Search Constraints

Start Over
Filtering by: Subject Pressurized water reactors -- Critical heat flux -- Measurement Remove constraint Subject: Pressurized water reactors -- Critical heat flux -- Measurement Degree Level Doctoral Remove constraint Degree Level: Doctoral

Search Results

  • 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

  • Out-of-pile blowdown critical heat flux experiment

    Creator:
    Moussaoui, Musa A.
    Abstract:
    In the most challenging nuclear power plant accidents, transient critical heat flux (CHF) is a primary phenomenon that drives peak cladding temperature and ultimately fuel failure. It is not yet determined whether the use of steady-state CHF methods can accurately predict transient CHF under the conditions of a blowdown due...
    Resource Type:
    Dissertation
    Full Text:
    . Title: Out-of-Pile Blowdown Critical Heat Flux Experiment Abstract approved

  • 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:
    The Development of a Mechanistic Critical Heat Flux