Program for simulation of transients in nuclear reactor cores under uniform pressure

A methodology for reactor transients analysis was developed and a computer algorithm was produced. The reactor core was represented by a fuel rod placed in an average cooling channel. The finite difference method was adopted for the solution of the point kinetics and thermohydraulic equations. The material properties, a function of the medium temperature, are calculated in the transient condition for temperatures in the immediately previous calculated time interval. Some reactivity transients calculations for thermal reactors were made. The model adopted for the simulation of transients was satisfactory when compared to results in the literature. (F.E.)