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TRUEX/SREX demonstration. Innovative technology summary report

Description: The tank waste at the Idaho National engineering and Environmental Laboratory (INEEL) must be removed from the tanks by 2012. Transuranic Extraction (TRUEX) and Strontium Extraction (SREX) are the preferred processes for treating INEEL tank waste. The demonstrations for both the TRUEX and SREX processes were carried out separately in the ICPP Remote Analytical Laboratory (RAL) shielded hot cell. A 24-stage bank of 2-cm diameter, centrifugal contactors was fabricated by Argonne National Laboratory. The contractors were modified at the ICPP for remote installation and operation in the RAL hot cell. An overall removal efficiency of 99.79% was obtained for the actinides using TRUEX. An overall removal efficiency of 94% was obtained for the actinides using SREX. The TRUEX and SREX processes will undergo further testing before full-scale processes are built. The experimental results are based on short-term testing (2--3 h). Longer testing times are needed. This report describes the technology, their performance, the application of the technology, costs, regulatory and policy issues, and lessons learned.
Date: December 1, 1998
Partner: UNT Libraries Government Documents Department

The extraction of rare earth elements from ICPP sodium-bearing waste and dissolved zirconium calcine by CMP and TRUEX solvents

Description: The extraction of stable isotopes of Eu and Ce was investigated from simulated sodium-bearing waste (SBW) and dissolved zirconium calcine by TRUEX and CMP solvents at the Idaho Chemical Processing Plant (ICPP). Single batch contacts were carried out in order to evaluate the rare earth behavior in the extraction, scrub, strip and wash sections for the proposed flowsheets. It has been shown that these lanthanides are efficiently extracted from the sodium-bearing wastes into either solvent, are not scrubbed and are stripped from both of the extractants with dilute HEDPA. The extraction distribution coefficients for Ce and Eu are higher in the TRUEX solvent (D{sub Ce} = 11.7, D{sub Eu} = 14.9) compared with CMP (D{sub Ce} = 9.3, D{sub Eu} = 7.23) for SBW. The extraction distribution coefficients for Ce and Eu are considerably less in the TRUEX solvent (D{sub Ce}=1.13, D{sub Eu}=1.8) than in the CMP solvent (D{sub Ce}=7.4, D{sub Eu=}6.1) for dissolved zirconium calcine feeds. The lower distribution coefficients for the extraction of lanthanides in the TRUEX/dissolved zirconium calcine system can be explained by zirconium loading of the solvent. The data obtained also confirmed that Ce and Eu can be used as non-radioactive surrogates for Am in separation experiments with acidic solutions.
Date: November 1, 1995
Creator: Todd, T.A.; Glagolenko, I.Y.; Herbst, R.S. & Brewer, K.N.
Partner: UNT Libraries Government Documents Department

Demonstration of the TRUEX process for partitioning of actinides from actual ICPP tank waste using centrifugal contactors in a shielded cell facility

Description: TRUEX is being evaluated at Idaho Chemical Processing Plant (ICPP) for separating actinides from acidic radioactive waste stored at ICPP; efforts have culminated in a recent demonstration with actual tank waste. A continuous countercurrent flowsheet test was successfully completed at ICPP using waste from tank WM-183. This demonstration was performed using 24 states of 2-cm dia centrifugal contactors in the shielded hot cell at the ICPP Remote Analytical Laboratory. The flowsheet had 8 extraction stages, 5 scrub stages, 6 strip stages, 3 solvent wash stages, and 2 acid rinse stages. A centrifugal contactor stage in the scrub section was not working during testing, and the scrub feed (aqueous) solution followed the solvent into the strip section, eliminating the scrub section in the flowsheet. An overall removal efficiency of 99.97% was obtained for the actinides, reducing the activity from 457 nCi/g in the feed to 0.12 nCi/g in the aqueous raffinate, well below the NRC Class A LLW requirement of 10 nCi/g for non-TRU waste.The 0.04 M HEDPA strip section back-extracted 99.9998% of the actinide from the TRUEX solvent. Removal efficiencies of >99. 90, 99.96, 99.98, >98.89, 93.3, and 89% were obtained for {sup 241}Am, {sup 238}Pu, {sup 239}Pu, {sup 235}U, {sup 238}U, and {sup 99}Tc. Fe was partially extracted by the TRUEX solvent, resulting in 23% of the Fe exiting in the strip product. Hg was also extracted by the TRUEX solvent (73%) and stripped from the solvent in the 0.25 M Na2CO3 wash section. Only 1.4% of the Hg exited with the high activity waste strip product.
Date: September 1, 1996
Creator: Law, J.D.; Brewer, K.N.; Herbst, R.S. & Todd, T.A.
Partner: UNT Libraries Government Documents Department

Elimination of phosphate and zirconium in the high-activity fraction resulting from TRUEX partitioning of ICPP zirconium calcines

Description: Laboratory testing was undertaken with the aim of developing a TRUEX flowsheet that would efficiently remove actinides from solutions of dissolved zirconium calcine and minimize the glass volume produced from the ensuing high-activity fraction. A TRUEX flowsheet is recommended for testing in the 2-cm centrifugal contactor pilot-plant based on the results from this testing. These laboratory tests show that zirconium recovery in the high activity fraction is minimized by scrubbing with an optimized NHF concentration of 0.2 M. This NH4F concentration in the scrub allowed the HEDPA strip concentration to be reduced from 0.04 M to 0.004 M because HEDPA is not consumed by zirconium. Complete TRU stripping was also achieved in these laboratory tests with 0.004 M HEDPA. Data from the small-scale laboratory batch contact tests were used in the Generic TRUEX Model (GTM) to evaluate the proposed flowsheet under counter-current conditions. GTM results indicate the raffinate will meet the Class A non-TRU limit of < 10 nCi/g in six extraction stages (O/A = 1), and quantitative actinide recovery will be achieved with the 0.004 M HEDPA in six strip stages (O/A = 1). Only 6.6 % of the initial zirconium concentration is anticipated to be recovered with the actinides, indicating the four scrub stages (O/A = 3) efficiently removes zirconium from the TRUEX solvent. In addition to recommending an improved TRUEX flowsheet for testing in the 2-cm centrifugal contactor pilot-plant, this work has shown that small reductions in zirconium extraction drastically improves flowsheet performance. These small changes in zirconium extraction can be accomplished by modifying the calcine dissolution parameters. Therefore, further calcine dissolution testing followed by TRUEX testing with the resulting feed solutions is also recommended.
Date: July 1, 1997
Creator: Brewer, K.N.; Tillotson, R.D. & Tullock, P.A.
Partner: UNT Libraries Government Documents Department

Present and future roles of solvent extraction in treatment of nuclear wastes

Description: Solvent extraction has played a major role in development of the nuclear industry and has recovered much of the uranium from raw materials and essentially all of the plutonium and uranium from spent fuels. These operations produced a wide variety of radioactive wastes as well as the uranium and plutonium products. Solvent extraction worked well in the earlier nuclear facilities and should play a significant role in future cleanup operations.
Date: December 31, 1995
Creator: Watson, J.S.
Partner: UNT Libraries Government Documents Department

Selective partitioning of mercury from co-extracted actinides in a simulated acidic ICPP waste stream

Description: The TRUEX process is being evaluated at the Idaho Chemical Processing Plant (ICPP) as a means to partition the actinides from acidic sodium-bearing waste (SBW). The mercury content of this waste averages 1 g/l. Because the chemistry of mercury has not been extensively evaluated in the TRUEX process, mercury was singled out as an element of interest. Radioactive mercury, {sup 203}Hg, was spiked into a simulated solution of SBW containing 1 g/l mercury. Successive extraction batch contacts with the mercury spiked waste simulant and successive scrubbing and stripping batch contacts of the mercury loaded TRUEX solvent (0.2 M CMPO-1.4 M TBP in dodecane) show that mercury will extract into and strip from the solvent. The extraction distribution coefficient for mercury, as HgCl{sub 2} from SBW having a nitric acid concentration of 1.4 M and a chloride concentration of 0.035 M was found to be 3. The stripping distribution coefficient was found to be 0.5 with 5 M HNO{sub 3} and 0.077 with 0.25 M Na{sub 2}CO{sub 3}. An experimental flowsheet was designed from the batch contact tests and tested counter-currently using 5.5 cm centrifugal contactors. Results from the counter-current test show that mercury can be removed from the acidic mixed SBW simulant and recovered separately from the actinides.
Date: December 1, 1995
Creator: Brewer, K.N.; Herbst, R.S. & Tranter, T.J.
Partner: UNT Libraries Government Documents Department

CMPO purity tests in the TRUEX solvent using americium-241

Description: The Transuranic Extraction (TRUEX) Process was developed by E.P. Horwitz and coworkers at Argonne National Laboratory (ANL) to separate the +4, +6, and +3 actinides from acidic aqueous solutions of nuclear wastes. Octyl (phenyl)-N-N-diisobutyl-carbamoylmethylphosphine oxide (CMPO) is the active actinide complexant used in the TRUEX solvent. CMPO is combined with tributyl phosphate (TBP) in an organic diluent, typically n-dodecane, to form the TRUEX solvent. Small quantities of impurities in the CMPO resulting from: (1) synthesis, (2) acid hydrolysis, or (3) radiolysis can result in actinide stripping problems from the solvent. The impurity, octylphenylphosphinic acid (POPPA), ia a powerful extractant at low acid concentrations which may be formed during CMPO synthesis. Consequently, commercial CMPO may contain sufficient quantities of POPPA to significantly impact the stripping of actinides from the TRUEX solvent. The purpose of these tests was to (1) determine if commercially available CMPO is sufficiently pure to alleviate actinide stripping problems from the TRUEX process and (2) to determine if solvent cleanup methods are sufficient to purify the commercially purchased CMPO. Extraction and solvent cleanup methodologies used by Horwitz and coworkers at ANL were used to determine CMPO purity with {sup 241}Am. The improvement of the americium distribution coefficient in dilute nitric acid resulting from further purifying this CMPO is not significant enough to warrant additional CMPO purifying steps. The commercially purchased CMPO is found to be acceptable to use, as received, in a full-scale TRUEX process.
Date: December 1, 1993
Creator: Brewer, K.N.; Herbst, R.S.; Tranter, T.J. & Todd, T.A.
Partner: UNT Libraries Government Documents Department

Centrifugal contractors for laboratory-scale solvent extraction tests

Description: A 2-cm contactor (minicontactor) was developed and used at Argonne National Laboratory for laboratory-scale testing of solvent extraction flowsheets. This new contactor requires only 1 L of simulated waste feed, which is significantly less than the 10 L required for the 4-cm unit that had previously been used. In addition, the volume requirements for the other aqueous and organic feeds are reduced correspondingly. This paper (1) discusses the design of the minicontactor, (2) describes results from having applied the minicontactor to testing various solvent extraction flowsheets, and (3) compares the minicontactor with the 4-cm contactor as a device for testing solvent extraction flowsheets on a laboratory scale.
Date: December 31, 1995
Creator: Leonard, R.A.; Chamberlain, D.B. & Conner, C.
Partner: UNT Libraries Government Documents Department

Application of single ion activity coefficients to determine solvent extraction mechanism for components of high level nuclear waste

Description: The TRUEX solvent extraction process is being developed to remove and concentrate transuranic (TRU) elements from high-level and TRU radioactive wastes currently stored at US Department of Energy sites. Phosphoric acid is one of the chemical species of concern at the Hanford site where bismuth phosphate was used to recover plutonium. The mechanism of phosphoric acid extraction with TRUEX-NPH solvent at 25{degrees}C was determined by phosphoric acid distribution ratios, which were measured by using phosphoric acid radiotracer and a variety of aqueous phases containing different concentrations of nitric acid and nitrate ions. A model was developed for predicting phosphoric acid distribution ratios as a function of the thermodynamic activities of nitrate ion and hydrogen ion. The Generic TRUEX Model (GTM) was used to calculate these activities based on the Bromley method. The derived model supports CMPO and TBP extraction of a phosphoric acid-nitric acid complex and a CMPO-phosphoric acid complex in TRUEX-NPH solvent.
Date: December 31, 1995
Creator: Nunez, L. & Vandegrift, G.F.
Partner: UNT Libraries Government Documents Department

TRUEX processing of plutonium analytical solutions at Argonne National Laboratory

Description: The TRUEX (TRansUranic EXtraction) solvent extraction process was developed at Argonne National Laboratory (ANL) for the Department of Energy. A TRUEX demonstration completed at ANL involved the processing of analytical and experimental waste generated there and at the New Brunswick Laboratory. A 20-stage centrifugal contactor was used to recover plutonium, americium, and uranium from the waste. Approximately 84 g of plutonium, 18 g of uranium, and 0.2 g of americium were recovered from about 118 liters of solution during four process runs. Alpha decontamination factors as high as 65,000 were attained, which was especially important because it allowed the disposal of the process raffinate as a low-level waste. The recovered plutonium and uranium were converted to oxide; the recovered americium solution was concentrated by evaporation to approximately 100 ml. The flowsheet and operational procedures were modified to overcome process difficulties. These difficulties included the presence of complexants in the feed, solvent degradation, plutonium precipitation, and inadequate decontamination factors during startup. This paper will discuss details of the experimental effort.
Date: December 31, 1995
Creator: Chamberlain, D.B.; Conner, C.; Hutter, J.C.; Leonard, R.A.; Wygmans, D.G. & Vandegrift, G.F.
Partner: UNT Libraries Government Documents Department

Advanced evaporator technology progress report FY 1992

Description: This report summarizes the work that was completed in FY 1992 on the program {open_quotes}Technology Development for Concentrating Process Streams.{close_quotes} The purpose of this program is to evaluate and develop evaporator technology for concentrating radioactive waste and product streams such as those generated by the TRUEX process. Concentrating these streams and minimizing the volume of waste generated can significantly reduce disposal costs; however, equipment to concentrate the streams and recycle the decontaminated condensates must be installed. LICON, Inc., is developing an evaporator that shows a great deal of potential for this application. In this report, concepts that need to be incorporated into the design of an evaporator operated in a radioactive environment are discussed. These concepts include criticality safety, remote operation and maintenance, and materials of construction. Both solubility and vapor-liquid equilibrium data are needed to design an effective process for concentrating process streams. Therefore, literature surveys were completed and are summarized in this report. A model that is being developed to predict vapor phase compositions is described. A laboratory-scale evaporator was purchased and installed to study the evaporation process and to collect additional data. This unit is described in detail. Two new LICON evaporators are being designed for installation at Argonne-East in FY 1993 to process low-level radioactive waste generated throughout the laboratory. They will also provide operating data from a full-sized evaporator processing radioactive solutions. Details on these evaporators are included in this report.
Date: January 1, 1995
Creator: Chamberlain, D.; Hutter, J.C. & Leonard, R.A.
Partner: UNT Libraries Government Documents Department

Demonstration of the TRUEX process for the treatment of actual high activity tank waste at the INEEL using centrifugal contactors

Description: The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Engineering and Environmental Laboratory (INEEL), formerly reprocessed spent nuclear fuel to recover fissionable uranium. The radioactive raffinates from the solvent extraction uranium recovery processes were converted to granular solids (calcine) in a high temperature fluidized bed. A secondary liquid waste stream was generated during the course of reprocessing, primarily from equipment decontamination between campaigns and solvent wash activities. This acidic tank waste cannot be directly calcined due to the high sodium content and has historically been blended with reprocessing raffinates or non-radioactive aluminum nitrate prior to calcination. Fuel reprocessing activities are no longer being performed at the ICPP, thereby eliminating the option of waste blending to deplete the waste inventory. Currently, approximately 5.7 million liters of high-activity waste are temporarily stored at the ICPP in large underground stainless-steel tanks. The United States Environmental Protection Agency and the Idaho Department of Health and Welfare filed a Notice of Noncompliance in 1992 contending some of the underground waste storage tanks do not meet secondary containment. As part of a 1995 agreement between the State of Idaho, the Department of Energy, and the Department of Navy, the waste must be removed from the tanks by 2012. Treatment of the tank waste inventories by partitioning the radionuclides and immobilizing the resulting high-activity and low-activity waste streams is currently under evaluation. A recent peer review identified the most promising radionuclide separation technologies for evaluation. The Transuranic Extraction-(TRUEX) process was identified as a primary candidate for separation of the actinides from ICPP tank waste.
Date: October 1, 1997
Creator: Law, J.D.; Brewer, K.N.; Todd, T.A. & Olson, L.G.
Partner: UNT Libraries Government Documents Department

Separation science and technology. Semiannual progress report, October 1993--March 1994

Description: This document reports on the work done by the Separations Science and Technology Programs of the Chemical Technology Division, Argonne National Laboratory (ANL), in the period October 1993-March 1994. This effort is mainly concerned with developing the TRUEX process for removing and concentrating actinides from acidic waste streams contaminated with transuranic (TRU) elements. The objectives of TRUEX processing are to recover valuable TRU elements and to lower disposal costs for the nonTRU waste product of the process. Other projects are underway with the objective of developing (1) evaporation technology for concentrating radioactive waste and product streams such as those generated by the TRUEX process, (2) treatment schemes for liquid wastes stored are being generated at ANL, (3) a process based on sorbing modified TRUEX solvent on magnetic beads to be used for separation of contaminants from radioactive and hazardous waste streams, and (4) a process that uses low-enriched uranium targets for production of {sup 99}Mo for nuclear medicine uses.
Date: December 1, 1997
Creator: Vandegrift, G.F.; Aase, S.B. & Buchholz, B.
Partner: UNT Libraries Government Documents Department

Actinide partitioning from actual Idaho chemical processing plant acidic tank waste using centrifugal contactors

Description: The TRUEX process is being evaluated at the Idaho Chemical Processing Plant (ICPP) for the separation of the actinides from acidic radioactive wastes stored at the ICPP. These efforts have culminated in a recent demonstration of the TRUEX process with actual tank waste. This demonstration was performed using 24 stages of 2-cm diameter centrifugal contactors installed in a shielded hot cell at the ICPP Remote Analytical Laboratory. An overall removal efficiency of 99.97% was obtained for the actinides. As a result, the activity of the actinides was reduced from 457 nCi/g in the feed to 0.12 nCi/g in the aqueous raffinate, which is well below the U.S. NRC Class A LLW requirement of 10 nCi/g for non-TRU waste. Iron was partially extracted by the TRUEX solvent, resulting in 23% of the Fe exiting in the strip product. Mercury was also extracted by the TRUEX solvent (76%) and stripped from the solvent in the 0.25 M Na{sub 2}CO{sub 3} wash section.
Date: October 1, 1997
Creator: Law, J.D.; Brewer, K.N. & Todd, T.A.
Partner: UNT Libraries Government Documents Department

Removal of actinides from dissolved ORNL MVST sludge using the TRUEX process

Description: Experiments were conducted to evaluate the transuranium extraction process for partitioning actinides from actual dissolved high-level radioactive waste sludge. All tests were performed at ambient temperature. Time and budget constraints permitted only two experimental campaigns. Samples of sludge from Melton Valley Storage Tank W-25 were rinsed with mild caustic (0.2 M NaOH) to reduce the concentrations of nitrates and fission products associated with the interstitial liquid. In one campaign, the rinsed sludge was dissolved in nitric acid to produce a solution containing total metal concentrations of ca. 1.8 M with a nitric acid concentration of ca. 2.9 M. About 50% of the dry mass of the sludge was dissolved. In the other campaign, the sludge was neutralized with nitric acid to destroy the carbonates, then leached with ca. 2.6 M NaOH for ca. 6 h before rinsing with the mild caustic. The sludge was then dissolved in nitric acid to produce a solution containing total metal concentrations of ca. 0.6 M with a nitric acid concentration of ca. 1.7 M. About 80% of the sludge dissolved. The dissolved sludge solution form the first campaign began gelling immediately, and a visible gel layer was observed after 8 days. In the second campaign, the solution became hazy after ca. 8 days, indicating gel formation, but did not display separated gel layers after aging for 20 days. Batch liquid-liquid equilibrium tests of both the extraction and stripping operations were conducted. Chemical analyses of both phases were used to evaluate the process. Evaluation was based on two metrics: the fraction of TRU elements removed from the dissolved sludge and comparison of the results with predictions made with the Generic TRUEX Model (GTM). The fractions of Eu, Pu, Cm, Th, and U species removed from aqueous solution in only one extraction stage were > 95% ...
Date: July 1, 1997
Creator: Spencer, B.B.; Egan, B.Z. & Chase, C.W.
Partner: UNT Libraries Government Documents Department

Separation Science and Technology. Semiannual progress report, April 1993--September 1993

Description: This document reports on the work done by the Separations Science and Technology Programs of the Chemical Technology Division, Argonne National Laboratory, in the period April-September 1993. This effort is mainly concerned with developing the TRUEX process for removing and concentrating actinides from acidic waste streams contaminated with transuranic (TRU) elements. The objectives of TRUEX processing are to recover valuable TRU elements and to lower disposal costs for the nonTRU waste product of the process. Other projects are underway with the objective of developing (1) evaporation technology for concentrating radioactive waste and product streams such as those generated by the TRUEX process, (2) treatment schemes for liquid wastes stored or being generated at Argonne, (3) a process based on sorbing modified TRUEX solvent on magnetic beads to be used for separation of contaminants from radioactive and hazardous waste streams, and (4) a process that uses low-enriched uranium targets for production of {sup 99}Mo for nuclear medicine uses.
Date: January 1, 1996
Creator: Vandegrift, G.F.; Chamberlain, D.B. & Conner, C.
Partner: UNT Libraries Government Documents Department

Countercurrent flowsheet testing of the TRUEX process with ICPP calcine

Description: Calcine was generated at the Idaho Chemical Processing Plant over several decades as a method of solidifying numerous raffinates and wastes from spent nuclear fuel reprocessing for convenient interim storage. Unfortunately, the bulk of the calcine is inert, with radionuclides comprising less than 1 weight percent of the total calcine mass. The bulk of the calcine currently stored at the ICPP was produced from wastes generated during reprocessing of zirconium clad fuels. Consequently, this material contains varying, but large quantities of zirconium oxide. Currently, separations options are being considered for acidic solutions of dissolved ICPP calcines to minimize high level waste volumes and economic penalties perceived for final disposal of these wastes. The actinide separation process being emphasized for the dissolved calcine solutions is the TRUEX process. Substantial problems have been encountered during TRUEX flowsheet development efforts for dissolved zirconium calcine simulant due to the high concentrations and subsequent extraction of zirconium from the feed. Alteration of the calcine dissolution parameters has resulted in the development of a successful TRUEX/Zr calcine baseline flowsheet. This flowsheet has been tested using 22 stages of a 2.0 centimeter diameter centrifugal contactor pilot plant using simulated dissolved Zr calcine solution. With this flowsheet, a removal efficiency of > 96% was obtained for {sup 241}Am (analytical detection limits were reached). Less than 0.25% of the {sup 95}Zr exited with the high-level waste strip product.
Date: July 1, 1998
Creator: Law, J.D.; Herbst, R.S.; Brewer, K.N. & Todd, T.A.
Partner: UNT Libraries Government Documents Department

Mercury extraction by the TRUEX process solvent: I. Kinetics, extractable species, dependence on nitric acid concentration and stoichiometry

Description: Mercury extraction from acidic aqueous solutions by the TRUEX process solvent (0.2 M CMPO, 1.4 M TBP in n-dodecane) has not extensively been examined. Research at the Idaho Chemical Processing Plant is currently in progress to evaluate the TRUEX process for actinide removal from several acidic waste streams, including liquid sodium-bearing waste (SBW), which contains significant quantities of mercury. Preliminary experiments were performed involving the extraction of Hg{sup 203}, added as HgCl{sub 2}, from 0.01 to 10 M HNO{sub 3} solutions. Mercury distribution coefficients (D{sub Hg}) range between 3 and 60 from 0.01 M to 2 M HNO{sub 3}. At higher nitric acid concentrations, i.e. 5 M HNO{sub 3} or greater, D{sub Hg} significantly decreases to values less than 1. These results indicate mercury is extracted from acidic solutions {<=}{approximately}2 M HNO{sub 3} and stripped with nitric acid solutions {>=}{approximately}5 M HNO{sub 3}. Experimental results indicate the extractable species is HgCl{sub 2} from nitrate media, i.e., chloride must be present in the nitrate feed to extract mercury. Extractions from Hg(NO{sub 3}){sub 2} solutions indicated substantially reduced distribution ratios, typically D{sub Hg}< 1, for the range of nitric acid concentrations examined (0.01 to 8 M HNO{sub 3}). Extraction of mercury, as HgCl{sub 2}, by the individual components of the TRUEX solvent was also examined from 2 M HNO{sub 3}. The diluent, n-dodecane, does not measurably extract mercury. With a 1.4 M TBP/n-dodecane solvent, D{sub Hg} {approximately}3.4 compared with D{sub Hg} {approximately}7 for the TRUEX solvent. Classical slope analysis techniques were utilized to evaluate the stoichiometric coefficients of Hg extraction independently for both CMPO and TBP.
Date: December 1, 1995
Creator: Herbst, R.S.; Brewer, K.N.; Tranter, T.J. & Todd, T.A.
Partner: UNT Libraries Government Documents Department

Quarterly progress report for the chemical development section of the Chemical Technology Division: October--December 1995

Description: This quarterly report is intended to provide a timely summary of the major activities being conducted in the Chemical Development Section of the Chemical Technology Division at the Oak Ridge National Laboratory (ORNL) during the period September-December 1995. The report summarizes ten major tasks conducted within five major areas of research and development within the section. The first major research area-Chemical Processes for Waste Management-includes the following tasks: Comprehensive Supernate Treatment, Partitioning of Sludge Compounds by Caustic Leaching, Studies on Treatment of Dissolved MVST Sludge Using TRUEX Process, ACT*DE*CON{sup SM} Test Program, Hot Demonstration of Proposed Commercial Nuclide Removal Technology, and Sludge Washing and Dissolution of ORNL Waste: Data for Modeling Sludge Science. The Comprehensive Supernate task is currently evaluating several sorbents in batch tests for removing strontium, technetium, and cesium from ORNL Melton Valley Storage Tank (MVST) supernatant solutions. Nine sorbents have been evaluated for removing strontium from MVST W-29 supernatant, and seven have been tested for technetium removal. All planned batch testing of cesium sorbents has been completed; however, additional cesium tests may be made as new sorbents become available. At the request of Hanford personnel, some batch tests were made to evaluate the effect on cesium distribution of selected sorbents which had been treated with an organic such as tributyl phosphate.
Date: March 1, 1996
Creator: Jubin, R.T.
Partner: UNT Libraries Government Documents Department

TRUEX process applied to radioactive Idaho Chemical Processing Plant high-level waste calcine

Description: Equal volume batch contact experiments were performed with dissolved, radioactive high-level waste (HLW) calcine and the TRansUranic EXtraction (TRUEX) process solvent. Extraction, scrub, and strip distribution coefficients (D) were obtained for the transuranic (TRU) elements in order to evaluate the efficiency of the TRUEX process in treating this waste. The extraction, scrub, and strip behavior of other elements, such as chromium, zirconium, and technetium, was also observed. A TRU alpha decontamination factor of >10,000 was achieved; after three extraction batch contacts TRU alpha activity was reduced from 1,420 nCi/g to 0.02 nCi/g. Dilute nitric acid was used to scrub extracted acid, zirconium, and iron from the solvent prior to stripping. Dilute 1-hydroxyethane, 1-1, diphosphonic acid (HEDPA) was used as a gross TRU stripping reagent to recover the extracted TRUs. Data from these batch contact experiments were used to develop a counter-current flowsheet for TRU removal using the Generic TRUEX Model (GTM). Process improvements and optimizations of the flowsheet have been evaluated using a non-radioactive dissolved calcine simulant spiked with tracers to obtain additional distribution coefficient data. These data were used in the GTM to refine the flowsheet. The flowsheet was then evaluated using a counter-current 5.5 cm centrifugal contactor pilot plant with a non-radioactive dissolved calcine simulant. The experiments involving radioactive waste provided crucial data for developing a baseline TRUEX process flowsheet which can effectively separate TRU components from ICPP high-level waste.
Date: May 1, 1996
Creator: Brewer, K.N.; Herbst, R.S.; Law, J.D.; Todd, T.A. & Olson, A.L.
Partner: UNT Libraries Government Documents Department

Validation of the generic TRUEX model using data from TRUEX demonstrations with actual high-level waste

Description: The main objective of the Generic TRUEX Model (GTM) is to calculate TRUEX solvent extraction flowsheets based on input of a specific feed and a specific set of process goals and constraints. The output will be (1) the compositions of all effluent streams, (2) the compositions of both phases (organic and aqueous) in each stage of the contacting equipment at steady state, and (3) estimates of the space and cost requirements for installing the flowsheet in a plant situation. Other options are available to calculate aqueous-phase speciation and thermodynamic activities, distribution ratios of extractable species, and solvent radiolytic and hydrolytic degradation. Calculation of these options is based on initial aqueous- and organic-phase compositions and other important variables supplied by the user. Three demonstrations of the TRUEX process have been run by Power Reactor and Nuclear Fuel Development Corp. (PNC) researchers at the Tokai Works using actual PUREX raffinates. A 19-stage mixer settler was used for the extraction and scrub sections, and a 16-to-19-stage unit for stripping. Stagewise data were collected on the behavior of nitric acid and several fission-product and actinide radioisotopes during these runs; Run 2 was the best documented and the one with which most comparisons were made. These data are important tools for validating predictions made by the GTM and understanding the intricacies of the TRUEX process. In this paper, results of the GTM calculations will be compared to the actual data published by PNC researchers. Differences between model predictions and experimental data were analyzed in terms of the process chemistry and demonstration conditions.
Date: December 31, 1995
Creator: Vandegrift, G.F. & Regalbuto, M.C.
Partner: UNT Libraries Government Documents Department

A comparision of TRUEX and CMP solvent extraction processes for actinide removal from ICPP wastes

Description: The Idaho Chemical Processing Plant (ICPP) is currently engaged in development efforts for the decontamination of high-level radioactive wastes generated from decades of nuclear fuel reprocessing. These wastes include several types of calcine, generated by high temperature solidification of reprocessing raffinates. In addition to calcine, there are smaller quantities of secondary wastes from decontamination and solvent wash activities which are typically referred to as sodium-bearing waste (SBW). Solvent extraction technologies based on octyl(phenyl)-N,N-diisobutyl-carbamoylmethylphosphine oxide (CMPO, the active extractant in the TRUEX process) and dihexyl-N,N-diethylcarbamoylmethylphosphonate (DHDECMP, the active extractant in the CMP process) are being evaluated for actinide partitioning from these waste streams. Calcines must first be dissolved in an appropriate acidic solution prior to treatment in solvent extraction based processes. The SBW is currently stored as an acidic solution and readily amenable to liquid extraction techniques. Development efforts to date have revolved around defining and refining baseline flowsheets with the TRUEX and CMP processes for each waste stream. Another objective of this work was to determine which of these technologies are best suited for the treatment of ICPP wastes. Laboratory batch contacts were performed to identify relevant chemistry and distribution coefficients. This information was then used to establish baseline flowsheet configuration with regard to chemistry. The laboratory data were used to model the behavior of the actinides and other constituents in the wastes in countercurrent, continuous processes based on centrifugal contactor technology. The laboratory data and modelling results form the basis for comparison of the two processes.
Date: April 1, 1996
Creator: Herbst, R.S.; Brewer, K.N.; Garn, T.G. & Law, J.D.
Partner: UNT Libraries Government Documents Department

TRUEX partitioning from radioactive ICPP sodium bearing waste

Description: The Idaho Chemical Processing Plant (ICPP) located at the Idaho National Engineering Laboratory in Southeast Idaho is currently evaluating several treatment technologies applicable to waste streams generated over several decades of-nuclear fuel reprocessing. Liquid sodium bearing waste (SBW), generated primarily during decontamination activities, is one of the waste streams of interest. The TRansUranic EXtraction (TRUEX) process developed at Argonne National Laboratory is currently being evaluated to separate the actinides from SBW. On a mass basis, the amount of the radioactive species in SBW are low relative to inert matrix components. Thus, the advantage of separations is a dramatic decrease in resulting volumes of high activity waste (HAW) which must be dispositioned. Numerous studies conducted at the ICPP indicate the applicability of the TRUEX process has been demonstrated; however, these studies relied on a simulated SBW surrogate for the real waste. Consequently, a series of batch contacts were performed on samples of radioactive ICPP SBW taken from tank WM-185 to verify that actual waste would behave similarly to the simulated waste. The test results with SBW from tank WM-185 indicate the TRUEX solvent effectively extracts the actinides from the samples of actual waste. Gross alpha radioactivity, attributed predominantly to Pu and Am, was reduced from 3.14E+04 dps/mL to 1.46 dps/mL in three successive batch contacts with fresh TRUEX solvent. This reduction corresponds to a decontamination factor of DF = 20,000 or 99.995% removal of the gross a activity in the feed. The TRUEX solvent also extracted the matrix components Zr, Fe, and Hg to an appreciable extent (D{sub Zr} > 10, D{sub Fe} {approx} 2, D{sub Hg} {approx}6). Iron co-extracted with the actinides can be successfully scrubbed from the organic with 0.2 M HNO{sub 3}. Mercury can be selectively partitioned from the actinides with either sodium carbonate or nitric acid ({ge} ...
Date: March 1995
Creator: Herbst, R. S.; Brewer, K. N.; Tranter, T. J. & Todd, T. A.
Partner: UNT Libraries Government Documents Department