• The KIPS Transactions:PartD
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• v.14D no.1 s.111
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• pp.121-130
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• 2007

in a software development, the design or architecture prior to implementing the software is essential for the success. This paper presents a case that we successfully designed a software architecture of radiation monitoring system (RMS) for HANARO research reactor currently operating in KAERI by applying the quality attribute-driven design method which is modified from the attribute-driven design (ADD) introduced by Bass[1]. The quality attribute-driven design method consists of following procedures: eliciting functionality and quality requirements of system as architecture drivers, selecting tactics to satisfy the drivers, determining architectures based on the tactics, and implementing and validating the architectures. The availability, maintainability, and interchangeability were elicited as duality requirements, hot-standby dual servers and weak-coupled modulization were selected as tactics, and client-server structure and object-oriented data processing structure were determined at architectures for the RMS. The architecture was implemented using Adroit which is a commercial off-the-shelf software tool and was validated based on performing the function-oriented testing. We found that the design method in this paper is an efficient method for a project which has constraints such as low budget and short period of development time. The architecture will be reused for the development of other RMS in KAERI. Further works are necessary to quantitatively evaluate the architecture.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.37 no.12
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• pp.1175-1183
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• 2013

Even if some CFD software developers and its users think that a state-of-the-art CFD software can be used to reasonably solve at least single-phase nuclear reactor safety problems, there remain limitations and uncertainties in the calculation result. From a regulatory perspective, the Korea Institute of Nuclear Safety (KINS) is presently conducting the performance assessment of commercial CFD software for nuclear reactor safety problems. In this study, to examine the prediction performance of commercial CFD software with the porous model in the analysis of the scale-down APR (Advanced Power Reactor Plus) internal flow, a simulation was conducted with the on-board numerical models in ANSYS CFX R.14 and FLUENT R.14. It was concluded that depending on the CFD software, the internal flow distribution of the scale-down APR was locally somewhat different. Although there was a limitation in estimating the prediction performance of the commercial CFD software owing to the limited amount of measured data, CFX R.14 showed more reasonable prediction results in comparison with FLUENT R.14. Meanwhile, owing to the difference in discretization methodology, FLUENT R.14 required more computational memory than CFX R.14 for the same grid system. Therefore, the CFD software suitable to the available computational resource should be selected for massively parallel computations.

• Journal of KIISE:Software and Applications
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• v.29 no.12
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• pp.860-872
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• 2002

Fault tree analysis is the most widely used saftly analysis technique in industry. However, the analysis is often applied manually, and there is no systematic and automated approach available to validate the analysis result. In this paper, we demonstrate that a real-time model checker UPPAAL is useful in formally specifying the required behavior of safety-critical software and to validate the accuracy of manually constructed fault trees. Functional requirements for emergency shutdown software for a nuclear power plant, named Wolsung SDS2, are used as an example. Fault trees were initially developed by a group of graduate students who possess detailed knowledge of Wolsung SDS2 and are familiar with safety analysis techniques including fault tree analysis. Functional requirements were manually translated in timed automata format accepted by UPPAAL, and the model checking was applied using property specifications to evaluate the correctness of the fault trees. Our application demonstrated that UPPAAL was able to detect subtle flaws or ambiguities present in fault trees. Therefore, we conclude that the proposed approach is useful in augmenting fault tree analysis.

• Journal of Convergence Society for SMB
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• v.2 no.1
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• pp.11-18
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• 2012

As scale of software has been expanded and complicated, it is difficult to detect hazards which induce functional failure of software. Functional failure of safety-critical system (nuclear power plant, air traffic control systems, railway operating system) could result in a disaster (personal injury, environmental pollution). Therefore, it is necessary to conduct a safety analysis for preventing functional failure and increasing safety of the software. However, there are some reasons (time and effort problem, low knowledge of various safety analysis techniques, selecting conventional technique in company, organization) which disturb selecting an apposite one. This paper presents some traditional safety analysis techniques, recently presented techniques and combined models. We expect that it helps stakeholders to choice adequate one for target system.

• Journal of the Korea Institute of Information Security & Cryptology
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• v.34 no.2
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• pp.229-244
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• 2024

Recently, supply chain attacks against nuclear facilities such as "Evil PLC" are increasing due to the application of digital technology in nuclear power plants such as the APR1400 reactor. Nuclear supply chain security requires a asset management system that can systematically manage a large number of providers due to the nature of the industry. However, due to the nature of the control system, there is a problem of inconsistent management of attribute information due to the long lifecycle of software assets. In addition, due to the availability of the operational technology, the introduction of automated configuration management is insufficient, and limitations such as input errors exist. This study proposes a systematic asset management system using SBOM(Software Bill Of Materials) and an improvement for input errors using natural language processing techniques.

• Journal of KIISE:Computing Practices and Letters
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• v.15 no.4
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• pp.265-269
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• 2009

Formal specification plays important roles in guaranteeing software safety of safety-critical systems such as nuclear power plant's digital control systems. We had developed a technique [1] which synthesizes Function Block Diagram(FBD) programs from NuSCR formal requirements specifications, but it did not be used widely as it had no automatic tool support. FBD is one of the programming languages for Programmable Logic Controllers(PLC) based system. This paper introduces a CASE tool, NuSCRtoFBD, developed to automate the synthesis procedure. The CASE tool NuSCRtoFBD can reduce a number of errors occurred in the process of manual FBD programming.

• Proceedings of the Korean Information Science Society Conference
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• 2005.07a
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• pp.853-855
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• 2005

본 논문은 Sequence Diagram을 이용한 안전등급 PLC(Safety-Grade Programmable Logic Controller) 운영 체제의 인터페이스 설계명세를 기술한다. 원자력 발전소에 사용하기 위한 안전 소프트웨어(Safety Software)의 규제 기준인 Reg. Guide는 IEEE Std 1016, IEEE Std 1016.1 의 설계명세서 작성 표준에 따라 작성하도록 요구하며, 이러한 규제기준과 기 표준을 안전등급 PLC를 위한 운영체제 소프트웨어의 설계명 세서도 만족해야 한다. 특히 운영 체제와 외부 장치들 사이의 인터페이스를 위해 실시간 특성의 표현에 장점을 갖는 Sequence Diagram을 적용함으로써 운영체제의 인터페이스에 대한 정확성, 완전성, 그리고 일관성을 향상시킬 수 있었다.

• Fire Science and Engineering
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• v.18 no.1
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• pp.24-30
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• 2004

In this work the dynamic heat transfer occurring in a cable penetration fire stop system built in the firewall of nuclear power plants is three-dimensionally investigated to develop a test-simulator that can be used to verify effectiveness of the sealants. Here was carried out an experiment to observe the heat transfer in the cable penetration fire stop system made of DOW CORNING products. The dynamic heat transfer occurring in the fire stop system is formulated in a parabolic partial differential equation subjected to a set of initial and boundary conditions. And it was modeled, simulated, and analyzed. The simulation results were illustrated in three-dimensional graphics and were compared with experimental data. Through the simulations, it was shown clearly that the temperature distribution was influenced very much by the number, position, and temperature of the cable streams. It also was found that the dynamic heat transfer through the cable streams was one of the most dominant factors, and the feature of heat conduction could be understood as an unsteady-state process. It is certain that these numerical results are useful for making a performance-based design for the cable penetration fire stop system.

• Proceedings of the KIPE Conference
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• 2000.07a
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• pp.249-252
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• 2000

원자력 교육원 #2(KNPEC-2) 시뮬레이터는 1980년도 중반에 웨스팅하우스에 의해 공급되어 계속 사용되어 오다가 현재 성능개선 연구가 진행 중이다. 이번 성능개선을 통해 기존의 컴퓨터 시스템(Gould MPX)와 소프트웨어의 전면 교체가 이루어지고 있으며 최적 계산 코드를 이용한 실시간 열수력 모델 (ARTS; Advanced Real-Time Thermal-Hydraulics Simulation) 개발 , 2-Group 3D 실시간 노심모델(REMARK ; REal Time Multigroup Advanced Reactor Kinetics)를 이용한 노심 주기개선 (Cycle Update) 가상현실 기술 등을 이용한 컴퓨터 교육지원 시스템(CATS: Computer Assister Training System)등 새로운 시도가 이루어지고 있으며 본 논문은 이러한 새로운 시도가 이루어지고 있으며 본 논문은 이러한 새로운 시도들 및 그 결과에 대해 기술하고 있다. 기준발전소(Reference Plant)인 영광 1호기 12주기의 노심모델로 주기개선(Cycle Update)을 위한 REMARK의 입력자료 생성을 위해 핵설계 전산체계인 APA(ALPHA-PHOENIX-ANC) 시스템의 출력으로부터 자동으로 REMARK 입력데이타를 생성하기 위한 GUI툴 개발하였다. 또 이를 이용하여 개발된 노심모델은 최적계산코드(RETRAn 3D) 의 열수력 해법을 이용하여 개발된 NSSS 열수력코드(ARTS) 와 결합(Integration) 되어 안정 및 과도 상태 시험에 사용되었으며 원자로 냉각재 펌프 정지등의 몇 가지 과도 시험 계산결과 기존 해석 결과와 잘 일치하였다 중앙제어실(MCR; Main Control Room)내의 운전원 행동만 훈련하도록 되어있는 기존시뮬레이터의 한계를 극복하기 위해 가상현실 (VR) 저작도구를 이용한 발전소 현장 내부를 표현하는 가상발전소 (Virtual Plant) 발전소 현장에 소재하여 기존 시뮬레이터의 모의한계 밖에 있던 패널을 표현한 가상판넬(Virtual Panel)등과 강의실에서 발전소 모의 훈련을 가능케 하기 위해 가상현실 기술을 이용한 컴퓨터 지원 교육훈력 시스템(CATS ; Computer Assister Training System)을 개발 중이며 일부 개발부분을 소개하였다.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.1
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• pp.47-52
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• 2011

In this paper, we describe the development of a finite-element model for seismic qualification. This paper presents finite-element analysis model of the electronic enclosure to be used at Arkansas nuclear power plant, USA. The verified model predicts natural frequencies within 5% error for all major modes below 50 Hz. The finite element lumped mass approach and the finite element stiffness approach using the COSMOSM finite element code is applied for static, eigenvalue, and dynamic analyses of the mathematical model of this system. The FEM model indicates that the stress levels corresponding to the specified loading conditions are below the allowable stress levels that have been specified in the AISC Code. The findings conclude that the electronic enclosure will withstand the seismic levels stated in the reference documents.