• Journal of Radiation Protection and Research
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• v.40 no.2
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• pp.92-100
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• 2015

For the purpose of improving the efficiency of the radiation impact assessment of dry interim storage facilities for the spent nuclear fuel of pressurized water reactors (PWRs), radiation impact assessment was performed after the application of sensitivity assessment according to the radiation source term designation method, development of a 2-step calculation technique, and cooling time credit. The present study successively designated radiation source terms in accordance with the cask arrangement order in the shielding building, assessed sensitivity, which affects direct dose, and confirmed that the radiation dosage of the external walls of the shielding building was dominantly affected by the two columns closest to the internal walls. In addition, in the case in which shielding buildings were introduced into storage facilities, the present study established and assessed the 2-step calculation technique, which can reduce the immense computational analysis time. Consequently, results similar to those from existing calculations were derived in approximately half the analysis time. Finally, when radiation source terms were established by adding the storage period of the storage casks successively stored in the storage facilities and the cooling period of the spent nuclear fuel, the radiation dose of the external walls of the buildings was confirmed to be approximately 40% lower than the calculated values; the cooling period was established as being identical. The present study was conducted to improve the efficiency of the Monte Carlo shielding analysis method for radiation impact assessment of interim storage facilities. If reliability is improved through the assessment of more diverse cases, the results of the present study can be used for the design of storage facilities and the establishment of site boundary standards.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.3
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• pp.239-245
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• 2010

There are 11,811 LWR spent fuels stored at reactor sites, as of 2009. Source terms based on reference spent fuel which represents entire spent fuels with bounding values in the aspect of source term has been applied to a design of nuclear installations, instead of those which are generated by weighting respective source term for each spent fuel. Simplified regression models to estimate total decay heat, radioactivity, and ingestion hazard index for spent fuel from Westinghouse-type reactors were developed in this study, because it can be used as a fundamental model for weighting source term for respective spent fuel to exclude conservativeness in source terms. It was found that the estimated source terms agreed with calculated value from ORIGEN-ARP within 5%. It was also found that the conservativeness could be excluded if the weight source terms were used as reference source term in the design. Therefore, it is expected that the developed regression model could be widely used in the conceptual design process of nuclear facilities related with storage and disposal of spent nuclear fuel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.1
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• pp.127-131
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• 2022

The structural safety of prototype transport and storage containers for very low-level radioactive liquid waste was experimentally estimated for its localization development. Transport containers for radioactive liquid waste have been researched and developed, however, there are no standardized commercial containers for very low-level radioactive waste in Korea. In this study, the structural safety of the designated IP-2 type container capable of transporting and temporarily storing large amounts of very low-level liquid waste, which is generated during the operation and decommissioning of nuclear power plants, was demonstrated. The stacking and drop tests, which were conducted to determine the structural integrity of the container, verified that there was no external leakage of the contents in spite of its structural deformation due to the drop impact. This study shows the effort required for the localization of the technology used in manufacturing transport and storage containers for very low-level radioactive liquid waste, and the additional structural reinforcement of the container in which the commercial intermediate bulk container (IBC) external frame was coupled.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.345-345
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• 2004

The spent fuel transportation casks have used as one of the most essential component in the nuclear industry. And, the number of the cask has been significantly increased in recent years. While the bulk amount of spent fuel in the world is still kept in the storage pool, the number of countries which have chosen the advantages of dual purpose cask for transportation and storage is rapidly increasing. The technical experience in the area of spent fuel transportation cask operation and maintenance for long period is also available and will be well utilized also in storage casks. The increasing use of casks for dual and multiple purposes raises an issue of long term consideration by international standardization. Accordingly IAEA is providing a regulatory requirements and guidelines as an effort for this standardization.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.234-244
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• 2017

Natural frequencies of immersed square sections decrease due to a contribution of added mass to the movement of square sections. In this study, natural frequencies of square sections are obtained as a function of gap size between the square section and a rigid square wall using the finite element method. Additionally, they are used to extract the added mass effect on translational and rotation motions. Published information and studies on the translational and torsional vibration of square beams are also examined for practical use. D coupling of a square section is also investigated for multiple square sections. The suggested added mass estimation can be applicable to the spent fuel storage design of a pressurized light water modulated nuclear power plant.

• Corrosion Science and Technology
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• v.11 no.6
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• pp.213-217
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• 2012

The amount of spent fuel from nuclear power plants has been increasing. An effective management plan of the spent fuel becomes a critical issue, because the storage capacity of each plant will reach its storage limit in a few years. The volume of high toxic spent fuel can be reduced through a fuel processing. Advanced Fuel Cycle (AFC) system is considered to be one of the options to reduce the toxicity and volume of the spent fuel. It is necessary to set up a test facility to demonstrate the feasibility of the process at the engineering scale. The objective of the work is a development of the safety evaluation technology for the AFC system. The evaluation technology of the AFC structural integrity and processes were surveyed and reviewed. Key evaluation parameters for the main processes such as electrolytic reduction, electrorefining, and electrowinning were obtained. The survey results may be used for the establishment of the AFC regulatory licensing procedure. The establishment of the licensing criteria minimizes the trials and errors of the AFC facility design. Issues taken from the survey on the regulatory procedure and design safety features for the AFC facility provide a chance to resolve potential issues in advance.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2007.05a
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• pp.190-191
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• 2007

Cleanout of the F Reactor Fuel Storage Basin (FSB) is an element of the FSB decontamination and decommissioning (D&D) and is required to complete interim safe storage (ISS) of the F Reactor. Following reactor shutdown and in preparation for a deactivation layaway action in 1970, the water level in the FReactor FSB was reduced to approximately 0.6 m (2 ft) over t]to floor. Basin components and other miscellaneous items were left or placed in the FSB. The item placement was performed with a sense of finality, and no attempt was made to place the items in an orderly manner. The F Reactor FSB was then filled to grade level with 6(20of local surface material (essentially a fine sand). The reactor FSB backfill cleanout has the potential of having to remove spent nuclear fuel (SNF) that may have been left unintentionally. Based on previous cleanout of six water-filled FSBs with similar designs (i.e., the B, C, D, and DR FSBs in the 1980's), it was estimated that up to five SNF elements could be discovered in the F FSB (I). In reality about 17 full SNF elements were found in the excavation. This paper covers the technical and programmatic challenges of performing this decommissioning effort with some of the controls used for SNF management. The paper also will highlight how many various technologies were married into a complete package to address the issue at hand and show how no one tools could complete the job, but combined, good progress is being made.

• Nuclear Engineering and Technology
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• v.28 no.4
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• pp.349-358
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• 1996

The leaching behavior of uranium from unirradiated CANDU UO$_2$ fuel pellet in the spent fuel wet storage and disposal conditions has been investigated. A modified IAEA leach test method was used, and then the extent of leaching was monitored by analysis for uranium in the leachant. The leach test has been performed in various leachants(demineralized water and boric acid solution at pH=6, synthetic granite groundwater) for a long-term period of 5.4 years, and the effect of temperature on the leach rate of uranium has been analyzed. The leach rates of uranium at $25^{\circ}C$ were dependent on the leachants. Over initial 100 days of leach periods, the leach rate in groundwater was the highest in three leachants and no significant differences of leach rates ore observed in the demineralized oater and boric acid solution. But these leach rates in three leachants around 2,000 days at $25^{\circ}C$ appeared to be reached the steady rates in the range of 1~5$\times$10$^{-8}$ g/$\textrm{cm}^2$ day. The leach rate of uranium in groundwater shooed to be independent of the temperature, but those in both demineralized water and boric acid solution increased with temperature. These results show that the leaching behavior of uranium from UO$_2$ fuel in both the demineralized water ann boric acid may be controlled tv the surface oxidative.dissolution reaction of UO$_2$ and the leach rate of uranium in groundwater at room temperature could mainly be controlled by the complex reaction of dissolved uranyl ions with carbonate ions and no variation of leach rate of UO$_2$ in groundwater with temperature may be due to the local deposition of passivating uranyl phases on the surface.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1009-1014
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• 2023

Muon tomography is a useful method for monitoring special nuclear materials (SNMs) such as spent nuclear fuel inside dry cask storage. Multiple Coulomb scattering of muons can be used to provide information about the 3-dimensional structure and atomic number(Z) of the inner materials. Tomography using muons is less affected by the shielding material and less harmful to health than other measurement methods. We developed a muon detector for muon tomography, which consists of a plastic scintillator, 64 long wavelength-shifting (WLS) fibers attached to the top of the plastic scintillator, and silicon photomultipliers (SiPMs) connected to both ends of each WLS fiber. The muon detector can acquire X and Y positions simultaneously using a position determination algorithm. The design parameters of the muon detector were optimized using DETECT2000 and Geant4 simulations, and a muon detector prototype was built based on the results. Spatial resolution measurement was performed using simulations and experiments to evaluate the feasibility of the muon detector. The experimental results were in good agreement with the simulation results. The muon detector has been confirmed for use in a muon tomography system.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2883-2897
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• 2022

According to the Renewable Energy 3020 Implementation Plan announced in 2017 by the South Korean government, the electricity share of renewable energy will be expanded to 20% of the total electricity generation by 2030. Given the intermittency of electricity generation from renewable energy, realization of such a plan presents challenges to managing South Korea's isolated national electric grid and implies potentially large excess electricity generation in certain situations. The purpose of this study is: 1) to develop a model to accurately simulate the effects of excess electricity generation from renewables which would arise during the transition, and 2) to propose strategies to manage excess electricity generation through effective utilization of domestic electricity generating capabilities. Our results show that in periods of greater PV and wind power, namely the spring and fall seasons, the frequency of excess electricity generation increases, while electricity demand decreases. This being the case, flexible operation of coal and nuclear power plants along with LNG and pumped-storage hydroelectricity can be used to counterbalance the excess electricity generation from renewables. In addition, nuclear energy plays an important role in reducing CO₂ emissions and electricity costs unlike the fossil fuel-based generation sources outlined in the 8th Basic Plan.