• 한국전산유체공학회지
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• 제12권4호
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• pp.44-53
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• 2007

A three-dimensional (3D) unstructured hydrodynamic solver for transient two-phase flows has been developed for a 3D component of a nuclear system code and a component-scale analysis tool. A two-fluid three-field model is used for the two-phase flows. The three fields represent a continuous liquid, an entrained liquid, and a vapour field. An unstructured grid is adopted for realistic simulations of the flows in a complicated geometry. The semi-implicit ICE (Implicit Continuous-fluid Eulerian) numerical scheme has been applied to the unstructured non-staggered grid. This paper presents the numerical method and the preliminary results of the calculations. The results show that the modified numerical scheme is robust and predicts the phase change and the flow transitions due to boiling and flashing very well.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2007년도 추계 학술대회논문집
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• pp.243-248
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• 2007

A three-dimensional (3D) unstructured hydrodynamic solver for transient two-phase flows has been developed. A two-fluid three-field model was adopted for the two-phase flows. The three fields represent a continuous liquid, an entrained liquid, and a vapour field. The hydrodynamic solver is for the 3D component of a nuclear system code and the component-scale analysis tools for transient two-phase flows. The finite volume method and unstructured grid are adopted, which are useful for the flows in a complicated geometry. The semi-implicit ICE (Implicit Continuous-fluid Eulerian) numerical scheme has been adapted to the unstructured non-staggered grid. This paper presents the numerical method and the preliminary results of the calculations. The results show that the numerical scheme is robust and predicts the phase change and the flow transitions due to boiling and flashing problems well.

• 한국전산유체공학회지
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• 제15권2호
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• pp.86-94
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• 2010

A thermal-hydraulic code, named CUPID, has been developed for the analysis of transient two-phase flows in nuclear reactor components. A two-fluid three-field model was used for steam-water two-phase flows. To obtain numerical solutions, the finite volume method was applied over unstructured cell-centered meshes. In steam-water two-phase flows, a phase change, i.e., evaporation or condensation, results in a great change in the flow field because of substantial density difference between liquid and vapor phases. Thus, two-phase flows are very sensitive to the local pressure distribution that determines the phase change. This in turn puts emphasis on the accurate evaluation of local pressure gradient. This paper presents a new reconstruction method to evaluate the pressure gradient at cell centers on unstructured meshes. The results of the new scheme for a simple test function, a gravity-driven cavity, and a wall boiling two-phase flow are compared with those of the previous schemes in the CUPID code.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2010년 춘계학술대회논문집
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• pp.380-388
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• 2010

A thermal-hydraulic code, named CUPID, has been developed for the analysis of transient two-phase flows in nuclear reactor components. A two-fluid three-field model was used for steam-water two-phase flows. To obtain numerical solutions, the finite volume method was applied over unstructured cell-centered meshes. In steam-water two-phase flows, a phase change, i.e., evaporation of condensation, results in a great change in the flow field because of substantial density difference between liquid and vapor phases. Thus, two-phase flows are very sensitive to the local pressure that determines the phase change. This in turn puts emphasis on the accurate evaluation of local pressure gradient. This paper presents a new numerical scheme to evaluate the pressure gradient at cell centers on unstructured meshes. The results of the new scheme for a simple test function a gravity-driven cavity, and a wall boiling two-phase flow are compared with those of the previous schemes in the cupid code.