• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1931-1938
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• 2021

Uranium ingot is produced by metallothermic reduction of uranium tetrafluoride using magnesium or calcium as reductant. Presence of oxygen containing compounds viz. uranyl fluoride and uranium oxide in the starting uranium fluoride has a significant effect on the firing time, final temperature of the charge, slag-metal separation and hence the metal recovery. As reported in the literature, the maximum tolerable limit for uranyl fluoride in the UF₄ is 2.5 wt% and limit for uranium oxide content is in the range 2-3 wt%. No theoretical or experimental basis is available till date for these limits. Analyses have been carried out in this study to understand the effect of UO₂F₂ concentration in the starting fluoride on the final temperature of the products and thus the reduction characteristics. UF₄ having uranyl fluoride concentration, less than as well as more than 2.5 wt%, have been investigated. Thermodynamic calculations have been carried out to arrive at a general expression for the final temperature attained by the products during calciothermic reduction of UF₄. Finally, an upper limit for the oxygen containing impurities has been estimated using the CaO-CaF₂ phase diagram.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1909-1913
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• 2022

This work shows the technical feasibility to obtain uranium tetrafluoride through an electrochemical mercury cell. This technique represents a custom scaling-up methodology from our previous studies to obtain UF₄ using the dropping mercury electrode cell. The UF₄ products were obtained from natural UF₆ gas, which was hydrolyzed to obtain a 50 g/L UO₂F₂ solution. The electrolysis cell was made using a mercury reservoir, to reach UF₄ production rates of 1 Kg UF₄/day. This custom design allowed a stable UF₄ production thanks to the mercury cathode, which do not permit the accumulation of solid products in its surface. The cell was tested using current densities from 5.000 to 17.500 A/m² and temperatures from 25 to 65 ℃. The maximum current efficiency achieved under these conditions was 80%. The UF₄ powders possessed spherical morphology, with diameters between 20 and 80 ㎛. Compared to the SnCl₂ precipitation, this process did not allow preferential growth of the precipitates. This improved the compaction of the UF₄ - Mg powders mixtures, with densities between 3.0 and 3.5 g/cm³. The purity of the UF₄ products was over 98%.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1711-1716
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• 2017

Uranium tetrafluoride ($UF_4$) is the most used nuclear material for producing metallic uranium by reduction with Ca or Mg. Metallic uranium is a raw material for the manufacture of uranium silicide, $U_3Si_2$, which is the most suitable uranium compound for use as nuclear fuel for research reactors. By contrast, ammonium uranyl carbonate is a traditional uranium compound used for manufacturing uranium dioxide $UO_2$ fuel for nuclear power reactors or $U_3O_8-Al$ dispersion fuel for nuclear research reactors. This work describes a procedure for recovering uranium and ammonium fluoride ($NH_4F$) from a liquid residue generated during the production routine of ammonium uranyl carbonate, ending with $UF_4$ as a final product. The residue, consisting of a solution containing high concentrations of ammonium ($NH_4^+$), fluoride ($F^-$), and carbonate ($CO_3^{2-}$), has significant concentrations of uranium as $UO_2^{2+}$. From this residue, the proposed procedure consists of precipitating ammonium peroxide fluorouranate (APOFU) and $NH_4F$, while recovering the major part of uranium. Further, the remaining solution is concentrated by heating, and ammonium bifluoride ($NH_4HF_2$) is precipitated. As a final step, $NH_4HF_2$ is added to $UO_2$, inducing fluoridation and decomposition, resulting in $UF_4$ with adequate properties for metallic uranium manufacture.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1560-1569
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• 2022

Liquid fueled molten salt reactors (MSRs) have seen renewed interest because of their inherent safety features, higher thermal efficiency and potential for efficient thorium utilisation for power generation. Thorium fluoride is one of the salts used in liquid fueled MSRs employing Th-U cycle. In the present study, ThF₄ was prepared by hydro-fluorination of ThO₂ using anhydrous HF gas. Process parameters viz. bed depth, hydrofluorination time and hydrofluorination temperature, were optimized for the preparation of ThF₄ in a static bed reactor setup. The products were characterized with X-Ray diffraction and experimental conditions for complete conversion to ThF₄ were established which also corroborated with the yield values. Hydrofluorination of ThO₂ at 450 ℃ for half an hour at a bed depth of 6 mm gave the best result, with a yield of about 99.36% ThF₄. No unconverted oxide or any other impurity was observed. Rietveld refinement was performed on the XRD data of this ThF₄, and Chi2 value of 3.54 indicated good agreement between observed and calculated profiles.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.109-114
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• 2020

This work presents the results of laboratory scale tests of the CARBOFLUOREX (CARBOnate FLUORide EXtraction) process - a novel technology for the recovery of U and Pu from the solid fluorides residue (fluorination ash) of Fluoride Volatility Method (FVM) reprocessing of spent nuclear fuel (SNF). To study the oxidative leaching of U from the fluorination ash (FA) by Na₂CO₃ or Na₂CO₃-H₂O₂ solutions followed by solvent extraction by methyltrioctylammonium carbonate in toluene and purification of U from the fission products (FPs) impurities we used a surrogate of FA consisting of UF₄ or UO₂F₂, and FPs fluorides with stable isotopes of Ce, Zr, Sr, Ba, Cs, Fe, Cr, Ni, La, Nd, Pr, Sm. Purification factors of U from impurities at the solvent extraction refining stage reached the values of 10⁴-10⁵, and up to 10⁶ upon the completion of the processing cycle. Obtained results showed a high efficiency of the CARBOFLUOREX process for recovery and separating of U from FPs contained in FA, which allows completing of the FVM cycle with recovery of U and Pu from hardly processed FA.