• Proceedings of the Computational Structural Engineering Institute Conference
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• 2006.04a
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• pp.473-480
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• 2006

Liquefaction of soil foundation is one of the major seismic damage types of infrastructures. In this paper, deterministic and probabilistic approaches for the evaluation of liquefaction potential are briefly summarized and the risk assessment method is newly proposed using seismic fragility and seismic hazard curves. Currently the deterministic approach is widely used to evaluate the liquefaction potential in Korea. However, the there are a certain degree of uncertainties in the soil properties such as elastic modulus and resistant capacity, therefore the probabilistic approach is more promising. Two types of probabilistic approach are introduced including (1) failure probability for a given design earthquake and (2) the seismic risk of liquefaction of soil for a given service life. The results from different methods show a similar trend, and the liquefaction potential can be more quantitatively evaluated using risk analysis method.

• 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.50 no.8
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• pp.1234-1245
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• 2018

The risk of multi-unit nuclear power plants (NPPs) at a site has received considerable critical attention recently. However, current probabilistic safety assessment (PSA) procedures and computer code do not support multi-unit PSA because the traditional PSA structure is mostly used for the quantification of single-unit NPP risk. In this study, the main purpose is to develop a multi-unit Level 2 PSA method and apply it to full-power operating six-unit OPR1000. Multi-unit Level 2 PSA method consists of three steps: (1) development of single-unit Level 2 PSA; (2) extracting the mapping data from plant damage state to source term category; and (3) combining multi-unit Level 1 PSA results and mapping fractions. By applying developed multi-unit Level 2 PSA method into six-unit OPR1000, site containment failure probabilities in case of loss of ultimate heat sink, loss of off-site power, tsunami, and seismic event were quantified.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.471-482
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• 2012

To assess the risk of nuclear power plant operation and to determine the risk impact of digital systems, there is a need to quantitatively assess the reliability of the digital systems in a justifiable manner. The Probabilistic Risk Analysis (PRA) is a tool which can reveal shortcomings of the NPP design in general and PRA analysts have not had sufficient guiding principles in modelling particular digital components malfunctions. Currently digital I&C systems are mostly analyzed simply and conventionally in PRA, based on failure mode and effects analysis and fault tree modelling. More dynamic approaches are still in the trial stage and can be difficult to apply in full scale PRA-models. As basic events CPU failures, application software failures and common cause failures (CCF) between identical components are modelled.The primary goal is to model dependencies. However, it is not clear which failure modes or system parts CCF:s should be postulated for. A clear distinction can be made between the treatment of protection and control systems. There is a general consensus that protection systems shall be included in PRA, while control systems can be treated in a limited manner. OECD/NEA CSNI Working Group on Risk Assessment (WGRisk) has set up a task group, called DIGREL, to develop taxonomy of failure modes of digital components for the purposes of PRA. The taxonomy is aimed to be the basis of future modelling and quantification efforts. It will also help to define a structure for data collection and to review PRA studies.

• Journal of Soil and Groundwater Environment
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• v.15 no.3
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• pp.7-14
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• 2010

Ecological Risk Assessment (ERA) supports a decision-making process such as establishment of environmental quality criteria. Soil quality criteria (SQC) are essential to protect soil organisms from the exposure to various soil contaminants. In this study, ERA methodologies of advanced countries for soil pollution were extensively compared to propose the ERA approach suitable for soil ecosystem in Korea. The soil ERAs in European Chemical Bureau(ECB), The Netherlands, and Canada can be classified as deterministic ecological risk assessment (DERA), and probabilistic ecological risk assessment (PERA) based on species sensitivity distribution (SSD). We propose three ERA methods according to abundance and reliability of soil ecotoxicity data. The method considered land use such as residential/agricultural, and industrial/commercial uses. The taxonomic groups of soil organism were classified as 'Class' level including different trophic levels (Magnoliopsida or Liliopsida, Clitellata, and Insecta or Secernentea). This study can be used to estimate the soil quality criteria to protect soil biota.

• Smart Structures and Systems
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• v.23 no.2
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• pp.185-194
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• 2019

The probabilistic tsunami risk assessment of large coastal areas is challenging because the inland propagation of a tsunami wave requires an accurate numerical model that takes into account the interaction between the ground, the infrastructures, and the wave itself. Classic mesh-based methods face many challenges in the propagation of a tsunami wave inland due to their ever-moving boundary conditions. In alternative, mesh-less based methods can be used, but they require too much computational power in the far-field. This study proposes a hybrid approach. A mesh-based method propagates the tsunami wave from the far-field to the near-field, where the influence of the sea floor is negligible, and a mesh-less based method, smooth particle hydrodynamic, propagates the wave onto the coast and inland, and takes into account the wave structure interaction. Nowadays, this can be done because the advent of general purpose GPUs made mesh-less methods computationally affordable. The method is used to simulate the inland propagation of the 2004 Indian Ocean tsunami off the coast of Indonesia.

• Earthquakes and Structures
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• v.17 no.1
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• pp.91-99
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• 2019

Fragility analysis is an effective tool that is frequently used for seismic risk assessment of bridges. There are three different approaches to derive a fragility curve: experimental, empirical and analytical. Both experimental and empirical methods to derive fragility curve are based on past earthquake reports and expert opinions which are not suitable for all bridges. Therefore, analytical fragility analysis becomes important. Nonlinear time history analysis is commonly used which is the most reliable method for determining probabilistic demand models. In this study, to determine the probabilistic demand models of bridges, time history analyses were performed considering both material and geometrical nonlinearities. Serviceability limit states for three different service velocities were considered as a performance goal. Also, support displacements, component yielding and collapse limits were taken into account. Both serviceability and component fragility were derived by using maximum likely hood methods. Finally, the seismic performance and critical members of the bridge were probabilistically determined and clearly presented.

• 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.

• Journal of Radiation Industry
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• v.17 no.1
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• pp.83-92
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• 2023

When assessing the risk of radioactive wastes transportation on land, computer codes such as RADTRAN and RISKIND are used as deterministic methods. Transportation risk assessment using the deterministic method requires a relatively short assessment time. On the other hand, transportation risk assessment using the probabilistic method requires a relatively long assessment time, but produces more reliable results. Therefore, a study is needed to evaluate the exposure dose using a deterministic method that can be evaluated relatively quickly, and to compare and analyze the exposure dose result using a probabilistic method. The purpose of this study is to evaluate the exposure dose during transportation of radioactive wastes using deterministic and probabilistic methods, and to compare and analyze them. For this purpose, the main exposure factors were selected and various exposure situations were set. The distance between the radioactive waste and the receptor, the size of the package, and the speed of vehicle were selected as the main exposure factors. The exposure situation was largely divided into when the radioactive wastes were stationary and when they were passing. And the dose (rate) model of the deterministic overland transportation risk assessment computer code was analyzed. Finally, the deterministic method of the RADTRAN computer code and the RISKIND computer code and the probabilistic method of the MCNP 6 computer code were used to evaluate the exposure dose in various exposure situations during transportation of radioactive wastes. Then we compared and analyzed them. As a result of the evaluation, the tendency of the exposure dose (rate) was similar when the radioactive wastes were stationary and passing. For the same situation, the evaluation results of the RADTRAN computer code were generally more conservative than the results of the RISKIND computer code and the MCNP 6 computer code. The evaluation results of the RISKIND computer code and the MCNP 6 computer code were relatively similar. The results of this study are expected to be used as basic data for establishing the radioactive wastes transportation risk assessment system in Korea in the future.

• Journal of the Korean Society of Systems Engineering
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• v.17 no.2
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• pp.82-90
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• 2021

The spent fuel pool (SFP) of a nuclear power plant functions to store the spent fuel. The spent fuel pool is designed to properly remove the decay heat generated from the spent fuel. If the cooling function is lost and proper operator action is not taken, the spent fuel in the storage pool can be damaged. Probabilistic safety assessment (PSA) is a safety evaluation method that can evaluate the risk of a large and complex system. So far, the probabilistic safety assessment of nuclear power plants has been mainly performed on the reactor. This study defined the requirements and the functional architecture for the probabilistic safety assessment of the spent fuel pool (SFP-PSA) by applying the systems engineering process. And, a systematic and efficient methodology was defined according to the architecture.