• 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.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4481-4490
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• 2022

In this study, a new rupture disk corrosion test (RDCT) method was developed for real-time detection of stress corrosion cracking (SCC) initiation of Alloy 600 in a primary water environment of pressurized water reactors. In the RDCT method, one side of a disk specimen was exposed to a simulated primary water at high temperature and pressure while the other side was maintained at ambient pressure, inducing a dome-shaped deformation and tensile stress on the specimen. When SCC occurs in the primary water environment, it leads to the specimen rupture or water leakage through the specimen, which can be detected in real-time using a pressure gauge. The tensile stress applied to the disk specimen was calculated using a finite element analysis. The tensile stress was calculated to increase as the specimen thickness decreased. The SCC initiation time of the specimen was evaluated by the RDCT method, from which result it was found that the crack initiation time decreased with the decrease of specimen thickness owing to the increase of applied stress. After the SCC initiation test, many cracks were observed on the specimen surface in an intergranular fracture mode, which is a typical characteristic of SCC in the primary water environment.

• Journal of the Korean Society of Safety
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• v.12 no.1
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• pp.19-28
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• 1997

This study was made of the effect of cold working on the stress corrosion cracking(SCC) of austenitlc 304 stainless steel in boiling 42% $MgCl_2$ solution. For this experiment, specimens cold-worked of 0%, 10%, 20%, 30%, 40% were fabricated respectively, and then experiments of mechanical properties and stress corrosion cracking(SCC) of these specimens were carried out. The results of these experiments indicate that the maximum resistance to SCC showed at 20% of cold working degree and that the SCC susceptibility depended on the volume fraction of deformation-induced martensite by cold working and the work hardening of matrix. On the other hand, the fracture mode was changed. This phenomenon was considered that deformation-induced martensite was grown from transgranular fracture mode to intergranular fracture mode and caused by increased of dislocation density along the slip planes.

• Corrosion Science and Technology
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• v.19 no.1
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• pp.37-42
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• 2020

Even though 304 low-carbon (304L) stainless steel was developed to enhance the resistance to intergranular corrosion and stress corrosion cracking, it is occasionally subject to degradation in harsh environments. The degree of sensitization (DOS) of 304L stainless steel was studied as a function of sensitization using a double-loop electrochemical potentiokinetic reactivation (DL-EPR) method. Sensitizing heat treatment was performed in an Ar atmosphere at 500℃, 600℃, and 700℃, with heat treatment times varying from 0 to 96 h. DOS was measured by the ratio of the peak current density value of the forward scan to that of the reverse scan. After the EPR experiment, the specimen surface was observed by scanning electron microscopy and energy dispersive spectroscopy. The DOS of the specimens heat-treated at 600℃ increased with heat treatment times up to 48 h and then decreased due to a self healing effect. The DOS was higher in specimens heat-treated at 600℃ than those at 500℃ or 700℃. Corrosion of the sensitized specimens occurred mainly at the δ-γ phase boundary. The corrosion morphology at the δ-γ phase boundary changed with sensitizing heat-treatment conditions due to differences in chromium activity in γ austenite and δ ferrite.

• Corrosion Science and Technology
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• v.19 no.3
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• pp.138-145
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• 2020

Immersion tests of Alloy 600 were conducted in simulated primary water environments of a pressurized water reactor at 325 ℃ for 10, 100, and 1000 h to obtain insight into effects of surface deformation on internal and intergranular (IG) oxidation behavior through precise characterization using various microscopic equipment. Oxidized samples after immersion tests were covered with polyhedral and filamentous oxides. It was found that oxides were abundant in mechanically ground (MG) samples the most. The number density of surface oxides increased with time irrespective of the method of surface finish. IG oxidation occurred in mechanically polished (MP) and chemically polished (CP) samples with thin internal oxidation layers. However, IG oxidation was suppressed with relatively thick internal oxidation layers in MG samples compared to MP and CP samples, suggesting that MG treatment could increase resistance to primary water stress corrosion cracking (PWSCC) from the standpoint of IG oxidation. As a result, appropriate surface treatment for Alloy 600 could prevent oxygen diffusion into grain boundaries, inhibit IG oxidation, and finally induce its high PWSCC resistance.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.111-116
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• 1996

IGSCC(Intergranular stress corrosion cracking) behaviors of Alloy 600 were studied by the electrchemical ten methods of controlling specimens electrode potentials in the active-passive transition region of anodic polarization curve. Anodic polarization and static potential tests of stressed C-ring type MA Alloy 600 were carried out in 10% NaOH at 300 $^{\circ}C$ for 7days. It was confirmed that IGSCC of Alloy 600 was accellerated by maintaining the specimen potential in the susceptible active-passive transition region of anodic polarization curve. An intergranular crack was initiated on the surface area of C-ring specimens where protective oxide layer was broken down. And the depth of the crack growth was about 100 ${\mu}$m during the testing periods.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.244-249
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• 2008

It happened a failure on special bolt which supported main landing gear actuator up-lock rod of 00 aircraft. Fracture was occurred at end of center drilled hole and thread machined on bolt. Metallographic, fractographic, and other characteristics of embrittlement analysis and experiments carried out on the failed bolt to find out the reason. Bolt surface was cadmium electroplated(EP) to give lubrication and provide excellent corrosion resistance. Resultly, Bolt was failed due to cadmium embrittlement occurred during baking treatment as well as center drilled hole. for the failure that are relevant to failure analysis and prevention. For their successful functional application, cadmium EP bolts require proper and adequate baking treatment after electroplating, and is complete with no center drilled hole

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1381-1389
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• 2019

Stress corrosion cracking (SCC) widely found in both primary and secondary sides of steam generator (SG) tubing in pressurized water reactors (PWR) has become an important safety issue. Using eddy-current tests (ECTs), non-destructive evaluations are performed for the integrity management of SG tubes against intergranular SCC. To enhance the reliability of ECT, this study investigates the effects of oxide films on ECT's detection capabilities for SCC in laboratory-degraded SG tubing in high temperature and high pressure aqueous environment.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.145-150
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• 1996

Acoustic Emission (AE) technique was applied to stress corrosion cracking of Inconel 600 to investigate the AE capability of detecting crack growth and to obtain the relation between AE characteristics and crack mechanism. The specimens were heat-treated in two conditions (600$^{\circ}C$ for 30 hrs or 700 $^{\circ}C$ for 1 hr) and undergone CERT at two extension rates ( 2.5${\times}$10$^{-5}$ or 1.25${\times}$10$^{－4}$(mm/s)). It was found that the AE peak amplitude from plastic deformation was generally smaller than about 48dB (0.25mV), while Intergranular stress corrosion cracking (IGSCC) and ductile fracture produced higher values of 49 to 70dB (0.3mV to 3mV). The slopes of cumulative amplitude distribution (b-values) were linearly dependent on IGSCC susceptibility and the higher the susceptibility, the smaller the b-value. The monitoring of combined AE parameters such as event rate, amplitude, count and energy can provide effective means to clearly identify the transition from crack initiation and small crack growth to rapid growth of dominant cracks.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.1-17
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• 2018

This article reviews the multiscale nature of stress corrosion cracking (SCC) observed by high-resolution characterizations in austenite stainless steels and Ni-base superalloys in light water reactors (including boiling water reactors, pressurized water reactors, and supercritical water reactors) with related opinions. A new statistical summary and comparison of observed degradation phenomena at different length scales is included. The intrinsic causes of this multiscale nature of SCC are discussed based on existing evidence and related opinions, ranging from materials theory to practical processing technologies. Questions of interest are then discussed to improve bottom-up understanding of the intrinsic causes. Last, a multiscale modeling and simulation methodology is proposed as a promising interdisciplinary solution to understand the intrinsic causes of the multiscale nature of SCC in light water reactors, based on a review of related supporting application evidence.