• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.05a
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• pp.191-192
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• 2017

• Journal of the Korean institute of surface engineering
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• v.42 no.1
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• pp.53-57
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• 2009

Recovery of heavy-metallic component from a waste solution of factory was undertaken by the solar cell electricity. The solution was obtained from an electrolytic etching process of 316L stainless steel. The electrolysis of the solution for recovery of heavy metallic components was made with platinum plated titanium mesh anode and copper plate cathode. Analysis for the solution and electro-winned materials were made by EDS, XRD and SEM. Iron, chromium, and sulfur components were recovered on the cathode from the solution. Result of EDS analysis for the electro-winned materials revealed that some metal oxide were contained in the recovered material. The recovered materials were expected to have metallic form only by the electrolysis, but metal compounds were contained because of weak solar cell power. Nickel and manganese component in the solution doesn't recovered by this electrolysis process, but they made a sludge with phosphoric acid in the solution.

• Resources Recycling
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• v.10 no.3
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• pp.29-36
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• 2001

Cyclic voltammetry has been applied in the basic studies for the treatment and recovery of heavy metal component contained in wastewater by electrochemical reduction. The electrochemical behaviors of zinc ion for zinc metal electrode and carbon elec-trode were characterized by voltammograms and it was showed that zinc ions were reduced to metallic form below -0.76 V vs SHE. The change in the features of crystalline form of metallic zinc upon oxidation and reduction reaction was observed by X-ray diffraction method and the Am analysis verified that zinc ions were reduced to metal on copper plate. The results of this study were regarded to be important and meaningful in the treatment of heavy metal containing wastewater and, as a result, in the obtainment of metallic product by electrochemical method.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.278-283
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• 2005

Cement and paraffin waste form were prepared with a acid extraction method for the analysis of radionuclides generated from nuclear power plants. The acid extraction method was carried out with $HNO_3-HCl$ acid. At first, we compared the method with the microwave acid digestion method using SRM. The solutions of decomposed SRM were then analyzed by AAS and ICP-AES. The acid extraction method had shown good results as microwave acid digestion method. This method provided recovery values greater than $80\%$ for metallic elements.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.125-133
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• 2017

KAERI is conducting research on spent cladding hulls and additive metals to generate a solidification host matrix for the noble metal fission product waste in anode sludge from the electro-refining process to minimize the volume of waste that needs to be disposed of. In this study, alloy compositions Zr-17Cr, Zr-22Cr, and Zr-27Cr were prepared with or without eight noble metals representing fuel waste using induction melting. The microstructures of the resulting alloys were characterized and electrochemical corrosion tests were conducted to evaluate their corrosion characteristics. All the compositions had better corrosion characteristics than other Zr-based alloys that were evaluated for comparison. Analysis of the leach solution after the corrosion test of the Zr-22Cr-8NM specimen indicated that the noble metals were not leached during corrosion under 500 mV imposed voltage, which simulates a highly oxidizing disposal environment. The results of this study confirm that Zr-Cr based compositions will likely serve as chemically stable waste forms.

• Journal of the Mineralogical Society of Korea
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• v.27 no.4
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• pp.207-222
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• 2014

In order to recover the metallic copper powder from the mine-waste rock, heap bioleaching, Fe removal and electrowinning experiments were carried out. The results of heap leaching with the mine-waste rock sample containing 0.034% Cu showed that, the leaching rate of Cu were 61% and 62% in the bacteria leaching and sulfuric acid leaching solution, respectively. Sodium hydroxide (NaOH), hydrogen peroxide ($H_2O_2$) and calcium hydroxide ($Ca(OH)_2$) were applied to effectively remov Fe from the heap leaching solution, and then $H_2O_2$ was selected for the most effective removing Fe agent. In order to prepare the electrolytic solution, $H_2O_2$ were again treated in the heap leaching, and Fe removal rates were 99% and 60%, whereas Cu removal rates were 5% and 7% in the bacteria and sulfuric acid leaching solutions, respectively. After electrowinning was examined in these leaching solution, the recovery rates of Cu were obtained 98% in bacteria and obtained 76% in the sulfuric leaching solution. The dendritic form of metallic copper powder was recovered in both leaching solutions.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.3
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• pp.373-382
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• 2020

This study proposes a method of separating uranium (U) and minor actinides from rare earth (RE) elements in the LiCl-KCl salt system. Several RE metals were used to reduce UCl₃ and MgCl₂ from the eutectic LiCl-KCl salt systems. Five experiments were performed on drawdown U and plutonium (Pu) surrogate elements from RECl₃-enriched LiCl-KCl salt systems at 773 K. Via the introduction of RE metals into the salt system, it was observed that the UCl₃ concentration can be lowered below 100 ppm. In addition, UCl₃ was reduced into a powdery form that easily settled at the bottom and was successfully collected by a salt distillation operation. When the RE metals come into contact with a metallic structure, a galvanic interaction occurs dominantly, seemingly accelerating the U recovery reaction. These results elucidate the development of an effective and simple process that selectively removes actinides from electrorefining salt, thus contributing to the minimization of the influx of actinides into the nuclear fuel waste stream.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.358-362
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• 2003

An electrochemical reduction technology which can reduce the decay heat, volume, and radioactivity of spent fuel by a factor of quarter by converting oxide type spent fuel to a metallic form in a molten salt was developed and mock-up test in a 5kg $U_3O_8$/batch scale was carried out. The electrochemical reaction was analyzed regarding the operational factors. The research efforts was also concentrated on the apparatus development for a hot test. Fresh $U_3O_8$ powder was metallized with a more than 99% yield via this electrochemical technology and design data for the 20kg $U_3O_8$/batch scale apparatus were also obtained.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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• pp.259-268
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• 2022

Molten salt consisting primarily of eutectic LiCl-KCl is currently being used in electrorefiners in the Fuel Conditioning Facility at Idaho National Laboratory. Options are currently being evaluated for storing this salt outside of the argon atmosphere hot cell. The hygroscopic nature of eutectic LiCl-KCl makes is susceptible to deliquescence in air followed by extreme corrosion of metallic cannisters. In this study, the effect of occluding the salt into a zeolite on water sorption/desorption was tested. Two zeolites were investigated: Na-Y and zeolite 4A. Na-Y was ineffective at occluding a high percentage of the salt at either 10 or 20wt% loading. Zeolite-4A was effective at occluding the salt with high efficiency at both loading levels. Weight gain in salt occluded zeolite-4A (SOZ) from water sorption at 20% relative humidity and 40℃ was 17wt% for 10% SOZ and 10wt% for 20% SOZ. In both cases, neither deliquescence nor corrosion occurred over a period of 31 days. After hydration, most of the water could be driven off by heating the hydrated salt occluded zeolite to 530℃. However, some HCl forms during dehydration due to salt hydrolysis. Over a wide range of temperatures (320-700℃) and ramp rates (5, 10, and 20℃ min^-1), HCl formation was no more than 0.6% of the Cl^- in the original salt.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.1 no.1
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• pp.47-53
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• 2003

An electrochemical reduction technology which can reduce the decay heat, volume, and radioactivity of spent fuel by a factor of quarter through converting oxide type spent fuel to a metallic form in a molten salt was developed and tests in a scale of g (3- 40g $U_{3}O_{8}$ batch) have been carried out by Korea Atomic Energy Research Institute. In this research, the reaction apparatus in a scale of 5kg $U_{3}O_{8}$ batch was designed and manufactured for the mock-up test to obtain design data of the apparatus which will be used for the hot test in a scale of 20kg $U_{3}O_{8}$ batch. The electrochemical reduction behavior of $U_{3}O_{8}$ was analyzed regarding the operational factors and fresh $U_{3}O_{8}$ powder was metallized with a more than 99% yield verifying the process validity of electrochemical reduction process in a kg scale.