W. R. Wilmarth

Stability Study Of Cs Extraction Solvent

Researchers at the Savannah River Site (SRS) and Oak Ridge National Laboratory (ORNL) examined the performance and stability to irradiation of an improved calix[4]arene-based paraffin extraction solvent consisting of a calix[4]arene chelator, 1-(2,2,3,3-tetrafluoropropoxy)-3-(4-sec-butylphenoxy)-2-propanol (Cs-7SB modifier) and trioctylamine (TOA) for the removal of cesium from high-level waste. As a result of testing performed in 1998, modifications to the solvent system were made to improve chemical stability. The robustness to irradiation of the current calix[4]arene-based solvent system was examined by exposing agitated solvent in contact with aqueous phases representing the composition on the stages of extraction (alkaline waste simulant), scrub (0.05-M nitric acid), and strip (0.001-M nitric acid) to a dose up to 50 Mrad by a 60Co gamma source. The solvent was also agitated for 101 days with high-level liquid waste from the Savannah River Site (SRS) tanks. After irradiation, the concentration of the components in the solvent system and the identification of degradation products were determined primarily using gas chromatography (GC) and high-performance liquid chromatography (HPLC). The expected yearly dose the solvent will receive is 100 krad/year. An approximately 10% concentration drop in trioctylamine was observed at a 2-Mrad dose of gamma radiation, an estimate 20-year radiation dose. At gamma-radiation doses as high as 16 Mrad, or an estimated 160-year-radiation dose, there was not a significant loss of the calix[4]arene chelator (less than 10%), and there was only a minor loss (less than 2%) of Cs-7SB modifier that yields the degradation product 4-sec-butylphenol. Additional testing indicated that this phenol is readily removed by a caustic wash. Distribution coefficients for Cs between the phases (DCs) were also determined using inductively coupled plasma mass spectroscopy (ICP-MS) and gamma-counting methods. Relative to unirradiated solvent, extraction, scrubbing, and stripping performance were not significantly affected by external gamma-radiation doses as high as 8 Mrad, an estimated 80-year dose. These findings indicate that the solvent is stable and retains its expected extraction, scrubbing, and stripping properties after exposure to relatively high-gamma-radiation doses (up to 8 Mrad). #The submitted manuscript has been authored by a contractor of the U.S. Government under contract No. DE-AC09-96SR18500. Accordingly, the U.S. Government retains a non-exclusive, royalty-free license to publish or reproduce the published form of this contribution, or allow others to do so, for U.S. Government purposes.

Nitric Acid Cleaning Of A Sodalite–sodium Diuranate Scale In High Level-waste Evaporators

The operation of a high-level radioactive waste evaporator was curtailed due to the presence of an aluminosilicate scale that contained sodium diuranate with a uranium-235 enrichment of approximately 3%. Utilizing pervious work, a plan to clean the evaporator system using a heated nitric acid and depleted uranium mixture was developed that addressed numerous safety issues. The sodium aluminosilicate scale was successfully removed during two cleaning cycles, but soluble silicon was not present in measurable quantities in the liquid samples taken during cleaning. Although this phenomenon was not observed in the laboratory tests, silicon was detected in a loose, solid phase discovered in the evaporator cone after the first cleaning cycle. #This report was prepared by Westinghouse Savannah River Company (WSRC) for the United States Department of Energy under Contract No. DE-AC09-88SR18035 and is an account of work performed under that contract. Accordingly, the U.S. Government retains a non-exclusive, royalty-free license to publish or reproduce the published form of this contribution, or allow others to do so, for U.S. Government purposes. Neither the United States Department of Energy, nor WSRC, nor any of their employees makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, or product or process, disclosed herein or represents that its use will not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trademark, name, manufacturer, or otherwise does not necessarily constitute or imply endorsement, recommendation, or favoring of same by WSRC or the United States Government or any agency thereof. The views and opinions of the authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

Sodium Aluminosilicate Formation in Tank 43H Simulants

This work studied the formation of a sodium aluminosilicate, Na{sub 8}Al{sub 6}Si{sub 6}O{sub 24}(NO{sub 3}){sub 2?4}H{sub 2}O, at 40{degree} 110{degree} C in simulated waste solutions with varied amounts of silicon and aluminum. The data agree well with literature solubility data for sodalite, the analogous chloride salt. The following conclusions result from this work: (1) The study shows, by calculation and experiments, that evaporation of the September 1997 Tank 43H inventory will only form minor quantities of the aluminosilicate. (2) The data indicate that the rate of formation of the nitrate enclathrated sodalite solid at these temperatures falls within the residence time ({lt}; 4 h) of liquid in the evaporator. (3) The silicon in entrained Frit 200 transferred to the evaporator with the Tank 43H salt solution will quantitatively convert to the sodium aluminosilicate. One kilogram of Frit 200 produces 2.1 kg of the sodium aluminosilicate.

WASTE TREATMENT TECHNOLOGY PROCESS DEVELOPMENT PLAN FOR HANFORD WASTE TREATMENT PLANT LOW ACTIVITY WASTE RECYCLE

The purpose of this Process Development Plan is to summarize the objectives and plans for the technology development activities for an alternative path for disposition of the recycle stream that will be generated in the Hanford Waste Treatment Plant Low Activity Waste (LAW) vitrification facility (LAW Recycle). This plan covers the first phase of the development activities. The baseline plan for disposition of this stream is to recycle it to the WTP Pretreatment Facility, where it will be concentrated by evaporation and returned to the LAW vitrification facility. Because this stream contains components that are volatile at melter temperatures and are also problematic for the glass waste form, they accumulate in the Recycle stream, exacerbating their impact on the number of LAW glass containers. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and reducing the halides in the Recycle is a key component of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, this stream does not have a proven disposition path, and resolving this gap becomes vitally important. This task seeks to examine the impact of potential future disposition of this stream in the Hanford tank farms, and to develop a process that will remove radionuclides from this stream and allow its diversion to another disposition path, greatly decreasing the LAW vitrification mission duration and quantity of glass waste. The origin of this LAW Recycle stream will be from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover or precipitates of scrubbed components (e.g. carbonates). The soluble components are mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet, and will not be available until the WTP begins operation, causing uncertainty in its composition, particularly the radionuclide content. This plan will provide an estimate of the likely composition and the basis for it, assess likely treatment technologies, identify potential disposition paths, establish target treatment limits, and recommend the testing needed to show feasibility. Two primary disposition options are proposed for investigation, one is concentration for storage in the tank farms, and the other is treatment prior to disposition in the Effluent Treatment Facility. One of the radionuclides that is volatile and expected to be in high concentration in this LAW Recycle stream is Technetium-99 ({sup 99}Tc), a long-lived radionuclide with a half-life of 210,000 years. Technetium will not be removed from the aqueous waste in the Hanford Waste Treatment and Immobilization Plant (WTP), and will primarily end up immobilized in the LAW glass, which will be disposed in the Integrated Disposal Facility (IDF). Because {sup 99}Tc has a very long half-life and is highly mobile, it is the largest dose contributor to the Performance Assessment (PA) of the IDF. Other radionuclides that are also expected to be in appreciable concentration in the LAW Recycle are {sup 129}I, {sup 90}Sr, {sup 137}Cs, and {sup 241}Am. The concentrations of these radionuclides in this stream will be much lower than in the LAW, but they will still be higher than limits for some of the other disposition pathways currently available. Although the baseline process will recycle this stream to the Pretreatment Facility, if the LAW facility begins operation first, this stream will not have a disposition path internal to WTP. One potential solution is to return the stream to the tank farms where it can be evaporated in the 242-A evaporator, or perhaps deploy an auxiliary evaporator to concentrate it prior to return to the tank farms. In either case, testing is needed to evaluate if this stream is compatible with the evaporator and the other wastes in the tank farm. It should be noted that prior experience in evaporation of another melter off-gas stream, the Recycle Stream at the SRS Defense Waste Processing Facility, unexpectedly caused deleterious impacts on evaporator scaling and formation of aluminosilicate solids before controls were implemented. The compatibility of this stream with other wastes and components in the tank farms has not been fully investigated, whether it is sent for storage in AW-102 in preparation for evaporation in 242-A evaporator, or if it is pre-concentrated in an auxiliary evaporator. This stream is expected to be unusual because it will be very high in corrosive species that are volatile in the melter (chloride, fluoride, sulfur), will have high ammonia, and will contain carryover particulates of glass-former chemicals. These species have potential to cause corrosion, precipitation, flammable gases, and scale in the tank farm system. Testing is needed to demonstrate acceptable conditions and limits for these compounds in wastes sent to the tank farms. Alternate disposition of this LAW Recycle stream could beneficially impact WTP, and may also remove a sizeable fraction of the 99Tc from the source term at the IDF. The alternative radionuclide removal process envisioned for this stream parallels the Actinide Removal Process that has been successfully used at SRS for several years. In that process, Monosodium Titanate (MST) is added to the tank waste to adsorb 90Sr and actinides, and then the MST and radionuclides are removed by filtration. The process proposed for investigation for the Hanford WTP LAW Recycle stream would similarly add MST to remove 90Sr and actinides, along with other absorbents or precipitating agents for the remaining radionuclides. These include inorganic reducing agents for Tc, and zeolites for 137Cs. After treatment, disposition of the decontaminated Recycle stream may be suitable for the Effluent Treatment Facility, where it could be evaporated and solidified. The contaminated slurry stream containing the absorbents and radionuclides will be preliminarily characterized in this phase of the program to evaluate disposal options, and disposition routes will be tested in the next phase. The testing described herein will aid in selection of the best disposal pathway. Several research tasks have been identified that are needed for this initial phase: Simulant formulation-  Concentration of Recycle to reduce storage volume; Blending of concentrated Recycle with tank waste; Sorption of radionuclides; Precipitation of radionuclides. After this initial phase of testing, additional tasks are expected to be identified for development. These tasks likely include evaluation and testing of applicable solid-liquid separation technologies, slurry rheology measurements, composition variability testing and evaluations, corrosion and erosion testing, slurry storage and immobilization investigations, and decontaminated Recycle evaporation and solidification. Although there are a number of unknown parameters listed in the technical details of the concepts described here, many of these parameters have precedence and do not generally require fundamental new scientific breakthroughs. Many of the materials and processes described are already used in radioactive applications in the DOE complex, or have been tested previously in comparable conditions. Some of these materials and equipment are already used in High Level Waste applications, which are much more complex and aggressive conditions than the LAW Recycle stream. In some cases, the unknown parameters are simply extensions of already studied conditions, such as tank waste corrosion chemistry. The list of testing needs at first appears daunting, but virtually all have been done before, although there are potential issues with compatibility with this unique waste stream. It is anticipated that the challenge will be more in integrating the system and complying with process limitations than in developing entirely new technologies. Several assumptions have been made in this document about the acceptability of radionuclide decontamination and potential waste forms for disposal. These assumptions have been used to define acceptability criteria for feasibility studies on removal. These limits are not intended to define regulatory or facility limits, but rather provide a starting point for evaluating various technologies.

Laboratory Scoping Tests Of Decontamination Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task seeks to examine the potential treatment of this stream to remove radionuclides and subsequently disposition the decontaminated stream elsewhere, such as the Effluent Treatment Facility (ETF), for example. The treatment process envisioned is very similar to that used for the Actinide Removal Process (ARP) that has been operating for years at the Savannah River Site (SRS), and focuses on using mature radionuclide removal technologies that are also compatible with longterm tank storage and immobilization methods. For this new application, testing is needed to demonstrate acceptable treatment sorbents and precipitating agents and measure decontamination factors for additional radionuclides in this unique waste stream. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet and will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. One of the radionuclides that is volatile and expected to be in high concentration in this LAW Off-Gas Condensate stream is Technetium-99 ({sup 99}Tc). Technetium will not be removed from the aqueous waste in the Hanford WTP, and will primarily end up immobilized in the LAW glass by repeated recycle of the off-gas condensate into the LAW melter. Other radionuclides that are also expected to be in appreciable concentration in the LAW Off-Gas Condensate are {sup 129}I, {sup 90}Sr, {sup 137}Cs, and {sup 241}Am. This report discusses results of preliminary radionuclide decontamination testing of the simulant. Testing examined use of Monosodium Titanate (MST) to remove {sup 90}Sr and actinides, inorganic reducing agents for {sup 99}Tc, and zeolites for {sup 137}Cs. Test results indicate that excellent removal of {sup 99}Tc was achieved using Sn(II)Cl{sub 2} as a reductant, coupled with sorption onto hydroxyapatite, even in the presence of air and at room temperature. This process was very effective at neutral pH, with a Decontamination Factor (DF) >577 in two hours. It was less effective at alkaline pH. Conversely, removal of the cesium was more effective at alkaline pH, with a DF of 17.9. As anticipated, ammonium ion probably interfered with the Ionsiv®a IE-95 zeolite uptake of {sup 137}Cs. Although this DF of {sup 137}Cs was moderate, additional testing is expected to identify more effective conditions. Similarly, Monosodium Titanate (MST) was more effective at alkaline pH at removing Sr, Pu, and U, with a DF of 319, 11.6, and 10.5, respectively, within 24 hours. Actually, the Ionsiv® IE-95, which was targeting removal of Cs, was also moderately effective for Sr, and highly effective for Pu and U at alkaline pH. The only deleterious effect observed was that the chromium co-precipitates with the {sup 99}Tc during the SnCl{sub 2} reduction. This effect was anticipated, and would have to be considered when managing disposition paths of this stream. Results of this separation testing indicate that sorption/precipitation was a viable concept and has the potential to decontaminate the stream. All radionuclides were at least partially removed by one or more of the materials tested. Based on the results, a possible treatment scenario could involve the use of a reductive precipitation agent (SnCl{sub 2}) and sorbent at neutral pH to remove the Tc, followed by pH adjustment and the addition of zeolite (Ionsiv® IE-95) to remove the Cs, Sr, and actinides. Addition of MST to remove Sr and actinides may not be needed. Since this was an initial phase of testing, additional tasks to improve separation methods were expected to be identified. Primarily, further testing is needed to identify the conditions for the decontamination process. Once these conditions are established, follow-on tasks likely include evaluation and testing of applicable solid-liquid separation technologies, slurry rheology measurements, composition variability testing and evaluations, corrosion and erosion testing, slurry storage and immobilization investigations, and decontaminated LAW Off-Gas Condensate evaporation and solidification.

Caustic-side Solvent Extraction Batch Distribution Coefficient Measurements For Savannah River Site High-level Wastes

The cesium distribution coefficients for extraction, scrubbing, and stripping of the caustic-side solvent extraction flowsheet were measured for a number of tank waste supernates from both the F- and H-Area Tank Farms. The measured distribution coefficients, DCss, indicate that the caustic-side solvent extraction flowsheet will successfully decontaminate the supernate and produce a concentrated, dilute nitric acid product. Because of the difficulty in performing the experiments on a 40-mL scale in the shielded cells, some caustic carryover into the scrub acid occurred and contributed to the higher-than-expected scrub distribution coefficients. Comparing the measured extraction distribution coefficients to recently published data by Oak Ridge National Laboratory (ORNL) personnel showed agreement between the model and experiments with two actual tank wastes. However, tests with three other tank wastes gave extraction DCss that deviated from the model predictions. Additional refinements of the model are planned in FY 2002. Two of these measurements were below the flowsheet requirement for extraction. Several aspects of the waste chemistry (e.g., anion concentrations) were examined; however, no distinct correlation was found for extraction distribution coefficient behavior. Testing also examined the affinity of the calixarene-solvent system for actinide removal. Plutonium and uranium extraction distribution coefficients measured around 2. However, additional analytical work is needed to measure the mass of the actinides in the organic phase (Wilmarth, W.R., Hobbs, D.T. Task Technical and Quality Assurance Plan Supporting CSSX Pilot Plant Critically Issues; WSRC-RP-2001-0076; Westinghouse Savannah River: Aiken, SC, 2001).[1] #This report was prepared for the United States Department of Energy under Contract No. DE-AC09-96SR18500 and is an account of work performed under that contract. Reference herein to any specific commercial product, process, or service trademark or otherwise does not necessarily constitute or imply endorsement, recommendation, or favoring of same by Westinghouse Savannah River Company or by the United States Government or any agency thereof. The views and opinions of the authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

Sr/TRU REMOVAL FROM HANFORD HIGH LEVEL WASTE

The originally proposed removal process for strontium and transuranic species from 241-AN-102 and 241-AN-107 Hanford waste tank supernates was a co-precipitation method. In initial testing, the slurry formed during the strontium and ferric nitrate co-precipitant additions was not filterable. A series of statistically designed tests were performed to evaluate the variables responsible for this poor filterability. These tests also explored strategies to improve the associated decontamination efficiency of the treatment process. These tests revealed that filterability is negatively influenced by ferric ion added to co-precipitate actinide and lanthanide species. The concentrations of sodium and organic complexants were observed to influence decontamination of the supernate. Furthermore, the amount of at least one organic complexant was correlated to the poor filterability. The americium decontamination factors measured following the maximum iron nitrate addition were marginal at levels of precipitant addition that are of practical interest. Based on these results, alternative precipitation schemes were investigated. Reported here is the chosen replacement process, a permanganate precipitation process. A statistically designed series of experiments examined the relationship between three responses and five precipitation parameters. The three responses are precipitate filterability, strontium decontamination, and plutonium decontamination. The parameters varied were the initial sodium and hydroxide concentrations of the waste, and the amount of calcium, strontium, and permanganate introduced. The results reveal an optimum set of conditions for decontamination of strontium and plutonium as well as improved filterability.

Strontium And Transuranic Precipitation And Crossflow Filtration Of A Large Hanford Tank 241-An-102 Sample

This work provides an important confirmation of the strontium/permanganate precipitation process to achieve both acceptable filterability and decontamination for Envelope C (Tanks 241-AN-102 and 241-AN-107) complexant wastes to be treated by the Hanford River Protection Project. This bench-scale demonstration contained a series of seven precipitation batches and crossflow filtration campaigns to decontaminate filtrate of Sr-90 and transuranics from 16.5 L of Tank 241-AN-102 supernatant liquid with entrained solids. Batches were caustic adjusted, strontium and permanganate precipitated, and crossflow filtered with entrained solids using a 2-ft long, 3/8″-internal diameter, 0.1-micron pore size Mott crossflow filter tube. Test ranges for the transmembrane pressures and crossflow velocities were in the range of 30 to 70 psid (2.07 to 4.83 bar) and 9 to 15 ft/s (2.7 to 4.6 m/s), respectively. #The submitted manuscript has been authored by a contractor of the U.S. Government under contract No. DE-AC09-96SR18500. Accordingly, the U.S. Government retains a non-exclusive, royalty-free license to publish or reproduce the published form of this contribution, or allow others to do so, for U.S. Government purposes.

THE EFFECT OF PRESSURE, HUMIDITY, CAUSTIC PRETREATMENT, AND ORGANIC CONSTITUENTS ON THE CESIUM ION EXCHANGE PERFORMANCE OF IONSIV® IE-911

We examined 137Cs exchange of crystalline silicotitanate (CST) in simulated waste solution. In particular, we focused on the effect of CST pretreatment on the kinetics and extent of cesium adsorption. We used IONSIV®IE-911 (UOP LLC, Molecular Sieves Division, Des Plaines, Ill), the engineered form of CST. Pretreatment steps examined include soaking IE-911 in 2 mol/L NaOH solution for 3 days, exposing IE-911 to 50% relative humidity for 1 week, and for 3 days, soaking IE-911 in organic-containing simulated salt solution or drying untreated IE-911 in air at 100°C. Some tests were conducted with the sample under 50 and 25 psig of argon. Pretreatment of IE-911 in 2 mol/L caustic solution for 72 hours yielded a slower approach to equilibrium cesium distribution in batch contact tests than the untreated IE-911 samples. Adsorbed carboxylates and carbonates likely affected the cesium transport by either increasing the path length or reducing mass transfer rate. However, the effect was completely removed when IE-911 was rinsed with deionized water. Heating IE-911, as received from the vendor, at 100°C for 72 hours significantly degraded its cesium-removal performance by a 40.7% reduction in capacity and a 43% reduction in sorption rate (relative to the untreated IE-911) over 1 week of testing. However, sodium conversion of these samples did not affect cesium sorption. The presence of organic chemicals (e.g., tri-n-butyl phosphate, dibutylphosphate, butanol, paraffin, and Dow Corning H-10 defoamer) in simulated salt solution did not affect cesium sorption on pretreated IE-911. Nearly identical (i.e., a difference of only 5.6%) distribution coefficients (K d) were found between lot no. 9990-9681-0004 and no. 9990-9881-0005. Increasing the atmospheric pressure from 0 to 50 psig had no effect on cesium sorption.

Transuranium removal from Hanford High Level Waste simulants using sodium permanganate and calcium

Hanford High Level Waste will require processing to reduce the concentration of various actinide elements prior to encapsulation into low activity glass for disposition. High level waste at Hanford contains elevated actinide concentrations in the supernate because of organic complexants present in the tanks. Traditional removal strategies are not viable processing sequences for the Hanford tanks containing complexants. Reported here is a novel actinide decontamination strategy. This pretreatment sequence consists of addition of calcium nitrate and sodium permanganate. The observed optimum decontamination efficiencies for plutonium and americium are more than 99.5%.