• Steel and Composite Structures
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• v.46 no.2
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• pp.221-236
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• 2023

In order to study the seismic response of the main plant of steel reinforced concrete (SRC) structure of the CAP1400 nuclear power plant under the influence of different high-mode vibration, the 1/7 model structure was manufactured and its dynamic characteristics was tested. Secondly, the finite element model of SRC frame-bent structure was established, the seismic response was analyzed by mode-superposition response spectrum method. Taking the combination result of the 500 vibration modes as the standard, the error of the base reactions, inter-story drift, bending moment and shear of different modes were calculated. Then, based on the results, the influence of high-mode vibration on the seismic response of the SRC frame-bent structure of the main plant was analyzed. The results show that when the 34 vibration modes were intercepted, the mass participation coefficient of the vertical and horizontal vibration mode was above 90%, which can meet the requirements of design code. There is a large error between the seismic response calculated by the 34 and 500 vibration modes, and the error decreases as the number of modes increases. When 60 modes were selected, the error can be reduced to about 1%. The error of the maximum bottom moment of the bottom column appeared in the position of the bent column. Finally, according to the characteristics of the seismic influence coefficient αj of each mode, the mode contribution coefficient γj•Xji was defined to reflect the contribution of each mode to the seismic action.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.544-551
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• 2020

Reactor control is a standout amongst the most vital issues in the nuclear power plant. In this paper, the optimal design of variable structure controller (VSC) based on the lightning search algorithm (LSA) is proposed for a nuclear reactor power system. The LSA is a new optimization algorithm. It is used to find the optimal parameters of the VSC instead of the trial and error method or experts of the designer. The proposed algorithm is used for the tuning of the feedback gains and the sliding equation gains of the VSC to prove a good performance. Furthermore, the parameters of the VSC are tuned by the genetic algorithm (GA). Simulation tests are carried out to verify the performance and robustness of the proposed LSA-based VSC compared with GA-based VSC. The results prove the high performance and the superiority of VSC based on LSA compared with VSC based on GA.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1210-1216
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• 2018

Site or multi-unit (MU) risk assessment has been a major issue in the field of nuclear safety study since the Fukushima accident in 2011. There have been few methods or experiences for MU risk assessment because the Fukushima accident was the first real MU accident and before the accident, there was little expectation of the possibility that an MU accident will occur. In addition to the lack of experience of MU risk assessment, since an MU nuclear power plant site is usually very complex to analyze as a whole, it was considered that a systematic method such as probabilistic safety assessment (PSA) is difficult to apply to MU risk assessment. This paper proposes a new MU risk assessment methodology by using the conventional PSA methodology which is widely used in nuclear power plant risk assessment. The logical failure structure of a site with multiple units is suggested from the definition of site risk, and a decomposition method is applied to identify specific MU failure scenarios.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2018.05a
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• pp.60-61
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• 2018

Finishing materials are very important to restrain fire spread from a compartment to another in a fire situation. Therefore, the evaluation of combustible properties for the combustible material is essential to apply finishing materials into a generic buillding or s special occupancy structure. In this study, the testing methods for evaluation of combustible performance of finishing materials of domestic were surveyed in order to prepare the guideline of application of finishing materials in nulear power plant.

• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 1999.10a
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• pp.186-192
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• 1999

Safety-related equipments of nuclear power plant must be seismically qualified to demonstrate their ability to function as required during and/or after the earthquake, The seismic qualification is usually achieved through analysis and testing. Analysis method is preferably adopted for structurally simple equipments which are easy to be mathematically modeled. However even for relatively complex equipments analysis method is occasionally used for computing the input motion or supporting information for the component test followed. Electrical cabinet is a typical example for which analysis method is combinedly used with test to get modal properties of the enclosing cabinet structure. Usually the structural elements and doors of the cabinet are loosely interconnected with small-size bolts or spot welding. Therefore cabinet-type equipment usually has high and complex nonlinear properties which are not easily idealized by simple practical modeling techniques. in this paper with respect to a typical cabinet-type structure(instrumentation cabinet of nuclear power plant) a comparative study has been performed between three different state-of-the -art modeling techniques: lumped mass model frame model and FEM modal. Form the study results it has been found that modal properties of the cabinet-type structure in the elastic behavior range can be reasonably computed through any type of modeling techniques in the practice with slight modification of model properties to get better accuracy. However it needs additional modeling techniques to get reasonable results up to nonlinear range.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.2
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• pp.215-223
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• 2021

The domestic CANDU nuclear power plants have been operated for a long time and various unforeseen spent fuel defects have been discovered. As the spent fuel defects are important factors in the safety of the nuclear power plant, a study on the analysis of the spent fuel defects to prevent their recurrence is necessary. However, in cases where the fuel rods inside the fuel assembly are defected, it is difficult to dismantle the fuel assembly owing to their welded structure and the facility conditions of the plant. Therefore, it is impossible to analyze the spent fuel defect because it is difficult to visually check the shape of the fuel defect. To resolve these problems, an analysis technology that can predict the number of defected fuel rods and defect size was developed. In this study, we developed a methodology for investigating the root cause of spent fuel defects using a database of the earlier fuel defects in the plants. It is anticipated that in the future this analysis technology will be applied when spent fuel defects occur.

• Journal of the Earthquake Engineering Society of Korea
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• v.25 no.3
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• pp.137-144
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• 2021

In this study, the seismic safety of nuclear power plant structures is evaluated and verified by performing a vibration test on a relatively simple shear wall structure. The shear walls are the prominent members of nuclear power plants and resist the seismic load. The shear wall structure is designed and manufactured to perform shaking table tests and is used to increase the accuracy of the analytical method by comparing them with the numerical analysis results. Different results will be checked and more efficient application methods will be studied depending on the method of designing reinforced concrete structures.

• Proceedings of the KIEE Conference
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• 1996.07b
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• pp.1275-1277
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• 1996

The Sequence Of Event (SOE) system used in nuclear power plants is a part of the Plant Data Acquisition System (PDAS). The SOE system of the existing nuclear power plant share the computer H/W and S/W, and required more complicated structure to provide the events or trip signals. Moreover, there are high potential of collision between synchronization signals and data transmitted to the Plant Computer System(PCS) when the synchronization signals are sent from PCS to the three SOE processors. When this collision happens the SOE system will break down, thus it is not possible to analyze the trend of events or trips. This paper issued the limitation items of the existing SOE system and proposed the revised SOE system.

• Journal of the Korean Society of Systems Engineering
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• v.15 no.1
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• pp.1-8
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• 2019

Reverse engineering involves examining a system or component so as to comprehend its structure, functionality, and operation. Creation of a system model in reverse engineering can serve several purposes: test generation, change impact analysis, and the creation of a new or modified system. When attempting to reverse engineering a system, often the most readily accessible information is the system description, which does not readily lend itself to use in Model Based System Engineering (MBSE). Therefore, it is necessary to be able to transform this description into a diagram, which clearly depicts the behavior of the system as well as the interaction between components. This study demonstrates how sequence diagrams can be extracted from the systems description. Using MBSE software, the sequence diagrams for the Engineered Safety Features Component Control System (ESF-CCS) of the Nuclear Power Plant are created. Sequence diagrams are chosen because they are a means of representing the systems behavior and the interaction between components. In addition, from these diagrams, the system's functional requirements can be elicited. These diagrams then serve as the baseline of the reverse engineering process and multiple system views are subsequently be created from them, thus speeding up the development process. In addition, the use of MBSE ensures that any additional information obtained from auxiliary sources can then be input into the system model, ensuring data consistency.

• Journal of the Earthquake Engineering Society of Korea
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• v.17 no.2
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• pp.53-59
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• 2013

Several researches have been studied to enhance the seismic performance of nuclear power plants (NPPs) by application of seismic isolation. If a seismic base isolation system is applied to NPPs, seismic performance of nuclear power plants should be reevaluated considering the soil-structure interaction effect. The seismic fragility analysis method has been used as a quantitative seismic safety evaluation method for the NPP structures and equipment. In this study, the seismic performance of an isolated NPP is evaluated by seismic fragility curves considering the soil-structure interaction effect. The designed seismic isolation is introduced to a containment building of Shin-Kori NPP which is KSNP (Korean Standard Nuclear Power Plant), to improve its seismic performance. The seismic analysis is performed considering the soil-structure interaction effect by using the linearized model of seismic isolation with SASSI (System for Analysis of Soil-Structure Interaction) program. Finally, the seismic fragility is evaluated based on soil-isolation-structure interaction analysis results.