• Journal of Radiation Protection and Research
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• v.30 no.1
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• pp.17-25
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• 2005

Since the safety of nuclear power plant has been becoming a big social issue the exposure dose of radiation for workers has been one of the important factors concerning the safety problem. The existing calculation methods of radiation dose used in the planning of radiation work assume that dose rate does not depend on the location within a work space thus the variation of exposure dose by different work path is not considered. In this study, a modified numerical method was presented to estimate the exposure dose during radiation work in radwaste storage considering the effects of the distance between a worker and sources. And a new numerical algorithm was suggested to search the optimal work path minimizing the exposure dose in pre-defined work space with given radiation sources. Finally, a virtual work simulation program was developed to visualize the exposure dose of radiation doting radiation works in radwaste storage and provide the capability of simulation for work planning. As a numerical example, a test radiation work was simulated under given space and two radiation sources, and the suggested optimal work path was compared with three predefined work paths. The optimal work path obtained in the study could reduce the exposure dose for the given test work. Based on the results, tile developed numerical method and simulation program could be useful tools in the planning of radiation work.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.319-327
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• 2010

A nuclear plant ESF ACS simulator was designed, built, and verified to perform experiment related to ESF ACS of nuclear power plants. The dimension of 3D CAD model was based on drawings of the main control room(MCR) of Yonggwang units 5 and 6. The CFD analysis was performed based on the measurement of the actual flow rate of ESF ACS. The air flowing in ACS was assumed to have $30^{\circ}C$ and uniform flow. The flow rate across the HEPA filter was estimated to be 1.83 m/s based on the MCR ACS flow rate of 12,986 CFM and HEPA filter area of 9 filters having effective area of $610{\times}610mm^2$ each. When MCR ACS was modeled, air flow blocking filter frames were considered for better simulation of the real ACS. In CFD analysis, the air flow rate in the lower part of the active carbon adsorber was simulated separately at higher than 7 m/s to reflect the measured value of 8 m/s. Through the CFD analyses of the ACSes of fuel building emergency ventilation system, emergency core cooling system equipment room ventilation cleanup system, it was confirmed that all three EFS ACSes can be simulated by controlling the flow rate of the simulator. After the CFD analysis, the simulator was built in nuclear grade and its reliability was verified through air flow distribution tests before it was used in main tests. The verification result showed that distribution of the internal flow was uniform except near the filter frames when medium filter was installed. The simulator was used in the tests to confirm the revised contents in Reg. Guide 1.52 (Rev. 3).

• Transactions of the Korean Society of Mechanical Engineers
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• v.18 no.3
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• pp.681-689
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• 1994

A model to predict the onset of significant void (OSV) in vertical flow between parallel plates has been developed. The model was compared to the experimental data of Whittle and Forgan (1967) and Dougherty et al. (1990), showing excellent agreement. The model was also compared with the Saha-Zuber(1974) correlation, which has been widely used in computer codes for nuclear safety analysis. The present theory is more conservative than this correlation, and further shows that, contrary to this correlation, the Stanton number is not solely related to the Peclet number. This may explain the large error margins required for the Saha-Zuber correlation, and also the scatter beyond the error margins specified by the authors. The steady-state OSV heat fluxes for equal and unequal heating cases between parallel plates were compared. The arithmetic mean of heat fluxes for unequal heating cases is less than the heat flux for equal heating cases. The result may imply that OSV is controlled by local thermal parameters rather than bulk parameters.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.4
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• pp.323-338
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• 2007

For the assessment of the performance and safety of a deep underground radioactive repository system, the thermal, hydraulic, mechanical, and chemical behaviors and their coupling should be studied. In order to analyze the THMC coupling behavior more effectively, which requires complex mathematical models and modelling techniques, DECOVALEX international cooperation project was launched in 1992. Since its beginning, four major stages of the project were successfully completed and THMC modelling techniques for various conditions could be developed. In this study, the current status and major achievements from the project were reviewed and possible benefits of the participation to the project were discussed.

• Journal of computational fluids engineering
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• v.21 no.2
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• pp.40-46
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• 2016

In the paper, a study on the analysis of fouling effect of the venturi flowmeter is described. In the research flow field solutions about the venturi flowmeter with fouling are obtained and then the effects on fouling states by inserting a ring into the throat of venturi flowmeter are studied. As the result shows, it is found that the inserted ring reduces the fouling effect due to the flow separation occurring at the ring. Consequently, a venturi flowmeter with ring shows smaller pressure loss differences than the original configuration with no ring on fouling state. This research suggests an efficient and economic method of inserting a ring to reduce the pressure loss effects due to fouling.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2005.06a
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• pp.1808-1811
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• 2005

The prediction of AOV(air-operated valve) performance is normally evaluated by the allowable opening thrust margin for the opening and closing stroke. However, it is not easy to carry out the dynamic test measurement for all the valves in the nuclear power plant due to the safety and operating conditions. The analysis of the available and required thrust for the valve is simulated as an alternate method to turn around this obstacle. The required air pressures to the stem displacement are discussed for differential valve pressure obtained by experiment. The result of the simulation is compared with that of the experiment. SIMULINK in MATLAB was used for the simulation and the results show good agreement with the actual test carried out with Fisher globe valve.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.67 no.11
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• pp.1447-1454
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• 2018

In this paper, the design of robust DSC-PLL(Delayed Signal Cancellation Phase Locked Loop) is proposed for coping with frequency variation. This method shows significant performance for detection of fundamental positive sequence component voltage when the grid voltage is polluted by grid unbalance and frequency variation. The feedback frequency estimation of DSC-PLL is tracking the drift in the phase by unbalance and frequency variation. The robust DSC PLL is to present the analysis on method and performance under frequency variations. These compensation algorithms can correct for discrepancies of changing the frequency within maximum 193[ms] and improve traditional DSC-PLL. Linear interpolation method is adopted to reduce the discretized errors in the digital implementation of the PLL. For verification of robust characteristic, PLL methods are implemented on FPGA with a discrete fixed point based. The proposed method is validated by both Matlab/Simulink and experimental results based on FPGA(XC7Z030).

• Earthquakes and Structures
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• v.14 no.5
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• pp.389-398
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• 2018

The seismic isolation system makes a structure isolated from ground motions to protect the structure from seismic events. Seismic isolation techniques have been implemented in full-scale buildings and bridges because of their simplicity, economic effectiveness, inherent stability and reliability. As for the responses of an isolated structure due to seismic events, it is well known that the most uncertain aspects are the seismic loading itself and structural properties. Due to the randomness of earthquakes and uncertainty of structures, seismic response distributions of an isolated structure are needed when evaluating the seismic fragility assessment (or probabilistic seismic safety assessment) of an isolated structure. Seismic response time histories are useful and often essential elements in its design or evaluation stage. Thus, a large number of non-linear dynamic analyses should be performed to evaluate the seismic performance of an isolated structure. However, it is a monumental task to gather the design or evaluation information of the isolated structure from too many seismic analyses, which is impractical. In this paper, a new methodology that can evaluate the seismic fragility assessment of an isolated structure is proposed by using stochastic response database, which is a device that can estimate the seismic response distributions of an isolated structure without any seismic response analyses. The seismic fragility assessment of the isolated nuclear power plant is performed using the proposed methodology. The proposed methodology is able to evaluate the seismic performance of isolated structures effectively and reduce the computational efforts tremendously.

• Journal of Radiation Protection and Research
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• v.28 no.4
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• pp.353-360
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• 2003

The concentration of $^{137}Cs$, the particle size, and the contents of TOC, H, N and S were measured for sediments collected in the adjacent sea to Yangnam, Korea. The concentrations of $^{137}Cs$ in sediments are in the range of $^{137}Cs$ for sediments strongly depend on particle size and TOC content of sediments. The results of multiple regression analysis suggest that humic substances may have great influence on the deposition of $^{137}Cs$ in sediment.

• Proceedings of the KSME Conference
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• 2000.11a
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• pp.843-848
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tubes, the integrity evaluation must be carried out, and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire integrity evaluation process. For this reason, an integrity evaluation system, which provides efficient way of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). Various analysis methods are provided for the integrity evaluation of pressure tube. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.