Uranium plutonium extracted from

Tracer techniques, for example, are used to obtain very small but representative and measurable samples of highly radioactive spent fuel solutions. One millilitre of the solution is then spiked with a known amount of uranium and plutonium tracer isotopes. A few microlitres of the spiked solution are dried and shipped to SAL. One to fifty nanograms of uranium or plutonium extracted from this tiny sample are sufficient for a complete analysis representing the composition of half a tonne of irradiated fuel with an accuracy of 0.3 to 0.5%. 

Applicability of SFE to nuclear fuel reprocessing has been proposed by Smart and Wai et al. (17, 18). We have developed a new process which employs a high pressure mixture of TBP-HNO3-H2O-CO2 as is described in this chapter and this approach has indicated a very efEcient extraction of uranium from UO2. Now, the nuclear industries have paid attention to the applications of SFE to future processes. In Japan, we have started a four-year project with nuclear plant construction companies to demonstrate uranium and plutonium extraction from a mixed oxide fuel using the high pressure mixture. On the other hand, uranium and plutonium will be extracted from the irradiated nuclear fuel with TBP(HN03)i.s(H20)o.6 in the same project. 

The most important TBP process is the PUREX process, in which 20—40 vol/% TBP in the hydrocarbon diluent is used. The flowsheet commonly used in the United States specifies 30 vol/% TBP (Long, 1978). A typical American flowsheet is shown in Figure 14.14. Uranium is extracted from the feed and decontaminated in the first contactor, separated from plutonium in the second contactor, and stripped back into an aqueous phase in the third contactor. A high aqueous phase to organic phase flow ratio... 

During the initial stages of development, it was assumed that the isotopic composition of plutonium in the fresh fuel would correspond to reactor-grade plutonium extracted from the cooled spent fuel of a typical light water reactor (e.g., from a 900 MW(e) PWR fuel irradiated up to a bum-up of 33 MW-d/kg U, reprocessed after 10 years of cooling) and loaded in the RBEC-M reactor in two years. The isotopic composition of uranium corresponds to depleted uranium with a content of 0.1 weight %... 

The last paragraph in this extract refers to work on the separation of uranium by ether extraction as a step toward obtaining a plutonium concentrate from a large sample of neutron-irradiated uranyl nitrate. 

Commercial-scale application of solvents coming under the category of neutral reagents is largely found as applied to the nuclear industry materials, as in example, for the separation and refining of uranium, plutonium, thorium, zirconium, and niobium. A process flowsheet for extracting niobium and tantalum from various resources is shown in Figure 5.23. It will... 

Solvent extraction, also called liquid-liquid extraction, can be used to separate a substance from a solution by extraction into another solvent. It can be used ether to recover a valuable substance from the original solution, or to purify the original solvent by removing an unwanted component. Examples of solvent extraction are the extraction of uranium and plutonium salts from solution in nitric acid, in the nuclear industry and the purification of water. 

A primary goal of chemical separation processes in the nuclear industry is to recover actinide isotopes contained in mixtures of fission products. To separate the actinide cations, advantage can be taken of their general chemical properties [18]. The different oxidation states of the actinide ions lead to ions of charges from +1 (e.g., NpOj) to +4 (e.g., Pu" " ) (see Fig. 12.1), which allows the design of processes based on oxidation reduction reactions. In the Purex process, for example, uranium is separated from plutonium by reducing extractable Pu(IV) to nonextractable Pu(III). Under these conditions, U(VI) (as U02 ) and also U(IV) (as if present, remain in the... 

The Purex process, ie, plutonium uranium reduction extraction, employs an organic phase consisting of 30 wt % TBP dissolved in a kerosene-type diluent. Purification and separation of U and Pu is achieved because of the extractability of U02+2 and Pu(IV) nitrates by TBP and the relative inextractability of Pu(III) and most fission product nitrates. Plutonium nitrate and U02(N03)2 are extracted into the organic phase by the formation of compounds, eg, Pu(N03)4 -2TBP. The plutonium is reduced to Pu(III) by treatment with ferrous sulfamate, hydrazine, or hydroxylamine and is transferred to the aqueous phase U remains in the organic phase. Further purification is achieved by oxidation of Pu(III) to Pu(IV) and re-extraction with TBP. The plutonium is transferred to an aqueous product. Plutonium recovery from the Purex process is ca 99.9 wt % (128). Decontamination factors are 106 — 10s (97,126,129). A flow sheet of the Purex process is shown in Figure 7. 

Suzuki, S., Tamura, K., Tachimori, S., Usui, Y. 1999. Extraction of uranium(VI) and plutonium(IV) from nitric acid solution by substituted cyclic amides. Solvent Extr. Res. Dev. Jpn. 6 72-79. 

The third fact is that spent nuclear fuel is not waste. Spent nuclear fuel contains 2% to 3% waste, but is about 97% recoverable uranium and plutonium. Each bundle has the potential electric energy equivalent of more than 10 million barrels of oil. High-level nuclear wastes consist of fission products and actinides that are extracted from spent fuel, but not saved for commercial use or research. Spent fuel may be temporarily stored until it is reprocessed to separate the waste from the valuable plutonium and uranium. The remaining glassified waste will then be permanently entombed. 

The control of the actinide metal ion valence state plays a pivotal role in the separation and purification of uranium and plutonium during the processing of spent nuclear fuel. Most commercial plants use the plutonium-uranium reduction extraction process (PUREX) [58], wherein spent fuel rods are initially dissolved in nitric acid. The dissolved U and Pu are subsequently extracted from the nitric solution into a non-aqueous phase of tributyl phosphate (TBP) dissolved in an inert hydrocarbon diluent such as dodecane or odourless kerosene (OK). The organic phase is then subjected to solvent extraction techniques to partition the U from the Pu, the extractability of the ions into the TBP/OK phase being strongly dependent upon the valence state of the actinide in question. 

Amine Extraction. Extraction of uranium (VI) and plutonium (VI) from acetic acid with tri-iso-octylamine in xylene has been studied previously by Moore (JJ), who suggests that the extracted species is the triacetate ion. The spectra in Figure 1 show this is not the predominate species extracted into TOA-xylene solutions (9). Comparing the spectrum of this extracted uranium with the spectra of pure tetraacetate and triacetate complexes in Figure 3 shows that the TOA-xylene extracts a mixture of these ions that is largely tetraacetate uranyl complex. 

Carrier (TOPO)-mediated transport of uranium(Vl) has been studied by Akiba and Hashrmoto [94] who have observed that uranium was extracted in the liquid membrane as U02(N03)2 2TOPO and stripped into the carbonate solutions as U02(C03)3. Using TOPO and HDEHP mixmre, liquid membrane technique was apphed for the recovery of uranium from WPPA [95]. Carrier-facilitated Pu(IV) pertraction through an SLM was standardized for its decontamination from oxalate wastes employing a commercially available Cyanex-923 (TOPO analog) in dodecane as the receptor [96,97]. More than 95% of plutonium could be easily recovered from Pu oxalate wastes solution during Pu reconversion operations. 

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