Salting agent

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. 

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. 

The organic phase solvent, diluent) is assumed to be inert (shaded area). The aqueous phase (nonshaded area) is unspecified, but may contain various salting agents, not considered here. 

Uses. Chemical intermediate for magnesium salts component of pharmaceuticals, cosmetics, dentifrices, free-running table salt agent in heat insulation and refractory applications... 

For the fluorometric method, uranium is concentrated by co-precipitation with aluminum phosphate, dissolved in diluted nitric acid containing magnesium nitrate as a salting agent, and the co-precipitated uranium is extracted into ethyl acetate and dried. The uranium is dissolved in nitric acid, sodium fluoride flux is added, and the samples fused over a heat source (EPA 1980). 

Therefore, if the feed acidity is decreased, the reduction rate should be enhanced and lower Pu losses should be achieved. This was confirmed by experiment where the loss was decreased to 2% Pu at a feed acidity of 0.8 mol/L. The calculated effect of feed acidity on Th and Pu losses from Contactor III is shown in Figure 6. The Pu oxidation model was used in deriving these results. An acidity of less than 0.6 mol/L would apparently be required to limit Pu losses to v 0.1%. At the same time, Th losses to the Pu product stream would be over 10%. Experimentally thorium loss was reduced to acceptable levels by adding a salting agent (NaNC ) to the 3AF and 3AS streams, but this solution unfortunately eliminates one of the major advantages of using HAN as reductant, viz, a decrease in the amount of solid wastes which must be treated. 

The thrust of the experimental program at ICPP was to find a separation procedure that would separate plutonium, americium, and curium from high-level first-cycle raffinate (see Table I) and leave behind the cladding elements, salting agents, and the bulk of the fission products. Fission-product lanthanides, because of their similar valence and ionic size, would be expected to follow americium in nearly any simple separation scheme. Americium and curium are present in ICPP waste as trivalent ions while plutonium is most likely present as both Pu(IV) and Pu(VI). Any separation scheme must be applicable to all these ionic actinide species. 

The TBP extraction process. The extraction of trivalent actinides by TBP may succeed, as well known, at low acidity conditions, provided a sufficient nitrate salt content is reached in the feed solution (13,14). However, the large amount of salting agents to be added to the HAW (tons of Na and Al nitrate per ton of spent... 

WA, W0 weight fractions of solute A and juexiraciabie dissolved species D (e.g., salting agent D) in the raffinate phase... 

As illustrated in Fig. 4.15, different equilibrium lines can exist for the extracting and scrubbing sections, as might occur if the scrub solution contains a different salting agent... 

The solvent used in the Redox process was hexone, methyl isobutyl ketone, an extractant already in use for purifying uranium ore concentrates (Chap. 5)., Hexone is immiscible with water and will extract uranyl nitrate and plutonyl nitrate selectively from fission-product nitrates if the aqueous solution has a sufficiently high nitrate ion concentration. In the Redox process, aluminum nitrate was used as salting agent because high concentrations of nitric acid would decompose the hexone solvent. 

A modification of the Redox process, the U-hexone process, was used at the Idaho Chemical Processing Plant of the U.S. AEC, to recover highly enriched uranium from U-A1 alloy fuel elements irradiated in the Materials Testing Reactor. The aluminum nitrate needed as salting agent was provided when the fuel was dissolved in nitric acid. The plutonium content of the fuel was too low to warrant recovery. Plutonium was made trivalent and inextractable before solvent extraction and thus routed to the aqueous high-level waste. 

The Purex process has four significant advantages over the Redox process (1) Waste volumes can be made much lower, as the nitric acid used as salting agent can be removed by evaporation. (2) The solvent, TBP, is less volatile and less flammable than hexone. (3) TBP is more stable against attack by nitric acid. (4) Operating costs are lower. 

The second separation plant used tributyl phosphate in kerosene (TBP/OK) as the solvent and mixer settlers with simple interstage lifts between cycles. The choice of this solvent enabled the further elimination of salting agents used in the later Butex cycles of the first plant with consequent benefit to effluent disposal. This plant had five times the output of the first yet was accommodated in a low, smaller building. Despite these marked changes between the two plants, many of the principles found to be satisfactory in the first plant were retained in the second, a tribute to the firm foundations laid, in a short time, by the original designers. 

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