• Nuclear Engineering and Technology
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• v.33 no.2
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• pp.241-253
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• 2001

Under severe accidents, the pressure and temperature response has an important role for the integrity of a nuclear power plant containment. The history of the pressure and temperature is characterized by the amount and state of steam/air mixture in a containment. Recently, the heat transfer rate to the structure surface is supposed to be increased by the wavy interface formed on condensate film. However, in the calculation by using CONTAIN code, the condensation heat transfer on a containment wall is calculated by assuming the smooth interface and has a tendency to be underestimated for safety. In order to obtain the best- estimate heat transfer calculation, we investigated the condensation heat transfer model in CONTAIN 1.2 code and adopted the new forced convection correlation which is considering wavy interface. By using the film tracking model in CONTAIN 1.2 code, the condensate film is treated to consider the effect of wavy interface. And also, it was carried out to investigate the effect of the different cell modelings - 5-cell and 10-cell modeling - for KNGR(Korean Next Generation Reactor) containment phenomena during a severe accident. The effect of wavy interface on condensate film appears to cause the decrease of peak temperature and pressure response . In order to obtain more adequate results, the proper cell modeling was required to consider the proper flow of steam/air mixture.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1140-1153
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• 2016

Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results.

• Proceedings of the Safety Management and Science Conference
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• 2001.11a
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• pp.193-202
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• 2001

The individual rotten water purification tank recently discharges wastewater and sewage through the outlet without purification ability. The outlet water and affiliated water purification tank with microorganism cultivator tank cultivates microorganism and then drops the value BOD, COD of sewage and discharges the quality of water into the outlet. The blower and water pump operating continuously prompts the waste of energy and deterioration of equipment. Each room of deposition tank, foaming tank, microorganism cultivator tank is equipment with the sludge detection senses so it can detect the density of each room. The power-drive plant of the blower and water pump which ate the system cultivating the microorganism must be made as fuzzy controlization (If the sludge amount of each room become higher, the rate of operation of blower and water pump must heighten, on the contrary, in case of row sludge amount, the total handling amount and microorganism handling amount of each room of control. Tank reducing the rate of operation must be DB. At present, the blower amount in proportion to the sludge and oxygen demanding amount has to control. Each mom must be checked outlet level of the outlet, also each room must flow backward discharge materials, and must operate feed-back control until we want to be come as a below value of BOD/COD(10PPM ; KS).

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.22 no.7
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• pp.454-459
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• 2010

This paper is a actual design case applied to make a bid for CHP plant construction in some country. The purpose of this study is to optimize the system performance for the requirement conditions written in ITB by the client. The system consists of gas turbine, steam turbine, heat recovery steam generator and heat exchangers for district heating. The performance analysis is conducted for various seasons conditions and heat load. As a result, air density and heat load is reduced in accordance with decreasing of the outdoor temperature, therefore the system power is reduced. Considering this, the design parameters to meet the requriement conditions are optimized.

• Corrosion Science and Technology
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• v.16 no.1
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• pp.8-14
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• 2017

The corrosion behavior of a ferritic/martensitic steel T91 exposed to an aqueous solution containing chloride and sulfate ions is investigated depending on the stimulated all-volatile treatment (AVT) and under oxygenated treatment (OT) conditions. The corrosion of T91 steel under OT condition is severe, while the corrosion under AVT condition is not. The co-existence of chloride and sulfate ions has antagonistic effect on the corrosion of T91 steel in both AVT and OT conditions. Unlike to corrosion resistance in the aqueous solution, OT pretreatment provides T91 steel lower oxidation-resistance than VAT pretreatment. From scanning electron microscopy/energy dispersive X-ray spectroscopy (SEM/EDS) and X-ray diffraction (XRD) analysis, the lower corrosion resistance in the aqueous solution by VAT conditions possibly is due to the formation of pits. In addition, the lower oxidation resistance of T91 steel pretreated by OT conditions is explained as follows: the cracks formed during the immersion under OT conditions accelerated peeling-off rate of the oxide film.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.982-996
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• 2016

In this paper, new tethered aerial robots including roaming tethered aerial robots (RTARs) for radioactive material sampling and stationary tethered aerial robots (STARs) for environment monitoring are proposed to meet extremely-long-endurance missions of nuclear power plants. The flight of the proposed tethered aerial robots may last for a few days or even a few months as long as the tethered cable provides continuous power. A high voltage AC or DC power system was newly adopted to reduce the mass of the tethered cable. The RTAR uses a tethered cable spooled from the aerial robot and an aerial tension control system. The aerial tension control system provides the appropriate tension to the tethered cable, which is accordingly laid down on the ground as the RTAR roams. The STAR includes a tethered cable spooled from the ground and a ground tension control system, which enables the STAR to reach high altitudes. Prototypes of the RTAR and STAR were designed and successfully demonstrated in outdoor environments, where the load power, power type, operating frequency, and flight attitude of the RTAR and STAR were: 180 W, AC 100 kHz, and 20 m; and 300 W, AC or DC 100 kHz, and 80 m, respectively.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.893-905
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• 2016

Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

• Proceedings of the KIPE Conference
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• 2008.06a
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• pp.320-322
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• 2008

The international oil price now has been going up every each day, about 120 USD per a gallon April 2008, so that most of countries in the world are concern of the the shortage of petroleum and the development of new and renewable energy resources. This paper presents optimal design and operational features of stand-alone 500W PEMFC(Proton Exchange Membrane Fuel Cell) system which can be a substitute instead fossil fuel. The stack of PEMFC is composed of 35 laminated graphites, and a unit cell of the stack has electrical characteristics as below; 14W, 0.9V, 15A. The other components of BOP(Balance of Plant) are composed of hydrogen and nitrogen tanks, regulators, 3way solenoid valves, mass flow meters, etc.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.40-47
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• 2011

KSNPs(Korea Standard Nuclear Power Plant) have been applied the break exclusion criteria to the high energy lines passing through containment penetration area to ensure that piping failures would not cause the loss of containment isolation function, and to reduce the resulting dynamic effects. Systems with the criteria are the Main Steam system, Feed Water system, Steam Generator Blowdown system, and Chemical & Volume Control system. In accordance with FSAR(Final Safety Analysis Report), a 100% volumetric examination by augmented in-service inspection of all pipe welds appled the break exclusion criteria is required for the break exclusion application piping. However, it is difficult to fully satisfy the requirements of inspection because 12", 8" and 6" weldolet weldments of Main Steam pipe line have complex structural shapes. To resolve the difficulty on the application of conventional UT(Ultrasonic Testing) technique, realistic mock-ups and UT calibration blocks were made. Simulations of conventional UT were performed utilizing CIVA, a commercial NDE(Nondestructive Examination) simulation software. Phased array UT experiments were performed through mock-up including artificial notch type flaws. A phased array UT technique is finally developed to improve the reliability of ultrasonic test at main steam line pipe to 12", 8" and 6" branch connection weld.

• Journal of Nuclear Fuel Cycle and Waste Technology
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• v.1 no.1
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• pp.9-27
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• 2013

Nagra was established in 1972 by the Swiss nuclear power plant operators and the Federal Government to implement permanent and safe disposal of all types of radioactive waste generated in Switzerland. The Swiss Nuclear Energy Act specifies that these shall be disposed of in deep geological repositories. A number of different geological formations and sites have been investigated to date and an extended database of geological characteristics as well as data and state-of-the-art methodologies required for the evaluation of the long-term safety of repository systems have been developed. The research, development, and demonstration activities are further supported by the two underground research facilities operating in Switzerland, the Grimsel Test Site and the Mont Terri Project, along with very active collaboration of Nagra with national and international partners. A new site selection process was approved by the Federal Government in 2008 and is ongoing. This process is driven by the long-term safety and feasibility of the geological repositories and is based on a step-wise decision-making approach with a strong participatory component from the affected communities and regions. In this paper a brief history and the current status of the Swiss radioactive waste management program are presented and special characteristics that may be useful beyond the Swiss program are highlighted and discussed.