两相流模型在直流蒸汽发生器热工水力分析中的比较研究

根据组成气液两相流基本场方程数量所反映的流动与传热特性的不同,两相流方程分为三方程、四方程、五方程和六方程模型,结合流动压降模型、传热模型、两相相互作用模型以及流动工质的状态参数和结构材料热物性等辅助关系式,可很好地对蒸汽产生系统进行设计和研究分析。本文分析了不同数量的两相流场方程的特点和局限性,结合直管式直流蒸汽发生器实验装置,分别选取最佳估算程序中4种不同的两相流场方程计算模型进行流动传热计算分析,重点比较了强制流动的单相过冷水被加热至单相过热蒸汽过程中的压力与传热特性,从而给出不同场方程的两相流模型在分析具有较大相变过程中的差异性,验证了RELAP5程序和RETRAN-3D程序计算分析直流蒸汽发生器的能力。结果表明,RELAP5程序的六方程模型更适合模拟直流蒸汽发生器。     

螺旋管直流蒸汽发生器两相流不稳定性

研究了螺旋管直流蒸汽发生器两相流不稳定性。阐述了两相流不稳定性机理。利用线性化频域理论，建立了螺旋管直流蒸汽发生器两相流不稳定性数学模型，编制了计算程序ＨＴＯＴＳＧＩＡ，分析了入口节流圈，系统压力及不同螺旋管圈等因素对螺旋管直流蒸汽发生器两个流不稳定性的影响，给出了螺秘管直流蒸汽发生器两相流稳定区域。计算值与实验值基本一致。     

蒸汽发生器二次侧两相流传热特性数值研究

以AP1000核电站蒸汽发生器为原型,建立蒸汽发生器二次侧"平均通道"模型,利用计算流体动力学软件ANSYS CFX,基于相界面模型对蒸汽发生器二次侧两相流流动和沸腾换热过程进行研究。结果表明:数值模拟计算方法能准确模拟蒸汽发生器二次侧汽液两相流沸腾和传热过程;满负荷运行时,流体由预热区经过泡核沸腾区过渡到稳定沸腾区,含汽率和传热系数沿流动方向逐渐增大,出口含汽率与该型号蒸汽发生器设计值符合较好,平均传热系数的模拟结果和JensLottes经验关联式的预测值基本一致。     

垂直上升管内对流沸腾瞬态数值模拟及其两相流不稳定性预测

核电站蒸汽发生器二次侧为两相对流沸腾换热过程,在设计过程中须保证其不发生两相流不稳定性。本工作采用时域法对垂直上升管内两相流不稳定性进行研究,建立了垂直上升直管内流动沸腾过程的一维模型,并编制计算程序。采用该程序模拟了流动沸腾过程气液两相流密度波的不稳定性,给出两相流波动过程瞬态参数分布,由此分析了密度波不稳定发生的机理,并分析了质量流速、系统压力、入口过冷度对不稳定的影响。结果表明,与已有实验及理论结果相比,瞬态参数计算结果与实验结果符合较好,可较好找到不同工况下直管内气液两相流发生不稳定的边界,结果优于Khabenski线算图方法。     

核电站蒸汽发生器研究设计中的几个问题

本文对蒸汽发生器中两相流动阻力，传热计算进行了分析探讨，并根据我国核电站蒸汽发生器的研究现状，指出了我们研究工作中存在的问题，并提出了几点建议。     

基于节点法的轴流式预热蒸汽发生器稳态热工水力分析

基于一维流动假设、传热假设和两相热平衡假设等,采用集总参数法和分布参数法相结合,建立了轴流式预热蒸汽发生器的一维稳态热工水力分析模型。采用C++语言编程,将计算结果与某典型轴流式预热蒸汽发生器热工水力参数的设计值进行对比,结果表明大部分总体参数计算结果的相对误差都在3%以内,验证了模型的合理性;蒸汽发生器中温度、空泡份额、压力等参数沿一次测流体流动方向的变化趋势,符合热工水力学及定性机理分析结果,说明所建立的模型和求解方法能够较准确预测轴流式预热蒸汽发生器稳态热工水力参数分布。      

蒸汽发生器U型管低含气率两相倒流现象研究

针对立式倒U型管自然循环蒸汽发生器传热管内的两相倒流现象，基于均相流模型，建立了U型管内低含气率两相流动传热理论模型，给出了U型管的进出口压降-质量流量曲线，分析了U型管内出现两相倒流现象的机理，研究了二次侧流体温度和入口含气率对倒流现象的影响规律，并与单相倒流进行了对比。利用RELAP5/MOD 3.3程序对相同条件下的倒流问题进行了计算。研究表明，提高蒸汽发生器二次侧工作压力可减少倒流，两相流入口含气率越高，倒流越易发生，两相流较单相流在U型管内更易倒流。     

自然循环蒸汽发生器倒U型管内倒流现象影响因素研究

在某些自然循环工况下,蒸汽发生器部分倒U型管内存在倒流现象。基于一维Oberbeck-Boussinesq方程,建立了蒸汽发生器并联倒U型管内单相水流动传热模型,并以两种尺寸的蒸汽发生器为例进行了计算。计算结果表明,小型蒸汽发生器内短管易发生倒流,大型蒸汽发生器内长管易发生倒流;蒸汽发生器进口水温对倒流现象的发生具有重要的影响。     

并列螺旋管圈管内两相流不稳定性程序OTSGI

研究了螺旋管直流蒸汽发生器两相流不稳定性。利用线性化频域理论，采用分相模型，考虑一邓管壁动态特性，建立了螺旋管直流蒸汽发生器两相充不稳定数学模型，编制了计算ＯＴＳＧＩ，分析了人口节流圈、系统压力等因素对不稳定性的影响，给出了螺旋管直流蒸汽发生器两相不稳定区域。     

高温气冷堆螺旋管式直流蒸汽发生器热工水力学

高温气冷堆蒸汽发生器具有一次侧氦气工质、二次侧直流、螺旋管结构、工作温度高等特点,其热工水力特性与传统压水堆自然循环蒸汽发生器存在很大区别。针对高温气冷堆蒸汽发生器的特点,对其基础热工水力及特有热工水力学问题进行了阐述,主要包括螺旋管内单相及两相流阻及换热计算、横掠螺旋管束流阻及换热计算、温度均匀性及两相流不稳定性等。同时介绍了清华大学核能与新能源技术研究院针对高温气冷堆蒸汽发生器热工设计、温度均匀性及两相流不稳定性等热工水力学问题所开发的一维稳态程序、一维瞬态程序、二维分析程序和方法,并对分析结果和结论进行了讨论。相关研究方法、程序和结论对其他相似参数螺旋管和直管式直流蒸汽发生器具有参考和借鉴意义。     

Experimental and Operational Verification of the HTR-10 Once-Through Steam Generator (SG)

In January 2003, the 10MW High-temperature Gas-cooled Reactor (HTR-10) reached its full power for continuous operation of seventy-two hours in the Institute of Nuclear Energy Technology, Tsinghua University. The reactor was operated smoothly at the designated parameters. The once-through steam generator (SG) is one of key equipments of the HTR-10 reactor. The SG includes 30 modular heating helical tube assemblies. Design of the SG includes hydraulics, heat transfer and stability designs. Based on the design requirement, it is necessary to ensure sufficient heat removal from the reactor in order to maintain stable operation. In order to confirm the thermal hydraulic reliability of the SG, a series of experiments had been carried out. The purpose of this paper is to introduce the design features and experimental verification of HTR-10 SG, and the research results of small bending radius helical coil-pipe used in HTR-10, for example, the heat transfer coefficient of water, superheat steam and the two phase flow in the helical tube, the heat transfer coefficient of the He flow across the helical tube, and the centrifugal force effect on the heat transfer for the helical tube. In the paper, some operational experimental data of the HTR-10 SG have been presented.     

快堆蒸汽发生器大泄漏钠-水反应计算

采用一维特征线方法建立了快堆蒸汽发生器单管发生双端断裂情况下 ,水从破裂传热管流出的泄漏率计算模型和钠 水反应气泡从球状到柱状的变温绝热生长模型。根据断裂处氢气压力变化 ,并考虑管内流体的压缩性建立了水的泄漏率模型。对分别发生在单相水区 ,单相蒸汽区和两相汽水混合区的大泄漏钠 水反应进行瞬态计算和定性分析。结果说明 ,传热管不同位置发生泄漏对二回路造成的影响不同     

高温气冷堆螺旋管式直流蒸汽发生器热工水力学

高温气冷堆蒸汽发生器具有一次侧氦气工质、二次侧直流、螺旋管结构、工作温度高等特点，其热工水力特性与传统压水堆自然循环蒸汽发生器存在很大区别。针对高温气冷堆蒸汽发生器的特点，对其基础热工水力及特有热工水力学问题进行了阐述，主要包括螺旋管内单相及两相流阻及换热计算、横掠螺旋管束流阻及换热计算、温度均匀性及两相流不稳定性等。同时介绍了清华大学核能与新能源技术研究院针对高温气冷堆蒸汽发生器热工设计、温度均匀性及两相流不稳定性等热工水力学问题所开发的一维稳态程序、一维瞬态程序、二维分析程序和方法，并对分析结果和结论进行了讨论。相关研究方法、程序和结论对其他相似参数螺旋管和直管式直流蒸汽发生器具有参考和借鉴意义。     

M310机组蒸汽发生器传热管5%破损率量化研究

M310核电机组《化学和放射化学技术规范》第3篇《放射化学规范》中规定,当“133Xe>92500 MBq/t或133Xe>37000 MBq/t和131I/133I>1.5”时,执行“如果至少1台蒸汽发生器（SG）传热管破损率超过5%,则应以50 MW/min 速率降负荷到NS/SG模式”指令。在实际运行过程中,由于无法判断传热管破损率是否超过5%,故无法确定是否执行“以50 MW/min 速率降负荷到NS/SG模式”的指令,因此,M310核电机组《放射化学规范》中该运行指令不具备可操作性。本文对国内外相关放射化学运行指令进行调研,充分理解“传热管破损率超过5%”的含义,并进行量化分析。通过对SG传热管在5、24、44 L/h泄漏率下的裂纹稳定性进行分析,发现在这些泄漏率下单根传热管都不会发生失稳断裂,进而给出了一个具有可操作性的建议,用指标“如果至少1台SG一次侧向二次侧的泄漏率超过5 L/h”,代替无法量化的“如果至少一台SG传热管破损率超过5%”,可保证核电厂运行的安全性和经济性。      

Application of flexibility model in modeling of flow boiling heat transfer

The mathematical modeling and computer simulation have been widely used in the analysis of system's dynamic characteristics, and often useful for system control. One of the popular methods for this purpose is the lumped parameter method. For flow boiling heat transfer system, the traditional lumped parameter modeling method has a problem that the heat transfer coefficients change suddenly at the boundary of coolant phase change. It can cause error. In this paper, an idea of flexibility model is developed to deal with the boundary problem and to improve the model of flow boiling heat transfer. The segments of coolant phase change's boundary are identified, and the membership functions which are derived from Fuzzy Mathematics are used to derive approximate expressions of heat transfer coefficient in those regions. The continuity of heat transfer coefficient can be described by those expressions. The membership functions are derived from mathematical analysis and transformation. The result shows that this idea is feasible and the conclusion is practicable.     

A new LMR steam generator free from SWR with a double tube bundle configuration

This paper describes the activities made at KAERI to develop an advanced liquid metal reactor (LMR) steam generator which is free from a sodium water reaction (SWR) to resolve the concern of the SWR possibility and improve the economic features of the LMR. The steam generator design houses two tube bundles that are functionally different and its tube bundle region is radially or vertically divided into two. The SG is equipped with hot and cold fluid tube bundles, a medium fluid and a pump. It prevents the occurrence of the sodium water reaction while sodium is still used as the coolant for the primary heat transport system. The feasibility of using the SG with a double tube bundle for an actual use in a LMR plant is evaluated by setting up the skeleton of the NSSS for various possible configurations of the SG tube bundles.Analysis was made for various types of the new steam generator with a double tube bundle. Since the heat transfer in the SG is made among three different fluids, it has unique heat transfer characteristics. The analysis showed the possibility of the occurrence of an undesirable reversed heat transfer not only in the integrated single-region bundle type but also in the integrated double-region bundle type. The performance analysis revealed practical performance characteristics for the various bundle configurations. Also the feasibility study for the various NSSS configurations with the new SG is described.     

蒸汽发生器传热管束过冷沸腾区两相流动数值模拟

基于两流体欧拉数学模型结合RPI壁面沸腾模型,利用大型商用CFD软件ANSYS CFX 12.0对蒸汽发生器传热管束过冷沸腾区一次侧、壁面和二次侧耦合传热过程进行了数值模拟。研究了三叶梅花孔支撑板和不同入口过冷度条件下蒸汽发生器传热管束内的流动沸腾现象,得到一、二次侧流场与温度场,二次侧空泡份额分布,支撑板梅花孔局部的流动状况及不同入口过冷度对蒸汽发生器热工水力特性的影响。数值模拟结果表明,三叶梅花孔支撑板的存在及不同入口过冷度对蒸汽发生器传热管束过冷沸腾区域的热工水力特性影响显著。     

一体化压水堆蒸汽发生器的热工水力瞬态特性分析

一体化压水堆的设计是将蒸汽发生器及稳压器等一回路所有部件都放入压力容器内，以提高安全性，采用可以精确模拟直汉蒸汽发生器二次侧水的饱和点，蒸干点位置等重要参数随时间变化的可移动边界并分法，选用适合各中换热工况的一整套换热关系式，建立了可以模拟一体化压水堆直流蒸汽发生器的稳态及瞬态热工不特性的物理及数学模型，并编制了计算程序，经对Ｂａｂｃｏｃｋ和Ｗｉｌｃｏｘ公司１９管直流蒸汽发生器实验装置进行了计算有     

Experiment and Verification Test of the Once-through Steam Generator of the 10 MW High-temperature Gas-cooled Reactor Flow Stability of the Once-through Steam Generator

In January 2003, the 10MW High-temperature Gas-cooled Reactor (HTR-10) reached its full power for continuous operation of seventy-two hours in the Institute of Nuclear Energy Technology, Tsinghua University. The reactor operated smoothlyqbthe design parameters were successfully attained.

The once-through steam generator (SG) is one of key equipments of the HTR-10 reactor. The SG includes 30 modular heating helical tube assemblies. There are two thermal hydraulic requirements to be satisfied for the once-through steam generator: (1) enough heat transfer surface; (2) qualified steam can be produced under rated electrical generation power, and water-steam two phase flow un-stability can be avoided. In order to obtain the thermal hydraulic characteristics of the SG reliably, before design, a numerical code was developed for the design, and a full-scale test loop with two heating tubes as model was established, and series experiments had been carried out.

The purpose of this paper is to introduce the design of SG and researches on the stability of small bending radius helical coil-pipe used in HTR-10, for exempla, the effects of outlet steam pressure, inlet water sub-cooling degree, thermal power and inlet throttling degree. Up to now, the SG has experienced full power operation smoothly, and approvingly reached its original design requirements.

In the paper, some operational experimental data of the HTR-10 S.G have been presented.     

适于窄缝流动沸腾传热的关系式

由于受窄缝间隙几何尺寸的限制，在加热的窄缝流道中沸腾气泡随着其长大和流动将受挤而变形，聚合，其表面张力的作用将大幅度改变，本文通过引入Bond无量纲数来考察窄缝流道的窄缝几何效应和气泡表面张力作用，修正了Chen沸腾传热关系式，并提出了相应的窄缝流动沸腾传热计算的关系式，通过与实验数据的验证比较，修正Chen沸腾传热关系式的计算结果与实验结果比较吻合，因此，该关系式能够用来预测在单面加热窄缝 道中的两相沸腾传热系数。