• Journal of Environmental Science International
• /
• v.22 no.8
• /
• pp.965-977
• /
• 2013

Reliable long-term streamflow forecasting is invaluable for water resource planning and management which allocates water supply according to the demand of water users. It is necessary to get probabilistic forecasts to establish risk-based reservoir operation policies. Probabilistic forecasts may be useful for the users who assess and manage risks according to decision-making responding forecasting results. Probabilistic forecasting of seasonal inflow to Andong dam is performed and assessed using selected predictors from sea surface temperature and 500 hPa geopotential height data. Categorical probability forecast by Piechota's method and logistic regression analysis, and probability forecast by conditional probability density function are used to forecast seasonal inflow. Kernel density function is used in categorical probability forecast by Piechota's method and probability forecast by conditional probability density function. The results of categorical probability forecasts are assessed by Brier skill score. The assessment reveals that the categorical probability forecasts are better than the reference forecasts. The results of forecasts using conditional probability density function are assessed by qualitative approach and transformed categorical probability forecasts. The assessment of the forecasts which are transformed to categorical probability forecasts shows that the results of the forecasts by conditional probability density function are much better than those of the forecasts by Piechota's method and logistic regression analysis except for winter season data.

• Nuclear Engineering and Technology
• /
• v.49 no.2
• /
• pp.373-379
• /
• 2017

Seismic probabilistic safety assessments are used to help understand the impact potential seismic events can have on the operation of a nuclear power plant. An important component to seismic probabilistic safety assessment is the seismic hazard curve which shows the frequency of seismic events. However, these hazard curves are estimated assuming a normal distribution of the seismic events. This may not be a strong assumption given the number of recorded events at each source-to-site distance. The use of a normal distribution makes the calculations significantly easier but may underestimate or overestimate the more rare events, which is of concern to nuclear power plants. This paper shows a preliminary exploration into the effect of using a distribution that perhaps more represents the distribution of events, such as the t-distribution to describe data. The integration of a probability distribution with potentially larger tails basically pushes the hazard curves outward, suggesting a different range of frequencies for use in seismic probabilistic safety assessments. Therefore the use of a more realistic distribution results in an increase in the frequency calculations suggesting rare events are less rare than thought in terms of seismic probabilistic safety assessment. However, the opposite was observed with the ground motion prediction equation considered.

• Nuclear Engineering and Technology
• /
• v.46 no.5
• /
• pp.699-706
• /
• 2014

The probabilistic seismic performance of a standard Korean nuclear power plant (NPP) with an idealized isolation is investigated in the present work. A probabilistic seismic hazard analysis (PSHA) of the Wolsong site on the Korean peninsula is performed by considering peak ground acceleration (PGA) as an earthquake intensity measure. A procedure is reported on the categorization and selection of two sets of ground motions of the Tohoku earthquake, i.e. long-period and common as Set A and Set B respectively, for the nonlinear time history response analysis of the base-isolated NPP. Limit state values as multiples of the displacement responses of the NPP base isolation are considered for the fragility estimation. The seismic risk of the NPP is further assessed by incorporation of the rate of frequency exceedance and conditional failure probability curves. Furthermore, this framework attempts to show the unacceptable performance of the isolated NPP in terms of the probabilistic distribution and annual probability of limit states. The comparative results for long and common ground motions are discussed to contribute to the future safety of nuclear facilities against drastic events like Tohoku.

• Journal of the Earthquake Engineering Society of Korea
• /
• v.8 no.6 s.40
• /
• pp.23-29
• /
• 2004

The design earthquake used for the seismic analysis and design of NPP (Nuclear Power Plant) is determined by the deterministic or probabilistic methods. The probabilistic seismic hazard analysis(PSHA) for the nuclear power plant sites was performed for the probabilistic seismic risk assessment. The probabilistic seismic hazard analysis for the nuclear power plant site had been completed as a part of the probabilistic seismic risk assessment. The probabilistic method become a resonable method to determine the design earthquakes for NPPs. In this study, the defining method of the probability based scenario earthquake was established, and as a sample calculation, the probability based scenario earthquakes were estimated by the de-aggregation of the probabilistic seismic hazard. By using this method, it is possible to define the probability based scenario earthquakes for the seismic design and seismic safety evaluation of structures. It is necessary to develop the rational seismic source map and the attenuation equations for the development of reasonable scenario earthquakes.

• Nuclear Engineering and Technology
• /
• v.49 no.2
• /
• pp.360-372
• /
• 2017

Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.

• Nuclear Engineering and Technology
• /
• v.49 no.2
• /
• pp.306-312
• /
• 2017

The purpose of this research is to introduce the technical standard of accident sequence precursor (ASP) analysis, and to propose a case study using the dynamic-probabilistic safety assessment (D-PSA) approach. The D-PSA approach can aid in the determination of high-risk/low-frequency accident scenarios from all potential scenarios. It can also be used to investigate the dynamic interaction between the physical state and the actions of the operator in an accident situation for risk quantification. This approach lends significant potential for safety analysis. Furthermore, the D-PSA approach provides a more realistic risk assessment by minimizing assumptions used in the conventional PSA model so-called the static-PSA model, which are relatively static in comparison. We performed risk quantification of a steam generator tube rupture (SGTR) accident using the dynamic event tree (DET) methodology, which is the most widely used methodology in D-PSA. The risk quantification results of D-PSA and S-PSA are compared and evaluated. Suggestions and recommendations for using D-PSA are described in order to provide a technical perspective.

• Nuclear Engineering and Technology
• /
• v.53 no.4
• /
• pp.1127-1133
• /
• 2021

A Spent Fuel Pool (SFP) is designed to store spent fuel assemblies in the pool. And, a SFP cooling and cleanup system cools the SFP coolant through a heat exchanger which exchanges heat with component cooling water. If the cooling system fails or interfacing pipe (e.g., suction or discharge pipe) breaks, the cooling function may be lost, probably leading to fuel damage. In order to prevent such an incident, it is required to properly cool the spent fuel assemblies in the SFP by either recovering the cooling system or injecting water into the SFP. Probabilistic safety assessment (PSA) is a good tool to assess the SFP risk when an initiating event for the SFP occurs. Since PSA has been focused on reactor-side so far, it is required to study on the framework of PSA approach for SFP and identify the key factors in terms of fuel damage frequency (FDF) through a case study. In this study, therefore, a case study of SFP-PSA on the basis of design information of APR-1400 has been conducted quantitatively, and several sensitivity analyses have been conducted to understand the impact of the key factors on FDF.

• Computers and Concrete
• /
• v.18 no.1
• /
• pp.69-97
• /
• 2016

Performance-based remaining life assessment of reinforced concrete bridge girders, subject to chloride-induced corrosion of reinforcement, is addressed in this paper. Towards this, a methodology that takes into consideration the human judgmental aspects in expert decision making regarding condition state assessment is proposed. The condition of the bridge girder is specified by the assignment of a condition state from a set of predefined condition states, considering both serviceability- and ultimate- limit states, and, the performance of the bridge girder is described using performability measure. A non-homogeneous Markov chain is used for modelling the stochastic evolution of condition state of the bridge girder with time. The thinking process of the expert in condition state assessment is modelled within a probabilistic framework using Brunswikian theory and probabilistic mental models. The remaining life is determined as the time over which the performance of the girder is above the required performance level. The usefulness of the methodology is illustrated through the remaining life assessment of a reinforced concrete T-beam bridge girder.

• Nuclear Engineering and Technology
• /
• v.41 no.4
• /
• pp.391-402
• /
• 2009

This paper introduces the development of safety assessment technology in Korea, focusing on the activities of the Korea Atomic Energy Research Institute in the areas of system thermal hydraulics, severe accidents and probabilistic safety assessment. In the 1970s and 1980s, safety analysis codes and methodologies were introduced from the United States, France, Canada and other developed countries along with technology related to the construction and operation of nuclear power plants. The main focus was on understanding and utilizing computer codes that were sourced from abroad up to the early 1990s, when efforts to develop domestic safety analysis codes and methodologies became active. Remarkable achievements have been made over the last 15 years in the development and application of safety analysis technologies. In addition, significant experimental work has been performed to verify the safety characteristics of reactors and fuels as well as to support the development and validation of analysis methods.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.13 no.1
• /
• pp.1-15
• /
• 2017

After Kyeongju earthquake occurred in September 12, 2016, the seismic safety of nuclear power plants became important issue in our country. The seismic safety of nuclear power plant against beyond design basis earthquake became very important to secure the public safety. In this paper, the current status of the seismic safety assessment methodology is reviewed and some aspects for the reliability improvement of the seismic safety assessment results are proposed. Seismic margin analysis and probabilistic seismic safety assessment have been used for the seismic safety evaluation of a nuclear power pant. The basic procedure and the related issues and proposals for the probabilistic seismic safety assessment are investigated.