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Filtering by: Subject Pressurized water reactors -- Critical heat flux -- Measurement Remove constraint Subject: Pressurized water reactors -- Critical heat flux -- Measurement Academic Affiliation Nuclear Engineering Remove constraint Academic Affiliation: Nuclear Engineering

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  • Critical heat flux estimation for annular channel geometry

    Creator:
    Pagh, Richard T.
    Abstract:
    Critical Heat Flux (CHF) is an important safety parameter for the design of nuclear reactors. The most commonly used predictive tool for determination of CHF is a look-up table developed using tube data with an average hydraulic test diameter of 8 mm. There exist in the world today nuclear reactors...
    Resource Type:
    Dissertation
    Full Text:
    Critical heat flux estimation for annular channel geometry

  • Assessment of level swell measurements in the OSU APEX Test Facility

    Creator:
    Ellis, Christopher P.
    Abstract:
    The purposes of this paper are to present the results of an experimental effort to measure the level swell in an air/water system and to generate a methodology for determining the volume-averaged void fraction within the Oregon State University (OSU) Advanced Plant Experiment (APEX) Test Facility. The results were then...
    Resource Type:
    Masters Thesis
    Full Text:
    rat ing conditions within a conven ti onal Pressuri zed Water Reactor (PWR). the core is completely