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Leaching of iron from Hanford tank sludge: Results of FY 1997 studies

Description: This report describes tests conducted at Pacific Northwest National Laboratory to investigate selective leaching of iron from Hanford tank sludges. As Fe represents a significant fraction of the sludge mass, its removal would reduce the mass of the material needed to be handled as high-level waste. Iron leaching can be viewed as an enhancement to the baseline method of caustic leaching, which primarily removes Al, P, and Cr. The U.S. Department of Energy funded the work through the Tanks Focus Area (TFA;EM-50).
Date: December 1, 1997
Creator: Lumetta, G.J.
Partner: UNT Libraries Government Documents Department

Pretreatment of neutralized cladding removal waste sludge: Results of the second design basis experiment

Description: For several years, the Pacific Northwest Laboratory (PNL) has been investigating methods to pretreat Hanford neutralized cladding removal waste (NCRW) sludge. In the past, Zircaloy-clad metallic U fuel was chemically decladded using the Zirflex process; NCRW sludge was formed when the decladding solution was neutralized for storage in carbon-steel tanks. This sludge, which is currently stored in Tanks 103-AW and 105-AW on the Hanford Site, primarily consists of insoluble Zr hydroxides and/or oxides and NaF. Significant quantities of Al, La, U, as well as other insoluble minor constituents are present in the sludge, along with sodium and potassium nitrates, nitrites, and hydroxides in the interstitial liquid. The sludge contains about 2,000 nCi of transuranic (TRU) material per gram of dry sludge, and mixed fission products. Therefore, the sludge must be handled as high-level waste (HLW). The NCRW sludge must be pretreated before treatment (e.g., vitrification) and disposal, so that the overall cost of disposal can be minimized. The NCRW pretreatment flowsheet was designed to achieve the following objectives: (a) to separate Am and Pu from the major sludge constituents (Na, Zr). (b) to separate Am and Pu from U. (c) to concentrate Am and Pu in a small volume for immobilization in borosilicate glass, based on Hanford Waste Vitrification Plant (HWVP). The flowsheet involves: (1) sludge washing, (2) sludge dissolution, (3) extraction of U with tributyl phosphate (TBP), and (4) extraction of TRUs with octyl(phenyl)-N,N-diisobutlycarbamoylmethyl-phosphine oxide (CMPO). As presented in the flowsheet, the NCRW sludge is first washed with 0.I M NaOH to remove interstitial liquid and soluble salts from the sludge including sodium and potassium fluorides, carbonates, hydroxides, nitrates, and nitrites. The washed sludge is then subjected to two dissolution steps to achieve near complete dissolution of Zr.
Date: May 1, 1994
Creator: Lumetta, G. J.
Partner: UNT Libraries Government Documents Department

Pretreatment of neutralized cladding removal waste (NCRW) sludge - results of FY 1991 studies

Description: Neutralized cladding removal waste (NCRW) sludge is a unique waste material that is stored in two underground double-shell tanks at the U.S. Department of Energy's Hanford Site. The NCRW sludge was formed by neutralization of the solution resulting from the chemical decladding of Zircaloy-clad metallic uranium fuel by the Zirflex process. The sludge consists of zirconium and sodium hydroxides and fluorides, with small amounts of potassium, nitrite, and other nonradioactive materials. The sludge also contains uranium, transuranic (TRU) elements, and mixed fission products typical of the nonvolatiles present in irradiated fuel. The NCRW sludge is considered a TRU waste, which must be vitrified for ultimate disposal in a geologic repository. The TRU portion of the waste may be separated from the larger amount of bulk waste material so only the TRU portion would require vitrification and geologic disposal. Separation would significantly reduce waste disposal costs. Work is underway to develop the transuranic extraction (TRUEX) process. This solvent extraction process has been demonstrated to separate a large percentage of the TRU elements from the bulk components of NCRW sludge. Earlier studies identified potential problems in the TRUEX processing of NCRW sludge: potential corrosion of imbedded piping in the facility initially planned for the process, instability of dissolved NCRW solutions towards precipitation, formation of interfacial crud during the TRUEX solvent extraction step, and the amount of phosphorus in the TRU product stream. These four problems were studied in FY 1991 and the results indicate that: a solution of 2 M HNO[sub 3] at a F/(Zr + Al) ratio of about 2 adequately dissolves washed NCRW sludge; such solutions should not be corrosive towards stainless steel materials; dissolved NCRW sludge solutions obtained by dissolution of washed sludge at low F/(Zr + Al) ratios (about 2) are much more stable with respect to precipitation.
Date: April 1, 1993
Creator: Lumetta, G.J. & Swanson, J.L.
Partner: UNT Libraries Government Documents Department

Pretreatment of Plutonium Finishing Plant (PFP) sludge: Report for the period October 1990--March 1992

Description: The current mission of the US Department of Energy's Hanford Site is one of environmental restoration. A major task within this mission is the disposal of large volumes of high-level wastes (HLW) that are stored in underground tanks on the site. Under the current planning assumptions, all high-level tank waste will be vitrified as borosilicate glass and then disposed of in a geologic repository. The costs associated with this disposal scheme are very high. Thus, methods to reduce the volume of glass required to vitrify these wastes are currently being investigated. Plutonium Finishing Plant (PFP) sludge is a unique transuranic waste that is stored in tank 241- SY-102 on the Hanford site. As the name implies, the bulk of this material consists of waste from operations at the Plutonium Finishing Plant; but, other wastes have also been added (e.g., wastes from decontamination activities). Because the quantities of plutonium and americium in the PFP sludge are greater than 100 nCi/g, this sludge must be handled as a HLW. Approximately 6000 glass canisters would result from vitrifying this waste directly. Sludge washing would reduce the required number of canisters to [approximately]2500, with the volume of glass being driven by the low allowable concentration limit for Cr in the vitrification plant feed. The cost of production and subsequent geologic disposal of each canister of glass is expected to be $0.5 M to $1 M. Thus, an economic incentive exists to develop methods of pretreating the sludge to reduce the number of glass canisters needed to contain the final vitrified product.
Date: April 1, 1993
Creator: Lumetta, G.J. & Swanson, J.L.
Partner: UNT Libraries Government Documents Department

Pretreatment of neutralized cladding removal waste sludge: Status Report

Description: This report describes the status of process development for pretreating Hanford neutralized cladding removal waste (NCRW) sludge, of which [approximately] 3.3 [times] 10[sup 6] L is stored in Tanks 103-AW and 105-AW at the Hanford Site. The initial baseline process chosen for pretreating NCRW sludge is to dissolve the sludge in nitric acid and extract the -transuranic (MU) elements from the dissolved sludge solution with octyl(phenyl)-N,N-diisobutylcarbamoyl methyl phosphine oxide (CNWO). This process converts the NCRW sludge into a relatively large volume of low-level waste (LLW) to be disposed of as grout, leaving only a small volume of high-level waste (HLW) requiring vitrification in the Hanford Waste Vitrification Plant (HWVP).
Date: March 1, 1993
Creator: Lumetta, G J & Swanson, J L
Partner: UNT Libraries Government Documents Department

Pretreatment of neutralized cladding removal waste (NCRW) sludge: Results of FY 1991 studies

Description: Neutralized cladding removal waste (NCRW) sludge is a unique waste material that is stored in two underground double-shell tanks at the US Department of Energy`s Hanford Site. The NCRW sludge was formed by neutralization of the solution resulting from the chemical decladding of Zircaloy-clad metallic uranium fuel by the Zirflex process. The sludge consists of zirconium and sodium hydroxides and fluorides, with small amounts of potassium, nitrite, and other nonradioactive materials. The sludge also contains uranium, transuranic (TRU) elements, and mixed fission products typical of the nonvolatiles present in irradiated fuel. The NCRW sludge is considered a TRU waste, which must be vitrified for ultimate disposal in a geologic repository. The TRU portion of the waste may be separated from the larger amount of bulk waste material so only the TRU portion would require vitrification and geologic disposal. Separation would significantly reduce waste disposal costs. Work is underway to develop the transuranic extraction (TRUEX) process. This solvent extraction process has been demonstrated to separate a large percentage of the TRU elements from the bulk components of NCRW sludge. Earlier studies identified potential problems in the TRUEX processing of NCRW sludge: potential corrosion of imbedded piping in the facility initially planned for the process, instability of dissolved NCRW solutions towards precipitation, formation of interfacial crud during the TRUEX solvent extraction step, and the amount of phosphorus in the TRU product stream. These four problems were studied in FY 1991 and the results indicate that: a solution of 2 M HNO{sub 3} at a F/(Zr + Al) ratio of about 2 adequately dissolves washed NCRW sludge; such solutions should not be corrosive towards stainless steel materials; dissolved NCRW sludge solutions obtained by dissolution of washed sludge at low F/(Zr + Al) ratios (about 2) are much more stable with respect to precipitation.
Date: April 1, 1993
Creator: Lumetta, G. J. & Swanson, J. L.
Partner: UNT Libraries Government Documents Department

Laboratory development of sludge washing and alkaline leaching processes: Test plan for FY 1994

Description: The US Department of Energy plans to vitrify (as borosilicate glass) the large volumes of high-level radioactive wastes at the Hanford site. To reduce costs, pretreatment processes will be used to reduce the volume of borosilicate glass required for disposal. Several options are being considered for the pretreatment processes: (1) sludge washing with water or dilute hydroxide: designed to remove most of the Na from the sludge, thus significantly reducing the volume of waste to be vitrified; (2) sludge washing plus caustic leaching and/or metathesis (alkaline sludge leaching): designed to dissolve large quantities of certain nonradioactive elements, such as Al, Cr and P, thus reducing the volume of waste even more; (3) sludge washing, sludge dissolution, and separation of radionuclides from the dissolved sludge solutions (advanced processing): designed to remove all radionuclides for concentration into a minimum waste volume. This report describes a test plan for work that will be performed in FY 1994 under the Sludge Washing and Caustic Leaching Studies Task (WBS 0402) of the Tank Waste Remediation System (TWRS) Pretreatment Project. The objectives of the work described here are to determine the effects of sludge washing and alkaline leaching on sludge composition and the physical properties of the washed sludge and to evaluate alkaline leaching methods for their impact on the volume of borosilicate glass required to dispose of certain Hanford tank sludges.
Date: July 1, 1994
Creator: Rapko, B. M. & Lumetta, G. J.
Partner: UNT Libraries Government Documents Department

Pretreatment of Plutonium Finishing Plant (PFP) sludge: Report for the period October 1990--March 1992

Description: The current mission of the US Department of Energy`s Hanford Site is one of environmental restoration. A major task within this mission is the disposal of large volumes of high-level wastes (HLW) that are stored in underground tanks on the site. Under the current planning assumptions, all high-level tank waste will be vitrified as borosilicate glass and then disposed of in a geologic repository. The costs associated with this disposal scheme are very high. Thus, methods to reduce the volume of glass required to vitrify these wastes are currently being investigated. Plutonium Finishing Plant (PFP) sludge is a unique transuranic waste that is stored in tank 241- SY-102 on the Hanford site. As the name implies, the bulk of this material consists of waste from operations at the Plutonium Finishing Plant; but, other wastes have also been added (e.g., wastes from decontamination activities). Because the quantities of plutonium and americium in the PFP sludge are greater than 100 nCi/g, this sludge must be handled as a HLW. Approximately 6000 glass canisters would result from vitrifying this waste directly. Sludge washing would reduce the required number of canisters to {approximately}2500, with the volume of glass being driven by the low allowable concentration limit for Cr in the vitrification plant feed. The cost of production and subsequent geologic disposal of each canister of glass is expected to be $0.5 M to $1 M. Thus, an economic incentive exists to develop methods of pretreating the sludge to reduce the number of glass canisters needed to contain the final vitrified product.
Date: April 1, 1993
Creator: Lumetta, G. J. & Swanson, J. L.
Partner: UNT Libraries Government Documents Department

The SX Solver: A New Computer Program for Analyzing Solvent-Extraction Equilibria

Description: A new computer program, the SX Solver, has been developed to analyze solvent-extraction equilibria. The program operates out of Microsoft Excel{reg_sign} and uses the built-in ''Solver'' function to minimize the sum of the square of the residuals between measured and calculated distribution coefficients. The extraction of nitric acid by tributylphosphate has been modeled to illustrate the program's use.
Date: February 9, 1999
Creator: McNamara, B.K.; Rapko, B.M. & Lumetta, G.J.
Partner: UNT Libraries Government Documents Department

Oxidative dissolution of chromium from Hanford Tank sludges under alkaline conditions

Description: Because of the expected high cost of vitrifying and disposing of high-level waste at the U.S. Department of Energy`s Hanford Site, pretreatment processes are being developed to reduce the anticipated volume of borosilicate glass. Sludge washing and caustic leaching, the baseline sludge pretreatment process, is expected to leach out a substantial portion of the {sup 137}Cs, possibly other radionuclides, and a significant portion of such major nonradionuclides as Al or P. The decontaminated solution will be routed to the low-level waste stream, where it will be immobilized in a glass matrix. The leached solids, which will contain the transuranic elements and {sup 90}Sr, will be handled as high-level waste. Previous studies indicate that poor removal of chromium in the +3 oxidation state [Cr(III)] occurs during baseline pretreatment. Because the concentration of Cr allowed in high level waste glass is low, a relatively small amount of Cr in the sludge can have a relatively large impact on the volume of high level waste glass produced. For this reason, additional leach steps to remove Cr would be desirable, and oxidative alkaline leaching has been proposed as a simple addition to the baseline sludge pretreatment. This report describes small-scale screening tests on the oxidative alkaline leaching of Cr performed with actual Hanford tank sludges.
Date: July 1, 1996
Creator: Rapko, B. M.; Lumetta, G.J. & Wagner, M.J.
Partner: UNT Libraries Government Documents Department

Washing and caustic leaching of Hanford tank sludges: Results of FY 1995 studies

Description: During the past few years, the primary mission at the US Department of Energy`s Hanford Site has changed from producing plutonium to environmental restoration. Large volumes of high-level radioactive wastes (HLW), generated during past Pu production and other operations, are stored in underground tanks on site. The current plan for remediating the Hanford tank farms consists of waste retrieval, pretreatment, treatment (immobilization), and disposal. The HLW will be immobilized in a borosilicate glass matrix; the resulting glass canisters will then be disposed of in a geologic repository. Because of the expected high cost of HLW immobilization and disposal, pretreatment processes will be implemented to reduce the volume of borosilicate glass produced in processing the tank wastes. This document describes sludge washing and caustic leaching tests conducted in FY 1995 at the Pacific Northwest Laboratory (PNL) at the request of Westinghouse Hanford Company. These tests were performed using sludges from seven Hanford waste tanks -- B-111, BX-107, C-103, S-104, SY-103, T-104, and T-111. The primary and secondary types of waste stored in each of these tanks are given in Table 1. 1. The data collected in this effort will be used to support the March 1998 Tri-Party Agreement decision on the extent of pretreatment to be performed on the Hanford tank sludges (Ecology, EPA, and DOE 1994).
Date: August 11, 1995
Creator: Rapko, B.M.; Lumetta, G.J. & Wagner, M.J.
Partner: UNT Libraries Government Documents Department

Separation of strontium-90 from Hanford high-level radioactive waste

Description: Current guidelines for disposing of high-level radioactive wastes stored in underground tanks at the US Department of Energy`s Hanford Site call for vitrifying high-level waste (HLW) in borosilicate glass and disposing the glass canisters in a deep geologic repository. Disposition of the low-level waste (LLW) is yet to be determined, but it will likely be immobilized in a glass matrix and disposed of on site. To lower the radiological risk associated with the LLW form, methods are being developed to separate {sup 90}Sr from the bulk waste material so this isotope can be routed to the HLW stream. A solvent extraction method is being investigated to separate {sup 90}Sr from acid-dissolved Hanford tank wastes. Results of experiments with actual tank waste indicate that this method can be used to achieve separation of {sup 90}Sr from the bulk waste components. Greater than 99% of the {sup 90}Sr was removed from an acidic dissolved sludge solution by extraction with di-tbutylcyclohexano-18-crown-6 in 1-octanol (the SREX process). The major sludge components were not extracted.
Date: October 1, 1993
Creator: Lumetta, G. J.; Wagner, M. J. & Jones, E. O.
Partner: UNT Libraries Government Documents Department

Process chemistry for the pretreatment of Hanford tank wastes

Description: Current guidelines for disposing radioactive wastes stored in underground tanks at the US Department of Energy`s Hanford Site call for the vitrification of high-level waste in borosilicate glass and disposal of the glass canisters in a deep geologic repository. Low-level waste is to be cast in grout and disposed of on site in shallow burial vaults. Because of the high cost of vitrification and geologic disposal, methods are currently being developed to minimize the volume of high-level waste requiring disposal. Two approaches are being considered for pretreating radioactive tank sludges: (1) leaching of selected components from the sludge and (2) acid dissolution of the sludge followed by separation of key radionuclides. The leaching approach offers the advantage of simplicity, but the acid dissolution/radionuclide extraction approach has the potential to produce the least number of glass canisters. Four critical components (Cr, P, S, and Al) were leached from an actual Hanford tank waste-Plutonium Finishing Plant sludge. The Al, P, and S were removed from the sludge by digestion of the sludge with 0.1 M NaOH at 100{degrees}C. The Cr was leached by treating the sludge with alkaline KMnO{sub 4} at 100{degrees}C. Removing these four components from the sludge will dramatically lower the number of glass canisters required to dispose of this waste. The transuranic extraction (TRUEX) solvent extraction process has been demonstrated at a bench scale using an actual Hanford tank waste. The process, which involves extraction of the transuranic elements with octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO), separated 99.9% of the transuranic elements from the bulk components of the waste. Several problems associated with the TRUEX processing of this waste have been addressed and solved.
Date: August 1, 1992
Creator: Lumetta, G. J.; Swanson, J. L. & Barker, S. A.
Partner: UNT Libraries Government Documents Department

Underground storage tank integrated demonstration: Evaluation of pretreatment options for Hanford tank wastes

Description: Separation science plays a central role inn the pretreatment and disposal of nuclear wastes. The potential benefits of applying chemical separations in the pretreatment of the radioactive wastes stored at the various US Department of Energy sites cover both economic and environmental incentives. This is especially true at the Hanford Site, where the huge volume (>60 Mgal) of radioactive wastes stored in underground tanks could be partitioned into a very small volume of high-level waste (HLW) and a relatively large volume of low-level waste (LLW). The cost associated with vitrifying and disposing of just the HLW fraction in a geologic repository would be much less than those associated with vitrifying and disposing of all the wastes directly. Futhermore, the quality of the LLW form (e.g., grout) would be improved due to the lower inventory of radionuclides present in the LLW stream. In this report, we present the results of an evaluation of the pretreatment options for sludge taken from two different single-shell tanks at the Hanford Site-Tanks 241-B-110 and 241-U-110 (referred to as B-110 and U-110, respectively). The pretreatment options examined for these wastes included (1) leaching of transuranic (TRU) elements from the sludge, and (2) dissolution of the sludge followed by extraction of TRUs and {sup 90}Sr. In addition, the TRU leaching approach was examined for a third tank waste type, neutralized cladding removal waste.
Date: June 1, 1993
Creator: Lumetta, G. J.; Wagner, M. J.; Colton, N. G. & Jones, E. O.
Partner: UNT Libraries Government Documents Department

Evaluation of solid-based separation materials for the pretreatment of radioactive wastes

Description: Separation science will play an important role in pretreating nuclear wastes stored at various US Department of Energy Sites. The application of separation processes offers potential economic and environmental benefits with regards to remediating these sites. For example, at the Hanford Site, the sizeable volume of radioactive wastes stored in underground tanks could be partitioned into a small volume of high-level waste (HLW) and a relatively large volume of low-level waste (LLW). After waste separation, only the smaller volume of HLW would require costly vitrification and geologic disposal. Furthermore, the quality of the remaining LLW form (e.g., grout) would be improved due to the lower inventory of radionuclides present in the LLW stream. This report investigates extraction chromatography as a possible separation process for Hanford wastes.
Date: May 1, 1993
Creator: Lumetta, G. J.; Wagner, M. J.; Wester, D. W. & Morrey, J. R.
Partner: UNT Libraries Government Documents Department

Washing and alkaline leaching of Hanford tank sludges: A status report

Description: Because of the assumed high cost of high-level waste (HLW) immobilization and disposal, pretreatment methods are being developed to minimize the volume of HLW requiring vitrification. Pacific Northwest Laboratory (PNL) is investigating several options for pretreating the radioactive wastes stored in underground tanks at the Hanford Site. The pretreatment methods under study for the tank sludges include: (1) simply washing the sludges with dilute NaOH, (2) performing caustic leaching (as well as washing) to remove certain wash components, and (3) dissolving the sludges in acid and extracting key radionuclides from the dissolved sludge solutions. The data collected in this effort will be used to support the March 1998 decision on the extent of pretreatment to be performed on the Hanford tank sludges. This document describes sludge washing and caustic leaching tests conducted in FY 1994. These tests were performed using sludges from single-shell tanks (SST) B-201 and U-110. A summary is given of all the sludge washing and caustic leaching studies conducted at PNL in the last few years.
Date: September 1, 1994
Creator: Lumetta, G. J. & Rapko, B. M.
Partner: UNT Libraries Government Documents Department

Sludge Treatment and Extraction Technology Development: Results of FY 1993 studies

Description: This report describes experimental results from work conducted in FY 1993 under the Sludge Treatment and Extraction Technology Development Task of the Tank Waste Remediation System (TWRS) Pretreatment Technology Development Project at Pacific Northwest Laboratory (PNL). Experiments were conducted in the following six general areas: (1) sludge washing, (2) sludge leaching, (3) sludge dissolution, (4) actinide separation by solvent extraction and extraction chromatography, (5) Sr separation by solvent extraction, and (6) extraction of Cs from acidic solution.
Date: March 1, 1994
Creator: Lumetta, G. J.; Wagner, M. J.; Barrington, R. J.; Rapko, B. M. & Carlson, C. D.
Partner: UNT Libraries Government Documents Department

Liquid membrane system for the removal and concentration of transuranic elements

Description: The goal of this program is to develop an efficient, reliable, and radiation-resistant modified liquid membrane system (MLMS) for the selective removal and concentration of transuranic elements (TRUs) and strontium-90 from dissolved Hanford sludge wastes. The efforts are divided into three categories: (1) demonstration and optimization of the MLMS for the TRUEX and SREX processes using simulant waste solution; (2) development of a radiation-resistant microporous divider and membrane module for testing with actual waste solutions; and (3) demonstration of the MLMS for the TRUEX and SREX processes using actual Hanford waste. Successful completion of these development efforts will yield a compact, versatile, and reliable MLMS for implementation with the TRUEX and SREX processes. The MLMS is simple, stable, more efficient, and easier to control and operate than conventional solvent-extraction processes, such as those employing centrifugal contactors. In addition, the MLMS process offers operational cost savings over the conventional technology, by exhibiting at least a 10% reduction in the consumption of extractant chemicals.
Date: December 31, 1996
Creator: Timmins, M.R.; Wysk, S.R.; Smolensky, L.A.; Jiang, D. & Lumetta, G.J.
Partner: UNT Libraries Government Documents Department

Solvent extraction of radionuclides from aqueous tank waste

Description: This task aims toward the development of efficient solvent-extraction processes for the removal of the fission products {sup 99}Tc, {sup 90}Sr, and {sup 137}Cs from alkaline tank wastes. Processes already developed or proposed entail direct treatment of the waste solution with the solvent and subsequent stripping of the extracted contaminants from the solvent into a dilute aqueous solution. Working processes to remove Tc(and SR) separately and Cs separately have been developed; the feasibility of a combined process is under investigation. Since Tc, Sr, and Cs will be vitrified together in the high-level fraction, however, a process that could separate Tc, Sr, and Cs simultaneously, as opposed to sequentially, potentially offers the greatest impact. A figure presents a simplified diagram of a proposed solvent-extraction cycle followed by three possible treatments for the stripping solution. Some degree of recycle of the stripping solution (option a) is expected. Simple evaporation (option c) is possible prior to vitrification; this offers the greatest possible volume reduction with simple operation and no consumption of chemicals, but it is energy intensive. However, if the contaminants are concentrated (option b) by fixed-bed technology, the energy penalty of evaporation can be avoided and vitrification facilitated without any additional secondary waste being produced.
Date: January 1, 1997
Creator: Moyer, B.A.; Bonnesen, P.V.; Sachleben, R.A.; Leonard, R.A. & Lumetta, G.J.
Partner: UNT Libraries Government Documents Department

The chemistry of sludge washing and caustic leaching processes for selected Hanford tank wastes

Description: A broad-based study on washing and caustic leaching of Hanford tank sludges was performed in FY 1995 to gain a better understanding of the basic chemical processes that underlie this process. This approach involved testing of the baseline sludge washing and caustic leaching method on several Hanford tank sludges, and characterization of the solids both before and after testing by electron microscopy, X-ray diffraction, and X-ray absorption spectroscopy. A thermodynamically based model was employed to help understand the factors involved in individual specie distribution in the various stages of the sludge washing and caustic leaching treatment. The behavior of the important chemical and radiochemical components throughout the testing is summarized and reviewed in this report.
Date: March 1, 1996
Creator: Rapko, B.M.; Blanchard, D.L.; Colton, N.G.; Felmy, A.R.; Liu, J. & Lumetta, G.J.
Partner: UNT Libraries Government Documents Department

Washing and caustic leaching of Hanford Tank C-106 sludge

Description: This report describes the results of a laboratory-scale washing and caustic leaching test performed on sludge from Hanford Tank C-106. The purpose of this test was to determine the behavior of important sludge components when subjected to washing with dilute or concentrated sodium hydroxide solutions. The results of this laboratory-scale test were used to support the design of a bench-scale washing and leaching process used to prepare several hundred grams of high-level waste solids for vitrification tests to be done by private contractors. The laboratory-scale test was conducted at Pacific Northwest Laboratory in FY 1996 as part of the Hanford privatization effort. The work was funded by the US Department of Energy through the Tank Waste Remediation System (TWRS; EM-30).
Date: October 1, 1996
Creator: Lumetta, G.J.; Wagner, M.J.; Hoopes, F.V. & Steele, R.T.
Partner: UNT Libraries Government Documents Department