• Journal of Energy Engineering
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• v.12 no.2
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• pp.74-83
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• 2003

In the advanced nuclear power plants including APR1400, the SDVS (Safety Depressurization and Vent System) is adopted to increase the plant safety using the concept of feed-and-bleed operation. In the case of the TLOFW (Total Loss of Feedwater), the POSRV (Power Operated Safety Relief Value) located at the top of the pressurizer is expected to open due to the pressurization of the reactor coolant system and discharges steam and/or water mixture into the water pool, where the mixture is condensed. During the condensation of the mixture, thermal-hydraulic loads such as pressure and temperature variations are induced to the pool structure. For the pool structure design, such thermal-hydraulic aspects should be considered. Understanding the phenomena of the submerged steam jet condensation in a water pool is helpful for system designers to design proper pool structure, sparger, and supports etc. This paper reviews and evaluates the steam jet condensation in a water pool on the physical phenomena of the steam condensation including condensation regime map, heat transfer coefficient, steam plume, steam jet condensation load, and steam jet induced flow.

• Tribology and Lubricants
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• v.25 no.5
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• pp.330-334
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• 2009

Hydraulic servo cylinders have been used to control accurately a large machine in power plant. Especially, Piston head and bush of servo cylinder is assembled sleeve and piston head and bush made of Cu alloy and pad sealing part. A damages of sleeve and piston head, bush are caused by friction and wear. Thus, It is necessary to examine friction and wear characteristics of Cu alloys for the piston head and bush. In this study, to be reliable on the piston and cylinder parts, dry friction and wear experiments were carried out with Cu alloys of four kinds of AlBC, PBC, BC and BS using reciprocating friction tester of pin on disk type. From this study, the result was shown that the AlBC and PBC with alloy elements were excellent to resistance wear. As the sliding speed was increased, the wear loss of PBC decreased than another Cu alloy.

• Nuclear Engineering and Technology
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• v.47 no.4
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• pp.416-423
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• 2015

It had been disputed how to apply wall drag to the dispersed phase in the framework of the conventional two-fluid model for two-phase flows. Recently, Kim et al. [1] introduced the volume-averaged momentum equation based on the equation of a solid/fluid particle motion. They showed theoretically that for dispersed two-phase flows, the overall two-phase pressure drop by wall friction must be apportioned to each phase, in proportion to each phase fraction. In this study, the validity of the proposed wall drag model is demonstrated though one-dimensional (1D) simulations. In addition, it is shown that the existing form loss model incorrectly predicts the motion of the dispersed phase. A new form loss model is proposed to overcome that problem. The newly proposed form loss model is tested in the region covering the lower plenum and the core in a nuclear power plant. As a result, it is shown that the new models can correctly predict the relative velocity of the dispersed phase to the surrounding fluid velocity in the core with spacer grids.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.469-474
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• 1997

To provide tile information ell severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore was performed to investigate the effect of thermal hydraulic design parameters ell severe accident progression of pressurized water reactors (PWRs), Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among mile parameters. For training. different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3&4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout(SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to tile other six parameters.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.402-409
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• 2019

When a loss of coolant accident (LOCA) occurs in a nuclear power plant, accident scenarios which can prevent core damage are defined based on break size. Current probabilistic safety assessment evaluates that core damage can be prevented under small-break LOCA (SBLOCA) and steam generator tube rupture (SGTR) with rapid cool down (RCD) strategy when all safety injection systems are unavailable. However, previous research has pointed out a limitation of RCD in terms of initiation time. Therefore, RCD success criteria estimation based on allowable coping time under a SBLOCA or SGTR when all safety injection systems are unavailable was performed based on time-line and thermal-hydraulic analyses. The time line analysis assumed a single emergency operating procedure flow, and the thermal hydraulic analysis utilized MARS-KS code with variables of break size, cooling rate, and operator allowable time. Results show while RCD is possible under SGTR, it is impossible under SBLOCA at the APR1400's current cooling rate limitation of 55 K/hr. A success criteria map for RCD under SBLOCA is suggested without cooling rate limitation.

• 한국신재생에너지학회:학술대회논문집
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• 2008.05a
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• pp.266-269
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• 2008

In this study, the results of experimental investigation on the water discharge capability of sluice caisson for tidal power plant were presented. In particular, the focus of the study was placed on the examination of change in water discharge capability of a sluice caisson according to the installation of rubble mound. For this purpose, a hydraulic experiment was carried out in an open channel flume with a great care to the measurement of discharge and water level in the flume since they greatly affects the estimation of the discharge capability of each sluice caisson. In the analysis, the experimental data of four different sluice models were used, which showed that the installation of rubble mound affects in different manner depending on each sluice caisson model. When each of the four sluice models were placed on the rubble mound respectively, the water discharge increased for one sluice caisson, whereas decreased for other three sluice caissons. Further detailed analysis is needed to quantitatively estimate the influence of installation of rubble mound on the water discharge capability of a sluice caisson.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.93-102
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• 2021

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

• 전기의세계
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• v.13 no.3
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• pp.28-37
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• 1964

It is the intention of this thesis to discriminate and investigate the cause of the shaft vibration of the vertical type hydroelectric power generator with respect to electrical, mechanical and hydraulic aspects, and to analyze the vibration which will occure by the each cause investigated above. In order to test the shaft vibration of No.1 generator in Hwachon, Korea new measurement method and measuring equipments were designed. In practice the shaft vibration of the generator was measured by above equipments and analyzed by the discriminative method. Detailed explanation for the designed measurement method and instruments is presented, and the results which I had tested three times for the generator No.1 in Hwachon power plant are added. As a appendix the mechanism and causes of the thrust bearing's wear and remarks for the runner are written.

• Proceedings of the KIEE Conference
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• 2005.07b
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• pp.1047-1049
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• 2005

Hydraulic power plant is operated for peak load and frequently start-stop because of no continuous operation. So the fault can happen due to field voltage swing in the middle of starting or reactive power swing on the line. On this research, we want to analyze that this status influence on line and generator. we hope this research can contribute to the power quality improvement.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2343-2351
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• 2024

Accurately predicting the thermal hydraulic parameters of a transient reactor core under different working conditions is the first step toward reactor safety. Mass flow rate and temperature are important parameters of core thermal hydraulics, which have often been modeled as time series prediction problems. This study aims to achieve accurate and continuous prediction of core thermal hydraulic parameters under instantaneous conditions, as well as test the feasibility of a newly constructed gated recurrent unit (GRU) model based on the soft attention mechanism for core parameter predictions. Herein, the China Experimental Fast Reactor (CEFR) is used as the research object, and CEFR 1/2 core was taken as subject to carry out continuous predictive analysis of thermal parameters under transient conditions., while the subchannel analysis code named SUBCHANFLOW is used to generate the time series of core thermal-hydraulic parameters. The GRU model is used to predict the mass flow and temperature time series of the core. The results show that compared to the adaptive radial basis function neural network, the GRU network model produces better prediction results. The average relative error for temperature is less than 0.5 % when the step size is 3, and the prediction effect is better within 15 s. The average relative error of mass flow rate is less than 5 % when the step size is 10, and the prediction effect is better in the subsequent 12 s. The GRU model not only shows a higher prediction accuracy, but also captures the trends of the dynamic time series, which is useful for maintaining reactor safety and preventing nuclear power plant accidents. Furthermore, it can provide long-term continuous predictions under transient reactor conditions, which is useful for engineering applications and improving reactor safety.