• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2017년도 제48회 춘계학술대회논문집
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• pp.112-115
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• 2017

고공엔진시험설비에서 지상 시험 조건의 시험 가능 여부와 설비 안전 문제를 점검하고자 로켓 엔진 유동을 해석하였다. 진공 챔버를 개방한 상태에서 냉각수를 초음속 디퓨저로 분사하면서 엔진이 작동하는 상황이며, 2차원 축대칭과 플룸, 공기, 냉각수의 3원 혼합물로 가정하였다. 해석 결과 냉각수 유량 200 kg/sec까지 지상 조건 시험이 가능하였다. 그러나 시동 초기 플룸의 역류로 인해 진공 챔버가 고온에 노출되고, 동시에 냉각수 역류로 인해 진공 챔버 내부가 오염되었다. 따라서 충분한 단열 대책과 오염 회피를 위한 작업이 선행되어야 한다.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2006년도 제26회 춘계학술대회논문집
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• pp.130-134
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• 2006

분사면 냉각성능을 개선하기 위한 액체로켓 엔진 연료 매니폴드내의 분리판 설치에 따른 분사균일성 변화를 관찰하였다. 3차원 전산유동해석으로 5개 후보 설계에 대하여 분사균일성을 비교하였으며 최적설계에 대하여 측정결과와 비교함으로써 해석방법을 검증하였다. case I과 II는 매니폴드로 공급되는 유량 전체가 분리판 아래로만 흘러 유속이 크게 증가한다. 하지만 분리판이 끝나는 지점에서의 유속변화와 분사면 중심에서의 유량의 집중으로 분사균일성이 크게 저하된다. 이에 비해 분리판이 입구에서 떨어져 장착된 case III와 IV는 유동이 분리판 위, 아래로 흐를 수 있어 유량집중이 완화되므로 균일한 분사특성을 가진다. 비교한 5가지 설계 중 냉각성능과 분사균일성 측면에서 case IV가 최적으로 판단된다.

• 한국산학기술학회논문지
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• 제21권10호
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• pp.379-385
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• 2020

본 논문은 전투차량에 장착되는 온수히터의 코어부위 냉각수 누수방지를 위한 설계에 관한 것이다. 온수히터는 가온된 냉각수를 승무원실의 히터 코어에 흘려 난방하는 장치로, 전투차량 운용 시 온수히터 코어 부위의 냉각수 누수현상이 발생하는 문제점이 확인되었다. 이 문제점은 주로 코어의 탱크와 튜브 부위의 접합부에서 발생하였는데, 이 부위가 취약하여 고압을 가압하였을 때 누수가 발생한 것으로 추정하였다. 이 문제를 개선하기 위하여 용접방식을 개선하고 온수히터 코어 말단 부위를 높은 압력에서 견딜 수 있는 구조로 변경하였다. 기존 코어와 개선 코어에 대하여 순차적으로 압력을 가하였을 때 기존 코어는 7.0 kgf/㎠에서 누수가 발생하였으며 개선 코어는 17.0 kgf/㎠까지 견고하게 구조를 유지하여 개선이 되었음을 입증하였다. 마지막으로 개선한 구조의 체계 적합성을 입증하기 위하여 성능시험 및 환경시험을 실시하였다. 본 논문의 연구결과를 바탕으로 제작된 개선 온수히터는 전투차량에 적용될 예정이며, 신뢰성 확보를 통한 방위력 향상과 유사 장비의 설계 및 고장분석에도 참고자료가 될 것으로 기대된다.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.1081-1090
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• 2024

The liquid Lead-bismuth eutectic is used as the coolant for Gen-IV reactor concepts. However, due to its strong corrosive and high operating temperature, it is difficult to accurately measure the flow rate in long-term operating conditions. Venturi flowmeter is a simple structured flowmeter, which plays a very important role in the flow measurement of high-temperature liquid metals, especially since the existing flowmeters are difficult to be competent. It has the advantages of easy maintenance and stable operation. Therefore, it is necessary to study the operating conditions of the venturi flowmeter under high-temperature conditions. This work performs a series of simulations of the fluid-solid interaction between the flow liquid metal and venturi flowmeter with COMSOL software, including the dimensional sensitivity analysis of the venturi flowmeter to explore the most suitable structure and parameters for liquid heavy metal, the sensitivity analysis of the geometric parameters of the venturi tube on the varying conditions. It shows that when the contraction angle of the venturi flowmeter is 33°, the diffusion angle is 13°, the diameter of the throat is 8 mm, and the temperature of the lead-bismuth eutectic is 733.15 K, it is most suitable for the measurement in the lead-bismuth circuit.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.949-958
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• 2024

In contrast to all-time flooded small modular reactor (SMR) systems, an in-kind flooding safety system (FSS) has been proposed as a passive safety system applicable to small modular reactors (SMRs) that adopt a metal containment vessel (MCV). Under transient conditions, the FSS can provide emergency cooling to dry reactor cavities and sustain long-term coolability using re-acquired evaporated steam in the reactor building on demand. When designing an FSS, the effect of the flooding flow area is vital as it affects the overall accident sequence and safety. Therefore, in this study, a MELCOR model of a reference SMR is developed and numerical analysis is performed under postulated accident scenarios. Without flooding, the MCV pressure of the reactor module exceeds the design pressure before core damage. To prevent core damage, an emergency flooding strategy is devised using various flow path parameters and requirements to ensure an adequate emergency coolant supply before the core damage is investigated. The results indicate that a flow area exceeding 0.02 m2 is required in the FSS to prevent MCV overpressure and core damage. This study is the first to report a strategic analysis for appropriately sizing an FSS flooding valve applicable to innovative SMRs.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4477-4490
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• 2023

Fuel locking mechanisms (FLMs) are essential in upward-flow research reactors to prevent accidental fuel separation from the core during reactor operation. This study presents a novel design concept for a remotely controlled plate-type nuclear fuel locking mechanism. By employing electromagnetic field analysis, we optimized the design of the electromagnet for fuel unlocking, allowing the FLM to adapt to various research reactor core designs, minimizing installation space, and reducing maintenance efforts. Computational flow analysis quantified the drag acting on the fuel assembly caused by coolant upflow. Subsequently, we performed finite element analysis and evaluated the structural integrity of the FLM based on the ASME boiler and pressure vessel (B&PV) code, considering design loads such as dead weight and flow drag. Our findings confirm that the new FLM design provides sufficient margins to withstand the specified loads. We fabricated a prototype comprising the driving part, a simplified moving part, and a dummy fuel assembly. Through basic operational tests on the assembled components, we verified that the manufactured products meet the performance requirements. This remote-controlled micro locking mechanism holds promise in enhancing the safety and efficiency of plate-type nuclear fuel operation in upflow research reactors.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.356-361
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• 1997

To investigate parametric effect on CHF and to get CHF data, experimental study has been performed with vertical round tubes under the condition of low pressure and low flow (LPLF). Test sections are made of Inconel-625 tube and have the geometry of 8 and 10 mm in diameter, and 0.5 and 1.0 m in heated length. All experiments have been conducted at the pressure of under 9 bar, the mass flux of under 250 kg/$m^2$ and the inlet subcooling of 350 and 450 kJ/kg, for stable upward flow with water as a coolant. Flow regime analysis has been performed for obtained CHF data with Mishima's flow regime map, which reveals that most of the CHF occur in the annular-mist flow regime. General parametric trends of the collected CHF data are consistent with those of previous studies. However, for the pressure effect on CHF, two different are observed; For relatively high mass flux, CHF increases with pressure and far lower mass flux, CHF decrease with pressure. Using modern data regression tool, ACE algorithm, two new CHF correlations for LPLF condition are developed based on local condition and inlet condition, respectively. The developed CHF correlations show better prediction accuracy compared with existing CHF prediction methods.

• Nuclear Engineering and Technology
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• 제12권3호
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• pp.153-162
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• 1980

냉각재상실사고의 재관수 단계중 연료봉 피복재의 온도거동 및 열전달 기구를 파악하는 것은 비상노심냉각계통 및 원자로의 안전성해석에 중요하다. 냉각재유동채널의 방위가 rewetting과정에 미치는 영향을 연구하기 위하여 수직 및 수경 유동채널을 이용한 실험을 수행하였으며, 노심이 수평압력관으로 구성되어 있는 CANDU원자로에 관한 실험을 중점적으로 수행하여 그 결과를 수직채널의 결과와 비교 하였다. 또한 rewetting현상을 육안관찰가기 위해 환상형 테스트부 및 외부에서 가열되는 석영관을 사용하였다. 실험결과로써 수평채널에서의 rewetting 속도는 유동의 층상 현상에 크게 영향을 받으나 그 평균값은 수직채널리 경우와 큰차이없음을 알 수 있었다.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.691-708
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• 2009

In order to analyze thermal hydraulic phenomena during a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in the APR1400 (Advanced Power Reactor 1400 MWe), we performed experimental studies with the SNUF (Seoul National University Facility), a reduced-height and reduce-pressure integral test loop with a scaled down APR1400. We performed experiments dealing with eight test cases under varied tests. As a result of the experiment, the primary system pressure, the coolant temperature, and the occurrence time of the downcomer seal clearing were affected significantly by the thermal power in the core and the SI flow rate. The break area played a dominant role in the vent of the steam. For our analytical investigation, we used the MARS code for simulation of the experiments to validate the calculation capability of the code. The results of the analysis showed good and sufficient agreement with the results of the experiment. However, the analysis revealed a weak capability in predicting the bypass flow of the SI water toward the broken DVI line, and it was insufficient to simulate the streamline contraction in the broken side. We, hence, need to improve the MARS code.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1955-1962
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• 2020

In a sudden shutdown of primary pump or coolant loss accident in a marine nuclear power plant, the primary flow decreases rapidly in a transition process from forced circulation (FC) to natural circulation (NC), and the lower flow enters the steam generator (SG) causing reverse flow in the U-tube. This can significantly compromise the safety of nuclear power plants. Based on the marine natural circulation steam generator (NCSG), an experimental loop is constructed to study the characteristics of reverse flow under middle-temperature and middle-pressure conditions. The transition from FC to NC is simulated experimentally, and the characteristics of SG reverse flow are studied. On this basis, the experimental loop is numerically modeled using RELAP5/MOD3.3 code for system analysis, and the accuracy of the model is verified according to the experimental data. The influence of the flow variation rate on the reverse flow phenomenon and flow distribution is investigated. The experimental and numerical results show that in comparison with the case of adjusting the mass flow discontinuously, the number of reverse flow tubes increases significantly during the transition from FC to NC, and the reverse flow has a more severe impact on the operating characteristics of the SG. With the increase of flow variation rate, the reverse flow is less likely to occur. The mass flow in the reverse flow U-tubes increases at first and then decreases. When the system is approximately stable, the reverse flow is slightly lower than obverse flow in the same U-tube, while the flow in the obverse flow U-tube increases.