• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.433-439
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• 1994

영광 3, 4호기는 ABB-CE 사의 System 80 원자로와 비교해서 원자로 내부 구조물의 수력학적 설계 목적과 방법 이 동일하고, 단지 원자로의 크기와 출력이 상대적으로 작아진 내부 구조물이 축소된 형태이다. 따라서 System 80 유동 모델 시험에서 측정된 실험 결과로부터 영광 3, 4호기 연료 집합체 수에 맞게 보간법을 사용하여 보수적으로 유량 분포를 구하고 영광 3, 4호기 유동 모델 시험에서 얻어진 유량 분포와 비교하여 원자로의 수력학적 특성을 검토하고 자각에 대해 열적 여유도를 구하여 이런 경우에 원자로 유동 모델 시험을 수행하지 않고 이전의 실험 결과를 설계에 사용할 수 있는 가에 대해 연구하였다.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.37 no.9
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• pp.855-862
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• 2013

A series of 1/5 scale-down reactor flow distribution tests had been conducted to determine the hydraulic characteristics of an APR+ (Advanced Power Reactor Plus), which were used as the input data for an open core thermal margin analysis code. In this study, to examine the applicability of computational fluid dynamics with the porous model to the analysis of APR+ internal flow, simulations were conducted using the commercial multi-purpose computational fluid dynamics software ANSYS CFX V.14. It was concluded that the porous domain approach for some reactor internal structures could adequately predict the flow characteristics inside a reactor in a qualitative manner. If sufficient computational resources are available, the predicted core inlet flow distribution is expected to be more accurate by considering the real geometry of the internal structures, especially upstream of the core inlet.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.38 no.3
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• pp.271-277
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• 2014

Internal structures, especially those located in the upstream of a reactor core, may have a significant influence on the core inlet flow rate distribution depending on both their shapes and the relative distance between the internal structures and the core inlet. In this study, to examine the effect of the reactor internal structure geometry treatment method on the prediction accuracy for the scale-down APR+ flow distribution, simulations with real geometry modeling were conducted using ANSYS CFX R.14, a commercial computational fluid dynamics software, and the predicted results were compared with those of the porous medium assumption. It was concluded that the core inlet flow distribution could be predicted more accurately by considering the real geometry of the internal structures located in the upstream of the core inlet. Therefore, if sufficient computational resources are available, an exact representation of these internal structures, for example, lower support structure bottom plate and ICI nozzle support plate, is needed for the accurate simulation of the reactor internal flow.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.37 no.12
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• pp.1175-1183
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• 2013

Even if some CFD software developers and its users think that a state-of-the-art CFD software can be used to reasonably solve at least single-phase nuclear reactor safety problems, there remain limitations and uncertainties in the calculation result. From a regulatory perspective, the Korea Institute of Nuclear Safety (KINS) is presently conducting the performance assessment of commercial CFD software for nuclear reactor safety problems. In this study, to examine the prediction performance of commercial CFD software with the porous model in the analysis of the scale-down APR (Advanced Power Reactor Plus) internal flow, a simulation was conducted with the on-board numerical models in ANSYS CFX R.14 and FLUENT R.14. It was concluded that depending on the CFD software, the internal flow distribution of the scale-down APR was locally somewhat different. Although there was a limitation in estimating the prediction performance of the commercial CFD software owing to the limited amount of measured data, CFX R.14 showed more reasonable prediction results in comparison with FLUENT R.14. Meanwhile, owing to the difference in discretization methodology, FLUENT R.14 required more computational memory than CFX R.14 for the same grid system. Therefore, the CFD software suitable to the available computational resource should be selected for massively parallel computations.

• Journal of Energy Engineering
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• v.23 no.4
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• pp.95-105
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• 2014

The CUPID code has been developed for a transient, three-dimensional, two-phase flow analysis at a component scale. It has been validated against a wide range of two-phase flow experiments. Especially, to assess its applicability to single- and two-phase flow analyses in the Calandria vessel of a CANDU nuclear reactor, it was validated using the experimental data of the 1/4-scaled facility of a Calandria vessel at the STERN laboratory. In this study, a preliminary thermal-hydraulic analysis of the CANDU reactor moderator tank using the CUPID code is carried out, which is based on the results of the previous studies. The complicated internal structure of the Calandria vessel and the inlet nozzle was modeled in a simplified manner by using a porous media approach. One of the most important factors in the analysis was found to be the modeling of the tank inlet nozzle. A calculation with a simple inlet nozzle modeling resulted in thermal stratification by buoyance, leading to a boiling from the top of the Calandria tank. This is not realistic at all and may occur due to the lack of inlet flow momentum. To improve this, a new nozzle modeling was used, which can preserve both mass flow and momentum flow at the inlet nozzle. This resulted in a realistic temperature distribution in the tank. In conclusion, it was shown that the CUPID code is applicable to thermal-hydraulic analysis of the CANDU reactor moderator tank using the cost-effective porous media approach and that the inlet nozzle modeling is very important for the flow analysis in the tank.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.130-130
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.590-595
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• 1998

차세대 원자로(KNGR : Korea Next Generation Reactor)에는 새로운 안전개념으로서 피동형 안전주입탱크(Safety Injection Tank. SIT)의 도입을 고려하고 있는데, 이러한 피동형 유량조절기능은 안전주입탱크내의 유체기구(Fluidic device)인 볼텍스챔버(vortex chamber)에 의해 이루어진다. 볼텍스챔버는 내부에서 발생되는 와류강도에 따라 유동저항의 강도가 달라짐을 이용하여 유량을 피동적으로 조절할 수 있는 유체기구이다. 본 연구에서는 볼텍스챔버의 유동특성을 관찰하기 위하여 소규모 실험장치를 구축하고, 이를 이용하여 실험을 수행하였다. 본 연구는 두 단계로 수행되었다. 제1단계 실험에서는 볼텍스챔버의 기하학적 특성이 안전주입탱크의 안전주입수 방출특성에 미치는 영향에 대한 거시적 관점에서의 연구로서. 볼텍스챔버의 기하학적 변수(유입구 및 방출구의 직경)가 안전주입수의 방출과정에서 발생되는 SIT 내의수위 거동, 안전주입수의 방출유량 특성등에 미치는 영향에 대해 중점적으로 고찰하였다 제2단계 실험에서는 1단계 실험에서 관찰된 안전주입탱크의 여러 가지 방출특성과 볼텍스챔버 내부 유동장의 유동특성과의 관련성을 규명하기 위해 PIV (Particle Image Velocimetry)를 이용하여 볼텍스챔버의 기하학적 변수에 따른 유동장 내부의 국소 유속분포를 측정하였다.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.2088-2093
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• 2008

In this study, flow field measurement of the Pebble Bed Reactor(PBR) for the High Temperature Gas-cooled Reactor(HTGR) was performed. Large number of pebbles in the core of PBR provides complicated flow channel. Due to the complicated geometries, numerical analysis has been intensively made rather than experimental observation. However, the justification of computational simulation by the experimental study is crucial to develop solid analysis of design method. In the present study, a wind tunnel installed with pebbles stacked was constructed and equipped with the Particle Image Velocimetry(PIV). We designed the system scaled up to realize the room temperature condition according to the similarity. The PIV observation gave us stagnation points, low speed region so that the suspected high temperature region can be identified. With the further supplementary experimental works, the present system may produce valuable data to justify the Computational Fluid Dynamics(CFD) simulation method.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.27 no.4
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• pp.220-227
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• 2015

Complex thermal hydraulic characteristics exist inside the reactor because the reactor internals consist of fuel assembly, internal structures and so on. In this study, to examine the effect of Reynolds-Averaged Navier-Stokes (RANS)-based two-equation turbulence models in the analysis of flow distribution inside a 1/5 scaled-down APR+, simulation was performed using the commercial computational fluid dynamics software, ANSYS CFX R.13 and the predicted results were compared with the measured data. It was concluded that reactor internal flow pattern was locally different depending on the turbulence models. In addition, the prediction accuracy of k-${\varepsilon}$ model was superior to that of other two-equation turbulence models and this model predicted the relatively uniform distribution of core inlet flow rate.

• Journal of Energy Engineering
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• v.21 no.4
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• pp.419-426
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• 2012

A transient, three-dimensional, two-phase flow analysis code, CUPID, has been developed in KAERI. In this work, we performed a preliminary analysis using the CUPID code to investigate the thermal-hydraulic behavior of the moderator in the Calandria vessel of a CANDU reactor. At first, we validated the CUPID code using the three experiments that were performed at Stern Laboratories Inc. To avoid the complexity to generate computational mesh around the Calandria tube bundles, a porous media approach was applied for the region. The pressure drop in the porous media zone was modeled by an empirical correlation. The results of the calculations showed that the CUPID code can predict the mixed flow pattern of forced and natural convection inside the Calandria vessel very well. Thereafter, the analysis was extended to a two-phase flow condition. Also, the local maximum temperature in the Calandria vessel was plotted as a function of the injection flow rate, which may be utilized to predict the local subcooling margin.