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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.
배준호(Jun Ho Bae),어동진(Dong Jin Euh),권태순(Tae Soon Kwon) 대한기계학회 2011 대한기계학회 춘추학술대회 Vol.2011 No.5
본 논문은 원자력 열수력 실험장치의 설계에 있어 CFD의 적용 예에 대해서 기술한다. 특히, 다음 주제에 관한 CFD 활용에 대해서 다룬다. 1) 노심 모의기의 유동 및 압력장 분포, 2) 원자로 강수부에서 압력파 교란 This paper describes the applicabi1ity of CFD for designing of nuclear thennal hydraulics test facility. Especially, CFD analysis will be discussed for the following subjects. 1) Flow and pressure distribution of core simulator, 2) Pressure Perturbation in Downcomer
원전 내 배관의 증기 누설 사고 시 누설 탐지 포집/이송 시스템 예비 해석
최대경,최청열,권태순,어동진,Choi, Dae Kyung,Choi, Choengryul,Kwon, Tae-Soon,Euh, Dong-Jin 한국압력기기공학회 2020 한국압력기기공학회 논문집 Vol.16 No.2
As leakage in nuclear power plants could cause a variety of problems, it is very critical to monitor leakage from the safety point of view. Accordingly, a new type of leak detection system is currently being developed and flow characteristics of the sampling and transportation system are investigated by using numerical analysis as a part of the development process in this study. The results showed that the steam mass fraction varied according to the effect of the gap between the insulation and piping component, transportation velocity, and material properties of porous media during the sampling and transportation process. The results of this study should be useful for understanding flow characteristics of the sampling and transportation system and its design and application.