Precipitation methods plutonium separation

The discovery of nuclear fission in 1938 proved the next driver in the development of coordination chemistry. Uranium-235 and plutonium-239 both undergo fission with slow neutrons, and can support neutron chain reactions, making them suitable for weaponization in the context of the Manhattan project. This rapidly drove the development of large-scale separation chemistry, as methods were developed to separate and purify these elements. While the first recovery processes employed precipitation methods (e.g., the bismuth phosphate cycle for plutonium isolation). 

The sucessful experiments for the retention of plutonium onto alumina from TTN0 -HF solution gave enough confidence to recomend the proposed method to separate traces of plutonium from waste solutions in the presence of macroamounts of uranium (VI). Of course, only macroamounts of thorium, uranium (IV) and rare earths are serious interfering ions, since they precipitate with HF. The behavior expected for neptunium in the same system should be similar to plutonium, thorium and rare earths. The retention of neptunium from HNO - HF solutions is in progress. The sorption yield for Pu was around 95%. The sorption mechanism is not well established. Figure 3 shows the proposed flowsheet for recovery of Pu traces from reprocessing waste solutions. 

The first microgram quantities of plutonium were produced [S6] in 1942 by irradiation of natural uranium with deuterons in the cyclotron of Washington University in St. Louis. This plutonium was separated at the Chicago Metallurgical Laboratory of the Manhattan Project by Seaborg and his collaborators, who employed the method of carrier precipitation frequently used by radiochemists to extract small amounts of radioactive material present at low concentration. As wartime urgency required that a plutonium separation plant be designed and built before macro quantities of plutonium could be available for process development, it was decided to use the same carrier precipitation process that had successfully produced the first small quantities of this element. 

The various processes available for fuel reprocessing are aqueous solvent extraction, precipitation, ion exchange, fractional distillation, pyrometal-lurgy, and fluoride volatility. Most of the commercial development experience has come from the solvent-extraction method for separation of uranium, plutonium, and fission products. 

Mullins and Leary [29] patented a method of separating americium from plutonium which involves bubbling a mixture of oxygen and argon gas into a molten salt containing both elements. Plutonium precipitates as PuOj, whereas americium stays in solution. 

Initially, the only means of obtaining elements higher than uranium was by a-particle bombardment of uranium in the cyclotron, and it was by this means that the first, exceedingly minute amounts of neptunium and plutonium were obtained. The separation of these elements from other products and from uranium was difficult methods were devised involving co-precipitation of the minute amounts of their salts on a larger amount of a precipitate with a similar crystal structure (the carrier ). The properties were studied, using quantities of the order of 10 g in volumes of... 

The wastes from uranium and plutonium processing of the reactor fuel usually contain the neptunium. Precipitation, solvent extraction, ion exchange, and volatihty procedures (see Diffusion separation methods) can be used to isolate and purify the neptunium. 

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. 

Plutonium Finishing. The separated plutonium was processed to Pu02 by conventional cation resin exchange, oxalate precipitation, and calcination methods. 

Electrodeposition is currently used in a minority of laboratories to prepare a thin, uniform, and reproducible source. The alpha-particle emitting isotopes of plutonium are electrodeposited on polished stainless steel, or platinum disk. In the co-precipitation technique, a small amount of a carrier (e.g., LaF 3) is used to co-precipitate the separated and purified plutonium from solution. The precipitate is then prepared for counting by either filtration or by evaporation of a slurry of the precipitate onto a stainless steel disk or planchet (ASTM 1982 1987). Recent methods use a glass fiber filter which can be used as the source for alpha counting techniques. It has been suggested that low yields result from electrodeposition due to the presence of traces of interfering elements (e.g., iron) (Bernhardt 1976). 

---