In this thesis are determined the temperatures in the thermocouple located at the exit of the coolant channels of the reactor core of a typical PWR, for identifying the occurrence of the control rod drop accident during the operation in power of the reactor. For that purpose, the multigroup diffusion model was used in cylindrical geometry, for calculation of neutron fluxes and consequently, power densities. Solution of the discretized multigroup neutron diffusion equation, was obtained using the SOR iterative method (Successive Over-Relaxation), in addition to the source extrapolation method of Chebyshev in the outer iterations. It also was developed in this thesis a cylindrization technique for the assemblies of the core, to consider the properties of the different materials of the assemblies. For calculation of the temperatures in the rods, use was made of a one dimensional heat transfer model in the pellet and cladding, assuming an average value for the thermal properties. An iterative step in a section axial of the rod was used, to update the thermal properties. The generated results showed the effectiveness of the model in identifying a control rod dropped into a position in the reactor core. The temperature distribution obtained with the developed system will be able to be used in an intelligent system for identification, in line and in real-time, of the control rod drop event.
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