Solvents Purex process

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. 

Chemistry. Chemical separation is achieved by countercurrent Hquid— Hquid extraction and involves the mass transfer of solutes between an aqueous phase and an immiscible organic phase. In the PUREX process, the organic phase is typically a mixture of 30% by volume tri- -butyl phosphate (solvent) and a normal paraffin hydrocarbon (diluent). The latter is typically dodecane or a high grade kerosene (20). A number of other solvent or diluent systems have been investigated, but none has proved to be a substantial improvement (21). 

These variations permit the separation of other components, if desired. Additional data on uranium, plutonium, and nitric acid distribution coefficients as a function of TBP concentration, solvent saturation, and salting strength are available (24,25). Algorithms have also been developed for the prediction of fission product distributions in the PUREX process (23). 

Historically, the Redox process was used to achieve the same purification as in the Purex process (97,129). The reagents were hexone (methyl isobutyl ketone) as the solvent, dichromate as an oxidant, and A1(N02)3 as the salting agent. The chief disadvantages of hexone are its flammability and its solubihty in water. However, because A1(N03)3 collects in the highly radioactive waste, thereby impeding the latter s further processing, the Redox process was abandoned in favor of the Purex process. 

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. 

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... 

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. 

Irradiated UO2 is dissolved in nitric acid, resulting in a dissolver solution with the approximate composition listed in Table 12.7. This is treated by the Purex process. The main steps in the conventional Purex process are shown schematically in Fig. 12.5. All existing plants listed in Table 12.8 use some variation of the Purex process. Typically, the extractant composition (percentage TBP, diluent) and the extraction equipment (i.e., pulse columns, mixer-settlers, etc.), vary from plant to plant. However, the upper concentration limit is 30% TBP to prevent a phase reversal due to the increased density of the fully loaded solvent phase. 

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. 

Schulz, W. W. Macroreticular Ion Exchange Resin Cleanup of Purex Process Solvent U.S. Atomic Energy Commission Report ARH-SA-58, Atlantic Richfield Hanford Co. Richland, Washington 1970. 

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. 

Uchiyama, G., Asakura, T., Hotoku, S., Fujine, S. 1998. The separation of neptunium and technetium in an advanced PUREX process. Solvent Extr. Ion Exch, 16 (5) 1191-1213. 

Anyun, Z., Jingxin, H., Xianye, Z., Fangding, W. 2001. Hydroxylamine derivatives in PUREX process, VI. Study on the partitioning of uranium/neptunium and uranium/ plutonium with N,N-diethylhydroxylamine in the purification cycle of uranium contactor. Solvent Extr. IonExch. 19 (6) 965-979. 

This review will exclusively deal with studies related to solvent-extraction processes (neither solid-phase precipitation nor ion-exchange chromatography) aiming at separating trivalent actinides from PUREX raffinates or spent-fuel dissolution... 

The SETFICS process (Solvent Extraction for Trivalent /-elements Intragroup Separation in CMPO-Complexant System) was initially proposed by research teams of the former Japan Nuclear Cycle Development Institute (JNC, today JAEA) to separate An(III) from PUREX raffinates. It uses a TRUEX solvent (composed of CMPO and TBP, respectively dissolved at 0.2 and 1.2 M in -dodecane) to coextract trivalent actinides and lanthanides, and a sodium nitrate concentrated solution (4 M NaN03) containing DTPA (0.05 M) to selectively strip the TPEs at pH 2 and keep the Ln(III) extracted by the TRUEX solvent (239). However, the DFs for heavy Ln(III) are rather poor. An optimized version of the SETFICS process has recently been proposed as an alternative process to extraction chromatography for the recovery of Am(III) and Cm(III) in the New Extraction System for TRU Recovery (NEXT) process. NEXT basically consists of a front-end crystallization of uranium, a simplified PUREX process using TBP for the recovery of U, Np, and Pu, and a back-end Am(III) + Cm(III) recovery step (240, 241). 

Reprocessing is based on liquid-liquid extraction for the recovery of uranium and plutonium from used nuclear fuel (PUREX process). The spent fuel is first dissolved in nitric acid. After the dissolution step and the removal of fine insoluble solids, an organic solvent composed of 30% TriButyl Phosphate (TBP) in TetraPropylene Hydrogenated (TPH) or Isopar L is used to recover both uranium and plutonium the great majority of fission products remain in the aqueous nitric acid phase. Once separated from the fission products, back-extraction combined with a reduction of Pu(I V) to Pu(III) allows plutonium to be separated from uranium these two compounds can be recycled.2... 

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