Projects
-
NSUF 17-1116: Investigation of irradiation-induced recrystallization in U-Mo fuel
Microstructural changes, characterized by grain subdivision and increased...
0 Datasets
View NSUF 17-1116: Investigation of irradiation-induced recrystallization in U-Mo fuel -
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...
0 Datasets
View NSUF 17-1118: IVEM investigation of defect evolution in FCC and BCC HEAs during heavy ion irradiation -
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...
0 Datasets
View 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 -
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...
0 Datasets
View NSUF 17-1121: Characterization of the Stability of the Microstructure of Novel ODS Alloys -
NSUF 17-1122: Enhanced irradiation tolerance of high-entropy alloys
Efforts are undergoing to extend the life of current reactors and develop...
0 Datasets
View NSUF 17-1122: Enhanced irradiation tolerance of high-entropy alloys -
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...
0 Datasets
View NSUF 17-728: Radiation-Hardening and Microstructural Stability of NF709 Austenitic Stainless Steel -
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...
0 Datasets
View NSUF 17-880: Mechanical characterization of neutron irradiated FSW ODS alloys -
NSUF 17-905: Enhancing radiation tolerance through increasing alloy complexity.
Project Objectives: The development of strategies to design structural alloys...
0 Datasets
View NSUF 17-905: Enhancing radiation tolerance through increasing alloy complexity. -
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...
0 Datasets
View NSUF 17-906: Radiation Tolerance of Friction Stir Welded Ferritic Oxide Dispersed Steel under Ion Irradiation -
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...
0 Datasets
View NSUF 17-908: Radiation induced segregation and phase separation in neutron irradiated FeCrAl alloys -
NSUF 17-909: Microstructural characterization of 3% burn-up MOX fuel
Attaining fundamental understanding of fuel performance requires detailed...
0 Datasets
View NSUF 17-909: Microstructural characterization of 3% burn-up MOX fuel -
NSUF 17-912: Investigation of gas bubble behavior under ion irradiation
The development of He and Xe bubbles in structural and fuel materials could...
0 Datasets
View NSUF 17-912: Investigation of gas bubble behavior under ion irradiation -
NSUF 17-917: Pore size distribution in U-Mo fuel irradiated to high burnup
Attaining fundamental understanding of fuel performance requires detailed...
0 Datasets
View NSUF 17-917: Pore size distribution in U-Mo fuel irradiated to high burnup -
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...
0 Datasets
View NSUF 17-922: Effect of grain boundary character and surface treatment on irradiation tolerance of nuclear alloys -
NSUF 17-923: a' precipitation in neutron-irradiated Fe-9/12Cr alloys
High chromium ferritic/martensitic (F-M) steels are one of the strong...
0 Datasets
View NSUF 17-923: a' precipitation in neutron-irradiated Fe-9/12Cr alloys -
NSUF 17-929: Radiation Stability Study on Nuclear Waste/Spent Fuel Materials
Since radiation damage and decay heat are significant issues for nuclear...
0 Datasets
View NSUF 17-929: Radiation Stability Study on Nuclear Waste/Spent Fuel Materials -
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...
0 Datasets
View NSUF 17-930: Characterization of ion irradiated 15-15Ti steel by APT -
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...
0 Datasets
View 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 -
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...
0 Datasets
View NSUF 17-935: In situ ion irradiation of second phase particles in zirconium fuel cladding -
NSUF 17-938: Nanoindentation testing of neutron irradiated 304 stainless steels hex-blocks
Investigation on neutron-induced changes in properties of reactor structural...
0 Datasets
View NSUF 17-938: Nanoindentation testing of neutron irradiated 304 stainless steels hex-blocks