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Purex actinide separation from

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]

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]

However, the conditions are often far from those of industrial situations. In order to better simulate solvent degradation during the PUREX process, a test loop was created in the 1990s in a CEA laboratory (Fontenay-aux-Roses, France), with the EDIT loop (Extraction Desextraction Irradiation Traitement) (21, 22). The laboratory simulation of industrial conditions consisted of a succession of representative physical and chemical treatments after the irradiation of the solvent (i.e., alkali and acid treatments, distillation). Indeed, these treatments can modify the final solvent composition because of the elimination of some compounds or the occurrence of secondary reactions. A few years later, the MARCEL (Module Avance de Radiolyse dans les Cycles d Extraction Lavage) test loop was built at Marcoule to follow the regeneration efficiencies of degraded solvents involved in actinide separation processes (4, 5). [Pg.439]

The CTH actinide separation process was developed as a possible means to reduce the expected long term dose to man from a geologic repository containing solidified radioactive waste from the reprocessing of spent nuclear fuel The distribution data for the elements present in significant amounts in the high level liquid waste (HLLW) from a Purex plant, the general principles and the flowsheet have been described in detail elsewhere A... [Pg.198]

During the aging of the HLLW solution from a Purex plant insoluble precipitates are known to form, which could endanger the operation of any actinide recovery process and increase actinide losses. It is therefore believed that an actinide separation process must use the HLLW solution as soon as possible after it is... [Pg.210]

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]

Origins. Most of the radioactive waste at SRP originates in the two separations plants, although some waste is produced in the reactor areas, laboratories, and peripheral installations. The principal processes used in the separations plants have been the Purex and the HM processes, but others have been used to process a variety of fuel and target elements. The Purex process recovers and purifies uranium and plutonium from neutron-irradiated natural uranium. The HM process recovers enriched uranium from uranium—aluminum alloys used as fuel in SRP reactors. Other processes that have been used include recovery of and thorium (from neutron-irradiated thorium), recovery of Np and Pu, separation of higher actinide elements from irradiated plutonium, and recovery of enriched uranium from stainless-steel-clad fuel elements from power reactors. Each of these processes produces a characteristic waste. [Pg.10]

For future advanced nuclear systems, minor actinides are considered more as a resource to be recycled and transmuted than to be disposed of directly into a nuclear repository. A key feature of advanced fuel cycles technologies would be to separate M A and ultimately americium from curium. Several countries are investigating the separation of MA from a PUREX/COEX based process raffinate or a modified PUREX process raffinate using new extractant molecules with two potential options for actinide separations ... [Pg.437]

In tlie PUREX process, the spent fuel and blanket materials are dissolved in nitric acid to form nitrates of plutonium and uranium. These are separated chemically from the other fission products, including the highly radioactive actinides, and then the two nitrates are separated into tv/o streams of partially purified plutonium and uranium. Additional processing will yield whatever purity of the two elements is desired. The process yields purified plutonium, purified uranium, and high-level wastes. See also Radioactive Wastes in the entry1 on Nuclear Power Technology. Because of the yield of purified plutonium, the PUREX process is most undesirable from a nuclear weapons proliferation standpoint,... [Pg.1647]

Mathur, J.N. Murali, M.S. Iyer, R.H. Ramanujam, A. Dhami, P.S. Gopalkrishnan, V. Rao, M.K. Badheka, L.P. Baneiji, A. Extraction chromatographic separation of minor actinides from purex high-level wastes using CMPO, Nucl. Technol. 109 (1995) 216-225. [Pg.113]

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... [Pg.130]

The ZEALEX Process Researchers from KRI have shown that the zirconium salt of dibutyl phosphoric acid (ZS-HDBP) was soluble in Isopar-L in the presence of 30% TBP. This super PUREX solvent, known as ZEALEX, extracts actinides (Np-Am) together with lanthanides and other fission products, such as Ba, Cs, Fe, Mo, and Sr from nitric acid solutions. The extraction yields depend on both the molar ratio between Zr and HDBP in the 30% TBP/Isopar-L mixture and the concentration of HN03 (232). Trivalent transplutonium and lanthanide elements can be stripped together from the loaded ZEALEX solvent by a complexing solution, mixing ammonium carbonate, (NH4)2C03, and ethylenediamine-N.N.N. N -tetraacetic acid (EDTA). An optimized version of the process should allow the separation of... [Pg.165]

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). [Pg.167]

Di(2-ethylhexyl) phosphoric acid (HDEHP) is an extractant molecule used for An(III)/Ln(III) separation. Used in TALSPEAK-type processes in a mixture with TBP, or in the DIAMEX-SANEX process in a mixture with a malonamide (154-157), it has also been proposed, in a mixture with TBP, to remove strontium from PUREX acid waste solution in the Hanford B plant (158). Therefore, numerous studies have focussed on the radiolytic degradation of HDEHP and its effects on the extraction of Sr(II), lanthanides(III), and actinides(III) (10, 158-163). [Pg.452]

Liljenzin, J. 0. Hagstrom, I. Persson, G. Svantesson, I., "Separation of Actinides from Purex Waste", Proc. ISEC 80,... [Pg.216]

The control of the actinide metal ion valence state plays a pivotal role in the separation and purification of uranium and plutonium during the processing of spent nuclear fuel. Most commercial plants use the plutonium-uranium reduction extraction process (PUREX) [58], wherein spent fuel rods are initially dissolved in nitric acid. The dissolved U and Pu are subsequently extracted from the nitric solution into a non-aqueous phase of tributyl phosphate (TBP) dissolved in an inert hydrocarbon diluent such as dodecane or odourless kerosene (OK). The organic phase is then subjected to solvent extraction techniques to partition the U from the Pu, the extractability of the ions into the TBP/OK phase being strongly dependent upon the valence state of the actinide in question. [Pg.453]

In summary, potential improvements could be made to the PUREX process in the following areas (1) separation of Np from U and Pu prior to the U/Pu split and (2) in the requirement to use a large excess of U(IV) reductant to reduce Pu(IV) to Pu(III). The majority of published work on the applications of photo catalysis in actinide redox chemistry has concentrated on solving the first of these difficulties through Np valence control. A smaller volume of literature exists on the applications of photocatalysis in valence state control of U and the radioactive d block metal, technetium. This section will review both of these aspects. [Pg.461]


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See also in sourсe #XX -- [ Pg.346 ]




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