• Proceedings of the Korean Institute of Building Construction Conference
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• 2018.05a
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• pp.318-319
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• 2018

Protective coatings at nuclear power plants should be designed to withstand exposure to ambient conditions during normal operation or design-basis accidents. However, there was a change in the perception of the protective coating to the revision of the Regulatory Guidelines by the NRC in July 2000. In other words, maintenance guidelines have been strengthened in order to minimize the clogging of the cooling water system due to the substances in the containment building. Therefore, KHNP, the contractor and operator of the nuclear power plant, plans to develop the coating system for nuclear power plants in accordance with the regulation, and plans to develop its own coating expert.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2018.11a
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• pp.140-141
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• 2018

Protective coatings at nuclear power plants should be designed to withstand exposure to ambient conditions during normal operation or design-basis accidents. However, there was a change in the perception of the protective coating to the revision of the Regulatory Guidelines by the NRC in July 2000. In other words, maintenance guidelines have been strengthened in order to minimize the clogging of the cooling water system due to the substances in the containment building. Therefore, KHNP, the contractor and operator of the nuclear power plant, plans to develop the coating system for nuclear power plants in accordance with the regulation, and plans to develop its own coating expert.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2017.11a
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• pp.63-64
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• 2017

Protective coatings at nuclear power plants should be designed to withstand exposure to ambient conditions during normal operation or design-basis accidents. However, there was a change in the perception of the protective coating to the revision of the Regulatory Guidelines by the NRC in July 2000. In other words, maintenance guidelines have been strengthened in order to minimize the clogging of the cooling water system due to the substances in the containment building. Therefore, KHNP, the contractor and operator of the nuclear power plant, plans to develop the coating system for nuclear power plants in accordance with the regulation, and plans to develop its own coating expert.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.1
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• pp.1-7
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• 2019

In this paper, the regulations and technical standards related to qualification of safety grade equipment in nuclear power plants are critically reviewed with the qualification procedure in terms of structures, systems, and equipment in nuclear power plants. These facilities should be designed and constructed to protect from natural conditions or disasters and to perform their safety functions even in case of postulated accidents. Equipment Qualification is to demonstrate that the safety related equipment is designed and constructed to perform their safety functions under normal and accident conditions. It is classified into environmental qualification and seismic qualification.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1776-1783
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• 2019

To accurately analyze the accidents in nuclear reactors, a thermohydraulic-neutronic coupling calculation is required to solve fluid dynamics and nuclear reactor kinetics equations in fine cells simultaneously and evaluate the local effects of neutronic and thermohydraulic parameters on each other. In the present study, a 3D thermohydraulic-neutronic coupling model is developed, validated and then applied for Isfahan MNSR (Miniature Neutron Source reactor) safety analysis. The proposed model is developed using FLUENT software and user defined functions (UDF) are applied to simulate the neutronic behavior of MNSR. The validation of the proposed model is first evaluated using 1mk reactivity insertion experiment into Isfahan MNSR core. Then, the developed coupling code is applied for a design basis accident (DBA) scenario analysis with the insertion of maximum allowed cold core reactivity of 4 mk. The results show that the proposed model is able to predict the behavior of the reactor core under normal and accident conditions with a good accuracy.

• Fire Science and Engineering
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• v.34 no.2
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• pp.72-85
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• 2020

This article is a study on the activity of rescue workers for designing simulation training scenarios assuming chemical accidents. On the basis of the complexity of the indoor scene in the case of chemical accidents and terrorism, we designed a 12-step simulation training scenario for two teams to analyze the improvement in firefighters' capabilities. On the basis of activity measurement in the simulation scenario, step 2 of training had the most drops in the maximum heart rate, as follow: N1, from 163 bpm to 153 bpm; N2, from 186 bpm to 151 bpm; N3, from 168 bpm to 162 bpm; and N4, from 166 bpm to 152 bpm. In terms of intensity level in the allowable activity time, it was found that in step 2 both N1 and N2 reduced from Level 5 to Level 3, N3 remained at Level 4, N4 reduced from Level 4 to Level 3, and the maximum allowable activity time increased.

• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.310-314
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• 2016

Background: FCTC10 container is designed to transport $^{60}Co$ radioactive sources used in irradiation industry. It belongs to Type B(U) Category III (yellow) package when being loaded with a $^{60}Co$ source of $1.8{\times}10^5$ Ci. Materials and Methods: The container is constituted of shielding container, basket, protective cover and bracket. Shielding ability is provided mainly by stainless steel shells, tungsten alloy and lead among steel shells. Radiation level around the container has been calculated with both Monte Carlo simulations and measurements. Results and Discussion: It is proven that the shielding performance of the container fulfills the requirements in GB11806-2004 (Regulations for the safe transport of radioactive material, China Standard Press). Exposure doses to workers and to critical groups of public were calculated based on hypothetical exposure scene according to transport practice experience. Conclusion: The results show that doses to workers and public are less than the constraint dose considered in design, and the radiation level would be increased less than a factor of 2 under design basis accidents.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.287-298
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• 2007

As part of the $AP1000^{TM}$ pressurized water reactor design certification program, a series of integral systems tests of the nuclear steam supply system was performed at the APEX-1000 test facility at Oregon State University. These tests provided data necessary to validate Westinghouse safety analysis computer codes for AP1000 applications. In addition, the tests provided the opportunity to investigate the thermal-hydraulic phenomena expected to be important in AP1000 small-break loss of coolant accidents (SBLOCAs). The APEX-1000 facility is a 1/4-scale pressure and 1/4-scale height simulation of the AP1000 nuclear steam supply system and passive safety features. A series of eleven tests was performed in the APEX-1000 facility as part of a U.S. Department of Energy contract. In all, four SBLOCA tests representing a spectrum of break sizes and locations were simulated along with tests to study specific phenomena of interest. The focus of this paper is the SBLOCA tests. The key thermal-hydraulic phenomena simulated in the APEX-1000 tests, and the performance and interactions of the passive safety-related systems that can be investigated through the APEX-1000 facility, are emphasized. The APEX-1000 tests demonstrate that the AP1000 passive safety-related systems successfully combine to provide a continuous removal of core decay heat and the reactor core remains covered with considerable margin for all small-break LOCA events.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.755-763
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• 2020

The critical flow phenomenon has been studied because of its significant effect for design basis accidents in nuclear power plants. Transition points from thermal non-equilibrium to equilibrium are different according to the geometric effect on the critical flow. This study evaluates the uncertainty parameters of the critical flow model for analysis of DBA (Design Basis Accident) with the MARS-KS (Multi-dimensional Analysis for Reactor Safety-KINS Standard) code used as an independent regulatory assessment. The uncertainty of the critical flow model is represented by three parameters including the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio, and their ranges are determined using large-scale Marviken test data. The uncertainty range of the thermal non-equilibrium factor is updated by the MCDA (Model Calibration through Data Assimilation) method. The updated uncertainty range is confirmed using an LBLOCA (Large Break Loss of Coolant Accident) experiment in the LOFT (Loss of Fluid Test) facility. The uncertainty ranges are also used to calculate an LBLOCA of the APR (Advanced Power Reactor) 1400 NPP (Nuclear Power Plants), focusing on the effect of the PCT (Peak Cladding Temperature). The results reveal that break flow is strongly dependent on the degree of the thermal non-equilibrium state in a ruptured pipe with a small L/D ratio. Moreover, this study provides the method to handle the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio in the system code.

• Korean Institute of Interior Design Journal
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• v.26 no.1
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• pp.53-62
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• 2017

It is very important for the elderly to maintain their independence and to live safe and comfortable daily lives in order to lead an independent life. The bathroom in the house has the greatest number of fall accidents and has a great influence on the independence of the elderly. Therefore, the purpose of this study is to develop a bathroom design method that improves the independence of the elderly whose physical function is deteriorated through preventing falls. In this study, first of all, we investigated the causes of falling in bathrooms based on the previous studies, and identified the intelligent services that can prevent falls and selected them according to individual needs. Secondly, we investigated the bathroom size of the apartment among the various housing type, analyzed the type of the bathroom, and developed a standard type of the bathroom. Thirdly, we analyzed the design guidelines for the elderly residents in Korea and proposed improvement directions. We also divided the intelligent sanitary appliances and fall-prevention products and proposed the installation method and directions. Therefore, the comprehensive bathroom design standards and proposals proposed in this study will be the basis of bathroom design to prevent falls and improve independence of the elderly, and it will be easy to understand and help the designer in designing. Finally, this study is meaningful in that it provides independence of the elderly through fall prevention and increases the easiness of hygiene action and suggests direction to maintain self-sustaining life of the elderly.