• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.4
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• pp.315-319
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• 2015

A high-throughput electrorefiner has been developed for commercialization use by enhancing the uranium recovery from the reduced metal which is produced from the oxide reduction process. It is necessary to scrap and effectively collect uranium dendrites from the surface of the solid cathode for high yield. When a steel electrode is used as the cathode in the electrorefining process, uranium is deposited and regularly stuck to the steel cathode during electrorefining. The sticking coefficient of a steel cathode is very high. In order to decrease the sticking coefficient of the steel cathode effectively, vibration mode was applied to the electrode in this study. Uranium dendrites were scraped and fell apart from the steel cathode by a vibration force. The vibrational scraping of the steel cathode was compared to the self-scraping of the graphite cathode. Effects of the applied current density and the vibration stroke on the scraping of the uranium dendrites were also investigated.

• Korean Journal of Materials Research
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• v.8 no.12
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• pp.1099-1109
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• 1998

To investigate the effect of V and Sb on the corrosion behavior of Zr- based alloys, corrosion tests were performed on 6 kinds of Zr alloys in an autoclave at $360^{\circ}C$ for 100 days. The transition of the corrosion rate occurred in the sample containing 0.1wt.%V after 10 days but did not occur in the samples containing 0.2wt.%V and 0.4wt.%V. The corrosion resistance of V containing alloys increased with increasing V contents from 0.1 to 0.4wt.% and the alloys containing 0.4wt.%V showed the best corrosion resistance. In the ternary alloys containing 0.1wt.%Sb and 0.4wt.%Sb, the corrosion rate increased significantly from the short exposure time. It was observed that the optimal Sb content for corrosion resistance was 0.2wt.%. The size and volume fraction of precipitates increased with increasing V and Sb contents. The superior corrosion resistance was observed in the Zr alloy having precipitate size of 0.11-0.13$\mu\textrm{m}$. From the result of corrosion behavior and the obserbation of precipitates, the optimal size of the precipitate appear to control the electron conduction in the cathodic reaction and play an important role in maintaining a stable oxide microstructure.

• Journal of the Korean Magnetics Society
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• v.22 no.2
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• pp.37-41
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• 2012

The Ag nanoparticles attached $La_{0.7}Sr_{0.3}Co_{0.3}Fe_{0.7}O_{3-{\delta}}$ (LSCF) perovskites were prepared by plasma method. The Ag nanoparticles with size of several nanometers deposited from the Ag target were coated on the surface of LSCF powders with size range from 0.2 to 3 ${\mu}m$. The agglomeration of Ag particles annealed at $800^{\circ}C$ under inert gas of Ar were rarely observed. The inter-diffusion between surface Ag and core LSCF is effectively strong to prevent aggregation of Ag nanoparticles. The wave number of FT-IR spectra for LSCF were largely shifted as the concentration of Ag on LSCF up to 2.11 wt.%. The ionic states of irons in LSCF were measured by M$\ddot{o}$ssbauer spectroscopy. The small amount of $Fe^{4+}$ ions are converted to $Fe^{3+}$ ions after Ag nanopartcles were coated on LSCF.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.3
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• pp.235-243
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• 2014

Surface oxidation behavior of ZIRLO (ZIRconium Low Oxidation) hulls was investigated using an X-ray photoelectron spectroscopy (XPS) technique. The effects of oxidation time (10-336 h at $500^{\circ}C$) and temperature ($400-700^{\circ}C$ for 10 h) were studied. Deconvolution results of the hulls oxidized at $500^{\circ}C$ revealed that a $ZrO_2$ phase appeared after 24 h (11.86%), and an increase in the $ZrO_2$ ratio was observed when the hulls were oxidized for 336 h (17.93%). On the other hand, the ZrO phase which employed 5.68% in the 10 h oxidized sample disappeared when the oxidation time increased to 24 h. The XPS results also showed that an increase in the oxidation temperature resulted in an increase in the ratio of ZrO, which increased from 0 to 5.68, 8.31, and 9.16% when the oxidation temperature increased from 400 to 500, 600, and $700^{\circ}C$, respectively. $ZrO_2$ phase was observed only in the sample that was oxidized at $700^{\circ}C$. The mechanism of ZrO formation was not conclusive, but it was suggested that a formation of hydroxide might have been accelerated at elevated temperatures leading to a formation of a $Zr(OH)_4$ phase. The relationship between the surface oxidation status of the hulls oxidized at $500^{\circ}C$ and their chlorination reaction feasibility was discussed, and it was suggested that the thickness of the oxide layer is an important parameter that determines the chlorination reaction feasibility.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.33-40
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• 2006

The characteristics of the aluminum waste melting and the distribution of the radioactive nuclides have been investigated for the estimation on the volume reduction and the decontamination of the aluminum wastes from the decommissioning of the TRIGA MARK it and III research reactors at the Korea Atomic Energy Research Institute(KAERI). The aluminum wastes were melted with the use of the fluxes such as flux $A:NaCl-KCl-Na_3AlF_6$, flux B:NaCl-NaF-KF, flux $C:CaF_2$, and flux $D:LiF-KCl-BaCl_2$ in the DC graphite arc furnace. For the assessment of the distribution of the radioactive nuclides during the melting of the aluminum, the aluminum materials were contaminated by the surrogate nuclides such as cobalt(Co), cesium(Cs) and strontium(Sr). The fluidity of aluminum melt was increased with the addition of the fluxes, which has slight difference according to the type of fluxes. The formation of the slag during the aluminum melting added the flux type C and D was larger than that with the flux A and B. The rate of the slag formation linearly increased with increasing the flux concentration. The results of the XRD analysis showed that the surrogate nuclide was transferred to the slag, which can be easily separated from the melt and then they combined with aluminum oxide to form a more stable compound. The distribution ratio of cobalt in ingot to that in slag was more than 40% at all types of fluxes. Since vapor pressures of cesium and strontium were higher than those that of the host metals at the melting temperature, their removal efficiency from the ingot phase to the slag and the dust phase was by up to 98%.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.27-36
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• 2012

Metal chloride waste is generated as a main waste streams in a series of electrolytic processes of a pyrochemical process. Different from carbonate or nitrate salt, metal chloride is not decomposed into oxide and chlorine but it is just vaporized. Also, it has low compatibility with conventional silicate glasses. Our research group adapted the dechlorination approach for the immobilization of waste salt. In this study, the composition of SAP ($SiO_2-Al_2O_3-P_2O_5$) was adjusted to enhance the reactivity and to simplify the solidification process as a subsequent research. The addition of $Fe_2O_3$ into the basic SAP decreased the SAP/Salt ratio in weight from 3 for SAP 1071 to 2.25 for M-SAP( Fe=0.1). The experimental results indicated that the addition of $Fe_2O_3$ increased the reactivity of M-SAP with LiCl-KCl but the reactivity gradually decreased above Fe=0.1. Also, introducing $B_2O_3$ into M-SAP requires no glass binder for the consolidation of reaction products. U-SAP ($SiO_2-Al_2O_3-Fe_2O_3-P_2O_5-B_2O_3$) could effectively dechlorinate the LiCl-KCl waste and its reaction product could be consolidated as a monolithic form without a glass binder. The leaching test result indicated that U-SAP 1071 was more durable than other SAPs wasteform. By using U-SAP, 1 g of waste salt could generated 3~4 g of wasteform for final disposal. The final volume would be about 3~4 times lower than the glass-bonded sodalite. From these results, it could be concluded that the dechlorination approach using U-SAP would be one of prospective methods to manage the volatile waste salt.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.93-100
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• 2009

This study has been carried out to look into the characteristics of an oxidative-dissolution of fission products (FP) co-dissolved with uranium (U) in a $(NH_4)_2CO_3$ carbonate solution. Simulated FP-oxides which contained 12 components have been added to the solution to examine their dissolution characteristics. It is found that $H_2O_2$ is an effective oxidant to minimize the oxidative-dissolution of FP. In the 0.5 M $(NH_4)_2CO_3$-0.5 M $H_2O_2$ solution, some elements such as Re, Te, Cs and Mo seem to be dissolved together with U, while 98${\pm}$2% for Re and Te, 94${\pm}$2% for Cs, and 29${\pm}$2 % for Mo are dissolved for 2 hours. It is revealed that dissolution rates of Re, Te and Cs are high (completely dissolved within 10${\sim}$20 minutes) due to their high solubility in the $(NH_4)_2CO_3$ solution regardless of the addition of $H_2O_2$, and independent of the concentrations of $Na_2CO_3$ and $H_2O_2$. However, the dissolution ratio of Mo seems to be slightly increased with time and about 33 % for 4 hours, indicating a very slow dissolution rate and also independent of the $(NH_4)_2CO_3$ concentration. It is found that the most important factor for the oxidative-dissolution of FP is the pH of the solution and an effective dissolution is achieved at a pH between 9${\sim}$10 in order to minimize the dissolution of FP.