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Filtering by: Subject Nuclear fuels -- Fluid dynamics Remove constraint Subject: Nuclear fuels -- Fluid dynamics Decade 2000-2009 Remove constraint Decade: 2000-2009

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  • 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

  • Liquid entrainment at an upward oriented vertical branch line from a horizontal pipe

    Creator:
    Welter, Kent B.
    Abstract:
    Under simulated accident conditions, tees in the primary coolant loop of a Pressurized Water Reactor (PWR) can deviate from their original design purpose and become separators that effectively remove core heat sink capacity. This method of primary coolant removal is a phenomelogical subset of phase separation known as liquid entrainment,...
    Resource Type:
    Dissertation
    Full Text:
    degree of Doctor of Philosophy in Nuclear Engineering presented on September 25, 2002. Title: Liquid

  • 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

  • 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