Sample records for actinide removal process

  1. Process to remove actinides from soil using magnetic separation

    DOE Patents [OSTI]

    Avens, Larry R. (Los Alamos, NM); Hill, Dallas D. (Los Alamos, NM); Prenger, F. Coyne (Los Alamos, NM); Stewart, Walter F. (Las Cruces, NM); Tolt, Thomas L. (Los Alamos, NM); Worl, Laura A. (Los Alamos, NM)

    1996-01-01T23:59:59.000Z

    A process of separating actinide-containing components from an admixture including forming a slurry including actinide-containing components within an admixture, said slurry including a dispersion-promoting surfactant, adjusting the pH of the slurry to within a desired range, and, passing said slurry through a pretreated matrix material, said matrix material adapted to generate high magnetic field gradients upon the application of a strong magnetic field exceeding about 0.1 Tesla whereupon a portion of said actinide-containing components are separated from said slurry and remain adhered upon said matrix material is provided.

  2. ACTINIDE REMOVAL PROCESS SAMPLE ANALYSIS, CHEMICAL MODELING, AND FILTRATION EVALUATION

    SciTech Connect (OSTI)

    Martino, C.; Herman, D.; Pike, J.; Peters, T.

    2014-06-05T23:59:59.000Z

    Filtration within the Actinide Removal Process (ARP) currently limits the throughput in interim salt processing at the Savannah River Site. In this process, batches of salt solution with Monosodium Titanate (MST) sorbent are concentrated by crossflow filtration. The filtrate is subsequently processed to remove cesium in the Modular Caustic Side Solvent Extraction Unit (MCU) followed by disposal in saltstone grout. The concentrated MST slurry is washed and sent to the Defense Waste Processing Facility (DWPF) for vitrification. During recent ARP processing, there has been a degradation of filter performance manifested as the inability to maintain high filtrate flux throughout a multi-batch cycle. The objectives of this effort were to characterize the feed streams, to determine if solids (in addition to MST) are precipitating and causing the degraded performance of the filters, and to assess the particle size and rheological data to address potential filtration impacts. Equilibrium modelling with OLI Analyzer{sup TM} and OLI ESP{sup TM} was performed to determine chemical components at risk of precipitation and to simulate the ARP process. The performance of ARP filtration was evaluated to review potential causes of the observed filter behavior. Task activities for this study included extensive physical and chemical analysis of samples from the Late Wash Pump Tank (LWPT) and the Late Wash Hold Tank (LWHT) within ARP as well as samples of the tank farm feed from Tank 49H. The samples from the LWPT and LWHT were obtained from several stages of processing of Salt Batch 6D, Cycle 6, Batch 16.

  3. Actinide recovery process

    DOE Patents [OSTI]

    Muscatello, Anthony C. (Arvada, CO); Navratil, James D. (Arvada, CO); Saba, Mark T. (Arvada, CO)

    1987-07-28T23:59:59.000Z

    Process for the removal of plutonium polymer and ionic actinides from aqueous solutions by absorption onto a solid extractant loaded on a solid inert support such as polystyrenedivinylbenzene. The absorbed actinides can then be recovered by incineration, by stripping with organic solvents, or by acid digestion. Preferred solid extractants are trioctylphosphine oxide and octylphenyl-N,N-diisobutylcarbamoylmethylphosphine oxide and the like.

  4. Actinide recovery process

    DOE Patents [OSTI]

    Muscatello, A.C.; Navratil, J.D.; Saba, M.T.

    1985-06-13T23:59:59.000Z

    Process for the removal of plutonium polymer and ionic actinides from aqueous solutions by absorption onto a solid extractant loaded on a solid inert support such as polystyrene-divinylbenzene. The absorbed actinides can then be recovered by incineration, by stripping with organic solvents, or by acid digestion. Preferred solid extractants are trioctylphosphine oxide and octylphenyl-N,N-diisobutylcarbamoylmethylphosphine oxide and the like. 2 tabs.

  5. Actinide metal processing

    DOE Patents [OSTI]

    Sauer, Nancy N. (Los Alamos, NM); Watkin, John G. (Los Alamos, NM)

    1992-01-01T23:59:59.000Z

    A process of converting an actinide metal such as thorium, uranium, or plnium to an actinide oxide material by admixing the actinide metal in an aqueous medium with a hypochlorite as an oxidizing agent for sufficient time to form the actinide oxide material and recovering the actinide oxide material is provided together with a low temperature process of preparing an actinide oxide nitrate such as uranyl nitrte. Additionally, a composition of matter comprising the reaction product of uranium metal and sodium hypochlorite is provided, the reaction product being an essentially insoluble uranium oxide material suitable for disposal or long term storage.

  6. Actinide metal processing

    DOE Patents [OSTI]

    Sauer, N.N.; Watkin, J.G.

    1992-03-24T23:59:59.000Z

    A process for converting an actinide metal such as thorium, uranium, or plutonium to an actinide oxide material by admixing the actinide metal in an aqueous medium with a hypochlorite as an oxidizing agent for sufficient time to form the actinide oxide material and recovering the actinide oxide material is described together with a low temperature process for preparing an actinide oxide nitrate such as uranyl nitrate. Additionally, a composition of matter comprising the reaction product of uranium metal and sodium hypochlorite is provided, the reaction product being an essentially insoluble uranium oxide material suitable for disposal or long term storage.

  7. Actinide removal from spent salts

    DOE Patents [OSTI]

    Hsu, Peter C. (Pleasanton, CA); von Holtz, Erica H. (Livermore, CA); Hipple, David L. (Livermore, CA); Summers, Leslie J. (Livermore, CA); Adamson, Martyn G. (Danville, CA)

    2002-01-01T23:59:59.000Z

    A method for removing actinide contaminants (uranium and thorium) from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents are added to precipitate the thorium as thorium oxide and/or the uranium as either uranium oxide or as a diuranate salt. The precipitated materials are filtered, dried and packaged for disposal as radioactive waste. About 90% of the thorium and/or uranium present is removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 0.1 ppm of thorium or uranium.

  8. Actinide Lanthanide Separation Process – ALSEP

    SciTech Connect (OSTI)

    Gelis, Artem V.; Lumetta, Gregg J.

    2014-01-29T23:59:59.000Z

    Separation of the minor actinides (Am, Cm) from the lanthanides at an industrial scale remains a significant technical challenge for closing the nuclear fuel cycle. To increase the safety of used nuclear fuel (UNF) reprocessing, as well as reduce associated costs, a novel solvent extraction process has been developed. The process allows for partitioning minor actinides, lanthanides and fission products following uranium/plutonium/neptunium removal; minimizing the number of separation steps, flowsheets, chemical consumption, and waste. This new process, Actinide Lanthanide SEParation (ALSEP), uses an organic solvent consisting of a neutral diglycolamide extractant, either N,N,N',N'-tetra(2 ethylhexyl)diglycolamide (T2EHDGA) or N,N,N',N'-tetraoctyldiglycolamide (TODGA), and an acidic extractant 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP]), dissolved in an aliphatic diluent (e.g. n-dodecane). The An/Ln co-extraction is conducted from moderate-to-strong nitric acid, while the selective stripping of the minor actinides from the lanthanides is carried out using a polyaminocarboxylic acid/citrate buffered solution at pH anywhere between 3 and 4.5. The extraction and separation of the actinides from the fission products is very effective in a wide range of HNO3 concentrations and the minimum separation factors for lanthanide/Am exceed 30 for Nd/Am, reaching > 60 for Eu/Am under some conditions. The experimental results presented here demonstrate the great potential for a combined system, consisting of a neutral extractant such as T2EHDGA or TODGA, and an acidic extractant such as HEH[EHP], for separating the minor actinides from the lanthanides.

  9. SALTSTONE VAULT CLASSIFICATION SAMPLES MODULAR CAUSTIC SIDE SOLVENT EXTRACTION UNIT/ACTINIDE REMOVAL PROCESS WASTE STREAM APRIL 2011

    SciTech Connect (OSTI)

    Eibling, R.

    2011-09-28T23:59:59.000Z

    Savannah River National Laboratory (SRNL) was asked to prepare saltstone from samples of Tank 50H obtained by SRNL on April 5, 2011 (Tank 50H sampling occurred on April 4, 2011) during 2QCY11 to determine the non-hazardous nature of the grout and for additional vault classification analyses. The samples were cured and shipped to Babcock & Wilcox Technical Services Group-Radioisotope and Analytical Chemistry Laboratory (B&W TSG-RACL) to perform the Toxic Characteristic Leaching Procedure (TCLP) and subsequent extract analysis on saltstone samples for the analytes required for the quarterly analysis saltstone sample. In addition to the eight toxic metals - arsenic, barium, cadmium, chromium, mercury, lead, selenium and silver - analytes included the underlying hazardous constituents (UHC) antimony, beryllium, nickel, and thallium which could not be eliminated from analysis by process knowledge. Additional inorganic species determined by B&W TSG-RACL include aluminum, boron, chloride, cobalt, copper, fluoride, iron, lithium, manganese, molybdenum, nitrate/nitrite as Nitrogen, strontium, sulfate, uranium, and zinc and the following radionuclides: gross alpha, gross beta/gamma, 3H, 60Co, 90Sr, 99Tc, 106Ru, 106Rh, 125Sb, 137Cs, 137mBa, 154Eu, 238Pu, 239/240Pu, 241Pu, 241Am, 242Cm, and 243/244Cm. B&W TSG-RACL provided subsamples to GEL Laboratories, LLC for analysis for the VOCs benzene, toluene, and 1-butanol. GEL also determines phenol (total) and the following radionuclides: 147Pm, 226Ra and 228Ra. Preparation of the 2QCY11 saltstone samples for the quarterly analysis and for vault classification purposes and the subsequent TCLP analyses of these samples showed that: (1) The saltstone waste form disposed of in the Saltstone Disposal Facility in 2QCY11 was not characteristically hazardous for toxicity. (2) The concentrations of the eight RCRA metals and UHCs identified as possible in the saltstone waste form were present at levels below the UTS. (3) Most of the inorganic species measured in the leachate do not exceed the MCL, SMCL or TW limits. (4) The inorganic waste species that exceeded the MCL by more than a factor of 10 were nitrate, nitrite and the sum of nitrate and nitrite. (5) Analyses met all quality assurance specifications of US EPA SW-846. (6) The organic species (benzene, toluene, 1-butanol, phenol) were either not detected or were less than reportable for the vault classification samples. (7) The gross alpha and radium isotopes could not be determined to the MCL because of the elevated background which raised the detection limits. (8) Most of the beta/gamma activity was from 137Cs and its daughter 137mBa. (9) The concentration of 137Cs and 90Sr were present in the leachate at concentrations 1/40th and 1/8th respectively than in the 2003 vault classification samples. The saltstone waste form placed in the Saltstone Disposal Facility in 2QCY11 met the SCHWMR R.61-79.261.24(b) RCRA metals requirements for a nonhazardous waste form. The TCLP leachate concentrations for nitrate, nitrite and the sum of nitrate and nitrite were greater than 10x the MCLs in SCDHEC Regulations R.61-107.19, Part I A, which confirms the Saltstone Disposal Facility classification as a Class 3 Landfill. The saltstone waste form placed in the Saltstone Disposal Facility in 2QCY11 met the R.61-79.268.48(a) non wastewater treatment standards.

  10. Actinide and lanthanide separation process (ALSEP)

    DOE Patents [OSTI]

    Guelis, Artem V.

    2013-01-15T23:59:59.000Z

    The process of the invention is the separation of minor actinides from lanthanides in a fluid mixture comprising, fission products, lanthanides, minor actinides, rare earth elements, nitric acid and water by addition of an organic chelating aid to the fluid; extracting the fluid with a solvent comprising a first extractant, a second extractant and an organic diluent to form an organic extractant stream and an aqueous raffinate. Scrubbing the organic stream with a dicarboxylic acid and a chelating agent to form a scrubber discharge. The scrubber discharge is stripped with a simple buffering agent and a second chelating agent in the pH range of 2.5 to 6.1 to produce actinide and lanthanide streams and spent organic diluents. The first extractant is selected from bis(2-ethylhexyl)hydrogen phosphate (HDEHP) and mono(2-ethylhexyl)2-ethylhexyl phosphonate (HEH(EHP)) and the second extractant is selected from N,N,N,N-tetra-2-ethylhexyl diglycol amide (TEHDGA) and N,N,N',N'-tetraoctyl-3-oxapentanediamide (TODGA).

  11. Process for making a ceramic composition for immobilization of actinides

    DOE Patents [OSTI]

    Ebbinghaus, Bartley B. (Livermore, CA); Van Konynenburg, Richard A. (Livermore, CA); Vance, Eric R. (Kirrawee, AU); Stewart, Martin W. (Barden Ridge, AU); Walls, Philip A. (Cronulla, AU); Brummond, William Allen (Livermore, CA); Armantrout, Guy A. (Livermore, CA); Herman, Connie Cicero (Pleasanton, CA); Hobson, Beverly F. (Livermore, CA); Herman, David Thomas (Pleasanton, CA); Curtis, Paul G. (Tracy, CA); Farmer, Joseph (Tracy, CA)

    2001-01-01T23:59:59.000Z

    Disclosed is a process for making a ceramic composition for the immobilization of actinides, particularly uranium and plutonium. The ceramic is a titanate material comprising pyrochlore, brannerite and rutile. The process comprises oxidizing the actinides, milling the oxides to a powder, blending them with ceramic precursors, cold pressing the blend and sintering the pressed material.

  12. Process to remove rare earth from IFR electrolyte

    DOE Patents [OSTI]

    Ackerman, J.P.; Johnson, T.R.

    1994-08-09T23:59:59.000Z

    The invention is a process for the removal of rare earths from molten chloride electrolyte salts used in the reprocessing of integrated fast reactor fuel (IFR). The process can be used either continuously during normal operation of the electrorefiner or as a batch process. The process consists of first separating the actinide values from the salt before purification by removal of the rare earths. After replacement of the actinides removed in the first step, the now-purified salt electrolyte has the same uranium and plutonium concentration and ratio as when the salt was removed from the electrorefiner. 1 fig.

  13. Process to remove rare earth from IFR electrolyte

    DOE Patents [OSTI]

    Ackerman, J.P.; Johnson, T.R.

    1992-01-01T23:59:59.000Z

    The invention is a process for the removal of rare earths from molten chloride electrolyte salts used in the reprocessing of integrated fast reactor fuel (IFR). The process can be used either continuously during normal operation of the electrorefiner or as a batch process. The process consists of first separating the actinide values from the salt before purification by removal of the rare earths. After replacement of the actinides removed in the first step, the now-purified salt electrolyte has the same uranium and plutonium concentration and ratio as when the salt was removed from the electrorefiner.

  14. Process to remove rare earth from IFR electrolyte

    DOE Patents [OSTI]

    Ackerman, John P. (Downers Grove, IL); Johnson, Terry R. (Wheaton, IL)

    1994-01-01T23:59:59.000Z

    The invention is a process for the removal of rare earths from molten chloride electrolyte salts used in the reprocessing of integrated fast reactor fuel (IFR). The process can be used either continuously during normal operation of the electrorefiner or as a batch process. The process consists of first separating the actinide values from the salt before purification by removal of the rare earths. After replacement of the actinides removed in the first step, the now-purified salt electrolyte has the same uranium and plutonium concentration and ratio as when the salt was removed from the electrorefiner.

  15. Continuous sulfur removal process

    DOE Patents [OSTI]

    Jalan, V.; Ryu, J.

    1994-04-26T23:59:59.000Z

    A continuous process for the removal of hydrogen sulfide from a gas stream using a membrane comprising a metal oxide deposited on a porous support is disclosed. 4 figures.

  16. Electrochemical Processes for Removing

    E-Print Network [OSTI]

    Fay, Noah

    Introduction Most unit operations for water treatment either add chemicals or produce a saline liquid waste one community is often the source of potable water for downstream locales. Water treatment processes for water treatment that address the problem of increasing watery supply salinity. The problem of water

  17. Carbon dioxide removal process

    DOE Patents [OSTI]

    Baker, Richard W.; Da Costa, Andre R.; Lokhandwala, Kaaeid A.

    2003-11-18T23:59:59.000Z

    A process and apparatus for separating carbon dioxide from gas, especially natural gas, that also contains C.sub.3+ hydrocarbons. The invention uses two or three membrane separation steps, optionally in conjunction with cooling/condensation under pressure, to yield a lighter, sweeter product natural gas stream, and/or a carbon dioxide stream of reinjection quality and/or a natural gas liquids (NGL) stream.

  18. Analysis of large soil samples for actinides

    SciTech Connect (OSTI)

    Maxwell, III; Sherrod L. (Aiken, SC)

    2009-03-24T23:59:59.000Z

    A method of analyzing relatively large soil samples for actinides by employing a separation process that includes cerium fluoride precipitation for removing the soil matrix and precipitates plutonium, americium, and curium with cerium and hydrofluoric acid followed by separating these actinides using chromatography cartridges.

  19. Status of development of actinide blanket processing flowsheets for accelerator transmutation of nuclear waste

    SciTech Connect (OSTI)

    Dewey, H.J.; Jarvinen, G.D.; Marsh, S.F.; Schroeder, N.C.; Smith, B.F.; Villarreal, R.; Walker, R.B.; Yarbro, S.L.; Yates, M.A.

    1993-09-01T23:59:59.000Z

    An accelerator-driven subcritical nuclear system is briefly described that transmutes actinides and selected long-lived fission products. An application of this accelerator transmutation of nuclear waste (ATW) concept to spent fuel from a commercial nuclear power plant is presented as an example. The emphasis here is on a possible aqueous processing flowsheet to separate the actinides and selected long-lived fission products from the remaining fission products within the transmutation system. In the proposed system the actinides circulate through the thermal neutron flux as a slurry of oxide particles in heavy water in two loops with different average residence times: one loop for neptunium and plutonium and one for americium and curium. Material from the Np/Pu loop is processed with a short cooling time (5-10 days) because of the need to keep the total actinide inventory, low for this particular ATW application. The high radiation and thermal load from the irradiated material places severe constraints on the separation processes that can be used. The oxide particles are dissolved in nitric acid and a quarternary, ammonium anion exchanger is used to extract neptunium, plutonium, technetium, and palladium. After further cooling (about 90 days), the Am, Cm and higher actinides are extracted using a TALSPEAK-type process. The proposed operations were chosen because they have been successfully tested for processing high-level radioactive fuels or wastes in gram to kilogram quantities.

  20. Toward understanding the thermodynamics of TALSPEAK process. Medium effects on actinide complexation

    SciTech Connect (OSTI)

    Peter R Zalupski; Leigh R Martin; Ken Nash; Yoshinobu Nakamura; Masahiko Yamamoto

    2009-07-01T23:59:59.000Z

    The ingenious combination of lactate and diethylenetriamine-N,N,N’,N”,N”-pentaacetic acid (DTPA) as an aqueous actinide-complexing medium forms the basis of the successful separation of americium and curium from lanthanides known as the TALSPEAK process. While numerous reports in the prior literature have focused on the optimization of this solvent extraction system, considerably less attention has been devoted to the understanding of the basic thermodynamic features of the complex fluids responsible for the separation. The available thermochemical information of both lactate and DTPA protonation and metal complexation reactions are representative of the behavior of these ions under idealized conditions. Our previous studies of medium effects on lactate protonation suggest that significant departures from the speciation predicted based on reported thermodynamic values should be expected in the TALSPEAK aqueous environment. Thermodynamic parameters describing the separation chemistry of this process thus require further examination at conditions significantly removed from conventional ideal systems commonly employed in fundamental solution chemistry. Such thermodynamic characterization is the key to predictive modelling of TALSPEAK. Improved understanding will, in principle, allow process technologists to more efficiently respond to off-normal conditions during large scale process operation. In this report, the results of calorimetric and potentiometric investigations of the effects of aqueous electrolytes on the thermodynamic parameters for lactate protonation and lactate complexation of americium and neodymium will be presented. Studies on the lactate protonation equilibrium will clearly illustrate distinct thermodynamic variations between strong electrolyte aqueous systems and buffered lactate environment.

  1. Processing and Disposition of Special Actinide Target Materials - 13138

    SciTech Connect (OSTI)

    Robinson, Sharon M.; Patton, Brad D. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)] [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Allender, Jeffrey S. [Savannah River National Laboratory (United States)] [Savannah River National Laboratory (United States)

    2013-07-01T23:59:59.000Z

    The Department of Energy (DOE) manages an inventory of materials that contains a range of long-lived radioactive isotopes that were produced from the 1960's through the 1980's by irradiating targets in high-flux reactors at the Savannah River Site (SRS) to produce special heavy isotopes for DOE programmatic use, scientific research, and industrial and medical applications. Among the products were californium-252, heavy curium (including Cm-246 through Cm-248), and plutonium-242 and -244. Many of the isotopes are still in demand today, and they can be recovered from the remaining targets previously irradiated at SRS or produced from the recovered isotopes. Should the existing target materials be discarded, the plutonium (Pu) and curium (Cm) isotopes cannot be replaced readily with existing production sources. Some of these targets are stored at SRS, while other target material is stored at Oak Ridge National Laboratory (ORNL) at several stages of processing. The materials cannot be stored in their present form indefinitely. Their long-term management involves processing items for beneficial use and/or for disposition, using storage and process facilities at SRS and ORNL. Evaluations are under way for disposition options for these materials, and demonstrations of improved flow sheets to process the materials are being conducted at ORNL and the Savannah River National Laboratory (SRNL). The disposition options and a management evaluation process have been developed. Processing demonstrations and evaluations for these unique materials are under way. (authors)

  2. Actinide extraction methods

    DOE Patents [OSTI]

    Peterman, Dean R. (Idaho Falls, ID) [Idaho Falls, ID; Klaehn, John R. (Idaho Falls, ID) [Idaho Falls, ID; Harrup, Mason K. (Idaho Falls, ID) [Idaho Falls, ID; Tillotson, Richard D. (Moore, ID) [Moore, ID; Law, Jack D. (Pocatello, ID) [Pocatello, ID

    2010-09-21T23:59:59.000Z

    Methods of separating actinides from lanthanides are disclosed. A regio-specific/stereo-specific dithiophosphinic acid having organic moieties is provided in an organic solvent that is then contacted with an acidic medium containing an actinide and a lanthanide. The method can extend to separating actinides from one another. Actinides are extracted as a complex with the dithiophosphinic acid. Separation compositions include an aqueous phase, an organic phase, dithiophosphinic acid, and at least one actinide. The compositions may include additional actinides and/or lanthanides. A method of producing a dithiophosphinic acid comprising at least two organic moieties selected from aromatics and alkyls, each moiety having at least one functional group is also disclosed. A source of sulfur is reacted with a halophosphine. An ammonium salt of the dithiophosphinic acid product is precipitated out of the reaction mixture. The precipitated salt is dissolved in ether. The ether is removed to yield the dithiophosphinic acid.

  3. Development of Dodecaniobate Keggin Chain Materials as Alternative Sorbents for SR and Actinide Removal from High-Level Nuclear Waste Solutions

    SciTech Connect (OSTI)

    Nyman, May; Bonhomme, Francois

    2004-03-28T23:59:59.000Z

    The current baseline sorbent (monosodium titanate) for Sr and actinide removal from Savannah River Site's high level wastes has excellent adsorption capabilities for Sr but poor performance for the actinides. We are currently investigating the development of alternative materials that sorb radionuclides based on chemical affinity and/or size selectivity. The polyoxometalates, negatively-charged metal oxo clusters, have known metal binding properties and are of interest for radionuclide sequestration. We have developed a class of Keggin-ion based materials, where the Keggin ions are linked in 1- dimensional chains separated by hydrated, charge-balancing cations. These Nb-based materials are stable in the highly basic nuclear waste solutions and show good selectivity for Sr and Pu. Synthesis, characterization and structure of these materials in their native forms and Sr-exchanged forms will be presented.

  4. Process for removing metals from water

    DOE Patents [OSTI]

    Napier, J.M.; Hancher, C.M.; Hackett, G.D.

    1987-06-29T23:59:59.000Z

    A process for removing metals from water including the steps of prefiltering solids from the water, adjusting the pH to between about 2 and 3, reducing the amount of dissolved oxygen in the water, increasing the pH to between about 6 and 8, adding water-soluble sulfide to precipitate insoluble sulfide- and hydroxide-forming metals, adding a containing floc, and postfiltering the resultant solution. The postfiltered solution may optionally be eluted through an ion exchange resin to remove residual metal ions. 2 tabs.

  5. Selective Separation of Trivalent Actinides from Lanthanides by Aqueous Processing with Introduction of Soft Donor Atoms

    SciTech Connect (OSTI)

    Kenneth L. Nash

    2009-09-22T23:59:59.000Z

    Implementation of a closed loop nuclear fuel cycle requires the utilization of Pu-containing MOX fuels with the important side effect of increased production of the transplutonium actinides, most importantly isotopes of Am and Cm. Because the presence of these isotopes significantly impacts the long-term radiotoxicity of high level waste, it is important that effective methods for their isolation and/or transmutation be developed. Furthermore, since transmutation is most efficiently done in the absence of lanthanide fission products (high yield species with large thermal neutron absorption cross sections) it is important to have efficient procedures for the mutual separation of Am and Cm from the lanthanides. The chemistries of these elements are nearly identical, differing only in the slightly stronger strength of interaction of trivalent actinides with ligand donor atoms softer than O (N, Cl-, S). Research being conducted around the world has led to the development of new reagents and processes with considerable potential for this task. However, pilot scale testing of these reagents and processes has demonstrated the susceptibility of the new classes of reagents to radiolytic and hydrolytic degradation. In this project, separations of trivalent actinides from fission product lanthanides have been investigated in studies of 1) the extraction and chemical stability properties of a class of soft-donor extractants that are adapted from water-soluble analogs, 2) the application of water soluble soft-donor complexing agents in tandem with conventional extractant molecules emphasizing fundamental studies of the TALSPEAK Process. This research was conducted principally in radiochemistry laboratories at Washington State University. Collaborators at the Radiological Processing Laboratory (RPL) at the Pacific Northwest National Laboratory (PNNL) have contributed their unique facilities and capabilities, and have supported student internships at PNNL to broaden their academic experience. New information has been developed to qualify the extraction potential of a class of pyridine-functionalized tetraaza complexants indicating potential single contact Am-Nd separation factors of about 40. The methodology developed for characterization will find further application in our continuing efforts to synthesize and characterize new reagents for this separation. Significant new insights into the performance envelope and supporting information on the TALSPEAK process has also been developed.

  6. E-Print Network 3.0 - actinide recovery process Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    report, April 2006 The report can... is a trivalent actinide and a chemical analog to curium, and it has many chemical similarities to trivalent... complexes stay in solution....

  7. Process for removing sulfur from coal

    DOE Patents [OSTI]

    Aida, T.; Squires, T.G.; Venier, C.G.

    1983-08-11T23:59:59.000Z

    A process is disclosed for the removal of divalent organic and inorganic sulfur compounds from coal and other carbonaceous material. A slurry of pulverized carbonaceous material is contacted with an electrophilic oxidant which selectively oxidizes the divalent organic and inorganic compounds to trivalent and tetravalent compounds. The carbonaceous material is then contacted with a molten caustic which dissolves the oxidized sulfur compounds away from the hydrocarbon matrix.

  8. Process for removing mercury from aqueous solutions

    DOE Patents [OSTI]

    Googin, J.M.; Napier, J.M.; Makarewicz, M.A.; Meredith, P.F.

    1985-03-04T23:59:59.000Z

    A process for removing mercury from water to a level not greater than two parts per billion wherein an anion exchange material that is insoluble in water is contacted first with a sulfide containing compound and second with a compound containing a bivalent metal ion forming an insoluble metal sulfide. To this treated exchange material is contacted water containing mercury. The water containing not more than two parts per billion of mercury is separated from the exchange material.

  9. Process for removing mercury from aqueous solutions

    DOE Patents [OSTI]

    Googin, John M. (Oak Ridge, TN); Napier, John M. (Oak Ridge, TN); Makarewicz, Mark A. (Knoxville, TN); Meredith, Paul F. (Knoxville, TN)

    1986-01-01T23:59:59.000Z

    A process for removing mercury from water to a level not greater than two parts per billion wherein an anion exchange material that is insoluble in water is contacted first with a sulfide containing compound and second with a compound containing a bivalent metal ion forming an insoluble metal sulfide. To this treated exchange material is contacted water containing mercury. The water containing not more than two parts per billion of mercury is separated from the exchange material.

  10. Fly ash enhanced metal removal process

    SciTech Connect (OSTI)

    Nonavinakere, S. [Plexus Scientific Corp., Annapolis, MD (United States); Reed, B.E. [West Virginia Univ., Morgantown, WV (United States). Dept. of Civil Engineering

    1995-12-31T23:59:59.000Z

    The primary objective of the study was to evaluate the effectiveness of fly ashes from local thermal power plants in the removal of cadmium, nickel, chromium, lead, and copper from aqueous waste streams. Physical and chemical characteristics of fly ashes were determined, batch isotherm studies were conducted. A practical application of using fly ash in treating spent electroless nickel (EN) plating baths by modified conventional precipitation or solid enhanced metal removal process (SEMR) was investigated. In addition to nickel the EN baths also contains completing agents such as ammonium citrate and succinic acid reducing agents such as phosphate and hypophosphite. SEMR experiments were conducted at different pHs, fly ash type and concentrations, and settling times.

  11. Process for removing polychlorinated biphenyls from soil

    DOE Patents [OSTI]

    Hancher, C.W.; Saunders, M.B.; Googin, J.M.

    1984-11-16T23:59:59.000Z

    The present invention relates to a method of removing polychlorinated biphenyls from soil. The polychlorinated biphenyls are extracted from the soil by employing a liquid organic solvent dispersed in water in the ratio of about 1:3 to 3:1. The organic solvent includes such materials as short-chain hydrocarbons including kerosene or gasoline which are immiscible with water and are nonpolar. The organic solvent has a greater affinity for the PCB's than the soil so as to extract the PCB's from the soil upon contact. The organic solvent phase is separated from the suspended soil and water phase and distilled for permitting the recycle of the organic solvent phase and the concentration of the PCB's in the remaining organic phase. The present process can be satisfactorily practiced with soil containing 10 to 20% petroleum-based oils and organic fluids such as used in transformers and cutting fluids, coolants and the like which contain PCB's. The subject method provides for the removal of a sufficient concentration of PCB's from the soil to provide the soil with a level of PCB's within the guidelines of the Environmental Protection Agency.

  12. IMPROVED PROCESSES TO REMOVE NAPHTHENIC ACIDS

    SciTech Connect (OSTI)

    Aihua Zhang; Qisheng Ma; William A. Goddard; Yongchun Tang

    2004-04-28T23:59:59.000Z

    In the first year of this project, we have established our experimental and theoretical methodologies for studies of the catalytic decarboxylation process. We have developed both glass and stainless steel micro batch type reactors for the fast screening of various catalysts with reaction substrates of model carboxylic acid compounds and crude oil samples. We also developed novel product analysis methods such as GC analyses for organic acids and gaseous products; and TAN measurements for crude oil. Our research revealed the effectiveness of several solid catalysts such as NA-Cat-1 and NA-Cat-2 for the catalytic decarboxylation of model compounds; and NA-Cat-5{approx}NA-Cat-9 for the acid removal from crude oil. Our theoretical calculations propose a three-step concerted oxidative decarboxylation mechanism for the NA-Cat-1 catalyst.

  13. Improved Processes to Remove Naphthenic Acids

    SciTech Connect (OSTI)

    Aihua Zhang; Qisheng Ma; Kangshi Wang; Yongchun Tang; William A. Goddard

    2005-12-09T23:59:59.000Z

    In the past three years, we followed the work plan as we suggested in the proposal and made every efforts to fulfill the project objectives. Based on our large amount of creative and productive work, including both of experimental and theoretic aspects, we received important technical breakthrough on naphthenic acid removal process and obtained deep insight on catalytic decarboxylation chemistry. In detail, we established an integrated methodology to serve for all of the experimental and theoretical work. Our experimental investigation results in discovery of four type effective catalysts to the reaction of decarboxylation of model carboxylic acid compounds. The adsorption experiment revealed the effectiveness of several solid materials to naphthenic acid adsorption and acidity reduction of crude oil, which can be either natural minerals or synthesized materials. The test with crude oil also received promising results, which can be potentially developed into a practical process for oil industry. The theoretical work predicted several possible catalytic decarboxylation mechanisms that would govern the decarboxylation pathways depending on the type of catalysts being used. The calculation for reaction activation energy was in good agreement with our experimental measurements.

  14. Demonstration of a TODGA/TBP process for recovery of trivalent actinides and lanthanides from a PUREX raffinate

    SciTech Connect (OSTI)

    Modolo, G.; Asp, H.; Vijgen, H. [Forschungszentrum Juelich GmbH, Institut fuer Energieforschung, 52425 Juelich (Germany); Malmbeck, R.; Magnusson, D. [European Commission, JRC, Institute for Transuranium Elements - ITU, 76125 Karlsruhe (Germany); Sorel, C. [Commissariat a l'Energie Atomique Valrho - CEA, DRCP/SCPS, BP17171, 30207 Bagnols-sur-Ceze (France)

    2007-07-01T23:59:59.000Z

    The efficiency of the partitioning of trivalent actinides from a PUREX raffinate has been demonstrated with a TODGA + TBP extractant mixture dissolved in an industrial aliphatic solvent TPH. Based on the results coming from cold and hot batch extraction studies and with the aid of computer code calculations a continuous counter current process have been developed and two flowsheets were tested using miniature centrifugal contactors. The feed solutions was a synthetic PUREX raffinate, spiked with {sup 241}Am, {sup 244}Cm, {sup 252}Cf, {sup 152}Eu and {sup 134}Cs. More than 99.9 % of the trivalent actinides and lanthanides were extracted and back-extracted and very high decontamination factors to most fission products were obtained. Co-extraction of zirconium, molybdenum and palladium was prevented using oxalic acid and HEDTA. However 10% of ruthenium was extracted and only 3 % could be back extracted using diluted nitric acid. (authors)

  15. Process for removing technetium from iron and other metals

    DOE Patents [OSTI]

    Leitnaker, J.M.; Trowbridge, L.D.

    1999-03-23T23:59:59.000Z

    A process for removing technetium from iron and other metals comprises the steps of converting the molten, alloyed technetium to a sulfide dissolved in manganese sulfide, and removing the sulfide from the molten metal as a slag. 4 figs.

  16. Process for removing technetium from iron and other metals

    DOE Patents [OSTI]

    Leitnaker, James M. (Kingston, TN); Trowbridge, Lee D. (Oak Ridge, TN)

    1999-01-01T23:59:59.000Z

    A process for removing technetium from iron and other metals comprises the steps of converting the molten, alloyed technetium to a sulfide dissolved in manganese sulfide, and removing the sulfide from the molten metal as a slag.

  17. Process for particulate removal from coal liquids

    DOE Patents [OSTI]

    Rappe, Gerald C. (Macungie, PA)

    1983-01-01T23:59:59.000Z

    Suspended solid particulates are removed from liquefied coal products by first subjecting such products to hydroclone action for removal in the underflow of the larger size particulates, and then subjecting the overflow from said hydroclone action, comprising the residual finer particulates, to an electrostatic field in an electrofilter wherein such finer particulates are deposited in the bed of beads of dielectric material on said filter. The beads are periodically cleaned by backwashing to remove the accumulated solids.

  18. Thermochemistry of the actinides

    SciTech Connect (OSTI)

    Kleinschmidt, P.D.

    1993-10-01T23:59:59.000Z

    The measurement of equilibria by Knudsen effusion techniques and the enthalpy of formation of the actinide atoms is briefly discussed. Thermochemical data on the sublimation of the actinide fluorides is used to calculate the enthalpies of formation and entropies of the gaseous species. Estimates are made for enthalpies and entropies of the tetrafluorides and trifluorides for those systems where data is not available. The pressure of important species in the tetrafluoride sublimation processes is calculated based on this thermochemical data.

  19. Actinides-1981

    SciTech Connect (OSTI)

    Not Available

    1981-09-01T23:59:59.000Z

    Abstracts of 134 papers which were presented at the Actinides-1981 conference are presented. Approximately half of these papers deal with electronic structure of the actinides. Others deal with solid state chemistry, nuclear physic, thermodynamic properties, solution chemistry, and applied chemistry.

  20. IMPROVED PROCESSES TO REMOVE NAPHTHENIC ACIDS

    SciTech Connect (OSTI)

    Aihua Zhang; Qisheng Ma; Kangshi Wang, William A. Goddard, Yongchun Tang

    2005-05-05T23:59:59.000Z

    In the second year of this project, we continued our effort to develop low temperature decarboxylation catalysts and investigate the behavior of these catalysts at different reaction conditions. We conducted a large number of dynamic measurements with crude oil and model compounds to obtain the information at different reaction stages, which was scheduled as the Task2 in our work plan. We developed a novel adsorption method to remove naphthenic acid from crude oil using naturally occurring materials such as clays. Our results show promise as an industrial application. The theoretical modeling proposed several possible reaction pathways and predicted the reactivity depending on the catalysts employed. From all of these studies, we obtained more comprehensive understanding about catalytic decarboxylation and oil upgrading based on the naphthenic acid removal concept.

  1. Process for removing pyritic sulfur from bituminous coals

    DOE Patents [OSTI]

    Pawlak, Wanda (Edmonton, CA); Janiak, Jerzy S. (Edmonton, CA); Turak, Ali A. (Edmonton, CA); Ignasiak, Boleslaw L. (Edmonton, CA)

    1990-01-01T23:59:59.000Z

    A process is provided for removing pyritic sulfur and lowering ash content of bituminous coals by grinding the feed coal, subjecting it to micro-agglomeration with a bridging liquid containing heavy oil, separating the microagglomerates and separating them to a water wash to remove suspended pyritic sulfur. In one embodiment the coal is subjected to a second micro-agglomeration step.

  2. Process for selected gas oxide removal by radiofrequency catalysts

    DOE Patents [OSTI]

    Cha, C.Y.

    1993-09-21T23:59:59.000Z

    This process to remove gas oxides from flue gas utilizes adsorption on a char bed subsequently followed by radiofrequency catalysis enhancing such removal through selected reactions. Common gas oxides include SO[sub 2] and NO[sub x]. 1 figure.

  3. E-Print Network 3.0 - actinide ma recycling Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    by a chemical process. These plants can however not separate neptunium, americium and curium (minor actinides... to developing a process for separation of the minor actinides...

  4. E-Print Network 3.0 - actinide removal process Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    no 5-6, Tome 33, Mai-Juin 1972,page C3-57 RELATIVISTIC ELECTRONIC BAND STRUCTURE OF THE HEAVY METALS Summary: -metalliques presentent beaucoup de proprieth interessantes mais ma1...

  5. Process for removing cadmium from scrap metal

    DOE Patents [OSTI]

    Kronberg, J.W.

    1994-01-01T23:59:59.000Z

    A process for the recovery of a metal, in particular, cadmium contained in scrap, in a stable form. The process comprises the steps of mixing the cadmium-containing scrap with an ammonium carbonate solution, preferably at least a stoichiometric amount of ammonium carbonate, and/or free ammonia, and an oxidizing agent to form a first mixture so that the cadmium will react with the ammonium carbonate to form a water-soluble ammine complex; evaporating the first mixture so that ammine complex dissociates from the first mixture leaving carbonate ions to react with the cadmium and form a second mixture that includes cadmium carbonate; optionally adding water to the second mixture to form a third mixture; adjusting the pH of the third mixture to the acid range whereby the cadmium carbonate will dissolve; and adding at least a stoichiometric amount of sulfide, preferably in the form of hydrogen sulfide or an aqueous ammonium sulfide solution, to the third mixture to precipitate cadmium sulfide. This mixture of sulfide is then preferably digested by heating to facilitate precipitation of large particles of cadmium sulfide. The scrap may be divided by shredding or breaking up to exposure additional surface area. Finally, the precipitated cadmium sulfide can be mixed with glass formers and vitrified for permanent disposal.

  6. Process for removing cadmium from scrap metal

    DOE Patents [OSTI]

    Kronberg, J.W.

    1995-04-11T23:59:59.000Z

    A process is described for the recovery of a metal, in particular, cadmium contained in scrap, in a stable form. The process comprises the steps of mixing the cadmium-containing scrap with an ammonium carbonate solution, preferably at least a stoichiometric amount of ammonium carbonate, and/or free ammonia, and an oxidizing agent to form a first mixture so that the cadmium will react with the ammonium carbonate to form a water-soluble ammine complex; evaporating the first mixture so that ammine complex dissociates from the first mixture leaving carbonate ions to react with the cadmium and form a second mixture that includes cadmium carbonate; optionally adding water to the second mixture to form a third mixture; adjusting the pH of the third mixture to the acid range whereby the cadmium carbonate will dissolve; and adding at least a stoichiometric amount of sulfide, preferably in the form of hydrogen sulfide or an aqueous ammonium sulfide solution, to the third mixture to precipitate cadmium sulfide. This mixture of sulfide is then preferably digested by heating to facilitate precipitation of large particles of cadmium sulfide. The scrap may be divided by shredding or breaking up to expose additional surface area. Finally, the precipitated cadmium sulfide can be mixed with glass formers and vitrified for permanent disposal. 2 figures.

  7. Process for removing heavy metal compounds from heavy crude oil

    DOE Patents [OSTI]

    Cha, Chang Y. (Golden, CO); Boysen, John E. (Laramie, WY); Branthaver, Jan F. (Laramie, WY)

    1991-01-01T23:59:59.000Z

    A process is provided for removing heavy metal compounds from heavy crude oil by mixing the heavy crude oil with tar sand; preheating the mixture to a temperature of about 650.degree. F.; heating said mixture to up to 800.degree. F.; and separating tar sand from the light oils formed during said heating. The heavy metals removed from the heavy oils can be recovered from the spent sand for other uses.

  8. RAPID SEPARATION OF ACTINIDES AND RADIOSTRONTIUM IN VEGETATION SAMPLES

    SciTech Connect (OSTI)

    Maxwell, S.

    2010-06-01T23:59:59.000Z

    A new rapid method for the determination of actinides and radiostrontium in vegetation samples has been developed at the Savannah River Site Environmental Lab (Aiken, SC, USA) that can be used in emergency response situations or for routine analysis. The actinides in vegetation method utilizes a rapid sodium hydroxide fusion method, a lanthanum fluoride matrix removal step, and a streamlined column separation process with stacked TEVA, TRU and DGA Resin cartridges. Lanthanum was separated rapidly and effectively from Am and Cm on DGA Resin. Alpha emitters are prepared using rare earth microprecipitation for counting by alpha spectrometry. The purified {sup 90}Sr fractions are mounted directly on planchets and counted by gas flow proportional counting. The method showed high chemical recoveries and effective removal of interferences. The actinide and {sup 90}Sr in vegetation sample analysis can be performed in less than 8 h with excellent quality for emergency samples. The rapid fusion technique is a rugged sample digestion method that ensures that any refractory actinide particles or vegetation residue after furnace heating is effectively digested.

  9. Managing Inventories of Heavy Actinides

    SciTech Connect (OSTI)

    Wham, Robert M [ORNL; Patton, Bradley D [ORNL

    2011-01-01T23:59:59.000Z

    The Department of Energy (DOE) has stored a limited inventory of heavy actinides contained in irradiated targets, some partially processed, at the Savannah River Site (SRS) and Oak Ridge National Laboratory (ORNL). The 'heavy actinides' of interest include plutonium, americium, and curium isotopes; specifically 242Pu and 244Pu, 243Am, and 244/246/248Cm. No alternate supplies of these heavy actinides and no other capabilities for producing them are currently available. Some of these heavy actinide materials are important for use as feedstock for producing heavy isotopes and elements needed for research and commercial application. The rare isotope 244Pu is valuable for research, environmental safeguards, and nuclear forensics. Because the production of these heavy actinides was made possible only by the enormous investment of time and money associated with defense production efforts, the remaining inventories of these rare nuclear materials are an important part of the legacy of the Nuclear Weapons Program. Significant unique heavy actinide inventories reside in irradiated Mark-18A and Mark-42 targets at SRS and ORNL, with no plans to separate and store the isotopes for future use. Although the costs of preserving these heavy actinide materials would be considerable, for all practical purposes they are irreplaceable. The effort required to reproduce these heavy actinides today would likely cost billions of dollars and encompass a series of irradiation and chemical separation cycles for at least 50 years; thus, reproduction is virtually impossible. DOE has a limited window of opportunity to recover and preserve these heavy actinides before they are disposed of as waste. A path forward is presented to recover and manage these irreplaceable National Asset materials for future use in research, nuclear forensics, and other potential applications.

  10. Process for removing an organic compound from water

    DOE Patents [OSTI]

    Baker, Richard W. (Palo Alto, CA); Kaschemekat, Jurgen (Palo Alto, CA); Wijmans, Johannes G. (Menlo Park, CA); Kamaruddin, Henky D. (San Francisco, CA)

    1993-12-28T23:59:59.000Z

    A process for removing organic compounds from water is disclosed. The process involves gas stripping followed by membrane separation treatment of the stripping gas. The stripping step can be carried out using one or multiple gas strippers and using air or any other gas as stripping gas. The membrane separation step can be carried out using a single-stage membrane unit or a multistage unit. Apparatus for carrying out the process is also disclosed. The process is particularly suited for treatment of contaminated groundwater or industrial wastewater.

  11. Metal chelate process to remove pollutants from fluids

    DOE Patents [OSTI]

    Chang, S.G.T.

    1994-12-06T23:59:59.000Z

    The present invention relates to improved methods using an organic iron chelate to remove pollutants from fluids, such as flue gas. Specifically, the present invention relates to a process to remove NO[sub x] and optionally SO[sub 2] from a fluid using a metal ion (Fe[sup 2+]) chelate wherein the ligand is a dimercapto compound wherein the --SH groups are attached to adjacent carbon atoms (HS--C--C--SH) or (SH--C--CCSH) and contain a polar functional group so that the ligand of DMC chelate is water soluble. Alternatively, the DMC is covalently attached to a water insoluble substrate such as a polymer or resin, e.g., polystyrene. The chelate is regenerated using electroreduction or a chemical additive. The dimercapto compound bonded to a water insoluble substrate is also useful to lower the concentration or remove hazardous metal ions from an aqueous solution. 26 figures.

  12. Actinide Chemistry

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth (AOD)ProductssondeadjustsondeadjustAbout the Building Technologies OfficeAccountingGuideON STELLAWongActinide

  13. Process for removing sulfate anions from waste water

    DOE Patents [OSTI]

    Nilsen, David N. (Lebanon, OR); Galvan, Gloria J. (Albany, OR); Hundley, Gary L. (Corvallis, OR); Wright, John B. (Albany, OR)

    1997-01-01T23:59:59.000Z

    A liquid emulsion membrane process for removing sulfate anions from waste water is disclosed. The liquid emulsion membrane process includes the steps of: (a) providing a liquid emulsion formed from an aqueous strip solution and an organic phase that contains an extractant capable of removing sulfate anions from waste water; (b) dispersing the liquid emulsion in globule form into a quantity of waste water containing sulfate anions to allow the organic phase in each globule of the emulsion to extract and absorb sulfate anions from the waste water and (c) separating the emulsion including its organic phase and absorbed sulfate anions from the waste water to provide waste water containing substantially no sulfate anions.

  14. Ammonia removal process upgrade to the Acme Steel Coke Plant

    SciTech Connect (OSTI)

    Harris, J.L. [Acme Steel Co., Chicago, IL (United States). Chicago Coke Plant

    1995-12-01T23:59:59.000Z

    The need to upgrade the ammonia removal process at the Acme Steel Coke Plant developed with the installation of the benzene NESHAP (National Emission Standard for Hazardous Air Pollutants) equipment, specifically the replacement of the final cooler. At Acme Steel it was decided to replace the existing open cooling tower type final cooler with a closed loop direct spray tar/water final cooler. This new cooler has greatly reduced the emissions of benzene, ammonia, hydrogen sulfide and hydrogen cyanide to the atmosphere, bringing them into environmental compliance. At the time of its installation it was not fully recognized as to the effect this would have on the coke oven gas composition. In the late seventies the decision had been made at Acme Steel to stop the production of ammonia sulfate salt crystals. The direction chosen was to make a liquid ammonia sulfate solution. This product was used as a pickle liquor at first and then as a liquid fertilizer as more markets were developed. In the fall of 1986 the ammonia still was brought on line. The vapors generated from the operation of the stripping still are directed to the inlet of the ammonia absorber. At that point in time it was decided that an improvement to the cyclical ammonia removal process was needed. The improvements made were minimal yet allowed the circulation of solution through the ammonia absorber on a continuous basis. The paper describes the original batch process and the modifications made which allowed continuous removal.

  15. 33rd Actinide Separations Conference

    SciTech Connect (OSTI)

    McDonald, L M; Wilk, P A

    2009-05-04T23:59:59.000Z

    Welcome to the 33rd Actinide Separations Conference hosted this year by the Lawrence Livermore National Laboratory. This annual conference is centered on the idea of networking and communication with scientists from throughout the United States, Britain, France and Japan who have expertise in nuclear material processing. This conference forum provides an excellent opportunity for bringing together experts in the fields of chemistry, nuclear and chemical engineering, and actinide processing to present and discuss experiences, research results, testing and application of actinide separation processes. The exchange of information that will take place between you, and other subject matter experts from around the nation and across the international boundaries, is a critical tool to assist in solving both national and international problems associated with the processing of nuclear materials used for both defense and energy purposes, as well as for the safe disposition of excess nuclear material. Granlibakken is a dedicated conference facility and training campus that is set up to provide the venue that supports communication between scientists and engineers attending the 33rd Actinide Separations Conference. We believe that you will find that Granlibakken and the Lake Tahoe views provide an atmosphere that is stimulating for fruitful discussions between participants from both government and private industry. We thank the Lawrence Livermore National Laboratory and the United States Department of Energy for their support of this conference. We especially thank you, the participants and subject matter experts, for your involvement in the 33rd Actinide Separations Conference.

  16. Removal of mercury from coal via a microbial pretreatment process

    SciTech Connect (OSTI)

    Borole, Abhijeet P. (Knoxville, TN); Hamilton, Choo Y. (Knoxville, TN)

    2011-08-16T23:59:59.000Z

    A process for the removal of mercury from coal prior to combustion is disclosed. The process is based on use of microorganisms to oxidize iron, sulfur and other species binding mercury within the coal, followed by volatilization of mercury by the microorganisms. The microorganisms are from a class of iron and/or sulfur oxidizing bacteria. The process involves contacting coal with the bacteria in a batch or continuous manner. The mercury is first solubilized from the coal, followed by microbial reduction to elemental mercury, which is stripped off by sparging gas and captured by a mercury recovery unit, giving mercury-free coal. The mercury can be recovered in pure form from the sorbents via additional processing.

  17. Process for removal of hazardous air pollutants from coal

    DOE Patents [OSTI]

    Akers, David J. (Indiana, PA); Ekechukwu, Kenneth N. (Silver Spring, MD); Aluko, Mobolaji E. (Burtonsville, MD); Lebowitz, Howard E. (Mountain View, CA)

    2000-01-01T23:59:59.000Z

    An improved process for removing mercury and other trace elements from coal containing pyrite by forming a slurry of finely divided coal in a liquid solvent capable of forming ions or radicals having a tendency to react with constituents of pyrite or to attack the bond between pyrite and coal and/or to react with mercury to form mercury vapors, and heating the slurry in a closed container to a temperature of at least about 50.degree. C. to produce vapors of the solvent and withdrawing vapors including solvent and mercury-containing vapors from the closed container, then separating mercury from the vapors withdrawn.

  18. Extraction process for removing metallic impurities from alkalide metals

    DOE Patents [OSTI]

    Royer, Lamar T. (Knoxville, TN)

    1988-01-01T23:59:59.000Z

    A development is described for removing metallic impurities from alkali metals by employing an extraction process wherein the metallic impurities are extracted from a molten alkali metal into molten lithium metal due to the immiscibility of the alkali metals in lithium and the miscibility of the metallic contaminants or impurities in the lithium. The purified alkali metal may be readily separated from the contaminant-containing lithium metal by simple decanting due to the differences in densities and melting temperatures of the alkali metals as compared to lithium.

  19. RAPID SEPARATION METHOD FOR ACTINIDES IN EMERGENCY SOIL SAMPLES

    SciTech Connect (OSTI)

    Maxwell, S.; Culligan, B.; Noyes, G.

    2009-11-09T23:59:59.000Z

    A new rapid method for the determination of actinides in soil and sediment samples has been developed at the Savannah River Site Environmental Lab (Aiken, SC, USA) that can be used for samples up to 2 grams in emergency response situations. The actinides in soil method utilizes a rapid sodium hydroxide fusion method, a lanthanum fluoride soil matrix removal step, and a streamlined column separation process with stacked TEVA, TRU and DGA Resin cartridges. Lanthanum was separated rapidly and effectively from Am and Cm on DGA Resin. Vacuum box technology and rapid flow rates are used to reduce analytical time. Alpha sources are prepared using cerium fluoride microprecipitation for counting by alpha spectrometry. The method showed high chemical recoveries and effective removal of interferences. This new procedure was applied to emergency soil samples received in the NRIP Emergency Response exercise administered by the National Institute for Standards and Technology (NIST) in April, 2009. The actinides in soil results were reported within 4-5 hours with excellent quality.

  20. Kinetics of actinide complexation reactions

    SciTech Connect (OSTI)

    Nash, K.L.; Sullivan, J.C.

    1997-09-01T23:59:59.000Z

    Though the literature records extensive compilations of the thermodynamics of actinide complexation reactions, the kinetics of complex formation and dissociation reactions of actinide ions in aqueous solutions have not been extensively investigated. In light of the central role played by such reactions in actinide process and environmental chemistry, this situation is somewhat surprising. The authors report herein a summary of what is known about actinide complexation kinetics. The systems include actinide ions in the four principal oxidation states (III, IV, V, and VI) and complex formation and dissociation rates with both simple and complex ligands. Most of the work reported was conducted in acidic media, but a few address reactions in neutral and alkaline solutions. Complex formation reactions tend in general to be rapid, accessible only to rapid-scan and equilibrium perturbation techniques. Complex dissociation reactions exhibit a wider range of rates and are generally more accessible using standard analytical methods. Literature results are described and correlated with the known properties of the individual ions.

  1. Extraction processes and solvents for recovery of cesium, strontium, rare earth elements, technetium and actinides from liquid radioactive waste

    DOE Patents [OSTI]

    Zaitsev, Boris N. (St. Petersburg, RU); Esimantovskiy, Vyacheslav M. (St. Petersburg, RU); Lazarev, Leonard N. (St. Petersburg, RU); Dzekun, Evgeniy G. (Ozersk, RU); Romanovskiy, Valeriy N. (St. Petersburg, RU); Todd, Terry A. (Aberdeen, ID); Brewer, Ken N. (Arco, ID); Herbst, Ronald S. (Idaho Falls, ID); Law, Jack D. (Pocatello, ID)

    2001-01-01T23:59:59.000Z

    Cesium and strontium are extracted from aqueous acidic radioactive waste containing rare earth elements, technetium and actinides, by contacting the waste with a composition of a complex organoboron compound and polyethylene glycol in an organofluorine diluent mixture. In a preferred embodiment the complex organoboron compound is chlorinated cobalt dicarbollide, the polyethylene glycol has the formula RC.sub.6 H.sub.4 (OCH.sub.2 CH.sub.2).sub.n OH, and the organofluorine diluent is a mixture of bis-tetrafluoropropyl ether of diethylene glycol with at least one of bis-tetrafluoropropyl ether of ethylene glycol and bis-tetrafluoropropyl formal. The rare earths, technetium and the actinides (especially uranium, plutonium and americium), are extracted from the aqueous phase using a phosphine oxide in a hydrocarbon diluent, and reextracted from the resulting organic phase into an aqueous phase by using a suitable strip reagent.

  2. Process and system for removing impurities from a gas

    DOE Patents [OSTI]

    Henningsen, Gunnar; Knowlton, Teddy Merrill; Findlay, John George; Schlather, Jerry Neal; Turk, Brian S

    2014-04-15T23:59:59.000Z

    A fluidized reactor system for removing impurities from a gas and an associated process are provided. The system includes a fluidized absorber for contacting a feed gas with a sorbent stream to reduce the impurity content of the feed gas; a fluidized solids regenerator for contacting an impurity loaded sorbent stream with a regeneration gas to reduce the impurity content of the sorbent stream; a first non-mechanical gas seal forming solids transfer device adapted to receive an impurity loaded sorbent stream from the absorber and transport the impurity loaded sorbent stream to the regenerator at a controllable flow rate in response to an aeration gas; and a second non-mechanical gas seal forming solids transfer device adapted to receive a sorbent stream of reduced impurity content from the regenerator and transfer the sorbent stream of reduced impurity content to the absorber without changing the flow rate of the sorbent stream.

  3. Process for removing naphthenic acids from petroleum distillates

    SciTech Connect (OSTI)

    Danzik, M.

    1987-01-06T23:59:59.000Z

    A liquid extraction process is described for removing naphthenic acids from naphthenic acid containing petroleum distillates boiling within the range of about 180/sup 0/-600/sup 0/C. and having an acid number of at least about 0.2 which process comprises the steps of: (a) intimately contacting the petroleum distillates with a solvent consisting essentially of methanol, water, and about from 2-20 wt. % ammonia and having a methanol: water ratio in the range of about from 0.2 to 3 parts by weight of methanol per part by weight of water and using an ammonia to petroleum distillate ratio of about 0.1-1 part by weight of ammonia per 100 parts by weight of the petroleum distillate. This selectively extracts the naphthenic acids into the solvent and yielding an immiscible two-phase liquid mixture, one of which is naphthenic acid-rich solvent phase and the other of which is a substantially napthenic acid-free petroleum distillate phase; and (b) separating and respectively recovering the naphtenic acid-rich solvent phase and petroleum distillate phase.

  4. Actinide halide complexes

    DOE Patents [OSTI]

    Avens, L.R.; Zwick, B.D.; Sattelberger, A.P.; Clark, D.L.; Watkin, J.G.

    1992-11-24T23:59:59.000Z

    A compound is described of the formula MX[sub n]L[sub m] wherein M is a metal atom selected from the group consisting of thorium, plutonium, neptunium or americium, X is a halide atom, n is an integer selected from the group of three or four, L is a coordinating ligand selected from the group consisting of aprotic Lewis bases having an oxygen-, nitrogen-, sulfur-, or phosphorus-donor, and m is an integer selected from the group of three or four for monodentate ligands or is the integer two for bidentate ligands, where the sum of n+m equals seven or eight for monodentate ligands or five or six for bidentate ligands. A compound of the formula MX[sub n] wherein M, X, and n are as previously defined, and a process of preparing such actinide metal compounds are described including admixing the actinide metal in an aprotic Lewis base as a coordinating solvent in the presence of a halogen-containing oxidant.

  5. Actinide halide complexes

    DOE Patents [OSTI]

    Avens, Larry R. (Los Alamos, NM); Zwick, Bill D. (Santa Fe, NM); Sattelberger, Alfred P. (Los Alamos, NM); Clark, David L. (Los Alamos, NM); Watkin, John G. (Los Alamos, NM)

    1992-01-01T23:59:59.000Z

    A compound of the formula MX.sub.n L.sub.m wherein M is a metal atom selected from the group consisting of thorium, plutonium, neptunium or americium, X is a halide atom, n is an integer selected from the group of three or four, L is a coordinating ligand selected from the group consisting of aprotic Lewis bases having an oxygen-, nitrogen-, sulfur-, or phosphorus-donor, and m is an integer selected from the group of three or four for monodentate ligands or is the integer two for bidentate ligands, where the sum of n+m equals seven or eight for monodentate ligands or five or six for bidentate ligands, a compound of the formula MX.sub.n wherein M, X, and n are as previously defined, and a process of preparing such actinide metal compounds including admixing the actinide metal in an aprotic Lewis base as a coordinating solvent in the presence of a halogen-containing oxidant, are provided.

  6. Process for removing sulfur dioxide from flue gases

    SciTech Connect (OSTI)

    Robinson, M.W. Jr.

    1989-08-29T23:59:59.000Z

    This patent describes an improvement in a dry process for the removal of sulfur dioxide from flue gases by the addition thereto of hydrated lime containing sugar in a coal combustion unit, wherein the flue gases result from the combustion of a coal in a combustion chamber, and the flue gases are treated in an electrostatic precipitator prior to discharge to the atmosphere the improvement comprising: passing the flue gases, after the addition of the hydrated lime is of fine particles of a specific surface of 7 to 25 square meters per gram, through a conduit towards the electrostatic precipitator; and adding an aqueous media to the flue gases in the conduit in an amount to increase the water content of the flue gases and cool the same by evaporative cooling to a temperature no lower than 20{sup 0}F. about the dew point of the gas, so as to avoid forming water droplets in the gas, so as to prevent condensation of water therefrom.

  7. RAPID SEPARATION METHOD FOR ACTINIDES IN EMERGENCY AIR FILTER SAMPLES

    SciTech Connect (OSTI)

    Maxwell, S.; Noyes, G.; Culligan, B.

    2010-02-03T23:59:59.000Z

    A new rapid method for the determination of actinides and strontium in air filter samples has been developed at the Savannah River Site Environmental Lab (Aiken, SC, USA) that can be used in emergency response situations. The actinides and strontium in air filter method utilizes a rapid acid digestion method and a streamlined column separation process with stacked TEVA, TRU and Sr Resin cartridges. Vacuum box technology and rapid flow rates are used to reduce analytical time. Alpha emitters are prepared using cerium fluoride microprecipitation for counting by alpha spectrometry. The purified {sup 90}Sr fractions are mounted directly on planchets and counted by gas flow proportional counting. The method showed high chemical recoveries and effective removal of interferences. This new procedure was applied to emergency air filter samples received in the NRIP Emergency Response exercise administered by the National Institute for Standards and Technology (NIST) in April, 2009. The actinide and {sup 90}Sr in air filter results were reported in {approx}4 hours with excellent quality.

  8. Research in actinide chemistry

    SciTech Connect (OSTI)

    Choppin, G.R.

    1993-01-01T23:59:59.000Z

    This research studies the behavior of the actinide elements in aqueous solution. The high radioactivity of the transuranium actinides limits the concentrations which can be studied and, consequently, limits the experimental techniques. However, oxidation state analogs (trivalent lanthanides, tetravalent thorium, and hexavalent uranium) do not suffer from these limitations. Behavior of actinides in the environment are a major USDOE concern, whether in connection with long-term releases from a repository, releases from stored defense wastes or accidental releases in reprocessing, etc. Principal goal of our research was expand the thermodynamic data base on complexation of actinides by natural ligands (e.g., OH[sup [minus

  9. Total nitrogen removal in a hybrid, membrane-aerated activated sludge process

    E-Print Network [OSTI]

    Nerenberg, Robert

    Total nitrogen removal in a hybrid, membrane-aerated activated sludge process Leon S. Downing wastewater. Air-filled hollow-fiber membranes are incorporated into an activated sludge tank removal in activated sludge. Ş 2008 Elsevier Ltd. All rights reserved. 1. Introduction The removal

  10. Fluidized bed gasification ash reduction and removal process

    DOE Patents [OSTI]

    Schenone, Carl E. (Madison, PA); Rosinski, Joseph (Vanderbilt, PA)

    1984-12-04T23:59:59.000Z

    In a fluidized bed gasification system an ash removal system to reduce the particulate ash to a maximum size or smaller, allow the ash to cool to a temperature lower than the gasifier and remove the ash from the gasifier system. The system consists of a crusher, a container containing level probes and a means for controlling the rotational speed of the crusher based on the level of ash within the container.

  11. Synthesis of actinide nitrides, phosphides, sulfides and oxides

    DOE Patents [OSTI]

    Van Der Sluys, William G. (Missoula, MT); Burns, Carol J. (Los Alamos, NM); Smith, David C. (Los Alamos, NM)

    1992-01-01T23:59:59.000Z

    A process of preparing an actinide compound of the formula An.sub.x Z.sub.y wherein An is an actinide metal atom selected from the group consisting of thorium, uranium, plutonium, neptunium, and americium, x is selected from the group consisting of one, two or three, Z is a main group element atom selected from the group consisting of nitrogen, phosphorus, oxygen and sulfur and y is selected from the group consisting of one, two, three or four, by admixing an actinide organometallic precursor wherein said actinide is selected from the group consisting of thorium, uranium, plutonium, neptunium, and americium, a suitable solvent and a protic Lewis base selected from the group consisting of ammonia, phosphine, hydrogen sulfide and water, at temperatures and for time sufficient to form an intermediate actinide complex, heating said intermediate actinide complex at temperatures and for time sufficient to form the actinide compound, and a process of depositing a thin film of such an actinide compound, e.g., uranium mononitride, by subliming an actinide organometallic precursor, e.g., a uranium amide precursor, in the presence of an effectgive amount of a protic Lewis base, e.g., ammonia, within a reactor at temperatures and for time sufficient to form a thin film of the actinide compound, are disclosed.

  12. SULFURIC ACID REMOVAL PROCESS EVALUATION: SHORT-TERM RESULTS

    SciTech Connect (OSTI)

    Gary M. Blythe; Richard McMillan

    2002-03-04T23:59:59.000Z

    The objective of this project is to demonstrate the use of alkaline reagents injected into the furnace of coal-fired boilers as a means of controlling sulfuric acid emissions. Sulfuric acid controls are becoming of increasing interest to utilities with coal-fired units for a number of reasons. Sulfuric acid is a Toxic Release Inventory species, a precursor to acid aerosol/condensable emissions, and can cause a variety of plant operation problems such as air heater plugging and fouling, back-end corrosion, and plume opacity. These issues will likely be exacerbated with the retrofit of SCR for NOX control on some coal-fired plants, as SCR catalysts are known to further oxidize a portion of the flue gas SO{sub 2} to SO{sub 3}. The project is testing the effectiveness of furnace injection of four different calcium- and/or magnesium-based alkaline sorbents on full-scale utility boilers. These reagents have been tested during four one- to two-week tests conducted on two FirstEnergy Bruce Mansfield Plant units. One of the sorbents tested was a magnesium hydroxide slurry produced from a wet flue gas desulfurization system waste stream, from a system that employs a Thiosorbic{reg_sign} Lime scrubbing process. The other three sorbents are available commercially and include dolomite, pressure-hydrated dolomitic lime, and commercial magnesium hydroxide. The dolomite reagent was injected as a dry powder through out-of-service burners, while the other three reagents were injected as slurries through air-atomizing nozzles into the front wall of upper furnace, either across from the nose of the furnace or across from the pendant superheater tubes. After completing the four one- to two-week tests, the most promising sorbents were selected for longer-term (approximately 25-day) full-scale tests. The longer-term tests are being conducted to confirm the effectiveness of the sorbents tested over extended operation and to determine balance-of-plant impacts. This reports presents the results of the short-term tests; the long-term test results will be reported in a later document. The short-term test results showed that three of the four reagents tested, dolomite powder, commercial magnesium hydroxide slurry, and byproduct magnesium hydroxide slurry, were able to achieve 90% or greater removal of sulfuric acid compared to baseline levels. The molar ratio of alkali to flue gas sulfuric acid content (under baseline conditions) required to achieve 90% sulfuric acid removal was lowest for the byproduct magnesium hydroxide slurry. However, this result may be confounded because this was the only one of the three slurries tested with injection near the top of the furnace across from the pendant superheater platens. Injection at the higher level was demonstrated to be advantageous for this reagent over injection lower in the furnace, where the other slurries were tested.

  13. SULFURIC ACID REMOVAL PROCESS EVALUATION: SHORT-TERM RESULTS

    SciTech Connect (OSTI)

    Gary M. Blythe; Richard McMillan

    2002-02-04T23:59:59.000Z

    The objective of this project is to demonstrate the use of alkaline reagents injected into the furnace of coal-fired boilers as a means of controlling sulfuric acid emissions. Sulfuric acid controls are becoming of increasing interest to utilities with coal-fired units for a number of reasons. Sulfuric acid is a Toxic Release Inventory species, a precursor to acid aerosol/condensable emissions, and can cause a variety of plant operation problems such as air heater plugging and fouling, back-end corrosion, and plume opacity. These issues will likely be exacerbated with the retrofit of SCR for NO{sub x} control on some coal-fired plants, as SCR catalysts are known to further oxidize a portion of the flue gas SO{sub 2} to SO{sub 3}. The project is testing the effectiveness of furnace injection of four different calcium- and/or magnesium-based alkaline sorbents on full-scale utility boilers. These reagents have been tested during four one- to two-week tests conducted on two First Energy Bruce Mansfield Plant units. One of the sorbents tested was a magnesium hydroxide slurry produced from a wet flue gas desulfurization system waste stream, from a system that employs a Thiosorbic{reg_sign} Lime scrubbing process. The other three sorbents are available commercially and include dolomite, pressure-hydrated dolomitic lime, and commercial magnesium hydroxide. The dolomite reagent was injected as a dry powder through out-of-service burners, while the other three reagents were injected as slurries through air-atomizing nozzles into the front wall of upper furnace, either across from the nose of the furnace or across from the pendant superheater tubes. After completing the four one- to two-week tests, the most promising sorbents were selected for longer-term (approximately 25-day) full-scale tests. The longer-term tests are being conducted to confirm the effectiveness of the sorbents tested over extended operation and to determine balance-of-plant impacts. This reports presents the results of the short-term tests; the long-term test results will be reported in a later document. The short-term test results showed that three of the four reagents tested, dolomite powder, commercial magnesium hydroxide slurry, and byproduct magnesium hydroxide slurry, were able to achieve 90% or greater removal of sulfuric acid compared to baseline levels. The molar ratio of alkali to flue gas sulfuric acid content (under baseline conditions) required to achieve 90% sulfuric acid removal was lowest for the byproduct magnesium hydroxide slurry. However, this result may be confounded because this was the only one of the three slurries tested with injection near the top of the furnace across from the pendant superheater platens. Injection at the higher level was demonstrated to be advantageous for this reagent over injection lower in the furnace, where the other slurries were tested.

  14. Advanced Aqueous Separation Systems for Actinide Partitioning

    SciTech Connect (OSTI)

    Nash, Kenneth L.; Sue Clark; G. Patrick Meier; Spiro Alexandratos; Robert Paine; Robert Hancock; Dale Ensor

    2012-03-21T23:59:59.000Z

    One of the most challenging aspects of advanced processing of spent nuclear fuel is the need to isolate transuranium elements from fission product lanthanides. This project expanded the scope of earlier investigations of americium (Am) partitioning from the lanthanides with the synthesis of new separations materials and a centralized focus on radiochemical characterization of the separation systems that could be developed based on these new materials. The primary objective of this program was to explore alternative materials for actinide separations and to link the design of new reagents for actinide separations to characterizations based on actinide chemistry. In the predominant trivalent oxidation state, the chemistry of lanthanides overlaps substantially with that of the trivalent actinides and their mutual separation is quite challenging.

  15. Advanced Extraction Methods for Actinide/Lanthanide Separations

    SciTech Connect (OSTI)

    Scott, M.J.

    2005-12-01T23:59:59.000Z

    The separation of An(III) ions from chemically similar Ln(III) ions is perhaps one of the most difficult problems encountered during the processing of nuclear waste. In the 3+ oxidation states, the metal ions have an identical charge and roughly the same ionic radius. They differ strictly in the relative energies of their f- and d-orbitals, and to separate these metal ions, ligands will need to be developed that take advantage of this small but important distinction. The extraction of uranium and plutonium from nitric acid solution can be performed quantitatively by the extraction with the TBP (tributyl phosphate). Commercially, this process has found wide use in the PUREX (plutonium uranium extraction) reprocessing method. The TRUEX (transuranium extraction) process is further used to coextract the trivalent lanthanides and actinides ions from HLLW generated during PUREX extraction. This method uses CMPO [(N, N-diisobutylcarbamoylmethyl) octylphenylphosphineoxide] intermixed with TBP as a synergistic agent. However, the final separation of trivalent actinides from trivalent lanthanides still remains a challenging task. In TRUEX nitric acid solution, the Am(III) ion is coordinated by three CMPO molecules and three nitrate anions. Taking inspiration from this data and previous work with calix[4]arene systems, researchers on this project have developed a C3-symmetric tris-CMPO ligand system using a triphenoxymethane platform as a base. The triphenoxymethane ligand systems have many advantages for the preparation of complex ligand systems. The compounds are very easy to prepare. The steric and solubility properties can be tuned through an extreme range by the inclusion of different alkoxy and alkyl groups such as methyoxy, ethoxy, t-butoxy, methyl, octyl, t-pentyl, or even t-pentyl at the ortho- and para-positions of the aryl rings. The triphenoxymethane ligand system shows promise as an improved extractant for both tetravalent and trivalent actinide recoveries form high level liquid wastes and a general actinide clean-up procedure. The selectivity of the standard extractant for tetravalent actinides, (N,N-diisobutylcarbamoylmethyl) octylphenylphosphineoxide (CMPO), was markedly improved by the attachment of three CMPO-like functions onto a triphenoxymethane platform, and a ligand that is both highly selective and effective for An(IV) ions was isolated. A 10 fold excess of ligand will remove virtually all of the 4+ actinides from the acidic layer without extracting appreciable quantities of An(III) and Ln(III) unlike simple CMPO ligands. Inspired by the success of the DIAMEX industrial process for extractions, three new tripodal chelates bearing three diglycolamide and thiodiglycolamide units precisely arranged on a triphenoxymethane platform have been synthesized for an highly efficient extraction of trivalent f-element cations from nitric acid media. A single equivalent of ligand will remove 80% of the Ln(III) ion from the acidic layer since the ligand is perfectly suited to accommodate the tricapped trigonal prismatic geometry preferred by the metal center. The ligand is perhaps the most efficient binder available for the heavier lanthanides and due to this unique attribute, the extraction event can be easily followed by 1H NMR spectroscopy confirming the formation of a TPP complex. The most lipophilic di-n-butyl tris-diglycolamide was found to be a significantly weaker extractant in comparison to the di-isopropyl analogs. The tris-thiodiglycolamide derivative proved to be an ineffective chelate for f-elements and demonstrated the importance of the etheric oxygens in the metal binding. The results presented herein clearly demonstrate a cooperative action of these three ligating groups within a single molecule, confirmed by composition and structure of the extracted complexes, and since actinides prefer to have high coordination numbers, the ligands should be particularly adept at binding with three arms. The use of such an extractant permits the extraction of metal ions form highly acidic environment through the ability

  16. Overview of actinide chemistry in the WIPP

    SciTech Connect (OSTI)

    Borkowski, Marian [Los Alamos National Laboratory; Lucchini, Jean - Francois [Los Alamos National Laboratory; Richmann, Michael K [Los Alamos National Laboratory; Reed, Donald T [Los Alamos National Laboratory; Khaing, Hnin [Los Alamos National Laboratory; Swanson, Juliet [Los Alamos National Laboratory

    2009-01-01T23:59:59.000Z

    The year 2009 celebrates 10 years of safe operations at the Waste Isolation Pilot Plant (WIPP), the only nuclear waste repository designated to dispose defense-related transuranic (TRU) waste in the United States. Many elements contributed to the success of this one-of-the-kind facility. One of the most important of these is the chemistry of the actinides under WIPP repository conditions. A reliable understanding of the potential release of actinides from the site to the accessible environment is important to the WIPP performance assessment (PA). The environmental chemistry of the major actinides disposed at the WIPP continues to be investigated as part of the ongoing recertification efforts of the WIPP project. This presentation provides an overview of the actinide chemistry for the WIPP repository conditions. The WIPP is a salt-based repository; therefore, the inflow of brine into the repository is minimized, due to the natural tendency of excavated salt to re-seal. Reducing anoxic conditions are expected in WIPP because of microbial activity and metal corrosion processes that consume the oxygen initially present. Should brine be introduced through an intrusion scenario, these same processes will re-establish reducing conditions. In the case of an intrusion scenario involving brine, the solubilization of actinides in brine is considered as a potential source of release to the accessible environment. The following key factors establish the concentrations of dissolved actinides under subsurface conditions: (1) Redox chemistry - The solubility of reduced actinides (III and IV oxidation states) is known to be significantly lower than the oxidized forms (V and/or VI oxidation states). In this context, the reducing conditions in the WIPP and the strong coupling of the chemistry for reduced metals and microbiological processes with actinides are important. (2) Complexation - For the anoxic, reducing and mildly basic brine systems in the WIPP, the most important inorganic complexants are expected to be carbonate/bicarbonate and hydroxide. There are also organic complexants in TRU waste with the potential to strongly influence actinide solubility. (3) Intrinsic and pseudo-actinide colloid formation - Many actinide species in their expected oxidation states tend to form colloids or strongly associate with non actinide colloids present (e.g., microbial, humic and organic). In this context, the relative importance of actinides, based on the TRU waste inventory, with respect to the potential release of actinides from the WIPP, is greater for plutonium and americium, and to less extent for uranium and thorium. The most important oxidation states for WIPP-relevant conditions are III and IV. We will present an update of the literature on WIPP-specific data, and a summary of the ongoing research related to actinide chemistry in the WIPP performed by the Los Alamos National Laboratory (LANL) Actinide Chemistry and Repository Science (ACRSP) team located in Carlsbad, NM [Reed 2007, Lucchini 2007, and Reed 2006].

  17. Thief process for the removal of mercury from flue gas

    DOE Patents [OSTI]

    Pennline, Henry W. (Bethel Park, PA); Granite, Evan J. (Wexford, PA); Freeman, Mark C. (South Park Township, PA); Hargis, Richard A. (Canonsburg, PA); O'Dowd, William J. (Charleroi, PA)

    2003-02-18T23:59:59.000Z

    A system and method for removing mercury from the flue gas of a coal-fired power plant is described. Mercury removal is by adsorption onto a thermally activated sorbent produced in-situ at the power plant. To obtain the thermally activated sorbent, a lance (thief) is inserted into a location within the combustion zone of the combustion chamber and extracts a mixture of semi-combusted coal and gas. The semi-combusted coal has adsorptive properties suitable for the removal of elemental and oxidized mercury. The mixture of semi-combusted coal and gas is separated into a stream of gas and semi-combusted coal that has been converted to a stream of thermally activated sorbent. The separated stream of gas is recycled to the combustion chamber. The thermally activated sorbent is injected into the duct work of the power plant at a location downstream from the exit port of the combustion chamber. Mercury within the flue gas contacts and adsorbs onto the thermally activated sorbent. The sorbent-mercury combination is removed from the plant by a particulate collection system.

  18. Process for removing polymer-forming impurities from naphtha fraction

    DOE Patents [OSTI]

    Kowalczyk, D.C.; Bricklemyer, B.A.; Svoboda, J.J.

    1983-12-27T23:59:59.000Z

    Polymer precursor materials are vaporized without polymerization or are removed from a raw naphtha fraction by passing the raw naphtha to a vaporization zone and vaporizing the naphtha in the presence of a wash oil while stripping with hot hydrogen to prevent polymer deposits in the equipment. 2 figs.

  19. Improved method for extracting lanthanides and actinides from acid solutions

    DOE Patents [OSTI]

    Horwitz, E.P.; Kalina, D.G.; Kaplan, L.; Mason, G.W.

    1983-07-26T23:59:59.000Z

    A process for the recovery of actinide and lanthanide values from aqueous acidic solutions uses a new series of neutral bi-functional extractants, the alkyl(phenyl)-N,N-dialkylcarbamoylmethylphosphine oxides. The process is suitable for the separation of actinide and lanthanide values from fission product values found together in high-level nuclear reprocessing waste solutions.

  20. High Metal Removal Rate Process for Machining Difficult Materials

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-UpHeatMulti-Dimensional Subject:Ground Hawaii CleanHeatinHighMetal Removal

  1. Process for off-gas particulate removal and apparatus therefor

    DOE Patents [OSTI]

    Carl, D.E.

    1997-10-21T23:59:59.000Z

    In the event of a breach in the off-gas line of a melter operation requiring closure of the line, a secondary vessel vent line is provided with a particulate collector utilizing atomization for removal of large particulates from the off-gas. The collector receives the gas containing particulates and directs a portion of the gas through outer and inner annular channels. The collector further receives a fluid, such as water, which is directed through the outer channel together with a second portion of the particulate-laden gas. The outer and inner channels have respective ring-like termination apertures concentrically disposed adjacent one another on the outer edge of the downstream side of the particulate collector. Each of the outer and inner channels curves outwardly away from the collector`s centerline in proceeding toward the downstream side of the collector. Gas flow in the outer channel maintains the fluid on the channel`s wall in the form of a ``wavy film,`` while the gas stream from the inner channel shears the fluid film as it exits the outer channel in reducing the fluid to small droplets. Droplets formed by the collector capture particulates in the gas stream by one of three mechanisms: impaction, interception or Brownian diffusion in removing the particulates. The particulate-laden droplets are removed from the fluid stream by a vessel vent condenser or mist eliminator. 4 figs.

  2. Process for off-gas particulate removal and apparatus therefor

    DOE Patents [OSTI]

    Carl, Daniel E. (Orchard Park, NY)

    1997-01-01T23:59:59.000Z

    In the event of a breach in the off-gas line of a melter operation requiring closure of the line, a secondary vessel vent line is provided with a particulate collector utilizing atomization for removal of large particulates from the off-gas. The collector receives the gas containing particulates and directs a portion of the gas through outer and inner annular channels. The collector further receives a fluid, such as water, which is directed through the outer channel together with a second portion of the particulate-laden gas. The outer and inner channels have respective ring-like termination apertures concentrically disposed adjacent one another on the outer edge of the downstream side of the particulate collector. Each of the outer and inner channels curves outwardly away from the collector's centerline in proceeding toward the downstream side of the collector. Gasflow in the outer channel maintains the fluid on the channel's wall in the form of a "wavy film," while the gas stream from the inner channel shears the fluid film as it exits the outer channel in reducing the fluid to small droplets. Droplets formed by the collector capture particulates in the gas stream by one of three mechanisms: impaction, interception or Brownian diffusion in removing the particulates. The particulate-laden droplets are removed from the fluid stream by a vessel vent condenser or mist eliminator.

  3. SULFURIC ACID REMOVAL PROCESS EVALUATION: LONG-TERM RESULTS

    SciTech Connect (OSTI)

    Gary M. Blythe; Richard McMillan

    2002-07-03T23:59:59.000Z

    The objective of this project is to demonstrate the use of alkaline reagents injected into the furnace of coal-fired boilers as a means of controlling sulfuric acid emissions. The project is being co-funded by the U.S. DOE National Energy Technology Laboratory, under Cooperative Agreement DE-FC26-99FT40718, along with EPRI, the American Electric Power Company (AEP), FirstEnergy Corp., the Tennessee Valley Authority, and Dravo Lime, Inc. Sulfuric acid controls are becoming of increasing interest to power generators with coal-fired units for a number of reasons. Sulfuric acid is a Toxic Release Inventory species and can cause a variety of plant operation problems such as air heater plugging and fouling, back-end corrosion, and plume opacity. These issues will likely be exacerbated with the retrofit of selective catalytic reduction (SCR) for NO{sub x} control on many coal-fired plants, as SCR catalysts are known to further oxidize a portion of the flue gas SO{sub 2} to SO{sub 3}. The project previously tested the effectiveness of furnace injection of four different calcium-and/or magnesium-based alkaline sorbents on full-scale utility boilers. These reagents were tested during four one- to two-week tests conducted on two FirstEnergy Bruce Mansfield Plant (BMP) units. One of the sorbents tested was a magnesium hydroxide byproduct slurry produced from a modified Thiosorbic{reg_sign} Lime wet flue gas desulfurization system. The other three sorbents are available commercially and include dolomite, pressure-hydrated dolomitic lime, and commercial magnesium hydroxide. The dolomite reagent was injected as a dry powder through out-of-service burners, while the other three reagents were injected as slurries through air-atomizing nozzles inserted through the front wall of the upper furnace, either across from the nose of the furnace or across from the pendant superheater tubes. After completing the four one- to two-week tests, the most promising sorbents were selected for longer-term (approximately 25-day) full-scale tests on two different units. The longer-term tests were conducted to confirm the effectiveness of the sorbents tested over extended operation on two different boilers, and to determine balance-of-plant impacts. The first long-term test was conducted on FirstEnergy's BMP, Unit 3, and the second test was conducted on AEP's Gavin Plant, Unit 1. The Gavin Plant testing provided an opportunity to evaluate the effects of sorbent injected into the furnace on SO{sub 3} formed across an operating SCR reactor. This report presents the results from those long-term tests. The tests determined the effectiveness of injecting commercially available magnesium hydroxide slurry (Gavin Plant) and byproduct magnesium hydroxide slurry (both Gavin Plant and BMP) for sulfuric acid control. The results show that injecting either slurry could achieve up to 70 to 75% overall sulfuric acid removal. At BMP, this overall removal was limited by the need to maintain acceptable electrostatic precipitator (ESP) particulate control performance. At Gavin Plant, the overall sulfuric acid removal was limited because the furnace injected sorbent was less effective at removing SO{sub 3} formed across the SCR system installed on the unit for NOX control than at removing SO{sub 3} formed in the furnace. The long-term tests also determined balance-of-plant impacts from slurry injection during the two tests. These include impacts on boiler back-end temperatures and pressure drops, SCR catalyst properties, ESP performance, removal of other flue gas species, and flue gas opacity. For the most part the balance-of-plant impacts were neutral to positive, although adverse effects on ESP performance became an issue during the BMP test.

  4. High Metal Removal Rate Process for Machining Difficult Materials

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-UpHeatMulti-Dimensional Subject:Ground Hawaii CleanHeatinHighMetal Removal ADVANCED

  5. High Metal Removal Rate Process for Machining Difficult Materials

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-UpHeatMulti-Dimensional Subject:Ground Hawaii CleanHeatinHighMetal Removal ADVANCEDHybrid

  6. acid removal process: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    synthesis from biomass pyrolysis with in situ carbon dioxideof pyrolysis, combustion and gasification of three biomassand biomass, undergoes several different processes andor...

  7. Determining the removal effectiveness of flame retardants from drinking water treatment processes

    E-Print Network [OSTI]

    Lin, Joseph C. (Joseph Chris), 1981-

    2004-01-01T23:59:59.000Z

    Low concentrations of xenobiotic chemicals have recently become a concern in the surface water environment. The concern expands to drinking water treatment processes, and whether or not they remove these chemicals while ...

  8. Water treatment process and system for metals removal using Saccharomyces cerevisiae

    DOE Patents [OSTI]

    Krauter, Paula A. W. (Livermore, CA); Krauter, Gordon W. (Livermore, CA)

    2002-01-01T23:59:59.000Z

    A process and a system for removal of metals from ground water or from soil by bioreducing or bioaccumulating the metals using metal tolerant microorganisms Saccharomyces cerevisiae. Saccharomyces cerevisiae is tolerant to the metals, able to bioreduce the metals to the less toxic state and to accumulate them. The process and the system is useful for removal or substantial reduction of levels of chromium, molybdenum, cobalt, zinc, nickel, calcium, strontium, mercury and copper in water.

  9. Processes to remove acid forming gases from exhaust gases

    DOE Patents [OSTI]

    Chang, S.G.

    1994-09-20T23:59:59.000Z

    The present invention relates to a process for reducing the concentration of NO in a gas, which process comprises: (A) contacting a gas sample containing NO with a gaseous oxidizing agent to oxidize the NO to NO[sub 2]; (B) contacting the gas sample of step (A) comprising NO[sub 2] with an aqueous reagent of bisulfite/sulfite and a compound selected from urea, sulfamic acid, hydrazinium ion, hydrazoic acid, nitroaniline, sulfanilamide, sulfanilic acid, mercaptopropanoic acid, mercaptosuccinic acid, cysteine or combinations thereof at between about 0 and 100 C at a pH of between about 1 and 7 for between about 0.01 and 60 sec; and (C) optionally contacting the reaction product of step (A) with conventional chemical reagents to reduce the concentrations of the organic products of the reaction in step (B) to environmentally acceptable levels. Urea or sulfamic acid are preferred, especially sulfamic acid, and step (C) is not necessary or performed. 16 figs.

  10. Development Of Chemical Reduction And Air Stripping Processes To Remove Mercury From Wastewater

    SciTech Connect (OSTI)

    Jackson, Dennis G.; Looney, Brian B.; Craig, Robert R.; Thompson, Martha C.; Kmetz, Thomas F.

    2013-07-10T23:59:59.000Z

    This study evaluates the removal of mercury from wastewater using chemical reduction and air stripping using a full-scale treatment system at the Savannah River Site. The existing water treatment system utilizes air stripping as the unit operation to remove organic compounds from groundwater that also contains mercury (C ~ 250 ng/L). The baseline air stripping process was ineffective in removing mercury and the water exceeded a proposed limit of 51 ng/L. To test an enhancement to the existing treatment modality a continuous dose of reducing agent was injected for 6-hours at the inlet of the air stripper. This action resulted in the chemical reduction of mercury to Hg(0), a species that is removable with the existing unit operation. During the injection period a 94% decrease in concentration was observed and the effluent satisfied proposed limits. The process was optimized over a 2-day period by sequentially evaluating dose rates ranging from 0.64X to 297X stoichiometry. A minimum dose of 16X stoichiometry was necessary to initiate the reduction reaction that facilitated the mercury removal. Competing electron acceptors likely inhibited the reaction at the lower 1 doses, which prevented removal by air stripping. These results indicate that chemical reduction coupled with air stripping can effectively treat large-volumes of water to emerging part per trillion regulatory standards for mercury.

  11. Solvent and water/surfactant process for removal of bitumen from tar sands contaminated with clay

    SciTech Connect (OSTI)

    Guymon, E.P.

    1990-11-06T23:59:59.000Z

    This patent describes a process for removing bitumen from a tar sand contaminated with clay. It comprises: obtaining a tar sand consisting of bitumen and clay mixed with sand; introducing the tar sand into a stripper vessel; dissolving the bitumen with a solvent, the solvent also removing the clay from the sand into a liquid medium formed with the solvent and bitumen; removing the liquid medium from the sand; and washing the sand with water to which a nonionic surface active agent has been added to remove residual bitumen from the sand, the surfactive agent comprising a linear alcohol having carbon atoms within the range on the order of about eight to fifteen carbon atoms and ethoxylate units on the carbon atoms within the range on the order of about two to eight ethoxylate units, the surfactant being present in the water in an effective amount less than about 0.5 percent by volume.

  12. E-Print Network 3.0 - actinide burner core Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the fuel... of minor actinides whose management would be problematic. Scenario with Light Water Reactors and Fast... and difficult process. Indeed, a 1 GWe reactor, whether it is...

  13. In-tank processes for destruction of organic complexants and removal of selected radionuclides

    SciTech Connect (OSTI)

    Schulz, W.W.; Kupfer, M.J.; McKeon, M.M.

    1995-02-01T23:59:59.000Z

    This report establishes the need and technical feasibility for using in-tank pretreatment processes for destruction of organic complexants and removal of {sup 90}Sr, transuranic (TRU) elements, and {sup 99}Tc from double-shell tank (DST) liquid wastes. Neither {sup 90}Sr nor {sup 99}{Tc} have to be removed from any DST solution to obtain vitrified product containing less than the Nuclear Regulatory Commission (NRC) criteria for Class C commercial low-level waste (LLW). To meet the NRC criterion for Class C LLW, TRU elements must be removed from liquid wastes in three (possibly five) DSTs. No {sup 90}Sr will have to be removed from any solution for the total vitrified waste from both DSTs and single-shell tanks to meet a goal of <7 MCi of radionuclides and a NRC ruling for Hanford Site Incidental Waste. Guidance from ALARA principles and the TWRS Environmental Impact Statement may dictate additional removal of radionuclides from DST supernatant liquids. Scavenging processes involving precipitation of strontium phosphate and/or hydrated iron oxide effectively remove {sup 90}Sr and/or TRU elements from actual DST wastes including complexant concentrate (CC) wastes. Destruction of organic complexants is not required for these scavenging processes to reduce the {sup 90}Sr and/or TRU element concentrations of DST waste solutions to or below the NRC criteria for Class C commercial LLW. However, substantially smaller amounts of scavenging agents would be required for removal of {sup 90}Sr and TRU elements from CC waste if organic complexants were destroyed. Low concentrations of added Sr(NO{sub 3}){sub 2} and Fe(NO{sub 3}){sub 3} are desirable to minimize the volume of HLW glass.

  14. Separating the Minor Actinides Through Advances in Selective Coordination Chemistry

    SciTech Connect (OSTI)

    Lumetta, Gregg J.; Braley, Jenifer C.; Sinkov, Sergey I.; Carter, Jennifer C.

    2012-08-22T23:59:59.000Z

    This report describes work conducted at the Pacific Northwest National Laboratory (PNNL) in Fiscal Year (FY) 2012 under the auspices of the Sigma Team for Minor Actinide Separation, funded by the U.S. Department of Energy Office of Nuclear Energy. Researchers at PNNL and Argonne National Laboratory (ANL) are investigating a simplified solvent extraction system for providing a single-step process to separate the minor actinide elements from acidic high-level liquid waste (HLW), including separating the minor actinides from the lanthanide fission products.

  15. Most modern wastewater treatment systems rely on microbial processes to remove contaminants. This makes wastewater

    E-Print Network [OSTI]

    Auckland, University of

    Most modern wastewater treatment systems rely on microbial processes to remove contaminants. This makes wastewater treatment one of the largest biotechnology industries in the world. In New Zealand alone, about 1.5 billion litres of treated domestic wastewater is discharged each day

  16. Process for removal of hydrogen halides or halogens from incinerator gas

    DOE Patents [OSTI]

    Huang, H.S.; Sather, N.F.

    1987-08-21T23:59:59.000Z

    A process for reducing the amount of halogens and halogen acids in high temperature combustion gas and through their removal, the formation of halogenated organics at lower temperatures, with the reduction being carried out electrochemically by contacting the combustion gas with the negative electrode of an electrochemical cell and with the halogen and/or halogen acid being recovered at the positive electrode.

  17. Development of Acetic Acid Removal Technology for the UREX+Process

    SciTech Connect (OSTI)

    Robert M. Counce; Jack S. Watson

    2009-06-30T23:59:59.000Z

    It is imperative that acetic acid is removed from a waste stream in the UREX+process so that nitric acid can be recycled and possible interference with downstreatm steps can be avoidec. Acetic acid arises from acetohydrozamic acid (AHA), and is used to suppress plutonium in the first step of the UREX+process. Later, it is hydrolyzed into hydroxyl amine nitrate and acetic acid. Many common separation technologies were examined, and solvent extraction was determined to be the best choice under process conditions. Solvents already used in the UREX+ process were then tested to determine if they would be sufficient for the removal of acetic acid. The tributyl phosphage (TBP)-dodecane diluent, used in both UREX and NPEX, was determined to be a solvent system that gave sufficient distribution coefficients for acetic acid in addition to a high separation factor from nitric acid.

  18. The ultra-high lime with aluminum process for removing chloride from recirculating cooling water

    E-Print Network [OSTI]

    Abdel-wahab, Ahmed Ibraheem Ali

    2004-09-30T23:59:59.000Z

    THE ULTRA-HIGH LIME WITH ALUMINUM PROCESS FOR REMOVING CHLORIDE FROM RECIRCULATING COOLING WATER A Dissertation by AHMED IBRAHEEM ALI ABDEL-WAHAB Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment...-WAHAB Submitted to Texas A&M University in partial fulfillment of the requirements for the degree of DOCTOR OF PHILOSOPHY Approved as to style and content by: Bill Batchelor (Chair of Committee) Robin L. Autenrieth (Member...

  19. Precipitate hydrolysis process for the removal of organic compounds from nuclear waste slurries

    DOE Patents [OSTI]

    Doherty, J.P.; Marek, J.C.

    1987-02-25T23:59:59.000Z

    A process for removing organic compounds from a nuclear waste slurry comprising reacting a mixture of radioactive waste precipitate slurry and an acid in the presence of a catalytically effective amount of a copper(II) catalyst whereby the organic compounds in the precipitate slurry are hydrolyzed to form volatile organic compounds which are separated from the reacting mixture. The resulting waste slurry, containing less than 10 percent of the original organic compounds, is subsequently blended with high level radioactive sludge land transferred to a vitrification facility for processing into borosilicate glass for long-term storage. 2 figs., 3 tabs.

  20. Precipitate hydrolysis process for the removal of organic compounds from nuclear waste slurries

    DOE Patents [OSTI]

    Doherty, Joseph P. (Elkton, MD); Marek, James C. (Augusta, GA)

    1989-01-01T23:59:59.000Z

    A process for removing organic compounds from a nuclear waste slurry comprising reacting a mixture of radioactive waste precipitate slurry and an acid in the presence of a catalytically effective amount of a copper (II) catalyst whereby the organic compounds in the precipitate slurry are hydrolyzed to form volatile organic compounds which are separated from the reacting mixture. The resulting waste slurry, containing less than 10 percent of the orginal organic compounds, is subsequently blended with high level radioactive sludge and transferred to a virtrification facility for processing into borosilicate glass for long-term storage.

  1. Process for removing sulfur from sulfur-containing gases: high calcium fly-ash

    DOE Patents [OSTI]

    Rochelle, Gary T. (Austin, TX); Chang, John C. S. (Cary, NC)

    1991-01-01T23:59:59.000Z

    The present disclosure relates to improved processes for treating hot sulfur-containing flue gas to remove sulfur therefrom. Processes in accordance with the present invention include preparing an aqueous slurry composed of a calcium alkali source and a source of reactive silica and/or alumina, heating the slurry to above-ambient temperatures for a period of time in order to facilitate the formation of sulfur-absorbing calcium silicates or aluminates, and treating the gas with the heat-treated slurry components. Examples disclosed herein demonstrate the utility of these processes in achieving improved sulfur-absorbing capabilities. Additionally, disclosure is provided which illustrates preferred configurations for employing the present processes both as a dry sorbent injection and for use in conjunction with a spray dryer and/or bagfilter. Retrofit application to existing systems is also addressed.

  2. Recycling of cleach plant filtrates by electrodialysis removal of inorganic non-process elements.

    SciTech Connect (OSTI)

    Tsai, S. P.; Pfromm, P.; Henry, M. P.; Fracaro, A. T.; Swanstrom, C. P.; Moon, P.; Energy Systems; Inst. of Paper Science and Tech.

    2000-11-01T23:59:59.000Z

    Water use in the pulp and paper industry is very significant, and the U.S. pulp and paper industries as well as other processing industries are actively pursuing water conservation and pollution prevention by in-process recycling of water. Bleach plant effluent is a large portion of the water discharged from a typical bleached kraft pulp mill. The recycling of bleach plant effluents to the kraft recovery cycle is widely regarded as an approach to low effluent bleached kraft pulp production. The focus of this work has been on developing an electrodialysis process for recycling the acidic bleach plant effluent of bleached Kraft pulp mills. Electrodialysis is uniquely suited as a selective kidney to remove non-process elements (NPEs) from bleach plant effluent before they reach the chemical recovery cycle. Using electrodialysis for selective NPE removal can prevent the problems caused by accumulation of inorganic NPEs in the pulping cycle and recovery boiler. In this work, acidic bleach plant filtrates from three mills using different bleaching sequences based on chlorine dioxide were characterized. The analyses showed no fundamental differences in the inorganic NPE composition or other characteristics among these filtrates. The majority of total dissolved solids in the effluents were found to be inorganic NPEs. Chloride and nitrate were present at significant levels in all effluent samples. Sodium was the predominant metal ion, while calcium and magnesium were also present at considerable levels. The feasibility of using electrodialysis to selectively remove inorganic NPEs from the acidic bleach effluent was successfully demonstrated in laboratory experiments with effluents from all these three mills. Although there were some variations in these effluents, chloride and potentially harmful cations, such as potassium, calcium, and magnesium, were removed efficiently from the bleach effluents into a small-volume, concentrated purge stream. This effective removal of inorganic NPEs can enable the mills to recycle bleach effluents to reduce water consumption. The electrodialysis process also effectively retained up to 98% of the organics and can reduce the organic discharge in the mill wastewater. By using suitable commercially available electrodialysis membranes, there were no indications of rapid or irreversible membrane fouling or scale formation, even in extended laboratory scale operations up to 100 hours. Results of laboratory experiments also showed that commercially available membranes properly selected for this process would have good stability to withstand the potentially oxidative conditions of the filtrate. A pilot-scale field demonstration was also conducted at a southern mill, using the D0 filtrate from the bleach plant. During the field demonstration we found serious membrane 2 stack clogging problems, which apparently were caused by fine fibers that escaped through the 5-micron pre-filters, although such a pre-filtration method had been satisfactory in the laboratory tests. Additional R&D is recommended to address this pre-filtration or clogging issue with systems approaches integrating pre-filtration, other separation methods, and stack design. After the pre-filtration/clogging issue is overcome, laboratory development and pilot demonstration are recommended to optimize the process parameters and to evaluate the long-term process parameters. The key technical issues here include membrane lives, control and mitigation of fouling and scaling, and cleaning-in-place protocols. From the data collected in this work, a preliminary process design and economic evaluations were performed for a model mill with 1,000-ton/day pulp production that uses a bleaching sequence based on chlorine dioxide. Assuming 3 m{sup 3} acidic effluents to be treated per ton of pulp produced, the electrodialysis process would require a membrane area of about 361 m{sup 2} for this model mill. The energy consumption of the electrodialytic stack for separation is estimated to be about $160/day, and the estimated capital cost of the electrodia

  3. Mercury Reduction and Removal from High Level Waste at the Defense Waste Processing Facility - 12511

    SciTech Connect (OSTI)

    Behrouzi, Aria [Savannah River Remediation, LLC (United States); Zamecnik, Jack [Savannah River National Laboratory, Aiken, South Carolina, 29808 (United States)

    2012-07-01T23:59:59.000Z

    The Defense Waste Processing Facility processes legacy nuclear waste generated at the Savannah River Site during production of enriched uranium and plutonium required by the Cold War. The nuclear waste is first treated via a complex sequence of controlled chemical reactions and then vitrified into a borosilicate glass form and poured into stainless steel canisters. Converting the nuclear waste into borosilicate glass is a safe, effective way to reduce the volume of the waste and stabilize the radionuclides. One of the constituents in the nuclear waste is mercury, which is present because it served as a catalyst in the dissolution of uranium-aluminum alloy fuel rods. At high temperatures mercury is corrosive to off-gas equipment, this poses a major challenge to the overall vitrification process in separating mercury from the waste stream prior to feeding the high temperature melter. Mercury is currently removed during the chemical process via formic acid reduction followed by steam stripping, which allows elemental mercury to be evaporated with the water vapor generated during boiling. The vapors are then condensed and sent to a hold tank where mercury coalesces and is recovered in the tank's sump via gravity settling. Next, mercury is transferred from the tank sump to a purification cell where it is washed with water and nitric acid and removed from the facility. Throughout the chemical processing cell, compounds of mercury exist in the sludge, condensate, and off-gas; all of which present unique challenges. Mercury removal from sludge waste being fed to the DWPF melter is required to avoid exhausting it to the environment or any negative impacts to the Melter Off-Gas system. The mercury concentration must be reduced to a level of 0.8 wt% or less before being introduced to the melter. Even though this is being successfully accomplished, the material balances accounting for incoming and collected mercury are not equal. In addition, mercury has not been effectively purified and collected in the Mercury Purification Cell (MPC) since 2008. A significant cleaning campaign aims to bring the MPC back up to facility housekeeping standards. Two significant investigations are being undertaken to restore mercury collection. The SMECT mercury pump has been removed from the tank and will be functionally tested. Also, research is being conducted by the Savannah River National Laboratory to determine the effects of antifoam addition on the behavior of mercury. These path forward items will help us better understand what is occurring in the mercury collection system and ultimately lead to an improved DWPF production rate and mercury recovery rate. (authors)

  4. Process for removal of ammonia and acid gases from contaminated waters

    DOE Patents [OSTI]

    King, C.J.; Mackenzie, P.D.

    1982-09-03T23:59:59.000Z

    Contaminating basic gases, i.e., ammonia and acid gases, e.g., carbon dioxide, are removed from process waters or waste waters in a combined extraction and stripping process. Ammonia in the form of ammonium ion is extracted by an immiscible organic phase comprising a liquid cation exchange component, especially an organic phosphoric acid derivative, and preferably di-2-ethyl hexyl phosphoric acid, dissolved in an alkyl hydrocarbon, aryl hydrocarbon, higher alcohol, oxygenated hydrocarbon, halogenated hydrocarbon, and mixtures thereof. Concurrently, the acidic gaseous contaminants are stripped from the process or waste waters by stripping with stream, air, nitrogen, or the like. The liquid cation exchange component has the ammonia stripped therefrom by heating, and the component may be recycled to extract additional amounts of ammonia.

  5. Process for removal of ammonia and acid gases from contaminated waters

    DOE Patents [OSTI]

    King, C. Judson (Kensington, CA); MacKenzie, Patricia D. (Berkeley, CA)

    1985-01-01T23:59:59.000Z

    Contaminating basic gases, i.e., ammonia, and acid gases, e.g., carbon dioxide, are removed from process waters or waste waters in a combined extraction and stripping process. Ammonia in the form of ammonium ion is extracted by an immiscible organic phase comprising a liquid cation exchange component, especially an organic phosphoric acid derivative, and preferably di-2-ethyl hexyl phosphoric acid, dissolved in an alkyl hydrocarbon, aryl hydrocarbon, higher alcohol, oxygenated hydrocarbon, halogenated hydrocarbon, and mixtures thereof. Concurrently, the acidic gaseous contaminants are stripped from the process or waste waters by stripping with steam, air, nitrogen, or the like. The liquid cation exchange component has the ammonia stripped therefrom by heating, and the component may be recycled to extract additional amounts of ammonia.

  6. Process for removing thorium and recovering vanadium from titanium chlorinator waste

    DOE Patents [OSTI]

    Olsen, Richard S. (Albany, OR); Banks, John T. (Corvallis, OR)

    1996-01-01T23:59:59.000Z

    A process for removal of thorium from titanium chlorinator waste comprising: (a) leaching an anhydrous titanium chlorinator waste in water or dilute hydrochloric acid solution and filtering to separate insoluble minerals and coke fractions from soluble metal chlorides; (b) beneficiating the insoluble fractions from step (a) on shaking tables to recover recyclable or otherwise useful TiO.sub.2 minerals and coke; and (c) treating filtrate from step (a) with reagents to precipitate and remove thorium by filtration along with acid metals of Ti, Zr, Nb, and Ta by the addition of the filtrate (a), a base and a precipitant to a boiling slurry of reaction products (d); treating filtrate from step (c) with reagents to precipitate and recover an iron vanadate product by the addition of the filtrate (c), a base and an oxidizing agent to a boiling slurry of reaction products; and (e) treating filtrate from step (d) to remove any remaining cations except Na by addition of Na.sub.2 CO.sub.3 and boiling.

  7. Method for removing volatile components from a ceramic article, and related processes

    DOE Patents [OSTI]

    Klug, Frederic Joseph (Schenectady, NY); DeCarr, Sylvia Marie (Waterford, NY)

    2002-01-01T23:59:59.000Z

    A method of removing substantially all of the volatile component in a green, volatile-containing ceramic article is disclosed. The method comprises freezing the ceramic article; and then subjecting the frozen article to a vacuum for a sufficient time to freeze-dry the article. Frequently, the article is heated while being freeze-dried. Use of this method efficiently reduces the propensity for any warpage of the article. The article is often formed from a ceramic slurry in a gel-casting process. A method for fabricating a ceramic core used in investment casting is also described.

  8. Cyclic process for producing methane in a tubular reactor with effective heat removal

    DOE Patents [OSTI]

    Frost, Albert C. (Congers, NY); Yang, Chang-Lee (Spring Valley, NY)

    1986-01-01T23:59:59.000Z

    Carbon monoxide-containing gas streams are converted to methane by a cyclic, essentially two-step process in which said carbon monoxide is disproportionated to form carbon dioxide and active surface carbon deposited on the surface of a catalyst, and said carbon is reacted with steam to form product methane and by-product carbon dioxide. The exothermic heat of reaction generated in each step is effectively removed during each complete cycle so as to avoid a build up of heat from cycle-to-cycle, with particularly advantageous techniques being employed for fixed bed, tubular and fluidized bed reactor operations.

  9. Cyclic process for producing methane from carbon monoxide with heat removal

    DOE Patents [OSTI]

    Frost, Albert C. (Congers, NY); Yang, Chang-lee (Spring Valley, NY)

    1982-01-01T23:59:59.000Z

    Carbon monoxide-containing gas streams are converted to methane by a cyclic, essentially two-step process in which said carbon monoxide is disproportionated to form carbon dioxide and active surface carbon deposited on the surface of a catalyst, and said carbon is reacted with steam to form product methane and by-product carbon dioxide. The exothermic heat of reaction generated in each step is effectively removed during each complete cycle so as to avoid a build up of heat from cycle-to-cycle, with particularly advantageous techniques being employed for fixed bed, tubular and fluidized bed reactor operations.

  10. Environmental research on actinide elements

    SciTech Connect (OSTI)

    Pinder, J.E. III; Alberts, J.J.; McLeod, K.W.; Schreckhise, R.G. (eds.)

    1987-08-01T23:59:59.000Z

    The papers synthesize the results of research sponsored by DOE's Office of Health and Environmental Research on the behavior of transuranic and actinide elements in the environment. Separate abstracts have been prepared for the 21 individual papers. (ACR)

  11. A high-speed photoresist removal process using multibubble microwave plasma under a mixture of multiphase plasma environment

    SciTech Connect (OSTI)

    Ishijima, Tatsuo [Research Center for Sustainable Energy and Technology, Kanazawa University, Kakuma-machi, Kanazawa, Ishikawa 920-1192 (Japan)] [Research Center for Sustainable Energy and Technology, Kanazawa University, Kakuma-machi, Kanazawa, Ishikawa 920-1192 (Japan); Nosaka, Kohei [Graduate School of Natural Science and Technology, Kanazawa University, Kakuma-machi, Kanazawa, Ishikawa 920-1192 (Japan)] [Graduate School of Natural Science and Technology, Kanazawa University, Kakuma-machi, Kanazawa, Ishikawa 920-1192 (Japan); Tanaka, Yasunori; Uesugi, Yoshihiko [Research Center for Sustainable Energy and Technology, Kanazawa University, Kakuma-machi, Kanazawa, Ishikawa 920-1192 (Japan) [Research Center for Sustainable Energy and Technology, Kanazawa University, Kakuma-machi, Kanazawa, Ishikawa 920-1192 (Japan); Graduate School of Natural Science and Technology, Kanazawa University, Kakuma-machi, Kanazawa, Ishikawa 920-1192 (Japan); Goto, Yousuke; Horibe, Hideo [Department of Applied Chemistry, Kanazawa Institute of Technology, 3-1 Yatsukaho, Hakusan, Ishikawa 924-0838 (Japan)] [Department of Applied Chemistry, Kanazawa Institute of Technology, 3-1 Yatsukaho, Hakusan, Ishikawa 924-0838 (Japan)

    2013-09-30T23:59:59.000Z

    This paper proposes a photoresist removal process that uses multibubble microwave plasma produced in ultrapure water. A non-implanted photoresist and various kinds of ion-implanted photoresists such as B, P, and As were treated with a high ion dose of 5 × 10{sup 15} atoms/cm{sup 2} at an acceleration energy of 70 keV; this resulted in fast removal rates of more than 1 ?m/min. When the distance between multibubble microwave plasma and the photoresist film was increased by a few millimeters, the photoresist removal rates drastically decreased; this suggests that short-lived radicals such as OH affect high-speed photoresist removal.

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

    SciTech Connect (OSTI)

    Brewer, K.N.; Herbst, R.S.; Tranter, T.J. [and others

    1995-12-01T23:59:59.000Z

    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.

  13. TECHNICAL AND OPERATING SUPPORT FOR PILOT DEMONSTRATION OF MORPHYSORB ACID GAS REMOVAL PROCESS

    SciTech Connect (OSTI)

    Nagaraju Palla; Dennis Leppin

    2004-02-01T23:59:59.000Z

    Over the past 14 years, the Gas Technology Institute and jointly with Uhde since 1997 developing Morphysorb{reg_sign} a new physical solvent-based acid gas removal process. Based on extensive laboratory, bench, pilot-plant scale experiments and computer simulations, DEGT Gas Transmission Company, Canada (DEGT) has chosen the process for use at its Kwoen processing facility near Chetwynd, British Columbia, Canada as the first commercial application for the Morphysorb process. DOE co-funded the development of the Morphysorb process in various stages of development. DOE funded the production of this report to ensure that the results of the work would be readily available to potential users of the process in the United States. The Kwoen Plant is designed to process 300 MMscfd of raw natural gas at 1,080-psia pressure. The sour natural gas contains 20 to 25 percent H{sub 2}S and CO{sub 2}. The plant reduces the acid gas content by about 50% and injects the removed H{sub 2}S and CO{sub 2} into an injection well. The Kwoen plant has been operating since August 2002. Morphysorb{reg_sign} is a physical solvent-based process used for the bulk removal of CO{sub 2} and/or H{sub 2}S from natural gas and other gaseous streams. The solvent consists of N-Formyl morpholine and other morpholine derivatives. This process is particularly effective for high-pressure and high acid-gas applications and offers substantial savings in investment and operating cost compared to competitive physical solvent-based processes. GTI and DEGT first entered into an agreement in 2002 to test the Morphysorb process at their Kwoen Gas Treating Plant in northern BC. The process is operating successfully without any solvent related problems and has between DEGTC and GTI. As of December 2003, about 90 Bcf of sour gas was processed. Of this about 8 Bcf of acid gas containing mainly H{sub 2}S and CO{sub 2} was injected back into the depleted reservoir and 82 Bcf sent for further processing at DEGTC's Pine River Plant. This report discusses the operational performance at Kwoen plant during the performance test as well as the solvent performance since the plant started up. The Morphysorb performance is assessed by Duke Energy according to five metrics: acid gas pickup, recycle gas flow, total hydrocarbon loss in acid gas stream, Morphysorb solvent losses and foaming related problems. Plant data over a period of one year show that the Morphysorb solvent has performed extremely well in four out of five of these categories. The fifth metric, Morphysorb solvent loss, is being evaluated over a longer-term period in order to accurately assess it. However, the preliminary indications based on makeup solvent used to date are that solvent losses will also be within expectations. The analysis of the solvent samples indicates that the solvent is very stable and did not show any sign of degradation. The operability of the solvent is good and no foaming related problems have been encountered. According to plant operators the Morphysorb unit runs smoothly and requires no special attention.

  14. Process for removing copper in a recoverable form from solid scrap metal

    DOE Patents [OSTI]

    Hartman, Alan D. (Albany, OR); Oden, Laurance L. (Albany, OR); White, Jack C. (Albany, OR)

    1995-01-01T23:59:59.000Z

    A process for removing copper in a recoverable form from a copper/solid ferrous scrap metal mix is disclosed. The process begins by placing a copper/solid ferrous scrap metal mix into a reactor vessel. The atmosphere within the reactor vessel is purged with an inert gas or oxidizing while the reactor vessel is heated in the area of the copper/solid ferrous scrap metal mix to raise the temperature within the reactor vessel to a selected elevated temperature. Air is introduced into the reactor vessel and thereafter hydrogen chloride is introduced into the reactor vessel to obtain a desired air-hydrogen chloride mix. The air-hydrogen chloride mix is operable to form an oxidizing and chloridizing atmosphere which provides a protective oxide coating on the surface of the solid ferrous scrap metal in the mix and simultaneously oxidizes/chloridizes the copper in the mix to convert the copper to a copper monochloride gas for transport away from the solid ferrous scrap metal. After the copper is completely removed from the copper/solid ferrous scrap metal mix, the flows of air and hydrogen chloride are stopped and the copper monochloride gas is collected for conversion to a recoverable copper species.

  15. A literature review of actinide-carbonate mineral interactions

    SciTech Connect (OSTI)

    Stout, D.L. [Missouri Univ., Columbia, MO (United States). Dept. of Geological Sciences; Carroll, S.A. [Lawrence Livermore National Lab., CA (United States)

    1993-10-01T23:59:59.000Z

    Chemical retardation of actinides in groundwater systems is a potentially important mechanism for assessing the performance of the Waste Isolation Pilot Plant (WIPP), a facility intended to demonstrate safe disposal of transuranic waste. Rigorous estimation of chemical retardation during transport through the Culebra Dolomite, a water-bearing unit overlying the WIPP, requires a mechanistic understanding of chemical reactions between dissolved elements and mineral surfaces. This report represents a first step toward this goal by examining the literature for pertinent experimental studies of actinide-carbonate interactions. A summary of existing models is given, along with the types of experiments on which these models are based. Articles pertaining to research into actinide interactions with carbonate minerals are summarized. Select articles involving trace element-carbonate mineral interactions are also reviewed and may serve as templates for future research. A bibliography of related articles is included. Americium(III), and its nonradioactive analog neodymium(III), partition strongly from aqueous solutions into carbonate minerals. Recent thermodynamic, kinetic, and surface studies show that Nd is preferentially removed from solution, forming a Nd-Ca carbonate solid solution. Neptunium(V) is rapidly removed from solution by carbonates. Plutonium incorporation into carbonates is complicated by multiple oxidation states. Little research has been done on the radium(H) and thorium(IV) carbonate systems. Removal of uranyl ion from solution by calcite is limited to monolayer surface coverage.

  16. Darlington tritium removal facility and station upgrading plant dynamic process simulation

    SciTech Connect (OSTI)

    Busigin, A. [NITEK USA, Inc., 6405 NW 77 PL, Parkland, FL 33067 (United States); Williams, G. I. D.; Wong, T. C. W.; Kulczynski, D.; Reid, A. [Ontario Power Generation Nuclear, Box 4000, Bowmanville, ON L1C 3Z8 (Canada)

    2008-07-15T23:59:59.000Z

    Ontario Power Generation Nuclear (OPGN) has a 4 x 880 MWe CANDU nuclear station at its Darlington Nuclear Div. located in Bowmanville. The station has been operating a Tritium Removal Facility (TRF) and a D{sub 2}O station Upgrading Plant (SUP) since 1989. Both facilities were designed with a Distributed Control System (DCS) and programmable logic controllers (PLC) for process control. This control system was replaced with a DCS only, in 1998. A dynamic plant simulator was developed for the Darlington TRF (DTRF) and the SUP, as part of the computer control system replacement. The simulator was used to test the new software, required to eliminate the PLCs. The simulator is now used for operator training and testing of process control software changes prior to field installation. Dynamic simulation will be essential for the ITER isotope separation system, where the process is more dynamic than the relatively steady-state DTRF process. This paper describes the development and application of the DTRF and SUP dynamic simulator, its benefits, architecture, and the operational experience with the simulator. (authors)

  17. Process for the removal of acid forming gases from exhaust gases

    DOE Patents [OSTI]

    Chang, S.G.; Liu, D.K.

    1992-11-17T23:59:59.000Z

    Exhaust gases are treated to remove NO or NO[sub x] and SO[sub 2] by contacting the gases with an aqueous emulsion or suspension of yellow phosphorus preferably in a wet scrubber. The pressure is not critical, and ambient pressures are used. Hot water temperatures are best, but economics suggest about 50 C is attractive. The amount of yellow phosphorus used will vary with the composition of the exhaust gas, less than 3% for small concentrations of NO, and 10% or higher for concentrations above say 1000 ppm. Similarly, the pH will vary with the composition being treated, and it is adjusted with a suitable alkali. For mixtures of NO[sub x] and SO[sub 2], alkalis that are used for flue gas desulfurization are preferred. With this process, 100% of the by-products created are usable, and close to 100% of the NO or NO[sub x] and SO[sub 2] can be removed in an economic fashion. 9 figs.

  18. Process for the removal of acid forming gases from exhaust gases

    DOE Patents [OSTI]

    Chang, Shih-Ger (El Cerrito, CA); Liu, David K. (San Pablo, CA)

    1992-01-01T23:59:59.000Z

    Exhaust gases are treated to remove NO or NO.sub.x and SO.sub.2 by contacting the gases with an aqueous emulsion or suspension of yellow phosphorus preferably in a wet scrubber. The pressure is not critical, and ambient pressures are used. Hot water temperatures are best, but economics suggest about 50.degree. C. are attractive. The amount of yellow phosphorus used will vary with the composition of the exhaust gas, less than 3% for small concentrations of NO, and 10% or higher for concentrations above say 1000 ppm. Similarly, the pH will vary with the composition being treated, and it is adjusted with a suitable alkali. For mixtures of NO.sub.x and SO.sub.2, alkalis that are used for flue gas desulfurization are preferred. With this process, 100% of the by-products created are usable, and close to 100% of the NO or NO and SO.sub.2 can be removed in an economic fashion.

  19. Evaluation of a Combined Cyclone and Gas Filtration System for Particulate Removal in the Gasification Process

    SciTech Connect (OSTI)

    Rizzo, Jeffrey J. [Phillips66 Company, West Terre Haute, IN (United States)

    2010-04-30T23:59:59.000Z

    The Wabash gasification facility, owned and operated by sgSolutions LLC, is one of the largest single train solid fuel gasification facilities in the world capable of transforming 2,000 tons per day of petroleum coke or 2,600 tons per day of bituminous coal into synthetic gas for electrical power generation. The Wabash plant utilizes Phillips66 proprietary E-Gas (TM) Gasification Process to convert solid fuels such as petroleum coke or coal into synthetic gas that is fed to a combined cycle combustion turbine power generation facility. During plant startup in 1995, reliability issues were realized in the gas filtration portion of the gasification process. To address these issues, a slipstream test unit was constructed at the Wabash facility to test various filter designs, materials and process conditions for potential reliability improvement. The char filtration slipstream unit provided a way of testing new materials, maintenance procedures, and process changes without the risk of stopping commercial production in the facility. It also greatly reduced maintenance expenditures associated with full scale testing in the commercial plant. This char filtration slipstream unit was installed with assistance from the United States Department of Energy (built under DOE Contract No. DE-FC26-97FT34158) and began initial testing in November of 1997. It has proven to be extremely beneficial in the advancement of the E-Gas (TM) char removal technology by accurately predicting filter behavior and potential failure mechanisms that would occur in the commercial process. After completing four (4) years of testing various filter types and configurations on numerous gasification feed stocks, a decision was made to investigate the economic and reliability effects of using a particulate removal gas cyclone upstream of the current gas filtration unit. A paper study had indicated that there was a real potential to lower both installed capital and operating costs by implementing a char cyclonefiltration hybrid unit in the E-Gas (TM) gasification process. These reductions would help to keep the E-Gas (TM) technology competitive among other coal-fired power generation technologies. The Wabash combined cyclone and gas filtration slipstream test program was developed to provide design information, equipment specification and process control parameters of a hybrid cyclone and candle filter particulate removal system in the E-Gas (TM) gasification process that would provide the optimum performance and reliability for future commercial use. The test program objectives were as follows: 1. Evaluate the use of various cyclone materials of construction; 2. Establish the optimal cyclone efficiency that provides stable long term gas filter operation; 3. Determine the particle size distribution of the char separated by both the cyclone and candle filters. This will provide insight into cyclone efficiency and potential future plant design; 4. Determine the optimum filter media size requirements for the cyclone-filtration hybrid unit; 5. Determine the appropriate char transfer rates for both the cyclone and filtration portions of the hybrid unit; 6. Develop operating procedures for the cyclone-filtration hybrid unit; and, 7. Compare the installed capital cost of a scaled-up commercial cyclone-filtration hybrid unit to the current gas filtration design without a cyclone unit, such as currently exists at the Wabash facility.

  20. Process and system for removing sulfur from sulfur-containing gaseous streams

    DOE Patents [OSTI]

    Basu, Arunabha (Aurora, IL); Meyer, Howard S. (Hoffman Estates, IL); Lynn, Scott (Pleasant Hill, CA); Leppin, Dennis (Chicago, IL); Wangerow, James R. (Medinah, IL)

    2012-08-14T23:59:59.000Z

    A multi-stage UCSRP process and system for removal of sulfur from a gaseous stream in which the gaseous stream, which contains a first amount of H.sub.2S, is provided to a first stage UCSRP reactor vessel operating in an excess SO.sub.2 mode at a first amount of SO.sub.2, producing an effluent gas having a reduced amount of SO.sub.2, and in which the effluent gas is provided to a second stage UCSRP reactor vessel operating in an excess H.sub.2S mode, producing a product gas having an amount of H.sub.2S less than said first amount of H.sub.2S.

  1. Catalytic two-stage coal liquefaction process having improved nitrogen removal

    DOE Patents [OSTI]

    Comolli, Alfred G. (Yardley, PA)

    1991-01-01T23:59:59.000Z

    A process for catalytic multi-stage hydrogenation and liquefaction of coal to produce high yields of low-boiling hydrocarbon liquids containing low concentrations of nitogen compounds. First stage catalytic reaction conditions are 700.degree.-800.degree. F. temperature, 1500-3500 psig hydrogen partial pressure, with the space velocity maintained in a critical range of 10-40 lb coal/hr ft.sup.3 catalyst settled volume. The first stage catalyst has 0.3-1.2 cc/gm total pore volume with at least 25% of the pore volume in pores having diameters of 200-2000 Angstroms. Second stage reaction conditions are 760.degree.-870.degree. F. temperature with space velocity exceeding that in the first stage reactor, so as to achieve increased hydrogenation yield of low-boiling hydrocarbon liquid products having at least 75% removal of nitrogen compounds from the coal-derived liquid products.

  2. Long-term risk from actinides in the environment: Modes of mobility. 1998 annual progress report

    SciTech Connect (OSTI)

    Breshears, D.D.; Whicker, J.J. [Los Alamos National Lab., NM (US); Ibrahim, S.A.; Whicker, F.W.; Hakonson, T.E. [Colorado State Univ., Fort Collins, CO (US); Kirchner, T. [New Mexico State Univ., Las Cruces, NM (US)

    1998-06-01T23:59:59.000Z

    'The mobility of actinides in surface soils is a key issue of concern at several DOE facilities in arid and semiarid environments, including Rocky Flats, Hanford, Nevada Test Site, Idaho National Engineering Laboratory, and Los Alamos National Laboratory and the Waste Isolation Pilot Plant (WIPP). Key sources of uncertainty in assessing Pu mobility are the magnitudes of mobility resulting from three modes of transport: (1) wind erosion, (2) water erosion, and (3) vertical migration. Each of these three processes depend on numerous environmental factors and they compete with one another, particularly for actinides in very shallow soils ({approximately} 1 \\265m). The overall goal of the study is to quantify the mobility of soil actinides from all three modes. The authors study is using field measurements, laboratory experiments, and ecological modeling to address these three processes at three DOE facilities where actinide kinetics are of concern: WIPP, Rocky Flats, and Hanford. Wind erosion is being measured with suite of monitoring equipment, water erosion is being studied with rainfall simulation experiments, vertical migration is being studied in controlled laboratory experiments, and the three processes are being integrated using ecological modeling. Estimates for clean up of soil actinides for the extensive tracts of DOE land range to hundreds of billion $ in the US Without studies of these processes, unnecessary clean-up of these areas may waste billions of dollars and cause irreparable ecological damage through the soil removal. Further, the outcomes of litigation against DOE are dependent on quantifying the mobility of actinides in surface soils. This report provides a summary of work for the first year of a 3-year project; subcontracts to collaborating institutions (Colorado State University and New Mexico State University) were not in place until late December 1997, and hence this report focuses on the results of the 5 months from January through May 1998. The major result to date is a review of literature on the potential for using soil concentrations of {sup 137}Cs and {sup 241}Am as tracers for plutonium in soil. Measurements of {sup 239}Pu contamination in the environment are expensive and time consuming, requiring radiochemical analysis and alpha spectroscopy. They evaluated the literature for measurements of {sup 137}Cs and {sup 241}Am, both of which are more cost-effectively measured by gamma spectrometry, as tracers for Pu in soil. Their results indicate that: significant positive correlation exists between Pu, Cs, and Am in soils and sediments at several locations including Rocky Flats, Los Alamos, and Hanford; atmospheric transport of Pu and Cs from worldwide fallout is essentially the same; the attachment of Pu and Cs to soil particles of various size is very similar; both Pu and Cs movement in the environment correlate well with soil and sediment particle movements; a significant correlation between Pu, Cs, and Am was found in soil as a function of depth, indicating similar vertical migration behavior (most of the activity of these radionuclides is confined to the top 10--20 cm of soil at virtually all locations); most Pu and Cs are strongly absorbed onto clay and organic matter in soils and there is essentially very little leaching of Pu, Am and Cs through soil columns. Based on the above information, they believe that {sup 137}Cs and {sup 241}Am are excellent tracers for both {sup 239}Pu and soil particle transport processes in clay, mineral bearing and/or organic soils. Therefore, Cs and Am would be good tracers for the proposed water erosion and vertical migration work, at least for both Rocky Flats and Hanford. The correlation between Pu and Cs may not be as strong in sandy soil (e.g. WIPP site), however, examination of more data is needed.'

  3. Process for removing and detoxifying cadmium from scrap metal including mixed waste

    SciTech Connect (OSTI)

    Kronberg, J.W.

    1994-07-01T23:59:59.000Z

    Cadmium-bearing scrap from nuclear applications, such as neutron shielding and reactor control and safety rods, must usually be handled as mixed waste since it is radioactive and the cadmium in it is both leachable and highly toxic. Removing the cadmium from this scrap, and converting it to a nonleachable and minimally radioactive form, would greatly simplify disposal or recycling. A process now under development will do this by shredding the scrap; leaching it with reagents which selectively dissolve out the cadmium; reprecipitating the cadmium as its highly insoluble sulfide; then fusing the sulfide into a glassy matrix to bring its leachability below EPA limits before disposal. Alternatively, the cadmium may be recovered for reuse. A particular advantage of the process is that all reagents (except the glass frit) can easily be recovered and reused in a nearly closed cycle, minimizing the risk of radioactive release. The process does not harm common metals such as aluminum, iron and stainless steel, and is also applicable to non-nuclear cadmium-bearing scrap such as nickel-cadmium batteries.

  4. Process for removing halogenated aliphatic and aromatic compounds from petroleum products

    DOE Patents [OSTI]

    Googin, John M. (Oak Ridge, TN); Napier, John M. (Oak Ridge, TN); Travaglini, Michael A. (Oliver Springs, TN)

    1983-01-01T23:59:59.000Z

    A process for removing halogenated aliphatic and aromatic compounds, e.g., polychlorinated biphenyls, from petroleum products by solvent extraction. The halogenated aliphatic and aromatic compounds are extracted from a petroleum product into a polar solvent by contacting the petroleum product with the polar solvent. The polar solvent is characterized by a high solubility for the extracted halogenated aliphatic and aromatic compounds, a low solubility for the petroleum product and considerable solvent power for polyhydroxy compound. The preferred polar solvent is dimethylformamide. A miscible compound, such as, water or a polyhydroxy compound, is added to the polar extraction solvent to increase the polarity of the polar extraction solvent. The halogenated aliphatic and aromatic compounds are extracted from the highly-polarized mixture of water or polyhydroxy compound and polar extraction solvent into a low polar or nonpolar solvent by contacting the water or polyhydroxy compound-polar solvent mixture with the low polar or nonpolar solvent. The halogenated aliphatic and aromatic compounds and the low polar or nonpolar solvent are separated by physical means, e.g., vacuum evaporation. The polar and nonpolar solvents are recovered from recycling. The process can easily be designed for continuous operation. Advantages of the process include that the polar solvent and a major portion of the nonpolar solvent can be recycled, the petroleum products are reclaimable and the cost for disposing of waste containing polychlorinated biphenyls is significantly reduced.

  5. Process for removing halogenated aliphatic and aromatic compounds from petroleum products

    DOE Patents [OSTI]

    Googin, J.M.; Napier, J.M.; Travaglini, M.A.

    1983-09-20T23:59:59.000Z

    A process is described for removing halogenated aliphatic and aromatic compounds, e.g., polychlorinated biphenyls, from petroleum products by solvent extraction. The halogenated aliphatic and aromatic compounds are extracted from a petroleum product into a polar solvent by contacting the petroleum product with the polar solvent. The polar solvent is characterized by a high solubility for the extracted halogenated aliphatic and aromatic compounds, a low solubility for the petroleum product and considerable solvent power for polyhydroxy compound. The preferred polar solvent is dimethylformamide. A miscible compound, such as, water or a polyhydroxy compound, is added to the polar extraction solvent to increase the polarity of the polar extraction solvent. The halogenated aliphatic and aromatic compounds are extracted from the highly-polarized mixture of water or polyhydroxy compound and polar extraction solvent into a low polar or nonpolar solvent by contacting the water or polyhydroxy compound-polar solvent mixture with the low polar or nonpolar solvent. The halogenated aliphatic and aromatic compounds and the low polar or nonpolar solvent are separated by physical means, e.g., vacuum evaporation. The polar and nonpolar solvents are recovered from recycling. The process can easily be designed for continuous operation. Advantages of the process include that the polar solvent and a major portion of the nonpolar solvent can be recycled, the petroleum products are reclaimable and the cost for disposing of waste containing polychlorinated biphenyls is significantly reduced. 1 fig.

  6. THERMAL PERFORMANCE ANALYSIS FOR SMALL ION-EXCHANGE CESIUM REMOVAL PROCESS

    SciTech Connect (OSTI)

    Lee, S.; King, W.

    2009-12-29T23:59:59.000Z

    The In-Riser Ion Exchange program focuses on the development of in-tank systems to decontaminate high level waste (HLW) salt solutions at the Savannah River Site (SRS) and the Hanford Site. Small Column Ion Exchange (SCIX) treatment for cesium removal is a primary in-riser technology for decontamination prior to final waste immobilization in Saltstone. Through this process, radioactive cesium from the salt solution is adsorbed onto the ion exchange media which is packed within a flow-through column. Spherical Resorcinol-Formaldehyde (RF) is being considered as the ion exchange media for the application of this technology at both sites. A packed column loaded with media containing radioactive cesium generates significant heat from radiolytic decay. Under normal operating conditions, process fluid flow through the column can provide adequate heat removal from the columns. However, in the unexpected event of loss of fluid flow or fluid drainage from the column, the design must be adequate to handle the thermal load to avoid unacceptable temperature excursions. Otherwise, hot spots may develop locally which could degrade the performance of the ion-exchange media or the temperature could rise above column safety limits. Data exists which indicates that performance degradation with regard to cesium removal occurs with RF at 65C. In addition, the waste supernate solution will boil around 130C. As a result, two temperature limits have been assumed for this analysis. An additional upset scenario was considered involving the loss of the supernate solution due to inadvertent fluid drainage through the column boundary. In this case, the column containing the loaded media could be completely dry. This event is expected to result in high temperatures that could damage the column or cause the RF sorbent material to undergo undesired physical changes. One objective of these calculations is to determine the range of temperatures that should be evaluated during testing with the RF media. Although, the safety temperature limit is based on the salt solution boiling point which does not apply in the air-filled case (because there is no liquid), this same limit (130C) is used as a measure for the evaluation of this condition as well. The primary objective of the present work is to develop models to simulate the thermal performance of the RF column design when the media is fully loaded with radioactive cesium and the central cooling tube is excluded. Previous analysis led to the consideration of this design simplification for RF, since the baseline column design with center cooling was developed assuming that CST media would be used for cesium removal which has a higher volumetric heat load. Temperature distributions and maximum temperatures across the column during SCIX process operations and upset conditions were conducted with a focus on SCIX implementation at Hanford. However, a feed composition and cesium loading were assumed which were known to be considerably higher than would typically be observed at Hanford. In order to evaluate the impact of this potentially highly conservative assumption, fractionally-reduced loading cases were also considered. A computational modeling approach was taken to include conservative, bounding estimates for key parameters so that the results would provide the maximum temperatures achievable under the design configurations.

  7. ACTINIDE-SPECIFIC SEQUESTERING AGENTS AND DECONTAMINATION APPLICATIONS

    SciTech Connect (OSTI)

    Smith, William L.; Raymond, Kenneth N.

    1980-07-01T23:59:59.000Z

    We have briefly reviewed the biological hazards associated with the actinide elements. The most abundant transuranium element produced by both industrial nuclear power plants and nuclear weapons programs is plutonium. It is also potentially the most toxic - particularly due to its long-term hazard as a carcinogen if it is introduced into the body. This toxicity is due in large part to the chemical and biochemical similarities of Pu(IV) and Fe(III). Thus in mammals plutonium is transported and stored by the transport and storage systems for iron. This results in the concentration and long-term retention of an alpha-emitting radionuclide ({sup 239}Pu) at sites such as the bone marrow where cell division occurs at a high rate. The earliest attempts at removal of actinide contamination by chelation therapy were essentially heuristic in that sequestering agents known to be effective at binding other elements were tried with plutonium. The research described here is intended to be a rational approach that begins with the observation that since Fe(III) and Pu(IV) are so similar, and since microbes produce agents called siderophores that are extremely effective and selective sequestering agents for Fe(III), the construction of similar chelating agents for the actinides should be possible using the same chelating groups found in the siderophores. The incorporation of four such groups (primarily catechol and hydroxamic acid) results in multidentate chelating agents that can completely encapsulate the central actinide(IV) ion and achieve the eight-coordinate environment most favored by such ions. The continuing development and improvement of such sequestering agents has produced compounds which remove significant amounts of plutonium deposited in bone and which remove a greater fraction of the total body burden than any other chelation therapy developed to date.

  8. Actinide Thermodynamics at Elevated Temperatures

    SciTech Connect (OSTI)

    Friese, Judah I.; Rao, Linfeng; Xia, Yuanxian; Bachelor, Paula P.; Tian, Guoxin

    2007-11-16T23:59:59.000Z

    The postclosure chemical environment in the proposed Yucca Mountain repository is expected to experience elevated temperatures. Predicting migration of actinides is possible if sufficient, reliable thermodynamic data on hydrolysis and complexation are available for these temperatures. Data are scarce and scattered for 25 degrees C, and nonexistent for elevated temperatures. This collaborative project between LBNL and PNNL collects thermodynamic data at elevated temperatures on actinide complexes with inorganic ligands that may be present in Yucca Mountain. The ligands include hydroxide, fluoride, sulfate, phosphate and carbonate. Thermodynamic parameters of complexation, including stability constants, enthalpy, entropy and heat capacity of complexation, are measured with a variety of techniques including solvent extraction, potentiometry, spectrophotometry and calorimetry

  9. Process for the removal of acid forming gases from exhaust gases and production of phosphoric acid

    DOE Patents [OSTI]

    Chang, Shih-Ger (El Cerrito, CA); Liu, David K. (San Pablo, CA)

    1992-01-01T23:59:59.000Z

    Exhaust gases are treated to remove NO or NO.sub.x and SO.sub.2 by contacting the gases with an aqueous emulsion or suspension of yellow phosphorous preferably in a wet scrubber. The addition of yellow phosphorous in the system induces the production of O.sub.3 which subsequently oxidizes NO to NO.sub.2. The resulting NO.sub.2 dissolves readily and can be reduced to form ammonium ions by dissolved SO.sub.2 under appropriate conditions. In a 20 acfm system, yellow phosphorous is oxidized to yield P.sub.2 O.sub.5 which picks up water to form H.sub.3 PO.sub.4 mists and can be collected as a valuable product. The pressure is not critical, and ambient pressures are used. Hot water temperatures are best, but economics suggest about 50.degree. C. The amount of yellow phosphorus used will vary with the composition of the exhaust gas, less than 3% for small concentrations of NO, and 10% or higher for concentrations above say 1000 ppm. Similarly, the pH will vary with the composition being treated, and it is adjusted with a suitable alkali. For mixtures of NO.sub.x and SO.sub.2, alkalis that are used for flue gas desulfurization are preferred. With this process, better than 90% of SO.sub.2 and NO in simulated flue gas can be removed. Stoichiometric ratios (P/NO) ranging between 0.6 and 1.5 were obtained.

  10. Modeling Ion-Exchange Processing With Spherical Resins For Cesium Removal

    SciTech Connect (OSTI)

    Hang, T.; Nash, C. A.; Aleman, S. E.

    2012-09-19T23:59:59.000Z

    The spherical Resorcinol-Formaldehyde and hypothetical spherical SuperLig(r) 644 ion-exchange resins are evaluated for cesium removal from radioactive waste solutions. Modeling results show that spherical SuperLig(r) 644 reduces column cycling by 50% for high-potassium solutions. Spherical Resorcinol Formaldehyde performs equally well for the lowest-potassium wastes. Less cycling reduces nitric acid usage during resin elution and sodium addition during resin regeneration, therefore, significantly decreasing life-cycle operational costs. A model assessment of the mechanism behind ''cesium bleed'' is also conducted. When a resin bed is eluted, a relatively small amount of cesium remains within resin particles. Cesium can bleed into otherwise decontaminated product in the next loading cycle. The bleed mechanism is shown to be fully isotherm-controlled vs. mass transfer controlled. Knowledge of residual post-elution cesium level and resin isotherm can be utilized to predict rate of cesium bleed in a mostly non-loaded column. Overall, this work demonstrates the versatility of the ion-exchange modeling to study the effects of resin characteristics on processing cycles, rates, and cold chemical consumption. This evaluation justifies further development of a spherical form of the SL644 resin.

  11. Conjugates of Actinide Chelator-Magnetic Nanoparticles for Used Fuel Separation Technology

    SciTech Connect (OSTI)

    You Qiang; Andrzej Paszczynski; Linfeng Rao

    2011-10-30T23:59:59.000Z

    The actinide separation method using magnetic nanoparticles (MNPs) functionalized with actinide specific chelators utilizes the separation capability of ligand and the ease of magnetic separation. This separation method eliminated the need of large quantity organic solutions used in the liquid-liquid extraction process. The MNPs could also be recycled for repeated separation, thus this separation method greatly reduces the generation of secondary waste compared to traditional liquid extraction technology. The high diffusivity of MNPs and the large surface area also facilitate high efficiency of actinide sorption by the ligands. This method could help in solving the nuclear waste remediation problem.

  12. CHARACTERIZATION OF ACTINIDES IN SIMULATED ALKALINE TANK WASTE SLUDGES AND LEACHATES

    SciTech Connect (OSTI)

    Nash, Kenneth L.

    2008-11-20T23:59:59.000Z

    In this project, both the fundamental chemistry of actinides in alkaline solutions (relevant to those present in Hanford-style waste storage tanks), and their dissolution from sludge simulants (and interactions with supernatants) have been investigated under representative sludge leaching procedures. The leaching protocols were designed to go beyond conventional alkaline sludge leaching limits, including the application of acidic leachants, oxidants and complexing agents. The simulant leaching studies confirm in most cases the basic premise that actinides will remain in the sludge during leaching with 2-3 M NaOH caustic leach solutions. However, they also confirm significant chances for increased mobility of actinides under oxidative leaching conditions. Thermodynamic data generated improves the general level of experiemental information available to predict actinide speciation in leach solutions. Additional information indicates that improved Al removal can be achieved with even dilute acid leaching and that acidic Al(NO3)3 solutions can be decontaminated of co-mobilized actinides using conventional separations methods. Both complexing agents and acidic leaching solutions have significant potential to improve the effectiveness of conventional alkaline leaching protocols. The prime objective of this program was to provide adequate insight into actinide behavior under these conditions to enable prudent decision making as tank waste treatment protocols develop.

  13. Processes for Removal and Immobilization of 14C, 129I, and 85Kr

    SciTech Connect (OSTI)

    Strachan, Denis M.; Bryan, Samuel A.; Henager, Charles H.; Levitskaia, Tatiana G.; Matyas, Josef; Thallapally, Praveen K.; Scheele, Randall D.; Weber, William J.; Zheng, Feng

    2009-10-05T23:59:59.000Z

    This is a white paper covering the results of a literature search and preliminary experiments on materials and methods to remove and immobilize gaseous radionuclided that come from the reprocessing of spent nuclear fuel.

  14. The removal of mercury from solid mixed waste using chemical leaching processes

    SciTech Connect (OSTI)

    Gates, D.D.; Chao, K.K.; Cameron, P.A.

    1995-07-01T23:59:59.000Z

    The focus of this research was to evaluate chemical leaching as a technique to treat soils, sediments, and glass contaminated with either elemental mercury or a combination of several mercury species. Potassium iodide/iodine solutions were investigated as chemical leaching agents for contaminated soils and sediments. Clean, synthetic soil material and surrogate storm sewer sediments contaminated with mercury were treated with KI/I{sub 2} solutions. It was observed that these leaching solutions could reduce the mercury concentration in soil and sediments by 99.8%. Evaluation of selected posttreatment sediment samples revealed that leachable mercury levels in the treated solids exceeded RCRA requirements. The results of these studies suggest that KI/I{sub 2} leaching is a treatment process that can be used to remove large quantities of mercury from contaminated soils and sediments and may be the only treatment required if treatment goals are established on Hg residual concentrations in solid matrices. Fluorescent bulbs were used to simulate mercury contaminated glass mixed waste. To achieve mercury contamination levels similar to those found in larger bulbs such as those used in DOE facilities a small amount of Hg was added to the crushed bulbs. The most effective agents for leaching mercury from the crushed fluorescent bulbs were KI/I{sub 2}, NaOCl, and NaBr + acid. Radionuclide surrogates were added to both the EPA synthetic soil material and the crushed fluorescent bulbs to determine the fate of radionuclides following chemical leaching with the leaching agents determined to be the most promising. These experiments revealed that although over 98% of the dosed mercury solubilized and was found in the leaching solution, no Cerium was measured in the posttreatment leaching solution. This finding suggest that Uranium, for which Ce was used as a surrogate, would not solubilize during leaching of mercury contaminated soil or glass.

  15. NOVEL PROCESS FOR REMOVAL AND RECOVERY OF VAPOR-PHASE MERCURY

    SciTech Connect (OSTI)

    Craig S. Turchi

    2000-09-29T23:59:59.000Z

    The goal of this project is to investigate the use of a regenerable sorbent for removing and recovering mercury from the flue gas of coal-fired power plants. The process is based on the sorption of mercury by noble metals and the thermal regeneration of the sorbent, recovering the desorbed mercury in a small volume for recycling or disposal. The project was carried out in two phases, covering five years. Phase I ran from September 1995 through September 1997 and involved development and testing of sorbent materials and field tests at a pilot coal-combustor. Phase II began in January 1998 and ended September 2000. Phase II culminated with pilot-scale testing at a coal-fired power plant. The use of regenerable sorbents holds the promise of capturing mercury in a small volume, suitable for either stable disposal or recycling. Unlike single-use injected sorbents such as activated carbon, there is no impact on the quality of the fly ash. During Phase II, tests were run with a 20-acfm pilot unit on coal-combustion flue gas at a 100 lb/hr pilot combustor and a utility boiler for four months and six months respectively. These studies, and subsequent laboratory comparisons, indicated that the sorbent capacity and life were detrimentally affected by the flue gas constituents. Sorbent capacity dropped by a factor of 20 to 35 during operations in flue gas versus air. Thus, a sorbent designed to last 24 hours between recycling lasted less than one hour. The effect resulted from an interaction between SO{sub 2} and either NO{sub 2} or HCl. When SO{sub 2} was combined with either of these two gases, total breakthrough was seen within one hour in flue gas. This behavior is similar to that reported by others with carbon adsorbents (Miller et al., 1998).

  16. CONTAMINATED PROCESS EQUIPMENT REMOVAL FOR THE D&D OF THE 232-Z CONTAMINATED WASTE RECOVERY PROCESS FACILITY AT THE PLUTONIUM FINISHING PLANT (PFP)

    SciTech Connect (OSTI)

    HOPKINS, A.M.; MINETTE, M.J.; KLOS, D.B.

    2007-01-25T23:59:59.000Z

    This paper describes the unique challenges encountered and subsequent resolutions to accomplish the deactivation and decontamination of a plutonium ash contaminated building. The 232-Z Contaminated Waste Recovery Process Facility at the Plutonium Finishing Plant was used to recover plutonium from process wastes such as rags, gloves, containers and other items by incinerating the items and dissolving the resulting ash. The incineration process resulted in a light-weight plutonium ash residue that was highly mobile in air. This light-weight ash coated the incinerator's process equipment, which included gloveboxes, blowers, filters, furnaces, ducts, and filter boxes. Significant airborne contamination (over 1 million derived air concentration hours [DAC]) was found in the scrubber cell of the facility. Over 1300 grams of plutonium held up in the process equipment and attached to the walls had to be removed, packaged and disposed. This ash had to be removed before demolition of the building could take place.

  17. H[sub 2]S-removal and sulfur-recovery processes using metal salts

    SciTech Connect (OSTI)

    Lynn, S.; Cairns, E.J.

    1992-01-01T23:59:59.000Z

    Scrubbing a sour gas stream with a solution of copper sulfate allows the clean-up temperature to be increased from ambient to the adiabatic saturation temperature of the gas. The copper ion in solution reacts with the H[sub 2]S to produce insoluble CuS. The choice of copper sulfate was set by the very low solubility of CuS and the very rapid kinetics of the Cus formation. Since the copper sulfate solutions used are acidic, CO[sub 2] will not be co-absorbed. In a subsequent step the solid CuS is oxidized by a solution of ferric sulfate. The copper sulfate is regenerated, and elemental sulfur is formed together with ferrous sulfate. The ferrous sulfate is reoxidized to ferric sulfate using air. Since the copper sulfate and ferric solutions are regenerated, the overall reaction in this process is the oxidation of hydrogen sulfide with oxygen to form sulfur. The use of copper sulfate has the further advantage that the presence of sulfuric acid, even as concentrated as 1 molar, does not inhibit the sorption of H[sub 2]S. Furthermore, the absorption reaction remains quite favorable thermodynamically over the temperature range of interest. Because the reaction goes to completion, only a single theoretical stage is required for complete H[sub 2]S removal and cocurrent gas/liquid contacting may be employed. The formation of solids precludes the use of a packed column for the contacting device. However, a venturi scrubber would be expected to perform satisfactorily. The kinetics of the oxidation of metal sulfides, in particular zinc and copper sulfide, is reported in the literature to be slow at near-ambient temperatures. The proposed process conditions for the oxidation step are different from those reported in the literature, most notably the higher temperature. The kinetics of the reaction must be studied at high temperatures and corresponding pressures. An important goal is to obtain sulfur of high purity, which is a salable product.

  18. H{sub 2}S-removal and sulfur-recovery processes using metal salts

    SciTech Connect (OSTI)

    Lynn, S.; Cairns, E.J.

    1992-11-01T23:59:59.000Z

    Scrubbing a sour gas stream with a solution of copper sulfate allows the clean-up temperature to be increased from ambient to the adiabatic saturation temperature of the gas. The copper ion in solution reacts with the H{sub 2}S to produce insoluble CuS. The choice of copper sulfate was set by the very low solubility of CuS and the very rapid kinetics of the Cus formation. Since the copper sulfate solutions used are acidic, CO{sub 2} will not be co-absorbed. In a subsequent step the solid CuS is oxidized by a solution of ferric sulfate. The copper sulfate is regenerated, and elemental sulfur is formed together with ferrous sulfate. The ferrous sulfate is reoxidized to ferric sulfate using air. Since the copper sulfate and ferric solutions are regenerated, the overall reaction in this process is the oxidation of hydrogen sulfide with oxygen to form sulfur. The use of copper sulfate has the further advantage that the presence of sulfuric acid, even as concentrated as 1 molar, does not inhibit the sorption of H{sub 2}S. Furthermore, the absorption reaction remains quite favorable thermodynamically over the temperature range of interest. Because the reaction goes to completion, only a single theoretical stage is required for complete H{sub 2}S removal and cocurrent gas/liquid contacting may be employed. The formation of solids precludes the use of a packed column for the contacting device. However, a venturi scrubber would be expected to perform satisfactorily. The kinetics of the oxidation of metal sulfides, in particular zinc and copper sulfide, is reported in the literature to be slow at near-ambient temperatures. The proposed process conditions for the oxidation step are different from those reported in the literature, most notably the higher temperature. The kinetics of the reaction must be studied at high temperatures and corresponding pressures. An important goal is to obtain sulfur of high purity, which is a salable product.

  19. Separations and Actinide Science -- 2005 Roadmap

    SciTech Connect (OSTI)

    Not Available

    2005-09-01T23:59:59.000Z

    The Separations and Actinide Science Roadmap presents a vision to establish a separations and actinide science research (SASR) base composed of people, facilities, and collaborations and provides new and innovative nuclear fuel cycle solutions to nuclear technology issues that preclude nuclear proliferation. This enabling science base will play a key role in ensuring that Idaho National Laboratory (INL) achieves its long-term vision of revitalizing nuclear energy by providing needed technologies to ensure our nation's energy sustainability and security. To that end, this roadmap suggests a 10-year journey to build a strong SASR technical capability with a clear mission to support nuclear technology development. If nuclear technology is to be used to satisfy the expected growth in U.S. electrical energy demand, the once-through fuel cycle currently in use should be reconsidered. Although the once-through fuel cycle is cost-effective and uranium is inexpensive, a once-through fuel cycle requires long-term disposal to protect the environment and public from long-lived radioactive species. The lack of a current disposal option (i.e., a licensed repository) has resulted in accumulation of more than 50,000 metric tons of spent nuclear fuel. The process required to transition the current once-through fuel cycle to full-recycle will require considerable time and significant technical advancement. INL's extensive expertise in aqueous separations will be used to develop advanced separations processes. Computational chemistry will be expanded to support development of future processing options. In the intermediate stage of this transition, reprocessing options will be deployed, waste forms with higher loading densities and greater stability will be developed, and transmutation of long-lived fission products will be explored. SASR will support these activities using its actinide science and aqueous separations expertise. In the final stage, full recycle will be enabled by advanced reactors and reprocessing methods based on pyrochemical methods and by using different fuel cycles that do not readily produce plutonium. SASR will facilitate the deployment of advanced pyrochemical separation technology and support development of reprocessing of thorium-based reactor fuels.

  20. Ceramic composition for immobilization of actinides

    DOE Patents [OSTI]

    Ebbinghaus, Bartley B. (Livermore, CA); Van Konynenburg, Richard A. (Livermore, CA); Vance, Eric R. (Kirrawee, AU); Stewart, Martin W. (Barden Ridge, AU); Jostsons, Adam (Eastwood, AU); Allender, Jeffrey S. (North Augusta, SC); Rankin, David Thomas (Aiken, SC)

    2000-01-01T23:59:59.000Z

    Disclosed is a ceramic composition for the immobilization of actinides, particularly uranium and plutonium. The ceramic is a titanate material comprising pyrochlore, brannerite and rutile.

  1. Removal of Radiocesium from Food by Processing: Data Collected after the Fukushima Daiichi Nuclear Power Plant Accident - 13167

    SciTech Connect (OSTI)

    Uchida, Shigeo; Tagami, Keiko [Office of Biospheric Assessment for Waste Disposal, National Institute of Radiological Sciences, Anagawa 4-9-1, Inage-ku, Chiba 263-8555 (Japan)] [Office of Biospheric Assessment for Waste Disposal, National Institute of Radiological Sciences, Anagawa 4-9-1, Inage-ku, Chiba 263-8555 (Japan)

    2013-07-01T23:59:59.000Z

    Removal of radiocesium from food by processing is of great concern following the accident of TEPCO's Fukushima Daiichi Nuclear Power Plant accident. Foods in markets are monitored and recent monitoring results have shown that almost all food materials were under the standard limit concentration levels for radiocesium (Cs-134+137), that is, 100 Bq kg{sup -1} in raw foods, 50 Bq kg{sup -1} in baby foods, and 10 Bq kg{sup -1} in drinking water; those food materials above the limit cannot be sold. However, one of the most frequently asked questions from the public is how much radiocesium in food would be removed by processing. Hence, information about radioactivity removal by processing of food crops native to Japan is actively sought by consumers. In this study, the food processing retention factor, F{sub r}, which is expressed as total activity in processed food divided by total activity in raw food, is reported for various types of corps. For white rice at a typical polishing yield of 90-92% from brown rice, the F{sub r} value range was 0.42-0.47. For leafy vegetable (indirect contamination), the average F{sub r} values were 0.92 (range: 0.27-1.2) after washing and 0.55 (range: 0.22-0.93) after washing and boiling. The data for some fruits are also reported. (authors)

  2. TAILORING INORGANIC SORBENTS FOR SRS STRONTIUM AND ACTINIDE SEPARATIONS: OPTIMIZED MONOSODIUM TITANATE PHASE II FINAL REPORT

    SciTech Connect (OSTI)

    Hobbs, D; Thomas Peters, T; Michael Poirier, M; Mark Barnes, M; Major Thompson, M; Samuel Fink, S

    2007-06-29T23:59:59.000Z

    This document provides a final report of Phase II testing activities for the development of a modified monosodium titanate (MST) that exhibits improved strontium and actinide removal characteristics compared to the baseline MST material. The activities included determining the key synthesis conditions for preparation of the modified MST, preparation of the modified MST at a larger scale by a commercial vendor, demonstration of the strontium and actinide removal characteristics with actual tank waste supernate and measurement of filtration characteristics. Key findings and conclusions include the following. Testing evaluated three synthetic methods and eleven process parameters for the optimum synthesis conditions for the preparation on an improved form of MST. We selected the post synthesis method (Method 3) for continued development based on overall sorbate removal performance. We successfully prepared three batches of the modified MST using Method 3 procedure at a 25-gram scale. The laboratory prepared modified MST exhibited increased sorption kinetics with simulated and actual waste solutions and similar filtration characteristics to the baseline MST. Characterization of the modified MST indicated that the post synthesis treatment did not significantly alter the particle size distribution, but did significantly increase the surface area and porosity compared to the original MST. Testing indicated that the modified MST exhibits reduced affinity for uranium compared to the baseline MST, reducing risk of fissile loading. Shelf-life testing indicated no change in strontium and actinide performance removal after storing the modified MST for 12-months at ambient laboratory temperature. The material releases oxygen during the synthesis and continues to offgas after the synthesis at a rapidly diminishing rate until below a measurable rate after 4 months. Optima Chemical Group LLC prepared a 15-kilogram batch of the modified MST using the post synthesis procedure (Method 3). Performance testing with simulated and actual waste solutions indicated that the material performs as well as or better than batches of modified MST prepared at the laboratory-scale. Particle size data of the vendor-prepared modified MST indicates a broader distribution centered at a larger particle size and microscopy shows more irregular particle morphology compared to the baseline MST and laboratory prepared modified MST. Stirred-cell (i.e., dead-end) filter testing revealed similar filtration rates relative to the baseline MST for both the laboratory and vendor-prepared modified MST materials. Crossflow filtration testing indicated that with MST-only slurries, the baseline MST produced between 30-100% higher flux than the vendor-prepared modified MST at lower solids loadings and comparable flux at higher solids loadings. With sludge-MST slurries, the modified MST produced 1.5-2.2 times higher flux than the baseline MST at all solids loadings. Based on these findings we conclude that the modified MST represents a much improved sorbent for the separation of strontium and actinides from alkaline waste solutions and recommend continued development of the material as a replacement for the baseline MST for waste treatment facilities at the Savannah River Site.

  3. ACTINIDES-1981. ABSTRACTS

    E-Print Network [OSTI]

    Authors, Various

    2010-01-01T23:59:59.000Z

    of binary and ternary metal, carbide and nitride systemsmetals are obtained by a van Arkel process starting frow the carbides.metal sublatti­ ce must be considered as "frozen". 1) Carbides.

  4. PROCESSING ALTERNATIVES FOR DESTRUCTION OF TETRAPHENYLBORATE

    SciTech Connect (OSTI)

    Lambert, D; Thomas Peters, T; Samuel Fink, S

    2007-02-27T23:59:59.000Z

    Two processes were chosen in the 1980's at the Savannah River Site (SRS) to decontaminate the soluble High Level Waste (HLW). The In Tank Precipitation (ITP) process (1,2) was developed at SRS for the removal of radioactive cesium and actinides from the soluble HLW. Sodium tetraphenylborate was added to the waste to precipitate cesium and monosodium titanate (MST) was added to adsorb actinides, primarily uranium and plutonium. Two products of this process were a low activity waste stream and a concentrated organic stream containing cesium tetraphenylborate and actinides adsorbed on monosodium titanate (MST). A copper catalyzed acid hydrolysis process was built to process (3, 4) the Tank 48H cesium tetraphenylborate waste in the SRS's Defense Waste Processing Facility (DWPF). Operation of the DWPF would have resulted in the production of benzene for incineration in SRS's Consolidated Incineration Facility. This process was abandoned together with the ITP process in 1998 due to high benzene in ITP caused by decomposition of excess sodium tetraphenylborate. Processing in ITP resulted in the production of approximately 1.0 million liters of HLW. SRS has chosen a solvent extraction process combined with adsorption of the actinides to decontaminate the soluble HLW stream (5). However, the waste in Tank 48H is incompatible with existing waste processing facilities. As a result, a processing facility is needed to disposition the HLW in Tank 48H. This paper will describe the process for searching for processing options by SRS task teams for the disposition of the waste in Tank 48H. In addition, attempts to develop a caustic hydrolysis process for in tank destruction of tetraphenylborate will be presented. Lastly, the development of both a caustic and acidic copper catalyzed peroxide oxidation process will be discussed.

  5. Precipitation process for the removal of technetium values from nuclear waste solutions

    DOE Patents [OSTI]

    Walker, D.D.; Ebra, M.A.

    1985-11-21T23:59:59.000Z

    High efficiency removal of techetium values from a nuclear waste stream is achieved by addition to the waste stream of a precipitant contributing tetraphenylphosphonium cation, such that a substantial portion of the technetium values are precipitated as an insoluble pertechnetate salt.

  6. Process for the conversion of and aqueous biomass hydrolyzate into fuels or chemicals by the selective removal of fermentation inhibitors

    DOE Patents [OSTI]

    Hames, Bonnie R. (Westminster, CO); Sluiter, Amie D. (Arvada, CO); Hayward, Tammy K. (Broomfield, CO); Nagle, Nicholas J. (Broomfield, CO)

    2004-05-18T23:59:59.000Z

    A process of making a fuel or chemical from a biomass hydrolyzate is provided which comprises the steps of providing a biomass hydrolyzate, adjusting the pH of the hydrolyzate, contacting a metal oxide having an affinity for guaiacyl or syringyl functional groups, or both and the hydrolyzate for a time sufficient to form an adsorption complex; removing the complex wherein a sugar fraction is provided, and converting the sugar fraction to fuels or chemicals using a microorganism.

  7. Evaluation of improved techniques for the removal of fission products from process wastewater and groundwater: FY 1996 status

    SciTech Connect (OSTI)

    Bostick, D.T. [Oak Ridge National Lab., TN (United States); Guo, B. [Oak Ridge Research Inst., TN (United States)

    1997-07-01T23:59:59.000Z

    This report describes laboratory results acquired in the course of evaluating new sorbents for the treatment of radiologically contaminated groundwater and process wastewater. During FY 1996, the evaluation of resorcinol-formaldehyde (R-F) resin for the removal of cesium and strontium from wastewaters was completed. Additionally, strontium sorption on sodium nonatitanate powder was characterized in a series of multicomponent batch studies. Both of these materials were evaluated in reference to a baseline sorbent, natural chabazite zeolite.

  8. MODELING AN ION EXCHANGE PROCESS FOR CESIUM REMOVAL FROM ALKALINE RADIOACTIVE WASTE SOLUTIONS

    SciTech Connect (OSTI)

    Smith, F; Luther Hamm, L; Sebastian Aleman, S; Johnston Michael, J

    2008-08-26T23:59:59.000Z

    The performance of spherical Resorcinol-Formaldehyde ion-exchange resin for the removal of cesium from alkaline radioactive waste solutions has been investigated through computer modeling. Cesium adsorption isotherms were obtained by fitting experimental data using a thermodynamic framework. Results show that ion-exchange is an efficient method for cesium removal from highly alkaline radioactive waste solutions. On average, two 1300 liter columns operating in series are able to treat 690,000 liters of waste with an initial cesium concentration of 0.09 mM in 11 days achieving a decontamination factor of over 50,000. The study also tested the sensitivity of ion-exchange column performance to variations in flow rate, temperature and column dimensions. Modeling results can be used to optimize design of the ion exchange system.

  9. Novel Sorbent-Based Process for High Temperature Trace Metal Removal

    SciTech Connect (OSTI)

    Gokhan Alptekin

    2008-09-30T23:59:59.000Z

    The objective of this project was to demonstrate the efficacy of a novel sorbent can effectively remove trace metal contaminants (Hg, As, Se and Cd) from actual coal-derived synthesis gas streams at high temperature (above the dew point of the gas). The performance of TDA's sorbent has been evaluated in several field demonstrations using synthesis gas generated by laboratory and pilot-scale coal gasifiers in a state-of-the-art test skid that houses the absorbent and all auxiliary equipment for monitoring and data logging of critical operating parameters. The test skid was originally designed to treat 10,000 SCFH gas at 250 psig and 350 C, however, because of the limited gas handling capabilities of the test sites, the capacity was downsized to 500 SCFH gas flow. As part of the test program, we carried out four demonstrations at two different sites using the synthesis gas generated by the gasification of various lignites and a bituminous coal. Two of these tests were conducted at the Power Systems Demonstration Facility (PSDF) in Wilsonville, Alabama; a Falkirk (North Dakota) lignite and a high sodium lignite (the PSDF operator Southern Company did not disclose the source of this lignite) were used as the feedstock. We also carried out two other demonstrations in collaboration with the University of North Dakota Energy Environmental Research Center (UNDEERC) using synthesis gas slipstreams generated by the gasification of Sufco (Utah) bituminous coal and Oak Hills (Texas) lignite. In the PSDF tests, we showed successful operation of the test system at the conditions of interest and showed the efficacy of sorbent in removing the mercury from synthesis gas. In Test Campaign No.1, TDA sorbent reduced Hg concentration of the synthesis gas to less than 5 {micro}g/m{sup 3} and achieved over 99% Hg removal efficiency for the entire test duration. Unfortunately, due to the relatively low concentration of the trace metals in the lignite feed and as a result of the intermittent operation of the PSDF gasifier (due to the difficulties in the handling of the low quality lignite), only a small fraction of the sorbent capacity was utilized (we measured a mercury capacity of 3.27 mg/kg, which is only a fraction of the 680 mg/kg Hg capacity measured for the same sorbent used at our bench-scale evaluations at TDA). Post reaction examination of the sorbent by chemical analysis also indicated some removal As and Se (we did not detect any significant amounts of Cd in the synthesis gas or over the sorbent). The tests at UNDEERC was more successful and showed clearly that the TDA sorbent can effectively remove Hg and other trace metals (As and Se) at high temperature. The on-line gas measurements carried out by TDA and UNDEERC separately showed that TDA sorbent can achieve greater than 95% Hg removal efficiency at 260 C ({approx}200g sorbent treated more than 15,000 SCF synthesis gas). Chemical analysis conducted following the tests also showed modest amounts of As and Se accumulation in the sorbent bed (the test durations were still short to show higher capacities to these contaminants). We also evaluated the stability of the sorbent and the fate of mercury (the most volatile and unstable of the trace metal compounds). The Synthetic Ground Water Leaching Procedure Test carried out by an independent environmental laboratory showed that the mercury will remain on the sorbent once the sorbent is disposed. Based on a preliminary engineering and cost analysis, TDA estimated the cost of mercury removal from coal-derived synthesis gas as $2,995/lb (this analysis assumes that this cost also includes the cost of removal of all other trace metal contaminants). The projected cost will result in a small increase (less than 1%) in the cost of energy.

  10. Removal of Chloride from Wastewater by Advanced Softening Process Using Electrochemically Generated Aluminum Hydroxide

    E-Print Network [OSTI]

    Mustafa, Syed Faisal

    2014-07-23T23:59:59.000Z

    produced mass of aluminum and theoretical mass as predicted by Faraday’s law vs time during electrolysis of 30 mM NaCl electrolyte solution. ........................................................................................... 35 Figure 8 Change... of pH versus time during electrolysis performed at different current values of 30mM NaCl electrolyte solution. ...................................................... 36 Figure 9 Removal of chloride during advanced softening experiment performed after...

  11. PREPARATION AND SPECTROSCOPIC PROPERTIES OF THREE NEW ACTINIDE (IV) BOROHYDRIDES

    E-Print Network [OSTI]

    Banks, Rodney Howard

    2010-01-01T23:59:59.000Z

    uranium tetrakis-borohydrides were prepared by a different reaction which involves the actinide tetrafluoride

  12. Use of the TRUEX process for the pretreatment of neutralized cladding removal waste (NCRW) sludge -- Results of FY 1990 studies

    SciTech Connect (OSTI)

    Swanson, J.L.

    1991-09-01T23:59:59.000Z

    The goal of this process is to separate the transuranic elements from the bulk components so that the bulk components can be disposed of as low-level waste with only a small transuranic-containing fraction requiring geologic disposal. The pretreatment process examined here is the one indicated to be most promising in the initial studies. It involves dissolving the unwashed sludge in nitric acid and then using the TRUEX solvent extraction process to remove the transuranic elements from the bulk components of the waste. The areas identified in this work that need additional information are gradual precipitate formation as dissolved sludge solutions age, and formation of solid material when the dissolved sludge solution is contacted with the solvent used in the TRUEX process. 5 refs., 71 figs., 10 tabs.

  13. Removal of Selenium from Wastewater using ZVI and Hybrid ZVI/Iron Oxide Process

    E-Print Network [OSTI]

    Yang, Zhen

    2012-12-20T23:59:59.000Z

    than 10 ug/L is possible within four month long time. Koren et al. also validated the effectiveness of P. stutzeri to convert selenium to elemental selenium (Koren et al., 1992). Maximum reduction rates were demonstrated to happen in pH of 7 to 9.5.../L was loaded together with solutions containing both Se(IV) and Se(VI). The removal rate of Se(IV) can be higher than 95% while that of Se(VI) is about 80%. Absorbing selenium onto a lanthanum oxide substrates was also investigated by researchers (Adutwum...

  14. Complexation of lanthanides and actinides by acetohydroxamic acid

    SciTech Connect (OSTI)

    Taylor, R.J. [British Technology Centre, Nexia Solutions, Sellafield, Seascale, CA20 1PG (United Kingdom); Sinkov, S.I. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Choppin, G.R. [Department of Chemistry and Biochemistry, Florida State University, Tallahassee, FL (United States)

    2008-07-01T23:59:59.000Z

    Acetohydroxamic acid (AHA) has been proposed as a suitable reagent for the complexant-based, as opposed to reductive, stripping of plutonium and neptunium ions from the tributylphosphate solvent phase in advanced PUREX or UREX processes designed for future nuclear-fuel reprocessing. Stripping is achieved by the formation of strong hydrophilic complexes with the tetravalent actinides in nitric acid solutions. To underpin such applications, knowledge of the complexation constants of AHA with all relevant actinide (5f) and lanthanide (4f) ions is therefore important. This paper reports the determination of stability constants of AHA with the heavier lanthanide ions (Dy-Yb) and also U(IV) and Th(IV) ions. Comparisons with our previously published AHA stability-constant data for 4f and 5f ions are made. (authors)

  15. E-Print Network 3.0 - actinides recycling studies Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    minor Actinides (predominantly... of data, the study of the possibility of transmuting heavy actinides in PWRs, the development of codes... on purely minor actinide fuel,...

  16. E-Print Network 3.0 - actinide burning experiment Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    actinide burning molten salt - liquid f... for fission products and actinides in subcritical cores with different neutron spectra. This experiment... on purely minor actinide...

  17. Experimental studies of actinides in molten salts

    SciTech Connect (OSTI)

    Reavis, J.G.

    1985-06-01T23:59:59.000Z

    This review stresses techniques used in studies of molten salts containing multigram amounts of actinides exhibiting intense alpha activity but little or no penetrating gamma radiation. The preponderance of studies have used halides because oxygen-containing actinide compounds (other than oxides) are generally unstable at high temperatures. Topics discussed here include special enclosures, materials problems, preparation and purification of actinide elements and compounds, and measurements of various properties of the molten volts. Property measurements discussed are phase relationships, vapor pressure, density, viscosity, absorption spectra, electromotive force, and conductance. 188 refs., 17 figs., 6 tabs.

  18. Removal of hydrophobic Volatile Organic Compounds1 in an integrated process coupling Absorption and2

    E-Print Network [OSTI]

    Boyer, Edmond

    technology like photochemical oxidation shows high efficiency70 but is also high energy-consuming; moreover processes involve water as absorbent, they appear not always really efficient for the treatment of24 of the process, hydrophobic VOC27 absorption in a gas-liquid contactor, and biodegradation in the TPPB. VOC

  19. RADIOLOGICAL CONTROLS FOR PLUTONIUM CONTAMINATED PROCESS EQUIPMENT REMOVAL FROM 232-Z CONTAMINATED WASTE RECOVERY PROCESS FACILITY AT THE PLUTONIUM FINSHING PLANT (PFP)

    SciTech Connect (OSTI)

    MINETTE, M.J.

    2007-05-30T23:59:59.000Z

    The 232-Z facility at Hanford's Plutonium Finishing Plant operated as a plutonium scrap incinerator for 11 years. Its mission was to recover residual plutonium through incinerating and/or leaching contaminated wastes and scrap material. Equipment failures, as well as spills, resulted in the release of radionuclides and other contamination to the building, along with small amounts to external soil. Based on the potential threat posed by the residual plutonium, the U.S. Department of Energy (DOE) issued an Action Memorandum to demolish Building 232-2, Comprehensive Environmental Response Compensation, and Liability Act (CERC1.A) Non-Time Critical Removal Action Memorandum for Removal of the 232-2 Waste Recovery Process Facility at the Plutonium Finishing Plant (04-AMCP-0486).

  20. Removal of pharmaceuticals and endocrine disrupting compounds in water recycling process using reverse osmosis systems 

    E-Print Network [OSTI]

    Al-Rifai, Jawad H.; Khabbazb, Hadi; Schäfer, Andrea

    2011-01-01T23:59:59.000Z

    A detailed investigation was carried out to evaluate the occurrence, persistence and fate of a range of micropollutants at different processing points at a full-scale water recycling plant (WRP) in Queensland, Australia. ...

  1. BWR Assembly Optimization for Minor Actinide Recycling

    SciTech Connect (OSTI)

    G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

    2010-03-22T23:59:59.000Z

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

  2. Actinide minimization using pressurized water reactors

    E-Print Network [OSTI]

    Visosky, Mark Michael

    2006-01-01T23:59:59.000Z

    Transuranic actinides dominate the long-term radiotoxity in spent LWR fuel. In an open fuel cycle, they impose a long-term burden on geologic repositories. Transmuting these materials in reactor systems is one way to ease ...

  3. Joint Actinide Shock Physics Experimental Research - JASPER

    ScienceCinema (OSTI)

    None

    2015-01-09T23:59:59.000Z

    Commonly known as JASPER the Joint Actinide Shock Physics Experimental Research facility is a two stage light gas gun used to study the behavior of plutonium and other materials under high pressures, temperatures, and strain rates.

  4. Joint Actinide Shock Physics Experimental Research - JASPER

    SciTech Connect (OSTI)

    None

    2014-10-31T23:59:59.000Z

    Commonly known as JASPER the Joint Actinide Shock Physics Experimental Research facility is a two stage light gas gun used to study the behavior of plutonium and other materials under high pressures, temperatures, and strain rates.

  5. The UV/H2O2 advanced oxidation process in UV disinfection units : removal of selected phosphate esters by hydroxyl radical

    E-Print Network [OSTI]

    Machairas, Alexandros, 1980-

    2004-01-01T23:59:59.000Z

    In this work, the issue of how to remove phosphate esters from drinking water is examined. From the various treatment processes available, the oxidation of phosphate esters through hydroxyl radical generated by the UV/H202 ...

  6. The carbon footprint analysis of wastewater treatment plants and nitrous oxide emissions from full-scale biological nitrogen removal processes in Spain

    E-Print Network [OSTI]

    Xu, Xin, S.M. Massachusetts Institute of Technology

    2013-01-01T23:59:59.000Z

    This thesis presents a general model for the carbon footprint analysis of advanced wastewater treatment plants (WWTPs) with biological nitrogen removal processes, using a life cycle assessment (LCA) approach. Literature ...

  7. MINOR ACTINIDE SEPARATIONS USING ION EXCHANGERS OR IONIC LIQUIDS

    SciTech Connect (OSTI)

    Hobbs, D.; Visser, A.; Bridges, N.

    2011-09-20T23:59:59.000Z

    This project seeks to determine if (1) inorganic-based ion exchange materials or (2) electrochemical methods in ionic liquids can be exploited to provide effective Am and Cm separations. Specifically, we seek to understand the fundamental structural and chemical factors responsible for the selectivity of inorganic-based ion-exchange materials for actinide and lanthanide ions. Furthermore, we seek to determine whether ionic liquids can serve as the electrolyte that would enable formation of higher oxidation states of Am and other actinides. Experiments indicated that pH, presence of complexants and Am oxidation state exhibit significant influence on the uptake of actinides and lanthanides by layered sodium titanate and hybrid zirconium and tin phosphonate ion exchangers. The affinity of the ion exchangers increased with increasing pH. Greater selectivity among Ln(III) ions with sodium titanate materials occurs at a pH close to the isoelectric potential of the ion exchanger. The addition of DTPA decreased uptake of Am and Ln, whereas the addition of TPEN generally increases uptake of Am and Ln ions by sodium titanate. Testing confirmed two different methods for producing Am(IV) by oxidation of Am(III) in ionic liquids (ILs). Experimental results suggest that the unique coordination environment of ionic liquids inhibits the direct electrochemical oxidation of Am(III). The non-coordinating environment increases the oxidation potential to a higher value, while making it difficult to remove the inner coordination of water. Both confirmed cases of Am(IV) were from the in-situ formation of strong chemical oxidizers.

  8. A simplified model of aerosol removal by natural processes in reactor containments

    SciTech Connect (OSTI)

    Powers, D.A.; Washington, K.E.; Sprung, J.L. [Sandia National Labs., Albuquerque, NM (United States); Burson, S.B. [Nuclear Regulatory Commission, Washington, DC (United States)

    1996-07-01T23:59:59.000Z

    Simplified formulae are developed for estimating the aerosol decontamination that can be achieved by natural processes in the containments of pressurized water reactors and in the drywells of boiling water reactors under severe accident conditions. These simplified formulae were derived by correlation of results of Monte Carlo uncertainty analyses of detailed models of aerosol behavior under accident conditions. Monte Carlo uncertainty analyses of decontamination by natural aerosol processes are reported for 1,000, 2,000, 3,000, and 4,000 MW(th) pressurized water reactors and for 1,500, 2,500, and 3,500 MW(th) boiling water reactors. Uncertainty distributions for the decontamination factors and decontamination coefficients as functions of time were developed in the Monte Carlo analyses by considering uncertainties in aerosol processes, material properties, reactor geometry and severe accident progression. Phenomenological uncertainties examined in this work included uncertainties in aerosol coagulation by gravitational collision, Brownian diffusion, turbulent diffusion and turbulent inertia. Uncertainties in aerosol deposition by gravitational settling, thermophoresis, diffusiophoresis, and turbulent diffusion were examined. Electrostatic charging of aerosol particles in severe accidents is discussed. Such charging could affect both the coagulation and deposition of aerosol particles. Electrostatic effects are not considered in most available models of aerosol behavior during severe accidents and cause uncertainties in predicted natural decontamination processes that could not be taken in to account in this work. Median (50%), 90 and 10% values of the uncertainty distributions for effective decontamination coefficients were correlated with time and reactor thermal power. These correlations constitute a simplified model that can be used to estimate the decontamination by natural aerosol processes at 3 levels of conservatism. Applications of the model are described.

  9. Field Demonstration of a Membrane Process to Recover Heavy Hydrocarbons and to Remove Water from Natural Gas

    SciTech Connect (OSTI)

    R. Baker; T. Hofmann; K. A. Lokhandwala

    2006-09-29T23:59:59.000Z

    The objective of this project is to design, construct and field demonstrate a membrane system to recover natural gas liquids (NGL) and remove water from raw natural gas. An extended field test to demonstrate system performance under real-world high-pressure conditions is being conducted to convince industry users of the efficiency and reliability of the process. The system was designed and fabricated by Membrane Technology and Research, Inc. (MTR) and installed and operated at BP Amoco's Pascagoula, MS plant. The Gas Research Institute is partially supporting the field demonstration and BP-Amoco helped install the unit and provides onsite operators and utilities. The gas processed by the membrane system meets pipeline specifications for dew point and BTU value and can be delivered without further treatment to the pipeline. Based on data from prior membrane module tests, the process is likely to be significantly less expensive than glycol dehydration followed by propane refrigeration, the principal competitive technology. During the course of this project, MTR has sold 13 commercial units related to the field test technology, and by the end of this demonstration project the process will be ready for broader commercialization. A route to commercialization has been developed during this project and involves collaboration with other companies already servicing the natural gas processing industry.

  10. Process Optimization for Solid Extraction, Flavor Improvement and Fat Removal in the Production of Soymilk From Full Fat Soy Flakes

    SciTech Connect (OSTI)

    Stanley Prawiradjaja

    2003-05-31T23:59:59.000Z

    Traditionally soymilk has been made with whole soybeans; however, there are other alternative raw ingredients for making soymilk, such as soy flour or full-fat soy flakes. US markets prefer soymilk with little or no beany flavor. modifying the process or using lipoxygenase-free soybeans can be used to achieve this. Unlike the dairy industry, fat reduction in soymilk has been done through formula modification instead of by conventional fat removal (skimming). This project reports the process optimization for solids and protein extraction, flavor improvement and fat removal in the production of 5, 8 and 12 {sup o}Brix soymilk from full fat soy flakes and whole soybeans using the Takai soymilk machine. Proximate analyses, and color measurement were conducted in 5, 8 and 12 {sup o}Brix soymilk. Descriptive analyses with trained panelists (n = 9) were conducted using 8 and 12 {sup o}Brix lipoxygenase-free and high protein blend soy flake soymilks. Rehydration of soy flakes is necessary to prevent agglomeration during processing and increase extractability. As the rehydration temperature increases from 15 to 50 to 85 C, the hexanal concentration was reduced. Enzyme inactivation in soy flakes milk production (measured by hexanal levels) is similar to previous reports with whole soybeans milk production; however, shorter rehydration times can be achieved with soy flakes (5 to 10 minutes) compared to whole beans (8 to 12 hours). Optimum rehydration conditions for a 5, 8 and 12 {sup o}Brix soymilk are 50 C for 5 minutes, 85 C for 5 minutes and 85 C for 10 minutes, respectively. In the flavor improvement study of soymilk, the hexanal date showed differences between undeodorized HPSF in contrast to triple null soymilk and no differences between deodorized HPSF in contrast to deodorized triple null. The panelists could not differentiate between the beany, cereal, and painty flavors. However, the panelists responded that the overall aroma of deodorized 8 {sup o}Brix triple null and HPSF soymilk are lower than the undeodorized triple null and HPSF soymilk. The triple null soymilk was perceived to be more bitter than the HPSF soymilk by the sensory panel due to oxidation on the triple null soy flakes. This oxidation may produce other aroma that was not analyzed using the GC but noticed by the panelists. The sensory evaluation results did show that the deodorizer was able to reduce the soymilk aroma in HPSF soymilk so it would be similar to triple null soymilk at 8 {sup o}Brix level. Regardless of skimming method and solids levels, the fat from the whole soybean milk was removed less efficiently than soy flake milk (7 to 30% fat extraction in contrast to 50 to 80% fat extraction respectively). In soy flake milk, less fat was removed as the % solid increases regardless of the processing method. In whole soybean milk, the fat was removed less efficiently at lower solids level milk using the commercial dairy skimmer and more efficient at lower solids level using the centrifuge-decant method. Based on the Hunter L, a, b measurement, the color of the reduced fat soy flake milk yielded a darker, greener and less yellow colored milk than whole soymilk ({alpha} < 0.05), whereas no differences were noticed in reduced fat soybean milk ({alpha} < 0.05). Color comparison of whole and skim cow's milk showed the same the same trend as in the soymilk.

  11. Signal processing method and system for noise removal and signal extraction

    DOE Patents [OSTI]

    Fu, Chi Yung (San Francisco, CA); Petrich, Loren (Lebanon, OR)

    2009-04-14T23:59:59.000Z

    A signal processing method and system combining smooth level wavelet pre-processing together with artificial neural networks all in the wavelet domain for signal denoising and extraction. Upon receiving a signal corrupted with noise, an n-level decomposition of the signal is performed using a discrete wavelet transform to produce a smooth component and a rough component for each decomposition level. The n.sup.th level smooth component is then inputted into a corresponding neural network pre-trained to filter out noise in that component by pattern recognition in the wavelet domain. Additional rough components, beginning at the highest level, may also be retained and inputted into corresponding neural networks pre-trained to filter out noise in those components also by pattern recognition in the wavelet domain. In any case, an inverse discrete wavelet transform is performed on the combined output from all the neural networks to recover a clean signal back in the time domain.

  12. New processes to recovery methanol and remove oxygenates from Valero MTBE unit

    SciTech Connect (OSTI)

    Hillen, P.; Clemmons, J.

    1987-01-01T23:59:59.000Z

    The refiner today has to evaluate every available option to increase octane in the gasoline pool to make up for the loss in octane created by lead phase down. Production of MTBE is one of the most attractive options. MTBE is produced by selectivity reacting isobutylene with methanol. Valero Refining's refinery at Corpus Christie, Texas (formerly Saber Refining) is one of the most modern refineries built in the last decade to upgrade resids. As part of the gasoline upgrading Valero had built a Butamer Unit to convert normal butane to isobutane upstream of their HF Alkylation Unit. In 1984 as an ongoing optimization of its operations, Valero Refining evaluated various processes to enable it to increase the octane output, and decided to build an MTBE unit. Valero selected the MTBE process licensed by Arco Technology, Inc. and contracted with Jacobs Engineering Group, Inc., Houston, Texas to provide detailed engineering and procurement services.

  13. Magnetically assisted chemical separation (MACS) process: Preparation and optimization of particles for removal of transuranic elements

    SciTech Connect (OSTI)

    Nunez, L.; Kaminski, M.; Bradley, C.; Buchholz, B.A.; Aase, S.B.; Tuazon, H.E.; Vandegrift, G.F. [Argonne National Lab., IL (United States); Landsberger, S. [Univ. of Illinois, Urbana, IL (United States)

    1995-05-01T23:59:59.000Z

    The Magnetically Assisted Chemical Separation (MACS) process combines the selectivity afforded by solvent extractants with magnetic separation by using specially coated magnetic particles to provide a more efficient chemical separation of transuranic (TRU) elements, other radionuclides, and heavy metals from waste streams. Development of the MACS process uses chemical and physical techniques to elucidate the properties of particle coatings and the extent of radiolytic and chemical damage to the particles, and to optimize the stages of loading, extraction, and particle regeneration. This report describes the development of a separation process for TRU elements from various high-level waste streams. Polymer-coated ferromagnetic particles with an adsorbed layer of octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) diluted with tributyl phosphate (TBP) were evaluated for use in the separation and recovery of americium and plutonium from nuclear waste solutions. Due to their chemical nature, these extractants selectively complex americium and plutonium contaminants onto the particles, which can then be recovered from the solution by using a magnet. The partition coefficients were larger than those expected based on liquid[liquid extractions, and the extraction proceeded with rapid kinetics. Extractants were stripped from the particles with alcohols and 400-fold volume reductions were achieved. Particles were more sensitive to acid hydrolysis than to radiolysis. Overall, the optimization of a suitable NMCS particle for TRU separation was achieved under simulant conditions, and a MACS unit is currently being designed for an in-lab demonstration.

  14. Strategic Design and Optimization of Inorganic Sorbents for Cesium, Strontium and Actinides

    SciTech Connect (OSTI)

    Maginn, Edward J.

    2005-07-01T23:59:59.000Z

    The basic science goal in this project is to identify structure/affinity relationships for selected radionuclides and existing sorbents. The research will then apply this knowledge to the design and synthesis of sorbents that will exhibit increased cesium, strontium and actinide removal. The target problem focuses on the treatment of high-level nuclear wastes. The general approach can likewise be applied to non-radioactive separations.

  15. Process for removal of mineral particulates from coal-derived liquids

    DOE Patents [OSTI]

    McDowell, William J. (Knoxville, TN)

    1980-01-01T23:59:59.000Z

    Suspended mineral solids are separated from a coal-derived liquid containing the solids by a process comprising the steps of: (a) contacting said coal-derived liquid containing solids with a molten additive having a melting point of 100.degree.-500.degree. C. in an amount of up to 50 wt. % with respect to said coal-derived liquid containing solids, said solids present in an amount effective to increase the particle size of said mineral solids and comprising material or mixtures of material selected from the group of alkali metal hydroxides and inorganic salts having antimony, tin, lithium, sodium, potassium, magnesium, calcium, beryllium, aluminum, zinc, molybdenum, cobalt, nickel, ruthenium, rhodium or iron cations and chloride, iodide, bromide, sulfate, phosphate, borate, carbonate, sulfite, or silicate anions; and (b) maintaining said coal-derived liquid in contact with said molten additive for sufficient time to permit said mineral matter to agglomerate, thereby increasing the mean particle size of said mineral solids; and (c) recovering a coal-derived liquid product having reduced mineral solids content. The process can be carried out with less than 5 wt. % additive and in the absence of hydrogen pressure.

  16. HIGH TEMPERATURE REMOVAL OF H{sub 2}S FROM COAL GASIFICATION PROCESS STREAMS USING AN ELECTROCHEMICAL MEMBRANE SYSTEM

    SciTech Connect (OSTI)

    Jack Winnick; Meilin Liu

    2003-06-01T23:59:59.000Z

    A bench scale set-up was constructed to test the cell performance at 600-700 C and 1 atm. The typical fuel stream inlet proportions were 34% CO, 22% CO{sub 2}, 35% H{sub 2}, 8% H{sub 2}O, and 450-2000 ppm H{sub 2}S. The fundamental transport restrictions for sulfur species in an electrochemical cell were examined. Temperature and membrane thickness were varied to examine how these parameters affect the maximum flux of H{sub 2}S removal. It was found that higher temperature allows more sulfide species to enter the electrolyte, thus increasing the sulfide flux across the membrane and raising the maximum flux of H{sub 2}S removal. The results identify sulfide diffusion across the membrane as the rate-limiting step in H{sub 2}S removal. The maximum H{sub 2}S removal flux of 1.1 x 10-6 gmol H{sub 2}S min{sup -1} cm{sup -2} (or 3.5 mA cm{sup -2}) was obtained at 650 C, with a membrane that was 0.9 mm thick, 36% porous, and had an estimated tortuosity of 3.6. Another focus of this thesis was to examine the stability of cathode materials in full cell trials. A major hurdle that remains in process scale-up is cathode selection, as the lifetime of the cell will depend heavily on the lifetime of the cathode material, which is exposed to very sour gas. Materials that showed success in the past (i.e. cobalt sulfides and Y{sub 0.9}Ca{sub 0.1}FeO{sub 3}) were examined but were seen to have limitations in operating environment and temperature. Therefore, other novel metal oxide compounds were studied to find possible candidates for full cell trials. Gd{sub 2}TiMoO{sub 7} and La{sub 0.7}Sr{sub 0.3}VO{sub 3} were the compounds that retained their structure best even when exposed to high H{sub 2}S, CO{sub 2}, and H{sub 2}O concentrations.

  17. Process for simultaneous removal of SO[sub 2] and NO[sub x] from gas streams

    DOE Patents [OSTI]

    Rosenberg, H.S.

    1987-02-03T23:59:59.000Z

    A process is described for simultaneous removal of SO[sub 2] and NO[sub x] from a gas stream that includes flowing the gas stream to a spray dryer and absorbing a portion of the SO[sub 2] content of the gas stream and a portion of the NO[sub x] content of the gas stream with ZnO by contacting the gas stream with a spray of an aqueous ZnO slurry; controlling the gas outlet temperature of the spray dryer to within the range of about a 0 to 125 F approach to the adiabatic saturation temperature; flowing the gas, unreacted ZnO and absorbed SO[sub 2] and NO[sub x] from the spray dryer to a fabric filter and collecting any solids therein and absorbing a portion of the SO[sub 2] remaining in the gas stream and a portion of the NO[sub x] remaining in the gas stream with ZnO; and controlling the ZnO content of the aqueous slurry so that sufficient unreacted ZnO is present in the solids collected in the fabric filter to react with SO[sub 2] and NO[sub x] as the gas passes through the fabric filter whereby the overall feed ratio of ZnO to SO[sub 2] plus NO[sub x] is about 1.0 to 4.0 moles of ZnO per of SO[sub 2] and about 0.5 to 2.0 moles of ZnO per mole of NO[sub x]. Particulates may be removed from the gas stream prior to treatment in the spray dryer. The process further allows regeneration of ZnO that has reacted to absorb SO[sub 2] and NO[sub x] from the gas stream and acid recovery. 4 figs.

  18. Process studies for a new method of removing H/sub 2/S from industrial gas streams

    SciTech Connect (OSTI)

    Neumann, D.W.; Lynn, S.

    1986-07-01T23:59:59.000Z

    A process for the removal of hydrogen sulfide from coal-derived gas streams has been developed. The basis for the process is the absorption of H/sub 2/S into a polar organic solvent where it is reacted with dissolved sulfur dioxide to form elemental sulfur. After sulfur is crystallized from solution, the solvent is stripped to remove dissolved gases and water formed by the reaction. The SO/sub 2/ is generated by burning a portion of the sulfur in a furnace where the heat of combustion is used to generate high pressure steam. The SO/sub 2/ is absorbed into part of the lean solvent to form the solution necessary for the first step. The kinetics of the reaction between H/sub 2/S and SO/sub 2/ dissolved in mixtures of N,N-Dimethylaniline (DMA)/ Diethylene Glycol Monomethyl Ether and DMA/Triethylene Glycol Dimethyl Ether was studied by following the temperature rise in an adiabatic calorimeter. This irreversible reaction was found to be first-order in both H/sub 2/S and SO/sub 2/, with an approximates heat of reaction of 28 kcal/mole of SO/sub 2/. The sole products of the reaction appear to be elemental sulfur and water. The presence of DMA increases the value of the second-order rate constant by an order of magnitude over that obtained in the glycol ethers alone. Addition of other tertiary aromatic amines enhances the observed kinetics; heterocyclic amines (e.g., pyridine derivatives) have been found to be 10 to 100 times more effective as catalysts when compared to DMA.

  19. Process for removing halogenated aliphatic and aromatic compounds from petroleum products. [Polychlorinated biphenyls; methylene chloride; perchloroethylene; trichlorofluoroethane; trichloroethylene; chlorobenzene

    DOE Patents [OSTI]

    Googin, J.M.; Napier, J.M.; Travaglini, M.A.

    1982-03-31T23:59:59.000Z

    A process for removing halogenated aliphatic and aromatic compounds, e.g., polychlorinated biphenyls, from petroleum products by solvent extraction. The halogenated aliphatic and aromatic compounds are extracted from a petroleum product into a polar solvent by contracting the petroleum product with the polar solvent. The polar solvent is characterized by a high solubility for the extracted halogenated aliphatic and aromatic compounds, a low solubility for the petroleum product and considerable solvent power for polyhydroxy compound. The preferred polar solvent is dimethylformamide. A miscible polyhydroxy compound, such as, water, is added to the polar extraction solvent to increase the polarity of the polar extraction solvent. The halogenated aliphatic and aromatic compounds are extracted from the highly-polarized mixture of polyhydroxy compound and polar extraction solvent into a low polar or nonpolar solvent by contacting the polyhydroxy compound-polar solvent mixture with the low polar or nonpolar solvent. The halogenated aliphatic and aromatic compounds in the low polar or nonpolar solvent by physical means, e.g., vacuum evaporation. The polar and nonpolar solvents are recovered for recycling. The process can easily be designed for continuous operation. Advantages of the process include that the polar solvent and a major portion of the nonpolar solvent can be recycled, the petroleum products are reclaimable and the cost for disposing of waste containing polychlorinated biphenyls is significantly reduced. 2 tables.

  20. Removal of organic and inorganic sulfur from Ohio coal by combined physical and chemical process. Final report

    SciTech Connect (OSTI)

    Attia, Y.A.; Zeky, M.El.; Lei, W.W.; Bavarian, F.; Yu, S. [Ohio State Univ., Columbus, OH (United States). Dept. of Materials Science and Engineering

    1989-04-28T23:59:59.000Z

    This project consisted of three sections. In the first part, the physical cleaning of Ohio coal by selective flocculation of ultrafine slurry was considered. In the second part, the mild oxidation process for removal of pyritic and organic sulfur.was investigated. Finally, in-the third part, the combined effects of these processes were studied. The physical cleaning and desulfurization of Ohio coal was achieved using selective flocculation of ultrafine coal slurry in conjunction with froth flotation as flocs separation method. The finely disseminated pyrite particles in Ohio coals, in particular Pittsburgh No.8 seam, make it necessary to use ultrafine ({minus}500 mesh) grinding to liberate the pyrite particles. Experiments were performed to identify the ``optimum`` operating conditions for selective flocculation process. The results indicated that the use of a totally hydrophobic flocculant (FR-7A) yielded the lowest levels of mineral matters and total sulfur contents. The use of a selective dispersant (PAAX) increased the rejection of pyritic sulfur further. In addition, different methods of floc separation techniques were tested. It was found that froth flotation system was the most efficient method for separation of small coal flocs.

  1. Comparative Study of f-Element Electronic Structure across a Series of Multimetallic Actinide, Lanthanide-Actinide and Lanthanum-Actinide Complexes Possessing Redox-Active Bridging Ligands

    SciTech Connect (OSTI)

    Schelter, Eric J.; Wu, Ruilian; Veauthier, Jacqueline M.; Bauer, Eric D.; Booth, Corwin H.; Thomson, Robert K.; Graves, Christopher R.; John, Kevin D.; Scott, Brian L.; Thompson, Joe D.; Morris, David E.; Kiplinger, Jaqueline L.

    2010-02-24T23:59:59.000Z

    A comparative examination of the electronic interactions across a series of trimetallic actinide and mixed lanthanide-actinide and lanthanum-actinide complexes is presented. Using reduced, radical terpyridyl ligands as conduits in a bridging framework to promote intramolecular metal-metal communication, studies containing structural, electrochemical, and X-ray absorption spectroscopy are presented for (C{sub 5}Me{sub 5}){sub 2}An[-N=C(Bn)(tpy-M{l_brace}C{sub 5}Me4R{r_brace}{sub 2})]{sub 2} (where An = Th{sup IV}, U{sup IV}; Bn = CH{sub 2}C{sub 6}H{sub 5}; M = La{sup III}, Sm{sup III}, Yb{sup III}, U{sup III}; R = H, Me, Et) to reveal effects dependent on the identities of the metal ions and R-groups. The electrochemical results show differences in redox energetics at the peripheral 'M' site between complexes and significant wave splitting of the metal- and ligand-based processes indicating substantial electronic interactions between multiple redox sites across the actinide-containing bridge. Most striking is the appearance of strong electronic coupling for the trimetallic Yb{sup III}-U{sup IV}-Yb{sup III}, Sm{sup III}-U{sup IV}-Sm{sup III}, and La{sup III}-U{sup IV}-La{sup III} complexes, [8]{sup -}, [9b]{sup -} and [10b]{sup -}, respectively, whose calculated comproportionation constant K{sub c} is slightly larger than that reported for the benchmark Creutz-Taube ion. X-ray absorption studies for monometallic metallocene complexes of U{sup III}, U{sup IV}, and U{sup V} reveal small but detectable energy differences in the 'white-line' feature of the uranium L{sub III}-edges consistent with these variations in nominal oxidation state. The sum of this data provides evidence of 5f/6d-orbital participation in bonding and electronic delocalization in these multimetallic f-element complexes. An improved, high-yielding synthesis of 4{prime}-cyano-2,2{prime}:6{prime},2{double_prime}-terpyridine is also reported.

  2. Ultratrace analysis of transuranic actinides by laser-induced fluorescence

    DOE Patents [OSTI]

    Miller, S.M.

    1983-10-31T23:59:59.000Z

    Ultratrace quantities of transuranic actinides are detected indirectly by their effect on the fluorescent emissions of a preselected fluorescent species. Transuranic actinides in a sample are coprecipitated with a host lattice material containing at least one preselected fluorescent species. The actinide either quenches or enhances the laser-induced fluorescence of the preselected fluorescent species. The degree of enhancement or quenching is quantitatively related to the concentration of actinide in the sample.

  3. Process for the combined removal of SO.sub.2 and NO.sub.x from flue gas

    DOE Patents [OSTI]

    Chang, Shih-Ger (El Cerrito, CA); Liu, David K. (Oakland, CA); Griffiths, Elizabeth A. (Neston, GB2); Littlejohn, David (Oakland, CA)

    1988-01-01T23:59:59.000Z

    The present invention in one aspect relates to a process for the simultaneous removal of NO.sub.x and SO.sub.2 from a fluid stream comprising mixtures thereof and in another aspect relates to the separation, use and/or regeneration of various chemicals contaminated or spent in the process and which includes the steps of: (A) contacting the fluid stream at a temperature of between about 105.degree. and 180.degree. C. with a liquid aqueous slurry or solution comprising an effective amount of an iron chelate of an amino acid moiety having at least one --SH group; (B) separating the fluid stream from the particulates formed in step (A) comprising the chelate of the amino acid moiety and fly ash; (C) washing and separating the particulates of step (B) with an aqueous solution having a pH value of between about 5 to 8; (D) subsequently washing and separating the particulates of step (C) with a strongly acidic aqueous solution having a pH value of between about 1 to 3; (E) washing and separating the particulates of step (D) with an basic aqueous solution having a pH value of between about 9 to 12; (F) optionally adding additional amino acid moiety, iron (II) and alkali to the aqueous liquid from step (D) to produce an aqueous solution or slurry similar to that in step (A) having a pH value of between about 4 to 12; and (G) recycling the aqueous slurry of step (F) to the contacting zone of step (A). Steps (D) and (E) can be carried out in the reverse sequence, however the preferred order is (D) and then (E). In another preferred embodiment the present invention provides a process for the removal of NO.sub.x, SO.sub.2 and particulates from a fluid stream which includes the steps of (A) injecting into a reaction zone an aqueous solution itself comprising (i) an amino acid moiety selected from those described above; (ii) iron (II) ion; and (iii) an alkali, wherein the aqueous solution has a pH of between about 4 and 11; followed by solids separation and washing as is described in steps (B), (C), (D) and (E) above. The overall process is useful to reduce acid rain components from combustion gas sources.

  4. Separation and analysis of actinides by extraction chromatography coupled with alpha-particle liquid scintillation spectrometry

    SciTech Connect (OSTI)

    Cadieux, J.R. Jr.; Reboul, S.H. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1996-11-01T23:59:59.000Z

    This work describes the development and testing of a new method for the separation and analysis of most actinides of interest in environmental samples. It combines simplified extraction chromatography using highly selective absorption resins (EiChrom columns) to partition the individual actinides with the measurement of their alpha-particle activities by liquid scintillation spectrometry using the Photon-Electron Rejecting Alpha Liquid Scintillation (PERALS{sup TM}) system. Water and soil samples along with environment quality-assurance standards are routinely processed by this method with an accuracy of {+-}5 to 20% at activity levels of 0.01 to 0.1 Bq.

  5. Evaluation of improved techniques for the removal of fission products from process wastewater and groundwater: FY 1997 status

    SciTech Connect (OSTI)

    Bostick, D.T.; DePaoli, S.M.; Guo, B.

    1998-02-01T23:59:59.000Z

    The primary goals of this effort in FY 1997 were to survey local end users of wastewater treatment technology and then to evaluate recently available treatment processes in light of user needs. Survey results indicate that local sites are confronted with a limited, and shrinking, budget for treating aqueous waste streams. Therefore, a process will be selected primarily on the basis of sorbent costs, use of existing equipment, and disposal costs for spent processing materials. Current laboratory testing and economic studies have been directed toward addressing the technical issues specific to the removal of {sup 90}Sr and {sup 137}Cs from groundwater and process wastewater. This year`s efforts have concentrated on evaluating the engineered form of crystalline silicotitanates (CSTs) for near neutral pH applications. Both powder and pellet forms of CST can be obtained through UOP; this task evaluated only the engineered form of the sorbent for wastewater remediation. Preliminary experimental efforts included measuring the average particle size, surface water content, total sodium content, ion exchange capacity, and equilibration mixing time. The as received material contains approximately 10% fines, which adhere to the CST pellet. The cesium and strontium ion-exchange capacities, based on multiple contacts with 50 ppm of the metal, are 0.8 meq/g and 1.1 meq/g, respectively. Batch tests indicated that an equilibrium mixing time of 100 h was required for cesium sorption. Group 2 cations (Sr, Ca, and Mg) required greater than 500 h. Particle diffusion coefficients were estimated for each of these cations from the batch studies.

  6. EXPERIENCE SUMMARY Development of the TALSqueak (Trivalent Actinide Lanthanide Separation using QUicker Extractants and

    E-Print Network [OSTI]

    trivalent actinides Development of Warm Water Oxidation chemistry for remediation of Hanford's K Basin Development of the LimeAid Process to compliment the efforts of Hanford's Waste Treatment Plant Investigation sludge remediation of the Hanford Site Determination of thermodynamic parameters for biphasic systems

  7. Silica Scaling Removal Process

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    systems Water treatment systems Water evaporation systems Potential mining applications (produced water) Industry applications for which silica scaling must be prevented Benefits:...

  8. Silica Scaling Removal Process

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administrationcontroller systemsBi (2)Sharing Smart GridShift EndSidneyChemistry » Silica Scaling

  9. Low Cost Chemical Feedstocks Using an Improved and Energy Efficient Natural Gas Liquid (NGL) Removal Process, Final Technical Report

    SciTech Connect (OSTI)

    Meyer, Howard, S.; Lu, Yingzhong

    2012-08-10T23:59:59.000Z

    The overall objective of this project is to develop a new low-cost and energy efficient Natural Gas Liquid (NGL) recovery process - through a combination of theoretical, bench-scale and pilot-scale testing - so that it could be offered to the natural gas industry for commercialization. The new process, known as the IROA process, is based on U.S. patent No. 6,553,784, which if commercialized, has the potential of achieving substantial energy savings compared to currently used cryogenic technology. When successfully developed, this technology will benefit the petrochemical industry, which uses NGL as feedstocks, and will also benefit other chemical industries that utilize gas-liquid separation and distillation under similar operating conditions. Specific goals and objectives of the overall program include: (i) collecting relevant physical property and Vapor Liquid Equilibrium (VLE) data for the design and evaluation of the new technology, (ii) solving critical R&D issues including the identification of suitable dehydration and NGL absorbing solvents, inhibiting corrosion, and specifying proper packing structure and materials, (iii) designing, construction and operation of bench and pilot-scale units to verify design performance, (iv) computer simulation of the process using commercial software simulation platforms such as Aspen-Plus and HYSYS, and (v) preparation of a commercialization plan and identification of industrial partners that are interested in utilizing the new technology. NGL is a collective term for C2+ hydrocarbons present in the natural gas. Historically, the commercial value of the separated NGL components has been greater than the thermal value of these liquids in the gas. The revenue derived from extracting NGLs is crucial to ensuring the overall profitability of the domestic natural gas production industry and therefore of ensuring a secure and reliable supply in the 48 contiguous states. However, rising natural gas prices have dramatically reduced the economic incentive to extract NGLs from domestically produced natural gas. Successful gas processors will be those who adopt technologies that are less energy intensive, have lower capital and operating costs and offer the flexibility to tailor the plant performance to maximize product revenue as market conditions change, while maintaining overall system efficiency. Presently, cryogenic turbo-expander technology is the dominant NGL recovery process and it is used throughout the world. This process is known to be highly energy intensive, as substantial energy is required to recompress the processed gas back to pipeline pressure. The purpose of this project is to develop a new NGL separation process that is flexible in terms of ethane rejection and can reduce energy consumption by 20-30% from current levels, particularly for ethane recoveries of less than 70%. The new process integrates the dehydration of the raw natural gas stream and the removal of NGLs in such a way that heat recovery is maximized and pressure losses are minimized so that high-value equipment such as the compressor, turbo-expander, and a separate dehydration unit are not required. GTI completed a techno-economic evaluation of the new process based on an Aspen-HYSYS simulation model. The evaluation incorporated purchased equipment cost estimates obtained from equipment suppliers and two different commercial software packages; namely, Aspen-Icarus and Preliminary Design and Quoting Service (PDQ$). For a 100 MMscfd gas processing plant, the annualized capital cost for the new technology was found to be about 10% lower than that of conventional technology for C2 recovery above 70% and about 40% lower than that of conventional technology for C2 recovery below 50%. It was also found that at around 40-50% C2 recovery (which is economically justifiable at the current natural gas prices), the energy cost to recover NGL using the new technology is about 50% of that of conventional cryogenic technology.

  10. Comparative evaluation of DHDECMP (dihexyl-N,N-diethylcarbamoyl-methylphosphonate) and CMPO (octylphenyl-N,N,-diisobutylcarbamoylmethylphosphine oxide) as extractants for recovering actinides from nitric acid waste streams

    SciTech Connect (OSTI)

    Marsh, S.F.; Yarbro, S.L.

    1988-02-01T23:59:59.000Z

    Certain neutral, bifunctional organophosphorous compounds are of special value to the nuclear industry. Dihexyl-N,N-diethylcarbomoylmethylphosphonate (DHDECMP) and octylphenyl-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) are highly selective extractants for removing actinide and lanthanide elements from nitric acid. We obtained these two extractants from newly available commercial sources and evaluated them for recovering Am(III), Pu(IV), and U(VI) from nitric acid waste streams of plutonium processing operations. Variables included the extractant (DHSECMP or CMPO), extractant/tributylphosphate ratio, diluent, nitrate concentration, nitrate salt/nitric acid ratio, fluoride concentration, and contact time. Based on these experimental data, we selected DHDECMP as the perferred extractant for this application. 18 refs., 30 figs.

  11. Rapid determination of actinides in asphalt samples

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Maxwell, Sherrod L.; Culligan, Brian K.; Hutchison, Jay B.

    2014-01-12T23:59:59.000Z

    A new rapid method for the determination of actinides in asphalt samples has been developed that can be used in emergency response situations or for routine analysis If a radiological dispersive device (RDD), Improvised Nuclear Device (IND) or a nuclear accident such as the accident at the Fukushima Nuclear Power Plant in March, 2011 occurs, there will be an urgent need for rapid analyses of many different environmental matrices, including asphalt materials, to support dose mitigation and environmental clean up. The new method for the determination of actinides in asphalt utilizes a rapid furnace step to destroy bitumen and organics present in the asphalt and sodium hydroxide fusion to digest the remaining sample. Sample preconcentration steps are used to collect the actinides and a new stacked TRU Resin + DGA Resin column method is employed to separate the actinide isotopes in the asphalt samples. The TRU Resin plus DGA Resin separation approach, which allows sequential separation of plutonium, uranium, americium and curium isotopes in asphalt samples, can be applied to soil samples as well.

  12. Characterization of transuranium actinide alloy phase diagrams

    SciTech Connect (OSTI)

    Gibson, J.K.; Haire, R.G.; Gensini, M.M. [Oak Ridge National Lab., TN (United States); Ogawa, T. [Japan Atomic Energy Research Inst., Tokai (Japan)

    1994-05-02T23:59:59.000Z

    Alloys of Np have been studied less than those,of the neighboring elements, U and Pu; the higher actinides have received even less attention. Recent interest in {sup 237}Np, {sup 241}Am and other actinide isotopes as significant, long-lived and highly radiotoxic nuclear waste components, and particularly the roles of metallic materials new handling/separations and remediation technologies, demands that this paucity of information concerning alloy behaviors be addressed. An additional interest in these arises from the possibility of revealing fundamental properties and bonding interactions, which would further characterize the unique electronic structures (e.g., 5f electrons) of the actinide elements. The small empirical knowledge basis presently available for understanding and modeling the alloying behavior of Np is summarized here, with emphasis on our recent results for the Np-Am, Np-Zr and Np-Fe phase diag rams. In view of the limited experimental data base for neptunium and the transplutonium metals, the value of semi-empirical intermetallic bonding models for predicting actinide alloy thermodynamics is evaluated.

  13. Evaluation of an alkaline-side solvent extraction process for cesium removal from SRS tank waste using laboratory-scale centrifugal contactors

    SciTech Connect (OSTI)

    Leonard, R. A.; Conner, C.; Liberatore, M. W.; Sedlet, J.; Aase, S. B.; Vandegrift, G. F.

    1999-11-29T23:59:59.000Z

    An alkaline-side solvent extraction process for cesium removal from Savannah River Site (SRS) tank waste was evaluated experimentally using a laboratory-scale centrifugal contactor. Single-stage and multistage tests were conducted with this contactor to determine hydraulic performance, stage efficiency, and general operability of the process flowsheet. The results and conclusions of these tests are reported along with those from various supporting tests. Also discussed is the ability to scale-up from laboratory- to plant-scale operation when centrifugal contractors are used to carry out the solvent extraction process. While some problems were encountered, a promising solution for each problem has been identified. Overall, this alkaline-side cesium extraction process appears to be an excellent candidate for removing cesium from SRS tank waste.

  14. JOWOG 22/2 - Actinide Chemical Technology (July 9-13, 2012)

    SciTech Connect (OSTI)

    Jackson, Jay M. [Los Alamos National Laboratory; Lopez, Jacquelyn C. [Los Alamos National Laboratory; Wayne, David M. [Los Alamos National Laboratory; Schulte, Louis D. [Los Alamos National Laboratory; Finstad, Casey C. [Los Alamos National Laboratory; Stroud, Mary Ann [Los Alamos National Laboratory; Mulford, Roberta Nancy [Los Alamos National Laboratory; MacDonald, John M. [Los Alamos National Laboratory; Turner, Cameron J. [Los Alamos National Laboratory; Lee, Sonya M. [Los Alamos National Laboratory

    2012-07-05T23:59:59.000Z

    The Plutonium Science and Manufacturing Directorate provides world-class, safe, secure, and reliable special nuclear material research, process development, technology demonstration, and manufacturing capabilities that support the nation's defense, energy, and environmental needs. We safely and efficiently process plutonium, uranium, and other actinide materials to meet national program requirements, while expanding the scientific and engineering basis of nuclear weapons-based manufacturing, and while producing the next generation of nuclear engineers and scientists. Actinide Process Chemistry (NCO-2) safely and efficiently processes plutonium and other actinide compounds to meet the nation's nuclear defense program needs. All of our processing activities are done in a world class and highly regulated nuclear facility. NCO-2's plutonium processing activities consist of direct oxide reduction, metal chlorination, americium extraction, and electrorefining. In addition, NCO-2 uses hydrochloric and nitric acid dissolutions for both plutonium processing and reduction of hazardous components in the waste streams. Finally, NCO-2 is a key team member in the processing of plutonium oxide from disassembled pits and the subsequent stabilization of plutonium oxide for safe and stable long-term storage.

  15. Development of an Integrated Multi-Contaminant Removal Process Applied to Warm Syngas Cleanup for Coal-Based Advanced Gasification Systems

    SciTech Connect (OSTI)

    Howard Meyer

    2010-11-30T23:59:59.000Z

    This project met the objective to further the development of an integrated multi-contaminant removal process in which H2S, NH3, HCl and heavy metals including Hg, As, Se and Cd present in the coal-derived syngas can be removed to specified levels in a single/integrated process step. The process supports the mission and goals of the Department of Energyâ??s Gasification Technologies Program, namely to enhance the performance of gasification systems, thus enabling U.S. industry to improve the competitiveness of gasification-based processes. The gasification program will reduce equipment costs, improve process environmental performance, and increase process reliability and flexibility. Two sulfur conversion concepts were tested in the laboratory under this project, i.e., the solventbased, high-pressure University of California Sulfur Recovery Process â?? High Pressure (UCSRP-HP) and the catalytic-based, direct oxidation (DO) section of the CrystaSulf-DO process. Each process required a polishing unit to meet the ultra-clean sulfur content goals of <50 ppbv (parts per billion by volume) as may be necessary for fuel cells or chemical production applications. UCSRP-HP was also tested for the removal of trace, non-sulfur contaminants, including ammonia, hydrogen chloride, and heavy metals. A bench-scale unit was commissioned and limited testing was performed with simulated syngas. Aspen-Plus®-based computer simulation models were prepared and the economics of the UCSRP-HP and CrystaSulf-DO processes were evaluated for a nominal 500 MWe, coal-based, IGCC power plant with carbon capture. This report covers the progress on the UCSRP-HP technology development and the CrystaSulf-DO technology.

  16. Molecular dynamics simulation and topological analysis of the network structure of actinide-bearing materials

    E-Print Network [OSTI]

    Dewan, Leslie

    2013-01-01T23:59:59.000Z

    Actinide waste production and storage is a complex problem, and a whole-cycle approach to actinide management is necessary to minimize the total volume of waste. In this dissertation, I examine three actinide-bearing ...

  17. E-Print Network 3.0 - actinide partitioning studies Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    MA : Minor Actinide LLFP : Long... of data, the study of the possibility of transmuting heavy actinides in PWRs, the development of codes... and transmutation were divided in...

  18. E-Print Network 3.0 - actinide separations final Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    us to explore alternatives to some... , of minor actinides, i.e. neptunium, americium and curium. If stored in geological depositories, plutonium... actinides than uranium fuels,...

  19. E-Print Network 3.0 - actinides conditioning synthese Sample...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AND CONCEPT EVOLUTIONS Summary: , of minor actinides, i.e. neptunium, americium and curium. If stored in geological depositories, plutonium... actinides than uranium fuels,...

  20. E-Print Network 3.0 - actinide target preparation Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    for such materials... ' actinides, to distinguish them from the larger quantities of uranium and plutonium also present in the fuel... to extract and recycle all actinides in...

  1. E-Print Network 3.0 - actinide compounds Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    concept of valence instabilities in light actinides has been... response of ICF rare earth compounds and actinide materials. -The important aspects of physical property......

  2. E-Print Network 3.0 - actinide halide complexes Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    It has been observed that complexes of lanthanide, actinide, and transition metal activate... that these actinide alkyl complexes undergo interesting C-H and C-N bond...

  3. E-Print Network 3.0 - actinide complexation kinetics Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    It has been observed that complexes of lanthanide, actinide, and transition metal activate... that these actinide alkyl complexes undergo interesting C-H and C-N bond...

  4. Progress toward Biomass and Coal-Derived Syngas Warm Cleanup: Proof-of-Concept Process Demonstration of Multicontaminant Removal for Biomass Application

    SciTech Connect (OSTI)

    Howard, Christopher J.; Dagle, Robert A.; Lebarbier, Vanessa MC; Rainbolt, James E.; Li, Liyu; King, David L.

    2013-06-19T23:59:59.000Z

    Systems comprising of multiple sorbent and catalytic beds have been developed for the warm syngas cleanup of coal- and biomass-derived syngas. Tailored specifically for biomass application the process described here consists of six primary unit operations: 1) Na2CO3 bed for HCl removal, 2) two regenerable ZnO beds for bulk H2S removal, 3) ZnO bed for H2S polishing, 4) NiCu/SBA-16 sorbent for trace metal (e.g. AsH3) removal, 5) steam reforming catalyst bed for tars and light hydrocarbons reformation and NH3 decomposition, and a 6) Cu-based LT-WGS catalyst bed. Simulated biomass-derived syngas containing a multitude of inorganic contaminants (H2S, AsH3, HCl, and NH3) and hydrocarbon additives (methane, ethylene, benzene, and naphthalene) was used to demonstrate process effectiveness. The efficiency of the process was demonstrated for a period of 175 hours, during which no signs of deactivation were observed. Post-run analysis revealed small levels of sulfur slipped through the sorbent bed train to the two downstream catalytic beds. Future improvements could be made to the trace metal polishing sorbent to ensure complete inorganic contaminant removal (to low ppb level) prior to the catalytic steps. However, dual, regenerating ZnO beds were effective for continuous removal for the vast majority of the sulfur present in the feed gas. The process was effective for complete AsH3 and HCl removal. The steam reforming catalyst completely reformed all the hydrocarbons present in the feed (methane, ethylene, benzene, and naphthalene) to additional syngas. However, post-run evaluation, under kinetically-controlled conditions, indicates deactivation of the steam reforming catalyst. Spent material characterization suggests this is attributed, in part, to coke formation, likely due to the presence of benzene and/or naphthalene in the feed. Future adaptation of this technology may require dual, regenerable steam reformers. The process and materials described in this report hold promise for a warm cleanup of a variety of contaminant species within warm syngas.

  5. Regenerative process and system for the simultaneous removal of particulates and the oxides of sulfur and nitrogen from a gas stream

    DOE Patents [OSTI]

    Cohen, M.R.; Gal, E.

    1993-04-13T23:59:59.000Z

    A process and system are described for simultaneously removing from a gaseous mixture, sulfur oxides by means of a solid sulfur oxide acceptor on a porous carrier, nitrogen oxides by means of ammonia gas and particulate matter by means of filtration and for the regeneration of loaded solid sulfur oxide acceptor. Finely-divided solid sulfur oxide acceptor is entrained in a gaseous mixture to deplete sulfur oxides from the gaseous mixture, the finely-divided solid sulfur oxide acceptor being dispersed on a porous carrier material having a particle size up to about 200 microns. In the process, the gaseous mixture is optionally pre-filtered to remove particulate matter and thereafter finely-divided solid sulfur oxide acceptor is injected into the gaseous mixture.

  6. Chemical decontamination of process equipment using recyclable chelating agents

    SciTech Connect (OSTI)

    Palmer, P.A.

    1994-10-01T23:59:59.000Z

    The Babcock and Wilcox Company is performing research and development in the application of chelating chemicals to dissolve uranium compounds and other actinide species from the surfaces of DOE process equipment. A chelating system specific for the removal of uranium and other actinides will be applied to the component selected for full-scale demonstration of the process. After application of the chelating solvent for an appropriate time period, the spent solvent will be removed to a waste processing facility, and the dissolved radioactive contaminants will be precipitated out of the solution. The regenerated chelating solvent will then be available for reuse in the cleaning system, thereby minimizing the amount of secondary waste generated by the process. Phase 1 activity has begun with bench-scale tests in the laboratory, to screen and optimize candidate solvent systems, and will proceed to development of a chemical cleaning process that will be tested in a pilot facility on an actual piece of contaminated equipment. The potential for application of the chelating agent as a foam rather than a liquid will also be investigated. The advantage of foaming application is a reduction of solvent volume requiring eventual treatment. The second phase of this program will be a full-scale demonstration of the developed technology on contaminated process equipment at a DOE site.

  7. POTENTIAL BENCHMARKS FOR ACTINIDE PRODUCTION IN HANFORD REACTORS

    SciTech Connect (OSTI)

    PUIGH RJ; TOFFER H

    2011-10-19T23:59:59.000Z

    A significant experimental program was conducted in the early Hanford reactors to understand the reactor production of actinides. These experiments were conducted with sufficient rigor, in some cases, to provide useful information that can be utilized today in development of benchmark experiments that may be used for the validation of present computer codes for the production of these actinides in low enriched uranium fuel.

  8. Dynamic tests for actinide/lanthanide separation by CMPO solvent in fluorinated diluents

    SciTech Connect (OSTI)

    Tkachenko, L.; Babain, V.; Alyapyshev, M.; Vizniy, A.; Il'in, A. [V. G. Khlopin Radium Institute, 2nd Murinskiy Ave 28., St. Petersburg 194021 (Russian Federation); Shadrin, A. [A.A. Bochvar High-technology Research Institute of Inorganic Materials, 5-a, Rogova str., Moscow 123098 (Russian Federation)

    2013-07-01T23:59:59.000Z

    Actinide and lanthanide extraction by new solvent: 0.2 M phenyl-octyl-N,N-diiso-butylcarbamoyl-phosphine oxide (CMPO) + 30% TBP + formal of octafluoro-pentanol was studied. A dynamic test with this solvent was performed. It was shown that americium and lanthanides are effectively extracted from PUREX process raffinate. The separation of americium from light lanthanides was confirmed in the modified SETFICS flowsheet with this new solvent. (authors)

  9. Recovery of minor actinides from spent fuel using TPEN-immobilized gels

    SciTech Connect (OSTI)

    Koyama, S.; Suto, M.; Ohbayashi, H. [Oarai Research and Development Center, Japan Atomic Energy Agency, Oarai (Japan); Oaki, H. [Solutions Research Organization, Tokyo Institute of Technology, Tokyo (Japan); Takeshita, K. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo (Japan)

    2013-07-01T23:59:59.000Z

    A series of separation experiments was performed in order to study the recovery process for minor actinides (MAs), such as americium (Am) and curium (Cm), from the actual spent fuel by using an extraction chromatographic technique. N,N,N',N'-tetrakis-(4-propenyloxy-2-pyridylmethyl) ethylenediamine (TPPEN) is an N,N,N',N'-tetrakis (2-pyridylmethyl) ethylenediamine (TPEN) analogue consisting of an incorporated pyridine ring that acts as not only a ligand but also as a site for polymerization and crosslinking of the gel. The TPPEN and N-isopropylacrylamide (NIPA) were dissolved into dimethylformamide (DMF, Wako Co., Ltd.) and a silica beads polymer, and then TTPEN was immobilized chemically in a polymer gel (so called TPEN-gel). Mixed oxide (MOX) fuel, which was highly irradiated up to 119 GWD/MTM in the experimental fast reactor Joyo, was used as a reference spent fuel. First, uranium (U) and plutonium (Pu) were separated from the irradiated fuel using an ion-exchange method, and then, the platinum group elements were removed by CMPO to leave a mixed solution of MAs and lanthanides. The 3 mol% TPPEN-gel was packed with as an extraction column (CV: 1 ml) and then rinsed by 0.1 M NaNO{sub 3}(pH 4.0) for pH adjustment. After washing the column by 0.01 M NaNO{sub 3} (pH 4.0), Eu was detected and the recovery rate reached 93%. The MAs were then recovered by changing the eluent to 0.01 M NaNO{sub 3} (pH 2.0), and the recovery rate of Am was 48 %. The 10 mol% TPPEN-gel was used to improve adsorption coefficient of Am and a condition of eluent temperature was changed in order to confirm the temperature swing effect on TPEN-gel for MA. More than 90% Eu was detected in the eluent after washing with 0.01 M NaNO{sub 3} (pH 3.5) at 5 Celsius degrees. Americium was backwardly detected and eluted continuously during the same condition. After removal of Eu, the eluent temperature was changed to 32 Celsius degrees, then Am was detected (pH 3.0). Finally remained Am could be stripped from TPPEN-gel by changing the pH of the eluent to 2.0. These results These results prove that the proposed recovery process for MAs is a potential candidate for future reprocessing methods based on the extraction chromatographic technique. (authors)

  10. Waste treatment process for removal of contaminants from aqueous, mixed-waste solutions using sequential chemical treatment and crossflow microfiltration, followed by dewatering

    DOE Patents [OSTI]

    Vijayan, Sivaraman (Deep River, CA); Wong, Chi F. (Pembroke, CA); Buckley, Leo P. (Deep River, CA)

    1994-01-01T23:59:59.000Z

    In processes of this invention aqueous waste solutions containing a variety of mixed waste contaminants are treated to remove the contaminants by a sequential addition of chemicals and adsorption/ion exchange powdered materials to remove the contaminants including lead, cadmium, uranium, cesium-137, strontium-85/90, trichloroethylene and benzene, and impurities including iron and calcium. Staged conditioning of the waste solution produces a polydisperse system of size enlarged complexes of the contaminants in three distinct configurations: water-soluble metal complexes, insoluble metal precipitation complexes, and contaminant-bearing particles of ion exchange and adsorbent materials. The volume of the waste is reduced by separation of the polydisperse system by cross-flow microfiltration, followed by low-temperature evaporation and/or filter pressing. The water produced as filtrate is discharged if it meets a specified target water quality, or else the filtrate is recycled until the target is achieved.

  11. Waste treatment process for removal of contaminants from aqueous, mixed-waste solutions using sequential chemical treatment and crossflow microfiltration, followed by dewatering

    DOE Patents [OSTI]

    Vijayan, S.; Wong, C.F.; Buckley, L.P.

    1994-11-22T23:59:59.000Z

    In processes of this invention aqueous waste solutions containing a variety of mixed waste contaminants are treated to remove the contaminants by a sequential addition of chemicals and adsorption/ion exchange powdered materials to remove the contaminants including lead, cadmium, uranium, cesium-137, strontium-85/90, trichloroethylene and benzene, and impurities including iron and calcium. Staged conditioning of the waste solution produces a polydisperse system of size enlarged complexes of the contaminants in three distinct configurations: water-soluble metal complexes, insoluble metal precipitation complexes, and contaminant-bearing particles of ion exchange and adsorbent materials. The volume of the waste is reduced by separation of the polydisperse system by cross-flow microfiltration, followed by low-temperature evaporation and/or filter pressing. The water produced as filtrate is discharged if it meets a specified target water quality, or else the filtrate is recycled until the target is achieved. 1 fig.

  12. Methods to estimate equipment and materials that are candidates for removal during the decontamination of fuel processing facilities

    SciTech Connect (OSTI)

    Duncan, D.R.; Valero, O.J. [Westinghouse Hanford Co., Richland, WA (United States); Hyre, R.A.; Pottmeyer, J.A.; Millar, J.S.; Reddick, J.A. [Los Alamos Technical Associates, Inc., Kennewick, WA (United States)

    1995-02-01T23:59:59.000Z

    The methodology presented in this report provides a model for estimating the volume and types of waste expected from the removal of equipment and other materials during Decontamination and Decommissioning (D and D) of canyon-type fuel reprocessing facilities. This methodology offers a rough estimation technique based on a comparative analysis for a similar, previously studied, reprocessing facility. This approach is especially useful as a planning tool to save time and money while preparing for final D and D. The basic methodology described here can be extended for use at other types of facilities, such as glovebox or reactor facilities.

  13. Overview of Fiscal Year 2002 Research and Development for Savannah River Site's Salt Waste Processing Facility

    SciTech Connect (OSTI)

    H. D. Harmon, R. Leugemors, PNNL; S. Fink, M. Thompson, D. Walker, WSRC; P. Suggs, W. D. Clark, Jr

    2003-02-26T23:59:59.000Z

    The Department of Energy's (DOE) Savannah River Site (SRS) high-level waste program is responsible for storage, treatment, and immobilization of high-level waste for disposal. The Salt Processing Program (SPP) is the salt (soluble) waste treatment portion of the SRS high-level waste effort. The overall SPP encompasses the selection, design, construction and operation of treatment technologies to prepare the salt waste feed material for the site's grout facility (Saltstone) and vitrification facility (Defense Waste Processing Facility). Major constituents that must be removed from the salt waste and sent as feed to Defense Waste Processing Facility include actinides, strontium, cesium, and entrained sludge. In fiscal year 2002 (FY02), research and development (R&D) on the actinide and strontium removal and Caustic-Side Solvent Extraction (CSSX) processes transitioned from technology development for baseline process selection to providing input for conceptual design of the Salt Waste Processing Facility. The SPP R&D focused on advancing the technical maturity, risk reduction, engineering development, and design support for DOE's engineering, procurement, and construction (EPC) contractors for the Salt Waste Processing Facility. Thus, R&D in FY02 addressed the areas of actual waste performance, process chemistry, engineering tests of equipment, and chemical and physical properties relevant to safety. All of the testing, studies, and reports were summarized and provided to the DOE to support the Salt Waste Processing Facility, which began conceptual design in September 2002.

  14. E-Print Network 3.0 - actinide iii cations Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Actinide(III) 2.26x10-7 log f(CO2) -5.50 Actinide(IV) 5.66x10... Actinide(III) case: Americium in WIPP Brine 12;Figure 2 Actinide(V) case: Neptunium in WIPP Brine 12... November...

  15. Use of the TRUEX process for the pretreatment of neutralized cladding removal waste (NCRW) sludge: Results of a design basis experiment

    SciTech Connect (OSTI)

    Swanson, J L

    1991-07-01T23:59:59.000Z

    This report presents the results of an experiment designed to demonstrate the feasibility of a sludge dissolution/solvent extraction process to separate transuranic elements from the bulk components of Hanford neutralized cladding removal waste (NCRW) sludge. Such a separation would allow the bulk of the waste to be disposed of as low-level waste, which is much less costly than geologic disposal as would be required for the waste in its current form. The results indicate that the proposed process is well suited to meet the desired objectives. A composite sample of NCRW sludge taken from Tank 103-AW in 1986 was dissolved in nitric acid at room temperature. Dissolution of bulk components and all radionuclides was {ge}95% complete; thus, {le}5% of the bulk components will require geologic disposal. The TRUEX (TRansUranium EXtraction) solvent extraction process gave very good separation of the transuranic from the bulk components of the waste.

  16. Potentiometric Sensor for Real-Time Remote Surveillance of Actinides in Molten Salts

    SciTech Connect (OSTI)

    Natalie J. Gese; Jan-Fong Jue; Brenda E. Serrano; Guy L. Fredrickson

    2012-07-01T23:59:59.000Z

    A potentiometric sensor is being developed at the Idaho National Laboratory for real-time remote surveillance of actinides during electrorefining of spent nuclear fuel. During electrorefining, fuel in metallic form is oxidized at the anode while refined uranium metal is reduced at the cathode in a high temperature electrochemical cell containing LiCl-KCl-UCl3 electrolyte. Actinides present in the fuel chemically react with UCl3 and form stable metal chlorides that accumulate in the electrolyte. This sensor will be used for process control and safeguarding of activities in the electrorefiner by monitoring the concentrations of actinides in the electrolyte. The work presented focuses on developing a solid-state cation conducting ceramic sensor for detecting varying concentrations of trivalent actinide metal cations in eutectic LiCl-KCl molten salt. To understand the basic mechanisms for actinide sensor applications in molten salts, gadolinium was used as a surrogate for actinides. The ß?-Al2O3 was selected as the solid-state electrolyte for sensor fabrication based on cationic conductivity and other factors. In the present work Gd3+-ß?-Al2O3 was prepared by ion exchange reactions between trivalent Gd3+ from GdCl3 and K+-, Na+-, and Sr2+-ß?-Al2O3 precursors. Scanning electron microscopy (SEM) was used for characterization of Gd3+-ß?-Al2O3 samples. Microfocus X-ray Diffraction (µ-XRD) was used in conjunction with SEM energy dispersive X-ray spectroscopy (EDS) to identify phase content and elemental composition. The Gd3+-ß?-Al2O3 materials were tested for mechanical and chemical stability by exposing them to molten LiCl-KCl based salts. The effect of annealing on the exchanged material was studied to determine improvements in material integrity post ion exchange. The stability of the ß?-Al2O3 phase after annealing was verified by µ-XRD. Preliminary sensor tests with different assembly designs will also be presented.

  17. Influence of microorganisms on the oxidation state distribution of multivalent actinides under anoxic conditions

    SciTech Connect (OSTI)

    Reed, Donald Timothy [Los Alamos National Laboratory; Borkowski, Marian [Los Alamos National Laboratory; Lucchini, Jean - Francois [Los Alamos National Laboratory; Ams, David [Los Alamos National Laboratory; Richmann, M. K. [Los Alamos National Laboratory; Khaing, H. [Los Alamos National Laboratory; Swanson, J. S. [Los Alamos National Laboratory

    2010-12-10T23:59:59.000Z

    The fate and potential mobility of multivalent actinides in the subsurface is receiving increased attention as the DOE looks to cleanup the many legacy nuclear waste sites and associated subsurface contamination. Plutonium, uranium and neptunium are the near-surface multivalent contaminants of concern and are also key contaminants for the deep geologic disposal of nuclear waste. Their mobility is highly dependent on their redox distribution at their contamination source as well as along their potential migration pathways. This redox distribution is often controlled, especially in the near-surface where organic/inorganic contaminants often coexist, by the direct and indirect effects of microbial activity. Under anoxic conditions, indirect and direct bioreduction mechanisms exist that promote the prevalence of lower-valent species for multivalent actinides. Oxidation-state-specific biosorption is also an important consideration for long-term migration and can influence oxidation state distribution. Results of ongoing studies to explore and establish the oxidation-state specific interactions of soil bacteria (metal reducers and sulfate reducers) as well as halo-tolerant bacteria and Archaea for uranium, neptunium and plutonium will be presented. Enzymatic reduction is a key process in the bioreduction of plutonium and uranium, but co-enzymatic processes predominate in neptunium systems. Strong sorptive interactions can occur for most actinide oxidation states but are likely a factor in the stabilization of lower-valent species when more than one oxidation state can persist under anaerobic microbiologically-active conditions. These results for microbiologically active systems are interpreted in the context of their overall importance in defining the potential migration of multivalent actinides in the subsurface.

  18. Gas separation process using membranes with permeate sweep to remove CO.sub.2 from gaseous fuel combustion exhaust

    DOE Patents [OSTI]

    Wijmans Johannes G. (Menlo Park, CA); Merkel, Timothy C. (Menlo Park, CA); Baker, Richard W. (Palo Alto, CA)

    2012-05-15T23:59:59.000Z

    A gas separation process for treating exhaust gases from the combustion of gaseous fuels, and gaseous fuel combustion processes including such gas separation. The invention involves routing a first portion of the exhaust stream to a carbon dioxide capture step, while simultaneously flowing a second portion of the exhaust gas stream across the feed side of a membrane, flowing a sweep gas stream, usually air, across the permeate side, then passing the permeate/sweep gas back to the combustor.

  19. Uranium Removal from Groundwater via In Situ Biostimulation: Field-Scale Modeling of Transport and Biological Processes

    SciTech Connect (OSTI)

    Yabusaki, Steven B.; Fang, Yilin; Long, Philip E.; Resch, Charles T.; Peacock, Aaron D.; Komlos, John; Jaffe, Peter R.; Morrison, Stan J.; Dayvault, Richard; White, David C.; Anderson, Robert T.

    2007-03-12T23:59:59.000Z

    During 2002 and 2003, bioremediation experiments in the unconfined aquifer of the Old Rifle UMTRA field site in western Colorado provided evidence for the immobilization of hexavalent uranium in groundwater by iron-reducing Geobacter sp. stimulated by acetate amendment. As the bioavailable Fe(III) terminal electron acceptor was depleted in the zone just downgradient of the acetate injection gallery, sulfate-reducing organisms came to dominate the microbial community. In the present study, we use multicomponent reactive transport modeling to analyze data from the 2002 field experiment to 1) identify the dominant transport and biological processes controlling uranium mobility during biostimulation, 2) determine field-scale parameters for these modeled processes, and 3) apply the calibrated process models to history match observations during the 2003 field experiment. In spite of temporally and spatially variable observations during the field-scale biostimulation experiments, the coupled process simulation approach was able to establish a quantitative characterization of the principal flow, transport, and reaction processes that could be applied without modification to describe the 2003 field experiment. Insights gained from this analysis include field-scale estimates of bioavailable Fe(III) mineral, and the magnitude of uranium bioreduction during biostimulated growth of the iron-reducing and sulfate-reducing microorganisms.

  20. Minor actinide waste disposal in deep geological boreholes

    E-Print Network [OSTI]

    Sizer, Calvin Gregory

    2006-01-01T23:59:59.000Z

    The purpose of this investigation was to evaluate a waste canister design suitable for the disposal of vitrified minor actinide waste in deep geological boreholes using conventional oil/gas/geothermal drilling technology. ...

  1. 30th Actinide Separations Conference, PNNL-SA-50126

    SciTech Connect (OSTI)

    Delegard, Calvin H.

    2006-05-25T23:59:59.000Z

    Program booklet for the 30th Actinide Separations Conference. Contains agenda and abstracts for 27 poster and 38 oral presentations to be made during the 3-day meeting, May 23-25, 2006.

  2. actinide consumption nuclear: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    in this work, fission cross-sections on 233U, the main fissile isotope of the ThU fuel cycle, and on the minor actinides 241Am, 243Am and 245Cm have been analyzed. Data on...

  3. Final Technical Progress Report Long term risk from actinides in the environment: Modes of mobility

    SciTech Connect (OSTI)

    Thomas B. Kirchner

    2002-03-22T23:59:59.000Z

    The key source of uncertainty in assessing actinide mobility is the relative importance of transport by: (1) wind erosion, (2) water erosion, and (3) vertical migration. Each of these three processes depends on several environmental factors and they compete with one another. A scientific assessment of the long-term risks associated with actinides in surface soils depends on better quantifying each of these three modes of mobility. The objective from our EMSP study was to quantify the mobility of soil actinides by wind erosion, water erosion, and vertical migration at three semiarid sites where actinide mobility is a key technical, social and legal issue. This EMSP project was the first to evaluate all three factors at a site. The approach has been to investigate both short- and long-term issues based on field and lab studies and model comparisons. Our results demonstrate the importance of incorporating threshold responses into a modeling framework that accounts for environmental factors and natural disturbances that trigger large changes in actinide mobility. The study measured erosional losses of sediment and fallout cesium (an actinide analogue) from field plots located near WIPP in 1998. The results highlight the large effect of burning as a disturbance on contaminant transport and mobility via runoff and erosion. The results show that runoff, erosion, and actinide transport are (1) strongly site specific-differences in radionuclide transport between WIPP and Rocky Flats differed by a factor of twelve because of soil and vegetation differences, and (2) are strongly impacted by disturbances such as fire, which can increase runoff, erosion, and actinide transport by more than an order of magnitude. In addition, a laboratory experiment using soil columns was conducted to investigate the vertical transport of contaminants in sandy soils. Nine columns of soil collected from the vicinity of the WIPP site were prepared. The column consisted of a piece of PVC pipe 20 cm in diameter and approximately 22 cm long. A thin ''marker layer'' of white soil was added to the top of each column followed by a thin layer of soil that had been spiked with 137Cs, cerium and lanthanum was applied to the surface. Approximately 900 cm of water (the equivalent of about 30 years of rainfall) was then applied at a rate of 3.2 L d-1. All of the activity contained in the soil core appeared to be in the top few mm of soil, i.e. there was virtually no movement of the 134Cs labeled particles. Finally, a library of object-oriented model components was created using Visual Basic to support the construction of contaminant transport models. These components greatly simplify the task of building 1- to 3- dimensional simulation models for risk assessment. The model components created under this funding were subsequently applied to help answer questions regarding risks from irrigation associated with potential releases from the Yucca Mountain waste repository.

  4. Actinide Production in the Reaction of Heavy Ions withCurium-248

    SciTech Connect (OSTI)

    Moody, K.J.

    1983-07-01T23:59:59.000Z

    Chemical experiments were performed to examine the usefulness of heavy ion transfer reactions in producing new, neutron-rich actinide nuclides. A general quasi-elastic to deep-inelastic mechanism is proposed, and the utility of this method as opposed to other methods (e.g. complete fusion) is discussed. The relative merits of various techniques of actinide target synthesis are discussed. A description is given of a target system designed to remove the large amounts of heat generated by the passage of a heavy ion beam through matter, thereby maximizing the beam intensity which can be safely used in an experiment. Also described is a general separation scheme for the actinide elements from protactinium (Z = 91) to mendelevium (Z = 101), and fast specific procedures for plutonium, americium and berkelium. The cross sections for the production of several nuclides from the bombardment of {sup 248}Cm with {sup 18}O, {sup 86}Kr and {sup 136}Xe projectiles at several energies near and below the Coulomb barrier were determined. The results are compared with yields from {sup 48}Ca and {sup 238}U bombardments of {sup 248}Cm. Simple extrapolation of the product yields into unknown regions of charge and mass indicates that the use of heavy ion transfer reactions to produce new, neutron-rich above-target species is limited. The substantial production of neutron-rich below-target species, however, indicates that with very heavy ions like {sup 136}Xe and {sup 238}U the new species {sup 248}Am, {sup 249}Am and {sup 247}Pu should be produced with large cross sections from a {sup 248}Cm target. A preliminary, unsuccessful attempt to isolate {sup 247}Pu is outlined. The failure is probably due to the half life of the decay, which is calculated to be less than 3 minutes. The absolute gamma ray intensities from {sup 251}Bk decay, necessary for calculating the {sup 251}Bk cross section, are also determined.

  5. Method and apparatus for removing and preventing window deposition during photochemical vapor deposition (photo-CVD) processes

    DOE Patents [OSTI]

    Tsuo, S.; Langford, A.A.

    1989-03-28T23:59:59.000Z

    Unwanted build-up of the film deposited on the transparent light-transmitting window of a photochemical vacuum deposition (photo-CVD) chamber is eliminated by flowing an etchant into the part of the photolysis region in the chamber immediately adjacent the window and remote from the substrate and from the process gas inlet. The respective flows of the etchant and the process gas are balanced to confine the etchant reaction to the part of the photolysis region proximate to the window and remote from the substrate. The etchant is preferably one that etches film deposit on the window, does not etch or affect the window itself, and does not produce reaction by-products that are deleterious to either the desired film deposited on the substrate or to the photolysis reaction adjacent the substrate. 3 figs.

  6. Method and apparatus for removing and preventing window deposition during photochemical vapor deposition (photo-CVD) processes

    DOE Patents [OSTI]

    Tsuo, Simon (Lakewood, CO); Langford, Alison A. (Boulder, CO)

    1989-01-01T23:59:59.000Z

    Unwanted build-up of the film deposited on the transparent light-transmitting window of a photochemical vacuum deposition (photo-CVD) chamber is eliminated by flowing an etchant into the part of the photolysis region in the chamber immediately adjacent the window and remote from the substrate and from the process gas inlet. The respective flows of the etchant and the process gas are balanced to confine the etchant reaction to the part of the photolysis region proximate to the window and remote from the substrate. The etchant is preferably one that etches film deposit on the window, does not etch or affect the window itself, and does not produce reaction by-products that are deleterious to either the desired film deposited on the substrate or to the photolysis reaction adjacent the substrate.

  7. Comparative studies of actinide and sub-actinide fission cross section calculation from MCNP6 and TALYS

    SciTech Connect (OSTI)

    Perkasa, Y. S. [Department of Physics, Sunan Gunung Djati State Islamic University Bandung, Jl. A.H Nasution No. 105 Cibiru, Bandung (Indonesia); Waris, A., E-mail: awaris@fi.itb.ac.id; Kurniadi, R., E-mail: awaris@fi.itb.ac.id; Su'ud, Z., E-mail: awaris@fi.itb.ac.id [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa No. 10 Bandung 40132 (Indonesia)

    2014-09-30T23:59:59.000Z

    Comparative studies of actinide and sub-actinide fission cross section calculation from MCNP6 and TALYS have been conducted. In this work, fission cross section resulted from MCNP6 prediction will be compared with result from TALYS calculation. MCNP6 with its event generator CEM03.03 and LAQGSM03.03 have been validated and verified for several intermediate and heavy nuclides fission reaction data and also has a good agreement with experimental data for fission reaction that induced by photons, pions, and nucleons at energy from several ten of MeV to about 1 TeV. The calculation that induced within TALYS will be focused mainly to several hundred MeV for actinide and sub-actinide nuclides and will be compared with MCNP6 code and several experimental data from other evaluator.

  8. A process for containment removal and waste volume reduction to remediate groundwater containing certain radionuclides, toxic metals and organics. Final report

    SciTech Connect (OSTI)

    Buckley, L.P.; Killey, D.R.W.; Vijayan, S.; Wong, P.C.F. [Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.

    1992-09-01T23:59:59.000Z

    A project to remove groundwater contaminants by an improved treatment process was performed during 1990 October--1992 March by Atomic Energy of Canada Limited for the United States Department of Energy, managed by Argonne National Laboratory. The goal was to generate high-quality effluent while minimizing secondary waste volume. Two effluent target levels, within an order of magnitude, or less than the US Drinking Water Limit, were set to judge the process effectiveness. The program employed mixed waste feeds containing cadmium, uranium, lead, iron, calcium, strontium-85-90, cesium-137, benzene and trichlorethylene in simulated and actual groundwater and soil leachate solutions. A combination of process steps consisting of sequential chemical conditioning, cross-flow microfiltration and dewatering by low temperature-evaporation, or filter pressing were effective for the treatment of mixed waste having diverse physico-chemical properties. A simplified single-stage version of the process was implemented to treat ground and surface waters contaminated with strontium-90 at the Chalk River Laboratories site. Effluent targets and project goals were met successfully.

  9. Use of once-through treat gas to remove the heat of reaction in solvent hydrogenation processes

    DOE Patents [OSTI]

    Nizamoff, Alan J. (Convent Station, NJ)

    1980-01-01T23:59:59.000Z

    In a coal liquefaction process wherein feed coal is contacted with molecular hydrogen and a hydrogen-donor solvent in a liquefaction zone to form coal liquids and vapors and coal liquids in the solvent boiling range are thereafter hydrogenated to produce recycle solvent and liquid products, the improvement which comprises separating the effluent from the liquefaction zone into a hot vapor stream and a liquid stream; cooling the entire hot vapor stream sufficiently to condense vaporized liquid hydrocarbons; separating condensed liquid hydrocarbons from the cooled vapor; fractionating the liquid stream to produce coal liquids in the solvent boiling range; dividing the cooled vapor into at least two streams; passing the cooling vapors from one of the streams, the coal liquids in the solvent boiling range, and makeup hydrogen to a solvent hydrogenation zone, catalytically hydrogenating the coal liquids in the solvent boiling range and quenching the hydrogenation zone with cooled vapors from the other cooled vapor stream.

  10. An Assessment of Spent Fuel Reprocessing for Actinide Destruction and Resource Sustainability.

    SciTech Connect (OSTI)

    Cipiti, Benjamin B.; Smith, James D.

    2008-09-01T23:59:59.000Z

    The reprocessing and recycling of spent nuclear fuel can benefit the nuclear fuel cycle by destroying actinides or extending fissionable resources if uranium supplies become limited. The purpose of this study was to assess reprocessing and recycling in both fast and thermal reactors to determine the effectiveness for actinide destruction and resource utilization. Fast reactor recycling will reduce both the mass and heat load of actinides by a factor of 2, but only after 3 recycles and many decades. Thermal reactor recycling is similarly effective for reducing actinide mass, but the heat load will increase by a factor of 2. Economically recoverable reserves of uranium are estimated to sustain the current global fleet for the next 100 years, and undiscovered reserves and lower quality ores are estimated to contain twice the amount of economically recoverable reserves--which delays the concern of resource utilization for many decades. Economic analysis reveals that reprocessed plutonium will become competitive only when uranium prices rise to about %24360 per kg. Alternative uranium sources are estimated to be competitive well below that price. Decisions regarding the development of a near term commercial-scale reprocessing fuel cycle must partially take into account the effectiveness of reactors for actnides destruction and the time scale for when uranium supplies may become limited. Long-term research and development is recommended in order to make more dramatic improvements in actinide destruction and cost reductions for advanced fuel cycle technologies.The original scope of this work was to optimize an advanced fuel cycle using a tool that couples a reprocessing plant simulation model with a depletion analysis code. Due to funding and time constraints of the late start LDRD process and a lack of support for follow-on work, the project focused instead on a comparison of different reprocessing and recycling options. This optimization study led to new insight into the fuel cycle. AcknowledgementThe authors would like to acknowledge the support of Laboratory Directed Research and Development Project 125862 for funding this research.

  11. The technical and economic impact of minor actinide transmutation in a sodium fast reactor

    SciTech Connect (OSTI)

    Gautier, G. M.; Morin, F. [Alternative Energy and Atomic Energy Commission, CEA, DEN, F - 13108 St Paul lez Durance (France); Dechelette, F.; Sanseigne, E. [Alternative Energy and Atomic Energy Commission, CEA, DEN DTN, F - 13108 St Paul lez Durance (France); Chabert, C. [Alternative Energy and Atomic Energy Commission, CEA, DEN, F - 13108 St Paul lez Durance (France)

    2012-07-01T23:59:59.000Z

    Within the frame work of the French National Act of June 28, 2006 pertaining to the management of high activity, long-lived radioactive waste, one of the proposed processes consists in transmuting the Minor Actinides (MA) in the radial blankets of a Sodium Fast Reactor (SFR). With this option, we may assess the additional cost of the reactor by comparing two SFR designs, one with no Minor Actinides, and the other involving their transmutation. To perform this exercise, we define a reference design called SFRref, of 1500 MWe that is considered to be representative of the Reactor System. The SFRref mainly features a pool architecture with three pumps, six loops with one steam generator per loop. The reference core is the V2B core that was defined by the CEA a few years ago for the Reactor System. This architecture is designed to meet current safety requirements. In the case of transmutation, for this exercise we consider that the fertile blanket is replaced by two rows of assemblies having either 20% of Minor Actinides or 20% of Americium. The assessment work is performed in two phases. - The first consists in identifying and quantifying the technical differences between the two designs: the reference design without Minor Actinides and the design with Minor Actinides. The main differences are located in the reactor vessel, in the fuel handling system and in the intermediate storage area for spent fuel. An assessment of the availability is also performed so that the impact of the transmutation can be known. - The second consists in making an economic appraisal of the two designs. This work is performed using the CEA's SEMER code. The economic results are shown in relative values. For a transmutation of 20% of MA in the assemblies (S/As) and a hypothesis of 4 kW allowable for the washing device, there is a large external storage demanding a very long cooling time of the S/As. In this case, the economic impact may reach 5% on the capital part of the Levelized Unit Electricity Cost (LUEC). A diminished concentration at 10% of MA, reduces the size of the external storage and the cooling time of the assemblies becomes compatible with the management of the irradiated fuel. Even with a low allowable power for the washing device, the economic impact on the capital cost is less than 2.5%. (authors)

  12. REVIEW OF EXPERIMENTAL STUDIES INVESTIGATING THE RATE OF STRONTIUM AND ACTINIDE ADSORPTION BY MONOSODIUM TITANATE

    SciTech Connect (OSTI)

    Hobbs, D.

    2010-10-01T23:59:59.000Z

    A number of laboratory studies have been conducted to determine the influence of mixing and mixing intensity, solution ionic strength, initial sorbate concentrations, temperature, and monosodium titanate (MST) concentration on the rates of sorbate removal by MST in high-level nuclear waste solutions. Of these parameters, initial sorbate concentrations, ionic strength, and MST concentration have the greater impact on sorbate removal rates. The lack of a significant influence of mixing and mixing intensity on sorbate removal rates indicates that bulk solution transport is not the rate controlling step in the removal of strontium and actinides over the range of conditions and laboratory-scales investigated. However, bulk solution transport may be a significant parameter upon use of MST in a 1.3 million-gallon waste tank such as that planned for the Small Column Ion Exchange (SCIX) program. Thus, Savannah River National Laboratory (SRNL) recommends completing the experiments in progress to determine if mixing intensity influences sorption rates under conditions appropriate for this program. Adsorption models have been developed from these experimental studies that allow prediction of strontium (Sr), plutonium (Pu), neptunium (Np) and uranium (U) concentrations as a function of contact time with MST. Fairly good agreement has been observed between the predicted and measured sorbate concentrations in the laboratory-scale experiments.

  13. Waste Minimization Study on Pyrochemical Reprocessing Processes

    SciTech Connect (OSTI)

    Boussier, H.; Conocar, O.; Lacquement, J. [CEA/DEN Valrho Marcoule/DRCP/SCPS/Pyrochemical Processes Laboratory, BP 17171 30207 Bagnols-sur-Ceze (France)

    2006-07-01T23:59:59.000Z

    Ideally a new pyro-process should not generate more waste, and should be at least as safe and cost effective as the hydrometallurgical processes currently implemented at industrial scale. This paper describes the thought process, the methodology and some results obtained by process integration studies to devise potential pyro-processes and to assess their capability of achieving this challenging objective. As example the assessment of a process based on salt/metal reductive extraction, designed for the reprocessing of Generation IV carbide spent fuels, is developed. Salt/metal reductive extraction uses the capability of some metals, aluminum in this case, to selectively reduce actinide fluorides previously dissolved in a fluoride salt bath. The reduced actinides enter the metal phase from which they are subsequently recovered; the fission products remain in the salt phase. In fact, the process is not so simple, as it requires upstream and downstream subsidiary steps. All these process steps generate secondary waste flows representing sources of actinide leakage and/or FP discharge. In aqueous processes the main solvent (nitric acid solution) has a low boiling point and evaporate easily or can be removed by distillation, thereby leaving limited flow containing the dissolved substance behind to be incorporated in a confinement matrix. From the point of view of waste generation, one main handicap of molten salt processes, is that the saline phase (fluoride in our case) used as solvent is of same nature than the solutes (radionuclides fluorides) and has a quite high boiling point. So it is not so easy, than it is with aqueous solutions, to separate solvent and solutes in order to confine only radioactive material and limit the final waste flows. Starting from the initial block diagram devised two years ago, the paper shows how process integration studies were able to propose process fittings which lead to a reduction of the waste variety and flows leading at an 'ideal' new block diagram allowing internal solvent recycling, and self eliminating reactants. This new flowsheet minimizes the quantity of inactive inlet flows that would have inevitably to be incorporated in a final waste form. The study identifies all knowledge gaps to be filled and suggest some possible R and D issues to confirm or infirm the feasibility of the proposed process fittings. (authors)

  14. Demonstration of Small Tank Tetraphenylborate Precipitation Process Using Savannah River Site High Level Waste

    SciTech Connect (OSTI)

    Peters, T.B.

    2001-09-10T23:59:59.000Z

    This report details the experimental effort to demonstrate the continuous precipitation of cesium from Savannah River Site High Level Waste using sodium tetraphenylborate. In addition, the experiments examined the removal of strontium and various actinides through addition of monosodium titanate.

  15. Actinide destruction and power peaking analysis in a 1000 MWt advanced burner reactor using moderated heterogeneous target assemblies

    SciTech Connect (OSTI)

    Kenneth Allen; Travis Knight; Samuel Bays

    2011-05-01T23:59:59.000Z

    The purpose of this research was to determine the effect of moderated heterogeneous subassemblies located in the core of a sodium-cooled fast reactor on minor actinide (MA) destruction rates over the lifecycle of the core. Additionally, particular emphasis was placed on the power peaking of the pins and the assemblies with the moderated targets as compared to standard unmoderated heterogeneous targets and a core without MA targets present. Power peaking analysis was performed on the target assemblies and on the fuel assemblies adjacent to the targets. The moderated subassemblies had a marked improvement in the overall destruction of heavy metals in the targets. The design with acceptable power peaking results had a 12.25% greater destruction of heavy metals than a similar ex-core unmoderated assembly. The increase in minor actinide destruction was most evident with americium where the moderated assemblies reduced the initial amount to less than 3% of the initial loading over a period of five years core residency. In order to take advantage of the high minor actinide destruction and minimize the power peaking effects, a hybrid scenario was devised where the targets resided ex-core in a moderated assembly for the first 506.9 effective full power days (EFPDs) and were moved to an in-core arrangement with the moderated targets removed for the remainder of the lifecycle. The hybrid model had an assembly and pin power peaking of less than 2.0 and a higher heavy metal and minor actinide destruction rate than the standard unmoderated heterogeneous targets either in-core or ex-core. The hybrid model has a 54.5% greater Am reduction over the standard ex-core model. It also had a 27.8% greater production of Cm and a 41.5% greater production of Pu than the standard ex-core model. The radiotoxicity of the targets in the hybrid design was 20% less than the discharged standard ex-core targets.

  16. Thermally unstable complexants: Stability of lanthanide/actinide complexes, thermal instability of the ligands, and applications in actinide separations

    SciTech Connect (OSTI)

    Nash, K.L.; Rickert, P.G.

    1991-01-01T23:59:59.000Z

    Water soluble complexing agents are commonly used in separations to enhance the selectivity of both ion exchange and solvent extraction processes. Applications of this type in the treatment of nuclear wastes using conventional complexing agents have found mixed success due to the nature of the complexants. In addition, the residual solutions containing these species have led to potentially serious complications in waste storage. To overcome some of the limitations of carboxylic acid and aminopolycarboxylate ligands, we have initiated a program to investigate the complexing ability, thermal/oxidative instability, and separation potential of a group of water soluble organophosphorus compounds which we call Thermally Unstable Complexants, or simply TUCS. Complexants of this type appear to be superior to conventional analogues in a number of respects. In this report, we will summarize our research to date on the actinide/lanthanide complexes with a series of substituted methanediphosphonic acids, the kinetics of their oxidative decomposition, and a few applications which have been developed for their use. 17 refs., 5 figs., 3 tab.

  17. E-Print Network 3.0 - actinide standard ii-iii Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    of a-U (*) Summary: and costly. Thus, while we were able to study the light and heavy actinide metals 1 in their cubic (high... applied to the study of some actinide...

  18. E-Print Network 3.0 - actinide compound ufe Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    49, dkcembre 1988 Summary: and actinide Laves phases. In comparison with RFE2 (R Rare earth) Laves phase compounds CeFe2 exhibits... M2 (Ac actinides) 5demon- strates that...

  19. E-Print Network 3.0 - actinide system inconsistencies Sample...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    of actinide metals Summary: heats of actinide metals M. J. Mortimer Chemistry Division, AERE Harwell, Didcot, Oxon, OX11 ORA, G... 'volution de ces grandeurs ainsi que celle de...

  20. E-Print Network 3.0 - actinides exposure review Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    of actinide metals Summary: heats of actinide metals M. J. Mortimer Chemistry Division, AERE Harwell, Didcot, Oxon, OX11 ORA, G... 'volution de ces grandeurs ainsi que celle de...

  1. E-Print Network 3.0 - actinide alloys Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    of actinide metals Summary: heats of actinide metals M. J. Mortimer Chemistry Division, AERE Harwell, Didcot, Oxon, OX11 ORA, G... 'volution de ces grandeurs ainsi que celle de...

  2. E-Print Network 3.0 - actinide chemistry Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Powered by Explorit Topic List Advanced Search Sample search results for: actinide chemistry Page: << < 1 2 3 4 5 > >> 1 www.emsl.pnl.gov ACTINIDE CHEMISTRY MEETS COMPUTATION...

  3. E-Print Network 3.0 - actinide complexes Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    (04-400) Summary: of the electronic properties and reactions of actinide and transition metal complexes. Hay, who is a Laboratory... as a leader in actinide chemistry, Burns was...

  4. E-Print Network 3.0 - actinide decay series Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    page C2-841 Summary: in the actinide series whereas the acti 1 es up to Pu metal are transition metal likeq" : The f states in Am... aux actinides metalliques et aux oxydes sont...

  5. E-Print Network 3.0 - actinide complexing agent Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    (04-400) Summary: of the electronic properties and reactions of actinide and transition metal complexes. Hay, who is a Laboratory... as a leader in actinide chemistry, Burns was...

  6. Sigma Team for Minor Actinide Separation: PNNL FY 2011 Status Report

    SciTech Connect (OSTI)

    Lumetta, Gregg J.; Braley, Jenifer C.; Sinkov, Sergey I.; Levitskaia, Tatiana G.; Carter, Jennifer C.; Warner, Marvin G.; Pittman, Jonathan W.

    2011-08-13T23:59:59.000Z

    This report summarizes work conducted in FY 2011 at PNNL to investigate new methods of separating the minor actinide elements (Am and Cm) from the trivalent lanthanide elements, and separation of Am from Cm. For the former, work focused on a solvent extraction system combining an acidic extractant (HDEHP) with a neutral extractant (CMPO) to form a hybrid solvent extraction system referred to as TRUSPEAK (combining the TRUEX and TALSPEAK processes). For the latter, ligands that strongly bing uranyl ion were investigated for stabilizing corresponding americyl ion.

  7. Evaluation of improved technologies for the removal of {sup 90}Sr and {sup 137}Cs from process wastewater and groundwater: FY 1995 status

    SciTech Connect (OSTI)

    Bostick, D.T.; Arnold, W.D. Jr.; Burgess, M.W.; McTaggart, D.R.; Taylor, P.A. [Oak Ridge National Lab., TN (United States); Guo, B. [Oak Ridge Research Inst., TN (United States)

    1996-03-01T23:59:59.000Z

    A number of new sorbents are currently being developed for the removal of {sup 90}Sr and {sup 137}Cs from contaminated, caustic low-level liquid waste (LLLW). These sorbents are potentially promising for use in the cleanup of contaminated groundwater and process wastewater containing the two radionuclides. The goal of this subtask is to evaluate the new sorbents to determine whether their associated treatment technology is more selective for the decontamination of wastewater streams than that of currently available processes. Activities during fiscal year 1995 have included completing the characterization of the standard treatment technology, ion exchange on chabazite zeolite. Strontium and cesium sorption on sodium-modified zeolite was observed in the presence of elevated concentrations of wastewater components: sodium, potassium, magnesium, and calcium. The most significant loss of nuclide sorption was noted in the first 0- to 4-meq/L addition of the cations to a wastewater simulant. Radionuclide sorption on the pretreated zeolite was also determined under dynamic flow conditions. Resorcinol-formaldehyde (R-F) resin, which was developed at the Savannah River Site, was selected as the first new sorbent to be evaluated for wastewater treatment. Nuclide sorption on this resin was greater when the resin had been washed with ultrapure water and air dried prior to use.

  8. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Timothy A. Hyde

    2012-06-01T23:59:59.000Z

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  9. Actinide (III) solubility in WIPP Brine: data summary and recommendations

    SciTech Connect (OSTI)

    Borkowski, Marian; Lucchini, Jean-Francois; Richmann, Michael K.; Reed, Donald T.

    2009-09-01T23:59:59.000Z

    The solubility of actinides in the +3 oxidation state is an important input into the Waste Isolation Pilot Plant (WIPP) performance assessment (PA) models that calculate potential actinide release from the WIPP repository. In this context, the solubility of neodymium(III) was determined as a function of pH, carbonate concentration, and WIPP brine composition. Additionally, we conducted a literature review on the solubility of +3 actinides under WIPP-related conditions. Neodymium(III) was used as a redox-invariant analog for the +3 oxidation state of americium and plutonium, which is the oxidation state that accounts for over 90% of the potential release from the WIPP through the dissolved brine release (DBR) mechanism, based on current WIPP performance assessment assumptions. These solubility data extend past studies to brine compositions that are more WIPP-relevant and cover a broader range of experimental conditions than past studies.

  10. actinide residue processing: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    The high power densities expected for the MIT microengine (silicon MEMS-based micro-gas turbine generator) require the turbine and compressor spool to rotate at a very high...

  11. Physics studies of higher actinide consumption in an LMR

    SciTech Connect (OSTI)

    Hill, R.N.; Wade, D.C.; Fujita, E.K.; Khalil, H.S.

    1990-01-01T23:59:59.000Z

    The core physics aspects of the transuranic burning potential of the Integral Fast Reactor (IFR) are assessed. The actinide behavior in fissile self-sufficient IFR closed cycles of 1200 MWt size is characterized, and the transuranic isotopics and risk potential of the working inventory are compared to those from a once-through LWR. The core neutronic performance effects of rare-earth impurities present in the recycled fuel are addressed. Fuel cycle strategies for burning transuranics from an external source are discussed, and specialized actinide burner designs are described. 4 refs., 4 figs., 3 tabs.

  12. Process for the recovery of curium-244 from nuclear waste

    SciTech Connect (OSTI)

    Posey, J.C.

    1980-10-01T23:59:59.000Z

    A process has been designed for the recovery of curium from purex waste. Curium and americium are separated from the lanthanides by a TALSPEAK extraction process using differential extraction. Equations were derived for the estimation of the economically optimum conditions for the extraction using laboratory batch extraction data. The preparation of feed for the extraction involves the removal of nitric acid from the Purex waste by vaporization under reduced pressure, the leaching of soluble nitrates from the resulting cake, and the oxalate precipitation of a pure lanthanide-actinide fraction. Final separation of the curium from americium is done by ion-exchange. The steps of the process, except ion-exchange, were tested on a laboratory scale and workable conditions were determined.

  13. Process for removing a nitrogen gas from mixture comprising N/sub 2/ and CO or CO/sub 2/ and CO

    SciTech Connect (OSTI)

    Matsui, S.; Hayashi, S.; Kumagai, M.; Tukahara, Y.

    1984-08-28T23:59:59.000Z

    Processes are disclosed for removing N/sub 2/ from a feed gas comprising CO+N/sub 2/ or CO, CO/sub 2/+N/sub 2/ through PSA by using at least two adsorption columns containing an adsorbent exhibiting selective adsorb property to carbon monoxide which comprises: a step of pressurizing an adsorption column by the feed gas; a step of introducing the feed gas into the adsorption column, in which step was previously completed, so as to adsorb CO, or CO+CO/sub 2/ on or in the adsorbent; a step of connecting the adsorption column, in which step was previously completed, to the other adsorption column in which step was previously completed, to reduce the pressure in the former adsorption column to one atmosphere or a pressure close to it; a step of purging nitrogen by passing product gas through the adsorption column; a step of desorbing carbon monoxide adsorbed on or in the adsorbent of the adsorption column, by vacuum pump to recover a product gas; and a step of a connecting the adsorption column, in which step was previously completed, to the other adsorption column in which step was previously completed to increase pressure in the former column, periodically switching the flow between or among said adsorption columns so as to repeat the above steps in all the adsorption columns.

  14. Vertical Extraction Process Implemented at the 118-K-1 Burial Ground for Removal of Irradiated Reactor Debris from Silo Structures - 12431

    SciTech Connect (OSTI)

    Teachout, Douglas B. [Vista Engineering Technologies, LLC, Richland, Washington, 99352 (United States); Adamson, Clinton J.; Zacharias, Ames [Washington Closure Hanford, LLC, Richland, Washington, 99352 (United States)

    2012-07-01T23:59:59.000Z

    The primary objective of a remediation project is the safe extraction and disposition of diverse waste forms and materials. Remediation of a solid waste burial ground containing reactor hardware and irradiated debris involves handling waste with the potential to expose workers to significantly elevated dose rates. Therefore, a major challenge confronted by any remediation project is developing work processes that facilitate compliant waste management practices while at the same time implementing controls to protect personnel. Traditional burial ground remediation is accomplished using standard excavators to remove materials from trenches and other excavation configurations often times with minimal knowledge of waste that will be encountered at a specific location. In the case of the 118-K-1 burial ground the isotopic activity postulated in historic documents to be contained in vertical cylindrical silos was sufficient to create the potential for a significant radiation hazard to project personnel. Additionally, certain reported waste forms posed an unacceptably high potential to contaminate the surrounding environment and/or workers. Based on process knowledge, waste management requirements, historic document review, and a lack of characterization data it was determined that traditional excavation techniques applied to remediation of vertical silos would expose workers to unacceptable risk. The challenging task for the 118-K-1 burial ground remediation project team then became defining an acceptable replacement technology or modification of an existing technology to complete the silo remediation. Early characterization data provided a good tool for evaluating the location of potential high exposure rate items in the silos. Quantitative characterization was a different case and proved difficult because of the large diameter of the silos and the potential for variable density of attenuating soils and waste forms in the silo. Consequently, the most relevant information supporting job planning and understanding of the conditions was the data obtained from the gross gamma meter that was inserted into each casing to provide a rough estimate of dose rates in the tubes. No added value was realized in attempting to quantify the source term and/or associate the isotopic activity with a particular actual waste form (e.g., sludge). Implementing the WRM system allowed monitoring of worker and boundary exposure rates from a distance, maintaining compliance with ALARA principles. This system also provided the project team early knowledge of items being removed that had high exposure rates associated with them, thus creating an efficient method of acknowledging an issue and arriving at a solution prior to having an upset condition. An electronic dosimeter with telemetry capability replaced the excavator mounted AMP-100 system approximately half way through remediation of the silos. Much higher connectivity efficiency was derived from this configuration. Increasing the data feed efficiency additionally led to less interruption of the remediation effort. Early in system testing process a process handicap on the excavator operator was acknowledged. A loss of depth perception resulted when maneuvering the excavator and bucket using the camera feed to an in-cab monitor. Considerable practice and mock-up testing allowed this handicap to be overcome. The most significant equipment failures involved the cable connection to the camera mounted between the clamshell bucket jaws and the video splitter in the excavator cab. Rotation of the clamshell bucket was identified as the cause of cable connection failures because of the cyclic twisting motion and continuous mechanical jarring of the connection. In-cab vibration was identified as the culprit in causing connection failures of the video splitter. While these failures were repaired, substantial production time was lost. Ultimately, the decision was made to purchase a second cable and higher quality video splitter eliminate the down time. An engineering improvement for future operations would be i

  15. Method for recovery of actinides from actinide-bearing scrap and waste nuclear material using O/sub 2/F/sub 2/

    DOE Patents [OSTI]

    Asprey, L.B.; Eller, P.G.

    1984-09-12T23:59:59.000Z

    Method for recovery of actinides from nuclear waste material containing sintered and other oxides thereof and from scrap materials containing the metal actinides using O/sub 2/F/sub 2/ to generate the hexafluorides of the actinides present therein. The fluorinating agent, O/sub 2/F/sub 2/, has been observed to perform the above-described tasks at sufficiently low temperatures that there is virtually no damage to the containment vessels. Moreover, the resulting actinide hexafluorides are not detroyed by high temperature reactions with the walls of the reaction vessel. Dioxygen difluoride is readily prepared, stored and transferred to the place of reaction.

  16. Technical and economic assessment of different options for minor actinide transmutation: the French case

    SciTech Connect (OSTI)

    Chabert, C.; Coquelet-Pascal, C. [CEA-Cadarache, DEN, Saint-Paul-lez-Durance (France); Saturnin, A. [CEA, DEN, Marcoule (France); Mathonniere, G.; Boullis, B.; Warin, D. [CEA-Saclay, DEN, Gif-sur-Yvette (France); Van Den Durpel, L. [AREVA-NC, Paris-la-Defense (France); Caron-Charles, M. [AREVA-NP, Paris-la-Defense (France); Garzenne, C. [EDF, Paris (France)

    2013-07-01T23:59:59.000Z

    Studies have been performed to assess the industrial perspectives of partitioning and transmutation of long-lived elements. These studies were carried out in tight connection with GEN-IV systems development. The results include the technical and economic evaluation of fuel cycle scenarios along with different options for optimizing the processes between the minor actinide transmutation in fast neutron reactors, their interim storage and geological disposal of ultimate waste. The results are analysed through several criteria (impacts on waste, on waste repository, on fuel cycle plants, on radiological exposure of workers, on costs and on industrial risks). These scenario evaluations take place in the French context which considers the deployment of the first Sodium-cooled Fast Reactor (SFR) in 2040. 3 management options of minor actinides have been studied: no transmutation, transmutation in SFR and transmutation in an accelerator-driven system (ADS). Concerning economics the study shows that the cost overrun related to the transmutation process could vary between 5 to 9% in SFR and 26 % in the case of ADS.

  17. Determining the dissolution rates of actinide glasses: A time and temperature Product Consistency Test study

    SciTech Connect (OSTI)

    Daniel, W.E.; Best, D.R.

    1995-12-01T23:59:59.000Z

    Vitrification has been identified as one potential option for the e materials such as Americium (Am), Curium (Cm), Neptunium (Np), and Plutonium (Pu). A process is being developed at the Savannah River Site to safely vitrify all of the highly radioactive Am/Cm material and a portion of the fissile (Pu) actinide materials stored on site. Vitrification of the Am/Cm will allow the material to be transported and easily stored at the Oak Ridge National Laboratory. The Am/Cm glass has been specifically designed to be (1) highly durable in aqueous environments and (2) selectively attacked by nitric acid to allow recovery of the valuable Am and Cm isotopes. A similar glass composition will allow for safe storage of surplus plutonium. This paper will address the composition, relative durability, and dissolution rate characteristics of the actinide glass, Loeffler Target, that will be used in the Americium/Curium Vitrification Project at Westinghouse Savannah River Company near Aiken, South Carolina. The first part discusses the tests performed on the Loeffler Target Glass concerning instantaneous dissolution rates. The second part presents information concerning pseudo-activation energy for the one week glass dissolution process.

  18. Review Article: The Effects of Radiation Chemistry on Solvent Extraction 3: A Review of Actinide and Lanthanide Extraction

    SciTech Connect (OSTI)

    Bruce J. Mincher; Giuseppe Modolo; Stephen P. Mezyk

    2009-12-01T23:59:59.000Z

    The partitioning of the long-lived ?-emitters and the high-yield fission products from dissolved nuclear fuel is a key component of processes envisioned for the safe recycling of nuclear fuel and the disposition of high-level waste. These future processes will likely be based on aqueous solvent extraction technologies for light water reactor fuel and consist of four main components for the sequential separation of uranium, fission products, group trivalent actinides and lanthanides, and then trivalent actinides from lanthanides. Since the solvent systems will be in contact with highly radioactive solutions, they must be robust toward radiolytic degradation in an irradiated mixed organic, aqueous acidic environment. Therefore, an understanding of their radiation chemistry is important to the design of a practical system. In the first paper in this series we reviewed the radiation chemistry of irradiated aqueous nitric acid and the tributyl phosphate ligand for uranium extraction in the first step of these extractions. In the second, we reviewed the radiation chemistry of the ligands proposed for use in the extraction of cesium and strontium fission products. Here, we review the radiation chemistry of the ligands that might be used in the third step in the series of separations, for the group extraction of the lanthanides and actinides. This includes traditional organophosphorous reagents such as CMPO and HDEHP, as well as novel reagents such as the amides and diamides currently being investigated.

  19. actinide materials annual: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    actinide materials annual First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Molecular dynamics simulation...

  20. Delayed neutron measurements from fast fission of actinide waste isotopes

    E-Print Network [OSTI]

    Charlton, William S.

    2012-06-07T23:59:59.000Z

    , was suggested which would yield a superior fit to the measured data. A series of measurements were performed to test the hypothesis suggested by this alternate group structure. Using a set of highly purified actinide samples (provided by Oak Ridge National...

  1. actinides loading optimization: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    actinides loading optimization First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Building load control...

  2. Ab Initio Enhanced calphad Modeling of Actinide-Rich Nuclear Fuels

    SciTech Connect (OSTI)

    Morgan, Dane [Univ. of Wisconsin, Madison, WI (United States); Yang, Yong Austin [Univ. of Wisconsin, Madison, WI (United States)

    2013-10-28T23:59:59.000Z

    The process of fuel recycling is central to the Advanced Fuel Cycle Initiative (AFCI), where plutonium and the minor actinides (MA) Am, Np, and Cm are extracted from spent fuel and fabricated into new fuel for a fast reactor. Metallic alloys of U-Pu-Zr-MA are leading candidates for fast reactor fuels and are the current basis for fast spectrum metal fuels in a fully recycled closed fuel cycle. Safe and optimal use of these fuels will require knowledge of their multicomponent phase stability and thermodynamics (Gibbs free energies). In additional to their use as nuclear fuels, U-Pu-Zr-MA contain elements and alloy phases that pose fundamental questions about electronic structure and energetics at the forefront of modern many-body electron theory. This project will validate state-of-the-art electronic structure approaches for these alloys and use the resulting energetics to model U-Pu-Zr-MA phase stability. In order to keep the work scope practical, researchers will focus on only U-Pu-Zr-{Np,Am}, leaving Cm for later study. The overall objectives of this project are to: Provide a thermodynamic model for U-Pu-Zr-MA for improving and controlling reactor fuels; and, Develop and validate an ab initio approach for predicting actinide alloy energetics for thermodynamic modeling.

  3. Strategic Design and Optimization of Inorganic Sorbents for Cesium, Strontium and Actinides

    SciTech Connect (OSTI)

    Clearfield, Abraham

    2005-07-01T23:59:59.000Z

    It has been determined that poorly crystalline CST and SNT prepared at low temperature (100-150 C) exhibit much faster kinetics in uptake of Sr2+. In-situ X-ray studies has shown that SNT is a precursor phase to the formation of CST. It is possible to form mixtures of CST and SNT in a single reactant mix by control of temperature and time of reaction. It has been found that addition of a small amount of Cs+ to the reactant mix for the preparation of Nb-CST allows formation of the crystals in one day rather than ten days at 200 C. These discoveries suggest that a proper mix of sorbents (SNT, CST, Nb-CST) can be made easily at low cost that would remove all the HLW at the Savannah River site with a single in-tank procedure. The basic science goal in this project is to identify structure/affinity relationships for selected radionuclides and existing sorbents. The research will then apply this knowledge to the design and synthesis of sorbents that will exhibit increased cesium, strontium and actinide removal. The target problem focuses on the treatment of high-level nuclear wastes. The general approach can likewise be applied to non-radioactive separations.

  4. Strategic Design and Optimization of Inorganic Sorbents for Cesium, Strontium and Actinides

    SciTech Connect (OSTI)

    Clearfield, Abraham

    2005-07-01T23:59:59.000Z

    It has been determined that poorly crystalline CST and SNT prepared at low temperature (100-150 deg. C) exhibit much faster kinetics in uptake of Sr2+. 2. In-situ X-ray studies has shown that SNT is a precursor phase to the formation of CST. 3. It is possible to form mixtures of CST and SNT in a single reactant mix by control of temperature and time of reaction. 4. It has been found that addition of a small amount of Cs+ to the reactant mix for the preparation of Nb-CST allows formation of the crystals in one day rather than ten days at 200 deg. C. 5. These discoveries suggest that a proper mix of sorbents (SNT, CST, Nb-CST) can be made easily at low cost that would remove all the HLW at the Savannah River site with a single in-tank procedure. Research Objective The basic science goal in this project is to identify structure/affinity relationships for selected radionuclides and existing sorbents. The research will then apply this knowledge to the design and synthesis of sorbents that will exhibit increased cesium, strontium and actinide removal. The target problem focuses on the treatment of high-level nuclear wastes. The general approach can likewise be applied to non-radioactive separations.

  5. Removal and recovery of radionuclides and toxic metals from wastes, soils and materials

    SciTech Connect (OSTI)

    Francis, A.J.

    1993-07-01T23:59:59.000Z

    A process has been developed at Brookhaven National Laboratory (BNL) for the removal of metals and radionuclides from contaminated materials, soils, and waste sites (Figure 1). In this process, citric acid, a naturally occurring organic complexing agent, is used to extract metals such as Ba, Cd, Cr, Ni, Zn, and radionuclides Co, Sr, Th, and U from solid wastes by formation of water soluble, metal-citrate complexes. Citric acid forms different types of complexes with the transition metals and actinides, and may involve formation of a bidentate, tridentate, binuclear, or polynuclear complex species. The extract containing radionuclide/metal complex is then subjected to microbiological degradation followed by photochemical degradation under aerobic conditions. Several metal citrate complexes are biodegraded and the metals are recovered in a concentrated form with the bacterial biomass. Uranium forms binuclear complex with citric acid and is not biodegraded. The supernatant containing uranium citrate complex is separated and upon exposure to light, undergoes rapid degradation resulting in the formation of an insoluble, stable polymeric form of uranium. Uranium is recovered as a precipitate (uranium trioxide) in a concentrated form for recycling or for appropriate disposal. This treatment process, unlike others which use caustic reagents, does not create additional hazardous wastes for disposal and causes little damage to soil which can then be returned to normal use.

  6. Geothermal hydrogen sulfide removal

    SciTech Connect (OSTI)

    Urban, P.

    1981-04-01T23:59:59.000Z

    UOP Sulfox technology successfully removed 500 ppM hydrogen sulfide from simulated mixed phase geothermal waters. The Sulfox process involves air oxidation of hydrogen sulfide using a fixed catalyst bed. The catalyst activity remained stable throughout the life of the program. The product stream composition was selected by controlling pH; low pH favored elemental sulfur, while high pH favored water soluble sulfate and thiosulfate. Operation with liquid water present assured full catalytic activity. Dissolved salts reduced catalyst activity somewhat. Application of Sulfox technology to geothermal waters resulted in a straightforward process. There were no requirements for auxiliary processes such as a chemical plant. Application of the process to various types of geothermal waters is discussed and plans for a field test pilot plant and a schedule for commercialization are outlined.

  7. Biomimetic Actinide Chelators: An Update on the Preclinical Development of the Orally Active Hydroxypyridonate Decorporation Agents 3,4,3-LI(1,2-HOPO) and 5-LIO(Me-3,2-HOPO)

    SciTech Connect (OSTI)

    Durbin, Patricia W.; Kullgren, Birgitta; Ebbe, Shirley N.; Xu, Jide; Chang, Polly Y.; Bunin, Deborah I.; Blakely, Eleanor A.; Bjornstad, Kathleen A.; Rosen, Chris J.; Shuh, David K.; Raymond, Kenneth N.

    2011-07-13T23:59:59.000Z

    The threat of a dirty bomb or other major radiological contamination presents a danger of large-scale radiation exposure of the population. Because major components of such contamination are likely to be actinides, actinide decorporation treatments that will reduce radiation exposure must be a priority. Current therapies for the treatment of radionuclide contamination are limited and extensive efforts must be dedicated to the development of therapeutic, orally bioavailable, actinide chelators for emergency medical use. Using a biomimetic approach based on the similar biochemical properties of plutonium(IV) and iron(III), siderophore-inspired multidentate hydroxypyridonate ligands have been designed and are unrivaled in terms of actinide-affinity, selectivity, and efficiency. A perspective on the preclinical development of two hydroxypyridonate actinide decorporation agents, 3,4,3-LI(1,2-HOPO) and 5-LIO(Me-3,2-HOPO), is presented. The chemical syntheses of both candidate compounds have been optimized for scale-up. Baseline preparation and analytical methods suitable for manufacturing large amounts have been established. Both ligands show much higher actinide-removal efficacy than the currently approved agent, diethylenetriaminepentaacetic acid (DTPA), with different selectivity for the tested isotopes of plutonium, americium, uranium and neptunium. No toxicity is observed in cells derived from three different human tissue sources treated in vitro up to ligand concentrations of 1 mM, and both ligands were well tolerated in rats when orally administered daily at high doses (>100 micromol kg d) over 28 d under good laboratory practice guidelines. Both compounds are on an accelerated development pathway towards clinical use.

  8. Development of a remote bushing for actinide vitrification

    SciTech Connect (OSTI)

    Schumacher, R.F.; Ramsey, W.G.; Johnson, F.M. [and others

    1996-12-31T23:59:59.000Z

    The Savannah River Site (SRS) and the Savannah River Technology Center (SRTC) are combining their existing experience in handling highly radioactive, special nuclear materials with commercial glass fiberization technology in order to assemble a small vitrification system for radioactive actinide solutions. The vitrification system or {open_quotes}brushing{close_quotes}, is fabricated from platinum-rhodium alloy and is based on early marble remelt fiberization technology. Advantages of this unique system include its relatively small size, reliable operation, geometrical safety (nuclear criticality), and high temperature capability. The bushing design should be capable of vitrifying a number of the actinide nuclear materials, including solutions of americium/curium, neptunium, and possibly plutonium. State of the art, mathematical and oil model studies are being combined with basic engineering evaluations to verify and improve the thermal and mechanical design concepts.

  9. Minor Actinides Transmutation Scenario Studies in PWR with Innovative Fuels

    SciTech Connect (OSTI)

    Grouiller, J. P.; Boucher, L.; Golfier, H.; Dolci, F.; Vasile, A.; Youinou, G.

    2003-02-26T23:59:59.000Z

    With the innovative fuels (CORAIL, APA, MIX, MOX-UE) in current PWRs, it is theoretically possible to obtain different plutonium and minor actinides transmutation scenarios, in homogeneous mode, with a significant reduction of the waste radio-toxicity inventory and of the thermal output of the high level waste. Regarding each minor actinide element transmutation in PWRs, conclusions are : neptunium : a solution exists but the gain on the waste radio-toxicity inventory is not significant, americium : a solution exists but it is necessary to transmute americium with curium to obtain a significant gain, curium: Cm244 has a large impact on radiation and residual power in the fuel cycle; a solution remains to be found, maybe separating it and keeping it in interim storage for decay into Pu240 able to be transmuted in reactor.

  10. Determination of actinides in urine and fecal samples

    DOE Patents [OSTI]

    McKibbin, Terry T. (Larimer County, CO)

    1993-01-01T23:59:59.000Z

    A method of determining the radioactivity of specific actinides that are carried in urine or fecal sample material is disclosed. The samples are ashed in a muffle furnace, dissolved in an acid, and then treated in a series of steps of reduction, oxidation, dissolution, and precipitation, including a unique step of passing a solution through a chloride form anion exchange resin for separation of uranium and plutonium from americium.

  11. Determination of actinides in urine and fecal samples

    DOE Patents [OSTI]

    McKibbin, T.T.

    1993-03-02T23:59:59.000Z

    A method of determining the radioactivity of specific actinides that are carried in urine or fecal sample material is disclosed. The samples are ashed in a muffle furnace, dissolved in an acid, and then treated in a series of steps of reduction, oxidation, dissolution, and precipitation, including a unique step of passing a solution through a chloride form anion exchange resin for separation of uranium and plutonium from americium.

  12. Chemical and Ceramic Methods Toward Safe Storage of Actinides

    SciTech Connect (OSTI)

    P.E.D. Morgan; R.M. Housley; J.B. Davis; M.L. DeHaan

    2005-08-19T23:59:59.000Z

    A very import, extremely-long-term, use for monazite as a radwaste encapsulant has been proposed. THe use of ceramic La-monazite for sequestering actinides (isolating them from the environment), especially plutonium and some other radioactive elements )e.g., fission-product rare earths), had been especially championed by Lynn Boatner of ORNL. Monazite may be used alone or, copying its compatibility with many other minerals in nature, may be used in diverse composite combinations.

  13. Method for the concentration and separation of actinides from biological and environmental samples

    DOE Patents [OSTI]

    Horwitz, E.P.; Dietz, M.L.

    1989-05-30T23:59:59.000Z

    A method and apparatus for the quantitative recover of actinide values from biological and environmental sample by passing appropriately prepared samples in a mineral acid solution through a separation column of a dialkyl(phenyl)-N,N-dialylcarbamoylmethylphosphine oxide dissolved in tri-n-butyl phosphate on an inert substrate which selectively extracts the actinide values. The actinide values can be eluted either as a group or individually and their presence quantitatively detected by alpha counting. 3 figs.

  14. Supercritical Fluid Extraction and Separation of Uranium from Other Actinides

    SciTech Connect (OSTI)

    Donna L. Quach; Bruce J. Mincher; Chien M. Wai

    2014-06-01T23:59:59.000Z

    This paper investigates the feasibility of separating uranium from other actinides by using supercritical fluid carbon dioxide (sc-CO2) as a solvent modified with tri-n-butylphosphate (TBP) for the development of an extraction and counter current stripping technique, which would be a more efficient and environmentally benign technology for used nuclear fuel reprocessing compared to traditional solvent extraction. Several actinides (U(VI), Np(VI), Pu(IV), and Am(III)) were extracted in sc-CO2 modified with TBP over a range of nitric acid concentrations and then the actinides were exposed to reducing and complexing agents to suppress their extractability. According to this study, the separation of uranium from plutonium in sc-CO2 modified with TBP was successful at nitric acid concentrations of less than 3 M in the presence of acetohydroxamic acid or oxalic acid, and the separation of uranium from neptunium was successful at nitric acid concentrations of less than 1 M in the presence of acetohydroxamic acid, oxalic acid, or sodium nitrite.

  15. The EBR-II X501 Minor Actinide Burning Experiment

    SciTech Connect (OSTI)

    M. K. Meyer; S. L. Hayes; W. J. Carmack; H. Tsai

    2009-07-01T23:59:59.000Z

    The X501 experiment was conducted in EBR-II as part of the IFR (Integral Fast Reactor) program to demonstrate minor actinide burning through the use of a homogeneous recycle scheme. The X501 subassembly contained two metallic fuel elements loaded with relatively small quantities of americium and neptunium. Interest in the behavior of minor actinides (MA) during fuel irradiation has prompted further examination of existing X501 data, and generation of new data where needed in support of the U.S. waste transmutation effort. The X501 experiment is one of the few minor actinide-bearing fuel irradiation tests conducted worldwide and knowledge can be gained by understanding the changes in fuel behavior due to addition of MA’s. Of primary interest are the affect of the MA’s on fuel-cladding-chemical-interaction, and the redistribution behavior of americium. The quantity of helium gas release from the fuel and any effects of helium on fuel performance are also of interest. It must be stressed that information presented at this time is based on the limited PIE conducted in 1995-1996, and currently represents a set of observations rather than a complete understanding of fuel behavior. This paper provides a summary of the X501 fabrication, characterization, irradiation, and post irradiation examination.

  16. The EBR-II X501 Minor Actinide Burning Experiment

    SciTech Connect (OSTI)

    Jon Carmack; S. L. Hayes; M. K. Meyer; H. Tsai

    2008-06-01T23:59:59.000Z

    The X501 experiment was conducted in EBR-II as part of the IFR (Integral Fast Reactor) program to demonstrate minor actinide burning through the use of a homogeneous recycle scheme. The X501 subassembly contained two metallic fuel elements loaded with relatively small quantities of americium and neptunium. Interest in the behavior of minor actinides (MA) during fuel irradiation has prompted further examination of existing X501 data, and generation of new data where needed in support of the U.S. waste transmutation effort. The X501 experiment is one of the few minor actinide-bearing fuel irradiation tests conducted worldwide and knowledge can be gained by understanding the changes in fuel behavior due to addition of MA’s. Of primary interest are the affect of the MA’s on fuel-cladding-chemical-interaction, and the redistribution behavior of americium. The quantity of helium gas release from the fuel and any effects of helium on fuel performance are also of interest. It must be stressed that information presented at this time is based on the limited PIE conducted in 1995-1996, and currently represents a set of observations rather than a complete understanding of fuel behavior.

  17. E-Print Network 3.0 - actinides storage host Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    of the most developed of the immobilisation technologies... - emitting radionuclides of heavy metals (actinides, Raand Po). The specific activity of the ash ... Source:...

  18. E-Print Network 3.0 - actinides including cm Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Summary: of waste actinides. Such damage can be studied by many techniques, including heavy-ionfast neutron... were presented and proposed, including; Fundamental studies...

  19. E-Print Network 3.0 - advanced actinide fuels Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Breeder Summary: energy moderation, actinide solubility, and initial fuel inventory. For heavy nuclei (HN) proportions... the fuel refabrication problem in the presence of...

  20. E-Print Network 3.0 - actinide decay heat Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    then on, it is the actinides... --specifically, isotopes of plutonium, americium, and curium--that will contribute most to radioactive ... Source: Massachusetts Institute of...

  1. E-Print Network 3.0 - actinide separations conference Sample...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    region 1, 2. 2. Separation in-29;ight and the parent21;daughter method for heavy... . Mnzenberg et al., Proc. Actinides-1981 Conference, Paci28;c Grove, Cali-...

  2. E-Print Network 3.0 - actinides review hyperfine Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Commission of the European Communities, Joint Research Centre, European... bande et la nature des liaisons chimiques sont tudies, pour les mtaux et composs d'actinides, ...

  3. E-Print Network 3.0 - actinide separations thermodynamic Sample...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    of SARS (severe acute respiratory syndrome) Summary: of the actinide elements such as uranium and plutonium is central to predicting nuclear weapons performance......

  4. E-Print Network 3.0 - actinides phosphinic resins Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Databases and Resources 11 Development of an Automatic Method for Americium and Plutonium Separation and Summary: tetravalent and hexavalent actinides present in the real sample,...

  5. E-Print Network 3.0 - actinides ix behavior Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Torstein - Institut for Fysik og Astronomi, Aarhus Universitet Collection: Physics ; Materials Science 5 One-electron physics of the actinides A. Toropova, C. A. Marianetti, K....

  6. E-Print Network 3.0 - actinide region progress Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Databases and Resources 71 The materials test station: A fast-spectrum irradiation facility Eric J. Pitcher Summary: with significant inclusion of plutonium and minor actinides....

  7. E-Print Network 3.0 - actinide intermetallic laves-phase Sample...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . - Essentially atomic electron-polaron mechanism reducing the magnetic moments of rare-earth and actinide... elements in intermetallic compounds is proposed. This mechanism is...

  8. Crystal Chemistry of Early Actinides (Thorium, Uranium, and Neptunium) and Uranium Mesoporous Materials.

    E-Print Network [OSTI]

    Sigmon, Ginger E.

    2010-01-01T23:59:59.000Z

    ??Despite their considerable global importance, the structural chemistry of actinides remains understudied. Thorium and uranium fuel cycles are used in commercial nuclear reactors in India… (more)

  9. MOLECULAR SPECTROSCPY AND REACTIONS OF ACTINIDES IN THE GAS PHASE AND CRYOGENIC MATRICES

    E-Print Network [OSTI]

    Heaven, Michael C.

    2011-01-01T23:59:59.000Z

    importance in the chemistry of uranium, and these species5f orbitals in the chemistry of uranium complexes. Using CHchemistry studies involving the actinides dealt with volatile uranium

  10. E-Print Network 3.0 - actinides bilan des Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    (March 1977) p. 43. 8... heats of actinide metals M. J. Mortimer Chemistry Division, AERE ... Source: Ecole Polytechnique, Centre de mathmatiques Collection: Mathematics 73...

  11. EA-1404: Actinide Chemistry and Repository Science Laboratory, Carlsbad, New Mexico

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts for the proposal to construct and operate an Actinide Chemistry and Repository Science Laboratory to support chemical research activities related to the...

  12. E-Print Network 3.0 - actinide partitioning part Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    lowest f... . 2. -Variation of the atomic volume along the actinide, lanthanide and transition metal series Source: Ecole Polytechnique, Centre de mathmatiques Collection:...

  13. E-Print Network 3.0 - actinide bearing nitride Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    > >> 41 Electronic structure and pairwise interactions in substoichiometric transition metal carbides and nitrides Summary: ) of transition metals, rare earths and actinides in...

  14. E-Print Network 3.0 - actinide transmutation reactor Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    All these names are used... . The idea of combining powerful accelerators - with a subcritical reactor for transmutation purposes... homogeneous fuel Actinides MgO Tc Fast...

  15. E-Print Network 3.0 - actinide based fuel Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    MA represents minor actinides such as Np, Am, and Cm. Fuel... of the performance on the subcritical level. Numerical experiments are carried out on a ... Source: Royal Institute...

  16. E-Print Network 3.0 - actinide measurement quality Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    4 LA-UR 09-03222 LOS ALAMOS NATIONAL LABORATORY Summary: assurance QAPD Quality Assurance Program Document SNL Sandia National Laboratories SOTERM Actinide Source... on the...

  17. Removal of radioactive materials and heavy metals from water using magnetic resin

    DOE Patents [OSTI]

    Kochen, R.L.; Navratil, J.D.

    1997-01-21T23:59:59.000Z

    Magnetic polymer resins capable of efficient removal of actinides and heavy metals from contaminated water are disclosed together with methods for making, using, and regenerating them. The resins comprise polyamine-epichlorohydrin resin beads with ferrites attached to the surfaces of the beads. Markedly improved water decontamination is demonstrated using these magnetic polymer resins of the invention in the presence of a magnetic field, as compared with water decontamination methods employing ordinary ion exchange resins or ferrites taken separately. 9 figs.

  18. Removal of radioactive materials and heavy metals from water using magnetic resin

    DOE Patents [OSTI]

    Kochen, Robert L. (Boulder, CO); Navratil, James D. (Simi Valley, CA)

    1997-01-21T23:59:59.000Z

    Magnetic polymer resins capable of efficient removal of actinides and heavy metals from contaminated water are disclosed together with methods for making, using, and regenerating them. The resins comprise polyamine-epichlorohydrin resin beads with ferrites attached to the surfaces of the beads. Markedly improved water decontamination is demonstrated using these magnetic polymer resins of the invention in the presence of a magnetic field, as compared with water decontamination methods employing ordinary ion exchange resins or ferrites taken separately.

  19. Cyclic Mode of Transmutation of Minor Actinides in Heavy-Water Reactor

    SciTech Connect (OSTI)

    Gerasimov, Aleksander S.; Kiselev, Gennady V.; Myrtsymova, Lidia A.; Zaritskaya, Tamara S. [Institute of Theoretical and Experimental Physics, SSC RF ITEP, Bolshaya Cheremushkinskaya, 25, 117218 Moscow (Russian Federation)

    2002-07-01T23:59:59.000Z

    Characteristics of process of transmutation of americium and curium from spent nuclear fuel in heavy-water reactor during first 10 lifetimes and at transition to equilibrium mode are calculated. During transmutation, dangerous nuclides, first of all, {sup 244}Cm and {sup 238}Pu are accumulated. They cause an increase of radiotoxicity. At first 10 cycles of a transmutation, the radiotoxicity is increased by 11 times in comparison with initial load of transmuted actinides. Heavy-water reactor with thermal power of 1000 MW can transmute americium and curium extracted from 7-8 VVER-1000 type reactors. It means that the required power of transmutation reactor makes about 4 % of thermal power of VVER-1000 type reactors. (authors)

  20. Fabrication of advanced oxide fuels containing minor actinide for use in fast reactors

    SciTech Connect (OSTI)

    Miwa, Shuhei; Osaka, Masahiko; Tanaka, Kosuke; Ishi, Yohei; Yoshimochi, Hiroshi; Tanaka, Kenya [Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Oarai-machi, Higashi-ibaraki-gun, Ibaraki, 311-1393 (Japan)

    2007-07-01T23:59:59.000Z

    R and D of advanced fuel containing minor actinide for use in fast reactors is described related to the composite fuel with MgO matrix. Fabrication tests of MgO composite fuels containing Am were done by a practical process that could be adapted to the presently used commercial manufacturing technology. Am-containing MgO composite fuels having good characteristics, i.e., having no defects, a high density, a homogeneous dispersion of host phase, were obtained. As related technology, burn-up characteristics of a fast reactor core loaded with the present MgO composite fuel were also analyzed, mainly in terms of core criticality. Furthermore, phase relations of MA oxide which was assumed to be contained in MgO matrix fuel were experimentally investigated. (authors)

  1. Feasibility of actinide separation from UREX-like raffinates using a combination of sulfur- and oxygen-donor extractants

    SciTech Connect (OSTI)

    Zalupski, P.R.; Peterman, D.R.; Riddle, C.L. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2013-07-01T23:59:59.000Z

    A synergistic combination of bis(o-trifluoromethylphenyl)dithios-phosphinic acid and trioctylphosphine oxide has been recently shown to selectively remove uranium, neptunium, plutonium and americium from aqueous environment containing up to 0.5 M nitric acid and 5.5 g/l fission products. Here the feasibility of performing this complete actinide recovery from aqueous mixtures is forecasted for a new organic formulation containing sulfur donor extractant of modified structure based on Am(III) and Eu(III) extraction data. A mixture of bis(bis-m,m-trifluoromethyl)phenyl)-dithios-phosphinic acid and TOPO in toluene enhances the extraction performance, accomplishing Am/Eu differentiation in aqueous mixtures up to 1 M nitric acid. The new organic recipe is also less susceptible to oxidative damage resulting from radiolysis. (authors)

  2. Feasibility of actinide separation from UREX-like raffinates using a combination of sulfur- and oxygen-donor extractants

    SciTech Connect (OSTI)

    Peter R. Zalupski; Dean R. Peterman; Catherine L. Riddle

    2013-09-01T23:59:59.000Z

    A synergistic combination of bis(o-trifluoromethylphenyl)dithiosphosphinic acid and trioctylphosphine oxide has been recently shown to selectively remove uranium, neptunium, plutonium and americium from aqueous environment containing up to 0.5 M nitric acid and 5.5 g/L fission products. Here the feasibility of performing this complete actinide recovery from aqueous mixtures is forecasted for a new organic formulation containing sulfur donor extractant of modified structure based on Am(III) and Eu(III) extraction data. A mixture of bis(bis-m,m-trifluoromethyl)phenyl)-dithiosphosphinic acid and TOPO in toluene enhances the extraction performance, accomplishing Am/Eu differentiation in aqueous mixtures up to 1 M nitric acid. The new organic recipe is also less susceptible to oxidative damage resulting from radiolysis.

  3. In situ removal of contamination from soil

    DOE Patents [OSTI]

    Lindgren, E.R.; Brady, P.V.

    1997-10-14T23:59:59.000Z

    A process of remediation of cationic heavy metal contamination from soil utilizes gas phase manipulation to inhibit biodegradation of a chelating agent that is used in an electrokinesis process to remove the contamination. The process also uses further gas phase manipulation to stimulate biodegradation of the chelating agent after the contamination has been removed. The process ensures that the chelating agent is not attacked by bioorganisms in the soil prior to removal of the contamination, and that the chelating agent does not remain as a new contaminant after the process is completed. 5 figs.

  4. Note LPSC 07-37 The TMSR as Actinide Burner and Thorium Breeder

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Note LPSC 07-37 The TMSR as Actinide Burner and Thorium Breeder E. Merle-Lucotte, D. Heuer, C. Le actinides. Studies [1] have thus been done on the Molten Salt Breeder Reactor (MSBR) [2] of Oak-Ridge to re fluoride salt LiF- ThF4 with 28%- mole 232 Th. This reflector, corresponding to a fertile blanket

  5. LIBS Spectral Data for a Mixed Actinide Fuel Pellet Containing Uranium, Plutonium, Neptunium and Americium

    SciTech Connect (OSTI)

    Judge, Elizabeth J. [Los Alamos National Laboratory; Berg, John M. [Los Alamos National Laboratory; Le, Loan A. [Los Alamos National Laboratory; Lopez, Leon N. [Los Alamos National Laboratory; Barefield, James E. [Los Alamos National Laboratory

    2012-06-18T23:59:59.000Z

    Laser-induced breakdown spectroscopy (LIBS) was used to analyze a mixed actinide fuel pellet containing 75% UO{sub 2}/20% PuO{sub 2}/3% AmO{sub 2}/2% NpO{sub 2}. The preliminary data shown here is the first report of LIBS analysis of a mixed actinide fuel pellet, to the authors knowledge. The LIBS spectral data was acquired in a plutonium facility at Los Alamos National Laboratory where the sample was contained within a glove box. The initial installation of the glove box was not intended for complete ultraviolet (UV), visible (VIS) and near infrared (NIR) transmission, therefore the LIBS spectrum is truncated in the UV and NIR regions due to the optical transmission of the window port and filters that were installed. The optical collection of the emission from the LIBS plasma will be optimized in the future. However, the preliminary LIBS data acquired is worth reporting due to the uniqueness of the sample and spectral data. The analysis of several actinides in the presence of each other is an important feature of this analysis since traditional methods must chemically separate uranium, plutonium, neptunium, and americium prior to analysis. Due to the historic nature of the sample fuel pellet analyzed, the provided sample composition of 75% UO{sub 2}/20% PuO{sub 2}/3% AmO{sub 2}/2% NpO{sub 2} cannot be confirm without further analytical processing. Uranium, plutonium, and americium emission lines were abundant and easily assigned while neptunium was more difficult to identify. There may be several reasons for this observation, other than knowing the exact sample composition of the fuel pellet. First, the atomic emission wavelength resources for neptunium are limited and such techniques as hollow cathode discharge lamp have different dynamics than the plasma used in LIBS which results in different emission spectra. Secondly, due to the complex sample of four actinide elements, which all have very dense electronic energy levels, there may be reactions and interactions occurring within the plasma, such as collisional energy transfer, that might be a factor in the reduction in neptunium emission lines. Neptunium has to be analyzed alone using LIBS to further understand the dynamics that may be occurring in the plasma of the mixed actinide fuel pellet sample. The LIBS data suggests that the emission spectrum for the mixed actinide fuel pellet is not simply the sum of the emission spectra of the pure samples but is dependent on the species present in the plasma and the interactions and reactions that occur within the plasma. Finally, many of the neptunium lines are in the near infrared region which is drastically reduced in intensity by the current optical setup and possibly the sensitivity of the emission detector in the spectral region. Once the optics are replaced and the optical collection system is modified and optimized, the probability of observing emission lines for neptunium might be increased significantly. The mixed actinide fuel pellet was analyzed under the experimental conditions listed in Table 1. The LIBS spectra of the fuel pellet are shown in Figures 1-49. The spectra are labeled with the observed wavelength and atomic species (both neutral (I) and ionic (II)). Table 2 is a complete list of the observed and literature based emission wavelengths. The literature wavelengths have references including NIST Atomic Spectra Database (NIST), B.A. Palmer et al. 'An Atlas of Uranium Emission Intensities in a Hollow Cathode Discharge' taken at the Kitt Peak National Observatory (KPNO), R.L. Kurucz 1995 Atomic Line Data from the Smithsonian Astrophysical Observatory (SAO), J. Blaise et al. 'The Atomic Spectrum of Plutonium' from Argonne National Laboratory (BFG), and M. Fred and F.S. Tomkins, 'Preliminary Term Analysis of Am I and Am II Spectra' (FT). The dash (-) shown under Ionic State indicates that the ionic state of the transition was not available. In the spectra, the dash (-) is replaced with a question mark (?). Peaks that are not assigned are most likely real features and not noise but cannot be confidently assi

  6. Final Report on Actinide Glass Scintillators for Fast Neutron Detection

    SciTech Connect (OSTI)

    Bliss, Mary; Stave, Jean A.

    2012-10-01T23:59:59.000Z

    This is the final report of an experimental investigation of actinide glass scintillators for fast-neutron detection. It covers work performed during FY2012. This supplements a previous report, PNNL-20854 “Initial Characterization of Thorium-loaded Glasses for Fast Neutron Detection” (October 2011). The work in FY2012 was done with funding remaining from FY2011. As noted in PNNL-20854, the glasses tested prior to July 2011 were erroneously identified as scintillators. The decision was then made to start from “scratch” with a literature survey and some test melts with a non-radioactive glass composition that could later be fabricated with select actinides, most likely thorium. The normal stand-in for thorium in radioactive waste glasses is cerium in the same oxidation state. Since cerium in the 3+ state is used as the light emitter in many scintillating glasses, the next most common substitute was used: hafnium. Three hafnium glasses were melted. Two melts were colored amber and a third was clear. It barely scintillated when exposed to alpha particles. The uses and applications for a scintillating fast neutron detector are important enough that the search for such a material should not be totally abandoned. This current effort focused on actinides that have very high neutron capture energy releases but low neutron capture cross sections. This results in very long counting times and poor signal to noise when working with sealed sources. These materials are best for high flux applications and access to neutron generators or reactors would enable better test scenarios. The total energy of the neutron capture reaction is not the only factor to focus on in isotope selection. Many neutron capture reactions result in energetic gamma rays that require large volumes or high densities to detect. If the scintillator is to separate neutrons from gamma rays, the capture reactions should produce heavy particles and few gamma rays. This would improve the detection of a signal for fast neutron capture.

  7. Microbial Transformation of TRU and Mixed Waste: Actinide Speciation and Waste Volume

    SciTech Connect (OSTI)

    Halada, Gary P

    2008-04-10T23:59:59.000Z

    In order to understand the susceptibility of transuranic and mixed waste to microbial degradation (as well as any mechanism which depends upon either complexation and/or redox of metal ions), it is essential to understand the association of metal ions with organic ligands present in mixed wastes. These ligands have been found in our previous EMSP study to limit electron transfer reactions and strongly affect transport and the eventual fate of radionuclides in the environment. As transuranic waste (and especially mixed waste) will be retained in burial sites and in legacy containment for (potentially) many years while awaiting treatment and removal (or remaining in place under stewardship agreements at government subsurface waste sites), it is also essential to understand the aging of mixed wastes and its implications for remediation and fate of radionuclides. Mixed waste containing actinides and organic materials are especially complex and require extensive study. The EMSP program described in this report is part of a joint program with the Environmental Sciences Department at Brookhaven National Laboratory. The Stony Brook University portion of this award has focused on the association of uranium (U(VI)) and transuranic analogs (Ce(III) and Eu(III)) with cellulosic materials and related compounds, with development of implications for microbial transformation of mixed wastes. The elucidation of the chemical nature of mixed waste is essential for the formulation of remediation and encapsulation technologies, for understanding the fate of contaminant exposed to the environment, and for development of meaningful models for contaminant storage and recovery.

  8. Plutonium and minor actinides utilization in Thorium molten salt reactor

    SciTech Connect (OSTI)

    Waris, Abdul; Aji, Indarta K.; Novitrian,; Kurniadi, Rizal; Su'ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jalan Ganesa 10 Bandung 40132 (Indonesia)

    2012-06-06T23:59:59.000Z

    FUJI-12 reactor is one of MSR systems that proposed by Japan. The original FUJI-12 design considers Th/{sup 233}U or Th/Pu as main fuel. In accordance with the currently suggestion to stay away from the separation of Pu and minor actinides (MA), in this study we evaluated the utilization of Pu and MA in FUJI-12. The reactor grade Pu was employed in the present study as a small effort of supporting THORIMS-NES scenario. The result shows that the reactor can achieve its criticality with the Pu and MA composition in the fuel of 5.96% or more.

  9. Actinide Neutron-Induced Fission Cross Section Measurements At LANSCE

    SciTech Connect (OSTI)

    Tovesson, F.; Laptev, A. B. [Los Alamos National Laboratory, Los Alamos NM 87545 (United States); Hill, T. S. [Idaho National Laboratory, Idaho Falls ID 83415 (United States)

    2011-06-01T23:59:59.000Z

    Fission cross sections of a range of actinides have been measured at the Los Alamos Neutron Science Center (LANSCE) in support of nuclear energy applications in a wide energy range from sub thermal energies up to 200 MeV. Parallel-plate ionization chambers are used to measure fission cross sections ratios relative to the {sup 235}U standard while incident neutron energies are determined using the time-of-flight method. Recent measurements include the {sup 233,238}U, {sup 239-242}Pu and {sup 243}Am neutron-induced fission cross sections. Obtained data are presented in comparison with existing evaluations and previous data.

  10. UCN Actinides | Ultracold Neutrons at Los Alamos National Laboratory

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLasDelivered energy consumption by sectorlong version) The U.S.1,summer gasoline priceActinides

  11. Gas Generation from Actinide Oxide Materials

    SciTech Connect (OSTI)

    George Bailey; Elizabeth Bluhm; John Lyman; Richard Mason; Mark Paffett; Gary Polansky; G. D. Roberson; Martin Sherman; Kirk Veirs; Laura Worl

    2000-12-01T23:59:59.000Z

    This document captures relevant work performed in support of stabilization, packaging, and long term storage of plutonium metals and oxides. It concentrates on the issue of gas generation with specific emphasis on gas pressure and composition. Even more specifically, it summarizes the basis for asserting that materials loaded into a 3013 container according to the requirements of the 3013 Standard (DOE-STD-3013-2000) cannot exceed the container design pressure within the time frames or environmental conditions of either storage or transportation. Presently, materials stabilized and packaged according to the 3013 Standard are to be transported in certified packages (the certification process for the 9975 and the SAFKEG has yet to be completed) that do not rely on the containment capabilities of the 3013 container. Even though no reliance is placed on that container, this document shows that it is highly likely that the containment function will be maintained not only in storage but also during transportation, including hypothetical accident conditions. Further, this document, by summarizing materials-related data on gas generation, can point those involved in preparing Safety Analysis Reports for Packages (SARPs) to additional information needed to assess the ability of the primary containment vessel to contain the contents and any reaction products that might reasonably be produced by the contents.

  12. In situ removal of contamination from soil

    DOE Patents [OSTI]

    Lindgren, Eric R. (Albuquerque, NM); Brady, Patrick V. (Albuquerque, NM)

    1997-01-01T23:59:59.000Z

    A process of remediation of cationic heavy metal contamination from soil utilizes gas phase manipulation to inhibit biodegradation of a chelating agent that is used in an electrokinesis process to remove the contamination, and further gas phase manipulation to stimulate biodegradation of the chelating agent after the contamination has been removed. The process ensures that the chelating agent is not attacked by bioorganisms in the soil prior to removal of the contamination, and that the chelating agent does not remain as a new contaminant after the process is completed.

  13. Actinide Foil Production for MPACT Research

    SciTech Connect (OSTI)

    Beller, Denis

    2012-10-31T23:59:59.000Z

    Sensitive fast-neutron detectors are required for use in lead slowing down spectrometry (LSDS), an active interrogation technique for used nuclear fuel assay for Materials Protection, Accounting, and Controls Technologies (MPACT). During the past several years UNLV sponsored a research project at RPI to investigate LSDS; began development of fission chamber detectors for use in LSDS experiments in collaboration with INL, LANL, and Oregon State U.; and participated in a LSDS experiment at LANL. In the LSDS technique, research has demonstrated that these fission chamber detectors must be sensitive to fission energy neutrons but insensitive to thermal-energy neutrons. Because most systems are highly sensitive to large thermal neutron populations due to the well-known large thermal cross section of 235U, even a miniscule amount of this isotope in a fission chamber will overwhelm the small population of higher-energy neutrons. Thus, fast-fission chamber detectors must be fabricated with highly depleted uranium (DU) or ultra-pure thorium (Th), which is about half as efficient as DU. Previous research conducted at RPI demonstrated that the required purity of DU for assay of used nuclear fuel using LSDS is less than 4 ppm 235U, material that until recently was not available in the U.S. In 2009 the PI purchased 3 grams of ultra-depleted uranium (uDU, 99.99998% 238U with just 0.2 ���± 0.1 ppm 235U) from VNIIEF in Sarov, Russia. We received the material in the form of U3O8 powder in August of 2009, and verified its purity and depletion in a FY10 MPACT collaboration project. In addition, chemical processing for use in FC R&D was initiated, fission chamber detectors and a scanning alpha-particle spectrometer were developed, and foils were used in a preliminary LSDS experiment at a LANL/LANSCE in Sept. of 2010. The as-received U3O8 powder must be chemically processed to convert it to another chemical form while maintaining its purity, which then must be used to electro-deposit U or UO2 in extremely thin layers (1 to 2 mg/cm2) on various media such as films, foils, or discs. After many months of investigation and trials in FY10 and 11, UNLV researchers developed a new method to produce pure UO2 deposits on foils using a unique approach, which has never been demonstrated, that involves dissolution of U3O8 directly into room temperature ionic liquid (RTIL) followed by electrodeposition from the RTIL-uDU solution (Th deposition from RTIL had been previously demonstrated). The high-purity dissolution of the U3O8 permits the use of RTIL solutions for deposition of U on metal foils in layers without introducing contamination. In FY10 and early FY11 a natural U surrogate for the uDU was used to investigate this and other techniques. In this research project UNLV will deposit directly from RTIL to produce uDU and Th foils devoid of possible contaminants. After these layers have been deposited, they will be examined for purity and uniformity. UNLV will complete the development and demonstration of the RTIL technology/ methodology to prepare uDU and Th samples for use in constructing fast-neutron detectors. Although this material was purchased for use in research using fast-fission chamber detectors for active inspection techniques for MPACT, it could also contribute to R&D for other applications, such as cross section measurements or neutron spectroscopy for national security

  14. Development of a Sorption Enhanced Steam Hydrogasification Process for In-situ Carbon Dioxide (CO2) Removal and Enhanced Synthetic Fuel Production

    E-Print Network [OSTI]

    Liu, Zhongzhe

    2013-01-01T23:59:59.000Z

    synthesis from biomass pyrolysis with in situ carbon dioxideof pyrolysis, combustion and gasification of three biomassand biomass, undergoes several different processes and/or reactions: dehydration, pyrolysis,

  15. Regenerative process for removal of mercury and other heavy metals from gases containing H.sub.2 and/or CO

    DOE Patents [OSTI]

    Jadhav, Raja A. (Naperville, IL)

    2009-07-07T23:59:59.000Z

    A method for removal of mercury from a gaseous stream containing the mercury, hydrogen and/or CO, and hydrogen sulfide and/or carbonyl sulfide in which a dispersed Cu-containing sorbent is contacted with the gaseous stream at a temperature in the range of about 25.degree. C. to about 300.degree. C. until the sorbent is spent. The spent sorbent is contacted with a desorbing gaseous stream at a temperature equal to or higher than the temperature at which the mercury adsorption is carried out, producing a regenerated sorbent and an exhaust gas comprising released mercury. The released mercury in the exhaust gas is captured using a high-capacity sorbent, such as sulfur-impregnated activated carbon, at a temperature less than about 100.degree. C. The regenerated sorbent may then be used to capture additional mercury from the mercury-containing gaseous stream.

  16. Actinide production from xenon bombardments of curium-248

    SciTech Connect (OSTI)

    Welch, R.B.

    1985-01-01T23:59:59.000Z

    Production cross sections for many actinide nuclides formed in the reaction of /sup 129/Xe and /sup 132/Xe with /sup 248/Cm at bombarding energies slightly above the coulomb barrier were determined using radiochemical techniques to isolate these products. These results are compared with cross sections from a /sup 136/Xe + /sup 248/Cm reaction at a similar energy. When compared to the reaction with /sup 136/Xe, the maxima in the production cross section distributions from the more neutron deficient projectiles are shifted to smaller mass numbers, and the total cross section increases for the production of elements with atomic numbers greater than that of the target, and decreases for lighter elements. These results can be explained by use of a potential energy surface (PES) which illustrates the effect of the available energy on the transfer of nucleons and describes the evolution of the di-nuclear complex, an essential feature of deep-inelastic reactions (DIR), during the interaction. The other principal reaction mechanism is the quasi-elastic transfer (QE). Analysis of data from a similar set of reactions, /sup 129/Xe, /sup 132/Xe, and /sup 136/Xe with /sup 197/Au, aids in explaining the features of the Xe + Cm product distributions, which are additionally affected by the depletion of actinide product yields due to deexcitation by fission. The PES is shown to be a useful tool to predict the general features of product distributions from heavy ion reactions.

  17. Large Component Removal/Disposal

    SciTech Connect (OSTI)

    Wheeler, D. M.

    2002-02-27T23:59:59.000Z

    This paper describes the removal and disposal of the large components from Maine Yankee Atomic Power Plant. The large components discussed include the three steam generators, pressurizer, and reactor pressure vessel. Two separate Exemption Requests, which included radiological characterizations, shielding evaluations, structural evaluations and transportation plans, were prepared and issued to the DOT for approval to ship these components; the first was for the three steam generators and one pressurizer, the second was for the reactor pressure vessel. Both Exemption Requests were submitted to the DOT in November 1999. The DOT approved the Exemption Requests in May and July of 2000, respectively. The steam generators and pressurizer have been removed from Maine Yankee and shipped to the processing facility. They were removed from Maine Yankee's Containment Building, loaded onto specially designed skid assemblies, transported onto two separate barges, tied down to the barges, th en shipped 2750 miles to Memphis, Tennessee for processing. The Reactor Pressure Vessel Removal Project is currently under way and scheduled to be completed by Fall of 2002. The planning, preparation and removal of these large components has required extensive efforts in planning and implementation on the part of all parties involved.

  18. High removal rate laser-based coating removal system

    DOE Patents [OSTI]

    Matthews, Dennis L. (Moss Beach, CA); Celliers, Peter M. (Berkeley, CA); Hackel, Lloyd (Livermore, CA); Da Silva, Luiz B. (Danville, CA); Dane, C. Brent (Livermore, CA); Mrowka, Stanley (Richmond, CA)

    1999-11-16T23:59:59.000Z

    A compact laser system that removes surface coatings (such as paint, dirt, etc.) at a removal rate as high as 1000 ft.sup.2 /hr or more without damaging the surface. A high repetition rate laser with multiple amplification passes propagating through at least one optical amplifier is used, along with a delivery system consisting of a telescoping and articulating tube which also contains an evacuation system for simultaneously sweeping up the debris produced in the process. The amplified beam can be converted to an output beam by passively switching the polarization of at least one amplified beam. The system also has a personal safety system which protects against accidental exposures.

  19. Downstream Processing of Recombinant Proteins from Transgenic Plant Systems: Phenolic Compounds Removal from Monoclonal Antibody Expressing Lemna minor and Purification of Recombinant Bovine Lysozyme from Sugarcane

    E-Print Network [OSTI]

    Barros, Georgia

    2012-07-16T23:59:59.000Z

    of the extraction condition, at least 47% of the starting BvLz was lost during the membrane processing. None of the evaluated extraction conditions caused a substantial recovery of BvLz in the concentrate. Alternative purification options for the IEX+HIC process...

  20. The Dirac equation in electronic structure calculations: Accurate evaluation of DFT predictions for actinides

    SciTech Connect (OSTI)

    Wills, John M [Los Alamos National Laboratory; Mattsson, Ann E [Sandia National Laboratories

    2012-06-06T23:59:59.000Z

    Brooks, Johansson, and Skriver, using the LMTO-ASA method and considerable insight, were able to explain many of the ground state properties of the actinides. In the many years since this work was done, electronic structure calculations of increasing sophistication have been applied to actinide elements and compounds, attempting to quantify the applicability of DFT to actinides and actinide compounds and to try to incorporate other methodologies (i.e. DMFT) into DFT calculations. Through these calculations, the limits of both available density functionals and ad hoc methodologies are starting to become clear. However, it has also become clear that approximations used to incorporate relativity are not adequate to provide rigorous tests of the underlying equations of DFT, not to mention ad hoc additions. In this talk, we describe the result of full-potential LMTO calculations for the elemental actinides, comparing results obtained with a full Dirac basis with those obtained from scalar-relativistic bases, with and without variational spin-orbit. This comparison shows that the scalar relativistic treatment of actinides does not have sufficient accuracy to provide a rigorous test of theory and that variational spin-orbit introduces uncontrolled errors in the results of electronic structure calculations on actinide elements.

  1. Turbomachinery debris remover

    DOE Patents [OSTI]

    Krawiec, Donald F. (Pittsburgh, PA); Kraf, Robert J. (North Huntingdon, PA); Houser, Robert J. (Monroeville, PA)

    1988-01-01T23:59:59.000Z

    An apparatus for removing debris from a turbomachine. The apparatus includes housing and remotely operable viewing and grappling mechanisms for the purpose of locating and removing debris lodged between adjacent blades in a turbomachine.

  2. Development of a Sorption Enhanced Steam Hydrogasification Process for In-situ Carbon Dioxide (CO2) Removal and Enhanced Synthetic Fuel Production

    E-Print Network [OSTI]

    Liu, Zhongzhe

    2013-01-01T23:59:59.000Z

    and steam-gasification of carbonaceous waste materials.L. Steam catalytic gasification of municipal solid waste forwaste are suitable and favorable for this process[55]. By contrast, conventional gasification

  3. Development of a Sorption Enhanced Steam Hydrogasification Process for In-situ Carbon Dioxide (CO2) Removal and Enhanced Synthetic Fuel Production

    E-Print Network [OSTI]

    Liu, Zhongzhe

    2013-01-01T23:59:59.000Z

    J. Different types of gasifiers and their integration withCO 2 in a pressurized-gasifier-based process. Energ Fuel.fluidized bed biomass steam gasifier-bed material and fuel

  4. Literature review of United States utilities computer codes for calculating actinide isotope content in irradiated fuel

    SciTech Connect (OSTI)

    Horak, W.C.; Lu, Ming-Shih

    1991-12-01T23:59:59.000Z

    This paper reviews the accuracy and precision of methods used by United States electric utilities to determine the actinide isotopic and element content of irradiated fuel. After an extensive literature search, three key code suites were selected for review. Two suites of computer codes, CASMO and ARMP, are used for reactor physics calculations; the ORIGEN code is used for spent fuel calculations. They are also the most widely used codes in the nuclear industry throughout the world. Although none of these codes calculate actinide isotopics as their primary variables intended for safeguards applications, accurate calculation of actinide isotopic content is necessary to fulfill their function.

  5. Integral Validation of Minor Actinide Nuclear Data by using Samples Irradiated at Dounreay Prototype Fast Reactor

    SciTech Connect (OSTI)

    Tsujimoto, Kazufumi; Oigawa, Hiroyuki; Shinohara, Nobuo [Japan Atomic Energy Research Institute, Shirakata Shirane 2-4, Tokai, Ibaraki 319-1195 (Japan)

    2005-05-24T23:59:59.000Z

    The reliability of nuclear data for minor actinides was evaluated by using the results of the post-irradiation experiment for actinide samples irradiated at the Dounreay Prototype Fast Reactor. The burnup calculations with JENDL-3.3, ENDF/B-VI.8, and JEFF-3.0 were performed. From the comparison between the experimental data and the calculational results, in general, the reliability of nuclear data for the minor actinides are at an adequate level for the conceptual design study of transmutation systems. It is, however, found that improvement of the accuracy is necessary for some nuclides, such as 238Pu, 242Pu, and 241Am.

  6. Delayed Neutron and Delayed Photon Characteristics from Photofission of Actinides

    SciTech Connect (OSTI)

    Dore, D.; Berthoumieux, E.; Leprince, A.; Ridikas, D. [DSM/IRFUS/PhN, CEA/Saclay, Gif-sur-Yvette, F-91191 (France); Ledoux, X. [CEA/DAM/DIF, Arpajon, F-91297 (France); Agelou, M.; Carrel, F.; Gmar, M. [CEA, LIST, Gif-sur-Yvette, F-91191 (France)

    2011-12-13T23:59:59.000Z

    Delayed neutron (DN) and delayed photon (DP) emissions from photofission reactions play an important role for applications involving nuclear material detection and characterization. To provide new, accurate, basic nuclear data for evaluations and data libraries, an experimental programme of DN and DP measurements has been undertaken for actinides with bremsstrahlung endpoint energy in the giant resonance region ({approx}15 MeV). In this paper, the experimental setup and the data analysis method will be described. Experimental results for DN and DP characteristics will be presented for {sup 232}Th, {sup 235,238}U, {sup 237}Np, and {sup 239}Pu. Finally, an example of an application to study the contents of nuclear waste packages will be briefly discussed.

  7. Flammability Analysis For Actinide Oxides Packaged In 9975 Shipping Containers

    SciTech Connect (OSTI)

    Laurinat, James E.; Askew, Neal M.; Hensel, Steve J.

    2013-03-21T23:59:59.000Z

    Packaging options are evaluated for compliance with safety requirements for shipment of mixed actinide oxides packaged in a 9975 Primary Containment Vessel (PCV). Radiolytic gas generation rates, PCV internal gas pressures, and shipping windows (times to reach unacceptable gas compositions or pressures after closure of the PCV) are calculated for shipment of a 9975 PCV containing a plastic bottle filled with plutonium and uranium oxides with a selected isotopic composition. G-values for radiolytic hydrogen generation from adsorbed moisture are estimated from the results of gas generation tests for plutonium oxide and uranium oxide doped with curium-244. The radiolytic generation of hydrogen from the plastic bottle is calculated using a geometric model for alpha particle deposition in the bottle wall. The temperature of the PCV during shipment is estimated from the results of finite element heat transfer analyses.

  8. E-Print Network 3.0 - actinide elements volume Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    no 5-6, Tome 33, Mai-Juin 1972,page C3-57 RELATIVISTIC ELECTRONIC BAND STRUCTURE OF THE HEAVY METALS Summary: and properties of the actinide elements before discussing the band...

  9. E-Print Network 3.0 - actinides np pu Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    NpTe, PuTe. Actinides compounds with the ThCr2Si2 type structure Since the discovery of heavy... . - The magneticand electricaltransport propertiesof the U, Np and Pu...

  10. E-Print Network 3.0 - actinide transmutation physics Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Fission Barriers and Half-Lives Summary: Acta Phys. Hung. A 251 (2006) 000-000 HEAVY ION PHYSICS On the Multiple-Humped Fission Barriers... and Half-Lives of Actinides...

  11. E-Print Network 3.0 - actinides fuel research Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Summary: in safety, proliferation resistance, and can be designed to breed fuel or burn heavy actinides. One... . The number of fuel pins in a fuel assembly of a PWR core is...

  12. E-Print Network 3.0 - application aux actinides Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    au Journal de Physique 111, Vol. 1,dCcembre 1991 Summary: and actinides : application to curium. Radiochim. Acta. 5253, part 1, 119 (1991). 12;... . MAUCHlEN CEADCCDPESPEA...

  13. E-Print Network 3.0 - actinide burning fuel Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Summary: in safety, proliferation resistance, and can be designed to breed fuel or burn heavy actinides. One... . The number of fuel pins in a fuel assembly of a PWR core is...

  14. Configuration adjustment potential of the Very High Temperature Reactor prismatic cores with advanced actinide fuels

    E-Print Network [OSTI]

    Ames, David E, II

    2006-10-30T23:59:59.000Z

    Minor actinides represent the long-term radiotoxicity of nuclear wastes. As one of their potential incineration options, partitioning and transmutation in fission reactors are seriously considered worldwide. If implemented, these technologies could...

  15. E-Print Network 3.0 - actinide waste forms Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    .-F. Lucchini, M.K. Richmann, and D.T. Reed. 2008. "Actinide (III) Solubility in WIPP Brine: Data Summary... ) but carbonate phases can be formed at the higher fugacities...

  16. E-Print Network 3.0 - actinide nuclei indirectly Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    40 (2009) ACTA PHYSICA POLONICA B No 3 PRODUCTION OF NEW SUPERHEAVY NUCLEI IN COMPLETE FUSION... physics. The cold Pb- and Bi-based 1 and hot actinide-based 2 complete...

  17. E-Print Network 3.0 - actinide nuclei induced Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    40 (2009) ACTA PHYSICA POLONICA B No 3 PRODUCTION OF NEW SUPERHEAVY NUCLEI IN COMPLETE FUSION... physics. The cold Pb- and Bi-based 1 and hot actinide-based 2 complete...

  18. E-Print Network 3.0 - actinide-based complete-fusion reactions...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    fusion reactions. The yields of superheavies with Z > 118 are sensitive... physics. The cold Pb- and Bi-based 1 and hot actinide-based 2 complete fusion reactions were...

  19. E-Print Network 3.0 - actinide nuclei Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    40 (2009) ACTA PHYSICA POLONICA B No 3 PRODUCTION OF NEW SUPERHEAVY NUCLEI IN COMPLETE FUSION... physics. The cold Pb- and Bi-based 1 and hot actinide-based 2 complete...

  20. Optimization of actinide transmutation in innovative lead-cooled fast reactors

    E-Print Network [OSTI]

    Romano, Antonino, 1972-

    2003-01-01T23:59:59.000Z

    The thesis investigates the potential of fertile free fast lead-cooled modular reactors as efficient incinerators of plutonium and minor actinides (MAs) for application to dedicated fuel cycles for transmutation. A methodology ...

  1. E-Print Network 3.0 - actinides separation chemistry Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    separation chemistry Search Powered by Explorit Topic List Advanced Search Sample search results for: actinides separation chemistry Page: << < 1 2 3 4 5 > >> 1 www.emsl.pnl.gov...

  2. E-Print Network 3.0 - actinides solution chemistry Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    solution chemistry Search Powered by Explorit Topic List Advanced Search Sample search results for: actinides solution chemistry Page: << < 1 2 3 4 5 > >> 1 www.emsl.pnl.gov...

  3. E-Print Network 3.0 - actinide metal compounds Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    effects in the X-ray photoemission spectra of the actinides Summary: in 3d transition metal compounds and ns (n 4 , 5) levels in rare earth systems, it is clear that ME......

  4. E-Print Network 3.0 - actinide elements progress Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    DE PHYSIQUE Colloque C4, supplment au n 4, Tome 40, avril 1979, page C4-207 Alkali metal actinide complex halides: thermochemical and structural Summary: Elements and...

  5. Actinide-lanthanide separation with solvents on the base of amides of heterocyclic diacids

    SciTech Connect (OSTI)

    Babain, V.A.; Alyapyshev, M.Y.; Tkachenko, L.I. [Khlopin Radium Institute, 28, 2ndMurinski pr., St-Petersburg, Russia 19402 (Russian Federation)

    2013-07-01T23:59:59.000Z

    The separation of actinides from lanthanides with a particular emphasis on Am(III) from Eu(III) with amides of heterocyclic dicarboxylic diacids was reviewed. It was shown that the di-amides of the 2,2'-dipyridyl-6,6'-dicarboxylic acid are the most promising ligands for the simultaneous selective recovery of actinides from HLLW (high level radioactive liquid waste) within the GANEX concept. (author)

  6. Extraction of trivalent lanthanides and actinides by ``CMPO-like`` calixarenes

    SciTech Connect (OSTI)

    Delmau, L.H.; Simon, N. [CEA Cadarache, St. Paul lez Durance (France)] [CEA Cadarache, St. Paul lez Durance (France); Schwing-Weill, M.J. [ECPM, Strasbourg (France)] [and others] [ECPM, Strasbourg (France); and others

    1999-04-01T23:59:59.000Z

    Extractive properties of calix[4]arenes bearing carbamoylmethylphosphine oxide moieties on their upper rim toward trivalent lanthanide and actinide cations were investigated. The study revealed that these molecules selectively extract light lanthanides and actinides from heavy lanthanides. All parameters present in the extraction system were varied to determine the origin of the selectivity. It was found that this selectivity requires a calix[4]arene platform and acetamidophosphine oxide groups containing phenyl substituents on the four phosphorus atoms.

  7. Part removal of 3D printed parts

    E-Print Network [OSTI]

    Peńa Doll, Mateo

    2014-01-01T23:59:59.000Z

    An experimental study was performed to understand the correlation between printing parameters in the FDM 3D printing process, and the force required to remove a part from the build platform of a 3D printing using a patent ...

  8. Management of Salt Waste from Electrochemical Processing of Used Nuclear Fuel

    SciTech Connect (OSTI)

    Michael F. Simpson; Michael N. Patterson; Joon Lee; Yifeng Wang; Joshua Versey; Ammon Williams; Supathorn Phongikaroon; James Allensworth; Man-Sung Yim

    2013-10-01T23:59:59.000Z

    Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electrorefiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form.

  9. Management of salt waste from electrochemical processing of used nuclear fuel

    SciTech Connect (OSTI)

    Simpson, M.F.; Patterson, M.N. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415 (United States); Lee, J.; Wang, Y. [Sandia National Laboratory, Albuquerque, NM (United States); Versey, J.; Phongikaroon, S. [University of Idaho, Idaho Falls, ID (United States)

    2013-07-01T23:59:59.000Z

    Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electro-refiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form. (authors)

  10. Method for fluorination of actinide fluorides and oxyfluorides using O/sub 2/F/sub 2/

    DOE Patents [OSTI]

    Eller, P.G.; Malm, J.G.; Penneman, R.A.

    1984-08-01T23:59:59.000Z

    The present invention relates generally to methods of fluorination and more particularly to the use of O/sub 2/F/sub 2/ for the preparation of actinide hexafluorides, and for the extraction of deposited actinides and fluorides and oxyfluorides thereof from reaction vessels. The experiments set forth hereinabove demonstrate that the room temperature or below use of O/sub 2/F/sub 2/ will be highly beneficial for the preparation of pure actinide hexafluorides from their respective tetrafluorides without traces of HF being present as occurs using other fluorinating agents: and decontamination of equipment previously exposed to actinides: e.g., walls, feed lines, etc.

  11. Proliferation Resistance Evaluation of ACR-1000 Fuel with Minor Actinides

    SciTech Connect (OSTI)

    Gray S. Chang

    2008-09-01T23:59:59.000Z

    The Global Nuclear Energy Partnership (GNEP) program is to significantly advance the science and technology of nuclear energy systems and to enhance the spent fuel proliferation resistance. It consists of both innovative nuclear reactors and innovative research in separation and transmutation. The merits of nuclear energy are high-density energy, with low environmental impacts (i.e. almost zero greenhouse gas emission). Planned efforts involve near-term and intermediate-term improvements in fuel utilization and recycling in current light water reactors (LWRs) as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs can play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In this work, an Advanced CANDU Reactor (ACR) fuel unit lattice cell model with 43 UO2 fuel rods will be used to investigate the effectiveness of a Minor Actinide Reduction Approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance. The main MARA objective is to increase the 238Pu / Pu isotope ratio by using the transuranic nuclides (237Np and 241Am) in the high burnup fuel and thereby increase the proliferation resistance even for a very low fuel burnup. As a result, MARA is a very effective approach to enhance the proliferation resistance for the on power refueling ACR system nuclear fuel. The MA transmutation characteristics at different MA loadings were compared and their impact on neutronics criticality assessed. The concept of MARA, significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy reconnaissance

  12. REVIEW OF ACTINIDE AND STRONTIUM LOADING DATA FOR MST AND MMST

    SciTech Connect (OSTI)

    Peters, T.; Hobbs, D.; Fink, S.

    2010-10-20T23:59:59.000Z

    SRNL reviewed the relevant data from MST and mMST fissile loading studies to determine if further studies were required. With respect to MST, SRNL found that the published results adequately bound the expected conditions that Small Column Ion Exchange (SCIX) process will operate under. The lack of strontium data does not represent an issue as strontium is not relevant to criticality. There is no threat to criticality safety from the lack of strontium loading data. However, SRNL proposes a single test with MST to ensure that future SCIX operations are conservatively bounded and strontium maximum loading is understood. With respect to attempts to maximally load mMST, SRNL's knowledge on actinide and strontium loading is limited to uranium behavior. mMST has a very weak affinity for uranium, and even extended contact time at high uranium concentration shows minimal loading onto mMST. This leaves questions about the ability to load plutonium, neptunium and strontium. SRNL proposes to perform two tests with mMST to ensure that questions on plutonium, neptunium, and strontium sorption are answered, as well as ensuring that future mMST operations are conservatively bounded.

  13. Improving the actinides recycling in closed fuel cycles, a major step towards nuclear energy sustainability

    SciTech Connect (OSTI)

    Poinssot, C.; Grandjean, S.; Masson, M. [RadioChemistry and Processes Department, CEA Marcoule, 30207 Bagnols sur Ceze (France); Bouillis, B.; Warin, D. [Innovation and Industrial Support Direction, CEA Saclay, F-91191 Gif-sur-Yvette (France)

    2013-07-01T23:59:59.000Z

    Increasing the sustainability of nuclear energy is a longstanding road that requires a stepwise approach to successively tackle the following 3 objectives. First of all, optimize the consumption of natural resource to preserve them for future generations and hence guarantee the energetic independence of the countries (no uranium ore is needed anymore). The current twice-through cycle of Pu implemented by France, UK, Japan and soon China is a first step in this direction and already allows the development and optimization of the relevant industrial processes. It also allows a major improvement regarding the conditioning of the ultimate waste in a durable and robust nuclear glass. Secondly, the recycling of americium could be an interesting option for the future with the deployment of FR fleet to save the repository resource and optimize its use by allowing a denser disposal. It would limit the burden towards the future generations and the need for additional repositories before several centuries. Thirdly, the recycling of the whole minor actinides inventory could be an interesting option for the far-future for strongly decreasing the waste long-term toxicity, down to a few centuries. It would bring the waste issue back within the human history, which should promote its acceptance by the social opinion.

  14. Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors

    SciTech Connect (OSTI)

    Bhatti, Zaki; Hyland, B.; Edwards, G.W.R. [Atomic Energy of Canada Limited, Chalk River Laboratories, 1 Plant Road, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01T23:59:59.000Z

    The irradiation of Th{sup 232} breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U{sup 238}. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in the Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction ?) for coolant voiding as standard NU fuel. (authors)

  15. Actinide production in /sup 136/Xe bombardments of /sup 249/Cf

    SciTech Connect (OSTI)

    Gregorich, K.E.

    1985-08-01T23:59:59.000Z

    The production cross sections for the actinide products from /sup 136/Xe bombardments of /sup 249/Cf at energies 1.02, 1.09, and 1.16 times the Coulomb barrier were determined. Fractions of the individual actinide elements were chemically separated from recoil catcher foils. The production cross sections of the actinide products were determined by measuring the radiations emitted from the nuclides within the chemical fractions. The chemical separation techniques used in this work are described in detail, and a description of the data analysis procedure is included. The actinide production cross section distributions from these /sup 136/Xe + /sup 249/Cf bombardments are compared with the production cross section distributions from other heavy ion bombardments of actinide targets, with emphasis on the comparison with the /sup 136/Xe + /sup 248/Cm reaction. A technique for modeling the final actinide cross section distributions has been developed and is presented. In this model, the initial (before deexcitation) cross section distribution with respect to the separation energy of a dinuclear complex and with respect to the Z of the target-like fragment is given by an empirical procedure. It is then assumed that the N/Z equilibration in the dinuclear complex occurs by the transfer of neutrons between the two participants in the dinuclear complex. The neutrons and the excitation energy are statistically distributed between the two fragments using a simple Fermi gas level density formalism. The resulting target-like fragment initial cross section distribution with respect to Z, N, and excitation energy is then allowed to deexcite by emission of neutrons in competition with fission. The result is a final cross section distribution with respect to Z and N for the actinide products. 68 refs., 33 figs., 6 tabs.

  16. MOLECULAR SPECTROSCPY AND REACTIONS OF ACTINIDES IN THE GAS PHASE AND CRYOGENIC MATRICES

    SciTech Connect (OSTI)

    Heaven, Michael C.; Gibson, John K.; Marcalo, Joaquim

    2009-02-01T23:59:59.000Z

    In this chapter we review the spectroscopic data for actinide molecules and the reaction dynamics for atomic and molecular actinides that have been examined in the gas phase or in inert cryogenic matrices. The motivation for this type of investigation is that physical properties and reactions can be studied in the absence of external perturbations (gas phase) or under minimally perturbing conditions (cryogenic matrices). This information can be compared directly with the results from high-level theoretical models. The interplay between experiment and theory is critically important for advancing our understanding of actinide chemistry. For example, elucidation of the role of the 5f electrons in bonding and reactivity can only be achieved through the application of experimentally verified theoretical models. Theoretical calculations for the actinides are challenging due the large numbers of electrons that must be treated explicitly and the presence of strong relativistic effects. This topic has been reviewed in depth in Chapter 17 of this series. One of the goals of the experimental work described in this chapter has been to provide benchmark data that can be used to evaluate both empirical and ab initio theoretical models. While gas-phase data are the most suitable for comparison with theoretical calculations, there are technical difficulties entailed in generating workable densities of gas-phase actinide molecules that have limited the range of species that have been characterized. Many of the compounds of interest are refractory, and problems associated with the use of high temperature vapors have complicated measurements of spectra, ionization energies, and reactions. One approach that has proved to be especially valuable in overcoming this difficulty has been the use of pulsed laser ablation to generate plumes of vapor from refractory actinide-containing materials. The vapor is entrained in an inert gas, which can be used to cool the actinide species to room temperature or below. For many spectroscopic measurements, low temperatures have been achieved by co-condensing the actinide vapor in rare gas or inert molecule host matrices. Spectra recorded in matrices are usually considered to be minimally perturbed. Trapping the products from gas-phase reactions that occur when trace quantities of reactants are added to the inert host gas has resulted in the discovery of many new actinide species. Selected aspects of the matrix isolation data were discussed in chapter 17. In the present chapter we review the spectroscopic matrix data in terms of its relationship to gas-phase measurements, and update the description of the new reaction products found in matrices to reflect the developments that have occurred during the past two years. Spectra recorded in matrix environments are usually considered to be minimally perturbed, and this expectation is borne out for many closed shell actinide molecules. However, there is growing evidence that significant perturbations can occur for open shell molecules, resulting in geometric distortions and/or electronic state reordering. Studies of actinide reactions in the gas phase provide an opportunity to probe the relationship between electronic structure and reactivity. Much of this work has focused on the reactions of ionic species, as these may be selected and controlled using various forms of mass spectrometry. As an example of the type of insight derived from reaction studies, it has been established that the reaction barriers for An+ ions are determined by the promotion energies required to achieve the 5fn6d7s configuration. Gas-phase reaction studies also provide fundamental thermodynamic properties such as bond dissociation and ionization energies. In recent years, an increased number of gas-phase ion chemistry studies of bare (atomic) and ligated (molecular) actinide ions have appeared, in which relevant contributions to fundamental actinide chemistry have been made. These studies were initiated in the 1970's and carried out in an uninterrupted way over the course of the past three d

  17. Fluorination process using catalysts

    DOE Patents [OSTI]

    Hochel, R.C.; Saturday, K.A.

    1983-08-25T23:59:59.000Z

    A process is given for converting an actinide compound selected from the group consisting of uranium oxides, plutonium oxides, uranium tetrafluorides, plutonium tetrafluorides and mixtures of said oxides and tetrafluorides, to the corresponding volatile actinide hexafluoride by fluorination with a stoichiometric excess of fluorine gas. The improvement involves conducting the fluorination of the plutonium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF/sub 3/, AgF/sub 2/ and NiF/sub 2/, whereby the fluorination is significantly enhanced. The improvement also involves conducting the fluorination of one of the uranium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF/sub 3/ and AgF/sub 2/, whereby the fluorination is significantly enhanced.

  18. Fluorination process using catalyst

    DOE Patents [OSTI]

    Hochel, Robert C. (Aiken, SC); Saturday, Kathy A. (Aiken, SC)

    1985-01-01T23:59:59.000Z

    A process for converting an actinide compound selected from the group consisting of uranium oxides, plutonium oxides, uranium tetrafluorides, plutonium tetrafluorides and mixtures of said oxides and tetrafluorides, to the corresponding volatile actinide hexafluoride by fluorination with a stoichiometric excess of fluorine gas. The improvement involves conducting the fluorination of the plutonium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF.sub.3, AgF.sub.2 and NiF.sub.2, whereby the fluorination is significantly enhanced. The improvement also involves conducting the fluorination of one of the uranium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF.sub.3 and AgF.sub.2, whereby the fluorination is significantly enhanced.

  19. Technical requirements for the actinide source-term waste test program

    SciTech Connect (OSTI)

    Phillips, M.L.F.; Molecke, M.A.

    1993-10-01T23:59:59.000Z

    This document defines the technical requirements for a test program designed to measure time-dependent concentrations of actinide elements from contact-handled transuranic (CH TRU) waste immersed in brines similar to those found in the underground workings of the Waste Isolation Pilot Plant (WIPP). This test program wig determine the influences of TRU waste constituents on the concentrations of dissolved and suspended actinides relevant to the performance of the WIPP. These influences (which include pH, Eh, complexing agents, sorbent phases, and colloidal particles) can affect solubilities and colloidal mobilization of actinides. The test concept involves fully inundating several TRU waste types with simulated WIPP brines in sealed containers and monitoring the concentrations of actinide species in the leachate as a function of time. The results from this program will be used to test numeric models of actinide concentrations derived from laboratory studies. The model is required for WIPP performance assessment with respect to the Environmental Protection Agency`s 40 CFR Part 191B.

  20. Reactor for removing ammonia

    DOE Patents [OSTI]

    Luo, Weifang (Livermore, CA); Stewart, Kenneth D. (Valley Springs, CA)

    2009-11-17T23:59:59.000Z

    Disclosed is a device for removing trace amounts of ammonia from a stream of gas, particularly hydrogen gas, prepared by a reformation apparatus. The apparatus is used to prevent PEM "poisoning" in a fuel cell receiving the incoming hydrogen stream.

  1. Synthesis and development of processes for the recovery of sulfur from acid gases. Part 1, Development of a high-temperature process for removal of H{sub 2}S from coal gas using limestone -- thermodynamic and kinetic considerations; Part 2, Development of a zero-emissions process for recovery of sulfur from acid gas streams

    SciTech Connect (OSTI)

    Towler, G.P.; Lynn, S.

    1993-05-01T23:59:59.000Z

    Limestone can be used more effectively as a sorbent for H{sub 2}S in high-temperature gas-cleaning applications if it is prevented from undergoing calcination. Sorption of H{sub 2}S by limestone is impeded by sintering of the product CaS layer. Sintering of CaS is catalyzed by CO{sub 2}, but is not affected by N{sub 2} or H{sub 2}. The kinetics of CaS sintering was determined for the temperature range 750--900{degrees}C. When hydrogen sulfide is heated above 600{degrees}C in the presence of carbon dioxide elemental sulfur is formed. The rate-limiting step of elemental sulfur formation is thermal decomposition of H{sub 2}S. Part of the hydrogen thereby produced reacts with CO{sub 2}, forming CO via the water-gas-shift reaction. The equilibrium of H{sub 2}S decomposition is therefore shifted to favor the formation of elemental sulfur. The main byproduct is COS, formed by a reaction between CO{sub 2} and H{sub 2}S that is analogous to the water-gas-shift reaction. Smaller amounts of SO{sub 2} and CS{sub 2} also form. Molybdenum disulfide is a strong catalyst for H{sub 2}S decomposition in the presence of CO{sub 2}. A process for recovery of sulfur from H{sub 2}S using this chemistry is as follows: Hydrogen sulfide is heated in a high-temperature reactor in the presence of CO{sub 2} and a suitable catalyst. The primary products of the overall reaction are S{sub 2}, CO, H{sub 2} and H{sub 2}O. Rapid quenching of the reaction mixture to roughly 600{degrees}C prevents loss Of S{sub 2} during cooling. Carbonyl sulfide is removed from the product gas by hydrolysis back to CO{sub 2} and H{sub 2}S. Unreacted CO{sub 2} and H{sub 2}S are removed from the product gas and recycled to the reactor, leaving a gas consisting chiefly of H{sub 2} and CO, which recovers the hydrogen value from the H{sub 2}S. This process is economically favorable compared to the existing sulfur-recovery technology and allows emissions of sulfur-containing gases to be controlled to very low levels.

  2. Integrated pollutant removal: modeling and experimentation

    SciTech Connect (OSTI)

    Ochs, Thomas L.; Oryshchyn, Danylo B.; Summers, Cathy A.

    2005-01-01T23:59:59.000Z

    Experimental and computational work at the Albany Research Center, USDOE is investigating an integrated pollutant removal (IPR) process which removes all pollutants from flue gas, including SOX, NOX, particulates, CO2, and Hg. In combination with flue gas recirculation, heat recovery, and oxy-fuel combustion, the process produces solid, gas, and liquid waste streams. The gas exhaust stream comprises O2 and N2. Liquid streams contain H2O, SOX, NOX, and CO2. Computer modeling and low to moderate pressure experimentation are defining system chemistry with respect to SOX and H2O as well as heat and mass transfer for the IPR process.

  3. Recovering Americium and Curium from Mark-42 Target Materials- New Processing Approaches to Enhance Separations and Integrate Waste Stream Disposition - 12228

    SciTech Connect (OSTI)

    Patton, Brad D.; Benker, Dennis; Collins, Emory D.; Mattus, Catherine H.; Robinson, Sharon M.; Wham, Robert M. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2012-07-01T23:59:59.000Z

    Oak Ridge National Laboratory (ORNL) is investigating flowsheets to enhance processing efficiencies and to address waste streams associated with recovery of americium (Am) and curium (Cm) from Mark-42 (Mk-42) target materials stored at ORNL. The objective of this work was to identify the most effective flowsheet with which to process the 104 Mk-42 oxide capsules holding a total of 80 g of plutonium (Pu), 190 g of Cm, 480 g of Am, and 5 kg of lanthanide (Ln) oxides for the recovery and purification of the Am/Cm for future use as feedstock for heavy actinide production. Studies were also conducted to solidify the process flowsheet waste streams for disposal. ORNL is investigating flowsheets to enhance processing efficiencies and address waste streams associated with recovery of Am and Cm from Mk-42 target materials stored at ORNL. A series of small-scale runs are being performed to demonstrate an improved process to recover Am/Cm and to optimize the separations of Ln fission products from the Am/Cm constituents. The first of these runs has been completed using one of the Am/Cm/Ln oxide capsules stored at ORNL. The demonstration run showed promising results with a Ln DF of 40 for the Am/Cm product and an Am/Cm DF of 75 for the Ln product. In addition, the total losses of Am, Cm, and Ln to the waste solvents and raffinates were very low at <0.2%, 0.02%, and 0.04%, respectively. However, the Ln-actinide separation was less than desired. For future Reverse TALSPEAK demonstration runs, several parameters will be adjusted (flow rates, the ratio of scrub to strip stages, etc.) to improve the removal of Ln from the actinides. The next step will also include scale-up of the processing flowsheet to use more concentrated solutions (15 g/L Ln versus 5 g/L) and larger volumes and to recycle the HDEHP solvent. This should improve the overall processing efficiency and further reduce losses to waste streams. Studies have been performed with simulated wastes to develop solidification processes for disposal of the secondary waste streams generated by this flowsheet. Formulations were successfully developed for all the waste simulants. Additional tests with actual waste will be the next step in this effort. Future plans are to process all of the remaining 103 capsules in storage at ORNL. A nine-capsule test is now under way to provide additional information to scale-up the process to a target 20-capsule batch size for future processing runs. (authors)

  4. Enhancing BWR Proliferation Resistance Fuel with Minor Actinides

    SciTech Connect (OSTI)

    Gray S. Chang

    2008-07-01T23:59:59.000Z

    Key aspects of the Global Nuclear Energy Partnership (GNEP) are to significantly advance the science and technology of nuclear energy systems and the Advanced Fuel Cycle (AFC) program. It consists of both innovative nuclear reactors and innovative research in separation and transmutation. To accomplish these goals, international cooperation is very important and public acceptance is crucial. The merits of nuclear energy are high-density energy, with low environmental impacts (i.e. almost zero greenhouse gas emission). Planned efforts involve near-term and intermediate-term improvements in fuel utilization and recycling in current light water reactors (LWRs) as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The challenges are solving the energy needs of the world, protection against nuclear proliferation, the problem of nuclear waste, and the global environmental problem. To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu and 240Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides (237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu /Pu. For future advanced nuclear systems, the minor actinides (MA) are viewed more as a resource to be recycled, or transmuted to less hazardous and possibly more useful forms, rather than simply as a waste stream to be disposed of in expensive repository facilities. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm3) to the top (0.35 g/cm3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. We concluded that the concept of MARA, which involves the use of transuranic nuclides (237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy rennaissance.

  5. Experimental level-structure determination in odd-odd actinide nuclei

    SciTech Connect (OSTI)

    Hoff, R.W.

    1985-04-04T23:59:59.000Z

    The status of experimental determination of level structure in odd-odd actinide nuclei is reviewed. A technique for modeling quasiparticle excitation energies and rotational parameters in odd-odd deformed nuclei is applied to actinide species where new experimental data have been obtained by use of neutron-capture gamma-ray spectroscopy. The input parameters required for the calculation are derived from empirical data on single-particle excitations in neighboring odd-mass nuclei. Calculated configuration-specific values for the Gallagher-Moszkowski splittings are used. Calculated and experimental level structures for /sup 238/Np, /sup 244/Am, and /sup 250/Bk are compared, as well as those for several nuclei in the rare-earth region. The agreement for the actinide species is excellent, with bandhead energies deviating 22 keV and rotational parameters 5%, on the average. Applications of this modeling technique are discussed.

  6. Minor Actinides Loading Optimization for Proliferation Resistant Fuel Design - BWR

    SciTech Connect (OSTI)

    G. S. Chang; Hongbin Zhang

    2009-09-01T23:59:59.000Z

    One approach to address the United States Nuclear Power (NP) 2010 program for the advanced light water reactor (LWR) (Gen-III+) intermediate-term spent fuel disposal need is to reduce spent fuel storage volume while enhancing proliferation resistance. One proposed solution includes increasing burnup of the discharged spent fuel and mixing minor actinide (MA) transuranic nuclides (237Np and 241Am) in the high burnup fuel. Thus, we can reduce the spent fuel volume while increasing the proliferation resistance by increasing the isotopic ratio of 238Pu/Pu. For future advanced nuclear systems, MAs are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. A typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of adding MAs (237Np and/or 241Am) to enhance proliferation resistance and improve fuel cycle performance for the intermediate-term goal of future nuclear energy systems. However, adding MAs will increase plutonium production in the discharged spent fuel. In this work, the Monte-Carlo coupling with ORIGEN-2.2 (MCWO) method was used to optimize the MA loading in the UO2 fuel such that the discharged spent fuel demonstrates enhanced proliferation resistance, while minimizing plutonium production. The axial averaged MA transmutation characteristics at different burnup were compared and their impact on neutronics criticality and the ratio of 238Pu/Pu discussed.

  7. Identification of process suitable diluent

    SciTech Connect (OSTI)

    Dean R. Peterman

    2014-01-01T23:59:59.000Z

    The Sigma Team for Minor Actinide Separation (STMAS) was formed within the USDOE Fuel Cycle Research and Development (FCRD) program in order to develop more efficient methods for the separation of americium and other minor actinides (MA) from used nuclear fuel. The development of processes for MA separations is driven by the potential benefits; reduced long-term radiotoxicty of waste placed in a geologic repository, reduced timeframe of waste storage, reduced repository heat load, the possibility of increased repository capacity, and increased utilization of energy potential of used nuclear fuel. The research conducted within the STMAS framework is focused upon the realization of significant simplifications to aqueous recycle processes proposed for MA separations. This report describes the research efforts focused upon the identification of a process suitable diluent for a flowsheet concept for the separation of MA which is based upon the dithiophosphinic acid (DPAH) extractants previously developed at the Idaho National Laboratory (INL).

  8. Development and Quantification of UV-Visible and Laser Spectroscopic Techniques for Materials Accountability and Process Control

    SciTech Connect (OSTI)

    Ken Czerwinski; Phil Weck; Frederic Poineau

    2010-12-29T23:59:59.000Z

    Ultraviolet-Visible Spectroscopy (UV-Visible) and Time Resolved Laser Fluorescence Spectroscopy (TRLFS) optical techniques can permit on-line, real-time analysis of the actinide elements in a solvent extraction process. UV-Visible and TRLFS techniques have been used for measuring the speciation and concentration of the actinides under laboratory conditions. These methods are easily adaptable to multiple sampling geometries, such as dip probes, fiber-optic sample cells, and flow-through cell geometries. To fully exploit these techniques for GNEP applications, the fundamental speciation of the target actinides and the resulting influence on 3 spectroscopic properties must be determined. Through this effort detection limits, process conditions, and speciation of key actinide components can be establish and utilized in a range of areas of interest to GNEP, especially in areas related to materials accountability and process control.

  9. Arsenic removal from water

    DOE Patents [OSTI]

    Moore, Robert C. (Edgewood, NM); Anderson, D. Richard (Albuquerque, NM)

    2007-07-24T23:59:59.000Z

    Methods for removing arsenic from water by addition of inexpensive and commonly available magnesium oxide, magnesium hydroxide, calcium oxide, or calcium hydroxide to the water. The hydroxide has a strong chemical affinity for arsenic and rapidly adsorbs arsenic, even in the presence of carbonate in the water. Simple and commercially available mechanical methods for removal of magnesium hydroxide particles with adsorbed arsenic from drinking water can be used, including filtration, dissolved air flotation, vortex separation, or centrifugal separation. A method for continuous removal of arsenic from water is provided. Also provided is a method for concentrating arsenic in a water sample to facilitate quantification of arsenic, by means of magnesium or calcium hydroxide adsorption.

  10. Technetium Removal Using Tc-Goethite Coprecipitation

    SciTech Connect (OSTI)

    Um, Wooyong; Wang, Guohui; Jung, Hun Bok; Peterson, Reid A.

    2013-11-18T23:59:59.000Z

    This report describes the results from laboratory tests performed at Pacific Northwest National Laboratory for the U.S. Department of Energy (DOE) EM-31 Support Program (EMSP) subtask, “Low temperature waste forms coupled with technetium removal using an alternative immobilization process such as Fe(II) treated-goethite precipitation” to increase our understanding of 99Tc long-term stability in goethite mineral form and the process that controls the 99Tc(VII) reduction and removal by the final Fe (oxy)hydroxide forms. The overall objectives of this task were to 1) evaluate the transformation process of Fe (oxy)hydroxide solids to the more crystalline goethite (?-FeOOH) mineral for 99Tc removal and 2) determine the mechanism that limits 99Tc(IV) reoxidation in Fe(II)-treated 99Tc-goethite mineral and 3) evaluate whether there is a long-term 99Tcoxidation state change for Tc sequestered in the iron solids.

  11. Heat recirculating cooler for fluid stream pollutant removal

    DOE Patents [OSTI]

    Richards, George A. (Morgantown, WV); Berry, David A. (Morgantown, WV)

    2008-10-28T23:59:59.000Z

    A process by which heat is removed from a reactant fluid to reach the operating temperature of a known pollutant removal method and said heat is recirculated to raise the temperature of the product fluid. The process can be utilized whenever an intermediate step reaction requires a lower reaction temperature than the prior and next steps. The benefits of a heat-recirculating cooler include the ability to use known pollutant removal methods and increased thermal efficiency of the system.

  12. Organic removal from domestic wastewater by activated alumina adsorption

    E-Print Network [OSTI]

    Yang, Pe-Der

    1982-01-01T23:59:59.000Z

    of the major groups of pollutants in wastewaters. Adsorption by granular activated carbon, a non-polar adsorbent, is now the primary treatment process for removal of residual organics from biologically treated wastewater. The ability of activated alumina... to human health if they exist in the water supply at relatively high concentrations. A wide variety of treatment processes are available to remove organic matter from wastewater. Biological treatment is the most cost effective method for removing oxygen...

  13. FY13 GLYCOLIC-NITRIC ACID FLOWSHEET DEMONSTRATIONS OF THE DWPF CHEMICAL PROCESS CELL WITH SIMULANTS

    SciTech Connect (OSTI)

    Lambert, D.; Zamecnik, J.; Best, D.

    2014-03-13T23:59:59.000Z

    Savannah River Remediation is evaluating changes to its current Defense Waste Processing Facility flowsheet to replace formic acid with glycolic acid in order to improve processing cycle times and decrease by approximately 100x the production of hydrogen, a potentially flammable gas. Higher throughput is needed in the Chemical Processing Cell since the installation of the bubblers into the melter has increased melt rate. Due to the significant maintenance required for the safety significant gas chromatographs and the potential for production of flammable quantities of hydrogen, eliminating the use of formic acid is highly desirable. Previous testing at the Savannah River National Laboratory has shown that replacing formic acid with glycolic acid allows the reduction and removal of mercury without significant catalytic hydrogen generation. Five back-to-back Sludge Receipt and Adjustment Tank (SRAT) cycles and four back-to-back Slurry Mix Evaporator (SME) cycles were successful in demonstrating the viability of the nitric/glycolic acid flowsheet. The testing was completed in FY13 to determine the impact of process heels (approximately 25% of the material is left behind after transfers). In addition, back-to-back experiments might identify longer-term processing problems. The testing was designed to be prototypic by including sludge simulant, Actinide Removal Product simulant, nitric acid, glycolic acid, and Strip Effluent simulant containing Next Generation Solvent in the SRAT processing and SRAT product simulant, decontamination frit slurry, and process frit slurry in the SME processing. A heel was produced in the first cycle and each subsequent cycle utilized the remaining heel from the previous cycle. Lower SRAT purges were utilized due to the low hydrogen generation. Design basis addition rates and boilup rates were used so the processing time was shorter than current processing rates.

  14. Electrochemical/Pyrometallurgical Waste Stream Processing and Waste Form Fabrication

    SciTech Connect (OSTI)

    Steven Frank; Hwan Seo Park; Yung Zun Cho; William Ebert; Brian Riley

    2014-12-01T23:59:59.000Z

    This report summarizes treatment and waste form options being evaluated for waste streams resulting from the electrochemical/pyrometallurgical (pyro ) processing of used oxide nuclear fuel. The technologies that are described are South Korean (Republic of Korea – ROK) and United States of America (US) ‘centric’ in the approach to treating pyroprocessing wastes and are based on the decade long collaborations between US and ROK researchers. Some of the general and advanced technologies described in this report will be demonstrated during the Integrated Recycle Test (IRT) to be conducted as a part of the Joint Fuel Cycle Study (JFCS) collaboration between US Department of Energy (DOE) and ROK national laboratories. The JFCS means to specifically address and evaluated the technological, economic, and safe guard issues associated with the treatment of used nuclear fuel by pyroprocessing. The IRT will involve the processing of commercial, used oxide fuel to recover uranium and transuranics. The recovered transuranics will then be fabricated into metallic fuel and irradiated to transmutate, or burn the transuranic elements to shorter lived radionuclides. In addition, the various process streams will be evaluated and tested for fission product removal, electrolytic salt recycle, minimization of actinide loss to waste streams and waste form fabrication and characterization. This report specifically addresses the production and testing of those waste forms to demonstrate their compatibility with treatment options and suitability for disposal.

  15. Countercurrent flowsheet testing of the TRUEX process with ICPP calcine

    SciTech Connect (OSTI)

    Law, J.D.; Herbst, R.S.; Brewer, K.N.; Todd, T.A.

    1998-07-01T23:59:59.000Z

    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.

  16. Removable feedwater sparger assembly

    DOE Patents [OSTI]

    Challberg, Roy C. (Livermore, CA)

    1994-01-01T23:59:59.000Z

    A removable feedwater sparger assembly includes a sparger having an inlet pipe disposed in flow communication with the outlet end of a supply pipe. A tubular coupling includes an annular band fixedly joined to the sparger inlet pipe and a plurality of fingers extending from the band which are removably joined to a retention flange extending from the supply pipe for maintaining the sparger inlet pipe in flow communication with the supply pipe. The fingers are elastically deflectable for allowing engagement of the sparger inlet pipe with the supply pipe and for disengagement therewith.

  17. Drum lid removal tool

    DOE Patents [OSTI]

    Pella, Bernard M. (Martinez, GA); Smith, Philip D. (North Augusta, SC)

    2010-08-24T23:59:59.000Z

    A tool for removing the lid of a metal drum wherein the lid is clamped over the drum rim without protruding edges, the tool having an elongated handle with a blade carried by an angularly positioned holder affixed to the midsection of the handle, the blade being of selected width to slice between lid lip and the drum rim and, when the blade is so positioned, upward motion of the blade handle will cause the blade to pry the lip from the rim and allow the lid to be removed.

  18. Removable feedwater sparger assembly

    DOE Patents [OSTI]

    Challberg, R.C.

    1994-10-04T23:59:59.000Z

    A removable feedwater sparger assembly includes a sparger having an inlet pipe disposed in flow communication with the outlet end of a supply pipe. A tubular coupling includes an annular band fixedly joined to the sparger inlet pipe and a plurality of fingers extending from the band which are removably joined to a retention flange extending from the supply pipe for maintaining the sparger inlet pipe in flow communication with the supply pipe. The fingers are elastically deflectable for allowing engagement of the sparger inlet pipe with the supply pipe and for disengagement therewith. 8 figs.

  19. Analysis on fuel breeding capability of FBR core region based on minor actinide recycling doping

    SciTech Connect (OSTI)

    Permana, Sidik; Novitrian,; Waris, Abdul [Nuclear Physics and Biophysics Research Division, Physics Department, Institut Teknologi Bandung (Indonesia); Ismail [Center for Technical Assessment of Nuclear Installation and Materials, Indonesian Nuclear Energy Regulatory (Indonesia); Suzuki, Mitsutoshi [Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA) (Japan); Saito, Masaki [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology (Japan)

    2014-09-30T23:59:59.000Z

    Nuclear fuel breeding based on the capability of fuel conversion capability can be achieved by conversion ratio of some fertile materials into fissile materials during nuclear reaction processes such as main fissile materials of U-233, U-235, Pu-239 and Pu-241 and for fertile materials of Th-232, U-238, and Pu-240 as well as Pu-238. Minor actinide (MA) loading option which consists of neptunium, americium and curium will gives some additional contribution from converted MA into plutonium such as conversion Np-237 into Pu-238 and it's produced Pu-238 converts to Pu-239 via neutron capture. Increasing composition of Pu-238 can be used to produce fissile material of Pu-239 as additional contribution. Trans-uranium (TRU) fuel (Mixed fuel loading of MOX (U-Pu) and MA composition) and mixed oxide (MOX) fuel compositions are analyzed for comparative analysis in order to show the effect of MA to the plutonium productions in core in term of reactor criticality condition and fuel breeding capability. In the present study, neptunium (Np) nuclide is used as a representative of MAin trans-uranium (TRU) fuel composition as Np-MOX fuel type. It was loaded into the core region gives significant contribution to reduce the excess reactivity in comparing to mixed oxide (MOX) fuel and in the same time it contributes to increase nuclear fuel breeding capability of the reactor. Neptunium fuel loading scheme in FBR core region gives significant production of Pu-238 as fertile material to absorp neutrons for reducing excess reactivity and additional contribution for fuel breeding.

  20. Membrane Based intensification of ammonia removal from wastewater

    E-Print Network [OSTI]

    Almutairi, Azel

    2011-12-31T23:59:59.000Z

    The aim of this research was to study a novel membrane based oxygen intensification system to enhance a biological wastewater treatment process for ammonia removal. Specifically, this work is concerned with the biological nitrification process which...

  1. Water Distribution and Removal Model

    SciTech Connect (OSTI)

    Y. Deng; N. Chipman; E.L. Hardin

    2005-08-26T23:59:59.000Z

    The design of the Yucca Mountain high level radioactive waste repository depends on the performance of the engineered barrier system (EBS). To support the total system performance assessment (TSPA), the Engineered Barrier System Degradation, Flow, and Transport Process Model Report (EBS PMR) is developed to describe the thermal, mechanical, chemical, hydrological, biological, and radionuclide transport processes within the emplacement drifts, which includes the following major analysis/model reports (AMRs): (1) EBS Water Distribution and Removal (WD&R) Model; (2) EBS Physical and Chemical Environment (P&CE) Model; (3) EBS Radionuclide Transport (EBS RNT) Model; and (4) EBS Multiscale Thermohydrologic (TH) Model. Technical information, including data, analyses, models, software, and supporting documents will be provided to defend the applicability of these models for their intended purpose of evaluating the postclosure performance of the Yucca Mountain repository system. The WD&R model ARM is important to the site recommendation. Water distribution and removal represents one component of the overall EBS. Under some conditions, liquid water will seep into emplacement drifts through fractures in the host rock and move generally downward, potentially contacting waste packages. After waste packages are breached by corrosion, some of this seepage water will contact the waste, dissolve or suspend radionuclides, and ultimately carry radionuclides through the EBS to the near-field host rock. Lateral diversion of liquid water within the drift will occur at the inner drift surface, and more significantly from the operation of engineered structures such as drip shields and the outer surface of waste packages. If most of the seepage flux can be diverted laterally and removed from the drifts before contacting the wastes, the release of radionuclides from the EBS can be controlled, resulting in a proportional reduction in dose release at the accessible environment. The purposes of this WD&R model (CRWMS M&O 2000b) are to quantify and evaluate the distribution and drainage of seepage water within emplacement drifts during the period of compliance for post-closure performance. The model bounds the fraction of water entering the drift that will be prevented from contacting the waste by the combined effects of engineered controls on water distribution and on water removal. For example, water can be removed during pre-closure operation by ventilation and after closure by natural drainage into the fractured rock. Engineered drains could be used, if demonstrated to be necessary and effective, to ensure that adequate drainage capacity is provided. This report provides the screening arguments for certain Features, Events, and Processes (FEPs) that are related to water distribution and removal in the EBS. Applicable acceptance criteria from the Issue Resolution Status Reports (IRSRs) developed by the U.S. Nuclear Regulatory Commission (NRC 1999a; 1999b; 1999c; and 1999d) are also addressed in this document.

  2. ELECTRONIC STRUCTURE IN METALS AND ALLOYS. ELECTRONIC STRUCTURE OF THE LIGHT ACTINIDES

    E-Print Network [OSTI]

    Boyer, Edmond

    ELECTRONIC STRUCTURE IN METALS AND ALLOYS. ELECTRONIC STRUCTURE OF THE LIGHT ACTINIDES B. D. DUNLAP electrons. A review is given of some areas of current interest, especially where hyperfine techniques have the 60 keV y-ray of 237Np[l]. At that time, our understanding of the electronic properties

  3. Assessment of SFR fuel pin performance codes under advanced fuel for minor actinide transmutation

    SciTech Connect (OSTI)

    Bouineau, V.; Lainet, M.; Chauvin, N.; Pelletier, M. [French Alternative Energies and Atomic Energy Commission - CEA, CEA Cadarache, DEN/DEC/SESC, 13108 Saint Paul lez Durance (France); Di Marcello, V.; Van Uffelen, P.; Walker, C. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz-Platz 1, D- 76344 Eggenstein-Leopoldshafen (Germany)

    2013-07-01T23:59:59.000Z

    Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like {sup 241}Am is, therefore, an option for the reduction of radiotoxicity and residual power packages as well as the repository area. In the SUPERFACT Experiment four different oxide fuels containing high and low concentrations of {sup 237}Np and {sup 241}Am, representing the homogeneous and heterogeneous in-pile recycling concepts, were irradiated in the PHENIX reactor. The behavior of advanced fuel materials with minor actinide needs to be fully characterized, understood and modeled in order to optimize the design of this kind of fuel elements and to evaluate its performances. This paper assesses the current predictability of fuel performance codes TRANSURANUS and GERMINAL V2 on the basis of post irradiation examinations of the SUPERFACT experiment for pins with low minor actinide content. Their predictions have been compared to measured data in terms of geometrical changes of fuel and cladding, fission gases behavior and actinide and fission product distributions. The results are in good agreement with the experimental results, although improvements are also pointed out for further studies, especially if larger content of minor actinide will be taken into account in the codes. (authors)

  4. Pneumatic soil removal tool

    DOE Patents [OSTI]

    Neuhaus, J.E.

    1992-10-13T23:59:59.000Z

    A soil removal tool is provided for removing radioactive soil, rock and other debris from the bottom of an excavation, while permitting the operator to be located outside of a containment for that excavation. The tool includes a fixed jaw, secured to one end of an elongate pipe, which cooperates with a movable jaw pivotably mounted on the pipe. Movement of the movable jaw is controlled by a pneumatic cylinder mounted on the pipe. The actuator rod of the pneumatic cylinder is connected to a collar which is slidably mounted on the pipe and forms part of the pivotable mounting assembly for the movable jaw. Air is supplied to the pneumatic cylinder through a handle connected to the pipe, under the control of an actuator valve mounted on the handle, to provide movement of the movable jaw. 3 figs.

  5. Pneumatic soil removal tool

    DOE Patents [OSTI]

    Neuhaus, John E. (Newport News, VA)

    1992-01-01T23:59:59.000Z

    A soil removal tool is provided for removing radioactive soil, rock and other debris from the bottom of an excavation, while permitting the operator to be located outside of a containment for that excavation. The tool includes a fixed jaw, secured to one end of an elongate pipe, which cooperates with a movable jaw pivotably mounted on the pipe. Movement of the movable jaw is controlled by a pneumatic cylinder mounted on the pipe. The actuator rod of the pneumatic cylinder is connected to a collar which is slidably mounted on the pipe and forms part of the pivotable mounting assembly for the movable jaw. Air is supplied to the pneumatic cylinder through a handle connected to the pipe, under the control of an actuator valve mounted on the handle, to provide movement of the movable jaw.

  6. Melter Glass Removal and Dismantlement

    SciTech Connect (OSTI)

    Richardson, BS

    2000-10-31T23:59:59.000Z

    The U.S. Department of Energy (DOE) has been using vitrification processes to convert high-level radioactive waste forms into a stable glass for disposal in waste repositories. Vitrification facilities at the Savannah River Site (SRS) and at the West Valley Demonstration Project (WVDP) are converting liquid high-level waste (HLW) by combining it with a glass-forming media to form a borosilicate glass, which will ensure safe long-term storage. Large, slurry fed melters, which are used for this process, were anticipated to have a finite life (on the order of two to three years) at which time they would have to be replaced using remote methods because of the high radiation fields. In actuality the melters useable life spans have, to date, exceeded original life-span estimates. Initial plans called for the removal of failed melters by placing the melter assembly into a container and storing the assembly in a concrete vault on the vitrification plant site pending size-reduction, segregation, containerization, and shipment to appropriate storage facilities. Separate facilities for the processing of the failed melters currently do not exist. Options for handling these melters include (1) locating a facility to conduct the size-reduction, characterization, and containerization as originally planned; (2) long-term storing or disposing of the complete melter assembly; and (3) attempting to refurbish the melter and to reuse the melter assembly. The focus of this report is to look at methods and issues pertinent to size-reduction and/or melter refurbishment in particular, removing the glass as a part of a refurbishment or to reduce contamination levels (thus allowing for disposal of a greater proportion of the melter as low level waste).

  7. KKG Group Paraffin Removal

    SciTech Connect (OSTI)

    Schulte, Ralph

    2001-12-01T23:59:59.000Z

    The Rocky Mountain Oilfield Testing Center (RMOTC) has recently completed a test of a paraffin removal system developed by the KKG Group utilizing the technology of two Russian scientists, Gennady Katzyn and Boris Koggi. The system consisting of chemical ''sticks'' that generate heat in-situ to melt the paraffin deposits in oilfield tubing. The melted paraffin is then brought to the surface utilizing the naturally flowing energy of the well.

  8. CX-002530: Categorical Exclusion Determination | Department of...

    Broader source: Energy.gov (indexed) [DOE]

    Salt Processing (MSP) Project seeks to deploy equipment to remove the cesium (Cs), strontium (Sr), and select actinides from the high level waste salt solutions using existing...

  9. Gadolinium speciation with Tetradentate, N-donor extractants for minor actinide/lanthanide separation: an XRD, mass spectrometry and EPR study

    SciTech Connect (OSTI)

    Whittaker, D.M. [School of Chemistry, The University of Manchester, Oxford Road, Manchester, M13 9PL (United Kingdom); Sharrad, C.A. [School of Chemical Engineering and Analytical Science, The University of Manchester, Oxford Road, Manchester M13 9PL (United Kingdom); Research Centre for Radwaste and Decommissioning, Dalton Nuclear Institute, The University of Manchester, Oxford Road, Manchester M13 9PL (United Kingdom); Sproules, S. [Photon Science Institute, The University of Manchester, Oxford Road, Manchester M13 9PL (United Kingdom); WestCHEM, School of Chemistry, University of Glasgow, Glasgow G12 8QQ (United Kingdom)

    2013-07-01T23:59:59.000Z

    The hydrophobic organic molecules CyMe{sub 4}-BTPhen (1) and CyMe{sub 4}-BTBP (2) have been developed and tuned over many years to be able to separate the trivalent actinides from the trivalent lanthanides (Ln) selectively in bi-phasic solvent extraction processes for the separation of the long-lived radio-toxic minor actinides from spent nuclear fuel. The ability of these N-donor ligands to perform this separation is poorly understood, as is their speciation with the metal ions when extracted into the organic phase. Our previous work has shown Ln{sup 3+} speciation to be largely 1:2 Ln:L in nature with another small molecule, either water or nitrate, occupying a cavity between the tetradentate bound N-donor ligands. The identity of the small molecule changes across the lanthanide series, and here we continue investigations into this speciation. Complexes of these N-donor ligands with Gd{sup 3+} have been synthesised and characterised by X-ray crystallography, mass spectrometry and EPR spectroscopy. We show that the N-donor ligands have no effect on the electronic configuration of Gd{sup 3+} and that the lanthanide contraction with the steric rigidity of the N-donor ligand appears to determine the size of the cavity between the coordinated ligands. This in turn appears to control the identity of the small molecule on the ninth site in the 1:2 Gd:L species. (authors)

  10. Thiacrown polymers for removal of mercury from waste streams

    DOE Patents [OSTI]

    Baumann, Theodore F. (Tracy, CA); Reynolds, John G. (San Ramon, CA); Fox, Glenn A. (Livermore, CA)

    2002-01-01T23:59:59.000Z

    Thiacrown polymers immobilized to a polystyrene-divinylbenzene matrix react with Hg.sup.2+ under a variety of conditions to efficiently and selectively remove Hg.sup.2+ ions from acidic aqueous solutions, even in the presence of a variety of other metal ions. The mercury can be recovered and the polymer regenerated. This mercury removal method has utility in the treatment of industrial wastewater, where a selective and cost-effective removal process is required.

  11. Heat treatment of exchangers to remove coke

    SciTech Connect (OSTI)

    Turner, J.D.

    1990-02-20T23:59:59.000Z

    This patent describes a process for preparing furfural coke for removal from metallic surfaces. It comprises: heating the furfural coke without causing an evolution of heat capable of undesirably altering metallurgical properties of the surfaces in the presence of a gas containing molecular oxygen at a sufficient temperature below 800{degrees}F (427{degrees}C) for a sufficient time to change the crush strength of the coke so as to permit removal with a water jet at a pressure of five thousand pounds per square inch.

  12. DISTRIBUTION OF LANTHANIDE AND ACTINIDE ELEMENTS BETWEEN BIS-(2-ETHYLHEXYL)PHOSPHORIC ACID AND BUFFERED LACTATE SOLUTIONS CONTAINING SELECTED COMPLEXANTS

    SciTech Connect (OSTI)

    Rudisill, Tracy S.; Diprete, David P.; Thompson, Major C.

    2013-04-15T23:59:59.000Z

    With the renewed interest in the closure of the nuclear fuel cycle, the TALSPEAK process is being considered for the separation of Am and Cm from the lanthanide fission products in a next generation reprocessing plant. However, an efficient separation requires tight control of the pH which likely will be difficult to achieve on a large scale. To address this issue, we measured the distribution of lanthanide and actinide elements between aqueous and organic phases in the presence of complexants which were potentially less sensitive to pH control than the diethylenetriaminepentaacetic (DTPA) used in the process. To perform the extractions, a rapid and accurate method was developed for measuring distribution coefficients based on the preparation of lanthanide tracers in the Savannah River National Laboratory neutron activation analysis facility. The complexants tested included aceto-, benzo-, and salicylhydroxamic acids, N,N,N',N'-tetrakis(2-pyridylmethyl)ethylenediamine (TPEN), and ammonium thiocyanate (NH{sub 4}SCN). The hydroxamic acids were the least effective of the complexants tested. The separation factors for TPEN and NH{sub 4}SCN were higher, especially for the heaviest lanthanides in the series; however, no conditions were identified which resulted in separations factors which consistently approached those measured for the use of DTPA.

  13. Method for removing chlorine compounds from hydrocarbon mixtures

    DOE Patents [OSTI]

    Janoski, E.J.; Hollstein, E.J.

    1984-09-29T23:59:59.000Z

    A process for removing halide ions from a hydrocarbon feedstream containing halogenated hydrocarbons wherein the contaminated feedstock is contacted with a solution of a suitable oxidizing acid containing a lanthanide oxide, the acid being present in a concentration of at least about 50 weight percent for a time sufficient to remove substantially all of the halide ion from the hydrocarbon feedstock.

  14. Test of Actinide-Lanthanide Separation in an Aluminum-Based Pyrochemical System

    SciTech Connect (OSTI)

    Rault, Laurence [Institut National Polytechnique de Grenoble (France); Heusch, Murielle [Institut National Polytechnique de Grenoble (France); Allibert, Michel [Institut National Polytechnique de Grenoble (France); Lemort, Florent [Commissariat a l'Energie Atomique (CEA) (France); Deschane, Xavier [Commissariat a l'Energie Atomique (CEA) (France); Boen, Roger [Commissariat a l'Energie Atomique (CEA) (France)

    2002-08-15T23:59:59.000Z

    The investigation of the actinide and lanthanide distribution between a liquid metal and a molten fluoride salt shows a significant increase of the separation coefficient by using an aluminum-based pyrochemical system instead of a zinc-based system. The obtained values partly depend on the LiF/AlF{sub 3} ratio and can reach more than 30 000 when AlF{sub 3} is in excess with regard to the formation of the cryolite (Li{sub 3} AlF{sub 6}). Furthermore, in the metal phase, the aluminum interacts with the lanthanides to a lesser extent than in other usual metallic solvents. This opens a new way to explore the feasibility of the separation of actinides and lanthanides in the field of nuclear fuel reprocessing.

  15. Evaluation of Fluid Conduction and Mixing within a Subassembly of the Actinide Burner Test Reactor

    SciTech Connect (OSTI)

    Cliff B. Davis

    2007-09-01T23:59:59.000Z

    The RELAP5-3D code is being considered as a thermal-hydraulic system code to support the development of the sodium-cooled Actinide Burner Test Reactor as part of the Global Nuclear Energy Partnership. An evaluation was performed to determine whether the control system could be used to simulate the effects of non-convective mechanisms of heat transport in the fluid, including axial and radial heat conduction and subchannel mixing, that are not currently represented with internal code models. The evaluation also determined the relative importance of axial and radial heat conduction and fluid mixing on peak cladding temperature for a wide range of steady conditions and during a representative loss-of-flow transient. The evaluation was performed using a RELAP5-3D model of a subassembly in the Experimental Breeder Reactor-II, which was used as a surrogate for the Actinide Burner Test Reactor.

  16. Strategic Design and Optimization of Inorganic Sorbents For Cesium, Strontium and Actinides

    SciTech Connect (OSTI)

    Hobbs, D.; Nyman, M.; Clearfield, A.; Maginn, E.

    2006-06-01T23:59:59.000Z

    The basic science goal in this project identifies structure/affinity relationships for selected radionuclides and existing sorbents. The task will apply this knowledge to the design and synthesis of new sorbents that will exhibit increased affinity for cesium, strontium and actinide separations. The target problem focuses on the treatment of high-level nuclear wastes. The general approach can likewise be applied to nonradioactive separations.

  17. Electronic Structure of Transition Metal Clusters and Actinide Complexes and Their Reactivity

    SciTech Connect (OSTI)

    Balasubramanian, K

    2008-10-06T23:59:59.000Z

    Our research in this area since October 2007 has resulted in seven completed publications and more papers of the completed work are in progress. Our work during this period principally focused on actinide complexes with secondary emphasis on spectroscopic properties and electronic structure of metal complexes. As the publications are available online with all of the details of the results, tables and figures, we are providing here only a brief summary of major highlights, in each of the categories.

  18. QUANTIFICATION OF ACTINIDE ALPHA-RADIATION DAMAGE IN MINERALS AND CERAMICS

    SciTech Connect (OSTI)

    Farnan, Ian E.; Cho, Herman M.; Weber, William J.

    2007-01-11T23:59:59.000Z

    There are large amounts of heavy alpha-emitters in nuclear waste and nuclear materials inventories stored in various sites around the world. These include plutonium and minor actinides such as americium and curium. In preparation for geological disposal there is a consensus that actinides that have been separated from spent nuclear fuel should be immobilised within mineral-based ceramics rather than glass. Over the long-term, the alpha-decay taking place in these ceramics will severely disrupt their crystalline structure and reduce their durability. A fundamental property in predicting cumulative radiation damage is the number of atoms permanently displaced per alpha–decay. Currently, this number is estimated as 1000-2000 atoms/alpha decay event. Here, we report nuclear magnetic resonance, spin-counting experiments that measure close to 5000 atoms/alpha decay event in radiation damaged natural zircons. New radiological NMR measurements on highly radioactive, 239Pu zircon show damage similar to that created by 238U and 232Th in mineral zircons at the same dose, indicating no significant effect of dose rate. Based on these measurements, the initially crystalline structure of a 10 wt% 239Pu zircon would be amorphous after only 1400 years in a geological repository. These measurements establish a basis for assessing the long-term structural durability of actinide-containing ceramics based on an atomistic understanding of the fundamental damage event.

  19. Dissolution of metal oxides and separation of uranium from lanthanides and actinides in supercritical carbon dioxide

    SciTech Connect (OSTI)

    Quach, D.L.; Wai, C.M. [Department of Chemistry, University of Idaho, Moscow, Idaho 83844 (United States); Mincher, B.J. [Idaho National Lab, Idaho Falls, Idaho (United States)

    2013-07-01T23:59:59.000Z

    This paper investigates the feasibility of extracting and separating uranium from lanthanides and other actinides by using supercritical fluid carbon dioxide (sc-CO{sub 2}) as a solvent modified with tri-n-butylphosphate (TBP) for the development of a counter current stripping technique, which would be a more efficient and environmentally benign technology for spent nuclear fuel reprocessing compared to traditional solvent extraction. Several actinides (U, Pu, and Np) and europium were extracted in sc-CO{sub 2} modified with TBP over a range of nitric acid concentrations and then the actinides were exposed to reducing and complexing agents to suppress their extractability. According to this study, uranium/europium and uranium/plutonium extraction and separation in sc-CO{sub 2} modified with TBP is successful at nitric acid concentrations of less than 6 M and at nitric acid concentrations of less than 3 M with acetohydroxamic acid or oxalic acid, respectively. A scheme for recycling uranium from spent nuclear fuel by using sc-CO{sub 2} and counter current stripping columns is presented. (authors)

  20. Actinide chemistry research supporting the Waste Isolation Pilot Plant (WIPP): FY94 results

    SciTech Connect (OSTI)

    Novak, C.F. [ed.

    1995-08-01T23:59:59.000Z

    This document contains six reports on actinide chemistry research supporting the Waste Isolation Pilot Plant (WIPP). These reports, completed in FY94, are relevant to the estimation of the potential dissolved actinide concentrations in WIPP brines under repository breach scenarios. Estimates of potential dissolved actinide concentrations are necessary for WIPP performance assessment calculations. The specific topics covered within this document are: the complexation of oxalate with Th(IV) and U(VI); the stability of Pu(VI) in one WIPP-specific brine environment both with and without carbonate present; the solubility of Nd(III) in a WIPP Salado brine surrogate as a function of hydrogen ion concentration; the steady-state dissolved plutonium concentrations in a synthetic WIPP Culebra brine surrogate; the development of a model for Nd(III) solubility and speciation in dilute to concentrated sodium carbonate and sodium bicarbonate solutions; and the development of a model for Np(V) solubility and speciation in dilute to concentrated sodium Perchlorate, sodium carbonate, and sodium chloride media.

  1. DISSOLUTION OF METAL OXIDES AND SEPARATION OF URANIUM FROM LANTHANIDES AND ACTINIDES IN SUPERCRITICAL CARBON DIOXIDE

    SciTech Connect (OSTI)

    Donna L. Quach; Bruce J. Mincher; Chien M. Wai

    2013-10-01T23:59:59.000Z

    This paper investigates the feasibility of extracting and separating uranium from lanthanides and other actinides by using supercritical fluid carbon dioxide (sc-CO2) as a solvent modified with tri-n-butylphosphate (TBP) for the development of a counter current stripping technique, which would be a more efficient and environmentally benign technology for spent nuclear fuel reprocessing compared to traditional solvent extraction. Several actinides (U, Pu, and Np) and europium were extracted in sc-CO2 modified with TBP over a range of nitric acid concentrations and then the actinides were exposed to reducing and complexing agents to suppress their extractability. According to this study, uranium/europium and uranium/plutonium extraction and separation in sc-CO2 modified with TBP is successful at nitric acid concentrations of less than 6 M and at nitric acid concentrations of less than 3 M with acetohydroxamic acid or oxalic acid, respectively. A scheme for recycling uranium from spent nuclear fuel by using sc-CO2 and counter current stripping columns is presented.

  2. Fundamental Thermodynamics of Actinide-Bearing Mineral Waste Forms - Final Report

    SciTech Connect (OSTI)

    Williamson, Mark A.; Ebbinghaus, Bartley B.; Navrotsky, Alexandra

    2001-03-01T23:59:59.000Z

    The end of the Cold War raised the need for the technical community to be concerned with the disposition of excess nuclear weapon material. The plutonium will either be converted into mixed-oxide fuel for use in nuclear reactors or immobilized in glass or ceramic waste forms and placed in a repository. The stability and behavior of plutonium in the ceramic materials as well as the phase behavior and stability of the ceramic material in the environment is not well established. In order to provide technically sound solutions to these issues, thermodynamic data are essential in developing an understanding of the chemistry and phase equilibria of the actinide-bearing mineral waste form materials proposed as immobilization matrices. Mineral materials of interest include zircon, zirconolite, and pyrochlore. High temperature solution calorimetry is one of the most powerful techniques, sometimes the only technique, for providing the fundamental thermodynamic data needed to establish optimum material fabrication parameters, and more importantly understand and predict the behavior of the mineral materials in the environment. The purpose of this project is to experimentally determine the enthalpy of formation of actinide orthosilicates, the enthalpies of formation of actinide substituted zirconolite and pyrochlore, and develop an understanding of the bonding characteristics and stabilities of these materials.

  3. Ionic Liquid and Supercritical Fluid Hyphenated Techniques for Dissolution and Separation of Lanthanides, Actinides, and Fission Products

    SciTech Connect (OSTI)

    Wai, Chien M. [Univ. of Idaho, Moscow, ID (United States); Bruce Mincher

    2012-12-01T23:59:59.000Z

    This project is investigating techniques involving ionic liquids (IL) and supercritical (SC) fluids for dissolution and separation of lanthanides, actinides, and fission products. The research project consists of the following tasks: Study direct dissolution of lanthanide oxides, uranium dioxide and other actinide oxides in [bmin][Tf{sub 2}N] with TBP(HNO{sub 3}){sub 1.8}(H{sub 2}O){sub 0.6} and similar types of Lewis acid-Lewis base complexing agents; Measure distributions of dissolved metal species between the IL and the sc-CO{sub 2} phases under various temperature and pressure conditions; Investigate the chemistry of the dissolved metal species in both IL and sc-CO{sub 2} phases using spectroscopic and chemical methods; Evaluate potential applications of the new extraction techniques for nuclear waste management and for other projects. Supercritical carbon dioxide (sc-CO{sub 2}) and ionic liquids are considered green solvents for chemical reactions and separations. Above the critical point, CO{sub 2} has both gas- and liquid-like properties, making it capable of penetrating small pores of solids and dissolving organic compounds in the solid matrix. One application of sc-CO{sub 2} extraction technology is nuclear waste management. Ionic liquids are low-melting salts composed of an organic cation and an anion of various forms, with unique properties making them attractive replacements for the volatile organic solvents traditionally used in liquid-liquid extraction processes. One type of room temperature ionic liquid (RTIL) based on the 1-alkyl-3-methylimidazolium cation [bmin] with bis(trifluoromethylsulfonyl)imide anion [Tf{sub 2}N] is of particular interest for extraction of metal ions due to its water stability, relative low viscosity, high conductivity, and good electrochemical and thermal stability. Recent studies indicate that a coupled IL sc-CO{sub 2} extraction system can effectively transfer trivalent lanthanide and uranyl ions from nitric acid solutions. Advantages of this technique include operation at ambient temperature and pressure, selective extraction due to tunable sc-CO{sub 2} solvation strength, no IL loss during back-extraction, and no organic solvent introduced into the IL phase.

  4. Method to remove uranium/vanadium contamination from groundwater

    DOE Patents [OSTI]

    Metzler, Donald R. (DeBeque, CO); Morrison, Stanley (Grand Junction, CO)

    2004-07-27T23:59:59.000Z

    A process for removing uranium/vanadium-based contaminants from groundwater using a primary in-ground treatment media and a pretreatment media that chemically adjusts the groundwater contaminant to provide for optimum treatment by the primary treatment media.

  5. Method to Remove Uranium/Vanadium Contamination from Groundwater

    DOE Patents [OSTI]

    Metzler, Donald R.; Morrison Stanley

    2004-07-27T23:59:59.000Z

    A process for removing uranium/vanadium-based contaminants from groundwater using a primary in-ground treatment media and a pretreatment media that chemically adjusts the groundwater contaminant to provide for optimum treatment by the primary treatment media.

  6. Removal of arsenic compounds from petroliferous liquids

    DOE Patents [OSTI]

    Fish, Richard H. (Berkeley, CA)

    1985-01-01T23:59:59.000Z

    Described is a process for removing arsenic from petroliferous derived liquids by contacting said liquid at an elevated temperature with a divinylbenzene-crosslinked polystyrene having catechol ligands anchored thereon. Also, described is a process for regenerating spent catecholated polystyrene by removal of the arsenic bound to it from contacting petroliferous liquid as described above and involves: a. treating said spent catecholated polystyrene, at a temperature in the range of about 20.degree. to 100.degree. C. with an aqueous solution of at least one carbonate and/or bicarbonate of ammonium, alkali and alkaline earth metals, said solution having a pH between about 8 and 10 and, b. separating the solids and liquids from each other. Preferably the regeneration treatment is in two steps wherein step (a) is carried out with an aqueous alcoholic carbonate solution containing lower alkyl alcohol, and, steps (a) and (b) are repeated using a bicarbonate.

  7. Removing Arsenic from Drinking Water

    ScienceCinema (OSTI)

    None

    2013-05-28T23:59:59.000Z

    See how INL scientists are using nanotechnology to remove arsenic from drinking water. For more INL research, visit http://www.facebook.com/idahonationallaboratory

  8. Removal of copper from ferrous scrap

    DOE Patents [OSTI]

    Blander, M.; Sinha, S.N.

    1987-07-30T23:59:59.000Z

    A process for removing copper from ferrous or other metal scrap in which the scrap is contacted with a polyvalent metal sulfide slag in the presence of an excess of copper-sulfide forming additive to convert the copper to copper sulfide which is extracted into the slag to provide a ratio of copper in the slag to copper in the metal scrap of at least about 10.

  9. Removal of copper from ferrous scrap

    DOE Patents [OSTI]

    Blander, M.; Sinha, S.N.

    1990-05-15T23:59:59.000Z

    A process for removing copper from ferrous or other metal scrap in which the scrap is contacted with a polyvalent metal sulfide slag in the presence of an excess of copper-sulfide forming additive to convert the copper to copper sulfide which is extracted into the slag to provide a ratio of copper in the slag to copper in the metal scrap of at least about 10.

  10. Removal of copper from ferrous scrap

    DOE Patents [OSTI]

    Blander, Milton (12833 S. 82nd Ct., Palos Park, IL 60464); Sinha, Shome N. (5748 Drexel, 2A, Chicago, IL 60637)

    1990-01-01T23:59:59.000Z

    A process for removing copper from ferrous or other metal scrap in which the scrap is contacted with a polyvalent metal sulfide slag in the presence of an excess of copper-sulfide forming additive to convert the copper to copper sulfide which is extracted into the slag to provide a ratio of copper in the slag to copper in the metal scrap of at least about 10.

  11. Utilization of Minor Actinides as a Fuel Component for Ultra-Long Life Bhr Configurations: Designs, Advantages and Limitations

    SciTech Connect (OSTI)

    Dr. Pavel V. Tsvetkov

    2009-05-20T23:59:59.000Z

    This project assessed the advantages and limitations of using minor actinides as a fuel component to achieve ultra-long life Very High Temperature Reactor (VHTR) configurations. Researchers considered and compared the capabilities of pebble-bed and prismatic core designs with advanced actinide fuels to achieve ultra-long operation without refueling. Since both core designs permit flexibility in component configuration, fuel utilization, and fuel management, it is possible to improve fissile properties of minor actinides by neutron spectrum shifting through configuration adjustments. The project studied advanced actinide fuels, which could reduce the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository and enable recovery of the energy contained in spent fuel. The ultra-long core life autonomous approach may reduce the technical need for additional repositories and is capable to improve marketability of the Generation IV VHTR by allowing worldwide deployment, including remote regions and regions with limited industrial resources. Utilization of minor actinides in nuclear reactors facilitates developments of new fuel cycles towards sustainable nuclear energy scenarios.

  12. THE HYDROTHERMAL REACTIONS OF MONOSODIUM TITANATE, CRYSTALLINE SILICOTITANATE AND SLUDGE IN THE MODULAR SALT PROCESS: A LITERATURE SURVEY

    SciTech Connect (OSTI)

    Fondeur, F.; Pennebaker, F.; Fink, S.

    2010-11-11T23:59:59.000Z

    The use of crystalline silicotitanate (CST) is proposed for an at-tank process to treat High Level Waste at the Savannah River Site. The proposed configuration includes deployment of ion exchange columns suspended in the risers of existing tanks to process salt waste without building a new facility. The CST is available in an engineered form, designated as IE-911-CW, from UOP. Prior data indicates CST has a proclivity to agglomerate from deposits of silica rich compounds present in the alkaline waste solutions. This report documents the prior literature and provides guidance for the design and operations that include CST to mitigate that risk. The proposed operation will also add monosodium titanate (MST) to the supernate of the tank prior to the ion exchange operation to remove strontium and select alpha-emitting actinides. The cesium loaded CST is ground and then passed forward to the sludge washing tank as feed to the Defense Waste Processing Facility (DWPF). Similarly, the MST will be transferred to the sludge washing tank. Sludge processing includes the potential to leach aluminum from the solids at elevated temperature (e.g., 65 C) using concentrated (3M) sodium hydroxide solutions. Prior literature indicates that both CST and MST will agglomerate and form higher yield stress slurries with exposure to elevated temperatures. This report assessed that data and provides guidance on minimizing the impact of CST and MST on sludge transfer and aluminum leaching sludge.

  13. Method of preparation of removable syntactic foam

    DOE Patents [OSTI]

    Arnold, C. Jr.; Derzon, D.K.; Nelson, J.S.; Rand, P.B.

    1995-07-11T23:59:59.000Z

    Easily removable, environmentally safe, low-density, syntactic foams are disclosed which are prepared by mixing insoluble microballoons with a solution of water and/or alcohol-soluble polymer to produce a pourable slurry, optionally vacuum filtering the slurry in varying degrees to remove unwanted solvent and solute polymer, and drying to remove residual solvent. The properties of the foams can be controlled by the concentration and physical properties of the polymer, and by the size and properties of the microballoons. The suggested solute polymers are non-toxic and soluble in environmentally safe solvents such as water or low-molecular weight alcohols. The syntactic foams produced by this process are particularly useful in those applications where ease of removability is beneficial, and could find use in packaging recoverable electronic components, in drilling and mining applications, in building trades, in art works, in the entertainment industry for special effects, in manufacturing as temporary fixtures, in agriculture as temporary supports and containers and for delivery of fertilizer, in medicine as casts and splints, as temporary thermal barriers, as temporary protective covers for fragile objects, as filters for particulate matter, which matter may be easily recovered upon exposure to a solvent, as in-situ valves (for one-time use) which go from maximum to minimum impedance when solvent flows through, and for the automatic opening or closing of spring-loaded, mechanical switches upon exposure to a solvent, among other applications. 1 fig.

  14. Method of preparation of removable syntactic foam

    DOE Patents [OSTI]

    Arnold, Jr., Charles (Albuquerque, NM); Derzon, Dora K. (Albuquerque, NM); Nelson, Jill S. (Albuquerque, NM); Rand, Peter B. (Albuquerque, NM)

    1995-01-01T23:59:59.000Z

    Easily removable, environmentally safe, low-density, syntactic foams are disclosed which are prepared by mixing insoluble microballoons with a solution of water and/or alcohol-soluble polymer to produce a pourable slurry, optionally vacuum filtering the slurry in varying degrees to remove unwanted solvent and solute polymer, and drying to remove residual solvent. The properties of the foams can be controlled by the concentration and physical properties of the polymer, and by the size and properties of the microballoons. The suggested solute polymers are non-toxic and soluble in environmentally safe solvents such as water or low-molecular weight alcohols. The syntactic foams produced by this process are particularly useful in those applications where ease of removability is beneficial, and could find use in packaging recoverable electronic components, in drilling and mining applications, in building trades, in art works, in the entertainment industry for special effects, in manufacturing as temporary fixtures, in agriculture as temporary supports and containers and for delivery of fertilizer, in medicine as casts and splints, as temporary thermal barriers, as temporary protective covers for fragile objects, as filters for particulate matter, which matter may be easily recovered upon exposure to a solvent, as in-situ valves (for one-time use) which go from maximum to minimum impedance when solvent flows through, and for the automatic opening or closing of spring-loaded, mechanical switches upon exposure to a solvent, among other applications.

  15. PILOT SCALE TESTING OF MONOSODIUM TITANATE MIXING FOR THE SRS SMALL COLUMN ION EXCHANGE PROCESS - 11224

    SciTech Connect (OSTI)

    Poirier, M.; Restivo, M.; Williams, M.; Herman, D.; Steeper, T.

    2011-01-25T23:59:59.000Z

    The Small Column Ion Exchange (SCIX) process is being developed to remove cesium, strontium, and select actinides from Savannah River Site (SRS) Liquid Waste using an existing waste tank (i.e., Tank 41H) to house the process. Savannah River National Laboratory (SRNL) is conducting pilot-scale mixing tests to determine the pump requirements for suspending monosodium titanate (MST), crystalline silicotitanate (CST), and simulated sludge. The purpose of this pilot scale testing is to determine the requirements for the pumps to suspend the MST particles so that they can contact the strontium and actinides in the liquid and be removed from the tank. The pilot-scale tank is a 1/10.85 linear scaled model of SRS Tank 41H. The tank diameter, tank liquid level, pump nozzle diameter, pump elevation, and cooling coil diameter are all 1/10.85 of their dimensions in Tank 41H. The pump locations correspond to the proposed locations in Tank 41H by the SCIX program (Risers B5 and B2 for two pump configurations and Risers B5, B3, and B1 for three pump configurations). The conclusions from this work follow: (i) Neither two standard slurry pumps nor two quad volute slurry pumps will provide sufficient power to initially suspend MST in an SRS waste tank. (ii) Two Submersible Mixer Pumps (SMPs) will provide sufficient power to initially suspend MST in an SRS waste tank. However, the testing shows the required pump discharge velocity is close to the maximum discharge velocity of the pump (within 12%). (iii) Three SMPs will provide sufficient power to initially suspend MST in an SRS waste tank. The testing shows the required pump discharge velocity is 66% of the maximum discharge velocity of the pump. (iv) Three SMPs are needed to resuspend MST that has settled in a waste tank at nominal 45 C for four weeks. The testing shows the required pump discharge velocity is 77% of the maximum discharge velocity of the pump. Two SMPs are not sufficient to resuspend MST that settled under these conditions.

  16. REMOVAL OF LEGACY PLUTONIUM MATERIALS FROM SWEDEN

    SciTech Connect (OSTI)

    Dunn, Kerry A. [Savannah River National Laboratory; Bellamy, J. Steve [Savannah River National Laboratory; Chandler, Greg T. [Savannah River National Laboratory; Iyer, Natraj C. [U.S. Department of Energy, National Nuclear Security Administration, Office of; Koenig, Rich E.; Leduc, D. [Savannah River National Laboratory; Hackney, B. [Savannah River National Laboratory; Leduc, Dan R. [Savannah River National Laboratory

    2013-08-18T23:59:59.000Z

    U.S. Department of Energy’s National Nuclear Security Administration (NNSA) Office of Global Threat Reduction (GTRI) recently removed legacy plutonium materials from Sweden in collaboration with AB SVAFO, Sweden. This paper details the activities undertaken through the U.S. receiving site (Savannah River Site (SRS)) to support the characterization, stabilization, packaging and removal of legacy plutonium materials from Sweden in 2012. This effort was undertaken as part of GTRI’s Gap Materials Program and culminated with the successful removal of plutonium from Sweden as announced at the 2012 Nuclear Security Summit. The removal and shipment of plutonium materials to the United States was the first of its kind under NNSA’s Global Threat Reduction Initiative. The Environmental Assessment for the U.S. receipt of gap plutonium material was approved in May 2010. Since then, the multi-year process yielded many first time accomplishments associated with plutonium packaging and transport activities including the application of the of DOE-STD-3013 stabilization requirements to treat plutonium materials outside the U.S., the development of an acceptance criteria for receipt of plutonium from a foreign country, the development and application of a versatile process flow sheet for the packaging of legacy plutonium materials, the identification of a plutonium container configuration, the first international certificate validation of the 9975 shipping package and the first intercontinental shipment using the 9975 shipping package. This paper will detail the technical considerations in developing the packaging process flow sheet, defining the key elements of the flow sheet and its implementation, determining the criteria used in the selection of the transport package, developing the technical basis for the package certificate amendment and the reviews with multiple licensing authorities and most importantly integrating the technical activities with the Swedish partners.

  17. Synthesis and Optimization of the Sintering Kinetics of Actinide Nitrides

    SciTech Connect (OSTI)

    Drryl P. Butt; Brian Jaques

    2009-03-31T23:59:59.000Z

    Research conducted for this NERI project has advanced the understanding and feasibility of nitride nuclear fuel processing. In order to perform this research, necessary laboratory infrastructure was developed; including basic facilities and experimental equipment. Notable accomplishments from this project include: the synthesis of uranium, dysprosium, and cerium nitrides using a novel, low-cost mechanical method at room temperature; the synthesis of phase pure UN, DyN, and CeN using thermal methods; and the sintering of UN and (Ux, Dy1-x)N (0.7 ? X ? 1) pellets from phase pure powder that was synthesized in the Advanced Materials Laboratory at Boise State University.

  18. Micro-Analysis of Actinide Minerals for Nuclear Forensics and Treaty Verification

    SciTech Connect (OSTI)

    M. Morey, M. Manard, R. Russo, G. Havrilla

    2012-03-22T23:59:59.000Z

    Micro-Raman spectroscopy has been demonstrated to be a viable tool for nondestructive determination of the crystal phase of relevant minerals. Collecting spectra on particles down to 5 microns in size was completed. Some minerals studied were weak scatterers and were better studied with the other techniques. A decent graphical software package should easily be able to compare collected spectra to a spectral library as well as subtract out matrix vibration peaks. Due to the success and unequivocal determination of the most common mineral false positive (zircon), it is clear that Raman has a future for complementary, rapid determination of unknown particulate samples containing actinides.

  19. Actinide extraction from simulated and irradiated spent nuclear fuel using TBP solutions in HFC-134a

    SciTech Connect (OSTI)

    Shadrin, A.; Babain, V.; Kamachev, V.; Murzin, A.; Shafikov, D.; Dormidonova, A. [Khlopin Radium Institute, RPA, 28, 2-nd Murinskii ay., St-Petersburg (Russian Federation)

    2008-07-01T23:59:59.000Z

    It was demonstrated that solutions of TBP-nitric acid adduct in liquid Freon HFC-134a (1.2 MPa, 25 deg. C) allowed for recovery of uranium with nearly the same effectiveness as supercritical CO{sub 2} at 30 MPa. At nearly quantitative recovery of U and Pu, a DF of ca. 10 can be attained on dissolution and extraction of simulated SNF samples. The possibility of recovery of actinides contained in cakes produced by oxide conversion of simulated and irradiated SNF with solutions of TBP and DBE in Freon HFC-134a was shown. (authors)

  20. Low-Temperature Synthesis of Actinide Tetraborides by Solid-State Metathesis Reactions

    SciTech Connect (OSTI)

    Lupinetti, Anthony J.; Garcia, Eduardo; Abney, Kent D.

    2004-12-14T23:59:59.000Z

    The synthesis of actinide tetraborides including uranium tetraboride (UB,), plutonium tetraboride (PUB,) and thorium tetraboride (ThB{sub 4}) by a solid-state metathesis reaction are demonstrated. The present method significantly lowers the temperature required to {approx_equal}850 C. As an example, when UCl{sub 4}, is reacted with an excess of MgB{sub 2}, at 850 C, crystalline UB, is formed. Powder X-ray diffraction and ICP-AES data support the reduction of UCl{sub 3}, as the initial step in the reaction. The UB, product is purified by washing water and drying.

  1. Conceptual configurations of an accelerator-driven subcritical system utilizing minor actinides

    SciTech Connect (OSTI)

    Cao, Y.; Gohar, Y. [Nuclear Engineering Div., Argonne National Laboratory, 9700 South Cass Ave., IL 60439 (United States)

    2012-07-01T23:59:59.000Z

    This paper purposes an Accelerator-Driven Subcritical (ADS) system which utilizes the Minor Actinides (MAs) from the US spent nuclear fuel inventory. A mobile fuel concept with micro-particles suspended in the liquid metal is adopted in the purposed system to avoid difficulties of developing and testing new MAs solid fuel forms. Three ADS configurations were developed and analyzed using the Monte Carlo fuel burnup methodology. The analyses demonstrated the capabilities of the proposed system to utilize the MAs and to dispose of the US spent nuclear fuels. (authors)

  2. Actinide partitioning from actual ICPP dissolved zirconium calcine using the TRUEX solvent

    SciTech Connect (OSTI)

    Brewer, K.N.; Herbst, R.S.; Tranter, T.J. [and others

    1995-05-01T23:59:59.000Z

    The TRansUranic EXtraction process (TRUEX), as developed by E.P. Horwitz and coworkers at Argonne National Laboratory (ANL), is being evaluated as a TRU extraction process for Idaho Chemical Processing Plant (ICPP) wastes. A criteria that must be met during this evaluation, is that the aqueous raffinate must be below the 10 nCi/g limit specified in 10 CFR 61.55. A test was performed where the TRUEX solvent (0.2 M octyl(phenyl)-N-N-diisobutyl-carbamoylmethyl-phosphine oxide (CMPO), and 1.4 M tributylphosphate (TBP) in an Isopar-L diluent) was contacted with actual ICPP dissolved zirconium calcine. Two experimental flowsheets were used to determine TRU decontamination factors, and TRU, Zr, Fe, Cr, and Tc extraction, scrub, and strip distribution coefficients. Results from these two flowsheets show that >99.99% of the TRU alpha activity was removed from the acidic feed after three contacts with the TRUEX solvent (fresh solvent being used for each contact). The resulting aqueous raffinate solution contained an approximate TRU alpha activity of 0.02 nCi/g, which is well below the non-TRU waste limit of 10 nCi/g specified in 10 CFR 61.55. Favorable scrub and strip distribution coefficients were also observed for Am-241, Pu-238, and Pu-239, indicating the feasibility of recovering these isotopes from the TRUTEX solvent. A solution of 0.04 M 1-hydroxyethane-1,1-diphosphonic acid (HEDPA) in 0.04 M HNO{sub 3} was used to successfully strip the TRUs from the TRUEX solvent. The results of the test using actual ICPP dissolved zirconium calcine, and subsequent GTM evaluation, show the feasibility of removing TRUs from the dissolved zirconium calcine with the TRUEX solvent and the deleterious effects zirconium poses with the ICPP zirconium calcine waste. Test results using actual ICPP zirconium calcine reveal the necessity of preventing zirconium from following the TRUs.

  3. Waste processing air cleaning

    SciTech Connect (OSTI)

    Kriskovich, J.R.

    1998-07-27T23:59:59.000Z

    Waste processing and preparing waste to support waste processing relies heavily on ventilation. Ventilation is used at the Hanford Site on the waste storage tanks to provide confinement, cooling, and removal of flammable gases.

  4. Decontamination of process equipment using recyclable chelating solvent

    SciTech Connect (OSTI)

    Jevec, J.; Lenore, C.; Ulbricht, S.

    1995-12-01T23:59:59.000Z

    The Department of Energy (DOE) is now faced with the task of meeting decontamination and decommissioning obligations at numerous facilities by the year 2019. Due to the tremendous volume of material involved, innovative decontamination technologies are being sought that can reduce the volumes of contaminated waste materials and secondary wastes requiring disposal. With sufficient decontamination, some of the material from DOE facilities could be released as scrap into the commercial sector for recycle, thereby reducing the volume of radioactive waste requiring disposal. Although recycling may initially prove to be more costly than current disposal practices, rapidly increasing disposal costs are expected to make recycling more and more cost effective. Additionally, recycling is now perceived as the ethical choice in a world where the consequences of replacing resources and throwing away reusable materials are impacting the well-being of the environment. Current approaches to the decontamination of metals most often involve one of four basic process types: (1) chemical, (2) manual and mechanical, (3) electrochemical, and (4) ultrasonic. {open_quotes}Hard{close_quotes} chemical decontamination solutions, capable of achieving decontamination factors (Df`s) of 50 to 100, generally involve reagent concentrations in excess of 5%, tend to physically degrade the surface treated, and generate relatively large volumes of secondary waste. {open_quotes}Soft{close_quotes} chemical decontamination solutions, capable of achieving Df`s of 5 to 10, normally consist of reagents at concentrations of 0.1 to 1%, generally leave treated surfaces in a usable condition, and generate relatively low secondary waste volumes. Under contract to the Department of Energy, the Babcock & Wilcox Company is developing a chemical decontamination process using chelating agents to remove uranium compounds and other actinide species from process equipment.

  5. DWPF FLOWSHEET STUDIES WITH SIMULANT TO DETERMINE THE IMPACT OF NEXT GENERATION SOLVENT ON THE CPC PROCESS AND GLASS FORMULATION

    SciTech Connect (OSTI)

    Newell, J.; Peeler, D.; Edwards, T.; Hay, M.; Stone, M.

    2011-06-29T23:59:59.000Z

    As a part of the Actinide Removal Process (ARP)/Modular Caustic Side Solvent Extraction Unit (MCU) Life Extension Project, a next generation solvent (NGS), a new strip acid, and modified monosodium titanate (mMST) will be deployed. The NGS is comprised of four components: 0.050 M MaxCalix (extractant), 0.50 M Cs-7SB (modifier), 0.003 M guanidine-LIX-79, with the balance ({approx}74 wt%) being Isopar{reg_sign} L. The strip acid will be changed from dilute nitric acid to dilute boric acid (0.01 M). Because of these changes, experimental testing with the next generation solvent and mMST was required to determine the impact of these changes in 512-S and Defense Waste Processing Facility (DWPF) operations, as well as Chemical Process Cell (CPC), glass formulation activities, and melter operations. Because of these changes, experimental testing with the next generation solvent and mMST is required to determine the impact of these changes. A Technical Task Request (TTR) was issued to support the assessments of the impact of the next generation solvent and mMST on the downstream DWPF flowsheet unit. The TTR identified five tasks to be investigated: (1) CPC Flowsheet Demonstration for NGS; (2) Solvent Stability for DWPF CPC Conditions; (3) Glass Formulation Studies; (4) Boron Volatility and Melt Rate; and (5) CPC Flowsheet Demonstration for mMST.

  6. Removal Rate Model for Magnetorheological Finishing of Glass

    SciTech Connect (OSTI)

    DeGroote, J.E.; Marino, A.E.; WIlson, J.P.; Bishop, A.L.; Lambropoulos, J.C.; Jacobs, S.D.

    2007-11-14T23:59:59.000Z

    Magnetorheological finishing (MRF) is a deterministic subaperture polishing process. The process uses a magntorheological (MR) fluid that consists of micrometer-sized, spherical, magnetic carbonyl iron (CI) particles, nonmagnetic polishing abrasives, water, and stabilizers. Material removal occurs when the CI and nonmagnetic polishing abrasives shear material off the surface being polished. We introduce a new MRF material removal rate model for glass. This model contains terms for the near surface mechanical properties of glass, drag force, polishing abrasive size and concentration, chemical durability of the glass, MR fluid pH, and the glass composition. We introduce quantitative chemical predictors for the first time, to the best of our knowledge, into an MRF removal rate model. We validate individual terms in our model separately and then combine all of the terms to show the whole MRF material removal model compared with experimental data. All of our experimental data were obtained using nanodiamond MR fluids and a set of six optical glasses.

  7. Beta-delayed fission and neutron emission calculations for the actinide cosmochronometers

    SciTech Connect (OSTI)

    Meyer, B.S.; Howard, W.M.; Mathews, G.J.; Takahashi, K.; Moeller, P.; Leander, G.A.

    1989-05-01T23:59:59.000Z

    The Gamow-Teller beta-strength distributions for 19 neutron-rich nuclei, including ten of interest for the production of the actinide cosmochronometers, are computed microscopically with a code that treats nuclear deformation explicitly. The strength distributions are then used to calculate the beta-delayed fission, neutron emission, and gamma deexcitation probabilities for these nuclei. Fission is treated both in the complete damping and WKB approximations for penetrabilities through the nuclear potential-energy surface. The resulting fission probabilities differ by factors of 2 to 3 or more from the results of previous calculations using microscopically computed beta-strength distributions around the region of greatest interest for production of the cosmochronometers. The indications are that a consistent treatment of nuclear deformation, fission barriers, and beta-strength functions is important in the calculation of delayed fission probabilities and the production of the actinide cosmochronometers. Since we show that the results are very sensitive to relatively small changes in model assumptions, large chronometric ages for the Galaxy based upon high beta-delayed fission probabilities derived from an inconsistent set of nuclear data calculations must be considered quite uncertain.

  8. Separation and Analysis of Actinides by Extraction Chromotography Coupled with Alpha Liquid Scintillation Spectrometry

    SciTech Connect (OSTI)

    Cadieux, J.R.; Reboul, S.H.

    1995-09-21T23:59:59.000Z

    This work describes the development and testing of a new method for the separation and analysis of most actinides of interest in environmental samples. It combines simplified extraction chromatography using highly selective absorption resins to partition the individual actinides with the measurement of their alpha activities by liquid scintillation spectrometry. The liquid scintillation counting technique pioneered by McDowell proved useful in determination of alpha emitting radionuclide in a wide variety of matrices. Alpha emitters are chemically extracted into an organic phase which also contains the scintillation cocktail. Oxygen is purged from the solution to improve the energy resolution of the measurement and the counting sample is sealed in a small glass tube for assay. The Photon-Electron Rejecting Alpha Liquid Scintillation (PERALS{trademark}) Spectrometer provides high counting efficiency, low background, pulse shape discrimination for photon/electron/{beta} particle rejection and moderate energy resolution in a compact package. Chemical extraction/liquid scintillation counting significantly reduces the extensive chemical purification and electroplating required for conventional alpha spectrometry with semiconductor detectors. PERALS{trademark} analyses have been used routinely for quickly surveying suspect samples and determining the source of unknown alpha activities.

  9. Conceptual design of minor actinides burner with an accelerator-driven subcritical system.

    SciTech Connect (OSTI)

    Cao, Y.; Gohar, Y. (Nuclear Engineering Division)

    2011-11-04T23:59:59.000Z

    In the environmental impact study of the Yucca Mountain nuclear waste repository, the limit of spent nuclear fuel (SNF) for disposal is assessed at 70,000 metric tons of heavy metal (MTHM), among which 63,000 MTHM are the projected SNF discharge from U.S. commercial nuclear power plants though 2011. Within the 70,000 MTHM of SNF in storage, approximately 115 tons would be minor actinides (MAs) and 585 tons would be plutonium. This study describes the conceptual design of an accelerator-driven subcritical (ADS) system intended to utilize (burn) the 115 tons of MAs. The ADS system consists of a subcritical fission blanket where the MAs fuel will be burned, a spallation neutron source to drive the fission blanket, and a radiation shield to reduce the radiation dose to an acceptable level. The spallation neutrons are generated from the interaction of a 1 GeV proton beam with a lead-bismuth eutectic (LBE) or liquid lead target. In this concept, the fission blanket consists of a liquid mobile fuel and the fuel carrier can be LBE, liquid lead, or molten salt. The actinide fuel materials are dissolved, mixed, or suspended in the liquid fuel carrier. Therefore, fresh fuel can be fed into the fission blanket to adjust its reactivity and to control system power during operation. Monte Carlo analyses were performed to determine the overall parameters of an ADS system utilizing LBE as an example. Steady-state Monte Carlo simulations were studied for three fission blanket configurations that are similar except that the loaded amount of actinide fuel in the LBE is either 5, 7, or 10% of the total volume of the blanket, respectively. The neutron multiplication factor values of the three configurations are all approximately 0.98 and the MA initial inventories are each approximately 10 tons. Monte Carlo burnup simulations using the MCB5 code were performed to analyze the performance of the three conceptual ADS systems. Preliminary burnup analysis shows that all three conceptual ADS systems consume about 1.2 tons of actinides per year and produce 3 GW thermal power, with a proton beam power of 25 MW. Total MA fuel that would be consumed in the first 10 years of operation is 9.85, 11.80, or 12.68 tons, respectively, for the systems with 5, 7, or 10% actinide fuel particles loaded in the LBE. The corresponding annual MA fuel transmutation rate after reaching equilibrium at 10 years of operation is 0.83, 0.94, or 1.02 tons/year, respectively. Assuming that the ADS systems can be operated for 35 full-power years, the total MAs consumed in the three ADS systems are 30.6, 35.3, and 37.2 tons, respectively. For the three configurations, it is estimated that 3.8, 3.3, or 3.1 ADS system units are required to utilize the entire 115 tons of MA fuel in the SNF inventory, respectively.

  10. Actinide Corroles: Synthesis and Characterization of Thorium(IV) and Uranium(IV) bis(-chloride) Dimers

    SciTech Connect (OSTI)

    Ward, Ashleigh L.; Buckley, Heather L.; Gryko, Daniel T.; Lukens, Wayne W.; Arnold, John

    2013-12-01T23:59:59.000Z

    The first synthesis and structural characterization of actinide corroles is presented. Thorium(IV) and uranium(IV) macrocycles of Mes2(p-OMePh)corrole were synthesised and characterized by single-crystal X-ray diffraction, UV-Visible spectroscopy, variable-temperature 1H NMR, ESI mass spectrometry and cyclic voltammetry.

  11. Removal of arsenic compounds from petroliferous liquids

    DOE Patents [OSTI]

    Fish, R.H.

    1984-04-06T23:59:59.000Z

    The present invention in one aspect comprises a process for removing arsenic from petroliferous-derived liquids by contacting said liquid with a divinylbenzene-crosslinked polystyrene polymer (i.e. PS-DVB) having catechol ligands anchored to said polymer, said contacting being at an elevated temperature. In another aspect, the invention is a process for regenerating spent catecholated polystyrene polymer by removal of the arsenic bound to it from contacting petroliferous liquid in accordance with the aspect described above which regenerating process comprises: (a) treating said spent catecholated polystyrene polymer with an aqueous solution of at least one member selected from the group consisting of carbonates and bicarbonates of ammonium, alkali metals, and alkaline earth metals, said solution having a pH between about 8 and 10, and said treating being at a temperature in the range of about 20/sup 0/ to 100/sup 0/C; (b) separating the solids and liquids from each other. In a preferred embodiment the regeneration treatment is in two steps wherein step: (a) is carried out with an aqueous alcoholic carbonate solution which includes at least one lower alkyl alcohol, and, steps (c) and (d) are added. Steps (c) and (d) comprise: (c) treating the solids with an aqueous alcoholic solution of at least one ammonium, alkali or alkaline earth metal bicarbonate at a temperature in the range of about 20 to 100/sup 0/C; and (d) separating the solids from the liquids.

  12. New density functional theory approaches for enabling prediction of chemical and physical properties of plutonium and other actinides.

    SciTech Connect (OSTI)

    Mattsson, Ann Elisabet

    2012-01-01T23:59:59.000Z

    Density Functional Theory (DFT) based Equation of State (EOS) construction is a prominent part of Sandia's capabilities to support engineering sciences. This capability is based on amending experimental data with information gained from computational investigations, in parts of the phase space where experimental data is hard, dangerous, or expensive to obtain. A prominent materials area where such computational investigations are hard to perform today because of limited accuracy is actinide and lanthanide materials. The Science of Extreme Environment Lab Directed Research and Development project described in this Report has had the aim to cure this accuracy problem. We have focused on the two major factors which would allow for accurate computational investigations of actinide and lanthanide materials: (1) The fully relativistic treatment needed for materials containing heavy atoms, and (2) the needed improved performance of DFT exchange-correlation functionals. We have implemented a fully relativistic treatment based on the Dirac Equation into the LANL code RSPt and we have shown that such a treatment is imperative when calculating properties of materials containing actinides and/or lanthanides. The present standard treatment that only includes some of the relativistic terms is not accurate enough and can even give misleading results. Compared to calculations previously considered state of the art, the Dirac treatment gives a substantial change in equilibrium volume predictions for materials with large spin-orbit coupling. For actinide and lanthanide materials, a Dirac treatment is thus a fundamental requirement in any computational investigation, including those for DFT-based EOS construction. For a full capability, a DFT functional capable of describing strongly correlated systems such as actinide materials need to be developed. Using the previously successful subsystem functional scheme developed by Mattsson et.al., we have created such a functional. In this functional the Harmonic Oscillator Gas is providing the necessary reference system for the strong correlation and localization occurring in actinides. Preliminary testing shows that the new Hao-Armiento-Mattsson (HAM) functional gives a trend towards improved results for the crystalline copper oxide test system we have chosen. This test system exhibits the same exchange-correlation physics as the actinide systems do, but without the relativistic effects, giving access to a pure testing ground for functionals. During the work important insights have been gained. An example is that currently available functionals, contrary to common belief, make large errors in so called hybridization regions where electrons from different ions interact and form new states. Together with the new understanding of functional issues, the Dirac implementation into the RSPt code will permit us to gain more fundamental understanding, both quantitatively and qualitatively, of materials of importance for Sandia and the rest of the Nuclear Weapons complex.

  13. RAPID DETERMINATION OF ACTINIDES IN URINE BY INDUCTIVELY-COUPLED PLASMA MASS SPECTROMETRY AND ALPHA SPECTROMETRY: A HYBRID APPROACH

    SciTech Connect (OSTI)

    Maxwell, S.; Jones, V.

    2009-05-27T23:59:59.000Z

    A new rapid separation method that allows separation and preconcentration of actinides in urine samples was developed for the measurement of longer lived actinides by inductively coupled plasma mass spectrometry (ICP-MS) and short-lived actinides by alpha spectrometry; a hybrid approach. This method uses stacked extraction chromatography cartridges and vacuum box technology to facilitate rapid separations. Preconcentration, if required, is performed using a streamlined calcium phosphate precipitation. Similar technology has been applied to separate actinides prior to measurement by alpha spectrometry, but this new method has been developed with elution reagents now compatible with ICP-MS as well. Purified solutions are split between ICP-MS and alpha spectrometry so that long- and short-lived actinide isotopes can be measured successfully. The method allows for simultaneous extraction of 24 samples (including QC samples) in less than 3 h. Simultaneous sample preparation can offer significant time savings over sequential sample preparation. For example, sequential sample preparation of 24 samples taking just 15 min each requires 6 h to complete. The simplicity and speed of this new method makes it attractive for radiological emergency response. If preconcentration is applied, the method is applicable to larger sample aliquots for occupational exposures as well. The chemical recoveries are typically greater than 90%, in contrast to other reported methods using flow injection separation techniques for urine samples where plutonium yields were 70-80%. This method allows measurement of both long-lived and short-lived actinide isotopes. 239Pu, 242Pu, 237Np, 243Am, 234U, 235U and 238U were measured by ICP-MS, while 236Pu, 238Pu, 239Pu, 241Am, 243Am and 244Cm were measured by alpha spectrometry. The method can also be adapted so that the separation of uranium isotopes for assay is not required, if uranium assay by direct dilution of the urine sample is preferred instead. Multiple vacuum box locations may be set-up to supply several ICP-MS units with purified sample fractions such that a high sample throughput may be achieved, while still allowing for rapid measurement of short-lived actinides by alpha spectrometry.

  14. Heavy Water Test Reactor Dome Removal

    SciTech Connect (OSTI)

    None

    2011-01-01T23:59:59.000Z

    A high speed look at the removal of the Heavy Water Test Reactor Dome Removal. A project sponsored by the Recovery Act on the Savannah River Site.

  15. Mercury and tritium removal from DOE waste oils

    SciTech Connect (OSTI)

    Klasson, E.T. [Oak Ridge National Lab., TN (United States)

    1997-10-01T23:59:59.000Z

    This work covers the investigation of vacuum extraction as a means to remove tritiated contamination as well as the removal via sorption of dissolved mercury from contaminated oils. The radiation damage in oils from tritium causes production of hydrogen, methane, and low-molecular-weight hydrocarbons. When tritium gas is present in the oil, the tritium atom is incorporated into the formed hydrocarbons. The transformer industry measures gas content/composition of transformer oils as a diagnostic tool for the transformers` condition. The analytical approach (ASTM D3612-90) used for these measurements is vacuum extraction of all gases (H{sub 2}, N{sub 2}, O{sub 2}, CO, CO{sub 2}, etc.) followed by analysis of the evolved gas mixture. This extraction method will be adapted to remove dissolved gases (including tritium) from the SRS vacuum pump oil. It may be necessary to heat (60{degrees}C to 70{degrees}C) the oil during vacuum extraction to remove tritiated water. A method described in the procedures is a stripper column extraction, in which a carrier gas (argon) is used to remove dissolved gases from oil that is dispersed on high surface area beads. This method appears promising for scale-up as a treatment process, and a modified process is also being used as a dewatering technique by SD Myers, Inc. (a transformer consulting company) for transformers in the field by a mobile unit. Although some mercury may be removed during the vacuum extraction, the most common technique for removing mercury from oil is by using sulfur-impregnated activated carbon (SIAC). SIAC is currently being used by the petroleum industry to remove mercury from hydrocarbon mixtures, but the sorbent has not been previously tested on DOE vacuum oil waste. It is anticipated that a final process will be similar to technologies used by the petroleum industry and is comparable to ion exchange operations in large column-type reactors.

  16. Actinide Chemistry

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc Documentation RUCProductstwrmrAre the EffectsAcknowledgment StatementGuidance »| Y-12

  17. Recovery of UO{sub 2}/PuO{sub 2} in IFR electrorefining process

    DOE Patents [OSTI]

    Tomczuk, Z.; Miller, W.E.

    1992-01-01T23:59:59.000Z

    This invention is comprised of a process for converting PuO{sub 2} and U0{sub 2} present in an electrorefiner to the chlorides, by contacting the PuO{sub 2} and U0{sub 2} with Li metal in the presence of an alkali metal chloride salt substantially free of rare earth and actinide chlorides for a time and at a temperature sufficient to convert the U0{sub 2} and PuO{sub 2} to metals while converting Li metal to Li{sub 2}O. Li{sub 2}O is removed either by reducing with rare earth metals or by providing an oxygen electrode for transporting 0{sub 2} out of the electrorefiner and a cathode, and thereafter applying an emf to the electrorefiner electrodes sufficient to cause the Li{sub 2}O to disassociate to 0{sub 2} and Li metal but insufficient to decompose the alkali metal chloride salt. The U and Pu and excess lithium are then converted to chlorides by reaction with CdCl{sub 2}.

  18. Recovery of UO[sub 2]/PuO[sub 2] in IFR electrorefining process

    DOE Patents [OSTI]

    Tomczuk, Z.; Miller, W.E.

    1994-10-18T23:59:59.000Z

    A process is described for converting PuO[sub 2] and UO[sub 2] present in an electrorefiner to the chlorides, by contacting the PuO[sub 2] and UO[sub 2] with Li metal in the presence of an alkali metal chloride salt substantially free of rare earth and actinide chlorides for a time and at a temperature sufficient to convert the UO[sub 2] and PuO[sub 2] to metals while converting Li metal to Li[sub 2]O. Li[sub 2]O is removed either by reducing with rare earth metals or by providing an oxygen electrode for transporting O[sub 2] out of the electrorefiner and a cathode, and thereafter applying an emf to the electrorefiner electrodes sufficient to cause the Li[sub 2]O to disassociate to O[sub 2] and Li metal but insufficient to decompose the alkali metal chloride salt. The U and Pu and excess lithium are then converted to chlorides by reaction with CdCl[sub 2].

  19. USE OF AN EQUILIBRIUM MODEL TO FORECAST DISSOLUTION EFFECTIVENESS, SAFETY IMPACTS, AND DOWNSTREAM PROCESSABILITY FROM OXALIC ACID AIDED SLUDGE REMOVAL IN SAVANNAH RIVER SITE HIGH LEVEL WASTE TANKS 1-15

    SciTech Connect (OSTI)

    KETUSKY, EDWARD

    2005-10-31T23:59:59.000Z

    This thesis details a graduate research effort written to fulfill the Magister of Technologiae in Chemical Engineering requirements at the University of South Africa. The research evaluates the ability of equilibrium based software to forecast dissolution, evaluate safety impacts, and determine downstream processability changes associated with using oxalic acid solutions to dissolve sludge heels in Savannah River Site High Level Waste (HLW) Tanks 1-15. First, a dissolution model is constructed and validated. Coupled with a model, a material balance determines the fate of hypothetical worst-case sludge in the treatment and neutralization tanks during each chemical adjustment. Although sludge is dissolved, after neutralization more is created within HLW. An energy balance determines overpressurization and overheating to be unlikely. Corrosion induced hydrogen may overwhelm the purge ventilation. Limiting the heel volume treated/acid added and processing the solids through vitrification is preferred and should not significantly increase the number of glass canisters.

  20. RECYCLING AND REMOVAL OF OFFSHORE WIND TURBINES AN INTERACTIVE METHOD FOR REDUCTION OF NEGATIVE ENVIRONMENTAL EFFECTS

    E-Print Network [OSTI]

    RECYCLING AND REMOVAL OF OFFSHORE WIND TURBINES ­ AN INTERACTIVE METHOD FOR REDUCTION OF NEGATIVE.borup@risoe.dk ABSTRACT: This paper describes a method for reduction of negative environmental impacts of wind turbines and an analysis of future removal and recycling processes of offshore wind turbines. The method is process

  1. High Pressure Phase Transformations in Heavy Rare Earth Metals and Connections to Actinide Crystal Structures

    SciTech Connect (OSTI)

    Vohra, Yogesh K.; Sangala, Bagvanth Reddy; Stemshorn, Andrew K. [Physics, University of Alabama at Birmingham (UAB), 310 Campbell Hall, 1300 University Boulevard, Birmingham, AL, 35294-1170 (United States); Hope, Kevin M. [Biology, Chemistry, and Mathematics, University of Montevallo, Harman Hall, Station 6480, Montevallo, AL, 35115 (United States)

    2008-07-01T23:59:59.000Z

    High-pressure studies have been performed on heavy rare earth metals Terbium (Tb) to 155 GPa and Holmium (Ho) to 134 GPa in a diamond anvil cell at room temperature. The following crystal structure sequence was observed in both metals hcp {yields} Sm-type {yields} dhcp {yields} distorted fcc (hR-24) {yields} monoclinic (C2/m) with increasing pressure. The last transformation to a low symmetry monoclinic phase is accompanied by a volume collapse of 5 % for Tb at 51 GPa and a volume collapse of 3 % for Ho at 103 GPa. This volume collapse under high pressure is reminiscent of f-shell delocalization in light rare earth metal Cerium (Ce), Praseodymium (Pr), and heavy actinide metals Americium (Am) and Curium (Cm). The orthorhombic Pnma phase that has been reported in Am and Cm after f-shell delocalization is not observed in heavy rare earth metals under high pressures. (authors)

  2. Pre-neutron emission mass distributions for low-energy neutron-induced actinide fission

    E-Print Network [OSTI]

    Xiaojun Sun; Chenggang Yu; Ning Wang

    2012-01-15T23:59:59.000Z

    According to the driving potential of a fissile system, we propose a phenomenological fission potential for a description of the pre-neutron emission mass distributions of neutron-induced actinide fission. Based on the nucleus-nucleus potential with the Skyrme energy-density functional, the driving potential of the fissile system is studied considering the deformations of nuclei. The energy dependence of the potential parameters is investigated based on the experimental data for the heights of the peak and valley of the mass distributions. The pre-neutron emission mass distributions for reactions 238U(n, f), 237Np(n, f), 235U(n, f), 232Th(n, f) and 239Pu(n, f) can be reasonably well reproduced. Some predictions for these reactions at unmeasured incident energies are also presented.

  3. Tunneling through equivalent multihumped fission barriers: Some implications for the actinide nuclei

    SciTech Connect (OSTI)

    Bhandari, B.S.; Al-Kharam, A.S.

    1989-03-01T23:59:59.000Z

    A comparison of the penetrabilities calculated in the Wentzel-Kramers-Brillouin approximation through equivalent multihumped fission barriers shows that the penetrability saturates to its maximum value much more slowly for a three-humped potential than that for comparable two-humped and single-humped potentials. An analysis of the slopes of the near-barrier photofission cross sections of actinides yields results that can be understood in terms of the predicted potential barrier shapes for these nuclei, and thus provides evidence in support of resolving the ''thorium anomaly'' along the lines suggested by Moeller and Nix. Our results further indicate that the uranium nuclei, and in particular /sup 236/U, may more likely exhibit three-humped potential shapes in which the apparent consequences of both the second and third minima may be observable.

  4. Method for digesting spent ion exchange resins and recovering actinides therefrom using microwave radiation

    DOE Patents [OSTI]

    Maxwell, III, Sherrod L. (Aiken, SC); Nichols, Sheldon T. (Augusta, GA)

    1999-01-01T23:59:59.000Z

    The present invention relates to methods for digesting diphosphonic acid substituted cation exchange resins that have become loaded with actinides, rare earth metals, or heavy metals, in a way that allows for downstream chromatographic analysis of the adsorbed species without damage to or inadequate elution from the downstream chromatographic resins. The methods of the present invention involve contacting the loaded diphosphonic acid resin with concentrated oxidizing acid in a closed vessel, and irradiating this mixture with microwave radiation. This efficiently increases the temperature of the mixture to a level suitable for digestion of the resin without the use of dehydrating acids that can damage downstream analytical resins. In order to ensure more complete digestion, the irradiated mixture can be mixed with hydrogen peroxide or other oxidant, and reirradiated with microwave radiation.

  5. Low-temperature synthesis of actinide tetraborides by solid-state metathesis reactions

    DOE Patents [OSTI]

    Lupinetti, Anthony J. (Los Alamos, NM); Garcia, Eduardo (Los Alamos, NM); Abney, Kent D. (Los Alamos, NM)

    2004-12-14T23:59:59.000Z

    The synthesis of actinide tetraborides including uranium tetraboride (UB.sub.4), plutonium tetraboride (PuB.sub.4) and thorium tetraboride (ThB.sub.4) by a solid-state metathesis reaction are demonstrated. The present method significantly lowers the temperature required to .ltoreq.850.degree. C. As an example, when UCl.sub.4 is reacted with an excess of MgB.sub.2, at 850.degree. C., crystalline UB.sub.4 is formed. Powder X-ray diffraction and ICP-AES data support the reduction of UCl.sub.3 as the initial step in the reaction. The UB.sub.4 product is purified by washing water and drying.

  6. Sub-barrier capture with quantum diffusion approach: actinide-based reactions

    E-Print Network [OSTI]

    V. V. Sargsyan; G. G. Adamian; N. V. Antonenko; W. Scheid; H. Q. Zhang

    2011-06-14T23:59:59.000Z

    With the quantum diffusion approach the behavior of capture cross sections and mean-square angular momenta of captured systems are revealed in the reactions with deformed nuclei at subbarrier energies. The calculated results are in a good agreement with existing experimental data. With decreasing bombarding energy under the barrier the external turning point of the nucleusnucleus potential leaves the region of short-range nuclear interaction and action of friction. Because of this change of the regime of interaction, an unexpected enhancement of the capture cross section is expected at bombarding energies far below the Coulomb barrier. This effect is shown its worth in the dependence of mean-square angular momentum of captured system on the bombarding energy. From the comparison of calculated and experimental capture cross sections, the importance of quasifission near the entrance channel is shown for the actinide-based reactions leading to superheavy nuclei.

  7. Yields of neutron-rich nuclei by actinide photofission in giant dipole resonance region

    E-Print Network [OSTI]

    Debasis Bhowmick; Debasis Atta; D. N. Basu; Alok Chakrabarti

    2015-01-19T23:59:59.000Z

    Photofission of actinides is studied in the region of nuclear excitation energies that covers the entire giant dipole resonance (GDR) region. A comparative analysis of the behavior of the symmetric and asymmetric modes of photon induced fission as a function of the average excitation energy of the fissioning nucleus is performed. The mass distributions of $^{238}$U photofission fragments are obtained at the endpoint bremsstrahlung energy of 29.1 MeV which corresponds to mean photon energy of 13.7$\\pm$0.3 MeV that coincides with GDR peak for $^{238}$U photofission. The integrated yield of $^{238}$U photofission as well as charge distribution of photofission products are calculated and its role in the production of neutron-rich nuclei and their exoticity is explored.

  8. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

    SciTech Connect (OSTI)

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-03-10T23:59:59.000Z

    The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

  9. Study on Equilibrium Characteristics of Thorium-Plutonium-Minor Actinides Mixed Oxides Fuel in PWR

    SciTech Connect (OSTI)

    Waris, A.; Permana, S.; Kurniadi, R.; Su'ud, Z. [Bosscha Laboratory, Department of Physics, Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Sekimoto, H. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology (Japan)

    2010-06-22T23:59:59.000Z

    A study on characteristics of thorium-plutonium-minor actinides utilization in the pressurized water reactor (PWR) with the equilibrium burnup model has been conducted. For a comprehensive evaluation, several fuel cycles scenario have been included in the present study with the variation of moderator-to-fuel volume ratio (MFR) of PWR core design. The results obviously exhibit that the neutron spectra grow to be harder with decreasing of the MFR. Moreover, the neutron spectra also turn into harder with the rising number of confined heavy nuclides. The required {sup 233}U concentration for criticality of reactor augments with the increasing of MFR for all heavy nuclides confinement and thorium and uranium confinement in PWR.

  10. Dynamical approach to heavy-ion induced fusion using actinide target

    SciTech Connect (OSTI)

    Aritomo, Y.; Hagino, K.; Chiba, S.; Nishio, K. [Flerov Laboratory of Nuclear Reactions, JINR, Dubna, 141980 (Russian Federation); Department of Physics, Tohoku University, Sendai 980-8578 (Japan); Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo 152-8550 (Japan); Japan Atomic Energy Agency, Tokai, Ibaraki, 319-1195 (Japan)

    2012-10-20T23:59:59.000Z

    To treat heavy-ion reactions using actinide target nucleus, we propose a model which takes into account the coupling to the collective states of interacting nuclei in the penetration of the Coulomb barrier and the dynamical evolution of nuclear shape from the contact configuration. A fluctuation-dissipation model (Langevin equation) was applied in the dynamical calculation, where effect of nuclear orientation at the initial impact on the prolately deformed target nucleus was considered. Using this model, we analyzed the experimental data for the mass distribution of fission fragments (MDFF) in the reaction of {sup 36}S+{sup 238}U at several incident energies. Fusion-fission, quasifission and deep-quasi-fission are separated as different trajectories on the potential energy surface. We estimated the fusion cross section of the reaction.

  11. Yields of neutron-rich nuclei by actinide photofission in giant dipole resonance region

    E-Print Network [OSTI]

    Bhowmick, Debasis; Basu, D N; Chakrabarti, Alok

    2015-01-01T23:59:59.000Z

    Photofission of actinides is studied in the region of nuclear excitation energies that covers the entire giant dipole resonance (GDR) region. A comparative analysis of the behavior of the symmetric and asymmetric modes of photon induced fission as a function of the average excitation energy of the fissioning nucleus is performed. The mass distributions of $^{238}$U photofission fragments are obtained at the endpoint bremsstrahlung energy of 29.1 MeV which corresponds to mean photon energy of 13.7$\\pm$0.3 MeV that coincides with GDR peak for $^{238}$U photofission. The integrated yield of $^{238}$U photofission as well as charge distribution of photofission products are calculated and its role in the production of neutron-rich nuclei and their exoticity is explored.

  12. Method for removing undesired particles from gas streams

    DOE Patents [OSTI]

    Durham, Michael Dean (Castle Rock, CO); Schlager, Richard John (Aurora, CO); Ebner, Timothy George (Westminster, CO); Stewart, Robin Michele (Arvada, CO); Hyatt, David E. (Denver, CO); Bustard, Cynthia Jean (Littleton, CO); Sjostrom, Sharon (Denver, CO)

    1998-01-01T23:59:59.000Z

    The present invention discloses a process for removing undesired particles from a gas stream including the steps of contacting a composition containing an adhesive with the gas stream; collecting the undesired particles and adhesive on a collection surface to form an aggregate comprising the adhesive and undesired particles on the collection surface; and removing the agglomerate from the collection zone. The composition may then be atomized and injected into the gas stream. The composition may include a liquid that vaporizes in the gas stream. After the liquid vaporizes, adhesive particles are entrained in the gas stream. The process may be applied to electrostatic precipitators and filtration systems to improve undesired particle collection efficiency.

  13. Comparison of actinide production in traveling wave and pressurized water reactors

    SciTech Connect (OSTI)

    Osborne, A.G.; Smith, T.A.; Deinert, M.R. [Department of Mechanical Engineering, University of Texas at Austin, Austin, TX (United States)

    2013-07-01T23:59:59.000Z

    The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactor cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)

  14. Hardening Neutron Spectrum for Advanced Actinides Transmutation Experiments in the ATR

    SciTech Connect (OSTI)

    G. S. Chang; R. G. Ambrosek

    2004-05-01T23:59:59.000Z

    The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast rest reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas released modelling, needs to be accurately predicted and the hardened neturon spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are peformed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neturon spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.

  15. Plant Mounds as Concentration and Stabilization Agents for Actinide Soil Contaminants in Nevada

    SciTech Connect (OSTI)

    D.S. Shafer; J. Gommes

    2009-02-03T23:59:59.000Z

    Plant mounds or blow-sand mounds are accumulations of soil particles and plant debris around the base of shrubs and are common features in deserts in the southwestern United States. An important factor in their formation is that shrubs create surface roughness that causes wind-suspended particles to be deposited and resist further suspension. Shrub mounds occur in some plant communities on the Nevada Test Site, the Nevada Test and Training Range (NTTR), and Tonopah Test Range (TTR), including areas of surface soil contamination from past nuclear testing. In the 1970s as part of early studies to understand properties of actinides in the environment, the Nevada Applied Ecology Group (NAEG) examined the accumulation of isotopes of Pu, 241Am, and U in plant mounds at safety experiment and storage-transportation test sites of nuclear devices. Although aerial concentrations of these contaminants were highest in the intershrub or desert pavement areas, the concentration in mounds were higher than in equal volumes of intershrub or desert pavement soil. The NAEG studies found the ratio of contaminant concentration of actinides in soil to be greater (1.6 to 2.0) in shrub mounds than in the surrounding areas of desert pavement. At Project 57 on the NTTR, 17 percent of the area was covered in mounds while at Clean Slate III on the TTR, 32 percent of the area was covered in mounds. If equivalent volumes of contaminated soil were compared between mounds and desert pavement areas at these sites, then the former might contain as much as 34 and 62 percent of the contaminant inventory, respectively. Not accounting for radionuclides associated with shrub mounds would cause the inventory of contaminants and potential exposure to be underestimated. In addition, preservation of shrub mounds could be important part of long-term stewardship if these sites are closed by fencing and posting with administrative controls.

  16. Forecast Technical Document Felling and Removals

    E-Print Network [OSTI]

    Forecast Technical Document Felling and Removals Forecasts A document describing how volume fellings and removals are handled in the 2011 Production Forecast system. Tom Jenkins Robert Matthews Ewan Mackie Lesley Halsall #12;PF2011 ­ Felling and removals forecasts Background A fellings and removals

  17. Removal of deposited copper from nuclear steam generators

    SciTech Connect (OSTI)

    McSweeney, P.

    1982-05-01T23:59:59.000Z

    A review of the copper-removal process implemented during the cleaning of the NPD nuclear steam generator in Ontario revealed that major shortcomings in the process were depletion of the strong ammonia solution and relatively poor copper removal. Tests have shown that the concentration of the ammonia solution can be preserved close to its initial value, and high concentrations of complexed copper obtained, by sparging the ammonia solution with oxygen recirculating through a gas recirculation loop. Using recirculating oxygen for sparging at ambient air temperature, approximately 11 g/l of copper were dissolved by 100 g/l ammonia solution while the gaseous ammonia content of the recirculating gas remained well below the lower flammability limit. The corrosion rates of mild steel and commonly used nuclear steam generator tube materials in oxygenated ammonia solution were less than 30 mil/yr and no intergranular attack of samples was observed during tests. A second technique studied for the removal of copper is to ammoniate the spent iron-removal solvent to approximately pH 9.5 and sparge with recirculating oxygen. Complexed ferric iron in the spent iron-removal solvent was found to be the major oxidizing agent for metallic copper. The ferric iron can be derived from oxidation of dissolved ferrous iron to the ferric state or from dissolved oxides of iron directly. To extract copper from the secondary sides of nuclear steam generators, strong ammonia solution sparged with recirculating oxygen is recommended as the first stage, while ammoniated spent iron-removal solvent sparged with recirculating oxygen may be used to remove the copper freshly exposed during the removal of iron.

  18. Removing sulphur oxides from a fluid stream

    DOE Patents [OSTI]

    Katz, Torsten; Riemann, Christian; Bartling, Karsten; Rigby, Sean Taylor; Coleman, Luke James Ivor; Lail, Marty Alan

    2014-04-08T23:59:59.000Z

    A process for removing sulphur oxides from a fluid stream, such as flue gas, comprising: providing a non-aqueous absorption liquid containing at least one hydrophobic amine, the liquid being incompletely miscible with water; treating the fluid stream in an absorption zone with the non-aqueous absorption liquid to transfer at least part of the sulphur oxides into the non-aqueous absorption liquid and to form a sulphur oxide-hydrophobic amine-complex; causing the non-aqueous absorption liquid to be in liquid-liquid contact with an aqueous liquid whereby at least part of the sulphur oxide-hydrophobic amine-complex is hydrolyzed to release the hydrophobic amine and sulphurous hydrolysis products, and at least part of the sulphurous hydrolysis products is transferred into the aqueous liquid; separating the aqueous liquid from the non-aqueous absorption liquid. The process mitigates absorbent degradation problems caused by sulphur dioxide and oxygen in flue gas.

  19. Removal of impurities from dry scrubbed fluoride enriched alumina

    SciTech Connect (OSTI)

    Schuh, L. [ABB Corporate Research Center, Heidelberg (Germany); Wedde, G. [ABB Environmental, Oslo (Norway)

    1996-10-01T23:59:59.000Z

    The pot-gas from an aluminum electrolytic cell is cleaned by a dry scrubbing process using fresh alumina as a scrubbing agent. This alumina is enriched with fluorides and trace impurities in a closed loop system with the pots. The only significant removal of the impurities is due to metal tapping. An improved technique has been developed that is more effective than earlier stripper systems. The impurity-rich fine fraction (< 10 {micro}m) of the enriched alumina is partly attached to the coarser alumina. That attachment has to be broken. Selective impact milling under special moderate conditions and air classifying have shown to be a cost effective process for the removal of impurities. For iron (Fe) and phosphorus (P) about 30--70% can be removed by the separation of 0.5--1% of the alumina. Full scale tests have successfully confirmed these results.

  20. Purification process

    SciTech Connect (OSTI)

    Marshall, A.

    1981-02-17T23:59:59.000Z

    A process for the removal of hydrogen sulphide from gases or liquid hydrocarbons, comprises contacting the gas or liquid hydrocarbon with an aqueous alkaline solution, preferably having a pH value of 8 to 10, comprising (A) an anthraquinone disulphonic acid or a water-soluble sulphonamide thereof (B) a compound of a metal which can exist in at least two valency states and (C) a sequestering agent.

  1. Speciation and structural characterization of plutonium and actinide-organic complexes in surface and groundwaters. 1998 annual progress report

    SciTech Connect (OSTI)

    Buesseler, K.O.; Repeta, D.J. [Woods Hole Oceanographic Inst., MA (US); Kelley, J.M. [Pacific Northwest National Lab., Richland, WA (US)

    1998-06-01T23:59:59.000Z

    'The authors proposed research is designed to study the association of actinides with dissolved organic complexes in subsurface waters. This study expands considerably on prior work due to the combination of Pu oxidation studies (for Pu speciation/chemical reactivity information), Pu isotope ratio work (for Pu source function information), and the detailed characterization of organic matter in size-fractionated groundwater samples. They have postulated that actinide associations with organic matter may be enhanced due to colloidal biopolymers. This report summarizes work completed after less than 2 years of a 3-year project. Activities thus far have included: (1) the development of sampling techniques to minimize contamination and artifact formation, (2) the separation of Pu isotopes by oxidation state in groundwater, (3) the development of techniques for the separation and identification of organic constituents from natural waters, (4) a study of background Pu and organic carbon concentrations at the proposed study sites, and (5) field work at the Savannah River site (SRS).'

  2. Removal of Natural Steroid Hormones from Wastewater Using

    E-Print Network [OSTI]

    Removal of Natural Steroid Hormones from Wastewater Using Membrane Contactor Processes J O S H U water resources and increased interest in wastewater reclamation for potable reuse. This interest has in the study of wastewater reuse in advanced life support systems (e.g., space missions) because

  3. Gas-phase energies of actinide oxides -- an assessment of neutral and cationic monoxides and dioxides from thorium to curium

    SciTech Connect (OSTI)

    Marcalo, Joaquim; Gibson, John K.

    2009-08-10T23:59:59.000Z

    An assessment of the gas-phase energetics of neutral and singly and doubly charged cationic actinide monoxides and dioxides of thorium, protactinium, uranium, neptunium, plutonium, americium, and curium is presented. A consistent set of metal-oxygen bond dissociation enthalpies, ionization energies, and enthalpies of formation, including new or revised values, is proposed, mainly based on recent experimental data and on correlations with the electronic energetics of the atoms or cations and with condensed-phase thermochemistry.

  4. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    SciTech Connect (OSTI)

    Koch, M.; Kazimi, M.S.

    1991-04-01T23:59:59.000Z

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  5. Removal of field and embedded metal by spin spray etching

    DOE Patents [OSTI]

    Contolini, R.J.; Mayer, S.T.; Tarte, L.A.

    1996-01-23T23:59:59.000Z

    A process of removing both the field metal, such as copper, and a metal, such as copper, embedded into a dielectric or substrate at substantially the same rate by dripping or spraying a suitable metal etchant onto a spinning wafer to etch the metal evenly on the entire surface of the wafer. By this process the field metal is etched away completely while etching of the metal inside patterned features in the dielectric at the same or a lesser rate. This process is dependent on the type of chemical etchant used, the concentration and the temperature of the solution, and also the rate of spin speed of the wafer during the etching. The process substantially reduces the metal removal time compared to mechanical polishing, for example, and can be carried out using significantly less expensive equipment. 6 figs.

  6. Test of the adequacy of using smoothly joined parabolic segments to parametrize the multihumped fission barriers in actinides

    SciTech Connect (OSTI)

    Bhandari, B.S. (Department of Physics, Faculty of Science, University of Garyounis, Benghazi (Libya))

    1990-10-01T23:59:59.000Z

    The adequacy of using smoothly joined parabolic segments to parametrize the multihumped fission barriers has been tested by examining its simultaneous consistency with the three relevant fission observables, namely, the near-barrier fission cross sections, isomeric half-lives, and the ground-state spontaneous fission half-lives of a wide variety of a total of 25 actinide nuclides. The penetrabilities through such multihumped fission barriers have been calculated in the Wentzel-Kramers-Brillouin approximation, and the various fission half-lives have been determined using the formalism given earlier by Nix and Walker. The results of our systematic analysis of these actinide nuclides suggest that such a parametrization is quite adequate at least for the even-even nuclei, as it reproduces satisfactorily their various observed fission characteristics. Major difficulties remain, however, for the odd mass and for the doubly odd nuclei where the calculated ground-state spontaneous fission half-lives are found to be several orders of magnitude larger than those measured. Possible reasons for such discrepancies are discussed. Fission branching ratios of the decay of the shape isomers in various actinide nuclides have also been calculated and are compared with their measured values.

  7. Measurements of actinide-fission product yields in Caliban and Prospero metallic core reactor fission neutron fields

    SciTech Connect (OSTI)

    Casoli, P.; Authier, N. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Laurec, J.; Bauge, E.; Granier, T. [CEA, Centre DIF, 91297 Arpajon (France)

    2011-07-01T23:59:59.000Z

    In the 1970's and early 1980's, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on {sup 233}U, {sup 235}U, and {sup 239}Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energy causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments. (authors)

  8. Nitrogen removal from natural gas using two types of membranes

    DOE Patents [OSTI]

    Baker, Richard W.; Lokhandwala, Kaaeid A.; Wijmans, Johannes G.; Da Costa, Andre R.

    2003-10-07T23:59:59.000Z

    A process for treating natural gas or other methane-rich gas to remove excess nitrogen. The invention relies on two-stage membrane separation, using methane-selective membranes for the first stage and nitrogen-selective membranes for the second stage. The process enables the nitrogen content of the gas to be substantially reduced, without requiring the membranes to be operated at very low temperatures.

  9. AX Tank Farm tank removal study

    SciTech Connect (OSTI)

    SKELLY, W.A.

    1999-02-24T23:59:59.000Z

    This report examines the feasibility of remediating ancillary equipment associated with the 241-AX Tank Farm at the Hanford Site. Ancillary equipment includes surface structures and equipment, process waste piping, ventilation components, wells, and pits, boxes, sumps, and tanks used to make waste transfers to/from the AX tanks and adjoining tank farms. Two remedial alternatives are considered: (1) excavation and removal of all ancillary equipment items, and (2) in-situ stabilization by grout filling, the 241-AX Tank Farm is being employed as a strawman in engineering studies evaluating clean and landfill closure options for Hanford single-shell tanks. This is one of several reports being prepared for use by the Hanford Tanks Initiative Project to explore potential closure options and to develop retrieval performance evaluation criteria for tank farms.

  10. Mexico HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Apply for Our Jobs Our Jobs Working at NNSA Blog Home content Four-Year Plan Mexico HEU Removal Mexico HEU Removal Location Mexico United States 24 24' 35.298" N, 102...

  11. Arsenic removal and stabilization by synthesized pyrite

    E-Print Network [OSTI]

    Song, Jin Kun

    2009-05-15T23:59:59.000Z

    hydride generation atomic absorption spectrometry method for measuring arsenic species (As(III), As(V)). The synthesized pyrite was applied to remove arsenic and its maximum capacity for arsenic removal was measured in batch adsorption experiments to be 3...

  12. Libya HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Plan Libya HEU Removal Libya HEU Removal Location Libya United States 27 34' 9.5448" N, 17 24' 8.4384" E See map: Google Maps Javascript is required to view this map....

  13. Canada HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Plan Canada HEU Removal Canada HEU Removal Location Canada United States 53 47' 24.972" N, 104 35' 23.4384" W See map: Google Maps Javascript is required to view this map....

  14. Israel HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Plan Israel HEU Removal Israel HEU Removal Location Israel United States 30 53' 18.2328" N, 34 52' 14.178" E See map: Google Maps Javascript is required to view this map....

  15. Turkey HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Plan Turkey HEU Removal Turkey HEU Removal Location Turkey United States 38 26' 50.2044" N, 40 15' 14.0616" E See map: Google Maps Javascript is required to view this map...

  16. Uzbekistan HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Uzbekistan HEU Removal Uzbekistan HEU Removal Location Uzbekistan United States 42 6' 56.196" N, 63 22' 8.9076" E See map: Google Maps Javascript is required to view this map...

  17. France HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Four-Year Plan France HEU Removal France HEU Removal Location United States 45 44' 20.0544" N, 2 17' 6.5616" E See map: Google Maps Javascript is required to view this map...

  18. Kazakhstan HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    HEU Removal Kazakhstan HEU Removal Location Kazakhstan United States 48 59' 44.1492" N, 67 3' 37.9692" E See map: Google Maps Javascript is required to view this map....

  19. Ukraine HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Plan Ukraine HEU Removal Ukraine HEU Removal Location Ukraine United States 50 12' 24.8688" N, 25 50' 23.4384" E See map: Google Maps Javascript is required to view this map...

  20. Chile HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Four-Year Plan Chile HEU Removal Chile HEU Removal Location United States 25 28' 1.4916" S, 69 33' 55.548" W See map: Google Maps Javascript is required to view this map...

  1. Taiwan HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Plan Taiwan HEU Removal Taiwan HEU Removal Location Taiwan United States 24 35' 37.4964" N, 120 53' 36.798" E See map: Google Maps Javascript is required to view this map....

  2. Romania HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Plan Romania HEU Removal Romania HEU Removal Location Romania United States 45 47' 1.932" N, 24 41' 50.1576" E See map: Google Maps Javascript is required to view this map....

  3. Hungary HEU removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Plan Hungary HEU removal Hungary HEU removal Location Hungary United States 47 11' 51.6336" N, 19 41' 15" E See map: Google Maps Javascript is required to view this map....

  4. Serbia HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Plan Serbia HEU Removal Serbia HEU Removal Location Serbia United States 44 22' 45.7068" N, 20 26' 4.452" E See map: Google Maps Javascript is required to view this map....

  5. Japan HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Plan Japan HEU Removal Japan HEU Removal Location Japan United States 37 36' 59.5872" N, 140 5' 51.5616" E See map: Google Maps Javascript is required to view this map....

  6. Poland HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Plan Poland HEU Removal Poland HEU Removal Location Poland United States 53 23' 50.2872" N, 17 50' 30.4692" E See map: Google Maps Javascript is required to view this map....

  7. Italy HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Plan Italy HEU Removal Italy HEU Removal Location Italy United States 43 41' 3.4548" N, 11 28' 11.0172" E See map: Google Maps Javascript is required to view this map...

  8. Vietnam HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Plan Vietnam HEU Removal Vietnam HEU Removal Location Vietnam United States 13 12' 30.8628" N, 108 19' 30.702" E See map: Google Maps Javascript is required to view this map....

  9. Method for removal of furfural coke from metal surfaces

    SciTech Connect (OSTI)

    Turner, J.D.

    1990-02-27T23:59:59.000Z

    This patent describes a process for preparing furfural coke for removal from metallic surfaces. It comprises: heating ship furfural coke without causing an evolution of heat capable of undesirably altering metallurgical properties of the surfaces in the presence of a gas with a total pressure of less than 100 psig containing molecular oxygen. The gas being at a sufficient temperature below 800{degrees}F. (427{degrees}C.) for a sufficient time to change the crush strength of the coke so as to permit removal with a water jet at a pressure of about 5000 psi.

  10. Method for removal of beryllium contamination from an article

    DOE Patents [OSTI]

    Simandl, Ronald F.; Hollenbeck, Scott M.

    2012-12-25T23:59:59.000Z

    A method of removal of beryllium contamination from an article is disclosed. The method typically involves dissolving polyisobutylene in a solvent such as hexane to form a tackifier solution, soaking the substrate in the tackifier to produce a preform, and then drying the preform to produce the cleaning medium. The cleaning media are typically used dry, without any liquid cleaning agent to rub the surface of the article and remove the beryllium contamination below a non-detect level. In some embodiments no detectible residue is transferred from the cleaning wipe to the article as a result of the cleaning process.

  11. Methods of hydrotreating a liquid stream to remove clogging compounds

    DOE Patents [OSTI]

    Minderhoud, Johannes Kornelis [Amsterdam, NL; Nelson, Richard Gene [Katy, TX; Roes, Augustinus Wilhelmus Maria [Houston, TX; Ryan, Robert Charles [Houston, TX; Nair, Vijay [Katy, TX

    2009-09-22T23:59:59.000Z

    A method includes producing formation fluid from a subsurface in situ heat treatment process. The formation fluid is separated to produce a liquid stream and a gas stream. At least a portion of the liquid stream is provided to a hydrotreating unit. At least a portion of selected in situ heat treatment clogging compositions in the liquid stream are removed to produce a hydrotreated liquid stream by hydrotreating at least a portion of the liquid stream at conditions sufficient to remove the selected in situ heat treatment clogging compositions.

  12. Removing Stains from Washable Fabrics.

    E-Print Network [OSTI]

    Beard, Ann Vanderpoorten

    1988-01-01T23:59:59.000Z

    Page Numbers Stain Page Numbers Acne medicine Blueberry Special 9 Wet 8 Adhesive tape Dye 8 Special 9 Butter Alcoholic beverages Dry 8 Wet 8 Oil 8 Tannin 8 Calamine lotion Asphalt Combination 8 Combination 8 Dye 8 Dye 8 Candle wax Automotive... the most gentle to the most harsh, so always stop treatments as soon as the stain has been removed. Dry Type Stains Dissolve the stain with a grease solvent. Lubricate the stain with dry spotter, coconut oil or mineral oil (sold in health food...

  13. Automatic Eyeglasses Removal from Face Images

    E-Print Network [OSTI]

    Narasayya, Vivek

    Automatic Eyeglasses Removal from Face Images Chenyu Wu, Ce Liu, Heung-Yueng Shum, Member, IEEE an intelligent image editing and face synthesis system that automatically removes eyeglasses from an input frontal face image. Although conventional image editing tools can be used to remove eyeglasses by pixel

  14. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    SciTech Connect (OSTI)

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01T23:59:59.000Z

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  15. Approach for Validating Actinide and Fission Product Compositions for Burnup Credit Criticality Safety Analyses

    SciTech Connect (OSTI)

    Radulescu, Georgeta [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Wagner, John C [ORNL

    2014-01-01T23:59:59.000Z

    This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application of the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. The validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.

  16. Solubility testing of actinides on breathing-zone and area air samples

    SciTech Connect (OSTI)

    Metzger, R.L.; Jessop, B.H.; McDowell, B.L. [Radiation Safety Engineering, Inc., Chandler, AZ (United States)

    1996-02-01T23:59:59.000Z

    A solubility testing method for several common actinides has been developed with sufficient sensitivity to allow profiles to be determined from routine breathing zone and area air samples in the workplace. Air samples are covered with a clean filter to form a filter-sample-filter sandwich which is immersed in an extracellular lung serum simulant solution. The sample is moved to a fresh beaker of the lung fluid simulant each day for one week, and then weekly until the end of the 28 day test period. The soak solutions are wet ashed with nitric acid and hydrogen peroxide to destroy the organic components of the lung simulant solution prior to extraction of the nuclides of interest directly into an extractive scintillator for subsequent counting on a Photon-Electron Rejecting Alpha Liquid Scintillation (PERALS{reg_sign}) spectrometer. Solvent extraction methods utilizing the extractive scintillators have been developed for the isotopes of uranium, plutonium, and curium. The procedures normally produce an isotopic recovery greater than 95% and have been used to develop solubility profiles from air samples with 40 pCi or less of U{sub 3}O{sub 8}. Profiles developed for U{sub 3}O{sub 8} samples show good agreement with in vitro and in vivo tests performed by other investigators on samples from the same uranium mills.

  17. Flotation machine and process for removing impurities from coals

    DOE Patents [OSTI]

    Szymocha, Kazimierz (Edmonton, CA); Ignasiak, Boleslaw (Edmonton, CA); Pawlak, Wanda (Edmonton, CA); Kulik, Conrad (Newark, CA); Lebowitz, Howard E. (Mountain View, CA)

    1997-01-01T23:59:59.000Z

    The present invention is directed to a type of flotation machine that combines three separate operations in a single unit. The flotation machine is a hydraulic separator that is capable of reducing the pyrite and other mineral matter content of a coal. When the hydraulic separator is used with a flotation system, the pyrite and certain other minerals particles that may have been entrained by hydrodynamic forces associated with conventional flotation machines and/or by the attachment forces associated with the formation of microagglomerates are washed and separated from the coal.

  18. Flotation machine and process for removing impurities from coals

    DOE Patents [OSTI]

    Szymocha, Kazimierz (Edmonton, CA); Ignasiak, Boleslaw (Edmonton, CA); Pawlak, Wanda (Edmonton, CA); Kulik, Conrad (Newark, CA); Lebowitz, Howard E. (Mountain View, CA)

    1995-01-01T23:59:59.000Z

    The present invention is directed to a type of flotation machine that combines three separate operations in a single unit. The flotation machine is a hydraulic separator that is capable of reducing the pyrite and other mineral matter content of a coal. When the hydraulic separator is used with a flotation system, the pyrite and certain other minerals particles that may have been entrained by hydrodynamic forces associated with conventional flotation machines and/or by the attachment forces associated with the formation of microagglomerates are washed and separated from the coal.

  19. High Metal Removal Rate Process for Machining Difficult Materials

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    precision to manufacture parts with complex shapes or micron-sized features. The use of ultrafast (femtosecond) lasers can overcome these limitations and machine advanced...

  20. ITER HEAT REMOVAL SYSTEM SYSTEM & PROCESS CONTROL DESIGN

    E-Print Network [OSTI]

    Raffray, A. René

    normal pulse operation, the heat deposited in the in-vessel components is released into the environment. Ito 1 , P. Lorenzetto 4 , Y. Okawa 5 1 ITER Joint Central Team, 11025 North Torrey Pines Road, La Jolla, CA, 92037, USA; 2 ITER Joint Central Team, Naka, Japan; 3 ITER Joint Central Team, Garching