• Proceedings of the Korean Radioactive Waste Society Conference
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• 2012.10a
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• pp.49-50
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• 2012

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2015.10a
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• pp.153-154
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• 2015

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2016.05a
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• pp.79-80
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• 2016

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2016.05a
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• pp.67-68
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• 2016

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2015.10a
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• pp.103-104
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• 2015

• Journal of computational fluids engineering
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• v.21 no.2
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• pp.47-53
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• 2016

This study investigated natural convection flow behavior in airflow path designed in concrete dry storage cask to remove the decay heat from spent nuclear fuels. Using FLUENT 16.1 code, thermal analysis for natural convection was carried out for three dimensional, 1/4 symmetry model under the normal condition that inlet ducts are 100% open. The maximum temperatures on other components except the fuel regions were satisfied with allowable values suggested in nuclear regulation-1536. From velocity and temperature distributions along the flow direction, the flow behavior in horizontal duct of air inlet and outlet duct, annular flow-path and bent pipe was delineated in detail. Theses results will be used as the theoretical background for the composing of airflow path for the designing of passive heat removal system by understanding the flow phenomena in airflow path.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2012.05a
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• pp.25-26
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• 2012

• Journal of computational fluids engineering
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• v.21 no.4
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• pp.33-39
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• 2016

Generally, thermal analysis of spent fuel storage cask has been conducted using the porous media and effective thermal conductivity model to simplify the structural complexity of spent fuel assemblies. As the fuel assembly is composed of two regions; active fuel region corresponding to UO2 pellets and unactive fuel region corresponding to the top and bottom nozzle, the heat transfer performance can be influenced depending on porous media application at these regions. In this study, numerical analysis on concrete storage cask of spent fuel was performed to investigate heat transfer effects for two cases; one was porous media application only to active fuel region(case 1) and the other one was porous media to whole length of fuel assembly(case 2). Using computational fluid dynamics code, the three dimensional, 1/4 symmetry model was constructed. For two cases, maximum temperatures for each component were evaluated below the allowable limits. For the case 1, maximum temperatures for fuel cladding, neutron absorber and baskets inside the canister were slightly higher than those for the case 2. In particular, even though the helium flows with low velocity due to buoyant forces occurred at the top and bottom of unactive fuel region, treating only active fuel region as the porous media was ineffective in respect of the heat removal performance of concrete storage cask, implying a conservative result.

• Journal of the Korea institute for structural maintenance and inspection
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• v.27 no.6
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• pp.152-161
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• 2023

In general, large-capacity hydrogen storage vessels, typically in the form of vertical cylindrical vessels, are constructed using steel materials. These vessels are anchored to foundation slabs that are specially designed to suit the environmental conditions. This anchoring method involves pre-installed anchors on top of the concrete foundation slab. However, it's important to note that such a design can result in concentrated stresses at the anchoring points when external forces, such as seismic events, are at play. This may lead to potential structural damage due to anchor and concrete damage. For this reason, in this study, it selected an vertical hydrogen storage vessel based on site observations and created a 3D finite element model. Artificial seismic motions made following the procedures specified in ICC-ES AC 156, as well as domestic recorded earthquakes with a magnitude greater than 5.0, were applied to analyze the structural behavior and performance of the target structures. Conducting experiments on a structure built to actual scale would be ideal, but due to practical constraints, it proved challenging to execute. Therefore, it opted for an analytical approach to assess the safety of the target structure. Regarding the structural response characteristics, the acceleration induced by seismic motion was observed to amplify by approximately ten times compared to the input seismic motions. Additionally, there was a tendency for a decrease in amplification as the response acceleration was transmitted to the point where the centre of gravity is located. For the vulnerable components, specifically the sub-system (support columns and anchorages), the stress levels were found to satisfy the allowable stress criteria. However, the concrete's tensile strength exhibited only about a 5% margin of safety compared to the allowable stress. This indicates the need for mitigation strategies in addressing these concerns. Based on the research findings presented in this paper, it is anticipated that predictable load information for the design of storage vessels required for future shaking table tests will be provided.

• Journal of Radiation Protection and Research
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• v.38 no.4
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• pp.189-193
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• 2013

Annual dose on the containment building wall of the interim storage facility at normal condition was calculated to estimate the dose rate transition of the facility of PWR spent nuclear fuel. In this study, source term was generated by ORIGEN-ARP with 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facility and the containment building and radiation shielding evaluations were conducted by MCNP code depending on the distance between the wall and the facility in the building. In the case of the centralized storage system, the distance required for the annual dose rate limit from 10CFR72 was estimated to be 50 m.