Steady State Thermal Hydraulic Model for PWR Light Water Reactors

Thermal-hydraulic analysis of nuclear reactors is very important to predict the temperature distribution in the reactor core elements, such as fuel cladding and fuel rods. Power reactors cores are composed of cylindrical fuel rods that contain fuel pellets, gap and cladding. Our goal will be to calculate the fuel and coolant temperature distribution and pressure gradient in an average and hot channel and others safety limits. The present paper introduces a simple one dimensional Fortran program called PWRTH for steady state Thermal Hydraulic (TH) calculations and fuel elements heat conduction for Pressurized Water Reactors. A parametric analysis for obtaining the maximum possible reactor power respecting safety limits is performed.