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Abstract  

A new two-step process was investigated to treat LiCl molten salt waste containing volatile radionuclides generated from an electro-metallurgical processing (pyro-processing) of spent oxide fuels. First, the chemical form of the soluble LiCl waste was transformed into a chloride-free and less soluble hydroxide compound by an electrochemical method, where an electrolytic de-chlorination was performed without adding any chemical salt. Then, a gelation process of the chemical form-changed Li compound, named gel-route stabilization/solidification (GRSS) system aimed to reduce the volatility of the radionuclides greatly, was introduced to stabilize/solidify the hydroxide salt wastes. The application of the electrochemical dechlorination/transformation process and the subsequent gel-route stabilization process to treat the soluble LiCl salt wastes was found to be effective.

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Abstract  

In order to remove U, Tc, and Np, which are positioning materials or target nuclides for transmutation, from the high-level radioactive waste, conditions of co-extraction and sequential stripping of the nuclides were studied by using 30 vol.% TBP. On the basis of the experiments performed on each element of U, Tc, and Np, a combination of co-extraction of U, Tc, Np Tc stripping Np stripping U stripping was suggested. To enhance the Np extraction yield, the electrolytic oxidation of Np(V) was required at the co-extraction step. For the stripping of Tc 5M HNO3, of Np the electrolytic reduction of Np(VI) to Np(V), and of U 0.3M sodium carbonate were used. Phase ratios (O/A or A/O) were recommended to be of 2-3, for co-extraction and for stripping.

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Abstract  

A new static contactor was developed for solvent extraction using capillary phenomena induced among clearances formed within a highly packed fiber bundle. Feeding two immiscible phases cocurrently into the fiber bundle generated a very large liquid-liquid contact area for mass transfer within the fiber bundle without any flow turbulence or drop phenomena. In order to test the characteristics and stability of the fiber bundle contactor, continuous extraction experiments were carried out using the fiber bundle contactor with a TBP-uranyl ion-nitric acid system. The fiber bundle contactor had the same extraction performance as that of an ideal batch extractor with good reproducibility due to the sufficient liquidliquid contact area generated by the packed fiber bundle. A minimum residence time of the aqueous phase within the fiber bundle contactor was required for the extraction system to reach an extraction equilibrium state. In the TBP-uranyl ion-nitric acid system, the residence time was about 1.9 minutes. This contactor was confirmed to be effective enough to perform solvent extraction and to study the extraction kinetics because of the stable and large static liquid-liquid contact area.

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Abstract  

Complexes of UO2 2+, Ce3+ and Nd3+ (M) with acetohydroxamic acid (AHA or L) in an aqueous solution have been investigated by the pH-spectral titration method at 25 °C in an aqueous medium of 1.0 M NaClO4 ionic strength. Cerium(III) and neodymium(III) form [ML]2+, [ML2]+, [ML3] complexes with acetohydroxamic acid, while in case of UO2 2+ form [UO2L]+, [UO2L2] complexes with acetohydroxamic acid. Data processing with SQUAD program calculates the best values for the stability constants from pH-spectrophotometric titration data. The protonation constant obtained was pK1 = 9.15 ± 0.04 at 25 °C. The stability constants for acetohydroxamic acid with UO2 2+, Ce3+ and Nd3+ were β1 = 7.22 ± 0.011, β2 = 14.89 ± 0.018 for UO2 2+ and β1 = 5.05 ± 0.062, β2 = 10.60 ± 0.076, β3 = 16.23 ± 0.088 for Ce3+ and β1 = 5.90 ± 0.028, β2 = 12.22 ± 0.038, β3 = 18.58 ± 0.042 for Nd3+, respectively.

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Abstract  

The removal of Cs and Re (as a surrogate for Tc) by selective precipitation from the simulated fission products which were co-dissolved with uranium during the oxidative dissolution of spent fuel in a Na2CO3–H2O2 solution was investigated in this study. The precipitations of Cs and Re were examined by introducing sodium tetraphenylborate (NaTPB) and tetraphenylohosponium chloride (TPPCl), respectively. The precipitation of Cs by NaTPB and that of Re by TPPCl each took place within 5 min, and an increase in temperature up to 50 °C and a stirring speed up to 1000 rpm hardly affected their precipitation rates. The most important factor in the precipitation with NaTPB and TPPCl was found to be a pH of the solution after precipitation. Since Mo tends to co-precipitate with Cs or Re at a lower pH, an effective precipitation with NaTPB and TPPCl was done at pH of above 9 without the co-precipitation of Mo. More than 99% of Cs and Re were precipitated when the initial concentration ratio of NaTPB to Cs was above 1 and when that of TPPCl to Re was above 1. The precipitation of Cs and Re was never affected by the concentration of Na2CO3 and H2O2, even though they were raised up to 1.5 and 1.0 M, respectively. Precipitation yields of Cs and Re in a Na2CO3–H2O2 solution were found to be dependent on the concentration ratios of [NaTBP]/[Cs] and [TPPCl]/[Re].

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Abstract  

The electrochemical redox behavior of nitric acid was studied using a glassy carbon fiber column electrode system, and its reaction mechanism was suggested and confirmed in several ways. Electrochemical reactions in less than 2.0M nitric acid was not observed. However, in more than 2.0M nitric acid, the reduction of nitric acid to nitrous acid occurred and the reduction rate was slow so that the nitric acid solution had to be in contact with an electrode for a period of time long enough for an apparent reduction current of nitric acid to nitrous acid to be observed. The nitrous acid generated in more than 2.0M nitric acid was rapidly and easily reduced to nitric oxide by an autocatalytic reaction. Sulfamic acid was confirmed to be effective to destroy the nitrous acid. At least 0.05M sulfamic acid was necessary to scavenge the nitrous acid generated in 3.5M nitric acid.

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Journal of Radioanalytical and Nuclear Chemistry
Authors: Kwang-Wook Kim, Kee-Chan Song, Eil-Hee Lee, In-Kyu Choi, and Jae-Hyung Yoo

Abstract  

The change of Np oxidation state in nitric acid and the effect of nitrous acid on the oxidation state were analyzed by spectrophotometry, solvent extraction, and electrochemical methods. The Np extraction with 30 vol.% TBP was enhanced by the adjustment of the Np oxidation state using a glassy carbon fiber column electrode system. The knowledge of electrolytic behavior of nitric acid was important because the nitrous acid affecting the Np redox reaction was generated during the adjustment of the Np oxidation state. The Np solution used in this work consisted of Np(V) and Np(VI) but no Np(IV). The ratio of Np(V) in the range of 0.5M5.5 M nitric acid was 32%19%. The electrolytic oxidation of Np(V) to Np(VI) in the solution enhanced the Np extraction efficiency about five times higher than without electrolytic oxidation. It was confirmed that the nitrous acid in a concentration of less than about 10–5 M acted as a catalyst to accelerate the chemical oxidation reaction of Np(V) to Np(VI).

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Journal of Radioanalytical and Nuclear Chemistry
Authors: Dong-Yong Chung, Heui-Seung Seo, Jae-Won Lee, Han-Beom Yang, Eil-Hee Lee, and Kwang-Wook Kim

Abstract  

A feasibility and basic study to find a possibility to develop such a process for recovering U alone from spent fuel by using the methods of an oxidative leaching and a precipitation of U in high alkaline carbonate media was newly suggested with the characteristics of a highly enhanced proliferation-resistance and more environmental friendliness. This study has focused on the examination of an oxidative leaching of uranium from SIMFUEL powders contained 16 elements (U, Ce, Gd, La, Nd, Pr, Sm, Eu, Y, Mo, Pd, Ru, Zr, Ba, Sr, and Te) using a Na2CO3 solution with hydrogen peroxide. U3O8 was dissolved more rapidly than UO2 in a carbonate solution. However, in the presence of H2O2, we can find out that the leaching rates of the reduced SIMFUEL powder are faster than the oxidized SIMFUEL powder. In carbonate solutions with hydrogen peroxide, uranium oxides were dissolved in the form of uranyl peroxo-carbonato complexes. UO2(O2)x(CO3)y 2−2x−2y, where x/y has 1/2, 2/1.

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Abstract  

This work studied a way to reclaim uranium from contaminated UO2 oxide scraps as a sinterable UO2 powder for UO2 fuel pellet fabrication, which included a dissolution of the uranium oxide scraps in a carbonate solution with hydrogen peroxide and a UO4 precipitation step. Dissolution characteristics of reduced and oxidized uranium oxides were evaluated in a carbonate solution with hydrogen peroxide, and the UO4 precipitation were confirmed by acidification of uranyl peroxo–carbonate complex solution. An agglomerated UO4 powder obtained by the dissolution and precipitation of uranium in the carbonate solution could not be pulverized into fine UO2 powder by the OREOX process, because of submicron-sized individual UO4 particles forming the agglomerated UO4 precipitate. The UO2 powder prepared from the UO4 precipitate could meet the UO2 powder specifications for UO2 fuel pellet fabrication by a series of steps such as dehydration of UO4 precipitate, reduction, and milling. The sinterability of the reclaimed UO2 powder for fuel pellet fabrication was improved by adding virgin UO2 powder in the reclaimed UO2 powder. A process to reclaim the contaminated uranium scraps as UO2 fuel powder using a carbonate solution was finally suggested.

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