控制棒导向筒周围流场的模拟实验

采用对流场无干扰的动态应变片测速技术和激光多普勒测速技术，在以水为介质的３００ＭＷ级核电站ＰＷＲ压力容器上腔室１：４可视化模拟研究装置中，测得了上腔室模拟体中各点的流速。从而弄清了作用在控制棒上的水力载荷，为分析控制棒能否按指令在导向管内自由升降和快速下插提供了实验依据，同时也为控制棒导向筒结构的优化设计提供了实验依据。     

压水堆上腔室流场的实验研究

ＰＷＲ作为核电发展的主要堆型，在全世界范围内得到了广泛的应用，也是我国的主要发展堆型。但是对关系到反应堆安全运行的、直接作用在控制棒导向筒上的上腔室流场的分析研究，长期以来由于紊流流动机制的复杂性和上腔室中控制棒导向筒组件布置的密集性，这方面的研究一直没有深入下去。在压水堆运行期间，作用在上腔室构件上的作用力与冷却剂的流动特性有很大的关系，通过模拟实验弄清上腔室的流速分布，对了解作用在控制棒上的水力载荷，以及控制棒能否按指令在导向筒内自由升降和快速下插具有十分重要的意义。本文在３００ＭＷｅ核电站ＰＷＲ上腔室１：４可视化模拟体中，以水为介质进行了上腔室流场的可视化实验研究。采用激光多普勒测速仪（ＬＤＶ）和Ｎ－Ｊ型应变片式测速仪测得了上腔室模拟体中的流速，并用归一化的数据处理方法，显示了整个流场的流速分布规律，找出了整个流速的最大区和最大值。从而为控制棒导向筒的结构力学分析和ＰＷＲ上腔室的数值模拟分析提供实验依据。     

CPWR640核电站严重事故的考虑

ＣＰＷＲ６４０核电站是由中国核动力学研究设计院（ＮＰＩＣ）和上海核工程研究设计院（ＳＮＥＲＤＩ）联合开发的６４０ＭＷ两环路压水堆核电站，该核电站比现有核电站更高的安全目标，严重事故管理已作为电站设计工作的一部分加以考虑，本文简要介绍了在ＣＰＷＲ６４０概念设计过程中对严重事故的考虑。     

压水堆上腔室芯部结构改进的流场研究

为了缓解对控制棒导向筒的横向水力载荷，确保控制棒能按指令在导向筒中自由升降和快速下插，结合服役中控制棒导向筒的结构，对上腔室芯部作了如下２种情况的改进：１对原ＰＷＲ靠近上腔室出口管嘴附近的控制棒导向筒组件（对称的４组：０２２６，０３２５，１１２９，１２２８）加设了保护套。２改用３３组控制棒导向筒组件的芯部，且对全部３３组控制棒导向筒组件分别加设保护套。在相同实验条件下，对以上２种改进情况的实验结果与原上腔室的实验结果作了比较分析，得到了这２种改进结构均能缓解流场对控制棒导向筒的水力载荷作用的结论。     

秦山300MW核电机组全范围仿真机安全壳模型

提出了一个多节点安全壳数学物理模型。该模型用于秦山３００ＭＷ核电机组全范围仿真机取得了良好的实时仿真效果。与ＣＯＮＴＥＭＰＴ－４／ＭＯＤ３程序计算出的秦山核电站安全壳在失水事故下的动态响应结果进行了比较两者符合良好。     

秦山300MW核电机组全范围仿真机RCS及SG的建模与仿真

本文介绍了一个用于ＰＷＲ一回路实时仿真及分析的两相热工水力模型。模型以五个基本守恒方程为基础，考虑了非均相，不平衡态的效应。该模型已成功地用于秦山３００ＭＷ核电机组全范围仿真机主冷却剂系统及Ｕ型管蒸汽发生器的仿真。验收测试证明，模型能够准确有效地仿真核汽供给系统的热工水力特性。     

用于压水堆核电站瞬态分析的微机程序──MACONP

提出了一个能在微机上运行的ＰＷＲ核电站瞬态分析程序ＭＡＣＯＮＰ。该程序可对有关运行瞬态和大部分设计基准事故进行分析计算。计算精度高、速度快、程序操作简单、使用方便，并给出了几种ＡＴＷＳ瞬态工况的分析结果，与大程序ＲＥＬＡＰ５／ＭＯＤ２和ＲＥＴＲＡＮ０２／ＭＯＤ２计算结果相比较，两者符合良好。     

反应堆流场分布优化实验研究和理论分析

采用法国TRIO-VF反应堆热工水力计算程序对反应堆流场进行数值计算,其几何模型与实验模型完全相同,获得了满意的结果.采用国内外最先进的PIV粒子成像测速仪对反应堆流场进行测量,获得了上腔室周边流场和上封头流场速度分布与压力分布以及吊篮上法兰喷嘴阻力系数.对测量结果进行了详细的分析讨论,并将计算结果与实验结果进行了对比分析.结果表明,上腔室和上封头内流场的速度和压力均相符较好.     

反应堆压力容器下腔室交混特性的数值模拟方法研究

反应堆冷却剂系统蒸汽管道发生破口事故后,硼溶液在反应堆压力容器下腔室的对流交混特性对于反应堆安全分析及事故后缓解与抑制策略制定均有重要作用。本文基于实验结果分析了反应堆压力容器下腔室的交混特性及浓度扩散过程,采用数值模拟方法结合实验数据比较了几种主要模型计算结果的准确性与可靠性。分析结果表明,压力容器下腔室的交混特性呈现出外围扩散特征,温度梯度法与组分输运模型具备描述浓度梯度扩散过程的能力,但在细节分布上仍存在进一步改善与优化的空间。     

大亚湾核电站换料堆芯关键安全参数检验

简要介绍了大亚湾核电站换料堆芯的安全评价范围及安全评价所需检验的关键安全参数和准则，还给出用引进的法国ＩＮＣＯＲＥ程序包对大亚湾二号堆第二循环堆芯安全评价的结果。其评价内容和方法不仅适用于大亚湾核电站换料堆芯，而且对秦山６００ＭＷ核电站及秦山３００ＭＷ等核电站反应堆换料堆芯的安全评价也都具有借鉴作用。     

Upper Plenum Dump during Reflood in PWR Loss-of-Coolant Accident

Upper plenum dump during reflood in a large break loss-of-coolant accident of PWR is studied with the emergency core coolant injection into the upper plenum in addition to the cold leg. Transient experiments were carried out by injecting water into the upper plenum and the simple analysis based on a one-dimensional model was done using the drift flux model in order to investigate the conditions under which water dump through the core occurs during reflood.

The most significant result is an upper plenum dump occurs when the pressure (hydrostatic head) in the upper plenum is greater than that in the lower plenum. Under those circumstances the flow regime isco-current down flow in which the upper plenum is rapidly emptied. On the other hand, when the upper plenum pressure (hydrostatic head) is less than the lower plenum pressure (hydrostatic head), the co-current down flow is not realized but a counter-current flow occurs. With subcooled water injection to the upper plenum, co-current down flow is realized even when the upper plenum hydrostatic head is less than the lower plenum hydrostatic head. The importance of this effect varies according to the magnitude of water subcooling.     

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。      

Measurement of turbulent flow phenomena for the lower plenum of a prismatic gas-cooled reactor

Mean velocity field and turbulence data are presented that measure turbulent flow phenomena in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics design (Gas-Turbine-Modular Helium Reactor). The data were obtained in the Matched-Index-of-Refraction (MIR) facility at Idaho National Laboratory (INL) and are offered as a benchmark for assessing computational fluid dynamics (CFD) software. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. The primary objective of this paper is to document the experiment and present a sample of the data set that has been established for this standard problem.Present results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). The flow in the lower plenum consists of multiple jets injected into a confined crossflow—with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate flow scaled to that expected from the staggered parallel rows of posts in the reactor design. Posts, side walls and end walls are fabricated from clear, fused quartz to match the refractive index of the mineral oil working fluid so that optical techniques may be employed for the measurements. The benefit of the MIR technique is that it permits optical measurements to determine flow characteristics in complex passages and around objects to be obtained without locating intrusive transducers that will disturb the flow field and without distortion of the optical paths. An advantage of the INL system is its large size, leading to improved spatial and temporal resolution compared to similar facilities at smaller scales. A three-dimensional (3D) particle image velocimetry (PIV) system was used to collect the data. Inlet-jet Reynolds numbers (based on the hydraulic diameter of the jet and the time-mean average flow rate) are approximately 4300 and 12,400. Uncertainty analysis and a discussion of the standard problem are included.The measurements reveal complicated flow patterns that include several large recirculation zones, reverse flow near the simulated reflector wall, recirculation zones in the upper portion of the plenum and complex flow patterns around the support posts. Data include three-dimensional PIV images of flow planes, data displays along the coordinate planes (slices) and presentations that describe the component flows at specific regions in the model.     

失水事故排放过程中的不稳定性

叙述了低温供热堆发生上空腔小破口失水事故后，自然循环系统的不稳定性，揭示了在排放过程中，由于冷却剂闪蒸现象引起的系统两相流不稳定性，以及在排放不同阶段中流量振荡特性。     

One-dimensional three-field model of condensation in horizontal countercurrent flow with supercritical liquid velocity

A one-dimensional three-field model was developed to predict the flow of liquid and vapor that results from countercurrent flow of water injected into the hot leg of a PWR and the oncoming steam flowing from the upper plenum. The model solves the conservation equations for mass, momentum, and energy in a continuous-vapor field, a continuous-liquid field, and a dispersed-liquid (entrained-droplet) field. Single-effect experiments performed in the upper plenum test facility (UPTF) of the former SIEMENS KWU (now AREVA) at Mannheim, Germany, were used to validate the countercurrent flow limitation (CCFL) model in case of emergency core cooling water injection into the hot legs. Subcooled water and saturated steam flowed countercurrent in a horizontal pipe with an inside diameter of 0.75 m. The flow of injected water was varied from 150 kg/s to 400 kg/s, and the flow of steam varied from 13 kg/s to 178 kg/s. The subcooling of the liquid ranged from 0 K to 104 K. The velocity of the water at the injection point was supercritical (greater than the celerity of a gravity wave) for all the experiments. The three-field model was successfully used to predict the experimental data, and the results from the model provide insight into the mechanisms that influence the flows of liquid and vapor during countercurrent flow in a hot leg. When the injected water was saturated and the flow of steam was small, all or most of the injected water flowed to the upper plenum. Because the velocity of the liquid remained supercritical, entrainment of droplets was suppressed. When the injected water was saturated and the flow of steam was large, the interfacial shear stress on the continuous liquid caused the velocity in the liquid to become subcritical, resulting in a hydraulic jump. Entrainment ensued, and the flow of liquid to the end of the hot leg was greatly reduced.The influence of condensation on the transition from supercritical to subcritical flow as observed in the experimental data is also predicted with the three-field model. When the injected water was subcooled, condensation on the flow of continuous liquid caused a reduction in the flow of vapor and, consequently, a reduction in the interfacial shear stress. Therefore, the flow of liquid remained supercritical to the end of the hot leg at the upper plenum. The entire flow of injected water flowed to the end of the hot leg at higher flows of steam when the injected water was subcooled than when it was saturated. When the flow of vapor was large enough to cause a hydraulic jump in the subcooled liquid, the rate of entrained droplets was greatly increased. The interfacial surface area of the droplets was several orders of magnitude greater than for the continuous-liquid field, and condensation rate on the droplet field was also several orders of magnitude greater. When the flow of vapor from the upper plenum was at its greatest, most of the flow in the continuous liquid was entrained before reaching the upper plenum. The large flow of subcooled droplets caused three-quarters of the steam to condense.     

CFD analysis of PWR core top and reactor vessel upper plenum internal subdomain models

One aspect of the Westinghouse AP1000™¹ reactor design is the reduction in the number of major components and simplification in manufacturing. One design change relative to current Westinghouse reactors of similar size is that AP1000 reactor vessel has two nozzles/hot legs instead of three. With regard to fuel performance, this design difference creates a different flow field in the reactor vessel upper plenum. The flow exiting from the core and entering the upper plenum must turn toward one of the two outlet nozzles and flow laterally around numerous control rod guide tubes and support columns. Also, below the upper plenum are the upper core plate and the top core region of the 157 fuel assemblies and 69 guidetube assemblies.To determine how the lateral flow in the top of the core and upper plenum compares to the current reactors a CFD model of the flow in the upper portion of the AP1000 reactor vessel was created.Before detailed CFD simulations of the flow in the entire upper plenum and top core regions were performed, conducting local simulations for smaller sections of the domain provided crucial and detailed physical aspects of the flow. These sub-domain models were used to perform mesh sensitivities and to assess what geometrical details may be eliminated from the larger model in order to reduce mesh size and computational requirements. In this paper, CFD analysis is presented for two subdomain models: the top core region and control rod guide tube region. These models are chosen for simulation because guide tube and top core region (including top grid, top nozzle, and hold-down device) are the major components of upper plenum effecting the flow patterns and pressure distribution.The top core region, corresponding to ¼ of fuel assembly, includes components as upper part of the fuel assemblies (top grid, fuel rods, top nozzle), core component hold-down devices, and upper core plates. These components distribute the core flow to different sections of guidetube regions. Mesh sensitivity studies have been conducted for each individual part in order to determine the proper geometrical simplifications. Pressure drop measurement data are compared with the predicted CFD results and act as a guideline for the mesh selection.The guidetube region includes control rod guidetubes themselves, adjacent support columns and open regions. In this study, two models of subdomains are analyzed: (1) a ¼ section of one control rod guide tube by itself and (2) a representative unit cell containing two ¼ sections of adjacent control rod guide tubes and one ¼ section of a neighboring support column.Predicted flow rates at each of the outflow locations in conjunction with results from the mesh sensitivity studies provide guidance on (1) what geometry to preserve or remove, (2) what geometry can be simplified to reduce the required mesh, and (3) an estimate of the total mesh required to model the entire upper plenum and top fuel domain.The commercial CFD code STAR-CCM+ is employed to generate the computational mesh, to solve the Reynolds-averaged Navier–Stokes equations for incompressible flow with a Realizable k–? turbulence model, and to post-process the results.     

AP1000上腔室夹带试验研究

为还原AP1000中上腔室夹带过程,以AP1000为原型按1∶5.6的模化比例建立了试验回路,研究不同蒸汽流量和压力容器液位下上腔室夹带的夹带率。结果表明：蒸汽流量对夹带率的影响很小,夹带率随压力容器液位的升高而增大;在较低液位,夹带率保持稳定,加入堆内构件后,上腔室夹带明显增强。     

Preliminary tests of particle image velocimetry for the upper plenum of a scaled model of a very high temperature gas cooled reactor

The Very High Temperature Reactor (VHTR) is a Generation IV nuclear reactor that is currently under design. During the design process multiple studies have been performed to develop safety codes for the reactor. Two major accidents of interest are the Pressurized Conduction Cooldown (PCC), and the Depressurized Conduction Cooldown (DCC) scenario. Both involve loss of forced coolant to the core, except the latter involves a pressure loss in the main coolant loop. During normal operation a circulator pumps the coolant into the upper plenum and down through the core, but following a loss of forced coolant the natural convection causes the flow to reverse to go through the core into the upper plenum. Computer codes may be developed to simulate the phenomenon that occurs in a PCC or DCC scenario, but benchmark data is needed to validate the simulations; previously there were no experimental test facilities to provide this. This study will cover the design, construction, and preliminary testing of a 1/16th scaled model of a VHTR that uses Particle Image Velocimetry (PIV) for flow visualization in the upper plenum. Three tests were run for a partially heated core at statistically steady state, and PIV was used to generate the velocity field of three naturally convective adjacent jets; the turbulent mixing of the jets was observed. After performing a sensitivity analysis the flow rate of a single pipe was extracted from the PIV flow field, and compared with an ultrasonic flowmeter and analytic flow rate. All the values lied within the uncertainty ranges, validating the test results.     

立式倒U型管蒸汽发生器倒流现象及初步分析

文章涉及中国核动力研究设计院自然循环实验装置单相稳态自然循环实验过程中立式倒U型管蒸汽发生器(UTSG)模拟体一次侧流体的流动特性。实验观察到：1)UTSG模拟体进口腔室压力低于出口腔室压力；2)UTSG模拟体入口腔室温度较热段温度有一陡降。通过对该实验现象的分析可以判定，在单相自然循环工况下，UTSG模拟体中某些传热管内出现了倒流。实验结果表明，倒流的出现使UTSG模拟体自然循环工况下的流动阻力系数较强迫循环工况下的明显增大。 