无内热源有序饱和多孔介质通道内气液两相流阻力特性研究

对无内热源有序饱和多孔介质内蒸汽-水两相流阻力特性进行了实验研究,在多孔介质通道内蒸汽-水两相受力分析的基础上,结合实验数据,得到了多孔介质内空泡份额及气液两相相间作用关系式,通过分析热工水力特征参数和多孔介质几何特征参数对两相流阻力特性的影响,得到了多孔介质内蒸汽-水两相流阻力关系式。结果表明,本文提出的两相流阻力关系式计算结果与实验结果符合良好,且优于其他关系式。本研究结果为进一步开展含内热源多孔介质内气液两相流阻力及传热特性研究提供了实验技术及理论依据。     

螺旋管蒸汽发生器三维热工水力程序HeTAF开发

针对核动力系统螺旋管蒸汽发生器,本文采用多孔介质方法对具有复杂换热组件区域多层螺旋套管结构进行简化,构建了壳侧工质流动换热特性数学物理模型,并基于均相流假设建立了管侧水-水蒸气两相流动沸腾换热特性分析模型,采用网格-节点映射方法实现了管壳两侧耦合传热计算,基于开源OpenFOAM平台开发了适用于螺旋管蒸汽发生器的三维全尺寸热工水力特性分析程序HeTAF。基于螺旋管两相流动沸腾换热实验开展了模型验证,并以高温气冷堆示范工程中螺旋管直流式蒸汽发生器为分析对象开展了单换热组件模拟,获得的氦气和蒸汽出口温度计算结果与设计值符合较好,表明HeTAF能有效预测换热组件内管壳两侧流动换热特性。本文的研究对螺旋管蒸汽发生器的设计和安全分析具有参考意义。     

基于JFNK的两流体三流场两相流模型的全隐算法研究

目前以两流体三流场两相流模型为数学模型的核反应堆安全分析程序大都采用半隐数值算法，数值稳定性受声速库伦特值的影响。少数以两流体三流场模型和全隐数值算法为基础的程序，采用经典牛顿迭代法求解，雅克比矩阵形成具有一定的难度。为了改善数值算法的稳定性且避免书写雅克比矩阵，一种无需形成雅克比矩阵的牛顿-Krylov迭代法（Jacobian-free Newton-Krylov,JFNK）被用于两相流全隐数值算法。两流体三流场两相流模型分别对汽相、液相和液滴相建立守恒方程，使用基于交错网格和有限体积差分全隐式离散守恒方程，线性方程组使用JFNK算法求解，当相缺失时，给缺失相一个很小的份额，以解决使用三流场模型计算单相、两相两流场时遇到的数值问题。程序模拟了Ransom水龙头数值基准题、过冷沸腾实验和Dryout传热实验，以验证了数值处理的可行性、数值算法的精度以及程序计算单相、单相到两相过渡以及两相流型之间过渡的可行性和模型计算精度。结果表明：数值算法精度较高，在从单相水、泡状流、弹状流、环雾流、弥散流等流型过渡时都表现得很好。     

Y型截止阀在液态铅铋中流阻测量与分析

液态铅铋合金是加速器驱动次临界系统(Accelerator Driven Sub-critical System,ADS)反应堆主选冷却剂材料之一,阀门是高温液态铅铋实验回路的重要组成部件之一,它的流动阻力大小直接影响整个回路装置的结构设计与安全运行。基于液态铅铋流体测量技术实验回路PREKY,开展了Y型截止阀在液态铅铋堆典型工况下的阻力特性测量实验研究,利用压差变送器获得压差实验数据,并与理论计算结果进行对比分析,验证了实验测量方法的可行性和测量结果的合理性。另外,获得了流速1.2–2.0 m·s-1内液态铅铋介质中Y型截止阀流阻理论计算指数x值为2.4,此值可直接应用于未来液态铅铋实验回路流阻计算与分析工作。     

非稳态流体力作用下两相流管束流弹失稳研究

流弹失稳会引起传热管振幅过大而发生磨损破坏,是两相流作用下蒸汽发生器管束流致振动的重要机理。为了较为准确地预测两相流作用下圆柱管的失稳临界流速,对试验测量的两相流非稳态流体力进行参数拟合,建立了气-水两相流作用下单管的动力学模型。通过无量纲化,运用Galerkin方法对方程变量进行离散后,联立求解方程得到了不同空泡份额的临界流速。数值结果表明,数值解与试验测得失稳临界流速较为吻合,验证了该模型可用于两相流传热管临界流速的预测。      

光纤双探针法测量液液两相流液滴直径

作为设计脉冲筛板柱的重要水力学特性参数,分散相液滴尺寸和液滴速率有重要的研究价值。使用内径为38mm的标准脉冲筛板柱,以水为连续相,纯煤油为分散相,利用光纤双探针法测量分散相表观流速、连续相表观流速和脉冲强度等因素对分散相液滴尺寸的影响。结果表明,光纤两相流参数测量系统能够很好地识别液液两相中的不同相,验证了光纤探针法应用于液液两相流的可行性。     

螺旋管直流蒸汽发生器两相流不稳定性

研究了螺旋管直流蒸汽发生器两相流不稳定性。阐述了两相流不稳定性机理。利用线性化频域理论，建立了螺旋管直流蒸汽发生器两相流不稳定性数学模型，编制了计算程序ＨＴＯＴＳＧＩＡ，分析了入口节流圈，系统压力及不同螺旋管圈等因素对螺旋管直流蒸汽发生器两个流不稳定性的影响，给出了螺秘管直流蒸汽发生器两相流稳定区域。计算值与实验值基本一致。     

基于分区密度补偿的稳压器液位测量研究

针对稳压器底部电热元件进行加热时，稳压器中上部和底部温度差异较大，导致传统稳压器差压法液位存在测量误差大的问题，提出了一种基于分区密度补偿的稳压器液测量方法。首先根据实际情况将稳压器分为饱和区和非饱和区，饱和区为饱和蒸汽所在区域，利用测量得到的温度对饱和蒸汽密度进行补偿；非饱和区域为介质水所在的区域，利用非饱和区域平均温度对介质水密度进行补偿。其次在稳压器饱和区和非饱和区，建立基于最小二乘法的多项式拟合模型，进行密度变量补偿，进而结合冷水段密度量进行液位计算。最后在实验装置上进行实验，并和基准液位进行比较，实验表明本文所提出的稳压器液位测量方法能够得到可靠的测量结果，因此本方法能够广泛应用于核工业等工业领域中压力容器液位测量。      

摇摆状态下两相流动局部参数光学探针测量实验

为了获取摇摆状态下气-液两相流动的局部界面信息,提出摇摆状态下通过探针获取两相流动局部时均界面参数的测量及信号处理方法。将自制的双传感器光学探针应用于摇摆状态下气-液两相流实验研究。通过实验验证摇摆状态下利用光学探针测量界面参数的可行性。相对于压降测量方法得出的空泡份额,探针测量方法的平均偏差仅为8%。     

射线法两相流测量的动态误差研究

讨论了基于射线法测量两相流参数与流体动态变化的关系．推导出由于介质动态变化导致射线计数误差与相含率测量误差的公式．并分析了动态误差与射线源及输油管道直径选择的关系．为利用射线检测两相流相含率设计方案提供了理论参考。     

Measurement System of Bubbly Flow Using Ultrasonic Velocity Profile Monitor and Video Data Processing Unit

The authors have been developing a measurement system for bubbly flow in order to clarify its multi-dimensional flow characteristics and to offer a data base to validate numerical codes for multi-dimensional two-phase flow. In this paper, the measurement system combining an ultrasonic velocity profile monitor with a video data processing unit is proposed, which can measure simultaneously velocity profiles in both gas and liquid phases, a void fraction profile for bubbly flow in a channel, and an average bubble diameter and void fraction. Furthermore, the proposed measurement system is applied to measure flow characteristics of a bubbly countercurrent flow in a vertical rectangular channel to verify its capability.     

核电汽轮机相对内效率测量方法研究

汽轮机的相对内效率是反映汽轮机运行经济性状态及通流部分运行状态的一项重要指标.核电汽轮机采用湿蒸汽作为工作介质,无法通过测量汽轮机各回热抽汽点和排汽点的湿度准确来确定汽轮机的相对内效率.通过对功率型相对内效率及焓降型相对内效率的分析比较发现,汽轮机通流部分运行状态发生变化时,两种相对内效率均可以反映出汽轮机通流部分的运行状态;汽轮机回热系统运行状态发生变化时,功率型与焓降型相对内效率在反映汽轮机通流部分运行经济状态方面是等效的,可以任选一种相对内效率作为汽轮机经济性能的评价指标.     

垂直上升管内对流沸腾瞬态数值模拟及其两相流不稳定性预测

核电站蒸汽发生器二次侧为两相对流沸腾换热过程,在设计过程中须保证其不发生两相流不稳定性。本工作采用时域法对垂直上升管内两相流不稳定性进行研究,建立了垂直上升直管内流动沸腾过程的一维模型,并编制计算程序。采用该程序模拟了流动沸腾过程气液两相流密度波的不稳定性,给出两相流波动过程瞬态参数分布,由此分析了密度波不稳定发生的机理,并分析了质量流速、系统压力、入口过冷度对不稳定的影响。结果表明,与已有实验及理论结果相比,瞬态参数计算结果与实验结果符合较好,可较好找到不同工况下直管内气液两相流发生不稳定的边界,结果优于Khabenski线算图方法。     

基于CFD方法的高温热管特性研究

高温热管运行特性的分析与预测,对热管设计和应用具有重要意义。为分析高温热管内两相流动传热特性,首先建立钠热管的计算流体力学（CFD）分析模型,并对模型计算值与钠热管稳态实验数据进行对比校核,模拟结果与实验测点温度的绝对误差小于40℃,相对误差在5%以内;其次,利用本文模型与方法对不同传热功率和倾角下的热管内部流场特性进行分析研究。研究表明,均匀加热条件下,蒸气腔内的速度在蒸发段接近线性变化,而在冷凝段,气体流速减小致使压强回升,同时,蒸气的流动压降和速度随加热功率增加呈下降趋势;在热管水平和倾角运行工况,热管内两相流动压降中液相压降均占主导;而气液间剪切效应中,气体流动速度为主导效应。本文模型可为热管堆等高温热管应用领域提供热管设计与分析方法。      

Measurement System of Bubbly Flow Using Ultrasonic Velocity Profile Monitor and Video Data Processing Unit, (II)

The authors have developed a measurement system which is composed of an ultrasonic velocity profile monitor and a video data processing unit in order to clarify its multi-dimensional flow characteristics in bubbly flows and to offer a data base to validate numerical codes for multi-dimensional two-phase flow. In this paper, the measurement system was applied for bubbly countercurrent flows in a vertical rectangular channel. At first, both bubble and water velocity profiles and void fraction profiles in the channel were investigated statistically. Next, turbulence intensity in a continuous liquid phase was defined as a standard deviation of velocity fluctuation, and the two-phase multiplier profile of turbulence intensity in the channel was clarified as a ratio of the standard deviation of flow fluctuation in a bubbly countercurrent flow to that in a water single phase flow. Finally, the distribution parameter and drift velocity used in the drift flux model for bubbly countercurrent flows were calculated from the obtained velocity profiles of both phases and void fraction profile, and were compared with the correlation proposed for bubbly countercurrent flows.     

Multi-dimensional modeling of two-phase flow in rod bundles and interpretation of velocities measured in BWRs by the cross-correlation technique

Two-phase flow in rod bundles is of the atmost importance in nuclear technology since it is a naturally occurring phenomenon in BWRs under normal operational conditions, or in PWRs undergoing a severe transient.It has recently been shown that by neutron noise analysis (cross-correlation) techniques, in the upper half of a normally operating BWR, one measures two or even three two-phase flow velocities (two or three peaks in the cross-correlation function); this was also found to be the case in measurements performed in simple air-water loops with different stationary and adiabatic two-phase flows, the direct consequence of these findings being that no cross-sectionally averaged two-phase flow models can be successfully employed for interpreting this kind of non-intrusive velocity measurements.It is the aim of this work to present an as precise as possible interpretation of velocity measurements in BWRs by the cross-correlation technique, which is based on the radially non-uniform quality and velocity distribution in BWR type bundles, as well as on our knowledge about the spatial ‘field of view’ of the in-core neutron detectors. After formulating the three-dimensional two-fluid model volume/time averaged equations and pointing out some problems associated with averaging, we expound a little on the turbulence mixing and void drift effects, as well as on the way they are modilled in advanced subchannel analysis codes like THERMIT or COBRA-TF. Subsequently, some comparisons are made between axial velocities measured in a commercial BWR by neutron noise analysis, and the steam velocities of the four subchannels nearest to the instrument tube of one of the four bundles as predicted by COBRA-III and by THERMIT. Although as expected, for well-known reasons, COBRA-III predicts subchannel steam velocities which are close to each other, THERMIT correctly predicts in the upper half of the core three largely different steam velocities in the three different types of BWR subchannels (corner, edge and interior).In the upper part of the core where a pronounced radial steam velocity and quality profile exists in the bundles, we associate the main peak of the cross-correlation function with the steam velocities in the edge subchannels, the second peak with the steam velocities in the corner subchannel, and the third small peak with the steam velocities in the interior subchannels. This interpretation is verified by a computer simulation with synthetic signals as well as by a simple phenomenological analytical model, and it opens the way for utilizing this kind of measurements (to a certain degree and within certain error bounds) for verification of advanced subchannel analysis codes like THERMIT-2 or COBRA-TF and in particular, for improving the two-phase mixing correlations employed in these codes.     

A simple model for vertical annular and horizontal stratified two-phase flows with liquid entrainment and phase transitions: one-dimensional steady state conditions

A simple one-dimensional three-fluid model is presented for the simulation and analyses of vertical annular and stratified horizontal or inclined two-phase flows. The model has been verified for various experimental data: developing annular flow, momentum transfer in an annular flow, plane flow with a hydraulic jump, flooding in a horizontal pipe, and stratified flow with direct steam condensation. Emphasis has been laid upon several mass, momentum and energy interfacial transfer processes. New correlations are proposed for the droplet entrainment intensity in annular flow and for steam direct contact condensation on the liquid film in a stratified flow. The liquid entrainment in the annular flow is correlated with the liquid film thickness. Direct contact condensation is correlated with the turbulent convective heat transfer in the liquid film. It has been shown that the present model is able to predict all dominant processes in both types of flow.     

基于RELAP5的LOCA喷放阶段下降段内CCFL特性研究

反应堆失水事故（LOCA）后下降段通道内形成的两相逆流状态极有可能引发汽-液逆向流动限制（CCFL）,不利于应急冷却水顺利进入堆芯,极大影响了核反应堆系统的安全性能。本研究基于RELAP5程序采用Wallis溢流关系式对UPFT实验装置进行建模并计算LOCA喷放阶段的下降段注水行为;通过对比下腔室蓄水量、下降段内压力及破口处蒸汽流量瞬态变化以验证模型的有效性,并对下降段通道内汽相速度场、液相体积分数分布特性进行分析。结果表明,由于下降段通道结构的三维特征引起的流动不均匀性影响了汽-液CCFL特性,随着蒸汽流量增大,在破口环路与下降段连接区域的压力梯度与向上流速度梯度越大,较少节点的划分方法很难真实反映下降段通道局部区域内汽-液溢流关系;在靠近破口的环路内注入的冷却水更难到达下腔室,而在远离破口环路的冷却水容易进入到下腔室;过热的蒸汽在流动过程中被冷却水冷却发生凝结现象,导致出口蒸汽流量小于进口蒸汽流量,且随着进口蒸汽流量的增大,凝结效应则随之减小。本研究所建立的模型与方法能够适用于LOCA喷放阶段下降段通道内的汽-液CCFL预测。      

Air/steam flow and steam wetness dependence on acoustic resonance in safety relief valves

Many experimental studies related to the flow-induced acoustic resonance closed side branches have been reported. However, few studies have reported on the effects of air/steam flow and steam wetness dependence on fluctuating pressure amplitude. Therefore, we investigated the effect of air/steam flow and steam wetness dependence on fluctuating pressure amplitude by conducting a high temperature and high pressure tests at the Hitachi Utility Steam Test Leading Facility (HUSTLE). The test section consisted of a main pipe and a side branch. The side branch was mounted on the long straight main pipe. Fluctuating pressures at the end face of the side branches were measured. The following two results were obtained; the first is that the air/steam flow had little effect on the fluctuating pressure amplitude normalized by dynamic pressure and frequency normalized by the resonance frequency; the second is that under the acoustic resonance (St = 0.41) and non-resonance (St = 0.55) conditions, fluctuating pressure and frequency changed little with steam wetness. The steam wetness during the boiling water reactor operation was less than 0.1%; thus, there was no effect of steam wetness on the acoustic pressure amplitude and the frequency under this operating condition.     

Safety Relief Valve Performance for Two-Phase Flow

The performance of main steam safety relief valve has been evaluated with respect only to the steam. In the present study, two-phase flow and subcooled water blow-out tests with model valves were performed in order to evaluate the valve's characteristics and performance. From the test results, it was made clear that not only for the steam but also for the two-phase flow the measurement data were hardly affected by scaling and also that the reaction force of the fluid to the valve stem was hardly dependent upon the void fraction. Analytical study was performed using the two-phase flow model in the valve. The results of the analysis showed good agreement with the test data. It was shown from the test and analysis results that the reaction force of the two-phase flow and subcooled water to the valve stem was almost as much as that of the steam flow, and the integrity of the safety relief valve could be maintained.