• Composites Research
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• v.28 no.2
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• pp.58-64
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• 2015

The purpose of this paper is to suggest the effective equivalent finite element model for the impact limiter of a nuclear spent fuel shipping cask made of sandwich composite panels. The sandwich composite panels were composed of a metallic facesheet and a core material made of urethane foam, balsa wood and red wood, respectively. The effective equivalent finite element model for the impact limiter was proposed by comparing the results of low-velocity impact test of sandwich panels. An explicit finite element analysis based on LS-DYNA 3D was done in this study. The results showed that the solid elements were recommended to model the facesheet and core of sandwich panels for impact limiter compared to combination modeling method, in which the layered shell element for facesheet and solid element for core material are used. In particular, the solid element for balsa and red wood core materials should be modeled by the element elimination approach.

• Nuclear Engineering and Technology
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• v.15 no.4
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• pp.248-255
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• 1983

The thermal analysis on the spent fuel shipping cask for a PWR fuel assembly is performed. Under the normal and fire-accident conditions the temperature distribution through a multilayer cask calculated in compliance with 10 CFR Part 71. A KNU 5&6 spent fuel assembly is assumed to be the decay heat source, which has the maximum discharge turnup of 45, 000MWD/MTU and has been stored in the spent fuel storage pool for 300 days. As a result of thermal analysis, the maximum cladding temperature in case of dry cavity under fire-accident conditions is calculated to be 455$^{\circ}C$. This value is much less than the limiting value specified in 10 CFR Part 50.46. It indicates that no fuel rod cladding rupture could occur under fire-accident conditions. It was also found that no melting of lead would take place in the major shield region.

• Journal of Radiation Protection and Research
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• v.18 no.2
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• pp.47-59
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• 1993

The criticality of the shipping cask(KSC-7) for transportion of 7PWR spent fuel assemblies has been calculated and analysised on the basis of neutron transport theory. For criticality analysis, effects of the rod pitches, the fixed neutron absorbers(borated sus+boral) were considered. The effective multiplication factor has been calculated by KENO-Va, Mote Carlo method computer code, with the HANSEN-ROACH 16 group cross section set, which was made for personal computer system. The criticality for the KSC-7 cask was calculated in terms of the fresh fuel which was conservative for the aspects of nuclear critility. From the results of criticality analysis, the calculated Keff is proved to be lower than subcritical limit during normal transportation and under hypothetical accident condition. The maximum calculated criticalities of the KSC-7 were lower the safety criticality limit 1.0 recommended by US 10CFR71 both under normal and hypothetical accident condition. Also, to verify the KSC-7 criticality calculation results by using KENO-Va, it was carried out benchmark calculation with experimental data of B & W(Bobcock and Wilcox) company. From the 3s series of calculation of the KSC-7 cask and benchmark calculation, the cask was safely designed in nuclear criticality, respectively.

• 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.

• Journal of Radiation Protection and Research
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• v.14 no.1
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• pp.56-65
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• 1989

The nuclear criticality of the KSC-4 shipping cask which can load four assemblies of PWR spent fuel was analyzed using KENO-IV computer code and 19-group nuclear cross section set generated from 218-group neutron cross section library(DLC-43/CSRL) using AMPX system. In accordance with 10CFR71, the analysis was performed for fresh fuel assemblies, instead of the spent fuels, under both norml transportation and hypothetical accident conditions. The calculated maximum multiplication factors(Keff) of the KSC-4 cask were 0.85289 and 0.94185 for the normal transportation and hypothetical accident conditions, respectively. The highest Keff of the KSC-4 cask is within the subcritical limit prescribed in l0CFR71 and accordingly the KSC-4 cask is safely designed in terms of nulear criticality.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2001.11b
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• pp.766-771
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• 2001

The package used to transport radioactive materials, which is called by 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 one for test. the verification by FE analysis is mainly used. But due to the complexity of mechanical behaviors. the results depends on how users apply the codes and it can cause severe errors during analysis. In this paper, finite element analysis is carried out for the 9 meters free drop and the puncture condition of the hypothetical accident conditions using LS-DYNA3D and ABAQUS/Explicit. We have investigated the analyzing technique for the free drop impact test of the cask and found several vulnerable cases to errors. 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.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.367-367
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• 2001

• Journal of Radiation Protection and Research
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• v.22 no.2
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• pp.77-83
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• 1997

Epoxy resin-type neutron shielding materials, KNS(Kaeri Neutron Shield)-101, KNS-102, and KNS-103 have been fabricated to be used in spent fuel shipping cask. The base material is epoxy resin, and polypropylene, aluminium hydroxide, and boron carbide are added. These shielding materials offer good fluidity at processing, which makes it possible to apply this resin shield to complicated geometric shapes such as shipping cask. The shielding property of these shielding materials for shipping cask for loading 28 PWR spent fuel assemblies has been evaluated. ANISN code is used to evaluate the shielding property of the shipping cask with the thickness of the three neutron shielding materials greater than 10 cm. As a result of analysis, the maximum calculated dose rate at the radial surface of the cask is determined to be $300{\mu}Sv/h$ and the maximum calculated dose rate at 100 cm from the cask is $97{\mu}Sv/h$. These dose rates remain within allowable values specified in related regulations.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1439-1447
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• 2023

The fuel cycle characteristics of the IVG.1M reactor were studied within the framework of the research reactor conversion program to modernize the IVG.1M reactor. Optimum use of the nuclear fuel and reactor was achieved through routine methods which included partial fuel reloading combined with scheduled maintenance operations. Since, the additional problem in planning the fuel cycle of the IVG.1M reactor was the poisoning of the beryllium parts of the core, reflector, and control system. An assessment of the residual power and composition of spent fuel is necessary for the selection and justification of the technology for its subsequent management. Computational studies were performed using the MCNP6.1 program and the neutronics model of the IVG.1M reactor. The proposed scheme of annual partial fuel reloading allows for maintaining a high reactor reactivity margin, stabilizing it within 2-4 β_eff for 20 years, and achieving a burnup of 9.9-10.8 MW × day/kg U in the steady state mode of fuel reloading. Spent fuel immediately after unloading from the reactor can be placed in a transport packaging cask for shipping or safely stored in dry storage at the research reactor site.