• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.1178-1183
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• 2004

Temperature distribution and heating characteristic of the panel heater for infrared heating have been investigated. The temperature variation with time is firstly measured with the thermocouple to figure out the response time of the heater to the power input. The heater reaches faster to the steady state in comparison to the ceramic heater. The infrared thermal imaging system is utilized to investigate the temperature distribution over the heater surface. The measured thermal images show that the thermal boundary layer induced by the free convection near the heater surface affects the temperature distribution on the surface. The images also show the fairly good uniformity of the temperature distribution in the core region of the surface.

• 한국조명전기설비학회:학술대회논문집
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• 한국조명전기설비학회 2009년도 추계학술대회 논문집
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• pp.423-427
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• 2009

전력계통이 초고압, 대규모화되고 이를 구성하고 있는 대용량 전력설비들의 고신뢰도 운전이 한층 요망되고 있는 환경에서 예방진단기술을 활용한 설비운영의 중요성은 나날이 커지고 있다. GIS 등 설비의 최적화 및 효율적인 관리를 위한 예방진단기술의 정착은 현재 유지보수 비용의 절감과 설비의 안정성 향상 및 체계적인 운영을 가능하게 한다는 점에서 전력설비의 성공적인 운영에 없어서는 안 될 핵심기술로서 자리매김하고 있으며, 설비의 잔존수명예측 등 향후 예방진단기술의 발전을 위한 지속적인 연구와 관심이 뒷받침되어져야 할 것이다.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.751-757
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• 2016

The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide ($UO_2$) and uranium zirconium hydride ($UZrH_{1.6}$) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with $UO_2$ contains $8{\times}8$ fuel rods while that fueled with $UZrH_{1.6}$ contains $9{\times}9$ fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. $UZrH_{1.6}$ fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2008년도 제39회 하계학술대회
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• pp.257-258
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• 2008

The inrush current of a transformer is a high-magnitude and harmonic-rich current generated when the transformer core is driven into saturation during energizing. The inrush current usually leads to undesirable effects, for example potential damage to the transformer, misoperation of a protective relay, and power quality deterioration in the distribution power system. Inrush current reduction is therefore important for power system operation. In this paper, to reduce the inrush current, the insertion resistance of the Superconducting Fault Current Limiter (SFCL) that is connected in series with the transformer in the distribution system is used. This paper implements the SFCL by using the Electromagnetic Transient Program-Restructured Version (EMTP-RV) to model the SFCL in the distribution system. The simulation results show the beneficial effects of the SFCL for reduction of the inrush current.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.43-54
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• 2016

In the present paper, development of the three-dimensional (3D) computational code based on Galerkin finite element method (GFEM) for solving the multigroup forward/adjoint diffusion equation in both rectangular and hexagonal geometries is reported. Linear approximation of shape functions in the GFEM with unstructured tetrahedron elements is used in the calculation. Both criticality and fixed source calculations may be performed using the developed GFEM-3D computational code. An acceptable level of accuracy at a low computational cost is the main advantage of applying the unstructured tetrahedron elements. The unstructured tetrahedron elements generated with Gambit software are used in the GFEM-3D computational code through a developed interface. The forward/adjoint multiplication factor, forward/adjoint flux distribution, and power distribution in the reactor core are calculated using the power iteration method. Criticality calculations are benchmarked against the valid solution of the neutron diffusion equation for International Atomic Energy Agency (IAEA)-3D and Water-Water Energetic Reactor (VVER)-1000 reactor cores. In addition, validation of the calculations against the $P_1$ approximation of the transport theory is investigated in relation to the liquid metal fast breeder reactor benchmark problem. The neutron fixed source calculations are benchmarked through a comparison with the results obtained from similar computational codes. Finally, an analysis of the sensitivity of calculations to the number of elements is performed.

• Nuclear Engineering and Technology
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• 제52권12호
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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.

• Journal of Electrical Engineering and Technology
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• 제10권1호
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• pp.176-187
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• 2015

DC bias will cause abnormal vibration of transformers. Aiming at such a problem, transformer vibration affected by DC bias has been studied combined with transformer core and winding vibration mechanism use multi-physical field simulation software COMSOL in this paper. Furthermore the coupling model of electromagnetic-structural force field has been established, and the variation pattern of inner flux density, distribution of mechanical stress, tension and displacement were analyzed based on the coupling model. Finally, an experiment platform has been built up which was employed to verify the correctness of model.

• 한국통신학회논문지
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• 제21권12호
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• pp.3277-3285
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• 1996

In this paper, w designed an opical power distribution device for application to an optical switching and an optical subscriber loop. We fabricated PSG thin film by LPCVD. Based on the measured index of fabricted thin film, rib-type waveguide was transformed to two-dimension by the effective index method and we simulated dispersion property to find asingle-mode condition. We found that the optimum design parameters of rib-type waveguide are:cladding layer of 3.mu.m, core layer of 3.mu.m, buffer layer of 10.mu.m, and core width of 4.mu.m. Each side of the guiding region was etched down to 4.mu.m to shape the core. We used these optimum parameters of the rib-type waveguide with branching angle of 0.5.deg. and simulted the Y-branch waveguide by the BPM simulation. Numerical loss in branching area was claculated to be 0.1581dB and equal to the total loss of the Y-branch. The loss of the fabricated Y-branch waveguide on PSG film ws 1.6dB at .lambda.=1.3.mu.m before annealing but was 1.2dB after annealing at 1000.deg. C for 10 minutes. Consequently, the loss of branching area from 3000.mu.m to 6000.mu.m in the z-direction was 0.8dB, and single-mode propagation was confirmed by measuring the near field pattern. For coupling the fabricated Y-branch waveguide with an optical fiber, we fabricated V-groove which was used as the upholder of optical fiber. An etching angle was 54.deg. and the width and depth of guiding groove was 150.mu.m, 70.mu.m, respectively. The optical fiber is inserted onto V-groove. Both the Y-branch and V-groove were connected through the index matching oil. Coupling loss after connecting Y-branch and the optical fiber on V-groove was 0.34dB and that after injecting index mateching oil was 0.14dB.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.351-359
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• 2016

The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

• 한국자기학회지
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• 제22권3호
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• pp.97-102
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• 2012

본 논문은 전선을 통하여 전력을 전달하는 일반적인 전력전송방식이 사용될 수 없는 환경에서 무선으로 전력을 전송하는 비접촉 전력전송기기 개발에 관한 것이다. 전력전송 방식은 자기공명방식보다 큰 전력을 전송하기에 적합한 전자기유도방식을 이용 하였다. 전력전송기기의 설계 방법은 전기장하(Electirc loading)와 자기장하(Magnetic loading)의 비율로 코어와 코일을 설계하는 장하분배법(Loading Distribution Method)으로 설계 하였고 유한요소법(Finite Elements Method)으로 기기에 발생하는 전자기장을 해석하여 설계한 전력전송기기의 적합성을 판단하고 적정한 설계치를 보정하였다. 본 연구를 통하여 개발된 전력전송방식은 비접촉식으로 수 mm 거리를 가지는 근거리에서 무선으로 전력을 전송하기에 적합함을 보였다.