Purex solvent extraction

S. B. Watson md R. H. Rmney, Modfcations of the SEPHIS Computer Code for Calculating the Purex Solvent Extraction System, ORNL/TM-5123, Oak Ridge National Laboratory, Oak Ridge, Term., 1975. 

Mitchell, A.D. 1979. SEPHIS-MOD4 A user s manual to revised model of the Purex solvent extraction system. ORNL-5471. 

Waste Compositions Selected for Study in the WFP. The major source of high-level wastes is the raflBnate or aqueous waste stream from the first cycle of the Purex solvent extraction process. The stream is a nitric acid solution containing over 99.9% of the nonvolatile fission... 

Wymer, R.G. Vondra, B.L. PUREX solvent extraction chemistry. In Light Water Reactor Nuclear Fuel Cycle CRC Press, Inc. Boca Raton, FL, 1981 103-162. 

G16. Gronier, W. S. Calculation of the Transient Behavior of a Dilute-Purex Solvent Extraction Process Having Application to the Reprocessing of LMFBR Fuels, Report ORNL 746, Apr. 1972. 

Geldard, J. and A. Beyerlein. 1985. PUBG PUREX solvent extraction process model. Report JkNL/ NESC-959. Argonne, IL Argonne national laboratory. 

Me Kibben, J.M. and J.E. Bercaw, 1971. Hydroxylamine nitrate as a plutonium reductant in the PUREX solvent extraction process. Report DP-1248. Aiken, SC Savannah River Laboratory. 

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. 

Solvent Extraction. A modified, one-cycle PUREX process is used at Rocky Flats to recover plutonium from miscellaneous Pu-U residues (11). The process utilizes the extraction of uranium (VI) into tributyl phosphate (TBP), leaving plutonium (III) in the raffinate. The plutonium is then sent to ion exchange for... 

Butex A process for separating the radioactive components of spent nuclear fuel by solvent extraction from nitric acid solution, using diethylene glycol dibutyl ether (also called Butex, or dibutyl carbitol) as the solvent. Developed by the Ministry of Supply (later the UK Atomic Energy Authority) in the late 1940s. Operated at Windscale from 1952 until 1964 when it was superseded by the Purex process. 

Purex [Plutonium and uranium recovery by extraction] A process for the solvent extraction of plutonium from solutions of uranium and fission products, obtained by dissolving spent nuclear fuel elements in nitric acid. The solvent is tri-n-butyl phosphate (TBP) in... 

Redox [Reduction oxidation] A process for separating the components of used nuclear fuel by solvent extraction. It was the first process to be used and was brought into operation at Hanford, United States, in 1951, but was superseded in 1954 by the Purex process. The key to the process was the alternate reduction and oxidation of the plutonium, hence the name. The solvent was Hexone (4-methyl-2-pentanone, methyl isobutyl ketone), so the process was also known as the Hexone process. The aqueous phase contained a high... 

Thorex [Thorium extraction] A process for separating the products from the nuclear breeder reaction in which uranium-233 is produced by the neutron bombardment of thorium-232. It uses solvent extraction into tri-n-butyl phosphate. Developed at the Oak Ridge National Laboratory, TN, in the early 1960s. See also Butex, Purex, Redox. 

The solvent extraction process that uses TBP solutions to recover plutonium and uranium from irradiated nuclear fuels is called Purex (plutonium uranium extraction). The Purex process provides recovery of more than 99% of both uranium and plutonium with excellent decontamination of both elements from fission products. The Purex process is used worldwide to reprocess spent reactor fuel. During the last several decades, many variations of the Purex process have been developed and demonstrated on a plant scale. 

Np, and fission products. The Thorex solvent extraction process is generally used to reprocess spent Th-based fuels. As in the Purex process, the solvent is TBP diluted in an appropriate mixture of aliphatic hydrocarbons. Figure 12.9 shows the Thorex process flow sheet used by Kuchler et al. [41] for reprocessing high-burn-up thorium fuel. 

Dietz, M. L., Horwitz, E. P, Rogers, R. D., Extraction of strontium from acidic nitrate media using a modified purex solvent, Solv. Extr. Ion Exch., 13, 1-17,1995. 

The uranium and thorium ore concentrates received by fuel fabrication plants still contain a variety of impurities, some of which may be quite effective neutron absorbers. Such impurities must be almost completely removed if they are not seriously to impair reactor performance. The thermal neutron capture cross sections of the more important contaminants, along with some typical maximum concentrations acceptable for fuel fabrication, are given in Table 9. The removal of these unwanted elements may be effected either by precipitation and fractional crystallization methods, or by solvent extraction. The former methods have been historically important but have now been superseded by solvent extraction with TBP. The thorium or uranium salts so produced are then of sufficient purity to be accepted for fuel preparation or uranium enrichment. Solvent extraction by TBP also forms the basis of the Purex process for separating uranium and plutonium, and the Thorex process for separating uranium and thorium, in irradiated fuels. These processes and the principles of solvent extraction are described in more detail in Section 65.2.4, but the chemistry of U022+ and Th4+ extraction by TBP is considered here. 

Many variants of the Purex (Plutonium Uranium Reduction Extraction) process23S based on TBP extraction have been developed but a basic outline flowsheet is illustrated in Figure 38. This shows the so-called early split flowsheet most commonly used in existing plants. It involves the separation of the uranium and plutonium using two different back-extractant streams during the first solvent extraction cycle. Additional solvent extraction cycles are then carried out independently on the uranium and plutonium streams to effect further purification. An alternative arrangement is the iate split flowsheet used at the Cap La Hague plant in France, and the... 

As a consequence, corporations operating PUREX plants have been using sophisticated process simulation codes, including the PAREX code in France (45-47), SpeedUp (Aspen Plus) in the UK (48), and SIMPSEX code in India (49-51). Argonne Model for Universal Solvent Extraction (AMUSE) code in the United States was contrived not only for PUREX, but for UREX+ processes (52), which will be mentioned later. In Japan, similar efforts have also been made (53-55). 

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