• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1318-1325
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• 2017

Direct Numerical Simulation (DNS) serves as an irreplaceable tool to probe the complexities of multiphase flow and identify turbulent mechanisms that elude conventional experimental measurement techniques. The insights unlocked via its careful analysis can be used to guide the formulation and development of turbulence models used in multiphase computational fluid dynamics simulations of nuclear reactor applications. Here, we perform statistical analyses of DNS bubbly flow data generated by Bolotnov ($Re_{\tau}=400$) and LueTryggvason ($Re_{\tau}=150$), examining single-point statistics of mean and turbulent liquid properties, turbulent kinetic energy budgets, and two-point correlations in space and time. Deformability of the bubble interface is shown to have a dramatic impact on the liquid turbulent stresses and energy budgets. A reduction in temporal and spatial correlations for the streamwise turbulent stress (uu) is also observed at wall-normal distances of $y^+=15$, $y/{\delta}=0.5$, and $y/{\delta}=1.0$. These observations motivate the need for adaptation of length and time scales for bubble-induced turbulence models and serve as guidelines for future analyses of DNS bubbly flow data.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2010년도 제35회 추계학술대회논문집
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• pp.386-392
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• 2010

상용 전산유체해석 프로그램을 사용하여 오리피스형 분사기에서 수직분사 액체 제트의 다상유동을 해석하였다. 본 연구의 목적은 오리피스형 분사기 분무 특성을 이해하고 연소기 내부 위치에 따른 유동조건와 항력계수와의 관계식을 구하는 것이다. 수치해석 결과 모형 램젯 연소기에서 수직분사 유동해석은 난류점성모델인 Realizable $k-{\varepsilon}$ 모델과 다상유동모델인 DPM 모델이 유효함을 확인하였다. 또한 오리피스형 분사기에서 레이저 흡수법을 사용하여 측정한 실험결과(증기농도분포)는 수치해석 결과와 잘 일치하였다.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1945-1954
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• 2020

Subchannel code is one of the effective simulation tools for thermal-hydraulic analysis in nuclear reactor core. In order to reduce the computational cost and improve the calculation efficiency, empirical correlation of turbulent mixing coefficient is employed to calculate the lateral mixing velocity between adjacent subchannels. However, correlations utilized currently are often fitted from data achieved in central channel of fuel assembly, which would simply neglect the wall effects. In this paper, the CFD approach based on spectral element method is employed to predict turbulent mixing phenomena through gaps in 3 × 3 bare tight lattice rod bundle and investigate the flow pulsation through gaps in different positions. Re = 5000,10000,20500 and P/D = 1.03 and 1.06 have been covered in the simulation cases. With a well verified mesh, lateral velocities at gap center between corner channel and wall channel (W-Co), wall channel and wall channel (W-W), wall channel and center channel (W-C) as well as center channel and center channel (C-C) are collected and compared with each other. The obvious turbulent mixing distributions are presented in the different channels of rod bundle. The peak frequency values at W-Co channel could have about 40%-50% reduction comparing with the C-C channel value and the turbulent mixing coefficient β could decrease around 25%. corrections for β should be performed in subchannel code at wall channel and corner channel for a reasonable prediction result. A preliminary analysis on fluctuation at channel gap has also performed. Eddy cascade should be considered carefully in detailed analysis for fluctuating in rod bundle.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3909-3918
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• 2022

In the event of a severe accident, the reactor core may melt due to insufficient cooling. the high-temperature core melt will have a strong interaction (FCI) with the coolant, which may lead to steam explosion. Steam explosion would pose a serious threat to the safety of the reactors. Therefore, the study of steam explosion is of great significance to the assessment of severe accidents in nuclear reactors. This research focuses on the development of a two-dimensional steam explosion integrated analysis code called SEINA. Based on the semi-implicit Euler scheme, the three-phase field was considered in this code. Besides, the influence of evaporation drag of melt and the influence of solidified shell during the process of melt droplet fragmentation were also considered. The code was simulated and validated by FARO L-14 and KROTOS KS-2 experiments. The calculation results of SEINA code are in good agreement with the experimental results, and the results show that if the effects of evaporation drag and melt solidification shell are considered, the FCI process can be described more accurately. Therefore, it is proved that SEINA has the potential to be a powerful and effective tool for the analysis of steam explosions in nuclear reactors.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2997-3009
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• 2023

The Passive containment Air-cooling System (PAS) can effectively remove the decay heat of the modular small nuclear reactor after an accident. The details of natural convection around the dome, which is a key part of PAS, were investigated numerically in the present study. The thermal dynamics around the dome were studied through the temperature, pressure and velocity contours and the streamlines. Additionally, the formation of the buoyant plume at the top of the dome was investigated. The results show that with the increase of Ra, the lift-off point moves toward the bottom of the dome, and the eddy under the buoyant plume grows larger gradually, which enhances the heat transfer. And the heat transfer along the dome surface with different truncation angles was investigated. As the angle increases, the heat transfer coefficient becomes stronger as well. Consequently, a newly developed heat transfer correlation considering the influence of truncation angle for the dome is proposed based on the simulated results. This study could provide a better understanding of natural convection around the dome of PAS and the proposed correlation could also offer more predictive value in the improvement of nuclear safety.

• 한국전산유체공학회지
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• 제9권1호
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• pp.18-24
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• 2004

The unified simulation for the multiphase flow by predictor-corrector scheme based on CIP method is introduced. In this algorithm, the interface between different phases is identified by a density function and tracked by solving an advection equation. Solid body motion is modeled by the translation and angular motion. The mathematical formulation and numerical results are also described. To verify the efficiency, accuracy and capability of proposed algorithm, two dimensional incompressible cavity flow, the motion of a floating ball into water and a single rising bubble by buoyancy force are numerically simulated by the present scheme. As results, it is confirmed that the present scheme gives an efficient, stable and reasonable solution in the multiphase flow problem.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3754-3767
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• 2023

Turbulent jet discharges in subcooled water pools are essential for safety systems in nuclear power plants, specifically in the pressure suppression pool of boiling water reactors and In-containment Refueling Water Storage Tank of advanced pressurized water reactors. The gas and liquid flow in these systems is investigated using multiphase flow analysis. This field has been extensively examined using a combination of experiments, theoretical models, and Computational Fluid Dynamics (CFD) simulations. ANSYS CFX offers two approaches to model multiphase flow behavior. The non-homogeneous Eulerian-Eulerian Model has been used in this work; it computes global information and is more convenient to study interpenetrated fluids. This study utilized the Large Eddy Simulation Model as the turbulence model, as it is better suited for non-stationary and buoyant flows. The CFD results of this study were validated with experimental data and theoretical results previously obtained. The figures of merit dimensionless penetration length and the dimensionless buoyancy length show good agreement with the experimental measurements. Correlations for these variables were obtained as a function of dimensionless numbers to give generality using only initial boundary conditions. CFD numerical model developed in this research has the capability to simulate the behavior of non-condensable gases discharged in water.

• 대한기계학회논문집B
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• 제33권6호
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• pp.443-453
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• 2009

The Rayleigh-Taylor instability, the bubble rising in both partially and fully filled containers and the droplet splash are simulated by an in-house solution code(PowerCFD), which are typical benchmark problems among multiphase flows with material interface due to density difference. The present method(code) employs an unstructured cell-centered method based on a conservative pressure-based finite-volume method with interface capturing method(CICSAM) in a volume of fluid(VOF) scheme for phase interface capturing. The present results are compared with other numerical solutions found in the literature. It is found that the present method simulates efficiently and accurately complex free surface flows such as multiphase flows with material interface due to both density difference and instability.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2008년도 춘계학술대회논문집
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• pp.572-575
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• 2008

Two-dimensional multiphase flows due to density difference such as the Rayleigh-Taylor instability problem and the droplet splash are simulated by an in-house solution code(PowerCFD). This code employs an unstructured cell-centered method based on a conservative pressure-based finite-volume method with interface capturing method in a volume of fluid(VOF) scheme for phase interface capturing. The present results are compared with other numerical solutions found in the literature. It is found that the present code simulates complex free surface flows such as multiphase flows due to density difference efficiently and accurately.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.162-176
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• 2022

In Lead-based reactor (LBR) severe accident, the meltdown and migration inside the reactor core will lead to fuel fragment concentration, which may further cause re-criticality and even core disintegration. Accurately predicting the migration and solidification behavior of melt in LBR severe accidents is of prime importance for safety analysis of LBR. In this study, the Moving Particle Semi-implicit (MPS) method is validated and used to simulate the migration and solidification behavior. Two main surface tension models are validated and compared. Meanwhile, the MPS method is validated by the L-plate solidification test. Based on the improved MPS method, the migration and solidification behavior of melt in LBR severe accident was studied furthermore. In the Pb-Bi coolant, the melt flows upward due to density difference. The migration and solidification behavior are greatly affected by the surface tension and viscous resistance varying with enthalpy. The whole movement process can be divided into three stages depending on the change in velocity. The heat transfer of core melt is determined jointly by two heat transfer modes: flow heat transfer and solid conductivity. Generally, the research results indicate that the MPS method has unique advantage in studying the migration and solidification behavior in LBR severe accident.