Plutonium processing crystallization

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 The former methods have been historically important but have now been superseded by solvent extraction with TBP.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. 

Americium was isolated first from plutonium, then from lanthanum and other impurities, by a combination of precipitation, solvent extraction, and ion exchange processes. Parallel with the separation, a vigorous program of research began. Beginning in 1950, a series of publications (1-24) on americium put into the world literature much of the classic chemistry of americium, including discussion of the hexavalent state, the soluble tetravalent state, oxidation potentials, disproportionation, the crystal structure(s) of the metal, and many compounds of americium. In particular, use of peroxydisulfate or ozone to oxidize americium to the (V) or (VI) states still provides the basis for americium removal from other elements. Irradiation of americium, first at Chalk River (Ontario, Canada) and later at the Materials Testing Reactor (Idaho), yielded curium for study. Indeed, the oxidation of americium and its separation from curium provided the clue utilized by others in a patented process for separation of americium from the rare earths. 

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. 

The influence of Pu valence in the feed solution was examined in the U crystallization process (Yano et al., 2004). When Pu(IV) existed in the feed solution, the yellow crystal was observed. On the other hand, the apvpearance of the crystal was orange in the feed solution adjusted so that Pu valence was Pu(VI), this color likely resulting from the mixture of the yellow crystal of UNH and the red crystal of PuNH. Plutonium(VI) in the feed solution was co-crystallized with U(VI) in the course of U crystallization. The crystal yields of Pu were smaller than those of U (Ohyama et al., 2005). The fact that the crystal ratio of Pu is smaller than that of U suggests a mechanism of U-Pu co-crystallization in which U begins to crystallize when the saturation point of U is reached by cooling the feed solution, and then Pu is crystallized on the UNH crystal. 

Plutonium behavior in the U crystallization process depends on the Pu valence in the feed solution. In this study, the Pu valence in the feed solution was changed to Pu(IV) by NO gas bubbling. After the crystallization, almost all the Pu remained in the mother liquor and attached to the surface of the UNH crystal. The mother liquor on the surface of crystal was efficiently removed after the UNH crystal was washed. 

Adachi, T. Ohnuki, M. Yoshida, N. Sonobe, T. Kawamura, W. Takeishi, H. Gunji, K. Kimura, T. Suzuki, T. Nakahara, Y. Muromura, T. Kobayashi, Y. Okashita, H. Yamamoto, T. (1990). Dissolution Study of Spend PWR Fuel Dissolution Behavior and Chemical Properties of Insoluble Residues. Journal of Nuclear Materials, Vol. 174, No. 1, (November 1990), pp. 60-71, ISSN 0022-3115 Anderson, H. H. (1949). Alkali Plutonium(IV) Nitrates, In The Transuranium Elements, National Nuclear Energy Series IV, Vol. 14B, G. T. Seaborg, J. J. Katz, W. M. Manning, (Eds.), pp. 964-967, McGraw-Hill Book Co. Inc., New York, USA Ebert, K Henrich, E. Stahl, R. Bauder, U. (1989). A Continuous Crystallization Process for Uranium and Plutonium Refinement, Proceedings of 2nd International Conference on Separation Science Technology, pp. 346-352, Paper No. S5b, Hamilton, Ontario, Canada, October 1-4,1989... 

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. 

---