• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.455-455
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.429.1-429
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.477.2-477.2
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• 2003

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.21-25
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• 2012

Functional Equipment Group(FEG) of Nuclear Power Plant(NPP) is the bundling of multiple equipment subcomponents around a major equipment with a common system isolation characteristic or out-of-service consideration. The main purpose of FEG is to reduce out-of-service time to enhance maintenance effectiveness. KHNP(Korea hydro & Nuclear Power Co.) needs to develop these FEGs to successfully perform On-line Maintenance(OLM) which will be expected to institute in the future. Recently, there has been growing interest in OLM in KHNP. The aim of this paper is to suggest the way of FEG development and to build the FEG of MCR(Main Control Room) HVAC(Heat, Ventilation and Air Conditioning) system. The results of this study might be make good use of OLM and be helpful for equipment maintenance for MCR HVAC system.

• Journal of Nuclear Fuel Cycle and Waste Technology
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• v.1 no.1
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• pp.83-92
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• 2013

A novel methodology to evaluate remote dismantling equipment for a reactor pressure vessel (RPV) in a decommissioning project is presented in this paper. The remote dismantling equipment, mainly composed of cutting tools and positioning equipment, is absolutely required to cut and handle highly radioactive and large components in nuclear power plants (NPPs); this equipment has a great effect on the overall success of the decommissioning project. Conventional evaluation methods have only focused on cutting technologies or positioning equipment, although remote dismantling equipment cannot achieve its goal without organic interaction between the cutting tools and the positioning equipment. In this paper, the cutting tools and the positioning equipment are evaluated by performance parameters according to their original characteristics, the relationship between the two systems, and common factors. Finally, the remote dismantling equipment used in recent decommissioning projects has been evaluated based on the proposed methodology. The results of this paper are expected to be useful for future decommissioning projects.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.109.1-109
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• 1999

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.51-58
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• 2006

Since 2002, more than 400 PWR spent nuclear fuel assemblies have been transported and stored on-site using transport casks in order to secure the storage capacity of PWR spent nuclear fuel of Kori nuclear power plant. The complete on-site transport system, which includes KN-12 transport casks, the related equipment and transport vehicles, had been developed and provided. KN-12 transport casks were designed, fabricated and licensed in accordance with Korean and IAEA's transport regulations, and the related equipment was also provided in accordance with the related regulations. The on-site transport and storage operation using two KN-12 casks and the related equipment has been conducted, and the strict Quality Control and Radiation Safety Management through the whole process has been carried out so as to achieve the required safety and reliability of the on-site transport of spent nuclear fuel.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2107-2119
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• 2022

Accurate remaining useful life (RUL) prediction for critical components of nuclear power equipment is an important way to realize aging management of nuclear power equipment. The electric gate valve is one of the most safety-critical and widely distributed mechanical equipment in nuclear power installations. However, the electric gate valve's extended service in nuclear installations causes aging and degradation induced by crack propagation and leakages. Hence, it is necessary to develop a robust RUL prediction method to evaluate its operating state. Although the particle filter(PF) algorithm and its variants can deal with this nonlinear problem effectively, they suffer from severe particle degeneracy and depletion, which leads to its sub-optimal performance. In this study, we combined the whale algorithm with regularized particle filtering(RPF) to rationalize the particle distribution before resampling, so as to solve the problem of particle degradation, and for valve RUL prediction. The valve's crack propagation is studied using the RPF approach, which takes the Paris Law as a condition function. The crack growth is observed and updated using the root-mean-square (RMS) signal collected from the acoustic emission sensor. At the same time, the proposed method is compared with other optimization algorithms, such as particle swarm optimization algorithm, and verified by the realistic valve aging experimental data. The conclusion shows that the proposed method can effectively predict and analyze the typical valve degradation patterns.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2682-2695
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• 2021

This paper aims to evaluate the structural dynamic responses and damage/failure of the nuclear fuel reprocessing plant under the free drop impact of spent fuel cask (SFC) and fuel assembly (FA) during the on-site transportation. At the present Part I of this paper, the large-scale SFC model free drop test and the corresponding numerical simulations are performed. Firstly, a composite target which is composed of the protective structure, i.e., a thin RC plate (representing the inverted U-shaped slab in the loading shaft) and/or an autoclaved aerated concrete (AAC) blocks sacrificial layer, as well as a thick RC plate (representing the bottom slab in the loading shaft) is designed and fabricated. Then, based on the large dropping tower, the free drop test of large-scale SFC model with the mass of 3 t is carried out from the height of 7 m-11 m. It indicates that the bottom slab in the loading shaft could not resist the free drop impact of SFC. The composite protective structure can effectively reduce the damage and vibrations of the bottom slab, and the inverted U-shaped slab could relieve the damage of the AAC blocks layer dramatically. Furthermore, based on the finite element (FE) program LS-DYNA, the corresponding refined numerical simulations are performed. By comparing the experimental and numerical damage and vibration accelerations of the composite structures, the present adopted numerical algorithms, constitutive models and parameters are validated, which will be applied in the further assessment of drop impact effects of full-scale SFC and FA on prototype nuclear fuel reprocessing plant in the next Part II of this paper.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.20 no.1
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• pp.7-15
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• 2024

Environmental qualification is required for safety related electrical equipment under harsh environments located in nuclear power plants according to 10 CFR 50.49 and RG 1.89. As analog technology has recently been replaced by digital technology, NRC established RG 1.209 as a regulatory guideline for environmental qualification of safety related computer-based I&C system located in mild environments, requiring evaluation for electromagnetic compatibility, smoke exposure and type test for actual service conditions such as temperature and humidity. In this paper, the trend of environmental qualification for digital equipment is analyzed by comparing the environmental qualification requirements between digital and analog equipment.