• Journal of Electrical Engineering and Technology
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• v.12 no.2
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• pp.959-968
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• 2017

Electrical safety monitoring System (ESMS) is a critical component in modern power systems, which is characterized by large-scale access points, massive users and versatile requirements. For convenience of the information integration and analysis, the software development, maintenance, and application in the system, the cloud platform based ESMS is established and assessed in this paper. Firstly the framework of the system is proposed, and then the assessment scheme with a set of evaluation indices are presented, by which the appropriate cloud product can be chosen to meet the requirements of a specific application. Moreover, to calculate the weights of the evaluation indices under uncertainty, an improved interval AHP method is adopted to take into consideration of the fuzziness of expert scoring, the qualitative consistency test, and the two normalizations in the process of eigenvectors. Case studies have been made to verify the feasibility of the assessment approach for ESMS.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.692-701
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• 2019

Recently, instrumentation and control (I&C) systems in nuclear power plants have undergone digitalization. Owing to the unique characteristics of digital I&C systems, the reliability analysis of digital systems has become an important element of probabilistic safety assessment (PSA). In a reliability analysis of digital systems, fault-tolerant techniques and their effectiveness must be considered. A fault injection experiment was performed on a safety-critical digital I&C system developed for nuclear power plants to evaluate the effectiveness of fault-tolerant techniques implemented in the target system. A software-implemented fault injection in which faults were injected into the memory area was used based on the assumption that all faults in the target system will be reflected in the faults in the memory. To reduce the number of required fault injection experiments, the memory assigned to the target software was analyzed. In addition, to observe the effect of the fault detection coverage of fault-tolerant techniques, a PSA model was developed. The analysis of the experimental result also can be used to identify weak points of fault-tolerant techniques for capability improvement of fault-tolerant techniques

• Journal of the Korean Society for Railway
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• v.12 no.6
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• pp.1043-1048
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• 2009

PES (Programmable Electronic System) is used by software development for the train control system. PES has been widely used in real world and consists of hardware, firmware and application software. The PES are easily apply to many applications because its implementation has high flexibility. Many safety critical functions are realized through software in safety critical system. Normally, it is difficult to detect failures for PES system because the PES is too sophisticated to identify sources of the failure. So, the reliability analysis is needed by using software fault tolerance techniques. Currently, there are the recovery block, distributed recovery block, N-version programming, N self-checking programming in fault tolerance techniques. In this paper, the models of recovery block and N-version programming in software fault tolerance techniques are suggested by using the Markov model. Also, the reliability in the train control system is analyzed through changing time. The fault occupancy rates of the program, adjustment test and voter are stationary. So, the relation between time and reliability is presented by using Matlab program. In the result of reliability, the reliability of recovery block is more high than N-version programming in case of the same number of substitution block.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.435-440
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• 1996

Large Loss-of-Coolant Accidents analyzed in Final Safety Analysis Reports are usually covered by Reactor Inlet Header. Reactor Outlet Header and Primary Pump Suction breaks as representative cases. In this study we analyze the total (guillotine) break of an Emergency Core Cooling System (ECCS) pipe located at the ECCS injection point into the Primary Heat Transport System (PHTS). It was expected that thermal-hydraulic behaviors in the PHT and ECC systems are different from those of a Reactor Inlet Header break, having an equivalent break size. The main purpose of this study is to get insights on the differences occurred between the two cases and to assess these differences from the phenomenon behavior point of view. It was also investigated whether the ECCS line break analysis results could be covered by header break analysis results. The study reveals that as the intact loop has almost the same behavior in both analyzed cases. broken loop behavior is different mostly regarding sheath temperature in the critical core pass and pressure decrease in the broken Reactor Inlet Header. Differences are also met in the ECCS behavior and in event sequences timings.

• Journal of the Korean Society of Systems Engineering
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• v.12 no.2
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• pp.81-90
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• 2016

The system performance of Engineered Safety Features is of utmost importance in a nuclear power plant. The human performance is identified as most critical to assurance of the optimal operability of safety systems during an emergency. The aim of this study is to determine how the performance of safety system could be evaluated using Augmented Reality technology. The paper presents a description of how a systems engineered approach could be used to develop the necessary operating conditions needed to conduct this measurement. Augmented Virtual Reality (AVR) interface technology is achieving ease of availability and widespread use in many applications today as illustrated by the launch of several AR and VR devices aimed at media consumption. As such, environments that incorporate such AVR hardware have become invaluable tools in designing human interface systems because of the high fidelity and intuitive response to natural human interaction that can be achieved [2]. The outcome of the measurement undertaken is to determine whether 1.) Operator(s) performance can be enhanced by introducing an improved cognitive method of monitoring plant information during an Emergency Operating Procedures (EOP) and 2.) In correlation, inform the performance of the diverse safety systems on the basis of human factors.

• Journal of Computing Science and Engineering
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• v.11 no.1
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• pp.9-23
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• 2017

NuDE 2.0 (Nuclear Development Environment 2.0) is a formal-method-based software development, verification and safety analysis environment for safety-critical digital I&Cs implemented with programmable logic controller (PLC) and field-programmable gate array (FPGA). It simultaneously develops PLC/FPGA software implementations from one requirement/design specification and also helps most of the development, verification, and safety analysis to be performed mechanically and in sequence. The NuDE 2.0 now consists of 25 CASE tools and also includes an in-depth solution for indirect commercial off-the-shelf (COTS) software dedication of new FPGA-based digital I&Cs. We expect that the NuDE 2.0 will be widely used as a means of diversifying software design/implementation and model-based software development methodology.

• Journal of the Architectural Institute of Korea Planning & Design
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• v.35 no.4
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• pp.37-44
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• 2019

According to the statistical surveys and studies, insufficient maintenance in the use of existing buildings caused fire and collapse accidents. In this respect, I analyzed the data managed by the current building maintenance and inspection system to find out the actual state of safety management and proposed two significant results. First, regarding the state of the buildings, the safety management status of the small-sized ones, where 20 years or more passed after construction, is the worst and a priority improvement plan is required. Second, there are eight deeply concerning factors for the fire incidents and collapse accidents of buildings. In the order of high risk, these factors are structural strength (seismic design), exterior wall finishing material, basement floor, interior finishing materials, other evacuation facilities, corridors stairs entrances, rooftop, fire partition. We need to have more special designs and management plans regarding high-risk factors as a system to prevent accidents in the building.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2147-2155
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• 2022

Nuclear safety-related underground liquid storage tanks, such as those used to store fuel for emergency diesel generators, are critical components for safety of hundreds of existing nuclear power plants (NPP) worldwide. Since most of those NPP will continue to operate for decades, a beyond design base (BDB) seismic screening of safety-related underground tanks in those NPP is beneficial and essential to public safety. The analytical methodology for buried tank subjected to seismic effect, including a BDB seismic evaluation, needs to consider both soil-structure and fluid-structure interaction effects. Comprehensive analysis of such a soil-structure-fluid system is costly and time consuming, often subjected to availability of state-of-art finite element tools. Simple, but practically and reasonably accurate techniques for seismic evaluation of underground liquid storage tanks have not been established. In this study, a mechanics based solution is proposed for the evaluation of a cylindrical underground liquid storage tank using hand calculation methods. For validation, a practical example of two underground diesel fuel tanks in an existing nuclear power plant is presented and application of the proposed method is confirmed by using published results of the computer-aided System for Analysis of Soil Structural Interaction (SASSI). The proposed approach provides an easy to use tool for BDB seismic assessment prior to making decision of applying more costly technique by owner of the nuclear facility.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.676-682
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• 2019

After the Fukushima Daiichi accident, there has been an increasing preference for passive safety features in the nuclear power industry. Some passive safety systems require limited active components to trigger subsequent passive operation. Under very serious accident conditions, passive safety features could be rendered inoperable or damaged. This study evaluates (i) the performance and effectiveness of the passive safety features of iPOWER (innovative Power Reactor), and (ii) whether a severe accident condition could be reached if the passive safety systems are damaged, namely the case of heat exchanger tube rupture. Analysis results show that the reactor coolant system remains in the hot shutdown condition without operator actions or electricity for over 72 h when the passive auxiliary feedwater systems (PAFSs) are operable without damage. However, heat exchanger tube rupture in the PAFS leads to core damage after about 18 h. Such results demonstrate that, to enhance the safety of iPOWER, maintaining the integrity of the PAFS is critical, and therefore additional protections for PAFS are necessary. To improve the reliability of iPOWER, additional battery sets are necessary for the passive safety systems using limited active components for accident mitigation under such extreme circumstances.

• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.32 no.7
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• pp.91-97
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• 2004

The first Korean liquid propellant rocket KSR-III successfully finished its flight test on Nov. 28, 2002. Herein, we summarize the results of range safety system operation which is employed for the first time in flight tests of rockets developed by Korea Aerospace Research Institute(KARI). During the flight, safety-critical flight data including instantaneous impact points are monitored in realtime by range safety officers utilizing Range Safety Display Systems. The recorded screen of the display system is presented for the explanation of safety operation. In addition, comparisons are made between onboard navigation system based and radar based results in calculating instantaneous impact points, and also errors from the finally recorded impact point are described.