蒸汽发生器传热管堵头有效性分析

蒸汽发生器传热管是核电厂—回路压力边界的薄弱环节,传热管的完整性直接影响到整个一次侧的安全。当传热管出现裂纹、腐蚀或磨损等缺陷时,在评定确认可能会发生一次侧流体进入二次侧情况下,需要对传热管进行堵管。利用有限元法对某蒸汽发生器传热管的滚压堵头进行分析评定,模拟计算在堵管时以及堵管后堵头、传热管接触力情况,通过计算及分析确认堵头在极限运行工况的有效性,计算显示此堵头满足强度要求。     

蒸汽发生器的衬管

早期，当蒸汽发生器传热管破损时，堵管是唯一的修理方法。但随着堵管率的不断提高，当大约20％的传热管被堵管时蒸汽发生器就会丧失其发电能力的裕量。20世纪70年代中期开发了衬管技术，甚至对过去已堵掉的管子也可采用衬管修理。     

蒸汽发生器二次侧清洁度的视频检查

立式平管板Ｕ形管束的压水堆核电站蒸汽发生器安装至瞬变工况期间，管板上表面的铁基金属残余物的堆积和外来物的存在，以及商业运行后管板、传热管、管子支撑板等的低流速区域里沉积物的堆积要求进行蒸汽发生器的清洁度检查。介绍了蒸汽发生器二次侧清洁度的视频检查技术。该技术适用于蒸汽发生器安装至瞬变工况和投入商业运行后可能发生污染的各阶段的清洁度检查。     

蒸汽发生器管板超差管孔胀管的非线性有限元分析

本工作利用三维非线性有限元分析了某核蒸汽发生器管板的胀管效果,研究了尺寸偏差和材料特性变化等对胀管效果的影响,计算了几种典型传热管间隙和刚度下的接触压力,并与非超差管孔胀接触压力进行了对比研究,最后对合适胀管压力的选择提出了建议.     

蒸汽发生器传热管焊接堵管用小型化激光头研制

核电站蒸汽发生器传热管破损将导致放射性冷却剂外泄，因此需进行堵管作业。激光焊接精度高，可用于蒸汽发生器的焊接堵管。然而，受限于蒸汽发生器的空间结构，现有的激光焊接头往往难以满足实际要求，因此需设计定制化的小型激光头。首先，分析和计算了小型化激光头的光路；其次，设计出小型激光头的机械结构，并完成其在蒸汽发生器内的运动仿真和干涉模拟；最后，对所设计的激光头装置进行了模拟传热管焊接验证。结果表明，所设计的小型化激光头具有良好的焊接性能和可操控性能，可用于蒸汽发生器的焊接堵管作业。     

AP1000及CAP1400蒸汽发生器管板钻孔缺陷研究

在核电蒸汽发生器管板钻孔过程中,不可避免地会出现少量的偏差或缺陷,管板管孔产生的偏差或缺陷直接影响传热管与管板的穿管和胀管。结合目前已经完成和正在完成中的AP1000及CAP1400的蒸汽发生器管板钻孔情况,从理论上分析造成管板管孔缺陷的原因缺陷的形式,通过工艺试验验证管孔缺陷的处理及胀管。针对不同的管孔缺陷,采用管板结构完整性的分析计算和模拟管孔缺陷胀接试验,形成AP1000及CAP1400蒸汽发生器管板钻孔缺陷的处理工艺。     

CPR1000蒸汽发生器二次侧三维稳态热工水力分析

采用三维稳态分析软件GENEPI,对CPR1000蒸汽发生器二次侧管束区进行了热工水力计算,利用多孔介质及局部阻力系数来表征传热管及各几何部件的复杂结构和压降影响,得到了二次侧管束区流场、温度场等的分布情况。计算结果表明:管束区最大干度为0.3;将典型传热管的动能数据提供给流致振动软件进行计算分析,结果显示在本工况下,传热管的流致振动在可接受范围内;对管板附近的流场及温度场进行分析,预测了此模型及工况下的泥渣沉积区域,为排污管的设计提供了输入数据。计算结果验证了CPR1000蒸汽发生器二次侧管束区设计的合理性。     

核电站蒸汽发生器一次侧整体及管板三维有限元应力分析

本文叙述了核电站蒸汽发生器一次侧整体(包括下封头、进出水接管、人孔、水室隔板、 管板、部分二次侧筒体)及管板,在内压和接管载荷作用下的三维有限元应力分析。得出一次侧各部件相互间的影响,管板的孔桥应力和孔边峰值应力。计算结果与轴对称有限元应力分析作了比较。对关键部位按ASME规范第Ⅲ篇的应力强度极限作了评价。     

核电站蒸汽发生器的更换

老一代核电站的蒸汽发生器相继发生传热管腐蚀破损，传热管泄漏率不断增高，计划外停堆的几率增加，维修费用也随之增加。当传热管的缺陷超过堵管限制时，一般要采用堵管的措施。随着堵管数量的增加，核电站难以维持其额定功率，这样就必须考虑更换蒸汽发生器。更换蒸汽发生器是保证核电站安全与经济效益的最好策略。在欧洲，广泛采取更换蒸汽发生器以增加传热面积提高热功率，热功率提高可以增加发电量，以便尽早收回更换所需费用。     

蒸汽发生器焊接堵管堵头的设计与评价

蒸汽发生器传热管是一回路承压边界中最薄弱的环节,当传热管发生严重降质时,通常采用堵管方式来避免二次侧流体受到放射污染。为克服现行堵管方法带来的高应力和塑性变形,或近程操作带来的人员受高剂量辐照问题,设计了一种新型蒸汽发生器自熔焊焊接堵管堵头,采用数值模拟方法进行了焊接残余应力研究和应力分析评价。结果表明该焊接堵管堵头是安全可靠的。     

Experience and status on primary system components after 15 years of operation in the nuclear power plant obrigheim

At Obrigheim the first large pressurized light water reactor built in Germany is operating with a nominal power of 345 MW. Since the beginning of electricity production in later 1968 the nuclear power plant Obrigheim (KWO) has proved a reliable, a safe and also an economical operation with a high availability (83%) over 15 years.KWO has shown that it is possible to prove and maintain the safety and reliability of the primary components on the basis of the present regulations and safety requirements. This was achieved by careful maintenance and by applying improved non-destructive test methodsThe reactor pressure vessel with one circumferential weld near the core could be qualified for future operation by means of inservice inspection, irradiation programs, and by implementation of technical changes for normal as well as for abnormal conditions. To maintain the reliability of the steam generators, extensive eddycurrent testing of the tubes has been performed every year. In order to reduce the corrosive attack on the tubes the secondary water chemistry was controlled very sensitively by minimizing the leakage through the condensor and by using all volatile treatment. The intergranular corrosion of the tubes above the tube sheet could be reduced strongly; but an increasing number of small leakages occurred in the tube sheet region. 458 tubes had been plugged in the old steam generators before they were replaced in 1983 by new ones.In summary it can be stated that the continuous effort to maintain a high quality status of the components is responsible for the high operation availability of the plant.     

Safety and availability of steam generator tubes affected by secondary side corrosion

The outside diameter stress corrosion cracking at tube support plates became the dominating ageing mechanism in steam generator tubes made of Inconel 600. A variety of maintenance approaches were developed and implemented world-wide to enable safe and reliable plant operation with affected tubes. Despite different philosophical and physical backgrounds involved, all applied approaches satisfy relevant regulatory requirements. The main goal followed in this paper is to quantify the degree of safety which is achieved through the implementation of selected maintenance approaches. A method is proposed which measures the operational safety and availability through three efficiency parameters: probability of steam generator tube rupture; predicted accidental leak rates through the defects in the tube bundle; and number of plugged tubes. An original probabilistic model quantifies the probability of tube rupture, while procedures available in literature were used to evaluate the accidental leak rates. A numerical example is based on data from the Kr ko NPP (PWR 623 MWe). The maintenance strategies analyzed are: (a) traditional defect depth (40%) plugging criterion; (b) alternate plugging criterion (bobbin coil voltage as defined by EPRI and US NRC); (c) combination of traditional and alternate plugging criteria; and (d) no plugging at all. Advantages of the defect specific approaches (b) and (c) over the traditional one (a) are clearly shown. The efficiency of the traditional approach (a) is shown to be comparable to the no plugging at all approach (d). Finally, a sensitivity analysis aimed at ranking of the input parameters is presented. Uncertain failure models are shown to be the major contributor to the scatter of obtained results.     

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.     

铅铋快堆螺旋管直流蒸汽发生器热工水力特性数值研究

本研究以铅铋快堆螺旋管直流蒸汽发生器（HOTSG）设计结构为研究对象,采用精细网格与多孔介质相结合的物理建模方法,通过一次侧三维湍流计算与二次侧用户自定义函数（UDF）分区传热计算相耦合的手段,在FLUENT求解器中开展了蒸汽发生器的热工水力特性数值分析研究。研究表明:铅铋入口附近的流量分配孔和腔室对应的直管段区域出现铅铋流速峰值,径向最大速度为0.431 m/s;入口腔室至管束区位置受到阻力突变的影响,压力、横流速度、轴向速度变化较大;热工参数变化符合流动与传热机理,临界热流密度（CHF）点附近一二次侧温差最大为109.61 K,此处最大热流密度为323.55 kW/m2。该研究将为铅铋快堆HOTSG结构设计、流致振动及安全评价提供重要的参考。      

Scoping assessment of design characteristics for an enhanced calandria tube

Over the lifetime of a CANada Deuterium Uranium (CANDU) type reactor, the pressure tubes and calandria tubes undergo creep deformation via static, dynamic and thermal stresses accelerated by neutron bombardment. Creep deformation leads to fuelling issues, potential contact between the calandria tube (CT) and the liquid injection shutdown system or between the CT and the pressure tube (PT). As such, this aging phenomenon limits the lifetime of these components. Also, in the event of Loss of Coolant Accident (LOCA) and Loss of Emergency Coolant Injection (LOECI) scenarios, PT/CT contact may occur and if sufficient cooling is not provided, PT/CT rupture may also occur. Conceptual designs were assessed to determine their potential for reducing the effects of aging by improving CT rigidity and thermal performance of the CT. Two different design options for a CT have been investigated using numerical simulation techniques. The CT design options include fins and ribs of different sizes and combinations. The fins and ribs provide improved structural integrity and improved thermal performance over the reactors lifetime. Analyzed results have shown that the design options yield an increased overall strength with a minimal impact on fuel efficiency. The analysis has determined that the finned design option is superior in terms of CT strength enhancement yet the ribbed design is superior for improving heat transfer in accident scenarios.     

Simplified theory for flow pattern prediction in perforated tubes

Perforated tubes are widely used in nuclear industries for many critical process and regulatory purpose. In pressurized heavy water reactors (PHWRs), perforated tubes are used in moderator system, in shutdown system for poison injection and in reactivity mechanism for housing the safety bank inside the core. In the moderator system and in poison injection loop the flow of fluid through perforated tubes match the requirement of the system. One is a steady state flow and the other is transient flow. The design of perforated tube involves arriving at the size of the holes, its spacing along the length and in the circumference. For PHWRs, the design is validated by rigorous experimentation. The paper offers a simplified analytical method along with detail theoretical steps required in arriving at the design of perforated tubes that can meet the end objective. For existing tubes, the given formulations can be used for estimating the flow profile. The paper presents case study of predicted flow through perforated tube and its validation by experimental results.     

Investigation on the interaction effect of two parallel axial through-wall cracks existing in steam generator tube

It is commonly required that steam generator tubes wall-thinned in excess of 40% should be plugged. However, the plugging criterion is known to be too conservative for some locations and types of defects and its application is confined to a single crack. In the previous study, the conservatism of the present plugging criterion of steam generator tubes was reviewed and a crack coalescence model applicable to steam generator tubes with two collinear axial through-wall cracks was proposed. Since parallel axial cracks are more frequently detected during in-service inspections than collinear axial cracks, the studies on parallel axial cracks spaced in circumferential direction are necessary. The objective of this paper is to investigate the interaction effect between two parallel axial through-wall cracks existing in a steam generator tube. Finite element analyses were performed and a new failure model of the steam generator tube with these types of cracks is suggested. Interaction effects between two adjacent cracks were investigated to explain the deformation behavior of cracked tubes.     

蒸汽发生器水室隔板的简化对下部组件等应力的影响

蒸汽发生器内部部件较为复杂,载荷作用下一次侧的力学行为通常会影响管板上部的部件,如支撑板和管束.通过对水室隔板的不同简化建立下部组件的轴对称和三维模型,考察对应力分布及支撑板和管束的影响.研究结果表明:轴对称模型能较为准确地得到应力强度分布,且较为保守.管板在不同模型下的位移差别较大,从而影响管板上部的支撑板以及管束的力学行为.     

The effect of tube rupture location on the consequences of multiple steam generator tube rupture event

A multiple steam generator tube rupture (MSGTR) event has never occurred in the commercial operation of nuclear reactors while single steam generator tube rupture (SGTR) events are reported to occur every 2 years. As there has been no occurrence of a MSGTR event, the understanding of transients and consequences of this event is very limited. In this study, a postulated MSGTR event in an advanced power reactor 1400 (APR1400) is analyzed using the thermal-hydraulic system code, MARS1.4. The APR1400 is a two-loop, 3893 MWt, PWR proposed to be built in 2010. The present study aims to understand the effects of rupture location in heat transfer tubes following a MSGTR event. The effects of five tube rupture locations are compared with each other. The comparison shows that the response of APR1400 allows the shortest time for operator action following a tube rupture in the vicinity of the hot-leg side tube sheet and allows the longest time following a tube rupture at the tube top. The MSSV lift time for rupture at the tube-top is evaluated as 24.5% larger than that for rupture at the hot-leg side tube sheet.