• 한국지반공학회:학술대회논문집
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• 한국지반공학회 2005년도 춘계 학술발표회 논문집
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• pp.1504-1511
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• 2005

The efficiency of the hydraulic transport of soil-water mixtures is an important factor in designing and operating a pump & pipeline system and is directly connected with dredging cost and working period. However, the hydraulic transport mechanism in the slurry flow inside the pipeline such as frictional losses, specific energy consumption, deposition velocity has not been well established. In this study a new dredging test loop system was designed and built. It is composed of a slurry pipeline with pipes of different diameters, a centrifugal slurry pump and a diesel engine connected with the slurry pump. and equipped with modern measuring facilities that enable to measure all important characteristics of a transportation system. The objective of this paper is to discuss the efficiency of the hydraulic transport of the Jumoonjin sand-water mixtures in the dredging test loop and to present simple equations induced from the test results of the loop that can express the transport product and the transport productivity.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2003년도 춘계학술대회 논문집
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• pp.1139-1143
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• 2003

HILS(Hardware-In-the-Loop Simulation) is a scheme that incorporates hardware components of primary concern in the numerical simulation environment. Due to its advantages over actual vehicle test and pure simulation, HILS is being widely accepted in automotive industries as test benches for vehicle control units. Developed in this study is a HILS system for EHB(Electro-Hydraulic Brake) systems that include a high pressure generator and a valve control system that independently modulates the brake pressures at four wheels. An EHB control logic was tested in the HILS system. Test results under various driving conditions are presented and compared with the VDC logic.

• 한국공작기계학회:학술대회논문집
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• 한국공작기계학회 2004년도 추계학술대회 논문집
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• pp.468-473
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• 2004

This paper deal with basic concept of Hardware In the Loop System(HILS) for hydraulic excavator. Hydraulic excavator has many nonlinearities because of P-Q diagram, dead zone and saturation of valve, single acting cylinder, heavy manipulator. So, actual test is needed when new component or control algorithm is developed but many restrictions exist. Hydraulic circuit of excavator is too complex to model mathematically but dynamic equation of manipulator has made good progress in previous studies. Basic concept of HILS and AMESim model of hydraulic components is contained in this paper.

• International Journal of Aeronautical and Space Sciences
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• 제11권4호
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• pp.303-312
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• 2010

This paper introduces the upper stage attitude control system of KSLV-I, which is the first space launch vehicle in Korea. The KSLV-I upper stage attitude control system consists of two electro-hydraulic actuators and a reaction control system using cold nitrogen gas. A proportional, derivative, and integral controller is designed for the electro-hydraulic thrust vectoring system, and Schmidt trigger ON/OFF controllers are designed for the reaction control system. Each attitude controller is designed to have enough stability margins. The stability and performance of KSLV-I upper stage attitude control system is verified via hardware in the loop tests. Hardware in the loop tests are accomplished for perturbed flight conditions as well as nominal flight condition. The test results show that the attitude control loop of KSLV-I upper stage is very stable and the attitude controllers perform well for all flight conditions. Attitude controllers designed in this paper have been successfully applied to the first flight of KSLV-I on August 25, 2009. The flight test results show that all attitude controllers of the KSLV-I upper stage performed well and satisfied the accuracy specifications even during abnormal flight conditions.

• 유공압시스템학회:학술대회논문집
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• 유공압시스템학회 2010년도 춘계학술대회
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• pp.116-125
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• 2010

In this study, a novel hydraulic energy-regenerative system was presented from its proposal through its modeling to its control. The system was based on a closed-loop hydrostatic transmission and used a hydraulic accumulator as the energy storage system in a novel configuration to recover the kinetic energy without any reversion of the fluid flow. The displacement variation in the secondary unit was reduced, which widened the uses of several types of hydraulic pump/motors for the secondary unit. The proposed system was modeled based on its physical attributes. Simulation and experiments were performed to evaluate the validity of the employed mathematical model and the energy recovery potential of the system. The experimental results indicated that the round trip recovery efficiency varied from 22% to 59% for the test bench.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1537-1546
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• 2017

Many thermal-hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal-hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the $5{\times}5$ rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.

• 한국유체기계학회 논문집
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• 제12권5호
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• pp.7-12
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• 2009

A nuclear fuel test loop(after below, FTL) is installed in the IRI of an irradiation hole in HANARO for testing the neutron irradiation characteristics and thermo hydraulic characteristics of a fuel loaded in a light water power reactor or a heavy water power reactor. There is an in-pile section(IPS) and an out-pile section(OPS) in this test loop. When HANARO is operated normally, the fuel loaded into the IPS has a nuclear reaction heat generated by a neutron irradiation. To remove the generated heat and to maintain the operation conditions of the test fuel, a main cooling water system(MCWS) is installed in the OPS of the FTL. The MCWS is composed of a main cooler, a pressurizer, two circulation pumps, a main heater, an interconnection pipe line and instruments. The interconnection pipeline is a closed loop which is connected to an inlet and an outlet of the IPS respectively. The MCWS is under a cold function test during a start-up period. This paper describes the system flow network analysis results of the flow control of a main cooling water system in the HANARO fuel test loop. It was confirmed through the results that the flow was met the system design requirements.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.1270-1275
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• 2004

Flow condensation heat transfer coefficients(HTCs) of R22 and R134a were measured on horizontal aluminum multi-channel tube. The experimental apparatus was composed of three main parts ; a refrigerant loop, a water loop and a water-ethylene glycol loop. The test section in the refrigerant loop was made of aluminum multi-channel tube of 1.4 mm hydraulic diameter and 0.53 m length. The refrigerant was cooled by passing cold water through an annulus surrounding the test section. The data scan vapor qualities $(0.1{\sim}0.9)$, mass flux ($200{\sim}400$ $kg/m^{2}s$) and heat flux ($7.3{\sim}7.7$ $kW/m^{2}$) at $40{\times}0.2^{\circ}C$ saturation temperature in small hydraulic diameter tube. It was found that some well-known previous correlations were not suitable for multichannel tube. So, It must develop new correlations for multi-channel tubes.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1083-1095
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• 2016

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V&V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodiumshowed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.37-38
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• 2005