• 제어로봇시스템학회:학술대회논문집
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• 2003.10a
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• pp.1244-1248
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• 2003

This paper describes visual tracking procedure of the underwater mobile robot for nuclear reactor vessel inspection, which is required to find the foreign objects such as loose parts. The yellowish underwater robot body tends to present a big contrast to boron solute cold water of nuclear reactor vessel, tinged with indigo by Cerenkov effect. In this paper, we have found and tracked the positions of underwater mobile robot using the two color information, yellow and indigo. The center coordinates extraction procedures are as follows. The first step is to segment the underwater robot body to cold water with indigo background. From the RGB color components of the entire monitoring image taken with the color CCD camera, we have selected the red color component. In the selected red image, we extracted the positions of the underwater mobile robot using the following process sequences; binarization, labelling, and centroid extraction techniques. In the experiment carried out at the Youngkwang unit 5 nuclear reactor vessel, we have tracked the center positions of the underwater robot submerged near the cold leg and the hot leg way, which is fathomed to 10m deep in depth.

• Journal of the Korean Society of Systems Engineering
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• v.9 no.1
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• pp.65-84
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• 2013

This paper introduces a systems engineering approach to reliability centered maintenance to address some of the weaknesses. Reliability centered maintenance is a systematic, disciplined process that produces an efficient equipment management strategy to reduce the probability of failure [1]. The study identifies the need for RCM, requirements analysis, design for RCM implementation. Value modeling is used to evaluate the value measures of RCM. The system boundary for the study has been selected as containment spray pump and its motor drive. Failure Mode and Criticality Effects analysis is applied to evaluate the failure modes while the logic tree diagram used to determine the optimum maintenance strategy. It is concluded that condition based maintenance tasks should be enhanced to reduce component degradation and thus improve reliability and availability of the component. It is recommended to apply time directed tasks to age related failures and failure finding tasks to hidden failures.

• Journal of the Korean Society for Nondestructive Testing
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• v.24 no.4
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• pp.362-370
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• 2004

In this work, the collapse moment due to wall-thinning defects is estimated by using fuzzy neural networks. The developed fuzzy neural networks have been applied to the numerical data obtained from the finite element analysis. Principal component analysis is used to preprocess the input signals into the fuzzy neural network to reduce the sensitivity to the input change and the fuzzy neural networks are trained by using the data set prepared for training (training data) and verified by using another data set different (independent) from the training data. Also, two fuzzy neural networks are trained for two data sets divided into the two classes of extrados and intrados defects, which is because they have different characteristics. The relative 2-sigma errors of the estimated collapse moment are 3.07% for the training data and 4.12% for the test data. It is known from this result that the fuzzy neural networks are sufficiently accurate to be used in the wall-thinning monitoring of elbows.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.471-476
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• 1996

In this study, the method to use the phase shift keying (PSK) communication technique in smart transmitter is presented. In nuclear applications. smart transmitters for various parameters are expected to improve the accuracy of measurement and to reduce the load of calibration work. The capability of communication in field level is the most important merit of the smart transmitter. The most popular method is using of digital and analog techniques simultaneously - transmitting measurements from the field at 4∼20mA while modulating the current to carry digital information in both directions over the same twisted pairs. Conventional smart transmitters use the frequency shift keying (FSK) method for digital communication. Generally, however, the FSK method has the speed limit at 1200 bps. Amount of information to transmit becomes increasing as the processing technique is improved. The PSK method is noticeable alternative for high speed digital communication, but it has non-zero DC component. In order to use the PSK method in the field transmission with smart transmitter, the method to remove the DC component is studied in this work.

• Nuclear Engineering and Technology
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• v.42 no.6
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• pp.608-619
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• 2010

A multi-scale analysis of water-cooled reactor thermalhydraulics can be used to take advantage of increased computer power and improved simulation tools, including Direct Numerical Simulation (DNS), Computational Fluid Dynamics (CFD) (in both open and porous mediums), and system thermalhydraulic codes. This paper presents a general strategy for this procedure for various thermalhydraulic scales. A short state of the art is given for each scale, and the role of the scale in the overall multi-scale analysis process is defined. System thermalhydraulic codes will remain a privileged tool for many investigations related to safety. CFD in porous medium is already being frequently used for core thermalhydraulics, either in 3D modules of system codes or in component codes. CFD in open medium allows zooming on some reactor components in specific situations, and may be coupled to the system and component scales. Various modeling approaches exist in the domain from DNS to CFD which may be used to improve the understanding of flow processes, and as a basis for developing more physically based models for macroscopic tools. A few examples are given to illustrate the multi-scale approach. Perspectives for the future are drawn from the present state of the art and directions for future research and development are given.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1863-1873
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• 2024

In the Korean domestic nuclear industry, to analyze the reliability of instrumentation and control (I&C) systems, the failure rates of the electronic components constituting the I&C systems are predicted based on the MIL-HDBK-217F standard titled 'Reliability Prediction of Electronic Equipment'. Based on these predicted failure rates, the mean time to failure of the I&C systems is calculated to determine the replacement period of the I&C systems. However, this conventional approach to the prediction of electronic component failure rates assumes that factors affecting the failure rates such as ambient temperature and operating voltage are static constants. In this regard, the objective of this study is to propose a prediction method for the remaining useful life (RUL) of electronic components considering mean time to failure calculations reflecting dynamic environments, such as changes in ambient temperature and operating voltage. Results of this study show that the RUL of electronic components can be estimated depending on time-varying temperature and electrical stress, implying that the RUL of electronic components can be predicted under dynamic stress conditions.

• Nuclear Engineering and Technology
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• v.9 no.3
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• pp.151-163
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• 1977

Being based on the Segal method, calculation was carried out for the natural uranium nuclear fuel cost with Zircaloy-4 cladding having design Parameters of Wolsung Nuclear Power Plant, CANDU-PHWR (Unit 1) , currently under construction in Korea aiming at its completion in 1982. An attempt was also made for tile sensitivity analysis of each fuel component; j. e., depreciation of fuel manufacturing plant caused by its life time, its load factor, production scale expansion of plant facilities, variations of construction and operating costs of fuel manufacturing plant, fluctuation of interest rates, extent of uranium ore price increases and effect of learning factor.

• Nuclear Medicine and Molecular Imaging
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• v.42 no.2
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• pp.88-97
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• 2008

In this review paper, basic configurations of gamma detectors in SPECT and PET systems were reviewed together with key performance parameters of the imaging system, such as the detection efficiency, the spatial resolution, the contrast resolution, and the data acquisition time for quick understanding of the system-component relationship and future design of advanced systems. Also key elements of SPECT and PET detectors, such as collimators, gamma detectors were discussed in conjunction with their current and future trend. Especially development trend of new scintillation crystals, innovative silicon-based photo-sensors and futuristic room-temperature semiconductor detectors were reviewed for researchers who are interested in the development of future nuclear medical imaging instruments.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.271-276
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• 1996

Coolant flow distribution to the assemblies and core coolant/component temperatures should be determined in LMR core steady state thermal-hydraulic performance analysis. Sodium flow is distributed to core assemblies with the overall goal of equalizing the peak cladding midwall temperatures for the peak temperature pin of each pin bundle, thus pin cladding damage accrual and pin reliability. The flow orificing analysis for conceptual design will be performed with Excel spreadsheet program ORFCE which was set up and tested, using the calibration factors based on available analyses data. For the verification of this program, flow orificing calculation for the MDP 840MWth core was performed. The calculational results are satisfactory compared to those of CRIEPI calculation.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2018.05a
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• pp.151-152
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• 2018

In the domestic nuclear power industry, concern about safety of nuclear power plants is continuously increased with the Fukushima nuclear power plant accident. In order to enhance the safety of nuclear power plants, it is important to ensure that the power plants are operating with proper margin within the original design bases. Margin management is the process of ensuring that the NPP designer and operator are aware of the physical and operating limits, and potential and probability of failure, for each component in the plant. All components are subject to margin considerations, but the most important components by scope and attention are those related to safety-related systems and NPP safe shutdown.