• 한국안전학회지
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• 제31권5호
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• pp.187-194
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• 2016

A single fire event within a fire area can cause multiple initiating events considered in internal events probabilistic safety assessment (PSA). For an example, a fire event in turbine building fire area can cause a loss of the main feed-water and loss of off-site power initiating events. This fire initiating event could result in special plant responses beyond the scope of the internal events PSA model. One approach to address a fire initiating event is to develop a specific fire event tree. However, the development of a specific fire event tree is difficult since the number of fire event trees may be several hundreds or more. Thus, internal fire events PSA model has been generally constructed by modifications of the pre-developed internal events PSA model. New accident sequence logics not covered in the internal events PSA model are separately developed to incorporate them into the fire PSA model. Recently, many fire PSA models have fire induced initiating event fault trees not shown in an internal event PSA model. Up to now, there has been no analytical comparative study on the constructions of fire events PSA model using internal events PSA model with and without fault trees of initiating events. In this study, the changing process of internal events PSA model to fire events PSA model is analytically presented and discussed.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2263-2275
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• 2023

In general, an internal fire events probabilistic safety assessment (PSA) model is quantified by modifying the pre-existing internal event PSA model. Because many pieces of equipment or cables can be damaged by a fire, a single fire event can lead to multiple internal events PSA initiating events (IEs). Consequently, when the fire events PSA model is quantified, inappropriate minimal cut sets (MCSs), such as duplicate MCSs, may be generated. This paper shows that single quantification of a hypothetical single-top fire event PSA model may generate the following four types of inappropriate MCSs: duplicate MCSs, MCSs subsumed by other MCSs, nonsense MCSs, and MCSs with over-counted fire frequencies. Among the inappropriate MCSs, the nonsense MCSs should be addressed first because they can interfere with the right interpretation of the other MCSs and prevent the resolution of the issues related to the other inappropriate MCSs. In addition, we propose a resolution process for each of the issues caused by these inappropriate MCSs and suggest an overall procedure for resolving them. The results of this study will contribute to the understanding and resolution of the inappropriate MCSs that may appear in the quantification of fire events PSA models.

• 대한안전경영과학회지
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• 제10권4호
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• pp.93-104
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• 2008

A successful fire risk assessment is depends on identification of risk, the analytical process of potential risk, on estimation of likelihood and the width and depth of consequence. Take the influence on enterprise into consideration, Fire risk assessment could carry out along the evaluation of the risk importance, the risk level and the risk acceptance. A large part of the limitation of choosing the risk assessment techniques impose restrictions on expense and time. If it is unnecessary high level risk assessment or Probabilistic Risk Assessment of buildings, in compliance with the Relative Ranking Method, Fire risk indexing and assessing is possible. As working-level technique, AHP method is useful with practical technique.

• 한국안전학회지
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• 제33권6호
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• pp.136-143
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• 2018

In this study, we conducted a pilot study on the multiple spurious operations (MSO) analysis in the fire probabilistic safety assessment (PSA) of domestic nuclear power plant (NPP) to identify the degree of influence of the operator actions used in the MSO mitigation strategies. The MSO scenario of the domestic reference NPP selected for this study is refueling water tank (RWT) drain down event. It could be caused by spurious operations of the containment spray system (CSS) of the reference NPP. The RWT drain down event can be stopped by the main control room (MCR) operator actions for stopping the operation of CSS pump or closing the CSS motor operated valve if the containment spray actuation signal (CSAS) is spuriously actuated. Outside the MCR, it can be stopped by operator actions for closing the CSS manual valves or motor operated valve or stopping the operation of CSS pump. The quantification result of a fire PSA model that takes into account all recovery actions for the RWT drain down event lead to risk reduction by about 95%, compared with quantification result of fire PSA model without considering them. Among the various operator actions, the recovery action for the spurious CSAS operations and the operator action for the manual valve are identified as the most important operator actions. This study quantitatively showed the extent to which the operator actions used as MSO countermeasures have affected the fire PSA quantification results. In addition, we can see the rank of importance among the operator recovery actions in quantitative terms.

• 대한조선학회논문집
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• 제38권1호
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• pp.116-122
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• 2001

본 논문에서는 화재안전공학에서 사용되는 모델링 기법과 통계자료들을 검토하고 구조 신뢰성 분야의 확률론적 안전도 평가방법을 도입하여 선박에 대한 방화설계나 선급 규정을 정량적으로 평가하는 방법을 마련하였다. 본 논문에서 개발된 FSEM(Fire Safety Evaluation Module)은 화재안전도의 척도로서 마지막으로 대피하는 사람의 사망확률을 산정하는 프로그램이며, 이를 이용해서 규정에 의한 방화설비를 갖춘 소규모의 화재실에서 대피자에 대한 사망확률을 추정해 보았고, 민감도 해석을 통해 선박에 대한 정량적인 화재안전 평가에의 적용가능성을 검토해 보았다. 계산결과와 검토로부터 선박화재에 대한 새로운 설계지표와 대피상황에서의 인간행동에 대한 폭넓은 연구의 필요성 및 선박의 특수 상황을 고려할 수 있는 새로운 화재진척 모델에 대한 개발이 수행되어야 함을 제시하였다.

• Structural Engineering and Mechanics
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• 제74권4호
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• pp.481-494
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• 2020

In this paper, a probability-based procedure to evaluate the performance of existing RC structures exposed to seismic and fire actions is presented. The procedure is demonstrated with reference to an existing old school building, located in Italy. The vulnerability assessment of the building highlights deficiencies under both static and seismic loads. Retrofit operations are designed to achieve the seismic safety. The idea of the work consists in assessing the performance of the existing and retrofitted building in terms of both the seismic and fire resistance. The seismic retrofit and fire resistance upgrading follow different paths, depending on the specific configuration of the building. A good seismic retrofit does not entail an improving of the fire resistance and vice versa. The goal of the current work is to study the variation of response due to the uncertainties considered in records/fire curves selection and to carry out the assessment of the studied RC structure by obtaining fragility curves under the effect of different records/temperature. The results show the fragility curves before and after retrofit operations and both in terms of seismic performance and fire resistance performance, measuring the percent improving for the different limit states.

• 한국안전학회지
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• 제35권1호
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• pp.87-96
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• 2020

The purpose of this paper is to develop a fire HRA (Human Reliability Analysis) procedure for full power operation of domestic NPPs (Nuclear Power Plants). For the development of fire HRA procedure, the recent research results of NUREG-1921 in an effort to meet the requirements of the ASME/ANS PRA Standard were reviewed. The K-HRA method, a standard method for HRA of a domestic level 1 PSA (Probabilistic Safety Assessment) and fire related procedures in domestic NPPs were reviewed. Based on the review, a procedure for the fire HRA required for a domestic fire PSA based on the K-HRA method was developed. To this end, HRA issues such as new operator actions required in the event of a fire and complexity of fire situations were considered. Based on the four kinds of HFE (Human Failure Event) developed for a fire HRA in this research, a qualitative analysis such as feasibility evaluation was suggested. And also a quantitative analysis process which consists of screening analysis and detailed analysis was proposed. For the qualitative analysis, a screening analysis by NUREG-1921 was used. In this research, the screening criteria for the screening analysis was modified to reduce vague description and to reflect recent experimental results. For a detailed analysis, the K-HRA method and scoping analysis by NUREG-1921 were adopted. To apply K-HRA to fire HRA for quantification, efforts to modify PSFs (Performance Shaping Factors) of K-HRA to reflect fire situation and effects were made. For example, an absence of STA (Shift Technical Advisor) to command a fire brigade at a fire area is considered and the absence time should be reflected for a HEP (Human Error Probability) quantification. Based on the fire HRA procedure developed in this paper, a case study for HEP quantification such as a screening analysis and detailed analysis with the modified K-HRA was performed. It is expected that the HRA procedure suggested in this paper will be utilized for fire PSA for domestic NPPs as it is the first attempt to establish an HRA process considering fire effects.

• 한국안전학회지
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• 제32권3호
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• pp.160-171
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• 2017

US national research laboratories developed the first Vital Area Identification (VAI) method for the physical protection of nuclear power plants that is based on Event Tree Analysis (ETA) and Fault Tree Analysis (FTA) techniques in 1970s. Then, Korea Atomic Energy Research Institute proposed advanced VAI method that takes advantage of fire and flooding Probabilistic Safety Assessment (PSA) results. In this study, in order to minimize the burden and difficulty of VAI, (1) a set of streamlined VAI rules were developed, and (2) this set of rules was applied to PSA fault tree and event tree at the initial stage of VAI process. This new rule-based VAI method is explained, and its efficiency and correctness are demonstrated throughout this paper. This new rule-based VAI method drastically reduces problem size by (1) performing PSA event tree simplification by applying VAI rules to the PSA event tree, (2) calculating preliminary prevention sets with event tree headings, (3) converting the shortest preliminary prevention set into a sabotage fault tree, and (4) performing usual VAI procedure. Since this new rule-based VAI method drastically reduces VAI problem size, it provides very quick and economical VAI procedure. In spite of an extremely reduced sabotage fault tree, this method generates identical vital areas to those by traditional VAI method. It is strongly recommended that this new rule-based VAI method be applied to the physical protection of nuclear power plants and other complex safety-critical systems such as chemical and military systems.

• 한국방재학회 논문집
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• 제9권6호
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• pp.57-68
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• 2009

최근 인적재난 발생의 불확실성에 대한 유연한 대처를 위해 확률 통계적 재난위험성 평가 및 위험관리 기술에 대한 필요성이 고조되고 있어 관련기술을 인적재난에 적용하기 위한 연구를 수행하였다. 먼저 재난위험성 평가 기법의 실효성, 경제성 및 지속가능한 시스템 구현을 위한 선제조건을 검토하였다. 이로부터 재난의 피해규모-발생확률 분포함수의 이론적 검토를 통해 확률 통계적인 재난위험 지표를 도출하고 재난안전(위험)도 평가에 활용함으로서 보다 간편한 정량적 재난위험도 평가기법을 개발하였기에 이를 소개한다. 또한 이를 활용하여 우리나라와 일본의 확률 통계적인 화재 안전유지 성능을 비교 분석하고 그 결과를 안전지수로 제시하였다. 향후 기존의 재난위험 평가기술과 융화 발전시켜 국내실정에 맞는 미래 재난 추정 및 예측 모델의 최적화 방안을 마련함으로써 지속적인 위험도 분석결과에 기반을 둔 합리적인 통합재난관리 방안 마련이 가능 할 것으로 기대된다.

• Earthquakes and Structures
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• 제19권6호
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• pp.471-484
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• 2020

Previous earthquakes show that, structural safety evaluations should include the evaluation of nonstructural components. Failure of nonstructural components can affect the operational capacity of critical facilities, such as hospitals and fire stations, which can cause an increase in number of deaths. Additionally, failure of nonstructural components may result in economic, architectural, and historical losses of community. Accelerations and random vibrations must be under the predefined limitations in structures with high technological equipment, data centers in this case. Failure of server equipment and anchored server racks are investigated in this study. A probabilistic study is completed for a low-rise rigid sample structure. The structure is investigated in two versions, (i) conventional fixed-based structure and (ii) with a base isolation system. Seismic hazard assessment is completed for the selected site. Monte Carlo simulations are generated with selected parameters. Uncertainties in both structural parameters and mechanical properties of isolation system are included in simulations. Anchorage failure and vibration failures are investigated. Different methods to generate fragility curves are used. The site-specific annual hazard curve is used to generate risk curves for two different structures. A risk matrix is proposed for the design of data centers. Results show that base isolation systems reduce the failure probability significantly in higher floors. It was also understood that, base isolation systems are highly sensitive to earthquake characteristics rather than variability in structural and mechanical properties, in terms of accelerations. Another outcome is that code-provided anchorage failure limitations are more vulnerable than the random vibration failure limitations of server equipment.