Uranium settler

The simple box-type mixer—settler (113) has been used extensively in the UK for the separation and purification of uranium and plutonium (114). In this type of extractor, interstage flow is handled through a partitioned box constmction. Interstage pumping is not needed because the driving force is provided by the density difference between solutions in successive stages (see Plutoniumand plutonium compounds Uraniumand uranium compounds). 

The development of the novel Davy-McKee combined mixer—settler (CMS) has been described (121). It consists of a single vessel (Fig. 13d) in which three 2ones coexist under operating conditions. A detailed description of units used for uranium recovery has been reported (122), and the units have also been studied at the laboratory scale (123). AppHcation of the Davy combined mixer electrostatically assisted settler (CMAS) to copper stripping from an organic solvent extraction solution has been reported (124). 

For vanadium solvent extraction, Hon powder can be added to reduce pentavalent vanadium to quadrivalent and trivalent Hon to divalent at a redox potential of —150 mV. The pH is adjusted to 2 by addition of NH, and an oxyvanadium cation is extracted in four countercurrent stages of mixer—settlers by a diesel oil solution of EHPA. Vanadium is stripped from the organic solvent with a 15 wt % sulfuric acid solution in four countercurrent stages. Addition of NH, steam, and sodium chlorate to the strip Hquor results in the precipitation of vanadium oxides, which are filtered, dried, fused, and flaked (22). Vanadium can also be extracted from oxidized uranium raffinate by solvent extraction with a tertiary amine, and ammonium metavanadate is produced from the soda-ash strip Hquor. Fused and flaked pentoxide is made from the ammonium metavanadate (23). 

FIG. 15-29 Kerr-McGee multistage mixer-settler, a) and (h) For uranium, (c) For vanadium extraction. 

Pandey, N.K. Koganti, S.B. Simulation of electro-mixer-settler for the partitioning of uranium and plutonium in PUREX process, Ind. J. Chem. Technol. 11 (2004) 535-547. 

The centrifugal contactor was first used to reprocess spent nuclear fuel at the SRS in 1966 (Webster et al., 1969). For almost 40 years, this 18-stage 25-cm SRL contactor was used for the extraction and scrub sections (the A-bank) of the PUREX (plutonium-uranium extraction) process at the SRS. Contactor operation stopped when the facility in which they were housed was shut down in 2003. This 18-stage contactor replaced a 24-stage mixer-settler. Mixer-settlers continued to be used for the rest of the processing, as most of the radiation was removed in the A-bank. The ability to... 

The uranium and plutonium are recovered for further use by first dissolving the spent fuel in nitric acid and subjecting the resulting solution to a solvent extraction process. Several different processes exist, the best known being the Purex process (Fig. 18), in which tributyl phosphate (TBP) (30% solution in kerosene) is the extractant. Extraction is carried out in compact mixer-settlers or air-pulsed columns fabricated of stainless steel, with about 99.9% removal of uranium and plutonium in the extract. 

The early experiments on solvent extraction directly from leached pulp were beset with problems such as losses of solvent in the aqueous phase and the formation of emulsions. The use of mixer-settler, pump mixer, and internal mixer-settler type contactors on a laboratory scale (Gil) has demonstrated the feasibility of uranium extraction from desanded slurries with 5-1. )% solids and from high-density slurries with 48-60% percent solids. The deemulsification rate of a synthetic slurry as a function of the temperature of the system and the pH of the slurries (T12) and the effect of extractant entrainment in the aqueous effluent on solvent extraction of uranium from slurries containing more than 40% solids (E6) have been studied. 

In 1968, an electrolytic reduction process was proposed by A. Schneider and A. L. Ayers (6) to circumvent the above disadvantages. A research program was carried out in the Allied Chemical Corporation s laboratories during the years 1968 to 1972 to develop the process and equipment. The work resulted in the development of the Electropulse Column ( 7) for the continuous (differential) electrolytic uranium-plutonium partition process, which was later scaled up, fabricated, and installed in the Allied-General Nuclear Services reprocessing plant at Barnwell, South Carolina. About the same time, a stagewise electrolytic uranium-plutonium partition process was tested on a mini mixer-settler unit in Germany. (8)... 

A laboratory study was undertaken to determine the behaviour of neptunium in the WAK flowsheet, and to devise a procedure for its recovery. Based on static ( ) and counter-current experiments (J5), the conclusion was reached that about half of the Np is co-extracted with the U and Pu in the HA-HS mixer-settlers of WAK while the other half is rejected to the HAW, see Fig.1. It could also be shown that an increase of the aqueous acidity, or the addition of pentavalent vanadium as an oxidant into the lower stages of the HA mixer-settler (6), would increase the Np yield in the organic solvent. In the 1BX-1BS mixer-settlers where the partitioning of U and Pu is carried out by use of uranium (IV)nitrate - hydrazine nitrate, a splitting of the coextracted Np between the two product streams was observed the proportions of the (co-extracted) Np which ended up in the 1CU (uranium product) stream fluctuated from 30 to 93 % while the difference amount (from 7 to 70 %) ended up in the 1 BP (plutonium product) stream. 

In the Purex process, irradiated UO2 is dissolved in nitric acid under such conditions that uranium is oxidized to uranyl nitrate and plutonium to Pu(N03)4. The resulting aqueous solution of uranyl, plutonium, and fission-product nitrates is fed to the center of countercurrent solvent extraction contactor I, which may be either a pulse column or a battery of mixer-settlers. This contactor is refluxed at one end by clean solvent and at the other by a dilute nitric acid scrub solution. The solvent extracts all the uranium and plutonium from the aqueous phase and some of the fission products. The fission products are removed from the solvent by the nitric acid scrub solution. Fission products leave contactor I in solution in aqueous nitric acid. 

Mixer-settler contactors of much larger scale are used in the solvent extraction operations associated with production of natural uranium (cf. Chap. 5), wherein nuclear criticality is not... 

The mixer-settler used at the Vitro uranium recovery operation near Salt Lake City, Utah, is shown schematically in Fig. 4.25. The contactor is a rubber-lined tank 6.1 m in diameter and... 

Leach liquor containing about 1 g UsOg/liter at a pH around 1.0 is fed at the rate of 3800 liters/min to the first of four mixer-settler states in series, where the uranium is extracted by a solution ccMitaining 3 v/o Alamine-336 (mixed -trioctyl- and n-tridecylamines) and 3 v/o isodecanol in a high-boiling kerosene diluent. These four stages reduce the uranium content of the aqueous stream from 1 g U3 0g/liter to around 0.001, while increasing that of the solvent from 0.002 g/liter to 3.33. 

Molybdenum is not stripped from the amine solvent by sodium chloride. If not kept below around 0.03 g MoAiter, it precipitates as a sludge and interferes with uranium extraction. To control molybdenum concentration, a portion of the solvent leaving the uranium stripping section is contacted in a sin e mixer-settler with an aqueous solution of Na2C03 and NH4OR This converts the molybdenum to sodium molybdate, Na2Mo04, and transfers it to the aqueous phase, from which molybdenum is recovered as a by-product. 

---