CFD analysis of flow and heat transfer of turbulent pulsating flow in a tube in rolling motion

The investigation of flow and heat transfer of turbulent pulsating flow is of vital importance to the nuclear reactor thermal hydraulic analysis in ocean environment. In this paper, the flow and heat transfer of turbulent pulsating flow is analyzed. The calculation results are firstly verified with experimental data. The agreement between them is satisfactory. The effect of spanwise and wall-normal additional forces is significant in small Reynolds number, and decreases with Reynolds number increasing. The rolling axis and rolling radius contribute slight to the flow and heat transfer. The effect of velocity oscillation period on the heat transfer is limited than that of Reynolds number and oscillating velocity Reynolds number. The traditional empirical correlations could not predict the flow and heat transfer of turbulent pulsating flow in rolling motion.     

Study on Pb-Bi-water direct contact boiling two-phase flow and heat transfer

For the development of 45w%Pb-55w%Bi cooled direct contact boiling water small fast reactor (PBWFR), Pb-Bi-water direct contact boiling two-phase flow loop has been fabricated and operated. The loop consists of a Pb-Bi flow loop (four heater pin bundle, a chimney, an upper plenum, a level meter tank, an air-water cooler, and an electromagnetic flow meter) and a water-steam flow loop (a pump, a preheated, an injection nozzle, the chimney, the upper plenum with mist separators and dryers, a condenser, a buffer tank, and an air-water cooler). At the rated operating condition system pressure is 7 MPa. The sub-cooled water was injected into a Pb-Bi flow in the chimney. A power of the heater pin bundle was controlled to obtain the inlet and outlet temperatures of the heater bundle. The Pb-Bi and steam flows were simulated analytically using one-dimensional models of frictional and form losses and a drag force. The Pb-Bi-steam two-phase frictional pressure loss was calculated by means of the two-phase flow multiplication factor of Lockhart-Martinelli model. It was found that Pb-Bi temperature decreased quickly in the chimney due to high heat transfer rate of Pb-Bi-water direct contact boiling. The volumetric overall heat transfer coefficient was 60–310 kW/m³K, and decreased with the superheat.     

The flow and heat transfer characteristics of turbulent flow in typical rod bundles in ocean environment

The flow and heat transfer characteristic of turbulent flow in typical 4 and 7 rod bundles in ocean environment is investigated theoretically. In ocean environment, the periodic variation of secondary flow in 7 rod bundles is not obvious. Because of the velocity oscillation, there is a periodic heat accumulation on the tube wall. And the restriction of the channel wall on the rolling motion is considerable. In 7 rod bundles, because of the restriction of the channel wall, the effect of the additional force perpendicular to flowing direction is limited, and the turbulent flowing and heat transfer is mainly determined by the axial turbulent intensity and inlet velocity. However, in the 4 rod bundles, the restriction of the channel wall is small. The effect of the additional force perpendicular to flowing direction on the flowing and heat transfer is significant. And the additional force perpendicular to flowing direction can also affect the Reynolds stress.     

ITER FW cooling by a flat channel, adapted to low flow rate and high pressure drop

Application of hypervapotron (HV) to cool in-vessel components of ITER - divertor and first wall (FW) - is characterized by the same design load (5 MW/m²) but water flow rate for FW is 8-9 times (almost by order!) less for parallel feeding elements so it seems it would be better to use other design. Several variants of a flat channel design different from HV are suggested that enable to adapt a channel to pressure quota up to 1 MPa and higher. A main feature of the suggested variants is a spiral or multi-spiral stream (flat multi spiral--FMS) that improves heat rejection and can be obtained both by exciting of such mode and forced by channel geometry. Comparison of the variants was carried out in simulations (Ansys CFX) as well as in experiments on the TSEFEY-M facility with electron-beam gun. It is shown that excitation of a spiral stream in a channel significantly reduces a temperature of a loaded surface of a channel. Miniature thermocouples were used to measure temperature near the surface.     

Numerical study on flow and heat transfer characteristics in the rod bundle channels under super critical pressure condition

In this study, the 3D flow and heat transfer characteristics in rod bundle channels of the super critical water-cooled reactor were numerically investigated using CFX codes. Different turbulent models were evaluated and the flow and heat transfer characteristics in different typical channels were obtained. The effect of pitch-to-diameter ratio (P/D) on the distributions of surface temperature and heat transfer coefficient (HTC) was analysed. For typical quadrilateral channel, it was found that HTC increases with P/D first and then decreases significantly when P/D is <1.4. There exists a “flat region” at the maximum value when P/D is 1.4. If P/D is larger than 1.4, heat transfer deterioration (HTD) occurs as main stream enthalpy is quite small. Furthermore, the HTD under low mass flow rate and the non-uniformity of circumferential temperature were also discussed.     

非能动余热排出热交换器流动和传热数值模拟

非能动余热排除系统(Passive Residual Heat Removal system,PRHR)是非能动核电厂的重要安全设施,在全厂断电事故下,大部分的堆芯衰变热是通过PRHR热交换器传递至内置换料水箱(In-containment Refueling Water Storage Tank,IRWST)。但PRHR热交换器属于大型非稳态换热器,其传热机理十分复杂。基于PRHR系统的重要性和复杂性,有必要研究PRHR系统的流动和传热特性。利用计算流体动力学(Computational Fluid Dynamics,CFD)软件针对非能动堆芯冷却系统试验装置中的PRHR系统进行建模计算,分析了PRHR热交换器及IRWST的流动和传热特性,发现IRWST内部沿垂直高度上呈现明显的温度分层现象,温度沿水平方向的分布趋于均匀;IRWST内部的流动主要是沿着C型传热管竖直段向上流动,流速逐渐增大,但在两相阶段,水箱上部区域流动明显增强;C型传热管上部水平段和竖直段上部区域的换热系数要明显高于其它区域,且在上部水平段与竖直段连接弯管处换热系数最大,在两相阶段,上部区域的换热系数明显增大。     

Numerical analysis of thermal-hydraulic behavior of supercritical water in vertical upward/downward flow channels

Investigations on the thermal-hydraulic behavior in the SCWR fuel assembly have obtained a significant attention in the international SCWR community. However, there is still a lack of understanding of the heat transfer behavior of supercritical fluids. In this paper, the numerical analysis is carried out to study the thermal-hydraulic behaviour in vertical sub-channels cooled by supercritical water. Remarkable differences in characteristics of secondary flow are found, especially in square lattice, between the upward flow and downward flow. The turbulence mixing across sub-channel gap for downward flow is much stronger than that for upward flow in wide lattice when the bulk temperature is lower than pseudo-critical point temperature. For downward flow, heat transfer deterioration phenomenon is suppressed with respect to the case of upward flow at the same conditions.     

Discussion on heat transfer correlation in turbulent channel flow imposed wall-normal magnetic field

Heat transfer degradation in high Prandtl number fluid was evaluated via direct numerical simulation (DNS). Target flow fields were fully developed turbulent channel flows imposed a wall-normal magnetic field in the high and low Prandtl number conditions (Pr = 5.25 and 0.025, respectively). Values of the bulk Reynolds number (Re_b = 14,000) and the Hartmann number (Ha = 0-32) were set to be equivalent to those of the previous experimental study by Yokomine et al. The numerical results of the Nusselt number for the high Prandtl number fluid were in good agreement with the experimental results by Yokomine et al. However, the magneto-hydrodynamics (MHD) effect on the heat transfer degradation was considerably larger than the empirical correlation proposed by Blum, particularly in the large interaction parameter range. On the other hand, the DNS results for the low Prandtl number fluid were consistent with the empirical correlation proposed by Blum and the experimental results by Gardner and Lykoudis.Therefore, we proposed a new correlation of the MHD heat transfer in high Prandtl number fluid (Pr = 5.25), and suggested that the empirical correlation proposed by Blum could be recommended for low Prandtl number fluid in the large interaction parameter range.     

Experimental study on friction and heat transfer characteristics of pulsating flow in rectangular channel under rolling motion

Friction and heat transfer characteristics of pulsating flow induced by rolling motion are experimentally studied. A series of single-phase forced circulation flow experiments are conducted in a vertical narrow channel. In the present study the flow rate is adjusted through control the impeller rotator speed of the pump. The results show that the flow rate pulsation simultaneously with the rolling motion and the relative amplitude of the flow rate pulsation decreases with the increasing flow rate. Accordingly, the relationships between the relative pulsation amplitude of friction factor, heat transfer coefficient and flow rate are classified. Therefore, the correlations have been developed to calculate the friction and heat transfer coefficient based on the relative pulsation amplitude of the flow rate.     

Numerical simulations on gaseous flow and heat transfer in multiple narrow channels of IFMIF High Flux Test Module

The International Fusion Materials Irradiation Facility (IFMIF) is designated to generate a materials irradiation database for the future fusion reactors. In the High Flux Test Module (HFTM) the test specimens will undergo a severe structural damage caused by neutron fluxes. The HFTM will be with helium gas. This paper presents the comprehensive thermo-hydraulic simulations of the HFTM as a part of the design activities. The turbulence models were assessed by comparing the simulations with in-house annular channel experiments. Since the required coolant flow rates are different for different compartments, multiple fluid domains were employed and simulated with appropriate turbulence (laminar) models individually. The flow distributions and heat transfer characteristics among various HFTM sub-channels will be discussed. Sensitivity study was carried out to assess the impacts of several factors on the simulation results.     

Fluid-to-fluid scaling of heat transfer in circular tubes cooled with supercritical fluids

Experimental investigations of heat transfer at prototypical conditions of supercritical water cooled reactors (SCWRs) are strongly limited due to their huge technical and financial efforts required. One of the possible solutions is the application of model fluids, which have much lower critical pressure and critical temperature. Model fluid technique has been widely applied in the thermal-hydraulic studies of nuclear engineering. In spite of growing activities of heat transfer at supercritical conditions using model fluids, there does still not exist any reliable fluid-to-fluid scaling methods, to transfer the test data in model fluids directly to the conditions of prototype fluid. This paper presents a fluid-to-fluid scaling method for heat transfer in circular tubes cooled with supercritical fluids. Based on conservation equations and boundary conditions, one set of dimensionless numbers and the requirements of a complete scaling are determined. Scaling of pressure and temperature ensures the similarity of thermo-physical properties of various fluids. A new dimensionless number, presenting the product of the so-called pseudo Boiling number, Reynolds number and Prandtl number, is applied to scale heat flux. The distortion approach is used to scale mass flux. The scaling of heat transfer coefficient is based on Nusselt number. In addition, a new approach is introduced to validate the scaling law. The validation results show good feasibility and reasonable accuracy of the proposed scaling law. Assessment of scaling factors of various parameters indicates the high feasibility of Freon-134a as model fluid for SC water. Some guidelines can be derived for the future experimental investigations on heat transfer at supercritical pressures using model fluid techniques.     

Simulation of heat transfer of supercritical water in obstacle-bearing vertical tube

The heat transfer coefficient is very low at bulk temperatures higher than the pseudo-critical point,because the supercritical pressure leads to a vapor-like fluid.In this paper,the heat transfer downstream an obstacle-bearing vertical tube is simulated by the CFD code of Fluent 6.1,using an adaptive grid in the supercritical condition.The reliable results are obtained by the RNG k-ε model using the enhanced wall treatment.The blockage ratio and local temperature of obstacle affect greatly the heat transfer enhancement,and the resultant influence region and decay trend are compared with the existing equations.     

Analysis of prompt supercritical process with heat transfer and temperature feedback

The prompt supercritical process of a nuclear reactor with temperature feedback and initial power as well as heat transfer with a big step reactivity (ρ0>β) is analyzed in this paper.Considering the effect of heat transfer on temperature of the reactor,a new model is set up.For any initial power,the variations of output power and reactivity with time are obtained by numerical method.The effects of the big inserted step reactivity and initial power on the prompt supercritical process are analyzed and discussed.It was found that the effect of heat transfer on the output power and reactivity can be neglected under any initial power,and the output power obtained by the adiabatic model is basically in accordance with that by the model of this paper,and the analytical solution can be adopted.The results provide a theoretical base for safety analysis and operation management of a power reactor.     

板状燃料组件流量分配CFD研究与优化

板状燃料组件被广泛应用于研究堆中,组件内的流量分配是设计时需要考虑的一项重要内容。计算流体动力学(Computational Fluid Dynamic,CFD)方法是研究流量分配的重要手段,但有限的计算资源限制了其在板状燃料组件流量分配研究中的推广。针对板状燃料组件冷却剂流道狭长、封闭的特点,提出了部分建模迭代求解的计算方式,将无流量分配组件与有流量分配组件两种工况下各流道流量的计算值与直接完整建模的结果进行了对比,最大误差分别为0.56%与0.81%。鉴于前者对计算资源的需求远小于后者,部分建模迭代求解可以作为板状燃料组件流量分配CFD研究的合理可信的优化方案。     

200MW低温供热堆盒间流道流动传热数值分析

采用三维CFD软件Phoenics-3.2,计算了200MW低温供热堆燃料组件盒间的流场及温场。研究了旁通流量、控制棒提升等因素的影响。在考虑这些因素之后,得出了最佳旁通入流方案。     

A quantitative estimate on the heat transfer in cylindrical fuel rods to account for flux depression inside fuel

In a nuclear reactor, the power is limited by thermal rather than by nuclear considerations. The reactor core must be operated at a power level that the temperatures of the fuel and cladding anywhere in the core must not exceed safety limits to avoid damages in the fuel elements.Heat transfer from fuel pins can be calculated analytically by using a flat power density in the fuel pin. In actual practice, the neutron flux distribution inside fuel pins results in a smaller effective distance for the heat to be transported to the coolant. This inherent phenomenon gives rise to a heat transfer benefit in fuel pin temperatures.In this research, a quantitative estimate for transferring heat from cylindrical fuel rods is accomplished by considering a non-uniform neutron flux, which leads to a flux depression factor. This, in turn, alters the temperature inside the fuel pin. A theoretical relationship combining the flux depression factor and a ratio of temperature gradients for uniform and non-uniform is derived, and a computational program, based on finite volume method and energy balance, is developed to validate the considered approximation.     

Visualization experiment of complex flow field in a sphere-packed pipe by detailed PIV measurement

A sphere-packed pipe has been proposed as a heat transfer promoter for the first wall cooling in a Flibe blanket. In this study, the flow field in a sphere-packed pipe was well investigated by means of two-dimensional PIV method by matching refractive index of a channel material and working fluid. Three-dimensional flow structure was clarified by integrating the obtained data. The feature of the flow was tortuous high-velocity region formed near pebbles and large velocity fluctuation in the vicinity of the channel wall. And, to apply this flow structure to the actual first wall cooling, a new cooling system using finger-stacked structure was proposed and discussed.     

Numerical simulation of heat transfer deterioration phenomenon in supercritical water through vertical tube

In this study, a numerical investigation of heat transfer deterioration (HTD) in supercritical water flowing through vertical tube is performed by using six low-Reynolds number turbulence models. All low-Reynolds models can be extended to reproduce the effect of buoyancy force on heat transfer and show the occurrence of localized HTD. However, most k–ε models seriously over-predict the deterioration and do not reproduce the subsequent recovery of heat transfer. The V2F and SST models perform better than other models in predicting the onset of deterioration due to strong buoyancy force. The SST model is able to quantitatively reproduce the two heat transfer deterioration phenomena with low mass flux which have been found in the present study.     

EAST-NBI反向电子吸收板换热计算与分析

中性束注入是托卡马克主要的辅助加热手段。目前先进实验超导托卡马克中性束注入(Experimental Advanced Superconducting Tokamak-Neutral Beam Injection,EAST-NBI)装置采用的是正离子源,运行过程中离子源的部分部件有很大的热沉积,其中反向电子吸收板最为严重。经实验研究,当束功率为3.5 MW时,沉积在反向电子吸收板上的平均功率密度高达4.4 MW·m~(-2),有着较大的换热负担。为探究反向电子吸收板的极限运行时间,采用数值模拟的方法,对反向电子吸收板在不同束功率、冷却水水压等多种工况进行了计算。结果表明,电子吸收板随着束功率的增大,换热负担加重、换热效果变差,通过增压泵提高冷却水入口压力一定程度上可以提高换热能力,使其极限运行时间延长。对该课题的研究可以指导EAST-NBI的运行,以保证离子源安全、稳定工作,此外还为反向电子吸收板的进一步结构优化奠定理论基础,对发展长脉冲、高功率的离子源具有重要意义。     

Liquid metal heat transfer in heat exchangers under low flow rate conditions

The present paper describes the liquid metal heat transfer in heat exchangers under low flow rate conditions. Measured data from some experiments indicate that heat transfer coefficients of liquid metals at very low Péclet number are much lower than what are predicted by the well-known empirical relations. The cause of this phenomenon was not fully understood for many years. In the present study, one countercurrent-type heat exchanger is analyzed using three, separated countercurrent heat exchanger models: one is a heat exchanger model in the tube bank region, while the upper and lower plena are modeled as two heat exchangers with a single heat transfer tube. In all three heat exchangers, the same empirical correlation is used in the heat transfer calculation on the tube and the shell sides. The Nusselt number, as a function of the Péclet number, calculated from measured temperature and flow rate data in a 50 MW experimental facility was correctly reproduced by the calculation result, when the calculated result is processed in the same way as the experiment. Finally, it is clarified that the deviation is a superficial phenomenon which is caused by the heat transfer in the plena of the heat exchanger.