• Journal of Power System Engineering
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• v.10 no.3
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• pp.38-44
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• 2006

Failure assessment of steam generator tube are very important for the integrity of energy plants. In pipes of energy plants, sometimes, the local wall thinning may result from severe erosion-corrosion damage. Recently, the effects of local wall thinning on fracture strength and fracture behavior of piping system have been well studied. In this paper, the elasto-plastic analysis is performed by FE code ANSIS on steam generator tube with wall thinning. We evaluated the failure mode, fracture strength and fracture behavior from FE analysis. It was possible to predict the crack initiation point by estimating true fracture ductility under multi-axial stress conditions at the center of the thinned area.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2004.10a
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• pp.245-253
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• 2004

There has been recent demand for extending the life of age-degraded structures and equipment by such techniques as diagnosis, maintenance, safety assessment, and estimating residual life on iron-making plants and hydraulic, thermal, and nuclear power plants. These techniques take into account comprehensive scenarios that may cause malfunction and structural damage and allow an assessment of risk based on the likely scenarios. In particular the safety assessment and residual life estimation of age-degraded ships and equipment facilities require consideration of various factors such as mechanical and thermal stresses, corrosion, hardness, load variation due to changes of operating condition, crack generation and strength reduction of material by fatigue. In this study, a detail thermal stress analysis, one of useful techniques of safety assessment and maintenance, is performed on a blast furnace by using general FEM code (MSC/NASTRAN).

• 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.

• Journal of the Korean Society for Railway
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• v.3 no.2
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• pp.77-83
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• 2000

This study proposes the integrity evaluation of the bogie frame using ultrasonic fractography analysis. Analysis objectives in this study are to investigate fracture planes of damaged zone by the A-scan method. The surface condition of fracture planes shows degree of degradation by the stress concentration. The detection of the natural defects in the bogie frame is performed using the characteristics of echodynamic pattern in ultrasonic signal. Results of ultrasonic testing agree fairly well with those of actual fracture plane. In quantitative fractography analysis, microstructures of actual fracture plane turned out to be intergranular and transgranular fracture. Proposed ultrasonic fractography analysis in this study can be used for the integrity evaluation of the bogie frame.

• 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.

• 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.

• Journal of the Korean institute of surface engineering
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• v.49 no.4
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• pp.389-394
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• 2016

The operation method of nuclear power plants is currently changing to high burn-up and long period that can enhance economics and efficiency of the plant. Since nuclear plant operation environment has been becoming severe, the amount of absorbed hydrogen also has increased. Absorbed hydrogen can be fatal securing safety of nuclear fuel cladding in case of Loss of Coolant Accidents(LOCA). In order to examine the impact of hydride on high-temperature oxidation, high-temperature oxidation experiment was performed on normal Zry-4 cladding and on Zry-4 cladding where hydrogen is charged in air pressure steam atmosphere under the $950^{\circ}C$ and $1000^{\circ}C$. According to the results, while oxidation acceleration due to charged hydrogen was not observed prior to breakaway oxidation creation, oxidation began to accelerate in cladding where hydrogens charged as soon as the breakaway oxidation started. If so much hydrogen are charged in the cladding, equiaxial monoclinic phase to unstable of stress is formed and it is presumed that oxidation is accelerated because nearby stress caused a crack in equiaxial phase, and that makes corrosion resistance decline sharply.

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.3
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• pp.248-255
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• 2009

After several damages by PWSCC were found in the world, USNRC and PNNL(Pacific Northwest National Laboratory) started the research on PWSCC under the project name of PINC. The aim of the project was 1) to fabricate representative NDE mock-ups with flaws to simulate PWSCCs, 2) to identify and quantitatively assess NDE methods for accurately detecting, sizing and characterizing PWSCCs, 3) to document the range of locations and morphologies of PWSCCs and 4) to incorporate results with other results of ongoing PWSCC research programs, as appropriate. Korea nuclear industries have also been participating in the project. Thermally and mechanically cracked-four mockups were prepared and phased array and manual ultrasonic testing(UT) techniques were applied. The results and lessons learned from the preliminary RRT are summarized as follows: 1) Korea RRT teams performed the RRT successfully. 2) Crack detection probability of the participating organizations was an average 87%, 80% and 80% respectively. 3) RMS error of the crack sizing showed comparatively good results. 4) The lessons learned may be helpful to perform the PINC RRT and PSI /ISI in Korea in the future.

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.2
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• pp.153-161
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• 2009

After several PWSCCs were found in Bugey(France), Ringhals(Sweden), Tihange(Belgium), Oconee, Arkansas, Crystal Fever, Davis-Basse, VC Summer(U.S.A.), Thuruga(Japan), USNRC and PNNL started the research on PWSCC, that is, PINC project. The aim of this project is to fabricate and obtain representative NDE mock-ups with flaws to simulate tight PWSCC cracks, to identify and quantitatively assess NDE methods for accurately detecting, sizing and characterizing tight cracks such as PWSCC, to document the range of locations and crack morphologies associated with PWSCC and observed responses and to incorporate findings from other ongoing PWSCC research programs, as appropriate. By participating in PINC project, Korean morphology technique about PWSCC and NDE technique have improved and become similar lever with other advanced country. Therefore, the evaluation technique of integrity for nickel alloy component has been improved by cooperation with university, research institute and industries.

• Proceedings of the KWS Conference
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• 2009.11a
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• pp.100-100
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• 2009

This paper is concerned with numerical analyses of residual stresses in welds and material's susceptibility to stress corrosion cracking (SCC) for the primary piping system in nuclear power plants: Both the dissimilar metal weld (DMW) for stainless steel to low alloy steel joints and the similar metal weld (SMW) for forged stainless steel to cast stainless steel joints are considered. Thermal elasto-plastic analyses using the finite element method (FEM) are performed to predict residual stresses generated in fabrication welding and its related processes for both the DMW and SMW, including effects of quenching for cast stainless steel piping, machining of the DMW root, and grinding of the SMW root. As a result, the effect of quenching should be included in the evaluation of residual stresses in the SMW for the cast stainless steel piping. It is deemed that residual stresses in both the DMW and SMW would not affect the SCC susceptibility of the welds providing that the welding processes are completed without any weld repair on the inside wall of the joint. However, the grinding process if performed on the safe-end to piping weld, would produce a high level of residual stresses in the inner surface region and thus a stress improvement process (e.g. buffing) should be considered to reduce susceptibilities to SCC.