• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.780-787
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• 2018

This article presents a security module based on a field programmable gate array (FPGA) to mitigate man-in-the-middle cyber attacks. Nowadays, the FPGA is considered to be the state of the art in nuclear power plants I&C systems due to its flexibility, reconfigurability, and maintainability of the FPGA technology; it also provides acceptable solutions for embedded computing applications that require cybersecurity. The proposed FPGA-based security module is developed to mitigate information-gathering attacks, which can be made by gaining physical access to the network, e.g., a man-in-the-middle attack, using a cryptographic process to ensure data confidentiality and integrity and prevent injecting malware or malicious data into the critical digital assets of a nuclear power plant data communication system. A model-based system engineering approach is applied. System requirements analysis and enhanced function flow block diagrams are created and simulated using CORE9 to compare the performance of the current and developed systems. Hardware description language code for encryption and serial communication is developed using Vivado Design Suite 2017.2 as a programming tool to run the system synthesis and implementation for performance simulation and design verification. Simple windows are developed using Java for physical testing and communication between a personal computer and the FPGA.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.471-482
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• 2012

To assess the risk of nuclear power plant operation and to determine the risk impact of digital systems, there is a need to quantitatively assess the reliability of the digital systems in a justifiable manner. The Probabilistic Risk Analysis (PRA) is a tool which can reveal shortcomings of the NPP design in general and PRA analysts have not had sufficient guiding principles in modelling particular digital components malfunctions. Currently digital I&C systems are mostly analyzed simply and conventionally in PRA, based on failure mode and effects analysis and fault tree modelling. More dynamic approaches are still in the trial stage and can be difficult to apply in full scale PRA-models. As basic events CPU failures, application software failures and common cause failures (CCF) between identical components are modelled.The primary goal is to model dependencies. However, it is not clear which failure modes or system parts CCF:s should be postulated for. A clear distinction can be made between the treatment of protection and control systems. There is a general consensus that protection systems shall be included in PRA, while control systems can be treated in a limited manner. OECD/NEA CSNI Working Group on Risk Assessment (WGRisk) has set up a task group, called DIGREL, to develop taxonomy of failure modes of digital components for the purposes of PRA. The taxonomy is aimed to be the basis of future modelling and quantification efforts. It will also help to define a structure for data collection and to review PRA studies.

• Proceedings of the KIEE Conference
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• 2004.07d
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• pp.2275-2277
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• 2004

Korea Next Generation Reactor(KNGR) is in the midst of being developed and will exceed Korea Standard Nuclear Power Plant(KSNP) economically. Domestic Instrumentation and Control(I&C) systems shall be applied to KNGR and the development of Control Element Drive Mechanism Control System(CEDMCS) considered as an essential part in nuclear I&C system will be dealt with in this paper. The newly developed CEDMCS has the control cabinet using the nuclear Distributed Control System(DCS) made in Korea and the power cabinet produced by our research institute and interfaced with the DCS control cabinet.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.273-278
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• 1998

Belief network(or Bayesian network) based on Bayes' rule in probabilistic theory can be applied to the reasoning of diagnostic systems. This paper describes the basic theory of concept and feasibility of using the network for diagnosis of nuclear power plants. An example shows that the probabilities of root causes of a failure are calculated from the measured or believed evidences.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.102.1-102
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• 2001

• Nuclear Engineering and Technology
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• v.44 no.8
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• pp.919-928
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• 2012

The applications of computers and communication system and network technologies in nuclear power plants have expanded recently. This application of digital technologies to the instrumentation and control systems of nuclear power plants brings with it the cyber security concerns similar to other critical infrastructures. Cyber security risk assessments for digital instrumentation and control systems have become more crucial in the development of new systems and in the operation of existing systems. Although the instrumentation and control systems of nuclear power plants are similar to industrial control systems, the former have specifications that differ from the latter in terms of architecture and function, in order to satisfy nuclear safety requirements, which need different methods for the application of cyber security risk assessment. In this paper, the characteristics of nuclear power plant instrumentation and control systems are described, and the considerations needed when conducting cyber security risk assessments in accordance with the lifecycle process of instrumentation and control systems are discussed. For cyber security risk assessments of instrumentation and control systems, the activities and considerations necessary for assessments during the system design phase or component design and equipment supply phase are presented in the following 6 steps: 1) System Identification and Cyber Security Modeling, 2) Asset and Impact Analysis, 3) Threat Analysis, 4) Vulnerability Analysis, 5) Security Control Design, and 6) Penetration test. The results from an application of the method to a digital reactor protection system are described.

• Proceedings of the KIEE Conference
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• 2005.05a
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• pp.75-77
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• 2005

An integrated I&C architecture for nuclear power plants is designed by the systems and devices being developed in a project. Its design reference is the APR1400 that was design certified in Korea. Digital equipment and several kinds of data communication networks (DCN) are used. To confirm the validity of DCN based architecture design, the traffic loads fur each network were calculated assuming the anticipated maximum traffic condition. The analysis showed that the utilizations of all networks satisfied the design requirements.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1589-1599
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• 2017

Field programmable gate arrays (FPGAs) are getting more attention in safety-related and safety-critical application development of nuclear power plant instrumentation and control systems. The high logic density and advancements in architectural features make static random access memory (SRAM)-based FPGAs suitable for complex design implementations. Devices deployed in the nuclear environment face radiation particle strike that causes transient and permanent failures. The major reasons for failures are total ionization dose effects, displacement damage dose effects, and single event effects. Different from the case of space applications, soft errors are the major concern in terrestrial applications. In this article, a review of radiation effects on FPGAs is presented, especially soft errors in SRAM-based FPGAs. Single event upset (SEU) shows a high probability of error in the dependable application development in FPGAs. This survey covers the main sources of radiation and its effects on FPGAs, with emphasis on SEUs as well as on the measurement of radiation upset sensitivity and irradiation experimental results at various facilities. This article also presents a comparison between the major SEU mitigation techniques in the configuration memory and user logics of SRAM-based FPGAs.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.166-172
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• 1995

It is a general tendency to digitalize the conventional relay based I&C systems in nuclear power plant. But, the digitalization of nuclear safety systems has many a difficulty to surmount. The typical on thing of many difficulties is the data communication problem between local controllers and systems. The network architecture built with LAN(Local Area Network) in digital systems of the other industries are general. But in case of nuclear safety systems many considerations in point of safety and license are required to implement it in the field. In this paper, some considerations for applying LAN in nuclear safety systems were reviewed.

• Journal of the Ergonomics Society of Korea
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• v.31 no.1
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• pp.261-269
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• 2012

Objective: The aim of this study is to develop an analysis method to extract plausible types of errors when using a smart mobile in nuclear power plants. Background: Smart mobiles such as a smart-phone and a tablet computer(smart-pad) are to be introduced to the various industries. Nuclear power plant like APR1400 already adopted many up-to-date digital devices within its main control room. With this trend, various types of smart mobiles will be inevitably introduced to the nuclear field in the near future. However nuclear power plants(NPPs) should be managed considering a big risk as a result of the trend not only economically but also socially compared to the other industrial systems. It is formally required to make sure to reasonably prevent the all hazards due to the introduction of new technologies and devices before the application to the specific tasks in nuclear power plants. Method: We define interaction segments(IS) as a main architect of interaction description, and enumerate all plausible error segments(ES) for a part of design evaluation of digital devices. Results: We identify various types of interaction errors which are coped with reasonably by interaction design using smart mobiles. Conclusion: According to the application result of the proposed method, we conclude that the proposed method can be utilized to specify the requirements to the human error hazards in digital devices, and to conduct a human factors review during the design of digital devices. Application: The proposed method can be applied to predict the human errors of the tasks related to the digital devices; therefore we can ensure the safety to apply the digital devices to be introduced to NPPs.