Kee-Chan Song

Effects of the different conditions of uranyl and hydrogen peroxide solutions on the behavior of the uranium peroxide precipitation.

The dynamic precipitation characteristics of UO(4) in different solution conditions (pH, ionic strength, uranium and H(2)O(2) concentrations) were characterized by measuring changes in the absorbance of the precipitation solution and by monitoring the change of particle size in a circulating particle size analyzer. The precipitation solution conditions affected the precipitation characteristics such as the induction time, precipitation rate, overall precipitation time, and particle size in a complex manner. With increases in both pH and ionic strength, the induction time was prolonged, and the individual particle size decreased, but the individual particles tended to grow by aggregation to form larger precipitates. The uranium concentration and the ionic strength of the solution affected the induction time and precipitation rate to the greatest extent.

A Conceptual Process Study for Recovery of Uranium Alone from Spent Nuclear Fuel by Using High-Alkaline Carbonate Media

Abstract This work studied a conceptual process to recover uranium alone from spent nuclear fuel using high-alkaline carbonate media with hydrogen peroxide for the purposes of reducing the volume of high-level active waste and recycling of uranium from the spent fuel with greatly enhanced proliferation resistance, environmental friendliness, and operational safety. The transuranium (TRU) elements were evaluated to be undissolved and precipitated together with other fission products during the oxidative leaching of uranium from the spent fuel. The leaching ratio of uranium dioxide to TRU dioxide from spent fuel in the carbonate solution with H2O2 was estimated to be more than about 108. Only Cs, Tc, Mo, and Te among the major fission products in the spent fuel were dissolved together in the carbonate solution. In the carbonate solution with H2O2, UO2 was dissolved in the form of uranyl peroxo-carbonato complex ions, which could be recovered in the form of uranium peroxide precipitate with a very low solubility by acidification of the solution in a succeeding step. All the inorganic salts of Na2CO3, NaOH, and HNO3 used in the process suggested could be almost completely recovered and recycled into the process again without any generation of secondary wastes.

Fabrication of UO2 Porous Pellets on a Scale of 30 kg-U/Batch at the PRIDE Facility

In the pyroprocess integrated inactive demonstration (PRIDE) facility at the Korea Atomic Energy Research Institute (KAERI), UO2 porous pellets were fabricated as a feed material for electrolytic reduction on an engineering scale of 30 kg-U/batch. To increase the batch size, we designed and modified the corresponding equipment for unit processes based on ceramic processing. In the course of pellet fabrication, the correlation between the green density and sintered density was investigated within a compaction pressure range of 106–206 MPa, in terms of the optimization of processing parameters. Analysis of the microstructures of the produced UO2 porous pellets suggested that the pellets were suitable for feed material in the subsequent electrolytic reduction process in pyroprocessing. This research puts forth modifications to the process and equipment to allow the safe mass production of UO2 porous pellets; we believe these results will have immense practical interest.

ADSORPTION CHARACTERISTICS OF RADIOTOXIC CESIUM AND IODINE FROM LOW-LEVEL LIQUID WASTES

In order to remove the radiotoxic nuclides, Cs+ and I−, from low-level liquid wastes, the adsorption characteristics have been studied using a mixed adsorbent of chabazite zeolite and activated carbon. The equilibrium data of each nuclide were well correlated with the DA equation in the wide range of equilibrium concentrations. The SEM-EDAX analysis provided precise understanding of the adsorption mechanism of each nuclide. A surface diffusion model was applied to estimate the intraparticle mass transfer and provided prediction results acceptable for practical implementation in the liquid waste treatment.

Oxidation State and Extraction of Neptunium with TBP

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 0.5M∼5.5 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).

Preparation of uranium oxide powder for nuclear fuel pellet fabrication with uranium peroxide recovered from uranium oxide scraps by using a carbonate–hydrogen peroxide solution

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.

Extraction and stripping behavior of U-Np-Tc ternary system to TBP

In order to remove U, Tc, and Np, which are positioning materials or target nuclides for transmutation, from the high-level radioactive waste, condition of co-extraction and sequential and sequential stripping of the nuclides wer studied by using 30 vol.% TBP. On the basis of the experiments ferformed 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 exidation 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.

Anodic Dissolution Characteristics of UO2 Electrode in the Potential Ranges over Oxygen Evolution in Carbonate Solutions at Several pHs

This work studied the anodic dissolution characteristics of a UO 2 electrode at several potentials which were much higher than the corrosion potential in carbonate solutions of a high concentration at several pHs. The cyclic voltammograms with several wide potential windows and dissolution rates at several constant applied potentials were measured in the carbonate solutions. The corrosion product at the electrode and the current efficiencies of the electrolytic dissolution were evaluated. The peak occurring after the potential region of the oxygen evolution in the voltammogram of the UO 2 electrode was ascribed to the suppression of the oxygen evolution by a corrosion product of UO 2 CO 3 . The effective dissolution of UO 2 in a carbonate solution could be obtained at an applied potential such as +4 V (vs SSE) or more, which was with an overpotential of oxygen evolution high enough to rupture the corrosion product at the electrode surface, rather than the potential for the oxidation of UO 2 grain as UO 2+ 2 before the oxygen evolution. The corrosion potential of UO 2 decreased with pH in the carbonate solution. And the dissolution rate and current efficiency of UO 2 increased with decrease of pH in the carbonate solution.

Continuous electrolytic decarbonation and recovery of a carbonate salt solution from a metal-contaminated carbonate solution.

This work studied the characteristic changes of a continuous electrolytic decarbonation and recovery of a carbonate salt solution from a metal-contaminated carbonate solution with changes of operational variables in an electrolytic system which consisted of a cell-stacked electrolyzer equipped with a cation exchange membrane and a gas absorber. The system could completely recover the carbonate salt solution from a uranyl carbonato complex solution in a continuous operation. The cathodic feed rate could control the carbonate concentration of the recovered solution and it affected the most transient pH drop phenomenon of a well type within the gas absorber before a steady state was reached, which caused the possibility of a CO(2) gas slip from the gas absorber. The pH drop problem could be overcome by temporarily increasing the OH(-) concentration of the cathodic solution flowing down within the gas absorber only during the time required for a steady state to be obtained in the case without the addition of outside NaOH. An overshooting peak of the carbonate concentration in the recovered solution before a steady state was observed, which was ascribed to the decarbonation of the initial solution filled within the stacked cells by a redundant current leftover from the complete decarbonation of the feeding carbonate solution.

Mitigation of Radiation Load by Trapping of Tritium in Off-gas during Dry-processes

Although tritium gas is produced a small amount comparing to the other fission products, it should be controlled for long half-life, high residence time, high isotopic exchange rate and ease of assimilation into living matter. A demand for lengthening the life time of disposal site needs treatment of spent nuclear fuel to reduce the volume of high level radwaste. This study is for trapping the tritium gas during dry processes carried out in oxygen condition, having a potential of exposure into operators and environment. The experiments were performed into two different hydrogen concentration ranges, i.e. 1000 ~ 9000 ppm of high range of hydrogen concentration and 25 ~ 100 ppm of low range of hydrogen concentration. At high range of hydrogen concentration, H2O conversion ratio at 400°C indicated above 98 % up to 7000 ppm, and 100% at 450°C. All the results of H2O conversion ratio at the low range of hydrogen concentration are represented close to 99%. Converted H2O vapor was adsorbed at the Molecular sieve close to 100%. More than 98% of H2O conversion ratio was attained up to 4 cm/s of linear gas velocity, whereas over 99% for the low range of hydrogen concentration. Some catalytic effect of ability of conversion hydrogen into H2O on stainless steel was studied.