Extraction with TBP

In solvent extraction with TBP, the fluoride complexes form solvated complexes with TBP. The stabilities of such complexes depend on the acidity/concentration of hydrofluoric... 

Raffinate phase cone., g l Fig. 10.9 Isotherms for citric acid extraction with TBP. 

The spent fuel element is still mainly UO2 and is dissolved in aqueous nitric acid, which is oxidizing enough to take the uranium to the VI oxidation state as UC>22+(aq) and Pu to Pu4+(aq) (the uranyl ion U022+ can be regarded as hydrolyzed U6+ see Section 13.6). Treatment of the solution of uranyl and plutonium(IV) nitrates with either an iron(II) salt or SO2 will reduce all the Pu to Pu3+(aq), which is not extractable with TBP, but will leave the uranium(VI) untouched (see Exercise 15.5). The solution is then equilibrated with TBP (which is immiscible with water) or TBP in an alkane solvent. The U022+ forms a neutral complex containing both TBP and the nitrate ions, which axe present in large excess ... 

The Purex process, ie, plutonium uranium reduction extraction, employs an organic phase consisting of 30 wt % TBP dissolved in a kerosene-type diluent. Purification and separation of U and Pu is achieved because of the extractability of U02+2 and Pu(IV) nitrates by TBP and the relative inextractability of Pu(III) and most fission product nitrates. Plutonium nitrate and U02(N03)2 are extracted into the organic phase by the formation of compounds, eg, Pu(N03)4 -2TBP. The plutonium is reduced to Pu(III) by treatment with ferrous sulfamate, hydrazine, or hydroxylamine and is transferred to the aqueous phase U remains in the organic phase. Further purification is achieved by oxidation of Pu(III) to Pu(IV) and re-extraction with TBP. The plutonium is transferred to an aqueous product. Plutonium recovery from the Purex process is ca 99.9 wt % (128). Decontamination factors are 106 — 10s (97,126,129). A flow sheet of the Purex process is shown in Figure 7. 

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. 

Advantage may be taken of these conditions by use of solvent extraction techniques. It is known (25-30) that Am extraction with TBP ( tributyl phosphate) or DBBP (dibutyl butyl phos-phonate) is enhanced by high nitrate salt concentrations in the aqueous phase, particularly at HN03 concentrations below 1.0 N. 

Co-location of the TBP and DBBP extraction processes in the same facility led inevitably to cross contamination of extractants. This problem was of greater consequence to the PRF system where small concentrations of DBBP in the TBP extractant interfered with plutonium stripping. No specific system malfunctions directly attributable to the presence of TBP in the DBBP solvent were identified. However, dilution of the DBBP extractant with TBP reduces its efficiency as an americium extractant. 

The efficient removal of actinides from the Na2C03 scrub waste solution presents several problems. Acidification of the carbonate solution with excess HNO3 followed by extraction with TBP (or, preferably, DHDECMP) (3) results in the rapid build-up of acidic degradation products (HDBP and H2MBP in the case of TBP) which prevent efficient back extraction. In addition, acidification of Na2C03 scrub waste results in the precipitation of actinide-DBP and -MBP complexes which are difficult to dissolve... 

Separation of hafnium from zirconium. In a zirconium-hafnium separation process developed by Eldorado Nuclear, a mixture of sodium zir-conate and hafnate can be obtained by fusing zircon sand with NaOH and dissolving the product in water. Acidification with nitric acid then gives aqueous zirconyl (ZrO +, or hydrolyzed Zr + ) and hafnyl (HfO " ") nitrates. Extraction with TBP then gives an extract containing mainly [Hf0(N03)2(TBP) ], but most of the ZrO " " remains in the aqueous phase and can be recovered on evaporation as essentially Hf-free Zr0(N03)2(s). The effectiveness of this process can be ascribed to compounding of factors... 

Larger amounts of some metals can be extracted as nitrates [U, Th, Ce(IV), Y, Se] or oxides (Os, Ru). Matrix yttrium or scandium are extracted with TBP from 12-13 M HNO3 [70],... 

The iron(III)-thiocyanate complexes can be extracted with oxygen-containing solvents such as ethers, higher alcohols, esters, and ketones. Depending on the solvent used, different species are extracted. The 1 4 Fe SCN complex is extracted with diethyl ether, while the 1 3 complex is extracted with TBP. 

Bromide complexes of Pt and other metals of this group are extracted from acid media with TBP, MIBK, amyl acetate [12] or TOPO [6]. Pt and Pd can be separated from Rh and Ir by extraction (with TBP) from iodide solution [13]. The same method has been used for separating platinum from copper [14]. 

Tellurium(IV) reacts with thiourea, S=C(NH2)2, (whose concentration in the final solution should be -10%) in 1 M H2SO4, HNO3, or H3PO4 to form a yellow cationic complex suitable for spectrophotometry [36]. The absorption maximum of the complex occurs at 320 nm. The cationic complex of tellurium with thiourea can be extracted with TBP as the ion-associate with thiocyanate ions. The following thiourea derivatives have been proposed for the spectrophotometric determination of tellurium l,4diphenylthiosemicarbazide [37], diphenyl-thiocarbazide, dinaphthylthiourea, and tetramethylthiourea [38]. 

Chemically complexing extractants can be much more selective than conventional solvents and therefore can reduce the amount of coextracled wmsr. On a solvnat-free basis, the extract in equilibrium with a 6.6% w/w aqueous acetic acid solution contains about 84% acetic acid for extraction with Alamine 336-DIBK, The selectivity For acetic acid over water is lower for extraction with TBP,5 but is still substantially higher than with conventions] solvents such as acetates or ketoses. Coextracled water can be removed in an extractive distillation column located before a solvent regeneration column, as shown in Fig. 13.2-3, in a henvy-solvenl analogue of the process shown in Fig. 15.1-2. 

Table 4.8 Material balance for zirconium-hafnium separation by fractional extraction with TBP+...

Figure 5.22 Purification of uranium ore concentrates by solvent extraction with TBP.

Uranium is recovered from the sulfate and phosphate filtrate by anion exchange (Chap. 5). Thorium and rare earths in the hydroxide precipitate are dissolved in nitric acid and separated by solvent extraction with TBP (Sec. 8.7). 

Attempts to separate thorium and uranium from sulfuric acid solution of monazite by solvent extraction with TBP were unsuccessful because distribution coefficients of uranium and thorium from monazite solutions were too low, as these elements are complexed by phosphate ion. Development of extractants with higher distribution coefficients for these metals has made solvent extraction a practical process for recovering uranium and thorium from monazite sulfate solutions and from sulfuric acid solutions of other thorium ores. This section describes processes tested on a pilot-plant scale by Oak Ridge National Laboratory [C5]. 

Solvent extraction with TBP has become the standard procedure for purifying thorium, just as for uranium. Processes used in different countries differ, however, in details such as the solvent used to dilute TBP, its concentration, and the means used to strip thorium and coextracted uranium from TBP. Table 6.20 summarizes the main features of processes used for purification of thorium on an industrial scale in the principal thorium-producing countries. Wylie [W5] gives more detail on early pilot-plant thorium-purification runs. Most of the published U.S. work on thorium purification on an industrial scale deals with irradiated thorium rather than natural this will be described under the Thorex process, in Sec. 5 of Chap. 10. 

Figure 6.8 Thorium purification by solvent extraction with TBP. Circles, relative flow ------------...

Figure 7.8 Pilot plant of French Atomic Energy Commission for separation of zirconium from hafnium by solvent extraction with TBP. Solid line, aqueous broken Une, organic.

The selective extraction of plutonium from uranium or fission products depends on proper adjustment of the valence state of plutonium relative to the other ions from which it is to be separated. For instance, in decontaminating plutonium by extraction with TBP, plutonium must be oxidized to the tetravalent state, without bringing cerium into the tetravalent, ceric state. Again, to separate plutonium from uranium and the fission products in the tributyl phosphate extraction process, plutonium must be trivalent and uranium hexavalent. 

Trivalent americium forms relatively unstable complexes with Cl and NOs and more stable complexes with the thiocyanate ion CNS. These americium complexes are more stable than those of the corresponding lanthanide compounds, so that americium can be separated from trivalent lanthanides by anion exchange with concentrated solutions of liQ, liNOs, or NH4CNS. Trivalent americium can be extracted with TBP from a concentrated nitrate solution. It can also be extracted with TBP from a molten LINO3 -KNOs eutectic at 150°C, with much higher distribution coefficients than in extraction from aqueous solutions. Americium is more readily extracted by this process than is trivalent curium [K2]. 

Culler and Blanco [CIS] have summarized other aqueous processes that have been studied for processing power reactor fuels not readily handled by the standard Purex or Thorex processes. Many of these require reagents other than nitric acid to dissolve either the cladding or the fuel, but finally use solvent extraction with TBP to separate and purify fissile materials. Details of these other processes are given in references cited by Culler and Blanco [Cl8]. 

Plutonium in oxide fuel dissolves as a mixture of tetravalent and hexavalent plutonyl nitrates, both of which are extractable with TBP. Neptunium dissolves as a mixture of inextractable pentavalent and extractable hexavalent nitrates. 

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. 

A technique for the separation of pertcchnetalc from mixed fission products by solvent extraction with TBP was described. The extraction was almost quantitative from a sulphuric acid solution. Sodium fluoride was used to provide the zirconium-niobium decontamination and a cation exchange column ensured the decontamination from metallic ions. Tc yields of 92 % were obtained [123. ... 

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