• Journal of Korean Society of Disaster and Security
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• 제16권4호
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• pp.85-99
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• 2023

By the social advancement, radiological disaster prevention planning is getting important considering complex disasters as in the Fukushima radiological disaster occurred by a chain of natural disasters. However, it has yet to be suggested the specific prevention plans for the complex disasters in the field of national radiological disaster prevention. This study aims to analyze the types of complex disasters in order to select the ones that are relatively more likely to occur in the domestic environment. It is also to analyze the impact on the radiological disaster prevention by searching damage spread of the classified natural disasters. We provides the necessary criterial for establishing disaster prevention plans through the scenarios for radiological emergency responses based on complex disasters. it is thought that these criteria can help prepare for the worst case scenario and implement effective resident protection.

• Journal of the Korea Concrete Institute
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• 제15권1호
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• pp.60-68
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• 2003

Tension tests of six half-thickness concrete containment wall elements were conducted as a part of Korea Atomic Energy Research Institute (KAERI) program. The aim of the KAERI test program is to provide a test-verified analytical method for estimating capacities of concrete reactor containment buildings under internal overpressurization from postulated degraded core accidents. The data from the tests reported herein should be useful for benchmarking analytical method that require modeling of material behavior including concrete cracking behavior and reinforcement/concrete interaction exhibited by the test. Major test variable is compressive strength of concrete, and its effect on the behavior of prestressed concrete panel subjected to biaxial tension is investigated.

• Journal of the Korean Society for Nondestructive Testing
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• 제24권2호
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• pp.186-196
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• 2004

In each country, the periodic ISI(In-Service Inspection) is required by the law to protect the public health and property from the potential accident of the nuclear facilities. To support the implementation of ISI program, the prescriptive ISI technical standards have been established. As the key parts of the ISI program, the non-destructive examination techniques are widely used to identify the degree of degradation of the pressure boundary components and welds. Recently, the risk informed-ISI program has been developed and implemented in several countries. Nonetheless, the existing ISI program which prescriptively decides the scope of inspection still has its own significance. In this article, the technical standards of ISI in leading countries like US, france, Canada, and Japan are reviewed and compared with the safety guide by IAEA. An outline to revise the domestic technical standards of ISI has been suggested.

• The Journal of the Korea institute of electronic communication sciences
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• 제9권2호
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• pp.267-271
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• 2014

Nuclear power plant data networks and their associated safety systems are being modernized to include many information technology (IT) networks and applications. Along with the advancement of plant data networks (PDN), instrumentation and control systems are being upgraded with modern digital, microprocessor-based systems. However, nuclear PDN is confronted significant side-effects, which PDN is exposed to prevalent cyber threats typically found in IT environments. Therefore, cyber security vulnerabilities and possibilities of cyber incidents are dramatically increased in nuclear PDN. Consequently, it should be designed fully ensuring the PDN meet all reliability, performance and security requirements in order to overcome the disadvantages raised from adaption of IT technology. In this paper, we provide technical security criteria should be used in design and evaluation of secure PDN. It is believed PDN, which is designed and operated along with these technical security critera, effectively protect against possible outside cyber threats.

• Fire Science and Engineering
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• 제30권4호
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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.

• Nuclear Engineering and Technology
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• 제22권4호
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• pp.359-370
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• 1990

This study concerns with comparing low-dimensional reactor kinetics methods with a three-dimensional kinetics method to be used for safety analysis of light water reactors in order to suggest means of preparing input parameters required for low-dimensional methods. For this purpose a one-dimensional finite difference two-group diffusion theory code ODTRAN and a third-order Hermit polynomial-based point kinetics code POTRAN are developed and used to obtain low-dimensional solutions to the LRA-BWR kinetics benchmark problem. The results are compared with a three-dimensional modified Borresen's coarse-mesh solution of the kinetics problem by CMSNACK code. Through this comparison some simple but practical means of preparing input parameters of low-dimensional kinetics analysis methods are suggested.

• Journal of the Korean Society for Railway
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• 제12권5호
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• pp.744-750
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• 2009

According to the result of related studies, the degradation of human performance has been revealed as one of the most significant causes resulting in the safety of any human-involved system. This means that preventing the occurrence of accidents/incidents through avoiding the degradation of human performance is prerequisite for their successive operation. To this end, it is necessary to develop a plausible tool to evaluate the complexity of a task, which has been known as one of the decisive factors affecting the human performance. For this reason, in this paper, the complexity of tasks to be conducted by KTX drivers was quantified by TACOM measure that is enable to quantify the complexity of proceduralized tasks being used in nuclear power plants. After that, TACOM scores about the tasks of KTX drivers were compared with NASA-TLX scores that are responsible for the level of a subjective workload to be felt by KTX drivers.

• Nuclear Engineering and Technology
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• 제25권3호
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• pp.353-360
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• 1993

Many types of pipe whip restraints are installed to protect the structural components from the anticipated pipe whip phenomena of high energy lines in nuclear power plants. It is necessary to investigate these phenomena accurately in order to design the pipe whip restraints properly and/or to evaluate the acceptability of the pipe whip restraint design. Various research programs have been conducted in many countries to develop analytical methods and to verify the validity of the methods. In this study, various types of finite elements in ANSYS[1], the general purpose finite element computer program, was used to simulate the postulated pipe whips to obtain impact loads and the calculated results were compared with the specific experimental results from the sample pipe whip test for the U-shaped pipe whip restraints. Some calculational models, having the gap element or the spring element between the pipe whip restraint and the pipe line, give reasonably good transient responses of the restraint forces compared with the experimental results, and could be useful in evaluating the acceptability of the pipe whip restraint design.

• Plant Journal
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• 제12권4호
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• pp.37-43
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• 2016

Power system which provides electricity to the accident mitigation load for nuclear power plant should be verified to maintain the proper voltage level under the various loading and source conditions. For this purpose, it was needed to collect the voltage data of safety related buses during operation of the Reactor Coolant Pump(RCP) motor and Component Cooling Water Pump(CCWP) motor, respectively, under the certain loading condition of the plant. The data (such as, voltage, current, power factor) collected from actual measurement were used to modify the existing ETAP model and then the reanalysis was conducted to simulate the testing conditions. Through these actual measurement and analysis, it ensures that the existing electrical system analysis including assumptions and methods was conducted properly. Finally, the voltage of safety related buses was not dropped below the acceptable level, and the discrepancy between two results was within the limit.

• Korean Chemical Engineering Research
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• 제57권1호
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• pp.28-41
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• 2019

We carried out performance analysis of Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor. We analyzed transient-dynamic behavior of fluids inside the steam generator to vent into a sodium dump tank or a water dump tank when tubes in the steam generator were broken to cause a large-water-leak accident. Accordingly, we preliminarily evaluated design requirements of our system. Our results showed that sodium in the shell side of the steam generator and in Intermediate Heat Transport System was completely vented within 50 s and feed water in the tube side of the steam generator was completely vented within 2.5 s. It was analyzed that pressure of the tube side of the steam generator was higher than pressure of the shell side of the steam generator, which showed that sodium in the shell side did not flow into the tube side. Our results are expected to be used as basis information to performance analysis of Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor.