• 설비공학논문집
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• 제15권11호
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• pp.943-956
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• 2003

A numerical program has been developed for the simulation of automotive engine cooling system. The program determines the mass flow rate of engine coolant circulating the engine cooling system and radiator cooling air when the engine speed is adopted by appropriate empirical correlation. The program used the method of thermal balance at individual element through the model for radiator component in radiator analysis. This study has developed the program that predicts the coolant mass flow rate, inlet and outlet temperatures of each component in the engine cooling system (engine, transmission, radiator and oil cooler) in its state of thermal equilibrium. This study also combined the individual programs and united into the total performance analysis program of the engine cooling system operating at a constant vehicle speed. An air conditioner system is also included in this engine cooling system so that the condenser of the air conditioner faces the radiator. The effect of air conditioner to the cooling performance, e.g., radiator inlet temperature, of the radiator and engine system was examined. This study could make standards of design of radiator capacity using heat rejection with respect to the mass flow rate of cooling air. This study is intended to predict the performance of each component at design step or to simulate the system when specification of the component is modified, and to analyze the performance of the total vehicle engine cooling system.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2001년도 춘계학술대회 논문집
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• pp.811-814
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• 2001

Warpage analysis of bobbin, molded by injection molding process was performed. Concerned with a mold design, cooling system was designed based on Taguchi method, the distance between cavity wall and cooling channel was most influent factor amongst four design variables like an inlet temperature of coolant, a coolant flow rate, a diameter of cooling channel, and the distance between cavity wall and cooling channel. Optimal packing processes to reduce the warpage of molded part was analyzed based on the response surface method by considering holding pressure. Their optimal processing conditions were 9.4 seconds, 5.3 seconds, 15.2 seconds, and 85MPa, respectively.

• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2183-2195
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• 2020

A mathematical model for the flowrate and rotation speed of RCP during idling was established. The numerical calculation method and dimensionless method were used to analyze the flow, head, torque and pressure and speed changes under idle conditions. Regularity, using the Q criterion vortex identification judgment method combined with surface flow spectrum morphology analysis to diagnose the vortex dynamic characteristics on RCP blade. On impeller blade, there is two oscillations in the pressure ratio on pressure surface in blade outlet region. The velocity on the suction surface is two times more oscillating than the inlet of blade, and there is an intersection with the velocity ratio curve on pressure surface. On blade of guide vane, the pressure ratio increases along the inlet to outlet direction, and the speed ratio decreases with the increase of idle time. There is a vortex that rotates counterclockwise on the suction surface, and the streamline on the suction surface of blade is subjected to the entrainment and blocking action of the vortex creates a large reverse flow in the main flow region. There are two vortices at the outlet of guide vane suction side and the vortices are in opposite directions.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.347-360
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• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2011년 춘계학술대회논문집
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• pp.579-586
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• 2011

During a reactor normal operation, a primary coolant was designed to remove the fission reaction heat of the reactor. When one pump is failure and the other pump shall supply the cooling water to cool the reduced power, it is necessary to estimate how much flow will be supplied to cool the reactor. We carried a flow net work analysis for two parallel pumping system as based on the piping net work of the primary cooling system in HANARO. As result, it is estimated that the flow of one pump increased than the rated flow of the pump below the cavitation critical flow.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.755-763
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• 2020

The critical flow phenomenon has been studied because of its significant effect for design basis accidents in nuclear power plants. Transition points from thermal non-equilibrium to equilibrium are different according to the geometric effect on the critical flow. This study evaluates the uncertainty parameters of the critical flow model for analysis of DBA (Design Basis Accident) with the MARS-KS (Multi-dimensional Analysis for Reactor Safety-KINS Standard) code used as an independent regulatory assessment. The uncertainty of the critical flow model is represented by three parameters including the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio, and their ranges are determined using large-scale Marviken test data. The uncertainty range of the thermal non-equilibrium factor is updated by the MCDA (Model Calibration through Data Assimilation) method. The updated uncertainty range is confirmed using an LBLOCA (Large Break Loss of Coolant Accident) experiment in the LOFT (Loss of Fluid Test) facility. The uncertainty ranges are also used to calculate an LBLOCA of the APR (Advanced Power Reactor) 1400 NPP (Nuclear Power Plants), focusing on the effect of the PCT (Peak Cladding Temperature). The results reveal that break flow is strongly dependent on the degree of the thermal non-equilibrium state in a ruptured pipe with a small L/D ratio. Moreover, this study provides the method to handle the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio in the system code.

• 한국산학기술학회논문지
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• 제17권10호
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• pp.725-731
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• 2016

본 연구의 목적은 복합열원을 이용하는 전기자동차용 부탄 연소식 히팅 시스템의 열적 성능을 수치적으로 연구하는 것이다. 복합열원 히팅 시스템은 승차공간 난방을 목적으로 하는 공기 가열부와 배터리 열관리를 위한 냉각수 가열부로 구성되어 있으며, 각 열원별 열적 성능을 분석하기 위하여 상용 수치해석 프로그램인 ANSYS CFX를 이용하여 공기 및 냉각수 유량변화에 따른 각 열원별 토출 온도를 도출하고 난방 용량을 계산하였다. 수치해석을 통하여 도출된 각 열원별 토출 온도는 이론적으로 계산한 토출 온도와 비교하였고, 약 0.15% 이하의 오차를 나타내었다. 결론적으로 외부공기의 유량을 0.005, 0.01, 0.015 kg/s로 증가시킬 경우 승차공간으로 유입되는 공기 온도는 감소하였으며, 배터리 열관리용으로 배출되는 냉각수 온도는 증가하였다. 또한 냉각수 유량을 0.005, 0.01, 0.015 kg/s로 증가시킬 경우 토출되는 난방 공기와 냉각수 온도는 감소하였다. 더불어 배터리 열관리를 위한 최적의 냉각수 온도와 승차공간을 위한 높은 난방 용량을 만족하기 위한 공기 및 냉각수 유량 조건은 각 0.01 kg/s 와 0.015 kg/s로 나타났다.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2003년도 제20회 춘계학술대회 논문집
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• pp.9-12
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• 2003

KSR-III 축소형 엔진을 원형으로 하는 8채널형 칼로리미터의 냉각성능해석을 수행하였다. 축대칭 압축성 해석에 의한 연소실 벽으로의 열유속을 이용하여 3차원 냉각유로 내부의 열전달 해석을 수행하였다. 연소실벽으로의 열유속은 문헌에서 제시하는 수준으로 확인되었으며 열전달 해석을 통하여 칼로리미터 개발과 운용에 필요한 냉각수의 압력강하, 냉각수 온도상승 및 연소실벽의 최고온도를 제시하였다.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3217-3228
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• 2021

Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local spots, which will result in cladding failure and threaten reactor safety. In this study, the flow blockage characteristics were analyzed with the sub-channel analysis method, and the circumferentially-varied method was employed for considering the non-uniform distribution of circumferential temperature. The developed sub-channel analysis code SACOS-PB was validated by a heat transfer experiment in a blocked 19-rod bundle cooled by lead-bismuth eutectic. The deviations between the predicted coolant temperature and experimental values are within ±5%, including small and large flow blockage scenarios. And the temperature distributions of the fuel rod could be better simulated by the circumferentially-varied method for the small blockage scenario. Based on the validated code, the analysis of blockage characteristics was conducted. It could be seen from the temperature and flow distributions that a large blockage accident is more destructive compared with a small one. The sensitivity analysis shows that the closer the blockage location is to the exit, the more dangerous the accident is. Similarly, a larger blockage length will lead to a more serious case. And a higher exit temperature will be generated resulting from a higher peak coolant temperature of the blocked region. This work could provide a reference for the future design and development of the LFR.

• 대한설비공학회:학술대회논문집
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• 대한설비공학회 2003년도 하계학술대회 논문집
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• pp.5-5
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• 2003