The effect of deformation temperature on the microstructure evolution of Inconel 625 superalloy

Hot compression tests of Inconel 625 superalloy were conducted using a Gleeble-1500 simulator between 900 °C and 1200 °C with different true strains and a strain rate of 0.1 s⁻¹. Scanning electron microscope (SEM) and electron backscatter diffraction technique (EBSD) were employed to investigate the effect of deformation temperature on the microstructure evolution and nucleation mechanisms of dynamic recrystallization (DRX). It is found that the relationship between the DRX grain size and the peak stress can be expressed by a power law function. Significant influence of deformation temperatures on the nucleation mechanisms of DRX are observed at different deformation stages. At lower deformation temperatures, continuous dynamic recrystallization (CDRX) characterized by progressive subgrain rotation is considered as the main mechanism of DRX at the early deformation stage. However, discontinuous dynamic recrystallization (DDRX) with bulging of the original grain boundaries becomes the operating mechanism of DRX at the later deformation stage. At higher deformation temperatures, DDRX is the primary mechanism of DRX, while CDRX can only be considered as an assistant mechanism at the early deformation stage. Nucleation of DRX can also be activated by the twinning formation. With increasing the deformation temperature, the effect of DDRX accompanied with twinning formation grows stronger, while the effect of CDRX grows weaker. Meanwhile, the position of subgrain formation shifts gradually from the interior of original grains to the vicinity of the original boundaries.     

Friction and wear of Inconel 690 and Inconel 600 for steam generator tube in room temperature water

In the steam generators of nuclear power plants, the flow of cooling water can cause the tubes to vibrate, resulting in fretting wear damage due to contacts between these tubes and their supports. The tubes are made of Inconel 690 and Inconel 600 and the supports are made of STS 304. In this paper, fretting wear tests in water were performed using the materials Inconel 690 and Inconel 600 in contact with STS 304. Fretting tests using a cross-cylinder type set up were conducted under various vibrating amplitudes and applied normal loads in order to measure friction forces and wear volumes. Also, conventional sliding tests using a pin-on-disk type set up were carried out to compare these test results.In the fretting tests, friction force was found to be strongly dependent on normal load and vibrating amplitude. Coefficients of friction decreased with an increase in the normal load and a decrease in the vibrating amplitude applied. Also, the wear of Inconel 600 and Inconel 690 was predicted using a work rate model. Depending on the normal load and vibrating amplitude applied, distinctively different wear mechanisms and often drastically different wear rates occurred. It was found that the fretting wear coefficients for Inconel 600 and Inconel 690 were 9.3×10⁻¹⁵ and 16.2×10⁻¹⁵ Pa⁻¹, respectively. This study shows that Inconel 690 can result in lesser friction forces and exhibits less wear resistance than Inconel 600 in room temperature water.     

Corrosion behavior of Inconel 690 and 693 in an iron phosphate melt

The corrosion resistance of Inconel 690 and 693 coupons submerged in an iron phosphate melt has been investigated. After 155 days in an iron phosphate melt at 1050 °C, which contained 30 wt% of a simulated low activity waste (LAW at Hanford), the weight loss of Inconel 690 and 693 was 14% and 8%, respectively. The overall corrosion rate, calculated from the initial and final dimensions of each coupon, was 1.3 and 0.7 μm/day for the Inconel 690 and 693, respectively. Scanning electron microscopy and X-ray diffraction of the submerged Inconel coupons after 155 days in the iron phosphate melt showed that an altered surface layer had formed which was depleted in nickel and consisted of a (Fe, Cr)₂O₃ solid solution. This altered layer appears to be chemically protective as indicated by the gradual reduction in weight loss which occurred with time in the iron phosphate melt. Inconel 693 appears to be a better candidate to use as an electrode in iron phosphate melts since its corrosion rate and weight loss was only about one half that of Inconel 690.     

低浓去污预氧化过程中因科镍690合金我因科镍600合金的电化学 …

介绍了低浓去污预氧化过程中两种蒸汽发生器材料因科镍６９０合金及因科镍６００合金（下面简称６９０合金及６００合金）在不同邓氧化剂中脱膜效果（以铬的释放曲线表征）实验、腐蚀电位迁移测试和极化曲线测试。结果表明,６００合金在碱性高锰酸钾（ＡＰ）中的脱膜效果好于酸性高锰酸钾（ＮＰ）中,而６９０合金则在（ＮＰ）中的脱膜效果好于ＡＰ中。并且在ＮＰ中,随着硝酸浓度的增加,６００合金的腐蚀电位向正方向移动,６９０     

An experimental study on fretting wear behavior of cross-contacting Inconel 690 tubes

A fretting wear test rig employing a piezoelectric actuator has been developed, which is equipped with a heating and water circulation system. The fretting wear tests of cross-contacting Inconel 690 tubes, which is widely used for power plant steam generator, have been carried out in room temperature ambient and 80 °C in-water conditions. Maximum normal load was 55 N, and the sliding amplitude was below 50 μm. Scars of the mixed-slip and the gross-slip fretting wear have been measured in terms of scar diameter and wear volume. From the relationship between the work rate and the wear rate, a threshold of work rate has been defined, and this is found to be closely related with fretting wear regimes. The wear coefficients have been evaluated in the gross-slip regime. Distinct fretting wear mechanisms have been observed for the two different test conditions from SEM microphotographs. The crack formation, large particle separation and resulting third body effect were significant in room temperature ambient condition. The protective nature of the tribologically transformed layers coupled with non-uniform contact results in the lower wear coefficient while smooth wear scar and extensive abrasion produces higher wear volume in the other condition.     

热处理对690合金抗腐蚀性能影响综述

研究了热处理工艺对６９０使 为微组织及其抗腐蚀和应务腐蚀性能的影响。将国产６９０合金样品与国外同类产品作了对比分析。结果显示，国产样品显微组织已与国外同类产品基本相同；对应不同热处理工艺，合金的显微结构有很大区别沿晶界析出的碳化物大小、形貌也不同，抗腐蚀和应力腐蚀的性能存在明显差异。     

含缺陷Inconel 690蒸汽发生器传热管的强度及堵管准则研究

针对缺陷对传热管强度的影响以及传热管判废准则问题展开研究,研制了适用于小管径蒸汽发生器传热管极限载荷及爆破压测试的实验装置,对含体积型缺陷及面型缺陷的Inconel 690蒸汽发生器传热管进行了实验研究,并采用有限元法对极限载荷及爆破压进行了估算.在此基础上,研究了传热管的堵管准则,提出了两级评定方法.该评定方法可根据缺陷的深度、轴向及环向长度来综合评价.     

国产690合金与321不锈钢异种金属焊缝的应力腐蚀行为研究

利用慢应变速率试验,采用非标准的漏斗状试样,对国产690合金与321不锈钢异种金属焊接部位(包括690合金热影响区、焊缝、321不锈钢热影响区)在100 mg/L Cl~(-1)除O_2条件下和100 mg/L Cl~(-1)饱和O_2条件下的应力腐蚀行为进行研究。并通过慢应变速率应力-位移曲线和断口形貌对微观组织、氯离子、氧含量对于材料的应力腐蚀(SCC)的影响进行分析。结果表明:690合金热影响区在100 mg/L Cl~(-1)除O_2条件下不易发生SCC,在100 mg/L Cl~(-1)饱和O_2条件下表现出一定的SCC倾向;321不锈钢热影响区在2种条件下均表现出明显的SCC倾向;690合金热影响区的粗大晶粒不利于塑性变形的晶粒间相互协调,导致了热影响区SCC的倾向增大。     

Inconel690(TT)合金裂纹尖端小范围屈服时的腐蚀疲劳裂纹扩展行为研究

对Inconel 690(TT)合金腐蚀疲劳裂纹尖端塑性区微观结构进行观察,并研究裂纹尖端塑性区及载荷比对模拟压水堆环境下的裂纹扩展行为的影响。裂纹尖端小范围屈服时,模拟压水堆环境对裂纹扩展速率有3倍左右的加速作用。。     

Inconel-690合金在核电厂水质环境中的均匀腐蚀性能研究

模拟核电厂水质环境,采用动水腐蚀回路研究3种蒸汽发生器传热管商用690材料的均匀腐蚀性能以及氧化膜的特性,并分别采用国家标准和美国标准对材料均匀腐蚀速率和腐蚀产物释放速率进行评价。结果表明:690合金管在核电厂水质环境中具有极低的腐蚀速率和腐蚀产物释放速率,日本住友管的腐蚀性能略优于宝钢管。     

Inconel690(TT)合金在压水堆二回路水环境下的腐蚀疲劳裂纹扩展行为研究

Inconel 690(TT)合金是压水反应堆蒸汽发生器传热管的关键材料之一,在压水反应堆工况下具有腐蚀疲劳开裂的风险。本文在裂纹尖端小范围屈服的条件下,研究了Inconel 690(TT)合金在模拟二回路水介质环境下的腐蚀疲劳裂纹扩展行为。研究发现:相对于室温情况下,模拟二回路水介质对疲劳裂纹扩展速率有最大3倍左右的加速作用;模拟二回水介质对疲劳裂纹扩展速率的加速作用受腐蚀疲劳裂纹非平面生长的影响,并与应力强度因子范围、最大应力强度因子及应力比密切相关。     

Formation of diffusion barrier coating on superalloy 690 substrate and its stability in borosilicate melt at elevated temperature

Aluminized and thermally oxidized superalloy 690 substrates forming Al₂O₃ layer on (NiCr)Al + Cr₅Al₈ types aluminides and bare substrates were exposed in sodium borosilicate melt at 1248 K for 192 h. SEM–EDXS analysis along the cross-section of bare substrate with adhered glass revealed formation of a continuous, thick Cr₂O₃ layer at the substrate/glass interface due to its low solubility in borosilicate melt. XRD on aluminide coated and thermally oxidized specimen revealed existence of Al₂O₃ along with NiAl and Cr₅Al₈ type phases after the exposure in borosilicate melt. SEM–EDXS analysis along the cross-section of aluminide coated and thermally oxidized sample with adhered glass indicated good stability of coating in borosilicate melt without any phase formation at the coating/glass interface. However, some Al enrichment in glass phase adjacent to interface was noticed without any significant Ni or Cr enrichment.     

Deformation mode maps for tensile deformation of neutron-irradiated structural alloys

The deformation microstructures of neutron-irradiated nuclear structural alloys, A533B steel, 316 stainless steel, and Zircaloy-4, have been investigated by tensile testing and transmission electron microscopy to map the extent of strain localization processes in plastic deformation. Miniature specimens with a thickness of 0.25 mm were irradiated to five levels of neutron dose in the range 0.0001-0.9 displacements per atom (dpa) at 65-100 °C and deformed at room temperature at a nominal strain rate of 10⁻³ s⁻¹. Four modes of deformation were identified, namely three-dimensional dislocation cell formation, planar dislocation activity, fine scale twinning, and dislocation channel deformation (DCD) in which the radiation damage structure has been swept away. The modes varied with material, dose, and strain level. These observations are used to construct the first strain-neutron fluence-deformation mode maps for the test materials. Overall, irradiation encourages planar deformation which is seen as a precursor to DCD and which contributes to changes in the tensile curve, particularly reduced work hardening and diminished uniform ductility. The fluence dependence of the increase in yield stress, ΔYS = α(?t)ⁿ had an exponent of 0.4-0.5 for fluences up to about 3 × 10²² n m⁻² (∼0.05 dpa) and 0.08-0.15 for higher fluences, consistent with estimated saturation in radiation damage microstructure but also concurrent with the acceleration of gross strain localization associated with DCD.     

Comparison of Inconel 625 and Inconel 600 in resistance to cavitation erosion and jet impingement erosion

Liquid droplet erosion (LDE), which often occurs in bellows made of nickel-based alloys, threatens the security operation of the nuclear power plant. As the candidate materials of the bellows, Inconel 600 and Inconel 625 were both tested for resistance to cavitation erosion (CE) and jet impingement erosion (JIE) through vibratory cavitation equipment and a jet apparatus for erosion-corrosion. Cumulative mass loss vs. exposure time was used to evaluate the erosion rate of the two alloys. The surface and cross-sectional morphologies before and after the erosion tests were observed by scanning electron microscopy (SEM), the inclusions were analyzed by an energy dispersive spectroscopy (EDS), and the surface roughness was also measured by surface roughness tester to illustrate the evolution of erosion process. The results show that the cumulative mass loss of CE of Inconel 625 is about 1/6 that of Inconel 600 and the CE incubation period of the Inconel 625 is 4 times as long as that of the Inconel 600. The micro-morphology evolution of CE process illustrates that the twinning and hardness of the Inconel 625 plays a significant role in CE. In addition, the cumulative mass loss of JIE of Inconel 625 is about 2/3 that of Inconel 600 at impacting angle of 90°, and almost equal to that of the Inconel 600 at impacting angle of 30°. Overall, the resistance to CE and JIE of Inconel 625 is much superior to that of Inconel 600.     

Interaction between borosilicate melt and Inconel

Understanding the mode of interaction between borosilicate melt and Inconel is important for long time usage of melter pot in vitrification plant. The present study shows that significant elemental exchanges take place across the borosilicate melt/Inconel interface resulting in the development of (Fe, Ni)CrO₄ needle and (Fe, Ni)Cr₂O₄ cubic phases. This results in significant depletion of Cr within Inconel near the interface. Beside these, CrB precipitates formed along the Inconel grain boundaries.     

Inconel 690合金氢扩散与渗透特性的研究

采用超高真空气相氢渗透技术,研究氢在Inconel 690合金的扩散和渗透行为,获得了不同厚度690合金样品在300~500℃温度范围内的氢渗透曲线,讨论了该合金厚度对扩散特性与渗透特性的影响。结果表明:在一定温度范围内,Inconel 690合金的氢扩散系数(D)、渗透系数(P)随温度变化关系遵循Arrhenius方程;在一定厚度范围内,氢扩散系数随样品厚度增加而增高并趋近于690合金的实际体扩散系数;样品厚度对氢渗透系数的变化影响较小。     

Thermal stability of the superalloys Inconel 625 and Nimonic 86

The thermal stability of Inconel 625 and Nimonic 86, as received, cold worked (10, 20, and 40%), and solution treated, was investigated in the temperature range 500–900°C. The annealing times varied from 0.3 (0.03) to 100 days.Precipitation hardening and recovery (recrystallisation) takes place in cold worked material, beginning after shorter times in cold worked material than in as received material. The temperature interval for precipitation hardening is extended in Nimonic 86, due to cold working, from about 500–600°C to about 450–700°C. It is possible to suppress or retard the precipitation hardening in solution treated Inconel 625 and Nimonic 86 by fast cooling after solution annealing. Hardness was measured at room temperature with five different loads, so that the parameters $\text{k}$ and $\text{n}$ from Meyer's-law, and the Brinell hardness number (for $\text{F ÷ D}{}^2\text{=30}$) could be determined.The lattice contraction of Inconel 625 due to ageing was investigated with X-ray measurements. The change of intensities of the diffractometer traces due to recovery was also determined.     

The IGSCC, sensitization, and microstructure study of Alloys 600 and 690

Both Alloys 600 and 690 were studied to understand the effect of heat treatment on the sensitization and SCC behavior of these alloys. The microstructural evolution and chromium depletion near the grain boundaries were carefully studied using analytical electron microscopy. The majority of the precipitates formed in Alloy 600 was found to be M₇C₃ with a hexagonal structure (a₀ = 1.398 nm, c₀ = 0.45 nm); whereas the carbides found in Alloy 690 were identified as M₂₃C₆ with an fcc structure (a₀ = 1.06 nm). Modified Huey test performed in boiling 40% HNO₃ was used to study the effect of heat treatment and degree of sensitization. Constant load tests and constant extension rate tests were performed in the solution containing sodium thiosulfate to study the SCC resistance of these alloys. The results of the constant load tests for Alloy 600 indicated that the susceptibility to SCC is sensitive to the chromium depletion depth at grain boundary, and the minimum value to prevent SCC failure is approximately 8 wt%. No SCC was observed for Alloy 690 tested using constand load and CERT in the same environments. All tests showed that Alloy 690 has a far better resistance to intergranular attack and SCC than Alloy 600, which is believed due to its high chromium content. It is therefore anticipated that Alloy 690 now a better substitute to Alloy 600 as a steam generator tubing material for pressurized water reactor will also offer a superior corrosion resistance when “sensitized” and in particular if exposed to sulfur containing media such as thiosulfate solutions.     

IGSCC of grain boundary engineered 316L and 690 in supercritical water

This study evaluated the influence of a high fraction of special grain boundaries on the intergranular stress corrosion cracking susceptibility of 316L stainless steel and nickel base alloy 690 in supercritical water. By thermomechanically processing the alloys to create specimens with largely different special boundary fractions, it was possible to isolate the effects of the grain boundary structure on the intergranular stress corrosion cracking behavior. Constant extension rate tensile experiments were performed in 500 °C deaerated supercritical water, and SEM analysis of the cracking behavior was performed on the gage surfaces of the specimens. Results indicate that the fraction of cracked grain boundary length in the specimens with higher fractions of special boundaries is reduced for 316L and 690 by factors of 9 and 5 at 15% strain, and 3 and 2 at 25% strain, respectively. This reduction is due to the special boundaries, which at 25% strain have a frequency of cracking that is 9-18 times lower than that for a random high angle boundary.     

热处理对690合金腐蚀性能影响的实验研究

采用适合高Ｃｒ含量合金的晶间腐蚀试验方法（沸腾６５％ＨＮＯ。＋０．１％ＨＦ溶夜浸渍试验）和在３１６℃、５０％ＮａＯＨ溶液中的慢应变速率试验（ＳＳＲＴ）．研究了热处理对６９０合金晶间腐蚀和碱应力腐蚀性能的影响．热处理包括不同固溶温度（９５０－１１５０℃）及特殊热处理（Ｔ．Ｔ７１５℃）时不同保持时间（２～３０ｈ），根据试验结果．推荐６９Ｏ合金的热处理条件是；固溶温度应＜１１００℃，在７１５℃特殊热处理保持时间１５ｈ。