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박군철(Goon-Cherl Park),전지한(Ji-Han Chun) 대한기계학회 2009 大韓機械學會論文集B Vol.33 No.9
This paper introduces the current nuclear experimental research activities in KAERI, the unique nuclear research institute in Korea, and the universities in Korea to solve and assess the issues which have been faced in the design of new reactors such as APR1400, SMART, GEN-IV reactors as well as fusion reactor. Also the experimental evaluations of safety for operating nuclear plants have been presented. The nuclear thermalhydraulic experiments performed in such organizations are classified the fundamental test, the separated effect test, and the integral effect test with ATLAS and SNUF. Introduction is deployed according to institutes. Finally, the future works and the direction of research voyage in the nuclear thermal-hydraulic field were suggested.
고정밀 2상유동 액막두께 측정을 위한 연성회로기판 기반 3-전극 센서 개발
이규병 ( Kyu Byoung Lee ),김종록 ( Jong Rok Kim ),어동진 ( Dong Jin Euh ),박군철 ( Goon Cherl Park ),조형규 ( Hyoung Kyu Cho ) 한국센서학회 2016 센서학회지 Vol.25 No.1
To understand a two-phase flow, a liquid film thickness is one of the important factors. A lot of researches have been performed to measure liquid film thickness with various approaches. Recently, an electrical conductance method which uses the conductivity of the liquid film has been widely applied on measuring the liquid film thickness. Though the electrical method has an advantage in high spatial resolution, as the conductivity of liquid can be affected by its temperature variation, the conventional electrical conductance methods have a limitation in being applied on varying temperature conditions where a heat transfer is involved. The purpose of this study is to develop a three-ring liquid film sensor that overcomes the limitation of the conventional method. The three-ring conductance method can measure the film thickness regardless of temperature variation by compensating the change of liquid conductivity. Considering its application on a wide range of conditions such as high temperature or curved surfaces, the sensor was fabricated on flexible printed circuit board (FPCB) in this study. This paper presents the concept of the measurement method, design procedure, prototype sensor fabrication and calibration results.
연성회로기판 기반 수평전열관 표면의 비등기포거동 가시화 실험 연구
김재순(Jae Soon Kim),김유나(Yu-Na Kim),박군철(Goon-Cherl Park),조형규(Hyoung Kyu Cho) 한국가시화정보학회 2016 한국가시화정보학회 학술발표대회 논문집 Vol.2016 No.4
The Passive Auxiliary Feedwater System(PAFS) is one of the advanced safety concepts adopted in the Advanced Power Reactor Plus(APR+). To validate the operational performance of the PAFS, detailed understanding of a boiling heat transfer on horizontal tube outside is of great importance. Especially, in the mechanistic boiling heat transfer model, it is important to visualize the phenomena but there are some limitations with conventional experimental approaches. In the present study, we devised a heater based on the Flexible Printed Circuit Board (FPCB) for a more comprehensive visualization and subsequently, a digital image processing technique for the bubble motion measurement was established. Using the measurement technique, important parameters of the nucleate boiling are analyzed.
연성회로기판 기반 수평전열관 표면의 비등기포거동 가시화 실험 연구
김재순(Jae Soon Kim),김유나(Yu-Na Kim),박군철(Goon-Cherl Park),조형규(Hyoung Kyu Cho) 한국가시화정보학회 2016 한국가시화정보학회지 Vol.14 No.2
The Passive Auxiliary Feedwater System(PAFS) is one of the advanced safety concepts adopted in the Advanced Power Reactor Plus(APR+). To validate the operational performance of the PAFS, detailed understanding of a boiling heat transfer on horizontal tube outside is of great importance. Especially, in the mechanistic boiling heat transfer model, it is important to visualize the phenomena but there are some limitations with conventional experimental approaches. In the present study, we devised a heater based on the Flexible Printed Circuit Board (FPCB) for a more comprehensive visualization and subsequently, a digital image processing technique for the bubble motion measurement was established. Using the measurement technique, important parameters of the nucleate boiling are analyzed.
액막유동 실험결과를 활용한 MARS 코드 다차원 모듈의 마찰모델 검증
양진화(Jin-Hwa Yang),어동진(Dong-Jin Euh),박현식(Hyun-Sik Park),조형규(Hyoung-Kyu Cho),박군철(Goon-Cherl Park) 대한기계학회 2016 대한기계학회 춘추학술대회 Vol.2016 No.12
With the advancements of local measurement technique and computational capability, high-precision experiments and analyses as to multi-dimensional thermal hydraulic phenomena in nuclear power plants have been carried out. One of the multi-dimensional phenomena is two-phase flow in the upper downcomer of nuclear reactor. In the reflood phase, the interaction between a downward liquid and a saturated transverse steam flow is important problem since it determines the bypass flow rate of the emergency core coolant (ECC). To simulate this two-phase film flow, ductshaped acrylic experimental facilities with 1/10 and 1/5 reduced scales were manufactured following the upper downcomer geometry. The air and water were used as operation fluids and the test conditions were selected as the velocity of lateral air velocity increased. From these data, it could be possible to validate the multi-dimensional modules of system analysis codes. The system analysis codes, for example, RELAP5/MOD3, MARS, SPACE, TRACE and CATHARE3 adapted multi-dimensional modules to simulate the two-phase flow more accurately. However, these modules in computational codes should be validated with multidimensional experimental study. In this study, MARS-MultiD was used to simulate the experiment, and obtained the local variables. Then, the friction models used in MARS-MultiD were validated by comparing the two-phase flow experimental results with the calculated local variables.
일체형 원자로 REX-10을 이용하는 지역에너지시스템의 운영 시나리오 개발
정종찬(Jong-Chan Jeong),김광호(Kwang-Ho Kim),박군철(Goon-Cherl Park) 대한전기학회 2009 대한전기학회 학술대회 논문집 Vol.2009 No.7
본 논문에서 일체형 원자로인 REX-10과 가스터빈발전기를 전원으로 이용하는 지역에너지시스템을 모델링 하고, 한국전력공사 전력계통과 연계운전 또는 독립운전하는 지역에너지시스템 안정적 운영방법 모의와 운전 중 발생할 수 있는 문제점을 모의하기 위해 운영 시나리오를 개발 적용하였다.
전성수(Seong Su Jeon),김지훈(Ji Hun Kim),김성진(Seong Jin Kim),박군철(Goon Cherl Park) 대한기계학회 2011 대한기계학회 춘추학술대회 Vol.2011 No.5
본 해석 연구에서는 상용 CFD 코드인 FLUENT의 VOF 모델을 이용하여 응축하는 기포의 거동을 분석하였다. FLUENT 코드로 응축하는 기포를 모사하기 위해서 사용자정의함수(UDF)를 이용하여 기포 응축을 모델링하였다. UDF 모델링을 평가하기 위해, CFD 해석 결과를 서울대학교에서 수행된 기포 응축 실험 결과와 비교하였다. UDF 모델링이 적용된 VOF 해석 결과는 실험 결과와 유사하였다. 이 방법을 이용해서, 다양한 조건에서 응축하는 기포 거동에 대한 기초 연구를 수행하였다. 해석 결과, 응축하는 기포는 속도, 이동거리, 경로 등 여러 면에서 단열기포와 다르다는 것을 확인할 수 있었다. In this numerical study, the behavior of condensing bubble was investigated using the Volume of Fluid (VOF) model in the FLUENT code. In order to simulate the condensing bubble with the FLUENT code, the bubble condensation was modeled by using the User-Defined Function (UDF). For the evaluation of the UDF modeling, the results of CFD simulation were compared with the results of a SNU bubble condensation experiment. Simulation results showed good agreements with the experimental data. Moreover, the fundamental behavior of the condensing bubble was investigated in various conditions. It was found that the behavior of the condensing bubble was different from that of the adiabatic bubble in many respects including the bubble shape, velocity, rise distance and moving trajectory.