• Journal of Electrochemical Science and Technology
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• v.10 no.2
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• pp.123-130
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• 2019

We investigated the electrochemical behavior of an electrode coated with a nickel hexacyanoferrate/graphene oxide (NiPB/GO) composite to evaluate its potential use for the electrochemical separation of radioactive Cs as a promising approach for reducing secondary Cs waste after decontamination. The NiPB/GO-modified electrode showed electrochemically switched ion exchange capability with excellent selectivity for Cs over other alkali metals. Furthermore, the repetitive ion insertion and desertion test for assessing the electrode stability showed that the electrochemical ion exchange capacity of the NiPB/GO-modified electrode increased further with potential cycling in 1 M of $NaNO_3$. In particular, this electrochemical treatment enhanced Cs uptake by nearly two times compared to that of NiPB/GO and still retained the ion selectivity of NiPB, suggesting that the electrochemically treated NiPB/GO composite shows promise for nuclear wastewater treatment.

• 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.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1957-1963
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• 2019

A UVC photo-Fenton advanced oxidation process (AOP) was studied to develop a process for the decomposition of oxalic acid waste generated in the chemical decontamination of nuclear power plants. The oxalate decomposition behavior was investigated by using a UVC photo-Fenton reactor system with a recirculation tank. The effects of the three operational variables-UVC irradiation, H₂O₂ and Fenton reagent-on the oxalate decomposition behavior were experimentally studied, and the behavior of the decomposition product, CO₂, was observed. UVC irradiation of oxalate resulted in vigorous CO₂ bubbling, and the irradiation dose was thought to be a rate-determining variable. Based on the above results, the oxalate decomposition kinetics were investigated from the viewpoint of radical formation, propagation, and termination reactions. The proposed UVC irradiation density model, expressed by the first-order reaction of oxalate with the same amount of H₂O₂ consumption, satisfactorily predicted the oxalate decomposition behavior, irrespective of the circulate rate in the reactor system within the experimental range.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.655-659
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• 2003

The research reactor (TRIGA Mark-III(KRR-2)) was constructed and had been operated in 1972. In 1999 the radioisotope process units had stopped its operation due to normal operation of HANARO. In 2003 the core assembly was decommissioned by D&D program. The contact exposure rate on the core assembly and the rotary specimen rack are from 300mSv/h to 700mSv/h. This report describes the decontaminationing procedures, the health physics programs, and the waste management.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2015.10a
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• pp.465-466
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• 2015

In this study, different amount of (fe(0)) were dissolve into different strength of phosphoric ($H_3PO_4$) acid and the optimum solubility was observed at 0.89M Fe(0) into 4M of $H_3PO_4$ acid. Different concentration of oxalic acid was added to determine the optimum precipitated condition. The dissolution kinetics of Fe(0) into $H_3PO_4$ acid was investigated at $40-50^{\circ}C$. The optimum Fe-oxalate precipitate was dried and thermal decomposition using DSC-TG was conducted. Approximately 52 wt(%) of oxalic acid was removed at $300^{\circ}C$. Iron oxides such as magnetite and hematite that may be formed on the surface of nuclear waste were also dissolved into the $H_3PO_4$ acid and the optimum solubility for magnetite is 0.005M while that for hematite is 0.02M in 8M $H_3PO_4$ acid, respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.305-320
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• 2016

The metal waste generated during nuclear power plant decommissioning constitutes a large proportion of the total radioactive waste. This study investigates the current status of domestic and international regulatory requirements for clearance and the clearance experience of domestic institutions. The RESRAD-RECYCLE code was used for analyzing the clearance of the metal wastes generated during actual nuclear power plant decommissioning, and assessment of the exposure dose of twenty-six scenarios was carried out. The evaluation results will be useful in preliminary analysis of clearance and recycling during nuclear power plant decommissioning. As a next step, the effects of reducing disposal costs by clearance can be studied.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.1 no.1
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• pp.11-23
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• 2003

A study on the electrolytic dissolution of SUS-304 and Inconel-600 specimen was carried out in neutral salt electrolyte to evaluate the applicability of electrochemical decontamination process for recycle or self disposal with authorization of large amount of metallic wastes contaminated with uranium compounds generated by dismantling a retired uranium conversion plant in Korea. Although the best electrolytic dissolution performance for the specimens was observed in a Na2s04 electrolyte, a NaNO$_3$ neutral salt electrolyte, in which about 30% for SUS-304 and the same for Inconel-600 in the weight loss was shown in comparison with that in a Na$_2$SO$_4$ solution, was selected as an electrolyte for the electrochemical decontamination of metallic wastes with the consideration on the surface of system components contacted with nitric acid and the compatibility with lagoon wastes generated during the facility operation. The effects of current density, electrolytic dissolution time, and concentration of NaNO$_3$ on the electrolytic dissolution of the specimens were investigated. On the basis of the results obtained through the basic inactive experiments, electrochemical decontamination tests using the specimens contaminated with uranium compounds such as UO$_2$, AUC (ammonium uranyl carbonate) and ADU (ammonium diuranate) taken from an uranium conversion facility were performed in 1M NaNO$_3$ solution with the current density or In mA/$\textrm{cm}^2$. it was verified that the electrochemical decontamination of the metallic wastes contaminated uranium compounds was quite successful in a NaNO$_3$ neutral salt electrolyte by reducing $\alpha$ and $\beta$ radioactivities below the level of self disposal within 10 minutes regardless of the type of contaminants and the degree of contamination.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.21-29
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• 2005

The HANARO, a multi-purpose research reactor of 30 MWth open-tank-in-pool type, has been under normal operation since its initial criticality in February, 1995. Recently, ten years after the initial operation of the HANARO, one of the two primary cooling pumps was decontaminated for overhaul maintenance in 2004. Before decontamination exposure doserate and surface contamination level of primary cooling pump measured at 4 points. After final decontamination exposure doserate and surface contamination level of primary cooling pump remeasured by same method done before. It is easy to decontaminate the out side exposed surfaces of the pump, but it is difficult to approach the inside surface due to double volute installed in the casing. Therefore, a new decontamination facility has been developed to solve this problem. A concentrated de-contaminant (DX-300) is rotated in the closed pump casing by the impeller actuated by a temporary motor. Nuclide particles are removed by the emulsification effect of the de-contaminant and the surface contaminants are chemically removed from the pump by the corrosion and dissolution effect. The inside surfaces of the primary cooling pump have been decontaminated by using the facility. As results, the contamination level of the inside surfaces was maintained below the surface contamination limit.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.421-429
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• 2018

Hydrazine based reductive dissolution applied on magnetite oxide was investigated. Dissolution of Fe(II) and Fe(III) from magnetite takes place either by protonation, surface complexation, or reduction. Solution containing hydrazine and sulfuric acid provides hydrogen to break bonds between Fe and oxygen by protonation and electrons for the reduction of insoluble Fe(III) to soluble Fe(II) in acidic solution of pH 3. In terms of dissolution rate, numerous transition metal ions were examined and Cu(II) ion was found to be the most effective to speed up the dissolution. During the cycle of Cu(I) ions to Cu(II) ions, the released electron promoted the reduction of Fe(III) and Cu(II) ions returned to Cu(I) ion due to the oxidation of hydrazine. In the experimental results, the addition of a very low amount of cupric ion (about 0.5 mM) to the solution increased the dissolution rate about 40% on average and up to 70% for certain specific conditions. It is confirmed that even though the coordination structure of copper ions with hydrazine is not clear, the $Cu(II)/H^+/N_2H_4$ system is acceptable regarding the dissolution performance as a decontamination reagent.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.516-522
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• 2023

The co-precipitation process plays a key role in the decontamination of radionuclides from low and intermediate levels of liquid waste. For that reason, the removal of Cs ions from waste solution by the co-precipitation method was carried out. A simulated liquid waste (¹³³Cs) was prepared from a 0.1 M CsCl solution, while wastewater generated by washing steel ash served as a representative of radioactive cesium solution (¹³⁷Cs). By co-precipitation, potassium ferrocyanide was applied for the adsorption of Cs ions, while nickel nitrate and iron sulfate were selected for supporting the precipitation. The amount of residual Cs ions in the CsCl solution after precipitation and filtration was determined by ICP-OES, while the radioactivity of ¹³⁷Cs was measured using a gamma-ray spectrometer. After cesium removal, the amount of cesium appearing in both XRD and SEM-EDS was analyzed. The removal efficiency of ¹³³Cs was 60.21% and 51.86% for nickel nitrate and iron sulfate, respectively. For the ash-washing solution, the removal efficiency of ¹³⁷Cs was revealed to be more than 99.91% by both chemical agents. This implied that the co-precipitation process is an excellent strategy for the effective removal of radioactive cesium in waste solution treatment.