• Tribology and Lubricants
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• 제39권4호
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• pp.123-132
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• 2023

In this study, the expansion, stress, and operating clearance of bearing elements during operation are observed using the inner/outer ring temperature test data of a 3.0×10⁶ DN-class roller bearing. The operating clearance characteristics of inner-/outer-ring cooling (IORC) bearings are compared to those of inner-ring cooling (IRC) bearings. For IRC bearings, the thermal expansion of the outer ring is the most important factor in clearance variation. As a result, the operating clearance is less than the initial clearance of 61 ㎛, and the operating clearance decreases to 0.5 ㎛ at 25,500 rpm. Conversely, the temperature of the outer ring of IORC bearings is lower than that of IRC bearings, so the operating clearance is kept smaller. When the coolant flow rate to the outer ring is approximately 1.5 to 2.0 L/min, the temperature difference between the inner and outer rings is minimized and the operating clearance is maintained at a significantly lower level than IRC bearings. Small operating clearances are expected to be effective in reducing cage slip and skid damage in roller bearings. The results and analysis procedures of this study can be utilized to design of bearing clearance, lubricant flow rate, and assembled interference in the early design stage of aerospace roller bearings.

• 전기학회논문지
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• 제63권3호
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• pp.431-436
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• 2014

In this paper, temperature change properties on the 210kW-class Interior Permanent Magnet Synchronous Motor (IPMSM) are performed with the cooling performance of a water cooling device through the thermal characteristic analysis of the IPMSM considering a real driving pattern of urban railway vehicles. First, the thermal analysis modeling of 210kW-class IPMSM, which is an alternative to the conventional induction motor, and its water cooling device is conducted. Next, the thermal characteristic analysis of the IPMSM considering a real driving pattern of urban railway vehicles is performed using 2-Dimensional FEM tool. Finally, the calculated characteristic results are analyzed. Consequently, it is confirmed that the internal temperature of the 210kW-class IPMSM may be lowered to about 42~52% by maintaining the coolant flow rate of the water cooling device (Cross sectional shape of the pipe has 220mm width and 10mm height) for 0.2kg/s level.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.402-409
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• 2019

When a loss of coolant accident (LOCA) occurs in a nuclear power plant, accident scenarios which can prevent core damage are defined based on break size. Current probabilistic safety assessment evaluates that core damage can be prevented under small-break LOCA (SBLOCA) and steam generator tube rupture (SGTR) with rapid cool down (RCD) strategy when all safety injection systems are unavailable. However, previous research has pointed out a limitation of RCD in terms of initiation time. Therefore, RCD success criteria estimation based on allowable coping time under a SBLOCA or SGTR when all safety injection systems are unavailable was performed based on time-line and thermal-hydraulic analyses. The time line analysis assumed a single emergency operating procedure flow, and the thermal hydraulic analysis utilized MARS-KS code with variables of break size, cooling rate, and operator allowable time. Results show while RCD is possible under SGTR, it is impossible under SBLOCA at the APR1400's current cooling rate limitation of 55 K/hr. A success criteria map for RCD under SBLOCA is suggested without cooling rate limitation.

• Nuclear Engineering and Technology
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• 제50권3호
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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.

• 대한기계학회논문집B
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• 제40권6호
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• pp.403-408
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• 2016

리튬 이온 배터리는 높은 에너지 밀도와 안정적인 충전/방전 특성을 내재하고 있어 하이드리드 및 전기자동차에 보편적으로 사용된다. 리튬 이온 배터리의 효율은 배터리 자체의 온도 특성에 직접적인 영향을 받으므로, 열을 효율적으로 냉각하는 기술이 요구된다. 본 논문에서는 수냉식 배터리 냉각 시스템의 냉각 성능과 펌프 소모동력에 관한 전산유체해석을 수행하였다. 이를 위해 배터리 셀의 냉각수 유량 및 냉각 채널의 특성에 따른 냉각 성능을 수치적으로 예측하였다. 이를 바탕으로 250개 배터리 셀을 기준으로 유량 및 차압에 의한 소모동력을 계산하였다. 이러한 연구는 차세대 하이브리드 및 전기자동차의 시간에 따른 배터리의 온도 변화 및 충/방전 효율 최적화 기술에 적용할 수 있는 기초 연구로 활용될 수 있을 것으로 기대된다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.915-920
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• 2001

A spent fuel problem has prevented the nuclear power from claiming to be a completely clean energy source. The nuclear transmutation technology to incinerate the long lived radioactive nuclides and produce energy during the incineration process is believed to be one or the best solutions. HYPER(Hybrid Power Extraction Reactor) is the accelerator driven transmutation system which is being developed by KAERI(Korea Atomic Energy Research Institute). Some major feature of HYPER have been developed and employed. On-power fueling concepts are employed to keep system power constant with minimum variation of accelerator power. A hollow cylinder-type metal fuel is designed for the on-line refueling concept. Lead-bismuth(Pb-Bi) is adopted as a coolant and Spallation target material. HYPER is a subcritical reactor which needs an external neutron source. 1GeV proton beam is irradiated to Lead-bismuth(Pb-Bi) target inside HYPER, and spallation neutrons are produced. When proton beams are irradiated, much heat is also deposited in the Pb-Bi target and beam window which separates Pb-Bi and accelerator vacuum. Therfore, an effective cooling is needed for HYPER target. In this paper, we performed the thermal-hydraulic analysis of HYPER target using FLUENT code, and also calculated thermal and mechanical stress of the beam window using ANSYS code.

• 한국안전학회지
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• 제30권5호
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• pp.131-137
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• 2015

For the evaluation of safety margin of a nuclear power plant using a conservative methodology, the influence of applied assumptions such as initial conditions and boundary conditions needs to be assessed deliberately. Usually, a combination of the most conservative initial conditions is determined, and the safety margin for the transient is evaluated through the analysis for this conservative conditions. In existing conservative methodologies, a most-conservative condition is searched through the analyses for the maximum, minimum, and nominal values of the major parameters. In the present study, we investigates a new approach which can be applied to choose a most-conservative initial condition effectively when a best-estimate computer code and a conservative evaluation methodology are utilized for the evaluation of safety margin of transients. By constituting the band of various initial conditions using the random sampling of input parameters, the sensitivity study for various parameters are performed systematically. A method of sampling the value of control or operation parameters for a certain range is adopted by use of MOSAIQUE program, which enables to minimize the efforts for achieving the steady-state for various different conditions. A representative control parameter is identified, which governs the reactor coolant flow rate, pressurizer pressure, pressurizer level, and steam generator level, respectively. It is shown that an appropriate distribution of input parameter is obtained by adjusting the range and distribution of the control parameter.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2001년도 하계학술대회 논문집 B
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• pp.714-716
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• 2001

This paper is about the characteristic analysis of a heater-trigger HTS switch. It is composed of a BSCCO tape wire and a heater as a thermal quench source. To minimize the heat flow to the liquid nitrigen, the epoxy was used around them. The heater-trigger switch operates in the process that the HTS tape is thermally quenched by the current flowing through the heater. Since the switch works in coolant, it needs to be designed by the accurate thermal characteristic analysis. Here, we focus on FEM for it.

• Nuclear Engineering and Technology
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• 제33권5호
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• pp.547-555
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• 2001

In pressurized light water reactor fuel assembly, spacer grids support nuclear fuel rods both laterally and vertically. The fuel rods are supported by spacer grid springs and grid dimples that are located in the grid cell. The support system allows for some thermal expansion and imbalance of the fuel rods. The imbalance is absorbed by elastic energy to prevent coolant flow- induced vibration damage. Design requirements are defined and a design process is established. The design process includes mathematical optimization as well as practical design method. The shape of the grid spring is designed to maintain its function during the lifetime of the fuel assembly. A structural optimization method is employed for the shape design. Since the optimization is carried out in the linear range of finite element analysis, the optimum solution is verified by nonlinear analysis. A good design is found and the final design is compared with the initial conceptual design. Commercial codes are utilized for structural analysis and optimization.

• 한국소음진동공학회논문집
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• 제22권5호
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• pp.474-479
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• 2012

A comprehensive vibration assessment program for the advanced power reactor 1400(APR1400) 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 of the inner barrel assembly top plate in APR1400. 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 APR1400 reactor vessel internals.