• Proceedings of the KIEE Conference
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• 2007.07a
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• pp.1201-1202
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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 introduces the framework for the software lifecycle. The licensing procedure for the railway software is also reviewed.

• Journal of the Korea Institute of Information Security & Cryptology
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• v.23 no.2
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• pp.243-250
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• 2013

We have entered into an era of smart software system where the many kinds of embedded software, especially SCADA and Automotive software not only require high reliability and safety but also high-security. Removing software weakness during the software development lifecycle is very important because hackers exploit weaknesses which are source of software vulnerabilities when attacking a system. Therefore the coding rule as like core functions of MISRA-C should expand their coding focus on security. In this paper, we used CERT-C secure coding rules for nuclear-related software being developed to demonstrate high-safety software, and proposed how to remove software weakness during development.

• Nuclear Engineering and Technology
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• v.26 no.4
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• pp.502-506
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• 1994

The design of safety computer system is described along with the case of software design and testing in the Core Protection Calculator System (CPCS). The application of computer system in safety system requires not only hardware qualification but thorough testing on software to verify its correctness and completeness. The testing on software for CPCS is performed by comparing the outputs of two versions of code. One is implemented in assembly language and the other is in Fortran. The testing is performed in sequencial and overlapping manner. Phase I test verifies that each software module is implemented correctly by executing every branch. Phase II test verifies that the integrated software is complete, meeting its requirements specification and also the integrated system meet its requirement and timing constraints. Through these testing, the Yonggwang Nuclear Power Plant Units (YGN) 3 and 4 CPCS software is verified to be correct and complete, and the integrated system is designed as in its requirements specification.

• KIISE Transactions on Computing Practices
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• v.23 no.6
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• pp.335-342
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• 2017

In the verification and validation procedures regarding the safety-critical software of nuclear power plants for the attainment of the requisite license from the regulatory body, it is difficult to judge the safety and dependability of the development, implementation, and validation activities through a simple reading and review of the documentation. Therefore, these activities, especially safety assurance activities, require systematic evaluation techniques to determine that software faults are acceptable level. In this study, a safety case methodology is applied in an assessment of the level and depth of the results of the development and validation of a manufacturer in its targeting of the bistable processor of a digital reactor protection system, and the evaluation results are analyzed. This study confirms the possibility of an effective supplementation of the existing safety demonstration method through the application of the employed safety case methodology.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.69-77
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• 2020

After the Fukushima Daiichi accidents, public concerns on nuclear safety and the corresponding burden of nuclear power plant licensees are increasing. In order to secure public trust and enhance the rationality of current safety regulation, we develop a risk-informed decision making (RIDM) framework for the Korean regulatory body. By analyzing all the regulatory activities for nuclear power plants in Korea, eight action items are selected for RIDM implementation, with appropriate procedures developed for each. For two items in particular - the accident sequence precursor analysis (ASPA) and the significance determination process (SDP) - two customized risk evaluation software has been developed for field inspectors and probabilistic safety assessment experts, respectively. The effectiveness of the proposed RIDM framework is demonstrated by applying the ASPA procedure to 35 unplanned scrams and the SDP to 24 findings from periodic inspections.

• Proceedings of the KIEE Conference
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• 2007.10a
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• pp.215-216
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• 2007

This paper presents a real-time scheduler for nuclear safety system. According to constraints and requirements of nuclear safety system, scheduler design analysis is done and algorithms are developed for implementation. Using DSP based hardware, a real-time scheduler is realized. Consequently, this paper shows the performance of periodical software through the monitoring program.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3787-3800
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• 2023

Reactivity Initiated Accident (RIA) scenarios require special attention using advanced simulation techniques due to their complexity and importance for nuclear power plant (NPP) safety. While the conservative approach has traditionally been used for safety analysis, it may lead to unrealistic results which calls for the use of best estimate plus uncertainty (BEPU) approach, especially with the current advances in computational power which makes the BEPU analysis feasible. In this work an Uncontrolled Control Element Assembly (CEA) Withdrawal at Full Power accident scenario is analyzed using the BEPU approach by loosely coupling the thermal hydraulics best-estimate system code (RELAP5/SCDAPSIM/MOD3.4) to the statistical analysis software (DAKOTA) using a Python interface. Results from the BEPU analysis indicate that a realistic treatment of the accident scenario yields a larger safety margin and is therefore encouraged for accident analysis as it may enable more economic and flexible operation.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.233-238
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• 1997

본 논문에서는 Safety-critical 소프트웨어를 위한 V'||'&'||'V 지침서(guideline) 개발 방법론을 제시한다. 즉, 기존의 산업계 표준인 IEEE Std-1012, IEEE Std-1059에서 논의되고 있는 개념을 근간으로 "독립성(independence)", "소프트웨어 안전성 분석(software safety analysis)", "COTS 평가(evaluation) 기준", "다른 보증(assurance) 조직들간의 관련성(relationship)" 등의 필수 안전 항목들을 추가하여 원전 안전성 시스템(NPP safety system)을 위한 V'||'&'||'V 지침서 개발 방법론을 제시하였다 제시된 방법론에는 V'||'&'||'V 지침서의 범위(scope), 승인기준(acceptance criteria) 부분인 지침서 프레임(guideline framework), V'||'&'||'V activities 및 methods 부분인 타스크(task) entrance 및 exit 기준(criteria), 리뷰 및 감사(review and audit), 테스팅 그리고 V'||'&'||'V material의 QA 레코드(records) 및 형상관리, 소프트웨어 검증 및 확인 계획서(Software Verification and Validation Plan : SVVP) 생성 등의 내용을 기술하고, Safety-critical 소프트웨어 V'||'&'||'V 방법론도 함께 제시하였다.

• KIPS Transactions on Computer and Communication Systems
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• v.5 no.9
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• pp.275-284
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• 2016

The software used in the nuclear safety field has been ensured through the development, validation, safety analysis, and quality assurance activities throughout the entire process life cycle from the planning phase to the installation phase. However, this evaluation through the development and validation process needs a lot of time and money, and there are limitations to ensure that the quality is improved enough. Therefore, the effort to calculate the reliability of the software continues for a quantitative evaluation instead of a qualitative evaluation. In this paper, we propose a reliability evaluation method for the software to be used for a specific operation of the digital controller in a nuclear power plant. After injecting weighted faults in the internal space of a developed controller and calculating the ability to detect the injected faults using diagnostic software, we can evaluate the software reliability of a digital controller in a nuclear power plant.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1635-1643
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• 2022

STPA (System-Theoretic Process Analysis) is a widely used safety analysis technique to identify UCAs (Unsafe Control Actions) resulting in potential losses. It is totally dependent on the experience and ability of analysts to construct an information model called Control Structures, upon which analysts try to identify unsafe controls between system components. This paper proposes a formal approach to support the manual identification of UCAs, effectively and systematically. It allows analysts to mechanically extract Process Model, an important element that makes up the Control Structures, from a formal requirements specification for a software controller. It then concisely constructs the contents of Context Tables, from which analysts can identify all relevant UCAs effectively, using a software fault tree analysis technique. The case study with a preliminary version of a Korean nuclear reactor protections system shows the proposed approach's effectiveness and applicability.