• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.51-55
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• 2014

Based on the results of Probabilistic Safety Assessment(PSA) for a Nuclear Power Plant (NPP), Common Cause Failure(CCF) events have been recognized as one of the main contributors to the risk. Also, the CCF data and estimation method used in domestic PSA models have been pointed out as an issue with respect to the quality. The existing method of MGL and non-staggered testing even widely used were considered conservative in estimating the safety and had a limited capability in uncertainty analyses. Therefore, this paper presents the CCF estimation using a new generic data source and Alpha factor method. The analyses showed that Alpha factor and staggered method are effective in estimating the CCF contribution and risk insights of reference plant. This method will be a common bases for the optimization of new design for the construction plants as well as for the updating of safety assessment on the operating nuclear power plants.

• Journal of Korean Institute of Industrial Engineers
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• v.4 no.2
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• pp.59-66
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• 1978

This paper is a work scheduling for design of piping Department of chemical plant using accumulated curve. Accumulated curve prepared by analysis of performance data, collected executed manhours for chemical plant of "D" Company during the past two years. It compared scheduled manhours with actual used manhours up to six months, put into the form of figures and charts the results can he summarized as below; 1) It can he found an important factor of critical control thus piping department got 30% of total scheduled menhours. 2) A plan of manpower mobilization can be scheduled before work starting. 3) Project progress can he found easily as put into the form of figures and charts for schedule to actual.

• International Journal of Reliability and Applications
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• v.5 no.1
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• pp.25-36
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• 2004

Ultrasonic inspection system consist of the examination procedures, equipment, and operators. The reliability of nondestructive testing is influenced by the inspection environment, materials and types of defect. It is very difficult to estimate the reliability of NDT due to the various factors. Piping inspection round robin was conducted to quantify the capability of ultrasonic inspection during in-service. In this study, the models for predicting the ultrasonic NDE reliability by logistic model and linear regression model are discussed. The utility of the NDT reliability assessment is verified by the analysis of the data from round robin test with these models.

• Transactions of the Korean Society of Mechanical Engineers
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• v.14 no.5
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• pp.1181-1185
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• 1990

A failure evaluation diagram to evaluate fatigue fracture was developed. The relation between the fatigue limit and the threshold stress intensity factor for the short-cracked specimens of various materials including a piping carbon steel can be rationally predicted by the proposed method. It is shown that the coupled failure evaluation diagram for fatigue and ductile fracture is expecially useful for evaluation of the flaw tolerance as well as the margin of the safety of the pressure vessel and piping. Further, accumulation of fatigue data will be needed to construct an accurate fatigue failure evaluation diagram.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.1
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• pp.44-51
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• 2023

The RCP is one of the main components in nuclear power plants and plays an important role in circulating coolant to the RCS system. Currently, nuclear plants are monitored using various monitoring systems. However, since they operate independently according to their functional purpose, it is not able to analyze vibration and operation/performance information comprehensively, and thus failure diagnosis accuracy is limited. In addition, these systems do not provide some important information (such as fault type, parts and cause) necessary for emergency actions, but provide only alarm information. To improve these technical problems, this study proposes a diagnosis technique (M/L, Rule-based model, Data-driven model, Narrow band model) and methodology for comprehensive analysis.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.2
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• pp.117-129
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• 2023

Hydrogen embrittlement of a pipe is an important factor in hydrogen transport. To characterize hydrogen embrittlement, tensile and fracture toughness tests should be conducted. However, in the case of hydrogen-embrittled materials, it is difficult to perform tests in hydrogen environment, particularly at low temperatures. It would be useful to develop a methodology to predict the fracture toughness of hydrogen-embrittled materials at low temperatures using more efficient tests. In this study, the fracture toughness of API X52 steels in hydrogen at low temperatures is predicted from numerical simulation using coupled finite element (FE) damage analyses with FE diffusion analysis, calibrated by analyzing small punch test data.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.15-22
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• 2011

The chemistry factor and RTPTS margin for domestic reactor pressure vessel materials were analyzed by using the surveillance data which have been obtained from 8 nuclear power plants in Korea. The surveillance data have been used to assess the integrity of the pressure vessel under the pressurized thermal shock (PTS) event. The chemistry factor, which is determined by the Cu and Ni contents of vessel materials, is considered a proper tool to assess the $RT_{PTS}$. The chemistry factors, which were obtained from the surveillance data of domestic reactor pressure vessels, were investigated and compared with those of Regulatory Guide 1.99 in this study. Regressions for ${\Delta}RT_{NDT}$ were performed to expect the chemistry factor as a function of Cu and Ni, and to estimate $RT_{PTS}$ margin. The margin analysis was performed by comparing the regression graphs and standard deviations with those of Regulatory Guide 1.99. The standard deviations calculated by using the domestic surveillance data for base metal and welds are almost same as the standard deviations which are suggested on Regulatory Guide 1.99, Rev.2.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2004.05a
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• pp.389-394
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• 2004

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.2
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• pp.40-49
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• 2016

Flow accelerated corrosion (FAC) of the carbon steel piping has been a significant problem in nuclear power plants. FAC occurs under certain hydrodynamic, environmental, and material conditions, and extensive research into the factors of FAC has been conducted. The basic process of FAC is now relatively well understood; however, a full mechanistic model has not yet been established. Recently, the Korea Atomic Energy Research Institute (KAERI) has built a large experiment loop system for FAC. To produce significant experimental results using this system, the factors affecting on FAC should be analyzed quantitatively, and a model needs to be developed. In this work, a statistical modeling methodology to develop an empirical model is described in detail, and a preliminary model is suggested. Firstly, FAC data were collected from the research literature in Japan and the results of domestic experiments. The flow rate, water temperature, pH at room temperature, and the Cr content are selected as major factors, and nonlinear regression is used to find the best fit of the available data. An iterative procedure between suggesting and evaluating a model is used until an optimum model is obtained. The developed model gives the FAC rate comparable to the measured FAC rate. The developed model is going to be refined using additional laboratory data in the future.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.1
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• pp.27-35
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• 2023

Pressure tubes are the main components of PHWR core and serve as the pressure boundary of the primary heat transport system. However, because pressure tubes have changed their geometrical dimensions under the severe operating conditions of high temperature, high pressure and neutron irradiation according to the increase of operation time, all dimensional changes should be predicted to ensure that dimensions remain within the allowable design ranges during the operation. Among the deformations, the diameter expansion due to creep leads to the increase of bypass flow which may not contribute to the fuel cooling, the decrease of critical channel power and finally the deration of the power to maintain the operational safety margin. This study is focused on the modeling of the expansion of the pressure tube diameter based on the operating conditions and measured diameter data. The pressure tube diameter expansion was modeled using the neutron flux and temperature distributions of each fuel channel and each fuel bundle as well as the measured diameter data. Although the basic concept of the current modeling approach is simple, the diameter prediction results using the developed methodology showed very good agreement with the real data, compared to the existing methodology.