直流蒸汽发生器蒸干点波动引起的热应力及疲劳分析

为研究蒸干点波动对蒸汽发生器传热管造成的损伤，以Babcock&Wilcox公司设计的直流蒸汽发生器为原型，首先利用一、二次侧耦合传热的方法得到相关热工水力参数，通过对比不同波动频率下蒸干点径向温度分布确定波动频率的影响，利用有限元分析得到传热管应力分布，最后根据S-N曲线对传热管进行疲劳评估，并探讨相关因素的影响。研究结果表明，蒸干点波动频率较低时径向温度分布与稳态相似，接触二次侧的传热管外壁面更容易发生疲劳损坏，虽然交变应力小于限值，但在堆内环境下存在一定运行隐患，温度波动幅值增大会导致传热管寿命明显下降，采用弹性约束有利于缓解蒸干点波动引起的疲劳。本研究为直流蒸汽发生器传热管在蒸干点波动条件下的寿命预测及安全运行提供了参考。     

铅铋快堆SGTR事故下高压过冷水注入高温铅铋合金流动传热数值模拟研究

铅铋快堆内蒸汽发生器传热管两侧为高压过冷水和高温铅铋冷却剂，传热管两侧较大的压差和温差以及液态铅铋合金（LBE）的腐蚀效应可能造成蒸汽发生器传热管破裂（SGTR）事故。深入研究事故后高压过冷水冲击高温液态LBE的射流沸腾和相变产物蒸汽扩散的特征，具有十分重要的学术意义和工程应用价值。为揭示事故工况下液态LBE与水相互作用的传热传质机理，基于流体体积（VOF）方法，结合LES湍流模型和Lee相变模型，建立了水/蒸汽-液态铅铋多相流动与传热的三维数值计算模型，系统研究了高压过冷水注入高温LBE内发生的相变传热过程。结合注入压力及过冷水温度等因素，分析了射流沸腾过程中不同工况对射流形态、迁移深度以及沸腾行为的影响，研究结果可为SGTR事故工况下堆芯安全性预测提供指导。     

卧式螺旋管式蒸汽发生器管内沸腾传热恶化的实验研究

报道了在高压汽水回路上进行的卧式螺旋管式蒸汽发生器管内沸腾传热恶化特性的实验结果，并对实验参数范围内出现的传热恶化进行了分类；分析和研究了出现恶化的条件及机理；给出了传热恶化时的热负荷关系式。     

管壁减薄对严重事故工况下传热管断裂的影响及缓解措施的有效性

严重事故的恶劣条件(反复的冷热交替及一、二回路之间的压差)可能导致蒸汽发生器(SG)传热管发生蠕变断裂。本文基于一级概率安全分析(PSA)的分析结果确定的典型事故序列,计算分析SG传热管壁减薄对严重事故工况下诱发蒸汽发生器传热管断裂(SGTR)的影响,给出严重事故缓解措施,例如一回路降压和给SG补水的有效性计算。     

基于流热固耦合的核电蒸汽发生器传热管热应力数值模拟

以大亚湾核电站蒸汽发生器为原型,基于相似模化原理建立了蒸汽发生器简化物理模型。采用两流体模型及热弹性力学基本关系式分别描述气液两相流沸腾相变过程和热应力变化规律。利用CFX对一、二回路侧流体流动传热及与传热管的耦合换热过程进行了数值模拟,并在ANSYS WORKBENCH中实现了流体温度场载荷向结构的传递,进而对传热管进行稳态热分析和热应力分析。计算结果表明：二回路出口质量含汽率为24.5%,冷却剂出口温度为296.2 ℃,均与大亚湾蒸汽发生器实际运行参数相符;传热管热应力与其壁面温差分布一致,且沿壁厚方向先减小后增大,并存在中性层,传热管最大热应力为54.5 MPa。研究结果为蒸汽发生器的优化设计及安全运行提供了一定的理论支撑。     

传热管泄漏对蒸汽发生器动态特性的影响

以大亚湾核电站蒸汽发生器为研究对象,建立了基于漂移流理论的蒸汽发生器一维动态数学模型及传热管泄漏模型,并进行了蒸汽发生器不同工况下的稳态仿真。在验证所建立漂移流模型和传热管泄漏模型的基础上,研究了不同工况下传热管泄漏位置及泄漏流量对蒸汽发生器关键参数的影响。研究结果表明,所建立的漂移流模型和传热管泄漏模型能准确反映不同泄漏情况下蒸汽发生器质量含汽率及蒸汽压力等关键参数的变化规律,泄漏发生在热端沸腾段入口处时各参数变化最显著,泄漏量为冷却剂流量的5%时出口质量含汽率由0.261降到0.163。基于漂移流理论传热管泄漏对蒸汽发生器动态特性影响的成功预测,为蒸汽发生器传热管泄漏事故的监测与防范措施的制定提供一定参考。     

蒸汽发生器传热管泵致脉动分析

主泵是核电厂主回路的核心设备之一,它产生的压力脉动会导致回路设备部件的振动,是设备发生疲劳失效的主要原因之一。对于AP1000核电厂,主泵和蒸汽发生器特殊的布置方式增加了泵致脉动对蒸汽发生器的影响。针对AP1000蒸汽发生器传热管泵致脉动分析建立了简化模型,计算结果可用于蒸汽发生器传热管的疲劳分析评定。     

压水堆核电厂二回路水质泥渣沉积分析与控制优化

核电厂蒸汽发生器相当于一个巨大的垃圾收集器,二回路系统的杂质及异物等均进入蒸汽发生器后,容易发生杂质沉积,并导致蒸汽发生器传热管传热效率降低,严重时甚至会引起蒸汽发生器传热管腐蚀破损。因此,本文从核电厂二回路各系统管道和容器的材质、二回路水质控制以及二回路腐蚀等方面出发,分析核电厂蒸汽发生器的泥渣含量高的原因,并提出合理的技术改进。最终达到降低蒸汽发生器泥渣量的沉积,提高蒸汽发生器的安全使用寿命的目的。     

蒸汽传热管破裂事故中颗粒物喷射的数值模拟

蒸汽传热管破裂事故发生时,一回路中的颗粒物随着高温高压流体一起喷射进入二次侧,会对蒸汽发生器二次侧产生冲击并加速二次侧腐蚀。根据大亚湾核电站蒸汽发生器传热管实际尺寸建模,利用FLUENT流体软件,对蒸汽传热管破裂事故发生时的颗粒物喷射过程进行模拟研究。对连续相采取k-ε模型预测湍流变化,多相流模型选择混合模型,对颗粒相采取DPM(Discrete Particle Model)模型跟踪颗粒运动轨迹。研究发现,破口附近是高速场和高温度场,而在破口上方,温度场和速度场逐渐恢复均匀性;在破口上方附近存在回流现象;破口附近颗粒随着流场运动对二次侧管壁产生冲击,破口附近壁面所受到的冲击最严重;在破口上方及背离破口的管道附近存在颗粒物聚集现象,可能导致传热恶化;随着高度的增加,破口的影响开始扩散。     

蒸汽发生器传热管束过冷沸腾区两相流动数值模拟

基于两流体欧拉数学模型结合RPI壁面沸腾模型,利用大型商用CFD软件ANSYS CFX 12.0对蒸汽发生器传热管束过冷沸腾区一次侧、壁面和二次侧耦合传热过程进行了数值模拟。研究了三叶梅花孔支撑板和不同入口过冷度条件下蒸汽发生器传热管束内的流动沸腾现象,得到一、二次侧流场与温度场,二次侧空泡份额分布,支撑板梅花孔局部的流动状况及不同入口过冷度对蒸汽发生器热工水力特性的影响。数值模拟结果表明,三叶梅花孔支撑板的存在及不同入口过冷度对蒸汽发生器传热管束过冷沸腾区域的热工水力特性影响显著。     

300MW核电汽轮机凝汽器动态过程的数值分析

凝汽器冷却管热应力直接影响到冷却管与管板之间连接的密封性，从而影响到蒸汽发生器的安全运行。通过对300MW核电汽轮机凝汽器动态过程数值仿真，分析了汽轮机真空系统严密性试验，冷却水中断以及汽轮机甩全负荷对凝汽器冷却管热应力的影响，为提高蒸汽发生器运行的安全性。奠定了理论基础。     

Thermal hydraulic investigations of intermediate heat exchanger in a pool-type fast breeder reactor

Intermediate heat exchanger (IHX) in a pool-type liquid metal cooled fast breeder reactor is an important heat exchanging component as it forms an intermediate boundary between the radioactive primary sodium in the pool and the non-radioactive secondary sodium in the steam generator (SG). The thermal loads during steady state and transient conditions impose thermal stresses on the heat exchanger tubes and on the shells which hold the tube bundle. Estimation of these thermal loads and achieving uniform temperature distribution in the tubes and shells by having uniform flow distributions are the major tasks of thermal hydraulic investigations of IHX. Through multi-dimensional thermal hydraulic investigations performed using commercially available computer codes such as PHOENICS, the flow and temperature distributions in the tubes and shells and in its secondary sodium inlet and outlet headers are obtained with and with out provisions of flow distribution devices. The effectiveness of these devices in achieving acceptably uniform flow and temperature distributions has been assessed and thermal loads on the tubes and shells for thermo mechanical analysis of the IHX have been defined. The predictions of the computational studies have been validated against simulated experiments.     

Residual stresses associated with the hydraulic expansion of steam generator tubing into tubesheets

Various methods are being used to expand heat transfer tubes into the thick tubesheets of nuclear steam generators. The residual stresses in the as-expanded tubes and methods for reducing these stresses are important because of the role which residual stresses play in stress corrosion cracking and stress assisted corrosion of the tubing. Of the various expansion processes, the hydraulic expansion process is most amenable to analytical study. This paper presents results on the residual stresses and strains in hydraulically expanded tubes and the tubesheet as computed by two different finite element codes with three different finite element models and by a theoretical incremental analysis method. The calculations include a sensitivity analysis to assess the effects of the expansion variables and the effect of stress relief heat treatments.     

高温气冷堆蒸汽发生器两相流不稳定性预报

论述了蒸汽发生器立式上升流动螺旋管内高压汽－水两相流不稳定性试验研究。研究结果表明，螺旋管中存在压力降型、密度波型和热力型脉动。采用无因次分析法得到了预测系统稳定的经验关系式，并得到了判断系统稳定性的界限图。同时对蒸汽发生器立式下降流动螺旋管与立式上升流动螺旋管的不稳定性进行了比较。最后对实际蒸汽发生器两相流不稳定性进行了预报。     

Assessment of actual loading conditions of steam generators and heat exchangers according to probabilistic criteria

Classical design codes are based on nominal conditions, e.g. pressure, temperature, material strength and tube diameter for tubes under pressure; local loads, hot spots, deviations in strength and geometry, etc. are compensated by the use of safety factors. More recent methods — in particular those for the assessment of steam generators and heat exchangers for nuclear applications — call for designs to meet actual loading conditions. Thus, detailed investigations are required to determine realistic operational characteristics of the components to be assessed. However, a worst case treatment on the basis of the above criteria would yield an over-conservative design which in special cases might even be worse than the classical one, e.g. in the case of heated tubes due to increased thermal stresses. As a consequence, with all the data or the spectra of data available, the assessment has to be executed according to probabilistic criteria resulting in a design with an acceptably low and, above all, known failure rate adjusted to fit into the overall plant concept.In components to which the above criteria have been applied before — mainly pressure vessels — the design variables pressure and temperature are generally considered uncorrelated and the temperature is in most cases assumed to be constant. For heat exchangers and steam generators these simplications are not acceptable on principle. On the contrary, even deviations in geometry (like tube tolerances) result in flow rate deviations with corresponding deviations in temperature and pressure. Thus a new procedure for assessment according to actual loading conditions, e.g. by the application of probabilistic criteria, is proposed. This procedure can be used for heat exchangers and steam generators. With this new method, components of nuclear power plants like steam generators and heat exchangers may be designed to meet the low failure probability required for satisfying the overall reliability concept of the plant.     

Calculation of tube degradation induced by dryout instability in sodium-heated steam generators

Analytical methods are adapted and presented for the calculation of thermal fluctuations and thermal stresses experienced by a tube wall in the region of departure from nucleate boiling (DNB) in a sodium-heated steam generator. Calculated results are presented using parameter ranges and geometry adopted from the Atomics International reference design for the steam generators of the Clinch River Breeder Reactor Plant. The physical phenomenon was modeled by subjecting the tube wall and the oxide scale layer on the water side of the tube (treated as either a powdery substance or a dense protective film) to periodic water heat transfer coefficient fluctuations simulating the effect of DNB in that area of the tube. The stresses obtained were employed in estimating (where possible) the most detrimental contribution to the life expectancy of the steam generator tube based on fatigue of the tube wall and repeated exfoliation of the oxide layers.     

快堆钠-水直流蒸汽发生器七管样机热工水力特性分析

钠-水直流蒸汽发生器是钠冷快堆主热传输系统的关键设备之一,其结构及内部的传热现象是十分复杂的。管内外侧的介质及压力不同,管内侧为高温高压的水/蒸汽,存在复杂的两相流动传热传质现象;管外侧为高温液态金属钠,沿换热管高度方向存在较大的钠温变化。本文以钠-水直流蒸汽发生器七管样机为研究对象,对其热工水力特性进行了CFD分析和实验研究,CFD分析结果和实验结果吻合较好,验证了CFD分析所采用的数学模型和数值方法的可靠性。结果表明,钠-水直流蒸汽发生器七管样机的传热面积是足够的,达到了设计指标要求,其界限质量含汽率约为0.42,临界热流密度约为451.98 kW/m²,从而确定了蒸干点的位置。     

高温气冷堆螺旋管式直流蒸汽发生器热工水力学

高温气冷堆蒸汽发生器具有一次侧氦气工质、二次侧直流、螺旋管结构、工作温度高等特点,其热工水力特性与传统压水堆自然循环蒸汽发生器存在很大区别。针对高温气冷堆蒸汽发生器的特点,对其基础热工水力及特有热工水力学问题进行了阐述,主要包括螺旋管内单相及两相流阻及换热计算、横掠螺旋管束流阻及换热计算、温度均匀性及两相流不稳定性等。同时介绍了清华大学核能与新能源技术研究院针对高温气冷堆蒸汽发生器热工设计、温度均匀性及两相流不稳定性等热工水力学问题所开发的一维稳态程序、一维瞬态程序、二维分析程序和方法,并对分析结果和结论进行了讨论。相关研究方法、程序和结论对其他相似参数螺旋管和直管式直流蒸汽发生器具有参考和借鉴意义。     

Analysis of EBR-II secondary sodium network for pressures pulses due to leaks of steam of water into sodium

A systematic procedure is presented for the safety analysis of the secondary sodium system of an LMFBR in the event of a leak of water or steam into sodium in an evaporator or a superheater. Using fracture mechanics, it is shown that the usual assumption of failure initiation by guillotine rupture of one or more water or steam tubes is unrealistic. A model is proposed for the gradual growth of leaks due to the phenomenon of wastage. The pressure rise in the system due to a sodium-water reaction is calculated solving one-dimensional hydrodynamic equations. By comparing results obtained for the proposed gradual growth model with those due to guillotine failure it is shown that the assumption of guillotine failure leads to a significant overestimation of pressures and stresses. Based upon the proposed leak progression model the stresses in the EBR-II secondary sodium system are shown to be within safe limits.