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.     

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.     

方形子通道内超临界流体流动传热CFD分析

国际上对超临界水冷堆进行了大量的研究,但对其堆芯内超临界流体流动传热特征的认识还十分欠缺.本研究采用CFX软件对典型超临界反应堆燃料组件子通道内的超临界热工水力特征进行了数值分析.研究了流动参数、边界条件和节径比(P/D)对子通道间交混现象和传热特性的影响.计算结果表明:燃料组件外围壁面子通道比内部子通道的湍流交混强烈;稠密栅格的湍流交混比宽栅格的湍流交混小.当P/D>1.2后,P/D比对湍流交混影响不再明显.研究还发现,在拟临界点附近区域,出现湍流交混系数的突变.     

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。     

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.     

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 analysis of thermal-hydraulic behavior of supercritical water in vertical upward/downward flow channels

Investigations on the thermal-hydraulic behavior in the SCWR fuel assembly have obtained a significant attention in the international SCWR community. However, there is still a lack of understanding of the heat transfer behavior of supercritical fluids. In this paper, the numerical analysis is carried out to study the thermal-hydraulic behaviour in vertical sub-channels cooled by supercritical water. Remarkable differences in characteristics of secondary flow are found, especially in square lattice, between the upward flow and downward flow. The turbulence mixing across sub-channel gap for downward flow is much stronger than that for upward flow in wide lattice when the bulk temperature is lower than pseudo-critical point temperature. For downward flow, heat transfer deterioration phenomenon is suppressed with respect to the case of upward flow at the same conditions.     

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.     

超临界水冷堆双排燃料组件子通道分析

研究基于Cobra-IV程序,开发了适用于超临界水冷堆燃料组件分析的子通道程序.针对超临界水冷堆慢谱双排组件,进行了稳态计算,获取了相关组件热工水力参数.在此基础上,针对单一通道进行了瞬态计算,分析了燃料棒线功率变化和冷却剂流量变化条件下,超临界水冷堆燃料组件的流动和传热的动态响应,为超临界水冷堆组件的优化设计提供了参考.     

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。      

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。     

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.     

CFD analysis of thermal-hydraulic behavior of supercritical water in sub-channels

Investigations on the thermal-hydraulic behavior in the supercritical water-cooled reactor (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 fluids. In this paper, computational fluid dynamics (CFD) analysis is carried out to study the thermal-hydraulic behavior of supercritical water flows in sub-channels of a typical SCWR fuel assembly using commercial CFD code CFX-5.6. Three types of sub-channels, e.g. regular sub-channel, wall sub-channel and corner sub-channel, are analyzed. Effects of various parameters, such as boundary conditions and pitch-to-diameter ratios, on the mixing phenomenon in sub-channels and heat transfer are investigated. The turbulent mixing in tight lattice (P/D = 1.1) is lower than that in wide lattice (P/D > 1.1), whereas, the effect of pitch-to-diameter ratio on the turbulent mixing is slight at P/D > 1.1. The amplitude of turbulent mixing in wall sub-channel is slightly higher than that in regular sub-channel and is close to that in corner sub-channel. The mixing coefficient in the sub-channel at P/D ≥ 1.2 is in the range from 0.022 to 0.028. The results also show unusual behavior of turbulent mixing in the vicinity of the pseudo-critical point, and further investigation is needed. The mass mixing due to cross flow in wall sub-channel is much stronger than that in regular sub-channel at a same pitch-to-diameter ratio. The mass mixing in wall and regular sub-channels, especially at small pitch-to-diameter ratio, brings an unfavorable feedback to the heat transfer and strengthens the non-uniformity of the circumferential distribution of heat transfer. The strong mass mixing in corner sub-channel should be paid attention.     

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.     

混合能谱超临界水堆堆芯热工-物理性能分析

针对一种新型的超临界水堆设计方案——混合能谱超临界水堆（SCWR-M）进行分析。混合能谱超临界水堆包括热谱区和快谱区两部分,分别布置在堆芯的外部与内部。它在继承了热谱与快谱超临界堆芯设计优点的同时,有效地克服了两者的不足。对于热谱区,冷却剂与慢化剂同向流动,大幅降低了燃料包壳的表面温度和组件的机械加工难度;对于快谱区,采用多层燃料组件和较大的栅距棒径比p/d,可得到较高的燃料转换比和较小的冷却剂负反应性系数。本工作采用自主开发的基于子通道分析和三维物理计算的耦合程序,对混合能谱超临界水堆的热工性能和中子物理性能（包括燃耗性能）进行研究。初步的耦合分析结果表明了混合能谱超临界水堆设计方案的可行性。     

Numerical investigation of acceleration effect on heat transfer deterioration phenomenon in supercritical water

It is important to understand the heat transfer deterioration (HTD) phenomenon for specifying cladding temperature limits in the fuel assembly design of supercritical water-cooled reactor (SCWR). In this study, a numerical investigation of heat transfer in supercritical water flowing through vertical tube with high mass flux and high heat flux is performed by using six low-Reynolds number turbulence models. The capabilities of the addressed models in predicting the observed phenomena of experimental study are shortly analyzed. Mechanisms of the effect of flow structures and fluid properties on heat transfer deterioration phenomenon are also discussed. Numerical results have shown that the turbulence is significantly suppressed when the large-property-variation region spreads to the buffer layer near the wall region, resulting in heat transfer deterioration phenomenon. The property variations of dynamic viscosity and specific heat capacity in supercritical water can impair the deterioration in heat transfer, while the decrease of thermal conductivity contributes to the deterioration.     

物理-热工耦合对超临界水堆系统特性的影响分析

物理-热工耦合是超临界水堆系统分析的关键问题之一。以日本超临界水冷热堆Super LWR的堆芯设计为例,借助Dragon编制中子截面数据库,建立双群中子扩散方程计算模块,联系同时建立的热工计算模块,得到超临界水堆的物理-热工耦合计算模型。通过对比稳态与瞬态工况下耦合前、后的热工工况,分析物理-热工耦合条件下的超临界水堆系统热工特性。结果表明：在稳态工况下,物理-热工耦合将导致内、外组件堆芯功率峰值沿轴向发生明显偏移,使得部分节点的包壳温度升高,但包壳最高温度降低;在瞬态工况下,物理-热工耦合将导致堆芯包壳最高温度的发生位置有所改变。发生给水加热丧失瞬态后,在某一时刻,外部组件的包壳最高温度将转而超过内部组件的包壳最高温度。可见,物理-热工耦合对包壳最高温度的大小和发生位置均可能产生明显影响。计算分析可为超临界水堆瞬态及安全分析提供相应理论参考。     

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.     

基于SCWR堆芯结构的子通道程序开发与应用

为能够对超临界水堆（SCWR）堆芯进行子通道分析,开发了新的子通道分析程序SABER。该程序在COBRA程序的基础上改进了网格结构和热传导模型,加入了新的边界条件和水物性模块,以适用于SCWR慢谱燃料组件的子通道分析。为评估程序的适用性,采用该程序对SCWR堆芯概念设计中的慢谱燃料组件进行子通道建模,并进行稳态计算。结果表明,该程序能够用于SCWR堆芯的子通道计算分析,并较好地解决了慢谱组件计算中慢化通道和冷却通道间的热耦合及逆向流动的模拟问题。     

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。