• Corrosion Science and Technology
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• v.3 no.4
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• pp.131-139
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• 2004

Effects of PbO on the repassivation kinetics and characteristics of passive film of Alloy 690 were examined to elucidate the influences of PbO on the SCC resistance of that alloy. The repassivation kinetics of the alloy was analyzed in terms of the current density flowing from the scratch, i(t), as a function of the charge density that has flowed from the scratch, q(t). Repassivation on the scratched surface of the alloy occurred in two kinetically different processes; passive film initially nucleated and grew according to the place exchange model in which log i(t) is linearly proportional to q(t), and then grew according to the high field ion conduction model in which log i(t) is linearly proportional to 1/q(t) with a slope of cBV. The cBV is found to be a parameter representing repassivation rate and hence SCC susceptibility of the alloy. The lower the value of cBV, the faster the repassivation rate and the higher the SCC resistance of an alloy. Addition of PbO to pH 4 and 10 solutions increased the value of cBV of alloy 690, reflecting slower repassivation rate than without PbO. The change in the value of cBV was grater in pH 10 than in pH 4. The increase in SCC susceptibility of alloy 690 with the addition of PbO to solution was presumably due to the Cr-depletion in the outer parts of passive film of the alloy with an incorporation of Pb compounds in the film, which was revealed by Mott-Schottky, AES and XPS analyses.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.2
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• pp.1-6
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• 2009

Nuclear power plants (NPP) using Alloy 600 (Ni 75wt%, Cr 15wt%, Fe 10wt%) as a heat exchanger tube of the steam generator (SG) have experienced various corrosion problems by ageing such as pitting, intergranular attack (IGA) and stress corrosion cracking (SCC). In spite of much effort to reduce the material degradations, SCC is still one of important problems to overcome. Especially lead is known to be one of the most deleterious species in the secondary system that cause SCC of the alloy. Even Alloy 690 (Ni 60wt%, Cr 30wt%, Fe 10wt%) as an alternative of Alloy 600 because of outstanding superiority to SCC is also susceptible to leaded environment. An oxide on SG tubing materials such as Alloy 600 and Alloy 690 is formed and modified expanding to complex sludge throughout hideout return (HOR) of various impurities including Pb. Oxide formation and breakdown is requisite for SCC initiation and propagation. Therefore it is expected that an oxide property such as a passivity of an oxide formed on steam generator tubing materials is deeply related to PbSCC and an inhibitor to hinder oxide modification by lead efficiently can be found. In the present work, the SCC susceptibility obtained by using a slow strain rate test (SSRT) in aqueous solutions with and without lead was discussed in view of the oxide property. The oxides formed on Alloy 600 and Alloy 690 in aqueous solutions with and without lead were examined by using a transmission electron microscopy (TEM), an energy dispersive x-ray spectroscopy (EDXS), an x-ray photoelectron spectroscopy (XPS) and an electrochemical impedance spectroscopy (EIS).

• Corrosion Science and Technology
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• v.5 no.4
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• pp.141-148
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• 2006

The chemical composition and the semiconducting properties of the passive films formed on Alloy 690 in various film formation conditions were investigated by XPS, photocurrent measurement, and Mott-Schottky analysis. The XPS and photocurrent spectra showed that the passive films formed on Alloy 690 in pH 8.5 buffer solution at ambient temperature, in air at $400^{\circ}C$, and in PWR condition comprise $Cr_2O_3$, $Cr(OH)_3$, ${\gamma}-Fe_2O_3$, NiO, and $Ni(OH)_2$. The thermally grown oxide in air and the passive film formed at high potential (0.3 $V_{SCE}$) in pH 8.5 buffer solution were highly Cr-enriched, whereas the films formed in PWR condition and that formed at low potential (-0.3 $V_{SCE}$) in pH 8.5 buffer solution showed relatively high Ni content and low Cr content. The Mott-Schottky plots exhibited n-type semiconductivity, inferring that the semiconducting properties of the passive films formed on Alloy 690 in various film formation conditions are dominated by Cr-substituted ${\gamma}-Fe_2O_3$. The donor density, i.e., concentration of oxygen vacancy, was measured to be $1.2{\times}10^{21}{\sim}4.6{\times}10^{21}cm^{-3}$ and lowered with increase in the Cr content in the passive film.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2304-2311
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• 2021

The dislocation density in strain-hardened Alloy 690 was analyzed using scanning transmission electron microscopy (STEM) to study the relationship between the local plastic strain and susceptibility to primary water stress corrosion cracking (PWSCC) in nuclear power plants. The test material was cold-rolled at various thickness reduction ratios from 10% to 40% to simulate the strain-hardening condition of plant components. The dislocation densities were measured at grain boundaries (GB) and in grain interiors of strain-hardened specimens from STEM images. The dislocation density in the grain interior monotonically increased as the strain-hardening proceeded, while the dislocation density at the GB increased with strain-hardening up to 20% but slightly decreases upon further deformation to 40%. The decreased dislocation density at the GB was attributed to the formation of deformation twins. After the PWSCC growth test of strain-hardened Alloy 690, the fraction of intergranular (IG) fracture was obtained from fractography. In contrast to the change in the dislocation density with strain-hardening, the fraction of IG fracture increased remarkably when strain-hardened over 20%. From the results, it was suggested that the PWSCC growth behavior of strain-hardened Alloy 690 not only depends on the dislocation density, but also on the microstructural defects at the GB.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.20 no.1
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• pp.16-24
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• 2024

Denting of steam generator (SG) tube is defined as the reduction in tube diameter due to the stresses exerted by the corrosion products formed on the outer diameter surface. This phenomenon is mostly observed in the crevices between SG tube and the top-of tubesheet or tube support plate. Despite the replacement of SG tube with Alloy 690, which has better corrosion resistance than Alloy 600, the denting of SG tube still remains a potential problem that could decrease the SG integrity. Deformation of SG tube by denting phenomenon can affect the surface properties and microstructure of SG tube. In this study, the effects of plastic deformation on surface properties and microstructure of Alloy 690 thermally treated (TT) tube was investigated by using the various analysis techniques. The plastic deformation of Alloy 690 increased the surface roughness and area. Many surface defects such as ripped surface and micro-cracks were observed on the deformed Alloy 690TT specimen. Based on the electron backscatter diffraction analysis, the dislocation density of deformed SG tube increased compared to non-deformed SG tube. In addition, the effects of changes in surface properties and microstructure of SG tube on general corrosion behavior were discussed.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.05a
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• pp.304-309
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• 2002

Fretting wear test in room temperature water was performed to evaluate the wear coefficient of Inconel 600,690 (Pressurized Water Reactor, PWR) and Alloy 800 (CANadian DeuteriumUranium, CANDU) steam generator (SG) tubes against ferritic and martensitic stainless steels. The main focus is to compare the wear behaviors between Alloy 800 and Inconel alloys. Test conditions are $10{\sim}30N$ of normal load, $200{\sim}450{\mu}m$ of sliding amplitude and 30Hz of frequency. The result indicated that the wear rate of Alloy 800 was higher than those of Inconel 690 at various test condition such as normal loads, sliding amplitudes etc. From the results of SEM observation, there was little evidence of plastic deformation layer that were dominantly formed on the worn surfaces of Inconel 690. Also, wear particles in Alloy 800 were released from contacting asperities deformed by severe plastic flow during fretting wear. Main cause of wear rate between Alloy 800 and Inconel 690 may be due to the difference of hardness between martensitic and ferritic stainless steel. The wear rate and wear mechanism of two tubes in room temperature water are discussed.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.286.2-286
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.239-239
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• 1998

• Proceedings of the Korean Powder Metallurgy Institute Conference
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• 2006.09a
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• pp.587-588
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• 2006

The stress corrosion cracking (SCC) susceptibility of Alloy 600 MA, Alloy 600 TT, Alloy 800, and Alloy 690 TT were investigated in a deaerated 0.01 M solution of sodium tetrathionate using reverse u-bend test samples at $340^{\circ}C$. The results showed that SCC occurred in all alloys, excluding Alloy 690 TT. The SCC susceptibility decreased with an increase in the chromium content of the alloys. The results of the deposits and spectra taken from an energy dispersive X-ray system confirmed the existence of a reduced sulfur causing SCC.

• Corrosion Science and Technology
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• v.7 no.3
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• pp.187-191
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• 2008

Transgranular stress corrosion cracking of nickel base alloys was reported by Copson and Dean in 1965. Study to establish this cracking mechanism needs to be carried out. Laboratory stress corrosion tests were performed for mill annealed(MA) or thermally treated(TT) steam generator tubing materials in a high temperature water containing lead. An electrochemical interaction of lead with the alloying elements of SG tubings was also investigated. Alloy 690 TT showed a transgranular stress corrosion cracking in a 40% NaOH solution with 5000 ppm of lead, while intergranular stress corrosion racking was observed in a 10% NaOH solution with 100 ppm lead. Lead seems to enhance the disruption of passive film and anodic dissolution of alloy 600 and alloy 690. Crack tip blunting at grain boundary carbides plays a role for the transgranular stress corrosion cracking.