The joint conversion of uranium and plutonium to the oxide form makes it possible to do away with the complicated step of blending and grinding the two distinct oxide powders, as currently employed for the purposes of MOX fuel fabrication. Over the past few years, a number of co-conversion routes have thus been revisited or devised on the basis of original chemical principles, in order to innovate in the area of uranium and plutonium management. Drawing on 50 years industrial experience and feedback, CEA/DEN, in partnership with Areva NC, are proposing to extend oxalic conversion of Pu(IV) to Pu oxide to the oxalic co-conversion of uranium and plutonium.

Of the various possible variants, the current reference route is the U(IV)- Pu(III) variant, owing to the following advantages (Grandjean 2005):

• Unexpectedly, U(IV) and Pu(III) co-crystallize into a single oxalate structure, thus ensuring homogeneous distribution of the two actinides, at the molecular scale (Arab-Chapelt 2007). This structure is an oxalate solid solution, over a broad range of Pu/(U + Pu) ratios: 0-50% (mol/ mol).

• The mixed oxalate exhibits very low solubility in a nitric acid solution carrying excess oxalic acid: solubility is about the same as that of the plutonium(IV) oxalate precipitate.

• Calcination of this mixed oxalate in an inert atmosphere yields a solid solution of (U, Pu)O2 oxide; the homogeneity of uranium and plutonium distribution, at the molecular level, is thus conserved, from the solution to the oxide.

• Technology similar to that operated in the current plutonium conversion workshops at La Hague may be used.

A major advantage of oxalic co-conversion is to produce mixed oxides exhibiting physicochemical characteristics directly tailored to fuel fabrica­tion, affording, in particular, the ability to use this raw material directly, with no further specific chemical, or mechanical treatment. More broadly speaking, from PUREX to COEXTM, the evolution to an oxalic U-Pu co­conversion process, combined with a process adapted from a MIMAS-type process, affords the following advantages:

• A process yielding a uranium-plutonium mixture, without at any point involving the handling of separated plutonium, thus reducing prolifera­tion risks.

• The (U, Pu)O2 oxide solid solution obtained from co-conversion, as the endproduct of spent fuel treatment by the COEXTM process, and raw material for the fabrication of advanced MOX fuel, results in an enhanced homogeneity of plutonium distribution within the fuel (com­pared with the current upstream grinding operations with mixtures of UO2 and PuO2 powders). This translates to the anticipation of improved in-reactor behavior for MOX fuel fabricated from such raw material, bringing benefits with regard to increased burnups.

• Reduced contamination potential and limited exposure to radiation, during fuel fabrication using COEXTM-produced powder.

• A further outcome would be simplified management of fabrication scrap, and a likely improvement in MOX fuel solubility, subsequent to irradiation (no PuO2 islands).

6.3 Conclusions

PUREX was developed in the US in the late 1940s originally as a separation system for production of pure plutonium for military applications. In a way, PUREX was an industrial revolution and the success of this process led to the development of modern liquid-liquid extraction processes. More than half a decade after its development, PUREX is still the foundation of all modern separation processes for spent nuclear fuel used on industrial scale. The system utilizes tributyl phosphate (TBP) to extract uranium and plu­tonium away from most other elements in the spent nuclear material. The preference of tetravalent and hexavalent (as MO22+) ions is the key to the selective extraction. The stability of UO22+ combined with the relative ease of changes to the oxidation state of plutonium, predominantly between the +3 and +4 state, is the other critical aspect of this process. The current state of the art PUREX processes for commercial reprocessing are used in four countries in the world (France, Japan, the UK, and Russia) and have an annual capacity of more than 5500 tonne heavy metal (including the new

Rokkasho-Mura plant). The reprocessing plants combine extraction (TBP) and redox reagents (e. g. U(IV), hydroxylamine, hydrazine) in several counter current liquid-liquid extraction batteries under different conditions to extract, scrub and strip selected elements to obtain a pure product. PUREX is not entirely without problems and elements such as Np, Tc, Zr, Nb and Ru put high demands on the process such as increased number of stages and the need for parallel decontamination batteries. Improvements and changes made to the PUREX process over the years include more advanced equipment to facilitate easier handling and increase throughput. Although PUREX has been used successfully for a long time, research has been continuous and a number of recent discoveries and ideas are close to being implemented to a future PUREX process, e. g. Advanced PUREX, SuperPUREX, COEX, UREX, etc. These new processes are being devel­oped to meet new or future policies regarding waste and product stream composition. One major issue has been to avoid a pure Pu product by routing either neptunium (UREX and NPEX) or uranium (COEX) into the Pu-product stream, thereby increasing proliferation resistance. The complicated issue of neptunium control have been addressed using simple hydroxamic acids and successful pilot scale processes has been adminis­tered in both U. S. (UREX+), the UK (Adv. PUREX) and France (COEX). All these processes look promising for the future successful use of PUREX — like chemistry in advanced nuclear fuel reprocessing and the adaptability of this process is all but proven to accommodate for future restrictions on waste and product stream composition.