• Journal of Institute of Convergence Technology
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• v.2 no.1
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• pp.7-12
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• 2012

Spent fuels are discharged from nuclear reactors as a result of power generations. The spent fuels would be considered as a useful resources because the main constituent is uranium and some other actinides are included in them. In order to utilize the resources chemical processes should be developed to treat the spent fuels and obtain uranium and other actinides to be fueled in a fast reactor. The technologies are categorized into wet and dry processes. In this study, the current status of such technologies is summarized to give a insight and a deep understanding on nuclear fuel cycles.

• Transactions of the Korean Society of Automotive Engineers
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• v.10 no.6
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• pp.72-79
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• 2002

The main objective of this study is to investigate the relationship between chemical structure and higher heating value of reformed diesel fuels by ultrasonic irradiation. In order to analyze the chemical structure changes of the reformed diesel fuels by ultrasonic irradiation, Proton nuclear magnetic resonance spectrometer(1H-NMR) was used and to analyze the effect of higher heating values of these diesel fuels, the bomb calorimeter was used. From the study, following conclusive remarks can be made. 1) The aromatic carbon percentages and higher heating values of the reformed diesel fuels by ultrasonic irradiation increased more than the conventional diesel ones. 2) The aromatics percentages and Branch Index(BI) of the reformed diesel fuels by ultrasonic irradiation decreased more than the conventional diesel ones. 3) The higher heating values on both for conventional fuel and reformed diesel fuels by ultrasonic energy irradiation is directly proportional to aromatic carbon percentages and inversely proportional to aromatic percentages and BI for these fuels.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.4
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• pp.503-516
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• 2023

A transfer cask serves as the container for transporting and handling canisters loaded with spent nuclear fuels from light water reactors. This study focuses on a cylindrical transfer cask, standing at 5,300 mm with an external diameter of 2,170 mm, featuring impact limiters on the top and bottom sides. The base of the cask body has an openable/closable lid for loading canisters with storage modules. The transfer cask houses a canister containing spent nuclear fuels from lightweight reactors, serving as the confinement boundary while the cask itself lacks the confinement structure. The objective of this study was to conduct a structural analysis evaluation of the transfer cask, currently under development in Korea, ensuring its safety. This evaluation encompasses analyses of loads under normal, off-normal, and accident conditions, adhering to NUREG-2215. Structural integrity was assessed by comparing combined results for each load against stress limits. The results confirm that the transfer cask meets stress limits across normal, off-normal, and accident conditions, establishing its structural safety.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.4
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• pp.451-463
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• 2023

With South Korea increasingly focusing on nuclear energy, the management of spent nuclear fuel has attracted considerable attention in South Korea. This study established a novel procedure for selecting safety-relevant radionuclides for long-term safety assessments of a deep geological repository in South Korea. Statistical evaluations were performed to identify the design basis reference spent nuclear fuels and evaluate the source term for up to one million years. Safety-relevant radionuclides were determined based on the half-life criteria, the projected activities for the design basis reference spent nuclear fuel, and the annual limit of ingestion set by the Nuclear Safety and Security Commission Notification No. 2019-10 without considering their chemical and hydrogeological properties. The proposed process was used to select 56 radionuclides, comprising 27 fission and activation products and 29 actinide nuclides. This study explains first the determination of the design basis reference spent nuclear fuels, followed by a comprehensive discussion on the selection criteria and methodology for safety-relevant radionuclides.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.1
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• pp.117-123
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• 2015

To remove heat generated during a burn-up test of nuclear fuels, the heat generation rate of nuclear fuels should be calculated accurately, and a coolant should be circulated in the test loop at an adequate flow rate. HANARO is an open pool-type reactor with an independent test loop for the burn-up test of nuclear fuels. A test rig is installed in the test loop, and a coolant is circulated through the test loop to maintain the temperature of the nuclear fuel rods within a desired temperature during an irradiation test. The components and sensors in the test rig can be broken or malfunction owing to the flow-induced vibration. In this study, a coolant flow simulation system was developed to verify and confirm the soundness of components and sensors assembled in the test rig with a high flow rate of the coolant.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.1166-1167
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• 2005

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1398-1405
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• 2019

We use a molecular dynamics simulation to explore thermal transport in oxide nuclear fuels with point defects. The effect of vacancy and substitutional defects on the thermal conductivity of plutonium dioxide and uranium dioxide is investigated. It is found that the thermal conductivities of these fuels are reduced significantly by the presence of small amount of vacancy defects; 0.1% oxygen vacancy reduces the thermal conductivity of plutonium dioxide by more than 10%. The missing of larger atoms has a more detrimental impact on the thermal conductivity of actinide oxides. In uranium dioxide, for example, 0.1% uranium vacancies decrease the thermal conductivity by 24.6% while the same concentration of oxygen vacancies decreases the thermal conductivity by 19.4%. However, uranium substitution has a minimal effect on the thermal conductivity; 1.0% uranium substitution decreases the thermal conductivity of plutonium dioxide only by 1.5%.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.2
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• pp.255-266
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• 2021

Solid-state mechanochemical reduction combined with subsequent melting consolidation was suggested as a technical option for the oxide reduction in pyroprocessing. Ni ingot was produced from NiO as a starting material through this technique while Li metal was used as a reducing agent. To determine the technical feasibility of this approach for pyroprocessing, which handles spent nuclear fuels, thermodynamic calculations of the phase stabilities of various metal oxides of U and other fission elements were made when several alkaline and alkali-earth metals were used as reducing agents. This technique is expected to be beneficial, not only for oxide reduction but also for other unit processes involved in pyroprocessing.

• Journal of the Korean Society for Precision Engineering
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• v.32 no.9
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• pp.815-824
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• 2015

In this study, we conducted the approximate multi-objective optimization of gap sizes of pressurized-water reactor (PWR) annular fuels. To determine the contacting tendency of the inner-outer gaps between the annular fuel pellets and cladding, thermoelastic-plastic-creep (TEPC)analysis of PWR annular fuels was performed, using in-house FE code. For the efficient heat transfer at certain levels of stress, we investigated the tensile, compressive hoop stress and temperature, and optimized the gap sizes using the non-dominant sorting genetic algorithm (NSGA-II). For this, response surface models of objective and constraint functions were generated, using central composite (CCD) and D-optimal design. The accuracy of approximate models was evaluated through $R^2$ value. The obtained optimal solutions by NSGA-II were verified through the TEPC analysis, and we compared the obtained optimum solutions and generated errors from the CCD and D-optimal design. We observed that optimum solutions differ, according to design of experiments (DOE) method.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.272.2-272
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• 2001