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Filtering by: Subject Pressurized water reactors -- Critical heat flux -- Measurement Remove constraint Subject: Pressurized water reactors -- Critical heat flux -- Measurement Advisor Klein, Andrew C. Remove constraint Advisor: Klein, Andrew C.

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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