Convection Heat 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:
0.3 ≤ P (MPa) ≤ 16.0
6.0 ≤ D (mm) ≤ 30.0
100.0 ≤ G (kg/m2s) ≤ 8000.0
-0.30 ≤ X ≤ 1.0
The correlations used as the foundation of this model are the
2) Biasi correlation, and the
3) Modified Barnett correlation
The mode 1 presented is comp a red 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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Gunnerson, Fred S.
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.
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