反应堆压力容器直接安注热分布特性研究

我国非能动系列压水堆将应急冷却系统冷却水的注入管道直接连接于压力容器上,与传统的冷管段安注不同,这种安注方式被称之为反应堆压力容器直接安注。本文以安注条件下的反应堆压力容器为研究对象,采用物理实验与数值分析结合的方法,对安注流体在压力容器表面形成的热分布形态进行研究。研究发现,不同于传统的主管道冷段斜接管安注方式,直接安注条件下安注流体在下降环腔中的分布形态接近于等腰三角形。以实验结果为基础,结合数值计算验证,发现了压力容器热分布角与流速比成正比关系,并进一步提出了安注流体分布计算模型,从而为反应堆安全设计提供参考。      

堆芯紧急冷却安注热混合实验研究

以核电厂压水堆中失水事故(LOCA)堆芯紧急安注系统(ECCS)启动后安注接管与冷管段的T型管处冷、热流体混合为研究对象,进行安注管和主管道内过冷水-高温冷却剂的热混合特性实验以及过冷水-汽水混合物直接接触冷凝特性实验,通过缩比尺寸实验对热混合相关现象进行研究。结果表明,单相热混合实验管内温度场随不同射流流型成一定分布;两相热混合工况安注后冷凝量随主管蒸汽量变化而成线性分布,并总结实验数据形成适用于本实验直接接触冷凝相关关系式。     

PTS瞬态流动与换热数值模拟研究

为了研究压水堆因“直接安注”冷水注入压力容器下降环腔而导致的承压热冲击（PTS）热工水力问题,基于1:10比例模型,应用计算流体力学软件FLUENT5.4进行了紊流流动换热的数值模拟分析,同时进行了常压瞬态传热实验研究。针对下降环腔折算流速0.5 m/s,安注流速10m/s的典型工况,研究了安注水开启后下降环腔内的瞬态流动换热特性,数值模拟与实验结果吻合良好。考察了压力容器安注接管出口区环形焊缝区及堆芯段筒体中子强辐照区所承受的热冲击状况,基于稳态流动研究了下降环腔内流体混合特性及流动机理,为热冲击分析提供参考。     

T型管热混合温度波动特性研究

在反应堆发生失水事故(LOCA)时,一回路系统压力降低,产生大量蒸汽,堆芯应急冷却系统(ECCS)启动后,安注水注入冷腿后在T型管处与蒸汽发生热混合,温度会出现明显波动,同时伴随有一定的回流。本文以T型管中冷热流体混合为研究对象,开展了安注过冷水与冷腿中的饱和蒸汽热混合实验。研究内容主要为过冷水与饱和蒸汽在水平T型管发生热混合之后的水跃和回流现象,基于动量分析的方法,分析了不同流型对热混合后温度分布的影响,提出了两相流动量比关系式用于分析T型管内温度波动特性。     

承压热冲击下压水堆压力容器壁面换热特性的数值模拟研究

为了研究压水堆因安注冷水直接注入反应堆压力容器下降环腔而导致的承压热冲击（PTS）热工水力问题，基于1∶10比例模型，应用计算流体力学商用软件FLUENT5．4进行了紊流流动换热的数值模拟分析，同时进行了常压传热实验研究。针对下降环腔折算流速0．5m/s，安注流速10m/s的典型工况，研究了压力容器下降环腔的壁面换热特性。通过分析下降环腔内的流动及混合特性，从流动机理上解释了压力容器内壁上准重接触点附近换热强烈的现象，并指出壁面换热强弱与近壁流体紊流脉动动能密切相关，为热冲击分析提供参考。     

承压热冲击下AP1000压力容器直接安注瞬态数值模拟研究

基于计算流体动力学(CFD)分析方法,采用流固共轭传热方式,对非能动堆芯冷却系统(PXS)的堆芯补水箱(CMT)热态功能试验、CMT注入同时自动减压系统(ADS)动作、蓄压安注箱(ACC)安注后CMT再注入以及常规余热排出系统运行等4种工况下反应堆压力容器(RPV)环腔内流动传热状态进行瞬态数值模拟,研究RPV壁面温度瞬态变化以及环腔下降段内流体的混合特性。结果表明:4种工况下直接安注(DVI)接管管嘴与RPV内壁面相交斜面处冷却水混合剧烈,冷段是否有流体注入环腔对其内流体温度分布变化影响巨大,且DVI接管管嘴局部区域将发生较大的温度变化。     

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.     

冷热流体混合特性的数值研究

应用三维热工水力程序AQUA对冷热流体混合时发生的热现象进行了数值分析，并用简化基本议程归一化方法对影响该热现象流动参数分布的因素进行了讨论。结果表明，流场与温度场计算结果与归一化分析相符，温度波动计算结果与自然规律一致。这表明AQUA程序能较好地模拟冷热流体混合时发生的热现象。     

高压安注管道热疲劳现象及改进措施分析

根据高压安注死管段内的热工水力特性,分析管道热疲劳发生的机理。提出大亚湾核电站和岭澳核电站一期高压安注死管段的改进方案,并对改进方案中的新增管线进行疲劳分析后可得,大亚湾核电站和岭澳核电站一期在改进中新增的管线,不会因机组各种运行工况引入的载荷而发生疲劳失效。压力井改进方案是有效的、安全的。     

承压热冲击下反应堆压力容器中流体的瞬态混合特性

热冲击下,反应堆压力容器中的热工水力特性是一个与反应堆安全密切相关的课题。本文在1/10的模型体上进行了高温高压下安全注水时流体的瞬态混合特性实验,得到了在有回路流动和无回路流动时以及不同的环腔流体温度下的混合特征。结果表明,环腔无流动时,随着安全注水流速的提高,混合函数下降得更快,幅度更大;环路有流动时,混合函数变化缓慢：当环腔内的流体温度达到一定的数值后,压力容器部分区域的混合函数发生明显变化。     

Boron dilution transients during natural circulation flow in PWR—Experiments and CFD simulations

Coolant mixing in the cold leg, downcomer and the lower plenum of pressurized water reactors is an important phenomenon mitigating the reactivity insertion into the core. Therefore, mixing of the de-borated slugs with the ambient coolant in the reactor pressure vessel was investigated at the four loops 1:5 scaled Rossendorf coolant mixing model (ROCOM) mixing test facility. In particular thermal hydraulics analyses have shown, that weakly borated condensate can accumulate in the pump loop seal of those loops, which do not receive a safety injection. After refilling of the primary circuit, natural circulation in the stagnant loops can re-establish simultaneously and the de-borated slugs are shifted towards the reactor pressure vessel (RPV).In the ROCOM experiments, the length of the flow ramp and the initial density difference between the slugs and the ambient coolant was varied. From the test matrix experiments with 0 resp. 2% density difference between the de-borated slugs and the ambient coolant were used to validate the CFD software ANSYS CFX. To model the effects of turbulence on the mean flow a higher order Reynolds stress turbulence model was employed and a mesh consisting of 6.4 million hybrid elements was utilized. Only the experiments and CFD calculations with modeled density differences show stratification in the downcomer. Depending on the degree of density differences the less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side, the lower borated coolant is entrained by the colder safety injection water and transported to the core. The validation proves that ANSYS CFX is able to simulate appropriately the flow field and mixing effects of coolant with different densities.     

Investigation of fluid mixing with direct vessel injection

The object of this work is to investigate fluid mixing phenomena as they related to pressurized thermal shock (PTS) in a pressurized water reactor vessel downcomer during transient cooldown with direct vessel injection (DVI), using test models. The test model designs were based on ABB Combustion Engineering (CE) System 80+ reactor geometry. A cold-leg, small-break loss-of-coolant accident (LOCA) and a main steam line break were selected as the potential PTS events for the ABB-CE System 80+. This work consists of two parts. The first part provides the visualization tests of the fluid mixing between DVI fluids and existing coolant in the downcomer region, and the second part presents the results of thermal mixing tests with DVI in the other test model. Flow visualization tests with DVI have clarified the physical interaction between DVI fluid and primary coolant during transient cooldown. A significant temperature drop was observed in the downcomer during the tests of a small-break LOCA. The measured transient temperature profiles compare well with the predictions from the REMIX code for a small-break LOCA, and with the calculations from the COMMIX-1B code for a stream line break event.     

Experiments on slug mixing under natural circulation conditions at the ROCOM test facility using high-resolution measurement techniques and numerical modeling

ROCOM is a four-loop test facility used for the investigation of coolant mixing in the primary circuit of pressurized water reactors. Recently, a new sensor was developed for an improved visualisation and quantification of the coolant mixing in the downcomer. This new sensor array spans a dense measuring grid and covers nearly the whole downcomer. In the presented work, special emphasis was given to the comparison of the data of this sensor with the results of calculations using the Computational Fluid Dynamics (CFD) code ANSYS CFX. A coolant mixing experiment during natural circulation conditions has been conducted. The underlying scenario of this experiment is based on a boron dilution scenario following a SBLOCA event. The corresponding CFD code solution has been obtained using the Best Practice Guidelines. All main effects observed in the measurement are described by the calculation. The detailed comparison reveals that the calculation underestimates the coolant mixing inside the reactor pressure vessel.The measurement data, boundary conditions of the experiment and facility geometry can be made available to other CFD code users for benchmarking.     

High-resolution boron dilution measurements using laser induced fluorescence (LIF)

The application of the laser induced fluorescence technique to the study of liquid mixing in the downcomer of a pressurized water reactor is presented. The scenario is that of a boron dilution event, in which a deborated slug is set in motion by the actuation of a reactor coolant pump. A separate effects test facility, built with transparent plexiglas, is used to conduct optical measurements of the slug mixing along its path to the core. The optical assembly is described and the conditions for the implementation of laser induced fluorescence as a quantitative measurement technique are discussed. Results from a slug injection experiment are shown which demonstrate the high-resolution capabilities of this procedure as applied to the study of liquid mixing in the complex geometry of a reactor vessel downcomer.     

Analysis of System Thermal Hydraulic Responses for Passive Safety Injection Experiment at ROSA-IV/Large Scale Test Facility

An experiment was conducted at the ROSA-IV/Large Scale Test Facility (LSTF) on the performance of a gravity-driven emergency core coolant (ECC) injection system attached to a pressurized water reactor (PWR). Such a gravity-driven injection system, though not used in the current-generation PWRs, is proposed for future reactor designs. The experiment was performed to identify key phenomena peculiar to the operation of a gravity injection system and to provide data base for code assessment against such phenomena. The simulated injection system consisted of a tank which was initially filled with cold water of the same pressure as the primary system. The tank was connected at its top and bottom, respectively, to the cold leg and the vessel downcomer. The injection into the downcomer was driven primarily by the static head difference between the cold water in the tank and the hot water in the pressure balance line (PBL) connecting the cold leg to the tank top. The injection flow was oscillatory after the flow through the PBL became two-phase flow. The experiment was post-test analyzed using a JAERI modified version of the RELAP5/MOD2 code. The code calculation simulated reasonably well the system responses observed in the experiment, and suggested that the oscillations in the injection flow was caused by oscillatory liquid holdup in the PBL connecting the cold leg to tank top.     

三环路压水堆压力容器上腔室交混矩阵数值研究

使用STAR-CCM+软件对三环路压水堆压力容器上腔室流场进行了大规模、精细化三维数值模拟,并采用组分跟踪方法分别对157个燃料组件出口冷却剂流动进行计算,构造了一个具有3×157个元素的“上腔室交混矩阵”,用该矩阵即可定量、精确地描述冷却剂从堆芯流出后,经上腔室内交混并再分配到各热管道的复杂流动过程。研究发现堆芯流出的冷却剂在压力容器上腔室内的交混是并不充分的,径向上不同位置燃料组件流出的冷却剂会在上腔室同热管道的接口区域存在明显的对应关系,而燃料组件径向功率分布的差异必然导致热管道中冷却剂热分层现象的产生。      

Experiments at the mixing test facility ROCOM for benchmarking of CFD codes

For the validation of computational fluid dynamics (CFD) codes, experimental data on fluid flow parameters with high resolution in time and space are needed.Rossendorf Coolant Mixing Model (ROCOM) is a test facility for the investigation of coolant mixing in the primary circuit of pressurized water reactors. This facility reproduces the primary circuit of a German KONVOI-type reactor. All important details of the reactor pressure vessel are modelled at a linear scale of 1:5. The facility is characterized by flexible possibilities of operation in a wide variety of flow regimes and boundary conditions. The flow path of the coolant from the cold legs through the downcomer until the inlet into the core is equipped with high-resolution detectors, in particular, wire mesh sensors in the downcomer of the vessel with a mesh of 64 × 32 measurement positions and in the core inlet plane with one measurement position for the entry into each fuel assembly, to enable high-level CFD code validation. Two different types of experiments at the ROCOM test facility have been proposed for this purpose. The first proposal concerns the transport of a slug of hot, under-borated condensate, which has formed in the cold leg after a small break LOCA, towards the reactor core under natural circulation. The propagation of the emergency core cooling water in the test facility under natural circulation or even stagnant flow conditions should be investigated in the second type of experiment. The measured data can contribute significantly to the validation of CFD codes for complex mixing processes with high relevance for nuclear safety.     

CFD simulation of Fortum PTS experiment

Detailed simulation of the thermal stresses of the reactor pressure vessel (RPV) wall in case of pressurized thermal shock (PTS) requires the simulation of the thermal mixing of cold high-pressure safety injection (HPI) water injected to the cold leg and flowing further to the downcomer. The simulation of the complex mixing phenomena including, e.g., stratification in the cold leg and buoyancy driven plume in the downcomer is a great challenge for CFD methods and requires careful validation of the used modelling methods.The selected experiment of Fortum mixing test facility modelling the Loviisa VVER-440 NPP has been used for the validation of CFD methods for thermal mixing phenomena related to PTS. The experimental data includes local temperature values measured in the cold leg and downcomer. Conclusions have been made on the applicability of used CFD method to thermal mixing simulations in case with stratification in the cold leg and buoyant plume in the downcomer.