Calculation of Transverse Acceptance of CYCIAE-100

CARR燃料组件是板状元件，共有21块燃料板和20个矩形冷却流道。流道宽度不均匀，从外到内分别有2．59、2．45、2．32和2．20mm4种宽度，长度均为66．6mm。燃料组件的主要特点是热流密度高、传热性能好。良好的组件流量分配特性是充分发挥这些优点的重要保障。对于这种窄缝流道的水力特性和流量分配实验，国内外均进行不多，加之结构复杂，流速很高，因此，很有必要对它们进行实验验证。     

矩形窄缝通道受热表面污垢沉积实验研究

失水事故下堆芯余热排出时，板状燃料元件表面因冷却剂引入负溶解度盐杂质而发生污垢沉积，导致传热性能恶化甚至堵塞通道。为研究矩形窄缝通道内受热表面污垢沉积特性，设计搭建了一种采集矩形窄缝通道表面特定方位处污垢样品的实验装置，以碳酸钙作为可溶性杂质，对60 mm×2 mm截面的矩形窄缝通道受热面污垢沉积过程进行实验测试，观测通道不同位置、不同时间下污垢沉积的微观形貌特征，测量污垢厚度的分布特性，并探究污垢沉积对壁面传热的影响规律。结果表明：污垢形态随沉积时间增加而改变，不同位置处的沉积厚度差异显著，通道传热系数随时间先下降后趋于稳定，下降幅度为26.49%。     

矩形通道速度边界层可视化实验研究

板状燃料组件结构紧凑、冷却剂通道狭窄，其边界层流场特性是决定矩形通道与常规通道内单相流动和传热特性存在差异的重要因素。本文采用粒子图像测速(PIV)技术，对间隙为2 mm和3 mm的矩形通道的速度边界层进行了可视化实验研究，分析了矩形通道边界层内速度分布、雷诺切应力等流场特性，探究了通道间隙对边界层的影响规律。实验结果表明，矩形通道的湍流边界层无量纲速度分布符合Spalding公式，在距离窄边壁面0.2～0.3 mm范围内存在雷诺切应力峰值区，随着雷诺数的增加，速度边界层的黏性底层逐渐减薄，对数律层占比增大，雷诺切应力峰值区向壁面方向靠近。减小矩形通道间隙，将会限制近壁面速度剖面的发展，使得近壁面速度梯度增大，湍流强度减小。     

窄间隙矩形通道间隙厚度对传热强化影响研究

根据窄间隙矩形通道的流道结构特点，参考圆管环状流临界热流密度（CHF）预测解析模型，得到了可以预测间隙厚度不小于0．5mm的窄间隙矩形通道内发生沸腾两相流环状流时的CHF解析模型。计算表明，当窄间隙矩形通道的进口截面宽度与间隙厚度比为25～85时，通道内的CHF值强化比较明显。根据汽-液两相介质的特点，推导出了在沸腾两相流系统中发生CHF时的传热强化判定准则。分析计算表明，这个判定准则是合理的，传热强化较好的进口截面宽度与间隙厚度比为45～75。综合两者的计算结果，窄间隙矩形通道内传热强化的参考进口截面宽度与间隙厚度比为45～75。     

矩形窄缝通道内再淹没过程骤冷温度研究

反应堆失水事故后，堆芯再淹没过程是维持燃料元件完整性以及缓解事故严重程度的重要手段之一。骤冷温度是再淹没过程关键参数，对了解再淹没过程先驱冷却与骤冷过程有着重要意义。本文基于双面加热的矩形窄缝通道试验装置，研究了矩形窄缝通道内再淹没过程，探究了初始壁面温度、加热功率、冷却剂流速、冷却剂过冷度、压力等对骤冷温度的影响，并通过量纲分析手段，提出了矩形窄缝通道内骤冷温度预测模型。结果表明，骤冷温度随初始壁面温度、加热功率、压力的升高而升高，与冷却剂流速与入口过冷度相关性较小，提出的骤冷温度预测模型预测效果良好。     

在垂直环形窄缝流道中的沸腾传热特性研究

为了弄清在窄缝环形流道中气泡的形成、聚合和变形的特性 ,以及气泡在聚合变形之后对传热特性的影响 ,在常压下用蒸馏水对窄缝间隙为 0 75mm的垂直环形流道 ,进行了可视化的流动沸腾传热实验研究 ;实验段的有效加热长度为 90 0mm ,其加热方式为单面内侧加热 ,实验的流量变化范围为 1 667× 1 0 - 5m3/s至 5 833× 1 0 - 5m3/s。实验得到了在不同质量流密度和热流密度下窄缝流道中的沸腾传热系数随干度变化的分布。通过与常规流道中的沸腾传热系数的比较 ,得到了在窄缝环形流道中沸腾传热系数比常规流道中的沸腾传热系数约高 1 5 %的结论。另外通过用高速摄像机对可视化的垂直环形流道中的流型进行的拍摄研究 ,分清了存在在窄缝环形流道中的四种流型     

中国先进研究堆矩形通道流场数值计算分析

通过SIMPLE数值方法,编制程序,对中国先进研究堆(CARR)全流道进行流场数值模拟.采用对CARR的单个冷却剂通道进行单相水的数值传热计算,并递增地改变流道入口流速,计算获得与入口流速对应的流道速度场与温度场分布,展现其变化规律,分析入口流速对流道热工水力参数分布的影响.采用所编制的程序,对板式燃料组件构成的窄矩形通道进行数值模拟,由此来确定热工水力设计需要的一些反应堆安全参数.这些安全参数为反应堆事故监测系统提供必要的热工过程状态信息,也为CARR提供必要的数据参考.     

双矩形通道密度波不稳定性研究

针对窄间隙矩形通道的密度波不稳定性进行了试验研究,试验采用了断面尺寸为25 mm×2 mm,加热长度为1 000 mm的双扁形矩形通道组成试验段.试验结果发现增加质量流速、入口过冷度、压力,均能增加稳定流动的范围.脉动周期随质量流速的增加而变短,随入口过冷度的增加而增加.压力对脉动周期的影响较小.以无量纲过冷度Nsub和相变数Npch比较不同长度试验段的结果,发现平行矩形通道的结果和平行直圆管的结果基本重合,长度和流道断面形状对流动不稳定性的影响较小.     

窄缝矩形通道单相流动及传热实验研究

以垂直向上窄缝矩形通道内去离子水为流动介质,对单相等温流动及恒热流密度条件下的单相传热进行了实验研究.结果表明,窄缝矩形通道内的单相等温流动特性及单相传热特性并未偏离常规尺度通道内的相关规律,采用经典理论解或关系式能获得较好的预测结果.     

在双面加热的窄缝环形流道中流动沸腾换热研究

在窄缝流道内发生沸腾换热现象时,由于沸腾产生的汽泡受窄缝流道的限制,受压变形而消除了汽泡表面张力对传热的影响。因此对此现象进行基础性理论研究具有很重要的意义。本文在常压下用蒸馏水对窄缝间隙为 0.75mm的垂直环形流道,进行了流动沸腾传热实验研究。实验段的有效加热长度为 900mm,其加热方式为内外侧双面加热,实验的流量变化范围在 1.67× 10－ 5～ 5.83× 10－ 5m3/s。通过实验得到了在不同质量流速和热流密度下双面加热的窄缝流道中内外侧沸腾换热系数随干度变化的分布和特点。研究结果表明,由于在窄缝流道中存在着大量的运动聚合受压变形汽泡,因此使内外侧沸腾换热系数都很高 (可达 105W· m－ 2· K－ 1以上 )。     

流量分配比对环形燃料芯块传热特性影响数值模拟研究

环形燃料具有两条冷却通道，外通道与内通道的冷却水流量分配比（φ）的变化可能会对芯块传热特性产生影响。本文建立了环形燃料单棒流固耦合CFD计算模型，在4种不同的流量分配比工况下，通过计算3个反映芯块传热特性的评价指标，研究了流量分配比变化对环形燃料芯块传热特性的影响。由分析计算结果可知，流量分配比变化不会对有间隙结构的环形燃料的芯块传热特性产生显著影响。     

Measurement of Temperature Fluctuations in a Natural Convection Water Loop

Investigations of reactor noise in water-cooled research reactors show that the power spectral density rises in the low frequency domain. The cause of this phenomenon is often attributed to fluctuations in the coolant temperature, but this has never been proved experimentally. The present experiment is an attempt in this direction.

The temperature fluctuation in a natural convection heat transfer loop decoupled from neutronics was measured and analyzed in the frequency domain. The test section of the loop had a rectangular channel measuring 5 mm × 50 mm in cross section and 500 mm in length. This configuration simulated a coolant channel of the MTR-type fuel element used in swimming-pool reactors. The power spectral density of the temperature fluctuation at the channel exit showed a shape similar to the power spectral density of the noise-equivalent source obtained in the Kyoto University Reactor at comparable power levels.     

基于OpenFOAM的铅基快堆流体通道内两相流研究

为研究铅基快堆中铅/铅铋的特殊热物性导致的在两相流情况下的热工水力特性，模拟流体通道中空泡存在对堆芯的输热能力以及安全性的影响，本文采用开源的CFD计算软件OpenFOAM，应用基于VOF方法的数值模拟，构建了铅基快堆中常见的三角形通道模型，通过与子通道程序的验证和单相条件下实验的校核，检验了所用代码的准确性，并对堆内冷却剂通道的两相流进行了模拟。模拟结果表明：随着两相流流速的增大，冷却剂出口温度降低。气液两相流在内通道流动过程中，气相基本在通道内部流动。随着轴向高度的升高，气泡会在内通道的中心区域聚合；燃料组件的角通道是气泡含量多的区域，会造成局部传热恶化，导致组件烧毁。     

Study on Two-phase Flow in Fluid Channel of Lead-based Fast Reactor Based on OpenFOAM

In order to study the thermal-hydraulic characteristics of two-phase flow caused by the special thermal properties of lead/lead-bismuth in lead-based fast reactors, the influence of bubble in fluid channel on the heat transfer capacity and safety of the core was simulated. In this paper the open source CFD calculation software OpenFOAM was adopted, and the numerical simulation was applied based on VOF method to construct a common triangular channel model in lead-based fast reactor. By simulating the two-phase flow of the coolant channel, it is found that as the flow rate increases, the outlet temperature of the coolant decreases. In the flow process of the gas-liquid two-phase flow in the channel, it can be found that the gas phase basically flows inside the channel. In the simulation of the fuel assembly, the corner channel is an area with a large amount of bubbles, which will cause the local heat transfer to deteriorate and cause the fuel assembly to burn out.     

空间堆辐射散热器设计分析

本文针对空间堆热管辐射散热器进行了初步设计分析，建立了单块辐射板传热模型，包括冷却剂与蒸发段的对流换热、热管内部由蒸发段到冷凝段的传热、冷凝段和C-C包壳之间的传热、C-C包壳辐射散热量等。选取了5种不同热管数目的方案进行计算，得到每种方案下冷却剂支管冷却剂温度沿流动方向的变化规律。结果表明，当热管根数为7 436时，满足设计要求。在热管根数固定的情况下，辐射散热器的最佳翅片宽度为30 mm，单块辐射板合适的冷却剂流量为0.5 kg/s。     

Numerical simulation and experimental research on CHF characteristics of AP1000 nuclear reactor

The present paper aims to investigate the critical heat flux (CHF) characteristics of AP1000 reactor based on the experimental and numerical researches, under normal operation and loop fault conditions, respectively. The differences of flow characteristics in these conditions were analyzed. It indicated that the flow features are very complicated in three dimensions and AP1000 has better self-regulation capability to distribute coolant flow compared to conventional reactors. Under normal operation condition, coolant of two loops is distributed along circumference of the reactor annular channel symmetrically. In case that one of the loops fails suddenly and the coolant is partially lost to total loss, the core flow distribution plate and lower grid plate cannot eliminate uneven flow immediately due to loop failure, also the nonuniformity of reactor coolant flow distribution increases gradually, which leads to the heat transfer deterioration easily. In addition, the reactor core departure from nuclear boiling ratio (DNBR) and CHF does not show a certain linear relation, and the DNBR and CHF of AP1000 are greater than that of conventional reactors which not only improve the reactor thermal efficiency, but also obviously reduce the probability of CHF phenomenon appear.     

Experimental and analytical dynamic analysis of coolant channel and its internals of Indian 540 MWe PHWR reactor

The horizontal coolant channel is one of the important parts of primary heat transport system in PHWR type of reactors. There are in all 392 channels in the core of Indian 540 MWe reactor. Each channel houses 13 natural uranium fuel bundles and shielding and sealing plugs one each on either side of the channel. The heavy water coolant flows through the coolant channel and carry the nuclear heat to outside the core for steam generation and power production in the turbo-generator. India has commissioned one 540 MWe PHWR reactor in September 2005 and another similar unit will be in operation very shortly. For a complete dynamic study of the channel and its internals under the influence of high coolant flow, experimental and analytical study has been carried out. The operating life of a typical coolant channel typically range from 10 to 15 full power years. Towards the end of its operating life, its health monitoring becomes an important activity. Vibration diagnosis plays an important role as a tool for life management of coolant. A good correlation has been achieved between the results of experimental and analytical models. Through the study dynamic characteristics of the coolant channel under simulated loading condition, an attempt has been made to develop a diagnostics to monitor the health of the coolant channel over its operating life. Experimental study has been also carried out to characterize PHWR fuel vibration under different flow conditions. Such results are published for the first time.     

Integral boundary layer heat transfer modeling for Tehran Research Reactor

In the present work the validity of applying the Boussinesq approximation in the analysis of natural convection heat transfer along nuclear fuel plates with large coolant channel aspect ratios is evaluated. The Boussinesq approximation is introduced into the integral boundary layer equations governing the system to describe the velocity and temperature distributions of the coolant in the cooling channels. The fuel plate temperature is related to the adjacent coolant fluid temperature by a fundamental law in conduction heat transfer. Air and water are considered as fluids. The coolant flow is assumed to be fully developed which is a convenient assumption for coolant channels having large aspect ratios. Obtained results indicate that the Boussinesq approximation is merely applicable over a limited range of coolant channel outlet fluid temperatures. The use of this approximation produces conservative estimation of the critical plate power for air flow and non-conservative estimation of the critical plate power for water flow.     

数值反应堆堆芯通道级三维热工水力程序CorTAF开发及初步验证

堆芯是核动力系统的核心部件,其完整性是反应堆安全运行的重要前提。传统核反应堆堆芯热工水力分析方法无法满足未来先进核动力系统的高精度模拟需求。本文依托开源CFD平台OpenFOAM,针对压水堆堆芯棒束结构特点建立了冷却剂流动换热模型、燃料棒导热模型和耦合换热模型,开发了一套基于有限体积法的压水堆全堆芯通道级热工水力特性分析程序CorTAF。选取GE3×3、Weiss和PNL2×6燃料组件流动换热实验开展模型验证,计算结果与实验数据基本符合,表明该程序适用于棒束燃料组件内冷却剂流动换热特性预测。本工作对压水堆堆芯安全分析工具开发具有参考和借鉴意义。     

Analysis, verification, and benchmarking of the transient thermal hydraulic ITERTHA code for the design of ITER divertor

An analytical study for the International Thermonuclear Experimental Reactor Thermal Hydraulic Analysis code (ITERTHA) is carried out for a copper divertor with a 5 mm tungsten tile. The influence of the incident heat flux, swirl-tape insertion in cooling channels as well as the coolant flow velocity on the divertor thermal response is analyzed and discussed. The ITERTHA code results are verified by the commercial finite element code, COSMOS. The heat transfer coefficients at the nodes located on the cooling channel-wall are determined outside COSMOS code by the same methodology used in ITERTHA. A good agreement is achieved under different incident heat fluxes. The ITERTHA code is also benchmarked against the thermal-hydraulic calculation of the outer divertor of the Fusion Ignition Research Experiment, FIRE for an incident heat flux of 20 MW/m² and coolant flow velocity of 10 m/s in a cooling channel of 8 mm diameter with swirl-tape inserts of 2 ratio and 1.5 mm thickness. The results show excellent agreement for both steady and transient states and prove the successful implementation of both the hydraulic and heated diameters of the swirl-tape channels in the used heat transfer correlations.