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Considerations Affecting Deep-Well Disposal of Tritium-Bearing Low-Level Aqueous Waste from Nuclear Fuel Reprocessing Plants

Description: Present concepts of disposal of low-level aqueous wastes (LLAW) that contain much of the fission-product tritium from light water reactors involve dispersal to the atmosphere or to surface streams at fuel reprocessing plants. These concepts have been challenged in recent years. Deep-well injection of low-level aqueous wastes, an alternative to biospheric dispersal, is the subject of this presentation. Many factors must be considered in assessing its feasibility, including technology, costs, env… more
Date: March 1977
Creator: Trevorrow, L. E.; Warner, D. L. & Steindler, M. J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Radioactive Liquid Waste Treatment Facility

Description: The Radioactive Liquid Waste Treatment Facility (RLWTF) at Argonne National Laboratory-West (ANL-W) in Idaho provides improved treatment for low-level aqueous waste compared to conventional systems. A unique, patented evaporated system is used in the RLWTF. SHADE (shielded hot air drum evaporator, US Patent No. 4,305,780) is a low-cost disposable unit constructed from standard components and is self-shielded. The results of testing and recent operations indicate that evaporation rates of 2 to 6… more
Date: July 1984
Creator: Black, Roger L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Nuclear Material Safeguards Surveillance and Accountancy by Isotope Correlation Techniques

Description: The purpose of this study is to investigate the applicability of isotope correlation techniques (ICT) to the Light Water Reactor (LWR) and the Liquid Metal Fast Breeder Reactor (LMFBR) fuel cycles for nuclear material accountancy and safeguards surveillance. The isotopic measurement of the inventory input to the reprocessing phase of the fuel cycle is the primary direct determination that an anomaly may exist in the fuel management of nuclear material. The nuclear materials accountancy gap whic… more
Date: November 1981
Creator: Persiani, P. J.; Goleb, J. A. & Kroc, T. K.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Nuclear Waste Programs Semiannual Progress Report: October 1991-March 1992

Description: This document reports on the work done by the Nuclear Waste Programs of the Chemical Technology Division (CMT), Argonne National Laboratory, in the period October 1991-March 1992. In these programs, studies are underway on the performance of waste glass and spent fuel in projected nuclear repository conditions to provide input to the licensing of the nation's high-level waste repositories.
Date: November 1993
Creator: Bates, John K.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Nuclear Waste Programs Semiannual Progress Report: April-September 1992

Description: This document reports on the work done by the Nuclear Waste Programs of the Chemical Technology Division (CMT), Argonne National Laboratory, in the period April-September 1992. In these programs, studies are underway on the performance of waste glass and spent fuel in projected nuclear repository conditions to provide input to the licensing of the nation's high-level waste repositories.
Date: May 1994
Creator: Bates, John K.; Bradley, C. R.; Buck, E. C.; Dietz, N. L.; Ebert, William L.; Emery, J. W. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Advanced Evaporator Technology Progress Report FY 1992

Description: This report summarizes the work that was completed in FY 1992 on the program "Technology Development for Concentrating Process Streams." 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 deconta… more
Date: January 1995
Creator: Chamberlain, D. B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Basic TRUEX Process for Rocky Flats Plant

Description: The Generic TRUEX Model was used to develop a TRUEX process flowsheet for recovering the transuranics (plutonium, americium) from a nitrate waste stream at Rocky Flats Plant. T\
Date: August 1994
Creator: Leonard, R. A.; Chamberlain, D. B.; Dow, J. A.; Farley, S. E.; Nuñez, Luis; Regalbuto, M. C. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Actinide Recovery Using Aqueous Biphasic Extraction: Initial Developmental Studies

Description: Aqueous biphasic extraction systems are being developed to treat radioactive wastes. The separation technique involves the selective partitioning of either solutes or colloid-size particles between two scible aqueous phases. Wet grinding of plutonium residues to an average particle size of one micron will be used to liberate the plutonium from the bulk of the particle matrix. The goal is to produce a plutonium concentrate that will integrate with existing and developing chemical recovery proces… more
Date: August 1992
Creator: Chaiko, David J.; Mensah-Biney, R.; Mertz, C. J. & Rollins, A. N.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Preliminary Plan for Treating Mixed Waste

Description: A preliminary waste treatment plan was developed for disposing of radioactive inorganic liquid wastes that contain hazardous metals and/or hazardous acid concentrations at Argonne National Laboratory. This plan, which involves neutralization and sulfide precipitation followed by filtration, reduces the concentration of hazardous metals and the acidity so that the filtrate liquid is simply a low-level radioactive waste that can be fed to a low-level waste evaporator.
Date: June 1993
Creator: Vandegrift, G. F.; Conner, C.; Hutter, Joseph C.; Leonard, R. A.; Nuñez, Luis; Sedlet, J. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Di-n-Amyl-n-Amylphosphonate and Tricaprylmethylammonium Nitrate as Potential Extractants for Reprocessing Th-U Fuels

Description: Both 2F di-n-amyl-n-amylphosphonate in n-dodecane (DA(AP)-DD) and 0.77F tricaprylmethylammonium nitrate in Aromatic 100 (TCMA.NO/sub 3/-AR100) extract uranium and thorium into relatively concentrated organic solutions. Countercurrent-flow tests with each extractant have demonstrated effective uranium-thorium separations by selective stripping from the organic phase. Both extractants offer advantageous alternatives to tributylphosphate (Thorex) in reprocessing irradiated mixed ThO2-UO2 fuels. Fo… more
Date: September 1979
Creator: Diamond, H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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TRUEX Hot Demonstration

Description: In FY 1987, a program was initiated to demonstrate technology for recovering transuranic (TRU) elements from defense wastes. This hot demonstration was to be carried out with solution from the dissolution of irradiated fuels. This recovery would be accomplished with both PUREX and TRUEX solvent extraction processes. Work planned for this program included preparation of a shielded-cell facility for the receipt and storage of spent fuel from commercial power reactors, dissolution of this fuel, op… more
Date: April 1990
Creator: Chamberlain, D. B.; Leonard, R. A.; Hoh, J. C.; Gay, E. C.; Kalina, D. G. & Vandegrift, G. F.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Dissolution Characteristics of Mixed UO₂ Powders in J-13 Water Under Saturated Conditions

Description: The Yucca Mountain Project/Spent Fuel program at Argonne National Laboratory is designed to determine radionuclide release rates by exposing high-level waste to repository-relevant groundwater. To gain experience for the tests with spent fuel, a scoping experiment was conducted at room temperature to determine the uranium release rate from an unirradiated uranium dioxide powder mixture (14.3 wt % enrichment in uranium-235) to J-13 water under saturated conditions. Another goal set for the exper… more
Date: March 1991
Creator: Veleckis, Ewald & Hoh, J. C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Leaching Action of EJ-13 Water on Unirradiated UO₂ Surfaces under Unsaturated Conditions at 90 C : Interim Report

Description: A set of experiments, based on the application of the Unsaturated Test method to the reaction of uranium dioxide with EJ-13 water, has been conducted over a period of 182.5 weeks. One half of the experiments have been terminated, while one half are still ongoing. Solutions that have dripped from uranium dioxide specimens have been analyzed for all experiments, while the reacted uranium dioxide surfaces have been examined for only the terminated experiments. A pulse of uranium release from the u… more
Date: July 1991
Creator: Wronkiewicz, D. J.; Bates, John K.; Gerding, Thomas J.; Veleckis, Ewald; Tani, B. & Hoh, J. C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Engineering Development of Fluid-Bed Fluoride Volatility Processes: Part 5. Description of a Pilot-Scale Facility for Uranium Dioxide-Plutonium Dioxide Processing Studies

Description: Report describing a pilot plant constructed at Argonne National Laboratory for studying two major process steps for the recovery of uranium and plutonium from spent nuclear fuels of power reactors. A major objective is the demonstration of optimum process conditions for the two steps for synthetic reactor fuel compositions, including those containing mixtures of inactive fission products.
Date: 1964
Creator: Vogel, G. J.; Carls, E. L. & Mecham, W. J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Studies on Uranium Recovery From Unirradiated Reactor Fuel Elements by Volatile Separation in Inert Fluidized Beds

Description: This technical report presents the results of studies on the recovery of uranium by fluoride volatilization from the following unirradiated reactor fuels: zirconium-uranium alloy, uranium oxide, uranium carbide, and uranium metal. The inert fluidized bed technique was found to be a promising approach to the volatile reprocessing of spent nuclear fuels. This report is accompanied by 9 tables and 10 figures.
Date: 1962
Creator: Reilly, J. J., (Of the Brookhaven National Laboratory, Associated Universities Incorporated); Regan, W. H.; Wirsing, E. & Hatch, L. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Dissolution of Zirconium Matrix Fuels in Molten Fluoride Salts

Description: The application of the Volatility Process for reprocessing spent reactor fuels to zirconium fuel elements requires the conversion of the fuel to a fluoride form. This conversion can be accomplished by the dissolution of such fuels in a bath of molten fluoride salts with a hydrogen fluoride sparge. Studies with dummy zirconium fuel elements showed dissolution rates averaging 2 mg/sq cm-min. The effect of submergence, geometrical configuration, and HF impingement rate appeared more significant th… more
Date: January 30, 1959
Creator: Horton, R. W. & Whatley, M. E. (Marvin E.), 1926-
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Unit Operations Section Monthly Progress Report, February 1959. Chemical Technology Division

Description: A gamma scintillation spectrometer was used to measure diffusivity of uranyl nitrate in water during preliminary capillary experiments. During Fluorox run FBR-22, 90.4% of the theoretical amount of UF formed was collected in cold traps and chemical traps. Toroid tests of flame calcined mixed Th-U oxide showed low corrosion rates, small changes in particle size and low solubilization of uranium, while denaturation of uranyl nitrate in a fluidized bed resulted in particle growth with uniform laye… more
Date: June 11, 1959
Creator: Bresee, J. C.; Haas, P. A.; Watson, C. D.; Whatley, M. E. (Marvin E.), 1926- & Horton, R. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Processing of Molten Salt Power Reactor Fuel

Description: Fuel reprocessing methods are being investigated for molten salt nuclear reactors which use LiF-BeF2 salt as a solvent for UF4 and ThF4. A liquid HF dissolution procedure coupled with fluorination has been developed for recovery of the uranium and LiF0BeF2 solvent salt which is highly enriched in Li-7. The recovered salt is decontaminated in the process from the major reactions poisons; namely, rare earths and neptunium. A brief investigation of alternate methods, including oxide precipitation,… more
Date: April 1, 1959
Creator: Campbell, David Owen, 1927- & Cathers, G. I.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Some Effects of Radiation on the Solvent Extraction Process

Description: Presented at the 135th National Meeting of the ACS Boston April 5 10 1959. The yield of total acid Gacid in the radiolysis of tributyl phosphate Amsco 125 82 solutions is 2.7 times the electron fraction of TBP or approximately 2.7 times the weight fraction of TBP per 100 ev of energy absorbed by the solution. Dibutyl phosphoric acid DBPA constitutes about 85% of the acid. Radiolysis of TBP also results in the conversion of about 0.9 molecules of TBP to a polymer per 100 ev of energy absorbed. U… more
Date: January 26, 1959
Creator: Davis, Wallace, Jr. & Wagner, Robert M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Some Effects of Radiation on Solvent Extraction Processes

Description: The yield of total acid Gacid in the radiolysis of tributyl phosphate Amsco 125 82 solutions is 2.7 times the electron fraction of TBP or approximately 2.7 times the weight fraction of TBP per 100 ev of energy absorbed by the solution. Dibutyl phosphoric acid DBPA constitutes about 85% of the acid. Radiolysis of TBP also results in the conversion of about 0.9 molecules of TBP to a polymer per 100 ev of energy absorbed. Uranium extraction stripping tests with an 8 stage spinner column have shown… more
Date: January 26, 1959
Creator: Davis, Wallace, Jr. & Wagner, Robert M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Unit Operations Section Monthly Progress Report, January 1959. Chemical Technology Division

Description: Two alternate systems, acetic acid-nickel and acetic acid-cobalt, were examined for possible replacement of the corrosion system: ferric chloride-nickel which is being used in the current transpiration corrosion protection studies. Two Fluorox fluidized bed runs were made, of 9 and 58 hr duration, in which dry air and oxygen were used as oxidization and fluidizing gases. Tests of the hydroclone classification of thoria indicated that more than 95% of the +.05 micron particles can be removed fro… more
Date: April 30, 1959
Creator: Bresee, J. C.; Haas, P. A.; Horton, R. W. & Watson, C. D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Some Effects of Radiation on Solvent Extraction Processes

Description: The yield of total acid Gacid in the radiolysis of tributyl phosphate Amsco 125 82 solutions is 2.7 times the electron fraction of TBP or approximately 2.7 times the weight fraction of TBP per 100 ev of energy absorbed by the solution. Dibutyl phosphoric acid DBPA constitutes about 85% of the acid. Radiolysis of TBP also results in the conversion of about 0.9 molecules of TBP to a polymer per 100 ev of energy absorbed. Uranium extraction stripping tests with an 8 stage spinner column have shown… more
Date: January 26, 1959
Creator: Davis, Wallace, Jr. & Wagner, Robert M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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