压水堆管道小破口蒸汽临界流泄漏实验研究

为探究压水堆核电厂小破口失水事故中管道小破口蒸汽临界流泄漏特性,开展了管道小破口泄漏实验,以探索饱和/过热蒸汽临界流泄漏特性。基于压力管道疲劳贯穿裂纹（微通道）,开展了流体压力3～12 MPa、流体温度240℃～320℃范围内的蒸汽临界流泄漏实验。实验结果表明,蒸汽临界质量流速与初始流体压力呈正相关关系,与初始流体过热度呈负相关关系。与过冷水临界流泄漏相比,蒸汽临界质量流速受入口压力损失、摩擦效应与加速效应的影响相对较弱。利用一维等熵模型预测了蒸汽临界质量流速,预测值与实验值平均相对偏差为14.17%,表明一维等熵模型具有良好的蒸汽临界质量流速预测精度。     

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。     

渐缩渐扩喷嘴临界流动的空穴核化模型

在西安交通大学高压汽水实验台架上进行了破口型式为渐缩渐扩喷嘴的临界流实验，实验参数为：入口压力３，０～１６．ＯＭＰａ，入口欠热度０～６０°Ｃ。对应临界质量流速为４０～１２０×１０３ｋｇ／ｍ２ｓ。实验表明，闪蒸起始点处的热力非平衡性随入口欠热度的增加而减小，入口为饱和状态时热力非平衡性最大。提出了一个新的空穴核化模型来预测闪蒸起始点处的减压值。该模型中含有一个由实验确定的系数，并被实验证实，该系数仅为入口欠热度的函数。将该模型计算的临界流量与临界流量的实验值进行了比较，取得了满意的结果。     

ACP100反应堆整体水力模拟试验研究

模块式先进小型压水堆（ACP100）是一种新型一体化小型反应堆，直流蒸汽发生器和主泵均直接集成在压力容器上，紧凑的结构导致其内部流场复杂。本研究应用1:3缩比模型模拟ACP100反应堆内部流场，开展反应堆整体水力模拟冷态试验。试验得到反应堆模型的总压降和分段压降，获得了反应堆模型总阻力系数以及主要流道分段阻力系数；并得到堆芯入口各燃料组件的流量分配因子。模型试验结果显示，主流道内的流动已进入第二自模区，流体的流型、流速分布以及阻力系数与原型反应堆相同；流动进入自模区后，反应堆模型的阻力系数为常数，阻力系数值为8.02，可直接用于原型反应堆压降计算；额定运行工况下，堆芯入口的流量分配因子值在0.91～1.08，满足设计需求；流量分配罩具有良好的整流作用，模拟失流事故工况下的流量分配仍较均匀。     

文丘里管空化限流现象数值模拟和实验研究

压水堆核电厂中发生超流量工况时,要求补水泵下游的文丘里流量计形成空化限流,以保护管道流量不超过限值。采用FLUENT数值模拟和高速摄像实验结合的方法,使用3种不同空化模型,对文丘里管的空化限流现象、空化发展规律和流动特性进行了研究。结果表明:采用Zwart-Gerber-Belamri（ZGB）空化模型和剪切应力输运（SST）k-ω湍流模型可对文丘里管空化限流现象进行较为准确的模拟;空化限流时文丘里管内部将发生周期性空化现象,同时将在壁面回射流的作用下发生小气泡脱落、尾部气泡脱落和空化云整体断裂式脱落等微观流动行为。       

离心式上充泵首级叶轮空化特性与试验研究

为研究和改善核电站离心式上充泵首级叶轮空化性能,采用数值模拟方法进行优化分析。将叶片数改为4片,研究了泵的最佳空化性能、扬程和效率。结果表明,最大流量工况点扬程模拟值与试验值的相对误差为2.9%,空化余量相对误差为3.6%,试验结果和模拟结果相吻合。将空化细分为初生空化、发展空化、临界空化、严重空化和断裂空化5个阶段,分析表明:初生空化时汽泡首先出现在叶片进口背面处,临界空化状态以后叶片工作面也开始出现汽泡;在发展空化到严重空化状态之间,空化和叶轮蜗壳动静干涉共同影响叶轮内的压力脉动规律;严重空化状态之后,空化成为主要影响因素,压力脉动变得相对稳定,叶轮进口和中部的压力脉动幅值明显减小,但叶轮出口处仍然保持较高幅值且比较规律的压力脉动。     

基于汽泡壅塞流模型的竖直圆管临界热流密度理论研究

为研究单管壅塞流的临界热流密度（CHF）现象,建立了基于近壁处汽泡壅塞机理的CHF计算模型。模型通过求解相应的质量、动量和能量方程,再结合汽泡直径脱离模型、壁面临界空泡份额等模型,从而计算得到CHF。将模型计算结果同实验值比较,吻合良好,验证了模型的正确性。在此基础上,以建立的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,为预测壅塞流CHF提供依据。      

基于FLUENT的核电用文丘里管稳流性能研究

采用Realizable k-?湍流模型和Zwart空化模型对某核电用空化型文丘里管的空化流动进行了数值模拟。模拟在特定工况条件下文丘里管内流动情况,得到流量变化曲线,预测空化区域,分析稳流原理和规律。模拟不同喉部直径文丘里管稳流性能,探究喉部直径变化对空化的影响。研究结果表明:随着入口压力的增大,文丘里管将发生空化塞流。将流量变化控制在一定范围内,达到相对稳流的作用。稳流时,管路压力每升高0.1 MPa,流量增加0.06 m3·h-1;喉部直径的尺寸直接影响水力空化初生与流量增幅;在一定范围内,文丘里管喉部直径大,空化流动发展迅速且流量增幅大,喉部直径小,管路流量增长幅值小。      

自然循环条件下倒U型管蒸汽发生器一次侧倒流现象关键影响因素研究

倒U型管蒸汽发生器（UTSG）在自然循环条件下存在倒流现象，影响一回路冷却剂系统载热能力及自然循环能力。本文参照芬兰压水堆热工实验装置（PWR PACTEL）中UTSG设计参数，利用计算流体力学（CFD）软件Fluent模拟流量匀速下降工况下UTSG中的倒流现象，研究一次侧运行参数、UTSG设计参数以及二次侧运行参数对于倒流现象的影响。结果表明，提高UTSG一次侧温度、一次侧运行压力、倒U型管热导率将增大UTSG的临界质量流量，使得UTSG更易发生倒流；提高UTSG二次侧给水量、二次侧温度以及倒U型管内壁粗糙度将使得UTSG临界质量流量下降，抑制倒流现象发生；而倒U型管壁厚对倒流现象几乎无影响；相较于改变二回路温度，改变一回路温度对于倒流现象的影响更为显著。本研究结果可为UTSG的参数优化提供一定参考。      

自然循环条件下蒸汽发生器倒U型管流量分配特性研究

以欧洲压水堆热工实验装置（PWR PACTEL）一回路系统蒸汽发生器为研究对象,首先,基于流体一维流动模型的质量、动量和能量守恒方程建立管道进出口压降以及传热与流体流量之间的关系;其次,以遗传算法为基础开发倒U型管蒸汽发生器流量分配计算程序,采用基准实验对程序正确性和可靠性开展验证;最后,利用流量分配程序计算蒸汽发生器倒U型管管组的流量分布情况,研究管高、管长以及一/二次侧换热系数对蒸汽发生器内流量分配的影响。结果表明,所开发流量分配程序计算结果与实验吻合良好;在选定的自然循环工况下,该蒸汽发生器中长管更易发生倒流,且倒流现象呈现分布范围广、单管流量低的特点;倒U型管内正流流速与管长成反比,与管高成正比,倒流流速随着管长的增加保持不变,与管高呈反比关系;传热系数较低时,总流量与传热系数成反比关系,当传热系数高于特定值后部分管内发生倒流,总流量骤降。      

Cavitation noise from butterfly valves

Cavitation in valves can produce levels of intense noise. It is possible to mathematically express a limit for a design level of cavitation noise in terms of the cavitation parameter σ. Using the cavitation parameter or limit, it is then possible to calculate the flow conditions at which a design level of cavitation noise will occur. However, the intensity of cavitation increases with the upstream pressure and valve size at a constant σ. Therefore, it is necessary to derive equations to correct or scale the cavitation limit for the effects of different upstream pressures and valve sizes.The following paper discusses and presents experimental data for the cavitation noise limit as well as the cavitation limits of incipient, critical, incipient damage, and choking cavitation for butterfly valves. The main emphasis is on the design limit of cavitation noise, and a noise level of 85 decibels was selected as the noise limit. Tables of data and scaling exponents are included for applying the design limits for the effects of upstream pressure and valve size.     

Modelling and computation of cavitation and boiling bubbly flows with the NEPTUNE_CFD code

This paper focuses on the modelling and the numerical simulation with the NEPTUNE_CFD code of cavitation phenomena and boiling bubbly flows.Compressible, unsteady, turbulent 3D two-phase flow is computed by the NEPTUNE_CFD solver, developed jointly by EDF R&D and CEA. The numerical approach is based on a finite-volume co-located cell-centred approach and makes use of an original pressure-based multi-field coupling algorithm [Mechitoua, N., et al., 2003. An unstructured finite volume solver for two-phase water/vapour flows modelling based on an elliptic oriented fractional step method. In: Proceedings of the 10th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH 10), Seoul, Korea].The cavitation nuclei come from wall nucleation or are pre-existing in the flow. Generated vapor bubbles are advected by the flow and expand in the regions where the local pressure is below the saturation with a tendency to agglomerate into slug bubbles.The model predictions compared with experimental data on enough selective local variables showed that satisfactory agreement could be obtained without any floating parameter to fit the data.The second part of the paper deals with boiling bubbly flow through a mixing device representing the effect of a fuel assembly spacer grid equipped with mixing blades (DEBORA-mixing experiment, CEA, Grenoble). Local measurements of the void fraction are provided downstream the mixing enhancer. The computations compare favourably with the experimental results; in particular, the global effect of the mixing blades was observed. A modification of the classical nucleate boiling model is proposed to overcome the strong model sensitivity with respect to near wall grid refinement.     

基于U型管式蒸汽发生器倒流理论模型的倒流临界简化判定准则

在U型管倒流问题中,倒流临界点的准确判断较为重要。本文建立了U型管倒流特性的理论分析模型,并基于理论分析模型对倒流临界判定准则进行了简化,将倒流理论分析模型和倒流临界简化判定准则的计算结果与实验结果进行对比。结果表明,倒流理论分析模型和倒流临界简化判定准则的计算结果与实验结果符合较好,平均相对误差的绝对值分别为3.4%和3.7%,说明倒流理论分析模型和倒流临界简化判定准则对倒流点预测结果较为准确。      

Two-phase flow regimes and mechanisms of critical heat flux under subcooled flow boiling conditions

A literature review of critical heat flux (CHF) experimental visualizations under subcooled flow boiling conditions was performed and systematically analyzed. Three major types of CHF flow regimes were identified (bubbly, vapor clot and slug flow regime) and a CHF flow regime map was developed, based on a dimensional analysis of the phenomena and available experimental information. It was found that for similar geometric characteristics and pressure, a Weber number (We)/thermodynamic quality (x) map can be used to predict the CHF flow regime.Based on the experimental observations and the review of the available CHF mechanistic models under subcooled flow boiling conditions, hypothetical CHF mechanisms were selected for each CHF flow regime, all based on a concept of wall dry spot overheating, rewetting prevention and subsequent dry spot spreading. Even though the selected concept has not received much attention (in term or theoretical developments and applications) as compared to other more popular DNB models, its basis have often been cited by experimental investigators and is considered by the authors as the “most-likely” mechanism based on the literature review and analysis performed in this work. The selected modeling concept has the potential to span the CHF conditions from highly subcooled bubbly flow to early stage of annular flow and has been numerically implemented and validated in bubbly flow and coupled with one- and three-dimensional (CFD) two-phase flow codes, in a companion paper. [Le Corre, J.M., Yao, S.C., Amon, C.H., in this issue. A mechanistic model of critical heat flux under subcooled flow boiling conditions for application to one and three-dimensional computer codes. Nucl. Eng. Des.].     

Multi-dimensional modelling of multiphase flow physics: high-speed nozzle and jet flows — a case study

Multi-dimensional modelling of multiphase flows has become more prevalent as computer capabilities have significantly expanded. Such analyses are necessary if the flow physics demonstrates behavior that is fundamentally different from the estimates of one-dimensional analyses. Multiphase multi-dimensional behavior may involve physical mechanisms that interact with the flow field transverse to the main fluid direction and feedback into downstream processes. Consider the physics of high-speed internal nozzle flow, downstream external jet flow and the dynamics of jet breakup. This is a prime example of a coupled problem where multi-dimensional aspects may need to be considered. This paper examines multiphase physics as an illustration of the conditions under which multi-dimensional modelling would be required. Internal nozzle flow can involve cavitation phenomena, and as the geometry becomes more abrupt or asymmetric, multi-dimensional modelling is required. High-speed simulations using our internal flow model, CAVALRY, indicate that cavitation behavior can become oscillatory as the nozzle shape is altered. This exiting internal flow emerges as a multi-dimensional external jet flow, whose downstream breakup can be noticeably influenced by the inlet conditions as well as the jet breakup mechanisms. Jet breakup models first developed for the TEXASV model are utilized in the multi-dimensional KIVA code simulations for gas–liquid flows. The simulation results suggest that similar jet breakup mechanisms are operative for a multi-fluid system. Our comparisons to particular sets of data for high-speed nozzle flow and jet breakup in a gas suggest that the approach can be extended to multiphase systems using similar concepts; i.e. TEXAS-3d.     

Type of flow, pressure drop, and critical heat flux of a two-phase sodium flow

Little is known about the two-phase pressure loss, the flow pattern, and the critical heat flux conditions for boiling sodium under forced convection. The specific thermohydraulic properties of sodium prohibit extrapolation to sodium of experimental data obtained for other liquids. Therefore, some new test series were carried out in a sodium loop with an induction heated test section of 9 mm inner diameter and 200 mm heated length. The two-phase pressure loss and the film thickness were measured up to the critical cooling conditions. The experimental results are compared with values predicted by known models on annular flow and annular mist flow, respectively. Satisfactory predictions of the flow pattern and the critical heat flux conditions could only be obtained using the measured two-phase pressure losses.     

An investigation of flow characteristics and critical heat flux in vertical upward round tube

Prediction of critical heat flux （CHF） in annular flow is important for the safety of once - through steam generator and the reactor core under accident conditions. The dryout in annular flow occurs at the point where the film is depleted due to entrainment, deposition, and evaporation. The film thickness, film mass flow rate along axial distribution, and CHF are calculated in vertical upward round tube on the basis of a separated flow modcl of annular flow. The theoretical CHF values are higher than those derived from experimental data, with error being within 30%.