Alloy 600(合金 600)研究综述
Alloy 600 合金 600 - The aim of the present work is to study the effect of grain boundary structure on the precipitation of chromium carbides in Alloy 600. [1] In this work, tungsten inert gas (TIG) welded joints of Alloy 600 were treated by LSP to enhance the mechanical properties. [2] Alloy 600 (equivalent to Inconel 600) has excellent corrosion resistance and is often used as a welding material in welded joints, but material properties of the alloy are heterogeneous in the welded zone due to the complex welding process. [3] The materials used previously for SG tubes around the world have been replaced and will be replaced by Alloy 690 given its improved corrosion resistance relative to that of Alloy 600. [4] Prospective Fe and Ni-base alloys, AISI 441, AISI 444, a FeCrAl alloy A197/Kanthal® EF101, alloy 600, and alloy 800H are investigated for their suitability to BOP components. [5] A study on UNSM among various techniques of adding compressive residual stress to Alloy 600 was conducted. [6] Fretting wear behaviors of Alloy 600MA in dry air and deionized water conditions were investigated. [7] Recently, our research team reported the effect of UNSM on the corrosion resistance of stainless steel and Alloy 600. [8] This study investigated the effect of temperature and time on the sensitization behavior of alloy 600. [9] The model versatility was also shown by simulating IGSCC in Alloy 600, also with good agreements. [10] In this work the crystallographic texture (deformation and annealing) of the nickel superalloy, Inconel 600 (alloy 600), was analyzed after ECAP processing, Equal Channel Angular Pressing, and subsequent heat treatments. [11] The electrochemical reactivity of Alloy 600 in the passive state was examined at the microstructural scale using a local-probe technique, the Scanning Electrochemical Microscopy (SECM). [12] The alloys 82 and 182 have been widely used as filler metal to join austenitic stainless steel with alloy 600 by a shielded metal arc welding process in the reactor pressure vessel and pressure vessel nozzles, which are both components in boiling water reactors. [13]本工作的目的是研究晶界结构对合金 600 中碳化铬析出的影响。 [1] 在这项工作中，合金 600 的钨极惰性气体 (TIG) 焊接接头通过 LSP 处理以提高机械性能。 [2] 合金600（相当于Inconel 600）具有优良的耐腐蚀性能，常被用作焊接接头中的焊接材料，但由于焊接工艺复杂，合金在焊接区的材料性能参差不齐。 [3] 以前用于全球 SG 管的材料已被替换，并将被 Alloy 690 取代，因为它相对于 Alloy 600 具有更高的耐腐蚀性。 [4] 研究了潜在的 Fe 和 Ni 基合金、AISI 441、AISI 444、FeCrAl 合金 A197/Kanthal® EF101、合金 600 和合金 800H 对防喷器部件的适用性。 [5] 对向合金 600 添加压缩残余应力的各种技术中的 UNSM 进行了研究。 [6] 研究了合金 600MA 在干燥空气和去离子水条件下的微动磨损行为。 [7] 最近，我们的研究团队报道了UNSM对不锈钢和合金600的耐腐蚀性能的影响。 [8] 本研究调查了温度和时间对合金 600 敏化行为的影响。 [9] 通过在合金 600 中模拟 IGSCC 也显示了模型的多功能性，也具有良好的一致性。 [10] 在这项工作中，镍超合金 Inconel 600（合金 600）的晶体织构（变形和退火）在 ECAP 加工、等通道角压和随后的热处理后进行了分析。 [11] 使用局部探针技术扫描电化学显微镜 (SECM) 在微观结构尺度上检查合金 600 在钝态状态下的电化学反应性。 [12] 合金 82 和 182 已被广泛用作填充金属，通过保护金属电弧焊工艺在反应堆压力容器和压力容器喷嘴中将奥氏体不锈钢与合金 600 连接起来，它们都是沸水反应堆中的组件。 [13]
stress corrosion cracking 应力腐蚀开裂
This paper provides a summary and analysis on the initiation of stress -corrosion cracking (SCC) of Alloy 600 (and related alloys) and stainless steels in nuclear reactor environments near 300 °C. [1] The stress corrosion cracking behavior of alloy 600 was studied in high temperature water at 288°C to 360°C. [2] Stress corrosion cracking (SCC) is one of serious aging degradation problems for the Alloy 600 components of pressurized water reactors (PWRs). [3] Historical views on stress corrosion cracking (SCC) of nickel-based alloys in light water nuclear power plants include the first cracks observed on Alloy 600 in pure water at 350°C by Henri Coriou. [4] The general corrosion and stress corrosion cracking behaviors of the nickel-base Alloy 600 in the primary high temperature aqueous coolants of light water cooled and moderated nuclear power reactors are reviewed. [5] The purpose of this work is to verify the effect of magnetite coupled with Alloy 600 on stress corrosion cracking (SCC) behavior in water containing 100 ppm PbO at 315 °C. [6] Stress corrosion cracking of Alloy 600 (A600) and 316 stainless steel (316SS) exposed to simulated pressurized water reactor primary water at temperatures of 320–360 °C has been investigated and compared. [7] This study aims to investigate and explain the magnetite-accelerated stress corrosion cracking phenomenon of Alloy 600 under caustic conditions, based on the electrochemical behavior. [8] The present work investigated the stress corrosion cracking (SCC) initiation behavior of Alloy 600 during the transition of hydrogenated/oxygenated water condition at evaluated temperature. [9] The mitigating effect introduced by intergranular Cr carbides on the stress corrosion cracking propagation of a cold-worked Alloy 600 has been firstly examined through high-resolution 3-dimensional (3D) sequential sectioning. [10]本文总结和分析了合金 600（及相关合金）和不锈钢在 300°C 附近的核反应堆环境中引发应力腐蚀开裂 (SCC)。 [1] 研究了合金 600 在 288°C 至 360°C 的高温水中的应力腐蚀开裂行为。 [2] nan [3] 关于轻水核电站中镍基合金应力腐蚀开裂 (SCC) 的历史观点包括 Henri Coriou 在 350°C 纯水中观察到的合金 600 上的第一个裂纹。 [4] 综述了镍基合金600在轻水冷慢化核动力堆一次高温水冷却剂中的一般腐蚀和应力腐蚀开裂行为。 [5] nan [6] nan [7] nan [8] nan [9] 首先通过高分辨率 3 维 (3D) 连续切片检查了晶间 Cr 碳化物对冷加工合金 600 应力腐蚀开裂扩展的缓解作用。 [10]
pressurized water reactor 压水反应堆
Alloy 600 is the material of choice in steam generator tubes in pressurized water reactors (PWR). [1] The present work will analyze the metallurgical phenomena involved in the application of ECAP in the Inconel 600 nickel superalloy (alloy 600), composed of 72% nickel, 14 ± 17% chromium and 6-10% iron, which is used in the nuclear industry as material of the Pressurized Water Reactors (PWR) steam generator primary circuit tubes, or Pressurized Water Reactors. [2] Intergranular oxidation occurred in a cold-worked Alloy 600 after exposure to pressurized water reactor primary water. [3] The crack initiation on a cold-worked surface of Alloy 600, exposed to simulated pressurized water reactor primary water, was mechanistically studied through high-resolution characterization. [4]合金 600 是压水反应堆 (PWR) 中蒸汽发生器管的首选材料。 [1] 目前的工作将分析 ECAP 在 Inconel 600 镍高温合金（合金 600）中的应用所涉及的冶金现象，该合金由 72% 的镍、14 ± 17% 的铬和 6-10% 的铁组成，用于核工业作为压水反应堆 (PWR) 蒸汽发生器主回路管或压水反应堆的材料。 [2] 冷加工合金 600 在暴露于压水反应堆一次水后发生晶间氧化。 [3] nan [4]
Base Alloy 600
The general corrosion and stress corrosion cracking behaviors of the nickel-base Alloy 600 in the primary high temperature aqueous coolants of light water cooled and moderated nuclear power reactors are reviewed. [1] A new multi-principal-component alloy (MPCA) filler metal with the composition Fe5Co20Ni20Mn35Cu20 was designed for brazing Ni-base Alloy 600 (Ni-Cr-Fe). [2]综述了镍基合金600在轻水冷慢化核动力堆一次高温水冷却剂中的一般腐蚀和应力腐蚀开裂行为。 [1] 一种新的多主成分合金 (MPCA) 填充金属，其成分为 Fe5Co20Ni20Mn35Cu20，用于钎焊镍基合金 600 (Ni-Cr-Fe)。 [2]
Worked Alloy 600
Intergranular oxidation occurred in a cold-worked Alloy 600 after exposure to pressurized water reactor primary water. [1] The mitigating effect introduced by intergranular Cr carbides on the stress corrosion cracking propagation of a cold-worked Alloy 600 has been firstly examined through high-resolution 3-dimensional (3D) sequential sectioning. [2]冷加工合金 600 在暴露于压水反应堆一次水后发生晶间氧化。 [1] 首先通过高分辨率 3 维 (3D) 连续切片检查了晶间 Cr 碳化物对冷加工合金 600 应力腐蚀开裂扩展的缓解作用。 [2]
alloy 600 exposed
In this work, a systematic evaluation of Alloy 600 exposed in low pressure H 2 -steam over a range of oxidizing potentials in the vicinity of the Ni/NiO transition has revealed a notable decrease in PIO for conditions more oxidizing than the Ni/NiO transition, whilst local diffusion-induced grain boundary migration occurs irrespective of the oxidizing potential. [1] Lastly, the paper introduces a new technique for measuring grain boundary strength, demonstrated on an oxidized grain boundary of Alloy 600 exposed to primary water chemistry environment. [2]在这项工作中，对暴露在低压 H 2 蒸汽中的合金 600 在 Ni/NiO 转变附近的一系列氧化电势的系统评估表明，在比 Ni/NiO 转变更具氧化性的条件下，PIO 显着降低，而无论氧化电位如何，都会发生局部扩散引起的晶界迁移。 [1] 最后，本文介绍了一种测量晶界强度的新技术，该技术在暴露于原始水化学环境的合金 600 的氧化晶界上进行了演示。 [2]