7棒束紧密栅元流体流动传热数值研究

紧密栅元内的流体流动传热研究对高转化比反应堆燃料组件的优化有十分重要的意义。本文采用CFD方法对7棒束紧密栅元棒束通道内流体流动传热现象进行了数值模拟,并与7棒束紧密栅元内氟利昂流体传热的实验结果进行对比分析,详细分析了定位格架对棒束内流体传热流动的影响。结果表明：数值计算所得的非加热棒的壁面温度和实验吻合良好,定位格架的存在对其下游流体流动、棒束最高温度分布及交混系数有明显的影响,棒束某些位置因流动滞止导致温度大幅上升,在设计中应加以注意。     

Single-phase convective heat transfer in rod bundles

The convective heat transfer for turbulent flow through rod bundles representative of nuclear fuel rods used in pressurized water reactors is examined. The rod bundles consist of a square array of parallel rods that are held on a constant pitch by support grids spaced axially along the rod bundle. Split-vane pair support grids, which create swirling flow in the rod bundle, as well as disc and standard support grids are investigated. Single-phase convective heat transfer coefficients are measured for flow downstream of support grids in a rod bundle. The rods are heated using direct resistance heating, and a bulk axial flow of air is used to cool the rods in the rod bundle. Air is used as the working fluid instead of water to reduce the power required to heat the rod bundle. Results indicate heat transfer enhancement for up to 10 hydraulic diameters downstream of the support grids. A general correlation is developed to predict the heat transfer development downstream of support grids. In addition, circumferential variations in heat transfer coefficients result in hot streaks that develop on the rods downstream of split-vane pair support grids.     

自然循环系统摇摆条件下棒束通道内传热特性研究

棒束燃料元件子通道间流体存在搅混与横向二次流,流动及阻力特性相较矩形通道、圆管等简单通道更为复杂。核动力舰船、船舶、小型浮动核电站等会受到海浪影响,经常处于倾斜、摇摆、垂荡等瞬变运动下。目前的相关研究多集中在低压工况的研究领域,高温高压自然循环运动条件下的研究较少。本文采用实验研究方法,对自然循环系统摇摆条件下棒束通道内流动传热特性进行了研究,获得了过冷沸腾和饱和沸腾两种条件下摇摆角度和摇摆周期对棒束壁面温度变化和传热系数的影响,并获得了摇摆周期内棒束通道内的传热系数计算关系式。结果表明,饱和沸腾传热系数变化比过冷沸腾的剧烈;在本文实验工况范围内,棒表面传热系数波动幅值随着摇摆幅度的增大而增大;摇摆条件下棒束通道过冷沸腾和饱和沸腾工况时均传热系数基本不变。     

Comparison of thermo-hydraulic performances of large scale vortex flow (LSVF) and small scale vortex flow (SSVF) mixing vanes in 17 × 17 nuclear rod bundle

Spacer grids in the nuclear fuel rod assembly maintain a constant distance between rods, secure flow passage and prevent the damage of the rod bundle from flow-induced vibration. The mixing vanes attached to the spacer grids generate vortex flows in the subchannels and enhance the heat transfer performance of the rod bundle. Various types of mixing vanes have been developed to produce cross flows between subchannels as well as vortex flows in the subchannels.The shapes of the mixing vane have been improved to generate larger turbulence and cross flow mixing. In the present study, two types of large scale vortex flow (LSVF) mixing vanes and two types of small scale vortex flow (SSVF) mixing vanes are examined. SSVF-single is conventional split type and SSVF-couple is split type with different arraying method. LSVF mixing vane has different geometry and arraying method to make large scale vortex. 17 × 17 rod bundle with eight spans of mixing vanes is simulated using the IBM 690 supercomputer. The FLUENT code and IBM supercomputer is employed to calculate the flow field and heat transfer in the subchannels.Turbulence intensities, maximum surface temperatures of the rod bundle, heat transfer coefficients and pressure drops of the four kinds of mixing vanes are compared. LSVF mixing vanes produced higher turbulence intensity and heat transfer coefficient than SSVF mixing vanes. Consequently, LSVF mixing vane increases the thermal efficiency and safety of the rod bundle.     

Numerical investigation of supercritical water-cooled nuclear reactor in horizontal rod bundles

The commercial CFD code STAR-CD v4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round rods and rod bundles. Reactors with vertical or horizontal flow in the core can be found. In vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal rods and rod bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. In the rod bundle simulations, it is found that the geometry and orientation of the rod bundle have strong effects on the wall temperature distributions and heat transfers. In one orientation the square bundle has a higher wall temperature difference than other bundles. However, when the bundles are rotated by 90° the highest wall temperature difference is found in the hexagon bundle. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR.     

Local thermal–hydraulic behaviour in tight 7-rod bundles

Advanced water-cooled reactor concepts with tight lattices have been proposed worldwide to improve the fuel utilization and the economic competitiveness. In the present work, experimental investigations were performed on thermal–hydraulic behaviour in tight hexagonal 7-rod bundles under both single-phase and two-phase conditions. Freon-12 was used as working fluid due to its convenient operating parameters. Tests were carried out under both single-phase and two-phase flow conditions. Rod surface temperatures are measured at a fixed axial elevation and in various circumferential positions. Test data with different radial power distributions are analyzed. Measured surface temperatures of unheated rods are used for the assessment of and comparison with numerical codes.In addition, numerical simulation using sub-channel analysis code MATRA and the computational fluid dynamics (CFD) code ANSYS-10 is carried out to understand the experimental data and to assess the validity of these codes in the prediction of flow and heat transfer behaviour in tight rod bundle geometries. Numerical results are compared with experimental data. A good agreement between the measured temperatures on the unheated rod surface and the CFD calculation is obtained. Both sub-channel analysis and CFD calculation indicates that the turbulent mixing in the tight rod bundle is significantly stronger than that computed with a well established correlation.     

5×5花瓣形燃料棒束组件内单相流动与换热特性数值模拟研究

相较于传统圆柱形燃料棒,花瓣形燃料棒具有安全裕量高等优点,研究其在压水堆运行工况下的热工水力特性具有重要意义。本文通过STAR-CCM+对5×5花瓣形燃料棒束组件进行数值模拟研究,计算并分析了组件内二次流速度、温度、换热系数等关键热工参数,获得了入口流速、螺旋节距对组件内部流动与换热特性的影响规律。计算结果表明：花瓣形燃料棒的螺旋结构可增强冷却剂的横向流动,同一高度上燃料棒表面温度分布具有周期性,增大入口流速可增强燃料棒的表面换热,消除温度分布的不均匀性。此外,螺旋节距大于750 mm,燃料棒换热性能与无扭转的燃料棒相差不大,甚至更低。     

湍流模型对5×5格架棒束通道流动传热数值模拟影响分析

同一软件工具采用不同湍流模型进行燃料组件格架棒束通道CFD分析时会得到不同的数值结果,本文采用ANSYS CFX软件,建立了包含典型5×5格架的棒束通道CFD模型,研究了涡粘和雷诺应力两大类6种典型湍流模型对燃料组件压降与换热特性数值结果的影响,计算了压降和Nu分布结果与相似的实验结果进行对比,通过分析3个典型搅混效果评价因子,探讨了搅混翼流动与换热的内在影响关系,同时对比了不同湍流模型对结果的影响。通过与相似实验数据对比分析,认为雷诺应力模型较适宜计算本文所研究的定位格架及棒束通道内流动传热特性。     

Single-phase CFD applicability for estimating fluid hot-spot locations in a 5 × 5 fuel rod bundle

High-thermal performance PWR spacer grids require both of low pressure loss and high critical heat flux (CHF) properties. Therefore, a numerical study using computational fluid dynamics (CFD) was carried out to estimate pressure loss in strap and mixing vane structures. Moreover, a CFD simulation under single-phase flow condition was conducted for one specific condition in a water departure from nucleate boiling (DNB) test to examine the applicability of the CFD model for predicting the CHF rod position. Energy flux around the rod surface in a water DNB test is the sum of the intrinsic energy flux from a rod and the extrinsic energy flux from other rods, and increments of the enthalpy and decrements of flow velocity near the rod surface are assumed to affect CHF performance. CFD makes it possible to model the complicated flow field consisting of a spacer grid and a rod bundle and evaluate the local velocity and enthalpy distribution around the rod surface, which are assumed to determine the initial conditions for the two-phase structure. The results of this study indicate that single-phase CFD can play a significant role in designing PWR spacer grids for improved CHF performance.     

方形组件紧密栅格内非稳态流动行为的研究

棒束子通道间冷却剂的交混作用能显著降低棒束周向壁面的温差,为进一步了解紧密栅棒束内特殊的流场结构,以水为工质,对P/D=1.1的双排六棒束方形通道内的流动进行了试验研究与数值模拟。采用流场示踪方法,在Re =2 000～40 000范围内拍摄了紧密栅内棒壁间瞬态流动可视化信息,捕捉到大尺度类周期性脉动结构,并获得了该脉动流的相关特征参数。结果表明：当Re≥5 000时,大尺度脉动流发生,并在实验工况内呈很强的周期性,脉动流的波长与Re无关,脉动主频率与Re成正比;采用SSG湍流模型对相同截面通道内的流动进行了非稳态计算,模拟出棒壁狭缝处的大尺度类周期性脉动行为,计算所得脉动流各项参数与试验值符合良好。     

Numerical Investigation of Channel for Helium Cooled China HCCB-TBM First Wall by Application Enhance Heat Transfer Method

To enhance heat transfer efficiency on first wall (FW) of ITER China Helium-Cooled Ceramic Breeder-Test Blanket Module (HCCB-TBM), CFD numerical simulation method is adopted. On the basis of calculating helium gas cooling scheme of FW smooth channel, FW structural temperature gradient, maximum wall temperature, average heat transfer coefficient, and pressure drop of channel are selected as evaluation indexes. Numerical simulation comparison are performed on heat transfer schemes like placing transversal ribs and V-shaped ribs in the flow channel of front wall and the helium gas turbulence intensity and the heat transfer area are improved through optimizing the distance and angle between V-shaped ribs and other parameters to enhance heat transfer. The optimization scheme of helium-cooled FW for HCCB-TBM through the three dimensional numerical simulation is: V-shaped ribs are placed on the inner surface of front wall, the rib cross section is 1 mm × 1 mm, the distance between rib pitches is 10 mm and the rib angle is 60°. Under the same helium cooling condition, compared with the FW smooth channel, the optimized V-shaped rib scheme enhances the average heat transfer efficiency by about 70 % and the FW maximum temperature drops by 349.3 K. The result provides support for further research on FW helium cooling heat transfer enhancement experiment and engineering design optimization for China HCCB-TBM.     

Heat transfer to a supercritical pressure fluid flowing in sub-bundle channels

Supercritical pressure water cooled reactor (SCWR) has been regarded as an innovative nuclear reactor. For the design and development of the SCWR, heat transfer performance under supercritical pressure is one of the most important indicators. In this paper, experimental data are presented on the heat transfer to a supercritical pressure fluid flowing vertically upward and downward in a small diameter heated tube and two sub-bundle channels with three heater rods and seven heater rods, using HCFC22 as the test fluid. Downstream of grid spacer for the sub-bundles, heat transfer enhancement was observed in the upward flow, but not in the downward flow. The enhancement was remarkable especially when the heat transfer deterioration occurs in the fully developed region unaffected by the spacer. The heat transfer correlation for the downstream region of the spacer was developed in the normal heat transfer of sub-bundles. In the fully developed region for the sub-bundle, occurrence of the heat transfer deterioration was suppressed or degree of the deterioration was moderated. At high mass velocity for downward flow in the seven rod sub-bundle, oscillation of heat transfer was observed in the region of the enthalpy over the pseudocritical point.     

矩形窄通道再淹没现象实验和数值研究

在反应堆发生大破口事故时,再淹没阶段可以有效地降低燃料元件温度,防止堆芯熔毁。为了预测再淹没过程中板状燃料元件的换热特性,进行了竖直矩形窄缝通道底部再淹没过程的实验研究。针对实验工况,基于商用软件CFX,通过耦合分析加热板和流体的方法研究竖直矩形窄缝通道底部再淹没过程。通过将数值模拟结果与实验结果进行对比,评价了相关模型的适用性,并验证了计算流体动力学(Computational Fluid Dynamics,CFD)方法在预测再淹没过程的有效性。基于验证后计算模型,对壁面初始温度、入口流速对再淹没过程的影响进行了分析,获得了相关初始条件对壁面温度变化的影响规律。     

Laminar free convection in open-ended vertical 7-rod bundles: experiments and porous model

The flow of ambient air induced solely by buoyancy, through a vertical rod bundle has been modelled as a phenomenon in a porous medium. The rods are at uniform heat flux condition and the circular shell adiabatic. The induced flow rate was found to be controlled by a parameter ψ dependent on the heat flux, rod diameter, length, fluid properties and the bundle permeability. Measurements performed on two 7-rod bundles corroborate the theoretical predictions. Longitudinally averaged heat transfer rates from the central and peripheral rods have also been measured and average information generated for the bundle.     

杆式支撑换热器换热的数值模拟

采用数值模拟方法研究杆式支撑换热器．可以克服实验研究方式的不足。利用相似理论确定数值模拟模型,分析了各结构参数变化对换热器传热和流动性能的影响．利用最小二乘法．进一步回归出杆式支撑换热器层流下的传热和流动阻力的准数关联式．设计,建造了一套可变结构参数的杆式支撑换热器热模实验装置．在壳程单排管间布杆或双排管间布杆、不同流速和不同折流栅结构参数情况下测得了100个实验值．将实验测量值和计算值进行了比较。结果表明：数值模拟计算的Nu值和△p值与实验测量值吻合很好,由实验数据回归得到的传热和流动阻力无因次准数关系式与数值模拟程序计算得到的准数关系式也吻合较好．显示出数值模拟方式进行换热器研究的显著优点.     

联箱匹配对U型布置并联管组换热器流量分配影响研究

借助ANSYS FLUENT软件,使用Realizable k-ε湍流模型对U型布置的动量恢复型并联管组模型进行数值计算,并从联箱内压力变化的角度分析推导得出最佳联箱截面比(OARH)的计算公式。结果表明,分配联箱和汇流联箱截面比对并联管组流量分配有很大影响,使用数值计算和分析方法得出的OARH均在0.6左右。本文提出的联箱尺寸匹配方法可增强并联管组换热器流量分配均匀性,并为换热器的联箱设计提供理论指导。     

转运通道中燃料组件自然对流传热特性试验研究

针对燃料组件滞留转运通道期间的自然循环传热过程开展了试验研究。获得了承载器顶角区域加热棒的试验数据,并拟合出传热经验关系式。计算结果与试验结果比较表明,该关系式能较好地计算顶角区域加热棒顶部局部努塞尔数Nu。并通过试验数据证实了在相同的燃料棒热流密度和承载器进口水温条件下,最靠近承载器顶角位置的1号棒的传热能力最差,壁温最高。     

过冷沸腾自然对流两相CFD模拟及应用

采用计算流体力学（CFD）方法,开展过冷沸腾自然对流两相模拟与应用研究。对侧壁加热圆柱水箱过冷沸腾自然对流实验采用两相CFD瞬态模拟,模拟时间为1 500 s,通过模型设置与模拟方法研究,再现了过冷沸腾发生后实验的温度阶跃,得到与实验较一致的温度分布、气泡产生时间与产生位置,确保了数值计算的合理性与准确性。在此基础上,对以欧洲ESBWR（经济简化沸水堆）非能动安全壳冷却系统（PCCS）为原型的ISP-42实验进行了两相CFD模拟,获得与实验一致的温度分布,确定采用两相CFD数值模拟对非能动安全壳冷却系统及非能动余热排出系统进行应用研究可行,为下一步计算传热系数、构建自然对流传热模型建立了良好基础。该项研究对工程应用中探寻非能动安全壳冷却系统及非能动余热排出系统的两相自然循环传热特性具有较大价值。     

Heat transfer in turbulent longitudinal flow through unbaffled assemblies of fuel rods

A numerical analysis of heat transfer in turbulent longitudinal flow through assemblies of unbaffled fuel rods is presented. The solution applies to triangular or rectangular arrays of fuel rods with fully developed velocity and temperature profiles, for fluids with Prandtl number 1 and « 1. In the case of liquid metals, the thermal resistance of the cladding and bond are considered, but the turbulent heat transport component is neglected. For common liquids the circumferential turbulent heat transfer is considered. Results are compared in the range of dimensionless rod spacing of 1.0–1.6. Theoretical predictions and experimental results of other authors dealing with the problem show relatively good agreement.     

带格架四棒束超临界水流动传热数值分析

棒束内超临界水流动传热是超临界水堆堆芯热工水力研究的重要内容,但对其认识还十分有限。本文针对四棒束内超临界水的流动传热现象开展数值模拟,特别分析了定位格架对棒束通道内流动和传热的影响。结果表明,采用SSG湍流模型计算所得到的棒束壁面温度和实验结果吻合良好,定位格架的存在影响下游流体的速度分布,显著提高格架下游的传热特性,交混系数有大幅上升,使得加热棒周向壁面温度分布更加平均,最高温度出现位置发生改变。