• Journal of the Earthquake Engineering Society of Korea
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• v.18 no.3
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• pp.151-160
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• 2014

Probabilistic tsunami hazard analysis (PTHA) is based on the approach of probabilistic seismic hazard analysis (PSHA) which is performed using various seismotectonic models and ground-motion prediction equations. The major difference between PTHA and PSHA is that PTHA requires the wave parameters of tsunami. The wave parameters can be estimated from tsunami propagation analysis. Therefore, a tsunami simulation analysis was conducted for the purpose of evaluating the wave parameters required for the PTHA of Uljin nuclear power plant (NPP) site. The tsunamigenic fault sources in the western part of Japan were chosen for the analysis. The wave heights for 80 rupture scenarios were numerically simulated. The synthetic tsunami waveforms were obtained around the Uljin NPP site. The results show that the wave heights are closely related with the location of the fault sources and the associated potential earthquake magnitudes. These wave parameters can be used as input data for the future PTHA study of the Uljin NPP site.

• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 2006.03a
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• pp.597-604
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• 2006

The seismic safety of the prestressed concrete containment building was evaluated by the seismic fragility analysis based on the nonlinear dynamic time-history analyses. Four kinds of earthquake ground motions were used for the seismic fragility analysis of the containment building to consider the potential earthquake hazard. The conventional seismic fragility analysis of the safety related structures in nuclear pouter plant have been performed by using the linear elastic analysis results for the seismic design. In this study, the displacement based seismic fragility analysis method was proposed.

• Journal of the Korean Society of Safety
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• v.31 no.6
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• pp.129-134
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• 2016

Standard design of APR+(advanced power reactor plus) was certified at 2014 by Korea regulatory body. Based on the experience gained from OPR1000 and APR1400, the APR1400 was being developed as a 1,500MWe class reactor using Korean technologies for design code, reactor coolant pump, and man-machine interface system. APR+ has been basically designed to have the seismic design basis of safe shutdown earthquake (SSE) 0.3g, a 4-train safety concept based on N+2 design philosophy, and a passive auxiliary feedwater system (PAFS). Also, safety issues on the Fukushima-type accidents have been extensively reviewed and applied to enhance APR+ safety. APR+ provides higher reliability and safety against tsunami and earthquake. The purpose of this paper is to implement probabilistic safety assessment considering these design features and to analyze sensitivity of core damage frequency for large loss of coolant accident of APR+.

• Journal of the Korean GEO-environmental Society
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• v.15 no.11
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• pp.67-75
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• 2014

In this study, the staged seismic performance evaluations were conducted to the 91 high speed railway tunnels in use for checking whether to comply with the recent design criteria or not. In addition, the seismic fragility functions of the tunnels were developed to allow the probabilistic risk assessment. The results of the staged seismic performance evaluations which consist of a preliminary assessment and a detailed assessment, show that the tunnels comply with the recent design criteria. With reference to the results of previous studies, a form of the proposed seismic fragility functions was set as a log-normal distribution by PGA, and the parameters of the functions were determined by using the probability of damage for the design PGA level. The seismic fragility functions were developed for each types (Cut & Cover, NATM) of tunnels. The seismic fragility functions from this study and the existing research results (FEMA, 2004) were compared to evaluate the seismic performance level of the tunnels, as a result the tunnels of this study were relatively superior to the ASSM tunnels on the seismic performance.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.14 no.5
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• pp.1097-1104
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• 1994

The external loads applied to a real structure may cause a severe damage and may eventually lead to total failure. It is thus the requirement that the structure must be designed to fulfil its safe function under any anticipated loads and must have the desired level of safety. The purpose of the present study is to propose a method of damage accumulation under seismic loadings to utilize it in the safety assessment of a reinforced concrete structure. To this end, the nonlinear hysteretic behavior of reinforced concrete structures is first modeled and the equivalent linearization technique is employed to solve numerically the probabilistic characteristics of response under random seismic loadings.

• Journal of the Korean Society of Safety
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• v.31 no.5
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• pp.95-101
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• 2016

This study proposes a multi-scale dynamic system reliability analysis of control system as a method of quantitative evaluation of its performance in probabilistic terms. In this second paper, we discuss the control effect of the viscous damper on the seismic performance of the structure-level failure. Since the failure of one structural member does not necessarily cause the collapse of the structural system, we need to consider a set of failure scenarios of the structural system and compute the sum of the failure probabilities of the failure scenarios where the statistical dependence between the failure scenarios should be taken into account. Therefore, this computation requires additional system reliability analysis. As a result, the proposed approach takes a hierarchial framework where the failure probability of a structural member is computed using a lower-scale system reliability with the union set of time-sequential member failures and their statistical dependence, and the failure probability of the structural system is again computed using a higher-scale system reliability with the member failure probabilities obtained by the lower-scale system reliability and their statistical dependence. Numerical results demonstrate that the proposed approach can provide an accurate and stable reliability assessment of the control performance of the viscous damper system on the system failure. Also, the parametric study of damper capacity on the seismic performance has been performed to demonstrate the applicability of the proposed approach through the probabilistic assessment of the seismic performance improvement of the damper system.

• 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.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.981-988
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• 2023

A seismic event caused an accident at the Fukushima Nuclear Power Plant, which further resulted in simultaneous accidents at several units. Consequently, this incident has aroused great interest in the safety of nuclear power plants worldwide. A reasonable safety evaluation of such an external event should appropriately consider the correlation between SSCs (structures, systems, and components) and the probability of failure. However, a probabilistic safety assessment in current nuclear industries is performed conservatively, assuming that the failure correlation between SSCs is independent or completely dependent. This is an extreme assumption; a reasonable risk can be calculated, or risk-based decision-making can be conducted only when the appropriate failure correlation between SSCs is considered. Thus, this study analyzed the effect of the failure correlation of SSCs on the safety of the system to realize rational safety assessment and decision-making. Consequently, the impact on the system differs according to the size of the failure probability of the SSCs and the AND and OR conditions.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.607-617
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• 2019

Global warming and climate change are increasing the intensity of typhoons and hurricanes and thus increasing the risk effects of typhoon and hurricane hazards on nuclear power plants (NPPs). To reflect these changes, a new NPP should be designed to endure design-basis hurricane wind speeds corresponding to an exceedance frequency of $10^{-7}/yr$. However, the short typhoon and hurricane observation records and uncertainties included in the inputs for an estimation cause significant uncertainty in the estimated wind speeds for return periods of longer than 100,000 years. A logic-tree framework is introduced to handle the epistemic uncertainty when estimating wind speeds. Three key parameters of a typhoon wind field model, i.e., the central pressure difference, pressure profile parameter, and radius to maximum wind, are used for constructing logic tree branches. The wind speeds of the simulated typhoons and the probable maximum wind speeds are estimated using Monte Carlo simulations, and wind hazard curves are derived as a function of the annual exceedance probability or return period. A logic tree decreases the epistemic uncertainty included in the wind intensity models and provides reasonably acceptable wind speeds.

• Journal of the Korean Society of Safety
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• v.36 no.5
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• pp.86-91
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• 2021

Calculating the scrutable core damage frequency (CDF) of nuclear power plants is an important component of the seismic probabilistic safety assessment (SPSA). In this work, a simple approach is developed to calculate CDF from minimal cut sets (MCSs) with non-rare events. When conventional calculation methods based on rare event approximations are employed, the CDF of industry SPSA models is significantly overestimated by non-rare events in the MCSs. Recently, quantification algorithms using binary decision diagrams (BDDs) have been introduced to prevent CDF overestimation in the SPSA. However, BDD structures are generated from a small part of whole MCSs due to limited computational memory, and they cannot be reviewed due to their complicated logic structure. This study suggests a simple approach for scrutinizing the CDF calculation based on whole MCSs in the SPSA system analysis model. The proposed approach compares the new results to outputs from existing algorithms, which helps in avoiding CDF overestimation.