AP1000稳压器波动管竖直管段空气-水CCFL实验研究

使用竖直管代替波动管模型开展稳压器波动管竖直管段内空气-水两相逆流限制（CCFL）特性可视化实验研究。实验现象表明:竖直管与上容器接口处的局部CCFL决定了进入竖直管内的液相流量;竖直管内的局部CCFL决定了从竖直管流出的液相流量;两处局部CCFL均随空气流量的增大而增强。在较低气量情况,进入竖直管内的液相能够完全或大部分流出,竖直管内的局部CCFL较弱,上容器和竖直管接口处的局部CCFL在整体CCFL中占主导地位,整体CCFL程度随着上容器液位升高而略有增强。在高气量情况,从上容器进入竖直管的液相大部分或者完全被限制而不能向下流出,竖直管内的局部CCFL强烈,在整体CCFL中占主导地位,整体CCFL特性不受上容器液位变化的影响。通过实验数据拟合得到了新的稳压器竖直管CCFL模型。稳压器波动管CCFL数据和稳压器竖直管CCFL数据基本重合,表明波动管CCFL主要由CCFL-U决定。      

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。      

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.     

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。     

Scaling in the safety of next generation reactors

A technique was developed to evaluate the applicability of data from small scale facilities for validation of codes for analysis of nuclear safety with emphasis on the next generation of reactors. The technique first divides an accident into phases based on the components that come into play as the accident evolves. Conservation equations, resolved to the component level and their interconnections, are derived for the active components in each phase. The equations are then nondimensionalized and reference parameters are selected such that the dependent variables, other than the system response of interest, are of order 1. Order of magnitude analysis is then performed for each equation and then between equations, based on the numerical values of the nondimensional coefficients for each term, with only the large order terms being retained. The resulting equations then contain terms whose impact on key system responses (e.g. reactor vessel level) are ordered in terms of the magnitude of the nondimensional groups multiplying the O[1] dependent variables. The reduced set of equations and nondimensional groups are validated with experimental data where possible. The validation process is meant to demonstrate that the important terms have been retained and enhance confidence in the system of equations used to capture the main processes occurring in each phase. The methodology was demonstrated by evaluating the applicability of small-scale facility data for next generation reactor SBLOCA. Based on the nondimensional equations, the dominant nondimensional groups, and hence the dominant physical mechanisms and their dependence on geometric and operational parameters, were identified for a particular scenario, an AP600 cold leg break, starting from the initiating event through long term cooling. The important parameters entering the groups included elevation differences between the reactor vessel and other components, PRHR heat transfer rates, fluid thermophysical properties, liquid levels in tanks, flow resistances in the CMT lines and IRWST lines, flow resistance in the pressurizer surge line, and pressurizer drain rate. It was also shown that, after the beginning of CMT draining and accumulator injection, the dominant processes do not depend on break size provided they are small. The dominant processes were dependent on plant geometry and the operation of engineered safety features, such as the automatic depressurization system. The same transient events were evaluated for three experimental facilities and the same nondimensional groups, and hence mechanisms, were shown to be important. It was found that these nondimensional groups covered the range expected in the AP600, indicating that while there may be some distortions in scaling for a particular facility, between them, the important phenomena were captured and the small-scale facility data appear applicable for SBLOCA in the AP600 system. In more general terms, the methodology appears suitable for assessing scaling of various facilities for other postulated accidents and for other reactor concepts.     

CFD studies on the phenomena around counter-current flow limitations of gas/liquid two-phase flow in a model of a PWR hot leg

In order to improve the understanding of counter-current two-phase flow and to validate new physical models, CFD simulations of a 1/3rd scale model of the hot leg of a German Konvoi pressurized water reactor (PWR) with rectangular cross section were performed. Selected counter-current flow limitation (CCFL) experiments conducted at Helmholtz-Zentrum Dresden-Rossendorf (HZDR) were calculated with ANSYS CFX using the multi-fluid Euler–Euler modelling approach. The transient calculations were carried out using a gas/liquid inhomogeneous multiphase flow model coupled with a shear stress transport (SST) turbulence model.In the simulation, the drag law was approached by a newly developed correlation of the drag coefficient (Höhne and Vallée, 2010) in the Algebraic Interfacial Area Density (AIAD) model. The model can distinguish the bubbles, droplets and the free surface using the local liquid phase volume fraction value. A comparison with the high-speed video observations shows a good qualitative agreement. The results indicate also a quantitative agreement between calculations and experimental data for the CCFL characteristics and the water level inside the hot leg channel.     

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.     

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。      

Experimental study on the air/water counter-current flow limitation in a model of the hot leg of a pressurized water reactor

An experimental investigation on the air/water counter-current two-phase flow in a horizontal rectangular channel connected to an inclined riser has been conducted. This test-section representing a model of the hot leg of a pressurized water reactor is mounted between two separators in a pressurized experimental vessel. The cross-section and length of the horizontal part of the test-section are (0.25 m × 0.05 m) and 2.59 m, respectively, whereas the inclination angle of the riser is 50°. The flow was captured by a high-speed camera in the bended region of the hot leg, delivering a detailed view of the stratified interface as well as of dispersed structures like bubbles and droplets. Countercurrent flow limitation (CCFL), or the onset of flooding, was found by analyzing the water levels measured in the separators. The counter-current flow limitation is defined as the maximum air mass flow rate at which the discharged water mass flow rate is equal to the inlet water mass flow rate.From the high-speed observations it was found that the initiation of flooding coincides with the formation of slug flow. Furthermore, a hysteresis was noticed between flooding and deflooding. The CCFL data was compared with similar experiments and empirical correlations available in the literature. Therefore, the Wallis-parameter was calculated for the rectangular cross-sections by using the channel height as length, instead of the diameter. The agreement of the CCFL curve is good, but the zero liquid penetration was found at lower values of the Wallis parameter than in most of the previous work. This deviation can be attributed to the special rectangular geometry of the hot leg model of FZD, since the other investigations were done for pipes.     

AP1000核电厂反应堆冷却剂系统布置设计

AP1000核电厂反应堆冷却剂系统布置设计,在满足系统功能的前提下,充分考虑了屏蔽防护、核级部件在役检查、模块化设计、内部灾害防护等方面的要求。反应堆冷却剂系统主设备及主回路采用了紧凑型的布置方式,改善了环路配置的经济性,波动管布置在考虑足够柔性的基础上采用了大倾斜角连续上坡的方式,降低了波动管在运行过程中出现热分层的可能性,稳压器安全阀及ADS第1、2、3级集中布置在稳压器顶部,组合成一体化的模块Q601,改善了反应堆冷却剂系统布置结构。     

压水堆稳压器波动管热分层的分析研究

热分层是管道水平管段中相对滞止或缓慢流动的冷、热流体因缺少混合而产生的不均匀温度分布现象.通过稳压器波动管热分层现象产生的原因和机理分析,并对稳压器波动管热分层现象进行数值模拟,建立了不同稳压器内部不同截面的热分层瞬态.     

基于运行数据的船用堆波动管热分层现象研究

基于运行数据将船用堆波动管热分层划分为升功率、降功率、变工况、小喷淋流量4类典型瞬态,对4类典型瞬态分别进行无量纲里查德森数（Ri）分析、瞬态工况数值模拟计算,得到波动管在4类典型瞬态下水平管段的热分层区间长度、持续时间和最大温差。结果表明,升功率和降功率瞬态热分层仅单次贯穿波动管,升功率瞬态的接头部位循环的热波动以及小喷淋流量瞬态水平段的长区间、长时间、大温差的热分层现象和变工况导致的热应力波动可能影响到波动管的安全。本文提出的基于运行数据的波动管热分层现象研究方法为后续热应力和热疲劳分析奠定了基础,同时可以为其他容积设备热分层研究提供参考。      

Analysis of thermal stratification in the primary circuit of a VVER-440 reactor with the CFX code

The thermal stratification can lead an important role in the aging of the NPP piping because of the stresses caused by the temperature differences and the cyclic temperature changes. These stresses can limit the lifetime of the piping, or lead to penetrating cracks. For the stress analyses, the determination of the thermal hydraulic parameters of the stratified flow is necessary, which can be simulated by computational fluid dynamics (CFD) codes. The results of the simulation show the time development and the breaking up of the stratification and the temperature distribution of the stratified flow. The main difficulty of these CFD simulations is the uncertainty of the boundary conditions because of the unknown flow circumstances. In this paper, some results of CFX simulations are presented concerning the pressurizer surge line, and the injection pipe of the HPIS for VVER-440 type reactors.     

Erratum to “Experimental study on the air/water counter-current flow limitation in a model of the hot leg of a pressurized water reactor”

This paper replaces the paper published in the journal by Deendarlianto et al. (2008). Because of an error in the implementation of the air flow meter some of the data given by Deendarlianto et al. (2008) are wrong. They are corrected within the present paper. The general results and conclusions remain unchanged.An experimental investigation on the air/water counter-current two-phase flow in a horizontal rectangular channel connected to an inclined riser has been conducted. This test-section representing a model of the hot leg of a pressurized water reactor is mounted between two separators in a pressurized experimental vessel. The cross-section and length of the horizontal part of the test-section are (0.25 m × 0.05 m) and 2.59 m, respectively, whereas the inclination angle of the riser is 50°. The flow was captured by a high speed camera in the bended region of the hot leg, delivering a detailed view of the stratified interface as well as of dispersed structures like bubbles and droplets. Counter-current flow limitation (CCFL), or the onset of flooding, was found by analyzing the water levels measured in the separators. The counter-current flow limitation is defined as the maximum air mass flow rate at which the discharged water mass flow rate is equal to the inlet water mass flow rate.From the high-speed observations it was found that the initiation of flooding coincides with the formation of slug flow. Furthermore, a slight hysteresis was noticed between flooding and deflooding. The CCFL data was compared with similar experiments and empirical correlations available in the literature. Therefore, the Wallis-parameter was calculated for the rectangular cross-sections by using the channel height as length, instead of the diameter. The agreement of the CCFL curve is good, but the zero liquid penetration was found at lower values of the Wallis parameter than in most of the previous work. This deviation can be attributed to the special rectangular geometry of the hot leg model of HZDR, since the other investigations were done for pipes.     

基于RELAP5的LOCA喷放阶段下降段内CCFL特性研究

反应堆失水事故（LOCA）后下降段通道内形成的两相逆流状态极有可能引发汽-液逆向流动限制（CCFL）,不利于应急冷却水顺利进入堆芯,极大影响了核反应堆系统的安全性能。本研究基于RELAP5程序采用Wallis溢流关系式对UPFT实验装置进行建模并计算LOCA喷放阶段的下降段注水行为;通过对比下腔室蓄水量、下降段内压力及破口处蒸汽流量瞬态变化以验证模型的有效性,并对下降段通道内汽相速度场、液相体积分数分布特性进行分析。结果表明,由于下降段通道结构的三维特征引起的流动不均匀性影响了汽-液CCFL特性,随着蒸汽流量增大,在破口环路与下降段连接区域的压力梯度与向上流速度梯度越大,较少节点的划分方法很难真实反映下降段通道局部区域内汽-液溢流关系;在靠近破口的环路内注入的冷却水更难到达下腔室,而在远离破口环路的冷却水容易进入到下腔室;过热的蒸汽在流动过程中被冷却水冷却发生凝结现象,导致出口蒸汽流量小于进口蒸汽流量,且随着进口蒸汽流量的增大,凝结效应则随之减小。本研究所建立的模型与方法能够适用于LOCA喷放阶段下降段通道内的汽-液CCFL预测。      

对NOTRUMP-AP600程序缺少动量通量项的评估与修正

由于NOTRUMP-AP600程序的动量守恒方程缺少动量通量项,在分析用于模拟AP600核电厂的APEX试验台架小破口事故时,安全壳内置换料水箱注射流量和稳压器混合水位等参数的预测值和实验值有较大偏离。本文对此进行评估：1) 采用均相流和分相流模型计算动量通量项对AP600核电厂自动卸压系统（ADS）管路压降的影响;2) 采用FLOAD4程序对需修正的第4级ADS(ADS4)管路的两相流压降进行计算,预测ADS4管路内的压力分布,并用作修正NOTRUMP-AP600程序ADS4管路压降的基准。结果表明,对于AP600核电厂ADS4管路,输入阻力系数需增加60%。     

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。     

模块式小型核反应堆稳压器波动管破口事故分析研究

针对模块式小型核反应堆（SMR）稳压器波动管破口事故建立了MELCOR计算模型,采用该模型对波动管破口触发的严重事故进程进行了模拟;并对其相关的热工水力参数进行分析研究,同时对比分析了不同破口面积对事故进程和结果的影响。分析结果表明:波动管破口尺寸为0.002 m2时,事故进程最为严重,该结果可为SMR的严重事故管理导则提供参考依据。      

Two-phase countercurrent flow in a model of a pressurized water reactor hot leg

The onset of flooding or countercurrent flow limitation (CCFL) determines the maximum rate at which one phase can flow countercurrently to another phase. In the present study, the experimental data of the CCFL for gas and liquid in a horizontal pipe with a bend are investigated. The different mechanisms that lead to flooding and that are dependent on the liquid flow rate are observed. For low and intermediate liquid flow rates, the onset of flooding appears simultaneously with the slugging of unstable waves that are formed at the crest of the hydraulic jump. At low liquid flow rates, slugging appears close to the bend; at higher liquid flow rates, it appears far away from the bend, in the horizontal section. For high liquid flow rates, no hydraulic jump is observed, and flooding occurs as a result of slug formation at the end of the horizontal pipe. The effects of the inclination angle of the bends, the liquid inlet conditions and the length of the horizontal pipes are of significance for the onset of flooding. A mathematical model of Ardron and Banerjee is modified to predict the onset of flooding. Flooding curves calculated by this model are compared with present experimental data and those of other researchers. The predictions of the onset of flooding as a function of the length-to-diameter ratio are in reasonable agreement with the experimental data.     

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。