• International Journal of Fuzzy Logic and Intelligent Systems
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• v.3 no.1
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• pp.112-120
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• 2003

This paper reports the development work carried out in respect of a proposed application of Neural Network approach for the Korean Next generation Reactor (KNGR) now referred as APR-1400. The emphasis is on establishing the methodology and the approach to be adopted towards realizing this application in the next generation reactors. Keeping in view the advantages and limitation of Artificial Neural Network Approach, the role of ANN has been limited to plant status or to be more precise plant transient monitoring. The simulation work carried out so far and the results obtained shows that artificial neural network approach caters to the requirements of plant status monitoring and qualifies to be incorporated as a part of proposed operator support systems of the referenced nuclear power plant.

• Proceedings of the KIEE Conference
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• 1998.07g
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• pp.2321-2323
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• 1998

A control system using a PLC has been developed for a dual tower drier(DTD) in a CANDU type nuclear power plant. This system will replace the existing DTD control system which was implemented with mechanical timers and relays. The new control system makes it possible for an operator to perform more precise time and dew point control for the DTD, thanks to the high efficiency and flexibility of the PLC. The operational cost for the control system is much reduced compared to the existing system.

• Proceedings of the ESK Conference
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• 1996.04a
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• pp.56-62
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• 1996

The main control room (MCR) of a nuclear power plant plays an important role in the operation of the plant. Since the traditional man-machine interface of the current MCR is old-fashioned, a next-generation MCR, that provides a VDT-based human-computer interface is being designed. This paper aims to provide a systematic and efficient method for converting a traditional man-machine interface of the MCR into a VDT-based one. Procedures and analysis methods are presented for efficient and effective development.

• Journal of the Korean Society of Safety
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• v.17 no.4
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• pp.173-177
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• 2002

THERP (Technique for Human Error Rate Prediction) methodology has been widely used for probabilistic safety assessments. The NUREG report involving this methodology is also called the HRA handbook. The THERP assumes that all actions involved in implementing a task are considered as components. In this paper human error rates associated with maintenance are evaluated by the THERP methodology. A gas governor system is used as an example which is also a risky system like nuclear power plants. It is also demonstrated that this approach is flexible in that it can be applied to any operator actions related to test and maintenance.

• Proceedings of the KIEE Conference
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• 1993.07a
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• pp.525-528
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• 1993

KEPCO Research Center is carrying out a simulator(full scope replica type) development project for two nuclear power plants(Kori-2, Younggwang-3,4) and one fossil power plant(Poryong-3,4). In this project, we aim not only the installation of high performance simulators at the power plant sites but also the realization of self reliance of power plant simulation technology in Korea. In the course of preparing procurement specification for the 3 simulators, the present status of power plant simulation technology has been surveyed and is presented in this paper. The fidelity of simulation and the automation of simulation model production has been greatly improved due to the ever increasing computing power of today's workstations. The need and importance of the application of high fidelity simulators to the operator training is refocused since the accident at TMI Nuclear Power Plant, U.S.A.

• Nuclear Engineering and Technology
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• v.29 no.1
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• pp.1-6
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• 1997

This paper presents a new dynamic approach for assessing feasibility associated with the implementation of accident management strategies by the operators. This approach includes the combined use of both the concept of reliability physics and a dynamic event tree generation scheme. The reliability physics is based on the concept of a comparison between two competing variables, i.e., the requirement and the achievement parameter, while the dynamic event tree generation scheme on the continuous generation of the possible event sequences at every branch point up to the desired solution. This approach is applied to a cavity flooding strategy in a reference plant, which is to supply water into the reactor cavity using emergency fire systems in the station blackout sequence. The MAAP code and Latin Hypercube sampling technique are used to determine the uncertainty of the requirement parameter. It has been demonstrated that this combined methodology may contribute to assessing the success likelihood of the operator actions required during accidents and therefore to developing the accident management procedures.

• Nuclear Engineering and Technology
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• v.34 no.2
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• pp.176-185
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• 2002

A 4-bundle shift refuelling method of CANDU6 F/H (Fuel Handling) System is analyzed to assess the operational flexibility and capacity of F/H system. The current 8-bundle shift refuelling scheme requires to refuel eight fuel bundles from a single fuel channel, and to refuel 14 fuel channels in a week on average assuming that the reactor is in a steady state. The analysis showed that the 4-bundle shift refuelling method increases F/M (Fuelling Machine) duty cycle and operator load. However, the fuellin’g method change from the 8- to 4-bundle shift refuelling ill not require additional team of operators. A marginal increase in the maintenance cost may be resulted in by the change of fuelling method and the increase of fatigue usage factors requires some components to be replaced during the FM lifetime.

• Journal of the Ergonomics Society of Korea
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• v.21 no.3
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• pp.13-24
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• 2002

Reduction and prevention of human error is one of the major interests for the enhancement of system safety and availability in Nuclear Power Plants (NPPs). As human beings have become the weak point in the system safety, a systematic evaluation on human performance during emergency situation should be performed in advance to identify the potential vulnerability of human tasks. Though the data gathering and analysis from real field is an important precondition, there were no available data in nuclear field of korea. This paper presents the result of human performance analysis on emergency tasks in NNPs. Firstly, a task analysis was performed to identify the characteristics of operator tasks during emergency condition and to classify them into a set of generic emergency tasks. Secondly, simulation data were collected and analyzed for the emergency tasks using the full scope simulator of Younggwang NPPs. The analyzed human performance information cover the event diagnosis time, the execution time of each procedural step, observation parameters, typer of irrelevant response, pattern of communication among staffs, and so on. These performance data would be used for human reliability analysis and the research of human error as technical bases.

• Journal of the Korean Society of Systems Engineering
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• v.12 no.2
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• pp.81-90
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• 2016

The system performance of Engineered Safety Features is of utmost importance in a nuclear power plant. The human performance is identified as most critical to assurance of the optimal operability of safety systems during an emergency. The aim of this study is to determine how the performance of safety system could be evaluated using Augmented Reality technology. The paper presents a description of how a systems engineered approach could be used to develop the necessary operating conditions needed to conduct this measurement. Augmented Virtual Reality (AVR) interface technology is achieving ease of availability and widespread use in many applications today as illustrated by the launch of several AR and VR devices aimed at media consumption. As such, environments that incorporate such AVR hardware have become invaluable tools in designing human interface systems because of the high fidelity and intuitive response to natural human interaction that can be achieved [2]. The outcome of the measurement undertaken is to determine whether 1.) Operator(s) performance can be enhanced by introducing an improved cognitive method of monitoring plant information during an Emergency Operating Procedures (EOP) and 2.) In correlation, inform the performance of the diverse safety systems on the basis of human factors.

• Nuclear Engineering and Technology
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• v.32 no.4
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• pp.316-327
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• 2000

The Emergency Core Cooling System in current nuclear power plants typically has a considerable number of complex functions and largely cumbersome operator interfaces. Functions for initiation, switch-over between various phases of operation, interlocks, monitoring, and alarming are usually performed by relays and analog comparator logic which are difficult to maintain and test. To improve problems of an analog based ECC (Emergency Core Cooling) System, the trip computer for ECCS based on Dynamic Safety System (DSS) is implemented. The DSS is a computer based reactor protection system that has fail-safe nature and performs a dynamic self-testing. The most important feature of the DSS is the introduction of test signal that send the system into a tripped state. The test signals are interleaved with the plant signals to produce an output which switches between a tripped and health state. The dynamic operation is a key feature of the failsafe design of the system. In this work, a possible implementation of the DSS using PLC is presented for a CANDU Reactor. ECC System of the CANDU Reactor is selected as the reference system.