(untitled) (Journal of Nuclear Technology in Applied Science, 2018)
Thermal-hydraulic analysis of nuclear reactors is very important topredict the temperature distribution in the reactor core elements, such asfuel cladding and fuel rods. Power reactors cores are composed of cylin-drical fuel rods that contain fuel pellets, gap and cladding. Our goal willbe 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 pro-gram called PWRTH for steady state Thermal Hydraulic (TH) calcula-tions and fuel elements heat conduction for Pressurized Water Reactors.A parametric analysis for obtaining the maximum possible reactor powerrespecting safety limits is performed.