• Korean Journal of Materials Research
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• v.9 no.2
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• pp.217-225
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• 1999

The effect on the pitting corrosion resistance of laser welding and surface treatment developed as a repair method of stream generator tubing material that was a major component of primary system at nuclear power plant was observed. Some heat-treated Alloy 600 tubing materials used at domestic nuclear power plants were laser-surface observed. Some heat-treated Alloy 600 tubing materials used at domestic nuclear power plants were laser-surface melted and the microstructural characteristics were examined. The pitting corrosion resistance was examined through Ep(pitting potential) and degree of pit generation by means of the electrochemical tests and the immersion tests respectively. The pit formation characteristics were investigated through microstructural changes and the pit initiation site and pit morphology. The test results showed that the pitting corrosion resistances was increased in the order of the followings; sensitized Alloy 600, solution annealed alloy600, and laser surface melted Alloy 600. Pits were initiated preferably at Ti-containing inclusions and their surroundings in all tested specimens and it is believed that higher pitting resistance of laser-surface treated Alloy 600 was caused by fine, homogeneous distribution of non-soluble inclusions, the disappearance of grain boundary, and the formation of dense, stable oxide film. The major element of corrosion products filled in the pit was Cr. On the other hand, Fe was enriched in the deposit formed on the pit.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.17-22
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• 2016

Alloy 600 widely used in nuclear power plant is susceptible to primary water stress corrosion cracking (PWSCC). It is important to prioritize the inspection of Alloy 600 components using PWSCC susceptibility index. Plant-specific model for the susceptibility index was reviewed. The normalized PWSCC susceptibility index to a reference value is suggested and applied. The result was found to be reasonable.

• Journal of the Korean Electrochemical Society
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• v.6 no.4
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• pp.271-281
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• 2003

The present article is concerned with pitting corrosion of Inconel alloy 600 at elevated solution temperatures. This article first summarized the previous works on the characteristics and the growth models of oxide film grown on alloy 600 at elevated solution temperatures. Thereafter, this article reviewed previous works on the morphological study on pitting corrosion of alloy 600 as functions of solution temperature and such anion additives as thiosulphate, sulphate, nitrate and bicarbonate ions in terms of pit morphology and its fractal dimension.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.1036-1046
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• 2016

We propose a primary water stress corrosion cracking (PWSCC) initiation model of Alloy 600 that considers the stress triaxiality factor to apply to finite element analysis. We investigated the correlation between stress triaxiality effects and PWSCC growth behavior in cold-worked Alloy 600 stream generator tubes, and identified an additional stress triaxiality factor that can be added to Garud's PWSCC initiation model. By applying the proposed PWSCC initiation model considering the stress triaxiality factor, PWSCC growth simulations based on the macroscopic phenomenological damage mechanics approach were carried out on the PWSCC growth tests of various cold-worked Alloy 600 steam generator tubes and compact tension specimens. As a result, PWSCC growth behavior results from the finite element prediction are in good agreement with the experimental results.

• Corrosion Science and Technology
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• v.7 no.1
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• pp.61-67
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• 2008

In this paper, electrochemical and microstructural characteristics of nickel-plated Alloy 600 were investigated in order to identify the performance of electroless Ni-plating on Alloy 600 in high-temperature aqueous condition with the comparison of electrolytic nickel-plating. For high temperature corrosion test of nickel-plated Alloy 600, specimens were exposed for 770 hours to typical PWR primary water condition. During the test, open circuit potentials (OCP's) of all specimens were measured using a reference electrode. Also, resistance to flow accelerated corrosion (FAC) test was examined in order to check the durability of plated layers in high-velocity flow environment at high temperature. After exposures to high flow rate aqueous condition, the integrity of surfaces was confirmed by using both scanning electron microscopy (SEM) and energy dispersive spectroscopy (EDS). For the field application, a remote process for electroless nickel-plating was demonstrated using a plate specimen with narrow gap on a laboratory scale. Finally, a practical seal design was suggested for more convenient application.

• Nuclear Engineering and Technology
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• v.30 no.6
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• pp.555-567
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• 1998

Alloy 600 and 690 steam generator tubes fabricated in Korea were evaluated by flare tests according to ASTM standards. The stress acting in the tube elements during the tests was predicted. All the tubes, including alleys 600 and 690, satisfied the requirement of a 30% or 35% O.D expansion. Flow curves obtained from the flare test were found to be higher in alloy 690 tubes than in alloy 600 ones. The difference between alloy 600 and 690 tubes increased gradually with flaring percentage (F.P,%). An effective stress corresponding to mean yield stress was introduced and calculated. It showed that the prediction values were in good agreement with the measured ones for all the 690 and 600 alloy tubes. It became possible to predict the amount of acting stresses within tubes during expansion process.

• 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).

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.228-237
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• 2019

The crack growth responses of as-received and as-welded Alloy 600/182 and Alloy 690/152 welds to constant loading were measured by a direct current potential drop method using compact tension specimens in primary water at $325^{\circ}C$ simulating the normal operating conditions of a nuclear power plant. The as-received Alloy 600 showed crack growth rates (CGRs) between $9.6{\times}10^{-9}mm/s$ and $3.8{\times}10^{-8}mm/s$, and the as-welded Alloy 182 had CGRs between $7.9{\times}10^{-8}mm/s$ and $7.5{\times}10^{-7}mm/s$ within the range of the applied loadings. These results indicate that Alloys 600 and 182 are susceptible to cracking. The average CGR of the as-welded Alloy 152 was found to be $2.8{\times}10^{-9}mm/s$. Therefore, Alloy 152 was proven to be highly resistant to cracking. The as-received Alloy 690 showed no crack growth even with an inhomogeneous banded microstructure. The cracking mode of Alloys 600 and 182 was an intergranular cracking; however, Alloy 152 was revealed to have a mixed (intergranular + transgranular) cracking mode. It appears that the Cr concentration and the microstructural features significantly affect the cracking resistance and the cracking behavior of Ni-base alloys in PWR primary water.

• Nuclear Engineering and Technology
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• v.29 no.2
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• pp.138-147
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• 1997

Korean-made alloys 600 and 690 tubes were evaluated by flare tests according to ASTM standards, and acting stresses during the test ore analyzed. All the tubes, including alloys 600 and 690 tubes with various heat treatment conditions, satisfied the requirement with 30 or 35750.D expansion. Axial stresses in alloy 690 tubes were higher than those in alloy 600 ones and the gap increased gradually with flaring percentage(F.P, %). Assuming the tubes as the rigid-perfectly plastic body, a stress equation was obtained using modified Tresca's yield criterion. Also microstructural change of the flared tubes was discussed with the acting stresses.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.3
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• pp.333-338
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• 2013

The compressive residual stress and fatigue behavior of shot peened alloy 600 under a high-temperature environment is investigated in this study. Alloy 600 is used in the main parts of nuclear power plants, and the compressive residual stress induced by the shot peening process is considered to prevent SCC (stress corrosion cracking). To obtain practical results, the fatigue characteristics and compressive residual stress are evaluated under the actual operating temperature of a domestic nuclear power plant, as well as a high-temperature environment. The experimental results show that the peening effects are valid at a high temperature lower than approximately $538^{\circ}C$, which is the threshold temperature. The fatigue life was maintained at temperatures lower than $538^{\circ}C$, and the compressive residual stress at $538^{\circ}C$ was 68.2% of that at room temperature. The present results are expected to be used to obtain basic safety and reliability data.