National Library of Energy BETA

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)

    1996-01-01

    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.

    2014-06-05

    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)

    1987-07-28

    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.

    1985-06-13

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

    1992-03-24

    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.

  6. Actinide metal processing

    DOE Patents [OSTI]

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

    1992-01-01

    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.

  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)

    2002-01-01

    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. Process for recovering actinide values

    DOE Patents [OSTI]

    Horwitz, E. Philip (Elmhurst, IL); Mason, George W. (Clarendon Hills, IL)

    1980-01-01

    A process for rendering actinide values recoverable from sodium carbonate scrub waste solutions containing these and other values along with organic compounds resulting from the radiolytic and hydrolytic degradation of neutral organophosphorous extractants such as tri-n butyl phosphate (TBP) and dihexyl-N,N-diethyl carbamylmethylene phosphonate (DHDECAMP) which have been used in the reprocessing of irradiated nuclear reactor fuels. The scrub waste solution is preferably made acidic with mineral acid, to form a feed solution which is then contacted with a water-immiscible, highly polar organic extractant which selectively extracts the degradation products from the feed solution. The feed solution can then be processed to recover the actinides for storage or recycled back into the high-level waste process stream. The extractant is recycled after stripping the degradation products with a neutral sodium carbonate solution.

  9. In vitro removal of actinide (IV) ions

    DOE Patents [OSTI]

    Weitl, Frederick L. (Martinez, CA); Raymond, Kenneth N. (Berkeley, CA)

    1982-01-01

    A compound of the formula: ##STR1## wherein X is hydrogen or a conventional electron-withdrawing group, particularly --SO.sub.3 H or a salt thereof; n is 2, 3, or 4; m is 2, 3, or 4; and p is 2 or 3. The present compounds are useful as specific sequestering agents for actinide (IV) ions. Also described is a method for the 2,3-dihydroxybenzamidation of azaalkanes.

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

    SciTech Connect (OSTI)

    Eibling, R.

    2011-09-28

    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.

  11. Actinide and lanthanide separation process (ALSEP)

    DOE Patents [OSTI]

    Guelis, Artem V.

    2013-01-15

    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).

  12. Pyrometallurgical processes for recovery of actinide elements

    SciTech Connect (OSTI)

    Battles, J.E.; Laidler, J.J.; McPheeters, C.C.; Miller, W.E.

    1994-01-01

    A metallic fuel alloy, nominally U-20-Pu-lOZr, is the key element of the Integral Fast Reactor (IFR) fuel cycle. Metallic fuel permits the use of an innovative, simple pyrometallurgical process, known as pyroprocessing, (the subject of this report), which features fused salt electrorefining of the spent fuel. Electrorefining separates the actinide elements from fission products, without producing a separate stream of plutonium. The plutonium-bearing product is contaminated with higher actinides and with a minor amount of rare earth fission products, making it diversion resistant while still suitable as a fuel material in the fast spectrum of the IFR core. The engineering-scale demonstration of this process will be conducted in the refurbished EBR-II Fuel Cycle Facility, which has entered the start-up phase. An additional pyrometallurgical process is under development for extracting transuranic (TRU) elements from Light Water Reactor (LWR) spent fuel in a form suitable for use as a feed to the IFR fuel cycle. Four candidate extraction processes have been investigated and shown to be chemically feasible. The main steps in each process are oxide reduction with calcium or lithium, regeneration of the reductant and recycle of the salt, and separation of the TRU product from the bulk uranium. Two processes, referred to as the lithium and salt transport (calcium reductant) processes, have been selected for engineering-scale demonstration, which is expected to start in late 1993. An integral part of pyroprocessing development is the treatment and packaging of high-level waste materials arising from the operations, along with the qualification of these waste forms for disposal in a geologic repository.

  13. 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)

    2001-01-01

    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.

  14. Process to remove rare earth from IFR electrolyte

    DOE Patents [OSTI]

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

    1992-01-01

    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. Process to remove rare earth from IFR electrolyte

    DOE Patents [OSTI]

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

    1994-01-01

    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.

  16. Process to remove rare earth from IFR electrolyte

    DOE Patents [OSTI]

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

    1994-08-09

    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.

  17. Continuous sulfur removal process

    DOE Patents [OSTI]

    Jalan, deceased, Vinod (late of Concord, MA); Ryu, Jae (Cambridge, MA)

    1994-01-01

    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.

  18. Continuous sulfur removal process

    DOE Patents [OSTI]

    Jalan, V.; Ryu, J.

    1994-04-26

    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.

  19. Key features of the Talspeak and similar trivalent actinide-lanthanide partitioning processes

    SciTech Connect (OSTI)

    Nash, Kenneth L.

    2008-07-01

    As closing of the nuclear-fuel cycle via the suite of UREX processes under development in the U.S. progresses, the Trivalent Actinide-Lanthanide Separation by Phosphorus Extractants and Aqueous Komplexants (TALSPEAK) process has been selected as the baseline process for partition of trivalent actinides away from fission-product lanthanides. In this report, selected features of the chemistry of the TALSPEAK process and the limited parallel information on other TALSPEAK-like processes are discussed. (author)

  20. Analysis of large soil samples for actinides

    DOE Patents [OSTI]

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

    2009-03-24

    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.

  1. Actinide solution processing at the Rocky Flats Environmental Technology Site

    SciTech Connect (OSTI)

    1995-04-01

    The Department of Energy (DOE) has prepared an Environmental Assessment (EA), DOE/EA-1039, for radioactive solution removal and processing at Rocky Flats Environmental Technology Site, Golden, Colorado. The proposal for solution removal and processing is in response to independent safety assessments and an agreement with the State of Colorado to remove mixed residues at Rocky Flats and reduce the risk of future accidents. Monthly public meetings were held during the scoping and preparation of the EA. The scope of the EA included evaluations of alternative methods and locations of solution processing. A comment period from February 20, 1995 through March 21, 1995 was provided to the public and the State of Colorado to offer written comment on the EA. Comments were received from the State of Colorado and the U.S. Environmental Protection Agency. A response to the agency comments is included in the Final EA.

  2. Carbon dioxide removal process

    DOE Patents [OSTI]

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

    2003-11-18

    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.

  3. 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

    2009-07-01

    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.

  4. The use of safeguards data for process monitoring in the Advanced Test Line for Actinide Separations

    SciTech Connect (OSTI)

    Barnes, J.W.; Yarbro, S.L.

    1987-01-01

    Los Alamos is constructing an integrated process monitoring/materials control and accounting (PM/MC and A) system in the Advanced Testing Line for Actinide Separations (ATLAS) at the Los Alamos Plutonium Facility. The ATLAS will test and demonstrate new methods for aqueous processing of plutonium. The ATLAS will also develop, test, and demonstrate the concepts for integrated process monitoring/materials control and accounting. We describe how this integrated PM/MC and A system will function and provide benefits to both process research and materials accounting personnel.

  5. 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)

    2013-07-01

    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)

  6. 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

    2010-09-21

    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.

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

    SciTech Connect (OSTI)

    Kenneth L. Nash; Sue B. Clark; Gregg Lumetta

    2009-09-23

    With increased application of MOX fuels and longer burnup times for conventional fuels, higher concentrations of the transplutonium actinides Am and Cm (and even heavier species like Bk and Cf) will be produced. The half-lives of the Am isotopes are significantly longer than those of the most important long-lived, high specific activity lanthanides or the most common Cm, Bk and Cf isotopes, thus the greatest concern as regards long-term radiotoxicity. With the removal and transmutation of Am isotopes, radiation levels of high level wastes are reduced to near uranium mineral levels within less than 1000 years as opposed to the time-fram if they remain in the wastes.

  8. Process for removing metals from water

    DOE Patents [OSTI]

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

    1987-06-29

    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.

  9. Process for removing mercury from aqueous solutions

    DOE Patents [OSTI]

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

    1985-03-04

    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. Process for removing sulfur from coal

    DOE Patents [OSTI]

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

    1983-08-11

    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.

  11. Partitioning of trivalent actinides from a Purex raffinate using a TODGA-based solvent-extraction process

    SciTech Connect (OSTI)

    Modolo, G.; Vijgen, H.; Malmbeck, R.; Magnusson, D.; Sorel, C.

    2008-07-01

    A TODGA/TBP process has been developed to separate trivalent actinides from a PUREX raffinate using a mixture of tetraoctyl-diglycolamide (TODGA) and tributylphosphate (TBP). Batch extraction experiments allowed us to choose and optimize the composition of the organic extractant and the aqueous feed solutions. With the aid of computer-code calculations, a countercurrent process has been developed, and an optimized flowsheet has been tested with a spiked feed solution and finally with a genuine PUREX raffinate. The results of the two tests were very promising, demonstrating that more than 99.9% of the trivalent actinides are extracted, and very high decontamination factors are obtained to the non-lanthanide fission products. Co-extracted ruthenium (10% during spiked test, 18% during hot test) is less efficiently back-extracted and therefore requires further process development. (authors)

  12. DISTRIBUTION OF ACTINIDES BETWEEN THE AQUEOUS AND ORGANIC PHASES IN THE TALSPEAK PROCESS

    SciTech Connect (OSTI)

    Rudisill, T.; Kyser, E.

    2010-09-02

    One objective of the US Department of Energy's Office of Nuclear Energy (DOE-NE) is the development of sustainable nuclear fuel cycles which improve uranium resource utilization, maximize energy generation, minimize waste generation, improve safety, and complement institutional measures limiting proliferation risks. Activities in progress which support this objective include the development of advanced separation technologies to recover the actinides from used nuclear fuels. With the increased interest in the development of technology to allow 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, at this time, the level of understanding associated with the chemistry and the control of the process variables is not acceptable for deployment of the process on an industrial scale. To address this issue, DOE-NE is supporting basic scientific studies focused on the TALSPEAK process through its Fuel Cycle Research and Development (R&D) program. One aspect of these studies is an experimental program at the Savannah River National Laboratory (SRNL) in which temperature-dependent distribution coefficients for the extraction of actinide elements in the TALSPEAK process were measured. The data were subsequently used to calculate conditional enthalpies and entropies of extraction by van't Hoff analysis to better understand the thermodynamic driving forces for the TALSPEAK process. In the SRNL studies, the distribution of Pu(III) in the TALSPEAK process was of particular interest. A small amount of Pu(III) would be present in the feed due to process losses and valence adjustment in prior recovery operations. Actinide elements such as Np and Pu have multiple stable oxidation states in aqueous solutions; therefore the oxidation state for these elements must be controlled in the TALSPEAK process, as the extraction chemistry is dependent upon the actinide's valence. Since our plans included the measurement of Pu(III) distribution coefficients using a Np(V) solution containing small amounts of {sup 238}Pu, it was necessary to demonstrate that the desired oxidation states of Np and Pu are produced and could be stabilized in a buffered lactate solution containing diethylenetriaminepentaacetic (DTPA). The stability of Np(V) and Pu(III) in lactic acid/DTPA solutions was evaluated by ultraviolet-visible (UV-vis) spectroscopy. To perform the evaluation, Np and Pu were added to solutions containing either hydroxylamine nitrate (HAN) or ferrous sulfamate (FS) as the reductant and nominally 1.5 M lactic acid/0.05 M DTPA. The pH of the solution was subsequently adjusted to nominally 2.8 as would be performed in the TALSPEAK process. In the valence adjustment study, we found that it was necessary to reduce Pu to Pu(III) prior to combining with the lactic acid and DTPA. The Pu reduction was performed using either HAN or FS. When FS was used, Np was reduced to Np(IV). The spectroscopic studies showed that Np(V) and Pu(III) are not stable in lactic acid/DTPA solutions. The stability of Np(IV)- and Pu(IV)-DTPA complexes are much greater than the stability of the Np(V)- and Pu(III)-DTPA complexes, and as a result, Np is slowly reduced to Np(IV) and Pu is slowly oxidized to Pu(IV) due to the reduced activity of the more stable complexes. When Np(V) was added to a solution containing a 1.5 M lactic acid/ammonium lactate buffer and 0.05 M DTPA, approximately 50% of the Np was reduced to Np(IV) in the first day. The fraction of Np(V) in the solution continued to diminish with time and was essentially reduced to Np(IV) after one week. When Pu(III) was added to a lactic acid/DTPA solution of the same composition, the spectrum recorded following at least two days after preparation of the solution continued to show some sign of Pu(III). The Pu(III) was completely oxidized to Pu(IV) after 3-4 days. The UV-vis spectroscopy demonstrated that Np(V) and Pu(III) were the predominate valences in the lactic acid/DTPA solution for th

  13. Thermochemistry of the actinides

    SciTech Connect (OSTI)

    Kleinschmidt, P.D.

    1993-10-01

    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.

  14. Actinides-1981

    SciTech Connect (OSTI)

    Not Available

    1981-09-01

    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.

  15. Process for removing polychlorinated biphenyls from soil

    DOE Patents [OSTI]

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

    1984-11-16

    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.

  16. IMPROVED PROCESSES TO REMOVE NAPHTHENIC ACIDS

    SciTech Connect (OSTI)

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

    2004-04-28

    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.

  17. Improved Processes to Remove Naphthenic Acids

    SciTech Connect (OSTI)

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

    2005-12-09

    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.

  18. Process for removing technetium from iron and other metals

    DOE Patents [OSTI]

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

    1999-03-23

    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.

  19. Process for particulate removal from coal liquids

    DOE Patents [OSTI]

    Rappe, Gerald C. (Macungie, PA)

    1983-01-01

    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.

  20. More Economical Sulfur Removal for Fuel Processing Plants

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

    crude oil it is present in sulfur-containing organic compounds which are converted into hydrocarbons and H 2 S during the removal process (hydrodesulfurization). In both cases,...

  1. Process for removing metal carbonyls from gaseous streams

    SciTech Connect (OSTI)

    Heyd, R.L.; Pignet, T.P.

    1988-04-26

    A process for removing metal carbonyl contaminates from a gaseous stream is described containing such contaminates and which is free from sulfur contaminates, which process comprises contacting the gaseous stream with a zinc sulfide absorbent to thereby remove metal carbonyl contaminates from the gaseous stream, and separating the gaseous stream from the zinc sulfide absorbent.

  2. IMPROVED PROCESSES TO REMOVE NAPHTHENIC ACIDS

    SciTech Connect (OSTI)

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

    2005-05-05

    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.

  3. 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)

    1990-01-01

    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.

  4. Process for selected gas oxide removal by radiofrequency catalysts

    DOE Patents [OSTI]

    Cha, C.Y.

    1993-09-21

    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.

  5. ITER HEAT REMOVAL SYSTEM SYSTEM & PROCESS CONTROL DESIGN

    E-Print Network [OSTI]

    Raffray, A. René

    in-vessel component cooling systems malfunction. Therefore, special design criteria are requestedITER HEAT REMOVAL SYSTEM SYSTEM & PROCESS CONTROL DESIGN Y. Hoshi 1 , Y. Kataoka 2 , V. Tanchuk 1 the vacuum vessel cooling system has a safety role of providing the ultimate decay heat removal system ,when

  6. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 OutreachProductswsicloudwsicloudden DocumentationAccommodations AccommodationsAcquisitionAcronymsActinide

  7. RAPID SEPARATION OF ACTINIDES AND RADIOSTRONTIUM IN VEGETATION SAMPLES

    SciTech Connect (OSTI)

    Maxwell, S.

    2010-06-01

    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.

  8. Process for removing cadmium from scrap metal

    DOE Patents [OSTI]

    Kronberg, J.W.

    1994-01-01

    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.

  9. Process for removing cadmium from scrap metal

    DOE Patents [OSTI]

    Kronberg, J.W.

    1995-04-11

    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.

  10. Process for removing cadmium from scrap metal

    DOE Patents [OSTI]

    Kronberg, James W. (Aiken, SC)

    1995-01-01

    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 expose additional surface area. Finally, the precipitated cadmium sulfide can be mixed with glass formers and vitrified for permanent disposal.

  11. 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)

    1991-01-01

    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.

  12. Process for removing sulfur from sulfur-containing gases

    DOE Patents [OSTI]

    Rochelle, Gary T. (Austin, TX); Jozewicz, Wojciech (Chapel Hill, NC)

    1989-01-01

    The present disclosure relates to improved processes for treating hot sulfur-containing flue gas to remove sulfur therefrom. Processes in accorda The government may own certain rights in the present invention pursuant to EPA Cooperative Agreement CR 81-1531.

  13. The CNG process: Acid gas removal with liquid carbon dioxide

    SciTech Connect (OSTI)

    Liu, Y.C.; Auyang, L.; Brown, W.R.

    1987-01-01

    The CNG acid gas removal process has two unique features: the absorption of sulfur-containing compounds and other trace contaminants with liquid carbon dioxide, and the regeneration of pure liquid carbon dioxide by triple-point crystallization. The process is especially suitable for treating gases which contain large amounts of carbon dioxide and much smaller amounts (relative to carbon dioxide) of hydrogen sulfide. Capital and energy costs are lower than conventional solvent processes. Further, products of the CNG process meet stringent purity specifications without undue cost penalties. A process demonstration unit has been constructed and operated to demonstrate the two key steps of the CNG process. Hydrogen sulfide and carbonyl sulfide removal from gas streams with liquid carbon dioxide absorbent to sub-ppm concentrations has been demonstrated. The production of highly purified liquid carbon dioxide (less than 0.1 ppm total contaminant) by triple-point crystallization also has been demonstrated.

  14. Isotope Tracer Studies of Diffusion in Sillicates and of Geological Transport Processes Using Actinide Elements

    SciTech Connect (OSTI)

    Wasserburg, Gerald J

    2008-07-31

    The objectives were directed toward understanding the transport of chemical species in nature, with particular emphasis on aqueous transport in solution, in colloids, and on particles. Major improvements in measuring ultra-low concentrations of rare elements were achieved. We focused on two areas of studies: (1) Field, laboratory, and theoretical studies of the transport and deposition of U, Th isotopes and their daughter products in natural systems; and (2) Study of calcium isotope fractionation effects in marine carbonates and in carbonates precipitated in the laboratory, under controlled temperature, pH, and rates of precipitation. A major study of isotopic fractionation of Ca during calcite growth from solution has been completed and published. It was found that the isotopic shifts widely reported in the literature and attributed to biological processes are in fact due to a small equilibrium fractionation factor that is suppressed by supersaturation of the solution. These effects were demonstrated in the laboratory and with consideration of the solution conditions in natural systems, where [Ca{sup 2+}] >> [CO{sub 3}{sup 2-}] + [HCO{sub 3}{sup -}]. The controlling rate is not the diffusion of Ca, as was earlier proposed, but rather the rate of supply of [CO{sub 3}{sup 2-}] ions to the interface. This now opens the issues of isotopic fractionation of many elements to a more physical-chemical approach. The isotopic composition of Ca {Delta}({sup 44}Ca/{sup 40}Ca) in calcite crystals has been determined relative to that in the parent solutions by TIMS using a double spike. Solutions were exposed to an atmosphere of NH{sub 3} and CO{sub 2}, provided by the decomposition of (NH4)2CO3. Alkalinity, pH, and concentrations of CO{sub 3}{sup 2-}, HCO{sub 3}{sup -}, and CO{sub 2} in solution were determined. The procedures permitted us to determine {Delta}({sup 44}Ca/{sup 40}Ca) over a range of pH conditions, with the associated ranges of alkalinity. Two solutions with greatly different Ca concentrations were used, but, in all cases, the condition [Ca] >> [CO{sub 3}{sup 2-}] was met. A wide range in {Delta}({sup 44}Ca/{sup 40}Ca) was found for the calcite crystals, extending from 0.04 {+-} 0.13 to -1.34 {+-} 0.15 {per_thousand}, generally anticorrelating with the amount of Ca removed from the solution. The results show that {Delta}({sup 44}Ca/{sup 40}Ca) is a linear function of the saturation state of the solution with respect to calcite ({Omega}). The two parameters are very well correlated over a wide range in {Omega} for each solution with a given [Ca]. Solutions, which were vigorously stirred, showed a much smaller range in {Delta}({sup 44}Ca/{sup 40}Ca) and gave values of -0.42 {+-} 0.14 {per_thousand}, with the largest effect at low {Omega}. It is concluded that the diffusive flow of CO{sub 3}{sup 2-} into the immediate neighborhood of the crystal-solution interface is the rate-controlling mechanism and that diffusive transport of Ca{sup 2+} is not a significant factor. The data are simply explained by the assumptions that: (a) the immediate interface of the crystal and the solution is at equilibrium with {Delta}({sup 44}Ca/{sup 40}Ca) {approx} -1.5 {+-} 0.25 {per_thousand}, and (b) diffusive inflow of CO{sub 3}{sup 2-} causes supersaturation, thus precipitating Ca from the regions, exterior to the narrow zone of equilibrium. We consider this model to be a plausible explanation of the available data reported in the literature. The well-resolved but small and regular isotope fractionation shifts in Ca are thus not related to the diffusion of very large hydrated Ca complexes, but rather due to the ready availability of Ca in the general neighborhood of the crystal solution interface. The largest isotopic shift which occurs is a small equilibrium effect which is then subdued by supersaturation precipitation for solutions where [Ca{sup 2+}] >> [CO{sub 3}{sup 2-}] + [HCO{sub 3}{sup -}]. It is shown that there is a clear temperature dependence of the net isotopic shifts, which is simply due to changes in {Omega}

  15. Minor actinide separation: simplification of the DIAMEX-SANEX strategy by means of novel SANEX processes

    SciTech Connect (OSTI)

    Geist, A. [Karlsruher Institut fuer Technologie - KIT, INE, P. O. Box 3640, 76021 Karlsruhe (Germany); Modolo, G.; Wilden, A.; Kaufholz, P. [Forschungszentrum Juelich GmbH, IEK-6, Juelich (Germany)

    2013-07-01

    The separation of An(III) from PUREX raffinate has previously been demonstrated by applying a DIAMEX process (i.e., co-extraction of An(III) and Ln(III) from HAR) followed by a SANEX process (i.e., selective extraction of An(III) from the DIAMEX product containing An(III) + Ln(III)). In line with process intensification issues, more compact processes have been developed: Recently, a 1c-SANEX process test was successfully performed, directly extracting An(III) from PUREX HAR. More recently, a new i-SANEX process was successfully tested. This process is based on the co-extraction of An(III) + Ln(III) into a TODGA solvent, followed by a selective back-extraction of An(III) by a water soluble complexing agent, in this case SO{sub 3}-Ph-BTP. In both cases, good recoveries were achieved, and very pure product solutions were obtained. However, both 1c-SANEX and i-SANEX used non-CHON chemicals. Nevertheless, these processes are a simplification to the DIAMEX + SANEX process as only one solvent is used. Finally, the new i-SANEX process is the most compact process. (authors)

  16. 33rd Actinide Separations Conference

    SciTech Connect (OSTI)

    McDonald, L M; Wilk, P A

    2009-05-04

    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.

  17. Magnetic process for removing heavy metals from water employing magnetites

    DOE Patents [OSTI]

    Prenger, F. Coyne; Hill, Dallas D.; Padilla, Dennis D.; Wingo, Robert M.; Worl, Laura A.; Johnson, Michael D.

    2003-07-22

    A process for removing heavy metals from water is provided. The process includes the steps of introducing magnetite to a quantity of water containing heavy metal. The magnetite is mixed with the water such that at least a portion of, and preferably the majority of, the heavy metal in the water is bound to the magnetite. Once this occurs the magnetite and absorbed metal is removed from the water by application of a magnetic field. In most applications the process is achieved by flowing the water through a solid magnetized matrix, such as steel wool, such that the magnetite magnetically binds to the solid matrix. The magnetized matrix preferably has remnant magnetism, but may also be subject to an externally applied magnetic field. Once the magnetite and associated heavy metal is bound to the matrix, it can be removed and disposed of, such as by reverse water or air and water flow through the matrix. The magnetite may be formed in-situ by the addition of the necessary quantities of Fe(II) and Fe(III) ions, or pre-formed magnetite may be added, or a combination of seed and in-situ formation may be used. The invention also relates to an apparatus for performing the removal of heavy metals from water using the process outlined above.

  18. Magnetic process for removing heavy metals from water employing magnetites

    DOE Patents [OSTI]

    Prenger, F. Coyne; Hill, Dallas D.

    2006-12-26

    A process for removing heavy metals from water is provided. The process includes the steps of introducing magnetite to a quantity of water containing heavy metal. The magnetite is mixed with the water such that at least a portion of, and preferably the majority of, the heavy metal in the water is bound to the magnetite. Once this occurs the magnetite and absorbed metal is removed from the water by application of a magnetic field. In most applications the process is achieved by flowing the water through a solid magnetized matrix, such as steel wool, such that the magnetite magnetically binds to the solid matrix. The magnetized matrix preferably has remnant magnetism, but may also be subject to an externally applied magnetic field. Once the magnetite and associated heavy metal is bound to the matrix, it can be removed and disposed of, such as by reverse water or air and water flow through the matrix. The magnetite may be formed in-situ by the addition of the necessary quantities of Fe(II) and Fe(III) ions, or pre-formed magnetite may be added, or a combination of seed and in-situ formation may be used. The invention also relates to an apparatus for performing the removal of heavy metals from water using the process outlined above.

  19. Sodium removal process development for LMFBR fuel subassemblies

    SciTech Connect (OSTI)

    Simmons, C.R.; Taylor, G.R.

    1981-10-01

    Two 37-pin scale models of Clinch River Breeder Reactor Plant fuel subassemblies were designed, fabricated and used at Westinghouse Advanced Reactors Division in the development and proof-testing of a rapid water-based sodium removal process for the ORNL Hot Experimental Facility, Liquid Metal Fast Breeder Reactor Fuel Reprocessing Cycle. Through a series of development tests on one of the models, including five (5) sodium wettings and three (3) high temperature sodium removal operations, optimum process parameters for a rapid water vapor-argon-water rinse process were identified and successfully proof-tested on a second model containing argon-pressurized, sodium-corroded model fuel pins simulating the gas plenum and cladding conditions expected for spent fuel pins in full scale subassemblies. Based on extrapolations of model proof test data, preliminary process parameters for a water vapor-nitrogen-water rinse process were calculated and recommended for use in processing full scale fuel subassemblies in the Sodium Removal Facility of the Fuel Receiving Cell, ORNL HEF.

  20. 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)

    1993-12-28

    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.

  1. Metal chelate process to remove pollutants from fluids

    DOE Patents [OSTI]

    Chang, S.G.T.

    1994-12-06

    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.

  2. Vaporization of actinide oxides in thermal treatment processes for mixed waste

    SciTech Connect (OSTI)

    Ebbinghaus, B.B.; Krikorian, O.H.; Adamson, M.G.

    1994-10-04

    The purpose of this study is to evaluate the volatilities of U, Pu, and Am in thermal treatment processes for mixed wastes. The thermodynamics of vaporization U and Pu oxides in the presence of oxygen and water vapor and of U oxide in the presence of oxygen and chlorine were studied. Experimental studies of U oxide volatilities by previous authors have also been reviewed. For species where data are unavailable estimation methods were used to obtain free energies of formation of the gaseous species, The data are applied to thermal treatment processes in general and then more specifically to conditions representative of the Rocky Flats Plant Fluidized Bed Unit. (RFP FBU), molten salt oxidizer, and an incinerator. U volatilities are greatest ranging from 2.67 x 10{sup -7} gU/h in the RFP FBU to 4. 00 gU/h for typical incinerator conditions. Pu volatilities are almost 5 orders of magnitude less than U and Am volatilities are about 3 orders of magnitude less than Pu.

  3. Actinide separations conference

    SciTech Connect (OSTI)

    Not Available

    1990-01-01

    This report contains the abstracts for 55 presentations given at the fourteenth annual Actinide Separations Conference. (JDL)

  4. 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)

    1997-01-01

    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.

  5. Process for removing carbonyl-sulfide from liquid hydrocarbon feedstocks

    SciTech Connect (OSTI)

    Debras, G.L.G.; DeClippeleir, G.E.M.J.; Cahen, R.M.

    1986-09-23

    A process is described for removing carbonyl sulfide from a liquid olefinic hydrocarbon feedstock comprising: (a) passing the hydrocarbon feedstock over an absorbent material comprising zinc oxide and a promoter selected from the group consisting of alumina, silico-aluminas and any combination thereof wherein the promoter is present in amounts from about 3 to about 15 percent by weight of the absorbent material; and (b) recovering a liquid olefinic hydrocarbon stream having a substantially reduced carbonyl sulfide content.

  6. 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

    1995-12-01

    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.

  7. Process for the removal of impurities from combustion fullerenes

    DOE Patents [OSTI]

    Alford, J. Michael; Bolskar, Robert

    2005-08-02

    The invention generally relates to purification of carbon nanomaterials, particularly fullerenes, by removal of PAHs and other hydrocarbon impurities. The inventive process involves extracting a sample containing carbon nanomaterials with a solvent in which the PAHs are substantially soluble but in which the carbon nanomaterials are not substantially soluble. The sample can be repeatedly or continuously extracted with one or more solvents to remove a greater amount of impurities. Preferred solvents include ethanol, diethyl ether, and acetone. The invention also provides a process for efficiently separating solvent extractable fullerenes from samples containing fullerenes and PAHs wherein the sample is extracted with a solvent in which both fullerenes and PAHs are substantially soluble and the sample extract then undergoes selective extraction to remove PAHs. Suitable solvents in which both fullerenes and PAHs are soluble include o-xylene, toluene, and o-dichlorobenzene. The purification process is capable of treating quantities of combustion soot in excess of one kilogram and can produce fullerenes or fullerenic soot of suitable purity for many applications.

  8. Process for removing carbonyl sulfide from gaseous streams

    SciTech Connect (OSTI)

    Tellis, C.

    1981-11-10

    This invention relates to a process for reducing the carbonyl sulfide content of a gaseous stream which has a concentration of carbonyl sulfide of from at least 1 to about 100 parts per million, by volume, which comprises providing an absorbent bed wherein the absorbent comprises zinc oxide and contains no more than 5%, by weight, of an oxide of an alkli or alkaline earth metal, and contacting said process stream with said adsorbent bed at a temperature of from about ambient to 250/sup 0/ C. For a period of time sufficient to remove at least 90% of the carbonyl sulfide content of said gaseous stream.

  9. 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)

    2001-01-01

    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.

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

    DOE Patents [OSTI]

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

    2011-08-16

    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.

  11. Process for removal of sulfur compounds from fuel gases

    DOE Patents [OSTI]

    Moore, Raymond H. (Richland, WA); Stegen, Gary E. (Richland, WA)

    1978-01-01

    Fuel gases such as those produced in the gasification of coal are stripped of sulfur compounds and particulate matter by contact with molten metal salt. The fuel gas and salt are intimately mixed by passage through a venturi or other constriction in which the fuel gas entrains the molten salt as dispersed droplets to a gas-liquid separator. The separated molten salt is divided into a major and a minor flow portion with the minor flow portion passing on to a regenerator in which it is contacted with steam and carbon dioxide as strip gas to remove sulfur compounds. The strip gas is further processed to recover sulfur. The depleted, minor flow portion of salt is passed again into contact with the fuel gas for further sulfur removal from the gas. The sulfur depleted, fuel gas then flows through a solid absorbent for removal of salt droplets. The minor flow portion of the molten salt is then recombined with the major flow portion for feed to the venturi.

  12. 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)

    2000-01-01

    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.

  13. Extraction process for removing metallic impurities from alkalide metals

    DOE Patents [OSTI]

    Royer, L.T.

    1987-03-20

    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.

  14. 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)

    1992-01-01

    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.

  15. Actinide halide complexes

    DOE Patents [OSTI]

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

    1992-11-24

    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.

  16. Chemical Addition prior to Membrane Processes for Natural Organic Matter (NOM) Removal 

    E-Print Network [OSTI]

    Schäfer, Andrea; Fane, Anthony G.; Waite, T. D.

    1998-01-01

    Membrane processes for surface water treatment include microfiltration (MF), ultrafiltration (UF) and nanofiltration (NF), depending on the target material to be removed and the limiting process economics. MF will remove ...

  17. 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

    2014-04-15

    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.

  18. 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)

    1992-01-01

    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.

  19. Advanced Aqueous Separation Systems for Actinide Partitioning

    SciTech Connect (OSTI)

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

    2012-03-21

    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.

  20. Overview of actinide chemistry in the WIPP

    SciTech Connect (OSTI)

    Borkowski, Marian; Lucchini, Jean - Francois; Richmann, Michael K; Reed, Donald T; Khaing, Hnin; Swanson, Juliet

    2009-01-01

    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].

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

    E-Print Network [OSTI]

    Mustafa, Syed Faisal

    2014-07-23

    solubility. Chloride can be removed from water and wastewater by precipitation as calcium chloroaluminate using advanced softening process. This research was conducted to evaluate chloride removal using electrochemically generated aluminum hydroxide and lime...

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

    DOE Patents [OSTI]

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

    1983-07-26

    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.

  3. Process for removing carbonyl sulfide from hydrocarbon feedstreams

    SciTech Connect (OSTI)

    Holmes, E.S.; Kosseim, A.J.

    1992-04-14

    This patent describes a process for removing carbon dioxide, hydrogen sulfide and carbonyl sulfide from a feedstream containing carbon dioxide, hydrogen sulfide and carbonyl sulfide and hydrocarbons. It comprises: contacting the feedstream in a hydrolysis zone with a first portion of a lean solution stream comprising an aqueous alkaline solution at an effective hydrolysis temperature to convert at least a portion of the carbonyl sulfide to carbon dioxide and hydrogen sulfide, withdrawing a first effluent stream containing a reduced concentration of carbonyl sulfide relative to the feedstream, and withdrawing a first rich solution stream comprising the aqueous alkaline solution, carbon dioxide and hydrogen sulfide; contacting the first effluent stream in an absorption zone with a second portion of the lean solution stream at an effective absorption temperature to absorb carbon dioxide and hydrogen sulfide, and withdrawing a second rich solution stream comprising the aqueous alkaline solution, carbon dioxide and hydrogen sulfide; combining at least a portion of the first rich solution stream and the second rich solution stream and contacting the combined rich solution stream in a regeneration zone at effective conditions to desorb carbon dioxide and hydrogen sulfide, withdrawing a vent gas stream comprising carbon dioxide and hydrogen sulfide, and withdrawing the lean solution stream; separating the lean solution stream into the first and second portions; and recycling the first portion of the lean solution stream to the hydrolysis zone and the second portion of the lean solution stream to the absorption zone.

  4. 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. ª 2008 Elsevier Ltd. All rights reserved. 1. Introduction The removal

  5. Process for the removal of acid gases from gaseous streams

    SciTech Connect (OSTI)

    Blytas, G.C.; Diaz, Z.

    1982-11-16

    Hydrogen sulfide, carbon dioxide, and carbonyl sulfide are removed from a gas stream in a staged procedure by: absorption of the CO/sub 2/ and COS; conversion of the hydrogen sulfide to produce sulfur in an absorbent mixture; hydrolysis of the carbonyl sulfide to produce a gas stream of hydrogen sulfide and carbon dioxide; and removal of the hydrogen sulfide from the gas stream.

  6. Fluidized bed gasification ash reduction and removal process

    DOE Patents [OSTI]

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

    1984-12-04

    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.

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

    SciTech Connect (OSTI)

    Gary M. Blythe; Richard McMillan

    2002-02-04

    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.

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

    SciTech Connect (OSTI)

    Gary M. Blythe; Richard McMillan

    2002-03-04

    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.

  9. 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)

    2003-02-18

    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.

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

    DOE Patents [OSTI]

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

    1983-12-27

    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.

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

    DOE Patents [OSTI]

    Kowalczyk, Dennis C. (Pittsburgh, PA); Bricklemyer, Bruce A. (Avonmore, PA); Svoboda, Joseph J. (Pittsburgh, PA)

    1983-01-01

    Polymer precursor materials are vaporized without polymerization or are removed from a raw naphtha fraction by passing the raw naphtha to a vaporization zone (24) and vaporizing the naphtha in the presence of a wash oil while stripping with hot hydrogen to prevent polymer deposits in the equipment.

  12. More Economical Sulfur Removal for Fuel Processing Plants

    Office of Energy Efficiency and Renewable Energy (EERE)

    Case study covering TDA Research, Inc and its direct oxidation process, which is economical enough for smaller fuel processing plants to use.

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

    DOE Patents [OSTI]

    Carl, D.E.

    1997-10-21

    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.

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

    DOE Patents [OSTI]

    Carl, Daniel E. (Orchard Park, NY)

    1997-01-01

    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.

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

    SciTech Connect (OSTI)

    Gary M. Blythe; Richard McMillan

    2002-07-03

    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.

  16. Biologically-based signal processing system applied to noise removal for signal extraction

    DOE Patents [OSTI]

    Fu, Chi Yung; Petrich, Loren I.

    2004-07-13

    The method and system described herein use a biologically-based signal processing system for noise removal for signal extraction. A wavelet transform may be used in conjunction with a neural network to imitate a biological system. The neural network may be trained using ideal data derived from physical principles or noiseless signals to determine to remove noise from the signal.

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

    E-Print Network [OSTI]

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

    2004-01-01

    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 ...

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

    E-Print Network [OSTI]

    Abdel-wahab, Ahmed Ibraheem Ali

    2004-09-30

    Chloride is a deleterious ionic species in cooling water systems because it is important in promoting corrosion. Chloride can be removed from cooling water by precipitation as calcium chloroaluminate using ultra-high lime with aluminum process (UHLA...

  19. 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)

    2002-01-01

    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.

  20. 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.

    2012-08-22

    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.

  1. Selection of an acid-gas removal process for an LNG plant

    SciTech Connect (OSTI)

    Stone, J.B.; Jones, G.N.; Denton, R.D.

    1996-12-31

    Acid gas contaminants, such as, CO{sub 2}, H{sub 2}S and mercaptans, must be removed to a very low level from a feed natural gas before it is liquefied. CO{sub 2} is typically removed to a level of about 100 ppm to prevent freezing during LNG processing. Sulfur compounds are removed to levels required by the eventual consumer of the gas. Acid-gas removal processes can be broadly classified as: solvent-based, adsorption, cryogenic or physical separation. The advantages and disadvantages of these processes will be discussed along with design and operating considerations. This paper will also discuss the important considerations affecting the choice of the best acid-gas removal process for LNG plants. Some of these considerations are: the remoteness of the LNG plant from the resource; the cost of the feed gas and the economics of minimizing capital expenditures; the ultimate disposition of the acid gas; potential for energy integration; and the composition, including LPG and conditions of the feed gas. The example of the selection of the acid-gas removal process for an LNG plant.

  2. Processes to remove acid forming gases from exhaust gases

    DOE Patents [OSTI]

    Chang, S.G.

    1994-09-20

    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.

  3. OPERATIONS REVIEW OF THE SAVANNAH RIVER SITE INTEGRATED SALT DISPOSITION PROCESS - 11327

    SciTech Connect (OSTI)

    Peters, T.; Poirier, M.; Fondeur, F.; Fink, S.; Brown, S.; Geeting, M.

    2011-02-07

    The Savannah River Site (SRS) is removing liquid radioactive waste from its Tank Farm. To treat waste streams that are low in Cs-137, Sr-90, and actinides, SRS developed the Actinide Removal Process and implemented the Modular Caustic Side Solvent Extraction (CSSX) Unit (MCU). The Actinide Removal Process contacts salt solution with monosodium titanate to sorb strontium and select actinides. After monosodium titanate contact, the resulting slurry is filtered to remove the monosodium titanate (and sorbed strontium and actinides) and entrained sludge. The filtrate is transferred to the MCU for further treatment to remove cesium. The solid particulates removed by the filter are concentrated to {approx} 5 wt %, washed to reduce the sodium concentration, and transferred to the Defense Waste Processing Facility for vitrification. The CSSX process extracts the cesium from the radioactive waste using a customized solvent to produce a Decontaminated Salt Solution (DSS), and strips and concentrates the cesium from the solvent with dilute nitric acid. The DSS is incorporated in grout while the strip acid solution is transferred to the Defense Waste Processing Facility for vitrification. The facilities began radiological processing in April 2008 and started processing of the third campaign ('MarcoBatch 3') of waste in June 2010. Campaigns to date have processed {approx}1.2 million gallons of dissolved saltcake. Savannah River National Laboratory (SRNL) personnel performed tests using actual radioactive samples for each waste batch prior to processing. Testing included monosodium titanate sorption of strontium and actinides followed by CSSX batch contact tests to verify expected cesium mass transfer. This paper describes the tests conducted and compares results from facility operations. The results include strontium, plutonium, and cesium removal, cesium concentration, and organic entrainment and recovery data. Additionally, the poster describes lessons learned during operation of the facility.

  4. Catalytic process for removing toxic gases from gas streams

    SciTech Connect (OSTI)

    Baglio, J.A.; Gaudet, G.G.; Palilla, F.C.

    1983-02-22

    A multi-stage process for reducing the content of sulfurcontaining gases-notably hydrogen sulfide, sulfur dioxide, carbonyl sulfide and carbon disulfide-in waste gas streams is provided. In the first stage, the gas stream is passed through a reaction zone at a temperature between about 150 and 350/sup 0/C in the presence of a pretreated novel catalyst of the formula xLn/sub 2/O/sub 3/ in which Ln is yttrium or a rare earth element and T is cobalt, iron or nickel, and each of x and y is independently a number from 0 to 3, said catalyst being substantially non-crystalline and having a surface area of from about 10 m/sup 2//g to about 40 m/sup 2//g. The preferred catalyst is one in which Ln is lanthanum, T is cobalt, and x and y range from 1 to 3, including non-integers. The first stage yields a product stream having a reduced content of sulfur-containing gases, including specifically, substantial reduction of carbonyl sulfide and virtual elimination of carbon disulfide. An intermediate stage is a claus reaction, which may take place in one or more reaction zones, at temperatures less than about 130/sup 0/ C, in the presence of known catalysts such as bauxite, alumina or cobalt molybdates. The final stage is the air oxidation of hydrogen sulfide at a temperature between about 150 and 300/sup 0/ C in the presence of a catalyst usable in first stage.

  5. PF-4 actinide disposition strategy

    SciTech Connect (OSTI)

    Margevicius, Robert W [Los Alamos National Laboratory

    2010-05-28

    The dwindling amount of Security Category I processing and storage space across the DOE Complex has driven the need for more effective storage of nuclear materials at LANL's Plutonium Facility's (PF-4's) vault. An effort was begun in 2009 to create a strategy, a roadmap, to identify all accountable nuclear material and determine their disposition paths, the PF-4 Actinide Disposition Strategy (PADS). Approximately seventy bins of nuclear materials with similar characteristics - in terms of isotope, chemical form, impurities, disposition location, etc. - were established in a database. The ultimate disposition paths include the material to remain at LANL, disposition to other DOE sites, and disposition to waste. If all the actions described in the document were taken, over half of the containers currently in the PF-4 vault would been eliminated. The actual amount of projected vault space will depend on budget and competing mission requirements, however, clearly a significant portion of the current LANL inventory can be either dispositioned or consolidated.

  6. Microbial Transformations of Actinides and Other Radionuclides

    SciTech Connect (OSTI)

    Francis,A.J.; Dodge, C. J.

    2009-01-07

    Microorganisms can affect the stability and mobility of the actinides and other radionuclides released from nuclear fuel cycle and from nuclear fuel reprocessing plants. Under appropriate conditions, microorganisms can alter the chemical speciation, solubility and sorption properties and thus could increase or decrease the concentrations of radionuclides in solution in the environment and the bioavailability. Dissolution or immobilization of radionuclides is brought about by direct enzymatic action or indirect non-enzymatic action of microorganisms. Although the physical, chemical, and geochemical processes affecting dissolution, precipitation, and mobilization of radionuclides have been extensively investigated, we have only limited information on the effects of microbial processes and biochemical mechanisms which affect the stability and mobility of radionuclides. The mechanisms of microbial transformations of the major and minor actinides U, Pu, Cm, Am, Np, the fission products and other radionuclides such as Ra, Tc, I, Cs, Sr, under aerobic and anaerobic conditions in the presence of electron donors and acceptors are reviewed.

  7. 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.

    2013-07-10

    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.

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

    DOE Patents [OSTI]

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

    1987-08-21

    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.

  9. Water Removal in Bioretention Devices by Evapotranspiration Processes and Related Issues Affecting Performance

    E-Print Network [OSTI]

    Pitt, Robert E.

    Water Removal in Bioretention Devices by Evapotranspiration Processes and Related Issues Affecting Performance Robert Pitt,1 Shirley Clark,2 and John Voorhees3 1 Cudworth Professor of Urban Water Systems 17057 USA. PH: (717) 948-6127, FAX: (717) 948-6580, e-mail: seclark@psu.edu 3 Senior Water Resources

  10. Method for fluorination of actinide fluorides and oxyfluorides thereof using O.sub.2 F.sub.2

    DOE Patents [OSTI]

    Eller, Phillip G. (Los Alamos, NM); Malm, John G. (Naperville, IL); Penneman, Robert A. (Albuquerque, NM)

    1988-01-01

    Method for fluorination of actinides and fluorides and oxyfluorides thereof using O.sub.2 F.sub.2 which generates actinide hexafluorides, and for removal of actinides and compounds thereof from surfaces upon which they appear as unwanted deposits. The fluorinating agent, O.sub.2 F.sub.2, has been observed to readily 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 thereby not destroyed by high temperature reactions with the walls of the reaction vessel. Dioxygen difluoride is easily prepared, stored and transferred to the desired place of reaction.

  11. Method for fluorination of actinide fluorides and oxyfluorides thereof using O[sub 2]F[sub 2

    DOE Patents [OSTI]

    Eller, P.G.; Malm, J.G.; Penneman, R.A.

    1988-11-08

    Method is described for fluorination of actinides and fluorides and oxyfluorides thereof using O[sub 2]F[sub 2] which generates actinide hexafluorides, and for removal of actinides and compounds thereof from surfaces upon which they appear as unwanted deposits. The fluorinating agent, O[sub 2]F[sub 2], has been observed to readily 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 thereby not destroyed by high temperature reactions with the walls of the reaction vessel. Dioxygen difluoride is easily prepared, stored and transferred to the desired place of reaction.

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

    SciTech Connect (OSTI)

    Robert M. Counce; Jack S. Watson

    2009-06-30

    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.

  13. Defining manganese(II) removal processes in passive coal mine drainage treatment systems through laboratory incubation experiments

    E-Print Network [OSTI]

    Burgos, William

    Defining manganese(II) removal processes in passive coal mine drainage treatment systems through Building, University Park, PA 16802, USA b Smithsonian Institution, PO Box 37012, MRC 119, Washington, DC for the passive removal of Mn(II) from coal mine drainage (CMD). Aqueous Mn(II) is removed via oxidative

  14. Environmental research on actinide elements

    SciTech Connect (OSTI)

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

    1987-08-01

    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)

  15. Removal and recovery of metal ions from process and waste streams using polymer filtration

    SciTech Connect (OSTI)

    Jarvinen, G.D.; Smith, B.F.; Robison, T.W.; Kraus, K.M.; Thompson, J.A.

    1999-06-13

    Polymer Filtration (PF) is an innovative, selective metal removal technology. Chelating, water-soluble polymers are used to selectively bind the desired metal ions and ultrafiltration is used to concentrate the polymer-metal complex producing a permeate with low levels of the targeted metal ion. When applied to the treatment of industrial metal-bearing aqueous process streams, the permeate water can often be reused within the process and the metal ions reclaimed. This technology is applicable to many types of industrial aqueous streams with widely varying chemistries. Application of PF to aqueous streams from nuclear materials processing and electroplating operations will be described.

  16. Simultaneous SO2, SO3 and NOx removal by the EBA Process

    SciTech Connect (OSTI)

    Hirano, S.; Aoki, S.; Izutsu, M.; Yuki, Y.

    1999-07-01

    The system for electron beam flue gas treatment, the EBA (Electron Beam with Ammonia) Process is an innovative, dry, air pollution control technology that, by ammonia injection and electron beam irradiation, can efficiently remove sulfur oxides, (SOx, i.e., SO2 and SO3), from partially humidified flue gas. This is accompanied by simultaneous nitrogen oxides, (NOx), emission reduction in an amount that is determined, optionally, by site-specific system design. The process operation converts these pollutants as well as gasborne hydrogen chloride into a dry by-product solids consisting principally of ammonium sulfate, ammonium nitrate and ammonium chloride that can be used, worldwide, as a plant nutrient stock in large-scale agriculture. Commercial application of the Process has now been achieved, utilizing engineering data gained over many years of experience in operation of pilot plant facilities (including a process development unit that was field installed in the 1980s under sponsorship by US DOE at a coal-fired powerplant of Indianapolis Power and Light Company). With the co-operation of the Chinese government, EBARA Corporation has in 1997 completed the first commercial EBA Process installation, which is sited at the coal-fired Chengdu Power Station in China. This retrofit facility as a whole in addition to its performance in this high-sulfur bituminous coal service, including the targeted SOx removal efficiency, incidental NOx abatement and the usability of the by-product, were assessed and fully accepted contractually by a diversely structured, Examination Committee established by Chinese government authorities and assisted by an Expert Group. Criteria and considerations of the Chinese officials, in evaluating adequacy of the technology for further commercial application in China, all assessed positively, encompassed design, installation, operation, performance, reliability, environmental impact, cost effectiveness and by-product utilization/marketing. To illustrate attractive process economics for SOx removal in high-sulfur service with only incidental NOx removal, i.e. FGD (flue gas desulfurization) only, evaluation of cost-effectiveness of the Chengdu facility (at its moderate design SO2 removal efficiency) is presented.

  17. Method for extracting lanthanides and actinides from acid solutions

    DOE Patents [OSTI]

    Horwitz, E. Philip (Naperville, IL); Kalina, Dale G. (Naperville, IL); Kaplan, Louis (Lombard, IL); Mason, George W. (Clarendon Hills, IL)

    1985-01-01

    A process for the recovery of actinide and lanthanide values from aqueous acidic solutions with an organic extractant having the formula: ##STR1## where .phi. is phenyl, R.sup.1 is a straight or branched alkyl or alkoxyalkyl containing from 6 to 12 carbon atoms and R.sup.2 is an alkyl containing from 3 to 6 carbon atoms. 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.

  18. Advancing the scientific basis of trivalent actinide-lanthanide separations

    SciTech Connect (OSTI)

    Nash, K.L. [Washington State University, Department of Chemistry, PO Box 644630, Pullman, WA 99164-4630 (United States)

    2013-07-01

    For advanced fuel cycles designed to support transmutation of transplutonium actinides, several options have been demonstrated for process-scale aqueous separations for U, Np, Pu management and for partitioning of trivalent actinides and fission product lanthanides away from other fission products. The more difficult mutual separation of Am/Cm from La-Tb remains the subject of considerable fundamental and applied research. The chemical separations literature teaches that the most productive alternatives to pursue are those based on ligand donor atoms less electronegative than O, specifically N- and S-containing complexants and chloride ion (Cl{sup -}). These 'soft-donor' atoms have exhibited usable selectivity in their bonding interactions with trivalent actinides relative to lanthanides. In this report, selected features of soft donor reagent design, characterization and application development will be discussed. The roles of thiocyanate, aminopoly-carboxylic acids and lactate in separation processes are detailed. (authors)

  19. 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)

    1989-01-01

    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.

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

    DOE Patents [OSTI]

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

    1987-02-25

    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.

  1. Method of removing sulfur emissions from a fluidized-bed combustion process

    DOE Patents [OSTI]

    Vogel, Gerhard John (Elmhurst, IL); Jonke, Albert A. (Elmhurst, IL); Snyder, Robert B. (Naperville, IL)

    1978-01-01

    Alkali metal or alkaline earth metal oxides are impregnated within refractory support material such as alumina and introduced into a fluidized-bed process for the combustion of coal. Sulfur dioxide produced during combustion reacts with the metal oxide to form metal sulfates within the porous support material. The support material is removed from the process and the metal sulfate regenerated to metal oxide by chemical reduction. Suitable pore sizes are originally developed within the support material by heat-treating to accommodate both the sulfation and regeneration while still maintaining good particle strength.

  2. Recovery of the actinides by electrochemical methods in molten chlorides using solid aluminium cathode

    SciTech Connect (OSTI)

    Malmbeck, R.; Mendes, E.; Serp, J.; Soucek, P.; Glatz, J.P. [European Commission, JRC, Institute for Transuranium Elements, Postfach 2340, 76125 Karlsruhe (Germany); Cassayre, L. [Laboratoire de Genie Chimique - LGC, Universite Paul Sabatier, UMR 5503, 118 route de Narbonne, 31062 Toulouse Cedex 04 (France)

    2007-07-01

    An electrorefining process in molten chloride salts is being developed at ITU to reprocess the spent nuclear fuel. According to the thermochemical properties of the system, aluminium is the most promising electrode material for the separation of actinides (An) from lanthanides (Ln). The actinides are selectively reduced from the fission products and stabilized by the formation of solid and compact actinide-aluminium alloys with the reactive cathode material. In this work, the maximum loading of aluminium with actinides was investigated by potentiostatic and galvano-static electrorefining of U-Pu- Zr alloys. A very high aluminium capacity was achieved, as the average loading was 1.6 g of U and Pu into 1 g of aluminium and the maximum achieved loading was 2.3 g. For recovery of the actinides from aluminium, a process based on chlorination and a subsequent sublimation of AlCl{sub 3} is proposed. (authors)

  3. Thermal spray vitrification process for the removal of lead oxide contained in organic paints

    SciTech Connect (OSTI)

    Karthikeyan, J.; Chen, J.; Bancke, G.A.; Herman, H.; Berndt, C.C.; Breslin, V.T.

    1995-12-31

    The US Environmental Protection Agency (US-EPA) regulations have necessitated the removal and containment of toxic lead from lead oxide containing paints. The Thermal Spray Vitrification Process (TSVP) is a novel technique in which a glass powder of appropriate composition is flame sprayed onto the painted surface to achieve removal and vitrification of the lead. Two different glass systems, i.e., alkali silicate and ferrous silicate, were chosen for detailed study. Appropriate amounts of raw materials were mixed, fused, quenched, ground and sieved to obtain the spray quality powders. Grit blasted mild steel coupons were used as test substrates for the spray parameter optimization studies; while those coupons with lead oxide containing organic paint were used for the lead removal experiments. The powders and deposits were investigated using Microtrac particle size analysis (for powders), optical microscopy, XRD and SEM. The remnant lead in the panel was measured using a specially prepared X-Ray Fluorescence (XRF) system. The lead leach rate was recorded as per US-EPA approved Toxicity Characteristic Leaching Procedure (TCLP). The results of this study have shown that lead oxide can be successfully removed form the paint by flame spraying a maximum of three layers of glass onto the painted surface. It is possible to obtain much higher lead removal rate with ferrous silicate glass as compared to alkali silicate glass is much higher than the ferrous silicate glass. The in situ vitrification has not been completely optimized; however, the lead containing glass coating can be remelted in situ or on site to enhance the vitrification of the lead which had been absorbed in the glass coating.

  4. 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)

    1991-01-01

    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.

  5. ZIRCONIUM AND FISSION PRODUCT MANAGEMENT IN THE ALSEP PROCESS

    SciTech Connect (OSTI)

    Lumetta, Gregg J.; Carter, Jennifer C.; Niver, Cynthia M.; Gelis, Artem V.

    2013-09-29

    Solvent extraction systems that combine neutral donor extractants and acidic extractants are being investigated to provide a single process solvent for separating Am and Cm from acidic high-level liquid waste, including their separation from the trivalent lanthanides. This approach of combining extractants is collectively referred to as the Actinide-Lanthanide SEParation (ALSEP) process. Managing Zr and other fission products is one of the critical factors in developing the ALSEP process. In this work, a strategy has been developed in which Zr(IV) is extracted into the process solvent, then it is stripped from the solvent after the actinides have been selectively stripped. Molybdenum is strongly extracted into ALSEP solvents. Scrubbing the solvent with a citrate buffer before the actinide stripping step effectively removes Mo. Distribution ratios for Ru and Fe are low for extraction from HNO3, so these components can easily be routed to the high-level waste raffinate.

  6. 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)

    2012-07-01

    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)

  7. INERT-MATRIX FUEL: ACTINIDE ''BURINGIN'' AND DIRECT DISPOSAL

    SciTech Connect (OSTI)

    Rodney C. Ewing; Lumin Wang

    2002-10-30

    Excess actinides result from the dismantlement of nuclear weapons (Pu) and the reprocessing of commercial spent nuclear fuel (mainly 241 Am, 244 Cm and 237 Np). In Europe, Canada and Japan studies have determined much improved efficiencies for burnup of actinides using inert-matrix fuels. This innovative approach also considers the properties of the inert-matrix fuel as a nuclear waste form for direct disposal after one-cycle of burn-up. Direct disposal can considerably reduce cost, processing requirements, and radiation exposure to workers.

  8. Orbital effects in actinide systems

    SciTech Connect (OSTI)

    Lander, G.H.

    1983-01-01

    Actinide magnetism presents a number of important challenges; in particular, the proximity of 5f band to the Fermi energy gives rise to strong interaction with both d and s like conduction electrons, and the extended nature of the 5f electrons means that they can interact with electron orbitals from neighboring atoms. Theory has recently addressed these problems. Often neglected, however, is the overwhelming evidence for large orbital contributions to the magnetic properties of actinides. Some experimental evidence for these effects are presented briefly in this paper. They point, clearly incorrectly, to a very localized picture for the 5f electrons. This dichotomy only enhances the nature of the challenge.

  9. Actinide Thermodynamics at Elevated Temperatures

    SciTech Connect (OSTI)

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

    2007-11-16

    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

  10. Bench Scale Application of the Hybridized Zero Valent Iron Process for the Removal of Dissolved Silica From Water 

    E-Print Network [OSTI]

    Morar, Nilesh Mohan

    2014-11-12

    is effective. A more robust and cost-effective dissolved silica removal technique is desirable. The hybridized zero-valent iron (hZVI) process, now commercially available as Pironox™, uses zero-valent iron (Fe^0 ) as its main reactive media developed to remove...

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

    DOE Patents [OSTI]

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

    1996-01-01

    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.

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

    DOE Patents [OSTI]

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

    1985-01-01

    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.

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

    DOE Patents [OSTI]

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

    1982-09-03

    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.

  14. 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)

    2002-01-01

    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.

  15. 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)

    1986-01-01

    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.

  16. 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)

    1982-01-01

    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.

  17. 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)

    1995-01-01

    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.

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

    DOE Patents [OSTI]

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

    1992-11-17

    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.

  19. 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)

    1992-01-01

    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.

  20. 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)

    2008-07-15

    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)

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

    DOE Patents [OSTI]

    Comolli, Alfred G. (Yardley, PA)

    1991-01-01

    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. 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)

    2012-08-14

    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.

  3. TESTING OF A FULL-SCALE ROTARY MICROFILTER FOR THE ENHANCED PROCESS FOR RADIONUCLIDES REMOVAL

    SciTech Connect (OSTI)

    Herman, D; David Stefanko, D; Michael Poirier, M; Samuel Fink, S

    2009-01-01

    Savannah River National Laboratory (SRNL) researchers are investigating and developing a rotary microfilter for solid-liquid separation applications in the Department of Energy (DOE) complex. One application involves use in the Enhanced Processes for Radionuclide Removal (EPRR) at the Savannah River Site (SRS). To assess this application, the authors performed rotary filter testing with a full-scale, 25-disk unit manufactured by SpinTek Filtration with 0.5 micron filter media manufactured by Pall Corporation. The filter includes proprietary enhancements by SRNL. The most recent enhancement is replacement of the filter's main shaft seal with a John Crane Type 28LD gas-cooled seal. The feed material was SRS Tank 8F simulated sludge blended with monosodium titanate (MST). Testing examined total insoluble solids concentrations of 0.06 wt % (126 hours of testing) and 5 wt % (82 hours of testing). The following are conclusions from this testing.

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

    SciTech Connect (OSTI)

    Nash, Kenneth L.

    2008-11-20

    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.

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

    SciTech Connect (OSTI)

    Qiang, You; Paszczynski, Andrzej; Rao, Linfeng

    2011-10-30

    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.

  6. Aqueous recovery of actinides from aluminum alloys

    SciTech Connect (OSTI)

    Gray, J.H.; Chostner, D.F.; Gray, L.W.

    1989-01-01

    Early in the 1980's, a joint Rocky Flats/Savannah River program was established to recover actinides from scraps and residues generated during Rocky Flats purification operations. The initial program involved pyrochemical treatment of Molten Salt Extraction (MSE) chloride salts and Electrorefining (ER) anode heel metal to form aluminum alloys suitable for aqueous processing at Savannah River. Recently Rocky Flats has expressed interest in expanding the aluminum alloy program to include treatment of chloride salt residues from a modified Molten Salt Extraction process and from the Electrorefining purification operations. Samples of the current aluminum alloy buttons were prepared at Rocky Flats and sent to Savannah River Laboratory for flowsheet development and characterization of the alloys. A summary of the scrub alloy-anode heel alloy program will be presented along with recent results from aqueous dissolution studies of the new aluminum alloys. 2 figs., 4 tabs.

  7. 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)

    1983-01-01

    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.

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

    DOE Patents [OSTI]

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

    1983-09-20

    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.

  9. Joint Actinide Shock Physics Experimental Research | National...

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

    Experimental Research Joint Actinide Shock Physics Experimental Research The JASPER gas gun at the Nevada National Security Site is used to fire a projectile at a plutonium...

  10. 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)

    2000-01-01

    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.

  11. 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)

    1992-01-01

    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.

  12. 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

    2007-06-29

    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.

  13. Actinide partitioning-transmutation program final report. I. Overall assessment

    SciTech Connect (OSTI)

    Croff, A.G.; Blomeke, J.O.; Finney, B.C.

    1980-06-01

    This report is concerned with an overall assessment of the feasibility of and incentives for partitioning (recovering) long-lived nuclides from fuel reprocessing and fuel refabrication plant radioactive wastes and transmuting them to shorter-lived or stable nuclides by neutron irradiation. The principal class of nuclides considered is the actinides, although a brief analysis is given of the partitioning and transmutation (P-T) of /sup 99/Tc and /sup 129/I. The results obtained in this program permit us to make a comparison of the impacts of waste management with and without actinide recovery and transmutation. Three major conclusions concerning technical feasibility can be drawn from the assessment: (1) actinide P-T is feasible, subject to the acceptability of fuels containing recycle actinides; (2) technetium P-T is feasible if satisfactory partitioning processes can be developed and satisfactory fuels identified (no studies have been made in this area); and (3) iodine P-T is marginally feasible at best because of the low transmutation rates, the high volatility, and the corrosiveness of iodine and iodine compounds. It was concluded on the basis of a very conservative repository risk analysis that there are no safety or cost incentives for actinide P-T. In fact, if nonradiological risks are included, the short-term risks of P-T exceed the long-term benefits integrated over a period of 1 million years. Incentives for technetium and iodine P-T exist only if extremely conservative long-term risk analyses are used. Further RD and D in support of P-T is not warranted.

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

    SciTech Connect (OSTI)

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

    2012-09-19

    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.

  15. Separation of actinides from lanthanides

    DOE Patents [OSTI]

    Smith, B.F.; Jarvinen, G.D.; Ryan, R.R.

    1988-03-31

    An organic extracting solution and an extraction method useful for separating elements of the actinide series of the periodic table from elements of the lanthanide series, where both are in trivalent form is described. The extracting solution consists of a primary ligand and a secondary ligand, preferably in an organic solvent. The primary ligand is a substituted monothio-1,3-dicarbonyl, which includes a substituted 4-acyl-2-pyrazolin-5-thione, such as 4-benzoyl-2,4- dihydro-5-methyl-2-phenyl-3H-pyrazol-3-thione (BMPPT). The secondary ligand is a substituted phosphine oxide, such as trioctylphosphine oxide (TOPO).

  16. Application of extraction chromatography to actinide decontamination of hydrochloric acid effluent streams

    SciTech Connect (OSTI)

    Schulte, L.D.; McKee, S.D.; Salazar, R.R.

    1996-05-01

    Extraction chromatography is under development as a method to lower actinide activity levels in effluent steams. Successful application of this technique for radioactive liquid waste treatment would provide a low activity feed stream for HCl recycle, reduce the loss of radioactivity to the environment in aqueous effluents, and would lower the quantity and reduce the hazard of the associated solid waste. The extraction of Pu and Am from HCl solutions was examined for several commercial and laboratory-produced sorbed resin materials. Inert supports included silica and polymer beads of differing mesh sizes. The support material was coated with either n-octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (O-CMPO) or di-(4-t-butylphenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (D-CMPO) as an extractant, and using either tributyl phosphate (TBP) or diamyl amylphosphonate (DAAP) as a diluent. Solutions tested were effluent streams generated by ion exchange and solvent extraction recovery of Pu. A finer mesh silica support material demonstrated advantages in removal of trivalent Am in some tests, but also showed a tendency toward plugging and channeling as column sizes and flow rates were increased. Larger bead sizes showed better physical properties as the process was scaled up to removal of gram quantities of Am from large effluent volumes. The ratio of extractant to diluent also appeared to play a role in the retention of Am. In direct comparative studies, when loaded on identical supports and diluent conditions, D-CMPO demonstrated better Am retention than O-CMPO from HCl process effluents.

  17. 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

    2009-10-05

    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.

  18. Importance of wet precipitation as a removal and transport process for atmospheric water soluble carbonyls

    E-Print Network [OSTI]

    Matsunaga, SN; Guenther, AB; Izawa, Y; Wiedinmyer, C; Greenberg, JP; Kawamura, K

    2007-01-01

    et al. , 2001). Wet precipitation samples contain almost alldeposition rate due to the precipitation Yield mg C m 2 h 1Importance of wet precipitation as a removal and transport

  19. 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.

    2007-01-25

    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.

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

    SciTech Connect (OSTI)

    Lynn, S.; Cairns, E.J.

    1992-01-01

    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.

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

    SciTech Connect (OSTI)

    Lynn, S.; Cairns, E.J.

    1992-11-01

    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.

  2. Complexation of lanthanides and actinides by acetohydroxamic acid

    SciTech Connect (OSTI)

    Taylor, R.J.; Sinkov, S.I.; Choppin, G.R.

    2008-07-01

    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)

  3. 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)

    2013-07-01

    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)

  4. Experimental studies of actinides in molten salts

    SciTech Connect (OSTI)

    Reavis, J.G.

    1985-06-01

    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.

  5. Tank 37H Salt Removal Batch Process and Salt Dissolution Mixing Study

    SciTech Connect (OSTI)

    Kwon, K.C.

    2001-09-18

    Tank 30H is the receipt tank for concentrate from the 3H Evaporator. Tank 30H has had problems, such as cooling coil failure, which limit its ability to receive concentrate from the 3H Evaporator. SRS High Level Waste wishes to use Tank 37H as the receipt tank for the 3H Evaporator concentrate. Prior to using Tank 37H as the 3H Evaporator concentrate receipt tank, HLW must remove 50 inches of salt cake from the tank. They requested SRTC to evaluate various salt removal methods for Tank 37H. These methods include slurry pumps, Flygt mixers, the modified density gradient method, and molecular diffusion.

  6. BWR Assembly Optimization for Minor Actinide Recycling

    SciTech Connect (OSTI)

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

    2010-03-22

    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).

  7. Joint Actinide Shock Physics Experimental Research - JASPER

    SciTech Connect (OSTI)

    2014-10-31

    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.

  8. Joint Actinide Shock Physics Experimental Research - JASPER

    ScienceCinema (OSTI)

    None

    2015-01-09

    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.

  9. Actinide minimization using pressurized water reactors

    E-Print Network [OSTI]

    Visosky, Mark Michael

    2006-01-01

    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 ...

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

    DOE Patents [OSTI]

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

    1985-11-21

    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.

  11. Actinide behavior in the Integral Fast Reactor. Final project report

    SciTech Connect (OSTI)

    Courtney, J.C.

    1994-11-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

  12. 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)

    2004-05-18

    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.

  13. Rapid determination of actinides in seawater samples

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

    Maxwell, Sherrod L.; Culligan, Brian K.; Hutchison, Jay B.; Utsey, Robin C.; McAlister, Daniel R.

    2014-03-09

    A new rapid method for the determination of actinides in seawater samples has been developed at the Savannah River National Laboratory. The actinides can be measured by alpha spectrometry or inductively-coupled plasma mass spectrometry. The new method employs novel pre-concentration steps to collect the actinide isotopes quickly from 80 L or more of seawater. Actinides are co-precipitated using an iron hydroxide co-precipitation step enhanced with Ti+3 reductant, followed by lanthanum fluoride co-precipitation. Stacked TEVA Resin and TRU Resin cartridges are used to rapidly separate Pu, U, and Np isotopes from seawater samples. TEVA Resin and DGA Resin were used tomore »separate and measure Pu, Am and Cm isotopes in seawater volumes up to 80 L. This robust method is ideal for emergency seawater samples following a radiological incident. It can also be used, however, for the routine analysis of seawater samples for oceanographic studies to enhance efficiency and productivity. In contrast, many current methods to determine actinides in seawater can take 1–2 weeks and provide chemical yields of ~30–60 %. This new sample preparation method can be performed in 4–8 h with tracer yields of ~85–95 %. By employing a rapid, robust sample preparation method with high chemical yields, less seawater is needed to achieve lower or comparable detection limits for actinide isotopes with less time and effort.« less

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

    SciTech Connect (OSTI)

    Gokhan Alptekin

    2008-09-30

    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.

  15. Method for extracting lanthanides and actinides from acid solutions by modification of Purex solvent

    DOE Patents [OSTI]

    Horwitz, E.P.; Kalina, D.G.

    1984-05-21

    A process has been developed for the extraction of multivalent lanthanide and actinide values from acidic waste solutions, and for the separation of these values from fission product and other values, which utilizes a new series of neutral bi-functional extractants, the alkyl(phenyl)-N, N-dialkylcarbamoylmethylphosphine oxides, in combination with a phase modifier to form an extraction solution. The addition of the extractant to the Purex process extractant, tri-n-butylphosphate in normal paraffin hydrocarbon diluent, will permit the extraction of multivalent lanthanide and actinide values from 0.1 to 12.0 molar acid solutions.

  16. 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.

    2010-02-24

    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.

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

    DOE Patents [OSTI]

    Miller, S.M.

    1983-10-31

    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.

  18. A model of material removal and post process surface topography for copper CMP

    E-Print Network [OSTI]

    Choi, Seungchoun; Doyle, Fiona M.; Dornfeld, David

    2011-01-01

    post process surface topography for copper CMP Seungchounis due to the uneven surface topography generated during theand the post CMP topography. Two synergistic mechanisms were

  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.

    2007-05-30

    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. Take care in picking membranes to combine with other processes for CO/sub 2/ removal

    SciTech Connect (OSTI)

    Schendel, R.; Seymour, J.

    1985-02-18

    Membranes are somewhat unique in gas processing. Membrane processing is different enough from conventional processing that each project must be examined on its own merits. Membranes should be considered based on their characteristics in combination with other technologies as well as by themselves, and used where they show advantage. Membranes may find use in unusual applications and may not be best in traditional applications. Membrane systems are easily simulated and evaluation of membrane applications can be made quickly. Using a mainframe computer, response is essentially instantaneous.

  1. Low Quality Natural Gas Sulfur Removal and Recovery CNG Claus Sulfur Recovery Process

    SciTech Connect (OSTI)

    Klint, V.W.; Dale, P.R.; Stephenson, C.

    1997-10-01

    Increased use of natural gas (methane) in the domestic energy market will force the development of large non-producing gas reserves now considered to be low quality. Large reserves of low quality natural gas (LQNG) contaminated with hydrogen sulfide (H{sub 2}S), carbon dioxide (CO{sub 2}) and nitrogen (N) are available but not suitable for treatment using current conventional gas treating methods due to economic and environmental constraints. A group of three technologies have been integrated to allow for processing of these LQNG reserves; the Controlled Freeze Zone (CFZ) process for hydrocarbon / acid gas separation; the Triple Point Crystallizer (TPC) process for H{sub 2}S / C0{sub 2} separation and the CNG Claus process for recovery of elemental sulfur from H{sub 2}S. The combined CFZ/TPC/CNG Claus group of processes is one program aimed at developing an alternative gas treating technology which is both economically and environmentally suitable for developing these low quality natural gas reserves. The CFZ/TPC/CNG Claus process is capable of treating low quality natural gas containing >10% C0{sub 2} and measurable levels of H{sub 2}S and N{sub 2} to pipeline specifications. The integrated CFZ / CNG Claus Process or the stand-alone CNG Claus Process has a number of attractive features for treating LQNG. The processes are capable of treating raw gas with a variety of trace contaminant components. The processes can also accommodate large changes in raw gas composition and flow rates. The combined processes are capable of achieving virtually undetectable levels of H{sub 2}S and significantly less than 2% CO in the product methane. The separation processes operate at pressure and deliver a high pressure (ca. 100 psia) acid gas (H{sub 2}S) stream for processing in the CNG Claus unit. This allows for substantial reductions in plant vessel size as compared to conventional Claus / Tail gas treating technologies. A close integration of the components of the CNG Claus process also allow for use of the methane/H{sub 2}S separation unit as a Claus tail gas treating unit by recycling the CNG Claus tail gas stream. This allows for virtually 100 percent sulfur recovery efficiency (virtually zero SO{sub 2} emissions) by recycling the sulfur laden tail gas to extinction. The use of the tail gas recycle scheme also deemphasizes the conventional requirement in Claus units to have high unit conversion efficiency and thereby make the operation much less affected by process upsets and feed gas composition changes. The development of these technologies has been ongoing for many years and both the CFZ and the TPC processes have been demonstrated at large pilot plant scales. On the other hand, prior to this project, the CNG Claus process had not been proven at any scale. Therefore, the primary objective of this portion of the program was to design, build and operate a pilot scale CNG Claus unit and demonstrate the required fundamental reaction chemistry and also demonstrate the viability of a reasonably sized working unit.

  2. 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

    2013-01-01

    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 ...

  3. 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-

    2004-01-01

    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 ...

  4. 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

    2011-01-01

    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. ...

  5. Operation of a bushing melter system designed for actinide vitrification

    SciTech Connect (OSTI)

    Ramsey, W.G.

    1996-03-01

    The Westinghouse Savannah River Company is developing a melter system to vitrify actinide materials. The melter system will used to vitrify the americium and curium solution which is currently stored in one of the Savannah River Site`s (SRS) processing canyons. This solution is one of the materials designated by the Defense Nuclear Facilities Safety Board (DNFSB) to be dispositioned as part of the DNFSB recommendation 94-1. The Am/Cm solution contains an extremely large fraction (>2 kilograms of Cm and 10 kilograms of Am) of t he United States`s total inventory of both elements. They have an estimated value on the order of one billion dollars - if they are processed through the DOE Isotope Sales program at the Oak Ridge National Laboratory. It is therefore deemed highly desirable to transfer the material to Oak Ridge in a form which can allow for recovery of the material. A commercial glass composition has been demonstrated to be compatible with up to 40 weight percent of the Am/Cm solution contents. This glass is also selectively attacked by nitric acid. This allows the actinide to be recovered by common separation processes.

  6. Synthesis of Functionalized Superparamagnetic Iron Oxide Nanoparticles from a Common Precursor and their Application as Heavy Metal and Actinide Sorbents

    SciTech Connect (OSTI)

    Warner, Marvin G.; Warner, Cynthia L.; Addleman, Raymond S.; Droubay, Timothy C.; Engelhard, Mark H.; Davidson, Joseph D.; Cinson, Anthony D.; Nash, Michael A.; Yantasee, Wassana

    2009-10-12

    We describe the use of a simple and versatile technique to generate a series of ligand stabilized iron oxide nanoparticles containing different ? functionalities with specificities toward heavy metals and actinides at the periphery of the stabilizing ligand shell from a common, easy to synthesize precursor nanoparticle. The resulting nanoparticles are designed to contain affinity ligands that make them excellent sorbent materials for a variety of heavy metals from contaminated aqueous systems such as river water and ground water as well as actinides from clinical samples such as blood and urine. Functionalized superparamagnetic nanoparticles make ideal reagents for extraction of heavy metal and actinide contaminants from environmental and clinical samples since they are easily removed from the media once bound to the contaminant by simply applying a magnetic field. In addition, these engineered nanomaterials have an inherently high active surface area (often > 100 m2/g) making them ideal sorbent materials for these types of applications

  7. FIELD DEMONSTRATION OF A MEMBRANE PROCESS TO RECOVER HEAVY HYDROCARBONS AND TO REMOVE WATER FROM NATURAL GAS

    SciTech Connect (OSTI)

    R. Baker; T. Hofmann; J. Kaschemekat; K.A. Lokhandwala; Membrane Group; Module Group; Systems Group

    2001-01-11

    The objective of this project is to design, construct and field demonstrate a 3-MMscfd 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 conditions is required to convince industry users of the efficiency and reliability of the process. The system will be designed and fabricated by Membrane Technology and Research, Inc. (MTR) and then installed and operated at British Petroleum (BP)-Amoco's Pascagoula, MS plant. The Gas Research Institute will partially support the field demonstration and BP-Amoco will help install the unit and provide onsite operators and utilities. The gas processed by the membrane system will meet pipeline specifications for dewpoint 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. At the end of this demonstration project the process will be ready for commercialization. The route to commercialization will be developed during this project and may involve collaboration with other companies already servicing the natural gas processing industry.

  8. 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)

    1994-05-02

    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.

  9. 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.

    2014-01-12

    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 organicsmore »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.« less

  10. Actinide Targets for Neutron Cross Section Measurements

    SciTech Connect (OSTI)

    John D. Baker; Christopher A. McGrath

    2006-10-01

    The Advanced Fuel Cycle Initiative (AFCI) and the Generation IV Reactor Initiative have demonstrated a lack of detailed neutron cross-sections for certain "minor" actinides, those other than the most common (235U, 238U, and 239Pu). For some closed-fuel-cycle reactor designs more than 50% of reactivity will, at some point, be derived from "minor" actinides that currently have poorly known or in some cases not measured (n,?) and (n,f) cross sections. A program of measurements under AFCI has begun to correct this. One of the initial hurdles has been to produce well-characterized, highly isotopically enriched, and chemically pure actinide targets on thin backings. Using a combination of resurrected techniques and new developments, we have made a series of targets including highly enriched 239Pu, 240Pu, and 242Pu. Thus far, we have electrodeposited these actinide targets. In the future, we plan to study reductive distillation to achieve homogeneous, adherent targets on thin metal foils and polymer backings. As we move forward, separated isotopes become scarcer, and safety concerns become greater. The chemical purification and electodeposition techniques will be described.

  11. Rapid determination of actinides in asphalt samples

    SciTech Connect (OSTI)

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

    2014-01-12

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

    DOE Patents [OSTI]

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

    2009-04-14

    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.

  13. The effect of actinides on the microstructural development in a metallic high-level nuclear waste form

    SciTech Connect (OSTI)

    Keiser, D. D., Jr.; Sinkler, W.; Abraham, D. P.; Richardson, J. W., Jr.; McDeavitt, S. M.

    1999-10-25

    Waste forms to contain material residual from an electrometallurgical treatment of spent nuclear fuel have been developed by Argonne National Laboratory. One of these waste forms contains waste stainless steel (SS), fission products that are noble to the process (e.g., Tc, Ru, Pd, Rh), Zr, and actinides. The baseline composition of this metallic waste form is SS-15wt.% Zr. The metallurgy of this baseline alloy has been well characterized. On the other hand, the effects of actinides on the alloy microstructure are not well understood. As a result, SS-Zr alloys with added U, Pu, and/or Np have been cast and then characterized, using scanning electron microscopy, transmission electron microscopy, and neutron diffraction, to investigate the microstructural development in SS-Zr alloys that contain actinides. Actinides were found to congregate non-uniformally in a Zr(Fe,Cr,Ni){sub 2+x} phase. Apparently, the actinides were contained in varying amounts in the different polytypes (C14, C15, and C36) of the Zr(Fe,Cr,Ni){sub 2+x} phase. Heat treatment of an actinide-containing SS-15 wt.% Zr alloy showed the observed microstructure to be stable.

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

    DOE Patents [OSTI]

    McDowell, William J. (Knoxville, TN)

    1980-01-01

    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.

  15. FIELD DEMONSTRATION OF A MEMBRANE PROCESS TO RECOVER HEAVY HYDROCARBONS AND TO REMOVE WATER FROM NATURAL GAS

    SciTech Connect (OSTI)

    Unknown

    2002-04-10

    The objective of this project is to design, construct and field demonstrate a 3-MMscfd membrane system to recover natural gas liquids (NGL) and remove water from raw natural gas. The gas processed by the membrane system will meet pipeline specifications for dew point and Btu value, and the process is likely to be significantly less expensive than glycol dehydration followed by propane refrigeration, the principal competitive technology. The BP-Amoco gas processing plant in Pascagoula, MS was finalized as the location for the field demonstration. Detailed drawings of the MTR membrane skid (already constructed) were submitted to the plant in February, 2000. However, problems in reaching an agreement on the specifications of the system compressor delayed the project significantly, so MTR requested (and was subsequently granted) a no-cost extension to the project. Following resolution of the compressor issues, the goal is to order the compressor during the first quarter of 2002, and to start field tests in mid-2002. Information from potential users of the membrane separation process in the natural gas processing industry suggests that applications such as fuel gas conditioning and wellhead gas processing are the most promising initial targets. Therefore, most of our commercialization effort is focused on promoting these applications. Requests for stream evaluations and for design and price quotations have been received through MTR's web site, from direct contact with potential users, and through announcements in industry publications. To date, about 90 commercial quotes have been supplied, and orders totaling about $1.13 million for equipment or rental of membrane units have been received.

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

    DOE Patents [OSTI]

    Rosenberg, H.S.

    1987-02-03

    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.

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

    SciTech Connect (OSTI)

    Neumann, D.W.; Lynn, S.

    1986-07-01

    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.

  18. 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.

    1989-04-28

    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.

  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.

    1982-03-31

    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. 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)

    1988-01-01

    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.

  1. The role of actinide burning and the Integral Fast Reactor in the future of nuclear power

    SciTech Connect (OSTI)

    Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.

    1990-12-01

    A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.

  2. 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

    2012-07-05

    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.

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

    E-Print Network [OSTI]

    Dewan, Leslie

    2013-01-01

    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 ...

  4. Minor actinides transmutation in SFR depleted uranium radial blanket, neutronic AND THERMAL HYDRAULIC EVALUATION

    SciTech Connect (OSTI)

    Buiron, L.; Varaine, F. [CEA/DER/SPRC/LEDC Commissariat a l'Energie Atomique (CEA), Cadarache Centre, 13108 Saint-Paul-lez-Durance Cedex (France); Lorenzo, D.; Palancher, H.; Valentin, B. [CEA/DEC/SESC/LC2I Commissariat a l'Energie Atomique (CEA), Cadarache Centre, 13108 Saint-Paul-lez-Durance Cedex (France)

    2007-07-01

    In the framework of next generation fast reactor design, the management of minor actinides (MA) is one of the fundamental issues. This can be made by either homogeneous or heterogeneous multi-recycling model. In the homogeneous process the minor actinides are diluted in the fuel assembly. In the so-called heterogeneous model, minor actinides are concentrated apart from the core fuel (in special pins within dedicated core fuel assemblies or in axial or radial blankets). Here, we proposed to see the transmutation performances of radial blankets loaded with a mixture of depleted uranium and minor actinides oxide. This particular heterogeneous multi-recycling model allows the loading a significantly higher mass of minor actinides in the core than the homogeneous multi-recycling model. The oxide matrix also allows to reprocess such S/A in the spent fuel standard flow. Starting from a preliminary design of a 3600 MW Sodium Fast Reactor (SFR) in progress at CEA, we investigated the transmutation performances of (U+MA)O{sub 2} fuel in radial blankets assemblies. Among all possibilities, we focused on two scenarios: a realistic case with MA enrichment close to 10% and a more optimistic one, near term technologically achievable, close to 40%. For an equilibrium core, the MA transmutation rate reaches 40% for total fuel life time around 11 years for both enrichments. For this particular heterogeneous model, the minor actinides equilibrium (production=destruction) can be achieved with only 23% of the SFR fleet using such 40% MA radial blankets. It represents a total fabrication of 50 of such S/A per year. Concerning non-proliferation issue, the discharged plutonium of these assemblies is highly degraded (contribution of {sup 238}Pu and {sup 240}Pu around 60%). From this starting point, a coupled neutronic-thermal hydraulic optimization based on a simple iterative process has been carried out to deal with minor actinides specific features: high specific decay heat, swelling and helium production. In this paper, we review the main characteristics of the optimized system complying with GEN IV-like images of sodium fast reactors. (authors)

  5. Electrochemical Processes for Removing

    E-Print Network [OSTI]

    Fay, Noah

    in boilers for electric power generation and evaporative cooling need to have potential scale of ion exchange media produces large quantities of brine. For example, a single regeneration cycle can

  6. Silica Scaling Removal Process

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

    millions of dollars each year in facility down time, maintenance, discharge water treatments (blowdown), and anti-scaling chemical additives. This is particularly true for...

  7. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power AdministrationRobust,Field-effect Photovoltaics -7541 UnlimitedShift EndShutdown 2013SignalSignal

  8. ROBUSTNESS OF THE CSSX PROCESS TO FEED VARIATION: EFFICIENT CESIUM REMOVAL FROM THE HIGH POTASSIUM WASTES AT HANFORD

    SciTech Connect (OSTI)

    Delmau, Laetitia Helene; Birdwell Jr, Joseph F; McFarlane, Joanna; Moyer, Bruce A

    2010-01-01

    This contribution finds the Caustic-Side Solvent Extraction (CSSX) process to be effective for the removal of cesium from the Hanford tank-waste supernatant solutions. The Hanford waste types are more challenging than those at the Savannah River Site (SRS) in that they contain significantly higher levels of potassium, the chief competing ion in the extraction of cesium. By use of a computerized CSSX thermodynamic model, it was calculated that the higher levels of potassium depress the cesium distribution ratio (D{sub Cs}), as validated to within {+-}11% by the measurement of D{sub Cs} values on various Hanford waste-simulant compositions. A simple analog model equation that can be readily applied in a spreadsheet for estimating the D{sub Cs} values for the varying waste compositions was developed and shown to yield nearly identical estimates as the computerized CSSX model. It is concluded from the batch distribution experiments, the physical-property measurements, the equilibrium modeling, the flowsheet calculations, and the contactor sizing that the CSSX process as currently formulated for cesium removal from alkaline salt waste at the SRS is capable of treating similar Hanford tank feeds, albeit with more stages. For the most challenging Hanford waste composition tested, 31 stages would be required to provide a cesium decontamination factor (DF) of 5000 and a concentration factor (CF) of 2. Commercial contacting equipment with rotor diameters of 10 in. for extraction and 5 in. for stripping should have the capacity to meet throughput requirements, but testing will be required to confirm that the needed efficiency and hydraulic performance are actually obtainable. Markedly improved flowsheet performance was calculated based on experimental distribution ratios determined for an improved solvent formulation employing the more soluble cesium extractant BEHBCalixC6 used with alternative scrub and strip solutions, respectively 0.1 M NaOH and 0.010 M boric acid. The improved solvent and flowsheet can meet minimum requirements (DF = 5000 and CF = 2) with 15 stages or more ambitious goals (DF = 40,000 and CF = 15) with 19 stages. Thus, a modular CSSX application for the Hanford waste seems readily obtainable with further short-term development.

  9. Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation

    E-Print Network [OSTI]

    Heidet, Florent

    2010-01-01

    Albright, D. , Plutonium and highly enriched uranium, 1996 :and swelling in uranium-plutonium mixed nitride fuels.products and to extract plutonium or any other actinide from

  10. 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

    2012-08-10

    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.

  11. Process for the selective removal of hydrogen sulphide and carbonyl sulfide from light hydrocarbon gases containing carbon dioxide

    SciTech Connect (OSTI)

    Bush, W.V.

    1988-06-07

    A process for the selective removal of H/sub 2/S and COS from a gas containing light hydrocarbons, H/sub 2/S, COS and CO/sub 2/, is described which comprises in a one step absorption, at treatment conditions, contacting the gas stream with a solvent stream consisting essentially of: (i) water, (ii) a bridgehead amine comprising a bicyclo tertiary amine or a bicyclo amidine to selectively hydrolyze the COS to H/sub 2/S and CO/sub 2/, (iii) a tertiary amine to selectively absorb the H/sub 2/S and to selectively exclude from absorption the CO/sub 2/ in the gas stream and the CO/sub 2/ produced by the hydrolysis of the COS, and (iv) a physical solvent acceptable for COS absorption wherein two streams are formed comprising: (1) a light hydrocarbon and CO/sub 2/-containing stream having 1 ppm to about 200 ppm H/sub 2/S and having 1 ppm COS to about 10 ppm COS and (2) a solvent stream rich in H/sub 2/S, water, tertiary amine and the bridgehead amine.

  12. Supercritical Carbon Dioxide Ligands for Extracting Actinide Metal Ions from Porous Solids

    SciTech Connect (OSTI)

    Albert W. Herlinger; Dr. Mark L. Dietz

    2003-03-06

    Numerous types of actinide-bearing waste materials are found throughout the DOE complex. Most of these wastes consist of large volumes of non-hazardous materials contaminated with relatively small quantities of actinide elements. Separation of these wastes into their inert and radioactive components would dramatically reduce the costs of stabilization and disposal. For example, the DOE is responsible for decontaminating concrete within 7000 surplus contaminated buildings. The best technology now available for removing surface contamination from concrete involves removing the surface layer by grit blasting, which produces a large volume of blasting residue containing a small amount of radioactive material. Disposal of this residue is expensive because of its large volume and fine particulate nature. Considerable cost savings would result from separation of the radioactive constituents and stabilization of the concrete dust. Similarly, gas diffusion plants for uranium enrichment contain valuable high-purity nickel in the form of diffusion barriers. Decontamination is complicated by the extremely fine pores in these barriers, which are not readily accessible by most cleaning techniques. A cost-effective method for the removal of radioactive contaminants would release this valuable material for salvage.

  13. Proceedings of the NSF Workshop on Research Needs in Thermal Aspects of Material Removal Processes, edited Ranga Komanduri, Oklahoma State University, Stillwater, OK, June 10-12, 2003

    E-Print Network [OSTI]

    Yao, Y. Lawrence

    Proceedings of the NSF Workshop on Research Needs in Thermal Aspects of Material Removal Processes laser energy has enough time to be transferred to lattice, electron and lattice can reach thermal, edited Ranga Komanduri, Oklahoma State University, Stillwater, OK, June 10-12, 2003 247 Thermal Aspects

  14. REMOVAL PROCESSES OF VOLCANIC ASH PARTICLES FROM THE ATMOSPHERE Gregg J.S. Bluth and William I. Rose, Michigan Technological University

    E-Print Network [OSTI]

    Bluth, Gregg

    REMOVAL PROCESSES OF VOLCANIC ASH PARTICLES FROM THE ATMOSPHERE Gregg J.S. Bluth and William I information for mapping ash hazards, as well as the means to study and predict the fates of volcanic clouds provide information on how the size, size distribution, and total mass of fine ash particles evolve

  15. POTENTIAL BENCHMARKS FOR ACTINIDE PRODUCTION IN HANFORD REACTORS

    SciTech Connect (OSTI)

    PUIGH RJ; TOFFER H

    2011-10-19

    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.

  16. Method for extracting lanthanides and actinides from acid solutions by modification of purex solvent

    DOE Patents [OSTI]

    Horwitz, E. Philip (Naperville, IL); Kalina, Dale G. (Naperville, IL)

    1986-01-01

    A process for the recovery of actinide and lanthanide values from aqueous solutions with an extraction solution containing an organic extractant having the formula: ##STR1## where .phi. is phenyl, R.sup.1 is a straight or branched alkyl or alkoxyalkyl containing from 6 to 12 carbon atoms and R.sup.2 is an alkyl containing from 3 to 6 carbon atoms and phase modifiers in a water-immiscible hydrocarbon diluent. The addition of the extractant to the Purex process extractant, tri-n-butylphosphate in normal paraffin hydrocarbon diluent, will permit the extraction of multivalent lanthanide and actinide values from 0.1 to 12.0 molar acid solutions.

  17. Method for extracting lanthanides and actinides from acid solutions by modification of Purex solvent

    DOE Patents [OSTI]

    Horwitz, E.P.; Kalina, D.G.

    1986-03-04

    A process is described for the recovery of actinide and lanthanide values from aqueous solutions with an extraction solution containing an organic extractant having the formula as shown in a diagram where [phi] is phenyl, R[sup 1] is a straight or branched alkyl or alkoxyalkyl containing from 6 to 12 carbon atoms and R[sup 2] is an alkyl containing from 3 to 6 carbon atoms and phase modifiers in a water-immiscible hydrocarbon diluent. The addition of the extractant to the Purex process extractant, tri-n-butylphosphate in normal paraffin hydrocarbon diluent, will permit the extraction of multivalent lanthanide and actinide values from 0.1 to 12.0 molar acid solutions. 6 figs.

  18. Covalent Bonding in Actinide Sandwich Molecules

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would like submit theCovalent Bonding in Actinide Sandwich Molecules Print Glenn Seaborg was one

  19. Covalent Bonding in Actinide Sandwich Molecules

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would like submit theCovalent Bonding in Actinide Sandwich Molecules Print Glenn Seaborg was

  20. Actinide-specific complexing agents: their structural and solution chemistry

    SciTech Connect (OSTI)

    Raymond, K.N.; Freeman, G.E.; Kappel, M.J.

    1983-07-01

    The synthesis of a series of tetracatecholate ligands designed to be specific for Pu(IV) and other actinide(IV) ions has been achieved. Although these compounds are very effective as in vivo plutonium removal agents, potentiometric and voltammetric data indicate that at neutral pH full complexation of the Pu(IV) ion by all four catecholate groups does not occur. Spectroscopic results indicate that the tetracatecholates, 3,4,3-LICAMS and 3,4,3-LICAMC, complex Am(III). The Am(IV)/(III)-catecholate couple (where catecholate = 3,4,3-LICAMS or 3,4,3-LICAMC) is not observed, but may not be observable due to the large currents associated with ligand oxidation. However, within the potential range where ligand oxidation does not occur, these experiments indicate that the reduction potential of free Am(IV)/(III) is probably greater than or equal to + 2.6 V vs NHE or higher. Proof of the complexation of americium in the trivalent oxidation state by 3,4,3-LICAMS and 3,4,3-LICAMC elimates the possibility of tetracatholates stabilizing Am(IV) in vivo.

  1. Separation of lanthanides from trivalent actinides, the role of aqueous-phase soft-donor complexing agents

    SciTech Connect (OSTI)

    Nilsson, Mikael; Hoch, Cortney; Meier, G. Patrick; Nash, Kenneth L.

    2008-07-01

    Closing the nuclear fuel cycle to reduce storage volumes and times requires advanced separation processes, among which is the separation of trivalent actinides from lanthanides that are present in the waste. A proven system is TALSPEAK, utilizing polyamino-carboxylates for this group separation. However, the narrow pH range these molecules require complicates their use. Soft-donor molecules that may complex actinides at low pH have been investigated. Results indicate that, although DTPA gives the best selectivity, all molecules tested showed preference for americium. The solubility of some reagents at low pH suggests the need for further development. (authors)

  2. Trivalent Lanthanide/Actinide Separation Using Aqueous-Modified TALSPEAK Chemistry

    SciTech Connect (OSTI)

    Travis S. Grimes; Richard D. Tillotson; Leigh R. Martin

    2014-05-01

    TALSPEAK is a liquid/liquid extraction process designed to separate trivalent lanthanides (Ln3+) from minor actinides (MAs) Am3+ and Cm3+. Traditional TALSPEAK organic phase is comprised of a monoacidic dialkyl bis(2-ethylhexyl)phosphoric acid extractant (HDEHP) in diisopropyl benzene (DIPB). The aqueous phase contains a soluble aminopolycarboxylate diethylenetriamine-N,N,N’,N”,N”-pentaacetic acid (DTPA) in a concentrated (1.0-2.0 M) lactic acid (HL) buffer with the aqueous acidity typically adjusted to pH 3.0. TALSPEAK balances the selective complexation of the actinides by DTPA against the electrostatic attraction of the lanthanides by the HDEHP extractant to achieve the desired trivalent lanthanide/actinide group separation. Although TALSPEAK is considered a successful separations scheme, recent fundamental studies have highlighted complex chemical interactions occurring in the aqueous and organic phases during the extraction process. Previous attempts to model the system have shown thermodynamic models do not accurately predict the observed extraction trends in the p[H+] range 2.5-4.8. In this study, the aqueous phase is modified by replacing the lactic acid buffer with a variety of simple and longer-chain amino acid buffers. The results show successful trivalent lanthanide/actinide group separation with the aqueous-modified TALSPEAK process at pH 2. The amino acid buffer concentrations were reduced to 0.5 M (at pH 2) and separations were performed without any effect on phase transfer kinetics. Successful modeling of the aqueous-modified TALSPEAK process (p[H+] 1.6-3.1) using a simplified thermodynamic model and an internally consistent set of thermodynamic data is presented.

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

    SciTech Connect (OSTI)

    Howard Meyer

    2010-11-30

    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.

  4. The removal of uranium from acidic media using ion exchange and/or extraction chromatography

    SciTech Connect (OSTI)

    FitzPatrick, J.R.; Schake, B.S.; Murphy, J.; Holmes, K; West, M.H.

    1996-06-01

    The separation and purification of uranium from either nitric acid or hydrochloric acid media can be accomplished by using either solvent extraction or ion-exchange. Over the past two years at Los Alamos, emerging programs are focused on recapturing the expertise required to do limited, small-quantity processing of enriched uranium. During this period of time, we have been investigating ion-addition, waste stream polishing is associated with this effort in order to achieve more complete removal of uranium prior to recycle of the acid. Extraction chromatography has been demonstrated to further polish the uranium from both nitric and hydrochloric acid media thus allowing for a more complete recovery of the actinide material and creation of less waste during the processing steps.

  5. material removal

    National Nuclear Security Administration (NNSA)

    %2A en Nuclear Material Removal http:www.nnsa.energy.govaboutusourprogramsdnnm3remove

    Pag...

  6. 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.

    2013-06-19

    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.

  7. 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

    2012-07-01

    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.

  8. REMOVAL OF PROCESSING AIDS FROM CERAMIC/POLYMER GREGORY C. STANGLE, DONG-JOO RHEE, AND ILHAN A. AKSAY

    E-Print Network [OSTI]

    Aksay, Ilhan A.

    . For this purpose, fabrication of ceramics by injection molding techniques is a favored approach since, injection molding, debindering, and densifying. Following consolidaticn and prior to sintering without any defect creation. This paper summarizes the results of our theoretical studies of the removal

  9. 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, Mitchell R. (Troy, NY); Gal, Eli (Lititz, PA)

    1993-01-01

    A process and system 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 The government of the United States of America has rights in this invention pursuant to Contract No. DE-AC21-88MC 23174 awarded by the U.S. Department of Energy.

  10. 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.

    1993-04-13

    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.

  11. 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

    2010-12-10

    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.

  12. Minor actinide waste disposal in deep geological boreholes

    E-Print Network [OSTI]

    Sizer, Calvin Gregory

    2006-01-01

    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. ...

  13. 30th Actinide Separations Conference, PNNL-SA-50126

    SciTech Connect (OSTI)

    Delegard, Calvin H.

    2006-05-25

    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.

  14. ACTINIDE-SPECIFIC SEQUESTERING AGENTS AND DECONTAMINATION APPLICATIONS

    E-Print Network [OSTI]

    Smith, William L.

    2013-01-01

    In: The Health Eff, of Plutonium and Radium, Proc. Sym. ,The Metabolism of Compounds Plutonium and Other Actinides.In: The Radiobiology of Plutonium. Stover, B. J. , Jee, H.

  15. 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.

    1994-11-22

    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.

  16. 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)

    1994-01-01

    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.

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

    SciTech Connect (OSTI)

    Perkasa, Y. S.; Waris, A. Kurniadi, R. Su'ud, Z.

    2014-09-30

    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.

  18. The removal of precious metals by conductive polymer filtration

    SciTech Connect (OSTI)

    Cournoyer, M.E.; Stark, P.C.; Trujillo, S.M.; Aguino, R.E.

    1997-12-31

    The growing demand for platinum-group metals (PGM) for several applications within the DOE complex and in industry, the need for modern and clean processes and the increasing volume of low grade material for secondary PGM recovery have a direct impact on the industrial practice of recovering and refining of precious metals. PGM recovery have a direct impact on the industrial practice of recovering and refining of precious metals. With precipitation-dissolution methods being the most common method of recovery, there is a tremendous need for advanced metal ion recovery and waste minimization techniques. Here at Los Alamos there is an integrated program in ligand-design and separation`s chemistry for recovery of actinide and toxic metals from variety of process streams. In the present investigations, a novel hollow fiber membrane (CPI) is characterized and its selectivity for PGM reported. In addition, a continuous single unit operation is proposed for the removal, concentration and recovery of platinum from catalytic and electroplating process streams is proposed.

  19. Characterization of host phases for actinides in simulated metallic waste forms by transmission electron microscopy.

    SciTech Connect (OSTI)

    Janney, D. E.

    2005-11-21

    Argonne National Laboratory has developed an electrometallurgical process for conditioning spent sodium-bonded metallic reactor fuel prior to disposal. A waste stream from this process consists of stainless steel cladding hulls that contain undissolved metal fission products such as Tc, Ru, Rh, Pd, and Ag; a small amount of undissolved actinides (U, Np, Pu) also remains with the hulls. These wastes will be immobilized in a waste form whose baseline composition is stainless steel alloyed with 15 wt% Zr (SS-15Zr). Scanning electron microscope (SEM) observations of simulated metal waste forms (SS-15Zr with added actinides) show eutectic intergrowths of iron solid-solution (''steel'') and Fe-Zr-Cr-Ni (''intermetallic'') materials. The actinide elements are almost entirely in the intermetallic materials, where they occur in concentrations as high as 20 at%. Neutron- and electron-diffraction studies of the simulated waste forms show materials with structures similar to those of Fe{sub 2}Zr and Fe{sub 23}Zr{sub 6}. New TEM observations of simulated waste form samples with compositions SS-15Zr-2Np, SS-15Zr-5U, SS-15Zr-11U-0.6Ru-0.3Tc-0.1Pd, and SS-15Zr-10Pu suggest that the major U- and Pu-bearing phase has a structure similar to that of the C15 (cubic, MgCu{sub 2}-type) polymorph of Fe{sub 2}Zr. Materials with this structure exhibit significant variability in chemical compositions and actinide concentrations up to 20 at% (normalized so that atomic fractions of Cr, Ni, Fe, and Zr add up to 1). A U-bearing material similar to the C36 (dihexagonal, MgNi{sub 2}-type) polymorph of Fe{sub 2}Zr was also observed. Chemical variability in materials with the C36 Fe{sub 2}Zr structure is smaller than in those with the C15 Fe{sub 2}Zr structure, and U concentrations are less than 5 at%. Uranium concentrations up to 5 at.% were observed in materials with the Fe{sub 23}Zr{sub 6} (cubic, Mn{sub 23}Th{sub 6}-type) structure. Microstructures similar to those produced during experimental deformation of Fe-10 at% Zr alloys were observed in intermetallic materials in all of the simulated waste form samples. Stacking faults and associated dislocations are common in samples with U, but rarely observed in those with Np and Pu, while twins occur in all samples. Previously reported differences in dissolution behavior between samples with different actinides may be related to increased defect-assisted dissolution in samples with U.

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

    SciTech Connect (OSTI)

    Gautier, G. M.; Morin, F.; Dechelette, F.; Sanseigne, E.; Chabert, C.

    2012-07-01

    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)

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

    SciTech Connect (OSTI)

    Cipiti, Benjamin B.; Smith, James D.

    2008-09-01

    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.

  2. Thermal-Hydraulic Analyses of Transients in an Actinide-Burner Reactor Cooled by Forced Convection of Lead Bismuth

    SciTech Connect (OSTI)

    Davis, Cliff Bybee

    2003-09-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) and the Massachusetts Institute of Technology (MIT) are investigating the suitability of lead or lead–bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The current analysis evaluated a pool type design that relies on forced circulation of the primary coolant, a conventional steam power conversion system, and a passive decay heat removal system. The ATHENA computer code was used to simulate various transients without reactor scram, including a primary coolant pump trip, a station blackout, and a step reactivity insertion. The reactor design successfully met identified temperature limits for each of the transients analyzed.

  3. 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)

    2012-05-15

    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.

  4. CRITICALITY SAFETY OF PROCESSING SALT SOLUTION AT SRS

    SciTech Connect (OSTI)

    Stephens, K; Davoud Eghbali, D; Michelle Abney, M

    2008-01-15

    High level radioactive liquid waste generated as a result of the production of nuclear material for the United States defense program at the Savannah River Site has been stored as 36 million gallons in underground tanks. About ten percent of the waste volume is sludge, composed of insoluble metal hydroxides primarily hydroxides of Mn, Fe, Al, Hg, and most radionuclides including fission products. The remaining ninety percent of the waste volume is saltcake, composed of primarily sodium (nitrites, nitrates, and aluminates) and hydroxides. Saltcakes account for 30% of the radioactivity while the sludge accounts for 70% of the radioactivity. A pilot plant salt disposition processing system has been designed at the Savannah River Site for interim processing of salt solution and is composed of two facilities: the Actinide Removal Process Facility (ARPF) and the Modular Caustic Side Solvent Extraction Unit (MCU). Data from the pilot plant salt processing system will be used for future processing salt at a much higher rate in a new salt processing facility. Saltcake contains significant amounts of actinides, and other long-lived radioactive nuclides such as strontium and cesium that must be extracted prior to disposal as low level waste. The extracted radioactive nuclides will be mixed with the sludge from waste tanks and vitrified in another facility. Because of the presence of highly enriched uranium in the saltcake, there is a criticality concern associated with concentration and/or accumulation of fissionable material in the ARP and MCU.

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

    SciTech Connect (OSTI)

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

    2009-09-01

    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.

  6. FY2010 Annual Report for the Actinide Isomer Detection Project

    SciTech Connect (OSTI)

    Warren, Glen A.; Francy, Christopher J.; Ressler, Jennifer J.; Erikson, Luke E.; Miller, Erin A.; Hatarik, R.

    2011-01-01

    This project seeks to identify a new signature for actinide element detection in active interrogation. This technique works by exciting and identifying long-lived nuclear excited states (isomers) in the actinide isotopes and/or primary fission products. Observation of isomers in the fission products will provide a signature for fissile material. For the actinide isomers, the decay time and energy of the isomeric state is unique to a particular isotope, providing an unambiguous signature for Special Nuclear Materials (SNM). Future work will include a follow-up measurement scheduled for December 2010 at LBNL. Lessons learned from the July 2010 measurements will be incorporated into these new measurements. Analysis of both the July and December experiments will be completed in a few months. A research paper to be submitted to a peer-reviewed journal will be drafted if the conclusions from the measurements warrant publication.

  7. DEVELOPMENT PROGRAM FOR PU-238 AQUEOUS RECOVERY PROCESS

    SciTech Connect (OSTI)

    M. PANSOY-HJELVIK; M. REIMUS; ET AL

    2000-10-01

    Aqueous processing is necessary for the removal of impurities from {sup 238}Pu dioxide ({sup 238}PuO{sub 2}) fuel due to unacceptable levels of {sup 234}U and other non-actinide impurities in the scrap fuel. Impurities at levels above General Purpose Heat Source (GPHS) fuel specifications may impair the performance.of the heat sources. Efforts at Los Alamos have focused on developing the bench scale methodology for the aqueous process steps which includes comminution, dissolution, ion exchange, precipitation, and calcination. Recently, work has been performed to qualify the bench scale methodology, to show that the developed process produces pure {sup 238}PuO{sub 2} meeting GPHS fuel specifications. In addition, this work has enabled us to determine how waste volumes may be minimized during full-scale processing. Results of process qualification for the bench scale aqueous recovery operation and waste minimization efforts are presented.

  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

    2012-06-01

    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 Ion Sensor For Pyroprocess Monitoring - Energy Innovation Portal

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 OutreachProductswsicloudwsicloudden DocumentationAccommodations AccommodationsAcquisitionAcronymsActinideActinide

  10. 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)

    1989-01-01

    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.

  11. 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.

    1992-09-01

    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.

  12. High-field magnetization processes in actinide intermetallics (invited)

    SciTech Connect (OSTI)

    Franse, J.J.M.; de Boer, F.R.; de Chatel, P.F.; Frings, P.H.; Menovsky, A.A. )

    1991-04-15

    In magnetically ordered intermetallics of uranium with {ital d} transition elements, the magnetic moment on the uranium site is often limited to values below 0.1{mu}{sub {ital B}}, with, in some cases, extremely large magnetic anisotropies. Several approaches are followed for explaining these small uranium moments: opposite directions and almost compensation of the spin and orbital moments, reduction of the uranium 5{ital f} moment by strong hybridization effects between the 5{ital f} and conduction electrons, and very weak itinerant magnetism of the 5{ital f} electrons. In the Laves-phase compounds UFe{sub 2} and UNi{sub 2}, the magnetic data have been explained in terms of opposite spin and orbital moments on the uranium sites. In the heavy-fermion compounds UPt{sub 3} and URu{sub 2}Si{sub 2}, on the contrary, a Kondo approach is followed, although coherence effects largely complicate a proper description. The experimental evidence for these different approaches will be reviewed.

  13. Methyltrihydroborate complexes of the lanthanides and actinides

    SciTech Connect (OSTI)

    Shinomoto, R.S.

    1984-11-01

    Reaction of MC1/sub 4/ (M = Zr, Hf, U, Th, Np) with LiBH/sub 3/CH/sub 3/ in chlorobenzene produces volatile, hexane-soluble M(BH/sub 3/CH/sub 3/)/sub 4/. Crystal structures are monomeric, tetrahedral species. Lewis base adducts prepared include U(BH/sub 3/CH/sub 3/)/sub 4/.THT, Th(BH/sub 3/CH/sub 3/)/sub 4/.L (L = THF (tetrahydrofuran), THT (tetrahydrothiophene), SMe/sub 2/, OMe/sub 2/), U(BH/sub 3/CH/sub 3/)/sub 4/.2L (L = THF, pyridine, NH/sub 3/), Th(BH/sub 3/CH/sub 3/)/sub 4/.2L (L = THF, THT, py, NH/sub 3/), M(BH/sub 3/CH/sub 3/)/sub 4/.L-L (M = U, Th; L-L = dme (1,2-dimethoxyethane), bmte (bis(1,2-methylthio)ethane), tmed (N,N,N',N'-tetramethylethylenediamine), dmpe (1,2-dimethylphosphinoethane)) and Th(BH/sub 3/CH/sub 3/)/sub 4/.1/2 OEt/sub 2/. Reaction of MC1/sub 3/ (M = Ho, Yb, Lu) with LiBH/sub 3/CH/sub 3/ in diethyl ether produces volatile, toluene-soluble M(BH/sub 3/CH/sub 3/)/sub 3/.OEt/sub 2/. Other Lewis base adducts prepared from M(BH/sub 3/CH/sub 3/)/sub 3/.OEt/sub 2/ include Ho(BH/sub 3/CH/sub 3/)/sub 3/.L (L = THT, THF, py), Ho(BH/sub 3/CH/sub 3/)/sub 3/.2L (L = THT, THF, py), Ho(BH/sub 3/CH/sub 3/)/sub 3/.tmed, Ho(BH/sub 3/CH/sub 3/)/sub 3/.3/2 L-L (L-L = dmpe, bmte), Yb(BH/sub 3/CH/sub 3/)/sub 3/.3/2 dmpe, Yb(BH/sub 3/Ch/sub 3/).L (L = THF, dme), Yb(BH/sub 3/CH/sub 3/)/sub 3/.2THF, and Lu(BH/sub 3/CH/sub 3/)/sub 3/.THF. By structural criteria, the bonding in actinide and lanthanide methyltrihydroborate complexes is primarily ionic in character even though they display covalent-like physical properties. Spectroscopic measurements indicate that there is some degree of covalent bonding in U(BH/sub 3/CH/sub 3/)/sub 4/.

  14. Literature review of intrinsic actinide colloids related to spent fuel waste package release rates

    SciTech Connect (OSTI)

    Zhao, P.; Steward, S.A.

    1997-01-01

    Existence of actinide colloids provides an important mechanism in the migration of radionuclides and will be important in performance of a geologic repository for high-level nuclear waste. Actinide colloids have been formed during long-term unsaturated dissolution of spent fuel by groundwater. This article summarizes a literature search of actinide colloids. This report emphasizes the formation of intrinsic actinide colloids, because they would have the opportunity to form soon after groundwater contact with the spent fuel and before actinide-bearing groundwater reaches the surrounding geologic formations.

  15. 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)

    1980-01-01

    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.

  16. Theoretical Design of a Thermosyphon for Efficient Process Heat Removal from Next Generation Nuclear Plant (NGNP) for Production of Hydrogen

    SciTech Connect (OSTI)

    Piyush Sabharwall; Fred Gunnerson; Akira Tokuhiro; Vivek Utgiker; Kevan Weaver; Steven Sherman

    2007-10-01

    The work reported here is the preliminary analysis of two-phase Thermosyphon heat transfer performance with various alkali metals. Thermosyphon is a device for transporting heat from one point to another with quite extraordinary properties. Heat transport occurs via evaporation and condensation, and the heat transport fluid is re-circulated by gravitational force. With this mode of heat transfer, the thermosyphon has the capability to transport heat at high rates over appreciable distances, virtually isothermally and without any requirement for external pumping devices. For process heat, intermediate heat exchangers (IHX) are required to transfer heat from the NGNP to the hydrogen plant in the most efficient way possible. The production of power at higher efficiency using Brayton Cycle, and hydrogen production requires both heat at higher temperatures (up to 1000oC) and high effectiveness compact heat exchangers to transfer heat to either the power or process cycle. The purpose for selecting a compact heat exchanger is to maximize the heat transfer surface area per volume of heat exchanger; this has the benefit of reducing heat exchanger size and heat losses. The IHX design requirements are governed by the allowable temperature drop between the outlet of the NGNP (900oC, based on the current capabilities of NGNP), and the temperatures in the hydrogen production plant. Spiral Heat Exchangers (SHE’s) have superior heat transfer characteristics, and are less susceptible to fouling. Further, heat losses to surroundings are minimized because of its compact configuration. SHEs have never been examined for phase-change heat transfer applications. The research presented provides useful information for thermosyphon design and Spiral Heat Exchanger.

  17. Isotopic validation for PWR actinide-only burnup credit using Yankee Rowe data

    SciTech Connect (OSTI)

    NONE

    1997-11-01

    Safety analyses of criticality control systems for transportation packages include an assumption that the spent nuclear fuel (SNF) loaded into the package is fresh or unirradiated. In other words, the spent fuel is assumed to have its original, as-manufactured U-235 isotopic content. The ``fresh fuel`` assumption is very conservative since the potential reactivity of the nuclear fuel is substantially reduced after being irradiated in the reactor core. The concept of taking credit for this reduction in nuclear fuel reactivity due to burnup of the fuel, instead of using the fresh fuel assumption in the criticality safety analysis, is referred to as ``Burnup Credit.`` Burnup credit uses the actual physical composition of the fuel and accounts for the net reduction of fissile material and the buildup of neutron absorbers in the fuel as it is irradiated. Neutron absorbers include actinides and other isotopes generated as a result of the fission process. Using only the change in actinide isotopes in the burnup credit criticality analysis is referred to as ``Actinide-Only Burnup Credit.`` The use of burnup credit in the design of criticality control systems enables more spent fuel to be placed in a package. Increased package capacity results in a reduced number of storage, shipping and disposal containers for a given number of SNF assemblies. Fewer shipments result in a lower risk of accidents associated with the handling and transportation of spent fuel, thus reducing both radiological and nonradiological risk to the public. This paper describes the modeling and the results of comparison between measured and calculated isotopic inventories for a selected number of samples taken from a Yankee Rowe spent fuel assembly.

  18. 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.

    1984-09-12

    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.

  19. Actinide Targets for Neutron Cross Section Measurements (C)

    SciTech Connect (OSTI)

    J. D. Baker; C. A. McGrath

    2006-04-01

    The Advanced Fuel Cycle Initiative (AFCI) and the Generation IV Reactor Initiative have demonstrated a lack of detailed neutron cross-sections for certain "minor" actinides, those other than the most common (235U, 238U, and 239Pu). For some closed-fuel-cycle reactor designs more than 50% of reactivity will, at some point, be derived from “minor” actinides that currently have poorly known (n,g) and (n,f) cross sections. A program of measurements under AFCI has begun to correct this. One of the initial hurdles has been to produce well-characterized, highly isotopically enriched, and chemically pure actinide targets on thin backings. Using a combination of resurrected techniques and new developments, we have made a series of targets including highly enriched 240Pu, and 242Pu. Thus far, we have electrodeposited these actinide targets. In the future, we plan to study reductive distillation to achieve homogeneous, adherent targets on thin metal foils and polymer backings. As we move forward, separated isotopes become scarcer, and safety concerns become greater. The chemical purification and electodeposition techniques will be described.

  20. Conceptual Design of a Simplified Skid-Mounted Caustic-Side Solvent Extraction Process for Removal of Cesium from Savannah Rive Site High-Level Waste

    SciTech Connect (OSTI)

    Birdwell, JR.J.F.

    2004-05-12

    This report presents the results of a conceptual design of a solvent extraction process for the selective removal of {sup 137}Cs from high-level radioactive waste currently stored in underground tanks at the U.S. Department of Energy's Savannah River Site (SRS). This study establishes the need for and feasibility of deploying a simplified version of the Caustic-Side Solvent Extraction (CSSX) process; cost/benefit ratios ranging from 33 to 55 strongly support the considered deployment. Based on projected compositions, 18 million gallons of dissolved salt cake waste has been identified as having {sup 137}Cs concentrations that are substantially lower than the worst-case design basis for the CSSX system that is to be deployed as part of the Salt Waste Processing Facility (SWPF) but that does not meet the waste acceptance criteria for immobilization as grout in the Saltstone Manufacturing and Disposal Facility at SRS. Absent deployment of an alternative cesium removal process, this material will require treatment in the SWPF CSSX system, even though the cesium decontamination factor required is far less than that provided by that system. A conceptual design of a CSSX processing system designed for rapid deployment and having reduced cesium decontamination factor capability has been performed. The proposed accelerated-deployment CSSX system (CSSX-A) has been designed to have a processing rate of 3 million gallons per year, assuming 90% availability. At a more conservative availability of 75% (reflecting the novelty of the process), the annual processing capacity is 2.5 million gallons. The primary component of the process is a 20-stage cascade of centrifugal solvent extraction contactors. The decontamination and concentration factors are 40 and 15, respectively. The solvent, scrub, strip, and wash solutions are to have the same compositions as those planned for the SWPF CSSX system. As in the SWPF CSSX system, the solvent and scrub flow rates are equal. The system is designed to facilitate remote operation and direct maintenance. Two general deployment concepts were considered: (1) deployment in an existing but unused SRS facility and (2) deployment in transportable containers. Deployment in three transportable containers was selected as the preferred option, based on concerns regarding facility availability (due to competition from other processing alternatives) and decontamination and renovation costs. A risk assessment identified environmental, safety, and health issues that exist. These concerns have been addressed in the conceptual design by inclusion of mitigating system features. Due to the highly developed state of CSSX technology, only a few technical issues remain unresolved; however, none of these issues have the potential to make the technology unviable. Recommended development tasks that need to be performed to address technical uncertainties are discussed in this report. Deployment of the proposed CSSX-A system provides significant qualitative and quantitative benefits. The qualitative benefits include (1) verification of full-scale contactor performance under CSSX conditions that will support SWPF CSSX design and deployment; (2) development of design, fabrication, and installation experience bases that will be at least partially applicable to the SWPF CSSX system; and (3) availability of the CSSX-A system as a means of providing contactor-based solvent extraction system operating experience to SWPF CSSX operating personnel. Estimates of fixed capital investment, development costs, and annual operating cost for SRS deployment of the CSSX-A system (in mid-2003 dollars) are $9,165,199, $2,734,801, and $2,108,820, respectively. When the economics of the CSSX-A system are compared with those of the baseline SWPF CSSX system, benefit-to-cost ratios ranging from 20 to 47 are obtained. The benefits in the cost/benefit comparison arise from expedited tank closure and reduced engineering, construction, and operating costs for the SWPF CSSX system. No significant impediments to deployment were determined in the reported a

  1. 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

    2009-12-01

    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.

  2. Spin and orbital moments in actinide compounds (invited)

    SciTech Connect (OSTI)

    Lebech, B. ); Wulff, M.; Lander, G.H. )

    1991-04-15

    The extended spatial distribution of both the transition-metal 3{ital d} electrons and the actinide 5{ital f} electrons results in a strong interaction between these electron states when the relevant elements are alloyed. A particular interesting feature of this hybridization, which is predicted by single-electron band-structure calculations, is that the orbital moments of the actinide 5{ital f} electrons are considerably reduced from the values anticipated by a simple application of Hund's rules. To test these ideas, and thus to obtain a measure of the hybridization, we have performed a series of neutron scattering experiments designed to determine the magnetic moments at the actinide and transition-metal sublattice sites in compounds such as UFe{sub 2}, NpCo{sub 2}, and PuFe{sub 2} and to separate the spin and orbital components at the actinide sites. The results show, indeed, that the ratio of the orbital to spin moment is reduced as compared to the free-ion expectations. In addition there is qualitative agreement with theory, although the latter predicts values of both components that are larger than those found by experiment. Because {bold L} and {bold S} are opposed in the light actinides, and {ital L} is usually greater than {ital S}, the reduction of {ital L} can result in a situation for which {ital L}{minus}{ital S}{congruent}0. This almost occurs in UFe{sub 2}. However, neutrons are capable of observing the individual components at finite wave vector ({bold Q}), although the total component (observed at {bold Q}={bold 0}) may indeed be close to zero.

  3. Development of the Next-Generation Caustic-Side Solvent Extraction (NG-CSSX) Process for Cesium Removal from High-Level Tank Waste

    SciTech Connect (OSTI)

    Moyer, Bruce A; Bonnesen, Peter V; Delmau, Laetitia Helene; Sloop Jr, Frederick {Fred} V; Williams, Neil J; Birdwell Jr, Joseph F; Lee, Denise L; Leonard, Ralph; Fink, Samuel D; Peters, Thomas B.; Geeting, Mark W

    2011-01-01

    This paper describes the chemical performance of the Next-Generation Caustic-Side Solvent Extraction (NG-CSSX) process in its current state of development for removal of cesium from the alkaline high-level tank wastes at the Savannah River Site (SRS) in the US Department of Energy (USDOE) complex. Overall, motivation for seeking a major enhancement in performance for the currently deployed CSSX process stems from needs for accelerating the cleanup schedule and reducing the cost of salt-waste disposition. The primary target of the NG-CSSX development campaign in the past year has been to formulate a solvent system and to design a corresponding flowsheet that boosts the performance of the SRS Modular CSSX Unit (MCU) from a current minimum decontamination factor of 12 to 40,000. The chemical approach entails use of a more soluble calixarene-crown ether, called MaxCalix, allowing the attainment of much higher cesium distribution ratios (DCs) on extraction. Concurrently decreasing the Cs-7SB modifier concentration is anticipated to promote better hydraulics. A new stripping chemistry has been devised using a vitrification-friendly aqueous boric acid strip solution and a guanidine suppressor in the solvent, resulting in sharply decreased DCs on stripping. Results are reported herein on solvent phase behavior and batch Cs distribution for waste simulants and real waste together with a preliminary flowsheet applicable for implementation in the MCU. The new solvent will enable MCU to process a much wider range of salt feeds and thereby extend its service lifetime beyond its design life of three years. Other potential benefits of NG-CSSX include increased throughput of the SRS Salt Waste Processing Facility (SWPF), currently under construction, and an alternative modular near-tank application at Hanford.

  4. Hollow-fiber gas-membrane process for removal of NH{sub 3} from solution of NH{sub 3} and CO{sub 2}

    SciTech Connect (OSTI)

    Qin, Y.; Cabral, J.M.S.; Wang, S.

    1996-07-01

    A hollow-fiber supported gas membrane process for the separation of NH{sub 3} from aqueous solutions containing both NH{sub 3} and CO{sub 2} was investigated theoretically and experimentally. A lumen laminar flow and radial diffusion model was applied to calculate the membrane wall transfer coefficient from the data stripping a single volatile component, NH{sub 3} or CO{sub 2}, from their individual aqueous solutions. Influence of the type of membranes and operating conditions on mass-transfer rate were discussed, especially the influence of the membrane transfer coefficient on the film mass-transfer coefficient in the lumen. Appropriate configurations of the hollow-fiber modules for stripping of a single component were analyzed to optimize mass transfer. To predict the stripping of NH{sub 3} from a solution containing NH{sub 3} and CO{sub 2}, a mathematical model incorporating local chemical equilibria and Nernst-Planck diffusion was developed to describe the mass transport. The models described the experimental data fairly well. The experimental results showed that the supported gas membrane process can be used to remove NH{sub 3} effectively from aqueous media containing NH{sub 3} and CO{sub 2}.

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

    SciTech Connect (OSTI)

    Morgan, Dane; Yang, Yong Austin

    2013-10-28

    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.

  6. Advancing Chemistry with the Lanthanide and Actinide Elements Final Report, September 2013

    SciTech Connect (OSTI)

    Evans, William John [Univ of California, Irvine

    2013-09-11

    The objective of this research is to use the unique chemistry available from complexes of the lanthanides and actinides, as well as related heavy metals such as scandium, yttrium, and bismuth to advance chemistry in energy-related areas. The lanthanides and actinides have a combination of properties in terms of size, charge, electropositive character, and f valence orbitals that provides special opportunities to probe reactivity and catalysis in ways not possible with the other metals in the periodic table. We seek to discover reaction pathways and structural types that reveal new options in reaction chemistry related to energy. Identification of new paradigms in structure and reactivity should stimulate efforts to develop new types of catalytic processes that at present are not under consideration because either the transformation or the necessary intermediates are unknown. This project is one half of my laboratory’s DOE research which was split 50:50 between Catalysis and Heavy Element Chemistry programs in 2010. Hence, this report is for a half-project.

  7. Determination of actinides in environmental samples with ICP-MS and automated batch preconcentration

    SciTech Connect (OSTI)

    Crain, J.S.; Smith, L.L.; Alvaradao, J.A. [Argonne National Laboratory, IL (United States)] [and others

    1994-12-31

    Thorium, uranium, and uranium progeny (e.g., {sup 226}Ra) were isolated from a variety of natural water matrices (well, spring, lake, river, and tap waters) using automated batch separation and a proprietary chelating resin. Minimal sample treatment was required in advance of the chemical separation procedure, i.e., samples were acidified, enriched isotopes were added and equilibrated, and, in certain cases, samples were filtered to remove biomass and other particulate matter. Major elemental constituents (e.g., Mg and Ca) were removed by the chemical separation. The isolated actinides and their detectable progeny were then determined using isotope dilution inductively coupled plasma-mass spectrometry combined with ultrasonic nebulization; flow injection was also used to improve sample utilization efficiency and increase the practical limit of preconcentration. Preconcentration factors of ca. 50 were achievable, thereby allowing determination of the isotopes TA interest at concentrations below 1 pg/L (i.e., 1 pCi/L of {sup 226}Ra). The internal and external reproducibility of these measurements will be described, as will their comparability to results obtained by radiochemical means. The authors will also discuss the utility of this procedure for environmental surveillance and the examination of uranium series disequilibria in natural aqueous systems.

  8. Removal of dissolved actinides from alkaline solutions by the method of appearing reagents

    DOE Patents [OSTI]

    Krot, Nikolai N. (Chelomiya St., 2, Apartment 15, Moscow, RU); Charushnikova, Iraida A. (Svoboda St., Bldg. 7A, Apartment 15, Moscow, RU)

    1997-01-01

    A method of reducing the concentration of neptunium and plutonium from alkaline radwastes containing plutonium and neptunium values along with other transuranic values produced during the course of plutonium production. The OH.sup.- concentration of the alkaline radwaste is adjusted to between about 0.1M and about 4M. [UO.sub.2 (O.sub.2).sub.3 ].sup.4- ion is added to the radwastes in the presence of catalytic amounts of Cu.sup.+2, Co.sup.+2 or Fe.sup.+2 with heating to a temperature in excess of about 60.degree. C. or 85.degree. C., depending on the catalyst, to coprecipitate plutonium and neptunium from the radwaste. Thereafter, the coprecipitate is separated from the alkaline radwaste.

  9. Modeling and simulation of material removal with particulate flows

    E-Print Network [OSTI]

    Arbelaez, D; Zohdi, T I; Dornfeld, David

    2008-01-01

    optimization In material removal processes there are a large number of input parameters which control the quality of the process.

  10. LANL Plutonium-Processing Facilities National Security

    E-Print Network [OSTI]

    range of national security programs that involve stockpile stewardship, plutonium processing, nuclear materials stabilization, materials disposition, nuclear forensics, nuclear counter-terrorism, and nuclear of technical capabilities. These capabilities form a center of excellence for actinide science and technology

  11. 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.

    2003-02-26

    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.

  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

    2005-08-19

    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. Determination of actinides in urine and fecal samples

    DOE Patents [OSTI]

    McKibbin, T.T.

    1993-03-02

    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.

  14. Determination of actinides in urine and fecal samples

    DOE Patents [OSTI]

    McKibbin, Terry T. (Larimer County, CO)

    1993-01-01

    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.

  15. 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.; Adamson, Clinton J.; Zacharias, Ames

    2012-07-01

    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

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

    DOE Patents [OSTI]

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

    1989-05-30

    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.

  17. 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.

    2011-07-13

    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.

  18. 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...

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

    E-Print Network [OSTI]

    Heaven, Michael C.

    2011-01-01

    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

  20. Laser Detection of Actinides and Other Elements | U.S. DOE Office...

    Office of Science (SC) Website

    Laser Detection of Actinides and Other Elements Nuclear Physics (NP) NP Home About Research Facilities Science Highlights Benefits of NP Funding Opportunities Nuclear Science...

  1. Radionuclide partitioning in the modified Unex process

    SciTech Connect (OSTI)

    Babain, V.; Smirnov, I.; Alyapyshev, M.; Todd, T.A.; Law, J.D.; Herbst, R.S.; Paulenova, A.

    2008-07-01

    The Universal Extraction (UNEX) process has been developed for simultaneous extraction of long-lived radionuclides (cesium, strontium, actinides, and lanthanides) from acidic solutions in one extraction cycle. Modification of this organic solvent through the use of diamides of dipicolinic acid instead of CMPO increases the extraction capacity of UNEX solvent toward lanthanides and actinide metals, allowing for the processing of spent nuclear fuel. The possibility of radionuclide group separation using the modified UNEX solvent [HCCD (chlorinated cobalt dicarbollide), TBDPA (tetrabutyl-diamide of dipicolinic acid), PEG in FS-1 3 (phenyl-trifluoromethyl-sulfone)] is being investigated. Individual strip products, including a) actinides and lanthanides, b) strontium, and c) cesium, can be obtained by selective stripping from UNEX solvent. Such partitioning will make it possible to transform the Cs/Sr product into the most stable matrices for long-term storage and to further process the actinide/lanthanide product for recycling to a nuclear reactor. (authors)

  2. 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)

    2013-07-01

    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)

  3. 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

    2013-09-01

    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.

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

    SciTech Connect (OSTI)

    Bliss, Mary; Stave, Jean A.

    2012-10-01

    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.

  5. Chemistry of lower valent actinide halides. Final report

    SciTech Connect (OSTI)

    Lau, K.H.; Hildenbrand, D.L.

    1992-01-01

    This research effort was concerned almost entirely with the first two members of the actinide series, thorium and uranium, although the work was later extended to some aspects of the neptunium-fluorine system in a collaborative program with Los Alamos National Laboratory. Detailed information about the lighter actinides will be helpful in modeling the properties of the heavier actinide compounds, which will be much more difficult to study experimentally. In this program, thermochemical information was obtained from high temperature equilibrium measurements made by effusion-beam mass spectrometry and by effusion-pressure techniques. Data were derived primarily from second-law analysis so as to avoid potential errors in third-law calculations resulting from uncertainties in spectroscopic and molecular constants. This approach has the additional advantage of yielding reaction entropies that can be checked for consistency with various molecular constant assignments for the species involved. In the U-F, U-Cl, and U-Br systems, all of the gaseous species UX, UX{sub 2}, UX{sub 3}, UX{sub 4}, and UX{sub 5}, where X represents the halogen, were identified and characterized; the corresponding species ThX, ThX{sub 2}, ThX{sub 3}, and ThX{sub 4} were studied in the Th-F, Th-Cl, and Th-Br systems. A number of oxyhalide species in the systems U-0-F, U-0-Cl, Th-0-F, and Th-O-Cl were studied thermochemically. Additionally, the sublimation thermodynamics of NpF{sub 4}(s) and NpO{sub 2}F{sub 2}(s) were studied by mass spectrometry.

  6. Selective extraction of trivalent actinides from lanthanides with dithiophosphinic acids and tributylphosphate

    SciTech Connect (OSTI)

    Jarvinen, G.; Barrans, R.; Schroeder, N.; Wade, K.; Jones, M.; Smith, B.F.; Mills, J.; Howard, G.; Freiser, H.; Muralidharan, S.

    1995-01-01

    A variety of chemical systems have been developed to separate trivalent actinides from lanthanides based on the slightly stronger complexation of the trivalent actinides with ligands that contain soft donor atoms. The greater stability of the actinide complexes in these systems has often been attributed to a slightly greater covalent bonding component for the actinide ions relative to the lanthanide ions. The authors have investigated several synergistic extraction systems that use ligands with a combination of oxygen and sulfur donor atoms that achieve a good group separation of the trivalent actinides and lanthanides. For example, the combination of dicyclohexyldithiophosphinic acid and tributylphosphate has shown separation factors of up to 800 for americium over europium in a single extraction stage. Such systems could find application in advanced partitioning schemes for nuclear waste.

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

    DOE Patents [OSTI]

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

    1997-01-21

    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.

  8. Actinide Foil Production for MPACT Research

    SciTech Connect (OSTI)

    Beller, Denis

    2012-10-31

    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

  9. Actinide production from xenon bombardments of curium-248

    SciTech Connect (OSTI)

    Welch, R.B.

    1985-01-01

    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.

  10. In situ removal of contamination from soil

    DOE Patents [OSTI]

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

    1997-10-14

    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.

  11. 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.

    1991-12-01

    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.

  12. 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)

    2005-05-24

    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.

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

    SciTech Connect (OSTI)

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

    2013-03-21

    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.

  14. Solid-state actinide acid phosphites from phosphorous acid melts

    SciTech Connect (OSTI)

    Oh, George N. [Department of Civil and Environmental Engineering and Earth Sciences, University of Notre Dame, 156 Fitzpatrick Hall, Notre Dame, IN 46556 (United States); Burns, Peter C., E-mail: pburns@nd.edu [Department of Civil and Environmental Engineering and Earth Sciences, University of Notre Dame, 156 Fitzpatrick Hall, Notre Dame, IN 46556 (United States); Department of Chemistry and Biochemistry, University of Notre Dame, Notre Dame, IN 46556 (United States)

    2014-07-01

    The reaction of UO{sub 3} and H{sub 3}PO{sub 3} at 100 °C and subsequent reaction with dimethylformamide (DMF) produces crystals of the compound (NH{sub 2}(CH{sub 3}){sub 2})[UO{sub 2}(HPO{sub 2}OH)(HPO{sub 3})]. This compound crystallizes in space group P2{sub 1}/n and consists of layers of uranyl pentagonal bipyramids that share equatorial vertices with phosphite units, separated by dimethylammonium. In contrast, the reaction of phosphorous acid and actinide oxides at 210 °C produces a viscous syrup. Subsequent dilution in solvents and use of standard solution-state methods results in the crystallization of two polymorphs of the actinide acid phosphites An(HPO{sub 2}OH){sub 4} (An=U, Th) and of the mixed acid phosphite–phosphite U(HPO{sub 3})(HPO{sub 2}OH){sub 2}(H{sub 2}O)·2(H{sub 2}O). ?- and ?-An(HPO{sub 2}OH){sub 4} crystallize in space groups C2/c and P2{sub 1}/n, respectively, and comprise a three-dimensional network of An{sup 4+} cations in square antiprismatic coordination corner-sharing with protonated phosphite units, whereas U(HPO{sub 3})(HPO{sub 2}OH){sub 2}(H{sub 2}O){sub 2}·(H{sub 2}O) crystallizes in a layered structure in space group Pbca that is composed of An{sup 4+} cations in square antiprismatic coordination corner-sharing with protonated phosphites and water ligands. We discuss our findings in using solid inorganic reagents to produce a solution-workable precursor from which solid-state compounds can be crystallized. - Graphical abstract: Reaction of UO{sub 3} and H{sub 3}PO{sub 3} at 100 °C and subsequent reaction with DMF produces crystals of (NH{sub 2}(CH{sub 3}){sub 2})[UO{sub 2}(HPO{sub 2}OH)(HPO{sub 3})] with a layered structure. Reaction of phosphorous acid and actinide oxides at 210 °C produces a viscous syrup and further solution-state reactions result in the crystallization of the actinide acid phosphites An(HPO{sub 2}OH){sub 4} (An=U, Th), with a three-dimensional network structure, and the mixed acid phosphite–phosphite U(HPO{sub 3})(HPO{sub 2}OH){sub 2}(H{sub 2}O){sub 2}·(H{sub 2}O) with a layered structure. - Highlights: • U(VI), U(IV) and Th(IV) phosphites were synthesized by solution-state methods. • A new uranyl phosphite structure is based upon uranyl phosphite anionic sheets. • New U and Th phosphites have framework structures.

  15. Process development for removal and recovery of cadmium from wastewater by a low-cost adsorbent: Adsorption rates and equilibrium studies

    SciTech Connect (OSTI)

    Periasamy, K.; Namasivayam, C. (Bharathiar Univ., Tamil Nadu (India). Dept. of Environmental Sciences)

    1994-02-01

    Activated carbon prepared from peanut hulls (PHC), an agricultural waste by-product, has been used for the adsorption of Cd(II) from synthetic wastewater. The adsorption data fit better with the Freundlich adsorption isotherm. The applicability of the Lagergren kinetic model has also been investigated. An almost quantitative removal of 20 mg/L Cd(II) by 0.7 g of PHC/L of aqueous solution was observed in the pH range 3.5--9.5. A comparative study with a commercial granular activated carbon (CAC) showed that the adsorption capacity (K[sub f]) of PHC was 31 times larger than that of CAC.

  16. In situ removal of contamination from soil

    DOE Patents [OSTI]

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

    1997-01-01

    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.

  17. Effect of Spin-Orbit Coupling on the Actinide Dioxides AnO2 ...

    Office of Scientific and Technical Information (OSTI)

    Coupling on the Actinide Dioxides AnO2 (AnTh, Pa, U, Np, Pu, and Am): A Screened Hybrid Density Functional Study Citation Details In-Document Search Title: Effect of...

  18. Large Component Removal/Disposal

    SciTech Connect (OSTI)

    Wheeler, D. M.

    2002-02-27

    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.

  19. Testing actinide fission yield treatment in CINDER90 for use in MCNP6 burnup calculations

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

    Fensin, Michael Lorne; Umbel, Marissa

    2015-09-18

    Most of the development of the MCNPX/6 burnup capability focused on features that were applied to the Boltzman transport or used to prepare coefficients for use in CINDER90, with little change to CINDER90 or the CINDER90 data. Though a scheme exists for best solving the coupled Boltzman and Bateman equations, the most significant approximation is that the employed nuclear data are correct and complete. Thus, the CINDER90 library file contains 60 different actinide fission yields encompassing 36 fissionable actinides (thermal, fast, high energy and spontaneous fission). Fission reaction data exists for more than 60 actinides and as a result, fissionmore »yield data must be approximated for actinides that do not possess fission yield information. Several types of approximations are used for estimating fission yields for actinides which do not possess explicit fission yield data. The objective of this study is to test whether or not certain approximations of fission yield selection have any impact on predictability of major actinides and fission products. Further we assess which other fission products, available in MCNP6 Tier 3, result in the largest difference in production. Because the CINDER90 library file is in ASCII format and therefore easily amendable, we assess reasons for choosing, as well as compare actinide and major fission product prediction for the H. B. Robinson benchmark for, three separate fission yield selection methods: (1) the current CINDER90 library file method (Base); (2) the element method (Element); and (3) the isobar method (Isobar). Results show that the three methods tested result in similar prediction of major actinides, Tc-99 and Cs-137; however, certain fission products resulted in significantly different production depending on the method of choice.« less

  20. Testing actinide fission yield treatment in CINDER90 for use in MCNP6 burnup calculations

    SciTech Connect (OSTI)

    Fensin, Michael Lorne; Umbel, Marissa

    2015-09-18

    Most of the development of the MCNPX/6 burnup capability focused on features that were applied to the Boltzman transport or used to prepare coefficients for use in CINDER90, with little change to CINDER90 or the CINDER90 data. Though a scheme exists for best solving the coupled Boltzman and Bateman equations, the most significant approximation is that the employed nuclear data are correct and complete. Thus, the CINDER90 library file contains 60 different actinide fission yields encompassing 36 fissionable actinides (thermal, fast, high energy and spontaneous fission). Fission reaction data exists for more than 60 actinides and as a result, fission yield data must be approximated for actinides that do not possess fission yield information. Several types of approximations are used for estimating fission yields for actinides which do not possess explicit fission yield data. The objective of this study is to test whether or not certain approximations of fission yield selection have any impact on predictability of major actinides and fission products. Further we assess which other fission products, available in MCNP6 Tier 3, result in the largest difference in production. Because the CINDER90 library file is in ASCII format and therefore easily amendable, we assess reasons for choosing, as well as compare actinide and major fission product prediction for the H. B. Robinson benchmark for, three separate fission yield selection methods: (1) the current CINDER90 library file method (Base); (2) the element method (Element); and (3) the isobar method (Isobar). Results show that the three methods tested result in similar prediction of major actinides, Tc-99 and Cs-137; however, certain fission products resulted in significantly different production depending on the method of choice.

  1. Enhancing the actinide sciences in Europe through hot laboratories networking and pooling: from ACTINET to TALISMAN

    SciTech Connect (OSTI)

    Bourg, S.; Poinssot, C.

    2013-07-01

    Since 2004, Europe supports the strengthening of the European actinides sciences scientific community through the funding of dedicated networks: (i) from 2004 to 2008, the ACTINET6 network of excellence (6. Framework Programme) gathered major laboratories involved in nuclear research and a wide range of academic research organisations and universities with the specific aims of funding and implementing joint research projects to be performed within the network of pooled facilities; (ii) from 2009 to 2013, the ACTINET-I3 integrated infrastructure initiative (I3) supports the cost of access of any academics in the pooled EU hot laboratories. In this continuation, TALISMAN (Trans-national Access to Large Infrastructures for a Safe Management of Actinides) gathers now the main European hot laboratories in actinides sciences in order to promote their opening to academics and universities and strengthen the EU-skills in actinides sciences. Furthermore, a specific focus is set on the development of advanced cutting-edge experimental and spectroscopic capabilities, the combination of state-of-the art experimental with theoretical first-principle methods on a quantum mechanical level and to benefit from the synergy between the different scientific and technical communities. ACTINET-I3 and TALISMAN attach a great importance and promote the Education and Training of the young generation of actinides scientists in the Trans-national access but also by organizing Schools (general Summer Schools or Theoretical User Lab Schools) or by granting students to attend International Conference on actinide sciences. (authors)

  2. 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)

    2009-07-07

    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.

  3. Graphitic packing removal tool

    DOE Patents [OSTI]

    Meyers, Kurt Edward (Avella, PA); Kolsun, George J. (Pittsburgh, PA)

    1997-01-01

    Graphitic packing removal tools for removal of the seal rings in one piece. he packing removal tool has a cylindrical base ring the same size as the packing ring with a surface finish, perforations, knurling or threads for adhesion to the seal ring. Elongated leg shanks are mounted axially along the circumferential center. A slit or slits permit insertion around shafts. A removal tool follower stabilizes the upper portion of the legs to allow a spanner wrench to be used for insertion and removal.

  4. Graphitic packing removal tool

    DOE Patents [OSTI]

    Meyers, K.E.; Kolsun, G.J.

    1997-11-11

    Graphitic packing removal tools for removal of the seal rings in one piece are disclosed. The packing removal tool has a cylindrical base ring the same size as the packing ring with a surface finish, perforations, knurling or threads for adhesion to the seal ring. Elongated leg shanks are mounted axially along the circumferential center. A slit or slits permit insertion around shafts. A removal tool follower stabilizes the upper portion of the legs to allow a spanner wrench to be used for insertion and removal. 5 figs.

  5. 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.

    1984-08-01

    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.

  6. 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

    2013-01-01

    JL. Kinetics of coal gasification. Ind Eng Chem Process DesK, Ryo Y, et al. Coal gasification with a subcritical steamkinetic analysis of coal char gasification reactions at high

  7. 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

    2013-01-01

    2 capture-a review. Energ Fuel. 2012; 26: 2751-7. 14. Yongfor CO 2 sequestration. Fuel Process Technol. 2005; 86:http://en.wikipedia.org/wiki/Synthetic_fuel 20. Phillips J.

  8. 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

    2013-01-01

    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

  9. 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)

    2013-07-01

    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)

  10. Actinide production in /sup 136/Xe bombardments of /sup 249/Cf

    SciTech Connect (OSTI)

    Gregorich, K.E.

    1985-08-01

    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.

  11. Actinides and radiation effects: impact on the back-end of the nuclear fuel cycle

    SciTech Connect (OSTI)

    Ewing, R. C. [Univ. of Michigan, Ann Arbor, MI (United States)

    2011-08-01

    During the past 70 years, more than 2000 metric tonnes of Pu, and substantial quantities of the 'minor' actinides such as Np, Am and Cm, have been generated in nuclear reactors. Some of these transuranium elements can be a source of energy in fission reactions (e.g. ²³?Pu), a source of fissile material for nuclear weapons (e.g. ²³?Pu and ²³?Np), and of environmental concern because of their long half-lives and radiotoxicity (e.g. ²³?Pu and ²³?Np). There are two basic strategies for the disposition of these transuranium elements: (1) to 'burn' or fission the actinides using nuclear reactors or accelerators; (2) to dispose of the actinides directly as spent nuclear fuel or to 'sequester' the actinides in chemically durable, radiation-resistant materials that are also suitable for geological disposal. For the latter strategy, there has been substantial interest in the use of actinide-bearing minerals, especially isometric pyrochlore, A ? B ?O? (A = rare earths; B = Ti, Zr, Sn, Hf), for the immobilization of actinides, particularly plutonium, both as inert matrix fuels and nuclear waste forms. Systematic studies of rare-earth pyrochlores have led to the discovery that certain compositions (B = Zr, Hf) are stable to very high doses of a -decay event damage. Recent developments in the understanding of the properties of actinide-bearing solids have opened up new possibilities for the design of advanced nuclear materials that can be used as fuels and waste forms. As an example, the amount of radiation damage that accumulates over time can be controlled by the selection of an appropriate composition for the pyrochlore and a consideration of the thermal environment of disposal. In the case of deep borehole disposal (3–5 km), the natural geothermal gradient may provide enough heat to reduce the amount of accumulated radiation damage bythermal annealing.

  12. Iron (III) Matrix Effects on Mineralization and Immobilization of Actinides

    SciTech Connect (OSTI)

    Cynthia-May S. Gong; Tyler A. Sullens; Kenneth R. Czerwinski

    2006-01-01

    Abstract - A number of models for the Yucca Mountain Project nuclear waste repository use studies of actinide sorption onto well-defined iron hydroxide materials. In the case of a waste containment leak, however, a complex interaction between dissolved waste forms and failed containment vessel components can lead to immediate precipitation of migratory iron and uranyl in the silicate rich near-field environment. Use of the Fe(III) and UO22+ complexing agent acetohydroxamic acid (AHA) as a colorimetric agent for visible spectrophotometry is well-known. Using the second derivative of these spectra a distinct shift in iron complexation in the presence of silicate is seen that is not seen with uranyl or alone. Silica also decreases the ability of uranyl and ferric solutions to absorb hydroxide, hastening precipitation. These ferric silicate precipitates are highly amorphous and soluble. Precipitates formed in the presence of uranyl below ~1 mol% exhibit lower solubility than precipitates from up to 50 mol % and of uranyl silicates alone.

  13. Part removal of 3D printed parts

    E-Print Network [OSTI]

    Peña Doll, Mateo

    2014-01-01

    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 ...

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

    SciTech Connect (OSTI)

    Heaven, Michael C.; Gibson, John K.; Marcalo, Joaquim

    2009-02-01

    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

  15. Final Project Report for ER15351 “A Study of New Actinide Zintl Ion Materials”

    SciTech Connect (OSTI)

    Peter K. Dorhout

    2007-11-12

    The structural chemistry of actinide main-group metal materials provides the fundamental basis for the understanding of structural coordination chemistry and the formation of materials with desired or predicted structural features. The main-group metal building blocks, comprising sulfur-group, phosphorous-group, or silicon-group elements, have shown versatility in oxidation state, coordination, and bonding preferences. These building blocks have allowed us to elucidate a series of structures that are unique to the actinide elements, although we can find structural relationships to transition metal and 4f-element materials. In the past year, we investigated controlled metathesis and self-propagating reactions between actinide metal halides and alkali metal salts of main-group metal chalcogenides such as K-P-S salts. Ternary plutonium thiophosphates have resulted from these reactions at low temperature in sealed ampules. we have also focused efforts to examine reactions of Th, U, and Pu halide salts with other alkali metal salts such as Na-Ge-S and Na-Si-Se and copper chloride to identify if self-propagating reactions may be used as a viable reaction to prepare new actinide materials and we prepared a series of U and Th copper chalcogenide materials. Magnetic measurements continued to be a focus of actinide materials prepared in our laboratory. We also contributed to the XANES work at Los Alamos by preparing materials for study and for comparison with environmental samples.

  16. 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

    2013-10-01

    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.

  17. Management of salt waste from electrochemical processing of used nuclear fuel

    SciTech Connect (OSTI)

    Simpson, M.F.; Patterson, M.N.; Lee, J.; Wang, Y.; Versey, J.; Phongikaroon, S.

    2013-07-01

    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)

  18. Technical feasibility of the Diamex process

    SciTech Connect (OSTI)

    Sorel, C.; Montuir, M.; Espinoux, D.; Lorrain, B.; Baron, P.

    2008-07-01

    The DIAMEX process was developed at the CEA as the first of the two-step strategy to separate trivalent actinides from trivalent lanthanides. It consists of co-extracting trivalent actinides and lanthanides using a diamide extractant, dimethyl-dioctyl-hexyl-ethoxy-malonamide (DMDOHEMA). To demonstrate the technical feasibility of this process, a test has been successfully carried out with a PUREX raffinate solution in pulsed columns and mixer settlers, which are representative of the solvent-extraction contactors that could be used at the industrial scale. At the end of the trial, the americium and curium recovery yield exceeded 99.9% with high decontamination factors. (authors)

  19. Production Pathways and Separation Procedures for High-Diagnostic-Value Activation Species, Fission Products, and Actinides Required for Preparation of Realistic Synthetic Post-Detonation Nuclear Debris

    SciTech Connect (OSTI)

    Faye, S A; Shaughnessy, D A

    2015-08-19

    The objective of this project is to provide a comprehensive study on the production routes and chemical separation requirements for activation products, fission products, and actinides required for the creation of realistic post-detonation surrogate debris. Isotopes that have been prioritized by debris diagnosticians will be examined for their ability to be produced at existing irradiation sources, production rates, and availability of target materials, and chemical separation procedures required to rapidly remove the products from the bulk target matrix for subsequent addition into synthetic debris samples. The characteristics and implications of the irradiation facilities on the isotopes of interest will be addressed in addition to a summary of the isotopes that are already regularly produced.

  20. Enhancing BWR Proliferation Resistance Fuel with Minor Actinides

    SciTech Connect (OSTI)

    Gray S. Chang

    2008-07-01

    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.

  1. Minor Actinides Loading Optimization for Proliferation Resistant Fuel Design - BWR

    SciTech Connect (OSTI)

    G. S. Chang; Hongbin Zhang

    2009-09-01

    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.

  2. Alternatives to Nitric Acid Stripping in the Caustic-Side Solvent Extraction (CSSX) Process for Cesium Removal from Alkaline High-Level Waste

    SciTech Connect (OSTI)

    Delmau, Laetitia Helene; Haverlock, Tamara; Bazelaire, Eve; Bonnesen, Peter V; Ditto, Mary E; Moyer, Bruce A

    2009-01-01

    Effective alternatives to nitric acid stripping in the Caustic-Side Solvent Extraction (CSSX) solvent have been demonstrated in this work. The CSSX solvent employs calix[4]arene-bis(tert-octylbenzo-18-crown-6) (BOBCalixC6) as the cesium extractant in a modified alkane diluent for decontamination of alkaline high-level wastes. Results reported in this paper support the idea that replacement of the nitrate anion by a much more hydrophilic anion like borate can substantially lower cesium distribution ratios on stripping. Without any other change in the CSSX flowsheet, however, the use of a boric acid stripping solution in place of the 1 mM nitric acid solution used in the CSSX process marginally, though perhaps still usefully, improves stripping. The less-than-expected improvement was explained by the carryover of nitrate from scrubbing into stripping. Accordingly, more effective stripping is obtained after a scrub of the solvent with 0.1 M sodium hydroxide. Functional alternatives to boric acid include sodium bicarbonate or cesium hydroxide as strip solutions. Profound stripping improvement is achieved when trioctylamine, one of the components of the CSSX solvent, is replaced with a commercial guanidine reagent (LIX 79). The more basic guanidine affords greater latitude in selection of aqueous conditions in that it protonates even at mildly alkaline pH values. Under process-relevant conditions, cesium distributions on stripping are decreased on the order of 100-fold compared with current CSSX performance. The extraction properties of the solvent were preserved unchanged over three successive extract-scrub-strip cycles. From the point of view of compatibility with downstream processing, boric acid represents an attractive stripping agent, as it is also a potentially ideal feed for borosilicate vitrification of the separated 137Cs product stream. Possibilities for use of these results toward a dramatically better next-generation CSSX process, possibly one employing the more soluble cesium extractant calix[4]arene-bis(2 ethylhexylbenzo-18-crown-6) (BEHBCalixC6) are discussed.

  3. Emergence of californium as the second transitional element in the actinide series

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

    Cary, Samantha K.; Vasiliu, Monica; Baumbach, Ryan E.; Stritzinger, Jared T.; Green, Thomas D.; Diefenbach, Kariem; Cross, Justin N.; Knappenberger, Kenneth L.; Liu, Guokui; Silver, Mark A.; et al

    2015-04-16

    A break in periodicity occurs in the actinide series between plutonium and americium as the result of the localization of 5f electrons. The subsequent chemistry of later actinides is thought to closely parallel lanthanides in that bonding is expected to be ionic and complexation should not substantially alter the electronic structure of the metal ions. Here we demonstrate that ligation of californium(III) by a pyridine derivative results in significant deviations in the properties of the resultant complex with respect to that predicted for the free ion. We expand on this by characterizing the americium and curium analogues for comparison, andmore »show that these pronounced effects result from a second transition in periodicity in the actinide series that occurs, in part, because of the stabilization of the divalent oxidation state. As a result, the metastability of californium(II) is responsible for many of the unusual properties of californium including the green photoluminescence.« less

  4. Emergence of californium as the second transitional element in the actinide series

    SciTech Connect (OSTI)

    Cary, Samantha K.; Vasiliu, Monica; Baumbach, Ryan E.; Stritzinger, Jared T.; Green, Thomas D.; Diefenbach, Kariem; Cross, Justin N.; Knappenberger, Kenneth L.; Liu, Guokui; Silver, Mark A.; DePrince, A. Eugene; Polinski, Matthew J.; Van Cleve, Shelley M.; House, Jane H.; Kikugawa, Naoki; Gallagher, Andrew; Arico, Alexandra A.; Dixon, David A.; Albrecht-Schmitt, Thomas E.

    2015-04-16

    A break in periodicity occurs in the actinide series between plutonium and americium as the result of the localization of 5f electrons. The subsequent chemistry of later actinides is thought to closely parallel lanthanides in that bonding is expected to be ionic and complexation should not substantially alter the electronic structure of the metal ions. Here we demonstrate that ligation of californium(III) by a pyridine derivative results in significant deviations in the properties of the resultant complex with respect to that predicted for the free ion. We expand on this by characterizing the americium and curium analogues for comparison, and show that these pronounced effects result from a second transition in periodicity in the actinide series that occurs, in part, because of the stabilization of the divalent oxidation state. As a result, the metastability of californium(II) is responsible for many of the unusual properties of californium including the green photoluminescence.

  5. Surface acoustic wave sensors/gas chromatography; and Low quality natural gas sulfur removal and recovery CNG Claus sulfur recovery process

    SciTech Connect (OSTI)

    Klint, B.W.; Dale, P.R.; Stephenson, C.

    1997-12-01

    This topical report consists of the two titled projects. Surface Acoustic Wave/Gas Chromatography (SAW/GC) provides a cost-effective system for collecting real-time field screening data for characterization of vapor streams contaminated with volatile organic compounds (VOCs). The Model 4100 can be used in a field screening mode to produce chromatograms in 10 seconds. This capability will allow a project manager to make immediate decisions and to avoid the long delays and high costs associated with analysis by off-site analytical laboratories. The Model 4100 is currently under evaluation by the California Environmental Protection Agency Technology Certification Program. Initial certification focuses upon the following organics: cis-dichloroethylene, chloroform, carbon tetrachloride, trichlorethylene, tetrachloroethylene, tetrachloroethane, benzene, ethylbenzene, toluene, and o-xylene. In the second study the CNG Claus process is being evaluated for conversion and recovery of elemental sulfur from hydrogen sulfide, especially found in low quality natural gas. This report describes the design, construction and operation of a pilot scale plant built to demonstrate the technical feasibility of the integrated CNG Claus process.

  6. 3-D Characterization of the Structure of Paper and Paperboard and Their Application to Optimize Drying and Water Removal Processes and End-Use Applications

    SciTech Connect (OSTI)

    Shri Ramaswamy, University of Minnesota; B.V. Ramarao, State University of New York

    2004-08-29

    The three dimensional structure of paper materials plays a critical role in the paper manufacturing process especially via its impact on the transport properties for fluids. Dewatering of the wet web, pressing and drying will benefit from knowledge of the relationships between the web structure and its transport coefficients. The structure of the pore space within a paper sheet is imaged in serial sections using x-ray micro computed tomography. The three dimensional structure is reconstructed from these sections using digital image processing techniques. The structure is then analyzed by measuring traditional descriptors for the pore space such as specific surface area and porosity. A sequence of microtomographs was imaged at approximately 2 ?m intervals and the three-dimensional pore-fiber structure was reconstructed. The pore size distributions for both in-plane as well as transverse pores were measured. Significant differences in the in-plane (XY) and the transverse directions in pore characteristics are found and may help partly explain the different liquid and vapor transport properties in the in-plane and transverse directions. Results with varying sheet structures compare favorably with conventional mercury intrusion porosimetry data. Interestingly, the transverse pore structure appears to be more open with larger pore size distribution compared to the in plane pore structure. This may help explain the differences in liquid and vapor transport through the in plane and transverse structures during the paper manufacturing process and during end-use application. Comparison of Z-directional structural details of hand sheet and commercially made fine paper samples show a distinct difference in pore size distribution both in the in-plane and transverse direction. Method presented here may provide a useful tool to the papermaker to truly engineer the structure of paper and board tailored to specific end-use applications. The difference in surface structure between the top and bottom sides of the porous material, i.e. "two-sidedness" due to processing and raw material characteristics may lead to differences in end-use performance. The measurements of surface structure characteristics include thickness distribution, surface volume distribution, contact fraction distribution and surface pit distribution. This complements our earlier method to analyze the bulk structure and Z-D structure of porous materials. As one would expect, the surface structure characteristics will be critically dependent on the quality and resolution of the images. This presents a useful tool to characterize and engineer the surface structure of porous materials such as paper and board tailored to specific end-use applications. This will also help troubleshoot problems related to manufacturing and end-use applications. This study attempted to identify the optimal resolution through a comparison between 3D images obtained by monochromatic synchrotron radiation X-?CT in phase contrast mode (resolution ? 1 ?m) and polychromatic radiation X-?CT in absorption mode (res. ? 5 ?m). It was found that both resolutions have the ability to show the expected trends when comparing different paper samples. The low resolution technique shows fewer details resulting in lower specific surface area, larger pore channels, characterized as hydraulic radii, and lower tortuosities, where differences between samples and principal directions are more difficult to detect. The disadvantages of the high resolution images are high cost and limited availability of hard x-ray beam time as well as the small size of the sample volumes imaged. The results show that the low resolution images can be used for comparative studies, whereas the high resolution images may be better suited for fundamental research on the paper structure and its influence on paper properties, as one gets more accurate physical measurements. In addition, pore space diffusion model has been developed to simulate simultaneous diffusion in heterogeneous porous materials such as paper containing cellu

  7. EGR Cooler Fouling- Visualization of Deposition and Removal Mechanis

    Broader source: Energy.gov [DOE]

    Presents experimental data on exhaust gas recirculation(EGR) cooler fouling using new test apparatus that allows for in-situ observation of deposition and removal processes

  8. Analysis on fuel breeding capability of FBR core region based on minor actinide recycling doping

    SciTech Connect (OSTI)

    Permana, Sidik; Novitrian,; Waris, Abdul; Ismail; Suzuki, Mitsutoshi; Saito, Masaki

    2014-09-30

    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.

  9. 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)

    2012-07-01

    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)

  10. Fluorination process using catalyst

    DOE Patents [OSTI]

    Hochel, Robert C. (Aiken, SC); Saturday, Kathy A. (Aiken, SC)

    1985-01-01

    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.

  11. Fluorination process using catalysts

    DOE Patents [OSTI]

    Hochel, R.C.; Saturday, K.A.

    1983-08-25

    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.

  12. Reactor for removing ammonia

    DOE Patents [OSTI]

    Luo, Weifang (Livermore, CA); Stewart, Kenneth D. (Valley Springs, CA)

    2009-11-17

    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.

  13. Ion Irradiation Effects in Synthetic Garnets Incorporating Actinides Satoshi Utsunomiya1

    E-Print Network [OSTI]

    Utsunomiya, Satoshi

    Ion Irradiation Effects in Synthetic Garnets Incorporating Actinides Satoshi Utsunomiya1 , Lu.0 MeV Kr2+ irradiation with in situ transmission electron microscopy over the temperature range of 50 the long term radiation effects due to radioactive decay can be simulated in short term with heavy ion-irradiation

  14. Microscopic Description of Nuclear Fission: Fission Barrier Heights of Even-Even Actinides

    E-Print Network [OSTI]

    J. McDonnell; N. Schunck; W. Nazarewicz

    2013-01-31

    We evaluate the performance of modern nuclear energy density functionals for predicting inner and outer fission barrier heights and energies of fission isomers of even-even actinides. For isomer energies and outer barrier heights, we find that the self-consistent theory at the HFB level is capable of providing quantitative agreement with empirical data.

  15. 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.

    2013-07-01

    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)

  16. Method for forming an extraction agent for the separation of actinides from lanthanides

    DOE Patents [OSTI]

    Klaehn, John R.; Harrup, Mason K.; Law, Jack D.; Peterman, Dean R.

    2010-04-27

    An extraction agent for the separation of trivalent actinides from lanthanides in an acidic media and a method for forming same are described, and wherein the methodology produces a stable regiospecific and/or stereospecific dithiophosphinic acid that can operate in an acidic media having a pH of less than about 7.

  17. Performance of the Lead-Alloy Cooled Concept Balanced for Actinide Burning and Electricity Production

    SciTech Connect (OSTI)

    Pavel Hejzlar; Cliff Davis

    2004-09-01

    A lead-bismuth-cooled fast reactor concept targeted for a balanced mission of actinide burning and low-cost electricity production is proposed and its performance analyzed. The design explores the potential benefits of thorium-based fuel in actinide-burning cores, in particular in terms of the reduction of the large reactivity swing and enhancement of the small Doppler coefficient typical of fertile-free actinide burners. Reduced electricity production cost is pursued through a longer cycle length than that used for fertile-free burners and thus a higher capacity factor. It is shown that the concept can achieve a high transuranics destruction rate, which is only 20% lower than that of an accelerator-driven system with fertile-free fuel. The small negative fuel temperature reactivity coefficient, small positive coolant temperature reactivity coefficient, and negative core radial expansion coefficient provide self-regulating characteristics so that the reactor is capable of inherent shutdown during major transients without scram, as in the Integral Fast Reactor. This is confirmed by thermal-hydraulic analysis of several transients without scram, including primary coolant pump trip, station blackout, and reactivity step insertion, which showed that the reactor was able to meet all identified thermal limits. However, the benefits of high actinide consumption and small reactivity swing can be attained only if the uranium from the discharged fuel is separated and not recycled. This additional uranium separation step and thorium reprocessing significantly increase the fuel cycle costs. Because the higher fuel cycle cost has a larger impact on the overall cost of electricity than the savings from the higher capacity factor afforded through use of thorium, this concept appears less promising than the fertile-free actinide burners.

  18. Performance of the Lead-Alloy-Cooled Reactor Concept Balanced for Actinide Burning and Electricity Production

    SciTech Connect (OSTI)

    Hejzlar, Pavel [Massachusetts Institute of Technology (United States); Davis, Cliff B. [Idaho National Engineering and Environmental Laboratory (United States)

    2004-09-15

    A lead-bismuth-cooled fast reactor concept targeted for a balanced mission of actinide burning and low-cost electricity production is proposed and its performance analyzed. The design explores the potential benefits of thorium-based fuel in actinide-burning cores, in particular in terms of the reduction of the large reactivity swing and enhancement of the small Doppler coefficient typical of fertile-free actinide burners. Reduced electricity production cost is pursued through a longer cycle length than that used for fertile-free burners and thus a higher capacity factor. It is shown that the concept can achieve a high transuranics destruction rate, which is only 20% lower than that of an accelerator-driven system with fertile-free fuel. The small negative fuel temperature reactivity coefficient, small positive coolant temperature reactivity coefficient, and negative core radial expansion coefficient provide self-regulating characteristics so that the reactor is capable of inherent shutdown during major transients without scram, as in the Integral Fast Reactor. This is confirmed by thermal-hydraulic analysis of several transients without scram, including primary coolant pump trip, station blackout, and reactivity step insertion, which showed that the reactor was able to meet all identified thermal limits. However, the benefits of high actinide consumption and small reactivity swing can be attained only if the uranium from the discharged fuel is separated and not recycled. This additional uranium separation step and thorium reprocessing significantly increase the fuel cycle costs. Because the higher fuel cycle cost has a larger impact on the overall cost of electricity than the savings from the higher capacity factor afforded through use of thorium, this concept appears less promising than the fertile-free actinide burners.

  19. Method for recovery of actinides from refractory oxides thereof using O.sub. F.sub.2

    DOE Patents [OSTI]

    Asprey, Larned B. (Los Alamos, NM); Eller, Phillip G. (Los Alamos, NM)

    1988-01-01

    Method for recovery of actinides from nuclear waste material containing sintered and other oxides thereof 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 destroyed by high temperature reactions with the walls of the reaction vessel. Dioxygen difluoride is readily prepared, stored and transferred to the place of reaction.

  20. Economic feasibility of biochemical processes for the upgrading of crudes and the removal of sulfur, nitrogen, and trace metals from crude oil -- Benchmark cost establishment of biochemical processes on the basis of conventional downstream technologies. Final report FY95

    SciTech Connect (OSTI)

    Premuzic, E.T.

    1996-08-01

    During the past several years, a considerable amount of work has been carried out showing that microbially enhanced oil recovery (MEOR) is promising and the resulting biotechnology may be deliverable. At Brookhaven National Laboratory (BNL), systematic studies have been conducted which dealt with the effects of thermophilic and thermoadapted bacteria on the chemical and physical properties of selected types of crude oils at elevated temperatures and pressures. Current studies indicate that during the biotreatment several chemical and physical properties of crude oils are affected. The oils are (1) emulsified; (2) acidified; (3) there is a qualitative and quantitative change in light and heavy fractions of the crudes; (4) there are chemical changes in fractions containing sulfur compounds; (5) there is an apparent reduction in the concentration of trace metals; and (6) the qualitative and quantitative changes appear to be microbial species dependent; and (7) there is a distinction between biodegraded and biotreated oils. The downstream biotechnological crude oil processing research performed thus far is of laboratory scale and has focused on demonstrating the technical feasibility of downstream processing with different types of biocatalysts under a variety of processing conditions. Quantitative economic analysis is the topic of the present project which investigates the economic feasibility of the various biochemical downstream processes which hold promise in upgrading of heavy crudes, such as those found in California, e.g., Monterey-type, Midway Sunset, Honda crudes, and others.

  1. Identification of process suitable diluent

    SciTech Connect (OSTI)

    Dean R. Peterman

    2014-01-01

    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).

  2. Heat recirculating cooler for fluid stream pollutant removal

    DOE Patents [OSTI]

    Richards, George A. (Morgantown, WV); Berry, David A. (Morgantown, WV)

    2008-10-28

    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.

  3. FY13 GLYCOLIC-NITRIC ACID FLOWSHEET DEMONSTRATIONS OF THE DWPF CHEMICAL PROCESS CELL WITH SIMULANTS

    SciTech Connect (OSTI)

    Lambert, D.; Zamecnik, J.; Best, D.

    2014-03-13

    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.

  4. Technetium Removal Using Tc-Goethite Coprecipitation

    SciTech Connect (OSTI)

    Um, Wooyong; Wang, Guohui; Jung, Hun Bok; Peterson, Reid A.

    2013-11-18

    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.

  5. Electrochemical/Pyrometallurgical Waste Stream Processing and Waste Form Fabrication

    SciTech Connect (OSTI)

    Steven Frank; Hwan Seo Park; Yung Zun Cho; William Ebert; Brian Riley

    2015-07-01

    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.

  6. Arsenic removal from water

    DOE Patents [OSTI]

    Moore, Robert C. (Edgewood, NM); Anderson, D. Richard (Albuquerque, NM)

    2007-07-24

    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.

  7. Removable feedwater sparger assembly

    DOE Patents [OSTI]

    Challberg, R.C.

    1994-10-04

    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.

  8. Drum lid removal tool

    DOE Patents [OSTI]

    Pella, Bernard M. (Martinez, GA); Smith, Philip D. (North Augusta, SC)

    2010-08-24

    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.

  9. Condensate removal device

    DOE Patents [OSTI]

    Maddox, James W. (Newport News, VA); Berger, David D. (Alexandria, VA)

    1984-01-01

    A condensate removal device is disclosed which incorporates a strainer in unit with an orifice. The strainer is cylindrical with its longitudinal axis transverse to that of the vapor conduit in which it is mounted. The orifice is positioned inside the strainer proximate the end which is remoter from the vapor conduit.

  10. 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.

    2013-04-15

    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.

  11. SCALING SOLID RESUSPENSION AND SORPTION FOR THE SMALL COLUMN ION EXCHANGE PROCESSING TANK

    SciTech Connect (OSTI)

    Poirier, M.; Qureshi, Z.

    2010-12-14

    The Small Column Ion Exchange (SCIX) process is being developed to remove cesium, strontium, and actinides from Savannah River Site (SRS) Liquid Waste using an existing 1.3 million gallon 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 and resuspending Monosodium Titanate (MST), Crystalline Silicotitanate (CST), and simulated sludge. In addition, SRNL will also be conducting pilot-scale tests to determine the mixing requirements for the strontium and actinide sorption. As part of this task, the results from the pilot-scale tests must be scaled up to a full-scale waste tank. This document describes the scaling approach. The pilot-scale tank is a 1/10.85 linear scale model of 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). MST additions are through Riser E1, the proposed MST addition riser in Tank 41H. To determine the approach to scaling the results from the pilot-scale tank to Tank 41H, the authors took the following approach. They reviewed the technical literature for methods to scale mixing with jets and suspension of solid particles with jets, and the technical literature on mass transfer from a liquid to a solid particle to develop approaches to scaling the test data. SRNL assembled a team of internal experts to review the scaling approach and to identify alternative approaches that should be considered.

  12. Evaluation of Fluid Conduction and Mixing within a Subassembly of the Actinide Burner Test Reactor

    SciTech Connect (OSTI)

    Cliff B. Davis

    2007-09-01

    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.

  13. High-burnup core design using minor actinide-containing metal fuel

    SciTech Connect (OSTI)

    Ohta, Hirokazu; Ogata, Takanari; Obara, T.

    2013-07-01

    A neutronic design study of metal fuel fast reactor (FR) cores is conducted on the basis of an innovative fuel design concept to achieve an extremely high burnup and realize an efficient fuel cycle system. Since it is expected that the burnup reactivity swing will become extremely large in an unprecedented high burnup core, minor actinides (MAs) from light water reactors (LWRs) are added to fresh fuel to improve the core internal conversion. Core neutronic analysis revealed that high burnups of about 200 MWd/kg for a small-scale core and about 300 MWd/kg for a large-scale core can be attained while suppressing the burnup reactivity swing to almost the same level as that of conventional cores with normal burnup. An actinide burnup analysis has shown that the MA consumption ratio is improved to about 60% and that the accumulated MAs originating from LWRs can be efficiently consumed by the high-burnup metal fuel FR. (authors)

  14. Extraction of actinides by multi-dentate diamides and their evaluation with computational molecular modeling

    SciTech Connect (OSTI)

    Sasaki, Y.; Kitatsuji, Y.; Hirata, M.; Kimura, T.; Yoshizuka, K.

    2008-07-01

    Multi-dentate diamides have been synthesized and examined for actinide (An) extractions. Bi- and tridentate extractants are the focus in this work. The extraction of actinides was performed from 0.1-6 M HNO{sub 3} to organic solvents. It was obvious that N,N,N',N'-tetra-alkyl-diglycolamide (DGA) derivatives, 2,2'-(methylimino)bis(N,N-dioctyl-acetamide) (MIDOA), and N,N'-dimethyl-N,N'-dioctyl-2-(3-oxa-pentadecane)-malonamide (DMDOOPDMA) have relatively high D values (D(Pu) > 70). The following notable results using DGA extractants were obtained: (1) DGAs with short alkyl chains give higher D values than those with long alkyl chain, (2) DGAs with long alkyl chain have high solubility in n-dodecane. Computational molecular modeling was also used to elucidate the effects of structural and electronic properties of the reagents on their different extractabilities. (authors)

  15. Electrodeposition of actinide compounds from an aqueous ammonium acetate matrix: Experimental development and optimization

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

    Boll, Rose Ann; Matos, Milan; Torrico, Matthew N.

    2015-03-27

    Electrodeposition is a technique that is routinely employed in nuclear research for the preparation of thin solid films of actinide materials which can be used in accelerator beam bombardments, irradiation studies, or as radioactive sources. The present study investigates the deposition of both lanthanides and actinides from an aqueous ammonium acetate electrolyte matrix. Electrodepositions were performed primarily on stainless steel disks; with yield analysis evaluated using -spectroscopy. Experimental parameters were studied and modified in order to optimize the uniformity and adherence of the deposition while maximizing the yield. The initial development utilized samarium as the plating material, with and withoutmore »a radioactive tracer. As a result, surface characterization studies were performed by scanning electron microscopy, electron microprobe analysis, radiographic imaging, and x-ray diffraction.« less

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

    E-Print Network [OSTI]

    Ames, David E, II

    2006-10-30

    : Chair of Committee, Pavel V. Tsvetkov Committee Members, Kenneth L. Peddicord Donald G. Allen Head of Department, William E. Burchill August 2006 Major Subject: Nuclear Engineering iii ABSTRACT Configuration.... The project assesses the prospective relation and application for Generation IV VHTRs in the AFCI Program [2, 3]. Utilization of minor actinides, from light water reactor (LWR) fuel, as a fuel component, would facilitate development of new fuel cycles...

  17. 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

    2013-10-01

    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.

  18. Fundamental Thermodynamics of Actinide-Bearing Mineral Waste Forms - Final Report

    SciTech Connect (OSTI)

    Williamson, Mark A.; Ebbinghaus, Bartley B.; Navrotsky, Alexandra

    2001-03-01

    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.

  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)

    2013-07-01

    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. Pneumatic soil removal tool

    DOE Patents [OSTI]

    Neuhaus, John E. (Newport News, VA)

    1992-01-01

    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.

  1. Pneumatic soil removal tool

    DOE Patents [OSTI]

    Neuhaus, J.E.

    1992-10-13

    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.

  2. Melter Glass Removal and Dismantlement

    SciTech Connect (OSTI)

    Richardson, BS

    2000-10-31

    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).

  3. Removal of Sarin Aerosol and Vapor by Water Sprays

    SciTech Connect (OSTI)

    Brockmann, John E.

    1998-09-01

    Falling water drops can collect particles and soluble or reactive vapor from the gas through which they fall. Rain is known to remove particles and vapors by the process of rainout. Water sprays can be used to remove radioactive aerosol from the atmosphere of a nuclear reactor containment building. There is a potential for water sprays to be used as a mitigation technique to remove chemical or bio- logical agents from the air. This paper is a quick-look at water spray removal. It is not definitive but rather provides a reasonable basic model for particle and gas removal and presents an example calcu- lation of sarin removal from a BART station. This work ~ a starting point and the results indicate that further modeling and exploration of additional mechanisms for particle and vapor removal may prove beneficial.

  4. Quick Stain Removal Guide 

    E-Print Network [OSTI]

    Brown, Pamela J.

    1998-07-29

    and dryer. ? Use the hottest wash temperature recommended on the garment care label. ? Use the right wash and dry cycle for the fabric. ? Avoid using chlorine bleach on silk, wool or spandex or on non-colorfast garments. ? Remove clothes... color-safe bleach or fabric softener. Bleaches are strong cleaning agents. Follow the in- structions on the containers. Use chlorine bleach on colorfast fabrics; it is most effective if added 5 to 6 minutes into the wash cycle. It is best...

  5. Thiacrown polymers for removal of mercury from waste streams

    DOE Patents [OSTI]

    Baumann, Theodore F.; Reynolds, John G.; Fox, Glenn A.

    2004-02-24

    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.

  6. 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)

    2002-01-01

    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.

  7. Sorbents for mercury removal from flue gas

    SciTech Connect (OSTI)

    Granite, Evan J.; Hargis, Richard A.; Pennline, Henry W.

    1998-01-01

    A review of the various promoters and sorbents examined for the removal of mercury from flue gas is presented. Commercial sorbent processes are described along with the chemistry of the various sorbent-mercury interactions. Novel sorbents for removing mercury from flue gas are suggested. Since activated carbons are expensive, alternate sorbents and/or improved activated carbons are needed. Because of their lower cost, sorbent development work can focus on base metal oxides and halides. Additionally, the long-term sequestration of the mercury on the sorbent needs to be addressed. Contacting methods between the flue gas and the sorbent also merit investigation.

  8. Electrochemically assisted paint removal

    SciTech Connect (OSTI)

    Keller, R.; Hydock, D.M.; Burleigh, T.D.

    1995-12-31

    A method to remove paint coatings from metal and other electronically conductive substrates is being studied. In particular, the remediation of objects coated with lead based paints is the focus of research. The approach also works very well with automotive coatings and may be competitive with sandblasting. To achieve debonding of the coating, the deteriorated or artifically damaged surface of the object is cathodically polarized. The object can be immersed in a benign aqueous electrolyte for treatment, or the electrolyte can be retained in an absorbent pad covering the surface to be treated.

  9. 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

    2009-05-20

    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.

  10. Method for removing chlorine compounds from hydrocarbon mixtures

    DOE Patents [OSTI]

    Janoski, E.J.; Hollstein, E.J.

    1984-09-29

    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.

  11. Optimized procedures for extractioin, purification and characterization of exopolymeric substances (eps) from two bacteria (sagittula stellata and pseudomonas fluorescens biovar ii) with relevance to the study of actinide binding in aquatic environments 

    E-Print Network [OSTI]

    Xu, Chen

    2009-05-15

    Flats Environmental Technology Site (RFETS) soil organic colloid spiked with Th showed similar activity distributions of both actinides along the pH gradient, with the activities of both actinides focusing on the low pH region. Characterizations...

  12. Synthesis and Optimization of the Sintering Kinetics of Actinide Nitrides

    SciTech Connect (OSTI)

    Drryl P. Butt; Brian Jaques

    2009-03-31

    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.

  13. 1. 7084 " The Carbon Matrices Made of Pyrolised Bis-Phtalocyanines as a Base for Encapsulation of the Long-Lived Nuclides of Iodine, Technetium and Minor Actinides" Valeri

    E-Print Network [OSTI]

    Titov, Anatoly

    for Encapsulation of the Long-Lived Nuclides of Iodine, Technetium and Minor Actinides" Valeri Tichonov, Pavel

  14. Detection limits for actinides in a monochromatic, wavelength-dispersive x-ray fluorescence instrument

    SciTech Connect (OSTI)

    Collins, Michael L [Los Alamos National Laboratory; Havrilla, George J [Los Alamos National Laboratory

    2009-01-01

    Recent developments in x-ray optics have made it possible to examine the L x-rays of actinides using doubly-curved crystals in a bench-top device. A doubly-curved crystal (DCC) acts as a focusing monochromatic filter for polychromatic x-rays. A Monochromatic, Wavelength-Dispersive X-Ray Fluorescence (MWDXRF) instrument that uses DCCs to measure Cm and Pu in reprocessing plant liquors was proposed in 2007 by the authors at Los Alamos National Laboratory. A prototype design of this MWDXRF instrument was developed in collaboration with X-ray Optical Systems Inc. (XOS), of East Greenbush, New York. In the MWDXRF instrument, x-rays from a Rhodium-anode x-ray tube are passed through a primary DCC to produce a monochromatic beam of 20.2-keV photons. This beam is focused on a specimen that may contain actinides. The 20.2-keV interrogating beam is just above the L3 edge of Californium; each actinide (with Z = 90 to 98) present in the specimen emits characteristic L x-rays as the result of L3-shell vacancies. In the LANL-XOS prototype MWDXRf, these x-rays enter a secondary DCC optic that preferentially passes 14.961-keV photons, corresponding to the L-alpha-1 x-ray peak of Curium. In the present stage of experimentation, Curium-bearing specimens have not been analyzed with the prototype MWDXRF instrument. Surrogate materials for Curium include Rubidium, which has a K-beta-l x-ray at 14.961 keV, and Yttrium, which has a K-alpha-1 x-ray at 14.958 keV. In this paper, the lower limit of detection for Curium in the LANL-XOS prototype MWDXRF instrument is estimated. The basis for this estimate is described, including a description of computational models and benchmarking techniques used. Detection limits for other actinides are considered, as well as future safeguards applications for MWDXRF instrumentation.

  15. 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.

    2008-07-01

    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)

  16. Lanthanide and actinide doped glasses as reference standards for dye doped systems

    SciTech Connect (OSTI)

    Pope, E.J.A.; Hentschel, A.

    1996-12-31

    Organic dye molecules are well known to be subject to chemical and optical bleaching damage, temperature instability, and other forms of optical degradation. Currently recognized methods of referencing rely upon fluorescent salt solutions, such as quinine sulfate. In this paper, optically-active lanthanide and actinide doped gel-glasses are compared as reference standards for dye doped polymers. Samples are subjected to continuous illumination by 254 nm UV radiation. While dye-doped polymers exhibited approximately 65 percent decline in fluorescence intensity after 96 hours of irradiation, glass samples and glass powder in resin showed no decline in fluorescence intensities.

  17. Micro-Analysis of Actinide Minerals for Nuclear Forensics and Treaty Verification

    SciTech Connect (OSTI)

    M. Morey, M. Manard, R. Russo, G. Havrilla

    2012-03-22

    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.

  18. Magnetocrystalline Anisotropy in UMn2Ge2 and Related Mn-based Actinide Ferromagnets

    SciTech Connect (OSTI)

    Parker, David S; Mandrus, D.; Ghimire, N J; Baumbach, Ryan; Singleton, John; Thompson, J.D.; Bauer, Eric D.; Li, Ling; Singh, David J

    2015-01-01

    We present magnetization isotherms in pulsed magnetic fields up to 62 Tesla, supported by first principles calculations, demonstrating a huge uniaxial magnetocrystalline anisotropy energy - approximately 20 MJ/m3 - in UMn2Ge2. This large anisotropy results from the extremely strong spin-orbit coupling affecting the uranium 5 f electrons, which in the calculations exhibit a substantial orbital moment exceeding 2 Bohr magnetons. We also find from theoretical calculations that a number of isostructural Mn-actinide compounds are expected to have similarly large anisotropy.

  19. Method to remove uranium/vanadium contamination from groundwater

    DOE Patents [OSTI]

    Metzler, Donald R. (DeBeque, CO); Morrison, Stanley (Grand Junction, CO)

    2004-07-27

    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.

  20. Method to Remove Uranium/Vanadium Contamination from Groundwater

    DOE Patents [OSTI]

    Metzler, Donald R.; Morrison Stanley

    2004-07-27

    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.

  1. Removal of mercury from waste gases

    SciTech Connect (OSTI)

    Muster, U.; Marr, R.; Pichler, G.; Kremshofer, S.; Wilferl, R.; Draxler, J.

    1996-12-31

    Waste and process gases from thermal power, incineration and metallurgical plants or those from cement and alkali chloride industries contain metallic, inorganic and organic mercury. Widespread processes to remove the major amount of mercury are absorption and adsorption. Caused by the lowering of the emission limit from 200 to 50 {mu}g/m{sup 3} [STP] by national and European legislators, considerable efforts were made to enhance the efficiency of the main separation units of flue gas cleaning plants. Specially impregnated ceramic carriers can be used for the selective separation of metallic, inorganic and organic mercury. Using the ceramic reactor removal rates lower than 5 {mu}g/m{sup 3} [STP] of gaseous mercury and its compounds can be achieved. The ceramic reactor is active, regenerable and stable for a long term operation. 4 refs., 7 figs.

  2. Tank 241-CX-70 waste removal and packaging

    SciTech Connect (OSTI)

    DuVon, D.K.

    1993-06-01

    Tank 241-CX-70, located on the Hanford Site in Washington State, is a 30,000 gal single-shell storage tank built in 1952 to hold high-level process waste from pilot tests of the reduction-oxidation process. In 1979 decommissioning operations were begun by pumping liquid waste from the tank to the double-shell tank (DST) 101-AY. Not all the waste was removed at that time. Approximately 10,300 gal of sludge remained. On September 25, 1987, operations were resumed to remove the remaining waste using a sluicing and pumping method. This report documents the final removal of waste from Tank 241-CX-70.

  3. Tank 241-CX-70 waste removal and packaging

    SciTech Connect (OSTI)

    DuVon, D.K.

    1993-01-01

    Tank 241-CX-70, located on the Hanford Site in Washington State, is a 30,000 gal single-shell storage tank built in 1952 to hold high-level process waste from pilot tests of the reduction-oxidation process. In 1979 decommissioning operations were begun by pumping liquid waste from the tank to the double-shell tank (DST) 101-AY. Not all the waste was removed at that time. Approximately 10,300 gal of sludge remained. On September 25, 1987, operations were resumed to remove the remaining waste using a sluicing and pumping method. This report documents the final removal of waste from Tank 241-CX-70.

  4. Removing Arsenic from Drinking Water

    ScienceCinema (OSTI)

    None

    2013-05-28

    See how INL scientists are using nanotechnology to remove arsenic from drinking water. For more INL research, visit http://www.facebook.com/idahonationallaboratory

  5. Removing Arsenic from Drinking Water

    SciTech Connect (OSTI)

    2011-01-01

    See how INL scientists are using nanotechnology to remove arsenic from drinking water. For more INL research, visit http://www.facebook.com/idahonationallaboratory

  6. Removal of copper from ferrous scrap

    DOE Patents [OSTI]

    Blander, M.; Sinha, S.N.

    1987-07-30

    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.

  7. 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)

    1990-01-01

    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.

  8. Removal of copper from ferrous scrap

    DOE Patents [OSTI]

    Blander, M.; Sinha, S.N.

    1990-05-15

    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. 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.

    2011-01-25

    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.

  10. Conceptual design of minor actinides burner with an accelerator-driven subcritical system.

    SciTech Connect (OSTI)

    Cao, Y.; Gohar, Y. (Nuclear Engineering Division)

    2011-11-04

    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.

  11. 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)

    1995-01-01

    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.

  12. Method of preparation of removable syntactic foam

    DOE Patents [OSTI]

    Arnold, C. Jr.; Derzon, D.K.; Nelson, J.S.; Rand, P.B.

    1995-07-11

    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.

  13. Preparation of actinide specimens for the US/UK joint experiment in the Dounreay Prototype Fast Reactor

    SciTech Connect (OSTI)

    Quinby, T C; Adair, H L; Kobisk, E H

    1982-05-01

    A joint research program involving the United States and the United Kingdom was initiated about four years ago for the purpose of studying the fuel behavior of higher actinides using in-core irradiation in the fast reactor at Dounreay, Scotland. Simultaneously, determination of integral cross sections of a wide variety of higher actinide isotopes (physics specimens) was proposed. Coincidental neutron flux and energy spectral measurements were to be made using vanadium encapsulated dosimetry materials in the immediate region of the fuel pellets and physics samples. The higher actinide samples chosen for the fuel study were /sup 241/Am and /sup 244/Cm in the forms of Am/sub 2/O/sub 3/, Cm/sub 2/O/sub 3/, and Am/sub 6/Cm(RE)/sub 7/O/sub 21/, where (RE) represents a mixture of lanthanides. Milligram quantities of actinide oxides of /sup 248/Cm, /sup 246/Cm, /sup 244/Cm, /sup 243/Cm, /sup 243/Am, /sup 241/Am, /sup 244/Pu, /sup 242/Pu, /sup 241/Pu, /sup 240/Pu, /sup 239/Pu, /sup 238/Pu, /sup 237/Np, /sup 238/U, /sup 236/U, /sup 235/U, /sup 234/U, /sup 233/U, /sup 232/Th, /sup 230/Th, and /sup 231/Pa were encapsulated to obtain nuclear cross section and reaction rate data for these materials.

  14. The Thermodynamic Properties of the f-Elements and their Compounds. Part 2. The Lanthanide and Actinide Oxides

    SciTech Connect (OSTI)

    Konings, Rudy J. M. Beneš, Ondrej; Kovács, Attila; Manara, Dario; Sedmidubský, David; Gorokhov, Lev; Iorish, Vladimir S.; Yungman, Vladimir; Shenyavskaya, E.; Osina, E.

    2014-03-15

    A comprehensive review of the thermodynamic properties of the oxide compounds of the lanthanide and actinide elements is presented. The available literature data for the solid, liquid, and gaseous state have been analysed and recommended values are presented. In case experimental data are missing, estimates have been made based on the trends in the two series, which are extensively discussed.

  15. 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

    2013-12-01

    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.

  16. REMOVAL OF LEGACY PLUTONIUM MATERIALS FROM SWEDEN

    SciTech Connect (OSTI)

    Dunn, Kerry A.; Bellamy, J. Steve; Chandler, Greg T.; Iyer, Natraj C.; Koenig, Rich E.; Leduc, D.; Hackney, B.; Leduc, Dan R.

    2013-08-18

    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. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room News Publications Traditional Knowledge KiosksAbout AwardedAcronyms This is a quick guide|

  18. Metathesis depolymerization for removable surfactant templates.

    SciTech Connect (OSTI)

    Zifer, Thomas (Sandia National Laboratories, Livermore, CA); Wheeler, David Roger; Rahimian, Kamayar; McElhanon, James Ross (Sandia National Laboratories, Livermore, CA); Long, Timothy Michael; Jamison, Gregory Marks; Loy, Douglas Anson (Los Alamos National Laboratories, Los Alamos, NM); Kline, Steven R. (National Institute of Standards and Technology, Gaithersburg, MD); Simmons, Blake Alexander (Sandia National Laboratories, Livermore, CA)

    2005-03-01

    Current methodologies for the production of meso- and nanoporous materials include the use of a surfactant to produce a self-assembled template around which the material is formed. However, post-production surfactant removal often requires centrifugation, calcination, and/or solvent washing which can damage the initially formed material architecture(s). Surfactants that can be disassembled into easily removable fragments following material preparation would minimize processing damage to the material structure, facilitating formation of templated hybrid architectures. Herein, we describe the design and synthesis of novel cationic and anionic surfactants with regularly spaced unsaturation in their hydrophobic hydrocarbon tails and the first application of ring closing metathesis depolymerization to surfactant degradation resulting in the mild, facile decomposition of these new compounds to produce relatively volatile nonsurface active remnants.

  19. Minor Actinide Recycle in Sodium Cooled Fast Reactors Using Heterogeneous Targets

    SciTech Connect (OSTI)

    Samuel Bays; Pavel Medvedev; Michael Pope; Rodolfo Ferrer; Benoit Forget; Mehdi Asgari

    2009-04-01

    This paper investigates the plausible design of transmutation target assemblies for minor actinides (MA) in Sodium Fast Reactors (SFR). A heterogeneous recycling strategy is investigated, whereby after each reactor pass, un-burned MAs from the targets are blended with MAs produced by the driver fuel and additional MAs from Spent Nuclear Fuel (SNF). A design iteration methodology was adopted for customizing the core design, target assembly design and matrix composition design. The overall design was constrained against allowable peak or maximum in-core performances. While respecting these criteria, the overall design was adjusted to reduce the total number of assemblies fabricated per refueling cycle. It was found that an inert metal-hydride MA-Zr-Hx target matrix gave the highest transmutation efficiency, thus allowing for the least number of targets to be fabricated per reactor cycle.

  20. 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)

    1999-01-01

    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.

  1. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

    SciTech Connect (OSTI)

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-03-10

    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.

  2. Pre-neutron emission mass distributions for low-energy neutron-induced actinide fission

    E-Print Network [OSTI]

    Xiaojun Sun; Chenggang Yu; Ning Wang

    2012-01-15

    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. 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)

    2004-12-14

    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.

  4. Dynamical approach to heavy-ion induced fusion using actinide target

    SciTech Connect (OSTI)

    Aritomo, Y.; Hagino, K.; Chiba, S.; Nishio, K.

    2012-10-20

    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.

  5. 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.

    2007-11-14

    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.

  6. Mechanical environmental transport of actinides and ¹³?Cs from an arid radioactive waste disposal site

    SciTech Connect (OSTI)

    Snow, Mathew S.; Clark, Sue B.; Morrison, Samuel S.; Watrous, Matthew G.; Olson, John E.; Snyder, Darin C.

    2015-10-01

    Particulate transport represents an important mechanism for actinides and fission products at the Earth's surface; soil samples taken in the early 1970's near the Subsurface Disposal Area (SDA) at Idaho National Laboratory (INL) provide a case study for examining the mechanisms and characteristics of actinide transport under arid conditions. Transuranic waste was disposed via shallow land burial at the SDA until shortly after a flooding event that occurred in 1969. In this study we analyze soils collected in the early 1970's for ¹³?Cs, ²?¹Am, and Pu using a combination of radiometric and mass spectrometric techniques. Two distinct ²??Pu/²³?Pu isotopic ratios are observed for contamination from the SDA, with values ranging from at least 0.059 to 0.069. ²?¹Am concentrations are observed to increase only slightly in 0-4 cm soils over the ~40 year period since soil sampling, contrary to Markham's previous hypothesis that ²?¹Pu is principally associated with the 0-4 cm soil fractions (Markham 1978). The lack of statistical difference in ²?¹Am/²³??²??Pu ratios with depth suggests mechanical transport and mixing discrete contaminated particles under arid conditions. Occasional samples beyond the northeastern corner are observed to contain anomalously high Pu concentrations with corresponding low ²??Pu/²³?Pu atoms ratios, suggesting the occurrence of "hot particles;" application of a background Pu subtraction results in calculated Pu atom ratios for the "hot particles" which are statistically similar to those observed in the northeastern corner. Taken together, our data suggests that flooding resulted in mechanical transport of contaminated particles into the area between the SDA and the flood containment dike in the northeastern corner, following which subsequent contamination spreading resulted from wind transport of discrete particles.

  7. Mechanical environmental transport of actinides and ¹³?Cs from an arid radioactive waste disposal site

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

    Snow, Mathew S.; Clark, Sue B.; Morrison, Samuel S.; Watrous, Matthew G.; Olson, John E.; Snyder, Darin C.

    2015-10-01

    Particulate transport represents an important mechanism for actinides and fission products at the Earth's surface; soil samples taken in the early 1970's near the Subsurface Disposal Area (SDA) at Idaho National Laboratory (INL) provide a case study for examining the mechanisms and characteristics of actinide transport under arid conditions. Transuranic waste was disposed via shallow land burial at the SDA until shortly after a flooding event that occurred in 1969. In this study we analyze soils collected in the early 1970's for ¹³?Cs, ²?¹Am, and Pu using a combination of radiometric and mass spectrometric techniques. Two distinct ²??Pu/²³?Pu isotopic ratiosmore »are observed for contamination from the SDA, with values ranging from at least 0.059 to 0.069. ²?¹Am concentrations are observed to increase only slightly in 0-4 cm soils over the ~40 year period since soil sampling, contrary to Markham's previous hypothesis that ²?¹Pu is principally associated with the 0-4 cm soil fractions (Markham 1978). The lack of statistical difference in ²?¹Am/²³??²??Pu ratios with depth suggests mechanical transport and mixing discrete contaminated particles under arid conditions. Occasional samples beyond the northeastern corner are observed to contain anomalously high Pu concentrations with corresponding low ²??Pu/²³?Pu atoms ratios, suggesting the occurrence of "hot particles;" application of a background Pu subtraction results in calculated Pu atom ratios for the "hot particles" which are statistically similar to those observed in the northeastern corner. Taken together, our data suggests that flooding resulted in mechanical transport of contaminated particles into the area between the SDA and the flood containment dike in the northeastern corner, following which subsequent contamination spreading resulted from wind transport of discrete particles.« less

  8. Comparison of actinide production in traveling wave and pressurized water reactors

    SciTech Connect (OSTI)

    Osborne, A.G.; Smith, T.A.; Deinert, M.R.

    2013-07-01

    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)

  9. Analysis of incident-energy dependence of delayed neutron yields in actinides

    SciTech Connect (OSTI)

    Nasir, Mohamad Nasrun bin Mohd Metorima, Kouhei Ohsawa, Takaaki Hashimoto, Kengo

    2015-04-29

    The changes of delayed neutron yields (?{sub d}) of Actinides have been analyzed for incident energy up to 20MeV using realized data of precursor after prompt neutron emission, from semi-empirical model, and delayed neutron emission probability data (P{sub n}) to carry out a summation method. The evaluated nuclear data of the delayed neutron yields of actinide nuclides are still uncertain at the present and the cause of the energy dependence has not been fully understood. In this study, the fission yields of precursor were calculated considering the change of the fission fragment mass yield based on the superposition of fives Gaussian distribution; and the change of the prompt neutrons number associated with the incident energy dependence. Thus, the incident energy dependent behavior of delayed neutron was analyzed.The total number of delayed neutron is expressed as ?{sub d}=?Y{sub i} • P{sub ni} in the summation method, where Y{sub i} is the mass yields of precursor i and P{sub ni} is the delayed neutron emission probability of precursor i. The value of Y{sub i} is derived from calculation of post neutron emission mass distribution using 5 Gaussian equations with the consideration of large distribution of the fission fragments. The prompt neutron emission ?{sub p} increases at higher incident-energy but there are two different models; one model says that the fission fragment mass dependence that prompt neutron emission increases uniformly regardless of the fission fragments mass; and the other says that the major increases occur at heavy fission fragments area. In this study, the changes of delayed neutron yields by the two models have been investigated.

  10. Duct Remediation Program: Material characterization and removal/handling

    SciTech Connect (OSTI)

    Beckman, T.d.; Davis, M.M.; Karas, T.M.

    1992-11-01

    Remediation efforts were successfully performed at Rocky Flats to locate, characterize, and remove plutonium holdup from process exhaust ducts. Non-Destructive Assay (NDA) techniques were used to determine holdup locations and quantities. Visual characterization using video probes helped determine the physical properties of the material, which were used for remediation planning. Assorted equipment types, such as vacuum systems, scoops, brushes, and a rotating removal system, were developed to remove specific material types. Personnel safety and material handling requirements were addressed throughout the project.

  11. Multipollutant Removal with WOWClean® System 

    E-Print Network [OSTI]

    Romero, M.

    2010-01-01

    from the flue gas of a power plant and demonstrate the technology. The system integrates proven emission reduction techniques into a single, multi-pollutant reduction system and is designed to remove Mercury, SOx, NOx, particulates, heavy metals...

  12. Removal of arsenic compounds from petroliferous liquids

    DOE Patents [OSTI]

    Fish, R.H.

    1984-04-06

    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.

  13. Protection #2: Trap and Remove Sediment

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

    Trap and Remove Sediment Protection 2: Trap and Remove Sediment The 3 Protections Defense in Depth August 1, 2013 Sediment behind LA Canyon weir is sampled and excavated...

  14. The removal of precious metals by conductive polymer filtration

    SciTech Connect (OSTI)

    Cournoyer, M.E.

    1996-10-01

    The growing demand for platinum-group metals (PGM) within the DOE complex and in industry, the need for modern and clean processes, and the increasing volume of low-grade material for secondary PGM recovery has a direct impact on the industrial practice of recovering and refining precious metals. There is a tremendous need for advanced metal ion recovery and waste minimization techniques, since the currently used method of precipitation-dissolution is inadequate. Los Alamos has an integrated program in ligand-design and separations chemistry which has developed and evaluated a series of water- soluble metal-binding polymers for recovering actinides and toxic metals from variety of process streams. A natural extension of this work is to fabricate these metal-selective polymers into membrane based separation unites, i.e., hollow-fiber membranes. In the present investigation, the material for a novel hollow-fiber membrane is characterized and its selectivity for PGM reported. Energy and waste savings and economic competitiveness are also described.

  15. BOA: Pipe asbestos insulation removal robot system

    SciTech Connect (OSTI)

    Schempf, H.; Bares, J.; Schnorr, W. [and others

    1995-12-31

    The BOA system is a mobile pipe-external robotic crawler used to remotely strip and bag asbestos-containing lagging and insulation materials (ACLIM) from various diameter pipes in (primarily) industrial installations. Steam and process lines within the DOE weapons complex warrant the use of a remote device due to the high labor costs and high level of radioactive contamination, making manual removal extremely costly and highly inefficient. Currently targeted facilities for demonstration and remediation are Fernald in Ohio and Oak Ridge in Tennessee.

  16. COD removal characteristics in air-cathode microbial fuel cells Bioresource Technology, Volume 176, 2015, Pages 23-31.

    E-Print Network [OSTI]

    such as activated sludge process require large amount of energy to remove them. Microbial fuel cell (MFC-cathode microbial fuel cells - Renewable Energy Glob.Search... COD removal characteristics in air-cathode microbial fuel cells Bioresource Technology

  17. Removal - An alternative to clearance

    SciTech Connect (OSTI)

    Feinhals, J.; Kelch, A.; Kunze, V.

    2007-07-01

    This presentation shows the differences between the application of clearance and removal, both being procedures for materials leaving radiation protection areas permanently. The differentiation will be done on the basis of the German legislation but may be also applicable for other national legislation. For clearance in Germany two basic requirements must be given, i.e. that the materials are activated or contaminated and that they result from the licensed use or can be assigned to the scope of the license. Clearance needs not to be applied to objects in Germany which are to be removed only temporarily from controlled areas with the purpose of repair or reuse in other controlled areas. In these cases only the requirements of contamination control apply. In the case of removal it must either be proved by measurements that the relevant materials are neither activated nor contaminated or that the materials result from areas where activation or contamination is impossible due to the operational history considering operational procedures and events. If the material is considered neither activated nor contaminated there is no need for a clearance procedure. Therefore, these materials can be removed from radiation protection areas and the removal is in the responsibility of the licensee. Nevertheless, the removal procedure and the measuring techniques to be applied for the different types of materials need an agreement from the competent authority. In Germany a maximum value of 10% of the clearance values has been established in different licenses as a criterion for the application of removal. As approximately 2/3 of the total mass of a nuclear power plant is not expected to be contaminated or activated there is a need for such a procedure of removal for this non contaminated material without any regulatory control especially in the case of decommissioning. A remarkable example is NPP Stade where in the last three years more than 8600 Mg were disposed of by removal and only 315 Mg were released by clearance, even before the decommissioning licensing procedure was finished. (authors)

  18. Recovery of UO{sub 2}/PuO{sub 2} in IFR electrorefining process

    DOE Patents [OSTI]

    Tomczuk, Z.; Miller, W.E.

    1992-01-01

    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}.

  19. Recovery of UO[sub 2]/PuO[sub 2] in IFR electrorefining process

    DOE Patents [OSTI]

    Tomczuk, Z.; Miller, W.E.

    1994-10-18

    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].

  20. Recovery of UO.sub.2 /Pu O.sub.2 in IFR electrorefining process

    DOE Patents [OSTI]

    Tomczuk, Zygmunt (Lockport, IL); Miller, William E. (Naperville, IL)

    1994-01-01

    A process 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.

  1. 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.

    1991-04-01

    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.

  2. Aqueous Biphasic Systems Based on Salting-Out Polyethylene Glycol or Ionic Solutions: Strategies for Actinide or Fission Product Separations

    SciTech Connect (OSTI)

    Rogers, Robin D.; Gutowski, Keith E.; Griffin, Scott T.; Holbrey, John D.

    2004-03-29

    Aqueous biphasic systems can be formed by salting-out (with kosmotropic, waterstructuring salts) water soluble polymers (e.g., polyethylene glycol) or aqueous solutions of a wide range of hydrophilic ionic liquids based on imidazolium, pyridinium, phosphonium and ammonium cations. The use of these novel liquid/liquid biphases for separation of actinides or other fission products associated with nuclear wastes (e.g., pertechnetate salts) has been demonstrated and will be described in this presentation.

  3. Concentration of Actinides in Plant Mounds at Safety Test Nuclear Sites in Nevada

    SciTech Connect (OSTI)

    David S. Shafer; Jenna Gommes

    2008-09-15

    Plant mounds or blow-sand mounds are accumulations of soil particles and plant debris around large shrubs and are common features in deserts in the southwestern United States. Believed to be an important factor in their formation, the 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, {sup 241}Am, and U in plant mounds at safety test sites. The NAEG studies found concentrations of these contaminants to be greater in shrub mounds than in the surrounding areas of desert pavement. For example, at Project 57 on the NTTR, it was estimated that 15 percent of the radionuclide inventory of the site was associated with shrub mounds, which accounted for 17 percent of the surface area of the site, a ratio of inventory to area of 0.85. At Clean Slate III at the TTR, 29 percent of the inventory was associated with approximately 32 percent of the site covered by shrub mounds, a ratio of 0.91. While the total inventory of radionuclides in intershrub areas was greater, the ratio of radionuclide inventory to area was 0.40 and 0.38, respectively, at the two sites. The comparison between the shrub mounds and adjacent desert pavement areas was made for only the top 5 cm since radionuclides at safety test sites are concentrated in the top 5 cm of intershrub areas. Not accounting for radionuclides associated with the shrub mounds would cause the inventory of contaminants and potential exposure to be underestimated. As part of its Environmental Restoration Soils Subproject, the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office has proposed that the majority of its contaminated soil 'Corrective Action Units', including the safety test sites, be closed by fencing and posting with administrative controls. The concentration of actinides in the shrub mounds has important implications for postclosure management of the safety test sites. Because resuspension factors at safety test sites can be three to four orders-of-magnitude higher than soil sites associated with atmospheric tests where criticality occurred, the shrub mounds are an important factor in stabilization of actinide contaminants. Loss of shrubs associated with mounds from fire or plant die-back from drought could cause radionuclides at these sites to become more prone to suspension and water erosion until the sites are stabilized. Alternatively, although shrub mounds are usually composed of predominantly fine sand size particles, smaller silt and clay size particles in them are often high in CaCO{sub 3} content. The CaCO{sub 3} may act as a cementing agent to limit erosion of the shrub mounds even if the vegetation cover is temporarily lost.

  4. 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)

    2011-07-01

    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)

  5. RECYCLING AND REMOVAL OF OFFSHORE WIND TURBINES AN INTERACTIVE METHOD FOR REDUCTION OF NEGATIVE ENVIRONMENTAL EFFECTS

    E-Print Network [OSTI]

    phases of new wind turbines. There are plans about offshore wind farms in many countries e.g. in northernRECYCLING AND REMOVAL OF OFFSHORE WIND TURBINES ­ AN INTERACTIVE METHOD FOR REDUCTION OF NEGATIVE and an analysis of future removal and recycling processes of offshore wind turbines. The method is process

  6. Microbial removal of no.sub.x from gases

    DOE Patents [OSTI]

    Sublette, Kerry L. (Tulsa, OK)

    1991-01-01

    Disclosed is a process by which a gas containing nitric oxide is contacted with an anaerobic microbial culture of denitrifying bacteria to effect the chemical reduction of the nitric oxide to elemental nitrogen. The process is particularly suited to the removal of nitric oxide from flue gas streams and gas streams from nitric acid plants. Thiobacillus dentrificians as well as other bacteria are disclosed for use in the process.

  7. 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)

    1998-01-01

    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.

  8. UNIVERSITY OF SUSSEX Removal expenses

    E-Print Network [OSTI]

    Sussex, University of

    scale: first year of employment: 100% second year of employment: 50% third year of employment: 25 be payable to staff whose contract is for one year or more. However, new staff employed on a contract of less than one year who contract is subsequently extended beyond one year will then be eligible for removal

  9. 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

    2014-04-08

    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.

  10. 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

    2002-01-01

    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.

  11. Hollow-fiber supported liquid membrane (HFSLM) for the separation of lanthanides and actinides

    SciTech Connect (OSTI)

    Mohapatra, P.K.; Ansari, S.A.; Bhattacharyya, A.; Manchanda, V.K.; Patil, C.B.

    2008-07-01

    The transport behavior of Nd(III) was investigated using hollow-fiber supported liquid membranes (HFSLM) from an acidic feed solution using N,N,N',N'-tetraoctyl-diglycolamide (TODGA) in normal paraffinic hydrocarbon (NPH) as the carrier. Near quantitative transport (>99%) of Nd(III) from 500 mL of feed containing 1 g/L Nd in 3.5 M HNO{sub 3} was possible in about 45 minutes. Quantitative transport time increased when the volume or Nd(III ) concentration in the feed was increased. The liquid membrane had excellent stability as indicated by eight consecutive runs that gave consistent transport rates. The HFSLM data using Cyanex- 301 in n-dodecane as carrier extractant for the lanthanide-actinide separation with the feed solution 1 M NaNO{sub 3} at pH 3.5 and stripping solution 0.01 M EDTA at a pH 3.5 were promising. (authors)

  12. Approach for Validating Actinide and Fission Product Compositions for Burnup Credit Criticality Safety Analyses

    SciTech Connect (OSTI)

    Radulescu, Georgeta; Gauld, Ian C; Ilas, Germina; Wagner, John C

    2014-01-01

    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.

  13. Industrial lead paint removal specifications

    SciTech Connect (OSTI)

    Stone, R.C.

    1997-06-01

    The purpose of this paper is to inform the reader as to some of the pertinent rules and regulations promulgated by the Environmental Protection Agency (EPA) and the Occupational Safety and Health Administration (OSHA) that may effect an industrial lead paint removal project. The paper discusses a recommended schedule of procedures and preparations to be followed by the lead paint removal specification writer when analyzing the possible impact of the project on the environment, the public and workers. Implications of the Clean Air Act, the Clean Water Act and the Resource Conservation and Recovery Act (RCRA) along with hazardous waste handling, manifesting, transporting and disposal procedures are discussed with special emphasis placed as to their impact on the writer and the facility owner. As the rules and regulations are highly complex, the writer has attempted to explain the methodology currently being used in state-of-the-art industrial lead abatement specifications.

  14. Removing Stains from Washable Fabrics. 

    E-Print Network [OSTI]

    Beard, Ann Vanderpoorten

    1988-01-01

    unique treatment be cause of chemical make-up or physical characteristics. (Examples: chewing gum, iodine, lead pencil) Stain Removal Products Bleaches Chlorine bleaches contain a hypochlorite com pound. Do not use them on wool, silk, polyurethane... foam, spandex or blends of these fibers; on finishes which are embossed or flame retardant; or on fabrics labeled no chlorine bleach. Avoid contact with metal. Always follow package instructions for dilu tion. Liquid chlorine bleaches lose strength...

  15. Finding mold removal directions using graphics hardware Rahul Khardekar

    E-Print Network [OSTI]

    McMains, Sara

    Finding mold removal directions using graphics hardware Rahul Khardekar University of California, Berkeley Sara McMains University of California, Berkeley 1 Introduction In molding and casting manufacturing processes, molten raw ma- terial is shaped in molds from which the resulting part must be re

  16. Removal of field and embedded metal by spin spray etching

    DOE Patents [OSTI]

    Contolini, R.J.; Mayer, S.T.; Tarte, L.A.

    1996-01-23

    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.

  17. 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.

    2003-10-07

    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.

  18. Electronic Structure of Transition Metal Clusters, Actinide Complexes and Their Reactivities

    SciTech Connect (OSTI)

    Krishnan Balasubramanian

    2009-07-18

    This is a continuing DOE-BES funded project on transition metal and actinide containing species, aimed at the electronic structure and spectroscopy of transition metal and actinide containing species. While a long term connection of these species is to catalysis and environmental management of high-level nuclear wastes, the immediate relevance is directly to other DOE-BES funded experimental projects at DOE-National labs and universities. There are a number of ongoing gas-phase spectroscopic studies of these species at various places, and our computational work has been inspired by these experimental studies and we have also inspired other experimental and theoretical studies. Thus our studies have varied from spectroscopy of diatomic transition metal carbides to large complexes containing transition metals, and actinide complexes that are critical to the environment. In addition, we are continuing to make code enhancements and modernization of ALCHEMY II set of codes and its interface with relativistic configuration interaction (RCI). At present these codes can carry out multi-reference computations that included up to 60 million configurations and multiple states from each such CI expansion. ALCHEMY II codes have been modernized and converted to a variety of platforms such as Windows XP, and Linux. We have revamped the symbolic CI code to automate the MRSDCI technique so that the references are automatically chosen with a given cutoff from the CASSCF and thus we are doing accurate MRSDCI computations with 10,000 or larger reference space of configurations. The RCI code can also handle a large number of reference configurations, which include up to 10,000 reference configurations. Another major progress is in routinely including larger basis sets up to 5g functions in thee computations. Of course higher angular momenta functions can also be handled using Gaussian and other codes with other methods such as DFT, MP2, CCSD(T), etc. We have also calibrated our RECP methods with all-electron Douglas-Kroll relativistic methods. We have the capabilities for computing full CI extrapolations including spin-orbit effects and several one-electron properties and electron density maps including spin-orbit effects. We are continuously collaborating with several experimental groups around the country and at National Labs to carry out computational studies on the DOE-BES funded projects. The past work in the last 3 years was primarily motivated and driven by the concurrent or recent experimental studies on these systems. We were thus significantly benefited by coordinating our computational efforts with experimental studies. The interaction between theory and experiment has resulted in some unique and exciting opportunities. For example, for the very first time ever, the upper spin-orbit component of a heavy trimer such as Au{sub 3} was experimentally observed as a result of our accurate computational study on the upper electronic states of gold trimer. Likewise for the first time AuH{sub 2} could be observed and interpreted clearly due to our computed potential energy surfaces that revealed the existence of a large barrier to convert the isolated AuH{sub 2} back to Au and H{sub 2}. We have also worked on yet to be observed systems and have made predictions for future experiments. We have computed the spectroscopic and thermodynamic properties of transition metal carbides transition metal clusters and compared our electronic states to the anion photodetachment spectra of Lai Sheng Wang. Prof Mike Morse and coworkers(funded also by DOE-BES) and Prof Stimle and coworkers(also funded by DOE-BES) are working on the spectroscopic properties of transition metal carbides and nitrides. Our predictions on the excited states of transition metal clusters such as Hf{sub 3}, Nb{sub 2}{sup +} etc., have been confirmed experimentally by Prof. Lombardi and coworkers using resonance Raman spectroscopy. We have also been studying larger complexes critical to the environmental management of high-level nuclear wastes. In collaboration with experimental co

  19. IDENTIFYING CANDIDATE PROTEIN FOR REMOVAL OF ENVIRONMENTALLY

    E-Print Network [OSTI]

    Uppsala Universitet

    IDENTIFYING CANDIDATE PROTEIN FOR REMOVAL OF ENVIRONMENTALLY HAZARDOUS SUBSTANCES Pharem Biotech products and technologies for removing environmental hazardous substances in our everyday life. The products can be applied in areas from the private customer up to the global corporate perspective

  20. 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...

  1. Local structure in solid solutions of stabilised zirconia with actinide dioxides (UO{sub 2}, NpO{sub 2})

    SciTech Connect (OSTI)

    Walter, Marcus; Somers, Joseph; Bouexiere, Daniel; Rothe, Joerg

    2011-04-15

    The local structure of (Zr,Lu,U)O{sub 2-x} and (Zr,Y,Np)O{sub 2-x} solid solutions has been investigated by extended X-ray absorption fine structure (EXAFS). Samples were prepared by mixing reactive (Zr,Lu)O{sub 2-x} and (Zr,Y)O{sub 2-x} precursor materials with the actinide oxide powders, respectively. Sintering at 1600 {sup o}C in Ar/H{sub 2} yields a fluorite structure with U(IV) and Np(IV). As typical for stabilised zirconia the metal-oxygen and metal-metal distances are characteristic for the different metal ions. The bond lengths increase with actinide concentration, whereas highest adaptation to the bulk stabilised zirconia structure was observed for U---O and Np---O bonds. The Zr---O bond shows only a slight increase from 2.14 A at 6 mol% actinide to 2.18 A at infinite dilution in UO{sub 2} and NpO{sub 2}. The short interatomic distance between Zr and the surrounding oxygen and metal atoms indicate a low relaxation of Zr with respect to the bulk structure, i.e. a strong Pauling behaviour. -- Graphical abstract: Metal-oxygen bond distances in (Zr,Lu,U)O{sub 2-x} solid solutions with different oxygen vacancy concentrations (Lu/Zr=1 and Lu/Zr=0.5). Display Omitted Research Highlights: {yields} EXAFS indicates high U and Np adaption to the bulk structure of stabilised zirconia. {yields} Zr---O bond length is 2.18 A at infinite Zr dilution in UO{sub 2} and NpO{sub 2}. {yields} Low relaxation (strong Pauling behaviour) of Zr explains its low solubility in UO{sub 2}.

  2. The HMDS Coating Flaw Removal Tool

    SciTech Connect (OSTI)

    Monticelli, M V; Nostrand, M C; Mehta, N; Kegelmeyer, L; Johnson, M A; Fair, J; Widmayer, C

    2008-10-24

    In many high energy laser systems, optics with HMDS sol gel antireflective coatings are placed in close proximity to each other making them particularly susceptible to certain types of strong optical interactions. During the coating process, halo shaped coating flaws develop around surface digs and particles. Depending on the shape and size of the flaw, the extent of laser light intensity modulation and consequent probability of damaging downstream optics may increase significantly. To prevent these defects from causing damage, a coating flaw removal tool was developed that deploys a spot of decane with a syringe and dissolves away the coating flaw. The residual liquid is evacuated leaving an uncoated circular spot approximately 1mm in diameter. The resulting uncoated region causes little light intensity modulation and thus has a low probability of causing damage in optics downstream from the mitigated flaw site.

  3. AX Tank Farm tank removal study

    SciTech Connect (OSTI)

    SKELLY, W.A.

    1999-02-24

    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.

  4. Removal of aqueous rinsable flux residues in a batch spray dishwater

    SciTech Connect (OSTI)

    Slanina, J.T.

    1992-02-01

    An alkaline detergent solution used in an industrial dishwasher was evaluated to remove aqueous rinsable flux residues on printed wiring boards (PWBs) after hot air solder leveling and hot oil solder dip and leveling. The dishwasher, a batch cleaning process, was compared to an existing conveyorized aqueous cleaning process. The aqueous soluble flux residues from both soldering processes were removed with a solution of a mild alkaline detergent dissolved in hot deionized (DI) water.

  5. Method for removal of beryllium contamination from an article

    SciTech Connect (OSTI)

    Simandl, Ronald F.; Hollenbeck, Scott M.

    2012-12-25

    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.

  6. 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

    2009-09-22

    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.

  7. Nuclear & Radiological Material Removal | National Nuclear Security...

    National Nuclear Security Administration (NNSA)

    & Radiological Material Removal | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation...

  8. Thermal analysis for fuel handling system for sodium cooled reactor considering minor actinide-bearing metal fuel.

    SciTech Connect (OSTI)

    Chikazawa, Y.; Grandy, C.; Nuclear Engineering Division

    2009-03-01

    The Advanced Burner Reactor (ABR) is one of the components of the Global Nuclear Energy Partnership (GNEP) used to close the fuel cycle. ABR is a sodium-cooled fast reactor that is used to consume transuranic elements resulting from the reprocessing of light water reactor spent nuclear fuel. ABR-1000 [1000 MW(thermal)] is a fast reactor concept created at Argonne National Laboratory to be used as a reference concept for various future trade-offs. ABR-1000 meets the GNEP goals although it uses what is considered base sodium fast reactor technology for its systems and components. One of the considerations of any fast reactor plant concept is the ability to perform fuel-handling operations with new and spent fast reactor fuel. The transmutation fuel proposed as the ABR fuel has a very little experience base, and thus, this paper investigates a fuel-handling concept and potential issues of handling fast reactor fuel containing minor actinides. In this study, two thermal analyses supporting a conceptual design study on the ABR-1000 fuel-handling system were carried out. One analysis investigated passive dry spent fuel storage, and the other analysis investigated a fresh fuel shipping cask. Passive dry storage can be made suitable for the ABR-1000 spent fuel storage with sodium-bonded metal fuel. The thermal analysis shows that spent fast reactor fuel with a decay heat of 2 kW or less can be stored passively in a helium atmosphere. The 2-kW value seems to be a reasonable and practical level, and a combination of reasonably-sized in-sodium storage followed by passive dry storage could be a candidate for spent fuel storage for the next-generation sodium-cooled reactor with sodium-bonded metal fuel. Requirements for the shipping casks for minor actinide-bearing fuel with a high decay heat level are also discussed in this paper. The shipping cask for fresh sodium-cooled-reactor fuel should be a dry type to reduce the reaction between residual moisture on fresh fuel and the sodium coolant. The cladding temperature requirement is maintained below the creep temperature limit to avoid any damage before core installation. The thermal analysis shows that a helium gas-filled cask can accommodate ABR-1000 fresh minor actinide-bearing fuel with 700-W decay heat. The above analysis results revealed the overall requirement for minor actinide-bearing metal fuel handling. The information is thought to be helpful in the design of the ABR-1000 and future sodium-cooled-reactor fuel-handling system.

  9. Synergistic extraction of trivalent lanthanides and actinides from acidic chloride media by tetra(n-octyl)diglycolamide

    SciTech Connect (OSTI)

    McAlister, D.R.; Horwitz, E. Philip

    2008-07-01

    Ferric chloride has been found to induce a significant synergistic enhancement of the extraction of trivalent lanthanides and actinides by tetra(n-octyl)diglycolamide (TODGA) from acidic chloride media. In this manuscript, results of a thorough investigation of the TODGA-HCl-Fe(III) system using solvent-extraction experiments designed to elucidate the stoichiometry of the synergistic species will be described. Results for Ac(III), Am(III), Eu(III), Pm(III), Y(III), Th(IV), Pu(IV), and U(VI) will be discussed. (authors)

  10. Atomistic Calculations of the Effect of Minor Actinides on Thermodynamic and Kinetic Properties of UO{sub 2{+-}x}

    SciTech Connect (OSTI)

    Deo, Chaitanya; Adnersson, Davis; Battaile, Corbett; uberuaga, Blas

    2012-10-30

    The team will examine how the incorporation of actinide species important for mixed oxide (MOX) and other advanced fuel designs impacts thermodynamic quantities of the host UO{sub 2} nuclear fuel and how Pu, Np, Cm and Am influence oxygen mobility. In many cases, the experimental data is either insufficient or missing. For example, in the case of pure NpO2, there is essentially no experimental data on the hyperstoichiometric form it is not even known if hyperstoichiometry NpO{sub 2{+-}x} is stable. The team will employ atomistic modeling tools to calculate these quantities

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

    SciTech Connect (OSTI)

    Nash, Kenneth L.; Rao, Linfeng

    2005-06-01

    Removal of waste-limiting components of sludge (Al, Cr, S, P) in underground tanks at Hanford by treatment with concentrated alkali has proven less efficacious for Al and Cr removal than had been hoped. More aggressive treatments of sludges, for example, contact with oxidants targeting Cr(III), have been tested in a limited number of samples and found to improve leaching efficiency for Cr. Oxidative alkaline leaching can be expected to have at best a secondary influence on the mobilization of Al. Our earlier explorations of Al leaching from sludge simulants indicated acidic and complexometric leaching can improve Al dissolution.

  12. Quick Stain Removal Guide (Spanish) 

    E-Print Network [OSTI]

    Brown, Pamela J.

    2001-08-22

    -5191S de la Oficina de Extensi?n, Etiquetas de Instrucciones y su Ropa). Lleve las prendas de vestir de lavado en seco a tintorer?as profe- sionales. Siga estas pautas para las prendas de vestir lavables: Sepa de qu? es la mancha. Si no sabe, intente...). The Soap and Detergent Association, Marzo/Abril. NY: New York. Laundering Facts. (1991). The Soap and Detergent Association. NY: New York. Beard, A.V. (1989). Removing stains from washable fabrics. Servicio de Extensi?n Agr?cola de Texas. TX: College...

  13. Section 46: Removal of Waste

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power AdministrationRobust, High-ThroughputUpcomingmagnetoresistanceand Governmentm D m r e DFigureEffects ofRemoval

  14. Removal of Separable Organic From Tank 241-C-103 Scoping Study

    SciTech Connect (OSTI)

    KOCH, M.R.

    2000-05-16

    This study is based on previous evaluations/proposals for removing the floating organic layer in C-103. A practical method is described with assumptions, cost and schedule estimates, and risks. Proposed operational steps include bulk organic removal, phase separation, organic washing and offsite disposal, followed by an in-situ polishing process.

  15. Balloon Dilatation: A Helpful Technique for Removal of a Stuck Dialysis Line

    SciTech Connect (OSTI)

    Farooq, Ammad Jones, Vaughan Agarwal, Sanjay

    2012-12-15

    We describe a useful technique for the removal of an irretrievable/stuck long-term intravenous catheter. The alternative would have meant removing it surgically or snaring it in a case of extremely difficult venous access. The process we used was effective in this particular case.

  16. Modeling removal of bacteriophages MS2 and PRD1 by dune recharge at Castricum, Netherlands

    E-Print Network [OSTI]

    Hassanizadeh, S. Majid

    Modeling removal of bacteriophages MS2 and PRD1 by dune recharge at Castricum, Netherlands Jack F by dune recharge was studied at a field site in the dune area of Castricum, Netherlands. Recharge water at investigating virus removal processes during dune recharge, are presented. In the Netherlands, about 14

  17. Removal of I, Rn, Xe and Kr from off gas streams using PTFE membranes

    DOE Patents [OSTI]

    Siemer, Darryl D. (Idaho Falls, ID); Lewis, Leroy C. (Idaho Falls, ID)

    1990-01-01

    A process for removing I, R, Xe and Kr which involves the passage of the off gas stream through a tube-in-shell assembly, whereby the tubing is a PTFE membrane which permits the selective passages of the gases for removing and isolating the gases.

  18. Conversion of oil shale ash into zeolite for cadmium and lead removal from wastewater

    E-Print Network [OSTI]

    Shawabkeh, Reyad A.

    Conversion of oil shale ash into zeolite for cadmium and lead removal from wastewater Reyad; available online 29 October 2003 Abstract A by-product fly ash from oil shale processing was converted shale; Ash; Zeolite; Cadmium and lead removal 1. Introduction Oil shale exists in Jordan with large

  19. Removing Weather Effects from Monochrome Images Srinivasa G. Narasimhan and Shree K. Nayar

    E-Print Network [OSTI]

    Narasimhan, Srinivasa G.

    Removing Weather Effects from Monochrome Images Srinivasa G. Narasimhan and Shree K. Nayar}@cs.columbia.edu Abstract Images of outdoor scenes captured in bad weather suffer from poor contrast. Under bad weather space invariant image processing techniques are not sufficient to remove weather effects from images

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

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    of the TMSR, including pumps and heat exchangers (IHX) The core structures are protected by reflectors which and reprocessing, drawbacks incompatible with any industrial development. On the other hand, the advantages flowing through the pumps, the gaseous extraction system which removes the gaseous and non soluble fission

  1. Method of making thermally removable epoxies

    SciTech Connect (OSTI)

    Loy, Douglas A. (Albuquerque, NM); Wheeler, David R. (Albuquerque, NM); Russick, Edward M. (Rio Rancho, NM); McElhanon, James R. (Albuquerque, NM); Saunders, Randall S. (late of Albuquerque, NM)

    2002-01-01

    A method of making a thermally-removable epoxy by mixing a bis(maleimide) compound to a monomeric furan compound containing an oxirane group to form a di-epoxy mixture and then adding a curing agent at temperatures from approximately room temperature to less than approximately 90.degree. C. to form a thermally-removable epoxy. The thermally-removable epoxy can be easily removed within approximately an hour by heating to temperatures greater than approximately 90.degree. C. in a polar solvent. The epoxy material can be used in protecting electronic components that may require subsequent removal of the solid material for component repair, modification or quality control.

  2. TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES

    SciTech Connect (OSTI)

    DOE

    1997-04-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading. A measurement of the average assembly burnup is required and that measurement must be within 10% of the utility burnup record for the assembly to be accepted. The measurement device must be accurate to within 10%. Each step is described in detail for use with any computer code system and is then demonstrated with the SCALE 4.2 computer code package using 27BURNUPLIB cross sections.

  3. Detroit Edison's Fermi 1 - Preparation for Reactor Removal

    SciTech Connect (OSTI)

    Swindle, Danny [Sargent and Lundy Engineers, LLC, 55 E. Monroe Street, Chicago, IL 60603 (United States)

    2008-01-15

    This paper is intended to provide information about the ongoing decommissioning tasks at Detroit Edison's Fermi 1 plant, and in particular, the work being performed to prepare the reactor for removal and disposal. In 1972 Fermi 1 was shutdown and the fuel returned to the Atomic Energy Commission. By the end of 1975, a retirement plan was prepared, the bulk sodium removed, and the plant placed in a safe store condition. The plant systems were left isolated with the sodium containing systems inert with carbon dioxide in an attempt to form a carbonate layer, thus passivating the underlying reactive sodium. In 1996, Detroit Edison determined to evaluate the condition of the plant and to make recommendations in relation to the Fermi 1 future plans. At the end of 1997 approval was obtained to remove the bulk asbestos and residual alkali-metals (i.e., sodium and sodium potassium (NaK)). In 2000, full nuclear decommissioning of the plant was approved. To date, the bulk asbestos insulation has been removed, and the only NaK remaining is located in six capillary instrument tubes. The remaining sodium is contained within the reactor, two of the three primary loops, and miscellaneous removed pipes and equipment to be processed. The preferred method for removing or reacting sodium at Fermi 1 is by injecting superheated steam into a heated, nitrogen inert system. The byproducts of this reaction are caustic sodium hydroxide, hydrogen gas, and heat. The decision was made to separate the three primary loops from the reactor for better control prior to processing each loop and the reactor separately. The first loop has already been processed. The main focus is now to process the reactor to allow removal and disposal of the Class C waste prior to the anticipated June 2008 closure of the Barnwell radioactive waste disposal facility located in South Carolina. Lessons learnt are summarized and concern: the realistic schedule and adherence to the schedule, time estimates, personnel accountability, back up or fill in work, work packages, condensation control, radiological contamination control, and organization of the waste stream.

  4. High Metal Removal Rate Process for Machining Difficult Materials

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

    PolyPlus Battery Company has developed the protected lithium electrode (PLE), an invention that enables the use of lithium metal as an electrode in a secondary battery....

  5. High Metal Removal Rate Process for Machining Difficult Materials

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

    for treating wastewater exist in a number of industries, including petroleum refining, pulp and paper, and chemical manufacturing. The sustainable low- energy approach to water...

  6. High Metal Removal Rate Process for Machining Difficult Materials

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

    technology (HSPT) (Completed). * Demonstrate targeted mechanical properties, including static tensile (yield strength: 950MPa and elongation of 13.1%), fracture toughness of K Ic...

  7. 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)

    1995-01-01

    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.

  8. Flotation machine and process for removing impurities from coals

    DOE Patents [OSTI]

    Szymocha, K.; Ignasiak, B.; Pawlak, W.; Kulik, C.; Lebowitz, H.E.

    1997-02-11

    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. 4 figs.

  9. 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)

    1997-01-01

    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.

  10. Flotation machine and process for removing impurities from coals

    DOE Patents [OSTI]

    Szymocha, K.; Ignasiak, B.; Pawlak, W.; Kulik, C.; Lebowitz, H.E.

    1995-12-05

    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 mineral 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. 4 figs.

  11. High Metal Removal Rate Process for Machining Difficult Materials

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

    tool, milling operations where the cutting tool rotates against the workpiece, and drilling operations where holes are produced or refined. However, traditional machining...

  12. Removal of Engineered Nanomaterials Through Conventional Water Treatment Processes

    E-Print Network [OSTI]

    Honda, Ryan

    2014-01-01

    Silver Released into Water from Commercially Available SockStability of commercial metal oxide nanoparticles in water.Water Res. 2008, 42 (8-9), 2204-12. Crittenden, J. C. ;

  13. Process for removing sulfur from sulfur-containing gases

    DOE Patents [OSTI]

    Rochelle, Gary T. (Austin, TX); Jozewicz, Wojciech (Chapel Hill, NC)

    1990-01-01

    The present disclosure relates to i The government may own certain rights in the present invention pursuant to EPA Cooperative Agreement CR 81-1531. This is a continuation of U.S. Ser. No. 928,337, filed Nov. 7, 1986, now U.S. Pat. No. 4,804,521.

  14. High Metal Removal Rate Process for Machining Difficult Materials

    Broader source: Energy.gov (indexed) [DOE]

    in high precision manufacturing environments: Fuel injector nozzle drilling (automotive industry) Ceramic hole drilling (electronics industry) Precious metal...

  15. Removal of Engineered Nanomaterials Through Conventional Water Treatment Processes

    E-Print Network [OSTI]

    Honda, Ryan

    2014-01-01

    nanocomposites used in sunscreen. Dispersion and fate of thetextiles, paints, and sunscreens 2,3,4 , 5 . The mostproducts (i.e. paints, sunscreens, toothpastes, and food-

  16. High Metal Removal Rate Process for Machining Difficult Materials

    Broader source: Energy.gov (indexed) [DOE]

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirleyEnergyTher i nA Guide to TappingWORK BREAKDOWNEnergy how toEM&High impactHigh

  17. High Metal Removal Rate Process for Machining Difficult Materials

    Broader source: Energy.gov (indexed) [DOE]

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirleyEnergyTher i nA Guide to TappingWORK BREAKDOWNEnergy how toEM&High impactHighAndy

  18. 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:FinancingPetroleum12,ExecutiveFinancing ProgramsDepartment of¡ ¢HelpHighJian

  19. Fuel removal, transport, and storage

    SciTech Connect (OSTI)

    Reno, H.W.

    1986-01-01

    The March 1979 accident at Unit 2 of the Three Mile Island Nuclear Power Station (TMI-2) which damaged the core of the reactor resulted in numerous scientific and technical challenges. Some of those challenges involve removing the core debris from the reactor, packaging it into canisters, loading canisters into a rail cask, and transporting the debris to the Idaho National Engineering Laboratory (INEL) for storage, examination, and preparation for final disposal. This paper highlights how some challenges were resolved, including lessons learned and benefits derived therefrom. Key to some success at TMI was designing, testing, fabricating, and licensing two rail casks, which each provide double containment of the damaged fuel. 10 refs., 12 figs.

  20. Removal of pollutant compounds from water supplies using ozone, ultraviolet light, and a counter, current packed column. Master's thesis

    SciTech Connect (OSTI)

    Kelly, E.L.

    1991-01-01

    Many water pollutants are determined to be carcinogenic and often appear in very low concentrations and still pose a health risk. Conventional water treatment processes cannot remove these contaminants and there is a great demand for the development of alternative removal technologies. The use of ozone and ultraviolet light in a counter current packed column could prove to be an effective treatment process to remove these contaminants.

  1. Removal of Pollutants by Atmospheric Non Thermal Plasmas Ahmed Khacef 1*

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    difficult to handle with conventional removal technologies like thermal and catalytic oxidation examples are hydrocarbons, chlorocarbons and chlorofluorocarbons (CFCs). Contamination of exhaust air streams with gaseous hydrocarbons or organic solvent vapours occurs in many industrial processes, e. g

  2. The Effectiveness of Full Actinide Recycle as a Nuclear Waste Management Strategy when Implemented over a Limited Timeframe – Part II: Thorium Fuel Cycle

    E-Print Network [OSTI]

    Lindley, Benjamin A.; Fiorina, Carlo; Gregg, Robert; Franceschini, Fausto; Parks, Geoffrey T.

    2014-12-06

    water reactors (LWRs) or sodium-cooled fast reactors (SFRs) is considered for uranium (U) fuel cycles. With full actinide recycling, at least 6 generations of SFRs are required in a gradual phase-out of nuclear power to achieve transmutation performance...

  3. Photochemical route to actinide-transition metal bonds: synthesis, characterization and reactivity of a series of thorium and uranium heterobimetallic complexes

    SciTech Connect (OSTI)

    Ward, Ashleigh; Lukens, Wayne; Lu, Connie; Arnold, John

    2014-04-01

    A series of actinide-transition metal heterobimetallics has been prepared, featuring thorium, uranium and cobalt. Complexes incorporating the binucleating ligand N[-(NHCH2PiPr2)C6H4]3 and Th(IV) (4) or U(IV) (5) with a carbonyl bridged [Co(CO)4]- unit were synthesized from the corresponding actinide chlorides (Th: 2; U: 3) and Na[Co(CO)4]. Irradiation of the isocarbonyls with ultraviolet light resulted in the formation of new species containing actinide-metal bonds in good yields (Th: 6; U: 7); this photolysis method provides a new approach to a relatively rare class of complexes. Characterization by single-crystal X-ray diffraction revealed that elimination of the bridging carbonyl is accompanied by coordination of a phosphine arm from the N4P3 ligand to the cobalt center. Additionally, actinide-cobalt bonds of 3.0771(5) and 3.0319(7) for the thorium and uranium complexes, respectively, were observed. The solution state behavior of the thorium complexes was evaluated using 1H, 1H-1H COSY, 31P and variable-temperature NMR spectroscopy. IR, UV-Vis/NIR, and variable-temperature magnetic susceptibility measurements are also reported.

  4. Method of CO.sub.2 removal from a gasesous stream at reduced temperature

    DOE Patents [OSTI]

    Fisher, James C; Siriwardane, Ranjani V; Berry, David A; Richards, George A

    2014-11-18

    A method for the removal of H.sub.2O and CO.sub.2 from a gaseous stream comprising H.sub.2O and CO.sub.2, such as a flue gas. The method initially utilizes an H.sub.2O removal sorbent to remove some portion of the H.sub.2O, producing a dry gaseous stream and a wet H.sub.2O removal sorbent. The dry gaseous stream is subsequently contacted with a CO.sub.2 removal sorbent to remove some portion of the CO.sub.2, generating a dry CO.sub.2 reduced stream and a loaded CO.sub.2 removal sorbent. The loaded CO.sub.2 removal sorbent is subsequently heated to produce a heated CO.sub.2 stream. The wet H.sub.2O removal sorbent and the dry CO.sub.2 reduced stream are contacted in a first regeneration stage, generating a partially regenerated H.sub.2O removal sorbent, and the partially regenerated H.sub.2O removal sorbent and the heated CO.sub.2 stream are subsequently contacted in a second regeneration stage. The first and second stage regeneration typically act to retain an initial monolayer of moisture on the various removal sorbents and only remove moisture layers bound to the initial monolayer, allowing for relatively low temperature and pressure operation. Generally the applicable H.sub.2O sorption/desorption processes may be conducted at temperatures less than about 70.degree. C. and pressures less than 1.5 atmospheres, with certain operations conducted at temperatures less than about 50.degree. C.

  5. Perchlorate ion (C104) removal using an electrochemically induced catalytic reaction on modified activated carbon 

    E-Print Network [OSTI]

    Langille, Meredith Caitlyn

    2009-05-15

    to remove perchlorate ion. Bioremediation has been used to remove perchlorate, generally with the addition of acetate and hydrogen as electron donors. Phytoremediation with plants like smartweed, watercress and the trees chinaberry, elm, willow, mulberry... treatment processes to chemically reduce (a transfer of electrons) perchlorate ion cannot do so effectively due to a high activation energy barrier. This unique energy barrier and electron transfer process is called the bridged electron transfer (6, 7, 8...

  6. Apparatus and method for loading and unloading multiple digital tape cassettes utilizing a removable magazine

    DOE Patents [OSTI]

    Lindenmeyer, C.W.

    1993-01-26

    An apparatus and method to automate the handling of multiple digital tape cassettes for processing by commercially available cassette tape readers and recorders. A removable magazine rack stores a plurality of tape cassettes, and cooperates with a shuttle device that automatically inserts and removes cassettes from the magazine to the reader and vice-versa. Photocells are used to identify and index to the desired tape cassette. The apparatus allows digital information stored on multiple cassettes to be processed without significant operator intervention.

  7. Savannah River Site Waste Removal Program - Past, Present and Future

    SciTech Connect (OSTI)

    Saldivar, E.

    2002-02-25

    The Savannah River Site has fifty-one high level waste tanks in various phases of operation and closure. These tanks were originally constructed to receive, store, and treat the high level waste (HLW) created in support of the missions assigned by the Department of Energy (DOE). The Federal Facilities Agreement (FFA) requires the high level waste to be removed from the tanks and stabilized into a final waste form. Additionally, closure of the tanks following waste removal must be completed. The SRS HLW System Plan identifies the interfaces of safe storage, waste removal, and stabilization of the high level waste and the schedule for the closure of each tank. HLW results from the dissolution of irradiated fuel components. Desired nuclear materials are recovered and the byproducts are neutralized with NaOH and sent to the High Level Waste Tank Farms at the SRS. The HLW process waste clarifies in the tanks as the sludge settles, resulting in a layer of dense sludge with salt supernate settling above the sludge. Salt supernate is concentrated via evaporation into saltcake and NaOH liquor. This paper discusses the history of SRS waste removal systems, recent waste removal experiences, and the challenges facing future removal operations to enhance efficiency and cost effectiveness. Specifically, topics will include the evolution and efficiency of systems used in the 1960's which required large volumes of water to current systems of large centrifugal slurry pumps, with significant supporting infrastructure and safety measures. Interactions of this equipment with the waste tank farm operations requirements will also be discussed. The cost and time improvements associated with these present-day systems is a primary focus for the HLW Program.

  8. Method for removing contaminants from plastic resin

    DOE Patents [OSTI]

    Bohnert, George W. (Harrisonville, MO); Hand, Thomas E. (Lee's Summit, MO); DeLaurentiis, Gary M. (Jamestown, CA)

    2008-12-09

    A resin recycling method that produces essentially contaminant-free synthetic resin material in an environmentally safe and economical manner. The method includes receiving the resin in container form. The containers are then ground into resin particles. The particles are exposed to a solvent, the solvent contacting the resin particles and substantially removing contaminants on the resin particles. After separating the particles and the resin, a solvent removing agent is used to remove any residual solvent remaining on the resin particles after separation.

  9. PRTR ion exchange vault water removal

    SciTech Connect (OSTI)

    Ham, J.E.

    1995-11-01

    This report documents the removal of radiologically contaminated water from the Plutonium Recycle Test Reactor (PRTR) ion exchange vault. Approximately 57,000 liters (15,000 gallons) of water had accumulated in the vault due to the absence of a rain cover. The water was removed and the vault inspected for signs of leakage. No evidence of leakage was found. The removal and disposal of the radiologically contaminated water decreased the risk of environmental contamination.

  10. Removal of arsenic compounds from spent catecholated polymer

    DOE Patents [OSTI]

    Fish, Richard H. (Berkeley, CA)

    1985-01-01

    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.

  11. Criticality safety aspects of K-25 Building uranium deposit removal

    SciTech Connect (OSTI)

    Haire, M.J.; Jordan, W.C. [Oak Ridge National Lab., TN (United States); Ingram, J.C. III; Stinnet, E.C. Jr. [Oak Ridge K-25 Site, TN (United States)

    1995-12-31

    The K-25 Building of the Oak Ridge Gaseous Diffusion Plant (now the K-25 Site) went into operation during World War II as the first large scale production plant to separate {sup 235}U from uranium by the gaseous diffusion process. It operated successfully until 1964, when it was placed in a stand-by mode. The Department of Energy has initiated a decontamination and decommissioning program. The primary objective of the Deposit Removal (DR) Project is to improve the nuclear criticality safety of the K-25 Building by removing enriched uranium deposits from unfavorable-geometry process equipment to below minimum critical mass. The method utilized to accomplish this are detailed in this report.

  12. Method for removing undesired particles from gas streams

    DOE Patents [OSTI]

    Durham, M.D.; Schlager, R.J.; Ebner, T.G.; Stewart, R.M.; Hyatt, D.E.; Bustard, C.J.; Sjostrom, S.

    1998-11-10

    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. 11 figs.

  13. System for removing contaminants from plastic resin

    DOE Patents [OSTI]

    Bohnert, George W. (Harrisonville, MO); Hand, Thomas E. (Lee's Summit, MO); DeLaurentiis, Gary M. (Jamestown, CA)

    2010-11-23

    A resin recycling system that produces essentially contaminant-free synthetic resin material in an environmentally safe and economical manner. The system includes receiving the resin in container form. A grinder grinds the containers into resin particles. The particles are exposed to a solvent in one or more solvent wash vessels, the solvent contacting the resin particles and substantially removing contaminants on the resin particles. A separator is used to separate the resin particles and the solvent. The resin particles are then placed in solvent removing element where they are exposed to a solvent removing agent which removes any residual solvent remaining on the resin particles after separation.

  14. Turkey HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Turkey HEU Removal | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the...

  15. General Counsel Legal Interpretation Regarding Medical Removal...

    Energy Savers [EERE]

    Regarding Medical Removal Protection Benefits Pursuant to 10 CFR Part 850, Chronic Beryllium Disease Prevention Program General Counsel Legal Interpretation Regarding Medical...

  16. Slag capture and removal during laser cutting

    DOE Patents [OSTI]

    Brown, Clyde O. (Newington, CT)

    1984-05-08

    Molten metal removed from a workpiece in a laser cutting operation is blown away from the cutting point by a gas jet and collected on an electromagnet.

  17. Method of making thermally removable polymeric encapsulants

    SciTech Connect (OSTI)

    Small, James H. (Santa Fe, NM); Loy, Douglas A. (Albuquerque, NM); Wheeler, David R. (Albuquerque, NM); McElhanon, James R. (Albuquerque, NM); Saunders, Randall S. (late of Albuquerque, NM)

    2001-01-01

    A method of making a thermally-removable encapsulant by heating a mixture of at least one bis(maleimide) compound and at least one monomeric tris(furan) or tetrakis(furan) compound at temperatures from above room temperature to less than approximately 90.degree. C. to form a gel and cooling the gel to form the thermally-removable encapsulant. The encapsulant can be easily removed within approximately an hour by heating to temperatures greater than approximately 90.degree. C., preferably in a polar solvent. The encapsulant can be used in protecting electronic components that may require subsequent removal of the encapsulant for component repair, modification or quality control.

  18. BOA II: pipe-asbestos insulation removal system

    SciTech Connect (OSTI)

    Schempf, H.; Mutschler; Boehmke, S.; Chemel, B.; Piepgras, C.

    1996-12-31

    BOA system is a mobile pipe-external robotic crawler used to remotely strip and bag asbestos-containing lagging and insulation materials from various diameter pipes in (primarily) industrial installations. Steam and process lines within the DOE weapons complex warrant the use of a remote device due to high labor costs and high level of radioactive contamination, making manual removal costly and inefficient. Currently targeted facilities for demonstration and remediation are Fernald in Ohio and Oak Ridge in Tennessee.

  19. CX-010025: Categorical Exclusion Determination

    Broader source: Energy.gov [DOE]

    Actinide Removal Process /Modular Caustic Side Solvent Extraction Unit Life Extension Support Testing CX(s) Applied: B3.6 Date: 01/24/2013 Location(s): South Carolina Offices(s): Savannah River Operations Office

  20. A Multi-Factor Analysis of Sustainable Agricultural Residue Removal Potential

    SciTech Connect (OSTI)

    Jared Abodeely; David Muth; Paul Adler; Eleanor Campbell; Kenneth Mark Bryden

    2012-10-01

    Agricultural residues have significant potential as a near term source of cellulosic biomass for bioenergy production, but sustainable removal of agricultural residues requires consideration of the critical roles that residues play in the agronomic system. Previous work has developed an integrated model to evaluate sustainable agricultural residue removal potential considering soil erosion, soil organic carbon, greenhouse gas emission, and long-term yield impacts of residue removal practices. The integrated model couples the environmental process models WEPS, RUSLE2, SCI, and DAYCENT. This study uses the integrated model to investigate the impact of interval removal practices in Boone County, Iowa, US. Residue removal of 4.5 Mg/ha was performed annually, bi-annually, and tri-annually and were compared to no residue removal. The study is performed at the soil type scale using a national soil survey database assuming a continuous corn rotation with reduced tillage. Results are aggregated across soil types to provide county level estimates of soil organic carbon changes and individual soil type soil organic matter content if interval residue removal were implemented. Results show interval residue removal is possible while improving soil organic matter. Implementation of interval removal practices provide greater increases in soil organic matter while still providing substantial residue for bioenergy production.