Americium recovery

Molten salt extraction residues are processed to recover plutonium by an aqueous precipitation process. The residues are dissolved in dilute HC1, the actinides are precipitated with potassium carbonate, and the precipitate redissolved in nitric acid (7M) to convert from a chloride to a nitrate system. The plutonium is then recovered from the 7M HNO3 by anion exchange and the effluent sent to waste or americium recovery. We are studying actinide (III) carbonate chemistry and looking at new... 

Plutonium and americium recovery from MSE Salt Stripping Product. 

Molten Salt Extraction Salts Plutonium and Americium Recovery. We have demonstrated the ability to successfully strip the plutonium from the MSE salts. The resulting metal product now contains as much as 10% americium and as a result cannot be fed directly into the metal processing sequence. To use the plutonium we must remove the americium. 

Rainey, R. H. Development of the Amex Process for Americium Recovery. 

Most of the studies focused on the influence of CMPO radiolysis on two important steps of the TRUEX process for americium recovery extraction and stripping. The effect of radiolysis on CMPO-diluent solutions was extremely different under these two process conditions (see Table 8.5), with a major increase in Z)Am values at pH 2, but a moderate decrease for 2-3 mol L 1 HN03. 

Additional work in progress includes optimization of parameters affecting the oxalate precipitation step this includes determination of the chloride concentration required to solubilize lead the oxalate ion concentration required for maximum americium recovery with minimum impurity precipitation precipitate aging and hydrogen ion concentrations that will minimize americium solubility yet maximize impurity solubilization. 

The aqueous waste from the CA Column (CAW) contains virtually all of the americium present in the feed. Table I shows the typical composition of the CAW stream. The exact composition of the stream depends on the composition of the CAF which is highly variable in a plutonium scrap processing plant. The CAW stream is the feed to the americium recovery solvent extraction system. 

Americium recovery development at Rocky Flats comprises work to improve the existing process as well as to introduce new methods, especially those that can partition americium and aluminum since the present cation exchange process can not do this. 

The problem with Dowex 50W-X8 is the slow recovery of plutonium. Again, it is suggested that this is because of the oxidation of Pu(lll) to Pu(IV) and formation of the hexanitrato plutonium complex Pu(lV) is more tightly held than Pu(lll) and the bulky hexanitrato complex has difficulty diffusing through the polymer matrix. We cannot explain why plutonium recovery is poorer from the Na-K-Mg residue than from the Ca-K-Mg residue whereas americium recovery is better from the Na-K-Mg system and worse from the Ca-K-Mg system. The mean + standard deviation for the duplicate tests suggests the effect is real. 

We have no explanation for the slower elution of americium from macroporous resin. Americium recovery from macroporous resin should be no worse than from gel-type resin one would expect it to be better. 

Figure 1 shows a simplified flow sheet for plutonium-239 recovery operations at Rocky Flats. Impure plutonium metal is sent through a pyrochemical process, called molten salt extraction (MSE), to remove the elemental impurity americium. The product plutonium metal, if it meets plant purity requirements, is sent to the foundry. Metal that does not meet foundry requirements is processed further, either through an aqueous process using ion exchange, or through a pyrochemical electrorefining process. The waste chloride salt from MSE is... 

The NaCl-KCl eutectic is used when the pregnant extraction salt is to be processed by aqueous recovery (this is the salt currently used at Rocky Flats because calcium follows americium in the present aqueous recovery process). The NaCl-CaCl system is used when the salt is processed by pyrochemical means to recover the americium and residual plutonium. When the pyrochemical recovery technique is used, the NaCl-CaCl2-MgCl2 salt is contacted with liquid calcium metal at approximately 850°C in a batch extractor. The calcium reduces A111CI3,... 

PUCI3, and MgCl2 to form a 50/50 mole % NaCl-CaCl salt phase and a molten Am-Pu-Mg-Ca alloy which is immiscible in the above salt(lO). After cooling, the metal phase is cleaved away from the salt phase and the salt phase is analyzed. Little, if any, Am or Pu remains in the salt phase and the salt residues can be discarded to waste. Metal recovery begins by evaporating magnesium and calcium from the residual metal button at about 800°C in vacuum. The americium can then be distilled away from the plutonium in a vacuum still operated at 1200°C, using yttria ceramic vessels to contain the molten metal fraction. The bottoms fraction contains the plutonium which is recycled back into the main plutonium stream. 

Molten Salt Extraction (MSE) - Salt Recovery. The salt residue from the americium extraction process is made up of NaCl, KC1, MgCl2, PuCl3, and AmCl3. A typical residue weighs approximately 2.0 kg and contains 200 g plutonium, 10-20 g americium, 50 g MgCl3 with the remaining equimolar NaCl-KCl. 

Zhu, Y., Song, C. 1992. Recovery of neptunium, plutonium and americium from highly active waste. Tri-alkyl phosphine oxide extraction. In Transuranium Elements A Half Century. Morss, L.R. Fuger, J., Eds. ACS, Washington, DC, pp. 318-330. 

Madic, C. Kertesz, C. Sontag, R. Koehly, G. Application of extraction chromatography to the recovery of neptunium, plutonium and americium from an industrial waste, Sep. Sci. Technol. 15 (1980) 745-762. 

Leonard, R. A., G. F. Vandegrift, D. G. Kalina, et al. 1985. The Extraction and Recovery of Plutonium and Americium from Nitric Acid Waste Solutions by the TRUEX Process— Continuing Development Studies. Argonne National Laboratory Report ANL-85-45, Argonne, IL. 

Sekine et al. [176] studied the liquid-liquid extraction separation and sequential determination of plutonium and americium in soils by alpha-spectrometry. The chemical recovery of plutonium from standard soil samples was 51-99% (average 81%) of the analytical level and for americium 60-70% of the analytical level. 

Sill et al. [26] have discussed a spectrometric method for the determination of americium and other alpha-emitting nuclides, including curium and californium, in potassium fluoride-pyrosulfate extracts of soils. Sekine [27] used a-spectrometry to determine americium in soils with a chemical recovery of 60-70%. Joshi [28] and Livens et al. [29] have discussed methods for the determination of241 americium in soils. 

Sekine et al. [27] used a-spectrometry to determine plutonium (and americium) in soil. The chemical recovery of plutonium was 51-99% and averaged 81%, while for americium the recovery was 60-70%. The method is coupled with the liquid-liquid extraction stage, taking about two days less than the ion exchange method a complete analysis takes about one week. 

This collection of the state-of-the-art papers emphasizes the continuing importance of industrial-scale production, separation, and recovery of transplutonium elements. Americium (At. No. 95) and curium (At. No. 96) were first isolated in weighable amounts during and immediately after World War II. Berkelium and californium were isolated in 1958 and einsteinium in 1961. These five man-made elements, in each case, subsequently became available in increasing quantities. 

In the case of the treatment of waste solutions for the recovery of americium 241, the distribution of tasks was based essentially on storage volume requirements. The first purification cycles were performed in Pollux which has capacity for storing large volumes of solutions final purification of americium 241 was performed in Irene. 

An initial experiment involving the treatment of small irradiated Pu/Al targets for the production of americium 243 and curium 244 was carried out in France in 1968 (2). The chemical process was based essentially on the use of a system comparable to the Talspeak system. After plutonium extraction by a 0.08 M trilaurylammonium nitrate solution in dodecane containing 3 vol % 2-octanol, the actinides (americium, curium) were coextracted with a fraction of the lanthanides by a 0.25 M HDEHP -dodecane solvent from an aqueous solution previously neutralized by A1(N0 ) x(0H)x and adjusted to 0.04 M DTPA. The actinides were selectively stripped by placing the organic phase in contact with an aqueous solution of the composition 3 M LiN0 -0.05 M DTPA. While this experiment achieved the recovery of 150 mg of americium 243 and 15 mg of curium 244 with good yields, the process presented a drawback due to the slow extraction of Al(III) which saturates the HDEHP. This process was therefore abandoned. 

Treatment of waste solution. The objective of the treatment of wastes of the type described in Table II is twofold first, the elimination of alpha-emitters from the waste, and secondly the recovery of americium 241 which can be utilized directly. Since all the waste solutions contain nitric acid, the only parameters which can conveniently be defined are ... 

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