• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.3
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• pp.181-188
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• 2010

Metal surface decontamination characteristics were investigated by using a laser ablation method. A second harmonic generation of a Q-switched Nd:YAG laser with a wave length of 532 nm, a pulse energy of 150 mJ and a pulse width of 5 ns was employed to assess the decontamination performance for metal surfaces contaminated with $CsNO_3$, $Co(NH_4)_2(SO_4)_2$, $Eu_2O_3$ and $CeO_2$. The ablation behavior was investigated for the decontamination variables such as a number of laser shots, laser fluence and an irradiation angle. Their optimum values were found to be 8, 13.3 J/$cm^2$ and $30^{\circ}$, respectively. The decontamination efficiency was different depending on the kinds of the contaminated ions, due to their different melting and boiling points and was in the order: $CsNO_3>Co(NH_4)_2(SO_4)_2>Eu_2O_3>CeO_2$. We also evaluated a correlation between the metal ablation thickness and the number of laser shots for the different laser fluences.

• Bulletin of the Korean Chemical Society
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• v.1 no.2
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• pp.49-54
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• 1980

Solutions of $Ni^{2+}$ and $Cd^{2+}$ were mixed with the solutions of various dibasic organic acids in the presence of cation exchange resin at room temperature. The distribution ratios of the metal ions between resin and solution were measured, using radioactive metal ions as tracer. From the observed variation of the distribution ratios with acid anion concentrations, it was concluded that $Ni^{2+}$ and $Cd^{2+}$ formed one-to-one complexes with succinate, malonate, o-phthalate and tartarate ions in aqueous, 20 % ethanol-water and 20 % acetone-water solutions. The results of the present study indicated that the relative stabilities of the complexes in solution increased generally in the order : $Ni^{2+}$ < $Cd^{2+}$ complexes. Succinate < malonate < o-phthalate < tartarate complexes. Aqueous < mixed solvent systems.

• Bulletin of the Korean Chemical Society
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• v.1 no.3
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• pp.78-82
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• 1980

Solutions of $Cd^{2+}$, $Co^{2+}$ and $Ni^{2+}$ were mixed with the solutions of hydroxycarboxylic acids such as salicylic, lactic and mandelic acids in the presence of cation exchange resin at room temperature. The distribution ratios of the metal ions between resin and solution were measured, using radioactive metal ions as tracer. From the observed variation of the distribution ratios with the acid anion concentrations, it was concluded that $Cd^{2+}$, $Co^{2+}$ and $Ni^{2+}$ formed the one-to-one complexes with salicylate, lactate and mandelate ions in aqueous, 20 % ethanol-water and 20 % acetone-water solutions. The results of the present study indicated that the relative stabilities of the metal-acid complexes in solution increased in the order: $Cd^{2+}$ <$Co^{2+}$ <$Ni^{2+}$ complexes. Salicylate

• Journal of Radiation Protection and Research
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• v.27 no.2
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• pp.95-100
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• 2002

Conventional and modified electrolytic decontamination experiment were performed in the 1.7 M solution of sodium sulfate and sodium nitrate tot decontamination of carbon steel as the simulated metal wastes which have been produced in large amounts from nuclear power plants. Anode ant cathode were used as inconel and titanium respective. The reaction time and temperature were 1 hr and $25^{\circ}C$ The analyses were performed of the characteristics such as weight loss arid thickness change of metal waste. suspended solid in electrolyte and SEM observation. In modified electrolyte decontamination system with increased current density ranged from 0.1 to $0.6A/cm^2$, the metal waste showed thickness changes of $0.48{\pm}0.005$ to $67.7{\pm}0.02{\mu}m$ in 1.7 M sodium sulfate and those of $0.06{\pm}0.005$ to $17.7{\pm}0.05{\mu}m$ in sodium nitrate. Metal waste in modified electrolyte decontamination system showed the thickness change of $9.8{\pm}0.01{\mu}m$ while it reacted up to $3.7{\pm}0.03{\mu}m$ in conventional system with $0.3 A/cm^2$ of current density and 1.7 M sodium sulfate. Decontamination efficiencies of modified electrolytic process ate much hither than that of conventional electrolytic process when both are applied to metal waste.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.2
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• pp.141-154
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• 2015

In general, conventional criticality analysis for spent fuel transport/storage systems have been performed based on the assumption of fresh fuel concerning the potential uncertainties from number density calculations of actinide nuclides and fission products in spent fuel. However, these evaluation methods cause financial losses due to an excessive criticality margin. In order to overcome this disadvantage, many studies have recently been conducted to design and commercialize a transportation and storage cask applied to the Burnup Credit (BUC). This study conducted an assessment to ensure criticality safety for reactor operating parameters, axial burn-up profiles and misload accident conditions, which are the factors that are likely to affect criticality safety when the BUC is applied to the dual-purpose cask under development at the KOrea RADioactive waste agency (KORAD). As a result, it was found that criticality resulting from specific power, changed substantially and relied on conditions of low enrichment and high burn-up. Considering the end effect in the case of high burn-up produced a positive-definite result. In particular, the increment of maximum effective multiplication factors due to misloading was 0.18467, confirming that misload is a factor that must be taken into account when applying the BUC. The results of this study may therefore be utilized as references in developing technologies to apply the BUC to domestic models and operational procedures or preventing any misload accidents during the process of spent fuel loading.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.291-303
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• 2020

The Kori-Unit 1 nuclear power plant, which is scheduled to be decommissioned after permanent shutdown, is expected to generate large amounts of various types of radioactive waste during the decommissioning process. Among these, nuclear reactors and internal structures have high levels of radioactivity and the dismantled structure must have the proper size and weight on the primary side. During decommissioning, it is important to prepare an appropriate and efficient disposal method through analysis of the disposal status and the legal restrictions on wastes generated from the reactors and internal structures. Nuclear reactors and internal structures generate radioactive wastes of various levels, such as medium, very low, and clearance. A radiation evaluation indicates that wastes in the clearance level are generated in the reactor head and upper head insulation. In this study, a clearance waste safety evaluation was conducted using the RESRAD-RECYCLE code, which is a safety evaluation code, based on the activation evaluation results for the clearance level wastes. The clearance scenario for the target radioactive waste was selected and the maximum individual and collective exposure doses at the time of clearance were calculated to determine whether the clearance criteria limit prescribed by the Nuclear Safety Act was satisfied. The evaluation results indicated that the doses were significantly low, and the clearance criteria were satisfied. Based on the safety assessment results, an appropriate metal recycle and disposal method were suggested for clearance, which are the subject of the deregulation of internal structures of nuclear power plant.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.3
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• pp.395-405
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• 2020

Compartment fire tests were performed using kerosene and Jet A-1 as fire sources to evaluate the relationship between flame temperature and opening size. The tests were performed for a fire caused by the release of kerosene owing to vehicle impact, and for a fire caused by the release of Jet-A-1 owing to airplane collision. The compartment fire tests were performed using a 1/3-scale model of a metal storage cask when the flame temperature was deemed to be the highest. We found the combustion time of Jet-A-1 to be shorter than that of kerosene, and consequently, the flame temperature of Jet-A-1 was measured to be higher than that of kerosene. When the opening was installed on the compartment roof, even though the area of the opening was small, the ventilation factor was large, resulting in a high flame temperature and long combustion. Therefore, the position of the opening is a crucial factor that affects the flame temperature. When the metal storage cask was stored in the compartment, the flame temperature decreased proportionally with the energy that the metal storage cask received from the flame.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.363-367
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• 2003

Much attention has been given to an electrochemical reduction process for converting uranium oxide to uranium metal in molten salt. The process has the versatility of being adopted for reducing other actinide and rare-earth metals from their oxides. Using the metal oxide to be reduced as a integrated cathode designed originally and inert conductors as anodes, oxygen anions are removed from the cathode and oxidized at the surface of the anodes in a molten salt cell. However, the electrochemical properties of alkali and alkali-earth metal oxides in molten salt have not been investigated thoroughly, which made the process incomplete when it is considered as a unit process in a back-end fuel cycle. It is well known that cesium and strontium Isotopes in spent fuel are main contributors for head load. The properties of cesium, strontium, and barium oxides such as the dissolution rates and reduction potentials in molten LiC1 dissolving $Li_2O$ are examined.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.4
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• pp.299-314
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• 2014

A lifting device is used to deal with transport cask for the transportation of spent fuels from nuclear power plants. This study performed theoretical analysis and numerical simulation to evaluate the structural integrity of the lifting device based on Nuclear Safety and Security Commission(NSSC) Notice No.2013-27 and US 10CFR Part 71 ${\S}71.45$. The results of theoretical analysis showed that the maximum stresses of all components were below the allowable values. This result confirmed that the lifting device was structurally safe during operation. The results of finite element analysis also showed that it was evaluated to satisfy the design criteria bothyielding and ultimate condition. All components have been shown to ensure the structural safety due to sufficient safety margins. In other words, the safety factor was 3 or more for the yielding condition and was 5 or more for the ultimate condition.

• Journal of the Mineralogical Society of Korea
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• v.26 no.4
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• pp.219-228
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• 2013

To understand characteristics of biogeochemical corrosion for the metal canisters that usually contain the radioactive wastes for a long-term period below the ground, some metal materials consisting of cast iron and copper were reacted for 3 months with D. desulfuricans, a sulfate-reducing bacterium, under a reducing condition. During the experiment, concentrations of dissolved metal ions were periodically measured, and then metal specimen and surface secondary products were examined using the electron microscopy to know the chemical and mineralogical changes of the original metal samples. The metal corrosion was not noticeable at the absence of D. desulfuricans, but it was relatively greater at the presence of the bacterium. In our experiment, darkish metal sulfides such as mackinawite and copper sulfide were the final products of biogeochemical metal corrosion, and they were easily scaled off the original specimen and suspended as colloids. For the copper specimen, in particular, there appeared an accelerated corrosion of copper in the presence of dissolved iron and bacteria in solution, probably due to a weakening of copper-copper binding caused by a growth of other phase, iron sulfide, on the copper surface.