• Journal of the Korean Society of Safety
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• v.30 no.3
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• pp.100-106
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• 2015

A module or independent subtree is a part of a fault tree whose child gates or basic events are not repeated in the remaining part of the fault tree. Modules are necessarily employed in order to reduce the computational costs of fault tree quantification. This quantification generates fault tree solutions such as minimal cut sets, minimal path sets, or binary decision diagrams (BDDs), and then, calculates top event probability and importance measures. This paper presents a new linear time algorithm to detect modules of large fault trees. It is shown through benchmark tests that the new method proposed in this study can very quickly detect the modules of a huge fault tree. It is recommended that this method be implemented into fault tree solvers for efficient probabilistic safety assessment (PSA) of nuclear power plants.

• Tunnel and Underground Space
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• v.6 no.2
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• pp.166-174
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• 1996

Stability of rock slope is greatly affected by the geometry and strength of discontinuities developed in the rock mass. In this study an analytical method which is capable of analyzing the effect of relative orientation between the discontinuities and the slope face on the safety of slope by assessing their vector components was used to evaluate the stability and the maximum cut-angle for the proposed slope design. The results of computerized vector analysis revealed that slope area under investigation might be divided into 3 sections of different face directions. The safety factors for benches in each 3 sections were calculated using the limit-equilibrium theory. Then, by utilizing the concept of probabilistic risk analysis, the susceptibility of entire slope failure was estimated. Based on the distribution of safety factor in each bench, the maximum cut angle of each section could be selected differently ot achieve the permanent stability of the entire slope.

• Journal of the Korean Society for Railway
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• v.10 no.2 s.39
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• pp.131-136
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• 2007

In this paper, the methodology to predict the number of deaths and possible fire propagation scenarios will be described in case of fire on a train in a tunnel. We use a probabilistic analysis method for the evaluation of possibility for each scenario and the deaths tolls are calculated with the help of the passenger evacuation simulation program. The resulting safety of passengers is displayed on a F/N graph, which could be used in part as a guideline to predict the safety level of the tunnel in fire.

• Nuclear Engineering and Technology
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• v.19 no.4
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• pp.279-291
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• 1987

The event tree/fault tree techniques used in the current probabilistic risk assessment (PRA) of nuclear power plants are based on the binary and static description of the components and the system. While these techniques Bay be adequate in most of the safety studies, more advanced techniques, e.g., the Markov reliability analysis, are required to accurately study such problems as the plant availability assessments and technical specifications evaluations that are becoming increasingly important. This paper describes a Markov model for the Reactor Protection System of a pressurized water reactor and presents results of model evaluations for two testing policies in technical specifications.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.899-905
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• 1998

The analysis priority makings the recommendation to reduce the total core damage frequency (CDF) of Wolsong nuclear Power Plant nits 2/3/4 was Performed in this paper. In order to derive the recommendation, the sensitivity analysis of CDF on which major contributors effect m performed based on the accident quantification results during Level 1 Probabilistic safety assessment (PSA). Priorities were ranked in tile way that compares the CDF reduction rate with efforts required to implement those recommendations using risk matrix

• Journal of Korean Society of Industrial and Systems Engineering
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• v.3 no.3
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• pp.13-22
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• 1980

Research and Development has a property that involves uncertainity and risk in itself. Therefore, in order to scheduling of R & D activity, it Is needed of a certain probabilistic network technique with due regard to feedback process used to occur in the R & D proceeding. It is GERT that was developed as the need arises . In this study, the network structure of GERT-I and GERT-II was combined and then simulation analysis was used to it. According to that analysis , an advanced GERT model which covers the following stochastic problems was examined. 1 Evaluating success feasibility under the complex condition (time and cost). 2 Selecting acceptance range for the worst. 3. Selecting optimum path on basis of time, cost and success. 4. Evaluating project utility among the project alternatives. It is for managing R&D projects more effectively.

• Journal of the Society of Naval Architects of Korea
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• v.57 no.1
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• pp.1-7
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• 2020

In general, an inspection schedule is established based on the long-term fatigue life during the design stage. However, in the design stage, it is difficult to clearly identify the uncertainty factors affecting long-term fatigue life. In this study, the probabilistic fatigue life assessment was conducted in accordance with the methodology of DNV-GL. Firstly, The initial crack distribution estimated through the initial crack propagation analysis was updated by reflecting the results of crack inspection. Secondly, the updated crack distribution was compared with the initial crack distribution, and the probability of failure was updated with the effect of crack inspection.

• Journal of the Korea Society of Computer and Information
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• v.11 no.6 s.44
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• pp.261-267
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• 2006

In order to perform a probabilistic safety assessment (PSA), it requires a large number of data for various fields. And the quality of a PSA results have become more important thing of the risk assessment. As part of enhancing the PSA qualify, Korea Atomic Energy Research Institute is developing a full power Human Reliability Analysis (HRA) calculator to manage human failure events (HFEs) and to calculate the diagnosis human error probabilities and execution human error probabilities. This paper introduces the development process and an overview of a standard HRA method for nuclear power plants. The study was carried out in three stages; 1) development of the procedures and rules for a standard HRA method. 2) design of a system structure, 3) development of the HRA calculator.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.26 no.4A
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• pp.647-657
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• 2006

Probabilistic Risk Assessment considering statistically random variables is performed for the preliminary design of a Cable Stayed Bridge, which is Prestressed Concrete Bridge consisted of cable and plate girders, based on the method of Working Stress Design and Strength Design. Component reliabilities of cables and girders have been evaluated using the response surface of the design variables at the selected critical sections based on the maximum shear, positive and negative moment locations. Response Surface Method (RSM) is successfully applied for reliability analyses for this relatively small probability of failure of the complex structure, which is hard to obtain through Monte-Carlo Simulations. or through First Order Second Moment Method that can not easily calculate the derivative terms of implicit limit state functions. For the analysis of system reliability, parallel resistance system consisting of cables and plate girder is changed into series connection system and the result of system reliability of total structure is presented. As a system reliability, the upper and lower probabilities of failure for the structural system have been evaluated and compared with the suggested prediction method for the combination of failure modes. The suggested prediction method for the combination of failure modes reveals the unexpected combinations of element failures in significantly reduced time and efforts compared with the previous permutation method or system reliability analysis method, which calculates upper and lower bound failure probabilities.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.33 no.2
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• pp.129-136
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• 2020

In this study, we develop a sampling-based seismic probabilistic risk assessment (SPRA) quantification technique that can accurately consider a partially dependent condition of component seismic fragility information. Specifically, the SPRA quantification method is proposed by combining the advantages of two representative methodologies: EPRI seismic fragility and JAERI seismic fragility input-based quantification. The most important feature of the proposed method is that it performs a SPRA using a sampling technique by transforming the EPRI seismic fragility input into JAERI seismic fragility input. When the proposed sampling-based approach was applied to an example of simple system and to a SPRA problem of a nuclear power plant, it was observed that the proposed method yields approximately similar system seismic fragility and seismic risk results as those of the exact solution. Therefore, it is believed that the approach proposed in this study can be used as a useful tool for accurately assessing seismic risks, considering the partial seismic dependence among the components; the existing SPRA method cannot handle such partial dependencies.