T Thomas Peters

Full-Scale Testing of a Caustic Side Solvent Extraction System to Remove Cesium from Savannah River Site Radioactive Waste

Abstract Savannah River Site (SRS) personnel have completed construction and assembly of the Modular Caustic Side Solvent Extraction Unit (MCU) facility. Following assembly, they conducted testing to evaluate the ability of the process to remove non-radioactive cesium and to separate the aqueous and organic phases. They conducted tests at salt solution flow rates of 3.5, 6.0, and 8.5 gpm. During testing, the MCU Facility collected samples and submitted them to Savannah River National Laboratory (SRNL) personnel for analysis of cesium, Isopar® L, and modifier [1-(2,2,3,3-tetrafluoropropoxy)-3-(4-sec-butylphenoxy)-2-propanol]. SRNL personnel analyzed the aqueous samples for cesium by Inductively-Coupled Plasma Mass Spectroscopy (ICP-MS) and the solvent samples for cesium using a Parr Bomb digestion followed by ICP-MS. They analyzed aqueous samples for Isopar® L and modifier by gas chromatography (GC). The conclusions from the cesium analyses follow. The cesium in the feed samples measured 15.8 mg/L, in agreement with expectations. The decontamination factor measured 181–1580 at a salt solution flow rate of 3.5 gpm, 211–252 at a salt solution flow rate of 6.0 gpm, and 275–878 at a salt solution flow rate of 8.5 gpm. The concentration factor measured 11.0–11.1 at 3.5 gpm salt solution flow rate, 12.8–13.2 at 6.0 gpm salt solution flow rate, and 12.0–13.2 at 8.5 gpm salt solution flow rate. The organic carryover from the final extraction contactor (#7) varied between 22 and 710 mg/L Isopar® L. The organic carryover was less at the lowest flow rate. The organic carryover from the final strip contactor (#7) varied between 80 and 180 mg/L Isopar® L. The organic carryover in the Decontaminated Salt Solution Hold Tank (DSSHT) and the Strip Effluent Hold Tank (SEHT) was less than 10 mg/L Isopar® L, indicating good recovery of the solvent by the coalescers and decanters.

Performance Testing of the Next-Generation CSSX Solvent with Actual SRS Tank Waste

Efforts are underway to qualify the Next-Generation Solvent for the Caustic Side Solvent Extraction (CSSX) process. Researchers at multiple national laboratories have been involved in this effort. As part of the effort to qualify the solvent extraction system at the Savannah River Site (SRS), Savannah River National Laboratory (SRNL) researchers performed a number of tests at various scales. A series of batch equilibrium, or Extraction-Scrub-Strip (ESS), tests were conducted first. These tests used ∼30 mL of Next-Generation Solvent and either actual SRS tank waste, or waste simulant solutions. The results from these cesium mass transfer tests were used to predict solvent behavior under a number of conditions. For larger-scale testing, twelve stages of 2-cm (diameter) centrifugal contactors were assembled. This rack of contactors is structurally similar to one tested in 2001 during the demonstration of the baseline CSSX process. No issues were encountered during assembly and mechanical testing. A non-radiological test was performed using 35 L of cesium-spiked caustic waste simulant followed by a test with 39 L of actual tank waste. Test results are discussed, particularly those related to the effectiveness of extraction.

Results of Supplemental MST Studies

The current design of the Salt Waste Processing Facility (SWPF) includes an auxiliary facility, the Actinide Finishing Facility, which provides a second contact of monosodium titanate (MST) to remove soluble actinides and strontium from waste if needed. This treatment will occur after cesium removal by Caustic-Side Solvent Extraction (CSSX). Although the process changes and safety basis implications have not yet been analyzed, provisions also exist to recover the MST from this operation and return to the initial actinide removal step in the SWPF for an additional (third) contact with fresh waste. A U.S. Department of Energy (DOE) request identified the need to study the following issues involving this application of MST: Determine the effect of organics from the solvent extraction (CSSX) process on radionuclide sorption by MST; Determine the efficiency of re-using MST for multiple contacts; and Examine fissile loading on MST under conditions using a waste containing significantly elevated concentrations of plutonium, uranium, neptunium, and strontium. This report describes the results of three experimental studies conducted to address these needs: (1) Addition of high concentrations of entrained CSSX solvent had no noticeable effect, over a two week period, on the sorption of the actinides and strontium by MST in a direct comparison experiment. (2) Test results show that MST still retains appreciable capacity after being used once. For instance, reused MST--in the presence of entrained solvent--continued to sorb actinides and strontium. (3) A single batch of MST was used to sequentially contact five volumes of a simulant solution containing elevated concentrations of the radionuclides of interest. After the five contacts, we measured the following solution actinide loadings on the MST: plutonium: 0.884 {+-} 0.00539 wt % or (1.02 {+-} 0.0112) E+04 {micro}g/g MST, uranium: 12.1 {+-} 0.786 wt % or (1.40 {+-} 0.104) E+05 {micro}g/g MST, and neptunium: 0.426 {+-} 0.00406 wt % or (4.92 {+-} 0.0923) E+03 {micro}g/g MST. (4) Over the duration of an experiment with the sequential strikes, the ability of MST to sorb actinides improved with additional strikes. This trend is counter-intuitive, but is confirmed by replicate experiments for plutonium, uranium, and neptunium. Conversely, over the duration of the experiment, the ability of MST to sorb strontium decreased the more it was used. This trend is confirmed by replicate experiment.

SODIUM ALUMINOSILICATE SOLIDS AFFINITY FOR CESIUM AND ACTINIDES

Washed sodium-aluminosilicate (NAS) solids at initial concentrations of 3.55 and 5.4 g/L sorb or uptake virtually no cesium over 288 hours, nor do any NAS solids generated during that time. These concentrations of solids are believed to conservatively bound current and near-term operations. Hence, the NAS solids should not have affected measurements of the cesium during the mass transfer tests and there is minimal risk of accumulating cesium during routine operations (and hence posing a gamma radiation exposure risk in maintenance). With respect to actinide uptake, it appears that NAS solids sorb minimal quantities of uranium - up to 58 mg U per kg NAS solid. The behavior with plutonium is less well understood. Additional study may be needed for radioactive operations relative to plutonium or other fissile component sorption or trapping by the solids. We recommend this testing be incorporated in the planned tests using samples from Tank 25F and Tank 49H to extend the duration to bound expected inventory time for solution.