Another version of the UREX+ process was tested in miniature centrifugal contactors at Argonne National lab in 2004 using the same spent fuel feed as previously described (Pereira 2005). Changes to the process included the coextraction of U, Tc, Np and Pu from the rest of the waste, see Fig. 6.6, routing the high active waste to further Cs/Sr removal. The benefit of this methodology, compared to the previous UREX+ flowsheet, is that the low specific activity of U, Tc, Np and Pu would be separated from the rest of the high active waste early in the process. This would make handling of the U/Tc + Np/Pu product streams much easier considering dose rates and shielding required. Another improvement is that less organic solvent is needed since only one extraction section is required followed by re-extrac­tion of uranium and technetium and reductive stripping of Np and Pu. Furthermore, no feed adjustment is needed for the second extraction of neptunium and plutonium where AHA is destroyed. The flowsheet for creating both U/Tc + Np/Pu product streams (Fig. 6.6) can be compared to only the upper part of the flowsheet of Fig. 6.5, for the UREX process. As in the first UREX+ test, the organic phase was not recycled. The recovery of neptunium and plutonium in the Np/Pu product was 87.2 and 99.8%, respectively. As before, some neptunium was lost to the raffinate but was recovered downstream. The largest impurity in the Np/Pu product was

zirconium, although this was claimed to relate to hydraulic problems in the Zr scrub section rather than chemical behavior.

The U/Tc separation was accomplished by anion exchange where the pertechnetate was retained by the resin and the uranium was effectively purified to a high degree, reaching all goals set for disposal as low level waste. The use of ion exchange as part of the PUREX process for Tc sepa­ration has been investigated separately (Dileep 2008) where anion exchange was used as a purification step for obtaining a pure plutonium product after the U/Pu split.

A number of flowsheets of the same general setup as those shown in Figs 6.5 and 6.6 have been tested in the UK since 2003 and are summarized in a publication by Birkett et al. (Birkett 2007). The U+Pu or U+Np product streams were first coextracted from the high active waste and transported by the loaded solvent to a separate stage for U/Np or U/Pu separation with the aid of AHA. Different tests were carried out to investigate the feasibil­ity of this process for different concentrations of Pu in the feed, and simu­lated 1) thermal MOX-fuel (7 wt.% Pu); 2) fuel from fast reactors (20 wt.% Pu); and 3) treatment of exotic Pu legacy waste (40 wt.% Pu). In all tests pure uranium product could be obtained. The purity of the plutonium product could potentially be controlled to accommodate proliferation resistance.