• Journal of Korean Association for Spatial Structures
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• v.24 no.3
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• pp.43-51
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• 2024

Dynamic responses of nuclear power plant structure subjected to earthquake loads should be carefully investigated for safety. Because nuclear power plant structure are usually constructed by material of reinforced concrete, the aging deterioration of R.C. have no small effect on structural behavior of nuclear power plant structure. Therefore, aging deterioration of R.C. nuclear power plant structure should be considered for exact prediction of seismic responses of the structure. In this study, a machine learning model for seismic response prediction of nuclear power plant structure was developed by considering aging deterioration. The OPR-1000 was selected as an example structure for numerical simulation. The OPR-1000 was originally designated as the Korean Standard Nuclear Power Plant (KSNP), and was re-designated as the OPR-1000 in 2005 for foreign sales. 500 artificial ground motions were generated based on site characteristics of Korea. Elastic modulus, damping ratio, poisson's ratio and density were selected to consider material property variation due to aging deterioration. Six machine learning algorithms such as, Decision Tree (DT), Random Forest (RF), Support Vector Machine (SVM), K-Nearest Neighbor (KNN), Artificial Neural Networks (ANN), eXtreme Gradient Boosting (XGBoost), were used t o construct seispic response prediction model. 13 intensity measures and 4 material properties were used input parameters of the training database. Performance evaluation was performed using metrics like root mean square error, mean square error, mean absolute error, and coefficient of determination. The optimization of hyperparameters was achieved through k-fold cross-validation and grid search techniques. The analysis results show that neural networks present good prediction performance considering aging deterioration.

• Proceedings of the KIEE Conference
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• 1997.11a
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• pp.588-590
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• 1997

In this paper, a highly reliable communication network for DCS in nuclear power plant is designed. The structure and characteristics of DCS in nuclear power plant is briefly explained. The features needed for a communication network for DCS in nuclear power plant is described. According to the abovo features, the layer structure of the communication network is determined and each layer is designed in detail.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.860-868
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• 2018

This article presents a new design of earthquake instrumentation that is suitable for quick decision-making after the seismic event at the nuclear power plant (NPP). The main objective of this work is to ensure more availability of the NPP by expediting walk-down period when the seismic wave is incident. In general, the decision-making to restart the NPP after the seismic event requires more than 1 month if an earthquake exceeds operating basis earthquake level. It affects to the plant availability significantly. Unnecessary shutdown can be skipped through quick assessments of operating basis earthquake, safe shutdown earthquake events, and damage status to structure, system, and components. Multidecision parameters such as cumulative absolute velocity, peak ground acceleration, Modified Mercalli Intensity Scale, floor response spectrum, and cumulative fatigue are discussed. The implementation scope on the field-programmable gate array platform of this work is limited to cumulative absolute velocity, peak ground acceleration, and Modified Mercalli Intensity. It can ensure better availability of the plant through integrated decision-making process by automatic assessment of NPP structure, system, and components.

• Nuclear Engineering and Technology
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• v.37 no.2
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• pp.191-200
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• 2005

Shaking table tests of the seismic behavior of a steel frame structure model were performed. The purpose of these tests was to estimate the effects of a near-fault ground motion and a scenario earthquake based on a probabilistic seismic hazard analysis for nuclear power plant structures. Three representative kinds of earthquake ground motions were used for the input motions: the design earthquake ground motion for the Korean nuclear power plants, the scenario earthquakes for Korean nuclear power plant sites, and the near-fault earthquake record from the Chi-Chi earthquake. The probability-based scenario earthquakes were developed for the Korean nuclear power plant sites using the PSHA data. A 4-story steel frame structure was fabricated to perform the tests. Test results showed that the high frequency ground motions of the scenario earthquake did not damage the structure at the nuclear power plant site; however, the ground motions had a serious effect on the equipment installed on the high floors of the building. This shows that the design earthquake is not conservative enough to demonstrate the actual danger to safety related nuclear power plant equipment.

• 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.

• Proceedings of the KIEE Conference
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• 1998.07b
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• pp.816-818
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• 1998

In this paper, a highly reliable communication network for DCS (Distributed Control System) in nuclear power plant is designed. The structure and characteristics of DCS in nuclear power plant is briefly explained. The features needed for a communication network for DCS in nuclear power plant is described. According to the above features, a layer structure for the communication network is determined and each layer is designed in detail. Accuracy of the model was evaluated by computer simulation.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2019.05a
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• pp.230-231
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• 2019

The nuclear power plant construction project is large-scale, has various stakeholders and computer system, and is produced and managed a large amount of information. The domestic nuclear power plant construction project has accumulated data based on many years of experience in the system. It has the competitiveness to suggeest alternatives that meet the requirement of the client in the overseas nuclear power plant project. Earned value management, which integrates schedule and cost, is possible to risk management. It was developed the earned value management system considering the the actual data properties and types of the preceding nuclear power plant construction projects. It will be able to increase the nuclaer power plant export competieiveness. Therefore, it was palned to carry out future stuies so that it be able to complement the measure to integrate cost and schedule in consideration of actual data(quantity of activity, etc.).

• Proceedings of the Computational Structural Engineering Institute Conference
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• 1992.10a
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• pp.148-155
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• 1992

A method to verify seismic qualification of the plant protection system cabinet for a nuclear power plant is presented. A finite element model of the cabinet is developed and correlated to the dynamic properties observed during in-situ vibration test of the actual structure. The results of the modal analysis provide insight into the fundamental dynamic properties of the structure. Techniques for verifying structural integrity and operability are exemplified by summarizing response spectrum and time history analyses of the structure.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2005.04a
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• pp.183-189
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• 2005

Nuclear containment structure is the last barrier for being secure from any nuclear power plant accident. Even though the safety requirements of nuclear power plant have been focused on removing accidental situations, nuclear containment structure must reserve the sufficient resisting capacity to any accident because it works as the last barrier. The acceptable nuclear containment structure makes possible to limit the effect of internal accidents and to avoid radioactive release. In this study, to conduct the numerical analysis for the structural safety of a containment structure, loss of coolant accident (LOCA) is considered as the basic accidental load, and Wolsong containment structure is considered as a target structure. The CANDU containment structure, such as Wolsong containment structure, is a prestressed concrete shell structure which has dome and is reinforced with bonded tendons. The evaluation of ultimate pressure capacity was conducted by nonlinear analysis of a prestressed concrete containment structure.

• Journal of Mechanical Science and Technology
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• v.15 no.1
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• pp.10-20
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• 2001

The leak before break(LBB) concept is difficult to apply to a structure with a thin tube that is immersed in a water environment. A heat exchanger in a nuclear power plant is such a structure. The present paper addresses an application of the LBB concept to a heat exchanger in a nuclear power plant. The minimum leaked coolant amount(approximately 37.9 liters) containing the radioactive material which can activate the radiation detector device installed in near the heat exchanger is assumed. A postulated initial flaw size that can not grow to a critical flaw size within the time period to activate the radiation detector is justified. In this case, the radiation detector can activate the warning signal caused by coolant leakage from initially postulated flaws of the heat exchanger. The nuclear plant can safely shutdown when this occurs. Since the postulated initial flaw size can not grow to the critical flaw size, the structural integrity of the heat exchanger is not impeded. Particularly the informational scenario presented in this paper discusses an actual nuclear plant.