• Proceedings of the Korea Concrete Institute Conference
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• 1994.10a
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• pp.285-290
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• 1994

The influence of elevated temperatures on the mechanical properties of concrete is important for fire-resistance studies and also for understanding the behavior of containment vessel, such as nuclear reactor pressure vessels, during service and ultimate condition. The present study is to clarify the damage/deterioration of concrete structures that are subjected to high temperature exposure. To this end, comprehensive experiments are conducted. The major test variables are the peak temperatures, rate of temperature increase, and sustained duration at peak temperature. The results include weight loss residual compressive strength and stress-strain curve. From those results, residua compressive strength formula and stress-strain relationship are proposed.

• Proceedings of the Korea Concrete Institute Conference
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• 1994.10a
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• pp.291-300
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• 1994

The influence of elevated temperatures on the mechanical properties of concrete is important for fire-resistance studies and also for understanding the behavior of containment vessel, such as nuclear reactor pressure vessels, during service and ultimate condition. The present study is to clarify the damage/deterioration of concrete structures that are subjected to high temperature exposure. To this end, comprehensive experiments are conducted. The major test variables are the peak temperatures, rate of temperature increase, and sustained duration at peak temperature. The results include weight loss residual compressive strength and stress-strain curve. From those results, residua compressive strength formula and stress-strain relationship are proposed.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.4
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• pp.355-361
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• 2014

Steam explosions can be caused by fuel-coolant interactions resulting from failure of the external vessel cooling system in a new nuclear power plant. This can threaten the integrity of structures, including the nuclear reactor and the containment building. In the present study, an improved technique for analyzing the steam explosion phenomenon was proposed on the basis of previous research and was verified by simulations involving alumina experiments. Also, the improved analysis technique was applied to determine the pressure history of the reactor cavity in accordance with postulated failure locations. The results of the analysis revealed that the effects of vessel side failure are more serious than those of vessel bottom failure, with approximately 70% higher maximum pressure.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4348-4358
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• 2022

The thermal-hydraulics phenomena in a containment during an accident will necessarily include radiative heat transfer (i) within the gas mixture due to the high radiative absorption and emission of steam and (ii) between the gas mixture and the surrounding structures. The analysis of some previous PANDA experiments (PSI, Switzerland) demonstrated the importance of the proper modelling of radiation for the benefit of numerical simulations. These results together with dedicated scoping calculations conducted for the present experiments indicated that the radiative heat transfer is considerable, even for a very low amount of steam (≈2%). The H2P2 series conducted in the large-scale PANDA facility at the Paul-Scherrer-Institut (PSI) in the framework of the OECD/NEA HYMERES-2 project is intended to enhance the understanding of thermal radiation phenomena and to provide a benchmark for corresponding numerical simulations. Thus, the test matrix was tailored around the two opposite extremes: either gas compositions with small steam content such that radiative heat transfer phenomena can be neglected. Or gas mixtures containing larger amounts of steam, so that radiative heat transfer is expected to play a dominant role. The H2P2 series consists of 5 experiments designed to isolate the radiation phenomena from convective and diffusive effects as much as possible. One vessel with a diameter of 4 m and a height of 8 m was preconditioned with different mixtures of air / steam at room and elevated temperatures. This was followed by the build-up of a stable helium stratification at constant pressure in the upper part of the vessel. After that, helium was injected from the top into the vessel which leads to an increase of the vessel pressure and a corresponding elevation-dependent and transient rise of the gas temperature. It is shown that even the addition of small amounts of steam in the initial gas atmosphere considerably impacts the radiative heat transport throughout all phases of the experiments and markedly influences i) the monitored gas peak temperature, ii) the temperature history during the compression and iii) the following relaxation phase after the compression was stopped. These PANDA experiments are the first of its kind conducted in a large scale thermal-hydraulic facility.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.411-425
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• 2017

Using a probabilistic safety assessment, a risk evaluation framework for an aircraft crash into an interim spent fuel storage facility is presented. Damage evaluation of a detailed generic cask model in a simplified building structure under an aircraft impact is discussed through a numerical structural analysis and an analytical fragility assessment. Sequences of the impact scenario are shown in a developed event tree, with uncertainties considered in the impact analysis and failure probabilities calculated. To evaluate the influence of parameters relevant to design safety, risks are estimated for three specification levels of cask and storage facility structures. The proposed assessment procedure includes the determination of the loading parameters, reference impact scenario, structural response analyses of facility walls, cask containment, and fuel assemblies, and a radiological consequence analysis with dose-risk estimation. The risk results for the proposed scenario in this study are expected to be small relative to those of design basis accidents for best-estimated conservative values. The importance of this framework is seen in its flexibility to evaluate the capability of the facility to withstand an aircraft impact and in its ability to anticipate potential realistic risks; the framework also provides insight into epistemic uncertainty in the available data and into the sensitivity of the design parameters for future research.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.78-91
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• 2024

Double steel plate concrete composite shear wall (SCSW) has been widely utilized in nuclear power plants and high-rise structures, and its shear connectors have a substantial impact on the seismic performance of SCSW. Therefore, in this study, the mechanical properties of SCSW with angle stiffening ribs as shear connections were parametrically examined for the reactor containment structure of nuclear power plants. The axial compression ratio of the SCSW, the spacing of the angle stiffening rib arrangement and the thickness of the angle stiffening rib steel plate were selected as the study parameters. Four finite element models were constructed by using the finite element program named ABAQUS to verify the experimental results of our team, and 13 finite element models were established to investigate the selected three parameters. Thus, the shear capacity, deformation capacity, ductility and energy dissipation capacity of SCSW were determined. The research results show that: compared with studs, using stiffened ribs as shear connectors can significantly enhance the mechanical properties of SCSW; When the axial compression ratio is 0.3-0.4, the seismic performance of SCSW can be maximized; with the lowering of stiffener gap, the shear bearing capacity is greatly enhanced, and when the gap is lowered to a specific distance, the shear bearing capacity has no major affect; in addition, increasing the thickness of stiffeners can significantly increase the shear capacity, ductility and energy dissipation capacity of SCSW. With the rise in the thickness of angle stiffening ribs, the improvement rate of each mechanical property index slows down. Finally, the shear bearing capacity calculation formula of SCSW with angle stiffening ribs as shear connectors is derived. The average error between the theoretical calculation formula and the finite element calculation results is 8% demonstrating that the theoretical formula is reliable. This study can provide reference for the design of SCSW.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.25 no.1
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• pp.57-64
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• 2012

Since a seismic exceeding design load can result in exposing radioactive material during disposal process of radioactive wastes, the repository should be designed with enough seismic margin. In this paper, a seismic fragility analysis was performed to evaluate the seismic capacity of surface facility structures. According to the analysis results, since inspection & store facility and radioactive waste facility have a rectangle geometry, the seismic capacity was differently presented about 23%~43% according to the axis of structures. The HCLPF capacity of inspection & store facility and radioactive waste facility was 0.52g and 0.93g, respectively. And it was observed that seismic capacity of radioactive waste facility was similar to that of a containment for nuclear power plants.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.26 no.2
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• pp.131-138
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• 2013

The high damping rubber bearings contributed to reduce a seismic force transmitted to upper structures, the material properties of rubber changes with time and the rubber with age-related degradation can affect the seismic response of structures and equipments. Therefore the seismic response of structure considering age-related degradation of isolators should be evaluated. In this paper, the stiffness and damping for isolators were defined using the aging data proposed by other researchers. The reactor containment building and the auxiliary building were selected to conduct the nonlinear analysis and the natural frequency, maximum responses, floor response spectrum(FRS) were evaluated with time using the four earthquakes with different frequency contents. According to the analysis results, the seismic responses are increased by the age-related degradation of isolators and the detail inspections should be conducted up to 20 years because it was presented that the change of FRS was high during this period.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.35 no.2
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• pp.83-91
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• 2022

Seismic design and analysis of nuclear power plant components are performed based on an decoupled model. However, this decoupled analysis has a limitation in that it generates inaccurate results compared to the coupled analysis because it cannot simulate actual phenomena such as the interaction between structures and components. Thus, this study performed seismic coupled and decoupled analysis on an existing nuclear containment structure and related components, considering the mass and natural frequency ratios. And based on these results, comparative analyses of responses of components were conducted. Consequently, the seismic coupled analysis result generally gave a smaller value than the decoupled analysis result. These results were similar to the analysis results for the simple coupled model, which was an existing study, but the difference in component responses was much more pronounced. Also, this was influenced by the installation location of the component rather than the influence of the input frequency of the input seismic motions. Finally, the difference between the decoupled and coupled seismic analysis occurred in the region where the mass ratio of the components was large, and the natural frequencies were almost similar due to the considerable dynamic interaction between the structure and the component in this realm.

• Magazine of the Korean Society of Agricultural Engineers
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• v.38 no.5
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• pp.74-84
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• 1996

Dynamic loading of structures often causes excursions of stresses well into the inelastic range and the influence of geometric changes on the dynamic response is also significant in many cases. Therefore, both material and geometric nonlinearity effects should be considered in case that a dynamic load acts on the structure. For developing a program to analyze the dynamic response of an axisymmetric shell in this study, the material nonlinearity effect on the dynamic response was formulated by the elasto-viscoplastic model highly corresponding to the real behavior of the material. Also, the geometrically nonlinear behavior is taken into account using a total Lagrangian coordinate system, and the equilibrium equation of motion was numerically solved by a central difference scheme. A complete finite element program has been developed and the results obtained by it are compared with those in the references 1 and 2. The results are in good agreement with each other. As a case study of its application, the developed program was applied to a dynamic response analysis of a nuclear reinforced concrete containment structure. The results obtained from the' numerical examples are summarized as follows : 1. The dynamic magnification factor of the displacement and the stress were unrelated with the concrete strength. 2. As shown by the results that the displacement dynamic magnification factor were form 1.7 to 2.3 and the stress dynamic magnification factor from 1.8 to 2.5, the dynamic magnification factor of stress were larger than that of displacement. 3. The dynamic magnification factor of stress on the exterior surface was larger than that on the interior surface of the structure.