• Fire Science and Engineering
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• v.30 no.4
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• pp.20-26
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• 2016

When the fire takes place in Nuclear Powr Plants(NPPs), the reactor should achieve and maintain safe shut-down conditions and minimize the radioactive material released to the environment. The U.S. Nuclear Regulatory Commission (NRC) has issued numerous generic communications related to fire protection over the past 20 years, after it issued its requirements in the Fire Protection Rule set forth in Title 10, Section 50.48 of the Code of Federal Regulations (10 CFR 50.48) and Appendix R to the 10 CFR 50. The and Nuclear Energy Institute (NEI) has developed a Methodology for Risk Informed Fire Safe-Shutdown Analysis, which is related to the Deterministic Method for Multiple Spurious Operations solutions. The aim of this study was to identify, achieve, and maintain Post-Fire Safe-Shutdown of the Main Control Room (MCR) of the CANDU reactor, even though one train of the multiple Safety Structures, Systems, and Components (SCCs) fail by the technical specification and analysis method.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2002.05a
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• pp.1210-1215
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• 2002

The spent fuel handling tool is used to handle the refuel bundle and treated by hoist rope on the bridge crane. The new developed handling tool of NPP(Nuclear Power Plant) should be conformed the structural stability under earthquake condition. In this study, the stress and seismic analysis of the handling tool are performed by finite element method. Using the Floor Response Spectrum(FRS) obtained through the time history analysis, the modal and seismic analysis under Operating Basis Earthquake(OBE) and Safe Shutdown Earthquake(SSE) load conditions are carried out. Total 4 cases of different locations of the trolly and the hook are investigated. With the spring-damper element, the tension analysis of hoist rope is conducted. The stability of handling tool under earthquake load condition is conformed with regulatory guide.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.129.2-129
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.991-992
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• 2005

• Proceedings of the KSME Conference
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• 2001.11a
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• pp.465-470
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• 2001

In this study, the seismic analysis of rack structure with fluid-structure interaction is performed through use of the Finite Element Method(FEM) code ANSYS. Fluid-structure interaction can specify in terms of an hydrodynamic effect which is defined as the added mass per unit length divided by the area of the cross section. Using the Floor Response Spectrum(FRS) obtained through the time-history analysis, modal analysis and seismic analysis under Operating Basis Earthquake(OBE) and Safe Shutdown Earthquake(SSE) condition is carried out. The fluid-structure interaction effects on the rack structure are investigated.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2011.10a
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• pp.480-485
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• 2011

In this study, the seismic analysis of power plant piping system was performed using finite element model. This study was performed by ANSYS 12.1. For qualification of power plant piping system, the response spectrum analysis was performed using the given operating basis earthquake(OBE) and safe shutdown earthquake(SSE) floor response spectrum. The maximum stresses of power plant piping system were 166 MPa under OBE condition and 281 MPa under SSE condition. Thus, it can shown that the structural integrity of tpower plant piping system has a stable structure for seismic load conditions.

• Fire Science and Engineering
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• v.32 no.1
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• pp.57-65
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• 2018

In this study, a post-fire safe shutdown analysis (PFSSA) including multiple spurious operation (MSO) treatments for cables was conducted with an imaginary nuclear power plant for training using a deterministic fire analysis code. The imaginary nuclear power plant for the training consisted of a reactor containment building and an auxiliary building, including a total of 22 fire areas. The equipment including valves, pumps, emergency diesel generators, switch gears, motor control centers, and logic controllers were located in each fire area of the imaginary plant. It was assumed that each equipment is connected with two cables and that each cable passes through the fire areas along the cable trays. A database containing the information on the equipment and cables for the imaginary plant was constructed for the fire area analysis. The fire area analysis was performed for several assumed MSO scenarios, equipment logics, and cable logics. A mitigation measure using a three hour rated wrap was applied to the failed cables and cable trays after the fire area analysis.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.24 no.3
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• pp.289-294
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• 2011

In this study, the seismic analysis of traveling sea water screen for nuclear power plant was performed using finite element model. For qualification of traveling sea water screen, the response spectrum analysis was performed using the given operating basis earthquake(OBE) and safe shutdown earthquake(SSE) floor response spectrum. Dynamic analysis of water sea traveling screen was performed using finite element method. The analytical maximum displacements of traveling sea water screen were 2.5 mm under OBE condition and 4.6 mm under SSE condition. The maximum stresses of traveling sea water screen were 24 MPa under OBE condition and 44 MPa under SSE condition, that this results were 18 %, 27% of yield strength of material. Thus, it can shown that the structural integrity of traveling sea water screen has a stable structure for seismic load conditions.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.654-659
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• 2020

The reliability of the spent fuel pool instrument is very important for the security of nuclear power plant, especially during the earthquake. The effect of the fluid force on the vibration characteristics of the flexible tube of the spent fuel pool instrument needs comprehensive analysis. In this paper, based on the potential flow theory, the hydrodynamic pressures acting on the flexible tube were obtained. A mathematical model of a flexible tube was constructed to obtain the dynamic response considering the effects of seismic load and fluid force, and a computer code was written. Based on the mathematical model and computer code, the maximum stresses of the flexible tube in both safe shutdown earthquake and operating basis earthquake events on the spent fuel pool with three typical water levels were calculated, respectively. The results show that the fluid force has an obvious effect on the stress and strain of the flexible tube in both safe shutdown earthquake and operating basis earthquake events.

• Journal of the Korean Society for Precision Engineering
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• v.22 no.12 s.177
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• pp.131-140
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• 2005

Responses of the cabinet of the reactor safety system under seismic leadings are analyzed, its dynamic characteristics and structural reliability being evaluated. Analyzed natural frequencies are compared with those measured from a resonance test. Structural safety of the cabinet is evaluated in consideration of the required response spectrums of the operation-base and safe-shutdown earthquakes. Transient responses of the cabinet are analyzed with input ground acceleration measured during the seismic test, accelerations being extracted at the locations of the main internal parts. The transient responses are compared with those from the seismic test, favorable results being shown.