AP1000波动管内空气-水液泛特性试验研究

在室温及常压条件下以空气-水为工质,对AP1000稳压器不同液位下的波动管内液泛过程及液泛特征点进行了试验现象研究、机理分析和数据分析,得到两相逆流液泛动态特性。研究结果表明:波动管竖直管部分是最容易发生液泛现象的位置;波动管内液泛特性符合Kutateladze关系式;当气体流量较小时,液相流量随着稳压器液位的增高而减小;当气体流量较大时,液相流量随着稳压器液位的增高而增大;液体完全滞止点基本与稳压器液位高度无关。     

非能动压水堆波动管内汽液逆流试验的研究

为了研究AP1000波动管中的CCFL,设立了以AP1000三代核电反应堆的波动管为原型的缩比试验台架,主要由模拟稳压器的上水箱、波动管、两相流测量装置、下水箱、供气系统、供水系统以及称重系统、摄像系统等组成,波动管采用可视化的亚克力材质,可方便对波动管内气液两相流进行观察及拍摄。试验表明,波动管内的气液两相流流动具有明显的周期性变化,在相同的液位高度下满足Wallis关系式;随着液位的升高,气体流量变化对液体流量的影响越来越明显;在不同液位下,溢流点气量不受液位高度的影响,有临界J_g~*=0.43;液位高度在较低的情况下,正向逆向过程对液泛特性基本没有影响,在液位较高的情况下,逆向过程对液泛特性有一定影响。     

基于修正AIAD模型的稳压器波动管CCFL现象数值模拟

现阶段采用的第三代核电技术广泛引入非能动自动卸压系统,提高了反应堆的安全性,但是破口事故后可能引发的气液相向流动限制现象(Counter-Current Flow Limitation,CCFL)会增加稳压器波动管自身的安全风险,因此对稳压器波动管中CCFL现象的研究非常重要。本文采用自由表面模型结合修正的AIAD(Algebraic Interfacial Area Density)模型对稳压器波动管CCFL现象进行了三维数值模拟。通过与之对应的实验现象比较,结果分析表明:所使用的模型可以正确模拟该现象下汽液两相的相间作用;并通过对气相流速和倾斜角的敏感性分析,可以得到如下结论:阻塞的推进主要受初始气相流速和稳压器波动管倾斜角的影响,在靠近管道起始点的位置主要受初始气相流速影响,远离管道起始点的位置主要受倾斜角的影响。     

稳压器波动管热分层应力及疲劳分析

稳压器波动管内流体的温度分层引起管壁温度分层,从而在管道截面产生整体弯曲应力、局部热应力以及管道系统超过预期的位移和支撑载荷.将稳压器波动管的热分层这种复杂的三维应力分析问题简化为一维和二维组合问题,利用SYSTUS程序和ROCOCO程序对秦山核电二期扩建工程稳压器波动管热分层的应力及疲劳进行了分析研究,计算了考虑热分...     

竖直向下大、小支管气相夹带起始可视化研究

已有气相夹带起始模型均是基于竖直向下的小支管夹带所得到的，用于分析大支管气相夹带时并不适用。本文基于可视化实验，对竖直向下的大、小尺寸T型管的气相夹带起始点进行对比研究。选择与主管道直径比为0.625和0.1的大、小支管，并采用有机玻璃作为实验管道材料，空气和水为实验工质。其实验现象表明：大、小支管夹带起始均有漩涡，且漩涡现象大致相同，但在大支管条件下会出现气室，夹带气体进入支管后也会在气室下翻腾，且在相同液相折算速度下，大支管夹带起始液位会高于小支管。     

垂直管内气液两相逆流极限管径效应实验

在压水堆安全性分析中,需准确预测气液逆流极限（CCFL）工况下两相流动关系。本文采用水下淹没排气的实验方法,对相同管长不同管径垂直管的CCFL特性进行可视化实验,并对垂直管CCFL关联式模型进行分析,主要结论有:①在CCFL工况下垂直管内流型为环状流动;表观气速较大时,大管径管内液膜呈搅拌状,小管径管内液膜呈波动状;随表观气速减小,均转为液面光滑的自由降膜流动;②Wallis数模型过度关联了管径变化对垂直管CCFL特性的影响;Kutateladze数和Froude-Ohnesorge数模型也不能良好关联垂直管CCFL特性的管径效应;③提出了新的CCFL无量纲参数和相应的实验关联式,由此可使垂直管CCFL特性的管径效应得以统一表征,还可以关联物性参数变化的影响。      

竖直向下大、小支管气相夹带起始可视化研究

已有气相夹带起始模型均是基于竖直向下的小支管夹带所得到的,用于分析大支管气相夹带时并不适用。本文基于可视化实验,对竖直向下的大、小尺寸T型管的气相夹带起始点进行对比研究。选择与主管道直径比为0.625和0.1的大、小支管,并采用有机玻璃作为实验管道材料,空气和水为实验工质。其实验现象表明:大、小支管夹带起始均有漩涡,且漩涡现象大致相同,但在大支管条件下会出现气室,夹带气体进入支管后也会在气室下翻腾,且在相同液相折算速度下,大支管夹带起始液位会高于小支管。     

基于分区密度补偿的稳压器液位测量研究

针对稳压器底部电热元件进行加热时，稳压器中上部和底部温度差异较大，导致传统稳压器差压法液位存在测量误差大的问题，提出了一种基于分区密度补偿的稳压器液测量方法。首先根据实际情况将稳压器分为饱和区和非饱和区，饱和区为饱和蒸汽所在区域，利用测量得到的温度对饱和蒸汽密度进行补偿；非饱和区域为介质水所在的区域，利用非饱和区域平均温度对介质水密度进行补偿。其次在稳压器饱和区和非饱和区，建立基于最小二乘法的多项式拟合模型，进行密度变量补偿，进而结合冷水段密度量进行液位计算。最后在实验装置上进行实验，并和基准液位进行比较，实验表明本文所提出的稳压器液位测量方法能够得到可靠的测量结果，因此本方法能够广泛应用于核工业等工业领域中压力容器液位测量。      

瞬态工况下压水堆稳压器波动管热分层现象数值模拟

利用计算流体动力学软件ANSYS/CFX,对秦山核电二期扩建工程2×650 MW压水堆核电站四号机组核岛厂房的稳压器波动管进行了三维全尺寸非稳态计算。建立了波动管整体和不同截面的热分层瞬态,对管内热分层流动与换热进行了研究。研究结果表明:同一截面内高温层流体和低温层流体的升温方式不同;不同截面位置的管内流动温度分布特性差别较大,但均呈现分层流体温差先增大后减小的趋势。计算结果可为后续波动管热应力分析及寿命评价提供一定基础。     

直角弯头连接的竖直管与水平管中的淹没问题

本文介绍了由直角弯头连接的竖直管与水平管内淹没与流向反转问题的实验研究，根据实验结果，分析了弯头对淹没与流向反转的影响，将所得实验结果与竖直管的实验结果进行了比较，分析了水平管长度对淹没过程的影响，以及这种管路中淹没消失滞后的问题，采用无因次量对实验数据进行了处理，给出了表达淹没开始点，淹没消失点和全部携带点的关系式。     

Flooding in the pressurizer surge line of AP600 plant and analyses of APEX data

For the passive AP600 plant, the three stages of ADS (automatic depressurization system) valves are attached to the top of pressurizer. The existence of these valves makes liquid flow into and out of the pressurizer an important part of the dynamics during a small break loss-of-coolant accident. In this paper, counter-current flow limit (CCFL) in the surge line was analyzed. Specifically, CCFL in vertical piping, in slightly inclined horizontal piping, and in horizontal and vertical elbows were compared. The CCFL in the vertical section of the surge line was found to be the most limiting section. That is, the vertical CCFL controls the pressurizer liquid drain rate. This conclusion was tested and verified by comparing the predicted vertical CCFL against the counter-current flow states in the surge line, observed in small break LOCA tests conducted at the AP600 scaled test facility (APEX).     

小型堆稳压器波动管双端破口试验研究

开展了模块化小堆稳压器波动管双端破口试验研究,获得了非能动安全系统的事故响应特性和一回路系统参数变化。试验研究结果表明,在稳压器波动管双端破口极端工况条件下,中压安注箱能在短时间内提供较大的稳定安注流量,及时补充系统水装量;高压安注系统运行过程比较复杂,安注流量与堆芯补水箱压力平衡管线内介质状态和中压安注系统运行状态密切相关,在1.7 h内呈间歇注入运行状态。在整个事故过程中,堆芯一直处于淹没状态,模块化小堆非能动安全系统能够确保稳压器波动管在双端破口极端工况条件下的堆芯安全。      

冷却剂丧失事故下稳压器波动管及人孔结构的失效机理研究

稳压器是核反应堆进行压力控制和保护的重要设备,冷却剂丧失事故（LOCA）产生的巨大冲击可能造成其关键部位的结构失效。通过多场耦合计算方法,对小破口LOCA下稳压器波动管的流动传热和结构应力、人孔结构的温度分布和密封性能进行了三维瞬态数值模拟,分析了其失效机理。结果表明：高温流体快速流入波动管形成了巨大的瞬时载荷,造成了管道短时间的强烈振动,管道中间部位变形最大,可能破坏管道支撑结构;各部位等效应力快速增大,与主管道的接管部位出现了集中应力现象,较大的应力波动会影响其寿命;人孔结构出现较大的温度分布不均匀性,密封结构下垫片的密封性能变化最大,在100 s前后其内、外侧密封面接触压力都降至设计密封比压值以下,即出现泄漏。本文根据分析结果提出了波动管和人孔结构的改进建议,可为船用核动力装置发生小破口LOCA后的事故缓解提供技术借鉴。     

Experimental investigation and CFD validation of countercurrent flow limitation (CCFL) in a large-diameter PWR hot-leg geometry

ABSTRACT

Countercurrent flow limitation (CCFL) is a phenomenon that consists of several flow patterns occurring simultaneously which produces a complex gas/liquid interface and interfacial momentum transfer, thus making it one of the most challenging two-phase flow configurations for computational fluid dynamics (CFD) validation. Numerous experimental investigations have been carried out in recent years regarding this, but most of those investigations were performed in small-diameter pipes or in non-pipe geometries (rectangular cross sections). A review of these experimental investigations has shown that the scale and geometry of the test section has a large impact upon the onset and characteristics of the CCFL. In order to provide a better understanding of this phenomenon in an actual pressurized water reactor (PWR) hot-leg geometry at a relatively large-diameter and scale, a test facility with a ～1/3.9 scale and a 190 mm inner diameter was constructed. Experiments were carried out at atmospheric pressure using water and air. High-speed recording was used to acquire high-quality images of the air/water interface. CCFL mechanisms, flow patterns, and the limits of the onset of CCFL and deflooding were experimentally identified. CFD simulations of two representative cases were carried out and assessed against experimental results. The analysis of the CFD simulations has provided insights into the improvements necessary for the accurate simulation of CCFL in large-scale geometries.     

管型对波动管热分层现象的影响分析

为分析评价压水堆核电厂稳压器波动管管型对热分层现象的影响,提出采用螺纹管来减弱热分层的措施。利用计算流体力学（CFD）分析方法,对升温、升压阶段波动管原型和改进模型的热分层现象进行数值模拟,得到两种模型不同波动流速下沿波动管轴线方向的截面最大温差分布以及流场分布。对比分析结果表明：波动管结构由光管改为螺纹管后流场紊动加强并出现涡流,冷热流体间的混合增强,与原型相比可使波动管的截面温差减小约1/3,从而有效地减弱热分层的影响。     

压水堆核电厂稳压器波动管热分层现象数值分析

为分析评价压水堆核电厂稳压器波动管热分层现象对波动管结构完整性的影响,采用计算流体力学(CFD)分析方法,对稳压器波动管热分层现象进行了数值模拟.研究了波动管内的流体流动,得到了稳压器波动管的传热特性、流体流场和温度分布,分析了稳压器波动管波动热分层现象与波动流速之间的关系.研究结果表明:波动流速在一定范围内变化时,管道最大截面温差随着波动流速的增大而增大.并且得到了不同波动流速下管道最大截面温差及其出现的位置,指出了热分层现象发生时波动管的薄弱环节.     

Stress analysis of a 900 MW pressurizer surge line including stratification effects

Following temperature monitoring programmes performed on 900 MW pressurized water reactor pressurizer surge lines, it has been reported that those lines are stratified in steady state, owing to their geometry. The highest temperature difference occurs during reactor heat-up and cool-down, reaching 110°C. Obviously, this phenomenon was not considered in nuclear steam supply system (NSSS) design transients and stress reports.Based on Electricité de France and FRAMATOME experiences, such as temperature measurements on site and mock-up, and thermal hydraulic computations, NSSS transients are updated. Stratification conditions are defined in different cross-sections of the line, using pressurizer temperature, hot leg temperature and flow rate, through the Froude number. A complete stress analysis of surge lines is performed including the updated transients and bending moment increase due to stratification. First of all different sensibility studies are carried out in order to simplify assumptions.Using a two-dimensional-one-dimensional method developed by FRAMATOME, the usage factor is then computed in different cross-sections, distinguishing upper and lower parts. In the presence of stratification, the surge line is subjected to thermal stresses following thermal shocks and to bending moment variation. These two load types are studied vs. time in order to reduce conservatism present in usual analyses.     

Transition from depressurization to long term cooling in AP600 scaled integral test facilities

A novel light water reactor design called the AP600 has been proposed by the Westinghouse Electric Corporation. In the evaluation of this plant’s behavior during a small break loss of coolant accident (LOCA), the crucial transition to low pressure, long-term cooling is marked by the injection of the gravitationally driven flow from the in-containment refueling water storage tank (IRWST). The onset of this injection is characterized by intermittency in the IRWST flow. This happens at a time when the reactor vessel reaches its minimum inventory. Therefore, it is important to understand and scale the behavior of the integral experimental test facilities during this portion of the transient. The explanation is that the periodic liquid drains and refills of the pressurizer are the reason for the intermittent behavior. The momentum balance for the surge line yields the nondimensional parameter controlling this process. Data from one of the three experimental facilities represent the phenomena well at the prototypical scale. The impact of the intermittent IRWST injection on the safe plant operation is assessed and its implications are successfully resolved. The oscillation is found to result from, in effect, excess water in the primary system and it is not of safety significance.