S Samuel Fink

INVESTIGATING HYDROGEN GENERATION AND CORROSION IN THE TREATMENT TANK AND THE POTENTIAL FORMATION OF A FLOATING LAYER IN NEUTRALIZATION TANK DURING WASTE TANK HEEL CHEMICAL CLEANING

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TESTING AND EVALUATION OF THE MODIFIED DESIGN OF THE 25-DISK ROTARY MICROFILTER

This report details redesign of a commercially available rotary microfilter to meet the operational and maintenance requirements for radioactive service. Personnel developed the design and coordinated procurement of two filters followed by testing of one unit. System testing examined the ability to rinse soluble material from the system, filtration performance using several insoluble solids loadings, effectiveness in washing sludge, amount of wear to parts and maintenance of the system including the insertion and removal of the filter stack, and the ability to flush solids from the system. The test program examined flushing the filter for soluble material by filling the system with a Rhodamine WT dye solution. Results showed that draining the system and rinsing with 50 gallons of water resulted in grater than 100X reduction of the dye concentration. Personnel determined filter performance using various amounts of insoluble sludge solids ranging from 0.06 to 15 weight percent (wt%) insoluble solids in a 3 molar (M) sodium simulated supernate. Through approximately 120 hours of start-and-stop (i.e., day shift) operation and various insoluble solids loadings, the filter produced filtration rates between 3 and 7 gallons per minute (gpm) (0.12-0.29 gpm/ft{sup 2}) for a 25-disk filter. Personnel washed approximately 80 gallons of simulated sludge using 207 gallons of inhibited water. Washing occurred at constant volume with wash water fed to a well mixed tank at the same rate as filtrate removal. Performance measurement involved collecting and analyzing samples throughout the washing for density and sodium content. Results showed an effective washing, mimicking a predicted dilution calculation for a well mixed tank and reducing the sodium concentration from 3.2 M to less than 0.3 M. Filtration rates during the washing process ranged between 3 and 4.3 gpm for one filter unit. The filter system then concentrated the washed 15 wt% insoluble solids slurry to approximately 20 wt% insoluble solids with no operational problems with the exception of the entrainment of air due to leaking packing in the feed pump. Prior to the air entrainment, the filtration rate was approximately 4.2 gpm for one filter assembly with the process fluid temperature adjusted to 35 C. Personnel measured the turbidity of filtrate samples from all phases of testing. All samples measured were less than 3 NTU, with the majority of samples less than 1 NTU. Thus, all measurements fell below the process acceptance criterion of less than 5 NTU. After slurry operations, personnel rinsed the filter with the equivalent of 250 gallons of water by re-circulating 50 gallons of water. The residual sludge solids remaining on the filter stack weighed approximately 685 grams. This amount of solids corresponds to an equivalent activity of 15.1 curies (Ci) beta and 0.38 Ci gamma radiation dose for Sludge Batch 4. Workers completely disassembled the filter system and examined it for signs of wear and component operation. An evaluation by a John Crane Inc. representative concluded that the wear observed on the mechanical seal resulted primarily from the numerous stops and starts, the abrasive nature of the process fluid and the possibility that the seal faces did not receive enough lubrication from the process fluid. No measurable slurry bypassed the mechanical seal. While it is extremely difficult to predict the life of the seal, the vendor representative indicates a minimum of one year in present service is reasonable. Changing the seal face material from silicon-carbide to a graphite-impregnated silicon-carbide is expected to double the life of the seal. Replacement with an air seal might be expected to increase lifetime to five years. The bottom bushing showed wear due to a misalignment during the manufacture of the filter tank. Minor adjustments to the alignment with shims and replacement of the graphite bushing with a superior material will greatly reduce this wear pattern.

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.

TESTING OF A FULL-SCALE ROTARY MICROFILTER FOR THE ENHANCED PROCESS FOR RADIONUCLIDES REMOVAL

Savannah River National Laboratory (SRNL) researchers are investigating and developing a rotary microfilter for solid-liquid separation applications in the Department of Energy (DOE) complex. One application involves use in the Enhanced Processes for Radionuclide Removal (EPRR) at the Savannah River Site (SRS). To assess this application, the authors performed rotary filter testing with a full-scale, 25-disk unit manufactured by SpinTek Filtration with 0.5 micron filter media manufactured by Pall Corporation. The filter includes proprietary enhancements by SRNL. The most recent enhancement is replacement of the filters main shaft seal with a John Crane Type 28LD gas-cooled seal. The feed material was SRS Tank 8F simulated sludge blended with monosodium titanate (MST). Testing examined total insoluble solids concentrations of 0.06 wt % (126 hours of testing) and 5 wt % (82 hours of testing). The following are conclusions from this testing.

THERMAL AND SPECTROSCOPIC ANALYSES OF CAUSTIC LIDE SOLVENT EXTRACTION SOLVENT CONTACTED WITH 16 MOLAR AND 8 MOLAR NITRIC ACID

Thermal and spectroscopic analyses were performed on multiple layers formed from contacting Caustic Side Solvent Extraction (CSSX) solvent with 1 M or 3 M nitric acid. A slow chemical reaction occurs (i.e., over several weeks) between the solvent and 1 M or 3 M nitric acid as evidenced by color changes and the detection of nitro groups in the infrared spectrum of the aged samples. Thermal analysis revealed that decomposition of the resulting mixture does not meet the definition of explosive or deflagrating material.

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.