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Filtering by: Creator French, Rod A. Remove constraint Creator: French, Rod A. Committee Member Woods, Brian Remove constraint Committee Member: Woods, Brian Committee Member Wu, Qiao Remove constraint Committee Member: Wu, Qiao

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  • Development of a Rapid Power Transient Critical Heat Flux Model

    Creator:
    Brown, Emory G.
    Abstract:
    The understanding of critical heat flux is crucial for the safety of nuclear reactors. To this end, many studies have been carried out to understand various aspects of the boiling crisis phenomenon. Power transient critical heat flux trigger mechanisms is a small category that is relatively under-studied and not well...
    Resource Type:
    Dissertation
    Full Text:
    PWR fuel rod during a RIA is a perfect example of this complication. 1.1 Motivation One scenario

  • 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:
    heat loss from vertical, heated cylinders by natural convection as a function of rod diameter and

  • 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:
    surround each heater rod in a hexagonal configuration for a total of 114 coolant channels in the GRTS

  • On the mechanism transition of power transient critical heat flux

    Creator:
    Yan, Yikuan
    Abstract:
    The reactivity-initiated accident (RIA) accident has stirred wide interest for the need of designing advanced and high tolerance fuels for next generation nuclear power plants. For pressurized water reactors (PWRs), the accidentally induced reactivity pulse adds transient energy input to the fuel. As a result, boiling may happen locally. Boiling...
    Resource Type:
    Dissertation
    Full Text:
    power plant by an ejected control rod while at power. When the control rod is ejected in an LWR, a