• Journal of the Korea Concrete Institute
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• v.26 no.1
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• pp.71-78
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• 2014

The long term integrity of concrete cask is very important for spent nuclear fuel dry storage system. However, there are serious concerns about early deterioration of concrete cask from creaking and corrosion of reinforcing steel by chloride ion because the cask is usually located in seaside, expecially by combined deterioration such as chloride ion and heat, carbonation. This study is to investigate the relation between temperature and chloride ion diffusion of concrete. Immersion tests using 3.5% NaCl solution that were controlled in four level of temperature, i.e. 20, 40, 65, and $90^{\circ}C$, were conducted for four months. The chloride ion diffusion coefficient of concrete was predicted based on the results of profiles of Cl- ion concentration with the depth direction of concrete specimens using the method of potentiometric titration by $AgNO_3$. Test results indicate that the diffusion coefficient of chloride ion increases remarkably with increasing temperature, and there was a linear relation between the natural logarithm values of the diffusion coefficients and the reciprocal of the temperature from the Arrhenius plots. Activation energy of concrete in this study was about 46.6 (W/C = 40%), 41.7 (W/C = 50%), 30.7 (W/C = 60%) kJ/mol under a temperature of up to $90^{\circ}C$, and concrete with lower water-cement ratio has a tendency towards having higher temperature dependency.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.3
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• pp.161-167
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• 2011

In case that the maximum temperature of any surface readily accessible during transport of a spent nuclear fuel (SNF) transport cask exceeds $85^{\circ}C$ in the absence of insolation under the ambient temperature of $38^{\circ}C$, personnel barriers or transport hood shall be used to prevent people from casual contact with the transport cask surface. Usually the air temperature within the hood and the hood surface temperature are calculated and further utilized as boundary conditions(free stream temperature and external radiation temperature) for thermal evaluation under normal conditions of transport. In this study, these temperatures are derived using the analytical method based on the heat transfer mechanism around the transport cask under transport hood assuming the thermal equilibrium. By comparing the analytical solutions with the results from the detailed calculations with CFD-computer-code FLUENT 12.1 it is verified that the analytical method is still efficient tool to estimate the temperatures and these temperatures can be further used as boundary conditions for thermal evaluation under normal conditions of transport.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.330-330
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.278.2-278
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• 2001

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.175-176
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.05a
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• pp.37-38
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.117-118
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• 2017

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.113-114
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• 2017

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.42-50
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• 2004

The pool $(3m{\times}6m{\times}10m{\times}$ in Irradiated Materials Examination Facility is generally used to transport irradiated materials between a moving cask and hot-cell. During the operation in the pool such as loading/unloading the cask, holding specimen and bucket elevation, water maybe contaminated by radioactive or contaminated impurities from irradiated materials. Then, it must be purified and filtered continuously to keep lower radioactivity than that of regulation prescribed by RCA Korea Activity in a part of radioactive contamination control. This paper described radioactive contamination distribution of water as transported materials, which is related to effective operation of purification and filtration system.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 1996.10a
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• pp.70-77
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• 1996

In this study, sliding response analysis of free standing structure such as multi-purpose nuclear spent fuel storage cask is peformed. The governing factors of sliding response are aspect ratio of structure and ground acceleration. The vertical acceleration component is very important factor in the sliding response of the structure. Based on the mathematical model, computer program is developed using direct forward integration method to predict the sliding response. Using the program, several parametric studies were made for sinusodial ground motion and for El Centre 1940 earthquake and Mexico 1973 earthquake. From the results, it is known that the frequency content and duration of strong motion affect the sliding of the structure.