Plutonium processing evaporation

Waste Treatment. Figure 2 outlines the current waste recovery and treatment processes, and proposed changes. Acid waste streams are sent through nitric acid and secondary plutonium recovery processes before being neutralized with potassium hydroxide and filtered. This stream and basic and laundry waste streams are sent to waste treatment. During waste treatment, the actinides in the aqueous waste are removed by three stages of hydroxide-iron carrier-flocculant precipitation. The filtrate solution is then evaporated to a solid with a spray dryer and the solids are cemented and sent to retrievable storage. 

The aqueous waste from the DBBP extraction process, still containing small but significant amounts of plutonium and 241Am, is diluted with a large volume of waste water from other PRF operations to produce the PRF salt waste (Table II). When evaporated to dry salts, the PRF salt waste contains, typically 1000 to 2000 nCi/g of alpha emitters. 

No definite reason for these fluctuations could be identified, but it is known that neptunium, due to its complicated redox chemistry, reacts in a very sensitive way to even minor process variations (7,8). Based on these results the proposal was made (J5) to recover the "co-extracted" portion of the neptunium by running the second plutonium and uranium purification cycles under conditions where the Np is directed into the aqueous raffinates (2AW and 2DW streams). In the Pu purification cycle, this can be done by adding sufficient nitrous acid to keep the Np pentavalent, while in the U purification cycle (which is run under slightly reducing conditions) a low acidity and a high loading help to reject Np into the aqueous 2DW stream. The two raffinate streams are combined in WAK in the 3W evaporator, and the Np is thus collected in the concentrate from this unit (3WW stream). Consequently the proposal was made to recover the Np from this 3WW stream by use of the well-known anion exchange process (9,J ). 

The scraps which arise during the fabrication of plutonium-containing nuclear fuels are collected and stored for some time before they are processed to recover the plutonium. Due to the decay of Pu-241, considerable amounts of Am-241 may build up in the stored material. At the Alkem company, plutonium is recovered from the scrap by anion exchange the americium which is not sorbed on the resin is collected in the combined effluents from the loading and wash steps. The effluents are concentrated by evaporation besides americium, the concentrated effluents contain major amounts of uranium, plutonium, corrosion products, and residues from chemical reagents. A typical composition is given below ... 

The hot run was made with the feed solution obtained by dissolving highly irradiated Pu-Al alloy in HNO3 with mercuric ion catalyst. Uranium was added to the solution to produce a typical Purex feed. Uranium and most of the plutonium were recovered by the normal Purex process. The aqueous waste containing Am, Cm, Cf, fission products, Al, and Hg was evaporated and acid was stripped to produce the feed (Table 2). The results are as expected from the laboratory tests excellent recovery of Pu, Am, and Cm but low decontamination factors (DF). 

An advantage of the Purex process is the low salt content of the aqueous waste stream so the liquid volume can be reduced by evaporation. The HNO3 values are recycled to the process. The Purex process produces nearly pure plutonium and recovers uranium and it is the process that has been adapted to treat domestic spent nuclear fuel. 

The tendency toward Pu(IV) polymerization is of considerable practical importance in process operations involving plutonium solutions. Dilution of an acidic plutonium solution with water can result in polymerization in localized regions of low acidity, so plutonium solutions should be diluted instead vdth acid solutions. Polymerization can result from leaks of steam or water into plutonium solutions or by overheating during evaporation. Polymer formation can clog transfer lines, interfere with ion-exchange separations, cause emulsification in solvent extraction and excessive foaming in evaporation, and can result in localized accumulation of plutonium that may create a criticality hazard [CS]. 

Fractional crystallization. Volatile metals with much lower boiling points than uranium, such as magnesium (1103°C), zinc (906°C), and cadmium (767°C), have been extensively studied as solvents for separating constituents of irradiated metal fuel by fractional crystallization, followed by evaporation of the solvent metal from the separated fractions. For example, in liquid magnesium, the solubility of plutonium or thorium is high, but uranium is very low. A process of this type was developed at Argonne National Laboratory [P6] for concentrating plutonium in the uranium metal blanket of a breeder reactor from 1 percent to 40 percent. 

Low-level aqueous wastes from steps 6 and 8 are processed for further recovery of plutonium and uranium, then concentrated for recovery of water and nitric acid. High-level aqueous wastes from step 4 are concentrated by evaporation, with recovery of condensed nitric acid in step 11. 

As in the Purex process, the Thorex process uses a solution of TBP in hydrocarbon diluent to extract the desired elements, uranium and thorium, from an aqueous solution of nitrates. Thorium nitrate however, has a much lower distribution coefficient between an aqueous solution and TBP than uranium or plutonium. To drive thorium into the TBP, the Thorex process as first developed at the Knolls Atomic Power Laboratory [HI] and the Oak Ridge National Laboratory [G14] added aluminum nitrate to the thorium nitrate in dissolved fuel. This had the disadvantage of increasing the bulk of the high-level wastes, which then contained almost as many moles of metallic elements as the original fuel. To reduce the metal content of the waste, the Oak Ridge National Laboratory in the late 1950s [Rl, R2] developed the acid Thorex process, in which nitric acid is substituted for most of the aluminum nitrate in the first extraction section. The nitric acid is later evaporated from the wastes, as in the Purex process. 

The uranium fraction is concentrated by evaporation and nitric acid is added to -S M. Uranium is then extracted into a TBP solution that is scrubbed with a reducing solution to remove traces of plutonium and neptunium. This process is repeated twice. Then uranium is back-extracted into 0.01 M nitric acid and converted to uranium oxide. 

The pure plutonium nitrate product from the solvent extraction process was concentrated in a titanium evaporator and conditioned with respect to nitric acid concentration and the plutonium IV valency state to reduce radiolytic off-gas production before loading to the transport flask for shipment to BNFL Sellaiield. 

As an example, most of the radioactivity from plutonium production is found in the liquid high-level waste from the first cycle of the Purex process. This liquid is neutralized with sodium hydroxide and stored in earth-shielded tanks. There a sludge settles out that contains most of the radioactivity. The residual liquor is partially evaporated to decrease the volume of the waste, and sodium nitrite crystallizes on top of the sludge. In a process to be used at the Savannah River plant, the sodium liquor fraction, containing most of the cesium fission product, is pumped from the tank and precipitated with tetra-phenyl borate. The precipitate will be calcined and packaged for disposal in a high-level repository, and the sodium nitrate crystallized from the residual liquor and sent to a low-level waste repository. 

The uranium ore concentrate or yellowcake is not a very pure substance. But, the dissolution, filtering, solvent extraction, and evaporation processes in the production of uranyl nitrate result in a reasonably pure uranium compound such as specified by the American Society for Testing and Materials standards. Thus, a chemically pure form of natural uranium is available as uranyl nitrate or UO3 prior to introduction of safeguards. Because of the high neutron-capture cross section for " N, uranyl nitrate is not a likely candidate as a fuel for a natural uranium reactor. However, pure UO3 could potentially be used as a ready source for fuel in a specially designed plutonium production reactor thus raising a potential diversion concern. 

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