Heat -- Convection -- Simulation methods
This thesis presents a model for predicting the forced convective critical heat flux (CHF) for water over a wide range of thermal-hydraulic conditions which might be encountered during normal and accident operations of a light water nuclear reactor. The model is primarily composed from existing steady-state CHF correlations for tubes or tube and rod bundle geometries, and encompasses the following parametric ranges: 03. Γëñ P (MPa) Γëñ 16.0; 6.0 Γëñ D (mm) Γëñ 30.0; 100.0 Γëñ G (kg/m2s) Γëñ 8000.0; -0.30 Γëñ X Γëñ 1.0. The correlations used as the foundation of this model are the 1) Westinghouse-3 correlation, 2) Biasi correlation, and the 3) Modified Barnett correlation. The model presented is compared with available data, and the resultant model is illustrated as a 3-D surface in mass flux, quality, and CHF space to represent general CHF behavior.
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Master of Science (M.S.)
College of Engineering
Length of Campus-only Access
Masters Thesis (Open Access)
Dahlquist, Joseph E., "Forced Convective Critical Heat Flux Modeling for Tubes and Rod Bundles" (1983). Retrospective Theses and Dissertations. 675.