• 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.49 no.3
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• pp.541-548
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• 2017

Uranium metal used for fabrication of fuel for research reactors in India is generally produced by magnesio-thermic reduction of $UF_4$. Performance of magnesio-thermic reaction and recovery and quality of uranium largely depends on properties of $UF_4$. As ammonium diuranate (ADU) is first product in powder form in the process flow-sheet, properties of $UF_4$ depend on properties of ADU. ADU is generally produced from uranyl nitrate solution (UNS) for natural uranium metal production and from uranyl fluoride solution (UFS) for low enriched uranium metal production. In present paper, ADU has been produced via both the routes. Variation of uranium recovery and crystal structure and composition of ADU with progress in precipitation reaction has been studied with special attention on first appearance of the precipitate Further, ADU produced by two routes have been calcined to $UO_3$, then reduced to $UO_2$ and hydroflorinated to $UF_4$. Effect of two different process routes of ADU precipitation on the characteristics of ADU, $UO_3$, $UO_2$ and $UF_4$ were studied here.

• 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.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.