• Journal of the Korean Society of Manufacturing Technology Engineers
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• v.23 no.1
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• pp.87-93
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• 2014

This paper provides a methodology for extracting design data from the three-dimensional design software SolidWorks$^{(R)}$, which is based on the existing trochoid pump design equations that are used by hydraulic field engineers. The data extracted from the SolidWorks$^{(R)}$ model are input to a hydraulic analysis software AMESim model to determine the design factors that can influence the properties of a trochoid pump. On the basis of the simulation results, this paper proposes a method to reduce the flow loss by adjusting the outlet angle of the trochoid pump. This proposal was verified by using actual experimental results, which confirmed that adjusting the outlet angle can increase the flow rate. Hence, the results presented in this paper can contribute to the prototyping of a trochoid pump by reducing the cost associated with a trial-and-error design.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.8
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• pp.1552-1564
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• 2002

Meter-in, meter-out and bleed-off circuits are widely utilized in order to adjust the speed of a hydraulic actuator by using a flow control valve and in order to regulate the pressure of a hydraulic volume by using a simple on-off valve. In these circuits, a relief valve serves either to maintain constant system pressure or to protect the system from over-pressure loading. The relief valve of a bleed-off circuit is the second case frequently undergoing on-off action during operation. It makes the analysis of the pressure control characteristics of the circuit highly difficult. In this paper, steady-state flow rate, pressure, heat loss and efficiency of the three circuits are reexamined and basic experiments far obtaining the characteristics of a pump and relief valve are conducted. Finally, simple empirical first-order dynamic models of decreasing and increasing pressure were separately proposed and verified by comparison with experiment. As the result, the basis for the theoretical analysis of the pressure control characteristics of a bleed-off circuit using a simple on-off valve is established.

• Journal of Korea Water Resources Association
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• v.53 no.10
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• pp.861-870
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• 2020

As the flow velocity and Reynolds number increase in rockfill porous media, the flow deviates from Darcy conditions. In this study, the permeability tests of rock column specimen and laboratory gabion weir model were carried out to investigate a head loss behaviour of flow through rockfill deposition in small river artificial recharge. Through column test, the nonlinear relationships between flow velocity and hydraulic gradient and coefficients were determined and the correlation formula of hydraulic mean radius and coefficients was proposed. The flow velocities and discharges in voids estimated by proposed equations were well matched with the measured values of laboratory gabion weir model.

• Nuclear Engineering and Technology
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• v.38 no.4
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• pp.343-352
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• 2006

Convenient analytical tools for evaluation of the aperiodic and the fluctuating instabilities of the passive residual heat removal system (PRHRS) of an integral reactor are developed and results are discussed from the viewpoint of the system design. First, a static model for the aperiodic instability using the system hydraulic loss relation and the downcomer feedwater heating equations is developed. The calculated hydraulic relation between the pressure drop and the feedwater flow rate shows that several static states can exist with various numbers of water-mode feedwater module pipes. It is shown that the most probable state can exist by basic physical reasoning, that there is no flow rate through the steam-mode feedwater module pipes. Second, a dynamic model for the fluctuating instability due to steam generation retardation in the steam generator and the dynamic interaction of two compressible volumes, that is, the steam volume of the main steam pipe lines and the gas volume of the compensating tank is formulated and the D-decomposition method is applied after linearization of the governing equations. The results show that the PRHRS becomes stabilized with a smaller volume compensating tank, a larger volume steam space and higher hydraulic resistance of the path $a_{ct}$. Increasing the operating steam pressure has a stabilizing effect. The analytical model and the results obtained from this study will be utilized for PRHRS performance improvement.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.581-588
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• 2013

This paper deals with the Safety Analysis for $CANDU^{(R)}$ 6 nuclear reactors as affected by main Heat Transport System (HTS) aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermal-hydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR) analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermal-hydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY) aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermal-hydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

• 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.

• Journal of Drive and Control
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• v.15 no.2
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• pp.46-51
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• 2018

Many research studies have been carried out related to saving energy and environmental pollution in the field of construction machinery. The best solution for reducing the related environmental pollution is to reduce fuel consumption by upgrading the energy efficiency of machinery used in this field. An efficiency upgrade in the field of construction machinery would mean minimizing the pressure loss in hydraulic pipe lines or achieving optimal operating conditions while responding to a load. One way to achieve this is to make an equivalent circuit, like an electrohydrostatic actuator, or to improve the spool type valve using the 4/3 way method. This study deals with an electrohydraulic proportional flow control valve. SimulationX software is used as a simulation tool for analyzing the dynamic characteristics. The analysis results, including the performance and characteristics of design parameters, are discussed and the validity of the theoretical analysis is also evaluated.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.287-298
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• 2007

As part of the $AP1000^{TM}$ pressurized water reactor design certification program, a series of integral systems tests of the nuclear steam supply system was performed at the APEX-1000 test facility at Oregon State University. These tests provided data necessary to validate Westinghouse safety analysis computer codes for AP1000 applications. In addition, the tests provided the opportunity to investigate the thermal-hydraulic phenomena expected to be important in AP1000 small-break loss of coolant accidents (SBLOCAs). The APEX-1000 facility is a 1/4-scale pressure and 1/4-scale height simulation of the AP1000 nuclear steam supply system and passive safety features. A series of eleven tests was performed in the APEX-1000 facility as part of a U.S. Department of Energy contract. In all, four SBLOCA tests representing a spectrum of break sizes and locations were simulated along with tests to study specific phenomena of interest. The focus of this paper is the SBLOCA tests. The key thermal-hydraulic phenomena simulated in the APEX-1000 tests, and the performance and interactions of the passive safety-related systems that can be investigated through the APEX-1000 facility, are emphasized. The APEX-1000 tests demonstrate that the AP1000 passive safety-related systems successfully combine to provide a continuous removal of core decay heat and the reactor core remains covered with considerable margin for all small-break LOCA events.

• 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.

• Journal of the Korean Society of Safety
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• v.36 no.2
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• pp.111-119
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• 2021

Multi-dimensional analysis of reactor safety-KINS standard (MARS-KS) is a thermal-hydraulic code to simulate multiple design basis accidents in reactors. The code has been essential to assess nuclear safety, but has mainly focused on light water reactors, which are in the majority in South Korea. Few previous studies considered pressurized heavy water reactor (PHWR) applications. To verify the code applicability for PHWRs, it is necessary to develop MARS-KS input decks under various transient conditions. This study proposes an input model to simulate small-break loss of coolant accidents for PHWRs. The input model includes major equipment and experimental conditions for test B9802. Calculation results for selected variables during steady-state closely follow test data within ±4%. We adopted the Henry-Fauske model to simulate break flow, with coefficients having similar trends to integrated break mass and trip time for the power supply. Transient calculation results for major thermal-hydraulic factors showed good agreement with experimental data, but further study is required to analyze heat transfer and void condensation inside steam generator u-tubes.