Fluid-to-fluid scaling for Critical heat flux researches at CAREM-25 Reactor

Critical heat flux (CHF) is the main limiting factor for the power obtained from a nuclear power plant. During CHF, a continuous steam film is created in a small region on the fuel element rod surface, producing a sharp loss on the heat transfer coefficients to the coolant and a sudden rise of the rod temperature in that particular zone. To avoid the occurrence of this phenomenon, safety factors are applied to operate the nuclear reactor at a power well below the corresponding to CHF conditions. Characterization studies of a certain fuel element concerning CHF allow not only to establish operational power and flow conditions for this phenomenon, but to propose design modifications to delay CHF's occurrence, enhancing the fuel element performance and allowing to increase the reactor's power generation. Analytical or available numeric techniques fail to obtain representative solutions for the phenomenon, which makes the development of experiments mandatory. While experimental tests must be performed in a full scale test section geometrically identical to the actual fuel element, it is possible to reduce the power and pressure requirements using fluid-to-fluid scaling. This technique has been successfully applied to the study of fuel elements of common power reactors, but there is no background experiences validating the use of this model in CAREM-25 reactor conditions. Experimental data obtained using water and Freon R-12 with similar geometrical and thermohydraulic conditions is presented to discuss these topics. (author)