A5073244814- Nuclear reactor physics and engineering
- Nuclear Materials and Properties
- Nuclear Engineering Thermal-Hydraulics
- Heat transfer and supercritical fluids
- Nuclear and radioactivity studies
- Vibration and Dynamic Analysis
- Hydraulic and Pneumatic Systems
- Mechanical stress and fatigue analysis
- Mechanical Failure Analysis and Simulation
- Graphite, nuclear technology, radiation studies
- Geotechnical Engineering and Underground Structures
- Advanced materials and composites
- Wind and Air Flow Studies
- Radioactive element chemistry and processing
- Engineering and Environmental Studies
- High-Temperature Coating Behaviors
- Metallurgy and Material Forming
- Particle Dynamics in Fluid Flows
- Biodiesel Production and Applications
- Cyclone Separators and Fluid Dynamics
- Fluid Dynamics and Vibration Analysis
- Fluid Dynamics and Mixing
- Superconducting Materials and Applications
- Heat Transfer and Boiling Studies
- Engineering Applied Research
Westinghouse Electric (United States)
2010-2024
Westinghouse Electric (Japan)
2007-2022
Bluffton University
2011
Westinghouse Electric (Sweden)
2006
This paper provides single and two phase rod bundle data to support verification of heat transfer models being used in steaming rate crud model predictions for bundles. The effort summarize this work was supported by the EPRI Robust Fuel Program is defined more detail report 1003383. Subcooled boiling tests were performed Combustion Engineering (CE) early 1980s provide insight on heavy deposits fuel failures observed peripheral rods bundles Maine Yankee cycle 4. Two 5×5 at Columbia...
Current Pressurized Water Reactors (PWR) fuel assembly thermal-hydraulic (T/H) analyses are performed on a subchannel basis that neglects detailed heat transfer and flow distributions surrounding rods. Subchannel codes such as VIPREW require input of thermal mixing hydraulic loss coefficients obtained from costly experiments. Fuel margin or performance is quantified in terms Departure Nuclear Boiling Ratio (DNBR) for PWR applications Critical Power (CPR) Reactors. DNBR CPR predictions...
Critical heat flux (CHF) is a primary parameter for nuclear fuel design and plant operation safety. CHF values are normally obtained from bundle integral departure nucleate boiling (DNB) or dryout experiments. These experiments expensive, detailed measurements (bubble dynamics, void fraction distribution, etc.) difficult to obtain, particularly under typical pressurized water reactor (PWR) conditions of high pressure temperature. Therefore, it highly desirable that computational tools such...
In order to understand crud formation on the fuel rod cladding surfaces of pressurized water reactors (PWRs), a Thermal-Hydraulic test facility referred as Westinghouse Advanced Loop Tester (WALT) was built at Science and Technology Department Laboratories in October 2005. Since then, number updates have been made are described here. These include heater improvements, system pressure stabilization, more effective protection systems. After these were made, WALT has operated with higher...
Researchers have performed many studies to try understand crud formation on the fuel pin clad surfaces, which has been observed in pressurized water reactors (PWR) as a result of sub-cooled nucleate boiling and precipitation reactions. Crud deposits, may cause an unexpected change core power distribution known induced shift (CIPS) or axial offset anomaly (AOA) if boron species accumulate deposits. If deposit is thick enough, potential exists for rod surface dryout. The Westinghouse Advanced...
A long-term flow-induced vibration and wear test was performed for a full-scale 17×17 PWR fuel mockup, the results were compared with numerical simulations. The on assembly or rods may cause Grid-to-Rod Fretting (GTRF) result in leakage of PWRs. GTRF involves non-linear rod due to excitation force induced by coolant flow around rod. So, simulation is VITRAN (Vibration Transient Analysis Non-linear) Computational Fluid Dynamics (CFD). code developed Westinghouse simulate GTRF. In this paper,...
This paper describes a laboratory test program to measure the thermal conductivity of corrosion product deposits on surface Pressurized Water Reactor (PWR) fuel rod under variety hydraulic conditions. information is necessary allow more accurate predictions temperatures in presence deposits, commonly known as crud. In this paper, four regime theory and methodology are proposed utilized for crud measurements calculations. The relevant were performed at Westinghouse Advanced Loop Tester (WALT)...
Heavy deposits (typically called crud) on fuel have had a significant impact plant operation, causing unexpected Crud Induced Power Shifts (CIPS) and in some cases crud induced cladding failure (CIF). The Westinghouse Advanced Loop Tester (WALT) at the George Science Technology Center (STC), which was completed October 2005, is designed to model top grid span of high power rod assembly under Pressurized Water Reactor (PWR) operating conditions. Both clean crudded surfaces can be simulated...
This paper presents a computational fluid dynamics (CFD) modeling methodology that has been developed to provide predictions of very local heat transfer variation in fuel rod assemblies. Results from the CFD analysis are used HIDUTYDRV and other advanced codes have internally by Westinghouse predict crud deposition dryout. is making EPRI Level IV corrosion guideline assessments, which were response INPO 0 2010 initiatives. production use for risk assessment CE-design 14×14 16×16 reloads. The...
A numerical investigation was performed to study the variation in axial and lateral velocity profiles occurring downstream of inlet nozzle a typical Westinghouse 17×17 PWR fuel assembly. Computational Fluid Dynamic (CFD) model developed with commercial CFD software. The comprised lower region assembly, including: Debris Filter Bottom Nozzle (DFBN), P-grid, Inconel grid, one half grid span, as well core plate hole. purpose obtain insight into flow redistribution resulting from interaction jet...