• Journal of Korean Society of Water and Wastewater
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• v.29 no.1
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• pp.65-75
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• 2015

In this study, cation and anion exchange process for performance evaluation was conducted. A pilot plant for the ultrpure water production was installed with the capacity of $25m^3/d$. The various production rate and regeneration of ion exchange rate were tested to investigate the design parameters. The test resulst was applied to calculate the operating costs. Changing the flow rate of the ion exchange capacity of the reproduction reviewed the cation exchange process as opposed to the design value is 120 to 164% efficiency, whereas both anion exchange process is 82 to 124% efficiency, respectively. This results can be applied for more large scale plant if the scale up parameters are consdiered. The ion exchange capacity of the application in accordance with the design value characteristic upon application equipment is expected to be needed. In this study, the performance of cation and anion exchange resin process was evaluated with pilot plant($25m^3/d$). The ion exchange capacity along with space velocity and regeneration volume was evaluated. In results, the operation results was compared with design parameters.

• The Journal of the Korea institute of electronic communication sciences
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• v.10 no.12
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• pp.1373-1380
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• 2015

The rapid spread and various design of solar power plants, development, construction, maintenance technology and business coexist variety and price decline and the recent solar power plants built since the early 2010 despite the decline in the power generation unit main equipment Construction of power plants to stabilize the technology has been continuously promoted. Recently, built in the form of a solar power plant to the construction and leasing the roof and upper structure to replace the limited ground space is increasing. Factory roof, studies such as warehouse and logistics center leased space in the roof installed solar plant status and skills, conducted a study to review and analyze the economy.

• Journal of Ocean Engineering and Technology
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• v.13 no.2 s.32
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• pp.99-107
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• 1999

Boiler high-temperature pipelines such as main steam pipe, header and steam drum in fossil power plants are degraded by creep damage due to severe operationg conditions which are high temperature and high pressure for an extended period time. Such material degradation leads to various component failures causing serious accidents at the plants. Conventional measurement techniques such as replica method, electric resistance method, and hardness test method have such disadvantages as complex preparation and measurement procedures, too many control parameters, and therefore, low practicality and they were applied only to component surfaces with good accessibility. In this paper, artificial creep degradation test and ultrasonic measurement for their creep degraded specimens have been carried out for the purpose of evaluation for creep damage which can occur in high-temperature pipeline of fossil power plant. Absolute measuring method of quantitative ultrasonic measurement for material degradation was established, and long term creep degradationtests using life prediction formula were carried out. As a result of ultrasonic tests for crept specimens, we confirmed that the sound velocity decreased and the attenuation coefficient linearly increased in proportion to the increase of creep fractiin(${\phi}$c).

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.429-434
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• 1996

K-REM［1］, which is under development as a realistic evaluation model of large break LOCA, is applied to the analysis of cold leg guillotine break of Ulchin 3&4. Fuel parameters on which statistical analysis of their effects on the peak cladding temperature (PCT) are made and system parameters on which the concept of limiting value approach (LVA) are applied, are determined from the single parameter sensitivity study. 3 parameters of fuel gap conductance, fuel thermal conductivity and power peaking factor are selected as fuel related ones and 4 parameters of axial power shape, reactor power, decay heat and the gas pressure of safety injection tank (SIT) are selected as plant system related ones. Response surface of PCT is generated from the plant calculation results and on which Monte Carlo sampling is made to get plant application uncertainty which is statistically combined with code uncertainty to produce the 95th percentile PCT. From the break spectrum analysis, blowdown PCT of 1350.23 K and reflood PCT of 1195.56 K are obtained for break discharge coefficients of 0.8 and 0.5, respectively.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1571-1584
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• 2023

In this study, a numerical analysis was performed as part of an international joint research project to reproduce a real cable tray fire that occurred in the heater bay area of the turbine building of a nuclear power plant. A sensitivity analysis was performed on various input parameters to derive results consistent with the sprinkler activation time obtained from the fire event analysis. For all sensitive parameters, the normalized sprinkler activation time correlated well with the power function of the normalized sprinkler height. A correlation equation was developed to identify the sprinkler activation time at any location when determining the slope or fire growth rate under the conditions assuming a linear or t-squared heat release rate (HRR) time curve. Various cable fire growth assumptions were used to determine which assumption was better to provide the prediction coincident with the information given from the fire event analysis in terms of the sprinkler activation time and total energy generated from cables damaged by fire. In the comprehensive analysis of all the sensitive parameters, the standard deviation of the input parameters increased as the sprinkler height decreased. Within the range of the sensitivity parameter values given in this study, when considering all sprinkler heights, the standard deviation of the cable model change was the largest and that of the overhang position change was the smallest.

• Proceedings of the ESK Conference
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• 1997.04a
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• pp.72-74
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• 1997

Integration and standardization of the related regulations and guidelines are necessary to evaluate a CFMS in a systematic manner. In this study, a hierachically categorized evaluation checklist was developed by collecting and structuring human factors issues listed in the various regulations, design requirements, and technical report. A PC-based database management system was developed to update and review the evaluation results. Standard interpretation examples, evaluation guidelines, and sample evaluation formats were also designed and integrated into the database system to reduce discrepancy in evaluation.

• Journal of Environmental Impact Assessment
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• v.6 no.2
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• pp.81-92
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• 1997

This paper presents the results from the comparison analysis and evaluation between the air pollutant dispersion modeling results and the observation data in the area within a 10 km radius from the Boryong thermal power plants. The observation data used in this study were the air pollutant concentrations which had been continuously measured from 8 locations around the Boryong power plants by TMS(tele-monitoring system) for 3 months from September to November, 1996. The short-term and long-term predictions were carried out using ISC3 model and LPDM(Lagrangian Panicle Dispersion Model). The results of ISC3 modeling in a short-term showed highly as 0.7 in a correlation coefficient, but in a long-term showed just 0.54. On the other hand, LPDM showed 0.78 in a correlation coefficient for a long-term, but in a short-term showed highly value than the observation concentrations.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2005.04a
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• pp.183-189
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• 2005

Nuclear containment structure is the last barrier for being secure from any nuclear power plant accident. Even though the safety requirements of nuclear power plant have been focused on removing accidental situations, nuclear containment structure must reserve the sufficient resisting capacity to any accident because it works as the last barrier. The acceptable nuclear containment structure makes possible to limit the effect of internal accidents and to avoid radioactive release. In this study, to conduct the numerical analysis for the structural safety of a containment structure, loss of coolant accident (LOCA) is considered as the basic accidental load, and Wolsong containment structure is considered as a target structure. The CANDU containment structure, such as Wolsong containment structure, is a prestressed concrete shell structure which has dome and is reinforced with bonded tendons. The evaluation of ultimate pressure capacity was conducted by nonlinear analysis of a prestressed concrete containment structure.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.343-347
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• 2001

Advanced Pressurized Reactor 1400(APR-1400), which is a standard evolutionary advanced light water reactor(ALWR), has been developed from 1992 as one of long-term Government Project(G-7). The APR-1400 is designed to operate at the rated output of 4000MWt to produce an electric power output of around 1450MWe. The balance of plant (BOP) for the secondary system consists of main steam, feedwater, condensate, turbine generator and auxiliary system. In this paper, we describe the major design features of secondary component, balance of plant configuration, and then the turbine cycle thermal performance evaluation using PEPSE code.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.458-463
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• 1997

A transient natural circulation behavior during a LOCA at hot-standby operation is evaluated for YGN Units 3/4. The plant initial condition is determined within the EOP limitation as suitable to hot-standby mode and the transient scenario is prepared as relevant to evaluation of transient natural circulation. A 0.4% cold leg break with loss of off-site power is calculated with RELAP5/MOD3.2, whose predictability has been verified for SBLOCA natural circulation test, S-NC-8B. Through one hour transient analysis, it is found that the plant has its own decay heat removal capability by natural circulation following a LOCA, at hot-standby mode. Additional calculation is performed to investigate an effect of HPSI flow on natural circulation.