• 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 $60^{\circ}C$.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.202-206
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• 2003

In this study, a modified bituminization technology has been developed, which needs no grinding of the granular resin waste, and enables the solid form to keep its shape stability as good as that of a cemented solid from Also, the study intended to apply the developed technology to the practical treatment of radioactive resin waste. In the experiment, the granular type resin was used and the straight-run distillation bitumen with penetration rate 60/70 was used as the solidifying agent. The PE was used as the additive. The shape stability increased remarkably with the additive of PE, which act as a binder in the solid form. The shape of the solid form was maintained without failure during the long-term exposure test when the additive content of spent PE is more than 10wt%. The proper ranges of bitumen content, PE content and operating temperature are 30-50wt%, 10-20wt% and $180^{\circ}C$ respectively. The bituminized solid form of radioactive resin waste by the technology of this study has the remarkably superior quality than the conventional solid forms, partially for the shape stability.

• Mycobiology
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• v.44 no.4
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• pp.338-342
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• 2016

The culture filtrate of Lentinula edodes shows potent antimicrobial activity against the plant pathogenic bacteria Ralstonia solanacearum. Bioassay-guided fractionation was conducted using Diaion HP-20 column chromatography, and the insoluble active compound was not adsorbed on the resin. Further fractionation by high-performance liquid chromatography (HPLC) suggested that the active compounds were organic acids. Nine organic acids were detected in the culture filtrate of L. edodes; oxalic acid was the major component and exhibited antibacterial activity against nine different phytopathogenic bacteria. Quantitative analysis by HPLC revealed that the content of oxalic acid was higher in the water extract from spent mushroom substrate than in liquid culture. This suggests that the water extract of spent L. edodes substrate is an eco-friendly control agent for plant diseases.

• Analytical Science and Technology
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• v.13 no.5
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• pp.584-591
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• 2000

The separation of Nd from the simulated $U_3Si/Al$ spent fuel solution with sequential two-step anion exchange separation has been studied. To prepare the simulated $U_3Si/Al$ spent nuclear fuel, unirradiated $U_3Si/Al$ whose composition consists of small $U_3Si$ particle dispersed in an Al matrix with Al cladding was dissolved with a mixture of 4 M HCl and 10 M $HNO_3$ and 8 or 15 fission product elements were added to the dissolved solution. The trace amount of silica in the solutions was removed by evaporating to dryness with HF and the U was adsorbed on the first anion exchange resin. Neodymium can be purely isolated from the fission product elements with a methanol-nitric acid eluent using the second anion exchange resin. A large excess of Al didn't influence on the elution velocity of Nd, but reduced the eluted contents of Nd, Al, Eu, Gd, Sm and Sr, A large amount of Al was removed first from the column with 3 mL of loading solution (0.8 M $HNO_3$/99.8% MeOH) before Nd elution by the eluent [0.04 M $HNO_3$-99.8% MeOH(1:9)]. The recovery of Nd was more than 94%, regardless of Al contents. Taking the 9 to 13 mL fraction of eluate was effective to purely isolate Nd.

• Analytical Science and Technology
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• v.13 no.4
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• pp.446-452
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• 2000

The simulated nuclear spent fuel (SIMFUEL) containing the platinum group elements which will not be dissolved in a nitric acid was completely dissolved with a acid digestion bomb. The metallic elements separated in the SIMFUEL were measured by inductively coupled plasma atomic emission spectrometry (ICP-AES). Because the peaks of metallic elements were spectrally interfered by uranium spectrum, uranium and metallic elements were separated by cation exchange resin for Mo, Pd, Rh and Ru and by anion exchange resin for Ba, Ce, La, Nd, Rh, Sr, Y and Zr, respectively. The recovery of Mo, Pd, Rh and Ru after separation by cation exchange chromatography found to be 99-103% and anion exchange separation showed 96.5-107% of recovery except Y with the simulated solution whose concentration was similar to the spent nuclear fuel. The relative standard deviation of this method showed 1.3-6.7% in the SIMFUEL whose concentrations of metallic elements were between several $10^2-10^3$ppm.

• Journal of the Korean Wood Science and Technology
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• v.51 no.1
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• pp.23-37
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• 2023

One type of biomass that has promising potential for bio pellet production is spent coffee grounds (SCGs). However, previous studies have shown that SCGs in bio pellets cause a lot of smoke. Therefore, they need to be mixed with a material that has a higher calorific value to produce better quality pellets. One material that can be used is pine wood because it has a natural resin content that can increase the calorific value. The aim of this study was to examine the quality of bio pellets produced with SCGs and pine wood charcoal at different particle sizes. The charcoal was ground using either a hammer mill (HM) or a ball mill (BM). Pine wood charcoal was mixed with SCGs at ratios of SCGs to pine wood charcoal of 4:6 and 6:4 by weight, respectively, and the adhesive used a tapioca with a composition ratio 5% of the raw material. The bio pellets were produced using a manual pellet press. The quality of the bio pellets was assessed based on Indonesian National Standard (SNI) 8021-2014, and the physical observations include flame length, burning rate, and compressive strength. The average water content, ash content, and calorific value of the bio pellets were in accordance with SNI 8021-2014, but the density and ash content values were below the standard values. The BM variation of bio pellets had a higher compressive strength than the HM variation, and the 4:6 BM variation had the longest burning time compared with 4:6 HM.

• Journal of Radiation Protection and Research
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• v.13 no.2
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• pp.52-56
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• 1988

An ion-exchange method for the detection of radioactivity in water using ion-exchange resin in concentrating radioactive nuclides was compared with an evaporation method. The loss of the radioactive materials in the sample treated by the ion-exchange method was less by about 20% than that by the evaporation method. In addition, the evaporation method needed about 20 hours for evaporating one liter of the sample at $70^{\circ}C$, while the ion-exchange method spent 6 hours to adsorb and adsorb the same amount of the sample on the resin. Consequently, the ion-exchange method is more effective than the evaporation method for the treatment of the radioactively contaminated water and is especially suitable for detecting the low-level radioactivity in water.

• Analytical Science and Technology
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• v.16 no.6
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• pp.450-457
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• 2003

The determination of uranium and its isotopes in spent nuclear fuels by isotope dilution mass spectrometry (IDMS) has been studied. The spent fuel samples were dissolved in 8 M $HNO_3$ or its mixture with 14 M $HNO_3-0.05M$ HF. The dissolved solutions were filterred on membrane filter with $1.2{\mu}m$ pore size. The uraniums in the spiked and unspiked sample solutions were quantitatively adsorbed by anion exchange resin, AG 1X8 and eluted with 0.1 M HCl. The contents of uranium and its isotopes ($^{234}U$, $^{235}U$, $^{236}U$ 및 $^{238}U$) in the spent fuel samples were determined by isotope dilution mass spectrometric method using $^{233}U$ as spike. The spike reference solution was standarized by reverse isotope dilution mass spectrometry (R-IDMS) using natural and depleted uranium. The results from IDMS were in average relative difference of 0.34% when compared with those by the potentiometric titration method.

• Resources Recycling
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• v.22 no.1
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• pp.55-63
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• 2013

Considering considerable contents of vanadium and tungsten in spent SCR DeNOx catalysts, separation and recovery of those metals are required. In this respect, commercial anion exchange resin (MP600) was employed to recover vanadium from the synthetic solution of ammonium metavanadate. Experimental results indicated that vanadium exist as anion under the acidic condition (pH 2 ~ 6) and adsorbed on the resin. Although the adsorption rate was increased with temperature, the maximum amount of adsorption was not affected by temperature. Desorption took place under either strong acidic (less than pH 1) or strong caustic (higher than pH 13) condition. However, desorption seldom took place under moderate conditions (pH 3~11). Furthermore, adsorption equilibrium results agreed well with Freundlich isotherm and pseudo-second-order reactions. And, adsorption energy was evaluated using Dubinin-Radushkevich and Temkin isotherm.

• Applied Chemistry for Engineering
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• v.8 no.1
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• pp.92-98
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• 1997

Effects of radiation dose on mechanical properties such as tensile strength, compressive strength, flexural strength, specific gravity and changes of weight and hydrogen content of epoxy resin-type neutron shielding materials to be used for spent fuel shipping casks have been investigated. At radiation dose up to 0.5MGy, the tensile strength, compressive strength and flexural strength of the shielding materials of KNS-115A, KNS-115B and KNS-115C have been increased with increase in the radiation dose. In contract, these mechanical properties have been decreased at radiation dose above 0.5MGy. The amount of radiation dose on the materials of KNS-115A, KNS-115B and KNS-115C has not resulted in a measurable loss of specific gravity and weight of them, whereas the reduction of hydrogen content has been observed.