• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.231-236
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• 2011

When the Fukushima nuclear power plant accident occurred in March of 2011, a hydrogen explosion in the reactor building at the 4th unit of Fukushima plants led to a big surprise because the full core of the unit 4 reactor had been moved and stored underwater at the spent nuclear fuel storage pool for periodic maintenance. It was because the possible criticality in the fuel storage pool by coolant loss may yield more severe situation than the similar accident happened inside the reactor vessel. Fortunately, it was assured to be evitable to an anxious situation by a look of water filled in the storage pool later. In the paper, the safety state of the spent fuel storage pool and rack structures of the domestic nuclear plants would be roughly reviewed and compared with the Fukushima plant case by engineering viewpoint of potential severe accidents.

• Journal of the Korean Society of Radiology
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• v.9 no.3
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• pp.125-130
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• 2015

RI & RG are used in various field such as medical field, industrial field, agricultural and food&life field. The number of small nuclear facilities is on the increase. We need to take an interest in decommissioning of small nuclear facility and predict the occurring problem from facility decommissioning. Because of the relatively low radiation risk, the preparation of the small nuclear facility dismantling is often neglected. As the accident in Goiania, Brazil showed, the impact of the decommissioning of small nuclear facilities is not less than the large nuclear facilities although it may seem dangerless. Therefore, we analyzed the each institutional characteristics of the decommissioning of small nuclear facilities through foreign case study on this research. Also, we proposed several considerations on decommissioning such as reuse of facility and source, lack of space, stakeholder involvement and failures of protection. Through these study, we tried to make guideline of the small nuclear facilities decommissioning.

• Journal of the Korean Society of Radiology
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• v.6 no.1
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• pp.79-82
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• 2012

Iodine is one of important nuclides to be checked for radiation exposure after nuclear power facility accidents. After Chernobyl accident, it was observed that there is a greater amount of organic iodine in the atmosphere than inorganic iodine. In this study, we not only varied the amount of sample being exposed to $^{131}I$ and the duration of exposure to $^{131}I$ but also diluted the sample in distilled water and mixed the sample in kelp and liquid $^{131}I$ to measure and analyze the radiation detection levels. We concluded that the radiation levels were not high enough to be harmful to human body. The radiation from $^{131}I$ decreased over time, and we calculated the half life at 7-9 days. We found that the radiation from any sample containing $^{131}I$ was halved by up to 7days.

• Journal of Korean Society of societal Security
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• v.1 no.3
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• pp.69-76
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• 2008

The object of this study is to develop a tool for quantifying risks related to the rail transportation of hazardous commodities and to present mitigation measures. In this study, the Quantitative Risk Assessment (QRA) is used as a risk analysis tool. Based on the previous explosion history (Iri explosion) and consideration of its high risk, Iksan-si is selected as a model city. The result, expressed as average individual risk for exposed people with various distance, indicates that the model city is considered to be safe according to the nuclear energy standard. Also, the mitigation measures are provided since Societal risk of Iksan-si is set within ALARP. Risk reduction measures include rail car design, rail transportation operation, demage spread control as well as derail prevention and alternative routes for reducing accident frequencies. Finally, it is expected to achieve high level of public safety by appling the risk reduction measures.

• Journal of Radiation Protection and Research
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• v.22 no.4
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• pp.251-261
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• 1997

A package to transport the high level radioactive materials is required to withstand the hypothetical accident conditions as well as normal transport conditions according to IAEA standards and domestic regulations. The regulations require that the package should maintain the shielding, thermal and structural integrities to release no radioactive material. In general, safety evaluation of packages is performed by experimental methods using scale model and/or analytical methods using computer codes. This paper presents the safety evaluation of package to transport the radioisotopes produced in the HANARO to the radioisotope production facility. Radiation shielding, thermal and structural analyses were peformed using the computer codes. It has been verified that the package is safe under hypothetical accident conditions as well as normal transport conditions.

• Journal of Convergence Society for SMB
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• v.6 no.1
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• pp.7-15
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• 2016

As most devices in a wide range of automotive, aerospace, and missile have built-in software that controls the system behaviors, the safety of the software is growing in its importance. That is, the software safety has emerged as one of big issues because the threat of accidents caused by software malfunction is rising. Accident by software can be occurred from user mal-operation, but the fundamental reason of the accident comes from insufficient verification of the safety in software development process. Therefore, this paper presents how the software safety analysis and management activities should be done in the development process. In particular, we propose how to apply the safety analysis and management activities in the prototype or incremental development process.

• Nuclear Engineering and Technology
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• v.13 no.3
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• pp.121-129
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• 1981

Three most outstanding maximum flow rate formulas are identified from many existing models. Outlines of the three limiting mass flow rate models are given along with computational procedures to estimate approximate amount of fission products released from a containment to environment for a given characteristic hole size for containment-isolation failure and containment pressure and temperature under a loss of coolant accident. Sample calculations are performed using the critical ideal gas flow rate model and the Moody's graphs for the maximum two-phase flow rates, and the results are compared with the values obtained from the mass leakage rate formula of CONTEMPT-LT code for converging nozzle and sonic flow. It is shown that the critical ideal gas flow rate formula gives almost comparable results as one can obtain from the Moody's model. It is also found that a more conservative approach to estimate leakage rate from a containment under a LOCA is to use the maximum ideal gas flow rate equation rather than tile mass leakage rate formula of CONTEMPT-LT.

• Journal of the Korea institute for structural maintenance and inspection
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• v.24 no.3
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• pp.122-129
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• 2020

To conduct probabilistic seismic fragility analysis for nuclear power plants, it is very important to define the failure modes and criteria that can represent actual serious accidents. The seismic design criteria for piping systems, however, cannot fully reflect serious accidents because they are based on plastic collapse and cannot express leakage, which is the actual limit state. Therefore, it is necessary to clearly define the limit state for reliable probabilistic seismic fragility analysis. Therefore, in this study, the limit state of the SCH80 3-inch steel pipe elbow, the vulnerable part of piping systems, was defined as leakage, and the in-plane cyclic loading test was conducted. Moreover, an attempt was made to quantify the failure criteria for the steel pipe elbow using the damage index, which was based on the dissipated energy that used the moment-deformation angle relationship.

• Nuclear Engineering and Technology
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• v.25 no.4
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• pp.552-560
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• 1993

In this study, the detailed fuel assembly stress analysis model to evaluate the structural integrity for seismic and blowdown accidents is developed. For this purpose, as the first step, the program MAIN which identifies the worst bending mode shaped fuel assembly(FA) in core model is made. And the finite element model for stress calculation of FA components is developed. In the model the fuel rods (FRs) and the guide thimbles are modelled by 3-dimensional beam elements, and the spacer grid spring is modelled by a linear and relational spring. The constraints come from the results of the program MAIN. The stress analysis of the 16$\times$16 type FA under arbitary seismic load is performed using the developed program and modelling technique as an example. The developed stress model is helpful for the stress calculation of FA components for seismic and blowdown loads to evaluate the structural integrity of FA.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.2
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• pp.110-115
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• 2010

This study was conducted to estimate the feasibility using thermal image measurement that is applicable to internal leak diagnosis for the power plant valve. Abnormal heating of valve surface associated with high temperature steam f10w toward valve outlet side in the condition of low temperature is a primary indicator of leakage problems in high temperature and pressure valves. Thermal imaging enables to see the invisible thermal radiation that may portend impending damage before their condition becomes critical. When steam flow in valve outlet side in the condition of low temperature is converted into heat transmitted through the valve body due to the internal leakage in valve. The existence of abnormally increasable leakage rate in the valve will result in abnormally high levels of heat to be generated that can be quickly identified with a thermal image avoiding energy loss or damage of valve component. From the experimental results, it was suggested that the thermal image measurement could be an effective way to precisely diagnose and evaluate internal leak situation of valve.