• Transactions of the Korean Society of Mechanical Engineers A
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• v.32 no.10
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• pp.873-881
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• 2008

Weld overlay is one of the residual stress mitigation methods which arrest crack initiation and crack growth. Therefore weld overlay can be applied to the region where cracking is likely to be. An overlay weld used in this manner is termed a preemptive weld overlay(PWOL). In pressurized water reactor(PWR) dissimilar metal weld is susceptible region for primary water stress corrosion cracking(PWSCC). In order to examine the effect of PWOL on residual stress mitigation, PWOL was applied to a specific dissimilar metal weld of Kori nuclear power plant by finite element analysis method. As a result, strong compressive residual stress was made in PWSCC susceptible region and PWOL was proved effective preemptive repair method for weldment.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.3
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• pp.7-12
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• 2012

Steam generator tubes are critical boundary of the primary and secondary side in nuclear power plants. Eddy current testing is commonly used as the method of non-destructive testing for the safety and integrity of steam generator tubes in the nuclear power plants. Changes in the geometric shape act as a stress concentration factor likely to cause a defect during the steam generator operation. The mixed-signals with the geometric shape are distorted and attributes that are difficult to detect signals. An example is bending stress due to compression process at a U-bend occurring in the intrados region which has a small radius of curvature. The resulting change in the geometric shape may lead to a dent like occurrences. The dent can cause stress concentration and generates stress corrosion cracks. In this study, the steam generator tubes of nuclear power plant were selected to study for analysis of mixed-signal containing dent and stress corrosion cracks.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1524-1536
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• 2017

During the fabrication process of reactor vessel head penetration (RVHP), the grain size of the tube material can be changed by hot or cold work and the inner side of the tube can also be shrunk due to welding outside of the tube. Several nonregular time-of-flight diffraction (TOFD) signals were found because of deformed grains. In this paper, an investigation of nonregular TOFD indications acquired from RVHP tubes using experiments and computer simulation was performed in order to identify and distinguish TOFD signals by coarse grains from those by Primary Water Stress Corrosion Crack (PWSCC). For proper understanding of the nonregular TOFD indications, microstructural analysis of the RVHP tubes and prediction of signals scattered from the grains using Finite Element Method (FEM) simulation were performed. Prediction of ultrasonic signals from the various sizes of side drilled holes to find equivalent flaws, determination of the size of the nonregular TOFD indications from the coarse grains, and experimental investigation of TOFD signals from coarse grain and shrinkage geometry to identify PWSCC signals were performed. From the computer simulation and experimental investigation results, it was possible to obtain the nonregular TOFD indications from the coarse grains in the alloy 690 penetration tube of RVHP; these nonregular indications may be classified as PWSCC. By comparing the computer simulation and experimental results, we were able to confirm a clear difference between the coarse grain signal and the PWSCC signal.

• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.28-33
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• 2010

As a part of license renewal for the continued operation of Wolsong unit 1, the periodic safety review report was submitted near the end of design lifetime, 2012, and now is under reviewing. Major components of primary system such as pressure tubes, feeder pipes and so on are being replaced and many components of secondary system are also being repaired. So the license renewal of Wolsong unit 1 is expected to be acquired without significant issues. But on the other hand, steam generators of Wolsong unit 1 had the good performance and therefore the replacement and repair for the steam generator are not needed. Recently it is reported that some cracks were detected in a few of european steam generator with Alloy 800 tubes and the cause of cracks was the outer diameter stress corrosion cracking(ODSCC) due to the concentration of chemical impurities on the outer surface of tube. Accordingly the overall review on this issue was performed. The long-term operation is likely to results to form the concentration mechanism for the tube corrosion as the sludge build-up in the secondary side of steam generator and the crack in the crevice between tube and tube-sheet and expansion transitions is apt to be occurred. In this paper, the history of steam generator inspection and operation of Wolsong unit 1 are reviewed and the reliability of steam generator tube is evaluated and the steam generator aging management program for Wolsong unit 1 is introduced.

• Corrosion Science and Technology
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• v.21 no.1
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• pp.62-67
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• 2022

The objective of this study was to perform failure analysis of an inlet pipe located in a governor valve of a steam turbine in a district heating system. During the operation, the temperature of the governor valve was increased to as high as ~500 ℃, which induced thermal expansion of the inlet pipe along both axial and radial directions. While the inlet pipe did not have contact with the valve seat, the side plane of the upside was constrained by the casing part, which led the inlet pipe to experience stress field in the form of fatigue and creep. The primary crack was initiated at about 30 mm below the top where the complex stress field was anticipated. These results suggest that the main failure mechanism is a combination of thermal fatigue and creep during the operation supported by the observation of apparent beach marks on the fracture surface and pores near the cracks, respectively.

• 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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.9
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• pp.1047-1052
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• 2012

Primary water stress corrosion cracking (PWSCC) instances have been reported in the Alloy 600 reactor pressure vessel head penetration nozzle and the Alloy 82/182 dissimilar metal butt weld nozzle in several PWRs. Therefore, in-service inspection programs have been adopted worldwide to prevent failure at the weld region. If a PWSCC is observed at the dissimilar metal weld region during inspection, its structural integrity should be evaluated; however, this requires considerable time and effort, and this might lead to a decrease in the plant utilization coefficient. To prevent this, KHNP-CRI have established integrity assessment criteria and developed a computer program for the fast evaluation and judgment of PWSCC. In this paper, the results and current status of the same are presented. Through this study, criteria for the structural integrity evaluation of PWSCC have been established, and a computer program has been developed to realize technical means for the evaluation of PWSCC structural integrity.

• Korean Journal of Materials Research
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• v.9 no.9
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• pp.919-925
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• 1999

Grain boundary characteristics and corrosion behavior of Alloy 600 material were investigated using the concept of grain boundary control by thermomechanical treatment(TMT). The grain boundary character distribution (GBCD) was analyzed by electron backscattered diffraction pattern. The effects of GBeD variation on intergranular at tack(JGA) and primary water stress corrosion cracking(PWSeC) were also evaluated. Changes in the fraction of coinci dence site lattice(CSL) boundaries in each cycle of TMT process were not distinguishable, but the total eSL boundary frequencies for TMT specimens increased about 10% compared with those of the commercial Alloy 600 material. It was found from IGA tests that the resistance to IGA was improved by TMT process. However, it was found from PWSCC test that repeating of TMT cycles resulted in the gradual decrease of the time to failure and the maximum load due to change in grain boundary characteristics, while the average crack propagation rate of primary crack increased mainly due to suppression of secondary crack propagation. It is considered that these corrosion characteristics in TMT specimens is attributed to 'fine tuning of grain boundary' mechanism.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.20 no.9
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• pp.2894-2900
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• 1996

The stream generator tubes represent an integral part of a major barrier against the fission product release to the environment. So, the rupture of these tubes could permit flow of reactor coolant into the secondary system and injure the safety of reactor coolant system. Therefore, if the crack was detected during In-Service Inspection of tubes the cracked tube should be evaluated by the pulgging criteria and plugged or not. In this study, the fracture mechanics evaluation is carried out on the thru-wall axial crack due to Primary Water Stress Corrosion Cracking in the roll transition aone of steam generator tube to help the assurence the integrity of tubes and estabilish the plugging criteria. Due to the Inconel which is used as tube material is more ductile than others, the plastic instability repture theory was used to calculate the critical and allowable crack length. Based on Leak Before Break concept the leak rate for the critical crack length and the allowable leak rate are compared and the safety of tubes was given.