• Structural Engineering and Mechanics
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• v.37 no.3
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• pp.257-265
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• 2011

Plant-specific analyses of 5 types of domestic reactors in Korea are performed to assure the structural integrity of the reactor pressure vessel (RPV) during transients which are expected to initiate pressurized thermal shock (PTS) events. The failure probability of the RPV due to PTS is obtained by performing probabilistic fracture mechanics analysis. The through-wall cracking frequency is calculated and compared to the acceptance criterion. Considering the fluence at the end of life expected by surveillance test, the sufficient safety margin is expected for the structural integrity of all reactor pressure vessels except for the oldest one during the pressurized thermal shock events. If the flaw with aspect ratio of 1/12 is considered to eliminate the conservatism, the acceptance criteria is not exceeded for all plants until the fluence level of $8{\times}10^{19}\;n/cm^2$, generating sufficient margin beyond the design life.

• 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.

• Journal of the Korea Society of Computer and Information
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• v.26 no.8
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• pp.139-147
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• 2021

In this study, we proposed the possibility of reliability analysis of risk index by using Monte Carlo simulation as basic data of safety accident occurrence data and expert opinion by construction industry type. Through this, it is expected that risk index for safety accidents according to detailed types of works will be presented stochastically and it will be possible to predict the risk factors and the expected range of damage based on the reliability analysis in the construction safety management plan. It will also reduce many of the planning risks that are common to decision makers in the field of construction management. In identifying risks, road bridge construction was classified into earthworks, drainage works, and bridge construction, and possible safety accidents were classified based on expert data. The risk index was calculated for each detailed construction of road and bridge construction, drainage construction, and bridge construction.

• Journal of Navigation and Port Research
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• v.33 no.4
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• pp.255-261
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• 2009

Prediction of the stability for ships is very complex in reality. In this paper, risk based approach is applied to predict the probability of capsize for a certified ship, which is effected by the forces of sea especially the wave loading Safety assessment and risk analysis process are also applied for the probabilistic prediction of stability for ships. The probability of shipsencountering different waves at sea is calculated by the existed statistics data and risk based models. Finally, ship capsizing probability is calculated according to single degree of freedom(SDF) rolling differential equation and basin erosion theory of nonlinear dynamics. Calculation results show that the survival probabilities of ship excited by the forces of the seas, especially in the beam seas status, can be predicted by the risk based method.

• Nuclear Engineering and Technology
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• v.34 no.6
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• pp.586-595
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• 2002

The Seismic probabilistic risk assessment (SPRA) or seismic margin assessment (SMA) have been used for the seismic safety evaluation of nuclear power plant structures and equipments. For the SPRA or SMA, the reference response spectrum should be defined. The site-specific median spectrum has been generally used for the seismic fragility analysis of structures and equipments in a Korean nuclear power plant Since the site-specific spectrum has been developed based on the peak ground motion parameter, the site-specific response spectrum does not represent the same probability of exceedance over the entire frequency range of interest. The uniform hazard spectrum is more appropriate to be used in seismic probabilistic risk assessment than the site- specific spectrum. A method for modifying the seismic fragility parameters that are calculated based on the site-specific median spectrum is described. This simple method was developed to incorporate the effects of the uniform hazard spectrum. The seismic fragility parameters of typical NPP components are modified using the uniform hazard spectrum. The modification factor is used to modify the original fragility parameters. An example uniform hazard spectrum is developed using the available seismic hazard data for the Korean nuclear power plant (NPP) site. This uniform hazard spectrum is used for the modification of fragility parameters.

• Nuclear Engineering and Technology
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• v.37 no.2
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• pp.191-200
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• 2005

Shaking table tests of the seismic behavior of a steel frame structure model were performed. The purpose of these tests was to estimate the effects of a near-fault ground motion and a scenario earthquake based on a probabilistic seismic hazard analysis for nuclear power plant structures. Three representative kinds of earthquake ground motions were used for the input motions: the design earthquake ground motion for the Korean nuclear power plants, the scenario earthquakes for Korean nuclear power plant sites, and the near-fault earthquake record from the Chi-Chi earthquake. The probability-based scenario earthquakes were developed for the Korean nuclear power plant sites using the PSHA data. A 4-story steel frame structure was fabricated to perform the tests. Test results showed that the high frequency ground motions of the scenario earthquake did not damage the structure at the nuclear power plant site; however, the ground motions had a serious effect on the equipment installed on the high floors of the building. This shows that the design earthquake is not conservative enough to demonstrate the actual danger to safety related nuclear power plant equipment.

• Smart Structures and Systems
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• v.22 no.4
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• pp.441-457
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• 2018

Characteristic parameter values of seismic isolators deviate from their nominal design values due to uncertainties and/or errors in their material properties and element dimensions, etc. Deviations may increase over service life due to environmental effects and service conditions. For accurate evaluation of the seismic safety level, all such effects, which would result in deviations in the structural response, need to be taken into account. In this study, the sensitivity of the probability of failure of the structures equipped with nonlinear base isolation systems to the uncertainties in various isolation system characteristic parameters is investigated in terms of various isolation system and superstructure response parameters in the context of a realistic three-dimensional base-isolated building model via Monte Carlo Simulations. The inherent record-to-record variability nature of the earthquake ground motions is also taken into account by carrying out analyses for a large number of ground motion records which are classified as those with and without forward-directivity effects. Two levels of nominal isolation periods each with three different levels of uncertainty are considered. Comparative plots of cumulative distribution functions and related statistical evaluation presented here portray the potential extent of the deviation of the structural response parameters resulting from the uncertainties and the uncertainty levels considered, which is expected to be useful for practicing engineers in evaluating isolator test results for their projects.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.386-403
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• 2016

The safety of nuclear power plants is analyzed by a probabilistic risk assessment, and the fault tree analysis is the most widely used method for a risk assessment with the event tree analysis. One of the well-known disadvantages of the fault tree is that drawing a fault tree for a complex system is a very cumbersome task. Thus, several graphical modeling methods have been proposed for the convenient and intuitive modeling of complex systems. In this paper, the reliability graph with general gates (RGGG) method, one of the intuitive graphical modeling methods based on Bayesian networks, is improved for the reliability analyses of dynamic systems that have various operation modes with time. A reliability matrix is proposed and it is explained how to utilize the reliability matrix in the RGGG for various cases of operation mode changes. The proposed RGGG with a reliability matrix provides a convenient and intuitive modeling of various operation modes of complex systems, and can also be utilized with dynamic nodes that analyze the failure sequences of subcomponents. The combinatorial use of a reliability matrix with dynamic nodes is illustrated through an application to a shutdown cooling system in a nuclear power plant.

• International journal of steel structures
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• v.18 no.4
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• pp.1397-1409
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• 2018

Stiffened plates with high slenderness parameters show large out-of-plane deflections, due to elastic buckling, which may occur before the plates reach their ultimate strength. From a serviceability point of view, restriction of out-of-plane deflections exceeding the fabrication tolerance is of primary importance. Compressive strength at the serviceability limit state (SLS) for slender stiffened plates under uniaxial stress was investigated through nonlinear elasto-plastic finite element analysis, considering both geometric and material nonlinearity. Both normal and high-performance steel were considered in the study. The SLS was defined based on a deflection limit and an elastic buckling strength. Probabilistic distributions of the SLS strengths were obtained through Monte Carlo simulations, in association with the response surface method. On the basis of the obtained statistical distributions, partial safety factors were proposed for SLS. Comparisons with the ultimate strength of different design codes e.g. Japanese Code, AASHTO, and Canadian Code indicate that AASHTO and Canadian Code provide significantly conservative design, while Japanese Code matches well with a 5% non-exceedance probability for compressive strength at SLS.

• Journal of the Korean Society of Safety
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• v.27 no.5
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• pp.224-228
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• 2012

According to the results of Probabilistic Safety Assessment(PSA) for a Nuclear Power Plant(NPP), an Emergency Power Supply(EPS) system has been considered as one of the most important safety system. Especially, the interests in the reliability of the EPS system have been increased after the severe accidents of Fukushima Daiichi. Firstly, we performed the risk assessment and the importance analysis of the EPS system based on the PSA models of the reference plant, which is the Korean standard NPP type. Considering a portable Diesel Generator(DG) system as the reliability reinforcement of the EPS system, we modified the PSA models and performed the risk impact assessment and the importance analysis. Although the reliability of the potable DG could be about 20% of the reliability of the alternative AC DG, we identified that Core Damage Frequency(CDF) was decreased by at least 4.6%. In addition, the risk impacts due to the unavailability of the EPS system on CDF were decreased.