• Journal of Mechanical Science and Technology
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• v.16 no.7
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• pp.975-985
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• 2002

This paper deals with theoretical model developed for analyzing the heat transfer of automotive cooling systems. The model has a modular structure which links various cooling system submodels. From the model, heat transfer rate of automotive cooling systems can be predicted, providing useful information at the early stages of the design and development. The aim of the study is to develop a simulation program for automotive cooling system analysis and a performance analysis program for analyzing heat exchanger. Heat release rate from combustion gas to coolant through the cylinder wall in engine cylinder was analysed by using an engine cycle simulation program. In this paper, details of each submodel are described together with the overall structure of the vehicle model.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.396-410
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• 2018

To assist operators to properly assess the current situation of the plant, accurate fault diagnosis methodology should be available and used. A reliable fault diagnosis method is beneficial for the safety of nuclear power plants. The major idea proposed in this work is integrating the merits of different fault diagnosis methodologies to offset their obvious disadvantages and enhance the accuracy and credibility of on-line fault diagnosis. This methodology uses the principle component analysis-based model and multi-flow model to diagnose fault type. To ensure the accuracy of results from the multi-flow model, a mechanical simulation model is implemented to do the quantitative calculation. More significantly, mechanism simulation is implemented to provide training data with fault signatures. Furthermore, one of the distance formulas in similarity measurement-Mahalanobis distance-is applied for on-line failure degree evaluation. The performance of this methodology was evaluated by applying it to the reactor coolant system of a pressurized water reactor. The results of simulation analysis show the effectiveness and accuracy of this methodology, leading to better confidence of it being integrated as a part of the computerized operator support system to assist operators in decision-making.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2008.04a
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• pp.641-646
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• 2008

Long and slender body with or without flexible supports under severe operating condition can be unstabilized even by the small cross flow. Turbulent flow mixer, which actually increases thermal-hydraulic performance of the nuclear fuel by boosting turbulence, disturbs the flow field around the fuel rod and affects dynamic behavior of the nuclear fuel rods. Few studies on this problem can be found in the literature because these effects depend on the specific natures of the support and the design of the system. This work shows how the dynamics of a multi-span fuel rod can be affected by the turbulent flow, which is discretely activated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was established. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.

• Nuclear Engineering and Technology
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• v.35 no.3
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• pp.251-259
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• 2003

The Taguchi method was applied to investigate the effect of design factors on the performance of the shutdown cooling heat exchanger in the SMART-P. This method provided the simulation matrix for the KDESCENT program and an efficient tool for analyzing the simulation results. Levels of the design factors were selected by the effectiveness-NTU method. From 18 runs with the KDESCENT program, it was found that the performance of the system was greatly influenced by the inlet temperature at the shell side and the mass flow rate of the reactor coolant at the tube side. After applying the Taguchi method, we identified the important design factor that should be controlled and designed carefully. This method provides an efficient way to estimate the influence of each design factor on a system performance.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.20 no.3
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• pp.215-222
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• 2010

Long and slender body, like a fuel rod, oscillating in axial flow can be unstabilized even by the small cross flow which can be activated by the flow mixer or turbulent generator. It is important to include these effects of flow disturbance in dynamic stability analysis of nuclear fuel rod. This work shows how eigen frequency of a multi-span fuel rod can be changed by the swirl flow, which is discretely generated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was calculated. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.

• Journal of the Korean Society of Manufacturing Technology Engineers
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• v.22 no.6
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• pp.908-912
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• 2013

Because of their longer operating times and larger size relative to conventional fans, the cooling fans mounted in buses consume larger amounts of energy. Most of the cooling fans mounted in a bus are connected to the engine by a viscous clutch. A viscous cooling fan's speed is determined by its fluid temperature, which is affected by the air flow through the radiator. The fan does not react immediately to the coolant temperature and in doing so causes unnecessary energy consumption. Therefore, the fuel economy of buses using viscous fans can be improved by changing to an electric cooling fan design, which can be actively controlled. In addition, electric power consumption is increased by using electric cooling fans. Thus, when electric fans are applied in conjunction with the alternator management system (AMS), the fuel economy is further enhanced. In this study, simulations were performed to predict coolant temperature and cooling fan speeds. Simulations were performed for both viscous and electric cooling fans, and power consumption was calculated. Additionally, fuel economy was calculated applying both the alternator management system and the electric cooling fan.

• Nuclear Engineering and Technology
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• v.19 no.3
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• pp.198-204
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• 1987

A plant simulation code, MCSIM (Micro-Computer SIMulator), has been developed to simulate plant transient accidents for pressurized water reactors. Reactor coolant system is modeled using decoupled energy and momentum equations, drift flux two-phase flow model and integral momentum equation. A two-fluid pressurizer model is used to simulate the pressurizer dynamics. Pot Boiler model is used for steam generator, steady-state decoupled energy and momentum equations for secondary side system, and point kinetics equations for nuclear power calculation. For test of the present version of MCSIM, complete loss of flow and RCCA withdrawal accidents are calculated with MCSIM. The results are compared with those in FSAR of KNU 5 & 6.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.953-967
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• 2017

In this paper, a complete station blackout (SBO) or complete loss of electrical power supplies is simulated and analyzed in a downward cooling 5-MW pool-type Material Testing Reactor (MTR). The scenario is traced in the absence of active cooling systems and operators. The code nodalization is successfully benchmarked against experimental data of the reactor's operating parameters. The passive heat removal system includes downward water cooling after pump breakdown by the force of gravity (where the coolant streams down to the unfilled portion of the holdup tank), safety flapper opening, flow reversal from a downward to an upward cooling direction, and then the upward free convection heat removal throughout the flapper safety valve, lower plenum, and fuel assemblies. Both short-term and long-term natural core cooling conditions are simulated and investigated using the RELAP5 code. Short-term analyses focus on the safety flapper valve operation and flow reversal mode. Long-term analyses include simulation of both complete SBO and long-term operation of the free convection mode. Results are promising for pool-type MTRs because this allows operators to investigate RELAP code abilities for MTR thermal-hydraulic simulations without any oscillation; moreover, the Tehran Research Reactor is conservatively safe against the complete SBO and long-term free convection operation.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3990-4002
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• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.72-83
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• 2022

This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.