A5101744631- Nuclear Materials and Properties
- Nuclear reactor physics and engineering
- Fusion materials and technologies
- Advanced ceramic materials synthesis
- High-Temperature Coating Behaviors
- Radioactive element chemistry and processing
- Nuclear and radioactivity studies
- ZnO doping and properties
- Nuclear materials and radiation effects
- Advanced materials and composites
- Graphite, nuclear technology, radiation studies
- Thermodynamic and Structural Properties of Metals and Alloys
- Diamond and Carbon-based Materials Research
- Luminescence Properties of Advanced Materials
- Aluminum Alloys Composites Properties
- High Temperature Alloys and Creep
- Nuclear Engineering Thermal-Hydraulics
- Perovskite Materials and Applications
- Intermetallics and Advanced Alloy Properties
- Electronic and Structural Properties of Oxides
- Thin-Film Transistor Technologies
- High Altitude and Hypoxia
- Orthopaedic implants and arthroplasty
- Metallurgy and Material Forming
- Coal Combustion and Slurry Processing
Idaho National Laboratory
2021-2024
Westinghouse Electric (United States)
2011-2023
Sichuan University
2021
Chengdu University
2021
Westinghouse Electric (Japan)
2020
Auburn University
2017
Chengdu Sport University
2012
University of Wisconsin–Madison
2011
University of Florida
2009
The University of Sydney
2002-2007
Silicon carbide (SiC)-based nuclear fission fuel rod cladding has been considered as one of the possible designs for accident tolerant fuels. It is in form a SiC fibre reinforced matrix composite tube (SiCf-SiCm) with monolithic outer and/or inner coating layers. This study focuses on deformation and fracture processes this material using situ X-ray micro-computed tomography (XCT) at room temperature (RT) 1200 °C an inert gas environment C-ring compression loading configuration. Prior to...
As part of Accident Tolerant Fuel initiative for light water reactors, uranium silicide and silicide-nitride are being considered as fuels that can be combined with a more robust cladding such ferritic (FeCrAl) alloy. Although these materials have been studied in the past, uncertainties remain concerning fundamental behaviour systems. In this study, four compositions between U3Si5 USi2 were fabricated by arc-melting. Additionally, an effort to understand UN–U3Si2 fuel system, unidentified...
Abstract A new, in situ hermeticity testing apparatus has been developed to allow helium leak evaluation of ceramic tubes, including nuclear‐grade SiC/SiC fuel cladding matrix composites (CMC), during four‐point bending with simultaneous monitoring local deformation and damage, using stereoscopic digital image correlation (DIC) acoustic emissions. The capabilities the experimental are demonstrated alumina, borosilicate glass, 4130 steel tubes representative dimensions then applied study...
In this work, we demonstrate and describe an effective method of protecting zirconium fuel cladding against oxygen hydrogen uptake at both accident working temperatures in water-cooled nuclear reactor environments. Zr alloy samples were coated with nanocrystalline diamond (NCD) layers different thicknesses, grown a microwave plasma chemical vapor deposition apparatus. addition to showing that such NCD layer prevents the from directly interacting water, show carbon released film enters...
Silicon carbide (SiC) ceramic matrix composite (CMC) cladding is currently being pursued as one of the leading candidates for accident-tolerant fuel (ATF) light water reactor applications. The morphology fabrication defects, including size and shape voids, key challenges that impacts performance guarantees safety. Therefore, quantification defects’ size, location, distribution, leak paths critical to determining SiC CMC in-core performance. This research aims provide quantitative insight...
We report on the fabrication and electrical characterization of bottom gate thin-film transistors (TFTs) based a sol-gel derived ZnO channel layer. The effect annealing active layers characteristics TFTs was systematically investigated. Photoluminescence (PL) spectra indicate that crystal quality improves with increasing temperature. Both device turn-on voltage (Von) threshold (VT) shift to positive As temperature is increased, both subthreshold slope interfacial defect density (Dit)...
The corrosion resistance of cerium silicide, a surrogate uranium is investigated to gain insight into the reaction silicide with water. As-received and proton-irradiated Ce3Si2, CeSi2, CeSi1.x monolithic pellets are subjected tests in water at 300°C 9 MPa for up 48 h. Results show that an oxide layer composed Ce4.67 (SiO4)3O forms on surface all samples, it grows thicker extended exposure times. Irradiated samples corrode greater extent than their unirradiated counterparts, which mainly...