• The Transactions of the Korean Institute of Electrical Engineers D
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• v.55 no.3
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• pp.131-137
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• 2006

The secondary system of nuclear power plants consists of sophisticated piping systems operating in very aggressive erosion and corrosion environments, which make a piping system vulnerable to the wear and degradation due to the several chemical components and high flow rate (~10 m/sec) of the coolant. To monitor the wear and degradation on a pipe, the vibration signals are measured from the pipe with an accelerometer For analyzing the vibration signal the time-frequency analysis (TFA) is used, which is known to be effective for the analysis of time-varying or transient signals. To reduce the inteferences (cross-terms) due to the bilinear structure of the time-frequency distribution, an adaptive cone-kernel distribution (ACKD) is proposed. The cone length of ACKD to determine the characteristics of distribution is optimally selected through an adaptive algorithm using the normalized Shannon's entropy And the ACKD's are compared with the results of other analyses based on the Fourier Transform (FT) and other TFA's. The ACKD shows a better signature for the wear/degradation within a pipe and provides the additional information in relation to the time that any analysis based on the conventional FT can not provide.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.162-176
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• 2022

In Lead-based reactor (LBR) severe accident, the meltdown and migration inside the reactor core will lead to fuel fragment concentration, which may further cause re-criticality and even core disintegration. Accurately predicting the migration and solidification behavior of melt in LBR severe accidents is of prime importance for safety analysis of LBR. In this study, the Moving Particle Semi-implicit (MPS) method is validated and used to simulate the migration and solidification behavior. Two main surface tension models are validated and compared. Meanwhile, the MPS method is validated by the L-plate solidification test. Based on the improved MPS method, the migration and solidification behavior of melt in LBR severe accident was studied furthermore. In the Pb-Bi coolant, the melt flows upward due to density difference. The migration and solidification behavior are greatly affected by the surface tension and viscous resistance varying with enthalpy. The whole movement process can be divided into three stages depending on the change in velocity. The heat transfer of core melt is determined jointly by two heat transfer modes: flow heat transfer and solid conductivity. Generally, the research results indicate that the MPS method has unique advantage in studying the migration and solidification behavior in LBR severe accident.

• Journal of the Korean Society of Safety
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• v.20 no.2 s.70
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• pp.32-37
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• 2005

The purpose of this study is for the sensitivity study f9r a Steam Generator Tube Rupture (SGTR) of the System-integrated Modular Advanced ReacTor for a Pilot (SMART-P) plant. The thermal hydraulic analysis of a SGIR for the Limiting Conditions for Operation (LCO) is performed using TASS/SMR code. The TASS/SMR code can calculate the core power, pressure, flow, temperature and other values of the primary and secondary system for the various initiating conditions. The major concern of this sensitivity study is not the minimum Critical Heat Flux Ratio(CHFR) but the maximum leakage amount from the primary to secondary sides at the steam generator. Therefore the break area causing the maximum accumulated break flow is researched for this reason. In the case of a SGIR for the SMART-p, the total integrated break flow is 11,740kg in the worst case scenario, the minimum CHFR is maintained at Over 1.3 and the hottest fuel rod temperature is below 606"I during the transient. It means that the integrity of the fuel rod is guaranteed. The reactor coolant system and the secondary system pressures are maintained below 18.7MPa, which is system design pressure.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1514-1524
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• 2019

Hydrogen-steam gas mixture may be injected into containment with flow regime varying both spatially and transiently due to wall effect and pressure difference between primary loop and containment in severe accidents induced by loss of coolant accident. Preliminary CFD analysis is conducted to gain information about the helium flow regime transition process from jet to buoyancy plume for forthcoming experimental study. Physical models of impinging jet and wall condensation are validated using separated effect experimental data, firstly. Then helium transportation is analyzed with the effect of jet momentum, buoyancy and wall cooling discussed. Result shows that helium distribution is totally dominated by impinging jet in the beginning, high concentration appears near gas source and wall where jet momentum is strong. With the jet weakening, stable light gas layer without recirculating eddy is established by buoyancy. Transient reversed helium distribution appears due to natural convection resulted from wall cooling, which delays the stratification. It is necessary to concern about hydrogen accumulation in lower space under the containment external cooling strategy. From the perspective of experiment design, measurement point should be set at the height of connecting pipe and near the wall for stratification stability criterion and impinging jet modelling validation.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2012.04a
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• pp.734-738
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• 2012

A comprehensive vibration assessment program for the Advanced Power Reactor 1400 reactor vessel internals is established in accordance with the United States Nuclear Regulatory Commission Regulatory Guide 1.20 Revision 3. This paper is related to instruments and measurement locations based on the vibration and stress response analysis results at the inner barrel assembly top plate in the reactor. The analysis results of the inner barrel assembly top plate in the reactor show that the deterministic stress and deformation due to the reactor coolant pump induced pressure pulsations are larger than the random stress and deformation induced by the flow turbulence. The selection of the instruments and measurement locations at Inner barrel assembly top plate in the reactor is essential requirements and very important study process for the vibration and stress measurement program in comprehensive vibration assessment program for the Advanced Power Reactor 1400 reactor vessel internals.

• Journal of the Korean Society of Propulsion Engineers
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• v.20 no.2
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• pp.1-10
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• 2016

Cooling characteristics are investigated numerically in the chamber for high-performance burnout of wastes with solid phase. Before the combustion chamber is manufactured, combustion analysis is performed for evaluation of burning rate and cooling performance. A water cooling method is applied and its feasibility for cooling is examined depending on coolant flow rate. Another method of complex cooling is adopted by combining air film cooling with water cooling, leading to improved cooling performance.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2002.05a
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• pp.705-709
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• 2002

The spacer grids in nuclear fuel assembly locate and align the fuel rods with respect to each other. They provide axial and lateral restraint against an excessive rod motion mainly caused by coolant flow. It is understood that each rod Is supported by multiple spacer grid. In such a case, it is important to determine spacer grid span so as to avoid resonance between the natural frequency of the fuel rods and excitation frequency. Actually dynamic characteristics of the fuel rods can be improved by assigning adequate spacer grid locations. When a dynamic performance of the structure is to be improved, design sensitivity analysis plays an important role as like many structural redesign problems. In this work, a shape design concept, different from conventional design, was applied to the problem. According to the theory shape can be a design parameter and optimal shape design can be found. This study concentrates on eigenvalue design sensitivity of the fuel rod supported by multiple spacer grids to determine optimal spacer grids positions.

• Journal of the korean Society of Automotive Engineers
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• v.2 no.1
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• pp.39-50
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• 1980

This paper presents the variations obtained in heat flow rate and engine performance of a four-stroke cycle Diesel engine when there were changes in the temperature of cooling water, compression ratio, injection timing of fuel, and other factors. Heat dissipation of engine cylinder was calculated by the heat transfer coefficient of Nusselt's empirical equation and the analysis of distribution of temperature in cylinder barrel was obtained by the finite element method of two-dimensional steady state heat conduction. In this experiment, the out side temperature of cylinder liner was measured by the data logger, and the temperature distribution of liner was computed by the analysis of triangular finite element model under the assumption due to surface heat flux of cylinder inner surface. The results obtained by this study are as follows. Under the given operating condition, the temperature distribution of cylinder liner by using finite element method shows that the mean temperature of barrel is in accordance with the experimental results of Eichelberg and temperature difference is lower than 4.23.deg. C. The heat dissipation of engine decrease in accordance with the decrease of piston mean velocity, compression ratio, and the increase of coolant temperature. Influence on the delay of injection timing of fuel brings about the decrease of heat rejection over the cylinder at constant test conditions.

• Journal of Electrical Engineering and Technology
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• v.9 no.3
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• pp.893-898
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• 2014

Recently, the interior permanent synchronous motor (IPMSM) has been applied to an integrated starter and generator (ISG) for hybrid electric vehicles. In the design of such a motor, thermal analysis is necessary to maximize the power density because the loss is proportional to the power of a motor. Therefore, a cooling device as a heat sink is required internally. Generally, a cooling system designed with a water jacket structure is widely used for electric motors because it has advantages of simple structure and cooling effectiveness. An effective approach to analyze an electric machine with a water jacket is a thermal equivalent network. This network is composed of thermal resistance, a heat source, and thermal capacitance that consider the conduction, convection, and radiation. In particular, modeling of the cooling channel in a network is challenging owing to the flow of the coolant. In this paper, temperature prediction using a thermal equivalent network is performed in an ISG that has a water cooled system. Then, an experiment is conducted to verify the thermal equivalent network.

• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.439-458
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• 2013

REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System) is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS) method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility). Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.