• Nuclear Engineering and Technology
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• v.5 no.4
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• pp.310-320
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• 1973

The sorption and fixation of cesium in dilute solutions by vermiculite saturated with Na or K were studied in order to investigate any possibibty of its use in radioactive effluent treatment. The cesium sorbed by vermiculite with the increase in pH is attributed to the increase of sorption surface as a result of the dispersion. The increased cesium sorption by Na-vermiculite is due to the different sorption rates by the different exchange sites : external surface and internal surface. It is shown that the larger amount of sorbed cesium was extracted by KCI rather than with any other extractants. It is suggested that the fixation of cesium by vermiculite occurs at the crystal edge where Cs may replace K. Domestic vermiculite is a valuable material for use in the cesium sorption and fixation, and might be useful as a good packing material outside the tank of highly radioactive liquid waste. And from these results one could suggest that the artificial alteration of the biotite to vermiculite might be occurring by treating with NaCl.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.283-288
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• 1997

고 준위 방사성 액체 패기물의 옥살산 침전 공정을 통하여 침전여액이 발생된다. 이 침전여액에는 옥살산과 같은 유기 물질이 함유되어 있으며 이 물질의 제거는 방사성 폐기물의 긍극적인 처리 처분에 대단히 중요하다 본 연구는 모의 침전여액을 제조하고 이로부터 옥살산 제거 연구가 두 가지 방법으로 수행되었다. 첫째 과산화수소를 이용한 제거 법으로서 침전여액 내 Ru$^{+4}$ 와 Fe$^{+3}$ 는 과산화수소를 자체 분해시키기 때문에 옥살산 제거를 방해함을 알 수 있었다 둘째는 NaOH를 용액 내 첨가하는 방법으로 pH증가에 따라 침전이 발생되면서 COD는 상당히 감소함을 보여주었다. 침전물은 $Na_2$C$_2$O$_4$로 확인되었고 pH9 이상에서 99%이상 감소함을 보여 주었다.다.

• Resources Recycling
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• v.7 no.3
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• pp.3-10
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• 1998

Simulated waste liquid containing 50 ppm cobalt ion was treated by adsorbing colloidal flotation using Fe(III) or Al(IlI) as flocclant and a sodium lamyl sulfate as a collector. Parameters such as pH, surfactant concentration, Fe(III) or Al(III) concentration, gas flow rate, etc., W앙e considered. The flotation with Fe(III) showed 99.8% removal efficiency of cohalt on the conditions of initial cobalt ion concentration 50 ppm, pH 9.5, gas flow rate 70 ml/min, and flotation time 30 min. When the waste solution, was treated with 35% $H_2O_2$ prior to adsorbing colloidal flotation, the optimal pH for removing cobalt shifted m to weak alkaline range and flotation could be applied in wider range of pH as compared to non-use of $H_2O_2$. Additional use of 20 ppm Al(III) after precipitation of 50 ppm Co(II) with 50 ppm Fe(III) made the optimal pH range for preferable flotation w wider. Foreign ions such as, $NO_3^-$, $SO_4^{2-}$, $Na^+$, $Ca^{2+}$ were adopted and their effects were observed. Of which sulfate ion was f found to be detrimental to removal of cob퍼t ion by flotation. Coprecipitation of Co ion with Fe(III) and Al(III) resulted in b better removal efficiency of cobalt IOn 피 the presence of sulfate ion.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.1
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• pp.1-6
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• 2014

A large amount of acidic waste solution is generated from the practical electrokinetic decontamination equipments for the remediation of soil contaminated with uranium. After filtration of uranium hydroxides formed by adding CaO into the waste solution, the filtrate was recycled in order to reduce the volume of waste solution. However, when the filtrate was used in an electrokinetic equipment, the low permeability of the filtrate from anode cell to cathode cell due to a high concentration of calcium made several problems such as the weakening of a fabric tamis, the corrosion of electric wire and the adhension of metallic oxides to the surface of cathode electrode. To solve these problems, sulfuric acid was added into the filtrate and calcium in the solution was removed as $CaSO_4$ precipitate. A decontamination test using a small electrokinetic equipment for 20 days indicated that Ca-removed waste solution decreased uranium concentration of the waste soil to 0.35 Bq/g, which is a similar to a decontamination result obtained by distilled water.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.139-165
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• 2019

The current status and prospect of decommissioning technology development at KAERI are reviewed here. Specifically, this review focuses on four key technologies: decontamination, remote dismantling, decommissioning waste treatments, and site remediation. The decontamination technologies described are component decontamination and system decontamination. A cutting method and a remote handling method together with a decommissioning simulation are described as remote dismantling technologies. Although there are various types of radioactive waste generated by decommissioning activities, this review focuses on the major types of waste, such as metal waste, concrete waste, and soil waste together with certain special types, such as high-level and high-salt liquid waste, organic mixed waste, and uranium complex waste, which are known to be difficult to treat. Finally, in a site remediation technology review, a measurement and safety evaluation related to site reuse and a site remediation technique are described.

• Resources Recycling
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• v.7 no.3
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• pp.11-16
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• 1998

Simulated waste liquid containing 50 ppm cobalt ion was t$\xi$sted by precipitate flotation using a sodium lauryl sulfate as a c collector. The effects of initial cobalt ion concentration, pH, surfactant concentration, flotation time, gas flow rate and foreign i ions on removal efficiency of cobalt ion were studied. Pretreatment of the waste liquid with 35% $H_2O_2$, prior to precipitate f flotation made shin of optimal flotation pH from the strong alkalinity to weak alkaline range and made a favorable flotation of c cobalt ion in wide range of pH. For the result of this experiment, 99.8% removal efficiency was obtained on the conditions of initial coball ion concentration 50 ppm, pH 9.5 gas flow rate 70 mllmin, flotation time 30 min. The simulate ion was fanned t to be the most harmful ion against removal of cobalt by precipitate flotation of the species which were tested The presence of 0.1 M of $SO_4^{2-}$ ion decreased remo,때 $\xi$폐iciency of cobalt to 90% while the cobalt were almost entirely removed in the a absence of sulfate ion.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.203-212
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• 2019

Disposal nonconformity of radioactive wastes refers to radioactive wastes that need to be treated, solidified and packaged during operation or decommissioning of NPPs, and are typically exemplified by particulate radioactive wastes with dispersion characteristics. These wastes include the dried powders of concentrated wastes generated in the process of operating NPPs, slurry and sludge, various powdered wastes generated in the decommissioning process (crushed concrete, decontamination sludge, etc.), and fine radioactive soil, which is not easy to decontaminate. As these particulate wastes must be packaged so that they become non-dispersive, they are solidified with solidification agents such as cement and polymer. If they are treated using existing solidification methods, however, the volume of the final wastes will increase. This drawback may increase the disposal cost and reduce the acceptability of disposal sites. Accordingly, to solve these problems, this study investigates the pelletization of particulate radioactive wastes in order to reduce final waste volume.

• Nuclear Engineering and Technology
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• v.9 no.4
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• pp.211-222
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• 1977

The decomposition of all the individual chemicals used in the Harwell inactive vitrification pilot plant has been studied by means of a thermal balance. Weight loss curves to 110$0^{\circ}C$ have been obtained. The four materials (sodium nitrate, cesium nitrate, lithium nitrate and ruthenium nitroso-nitrate solution) showed a greater weight loss than that based on an oxide yield, and hence these compounds of their products of decomposition are volatile below 110$0^{\circ}C$. The remaining materials suffered a weight loss no more than that corresponding to a full yield of the oxide, and hence they were not volatile below 110$0^{\circ}C$. Most of chemicals begin to decompose at less than 75$^{\circ}C$ but the nitrates of cesium, strontium, barium and sodium not until 295$^{\circ}C$ to 59$0^{\circ}C$. The results obtained can be used in the analysis of process conditions in the vitrification and calcination of highly radioactive wastes and also of the thermal decomposition behaviour of mixtures containing those materials.

• Journal of Energy Engineering
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• v.3 no.1
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• pp.70-76
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• 1994

본 연구는 방사성 폐액의 처리를 목적으로 화학처리와 한외여과막(UF)의 결합공정에 의해서 세슘이온의 제거특성을 조사하였다. 이 공정은 대상핵종과 선택성이 크고 한외여과막에 의해서 분리가 가능한 거대분자를 주입하여 핵종을 결합시키고, 이를 한외여과막에 의해서 분리 제거하는 개념이다. 실험은 흡착제로서 K2Cu3(Fe(CN6)2)를 제조하여 주입하였고 회분식 UF stirred cell를 이용하였으며, 용액의 pH, 세슘이온의 농도 및 K2Cu3(Fe(CN6)2)의 농도에 따라 세슘이온의 제거효율을 측정하였다. 세슘의 제거효율은 pH 및 K2Cu3(Fe(CN6)2)의 몰비에 따라 결정되며, pH가 5∼6에서 높은 제거율을 나타내었고 Cs/K2Cu3(Fe(CN6)2)의 몰비가 1.5 이하에서 90% 이상 제거되었다. K2Cu3(Fe(CN6)2)에 대한 Cs의 결합특성은 Langmuir isotherm형태의 식으로 나타내어 평가하였으며, 이때 세슘이온의 최대 흡착용량은 1.72 mM/mM K2Cu3(Fe(CN6)2) 이었다.

• Economic and Environmental Geology
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• v.51 no.2
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• pp.177-183
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• 2018

This study assessed an application of electrocoagulation (EC) for the removal of cobalt (Co) in radioactive liquid waste from nuclear power plant. The EC process is an electrochemical means to remove a contaminant in wastewater and a novel process to complement the disadvantage of chemical treatment and membrane process. Radioactive liquid waste has been produced from washing process of radio nuclide power plant cooling system. The EC process eliminates Co from the electrolyte within 10 min; in addition, the dewatered sludge produced in EC process is only 0.2 g. Therefore, the EC process is a promising technique for the removal of Co in radioactive liquid waste and volume reduction of wastes.