• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1487-1503
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• 2019

The methods and performance of a pin-level nuclear reactor core thermal-hydraulics (T/H) code ESCOT employing the drift-flux model are presented. This code aims at providing an accurate yet fast core thermal-hydraulics solution capability to high-fidelity multiphysics core analysis systems targeting massively parallel computing platforms. The four equation drift-flux model is adopted for two-phase calculations, and numerical solutions are obtained by applying the Finite Volume Method (FVM) and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE)-like algorithm in a staggered grid system. Constitutive models involving turbulent mixing, pressure drop, and vapor generation are employed to simulate key phenomena in subchannel-scale analyses. ESCOT is parallelized by a domain decomposition scheme that involves both radial and axial decomposition to enable highly parallelized execution. The ESCOT solutions are validated through the applications to various experiments which include CNEN $4{\times}4$, Weiss et al. two assemblies, PNNL $2{\times}6$, RPI $2{\times}2$ air-water, and PSBT covering single/two-phase and unheated/heated conditions. The parameters of interest for validation include various flow characteristics such as turbulent mixing, spacer grid pressure drop, cross-flow, reverse flow, buoyancy effect, void drift, and bubble generation. For all the validation tests, ESCOT shows good agreements with measured data in the extent comparable to those of other subchannel-scale codes: COBRA-TF, MATRA and/or CUPID. The execution performance is examined with a mini-sized whole core consisting of 89 fuel assemblies and for an OPR1000 core. It turns out that it is about 1.5 times faster than a subchannel code based on the two-fluid three field model and the axial domain decomposition scheme works as well as the radial one yielding a steady-state solution for the OPR1000 core within 30 s with 104 processors.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회B
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• pp.2676-2681
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• 2008

In previous studies, Convex cyclone are proposed to reduce pressure loss which are design cyclone wall with a single continuous curve. Studies about a prediction model for pressure loss and cut-size has focused on conventional cylinder-on-con cyclone. Therefore, the models do not perform well for uncommon design. In this study, a predict model for pressure loss and cut-size depend on cyclone wall curvature are developed. The tangential velocity below vortex-finder is obtained with consideration about friction area and momentum loss on the cyclone wall, and with this the variation of vortex-core and core velocity is obtained. Pressure loss is predicted using a Rankine vortex hypothesis. The prediction results are well agreed with experiments and CFD results.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2008년도 춘계학술대회논문집
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• pp.292-295
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• 2008

In a SCWR (SuperCritical pressure Water cooled Reactor), the coolant temperature initially at below the pseudo-critical temperature at the bottom of a reactor core increases as the coolant flows upward through the sub-channels of the fuel assemblies, and it finally becomes higher than the pseudo-critical temperature when it leaves the reactor core. At certain conditions, heat transfer deterioration occurs near the pseudo-critical temperature and it may cause a drastic rise of the fuel surface temperature resulting a fuel failure. Therefore, an accurate estimation of the heat transfer coefficient is very important for the thermal-hydraulic design of a reactor core. An experiment on heat transfer to the vertically upward flowing $CO_2$ at a supercritical pressure in a circular tube were performed at KAERI. The internal diameter of the test section is 6.32 mm, which corresponds to the hydraulic diameter of a sub-channel in the conceptional design proposed by KAERI. The test range of the mass flux is 285 to 1200 kg/m$^2$s and the maximum heat flux is 170 kW/m$^2$. The inlet pressure is maintained at 8.12 MPa, which is 1.1 times the critical pressure. A new correlation, which covers both the normal and deterioration heat transfer regimes was proposed and compared with the estimations by exiting correlations.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.273-286
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• 2007

This paper presents the research activities performed to date for the development of a supercritical pressure water-cooled reactor (SCWR) in Korea. The research areas include a conceptual design of an SCWR with an internal flow recirculation, a reactor core conceptual design, a heat transfer test with supercritical $CO_2$, an adaptation of an existing safety analysis code to the supercritical pressure condition, and an evaluation of candidate materials through a corrosion study. Methods to reduce the cladding temperature are introduced from two different perspectives, namely, thermal-hydraulics and core neutronics. Briefly described are the results of an experiment on the heat transfer at a supercritical pressure, an experiment that is essential for the analysis of the subchannels of fuel assemblies and the analysis of a system safety. An existing system code has been adapted to SCWR conditions, and the process of a first-hand validation is presented. Finally, the corrosion test results of the candidate materials for an SCWR are introduced.

• 한국자동차공학회논문집
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• 제9권5호
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• pp.9-16
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• 2001

End-gas temperatures were measured using CARS technique in a conventional DOHC spark- ignition engine fueled with PRF80. The measured pressure data were analyzed using band pass filter method. The measured CARS temperatures were compared with adiabatic core temperatures calculated from measured pressures. Significant heating by pre-flame reaction in the end gas zone was observed in the late part of compression stroke under both knocking and non-knocking conditions. CARS temperatures measured at 10 crank angle degree before knock occurrence was higher than adiabatic core temperatures. These results indicate that there exist some exothermic reactions in low pressure and temperature region. CARS temperatures began to be higher than the adiabatic core temperature when the end-gas temperatures reached look. The temperature elevation due to the pre-flame reaction correlated better with CARS temperature than with cylinder pressure.

• 소성∙가공
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• 제15권9호
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• pp.635-640
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• 2006

Sandwich sheet with inner structure is expected to find many applications because of high stiffness to mass ratio. However, low resistance to the compressive pressure in the thickness direction is a dominating factor in the formability of sandwich sheet. In this study, sandwich sheet with pyramid type core is considered. For the compressive characteristics in the thickness direction, experiments and finite element simulations are carried out. In the experiment, deformation behavior is observed and discussed as the compression proceeds. It is shown that a corresponding finite element simulation can give a reasonable agreement with experiment in terms of maximum pressure. However, simulation shows some discrepancy from the experiment in terms of compressive pressure-displacement characteristics. The reasons for this discrepancy are studied in the geometrical imperfectness of sandwich sheet. It is also observed that most of deformation is dominated by buckling mode of pyramid legs.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.237-241
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• 2001

The uncertainties associated with fluence calculation at the pressure vessel have been evaluated for the Korean Next Generation Reactor, APR1400. To obtain uncertainties, sensitivity analyses were performed for each of the parameters important to calculated fast neutron fluence. Among the important parameters to the overall uncertainties, reactor modeling and core neutron source were examined. Mechanical tolerances, composition and density variations in the reactor materials as well as application of $r-{\theta}$ geometry in rectilinear region contribute to uncertainty in the reactor modeling. Depletion and buildup of fissile nuclides, instrument error related to core power level, uncertainty of fuel pin burnup, and variation of long-term axial peaking factors are main contributors to the core neutron source uncertainty. The sensitivity analyses have shown that the uncertainty in the fluence calculation at the reactor pressure vessel is +12%.

• Fibers and Polymers
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• 제6권2호
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• pp.146-150
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• 2005

In this paper, friction of air-jet textured yams is investigated. Using a friction measuring apparatus fabricated inhouse, dynamic friction forces of the yams under yarn-to-metal (YM) and yam-to-yam (YY) rubbing modes are measured. The influence of processing variables of air-jet texturing viz., overfeed, air pressure, dry/wet texturing and normal/core-and-effect texturing on dynamic friction is analysed. The results indicate that friction force increases with increasing rubbing speeds and yam input tension. YM dynamic friction decreases initially and then starts to increase at higher overfeeds. YY dynamic friction increases with increasing overfeed. YM dynamic friction decreases with an increase in air pressure while an opposite trend is observed for YY friction. Wet textured yams have higher friction than dry textured yams. Core wetted coreand-effect textured yams have higher friction than normal textured yams.

• 지구물리와물리탐사
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• 제15권2호
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• pp.102-115
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• 2012

이 논문은 퇴적물내 포함되어 있는 가스하이드레이트의 포화도를 계산하는 여러 방법에 대하여 토론하고자 한다. 가스하이드레이트의 포화도를 계산하는 방법은 물리검층 자료를 이용하는 방법과 코어자료(압력코어 포함)를 이용하는 방법, 그리고 탄성파 탐사자료로부터 얻을 수 있는 속도 자료를 이용하는 방법 등 크게 세가지 방법으로 나눌 수 있다. 물리검층 자료중 전기비저항 자료를 이용하는 방법의 경우 Archie 식을 주로 이용하는데 이 경우 각각의 변수 값을 정확하게 정의하는 게 중요하다. 또한 가스하이드레이트의 산출형태도 포화도 계산에 큰 영향을 주기 때문에 주의해야 한다. 코어자료를 이용하는 경우 공극수의 염소량을 측정하는 방법과 압력코어를 취득할 경우 이를 이용하는 방법이 있다. 지금까지 발표된 정량적이고 가장 정확한 가스하이드레이트 포화도값을 구할 수 있는 방법이 압력코어를 이용하는 것이다. 그러나 이는 비용과 시간이 많이 소요되기 때문에 연속적인 자료를 얻기가 어렵다는 단점이 있다. 지금까지 발표된 가스하이드레이트 포화도 값을 비교해 보면 전기비저항 값을 이용한 경우가 가장 높은 값을 압력코어를 이용하여 측정한 경우가 가장 낮은 값을 보여주는 경향이 있다. 그러나 이러한 값이 모든 경우에 있어서 절대적인 경향을 보여준다고 볼 수는 없다. 그러므로 가스하이드레이트의 포화도를 정확하게 계산하기 위해서는 여러 가지 방법을 이용하여 계산해야 하며 이를 비교하여 가장 적절한 값을 사용해야 할 것이다.

• 품질경영학회지
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• 제43권3호
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• pp.299-312
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• 2015

Purpose: This paper analyzes the phenomenon on the degradation of the electrical performance by the pressure in the manufacturing process of sandwich type radomes. Methods: This paper consists of two steps to analyze the relation between the electrical performance and the pressure. First, the thickness of the core of the flat panels which were fabricated with different pressure was measured with the microscope, and then the electrical performance of the flat panels was analyzed with simulation and experiment. Based on the results of the electrical performance and the measured thickness with respect to the flat panels, the relation between the electrical performance and the applied pressure in the manufacturing process was analyzed. Results: The simulated and measured results with respect to the flat panel shows that the high pressure results in the nonuniform thickness of the core, which is applied to the radome fabrication. As a result, the degradation of the electrical performance occurs because the unintended scattered field is generated as the electromagnetic wave transmits (or impinges upon) the radome. Furthermore, we observed that the electrical performance of both the flat panel and the radome got worse as the high pressure was applied. Conclusion: Through simulation and experiment, therefore, it is demonstrated that for the high pressure in the manufacturing process the nonuniform thickness of the core increases and results in the degradation of the electrical performance of the radome.