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

• Economic and Environmental Geology
• /
• v.41 no.6
• /
• pp.719-729
• /
• 2008

In this study, the speciation, adsorption, and transport of uranium in groundwater environments were simulated using geochemical models. The retarded transport of uranium by adsortption was effectively simulated using 3-D groundwater flow and reactive transport models. The results showed that most uranium was adsorbed(up to 99.5%) in a neutral pH(5.5$pCO_2(10^{-3.6}atm)$ condition. Under the higher $pCO_2(10^{-2.5}atm)$ condition, however, the pH range where most uranium was absorbed was narrow from 6 to 7. Under very low $pCO_2(10^{-4.5}atm)$ condition, uranium was mostly absorbed in the relatively wide pH range between 5.5 and 8.5. In the model including anion complexes, the uranium adsorption decreased by fluoride complex below the pH of 6. The results of this study showed that uranium transport is strongly affected by hydrochemical conditions such as pH, $pCO_2$, and the kinds and concentrations of anions($Cl^-$, ${SO_4}^{2-}$, $F^-$). Therefore, geochemical models should be used as an important tool to predict the environmental impacts of uranium and other hazardous compounds in many site investigations.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2004.06a
• /
• pp.72-80
• /
• 2004

Electrolytic dissolution study was carried out to evaluate the applicability of electrochemical decontamination process using a neutral salt electrolyte as a decontamination technology for the recycle or self disposal with authorization of large amount of metallic wastes contaminated with uranium compounds generated by dismantling a retired uranium conversion plant using SUS-304 and Inconel-600 specimen as the main materials of internal system components of the plant. The effects of type of neutral salt as an electrolyte, current density, and concentration of electrolyte on the dissolution of the materials were evaluated. 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 plant were peformed in $Na_2SO_4$ and $NaNO_3$ solution. It was verified that the electrochemical decontamination of the dismantled metallic wastes was quite successful in $Na_2SO_4$ and $NaNO_3$ neutral salt electrolyte by reducing $\beta$ radioactivities below the level of self disposal with authorization within 10 minutes regardless of the type of contaminants and the degree of contamination.

• Nuclear Engineering and Technology
• /
• v.51 no.1
• /
• pp.214-220
• /
• 2019

Thermochemical cycles have been predominantly used for energy transformation from heat to stored chemical free energy in the form of hydrogen. The thermochemical cycle based on uranium (UTC), proposed by Oak Ridge National Laboratory, has been considered as a better alternative compared to other thermochemical cycles mainly due to its safety and high efficiency. UTC process includes three steps, in which only the first step is unique. Hydrogen production apparatus with hectogram reactants was designed in this study. The results showed that high yield hydrogen was obtained, which was determined by drainage method. The results also indicated that the chemical conversion rate of hydrogen production was in direct proportion to the mass of $Na_2CO_3$, while the solid product was $Na_2UO_4$, instead of $Na_2U_2O_7$. Nevertheless the thermochemical cycle used for hydrogen generation can be closed, and chemical compounds used in these processes can also be recycled. So the cycle with $Na_2UO_4$ as its first reaction product has an advantage over the proposed UTC process, attributed to the fast reaction rate and high hydrogen yield in the first reaction step.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.7 no.3
• /
• pp.133-141
• /
• 2009

Study on the chemical speciation of uranium(VI) species, ${UO_2}^{2+}$, $UO_2(OH)^+$, ${(UO_2)}_2{(OH)_2}^{2+}$, ${(UO_2)}_3{(OH)_5}^+$, has been peformed by using time-resolved laser-induced fluorescence spectroscopy. Speciation sensitivity which depends on the excitation wavelength was investigated. We obtained the speciation sensitivity in the order of $10^{-9}$ M concentration of U(VI) compounds at the excitation wavelength of 266 nm. The fluorescence spectrum and lifetime of ${UO_2}^{2+}$ were carefully measured at pH 1 and ion strength of 0.1 M. The spectrum showed the four characteristic peaks located around 488, 509, 533, 559nm and the fluorescence lifetime of $1.92{\pm}0.17{\mu}s$. The wavelength shifts of fluorescence peaks and the change of lifetimes for uranium hydrolysis compounds were compared with those of ${UO_2}^{2+}$. We report on the characteristic features, the shifts of peaks to the longer wavelength direction and the prolonged lifetimes, in the fluorescence of the U(VI) hydrolysis compounds.

• Analytical Science and Technology
• /
• v.14 no.3
• /
• pp.203-211
• /
• 2001

In this paper it was described on the intereference effect of uranium to analyze Cs in $UO_2$ by Electron Probe Micro Analysis(EPMA) and the beam stability of Cs $L_{\alpha}$ X-ray intensity for some Cs compounds. According to the experimental results, the CsI showed the highest $L_{\alpha}$ X-ray intensity among the tested Cs compounds at the experimental condition; 15~30 kV of accelerating voltage and PET, LiF crystal. When 100 nA of beam current was applied to Cs compounds, Cs $L_{\alpha}$ X-ray intensity was continuously decreased with increasing time. The decreasing rate of Cs $L_{\alpha}$ X-ray intensity was directly proportional to the applied beam current and accelerating voltage but inversely proportional to the applied beam size. It was found that uranium interference can be prevented by using Cs $L_{\alpha}$ X-ray wavelength of Lif crytal for Cs analysis in $UO_2$ by EPMA.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.20 no.2
• /
• pp.193-207
• /
• 2022

Thermodynamically, TRUO_x, REO_x, and SrO_x can be chlorinated using ammonium chloride (NH₄Cl) as a chlorinating agent, whereas uranium oxides (U₃O₈ and UO₂) remain in the oxide form. In the preliminary experiments of this study, U₃O₈ and CeO₂ are reacted separately with NH₄Cl at 623 K in a sealed reactor. CeO₂ is highly reactive with NH₄Cl and becomes chlorinated into CeCl₃. The chlorination yield ranges from 96% to 100%. By contrast, U₃O₈ remains as UO₂ even after chlorination. We produced U/REO_x- and U/SrO_x-simulated fuels to understand the chlorination characteristics of the oxide compounds. Each simulated fuel is chlorinated with NH₄Cl, and the products are dissolved in LiCl-KCl salt to separate the oxide compounds from the chloride salt. The oxide compounds precipitate at the bottom. The precipitate and salt phases are sampled and analyzed via X-ray diffraction, scanning electron microscope-energy dispersive spectroscopy, and inductively coupled plasma-optical emission spectroscopy. The analysis results indicate that REO_x and SrO_x can be easily chlorinated from the simulated fuels; however, only a few of U oxide phases is chlorinated, particularly from the U/SrO_x-simulated fuels.

• Journal of Powder Materials
• /
• v.16 no.2
• /
• pp.115-121
• /
• 2009

The effects of thermal treatment conditions on ADU (ammonium diuranate) prepared by SOL-GEL method, so-called GSP (Gel supported precipitation) process, were investigated for $UO_2$ kernel preparation. In this study, ADU compound particles were calcined to $UO_3$ particles in air and Ar atmospheres, and these $UO_3$ particles were reduced and sintered in 4%-$H_2$/Ar. During the thermal calcining treatment in air, ADU compound was slightly decomposed, and then converted to $UO_3$ phases at $500^{\circ}C$. At $600^{\circ}C$, the $U_3O_8$ phase appeared together with $UO_3$. After sintering of theses particles, the uranium oxide phases were reduced to a stoichiometric $UO_2$. As a result of the calcining treatment in Ar, more reduced-form of uranium oxide was observed than that treated in air atmosphere by XRD analysis. The final phases of these particles were estimated as a mixture of $U_3O_7$ and $U_4O_9$.

• Nuclear Engineering and Technology
• /
• v.52 no.5
• /
• pp.1013-1021
• /
• 2020

A combination of simultaneous thermal analysis, evolved gas analysis and non-ambient XRD techniques was used to characterise and investigate the conversion reactions of ammonium uranates into uranium oxides. Two solid phases of the ternary system NH₃ - UO₃ - H₂O were synthesised under specified conditions. Microspheres prepared by the sol-gel method via internal gelation were identified as 3UO₃·2NH₃·4H₂O, whereas the product of a typical ammonium diuranate precipitation reaction was associated to the composition 3UO₃·NH₃·5H₂O. The thermal decomposition profile of both compounds in air feature distinct reaction steps towards the conversion to U₃O₈, owing to the successive release of water and ammonia molecules. Both compounds are converted into α-U₃O₈ above 550 ℃, but the crystallographic transition occurs differently. In compound 3UO₃·NH₃·5H₂O (ADU) the transformation occurs via the crystalline β-UO₃ phase, whereas in compound 3UO₃·2NH₃·4H₂O (microspheres) an amorphous UO₃ intermediate was observed. The new insights obtained on these uranate systems improve the information base for designing and synthesising minor actinide-containing target materials in future applications.

• Nuclear Engineering and Technology
• /
• v.51 no.7
• /
• pp.1799-1804
• /
• 2019

The concept of a new method, the CARBEX (CARBonate EXtraction) process, was proposed for reprocessing of spent uranium oxide fuel. The proposed process is based on use of water solutions of $Na_2CO_3$ or $(NH_4)_2CO_3$ and solvent extraction (SE) by the quaternary ammonium compounds for selective recovery and purification of U from the fission products (FPs). Applying of SE allows to reach high degree of purification of U from FPs. Carrying out the processes in poorly aggressive alkaline carbonate media leads to increasing safety of SNF's reprocessing and better selectivity of separation of lanthanides and actinides. Moreover carbonate reprocessing media allows to carry out a recycling and regeneration of reagents. We have been done laboratory scale experiments on the extraction components of simulated voloxidated spent fuel in the solutions of NaOH or $Na_2CO_3-H_2O_2$ and recovery of U from carbonate solutions by SE method using carbonate of methyltrioctylammonium in toluene. It was shown that the purification factors of U from impurities of simulated FPs reached values $10^3-10^5$. The received results support our opinion that CARBEX after the further development can become more safe, simple and profitable method of spent fuel reprocessing.