Experimental and numerical investigations of three-dimensional turbulent flow of water surrounding a CANDU simulation fuel bundle structure inside a channel

Computational fluid dynamics (CFD) is used to simulate highly turbulent coolant flows surrounding a simulation CANDU^® fuel bundle structure inside a flow channel. Three CFD methods are used: large eddy simulation (LES), detached eddy simulation (DES), and Reynolds stress model (RSM). The outcome of the simulations is compared with the experimental pressure data measured using an in-water microphone and a miniature pressure transducer placed at various locations in the vicinity of the bundle structure. Among all the three methods employed in developing computational models, LES provides the most accurate results for turbulent pressures.     

Turbulence-induced Heat Transfer in PBMR Core Using LES and RANS

This paper introduces the results of numerical simulations on flow fields and relevant heat transfer in the pebble bed reactor (PBR) core. In the core, since the coolant passes a highly complicated random flow path with a high Reynolds number, an appropriate treatment of the turbulence is required. A set of simple experiments for the flow over a circular cylinder with heat transfer was conducted to finally select the large eddy simulation (LES) and k-ω model among the considering Reynolds-averaged Navier-Stokes (RANS) models for PBR application. Using these models, the PBR cores, whose geometries were simplified to the body-centered cubical (BCC) and face-centered cubical (FCC) structures, were simulated. A larger pressure drop, a more random flow field, a higher vorticity magnitude and a higher temperature at the local hot spots on the pebble surface were found in the results of the LES than in those of RANS for both geometries. In cases of the LES, the flow structures were resolved up to the grid scales. Irregular distributions of the flow and local heat transfer were found in the BCC core, while relatively regular distributions for the FCC core. The turbulent nature of the coolant flow in the pebble core evidently affected the fuel surface temperature distribution.     

Numerical Study of Particle Concentration Effect on Deposition Characteristics in Turbulent Pipe Flows

The present study focuses on the particle concentration effect on the particle deposition onto a channel wall in numerically simulated turbulent pipe flows. Large eddy simulation of the incompressible Navier-Stokes equations was performed to calculate the time-dependent turbulent flow field of continuous gas phase. Considering the technical application to the droplet motion in boiling water reactor subchannels, the flow direction was set to be vertically upward. The particles were placed at random initial locations in the pipe. The subsequent particle motion was tracked individually using a simple Lagrangian equation. To investigate the particle concentration effect on the rate of particle deposition, “two-way method” was applied, in which the two-way interactions between the continuous phase and the particles were taken into consideration. The calculated results showed that the mass transfer coefficient of particle deposition decreased noticeably with an increase in the particle concentration. This tendency was consistent with the experimental observations and empirical correlations. The present numerical results indicated that the turbulence modulation in the continuous gas phase is one of the primary causes of the reduction of the deposition mass transfer coefficient observed under high-droplet-concentration conditions.     

铅铋冷却快堆含绕丝燃料组件子通道程序开发与验证

由于铅铋冷却剂流动传热现象的复杂性,准确计算铅铋冷却含绕丝燃料组件的冷却剂和包壳温度是液态金属冷却快堆燃料组件热工分析的重点。本文基于集总参数法对守恒方程进行求解,开发了适用于铅铋冷却快堆的子通道分析程序,对液态铅铋在棒束燃料组件中的摩擦阻力模型、湍流交混模型和对流换热模型进行了适用性分析,并对7棒束大涡模拟和19棒束含绕丝传热实验进行了对比验证。结果表明：包壳和冷却剂温度的最大相对误差低于5%。程序能较好完成铅铋冷却含绕丝燃料组件的热工水力计算,可为铅铋冷却快堆设计提供支持。     

数值堆热工流体程序CVR-PACA验证及典型应用

谱元方法是一种高精度的数值计算方法,采用该方法开发了数值堆高精度热工水力并行CFD计算程序CVR-PACA。应用CVR-PACA对单棒光棒通道湍流流场、3×3光棒棒束湍流流场、Matis-H压水堆棒束通道基准题、19棒带绕丝组件通道湍流流场进行了仿真计算。通过与实验测量值对比,研究定量验证了大涡模拟(LES)模型及非稳态雷诺时均(URANS)模型对各类棒束通道流场预测的准确性。算例在建模过程中采用网格分裂技术实现了复杂几何的纯六面体网格划分,用于支撑谱元方法计算。研究较为全面地积累了高精度谱元方法模拟流场流动及换热的建模经验,获取了各类棒束通道内丰富的流动和换热细节,获得的建模经验能更加精准有力地指导相关设计的优化改进。     

液态铅铋合金湍流普朗特数及RANS模型优选

工程上常采用RANS湍流模型进行热工水力相关的数值模拟,然而液态铅铋合金（LBE）具有独特的热物性,常规湍流普朗特数模型和RANS湍流模型对其流动与传热模拟的适用性有待研究。为更准确地描述绕丝燃料组件内LBE的流动与换热过程,本文基于大涡模拟对湍流普朗特数模型和RANS湍流模型进行优选。首先,采用四种湍流普朗特数模型对绕丝燃料组件内LBE的流动与传热过程进行大涡模拟,对比分析实验数据和模拟结果并进行模型优选。基于优选的湍流普朗特数模型,评价RANS湍流模型对LBE数值模拟的适用性和准确性。结果表明,Cheng湍流普朗特数模型和SST k-ω模型对LBE流动与传热模拟的准确性和适用性最高。     

加/减速流动下棒束通道内速度分布和湍流特性研究

事故工况及海洋条件下反应堆处于非稳态工况,堆芯燃料组件内热工水力行为复杂多变,对反应堆安全提出了更高挑战,因此有必要对非稳态下燃料组件内流动换热特性开展研究。基于粒子图像测速（PIV）技术,结合远心镜头和脉冲控制器,实现对燃料组件内复杂流场的高时空分辨率、长时间的连续测量,获得了流量波动下燃料组件内时空演变的流场结构,分析了棒束通道内速度分布、湍流强度、雷诺应力等瞬时流场信息的空间演变特性。以定常流动下流场分布特性为基准,对比分析了加速度对燃料组件内空间流场分布的贡献特点。实验结果表明：加速流动提高了棒束通道内流层之间的速度梯度,抑制了横向速度和湍流强度;减速流动减弱了棒束通道内流层之间的速度梯度,提高了横向速度和湍流强度。实验结果有助于揭示燃料组件在非稳态条件下的瞬态特性,并为燃料组件的设计和优化奠定基础。     

高温气冷堆热气导管中石墨粉尘沉积特性分析

高温气冷堆中石墨粉尘的运动规律对反应堆安全具有重要意义。本工作采用计算流体力学方法得到热气导管中的温度场分布,在此基础上分析了热气导管中热泳沉积与湍流沉积的规律。结果表明,随着颗粒粒径的增大,热泳沉积率下降,而湍流沉积率则先减少后增大。通过比较30%FP及100%FP两种反应堆功率下的计算结果发现：反应堆功率为30%FP时对应的热泳沉积率更高,而反应堆功率为100%FP时,湍流沉积率增长更加迅速。当颗粒粒径较小时,热泳沉积与湍流沉积作用相当,颗粒粒径较大时,湍流沉积明显占主导地位。最后,采用最可几分布的粒径计算了热气导管中石墨粉尘总的沉积量,计算结果表明石墨粉尘沉积总量有限。     

Development,computer simulation and performance testing in sodium of an eddy current flowmeter

Sodium is used as a coolant in Liquid Metal Fast Breeder Reactor (LMFBR). Sodium flow measurement is of prime importance both from the operational and safety aspects of a fast reactor. Various types of flowmeters namely permanent magnet, saddle type and eddy current flowmeters are used in FBRs. From the safety point of view flow through the core should be assured under all operating conditions. This requires a flow sensor which can withstand the high temperature sodium environment and can meet the dimensional constraints and be amenable to maintenance. Eddy current flowmeter (ECFM) is one such device which meets these requirements. It is meant for measuring flow in PFBR primary pump and also at the outlets of the fuel sub-assemblies to detect flow blockage. A simulation model of ECFM was made and output of ECFM was predicted for various flowrates and temperatures. The simulation model was validated by testing in a sodium loop. This paper deals with the design, simulation and tests conducted in sodium for the eddy current flowmeter for use in the Prototype Fast Breeder Reactor (PFBR).     

Turbulent velocity and temperature distribution in the central subchannel of rod bundles

In this paper a method is given for solving the momentum and heat transfer equations for the central subchannel of a reactor subassembly in general curvilinear orthogonal coordinates. For turbulent flow, the eddy diffusivities are determined by Prandtl's ‘mixing length’ hypothesis. A new method is proposed to determine eddy diffusivities parallel to the wall. The eddy diffusivities of heat are calculated from those of momentum using the relations obtained by various authors, and the results are compared in the case of sodium. To show the capability of the computer codes developed, the three-dimensional temperature field is calculated in the central subchannel of a fuel element cooled by sodium and helium. The agreement between calculated and experimental results is satisfactory.     

Large eddy simulation of turbulent flow surrounding two simulated CANDU fuel bundles

Three-dimensional turbulent flow in one and a half simulated 43-element CANDU^®1 fuel bundle at the inlet of a fuel channel is solved using large eddy simulation. Wake generated after the endplates and the flow development in the inlet bundle are investigated based on the simulation results. Spatial distribution and frequency spectra of fluid force components are also examined. The simulation results are compared to experimental data available in the literature as well as the measurements conducted by the authors. The current investigation provides a basic understanding of the fluid excitation in a simulated 43-element fuel bundle. The results may be used in a flow-induced vibration analysis for fuel bundles.     

湍流模型在堆芯热工水力性能分析中的应用

针对正三角形布置堆芯棒束燃料通道内冷却剂充分发展湍流流场模拟,对比分析了计算流体动力学软件湍流模型对复杂流道内湍流流场模拟结果的影响。结果表明：湍流模型选取的不同对模拟结果有着显著影响,由于堆芯几何结构复杂,冷却剂流动为复杂三维流动,湍流呈高度各向异性。基于各向同性假设的湍流模型不能准确捕捉堆芯内冷却剂的二次流现象。基于求解雷诺应力输运方程的雷诺应力模型（RSM）能够较好地预测复杂流道内的二次流。本工作的研究结果为复杂流道流动换热模拟及深入研究分析堆芯热工水力性能提供了一定借鉴和指导。     

基于γ测量的球床式高温气冷堆内乏燃料球探测方案的模拟分析

在球床式高温气冷堆中,对排出堆芯的乏燃料球的探测和数量统计是燃料监测的重要内容。按国际原子能机构针对球床式高温气冷堆核安保的要求,对于燃料装卸系统管道内的燃料监测应开发一种独立于现有涡流检测原理的新监测方案。本文提出了一种基于γ测量原理的新探测方案,设计了探测器构型,对其在堆稳态运行时的探测功能进行验证。结合球床式高温气冷堆HTR-10的燃料球放射性核素数据,及对不同球速不同燃耗的燃料球经过探测区域过程的蒙特卡罗模拟分析,验证了此方案对单个燃料球鉴别和计数的可靠性,同时证明了该方案对于燃料球球流探测的可行性,为今后该探测方案的完善和实际装置的制作提供了设计基础。     

Performance evaluation of eddy current flowmeter in Monju

Measurement of the temperature and flow rate at each fuel subassembly outlet is an effective way for a liquid metal fast breeder reactor to detect a loss of coolant accident or reactivity-initiated accident in the early stage and to understand the reactor’s thermal hydrodynamic performance. Japan Atomic Energy Agency has developed the eddy current flowmeter in practical use and installed 34 of them in the upper core structure of fast breeder reactor, Monju. This report presents data obtained by using the flowmeters in Monju. We observed high linearity between each of the flowmeter’s signal intensity and the primary sodium’s flow rate under 10–100% flow rate condition. High linearity was also observed in a region of low velocity (approx. 0.25 m/s). The fluctuation of flow rate observed by the flowmeters was below 0.2 m/s which is 5% of the time-averaged velocity under a rated condition. These experimental results show that the eddy current flowmeter is an effective tool to detect the changes in relative flow rate.     

基于开放栅格结构的空间堆三维热工水力特性研究

高温气冷堆结合磁流体发电是一种高效的空间电源系统,可以满足空间任务对于大功率、高效率的需求,具有广阔的应用前景。本文参考美国普罗米修斯计划中的开放栅格方案,结合磁流体发电需满足的设计条件,提出了一种由三角形布置、217根燃料棒构成的堆芯方案。在通过试验数据确定流动模型后,对该空间堆进行了三维建模,并在考虑气隙结构、燃料棒功率分布及堆内辐射的基础上研究其热工水力特性,重点针对环境温度及外壁面发射率展开了热工参数敏感性分析。计算结果表明,该堆芯热工设计满足材料温度、压降限值等指标要求。冷却剂在燃料区横向流动不明显,不存在复杂涡结构,流动现象相对较为简单。稳态热工计算结果对环境温度的改变并不敏感,但发射率的改变影响相对较大。     

Investigation on Slit Jet through Upper Internal Structure (UIS) in Highly Compact Vessel of Sodium-Cooled Fast Reactor

An advanced loop-type sodium-cooled fast reactor has been developed by the Japan Atomic Energy Agency. The upper internal structure (UIS) above the core is a key component where control rod guide tubes are housed. A radial slit is set in the UIS to simplify the fuel-handling system and to reduce the reactor vessel diameter. A high-velocity upward flow is formed in the UIS slit. This slit jet influences thermal hydraulic issues in the reactor vessel. A water experiment was carried out to understand the flow field in the UIS, which is composed of the control rod guide tubes and several horizontal perforated plates with a slit. A refractive index matching method was applied to visualize the flow in such a complex geometry. Velocity measurement using particle image velocimetry showed that the velocity in the UIS slit was accelerated by the multiple slits and kept at a high value at the mid-height of the reactor upper plenum. A numerical simulation was carried out for this complex geometry of the UIS to obtain an adequate simulation method. A comparison between the experimental and analytical velocity profiles showed that the numerical simulation is highly applicable.     

基于格子Boltzmann方法的反应堆核-热-流耦合模拟

反应堆堆芯内部存在多种不同物理场之间的相互作用和反馈,对其准确模拟需要考虑这些物理过程之间的耦合。为了降低堆芯核 热 流耦合模拟的实现难度,消除不同物理场之间的外部插值过程,本文构建了核 热 流耦合模拟的格子Boltzmann方法（LBM）,将中子输运（包括SN方程、SP3方程以及扩散方程）、考虑燃料流动效应的缓发中子先驱核守恒方程以及流动传热方程统一到相似的LBM格式下,采用统一的LBM碰撞 迁移过程进行求解,有效降低了堆芯多物理耦合模拟的实现难度。计算结果表明：本文建立的核 热 流耦合LBM模型对不同雷诺数下的流动效应均能准确模拟,同时温度反馈在高温熔盐堆低速流动条件下有较为明显的影响,不能忽略;提高堆芯熔盐流速能够有效地展平功率及温度分布。     

Estimation of fuel and clad temperature of a research reactor during dry period of de-fuelling operation

Heat transfer and fluid flow studies related to spent fuel bundle of a research reactor in fuelling machine has been carried out. When the fuel is in reactor core, the heat generated in the fuel bundle is removed by heavy water under normal reactor operation. However, during the de-fuelling operation, the fuel bundle is exposed to air for some period called dry period. During this period, the decay heat from fuel bundle has to be removed by air flow. This flow of air is induced by natural convection only. In this period, the temperatures of fuel and clad rise. If clad temperature rises beyond a certain limit, structural failure may occur. This failure can result into release of fission products from fuel rod. Hence the temperature of clad has to be within specified limit under all conditions. The objective of this study is to estimate the clad temperature rise during the dry period.In the CFD simulation, the turbulent natural convection flow over fuel and radiation heat transfer are accounted. Standard k-? model for turbulence, Boussinesq approximation for computing the natural convection flow and IMMERSOL model for radiation are used.The steady state and transient CFD simulation of flow and heat is performed, using the CFD code PHOENICS. The steady state analysis provides the maximum temperature the clad will attain if fuel bundle is left exposed to air for sufficiently long time. For safe operation, the clad temperature should be limited to a specified value. From steady state CFD analysis, it is found that steady state clad temperature for various decay powers is higher than the limiting value. Hence transient analysis is also performed. In the transient analysis, the variation of clad temperature with time is predicted for various decay powers. Safe dry time, i.e. the time required for clad to reach the limiting value, is predicted for various decay powers. Determination of safe dry time helps in deciding the time available to the operator to drop the bundle in light water pool for storage. The analysis is found useful in optimizing the de-fuelling process.     

Numerical Simulation of Fluid Flow and Heat Transfer of Supercritical Fluids in Fuel Bundles

A supercritical water-cooled reactor (SCWR) was proposed as a kind of generation IV reactor in order to improve the efficiency of nuclear reactors. Although investigations on the thermal-hydraulic behavior in SCWR have attracted much attention, there is still a lack of CFD study on the heat transfer of supercritical water in fuel channels. In order to understand the thermal-hydraulic behavior of supercritical fluids in nuclear reactors, the local fluid flow and heat transfer of supercritical water in a 37-element fuel bundle has been studied numerically in this work. Results show that secondary flow appears and the cladding surface temperature (CST) is very nonuniform in the fuel bundle. The maximum cladding surface temperature (MaxCST), which is an important design parameter for SCWR, can be predicted and analyzed using the CFD method. Due to a very large circumferential temperature gradient in cladding surfaces of the fuel bundle, the precise cladding temperature distributions using the CFD method is highly recommended.