Uranium separation from

Efimova, Z.I., Smirnov, Yu.V. and Sokolova, I.D., Development of new methods of Uranium separation from non-tradition sources. (In Russian) At. Tekh. Rubezhom. 4 (1987) 3-9. 

Zaheri, A., Moheb, A., Keshtkar, A.R. Shirani, A.S. (2010) Uranium separation from wastewater by electrodialysis. Iranian Journal of Environmental Health Science Engineering, 7,429-436. 

Uranium is among the major elements in most fuel cycles and the effective separation of actinides from lanthanides is highly desirable mainly because of the neutron poisoning properties of the fission products. Electroseparation processes are investigated widely in molten chlorides and fluorides. For example, U separation by electrolysis on a solid inert cathode is used within the integral fast reactor (IFR) concept [1]. It is the main goal of this work to extend knowledge about the possibilities of uranium separation from molten fluoride systems. 

MIBK is a highly effective separating agent for metals from solutions of their salts and is used in the mining industries to extract plutonium from uranium, niobium from tantalum, and zirconium from hafnium (112,113). MIBK is also used in the production of specialty surfactants for inks (qv), paints, and pesticide formulations, examples of which are 2,4,7,9-tetramethyl-5-decyn-4,7-diol and its ethoxylated adduct. Other appHcations include as a solvent for adhesives and wax/oil separation (114), in leather (qv) finishing, textile coating, and as a denaturant for ethanol formulations. 

Laser isotope separation techniques have been demonstrated for many elements, including hydrogen, boron, carbon, nitrogen, oxygen, sHicon, sulfur, chlorine, titanium, selenium, bromine, molybdenum, barium, osmium, mercury, and some of the rare-earth elements. The most significant separation involves uranium, separating uranium-235 [15117-96-1], from uranium-238 [7440-61-1], (see Uranium and uranium compounds). The... 

Uranium-239 [13982-01 -9] has a half-life of 23.5 min neptunium-239 [13968-59-7] has a half-life of 2.355 d. Recycling or reprocessing of spent fuel involves separation of plutonium from uranium and from bulk fission product isotopes (see Nuclearreactors, chemical reprocessing). 

An improved solvent extraction process, PUREX, utilizes an organic mixture of tributyl phosphate solvent dissolved in a hydrocarbon diluent, typically dodecane. This was used at Savannah River, Georgia, ca 1955 and Hanford, Washington, ca 1956. Waste volumes were reduced by using recoverable nitric acid as the salting agent. A hybrid REDOX/PUREX process was developed in Idaho Falls, Idaho, ca 1956 to reprocess high bum-up, fuUy enriched (97% u) uranium fuel from naval reactors. Other separations processes have been developed. The desirable features are compared in Table 1. 

Uranium Purification. Subsequent uranium cycles provide additional separation from residual plutonium and fission products, particularly zirconium— niobium and mthenium (30). This is accompHshed by repeating the extraction/stripping cycle. Decontamination factors greater than 10 at losses of less than 0.1 wt % are routinely attainable. However, mthenium can exist in several valence states simultaneously and can form several nitrosyl—nitrate complexes, some for which are extracted readily by TBP. Under certain conditions, the nitrates of zirconium and niobium form soluble compounds or hydrous coUoids that compHcate the Hquid—Hquid extraction. SiUca-gel adsorption or one of the similar Hquid—soHd techniques may also be used to further purify the product streams. 

The same chemical separation research was done on thorium ores, leading to the discovery of a completely different set of radioactivities. Although the chemists made fundamental distinctions among the radioactivities based on chemical properties, it was often simpler to distinguish the radiation by the rate at which the radioactivity decayed. For uranium and thorium the level of radioactivity was independent of time. For most of the radioactivities separated from these elements, however, the activity showed an observable decrease with time and it was found that the rate of decrease was characteristic of each radioactive species. Each species had a unique half-life, ie, the time during which the activity was reduced to half of its initial value. 

Carbonate leaching under ambient conditions is extremely slow with poor recoveries. Therefore, the ore is typically leached in an autoclave with air providing most of the needed oxygen. The leach Hquor is separated from the soHd in a countercurrent—decantation system of thickeners, and the uranium is precipitated from the clarified sodium carbonate solution with addition of sodium hydroxide (eq. 9) (23). 

Solvent Extraction. Solvent extraction has widespread appHcation for uranium recovery from ores. In contrast to ion exchange, which is a batch process, solvent extraction can be operated in a continuous countercurrent-fiow manner. However, solvent extraction has a large disadvantage, owing to incomplete phase separation because of solubihty and the formation of emulsions. These effects, as well as solvent losses, result in financial losses and a potential pollution problem inherent in the disposal of spent leach solutions. For leach solutions with a concentration greater than 1 g U/L, solvent extraction is preferred. For low grade solutions with <1 g U/L and carbonate leach solutions, ion exchange is preferred (23). Solvent extraction has not proven economically useful for carbonate solutions. 

Extraction of Bertrandite. Bertrandite-containing tuff from the Spor Mountain deposits is wet milled to provide a thixotropic, pumpable slurry of below 840 p.m (—20 mesh) particles. This slurry is leached with sulfuric acid at temperatures near the boiling point. The resulting beryUium sulfate [13510-49-1] solution is separated from unreacted soflds by countercurrent decantation thickener operations. The solution contains 0.4—0.7 g/L Be, 4.7 g/L Al, 3—5 g/L Mg, and 1.5 g/L Fe, plus minor impurities including uranium [7440-61-1/, rare earths, zirconium [7440-67-7] titanium [7440-32-6] and zinc [7440-66-6]. Water conservation practices are essential in semiarid Utah, so the wash water introduced in the countercurrent decantation separation of beryUium solutions from soflds is utilized in the wet milling operation. 

Assay of beryUium metal and beryUium compounds is usuaUy accompHshed by titration. The sample is dissolved in sulfuric acid. Solution pH is adjusted to 8.5 using sodium hydroxide. The beryUium hydroxide precipitate is redissolved by addition of excess sodium fluoride. Liberated hydroxide is titrated with sulfuric acid. The beryUium content of the sample is calculated from the titration volume. Standards containing known beryUium concentrations must be analyzed along with the samples, as complexation of beryUium by fluoride is not quantitative. Titration rate and hold times ate critical therefore use of an automatic titrator is recommended. Other fluotide-complexing elements such as aluminum, sUicon, zirconium, hafnium, uranium, thorium, and rate earth elements must be absent, or must be corrected for if present in smaU amounts. Copper-beryUium and nickel—beryUium aUoys can be analyzed by titration if the beryUium is first separated from copper, nickel, and cobalt by ammonium hydroxide precipitation (15,16). 

The separation of basic precipitates of hydrous Th02 from the lanthanides in monazite sands has been outlined in Fig. 30.1 (p. 1230). These precipitates may then be dissolved in nitric acid and the thorium extracted into tributyl phosphate, (Bu"0)3PO, diluted with kerosene. In the case of Canadian production, the uranium ores are leached with sulfuric acid and the anionic sulfato complex of U preferentially absorbed onto an anion exchange resin. The Th is separated from Fe, A1 and other metals in the liquor by solvent extraction. 

Many of the fission products formed in a nuclear reactor are themselves strong neutron absorbers (i.e. poisons ) and so will stop the chain reaction before all the (and Pu which has also been formed) has been consumed. If this wastage is to be avoided the irradiated fuel elements must be removed periodically and the fission products separated from the remaining uranium and the plutonijjm. Such reprocessing is of course inherent in the operation of fast-breeder reactors, but whether or not it is used for thermal reactors depends on economic and political factors. Reprocessing is currently undertaken in the UK, France and Russia but is not considered to be economic in the USA. 

Other methods reported for the determination of beryllium include UV-visible spectrophotometry [80,81,83], gas chromatography (GC) [82], flame atomic absorption spectrometry (AAS) [84-88] and graphite furnace (GF) AAS [89-96]. The ligand acetylacetone (acac) reacts with beryllium to form a beryllium-acac complex, and has been extensively used as an extracting reagent of beryllium. Indeed, the solvent extraction of beryllium as the acety-lacetonate complex in the presence of EDTA has been used as a pretreatment method prior to atomic absorption spectrometry [85-87]. Less than 1 p,g of beryllium can be separated from milligram levels of iron, aluminium, chromium, zinc, copper, manganese, silver, selenium, and uranium by this method. See also Sect. 5.74.9. 

From Figs. 1.1 and 1.2 it follows that the main end products of the uranium and thorium series are isotopes of lead (at the time referred to as Pb206.5 and ThO2208.4). The end products are thus isotopes of lead differing by two mass units. This observation became the motivation for the measurement of atomic weights of lead samples separated from thorium and uranium minerals. In his Nobel Lecture, Soddy describes this work as follows ... 

William Crookes was able to separate from uranium salts by chemical means a small quantity of an intensely radioactive substance, which he called Uranium X, the residual uranium having lost most of its activity and M. 

For the extraction of Tc from molybdemun irradiated by neutrons or separated from uranium fission products, inorganic sorbents, especially aliuninum oxide have widely been applied. In preparing a Tc generator from irradiated molybdenum , MoOj is dissolved in cone, nitric acid, the solution is diluted and passed through an aluminum oxide column. The column is then eluted by 0.2 N H2SO4 to extract Tc. If molybdenum is adsorbed by AljOj as molybdatophos-phate instead of molybdate, the exchange capacity of molybdenum increases from... 

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... 

Uranium mineral first is digested with hot nitric acid. AH uranium and radium compounds dissolve in the acid. The solution is filtered to separate insoluble residues. The acid extract is then treated with sulfate ions to separate radium sulfate, which is co-precipitated with the sulfates of barium, strontium, calcium, and lead. The precipitate is boiled in an aqueous solution of sodium chloride or sodium hydroxide to form water-soluble salts. The solution is filtered and the residue containing radium is washed with boiling water. This residue also contains sulfates of other alkahne earth metals. The sohd sulfate mixture of radium and other alkahne earth metals is fused with sodium carbonate to convert these metals into carbonates. Treatment with hydrochloric acid converts radium and other carbonates into chlorides, all of which are water-soluble. Radium is separated from this solution as its chloride salt by fractional crystallization. Much of the barium, chemically similar to radium, is removed at this stage. Final separation is carried out by treating radium chloride with hydrobromic acid and isolating the bromide by fractional crystallization. 

Finely-ground monazite is treated with a 45% NaOH solution and heated at 138°C to open the ore. This converts thorium, uranium, and the rare earths to their water-insoluble oxides. The insoluble residues are filtered, dissolved in 37% HCl, and heated at 80°C. The oxides are converted into their soluble chlorides. The pH of the solution is adjusted to 5.8 with NaOH. Thorium and uranium are precipitated along with small quantities of rare earths. The precipitate is washed and dissolved in concentrated nitric acid. Thorium and uranium are separated from the rare earths by solvent extraction using an aqueous solution of tributyl phosphate. The two metals are separated from the organic phase by fractional crystallization or reduction. 

Radioactivity of uranium can be measured by alpha counters. The metal is digested in nitric acid. Alpha activity is measured by a counting instrument, such as an alpha scintillation counter or gas-flow proportional counter. Uranium may be separated from the other radioactive substances by radiochemical methods. The metal or its compound(s) is first dissolved. Uranium is coprecipitated with ferric hydroxide. Precipitate is dissolved in an acid and the solution passed through an anion exchange column. Uranium is eluted with dilute hydrochloric acid. The solution is evaporated to near dryness. Uranium is converted to its nitrate and alpha activity is counted. Alternatively, uranium is separated and electrodeposited onto a stainless steel disk and alpha particles counted by alpha pulse height analysis using a silicon surface barrier detector, a semiconductor particle-type detector. 

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