• Corrosion Science and Technology
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• 제6권6호
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• pp.316-321
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• 2007

The effect of annealing on the pitting corrosion resistance of anodized Al-Mg alloy (AA5052) processed by equal-channel angular pressing (ECAP) was investigated by electrochemical techniques in a solution containing 0.2 mol/L of $AlCl_3$ and also by surface analysis. The Al-Mg alloy was annealed at a fixed temperature between 473 and 573 K for 120 min in air after ECAP. Anodizing was conducted for 40 min at $100-400A/m^2$ at 293 K in a solution containing 1.53 mol/L of $H_2SO_4$ and 0.0185 mol/L of $Al_2(SO_4)_3$. The internal stress generated in anodic oxide films during anodization was measured with a strain gauge to clarify the effect of ECAP on the pitting corrosion resistance of anodized Al-Mg alloy. The time required to initiate the pitting corrosion of anodized Al-Mg alloy was shorter in samples subjected to ECAP, indicating that ECAP decreased the pitting corrosion resistance. However, the pitting corrosion resistance was greatly improved by annealing after ECAP. The time required to initiate pitting corrosion increased with increasing annealing temperature. The strain gauge attached to Al-Mg alloy revealed that the internal stress present in the anodic oxide films was compressive stress, and that the stress was larger with ECAP than without. The compressive internal stress gradually decreased with increasing annealing temperature. Scanning electron microscopy showed that cracks occurred in the anodic oxide film on Al-Mg alloy during initial corrosion and that the cracks were larger with ECAP than without. The ECAP process of severe plastic deformation produces large internal stresses in the Al-Mg alloy; the stresses remain in the anodic oxide films, increasingthe likelihood of cracks. It is assumed that the pitting corrosion is promoted by these cracks as a result of the higher internal stress resulting from ECAP. The improvement in the pitting corrosion resistance of anodized AlMg alloy as a result of annealing appears to be attributable to a decrease in the internal stresses in anodic oxide films

• 대한기계학회논문집A
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• 제20권4호
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• pp.1123-1132
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• 1996

It has been reported that high temperature structural components represent the phenomenon of material degradation according to a long term service under high temperature and pressure. Especially, fossile power plant components using the fossil fuel and heavy oil are affected by dewpoint corrosion of $H_2SO_4$produced during a combustion. Therefore, the service materials subjected to high temperature and pressure may occur the stress corrosion cracking. The object of this paper is to investigate SCC susceptibility according to the material degradation of the high temperature structural materials in dewpoint corrosive environment-$H_2SO_4$.The obtained results are summarized as follows : 1) In case of secondary superheater tube, the fractograph of dimple is observed at the concentration of $H_2SO_4$-5%. When the concentration of $H_2SO_4$ is above 10%, the fracture mode is shifted from a transgranular fracture to an quasi-intergranular fracture according to the increment of concentration. 2) In the relationship between [$\Delta$DBTT]$_sp$ and SCC susceptibility, it is confirmed that the greater material degradation degree is, the higher SCC susceptibility is. In addition, it can be known that SP test is useful test method to evaluate SCC susceptibility for high temperature structural components. 3) When [$\Delta$DBTT]$_sp$ is above 17$17^{\circ}C$ the SCC fracture behavior is definitely observed with SCC susceptibility of above 0.4.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3183-3193
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• 2023

The water radiolysis in-core at light water reactors (LWRs) produces various radicals with other ionic species/molecules and radioactive nitrogen species in the reactor coolant. Nitrogen species can exist in many different chemical forms and recirculate in water and steam, and consequently contribute to what extent the environmental safety at nuclear power plants. Therefore, a clear understanding of formation kinetics and chemical behaviors of nitrogen species under irradiation is crucial for better insight into the characteristics of major radioactive species released to the main steam or relevant coolant systems and eventually development of advanced processes/methodologies to enhance the environmental safety at nuclear power plants. This paper thus focuses on basic principles on electrochemical interaction kinetics of radiolytic molecules and various nitrogen species in high temperature water, fundamental approaches for calculating thermodynamic values to predict their stability and domain in LWRs, and the effect of nitrogen species on crevice chemistry/corrosion and intergranular stress corrosion cracking (IGSCC) susceptibility of structure materials in high temperature water.

• 한국압력기기공학회 논문집
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• 제5권2호
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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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• 제20권4호
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• pp.210-229
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• 2021

To safely operate domestic nuclear power plants approaching the end of their design life, the material degradation management strategy of the components is important. Among studies conducted to improve the soundness of nuclear reactor components, research methods for understanding the degradation of reactor internals and preparing management strategies were surveyed. Since the IGSCC (Intergranular Stress Corrosion Cracking) initiation and propagation process is associated with metal dissolution at the crack tip, crack initiation sensitivity was decreased in the hydrogenated water with decreased crack sensitivity but occurrence of small surface cracks increased. A stress of 50 to 55% of the yield strength of the irradiated materials was required to cause IASCC (Irradiation Assisted Stress Corrosion Cracking) failure at the end of the reactor operating life. In the threshold-stress analysis, IASCC cracks were not expected to occur until the end of life at a stress of less than 62% of the investigated yield strength, and the IASCC critical dose was determined to be 4 dpa (Displacement Per Atom). The stainless steel surface oxide was composed of an internal Cr-rich spinel oxide and an external Fe and Ni-rich oxide, regardless of the dose and applied strain level.

• Corrosion Science and Technology
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• 제17권4호
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• pp.147-153
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• 2018

Copper is used in various applications in environments favoring and enabling formation of biofilms by naturally occurring microbes. Copper is also the chosen corrosion barrier for nuclear waste in Finland. The copper canisters should have lifetimes of 100,000 years. Copper is commonly considered to be resistant to corrosion in oxygen-free water. This is an important argument for using copper as a corrosion protection in the planned canisters for spent nuclear-fuel encapsulation. However, microbial biofilm formation on metal surfaces can increase corrosion in various conditions and provide conditions where corrosion would not otherwise occur. Microbes can alter pH and redox potential, excrete corrosion-inducing metabolites, directly or indirectly reduce or oxidize the corrosion products, and form biofilms that create corrosive microenvironments. Microbial metabolites are known to initiate, facilitate, or accelerate general or localized corrosion, galvanic corrosion, and intergranular corrosion, as well as enable stress-corrosion cracking. Sulfate-reducing bacteria (SRB) are present in the repository environment. Sulfide is known to be a corrosive agent for copper. Here we show results from corrosion of copper in anoxic simulated ground water in the presence of SRB enriched from the planned disposal site.

• 에너지공학
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• 제23권2호
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• pp.149-157
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• 2014

원자력발전소의 증기발생기 전열관은 가동 중에 다양한 형태의 부식 손상이 발생한다. 전열관의 외면에 발생하는 응력부식균열은 2차측 응력부식균열이라 불리는데 주로 전열관의 확관천이지역에서 발생한다. 그 원인은 이 지역의 기하학적 특성과 관련된 슬러지의 침적에 의한 불순물의 농축과 증기 발생기 제작과정에서 확관에 의한 잔류응력이다. 특히 잔류응력은 확관방법에 따라 방향성 및 그 크기가 달라지는데 전열관에 발생하는 균열의 방향 및 발생빈도는 이와 관련이 있다. 현장 경험에 따르면, 폭발확관된 전열관은 수압확관된 전열관에 비해 확관천이 부위에서 원주방향 균열이 잘 발생하는 것으로 나타났다. 따라서 본 연구에서는 예민화된 증기발생기 전열관에 대한 응력부식균열 시험을 통해 확관법에 따른 특정방향 균열의 발생빈도 및 균열 크기를 비교하였다. 또한 균열이 발생된 전열관의 파단면 검사를 통해 균열 양상과 수화학 환경 중의 특정 성분의 영향을 관찰하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(3)
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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.

• 한국압력기기공학회 논문집
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• 제11권1호
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• pp.20-24
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• 2015

Steam generator(S/G) tubes in pressurized water reactor (PWR's) are subject to several types of degradation. This degradation includes denting, pitting, intergranular attack(IGA), intergranular stress corrosion cracking(IGSCC), fatigue, fretting and wear. Degradation can be derived from either the primary side(inside) or the secondary side(outside) of the tube. Recent issue for tube degradation in domestic steam generator is the tube end cracking on seal weld region. The seal weld region at the tube end and tube itself is regarded as a pressure boundary between the primary side and the secondary side. One of the Westinghouse Model-F S/G has experienced tube end cracking and its number of plugging approximately becomes to the operating limit up to 5% due to tube end cracking which was reported as SAI/MAI(single/multiple axial indication) or SCI/MCI(Single/multiple circumferential indication) from the results of eddy current testing. Eddy current mock-up test was carried out to determine the origin of cracking whether it is from weld zone area or parent tube. This result was helpful to analyze crack location on ECT data. Correct action on this problem was the installation of tube-end sleeve. Last year, after removing 340 installed plugs from tubes, selected 269 tubes took tube-end sleeve installation. Tube-end sleeve brought pressure boundary from parent tube to installed sleeve tube. Tube-end sleeve has the benefit of reducing outage period and increasing more revenue than replacing S/G. This paper is provided to assist interest parties in effectively understanding this issue.

• 한국압력기기공학회 논문집
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• 제6권2호
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• pp.34-41
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• 2010

This work is concerned with the analysis of stress corrosion cracking(SCC) behavior of Alloy 600 nozzle penetration mock-up according to a residual stress induced by a dissimilar metal welding(DMW) in a nuclear reactor pressure vessel. The effects of the dimension and materials of the nozzle penetration on the deformation and the residual stress induced by DMW were investigated using a finite element analysis(FEA). The inner diameter(ID) change of the nozzle by DMW and its dependance on the design variables, calculated by FEA, were well consistent with those measured from the mock-up. Accelerated SCC tests were performed for three mock-ups with different wall thicknesses in a highly acidic solution to investigate mainly the effect of the residual stress on the SCC behavior of Alloy 600 nozzle. From a destructive examination of the mock-up after the tests, the SCC behavior of the nozzle was fairly related with the residual stress induced by DMW : axial cracks were found in the ID surface of the nozzle within the J-weld region where the highest tensile hoop stress was predicted by FEA, while circumferential cracks were observed beyond both J-weld root and toe where the highest tensile axial stress was expected.