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，于2002年对沸腾气泡在受热壁和湍流场中的详细行为、不同流型下泡状流中含气率的分布机理以及环状两相流中气核内的气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预测进行了较为详细的建模和分析，包括核反应堆的实际运行条件；在反应堆的实际运行条件(高温高压)下，对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，当压力较低(5 Mpa)时，空泡分数峰值出现在壁面附近，而在较高压力(15 Mpa)时，空泡分数峰值出现在子流道中心。为了解释这些现象，建立了考虑压力影响的气泡弥散系数的力学模型，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。预测结果与单通道和5×5棒束的空泡率分布和CHF的实验数据吻合较好。考虑到反应堆实际情况下热流的不均匀，预测了环状两相流中的干涸热流。在预测中，对雾化速率、沉积速率、质量、环状流开始时的液滴分数和干燥时的液膜厚度之间的关系进行了详细的模拟和评估。利用这些关联式，可以准确地预测均匀热流密度分布和非均匀热流密度分布条件下压力、流道当量直径和长度以及液体流量的较大范围内的干涸热流密度。

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.