A Study on turbulence structure and critical heat flux in boiling two-phase flow

为了建立基于沸腾两相流湍流机理的CHF通用预测方法，2002年对沸腾汽泡在加热壁面上的详细行为和湍流场、不同流型下泡状流含气率分布的机理以及环状两相流内气核的膜行为和湍流结构进行了研究。2003年，对沸腾两相流的湍流结构和CHF预测进行了更详细的模拟和分析，包括核反应堆的实际运行工况，在核反应堆的实际运行工况（高温高压）下，对核反应堆燃料组件进行了湍流结构分析和CHF预测。实验结果表明，在较低压力（5MPa）下，空泡率峰值出现在近壁处，而在较高压力（15MPa）下，空泡率峰值出现在子通道中心。为了解释这一现象，建立了考虑压力影响的气泡弥散系数的机理模型，并推导了修正的气泡弥散系数关联式。利用该关联式，预测了棒束的空泡率分布和CHF。计算结果与单通道和5 × 5棒束的空泡率分布和CHF实验数据吻合较好。考虑到核反应堆实际工况下热流的不均匀性，对环形两相流中的干涸热流进行了预测。在预测中，卷吸速率，沉积速率，质量和液滴分数在环形流和干膜厚度的开始的相关性进行了建模和评估的细节。利用这些关联式，在较宽的压力、流道当量直径、流道长度和液体流量范围内，对均匀和非均匀热流密度分布下的干涸热流密度进行了准确预测。

In order to establish general predictive method of CHF based on turbulence mechanisms of boiling two-phase flow, investigation was carried out on detailed behavior of boiling bubbles at heated wall and turbulence field, mechanisms of void fraction distribution in bubbly flow in various flow geometries and film behavior and turbulence structure in gas core in annular two-phase flow in 2002. In 2003, more detailed modeling and analyses were carried out on turbulence structure in boiling two-phase flow and prediction of CHF, including actual operating conditions of nuclear reactor.In actual operating conditions of nuclear reactors (high temperature and high pressure), analyses of turbulence structure and prediction of CHF were carried out for fuel assembly of nuclear reactor. Experimental results showed that peak void fraction was observed near wall for lower pressure (5Mpa) while peak void fraction appears in the center of subchannel for higher pressure (15Mpa). In order to explain these phenomena, mechanistic model of bubble dispersion coefficient was developed considering the effect of pressure and modified correlation of bubble dispersion coefficient was derived. Using this correlation, void fraction distribution and CHF for rod bundle were predicted.. The prediction agreed well with the experimental data of void fraction distribution and CHF for single channel and 5x5rod bundle.In consideration of nonuniformity of heat flux in actual conditions of nuclear reactor, dryout heat flux was predicted in annular two-phase flow. In the prediction, the correlations of entrainment rate, deposition rate, quality and droplet fraction at the onset of annular flow and liquid film thickness at dryout were modeled and evaluated in details. Using these correlations, dryout heat fluxes ware accurately predicted for wide ranges of pressure, equivalent diameter and length of flow channel and liquid flow rate under the conditions of uniform and nonuniform heat flux distributions.