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Filtering by: Alternative title ssim FG 7 Remove constraint Alternative title ssim: FG 7 Advisor Wu, Qiao Remove constraint Advisor: Wu, Qiao Committee Member Woods, Brian Remove constraint Committee Member: Woods, Brian

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  • Quantified PIRT and uncertainty quantification for computer code validation

    Creator:
    Luo, Hu
    Abstract:
    This study is intended to investigate and propose a systematic method for uncertainty quantification for the computer code validation application. Uncertainty quantification has gained more and more attentions in recent years. U.S. Nuclear Regulatory Commission (NRC) requires the use of realistic best estimate (BE) computer code to follow the rigorous...
    Resource Type:
    Dissertation
    Full Text:
    Mixture Pressure and Body Force Π Values ........................... 75 Figure 4-7 Primary Loop Mixture

  • The Development of a Mechanistic Critical Heat Flux Correlation

    Creator:
    Luitjens, Jeffrey M.
    Abstract:
    A mechanistic based critical heat flux (CHF) correlation has been developed derived from a mass, momentum, and energy balance at a nucleation site for water. The trigger mechanism for the CHF condition is postulated to be related to the maximum boiling frequency on a heated surface. It is observed that...
    Resource Type:
    Dissertation
    Full Text:
    . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2 The Boiling Process 7 2.1 Nucleation in a Pure Liquid (Bulk Boiling

  • Dynamic System Scaling Application to Accelerated Nuclear Fuel Testing

    Creator:
    Yoshiura, Ramon K.
    Abstract:
    The development of nuclear fuel and materials requires is a continuous effort to investigate the acute and prolonged effects of irradiation, thermal-material stress, chemical change, or other conceivable damage mechanics acquired during normal operation or accident scenarios throughout its lifetime. As in-core fuel property measurement techniques advance to support in...
    Resource Type:
    Dissertation
    Full Text:
    the degree of Doctor of Philosophy in Nuclear Engineering presented on June 7, 2021. Title: Dynamic

  • Stability improvement of the one-dimensional two-fluid model for horizontal two-phase flow with model unification

    Creator:
    Abel, Kent C.
    Abstract:
    The next generation of nuclear safety analysis computer codes will require detailed modeling of two-phase fluid flow. The most complete and fundamental model used for these calculations is known as the two-fluid model. It is the most accurate of the two-phase models since it considers each phase independently and links...
    Resource Type:
    Dissertation
    Full Text:
    2.3: Flow map for horizontal flow (Mandhane et al., 1974) .......................................... 7