核电结构材料应力腐蚀裂的研究现状与进展简

围绕应力腐蚀行为的实验研究方法、影响因素以及应力腐蚀机制的理论分析等几个方面综述了核电结构材料应力腐蚀研究的现状,讨论了研究中亟待解决的问题,指出了研究的发展方向与趋势。     

Corrosion behaviour of reactor coolant system materials in nuclear power plants

The materials used for the pressure‐retaining parts of reactor coolant system components in light water reactor nuclear power plants have to meet special requirements in terms of their mechanical properties, workability and in‐service performance. Corrosion issues play an important role in connection with plant operating conditions. While giving consideration to the specific service environment of the reactor whether a pressurized or boiling water reactor – the materials used for the individual components and the water chemistries employed in the various systems are selected such that metal loss due to general corrosion will remain very low. Thus the materials used in light water reactor plants exhibit a high general resistance to corrosion for their specified service conditions, material conditions and mechanical loads. However, under certain operating conditions other corrosion mechanisms may be found to induce damage. This paper uses data from the literature, published results of national and international research programs, information on damage which has actually occurred world‐wide and experience gained by Framatome ANP GmbH (former Siemens/KWU) in this field as a basis for discussing these mostly localised corrosion phenomena in terms of “classical” corrosion systems. Aspects associated with irradiation and its effects are not considered. Suitable remedial actions are, however, addressed wherever these are of relevance. The materials considered comprise unalloyed and low‐alloy steels, austenitic chromium‐nickel steels as well as high‐nickel steels and nickel‐base alloys which are exposed to the reactor coolant environment of boiling water reactor or pressurized water reactor plants, including materials investigated in corresponding water environments simulated in the laboratory.     

Corrosion resistance of SG tubing material Incoloy 800 mod. and Inconel 690 TT

The corrosion resistance of the SG tubing material Incoloy 800 mod. and Inconel 690 TT is considered with respect to pitting, sensitization characteristics and stress corrosion cracking, especially chloride-induced cracking, pure water cracking and caustic cracking. Based on laboratory data Incoloy 800 mod. as well as Inconel 690 TT show very good corrosion resistance under specified and faulted SG water conditions. The published results often refer to Incoloy 800 according to ASTM B 163–66. It is shown that the Incoloy 800 grade used by nuclear power suppliers is optimized for this application. The restricted chemical composition of the modified grade of Incoloy 800 vs. the standard grade is discussed and it is shown that this results in a significantly higher resistance to the above mentioned corrosion phenomena. The operating experience with Incoloy 800 mod. heat transfer tubes in nuclear steam generators is discussed briefly with respect to corrosion. No materrial-specific weaknesses have been observed with Incoloy 800 mod. Since more than 19 years of operation only at one tube out of 235 000 SG tubes made of Incoloy 800 mod. a leakage caused by corrosion has been occurred. This finding is independent of different tube manufacturers and different water treatments of the various plants.     

应力腐蚀机理和基于FEM的应力腐蚀裂纹扩展模拟技术的研究进展

本研究总结了应力腐蚀研究的5个经典理论:活性通路理论、钝化膜破坏理论、氢脆理论、腐蚀产物楔入模型和环境破裂三阶段理论。详细介绍了最为广泛接受的钝化膜破坏理论,在此基础上,从单一参数计算和应力腐蚀裂纹扩展模拟两个方面着重介绍了基于有限元方法对应力腐蚀裂纹扩展的研究方法和技术,单一参数计算可以满足获取应力腐蚀裂纹扩展预测模型关键参数,但是将裂纹尖端认为是一个点的假设存在不合理,应力腐蚀裂纹扩展模拟可以将载荷和环境因素综合考虑在内,但还仍然不能仿真出裂纹微观扩展现象。研究结合实践总结提出了应力腐蚀裂纹扩展模拟的工程技术方法,讨论了有限元技术在应力腐蚀研究方面不能实现微观物理过程、时间相关性等存在的问题,指出了基于应力腐蚀时间相关性的应力腐蚀裂纹扩展模拟研究是应力腐蚀破坏预防及预测研究工作的重点。     

划伤690TT合金在高温含氧水中应力腐蚀裂纹萌生的研究

利用划伤技术研究了690TT合金在325 ℃高温含氧硼锂水中的裂纹萌生和生长情况。试样表面和截面显微分析的结果表明,划伤沟槽底部局部萌生了典型的沿晶应力腐蚀裂纹。由于应力集中,在慢速率拉伸阶段划伤沟槽底部产生了机械裂纹,而机械裂纹成为恒载过程中690TT合金沿晶应力腐蚀裂纹萌生和生长的先导。尖端非常接近晶界或者沿着晶界的机械裂纹可继续形成沿晶应力腐蚀裂纹。690TT合金在恒载荷条件下对应力腐蚀开裂仍有一定的敏感性。     

高钼镍基合金在熔盐中脱溶腐蚀的研究

镍基合金、特别是Hastelloy N(16Mo—7Cr—5Fe—Ni)为美国发展熔盐反应堆研制的一种耐蚀材料。大量文献是针对在高纯熔融氟化物燃料盐系中的试验结果,并证实一般主要发生的是选择性脱铬反应。 本试验(静态)是在普通(未经特殊净化处理)熔融氟化物盐或硝酸盐中进行。通过金属物理等手段,详细研究了表面腐蚀层的结构、钼对晶界脱溶腐蚀的影响及应力腐蚀破裂的过程。表明除铬元素外,也存在钼元素的选择性脱溶腐蚀,并构成上述腐蚀现象的主要原因。本文着重说明了钼的双重作用:一方面随着合金中钼含量的提高,明显改变了钼、铬元素沿晶粒边界和通过晶粒内部脱溶腐蚀的相对速度,从而抑制了在高温氟盐中出现晶间腐蚀;另一方面钼、铬元素的脱溶反应造成表面腐蚀层基本上是一种疏松多孔的纯镍层。在熔融硝酸盐中,尽管Hastelloy N合金的一般腐蚀速度极低,但发现在张应力的作用下加速了局部晶界的脱溶腐蚀;而钼、铬元素沿晶界的选择性脱溶反过来又促进了裂纹的形成和扩展,引起晶间应力腐蚀破裂。     

高温高压水环境中腐蚀产物膜的研究现状

高温高压水是轻水核反应堆冷却系统的主要服役环境,反应堆压力容器、管道及蒸汽发生器等构件材料在高温水中的环境损伤是影响核电安全的重要因素.材料在高温水中形成的腐蚀产物膜是影响其服役稳定与环境失效的关键.本文介绍了高温高压水环境中不锈钢和镍基合金腐蚀产物膜的形貌、结构、影响因素及形成机制,并对当前研究中存在的主要问题进行了讨论.     

Effect of Hydrogen on Corrosion and Stress Corrosion Cracking of AZ91 Alloy in Aqueous Solutions

The effect of hydrogen on the corrosion and stress corrosion cracking of the magnesium AZ91 alloy has been investigated in aqueous solutions.Hydrogen produced by corrosion in water diffuses into,and reacts with the Mg matrix to form hydride.Some of the hydrogen accumulates at hydride/Mg matrix(or secondary phase) interfaces as a consequence of slow hydride formation and the incompatibility of the hydride with the Mg matrix(or secondary phase),and combines to form molecular hydrogen.This leads to the development of a local pressure at the hydride/Mg matrix(or secondary phase)interface.The expansion stress caused by hydride formation and the local hydrogen pressure due to its accumulation result in brittle fracture of hydride.These two combined effects promote both the corrosion rate of the AZ91 alloy,and crack initiation and propagation even in the absence of an external load.Hydrogen absorption leads to a dramatic deterioration in the mechanical properties of the AZ91 alloy,indicating that hydrogen embrittlement is responsible for transgulanar stress corrosion cracking in aqueous solutions.     

不锈钢应力腐蚀开裂综述

应力腐蚀开裂一直以来是不锈钢领域的重要研究课题,也是许多行业亟需解决的工程问题。应力腐蚀开裂是材料、环境和应力三者相互作用的结果,由于其复杂性,目前人们对不锈钢发生应力腐蚀开裂的机理尚存在许多不同的见解,但是经过近一个世纪的研究,从材料选择、环境控制等方面入手,预防不锈钢发生应力腐蚀是能够达到的。综述了应力腐蚀开裂的特征、机理和三个影响因素(应力、材料和环境)。对应力腐蚀的阳极溶解机理和氢致开裂机理进行了概述,阐述并探讨了不锈钢应力腐蚀开裂的滑移溶解机理、氧化膜开裂机理以及氢致开裂机理。归纳了组织结构对不锈钢应力腐蚀的影响,分析了材料成分如(Ni、Mo和N)的添加与应力腐蚀敏感性的关系,总结了环境因素在应力腐蚀中的作用,对特定介质中不锈钢的应力腐蚀规律进行了归纳,并探讨了温度变化对不锈钢应力腐蚀的影响。介绍了近年来关于控制不锈钢应力腐蚀开裂方法的研究进展,如晶界工程、细化晶粒以及涂层等。最后展望了不锈钢应力腐蚀开裂未来的研究方向。     

690TT合金在特殊环境下的腐蚀问题

在广泛的电厂运行及实验室研究基础上,690TT合金被证明是目前最佳的蒸汽发生器（SGs）管材之一,690TT合金的使用有效地提高了PWR蒸汽发生器的可靠性,因而成为在役第二代核电站中最常用的传热管管材,并将大量应用于第三代商用核电厂。然而在水质恶化以及随服役时间的增加,690TT合金不可避免的也会遭遇腐蚀。本文对690TT合金使用安全性造成潜在威胁的脱合金成分腐蚀（Cr贫化）、铅致应力腐蚀破裂（PbSCC）、低价硫应力腐蚀开裂（Sy-SCC）的情况加以较详细的介绍。     

Analysis of failed nuclear plant components

Argonne National Laboratory has conducted analyses of failed components from nuclear power- gener-ating stations since 1974. The considerations involved in working with and analyzing radioactive compo-nents are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in serv-ice. The failures discussed are (1) intergranular stress- corrosion cracking of core spray injection piping in a boiling water reactor, (2) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressurized water reactor, (3) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (4) failure of pump seal wear rings by nickel leaching in a boiling water reactor. Work supported by Commonwealth Edison Company under ACK 85026.     

海水泄漏对SG二次侧金属腐蚀的实验分析

通过给水中掺入一定比例的厂区海水,采用静态高压釜真实模拟凝汽器海水泄漏后的SG二次侧水质,研究了不同海水污染程度下SG二次侧内部构件金属材料的均匀腐蚀、点蚀和应力腐蚀性能.海水泄漏加剧了SG内碳钢和低碳钢部件的腐蚀,抑制了马氏体、奥氏体不锈钢钝化膜的生长,加剧了镍基合金表面氧化物的沉积,为核电厂治理凝汽器海水泄漏事件的...     

膜致应力对应力腐蚀裂尖力学特性的影响

氧化膜破裂理论是目前定量预测核电高温水环境中镍基合金应力腐蚀开裂速率应用最为广泛的理论模型之一,其中应力强度因子是衡量应力腐蚀开裂速率的重要参量。为进一步了解氧化膜破裂机理及裂纹扩展驱动力特性,提出了膜致应力强度因子。为了深入了解膜致应力强度因子在 EAC 裂纹扩展过程中裂尖的力学状况,在不考虑外载的情况下,从理论和数值模拟两方面分析研究了EAC 裂尖基体金属区域的应力应变分布状态,得出了膜致应力强度因子对裂尖Mises应力、等效塑性应变、拉伸应力、拉伸应变及拉伸应变梯度的影响规律,为提高定量预测高温高压水环境中镍基合金及不锈钢 EAC 扩展速率精度奠定基础,进而完善了氧化膜破裂机理。     

Effects of Stress Intensity Factor on Electrochemical Corrosion Potential at Crack Tip of Nickel-Based Alloys in High Temperature Water Environments

在高温水环境中,应力会提高镍基合金裂纹尖端的阳极溶解速率并加速裂纹扩展。采用弹塑性有限元方法,对高温水环境中镍基合金裂纹尖端应力和电化学腐蚀的关系进行研究。分析了应力强度因子对模拟高温水环境中600合金1T-CT试样裂纹尖端表面电化学腐蚀电位的影响,并讨论了弹性变形和塑性变形对裂纹尖端电化学腐蚀电位变化的影响。     

Stress Corrosion Cracking of Turbine Disc and Rotor Alloys

Abstract

Investigations have been carried out to examine the stress corrosion behaviour of some medium strength low alloy steels in high purity wet steam and water. The alloys involved are used for the manufacture of steam turbine discs and rotors.

Stress corrosion cracking occurred in both 3% CrMo and 31/2% NiCrMoV steels. Crack growth rates were measured, after exposure in on-site rigs, for periods up to 20,000 h. The effects of applied stress, stress intensity, and strength level were studied in addition to microstructure. The possible role of certain non-metallic incluszons upon the cracking process is highlighted, as is the presence of chromium.     

Contribution of research to the understanding and mitigation of environmentally assisted cracking in structural components of light water reactors

Environmentally assisted cracking (EAC) is a potential threat to the safety and integrity of water-wetted components in operating water-cooled nuclear power plants. Two forms of EAC are commonly distinguished, depending on the form of loading contributing to damage: stress corrosion cracking (SCC) and corrosion fatigue. A number of instances of in-service degradation due to EAC have occurred in operating plants worldwide, often leading to unplanned plant outages. Understanding the causes of EAC is essential to minimise the loss of plant availability due to its occurrence and to avoid the possibility of catastrophic failure, for example, if a crack grew to a critical size in a major pressure boundary component. This paper will describe some examples of these phenomena in the main materials of construction of pressure boundary and other critical components in pressurised and boiling water reactors (BWRs). Over the last several decades, substantial research programmes have been carried out in a number of laboratories worldwide, aimed at further understanding of the processes leading to EAC to manage occurrences in plant and minimise future failures. Selected areas of research on EAC in light water reactor environments are discussed. Corrosion fatigue in low-alloy pressure vessel steels was the subject of considerable attention in the 1980s and early 1990s because of its potential threat to pressure vessel integrity and the publication of data, suggesting that there is a major influence of environment on fatigue crack growth in some laboratory tests. The author’s research provided insight into the conditions under which the major environmental effects occur and contributed to the development of an ASME Code Case for pressurised water reactor (PWR) conditions which provided a means of screening based on steel sulphur content and loading conditions. More recently, the research focus in this area has moved to austenitic stainless steels, again providing support to Code Case development and furthering mechanistic understanding. A recent review of knowledge gaps for EPRI provides a basis for future research on environmentally assisted fatigue and will inform the development of new assessment methodologies. A key area of the current study concerns differences in loading conditions between specimens in laboratory tests and plant components subject to transient loading. In the case of SCC, stainless steels have shown the greatest propensity to cracking in BWRs, while Alloy 600 has been a major cause of in-service failures in PWRs, both on the primary side, as recognised by Coriou in the early 1960s, and in secondary environments where a number of different corrosion-related failure processes have been identified. High-strength alloys, such as Alloy X-750 used for fastener applications, have also caused failures in both reactor types. For austenitic materials, SCC susceptibility is enhanced by irradiation, resulting in failures in core internals components. Ferritic stainless steels also undergo SCC under some specific circumstances but are generally more resistant than the lower chromium austenitic materials.     

Al-Zn-Cu-Mg及Al-Cu-Mg在氯化物水溶液中应力腐蚀开裂的扫描电镜断口形貌

本文研究了Al—Zn—Cu—Mg及Al—Mg—Cu在氯化物水溶液中用慢应变速率方法得到的应力腐蚀开裂的扫描电镜断口形貌。结果发现两种铝合金在氯化物水溶液中的应力腐蚀开裂断口复盖着—层腐蚀产物,这层腐蚀产物可以呈现不同的扫描电镜断口形貌,如类似解理的“平滑”形貌,“龟裂泥巴状”形貌等等。遮盖了应力腐蚀开裂固有的真实形貌;应力腐蚀开裂固有的真实形貌是相同的,呈典型的沿晶开裂形貌,盖有腐蚀产物层的形貌则取决于介质pH和电位。因此,应该根据铝合金应力腐蚀开裂固有的真实扫描电镜断口形貌来鉴别铝合金的应力腐蚀。将断口进行化学去膜处理可以得到铝合金应力腐蚀开裂的固有的真实的扫描电镜断口形貌。     

The effect of prior deformation on stress corrosion cracking growth rates of Alloy 600 materials in a simulated pressurized water reactor primary water

The effect of prior deformation on stress corrosion cracking (SCC) growth rates of Alloy 600 materials in a simulated pressurized water reactor primary water environment is studied. The prior deformation was introduced by welding procedure or by cold working. Values of Vickers hardness in the Alloy 600 weld heat-affected zone (HAZ) and in the cold worked (CW) Alloy 600 materials are higher than that in the base metal. The significantly hardened area in the HAZ is within a distance of about 2-3 mm away from the fusion line. Electron backscatter diffraction (EPSD) results show significant amounts of plastic strain in the Alloy 600 HAZ and in the cold worked Alloy 600 materials. Stress corrosion cracking growth rate tests were performed in a simulated pressurized water reactor primary water environment. Extensive intergranular stress corrosion cracking (IGSCC) was found in the Alloy 600 HAZ, 8% and 20% CW Alloy 600 specimens. The crack growth rate in the Alloy 600 HAZ is close to that in the 8% CW base metal, which is significantly lower than that in the 20% CW base metal, but much higher than that in the as-received base metal. Mixed intergranular and transgranular SCC was found in the 40% CW Alloy 600 specimen. The crack growth rate in the 40% CW Alloy 600 was lower than that in the 20% CW Alloy 600. The effect of hardening on crack growth rate can be related to the crack tip mechanics, the sub-microstructure (or subdivision of grain) after cross-rolling, and their interactions with the oxidation kinetics.     

Corrosion control in electric power plants—success stories

Corrosion-related failures in electric power plants often stimulate research aimed at understanding the cause of the failures and at developing cost-effective corrosion control options. This paper describes some of the success stories. Over the last few years, EPRI has helped solve several aqueous corrosion problems using the following approaches: materials selection, coatings, water chemistry control, residual stress modification and corrosion monitoring. These approaches were applied successfully to flue gas desulfurization systems, low pressure steam turbines and boiling water reactor piping. A measure of the success of these corrosion control options is the cost savings of over $570 million documented by only seven of EPRI's over 700 utility members after applying these techniques.     

Corrosion susceptibility of peened friction stir welded 7075 aluminum alloy joints

Effects of surface treatment techniques like laser and shot peening on stress corrosion cracking (SCC) susceptibility of friction stir welded (FSW) 7075 aluminum alloy joints were investigated. This study had two parts; the first part investigated the peening effects on stress corrosion cracking susceptibility in FSW samples by slow strain rate testing in a 3.5% NaCl solution. The second part of the study investigated the effects of peening on corrosion while submerged in a 3.5% NaCl solution with no external loads applied. No signs of corrosion pitting or SCC were evident on any of the tensile samples during the slow strain rate testing. The FSW plates exposed in 3.5% NaCl solution for 60 days were inspected periodically for signs of corrosion and stress corrosion cracking in the areas expected to have residual stresses due to welding. Pitting corrosion was seen on the samples, but even after 60 day exposure no stress corrosion cracking was detected on any of the peened or unpeened samples.