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Reprocessing of fluorination ash surrogate in the CARBOFLUOREX process
Boyarintsev, Alexander V.,Stepanov, Sergei I.,Chekmarev, Alexander M.,Tsivadze, Aslan Yu. Korean Nuclear Society 2020 Nuclear Engineering and Technology Vol.52 No.1
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 Na<sub>2</sub>CO<sub>3</sub> or Na<sub>2</sub>CO<sub>3</sub>-H<sub>2</sub>O<sub>2</sub> 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 UF<sub>4</sub> or UO<sub>2</sub>F<sub>2</sub>, 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 10<sup>4</sup>-10<sup>5</sup>, and up to 10<sup>6</sup> 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.
Separation and purification of elements from alkaline and carbonate nuclear waste solutions
Boyarintsev Alexander V.,Stepanov Sergei I.,Kostikova Galina V.,Zhilov Valeriy I.,Safiulina Alfiya M.,Tsivadze Aslan Yu 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.2
This article provides a survey of wet (aqueous) methods for recovery, separation, and purification of uranium from fission products in carbonate solutions during the reprocessing of spent nuclear fuel and methods for removal of radionuclides from alkaline radioactive waste. The main methods such as selective direct precipitation, ion exchange, and solvent extraction are considered. These methods were compared and evaluated for reprocessing of spent nuclear fuel in carbonate media according to novel alternative non-acidic methods and for treatment processes of alkaline radioactive waste.
Reprocessing of simulated voloxidized uranium–oxide SNF in the CARBEX process
Alexander V. Boyarintsev,Sergei I. Stepanov,Galina V. Kostikova,Valeriy I. Zhilov,Alexander M. Chekmarev,Aslan Yu. Tsivadze 한국원자력학회 2019 Nuclear Engineering and Technology Vol.51 No.7
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 Na2CO3 or (NH4)2CO3 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 Na2CO3–H2O2 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 103–105. The received results support our opinion that CARBEX after the further development can become more safe, simple and profitable method of spent fuel reprocessing.
Reprocessing of spent nuclear fuel in carbonate media: Problems, achievements, and prospects
Stepanov Sergei I.,Boyarintsev Alexander V. 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.7
The review discusses various alternative approaches for spent nuclear fuel (SNF) reprocessing in aqueous carbonate media. The main stages, schemes, and methods of the most well-known and well-described processes for reprocessing SNF and some high-level radioactive waste using carbonate systems developed by research groups in Japan, the United States of America, the Republic of Korea, and the Russian Federation described and compared. The main advantages of such methods are outlined compared to the SNF reprocessing in nitric acid media. The levels of development and proximity of the designed processes to the industrial implementation are shown. The main principle achievements, prospects, and routes for the refinement of such methods for the technology of SNF reprocessing and handling of high-level radioactive waste formulated