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Sample records for actinide removal process

  1. Process to remove actinides from soil using magnetic separation

    DOE Patents [OSTI]

    Avens, Larry R.; Hill, Dallas D.; Prenger, F. Coyne; Stewart, Walter F.; Tolt, Thomas L.; Worl, Laura A.

    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, 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.

  6. 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.

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

    SciTech Connect (OSTI)

    Spencer, B.B.; Egan, B.Z.; Chase, C.W.

    1997-07-01

    Experiments were conducted to evaluate the transuranium extraction process for partitioning actinides from actual dissolved high-level radioactive waste sludge. All tests were performed at ambient temperature. Time and budget constraints permitted only two experimental campaigns. Samples of sludge from Melton Valley Storage Tank W-25 were rinsed with mild caustic (0.2 M NaOH) to reduce the concentrations of nitrates and fission products associated with the interstitial liquid. In one campaign, the rinsed sludge was dissolved in nitric acid to produce a solution containing total metal concentrations of ca. 1.8 M with a nitric acid concentration of ca. 2.9 M. About 50% of the dry mass of the sludge was dissolved. In the other campaign, the sludge was neutralized with nitric acid to destroy the carbonates, then leached with ca. 2.6 M NaOH for ca. 6 h before rinsing with the mild caustic. The sludge was then dissolved in nitric acid to produce a solution containing total metal concentrations of ca. 0.6 M with a nitric acid concentration of ca. 1.7 M. About 80% of the sludge dissolved. The dissolved sludge solution form the first campaign began gelling immediately, and a visible gel layer was observed after 8 days. In the second campaign, the solution became hazy after ca. 8 days, indicating gel formation, but did not display separated gel layers after aging for 20 days. Batch liquid-liquid equilibrium tests of both the extraction and stripping operations were conducted. Chemical analyses of both phases were used to evaluate the process. Evaluation was based on two metrics: the fraction of TRU elements removed from the dissolved sludge and comparison of the results with predictions made with the Generic TRUEX Model (GTM). The fractions of Eu, Pu, Cm, Th, and U species removed from aqueous solution in only one extraction stage were > 95% and were close to the values predicted by the GTM. Mercury was also found to be strongly extracted, with a one-stage removal of > 92%.

  8. Actinide removal from spent salts

    DOE Patents [OSTI]

    Hsu, Peter C.; von Holtz, Erica H.; Hipple, David L.; Summers, Leslie J.; Adamson, Martyn G.

    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.

  9. Bidentate organophosphorus solvent extraction process for actinide recovery and partition

    DOE Patents [OSTI]

    Schulz, Wallace W.

    1976-01-01

    A liquid-liquid extraction process for the recovery and partitioning of actinide values from acidic nuclear waste aqueous solutions, the actinide values including trivalent, tetravalent and hexavalent oxidation states is provided and includes the steps of contacting the aqueous solution with a bidentate organophosphorous extractant to extract essentially all of the actinide values into the organic phase. Thereafter the respective actinide fractions are selectively partitioned into separate aqueous solutions by contact with dilute nitric or nitric-hydrofluoric acid solutions. The hexavalent uranium is finally removed from the organic phase by contact with a dilute sodium carbonate solution.

  10. Process for recovering actinide values

    DOE Patents [OSTI]

    Horwitz, E. Philip; Mason, George W.

    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.

  11. In vitro removal of actinide (IV) ions

    DOE Patents [OSTI]

    Weitl, Frederick L.; Raymond, Kenneth N.

    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.

  12. 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.

  13. 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).

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

    DOE Patents [OSTI]

    Ebbinghaus, Bartley B.; Van Konynenburg, Richard A.; Vance, Eric R.; Stewart, Martin W.; Walls, Philip A.; Brummond, William Allen; Armantrout, Guy A.; Herman, Connie Cicero; Hobson, Beverly F.; Herman, David Thomas; Curtis, Paul G.; Farmer, Joseph

    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.

  15. Development of the Actinide-Lanthanide Separation (ALSEP) Process

    SciTech Connect (OSTI)

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

    2014-09-30

    Separating the minor actinide elements (Am and Cm) from acidic high-level raffinates arising from the reprocessing of irradiated nuclear fuel is an important step in closing the nuclear fuel cycle. Most proposed approaches to this problem involve two solvent extraction steps: 1) co-extraction of the trivalent lanthanides and actinides, followed by 2) separation of the actinides from the lanthanides. The objective of our work is to develop a single solvent-extraction process for isolating the minor actinide elements. We report here a solvent containing N,N,N',N'-tetra(2 ethylhexyl)diglycolamide (T2EHDGA) combined with 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) that can be used to separate the minor actinides in a single solvent-extraction process. T2EHDGA serves to co-extract the trivalent actinide and lanthanide ions from nitric acid solution. Switching the aqueous phase chemistry to a citrate buffered solution of N-(2-hydroxyethyl)ethylenediamine-N,N',N'-triacetic acid at pH 2.5 to 4 results in selective transfer of the actinides to the aqueous phase, thus affecting separation of the actinides from the lanthanides. Separation factors between the lanthanides and actinides are approximately 20 in the pH range of 3 to 4, and the distribution ratios are not highly dependent on the pH in this system.

  16. Silica Scaling Removal Process

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

    Scaling Removal Process Scientists at Los Alamos National Laboratory have developed a novel technology to remove both dissolved and colloidal silica using small gel particles....

  17. 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.

  18. 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.

  19. Process to remove rare earth from IFR electrolyte

    DOE Patents [OSTI]

    Ackerman, John P.; Johnson, Terry R.

    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.

  20. Processing and Disposition of Special Actinide Target Materials

    Office of Scientific and Technical Information (OSTI)

    (Conference) | SciTech Connect SciTech Connect Search Results Conference: Processing and Disposition of Special Actinide Target Materials Citation Details In-Document Search Title: Processing and Disposition of Special Actinide Target Materials Authors: Robinson, Sharon M [1] ; Patton, Bradley D [1] + Show Author Affiliations ORNL Publication Date: 2013-01-01 OSTI Identifier: 1088123 DOE Contract Number: DE-AC05-00OR22725 Resource Type: Conference Resource Relation: Conference: WM2013,

  1. Phenol removal pretreatment process

    DOE Patents [OSTI]

    Hames, Bonnie R.

    2004-04-13

    A process for removing phenols from an aqueous solution is provided, which comprises the steps of contacting a mixture comprising the solution and a metal oxide, forming a phenol metal oxide complex, and removing the complex from the mixture.

  2. 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.

  3. SRNL Development of Recovery Processes for Mark-18A Heavy Actinide...

    Office of Scientific and Technical Information (OSTI)

    SRNL Development of Recovery Processes for Mark-18A Heavy Actinide Targets Citation ... Visit OSTI to utilize additional information resources in energy science and technology. A ...

  4. SRNL Development of Recovery Processes for Mark-18A Heavy Actinide Targets

    Office of Scientific and Technical Information (OSTI)

    (Conference) | SciTech Connect SRNL Development of Recovery Processes for Mark-18A Heavy Actinide Targets Citation Details In-Document Search Title: SRNL Development of Recovery Processes for Mark-18A Heavy Actinide Targets Savannah River National Laboratory (SRNL) and Oak Ridge National Laboratory (ORNL) are developing plans for the recovery of rare and unique isotopes contained within heavy-actinide target assemblies, specifically the Mark-18A. Mark-18A assemblies were irradiated in

  5. 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)

  6. Development and validation of process models for minor actinide separations processes using centrifugal contactors

    SciTech Connect (OSTI)

    Fox, O.D.; Carrott, M.J.; Gaubert, E.; Maher, C.J.; Mason, C.; Taylor, R.J.; Woodhead, D.A.

    2007-07-01

    As any future spent fuel treatment facility is likely to be based on intensified solvent extraction equipment it is important to understand the chemical and mass transfer kinetics of the processes involved. Two candidate minor actinide separations processes have been examined through a programme of modeling and experimental work to illustrate some of the issues to address in turning these technologies in to fully optimized processes suitable for industrialization. (authors)

  7. 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.

  8. 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.

  9. Analysis of large soil samples for actinides

    DOE Patents [OSTI]

    Maxwell, III; Sherrod L.

    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.

  10. Partitioning of minor actinides from PUREX raffinate by the TODGA process

    SciTech Connect (OSTI)

    Magnusson, D.; Christiansen, B.; Glatz, J.P.; Malmbeck, R.; Serrano Purroy, D.; Modolo, G.; Sorel, C.

    2007-07-01

    A genuine High Active Raffinate (HAR) was produced from small scale PUREX reprocessing of a UO{sub 2} spent fuel solution as feed for a subsequent TODGA/TBP process. In this process, efficient recovery of the trivalent Minor Actinides (MA) actinides could be demonstrated using a hot cell set-up of 32 centrifugal contactor stages. The feed decontamination factors obtained for Am and Cm were in the range of 4 x 10{sup 4} which corresponds to a recovery of more than 99.99 % in the product fraction. Trivalent lanthanides and Y were co-extracted, otherwise only a small part of the Ru ended up in the product. The collected actinide/lanthanide fraction can be used as feed for a SANEX (separation actinides from lanthanides) with some modification of the acidity depending on the extracting molecule. (authors)

  11. Actinide extraction methods

    DOE Patents [OSTI]

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

  12. Catalyst regeneration process including metal contaminants removal

    DOE Patents [OSTI]

    Ganguli, Partha S.

    1984-01-01

    Spent catalysts removed from a catalytic hydrogenation process for hydrocarbon feedstocks, and containing undesired metals contaminants deposits, are regenerated. Following solvent washing to remove process oils, the catalyst is treated either with chemicals which form sulfate or oxysulfate compounds with the metals contaminants, or with acids which remove the metal contaminants, such as 5-50 W % sulfuric acid in aqueous solution and 0-10 W % ammonium ion solutions to substantially remove the metals deposits. The acid treating occurs within the temperature range of 60.degree.-250.degree. F. for 5-120 minutes at substantially atmospheric pressure. Carbon deposits are removed from the treated catalyst by carbon burnoff at 800.degree.-900.degree. F. temperature, using 1-6 V % oxygen in an inert gas mixture, after which the regenerated catalyst can be effectively reused in the catalytic process.

  13. 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.

  14. GANEX: Adaptation of the DIAMEX-SANEX Process for the Group Actinide Separation

    SciTech Connect (OSTI)

    Miguirditchian, M.; Chareyre, L.; Heres, X.; Hill, C.; Baron, P.; Masson, M.

    2007-07-01

    The DIAMEX-SANEX process using the solvent HDEHP/DMDOHEMA/TPH was adapted to manage the separation of neptunium and plutonium along with americium and curium in the second cycle of the GANEX process. Distribution ratios of Np and Pu depending on their initial oxidation states, actinide/lanthanide separation factor and loading capacity of the solvent were measured after batch experiments in order to verify the behaviour of Np and Pu in this process and check their impact on the hydrodynamics. Results show that after some experimental optimizations, the group separation seems possible using this process. A demonstrative test will be carried out on a high active feed in 2008 at CEA Marcoule. (authors)

  15. Actinide Research Quarterly

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

    ARQ A publication of the Glenn T. Seaborg Institute for Transactinium Science. Latest Issue:August 2015 ARQ past issues covers All Issues » submit Actinide Research Quarterly (ARQ) ARQ is a publication of the Glenn T. Seaborg Institute for Transactinium Science, a part of the LANL National Security Education Center. The Actinide Research Quarterly reports on research in actinide science in areas such as process chemistry, metallurgy, surface and separation sciences, atomic and molecular

  16. Actinide Chemistry

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

    Actinide Chemistry Actinide Chemistry Research into alternative forms of energy, especially energy security, is one of the major national security imperatives of this century. Get Expertise David Gallimore Actinide Analytical Chemistry Email Rebecca Chamberlin Actinide Analytical Chemistry Email Josh Smith Chemistry Communications Email Along with the lanthanides, they are often called "the f-elements" because they have valence electrons in the f shell. Actinide chemistry serves a

  17. 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.

  18. Process for removing metals from water

    DOE Patents [OSTI]

    Napier, John M.; Hancher, Charles M.; Hackett, Gail D.

    1989-01-01

    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 flocculating agent, separating precipitate-containing floc, and postfiltering the resultant solution. The postfiltered solution may optionally be eluted through an ion exchange resin to remove residual metal ions.

  19. Actinide Chemistry

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

    Actinide Chemistry Actinide chemistry serves a critical role in addressing global threats Project Description At Los Alamos, scientists are using actinide analytical chemistry to identify and quantify the chemical and isotopic composition of materials. Since the Manhattan Project, such work has supported the Laboratory's mission. Today, actinide analytical chemistry plays a critical role in ensuring the national security of the United States through nuclear science and technology. The scope of

  20. Process for removing sulfur from coal

    DOE Patents [OSTI]

    Aida, Tetsuo; Squires, Thomas G.; Venier, Clifford G.

    1985-02-05

    A process 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.

  1. 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.

  2. Process for removing mercury from aqueous solutions

    DOE Patents [OSTI]

    Googin, John M.; Napier, John M.; Makarewicz, Mark A.; Meredith, Paul F.

    1986-01-01

    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.

  3. 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.

  4. 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)

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Process for selected gas oxide removal by radiofrequency catalysts

    DOE Patents [OSTI]

    Cha, Chang Y.

    1993-01-01

    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.

  10. 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.

  11. Process for removing technetium from iron and other metals

    DOE Patents [OSTI]

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

    1999-01-01

    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.

  12. High Metal Removal Rate Process for Machining Difficult Materials

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

    Currently, most industries treat water to meet standards for direct discharge to surface water. The process includes a primary treatment to remove free oils and solids and ...

  13. Process for removing carbon from uranium

    DOE Patents [OSTI]

    Powell, George L.; Holcombe, Jr., Cressie E.

    1976-01-01

    Carbon contamination is removed from uranium and uranium alloys by heating in inert atmosphere to 700.degree.-1900.degree.C in effective contact with yttrium to cause carbon in the uranium to react with the yttrium. The yttrium is either in direct contact with the contaminated uranium or in indirect contact by means of an intermediate transport medium.

  14. 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.

  15. 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.

  16. 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.

  17. Process for removing pyritic sulfur from bituminous coals

    DOE Patents [OSTI]

    Pawlak, Wanda; Janiak, Jerzy S.; Turak, Ali A.; Ignasiak, Boleslaw L.

    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.

  18. Removal of Process Gas Equipment Marks Portsmouth Site Cleanup Milestone

    Broader source: Energy.gov [DOE]

    PIKE COUNTY, Ohio – The U.S. Department of Energy (DOE) recently met a significant milestone in the Portsmouth Site (PORTS) deactivation effort by removing the final component of process gas...

  19. Removal of Process Gas Equipment Marks Portsmouth Site Cleanup...

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

    ... "The removal of more than 7,000 components brings us to another step in DOE's mission to safely D&D the process buildings and clean up the site," said Dr. Vince Adams, DOE site ...

  20. 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.

  1. 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.

  2. 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.

  3. Mixed monofunctional extractants for trivalent actinide/lanthanide separations: TALSPEAK-MME

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

    Johnson, Aaron T.; Nash, Kenneth L.

    2015-08-20

    The basic features of an f-element extraction process based on a solvent composed of equimolar mixtures of Cyanex-923 (a mixed trialkyl phosphine oxide) and 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) extractants in n-dodecane are investigated in this report. This system, which combines features of the TRPO and TALSPEAK processes, is based on co-extraction of trivalent lanthanides and actinides from 0.1 to 1.0 M HNO3 followed by application of a buffered aminopolycarboxylate solution strip to accomplish a Reverse TALSPEAK selective removal of actinides. This mixed-extractant medium could enable a simplified approach to selective trivalent f-element extraction and actinide partitioning in a singlemore » process. As compared with other combined process applications in development for more compact actinide partitioning processes (DIAMEX-SANEX, GANEX, TRUSPEAK, ALSEP), this combination features only monofunctional extractants with high solubility limits and comparatively low molar mass. Selective actinide stripping from the loaded extractant phase is done using a glycine-buffered solution containing N-(2-hydroxyethyl)ethylenediaminetriacetic acid (HEDTA) or triethylenetetramine-N,N,N',N'',N''',N'''-hexaacetic acid (TTHA). Lastly, the results reported provide evidence for simplified interactions between the two extractants and demonstrate a pathway toward using mixed monofunctional extractants to separate trivalent actinides (An) from fission product lanthanides (Ln).« less

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

    DOE Patents [OSTI]

    Cha, Chang Y.; Boysen, John E.; Branthaver, Jan F.

    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.

  5. Process for removing sulfur from sulfur-containing gases

    DOE Patents [OSTI]

    Rochelle, Gary T.; Jozewicz, Wojciech

    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.

  6. 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.

  7. Advanced Aqueous Separation Systems for Actinide Partitioning

    SciTech Connect (OSTI)

    Nash, Ken; Martin, Leigh; Lumetta, Gregg

    2015-04-02

    One of the most challenging aspects of advanced processing of used nuclear fuel is the separation of transplutonium actinides from fission product lanthanides. This separation is essential if actinide transmutation options are to be pursued in advanced fuel cycles, as lanthanides compete with actinides for neutrons in both thermal and fast reactors, thus limiting efficiency. The separation is difficult because the chemistry of Am3+ and Cm3+ is nearly identical to that of the trivalent lanthanides (Ln3+). The prior literature teaches that two approaches offer the greatest probability of devising a successful group separation process based on aqueous processes: 1) the application of complexing agents containing ligand donor atoms that are softer than oxygen (N, S, Cl-) or 2) changing the oxidation state of Am to the IV, V, or VI state to increase the essential differences between Am and lanthanide chemistry (an approach utilized in the PUREX process to selectively remove Pu4+ and UO22+ from fission products). The latter approach offers the additional benefit of enabling a separation of Am from Cm, as Cm(III) is resistant to oxidation and so can easily be made to follow the lanthanides. The fundamental limitations of these approaches are that 1) the soft(er) donor atoms that interact more strongly with actinide cations than lanthanides form substantially weaker bonds than oxygen atoms, thus necessitating modification of extraction conditions for adequate phase transfer efficiency, 2) soft donor reagents have been seen to suffer slow phase transfer kinetics and hydro-/radiolytic stability limitations and 3) the upper oxidation states of Am are all moderately strong oxidants, hence of only transient stability in media representative of conventional aqueous separations systems. There are examples in the literature of both approaches having been described. However, it is not clear at present that any extant process is sufficiently robust for application at the scale necessary for commercial fuel processing supporting transmutation of transplutonium elements. This research project continued basic themes investigated by this research group during the past decade. In the Fuel Cycle Research and Development program at DOE, the current favorite process for accomplishing the separation of trivalent actinides from fission product lanthanides is the TALSPEAK process. TALSPEAK is a solvent extraction method (developed at Oak Ridge National Lab in the 1960s) based on the combination of a cation exchanging extractant (e.g., HDEHP), an actinide-selective aminopolycarboxylate complexing agent (e.g., DTPA), and a carboxylic acid buffer to control pH in the range of 3-4. Considerable effort has been expended in this research group during the past 8 years to elaborate the details of TALSPEAK in the interest of developing improved approaches to the operation of TALSPEAK-like systems. In this project we focused on defining aggregation phenomena in conventional TALSPEAK separations, on supporting the development of Advanced TALSPEAK processes, on profiling the aqueous complexation kinetics of lanthanides in TALSPEAK relevant aqueous media, on the design of new diglycolamide and N-donor extractants, and on characterizing cation-cation complexes of pentavalent actinides.

  8. 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.

  9. 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.

  10. 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.

  11. Process for removing an organic compound from water

    DOE Patents [OSTI]

    Baker, Richard W.; Kaschemekat, Jurgen; Wijmans, Johannes G.; Kamaruddin, Henky D.

    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.

  12. 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.

  13. 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.

  14. Metal chelate process to remove pollutants from fluids

    DOE Patents [OSTI]

    Chang, Shih-Ger T.

    1994-01-01

    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.

  15. Process for removing sulfate anions from waste water

    DOE Patents [OSTI]

    Nilsen, David N.; Galvan, Gloria J.; Hundley, Gary L.; Wright, John B.

    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.

  16. RECOVERY OF ACTINIDES FROM AQUEOUS NITRIC ACID SOLUTIONS

    DOE Patents [OSTI]

    Ader, M.

    1963-11-19

    A process of recovering actinides is presented. Tetravalent actinides are extracted from rare earths in an aqueous nitric acid solution with a ketone and back-extracted from the ketone into an aqueous medium. The aqueous actinide solution thus obtained, prior to concentration by boiling, is sparged with steam to reduce its ketone to a maximum content of 3 grams per liter. (AEC)

  17. 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.

  18. Managing Inventories of Heavy Actinides (Conference) | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    SciTech Connect Search Results Conference: Managing Inventories of Heavy Actinides Citation Details In-Document Search Title: Managing Inventories of Heavy Actinides The Department of Energy (DOE) has stored a limited inventory of heavy actinides contained in irradiated targets, some partially processed, at the Savannah River Site (SRS) and Oak Ridge National Laboratory (ORNL). The 'heavy actinides' of interest include plutonium, americium, and curium isotopes; specifically 242Pu and 244Pu,

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

    SciTech Connect (OSTI)

    Harris, J.L.

    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.

  20. 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.

  1. 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.

  2. 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.; Esimantovskiy, Vyacheslav M.; Lazarev, Leonard N.; Dzekun, Evgeniy G.; Romanovskiy, Valeriy N.; Todd, Terry A.; Brewer, Ken N.; Herbst, Ronald S.; Law, Jack D.

    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.

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

    DOE Patents [OSTI]

    Borole, Abhijeet P.; Hamilton, Choo Y.

    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.

  4. Process for removal of sulfur compounds from fuel gases

    DOE Patents [OSTI]

    Moore, Raymond H.; Stegen, Gary E.

    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.

  5. Process for removal of hazardous air pollutants from coal

    DOE Patents [OSTI]

    Akers, David J.; Ekechukwu, Kenneth N.; Aluko, Mobolaji E.; Lebowitz, Howard E.

    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.

  6. 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.

  7. Actinide-ion sensor

    DOE Patents [OSTI]

    Li, Shelly X; Jue, Jan-fong; Herbst, Ronald Scott; Herrmann, Steven Douglas

    2015-01-13

    An apparatus for the real-time, in-situ monitoring of actinide-ion concentrations. A working electrolyte is positioned within the interior of a container. The working electrolyte is separated from a reference electrolyte by a separator. A working electrode is at least partially in contact with the working electrolyte. A reference electrode is at least partially in contact with the reference electrolyte. A voltmeter is electrically connected to the working electrode and the reference electrode. The working electrolyte comprises an actinide-ion of interest. The separator is ionically conductive to the actinide-ion of interest. The separator comprises an actinide, Zr, and Nb. Preferably, the actinide of the separator is Am or Np, more preferably Pu. In one embodiment, the actinide of the separator is the actinide of interest. In another embodiment, the separator further comprises P and O.

  8. Actinide halide complexes

    DOE Patents [OSTI]

    Avens, Larry R.; Zwick, Bill D.; Sattelberger, Alfred P.; Clark, David L.; Watkin, John G.

    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.

  9. 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.

  10. 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.

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

    SciTech Connect (OSTI)

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

    2010-02-03

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

  12. PREPARATION OF ACTINIDE-ALUMINUM ALLOYS

    DOE Patents [OSTI]

    Moore, R.H.

    1962-09-01

    BS>A process is given for preparing alloys of aluminum with plutonium, uranium, and/or thorium by chlorinating actinide oxide dissolved in molten alkali metal chloride with hydrochloric acid, chlorine, and/or phosgene, adding aluminum metal, and passing air and/or water vapor through the mass. Actinide metal is formed and alloyed with the aluminum. After cooling to solidification, the alloy is separated from the salt. (AEC)

  13. Appendix SOTERM: Actinide Chemistry Source Term

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

    SOTERM-2014 Actinide Chemistry Source Term United States Department of Energy Waste Isolation Pilot Plant Carlsbad Field Office Carlsbad, New Mexico Compliance Recertification Application 2014 Appendix SOTERM-2014 Actinide Chemistry Source Term Table of Contents SOTERM-1.0 Introduction SOTERM-2.0 Expected WIPP Repository Conditions, Chemistry, and Processes SOTERM-2.1 Ambient Geochemical Conditions SOTERM-2.2 Repository Conditions SOTERM-2.2.1 Repository Pressure SOTERM-2.2.2 Repository

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

    DOE Patents [OSTI]

    Van Der Sluys, William G.; Burns, Carol J.; Smith, David C.

    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.

  15. 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.

  16. 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.

  17. 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].

  18. Method for preparing actinide nitrides

    DOE Patents [OSTI]

    Bryan, G.H.; Cleveland, J.M.; Heiple, C.R.

    1975-12-01

    Actinide nitrides, and particularly plutonium and uranium nitrides, are prepared by reacting an ammonia solution of an actinide compound with an ammonia solution of a reactant or reductant metal, to form finely divided actinide nitride precipitate which may then be appropriately separated from the solution. The actinide nitride precipitate is particularly suitable for forming nuclear fuels.

  19. 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.

  20. Fluidized bed gasification ash reduction and removal process

    DOE Patents [OSTI]

    Schenone, Carl E.; Rosinski, Joseph

    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.

  1. Removal of heavy metal ions from oil shale beneficiation process water by ferrite process

    SciTech Connect (OSTI)

    Mehta, R.K.; Zhang, L.; Lamont, W.E.; Schultz, C.W.

    1991-12-31

    The ferrite process is an established technique for removing heavy metals from waste water. Because the process water resulting from oil shale beneficiation falls into the category of industrial waste water, it is anticipated that this process may turn out to be a potential viable treatment for oil shale beneficiation process water containing many heave metal ions. The process is chemoremedial because not only effluent water comply with quality standards, but harmful heavy metals are converted into a valuable, chemically stable by-product known as ferrite. These spinel ferrites have magnetic properties, and therefore can be use in applications such as magnetic marker, ferrofluid, microwave absorbing and scavenging material. Experimental results from this process are presented along with results of treatment technique such as sulfide precipitation.

  2. Removal of heavy metal ions from oil shale beneficiation process water by ferrite process

    SciTech Connect (OSTI)

    Mehta, R.K.; Zhang, L.; Lamont, W.E.; Schultz, C.W. . Mineral Resources Inst.)

    1991-01-01

    The ferrite process is an established technique for removing heavy metals from waste water. Because the process water resulting from oil shale beneficiation falls into the category of industrial waste water, it is anticipated that this process may turn out to be a potential viable treatment for oil shale beneficiation process water containing many heave metal ions. The process is chemoremedial because not only effluent water comply with quality standards, but harmful heavy metals are converted into a valuable, chemically stable by-product known as ferrite. These spinel ferrites have magnetic properties, and therefore can be use in applications such as magnetic marker, ferrofluid, microwave absorbing and scavenging material. Experimental results from this process are presented along with results of treatment technique such as sulfide precipitation.

  3. 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.

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

    DOE Patents [OSTI]

    Kowalczyk, Dennis C.; Bricklemyer, Bruce A.; Svoboda, Joseph J.

    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.

  5. 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.

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

    DOE Patents [OSTI]

    Pennline, Henry W.; Granite, Evan J.; Freeman, Mark C.; Hargis, Richard A.; O'Dowd, William J.

    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.

  7. 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.

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

    DOE Patents [OSTI]

    Carl, Daniel E.

    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.

  9. CNG process, a new approach to physical-absorption acid-gas removal

    SciTech Connect (OSTI)

    Hise, R.E.; Massey, L.G.; Adler, R.J.; Brosilow, C.B.; Gardner, N.C.; Brown, W.R.; Cook, W.J.; Petrik, M.

    1982-01-01

    The CNG acid gas removal process embodies three novel features: (1) scrubbing with liquid carbon dioxide to remove all sulfurous molecules and other trace contaminants; (2) triple-point crystallization of carbon dioxide to concentrate sulfurous molecules and produce pure carbon dioxide; and (3) absorption of carbon dioxide with a slurry of solid carbon dioxide in organic carrier liquid. The CNG process is discussed and contrasted with existing acid gas removal technology as represented by the Benfield, Rectisol, and Selexol acid gas removal processes.

  10. Nonaqueous actinide hydride dissolution and production of actinide $beta$- diketonates

    DOE Patents [OSTI]

    Crisler, L.R.

    1975-11-11

    Actinide beta-diketonate complex molecular compounds are produced by reacting a beta-diketone compound with a hydride of the actinide material in a mixture of carbon tetrachloride and methanol. (auth)

  11. SOLVENT FOR EXTRACTING ACTINIDE SALTS

    DOE Patents [OSTI]

    Kaplan, L.

    1959-10-27

    BS>A mixture of hexone and 2-hexylpyridine can be used for the selective extraction of actinide values.

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

    Office of Scientific and Technical Information (OSTI)

    The mercury can be recovered in pure form from the sorbents via additional processing. Inventors: Borole, Abhijeet P. 1 ; Hamilton, Choo Y. 1 + Show Author Affiliations ...

  13. 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.

  14. 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.

  15. Processes to remove acid forming gases from exhaust gases

    DOE Patents [OSTI]

    Chang, Shih-Ger

    1994-01-01

    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.degree. and 100.degree. 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 environ-mentally acceptable levels. Urea or sulfamic acid are preferred, especially sulfamic acid, and step (C) is not necessary or performed.

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

    DOE Patents [OSTI]

    Krauter, Paula A. W.; Krauter, Gordon W.

    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.

  17. 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.

  18. 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.

  19. Method and system for the removal of oxides of nitrogen and sulfur from combustion processes

    DOE Patents [OSTI]

    Walsh, John V.

    1987-12-15

    A process for removing oxide contaminants from combustion gas, and employing a solid electrolyte reactor, includes: (a) flowing the combustion gas into a zone containing a solid electrolyte and applying a voltage and at elevated temperature to thereby separate oxygen via the solid electrolyte, (b) removing oxygen from that zone in a first stream and removing hot effluent gas from that zone in a second stream, the effluent gas containing contaminant, (c) and pre-heating the combustion gas flowing to that zone by passing it in heat exchange relation with the hot effluent gas.

  20. Actinide targets for the synthesis of super-heavy elements

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

    Roberto, J.; Alexander, Charles W.; Boll, Rose Ann; Ezold, Julie G.; Felker, Leslie Kevin; Rykaczewski, Krzysztof Piotr; Hogle, Susan L.

    2015-06-18

    Since 2000, six new super-heavy elements with atomic numbers 113 through 118 have been synthesized in hot fusion reactions of 48Ca beams on actinide targets. These target materials, including 242Pu, 244Pu, 243Am, 245Cm, 248Cm, 249Cf, and 249Bk, are available in very limited quantities and require specialized production and processing facilities resident in only a few research centers worldwide. This report describes the production and chemical processing of heavy actinide materials for super-heavy element research, current availabilities of these materials, and related target fabrication techniques. The impact of actinide materials in super-heavy element discovery is reviewed, and strategies for enhancing themore » production of rare actinides including 249Bk, 251Cf, and 254Es are described.« less

  1. 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.

  2. Actinide Ion Sensor For Pyroprocess Monitoring - Energy Innovation Portal

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

    Actinide Ion Sensor For Pyroprocess Monitoring DOE Grant Recipients Idaho National Laboratory Contact GRANT About This Technology Technology Marketing Summary Idaho National Laboratory (INL) has created a novel method and apparatus for monitoring plutonium concentration during pyroprocessing to ensure that the refining process is efficient at collecting actinides. Unlike existing solutions, which require complex, time-consuming remote testing, this unique sensor is capable of providing real-time

  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. 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.

  5. EM Celebrates Milestone with Removal of Last Waste Tank at Separations Process Research Unit

    Broader source: Energy.gov [DOE]

    NISKAYUNA, N.Y. – EM recently marked a notable milestone at the Separations Process Research Unit (SPRU) when workers removed the last of seven large waste storage tanks from a vault and shipped it to an offsite low-level radioactive waste disposal facility.

  6. 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.

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

    DOE Patents [OSTI]

    Huang, Hann S.; Sather, Norman F.

    1988-01-01

    A process for reducing the amount of halogens and halogen acids in high temperature combustion gases 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.

  8. 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.

  9. Process for the removal of radium from acidic solutions containing same

    DOE Patents [OSTI]

    Scheitlin, F.M.

    The invention is a process for the removal of radium from acidic aqueous solutions. In one aspect, the invention is a process for removing radium from an inorganic-acid solution. The process comprises contacting the solution with coal fly ash to effect adsorption of the radium on the ash. The radium-containing ash then is separated from the solution. The process is simple, comparatively inexpensive, and efficient. High radium-distribution coefficients are obtained even at room temperature. Coal fly ash is an inexpensive, acid-resistant, high-surface-area material which is available in large quantities throughout the United States. The invention is applicable, for example, to the recovery of /sup 226/Ra from nitric acid solutions which have been used to leach radium from uranium-mill tailings.

  10. Actinide Burning in CANDU Reactors

    SciTech Connect (OSTI)

    Hyland, B.; Dyck, G.R.

    2007-07-01

    Actinide burning in CANDU reactors has been studied as a method of reducing the actinide content of spent nuclear fuel from light water reactors, and thereby decreasing the associated long term decay heat load. In this work simulations were performed of actinides mixed with natural uranium to form a mixed oxide (MOX) fuel, and also mixed with silicon carbide to form an inert matrix (IMF) fuel. Both of these fuels were taken to a higher burnup than has previously been studied. The total transuranic element destruction calculated was 40% for the MOX fuel and 71% for the IMF. (authors)

  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. 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.

  13. 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.

  14. Environmental research on actinide elements

    SciTech Connect (OSTI)

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

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

    DOE Patents [OSTI]

    Doherty, Joseph P.; Marek, James C.

    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.

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

    DOE Patents [OSTI]

    Vogel, Gerhard John; Jonke, Albert A.; Snyder, Robert B.

    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.

  17. 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.

  18. 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.

  19. METAL REMOVAL FROM PROCESS AND STORMWATER DISCHARGES BY CONSTRUCTED TREATMENT WETLANDS

    SciTech Connect (OSTI)

    NELSON, ERIC

    2004-11-02

    The A-01 NPDES outfall at the Savannah River Site receives process wastewater and stormwater which passes through a wetland treatment system (WTS) prior to discharge. The overall objective of our research is to better understand the mechanisms of operation of the A-01 WTS in order to provide better input to the design of future systems. The system is a vegetated surface flow wetland and has a retention time of approximately 48 hours. Sampling conducted during the fourth year of operation validated continued wetland performance, and assessed the fate of a larger suite of metals present in the water. Copper and mercury removal efficiencies were still very high, both in excess of 80 per cent removal from the water after passage through the wetland system. Lead removal from the water by the system was 83 per cent, zinc removal was 60 per cent, and nickel was generally unaffected. Nitrates entering into the wetland cells are almost immediately removed from the water column and generally no nitrates are discharged from the A cells. The wetland cells are very anaerobic and the sediments have negative redox potentials. As a result, manganese and iron mineral phases in the sediments have been reduced to soluble forms and increase in the water during passage through the wetland system. Dissolved organic carbon in the water column is also increased by the system and reduces toxicity of the effluent. Operation and maintenance of the system is minimal, and consists of checking for growth of the vegetation and free flow of the water through the system.

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

    DOE Patents [OSTI]

    Rochelle, Gary T.; Chang, John C. S.

    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.

  1. Method for extracting lanthanides and actinides from acid solutions

    DOE Patents [OSTI]

    Horwitz, E. Philip; Kalina, Dale G.; Kaplan, Louis; Mason, George W.

    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.

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

    DOE Patents [OSTI]

    Olsen, Richard S.; Banks, John T.

    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.

  3. 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.

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

    DOE Patents [OSTI]

    King, C. Judson; MacKenzie, Patricia D.

    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.

  5. PROCESS OF REMOVING PLUTONIUM VALUES FROM SOLUTION WITH GROUP IVB METAL PHOSPHO-SILICATE COMPOSITIONS

    DOE Patents [OSTI]

    Russell, E.R.; Adamson, A.W.; Schubert, J.; Boyd, G.E.

    1957-10-29

    A process for separating plutonium values from aqueous solutions which contain the plutonium in minute concentrations is described. These values can be removed from an aqueous solution by taking an aqueous solution containing a salt of zirconium, titanium, hafnium or thorium, adding an aqueous solution of silicate and phosphoric acid anions to the metal salt solution, and separating, washing and drying the precipitate which forms when the two solutions are mixed. The aqueous plutonium containing solution is then acidified and passed over the above described precipi-tate causing the plutonium values to be adsorbed by the precipitate.

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

    DOE Patents [OSTI]

    Frost, Albert C.; Yang, Chang-Lee

    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.

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

    DOE Patents [OSTI]

    Frost, Albert C.; Yang, Chang-lee

    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.

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

    DOE Patents [OSTI]

    Klug, Frederic Joseph; DeCarr, Sylvia Marie

    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.

  9. 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.; Cassayre, L.

    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)

  10. 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

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

    DOE Patents [OSTI]

    Hartman, Alan D.; Oden, Laurance L.; White, Jack C.

    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.

  12. CNG Acid gas removal process. Technical progress report 2, 1 December 1983-29 February 1984

    SciTech Connect (OSTI)

    Auyang, L.; Liu, Y.C.

    1985-01-01

    Development work on the CNG acid gas removal process under DOE Contract No. AC21-83MC20230 continued during the period December 1, 1983 through February 29, 1984. Two tasks were active during this time: Task 1 hydrogen sulfide absorber (design and construction of hydrogen sulfide absorber); and Task 4 technology transfer. Within Subtask 1.1, the flow sheet of the integrated hydrogen sulfide absorber and the carbon dioxide triple-point crystallizer is reviewed. Control objectives of the hydrogen sulfide absorber and control strategies were established and are discussed. Within Subtask 1.2, detailed engineering designs have been completed for the absorption column, the light ends flasher, cooler/condenser, and the liquid carbon dioxide surge tank. This equipment is now in various stages of construction. Other process equipment specified and placed on order includes the main gas compressor, recycle light ends gas compressor, liquid carbon dioxide absorbent pump, and the concentrated acid gas stream pump. Within Task 4, two papers discussing the CNG acid gas removal technology have been prepared. One paper will be presented in the Acid and Sour Gas Symposium at the AIChE Winter National Meeting, Atlanta, Georgia. The other paper will be presented at the Eleventh Energy Technology Conference, Washington, DC. 10 figs., 5 tabs.

  13. 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.

  14. SOLUBILIZATION OF ACTINIDE METAL-CONTAINING SLAG

    DOE Patents [OSTI]

    Hopkins, H.H. Jr.

    1959-08-01

    This patent relates to solubilization of the actinide rare earths valves contained in the slag materials resulting from the reduction of actinide salts, such as plutonium tetrafluoride. According to the invention the slag is subjected to a high temperature chloridizing roast, preferably from the reduction of actinide salts, such as plutonium tetrafluoride. According to the invention the slag is subjected to a high temperature chloridizing roast, preferably at about 700 deg C with gaseous hydrogen chloride, until the actinides within the slag are substantially convented to the chlorides. The resultant chlorinated actinides are then leached from the cooled roasted mass by treating with aqueous 0.01 M nitric acid.

  15. Separations of actinides, lanthanides and other metals

    DOE Patents [OSTI]

    Smith, Barbara F.; Jarvinen, Gordon D.; Ensor, Dale D.

    1995-01-01

    An organic extracting solution comprised of a bis(acylpyrazolone or a substituted bis(acylpyrazolone) and an extraction method useful for separating certain elements of the actinide series of the periodic table having a valence of four from one other, and also from one or more of the substances in a group consisting of hexavalent actinides, trivalent actinides, trivalent lanthanides, trivalent iron, trivalent aluminum, divalent metals, and monovalent metals and also from one or more of the substances in a group consisting of hexavalent actinides, trivalent actinides, trivalent lanthanides, trivalent iron, trivalent aluminum, divalent metals, and monovalent metals and also useful for separating hexavalent actinides from one or more of the substances in a group consisting of trivalent actinides, trivalent lanthanides, trivalent iron, trivalent aluminum, divalent metals, and monovalent metals.

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

    DOE Patents [OSTI]

    Chang, Shih-Ger; Liu, David K.

    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.

  17. 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.

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

    SciTech Connect (OSTI)

    Rizzo, Jeffrey J.

    2010-04-30

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

  19. 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.

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

    DOE Patents [OSTI]

    Basu, Arunabha; Meyer, Howard S.; Lynn, Scott; Leppin, Dennis; Wangerow, James R.

    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.

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

    DOE Patents [OSTI]

    Comolli, Alfred G.

    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. Apparatus and process for removing a predetermined portion of reflective material from mirror

    DOE Patents [OSTI]

    Perry, Stephen J.; Steinmetz, Lloyd L.

    1994-01-01

    An apparatus and process are disclosed for removal of a stripe of soft reflective material of uniform width from the surface of a mirror by using a blade having a large included angle to inhibit curling of the blade during the cutting operation which could result in damage to the glass substrate of the mirror. The cutting blade is maintained at a low blade angle with respect to the mirror surface to produce minimal chipping along the cut edge and to minimize the force exerted on the coating normal to the glass surface which could deform the flat mirror. The mirror is mounted in a cutting mechanism containing a movable carriage on which the blade is mounted to provide very accurate straightness of the travel of the blade along the mirror.

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

    DOE Patents [OSTI]

    Googin, John M.; Napier, John M.; Travaglini, Michael A.

    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.

  4. 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.

  5. Flotation process for removal of precipitates from electrochemical chromate reduction unit

    DOE Patents [OSTI]

    DeMonbrun, James R.; Schmitt, Charles R.; Williams, Everett H.

    1976-01-01

    This invention is an improved form of a conventional electrochemical process for removing hexavalent chromium or other metal-ion contaminants from cooling-tower blowdown water. In the conventional process, the contaminant is reduced and precipitated at an iron anode, thus forming a mixed precipitate of iron and chromium hydroxides, while hydrogen being evolved copiously at a cathode is vented from the electrochemical cell. In the conventional process, subsequent separation of the fine precipitate has proved to be difficult and inefficient. In accordance with this invention, the electrochemical operation is conducted in a novel manner permitting a much more efficient and less expensive precipitate-recovery operation. That is, the electrochemical operation is conducted under an evolved-hydrogen partial pressure exceeding atmospheric pressure. As a result, most of the evolved hydrogen is entrained as bubbles in the blowdown in the cell. The resulting hydrogen-rich blowdown is introduced to a vented chamber, where the entrained hydrogen combines with the precipitate to form a froth which can be separated by conventional techniques. In addition to the hydrogen, two materials present in most blowdown act as flotation promoters for the precipitate. These are (1) air, with which the blowdown water becomes saturated in the course of normal cooling-tower operation, and (2) surfactants which commonly are added to cooling-tower recirculating-water systems to inhibit the growth of certain organisms or prevent the deposition of insoluble particulates.

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

    DOE Patents [OSTI]

    Chang, Shih-Ger; Liu, David K.

    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.

  7. Composition, process, and apparatus, for removal of water and silicon mu-oxides from chlorosilanes

    DOE Patents [OSTI]

    Tom, Glenn M.; McManus, James V.

    1991-10-15

    A scavenger composition having utility for removal of water and silicon mu-oxide impurities from chlorosilanes, such scavenger composition comprising: (a) a support; and (b) associated with the support, one or more compound(s) selected from the group consisting of compounds of the formula: R.sub.a-x MCl.sub.x wherein: M is a metal selected from the group consisting of the monovalent metals lithium, sodium, and potassium; the divalent metals magnesium, strontium, barium, and calcium; and the trivalent metal aluminum; R is alkyl; a is a number equal to the valency of metal M; and x is a number having a value from 0 to a, inclusive; and wherein said compound(s) of the formula R.sub.a-x MCl.sub.x have been activated for impurity-removal service by a reaction scheme selected from those of the group consisting of: (i) reaction of such compound(s) with hydrogen chloride to form a first reaction product therefrom, followed by reaction of the first reaction product with a chlorosilane of the formula: SiH.sub.4"y Cl.sub.y, wherein y is a number having a value of from 1 to 3, inclusive; and (ii) reaction of such compound(s) with a chlorosilane of the formula: SiH.sub.4-y Cl.sub.y wherein y is a number having a value of 1 to 3, inclusive. A corresponding method of making the scavenger composition, and of purifying a chlorosilane which contains oxygen and silicon mu-oxide impurities, likewise are disclosed, together with a purifier apparatus, in which a bed of the scavenger composition is disposed. The composition, purification process, and purifier apparatus of the invention have utility in purifying gaseous chlorosilanes which are employed in the semiconductor industry as silicon source reagents for forming epitaxial silicon layers.

  8. Process for removal of water and silicon mu-oxides from chlorosilanes

    DOE Patents [OSTI]

    Tom, Glenn M.; McManus, James V.

    1992-03-10

    A scavenger composition having utility for removal of water and silicon mu-oxide impurities from chlorosilanes, such scavenger composition comprising: (a) a support; and (b) associated with the support, one or more compound(s) selected from the group consisting of compounds of the formula: R.sub.a-x MCl.sub.x wherein: M is a metal selected from the group consisting of the monovalent metals lithium, sodium, and potassium; the divalent metals magnesium, strontium, barium, and calcium; and the trivalent metal aluminum; R is alkyl; a is a number equal to the valency of metal M; and x is a number having a value of from 0 to a, inclusive; and wherein said compound(s) of the formula R.sub.a-x MCl.sub.x have been activated for impurity-removal service by a reaction scheme selected from those of the group consisting of: (i) reaction of such compound(s) with hydrogen chloride to form a first reaction product therefrom, followed by reaction of the first reaction product with a chlorosilane of the formula: SiH.sub.4-y Cl.sub.y, wherein y is a number having a value of from 1 to 3, inclusive; and (ii) reaction of such compound(s) with a chlorosilane of the formula: SiH.sub.4-y Cl.sub.y wherein y is a number having a value of 1 to 3, inclusive. A corresponding method of making the scavenger composition, and of purifying a chlorosilane which contains oxygen and silicon mu-oxide impurities, likewise are disclosed, together with a purifier apparatus, in which a bed of the scavenger composition is disposed. The composition, purification process, and purifier apparatus of the invention have utility in purifying gaseous chlorosilanes which are employed in the semiconductor industry as silicon source reagents for forming epitaxial silicon layers.

  9. 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.

  10. Covalent Bonding in Actinide Sandwich Molecules

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

    Covalent Bonding in Actinide Sandwich Molecules Print Glenn Seaborg was one of the first scientists to recognize that differences in the degree of covalent bonding in lanthanide and actinide compounds could have profound consequences for their unique chemical reactivity and physical properties. Now, researchers working at ALS Beamline 11.0.2 have found evidence for unexpected bonding interactions in two organometallic actinide "sandwich" complexes that have been lightning rods in

  11. Covalent Bonding in Actinide Sandwich Molecules

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

    Covalent Bonding in Actinide Sandwich Molecules Print Glenn Seaborg was one of the first scientists to recognize that differences in the degree of covalent bonding in lanthanide and actinide compounds could have profound consequences for their unique chemical reactivity and physical properties. Now, researchers working at ALS Beamline 11.0.2 have found evidence for unexpected bonding interactions in two organometallic actinide "sandwich" complexes that have been lightning rods in

  12. Covalent Bonding in Actinide Sandwich Molecules

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

    Covalent Bonding in Actinide Sandwich Molecules Print Glenn Seaborg was one of the first scientists to recognize that differences in the degree of covalent bonding in lanthanide and actinide compounds could have profound consequences for their unique chemical reactivity and physical properties. Now, researchers working at ALS Beamline 11.0.2 have found evidence for unexpected bonding interactions in two organometallic actinide "sandwich" complexes that have been lightning rods in

  13. Covalent Bonding in Actinide Sandwich Molecules

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

    Covalent Bonding in Actinide Sandwich Molecules Print Glenn Seaborg was one of the first scientists to recognize that differences in the degree of covalent bonding in lanthanide and actinide compounds could have profound consequences for their unique chemical reactivity and physical properties. Now, researchers working at ALS Beamline 11.0.2 have found evidence for unexpected bonding interactions in two organometallic actinide "sandwich" complexes that have been lightning rods in

  14. Covalent Bonding in Actinide Sandwich Molecules

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

    Covalent Bonding in Actinide Sandwich Molecules Covalent Bonding in Actinide Sandwich Molecules Print Wednesday, 28 May 2014 00:00 Glenn Seaborg was one of the first scientists to recognize that differences in the degree of covalent bonding in lanthanide and actinide compounds could have profound consequences for their unique chemical reactivity and physical properties. Now, researchers working at ALS Beamline 11.0.2 have found evidence for unexpected bonding interactions in two organometallic

  15. Covalent Bonding in Actinide Sandwich Molecules

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

    Covalent Bonding in Actinide Sandwich Molecules Print Glenn Seaborg was one of the first scientists to recognize that differences in the degree of covalent bonding in lanthanide and actinide compounds could have profound consequences for their unique chemical reactivity and physical properties. Now, researchers working at ALS Beamline 11.0.2 have found evidence for unexpected bonding interactions in two organometallic actinide "sandwich" complexes that have been lightning rods in

  16. Covalent Bonding in Actinide Sandwich Molecules

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

    Covalent Bonding in Actinide Sandwich Molecules Print Glenn Seaborg was one of the first scientists to recognize that differences in the degree of covalent bonding in lanthanide and actinide compounds could have profound consequences for their unique chemical reactivity and physical properties. Now, researchers working at ALS Beamline 11.0.2 have found evidence for unexpected bonding interactions in two organometallic actinide "sandwich" complexes that have been lightning rods in

  17. 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.

  18. Ceramic composition for immobilization of actinides

    DOE Patents [OSTI]

    Ebbinghaus, Bartley B.; Van Konynenburg, Richard A.; Vance, Eric R.; Stewart, Martin W.; Jostsons, Adam; Allender, Jeffrey S.; Rankin, David Thomas

    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.

  19. Covalent Bonding in Actinide Sandwich Molecules

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

    ALS Beamline 11.0.2 have found evidence for unexpected bonding interactions in two organometallic actinide "sandwich" complexes that have been lightning rods in discussions of...

  20. 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.

  1. 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.

  2. 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.

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

    SciTech Connect (OSTI)

    Craig S. Turchi

    2000-09-29

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

  4. Prompt fission neutron spectra of actinides

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

    Capote, R.; Chen, Y. -J.; Hambsch, F. -J.; Kornilov, N. V.; Lestone, J. P.; Litaize, O.; Morillon, B.; Neudecker, D.; Oberstedt, S.; Ohsawa, T.; et al

    2016-01-06

    Here, the energy spectrum of prompt neutrons emitted in fission (PFNS) plays a very important role in nuclear science and technology. A Coordinated Research Project (CRP) "Evaluation of Prompt Fission Neutron Spectra of Actinides" was established by the IAEA Nuclear Data Section in 2009, with the major goal to produce new PFNS evaluations with uncertainties for actinide nuclei.

  5. Method of removing the effects of electrical shorts and shunts created during the fabrication process of a solar cell

    DOE Patents [OSTI]

    Nostrand, Gerald E.; Hanak, Joseph J.

    1979-01-01

    A method of removing the effects of electrical shorts and shunts created during the fabrication process and improving the performance of a solar cell with a thick film cermet electrode opposite to the incident surface by applying a reverse bias voltage of sufficient magnitude to burn out the electrical shorts and shunts but less than the break down voltage of the solar cell.

  6. 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).

  7. Separation of actinides from lanthanides

    DOE Patents [OSTI]

    Smith, Barbara F.; Jarvinen, Gordon D.; Ryan, Robert R.

    1989-01-01

    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. 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).

  8. Actinide ion sensor for pyroprocess monitoring

    DOE Patents [OSTI]

    Jue, Jan-fong; Li, Shelly X.

    2014-06-03

    An apparatus for real-time, in-situ monitoring of actinide ion concentrations which comprises a working electrode, a reference electrode, a container, a working electrolyte, a separator, a reference electrolyte, and a voltmeter. The container holds the working electrolyte. The voltmeter is electrically connected to the working electrode and the reference electrode and measures the voltage between those electrodes. The working electrode contacts the working electrolyte. The working electrolyte comprises an actinide ion of interest. The reference electrode contacts the reference electrolyte. The reference electrolyte is separated from the working electrolyte by the separator. The separator contacts both the working electrolyte and the reference electrolyte. The separator is ionically conductive to the actinide ion of interest. The reference electrolyte comprises a known concentration of the actinide ion of interest. The separator comprises a beta double prime alumina exchanged with the actinide ion of interest.

  9. 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.

  10. 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.

  11. Rapid method to determine actinides and 89/90Sr in limestone and marble samples

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

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

    2016-04-12

    A new method for the determination of actinides and radiostrontium in limestone and marble samples has been developed that utilizes a rapid sodium hydroxide fusion to digest the sample. Following rapid pre-concentration steps to remove sample matrix interferences, the actinides and 89/90Sr are separated using extraction chromatographic resins and measured radiometrically. The advantages of sodium hydroxide fusion versus other fusion techniques will be discussed. This approach has a sample preparation time for limestone and marble samples of <4 hours.

  12. 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.

  13. 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.

  14. 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

    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)

  15. 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.

  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. Fusion Techniques for the Oxidation of Refractory Actinide Oxides

    SciTech Connect (OSTI)

    Rudisill, T.S.

    1999-04-15

    Small-scale experiments were performed to demonstrate the feasibility of fusing refractory actinide oxides with a series of materials commonly used to decompose minerals, glasses, and other refractories as a pretreatment to dissolution and subsequent recovery operations. In these experiments, 1-2 g of plutonium or neptunium oxide (PuO2 or NpO2) were calcined at 900 degrees Celsius, mixed and heated with the fusing reagent(s), and dissolved. For refractory PuO2, the most effective material tested was a lithium carbonate (Li2CO3)/sodium tetraborate (Na2B4O7) mixture which aided in the recovery of 90 percent of the plutonium. The fused product was identified as a lithium plutonate (Li3PuO4) by x-ray diffraction. The use of a Li2CO3/Na2B4O7 mixture to solubilize high-fired NpO2 was not as effective as demonstrated for refractory PuO2. In a small-scale experiment, 25 percent of the NpO2 was oxidized to a neptunium (VI) species that dissolved in nitric acid. The remaining neptunium was then easily recovered from the residue by fusing with sodium peroxide (Na2O2). Approximately 70 percent of the neptunium dissolved in water to yield a basic solution of neptunium (VII). The remainder was recovered as a neptunium (VI) solution by dissolving the residue in 8M nitric acid. In subsequent experiments with Na2O2, the ratio of neptunium (VII) to (VI) was shown to be a function of the fusion temperature, with higher temperatures (greater than approximately 400 degrees C) favoring the formation of neptunium (VII). The fusion of an actual plutonium-containing residue with Na2O2 and subsequent dissolution was performed to demonstrate the feasibility of a pretreatment process on a larger scale. Sodium peroxide was chosen due to the potential of achieving higher actinide recoveries from refractory materials. In this experiment, nominally 10 g of a graphite-containing residue generated during plutonium casting operations was initially calcined to remove the graphite. Removal of combustible material prior to a large-scale fusion with Na2O2 is needed due to the large amount of heat liberated during oxidation. Two successive fusions using the residue from the calcination and the residue generated from the initial dissolution allowed recovery of 98 percent of the plutonium. The fusion of the residue following the first dissolution was performed at a higher temperature (600 degrees Celsius versus 450 degrees Celsius during the first fusion). The ability to recover most of the remaining plutonium from the residue suggest the oxidation efficiency of the Na2O2 fusion improves with higher temperatures similar to results observed with NpO2 fusion.

  18. 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)

  19. PROCESSING ALTERNATIVES FOR DESTRUCTION OF TETRAPHENYLBORATE

    SciTech Connect (OSTI)

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

    2007-02-27

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

  20. Factors influencing the transport of actinides in the groundwater environment. Final report

    SciTech Connect (OSTI)

    Sheppard, J.C.; Kittrick, J.A.

    1983-07-31

    This report summarizes investigations of factors that significantly influence the transport of actinide cations in the groundwater environment. Briefly, measurements of diffusion coefficients for Am(III), Cm(III), and Np(V) in moist US soils indicated that diffusion is negligible compared to mass transport in flowing groundwater. Diffusion coefficients do, however, indicate that, in the absence of flowing water, actinide elements will migrate only a few centimeters in a thousand years. The remaining investigations were devoted to the determination of distribution ratios (K/sub d/s) for representative US soils, factors influencing them, and chemical and physical processes related to transport of actinides in groundwaters. The computer code GARD was modified to include complex formation to test the importance of humic acid complexing on the rate of transport of actinides in groundwaters. Use of the formation constant and a range of humic acid, even at rather low concentrations of 10/sup -5/ to 10/sup -6/ molar, significantly increases the actinide transport rate in a flowing aquifer. These computer calculations show that any strong complexing agent will have a similar effect on actinide transport in the groundwater environment. 32 references, 9 figures.

  1. 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.

  2. 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.; Sluiter, Amie D.; Hayward, Tammy K.; Nagle, Nicholas J.

    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.

  3. Summary - Salt Waste Processing Facility Design at the Savannah River Site

    Office of Environmental Management (EM)

    Salt Waste Processing Facility ETR Report Date: November 2006 ETR-4 United States Department of Energy Office of Environmental Management (DOE-EM) External Technical Review of the Salt Waste Processing Facility Design at the Savannah River Site (SRS) Why DOE-EM Did This Review The Salt Waste Processing Facility (SWPF) is intended to remove and concentrate the radioactive strontium (Sr), actinides, and cesium (Cs) from the bulk salt waste solutions in the SRS high-level waste tanks. The sludge

  4. 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).

  5. 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.

  6. 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.

  7. 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.

  8. 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

  9. 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.

  10. 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.

  11. 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.

  12. Nonaqueous method for dissolving lanthanide and actinide metals

    DOE Patents [OSTI]

    Crisler, L.R.

    1975-11-11

    Lanthanide and actinide beta-diketonate complex molecular compounds are produced by reacting a beta-diketone compound with a lanthanide or actinide element in the elemental metallic state in a mixture of carbon tetrachloride and methanol.

  13. Low Cost Chemical Feedstocks Using an Improved and Energy Efficient Natural Gas Liquid Removal Process

    SciTech Connect (OSTI)

    2004-07-01

    This factsheet describes a research project whose goal is to develop a new low-cost and energy efficient NGL recovery process - through a combination of theoretical, bench-scale, and pilot-scale testing - so that it can be offered to the natural gas industry for commercialization.

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

    DOE Patents [OSTI]

    Miller, Steven M.

    1988-01-01

    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.

  15. 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.

  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. 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.

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

    SciTech Connect (OSTI)

    Stanley Prawiradjaja

    2003-05-31

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

  19. 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.

  20. SEPARATION OF PLUTONIUM FROM FISSION PRODUCTS BY A COLLOID REMOVAL PROCESS

    DOE Patents [OSTI]

    Schubert, J.

    1960-05-24

    A method is given for separating plutonium from uranium fission products. An acidic aqueous solution containing plutonium and uranium fission products is subjected to a process for separating ionic values from colloidal matter suspended therein while the pH of the solution is maintained between 0 and 4. Certain of the fission products, and in particular, zirconium, niobium, lanthanum, and barium are in a colloidal state within this pH range, while plutonium remains in an ionic form, Dialysis, ultracontrifugation, and ultrafiltration are suitable methods of separating plutonium ions from the colloids.

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

    DOE Patents [OSTI]

    McDowell, William J.

    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.

  2. 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.

  3. 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.

  4. 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

  5. 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.

  6. 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.

  7. Process for simultaneous removal of SO.sub.2 and NO.sub.x from gas streams

    DOE Patents [OSTI]

    Rosenberg, Harvey S.

    1987-01-01

    A process 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.degree. to 125.degree. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

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

    DOE Patents [OSTI]

    Chang, Shih-Ger; Liu, David K.; Griffiths, Elizabeth A.; Littlejohn, David

    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.

  13. Simultaneous SO{sub 2}, SO{sub 3} and NOx removal with ammonia and electron beam irradiation by the EBA process

    SciTech Connect (OSTI)

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

    1999-07-01

    Details are presented of design, performance, operational experience and cost-effectiveness of an ammonium sulfate/nitrate yielding, 90 MWe-capacity, Electron Beam system retrofit, an initial commercial installation of the EBA Process in high-sulfur bituminous coal service at the Chengdu Power Station of Sichuan Electric Power Administration in China. 1997/1998 system startup and commissioning activities leading to successful acceptance tests in 1998 are reviewed to indicate the scope of problems addressed and overcome, and the resulting broad applicability for low-grade fuel service, e.g. in Asia and North America, is illustrated. A retrofit installation of 220 MWe capacity at a powerplant of Chubu Electric Power Company, Inc., in Japan, 92+% SO{sub 2} and SO{sub 3} removal w/60+% NOx removal, that will start up in late 1999 is reviewed. Process economics, i.e. cost/ton SO{sub 2} removal are presented for: The Chengdu Station installation, 80+% SO{sub 2} and SO{sub 3} removal w/20% NOx removal; and a commonly referenced, hypothetical application on a new unit in the U.S. of 300 MWe capacity with 2.6% sulfur bituminous coal fueling designed for performance of 90% SO{sub 2} and SO{sub 3} removal w/65% NOx removal.

  14. 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.

  15. 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.

  16. Separation of Californium from other Actinides

    DOE Patents [OSTI]

    Mailen, J.C.; Ferris, L.M.

    1973-09-25

    A method is provided for separating californium from a fused fluoride composition containing californium and at least one element selected from the group consisting of plutonium, americium, curium, uranium, thorium, and protactinium which comprises contacting said fluoride composition with a liquid bismuth phase containing sufficient lithium or thorium to effect transfer of said actinides to the bismuth phase and then contacting the liquid bismuth phase with molten LiCl to effect selective transfer of californium to the chloride phase.

  17. 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.

  18. 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.

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

    SciTech Connect (OSTI)

    Kaaeid Lokhandwala

    2003-09-29

    The objective of this project is to design, construct and field demonstrate a membrane system to recover natural gas liquids (NGLs) and remove water from raw natural gas. To convince industry users of the efficiency and reliability of the process, we plan to conduct an extended field test to demonstrate system performance under real-world conditions. The membrane system has been designed and fabricated by Membrane Technology and Research, Inc. (MTR). The MTR membrane system and the compressor are now onsite at BP's Pascagoula, MS plant. The plant is undergoing a very significant expansion and the installation of the membrane unit into the test location is being implemented, albeit at a slower rate than we expected. The startup of the system and conducting of tests will occur in the next six months, depending on the availability of the remaining budget. In the interim, significant commercial progress has been made regarding the introduction of the NGL membrane and systems into the natural gas market.

  20. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER

    Office of Scientific and Technical Information (OSTI)

    REACTORS USING HYDRIDE FUEL (Technical Report) | SciTech Connect FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL Citation Details In-Document Search Title: FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL 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

  1. Theory in evaluation of actinide fission and capture cross sections.

    Office of Scientific and Technical Information (OSTI)

    (Conference) | SciTech Connect Theory in evaluation of actinide fission and capture cross sections. Citation Details In-Document Search Title: Theory in evaluation of actinide fission and capture cross sections. The authors discuss the possibilities and limitations of the use of theory as a tool in the evaluation of actinide fission and capture cross-sections. They consider especially the target {sup 235}U as an example. They emphasize the roles of intermediate structure in the fission

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

    SciTech Connect (OSTI)

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

    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. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT...

    Office of Scientific and Technical Information (OSTI)

    IN LIGHT WATER REACTORS USING HYDRIDE FUEL Citation Details In-Document Search Title: FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING ...

  4. Actinide Research Quarterly (Technical Report) | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    Technical Report: Actinide Research Quarterly Citation Details In-Document Search Title: ... This site is a product of DOE's Office of Scientific and Technical Information (OSTI) and ...

  5. Photochemical route to actinide-transition metal bonds: synthesis...

    Office of Scientific and Technical Information (OSTI)

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

  6. Method for decontamination of nickel-fluoride-coated nickel containing actinide-metal fluorides

    DOE Patents [OSTI]

    Windt, Norman F.; Williams, Joe L.

    1983-01-01

    The invention is a process for decontaminating particulate nickel contaminated with actinide-metal fluorides. In one aspect, the invention comprises contacting nickel-fluoride-coated nickel with gaseous ammonia at a temperature effecting nickel-catalyzed dissociation thereof and effecting hydrogen-reduction of the nickel fluoride. The resulting nickel is heated to form a melt and a slag and to effect transfer of actinide metals from the melt into the slag. The melt and slag are then separated. In another aspect, nickel containing nickel oxide and actinide metals is contacted with ammonia at a temperature effecting nickel-catalyzed dissociation to effect conversion of the nickel oxide to the metal. The resulting nickel is then melted and separated as described. In another aspect nickel-fluoride-coated nickel containing actinide-metal fluorides is contacted with both steam and ammonia. The resulting nickel then is melted and separated as described. The invention is characterized by higher nickel recovery, efficient use of ammonia, a substantial decrease in slag formation and fuming, and a valuable increase in the service life of the furnace liners used for melting.

  7. A Combined Nonfertile and UO{sub 2} PWR Fuel Assembly for Actinide...

    Office of Scientific and Technical Information (OSTI)

    Assembly for Actinide Waste Minimization Citation Details In-Document Search Title: A Combined Nonfertile and UOsub 2 PWR Fuel Assembly for Actinide Waste Minimization A new ...

  8. Inorganic, Isotope, and Actinide Chemistry

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

    | Department of Energy Innovative, Lower Cost Sensors and Controls Yield Better Energy Efficiency Innovative, Lower Cost Sensors and Controls Yield Better Energy Efficiency March 23, 2015 - 1:05pm Addthis ORNL researchers are experimenting with additive roll-to-roll manufacturing techniques to develop low-cost wireless sensors. ORNL’s Pooran Joshi shows how the process enables electronics components to be printed on flexible plastic substrates. Credit: Oak Ridge National Lab ORNL

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

    SciTech Connect (OSTI)

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

    1999-11-29

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

  10. Hanford Site production reactor data pertinent to actinide burning

    SciTech Connect (OSTI)

    Toffer, H.; Roblyer, S.P.

    1993-06-01

    During the 44 years of operation, irradiation of special actinides occurred in the Hanford Site production reactors. The data derived from such irradiations could be of value to advanced actinide burners having a thermal neutron spectrum. Recently, such information has become unclassified and, therefore available for public release. This data is discussed in this report.

  11. 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.

  12. Actinide-Aluminate Speciation in Alkaline Radioactive Waste

    SciTech Connect (OSTI)

    Dr. David L. Clark; Dr. Alexander M. Fedosseev

    2001-12-21

    Investigation of behavior of actinides in alkaline media containing AL(III) showed that no aluminate complexes of actinides in oxidation states (IIII-VIII) were formed in alkaline solutions. At alkaline precipitation IPH (10-14) of actinides in presence of AL(III) formation of aluminate compounds is not observed. However, in precipitates contained actinides (IIV)<(VI), and to a lesser degree actinides (III), some interference of components takes place that is reflected in change of solid phase properties in comparison with pure components or their mechanical mixture. The interference decreases with rise of precipitation PH and at PH 14 is exhibited very feebly. In the case of NP(VII) the individual compound with AL(III) is obtained, however it is not aluminate of neptunium(VII), but neptunate of aluminium(III) similar to neptunates of other metals obtained earlier.

  13. 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 Energy’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 – 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.

  14. 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.

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

    DOE Patents [OSTI]

    Horwitz, E. Philip; Kalina, Dale G.

    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.

  16. METHOD FOR THE PREPARATION OF STABLE ACTINIDE METAL OXIDE-CONTAINING SLURRIES AND OF THE OXIDES THEREFOR

    DOE Patents [OSTI]

    Hansen, R.S.; Minturn, R.E.

    1958-02-25

    This patent deals with a method of preparing actinide metal oxides of a very fine particle size and of forming stable suspensions therefrom. The process consists of dissolving the nitrate of the actinide element in a combustible organic solvent, converting the solution obtained into a spray, and igniting the spray whereby an oxide powder is obtained. The oxide powder is then slurried in an aqueous soiution of a substance which is adsorbable by said oxides, dspersed in a colloid mill whereby a suspension is obtained, and electrodialyzed until a low spectiic conductance is reached.

  17. 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)

  18. 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.

  19. 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.

  20. 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.

  1. 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.; Gal, Eli

    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.

  2. material removal

    National Nuclear Security Administration (NNSA)

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

    Page...

  3. material removal

    National Nuclear Security Administration (NNSA)

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

    Pag...

  4. 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 ?-Al2O3 was selected as the solid-state electrolyte for sensor fabrication based on cationic conductivity and other factors. In the present work Gd3+-?-Al2O3 was prepared by ion exchange reactions between trivalent Gd3+ from GdCl3 and K+-, Na+-, and Sr2+-?-Al2O3 precursors. Scanning electron microscopy (SEM) was used for characterization of Gd3+-?-Al2O3 samples. Microfocus X-ray Diffraction (-XRD) was used in conjunction with SEM energy dispersive X-ray spectroscopy (EDS) to identify phase content and elemental composition. The Gd3+-?-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 ?-Al2O3 phase after annealing was verified by -XRD. Preliminary sensor tests with different assembly designs will also be presented.

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

    SciTech Connect (OSTI)

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

    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.

  6. 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.

  7. 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; Wong, Chi F.; Buckley, Leo P.

    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.

  8. Thorium and Uranium: Elements of Opportunity in Actinide Organometalli...

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

    Thorium and Uranium: Elements of Opportunity in Actinide Organometallic Chemistry January 12, 2016 11:00AM to 12:00PM Presenter Jaqueline L. Kiplinger, Los Alamos National...

  9. Spectroscopy and Structure of the Simplest Actinide Bonds (Journal Article)

    Office of Scientific and Technical Information (OSTI)

    | DOE PAGES Published Article: Spectroscopy and Structure of the Simplest Actinide Bonds « Prev Next » Title: Spectroscopy and Structure of the Simplest Actinide Bonds Authors: Heaven, Michael C. ; Barker, Beau J. ; Antonov, Ivan O. Publication Date: 2014-11-20 OSTI Identifier: 1159530 Grant/Contract Number: FG02-01ER15153 Type: Published Article Journal Name: Journal of Physical Chemistry. A, Molecules, Spectroscopy, Kinetics, Environment, and General Theory Additional Journal

  10. Actinide Research Quarterly (Technical Report) | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    Actinide Research Quarterly Citation Details In-Document Search Title: Actinide Research Quarterly Authors: Migliori, Albert [1] + Show Author Affiliations Los Alamos National Lab. (LANL), Los Alamos, NM (United States) Publication Date: 2015-06-29 OSTI Identifier: 1188164 Report Number(s): LA-UR--15-24862 DOE Contract Number: AC52-06NA25396 Resource Type: Technical Report Research Org: Los Alamos National Lab. (LANL), Los Alamos, NM (United States) Sponsoring Org: USDOE Country of Publication:

  11. Environmental Assessment for Actinide Chemistry and Repository Science

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

    Laboratory questions on the Environmental Assessment for Actinide Chemistry and Repository Science Laboratory, email Harold.Johnson@wipp.ws or call (505) 234-7349. Environmental Assessment for Actinide Chemistry and Repository Science Laboratory Final - January, 2006 This document has been provided to you in PDF format. Please install Adobe Acrobat Reader before accessing these documents. Some of the Chapters containing complex graphics have been split into multiple parts to allow for more

  12. Photochemical route to actinide-transition metal bonds: synthesis,

    Office of Scientific and Technical Information (OSTI)

    characterization and reactivity of a series of thorium and uranium heterobimetallic complexes (Journal Article) | SciTech Connect Journal Article: Photochemical route to actinide-transition metal bonds: synthesis, characterization and reactivity of a series of thorium and uranium heterobimetallic complexes Citation Details In-Document Search Title: Photochemical route to actinide-transition metal bonds: synthesis, characterization and reactivity of a series of thorium and uranium

  13. Theory in evaluation of actinide fission and capture cross sections.

    Office of Scientific and Technical Information (OSTI)

    (Conference) | SciTech Connect Theory in evaluation of actinide fission and capture cross sections. Citation Details In-Document Search Title: Theory in evaluation of actinide fission and capture cross sections. × You are accessing a document from the Department of Energy's (DOE) SciTech Connect. This site is a product of DOE's Office of Scientific and Technical Information (OSTI) and is provided as a public service. Visit OSTI to utilize additional information resources in energy science

  14. Thermodynamic stability of actinide pyrochlore minerals in deep geologic

    Office of Scientific and Technical Information (OSTI)

    repository environments (Conference) | SciTech Connect Thermodynamic stability of actinide pyrochlore minerals in deep geologic repository environments Citation Details In-Document Search Title: Thermodynamic stability of actinide pyrochlore minerals in deep geologic repository environments Crystalline phases of pyrochlore (e.g., CaPuTi{sub 2}O{sub 7}, CaUTi{sub 2}O{sub 7}) have been proposed as a durable ceramic waste form for disposal of high level radioactive wastes including surplus

  15. Yields of fission products from various actinide targets (Conference) |

    Office of Scientific and Technical Information (OSTI)

    SciTech Connect fission products from various actinide targets Citation Details In-Document Search Title: Yields of fission products from various actinide targets No abstract prepared. Authors: Spejewski, Eugene H. [1] ; Carter, H Kennon [1] ; Kronenberg, Andreas [1] ; Stracener, Daniel W [2] ; Greene, John P. [3] ; Nolen, Jerry A. [3] ; Talbert, Willard L. [4] + Show Author Affiliations Oak Ridge Associated Universities (ORAU) ORNL Argonne National Laboratory (ANL) TechSource, Inc.

  16. Actinide Source Term Program, position paper. Revision 1

    SciTech Connect (OSTI)

    Novak, C.F.; Papenguth, H.W.; Crafts, C.C.; Dhooge, N.J.

    1994-11-15

    The Actinide Source Term represents the quantity of actinides that could be mobilized within WIPP brines and could migrate with the brines away from the disposal room vicinity. This document presents the various proposed methods for estimating this source term, with a particular focus on defining these methods and evaluating the defensibility of the models for mobile actinide concentrations. The conclusions reached in this document are: the 92 PA {open_quotes}expert panel{close_quotes} model for mobile actinide concentrations is not defensible; and, although it is extremely conservative, the {open_quotes}inventory limits{close_quotes} model is the only existing defensible model for the actinide source term. The model effort in progress, {open_quotes}chemical modeling of mobile actinide concentrations{close_quotes}, supported by a laboratory effort that is also in progress, is designed to provide a reasonable description of the system and be scientifically realistic and supplant the {open_quotes}Inventory limits{close_quotes} model.

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

    DOE Patents [OSTI]

    Wijmans Johannes G.; Merkel, Timothy C.; Baker, Richard W.

    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.

  19. 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.

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

    DOE Patents [OSTI]

    Tsuo, Simon; Langford, Alison A.

    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.

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

    DOE Patents [OSTI]

    Tsuo, S.; Langford, A.A.

    1989-03-28

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

  3. 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.

  4. 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 leadbismuth 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.

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

    DOE Patents [OSTI]

    Nizamoff, Alan J.

    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.

  6. 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 (SHEs) 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.

  7. Actinide Dioxides in Water: Interactions at the Interface

    SciTech Connect (OSTI)

    Alexandrov, Vitaly; Shvareva, Tatiana Y.; Hayun, Shmuel; Asta, Mark; Navrotsky, Alexandra

    2011-12-15

    A comprehensive understanding of chemical interactions between water and actinide dioxide surfaces is critical for safe operation and storage of nuclear fuels. Despite substantial previous research, understanding the nature of these interactions remains incomplete. In this work, we combine accurate calorimetric measurements with first-principles computational studies to characterize surface energies and adsorption enthalpies of water on two fluorite-structured compounds, ThO? and CeO?, that are relevant for understanding the behavior of water on actinide oxide surfaces more generally. We determine coverage-dependent adsorption enthalpies and demonstrate a mixed molecular and dissociative structure for the first hydration layer. The results show a correlation between the magnitude of the anhydrous surface energy and the water adsorption enthalpy. Further, they suggest a structural model featuring one adsorbed water molecule per one surface cation on the most stable facet that is expected to be a common structural signature of water adsorbed on actinide dioxide compounds.

  8. Nuclear waste actinides as fissile fuel in hybrid blankets

    SciTech Connect (OSTI)

    Sahin, S.; Al-Kusayer, T.A.

    1983-12-01

    The widespread use of the present LWRs produces substantial quantities of nuclear waste materials. Among those, actinide nuclear waste poses a serious problem of stockage because the associated half life times for actinides is measured in terms of geological time periods (several millions of years) so that no waste disposal guarantee over such time intervals can be given, except for space disposal. On the other hand, these nuclear waste actinides are very good fissionable materials for high energetic (D,T) fusion neutrons. It is therefore worthwhile to investigate their quality as potential nuclear fuel in hybrid blankets. The present study investigates the neutronic performance of hybrid blankets containing Np/sup 237/ and Cm/sup 244/ as fissile materials. The isotopic composition of Americium has been adjusted to the spent fuel isotope composition of a LWR. The geometrical design has been made, according to the AYMAN fussion-fission (hybrid) experimental facility, now in the very early phase of planning.

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

    SciTech Connect (OSTI)

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

    2009-09-01

    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.

  10. 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.

  11. 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.

  12. Ba(OH)/sub 2/. 8H/sub 2/O process for the removal and immobilization of carbon-14. Final report

    SciTech Connect (OSTI)

    Haag, G.L.; Holladay, D.W.; Pitt, W.W. Jr.; Young, G.C.

    1986-01-01

    The airborne release of /sup 14/C from various nuclear facilities has been identified as a potential biohazard due to the long half-life of /sup 14/C (5730 years) and the ease with which it may be assimilated into the biosphere. At ORNL, technology has been developed for the removal and immobilization of this radionuclide. Prior studies have indicated that /sup 14/C will likely exist in the oxidized form as CO/sub 2/ and will contribute slightly to the bulk CO/sub 2/ concentration of the gas stream, which is air-like in nature (approx.300 ppM/sub v/ CO/sub 2/). The technology that has been developed utilizes the CO/sub 2/-Ba(OH)/sub 2/.8H/sub 2/O gas-solid reaction with the mode of gas-solid contacting being a fixed bed. The product, BaCO/sub 3/, possesses excellent thermal and chemical stability, prerequisites for the long-term disposal of nuclear wastes. For optimal process operation, studies have indicated that an operating window of adequate size does exist. When operating within the window, high CO/sub 2/ removal efficiency (effluent concentrations <100 ppB/sub v/), high reactant utilization (>99%), and an acceptable pressure drop across the bed (3 kPa/m at a superficial velocity of 13 cm/s) are possible. Three areas of experimental investigation are reported: (1) microscale studies on 150-mg samples to provide information concerning surface properties, kinetics, and equilibrium vapor pressures; (2) macroscale studies on large fixed beds (4.2 kg of reactant) to determine the effects of humidity, temperature, and gas flow rate upon bed pressure drop and CO/sub 2/ breakthrough; and (3) design, construction, and operation of a pilot unit capable of continuously processing a 34-m/sup 3//h (20-ft/sup 3//min) air-based gas stream.

  13. 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.

  14. 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.

  15. 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/.

  16. 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

  17. 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.

  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. Experimental Evaluation of Actinide Transport in a Fractured Granodiorite

    SciTech Connect (OSTI)

    Dittrich, Timothy M.; Reimus, Paul W.

    2015-03-16

    The objective of this study was to demonstrate and evaluate new experimental methods for quantifying the potential for actinide transport in deep fractured crystalline rock formations. We selected a fractured granodiorite at the Grimsel Test Site (GTS) in Switzerland as a model system because field experiments have already been conducted with uranium and additional field experiments using other actinides are planned at the site. Thus, working on this system provides a unique opportunity to compare lab experiment results with fieldscale observations. Rock cores drilled from the GTS were shipped to Los Alamos National Laboratory, characterized by x-ray diffraction and microscopy, and used in batch sorption and column breakthrough experiments. Solutions with pH 6.8 and 8.8 were tested. Solutions were switched to radionuclide-free synthetic Grimsel groundwater after near-steady actinide/colloid breakthrough occurred in column experiments. We are currently evaluating actinide adsorption/desorption rates as a function of water chemistry (initial focus on pH), with future testing planned to evaluate the influence of carbonate concentrations, flow rates, and mineralogy in solutions and suspensions with bentonite colloids. (auth)

  20. 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.

  1. 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}.

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

    DOE Patents [OSTI]

    Krot, Nikolai N.; Charushnikova, Iraida A.

    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.

  4. 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

  5. Method for the recovery of actinide elements from nuclear reactor waste

    DOE Patents [OSTI]

    Horwitz, E. Philip; Delphin, Walter H.; Mason, George W.

    1979-01-01

    A process for partitioning and recovering actinide values from acidic waste solutions resulting from reprocessing of irradiated nuclear fuels by adding hydroxylammonium nitrate and hydrazine to the waste solution to adjust the valence of the neptunium and plutonium values in the solution to the +4 oxidation state, thus forming a feed solution and contacting the feed solution with an extractant of dihexoxyethyl phosphoric acid in an organic diluent whereby the actinide values, most of the rare earth values and some fission product values are taken up by the extractant. Separation is achieved by contacting the loaded extractant with two aqueous strip solutions, a nitric acid solution to selectively strip the americium, curium and rare earth values and an oxalate solution of tetramethylammonium hydrogen oxalate and oxalic acid or trimethylammonium hydrogen oxalate to selectively strip the neptunium, plutonium and fission product values. Uranium values remain in the extractant and may be recovered with a phosphoric acid strip. The neptunium and plutonium values are recovered from the oxalate by adding sufficient nitric acid to destroy the complexing ability of the oxalate, forming a second feed, and contacting the second feed with a second extractant of tricaprylmethylammonium nitrate in an inert diluent whereby the neptunium and plutonium values are selectively extracted. The values are recovered from the extractant with formic acid.

  6. 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.

  7. 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.

  8. 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.

  9. Determination of actinides in urine and fecal samples

    DOE Patents [OSTI]

    McKibbin, Terry T.

    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.

  10. Chemical properties of the heavier actinides and transactinides

    SciTech Connect (OSTI)

    Hulet, E.K.

    1981-01-01

    The chemical properties of each of the elements 99 (Es) through 105 are reviewed and their properties correlated with the electronic structure expected for 5f and 6d elements. A major feature of the heavier actinides, which differentiates them from the comparable lanthanides, is the increasing stability of the divalent oxidation state with increasing atomic number. The divalent oxidation state first becomes observable in the anhydrous halides of californium and increases in stability through the series to nobelium, where this valency becomes predominant in aqueous solution. In comparison with the analogous 4f electrons, the 5f electrons in the latter part of the series are more tightly bound. Thus, there is a lowering of the 5f energy levels with respect to the Fermi level as the atomic number increases. The metallic state of the heavier actinides has not been investigated except from the viewpoint of the relative volatility among members of the series. In aqueous solutions, ions of these elements behave as a normal trivalent actinides and lanthanides (except for nobelium). Their ionic radii decrease with increasing nuclear charge which is moderated because of increased screening of the outer 6p electrons by the 5f electrons. The actinide series of elements is completed with the element lawrencium (Lr) in which the electronic configuration is 5f/sup 14/7s/sup 2/7p. From Mendeleev's periodicity and Dirac-Fock calculations, the next group of elements is expected to be a d-transition series corresponding to the elements Hf through Hg. The chemical properties of elements 104 and 105 only have been studied and they indeed appear to show the properties expected of eka-Hf and eka-Ta. However, their nuclear lifetimes are so short and so few atoms can be produced that a rich variety of chemical information is probably unobtainable.

  11. Joint Actinide Shock Physics Experimental Research | National Nuclear

    National Nuclear Security Administration (NNSA)

    Security Administration 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 target. The shock wave produced by the impact passes through the plutonium, and diagnostic equipment measures the properties of the shocked plutonium. Shock physics experiments such as this are critical to maintaining the safety and security of the nation's stockpile in the absence of underground nuclear testing. For

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

    SciTech Connect (OSTI)

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

    2014-06-01

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

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

    DOE Patents [OSTI]

    Horwitz, E. Philip; Dietz, Mark L.

    1989-01-01

    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.

  14. 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.

  15. Use of Thorium for Transmutation of Plutonium and Minor Actinides in PWRs

    Office of Scientific and Technical Information (OSTI)

    (Journal Article) | SciTech Connect Use of Thorium for Transmutation of Plutonium and Minor Actinides in PWRs Citation Details In-Document Search Title: Use of Thorium for Transmutation of Plutonium and Minor Actinides in PWRs An assessment is made of the potential for Th-based fuel to minimize Pu and minor actinide (MA) production in pressurized water reactors (PWRs). Destruction rates and residual amounts of Pu and MA in the fuel used for transmutation are examined. In particular,

  16. 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)

  17. 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...

  18. Analyses in Support of Z-Pinch IFE and Actinide Transmutation...

    Office of Scientific and Technical Information (OSTI)

    29 ENERGY PLANNING, POLICY AND ECONOMY; 22 GENERAL STUDIES OF NUCLEAR REACTORS; 70 PLASMA PHYSICS AND FUSION TECHNOLOGY; ACTINIDES; DESIGN; DOCUMENTATION; ELECTRIC ...

  19. Analyses in Support of Z-Pinch IFE and Actinide Transmutation...

    Office of Scientific and Technical Information (OSTI)

    ... 29 ENERGY PLANNING, POLICY AND ECONOMY; 22 GENERAL STUDIES OF NUCLEAR REACTORS; 70 PLASMA PHYSICS AND FUSION TECHNOLOGY; ACTINIDES; DESIGN; DOCUMENTATION; ELECTRIC ...

  20. 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.

  1. Progress of nitride fuel cycle research for transmutation of minor actinides

    SciTech Connect (OSTI)

    Arai, Yasuo; Akabori, Mitsuo; Minato, Kazuo

    2007-07-01

    Recent progress of nitride fuel cycle research for transmutation of MA is summarized. Preparation of MA-bearing nitride pellets, such as (Np,Am)N, (Am,Pu)N and (Np,Pu,Am,Cm)N, was carried out. Irradiation behavior of U-free nitride fuel was investigated by the irradiation test of (Pu,Zr)N and PuN+TiN fuels, in which ZrN and TiN were added as a possible diluent material. Further, pyrochemical process of spent nitride fuel was developed by electrorefining in a molten chloride salt and subsequent re-nitridation of actinides in liquid Cd cathode electro-deposits. Nitride fuel cycle for transmutation of MA has been demonstrated in a laboratory scale by the experimental study with MA and Pu. (authors)

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

    SciTech Connect (OSTI)

    Miwa, Shuhei; Osaka, Masahiko; Tanaka, Kosuke; Ishi, Yohei; Yoshimochi, Hiroshi; Tanaka, Kenya

    2007-07-01

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

  3. Lessons Learned from Characterization, Performance Assessment, and EPA Regulatory Review of the 1996 Actinide Source Term for the Waste Isolation Pilot Plant

    SciTech Connect (OSTI)

    Larson, K.W.; Moore, R.C.; Nowak, E.J.; Papenguth, H.W.; Jow, H.

    1999-03-22

    The Waste Isolation Pilot Plant (WIPP) is a US Department of Energy (DOE) facility for the permanent disposal of transuranic waste from defense activities. In 1996, the DOE submitted the Title 40 CFR Part 191 Compliance Certification Application for the Waste Isolation Pilot Plant (CCA) to the US Environmental Protection Agency (EPA). The CCA included a probabilistic performance assessment (PA) conducted by Sandia National Laboratories to establish compliance with the quantitative release limits defined in 40 CFR 191.13. An experimental program to collect data relevant to the actinide source term began around 1989, which eventually supported the 1996 CCA PA actinide source term model. The actinide source term provided an estimate of mobile dissolved and colloidal Pu, Am, U, Th, and Np concentrations in their stable oxidation states, and accounted for effects of uncertainty in the chemistry of brines in waste disposal areas. The experimental program and the actinide source term included in the CCA PA underwent EPA review lasting more than 1 year. Experiments were initially conducted to develop data relevant to the wide range of potential future conditions in waste disposal areas. Interim, preliminary performance assessments and actinide source term models provided insight allowing refinement of experiments and models. Expert peer review provided additional feedback and confidence in the evolving experimental program. By 1995, the chemical database and PA predictions of WIPP performance were considered reliable enough to support the decision to add an MgO backfill to waste rooms to control chemical conditions and reduce uncertainty in actinide concentrations, especially for Pu and Am. Important lessons learned through the characterization, PA modeling, and regulatory review of the actinide source term are (1) experimental characterization and PA should evolve together, with neither activity completely dominating the other, (2) the understanding of physical processes required to develop conceptual models is greater than can be represented in PA models, (3) experimentalists should be directly involved in model and parameter abstraction and simplification for PA, and (4) external expert review should be incorporated early in a project to increase confidence long before regulatory reviews begin.

  4. One of the Largest Pieces of Processing Equipment Removed from Plutonium Finishing Plant- Worker involvement led to safe completion of high-hazard work

    Broader source: Energy.gov [DOE]

    RICHLAND, WASH. – U.S. Department of Energy (DOE) contractor CH2M HILL Plateau Remediation Company (CH2M HILL) announced today the successful removal of one of the largest, most complex pieces of equipment from the Plutonium Finishing Plant (PFP) at the Hanford Site in southeast Washington State.

  5. 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.

  6. The Need for US Actinide Enrichment Capability (Conference) | SciTech

    Office of Scientific and Technical Information (OSTI)

    Connect The Need for US Actinide Enrichment Capability Citation Details In-Document Search Title: The Need for US Actinide Enrichment Capability Authors: Patton, Bradley D [1] ; Robinson, Sharon M [1] + Show Author Affiliations ORNL Publication Date: 2014-01-01 OSTI Identifier: 1156743

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

    SciTech Connect (OSTI)

    Waris, Abdul; Aji, Indarta K.; Novitrian,; Kurniadi, Rizal; Su'ud, Zaki

    2012-06-06

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

  8. 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.

    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)

  9. 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.

  10. 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.

  11. Gas Generation from Actinide Oxide Materials

    SciTech Connect (OSTI)

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

    2000-12-01

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

  12. 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

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

    SciTech Connect (OSTI)

    Halada, Gary P

    2008-04-10

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

  14. 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.

  15. 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.

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

    DOE Patents [OSTI]

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

    1997-01-21

    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.

  17. A Combined Nonfertile and UO{sub 2} PWR Fuel Assembly for Actinide Waste

    Office of Scientific and Technical Information (OSTI)

    Minimization (Journal Article) | SciTech Connect A Combined Nonfertile and UO{sub 2} PWR Fuel Assembly for Actinide Waste Minimization Citation Details In-Document Search Title: A Combined Nonfertile and UO{sub 2} PWR Fuel Assembly for Actinide Waste Minimization A new COmbined NonFertile and Uranium (CONFU) fuel assembly is proposed to limit the actinides that need long-term high-level waste storage from the pressurized water reactor (PWR) fuel cycle. In the CONFU assembly concept,

  18. Irradiation of Metallic and Oxide Fuels for Actinide Transmutation in the ATR

    SciTech Connect (OSTI)

    MacLean, Heather J.; Hayes, Steven L.

    2007-07-01

    Metallic fuels containing minor actinides and rare earth additions have been fabricated and are prepared for irradiation in the ATR, scheduled to begin during the summer of 2007. Oxide fuels containing minor actinides are being fabricated and will be ready for irradiation in ATR, scheduled to begin during the summer of 2008. Fabrication and irradiation of these fuels will provide detailed studies of actinide transmutation in support of the Global Nuclear Energy Partnership. These fuel irradiations include new fuel compositions that have never before been tested. Results from these tests will provide fundamental data on fuel irradiation performance and will advance the state of knowledge for transmutation fuels. (authors)

  19. Practical combinations of light-water reactors and fast reactors for future actinide transmutation

    SciTech Connect (OSTI)

    Collins, Emory D.; Renier, John-Paul

    2007-07-01

    Multicycle partitioning-transmutation (P-T) studies continue to show that use of existing light-water reactors (LWRs) and new advanced light-water reactors (ALWRs) can effectively transmute transuranic (TRU) actinides, enabling initiation of full actinide recycle much earlier than waiting for the development and deployment of sufficient fast reactor (FR) capacity. The combination of initial P-T cycles using LWRs/ALWRs in parallel with economic improvements to FR usage for electricity production, and a follow-on transition period in which FRs are deployed, is a practical approach to near-term closure of the nuclear fuel cycle with full actinide recycle. (authors)

  20. 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.

  1. 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.

  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.

    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. ENHANCING ADVANCED CANDU PROLIFERATION RESISTANCE FUEL WITH MINOR ACTINIDES

    SciTech Connect (OSTI)

    Gray S. Chang

    2010-05-01

    The advanced nuclear system will significantly advance the science and technology of nuclear energy systems and to enhance the spent fuel proliferation resistance. Minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs can play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In this work, an Advanced CANDU Reactor (ACR) fuel unit lattice cell model with 43 UO2 fuel rods will be used to investigate the effectiveness of a Minor Actinide Reduction Approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance. The main MARA objective is to increase the 238Pu / Pu isotope ratio by using the transuranic nuclides (237Np and 241Am) in the high burnup fuel and thereby increase the proliferation resistance even for a very low fuel burnup. As a result, MARA is a very effective approach to enhance the proliferation resistance for the on power refueling ACR system nuclear fuel. The MA transmutation characteristics at different MA loadings were compared and their impact on neutronics criticality assessed. The concept of MARA, significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy reconnaissance.

  4. The Effects of Radiation Chemistry on Solvent Extraction 4. Separation of the Trivalent Actinides and Considerations for Radiation-Resistant Solvent Systems

    SciTech Connect (OSTI)

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

    2010-07-01

    The separation of the minor actinides from dissolved nuclear fuel is one of the more formidable challenges associated with the design of the advanced fuel cycle. The partitioning of americium and its transmutation in fast reactor fuel would reduce high-level-waste long-term storage requirements by as much as two orders of magnitude. However, the lanthanides have very similar chemistry. They also have large neutron capture cross sections and poor metal alloy properties and thus they can not be incorporated into fast reactor fuel. A separation amenable to currently existing aqueous solvent extraction processes is therefore desired, and research is underway in Europe, Asia and the USA toward this end. Current concepts for this final separation rely on the use of soft-donor nitrogen or sulfur-containing ligands that favor complexation with the 5f orbitals of the actinides. In the USA, the most developed process is the TALSPEAK (Trivalent Actinide Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes) process, based upon the competition between bis(2-ethylhexyl)phosphoric acid (HDEHP) in the organic phase and lactate-buffered diethylenetriamine pentaacetic acid (DTPA) in the aqueous phase. In Europe and Japan, current investigation is focused on the BTP diamide mixtures or dithiophosphinic acids. Any process eventually adopted must be robust under conditions of high-radiation dose-rates and acid hydrolysis. The effects of irradiation on solvent extraction formulations may result in: 1) decreased ligand concentrations resulting in lower metal distribution ratios, 2) decreased selectivity due to the generation of ligand radiolysis products that are complexing agents, 3) decreased selectivity due to the generation of diluent radiolysis products that are complexing agents, and 4) altered solvent performance due to films, precipitates, and increased viscocity. Many of the ligands associated with minor actinide/lanthanide separations are relatively new. Unlike their predecessors, many have been designed with radiation chemical principals in mind. The evolution of the BTPs and diamides show attention to these details, and a series of structural modifications have evolved that are meant to address them. In this fourth and final part in the series we report on the radiation chemistry of the minor actinide separations processes. We also provide a summary of the general radiolysis reactions that have implications for all ligand and diluent systems.

  5. Soft X-Ray and Vacuum Ultraviolet Based Spectroscopy of the Actinides...

    Office of Scientific and Technical Information (OSTI)

    Conference: Soft X-Ray and Vacuum Ultraviolet Based Spectroscopy of the Actinides Citation Details In-Document Search Title: Soft X-Ray and Vacuum Ultraviolet Based Spectroscopy of ...

  6. Soft X-Ray and Vacuum Ultraviolet Based Spectroscopy of the Actinides...

    Office of Scientific and Technical Information (OSTI)

    Conference: Soft X-Ray and Vacuum Ultraviolet Based Spectroscopy of the Actinides Citation Details In-Document Search Title: Soft X-Ray and Vacuum Ultraviolet Based Spectroscopy of...

  7. 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.

  8. An instrument for the investigation of actinides with spin resolved photoelectron spectroscopy and bremsstrahlung isochromat spectroscopy

    SciTech Connect (OSTI)

    Yu, S.-W.; Tobin, J. G.; Chung, B. W.

    2011-01-01

    A new system for spin resolved photoelectron spectroscopy and bremsstrahlung isochromat spectroscopy has been built and commissioned at Lawrence Livermore National Laboratory for the investigation of the electronic structure of the actinides.Actinide materials are very toxic and radioactive and therefore cannot be brought to most general user facilities for spectroscopic studies. The technical details of the new system and preliminary data obtained therein will be presented and discussed.

  9. 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.

  10. 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

  11. Mass balance and separation factor of actinides through series process test on pyro-process

    SciTech Connect (OSTI)

    Kitawaki, Shinichi; Fukushima, Mineo; Yahagi, Noboru; Kurata, Masaki

    2007-07-01

    An electrolysis test in a sequential condition was performed using U-Pu alloy and liquid Cd as an anode and a cathode material, respectively, in which the anode and cathode was changed three times. Mass balance of U and Pu after three-time electrolysis was determined from both the chemical analysis of salt and anode residue, and the weight increase in the cathode. Approximately 95% of U and 100% of Pu was detected even after three times electrolysis. The loss of U is more significant than that of Pu. The mass balance was also evaluated for the intermediate steps of the sequence. More than 90% of U and Pu with respect to the initial amount were constantly caught up with in each step. The separation factor of U/Pu in three cathodes were varied 1.62-2.06. The chemical form of the anode residue was UO, PuOCl and PuO{sub 2}, which can be converted to the chloride by a reaction with ZrCl{sub 4}. (authors)

  12. Calculation of binary phase diagrams between the actinide elements, rare earth elements, and transition metal elements

    SciTech Connect (OSTI)

    Selle, J E

    1992-06-26

    Attempts were made to apply the Kaufman method of calculating binary phase diagrams to the calculation of binary phase diagrams between the rare earths, actinides, and the refractory transition metals. Difficulties were encountered in applying the method to the rare earths and actinides, and modifications were necessary to provide accurate representation of known diagrams. To calculate the interaction parameters for rare earth-rare earth diagrams, it was necessary to use the atomic volumes for each of the phases: liquid, body-centered cubic, hexagonal close-packed, and face-centered cubic. Determination of the atomic volumes of each of these phases for each element is discussed in detail. In some cases, empirical means were necessary. Results are presented on the calculation of rare earth-rare earth, rare earth-actinide, and actinide-actinide diagrams. For rare earth-refractory transition metal diagrams and actinide-refractory transition metal diagrams, empirical means were required to develop values for the enthalpy of vaporization for rare earth elements and values for the constant (C) required when intermediate phases are present. Results of using the values determined for each element are presented.

  13. 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)

  14. 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.

  15. High removal rate laser-based coating removal system

    DOE Patents [OSTI]

    Matthews, Dennis L.; Celliers, Peter M.; Hackel, Lloyd; Da Silva, Luiz B.; Dane, C. Brent; Mrowka, Stanley

    1999-11-16

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

  16. Turbomachinery debris remover

    DOE Patents [OSTI]

    Krawiec, Donald F.; Kraf, Robert J.; Houser, Robert J.

    1988-01-01

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

  17. In situ removal of contamination from soil

    DOE Patents [OSTI]

    Lindgren, Eric R.; Brady, Patrick V.

    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.

  18. 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.

  19. Graphitic packing removal tool

    DOE Patents [OSTI]

    Meyers, Kurt Edward; Kolsun, George J.

    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.

  20. 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.

  1. On the valence fluctuation in the early actinide metals

    SciTech Connect (OSTI)

    Soderlind, P.; Landa, A.; Tobin, J. G.; Allen, P.; Medling, S.; Booth, C. H.; Bauer, E. D.; Cooley, J. C.; Sokaras, D.; Weng, T. -C.; Nordlund, D.

    2015-12-15

    In this study, recent X-ray measurements suggest a degree of valence fluctuation in plutonium and uranium intermetallics. We are applying a novel scheme, in conjunction with density functional theory, to predict 5f configuration fractions of states with valence fluctuations for the early actinide metals. For this purpose we perform constrained integer f-occupation calculations for the α phases of uranium, neptunium, and plutonium metals. For plutonium we also investigate the δ phase. The model predicts uranium and neptunium to be dominated by the f3 and f4 configurations, respectively, with only minor contributions from other configurations. For plutonium (both α and δ phase) the scenario is dramatically different. Here, the calculations predict a relatively even distribution between three valence configurations. The δ phase has a greater configuration fraction of f6 compared to that of the α phase. The theory is consistent with the interpretations of modern X-ray experiments and we present resonant X-ray emission spectroscopy results for α-uranium.

  2. On the valence fluctuation in the early actinide metals

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

    Soderlind, P.; Landa, A.; Tobin, J. G.; Allen, P.; Medling, S.; Booth, C. H.; Bauer, E. D.; Cooley, J. C.; Sokaras, D.; Weng, T. -C.; et al

    2015-12-15

    In this study, recent X-ray measurements suggest a degree of valence fluctuation in plutonium and uranium intermetallics. We are applying a novel scheme, in conjunction with density functional theory, to predict 5f configuration fractions of states with valence fluctuations for the early actinide metals. For this purpose we perform constrained integer f-occupation calculations for the α phases of uranium, neptunium, and plutonium metals. For plutonium we also investigate the δ phase. The model predicts uranium and neptunium to be dominated by the f3 and f4 configurations, respectively, with only minor contributions from other configurations. For plutonium (both α and δmore » phase) the scenario is dramatically different. Here, the calculations predict a relatively even distribution between three valence configurations. The δ phase has a greater configuration fraction of f6 compared to that of the α phase. The theory is consistent with the interpretations of modern X-ray experiments and we present resonant X-ray emission spectroscopy results for α-uranium.« less

  3. 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.

  4. Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors

    SciTech Connect (OSTI)

    Bhatti, Zaki; Hyland, B.; Edwards, G.W.R.

    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 β) for coolant voiding as standard NU fuel. (authors)

  5. Improving the actinides recycling in closed fuel cycles, a major step towards nuclear energy sustainability

    SciTech Connect (OSTI)

    Poinssot, C.; Grandjean, S.; Masson, M.; Bouillis, B.; Warin, D.

    2013-07-01

    Increasing the sustainability of nuclear energy is a longstanding road that requires a stepwise approach to successively tackle the following 3 objectives. First of all, optimize the consumption of natural resource to preserve them for future generations and hence guarantee the energetic independence of the countries (no uranium ore is needed anymore). The current twice-through cycle of Pu implemented by France, UK, Japan and soon China is a first step in this direction and already allows the development and optimization of the relevant industrial processes. It also allows a major improvement regarding the conditioning of the ultimate waste in a durable and robust nuclear glass. Secondly, the recycling of americium could be an interesting option for the future with the deployment of FR fleet to save the repository resource and optimize its use by allowing a denser disposal. It would limit the burden towards the future generations and the need for additional repositories before several centuries. Thirdly, the recycling of the whole minor actinides inventory could be an interesting option for the far-future for strongly decreasing the waste long-term toxicity, down to a few centuries. It would bring the waste issue back within the human history, which should promote its acceptance by the social opinion.

  6. 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.

  7. 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.

  8. 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

  9. 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)

  10. 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.

  11. Coupled Hybrid Monte Carlo: Deterministic Analysis of VHTR Configurations with Advanced Actinide Fuels

    SciTech Connect (OSTI)

    Tsvetkov, Pavel V.; Ames II, David E.; Alajo, Ayodeji B.; Pritchard, Megan L.

    2006-07-01

    Partitioning and transmutation of minor actinides are expected to have a positive impact on the future of nuclear technology. Their deployment would lead to incineration of hazardous nuclides and could potentially provide additional fuel supply. The U.S. DOE NERI Project assesses the possibility, advantages and limitations of involving minor actinides as a fuel component. The analysis takes into consideration and compares capabilities of actinide-fueled VHTRs with pebble-bed and prismatic cores to approach a reactor lifetime long operation without intermediate refueling. A hybrid Monte Carlo-deterministic methodology has been adopted for coupled neutronics-thermal hydraulics design studies of VHTRs. Within the computational scheme, the key technical issues are being addressed and resolved by implementing efficient automated modeling procedures and sequences, combining Monte Carlo and deterministic approaches, developing and applying realistic 3D coupled neutronics-thermal-hydraulics models with multi-heterogeneity treatments, developing and performing experimental/computational benchmarks for model verification and validation, analyzing uncertainty effects and error propagation. This paper introduces the suggested modeling approach, discusses benchmark results and the preliminary analysis of actinide-fueled VHTRs. The presented up-to-date results are in agreement with the available experimental data. Studies of VHTRs with minor actinides suggest promising performance. (authors)

  12. 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.

  13. 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.

  14. 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.

  15. 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

  16. Fluorination process using catalyst

    DOE Patents [OSTI]

    Hochel, Robert C.; Saturday, Kathy A.

    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.

  17. 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.

  18. 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.

  19. 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.

  20. 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)

  1. Device for removing blackheads

    DOE Patents [OSTI]

    Berkovich, Tamara

    1995-03-07

    A device for removing blackheads from pores in the skin having a elongated handle with a spoon shaped portion mounted on one end thereof, the spoon having multiple small holes piercing therethrough. Also covered is method for using the device to remove blackheads.

  2. Synthesis and development of processes for the recovery of sulfur from acid gases. Part 1, Development of a high-temperature process for removal of H{sub 2}S from coal gas using limestone -- thermodynamic and kinetic considerations; Part 2, Development of a zero-emissions process for recovery of sulfur from acid gas streams

    SciTech Connect (OSTI)

    Towler, G.P.; Lynn, S.

    1993-05-01

    Limestone can be used more effectively as a sorbent for H{sub 2}S in high-temperature gas-cleaning applications if it is prevented from undergoing calcination. Sorption of H{sub 2}S by limestone is impeded by sintering of the product CaS layer. Sintering of CaS is catalyzed by CO{sub 2}, but is not affected by N{sub 2} or H{sub 2}. The kinetics of CaS sintering was determined for the temperature range 750--900{degrees}C. When hydrogen sulfide is heated above 600{degrees}C in the presence of carbon dioxide elemental sulfur is formed. The rate-limiting step of elemental sulfur formation is thermal decomposition of H{sub 2}S. Part of the hydrogen thereby produced reacts with CO{sub 2}, forming CO via the water-gas-shift reaction. The equilibrium of H{sub 2}S decomposition is therefore shifted to favor the formation of elemental sulfur. The main byproduct is COS, formed by a reaction between CO{sub 2} and H{sub 2}S that is analogous to the water-gas-shift reaction. Smaller amounts of SO{sub 2} and CS{sub 2} also form. Molybdenum disulfide is a strong catalyst for H{sub 2}S decomposition in the presence of CO{sub 2}. A process for recovery of sulfur from H{sub 2}S using this chemistry is as follows: Hydrogen sulfide is heated in a high-temperature reactor in the presence of CO{sub 2} and a suitable catalyst. The primary products of the overall reaction are S{sub 2}, CO, H{sub 2} and H{sub 2}O. Rapid quenching of the reaction mixture to roughly 600{degrees}C prevents loss Of S{sub 2} during cooling. Carbonyl sulfide is removed from the product gas by hydrolysis back to CO{sub 2} and H{sub 2}S. Unreacted CO{sub 2} and H{sub 2}S are removed from the product gas and recycled to the reactor, leaving a gas consisting chiefly of H{sub 2} and CO, which recovers the hydrogen value from the H{sub 2}S. This process is economically favorable compared to the existing sulfur-recovery technology and allows emissions of sulfur-containing gases to be controlled to very low levels.

  3. 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.

  4. Method of loading organic materials with group III plus lanthanide and actinide elements

    DOE Patents [OSTI]

    Bell, Zane W.; Huei-Ho, Chuen; Brown, Gilbert M.; Hurlbut, Charles

    2003-04-08

    Disclosed is a composition of matter comprising a tributyl phosphate complex of a group 3, lanthanide, actinide, or group 13 salt in an organic carrier and a method of making the complex. These materials are suitable for use in solid or liquid organic scintillators, as in x-ray absorption standards, x-ray fluorescence standards, and neutron detector calibration standards.

  5. Performance Comparison of Metallic, Actinide Burning Fuel in Lead-Bismuth and Sodium Cooled Fast Reactors

    SciTech Connect (OSTI)

    Weaver, Kevan Dean; Herring, James Stephen; Mac Donald, Philip Elsworth

    2001-04-01

    Various methods have been proposed to incinerate or transmutate the current inventory of trans-uranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non-fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years.

  6. 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)

  7. 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.

  8. Reactor for removing ammonia

    DOE Patents [OSTI]

    Luo, Weifang; Stewart, Kenneth D.

    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.

  9. 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.

  10. 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.

  11. 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.

  12. Workers Remove Glove Boxes from Ventilation at Hanford's Plutonium...

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

    processing area have been cleaned, allowing for their removal from ventilation used to control contamination. Addthis Related Articles Employees cut a ventilation duct attached...

  13. THERMALLY SHIELDED MOISTURE REMOVAL DEVICE

    DOE Patents [OSTI]

    Miller, O.E.

    1958-08-26

    An apparatus is presented for removing moisture from the air within tanks by condensation upon a cartridge containing liquid air. An insulating shell made in two halves covers the cartridge within the evacuated system. The shell halves are hinged together and are operated by a system of levers from outside the tank with the motion translated through a sylphon bellows to cover and uncover the cartridge. When the condensation of moisture is in process, the insulative shell is moved away from the liquid air cartridge, and during that part of the process when there is no freezing out of moisture, the shell halves are closed on the cell so thnt the accumulated frost is not evaporated. This insulating shell greatly reduces the consumption of liquid air in this condensation process.

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

    SciTech Connect (OSTI)

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

    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.

  15. Arsenic removal from water

    DOE Patents [OSTI]

    Moore, Robert C.; Anderson, D. Richard

    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.

  16. 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.

  17. Drum lid removal tool

    DOE Patents [OSTI]

    Pella, Bernard M.; Smith, Philip D.

    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.

  18. Removable feedwater sparger assembly

    DOE Patents [OSTI]

    Challberg, Roy C. (Livermore, CA)

    1994-01-01

    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.

  19. 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.

  20. Condensate removal device

    DOE Patents [OSTI]

    Maddox, James W.; Berger, David D.

    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.

  1. 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.

  2. 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)

  3. 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

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

    SciTech Connect (OSTI)

    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 without a radioactive tracer. As a result, surface characterization studies were performed by scanning electron microscopy, electron microprobe analysis, radiographic imaging, and x-ray diffraction.

  5. 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.

  6. 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)

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

    SciTech Connect (OSTI)

    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 without a radioactive tracer. As a result, surface characterization studies were performed by scanning electron microscopy, electron microprobe analysis, radiographic imaging, and x-ray diffraction.

  8. Heat recirculating cooler for fluid stream pollutant removal

    DOE Patents [OSTI]

    Richards, George A.; Berry, David A.

    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.

  9. Water Distribution and Removal Model

    SciTech Connect (OSTI)

    Y. Deng; N. Chipman; E.L. Hardin

    2005-08-26

    The design of the Yucca Mountain high level radioactive waste repository depends on the performance of the engineered barrier system (EBS). To support the total system performance assessment (TSPA), the Engineered Barrier System Degradation, Flow, and Transport Process Model Report (EBS PMR) is developed to describe the thermal, mechanical, chemical, hydrological, biological, and radionuclide transport processes within the emplacement drifts, which includes the following major analysis/model reports (AMRs): (1) EBS Water Distribution and Removal (WD&R) Model; (2) EBS Physical and Chemical Environment (P&CE) Model; (3) EBS Radionuclide Transport (EBS RNT) Model; and (4) EBS Multiscale Thermohydrologic (TH) Model. Technical information, including data, analyses, models, software, and supporting documents will be provided to defend the applicability of these models for their intended purpose of evaluating the postclosure performance of the Yucca Mountain repository system. The WD&R model ARM is important to the site recommendation. Water distribution and removal represents one component of the overall EBS. Under some conditions, liquid water will seep into emplacement drifts through fractures in the host rock and move generally downward, potentially contacting waste packages. After waste packages are breached by corrosion, some of this seepage water will contact the waste, dissolve or suspend radionuclides, and ultimately carry radionuclides through the EBS to the near-field host rock. Lateral diversion of liquid water within the drift will occur at the inner drift surface, and more significantly from the operation of engineered structures such as drip shields and the outer surface of waste packages. If most of the seepage flux can be diverted laterally and removed from the drifts before contacting the wastes, the release of radionuclides from the EBS can be controlled, resulting in a proportional reduction in dose release at the accessible environment. The purposes of this WD&R model (CRWMS M&O 2000b) are to quantify and evaluate the distribution and drainage of seepage water within emplacement drifts during the period of compliance for post-closure performance. The model bounds the fraction of water entering the drift that will be prevented from contacting the waste by the combined effects of engineered controls on water distribution and on water removal. For example, water can be removed during pre-closure operation by ventilation and after closure by natural drainage into the fractured rock. Engineered drains could be used, if demonstrated to be necessary and effective, to ensure that adequate drainage capacity is provided. This report provides the screening arguments for certain Features, Events, and Processes (FEPs) that are related to water distribution and removal in the EBS. Applicable acceptance criteria from the Issue Resolution Status Reports (IRSRs) developed by the U.S. Nuclear Regulatory Commission (NRC 1999a; 1999b; 1999c; and 1999d) are also addressed in this document.

  10. In-Situ Production of Microbial Pigments for Metal and Actinide

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

    Immobilization - Energy Innovation Portal In-Situ Production of Microbial Pigments for Metal and Actinide Immobilization Unique in-situ method shown to dramatically reduce the mobility of contaminants in the soil without need for excavation Savannah River National Laboratory Contact SRNL About This Technology The stimulation of melanin production by subsurface bacteria offers a means to accelerate the immobilization rates of metal and radionuclide contaminants in the subsurface, even at low

  11. SRNL Development of Recovery Processes for Mark-18A Heavy Actinide...

    Office of Scientific and Technical Information (OSTI)

    A status update is provided for the characterization, including modeling using the Monte ... Resource Relation: Conference: Institute of Nuclear Materials Management 46th Annual ...

  12. 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.

  13. Method for decontamination of nickel-fluoride-coated nickel containing actinide-metal fluorides

    DOE Patents [OSTI]

    Windt, N.F.; Williams, J.L.

    In one aspect, the invention comprises contacting nickel-fluoride-coated nickel with gaseous ammonia at a temperature effecting nickel-catalyzed dissociation thereof and effecting hydrogen-reduction of the nickel fluoride. The resulting nickel is heated to form a melt and a slag and to effect transfer of actinide metals from the melt into the slag. The melt and slag are then separated. In another aspect, nickel contianing nickel oxide and actinide metals is contacted with ammonia at a temperature effecting nickel-catalyzed dissociation to effect conversion of the nickel oxide to the metal. The resulting nickel is then melted and separated as described. In another aspect nickel-fluoride-coated nickel containing actinide-metal fluorides is contacted with both steam and ammonia. The resulting nickel then is melted and separated as described. The invention is characterized by higher nickel recovery, efficient use of ammonia, a substantial decrease in slag formation and fuming, and a valuable increase in the service life of the furnace liners used for melting.

  14. Radiolytic Degradation in Lanthanide/Actinide Separation LigandsNOPOPO: Radical Kinetics and Efficiencies Determinations

    SciTech Connect (OSTI)

    Katy L. Swancutt; Stephen P. Mezyk; Richard D. Tillotson; Sylvie Pailloux; Manab Chakravarty; Robert T. Paine; Leigh R. Martin

    2011-07-01

    Trivalent lanthanide/actinide separations from used nuclear fuel occurs in the presence radiation fields that degrades the extraction ligands and solvents. Here we have investigated the stability of a new ligand for lanthanide/actinide separation; 2,6-bis[(di(2-ethylhexyl)phosphino)methyl] pyridine N,P,P-trioxide, TEH(NOPOPO). The impact of {gamma}-radiolysis on the distribution ratios for actinide (Am) and Lanthanide (Eu) extraction both in the presence and absence of an acidic aqueous phase by TEH(NOPOPO) was determined. Corresponding reaction rate constants for the two major radicals, hydroxyl and nitrate, were determined for TEH(NOPOPO) in the aqueous phase, with room temperature values of (3.49 {+-} 0.10) x 10{sup 9} and (1.95 {+-} 0.15) x 10{sup 8} M{sup -1} s{sup -1}, respectively. The activation energy for this reaction was found to be 30.2 {+-} 4.1 kJ mol{sup -1}. Rate constants for two analogues (2-methylphosphonic acid pyridine N,P-dioxide and 2,6-bis(methylphosphonic acid) pyridine N,P,P-trioxide) were also determined to assist in determining the major reaction pathways.

  15. 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.

  16. Pneumatic soil removal tool

    DOE Patents [OSTI]

    Neuhaus, John E.

    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.

  17. 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).

  18. REMOVAL OF CHLORIDE FROM AQUEOUS SOLUTIONS

    DOE Patents [OSTI]

    Schulz, W.W.

    1959-08-01

    The removal of chlorides from aqueons solutions is described. The process involves contacting the aqueous chloride containing solution with a benzene solution about 0.005 M in phenyl mercuric acetate whereby the chloride anions are taken up by the organic phase and separating the organic phase from the aqueous solutions.

  19. 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.

  20. 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.

  1. 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.

  2. Mercury removal sorbents

    DOE Patents [OSTI]

    Alptekin, Gokhan

    2016-03-29

    Sorbents and methods of using them for removing mercury from flue gases over a wide range of temperatures are disclosed. Sorbent materials of this invention comprise oxy- or hydroxyl-halogen (chlorides and bromides) of manganese, copper and calcium as the active phase for Hg.sup.0 oxidation, and are dispersed on a high surface porous supports. In addition to the powder activated carbons (PACs), this support material can be comprised of commercial ceramic supports such as silica (SiO.sub.2), alumina (Al.sub.2O.sub.3), zeolites and clays. The support material may also comprise of oxides of various metals such as iron, manganese, and calcium. The non-carbon sorbents of the invention can be easily injected into the flue gas and recovered in the Particulate Control Device (PCD) along with the fly ash without altering the properties of the by-product fly ash enabling its use as a cement additive. Sorbent materials of this invention effectively remove both elemental and oxidized forms of mercury from flue gases and can be used at elevated temperatures. The sorbent combines an oxidation catalyst and a sorbent in the same particle to both oxidize the mercury and then immobilize it.

  3. 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.

  4. 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.

  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.; Reynolds, John G.; Fox, Glenn A.

    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. Rubber stopper remover

    DOE Patents [OSTI]

    Stitt, Robert R.

    1994-01-01

    A device for removing a rubber stopper from a test tube is mountable to an upright wall, has a generally horizontal splash guard, and a lower plate spaced parallel to and below the splash guard. A slot in the lower plate has spaced-apart opposing edges that converge towards each other from the plate outer edge to a narrowed portion, the opposing edges shaped to make engagement between the bottom of the stopper flange and the top edge of the test tube to wedge therebetween and to grasp the stopper in the slot narrowed portion to hold the stopper as the test tube is manipulated downwardly and pulled from the stopper. The opposing edges extend inwardly to adjoin an opening having a diameter significantly larger than that of the stopper flange.

  8. Summary - SRS Salt Waste Processing Facility

    Office of Environmental Management (EM)

    SRS Co DOE S Proces concen actinid in a se remov adjustm sorben sorben solutio passed separa stream extract sufficie separa (with S vitrifica (DWP Sr/acti federa assure and ha Critica The te (CTE) descrip Readin The Ele Site: S roject: S F Report Date: J ited States Why DOE omposite High Lev Savannah Rive ssing Facility (S ntrate targeted des) from High eries of unit ope ved by contactin ment) with a m nt in a batch m nt (containing S on by cross flow d to a solvent e ated to an aque m. The bulk

  9. 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.

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

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

    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.

  11. 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.

  12. 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.

  13. 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)

  14. EM Celebrates Milestone with Removal of Last Waste Tank at Separations...

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

    Tank at Separations Process Research Unit EM Celebrates Milestone with Removal of Last ... NISKAYUNA, N.Y. - EM recently marked a notable milestone at the Separations Process ...

  15. Final report on solid ferrous scrap copper removal

    SciTech Connect (OSTI)

    Hartman, A.D.; Williamson, C.A.; Davis. D.L.

    1996-08-01

    Research has shown that physically distinct impurities in shredded ferrous scrap can be removed, and that metallic values can be recovered from the removed impurities. Although the closing of the U.S. Bureau of Mines terminated this research, it should be continued by others. Areas for continued research consideration could include further scrap testing to optimize process parameters, among others.

  16. Method for removing metals from a cleaning solution

    DOE Patents [OSTI]

    Deacon, Lewis E.

    2002-01-01

    A method for removing accumulated metals from a cleaning solution is provided. After removal of the metals, the cleaning solution can be discharged or recycled. The process manipulates the pH levels of the solution as a means of precipitating solids. Preferably a dual phase separation at two different pH levels is utilized.

  17. Method for removing chlorine compounds from hydrocarbon mixtures

    DOE Patents [OSTI]

    Janoski, Edward J. (Havertown, PA); Hollstein, Elmer J. (Wilmington, DE)

    1985-12-31

    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.

  18. 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.

  19. 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

  20. removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    removal | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy...

  1. 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

  2. Protection #1: Remove the Source

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

    Remove the Source Protection #1: Remove the Source The 3 Protections = Defense in Depth August 1, 2013 Waste being removed from MDA-B inside a metal building Excavation of waste from MDA-B thumbnail of Removing the source means excavating contaminants, sorting these by waste type, and transporting to a disposal area in which contaminants are contained. RELATED IMAGES http://farm8.staticflickr.com/7388/9571274521_679fe1e34a_t.jpg Enlarge http://farm4.staticflickr.com/3726/9571272211_6873a5717f

  3. Removal to Maximum Extent Practical

    Broader source: Energy.gov [DOE]

    Summary Notes from 1 November 2007 Generic Technical Issue Discussion on Removal of Highly Radioactive Radionuclides/Key Radionuclides to the Maximum Extent Practical

  4. 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.

  5. 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.

  6. 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.

  7. Glovebox Removal at Hanford Site's Plutonium Finishing Plant Winding Down

    Broader source: Energy.gov [DOE]

    RICHLAND, Wash. – At the Plutonium Finishing Plant on the Hanford Site, crews with EM contractor CH2M HILL Plateau Remediation Company are in the process of removing the last of the gloveboxes from the facility before demolition begins.

  8. Removal of arsenic compounds from petroliferous liquids

    DOE Patents [OSTI]

    Fish, Richard H.

    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.

  9. 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.

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

    SciTech Connect (OSTI)

    Konings, Rudy J. M. Bene, Ondrej; Kovcs, Attila; Manara, Dario; Sedmidubsk, 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.

  11. 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.

  12. 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.

  13. Removal of copper from ferrous scrap

    DOE Patents [OSTI]

    Blander, Milton; Sinha, Shome N.

    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.

  14. 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.

  15. 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.

  16. THE HYDROTHERMAL REACTIONS OF MONOSODIUM TITANATE, CRYSTALLINE SILICOTITANATE AND SLUDGE IN THE MODULAR SALT PROCESS: A LITERATURE SURVEY

    SciTech Connect (OSTI)

    Fondeur, F.; Pennebaker, F.; Fink, S.

    2010-11-11

    The use of crystalline silicotitanate (CST) is proposed for an at-tank process to treat High Level Waste at the Savannah River Site. The proposed configuration includes deployment of ion exchange columns suspended in the risers of existing tanks to process salt waste without building a new facility. The CST is available in an engineered form, designated as IE-911-CW, from UOP. Prior data indicates CST has a proclivity to agglomerate from deposits of silica rich compounds present in the alkaline waste solutions. This report documents the prior literature and provides guidance for the design and operations that include CST to mitigate that risk. The proposed operation will also add monosodium titanate (MST) to the supernate of the tank prior to the ion exchange operation to remove strontium and select alpha-emitting actinides. The cesium loaded CST is ground and then passed forward to the sludge washing tank as feed to the Defense Waste Processing Facility (DWPF). Similarly, the MST will be transferred to the sludge washing tank. Sludge processing includes the potential to leach aluminum from the solids at elevated temperature (e.g., 65 C) using concentrated (3M) sodium hydroxide solutions. Prior literature indicates that both CST and MST will agglomerate and form higher yield stress slurries with exposure to elevated temperatures. This report assessed that data and provides guidance on minimizing the impact of CST and MST on sludge transfer and aluminum leaching sludge.

  17. RAPID DETERMINATION OF ACTINIDES IN URINE BY INDUCTIVELY-COUPLED PLASMA MASS SPECTROMETRY AND ALPHA SPECTROMETRY: A HYBRID APPROACH

    SciTech Connect (OSTI)

    Maxwell, S.; Jones, V.

    2009-05-27

    A new rapid separation method that allows separation and preconcentration of actinides in urine samples was developed for the measurement of longer lived actinides by inductively coupled plasma mass spectrometry (ICP-MS) and short-lived actinides by alpha spectrometry; a hybrid approach. This method uses stacked extraction chromatography cartridges and vacuum box technology to facilitate rapid separations. Preconcentration, if required, is performed using a streamlined calcium phosphate precipitation. Similar technology has been applied to separate actinides prior to measurement by alpha spectrometry, but this new method has been developed with elution reagents now compatible with ICP-MS as well. Purified solutions are split between ICP-MS and alpha spectrometry so that long- and short-lived actinide isotopes can be measured successfully. The method allows for simultaneous extraction of 24 samples (including QC samples) in less than 3 h. Simultaneous sample preparation can offer significant time savings over sequential sample preparation. For example, sequential sample preparation of 24 samples taking just 15 min each requires 6 h to complete. The simplicity and speed of this new method makes it attractive for radiological emergency response. If preconcentration is applied, the method is applicable to larger sample aliquots for occupational exposures as well. The chemical recoveries are typically greater than 90%, in contrast to other reported methods using flow injection separation techniques for urine samples where plutonium yields were 70-80%. This method allows measurement of both long-lived and short-lived actinide isotopes. 239Pu, 242Pu, 237Np, 243Am, 234U, 235U and 238U were measured by ICP-MS, while 236Pu, 238Pu, 239Pu, 241Am, 243Am and 244Cm were measured by alpha spectrometry. The method can also be adapted so that the separation of uranium isotopes for assay is not required, if uranium assay by direct dilution of the urine sample is preferred instead. Multiple vacuum box locations may be set-up to supply several ICP-MS units with purified sample fractions such that a high sample throughput may be achieved, while still allowing for rapid measurement of short-lived actinides by alpha spectrometry.

  18. Method of preparation of removable syntactic foam

    DOE Patents [OSTI]

    Arnold, Jr., Charles; Derzon, Dora K.; Nelson, Jill S.; Rand, Peter B.

    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.

  19. 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.

  20. 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 Energys 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 GTRIs 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 NNSAs 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.

  1. Low-temperature synthesis of actinide tetraborides by solid-state metathesis reactions

    DOE Patents [OSTI]

    Lupinetti, Anthony J.; Garcia, Eduardo; Abney, Kent D.

    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.

  2. 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.

  3. Study on Equilibrium Characteristics of Thorium-Plutonium-Minor Actinides Mixed Oxides Fuel in PWR

    SciTech Connect (OSTI)

    Waris, A.; Permana, S.; Kurniadi, R.; Su'ud, Z.; Sekimoto, H.

    2010-06-22

    A study on characteristics of thorium-plutonium-minor actinides utilization in the pressurized water reactor (PWR) with the equilibrium burnup model has been conducted. For a comprehensive evaluation, several fuel cycles scenario have been included in the present study with the variation of moderator-to-fuel volume ratio (MFR) of PWR core design. The results obviously exhibit that the neutron spectra grow to be harder with decreasing of the MFR. Moreover, the neutron spectra also turn into harder with the rising number of confined heavy nuclides. The required {sup 233}U concentration for criticality of reactor augments with the increasing of MFR for all heavy nuclides confinement and thorium and uranium confinement in PWR.

  4. Method for digesting spent ion exchange resins and recovering actinides therefrom using microwave radiation

    DOE Patents [OSTI]

    Maxwell, III, Sherrod L.; Nichols, Sheldon T.

    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.

  5. 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.

  6. 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)

  7. 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.

  8. Waste processing air cleaning

    SciTech Connect (OSTI)

    Kriskovich, J.R.

    1998-07-27

    Waste processing and preparing waste to support waste processing relies heavily on ventilation. Ventilation is used at the Hanford Site on the waste storage tanks to provide confinement, cooling, and removal of flammable gases.

  9. Section 46: Removal of Waste

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

    in and around the WIPP site, the EPA did not identify any significant changes in the planning and execution of the DOE's strategy for removal of waste since the 1998...

  10. Ion Removal - Energy Innovation Portal

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

    Advanced Materials Advanced Materials Find More Like This Return to Search Ion Removal Idaho National Laboratory Contact INL About This Technology Technology Marketing Summary INL's ion removal technology leverages the ability of phosphazene polymers discriminate between water and metal ions, which allows water to pass through the membrane while retaining the ions. Description The inherent chemical and thermal stability of the phosphazene polymers are an added strengths for separating and

  11. 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.

  12. 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.

  13. USE OF AN EQUILIBRIUM MODEL TO FORECAST DISSOLUTION EFFECTIVENESS, SAFETY IMPACTS, AND DOWNSTREAM PROCESSABILITY FROM OXALIC ACID AIDED SLUDGE REMOVAL IN SAVANNAH RIVER SITE HIGH LEVEL WASTE TANKS 1-15

    SciTech Connect (OSTI)

    KETUSKY, EDWARD

    2005-10-31

    This thesis details a graduate research effort written to fulfill the Magister of Technologiae in Chemical Engineering requirements at the University of South Africa. The research evaluates the ability of equilibrium based software to forecast dissolution, evaluate safety impacts, and determine downstream processability changes associated with using oxalic acid solutions to dissolve sludge heels in Savannah River Site High Level Waste (HLW) Tanks 1-15. First, a dissolution model is constructed and validated. Coupled with a model, a material balance determines the fate of hypothetical worst-case sludge in the treatment and neutralization tanks during each chemical adjustment. Although sludge is dissolved, after neutralization more is created within HLW. An energy balance determines overpressurization and overheating to be unlikely. Corrosion induced hydrogen may overwhelm the purge ventilation. Limiting the heel volume treated/acid added and processing the solids through vitrification is preferred and should not significantly increase the number of glass canisters.

  14. Removal of sulfur compounds from combustion product exhaust

    DOE Patents [OSTI]

    Cheng, Dah Y. (Palo Alto, CA)

    1982-01-01

    A method and device are disclosed for removing sulfur containing contaminents from a combustion product exhaust. The removal process is carried out in two stages wherein the combustion product exhaust is dissolved in water, the water being then heated to drive off the sulfur containing contaminents. The sulfur containing gases are then resolublized in a cold water trap to form a concentrated solution which can then be used as a commercial product.

  15. 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)

  16. 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.

  17. 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].

  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 /Pu O.sub.2 in IFR electrorefining process

    DOE Patents [OSTI]

    Tomczuk, Zygmunt; Miller, William E.

    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.

  20. 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.

  1. ,"Virginia Natural Gas Nonhydrocarbon Gases Removed (Million...

    U.S. Energy Information Administration (EIA) Indexed Site

    Data for" ,"Data 1","Virginia Natural Gas Nonhydrocarbon Gases Removed ... 2:52:09 AM" "Back to Contents","Data 1: Virginia Natural Gas Nonhydrocarbon Gases Removed ...

  2. Photovoltaic module with removable wind deflector (Patent) |...

    Office of Scientific and Technical Information (OSTI)

    Patent: Photovoltaic module with removable wind deflector Citation Details In-Document Search Title: Photovoltaic module with removable wind deflector A photovoltaic (PV) module ...

  3. Heavy Water Test Reactor Dome Removal

    SciTech Connect (OSTI)

    2011-01-01

    A high speed look at the removal of the Heavy Water Test Reactor Dome Removal. A project sponsored by the Recovery Act on the Savannah River Site.

  4. Nuclear Material Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Nuclear Material Removal Once weapons-usable nuclear material is no longer required, the Office of Nuclear Material Removal works with global partners and facilities to ...

  5. 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...

  6. Photovoltaic module with removable wind deflector (Patent) |...

    Office of Scientific and Technical Information (OSTI)

    Photovoltaic module with removable wind deflector Title: Photovoltaic module with removable wind deflector A photovoltaic (PV) module assembly including a PV module, a deflector, ...

  7. 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.

  8. Metals removal from spent salts

    DOE Patents [OSTI]

    Hsu, Peter C.; Von Holtz, Erica H.; Hipple, David L.; Summers, Leslie J.; Brummond, William A.; Adamson, Martyn G.

    2002-01-01

    A method and apparatus for removing metal contaminants 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 may be added to precipitate the metal oxide and/or the metal as either metal oxide, metal hydroxide, or as a salt. The precipitated materials are filtered, dried and packaged for disposal as waste or can be immobilized as ceramic pellets. More than about 90% of the metals and mineral residues (ashes) present are removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be spray-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 1.0 ppm of contaminants.

  9. 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)

  10. Correlation consistent basis sets for actinides. I. The Th and U atoms

    SciTech Connect (OSTI)

    Peterson, Kirk A.

    2015-02-21

    New correlation consistent basis sets based on both pseudopotential (PP) and all-electron Douglas-Kroll-Hess (DKH) Hamiltonians have been developed from double- to quadruple-zeta quality for the actinide atoms thorium and uranium. Sets for valence electron correlation (5f6s6p6d), cc − pV nZ − PP and cc − pV nZ − DK3, as well as outer-core correlation (valence + 5s5p5d), cc − pwCV nZ − PP and cc − pwCV nZ − DK3, are reported (n = D, T, Q). The -PP sets are constructed in conjunction with small-core, 60-electron PPs, while the -DK3 sets utilized the 3rd-order Douglas-Kroll-Hess scalar relativistic Hamiltonian. Both series of basis sets show systematic convergence towards the complete basis set limit, both at the Hartree-Fock and correlated levels of theory, making them amenable to standard basis set extrapolation techniques. To assess the utility of the new basis sets, extensive coupled cluster composite thermochemistry calculations of ThF{sub n} (n = 2 − 4), ThO{sub 2}, and UF{sub n} (n = 4 − 6) have been carried out. After accurately accounting for valence and outer-core correlation, spin-orbit coupling, and even Lamb shift effects, the final 298 K atomization enthalpies of ThF{sub 4}, ThF{sub 3}, ThF{sub 2}, and ThO{sub 2} are all within their experimental uncertainties. Bond dissociation energies of ThF{sub 4} and ThF{sub 3}, as well as UF{sub 6} and UF{sub 5}, were similarly accurate. The derived enthalpies of formation for these species also showed a very satisfactory agreement with experiment, demonstrating that the new basis sets allow for the use of accurate composite schemes just as in molecular systems composed only of lighter atoms. The differences between the PP and DK3 approaches were found to increase with the change in formal oxidation state on the actinide atom, approaching 5-6 kcal/mol for the atomization enthalpies of ThF{sub 4} and ThO{sub 2}. The DKH3 atomization energy of ThO{sub 2} was calculated to be smaller than the DKH2 value by ∼1 kcal/mol.

  11. removal

    National Nuclear Security Administration (NNSA)

    80 pounds) of highly enriched uranium (HEU) spent fuel from the Institute of Nuclear Physics (INP) in Almaty, Kazakhstan. The HEU was transported via two air shipments to a...

  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.

    2014-11-01

    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. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.

  13. 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.

  14. Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

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

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

    2014-11-01

    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 ofmore » 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. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.« less

  15. 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)

  16. CPP-603 Chloride Removal System Decontamination and Decommissioning. Final report

    SciTech Connect (OSTI)

    Moser, C.L.

    1993-02-01

    The CPP-603 (annex) Chloride Removal System (CRS) Decontamination and Decommissioning (D&D) Project is described in this report. The CRS was used for removing Chloride ions and other contaminants that were suspended in the waters of the underwater fuel storage basins in the CPP-603 Fuel Receiving and Storage Facility (FRSF) from 1975 to 1981. The Environmental Checklist and related documents, facility characterization, decision analysis`, and D&D plans` were prepared in 1991. Physical D&D activities were begun in mid summer of 1992 and were completed by the end of November 1992. All process equipment and electrical equipment were removed from the annex following accepted asbestos and radiological contamination removal practices. The D&D activities were performed in a manner such that no radiological health or safety hazard to the public or to personnel at the Idaho National Engineering Laboratory (INEL) occurred.

  17. Massive Hanford Test Reactor Removed - Plutonium Recycle Test...

    Office of Environmental Management (EM)

    Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor removed from Hanford's 300 Area Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor removed ...

  18. Removing sulphur oxides from a fluid stream

    DOE Patents [OSTI]

    Katz, Torsten; Riemann, Christian; Bartling, Karsten; Rigby, Sean Taylor; Coleman, Luke James Ivor; Lail, Marty Alan

    2014-04-08

    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.

  19. Part 3: Removal Action | Department of Energy

    Office of Environmental Management (EM)

    3: Removal Action Part 3: Removal Action Question: When may removal actions be initiated? Answer: Removal actions may be initiated when DOE determines that the action will prevent, minimize, stabilize, or eliminate a risk to health or the environment. The NCP specifies that the determination that a risk to health or the environment is appropriate for removal action should be based on: actual or potential exposure of humans, animals, or the food chain the presence of contained hazardous

  20. Method for removing undesired particles from gas streams

    DOE Patents [OSTI]

    Durham, Michael Dean; Schlager, Richard John; Ebner, Timothy George; Stewart, Robin Michele; Hyatt, David E.; Bustard, Cynthia Jean; Sjostrom, Sharon

    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.