CFD in supercritical water-cooled nuclear reactor (SCWR) with horizontal tube bundles

The commercial CFD code STAR-CD 4.02 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 tubes and tube bundles. Reactors with vertical or horizontal flow in the core can be found. In a vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in a horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal tubes and tube 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. From the study of single round tubes, the Speziale quadratic non-linear high-Re k-? turbulence model with the two-layer model for near wall treatment is found to produce the best results in comparison with experimental data. In tube bundle simulations, it is found that the temperature is higher in the top half of the bundle and the highest tube wall temperature is located at the outside tubes where the flow rate is the lowest. The secondary flows across the bundle are highly complex. Their main effect is to even out the temperature over the area within each individual recirculation region. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR.     

CFD investigation of vertical rod bundles of supercritical water-cooled nuclear reactor

The commercial CFD code STAR-CD v4.02 is used as the numerical simulation tool for the supercritical water-cooled nuclear reactor (SCWR). The numerical simulation is based on the real full 3D rod bundles’ geometry of the nuclear reactors. For satisfying the near-wall resolution of y⁺ ≤ 1, the structure mesh with the stretched fine mesh near wall is employed. The validation of the numerical simulation for mesh generation strategy and the turbulence model for the heat transfer of supercritical water is carried out to compare with 3D tube experiments. After the validation, the same mesh generation strategy and the turbulence model are employed to study three types of the geometry frame of the real rod bundles. Through the numerical investigations, it is found that the different arrangement of the rod bundles will induce the different temperature distribution at the rods’ walls. The wall temperature distributions are non-uniform along the wall and the values depend on the geometry frame. At the same flow conditions, downward flow gets higher wall temperature than upward flow. The hexagon geometry frame has the smallest wall temperature difference comparing with the others. The heat transfer is controlled by P/D ratio of the bundles.     

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。     

Assessment of Applicability of Two-Fluid Model Code ACE-3D to Heat Transfer Test of Supercritical Water Flowing in an Annular Channel

A supercritical-water-cooled reactor (SCWR) is a high-temperature, high-pressure water cooled reactor that operates above the critical pressure of water. In order to perform efficiently the thermal design of the SCWR, it is important to assess the thermal hydraulics in rod bundles of the core. Experimental conditions of mockup tests, however, may be limited because of technical and financial reasons. Therefore, it is required to establish an analytical design technique that can extrapolate experimental data to various design conditions of the reactor. Japan Atomic Energy Agency (JAEA) has improved the three-dimensional two-fluid model analysis code ACE-3D, which was originally developed for the two-phase flow thermal hydraulics of light water reactors, to handle the thermal hydraulic properties of water in the supercritical region. In the present study, heat transfer experiments of supercritical water flowing in a vertical annular channel around a heater pin, which were performed at JAEA, were analyzed with the improved ACE-3D to assess the prediction performance of the code. As a result, it was implied that the ACE-3D code is applicable to the prediction of wall temperatures of a single rod that simulates the fuel bundle geometry of the SCWR core.     

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。     

Numerical analysis of heat transfer in supercritical water cooled flow channels

Research activities are ongoing worldwide to develop nuclear power plants with a supercritical water cooled reactor (SCWR) with the purpose to achieve a high thermal efficiency and to improve their economical competitiveness. However, there is still a big deficiency in understanding and prediction of heat transfer in supercritical fluids. In this paper, heat transfer of supercritical water has been investigated in various flow channels using the computational fluid dynamics (CFD) code CFX-5.6 to provide basic knowledge of the heat transfer behaviour and to gather the first experience in the application of CFD codes to heat transfer in supercritical fluids. Three different flow channels are selected, i.e. circular tubes, the sub-channel of a square-array rod bundle and the sub-channel of a triangular-array rod bundle. The effect of mesh structures, turbulence models, as well as flow channel configurations is analysed. Based on the present results, recommendations are made on the application of turbulence models to the heat transfer of supercritical fluids in various flow channels. A new definition for the onset of heat transfer deterioration is proposed. A strong non-uniformity of heat transfer is observed in sub-channel geometries. This non-uniformity has to be taken into account in the design of fuel assemblies of SCWR.     

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.     

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.     

CFD analyses in tight-lattice subchannels and seven-rod bundle geometries of a Super Fast Reactor

This paper presents CFD analyses in heat unsymmetric subchannels and heat symmetric seven-rod bundle geometries of a Super Fast Reactor (Super FR) fuel assembly using STAR-CD. The purpose of CFD analyses in heat unsymmetric subchannels is to evaluate the effect of the power differences on the heat transfer in subchannels of the Super Fast Reactor. For heat symmetric seven-rod bundles, the effects of the gap clearance between the fuel rod and the assembly wall and the displacement of the fuel rod on the circumferential temperature distributions and Maximum Cladding Surface Temperature (MCST) are analyzed. The results show that larger power difference between fuel rods gives larger circumferential temperature difference of the hottest fuel rods. Considering cross flow between edge and ordinary subchannels, 1 mm gap between the fuel rod and the assembly wall is better for small MCST although the circumferential temperature difference in edge subchannel is large. MCST increases exponentially with the displacement. The relative error of displacement should be less than 1% if the allowable increment of MCST due to displacement is less than 6 °C.     

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.     

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.     

Numerical investigation of heat transfer in upward flows of supercritical water in circular tubes and tight fuel rod bundles

Heat transfer in upward flows of supercritical water in circular tubes and in tight fuel rod bundles is numerically investigated by using the commercial CFD code STAR-CD 3.24. The objective is to have more understandings about the phenomena happening in supercritical water and for designs of supercritical water cooled reactors. Some turbulence models are selected to carry out numerical simulations and the results are compared with experimental data and other correlations to find suitable models to predict heat transfer in supercritical water. The comparisons are not only in the low bulk temperature region, but also in the high bulk temperature region. The two-layer model (Hassid and Poreh) gives a better prediction to the heat transfer than other models, and the standard k– high Re model with the standard wall function also shows an acceptable predicting capability. Three-dimensional simulations are carried out in sub-channels of tight square lattice and triangular lattice fuel rod bundles at supercritical pressure. Results show that there is a strong non-uniformity of the circumferential distribution of the cladding surface temperature, in the square lattice bundle with a small pitch-to-diameter ratio (P/D). However, it does not occur in the triangular lattice bundle with a small P/D. It is found that this phenomenon is caused by the large non-uniformity of the flow area in the cross-section of sub-channels. Some improved designs are numerically studied and proved to be effective to avoid the large circumferential temperature gradient at the cladding surface.     

超临界水堆子通道分析

超临界水堆作为6种第4代未来堆型中唯一的水冷堆,具有一些独特的特点,受到了广泛重视。本工作以上海核工程研究设计院的常规压水堆子通道程序为基础,开发编制了适用于超临界水堆的子通道程序,并对典型带有慢化剂水棒的超临界水堆燃料组件进行了模拟计算,得到了堆芯子通道内的温度、燃料棒包壳温度、表面传热系数等参数的分布规律。此外,研究了不同超临界流体换热关系式对计算结果的影响,结果显示,各传热关系式的计算结果存在一定差异。     

超临界水冷堆堆芯子通道稳态热工分析

超临界水冷堆(SCWR)作为6种第四代未来堆型中唯一的水冷堆,冷却剂出口温度可达500℃,具有良好的经济性.本文采用改进的COBRA-IV程序对超临界水冷堆方形组件子通道进行稳态热工分析.对计算结果进行分析可知:减小慢化剂通道中给水质量流量份额和加大慢化剂通道与相邻子通道之间的热阻,可以降低热管焓升,后者还可以得到较好的慢化效果.通过热通道的传热恶化分析发现,超临界水冷堆的设计不能避免传热恶化,必须精确计算传热恶化条件下的包壳温度才能确定包壳能否保证其完整性.     

SCWR subchannel code ATHAS development and CANDU-SCWR analysis

A subchannel code (ATHAS) is developed for preliminary analyses of flow and enthalpy distributions and cladding temperatures at supercritical water conditions. The code is applicable for transient and steady state calculations. A number of heat-transfer correlations, frictional resistance correlations, and mixing models have been implemented into the code as options for sensitivity analyses. In addition, a 3D heat conduction model has been introduced to establish the cladding temperature. The results show that (1) a CANFLEX [Inch, W.W.T., Thompson, P.D., Suk, H.C., 2000. Introduction of the new fuel bundle CANFLEX into an existing CANDU reactor. In: Proceedings of the 12th Pacific Basin Nuclear Conference, October 29–November 2, Seoul, Korea.] bundle is appropriate for use in the CANDU supercritical water-cooled reactor (SCWR) based on heat transfer analysis, (2) the selection of heat transfer, friction, and mixing correlations has a significant impact on the prediction of the maximum cladding-surface temperature, and (3) the inclusion of the 3D heat conduction in the calculation has provided a more realistic prediction of the maximum cladding-surface temperature than assuming a uniform cladding temperature due to the heterogeneous characteristic of rods.     

含绕丝2×2棒束内超临界水传热试验研究

以超临界水冷堆燃料性能验证试验为背景,对带有螺旋绕丝的2×2棒束内超临界水的传热特性进行了试验研究。试验参数范围为：压力23～28 MPa,质量流速400～1 000 kg/(m²•s),壁面热流密度200～1 000 kW/m²。通过试验,获得了加热管周向壁温的分布规律,并分析了热流密度、质量流速、压力、螺旋绕丝对壁温和换热系数的影响。研究结果表明,加热管周向壁温呈现非均匀、非对称分布的特性,最高壁温出现在边角子通道或螺旋绕丝覆盖的位置。在拟临界区,换热系数随热流密度的升高或质量流速的降低而迅速减小,而随压力的变化较微弱。相对于光滑2×2棒束,螺旋绕丝不仅改变了周向壁温分布规律,同时也提高了平均换热系数。     

基于加速度效应的超临界水传热特性模型研究

超临界水冷堆（SCWR）运行在水的热力学临界点（22.1 MPa,374℃）之上,堆内冷却剂处于超临界状态,物性变化剧烈,与常规压水堆临界热流密度（CHF）导致包壳表面壁温飞升不同,超临界压力下的传热恶化是在变物性的影响下使得包壳表面温度相对缓慢上升,传统的热点判定方法和偏离泡核沸腾比（DNBR）限值等传热特性分析方法不再完全适用,因此,预测超临界水传热恶化时包壳壁温对SCWR的安全分析相当重要。本文基于边界层方程推导了超临界水传热关系式的加速度效应修正项,基于圆管实验数据,对加速度效应修正项的相关系数进行拟合获得超临界水传热特性半经验关系式,通过数据对比,该关系式在正常传热和传热恶化工况下均具有较好的适用性。本文获得的超临界水传热特性半经验关系式可为SCWR堆芯设计分析提供支持。      

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.     

数值反应堆堆芯通道级三维热工水力程序CorTAF开发及初步验证

堆芯是核动力系统的核心部件,其完整性是反应堆安全运行的重要前提。传统核反应堆堆芯热工水力分析方法无法满足未来先进核动力系统的高精度模拟需求。本文依托开源CFD平台OpenFOAM,针对压水堆堆芯棒束结构特点建立了冷却剂流动换热模型、燃料棒导热模型和耦合换热模型,开发了一套基于有限体积法的压水堆全堆芯通道级热工水力特性分析程序CorTAF。选取GE3×3、Weiss和PNL2×6燃料组件流动换热实验开展模型验证,计算结果与实验数据基本符合,表明该程序适用于棒束燃料组件内冷却剂流动换热特性预测。本工作对压水堆堆芯安全分析工具开发具有参考和借鉴意义。     

超临界水冷堆的安全分析

应用RELAP5-3D程序建立了超临界水冷堆(SCWR)的稳态模型,并在此基础上,分别对SCWR的两种瞬态和两种事故工况进行了分析。汽轮机旁路系统的存在可有效维持反应堆压力,保证反应堆安全。若SCWR失去给水,在辅助给水系统启动之前,向下流的水棒可通过热传导带走堆芯热量,并向燃料通道内提供冷却剂,缓解堆芯升温。因而,向下流的水棒体现了SCWR的安全性。主泵卡轴事故由于没有惰转,最热包壳温度值最大,因而主泵惰转可有效缓解包壳温度的升高。