Keywords

Heat -- Convection -- Simulation methods

Abstract

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.

Notes

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Graduation Date

Fall 1983

Advisor

Gunnerson, Fred

Degree

Master of Science (M.S.)

College

College of Engineering

Degree Program

Engineering

Format

PDF

Pages

73 p.

Language

English

Rights

Public Domain

Length of Campus-only Access

None

Access Status

Masters Thesis (Open Access)

Identifier

DP0014091

Included in

Engineering Commons

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