• 대한기계학회논문집
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• 제16권9호
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• pp.1780-1787
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• 1992

본 연구에서는 대수층을 시뮬레이션한 실험장치를 통하여 온도분포를 관찰하 고 자연대류에 관한 단순화된 이론적 모델을 개발하여 실험치와 비교 검토하고자 한 다.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3320-3325
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• 2023

The scaled water-cooled Reactor Cavity Cooling System (RCCS) experimental facility reproduces a passive safety feature to be implemented in Generation IV nuclear reactors. It keeps the reactor cavity and other internal structures in operational conditions by removing heat leakage from the reactor pressure vessel. The present work uses Flownex one-dimensional thermal-fluid code to model the facility and predict the experimental thermal-hydraulic behavior. Two representative steady-state cases defined by the bulk volumetric flow rate are simulated (Re = 2,409 and Re = 11,524). Results of the cavity outlet temperature, risers' temperature profile, and volumetric flow split in the cooling panel are also compared with the experimental data and RELAP system code simulations. The comparisons are in reasonable agreement with the previous studies, demonstrating the ability of Flownex to simulate the RCCS behavior. It is found that the low Re case of 2,409, temperature and flow split are evenly distributed across the risers. On the contrary, there's an asymmetry trend in both temperature and flow split distributions for the high Re case of 11,524.

• 한국산업융합학회 논문집
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• 제26권6_2호
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• pp.1109-1115
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• 2023

The shell and tube-type heat exchanger was the most utilized in industrial field because of its simple structure and wide operation conditions and so on. This study was performed to investigate the characteristics of behavior of thermal flow according to operation condition of small-sized shell and tube-type heat exchanger. The operation conditions, here, were set up to flow rate of hot air with temperature of 100℃, number of baffle and cut rate of baffle(BCR) using numerical analysis. As the results, both mean relative pressure and relative pressure drop was increased with quadratic curve in case of less than BCR 25%, however, decreased linearly in case of more than BCR 25%. The collision with first baffle by flow velocity and temperature, of hot air, respectively, was depended on BCR. Further it showed that the behaviors between flow velocity and temperature were almost similar.

• International Journal of Air-Conditioning and Refrigeration
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• 제16권1호
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• pp.22-29
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• 2008

A sequence of visually experimental observations was conducted to investigate the flow boiling and two-phase flow in a coiled tube. Different boiling modes and bubble dynamical evolutions were identified for better recognizing the phenomena and understanding the two-phase flow evolution and heat transfer mechanisms. The dissolved gases and remained vapor would serve as foreign nucleation sites, and together with the effect of buoyancy, centrifugal force and liquid flow, these also induce very different flow boiling nucleation, boiling modes, bubble dynamical behavior, and further the boiling heat transfer performance. Bubbly flow, plug flow, slug flow, stratified/wavy flow and annular flow were observed during the boiling process in the coiled tube. Particularly the effects of flow reconstructing and thermal non-equilibrium release in the bends were noted and discussed with the physical understanding. Coupled with the effects of the buoyancy, centrifugal force and inertia or momentum ratio of the two fluids, the flow reconstructing and thermal non-equilibrium release effects have critical importance for flow pattern in the bends and flow evolution in next straight sections.

• 한국태양에너지학회:학술대회논문집
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• 한국태양에너지학회 2008년도 추계학술발표대회 논문집
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• pp.273-279
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• 2008

The purpose of this study is not only to evaluate thermal performance but also to find the stress behavior of heat transfer tubes under the part load operation in Heat Recovery Steam Generator. Flow analysis was performed to know the behavior of exhaust gas from gas turbine and thermal performance was calculated using distribution of hot exhaust velocity. In addition, tubes temperature during operation were gathered from actual plant to verify the uneven flow distribution under part load operation. Stress analysis was performed using tubes temperature data gathered from actual plant under both part and full load operations to know the stress behavior of tubes.

• International Journal of Air-Conditioning and Refrigeration
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• 제11권1호
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• pp.10-16
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• 2003

The present study performs an experimental measurement on transient thermal response of an air-conditioned space by a variable air volume (VAV) system with a PI(pro-portional-integral) control logic. A thermal chamber with a PI controlled VAV unit is constructed to verify the previously suggested stratified multi-zone model. The effects of thermal parameters and control parameters such as supply air temperature and PI control factor are investigated by implementing the thermal chamber test. The experimental results obtained show that transient behavior of the air-conditioned space-temperature is in good accordance with the simulation results of the stratified thermal model.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2332-2342
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• 2024

Lead-cooled fast reactors (LFRs) have a wide range of application scenarios, which require the thermal-hydraulic characteristics of LFRs to be reliable. In the present paper, the Lead-cooled fast reactor Thermal-Hydraulic Analysis Code LETHAC was developed, including the models of pipe, heat exchanger, and pool. To verify the correctness of LETHAC, two experimental facilities and three experimental cases were selected, including GFT and PLOFA tests for NACIE-UP and Test-1 for CIRCE. The calculated results show the same and consistent trend with the experimental data, but there are some discrepancies. It can be found that LETHAC is suitable and reliable in predicting the transient behavior of lead-cooled system.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.641-646
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• 1998

In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUIN $T^{™}$ has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA).

• 한국공작기계학회논문집
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• 제17권3호
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• pp.20-26
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• 2008

Nanoimprint lithography(NIL) is becoming next generation lithography of significant interest due to its low cost and a potential patterning resolution of 10nm or less. Success of the NIL relies on the adequate conditions of pressure, temperature and time. To have the adequate conditions for NIL, one has to understand the polymer flowing behavior during the imprinting process. In this paper, an analytical approach of polymer flow in thermal NIL was performed based on the squeeze flow with partial slip boundary conditions. Velocity profiles and pressure distributions of the polymer flow were obtained and imprinting forces and residual thickness were predicted with the consideration of the slip velocity between the polymer and the mold/substrate. The results show that the consideration of the slip is very important for investigating the polymer flow in Thermal NIL.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1995년도 춘계학술발표회 초록집
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• pp.113-116
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• 1995

Simplified approach has been adopted for the prediction of the thermal behavior of CANDU reactor core during power transients. Based on the assumption that the ratio of mass flow rate for each core channel does not vary during the transient, quasy-steady state analysis technique is applied with predicted core inlet boundary conditions(total mass flow rate and specific enthalpy). For restricted transient case, the presented method shows functionally reasonable estimation of core thermal behavior which could be implemented in the fast running reactor simulation program.