• Nuclear Engineering and Technology
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• 제44권8호
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• pp.871-888
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• 2012

This paper aims at presenting the state of the art, the recent progress, and the perspective for the future, in the modelling of two-phase flow in the horizontal legs of a PWR. All phenomena relevant for safety analysis are listed first. The selection of the modelling approach for system codes is then discussed, including the number of fluids or fields, the space and time resolution, and the use of flow regime maps. The classical two-fluid six-equation one-pressure model as it is implemented in the CATHARE code is then presented and its properties are described. It is shown that the axial effects of gravity forces may be correctly taken into account even in the case of change of the cross section area or of the pipe orientation. It is also shown that it can predict both fluvial and torrential flow with a possible hydraulic jump. Since phase stratification plays a dominant role, the Kelvin-Helmholtz instability and the stability of bubbly flow regime are discussed. A transition criterion based on a stability analysis of shallow water waves may be used to predict the Kelvin-Helmholtz instability. Recent experimental data obtained in the METERO test facility are analysed to model the transition from a bubbly to stratified flow regime. Finally, perspectives for further improvement of the modelling are drawn including dynamic modelling of turbulence and interfacial area and multi-field models.

• Nuclear Engineering and Technology
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• 제37권6호
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• pp.511-524
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• 2005

Thermalhydraulic reactor simulation of tomorrow will require a new generation of codes combining at least three scales, the CFD scale in open medium, the component scale and the system scale. DNS will be used as a support for modelling more macroscopic models. NEPTUNE is such a new generation multi-scale platform developed jointly by CEA-DEN and EDF-R&D and also supported by IRSN and FRAMATOME-ANP. The major steps towards the next generation lie in new physical models and improved numerical methods. This paper presents the advances obtained so far in physical modelling for each scale. Macroscopic models of system and component scales include multi-field modelling, transport of interfacial area, and turbulence modelling. Two-phase CFD or CMFD was first applied to boiling bubbly flow for departure from nucleate boiling investigations and to stratified flow for pressurised thermal shock investigations. The main challenges of the project are presented, some selected results are shown for each scale, and the perspectives for future are also drawn. Direct Numerical Simulation tools with Interface Tracking Techniques are also developed for even smaller scale investigations leading to a better understanding of basic physical processes and allowing the development of closure relations for macroscopic and CFD models.

• Nuclear Engineering and Technology
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• 제42권4호
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• pp.365-376
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• 2010

This paper summarizes the results of a Writing Group on the Extension of CFD codes to two-phase flow safety problems, which was created by the Group for Analysis and Management of Accidents of the Nuclear Energy Agency' Committee on the Safety of Nuclear Installations (NEA-CSNI). Two-phase CFD used for safety investigations may predict small scale flow processes, which are not seen by system thermalhydraulic codes. However, the two-phase CFD models are not as mature as those in the single phase CFD and potential users need some guidance for proper application. In this paper, a classification of various modelling approaches is proposed. Then, a general multi-step methodology for using two-phase-CFD is explained, including a preliminary identification of flow processes, a model selection, and a verification and validation process. A list of 26 nuclear reactor safety issues that could benefit from investigations at the CFD scale is identified. Then, a few issues are analyzed in more detail, and a preliminary state-of-the-art is proposed and the remaining gaps in the existing approaches are identified. Finally, guidelines for users are proposed.

• Advances in concrete construction
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• 제13권5호
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• pp.361-365
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• 2022

The objective of this study is to simulate the two-phase flow in pipes with various two-fluid models and determinate the shear stress. A hyperbolic shear deformation theory is used for modelling of the pipe. Two-fluid models are solved by using the conservative shock capturing method. Energy relations are used for deriving the motion equations. When the initial conditions of problem satisfied the Kelvin Helmholtz instability conditions, the free-pressure two-fluid model could accurately predict discontinuities in the solution field. A numerical solution is applied for computing the shear stress. The two-pressure two-fluid model produces more numerical diffusion compared to the free-pressure two-fluid and single-pressure two-fluid models. Results show that with increasing the two-phase percent, the shear stress is reduced.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.1152-1166
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• 2023

The bubbly flow of air or steam in subcooled water are investigated here in several test cases, characterized by different pipe sizes, bubble dimensions and flow rates, by means of CFD using a Eulerian-Eulerian approach. The performance of models that differ for the turbulence closure in the continuous phase, as well as for the description of the lift force on the dispersed phase, are compared in detail. When air is considered, the space of the experimental parameters leading to a reasonable performance for the selected models are identified and discussed, while the issues left in the modelling of the concurrent condensation are highlighted for the cases where steam is used.

• 한국정밀공학회지
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• 제12권1호
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• pp.68-77
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• 1995

The two-dimensional pressure distribution, which is the most important parameter in the casting cleaning installations(CCI), was analyzed using the K-FIX computer program for two-phase flow. Modelling was done using R-Z coordinates for the initial and boundary conditions which don't have periodic influx and efflux, and also there was the electric discharge due to high pressure and temperature. The marked particles were introduced to prodict the structure and the size of main and local moving surfaces. The initial and boundary conditions were modified due to the internal structure of CCI.From the results of numerical analysis, it was shown that the maximum pressure on casting was increased with the increase of a water level. The pressure on casting in the radial direction was higher than that in axial direction. Also, it was proved that by introducing the marked particles it was possible to predict the surface structure in case of two-phase flow.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3166-3175
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• 2022

The most numerical investigations of the thermal-hydraulic phenomena following the loss of the residual heat removal capability during the mid-loop operation of the pressurized water reactor were performed according to simplifications and are not sufficiently accurate. To perform more accurate and more reliable predictions of thermal-hydraulic accidents in a nuclear power plant using computational fluid dynamics codes, a more detailed methodology is needed. Modelling results identified that thermal stratification and natural convection are observed. Temperatures of lower monitoring points remain low, while temperatures of upper monitoring points increase over time. The water in the heated region, in the upper unheated region and the pipe region was well mixed due to natural convection, meanwhile, there is no natural convection in the lower unheated region. Water temperature in the pipe region increased after a certain time delay due to circulation of flow induced by natural convection in the heated and upper unheated regions. The modelling results correspond to the experimental data. The developed computational fluid dynamics methodology could be applied for modelling of two-component single/two-phase natural convection and thermal stratification phenomena during the mid-loop operation of the pressurized water reactor or other nuclear and non-nuclear installations at similar conditions.

• 대한기계학회논문집
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• 제11권1호
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• pp.1-18
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• 1987

본 연구에서는 입자가 부상된 이상난류제트유동에 Einstein의 확산모형, Pes- kin모형, 3-방정식 모형, 4-방정식 모형, 대수응력모형 등을 적용하여 해석하고 각 모 형들의 결과를 비교 분석하였다. 이상난류유동의 수치해석에서 공기는 제1유체유동 으로 하고 첨가되는 고체분말의 흐름은 밀도(.rho.$_{p}$), 층류동점성계수(.nu.$_{p}$), 과점성계수(.nu.$_{pt}$ )를 갖는 제2유체유동의 흐름으로 간주하였다.

• 대한설비공학회지:설비저널
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• 제17권3호
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• pp.238-248
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• 1988

The flow pattern inside the power condenser is generally known to be very complicated due to the phase change and turbulence effects as well as the effect of condenser geometry. In the present study, the flow pattern inside the power condenser was numerically simulated with a personal computer. The widely known CHAMPION 2/E/FIX(Concentration, Heat and Momentum Program Instruction Outfit, 2D/Elliptic/Fixed grid) computer code was modified for this purpose. The flow was asssumed to be two-dimensional and steady-state, and the tube bank was considered to be homogeneous porous medium. Simple turbulent diffusion coefficients based on the appropriate experiments were obtained for the computation. Through this analytical approach, the flow pattern could be predicted fairly well. The computational results also show that the location of the air vent plays an important key role in determining the efficiency of the condenser.

• 대한기계학회논문집
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• 제12권1호
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• pp.1-7
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• 1988

본 연구에서는 컬럼모델을 보울트로 고정할 때 접합면에 알루미늄판, 황동판, 스테인리스판 등을 삽입하고 정적강성과 동적특성을 검토하여 이것을 기초로 공작물- 주축대-공구로 형성되고 있는 사이클중에서 선반의 주축대와 베드를 연결하는 결합부에 모델실험을 사용한 게재물을 삽입하고 선반구축계의 동적특성을 검토하였다.