• 산업경영시스템학회지
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• 제46권2호
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• pp.133-142
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• 2023

The large process plant is currently implementing predictive maintenance technology to transition from the traditional Time-Based Maintenance (TBM) approach to the Condition-Based Maintenance (CBM) approach in order to improve equipment maintenance and productivity. The traditional techniques for predictive maintenance involved managing upper/lower thresholds (Set-Point) of equipment signals or identifying anomalies through control charts. Recently, with the development of techniques for big analysis, machine learning-based AAKR (Auto-Associative Kernel Regression) and deep learning-based VAE (Variation Auto-Encoder) techniques are being actively applied for predictive maintenance. However, this predictive maintenance techniques is only effective during steady-state operation of plant equipment, and it is difficult to apply them during start-up and shutdown periods when rises or falls. In addition, unlike processes such as nuclear and thermal power plants, which operate for hundreds of days after a single start-up, because the pumped power plant involves repeated start-ups and shutdowns 4-5 times a day, it is needed the prediction and alarm algorithm suitable for its characteristics. In this study, we aim to propose an approach to apply the optimal predictive alarm algorithm that is suitable for the characteristics of Pumped Storage Power Plant(PSPP) facilities to the system by analyzing the predictive maintenance techniques used in existing nuclear and coal power plants.

• Nuclear Engineering and Technology
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• 제39권2호
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• pp.123-128
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• 2007

In order to apply a static fault-tree (FT) method to a system or a plant whose configuration changes dynamically, condition gates and a post processing method are used to effectively accommodate these changes. An operator's performance change, which can be caused by these configuration changes, should also be considered to assess the risk to a plant in a more realistic manner. This study aims to develop an integrated framework to accommodate various configuration changes and their effect on an operator’s performance by using the FT model. We applied a condition-based human reliability assessment (CBHRA) method to consider various conditions endured by an operator. That is, we integrated the CBHRA method with the conventional post processing method for modeling the system configuration changes. The effect of the condition monitoring systems installed in a plant is also considered. In this study, we show an example application of the integrated framework to a probabilistic safety assessment for the shutdown phase of a nuclear power plant.

• 한국멀티미디어학회논문지
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• 제19권2호
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• pp.352-359
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• 2016

This paper presents high efficiency 5A synchronous DC-DC buck converter. The proposed DC-DC buck converter works from 4.5V to 18V input voltage range, and provides up to 5A of continuous output current and output voltage adjustable down to 0.8V. This chip is packaged MCP(multi-chip package) with control chip, top side P-CH switch, and bottom side N-CH switch. This chip is designed in a 25V high voltage CMOS 0.35um technology. It has a maximum power efficiency of up to 94% and internal 3msec soft start and fixed 500KHz PWM(Pulse Width Modulation) operations. It also includes cycle by cycle current limit function, short and thermal shutdown protection circuit at 150℃. This chip size is 2190um*1130um includes scribe lane 10um.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.683-690
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• 1996

The MIDLOOP code has been developed for the evaluation of RES pressurization transients initiated from a loss-of-Residual Heat Removal System (RHRS) during mid-loop operation after reactor shutdown. It provides a fast running and realistic tool for studying parametrically the response of important plant parameters such as pressure, temperature, and level to various plant combinations of the primary side vent, makeup, and leakage procedures and the steam generator (SG) conditions. The code consists of ten nodes representing the primary and secondary sides of a nuclear power plant and can analyze the effect of air on the primary system pressurization and primary to secondary heat transfer. The analysis results of the MIDLOOP code are in good agreement with the ROSA-IV/LSTF experiment without opening in the RCS.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2008년도 제30회 춘계학술대회논문집
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• pp.289-292
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• 2008

Highly over-expanded nozzle of the rocket engines will be excited by non-axial forces due to flow separation at sea level operations. Since rocket engines are designed to produce axial thrust to power the vehicle, non-axial static and/or dynamic forces are not desirable. Several engine failures were attributed to the side loads. Present work investigate the unsteady flow in an over-expanded rocket nozzle in order to estimate side load during a shutdown/starting. Numerical computations has been carried out with density based solver on multi-block structured grid. Present solver is explicit in time and unsteady time step is calculated using dual time step approach. AUSMDV is considered as a numerical scheme for the flux calculations. One equation Spalart-Allmaras turbulence model is selected. Results presented here is for two nozzle pressure ratio i.e. 100 and 20. At 100 NPR, restricted shock separation (RSS) pattern is observed while, 20 NPR shows free shock separation (FSS) pattern. Side load is observed during the transition of separation pattern at different NPR.

• Structural Engineering and Mechanics
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• 제2권2호
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• pp.157-171
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• 1994

A general approach to the dynamic time-history analysis of the reactor core is presented in this paper as a part of the fuel assembly qualification program. Several detailed core models are set up to reflect the placement of the fuel assemblies within the core shroud. Peak horizontal responses are obtained for each model for the motions induced form earthquake. The dynamic responses such as fuel assembly deflected shapes and spacer grid impact loads are carefully investigated. Also, the sensitivity responses are obtained for the earthquake motions and the fuel assembly non-linear response characteristics are discussed.

• 한국유체기계학회 논문집
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• 제14권2호
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• pp.25-28
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• 2011

The control rod assemblies do freely fall into the reactor core by the gravity from the control rod drive mechanism. In order to achieve a rapid shutdown and control the reactor power, it is required to insert control rod assemblies as soon as possible. In this paper, we evaluated the drop time and flow characteristics caused around guide tube for SMART(System-integrated modular advanced reactor) control rod assembly. Numerical analyses are carried out with FLUENT program of computational fluid dynamics. This study results show that the drop time of the control rod assembly in the operating condition of SMART is more 20 percent rapidly than the drop time of the room temperature and ambient atmosphere condition.

• The Journal of the Acoustical Society of Korea
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• 제16권1E호
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• pp.9-14
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• 1997

Check valves used in industrial and Nuclear Power Plant safety systems are susceptible to failure modes generally associated with wear of internal parts. Specifically, hinge pins, disc studs, pistons, and other mechanical parts may degrade over time, and in some cases, may which might produce a disabling event leading to plant or process shutdown. The primary diagnostic technique in the past has been to disassemble the valves. This procedure is costly, time consuming, and in the nuclear industry, it can lead to radiation exposure in some situations. Additionally repair and reassembly of a valve does not ensure proper operation. Non-intrusive diagnostic technologies including acoustics and magnetics with a digital signal analysis allow to evaluate check valve performance without a disassembly and is able to help the user detect degraded valve conditions.

• Nuclear Engineering and Technology
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• 제47권2호
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• pp.148-156
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• 2015

Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs) that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C) strategies for a large 1,000 MWe iPWR is described. Reactor system modeling-which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum-is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

• 전력전자학회:학술대회논문집
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• 전력전자학회 2002년도 전력전자학술대회 논문집
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• pp.77-81
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• 2002

In many high performance engineering systems such as automated production system and transportation systems, AC-servo drives are employed as the most Important driving parts. And the faults of servo drives result in overall system performance deterioration or an unscheduled shutdown In critical situations. The real-time fault detection and isolation(FDI) scheme Is very useful to prevent them and to guarantee the desired reliability of the overall system. In this paper, the FDI schemes which can be applied to AC servo drives are introduced and some new results are presented.