Projects
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NSUF 24-5122: Probing the effect of specific chemical elements on the irradiation induced defects formation and evolution in Multi-Principal Element Alloys (MPEAs) at high temperatures
In situ ion irradiations will be conducted on two families of high entropy...
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Pregledaj NSUF 24-5122: Probing the effect of specific chemical elements on the irradiation induced defects formation and evolution in Multi-Principal Element Alloys (MPEAs) at high temperatures -
NSUF 24-5129: Visualizing the impact of irradiation damage on alloy element redistribution accompanying Zr alloy corrosion via atom probe tomography
Dilute Zr alloys (e.g., Zircaloy-4, ZIRLO) are widely used as nuclear fuel...
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Pregledaj NSUF 24-5129: Visualizing the impact of irradiation damage on alloy element redistribution accompanying Zr alloy corrosion via atom probe tomography -
NSUF 24-5130: Unveiling Long-Term Irradiation and Thermal Aging Effects on Solute Segregation in Commercial Reactor Pressure Boundary Material
We will seek in both the pressurizer and surveillance material to select APT...
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Pregledaj NSUF 24-5130: Unveiling Long-Term Irradiation and Thermal Aging Effects on Solute Segregation in Commercial Reactor Pressure Boundary Material -
NSUF 24-5134: Nanostructural Analysis of Cold-Sprayed Cr-Coated Zircaloy-4 Under Proton Irradiation and Corrosion
The cold-sprayed Cr coating on Zircaloy-4 is a promising, cost-effective...
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Pregledaj NSUF 24-5134: Nanostructural Analysis of Cold-Sprayed Cr-Coated Zircaloy-4 Under Proton Irradiation and Corrosion -
NSUF 24-5135: Quantification of Radial Constituent Redistribution in Annular U-10Zr Irradiated Fuel Using EPMA
The objective of the proposed work is to determine the composition of each...
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Pregledaj NSUF 24-5135: Quantification of Radial Constituent Redistribution in Annular U-10Zr Irradiated Fuel Using EPMA -
NSUF 24-5142: In-situ TEM study of microstructural evolution and deformation in FFTF irradiated HT-9 cladding
HT-9, a tempered martensitic (TM) alloy with a nominal composition of...
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Pregledaj NSUF 24-5142: In-situ TEM study of microstructural evolution and deformation in FFTF irradiated HT-9 cladding -
NSUF 24-5145: Natural zirconolites and pyrochlores as test cases for wasteform matrices
Minerals with zirconolite or pyrochlore structures have been proposed as host...
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Pregledaj NSUF 24-5145: Natural zirconolites and pyrochlores as test cases for wasteform matrices -
NSUF 24-5146: Irradiation effects on thermal conductivity and phase evolution in low burnup U-10Pu-10Zr
The objective of this experiment is to characterize the microstructure and...
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Pregledaj NSUF 24-5146: Irradiation effects on thermal conductivity and phase evolution in low burnup U-10Pu-10Zr -
NSUF 24-5155: Formation and dissolution of carbides and precipitates in self-ion irradiated HT9 alloy, and the impact on its mechanical properties
We have recently shown that ion irradiation can induce the formation of...
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Pregledaj NSUF 24-5155: Formation and dissolution of carbides and precipitates in self-ion irradiated HT9 alloy, and the impact on its mechanical properties -
NSUF 24-5157: TEM determination of dislocation structure formation at 4% tensile deformation in a neutron irradiated Fe-9Cr Model Alloy
This work will build on in situ tensile deformation in the ANL Advanced...
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Pregledaj NSUF 24-5157: TEM determination of dislocation structure formation at 4% tensile deformation in a neutron irradiated Fe-9Cr Model Alloy -
NSUF 24-5164: Investigating effect of dose rate on microstructure evolution in 800H alloy at high doses
The proposed Rapid Turnaround Experiment (RTE) aims at elucidating the effect...
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Pregledaj NSUF 24-5164: Investigating effect of dose rate on microstructure evolution in 800H alloy at high doses -
NSUF 24-5166: In-situ TEM study of irradiation effects in zirconium oxides
Zirconium alloys have been widely used as fuel cladding materials in light...
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Pregledaj NSUF 24-5166: In-situ TEM study of irradiation effects in zirconium oxides -
NSUF 24-5167: Comparative Analysis of Hastelloy X and 316L Stainless Steel Under Simultaneous Irradiation and Corrosion for Advanced Reactor Applications
This research investigates the performance of Hastelloy X and 316L stainless...
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Pregledaj NSUF 24-5167: Comparative Analysis of Hastelloy X and 316L Stainless Steel Under Simultaneous Irradiation and Corrosion for Advanced Reactor Applications -
NSUF 24-5173: High-throughput ion irradiation approach for temperature-dependent swelling measurement in additively manufactured 316L alloy
The project will develop a high-throughput testing approach that allows for...
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Pregledaj NSUF 24-5173: High-throughput ion irradiation approach for temperature-dependent swelling measurement in additively manufactured 316L alloy -
NSUF 24-5177: In-Situ TEM Nanomechanical Testing of Neutron Irradiated U-10Zr
Based on current understanding, irradiated U-10Zr fuels have four concentric...
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Pregledaj NSUF 24-5177: In-Situ TEM Nanomechanical Testing of Neutron Irradiated U-10Zr -
NSUF 24-5181: Nano indentation hardness of U10Zr
This proposal aims to investigate the mechanical properties of irradiated...
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Pregledaj NSUF 24-5181: Nano indentation hardness of U10Zr -
NSUF 24-5182: Mechanical and microstructural characterization of neutron irradiated ultra-fine-grained and conventional ferritic steel
The major objective of the proposed work is to characterize the effect of...
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Pregledaj NSUF 24-5182: Mechanical and microstructural characterization of neutron irradiated ultra-fine-grained and conventional ferritic steel -
NSUF 24-MOD1A: MOD1A Yttrium Hydride
The use and properties of YHx as a high-temperature moderator have been...
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Pregledaj NSUF 24-MOD1A: MOD1A Yttrium Hydride -
NSUF 24-TTUSC: Tensile Testing Utilizing the Standard Capsule (TTUSC)
The TTUSC samples were irradiated in support of LDRD 20A44-046FP...
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Pregledaj NSUF 24-TTUSC: Tensile Testing Utilizing the Standard Capsule (TTUSC) -
NSUF 24-Zion: Harvested Reactor Pressure Vessel Material from Zion Unit 1
The decommissioning of the Zion Units 1 and 2 Nuclear Generating Station in...
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Pregledaj NSUF 24-Zion: Harvested Reactor Pressure Vessel Material from Zion Unit 1