From 2003 to 2007 a number of potential flowsheets for the processing of commercial LWR spent fuel were developed and tested with actual spent fuel at laboratory-scale. These processes are outlined in Fig. 7.1. The UREX+1 series was intended for extraction of the TRU elements as a

7.7 UREX+ flowsheets for the treatment of LWR SNF based on the GNEP identified repository benefits given in Table 7.1.

group; there is no separation among the TRUs which are to be burned as fuel in fast spectrum reactors (FR). The UREX+2 and UREX+3 process series were intended for LWR or FR reactor recycle of plutonium and neptunium. The combined product is a recycle fuel feedstock that could be manipulated and fabricated into fuel in a glovebox facility rather than a shielded cell. Because of the large relevant database on fuel performance from prior testing programs, such a fuel may be easier to qualify than the TRU fuel. The UREX+4 process series was selected as it provides the added option of burning americium in specially designed target assemblies. Separation of Cm from the Am reduces shielding requirements for target fabrication. With the isolation of Cs and Sr, the other fission products remaining after any of the UREX+ process strategies have relatively low radiation levels and heat generation rates, and can therefore be immobi­lized at high concentrations in durable ceramic or glass waste forms. The approximately 30-year half-lives of 137Cs and 90Sr isotopes imply that storage is required for 500 years or less.

Many of the head-end operations necessary for an advanced reprocessing plant using the UREX+ strategy are similar to those currently used at the commercial spent fuel reprocessing facilities at LaHague (France), Sellafield (United Kingdom), and Rokkasho (Japan). While there may be qualities unique to the fuel to be processed that require some head-end conditioning differing from that used at these facilities, the major operations and process streams will be similar. As these facilities are designed for PUREX, the major change would be adjusting the dissolver product conditions to meet the requirements of the UREX process.

Off-gases emanating from the fuel-chopper and the dissolver will contain radioactive 129I, tritium, 14C, and 85Kr. In the current reprocessing plants in Europe and Asia, these are not fully captured. If capture and sequestration of volatile radionuclides is mandated, the effectiveness of capture by various adsorbents or scrubbers must be evaluated, and methods to recover the radionuclides and sequester them in waste or storage forms must be developed.

Post-process treatment and solidification of effluents is an area where data are limited for many of the UREX+ separations. There is extensive plant-scale experience in converting uranium and plutonium to solid oxides but not for mixed actinide streams. Conversion of high-level liquid wastes to solid forms by either calcination or vitrification has been demonstrated at industrial scales but not for high activity Cs/Sr and Am/Cm solutions. Extensive data exist for acid-water evaporation, acid recycling, and waste­water treatment. However, information on reagent cleanup and recycle is less extensive. Data on recovery and calcination of trace organics in aqueous streams is likely available, though not for all of the organic species used in a UREX+ processes.

Design of the UREX+ flowsheets developed for the treatment of LWR SNF to achieve the repository benefits listed in Table 7.1 was done by com­bining solvent extraction systems in series. Figure 7.1 shows how several different solvent extraction modules were combined for several of the options. The solvent extraction process modules were tuned to meet the disposition goals of the program, also given in Table 7.1. It should be noted that isolation of any of the products listed in Table 7.2 could be omitted by recombination or exclusion of specific separation modules. In achieving these product goals, the developer is free to select any extraction process that will yield the desired products and can be made compatible with other process modules through intermediate processing.

The PUREX process is fully developed and will yield separate Pu and U streams. Recent modifications to the basic PUREX process are intended to result in incomplete separations, so that Pu is recovered with neptunium and/or uranium. These mixed PUREX separations, have been developed by both DOE and commercial companies, and are referred to generically as co-decontamination processes. Specific examples include, COEX® devel­oped by Areva, and the first separation in Energy Solutions’ NUEX process. The concept has been demonstrated at both Argonne and Oak Ridge National Laboratories with actual spent fuel.

The UREX process was uniquely developed to recover technetium, in addition to uranium. The solvent for the UREX process is the typical PUREX solvent, tributyl phosphate (TBP) dissolved in n-dodecane. Uranium and technetium are extracted from the bulk of the dissolved fuel, but co-extraction of Pu and Np is prevented by introduction of a complex — ant/reductant. Once stripped from the solvent, uranium is separated from Tc by anion exchange. Technetium was targeted because its mobility in the environment, as pertechnetate, contributed to the potential long-term dose to the public from the Yucca Mountain Repository. With UREX, it can be recovered in high yield, >95%. This level of recovery of technetium with a co-decontamination-type separation is difficult.

After the lead UREX separation, there is flexibility in the follow-on separations modules. Using the UREX+3a process, as an example, we can follow a typical separations scheme based solely on aqueous solvent extrac­tion to achieve the desired separations. UREX+3a is a cascade of five solvent separate extraction processes, referred to here as “process modules” and one ion exchange process. Following UREX and ion exchange, the process steps include: (1) recovery of Cs and Sr (CCD-PEG or FPEX),

(2) separation and recovery of a mixed U, Pu and Np product (NPEX),

(3) separation of transition metal fission products from lanthanide fission products and Am and Cm (TRUEX), and (4) separation and recovery of Am and Cm from lanthanides (TALSPEAK). Similar separations can be done for a co-decontamination front end, but in this case, the NPEX segment is omitted as the Pu and Np are recovered in the initial separation segment as shown in Fig. 7.1 (2004).

There is some flexibility in both the sequencing and selecting the process for Cs and Sr recovery. Cs and Sr can be recovered immediately after UREX using either the FPEX or the CCD-PEG process with appropriate feed adjustment. The CCD-PEG solvent is a mixture of chlorinated cobalt dicarbollide (CCD) and polyethylene glycol (PEG); the diluent is phenyl — trifluoromethyl sulfone (FS-13). An alternative process is FPEX, which uses a solvent containing a calixarene, a crown ether, and a modifier, all in Isopar-L® (a refined kerosene) diluent. The FPEX diluent is more compat­ible with the other processes, which are based on n-dodecane, but the CCD-PEG process is more mature. In all of the schemes shown in Fig. 7.1, the raffinate from the UREX segment is the feed to the CCD-PEG or FPEX process segment. However, these processes can be run to recover Cs and Sr from the TRUEX raffinate rather than the UREX raffinate. Reducing the amount of processing prior to TRU recovery should result in a purer product, though shielding requirements are higher. This would remove one process from the main sequence, which may be beneficial for actinide product purity for recycle in reactor fuel.

The NPEX process separates plutonium and neptunium from the other components in the FPEX or CCD-PEG raffinate. Like UREX, the solvent for NPEX is the typical PUREX solvent. Although the process was designed to extract Pu and Np, uranium is added to the extracted Pu and Np in the strip section to yield a U-Pu-Np product with a controllable elemental ratio.

The TRUEX process separates the lanthanide and transuranic actinides from the other components in the NPEX raffinate. The solvent is a mixture of CMPO (octyl(phenyl)-N, N — diisobutylcarbamoylmethyl phosphine oxide) and TBP in n-dodecane. If TRUEX follows NPEX, Am and Cm are recovered with the lanthanides. If NPEX is omitted, Pu and Np are also contained in the “grouped TRU” product.

The actinide:lanthanide separation is achieved using TALSPEAK. TALSPEAK solvent is HDEHP (bis(2-ethylhexyl)phosphoric acid dis­solved in n-dodecane while the aqueous phase is a buffered lactate solution containing DTPA. Extraction of the lanthanides over actinides is regulated by control of pH and operating conditions. The product contains Am and Cm for UREX+2, 3 and 4 series, and all of the transuranics in the case of the UREX+1 series. If a suitable separation is devised, Am can be further separated from Cm, yielding UREX+4 in Table 7.1. At the moment there are no proven candidates for the separation of Am from Cm at an industrial scale.

These separation module sequences (except for Am from Cm separation) are based on proven or at least lab-tested processes that can be run on prototypical equipment. Other aqueous processing concepts have been proposed, but are not currently suitable for bench-scale process tests. As it is attractive to reduce the number of processing steps required to achieve the separations, much current focus is on finding extractants that are highly elective for actinides. One approach under development which has shown promise is a TRUEX-TALSPEAK hybrid, christened TRUSPEAK, which separates actinides from all fission products in a single process module rather than two sequential modules.