• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1834-1846
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• 2020

The SMART (System-integrated Modular Advanced ReacTor) is an integral-type small modular reactor developed by KAERI (Korea Atomic Energy Research Institute). This paper discusses the development of a dynamic equivalent model of the SMART control rod drive mechanism that can be efficiently utilized for complicated analysis during the design of the SMART. A semi-empirical approach is used to develop the equivalent model; that is, the equivalent model is defined analytically and verified empirically. Two types of tests, dynamic characteristics tests and seismic loading tests, are conducted for the development and verification of the dynamic equivalent model, respectively. Acceleration response spectra from the seismic analysis based on the developed equivalent model show good agreement with those from the seismic loading tests.

• Structural Engineering and Mechanics
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• v.3 no.6
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• pp.569-581
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• 1995

In this paper, the seismic behavior of a 10-story building equipped with viscoelastic dampers is analyzed. The effects of ambient temperature, the thickness, the total area, and the position of the viscoelastic dampers are studied. Results indicate that the energy-absorbing capacity of viscoelastic damper decreases with increasing the ambient temperature. The thickness and the total area of viscoelastic dampers also affect the seismic mitigation capacity. The thickness cannot be too small, which is not effective in vibration reduction, nor can it be too large, which not only increases the cost but also reduces the seismic resistance. The total area of viscoelastic dampers should be determined properly for optimum damper performance at the most economical design. The mounting position of viscoelastic dampers also influences the structure's seismic performance. Numerical results show that, if properly equipped, the VE dampers can reduce the structural response both floor displacement and story shear force and increase the overall level of damping in structures during earthquakes.

• International journal of steel structures
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• v.18 no.4
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• pp.1420-1430
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• 2018

Earthquakes of 5.8 and 5.4 Richter scale recently occurred one after another in Korea, changing the Korean peninsula from an earthquake safe zone but 'earthquake danger zone'. Therefore, seismic reinforcements must expand to include structures with low seismic resistance in order to prepare for earthquakes on a larger scale in the future. This study investigated the performances of various seismic reinforcement systems such as X-braced steel rod reinforcement, steel shear wall with circular opening reinforcement, and slit damper reinforcement using shaking table test and computational analyses of seismic data in order to establish a proper seismic reinforcement plan. These three seismic reinforcement systems could increase the stiffness and strength of existing structures and reduce maximum drift ratio in the event of an earthquake.

• Journal of the Earthquake Engineering Society of Korea
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• v.22 no.3
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• pp.113-119
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• 2018

The probabilistic seismic safety assessment is one of the methodology to evaluate the seismic safety of the nuclear power plants. The site characteristics of the nuclear power plant should be reflected when evaluating the seismic safety of the nuclear power plant. The Korea seismic characteristics are strong in high frequency region and may be different from NRC Regulatory Guide 1.60, which is the design spectrum of nuclear power plants. In this study, seismic response of a nuclear power plant structure by Pohang earthquake (2017.11.15. (KST)) is investigated. The Pohang earthquake measured at the Cheongsong seismic observation station (CHS) is scaled to the peak ground acceleration (PGA) of 0.2 g and the seismic acceleration time history curve corresponding to the design spectrum is created. A nuclear power plant of the containment building and the auxiliary buildings are modeled using OPENSEES to analyze the seismic response of the Pohang earthquake. The seismic behavior of the nuclear power plant due to the Pohang earthquake is investigated. And the seismic performances of the equipment of a nuclear power plant are evaluated by the HCLPF. As a result, the seismic safety evaluation of nuclear power plants should be evaluated based on site-specific characteristics of nuclear power plants.

• Journal of the Earthquake Engineering Society of Korea
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• v.18 no.3
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• pp.141-150
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• 2014

A methodology to evaluate the seismic performance of interface piping systems that cross the isolation interface in the seismically isolated nuclear power plant (NPP) was developed. The developed methodology was applied to the safety-related interface piping system to demonstrate the seismic performance of the target piping system. Not only the seismic performance for the design level earthquakes but also the performance for the beyond design level earthquakes were evaluated. Two artificial seismic ground input motions which were matched to the design response spectra and two historical earthquake ground motions were used for the seismic analysis of piping system. The preliminary performance evaluation results show that the excessive relative displacements can occur in the seismically isolated piping system. If the input ground motion contained relatively high energy in the low frequency region, we could find that the stress response of the piping system exceed the allowable stress level even though the intensity of the input ground motion is equal to the design level earthquake. The structural responses and seismic performances of piping system were varied sensitively with respect to the intensities and frequency contents of input ground motions. Therefore, for the application of isolation system to NPPs and the verification of the safety of piping system, the seismic performance of the piping system subjected to the earthquake at the target NPP site should be evaluated firstly.

• Structural Engineering and Mechanics
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• v.43 no.4
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• pp.475-502
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• 2012

In this paper, the seismic analysis modeling technologies for sodium-cooled fast reactor (SFR) are presented with detailed descriptions for each structure, system and component (SSC) model. The complicated reactor system of pool type SFR, which is composed of the reactor vessel, internal structures, intermediate heat exchangers, primary pumps, core assemblies, and core support structures, is mathematically described with simple stick models which can represent fundamental frequencies of SSC. To do this, detailed finite element analyses were carried out to identify fundamental beam frequencies with consideration of fluid added mass effects caused by primary sodium coolant contained in the reactor vessel. The calculation of fluid added masses is performed by detailed finite element analyses using FAMD computer program and the results are discussed in terms of the ways to be considered in a seismic modeling. Based on the results of seismic time history analyses for both seismic isolation and non-isolation design, the functional requirements for relative deflections are discussed, and the design floor response spectra are proposed that can be used for subsystem seismic design.

• Structural Engineering and Mechanics
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• v.31 no.1
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• pp.93-112
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• 2009

In this study seismic analyses of steel structures were carried out to examine the effect of ground motion characteristics and structural properties on energy demands using 100 earthquake ground motions recorded in different soil conditions, and the results were compared with those of previous works. Analysis results show that ductility ratios and the site conditions have significant influence on input energy. The ratio of hysteretic to input energy is considerably influenced by the ductility ratio and the strong motion duration. It is also observed that as the predominant periods of the input energy spectra are significantly larger than those of acceleration response spectra used in the strength design, the strength demand on a structure designed based on energy should be checked especially in short period structures. For that reason framed structures with buckling-restrained-braces (BRBs) were designed in such a way that all the input energy was dissipated by the hysteretic energy of the BRBs, and the results were compared with those designed by conventional strength-based design procedure.

• Structural Engineering and Mechanics
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• v.85 no.1
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• pp.55-64
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• 2023

Energy-based seismic design and evaluation methods are promising to be involved in the next generation design codes. Accordingly, determining the distribution of earthquake input energy demand among floor levels is quite imperative in order to develop an energy-based seismic design procedure. In this paper, peak floor input energy demands are achieved from relative input energy response histories of several reinforced concrete (RC) frames. A set of 22 horizontal acceleration histories selected from recorded near-fault earthquakes and scaled in time domain to be compatible with the elastic acceleration design spectra of Turkish Seismic Design Code are used in time history analyses. The distribution of the computed input energy per mass values and the arithmetic means through the height of the considered RC frames are presented as a result. It is found that spatial distribution of input energy per mass is highly affected by the number of stories. Very practical yet consistent formulation of distributing the total input energy to story levels is achieved, as a most important contribution of the study.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.600-606
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• 2019

Studies on the application of the lead rubber bearing (LRB) isolation system to nuclear power plants are being carried out as one of the measures to improve seismic performance. Nuclear power plants with isolation systems require seismic probabilistic safety assessments, for which the seismic fragility of the structures, systems, and components needs be calculated, including for beyond design basis earthquakes. To this end, seismic response analyses are required, where it can be seen that the behaviors of the isolation system components govern the overall seismic response of an isolated plant. The numerical model of the LRB used in these seismic response analyses plays an important role, but in most cases, the extreme performance of the LRB has not been well studied. The current work therefore develops an extreme nonlinear numerical model that can express the seismic response of the LRB for beyond design basis earthquakes. A full-scale LRB was fabricated and dynamically tested with various input conditions, and test results confirmed that the developed numerical model better represents the behavior of the LRB over previous models. Subsequent seismic response analyses of isolated nuclear power plants using the model developed here are expected to provide more accurate results for seismic probabilistic safety assessments.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1420-1427
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• 2023

Seismic risk has received increased attention since the 2011 Fukushima accident in Japan. The seismic risk of a nuclear power plant is evaluated via seismic probabilistic safety assessment (PSA), for which several methods are available. Recently, the discrete approach has become widely used. This approximates the seismic risk by discretizing the ground motion level interval into a small number of subintervals with the expectation of providing a conservative result. The present study examines the effect of the number of subintervals upon the results of seismic risk quantification. It is demonstrated that a small number of subintervals may lead to either an underestimation or overestimation of the seismic risk depending on the ground motion level. The present paper also provides a method for finding the boundaries between overestimation and underestimation regions, and illustrates the effect of the number of subintervals upon the seismic risk evaluation with an example. By providing a method for determining the effect of a small number of subintervals upon the results of seismic risk quantification, the present study will assist seismic PSA analysts to determine the appropriate number of subintervals and to better understand seismic risk quantification.