• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.755-763
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• 2020

The critical flow phenomenon has been studied because of its significant effect for design basis accidents in nuclear power plants. Transition points from thermal non-equilibrium to equilibrium are different according to the geometric effect on the critical flow. This study evaluates the uncertainty parameters of the critical flow model for analysis of DBA (Design Basis Accident) with the MARS-KS (Multi-dimensional Analysis for Reactor Safety-KINS Standard) code used as an independent regulatory assessment. The uncertainty of the critical flow model is represented by three parameters including the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio, and their ranges are determined using large-scale Marviken test data. The uncertainty range of the thermal non-equilibrium factor is updated by the MCDA (Model Calibration through Data Assimilation) method. The updated uncertainty range is confirmed using an LBLOCA (Large Break Loss of Coolant Accident) experiment in the LOFT (Loss of Fluid Test) facility. The uncertainty ranges are also used to calculate an LBLOCA of the APR (Advanced Power Reactor) 1400 NPP (Nuclear Power Plants), focusing on the effect of the PCT (Peak Cladding Temperature). The results reveal that break flow is strongly dependent on the degree of the thermal non-equilibrium state in a ruptured pipe with a small L/D ratio. Moreover, this study provides the method to handle the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio in the system code.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.759-766
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• 2013

As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the Advanced Power Reactor Plus (APR+), and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. The heat removal capability of the PAFS is strongly dependent on the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX). To evaluate the model of heat transfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used to simulate the experimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL). The Shah model, a default model for condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. To improve the calculation result, The Thome model and the new version of the Shah model are implemented and compared with the experimental data.

• Journal of Ocean Engineering and Technology
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• v.28 no.3
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• pp.245-253
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• 2014

This research analyzed the current situation and problems with an environmental impact assessment to provide a rational ocean physics assessment technique for power plant thermal effluent. This research also tried to create an improvement plan for heated effluent diffusion impact assessment by examining the reporting regulations for environmental impact assessment, national and international evaluation guidelines, etc. In the case of evaluating the oceanographic impact of heated effluent discharged from power plants, a pre-investigation is necessary before a full-scale presentence investigation, to accurately predict and minimize power plant construction effects on the surrounding environments. Before this presentence investigation, moreover, an integrated presentence plan, which agrees with the business plan, effect prediction, and post-investigation, needs to be established. A sufficient summit investigation must be made, which considers climate changes, and new and additional power plant construction. For accurate long-term oceanic environmental change prediction, the credibility of effect prediction must be elevated by presenting an evaluation method that is categorized by numerical organization models, verification methods, result presentation, and other things. Furthermore, unproductive conflicts between the people involved in heated effluent evaluation should be reduced by these improvement plans.

• 제어로봇시스템학회:학술대회논문집
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• 2001.10a
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• pp.157.3-157
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• 2001

In this paper, the optimal maintenance scheduling for turbine with considering maintenance replacement rate was proposed in order to reduce the maintenance cost during the whole period of operation, meanwhile keeping current reliability of turbine. The proposed method is only based on a few limited available data with various factors relating to maintenance replacement and repair of turbine. The proposed method will be adopted by Kyushu Electric Power Co., Inc. from April in 2002 to determine the maintenance schedule of thermal power plants.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.33 no.8
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• pp.849-857
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• 2009

To guarantee the safety and reliability of obsolete thermal power plants, on site routine patrol in their facilities has been done by human workers. Due to their poor working environments, however, a patrol robot system has been gradually required instead of the human workers from the viewpoint of the workers' safety and work efficiency. For this purpose, this paper presents a patrol robot, controllers, and its control scheme. Especially, this robot system uses a line tracing algorithm, which uses a vision camera instead of IR sensors, and an RFID system for its patrol operation. We confirmed its effectiveness through experiments.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.25 no.4
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• pp.451-460
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• 2001

An analysis program for the thermal performance prediction of steam turbine cogeneration systems with multi-extraction, reheat and regeneration has been developed on the basis of the thermodynamic heat balance method. Heat balance analyses were performed for a commercial cogeneration power plant using the program. Its appropriateness was verified by comparing its heat balance results with those of other commercial programs and those provided by the original system designer. Further parametric analyses were carried out and performance improvement measures in designing the plant were suggested.

• Proceedings of the KIEE Conference
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• 2006.07a
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• pp.331-332
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• 2006

This paper proposes the fuel cell power plants as alternative energy sources for distributed generation in Jeju Island, Korea. This will help to increase the fuel efficiency, at least double the current thermal power plants, decrease environmental pollution, virtually to none, increase the reliability of power supply, reducing the dependency of the HVDC link, and provide quality power to the growing infrastructure in meeting the requirements for the free-trade international island.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.20-24
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• 2014

Recently, it is getting more and more important ensuring the integrity for the equipment degradation according to the increase of nuclear power plant operating period. In many equipment of the nuclear power plant, snubbers mainly installed in reactor coolant pumps, steam generators and piping protected the equipment and piping from the occurrence of transient dynamic loads such as the earthquake, thermal load during the plant operation. This report describes the function, regulation, inspection requirements and management status of the snubbers installed in domestic nuclear power plants.

• Proceedings of the KIEE Conference
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• 2007.10a
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• pp.253-254
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• 2007

This paper deals with the development of power control system(PCS) for nuclear power plants. The PCS provides the control motive power to operate the CEDMs(Control Element Drive Mechanism) for reactivity control inside the reactor vessel. The CEDM is raise and lower the CEAs( Control Element Assemblies) inside the reactor core. The CEAs are constructed with the Boron-10 isotope which has a high microscopic cross section of absorption for thermal neutrons. This characteristic causes the addition of negative reactivity when a CEA is inserted and positive reactivity when it is withdrawn from the reactor core.

• Proceedings of the KIEE Conference
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• 2001.07d
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• pp.2314-2316
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• 2001

The steam gererator which produces thermal energy from coal or gas is a very important device in power plants, including the turbine driving synchronous generator which transforms kinetic energy into electrical energy. The turbine and the generator are driven by many kinds of media according to the types of which power plants are classified into steam turbine generator, gas turbine generator, water turbine generator and so on. This paper introduces the overspeed protection as well as the various speed and load control methods of some types of turbines.