• Journal of the Korean Society for Precision Engineering
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• v.16 no.6
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• pp.100-107
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• 1999

This paper describes a probabilistic remaining life assessment program for the creep crack growth. The probabilistic life assessment program is developed to increase the reliability of life assessment. The probabilistic life assessment involves some uncertainties, such as, initial crack size, material properties, and loading condition, and a triangle distribution function is used for random variable generation. The resulting information provides the engineer with an assessment of the probability of structural failure as a function of operating time given the uncertainties in the input data. This study forms basis of the probabilistic life assessment technique and will be extended to other damage mechanisms.

• Earthquakes and Structures
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• v.6 no.6
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• pp.665-687
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• 2014

This paper presents a framework for analytical seismic vulnerability assessment of substandard reinforced concrete (RC) structures in developing countries. Amodified capacity-demand diagram method is used to predict the response of RC structures with degrading behaviour. A damage index based on period change is used to quantify the evolution of damage. To demonstrate the framework, a class of substandard RC buildings is examined. Abrupt accumulation of damage is observed due to the brittle failure modes and this is reflected in the developed vulnerability curves, which differ substantially from the curves of ductile structures.

• 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.

• Nuclear Engineering and Technology
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• v.41 no.10
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• pp.1235-1242
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• 2009

Probabilistic seismic hazard analysis (PSHA) is routinely conducted in the US for nuclear plants, for the determination of appropriate seismic design levels. These analyses incorporate uncertainties in earthquake characteristics in stable continental regions (where direct observations of large earthquakes are rare), in estimates of rock motions, in site effects on strong shaking, and in the damage potential of seismic shaking for engineered facilities. Performance goals related to the inelastic deformation of individual components, and related to overall seismic core damage frequency, are used to determine design levels. PSHA has the ability to quantify and document the important uncertainties that affect seismic design levels, and future work can be guided toward reducing those uncertainties.

• Smart Structures and Systems
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• v.15 no.4
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• pp.1121-1137
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• 2015

This study presents probabilistic-based damage identification technique for highlighting damage in metallic structures. This technique utilizes distributed piezoelectric transducers to generate and monitor the ultrasonic Lamb wave with narrowband frequency. Diagnostic signals were used to define the scatter signals of different paths. The energy of scatter signals till different times were calculated by taking root mean square of the scatter signals. For each pair of parallel paths an error function based on the energy of scatter signals is introduced. The resultant error function then is used to estimate the probability of the presence of damage in the monitoring area. The presented method with an autofocusing feature is applied to aluminum plates for method verification. The results identified using both simulation and experimental Lamb wave signals at different central frequencies agreed well with the actual situations, demonstrating the potential of the presented algorithm for identification of damage in metallic structures. An obvious merit of the presented technique is that in addition to damages located inside the region between transducers; those who are outside this region can also be monitored without any interpretation of signals. This novelty qualifies this method for online structural health monitoring.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.11 no.9
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• pp.3148-3154
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• 2010

This paper presents the probabilistic analysis for fatigue life of Glass/Epoxy laminates with impact-induced damage. For this, a series of impact tests were perfomed on the Glass/Epoxy laminates using instrumented impact testing machine. Then, tensile and fatigue tests carried out so as to generate post-impact residual strength and fatigue life. Two Parameter Weibull distribution was used to fit the residual strength and fatigue life data of Glass/Epoxy composite laminates. The residual strength was affected by impact energy and their variance decreased with increasing of impact energy. The fatigue life of impacted laminates was greatly reduced by impact energy and this trend depended on applied stress amplitude. Additionally, the variation of fatigue life was gradually decreased with the applied stress amplitude.

• Journal of the Korean Society of Safety
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• v.36 no.3
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• pp.66-73
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• 2021

This report documents the results of an at-power internal events Level 1 Probabilistic Safety Assessment (PSA) for a Korea research reactor (KRR). The aim of the study is to determine the accident sequences, construct an internal level 1 PSA model, and estimate the core damage frequency (CDF). The accident quantification is performed using the AIMS-PSA software version 1.2c along with a fault tree reliability evaluation expert (FTREX) quantification engine. The KRR PSA model is quantified using a cut-off value of 1.0E-15/yr to eliminate the non-effective minimal cut sets (MCSs). The final result indicates a point estimate of 4.55E-06/yr for the overall CDF attributable to internal initiating events in the core damage state for the KRR. Loss of Electric Power (LOEP) is the predominant contributor to the total CDF via a single initiating event (3.68E-6/yr), providing 80.9% of the CDF. The second largest contributor is the beam tube loss of coolant accident (LOCA), which accounts for 9.9% (4.49E-07/yr) of the CDF.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.211-221
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• 2020

One fundamental element of probabilistic safety assessment (PSA) is the initiating event (IE) analysis. Since IE frequencies can change over time, time-trend analysis is required to obtain optimized IE frequencies. Accordingly, such time-trend analyses have been employed to estimate industry-average IE frequencies for use in the PSAs of U.S. nuclear power plants (NPPs); existing PSAs of Korean NPPs, however, neglect such analysis in the estimation of IE frequencies. This article therefore provides the method for and results of estimating Korean industry-average IE frequencies using time-trend analysis. It also examines the effects of the IE frequencies obtained from this study on risk insights by applying them to recently updated internal events Level 1 PSA models (at-power and shutdown) for an OPR-1000 plant. As a result, at-power core damage frequency decreased while shutdown core damage frequency increased, with the related contributions from each IE category changing accordingly. These results imply that the incorporation of time-trend analysis leads to different IE frequencies and resulting risk insights. The IE frequency distributions presented in this study can be used in future PSA updates for Korean NPPs, and should be further updated themselves by adding more recent data.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2092-2098
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• 2024

Probabilistic safety assessment (PSA) plays a critical role in ensuring the safe operation of nuclear power plants. In PSA, event trees are developed to identify accident sequences that could lead to core damage. These event trees are then transformed into a core-damage fault tree, wherein the accident sequences are represented by usual and complemented logic gates representing failed and successful operations of safety systems, respectively. The core damage frequency (CDF) is estimated by calculating the minimal cut sets (MCSs) of the core-damage fault tree. Delete-term approximation (DTA) is commonly employed to approximately solve MCSs representing accident sequence logics from noncoherent core-damage fault trees. However, DTA can lead to an overestimation of CDF, particularly when fault trees contain many nonrare events. To address this issue, the present study introduces a new zero-suppressed ternary decision diagram (ZTDD) algorithm that averts the CDF overestimation caused by DTA. This ZTDD algorithm can optionally calculate MCSs with DTA or prime implicants (PIs) without any approximation from the core-damage fault tree. By calculating PIs, accurate CDF can be calculated. The present study provides a comprehensive explanation of the ZTDD structure, formula of the ZTDD algorithm, ZTDD minimization, probability calculation from ZTDD, strength of the ZTDD algorithm, and ZTDD application results. Results reveal that the ZTDD algorithm is a powerful tool that can quickly and accurately calculate CDF and drastically improve the safety of nuclear power plants.

• Journal of the Earthquake Engineering Society of Korea
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• v.25 no.2
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• pp.53-58
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• 2021

Nuclear power plant's safety against seismic events is evaluated as risk values by probabilistic seismic safety assessment. The risk values vary by the seismic failure correlation between the structures, systems, and components (SSCs). However, most probabilistic seismic safety assessments idealized the seismic failure correlation between the SSCs as entirely dependent or independent. Such a consideration results in an inaccurate assessment result not reflecting real physical phenomenon. A nuclear power plant's seismic risk should be calculated with the appropriate seismic failure correlation coefficient between the SSCs for a reasonable outcome. An accident scenario that has an enormous impact on a nuclear power plant's seismic risk was selected. Moreover, the probabilistic seismic response analyses of a nuclear power plant were performed to derive appropriate seismic failure correlations between SSCs. Based on the analysis results, the seismic failure correlation coefficient between SSCs was derived, and the seismic fragility curve and core damage frequency of the loss of essential power event were calculated. Results were compared with the seismic fragility and core damage frequency of assuming the seismic failure correlations between SSCs were independent and entirely dependent.