• Journal of Welding and Joining
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• v.33 no.1
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• pp.87-93
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• 2015

This study was performed to develop repair technology for J-Groove dissimilar metal weld of pressurizer heater sleeve in nuclear power plants. Pad, J-Groove automatic welding and mechanical machining equipments to develop repair technology using 'Half Nozzle Repair' were designed and manufactured. To obtain the optimum welding process parameters during Pad temperbead overlay welding, several welding experiments using Taguchi method were conducted. Weldability of Pad overlay weld specimens was estimated by PT/RT test, FE-SEM, EDS and Vickers hardness test. Also, J-Groove welding to adjust weld shape conditions requiring in ASME Code was carried out and its integrity of weld specimens was evaluated through PT/RT test and optical microscope. Consequently, it was revealed that Pad and J-Groove overlay welding for dissimilar metal weld of pressurizer heater sleeve could be possible to meet Code standard without weld defect.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.15-19
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• 2014

A Failure or degradation of reactor upper head penetration is a recurring problem due to long term operation at nuclear power plants. And a flaw in the reactor upper head penetration has caused unplanned plant shutdown for repair as well as high economic impact on the plants. Consequently, a detection of flaws is of the utmost importance. Prior to the replacement of reactor upper head penetration, some utilities have repaired the flaws of reactor upper head penetration generated by overlay weld. Until now, only the base metal in reactor upper head penetration has been inspected according to 10 CFR 50.55a and ASME code case N-729-1. Accordingly, it is difficult to detect manufacturing defects and repair defects in overlay weld. This paper presents a case study on the application of Time of Flight Diffraction technique for reactor head penetration mockup with artificial flaws in overlay weld. This study offers a way to understand the flaws detected in reactor upper head penetration overlay weld.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.94-97
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• 2008

In the welding process, weldments usually include repair weld during the manufacturing process. Repair welds is supposed to cause strong tensile residual stress. Moreover weldments, usually made by Alloy 82/182, is susceptible to PWSCC. Therefore, mitigation of welding residual stress in weldments is important for reliable operating. PWOL is one of the methods for mitigation and verified for over twenty years. In this paper, residual stress distribution of repaired weldments and the effect of PWOL on mitigation is examined for surge nozzle.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.22 no.6
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• pp.557-564
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• 2009

Welding residual stress is occurred after welding process. Tensile residual stress is one factor of PWSCC. Repair welding usually happened during the manufacturing welding process. Repair welds cause strong tensile residual stress. In PWR, Repair weldments made by Alloy 82/182 is susceptible to PWSCC caused by tensile stress, material and environment. Therefore, mitigation of welding residual stress in weldments is important for reliable operating. PWOL is one of the methods for mitigation and verified for over twenty years. In this paper, residual stress distribution of repaired weldments and the effect of PWOL on mitigation is examined for surge nozzle.

• 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.

• Nuclear Engineering and Technology
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• v.32 no.1
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• pp.67-76
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• 2000

Significant wall thinning due to flow accelerated corrosion(FAC）was recently reported in main feedwater pipes in 3 Korean pressurized water reactor(PWR) plants. The main feedwater pipes in one plant were repaired using overlay weld method at the outside of pipe, while those in 2 other plants were replaced with new pipes. In this study, the effect of the wall thinning in the main feedwater pipes on piping integrity was evaluated using finite element method. Especially, the effects of both the overlay weld repair and the stress concentration in notch-type thinned area on the piping integrity were investigated. The results are as follows : (1) The piping load carrying capacity may significantly decrease due to FAC. In special, the load carrying capacity of the main feedwater pipe was reduced by about 40% during about 140 months operation in Korean PWR plants. (2) By performing overlay weld repair at the outside of pipe, the piping load carrying capacity can increase and the stress concentration level in the thinned area can be reduced.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.534-537
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• 2008

In recent years, the dissimilar metal, Alloy 82/182 welds used to connect stainless steel piping and low alloy steel or carbon steel components in nuclear reactor piping system have experienced cracking due to primary water stress corrosion(PWSCC). It is well known that one reason of the cracking is the residual stress by the weld. But, it is difficult to estimate exactly weld residual stress due to many parameters of welding. In this paper, the analysis of 3 FEM models made by ABAQUS Code is performed to estimate exactly the weld residual stress on the dissimilar metal weld. 3 FEM models are Butt model, Repair model and Overlay model and are the plane.strain 2D model. The thermal analysis and the stress analysis are performed on each model and the residual stresses on each model were calculated and compared respectively. Also, the specimen of Butt model was made and the residual stresses were measured by X-Ray method and Hole Drilling Technique. These results were compared with the FEM result of Butt model.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3837-3851
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• 2021

To investigate how the fabrication and repair of socket welded joints could be used to enhance fatigue resistance under vibration condition, experimental test data of installation conditions that potentially influence fatigue strength were analyzed with the S-N curve. It was found that the decreasing fatigue strength of stainless steel socket welded joints was attributed to the effect of high heat input of welding process. The effect of welding method, slip-on gap and radial-gap conditions on fatigue strength was insignificant. The test data of repair technique application, 2 × 1 leg length and of socket weld overlay, clearly showed higher fatigue strength but there was a limitation for higher stress region because of the weld toe crack.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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• pp.51-60
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• 2015

Local wall thinning and integrity degradation caused by several mechanisms, such as flow accelerated corrosion, cavitation, flashing and/or liquid droplet impingement, is a main concern in secondary steam cycle piping system of nuclear power plants in terms of safety and operability. Thinned pipe management program (TPMP) has being developed and optimized to reduce the possibility of unplanned shutdown and/or power reduction due to pipe failure caused by wall thinning. In this paper, newest technologies, standards and regulations related to the integrity assessment, repair and replacement of thinned pipe component are reviewed. And technical improvement items in TPMP to secure the reliability and effectiveness are also presented.