• 한국전산유체공학회:학술대회논문집
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• 2007.10a
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• pp.194-199
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• 2007

IB (immersed boundary) method is one of the prominent tool in computational fluid dynamics for the analysis of flows over complex geometries. The IB technique simplyfies the solution procedure by eliminating the requirement of complex body fitted grids and it is also superior in terms of memory requirement. In this study we have developed numerical code (FOTRAN) for the analysis of 2D flow over a cylinder using IB technique. The code is validated by comparing the wake lengths and separation angles given by Guo et. al. We employed fractional-step procedure for solving the Navier-Stokes equations governing the flow and discrete forcing IB technique for imposing boundary conditions. Also we have developed a 3D code for the backward-facing-step flow and flow over a sphere. The reattachment length in backward-facing-step flow was compared with the one given by Nie and Armaly, which has proven the validity of our code.

• The KSFM Journal of Fluid Machinery
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• v.6 no.2 s.19
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• pp.41-46
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• 2003

The hydraulic design optimization and performance analysis of mixed-flow pumps for waterjet marine vehicle propulsion has been carried out using mean streamline analysis and three-dimensional computational fluid dynamics (CFD) code. In the present study, the conceptual design optimization has been formulated with a non-linear objective function to minimize the fluid dynamic losses, and then the commercial CFD code was incorporated to allow for detailed flow dynamic phenomena in the pump system. Newly designed mixed-flow model pump has been tested in the laboratory. Predicted performance curves by the CFD code agree very well with experimental data for a newly designed mixed-flow pump over the normal operating conditions. The design and prediction method presented herein can be used efficiently as a unified hydraulic design process of mired-flow pumps for waterjet marine vehicle propulsion.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.665-674
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• 2013

This paper presents the development and validation methods of the FUTURE (FUel cycle analysis Tool for nUcleaR Energy) code, which was developed for a dynamic material flow evaluation and economic analysis of the nuclear fuel cycle. This code enables an evaluation of a nuclear material flow and its economy for diverse nuclear fuel cycles based on a predictable scenario. The most notable virtue of this FUTURE code, which was developed using C# and MICROSOFT SQL DBMS, is that a program user can design a nuclear fuel cycle process easily using a standard process on the canvas screen through a drag-and-drop method. From the user's point of view, this code is very easy to use thanks to its high flexibility. In addition, the new code also enables the maintenance of data integrity by constructing a database environment of the results of the nuclear fuel cycle analyses.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.19-33
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• 2024

In nuclear thermal-hydraulic system codes, most correlations used for vertical pipes, under downward two-phase flow, have been developed considering small pipes or pool systems. This suggests that there could be uncertainties in applying the correlations to accident scenarios involving large vertical pipes owing to the difference in the characteristics of two-phase flows, or flow conditions, between large and small pipes. In this study, we modified the Multi-dimensional Analysis of Reactor Safety KINS Standard (MARS-KS) code using correlations, such as the drift-flux model and two-phase multiplier, developed in a plant-scale air-inflow experiment conducted for a pipe of diameter 600 mm under downward two-phase flow. The results were then analyzed and compared with those based on previous correlations developed for small pipes and pool conditions. The modified code indicated a good estimation performance in two plant-scale experiments with large pipes. For the siphon-breaking experiment, the maximum errors in water flow for modified and original codes were 2.2% and 30.3%, respectively. For the air-inflow accident experiment, the original code could not predict the trend of frictional pressure gradient in two-phase flow as / increased, while the modified MARS-KS code showed a good estimation performance of the gradient with maximum error of 3.5%.

• 유체기계공업학회:학술대회논문집
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• 2002.12a
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• pp.47-53
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• 2002

The hydraulic design optimization and performance analysis of mixed-flow pumps for waterjet marine vehicle propulsion has been carried out using mean streamline analysis and three-dimensional computational fluid dynamics (CFD) code. In the present study the conceptual design optimization has been formulated with a non-linear objective function to minimize the fluid dynamic losses and then the commercial CFD code was incorporated to allow for detailed flow dynamic phenomena in the pump system. New designed mixed-flow model pump has been tested in the laboratory. Predicted performance curves by the CFD code agree very well with experimental data for a newly designed mixed-flow pump over the normal operating conditions. The design and prediction methods presented herein can be used efficiently as a unified hydraulic design process of mixed-flow pumps for waterjet marine vehicle propulsion.

• Journal of computational fluids engineering
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• v.21 no.3
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• pp.89-97
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• 2016

In this study, direct code coupling, in which two codes share a single flow field, was conducted using 3-dimensional high resolution thermal hydraulics code, CUPID and 1-dimensional system analysis code, MARS. This approach provide the merit to use versatile capability of MARS for nuclear power plants and 3-dimensional T/H analysis capability of CUPID. Numerical Method to directly couple CUPID and MARS was described in this paper. The straight flow and manometer flow oscillation were calculated to verify conservation of coupled CUPID/MARS code in mass, momentum, and energy. This verification calculations indicates that the CUPID/MARS is coupled appropriately in numerical aspect and the coupled code can be applied to nuclear reactor thermal hydraulics after validation against integral transient experiments.

• 한국전산유체공학회:학술대회논문집
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• 1998.05a
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• pp.162-167
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• 1998

A computer code has been developed for the analysis of 2D viscous flow with free surface. VOF method and higher order upwind scheme have been employed for the accurate prediction of free surface motion. Surface tension effect and axisymmetric flow can be computed by the present code.

• Nuclear Engineering and Technology
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• v.46 no.5
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• pp.655-666
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• 2014

The CUPID code has been developed at KAERI for a transient, three-dimensional analysis of a two-phase flow in light water nuclear reactor components. It can provide both a component-scale and a CFD-scale simulation by using a porous media or an open media model for a two-phase flow. In this paper, recent advances in the CUPID code are presented in three sections. First, the domain decomposition parallel method implemented in the CUPID code is described with the parallel efficiency test for multiple processors. Then, the coupling of CUPID-MARS via heat structure is introduced, where CUPID has been coupled with a system-scale thermal-hydraulics code, MARS, through the heat structure. The coupled code has been applied to a multi-scale thermal-hydraulic analysis of a pool mixing test. Finally, CUPID-SG is developed for analyzing two-phase flows in PWR steam generators. Physical models and validation results of CUPID-SG are discussed.

• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.797-806
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• 2009

The Multichannel Analyzer for Transient and steady-state in Rod Array - Liquid Metal Reactor for Flow Blockage analysis (MATRA-LMR-FB) code for the analysis of a subchannel blockage has been developed and evaluated through several experiments. The current version of the code is improved here by the implementation of a distributed resistance model which accurately considers the effect of flow resistance on wire spacers, by the addition of a turbulent mixing model, and by the application of a hybrid scheme for low flow regions. Validation calculations for the MATRA-LMR-FB code were performed for Oak Ridge National Laboratory (ORNL) 19-pin tests with wire spacers and Karlsruhe 169-pin tests with grid spacers. The analysis of the ORNL 19-pin tests conducted using the code reveals that the code has sufficient predictive accuracy, within a range of 5 $^{\circ}C$, for the experimental data with a blockage. As for the results of the analyses, the standard deviation for the Karlsruhe 169-pin tests, 0.316, was larger than the standard deviation for the ORNL 19-pin tests, 0.047.

• 한국연소학회:학술대회논문집
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• 2001.06a
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• pp.90-98
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• 2001

Characteristics of high temperature rocket nozzle flow is discussed along with the aspects of computational analysis. Three methods of nozzle flow analysis, frozen-equilibrium, shifting-equilibrium and non-equilibrium approaches, were discussed, those were coupled with the methods of computational fluid dynamics code. A chemical equilibrium code developed for the analysis of general hydrocarbon fuel was coupled with three approaches of nozzle flow analysis. The approaches were used for the performance prediction of KSR-III Rocket, and compared with the theoretical results from NASA CEA (Chemical Equilibrium with Applications) code.