• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.423-434
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• 2016

The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.426-433
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• 2017

This paper describes the development process, the innovative techniques used and insights gained from the latest integrated, full scope, multistate Level 2 PSA analysis conducted at the Leibstadt Nuclear Power Plant (KKL), Switzerland. KKL is a modern single-unit General Electric Boiling Water Reactor (BWR/6) with Mark III Containment, and a power output of $3600MW_{th}/1200MW_e$, the highest among the five operating reactors in Switzerland. A Level 2 Probabilistic Safety Assessment (PSA) analyses accident phenomena in nuclear power plants, identifies ways in which radioactive releases from plants can occur and estimates release pathways, magnitude and frequency. This paper attempts to give an overview of the advanced modeling techniques that have been developed and implemented for the recent KKL Level 2 PSA update, with the aim of systematizing the analysis and modeling processes, as well as complying with the relatively prescriptive Swiss requirements for PSA. The analysis provides significant insights into the absolute and relative importances of risk contributors and accident prevention and mitigation measures. Thanks to several newly developed techniques and an integrated approach, the KKL Level 2 PSA report exhibits a high degree of reviewability and maintainability, and transparently highlights the most important risk contributors to Large Early Release Frequency (LERF) with respect to initiating events, components, operator actions or seismic component failure probabilities (fragilities).

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1297-1306
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• 2019

Probabilistic safety assessments (PSA) have been used for several decades to visualize the risk level of commercial nuclear power plants (NPPs). Since the role of a human reliability analysis (HRA) is to provide human error probabilities for safety critical tasks to support PSA, PSA quality is strongly affected by HRA quality. Therefore, it is important to understand the underlying limitations or problems of HRA techniques. For this reason, this study conducted a survey among 14 subject matter experts who represent the HRA community of domestic Korean NPPs. As a result, five significant HRA issues were identified: (1) providing a technical basis for the K-HRA (Korean HRA) method, and developing dedicated HRA methods applicable to (2) diverse external events to support Level 1 PSA, (3) digital environments, (4) mobile equipment, and (5) severe accident management guideline tasks to support Level 2 PSA. In addition, an HRA method to support multi-unit PSA was emphasized because it plays an important role in the evaluation of site risk, which is one of the hottest current issues. It is believed that creating such a catalog of prioritized issues will be a good indication of research direction to improve HRA and therefore PSA quality.

• Journal of the Korean Geotechnical Society
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• v.30 no.10
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• pp.5-18
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• 2014

Many regions around the world are vulnerable to rainfall-induced slope failures. A variety of methods have been proposed for revealing the mechanism of slope failure initiation. Current analysis methods, however, do not consider the effects of non-homogeneous soil profiles and variable hydraulic responses on rainfall-induced slope failures. In this study, probabilistic stability analyses were conducted for weathered residual soil slopes with different soil thickness overlying impermeable bedrock to study the rainfall-induced failure mechanisms depending on the soil thickness. A series of seepage and stability analyses of an infinite slope based on one-dimensional random fields were performed to consider the effects of uncertainty due to the spatial heterogeneity of hydraulic conductivity on the failure of unsaturated slopes due to rainfall infiltration. The results showed that a probabilistic framework can be used to efficiently consider various failure patterns caused by spatial variability of hydraulic conductivity in rainfall infiltration assessment for a infinite slope.

• Earthquakes and Structures
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• v.22 no.2
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• pp.169-184
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• 2022

The importance of seismicity in developing countries and the strengthening of buildings is a topic of major importance. Therefore, the study of several solutions with the development of new technologies is of great importance to investigate the damage on retrofitted structures by using probabilistic methods. The Federal Emergency Management Agency considers three types of performance levels by considering different scenarios, intensity and duration. The selection and scaling of ground motions mainly depends on the aim of the study. Intensity-based assessments are the most common and compute the response of buildings for a specified seismic intensity. Assessments based on scenarios estimate the response of buildings to different earthquake scenarios. A risk-based assessment is considered as one of the most effective. This research represents a practical method for developing countries where exists many active faults, tall buildings and lack of good implementable approaches. Therefore, to achieve the main goal, two high-rise steel buildings have been modeled and assessed. The contribution of buckling-restrained braces in the elastic design of both buildings is firstly verified. In the nonlinear static range, both buildings presented repairable damage at the central top part and some life safety hinges at the bottom. The nonlinear incremental dynamic analysis was applied by 15 representative/scaled accelerograms to obtain levels of performance and fragility curves. The results shown that by using probabilistic methods, it is possible to estimate the probability of collapse of retrofitted buildings by buckling-restrained braces and tuned mass dampers, which are practical retrofitting options to protect existing structures against earthquakes.

• Journal of the Earthquake Engineering Society of Korea
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• v.10 no.1 s.47
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• pp.63-74
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• 2006

The seismic fragility analysis method has been used as a quantitative seismic safety evaluation method for the NPP(Nuclear Power Plant) structures and equipments. The seismic fragility analysis gives a realistic seismic capacity excluding the convertism included in the design stage. The conservatism is considered as the probabilistic parameters related to the response and capacity in the seismic fragility analysis. In this study, the displacement based seismic fragility analysis method was proposed based on the nonlinear dynamic analysis results. In this study, the seismic safety of the prestressed concrete containment building of KSNP(Korean Standard Nuclear Power Plant) was evaluated for the scenario earthquakes, neat-fault, far-fault, design earthquake and probability based scenario earthquake, which can be occurred in the NPP sites.

• Journal of the Korean Society of Safety
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• v.15 no.2
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• pp.126-135
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• 2000

Probabilistic Safety Assessment (PSA) is an engineering analysis method to identify possible contributors to the risk from a nuclear power plant and now it has become a standard tool in safety evaluation of nuclear power plants. PSA consists of three phases named as Level 1, 2 and 3. Level 2 PSA, mainly focused in this paper, uses a step-wise approach. At first, plant damage states (PDSs) are defined from the Level 1 PSA results and they are quantified. Containment event tree (CET) is then constructed considering the physico-chemical phenomena in the containment. The quantification of CET can be assisted by a decomposition event tree (DET). Finally, source terms are quantitatively characterized by the containment failure mode. As the main benefit of PSA is to provide insights into plant design, performance and environmental impacts, including the identification of the dominant risk contributors and the comparison of options for reducing risk, this technique is expected to be applied to the industrial safety area.

• Journal of Korean Society of Steel Construction
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• v.15 no.2
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• pp.187-196
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• 2003

Today, the Working Stress Rating (WSR) is being widely used for the capacity-rating and the safety assessment of railroad steel bridges. Since it cannot incorporate the uncertainties, several studies have been carried out in order to get over the incompleteness of the conventional capacity-rating and safety assessment. A system reliability-based equivalent capacity-rating method, which can evaluate the capacity of existing bridges, has been recently proposed. For more efficient reliability analysis, probabilistic parameters of the random variables in the limit-state models should be reasonably evaluated. Especially, uncertainties for live load effects must be realistically included. In this study, an improved limit-state model was used for the system reliability-based equivalent strength method. This model can incorporate the probabilistic parameters obtained from ambient measurement data. To demonstrate the applicability of the improved system reliability-based equivalent capacity rating method, this was applied to the existing steel plate girder bridge for comparison with the conventional capacity-rating and safety assessment.

• Journal of the Society of Disaster Information
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• v.18 no.2
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• pp.252-260
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• 2022

Purpose: As the intensity and frequency of natural hazards are increasing due to climate change, external events that affecting nuclear power plants(NPPs) may increase. NPPs must be protected from external events such as natural hazards and human-induced hazards. External events that may occur in NPPs should be identified, and external events that may affect NPPs should be identified. This study introduces the methodology of identification and screening methods for external events by literature review. Method: The literature survey was conducted on the identification and screening methods of external events for probabilistic safety assessment of NPPs. In addition, the regulations on the identification and screening of external events were investigated. Result: In order to minimize the cost of external event impact analysis of nuclear power plants, research on identifying and screening external events is being conducted. In general, in the identification process, all events that can occur at the NPPs are identified. In the screening process, external events are selected based on qualitative and quantitative criteria in most studies. Conclusions: The process of identifying and screening external events affecting NPPs is becoming important. This paper, summarize on how to identify and screen external events for a probabilistic safety assessment of NPPs. It is judged that research on bounding analysis and conservative analysis methods performed in the quantitative screening process of external events is necessary.

• Journal of the Korean Society of Safety
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• v.31 no.4
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• pp.90-96
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• 2016

This study proposes a dynamic reliability analysis of control system as a method of quantitative evaluation of its performance in probabilistic terms. In this dynamic reliability analysis, the failure event is defined as an event that the dynamic response of the structural system exceeds a displacement limit, whereas the conventional reliability analysis method has limitations that do not properly assess the actual time history response of the structure subjected to dynamic loads, such as earthquakes and high winds, by taking the static response into account in the failure event. In this first paper, we discuss the control effect of the viscous damper on the seismic performance of the member-level failure where the failure event of the structural member consists of the union set of time-sequential member failures during the earthquake excitations and the failure probability of the earthquake-excited structural member is computed using system reliability approach to consider the statistical dependence of member failures between the subsequent time points. Numerical results demonstrate that the proposed approach can present a reliable assessment of the control performance of the viscous damper system in comparison with MCS method. The most important advantage of the proposed approach can provide us more accurate estimate of failure probability of the structural control system by using the actual time-history responses obtained by dynamic response analysis.