• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.137-138
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.119-120
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• 2018

• 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.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.601-607
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• 2022

In recent years, monitoring of spent nuclear fuel inside dry cask storage has become an important area of national security. Muon tomography is a useful method for monitoring spent nuclear fuel because it uses high energy muons that penetrate deep into the target material and provides a 3-D structure of the inner materials. We designed a muon tomography system consisting of four 2-D position sensitive detector and characterized and optimized the system parameters. Each detector, measuring 200 × 200 cm², consists of a plastic scintillator, wavelength shifting (WLS) fibers and, SiPMs. The reconstructed image is obtained by extracting the intersection of the incoming and outgoing muon tracks using a Point-of-Closest-Approach (PoCA) algorithm. The Geant4 simulation was used to evaluate the performance of the muon tomography system and to optimize the design parameters including the pixel size of the muon detector, the field of view (FOV), and the distance between detectors. Based on the optimized design parameters, the spent fuel assemblies were modeled and the line profile was analyzed to conduct a feasibility study. Line profile analysis confirmed that muon tomography system can monitor nuclear spent fuel in dry storage container.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.4
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• pp.299-314
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• 2014

A lifting device is used to deal with transport cask for the transportation of spent fuels from nuclear power plants. This study performed theoretical analysis and numerical simulation to evaluate the structural integrity of the lifting device based on Nuclear Safety and Security Commission(NSSC) Notice No.2013-27 and US 10CFR Part 71 ${\S}71.45$. The results of theoretical analysis showed that the maximum stresses of all components were below the allowable values. This result confirmed that the lifting device was structurally safe during operation. The results of finite element analysis also showed that it was evaluated to satisfy the design criteria bothyielding and ultimate condition. All components have been shown to ensure the structural safety due to sufficient safety margins. In other words, the safety factor was 3 or more for the yielding condition and was 5 or more for the ultimate condition.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.18 no.1
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• pp.43-49
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• 2005

The package used to transport radioactive materials, which is called by the shipping cask, must be safe under normal and hypothetical accident conditions. These requirements for the cask design must be verified through test or finite element analysis. Since the cost for FE analysis is less than the one for test, the verification by FE analysis is mainly used. But due to the complexity of mechanical behaviors, the results depend on how users apply the codes and can cause severe errors during analysis. In this paper, finite element analysis is carried out for the 9 meters free drop condition of the hypothetical accident conditions using LS-DYNA3D and ABAQUS/Explicit. We have investigated the analyzing technique lot the free drop impact test of the cask and investigated several vulnerable cases. The analyzed results were compared with each other. We have suggested a reliable and relatively simple analysis technique for the drop test of spent nuclear fuel casks.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2002.05a
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• pp.98-102
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• 2002

As the nuclear power plant has been operated continuously and increased gradually, transportation and storage of spent fuel are seriously considered nowadays. The transportation cask which contains radioactive material needs to be inspected about structural safety. About safety verification, description of IAEA Safety Standards states that cask must withstand hypothetical accident conditions. In this paper, 9m free drop impact analysis was performed for transportation cask and impact limiter by using the finite element methods. Furthermore, we obtained the dynamic behavior of wood to as compared with safety test results, and verified the safety of transportation cask.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.05a
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• pp.153-154
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• 2017

The thermal analysis of transportation and storage cask for SNF was conducted during short term loading operations for forced gas drying condition. The fuel cladding temperature in 6 regions of SNF in the cask during the short term loading operations for forced gas drying condition is shown in the Fig. 3. The thermal analysis results of calculated maximum cladding temperature in each process demonstrate that operating scenario of TFD in detailed design maintain well below the temperature limits of $400^{\circ}C$.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.11
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• pp.1345-1352
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• 2012

This paper aims to evaluate the low-velocity impact responses and mechanical properties of balsa-wood and urethane-foam core materials and their sandwich panels, which are applied as the impact limiter of a nuclear spent fuel shipping cask. For the urethane-foam core, which is isotropic, tensile, compressive, and shear mechanical tests were conducted. For the balsa-wood core, which is orthotropic and shows different material properties in different orthogonal directions, nine mechanical properties were determined. The impact test specimens for the core material and their sandwich panel were subjected to low-velocity impact loads using an instrumented testing machine at impact energy levels of 1, 3, and 5 J. The experimental results showed that both the urethane-foam and the balsa-wood core except in the growth direction (z-direction) had a similar impact response for the energy absorbing capacity, contact force, and indentation. Furthermore, it was found that the urethane-foam core was suitable as an impact limiter material owing to its resistance to fire and low cost, and the balsa-wood core could also be strongly considered as an impact limiter material for a lightweight nuclear spent fuel shipping cask.

• Journal of Radiation Protection and Research
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• v.34 no.1
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• pp.1-7
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• 2009

It is expected that a substantial amount of spent fuels will be transported from the four nuclear power plant (NPP) sites in Korea to a hypothetical centralized interim storage facility or a final repository in the near future. The cost for the transportation is proportional to the amount of spent fuels. In this paper, a cost estimation program is developed based on the conceptual design of a transportation system and a logistics analysis. Using the developed computer program, named as CASK, the minimum capacity of a centralized interim storage facility (CISF) and the transportation cost for PWR spent fuels are calculated. The PWR spent fuels are transported from 4 NPP sites to a final repository (FR) via the CISF. Since NPP sites and the CISF are located along the coast, a sea-transportation is considered and a road-transportation is considered between the CISF and the FR. The result shows that the minimum capacity of the interim storage facility is 15,000 MTU.