• Journal of Radiopharmaceuticals and Molecular Probes
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• v.7 no.2
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• pp.69-77
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• 2021

No-carrier-added holmium-166 (n.c.a ¹⁶⁶Ho) separation is performed based on the results of separation conditions using stable isotopes dysprosium (Dy) and holmium (Ho) to minimize radioactive waste from separation optimization procedures. Successful separation of two adjacent lanthanides was achieved by cation-exchange chromatography using a sulfonated resin in the H+ form (BP-800) and α-hydroxyisobutyric acid (α-HIBA) as eluent. For the identification process after separation of stable isotopes, the use of chromogenic reagents alternatively enables on-line detection because the lanthanides are hardly absorb light in the UV-vis region or exhibit radioactivity. Four different chromogenic reagents were pre-tested to evaluate suitable coloring reagents, of which 4-(2-Pyridylazo)resorcinol is the most recommendable considering the sensitivity and specificity for lanthanides. Lanthanide radioisotopes (RI) were monitored for separation with an RI detector using a lab-made separation LC system. Under the proper separation conditions, the n.c.a ¹⁶⁶Ho was effectively obtained from a large amount of 100 mg dysprosium target within 2 hrs.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.3
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• pp.175-182
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• 2009

The spent cationic exchange resins and anionic exchange resins were separated from mixed spent exchange resins by a fluidized bed gravimetric separator. The separated resins were identified by an elemental analysis and thermogravimetric analysis. The each test sample was prepared by diluting the slurry made by wet ball milling the cationic exchange resins and the anionic exchange resins separated as a spherical granular form for 24 hours. The resulting test samples showed a slurry form of less than $75{\mu}m$ of particle size and 25,000ppm of $COD_{cr}$. The decomposition conditions of each test samples from a thermal power plant were obtained with a lab-scale(reactor volume : 220mL) supercritical water oxidation(SCWO) facility. Then pilot plant(reactor volume : 24 L) tests were performed with the test samples from a thermal power plant and a nuclear power plant successively. Based on the optimal decomposition conditions and the operation experiences by lab-scale facility and the pilot plant, a commercial plant(capacity : 150kg/h) can be installed in a nuclear power plant was designed.

• Proceedings of the Korean Institute of Resources Recycling Conference
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• 2006.05a
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• pp.90-94
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• 2006

The waste solution discharged from the LCD manufacturing process contains acids like nitric, acetic and phosphoric acid and metal ions such as Al, Mo and other impurities. It is important to removal of impurities to tess than 1ppm in phosphoric acid to reuse as an etchant because the residual impurities even in sub-ppm concentration in semiconductor materials play a major role on the electronic properties. In this study, we have been clearly established that a mixed system of solvent extraction, diffusion dialysis and ion-exchange technique, which made individually the most of characteristics is developed to commercialize in an efficient system for recovering the high-purity phosphoric acid. By applying vacuum evaporation, the yield of the process are almost 99% removal of nitric acid and acetic acid was achieved. And by applying the solvent extraction method with tri-octyl phosphate(TOP) as an extractant, the removal of acetic and nitric acid from the acid mixture was achieved effectively at the ratio O/A=1/3 with four stages and the stripping of nitric acid from organic phase is attained at a ration of O/A=1 with six stages by distilled water. About 97% and 76% removal of Al and Mo were achieved by diffusion dialysis. Essentially complete less than 1ppm removal of Al, Mo by using ion exchange ion resin and purification of the phosphoric acid was obtain.

• Resources Recycling
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• v.27 no.6
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• pp.23-29
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• 2018

Recently rapid economic growth and technological development have led to an increase in the generation of waste electrical and electronic equipment (WEEE). As the amount of electric and electronic waste generated increases, the importance of processing waste printed circuit boards (PCB) is also increasing. Various studies have been conducted to recycle various valuable metals contained in a waste PCB in an environmentally friendly and economical manner. To get anode slime containing Ag and Au, Anode copper prepared from PCB scraps was used by means of electro-refining. Ag and Au recovery was conducted by leaching, direct reduction, and ion exchange method. In the case of silver, the anode slime was leached at 3 M $HNO_3$, 100 g/L, $70^{\circ}C$, and Ag was recovered by precipitation, alkali dissolution, and reduction method. In the case of gold, the nitrate leaching residues of the anode slime was leached at 25% aqua regia, 200 g/L, $70^{\circ}C$, and Au was recovered by pH adjustment, ion exchange resin adsorption, desorption and reduction method. The purity of the obtained Au and Ag were confirmed to be 99.99%.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.158-166
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• 2004

The optimal conditions are obtained for the decomposition of solid radioactive wastes, including ion exchange resin, zeolite, active charcoal, and sludge from nuclear power plant. In the process of decomposing the radioactive wastes were used the microwave acid digestion method with mixed acid. The solution after acid digestion by the following method was colorless and transparent. Each solution was analyzed with ICP-AES and AAS and the recovery yield for 5 different elements added the simulated radioactive wastes were over 94%. As an effective pre-treatment, the proposed microwave acid digestion conditions concerning the chemical trait of each radioactive waste are expected to be generally applied to above-mentioned radioactive wastes from nuclear power plant hereafter.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.4
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• pp.231-238
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• 2004

The conditions are obtained for the decomposition of solid radioactive wastes, including ion exchange resin, zeolite, charcoal, and sludge from nuclear power plant. In the process of decomposing the radioactive wastes was used the microwave acid digestion method with mixed acid. The solution after acid digestion by the following method was colorless and transparent. Each solution was analyzed with ICP-AES and AAS and the recovery yield for 5 different elements added into the simulated radioactive wastes were over $94{\%}$. The elemental analysis of destructive low and medium level radioactive wastes by the proposed microwave acid digestion conditions concerning the chemical characteristics of each radioactive waste are expected to be useful basic data for development of optimal glass formulation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.179-188
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• 2007

In this work $^{14}C\;and\;^3H$ distribution characteristics of spent resins from nuclear power plants(NPPs), pressurized water reactors(PWRs), was investigated. It was found that the recovery percent of $^{14}C$ by the wet oxidation-acid stripping was $81%{\sim}100%$ for the added activity range of $^{14}C,\;0.72\;Bq{\sim}460\;Bq$, and it was not affected by the kinds of stripping acids, 3N-HCl, $3\;N-HNO_3\;and\;3\;N-H_2SO_4$. And the recovery percent of $^3H$ by distillation using the same apparatus was $81%{\sim}101%$ for the added activity range of $^3H,\;0.60\;Bq{\sim}435\;Bq$. Among the tested stripping acids, 3\;N-HCl, $3\;N-HNO_3\;and\;3\;N-H_2SO_4$, only the trapped $^3H$ solution by distillation in $3\;N-H_2SO_4$ was compatible with the 3H scintillator, Ultimagold XR. Neither of the $^{14}C\;and\;^3H$ trapping solutions from the spent ion exchange resin samples by the wet oxidation-3 $N-H_2SO_4$ stripping contained gamma nuclides. However, some gamma nuclides, $^{60}Co,\;^{134}Cs,\;^{137}Cs\;and\;^{54}Mn$, were found in the trapped $^3H$ solutions of the spent resins by the wet oxidation-3 N-HCl stripping. It was the same for the $^3H$ trapping solutions of the spent resins by Sample Oxidizer(PACKARD MODEL 307). Meanwhile only two nuclides, $^{134}Cs,\;and\;^{134}Cs$, were found in the $^{14}C$ trapping solutions of the spent resins by Sample Oxidizer(PACKARD MODEL 307). It was found that most of the $^{14}C$ in the spent resins existed as inorganic carbon form, more than about 70% of the total $^{14}C$ content. Among the analyzed 30 spent ion exchange resin samples, the average concentration of $^{14}C$ and $^3C$ for the high radioactive samples, 8 samples, was $19000\;Bq/g{\pm}41000\;Bq/g,\;670\;Bq/g{\pm}460\;Bq/g$ and that for the low radioactive samples, 22 samples, was $4.2\;Bq/g{\pm}4.3\;Bq/g,\;6.0\;Bq/g{\pm}5.3\;Bq/g$, respectively. And the average $^{14}C/^3H$ ratio for the high radioactive samples, was higher, 28, than that of low radioactive samples, 0.70. Some linear relationship trend was found between the activity concentrations of $^{14}C\;and\;^3H$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.3
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• pp.165-174
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• 2004

In order to simultaneously vitrify the ion exchange resin(IER) and combustible dry active waste(DAW) generated from Korean nuclear power plants, a vitrification pilot test was conducted using an induction cold crucible melter(CCM) . The energy necessary for startup of the glass using a Ti-ring was evaluated as about 290 kWh. The power supplied from a high frequency generator to melt the glass properly was ranged from 160 to 190 kW without any interruption. When the mixture of the IER and DAW was fed into the CCM, the concentration of CO was lowered up to 1/40 compared to feeding the IER solely. It may be caused by the DAW which can produce about 1.8 times higher heat compared to the IER. When the swelling phenomenon occurred in the glass melt, the concentration of $NO_2$, oxidizing gas, was higher than NO, reducing gas. Total feed amounts of the IER and DAW were 368 and 751 kg, respectively. And then, about 74 of volume reduction factor was achieved.

• Journal of the Korean Ceramic Society
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• v.38 no.9
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• pp.839-845
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• 2001

In order to vitrify the Ion-Exchange Resin(IER), Dry Active Waste(DAW), and borate concentrate generated from the commercial nuclear facilities, the glass formulation study based on the their compositions was performed. Two glasses named as RG-1 and DG-1 were formulated as the candidate glasses for the vitrification of hte IER and DAW, respectively. A glass named as MG-1 was also formulated as a candidate glass for the vitrification of the mixed wastes containing the IER, DAW, and borate concentrate. The process parameters, product qualities, and economics were evaluated for the candidate glasses and confirmed experimentally for the some properties. The glass viscosity and electrical conductivity as the process parameters were in the desired ranges. the product qualities such as glass density, chemical durability, phase stability, etc. were satisfactory. In case of vitrifying the wastes using our developed glass formulation study, the volume reduction factors for the IER, DAW and mixed wastes were evaluated as 21, 89 and 75, respectively.

• Applied Chemistry for Engineering
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• v.31 no.5
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• pp.575-580
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• 2020

In this study, we simulated an amine regeneration process with heat-stable salts removal unit. We derived the optimal operating conditions considering the flow rate of waste, the removal rate of heat-stable salts, and the loss rate of MDEA (methyl diethanolamine). In the amine regeneration process that absorbs and removes acid gas, heat-stable salt impairs the absorption efficiency of process equipment and amine solution. An ion exchange resin method is to remove heat-stable salts through neutralization by using a strong base solution such as NaOH. The acid gas removal process was established using the Radfrac model, and the equilibrium constant of the reaction was calculated using Gibbs free energy. The removed amine solution is separated and flows to the heat-stable salts remover which is modeled by using the Rstoic model with neutralization reaction. Actual operation data and simulation results were compared and verified, and also a case study was conducted by adjusting the inflow mass of removal unit followed by suggesting optimal conditions.