Single-phase convective heat transfer enhancement by spacer grids in a rod bundle

The spacer grids within a fuel assembly of a nuclear reactor core disrupt and re-establish the momentum and thermal boundary layers so that they enhance the local heat transfer within and downstream of the spacer grids. An experimental study in a 6×6 rod bundle has been performed to investigate the effects of spacer grids on the single-phase convective heat transfer enhancement. The experimental data showed that the Reynolds number has a significant impact on the heat transfer enhancement only when the Reynolds numbers are lower than about 10,000. The conventional correlations showed poor predictions of the heat transfer enhancement by spacer grids at low Reynolds numbers; in particular, the maximum heat transfer rate at the top end of the spacer grids was significantly overestimated. Furthermore, the conventional correlations did not properly account for the effects of the Reynolds numbers on the heat transfer enhancement. Therefore, more systematic experiments should be performed using various spacer grids with large blockage ratios at low Reynolds numbers, considering an early phase of the reflood conditions.     

肋片对圆环通道内超临界水传热特性影响的数值分析

以计算流体力学(CFD)商业软件CFX为计算平台,对圆环通道内超临界水(SCW)的传热特性进行数值模拟。通过对几种湍流模型的对比,选取对超临界条件适用性相对较好的SST模型进行计算;比较光滑圆环通道与带肋片圆环通道内的壁温分布及传热系数的变化;研究肋片形状、高度、间距及宽度等因素对传热特性的影响。结果表明:圆环通道内肋片的存在会导致局部传热强化;在本文计算的尺寸范围内,肋片形状和宽度对传热的影响很小;一定范围内,增加肋片高度以及减小肋片间距都有利于传热强化。     

CFD investigations of helical wire-wrap fuel pin bundle and its comparison with straight wire bundle

Computational Fluid Dynamics (CFD) investigations of a fast reactor fuel pin bundle wrapped with helical and straight spacer wires have been carried out and the advantages of using helical spacer wire have been assessed. The flow and temperature distributions in the fuel pin bundle are obtained by solving the statistically averaged 3-Dimensional conservation equations of mass, momentum and energy along with high Reynolds number k-ε turbulence model using a customized CFD code CFDEXPERT. It is seen that due to the helical wire-wrap spacer, the coolant sodium not only flows in axial direction in the fuel pin bundle but also in a transverse direction. This transverse flow enhances mixing of coolant among the sub channels and due to this, the friction factor and heat transfer coefficient of the coolant increase. Estimation of friction factor, Nusselt number, sodium temperature uniformity at the outlet of the bundle and clad hot spot factor which are measures of the extent of coolant mixing and non-homogeneity in heat transfer coefficient around fuel pin are paid critical attention. It is seen that the friction factor and Nusselt number are higher (by 25% and 15% respectively) for the helical wire wrap pin bundle compared to straight wire bundle. It is seen that for 217 fuel pin bundle the maximum clad temperature is 750 K for straight wire case and the same for helical wire is 720 K due to the presence of transverse flow. The maximum temperature occurs at the location of the gap between pin and wire. The ΔT between the bulk sodium in the central sub-channel and peripheral sub-channel is 30 K for straight wire and the same for helical wire is 18 K due to the presence of secondary transverse flow which makes the outlet temperature more uniform. The hotspot factor and the hot channel factors predicted by CFD simulation are 10% lower than that used in conventional safety analysis indicating the conservatism in the safety analysis.     

类四边形堆芯子通道内超临界水传热特性研究

基于类四边形堆芯子通道超临界水的传热试验,建立棒径为8 mm、栅距比为1.2的超临界水冷反应堆（SCWR）类四边形堆芯子通道物理模型,采用SSG湍流模型,在p=23～28 MPa超临界压力范围,研究了子通道内超临界水的传热特性,分析了压力、质量流速和热流密度等热工参数对类四边形子通道内超临界水传热特性的影响。研究结果表明：采用SSG湍流模型数值研究计算得到的内壁温度与试验结果变化趋势一致。在拟临界区,随压力的增大,相应的换热系数峰值逐渐降低。质量流速的增加,在整个焓区均能明显加强子通道内传热现象。随热流密度的增加,内壁温度逐渐升高,对应的换热系数峰值降低,同时逐渐向低焓区方向移动。     

环形燃料棒束再淹没行为实验研究

环形燃料是一种可在维持或提升安全裕度的前提下大幅提高反应堆经济效益的新型压水堆燃料,由于其双面冷却的特点,环形燃料在LOCA再淹没阶段的热工水力行为与传统实心燃料存在显著差异。现有关于环形燃料再淹没行为的实验研究鲜有报道。本研究基于自主设计的高温环形电加热棒建立了环形棒束再淹没实验装置,开展了3×3环形棒束底部再淹没实验研究,探究了环形棒束再淹没典型物理过程及不同工况下再淹没关键参数的变化规律。结果表明,环形棒束再淹没物理过程与传统实心棒束类似,且内外通道的骤冷前沿推进和传热模式变化趋于同步。在同一时刻下,环形棒内外壁面间存在温度梯度。骤冷前沿推进速度随再淹没速度和过冷度的增大而增大,随峰值包壳温度和线功率密度的增大而减小。此外,定位格架在低流速、低过冷度与高壁温工况下能显著提升下游的骤冷前沿推进速度。     

流量分配比对环形燃料芯块传热特性影响数值模拟研究

环形燃料具有两条冷却通道,外通道与内通道的冷却水流量分配比（φ）的变化可能会对芯块传热特性产生影响。本文建立了环形燃料单棒流固耦合CFD计算模型,在4种不同的流量分配比工况下,通过计算3个反映芯块传热特性的评价指标,研究了流量分配比变化对环形燃料芯块传热特性的影响。由分析计算结果可知,流量分配比变化不会对有间隙结构的环形燃料的芯块传热特性产生显著影响。     

Thermal-hydraulic performance analysis of a subchannel with square and triangle fuel rod arrangements using the entropy generation approach

The present paper discusses entropy generation in fully developed turbulent flows through a subchannel,arranged in square and triangle arrays. Entropy generation is due to contribution of both heat transfer and pressure drop. Our main objective is to study the effect of key parameters such as spacer grid, fuel rod power distribution,Reynolds number Re, dimensionless heat power ω, lengthto-fuel-diameter ratio λ, and pitch-to-diameter ratio ξ on subchannel entropy generation. The analysis explicitly shows the contribution of heat transfer and pressure drop to the total entropy generation. An analytical formulation is introduced to total entropy generation for situations with uniform and sinusoidal rod power distribution. It is concluded that power distribution affects entropy generation.A smoother power profile leads to less entropy generation.The entropy generation of square rod array bundles is more efficient than that of triangular rod arrays, and spacer grids generate more entropy.     

Measurements of the flow characteristics of the lateral flow in the 6 × 6 rod bundles with Tandem Arrangement Vanes

It is very important to increase the heat transfer efficiency in rod bundles in order to prevent the hot spot on the surface of fuel rods in view of the thermal hydraulic safety of nuclear power plants. It is representative to mount vanes in the support grid, which generate swirling flow. It is necessary to measure the flow pattern for investigating the thermal hydraulic flow characteristics in subchannels. In this study, it is performed to measure experimentally the flow field in cross-sections of the 6 × 6 rod bundles with new type vanes - Tandem Arrangement Vanes (TAV) by using Laser Doppler Anemometry. Through measurements, data are acquired at a nominal Reynolds number of 50,000 and for three streamwise locations at 3, 10, and 20 hydraulic diameters. Many previous experimental studies by the existing split mixing vanes show small turbulent length scales and short retention time till 10D_h after spacer grid. On the other hand, the TAVs proposed in the present study generate the big enforced swirl flow more than 20D_h after spacer grid and heat transfer effect are maintained through this distance.     

Evaluation of a numeric procedure for flow simulation of a 5 × 5 PWR rod bundle with a mixing vane spacer

The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The presence of the spacer grids promote two antagonist effects on the core: a desirable increase of the local heat transfer downstream the grids and an adverse increase of the pressure drop due to the constriction on the coolant flow area. Most spacer grids are designed with mixing vanes which cause a cross and swirl flow between and within the subchannels, enhancing even more the heat transfer performance in the grid vicinity. The improvement of the heat transfer increases the departure from the nucleate boiling ratio, allowing higher operating power in the reactor. Due to these important thermal and fluid dynamic features, experimental and theoretical investigations have been carried out in the past years for the development of spacer grid design. More recently, the Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently for this purpose. Many computational works have been performed, but the appropriate numerical procedure for the flow in rod bundle simulations is not yet a consensus. This work presents results of flow simulations performed with the commercial code CFX 11.0 in a PWR 5 × 5 rod bundle segment with a split vane spacer grid. The geometrical configuration and flow conditions used in the experimental studies performed by Karoutas et al. were assumed in the simulations. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 sub-domains. The sub-domains were simulated sequentially applying the outlet results of a previous sub-domain as inlet condition for the next. In this study the k-ε turbulence model was used. The simulations were also compared with those performed by Karoutas et al. in half a subchannel and In et al. in one subchannel computational domains. Comparison between numerical and experimental results of lateral and axial velocities along of the rod bundle show good agreement for all evaluated heights downstream the spacer grid. The present numerical procedure shows better predictions than Karoutas et al. model especially further from the spacer grid where the peripheral subchannels have more influence in the average flow.     

高压条件下燃料组件内过冷沸腾过程传热不均匀性研究

基于流体动力学软件Fluent中的流体体积函数（VOF）两相流模型,通过编写用户自定义函数（UDF）程序添加控制方程源项,建立过冷沸腾模型,对压水堆带定位格架的5×5燃料组件棒束通道内的过冷沸腾现象进行数值模拟。根据模拟结果,从空泡份额、燃料棒周向传热方面对比分析各个子通道内传热特性。研究发现各子通道内空泡份额的分布不均匀性较大,同样加热条件下,边通道的沸腾程度高于角通道。此外,对棒束周向的传热特性进行了分析,燃料棒周向努塞尔数呈不均匀性分布,燃料棒0°、90°、180°、270°等方向附近的传热能力较强,其相应的横向速度较大,对应的沸腾程度较强。      

Evaluation of a numeric procedure for flow simulation of a 5 × 5 PWR rod bundle with a mixing vane spacer

The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The presence of the spacer grids promote two antagonist effects on the core: a desirable increase of the local heat transfer downstream the grids and an adverse increase of the pressure drop due to the constriction on the coolant flow area. Most spacer grids are designed with mixing vanes which cause a cross and swirl flow between and within the subchannels, enhancing even more the heat transfer performance in the grid vicinity. The improvement of the heat transfer increases the departure from the nucleate boiling ratio, allowing higher operating power in the reactor. Due to these important thermal and fluid dynamic features, experimental and theoretical investigations have been carried out in the past years for the development of spacer grid design. More recently, the Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently for this purpose. Many computational works have been performed, but the appropriate numerical procedure for the flow in rod bundle simulations is not yet a consensus. This work presents results of flow simulations performed with the commercial code CFX 11.0 in a PWR 5 × 5 rod bundle segment with a split vane spacer grid. The geometrical configuration and flow conditions used in the experimental studies performed by Karoutas et al. were assumed in the simulations. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 sub-domains. The sub-domains were simulated sequentially applying the outlet results of a previous sub-domain as inlet condition for the next. In this study the k–ε turbulence model was used. The simulations were also compared with those performed by Karoutas et al. in half a subchannel and In et al. in one subchannel computational domains. Comparison between numerical and experimental results of lateral and axial velocities along of the rod bundle show good agreement for all evaluated heights downstream the spacer grid. The present numerical procedure shows better predictions than Karoutas et al. model especially further from the spacer grid where the peripheral subchannels have more influence in the average flow.     

Experimental investigation of heat transfer for supercritical pressure water flowing in vertical annular channels

An experiment has recently been completed at Xi’an Jiaotong University (XJTU) to obtain wall-temperature measurements at supercritical pressures with upward flow of water inside vertical annuli. Two annular test sections were constructed with annular gaps of 4 and 6 mm, respectively, and an internal heater of 8 mm outer diameter. Experimental-parameter ranges covered pressures of 23-28 MPa, mass fluxes of 350-1000 kg/m²/s, heat fluxes of 200-1000 kW/m², and bulk inlet temperatures up to 400 °C. Depending on the flow conditions and heat fluxes, two distinctive heat transfer regimes, referring to as the normal heat transfer and deteriorated heat transfer, have been observed. At similar flow conditions, the heat transfer coefficients for the 6 mm gap annular channel are larger than those for the 4 mm gap annular channel. A strong effect of spiral spacer on heat transfer has been observed with a drastic reduction in wall temperature at locations downstream of the device in the annuli. Two tube-data-based correlations have been assessed against the experimental heat transfer results. The Jackson correlation agrees with the experimental trends and overpredicts slightly the heat transfer coefficients. The Dittus-Boelter correlation is applicable only for the normal heat transfer region but not for the deteriorated heat transfer region.     

Analysis of radiant heat transfer in a BWR fuel assembly

A computer code ‘CIDER’ was developed which analyzes radiant heat transfer in a BWR fuel rod bundle under loss of coolant conditions. In the code, (1) a channel box and fuel rods are considered to be gray bodies, (2) reflection and absorption of radiation beams in the atmosphere is neglected, (3) a fuel rod is approximated by a regular polygonal rod, and (4) radiant heat flux is calculated considering circumferential temperature distribution on each fuel rod surface, which is determined from radial and circumferential heat conduction calculations in a fuel rod. It was found that the conventional model with uniform cladding temperature overestimated heat flux about 30% in a typical situation, or correspondingly underestimated the temperature rises.     

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。     

Experimental and numerical study on lead-bismuth heat transfer in a fuel rod simulator

As a task of the EU project IP EUROTRANS towards development of an Accelerator Driven System (ADS) dedicated to the transmutation of long-lived fission products, experiments and simulations were performed on the TALL test facility at KTH to investigate thermal hydraulics along a single fuel rod simulator cooled by lead-bismuth eutectic (LBE). The fuel rod simulator is concentrically inserted in a tube, so that an annular channel is formed for LBE flow. This paper presents the measured temperature profiles in the annular channel, and the comparisons with the simulation results of the CFX code. The primary objective is to help understanding the LBE heat transfer characteristics and qualifying the turbulence and heat transfer modeling for LBE application. The quantitative comparison between the calculated and measured temperatures of the LBE indicates that the simulation underestimates the experiment at most radial and axial positions. Finally the uncertainties in measurement and the deficiency in turbulence models resulting in such a disagreement were discussed, which will be directive and beneficial to future work in the field.     

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。     

CFD analysis of thermal-hydraulic behavior in SCWR typical flow channels

Investigations on thermal-hydraulic behavior in SCWR fuel assembly have obtained a significant attention in the international SCWR community. However, there is still a lack of understanding and ability to predict the heat transfer behavior of supercritical water. In this paper, CFD analysis is carried out to study the flow and heat transfer behavior of supercritical water in sub-channels of both square and triangular rod bundles. Effect of various parameters, e.g. thermal boundary conditions and pitch-to-diameter ratio on the thermal-hydraulic behavior is investigated. Two boundary conditions, i.e., constant heat flux at the outer surface of cladding and constant heat density in the fuel pin are applied. The results show that the structure of the secondary flow mainly depends on the rod bundle configuration as well as the pitch-to-diameter ratio, whereas, the amplitude of the secondary flow is affected by the thermal boundary conditions, as well. The secondary flow is much stronger in a square lattice than that in a triangular lattice. The turbulence behavior is similar in both square and triangular lattices. The dependence of the amplitude of the turbulent velocity fluctuation across the gap on Reynolds number becomes prominent in both lattices as the pitch-to-diameter ratio increases. The effect of thermal boundary conditions on turbulent velocity fluctuation is negligibly small. For both lattices with small pitch-to-diameter ratios (P/D < 1.3), the mixing coefficient is about 0.022. Both secondary flow and turbulent mixing show unusual behavior in the vicinity of the pseudo-critical point. Further investigation is needed. A strong circumferential non-uniformity of wall temperature and heat transfer is observed in tight lattices at constant heat flux boundary conditions, especially in square lattices. In the case with constant heat density of fuel pin, the circumferential conductive heat transfer significantly reduces the non-uniformity of circumferential distribution of wall temperature and heat transfer, which is favorable for the design of SCWR fuel assemblies.     

带交混翼矩形流道内超临界流动传热CFD研究

针对超临界水堆使用的棒束,取1根加热棒为中心的矩形区域作为研究对象,采用CFD中CFX10作为数值计算方法,研究了25MPa超临界压力下,有无交混翼时流体的流动和传热特点。计算结果表明：交混翼可增加流道内温度分布的均匀性,同时具有强化传热作用。     

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。