镍基合金环境致裂过程中裂尖微观力学特性分析（英文）

核电站-回路压力容器、管道及蒸汽发生器等设备和结构中广泛采用镍基合金和奥氏体不锈钢,而这些材料的环境致裂(EAC)却是核电结构的主要安全隐患之一。研究表明,核电高温高压水环境中镍基合金的EAC是裂尖氧化膜破裂和再生成的一个过程。为了深入了解镍基合金EAC裂纹扩展过程中裂尖的力学状况,从理论和数值模拟两方面分析研究了EAC裂尖氧化膜和基体金属区域的应力分布规律,为提高定量预测高温高压水环境中镍基合金EAC扩展速率精度奠定基础。     

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。     

膜致应力对应力腐蚀裂尖力学特性的影响（英文）

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(环境致裂)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。     

镍基合金应力腐蚀裂尖氧化膜力学特性分析

裂纹尖端氧化膜形成与破裂是核电站压力容器高温水环境中镍基合金材料应力腐蚀开裂(SCC)的主要过程之一。由于应力腐蚀裂纹尖端形貌和扩展方式的特殊性,本研究利用ABAQUS有限元软件的子模型技术,在微观尺度下对由裂尖氧化膜和基体金属共同构成的应力腐蚀裂尖应力应变场进行了分析。结果表明,SCC裂尖氧化膜前端沟形裂纹的存在,会造成氧化膜中应力和应变的很大变化,且随着沟形裂纹的长度增加,这种变化越加明显;另一方面,与氧化膜中应力相比,塑性应变对裂尖形貌变化更加敏感,从一个侧面说明,裂尖塑性应变是研究SCC裂尖氧化膜形成与破裂比较理想的力学参量。     

材料力学性能对高温水环境中镍基合金裂尖应力分布的影响

借助有限元数值模拟方法,分析了力学性能参数对镍基合金裂尖应力分布的影响。结果表明:随着材料力学性能的变化,镍基合金和氧化膜的裂尖应力分布会发生不同的变化,由此得到了镍基合金和氧化膜的裂尖应力分布规律。     

安全端焊接接头镍基合金600SCC裂纹扩展行为研究

镍基合金作为压水堆一回路安全端焊接接头焊缝的常用材料,由于严苛的服役环境以及焊缝处材料力学性能的不均匀使得镍基合金极易发生应力腐蚀开裂现象,对核电安全运行造成很大影响。为了解材料宏观结构参量变化(包括材料塑性性能以及应力强度因子K)对SCC裂纹扩展速率的变化,本文通过建立镍基合金600不同宏观结构参量下的SCC裂纹扩展有限元模型,分析了镍基合金600不同塑性以及载荷参数变化对裂尖塑性区和拉伸塑性应变的影响,结果表明塑性区尺寸及裂尖拉伸应变受到裂尖应力强度因子、屈服强度及硬化指数的影响,其中裂尖应力强度因子的影响较大,同时与屈服强度成反比,应力强度因子和硬化指数成正比;通过比较不同应力强度因子下计算所得SCC扩展速率结果和高温水环境下SCC扩展速率实验,获得了符合镍基合金600的特征距离r₀的取值范围;研究结果能为核电镍基合金600的高温水环境下SCC速率预测提供一定的科学依据。     

安全端焊接接头镍基合金600应力腐蚀裂纹的扩展行为（英文）

镍基合金作为压水堆一回路安全端焊接接头焊缝的常用材料,由于严苛的服役环境以及焊缝处材料力学性能的不均匀使得镍基合金极易发生应力腐蚀开裂(SCC)现象,对核电安全运行造成很大影响。为了解材料宏观结构参量变化(包括材料塑性性能以及应力强度因子K)对SCC裂纹扩展速率的变化,通过建立镍基合金600在不同宏观结构参量下的SCC裂纹扩展有限元模型,分析了镍基合金600不同塑性以及载荷参数变化对裂尖塑性区和拉伸塑性应变的影响。结果表明:塑性区尺寸及裂尖拉伸应变受到裂尖应力强度因子、屈服强度及硬化指数的影响,其中裂尖应力强度因子的影响较大,同时与屈服强度成反比,应力强度因子和硬化指数成正比;通过比较不同应力强度因子下计算所得SCC扩展速率结果和高温水环境下SCC扩展速率实验,获得了符合镍基合金600的特征距离r0的取值范围。研究结果能为核电镍基合金600在高温水环境下SCC速率预测提供一定的科学依据。     

核电结构材料应力腐蚀开裂初始阶段裂尖驱动力的研究（英文）

裂尖力学状态是影响核电结构材料应力腐蚀开裂(SCC)扩展速率的主要因素之一。为了研究SCC不同扩展阶段裂尖驱动力的变化及其对SCC扩展速率的影响,建立了SCC扩展不同阶段的有限元模型,详细分析了裂纹初始阶段影响裂尖应力状态的工作载荷、残余应力,以及氧化膜形成过程中产生的膜致应力。结果表明,在SCC裂纹初始阶段,裂尖氧化膜形成所产生的"锲入张力"是SCC的主要驱动力;随着裂纹的扩展,工作载荷和残余应力逐渐成为SCC裂纹扩展的主要驱动力。     

核电结构材料应力腐蚀开裂初始阶段裂尖驱动力的研究

裂尖力学状态是影响核电结构材料应力腐蚀开裂(SCC)扩展速率的主要因素之一。为了搞清SCC不同扩展阶段裂尖驱动力的变化及其对SCC扩展速率的影响,本文建立了SCC扩展不同阶段的有限元模型,详细分析了裂纹初始阶段影响裂尖应力状态的工作载荷、残余应力,以及氧化膜形成过程中产生的膜致应力。结果表明,在SCC裂纹初始阶段,裂尖氧化膜形成所产生的“锲入张力”是SCC的主要驱动力;随着裂纹的扩展,工作载荷和残余应力逐渐成为SCC裂纹扩展的主要驱动力。     

膜致应力对应力腐蚀裂纹尖端蠕变特性的影响

应力腐蚀开裂(SCC)是核电一回路安全端在服役过程中的一种重要失效形式。针对SCC裂尖基体金属在高温高压条件下发生的蠕变现象,利用ABAQUS模拟了在高温水环境下膜致应力对316L不锈钢应力腐蚀裂纹尖端蠕变特性的影响。结果表明:蠕变后,基体金属裂尖的应力、塑性变形、蠕变率和裂纹扩展率都会随之增大,进而使裂尖氧化膜破裂的倾向更加明显。     

Strain/Oxidation Interactions in Steels and Model Alloys

Strain/oxidation interactions in steels and model alloys The strains generated in oxide films on a number of materials of interest to the power generation industry are considered in relation to their operating conditions. Topics discussed the growth of nuclear canning alloys, oxide spalling, the formation of multilayer oxides during internal boiler corrosion and low ductility failures in austenitic stainless steels. The strains induced during thermal cycling are considered in detail and the temperature changes required to cause film craking are estimated and compared with observed spalling behaviour.     

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.     

The development of alumina-forming austenitic stainless steels for high-temperature structural use

A new family of alumina-forming austenitic stainless steels is under development at Oak Ridge National Laboratory for structural use in aggressive oxidizing environments at 600–900°C. Data obtained to date indicate the potential to achieve superior oxidation resistance compared to conventional Cr₂O₃-forming iron-and nickel-based heat-resistant alloys, with creep strength comparable to state-of-the-art advanced austenitic stainless steels. A preliminary assessment also indicated that the newly developed alloys are amenable to welding. Details of the alloy design approach and composition-microstructure-property relationships are presented. Author’s Note: Part of this research summary is based on a recent review paper (see Reference 1) and findings first reported in References 2–7.     

A new approach for DL-EPR testing of thermo-mechanically processed austenitic stainless steel

Standard test methods such as the electrochemical potentiokinetic reactivation test (EPR) and double-loop EPR test (DL-EPR) are commonly used to characterise sensitisation behaviour in austenitic stainless steels and nickel-based alloys. In this study, the DL-EPR test is augmented by large-area image analysis (IA) to characterise and quantify the networks of attacked grain boundaries. A new analysis approach that is based on a grain boundary cluster parameter is proposed to describe the network of corrosion susceptible grain boundaries, which may be estimated from electron backscatter diffraction (EBSD) data. This method may provide a better assessment of the relative DOS of different heats of austenitic stainless steels.     

Weldability and weld performance of candidate nickel base superalloys for advanced ultrasupercritical fossil power plants part I: fundamentals

Abstract

Fossil fuel will continue to be the major source of energy for the foreseeable future. To meet the demand for clean and affordable energy, an increase in the operating efficiency of fossil fired power plants is necessary. There are several initiatives worldwide to achieve efficiencies >45% higher heating value (HHV) through an increase in steam temperature (700 to 760°C) and pressure (27.6 to 34.5?MPa). Realising this goal requires materials with excellent creep rupture properties and corrosion resistance at elevated temperatures. In order to accomplish this, three classes of materials have been identified: creep strength enhanced ferritic steels, austenitic stainless steels and nickel base superalloys. Although new alloys have been designed and developed to meet this need, welding can have a significant and often detrimental effect on the required mechanical and corrosion resistant properties. Two previous papers addressed the welding and weldability of ferritic and austenitic stainless steels. Welding and weldability of nickel base alloys will be discussed in a two part paper. In this paper, the primary focus will be on the fundamentals of welding and weldability of Ni base superalloys.     

Korrosionsuntersuchungen an Reaktorwerkstoffen in strömender Natriumschmelze bei hohen Temperaturen

Research on corrosion of reactor materials in flowing sodium melt at high temperatures Corrosion tests on materials for fuel elements of a high-capacity sodium-cooled reacto, in sodium flowing at a rate of 0.5 metres per second at temperatures ranging from 500 to 600°C have been carried out in a circulation apparatur's of high-grade steel, containing some 80 litres of sodium melt. The electromagnetically circulated sodium is fine-cleaned internally, the purity being measured in a shunt circuit. The degree of sodium corrosion of high-grade austenitic steels and nickel-based alloys is relatively low and obeys different laws from those known from the reaction with vanadium-based alloys, which is greatly influenced by low oxide contents in the sodium. The results are in broad agreement with the corrosion rates measured in the USA., France and Britain.     

动态应变时效对核电材料环境致裂影响的研究现状与进展

综述了核级碳钢、低合金钢、不锈钢发生动态应变时效(DSA)的反常特征、影响因素及机制,讨论了DSA与高温高压水环境因素的交互作用对核电材料环境致裂的可能影响。指出了当前研究中存在的问题及进一步的研究方向。     

Weldability and weld performance of candidate nickel based superalloys for advanced ultrasupercritical fossil power plants Part II: weldability and cross-weld creep performance

Fossil fuel will continue to be the major source of energy for the foreseeable future. To meet the demand for clean and affordable energy, an increase in the operating efficiency of fossil fired power plants is necessary. There are several initiatives worldwide to achieve efficiencies >45%HHV (higher heating value) through an increase in steam temperature (700–760°C) and pressure (27.6–34.5?MPa). Realising this goal requires materials with excellent creep rupture properties and corrosion resistance at elevated temperatures. Two previous papers addressed the welding and weldability of ferritic and austenitic stainless steels. Welding and weldability of nickel based alloys will be discussed in a two-part paper. In this paper, the primary focus will be on the behaviour of candidate nickel based alloys that are being proposed in advanced ultrasupercritical power plants and with regard to weldability (Part I) and cross-weld creep performance (Part II).