• Nuclear Engineering and Technology
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• v.23 no.3
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• pp.285-298
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• 1991

An example application of the fuzzy set theory is first made to a simple portion of a given accident progression event tree with typical qualitative fuzzy input data, and thereby computational algorithms suitable for application of the fuzzy set theory to the accident progression event tree analysis are identified and illustrated with example applications. Then the procedure used in the simple example is extended to extremely complex accident progression event trees with a number of phenomenological uncertainty issues, i.e., a typical plant damage state‘SEC’of the Zion Nuclear Power Plant risk assessment. The results show that the fuzzy averages of the fuzzy outcomes are very close to the mean values obtained by current methods. The main purpose of this paper is to provide a formal procedure for application of the fuzzy set theory to accident progression event trees with imprecise and qualitative branch probabilities and/or with a number of phenomenological uncertainty issues.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2888-2898
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• 2021

Vital area identification (VAI) is an essential procedure for the design of physical protection systems (PPSs) for nuclear power plants (NPPs). The purpose of PPS design is to protect vital areas. VAI has been improved continuously to overcome the shortcomings of previous VAI generations. In first-generation VAI, a sabotage fault tree was developed directly without reusing probabilistic safety assessment (PSA) results or information. In second-generation VAI, VAI model was constructed from all PSA event trees and fault trees. While in third-generation VAI, it was developed from the simplified PSA event trees and fault trees. While VAIs have been performed for NPPs in full-power operations, VAI for NPPs in low-power and shutdown (LPSD) operations has not been studied and performed, even though NPPs in LPSD operations are very vulnerable to sabotage due to the very crowded nature of NPP maintenance. This study is the first to research and apply VAI to LPSD operation of NPP. Here, the third-generation VAI method for full-power operation of NPP was adapted to the VAI of LPSD operation. In this study, LPSD VAI for a few plant operational states (POSs) was performed. Furthermore, the operation strategy of vital areas for both full-power and LPSD operations was discussed. The LPSD VAI method discussed in this paper can be easily applied to all POSs. The method and insights in this study can be important for future LPSD VAI that reflects various LPSD operational states. Regulatory bodies and electric utilities can take advantage of this LPSD VAI method.

• Journal of the Korean Society for Railway
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• v.11 no.6
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• pp.530-535
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• 2008

After train fire accident in Daegue, many research on train fire safety improvement have been carrying out. Since many alternative fire safety measures can be applied in our railway system, the effect of the each safety measure must be quantified prior to the safety investment. In order to estimate the effects of each safety measure quantitatively, fault trees and event trees are constructed in this study. Results can be applied for cost-benefit analysis or sensitivity analysis for safety measures in risk assessment process.

• Journal of the Korean Society of Safety
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• v.31 no.3
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• pp.34-41
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• 2016

The purpose of this paper is to conduct ETA on six items of ESS: the whole system, battery, BMS, PCS, ESS and cable. To achieve that, ESS work flow and its components are categorized. Based on performance, human, environmental, management, and safety, this paper drew initiation events (IE) and end states (ES). ETA is applied to the main functions of each item, and the end states that may occur in one initiation event are suggested. In addition, detailed classification was performed to induce various end states on the basis of the suggested initiation events ; loss of grid electricity of ESS, loss of battery electricity(DC) of battery, impairment of electric function of BMS, loss of grid electricity(AC) of PCS, loss of data of EMS, Mechanical damage of cable, event sequence analysis conducted on the basis of event trees. If the suggested IEs and ESs are applied on the basis of ESS event cases, it is expected to prevent the same kinds of accident and operate ESS safely.

• The KIPS Transactions:PartA
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• v.9A no.1
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• pp.19-28
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• 2002

A new class of applications based on event interactions are emerging for the wide-area network such as Internet, which is characterized as loose coupling, heterogeneity. and asynchrony. Content-barred publish/subscribe systems are widely being studied to implement the extent notification service for wide-area networks. In this paper, we analyze some problems of the content-based routing algorithm in SIENA, a recently developed as a representative event notification service architecture, and develop an enhanced routing algorithm.

• Journal of Multimedia Information System
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• v.7 no.4
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• pp.263-268
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• 2020

This paper introduces an event modeling and simulation system using behavior trees. The system aims to overcome the limitations of existing fixed, simple scenario-based training content, and to extend the behavior of objects to enable various experience deployments. To achieve this goal, we made specific tasks of behavior trees can change according to users' reaction and developed an adaptive simulation module that can analyze and execute behavior trees that changes at runtime. In order to validate our approach, we applied the adaptive behavior tree simulation to the scenarios in our virtual reality simulation-based fire training system we have been developing and demonstrated the implementation results.

• Journal of Korean Society of Industrial and Systems Engineering
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• v.30 no.3
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• pp.37-43
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• 2007

We use influence diagrams to describe event trees used in safety analyses of low-probability high-risk incidents. This paper shows how the branch parameters used in the event tree models can be updated by a bayesian method based on the observed counts of certain well-defined subsets of accident sequences. We focus on the analysis of the shared branch parameters, which may frequently often in the real accident initiation and propagation to more severe accident. We also suggest the way to utilize different levels of accident data to forecast low-probability high-risk accidents.

• International Journal of Reliability and Applications
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• v.2 no.1
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• pp.27-36
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• 2001

This Paper proposes a new methodology for assessing the reliability of an accident management, which Is based on the reliability physics and the scheme to generate dynamic event tree. The methodology consists of 3 main steps: screening; uncertainty propagation; and probability estimation. Sensitivity analysis is used for screening the variables of significance. Latin Hypercube sampling technique and MAAP code are used for uncertainty propagation, and the dynamic event tree generation method is used for the estimation of non-success probability of implementing an accident management strategy. This approach is applied in assessing the non-success probability of implementing a cavity flooding strategy, which is to supply water into the reactor cavity using emergency fire systems during the sequence of station blackout at the reference plant.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.627-638
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• 2018

This article describes an integrated Level 1-Level 2 probabilistic risk assessment (PRA) methodology to evaluate the radiological risk during postulated accident scenarios initiated during the decommissioning phase of a typical Mark I containment boiling water reactor. The fuel damage scenarios include those initiated while the reactor is permanently shut down, defueled, and the spent fuel is located into the spent fuel storage pool. This article focuses on the integrated Level 1-Level 2 PRA aspects of the analysis, from the beginning of the accident to the radiological release into the environment. The integrated Level 1-Level 2 decommissioning PRA uses event trees and fault trees that assess the accident progression until and after fuel damage. Detailed deterministic severe accident analyses are performed to support the fault tree/event tree development and to provide source term information for the various pieces of the Level 1-Level 2 model. Source terms information is collected from accidents occurring in both the reactor pressure vessel and the spent fuel pool, including simultaneous accidents. The Level 1-Level 2 PRA model evaluates the temporal and physical changes in plant conditions including consideration of major uncertainties. The goal of this article is to provide a methodology framework to perform a decommissioning Probabilistic Risk Assessment (PRA), and an application to a real case study is provided to show the use of the methodology. Results will be derived from the integrated Level 1-Level 2 decommissioning PSA event tree in terms of fuel damage frequency, large release frequency, and large early release frequency, including uncertainties.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.110-116
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• 2022

Human failure event (HFE) dependency analysis is a part of human reliability analysis (HRA). For efficient HFE dependency analysis, a maximum number of minimal cut sets (MCSs) that have HFE combinations are generated from the fault trees for the probabilistic safety assessment (PSA) of nuclear power plants (NPPs). After collecting potential HFE combinations, dependency levels of subsequent HFEs on the preceding HFEs in each MCS are analyzed and assigned as conditional probabilities. Then, HFE recovery is performed to reflect these conditional probabilities in MCSs by modifying MCSs. Inappropriate HFE dependency analysis and HFE recovery might lead to an inaccurate core damage frequency (CDF). Using the above process, HFE recovery is performed on MCSs that are generated with a non-zero truncation limit, where many MCSs that have HFE combinations are truncated. As a result, the resultant CDF might be underestimated. In this paper, a new method is suggested to incorporate HFE recovery into the MCS generation stage. Compared to the current approach with a separate HFE recovery after MCS generation, this new method can (1) reduce the total time and burden for MCS generation and HFE recovery, (2) prevent the truncation of MCSs that have dependent HFEs, and (3) avoid CDF underestimation. This new method is a simple but very effective means of performing MCS generation and HFE recovery simultaneously and improving CDF accuracy. The effectiveness and strength of the new method are clearly demonstrated and discussed with fault trees and HFE combinations that have joint probabilities.