• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.329-345
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• 2023

This paper described a method for analyzing the structural performance of a metal container used for disposing radioactive waste generated during the decommissioning of a nuclear power plant, and numerical analysis results of a method for reinforcing the container. The containers to be analyzed were those that can be used in near-surface and landfill disposal facilities scheduled to be operated at the Gyeongju radioactive waste disposal facility. Structural reinforcement of the container was performed by lattice reinforcement, column reinforcement, and bottom plate reinforcement. Accordingly, a total of 14 reinforcement cases were modeled. The external force causing damage to the container was set equivalent to the impact of a 9-m fall, accounting for the height of the vault at the near-surface disposal facility. The reinforcement methods with a high contribution to the structural performance of the container were concluded to be lattice and column reinforcements.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.1-19
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• 2020

Based on the current high-level radioactive waste management basic plan and the analysis results of spent nuclear fuel characteristics, such as dimensions and decay heat, an improved geological disposal concept for spent nuclear fuel from domestic nuclear power plants was proposed in this study. To this end, disposal container concepts for spent nuclear fuel from two types of reactors, pressurized water reactor (PWR) and Canada deuterium uranium (CANDU), considering the dimensions and interim storage method, were derived. In addition, considering the cooling time of the spent nuclear fuel at the time of disposal, according to the current basic plan-based scenarios, the amount of decay heat capacity for a disposal container was determined. Furthermore, improved disposal concepts for each disposal container were proposed, and analyses were conducted to determine whether the design requirements for the temperature limit were satisfied. Then, the disposal efficiencies of these disposal concepts were compared with those of the existing disposal concepts. The results indicated that the disposal area was reduced by approximately 20%, and the disposal density was increased by more than 20%.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.1
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• pp.133-140
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• 2021

This study evaluates the radioactivity of concrete waste that occurs due to large amounts of decommissioned nuclear wastes and then determines the surface dose rate when the waste is packaged in a disposal container. The radiation assessment was conducted under the presumption that impurities included in the bio-shielded concrete contain the highest amount of radioactivity among all the concrete wastes. Neutron flux was applied using the simplified model approach in a sample containing the most Co and Eu impurities, and a maximum of 9.8×104 Bq·g-1 60Co and 2.63×105 Bq·g-1 152Eu was determined. Subsequently, the surface dose rate of the container was measured assuming that the bio-shield concrete waste would be packaged in a newly developed disposal container. Results showed that most of the concrete wastes with a depth of 20 cm or higher from the concrete surface was found to have less than 1.8 mSv·hr-1 in the surface dose of the new-type disposal container. Hence, when bio-shielded concrete wastes, having the highest radioactivity, is disposed in the new disposal container, it satisfies the limit of the surface dose rate (i.e., 2 mSv·hr-1) as per global standards.

• Nuclear Engineering and Technology
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• v.31 no.6
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• pp.83-100
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• 1999

As disposal packaging concepts of spent fuels generated from the domestic NPP, two types, one is to package PWR and CANDU spent fuels in different containers and the other is to package them together, were proposed. The configuration of the containers and the layout of underground repository, such as the container spacing and the deposition tunnel spacing, were developed. The layout of underground repository satisfies the thermal constraint of the bentonite buffer surrounding disposal container, which should be lower than $100^{\circ}C$ in order to keep the physical and chemical properties of bentonite From the spent fuel packaging concepts and container emplacement methods, seven options were developed. With a typical pair-wise comparison methods, AHP, the most promising disposal concept was selected based on the technology Point of view.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.21-36
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• 2020

With respect to spent nuclear fuels, disposal containers and bentonite buffer blocks in deep geological disposal systems are the primary engineered barrier elements that are required to isolate radioactive toxicity for a long period of time and delay the leakage of radio nuclides such that they do not affect human and natural environments. Therefore, the thermal stability of the bentonite buffer and structural integrity of the disposal container are essential factors for maintaining the safety of a deep geological disposal system. The most important requirement in the design of such a system involves ensuring that the temperature of the buffer does not exceed 100℃ because of the decay heat emitted from high-level wastes loaded in the disposal container. In addition, the disposal containers should maintain structural integrity under loads, such as hydraulic pressure, at an underground depth of 500 m and swelling pressure of the bentonite buffer. In this study, we analyzed the thermal stability and structural integrity in a deep geological disposal environment of the improved deep geological disposal systems for domestic light-water and heavy-water reactor types of spent nuclear fuels, which were considered to be subject to direct disposal. The results of the thermal stability and structural integrity assessments indicated that the improved disposal systems for each type of spent nuclear fuel satisfied the temperature limit requirement (< 100℃) of the disposal system, and the disposal containers were observed to maintain their integrity with a safety ratio of 2.0 or higher in the environment of deep disposal.

• Transactions of Materials Processing
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• v.12 no.5
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• pp.479-486
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• 2003

This paper presents an elastoplastic analysis for spent nuclear fuel disposal container and its 50 cm thick bentonite buffer to predict the collapse of the container while the horizontal asymmetric sudden rock movement of 10 cm is applied on the composite structure. This sudden rock movement is anticipated by the earthquake etc. at a deep underground. Elastoplastic material model is adopted. Drucker-Prager yield criterion is used for the material yield prediction of the bentonite buffer and von-Mises yield criterion is used for the material yield prediction of the container. Analysis results show that even though very large deformations occur beyond the yield point in the bentonite buffer, the container structure still endures elastic small strains and stresses below the yield strength. Hence, the asymmetric 50 cm thick bentonite buffer can protect the container safely against the 10 cm sudden rock movement by earthquake etc.. Analysis results also show that bending deformations occur in the container structure due to the shear deformation of the bentonite buffer. The finite element analysis code, NISA, is used for the analysis.

• Journal of Radiation Protection and Research
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• v.31 no.1
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• pp.37-46
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• 2006

For the first step to develop a reference disposal container of spent fuel to be used in a deep geological repository, this paper examined safe dimensions of the disposal container on the points of nuclear criticality and radiation safety and mechanical structural safety and provided basic information for dimensioning the container and configuration of the container components, and establishing the favorable and safe disposal conditions. When the safety factor for stress due to the external loads (hydrostatic and swelling pressure) is taken as 2.0, the safe diameter of the filler material to provide enough container strength under the assumed external loads is found to be 112cm with 13cm spacing between inner baskets in PWR container. Also the thickness of the thinner section between the fuel basket and the surface of the cast insert is determined to be 150 mm. Regarding these dimensions of the container, the PWR fuel container is sketched to accommodate 4 square assemblies or 297 CANDU fuel 297 bundles (33 circle tubes x 9 stacks). However the top and bottom parts need to be checked again through the detail radiation shielding analysis with respects to the emplacement position and handling processes of the disposal container.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.37-50
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• 2020

In this paper, an overview of the scoping calculation results is provided with respect to criticality and radiation shielding of two KBS-3V type PWR SNF disposal systems and one NWMO-type CANDU SNF disposal system of the improved KAERI reference disposal system for SNFs (KRS⁺). The results confirmed that the calculated effective multiplication factors (k_eff) of each disposal system comply with the design criteria (< 0.95). Based on a sensitivity study, the bounding conditions for criticality assumed a flooded container, actinide-only fuel composition, and a decay time of tens of thousands of years. The necessity of mixed loading for some PWR SNFs with high enrichment and low discharge burnup was identified from the evaluated preliminary possible loading area. Furthermore, the absorbed dose rate in the bentonite region was confirmed to be considerably lower than the design criterion (< 1 Gy·hr^-1). Entire PWR SNFs with various enrichment and discharge burnup can be deposited in the KRS⁺ system without any shielding issues. The container thickness applied to the current KRS⁺ design was clarified as sufficient considering the minimum thickness of the container to satisfy the shielding criterion. In conclusion, the current KRS⁺ design is suitable in terms of nuclear criticality and radiation shielding.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.4
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• pp.469-488
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• 2022

Technology for high-level-waste disposal employing a multibarrier concept using engineered and natural barrier in stable bedrock at 300-1,000 m depth is being commercialized as a safe, long-term isolation method for high-level waste, including spent nuclear fuel. Managing heat generated from waste is important for improving disposal efficiency; thus, research on efficient heat management is required. In this study, thermal management methods to maximize disposal efficiency in terms of the disposal area required were developed. They efficiently use the land in an environment, such as Korea, where the land area is small and the amount of waste is large. The thermal effects of engineered barriers and natural barriers in a high-level waste disposal repository were analyzed. The research status of thermal management for the main bedrocks of the repository, such as crystalline, clay, salt, and other rocks, were reviewed. Based on a characteristics analysis of various heat management approaches, the spent nuclear fuel cooling time, buffer bentonite thermal conductivity, and disposal container size were chosen as efficient heat management methods applicable in Korea. For each method, thermal analyses of the disposal repository were performed. Based on the results, the disposal efficiency was evaluated preliminarily. Necessary future research is suggested.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.41-50
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• 2006

In this paper, sets of engineering analyses were conducted to renew the overall dimensions and configurations of a disposal container proposed as a prototype in the previous study. Such efforts and calculation results can provide new design variables such as the inner basket array type and thickness of the outer shell and the lid & bottom of a spent nuclear fuel disposal container. These efforts include radiation shielding and nuclear criticality analyses to check to see whether the dimensions of the container proposed from the mechanical structural analyses can provide a nuclear safety or not. According to the results of the structural analysis of a PWR disposal container by varying the diameter of the container insert, the Maximum Von Mises stress from the 102 cm-container meets the safety factor of 2.0 for both extreme and normal load conditions. This container also satisfies the nuclear criticality and radiation safety limits. This decrease in the diameter results in a weight loss of a container by $\sim20$ tons.