蒸汽发生器传热管腐蚀监测技术研究

蒸汽发生器传热管的腐蚀监测是保障核动力装置安全运行的重要问题之一.在对声发射技术原理进行介绍的基础上,采集和分析利用声发射仪对传热管进行加压实验时的声发射信号.实验结果表明:在传热管上形成微小直径穿孔性管壁腐蚀点,这种点蚀形成后,腐蚀先是向深处发展,形成尖端腐蚀,然后再向两侧延伸,逐渐形成裂纹,随着裂纹的进一步发展便形成泄漏事故;利用声发射仪对蒸汽发生器的传热管工作过程进行实时监控,可实时判断传热管材料的应力腐蚀情况,操作人员可根据声发射信号强度及其变化控制核动力装置的运行情况.     

蒸汽发生器传热管失效分析及防护措施

蒸汽发生器是核动力装置中一个非常重要的热量交换设备,也是整个核动力装置中的薄弱环节。在核动力装置的停堆事故中,有一半以上是由于蒸汽发生器破损引起的,严重影响到整个核动力装置的安全性和可靠性。如何使传热管不过早破损,延长蒸汽发生器的使用寿命是当前面临的重要课题之一。文章通过对蒸汽发生器传热管所用奥氏体不锈钢材料的分析,结合国内外相关事故分析报告,确定了传热管破损的类型为穿晶型应力腐蚀破坏。通过对腐蚀产物、工作环境以及管束应力的分析,找出蒸汽发生器传热管发生应力腐蚀开裂的影响因素,同时提出相应的防护措施。     

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。     

蒸汽发生器传热管的降质及对其完整性的评估

蒸汽发生器传热管是反应堆冷却剂压力边界的主要组成部分，这就意味着必须保持传热管的完整性。然而，运行经验表明，蒸汽发生器传热管会出现各种降质。这些降质可能会导致管子的泄漏或破裂，使反应堆冷却剂丧失，并提供了直接通向二回路和释放到环境中去的途径。本文将介绍几种已知的传热管降质，传热管完整性性能准则．并对蒸汽发生器传热管完整性进行评估。     

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。     

压水堆活化腐蚀产物源项~(58)Co和~(60)Co敏感性研究

以某压水堆核电厂为例,采用CORA程序分析压水堆核电厂一回路材料组成、蒸汽发生器传热管材料钴含量、冷却剂氢氧化锂浓度、净化效率和反应堆运行功率等因素变化对一回路腐蚀产物58Co和60Co活度浓度的影响。计算结果表明:通过限制蒸汽发生器传热管材料中钴元素的含量、提高冷却剂中氢氧化锂浓度、提高冷却剂净化效率和降低功率等措施可以有效降低活化腐蚀产物的活度浓度,为压水堆核电厂辐射剂量控制提供参考。     

压力诱发SGTR的概率计算方法探讨

诱发蒸汽发生器传热管破裂现象对核电厂堆芯损伤和放射性早期大量释放风险有非常大的影响。准确地对诱发蒸汽发生器传热管断裂概率进行计算和定值对正确认知核电厂的核安全风险非常重要。文章调研了已有压力诱发蒸汽发生器传热管破裂概率的取值计算方法,通过对不同取值计算方法的对比分析,结合国内实际情况,提出了一套较为合理可行的压力诱发蒸汽发生器传热管破裂概率的计算方法。文章推荐的诱发蒸汽发生器传热管破裂数据采集和分析计算方法为后续国内核电厂概率安全分析应用和安全监管提供了参考。     

基于时间序列神经网络的蒸汽发生器传热管泄漏程度诊断研究

针对蒸汽发生器U形传热管泄漏，本文提出了一种基于时间序列神经网络对蒸汽发生器传热管泄漏程度进行诊断研究的方法。首先，对核电厂蒸汽发生器U型传热管泄漏进行机理分析，构建其数学模型，提取其泄漏的直接特征参数，再依据Fisher得分法，提取其间接特征参数；其次，通过滑动时间窗口法从预处理后的时间序列数据中生成数据样本，作为时间序列神经网络的输入，并以蒸汽发生器U形传热管泄漏程度信息为标注，基于反向传播（BP）算法对五层神经网络系统进行训练，得到蒸汽发生器U形传热管泄漏的时间序列神经网络模型；最后，模拟核电厂运行过程蒸汽发生器U形传热管泄漏时的时间序列测试数据。仿真结果表明，时间序列神经网络对演变事件的处理具有较好的有效性和较高的泛化能力，对故障程度的诊断研究具有参考价值。      

蒸汽发生器传热管结垢厚度的涡流检测方法与应用

蒸汽发生器二回路中有较多的沉积物存在并危害传热管安全,利用涡流检测方法可以对传热管二次侧泥渣进行有效检测。通过模拟传热管结垢的不同厚度并进行实验,可获得厚度与幅值的对应关系。本文描述了对蒸汽发生器传热管结垢的检测方法及幅值与厚度的对应关系,为统计蒸汽发生器传热管外壁结垢情况提供了较为有效的参考基准量。     

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。     

Behaviour of the steam generator tubing in water with different pH values

Steam generators are crucial components of pressurized water reactors. The failure of the steam generator as a result of tube degradation by corrosion has been a major cause of Pressurized Heavy Water Reactor (PHWR) plant unavailability. Steam generator problems have caused major economic losses in terms of lost electricity production through forced unit outages and, in cases of extreme damage, as additional direct cost for large-scale repair or replacement of steam generators.The excellent performance to date of CANDU steam generators can be attributed, in part, to their design and performance characteristics, which typically involve higher recirculation ratios and lower heat fluxes and temperatures. However, the steam generator tubes are susceptible to failure by a variety of mechanisms, the vast majority of which are related to corrosion.The generalized corrosion is an undesirable process because it is accompanied by deposition of the corrosion products which affect the steam generator performances. It is very important to understand the generalized corrosion mechanism with the purpose of evaluating the quantities of corrosion products which exist in the steam generator after a determined period of operation (IAEA, 1997).The purpose of the experimental research consists in the assessment of corrosion behaviour of the tubes material, Incoloy-800, at normal secondary circuit parameters (temperature—260 °C, pressure—5.1 MPa). The testing environment was the demineralised water without impurities, at different pH values regulated with morpholine and cyclohexylamine (all volatile treatment—AVT).The results are presented like micrographics and graphics representing weight loss of metal due to corrosion, corrosion rate, total corrosion products formed, the adherent corrosion products, released corrosion products, release rate of corrosion products and release rate of the metal.     

FRAMATOME's continous efforts to improve steam generator corrosion resistance

The steam generators of PWR nuclear reactors are among the primary components most affected by corrosion problems. Corrosion of the steam generator tubes, which assure heat transfer between the primary and secondary circuits, have been observed on a large number of operating steam generators, especially in the United States. According to an NRC survey, as of November 1981, forty PWR units with steam generators of the recirculation type were in operation in the US. Of these, 32 have been found to have one or more forms of tube degradation.Construction of the French PWR nuclear program started in the early 70s, at the time a number of operating plants in the US were being affected by the first corrosion problems. Since, at that time, its construction program was in an early stage, FRAMATOME was able to make modifications on the first units to improve steam generator resistance to corrosion. For instance, full depth expansion of the tubes in the tube-sheet using an explosive process (Westex) was performed on Fessenheim 1 steam generators already installed on site. Later on, continuous operating experience was being obtained in the US, before startup of the French units. This allowed FRAMATOME to react rapidly and take immediate corrective actions at the design stage, during fabrication and sometimes even on site in order to mitigate the risk of corrosion in the steam generators.FRAMTOME is confident that the present design of its steam generator models, including a large number of major improvements is adequate to prevent major corrosion problems to occur during operation. However, the company has embarked on an important development program to further improve the corrosion resistance and thereby the reliability of its steam generators. This program includes studies on new tube expansion techniques, alternate materials for steam generator tubes (Inconel 690), improved tube inspection methods, local thermohydraulic flow, tube vibrations, etc.     

Evaluation of the burst characteristics for axial notches on SG tubings

Some events of steam generator tubes have been reported in some nuclear power plants around the world. Main causes of the leakage are from various types of corrosion in the steam generator (SG) tubing. Primary water stress corrosion cracking (PWSCC) of steam generator tubing have occurred in many tubes in Korean plants, and they were repaired using sleeves or plugs. In order to develop proper repair criteria, it is necessary to ascertain the leak behavior of the tubings. A high-pressure leak and burst testing system was manufactured. Various types of electro-discharged-machined (EDM) notches having different lengths were machined on the o.d. of test tubes to study SG tube behavior. Leak rate and ligament rupture pressure as well as the burst pressure were measured for the tubes at room temperature. Rupture pressure of the part through-wall defect tubes depends on the defect depth and length. Water flow rates after the rupture were independent of the flaw types; tubes having 20–60 mm long EDM notches showed similar flow rates regardless of the initial defect depth. A fast pressurization rate generated a lower burst pressure than the case of a slow pressurization.     

蒸汽发生器里的泥渣收集器

泥渣收集器能够有效地捕集悬浮于再循环水中的粒子，大大减轻传热管表面结垢和管板上泥渣的沉积，从而减少管子腐蚀的可能性，并有助于改善蒸汽发生器水位的稳定性和二次侧的水化学控制。泥渣收集器还可以加装在已投入运行的蒸汽发生器上。本文介绍了国外对泥渣收集器的研究和其巨大经济效益，提出了国内自主开发泥渣收集器的初步设想。     

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.     

小型模块式反应堆螺旋管蒸汽发生器设计和热工水力分析

近年来，国际上一体化小型模块式反应堆发展飞速，我国也正在加速研制一体化小型模块式反应堆。本文针对15 MW的一体化小型模块式反应堆，设计一种螺旋管式蒸汽发生器，共12个蒸汽发生器组件均匀分布在反应堆堆芯围板外侧和压力容器内侧壁的环形空间中，每个组件含5层、25根螺旋管，整个蒸汽发生器共300根螺旋管。给出了蒸汽发生器的具体参数，分析了蒸汽发生器组件中换热系数、温度、温差和热流密度等沿管长的变化，并给出了螺旋管内流体的动力特性曲线。     

PWR核电站蒸汽发生器传热管和主管道的应力腐蚀破裂研究

用慢应变速率试验(SSRT)、恒载荷试验(CLT)和低周循环载荷试验方法研究以秦山和大亚湾核电站安全为目的的有关压力边界管道破裂始发事件应力腐蚀破裂(SCC)的行为,为评价管道的结构完整性提供支持性实验数据。研究的材料有核等级主管道焊接热影响区(WHAZ)316不锈钢(SS),核等级蒸汽发生器(SG)传热管材Incoloy-800、Inconel-600、Inconel-690和321SS。研究的影响因素包括材料冶金、表面喷丸处理、载荷、应变速率、循环载荷以及水化学条件对SCC的影响规律。     

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.     

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。     

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。