• 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.11 no.2
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• pp.1-5
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• 2015

A failure or degradation of reactor upper head penetration is a troublesome problem at Nuclear Power Plants. A flaw in the reactor upper head penetration can result in unplanned plant shutdown for repair, and cause serious economic losses on the plants. Consequently, a detection of flaws is a matter of more importance. Until now, only the base metal, not including J-groove weld, in reactor upper head penetration has been inspected in accordance with 10 CFR 50.55a and ASME code case N-729-1 requirements. Accordingly, it is rather difficult to detect manufacturing defects and repair defects in J-groove weld. This paper presents a case study on the application of Time of Flight Diffraction UT technique to examine the J-groove weld in reactor head penetration using reactor head penetration mockup with artificial flaws. We expect that this study result will offer a way to understand the non-destructive examination technology for J-groove weld in reactor upper head penetration.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.100-109
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• 2020

Pipelines subjected to ground movement would be easily exposed to large-scale deformation. Since such deformations may cause the pipeline failure, it is important to ensure the safety of pipelines in various operation conditions. However, crack in weld metal have been considered as one of the main causes that can deteriorate the structural integrity of the pipeline. For this reason, the structural integrity of the pipe containing the crack in the weld should be obtained. In order to assess cracked pipe, J-integral and crack-tip opening displacement(CTOD) have been applied widely as the elastic-plastic fracture mechanics parameters representing crack driving force. In this study, engineering solutions to calculate the J-integral and CTOD of pipes with a circumferential outer surface crack in the weld are proposed. For this purpose, 3-dimensional elastic-plastic finite element(FE) analyses have been performed considering the effect of overmatch and width of weld. The shape of the weld was simplified to I-groove, and axial displacement was employed as for loading condition. Based on FE results, the effects of crack size, material properties and width of weldment on J-integral and CTOD were investigated. Additionally, the J-integral and CTOD for I-groove were compared with those for V-groove to examine the effects of the weld shape, and a proportionality coefficient of J-integral and CTOD was calculated from the results of this paper.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.20-27
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• 2010

In order to prevent the corrosion of component contacted primary water designed alloy 600 material in the nuclear power plant. But the primary water stress corrosion cracking(PWSCC) of alloy 600 and weld area occurs continuously due to the residual stress. The leakage accident resulted from PWSCC in the drain nozzle of the steam generator of domestic power plants. Heater sleeves of the pressurizer are welded with alloy 600 weld material and therefore exposed to the primary water environment. PWSCC occurred in heater sleeve material and weld area of many foreign power plants. The current issue of domestic nuclear power plants are consequently concentrated to PWSCC of similar material. In order to improve the detection and the sizing of the PWSCC in the welding sleeve of the pressurizer, the automatic UT system and multi-directions probe sets have been developed. The experimental studies have been performed using the mock-up block containing artificial reflectors(ID connected EDM notch) and semi-artificial cracks made from thermal fatigue. The automatic UT System is applied in the detection and the length sizing of the ID/OD on the tube and the J-groove weld area of the artificial reflectors and results of the detection and the sizing are compared respectively. Also, the developed automatic UT system is successfully accomplished to inspect the heater sleeve and the J-groove weld area on the pressurizer for the detection of PWSCC.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.10
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• pp.1115-1130
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• 2012

In nuclear power plants, the reactor pressure vessel (RPV) upper head control rod drive mechanism (CRDM) penetration nozzles are fabricated using J-groove weld geometry. Recently, the incidences of cracking in Alloy 600 CRDM nozzles and their associated welds have increased significantly. The cracking mechanism has been attributed to primary water stress corrosion cracking (PWSCC), and it has been shown to be driven by welding residual stresses and operational stresses in the weld region. The weld-induced residual stress is the main factor contributing to crack growth. Therefore, an exact estimation of the residual stress is important for ensuring reliable operation. This study presents the residual stress computation performed for an RPV CRDM penetration nozzle in Korea. Based on two and three dimensional finite element analyses, the effect of welding variables on the residual stress variation is estimated for sensitivity analysis.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.6
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• pp.637-647
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• 2014

In pressurized water reactors, the upper head of the reactor pressure vessel (RPV) contains numerous control rod drive mechanism (CRDM) nozzles. These nozzles are fabricated by welding after being inserted into the RPV head with a room temperature shrink fit. The tensile residual stresses caused by this welding are a major factor in primary water stress corrosion cracking (PWSCC). Over the last 15 years, the incidences of cracking in alloy 600 CRDM nozzles have increased significantly. These cracks are caused by PWSCC and have been shown to be driven by the welding residual stresses and operational stresses in the weld region. Various measures are being sought to overcome these problems. The defects resulting from the welding process are often the cause of PWSCC acceleration. Therefore, any weld defects found in the RPV manufacturing process are immediately repaired by repair welding. Detailed finite-element simulations for the Korea Nuclear Reactor Pressure Vessel were conducted in order to predict the magnitudes of the repair weld residual stresses in the tube materials.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.2
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• pp.300-307
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• 2002

Partial penetration welding joint is defined as groove welds welded from one side, without steel backing or groove welds welded from both sides but without back gouging. So it has an unwelded portion at the root of the weld. Study of partial penetration weldment fracture behavior includes residual stress analysis and fracture analysis. The J-integral loses its path independency in residual stress field. Therefore, it is necessary to introduce a new J-integral, J, which is defined including the effect of plastic deformation and thermal strain. In this study, theoretical formulation and program were developed for the evaluation of J-integral for the crack tip located in the weldment. Evaluations of fracture behavior were performed for partial penetration multi-pass weldment of 25.4mm thick plate by J-integral. From a point of fracture in partial penetration multi-pass welding, it seemed to be better to control root face smaller than 6.35mm.

• Journal of Welding and Joining
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• v.22 no.5
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• pp.38-45
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• 2004

The present study is to investigate the effect of heat input on the microstructure, tensile properties and toughness of single-pass submerged arc bead-in-groove welds produced on SA508 class 3 steels. The heat input was varied in the range of 1.6, 3.2 and 5.0 kJ/mm. The toughness of weld metals was evaluated by using subsize Charpy V-notch specimens in the temperature range of -19$0^{\circ}C$ to 2$0^{\circ}C$. The weld microstructure and fractography were observed by optical and scanning electron microscopies, respectively. With increasing heat inputs, tensile strength and hardness of weld metals were decreased while elongation was increased. The poor notch toughness at 1.6 kJ/mm was attributed to the formation of ferrite with aligned second phase and banitic microstructure with high yield strength while that at 5.0 kJ/mm was due to the presence of grain boundary and polygonal ferrites. The microstructure of the intermediate energy input welds consisted of a high proportion of acicular ferrite with limited polygonal ferrites, which provide improved notch toughness.

• Journal of the Korean Society for Nondestructive Testing
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• v.5 no.2
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• pp.12-21
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• 1986

Generally, butt welds with plate thickness $30{\sim}40mm$ are welded with groove angle $40^{\circ},\;60^{\circ},\;70^{\circ}$, etc. In the detection of internal weld defects, oblique testing with single probe has been mainly used. But, recently, in acccordance with enlargement of welded structure, thick plate with 100-200mm are frequently required. Thus I-groove welding method was lately developed and often used. In this case, most frequently generated defects are the lack of weld penetration and incomplete fusion between base metal and welding material. If we would detect by oblique testing with single probe, detecting flaw is occassionally impossible or very underestimated. In this study, the limit for applying tandem method was studied in dise and band type flaws. The estimation of flaw size could be within 10% error compared to real flaws.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.10
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• pp.1337-1345
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• 2011

In pressurized water reactors (PWRs), the reactor pressure vessel (RPV) upper head contains numerous control rod drive mechanism (CRDM) nozzles. In the last 10 years, the incidences of cracking in alloy 600 CRDM nozzles and their associated welds has increased significantly. Several axial and circumferential cracks have been found in CRDM nozzles in European PWRs and U.S. nuclear power plants. These cracks are caused by primary water stress corrosion cracking (PWSCC) and have been shown to be driven by welding residual stresses and operational stresses in the weld region. Therefore, detailed finite-element (FE) simulations for the Korea Nuclear Reactor Pressure Vessel have been conducted in order to predict the magnitudes of the weld residual stresses in the tube materials. In particular, the weld residual stress results are compared in terms for nozzle location, geometry factor$r_o$/t, geometry of fillet, and adjacent nozzle.