压水堆蒸汽发生器水滴重力分离的理论研究(英文)

压水堆蒸汽发生器水滴重力分离空间是汽水分离装置的重要组成部分,是连接汽水分离器和蒸汽干燥器的纽带。蒸汽发生器的设计,除了要选用分离效率高,结构紧凑的汽水分离器和干燥器外,还要合理选择重力分离空间高度。高度太低,湿分得不到充分的重力分离,将会影响干燥效果;高度太高,将会增加蒸汽发生器和核岛设施的投资。文中对重力分离空间蒸汽携带水滴的过程进行了理论研究,推导出了水滴重力分离高度,水滴的飞升直径和飞升速度的一般表达式,在分析中还考虑了汽液两相间速度的不一致(即滑动比S),并提出了直径-粘度组合项。理论研究得出的结论可为确定蒸汽发生器重力分离空间高度提供依据。     

压水堆蒸汽发生器水滴重力分离的理论研究

压水堆蒸汽发生器水滴重力分离空间是汽水分离装置的重要组成部分,是连接汽水分离器和蒸汽干燥器的纽带。蒸汽发生器的设计,除了要选用分离效率高,结构紧凑的汽水分离器和干燥器外,还要合理选择重力分离空间高度。高度太低,湿分得不到充分的重力分离,将会影响干燥效果;高度太高,将会增加蒸汽发生器和核岛设施的投资。文中对重     

重力场液滴旋转问题探究

对重力分离空间蒸汽流场单液滴旋转运动问题进行探究。根据液滴重力分离机理,采用考虑转动的单液滴模型,求解及分析了液滴在蒸汽流中受力、力矩及运动问题。描述了蒸汽流场及液滴旋转、液滴直径和速度等参数对于液滴轨迹的影响,为单液滴模型建立奠定了基础。     

液滴旋转模型在重力场和均匀流场中的应用

本文主要应用单液滴旋转模型,研究了重力分离空间均匀蒸汽流场中液滴行为,定性描述了均匀蒸汽流场中液滴轨迹,分析了液滴在重力场和均匀蒸汽流场中运动的影响因素。主要讨论了液滴旋转及升力的影响,为液滴旋转模型的定量分析及其应用奠定了基础。     

重力分离空间均匀流中液滴行为问题研究

本文针对汽水分离器均匀流场中单个液滴的运动机理进行研究,分析了液滴在蒸汽流中所受的力;建立了单个液滴运动模型,并进行了数值求解;分析了其主要参数对液滴运动轨迹的影响,包括蒸汽流速、液滴直径和初速度等。定性描述了蒸汽流场中液滴重力分离之机理,为单液滴动力学模型定量分析奠定了基础。     

水滴与液态金属锡相互作用实验研究

相对低温的水喷射进入高温液态金属池中与液态金属锡相互接触产生剧烈沸腾和蒸发,并可能导致蒸汽爆炸从而引发安全问题。为了研究不同热工参数对水滴与液态金属锡相互作用的影响,本研究采用高速摄像仪对水滴落入高温熔融锡中相互作用过程进行了可视化研究。首先,对水滴在金属锡表面的蒸发时间进行了研究,探讨了不同金属温度条件下水滴与金属锡表面传热机理。其次,引入无量纲直径参数来定量水滴与熔融金属的反应剧烈程度,研究了水滴下落速度和金属锡温度对水滴与液态金属锡相互作用激烈程度的影响。结果表明:最小膜态沸腾温度大约为210℃左右。当金属超过270℃时,水滴与金属会发生激烈反应,随着金属温度和We数的增加,水滴与熔融金属锡蒸汽爆炸激烈程度增加。     

单钩波形板分离器内二次携带机理分析

通过理论分析与三维数值模拟研究了实际运行工况下单钩波形板分离器内二次携带机理。采用Realizable k-ε湍流模型对波形板内气相流场进行数值模拟,利用离散相模型结合涡交互模型对水滴运动进行计算,根据壁面水膜运动方程求解水膜的速度与厚度分布;依据理论分析,对可能形成二次水滴的4种方式进行判定。结果表明：波形板内发生气体动力造成的水滴破碎与水滴撞击水膜产生飞溅的可能性较低,但较冷态工况的可能性高;水膜主要集中在波形板的前两级,随入口蒸汽速度或湿度的增加,水膜增厚并向下游移动;将水膜剥落和水膜分离的判别式进行统一,并证实波形板二次携带主要由水膜的剥落和分离造成,且相较水膜剥落,钩峰处的水膜分离更易发生。     

秦山核电厂蒸汽发生器汽水分离装置的研制

本文介绍了为秦山核电厂蒸汽发生器汽水分离装置所进行的空气-水冷态选型试验和蒸汽-水热态考核试验及该汽水分离装置的结构。综合性能试验表明干燥后的蒸汽湿度远低于0.25％的规定值。     

立式蒸汽发生器汽水分离装置的试验研究

本文介绍了立式蒸汽发生器汽水分离装置所进行的空气-水冷态选型试验。着重介绍了在高温高压水试验台上进行的立式蒸汽发生器汽水分离装置的热态考核试验。综合性能试验表明,经干燥后的蒸汽湿度远低于0.25%的规定值。     

重力注水过程流动不稳定现象关键影响因素研究

重力驱动注水过程中由于流量较小以及蒸汽的积聚可能导致流动不稳定现象的发生,对核反应堆安全运行具有重要的影响。通过实验研究的方法,搭建了重力注水模拟实验装置,研究了不同蒸汽出口形阻、高位储水箱水位和加热棒初始温度下流动不稳定现象的变化规律。结果表明,重力驱动注水过程流动不稳定现象包含冷却水初次注入阶段、注入水逐出阶段和冷却水再注入阶段等。在一定冷却水初始温度、冷却水入口形阻以及系统压力下,蒸汽排出速度以及实验本体内筒顶部的聚集情况取决于蒸汽出口形阻,减小蒸汽出口形阻可加快蒸汽排放速度,压力峰峰值降低、振荡周期变长,有利于系统稳定;提高高位储水箱水位加快了冷却水注入速率,增加了加热棒被淹没率,降低了流动不稳定现象的发生次数和持续时间;随加热棒初始温度的升高,冷却水流量出现了波动向停滞的转变,流动不稳定现象发生的次数增加且持续时间加长。     

均相流蒸汽发生器瞬态分析模型

本文介绍了压水堆核电厂蒸汽发生器的一维均相流瞬态分析模型。基于该模型开发的程序计算结果与法国ＢＵＧＥＹ４蒸汽发生器上的试验结果及ＡＴＨＯＳ程序的计算结果较好符合，该模型可用于分析压堆核电厂Ｕ型管自然循环式蒸汽发生器的热工水力瞬态过程。     

Particle deposition and fouling in PWR steam generators

Commercial PWR steam generators have experienced reliability problems within the first decade of operation associated with material degradation, one of the causes of which is particle deposition and tube fouling. As a result steam generators often require costly outages for inspection and cleaning of fouling deposits. Knowledge of locations where sludge has accumulated in the steam generator can aid in planning and targeting locations for cleaning and removal of deposits. A particulate deposition model has been developed and implemented in the three dimensional thermal hydraulics computer code, ATHOS3 to calculate sludge and fouling regions within the steam generators during operation. This transient particle deposition model uses the thermal hydraulic field calculated by the ATHOS3 code, and the concentration of magnetite particles entering the steam generator to calculate the particle distributions and deposition on vertical and horizontal surfaces within the steam generator. Results of some simulations of operating steam generator designs are presented in this paper. These results show that preferred regions for deposition include hot side upper bundle and a kidney shaped region on top of the tube sheet.     

PACTEL integral test facility – Description of versatile applications

Regardless of the large number of thermal hydraulic experiments conducted with many different facilities, the need for good quality data from integral test facilities has not yet reached saturation. The parallel channel test loop (PACTEL) facility is one of the largest facilities of its kind. It was originally designed to model the thermal-hydraulic behavior of VVER-440 type pressurized water reactors (PWRs) currently used in Finland. Nevertheless, the PACTEL facility has served also in many other purposes than for VVERs only. The facility has been modified on a case-by-case basis according to the needs in configuration and positioning of auxiliary equipment. The newest plan is to modify PACTEL to be the PWR PACTEL, a facility with vertical steam generators for EPR applications topical in Finland. This paper describes the versatile use of the PACTEL facility for a large spectrum of thermal hydraulic research. The PACTEL facility is ideal for investigating planned recovery procedures during accidents and operational transients. For this purpose experimental series among others on small break loss-of-coolant accidents (SBLOCA), primary-to-secondary leakages (PRISE), and on anticipated transients without scram (ATWS) have been carried out. The PACTEL natural circulation experiment with stepwise coolant inventory reduction formed the basis for the OECD International Standard Problem (ISP-33). In addition, many other one-phase and two-phase natural circulation tests have been executed.     

Steam drum level dynamics in a multiple loop natural circulation system of a pressure-tube type boiling water reactor

Advanced Heavy Water Reactor (AHWR) is a pressure tube type boiling water reactor employing natural circulation as the mode of heat removal under all the operating conditions. Main heat transport system (MHTS) of AHWR is essentially a multi-loop natural circulation system with all the loops connected to each other. Each loop of MHTS has a steam drum that provides for gravity based steam–water separation. Steam drum level is a very critical parameter especially in multi-loop natural circulation systems as large departures from the set point may lead to ineffective separation of steam–water or may affect the driving head. However, such a system is susceptible to steam drum level anomalies under postulated asymmetrical operating conditions among the different quadrants of the core like feedwater flow distribution anomaly among the steam drums or power anomaly among the core quadrants. Analyses were carried out to probe such scenarios and unravel the underlying dynamics of steam drum level using system code RELAP5/Mod3.2. In addition, a scheme to obviate such problem in a passive manner without dependence on level controller was examined. It was concluded that steam drums need to be connected in the liquid as well as steam space to make the system tolerant to asymmetrical operating conditions.     

立式自然循环蒸汽发生器的重量优化设计

蒸汽发生器（SG）是压水堆核动力装置中重量及尺寸较大的关键设备之一,且随着SG大容量、大型化的发展趋势使其重量和体积进一步增加,这给设备的运输和布置带来了困难,因此,有必要在SG设计中,在满足换热性能及安全要求的条件下,寻找设计参数的最佳组合,以减小SG重量或体积。本文建立了SG重量评价程序,并在对影响SG重量的热工参数和结构参数进行敏感性分析的基础上,采用复合形 遗传优化算法,以SG重量为目标进行了参数优化设计。优化结果显示：与原始方案相比,优化设计方案的重量减小了17.16%,优化效果显著,并从理论上证明了优化的可能性,研究结果可为实际工程设计提供参考。     

坎杜堆蒸汽发生器设计综述

国际核电站的运行经验表明，坎杜堆蒸汽发生器具有较高的可靠性，坎杜堆蒸汽发生器的主要型号有：皮克灵Ａ，皮克灵Ｂ，布鲁斯Ａ，布鲁斯Ｂ，根蒂莱２和达灵顿核电站的蒸汽发生器，根据坎杜堆多年的运行经验，其蒸汽发生器设计经历了多次改进，改进后的坎杜型蒸汽发生器使用了能够抵抗各种腐蚀的管材，具有高的循环倍率，柔性支撑板，高效汽水分离装置和方便的维修特性。     

Steam generator tube rupture (SGTR) scenarios

The steam generator tube rupture (SGTR) scenarios project was carried out in the EU 5th framework programme in the field of nuclear safety during years 2000–2002. The first objective of the project was to generate a comprehensive database on fission product retention in a steam generator. The second objective was to verify and develop predictive models to support accident management interventions in steam generator tube rupture sequences, which either directly lead to severe accident conditions or are induced by other sequences leading to severe accidents. The models developed for fission product retention were to be included in severe accident codes. In addition, it was shown that existing models for turbulent deposition, which is the dominating deposition mechanism in dry conditions and at high flow rates, contain large uncertainties. The results of the project are applicable to various pressurised water reactors, including vertical steam generators (western PWR) and horizontal steam generators (VVER).     

900MWe压水堆蒸汽发生器部分给水管线破口情况下的给水流量分析

在900 MWe压水堆中,蒸汽发生器的给水是非常重要的,特别是在1根给水管线破口的情况下,给水泵必须通过另外2根未破损的给水管线向蒸汽发生器提供足够的流量.本工作对上述情况下未受影响的蒸汽发生器的给水流量进行了分析,并阐述了对其进行测量、计算及误差处理的主要原理和方法.