A subchannel program was designed to make a steady-state core-wide analysis for PWR core. Equations of continuity, energy, axial and transverse momentum, which are the main elements of the program, were presented in the paper. The spatial forward differentiation method was used to get the thermal hydraulic parameters, such as temperature, pressure, enthalpy of different spatial steps along the axial direction in each channel. In order to validate the program, a physical model of the reactor core, which is based on 900MW PWR nuclear plant, was established and computed in the program. In the program, the core was radially divided into 8 subchannels. Variations of the temperature, pressure and enthalpy in each coolant subchannel were computed. The exterior surfaces’ temperatures of the fuel rods in subchannels were computed as well. The computation shows that outlet pressure of the core is about 15.26 MPa; the average outlet temperature of each channel is about 325.34°. These results are within reasonable range based on comparing with published datum [1].

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