Nonlinear effects in two-phase flow dynamics

Thermally induced two-phase flow oscillations in uniformly heated boiling channels have been analyzed numerically using a one-dimensional model of two-phase flow. Two different approaches to modeling of subcooled boiling have been considered: a mechanistic model and a profile-fit model. The overall model has been numerically implemented as a computer code, DYNOBOSS, which has been validated against a linear stability analysis code and experimental data.The effects of both modeling assumptions and numerical methods of solution have been studied. It has been shown that the calculated transient response of the boiling channel may be very sensitive to the numerical scheme and spatial discretization, especially for operating conditions in the linearly unstable region. For the range of operating parameters studied, phasic slip has shown a significant stabilizing effect on the system, whereas subcooled boiling has indicated smaller influence. Furthermore, it has been shown that the rate of increase of limit cycle amplitude with channel exit quality is higher for low than high inlet subcoolings.     

竖直管道内间歇式两相流动沸腾特性分析

自然循环或重力注水过程的热功率、冷却剂流量等操作条件较小,易出现各种流动不稳定现象,影响核反应堆事故的发展进程,间歇式流动沸腾现象就属于其中的一种。以去离子水为工质,采用2×2加热棒束,对内径为32 mm竖直通道内的间歇式流动沸腾现象进行了实验研究,分析了不同热流密度下间歇式流动沸腾不稳定现象的变化规律,讨论了热流密度对间歇式沸腾周期的影响。结果表明,在一定的热流密度条件下,当加热通道内流体达到饱和并过热时,会发生周期性地剧烈喷涌及冷液回流现象,期间伴随泡状流、弹状流、搅混流及环状流等多种流动形态;间歇喷涌周期取决于沸腾停滞时间,随热流密度的不断增大,沸腾停滞时间缩短,间歇喷涌周期也缩短。当热流密度增大到一定程度时,间歇式流动沸腾现象消失,从而转变为另一种两相流动不稳定现象。     

中低压条件下矩形窄缝通道两相流动传热试验研究

在中低压条件下,对矩形窄缝通道两相流动传热进行试验研究,分析两相流动传热的变化规律,拟合出饱和沸腾传热系数计算关系式,并采用简化的一维分析方法对两相压降进行分析计算。试验结果表明:在相同热平衡含汽率(x)情况下,两相流动压降随系统压力(p)的降低而增大,随系统流量的增大而增大的变化规律;p越低,两相流动压降随x的增加而增大越剧烈;流量越大,两相流动压降随x的增加而增大越剧烈。通过数据回归方法得到汽相湿周长比例因子F并拟合了计算关系式,其计算值与试验值符合得较好。矩形窄缝通道内饱和沸腾平均传热系数受p、质量流量及热流密度的影响较大。     

Applications of ANNs in flow and heat transfer problems in nuclear engineering: A review work

Artificial Neural Networks (ANNs) have been applied to deal with flow and heat transfer problems over the past two decades. In the present paper, recent work on the applications of ANNs for predicting the flow regime, pressure drop, void fraction, critical heat flux, onset of nucleate boiling, heat transfer coefficient and boiling curve has been reviewed, respectively. As can be noted in this review work, various types of ANNs can be employed as predictors with acceptable precisions. At the end of this review, methods to improve performance of ANNs and further applications of ANNs in flow and heat transfer problems were introduced.     

Flow excursion phenomenon and its mechanism in natural circulation

An experiment was performed on the test loop (HRTL-5), which simulates the geometry and system design of a 5-MW nuclear heating reactor. In a wide range of inlet subcoolings, different flow modes, such as single-phase stable flow, subcooled boiling stable flow, subcooled boiling static flow excursion, density-wave oscillation and stable two-phase flow in the natural circulation system have been described. The phenomenon and mechanism of the static flow-excursion, which has never been studied well on this field, is especially interpreted. The experimental results show that, in the process of flow excursion, the mass flow rate and the inlet temperature decreases, while the exit temperature increases smoothly. As the process of the excursion continues for about 1 h, short period dynamic flow oscillation occurs, which can only be seen in the process of this static flow excursion, and has also never been studied well. These static and dynamic flow instabilities combine together and continue for about 2 h, then a point is reached, at which the static flow excursion disappears, but the dynamic flow oscillation continues. The mechanism of the static flow excursion is interpreted through two sets of curves for flow resistance pressure drop and driven head in natural circulation, and one curve for the natural circulation operation under special thermohydraulic condition. The study of the flow excursion and its concerned dynamic flow oscillation is of great significance for the development of the nuclear heating reactor under natural circulation.     

Sabena: Subassembly boiling evolution numerical analysis

This paper describes the computer code SABENA that has been used in subassembly sodium boiling evolution numerical analysis as a contribution to fast breeder reactor safety analysis. SABENA is a two-fluid model subchannel code system to calculate coolant boiling and two-phase flow in a rod bundle together with external loop characteristics which affects the overall boiling behavior in the bundle section. With the use of relatively simple but reasonable constitutive models, the SABENA code has been applied to and validated against many multi-pin sodium boiling problems. The results have shown excellent agreement with the experiments. The numerical methods and models employed in the code have proven to be robust and efficient in light of the extreme severity of the conditions characterizing low-pressure sodium boiling.     

On the space cross-correlation of noise sources in boiling water reactors

In this paper, an attempt has been made to investigate the space cross-correlation of boiling noise in boiling water reactors by assuming the slip ratio to be unity and power profile to be constant along the core height of the reactor. Two cases have been considered: One, no flow fluctuations (ΔV = 0) caused by boiling noise and two, finite flow fluctuations (ΔV ≠ 0) caused by the boiling noise. It has been found that finite range space cross-correlation of noise sources exists only for ΔV ≠ 0 case and not for ΔV = 0 case. Auto power spectral density of steam content fluctuations, Δα have a break frequency which is directly related with the flow velocity and attenuation coefficient, μ of the cross-correlation of noise sources in BWRs. Normalized root mean square value of Δα is more sensitive to μ in the upper half of the core for μ ≤ 10. For $\text{μ ? 20}$, it is more sensitive to μ in the lower half of the core.     

On the dynamics of nucleate region flow boiling

A transfer function to describe boiling dynamics has been developed analytically. Dynamic measurements were performed for pool biling and experimental transfer functions were developed. Flow boiling dynamics were then investigated. Results obtained for negative h region flow boiling dynamics have been presented. The present study concerns nucleate region flow boiling dynamics.     

Influence of near-wall modelling on boiling flow simulation

The capability of the three-dimensional two-fluid codes to simulate the local boiling flow processes has been assessed. Boiling flow experiments of Roy et al. [Roy, R.P., Kang, S., Zarate, J.A., Laporta, A., 2002. Turbulent subcooled boiling flow—experiments and simulations, J. Heat Transfer 124, 73-93] and Lee et al. [Lee, T.H., Park, G.C., Lee, D.J., 2002. Local flow characteristics of subcooled boiling flow of water in a vertical concentric annulus. Int. J. Multiphase Flow 28, 1351-1368], both performed in annular vertical channels were used as an experimental benchmark data set. The boiling flow is strongly affected by local mechanisms in the boundary layer near the heated wall. In this paper, the influence of near-wall modelling on the distribution of flow parameters at flow boiling has been analyzed. A generic wall function model for 3D two-fluid codes, based on surface roughness analogy has been proposed instead of commonly used single-phase log-law model. The new model has been implemented in the code CFX-4.4. In general, better agreement of phase velocities with experimental data were obtained with the new model. Presented results show that the influence of nucleating bubbles on the near-wall velocity profile should be taken into account. The second goal of this paper is to compare the NEPTUNE_CFD simulations against CFX-4.4 results and experimental data.     

在双面加热的窄缝环形流道中流动沸腾换热研究

在窄缝流道内发生沸腾换热现象时,由于沸腾产生的汽泡受窄缝流道的限制,受压变形而消除了汽泡表面张力对传热的影响。因此对此现象进行基础性理论研究具有很重要的意义。本文在常压下用蒸馏水对窄缝间隙为 0.75mm的垂直环形流道,进行了流动沸腾传热实验研究。实验段的有效加热长度为 900mm,其加热方式为内外侧双面加热,实验的流量变化范围在 1.67× 10－ 5～ 5.83× 10－ 5m3/s。通过实验得到了在不同质量流速和热流密度下双面加热的窄缝流道中内外侧沸腾换热系数随干度变化的分布和特点。研究结果表明,由于在窄缝流道中存在着大量的运动聚合受压变形汽泡,因此使内外侧沸腾换热系数都很高 (可达 105W· m－ 2· K－ 1以上 )。     

Flow rundown experiments in a seven pin bundle

A series of experiments has been carried out using an electrically heated seven pin bundle to simulate the conditions under which boiling of the sodium coolant could occur in the event of the loss of power to the circulator pumps of a fast reactor, coincident with a failure of the reactor to trip. Although it was not possible to represent the conditions of the reactor exactly, nor to continue the tests far into dryout, the results nevertheless give valuable qualitative information on the course of boiling development as well as useful quantitative information against which the predictions of computer codes can be checked. In particular, data have been obtained relating to the incidence of superheat, the location and time to dryout of the residual liquid films, the void fraction within the boiling region, and the types of flow regimes which may be expected within different parts of the boiling region at various stages of the transient.     

Advances on the research on nonlinear phenomena in boiling natural circulation loop

The state-of-the-art of theoretical investigations on the flow oscillations that occur in a boiling natural circulation loop has been presented here. Motivation behind the work is to develop a high-fidelity model that is capable of predicting nature of flow instabilities more accurately. At the low pressures and low heat fluxes conditions, the major four types of instabilities may occur in boiling natural circulation loop depending on operating conditions: Flow excursion, Geysering instability, Flashing-induced instability and Type I density-wave oscillations. The characteristics of different instabilities as well as the effects of different operating and geometric parameters on them have been reviewed. The objective of this review is to gather the research findings on the nonlinear stability phenomena in various boiling flow channel systems over a period of several years. This review indicates that most of the theoretical predictions of amplitudes and periods of the sustained oscillations are carried out using two models, namely, homogeneous equilibrium model (HEM) (still debatable) and drift-flux model (DFM) (more realistic) and are validated by experimental findings. This review work on theoretical investigations presented in this paper indicates that there are enough scopes for improving mathematical formulations of the natural circulation boiling loop (NCBL) for thermohydraulic instabilities.     

管内过冷流动沸腾CFD模型参数敏感性研究

采用CFD方法对燃料组件进行过冷流动沸腾数值模拟研究是反应堆热工水力分析的一项重要内容。本研究使用STAR CCM+基于欧拉双流体模型结合壁面沸腾模型对管内过冷流动沸腾进行数值模拟,得到了壁面温度、主流温度及空泡份额的分布。基于实验结果对网格模型、湍流模型、壁面沸腾模型及相间作用力模型的参数设置进行了敏感性分析。研究结果表明,对于欧拉双流体模型,并非网格量越多结果越准确,加热面第1层网格的高度对结果影响显著。湍流模型和曳力模型对计算结果影响较小,非曳力中的湍流耗散力及升力对结果影响较大。Li Quan或Hibiki Ishii汽化核心密度模型与Kocamustafaogullari气泡脱离直径模型组合对壁面温度及空泡份额的计算较准确。本研究可为反应堆燃料组件内过冷流动沸腾数值模拟提供参考依据。     

Study on the flow-pattern-transition instability in a natural circulation heavy water moderated boiling light water cooled reactor

A mathematical model has been developed to study the flow pattern transition instability which may occur in a boiling two-phase system. The model considers flow pattern transition criteria for vertical upward and horizontal flow in pipes to identify the flow pattern transition and flow pattern specific pressure drop models. It also considers the drift flux model to estimate the void fraction in the two-phase region. The model has been applied to predict the flow pattern transition instability in a natural circulation heavy water moderated boiling light water cooled reactor. It is found that the instability characteristics is similar to that of the Ledinegg-type instability. However, the number of multiple steady states for a given operating power can be much larger in the flow pattern transition instability as compared to that of the Ledinegg-type instability. Stability maps were plotted and compared for both the flow pattern transition instability and that of the Ledinegg-type instability. The influence of various geometric and operating parameters on this instability were investigated.     

Wall function approach for boiling two-phase flows

One of the important goals of the NURESIM project is to assess and improve the simulation capability of the three-dimensional two-fluid codes for prediction of local boiling flow processes. The boiling flow is strongly affected by local mechanisms in the turbulent boundary layer near the heated wall. Wall-to-fluid transfer models for boiling flow with the emphasis on near-wall treatment are being addressed in the paper. Since the computational grid of the 3D two-fluid models is too coarse to resolve the variable gradients in the near-wall region, the use of wall functions is a common approach to model the liquid velocity and temperature profile adjacent to the heated wall.The wall function model for momentum, based on the surface roughness analogy has been discussed and implemented in the NEPTUNE_CFD code. The model has been validated on several upward boiling flow experiments, differing in the geometry, working fluid and operating conditions. The simulations with the new wall function model show an improved prediction of flow parameters over the boiling boundary layer. Furthermore, a wall function model for the energy equation, based on enhanced two-phase wall friction has been derived and validated.     

Fundamentals of liquid metal boiling thermohydraulics

Developments in space and in nuclear-energy have demonstrated the heat-transfer utility of liquid metals in both low-pressure and high-temperature applications.Compared with water or organic liquids, liquid metals are advantageous in the design of thermal features, as for heat transport at high-temperatures at normal pressures.Nowadays, sodium is used in fast breeder reactors, because of its low neutron absorption. Furthermore, liquid metals have also been used in cooling valves, in heat pipes, and for driving Stirling engines.In fast reactors, boiling represents a disturbing occurrence which must be avoided, or controlled as much as possible.The main factors determining the course of the disturbance are the inception superheat, the flow patterns and the boiling characteristics. These factors are connected in a rather complex way. This paper deals with basic research in the liquid-metals boiling field, including the inherent tendency of liquid metals to superheat before the initial boiling point, the flow pattern during boiling, and the two-phase flow boiling characteristics.     

Analysis of low-heat-flux sodium boiling test in a 37-pin bundle by the two-fluid model computer code SABENA

This paper summarizes the development of a new detailed multi-dimensional multi-field computer code SABENA and its application to an out-of-pile low-heat-flux sodium boiling test in a 37-pin bundle. The semi-implicit numerical method employed in the two-fluid six-equation two-phase flow model has proved in solving a wide spectrum of sodium boiling transients in a rod bundle under low pressure conditions. The code is capable of predicting the spatial incoherency of the boiling, dryout on fuel cladding surfaces and fuel pin heat transfer. Essential to the successful application of such a mechanistic model computer code are validational efforts aimed at the LMFBR accident phenomenology analyses. Through the simulation of the natural circulation boiling conditions, this study provides a consistent analytical interpretation of the experimental data. The important influences of such parameters as the inlet flow restriction and bundle geometry have been examined through interpretations of two-phase flow analysis including considerations of the flow instability induced dryout mechanism.     

Computation of single- and two-phase heat transfer rates suitable for water-cooled tubes and subchannels

A computational method for predicting heat transfer, valid for a wide range of flow conditions (from pool boiling and laminar flow conditions to highly turbulent flow), has been developed. It correctly identifies the heat transfer modes and predicts the heat transfer rates as well as transition points (such as the critical heat flux point) on the boiling curve.The computational heat transfer method consists of a combination of carefully chosen heat transfer equations for each heat transfer mode. Each of these equations has been selected because of their accuracy, wide range of application, and correct asymptotic trends. Using a mechanistically-based heat transfer logic, these equations have been combined in a convenient software package suitable for PC or mainframe application. The computational method has been thoroughly tested against many sets of experimental data. The parametric and asymptotic trends of the prediction method have been examined in detail. Correction factors are proposed for extending the use of individual predictive techniques to various geometric configurations and upstream conditions.     

On the dynamics of film region flow boiling

A transfer function to describe boiling dynamics has been developed analytically. A system for control of the heater temperature in a boiling process was developed and constructed. Using this system, dynamic measurements were performed for pool boiling and experimental transfer functions were developed. Flow boiling dynamics were then investigated. Results obtained for nucleate and negative h region operation have been presented. The present study concerns film region flow boiling dynamics.