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Purex separation 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... [Pg.511]

Following decontamination of the uranium/plutonium from the fission products, the plutonium is separated from the uranium. This is done by reducing the Pu(IV) to nonextractable Pu(III), leaving uranium in the hexavalent state. In the older Purex plants, this was done using Fe2+ while the newer plants add U4+. The plutonium thus ends up in an aqueous phase while the uranium remains in the organic phase. [Pg.483]

Magnusson, D., Christiansen, B., Glatz, J.P., Malmbeck, R., Modolo, G., Serrano Purroy, D., Sorel, C. 2008. Demonstration of minor actinide separation from a genuine PUREX raffinate by TODGA/TBP and SANEX reprocessing. ATALANTE 2008 Nuclear Fuel Cycles for a Sustainable Future, May, Montpellier, France. [Pg.188]

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... [Pg.198]

In one modification of the Purex process, the plutonium is not separated from the uranium. In this version, the first cycle has only two columns instead of three. In addition to reducing the criticality risk, this modification reduces the risk of unauthorized diversion of the plutonium. [Pg.975]

The PUREX process exploits two features of U chemistry (1) the UC>22+ ion is the thermodynamically most stable form of U in aqueous solution both Pu022+ and Np022+ are easily reduced to Pu4+ and Np02+ under similar conditions (vide infra) and (2) in general, the actinide MC>22+ ions can be extracted from nitrate solutions into non-polar organic solvents [75] such as the phosphate esters, e.g. TBP. Since most other metal ions are not extracted under similar conditions, solvent extraction provides a convenient route for the purification of U and Pu from practically all other metals. Np can also be rendered extractable by manipulation of its oxidation state. Similarly, U can be separated from Pu by the selective reduction of Pu(IV) to Pu(III), rendering it inextractable into TBP/OK. [Pg.457]

It is also of interest to note that the effect of DBP on zirconium separation from thorium in the Acid-Thorex system is different than zirconium separation from uranium in the Purex system.(Figure k) The Purex data are from reference 6 and the Acid-Thorex data are from General Atomic Company pilot plant studies. The thorium probably forms a stronger DBP complex than does uranyl ion and, therefore, the amount of uncomplexed DBP available for raising the equilibrium distribution of zirconium would be less in the Acid-Thorex process. [Pg.363]

Dissolution, described in Sec. 4.4, produces an aqueous solution of uranyl nitrate, plutonium(IV) nitrate, nitric acid, small concentrations of neptunium, americium, and curium nitrates, and almost all of the nonvolatile fission products in the fuel. With fuel cooled 150 days after bumup of 33,000 MWd/MT, the fission-product concentration is around 1700 Ci/liter. The fint step in the solvent extraction portion of the Purex process is primary decontamination, in which from 99 to 99.9 percent of these fission products are separated from the uranium and plutonium. Early removal of the fission products reduces the amount of required shielding, simplifies maintenance, and facilitates later process operations by reducing solvent degradation from radiolysis. [Pg.484]

Figure 10.29 shows the principal steps in applying the Purex process to irradiated LMFBR fuel, step 7 of Fig. 10.28. The flow scheme and the compositions and locations of solvent, scrubbing, and stripping streams have been taken from the process flow sheet of a 1978 Oak Ridge report [Oil] describing a planned experimental reprocessing facility designed for 0.5 MT of uranium-plutonium fuel or 0.2 MT of uranium-plutonium-thoiium fuel per day. As that report gave process flow rates only for the uranium-plutonium-thorium fuel. Fig. 10.29 does not give flow rates for the uranium-plutonium fuel of present interest. This flow sheet shows the codecontamination step, in which flssion products are separated from uranium and plutonium the partitioning step, which produces an aqueous stream of partially decontaminated... Figure 10.29 shows the principal steps in applying the Purex process to irradiated LMFBR fuel, step 7 of Fig. 10.28. The flow scheme and the compositions and locations of solvent, scrubbing, and stripping streams have been taken from the process flow sheet of a 1978 Oak Ridge report [Oil] describing a planned experimental reprocessing facility designed for 0.5 MT of uranium-plutonium fuel or 0.2 MT of uranium-plutonium-thoiium fuel per day. As that report gave process flow rates only for the uranium-plutonium-thorium fuel. Fig. 10.29 does not give flow rates for the uranium-plutonium fuel of present interest. This flow sheet shows the codecontamination step, in which flssion products are separated from uranium and plutonium the partitioning step, which produces an aqueous stream of partially decontaminated...
Hanfoid [D3]. Nitrite concentration in feed to the HA column of a standard Purex plant was adjusted to route most of the neptunium in inadiated natural uranium into the extract from the HS scrubbing column. Sufficient ferrous sulfamate was used in the partitioning column to reduce neptunium to Np(IV), which followed uranium. This neptunium was separated from uranium by fractional extraction with TBP in the second uranium cycle. The dilute neptunium product was recycled to HA column feed, to build up its concentration. Periodically, irradiated uranium feed was replaced by unirradiated uranium, which flushed plutonium and fission products from the system. The impure neptunium remaining was concentrated and purified by solvent extraction and ion exchange. [Pg.545]

In the Purex process for the separation of nranium from plutonium, complexes of type M02(N03)2 2TBP org , where M = U or Pu, and org = kerosine, are first formed by dissolving the spent nuclear fuel in HNO3 and kerosine. A reducing agent is then added to convert Pu to Pu which forms a weaker complex which can be separated from the complex. [Pg.1096]

Some part of the spent fuel of atomic reactors is reprocessed separating uranium, plutonium, and the fission products, in order to produce new fissionable fuel or to collect some part of the valuable fission products. While several reprocessing methods have been proposed, the Purex process is the most widely used all over the world. The process uses 30% tributyl phosphate, TBP, as extractant in dodecane or kerosene solvent that is used to decrease the viscosity and the density of the liquid. The mixture is easily separated from water. The spent fuel is dissolved in concentrated nitric acid and the aqueous solution is mixed with the organic extractant. U and Pu present in the aqueous phase in the forms U02 and Pu are extracted to the organic phase, the fission products remain in the aqueous solution. After reduction of Pu by chemical or electrochemical method, Pu goes back to the aqueous phase, while the uranium remains in the organic phase (Benndict et al., 1981 Choppin et al. 1995 Katsumura 2004). [Pg.1315]

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... [Pg.411]

Once plutonium and uranium are coextracted and codecontaminated, plutonium is separated from uranium in the partitioning contactor by reduction to Pu(III) with a reduc-tant. Over the years, a number of plutonium reductants have been proposed. The most widely used reductant to partition plutonium from uranium in the PUREX process was (Fe(S03NH2)2) other alternates were proposed such as hydrazine-stabilized ferrous nitrate or uranous nitrate, and hydroxylamine salts. [Pg.413]

Each of these elements may be used for production of nuclear fuel or other purposes. The recovery efficiency for uranium is reported as 99.87% and for plutonium 99.36%-99.51% (NEA 2012). The extended PUREX includes separation of neptunium and technetium as well as recovery of americium and curium that are also separated from each other by additional extraction stages as given in detail in the flowsheet (NEA 2012). The advanced UREX-i-3 process generates six streams after separation uranium for re-enrichment Pu-U-Np for mixed oxide fuel c for managed disposal Am-Cm to be used as burnable poisons and for transmutation high-heat-generating products (Cs and Sr) and a composite vitrified waste with all other fission products. Some fuel types may require preliminary steps like grinding to enable their dissolution. [Pg.104]

The fuel cycle is shown in Fig. 10-5. Slurry is removed from the reactor at the rate required to maintain a specified poison level. The fuel is separated from the D2O, cooled for 20 days while the neptunium decays, and partially decontaminated in a Purex plant. Part of the fuel is then sent directly to the reactor feed-preparation equipment. Plutonium is separated from the remainder of the uranium and added to the reactor feed. The uranium is completely decontaminated, stored for 100 days until the U " decays, and sent to the diffusion plant. A 30-day reserve of reactor feed is kept on hand. [Pg.531]

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. [Pg.202]


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