• Fire Science and Engineering
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• v.33 no.6
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• pp.43-52
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• 2019

A walk-down procedure for fire modeling of nuclear power plants, based on deterministic fire protection requirements, was developed. The walk-down procedure includes checking the locations of safety shutdown equipment and cables that are not correctly indicated on drawings and identifying the existence and location of combustibles and ignition sources. In order to verify the performance of the walk-down procedure developed in this study, a sample of important equipment and cables were selected for hypothetical multiple spurious operation (MSO) scenarios. In addition, the hypothetical fire modeling scenarios were derived from the selected safe shutdown equipment and cables and an actual walk-down was conducted. The plant information collected through the walk-down was compared to the information obtained from the drawings, so that the collected information may be used as input values for the fire modeling.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2652-2659
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• 2020

The methyl iodide (CH₃I) removal performance of tri-ethylene-di-amine impregnated activated carbon (TEDA-AC) used in the air cleaning unit of nuclear power plants (NPPs) should be maintained at least 99% between 24 month-performance test period. In order for evaluating the effectiveness of TEDA-AC on the removal performance of CH₃I in nuclear power plant during the operation of NPPs, the long-term test for up to 15 months was carried out under the simulated operating conditions (e.g., 25 ℃, RH 50%, ppb level poisoning gases injection) at nuclear power plants (NPPs). The TEDA-AC samples were analyzed with the Brunauer-Emmett-Teller (BET) specific surface area and TEDA content as well as CH₃I penetration test. It is clearly evident that more than 99% of CH₃I removal performance of TEDA-AC was observed in the TEDA-AC samples during 15 months of long-term operation under the simulated NPP operating conditions including the ppb level of organic and oxide form of poisoning gases. BET specific surface area and TEDA content that can affect the CH₃I removal performance of TEDA-AC were also maintained as those in new TEDA-AC during 15 months of long-term operation.

• Journal of Institute of Control, Robotics and Systems
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• v.4 no.4
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• pp.551-560
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• 1998

This paper presents a neural network model representing complex hydro-thermo-dynamic characteristics of a steam generator in nuclear power plants. The key modeling processes include training data gathering process, analysis of system dynamics and determining of the neural network structure, training process, and the final process for validation of the trained model. In this paper, we suggest a training data gathering method from an unstable steam generator so that the data sufficiently represent the dynamic characteristics of the plant over a wide operating range. In addition, we define the inputs and outputs of neural network model by analyzing the system dimension, relative degree, and inputs/outputs of the plant. Several types of neural networks are applied to the modeling and training process. The trained networks are verified by using a class of test data, and their performances are discussed.

• Proceedings of the KIEE Conference
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• 2002.07d
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• pp.2322-2324
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• 2002

Korea Water Resources Corporation (KOWACO) operates nine hydroelectric power stations on the Han River, Nakdong River, Geum River & Seomjin River Areas. KOWACO will take the lead in improving the efficiency of water resources supply in this parer, KOWACO Generation Intergrated Operating System(GIOS) project will improve the generating plant, hydraulic structures, switchgear, and station controls, alarms and services for unmanned remote controlled operation by modern intergraed SCADA system. Mordern SCADA is used to cover all computer systems designed to perform the following main functions: - to collect data from industrial plant devices. - to process and perform calculations on the collected. - to present collected and derived data on displays on. - to accept commands entered by human operators and act on them such as sending control commands to plant devices.

• Journal of the Korean Society of Safety
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• v.27 no.5
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• pp.229-234
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• 2012

Operator's action time is evaluated from MAAP4 analysis used in conventional probabilistic safety assessment(PSA) of a nuclear power plant. MAAP4 code which was developed for severe accident analysis is too conservative to perform a realistic PSA. A best-estimate code such as RELAP5/MOD3, MARS has been used to reduce the conservatism of thermal hydraulic analysis. In this study, operator's action time of core cooling recovery operation is evaluated by using the MARS code, which its Fussell-Vessely(F-V) value was evaluated as highly important in a small break loss of coolant(SBLOCA) event and loss of component cooling water(LOCCW) event in previous PSA. The main conclusions were elicited : (1) MARS analysis provides larger time window for operator's action time than MAAP4 analysis and gives the more realistic time window in PSA (2) Sufficient operator's action time can reduce human error probability and core damage frequency in PSA.

• The KSFM Journal of Fluid Machinery
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• v.16 no.6
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• pp.26-31
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• 2013

Last blades of low-pressure turbine in nuclear power plant are the highly damaged part and always suffered from different types of loadings leading to various stress components, stresses due to centrifugal force and steam flow loading. Especially, centrifugal stress generated by turbine rotation is one of the main problems and more significant than other stresses as they have the greatest effect on total stress. Therefore, this study was performed to obtain the important information for estimation model development of the blade centrifugal stress level and distribution.

• Journal of the Korean Society of Safety
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• v.26 no.5
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• pp.41-45
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• 2011

The fire event occurred in fire proof zone often causes serious electrical problems such as shorts, ground faults, or open circuits in nuclear power plants. These would be directed to the loss of safe shutdown capabilities performed by safety related systems and equipments The fire event can treat the basic design principle that safety systems should keep their functions with redundancy and independency. In case of a cable fire, operators can not perform their mission properly and can misjudge the situation because of spurious operation, wrong indication or instrument. These would deteriorate the plant capabilities of safety shutdown and make disastrous conditions. In this paper, the cables of the representative nuclear power plant in korea is selected and the cable functional failure temperature by exposed fire using Cable Response to Live Fire(CAROLFIRE) is studied. It is expected that the results are very useful to know the cable failure temperature by exposed fire. We confirmed the safety and integrity of the cable by exposed fire and those results will use the based data of cable exposed fire characteristics.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.64-72
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• 2004

The dryer (ten per unit) are operating to remove tritium in PHWR(Pressurized Heavy Water Reactor). There are coming out heavy water adsorbent from operating the dryer (95 drums for ten year per unit) The amount of radioactivity of heavy water adsorbent almost exceed ninety times more than disposal limit-in-itself showed by The Ministry of Science and Technology. It has to be disposed whole radioactive waste products, however there are problems of increase at the expense of their permanent disposal. In this research, We have studied how to remove kinds of nuclear materials and amount of tritium with in heavy water adsorbent. As the result we could develop disposal equipment and apply it. D20 adsorbent have to contain below Gamma nuclide O.3Bq/g and tritium 100Bq/g "The Regulation for disposal of the radioactivity wastes" showed by The Ministry of Science and Technology. There fore. So as to remove amount of tritium and kinds of nuclear materials (DTO) we needed a equipment. Also we have studied how to remove effectively radioactivity with in Adsorbent. As cleaning heavy water adsorbent and drying on each condition (temperature for drying and hours for cleaning). Because there is something to return heavy water adsorbent by removing impurities within adsorbent when it is dried o high temperature. After operating, we have been applying this research to the way to dispose heavy water adsorbent. Through this we could reduce solid waste products and the expense of permanent disposal of radioactive waste products and also we could contribute nuclear power plant run safely. According to the result we could keep the best condition of radiation safety super vision and we could help people believe in safety with Radioactivity wastes control for harmony with Environment.

• Journal of Energy Engineering
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• v.24 no.2
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• pp.40-44
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• 2015

Sensitivity on the reactor safety was evaluated for the safety margin and time delay applied to the opening pressure of liquid relief valve(LRV) of the primary heat transport system(PHTS) in the pressurized heavy water reactor(PHWR) type nuclear power plant. Since the LRV is the pressure boundary for the PHTS in the safety analysis, the operating of LRV has a significant effect on the safety analysis results. Therefore it is required during the regulatory review of Wolsong Unit 1 safety analysis to find the safety effect of the application of safety margin and time delay to the LRV opening pressure for the safety analysis of PHTS pressurizing events.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.40 no.6
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• pp.613-619
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• 2016

A pneumatic control valve performs a major role in controlling the flow of a system or the level of a key tank in many power plants, and its performance should be guaranteed during the plant's lifetime. Its operation starts by supplying air to the pneumatic actuator or by exhausting the air from the actuator. To control the valve position, the amount of air supply or exhaust is adjusted by a control loop where various accessaries are equipped. In this paper, air leakage in the air supply line, changes in the valve packing force, and false adjustments of zero and the span of the positioner are simulated and analyzed using a 2-in pneumatic valve with a position control loop including an I/P converter and positioner, where the valve position is controlled within ${\pm}2%$ of the control pressure at 67% opening position.