• Proceedings of the KIEE Conference
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• 2007.10c
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• pp.201-203
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• 2007

Safety critical systems are those in which a failure can have serious and irreversible consequences. Nowadays digital technology has been rapidly applied to critical system such as railways, airplanes, nuclear power plants, vehicles. The main difference between analog system and digital system is that the software is the key component of the digital system. The digital system performs more varying and highly complex functions efficiently compared to the existing analog system because software can be flexibly designed and implemented. The flexible design make it difficult to predict the software failures. This paper reviews safety standard and criteria for safety critical system such as railway system and suggests software development methodology for more detail description.

• Electronics and Telecommunications Trends
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• v.8 no.3
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• pp.39-52
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• 1993

This study involves a Common Cause Failure (CCF) problem on the ultra-high reliability required system development such as war game operations, nuclear power control, air traffic control, space shuttle missions, and large scale network communication system. The system situation problems are defined according to CCF, reliability and system fault identifications for the development cast verifications in the multi-version redundant software system. Then, CCF analysis of redundant system, system principles and statistical dependence are also described. This validation oh the CCF in the human software interaction system will notify software engineers to conceive what really is CCF contribution factor, not only the internal but the external ones.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.41-42
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• 2017

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3653-3664
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• 2021

In PWR three-dimensional transient coupling calculation software CORCA-K, the nodal Green's function method and diagonal implicit Runge Kutta method are used to solve the spatiotemporal neutron dynamic diffusion equation, and the single-phase closed channel model and one-dimensional cylindrical heat conduction transient model are used to calculate the coolant temperature and fuel temperature. The LMW, NEACRP and PWR MOX/UO₂ benchmarks and FangJiaShan (FJS) nuclear power plant (NPP) transient control rod move cases are used to verify the CORCA-K. The effects of burnup, fuel effective temperature and ejection rate on the control rod ejection process of PWR are analyzed. The conclusions are as follows: (1) core relative power and fuel Doppler temperature are in good agreement with the results of benchmark and ADPRES, and the deviation between with the reference results is within 3.0% in LMW and NEACRP benchmarks; 2) the variation trend of FJS NPP core transient parameters is consistent with the results of SMART and ADPRES. And the core relative power is in better agreement with the SMART when weighting coefficient is 0.7. Compared with SMART, the maximum deviation is -5.08% in the rod ejection condition and while -5.09% in the control rod complex movement condition.

• Journal of Computing Science and Engineering
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• v.11 no.1
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• pp.9-23
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• 2017

NuDE 2.0 (Nuclear Development Environment 2.0) is a formal-method-based software development, verification and safety analysis environment for safety-critical digital I&Cs implemented with programmable logic controller (PLC) and field-programmable gate array (FPGA). It simultaneously develops PLC/FPGA software implementations from one requirement/design specification and also helps most of the development, verification, and safety analysis to be performed mechanically and in sequence. The NuDE 2.0 now consists of 25 CASE tools and also includes an in-depth solution for indirect commercial off-the-shelf (COTS) software dedication of new FPGA-based digital I&Cs. We expect that the NuDE 2.0 will be widely used as a means of diversifying software design/implementation and model-based software development methodology.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.335-340
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• 1998

In this paper a prototype of Requirement Tracking and Verification System(RTVS) for a Distributed Control System was implemented and tested. The RTVS is a software design and verification tool. The main functions required by the RTVS are managing, tracking and verification of the software requirements listed in the documentation of the DCS. The analysis of DCS software design procedures and inter(aces with documents were performed to define the user of the RTVS, and the design requirements for RTVS were developed.

• Journal of the Korea Institute of Information Security & Cryptology
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• v.25 no.2
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• pp.363-374
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• 2015

Instrumentation and control(I&C) system has been mainly designed and operated based on analog technologies in existing Nuclear Power Plants(NPPs). However, As the development of Information Technology(IT), digital technologies are gradually being adopted in newly built NPPs. I&C System based on digital technologies has many advantages but it is vulnerable to cyber threat. For this reason, cyber threat adversely affects on safety and reliability of I&C system as well as the entire NPPs. Therefore, the software equipped to NPPs should be developed with cyber security attributes from the initiation phase of software development life cycle. Moreover through cyber security assessment, the degree of confidence concerning cyber security should be measured and if managerial, technical and operational work measures are implemented as intended should be reviewed in order to protect the I&C systems and information. Currently the overall cyber security program, including cyber security assessment, is not established on I&C systems. In this paper, we propose cyber security assessment methods in the Software Development Life Cycle by drawing cyber security activities and assessment items based on regulatory guides and standard technologies concerned with NPPs.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.593-594
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.87.1-87.1
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.87-87
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• 1999