• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.390.2-390
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• 2002

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.2
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• pp.255-269
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• 2023

Compared to operational wastes, nuclear power plant (NPP) decommissioning wastes are generated in larger quantities within a short time and include diverse types with a wider range of radiation characteristics. Currently used 200 L drums and IP-2 type transport containers are inefficient and restrictive in packaging and transporting decommissioning wastes. Therefore, new packaging and transport containers with greater size, loading weight, and shielding performance have been developed. When transporting radioactive materials, radiological safety should be assessed by reflecting parameters such as the type and quantity of the package, transport route, and transport environment. Thus far, safety evaluations of radioactive waste transport have mainly targeted operational wastes, that have less radioactivity and a smaller amount per transport than decommissioning wastes. Therefore, in this study, the possible radiation effects during the transport from NPP to disposal facilities were evaluated to reflect the characteristics of the newly developed containers and decommissioning wastes. According to the evaluation results, the exposure dose to transport workers, handling workers, and the public was lower than the domestic regulatory limit. In addition, all exposure dose results were confirmed, through sensitivity analysis, to satisfy the evaluation criteria even under circumstances when radioactive materials were released 100% from the container.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1878-1886
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• 2021

For industrial package (IP)-type transport containers for radioactive materials, a free drop test should be conducted under regulatory conditions. Owing to various uncertainties observed during the drop test, errors in drop angles inevitably occur. In IP-type metal transport containers in which the container directly impacts onto a rigid target without any shock absorbing materials, the error in the drop angle due to a slight misalignment makes a significant difference from the ideal drop. In particular, in a vertical drop, the error in the drop angle causes a strong secondary impact. In this paper, a numerical method is proposed to estimate the error in the drop angle occurring during the test. To determine this error, an optimization method accompanying a computational drop analysis is proposed, and a surrogate model is introduced to ensure calculation efficiency. Effectiveness of the proposed method is validated by performing the verification and comparison between the test and the analysis applied with the drop angle error.

• Journal of Radiation Protection and Research
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• v.28 no.4
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• pp.291-298
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• 2003

The KN-12 transport cask has been designed to transport 12 PWR spent nuclear fuel assemblies and to comply with the regulatory requirements for a Type B(U) package. The containment boundary of the cask is defined by a cask body, a cask lid, lid bolts with nuts, O-ring seals and a bolted closure lid. The containment vessel for the cask consists of a forged thick-walled carbon steel cylindrical body with an integrally-welded carbon steel bottom and is closed by a lid made of stainless steel, which is fastened to the cask body by lid bolts with nuts and sealed by double elastomer O-rings. In the cask lid an opening is closed by a plug with an O-ring seal and covered by the bolted closure lid sealed with an O-ring. The cask must maintain a radioactivity release rate of not more than the regulatory limit for normal transport conditions and for hypothetical accident conditions, as required by the related regulations. The containment requirements of the cask are satisfied by maintaining a maximum air reference leak rate of $2.7{\times}10^{-4}ref.cm^3s^{-1}$ or a helium leak rate of $3.3{\times}10^{-4}cm^3s^{-1}$ for normal transport conditions and for hypothetical accident conditions.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.3
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• pp.231-236
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• 2005

HIC transport packaging to transport a high integrity container(HIC) containing dry spent resin generated from nuclear power plants is to comply with the regulatory requirements of Korea and IAEA for Type B packaging due to the high radioactivity of the content, and to maintain the structural integrity under normal and accident conditions. It must withstand 9 m free drop impact onto an unyielding surface and 1 m drop impact onto a mild steel bar in a position causing maximum damage. For the conceptual design of a cylindrical HIC transport package, three dimensional dynamic structural analysis to ensure that the integrity of the package is maintained under all credible loads for 9 m free drop and 1 m puncture conditions were carried out using ABAQUS code.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.20-28
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• 2008

According to IAEA regulations, a transportation package of radioactive material should perform its intended function of containing the radioactive contents after the drop test, which is one of hypothetical accident conditions. Impact limiters attached to a transport cask absorb the most of impact energy. So, it is appreciated to determine properly the shape, size and material of impact limiters. A material data needed in this determination is a dynamic one. In this study, several materials considered as those of impact limiters were tested by a drop weight facility to acquire dynamic material characteristics data. Impact absorbing volume of the impact limiter was derived mathematically for each drop condition. A size optimization of impact limiter was conducted. The derived impact absorbing volumes were applied as constraints. These volumes should be less than critical volumes generated based on the dynamic material characteristics. The derived procedure to decide the shape of impact limiter can be useful at the preliminary design stage when the transportation package's outline is roughly determined and applied as input value.

• Journal of Fisheries and Marine Sciences Education
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• v.28 no.6
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• pp.1772-1780
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• 2016

NCS development for the treatment of waste oils from ship was carried out through steps such as analysis on characteristics, development of competency standard, utilizing package, and validation of industry sites. The results were as follows. Firstly, duty competency was classified as levels from 2 to 6. Educational training institutions were followed by 75 graduate schools, 73 universities, 54 colleges, and 37 high schools. Secondly, developed standards were consisted of duty and competency unit. The name of duty was the treatment of waste oils from ship and competency units were consisted of 8 items as classification of waste oils from ship, pickup and transport of waste oils from ship, warehousing of waste oils from ship to marine disposal company, transport of waste oils from ship to land, warehousing of waste oils from to disposal company, determination of disposal method and plant recycling treatment, and incineration treatment. 28 competency unit elements below 8 competency units were developed. Thirdly, utilizing package was developed into 3 areas of life-long career path, training criteria, and guidelines for exam according to national competency standards in order to develop development of labor's career and perform personal management such as hiring and promotion in industry sites.

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 2011.04a
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• pp.89-92
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• 2011

The modeling of thermodynamic non-idealities and transport anomalies is a crucial prerequisite to realistically simulate the mixing and combustion processes of liquid propellants injected above critical pressures. This study has developed a specific set of subroutines to calculate the thermodynamic and transport properties based on the generalized cubic equation of state (EoS) in a coupled manner with the standard chemical kinetics packages (Chemkin). The existing flamelet analysis code is extended with the real-fluid package and applied to numerical investigation of local flame structures of kerosene and liquid oxygen at high pressure conditions relevant to the actual rocket engines.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1079-1087
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• 2021

NTP-ERSN is a package developed for solving the multigroup form of the discrete ordinates, characteristics and collision probability of the Boltzmann transport equation in one-dimensional cartesian geometry, by combining pin cells. In this work, C5G7 MOX benchmark is used to verify the accuracy and efficiency of NTP-ERSN package, by treating reactor core problems without spatial homogenization. This benchmark requires solutions in the form of normalized pin powers as well as the vectors and the eigenvalue. All NTP-ERSN simulations are carried out with appropriate spatial and angular approximations. A good agreement between NTP-ERSN results with those obtained with OpenMC calculation code for seven energy groups. In addition, our studies about angular and mesh refinements are carried out to produce better quality solution. Moreover, NTP-ERSN GUI has also been updated and adapted to python 3 programming language.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.375-388
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• 2024

In this study, to evaluate the shock and vibration load characteristics of used fuel, a sea transportation test was conducted using simulated fuel assemblies under normal transport conditions. An overall test data analysis was performed based on the measured strain and acceleration data obtained from cruise, rotation, acceleration, braking, depth of water, and rolling tests. In addition, shock response spectrum and power spectral densities were obtained for each test case. Amplification and attenuation characteristics were investigated based on the load path. The load was amplified as it passed from the overpack to the simulated used fuel-assembly. As a result of the RMS trend analysis, the fuel-loading position of the transportation package affected the measured strain in the fuel rod, and the maximum strains were obtained at the spans with large spacing. However, even these maximum strains were very small compared to the fatigue strength and the cladding yield strength. Moreover, the fuel rods located on the side exhibited a larger strain value than those at the center.