• Proceedings of the KSR Conference
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• 2008.11b
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• pp.1487-1493
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• 2008

Railway vehicle is very important to implement the effective maintenance in proper to prevent any failure during operation period. Many railway authorities are making efforts to maintain the railway vehicle through scientific and systematic procedure. To achieve this, Reliability Centered Maintenance(RCM) is partially applied. The efficiency of RCM has proven and its terminology was familiar with nuclear power, military and chemical plant etc. since the commercial aircraft's industries has introduced the maintenance program based on the target of reliability. The application of RCM on railway vehicle can be utilized with systematic analysis method to select the best effective maintenance period and action to prevent the failures by selecting the equipment affecting the its safety and reliability. This paper is presented that the procedure of adequate and effective maintenance for railway vehicle by comparing among the related standards in example IEC60300-3,11, MIL-STD-2173, and technical documents or papers. In accordance with above result, RCM procedure is proposed to apply effectively for maintenance of railway vehicle. That is, (1) Analysis of data and Calculation of criticality per equipment (2) Selection of equipment to analyze (3) Analysis of failure mode and effect (4) Evaluation of maintenance method and period (5) Optimization of maintenance program through renewal of maintenance method and period.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.413-418
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• 1996

The main purpose of this paper is to develop the modified LTC code for accurate analysis of the boron concentration behavior of all components in the Nuclear Steam Supply System (NSSS). This is achieved by adapting a multi-cell mad to the existing Long Term Cooling (LTC) code. To verify the modified LTC, the simulated results were compared with the actual test results measured during YGN 4 initial criticality test. It was shown that the simulated results of this modified LTC were in good agreement with the actual test results. Also, the boron concentration behavior analysis were performed using the modified LTC code for both direct and indirect dilution/boration nude using YGN 3,4 design data. This modified LTC code can provide a valuable information in predicting boron concentration behavior during power maneuvering such as startup operation, shutdown operation and load follow operation. It is expected that the modified LTC can be applied to both on-line and off-line mode using Plant Computer System(PCS).

• Journal of the Korean Society for Aviation and Aeronautics
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• v.26 no.4
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• pp.116-121
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• 2018

According to the incresement of aging aircraft, Republic of Korea Air Force (RoKAF) conducted a teardown inspection of aircraft's EWIS (electrical wiring interconnection system) to determine the status of deterioration and the influence of failure occurrences due to it. The inspected aircraft were the retired fighter jets that had been used for more than 40 years. By analyzing defect type and the defect tendency, RoKAF can establish the necessary measures for the usage extension of their fleet and furthermore, the analysis results can be used as a basic data for the preparation of it's aircraft aging. EWIS inspection was done throughout careful visual inspection technique by removing all the ducts and pipes located in the fuselage and wings. For the aircraft wiring where no damage was found, the elongation tests were performed to determine the deterioration of wiring according to the location of the aircraft. The connectors, which is the main cause of intermittent failure, were completely disassembled and inspected for internal damage such as corrosion, abrasion, and traces of foreign objects. The detected defects were classified into 4 severity levels based on the type of damage, and the classified defects were weighted according to the criticality which may affects to it's system to establish the action plan.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.790-802
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• 2022

Large-scale reactor simulation often requires the use of Monte Carlo calculation techniques to estimate important reactor parameters. One drawback of these Monte Carlo calculation techniques is they inevitably result in some uncertainty in calculated quantities. The present study includes parametric uncertainty quantification (UQ) and sensitivity analysis (SA) on the Advanced Test Reactor Critical (ATRC) facility housed at Idaho National Laboratory (INL) and addresses some complications due to Monte Carlo uncertainty when performing these analyses. This approach for UQ/SA includes consideration of Monte Carlo code uncertainty in computed sensitivities, consideration of uncertainty from directly measured parameters and a comparison of results obtained from brute-force Monte Carlo UQ versus UQ obtained from a surrogate model. These methodologies are applied to the uncertainty and sensitivity of k_eff for two sets of uncertain parameters involving fuel plate geometry and fuel plate composition. Results indicate that the less computationally-expensive method for uncertainty quantification involving a linear surrogate model provides accurate estimations for k_eff uncertainty and the Monte Carlo uncertainty in calculated k_eff values can have a large effect on computed linear model parameters for parameters with low influence on k_eff.

• Journal of the Korean Society for Railway
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• v.13 no.1
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• pp.58-64
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• 2010

As a modem urban train is getting complex in terms of high-technology in its systems and components, the failure management should be performed with scientific and systematic technique. FMEA is a technique to analyze the failure trends of component parts and influences to the higher level system in order to discover the design incompleteness and potential defects, which is for improving reliability. Especially, FMECA (Failure Mode Effects, and Criticality Analysis) is used in case that the criticality that has an immense influence to the system is important. In case of urban train, in its design and manufacturing steps, FMEA is frequently used as an analysis technique to meet the safety objectives and eliminate potential hazards/failures since the concepts of reliability of train is introduced these days. Though, FMEA technique in the maintenances steps lacks in its investigation and applications yet. FMEA is also not applied to the trains operated by Seoul metro in the design and manufacture steps excepts the newest trains. In this paper, through analyzing the failures/maintenance data of the belt-type door systems used in trains operated in Seoul metro Line 1, which is accumulated in RIMS (Rolling-stock Information Maintenance System), FMEA procedures to the belt-type door engines are proposed. Especially, an effort is made, to approach the detailed FMECA procedures to the door magnet valve and switch and door engine devices which vastly influences the customer safety and satisfaction.

• Journal of Radiation Protection and Research
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• v.31 no.1
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• pp.37-46
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• 2006

For the first step to develop a reference disposal container of spent fuel to be used in a deep geological repository, this paper examined safe dimensions of the disposal container on the points of nuclear criticality and radiation safety and mechanical structural safety and provided basic information for dimensioning the container and configuration of the container components, and establishing the favorable and safe disposal conditions. When the safety factor for stress due to the external loads (hydrostatic and swelling pressure) is taken as 2.0, the safe diameter of the filler material to provide enough container strength under the assumed external loads is found to be 112cm with 13cm spacing between inner baskets in PWR container. Also the thickness of the thinner section between the fuel basket and the surface of the cast insert is determined to be 150 mm. Regarding these dimensions of the container, the PWR fuel container is sketched to accommodate 4 square assemblies or 297 CANDU fuel 297 bundles (33 circle tubes x 9 stacks). However the top and bottom parts need to be checked again through the detail radiation shielding analysis with respects to the emplacement position and handling processes of the disposal container.

• Journal of the Korean Society of Marine Environment & Safety
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• v.29 no.1
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• pp.76-85
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• 2023

Demand for hydrogen as a renewable energy resource is increasing. However, unlike conventional fossil fuels, hydrogen requires a dedicated refueling station for fuel supply. A risk assessment of hydrogen refueling stations must be undertaken to secure the infrastructure. Therefore, in this study, a risk assessment for hydrogen refueling stations was conducted through both qualitative and quantitative risk assessments. For the qualitative evaluation, the hydrogen dispenser module was evaluated as two nodes using the hazard and operability (HAZOP) analysis. The risk due to filter clogging and high-pressure accidents was evaluated to be high according to the criticality estimation matrix. For the quantitative risk assessment, the Hydrogen Korea Risk Assessment Module (Hy-KoRAM) was used to indicate the shape of the fire and the range of damage impact, and to evaluate the individual and social risks. The individual risk level was determined of to be as low as reasonably practicable (ALARP). Additional safety measures proposed include placing the hydrogen refueling station about 100m away from public facilities. The social risk level was derived as 1E-04/year, with a frequency of approximately 10 deaths, falling within the ALARP range. As a result of the qualitative and quantitative risk assessments, additional safety measures for the process and a safety improvement plan are proposed through the establishment of a restricted area near the hydrogen refueling station.

• Korean Journal of Computational Design and Engineering
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• v.18 no.1
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• pp.36-48
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• 2013

In general, the product is used under several circumstances including environmental and usage conditions. According to the circumstances, the product has various performance degradation processes. In order to optimize the lifecycle of product usage, it is important to observe the degradation process and make suitable decisions on product operations. However, there are not much research works in evaluating the degree of product degradation based on product usage data. Recently, due to emerging ICT (Information and Communication Technology) technologies, it becomes possible to get the product usage data. Based on the gathered data, it is possible to analyze the degree of product degradation. The analysis of product usage data can improve product use and product design with advanced decisions. To this end, this study addresses one approach based on FMEA/FMECA method, called PDMCA (Performance, Degradation Modes and Criticality Analysis) for evaluating product degradation status and making suitable decisions.

• The Journal of the Institute of Internet, Broadcasting and Communication
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• v.11 no.2
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• pp.135-143
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• 2011

Currently network management systems(NMS) just give useful information about the criticality of the alarms and the process of the fault analysis is mainly dependent on the experts who have many years experiences in the field. Due to these reasons it takes very much time and manpower cost to localize the real root of the fault from the alarm information. Therefore, to solve these problems in this research we analyze the probability of the fault for each alarm and provide how to give the problem solving procedure with confidence level and give idea to build a system to realize the problem solving procedure. In addition, we give a case study to show how to use the proposed ideas.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.1835-1840
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• 2004

The HANARO, a multi-purpose research reactor of 30 MWth, open-tank-in-pool type, has been under normal operation since its initial criticality in February, 1995. The HANARO is composed of inlet plenum, grid plate, core channel with flow tubes and chimney. The reactor core channel is located at about twelve meters (12 m) depth of the reactor pool and cooled by the upward flow that the coolant enters the lower inlet of the plenum,. rises up through the grid plate and the core channel and comes out from the outlet of chimney. A guide tube is extended from the reactor core to the top of the reactor chimney for easily un/loading a target under the reactor normal operation. But active coolant through the core can be quickly raised up to the top of the chimney through the guide tube by a jet flow. This paper describes an analytical analysis that is the study of the flow behavior through the guide tube under reactor normal operation and unloading the target. As results, it was conformed through the analysis results that the guide jet is suppressed under the top of the chimney after modifying the orifice diameter of 37.5 mm to 31 mm.