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A Computational Fluid Dynamic Study on Flow Induced Vibration in Helically Wire Wrapped Rods Under Axial Turbulent Flow

Creator:: Winkle, Tristan
Abstract:: Vibration of nuclear fuel rods is an area of significance for the long-term safe operation of any nuclear power plant. One mechanism of vibration, flow induced vibration (FIV), must be quantified when designing a new core, such as a sodium fast reactor which employs helically-wrapped wire spacers. Experimental work undertaken...
Resource Type:: Masters Thesis
Full Text:: compares the amplitude and frequency of vibration between simulated and experimental wire-wrapped rod

Design, Fabrication, and Evaluation of a Novel High-Performance Resistance Heater Rod for Nuclear Thermal Hydraulic Experimentation

Creator:: Sharaf, Salem A.
Abstract:: The Accident Tolerant Fuels campaign demonstrated a need for nuclear fuel pin simulators to further research fuel and cladding combinations that can withstand off normal conditions in light water reactors. The use of a heater rod as a fuel pin simulator greatly reduces the risk involved during testing as no...
Resource Type:: Masters Thesis
Full Text:: , Fabrication, and Evaluation of a Novel High-Performance Resistance Heater Rod for Nuclear Thermal Hydraulic

Design, fabrication, and characterization of the laser-imaged natural circulation (LINC) facility

Creator:: Goodrich, Sam S.
Abstract:: Currently there is a great amount of interest in the phenomena of natural circulation as a cooling mechanism for normal operation as well as emergency conditions in nuclear reactors and spent fuel pools. In order to better understand this phenomena for the specific geometry of vertical, heated rods in water,...
Resource Type:: Masters Thesis
Full Text:: of 100 mm along the entrance region between the two rods at one symmetrical rod power. In addition

Fuel Dispersal Under LOCA Conditions

Creator:: Latimer, Griffen D.
Abstract:: Most operating power reactors use oxide fuel in the form of uranium oxide (UO2) which is robust against high temperatures but suffers from poor mechanical performance at high levels of burnup. Currently, the maximum average burnup in light water reactors in the US is 62.5 GWd/t, which is partially imposed...
Resource Type:: Dissertation
Full Text:: ), unirradiated rod testing (group B), and irradiated rod testing (groups C, D, E, F, and G). For the

Computational Fluid Dynamics Modeling of the Density Evolution Inside of the Helium-3 Enhanced Negative Reactivity Insertion (HENRI) System

Creator:: Balderrama Prieto, Silvino Alejandro
Abstract:: The testing of nuclear fuel under reactivity-initiated accident (RIA) conditions is paramount for the better understanding of the fuel’s behavior during this transient accident events. The Transient Reactor Test (TREAT) facility is a nuclear reactor that will be capable of recreating the thermal-hydraulic and neutronic boundary conditions representative of RIA...
Resource Type:: Masters Thesis
Full Text:: strong neutron absorber, into an annular control rod using a gas injection system known as the Helium-3

Out-of-pile blowdown critical heat flux experiment

Creator:: Moussaoui, Musa A.
Abstract:: In the most challenging nuclear power plant accidents, transient critical heat flux (CHF) is a primary phenomenon that drives peak cladding temperature and ultimately fuel failure. It is not yet determined whether the use of steady-state CHF methods can accurately predict transient CHF under the conditions of a blowdown due...
Resource Type:: Dissertation
Full Text:: was built to simulate an electrically heated rod under blowdown conditions. The test facility

Natural Convection Heat Transfer and Boundary Layer Transition for Vertical Heated Cylinders

Creator:: Goodrich, Sam S.
Abstract:: In recent years, the global nuclear industry has placed a greater emphasis on passively safe reactor designs. In particular, much attention and design work has been applied to engineered systems for passive cooling of nuclear fuel, whether it be loaded in a core or stored in spent fuel pools. The...
Resource Type:: Dissertation
Full Text:: ..................... 130 Figure 5.6 Custom heater rod cross-sectional diagram example ............................ 134

Applying uncertainty and sensitivity on thermal hydraulic subchannel analysis for the multi-application small light water reactor

Creator:: Brigantic, Adam
Abstract:: Small modular reactors (SMRs) are a recent advancement in commercial nuclear reactor design with growing interest worldwide. New SMR concepts, such as the Multi-Application Small Light Water Reactor (MASLWR), must undergo a licensing processes established by the U.S. Nuclear Regulatory Commission (NRC) prior to commercial operation. Given the lack of...
Resource Type:: Masters Thesis
Full Text:: ......................................................................... 26 Nuclear Fuel Rod

Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA® reactor

Creator:: Schickler, Robert
Abstract:: In 2008, the Oregon State TRIGA® Reactor (OSTR) was converted from highly-enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the...
Resource Type:: Masters Thesis
Full Text:: .......................................... 6 2-2: Differences in critical rod heights (percent withdrawn from core

Blowdown transient critical heat flux

Creator:: Moussaoui, Musa A.
Abstract:: Occurring in the most challenging nuclear power plant accidents, transient critical heat flux (CHF) is the primary phenomenon that drives peak cladding temperature and ultimately fuel failure. It is unclear whether the use of steady-state CHF correlations can accurately predict the gross thermal-hydraulic perturbations of a blowdown in a large...
Resource Type:: Masters Thesis
Full Text:: : Center fuel bundle, instrumented fuel rod locations (page 234 of [71]). ....... 45 Figure 3.5

ScholarsArchive@OSU

A Computational Fluid Dynamic Study on Flow Induced Vibration in Helically Wire Wrapped Rods Under Axial Turbulent Flow

Design, Fabrication, and Evaluation of a Novel High-Performance Resistance Heater Rod for Nuclear Thermal Hydraulic Experimentation

Design, fabrication, and characterization of the laser-imaged natural circulation (LINC) facility

Fuel Dispersal Under LOCA Conditions

Computational Fluid Dynamics Modeling of the Density Evolution Inside of the Helium-3 Enhanced Negative Reactivity Insertion (HENRI) System

Out-of-pile blowdown critical heat flux experiment

Natural Convection Heat Transfer and Boundary Layer Transition for Vertical Heated Cylinders

Applying uncertainty and sensitivity on thermal hydraulic subchannel analysis for the multi-application small light water reactor

Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA® reactor

Blowdown transient critical heat flux

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