• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.28-41
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• 2004

The study is aimed to assess the usefulness of the mixture design for spent resin immobilization in cement. Although a considerable amount of research has been carried out to determine the limits for the composition of an acceptable resin-cement mixture, no efficient experimental strategy exists that explores the full properties of waste form against composition relationship. In order to gain an overall view, this report introduces the method of mixture design and mixture analysis, and describes the design of experiment of the 5-component mixture with the constraint conditions. The mathematic models of 28-day compressive strength varying with the ingredients are fitted, and the main effect and interaction effect of two ingredients are identified quantitatively along with the graphical interpretation using the response trace plot and contour plots.

• Journal of the Korean Electrochemical Society
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• v.2 no.4
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• pp.183-189
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• 1999

A study on the destruction of organic cation and anion exchange resins by electro-generated Ag(II) as a mediator was carried out to develop the ambient-temperature aqueous process, known as Ag(II)-mediated electro-chemical oxidation (MEO) process, for the treatment of a large quantity of spent organic ion exchange resins as the low and Intermediated-level radioactive wastes arising from the operation, maintenance and repairs of nuclear facilities. The effects of controllable process parameters such as applied current density, temperature, and nitric acid concentration on the MEO of organic ion exchange resins were investigated. The cation exchange resin was completely decomposed to $CO_2$. The current efficiency increased with a decrease in applied current density while nitric acid concentration and temperature on the MEO of cation exchange resin did not affect the MEO. On the other hand, anion exchange resins were decomposed to CO and $CO_2$. The ultimate conversion to CO was about $10\%$ regardless of temperature. The destruction efficiencies to $CO_2$ were dependent upon temperature and the effective destruction of anion exchange resin could be obtained above $60^{\circ}C$.

• Journal of Korean Society on Water Environment
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• v.29 no.6
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• pp.777-781
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• 2013

Ion-exchange technology is one of the best for removing nitrate from drinking water. However, problems related to the disposal of spent brine from regeneration of exhausted resins must be overcome so that ion exchange can be applied more widely and economically, especially in small communities. In this background, a combined bio-regeneration and ion-exchange system was operated in order to prove that nitrate-laden resins could be bio-regenerated through direct contact with denitrifying bacteria. A nitrate-selective A520E resin was successfully regenerated by denitrifying bacteria. The bio-regeneration efficiency of nitrate-laden resins increased with the amount of flow passed through the ion-exchange column. When the fully exhausted resin was bio-regenerated for 5 days at the flowrate of 30 BV/hr and MLSS concentration of $125{\pm}25mg/L$, 97.5% of ion-exchange capacity was recovered. Measurement of nitrate concentrations in the column effluents also revealed that less than 5% of nitrate was eluted from the resin during 5 days of bio-regeneration. This result indicates that the main mechanism of bio-regeneration is the direct reduction of nitrate by denitrifying bacteria on the resin.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.293-294
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• 2005

• Analytical Science and Technology
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• v.18 no.4
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• pp.298-308
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• 2005

The methods for determining C-14 and tritium contents in the spent nuclear fuel sample were developed. The carbon-14($^{14}CO_2$) released during the dissolution of the spent fuel sample and $CaCO_3$ ($CO_2$ carrier) with 8 M $HNO_3$ at $90^{\circ}C$ was collected in trap containing 1.5 M NaOH. The volatile radioactive iodine evolved when the spent fuel was dissolved, was trapped on to Ag-silicagel (Ag-impregnated silicagel) adsorbent in column which is connected to two NaOH traps. The solutions which contain tritium as HTO after fuel dissolution were decontaminated by deionization with a mixture of cation and anion exchange resins and inorganic ionexchangers. The amount of C-14 in the trap solutions and the HTO concentration in the resulting deionization water were then determined by liquid scintillation counting.

• Resources Recycling
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• v.31 no.3
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• pp.73-80
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• 2022

Reduction smelting of spent lithium-ion batteries at high temperature produces metallic alloys. Following solvent extraction of the leaching solutions of these metallic alloys with either sulfuric or hydrochloric acid, the raffinate is found to contain Ni(II), Co(II), Mn(II), and Si(IV). In this study, two cationic exchange resins (Diphonix and P204) were employed to investigate the loading behavior of these ions from synthetic sulfate and chloride solutions. Experimental results showed that Ni(II), Co(II), and Mn(II) could be selectively loaded onto the Diphonix resin from a sulfate solution of pH 3.0. With a chloride solution of pH 6.0, Mn(II) was selectively loaded onto the P204 resin, leaving Ni(II) and Si(IV) in the effluent. Elution experiments with H₂SO₄ and/or HCl resulted in the complete recovery of metal ions from the loaded resin.

• Resources Recycling
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• v.31 no.6
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• pp.25-35
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• 2022

The adsorption/desorption behavior and separation conditions of vanadium and tungsten ions were investigated using a gel-type anion-exchange resin. In the adsorption experiment with the initial acidity of the solution, the adsorption rate of vanadium was remarkably low in strong acids and bases. Additionally, the adsorption rate of tungsten was low in a strong base. An increase in the reaction temperature increased the adsorption reaction rate and maximum adsorption. The effect of tungsten on the maximum adsorption was minimal. The adsorption isotherms of vanadium and tungsten on the ion-exchange resin were suitable for the Langmuir adsorption isotherms of both the ions. For tungsten, the adsorption isotherms of vanadium and tungsten were polyoxometalate. Both ion-exchange resins were simulated using similar quadratic reaction rate models. Vanadium was desorbed in the aqueous solutions of HCl or NaOH, the desorption characteristics of vanadium and tungsten depended on the desorption solution, and tungsten was desorbed in the aqueous solution of NaOH. It was possible to separate the two ions using the desorption process. The desorption reaction reached equilibrium within 30 min, and more than 90% recovery was possible.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.4
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• pp.345-352
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• 2006

The cemented and paraffin wastes form which are incorporated the concentrated wastes, the cemented waste form which is incorporated the spent ion-exchange resins, and the miscellaneous waste(decontamination paper) were irradiated up to $10^{+8}$ rads at $5.43{\times}10^{+5}$ rads/hr with Co-60(72,023.9 Ci) as an external irradiation source. As a result, the radiolysis gases such as $H_2,\;CH_4,\;N_2,\;C_2H_6,\;O_2,\;CO\;and\;CO_2$, were measured in all the wastes. The major gas which was generated in all the wastes was hydrogen($H_2$). The volume of the generated gases showed a difference from $0.029{\sim}0.788\;cm^3.atm/1.1g$ according to the type of wastes, and more was generated in the cemented waste form incorporated a spent ion-exchange resin than in the other wastes. More hydrogen($H_2$) gas was generated in the decontamination paper waste than in the other wastes, and the G($H_2$) value was 0.12.

• Analytical Science and Technology
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• v.21 no.2
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• pp.143-147
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• 2008

U(IV) ion, the valance state of uranium presumed at in a deep-depth disposal of a spent fuel, was prepared and separated from U(VI) ion. In order to prepare U(IV) ion, tests were performed by adding several reducing agents into a uranyl solution or by dissolution of uranium oxide in a mixed acid added with a reducing agent. The valance states of the uranium in the prepared solutions were identified by separating two ions with a Dowex AG 50W-X8 cation exchange resins and measuring the solutions using a laser-induced fluorescence spectroscopy. However, U(IV) and U(VI) were not separated by a Lichroprep Si60 exchange resin in the same separation condition of Pu(IV) and Pu(VI).