环形通道内液态金属钠流动换热特性实验研究

对液态金属钠在环形通道内的单相流动换热特性进行了实验研究。结合实验数据,将液态金属钠单相流动分为层流区(Re≤2 000)、过渡区(2 000Re≤4 000)及湍流区(Re4 000),分别拟合得到不同流态下摩擦系数的计算关系式,并拟合得到液态金属钠环形通道内换热特性的相应关系式。结果表明:液态金属钠单相流动特性与常规流体(如水)类似,其层流区摩擦系数略大于水,湍流区与水的很接近。液态金属钠对流换热过程中,导热项占较大份额,同时Nu随Pe的增大而略有增大。     

环形通道内液态金属钠沸腾两相流动特性实验研究

对环形通道内液态金属钠沸腾两相流动特性进行了实验研究。实验中质量流速G≤2 000kg·m-2·s-1,系统压力p≤0.1 MPa,热流密度q≤550kW·m-2。两相流动摩擦压降通过在相同质量流量的单相流动摩擦阻力系数的基础上引入两相摩擦倍增因子来考虑两相的影响。实验结果表明:环形通道内液态金属钠两相摩擦倍增因子随Martinelli参数的增大有减小趋势。综合本文实验数据、Lurie等的实验数据以及Kaiser等的棒束实验数据,拟合得到了计算液态金属钠沸腾两相流动摩擦倍增因子的关系式。计算了本文拟合得到的关系式与各组实验数据间的相对标准偏差(RSD),表明本文关系式适用于计算环形通道内液态金属钠沸腾两相流动特性。     

棒束通道内过冷沸腾起始点的实验研究

本文以去离子水为实验介质,在进口温度80～100 ℃、质量流速0～100 kg/(m²•s)、热流密度0～80 kW/m²的条件下对棒束通道内的过冷沸腾起始点（ONB）进行了实验研究。分析了部分热工参数和棒束特殊的几何结构对ONB的影响,通过引入雷诺数,对棒束通道内ONB的数据进行非线性回归分析,得到适用于棒束通道ONB的经验关系式。结果表明：新拟合得到的关系式能较准确地预测棒束通道内ONB的热流密度,其预测值的相对误差为14.75%。     

环形通道内液态铅铋合金流动换热特性实验研究

通过对环形通道内液态铅铋合金的流动换热特性进行实验研究,得到了气泡泵注气对液态金属流动的影响,并拟合出环形通道内液态铅铋合金的摩擦系数关系式和换热特性关系式。结果表明：采用气泡泵注气能有效提升铅铋合金的质量流速;相同Reynolds数下环形通道内液态铅铋合金的摩擦系数大于由布拉休斯公式计算得到的摩擦系数;液态铅铋合金对流换热过程中,导热项占主导地位,并且Nusselt数随Peclet数的增大而增大。     

环形通道内液态铅铋合金流动换热特性实验研究

通过对环形通道内液态铅铋合金的流动换热特性进行实验研究,得到了气泡泵注气对液态金属流动的影响,并拟合出环形通道内液态铅铋合金的摩擦系数关系式和换热特性关系式。结果表明:采用气泡泵注气能有效提升铅铋合金的质量流速;相同Reynolds数下环形通道内液态铅铋合金的摩擦系数大于由布拉休斯公式计算得到的摩擦系数;液态铅铋合金对流换热过程中,导热项占主导地位,并且Nusselt数随Peclet数的增大而增大。     

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。     

环形通道内液态金属钠沸腾两相传热特性实验研究

对环形通道内液态金属钠沸腾两相流动特性进行了实验研究。实验中,系统压力为3.6~110.0kPa,热流密度为11~600kW·m~(-2),流速为0.02~0.45m·s~(-1)。实验结果表明,液态金属钠沸腾传热系数与壁面热流密度和系统压力有强烈关系,而与入口过冷度和质量流速无关。在本文实验数据基础上,拟合得到了计算液态金属钠沸腾两相传热系数的关系式,通过与各组实验数据间的比较,证明本文关系式适用于计算环形通道内液态金属钠沸腾两相传热系数。     

2×2棒束内超临界水传热特性实验研究

以中国超临界水冷堆（CSR1000）燃料组件研发为研究背景,采用实验辅以理论分析的方法,开展2×2棒束结构内超临界水工质的传热特性研究。实验工况范围为:压力（P）23~25 MPa;质量流速（G）680~1400 kg/(m2?s);热流密度（q）174~968 kW/m2。实验结果表明,随着q的增加、G的减小,2×2棒束的传热性能减弱;随着P从23 MPa变化到25 MPa,2×2棒束的传热性能变化微弱; 2×2棒束内超临界水的传热特性既与边界层和主流的物性差异程度有关,又受流道各子通道之间的流动传热不均匀性影响;基于实验数据进行多元线性回归分析,获得2×2棒束内超临界水换热关系式,约88.9%的实验数据与该换热关系式的计算值偏差范围在±25%内。      

TOPAZⅡ反应堆本体流固共轭传热数值模拟

TOPAZⅡ是用于为空间探索提供动力的一种反应堆,TOPAZⅡ堆本体内涉及液态钠钾合金流动,流体和冷却剂套管之间的换热、堆本体零部件的固体导热,反射层外壁面与外界的辐射换热等问题,本文利用计算流体力学程序CFX对TOPAZⅡ反应堆堆本体流固共轭传热进行数值模拟,数值模拟计算得到了全堆芯的流量分配数据,数据表明各通道流量分配因子偏差非常小;得到了活性区环形通道的壁面摩擦系数,摩擦系数反映了压降与流量的关系,与经验关系式计算得到的摩擦系数进行了对比,额定流量下的数值结果与经验关系式的偏差不到5%;数值模拟得到的努赛尔数与已发表的经验关系式进行了比较,最大偏差小于1%,验证了液态钠钾合金环管内的流动与换热数值模拟的可靠性与准确性。计算得到了详细的活性区慢化剂、端部铍反射层、侧铍反射层等的温度分布,所获得的计算结果可以为力学分析提供设计依据。     

棒束通道内混合对流传热实验研究

以自然循环下堆芯内可能会发生的低流量传热为研究背景,对5×5棒束通道内的混合对流传热现象进行了实验研究。实验压力为6 MPa, 质量流量为25~150 kg/(m2·s),热流密度为25~300 kW/m2,实验雷诺数Re为1000~30000,浮升力参数Bo*为2×10-7~3×10-3。实验发现,随着Bo*的增大,棒束通道内传热产生先弱化后强化的趋势。浮升力对棒束通道内传热造成影响的起始点为Bo*=3.5×10-6,当Re >15000时,浮升力依然可对传热造成弱化现象。基于实验数据,提出了适用于棒束通道的混合对流经验关系式。      

Experimental Study of Liquid Sodium Flow and Heat Transfer Characteristics Along Hexagonal 7-rod Bundle Channel

The flow and heat transfer characteristics of single-phase liquid sodium were experimentally investigated in a hexagonal 7-rod bundle channel with the velocity of 0-4 m/s, the heat flux of 0-120 kW/m² and the absolute pressure of 1.5-200 kPa. The corresponding Reynolds number ranges from 4 000 to 60 000, and the Peclet number varies from 0 to 340. The influence of some thermal parameters on the heat transfer characteristics of liquid sodium flow in a hexagonal 7-rod bundle channel was analyzed in depth. Empirical correlations of liquid sodium flow and heat transfer in a hexagonal 7-rod bundle channel were obtained by nonlinear regression analysis for experimental data. The results show that these correlations can accurately predict the friction coefficient and Nu in a hexagonal 7-rod bundle channel. The prediction error for flow and heat transfer is less than 5% and 6%, respectively. The new equation was compared with other results, and the error is within 30%. It is shown that the new empirical correlation is suitable for the flow heat transfer of liquid sodium in a hexagonal 7-rod bundle channel.     

Heat transfer from a totally blocked fuel subassembly of a liquid metal fast breeder reactor: Part I. Experimental investigation

The safety issues of liquid metal fast breeder reactors (LMFBR) are crucial due to the fact that a highly reactive and hazardous fluid like liquid sodium is used as coolant. One of the extreme cases, which can occur in a fuel subassembly of an LMFBR, is a total blockage of liquid inside the subassembly, which may lead to boiling of sodium. The present study addresses this problem by conducting experiments on a 19-rod bundle assembly enclosed inside a tall hexagonal enclosure. Liquid sodium is used as the heat transfer fluid. The natural convection mode of heat transfer is the main focus of investigation with a co-flowing air through an annular packed bed to simulate the neighbouring subassemblies. The maximum temperature achieved under different rates of power generations and air flow conditions are observed. Also the radial temperature distributions at different planes under different operating conditions of power and air flow rates have been observed. The results are of significant importance for validating analysis for the purpose of prediction of boiling incipience in an LMFBR subassembly under totally blocked condition.     

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.     

Distributed parameter analysis for the prediction of the fine structure of flow and temperature fields in wire-wrapped fuel pin bundle geometries

This paper describes the numerical method of a distributed parameter analysis code SPIRAL for the calculation of fluid flow and temperature in arbitrary channel geometries, discusses the numerical method in the modeling and solution of the problem, and presents some results, including comparison with experiments.

The derivation and solution of the finite element equations is discussed. In order to overcome difficulties arising from the geometry, the Galerkin finite element method using isoparametric elements was employed, and a procedure of finite element generation using curvilinear coordinate system was developed.

The SPIRAL code permits calculation of the fine structure of the multi-dimensional steady-state single-phase fluid flow and temperature fields in LMFBR fuel pin subassemblies in the presence of wire spacers. Calculated results are presented for crossflow velocity distributions and crossflow pressure drop characteristics in a tube bundle geometry with and without wire spacers, natural convection and heat transfer in horizontal annuli, flow in a wire-wrapped 7-pin bundle geometry and fully developed turbulent flow in a parallel 4-rod array contained in a rectangular duct.     

Flow and Heat Transfer Simulation of a Liquid-Metallic Coolant in Experimental FA Using the Sub-Channel Module Celsist

Atomic Energy - The results of modeling the flow and heat transfer of a liquid metal coolant in experimental fuel assemblies using the CELSIST sub-channel module are presented: lead in a 169-rod...     

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。     

Characterization of heat transfer and temperature distributions in an electrically heated model of an LMFBR blanket assembly

This paper describes results of an experimental program to reduce uncertainties associated with the thermal-hydraulic design and analysis of LMFBR blanket assemblies. These assemblies differ significantly from fuel assemblies in design detail and operating conditions. In blanket assemblies, heat transfer occurs over a wide range of complex operating conditions. The range and complexity of conditions are the result of flux and power gradients which are an inherent feature of the blanket region and the power generation level in an assembly which can vary from 20 kW to 2 MW. To provide effective cooling of all assemblies and economical operation, coolant is metered to groups of assemblies in proportion to their ultimate power level. As a result, the assembly flow can be in the laminar, transition or turbulent range. Because of the wide range of heat generation rates and the range of coolant flow velocities, heat transfer from rods to coolant may take place in the forced, natural or mixed convection mode. Under low flow conditions, buoyancy affects the flow pattern in the bundle, and thus, alters the temperature distribution. The complexities are further compounded since, in addition to temperature gradients within an assembly, there are also significant temperature differences between adjacent assemblies. This results in heat transfer by conduction between adjacent assemblies, which tends to further distort flow and temperature patterns.Since these effects cannot be accurately predicted analytically, full-size radial blanket assembly heat transfer tests are being conducted using electrically heated fuel rod simulators in flowing sodium. A 61-rod electrically heated radial blanket assembly mockup of prototypic dimensions was designed, constructed and installed in a 200 gpm (45 m³/hr) sodium test loop.Heat transfer tests are being conducted over a wide range of power and sodium flow rates with this full-scale, vertical, electrical-resistance-heated rod bundle. The rod bundle is extensively instrumented by thermocouples located at six distinct elevations in the wire wrap and inside the heater cladding. Tests were conducted covering the flow range from fully turbulent to fully laminar with approximately constant power-to-flow ratio. The power input patterns included across bundle gradients of 2.8 to 1 and 2.0 to 1 maximum to minimum, uniform power input to all rods and a dished distribution with low power in the central row and high power in the two rows of rods adjacent to the duct walls.The test program provided experimentally measured axial and transverse temperature profiles for the test model over a range of anticipated plant operating conditions. The data were used to (a) determine the effect of Reynolds Number, power gradients and power-to-flow ratio on transverse and axial temperature profiles and particularly on peak and peripheral channel temperatures; (b) determine the effect of inter-assembly heat transfer on peak temperatures and temperature distributions; and (c) determine the effect of buoyancy on temperature profiles.     

超临界水四棒束传热数值分析

超临界水冷堆(SCWR)开发的关键是棒束内超临界水(SCW)的热工水力特性。本文针对超临界水四棒束流动传热实验进行CFD数值模拟,SSG湍流模型的计算结果与实验结果吻合良好。分析结果表明,流动方向对棒束截面内流量分布有显著影响。与下降流相比,尽管上升流时棒束间流动搅混较弱,但上升流时棒束截面流量及壁面周向温度分布更加均匀,加热棒壁面温度更低。可见,棒束横截面上的流量分布是影响加热棒壁面流动传热的主要因素。     

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.