Boiling and two-phase flows

改善电力工业中使用的热力设备的性能可以大大有助于减少燃烧燃料对环境的影响。此外，核电站发电没有排放。因此，核技术现在处于电力工业的前景。更好的热力系统设计有助于减少传统电厂的排放，并保证核电站的安全运行。因此，本提案提出了一个旨在研究几个主题的计划，这些主题可以帮助提高常规和核电站的效率。这些科目是：i）沸腾流体中的压力损失：它们的预测对于热力设备的尺寸设计和减少能源浪费至关重要。它们是用从数据中得出的经验相关性计算的。在沸腾实验中，水通常被其他流体代替。通过仅在临界热通量（CHF）流动条件下验证的标度律，将数据转换为水当量值。氟利昂回路将用于在不同于CHF的高压水等效流量条件下进行实验。ii）全尺寸棒束组件中的流量分布：燃料通道中的压力损失和热传递取决于两相流模式。将制造一个透明的棒束模拟组件，用于空气-水混合物的两相分布实验。通过测量局部空隙率等参数来表征流动。iii）子通道分析：燃料通道热工水力分析采用子通道代码进行。为了计算横向混合效应，这些代码需要由用户选择经验常数。然而，这些参数取决于局部条件先验未知。将在水平子通道模型中使用与目标函数相耦合的遗传算法，以在没有用户干预的情况下优化横向混合。iv）Candu Calandria容器的三维建模：其目的是引入比核工业中常用的模拟技术更现代的模拟技术。模拟的局部流动和温度将用于核安全分析。

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.