• Proceedings of the Korean Institute of Building Construction Conference
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• 2020.06a
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• pp.113-114
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• 2020

In the design and maintenance of buildings, identifying the degree of damage in the event of a fire is an important factor in fire prevention and fire safety design. In order to predict fire damage, safety measures should be established by predicting the nature of evacuation according to fire, smoke and in-house characteristics, and the effects of the operation of fire safety facilities should also be considered, but in Korea, the risk analysis due to the operation of fire safety facilities is insufficient. Accordingly, this study uses fire statistics and sprinkler inspection data to analyze the degree of fire damage caused by the operation of sprinkler facilities in a probabilistic manner.

• IE interfaces
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• v.25 no.4
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• pp.382-392
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• 2012

Reactor protection system to keep nuclear safety and operational economy of plants requires high reliability. Such a high reliability of the system can be achieved through the redundant design of components. However, common cause failures of components reduce the benefits of redundant design. Thus, the common cause failure analysis, to accurately calculate the reliability of the reactor protection system, is carried out using alpha-factor model. Analysis results to 24 operating months are that 1) the system reliability satisfies the reliability goal of EPRI-URD and 2) the common cause failure contributes 90% of the system unreliability. The uncertainty analysis using alpha factor parameters of 0.05 and 0.95 quantile values shows significantly large difference in the system unreliability.

• Journal of the Korean Society for Railway
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• v.8 no.2
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• pp.161-169
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• 2005

In this paper, we provide safety requirements and advices to guarantee the safety of an interface in a level crossing system which is an interface between the conventional facilities and the new ATP (Automatic Train Protection) system, as well as we accomplish a safety management for the facilities of a country that has a different standard with already standardized ATP system. The system model has been made based on a safety activity of the international standard, and then a tolerance of a risk by the safety activity through PHA (Preliminary Hazard Analysis) has been analyzed. finally we achieved HIA (Hazard Identification and Analysis) for the assumptions that have been produced from a operating scenario and a functional interface. Thus, the safety requirements for the interface has been provided from the safety plan of HIA, and we showed the safety activity to guarantee the system safety through HIA which was depend on the design.

• Nuclear Engineering and Technology
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• v.16 no.2
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• pp.106-115
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• 1984

Core safety limits define reactor operating conditions and parameters that will assure fuel rod and reactor system's integrity. Limiting safety system settings (LSSS) programmed into reactor protection system (RPS) then ensure a rapid reactor trip to prevent or suppress conditions which might violate the core safety limits. Generation of the LSSS must properly take into account uncertainties in both calculated and measured parameters in order to assure, with an appropriate degree of confidence, that the RPS will protect the core safety limits. Reviewed in this report are Westinghouse RPS setpoint generation philosophy, methodology of safety limit development and LSSS generation procedure. The Westinghouse RPS trip setpoint generation methodology has been established based on the calculation of core safety limits and the selection of LSSS allowing appropriate uncertainties in a conservative manner. Such conservative values of setpoint assure a high degree of core protection against fuel melting and occurrence of DNB.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.45-57
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• 1990

A steady-state margin comparison study was performed between analog and digital protection systems. The systems compared are the thermal overpower and overtemperature delta T system of Westinghouse, and Core Protection Calculator System of Combustion Engineering, Inc. No dynamic offset was considered to eliminate the margin differences by different safety analysis methodologies. The result shows that the digital protection system has about 30% more rated power margin than the analog system in protecting against the fuel rod centerline melting. The digital protection system is shown to have almost same margin with the analog protection system in preventing the DNB at EOC (End of Cycle) even if the digital protection system has about 10% more margin at BOC(Beginning of Cycle).

• Fire Science and Engineering
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• v.30 no.4
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• pp.103-110
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• 2016

While the demand for the gas system fire extinguishers increases every year, there are insufficient safety measures for assessing the extinguishing performance, such as system safety and reliability in the preparation of increasing demand, which has emerged as a social problem. One of the most critical causes of accidents occurring with the gas extinguishing system is pressure leakage from the extinguishing agent storage container. This is considered to be one of the critical factors on which the success of fire suppression depends. In this study, its safety measure was studied, Because it was deemed urgently necessary. The newly developed pressure leakage monitoring system is a system monitoring storage condition, pressure, leakage and discharge of the storage container related to agent concentration, which is one of the critical factors for fire suppression. This was developed to be applicable to the $CO_2$ and HFC-23 systems. Therefore, for structural safety analysis, the safety performance was verified by the fluid structure coupling analysis of the safety problems that may occur when the pressure leakage monitoring system is applied to the gas fire extinguisher. For analysis programs, the FloEFD program from Mentor Graphics was used for computational fluid dynamics analysis and ABAQUS from Dassault Systems was used for structural analysis. From the result of numerical analysis, the structure of $CO_2$ did not develop plastic deformation and its safety was verified. However, plastic deformation and deviation issue occurred with the HFC-23 monitoring system and therefore verified the structural safety of pressure leakage monitoring system by data obtained from redesigning and adjusting the condition of numerical interpretation three times.

• Proceedings of KOSOMES biannual meeting
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• 2005.05a
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• pp.23-29
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• 2005

Our country marine transport enterprise the international from competitive power from the hazard which develops at marine transport powerful country the government and marine transport presentation with activation of safety management protection of life Protection and oceanic environment from the sea hazard at the place where compared to it is endeavoring more the activation plan so far regarding the safety management from the coastal line insufficient are and the system and actuality safety management activation plan which is suitable in the coastal line is necessary. From the research which it sees consequently today a result analysis of coastal line safety management and the development direction of activation plan importance from it let in the immediacy which from the marine transport enterprise is important with the international and oceanic environmental protection it presented.

• Journal of KIISE:Software and Applications
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• v.31 no.6
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• pp.750-759
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• 2004

The software of the nuclear power plant digital control system is a safety-critical system where many techniques must be applied to it in order to preserve safety in the whole system. Formal specifications especially allow the system to be clearly and completely specified in the early requirements specification phase therefore making it a trusted method for increasing safety. In this paper, we discuss the NuSCR, which is a qualified formal specification method for specifying nuclear power plant digital control system software requirements. To investigate the application of NuSCR, we introduce the experience of using NuSCR in formally specifying the plant protection system's software requirements, which is presently being developed at KNICS. Case study that shows that the formal specification approach NuSCR is very much qualified and specialized for the nuclear domain is also shown.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.1313-1314
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• 2005

• Fire Science and Engineering
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• v.26 no.6
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• pp.38-44
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• 2012

The purpose of fire protection for the nuclear power plants (NPPs) is to ensure safe shutdown state of the reactor, to minimize the release of radioactive materials to the environment, to provide physical safety of the on-site personnel, and to limit the property damage. Fire protection and extinguishing equipments are one of the important protection measures based on the defense-in-depth concept, which can promptly detect and control and extinguish those fires that do occur, thereby limiting fire damage. However, a separate evaluation process might be additionally necessary for the construction permit and operating license because the fire protection laws of the NEMA for installation standards of the fire protection systems is not fully characterized for the NPPs. It is also not easy to implement the regulations such as the performance based design concept for fire protection system of the NPPs which are characterized for a relatively low density of employee. This study suggests a guideline for the improvement of the technical standards for fire protection systems of the NPPs by evaluating the fundamental problems drawn by reviewing laws and regulatory guides relevant to fire protection and by evaluating the applicability of the KEPIC FPN in domestic nuclear power plants.