• Proceedings of the KSR Conference
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• 2008.11b
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• pp.214-218
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• 2008

Because on the high-tension underground distribution line of an electric railway high voltage XLPE Cable two or three circuits between railway stations with a standard as receiving transformer facilities are established at a $30km{\sim}50km$ interval, reactive power in which the phase of a current is larger than that of a voltage is supplied when trains are not working, so when there are no loading or low loading as night. Due to the long-distance trend of the underground distribution system on an alternating current railway distribution line, the terminal voltage of a transformer is over the standard voltage, and after all, commercial cycle overvoltage is continued. To solve this problem, the shunt reactor is installed in middle of power distribution lines to maintain receiver voltage meted under the allowance regulation through control of the reactive power. Also, in case that the thickness of single cable is over $60mm^2$ and length of line is about over 30km, a circuit breaker is broken by shorting shunt ability of charging current in excess of shunt current(31.5A.rms). Therefore, this thesis presents installing the location of shunt reactor for quantitative analysis by using optimum algorism for compensation and control of the charging current.

• 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.

• Journal of the Korean Society of Safety
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• v.32 no.1
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• pp.140-144
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• 2017

Fire protection for nuclear power plants should be designed according to the concept of "Defense in Depth" to achieve the reactor safety shutdown. This concept focuses on fire prevention, fire suppression and safe shutdown. Fire prevention is the first line of "Defense in Depth" and the licensee should establish administrative measures to minimize the potential for fire to occur. Administrative measures should include procedures to control handling and use of combustibles. Electrical cables is the major contributor of fire loads in nuclear power plants, therefore electrical cables should be fire retardant. Electrical cables installed in nuclear power plants should pass the flame test in IEEE-383 standard in accordance with NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants". To assure the fire retardant of electrical cables during design life, both aged and unaged cable specimens should be tested in accordance with IEEE-383. It can be generally thought that the flammability of electrical cables has been increased by wearing as time passed, however the results from fire retardant tests performed in U.S.A and Korea indicate the inconsistent tendency of aging and consequential decrease in flammability. In this study, it is expected that the effective methodology for validation of fire retardant performance would be identified through the review of the results from fire retardant tests.

• Journal of the Korean Society of Safety
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• v.23 no.5
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• pp.54-60
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• 2008

The objective of this study is to obtain the optimal process condition and the maximum decomposition efficiency by measuring the decomposition efficiency, electricity consumption, and voltage in accordance with the change of the process variables such as the frequency, maintaining time period, concentration, electrode material, thickness of the electrode, the number of windings of the electrode, and added materials etc. of the harmful atmospheric contamination gases such as NO, $NO_2$, and $SO_2$ etc. with the plasma which is generated by the discharging of the specially designed and manufactured $TiO_2$ catalysis reactor and SPCP reactor. The decomposition efficiency of the NO, the standard samples, is obtained with the plasma which is being generated by the discharge of the combination effect of the $TiO_2$ catalysis reactor and SPCP reactor with the variation of those process variables such as the frequency of the high voltage generator($5{\sim}50kHz$), maintaining time of the harmful gases($1{\sim}10.5sec$), initial concentration($100{\sim}1,000ppm$), the material of the electrode(W, Cu, Al), the thickness of the electrode(1, 2, 3mm), the number of the windings of the electrode(7, 9, 11turns), basic gases($N_2$, $O_2$, air), and the simulated gas($CO_2$) and the resulting substances are analyzed by utilizing FT-IR & GC.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1651-1664
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• 2023

It is very difficult to capture the multi-dimensional phenomena such as asymmetric flow and temperature distributions with the one-dimensional (1D) model, obviously, due to its inherent limitation. In order to overcome such a limitation of the 1D representation, many state-of-the-art system codes have equipped a three-dimensional (3D) component for multi-dimensional analysis capability. In this study, a standard multi-dimensional analysis model of APR1400 (Advanced Power Reactor 1400) has been developed using TRACE (TRAC/RELAP Advanced Computational Engine). The entire reactor pressure vessel (RPV) of APR1400 has been modeled using a single 3D component. The fuels in the reactor core have been described with detailed and coarse representations, respectively, to figure out the impact of the fuel description. Using both 3D RPV models, a comparative analysis has been performed postulating a double-ended guillotine break at a cold leg. Based on the results of comparative analysis, it is revealed that both models show no significant difference in general plant behavior and the model with coarse fuel model could be used for faster transient analysis without reactor kinetics coupling. The analysis indicates that the asymmetric temperature and flow distributions are captured during the transient, and such nonuniform distributions contribute to asymmetric quenching behaviors during blowdown and reflood phases. Such asymmetries are directly connected to the figure of merits in the LBLOCA analysis. Therefore, it is recommended to employ a multi-dimensional RPV model with a detailed fuel description for a realistic safety analysis with the consideration of the spatial configuration of the reactor core.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1379-1387
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• 2017

This paper presents a comparison of homogenization techniques implemented in the APOLLO3 platform for transport core calculations: standard scalar flux weighting and new flux-moment homogenization, in different combinations with (or without) leakage models. Besides the historical B1-homogeneous model, a new B-heterogeneous one has indeed been implemented recently in the two/three-dimensional-transport solver using the method of characteristics. First analyses have been performed on a very simple Sodium Fast Reactor core with a regular hexagonal lattice. They show that using the heterogeneous leakage model in association with flux-moment homogenization strongly improves the prediction of $k_{eff}$ and void reactivity effects. These good results are confirmed when the application is done to the fissile assemblies of the more complex CFV (Low Void Effect) core of the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project of sodium-cooled fast breeder reactor (Generation IV).

• Journal of Power System Engineering
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• v.19 no.5
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• pp.45-51
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• 2015

Grade 91 steel is used for the major structural components of Generation-IV reactor systems such as a very high temperature reactor (VHTR) and sodium-cooled fast reactor (SFR). Since these structures are designed for up to 60 years at elevated temperatures, the prediction of long-term creep life is very important to determine an allowable design stress of elevated temperature structural component. In this study, a large body of creep rupture data was collected through world-wide literature surveys, and using these data, the long-term creep life was predicted in terms of three methods: Larson-Miller (L-M), Manson-Haferd (M-H) and Wilshire methods. The results for each method was compared using the standard deviation of error. The L-M method was overestimated in the longer time of a low stress. The Wilshire method was superior agreement in the long-term life prediction to the L-M and M-H methods.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.64-69
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• 2014

Gr. 91 steel is used for the major structural components of Generation-IV reactor systems, such as a very high temperature reactor(VHTR) and sodium-cooled fast reactor(SFR). Since these structures are designed for up to 60 years at elevated temperatures, the prediction of long-term creep life is important for a design application of Gr. 91 steel. In this study, a number of creep rupture data were collected through world-wide literature surveys, and using these data, the long-term creep life was predicted in terms of three methods: the single-C method in Larson-Miller(L-M) parameter, multi-C constant method in the L-M parameter, and a modified method("sinh" equation) in the L-M parameter. The results of the creep-life prediction were compared using the standard deviation of error value, respectively. Modified method proposed by the "sinh" equation revealed better agreement in creep life prediction than the single-C L-M method.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.1
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• pp.58-65
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• 2018

The purpose of this study is to develop a Half Nozzle Repair(HNR) process to prevent the leakage from welds on small bore piping in Reactor Coolant System. The Codes & Standards of tempered bead and design requirements of J-Groove welds are reviewed. Automatic machine GTAW welding and machining equipments are developed to perform HNR process. Single pass welding and overlay welding equipments are conducted in order to obtain the optimal temper bead welding process parameters with Alloy 52M filler wire. Coarse grain heat affected zone(CGHAZ) is formed by rapid cooling rate in heat affected zone after welding. Accordingly, a proper temper bead technique is required to reduce CGHAZ in 1-Layer of welds by 2- and 3-Layers. Mock-up tests show that the developed HNR process is possible to meet ASME Code & Standard requirements without any defect.

• Computational Structural Engineering
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• v.10 no.3
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• pp.225-231
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• 1997

The dynamic characteristics of Control Element Drive Mechanism(CEDM) for Korea Standard Nuclear Power Plant are studied with the CEDM modeled as a secondary mass in a simplified two degree of freedom system, while the reactor vessel as a primary mass. The optimal .mu.-f curve is developed to reduce the response amplitudes of both primary and secondary masses. In order to improve a design it is proposed that the natural frequency ratio, f, should be converged to 0.93, the mass ratio, .mu., should not be reduced, and the result should be converged to the optimal .mu.-f curve. Optimal design for CEDM components has been carried out and the response amplitude ratios of reactor are reduced 10.5 - 19.7% while those of CEDM are reduced 6.3 - 3.4%.