• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.508-509
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• 2005

• Transactions of the Korean Society of Automotive Engineers
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• v.25 no.1
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• pp.103-110
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• 2017

Most commercial vehicles have adopted the hydraulic power steering system. To reduce fuel consumption and to improve steering controllability, a hybrid electric power steering system is being developed for commercial vehicles. In this study, the HILS (Hardware In the Loop Simulation) system equipped with a commercial vehicle hybrid electric power steering system was developed and the vehicle dynamic performance of a truck with the steering system was evaluated. The hybrid electric power steering system is composed of the EHPS motor pump, column mounted EPS system, and ball nut steering gear box for heavy commercial vehicles. The accuracy of vehicle models equipped with the HILS system was verified with comparisons between the simulation results and field test results. The road reaction forces of the steering system were generated from the vehicle model and verified using field test results. Step steering tests using the verified HILS system were carried out and the performance of a newly developed commercial vehicle hybrid electric power steering system was evaluated.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.968-978
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• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

• 제어로봇시스템학회:학술대회논문집
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• 2000.10a
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• pp.191-191
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• 2000

Hydraulic servo system is difficult to be made up and each component is very expensive, it takes long for actual system to make and test and it costs a high price. Because of these characteristics of hydraulic servo system, a real time simulator that could describe behavior of real system is highly demanded, without composing real hydraulic system. So, many studies have been (lone on these subjects and many simulators are developed with superiority. Since the nonlinearity of a hydraulic system common simulator have composed of many calculative times byusing DSP(Digital Signal processing) and have made it possible to find the situations of the system in real time, calculating hydraulic simulation and controller separately. In this study, we suggest real-time simulator that could describe real system without ordinary DSP card. This simulator is composed of 80196kc and personal computer. DSP card that has calculated complex numerical equation is supplanted by personal computer and 80196kc generates control signals independently out of the personal computer. In all process, personal computer is synchronized with one-board microprocessor within sampling time in the closed loop system. This makes it possible to be described in hydraulic servo system in real time. And to make a comparison between the result of the real-time simulator and a hydraulic servo system.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2000.05a
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• pp.557-560
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• 2000

The Purpose of this study is analysis of kinematic for a modified manipulator and experimental test to certify auto-balancing operation. The test is carried out as follows. First, we solve the inverse kinematics and then do a closed loop control. Second we confirm translation displacement and rotation angle of a manipulator.

• Journal of the Korean Society of Mechanical Technology
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• v.13 no.2
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• pp.137-143
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• 2011

Fuel Test Loop(FTL) is a facility which could conduct a fuel irradiation test at HANARO(High-flux Advanced Neutron Application Reactor). FTL simulates commercial NPP's operating conditions such as the pressure, temperature and neutron flux levels to conduct the irradiation and thermo-hydraulic tests. The In-Pile Test Section(IPS) installed in HANARO FTL is designed as a pressure vessel design conditions of $350^{\circ}C$, 17.5MPa. The instrumentation MI-cables for thermocouples, SPND and LVDT are passed through the sealing plug, which is in the pressure boundary region and is a part of instrumentation feedthrough of MI-cable. In this study, the brazing method and performance test results are introduced to the sealing plug with BNi-2 filler metal, which is selected with consideration of the compatibility for the coolant. The performance was verified through the insulation resistance test, hydrostatic test, and helium leak test.

• Journal of Drive and Control
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• v.15 no.4
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• pp.55-60
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• 2018

Force-controlled electro-hydrostatic actuators have to exhibit high backdrivability, to quickly compensate for force control errors caused by externally disturbed rod movement. To obtain high backdrivability, the servomotor for driving the hydraulic pump, should rotate exactly to such a revolution to compensate for force control errors, compressing or decompressing cylinder chambers. In this study, we proposed a modified velocity control structure, including a robust internal-loop compensator (RIC)-based velocity controller, for the servomotor to improve backdrivability of a force-controlled EHA. Performance improvement was confirmed experimentally, wherein sinusoidal velocity disturbance was applied to the force-controlled EHA, with constant reference input. Its dynamic force control errors reduced effectively, with the proposed control scheme, compared to test results with a conventional motordriver, for motor velocity control.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.776-792
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• 2021

Conventional integral effect test facilities were constructed to enable the precise observation of thermal-hydraulic phenomena and reactor behaviors under postulated accident conditions to prove reactor safety. Although these facilities improved the understanding of thermal-hydraulic phenomena and reactor safety, applications of new technologies and their performance tests have been limited owing to the cost and large scale of the facilities. Various nuclear technologies converging 4th industrial revolution technologies such as artificial intelligence, drone, and 3D printing, are being developed to improve plant management strategies. Additionally, new conceptual passive safety systems are being developed to enhance reactor safety. A new integral effect test facility having a noticeable scaling ratio, i.e., the (UNIST reactor innovation loop (URI-LO), is designed and constructed to improve the technical quality of these technologies by performance and feasibility tests. In particular, the URI-LO, which is constructed using a transparent material, enables better visualization and provides physical insights on multidimensional phenomena inside the reactor system. The facility design based on three-level approach is qualitatively validated with preliminary analyses, and its functionality as a test facility is confirmed through a series of experiments. The design feature, design validation, functionality test, and future utilization of the URI-LO are introduced.

• Nuclear Engineering and Technology
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• v.42 no.6
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• pp.656-661
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• 2010

Lead-alloys are very attractive nuclear coolants due to their thermo-hydraulic, chemical, and neutronic properties. By utilizing the HELIOS (Heavy Eutectic liquid metal Loop for Integral test of Operability and Safety of PEACER$^2$) facility, a thermal hydraulic benchmarking study has been conducted for the prediction of pressure loss in lead-alloy cooled advanced nuclear energy systems (LACANES). The loop has several complex components that cannot be readily characterized with available pressure loss correlations. Among these components is the core, composed of a vessel, a barrel, heaters separated by complex spacers, and the plenum. Due to the complex shape of the core, its pressure loss is comparable to that of the rest of the loop. Detailed CFD simulations employing different CFD codes are used to determine the pressure loss, and it is found that the spacers contribute to nearly 90 percent of the total pressure loss. In the system codes, spacers are usually accounted for; however, due to the lack of correlations for the exact spacer geometry, the accuracy of models relies strongly on assumptions used for modeling spacers. CFD can be used to determine an appropriate correlation. However, application of CFD also requires careful choice of turbulence models and numerical meshes, which are selected based on extensive experience with liquid metal flow simulations for the KALLA lab. In this paper consistent results of CFX and Star-CD are obtained and compared to measured data. Measured data of the pressure loss of the core are obtained with a differential pressure transducer located between the core inlet and outlet at a flow rate of 13.57kg/s.

• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.561-586
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• 2012

This paper describes a summary of final prediction results by system-scale safety analysis codes during the OECD/NEA/CSNI ISP-50 exercise, targeting a 50% Direct Vessel Injection (DVI) line break integral effect test performed with the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS). This ISP-50 exercise has been performed in two consecutive phases: "blind" and "open" phases. Quantitative comparisons were performed using the Fast Fourier Transform Based Method (FFTBM) to compare the overall accuracy of the collected calculations. Great user effects resulting from the combination of the possible reasons were found in the blind phase, confirming that user effect is still one of the major issues in connection with the system thermal-hydraulic code application. Open calculations showed better prediction accuracy than the blind calculations in terms of average amplitude (AA) value. A total of nineteen organizations from eleven countries participated in this ISP-50 program and eight leading thermal-hydraulic system analysis codes were used: APROS, ATHLET, CATHARE, KORSAR, MARS-KS, RELAP5/MOD3.3, TECH-M-97, and TRACE.