• Journal of the Korea Institute of Military Science and Technology
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• v.6 no.1
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• pp.30-38
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• 2003

The measurement on the working gas flow in the closed cycle diesel engine is insufficient to Identify its operating condition because of lack of sensors available. For this reason, the determination of optimum operating condition and controller parameters requires heavy experimental efforts. A simulation program is developed to minimize the experimental efforts. For the development of the simulation program, mathematical models are derived based on the physical laws and linear regression. The validity of the simulation program is demonstrated using the measured data of the closed cycle diesel engine.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1842-1852
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• 2019

The AP1000 reactor coolant pump is a vertical shielded-mixed flow pump, is the most important coolant power supply and energy exchange equipment in nuclear reactor primary circuit system, whose steadystate and transient performance affect the safety of the whole nuclear island. Moreover, safety demonstration of reactor coolant pump is the most important step to judge whether it can be practiced, among which software simulation is the first step of theoretical verification. This paper mainly introduces the fluid-solid coupling simulation method applied to reactor coolant pump, studying the feasibility of simulation results based on workbench fluid-solid coupling technology. The study found that: for the unsteady calculations of the pure liquid media, the average head of the reactor coolant pump with bidirectional fluid-solid coupling decreases to a certain extent. And the coupling result is closer to the real experimental value. The large stress and deformation of rotor under different flow conditions are mainly distributed on impeller and idler, and the stress concentration mainly occurs at the junction of front cover plate and blade outlet. Among the factors that affect the dynamic stress change of rotor, the pressure load takes a dominant position.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.34 no.11
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• pp.991-998
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• 2010

The effect of increasing the elevation of an IHX (intermediate heat exchanger) on the transient performance of the KALIMER-600 reactor pool during the early phase of a loss of normal heat sink accident was investigated. Three reactors equipped with IHXs that were elevated to different heights were designed, and the thermal-hydraulic analyses were carried out for the steady and transient state by using the COMMIX-1AR/P code. In order to analyze the effects of the elevation of an IHX between reactors, various thermal-hydraulic properties such as mass flow rate, core peak temperature, RmfQ (ratio of mass flow over Q) and initiation time of decay heat removal via DHX (decay heat exchanger) were evaluated. It was found that with an increase in the IHX elevation, the circulation flow rate increases and a steep rise in the core peak temperature under the same coastdown flow condition is prevented without a delay in the initiation of the second stage of cooling. The available coastdown flow range in the reactor could be increased by increasing the elevation of the IHX.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.8
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• pp.535-542
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• 2016

In this study, a model reference adaptive controller is developed to regulate the outlet air flow rate of centrifugal compressor for automotive supercharger. The centrifugal compressor is developed using the analytical based method to predict the transient behavior of operating and the designed model is validated with experimental data to confirm the system accuracy. The model reference adaptive control structure consists of a compressor model and a MRAC(model reference adaptive control) mechanism. The feedback control do not robust with variation of system parameter but the applied adaptive control is robust even if the system parameter is changed. As a result, the MRAC was regulated to reference air flow rate. Also MRAC was found to be more robust control compared with the feedback control even if the system parameter is changed.

• 한국전산유체공학회:학술대회논문집
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• 2000.05a
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• pp.39-44
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• 2000

Two-dimensional Navier-Stokes codes are used to simulate the shock tunnel in Konkuk university. In order to design experiments in impulse facilities properly and to interpret data from such facilities, it is necessary to understand how the flow approaches steady state. This is done by determining the transient flow field and flow establishment time around a given model. This will be accomplished by developing appropriate CFD codes which solve the Navier-Stokes equations, and simulating the starting process and resulting unsteady viscous flow phenomena. The starting process in a shock tunnel consists of multiple shock interactions and contact discontinuities, which are difficult to solve with the classical shock capturing schemes. A recently developed high resolution scheme is adapted for resolving the unsteady phenomena of those multiple shock interactions and contact surfaces during the starting process. The bifurcation phenomenon due to the interactions of the reflected shock from the end of the shock tube with the boundary layer generated by the incident shock becomes of particular interest. By comparing with the experiment results, the accuracy of the numerical analysis is validated and it is demonstrated that the properties which can hardly be obtained through the experiment can be estimated.

• Advances in Energy Research
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• v.5 no.3
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• pp.255-275
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• 2017

The advantages of using natural circulation (NC) as a cooling system, has prompted the worldwide development to investigate this phenomenon more than before. The interesting application of the NC in low power experimental facilities and research reactors, highlights the obligation of study in these laminar flows. The inherent oscillations of NC between hot source and cold sink in low Grashof numbers necessitates stability analysis of cooling flow with experimental or numerical schemes. For this type of analysis, numerical methods could be implemented to desired mass, momentum and energy equations as an efficient instrument for predicting the behavior of the flow field. In this work, using the explicit, implicit and Crank-Nicolson methods, the fluid flow parameters in a natural circulation experimental test loop are obtained and the sensitivity of solving approaches are discussed. In this way, at first, the steady state and transient results from explicit are obtained and compared with experimental data. The implicit and crank-Nicolson scheme is investigated in next steps and in subsequent this research is focused on the numerical aspects of instability prediction for these schemes. In the following, the assessment of the flow behavior with coarse and fine mesh sizes and time-steps has been reported and the numerical schemes convergence are compared. For more detail research, the natural circulation of fluid was modeled by ANSYS-CFX software and results for the experimental loop are shown. Finally, the stability map for rectangular closed loop was obtained with employing the Nyquist criterion.

• 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.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2757-2772
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• 2023

The fluid retention effect of the Volume Control Tank (VCT) leads to a long time delay in Reactor Coolant System (RCS) concentration during the boration process. A coupled model combining a lumped-parameter sub-model and a computational fluid dynamics sub-model is currently used to investigate the concentration dynamic characteristic of RCS during the boration process. This model is validated by comparison with experimental data, and the predicted results show excellent agreement with experimental data. We provide detailed fields in VCT and concentration variations of RCS to study the interaction between mixing in VCT and the transient responses of RCS. Moreover, the impacts of the inlet flow rate, inlet nozzle diameter, original concentration, and replenishing temperature of VCT on the RCS concentration characteristic are studied. The inlet flow rate and nozzle diameter of VCT remarkably affect the RCS concentration characteristic. Too-large or too-small inlet flow rates and nozzle diameters will lead to unacceptable long delays. In this work, the optimal inlet flow rate and nozzle diameter of VCT are 5 m³/h and 58.8 mm, respectively. Besides, the impacts of the original concentration and replenishing temperature of VCT are negligible under normal operating conditions.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2743-2759
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• 2020

A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the T_hot and T_cold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1905-1915
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• 2019

This paper investigated the influence of simple support girds on flow, irrespective of having mixing vanes, in a 1 × 3 array rod bundle by using CFD methodology and the most accurate turbulence model which could reflect the actual physics of the flow was determined. In this context, a CFD model was created simulating the experimental studies on a single-phase flow [1] and the results were compared with the experimental data. In the first part of the study, influence of mesh was examined. Tetra, hybrid and poly type meshes were analyzed and convergence study was carried out on each in order to determine the most appropriate type and density. k - ε Standard and RSM LPS turbulence models were used in this section. In the second part of the study, the most appropriate turbulence model that could reflect the physics of the actual flow was investigated. RANS based turbulence models were examined using the mesh that was determined in the first part. Velocity and turbulence intensity results obtained on the upstream and downstream of the spacer grid at -3d_h, +3d_h and +40d_h locations were compared with the experimental data. In the last section of the study, the behavior of flow through the spacer grid was examined and its prominent aspects were highlighted on the most appropriate turbulence model determined in the second part. Results of the study revealed the importance of mesh type. Hybrid mesh having the largest number of structured elements performed remarkably better than the other two on results. While comparisons of numerical and experimental results showed an overall agreement within all turbulence models, RSM LPS presented better results than the others. Lastly, physical appearance of the flow through spacer grids revealed that springs has more influence on flow than dimples and induces transient flow behaviors. As a result, flow through a simple support grid was examined and the most appropriate turbulence model reflecting the actual physics of the flow was determined.