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

THERMAL AND SPECTROSCOPIC ANALYSES OF CAUSTIC SIDE SOLVENT EXTRACTION SOLVENT CONTACTED WITH 1 MOLARAND 3 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

DEVELOPMENT OF A ROTARY MICROFILTER FOR RADIOACTIVE WASTE APPLICATIONS

The processing rate of Savannah River Site (SRS) high-level waste decontamination processes are limited by the flow rate of the solid-liquid separation. The baseline process, using a 0.1 micron cross-flow filter, produces {approx}0.02 gpm/sq. ft. of filtrate under expected operating conditions. Savannah River National Laboratory (SRNL) demonstrated significantly higher filter flux for actual waste samples using a small-scale rotary filter. With funding from the U. S. Department of Energy Office of Cleanup Technology, SRNL personnel are evaluating and developing the rotary microfilter for radioactive service at SRS. The authors improved the design for the disks and filter unit to make them suitable for high-level radioactive service. They procured two units using the new design, tested them with simulated SRS wastes, and evaluated the operation of the units. Work to date provides the following conclusions and program status: (1) The authors modified the design of the filter disks to remove epoxy and Ryton{reg_sign}. The new design includes welding both stainless steel and ceramic coated stainless steel filter media to a stainless steel support plate. The welded disks were tested in the full-scale unit. They showed good reliability and met filtrate quality requirements. (2) The authors modified the design of the unit, making installation and removal easier. The new design uses a modular, one-piece filter stack that is removed simply by disassembly of a flange on the upper (inlet) side of the filter housing. All seals and rotary unions are contained within the removable stack. (3) While it is extremely difficult to predict the life of the seal, the vendor representative indicates a minimum of one year in present service conditions 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 of the current seal with an air seal could increase the lifetime to 5 years and is undergoing testing in the current work. (4) The bottom bushing showed wear due to a misalignment during the manufacture of the filter tank. Replacing the graphite bushing with a more wear resistant material such as a carbide material will increase the lifetime of the bushing. This replacement requires a more wear resistant part or coating to prevent excessive wear of the shaft. The authors are currently conducting testing with the more wear resistant bushing. (5) The project team plans to use the rotary microfilter as a filter in advance of an ion exchange process under development for potential deployment in SRS waste tank risers

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

PROCESSING ALTERNATIVES FOR DESTRUCTION OF TETRAPHENYLBORATE

Two processes were chosen in the 1980's at the Savannah River Site (SRS) to decontaminate the soluble High Level Waste (HLW). The In Tank Precipitation (ITP) process (1,2) was developed at SRS for the removal of radioactive cesium and actinides from the soluble HLW. Sodium tetraphenylborate was added to the waste to precipitate cesium and monosodium titanate (MST) was added to adsorb actinides, primarily uranium and plutonium. Two products of this process were a low activity waste stream and a concentrated organic stream containing cesium tetraphenylborate and actinides adsorbed on monosodium titanate (MST). A copper catalyzed acid hydrolysis process was built to process (3, 4) the Tank 48H cesium tetraphenylborate waste in the SRS's Defense Waste Processing Facility (DWPF). Operation of the DWPF would have resulted in the production of benzene for incineration in SRS's Consolidated Incineration Facility. This process was abandoned together with the ITP process in 1998 due to high benzene in ITP caused by decomposition of excess sodium tetraphenylborate. Processing in ITP resulted in the production of approximately 1.0 million liters of HLW. SRS has chosen a solvent extraction process combined with adsorption of the actinides to decontaminate the soluble HLW stream (5). However, the waste in Tank 48H is incompatible with existing waste processing facilities. As a result, a processing facility is needed to disposition the HLW in Tank 48H. This paper will describe the process for searching for processing options by SRS task teams for the disposition of the waste in Tank 48H. In addition, attempts to develop a caustic hydrolysis process for in tank destruction of tetraphenylborate will be presented. Lastly, the development of both a caustic and acidic copper catalyzed peroxide oxidation process will be discussed

TESTING OF A ROTARY MICROFILTER TO SUPPORT HANFORD APPLICATIONS

Savannah River National Laboratory (SRNL) researchers are investigating and developing a rotary microfilter for solid-liquid separation applications at the Savannah River Site (SRS). Because of the success of that work, the Hanford Site is evaluating the use of the rotary microfilter for its Supplemental Pretreatment process. The authors performed rotary filter testing with a full-scale, 25-disk unit with 0.5 {micro} filter media manufactured by Pall Corporation using a Hanford AN-105 simulant at solids loadings of 0.06, 0.29, and 1.29 wt%. The conclusions from this testing are: (1) The filter flux at 0.06 wt% solids reached a near constant value at an average of 0.26 gpm/ft{sup 2} (6.25 gpm total). (2) The filter flux at 0.29 wt% solids reached a near constant value at an average of 0.17 gpm/ft{sup 2} (4 gpm total). (3) The filter flux at 1.29 wt% solids reached a near constant value at an average of 0.10 gpm/ft{sup 2} (2.4 gpm total). (4) Because of differences in solids loadings, a direct comparison between crossflow filter flux and rotary filter flux is not possible. The data show the rotary filter produces a higher flux than the crossflow filter, but the improvement is not as large as seen in previous testing. (5) Filtrate turbidity measured < 4 NTU in all samples collected. (6) During production, the filter should be rinsed with filtrate or dilute caustic and drained prior to an extended shutdown to prevent the formation of a layer of settled solids on top of the filter disks. (7) Inspection of the seal faces after {approx} 140 hours of operation showed an expected amount of initial wear, no passing of process fluid through the seal faces, and very little change in the air channeling grooves on the stationary face. (8) Some polishing was observed at the bottom of the shaft bushing. The authors recommend improving the shaft bushing by holding it in place with a locking ring and incorporated grooves to provide additional cooling. (9) The authors recommend that CH2MHill Hanford test other pore size media to determine the optimum pore size for Hanford waste

DEVELOPMENT OF A CROSSFLOW FILTER TO REMOVE SOLIDS FROM RADIOACTIVE LIQUID WASTE: COMPARISON OF TEST DATA WITH OPERATING EXPERIENCE - 9119

In 2008, the Savannah River Site (SRS) began treatment of liquid radioactive waste from its Tank Farms. To treat waste streams containing {sup 137}Cs, {sup 90}Sr, and actinides, SRS developed the Actinide Removal Process (ARP) and the Modular Caustic Side Solvent Extraction Unit (MCU). The Actinide Removal Process contacts the waste with monosodium titanate (MST) to sorb strontium and select actinides. After MST contact, the process filters the resulting slurry to remove the MST (with sorbed strontium and actinides) and any entrained sludge. The filtrate is transported to the MCU to remove cesium. The solid particle removed by the filter are concentrated to {approx} 5 wt %, washed to reduce the concentration of dissolved sodium, and transported to the Defense Waste Processing Facility (DWPF) for vitrification. The authors conducted tests with 0.5 {micro} and 0.1 {micro} Mott sintered stainless steel crossflow filter at bench-scale (0.19 ft{sup 2} surface area) and pilot-scale (11.2 ft{sup 2}). The collected data supported design of the filter for the process and identified preferred operating conditions for the full-scale process (230 ft{sup 2}). The testing investigated the influence of operating parameters, such as filter pore size, axial velocity, transmembrane pressure, and solids loading, on filter flux, and validated the simulant used for pilot-scale testing. The conclusions from this work follow: (1) The 0.1 {micro} Mott sintered stainless steel filter produced higher flux than the 0.5 {micro} filter. (2) The filtrate samples collected showed no visible solids. (3) The filter flux with actual waste is comparable to the filter flux with simulated waste, with the simulated waste being conservative. This result shows the simulated sludge is representative of the actual sludge. (4) When the data is adjusted for differences in transmembrane pressure, the filter flux in the Actinide Removal Process is comparable to the filter flux in the bench-scale and pilot-scale testing. (5) Filter flux increased with transmembrane pressure, increased with axial velocity, and decreased with concentration in agreement with classical crossflow filtration theories

Impact of Irradiation on Solvent used in SRS Waste Treatment Processes -9122

ABSTRACT Savannah River Site (SRS) will use a Caustic Side Solvent Extraction (CSSX) process to selectively remove radioactive Cs-137 from the caustic High Level Waste (HLW) salt solutions stored in the large carbon steel waste tanks in the SRS Tank Farm. This HLW resulted from several decades of operations at SRS to produce nuclear materials for the United States Government. The removed Cs-137 will be sent to the Defense Waste Processing Facility (DWPF) where it will be immobilized along with the HLW sludges from the SRS Tank Farm into a borosilicate glass that will be put into permanent disposal. Currently the CSSX process is operating on an interim basis in the Modular Caustic Side Solvent Extraction Unit (MCU) facility. Eventually the process will occur in the full scale Salt Waste Processing Facility (SWPF) currently being built. The organic solvent developed for the process is primarily a mixture of the Isopar ® L (a blend of C 10 -C 12 branched alkanes such as dodecane) and an alkyl aryl polyether added as a Modifier (commonly called Cs-7SB) to enhance the solubility of the extractant which is a calixarene-crown ether. The solvent also includes trioctylamine to mitigate the adverse impact of lipophilic agents on the stripping of the cesium into nitric acid. Since the mixture is primarily organic hydrocarbons, it is expected that radiolysis of the mixture with gamma rays and beta particles from the Cs-137 will produce the flammable gas H 2 and also eventually degrade the solvent. For example, much research has been performed on the radiolysis of the organic solvent used in the tributylphosphate (TPB) extraction process (PUREX process) that has been used at SRS and in many other countries for several decades to separate U and Pu from radioactive U-235 fission products such as Cs-137. [1] The purpose of this study was to investigate the radiolysis of the organic solvent for the CSSX process. Researchers at Savannah River National Laboratory (SRNL) irradiated samples of solvent with Co-60 gamma rays. Prior to the irradiation, the solvent was contacted with the aqueous solutions that will be used in the MCU and SWPF facilities. These were the aqueous caustic salt feed, the scrub solution, and wash water. The rates of radiolytic H 2 production were measured both by determining the composition of the gases produced and by measuring pressures produced during radiolysis. The irradiated solvents were then analyzed by various analytical techniques to assess how much of the Isopar ® L, the Modifier, and the extractant had decomposed

STATUS OF CHEMICAL CLEANING OF WASTE TANKS AT THE SAVANNAH RIVER SITE F TANK FARM CLOSURE PROJECT - 9114

Chemical Cleaning is currently in progress for Tanks 5 and 6 at the Savannah River Site. The Chemical Cleaning process is being utilized to remove the residual waste heel remaining after completion of Mechanical Sludge Removal. This work is required to prepare the tanks for closure. Tanks 5 and 6 are 1950s vintage carbon steel waste tanks that do not meet current containment standards. These tanks are 22.9 meters (75 feet) in diameter, 7.5 meters (24.5 feet) in height, and have a capacity of 2.84E+6 liters (750,000 gallons). Chemical Cleaning adds 8 wt % oxalic acid to the carbon steel tank to dissolve the remaining sludge heel. The resulting acidic waste solution is transferred to Tank 7 where it is pH adjusted to minimize corrosion of the carbon steel tank. The Chemical Cleaning flowsheet includes multiple strikes of acid in each tank. Acid is delivered by tanker truck and is added to the tanks through a hose assembly connected to a pipe penetration through the tank top. The flowsheet also includes spray washing with acid and water. This paper includes an overview of the configuration required for Chemical Cleaning, the planned flowsheet, and an overview of technical concerns associated with the process. In addition, the current status of the Chemical Cleaning process in Tanks 5 and 6, lessons learned from the execution of the process, and the path forward for completion of cleaning in Tanks 5 and 6 will also be discussed

Plutonium dissolution process

A two-step process for dissolving Pu metal is disclosed in which two steps can be carried out sequentially or simultaneously. Pu metal is exposed to a first mixture of 1.0-1.67 M sulfamic acid and 0.0025-0.1 M fluoride, the mixture having been heated to 45-70 C. The mixture will dissolve a first portion of the Pu metal but leave a portion of the Pu in an oxide residue. Then, a mineral acid and additional fluoride are added to dissolve the residue. Alternatively, nitric acid between 0.05 and 0.067 M is added to the first mixture to dissolve the residue as it is produced. Hydrogen released during the dissolution is diluted with nitrogen