Frictional pressure drop of gas liquid two-phase flow in pipes

Experiments of air water two-phase flow frictional pressure drop of vertical and horizontal smooth and relatively rough pipes were conducted, respectively. The result demonstrated that the frictional pressure drop increases with increasing relative roughness of the pipe. However, the influence of the relative roughness becomes more evident at higher vapour quality and higher mass flux. A new prediction model for frictional pressure drop of two-phase flow in pipes is proposed. The model includes a new definition of the Reynolds number and the friction factor of two-phase flow. The proposed model fits the presented experimental data very well, for vertical, horizontal, smooth and rough pipes. Therefore, the reproductive accuracy of the model is tested on the experimental data existing in the open literature and compared with the most common models. The statistical comparison, based on the Friedel’s Data-Bank containing of about 16,000 measured data, demonstrated that the proposed model is the best overall agreement with the data. The model was tested for a wide range of flow types, fluid systems, physical properties and geometrical parameters, typically encountered in industrial piping systems. Hence, calculating based on the new approach is sufficiently accurate for engineering purposes.     

Pressure drop and velocity distribution in rod bundles with spacer grids

The paper contains experimental data and analysis of the pressure drop of turbulent flow through rod bundles. For laminar flow the dependence of the pressure drop on the pitch-to-diameter and wall-to-diameter ratios is discussed on the basis of theoretical analysis. In addition, correlations for the calculation of the pressure loss due to spacer grids are presented and compared with experimental data.Detailed measurements of the velocity distribution in a full bundle of 19 rods are compared with predictions for fully developed turbulent flow. Moreover, detailed measurements of the velocity distributions upstream and downstream of spacer grids typical for LMFBRs are discussed together with the mass flow separation and redistribution between the subchannels. The mass flow distribution found experimentally is compared with the predictions by a subchannel code. The status of experimental knowledge is shown.     

Pressure Drop Experiments using Tight-Lattice 37-Rod Bundles

In order to design a Reduced-Moderation Water Reactor (RMWR) core from a thermal-hydraulic point of view, an evaluation method on the pressure drop in a tight-lattice rod bundle is required. In this study, axial pressure drops in tight-lattice 37-rod bundles were measured under conditions of 2-9 MPa in exit pressure and 200-1,000 kg/(m²·s) in mass velocity. The measured pressure drops were compared with calculated ones by the evaluation method with the Martinelli-Nelson's correlation. The comparison shows that a single-phase friction factor can be applied not only to a circular tube but also to a tight-lattice bundle except for an extremely small gap width. Then two-phase friction loss is a dominant component and accounts for about 60% of the pressure drop under an RMWR nominal operating condition. The evaluation method can evaluate effects of the flow area configuration (rod number, rod diameter, gap width) and axial power distribution under a wide range of flow conditions, and it can yield a good prediction of the pressure drop in a tight-lattice bundle.     

压水堆驱动线落棒历程计算

控制棒落棒性能验证是核电厂安全分析的重要部分,研制驱动线落棒历程计算程序有利于验证和改进控制棒驱动线设计。基于驱动线结构特点,分析运动组件的受力情况并进行分解,选择理论或数值方法逐一求取各分力的瞬态值,从而建立驱动线落棒历程的循环步进计算程序。利用秦山核电二期工程驱动线落棒性能试验数据对理论模型和程序计算结果进行对比验证。结果证明：所建立的驱动线落棒历程计算程序适用于压水堆驱动线系统,能正确地对运动组件落棒受力与运动历程进行模拟。     

Longitudinal laminar flow in asymmetrical finite bundles of rods

Theoretical analysis is presented on the asymmetrical effects, i.e. peripheral displacement of rods or unequal rod diameters, on longitudinal laminar flow in a finite bundle of rods. Solution is by the method of superposition. Numerical results are obtained for shear stress and velocity distributions when only one rod is displaced in a seven rod cluster of equal-diameter rods.     

Benchmark Analysis of TRIGA Mark II Reactivity Experiment Using a Continuous Energy Monte Carlo Code MCNP

The benchmark analysis of reactivity experiments in the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by a three-dimensional continuous-energy Monte Carlo code MCNP4A. The reactivity worth and integral reactivity curves of the control rods as well as the reactivity worth distributions of fuel and graphite elements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated values of integral reactivity curves of the control rods were in agreement with the experimental data obtained by the period method. The integral worth measured by the rod drop method was also consistent with the calculation. The calculated values of the fuel and the graphite element worth distributions were consistent with the measured ones within the statistical error estimates. These results showed that the exact core configuration including the control rod positions to reproduce the fission source distribution in the experiment must be introduced into the calculation core for obtaining the precise solution. It can be concluded that our simulation model of the TRIGA-II core is precise enough to reproduce the control rod worth, fuel and graphite elements reactivity worth distributions.     

Experimental and numerical simulation of air-water two-phase flow in the rod bundle with grid spacer

1 Introduction Grid spacer is the key part of reactor fuel assem-bly. The presence of spacers in fuel assemblies affectsvarious thermal-hydraulic characteristics of the reactorcore. The grid spacer with fine performance can im-prove thermal-hydraulic performance of the core fuelassembly and enhance the critical heat flux withouttoo much augment of the pressure loss. As a result,the implementation of grid spacer with high thermalperformance provides more thermal margin, then in-creases s…     

The SABRE code for fuel rod cluster thermohydraulics

This paper describes the capabilities of the SABRE code for the calculation of single phase and two phase fluid flow and temperature in fuel pin bundles, discusses the methods used in the modelling and solution of the problem, and presents some results including comparison with experiments.The SABRE code permits calculation of steady-state or transient, single or two phase, flows and the geometrical options include general representation of grids, wire wraps, multiple blockages, bowed pins, etc. Transient flows may be calculated using either semi-implicit or fully implicit time solution methods and the temperature distributions within the fuel pins are determined as well as the velocity and temperature of the coolant. Two phase flows are calculated using a homogeneous boiling model, with the possibility of a specified slip between the two phases. General inlet boundary conditions are available (including pressure, velocity, total mass flow) and these may vary linearly with time; the outlet boundary condition is taken as constant pressure. The treatment of grids allows for irreversible head losses at entry and exit. The wire wrap model introduces a grid resistance tensor with its principal axes along and perpendicular to the wire, resulting in a very satisfactory modelling of inducement of swirl.The derivation and solution of the difference equations is discussed. Emphasis is given to the derivation of the spatial differences in triangular subchannel geometry, and the use of central, upwind or vector upwind schemes. The method of solution of the difference equations is described for both steady state and transient problems. Together with these topics we consider the problems involved in turbulence modelling and how it is implemented in SABRE. This includes supporting work with a fine scale curvilinear coordinate programme to provide turbulence source data. The problem of modelling boiling flows is discussed, with particular reference to the numerical problems caused by the rapid density change on boiling.The final part of the paper presents applications of the code to the analysis of blockage situations, the study of flow and power transients and analysis of natural circulation within clusters to demonstrate the scope of the code and compare with available experimental results. The comparisons include the calculation of a flow pressure drop characteristic of a boiling channel showing the Ledinegg instability, examples of overpower and flow rundown transients which lead to coolant boiling, and calculation of natural circulation within a rod cluster.     

控制棒水压驱动系统落棒减速性能实验研究

控制棒水压驱动系统是清华大学为低温核供热堆发明的新型的内置式控制棒驱动技术,控制棒水力减速部件是水压驱动系统的关键部件之一,通过其对控制棒落棒过程进行减速,在保证落棒时间的前提下,降低控制棒快速落棒过程的冲击力。分析了水力减速部件组成和工作原理,确定了水力减速箱侧壁开孔方案,完成了不同开孔方案工况下控制棒水压驱动系统冷态落棒减速性能实验,在实验结果的基础上对比和分析了不同方案下的落棒减速机理和落棒过程特征参数随开孔方案的变化规律。分析结果表明:随开孔面积的增大,落棒时间逐渐减小,落棒峰值速度逐渐增大。在开孔面积大于0.004 m~2时,随开孔面积的增大,落棒峰值速度增大过程趋于平缓,落棒稳定速度和落棒延迟时间变化不大,控制棒触碰碟簧速度缓慢增大。实验研究成果为控制棒水压驱动系统落棒减速部件的理论建模和设计优化提供了基础。     

高压及超临界压力下螺旋管内单相水摩擦压降特性实验研究

在较宽的参数范围内对高压及超临界压力下单相水流经立式螺旋管的摩擦压降特性进行实验研究。研究表明,对于高压工况,在高雷诺数Re条件下的水力粗糙区,管壁粗糙度对摩擦压降的影响要远大于二次流的影响,故单相水在螺旋管内的摩擦压降基本可直接采用直管的相关公式来预测;对于超临界压力工况,在物性变化剧烈的拟临界区内,采用高压工况计算公式所得的摩擦压降预测值要明显偏低于实验值。     

水压驱动控制棒快速落棒冲击研究

控制棒水压驱动系统是由清华大学核能与新能源技术研究院发明的一项新型驱动技术。在快速落棒过程中,控制棒通过水力缓冲器进行缓冲。进行了控制棒水力缓冲性能实验,得到快速落棒过程中控制棒的关键缓冲性能参数;在实验结果基础上,运用达朗贝尔原理,将控制棒在冲击过程中所受的最大惯性力作为等效静载荷作用到控制棒上,利用有限元软件ABAQUS计算控制棒在最恶劣情况下的变形和应力分布,将计算结果与实验结果比较,验证了用简化模型代替非线性瞬态动力学分析的可行性。同时得到了控制棒在快速落棒冲击过程中不发生破坏的判据,为控制棒和水力缓冲器的设计和优化奠定了基础。     

控制棒水压驱动系统冷态性能实验研究

控制棒水压驱动系统是清华大学为低温核供热堆NHR200发明的新型的内置式控制棒驱动技术,该驱动系统由水压驱动机构、组合阀、控制棒和缓冲器等组成。控制棒水压驱动系统冷态性能是控制棒步进时间和系统驱动压力选取的基础。本文分析了控制棒水压驱动系统的组成和工作原理,完成了全尺寸控制棒水压驱动系统冷态性能实验,包括水压缸最小落棒压力实验、提升缸带载步进实验和快速落棒实验等。在实验结果的基础上分析了关键特性参数的变化规律和机理。结果表明：最小落棒压力是保持驱动机构销爪正常工作所需的最小驱动压力,其对应于压力时程曲线上峰值波动过程的变化起点;步升和步降过程压力拐点分别对应位移到位点,随着驱动水压的增加,水压缸充压拐点压力逐渐增加,步升时间、充压拐点时间逐渐减少。实验研究成果为控制棒水压驱动系统的设计、优化和工程应用奠定基础。     

Experimental and numerical studies of pressure drop in PbLi flows in a circular duct under non-uniform transverse magnetic field

Experiments and three-dimensional (3D) numerical simulations are performed to investigate the magnetohydrodynamic (MHD) characteristics of liquid metal (LM) flows of molten lead-lithium (PbLi) eutectic alloy in an electrically conducting circular duct subjected to a transverse non-uniform (fringing) magnetic field. An indirect measurement approach for differential pressure in high temperature LM PbLi is first developed, and then detailed data on pressure drop in this PbLi MHD flow are measured. The obtained experimental results for the pressure distribution are in good agreement with numerical simulations. Using the numerical simulation results, the 3D effects caused by fringing magnetic field on the LM flow are illustrated via distributions for the axial pressure gradients and transverse pressure differences. It has been verified that a simple approach for estimation of pressure drop in LM MHD flow in a fringing magnetic field proposed by Miyazaki et al. [22] i.e., a simple integral of pressure gradient along the fringing field zone using a quasi-fully-developed flow assumption, is also applicable to the conditions of the present experiment providing the magnetic interaction parameter is large enough. Furthermore, for two different sections of the LM flow at the entry to and at the exit from the magnet, it is found that the pressure distributions in the duct cross sections in these two regions are different.     

An experimental study and assessment of existing friction factor correlations for wire-wrapped fuel assemblies

The main objective of the present study is to perform a comparative study of five existing correlations that have been selected and identify the best performing correlations in the subchannel pressure drop analysis of a wire-wrapped fuel assembly by means of directly comparing with experimental data obtained in the present work. For this purpose, a series of water experiments have been performed using a helical wire-wrapped 19-pin fuel assembly for various combinations of test parameters. Four different test sections that have different pitch to rod diameter ratios (P/D) and wire lead length to rod diameter ratios (H/D) have been fabricated. A series of pressure drop measurements were made to obtain friction factors for these four test sections. A total of 293 data were obtained and the present along with existing data are used in the present comparative study of existing correlations. The results of this study show that both the original and the simplified Cheng and Todreas correlations give the best agreement with experimental data for all flow regions.     

棒束通道内两相流动摩擦阻力特性分析

常温常压下,对竖直3×3棒束通道内气液两相流动阻力特性进行了实验研究。利用所获得的实验数据,对8种典型的两相流动摩擦压降计算模型进行了评价。结果表明,均相模型在两相流速较高时精度较高,在两相流速较低时则偏差较大。分相模型中,Friedel模型和Lombodi-Pedrocchi模型不适用于本实验条件下棒束通道内气液两相流动摩擦压降的计算。Chisholm C模型、Zhang-Mishima模型、ChisholmB模型、Mishima-Hibiki模型及L.Sun模型的预测值与实验值的平均相对误差介于20%~30%之间。基于实验数据,通过修正ChisholmC模型的C系数,给出一个新的修正模型,其计算值与实验值符合良好。     

Pressure drop along longitudinally-finned seven-rod cluster nuclear fuel elements

Pressure drop measurements have been carried out on three different designs of seven-rod bundles having longitudinally-finned rods. Experiments have also been carried out for different shapes and lengths of spacers. Results obtained are reported in this paper. An analysis of the results indicates that in addition to the pitch-to-diameter ratio, wall-distance to diameter ratio may also have an important influence on the friction factors. For clusters with finned rods an equivalent rod diameter as postulated in this article may be a more appropriate parameter for comparison of friction factor for rod clusters with that for smooth tubes on the basis of pitch-to-diameter ratio. The length of the spacer has been found to have a significant influence on the pressure drop across the spacer.     

Experiment and numerical simulation of bubbly two-phase flow across horizontal and inclined rod bundles

Experimental and numerical analyses were carried out on vertically upward air-water bubbly two-phase flow behavior in both horizontal and inclined rod bundles with either in-line or staggered array. The inclination angle of the rod bundle varied from 0 to 60° with respect to the horizontal. The measured phase distributions indicated non-uniform characteristics, particularly in the direction of the rod axis when the rods were inclined. The mechanisms for this non-uniform phase distribution is supposed to be due to: (1) Bubble segregation phenomenon which depends on the bubble size and shape; (2) bubble entrainment by the large scale secondary flow induced by the pressure gradient in the horizontal direction which crosses the rod bundle; (3) effects of bubble entrapment by vortices generated in the wake behind the rods which travel upward along the rod axis; and (4) effect of bubble entrainment by local flows sliding up along the front surface of the rods. The liquid velocity and turbulence distributions were also measured and discussed. In these speculations, the mechanisms for bubble bouncing at the curved rod surface and turbulence production induced by a bubble were discussed, based on visual observations. Finally, the bubble behaviors in vertically upward bubbly two-phase flow across horizontal rod bundle were analyzed based on a particle tracking method (one-way coupling). The predicted bubble trajectories clearly indicated the bubble entrapment by vortices in the wake region.     

Control rod drop analysis by finite element method using fluid–structure interaction for a pressurized water reactor power plant

The control rod drop analysis is very important for safety analysis. For seismic and loss of coolant accident event, the control rod assemblies shall be capable of traveling from a fully withdrawn position to 90% insertion without any blockage and within specified time and displacement limits. The analysis has been executed by analytical method using in-house code. In this method, several field data are needed. These data are obtained from nuclear, thermal–hydraulic and mechanical design groups, peculiar codes, those work groups need to cooperate together.Following the enhancement of a computer and development of the multi-physics analysis code, a new method for the control rod drop analysis is proposed by finite element method. This analysis model incorporates the structure and fluid parts, termed as a fluid and structure interaction (FSI). Because a control rod is submerged inside a guide tube of a fuel assembly, the FSI boundary condition is applied. In this model, it is assumed that the fluid is incompressible laminar flow. The structures are modeled with the solid elements because there is no deformation due to the fluid flow. The analysis two-dimensional plane model is created in the analysis with considering an axi-symmetric geometry. Therefore, the proposed analysis model will be very simple and the design data from other fields will be unnecessary.The analysis results are compared with those of the in-house code, which have been used for a commercial design. After validation, it is found that the present analysis gives a useful tool in the design of the control rod and fuel assembly.     

Loading of tubular filters in AGR

This paper presents the results of experimental work concerned with the collection of dust particulate by tubular metallic filters and the effect of loading on pressure drop and flow distributions. Two different filter materials, one particular the other fibrous, and three types of filter geometries were tested. Air of atmospheric pressure and temperature was used as carrier gas for Test Dust No. 2. From the experimental results empirical equations were derived which describe the change of pressure drop across the filter wall and the variations of flow non-uniformity along the filter as a function of loading and other parameters, and also effects of suction and injection on the friction factor for the three tested candle geometries. These correlations have been incorporated in a computer code which predicts the pressure distributions, cross flow velocities and masses deposited along the filter in the loading process.     

铅铋堆零功率实验装置控制棒价值测量及方法优化

零功率实验装置的控制棒价值测量一般采用周期法、置换法或落棒法对刻棒实验进行简单处理。为提高刻棒效率,本文提出了无补偿的多步降棒刻棒方法,采用该方法对我国首个铅铋堆零功率实验装置控制棒价值进行了测量,与补偿刻棒方式及落棒法测量结果进行了对比,并通过理论计算验证了该方法的准确性。结果表明：本文方法有效降低了空间效应对测量值的影响,控制棒价值测量结果准确可靠,可在较短时间内完成较高精度的刻棒实验,适用于需经常更换装料方案的临界实验装置。