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Researchers investigated the neutralization of neptunium solution from H-Canyon Tank 16.4 and the properties of the resulting slurry. This work investigated slurry properties from a single neutralization protocol and limited storage times.

The present study identifies organic compounds or classes of compounds that may occur in SRS high-level waste (HLW) based on a review of previous sample analyses and records of disposals of organic compounds to the HLW tank farms.

solutions in two laboratory experiments. The first experiment tested four waste solutions for supersaturation of aluminum by monitoring the aluminum concentration after seeding with gibbsite. The second experiment tested two waste samples for precipitation of aluminosilicates by heating the solutions to accelerate solids formation. The results of the experiments with actual waste solutions are supported in this report.

Researchers characterized high-level waste derived from dissolved salt cake from the Savannah River Site (SRS) tank farm and measured the cesium distribution coefficients (DCs) for extraction, scrub, and stripping steps of the caustic-side solvent extraction (CSSX) flowsheet. The measurements used two SRS high-level waste samples derived entirely or in part from salt cake. The chemical compositions of both samples are reported. Dissolved salt cake waste contained less Cs-137 and more dianions than is typical of supernate samples. Extraction, scrub, and strip DCs values for both samples exceeded process requirements and agreed well with model predictions. The results indicate no significant problems processing dissolved salt cake compared to supernate. During the course of testing, researchers observed sorption of cesium on glass sample vials containing strip solutions. The problem was detected in the material balance calculated for each organic/aqueous contact. Methods and recommendations for avoiding this problem are discussed.

This report details the results of analyses for organic compounds and trace components in various process streams based on samples from the 48-hour radioactive waste test.

Researchers successfully demonstrated the chemistry and process equipment of the Caustic-Side Solvent Extraction (CSSX) flowsheet for the decontamination of high level waste using a 33-stage, 2-cm centrifugal contactor apparatus at the Savannah River Technology Center. This represents the first CSSX process demonstration using Savannah River Site (SRS) high level waste. Three tests lasting 6, 12, and 48 hours processed simulated average SRS waste, simulated Tank 37H/44F composite waste, and Tank 37H/44F high level waste, respectively.

This report describes cesium desorption from IONSIV IE-911 during ambient temperature storage and following temperature increases to 35 and 55 degrees C. This report also describes cesium sorption following return to ambient temperature. The IONSIV IE-911 used in these tests was loaded with cesium from Tank 44F radioactive waste in an ion exchange column test in 1999. Cesium desorbed and resorbed in the presence of Tank 44F waste and simulated waste solutions.

Previous studies indicate that palladium catalyzes rapid decomposition of alkaline tetraphenylborate slurries. Oxygen inhibits the reaction at low temperature (25 C), presumably by preventing activation of the catalyst. The present study investigated oxygen's inhibiting effectiveness at higher temperature (45 C) and catalyst concentrations.

This study measured the ability of crystalline silicotitanate to remove cesium from Savannah River Site radioactive waste.

Observations during experiments with UOP IONSIV(TM) IE-911 during the past year and information obtained during technical discussions with UOP, LLC, personnel in June 8, 1999, generated technical questions concerning the application of crystalline silicotitanate (CST) to Savannah River Site (SRS) soluble waste decontamination.

Waste Processing Technology personnel routinely prepare 0.5 to 10 L batches of salt solutions simulating Savannah River Site (SRS) soluble waste. This report describes the compositions and preparation methods.

Non-elutable ion exchange using crystalline silicotitanate is being considered for removing cesium from Savannah River Site radioactive waste. The construction cost of this process depends strongly on the size of the ion exchange column required to meet product specifications.

Non-elutable ion exchange is being considered as a potential replacement for the In-Tank Precipitation process for removing cesium from Savannah River Site (SRS) radioactive waste. Crystalline silicotitanate (CST) particles are the reference ion exchange medium for the process. A major factor in the construction cost of this process is the size of the ion exchange column required to meet product specifications for decontaminated waste. To validate SRS column sizing calculations, SRS subcontracted two reknowned experts in this field to perform similar calculations: Professor R. G. Anthony, Department of Chemical Engineering, Texas A&038;M University, and Professor S. W. Wang, Department of Chemical Engineering, Purdue University. The appendices of this document contain reports from the two subcontractors. Definition of the design problem came through several meetings and conference calls between the participants and SRS personnel over the past few months. This document summarizes the problem definition and results from the two reports.

This study measured the adsorption of cesium from simulated Savannah River Site liquid waste onto crystalline silicotitanate (CST) in equilibrium (Kd) and ion exchange column tests.

Addition of sodium tetraphenylborate to Tank 48H will reduce the soluble Cs-137 concentration to below 12 nCi/g, well within the pending Process Requirement limit of 85 nCi/g.

A sorbent containing crystalline silicotitanate (CST) tested for cesium removal from simulated Savannah River Site (SRS) soluble high activity waste showed rapid kinetics (1 h contact time) and high distribution coefficients (Kd 4000 mL/g of CST). The sorbent was prepared by Selective Environmental Technologies, Inc., (Selentec) as a MAG*SEP particle containing CST obtained from the Molecular Sieve Department of UOP, LLC, Results of preliminary tests suggest potential applications of the Selentec MAG*SEP particles to radioactive waste decontamination at SRS.

This report serves two purposes. First, it documents the analytical results of Tank 48H samples taken between April and August 1996. Second, it describes investigations of the precision of the sampling and analytical methods used on the Tank 48H samples.

High-level radioactive wastes are stored in large, steel tanks in the Savannah River Site`s Tank Farms. The liquid levels in these tanks are monitored to detect leakage of waste out of tanks or leakage of liquids into the tanks. Recent unexplained level fluctuations in high-level waste (HLW) tanks have caused High Level Waste Engineering (HLWE) to develop a program to better understand tank level behavioral Interim Waste Technology (IWT) has been requested by HLWE to obtain data which will lead to a better understanding of the radiolytic generation of gases in salt cake. A task plan has been written in response to this request. This document details the controls necessary to ensure the quality of the results of the activities described in the task plan.

High-level radioactive wastes are stored in large, steel tanks in the Savannah River Site Tank Farms. The liquid levels in these tanks are monitored to detect leakage of waste out of tanks or leakage of liquids into the tanks. Recent unexplained level fluctuations in high-level waste (HLW) tanks have caused High Level Waste Engineering (HLWE) to develop a program to better understand tank level behavior. Interim Waste Technology (IWT) has been requested by HLWE to obtain data which will lead to a better understanding of the radiolytic generations of gases in salt cake. The task described below will provide data from laboratory experiments with simulated wastes which can be used in tank level fluctuation modeling. The following experimental programs have been formulated to meet the task requirements of the customer: (A) determine whether radiolytically generated gas bubbles can be trapped in salt cake; (B) determine the composition of gases produced by radiolysis; (C) determine the yield of radiolysis gases as a function of radiation dose; (D) determine bubble distribution.

During the ITP cold chemical testing program, the efficiency of the benzene strippers will be measured and evaluated. Since the stripping efficiency is partially dependent upon the dynamic viscosity of the liquid phase, this property must be measured on samples taken during the test program. A procedure to measure the dynamic viscosity of salt solutions was developed from standard American Society of Testing and materials (ASTM) methods. The SRS procedure differs from the ASTM procedure and, therefore, a test program was initiated to determine its accuracy and precision. The results of these statistically designed tests are reported elsewhere, but supporting information on the experimental procedures, standards, and equipment are given in this report.

If late washing of precipitate is available, it is possible to modify the normal washing phase of the ITP process so that tank corrosion is prevented by inhibiting with sodium hydroxide rather than sodium nitrite. Hydroxide inhibition has numerous advantages to a hydroxide/nitrite flowsheet.1 However, the rate of hydroxide depletion due to radiolysis and C0{sub 2} absorption were uncertainties. Based on recent experiments and calculations: hydroxide consumption by radiolysis will be 0.01 molar per month during Tank 49 storage, hydroxide depletion due to C0{sub 2} absorption will vary from 0.0006 to 0.025 molar per month for waste volumes between 50,000 and 1 million gallons and air flowrates between 100 and 200 cfm. A nominal rate of 0.006 molar/month (or less) is expected in Tank 49 after the first two ITP cycles have been completed. A material balance for the ITP process based on hydroxide inhibition has been calculated and the potential savings have been estimated.

Adsorption of Pu and U onto monosodium titanate (MST) in alkaline salt solution was measured. Changes in MST particle size distribution do not significantly affect the loadings of actinides. Max loading of plutonium-239 is 0.68 wt% at 0.79 mg/L Pu, below the infinitely safe value of 0.80 wt%. Max loading of uranium-235 onto MST is 1.4 wt% at a concentration of 20.3 {plus_minus} 2.0 mg/L U, slightly higher than the calculated infinitely safe limit of 1.2 wt%. Experimental data indicated there is competition between plutonium and uranium for sites on the MST, and that the loading will favor the higher concentration species. Since the solubility of uranium is 10--100 times higher than of plutonium, uranium will be loaded to its maximum limit, but plutonium will be below its maximum limit. To ensure that the concentration of fissile materials cannot exceed nuclear safety limits, it is recommended that plutonium and uranium solubility tests be conducted with solutions which bound the compositions of waste which will be treated in ITP process.

The rate at which hydrogen is produced from the radiolysis of high-activity waste solutions depends on the radiation dose rate to the solution and the concentration of nitrate ion in solution. At a constant dose rate, the hydrogen generation rate decreases as the nitrate ion concentration increases. Using previously measured rates of hydrogen production, an equation has been derived which calculates the hydrogen generation rate when the nitrate ion concentration is known. The hydrogen generation rate from this equation can be used to more accurately predict hydrogen production in waste tanks than is currently obtained assuming a fixed generation rate. The current fixed rate underestimates the hydrogen produced in dilute waste solutions.

Borosilicate glass containing radioactive wastes from the Savannah River Plant have been leached for 900 days. The International Standards Organization's (ISO) static leach test procedure was used on glass buttons in various leachants. Leach rates based on /sup 90/Sr and /sup 137/Cs analyses were similar: 2 x 10/sup -8/ to 3 x 10/sup -8/ g/(cm/sup 2/)(d) in distilled water, 1 x 10/sup -8/ to 3 x 10/sup -7/ g/(cm/sup 2/)(d) in pH 7 buffer, 3 x 10/sup -7/ to 7 x 10/sup -7/ g/(cm/sup 2/)(d) in pH 9 buffer, and 7 x 10/sup -6/ to 8 x 10/sup -5/ g/(cm/sup 2/)(d) in pH 4 buffer. Rates based on Pu analyses were the same as above in distilled water and pH 9 buffer, but were lower by an order of magnitude in pH 4 and pH 7 buffers. Almost all leach rates remained constant between 200 and 900 days of leaching. Increasing the concentration of the buffering agents had no effect on the leach rates at pH 7 (phosphate) and pH 9 (carbonate), but dramatically increased the rates at pH 4 (acetate). Leach rates did not differ significantly between high aluminum and high iron waste glasses.

Liquid slurries expected under normal in-tank processing (ITP) operations are not ignitible because of their high water content. However, deposits of dry solids from the slurries are combustible and produce dense, black smoke when burned. The dry solids burn similarly to Styrofoam and more easily than sawdust. It is the opinion of fire hazard experts that a benzene vapor deflagration could ignite the dry solids. A tetraphenylborate solids fire will rapidly plug the waste tank HEPA ventilation filters due to the nature of the smoke produced. To prevent ignition and combustion of these solids, the waste tanks have been equipped with a nitrogen inerting system.