Projects
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NSUF 17-1115: Heavy ion irradiation and ex situ transmission electron microscopy study of the effectiveness of twin boundaries in alleviating radiation damage in 316 austenitic stainless steels
Interfaces can act as radiation defect sinks and thus reduce defect density...
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NSUF 17-1115: Heavy ion irradiation and ex situ transmission electron microscopy study of the effectiveness of twin boundaries in alleviating radiation damage in 316 austenitic stainless steels ansehen -
NSUF 17-1116: Investigation of irradiation-induced recrystallization in U-Mo fuel
Microstructural changes, characterized by grain subdivision and increased...
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NSUF 17-1116: Investigation of irradiation-induced recrystallization in U-Mo fuel ansehen -
NSUF 17-1118: IVEM investigation of defect evolution in FCC and BCC HEAs during heavy ion irradiation
The proposed study will investigate the microstructural evolution of two...
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NSUF 17-1118: IVEM investigation of defect evolution in FCC and BCC HEAs during heavy ion irradiation ansehen -
NSUF 17-1119: Characterization of Oxide Layer on the Surface of High Temperature Water Corroded Zircaloy-4 In the Presence of Neutron+Gamma and Gamma Only
Although there is a correlation between autoclave and in-reactor zirconium...
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NSUF 17-1119: Characterization of Oxide Layer on the Surface of High Temperature Water Corroded Zircaloy-4 In the Presence of Neutron+Gamma and Gamma Only ansehen -
NSUF 17-1121: Characterization of the Stability of the Microstructure of Novel ODS Alloys
The objective of this work is to characterize the irradiated microstructure...
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NSUF 17-1121: Characterization of the Stability of the Microstructure of Novel ODS Alloys ansehen -
NSUF 17-1122: Enhanced irradiation tolerance of high-entropy alloys
Efforts are undergoing to extend the life of current reactors and develop...
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NSUF 17-1122: Enhanced irradiation tolerance of high-entropy alloys ansehen -
NSUF 17-728: Radiation-Hardening and Microstructural Stability of NF709 Austenitic Stainless Steel
Austenitic stainless steel NF709 has been down-selected for development as a...
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NSUF 17-728: Radiation-Hardening and Microstructural Stability of NF709 Austenitic Stainless Steel ansehen -
NSUF 17-880: Mechanical characterization of neutron irradiated FSW ODS alloys
One of the goals of the Fuel Cycle R&D program is the development of...
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NSUF 17-880: Mechanical characterization of neutron irradiated FSW ODS alloys ansehen -
NSUF 17-905: Enhancing radiation tolerance through increasing alloy complexity.
Project Objectives: The development of strategies to design structural alloys...
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NSUF 17-905: Enhancing radiation tolerance through increasing alloy complexity. ansehen -
NSUF 17-906: Radiation Tolerance of Friction Stir Welded Ferritic Oxide Dispersed Steel under Ion Irradiation
The objective of this work is to examine the effect of high damage levels of...
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NSUF 17-906: Radiation Tolerance of Friction Stir Welded Ferritic Oxide Dispersed Steel under Ion Irradiation ansehen -
NSUF 17-908: Radiation induced segregation and phase separation in neutron irradiated FeCrAl alloys
The objective of this project is to understand the radiation induced...
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NSUF 17-908: Radiation induced segregation and phase separation in neutron irradiated FeCrAl alloys ansehen -
NSUF 17-909: Microstructural characterization of 3% burn-up MOX fuel
Attaining fundamental understanding of fuel performance requires detailed...
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NSUF 17-909: Microstructural characterization of 3% burn-up MOX fuel ansehen -
NSUF 17-912: Investigation of gas bubble behavior under ion irradiation
The development of He and Xe bubbles in structural and fuel materials could...
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NSUF 17-912: Investigation of gas bubble behavior under ion irradiation ansehen -
NSUF 17-917: Pore size distribution in U-Mo fuel irradiated to high burnup
Attaining fundamental understanding of fuel performance requires detailed...
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NSUF 17-917: Pore size distribution in U-Mo fuel irradiated to high burnup ansehen -
NSUF 17-922: Effect of grain boundary character and surface treatment on irradiation tolerance of nuclear alloys
The objective of this project is to determine the influence of the grain...
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NSUF 17-922: Effect of grain boundary character and surface treatment on irradiation tolerance of nuclear alloys ansehen -
NSUF 17-923: a' precipitation in neutron-irradiated Fe-9/12Cr alloys
High chromium ferritic/martensitic (F-M) steels are one of the strong...
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NSUF 17-923: a' precipitation in neutron-irradiated Fe-9/12Cr alloys ansehen -
NSUF 17-929: Radiation Stability Study on Nuclear Waste/Spent Fuel Materials
Since radiation damage and decay heat are significant issues for nuclear...
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NSUF 17-929: Radiation Stability Study on Nuclear Waste/Spent Fuel Materials ansehen -
NSUF 17-930: Characterization of ion irradiated 15-15Ti steel by APT
This work proposes APT analysis of self-ion irradiated 15-15Ti cladding steel...
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NSUF 17-930: Characterization of ion irradiated 15-15Ti steel by APT ansehen -
NSUF 17-933: Sample Preparation for Ex-situ Transmission Electron Microscopy Study of Deformation-induced Twinning and Martensite in Two 316L Austenitic Stainless Steels: Role of Stacking Fault Energy and Grain Orientation
A preliminary experiment was conducted at Oak Ridge National Laboratory’s...
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NSUF 17-933: Sample Preparation for Ex-situ Transmission Electron Microscopy Study of Deformation-induced Twinning and Martensite in Two 316L Austenitic Stainless Steels: Role of Stacking Fault Energy and Grain Orientation ansehen -
NSUF 17-935: In situ ion irradiation of second phase particles in zirconium fuel cladding
Zirconium alloys are universally used as fuel cladding and support structures...
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NSUF 17-935: In situ ion irradiation of second phase particles in zirconium fuel cladding ansehen