• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.313-326
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• 2017

The systematic management of plant risk is crucial for enhancing the safety of nuclear power plants and for designing new nuclear power plants. Accident sequence precursor (ASP) analysis may be able to provide risk significance of operational experience by using probabilistic risk assessment to evaluate an operational event quantitatively in terms of its impact on core damage. In this study, an ASP methodology for two operation mode, full power and low power/shutdown operation, has been developed and applied to significant accident precursors that may occur during the operation of nuclear power plants. Two operational events, loss of feedwater and steam generator tube rupture, are identified as ASPs. Therefore, the ASP methodology developed in this study may contribute to identifying plant risk significance as well as to enhancing the safety of nuclear power plants by applying this methodology systematically.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.20 no.2
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• pp.689-696
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• 1996

A comparison is made of the heat loss from a hollow cylinder, computed using an one-dimensional analytic method and a two-dimensional separation of variables scheme. For a two-dimensional analysis, the temperature of the inner surface as a boundary condition can be varied along the length of the cylinder by varing the temperature variation factor, b. Comparisons of the heat loss from the hollow cylinder using these two methods are given as a function of non-dimensional cylinder length, the ratio of the outer radius to the inner radius, temperature variation factor and Biot number. The result shows that the value of the heat loss from the hollow cylinder obtained using the one-dimensional analytic method becomes close to the value given by the two-dimensional separation of variables scheme as the value of Biot number and the non-dimensional hollow cylinder length increase and as the ratio of the outer radius to the inner radius decreases.

• Journal of Welding and Joining
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• v.19 no.3
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• pp.334-341
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• 2001

The chemical and geometric effects of weld on flow-accelerated corrosion (FAC) of SA106 Gr.C low alloy steel pipe in 3.5wt% NaCl and simulated feedwater of nuclear power plant have been investigated by using rotating cylinder electrode. Polarization test and weight loss test were conducted and compared at rotating speed of 2000rpm (3.14m/s) with the variables of chemical and geometric parameters. The results showed that the chemical effects were relatively larger than the geometric effects, and the welded parts were the local anode and preferentially corroded, which could be explained by the differences between microstructural and compositional parameters. On the other hand, under active corrosion conditions, the heat affected zone were severely corroded and microstructural effects became the important role in the whole process.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.562-567
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• 1995

The concept of VLDT (Variable Low DNBR Trip), a new CPC trip function, was proposed and applied to the events of increase in secondary heat removal, such as an excess feedwater event anti an IOSGADV (Inadvertent Opening S/G Atmospheric Dump Valve). Major assumption used in this study was no time delay to LOOP (Loss of Offsite Power) after turbine trip. In case of using this VLDT function, safety criterion of DNB would not be violated under the same condition as previous analysis without any change in thermal margin.

• Nuclear Engineering and Technology
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• v.15 no.1
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• pp.41-49
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• 1983

A computer code, BERD (Bayesian Estimation of Reliability Data), has been developed and tested in order to update the data for the reliability analysis of safety related systems in a specific nuclear power plant. The code has been used to derive the plant-specific data for reliability analysis of the auxiliary feedwater system of a pressurized water reactor. The prior information for components selected was taken from the U.S. Reactor Safety Study, WASH-1400, and the operating experiences from published licensee event reports. The results show that the updated data are well fitted to log-normal distribution curves and the error factors are reduced significantly.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.437-442
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• 1997

To determine the optimum pressurizer water level during normal operation for Kori unit 1, performance and safety analysis are peformed. The methodology is developed by evaluating "decrease in secondary heat removal" events such as Loss of Normal Feedwater accident. To demonstrate optimum pressurizer level setpoint, RETRAN-03 code is used for performance analysis. Analysis results of RETRAN following reactor trip are compared with the actual plant data to justify RETRAN code modelling, The results of performance and safety analyses show that the newly established level setpoints not only improve the performance of pressurizer during transient including reactor trip but also meet the design bases of the pressurizer volume and pressure. pressure.

• Proceedings of the Membrane Society of Korea Conference
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• 1991.04a
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• pp.9-9
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• 1991

This paper will begin by describing osmosis and how reverse osmosis works. It will show how osmotic pressure affects reverse osmosis operations. It uill explain salt rejection, membrane flux, and recovery rates and the affect that salt built up has on membrane performance. It wil 1 explain the limitations of RO performance and why pretreatment is important. It will describe the two basic types of membrane, asymmetric and thin-film composite and explain the difference between these types plus compare cellulose acetate types to aromatic polyamide type membranes. It will discuss operating efficiences as it compares to feedwater pressure, concentration, temperature and pH. Finally, it will discuss the differences between tubular, plate and frame, hollow fiber and spiral wound element design. It will be a paper that talks about the basics of RO systems and should give a person who is unfamiliar with RO a basic introduction to this type of separation technology.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2003.05a
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• pp.670-676
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• 2003

To investigate the group plugging effect the fluid-elastic instability analysis has been performed on various column and row number of the KSNP steam generator lutes. This study compares the stability ratio of the plugged tube with that of the intact one. The information on the thermal-hydraulic data of the steam generator have been obtained by using the ATHOS3-MOD1 code with and without the thermal energy transfer at the plugged region. Both of the boundary conditions of hot-leg temperature and feedwater mass flow rate are fixed for this investigation. From the results of this study the stability ratios inside the group plugging zone are decreased slightly. At the outside of group plugging zone, however, most of the stability ratios tend to be increased.

• Proceedings of the KIEE Conference
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• 1998.07a
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• pp.189-192
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• 1998

This paper presents the basis for the need of feed-water deaerator installation in nuclear power plant and has been conducting reviews to understand the mechanism of dissolved oxygen(DO) formation as well as removal. DO is produced by feedwater make-up, air inleakage in vacuum system and condensation in condenser, etc. and removed by mechanical method and chemical method. DO has an influence on S/G in the form of denting, pitting, fretting, etc. DO control by deaerator has an great effect on the integrity of S/G in nuclear power plant.

• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.394-402
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• 1993

A novel framework for quantitative estimates of the benefits of signals on nuclear power plant availability and economics has been developed in this work. The models developed in this work quantify how the perfect signals affect the human operator's success in restoring the power plant to the desired state when it enters undesirable transients. Also, the models quantify the economic benefits of these perfect signals. The models have been applied to the condensate feedwater system of the nuclear power plant for demonstration.