• Journal of the Korean Society of Safety
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• v.31 no.5
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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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• v.55 no.6
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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.

• Journal of the Korea Safety Management & Science
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• v.10 no.4
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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.

• Journal of the Korean Society of Safety
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• v.33 no.6
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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.

• Journal of the Society of Naval Architects of Korea
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• v.38 no.1
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• pp.116-122
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• 2001

In this paper, to make a base of the fire safety assessment about ship's fire protection design and Classification Society rule, statistical informations and modeling techniques for the fire safety engineering are investigated and probabilistic safety assessment methods in the structural reliability engineering are introduced. FSEM(Fire Safety Evaluation Module) developed in this paper calculates the probability of fatality, which can be used as an index of fire safety. FSEM is used to calculate the probability of fatality of the evacuees in a small room installed according to the rules for fire-proof. Sensitivity analysis is executed to investigate FSEM's applicability to ship. From results, the necessity of new criterion for ship's fire safety design, the need to study the human behavior in the evacuation from fire, and the development of new fire progress model considering special situations in ships are acknowledged.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.37 no.3
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• pp.205-216
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• 2024

Recently, seismic Probabilistic Safety Assessment (PSA) methods have been developed for process plants, such as gas plants, oil refineries, and chemical plants. The framework originated from the PSA of nuclear power plants, which aims to assess the risk of reactor core damage. The original PSA method was modified to adopt the characteristics of a process plant whose purpose is continuous operation without shutdown. Therefore, a fault tree, whose top event is shut down, was constructed and transformed into a Bayesian Network (BN), a probabilistic graph model, for efficient risk-informed decision-making. In this research, the fault tree-based BN from the previous research is further developed to consider the multi-hazard of earthquake-induced fire and explosion (EQ-induced F&E). For this purpose, an event tree describing the occurrence of fire and explosion from a release is first constructed and transformed into a BN. And then, this BN is connected to the previous BN model developed for seismic PSA. A virtual plot plan of a gas plant is introduced as a basis for the construction of the specific EQ-induced F&E BN to test the proposed BN framework. The paper demonstrates the method through two examples of risk-informed decision-making. In particular, the second example verifies how the proposed method can establish a repair and retrofit strategy when a shutdown occurs in a process plant.

• Structural Engineering and Mechanics
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• v.74 no.4
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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.

• Journal of the Korean Society of Safety
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• v.35 no.1
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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.

• Journal of the Korean Society of Safety
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• v.32 no.3
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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.

• Journal of the Korean Society of Hazard Mitigation
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• v.9 no.6
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• pp.57-68
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• 2009

Recently, damage scale of human-induced disaster is sharply increased but its occurrences and damages are so uncertain that it is hard to construct a resonable response & mitigation plan for infrastructures. Therefore, the needs for a advanced risk management technique based on a probabilistic and stochastic risk evaluation theory is increased. In this study, these evaluation methods were investigated and a advanced disaster risk evaluation method, which is based on the probabilistic or stochastic risk assessment theory and also is a quantitative evaluation technique, was suggested. With this method, the safety changes as the result of fire damage management for recent 40 years was analyzed. And the result was compared with that of Japan. Through the consilience of the traditional risk assessment method and this method, a stochastical estimation technique for the uncertainty of future disaster's damage could support a cost-effective information for a resonable decision making on disaster mitigation.