基于加热壁面能量平衡的窄矩形通道内CHF机理模型

窄矩形通道因具有结构紧凑、换热面积大等优点而被广泛应用于各个领域。通过完善窄矩形通道中临界热流密度（CHF）的预测方法,建立CHF机理模型,可以提高反应堆的安全性和经济性。本文对窄矩形通道内竖直向上流动CHF进行了可视化实验研究,在此基础上开发了一种基于加热壁面能量平衡的CHF机理模型,并提供一组本构关系用于封闭所开发的新模型,同时使用实验数据对新模型进行对比评价,对比结果发现,新模型在窄矩形通道中模拟结果良好,偏差基本都在±20%之间。     

基于矩形窄缝通道实验数据的DNB机理模型评价

偏离泡核沸腾（DNB）对于压水堆安全具有重要意义。已有机理模型能否适用于矩形窄缝通道缺乏足够的实验验证。本文基于矩形窄缝通道实验数据,对微液层蒸干模型和汽泡壅塞模型两类DNB机理模型进行了计算评价。结果显示：汽泡壅塞模型适用范围较微液层蒸干模型宽;部分热工参数对模型计算性能有系统性影响。随空泡份额的增大,各模型的计算性能均变差,可能是通道几何差异所致。     

竖直加热圆管内过冷沸腾及CHF数值模拟

基于欧拉两流体模型和非平衡过冷沸腾模型,完成过冷沸腾数值模型的构建,并通过与Bartolomei单管过冷沸腾实验进行对比,验证模型的正确性。利用该模型计算得到圆管的沸腾曲线,将进入"临界区"后的第一个点作为偏离泡核沸腾(DNB)判定的标准,对高压、高流量下圆管内的DNB型临界热流密度(CHF)进行数值模拟,CHF数据取自最新(2006年)的查询表;计算中考虑质量流量、平衡含汽率和压力对CHF的影响,最终预测值与实际值符合良好,误差在15%以内。预测CHF出现的位置也与实际相符,表明本文提出的方法能够很好地模拟高压、高流量下圆管内的DNB型CHF。     

基于逐步回归无量纲棒束CHF关系式开发与DNBR限值确定

针对目前国内外先进压水堆棒束临界热流密度（CHF）经验关系式普遍存在数学形式复杂、自变量系数众多且缺乏物理意义的共性问题,以美国电力研究院（EPRI）棒束CHF数据库中遴选的485个5×5压水堆棒束CHF数据点为基础,基于逐步回归分析开发了一套新型无量纲棒束CHF关系式。考虑了导向管冷壁效应与轴向非均匀加热效应后,实测CHF与预测CHF之比M/P的平均值为0.998,均方根偏差为0.0546,标准差为0.0546,基于分组法确定了关系式的95/95偏离泡核沸腾比（DNBR）限值为1.16。     

圆管内偏离泡核沸腾机理模型的比较分析

分析比较了基于壁面汽泡壅塞模型和汽泡下液膜烧干两种机理的圆管内偏离泡核沸腾机理的5个模型，并与偏离泡核沸腾（DNB）实验数据进行了比较分析，得出了各种模型的优缺点。     

扁矩形通道CHF试验研究及其CHF关系式的分析评估

间隙在 1～ 3mm的具有大宽度的扁矩形通道 ,其临界热流密度与通常圆管及方管内的相比 ,具有自己不同的特点。对此在尺寸为 80mm× 3mm ,加热长度为 4 0 0mm和 80 0mm的扁矩形通道上 ,进行了上升流及下降流临界热流密度试验研究。并在此基础上 ,结合试验数据 ,对一些CHF预测关系式进行了分析评估。     

内螺纹管临界热流密度预测与数值计算

基于壁面汽泡壅塞理论,针对近临界压力区两相流动沸腾的偏离泡核沸腾(DNB)现象,对垂直上升内螺纹管的DNB型临界热流密度(CHF)进行了数值计算研究。以内螺纹管为分析对象改进已有的汽泡壅塞模型,计算了汽泡层区与主流区的极限传递质量流量、湍流速度分布、汽泡层区临界截面含气率等本构关系,汽泡脱离直径的计算考虑了汽泡接触角的影响。本文模型还根据大量CHF实验数据拟合得到了新的α_b关联式。最后,基于Fortran语言编制了CHF的理论预测数值模型程序,研究分析了压力、质量流速、热平衡干度及进口欠焓对CHF的影响,并根据CHF查表值对本文模型进行评估,同时将实验得到的内螺纹管CHF数据与采用Bowring模型、Katto模型、Shah模型和本文模型计算的CHF进行比较,发现本文模型的误差最小,与实验值吻合结果较好,说明本文模型能较好地对垂直上升内螺纹管DNB型CHF进行预测。     

近临界压力区垂直内螺纹管临界热流密度实验研究

在近临界压力区,对垂直上升内螺纹管流动沸腾的偏离泡核沸腾(DNB)型临界热流密度(CHF)现象进行了实验研究。试验段采用ф35 mm×5.67 mm六头内螺纹管。实验参数范围为:压力18~21 MPa,质量流速500~1 000kg/(m~2·s),进口过冷度3~5℃,内壁热负荷40~960kW/m~2。实验得到了不同工况下的内壁温度和传热系数分布特性,分析了流动参数对内螺纹管中DNB型CHF的影响,并根据实验数据拟合出两相区的传热关联式与临界热流密度(qCHF)预测关联式。内螺纹管的qCHF实验数据被用于与光管的qCHF预测值进行对比,发现内螺纹管具有一定的CHF强化作用,但当压力越靠近临界压力时这种作用会被抑制甚至消失。实验结果表明:在近临界压力下,内螺纹管会在低干度区甚至过冷区发生DNB现象,压力的增大和质量流速的减小均会使DNB提前发生。qCHF随压力的减小和质量流速的增大而增大。在特定工况下,试验段不同截面会分别发生偏离泡核沸腾与蒸干。     

一体化CHF关系式开发系统研制及验证

针对手动开发临界热流密度（CHF）关系式中环节众多、过程繁琐、数据量大、出错率高、耗时较长等缺点和不足,中国核动力研究设计院（NPIC）进行了一体化CHF关系式开发系统（ICODES）的研制。本文对ICODES的理论基础、系统结构等进行了说明,并基于自主化燃料组件的CHF关系式（CF-DRW关系式）形式和NPIC的棒束CHF实验数据对ICODES进行了验证。结果表明,ICODES能够满足燃料组件CHF关系式的开发需求。      

Numerical simulation and experimental research on CHF characteristics of AP1000 nuclear reactor

The present paper aims to investigate the critical heat flux (CHF) characteristics of AP1000 reactor based on the experimental and numerical researches, under normal operation and loop fault conditions, respectively. The differences of flow characteristics in these conditions were analyzed. It indicated that the flow features are very complicated in three dimensions and AP1000 has better self-regulation capability to distribute coolant flow compared to conventional reactors. Under normal operation condition, coolant of two loops is distributed along circumference of the reactor annular channel symmetrically. In case that one of the loops fails suddenly and the coolant is partially lost to total loss, the core flow distribution plate and lower grid plate cannot eliminate uneven flow immediately due to loop failure, also the nonuniformity of reactor coolant flow distribution increases gradually, which leads to the heat transfer deterioration easily. In addition, the reactor core departure from nuclear boiling ratio (DNBR) and CHF does not show a certain linear relation, and the DNBR and CHF of AP1000 are greater than that of conventional reactors which not only improve the reactor thermal efficiency, but also obviously reduce the probability of CHF phenomenon appear.     

Comparison of pool boiling CHF of a polished copper block and carbon steel block on a declined slope

This study conducts a critical heat flux (CHF) experiment on a carbon steel block, and the block is positioned on slope that is declined at angles of 5° and 10°. The results of the carbon steel block experiment were then analyzed and compared with the results obtained from a copper block experiment that had been conducted previously at the same test facility. The comparison showed that several different types of phenomena had occurred, and the carbon steel block CHF at both 5° and 10° was much lower than that of the copper block. Detailed images of the heating surface of each material were acquired by a high-speed camera under different heat fluxes and analyzed. The carbon steel block surface generates more bubbles compared to the copper block under the same heat flux, which indicates that the carbon steel block should have a large number of nucleation sites. This causes a higher CHF. Finally, several existing theories on CHF mechanisms were also analyzed in an attempt to explain the difference of copper and carbon steel. It seemed that the contact angle alone was not sufficient to explain the large CHF decrease in the carbon steel block.     

基于液膜蒸干模型的低压低流量环状流CHF理论研究

为对低压低流量下的环状流临界热流密度（CHF）进行预测,建立了考虑液膜蒸发、液滴沉积和夹带的液膜蒸干模型,并用已有的实验数据对其进行验证。计算结果表明：在实验参数范围内,CHF计算值与实验值相对偏差在25%以内,两者符合较好。以建立的环状流CHF模型为基础,研究了进口焓差、质量流速、管径和加热长度对CHF的影响。该模型能够有效地计算低压低流量环状流CHF和分析CHF随不同参数的变化趋势。     

Rod Bowed to Contact Departure from Nucleate Tests in Coldwall Thimble Cell Geometry

Departure from nucleate boiling (DNB) tests were conducted to measure the effect of a rod bowed to contact in a coldwall thimble cell of a pressurized water reactor. The tests were conducted using electrically heated 4×4 rod bundles with 4,270mm heated length and with axially and radially non-uniform heat flux, locating mixing vane grids on a 660 mm spacing. One of central heater rods was bowed into contact at the topmost grid span where most of DNB were observed in the unbowed geometry test. The tests were resulted in no effect of a rod bowed to contact at low pressure (105 and 127 ata) but considerable effect at high pressure (148 and 169 ata). The reduction of DNB heat flux was given by a function of rod average heat flux and pressure.     

纳米流体饱和池沸腾传热及CHF模型研究

纳米流体传热是一种新兴的换热方式,目前研究多集中在单相研究领域,而纳米流体沸腾传热特性的相关研究较少。本文采用热通量拆分方法,将壁面传热分为4种模型(微液层蒸发、气泡脱离前的瞬态导热、气泡脱离后的瞬态导热以及微对流换热),对这4种模型的传热量分别进行计算,结合壁面核化中心密度等参数,计算了壁面平均传热系数和CHF。结果表明,本文计算结果与国际上已发表的实验数据符合较好,充分验证了所建立模型的合理性。     

基于低压过冷流动沸腾临界热流密度数据评价各种公式、模型及95 CHF表

以低压过冷流动沸腾的临界热流密度 (CHF)的实验数据为基础 ,对典型的CHF计算公式、模型及 95CHF表作了评价。所用数据共 2 37点 ,范围如下 :压力 0 1 3～ 1 92MPa;速度 1 4 7～2 2 32m/s;出口过冷度 6 2～ 1 0 8 7K ;加热长度 2 5 5～ 40 0mm ;管径 5 1 7～ 1 6 0mm。