The COEXTM or COEXtraction process has been developed by French workers, and is designed to bleed some of the U (and possibly Np) into the Pu product to eliminate the pure Pu stream (Baron 2007). The development of this new reprocessing and recycling COEXTM process was aimed at:

• further enhancing proliferation resistance

• maintaining a high level of process performance

• minimizing both investment (capital) and operational costs

• taking full advantage of present industrial experience

• keeping open possible evolutions to take into account new types of reactors or future changes in management strategies of the transuranic elements.

The separation flowsheets used in the COEXTM process (Drain 2008) are based on the expertise accruing from the design, and deployment of the PUREX process, but equally from operational feedback with that process. The design for the extraction cycles has extensively relied on a simulation tool, validated by experimental investigations carried out in the workshops at the La Hague plants, and equally by comparisons with findings from industrial operations. Consequently, the COEXTM process is basically an

evolution of the PUREX process, by modifying it to produce a U + Pu mixture (U/Pu >20%), rather than pure plutonium: no pure plutonium is separated at any point of the process (see Fig. 6.7). The advantage of so doing is to curb proliferation risks and to produce a perfectly homogeneous mixed oxide, for the purposes of MOX fuel fabrication, affording enhanced performance.

The proposed COEXTM flowsheet is indicated in Fig. 6.7, and involves an initial U/Pu codecontamination step that is fairly similar to the first purifica­tion cycle, as implemented in the La Hague and Rokkashomura plants (see Fig. 6.2). The codecontamination step is unchanged: this covers the extrac­tion flowsheet, including measures to allow removal of technetium. The plutonium stripping part is modified by including conditions that results in uranium scrubbing into the Pu strip product. This function is slightly tweaked from uranium scrub to neptunium scrub, to allow some uranium to remain in the plutonium stream. The function further allows neptunium to be extracted from the plutonium stream, directing the neptunium thus extracted to the cycle’s uranium product stream, thanks to higher distribu­tion coefficient of Np(IV) compared to U(IV) (see Table 6.1).

To ensure the plutonium becomes extractable by the solvent phase again, an adjustment is made to the solution yielded by the first cycle, by raising the nitric acid concentration, concomitant with reoxidation of plutonium to oxidation state IV, uranium being adjusted to oxidation state VI. The stream then undergoes treatment in the form of a U-Pu cycle, only involving
plutonium extraction, and back-extraction (stripping) functions. Reductive plutonium stripping is effected using hydroxylamine nitrate.

Topping up with uranium is provided for, at the head end of this opera­tion, to adjust the Pu/U ratio in the production stream. Finally, the stripped solvent is recycled in the plutonium stripping operation carried out upstream, in the first cycle. This measure makes it possible to tolerate some loss of plutonium to the stripped solvent, and thus limit the number of stages used for the plutonium stripping operation (hence the term “mini­cycle,” for this complementary purification cycle).

A uranium purification cycle, identical to the cycles implemented at the La Hague plant, is further provided for — on the one hand, to complete decontamination of the uranium stream, with respect to в and у emitters, and, on the other hand, to ensure removal of neptunium.