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。     

共用凝汽器建模与仿真

热力系统的主汽轮机与辅汽轮机共用一个凝汽器,称其为共用凝汽器,在共用凝汽器工作原理的基础上,依据凝汽器结构特点建立共用凝汽器的动态数学模型。仿真分析共用凝汽器典型工况下的运行特性,得到共用凝汽器各子区间不同的热力特性,对探究共用凝汽器的热力特性具有一定意义。     

基于小波神经网络的蒸汽发生器动态过程辨识

在核动力蒸汽发生器(SG)运行过程中，其逆动力学效应使其动态特性难以辨识。为提高蒸汽发生器动态特性辨识的效果，提出了基于小波神经网络的蒸汽发生器动态过程辨识的新方法。辨识模型采用串并联型辨识结构，网络训练采用Levenberg Marququardt学习算法（LMBP）。对蒸汽发生器典型运行工况的辨识结果表明，所提出的方法能够正确地辨识蒸汽发生器的动态特性且具有较高的辨识精度。     

低参数直流蒸汽发生器稳定性实验研究

针对核供热堆的低温、低压运行条件，建立了大盘管直流蒸汽发生器热工水力学实验系统，实验研究了低参数条件下入口流动阻力、运行负荷和蒸汽发生器出口压力对直流蒸汽发生器稳定性的影响。结果表明：入口节流阻力比、运行负荷、系统压力对直流蒸汽发生器稳定性有影响，通过调整入口节流阻力比，直流蒸汽发生器可在低压参数下稳定运行。     

直流蒸汽发生器动态特性与小破口故障仿真研究

以B&W直流蒸汽发生器为对象,基于过程仿真软件(APROS)支撑平台中的基本模块,建立了图形化的直流蒸汽发生器仿真模型。对模型进行变工况下的稳态和动态仿真,由结果可知,一次侧入口焓值与二次侧出口压力对稳态特性影响最大,一次侧入口温度对动态特性影响最大。进一步研究直流蒸汽发生器发生换热管破裂事故时,破口位置和破裂程度对其运行特性的影响。结果表明,破口发生位置接近一次侧入口时,对直流蒸汽发生器运行影响最大;换热管破裂对直流蒸汽发生器运行特性的影响随着破裂程度的增加而增大。     

1000 MW核电汽轮机组凝汽器冷却管漏水原因分析及设计改进

文章介绍了某核电站凝汽器冷却管大面积漏水情况以及凝汽器冷却管漏水处理方案,并对漏水原因进行了排查总结,指出了某核电站凝汽器冷却管大面积漏水的根本原因是高速汽流对凝汽器冷却管造成的激振而使冷却管损坏。在总结某核电站凝汽器冷却管漏水经验反馈基础上对福清核电站二期工程凝汽器进行了设计改进,主要通过增加中间隔板数量,改变冷却管振动特性、在汽道迎流面设置不锈钢假管、增加厚壁管数量、设置导流板、减小中间管板孔径等五方面,对凝汽器进行设计改进,通过这些改进可以避免凝汽器冷却管发生大面积泄漏。     

M310核电机组凝汽器运行方式优化

国内核电厂的常规岛循环冷却水系统通常是采用海水直流供水方式,主要由凝汽器、循环水泵及管网系统组成,是保证汽轮发电机组经济、安全、稳定运行的关键设备。针对凝汽器单侧运行方式对机组产生的不利影响,系统性地分析了凝汽器、循环水泵设备运行特性及相互之间的影响,并通过凝汽器单侧运行现场操作试验和数据分析循环水系统功能设计改进的影响。经过现场试验验证与分析计算,确定了循环水泵运行的实际工况点,分析了凝汽器实际运行冷却管内流速变化的影响,为凝汽器循环水系统运行方式优化提供了数据依据。     

秦一厂320 MW机组增容改造蒸汽发生器热工计算研究

为配合秦一厂320 MW机组运行许可证延续专项汽轮机增容改造后核岛运行总参数(NSSS功率、一回路平均温度)的分析计算,保证蒸汽发生器出口蒸汽压力与流量满足常规岛增容后的新要求.本研究基于THETA程序建立了蒸汽发生器热工模型,通过迭代计算,优化调整蒸汽发生器出口参数来确定新的核岛运行总参数.研究结果表明,在NSSS功...     

CAP1400蒸汽发生器汽水分离装置研究

汽水分离装置是蒸汽发生器中的主要部件,其性能不仅会影响蒸汽发生器水力循环特性及水位适用性,决定上部尺寸大小,而且会影响汽轮机的正常运行。对CAP1400核电厂蒸汽发生器汽水分离装置进行了不同蒸汽负荷、饱和水量及高水位和正常水位等试验工况下的热态性能试验,获得了SP3型初级分离器与P3X型干燥器组合随蒸汽负荷、饱和水量、水位变化的分离特性。通过初级分离器和干燥器的阻力测量,分别获得了分离器和干燥器的阻力特性,对CAP1400蒸汽发生器的设计研发起到支撑作用。     

受DNB限制的压水堆运行工况的优化

压水反应堆中蒸汽发生器的更换要求新蒸汽发生器的变更性能同要求的额定运行工况相匹配。为了补偿蒸汽发生器更换的费用，常常通过对蒸汽发生器的热交换性能做重大的改进，以及修改汽轮机控制阀的位置使得提升电站的功率成为可能。该文讨论由Tractebel提出的方法，该方法用于解决受DNB（偏离泡核沸腾）限制的现代压水堆的这类问题。已开发并认证了一个程序，该程序已用于准备更换蒸汽发生器比利时的多伊尔（Doel)3号蒂昂热（Tihanqe)1两台机组。该程序基于这样一个基本准则，即对大部分的二类事故，最大反应堆功率和进口温度都不会违背超功率限值（目前是118％的额定功率），在该事故过程中，也不会违背DNBR限值。该程序可用于设计者评价各种关键的安全参数以及评估热工水力设计所必须的基本假定对运行工况的影响。该文给出了这种研究的例子，重点放在热工水力设计程序、电站主要的热工水力参数以及新蒸汽发生器的特点上。DNBOPT程序的研制也是期望完成监查、敏感性计算、以及确认由蒸汽发生器供货商完成的计算。     

核电站凝汽器设备运行性能监督及诊断分析

发电厂热力循环中的凝汽器及循环冷却水系统,是保证汽轮发电机组经济、安全、稳定运行的重要系统设备。在汽轮发电机组的实际运行中,由于凝汽器压力变化受到蒸汽热负荷、冷却水进水温度、流量,以及真空系统严密性等诸多因素的影响,运行人员无法通过凝汽器运行压力、温度等直接监测数据来分析判断设备运行性能的好坏。本文依据制造厂家提供的凝汽器特性曲线,通过对某一核电机组的实际运行数据进行分析,并考虑实际运行中海水温度变化引起的凝汽器的变工况运行,提出了凝汽器冷却水温升、传热端差和运行压力等设备运行性能指标参数的基准曲线的确定方法。利用基准曲线可方便获取凝汽器设备运行性能指标参数的应达值,为凝汽器性能变化趋势分析及设备故障原因和诊断提供依据。在核电机组实际运行中,快速识别设备实际运行状态是否偏离设计工况,及时查找凝汽器设备性能劣化原因和处理设备故障,对于保障机组设备安全可靠运行至关重要。     

Analysis of Overheating Rupture in Heat-Transfer Tubes Causing Corrosive High-Temperature Reaction

In the steam generator of a liquid metal fast breeder reactor, a defect penetrating through heat-transfer tube will cause high-pressure water/steam to spout into the low-pressure sodium filling the space outside the tube, to initiate sodium-water reactions. If the leak exceeds an intermediate level (~2kg/s), the reaction jet may rupture adjoining tubes with overheating in the event of insufficient cooling available inside the tubes. Such phenomenon of overheating tube rupture presents a serious problem to the economy and safety of steam generator. With a view to clarifying the failure behavior of steam generator heat-transfer tubes under such condition a model of the phenomenon is derived through a series of tests on sodium-water reactions making use of a test loop representing the scale model of an actual fast breeder steam generator. Comparison of actual test data with analysis based on the model has yielded the following information: The failure behavior of gas-pressurized tubes fall into two categories: (a) by creep failure—occurring upon increase of cumulative damage with tube wall wastage caused by the reaction jet and (b) by ductile failure accompanied by creep—upon tube heating with the reaction jet to the extent of lowering tube wall strength below the hoop stress exerted by tube pressure. Analysis of the two categories of failure results in estimation of the percentage difference between analyzed and measured times to failure of 35–50% in the case of creep failure and of 20–50% in the case of ductile failure accompanied by creep. In practical application to steam generators in order to provide a safety margin a time factor—i.e., the safety factor indicating multiple of actual time to failure—of 3 is adopted against 1.5–2 indicated from test to be the actually applicable value.     

Experimental Study on Heat Removal during Cold Leg Small Break LOCAs in PWRs

Experiments which simulated small break loss-of-coolant accidents (SBLOCAs) resulting from 2.1–0.13% break in the cold leg of a PWR were conducted with an apparatus of 1/270 scale in volume. In the large break size case, the decay heat was mainly removed by the break flow and in the case of a small break, the steam generator played an important role. In this case, thermal hydraulic behaviors such as natural circulation and reflux condensation cooling were important during the transient. Depressurization in the secondary system due to bleeding steam from the steam generator by an operator action was so effective to make the accident to come to an end. The operation to depressurize the secondary system was also efficient to rewet the core which had been uncovered due to a loop seal formation in a cross-over leg.

No effects of initial 200 ppm dissolved gas in the coolant were observed on the cooling performance of the steam generator. It was considered that it was because the gas which came from the coolant into the steam during the depressurization transient did not remain in the tubes of the steam generator.     

危冷系统对蒸汽发生器传热管破损事故缓解能力分析

针对蒸汽发生器传热管破损事故后果严重,危冷系统投入对事故缓解能力认识不清的问题,为提高危冷系统对事故缓解能力的认识,增强传热管破损事故处置能力,利用MELCOR程序建立了破损安全分析模型。通过计算对比分析了危冷系统投入与否对事故后果的影响,并比较了危冷系统对不同尺寸传热管破损事故的缓解能力。经仿真分析,明确了危冷系统对传热管破损事故的缓解能力,对提高运行人员事故处置能力及保证反应堆运行安全有重要意义。      

快堆蒸汽发生器沸腾传热恶化引起的热应力计算和分析

快堆管壳式直流蒸汽发生器发生沸腾传热恶化是不可避免的,由此引起的传热管管壁温度波动会使传热管受到疲劳破坏。研究蒸汽发生器的沸腾传热恶化及热疲劳破坏的实验昂贵,难度较大。本文依据国外已发表的实验结果,建立蒸汽发生器沸腾传热恶化发生时传热管管壁温度及热应力的分析模型,应用数值方法求解,对蒸汽压力、质量流速、钠汽温差变动的影响进行了讨论并给出了主要结论。     

Determination of equivalent single crack based on coalescence criterion of collinear axial cracks

In a nuclear power plant the steam generator tubes cover a major portion of the primary pressure-retaining boundary. Thus, very conservative approaches have been taken in the light of steam generator tube integrity. According to the present criteria, tubes wall-thinned in excess of 40% should be plugged whatever the cause. However, many analytical and experimental results have shown that no safety problems exist even with thickness reductions greater than 40%. The present criterion was developed about 20 years ago when wear and pitting were dominant causes for steam generator tube degradation, and it is based on tubes with single cracks regardless of the fact that the appearance of multiple cracks is more common in general. The objective of this study is to review the conservatism of the present plugging criteria of steam generator tubes and to propose a new coalescence model for two adjacent through-wall cracks existing in steam generator tubes. Using the existing failure models and experimental results, we reviewed the conservatism of the present plugging criteria. In order to verify the usefulness of the proposed new coalescence model, we performed finite element analysis and some parametric studies. Then, we developed a coalescence evaluation diagram.     

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。     

蒸汽发生器传热管的微动磨损损伤及预测模型

蒸汽发生器是核电厂中能量转换的关键装备,内部高速流经的高温、高压流体引起传热管流激振动,造成传热管微动磨损损伤,严重时发生管道破裂。文章介绍了传热管典型的微动磨损失效案例,相应的模拟实验研究结果,以及机械磨损与冲蚀-腐蚀共同作用的损伤机制。采用工作率模型可对传热管的磨损失效进行合理的寿命预测评估,该预测模型已经在核电厂安全评估方面应用。     

Restraining effect of support plates on the limit loads for circumferential cracks in the steam generator tube

The steam generator in a nuclear power plant is a large heat exchanger that uses heat from a reactor to generate steam to drive the turbine generator. Rupture of a steam generator tube can result in release of fission products to environment outside. Therefore, an accurate integrity assessment of the steam generator tubes with cracks is of great importance for maintaining the safety of a nuclear power plant. The steam generator tubes are supported at regular intervals by tube support plates and rotations of the tubes are restrained. Although it has been reported that the limit load for a circumferential crack is significantly affected by boundary condition of the tube, existing limit load solutions do not consider the restraining effect of tube support plate correctly. In addition, there are no limit load solutions for circumferential cracks in U-bend region with the effect of tube support plate. This paper provides detailed limit load solutions for circumferential cracks in top of tube sheet and the U-bend regions of the steam generator tube with the actual boundary conditions to simulate the restraining effect of the tube support plate. Such solutions are developed based on three-dimensional (3-D) finite element analyses. The resulting limit load solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.