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 1) Westinghouse-3 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.