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  • Application of sandwich structure analysis in predicting critical flow velocity for a laminated flat plate

    Creator:
    Jensen, Philip
    Abstract:
    The Oregon State University (OSU), Hydro Mechanical Fuel test Facility (HMFTF) is designed to hydro-mechanically test prototypical plate type fuel. OSU's fuel test program is a part of the Global Threat Reduction Initiative (GTRI), formerly known as the Reduced Enrichment for Research and Test Reactor program. One of the GTRI's...
    Resource Type:
    Masters Thesis
    Full Text:
    enriched uranium (HEU) to a low enriched uranium (LEU) fuel in an effort to reduce nuclear proliferation

  • Vibration Analysis of Advanced Test Reactor Miniplate Hydraulic Test in the OSU HMFTF

    Creator:
    Latimer, Griffen D.
    Abstract:
    Current research on the topic of advanced reactor fuel types include the use of ultrahigh density Uranium-Molybdenum fuels, towards their use in high-performance research reactors. These reactors operate with high power densities, and the increased cooling requirements therefore place high relevance on the fluid-structure interaction with these fuel elements; therefore...
    Resource Type:
    Masters Thesis
    Full Text:
    conveying fluid. Numerous studies have investigated the dynamic response of shells under these conditions

  • Modification of Lower Plenum Structure for Mixing Promotion during Normal Operation of the High Temperature Gas Cooled Reactor

    Creator:
    Gradecka, Malwina
    Abstract:
    Since High Temperature Gas-cooled Reactors are being considered as most promising design of upcoming IV Gen reactors, thus key research areas were identified to address safety aspects of this design. Number of simulations and experiments need to be conducted in this field. In this study thermal-hydraulics aspects of coolant flow...
    Resource Type:
    Dissertation
    Full Text:
    production of synthetic fuels. To satisfy future demand for transportation fuel, nuclear process heat may

  • A study of opposing mixed convection in the GRTS and in downward pipe flows using the FLUENT CFD code

    Creator:
    Jackson, R. Brian
    Abstract:
    Downward flow opposing mixed convection conditions have been studied using the FLUENT CFD code. This work was comprised of two primary objectives. The first was to study core channel flow reversal susceptibility in the Gas Reactor Test Section (GRTS) at low flow conditions. The second objective was the development of...
    Resource Type:
    Masters Thesis
    Full Text:
    Jackson for the degree of Master of Science in Nuclear Engineering presented on March 17, 2008

  • Flow Dynamics and Condensation of Film Flows in Small Modular Reactors

    Creator:
    Lee, Dongyoung
    Abstract:
    There is renewed interest in the reliability and safety of nuclear power plants following the Fukushima Daiichi nuclear accident followed by 8.9 magnitude earthquake and Tsunami with the height of 15 m on March 11, 2011. Small Modular Reactors (SMRs) have been developed to improve safety systems by utilizing passive...
    Resource Type:
    Dissertation
    Full Text:
    integral PWR, which consists of nuclear fuel, control rod drives, SG, pressurizer and reactor coolant

  • Pressure losses across a 180 degree turn in the AP1000 cooling annulus

    Creator:
    Nelson, Roy K.
    Abstract:
    A one-sixth scaled facility was constructed at Oregon State University to determine the pressure losses around Westinghouse's AP1000 Passive Containment Cooling System. An explanation of the AP1000 Passive Containment Cooling System is given. This study examined the pressure loss and loss coefficient around the 180 degree turn of the Passive...
    Resource Type:
    Masters Thesis
    Full Text:
    Master of Science in Nuclear Engineering presented on October 2, 2008. Title: Pressure Losses Across a

  • Comparative analysis of the Žukauskas method and data from the OSU MASLWR test facility steam generator

    Creator:
    Wuth, Bradyn J.
    Abstract:
    Some of the next generation of small modular reactors have been designed with helically coiled steam generators integrated into the reactor vessel to enhance natural circulation. According to the developers of TASS/SMR, a one-dimensional thermal hydraulic code written to model the Korean Atomic Energy Research Institute's SMART reactor, the Žukauskas...
    Resource Type:
    Masters Thesis
    Full Text:
    to a nuclear core's fuel assemblies. The lower core plate and shroud around the core keep the

  • An Approach to Modeling Fuel Plate Deformations in Fluid-Structure Interactions

    Creator:
    Howard, Trevor
    Abstract:
    As computational power increases, so does the desire to use fluid-structure interaction (FSI) software to design complex systems and components such as fuel plates. Presently, an effort is underway to support the design and qualification of a new nuclear fuel material which is intended for applications in select, plate-type, research...
    Resource Type:
    Masters Thesis
    Full Text:
    An Approach to Modeling Fuel Plate Deformations in Fluid

  • A thermal hydraulics analysis of a molybdenum production element for implementation in TRIGA reactors

    Creator:
    Byfield, Patrick
    Abstract:
    The metastable isotope of technetium-99 (Tc-99m) is an important diagnostic tool used in the field of nuclear medicine due to the isotope's 6.0 hour half-life, 140.5 keV γ-decay mechanism, and multiple oxidation states [1,2]. Approximately 70% of the world’s nuclear medicine procedures involve the use of Tc-99m [3]. The conventional...
    Resource Type:
    Masters Thesis
    Full Text:
    , wherein the fluid is heated by fuel rods near the bottom of the reactor pool and cooling is provided by

  • Thermal hydraulic analysis of the Oregon State TRIGA Reactor using RELAP5-3D

    Creator:
    Marcum, Wade R.
    Abstract:
    Oregon State University has recently conducted a complete core conversion analysis as part of the Reduced Enrichment for Research and Test Reactors Pprogram. The goals of the thermal hydraulic analyses were to calculate natural circulation flow rates, coolant temperatures and fuel temperatures as a function of core power for both...
    Resource Type:
    Masters Thesis
    Full Text:
    Wade R. Marcum for the degree of Master of Science in Nuclear Engineering presented on January 23