• Nuclear Engineering and Technology
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• v.40 no.5
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• pp.397-408
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• 2008

This paper describes the application of a software fault tree analysis (FTA) as one of the analysis techniques for a software safety analysis (SSA) at the design phase and its analysis results for the safety-critical software of a digital reactor protection system, which is called the KNICS RPS, being developed in the KNICS (Korea Nuclear Instrumentation & Control Systems) project. The software modules in the design description were represented by function blocks (FBs), and the software FTA was performed based on the well-defined fault tree templates for the FBs. The SSA, which is part of the verification and validation (V&V) activities, was activated at each phase of the software lifecycle for the KNICS RPS. At the design phase, the software HAZOP (Hazard and Operability) and the software FTA were employed in the SSA in such a way that the software HAZOP was performed first and then the software FTA was applied. The software FTA was applied to some critical modules selected from the software HAZOP analysis.

• Journal of the Korea Society for Simulation
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• v.24 no.4
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• pp.1-9
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• 2015

As the use of software is more wider in the safety-critical nuclear fields, so study to improve safety and quality of the software has been actively carried out for more than the past decade. In the nuclear power plant, nuclear man-machine interface systems (MMIS) performs the function of the brain and neural networks of human and consists of fully digitalized equipments. Therefore, errors in the software for nuclear MMIS may occur an abnormal operation of nuclear power plant, can result in economic loss due to the consequential trip of the nuclear power plant. Verification and validation (V&V) is a software-engineering discipline that helps to build quality into software, and the nuclear industry has been defined by laws and regulations to implement and adhere to a through verification and validation activities along the software lifecycle. V&V is a collection of analysis and testing activities across the full lifecycle and complements the efforts of other quality-engineering functions. This study propose a methodology based on V&V activities and related tool-chain to improve quality for software in the nuclear power plant. The optimized methodology consists of a document evaluation, requirement traceability, source code review, and software testing. The proposed methodology has been applied and approved to the real MMIS project for Shin-Hanul units 1&2.

• Proceedings of the Korean Society for Quality Management Conference
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• 2004.04a
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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.

• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.165-178
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• 2013

A comprehensive review on the technical standards about human factors (HF) design and software reliability maintenance for digital instrumentation and control (I&C) and human-machine interface technology (HMIT) in Japanese light water reactor nuclear power plants (NPPs) was given in this paper mainly by introducing the relevant activities at the Japan Electric Association to set up many industrial standards within the traditional framework of nuclear safety regulation in Japan. In Japan, the Fukushima Daiichi accident that occurred on March 11, 2011 has great impact on nuclear regulation and nuclear industries where concerns by the general public about safety have heightened significantly. However for the part of HF design and software reliability maintenance of digital I&C and HMIT for NPP, the author believes that the past practice of Japanese activities with the related technical standards can be successfully inherited in the future, by reinforcing the technical preparedness for the prevention and mitigation against any types of severe accident occurrence.

• Journal of Korean Society of Industrial and Systems Engineering
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• v.19 no.38
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• pp.61-68
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• 1996

Quality Assurance Principles must be effectively implemented in developments and in use of safety critical software in nuclear industry. Brief definitions related to computer software and quality assurance were defined and several methods for evaluating software quality were proposed herewith. Independent verification and validation was suggested to assure the quality of safety critical software.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.1047-1048
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• 2005

• Nuclear Engineering and Technology
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• v.19 no.1
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• pp.42-50
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• 1987

Through a project "Development of Nuclear Safety Parameter Monitoring System", a nuclear data link system was established between Kori nuclear unit 2 and Nuclear Safety Center. We present in this paper the selected parameter sets, a description of the developed pseudo-network software and the functional descriptions of the equipments involved. We also include the conceptual design of the Kori four unit ERF/SPDS system, along with the localization direction for the related software and hardware. hardware.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.602-607
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• 1995

This article presents safety review experience of microprocessor-based Interposing Logic System(ILS) of Engineering Safety Feature Actuation System(ESFAS). The ILS is the first application of computerized logic design to safety system in Korean nuclear power plants without verification of the system reliability by proven technology concept. As a result of evaluation for the ILS, Korea Institute of Nuclear Safety(KINS) concluded that the microprocessor-based ILS is not acceptable in some features detailed enough to defend against software common mode failures(CMF). Therefore, we required licensee to install hardwired interlock signal configuration and a Hardwired Backup Panel to control safety-related equipment. We believe that the microprocessor-based ILS with the hardwired backup panel and inter-connection of interlock signal by hardwired configuration will improve the plant safety.

• Nuclear Engineering and Technology
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• v.42 no.4
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• pp.377-381
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• 2010

In 2007 the Nuclear Energy Agency's Committee on the Safety of Nuclear Installations published Best Practice Guidelines for the use of CFD in Nuclear Reactor Safety. This paper provides an overview of the document' contents and highlights a few of its recommendations. The document covers the full extent of a CFD analysis from initial problem definition and selection of an appropriate tool for the analysis, through final documentation of results. It provides advice on selection of appropriate simulation software, mesh construction, and selection of physical models. In addition it contains extensive discussion of the verification and validation process that should accompany any high-quality CFD analysis.

• Nuclear Engineering and Technology
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• v.29 no.6
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• pp.433-443
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• 1997

In this work, an analytic approach to the dependability of software in the operational phase is suggested with special attention to the hardware fault effects on the software behavior : The hardware faults considered are memory faults and the dependability measure in question is the reliability. The model is based on the simple reliability theory and the graph theory which represents the software with graph composed of nodes and arcs. Through proper transformation, the graph can be reduced to a simple two-node graph and the software reliability is derived from this graph. Using this model, we predict the reliability of an application software in the digital system (ILS) in the nuclear power plant and show the sensitivity of the software reliability to the major physical parameters which affect the software failure in the normal operation phase. We also found that the effects of the hardware faults on the software failure should be considered for predicting the software dependability accurately in operation phase, especially for the software which is executed frequently. This modeling method is particularly attractive for the medium size programs such as the microprocessor-based nuclear safety logic program.