• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1289-1296
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• 2019

Non-functional requirements plays a critical role in designing variety of applications domain ranges from safety-critical systems to simple gaming applications. Performance is one of the crucial non-functional requirement, especially in control and safety systems, that validates the design. System risk can be quantified as a product of probability of system failure and severity of its impact. In this paper, we devise a technique to do the performance analysis of safety critical and control systems and to estimate performance based risk factor. The technique elaborates Petri nets to estimate performability to ensure system dependability requirements. We illustrate the technique on a case study of Nuclear Power Plant system. The technique has been validated on 17 safety critical and control systems of Nuclear Power Plant.

• 대한전기학회논문지:전기기기및에너지변환시스템부문B
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• 제54권6호
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• pp.261-268
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• 2005

This paper propose the safety design of automatic train control system which is used for controlling and monitoring train speed not to excess a permitted speed. Safety activities are shown for the computerized system to achieve a required safety requirement. The safety activities are composed of system dynamic modelling to identify potential hazards contained in the target system, to analyze sub system faults to provoke the hazards. Risks analysis are carried out to estimate losses caused from the hazards to allocate safety requirement. We Proposed design solutions for sub system to meet safety requirement.

• Safety and Health at Work
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• 제8권3호
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• pp.267-275
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• 2017

Background: The mining industry is known worldwide for its highly risky and hazardous working environment. Technological advancement in ore extraction techniques for proliferation of production levels has caused further concern for safety in this industry. Research so far in the area of safety has revealed that the majority of incidents in hazardous industry take place because of human error, the control of which would enhance safety levels in working sites to a considerable extent. Methods: The present work focuses upon the analysis of human factors such as unsafe acts, preconditions for unsafe acts, unsafe leadership, and organizational influences. A modified human factor analysis and classification system (HFACS) was adopted and an accident predictive fuzzy reasoning approach (FRA)-based system was developed to predict the likelihood of accidents for manganese mines in India, using analysis of factors such as age, experience of worker, shift of work, etc. Results: The outcome of the analysis indicated that skill-based errors are most critical and require immediate attention for mitigation. The FRA-based accident prediction system developed gives an outcome as an indicative risk score associated with the identified accident-prone situation, based upon which a suitable plan for mitigation can be developed. Conclusion: Unsafe acts of the worker are the most critical human factors identified to be controlled on priority basis. A significant association of factors (namely age, experience of the worker, and shift of work) with unsafe acts performed by the operator is identified based upon which the FRA-based accident prediction model is proposed.

• 항공우주기술
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• 제7권2호
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• pp.123-130
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• 2008

비행안전분석은 우주발사체 발사운용시의 안전보장을 위하여 반드시 수행되어야 하며, 여러 가지 비정상 비행상황에 대한 위험분석을 포함한다. 비행안전 분석시스템은 요구되는 다양한 반복적 계산과 결과 생성 및 도시화를 한 시스템 내에서 처리할 수 있도록 하여 사용자 편의성을 증가시키도록 개발되었으며, 또한 삼차원 지구면 상에 결과를 표시하도록하여 보다 실질적인 결과 표현이 가능하도록 하였다. 개발된 비행안전 분석시스템은 소형위성발사체 발사운용을 위해 사용된다.

• 중소기업융합학회논문지
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• 제2권1호
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• pp.11-18
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• 2012

소프트웨어의 규모가 커지고 복잡해지면서 소프트웨어의 기능적 실패(Functional Failure)를 만들어 내는 위험(Hazard) 요소들을 분석하기가 어려워지고 있다. 안전 필수 시스템(원자력 발전소, 항공관제 시스템, 철도 운영 시스템)에서 이런 기능적 실패는 곧 큰 사고 (인명피해, 환경오염) 로 이어지게 된다. 따라서 이러한 기능적 실패를 방지하고 소프트웨어의 안전성을 높이기 위해서는 소프트웨어 안전성 분석이 필요하다. 하지만 몇 가지 이유 (시간과 노력, 안전성 분석 기법의 다양한 지식 부족, 기관이나 회사의 관습적인 방법 사용) 로 적절하지 못한 안전성 분석 기법을 선택하게 되는 경우가 있다. 따라서 본 논문에서는 기존 안전성 분석 기법과 최신 안전성 분석 기법, 통합 모델 몇 가지를 소개 하겠다. 이것을 통해 전문가는 여러 종류의 안전성 기법을 간략하게 확인 할 수 있을 것이고, 시스템에 맞는 안전성 분석 기법을 선택하는데 도움을 받을 수 있을 것이다.

• 한국안전학회지
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• 제34권5호
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• pp.159-166
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• 2019

The seismic PSA is to probabilistically estimate the potential damage that a large earthquake will cause to a nuclear power plant. It integrates the probabilistic seismic hazard analysis, seismic fragility analysis, and system analysis and is utilized to identify seismic vulnerability and improve seismic capacity of nuclear power plants. Recently, the seismic risk of domestic multi-unit nuclear power plant sites has been evaluated after the Great East Japan Earthquake and Gyeongju Earthquake in Korea. However, while the currently available methods for system analysis can derive basic required results of seismic PSA, they do not provide the detailed results required for the efficient improvement of seismic capacity. Therefore, for in-depth seismic risk evaluation, improved system analysis method for seismic PSA has become necessary. This study develops a system analysis method that is not only suitable for multi-unit seismic PSA but also provides risk information for the seismic capacity improvements. It will also contribute to the enhancement of the safety of nuclear power plants by identifying the seismic vulnerability using the detailed results of seismic PSA. In addition, this system analysis method can be applied to other external event PSAs, such as fire PSA and tsunami PSA, which require similar analysis.

• 한국산업안전학회:학술대회논문집
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• 한국안전학회 1997년도 추계 학술논문발표회 논문집
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• pp.117-122
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• 1997

This paper presents a study on the fuzzy based approach for the safety assessment of human body under ELF electric and magnetic(EM) field considering power system states. The analysis of ELF EM field based on quasi-static method is introduced. UP to the present, the analysis of ELF EM field has been conducted with the consideration of one transmission line, or a power line model only In this paper, however, the power system is included to model the expected and/or unexpected uncertainty caused by the load fluctuation and parameter changes and the states are classified into two types, normal state resulting from normal operation and emergency state from outages. In order to analyze the uncertainty in the normal state, the Monte Carlo Simulation, a statistic approach was introduced and line current and bus voltage distribution are calculated by a contingency analysis method, in the emergency state. To access the safety of human body, the approach based on fuzzy linguistic variable is adopted to overcome the shortcomings of the assessment by a crisp set concept. In order to validate the usefulness of the approach suggested herein, the case study using a sample system with 765(kV) was done. The results are presented and discussed.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.498-508
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• 2021

Micro reactors are increasingly being considered for utilization as distributed power sources. Hence, the probabilistic safety assessment (PSA) of a direct supercritical-CO₂-cooled fast reactor, called micro modular reactor (MMR), was performed in this study; this reactor was developed using innovative design concepts. It adopted a modular design and passive safety systems to minimize site constraints. As the MMR is in its conceptual design phase, design weaknesses and valuable safety insights could be identified during PSA. Level 1 internal event PSA was carried out involving literature survey, system characterization, identification of initiating events, transient analyses, development of event trees and fault trees, and quantification. The initiating events and scenarios significantly contributing to core damage frequency (CDF) were determined to identify design weaknesses in MMR. The most significant initiating event category contributing to CDF was the transients with the power conversion system initially available category, owing to its relatively high occurrence frequency. Further, an importance analysis revealed that the safety of MMR can be significantly improved by improving the reliability of reactor trip and passive decay heat removal system operation. The findings presented in this paper are expected to contribute toward future applications of PSA for assessing unconventional nuclear reactors in their conceptual design phases.

• 한국전자통신학회논문지
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• 제10권10호
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• pp.1079-1086
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• 2015

본 논문은 안전 무결성 레벨이 4(: Safety Integrity Level 4)인 안전 관련 시스템에 적합한 출력 모듈의 구조를 설계하였고, 설계된 출력 모듈에 요구되는 정량적 목표치의 수준을 제시하였다. 특히 다양한 출력 방식 중에서 릴레이 출력 신호와 아날로그 신호로 구성된 출력 모듈이 안전 기능을 수행하는 시스템의 구성 요소인 경우를 적용하였다. 분석 방법은 FMEA(: Failure Modes and Effect Analysis), FTA(: Fault Tree Analysis) 방식을 이용하였다. 그 결과로 철도 응용 규격인 IEC 62425의 요구사항을 준수하는 SIL4 안전 관련 시스템의 출력 모듈의 구조와 고장빈도인 정량적 목표치를 제시하였다.

• 한국안전학회지
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• 제28권3호
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• pp.123-128
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• 2013

The Advanced Power Reactor Plus(APR+), which is a GEN III+ reactor based on the APR1400, is being developed in Korea. In order to enhance the safety of the APR+, a passive auxiliary feedwater system(PAFS) has been adopted in the APR+. The PAFS replaces the conventional active auxiliary feedwater system(AFWS) by introducing a natural driving force mechanism while maintaining the system function of cooling the primary side and removing the decay heat. As the PAFS completely replaces the conventional AFWS, it is required to verify the cooling capacity of PAFS for the core damage frequency(CDF) evaluation. For this reason, this paper discusses the cooling performance of the PAFS during transient accidents. The test case and scenarios were picked from the result of the sensitivity analysis in APR+ Probabilistic Safety Assessment(PSA). The analysis was performed by the best estimate thermal-hydraulic code, RELAP5/.MOD3.3. This study shows that the plant maintains the stable state without the core damages under the given test scenarios. The results of PSA considering this analysis' results shows that the CDF values are decreased. The analysis results can be used for more realistic and accurate performance of a PSA.