• Proceedings of the Korean Radioactive Waste Society Conference
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• 2019.05a
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• pp.213-214
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• 2019

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1577-1586
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• 2023

Safety qualification of a steam generator is a crucial issue related to faulted condition design loads, including earthquake loads, and it should be ensured that the structural integrity of a steam generator does not exceed its design load. Using data from the Gyeongju and Pohang earthquakes, the two most powerful recorded seismic events in Korea, seismic analyses of a typical steam generator are conducted in this study. The modal characteristics are used to develop an input deck for these analyses. With a time history analysis, the responses of the steam generator in the event of an earthquake are obtained. In particular, the displacement, velocity, and acceleration responses are obtained in the time domain, with these outcomes then used for a detailed structural analysis as part of the ensuing assessment. The response spectra are also generated to determine the response characteristics in the frequency domain, focusing on the response comparisons between the Gyeongju and Pohang earthquakes. Structural integrity can be ensured by performing additional analysis using results obtained from the time history analysis considering the input excitations of various earthquakes considered in the design.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.335-336
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• 2017

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.4
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• pp.523-531
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• 2022

The International Atomic Energy Agency (IAEA) entails independent decision-making for the safety supervision of civil nuclear facilities. To evaluate and review the safety of nuclear facilities, the national regulatory body usually consults independent institutions or external committees. Technical Support Organizations (TSOs) include national laboratories, research institutions, and consulting organizations. Support from professional organizations in other countries may also be required occasionally. Most of the world's major nuclear power countries adopt an independent nuclear safety supervision model. Accordingly, China has continuously improved upon the construction of such a system by establishing the National Nuclear Safety Administration (NNSA) as the decision-making department for nuclear and radiation safety supervision, six regional safety supervision stations, the Nuclear and Radiation Safety Center (NSC), a nuclear safety expert committee, and the National Nuclear and Radiation Safety Supervision Technology R&D Base, which serves as the test, verification, and R&D platform for providing consultation and technical support. An R&D system, however, remains to be formed. Future endeavors must focus on improving the technical support capacity of these systems. As an enhancement from institutional independence to capability independence is necessary for ensuring the independence of China's nuclear safety regulatory institution, its regulatory capacity must be improved in the future.

• Journal of the Korean Society of Safety
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• v.30 no.3
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• pp.114-122
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• 2015

In a fire hazard analysis (FHA) for nuclear power plant, various electrical circuit analyses are performed in the parts of fire loading analysis, fire modeling analysis, separation criteria analysis, associated circuit analysis, and multiple spurious operation analysis. Thus, electrical circuit analyses are very important areas so that reliability of the analysis results should be assured. This study is to establish essential electrical elements for each analysis for verification of the reliability of the electrical circuit analyses in the fire hazard analysis for nuclear power plants. Applying the results derived by the study to domestic nuclear power plants, it is expected to determine the adequacy of the fire hazard analysis report and contribute to the reliability of the fire hazard analysis of those plants.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.118-118
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.147-147
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.180-181
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.105.1-105
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• 2001

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2716-2727
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• 2021

The core shroud is one of the most important internal components of the reactor vessel internals because it meets the neutron fluence directly emitted by the nuclear fuel. In particular, dynamic effects for an earthquake should be evaluated with respect to the neutron irradiation flux. As a prerequisite to this study, simplified and detailed finite element models are developed for the core shroud using the ANSYS Design Parametric Language. Using the El Centro earthquake, seismic analyses are performed for the simplified and detailed core shroud models. Modal characteristics are obtained and their results are used for a time history analysis. Response spectrum analyses are also performed to access the degree of seismic excitation. The results of these analyses are compared to investigate the response characteristics between the simplified and detailed core shroud models from the time history and response spectrum analyses.