• Proceedings of the KWS Conference
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• 2005.06a
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• pp.209-211
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• 2005

Super duplex stainless steels have been used for offshore oil and gas piping systems which are subject to corrosion atmosphere, because they have excellent resistance to Stress Corrosion Cracking (SCC) and Pitting corrosion and high strength/weight ratio. Normally, the welding for duplex stainless steels has been peformed using GTAW with Ar shielding gas. However, in case of using Ar as shielding gas, the corrosion resistance at root weld metal will be deteriorated due to loss of nitrogen from weld deposit during welding. It is wellknown that the corrosion resistance of super duplex stainless can be restored by addition of nitrogen as shielding gas. In this study, we made super duplex welding with using several kinds of shielding and purging gases and investigated the relationship between shielding gas and corrosion resistance. Consequently, it was shown that corrosion resistance of weld deposit can be restored by addition of $N_{2}$ as shielding gas.

• Nuclear Engineering and Technology
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• v.17 no.3
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• pp.193-199
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• 1985

Iodine-induced stress corrosion cracking tests of Zircaloy-4 cladding were undertaken using the modified infernal pressurization method. The effects of iodine concentration and applied stress were studied. The critical iodine concentration for SCC was found to be about 0.2 mg/$\textrm{cm}^2$ at 603$^{\circ}$K. The threshold stress was dependent on the test temperature and the mechanical properties of the specimen. The fracture surface showed that the crack propagated stepwise iron one grain to others until the material was unstable and then ruptured mechanically. The initial region showed the transgranular feature and the wedge-shaped cracks. As the crack proceeded, the transgranular and ductile-tearing mired feature appeared in the middle region.

• Journal of the Korean Society for Nondestructive Testing
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• v.21 no.1
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• pp.62-68
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• 2001

In order to characterize the microscopic fracture behaviour of the weldment din stress corrosion cracking(SCC) phenomena, SCC and acoustic emission(AE) tests were carried out simultaneously and the correlation between mechanical paramenters obtained from SCC and AE tests was investigated. In the case of base metal, much more AE events were produced at -0.5V than at -0.8V because of the dissolution mechanism before the maximum load. Regardless of the applied voltages to the specimens, however, AE events decreased after the maximum load. In the case of weldment, lots of AE events with larger amplitude $range(40{\sim}100dB)$ were produced because of the singularities of weld HAZ in comparision to the base metal and post-weld heat-treated(PWHT) specimens. Numerous and larger cracks for the weldment were observed on the fractured surfaces by SEM examination. From these results, it was concluded that SCC for the weldment appeared most severely in synthetic seawater. Weld HAZ was softened by PWHT which also contributed to the reduced susceptibility to corrosive environment in comparison to the weldment.

• Proceedings of the Korean Institute of Industrial Safety Conference
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• 2000.11a
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• pp.257-262
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• 2000

원전을 가동함에 따라 전열관에서는 SCC(stress corrosion cracking), 프레팅(fretting) 등과 같은 다양한 종류의 결함이 발생된다. 이러한 결함이 발생된 전열관에 대해서는 건정성 평가를 수행하여 계속 가동을 수행하던가, 전열관 막음(plugging) 또는 재생보수(sleeve) 등의 보수 작업을 수행하게 된다. 현행 전열관의 구조 건정성 확보를 위한 방안 중의 하나로 결함의 종류, 위치 등에 관계없이 모든 결함에 대하여 40% 관두께 기준을 적용하고 있다［1］. 그러나 현재 적용되고 있는 40% 관두께 보수기준은 전열관의 파열사고 가능성을 완벽하게 차단하지 못하면서도 과도하게 보수적인 측면이 있다.(중략)

• Journal of the Korean Society for Nondestructive Testing
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• v.33 no.6
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• pp.511-518
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• 2013

Steam Generator (SG) tube, as a barrier isolating the primary coolant system from the secondary side of nuclear power plants (NPP), must maintain the structural integrity for the public safety and their efficient power generation. So, SG tubes are subject to the periodic examination and the repairs if needed so that any defective tubes are not in service. Recently, corrosion related degradations were detected in the tubes of the domestic OPR-1000 NPP, as a form of axially oriented outer diameter stress corrosion cracking (ODSCC). According to the studies on the factors causing the heat fouling as well as developing corrosion cracking, densely scaled deposits on the secondary side of the SG tubes are mainly known to be problematic causing the adverse impacts against the soundness of the SG tubes [1]. Therefore, the processes of various cleaning methods efficiently to dissolve and remove the deposits have been applied as well as it is imperative to maintain the structural integrity of the tubes after exposing to the cleaning agent. So qualification test (QT) should be carried out to assess the perfection of the chemical cleaning and QT is to apply the processes and to do ECT. In this paper, the chemical cleaning processes to dissolve and remove the scaled deposits are introduced and results of ECT on the artificial crack specimens to determine the effectiveness of those processes are represented.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.78-83
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• 1998

Alloy 600과 Alloy 690의 응력부식파괴(Stress corrosion cracking, SCC)에 미치는 TiO$_2$의 영향을 315$^{\circ}C$의 10%NaOH 수용액에서 RUB(reverse U-bend) 시편, C-Ring 시편과 CT(compact tension)시편을 사용하여 평가하였다. 시편은 alloy 600 MA(mill anneal), alloy 600 TT(thermal treatment) 그리고 alloy 690 TT로 제작하였다. SCC 시험은 탈산된 10%NaOH 수용액에 2 g/1 TiO$_2$를 첨가한 용액과 첨가하지 않은 용액에서 수행하였으며, 이 조건에서 분극곡선도 얻었다. SCC 시험시 시편을 부식전위로부터 +150 ㎷ 양극분극을 가하였다. 기준전극으로 external Ag/AgCl electrode를 사용하였다. Alloy 600 MA로 제작한 RUB 시편은 TiO$_2$가 없는 용액에서 5일 안에 벽 관통 균열을 보였으나 TiO$_2$가 첨가된 용액에서는 균열을 관찰할 수 없었다. TiO$_2$가 첨가됨에 따라 alloy 600과 alloy 690의 임계전류밀도는 크게 감소하였고 또한 부동태 전류밀도도 감소하였다. 부동테 영역에서 TiO$_2$가 있는 용액의 경우 여러 peak가 있는 반면에 TiO$_2$가 없는 용액은 peak가 뚜렷하지 않았다. 이런 결과는 TiO$_2$가 첨가점에 따라 active region에서도 안정한 부동태 피막이 존재한다는 것을 시사한다. 또한 TiO$_2$가 없는 경우 SCC가 잘 일어나는 영역에 존재하는 부동태 피막이 TiO$_2$ 첨가에 따라 repassivation kinetics 등의 성질이 변화한 것으로 판단된다.

• Nuclear Engineering and Technology
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• v.30 no.4
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• pp.377-385
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• 1998

Many nuclear power plants have experienced unscheduled shutdown due to the leakage of steam generator tubes. The leakages are normally due to the crack, possibly stress corrosion cracking (SCC) near the tube expansion at the top of tubesheet or at the tangential point of the row-1 U-bend region. The conventional eddy current technique, which makes use of a differential bobbin coil, has been found to be inadequate for the early detection of SCC. During the in-service inspection, therefore, it is a general practice that the rotating pancake coil (RPC) is used for detecting the cracks. Even in using RPC, however, it is difficult to determine the depth of the cracks quantitatively. This paper attempts to determine the detectability and sizing ability of RPC technique for axial or circumferential cracks at the tube expansion region. The simulated cracks with various dimensions were fabricated by electro-discharge machining (EDM) method. Experimental results are discussed with theoretical calculations.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.7
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• pp.875-881
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• 2013

The failure of a bolted joint owing to stress corrosion cracking (SCC) has been considered one of the most important structural integrity issues in a nuclear power plant. In this study, the failure possibility of bolting, which is used to support the steam generator of a pressurized water reactor, owing to SCC and brittle fracture was evaluated in accordance with guidelines proposed by the Electric Power Research Institute, which are called the Reference Flaw Factor method. For this evaluation, first, detailed finite element stress analyses were conducted to obtain the actual nominal stresses of bolting in which either service loads or bolt preloads were considered. Based on these nominal stresses, the structural integrity of bolting was addressed from the viewpoints of SCC and toughness. In addition, the accuracy of the EPRI Reference Flaw Factor for assessing bolting failure was investigated using finite element fracture mechanics analyses.

• Corrosion Science and Technology
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• v.11 no.3
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• pp.96-102
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• 2012

Outer diameter stress corrosion cracking (ODSCC) has occurred for Alloy 600 (Ni 75 wt%, Cr 15 wt%, Fe 10 wt%) as a heat exchanger tube of the steam generator (SG) in nuclear power plants (NPP) during long term operation. Among many causes for SCC, lead (Pb) is known to be one of the most deleterious species in the secondary system. In the present work, the oxide formed on Alloy 600 was characterized as a function of the PbO content in 0.1 M NaOH at $315^{\circ}C$ by using an electrochemical impedance spectroscopy (EIS), a transmission electron microscopy (TEM), equipped with an energy dispersive x-ray spectroscopy (EDS). The oxide property was analyzed in view of SCC susceptibility.

• Corrosion Science and Technology
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• v.3 no.5
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• pp.198-208
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• 2004

Although there has been no general agreement on the mechanism of primary water stress corrosion cracking (PWSCC) as one of major degradation modes of Ni-base alloys in pressurized water reactors (PWR's), common postulation derived from previous studies is that the damage to the alloy substrate can be related to mass transport characteristics and/or repair properties of overlaid oxide film. Recently, it was shown that the oxide film structure and PWSCC initiation time as well as crack growth rate were systematically varied as a function of dissolved hydrogen concentration in high temperature water, supporting the postulation. In order to understand how the oxide film composition can vary with water chemistry, this study was conducted to characterize oxide films on Alloy 600 by an in-situ Raman spectroscopy. Based on both experimental and thermodynamic prediction results, Ni/NiO thermodynamic equilibrium condition was defined as a function of electrochemical potential and temperature. The results agree well with Attanasio et al.'s data by contact electrical resistance measurements. The anomalously high PWSCC growth rate consistently observed in the vicinity of Ni/NiO equilibrium is then attributed to weak thermodynamic stability of NiO. Redox-induced phase transition between Ni metal and NiO may undermine the integrity of NiO and enhance presumably the percolation of oxidizing environment through the oxide film, especially along grain boundaries. The redox-induced grain boundary oxide degradation mechanism has been postulated and will be tested by using the in-situ Raman facility.