• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.191-196
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• 1996

The potential value of the Application Specific Integrated Circuits(ASIC's) in safety systems of Nuclear Power Plants(NPP's) is being increasingly recognized because they are essentially hardwired circuitry on a chip, the reliability of the system can be proved more easily than that of software based systems which is difficult in point of software V&V(Verification and Validation). There are two types of ASIC, one is a full customized type, the other is a half customized type. PLD(Programmable Logic Device) used in this paper is a half customized ASIC which is a device consisting of blocks of logic connected with programmable interconnections that are customized in the package by end users. This paper describes the RPS(Reactor Protection System) composed of ASICs which provides emergency shutdown of the reactor to protect the core and the pressure boundary of RCS(Reactor Coolant System) in NPP's. The RPS is largely composed of five logic blocks, each of them was implemented in one PLD, as the followings. A). Bistable Logic B). Matrix Logic C).Initiation Logic D). MMI(Man Machine Interface) Logic E). Test Logic.

• 제어로봇시스템학회논문지
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• 제14권1호
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• pp.36-41
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• 2008

Cooling tower is an important equipment of the cooling systems for large buildings like factory and department store. Water used for cooling in cooling tower is reused continuously. If the water is polluted, corrosion and scale can happen at equipments and pipes. In order to prevent this problem, it is necessary to control the water quality using chemicals. To control the water quality, an automatic control system is designed, fabricated, and experimented. The control system is based on an imbedded microcontroller. Relays are used for power driving, an LCD and LED for display, and RS485 for remote data acquisition. Monitoring program is also developed for easy man-machine interface and extraction of data stored in the imbedded processor and EEPROM. The control system calculates amounts of chemicals necessary using sensor data and injects the chemicals into the cooling tower on proper time. The developed water quality control system is expected to reduce cost of maintenance and extend the lifetime of the cooling systems with low cost.

• 한국안전학회지
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• 제31권6호
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• pp.129-134
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• 2016

Standard design of APR+(advanced power reactor plus) was certified at 2014 by Korea regulatory body. Based on the experience gained from OPR1000 and APR1400, the APR1400 was being developed as a 1,500MWe class reactor using Korean technologies for design code, reactor coolant pump, and man-machine interface system. APR+ has been basically designed to have the seismic design basis of safe shutdown earthquake (SSE) 0.3g, a 4-train safety concept based on N+2 design philosophy, and a passive auxiliary feedwater system (PAFS). Also, safety issues on the Fukushima-type accidents have been extensively reviewed and applied to enhance APR+ safety. APR+ provides higher reliability and safety against tsunami and earthquake. The purpose of this paper is to implement probabilistic safety assessment considering these design features and to analyze sensitivity of core damage frequency for large loss of coolant accident of APR+.

• 대한전기학회논문지:전기물성ㆍ응용부문C
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• 제50권11호
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• pp.562-568
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• 2001

In this paper, we described the export system to monitor and diagnose 765KV power apparatus. To develop this expert system, we studied the knowledge bases and data bases for 765KV transformer and GIS. In order to make the reliable inference of knowledge base and the good MMI(Man Machine Interface), the data bases were consisted of the tables of power apparatus information, limit level value, measured input data, inference result and diagnosis result. The knowledge base had various rules to infer the conditions of transformer and GIS. We applied both the forward chaining and backward chaining methods to these rules of system for good inferences. This paper describes the applied methods for expert system. Also, this developed system was tested with dissolved gas analyzing result and the result was shown.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 2000년도 제15차 학술회의논문집
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• pp.180-180
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• 2000

The shape of cold strip for the stainless process has been become issue in quality recently, and hence POSCO (Pohang Iron & Steel Co., Ltd) developed an automatic control system for strip shape in the sendzimir mill. The strip shape is measured by an outward measuring roll and is controlled by As_U roll and first intermediate roll. As_U roll consists of 8 saddles, which are controlled vertically. The fist intermediate rolls, which are controlled horizontally, consist of two pairs of rolls up and down. A developed shape control system is applied to real plant by using fuzzy logic and neural network method to control actuators; As_U roll and first intermediate roll. This system composes mainly of three parts as a real-time system, input to output conditioner board, and man-machine interface. The actual shape is recognized by neural network and converted into symmetric shape. The fuzzy controller, based on the shape from neural network and sensor, controls positions of the As_U roll and first intermediate roll. This paper verifies the shape controller performance. The experiments are made on line for the sendzimir mill. The shape control performance shows very efficient for the target tracking, shape symmetry, and fluctuation of shape.

• 한국품질경영학회:학술대회논문집
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• 한국품질경영학회 2004년도 품질경영모델을 통한 가치 창출
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• pp.114-119
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• 2004

Software verification and validation(V&V) is a means to develop high-quality software and assure safety and reliability for software. Also, we can achieve the desired software quality through systematic V&V activities. The software to be applied safety critical system like nuclear power plants is required to setup the V&V methodology that comply with licensing requirements for nuclear power plants and should be performed V&V activities according to it. In this paper, we classified safety-critical, safety-related and non-safety for software according to safety function to be peformed and define V&V activities to be applied software grade. Also, we defined V&V activities, procedures and documentation for each phase of software development life cycle and showed techniques and management to perform V&V. Finally, we propose the V&V framework to be applied software development of SMART(System-integrated Modular Advanced ReacTor) MMIS (Man-Machine Interface System) and to comply with domestic licensing requirements.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2014년도 추계학술대회 논문집
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• pp.110-115
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• 2014

The nuclear power plant was launched by Kori unit 1 in 1978 years. Currently, 23 nuclear power plants have been operating in Korea since 1978 years. The localization was completed for most of the reactor facility from Hanbit(Youngkwang) unit 3&4. However, RCP(Reactor Coolant Pump) and MMIS(Man Machine Interface System) is an important technology that has been excluded from the scope of the technical transfer has been dependent on a specific overseas vendor. Recent success in RCP development through co-operation with government and industries. Developed RCP will be applied to Shin-Hanul unit 1&2 nuclear power plants. The RCP operates in high speed and high pressure condition and only rotating component in the NSSS(Nuclear Steam Supply System). Therefore, the problem of vibration has arisen caused by the hydraulic forces of the working fluid. These forces can influence on the stability characteristics for entire RCS(Reactor Coolant System) loop, and can act as significant destabilizing forces. In this study, vibration evaluation of the pump shaft of development RCP estimated under normal operation and over speed conditions. In order to predict the vibration characteristics and dynamic behavior, modal analysis, critical speed analysis and unbalance response spectrum analysis were performed.

• 대한인간공학회지
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• 제25권3호
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• pp.85-96
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• 2006

Main control room(MCR) man-machine interface(MMI) design of advanced nuclear power plants(NPPs) such as APR(advanced power reactor)-1400 can be validated through performance-based tests to determine whether it acceptably supports safe operation of the plant. In this work, plant performance, personnel task, situation awareness, workload, teamwork, and anthropometric/physiological factor are considered as factors for the human performance evaluation. For development of measures in each of the factors, techniques generally used in various industries and empirically proven to be useful are adopted as main measures and some helpful techniques are developed in order to complement the main measures. Also the development of the measures is addressed based on the theoretical background. Finally we discuss the way in which the measures can be effectively integrated and then HUPESS(HUman Performance Evaluation Support System) which is in development based on the integrated way is briefly introduced.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.477-488
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• 2013

The PLC (Programmable Logic Controller) has been widely used to implement real-time controllers in nuclear RPSs (Reactor Protection Systems). Increasing complexity and maintenance cost, however, are now demanding more powerful and cost-effective implementation such as FPGA (Field-Programmable Gate Array). Abandoning all experience and knowledge accumulated over the decades and starting an all-new development approach is too risky for such safety-critical systems. This paper proposes an RPS software development process with a platform change from PLC to FPGA, while retaining all outputs from the established development. This paper transforms FBD designs of the PLC-based software development into a behaviorally-equivalent Verilog program, which is a starting point of a typical FPGA-based hardware development. We expect that the proposed software development process can bridge the gap between two software developing approaches with different platforms, such as PLC and FPGA. This paper also demonstrates its effectiveness using an example of a prototype version of a real-world RPS in Korea.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2237-2250
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• 2021

The analog-based Ex-core Neutron Flux Monitoring System (ENFMS) in Korean Nuclear Power Plants (NPPs) has been performing its intended functions successfully for a long time. On the other hand, the primary concern with the extended use of analog systems is the aging effect, such as mechanical failures, environmental degradation, and obsolescence. The transition to a digital-based Man-Machine Interface Systems (MMIS) in Korea and other countries has been accelerating, but some systems are still analog-based IC systems, such as the ENFMS in APR1400 NPPs. Digitalized ENFMS can become a reality using computers and microprocessors owing to the progress in digital electronics and information technology. This paper presents the result of the first phase of the research on the digitalization of the ENFMS signal processing electronics for NPPs operated or produced in Korea. It has two main parts: (1) review engineering bases of ex-core neutron flux monitoring system, including nuclear engineering, instrumentation techniques, and analog and digital signal processing techniques, and (2) analysis of analog signal processing electronics of ENFMS for OPR1000 and APR1400 power plants. They are prerequisite to the second phase of the research which is the detailed implementation of the digitalization.