• 비파괴검사학회지
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• 제9권1호
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• pp.30-38
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• 1989

Since the nuclear power plants in Korea have been operated in 1979, the nondestructive testing (NDT) of pressure vessels and/or piping welds plays an important role for maintaining the safety and integrity of the plants. Ultrasonic method is superior to the other NDT method in the viewpoint of the detectability of small flaw and accuracy to determine the locations, sizes, orientations, and shapes. As the service time of the nuclear power plants is increased, the radiation level from the components is getting higher. In order to get more quantitative and reliable results and secure the inspector from the exposure to high radiation level, automation of the ultrasonic equipments has been one of the important research and development(R & D) subject. In this research, it was attempted to visualize the shape of flaws presented inside the specimen using a Modified C-Scan technique. In order to develope Modified C-Scan technique, an automatic ultrasonic scanner and a module to control the scanner were designed and fabricated. IBM-PC/XT was interfaced to the module to control the scanner. Analog signals from the SONIC MARK II were digitized by Analog-Digital Converter(ADC 0800) for Modified C-Scan display. A computer program has been developed and has capability of automatic data acquisition and processing from the digital data, which consist of maximum amplitudes in each gate range and locations. The data from Modified C-Scan results was compared with shape from artificial defects using the developed system. Focal length of focused transducer was measured. The automatic ultrasonic equipment developed through this study is essential for more accurate, reliable, and repeatable ultrasonic experiments. If the scanner are modified to meet to appropriate purposes, it can be applied to automation of ultrasonic examination of nuclear power plants and helpful to the research on ultrasonic characterization of the materials.

• 복식문화연구
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• 제15권5호
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• pp.902-915
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• 2007

This study examines decorative features found in the traditional costume of the Miao, one of the minority groups in China, and adapts them for the development of modern Korean-style fashion designs in order to present visual sample data and to introduce techniques. As for the research method, both literature and visual data are examined: the former includes Miao-related books and theses; the latter, pictures from various documents and real-life data in the possession of Seoul Women's University. The study result is as follows: 1. The basic structure of Miao costume is consist of a Caftan-type upper garment and an accentuated lower garment, pleats skirt although they are decorated luxuriously with various techniques such as delicate embroidery and dyeing. 2. Details and trimmings in modern shapes are developed and presented by adding modern materials to satin, twists, cross-stitch embroidery, stripe patterns, geometrical patchwork, machine pleats and smoking, trimmings of fringe & feather, and the border line decoration of coloration, braids, piping, embroidery, belts tied and wound with colorful threads. 3. Detachable accessaries such as front reinforcement strips, back reinforcement strips, belts, decorative aprons and patchwork skirts are developed and designed by adapting highlighted layered effects. 4. The silver decoration used as an necessary by the Miao is suitable for modern fashion design in futuristic sense. Various techniques and decorative features of Miao costume display their true value all the more in modern fashion stream in which ethnic trends draw much attention.

• Nuclear Engineering and Technology
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• 제44권3호
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• pp.297-310
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• 2012

Aging-related degradation of nuclear power plant components is an important aspect to consider in securing the long term safety of the plant, especially the seismic safety, since the degradation of the components affects not only their seismic capacity but their response. This can cause a change in the seismic margin of a component and the overall seismic safety of a system. To better understand the status and characteristics of degradation of components in Nuclear Power Plants (NPPs), the degradation occurrences of components in the U.S. NPPs were identified by reviewing recent publicly available information sources and the characteristics of these occurrences were evaluated and compared to observations from the past. Ten categories of components that are of high risk significance in Korean NPPs were identified, comprising anchorage, concrete, containment, exchanger, filter, piping systems, reactor pressure vessels, structural steel, tanks, and vessels. Software tools were developed to expedite the review process. Results from this review effort were compared to previous data in the literature to characterize the overall degradation trends.

• Steel and Composite Structures
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• 제18권1호
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• pp.29-50
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• 2015

With a quasi-three point bending apparatus, ratcheting deformation is studied experimentally on a pressurized austenitic stainless steel Z2CND18.12N pipe under bending load and vertical displacement control, respectively. The characteristic of ratcheting behavior of straight pipe under both control methods is achieved and compared. The cyclic bending loading and internal pressure influence ratcheting behavior of pressurized straight pipe significantly under loading control and the ratcheting characteristics are also highly associated with the cyclic displacement and internal pressure under displacement control. They all affect not only the saturation of the ratcheting strain but the ratcheting strain rate. In addition, ratcheting simulation is performed by elastic-plastic finite element analysis with ANSYS in which the bilinear model, Chaboche model, Ohno-Wang model and modified Ohno-Wang model are applied. By comparison with the experimental data, it is found that the CJK model gives reasonable simulation. Ratcheting boundaries under two control modes are almost same.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.2119-2129
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• 2020

Flow-accelerated corrosion (FAC), liquid droplet impingement erosion (LDIE), cavitation and flashing can cause continuous wall-thinning in nuclear secondary pipes. In order to prevent pipe rupture events resulting from the wall-thinning, most NPPs (nuclear power plants) implement their management programs, which include periodic thickness inspection using UT (ultrasonic test). Meanwhile, it is well known in field experiences that the thickness measurement errors (or deviations) are often comparable with the amount of thickness reduction. Because of these errors, it is difficult to estimate wall-thinning exactly whether the significant thinning has occurred in the inspected components or not. In the previous study, the authors presented an approximate estimation procedure as the first step for thickness measurement deviations at each inspected component and the statistical & quantitative characteristics of the measurement deviations using plant experience data. In this study, statistical significance was quantified for the current methods used for wall-thinning determination. Also, the authors proposed new estimation procedures for determining local wall-thinning to overcome the weakness of the current methods, in which the proposed procedure is based on analysis of variance (ANOVA) method using subgrouping of measured thinning values at all measurement grids. The new procedures were also quantified for their statistical significance. As the results, it is confirmed that the new methods have better estimation confidence than the methods having used until now.

• 대한기계학회논문집A
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• 제26권11호
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• pp.2442-2449
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• 2002

This paper provides an reference stress based J estimation equation fur cylinders with finite internal axial surface cracks under internal pressure. In part 1, the J estimation equation based on deformation plasticity using Ramberg-Osgood (R-0) materials is proposed. In this paper, the developed CE/EPRI -type solutions ale then re-formulated based on the reference stress concept. Such a re-formulation provides a simpler equation for J. estimation are then further extended to combined internal pressure and bending. The proposed reference stress based J estimation equation is compared with elastic-plastic 3-D FE results using actual stress-strain data for a Type 304 stainless steel. Good agreement between the FE results and the proposed reference stress based J estimations provides confidence in the use of the proposed method to elastic -plastic fracture mechanics of pressurised piping.

• 한국정밀공학회지
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• 제18권6호
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• pp.67-73
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• 2001

Three dimensional, elastic-plastic finite element(FE) analyses for circumferential through-wall cracked pipes under combined tension and bending are performed using actual tensile data of stainless steel, for two purposes. The first one is to validate the recently-proposed enhanced reference stress (ERS) method to estimate the J-integral and COD for circumferential through-wall cracked pipes under combined tension and bending. The second one is to compare those results with the GE/EPRI estimations. The FE results of the J-integral and the COD, resulting from six cases of proportional and non-proportional combined tension and bending, compare very well with those estimated from the proposed method. Excellent agreements of the proposed method with the detailed FE results provide sufficient confidence in the use of the proposed method to the Leak-Before-Break(LBB) analysis of through-wall cracked pipes under combined tension and bending.

• 한국지반공학회:학술대회논문집
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• 한국지반공학회 2010년도 추계 학술발표회
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• pp.60-68
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• 2010

In this study, the seepage flow monitoring method by hydaulic head loss rate graph was developed for the purpose of monitoring the seepage flow from the see side or from the lake on sea dike in which seepage force was varied periodically. The hydraulic head loss rate was defined in this method. The value of the rate is in the range from 0 to 1. the value of 0 means perfectly free flow of seepage. the value of 1 means perfect waterproofing. The value of coefficient of determination in the hydraulic head loss rate graph closer to 1 means that the seepage flow way is stable. The value of coefficient of determination in the hydraulic head loss rate graph closer to 0 means that the hole may exist or the piping may be in the progress. The pore water pressure data measured in Saemangeum sea dike was analyzed with the developed method The result showed that the variation of seepage flow state was detected sensitively by this method and the interception effect of sea dike could be estimated quantitatively.

• 한국해양공학회지
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• 제16권3호
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• pp.54-58
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• 2002

This study investigates a relationship between fracture mechanics parameters (Stress Intensity Factor Range: ΔK, Maximum Stress Intensity Factor; Kmax) and X-ray parameters (residual stress:$\sigma$r half-value breadth: B) for SG365 steel at elevated temperature up to 30$0^{\circ}C$. The fatigue crack propagation test were carried out and X-ray diffraction technique according to the direction of crack length was applied to fatigue fractured surface. The residual stress on the fracture surface was found to increase at low ΔK region, to reach a maximum value at a certain value of Kmax or ΔK and then to decrease. Residual stress was independent of stress ratio by arrangement of ΔK and half value breadth were independent of the arrangement of Kmax. The equation of $\sigma$r-ΔK was established by the experimental data. therefore, fracture mechanics parameters could be estimated by the measurement of X-ray parameters.

• 한국분무공학회지
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• 제16권2호
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• pp.82-89
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• 2011

When pipe components made of carbon steel in nuclear, fossil, and industry plants are exposed to flowing fluid, wall thinning caused by FAC(flow accelerated corrosion) can be generated and eventually ruptured at the position of pressure boundary. The aim of this study is to identify the locations at which local wall thinning occurs and to determine the turbulence coefficient related to local wall thinning. Experiment and numerical analyses for the tee sections of down scaled piping components were performed and the results were compared. In particular, flow visualization experiment which was used alkali metallic salt was performed to find actual location of local wall thinning inside tee components. In order to determine the relationship between turbulence coefficients and local wall thinning, numerical analyses were performed for tee components in the main feedwater systems. The turbulence coefficients based on the numerical analyses were compared with the local wall thinning based on the measured data. From the comparison of the results, the vertical flow velocity component(Vr) flowing to the wall after separating in the wall due to the geometrical configuration and colliding with the wall directly at an angle of some degree was analogous to the configuration of local wall thinning.