• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.3
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• pp.189-199
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• 2004

A bridge transported servo manipulator (BTSM) system for the advanced spent fuel conditioning process (ACP) has been developed to overcome the limitation of access, which is a drawback of mechanical master-slave manipulators (MSM) for the equipment maintenance. The servo manipulator is composed of a slave manipulator attached to the telescoping tubesets equipped with the overhead bridge installed at a hot cell and a master manipulator installed at an out-of-hot cell. Each manipulator has 7 degrees-of-freedom (DOF): a body rotation, an upper-arm tilt, a lower-arm tilt, a lower-arm rotation, a wrist pan & tilt, and a grasp motion. A wire-driven mechanism for a lower-arm rotation, a wrist pan and tilt, and a grasp motion of the manipulator has been adopted to increase the handling capacity compared to the manipulator weight and decrease the friction. The main disadvantage of the wire-driven mechanism is that if one link is in motion, other links can be affected. In this paper, the transmission characteristics among the wire-driven links have been formulated to overcome this drawback. The unexpected behaviors are confirmed by analyses of transmission characteristics as well as experiments. Also, the experimental results show that the unexpected behaviors are greatly decreased by the proposed compensation equations.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.33-39
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• 2010

The electrolytic reduction of a spent oxide fuel involves a liberation of the oxygen in a molten LiCl electrolyte, which results in a chemically aggressive environment that is too corrosive for typical structural materials. Accordingly, it is essential to choose the optimum material for the processing equipment that handles the high molten salt. In this study, hot corrosion studies were performed on bare as well as coated superalloy specimens after exposure to lithium molten salt at $675^{\circ}C$ for 216 h under an oxidizing atmosphere. The IN713LC superalloy specimens were sprayed with an aluminized NiCrAlY bond coat and then with an $Y_2O_3$ top coat. The bare superalloy reveals an obvious weight loss due to spalling of the scale by the rapid scale growth and thermal stress. The chemical and thermal stability of the top coat has been found to be beneficial for increasing to the corrosion resistance of the structural materials for handling high temperature lithium molten salts.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.3
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• pp.199-206
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• 2017

A geological repository for the disposal of high-level radioactive waste is generally constructed in host rock at depths of 500~1,000 meters below the ground surface. A geological repository system consists of a disposal canister with packed spent fuel, buffer material, backfill material, and intact rock. The buffer is indispensable to assure the disposal safety of high-level radioactive waste, and it can restrain the release of radionuclides and protect the canister from the inflow of groundwater. Since high temperature in a disposal canister is released to the surrounding buffer material, the thermal properties of the buffer material are very important in determining the entire disposal safety. Even though there have been many studies on thermal conductivity, there have been only few studies that have investigates the specific heat capacity of the bentonite buffer. Therefore, this paper presents a specific heat capacity prediction model for compacted Gyeongju bentonite buffer material, which is a Ca-bentonite produced in Korea. Specific heat capacity of the compacted bentonite buffer was measured using a dual probe method according to various degrees of saturation and dry density. A regression model to predict the specific heat capacity of the compacted bentonite buffer was suggested and fitted using 33 sets of data obtained by the dual probe method.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.115-122
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• 2010

In this study, the removal of Cs and Tc from a simulated fission products (FP) solution which were co-dissolved with U during the oxidative-dissolution of spent fuel in a mixed carbonate solution of $(Na_2CO_3-NaHCO_3)-H_2O_2$ was investigated by using a selective precipitation method. As Cs and Tc might cause an unstable behavior due to the high decay heat emission of Cs as well as the fast migration of Tc when disposed of underground, it is one of the important issues to removal them in views of the increase of disposal safety. The precipitation of Cs and Re (as a surrogate for Tc) was examined by introducing sodium tetraphenylborate (NaTPB) and tetraphenylphosponium chloride (TPPCl), respectively. Precipitation of Cs by NaTPB and that of Re by TPPCl were completed within 5 minutes. Their precipitation rates were not influenced so much by the temperature and stirring speed even if they were increased by up to $50^{\circ}C$ and 1,000 rpm. However, the pH of the solution was found to have a great influence on the precipitation with NaTPB and TPPCl. Since Mo tends to co-precipitate with Re at a lower pH, especially, it was effective that a selective precipitation of Re by TPPCl was carried out at pH of above 9 without co-precipitation of Mo and Re. Over 99% of Cs was precipitated when the ratio of [NaTPB]/[Cs]>1 and more than 99% of Re, likewise, was precipitated when the ratio of [TPPCl]/[Re]>1.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.75-94
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• 2019

As the present single-layer repository concept requires too large an area for the site of the repository, a multi-layer repository concept has been suggested to improve the disposal density. The effects of the excavation damaged zone around the multi-layer repository constructed in the deep host rock on the temperature distribution in the repository were analyzed. For the thermal analysis of the multi-layer repository, the hydrothermal model was used to consider the resaturation process occurring in the buffer, backfill and rock. The existence of an excavation damaged zone has a significant effect on the temperature distribution in the repository, and the maximum peak temperatures of double-layer and triple-layer repositories can rise to $7^{\circ}C$ and $12^{\circ}C$, respectively depending on the size of the excavation damaged zone and the degree of decrease of the thermal conductivity. The dominant factor affecting the peak temperature in the multi-layer repository is the decrease of thermal conductivity in the excavation damaged zone, and the excavation damaged zone formed around the deposition hole has more significant effects on the peak temperature than does the excavation damaged zone formed around the disposal tunnel.

• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.67-73
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• 1995

To investigate the storage behavior of the defective LWR spent fuel, air-oxidation tests on non-irradiated and irradiated U $O_2$ were performed. The weight gains of non-irradiated U $O_2$ specimens are characterized by the S-shape curves at 250-40$0^{\circ}C$ temperature range. One hundred percent conversion of U $O_2$ to U$_3$ $O_{8}$ corresponds with about 4％ weight increase. The activation energies are 110 kJ/mol above 35$0^{\circ}C$ and 153 kJ/mol below 35$0^{\circ}C$. The irradiated U $O_2$ specimens with about 35 GWD/MTU burnup were oxidized at 300-40$0^{\circ}C$ in air. They show a rapid increase of weight gain at the initial stage and a slow increase at the later stage when compared with non-irradiated U $O_2$. The activation energy under these conditions is 95 kJ/mol. Burnup and aging effects of irradiated U $O_2$ were also investigated at 35$0^{\circ}C$ in air environment, but the specimens appears insensitive to these variables.s.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.3
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• pp.253-257
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• 2013

Geology, hydrogeology, and geochemistry are the main technical siting factors of a geological repository for spent nuclear fuels. This paper evaluated the siting criteria of overseas geological repository with related to hydrogeologic properties, such as hydraulic conductivity, partitioning coefficient, dispersion coefficient, boundary condition, and water age. Each country establishes the siting criteria based on its important geological backgrounds and information, and social environment. For example, Sweden and Finland that have decided a crystalline rock as a host rock of a geological repository show different siting criteria for hydraulic conductivity. In Sweden, it is preferable to avoid area where the hydraulic conductivity on a deposition hole scale (~30m) exceeds $10^{-8}m/s$, whereas Finland does not decide any criterion for the hydraulic conductivity because of limited data for it. In addition, partitioning coefficients should be less than 10-1 of average value in Swedish crystalline bedrock. However, the area where shows 100 times less than average partitioning coefficients of radionuclides in crystalline rock should be avoided in Sweden. In German, the partitioning coefficients for the majority of the long-term-relevant radionuclides should be greater than or equal to $0.001m^3/kg$. Therefore, it is strongly required to collect much and exact information for the hydrogeologic properties in order to set up the siting criteria.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.2
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• pp.77-84
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• 2005

A voloxidizer for a hot cell demonstration, that handles spent fuels of a high radiation level in a limited space should be small and spent fuel powders should not be dispersed out of the equipment involved. In this study a density rate equation as well as the Stokes'equation has been proposed in order to obtain the theoretical terminal velocity of powders. The terminal velocity of U$_{3}$O$_{8}$ has been predicted by using the terminal velocity of SiO$_{2}$, and then determination has been the optimum air flow rate which is able to prevent powders from scattering. An equation which has shown a relationship between theoretical terminal velocities of U$_{3}$O$_{8}$ and SiO$_{2}$ has been derived with the help of the Stokes'equation, and then an experimental verification made for the theoretical Stokes' equation of SiO$_{2}$ by means of an experimental device made of acryl. The theoretical terminal velocity based on the proposed density rate equation has been verified by detecting U$_{3}$O$_{8}$ powders in a filter installed in the mock-up voloxidizer. As the results, the optimum air flow rates seem to be 20 LPM by the Stokes'equation while they are 14.5 L/min by the density rate equation. At the experiments with the mock-up voloxidizer, a trace amount of U$_{3}$O$_{8}$ seems to be detectable at the air flow rate of 14.5 L/min by the density rate equation, but U$_{3}$O$_{8}$ powders of 7$\mu$m diameter seem detectable at the air flow rate of 20 L/min by the Stokes'equation. It is revealed that 14.5 L/min is the optimum air flowe rate which is capable of preventing U$_{3}$O$_{8}$ powders from scattering in the UO$_{2}$ voloxidizer and the proposed density rate equation is proper to calculate the terminal velocity of U$_{3}$O$_{8}$ powders.

• Journal of Korean Tunnelling and Underground Space Association
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• v.24 no.6
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• pp.451-464
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• 2022

The high-level nuclear waste repository is a deep geological disposal system exposed to complex environmental conditions such as high temperature, radiation, and ground-water due to handling spent nuclear fuel. Continuous exposure can lead to cracking and deterioration of the structure over time. On the other hand, the high-level nuclear waste repository requires an ultra-long life expectancy. Thus long-term structural health monitoring is essential. Various sensors such as an accelerometer, earth pressure gauge, and displacement meter can be used to monitor the health of a structure, and a piezoelectric sensor is generally used. Therefore, it is necessary to develop a highly durable sensor based on the durability assessment of the piezoelectric sensor. This study designed an accelerated life test for durability assessment and life prediction of the piezoelectric sensor. Based on the literature review, the number of accelerated stress levels for a single stress factor, and the number of samples for each level were selected. The failure mode and mechanism of the piezoelectric sensor that can occur in the environmental conditions of the high-level waste repository were analyzed. In addition, two methods were proposed to investigate the maximum harsh condition for the temperature stress factor. The reliable operating limit of the piezoelectric sensor was derived, and a reasonable accelerated stress level was set for the accelerated life test. The suggested methods contain economical and practical ideas and can be widely used in designing accelerated life tests of piezoelectric sensors.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.9-17
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• 2007

In order to compare the correlation of radioactivity ratio between the radwaste streams and the primary coolant of PWR NPPs, A RCS sampling kit was installed to primary coolant system for the collection of the radionuclides during the normal operation of NPPs. RCS samples were collected from PWR type of domestic NPPs through 2004 to 2005, and pretreated with acid microwave digestion or leaching method to assay quantitatively of several interesting radionuclides. The radioactivity ratios of $^{137}Cs\;to\;^{60}Co$ in a filter cartridge and a resin cartridge were 2.3E-2 and 7.3E-1, respectively. At a same period of the reactor operating cycle, the radioactivity ratios of $^{137}Cs\;to\;^{60}Co$ were 6.3E-1 for a evaporator bottom, 6.7E-1 for a spent resin, and 5.6E-2 for a dry active waste, so that these radwaste streams were identified as having similar characteristics with the corresponding RCS samples.