Purex solutions, concentration

The plutonium extracted by the Purex process usually has been in the form of a concentrated nitrate solution or symp, which must be converted to anhydrous PuF [13842-83-6] or PuF, which are charge materials for metal production. The nitrate solution is sufficientiy pure for the processing to be conducted in gloveboxes without P- or y-shielding (130). The Pu is first precipitated as plutonium(IV) peroxide [12412-68-9], plutonium(Ill) oxalate [56609-10-0], plutonium(IV) oxalate [13278-81-4], or plutonium(Ill) fluoride. These precipitates are converted to anhydrous PuF or PuF. The precipitation process used depends on numerous factors, eg, derived purity of product, safety considerations, ease of recovering wastes, and required process equipment. The peroxide precipitation yields the purest product and generally is the preferred route (131). The peroxide precipitate is converted to PuF by HF—O2 gas or to PuF by HF—H2 gas (31,132). 

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. 

As described in Figure 3.7, TRU separation is performed by implementing the DIDPA process on pretreated PUREX raffinates. A front-end denitration step by formic acid is thus required to reduce the nitric acid concentration of the feed down to 0.5 M to allow the TRU elements to be extracted by the cation exchanger di-fvo-dccyl-phosphoric acid (DIDPA). This preliminary step, however, induces the precipitation of Mo and Zr (and thus the potential carrying of Pu), which requires filtration steps. The TRU and Ln(III) elements are coextracted by a solvent composed of the dimerized DIDPA and TBP, dissolved at 0.5 and 0.1 M, respectively, in n-dodecane. The An(III) + Ln(III) fraction is back-extracted into a concentrated 4 M nitric acid solution, whereas Np and Pu are selectively stripped by oxalic acid. 

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... 

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). 

In addition to the aqueous raffinates from the solvent extraction cycles of the Purex process, an actinide bearing waste stream will arise from the washing of the TBP/OK solvent prior to its recycle to the first cycle. These wastes will typically contain actinides in a mixed NajCOs/NaNOs solution which also contains HjMBP and HDBP. The uranium present will form soluble U complexes with carbonate, as discussed in Section 65.2.2.l(i). Carbonate complexation of Pu also leads to solubility in alkaline solutions and in Na2C03 media precipitation did not occur below pH 11.4, although precipitates did form on reduction to Pu One Pu" species precipitated from carbonate media has been identified as Pu(0H)3-Pu2(C03)3 H20. In 2M Na2C03 media, Np is oxidized by air to Np above pH 11.7 while Np either precipitates or is reduced above pH 13. The potential of the Am /Am " couple, in common with those of other actinides, becomes more cathodic with increasing carbonate concentration. In the total bicarbonate plus carbonate concentration range 1.2-2.3 M all the americium oxidation states from (III) to (VI)... 

The redox methods are well known for the purification and concentration of plutonium from the Purex process solutions. 

Pu(IV) reduction rate were studied Pig.8 shows that at large excess of U and low concentration of Pu, no effect of acid concentration on Pu(IT) reduction rate could he observed. After a period of electrolysis of less than 30 seconds, nearly all of the Pu(IV) could be converted to Pu(III). This fact corresponds to the change of the potential of the electrolyte solution with time, which drops very rapidly after the start of the electrolysis. The effect of 0-concentration on the reduction rate of Pu(IV) is shown in Fig.9, from which it is clearly shown that the reduction rate of Pu(IT) depends very much on the amount of U relative to that of Pu in the electrolyte solution. The upper two curves showed that if the weight ratio of U/Pu is near or more than one, the reduction rate of Pu(IV) could be greatly accelerated. This fact indicates clearly that here U(IV) plays an important role in the reduction of Pu(IV). On the other hand, if the U-oontent in the solution is small compared to that of Pu, the rate of reduction of Pu(IV) is determined chiefly by the electrolytic reduction of Pu(IV) itself which is rather slow. This fact should be borne in mind in designing electrolytic reduction equipments in the purex process. 

Cobalt (II) acetate tetrahydrate, Cobalt (II) nitrate hexahydrate, cobalt chloride, sodium aluminate (Na2O.Al2O3.3H2O), boric acid (H3BO3) and tetraethylammonium hydroxide (Aldrich), acetic acid purex and cyclohexane for analysis (SDS) were used as received. Ludox HS-40 colloidal silica solution was obtained from Dupont. Zeolites BEA 15 and BEA 27 were synthesized in the presence of tetraethylammonium hydroxide (TEAOH) according to the procedure described by Wadlinger and al. [14]. Dealuminated BEA 1100 was obtained by treating BEA 15 with concentrated nitric acid [15],... 

Simplified flow sheets are given in Figs. 2 and 3 for two of the basic types of waste that will be encountered. The Purex type of waste is the simplest of all of the wastes to process, being a nitric acid solution of fission products, corrosion products, and a small amount of other impurities. Advantage is taken of the fact that one can evaporate this waste, and thereby achieve a greatly increased concentration of material, before actual chemical separation of the constituents is started. 

The principle of the Purex process, now commonly used for processing irradiated uranium by solvent extraction, is illustrated in Fig. 1.18. The solvent used in this process is a solution of tributyl phosphate (TBP) in a high-boiling hydrocarbon, frequently n-dodecane or a mixture of similar hydrocarbons. TBP forms complexes with uranyl nitrate [U0i(N03)2] and tetravalent plutonium nitrate [Pu(N03)4] whose concentration in the hydrocarbon phase is higher than in an aqueous solution of nitric acid in equilibrium with the hydrocarbon phase. On the other hand, TBP complexes of most fission products and trivalent plutonium nitrate have lower concentrations in the hydrocarbon phase than in the aqueous phase in equilibrium. 

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