The IPR-R1 TRIGA nuclear research reactor at the Nuclear Technology Development Center (CDTN), in Belo Horizonte (Brazil), is a pool type reactor. The maximum core power is 250 kW cooled by natural circulation of light water and an open surface. Fuel to coolant heat transfer patterns must be evaluated as function of the reactor power in order to assess the thermal hydraulic performance of the core. The heat generated by nuclear fission in the reactor core is transferred from fuel elements to the cooling system through the fuel/cladding (gap) and the cladding to coolant interfaces. The objective of the thermal and hydrodynamic projects of the reactors is to remove the heat safely, without producing excessive temperature in the fuel elements. The regions of the reactor core where boiling occurs at many different power levels can be determined from the heat transfer coefficient data. As the reactor core power increases the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. Experimental results indicated that subcooled pool boiling occurs at the cladding surface in the reactor core central channels at power levels in excess of 60 kW. However, due to the high heat transfer coefficient in subcooled boiling the cladding temperature is quite uniform along most of the active fuel rod region and do not increase very much with the reactor power. An operational computer program and a data acquisition and signal processing system were developed as part of this research project to allow on line monitoring of the operational parameters.