irradiated Pyrochlores and Fluorites
Direct Observation of Radiation Response in Ni and Ni-base Concentrated Solid-solution Alloys
Effect of Helium Implantation Mode on Void Formation in Ion-Irradiated T91 Steel
Effects of Electronic Energy Loss on Damage Evolution in Ion-irradiated Ceramics
Energy Dissipation and Defect Evolution in Concentrated Solid-solution Alloys
Evaluation of Radiation effects in FeMnNiCr High Entropy Alloy
Evidence of Accelerated Oxide Dissolution during Irradiation-Corrosion of 316L Stainless Steel in Primary Water
Finite Element Analysis of Micro-cantilever Beam Experiments in UO2
High-energy Synchrotron Radiation Study of Heavy Ion Irradiated U-Mo/Al Dispersion Fuel
In-situ High-Energy X-ray Study of Neutron Irradiation Effect on Tensile Deformation Behavior of an Fe-Cr Model Alloy
In-Situ Measurement of Tritium Released from Gamma-LiAlO2 Pellets Irradiated in the Advanced Test Reactor
In Situ Corrosion Studies of Nuclear Claddings in Extreme Environments
Influence of Microstructural Features on Void Evolution in Self-Ion Irradiated HT9 at Very High Dose
Initial Post Irradiation Examination Results of a Novel Fuel Concept with Enhanced Thermal Properties
Ion Irradiation Damage in Ferritic/Martensitic Steel T91
Ion Irradiation Induced Defect Evolution in Ni and Ni-Based FCC Binary Alloys
Ion Irradiation of Thin Foils: Mechanisms, Modeling, and Prediction of Neutron Damage
Mechanical Behavior of UO2 at Sub-Grain Length Scales: A Quantification of Creep Properties via High Temperature Mechanical Testing
Microstructural and Nanomechanical Characteristics of an Ion-Irradiated Lanthana-Bearing Nanostructured Ferritic Steel
Microstructural Characterization of ATR Irradiated Cu/Nb Nanolayered Composites
Modeling Microstructural Evolution in Neutron Irradiated Tungsten during Isochronal Annealing Process
Multiscale Modeling of Defect Cluster Evolution in Irradiated Structural Materials
Nanoindentation and In Situ Microcompression Testing in Various Dose Regimes of Proton-beam Irradiated 304 SS
Noble Gas Behavior in Nuclear Fuel and Ceramic Nuclear Waste Forms
Non-contact Analysis of Dislocation Effects in Single Crystal Niobium and Vacancy Effects in Intermetallic NiAl
Non-contact Determination of Ion Irradiation Effects in Pure Polycrystalline Copper
Observed U-Mo Alloy Microstructures After Irradiation in the Advanced Test Reactor
On a Precipitation Damage Meter to Quantify Dose Rate and Damaging Particle Effects on Ion and Neutron Irradiated RPV Steels
Optimization of the Composition of FeCrAl Alloys for Radiation Environments
Oxidation of FeCrAl Alloys in Simulated PWR Environments during In-situ Proton Irradiation
Phase Field Modeling of Void Growth and Coarsening in Irradiated Materials
Production of Microstructure to Mimic Key Effects of Neutron Irradiation Damage in Core Materials
Self-ion Irradiation Induced Dispersoid Instabilities and Dispersiod-defect Interactions in ODS Alloys
Solute Redistribution Processes in Neutron-irradiated Model FeCrAl Alloys
Structural Characterization of Nanoscale Intermetallic Precipitates in Highly Neutron Irradiated Reactor Pressure Vessel Steels
Suppression of Void Nucleation during Self-ion Irradiation by Interaction of Injected Interstitial Effect and Ion Beam Rastering
TEM Characterization of Neutron-irradiated Cast Austenitic Stainless Steel at 320C to 0.08 dpa
The Effect of Pre-implanted Helium on Void Incubation and Growth in Ferritic-Martensitic Steels
Thermal Aging and Low Dose Neutron Irradiation Effect on the Microstructural Stability of Delta Ferrite in a 308L Weld
Utilizing Sandia’s In-situ Ion Irradiation TEM to Elucidate Governing Mechanisms in Complex Environments
Y-10: Swift Heavy Ion Irradiation Damage in Ti-6Al-4V: Characterization of the Microstructure and Mechanical Properties
Y-11: X-ray Micro-computed Tomography for Nondestructive Examination of Nuclear Materials
Y-1: A Combined Radiation and Corrosion Experiment for Molten Salt Reactor (MSR)
Y-2: Comparison of Nanoindentation, Microhardness, and Tensile Testing on Neutron Irradiated Ferritic/Martensitic Steels
Y-3: Effects of Neutron Irradiation on Zr52.5Cu17.9Ni14.6Al10Ti5 (BAM-11) Bulk Metallic Glass
Y-4: Grain Boundary Character Effect on Radiation Induced Defect Distribution in Nanocrystalline Nickel and Nickel-Chromium Thin Films
Y-5: Kinetics of Defect Formation in Advanced F/M Steels Under Ion-Beam Irradiation Using In-situ TEM
Y-6: Preliminary Experiments to Develop a He-W Calibration Standard for Laser Induced Breakdown Spectroscopy
Y-7: Reexamination of the “Temperature-shift” Arising from Increases in dpa-rate during Ion Bombardment
Y-8: Room Temperature Au2+ Irradiation of Ni, Ni-Co and Ni-Fe Single Phase Alloys
Y-9: Study of Thermal Aging on Corrosion Fatigue of Z3CN20.09M Duplex Stainless Steel in High Temperature Water

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