Projects
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NSUF 19-1663: Ion irradiation of ThO2 and UO2 single crystals
The technical objective of this RTE is to generate dislocation loop dominated...
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瀏覽 NSUF 19-1663: Ion irradiation of ThO2 and UO2 single crystals -
NSUF 19-1665: Defect Clustering in 316H Stainless Steel and High Entropy Alloy Under In-situ Irradiation at 600-700°C
Despite the promising results about the irradiation resistance of high...
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瀏覽 NSUF 19-1665: Defect Clustering in 316H Stainless Steel and High Entropy Alloy Under In-situ Irradiation at 600-700°C -
NSUF 19-1666: Nanoindentation of Phases in Irradiated and Control U-10Zr Fuels
A subset of the Mechanistic Fuel Failure-3 (MFF-3) irradiations included...
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瀏覽 NSUF 19-1666: Nanoindentation of Phases in Irradiated and Control U-10Zr Fuels -
NSUF 19-1670: A study of the tensile response of HT-9 alloys following ATR irradiation to doses between 0.01 and 10 dpa at 300, 450 and 550C
The objective of this program is to investigate the tensile properties of...
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瀏覽 NSUF 19-1670: A study of the tensile response of HT-9 alloys following ATR irradiation to doses between 0.01 and 10 dpa at 300, 450 and 550C -
NSUF 19-1672: Microstructure characterization on neutron irradiated and post-tensile duplex stainless steels
The work is planned to systemically understand the microstructural evolution...
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瀏覽 NSUF 19-1672: Microstructure characterization on neutron irradiated and post-tensile duplex stainless steels -
NSUF 19-1674: Radiation tolerance of Ln3TaO7 weberite-type nuclear waste matrix materials
In this project, 1 MeV Kr ions will be used to irradiate La3TaO7, Sm3TaO7,...
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瀏覽 NSUF 19-1674: Radiation tolerance of Ln3TaO7 weberite-type nuclear waste matrix materials -
NSUF 19-1676: Thermal diffusivity and microstructure analysis of in-core molten salt irradiated graphite
From the success of the Molten Salt Reactor Experiment (MSRE) operated at Oak...
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瀏覽 NSUF 19-1676: Thermal diffusivity and microstructure analysis of in-core molten salt irradiated graphite -
NSUF 19-1680: In-situ small-scale mechanical testing of fast reactor advance metallic fuel alloy.
The main objective of this project is the determination of hardness, elastic...
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瀏覽 NSUF 19-1680: In-situ small-scale mechanical testing of fast reactor advance metallic fuel alloy. -
NSUF 19-1681: Effects of dose and temperature on microstructural evolution of Zircaloy-4 alloys during proton irradiation
Proton irradiation is commonly used to mimic neutron irradiation since it is...
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瀏覽 NSUF 19-1681: Effects of dose and temperature on microstructural evolution of Zircaloy-4 alloys during proton irradiation -
NSUF 19-1683: Resolving the Puzzle of Flux Effects on High Fluence Precipitation and Embrittlement of RPV Steels
This proposal covers characterization 5 alloys by Atom Probe Tomography (APT)...
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瀏覽 NSUF 19-1683: Resolving the Puzzle of Flux Effects on High Fluence Precipitation and Embrittlement of RPV Steels -
NSUF 19-1687: Mechanical characterization of three lower dose HT-9 heats (ORNL, LANL and EBR II) after side-by-side neutron irradiation at LWR and fast reactor relevant temperatures
HT-9 is being considered as a candidate structural material for advanced...
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瀏覽 NSUF 19-1687: Mechanical characterization of three lower dose HT-9 heats (ORNL, LANL and EBR II) after side-by-side neutron irradiation at LWR and fast reactor relevant temperatures -
NSUF 19-1691: Thermal Driven Grain Growth and Fission Gas Bubble Coarsening in Nano-grain Sized U3Si2
U3Si2 is an advanced fuel form with enhanced thermal conductivity and higher...
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瀏覽 NSUF 19-1691: Thermal Driven Grain Growth and Fission Gas Bubble Coarsening in Nano-grain Sized U3Si2 -
NSUF 19-1692: He++ Irradiation of Aerosol Jet Printed Silver Structures
A fundamental challenge that this project intends to examine is the...
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瀏覽 NSUF 19-1692: He++ Irradiation of Aerosol Jet Printed Silver Structures -
NSUF 19-1693: EBSD characterization of neutron irradiated mineral concrete aggregates
The Radiation Induced Volumetric Expansion (RIVE) may produce cracking and...
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瀏覽 NSUF 19-1693: EBSD characterization of neutron irradiated mineral concrete aggregates -
NSUF 19-1694: In-situ irradiation study of carbides/nitrides/carbo-nitrides in additively manufactured ferritic-martensitic steels.
Ferritic-martensitic (F-M) steels are promising structural material...
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瀏覽 NSUF 19-1694: In-situ irradiation study of carbides/nitrides/carbo-nitrides in additively manufactured ferritic-martensitic steels. -
NSUF 19-1695: TEM Characterization of Highly Irradiated Stainless Steel
The microstructural evolution of key reactor components, and the associated...
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瀏覽 NSUF 19-1695: TEM Characterization of Highly Irradiated Stainless Steel -
NSUF 19-1698: Mechanical property and microstructural characterization of irradiated stainless steel via in situ SEM-EBSD mechanical testing.
Metallic systems exposed to high loads and elevated temperatures are...
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瀏覽 NSUF 19-1698: Mechanical property and microstructural characterization of irradiated stainless steel via in situ SEM-EBSD mechanical testing. -
NSUF 19-1700: In-Situ Observation of Radiation-Induced Phase Transformation in U-Mo
It was speculated that recrystallization in U-Mo fuels may be related to the...
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瀏覽 NSUF 19-1700: In-Situ Observation of Radiation-Induced Phase Transformation in U-Mo -
NSUF 19-1720: Characterization of alpha irradiated and control cementitious grouts / grout components used for nuclear waste encapsulation
This project aims to investigate changes in cement microstructure and...
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瀏覽 NSUF 19-1720: Characterization of alpha irradiated and control cementitious grouts / grout components used for nuclear waste encapsulation -
NSUF 19-1721: Nanoscale analysis of Mn-Si-Ni phase in neutron irradiated T91 at 320C
The fuel cladding material of Generation IV nuclear reactors experience...
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瀏覽 NSUF 19-1721: Nanoscale analysis of Mn-Si-Ni phase in neutron irradiated T91 at 320C