Boiling and two-phase flows

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Improving the performance of thermal equipment used in the power industry can strongly contribute in re- ducing environmental impacts of burning fuels. In addition, nuclear power plants produce electricity without emissions. Thus, the nuclear technology is now at the foreground of the power industry. A better design of thermal systems can help in reducing conventional plant emissions as well as to guarantee the safe operation of nuclear stations. Therefore, this proposal presents a program intended to study several subjects that could help in improving the efficiency of conventional and nuclear power plants. These subjects are: i) Pressure Losses in Boiling Fluids: their prediction is essential for dimensioning thermal equipment and consequently reducing energy waste. They are calculated with empirical correlations developed from data. In boiling experiments, water is commonly replaced by other fluids. The data is converted into water equivalent values by scaling laws that have been verified only under Critical Heat Flux (CHF) flow conditions. A Freon loop will be used for performing experiments under high-pressure water equivalent flow conditions different from CHF. ii) Flow Distribution in a Full Size Rod Bundle Assembly: pressure losses and heat transfer in fuel channels depend on two-phase flow patterns. A transparent rod-bundle simulation assembly will be manufactured for performing two-phase distribution experiments with air-water mixtures. The flow will be characterized by measuring among other parameters, the local void fractions. iii) Subchannel Analysis: fuel channel thermalhydraulic analyses are carried out with subchannel codes. For calculating lateral mixing effects, these codes require empirical constants to be selected by the user. These parameters, however, depend on local conditions a priori not known. A genetic algorithm coupled to an objective function will be used in a horizontal subchannel model for optimizing lateral mixing without user intervention. iv) 3-D Modeling of the Candu Calandria Vessel: it is intended to introduce more modern simulation techniques than those commonly employed in the nuclear industry. Simulated local flow and temperatures will be used in nuclear safety analyses.