• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.849-858
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• 2009

Risk caused by safety-critical instrumentation and control (I&C) systems considerably affects overall plant risk. As digitalization of safety-critical systems in nuclear power plants progresses, a risk model of a digitalized safety system is required and must be included in a plant safety model in order to assess this risk effect on the plant. Unique features of a digital system cause some challenges in risk modeling. This article aims at providing an overview of the issues related to the development of a static fault-tree-based risk model. We categorize the complicated issues of digital system probabilistic risk assessment (PRA) into four groups based on their characteristics: hardware module issues, software issues, system issues, and safety function issues. Quantification of the effect of these issues dominates the quality of a developed risk model. Recent research activities for addressing various issues, such as the modeling framework of a software-based system, the software failure probability and the fault coverage of a self monitoring mechanism, are discussed. Although these issues are interrelated and affect each other, the categorized and systematic approach suggested here will provide a proper insight for analyzing risk from a digital system.

• Proceedings of the Korea Society for Simulation Conference
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• 1998.03a
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• pp.86-90
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• 1998

The safety and reliability of nuclear power plant operations relies heavily on the plant operators ability to respond to various emergency situations. It has become standard industry practice to utilize simulators to improve the safety and reliability of nuclear power plants operations. The simulators built for Younggwang#3,4, which is the basic model of the Korean Nuclear Power Plant design, has been developed precisely for this purpose. Dynamic Model and Display Model are developed under US3(UNIX Simulation Software Support System) environment in simulator for Younggwang#3,4. Since these two models are developed under each own operating system, it is necessary to develop a method for transporting data between these two systems. This paper descirves communication environment between Dynamic Model and Display Model, and addresses a file generation method for the Display Model, which will be necessary for designing MMI of MCR(Main Control Room) in the furture.

• Journal of the Korean Society of Systems Engineering
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• v.14 no.2
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• pp.73-82
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• 2018

In this work, a hardware based cryptographic module for the cyber security of nuclear power plant is developed using a system engineering approach. Nuclear power plants are isolated from the Internet, but as shown in the case of Iran, Man-in-the-middle attacks (MITM) could be a threat to the safety of the nuclear facilities. This FPGA-based module does not have an operating system and it provides protection as a firewall and mitigates the cyber threats. The encryption equipment consists of an encryption module, a decryption module, and interfaces for communication between modules and systems. The Advanced Encryption Standard (AES)-128, which is formally approved as top level by U.S. National Security Agency for cryptographic algorithms, is adopted. The development of the cyber security module is implemented in two main phases: reverse engineering and re-engineering. In the reverse engineering phase, the cyber security plan and system requirements are analyzed, and the AES algorithm is decomposed into functional units. In the re-engineering phase, we model the logical architecture using Vitech CORE9 software and simulate it with the Enhanced Functional Flow Block Diagram (EFFBD), which confirms the performance improvements of the hardware-based cryptographic module as compared to software based cryptography. Following this, the Hardware description language (HDL) code is developed and tested to verify the integrity of the code. Then, the developed code is implemented on the FPGA and connected to the personal computer through Recommended Standard (RS)-232 communication to perform validation of the developed component. For the future work, the developed FPGA based encryption equipment will be verified and validated in its expected operating environment by connecting it to the Advanced power reactor (APR)-1400 simulator.

• Journal of Energy Engineering
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• v.7 no.1
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• pp.96-102
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• 1998

The use of computers as part of nuclear safety systems elicits additional requirements-software verification and validation (v/v), hardware qualification-not specifically addressed in general industry fields. The computer used in nuclear power plants is a system that includes computer hardware, software, firmware, and interfaces. To develop the computer systems graded with nuclear safety class, the developing environments have to be required in advance and the developed software have to be verified and validated in accordance with nuclear code and standards. With this requirements, the test facility for Inadequate Core Cooling Monitoring System (ICCMS) as one of safety systems in the nuclear power plants was developed. The test facility consists of three(3) parts such as Input/Output (I/O) simulator, Plant Data Acqusition System (PDAS) cabinets and supervisory computer. The performance of the system was validated by manual test procedure.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.587-591
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• 1997

The DGs (Diesel Generators) in NPP (Nuclear Power Plant) has been used for the emergency electric power source to shut down the nuclear reactor safely in case of station blackout. The RCM (Reliability Centered Maintenance) has been applied to DGs for increasing the safety of NPP. The structured defects of DG were not remedied by the improvement of maintenance method. As the first stage of RCM, to find the structured defects, its failure modes were searched and analyzed through the ten year maintenance information. The structured defects such as the air compressor, the lubricating oil pressure, and the insufficient load were the root causes of main failures. The air reservoir reinstallation, the lubricating oil tube modification, the load bank installation, and the qualitative instrumentation were the solutions for the hardware oriented RCM of DGs. There remains the software oriented RCM such as the rejection of useless maintenance, the preventive maintenance, the database of maintenance information, and the predictive maintenance.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.63 no.12
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• pp.1637-1639
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• 2014

In the past, the analog protection relays have been widely used for the safety-related systems in the nuclear power plants due to their stability and reliability. Meanwhile, as the high performance digital system has been developed, the digital systems have been adopted in the non-safety systems. However, since the digital systems currently used in the non-safety systems were not developed according to Q-class standard, Commercial Grade Item Dedication (CGID) procedure should be performed in order to apply them to the safety-related system. The purpose of this paper is to describe the CGID procedure including the analysis of the hardware architecture as well as the software embedded in protective relay to apply to the emergency diesel generator in the nuclear power plant. The entire CGID procedure was performed strictly according to the international standard and regulations.

• KNOM Review
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• v.22 no.3
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• pp.31-39
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• 2019

This paper introduces our research results of SDN/SON-based emergeny communications system for nuclear power plant under extreme natural diaster like Fukusima nuclear power plant incident. In such a extreme situation, nuclear power plant operational staffs which do not have technical knowledge of network provisioning and operations have to deploy the emergency network. Thus our proposed system provides capabilities to provision an emergency network autonomically and enable voice and data services. It also describes our PoC system developed based on the proposed core technology and architecture with the results of system verification testing.

• Journal of Energy Engineering
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• v.8 no.2
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• pp.213-223
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• 1999

In recently, the safety system software of the nuclear power plant has been verified and validated according to ANSI/IEEE-ANS-7-4.3.2-1982 to improve the reliability. This standard requires that safety-related software should be tested in the static and dynamic environments. In case of Inadequate Core Cooling Monitoring System (ICCMS), the static test procedure and related techniques are developed but the dynamic test procedure and related techniques are not developed. Therefore, this paper discusses the undeveloped techniques, and suggests the dynamic test procedure and the program for generation of test input data. The performance of the program was identified using accident analysis report of Ulchin 3&4 Final Safety Analysis Report (FSAR).

• Journal of the Korean Society of Systems Engineering
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• v.15 no.1
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• pp.1-8
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• 2019

Reverse engineering involves examining a system or component so as to comprehend its structure, functionality, and operation. Creation of a system model in reverse engineering can serve several purposes: test generation, change impact analysis, and the creation of a new or modified system. When attempting to reverse engineering a system, often the most readily accessible information is the system description, which does not readily lend itself to use in Model Based System Engineering (MBSE). Therefore, it is necessary to be able to transform this description into a diagram, which clearly depicts the behavior of the system as well as the interaction between components. This study demonstrates how sequence diagrams can be extracted from the systems description. Using MBSE software, the sequence diagrams for the Engineered Safety Features Component Control System (ESF-CCS) of the Nuclear Power Plant are created. Sequence diagrams are chosen because they are a means of representing the systems behavior and the interaction between components. In addition, from these diagrams, the system's functional requirements can be elicited. These diagrams then serve as the baseline of the reverse engineering process and multiple system views are subsequently be created from them, thus speeding up the development process. In addition, the use of MBSE ensures that any additional information obtained from auxiliary sources can then be input into the system model, ensuring data consistency.

• Journal of Applied Reliability
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• v.14 no.4
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• pp.248-255
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• 2014

Nuclear power plants may use commercial grade items if they can pass special tests or inspections, which is called Commercial Grade Item Dedication (CGID). Digital items, however, should follow EPRI TR-106439 to be applied to nuclear power plants. This paper explains the scheme and requirements of the EPRI TR-106439 and introduces some guidelines. Firstly, in order to clarify requirements of the EPRI TR-106439, code interpretation is performed. And through case study of digital indicator, limitations of EPRI TR-106439 are mentioned, and improvement direction is proposed.