Plutonium recovery from

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. 

Large-scale plutonium recovery/processing facilities originated at Los Alamos and Hanford as part of the Manhattan Project in 1943. Hanford Operations separated plutonium from irradiated reactor fuel, whereas Los Alamos purified plutonium, as well as recovered the plutonium from scrap and residues. In the 1950 s, similar processing facilities were constructed at Rocky Flats and Savannah River. 

The original plutonium recovery and purification processes were adopted from Los Alamos processes in 1950. The processes at Rocky Flats are still similar today, in many respects, to the Los Alamos processes. 

Mullins, L.J. Christensen, D.C. Babcock, B.R. "Fused Salt Processing of Impure Plutonium Dioxide to High Purity Metal", Los Alamos Nat. Lab. Report LA-9154-MS also Symposium on Actinide Recovery from Waste and Low Grade Sources, ACS, New York City August 23-28, 1981 (in press). 

Plutonium and americium recovery from MSE Salt Stripping Product. 

This latter situation affords a good method for separating uranium from plutonium. Hydroxylammonium formate (HAF) and hydrazium formate (NHF) were added to the formic acid to reduce Pu(IV) to Pu(III) to aid in plutonium recovery, although formic acid alone will strip tetravalent actinides, e.g., Th(IV) from 0D[IB]CMP0, once excess HNO3 present in the organic phase is removed. Thus, formic acid with HAF and NHF affords an excellent method for stripping all the actinides from these very powerful CMP extractants. Under the above conditions Am(III) and Cm(III) present in... 

Am-Cm Recovery from Plutonium Process Waste. Trans. Amer. Nucl. Soc. 

The ability of 2,3-dihydroxybenzoyl-N-glycine to form a stable complex with plutonium is demonstrated in Fig. 2. In the absence of 2,3-dihydroxybenzoyl-N-glycine plutonium nitrate hydrolyzed and could not be eluted from the column. In the presence of 2,3-dihydroxybenzoyl-N-glycine the plutonium recovery averaged 80%. 

Practical Utilization, Since the potential reserves of 235U are limited, some point will be reached where this power source no longer will be competitive with fossil fuels, synthetic fuels, solar power plants, etc.—unless the development of means for the practical utilization of plutonium can be achieved. An important element of nuclear fuel cost is the credit received from the sale or future utilization of plutonium after its recovery from spent fuel. The plutonium credit is realistic only if the plutonium is used for power production, since, at present, there are few commercial uses envisioned where it would yield a similar economic return. 

High-level radioactive defense waste solutions, originating from plutonium recovery and waste processing operations at the U.S. Department of Energy s Hanford Site, currently are stored in mild steel-lined concrete tanks located in thick sedimentary beds of sand and gravel. Statistically designed experiments were used to identify the effects of 12 major chemical components of Hanford waste solution on radionuclide solubility and sorption. 

The actinides are extracted from high acid (e.g., 7N HNO3) solutions and can be back-extracted with dilute acid. The method is therefore easily used with the column effluent from the anion exchange plutonium recovery step. 

Recovery of Plutonium Traces from Nitric Acid-Fluorhydric Acid Solutions by Sorption onto Alumina... 

Sorption of plutonium traces onto alumina from uranyl nitra-te solutions has been investigated. Several methods have been previously proposed for the recovery of plutonium traces from reprocessing solutions. Those methods include ion exchange (1 2), solvent extraction and, extraction chromatography 0,6). 

Table II shows that the plutonium sorption onto A190 is around 87% and it was demonstrated that with the aid of fluoride ions the recovery of plutonium traces from waste solutions was possible.

The Rocky Flats Plant (RFP) has a large recovery facility to recover plutonium from miscellaneous scraps and residues a byproduct in the plutonium recovery stream is americium from the decay of plutonium-241. Currently a NaCl-KCl-MgClo eutectic salt is used at Rocky Flats to separate americium from plutonium in a molten salt extraction (MSE) process. 

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. 

Another important consideration is the problem of dissolving the mixed-oxide fuel for subsequent reprocessing and plutonium recovery after the irradiated mixed-oxide fuel has been discharged from the reactor. When plutonium dioxide is in solid solution with uranium dioxide at low concentrations, as in the case of plutonium created during the irradiation of uranium dioxide... 

Chemical separation. Current concepts for high-efficiency separation of actinides call for improved plutonium recovery, coextraction of uranium and neptunium with subsequent partitioning by valence control, and extraction of amercium and curium from the HAW stream. There are a number of major problems to be solved before a technically feasible process will be available. 

Figure 11.22 Typical plutonium-recovery operations (Hanford Engineering Development Laboratory). (From Richardson (RlJ.)...

The acid flowsheet was found to give a better plutonium recovery than the acid deficient process. However, the latter gave better decontamination from gross activity, especially for uranium, with DF/ y figures of 3.1 x 10 for uranium and 2x 10 for plutonium being attained after two cycles. Comparable values for the acid flowsheet were respectively 1.6 x 10" and 6.3 X 10. Although... 

Separations research at the Rocky Flats Plant (United States) has found ways to significantly improve Pu recovery from HNO waste streams by using extraction chromatography. Several extractants, including TBP (tribu-tylphosphate), TOPO (tri- -octylphosphine oxide), DHDECMP (dihexyl N,N-diethylcarbamoylmethylenephosphonate), and CMPO (octylphenyl-N,N-diisobutylcarbamoylmethylphosphine oxide) and their mixtures supported on Amberlite XAD4 and anion exchange resins (Dowex 1x4 and Amberlite IRA 938), were evaluated for their ability to remove low concentrations of plutonium from 7 M acid nitric [71-73],... 

In addition, TOPO has the advantage of being able to sorb the hydrolytic plutonium polymer, which is a common component of plutonium waste streams, whereas anion exchange resins cannot do so [72]. Fortunately, American Cyanamid has introduced a low-cost alternative to TOPO, which also extracts plutonium polymer from nitric acid [73]. This material, called Cyanex 923, is a mixture of trialkyIphosphine oxides, and its low cost makes this approach for the recovery of plutonium more realistic. 

Reactor plutonium recovered from low enriched uranium spent fuel (less than 5% U-235) constitutes a typical example of a mixture of radionuclides with known identity and quantity for each constituent. Calculations according to para. 404 of the Regulations result in activity limits independent of the abundance of the plutonium radionuclides and the bumup within the range 10 000-40 000 MW d/t. The following values for reactor plutonium can be used within the above range of bumup, the Am-241 buildup taken into account, up to five years after recovery ... 

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