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황돈관,최낙준,정우현,김태일,이요한,조항진 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.5
In an open-pool type research reactor with a downward forced flow in the core, pipes can be under subatmospheric pressure because of the large pressure drop at the reactor core in the atmospheric pool. Subatmospheric pressure can result in air inflow into the pipe from the pressure difference between the atmosphere and the inside of the pipe, which in a postulated pipe break scenario can lead to the breakdown of the cooling pump. In this study, a plant-scale experiment was conducted to study air inflow in large piping systems by considering the actual operational conditions of an advanced research reactor. The air inflow rate was measured, and the entrained air was visualized to investigate the behavior of air inflow and flow regime depending on the pipe break size. In addition, the developed driftflux model for a large vertical pipe with a diameter of 600 mm was compared with other correlations. The flow regime transition in a large vertical pipe under downward flow was also studied using the newly developed drift-flux model. Consequently, the characteristics of two-phase flow in a large vertical pipe were found to differ from those in small vertical pipes where liquid recirculation was not dominant.
황돈관,강순호,최낙준,조항진 한국원자력학회 2024 Nuclear Engineering and Technology Vol.56 No.1
In nuclear thermal-hydraulic system codes, most correlations used for vertical pipes, under downward two-phase flow, have been developed considering small pipes or pool systems. This suggests that there could be uncertainties in applying the correlations to accident scenarios involving large vertical pipes owing to the difference in the characteristics of two-phase flows, or flow conditions, between large and small pipes. In this study, we modified the Multi-dimensional Analysis of Reactor Safety KINS Standard (MARS-KS) code using correlations, such as the drift-flux model and two-phase multiplier, developed in a plant-scale air-inflow experiment conducted for a pipe of diameter 600 mm under downward two-phase flow. The results were then analyzed and compared with those based on previous correlations developed for small pipes and pool conditions. The modified code indicated a good estimation performance in two plant-scale experiments with large pipes. For the siphonbreaking experiment, the maximum errors in water flow for modified and original codes were 2.2% and 30.3%, respectively. For the air-inflow accident experiment, the original code could not predict the trend of frictional pressure gradient in two-phase flow as 〈jg〉/〈j〉 increased, while the modified MARS-KS code showed a good estimation performance of the gradient with maximum error of 3.5%.
단면 고열속부하된 원형관과 Swirl tube의 열수력학적 성능 비교
임지환(Ji Hwan Lim),황돈관(Donkoan Hwang),박민규(Minkyu Park),오훈교(Hoongyo Oh),김무환(Moo Hwan Kim),조항진(HangJin Jo) 대한기계학회 2020 대한기계학회 춘추학술대회 Vol.2020 No.12
Currently, nuclear fusion power is in the spotlight as a future energy source with high efficiency energy, fuel abundance, eco-friendliness, and stability of power generation. In this study, sub-cooled flow boiling heat transfer performance was compared and evaluated for smooth channel and swirl tube(twist ratio=3, 5) that are expected to be utilized as divertor cooling channels. To simulate the fusion thermal condition, a test loop was constructed to withstand high temperature (150℃) and high pressure (20 bar). In addition, in order to load a high heat flux under one-side heating condition, the authors developed one-side joule heating system capable of loading heat flux up to an effective heat flux of 10.45MW/�㎡. Using these equipment, authors compared and analyzed the thermal-hydraulic performance in various ways by evaluating the critical heat flux (CHF), heat transfer coefficient (HTC), and differential pressure for the smooth channel and swirl tube.
단면 열속부하된 Hypervapotron 냉각채널의 과냉비등상황 임계열유속 실험적 연구
임지환(Ji Hwan Lim),박민규(Min Gyu Park),오훈교(Hun Gyo Oh),황돈관(Don Koan Hwang),김무환(Moo Hwan Kim),조항진(Hang Jin Jo) 대한기계학회 2019 대한기계학회 춘추학술대회 Vol.2019 No.11
In Korea, KSTAR’s tokamak long-time operation technique has reached 70 seconds in 2016, and steady-state plasma control technique is estimated to reach world-class level. However, the lack of cooling capacity to maintain the thermal steady state of the tokamak is becoming a severe engineering problem in preparation for K-DEMO. To simulate the fusion thermal condition, a test loop was constructed to withstand high temperature (150℃) and high pressure (20 bar). Through this loop, CHF experiments were carried out for the Hypervapotron channel and the results were compared with Baxi Hypervapotron CHF correlation. But, a high error average of 173.46% was found, and the Ja number index was changed to 0.45 in consideration of the 10 bar pressure to enable a reasonable prediction of 6.85% lower error. If CHF modeling and verification test is performed later, it will be possible to predict CHF over a wider range.
과냉유동비등 상황에서의 직사각형 채널의 비등시작점에 대한 실험적 연구
임지환(Ji Hwan Lim),이수원(Su Won Lee),오훈교(Hoongyo Oh),박민규(Minkyu Park),황돈관(Donkoan Hwang),김무환(Moo Hwan Kim),조항진(HangJin Jo) 대한기계학회 2021 대한기계학회 춘추학술대회 Vol.2021 No.11
The cooling system of a fusion power plant is safe to operate in single-phase to avoid flow instability due to vapour or reaching a potential critical heat flux. However, there are very few cases of experimental evaluation of onset of nucleate boiling (ONB) under one-side high heat load conditions inside the tokamak. Therefore, ONB of a rectangular channel was evaluated under one-side heating condition. As a result of analyzing the effect of system parameters on ONB heat flux, it was confirmed that ONB heat flux had a proportional relationship with flow and sub-cooling, and an inverse relationship with pressure. In addition, as a result of evaluating various types of ONB correlations, it was difficult to find correlations that predict ONB with high accuracy under one-side high heat load conditions. This proves that it is essential to develop a new ONB correlation for application to the tokamak cooling system.