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Sample records for plutonium oxide containers

  1. PRESSURIZATION OF CONTAINMENT VESSELS FROM PLUTONIUM OXIDE CONTENTS

    SciTech Connect

    Hensel, S.

    2012-03-27

    Transportation and storage of plutonium oxide is typically done using a convenience container to hold the oxide powder which is then placed inside a containment vessel. Intermediate containers which act as uncredited confinement barriers may also be used. The containment vessel is subject to an internal pressure due to several sources including; (1) plutonium oxide provides a heat source which raises the temperature of the gas space, (2) helium generation due to alpha decay of the plutonium, (3) hydrogen generation due to radiolysis of the water which has been adsorbed onto the plutonium oxide, and (4) degradation of plastic bags which may be used to bag out the convenience can from a glove box. The contributions of these sources are evaluated in a reasonably conservative manner.

  2. Plutonium metal and oxide container weld development and qualification

    SciTech Connect

    Fernandez, R.; Horrell, D.R.; Hoth, C.W.; Pierce, S.W.; Rink, N.A.; Rivera, Y.M.; Sandoval, V.D.

    1996-01-01

    Welds were qualified for a container system to be used for long-term storage of plutonium metal and oxide. Inner and outer containers are formed of standard tubing with stamped end pieces gas-tungsten-arc (GTA) welded onto both ends. The weld qualification identified GTA parameters to produce a robust weld that meets the requirements of the Department of Energy standard DOE-STD-3013-94, ``Criteria for the Safe Storage of Plutonium Metals and Oxides.``

  3. STAINLESS STEEL INTERACTIONS WITH SALT CONTAINING PLUTONIUM OXIDES

    SciTech Connect

    Nelson, Z.; Chandler, G.; Dunn, K.; Stefek, T.; Summer, M.

    2010-02-01

    Salt containing plutonium oxide materials are treated, packaged and stored within nested, stainless steel containers based on requirements established in the DOE 3013 Standard. The moisture limit for the stored materials is less than 0.5 weight %. Surveillance activities which are conducted to assess the condition of the containers and assure continuing 3013 container integrity include the destructive examination of a select number of containers to determine whether corrosion attack has occurred as a result of stainless steel interactions with salt containing plutonium oxides. To date, some corrosion has been observed on the innermost containers, however, no corrosion has been noted on the outer containers and the integrity of the 3013 container systems is not expected to be compromised over a 50 year storage lifetime.

  4. Simulation and analysis of the plutonium oxide/metal storage containers subject to various loading conditions

    SciTech Connect

    Gong, C.; Miller, R.F.

    1995-05-01

    The structural and functional requirements of the Plutonium Oxide/Metal Storage Containers are specified in the Report ``Complex 21 Plutonium Storage Facility Material Containment Team Technical Data Report`` [Complex 21, 1993]. There are no existing storage containers designed for long term storage of plutonium and current codes, standards or regulations do not adequately cover this case. As there is no extensive experience with the long term (50+ years) storage of plutonium, the design of high integrity storage containers must address many technical considerations. This analysis discusses a few potential natural phenomena that could theoretically adversely affect the container integrity over time. The plutonium oxide/metal storage container consists of a primary containment vessel (the outer container), a bagless transfer can (the inner container), two vertical plates on top of the primary containment vessel, a circular plate (the flange) supported by the two plates, tube for gas sampling operations mounted at the center of the primary containment vessel top and a spring system being inserted in the cavity between the primary containment vessel and the cap of the bagless transfer can. The dimensions of the plutonium oxide/metal storage container assembly can be found in Figure 2-1. The primary container, the bagless transfer can, and all the attached components are made of Type 304L stainless steel.

  5. Production of microspheres of thorium oxide, uranium oxide and plutonium oxide and their mixtures containing carbon

    SciTech Connect

    Bezzi, G.; Facchini, A.; Martignani, G.; Pastore, M.

    1980-05-13

    A process is desclosed for the production of microspheres of thorium oxide, uranium oxide or plutonium oxide and mixtures thereof, containing carbon. The microspheres are prepared by first forming an aqueous alkaline solution of a salt of said metals which will precipitate to a solid form in a concentrated alkaline medium together with an alkaline polymerizable organic monomer and carbon, thereafter dripping this solution into a concentrated alkaline medium wherein metal hydroxide is precipitated out and said monomers polymerized to thereby form said microspheres. The precipitated metal compounds with polymerized monomer and carbon are washed to remove unwanted ions and dried by various methods such as air currents, infrared lamps, high frequency microwaves or preferably by azeotropic distillation whereby the washed particles are dispersed in an organic liquid immiscible with water which is then boiled off. Simple liquids include 1,1,1-trichloroethane, carbontetrachloride and xylene. The particles are then subjected to thermal treatments which vary according to the particular composition and the properties required in the final product. The microspheres obtained according to the process of the invention possess excellent properties of high density, good homogeneity, and high tensile strength and are useful as fuel for high temperature gas-cooled nuclear reactors. In a preferred embodiment, uranyl nitrate is mixed with acrolein in dilute ammonia solution to which carbon is added and the solution dripped into a concentrated alkaline medium, and the microspheres thereby formed washed with water, dried by azeotropic distillation with an organic liquid, and thermally cycled.

  6. DOE nuclear material packaging manual: storage container requirements for plutonium oxide materials

    SciTech Connect

    Veirs, D Kirk

    2009-01-01

    Loss of containment of nuclear material stored in containers such as food-pack cans, paint cans, or taped slip lid cans has generated concern about packaging requirements for interim storage of nuclear materials in working facilities such as the plutonium facility at Los Alamos National Laboratory (LANL). In response, DOE has recently issued DOE M 441.1 'Nuclear Material Packaging Manual' with encouragement from the Defense Nuclear Facilities Safety Board. A unique feature compared to transportation containers is the allowance of filters to vent flammable gases during storage. Defining commonly used concepts such as maximum allowable working pressure and He leak rate criteria become problematic when considering vented containers. Los Alamos has developed a set of container requirements that are in compliance with 441.1 based upon the activity of heat-source plutonium (90% Pu-238) oxide, which bounds the requirements for weapons-grade plutonium oxide. The pre and post drop-test He leak rates depend upon container size as well as the material contents. For containers that are routinely handled, ease of handling and weight are a major consideration. Relatively thin-walled containers with flat bottoms are desired yet they cannot be He leak tested at a differential pressure of one atmosphere due to the potential for plastic deformation of the flat bottom during testing. The He leak rates and He leak testing configuration for containers designed for plutonium bearing materials will be presented. The approach to meeting the other manual requirements such as corrosion and thermal degradation resistance will be addressed. The information presented can be used by other sites to evaluate if their conditions are bounded by LANL requirements when considering procurement of 441.1 compliant containers.

  7. PROCESS OF OXIDIZING PLUTONIUM

    DOEpatents

    Coryell, C.D.

    1959-08-25

    The oxidation of plutonium to the plus six valence state is described. The oxidation is accomplished by treating the plutonium in aqueous solution with a solution above 0.01 molar in argentic ion, above 1.1 molar in nitric acid, and above 0.02 molar in argentous ion.

  8. Dynamic and quasi-static simulation and analysis of the plutonium oxide/metal containers subject to 30-foot dropping

    SciTech Connect

    Gong, C.; Miller, R.F.

    1995-01-01

    This analysis of the plutonium oxide/metal storage containers is in support of the design and testing project The results from the dynamic analysis show some important facts that have not been considered before. The internal bagless transfer can will have higher stress than the primary container. The quasi-static analysis provides a conservative solution. In both vertical upright drop (dynamic) and inclined upside down drop (quasi-static) the containers are structurally sound.

  9. Plutonium oxide dissolution

    SciTech Connect

    Gray, J.H.

    1992-09-30

    Several processing options for dissolving plutonium oxide (PuO[sub 2]) from high-fired materials have been studied. The scoping studies performed on these options were focused on PuO[sub 2] typically generated by burning plutonium metal and PuO[sub 2] produced during incineration of alpha contaminated waste. At least two processing options remain applicable for dissolving high-fired PuO[sub 2] in canyon dissolvers. The options involve solid solution formation of PuO[sub 2] With uranium oxide (UO[sub 2]) and alloying incinerator ash with aluminum. An oxidative dissolution process involving nitric acid solutions containing a strong oxidizing agent, such as cerium (IV), was neither proven nor rejected. This uncertainty was due to difficulty in regenerating cerium (IV) ions during dissolution. However, recent work on silver-catalyzed dissolution of PuO[sub 2] with persulfate has demonstrated that persulfate ions regenerate silver (II). Use of persulfate to regenerate cerium (IV) or bismuth (V) ions during dissolution of PuO[sub 2] materials may warrant further study.

  10. Plutonium oxide dissolution

    SciTech Connect

    Gray, J.H.

    1992-09-30

    Several processing options for dissolving plutonium oxide (PuO{sub 2}) from high-fired materials have been studied. The scoping studies performed on these options were focused on PuO{sub 2} typically generated by burning plutonium metal and PuO{sub 2} produced during incineration of alpha contaminated waste. At least two processing options remain applicable for dissolving high-fired PuO{sub 2} in canyon dissolvers. The options involve solid solution formation of PuO{sub 2} With uranium oxide (UO{sub 2}) and alloying incinerator ash with aluminum. An oxidative dissolution process involving nitric acid solutions containing a strong oxidizing agent, such as cerium (IV), was neither proven nor rejected. This uncertainty was due to difficulty in regenerating cerium (IV) ions during dissolution. However, recent work on silver-catalyzed dissolution of PuO{sub 2} with persulfate has demonstrated that persulfate ions regenerate silver (II). Use of persulfate to regenerate cerium (IV) or bismuth (V) ions during dissolution of PuO{sub 2} materials may warrant further study.

  11. Conversion of plutonium-containing materials into borosilicate glass using the glass material oxidation and dissolution system

    SciTech Connect

    Forsberg, C.W.; Beahm, E.C.; Parker, G.W.

    1996-01-27

    The end of the cold war has resulted in excess plutonium-containing materials (PCMs) in multiple chemical forms. Major problems are associated with the long-term management of these materials: safeguards and nonproliferation issues; health, environment, and safety concerns; waste management requirements; and high storage costs. These issues can be addressed by conversion of the PCMs to glass: however, conventional glass processes require oxide-like feed materials. Conversion of PCMs to oxide-like materials followed by vitrification is a complex and expensive process. A new vitrification process has been invented, the Glass Material Oxidation and Dissolution System (GMODS) to allow direct conversion of PCMs to glass. GMODS directly converts metals, ceramics, and amorphous solids to glass; oxidizes organics with the residue converted to glass; and converts chlorides to borosilicate glass and a secondary sodium chloride stream. Laboratory work has demonstrated the conversion of cerium (a plutonium surrogate), uranium (a plutonium surrogate), Zircaloy, stainless steel, multiple oxides, and other materials to glass. Equipment options have been identified for processing rates between 1 and 100,000 t/y. Significant work, including a pilot plant, is required to develop GMODS for applications at an industrial scale.

  12. TERNARY ALLOY-CONTAINING PLUTONIUM

    DOEpatents

    Waber, J.T.

    1960-02-23

    Ternary alloys of uranium and plutonium containing as the third element either molybdenum or zirconium are reported. Such alloys are particularly useful as reactor fuels in fast breeder reactors. The alloy contains from 2 to 25 at.% of molybdenum or zirconium, the balance being a combination of uranium and plutonium in the ratio of from 1 to 9 atoms of uranlum for each atom of plutonium. These alloys are prepared by melting the constituent elements, treating them at an elevated temperature for homogenization, and cooling them to room temperature, the rate of cooling varying with the oomposition and the desired phase structure. The preferred embodiment contains 12 to 25 at.% of molybdenum and is treated by quenching to obtain a body centered cubic crystal structure. The most important advantage of these alloys over prior binary alloys of both plutonium and uranium is the lack of cracking during casting and their ready machinability.

  13. OXIDATIVE METHOD OF SEPARATING PLUTONIUM FROM NEPTUNIUM

    DOEpatents

    Beaufait, L.J. Jr.

    1958-06-10

    A method is described of separating neptunium from plutonium in an aqueous solution containing neptunium and plutonium in valence states not greater than +4. This may be accomplished by contacting the solution with dichromate ions, thus oxidizing the neptunium to a valence state greater than +4 without oxidizing any substantial amount of plutonium, and then forming a carrier precipitate which carries the plutonium from solution, leaving the neptunium behind. A preferred embodiment of this invention covers the use of lanthanum fluoride as the carrier precipitate.

  14. Oxidation of plutonium dioxide.

    PubMed

    Korzhavyi, Pavel A; Vitos, Levente; Andersson, David A; Johansson, Börje

    2004-04-01

    The physics and chemistry of the actinide elements form the scientific basis for rational handling of nuclear materials. In recent experiments, most unexpectedly, plutonium dioxide has been found to react with water to form higher oxides up to PuO(2.27), whereas PuO(2) had always been thought to be the highest stable oxide of plutonium. We perform a theoretical analysis of this complicated situation on the basis of total energies calculated within density functional theory combined with well-established thermodynamic data. The reactions of PuO(2) with either O(2) or H(2)O to form PuO(2+delta) are calculated to be endothermic: that is, in order to occur they require a supply of energy. However, our calculations show that PuO(2+delta) can be formed, as an intermediate product, by reactions with the products of radiolysis of water, such as H(2)O(2). PMID:15034561

  15. Oxygen potential of a prototypic Mo-cermet fuel containing plutonium oxide

    NASA Astrophysics Data System (ADS)

    Miwa, Shuhei; Osaka, Masahiko; Nozaki, Takahiro; Arima, Tatsumi; Idemitsu, Kazuya

    2015-10-01

    Oxygen potential of a prototypic Mo-cermet fuel containing 50 vol.% PuO2-x were investigated by the thermogravimetric analysis in the temperature range from 1273 K to 1473 K. It was shown that the oxygen potential and oxidation rate of the Mo-cermet were the same as those of pure PuO2-x below the oxygen potential of Mo/MoO2 oxidation reaction. The same features of the Mo-cermet sample containing 50 vol.% PuO2-x with those of pure PuO2-x were discussed in terms of the microstructure.

  16. MATERIAL PROPERTIES OF PLUTONIUM-BEARING OXIDES STORED IN STAINLESS STEEL CONTAINERS

    SciTech Connect

    Kessinger, G.; Almond, P.; Bridges, N.; Bronikowski, M.; Crowder, M.; Duffey, J.; Livingston, R.; Mcelwee, M.; Missimer, D.; Scogin, J.; Summer, M.; Jurgensen, A.

    2010-02-01

    The destructive examination (DE) of 3013 containers after storage is part of the Surveillance and Monitoring Program based on the Department of Energy's standard for long-term storage of Pu (DOE-STD-3013). The stored, Pu-bearing materials may contain alkali halide contamination that varies from trace amounts of salt to about 50 weight percent, with smaller fractions of other compounds and oxides. These materials were characterized prior to packaging, and surveillance characterizations are conducted to determine the behavior of the materials during long term storage. The surveillance characterization results are generally in agreement with the pre-surveillance data. The predominant phases identified by X-ray diffraction are in agreement with the expected phase assemblages of the as-received materials. The measured densities are in reasonable agreement with the expected densities of materials containing the fraction of salts and actinide oxide specified by the pre-surveillance data. The radiochemical results are generally in good agreement with the pre-surveillance data for mixtures containing 'weapons grade' Pu (nominally 94% {sup 239}Pu and 6% {sup 240}Pu); however, the ICP-MS results from the present investigation generally produce lower concentrations of Pu than the pre-surveillance analyses. For mixtures containing 'fuel grade' Pu (nominally 81-93% {sup 239}Pu and 7-19% {sup 240}Pu), the ICP-MS results from the present investigation appear to be in better agreement with the pre-surveillance data than the radiochemistry results.

  17. Low temperature oxidation of plutonium

    SciTech Connect

    Nelson, Art J.; Roussel, Paul

    2013-05-15

    The initial oxidation of gallium stabilized {delta}-plutonium metal at 193 K has been followed using x-ray photoelectron spectroscopy. On exposure to Langmuir quantities of oxygen, plutonium rapidly forms a trivalent oxide followed by a tetravalent plutonium oxide. The growth modes of both oxides have been determined. Warming the sample in vacuum, the tetravalent oxide reduces to the trivalent oxide. The kinetics of this reduction reaction have followed and the activation energy has been determined to be 38.8 kJ mol{sup -1}.

  18. Direct vitrification of plutonium-containing materials (PCM`s) with the glass material oxidation and dissolution system (GMODS)

    SciTech Connect

    Forsberg, C.W. Beahm, E.C.; Parker, G.W.; Rudolph, J.C.; Haas, P.A.; Malling, G.F.; Elam, K.; Ott, L.

    1995-10-30

    The end of the cold war has resulted in excess PCMs from nuclear weapons and associated production facilities. Consequently, the US government has undertaken studies to determine how best to manage and dispose of this excess material. The issues include (a) ensurance of domestic health, environment, and safety in handling, storage, and disposition, (b) international arms control agreements with Russia and other countries, and (c) economics. One major set of options is to convert the PCMs into glass for storage or disposal. The chemically inert characteristics of glasses make them a desirable chemical form for storage or disposal of radioactive materials. A glass may contain only plutonium, or it may contain plutonium along with other radioactive materials and nonradioactive materials. GMODS is a new process for the direct conversion of PCMs (i.e., plutonium metal, scrap, and residues) to glass. The plutonium content of these materials varies from a fraction of a percent to pure plutonium. GMODS has the capability to also convert other metals, ceramics, and amorphous solids to glass, destroy organics, and convert chloride-containing materials into a low-chloride glass and a secondary clean chloride salt strewn. This report is the initial study of GMODS for vitrification of PCMs as input to ongoing studies of plutonium management options. Several tasks were completed: initial analysis of process thermodynamics, initial flowsheet analysis, identification of equipment options, proof-of-principle experiments, and identification of uncertainties.

  19. COLUMBIC OXIDE ADSORPTION PROCESS FOR SEPARATING URANIUM AND PLUTONIUM IONS

    DOEpatents

    Beaton, R.H.

    1959-07-14

    A process is described for separating plutonium ions from a solution of neutron irradiated uranium in which columbic oxide is used as an adsorbert. According to the invention the plutonium ion is selectively adsorbed by Passing a solution containing the plutonium in a valence state not higher than 4 through a porous bed or column of granules of hydrated columbic oxide. The adsorbed plutonium is then desorbed by elution with 3 N nitric acid.

  20. Lithium metal reduction of plutonium oxide to produce plutonium metal

    DOEpatents

    Coops, Melvin S.

    1992-01-01

    A method is described for the chemical reduction of plutonium oxides to plutonium metal by the use of pure lithium metal. Lithium metal is used to reduce plutonium oxide to alpha plutonium metal (alpha-Pu). The lithium oxide by-product is reclaimed by sublimation and converted to the chloride salt, and after electrolysis, is removed as lithium metal. Zinc may be used as a solvent metal to improve thermodynamics of the reduction reaction at lower temperatures. Lithium metal reduction enables plutonium oxide reduction without the production of huge quantities of CaO--CaCl.sub.2 residues normally produced in conventional direct oxide reduction processes.

  1. EVALUATION OF PLUTONIUM OXIDE DESTRUCTIVE CHEMICAL ANALYSES FOR VALIDITY OF ORIGINAL 3013 CONTAINER BINNING

    SciTech Connect

    Mcclard, J.; Kessinger, G.

    2010-02-01

    The surveillance program for 3013 containers is based, in part, on the separation of containers into various bins related to potential container failure mechanisms. The containers are assigned to bins based on moisture content and pre-storage estimates of content chemistry. While moisture content is measured during the packaging of each container, chemistry estimates are made by using a combination of process knowledge, packaging data and prompt gamma analyses to establish the moisture and chloride/fluoride content of the materials. Packages with high moisture and chloride/fluoride contents receive more detailed surveillances than packages with less chloride/fluoride and/or moisture. Moisture verification measurements and chemical analyses performed during the surveillance program provided an opportunity to validate the binning process. Validation results demonstrated that the binning effort was generally successful in placing the containers in the appropriate bin for surveillance and analysis.

  2. Study of Plutonium Oxide Powder Emissions from Simulated Shipping Container Leaks

    SciTech Connect

    Yesso, J. D.; Madia, W. J.; Beatty, G. H.; Schmidt, E. W.; Schwendiman, L C; Mishima, J.

    1980-08-01

    To provide data to facilitate the predictions of PuO{sub 2} emissions through leaks in PuO{sub 2} shipping containers under accident conditions, a series of experiments was conducted using PuO{sub 2} powder and an experimental system designed to simulate a shipping container leak. Over two hundred experiments were completed. The experimental parameters investigated were the leak size/type, internal system pressure, agitation of the apparatus, leak orientation with respect to the powder location and the run time. No single parameter appeared to have any observable effect on the quantities of PuO{sub 2} emitted. However, there was an apparent dependency on the interaction between the orifice area and the internal pressure. The dependency took the form of a function of A{radical}P. Although this functional form was suggested by the data, the data were not sufficient to allow a more detailed function to be determined. The results of experiments in which the run time was variable produced the observation that changes in the run time did not result in changes in the quantities of PuO{sub 2} emitted. This observation led to the conclusion that the majority of PuO{sub 2} observed is emitted during the initial pressurization of the leak tube.

  3. Lawrence Livermore National Laboratory Measurements of Plutonium-bearing Oxide in DOE-STD-3013-2000 Containers Using Calorimetry and Gamma Isotopic Analyses

    SciTech Connect

    Dearborn, D M; Keeton, S C

    2004-06-23

    Lawrence Livermore National Laboratory (LLNL) routinely uses calorimetry and gamma isotopic analyses (Cal/Iso) for the accountability measurement of plutonium (Pu) bearing items. In the past 15 years, the vast majority of those items measured by Cal/Iso were contained in a thin-walled convenience can enclosed in another thin-walled outer container. However, LLNL has recently begun to use DOE-STD-3013-2000 containers as well. These DOE-STD-3013-2000 containers are comprised of a stainless steel convenience can enclosed in welded stainless steel primary and secondary containers. In addition to the fact that the wall thickness of the DOE-STD-3013-2000 containers is much greater than that of other containers in our experience, the DOE-STD-3013-2000 containers appear to have larger thermal insulation characteristics. To date, we have derived Pu-mass values from Cal/Iso measurements of 74 different DOE-STD-3013-2000 containers filled with Pu-bearing oxide or mixed uranium-plutonium (U-Pu) oxide material. Both water-bath and air-bath calorimeters were used for these measurements and both use software to predict when thermal equilibrium is attained. Our experience has shown that after apparent equilibrium has been attained, at least one more complete cycle, and sometimes two or three more complete cycles, is required to gain a measure of true thermal equilibrium. Otherwise, the derived Pu-mass values are less than would be expected from a combination of previously measured Pu-bearing items and would contribute to increased loss in our inventory difference determinations. Conclusions and recommendations drawn from LLNL experience with measurements of Pu mass in Pu-bearing oxide or mixed U-Pu oxide in DOE-STD-3013-2000 containers using the Cal/Iso technique are included.

  4. Plutonium Oxide Process Capability Work Plan

    SciTech Connect

    Meier, David E.; Tingey, Joel M.

    2014-02-28

    Pacific Northwest National Laboratory (PNNL) has been tasked to develop a Pilot-scale Plutonium-oxide Processing Unit (P3U) providing a flexible capability to produce 200g (Pu basis) samples of plutonium oxide using different chemical processes for use in identifying and validating nuclear forensics signatures associated with plutonium production. Materials produced can also be used as exercise and reference materials.

  5. Closure Welding of Plutonium Bearing Storage Containers

    SciTech Connect

    Cannell, G.R.

    2002-02-28

    A key element in the Department of Energy (DOE) strategy for the stabilization, packaging and storage of plutonium-bearing materials involves closure welding of DOE-STD-3013 Outer Containers (3013 container). The 3013 container provides the primary barrier and pressure boundary preventing release of plutonium-bearing materials to the environment. The final closure (closure weld) of the 3013 container must be leaktight, structurally sound and meet DOE STD 3013 specified criteria. This paper focuses on the development, qualification and demonstration of the welding process for the closure welding of Hanford PFP 3013 outer containers.

  6. PRESSURE DEVELOPMENT IN SEALED CONTAINERS WITH PLUTONIUM BEARING MATERIALS

    SciTech Connect

    Duffey, J.; Livingston, R.

    2010-02-01

    Gas generation by plutonium-bearing materials in sealed containers has been studied. The gas composition and pressure are determined over periods from months to years. The Pu-bearing materials studied represent those produced by all of the major processes used by DOE in the processing of plutonium and include the maximum amount of water (0.5% by weight) allowed by DOE's 3013 Standard. Hydrogen generation is of high interest and the Pu-bearing materials can be classed according to how much hydrogen is generated. Hydrogen generation by high-purity plutonium oxides packaged under conditions typical for actual 3013 materials is minimal, with very low generation rates and low equilibrium pressures. Materials with chloride salt impurities have much higher hydrogen gas generation rates and result in the highest observed equilibrium hydrogen pressures. Other materials such as those with high metal oxide impurities generate hydrogen at rates in between these extremes. The fraction of water that is converted to hydrogen gas as equilibrium is approached ranges from 0% to 25% under conditions typical of materials packaged to the 3013 Standard. Generation of both hydrogen and oxygen occurs when liquid water is present. The material and moisture conditions that result in hydrogen and oxygen generation for high-purity plutonium oxide and chloride salt-bearing plutonium oxide materials have been characterized. Other gases that are observed include nitrous oxide, carbon dioxide, carbon monoxide, and methane.

  7. REVIEW OF PLUTONIUM OXIDATION LITERATURE

    SciTech Connect

    Korinko, P.

    2009-11-12

    A brief review of plutonium oxidation literature was conducted. The purpose of the review was to ascertain the effect of oxidation conditions on oxide morphology to support the design and operation of the PDCF direct metal oxidation (DMO) furnace. The interest in the review was due to a new furnace design that resulted in oxide characteristics that are different than those of the original furnace. Very little of the published literature is directly relevant to the DMO furnace operation, which makes assimilation of the literature data with operating conditions and data a convoluted task. The oxidation behavior can be distilled into three regimes, a low temperature regime (RT to 350 C) with a relatively slow oxidation rate that is influenced by moisture, a moderate temperature regime (350-450 C) that is temperature dependent and relies on more or less conventional oxidation growth of a partially protective oxide scale, and high temperature oxidation (> 500 C) where the metal autocatalytically combusts and oxidizes. The particle sizes obtained from these three regimes vary with the finest being from the lowest temperature. It is surmised that the slow growth rate permits significant stress levels to be achieved that help break up the oxides. The intermediate temperatures result in a fairly compact scale that is partially protective and that grows to critical thickness prior to fracturing. The growth rate in this regime may be parabolic or paralinear, depending on the oxidation time and consequently the oxide thickness. The high temperature oxidation is invariant in quiescent or nearly quiescent conditions due to gas blanketing while it accelerates with temperature under flowing conditions. The oxide morphology will generally consist of fine particles (<15 {micro}m), moderately sized particles (15 < x < 250 {micro}m) and large particles (> 250 {micro}m). The particle size ratio is expected to be < 5%, 25%, and 70% for fine, medium and large particles, respectively, for

  8. PLUTONIUM METAL: OXIDATION CONSIDERATIONS AND APPROACH

    SciTech Connect

    Estochen, E.

    2013-03-20

    Plutonium is arguably the most unique of all metals when considered in the combined context of metallurgical, chemical, and nuclear behavior. Much of the research in understanding behavior and characteristics of plutonium materials has its genesis in work associated with nuclear weapons systems. However, with the advent of applications in fuel materials, the focus in plutonium science has been more towards nuclear fuel applications, as well as long term storage and disposition. The focus of discussion included herein is related to preparing plutonium materials to meet goals consistent with non-proliferation. More specifically, the emphasis is on the treatment of legacy plutonium, in primarily metallic form, and safe handling, packaging, and transport to meet non-proliferation goals of safe/secure storage. Elevated temperature oxidation of plutonium metal is the treatment of choice, due to extensive experiential data related to the method, as the oxide form of plutonium is one of only a few compounds that is relatively simple to produce, and stable over a large temperature range. Despite the simplicity of the steps required to oxidize plutonium metal, it is important to understand the behavior of plutonium to ensure that oxidation is conducted in a safe and effective manner. It is important to understand the effect of changes in environmental variables on the oxidation characteristics of plutonium. The primary purpose of this report is to present a brief summary of information related to plutonium metal attributes, behavior, methods for conversion to oxide, and the ancillary considerations related to processing and facility safety. The information provided is based on data available in the public domain and from experience in oxidation of such materials at various facilities in the United States. The report is provided as a general reference for implementation of a simple and safe plutonium metal oxidation technique.

  9. Dissolution behavior of plutonium containing zirconia-magnesia ceramics

    NASA Astrophysics Data System (ADS)

    Holliday, Kiel; Hartmann, Thomas; Cerefice, Gary; Czerwinski, Ken

    2012-03-01

    This study explores the dissolution properties of zirconia-magnesia ceramics containing plutonium as the basis of an inert matrix nuclear fuel. The magnesium oxide phase remains pure MgO, while the zirconia incorporates a small amount of magnesium oxide along with all of the plutonium oxide and erbium oxide. The performance of the material under reactor and repository environments was examined. Reactor conditions are examined using a pressure vessel to expose the material to 300 °C water. To assess the performance of the material as a waste form it was submerged in 90 °C water for 1000 h. In both aqueous dissolution studies there was minimal release of less than 0.8 wt.% of plutonium from the material. To examine the potential for recycling, the dissolution behavior of the fuel matrix was examined in acidic solutions: pure nitric acid and a nitric acid-hydrofluoric acid-peroxide solution. Both acidic media exhibit potential for dissolving plutonium from the zirconia matrix. The experiments performed in this study are meant to lay a foundation for the chemical performance of zirconia-magnesia inert matrix fuel containing fissile material and burnable poison.

  10. Dissolution Behavior of Plutonium Containing Zirconia-Magnesia Ceramics

    SciTech Connect

    Kiel Holliday; Thomas Hartmann; Gary Cerefice; Ken Czerwinski

    2012-03-01

    This study explores the dissolution properties of zirconia-magnesia ceramics containing plutonium as the basis of an inert atrix nuclear fuel. The magnesium oxide phase remains pure MgO, while the zirconia incorporates a small amount of magnesium oxide along with all of the plutonium oxide and erbium oxide. The performance of the material under reactor and repository environments was examined. Reactor conditions are examined using a pressure vessel to expose the material to 300 degrees C water. To assess the performance of the material as a waste form it was submerged in 90 degrees C water for 1000 h. In both aqueous dissolution studies there was minimal release of less than 0.8 wt.% of plutonium from the material. To examine the potential for recycling, the dissolution behavior of the fuel matrix was examined in acidic solutions: pure nitric acid and a nitric acid-hydrofluoric acid-peroxide solution. Both acidic media exhibit potential for dissolving plutonium from the zirconia matrix. The experiments performed in this study are meant to lay a foundation for the chemical performance of zirconia-magnesia inert matrix fuel containing fissile material and burnable poison.

  11. Processing of Non-PFP Plutonium Oxide in Hanford Plants

    SciTech Connect

    Jones, Susan A.; Delegard, Calvin H.

    2011-03-10

    Processing of non-irradiated plutonium oxide, PuO2, scrap for recovery of plutonium values occurred routinely at Hanford’s Plutonium Finishing Plant (PFP) in glovebox line operations. Plutonium oxide is difficult to dissolve, particularly if it has been high-fired; i.e., calcined to temperatures above about 400°C and much of it was. Dissolution of the PuO2 in the scrap typically was performed in PFP’s Miscellaneous Treatment line using nitric acid (HNO3) containing some source of fluoride ion, F-, such as hydrofluoric acid (HF), sodium fluoride (NaF), or calcium fluoride (CaF2). The HNO3 concentration generally was 6 M or higher whereas the fluoride concentration was ~0.5 M or lower. At higher fluoride concentrations, plutonium fluoride (PuF4) would precipitate, thus limiting the plutonium dissolution. Some plutonium-bearing scrap also contained PuF4 and thus required no added fluoride. Once the plutonium scrap was dissolved, the excess fluoride was complexed with aluminum ion, Al3+, added as aluminum nitrate, Al(NO3)3•9H2O, to limit collateral damage to the process equipment by the corrosive fluoride. Aluminum nitrate also was added in low quantities in processing PuF4.

  12. METHOD FOR SEPARATING PLUTONIUM AND FISSION PRODUCTS EMPLOYING AN OXIDE AS A CARRIER FOR FISSION PRODUCTS

    DOEpatents

    Davies, T.H.

    1961-07-18

    Carrier precipitation processes for separating plutonium values from uranium fission products are described. Silicon dioxide or titanium dioxide in a finely divided state is added to an acidic aqueous solution containing hexavalent plutonium ions together with ions of uranium fission products. The supernatant solution containing plutonium ions is then separated from the oxide and the fission products associated therewith.

  13. PROCESS OF ELIMINATING HYDROGEN PEROXIDE IN SOLUTIONS CONTAINING PLUTONIUM VALUES

    DOEpatents

    Barrick, J.G.; Fries, B.A.

    1960-09-27

    A procedure is given for peroxide precipitation processes for separating and recovering plutonium values contained in an aqueous solution. When plutonium peroxide is precipitated from an aqueous solution, the supernatant contains appreciable quantities of plutonium and peroxide. It is desirable to process this solution further to recover plutonium contained therein, but the presence of the peroxide introduces difficulties; residual hydrogen peroxide contained in the supernatant solution is eliminated by adding a nitrite or a sulfite to this solution.

  14. HB-Line Plutonium Oxide Data Collection Strategy

    SciTech Connect

    Watkins, R.; Varble, J.; Jordan, J.

    2015-05-26

    HB-Line and H-Canyon will handle and process plutonium material to produce plutonium oxide for feed to the Mixed Oxide Fuel Fabrication Facility (MFFF). However, the plutonium oxide product will not be transferred to the MFFF directly from HB-Line until it is packaged into a qualified DOE-STD-3013-2012 container. In the interim, HB-Line will load plutonium oxide into an inner, filtered can. The inner can will be placed in a filtered bag, which will be loaded into a filtered outer can. The outer can will be loaded into a certified 9975 with getter assembly in compliance with onsite transportation requirement, for subsequent storage and transfer to the K-Area Complex (KAC). After DOE-STD-3013-2012 container packaging capabilities are established, the product will be returned to HB-Line to be packaged into a qualified DOE-STD-3013-2012 container. To support the transfer of plutonium oxide to KAC and then eventually to MFFF, various material and packaging data will have to be collected and retained. In addition, data from initial HB-Line processing operations will be needed to support future DOE-STD-3013-2012 qualification as amended by the HB-Line DOE Standard equivalency. As production increases, the volume of data to collect will increase. The HB-Line data collected will be in the form of paper copies and electronic media. Paper copy data will, at a minimum, consist of facility procedures, nonconformance reports (NCRs), and DCS print outs. Electronic data will be in the form of Adobe portable document formats (PDFs). Collecting all the required data for each plutonium oxide can will be no small effort for HB-Line, and will become more challenging once the maximum annual oxide production throughput is achieved due to the sheer volume of data to be collected. The majority of the data collected will be in the form of facility procedures, DCS print outs, and laboratory results. To facilitate complete collection of this data, a traveler form will be developed which

  15. METHOD OF MAINTAINING PLUTONIUM IN A HIGHER STATE OF OXIDATION DURING PROCESSING

    DOEpatents

    Thompson, S.G.; Miller, D.R.

    1959-06-30

    This patent deals with the oxidation of tetravalent plutonium contained in an aqueous acid solution together with fission products to the hexavalent state, prior to selective fission product precipitation, by adding to the solution bismuthate or ceric ions as the oxidant and a water-soluble dichromate as a holding oxidant. Both oxidant and holding oxidant are preferably added in greater than stoichiometric quantities with regard to the plutonium present.

  16. Nondestructive assay methods for solids containing plutonium

    SciTech Connect

    Macmurdo, K.W.; Gray, L.W.; Gibbs, A.

    1984-06-01

    Specific nondestructive assay (NDA) methods, e.g. calorimetry, coincidence neutron counting, singles neutron counting, and gamma ray spectrometry, were studied to provide the Savannah River Plant with an NDA method to measure the plutonium content of solid scrap (slag and crucible) generated in the JB-Line plutonium metal production process. Results indicate that calorimetry can be used to measure the plutonium content to within about 3% in 4 to 6 hours by using computerized equilibrium sample power predictive models. Calorimetry results confirm that a bias exists in the present indirect measurement method used to estimate the plutonium content of slag and crucible. Singles neutron counting of slag and crucible can measure plutonium to only +-30%, but coincidence neutron counting methods improve measurement precision to better than +-10% in less than ten minutes. Only four portions of a single slag and crucible sample were assayed, and further study is recommended.

  17. Pulmonary carcinogenesis from plutonium-containing particles

    SciTech Connect

    Thomas, R.G.; Smith, D.M.; Anderson, E.C.

    1980-01-01

    Plutonium administered as an alpha radiation source to the respiratory tracts of Syrian hamsters has resulted in various incidences of neoplasia. Adenomas are the primary lung tumor observed, but adenocarcinomas are also prevalent.

  18. CONCENTRATION AND DECONTAMINATION OF SOLUTIONS CONTAINING PLUTONIUM VALUES BY BISMUTH PHOSPHATE CARRIER PRECIPITATION METHODS

    DOEpatents

    Seaborg, G.T.; Thompson, S.G.

    1960-08-23

    A process is given for isolating plutonium present in the tetravalent state in an aqueous solution together with fission products. First, the plutonium and fission products are coprecipitated on a bismuth phosphate carrier. The precipitate obtained is dissolved, and the plutonium in the solution is oxidized to the hexavalent state (with ceric nitrate, potassium dichromate, Pb/ sub 3/O/sub 4/, sodium bismuthate and/or potassium dichromate). Thereafter a carrier for fission products is added (bismuth phosphate, lanthanum fluoride, ceric phosphate, bismuth oxalate, thorium iodate, or thorium oxalate), and the fission-product precipitation can be repeated with one other of these carriers. After removal of the fission-product-containing precipitate or precipitates. the plutonium in the supernatant is reduced to the tetravalent state (with sulfur dioxide, hydrogen peroxide. or sodium nitrate), and a carrier for tetravalent plutonium is added (lanthanum fluoride, lanthanum hydroxide, lanthanum phosphate, ceric phosphate, thorium iodate, thorium oxalate, bismuth oxalate, or niobium pentoxide). The plutonium-containing precipitate is then dissolved in a relatively small volume of liquid so as to obtain a concentrated solution. Prior to dissolution, the bismuth phosphate precipitates first formed can be metathesized with a mixture of sodium hydroxide and potassium carbonate and plutonium-containing lanthanum fluorides with alkali-metal hydroxide. In the solutions formed from a plutonium-containing lanthanum fluoride carrier the plutonium can be selectively precipitated with a peroxide after the pH was adjusted preferably to a value of between 1 and 2. Various combinations of second, third, and fourth carriers are discussed.

  19. Bulging of cans containing plutonium residues. Summary report

    SciTech Connect

    Van Konynenburg, R.A.; Wood, D.H.; Condit, R.H.; Shikany, S.D.

    1996-03-01

    In 1994, two cans in the Lawrence Livermore National Laboratory Plutonium Facility were found to be bulging as a result of the generation of gases form the plutonium ash residues contained in the cans. This report describes the chronology of this discovery, the response actions that revealed other pressurized cans, the analysis of the causes, the short-term remedial action, a followup inspection of the short-term storage packages, and a review of proposed long-term remedial options.

  20. Characterization and stability of thin oxide films on plutonium surfaces

    NASA Astrophysics Data System (ADS)

    Flores, H. G. García; Roussel, P.; Moore, D. P.; Pugmire, D. L.

    2011-02-01

    X-ray photoelectron spectroscopy (XPS) and Auger electron spectroscopy (AES) were employed to study oxide films on plutonium metal surfaces. Measurements of the relative concentrations of oxygen and plutonium, as well as the resulting oxidation states of the plutonium (Pu) species in the near-surface region are presented. The oxide product of the auto-reduction (AR) of plutonium dioxide films is evaluated and found to be an oxide species which is reduced further than what is expected. The results of this study show a much greater than anticipated extent of auto-reduction and challenge the commonly held notion of the stoichiometric stability of Pu 2O 3 thin-films. The data indicates that a sub-stoichiometric plutonium oxide (Pu 2O 3 - y ) exists at the metal-oxide interface. The level of sub-stoichiometry is shown to depend, in part, on the carbidic contamination of the metal surface.

  1. CORROSION MONITORING OF PLUTONIUM OXIDE AND SNF

    SciTech Connect

    Douglas, D.G.; Haas, C.M.; Smith, C.M.; Ohl, P.C.

    2003-02-27

    While developing a method to measure pressure in totally sealed stainless steel containers holding spent nuclear fuel at the U.S. DOE Hanford Site, Vista Engineering Technologies, LLC (Vista Engineering) personnel adapted the central concept to corrosion monitoring techniques for the same containers. The ability to monitor corrosion within vessels containing spent nuclear fuel, plutonium and other hazardous materials is imperative for safe storage. Vista Engineering personnel have devised a way to monitor corrosion in a totally sealed stainless steel container using a Magnetically Coupled Corrosion Gauge (MCCG) Patent Pending. The MCCG can be used to detect corrosion as well as measure corrosion rate and does not require any penetration of the containment vessel, which minimizes pressure boundary surface area and sensitive weld materials in the vessels.

  2. MOISTURE AND SURFACE AREA MEASUREMENTS OF PLUTONIUM-BEARING OXIDES

    SciTech Connect

    Crowder, M.; Duffey, J.; Livingston, R.; Scogin, J.; Kessinger, G.; Almond, P.

    2009-09-28

    To ensure safe storage, plutonium-bearing oxides are stabilized at 950 C for at least two hours in an oxidizing atmosphere. Stabilization conditions are expected to decompose organic impurities, convert metals to oxides, and result in moisture content below 0.5 wt%. During stabilization, the specific surface area is reduced, which minimizes readsorption of water onto the oxide surface. Plutonium oxides stabilized according to these criteria were sampled and analyzed to determine moisture content and surface area. In addition, samples were leached in water to identify water-soluble chloride impurity content. Results of these analyses for seven samples showed that the stabilization process produced low moisture materials (< 0.2 wt %) with low surface area ({le} 1 m{sup 2}/g). For relatively pure materials, the amount of water per unit surface area corresponded to 1.5 to 3.5 molecular layers of water. For materials with chloride content > 360 ppm, the calculated amount of water per unit surface area increased with chloride content, indicating hydration of hygroscopic salts present in the impure PuO{sub 2}-containing materials. The low moisture, low surface area materials in this study did not generate detectable hydrogen during storage of four or more years.

  3. Historical review of plutonium storage container failures at Lawrence Livermore National Laboratory

    SciTech Connect

    Dodson, K.E.

    1994-05-01

    As part of the DOE Plutonium Vulnerability Assessment, an investigation was made to characterize the can failures at LLNL. Since the LLNL Plutonium Facility was opened for plutonium operations in 1961, there have only been three can failures that could be remembered by plutonium handlers, vault workers, chemical analysts, and material managers. Only one of these can failures was discovered during the processing of more than 606 packages containing plutonium as part of the LLNL Plutonium Inventory Reduction Program. A very low failure rate, especially since some of the 606 cans had been in storage for two to three decades. Two of the three containers that failed were made of aluminum and were packaged with 1.25 inch diameter plutonium metal spheres. The cans were split down their entire length and the plutonium metal was heavily oxidized. The secondary gallon container of the third package failure was found to be imploded in the storage vault. Upon closer examination, the plastic bags around the inner pint can were badly melted and the lid on the can was loose. Like the other two failures, the metal was heavily oxidized. In all three of the can failures, it is theorized that air entered the inner can through incomplete sealing and the oxygen in the air then reacted with the plutonium metal to produce plutonium oxide. Air was supplied to the inner can by permeation through the surrounding plastic bag. The air could have either diffused through the bag or could have been pumped through the twisted and taped ends of the inner most bag. The inner bags and cans were packaged into second bags and cans in an air atmosphere; therefore, trapping air inside the packaging configuration that could have passed through the bags. A failure of the inner can integrity would be necessary for the air to pass into it. In all three LLNL can failure cases, it is believed that the seal of the inner can was not sufficient to prevent a breach of the can environment.

  4. Evaluation of plutonium oxidation using pulsed neutron measurements with {sup 252}Cf

    SciTech Connect

    Valentine, T.E.; Mihalczo, J.T.

    1997-09-01

    The unrecognized oxidation of plutonium in {open_quotes}sealed{close_quotes} canisters poses a unique problem for both material control and accountability. A feasibility study was performed to address the use of randomly pulsed neutron measurements with {sup 252}Cf to determine if plutonium metal in a canister has oxidized without opening the container. The Monte Carlo code MCNP-DSP was used to determine if time-of-flight transmission measurements could be used to determine oxidation of plutonium in {open_quotes}sealed{close_quotes} cans. In the Monte Carlo models, a plutonium button in a can was positioned between a {sup 252}Cf source and a scintillation detector, and the time distribution of counts after {sup 252}Cf fission in the detector was calculated. The time distribution of counts after {sup 252}Cf fission differs between plutonium metal and plutonium oxide because resonances in oxygen will affect transmission of certain energy neutrons from {sup 252}Cf sources in ionization chambers. This method could be used to determine the presence of other materials that react with plutonium in {open_quotes}sealed{close_quotes} cans.

  5. Direct conversion of plutonium-containing materials to borosilicate glass for storage or disposal

    SciTech Connect

    Forsberg, C.W.; Beahm, E.C.

    1995-06-27

    A new process, the Glass Material Oxidation and Dissolution System (GMODS), has been invented for the direct conversion of plutonium metal, scrap, and residue into borosilicate glass. The glass should be acceptable for either the long-term storage or disposition of plutonium. Conversion of plutonium from complex chemical mixtures and variable geometries into homogeneous glass (1) simplifies safeguards and security; (2) creates a stable chemical form that meets health, safety, and environmental concerns; (3) provides an easy storage form; (4) may lower storage costs; and (5) allows for future disposition options. In the GMODS process, mixtures of metals, ceramics, organics, and amorphous solids containing plutonium are fed directly into a glass melter where they are directly converted to glass. Conventional glass melters can accept materials only in oxide form; thus, it is its ability to accept materials in multiple chemical forms that makes GMODS a unique glass making process. Initial proof-of-principle experiments have converted cerium (plutonium surrogate), uranium, stainless steel, aluminum, and other materials to glass. Significant technical uncertainties remain because of the early nature of process development.

  6. PROCESS FOR PRODUCTION OF PLUTONIUM FROM ITS OXIDES

    DOEpatents

    Weissman, S.I.; Perlman, M.L.; Lipkin, D.

    1959-10-13

    A method is described for obtaining a carbide of plutonium and two methods for obtaining plutonium metal from its oxides. One of the latter involves heating the oxide, in particular PuO/sub 2/, to a temperature of 1200 to 1500 deg C with the stoichiometrical amount of carbon to fornn CO in a hard vacuum (3 to 10 microns Hg), the reduced and vaporized plutonium being collected on a condensing surface above the reaction crucible. When an excess of carbon is used with the PuO/sub 2/, a carbide of plutonium is formed at a crucible temperature of 1400 to 1500 deg C. The process may be halted and the carbide removed, or the reaction temperature can be increased to 1900 to 2100 deg C at the same low pressure to dissociate the carbide, in which case the plutonium is distilled out and collected on the same condensing surface.

  7. Chemical and Radiochemical Composition of Thermally Stabilized Plutonium Oxide from the Plutonium Finishing Plant Considered as Alternate Feedstock for the Mixed Oxide Fuel Fabrication Facility

    SciTech Connect

    Tingey, Joel M.; Jones, Susan A.

    2005-07-01

    Eighteen plutonium oxide samples originating from the Plutonium Finishing Plant (PFP) on the Hanford Site were analyzed to provide additional data on the suitability of PFP thermally stabilized plutonium oxides and Rocky Flats oxides as alternate feedstock to the Mixed Oxide Fuel Fabrication Facility (MFFF). Radiochemical and chemical analyses were performed on fusions, acid leaches, and water leaches of these 18 samples. The results from these destructive analyses were compared with nondestructive analyses (NDA) performed at PFP and the acceptance criteria for the alternate feedstock. The plutonium oxide materials considered as alternate feedstock at Hanford originated from several different sources including Rocky Flats oxide, scrap from the Remote Mechanical C-Line (RMC) and the Plutonium Reclamation Facility (PRF), and materials from other plutonium conversion processes at Hanford. These materials were received at PFP as metals, oxides, and solutions. All of the material considered as alternate feedstock was converted to PuO2 and thermally stabilized by heating the PuO2 powder at 950 C in an oxidizing environment. The two samples from solutions were converted to PuO2 by precipitation with Mg(OH)2. The 18 plutonium oxide samples were grouped into four categories based on their origin. The Rocky Flats oxide was divided into two categories, low- and high-chloride Rocky Flats oxides. The other two categories were PRF/RMC scrap oxides, which included scrap from both process lines and oxides produced from solutions. The two solution samples came from samples that were being tested at Pacific Northwest National Laboratory because all of the plutonium oxide from solutions at PFP had already been processed and placed in 3013 containers. These samples originated at the PFP and are from plutonium nitrate product and double-pass filtrate solutions after they had been thermally stabilized. The other 16 samples originated from thermal stabilization batches before canning at

  8. Materials identification and surveillance project item evaluation: Items, impure plutonium oxide (ATL27960) and pure plutonium oxide (PEOR3258)

    SciTech Connect

    Allen, T.; Appert, Q.; Davis, C.

    1997-03-01

    In this report, Los Alamos scientists characterize properties relevant to storage of an impure plutonium oxide (74 mass % plutonium) in accordance with the Department of Energy (DOE) standard DOE-STD-3013-96. This oxide is of interest because it is the first impure plutonium oxide sample to be evaluated and it is similar to other materials that must be stored. Methods used to characterize the oxide at certain points during calcination include surface-area analyses, mass loss-on-ignition (LOI) measurements, elemental analysis, moisture-adsorption measurements, and quantitative supercritical-CO{sub 2} extraction of adsorbed water. Significant decreases in the LOI and surface area occurred as the oxide was calcined at progressively increasing temperatures. Studies indicate that supercritical-CO{sub 2} extraction is an effective method for removing adsorbed water from oxides. We extracted the water from powdered oxides (high-purity ZrO{sub 2}, pure PuO{sub 2}, and impure plutonium oxide) using CO{sub 2} at 3000 psi pressure and 75{degrees}C, and we quantitatively determined it by using gravimetric and dew-point procedures. The effectiveness of the extraction method is demonstrated by good agreement between the amounts of water extracted from pure zirconium and plutonium dioxides and the mass changes obtained from LOI analyses. However, the amount of moisture (0.025 mass %) extracted from the impure plutonium oxide after it had been calcined at 950{degrees}C and stored for a period of months is much less than the LOI value (0.97 mass %). These results imply that the impure plutonium oxide is free of adsorbed water after calcination at 950{degrees}C, even though the sample does not satisfy the LOI requirement of <0.50 mass % for storage.

  9. EXTRACTION METHOD FOR SEPARATING URANIUM, PLUTONIUM, AND FISSION PRODUCTS FROM COMPOSITIONS CONTAINING SAME

    DOEpatents

    Seaborg, G.T.

    1957-10-29

    Methods for separating plutonium from the fission products present in masses of neutron irradiated uranium are reported. The neutron irradiated uranium is first dissolved in an aqueous solution of nitric acid. The plutonium in this solution is present as plutonous nitrate. The aqueous solution is then agitated with an organic solvent, which is not miscible with water, such as diethyl ether. The ether extracts 90% of the uraryl nitrate leaving, substantially all of the plutonium in the aqueous phase. The aqueous solution of plutonous nitrate is then oxidized to the hexavalent state, and agitated with diethyl ether again. In the ether phase there is then obtained 90% of plutonium as a solution of plutonyl nitrate. The ether solution of plutonyl nitrate is then agitated with water containing a reducing agent such as sulfur dioxide, and the plutonium dissolves in the water and is reduced to the plutonous state. The uranyl nitrate remains in the ether. The plutonous nitrate in the water may be recovered by precipitation.

  10. A contained sealed reflux dissolution apparatus for plutonium materials

    SciTech Connect

    Oldham, R.D.; Mitchell, W.G.; Leahy, C.K.; Narayanan, U.I.; Lewis, K.

    1991-12-01

    A containment apparatus has been designed and a procedure developed which permits the overnight operation in a tornado prone area of the sealed reflux dissolution system for the dissolution of hard-to-dissolve plutonium containing materials. A historical review of the development of the apparatus and the procedure used at the New Brunswick Laboratory is presented. The detailed operating procedure, the engineering drawings necessary to fabricate the apparatus, and the Safety Analysis Report containing a worst-case, single occurrence failure analysis are provided in the Appendices. 3 refs.

  11. BAR-CODE BASED WEIGHT MEASUREMENT STATION FOR PHYSICAL INVENTORY TAKING OF PLUTONIUM OXIDE CONTAINERS AT THE MINING AND CHEMICAL COMBINE RADIOCHEMICAL REPROCESSING PLANT NEAR KRASNOYARSK, SIBERIA.

    SciTech Connect

    SUDA,S.

    1999-09-20

    This paper describes the technical tasks being implemented to computerize the physical inventory taking (PIT) at the Mining and Chemical Combine (Gorno-Khimichesky Kombinat, GKhK) radiochemical plant under the US/Russian cooperative nuclear material protection, control, and accounting (MPC and A) program. Under the MPC and A program, Lab-to-Lab task agreements with GKhK were negotiated that involved computerized equipment for item verification and confirmatory measurement of the Pu containers. Tasks under Phase I cover the work for demonstrating the plan and procedures for carrying out the comparison of the Pu container identification on the container with the computerized inventory records. In addition to the records validation, the verification procedures include the application of bar codes and bar coded TIDs to the Pu containers. Phase II involves the verification of the Pu content. A plan and procedures are being written for carrying out confirmatory measurements on the Pu containers.

  12. Conversion of mixed plutonium-uranium oxides. [COPRECAL

    SciTech Connect

    Thomas, L.L.

    1980-04-01

    Coprocessing is among the several reprocessing schemes being considered to improve the proliferation resistance of the back end of the nuclear fuel cycle. Coconversion of mixed oxides has been developed but not demonstrated on a production scale. AGNS developed a preliminary conceptual design for a production scale facility to convert mixed plutonium-uranium nitrate to the mixed oxide.

  13. Method for dissolving plutonium oxide with HI and separating plutonium

    DOEpatents

    Vondra, Benedict L.; Tallent, Othar K.; Mailen, James C.

    1979-01-01

    PuO.sub.2 -containing solids, particularly residues from incomplete HNO.sub.3 dissolution of irradiated nuclear fuels, are dissolved in aqueous HI. The resulting solution is evaporated to dryness and the solids are dissolved in HNO.sub.3 for further chemical reprocessing. Alternatively, the HI solution containing dissolved Pu values, can be contacted with a cation exchange resin causing the Pu values to load the resin. The Pu values are selectively eluted from the resin with more concentrated HI.

  14. Hydrolysis of plutonium: Corrosion kinetics in DMSO solutions containing simulated high explosive and water

    SciTech Connect

    Haschke, J.M.; Pruner, R.E. II

    1995-01-01

    A sequence of experiments is described that address the compatibility of plutonium metal with dimethyl sulfoxide solvent and with solutions containing simulated HMX explosive and simulated explosive plus water. In the absence of water, reaction is slow and forms a thin adherent product layer on clean metal surfaces. Corrosion of oxide-coated plutonium is observed after 15 to 20 days in a solution containing 0.18 mass % (0.11 M) water. After corrosion initiates, the rate accelerates rapidly and attains a value of 0.13 mg Pu/cm{sup 2} h with a surface that is approximately one percent active. Dependence of the Pu + H{sub 2}O reaction on water concentration is evaluated using the data from literature sources. Hazards associated with the use of wet dimethyl sulfoxide as a solvent for removing explosives during weapon dismantlement are identified and a simple method for their mitigation is outlined.

  15. Plutonium Speciation in Support of Oxidative-Leaching Demonstration Test

    SciTech Connect

    Sinkov, Sergey I.

    2007-10-31

    Bechtel National, Inc. (BNI) is evaluating the plutonium speciation in caustic solutions that reasonably represent the process streams from the oxidative-leaching demonstration test. Battelle—Pacific Northwest Division (PNWD) was contracted to develop a spectrophotometric method to measure plutonium speciation at submicromolar (< 10-6 M) concentrations in alkaline solutions in the presence of chromate and carbonate. Data obtained from the testing will be used to identify the oxidation state of Pu(IV), Pu(V), and Pu(VI) species, which potentially could exist in caustic leachates. Work was initially conducted under contract number 24590-101-TSA-W000-00004 satisfying the needs defined in Appendix C of the Research and Technology Plan TSS A-219 to evaluate the speciation of chromium, plutonium, and manganese before and after oxidative leaching. In February 2007, the contract mechanism was switched to Pacific Northwest National Laboratory (PNNL) Operating Contract MOA: 24590-QL-HC9-WA49-00001.

  16. 1. West facade of Plutonium Concentration Facility (Building 233S), ReductionOxidation ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    1. West facade of Plutonium Concentration Facility (Building 233-S), Reduction-Oxidation Building (REDOX-202-S) to the right. Looking east. - Reduction-Oxidation Complex, Plutonium Concentration Facility, 200 West Area, Richland, Benton County, WA

  17. Characterization of representative materials in support of safe, long term storage of surplus plutonium in DOE-STD-3013 containers

    SciTech Connect

    Smith, Paul H; Narlesky, Joshua E; Worl, Laura A; Gillispie, Obie W

    2010-01-01

    The Surveillance and Monitoring Program (SMP) is a joint LANL/SRS effort funded by DOE/EM to provide the technical basis for the safe, long-term storage (up to 50 years) of over 6 metric tons of plutonium stored in over 5000 DOE-STD-3013 containers at various facilities around the DOE complex. The majority of this material is plutonium that is surplus to the nuclear weapons program, and much of it is destined for conversion to mixed oxide fuel for use in US nuclear power plants. The form of the plutonium ranges from relatively pure metal and oxide to very impure oxide. The performance of the 3013 containers has been shown to depend on moisture content and on the levels, types and chemical forms of the impurities. The oxide materials that present the greatest challenge to the storage container are those that contain chloride salts. The chlorides (NaCl, KCl, CaCl{sub 2}, and MgCl{sub 2}) range from less than half of the impurities present to nearly all the impurities. Other common impurities include oxides and other compounds of calcium, magnesium, iron, and nickel. Over the past 15 years the program has collected a large body of experimental data on over 60 samples of plutonium chosen to represent the broader population of materials in storage. This paper will summarize the characterization data, including the origin and process history, particle size, surface area, density, calorimetry, chemical analysis, moisture analysis, prompt gamma, gas generation and corrosion behavior.

  18. Characterization of Representative Materials in Support of Safe, Long Term Storage of Surplus Plutonium in DOE-STD-3013 Containers

    SciTech Connect

    Narlesky, Joshua E.; Stroud, Mary Ann; Smith, Paul Herrick; Wayne, David M.; Mason, Richard E.; Worl, Laura A.

    2013-02-15

    The Surveillance and Monitoring Program is a joint Los Alamos National Laboratory/Savannah River Site effort funded by the Department of Energy-Environmental Management to provide the technical basis for the safe, long-term storage (up to 50 years) of over 6 metric tons of plutonium stored in over 5,000 DOE-STD-3013 containers at various facilities around the DOE complex. The majority of this material is plutonium that is surplus to the nuclear weapons program, and much of it is destined for conversion to mixed oxide fuel for use in US nuclear power plants. The form of the plutonium ranges from relatively pure metal and oxide to very impure oxide. The performance of the 3013 containers has been shown to depend on moisture content and on the levels, types and chemical forms of the impurities. The oxide materials that present the greatest challenge to the storage container are those that contain chloride salts. Other common impurities include oxides and other compounds of calcium, magnesium, iron, and nickel. Over the past 15 years the program has collected a large body of experimental data on 54 samples of plutonium, with 53 chosen to represent the broader population of materials in storage. This paper summarizes the characterization data, moisture analysis, particle size, surface area, density, wattage, actinide composition, trace element impurity analysis, and shelf life surveillance data and includes origin and process history information. Limited characterization data on fourteen nonrepresentative samples is also presented.

  19. CSER 96-027: storage of cemented plutonium residue containers in 55 gallon drums

    SciTech Connect

    Watson, W.T.

    1997-01-20

    A nuclear criticality safety analysis has been performed for the storage of residual plutonium cementation containers, produced at the Plutonium Finishing Plant, in 55 gallon drums. This CSER increases the limit of total plutonium stored in each 55 gallon drum from 100 to 200 grams.

  20. Plutonium release from pressed plutonium oxide fuel pellets in aquatic environments

    SciTech Connect

    Patterson, J.H.; Steinkruger, F.J.; Matlack, G.M.; Heaton, R.C.; Coffelt, K.P.; Herrera, B.

    1983-12-01

    Plutonium oxide pellets (80% /sup 238/Pu, 40 g each) were exposed to fresh water and sea water at two temperatures for 3 y in enclosed glass chambers. The concentrations of plutonium observed in the waters increased linearly with time throughout the experiment. However, the observed release rates were inversely dependent on temperature and salinity, ranging from 160 ..mu..Ci/day for cold fresh water to 1.4 ..mu..Ci/day for warm sea water. The total releases, including the chamber residues, showed similar dependencies. A major portion (typically greater than 50%) of the released plutonium passed through a 0.1-..mu..m filter, with even larger fractions (greater than 80%) for the fresh water systems.

  1. Behavior of plutonium oxide particulates in a simulated Florida environment

    SciTech Connect

    Heaton, R.C.; Patterson, J.H.; Coffelt, K.P.

    1985-08-01

    The behavior of /sup 238/Pu oxide particles (20 to 74 ..mu..m in diameter) deposited on a soil surface was studied by using an environmental test chamber. The soil was obtained from Florida orange groves, and the chamber was set up to simulate a Florida climate. After more than 9 months and more than 60 simulated rainfalls, the plutonium oxide particles remained on top of the soil and showed no evidence of having moved down into the soil column. Plutonium was released into the soil drainages at the rate of 18 ng/m/sup 2//L. This release, which represents a minute portion of the source, appears to correlate with the volume of the drainage rather than with time and probably consists of plutonium attached to very fine soil particles. The average concentration of plutonium observed in the air was 7 fCi/L, which on an absolute basis, represents 8 x 10/sup -12/% of the source material. Thus the generation of airborne plutonium constitutes an insignificant release pathway in terms of the original source. However, the air concentration during, and especially at the beginning of, a rainfall was typically much higher (1400 fCi/L). This concentration decayed rapidly after the end of the rainfall. These results are compared with those from past experiments, and their implications are discussed.

  2. LITERATURE REVIEW FOR OXALATE OXIDATION PROCESSES AND PLUTONIUM OXALATE SOLUBILITY

    SciTech Connect

    Nash, C.

    2012-02-03

    A literature review of oxalate oxidation processes finds that manganese(II)-catalyzed nitric acid oxidation of oxalate in precipitate filtrate is a viable and well-documented process. The process has been operated on the large scale at Savannah River in the past, including oxidation of 20 tons of oxalic acid in F-Canyon. Research data under a variety of conditions show the process to be robust. This process is recommended for oxalate destruction in H-Canyon in the upcoming program to produce feed for the MOX facility. Prevention of plutonium oxalate precipitation in filtrate can be achieved by concentrated nitric acid/ferric nitrate sequestration of oxalate. Organic complexants do not appear practical to sequester plutonium. Testing is proposed to confirm the literature and calculation findings of this review at projected operating conditions for the upcoming campaign. H Canyon plans to commence conversion of plutonium metal to low-fired plutonium oxide in 2012 for eventual use in the Mixed Oxide Fuel (MOX) Facility. The flowsheet includes sequential operations of metal dissolution, ion exchange, elution, oxalate precipitation, filtration, and calcination. All processes beyond dissolution will occur in HB-Line. The filtration step produces an aqueous filtrate that may have as much as 4 M nitric acid and 0.15 M oxalate. The oxalate needs to be removed from the stream to prevent possible downstream precipitation of residual plutonium when the solution is processed in H Canyon. In addition, sending the oxalate to the waste tank farm is undesirable. This report addresses the processing options for destroying the oxalate in existing H Canyon equipment.

  3. NONDESTRUCTIVE EXAMINATION OF PLUTONIUM-BEARING MATERIAL CONTAINERS

    SciTech Connect

    Yerger, L.; Mcclard, J.; Traver, L.; Grim, T.

    2010-02-01

    The first nondestructive examination (NDE) of 3013-type containers as part of the Department of Energy's (DOE's) Integrated Surveillance Program (ISP) was performed in February, 2005. Since that date 280 NDE surveillances on 255 containers have been conducted. These containers were packaged with plutonium-bearing materials at multiple DOE sites. The NDE surveillances were conducted at Hanford, Lawrence Livermore National Laboratory (LLNL), and Savannah River Site (SRS). These NDEs consisted of visual inspection, mass verification, radiological surveys, prompt gamma analysis, and radiography. The primary purpose of performing NDE surveillances is to determine if there has been a significant pressure buildup inside the inner 3013 container. This is done by measuring the lid deflection of the inner 3013 container using radiography images. These lid deflection measurements are converted to pressure measurements to determine if a container has a pressure of a 100 psig or greater. Making this determination is required by Surveillance and Monitoring Plan (S&MP). All 3013 containers are designed to withstand at least 699 psig as specified by DOE-STD-3013. To date, all containers evaluated have pressures under 50 psig. In addition, the radiography is useful in evaluating the contents of the 3013 container as well as determining the condition of the walls of the inner 3013 container and the convenience containers. The radiography has shown no signs of degradation of any container, but has revealed two packaging anomalies. Quantitative pressure measurements based on lid deflections, which give more information than the 'less than or greater than 100 psig' (pass/fail) data are also available for many containers. Statistical analyses of the pass/fail data combined with analysis of the quantitative data show that it is extremely unlikely that any container in the population of 3013 containers considered in this study (e.g., containers packaged according to the DOE-STD-3013 by

  4. Immobilization of plutonium-containing waste into borobasalt, piroxen and andradite mineral-like compositions

    NASA Astrophysics Data System (ADS)

    Matyunin, Yu. I.; Yudintsev, S. V.; Jardine, L. J.

    2000-07-01

    Immobilization of plutonium-containing waste with obtaining stable and solid compositions is one of the problems that require a solution while managing radioactive waste. At VNIINM work is under way to select and synthesize matrix compositions for the immobilization of various-origin plutonium waste with the use of a cold crucible induction melting technology (CCIM). This paper presents information on the synthesis in a muffle furnace and in the CCIM zerium-, uranium- and plutonium-containing borobasalt, piroxen and andradite materials.

  5. Antineutrino monitoring of burning mixed oxide plutonium fuels

    NASA Astrophysics Data System (ADS)

    Hayes, A. C.; Trellue, H. R.; Nieto, Michael Martin; Wilson, W. B.

    2012-02-01

    Background: Antineutrino monitoring of reactors is an enhanced nuclear safeguard that is being explored by several international groups. A key question is whether such a scheme could be used to verify the destruction of plutonium loaded in a reactor as mixed oxide (MOX) fuel.Purpose: To explore the effectiveness of antineutrino monitoring for the purposes of nuclear accountability and safeguarding of MOX plutonium, we examine the magnitude and temporal variation in the antineutrino signals expected for different loadings of MOX fuels.Methods: Reactor burn simulations are carried out for four different MOX fuel loadings and the antineutrino signals as a function of fuel burnup are computed and compared.Results: The antineutrino signals from reactor-grade and weapons-grade MOX are shown to be distinct from those from burning low enriched uranium, and this signal difference increases as the MOX plutonium fraction of the reactor core increases.Conclusion: Antineutrino monitoring could be used to verify the destruction of plutonium in reactors, although verifying the grade of the plutonium being burned is found to be more challenging.

  6. Literature review for oxalate oxidation processes and plutonium oxalate solubility

    SciTech Connect

    Nash, C. A.

    2015-10-01

    A literature review of oxalate oxidation processes finds that manganese(II)-catalyzed nitric acid oxidation of oxalate in precipitate filtrate is a viable and well-documented process. The process has been operated on the large scale at Savannah River in the past, including oxidation of 20 tons of oxalic acid in F-Canyon. Research data under a variety of conditions show the process to be robust. This process is recommended for oxalate destruction in H-Canyon in the upcoming program to produce feed for the MOX facility. Prevention of plutonium oxalate precipitation in filtrate can be achieved by concentrated nitric acid/ferric nitrate sequestration of oxalate. Organic complexants do not appear practical to sequester plutonium. Testing is proposed to confirm the literature and calculation findings of this review at projected operating conditions for the upcoming campaign.

  7. PRODUCTION OF PLUTONIUM FLUORIDE FROM BISMUTH PHOSPHATE PRECIPITATE CONTAINING PLUTONIUM VALUES

    DOEpatents

    Brown, H.S.; Bohlmann, E.G.

    1961-05-01

    A process is given for separating plutonium from fission products present on a bismuth phosphate carrier. The dried carrier is first treated with hydrogen fluoride at between 500 and 600 deg C whereby some fission product fluorides volatilize away from plutonium tetrafluoride, and nonvolatile fission product fluorides are formed then with anhydrous fluorine at between 400 and 500 deg C. Bismuth and plutonium distill in the form of volatile fluorides away from the nonvolatile fission product fluorides. The bismuth and plutonium fluorides are condensed at below 290 deg C.

  8. Plutonium Immobilization Puck Handling

    SciTech Connect

    Kriikku, E.

    1999-01-26

    The Plutonium Immobilization Project (PIP) will immobilize excess plutonium and store the plutonium in a high level waste radiation field. To accomplish these goals, the PIP will process various forms of plutonium into plutonium oxide, mix the oxide powder with ceramic precursors, press the mixture into pucks, sinter the pucks into a ceramic puck, load the pucks into metal cans, seal the cans, load the cans into magazines, and load the magazines into a Defense Waste Processing Facility (DPWF) canister. These canisters will be sent to the DWPF, an existing Savannah River Site (SRS) facility, where molten high level waste glass will be poured into the canisters encapsulating the ceramic pucks. Due to the plutonium radiation, remote equipment will perform these operations in a contained environment. The Plutonium Immobilization Project is in the early design stages and the facility will begin operation in 2005. This paper will discuss the Plutonium Immobilization puck handling conceptual design and the puck handling equipment testing.

  9. Plutonium oxidation and subsequent reduction by Mn(IV) minerals in Yucca Mountain tuff.

    PubMed

    Powell, Brian A; Duff, Martine C; Kaplan, Daniel I; Fjeld, Robert A; Newville, Matthew; Hunter, Douglas B; Bertsch, Paul M; Coates, John T; Eng, Peter; Rivers, Mark L; Serkiz, Steven M; Sutton, Stephen R; Triay, Ines R; Vaniman, David T

    2006-06-01

    Plutonium oxidation state distribution on Yucca Mountain tuff and synthetic pyrolusite (beta-MnO2) suspensions was measured using synchrotron X-ray micro-spectroscopy and microimaging techniques as well as ultrafiltration/solventextraction techniques. Plutonium sorbed to the tuff was preferentially associated with manganese oxides. For both Yucca Mountain tuff and synthetic pyrolusite, Pu(IV) or Pu(V) was initially oxidized to more mobile Pu(V/VI), but over time, the less mobile Pu(IV) became the predominant oxidation state of the sorbed Pu. The observed stability of Pu(IV) on oxidizing surfaces (e.g., pyrolusite), is proposed to be due to the formation of a stable hydrolyzed Pu(IV) surface species. These findings have important implications in estimating the risk associated with the geological burial of radiological waste in areas containing Mn-bearing minerals, such as at the Yucca Mountain or the Hanford Sites, because plutonium will be predominantly in a much less mobile oxidation state (i.e., Pu(IV)) than previously suggested (i.e., Pu(V/VI). PMID:16786687

  10. PLATINUM HEXAFLUORIDE AND METHOD OF FLUORINATING PLUTONIUM CONTAINING MIXTURES THERE-WITH

    DOEpatents

    Malm, J.G.; Weinstock, B.; Claassen, H.H.

    1959-07-01

    The preparation of platinum hexafluoride and its use as a fluorinating agent in a process for separating plutonium from fission products is presented. According to the invention, platinum is reacted with fluorine gas at from 900 to 1100 deg C to form platinum hexafluoride. The platinum hexafluoride is then contacted with the plutonium containing mixture at room temperature to form plutonium hexafluoride which is more volatile than the fission products fluorides and therefore can be isolated by distillation.

  11. Experience making mixed oxide fuel with plutonium from dismantled weapons

    SciTech Connect

    Blair, H.T.; Ramsey, K.B.

    1995-12-31

    Mixed depleted UO{sub 2} and PuO{sub 2} (MOX) pellets prototypic of fuel proposed for use in commercial power reactors were made with plutonium recovered from dismantled weapons. We characterized plutonium dioxide powders that were produced at the Los Alamos and Lawrence Livermore National Laboratories (LANL and LLNL) using various methods to recover the plutonium from weapons parts and to convert It to oxide. The gallium content of the PUO{sub 2} prepared at LANL was the same as in the weapon alloy while the content of that prepared at LLNL was less. The MOX was prepared with a five weight percent plutonium content. We tested various MOX powders milling methods to improve homogeneity and found vibratory milling superior to ball milling. The sintering behavior of pellets made with the PuO{sub 2} from the two laboratories was similar. We evaluated the effects of gallium and of erbium and gadolinium, that are added to the MOX fuel as deplorable neutron absorbers, on the pellet fabrication process and an the sintered pellets. The gallium content of the sintered pellets was <10 ppm, suggesting that the gallium will not be an issue in the reactor, but that it will be an Issue in the operation of the fuel fabrication processing equipment unless it is removed from the PuO{sub 2} before it is blended with the UO{sub 2}.

  12. PLUTONIUM ALLOYS CONTAINING CONTROLLED AMOUNTS OF PLUTONIUM ALLOTROPES OBTAINED BY APPLICATION OF HIGH PRESSURES

    DOEpatents

    Elliott, R.O.; Gschneidner, K.A. Jr.

    1962-07-10

    A method of making stabilized plutonium alloys which are free of voids and cracks and have a controlled amount of plutonium allotropes is described. The steps include adding at least 4.5 at.% of hafnium, indium, or erbium to the melted plutonium metal, homogenizing the resulting alloy at a temperature of 450 deg C, cooling to room temperature, and subjecting the alloy to a pressure which produces a rapid increase in density with a negligible increase in pressure. The pressure required to cause this rapid change in density or transformation ranges from about 800 to 2400 atmospheres, and is dependent on the alloying element. (AEC)

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

    SciTech Connect

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

    2013-03-21

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

  14. Late-occurring pulmonary pathologies following inhalation of mixed oxide (uranium + plutonium oxide) aerosol in the rat.

    PubMed

    Griffiths, N M; Van der Meeren, A; Fritsch, P; Abram, M-C; Bernaudin, J-F; Poncy, J L

    2010-09-01

    Accidental exposure by inhalation to alpha-emitting particles from mixed oxide (MOX: uranium and plutonium oxide) fuels is a potential long-term health risk to workers in nuclear fuel fabrication plants. For MOX fuels, the risk of lung cancer development may be different from that assigned to individual components (plutonium, uranium) given different physico-chemical characteristics. The objective of this study was to investigate late effects in rat lungs following inhalation of MOX aerosols of similar particle size containing 2.5 or 7.1% plutonium. Conscious rats were exposed to MOX aerosols and kept for their entire lifespan. Different initial lung burdens (ILBs) were obtained using different amounts of MOX. Lung total alpha activity was determined by external counting and at autopsy for total lung dose calculation. Fixed lung tissue was used for anatomopathological, autoradiographical, and immunohistochemical analyses. Inhalation of MOX at ILBs ranging from 1-20 kBq resulted in lung pathologies (90% of rats) including fibrosis (70%) and malignant lung tumors (45%). High ILBs (4-20 kBq) resulted in reduced survival time (N = 102; p < 0.05) frequently associated with lung fibrosis. Malignant tumor incidence increased linearly with dose (up to 60 Gy) with a risk of 1-1.6% Gy for MOX, similar to results for industrial plutonium oxide alone (1.9% Gy). Staining with antibodies against Surfactant Protein-C, Thyroid Transcription Factor-1, or Oct-4 showed differential labeling of tumor types. In conclusion, late effects following MOX inhalation result in similar risk for development of lung tumors as compared with industrial plutonium oxide. PMID:20699696

  15. A Study of the Stability and Characterization Plutonium Dioxide and Chemical Characterization [of] Rocky Flats and Los Alamos Plutonium-Containing Incinerator Ash

    SciTech Connect

    Ray, A.K.; Boettger, J.C.; Behrens, Robert G.

    1999-11-29

    In the presentation ''A Study of the Stability and Characterization of Plutonium Dioxide'', the authors discuss their recent work on actinide stabilities and characterization, in particular, plutonium dioxide PuO{sub 2}. Earlier studies have indicated that PuO{sub 2} has the fluorite structure of CaF{sub 2} and typical oxide semiconductor properties. However, detailed results on the bulk electronic structure of this important actinide oxide have not been available. The authors have used all-electron, full potential linear combinations Gaussian type orbitals fitting function (LCGTO-FF) method to study PuO{sub 2}. The LCGTO-FF technique characterized by its use of three independent GTO basis sets to expand the orbitals, charge density, and exchange-correlation integral kernels. Results will be presented on zero pressure using both the Hedin-Lundquist local density approximation (LDA) model or the Perdew-Wang generalized gradient approximation (GGA) model. Possibilities of different characterizations of PuO{sub 2} will be explored. The paper ''Chemical Characterization Rocky Flats and Los Alamos Plutonium-Containing Incinerator Ash'' describes the results of a comprehensive study of the chemical characteristics of virgin, calcined and fluorinated incinerator ash produced at the Rocky Flats Plant and at the Los Alamos National Laboratory prior to 1988. The Rocky Flats and Los Alamos virgin, calcined, and fluorinated ashes were also dissolved using standard nitrate dissolution chemistry. Corresponding chemical evaluations were preformed on the resultant ash heel and the results compared with those of the virgin ash. Fluorination studies using FT spectroscopy as a diagnostic tool were also performed to evaluate the chemistry of phosphorus, sulfur, carbon, and silicon containing species in the ash. The distribution of plutonium and other chemical elements with the virgin ash, ash heel, fluorinated ash, and fluorinated ash heel particulates were studied in detail using

  16. Redox state of plutonium in irradiated mixed oxide fuels

    NASA Astrophysics Data System (ADS)

    Degueldre, C.; Pin, S.; Poonoosamy, J.; Kulik, D. A.

    2014-03-01

    Nowadays, MOX fuels are used in about 20 nuclear power plants around the world. After irradiation, plutonium co-exists with uranium oxide. Due to the redox sensitive nature of UO2 other plutonium oxides than PuO2 potentially present in the fuel may interact with the matrix. The aim of this study is to determine which plutonium species are present in heterogeneous and homogeneous MOX. The results provided by X-ray Absorption Near Edge Spectroscopy (XANES) for non-irradiated as well as irradiated (center and periphery) homogeneous MOX fuel were published earlier and are completed by Extended X-ray Fine Structure (EXAFS) analysis in this work. The EXAFS signals have been extracted using the ATHENA code and the analyses were carried using EXCURE98 as performed earlier for an analogous element. EXAFS shows that plutonium redox state remains tetravalent in the solid solution and that the minor fraction of trivalent Pu must be below 10%. Independently, the study of homogeneous MOX was also approached by thermodynamics of solid solution of (U,Pu)O2. Such solid solutions were modeled using the Gibbs Energy Minimisation (GEM)-Selektor code (developed at LES, NES, PSI) supported by the literature data on such solid solutions. A comparative study was performed showing which plutonium oxides in their respective mole fractions are more likely to occur in (U,Pu)O2. In the modeling, these oxides were set as ideal and non-ideal solid solutions, as well as separate pure phases. Pu exists mainly as PuO2 in the case of separate phases, but can exist under its reduced forms, PuO1.61 and PuO1.5 in minor fraction i.e. ~15% in ideal solid solution (unlikely) and ~10% in non-ideal solid solution (likely) and at temperature around 1300 K. This combined thermodynamic and EXAFS studies confirm independently the results obtained so far by Pu XANES for the same MOX samples.

  17. GTA Welding Research and Development for Plutonium Containment

    SciTech Connect

    Sessions, C.E.

    2002-02-21

    This paper discusses the development of two welding systems that are used to contain actinide metals and oxides for long term storage. The systems are termed the bagless transfer system (BTS) and the outer container welder (OCW) system. The BTS is so named because it permits the containment of actinides without a polymeric package (i.e., bag). The development of these two systems was directed by Department of Energy Standard 3013, hereafter referred to as DOE 3013. This document defines the product and container requirements. In addition, it references national codes and standards for leak rates, ANSI N14.5, and design, American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section VIII (BandPVC).

  18. Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

    SciTech Connect

    Kudinov, K. G.; Tretyakov, A. A.; Sorokin, Yu. P.; Bondin, V. V.; Manakova, L. F.; Jardine, L. J.

    2002-02-26

    In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on a production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration in Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is

  19. Vitrification of simulated radioactive Rocky Flats plutonium containing ash residue with a Stir Melter System

    SciTech Connect

    Marra, J.C.; Kormanyos, K.R.; Overcamp, T.J.

    1996-10-01

    A demonstration trial has been completed in which a simulated Rocky Flats ash consisting of an industrial fly-ash material doped with cerium oxide was vitrified in an alloy tank Stir-Melter{trademark} System. The cerium oxide served as a substitute for plutonium oxide present in the actual Rocky Flats residue stream. The glass developed falls within the SiO{sub 2} + Al{sub 2}O{sub 3}/{Sigma}Alkali/B{sub 2}O{sub 3} system. The glass batch contained approximately 40 wt% of ash, the ash was modified to contain {approximately} 5 wt% CeO{sub 2} to simulate plutonium chemistry in the glass. The ash simulant was mixed with water and fed to the Stir-Melter as a slurry with a 60 wt% water to 40 wt% solids ratio. Glass melting temperature was maintained at approximately 1,050 C during the melting trials. Melting rates as functions of impeller speed and slurry feed rate were determined. An optimal melting rate was established through a series of evolutionary variations of the control variables` settings. The optimal melting rate condition was used for a continuous six hour steady state run of the vitrification system. Glass mass flow rates of the melter were measured and correlated with the slurry feed mass flow. Melter off-gas was sampled for particulate and volatile species over a period of four hours during the steady state run. Glass composition and durability studies were run on samples collected during the steady state run.

  20. Thermal and Physical Properties of Plutonium Dioxide Produced from the Oxidation of Metal: a Data Summary

    SciTech Connect

    Wayne, David M.

    2014-01-13

    The ARIES Program at the Los Alamos National Laboratory removes plutonium metal from decommissioned nuclear weapons, and converts it to plutonium dioxide in a specially-designed Direct Metal Oxidation furnace. The plutonium dioxide is analyzed for specific surface area, particle size distribution, and moisture content. The purpose of these analyses is to certify that the plutonium dioxide powder meets or exceeds the specifications of the end-user, and the specifications for the packaging and transport of nuclear materials. Analytical results from plutonium dioxide from ARIES development activities, from ARIES production activities, from muffle furnace oxidation of metal, and from metal that was oxidized over a lengthy time interval in air at room temperature, are presented. The processes studied produce plutonium dioxide powder with distinct differences in measured properties, indicating the significant influence of oxidation conditions on physical properties.

  1. Criticality experiments with mixed oxide fuel pin arrays in plutonium-uranium nitrate solution

    SciTech Connect

    Lloyd, R.C. ); Smolen, G.R. )

    1988-08-01

    A series of critical experiments was completed with mixed plutonium-uranium solutions having a Pu/(Pu + U) ratio of approximately 0.22 in a boiler tube-type lattice assembly. These experiments were conducted as part of the Criticality Data Development Program between the United States Department of Energy (USDOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. A complete description of the experiments and data are included in this report. The experiments were performed with an array of mixed oxide fuel pins in aqueous plutonium-uranium solutions. The fuel pins were contained in a boiler tube-type tank and arranged in a 1.4 cm square pitch array which resembled cylindrical geometry. One experiment was perfomed with the fuel pins removed from the vessel. The experiments were performed with a water reflector. The concentration of the solutions in the boiler tube-type tank was varied from 4 to 468 g (Pu + U)/liter. The ratio of plutonium to total heavy metal (plutonium plus uranium) was approximately 0.22 for all experiments.

  2. PLUTONIUM UPTAKE BY PLANTS FROM SOIL CONTAINING PLUTONIUM-238 DIOXIDE PARTICLES

    EPA Science Inventory

    Three plant species--alfalfa, lettuce, and radishes were grown in soils contaminated with plutonium-238 dioxide (238)PuO2 at concentrations of 23, 69, 92, and 342 nanocuries per gram (nCi/g). The length of exposure varied from 60 days for the lettuce and radishes to 358 days for ...

  3. Conceptual Design for the Pilot-Scale Plutonium Oxide Processing Unit in the Radiochemical Processing Laboratory

    SciTech Connect

    Lumetta, Gregg J.; Meier, David E.; Tingey, Joel M.; Casella, Amanda J.; Delegard, Calvin H.; Edwards, Matthew K.; Jones, Susan A.; Rapko, Brian M.

    2014-08-05

    This report describes a conceptual design for a pilot-scale capability to produce plutonium oxide for use as exercise and reference materials, and for use in identifying and validating nuclear forensics signatures associated with plutonium production. This capability is referred to as the Pilot-scale Plutonium oxide Processing Unit (P3U), and it will be located in the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. The key unit operations are described, including plutonium dioxide (PuO2) dissolution, purification of the Pu by ion exchange, precipitation, and conversion to oxide by calcination.

  4. Effect of microorganisms on the plutonium oxidation states.

    PubMed

    Lukšienė, Benedikta; Druteikienė, Rūta; Pečiulytė, Dalia; Baltrūnas, Dalis; Remeikis, Vidmantas; Paškevičius, Algimantas

    2012-03-01

    Particular microbes from substrates at the low-level radioactive waste repository in the Ignalina NPP territory were exposed to (239)Pu (IV) at low pH under aerobic conditions. Pu(III) and Pu(IV) were separated and quantitatively evaluated using the modified anion exchange method and alpha spectrometry. Tested bacteria Bacillus mycoides and Serratia marcescens were more effective in Pu reduction than Rhodococcus fascians. Fungi Paecillomyces lilacinus and Absidia spinosa var. spinosa as well as bacterium Rhodococcus fascians did not alter the plutonium oxidation state. PMID:22112595

  5. Direct determination of the intracellular oxidation state of plutonium.

    PubMed

    Gorman-Lewis, Drew; Aryal, Baikuntha P; Paunesku, Tatjana; Vogt, Stefan; Lai, Barry; Woloschak, Gayle E; Jensen, Mark P

    2011-08-15

    Microprobe X-ray absorption near edge structure (μ-XANES) measurements were used to determine directly, for the first time, the oxidation state of intracellular plutonium in individual 0.1-μm(2) areas within single rat pheochromocytoma cells (PC12). The living cells were incubated in vitro for 3 h in the presence of Pu added to the media in different oxidation states (Pu(III), Pu(IV), and Pu(VI)) and in different chemical forms. Regardless of the initial oxidation state or chemical form of Pu presented to the cells, the XANES spectra of the intracellular Pu deposits were always consistent with tetravalent Pu even though the intracellular milieu is generally reducing. PMID:21755934

  6. Direct Determination of the Intracellular Oxidation State of Plutonium

    PubMed Central

    Gorman-Lewis, Drew; Aryal, Baikuntha P.; Paunesku, Tatjana; Vogt, Stefan; Lai, Barry; Woloschak, Gayle E.; Jensen, Mark P.

    2013-01-01

    Microprobe X-ray absorption near edge structure (μ-XANES) measurements were used to determine directly, for the first time, the oxidation state of intracellular plutonium in individual 0.1 μm2 areas within single rat pheochromocytoma cells (PC12). The living cells were incubated in vitro for 3 hours in the presence of Pu added to the media in different oxidation states (Pu(III), Pu(IV), and Pu(VI)) and in different chemical forms. Regardless of the initial oxidation state or chemical form of Pu presented to the cells, the XANES spectra of the intracellular Pu deposits was always consistent with tetravalent Pu even though the intracellular milieu is generally reducing. PMID:21755934

  7. Apparatus and process for the electrolytic reduction of uranium and plutonium oxides

    DOEpatents

    Poa, David S.; Burris, Leslie; Steunenberg, Robert K.; Tomczuk, Zygmunt

    1991-01-01

    An apparatus and process for reducing uranium and/or plutonium oxides to produce a solid, high-purity metal. The apparatus is an electrolyte cell consisting of a first container, and a smaller second container within the first container. An electrolyte fills both containers, the level of the electrolyte in the first container being above the top of the second container so that the electrolyte can be circulated between the containers. The anode is positioned in the first container while the cathode is located in the second container. Means are provided for passing an inert gas into the electrolyte near the lower end of the anode to sparge the electrolyte and to remove gases which form on the anode during the reduction operation. Means are also provided for mixing and stirring the electrolyte in the first container to solubilize the metal oxide in the electrolyte and to transport the electrolyte containing dissolved oxide into contact with the cathode in the second container. The cell is operated at a temperature below the melting temperature of the metal product so that the metal forms as a solid on the cathode.

  8. Flexible process options for the immobilisation of residues and wastes containing plutonium

    SciTech Connect

    Stewart, M.W.A.; Moricca, S.A.; Day, R. A.; Begg, B. D.; Scales, C. R.; Maddrell, E. R.; Eilbeck, A. B.

    2007-07-01

    Residues and waste streams containing plutonium present unique technical, safety, regulatory, security, and socio-political challenges. In the UK these streams range from lightly plutonium contaminated materials (PCM) through to residue s resulting directly from Pu processing operations. In addition there are potentially stocks of Pu oxide powders whose future designation may be either a waste or an asset, due to their levels of contamination making their reuse uneconomic, or to changes in nuclear policy. While waste management routes exist for PCM, an immobilisation process is required for streams containing higher levels of Pu. Such a process is being developed by Nexia Solutions and ANSTO to treat and immobilise Pu waste and residues currently stored on the Sellafield site. The characteristics of these Pu waste streams are highly variable. The physical form of the Pu waste ranges from liquids, sludges, powders/granules, to solid components (e.g., test fuels), with the Pu present as an ion in solution, as a salt, metal, oxide or other compound. The chemistry of the Pu waste streams also varies considerably with a variety of impurities present in many waste streams. Furthermore, with fissile isotopes present, criticality is an issue during operations and in the store or repository. Safeguards and security concerns must be assessed and controlled. The process under development, by using a combination of tailored waste form chemistry combined with flexible process technology aims to develop a process line to handle a broad range of Pu waste streams. It aims to be capable of dealing with not only current arisings but those anticipated to arise as a result of future operations or policy changes. (authors)

  9. Furnace System Testing to Support Lower-Temperature Stabilization of High Chloride Plutonium Oxide Items at the Hanford Plutonium Finishing Plant

    SciTech Connect

    Schmidt, Andrew J.; Gerber, Mark A.; Fischer, Christopher M.; Elmore, Monte R.

    2003-04-16

    High chloride content plutonium (HCP) oxides are impure plutonium oxide scrap which contains NaCl, KCl, MgCl2 and/or CaCl2 salts at potentially high concentrations and must be stabilized at 950 C per the DOE Standard, DOE-STD-3013-2000. The chlorides pose challenges to stabilization because volatile chloride salts and decomposition products can corrode furnace heating elements and downstream ventilation components. Thermal stabilization of HCP items at 750 C (without water washing) is being investigated as an alternative method for meeting the intent of DOE STD 3013-2000. This report presents the results from a series of furnace tests conducted to develop material balance and system operability data for supporting the evaluation of lower-temperature thermal stabilization.

  10. Treatment of plutonium process residues by molten salt oxidation

    SciTech Connect

    Stimmel, J.; Wishau, R.; Ramsey, K.B.; Montoya, A.; Brock, J.; Heslop, M.; Wernly, K.

    1999-04-01

    Molten Salt Oxidation (MSO) is a thermal process that can remove more than 99.999% of the organic matrix from combustible {sup 238}Pu material. Plutonium processing residues are injected into a molten salt bed with an excess of air. The salt (sodium carbonate) functions as a catalyst for the conversion of the organic material to carbon dioxide and water. Reactive species such as fluorine, chlorine, bromine, iodine, sulfur, phosphorous and arsenic in the organic waste react with the molten salt to form the corresponding neutralized salts, NaF, NaCl, NaBr, NaI, Na{sub 2}SO{sub 4}, Na{sub 3}PO{sub 4} and NaAsO{sub 2} or Na{sub 3}AsO4. Plutonium and other metals react with the molten salt and air to form metal salts or oxides. Saturated salt will be recycled and aqueous chemical separation will be used to recover the {sup 238}Pu. The Los Alamos National Laboratory system, which is currently in the conceptual design stage, will be scaled down from current systems for use inside a glovebox.

  11. Thermal Property Evaluation of Cerium Dioxide and Cerium Dioxide Magnesium Oxide Powders for Testing Plutonium

    SciTech Connect

    HOYT, R C

    2002-06-01

    Ceric oxide (CeO{sub 2}) and mixtures of CeO{sub 2} -magnesium oxide (MgO) have been utilized at the Plutonium Finishing Plant (PFP) as surrogate materials to represent plutonium dioxide (PuO{sub 2}) and impure PuO{sub 2} containing impurities such as MgO during verification tests on PFP's stabilization furnaces. Magnesium oxide was selected during furnace testing as the impurity of interest since much of the impure PuO{sub 2} to be stabilized and packaged at the PFP contains significant amounts of MgO from solution stabilization work. The issue being addressed in this study is whether or not heating the surrogate materials to 950 C adequately simulates heating PuO{sub 2} powders to 950 C. This paper evaluates some of the thermal properties of these oxides, as related to the heating of powders of these materials where heat transfer within the powders is governed primarily by conduction. Detailed heat transfer modeling was outside the scope of this paper.

  12. Plutonium

    NASA Astrophysics Data System (ADS)

    Clark, David L.; Hecker, Siegfried S.; Jarvinen, Gordon D.; Neu, Mary P.

    The element plutonium occupies a unique place in the history of chemistry, physics, technology, and international relations. After the initial discovery based on submicrogram amounts, it is now generated by transmutation of uranium in nuclear reactors on a large scale, and has been separated in ton quantities in large industrial facilities. The intense interest in plutonium resulted fromthe dual-use scenario of domestic power production and nuclear weapons - drawing energy from an atomic nucleus that can produce a factor of millions in energy output relative to chemical energy sources. Indeed, within 5 years of its original synthesis, the primary use of plutonium was for the release of nuclear energy in weapons of unprecedented power, and it seemed that the new element might lead the human race to the brink of self-annihilation. Instead, it has forced the human race to govern itself without resorting to nuclear war over the past 60 years. Plutonium evokes the entire gamut of human emotions, from good to evil, from hope to despair, from the salvation of humanity to its utter destruction. There is no other element in the periodic table that has had such a profound impact on the consciousness of mankind.

  13. The structure of plutonium(IV) oxide as hydrolysed clusters in aqueous suspensions.

    PubMed

    Ekberg, Christian; Larsson, Kristian; Skarnemark, Gunnar; Ödegaard-Jensen, Arvid; Persson, Ingmar

    2013-02-14

    The behavior of plutonium still puzzles scientists 70 years after its discovery. There are several factors making the chemistry of plutonium interesting including its ability to keep several oxidation states. Another unique property is that the oxidation states +III, +IV, +V and +VI may exist simultaneously in solution. Another property plutonium shares with some other tetravalent metal ions is the ability to form stable polynuclear complexes or colloids. The structures of freshly prepared and five-year old plutonium(IV) colloids are compared with crystalline plutonium(IV) oxide using Pu L(3)-edge EXAFS. It was shown that as the plutonium colloids age they do in fact shrink in size, contrary to previous expectations. The aged colloidal particles are indeed very small with only 3-4 plutonium atoms, and with a structure very similar to solid plutonium(IV) oxide, but with somewhat shorter mean Pu-O bond and Pu···Pu distances indicating a partial oxidation. The very small size of the colloidal particles is further supported by the fact that they do not sediment on heavy ultra-centrifugation. PMID:23175453

  14. Method for removal of plutonium impurity from americium oxides and fluorides

    DOEpatents

    FitzPatrick, John R.; Dunn, Jerry G.; Avens, Larry R.

    1987-01-01

    Method for removal of plutonium impurity from americium oxides and fluorides. AmF.sub.4 is not further oxidized to AmF.sub.6 by the application of O.sub.2 F at room temperature, while plutonium compounds present in the americium sample are fluorinated to volatile PuF.sub.6, which can readily be separated therefrom, leaving the purified americium oxides and/or fluorides as the solid tetrafluoride.

  15. Method for removal of plutonium impurity from americium oxides and fluorides

    DOEpatents

    FitzPatrick, J.R.; Dunn, J.G.; Avens, L.R.

    1987-02-13

    Method for removal of plutonium impurity from americium oxides and fluorides. AmF/sub 4/ is not further oxidized to AmF/sub 6/ by the application of O/sub 2/F at room temperature thereto, while plutonium compounds present in the americium sample are fluorinated to volatile PuF/sub 6/, which can readily be separated therefrom, leaving the purified americium oxides and/or fluorides as the solid tetrafluoride thereof.

  16. Redox Bias in Loss on Ignition Moisture Measurement for Relatively Pure Plutonium-Bearing Oxide Materials

    SciTech Connect

    Eller, P. G.; Stakebake, J. L.; Cooper, T. D.

    2002-02-26

    This paper evaluates potential analytical bias in application of the Loss on Ignition (LOI) technique for moisture measurement to relatively pure (plutonium assay of 80 wt.% or higher) oxides containing uranium that have been stabilized according to stabilization and storage standard DOE-STD-3013-2000 (STD- 3013). An immediate application is to Rocky Flats (RF) materials derived from high-grade metal hydriding separations subsequently treated by multiple calcination cycles. Specifically evaluated are weight changes due to oxidation/reduction of multivalent impurity oxides that could mask true moisture equivalent content measurement. Process knowledge and characterization of materials representing complex-wide materials to be stabilized and packaged according to STD-3013, and particularly for the immediate RF target stream, indicate that oxides of uranium, iron and gallium are the only potential multivalent constituents expected to be present above 0.5 wt.%. The evaluation show s that of these constituents, with few exceptions, only uranium oxides can be present at a sufficient level to produce weight gain biases significant with respect to the LOI stability test. In general, these formerly high-value, high-actinide content materials are reliably identifiable by process knowledge and measurement. Significant bias also requires that UO2 components remain largely unoxidized after calcination and are largely converted to U3O8 during LOI testing at only slightly higher temperatures. Based on well-established literature, it is judged unlikely that this set of conditions will be realized in practice. We conclude that it is very likely that LOI weight gain bias will be small for the immediate target RF oxide materials containing greater than 80 wt.% plutonium plus a much smaller uranium content. Recommended tests are in progress to confirm these expectations and to provide a more authoritative basis for bounding LOI oxidation/reduction biases. LOI bias evaluation is more

  17. PROCESS OF SECURING PLUTONIUM IN NITRIC ACID SOLUTIONS IN ITS TRIVALENT OXIDATION STATE

    DOEpatents

    Thomas, J.R.

    1958-08-26

    >Various processes for the recovery of plutonium require that the plutonium be obtalned and maintained in the reduced or trivalent state in solution. Ferrous ions are commonly used as the reducing agent for this purpose, but it is difficult to maintain the plutonium in a reduced state in nitric acid solutions due to the oxidizing effects of the acid. It has been found that the addition of a stabilizing or holding reductant to such solution prevents reoxidation of the plutonium. Sulfamate ions have been found to be ideally suitable as such a stabilizer even in the presence of nitric acid.

  18. CONCENTRATION PROCESS FOR PLUTONIUM IONS, IN AN OXIDATION STATE NOT GREATER THAN +4, IN AQUEOUS ACID SOLUTION

    DOEpatents

    Seaborg, G.T.; Thompson, S.G.

    1960-06-14

    A process for concentrating plutonium is given in which plutonium is first precipitated with bismuth phosphate and then, after redissolution, precipitated with a different carrier such as lanthanum fluoride, uranium acetate, bismuth hydroxide, or niobic oxide.

  19. An assessment of the validity of cerium oxide as a surrogate for plutonium oxide gallium removal studies

    SciTech Connect

    Kolman, D.G.; Park, Y.; Stan, M.; Hanrahan, R.J. Jr.; Butt, D.P.

    1999-03-01

    Methods for purifying plutonium metal have long been established. These methods use acid solutions to dissolve and concentrate the metal. However, these methods can produce significant mixed waste, that is, waste containing both radioactive and chemical hazards. The volume of waste produced from the aqueous purification of thousands of weapons would be expensive to treat and dispose. Therefore, a dry method of purification is highly desirable. Recently, a dry gallium removal research program commenced. Based on initial calculations, it appeared that a particular form of gallium (gallium suboxide, Ga{sub 2}O) could be evaporated from plutonium oxide in the presence of a reducing agent, such as small amounts of hydrogen dry gas within an inert environment. Initial tests using ceria-based material (as a surrogate for PuO{sub 2}) showed that thermally-induced gallium removal (TIGR) from small samples (on the order of one gram) was indeed viable. Because of the expense and difficulty of optimizing TIGR from plutonium dioxide, TIGR optimization tests using ceria have continued. This document details the relationship between the ceria surrogate tests and those conducted using plutonia.

  20. Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

    SciTech Connect

    Ade, Brian J; Gauld, Ian C

    2011-10-01

    In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium

  1. Evaluation of filter media for clarification of partially dissolved residues containing plutonium

    SciTech Connect

    Foley, E.S.

    1989-10-09

    A common process in the chemical industry employs the leaching of a desirable component from an insoluble substrate, followed by filtration to produce a clarified solution of the desirable component and a discardable residue. The work described here involved evaluating sintered metal filter media for separating dissolved plutonium from undissolved residues generated at various locations owned by the Department of Energy throughout the United States. The work was performed during a six-week assignment at the Savannah River Laboratory as part of a high school science enrichment program conducted in the summer of 1989. The leach step used included dissolving the plutonium-containing solids in a solution of nitric-hydrofluoric acid. To simulate the partial solubility of the actual plutonium-containing residues, a non-radioactive power plant flyash was used. 6 refs., 14 figs., 1 tab.

  2. Plutonium isotopic analysis system for plutonium samples enriched in sup 238 Pu in EP 60/61 containers

    SciTech Connect

    Ruhter, W.D.

    1990-06-01

    This user's manual is addressed to the Savannah River Site personnel (routine operators and supervisors) who perform measurements with the Pu-238 isotopic analysis system. Each chapter begins with a table of contents that lists the section title, illustrations, and tabular data presented in that chapter. The first chapter in this manual is an introduction to the system. Chapter 2 lists required settings for the system's commercial nuclear instrument modules. System operating procedures are given in Chapter 3. Chapter 4 contains routine and supervisorial operator interactions. Chapter 5 describes the system's short- and long-printout output formats. Chapter 6 gives instructions for changing system parameters. Error messages are listed and described Chapter 7. Chapter 8 contains a reference article on measuring relative plutonium isotopics in plutonium samples enriched in Pu-238. All commercial items mentioned in this manual are assumed to be functioning correctly for the purposes of system operation. Users are referred to individual equipment manufacturers' manuals for details of operation, trouble-shooting, and maintenance of this commercial equipment.

  3. CONVERSION OF RUSSIAN WEAPON-GRADE PLUTONIUM INTO OXIDE FOR MIXED OXIDE (MOX) FUEL FABRICATION.

    SciTech Connect

    Glagovski, E.; Kolotilov, Y.; Glagolenko, Y.; Zygmunt, Stanley J.; Mason, C. F. V.; Hahn, W. K.; Durrer, R. E.; Thomas, S.; Sicard, B.; Herlet, N.; Fraize, G.; Villa, A.

    2001-01-01

    Progress has been made in the Russian Federation towards the conversion of weapons-grade plutonium (w-Pu) into plutonium oxide (PuO{sub 2}) suitable for further manufacture into mixed oxide (MOX) fuels. This program is funded both by French Commissariat x 1'Energie Atomique (CEA) and the US National Nuclear Security Administration (NNSA). The French program was started as a way to make available their expertise gained from manufacturing MOX fuel. The US program was started in 1998 in response to US proliferation concerns and the acknowledged international need to decrease available w-Pu. Russia has selected both the conversion process and the manufacturing site. This paper discusses the present state of development towards fulfilling this mission: the demonstration plant designed to process small amounts of Pu and validate all process stages and the industrial plant that will process up to 5 metric tons of Pu per year.

  4. Redox bias in loss of ignition moisture measurement for relatively pure plutonium-bearing oxide materials.

    SciTech Connect

    Eller, P. G.; Stakebake, J. L.; Cooper, T. D.

    2001-01-01

    This paper evaluates potential analytical bias in application of the Loss on Ignition (LOI) technique for moisture measurement to relatively pure (plutonium assay of 80 wt.% or higher) oxides containing uranium that have been stabilized according to stabilization and storage standard DOE-STD-3013-2000 (STD-3013). An immediate application is to Rocky Flats (RF) materials derived from highgrade metal hydriding separations subsequently treated by multiple calcination cycles. Specifically evaluated are weight changes due to oxidatiodreduction of multivalent impurity oxides that could mask true moisture equivalent content measurement. Process knowledge and characterization of materials representing complex-wide materials to be stabilized and packaged according to STD-3013, and particularly for the immediate RF target stream, indicate that oxides of uranium, iron and gallium are the only potential multivalent constituents expected to be present above 0.5 wt.%. The evaluation shows that of these constituents, with few exceptions, only uranium oxides can be present at a sufficient level to produce weight gain biases significant with respect to the LO1 stability test. In general, these formerly high-value, high-actinide content materials are reliably identifiable by process knowledge and measurement. Si&icant bias also requires that UO1 components remain largely unoxidized after calcination and are largely converted to U30s clsning LO1 testing at only slightly higher temperatures. Based on wellestablished literature, it is judged unlikely that this set of conditions will be realized in practice. We conclude that it is very likely that LO1 weight gain bias will be small for the immediate target RF oxide materials containing greater than 80 wt.% plutonium plus a much smaller uranium content. Recommended tests are in progress to confum these expectations and to provide a more authoritative basis for bounding LO1 oxidatiodreduction biases. LO1 bias evaluation is more difficult

  5. High temperature X-ray diffraction study of the oxidation products and kinetics of uranium-plutonium mixed oxides.

    PubMed

    Strach, Michal; Belin, Renaud C; Richaud, Jean-Christophe; Rogez, Jacques

    2014-12-15

    The oxidation products and kinetics of two sets of mixed uranium-plutonium dioxides containing 14%, 24%, 35%, 46%, 54%, and 62% plutonium treated in air were studied by means of in situ X-ray diffraction (XRD) from 300 to 1773 K every 100 K. The first set consisted of samples annealed 2 weeks before performing the experiments. The second one consisted of powdered samples that sustained self-irradiation damage. Results were compared with chosen literature data and kinetic models established for UO2. The obtained diffraction patterns were used to determine the temperature of the hexagonal M3O8 (M for metal) phase formation, which was found to increase with Pu content. The maximum observed amount of the hexagonal phase in wt % was found to decrease with Pu addition. We conclude that plutonium stabilizes the cubic phases during oxidation, but the hexagonal phase was observed even for the compositions with 62 mol % Pu. The results indicate that self-irradiation defects have a slight impact on the kinetics of oxidation and the lattice parameter even after the phase transformation. It was concluded that the lattice constant of the high oxygen phase was unaffected by the changes in the overall O/M when it was in equilibrium with small quantities of M3O8. We propose that the observed changes in the high oxygen cubic phase lattice parameter are a result of either cation migration or an increase in the miscibility of oxygen in this phase. The solubility of Pu in the hexagonal phase was estimated to be below 14 mol % even at elevated temperatures. PMID:25412433

  6. METHOD OF PREPARING URANIUM, THORIUM, OR PLUTONIUM OXIDES IN LIQUID BISMUTH

    DOEpatents

    Davidson, J.K.; Robb, W.L.; Salmon, O.N.

    1960-11-22

    A method is given for forming compositions, as well as the compositions themselves, employing uranium hydride in a liquid bismuth composition to increase the solubility of uranium, plutonium and thorium oxides in the liquid bismuth. The finely divided oxide of uranium, plutonium. or thorium is mixed with the liquid bismuth and uranium hydride, the hydride being present in an amount equal to about 3 at. %, heated to about 5OO deg C, agitated and thereafter cooled and excess resultant hydrogen removed therefrom.

  7. Analysis of plutonium oxide surrogate residue using laser-induced breakdown spectroscopy

    NASA Astrophysics Data System (ADS)

    Zheng, Hongbo; Yueh, Fang Yu; Miller, Tracy; Singh, Jagdish P.; Zeigler, Kristine E.; Marra, James C.

    2008-09-01

    Laser-induced breakdown spectroscopy was used to determine the elemental composition of a CeO 2 composite powder for process control verification during lanthanide borosilicate glass fabrication. Cerium oxide is used as a surrogate for plutonium oxide, which along with other canister contents will be combined with frit to make glass. Laser-induced breakdown spectroscopy data for the composition of the CeO 2 batch containing concentrations of Ce, Cr, Si, Fe, Ta, Ni, Zn, Al Mg, Gd, and W were quantitatively determined from laser-induced breakdown spectroscopy spectra of both pellet and powder samples. The results of both forms were compared and it was determined that the pellet data gave slightly better precision than the powder sample.

  8. Radiolysis of Salts and Long-Term Storage Issues for Both Pure and Impure PuO{sub 2} Materials in Plutonium Storage Containers

    SciTech Connect

    Lav Tandon

    2000-05-01

    The Material Identification and Surveillance (MIS) project sponsored a literature search on the effects of radiation on salts, with focus on alkali chlorides. The goal of the survey was to provide a basis for estimating the magnitude of {alpha} radiation effects on alkali chlorides that can accompany plutonium oxide (PuO{sub 2}) into storage. Chloride radiolysis can yield potentially corrosive gases in plutonium storage containers that can adversely affect long-term stability. This literature search was primarily done to provide a tutorial on this topic, especially for personnel with nonradiation chemistry backgrounds.

  9. Containment of nitric acid solutions of Plutonium-238

    NASA Astrophysics Data System (ADS)

    Reimus, M. A. H.; Silver, G. L.; Pansoy-Hjelvik, L.; Ramsey, K. B.

    1999-01-01

    The corrosion of various metals that could be used to contain nitric acid solutions of Pu-238 has been studied. Tantalum and tantalum/2.5% tungsten resisted the test solvent better than 304L stainless steel and several INCONEL alloys. The solvent used to imitate nitric acid solutions of Pu-238 contained 70% nitric acid, hydrofluoric acid, and ammonium hexanitratocerate.

  10. Quantitative NDA measurements of advanced reprocessing product materials containing uranium, neptunium, plutonium, and americium

    NASA Astrophysics Data System (ADS)

    Goddard, Braden

    The ability of inspection agencies and facility operators to measure powders containing several actinides is increasingly necessary as new reprocessing techniques and fuel forms are being developed. These powders are difficult to measure with nondestructive assay (NDA) techniques because neutrons emitted from induced and spontaneous fission of different nuclides are very similar. A neutron multiplicity technique based on first principle methods was developed to measure these powders by exploiting isotope-specific nuclear properties, such as the energy-dependent fission cross sections and the neutron induced fission neutron multiplicity. This technique was tested through extensive simulations using the Monte Carlo N-Particle eXtended (MCNPX) code and by one measurement campaign using the Active Well Coincidence Counter (AWCC) and two measurement campaigns using the Epithermal Neutron Multiplicity Counter (ENMC) with various (alpha,n) sources and actinide materials. Four potential applications of this first principle technique have been identified: (1) quantitative measurement of uranium, neptunium, plutonium, and americium materials; (2) quantitative measurement of mixed oxide (MOX) materials; (3) quantitative measurement of uranium materials; and (4) weapons verification in arms control agreements. This technique still has several challenges which need to be overcome, the largest of these being the challenge of having high-precision active and passive measurements to produce results with acceptably small uncertainties.

  11. Use of sulfur hexafluoride and perfluorocarbon tracers in plutonium storage containers for leak detection

    SciTech Connect

    Kung, J.K.

    1998-05-01

    This study involves an investigation of the feasibility of a tracer-based leak detection system for long-term interim plutonium storage. In particular, a protocol has been developed based on the use of inert tracers with varying concentrations in order to {open_quotes}fingerprint{close_quotes} or {open_quotes}tag{close_quotes} specific containers. A particular combination of tracers at specific ratios could be injected into the free volume of each container, allowing for the detection of leaks as well as determination of the location of leaking containers. Based on plutonium storage considerations, sulfur hexafluoride and four perfluorocarbon tracers were selected and should allow a wide range of viable fingerprinting combinations. A {open_quotes}high-low{close_quotes} protocol which uses two distinct chromatographic peak areas or concentration levels, is recommended. Combinations of air exchange rates, detection durations, and detectability limits are examined in order to predict minimum tracer concentrations required for injection in storage containers.

  12. GAS ANALYSES FROM HEADSPACE OF PLUTONIUM-BEARING MATERIALS CONTAINERS

    SciTech Connect

    Almond, P.; Livingston, R.; Traver, L.; Arnold, M.; Bridges, N.; Kessinger, G.; Duffey, J.

    2010-02-01

    The Savannah River National Laboratory (SRNL) 3013 destructive examination program performs surveillances on 3013 containers originating from multiple sites across the DOE complex. The bases for the packaging, storage, and surveillance activities are derived from the Department of Energy's 3013 Standard (DOE-STD-3013-2004). During destructive examination, headspace gas samples are obtained from the 3013 inner container and the annulus between the outer and inner containers. To characterize gas species, the samples are analyzed by gas chromatography (GC), direct-inlet mass spectrometry (DIMS), and Fourier-transform infrared spectroscopy (FTIR). The GC results, as well as other parameters, are utilized as input into the gas evaluation software tool (GEST) program for computation of pre-puncture gas compositions and pressures. Over 30 containers from the Hanford Site and the Rocky Flats Environmental Technology Site (RFETS) have been examined in the first three years of the surveillance program. Several containers were shown to have appreciable hydrogen content (some greater than 30 mol %), yet little or no oxygen was detected in any of the containers, including those exhibiting high hydrogen concentrations. Characteristics including moisture content, surface area, and material composition, along with the headspace gas composition, are utilized in an attempt to explain the chemical behavior of the packaged materials.

  13. HEADSPACE GAS EVALUATION OF WELDED PLUTONIUM STORAGE CONTAINERS

    SciTech Connect

    Hardy, B; Stephen Harris, S; Matthew Arnold, M; Steve Hensel, S

    2008-04-01

    The Can Puncture Device (CPD) serves as a containment vessel during the puncture of nested 3013 containers as part of surveillance operations in K-Area. The purpose of the CPD sampling process is to determine the original pressure and composition of gases within the inner 3013 container. The relation between the composition of the gas sample drawn from the CPD and that originally in the inner 3013 container depends on the degree of mixing that occurs over the interval of time from the puncture to drawing the sample. Gas mixing is bounded by the extremes of no mixing of gases in the inner container and that of complete mixing, in which case the entire CPD system is of uniform composition. Models relating the sample composition and pressure to the initial (pre-puncture) inner can composition and pressure for each of these extremes were developed. Predictions from both models were compared to data from characterization experiments. In the comparison, it was found that the model that assumed complete gas mixing after puncture, the Uniform Mixing Model, showed significantly better agreement with the data than the model that assumed no change in the composition of the inner container, referred to as the Non-Uniform Mixing Model. Both models were implemented as Microsoft{reg_sign} Excel spreadsheet calculations, which utilize macros, to include the effects of uncertainties and biases in the measurements of process parameters and in the models. Potential inleakage of gas from the glovebox is also addressed. The spreadsheet utilizing the Uniform Mixing Model, which was validated by data from the characterization tests, is used to evaluate the pre-puncture composition and pressure within the inner 3013 container. This spreadsheet model is called the Gas Evaluation Software Tool (GEST).

  14. Relativistic density functional theory modeling of plutonium and americium higher oxide molecules.

    PubMed

    Zaitsevskii, Andréi; Mosyagin, Nikolai S; Titov, Anatoly V; Kiselev, Yuri M

    2013-07-21

    The results of electronic structure modeling of plutonium and americium higher oxide molecules (actinide oxidation states VI through VIII) by two-component relativistic density functional theory are presented. Ground-state equilibrium molecular structures, main features of charge distributions, and energetics of AnO3, AnO4, An2On (An=Pu, Am), and PuAmOn, n = 6-8, are determined. In all cases, molecular geometries of americium and mixed plutonium-americium oxides are similar to those of the corresponding plutonium compounds, though chemical bonding in americium oxides is markedly weaker. Relatively high stability of the mixed heptoxide PuAmO7 is noticed; the Pu(VIII) and especially Am(VIII) oxides are expected to be unstable. PMID:23883027

  15. Relativistic density functional theory modeling of plutonium and americium higher oxide molecules

    NASA Astrophysics Data System (ADS)

    Zaitsevskii, Andréi; Mosyagin, Nikolai S.; Titov, Anatoly V.; Kiselev, Yuri M.

    2013-07-01

    The results of electronic structure modeling of plutonium and americium higher oxide molecules (actinide oxidation states VI through VIII) by two-component relativistic density functional theory are presented. Ground-state equilibrium molecular structures, main features of charge distributions, and energetics of AnO3, AnO4, An2On (An=Pu, Am), and PuAmOn, n = 6-8, are determined. In all cases, molecular geometries of americium and mixed plutonium-americium oxides are similar to those of the corresponding plutonium compounds, though chemical bonding in americium oxides is markedly weaker. Relatively high stability of the mixed heptoxide PuAmO7 is noticed; the Pu(VIII) and especially Am(VIII) oxides are expected to be unstable.

  16. Radial plutonium redistribution in mixed-oxide fuel. [LMFBR

    SciTech Connect

    Lawrence, L.A.; Schwinkendorf, K.N.; Karnesky, R.A.

    1981-10-01

    Alpha autoradiographs from all HEDL fuel pin metallography samples are evaluated and catalogued according to different plutonium distribution patterns. The data base is analyzed for effects of fabrication and operating parameters on redistribution.

  17. DISSOLUTION OF PLUTONIUM CONTAINING CARRIER PRECIPITATE BY CARBONATE METATHESIS AND SEPARATION OF SULFIDE IMPURITIES THEREFROM BY SULFIDE PRECIPITATION

    DOEpatents

    Duffield, R.B.

    1959-07-14

    A process is described for recovering plutonium from foreign products wherein a carrier precipitate of lanthanum fluoride containing plutonium is obtained and includes the steps of dissolving the carrier precipitate in an alkali metal carbonate solution, adding a soluble sulfide, separating the sulfide precipitate, adding an alkali metal hydroxide, separating the resulting precipitate, washing, and dissolving in a strong acid.

  18. Development of a container for the transportation and storage of plutonium bearing materials

    SciTech Connect

    Ammerman, D.; Geinitz, R.; Thorp, D.; Rivera, M.

    1998-03-01

    There is a large backlog of plutonium contaminated materials at the Rocky Flats Environmental Technology Site near Denver, Colorado, USA. The clean-up of this site requires this material to be packaged in such a way as to allow for efficient transportation to other sites or to a permanent geologic repository. Prior to off-site shipment of the material, it may be stored on-site for a period of time. For this reason, it is desirable to have a container capable of meeting the requirements for storage as well as the requirements for transportation. Most of the off-site transportation is envisioned to take place using the TRUPACT-II Type B package, with the Waste Isolation Pilot Plant (WIPP) as the destination. Prior to the development of this new container, the TRUPACT-II had a limit of 325 FGE (fissile gram equivalents) of plutonium due to criticality control concerns. Because of the relatively high plutonium content in the material to be transported, transporting 325 FGE per TRUPACT-II is uneconomical. Thus, the purpose of the new containers is to provide criticality control to increase the allowed TRUPACT-II payload and to provide a safe method for on-site storage prior to transport. This paper will describe the analysis and testing used to demonstrate that the Pipe Overpack Container provides safe on-site storage of plutonium bearing materials in unhardened buildings and provides criticality control during transportation within the TRUPACT-II. Analyses included worst-case criticality analyses, analyses of fork-lift time impacts, and analyses of roof structure collapse onto the container. Testing included dynamic crush tests, bare pipe impact tests, a 30-minute totally engulfing pool-fire test, and multiple package impact tests in end-on and side-on orientations.

  19. Evaluation of Possible Surrogates for Validation of the Oxidation Furnace for the Plutonium Disposition Project

    SciTech Connect

    Duncan, A.

    2007-12-31

    The Plutonium Disposition project (PuD) is considering an alternative furnace design for direct metal oxidation (DMO) of plutonium metal to use as a feed for potential disposition routes. The proposed design will use a retort to oxidize the feed at temperatures up to 500 C. The atmosphere will be controlled using a metered mixture of oxygen, helium and argon to control the oxidation at approximately 400 torr. Since plutonium melts at 664 C, and may potentially react with retort material to form a lower melting point eutectic, the oxidation process will be controlled by metering the flow of oxygen to ensure that the bulk temperature of the material does not exceed this temperature. A batch processing time of <24 hours is desirable to meet anticipated furnace throughput requirements. The design project includes demonstration of concept in a small-scale demonstration test (i.e., small scale) and validation of design in a full-scale test. These tests are recommended to be performed using Pu surrogates due to challenges in consideration of the nature of plutonium and operational constraints required when handling large quantities of accountable material. The potential for spreading contamination and exposing workers to harmful levels of cumulative radioactive dose are motivation to utilize non-radioactive surrogates. Once the design is demonstrated and optimized, implementation would take place in a facility designed to accommodate these constraints. Until then, the use of surrogates would be a safer, less expensive option for the validation phase of the project. This report examines the potential for use of surrogates in the demonstration and validation of the DMO furnace for PuD. This report provides a compilation of the technical information and process requirements for the conversion of plutonium metal to oxide by burning in dry environments. Several potential surrogates were evaluated by various criteria in order to select a suitable candidate for large scale

  20. Far-Field Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

    SciTech Connect

    J.P. Nicot

    2000-09-29

    The objective of this calculation is to estimate the quantity of fissile material that could accumulate in fractures in the rock beneath plutonium-ceramic (Pu-ceramic) and Mixed-Oxide (MOX) waste packages (WPs) as they degrade in the potential monitored geologic repository at Yucca Mountain. This calculation is to feed another calculation (Ref. 31) computing the probability of criticality in the systems described in Section 6 and then ultimately to a more general report on the impact of plutonium on the performance of the proposed repository (Ref. 32), both developed concurrently to this work. This calculation is done in accordance with the development plan TDP-DDC-MD-000001 (Ref. 9), item 5. The original document described in item 5 has been split into two documents: this calculation and Ref. 4. The scope of the calculation is limited to only very low flow rates because they lead to the most conservative cases for Pu accumulation and more generally are consistent with the way the effluent from the WP (called source term in this calculation) was calculated (Ref. 4). Ref. 4 (''In-Drift Accumulation of Fissile Material from WPs Containing Plutonium Disposition Waste Forms'') details the evolution through time (breach time is initial time) of the chemical composition of the solution inside the WP as degradation of the fuel and other materials proceed. It is the chemical solution used as a source term in this calculation. Ref. 4 takes that same source term and reacts it with the invert; this calculation reacts it with the rock. In addition to reactions with the rock minerals (that release Si and Ca), the basic mechanisms for actinide precipitation are dilution and mixing with resident water as explained in Section 2.1.4. No other potential mechanism such as flow through a reducing zone is investigated in this calculation. No attempt was made to use the effluent water from the bottom of the invert instead of using directly the effluent water from the WP. This

  1. Plutonium Oxidation State Distribution under Aerobic and Anaerobic Subsurface Conditions for Metal-Reducing Bacteria

    NASA Astrophysics Data System (ADS)

    Reed, D. T.; Swanson, J.; Khaing, H.; Deo, R.; Rittmann, B.

    2009-12-01

    The fate and potential mobility of plutonium in the subsurface is receiving increased attention as the DOE looks to cleanup the many legacy nuclear waste sites and associated subsurface contamination. Plutonium is the near-surface contaminant of concern at several DOE sites and continues to be the contaminant of concern for the permanent disposal of nuclear waste. The mobility of plutonium is highly dependent on its redox distribution at its contamination source and along its 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. The redox distribution of plutonium in the presence of facultative metal reducing bacteria (specifically Shewanella and Geobacter species) was established in a concurrent experimental and modeling study under aerobic and anaerobic conditions. Pu(VI), although relatively soluble under oxidizing conditions at near-neutral pH, does not persist under a wide range of the oxic and anoxic conditions investigated in microbiologically active systems. Pu(V) complexes, which exhibit high chemical toxicity towards microorganisms, are relatively stable under oxic conditions but are reduced by metal reducing bacteria under anaerobic conditions. These facultative metal-reducing bacteria led to the rapid reduction of higher valent plutonium to form Pu(III/IV) species depending on nature of the starting plutonium species and chelating agents present in solution. Redox cycling of these lower oxidation states is likely a critical step in the formation of pseudo colloids that may lead to long-range subsurface transport. The CCBATCH biogeochemical model is used to explain the redox mechanisms and final speciation of the plutonium oxidation state distributions observed. These results for microbiologically active systems are interpreted in the context of their importance in defining the overall migration

  2. Complexation and redox interactions between aqueous plutonium and manganese oxide interfaces

    SciTech Connect

    Shaughnessy, Dawn A.; Nitsche, Heino; Booth, Corwin H.; Shuh, David K.; Waychunas, Glenn A.; Wilson, Richard E.; Cantrell, Kirk J.; Serne, R. Jeffrey

    2001-11-01

    The sorption of Pu(VI) and Pu(V) onto manganite (MnOOH) and Hausmannite (Mn3O4) was studied at pH 5. Manganite sorbed 21-24% from a 1x10-4 M plutonium solution and the hausmannite removed between 43-66% of the plutonium. The increased sorption by hausmannite results from its larger surface area (about twice that of manganite) plus a larger number of active surface sites. X-ray absorption near-edge structure (XANES) spectra taken at the Pu LIII edge were compared to standard spectra of plutonium in single oxidation states. Based on these spectra, it appears that both manganite and hausmannite reduce the higher valent plutonium species to Pu(IV). Between 53-59% of the plutonium was present as Pu(IV) in the manganite samples while 55-61% of the plutonium complexed to the hausmannite had also been reduced to Pu(IV). The exact mechanism behind this redox interaction between the plutonium and the manganese needs to be identified.

  3. High-speed impact test of an air-transportable plutonium nitrate shipping container

    SciTech Connect

    Yoshimura, H.R.; Pope, R.B.; Leisher, W.B.; Joseph, B.J.

    1980-04-01

    To obtain information on package response for comparison with other test environments, a high-speed impact test was performed on a modified Federal Republic of Germany 18B plutonium nitrate air-transportable container. The container, modified with reinforcing rings for improved crush resistance around the inner tube assembly, was impacted at a velocity of 137 m/s onto an unyielding surface. Substantial crushing of the foam overpack and extensive deformation of the container cavity occurred, causing release of the liquid surrogate contents from the titanium shipping container. The container damage resulting from the high-speed pulldown test was more severe than that from a 185-m free fall onto a semirigid surface by a similar container or the crush environment produced by a 9-m drop of a 2-Mg block onto the container resting on an unyielding surface.

  4. Plutonium inventories for stabilization and stabilized materials

    SciTech Connect

    Williams, A.K.

    1996-05-01

    The objective of the breakout session was to identify characteristics of materials containing plutonium, the need to stabilize these materials for storage, and plans to accomplish the stabilization activities. All current stabilization activities are driven by the Defense Nuclear Facilities Safety Board Recommendation 94-1 (May 26, 1994) and by the recently completed Plutonium ES&H Vulnerability Assessment (DOE-EH-0415). The Implementation Plan for accomplishing stabilization of plutonium-bearing residues in response to the Recommendation and the Assessment was published by DOE on February 28, 1995. This Implementation Plan (IP) commits to stabilizing problem materials within 3 years, and stabilizing all other materials within 8 years. The IP identifies approximately 20 metric tons of plutonium requiring stabilization and/or repackaging. A further breakdown shows this material to consist of 8.5 metric tons of plutonium metal and alloys, 5.5 metric tons of plutonium as oxide, and 6 metric tons of plutonium as residues. Stabilization of the metal and oxide categories containing greater than 50 weight percent plutonium is covered by DOE Standard {open_quotes}Criteria for Safe Storage of Plutonium Metals and Oxides{close_quotes} December, 1994 (DOE-STD-3013-94). This standard establishes criteria for safe storage of stabilized plutonium metals and oxides for up to 50 years. Each of the DOE sites and contractors with large plutonium inventories has either started or is preparing to start stabilization activities to meet these criteria.

  5. Plutonium metal vs. oxide determination with the pulse-shape-discrimination-capable plastic scintillator EJ-299-33

    NASA Astrophysics Data System (ADS)

    Pozzi, S. A.; Bourne, M. M.; Dolan, J. L.; Polack, K.; Lawrence, C.; Flaska, M.; Clarke, S. D.; Tomanin, A.; Peerani, P.

    2014-12-01

    Neutron measurements can be used to distinguish plutonium in metal or oxide form, a capability that is of great interest in nuclear nonproliferation, treaty verification, and other applications. This paper describes measurements performed on well-characterized samples of plutonium oxide and plutonium metal using the pulse-shape-discrimination-capable plastic scintillator EJ-299-33. Results are compared to those obtained with a same-sized detector cell using the liquid scintillator EJ-309. The same optimized, digital pulse shape discrimination technique is applied to both detectors and the neutron pulse height distributions are compared. Results show that the EJ-299-33 plastics can be successfully used for plutonium measurements, where the gamma ray to neutron detection ratio is much higher than for typical radioactive sources. Results also show that EJ-299-33 detectors can be used to characterize plutonium samples, specifically to discriminate between plutonium metal and oxide.

  6. The growth and evolution of thin oxide films on delta-plutonium surfaces

    SciTech Connect

    Garcia Flores, Harry G; Pugmire, David L

    2009-01-01

    The common oxides of plutonium are the dioxide (PuO{sub 2}) and the sesquioxide (Pu{sub 2}O{sub 3}). The structure of an oxide on plutonium metal under air at room temperature is typically described as a thick PuO{sub 2} film at the gas-oxide interface with a thinner PuO{sub 2} film near the oxide-metal substrate interface. In a reducing environment, such as ultra high vacuum, the dioxide (Pu{sup 4+}; O/Pu = 2.0) readily converts to the sesquioxide (Pu{sup 3+}; O/Pu = 1.5) with time. In this work, the growth and evolution of thin plutonium oxide films is studied with x-ray photoelectron spectroscopy (XPS) under varying conditions. The results indicate that, like the dioxide, the sesquioxide is not stable on a very clean metal substrate under reducing conditions, resulting in substoichiometric films (Pu{sub 2}O{sub 3-y}). The Pu{sub 2}O{sub 3-y} films prepared exhibit a variety of stoichiometries (y = 0.2-1) as a function of preparation conditions, highlighting the fact that caution must be exercised when studying plutonium oxide surfaces under these conditions and interpreting resulting data.

  7. Doped palladium containing oxidation catalysts

    DOEpatents

    Mohajeri, Nahid

    2014-02-18

    A supported oxidation catalyst includes a support having a metal oxide or metal salt, and mixed metal particles thereon. The mixed metal particles include first particles including a palladium compound, and second particles including a precious metal group (PMG) metal or PMG metal compound, wherein the PMG metal is not palladium. The oxidation catalyst may also be used as a gas sensor.

  8. Long-term exposure of pressed plutonium oxide heat sources to aquatic environments

    SciTech Connect

    Heaton, R.C.; Patterson, J.H.; Kosiewicz, S.T.; Matlack, G.M.; Steinkruger, F.J.; Nelson, G.B.; Vanderborgh, N.E.; Coffelt, K.P.; Herrera, B.

    1984-11-01

    Plutonium-238 oxide fuel pellets were exposed to water for 2.5 to 6.4 yr, and the concentration of plutonium in the water was monitored. Water composition and temperature were found to be important factors in determining the rate of plutonium release into the water. Typical release rates ranged from 10 to 40 ng/m/sup 2//s in cold fresh water and from 0.3 to 11 ng/m/sup 2//s in cold sea water. Release rates in sea water varied over time and sometimes were erratic. The plutonium release per unit area did not depend on the size of the PuO/sub 2/ source. The released plutonium was in an extremely fine form, able to pass through 10,000 molecular weight cutoff filters. Apparent differences in the fuel pellet surfaces after exposure suggest that plutonium release is controlled by physical and chemical processes occurring at the solid-liquid interface. Release mechanisms and their implications are discussed.

  9. Repackaging of Two Bulged Crimp Sealed Cans Containing Plutonium Bearing Materials

    SciTech Connect

    Watkins, R.W.

    2003-04-29

    Two cans containing plutonium bearing materials were found during radiography surveillance activities to be bulged. The cans had been stored in DOT 6M shipping containers at the Savannah River Site. The material in the first can (Item CZA96-179) was packaged can/bag/can configuration with the inner and outer cans being crimp sealed. The crimp sealed innermost can was clearly deformed from the radiography picture taken for surveillance purposes. This material had been stored in the shipping container since the mid 1970s. The second can (Item 50014440) contained plutonium bearing material of a different origin. The material had been repackaged at the Savannah River Site in the mid 1990's, and the repackaged can was stored in a 6M shipping drum. A special puncturing tool, which secured the can and allowed for a very controlled puncture of both outer and inner cans was used in a glovebox. The glovebox has a dry air system and an argon supply. The puncturing tool utilized a non-sparking punch and an argon purge. The cans were repackaged into filtered outer cans. A description of the puncturing tool, repackaging activities, and of the materials will be provided.

  10. PROCESS FOR SEPARATING PLUTONIUM FROM IMPURITIES

    DOEpatents

    Wahl, A.C.

    1957-11-12

    A method is described for separating plutonium from aqueous solutions containing uranium. It has been found that if the plutonium is reduced to its 3+ valence state, and the uranium present is left in its higher valence state, then the differences in solubility between certain salts (e.g., oxalates) of the trivalent plutonium and the hexavalent uranium can be used to separate the metals. This selective reduction of plutonium is accomplished by adding iodide ion to the solution, since iodide possesses an oxidation potential sufficient to reduce plutonium but not sufficient to reduce uranium.

  11. Removal of Uranium from Plutonium Solutions by Anion Exchange

    SciTech Connect

    Rudisill, T.S.

    2002-03-22

    The anion exchange capacity in the HB-Line Phase II Facility will be used to purify plutonium solutions potentially containing significant quantities of depleted uranium. Following purification, the plutonium will be precipitated as an oxalate and calcined to plutonium oxide (PuO2) for storage until final disposition.

  12. Resumption of thermal stabilization of plutonium oxide in Building 707, Rocky Flats Plant, Golden, Colorado. Environmental Assessment

    SciTech Connect

    Not Available

    1994-02-01

    The Department of Energy is proposing thermal stabilization to enhance the safe storage of plutonium at Rocky Flats Plant until decisions are made on long-term storage and disposition of the material. The proposed action is to resume thermal stabilization of pyrophoric plutonium in Building 707 at Rocky Flats Plant. Thermal stabilization would heat the pyrophoric plutonium under controlled conditions in a glovebox furnace to promote full oxidation and convert the material into stable plutonium oxide in the form of PuO{sub 2}. Other activities associated with thermal stabilization would include post-stabilization characterization of non-pyrophoric plutonium and on-site movement of pyrophoric and non-pyrophoric plutonium. This report covers; purpose and need; proposed action; alternatives to the proposed action; affected environment; environmental effects of proposed action and no action alternative; agencies and person consulted; and public participation.

  13. A rationale for maintaining the double containment requirement for plutonium shipments

    SciTech Connect

    Channell, James K.; Anastas, George

    2003-12-31

    Current U.S. Nuclear Regulatory Commission (NRC) transportation regulations (10 CFR 71.63 (b)) require that all shipments containing more than 20 curies of plutonium must be transported in packages that provide double containment. On April 30, 2002 the NRC issued a proposed rule that would eliminate §71.63(b) and the double containment requirement. NRC’s reasons for proposing elimination of §71.63(b) are: (1) compatibility with International Atomic Energy Agency Transportation Safety Standards (which do not have the requirement); (2) the current rule is inconsistent with the A1/A2 system since it applies only to plutonium; (3) double containment causes a heavier package and results in higher transportation costs; (4) the separate inner containment results in additional radiation exposure; and (5) while there would be additional protection from a separate inner container in an accident; this type of approach is not “risk informed nor performance based.” The Environmental Evaluation Group (EEG) has been a proponent of the double containment requirement for the Waste Isolation Pilot Plant (WIPP) shipments for twenty years. This requirement affects shipments to WIPP much more than any other current or planned shipping campaign because reactor fuel elements, metal or metal alloy, and vitrified high-level waste are exempt from §71.63(b). EEG submitted comments on the Proposed Rule on July 26, 2002 (Appendix C). This report is an update and expansion of the July 26, 2002 comments. Actual WIPP experience with shipments in the double contained TRUPACT-II package is used to respond to NRC arguments for deletion of §71.63(b) and offers a rationale for maintaining the current requirement.

  14. Mixed oxide fuels testing in the advanced test reactor to support plutonium disposition

    SciTech Connect

    Ryskamp, J.M.; Sterbentz, J.W.; Chang, G.S.

    1995-09-01

    An intense worldwide effort is now under way to find means of reducing the stockpile of weapons-grade plutonium. One of the most attractive solutions would be to use WGPu as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PUO{sub 2}) mixed with urania (UO{sub 2}). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification, (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania, (3) The effects of WGPu isotopic composition, (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight, (5) The effects of americium and gallium in WGPu, (6) Fission gas release from MOX fuel pellets made from WGPu, (7) Fuel/cladding gap closure, (8) The effects of power cycling and off-normal events on fuel integrity, (9) Development of radial distributions of burnup and fission products, (10) Power spiking near the interfaces of MOX and urania fuel assemblies, and (11) Fuel performance code validation. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified.

  15. Design of the Laboratory-Scale Plutonium Oxide Processing Unit in the Radiochemical Processing Laboratory

    SciTech Connect

    Lumetta, Gregg J.; Meier, David E.; Tingey, Joel M.; Casella, Amanda J.; Delegard, Calvin H.; Edwards, Matthew K.; Orton, Robert D.; Rapko, Brian M.; Smart, John E.

    2015-05-01

    This report describes a design for a laboratory-scale capability to produce plutonium oxide (PuO2) for use in identifying and validating nuclear forensics signatures associated with plutonium production, as well as for use as exercise and reference materials. This capability will be located in the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. The key unit operations are described, including PuO2 dissolution, purification of the Pu by ion exchange, precipitation, and re-conversion to PuO2 by calcination.

  16. Separation medium containing thermally exfoliated graphite oxide

    NASA Technical Reports Server (NTRS)

    Prud'homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor); Herrera-Alonso, Margarita (Inventor)

    2012-01-01

    A separation medium, such as a chromatography filling or packing, containing a modified graphite oxide material, which is a thermally exfoliated graphite oxide with a surface area of from about 300 m.sup.2/g to 2600 m.sup.2/g, wherein the thermally exfoliated graphite oxide has a surface that has been at least partially functionalized.

  17. Oxidation-state distribution of plutonium in surface and subsurface waters at Thule, northwest Greenland.

    PubMed

    McMahon, C A; León Vintró, L L; Mitchell, P I; Dahlgaard, H

    2000-03-01

    The speciation of plutonium in Arctic waters sampled on the northwest Greenland shelf in August 1997 is discussed in this paper. Specifically, we report the results of analyses carried out on seawater sampled (a) close to the Thule air base where, in 1968, a US military aircraft carrying four nuclear weapons crashed on sea ice, releasing kilogram quantities of plutonium to the snow pack and underlying seabed sediments, and (b) at a reference station (Upernavik) located approximately 400 km to the south. The data show that most of the plutonium in the dissolved phase at Thule is in the form of Pu(V, VI) (mean: 68+/-6%; n = 6), with little if any distinction apparent between surface and bottom waters. Further, the oxidation state distribution at stations close to the accident site is similar to that measured at Upernavik, remote from this site. It is also similar to the distribution observed in shelf waters at mid-latitudes, suggesting that the underlying processes controlling plutonium speciation are insensitive to temperature over the range 0-25 degrees C. Measurements using tangential-flow ultrafiltration indicate that virtually all of the plutonium (including the fraction in a reduced chemical form) is present as fully dissolved species. Most of this plutonium would seem to be of weapons fallout origin, as the mean 238Pu/239,240Pu activity ratio in the water column (dissolved phase) at Thule (0.06+/-0.02; n = 10) is similar to the global fallout ratio at this latitude (approximately 0.04). Thus, there is little evidence of weapons-grade plutonium in the water column at Thule at the present time. PMID:10724428

  18. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    SciTech Connect

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  19. Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

    SciTech Connect

    Kudinov, K.G.; Tretyakov, A.A.; Sorokin, Y.P.; Bondin, V.V.; Manakova, L.F.; Jardine, L.J.

    2001-12-01

    In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is incineration

  20. Effect of cooling rate on achieving thermodynamic equilibrium in uranium-plutonium mixed oxides

    NASA Astrophysics Data System (ADS)

    Vauchy, Romain; Belin, Renaud C.; Robisson, Anne-Charlotte; Hodaj, Fiqiri

    2016-02-01

    In situ X-ray diffraction was used to study the structural changes occurring in uranium-plutonium mixed oxides U1-yPuyO2-x with y = 0.15; 0.28 and 0.45 during cooling from 1773 K to room-temperature under He + 5% H2 atmosphere. We compare the fastest and slowest cooling rates allowed by our apparatus i.e. 2 K s-1 and 0.005 K s-1, respectively. The promptly cooled samples evidenced a phase separation whereas samples cooled slowly did not due to their complete oxidation in contact with the atmosphere during cooling. Besides the composition of the annealing gas mixture, the cooling rate plays a major role on the control of the Oxygen/Metal ratio (O/M) and then on the crystallographic properties of the U1-yPuyO2-x uranium-plutonium mixed oxides.

  1. Plutonium dissolution process

    DOEpatents

    Vest, Michael A.; Fink, Samuel D.; Karraker, David G.; Moore, Edwin N.; Holcomb, H. Perry

    1996-01-01

    A two-step process for dissolving plutonium metal, which two steps can be carried out sequentially or simultaneously. Plutonium metal is exposed to a first mixture containing approximately 1.0M-1.67M sulfamic acid and 0.0025M-0.1M fluoride, the mixture having been heated to a temperature between 45.degree. C. and 70.degree. C. The mixture will dissolve a first portion of the plutonium metal but leave a portion of the plutonium in an oxide residue. Then, a mineral acid and additional fluoride are added to dissolve the residue. Alteratively, nitric acid in a concentration between approximately 0.05M and 0.067M is added to the first mixture to dissolve the residue as it is produced. Hydrogen released during the dissolution process is diluted with nitrogen.

  2. Recycle of scrap plutonium-238 oxide fuel to support future radioisotope applications

    NASA Astrophysics Data System (ADS)

    Schulte, Louis D.; Purdy, Geraldine M.; Jarvinen, Gordon D.; Ramsey, Kevin; Silver, Gary L.; Espinoza, Jacob; Rinehart, Gary H.

    1998-01-01

    The Nuclear Materials Technology (NMT) Division of Los Alamos National Laboratory has initiated a development program to recover & purify plutonium-238 oxide from impure feed sources in a glove box environment. A glove box line has been designed and a chemistry flowsheet developed to perform this recovery task at large scale. The initial demonstration effort focused on purification of 238PuO2 fuel by HNO3/HF dissolution, followed by plutonium(III) oxalate precipitation and calcination to an oxide. Decontamination factors for most impurities of concern in the fuel were very good, producing 238PuO2 fuel significantly better in purity than specified by General Purpose Heat Source (GPHS) fuel powder specifications. A sufficient quantity of purified 238PuO2 fuel was recovered from the process to allow fabrication of a GPHS unit for testing. The results are encouraging for recycle of relatively impure plutonium-238 oxide and scrap residue items into fuel for useful applications. The high specific activity of plutonium-238 magnifies the consequences and concerns of radioactive waste generation. This work places an emphasis on development of waste minimization technologies to complement the aqueous processing operation. Results from experiments on neutralized solutions of plutonium-238 resulted in decontamination to about 1 millicurie/L. Combining ultrafiltration treatment with addition of a water-soluble polymer designed to coordinate Pu, allowed solutions to be decontaminated to about 1 microcurie/L. Efforts continue to develop a capability for efficient, safe, cost-effective, and environmentally acceptable methods to recover and purify 238PuO2 fuel.

  3. Development program to recycle and purify plutonium-238 oxide fuel from scrap

    NASA Astrophysics Data System (ADS)

    Schulte, Louis D.; Silver, Gary L.; Avens, Larry R.; Jarvinen, Gordon D.; Espinoza, Jacob; Foltyn, Elizabeth M.; Rinehart, Gary H.

    1997-01-01

    Nuclear Materials Technology (NMT) Division of Los Alamos National Laboratory (LANL) has initiated a development program to recover & purify plutonium-238 oxide from impure sources. A glove box line has been designed and a process flowsheet developed to perform this task on a large scale. Our initial effort has focused on purification of 238PuO2 fuel that fails to meet General Purpose Heat Source (GPHS) specifications because of impurities. The most notable non-actinide impurity was silicon, but aluminum, chromium, iron and nickel were also near or in excess of limits specified by GPHS fuel powder specifications. 234U was by far the largest actinide impurity observed in the feed material because it is the daughter product of 238Pu by alpha decay. An aqueous method based on nitric acid was selected for purification of the 238PuO2 fuel. All aqueous processing used high purity reagents, and was performed in PTFE apparatus to minimize introduction of new contaminants. Impure 238PuO2 was finely milled, then dissolved in refluxing HNO3/HF and the solution filtered. The dissolved 238Pu was adjusted to the trivalent state by an excess of reducing reagents to compensate for radiolytic effects, precipitated as plutonium(III) oxalate, and recovered by filtration. The plutonium(III) oxalate was subsequently calcined to convert the plutonium to the oxide. Decontamination factors for silicon, phosphorus and uranium were excellent. Decontamination factors for aluminum, chromium, iron and nickel were very good. The purity of the 238PuO2 recovered from this operation was significantly better than specifications. Efforts continue to develop the capability for efficient, safe, cost-effective, and environmentally acceptable methods to recover and purify 238PuO2 fuel in a glove box environment. Plutonium-238 materials targeted for recovery includes impure oxide and scrap items that are lean in 238Pu values.

  4. The AL-R8 SI: The next generation staging container for plutonium pits at the USDOE Pantex Plant

    SciTech Connect

    Eifert, E.J.; Vickers, L.D.

    1999-11-01

    The AL-R8 SI (sealed insert) is the next generation staging container for plutonium pits at the US DOE Pantex Plant. The sealed insert is a stainless steel container that will be placed inside a modified AL-R8 container to stage pits. A pit is a hollow sphere of plutonium metal which is the primary fissionable material in nuclear weapons (warheads and bombs). It is hermetically sealed by a cladding material, which is usually stainless steel. Personnel exposures to ionizing radiation from the pits in storage are expected to decrease due to the attenuation provided by the new SI. All personnel exposures to ionizing radiation at Pantex Plant are As Low As Reasonably Achievable (ALARA). Pantex Plant secures the common defense and national security of the US by safely staging plutonium pits in a manner that protests the health and safety of employees, the public, and the environment.

  5. The AL-R8 SI: the next generation staging container for plutonium pits at the USDOE Pantex Plant.

    PubMed

    Eifert, E J; Vickers, L D

    1999-11-01

    The AL-R8 SI (sealed insert) is the next generation staging container for plutonium pits at the U.S. DOE Pantex Plant. The sealed insert is a stainless steel container that will be placed inside a modified AL-R8 container to stagepits. A pit is a hollow sphere of plutonium metal which is the primary fissionable material in nuclear weapons (warheads and bombs). It is hermetically sealed by a cladding material, which is usually stainless steel. Personnel exposures to ionizing radiation from the pits in storage are expected to decrease due to the attenuation provided by the new SI. All personnel exposures to ionizing radiation at Pantex Plant are As Low As Reasonably Achievable (ALARA). Pantex Plant secures the common defense and national security of the United States by safely staging plutonium pits in a manner that protects the health and safety of employees, the public, and the environment. PMID:10527158

  6. SMALL-SCALE TESTING OF PLUTONIUM (IV) OXALATE PRECIPITATION AND CALCINATION TO PLUTONIUM OXIDE TO SUPPORT THE MOX FEED MISSION

    SciTech Connect

    Crowder, M.; Pierce, R.; Scogin, J.; Daniel, G.; King, W.

    2012-06-25

    The H-Canyon facility will be used to dissolve Pu metal for subsequent purification and conversion to plutonium dioxide (PuO{sub 2}) using Phase II of HB-Line. To support the new mission, SRNL conducted a series of experiments to produce calcined plutonium (Pu) oxide and measure the physical properties and water adsorption of that material. This data will help define the process operating conditions and material handling steps for HB-Line. An anion exchange column experiment produced 1.4 L of a purified 52.6 g/L Pu solution. Over the next nine weeks, seven Pu(IV) oxalate precipitations were performed using the same stock Pu solution, with precipitator feed acidities ranging from 0.77 M to 3.0 M nitric acid and digestion times ranging from 5 to 30 minutes. Analysis of precipitator filtrate solutions showed Pu losses below 1% for all precipitations. The four larger precipitation batches matched the target oxalic acid addition time of 44 minutes within 4 minutes. The three smaller precipitation batches focused on evaluation of digestion time and the oxalic acid addition step ranged from 25-34 minutes because of pump limitations in the low flow range. Following the precipitations, 22 calcinations were performed in the range of 610-690 C, with the largest number of samples calcined at either 650 or 635 C. Characterization of the resulting PuO{sub 2} batches showed specific surface areas in the range of 5-14 m{sup 2}/g, with 16 of the 22 samples in the range of 5-10 m2/g. For samples analyzed with typical handling (exposed to ambient air for 15-45 minutes with relative humidities of 20-55%), the moisture content as measured by Mass Spectrometry ranged from 0.15 to 0.45 wt % and the total mass loss at 1000 C, as measured by TGA, ranged from 0.21 to 0.58 wt %. For the samples calcined between 635 and 650 C, the moisture content without extended exposure ranged from 0.20 to 0.38 wt %, and the TGA mass loss ranged from 0.26 to 0.46 wt %. Of these latter samples, the samples

  7. Tire containing thermally exfoliated graphite oxide

    NASA Technical Reports Server (NTRS)

    Prud'homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor)

    2011-01-01

    A tire, tire lining or inner tube, containing a polymer composite, made of at least one rubber and/or at least one elastomer and a modified graphite oxide material, which is a thermally exfoliated graphite oxide with a surface area of from about 300 sq m/g to 2600 sq m/g.

  8. CONVERSION OF PLUTONIUM TRIFLUORIDE TO PLUTONIUM TETRAFLUORIDE

    DOEpatents

    Fried, S.; Davidson, N.R.

    1957-09-10

    A large proportion of the trifluoride of plutonium can be converted, in the absence of hydrogen fluoride, to the tetrafiuoride of plutonium. This is done by heating plutonium trifluoride with oxygen at temperatures between 250 and 900 deg C. The trifiuoride of plutonium reacts with oxygen to form plutonium tetrafluoride and plutonium oxide, in a ratio of about 3 to 1. In the presence of moisture, plutonium tetrafluoride tends to hydrolyze at elevated temperatures and therefore it is desirable to have the process take place under anhydrous conditions.

  9. Physics and thermal hydraulics design of a small water cooled reactor fuelled with plutonium in rock-like oxide (ROX) form

    SciTech Connect

    Gaultier, M.; Danguy, G.; Perry, A.; Williams, A.; Brushwood, J.; Thompson, A.; Beeley, P. A.

    2006-07-01

    This paper describes the Physics and Thermal Hydraulics areas of a design study for a small water-cooled reactor. The aim was to design a Pressurised Water Reactor (PWR) of maximum power 80 MWt, using a dispersed layout, capable of maximising primary natural circulation flow. The reactor fuel consists of plutonium contained in granular form within a Rock-like Oxide (ROX) pellet structure. (authors)

  10. LOCALIZED CORROSION OF AUSTENITIC STAINLESS STEELEXPOSED TO MIXTURES OF PLUTONIUM OXIDE AND CHLORIDE SALTS

    SciTech Connect

    Zapp, P; Kerry Dunn, K; Jonathan Duffey, J; Ron Livingston, R; Zane Nelson, Z

    2008-11-21

    Laboratory corrosion tests were conducted to investigate the corrosivity of moist plutonium oxide/chloride (PuO{sub 2}/Cl-) salt mixtures on 304L and 316L stainless steel coupons. The tests exposed flat coupons for pitting evaluation and 'teardrop' stressed coupons for stress corrosion cracking (SCC) evaluation at room temperature to various mixtures of PuO{sub 2} and chloride-bearing salts for periods up to 500 days. The two flat coupons were placed so that the solid oxide/salt mixture contacted about one half of the coupon surface. One teardrop coupon was placed in contact with solid mixture; the second teardrop was in contact with the headspace gas only. The mixtures were loaded with nominally 0.5 wt % water under a helium atmosphere. Observations of corrosion ranged from superficial staining to pitting and SCC. The extent of corrosion depended on the total salt concentration and on the composition of the salt. The most significant corrosion was found in coupons that were exposed to 98 wt % PuO{sub 2}, 2 wt % chloride salt mixtures that contained calcium chloride. SCC was observed in two 304L stainless steel teardrop coupons exposed in solid contact to a mixture of 98 wt % PuO{sub 2}, 0.9 wt % NaCl, 0.9 wt % KCl, and 0.2 wt % CaCl{sub 2}. The cracking was associated with the heat-affected zone of an autogenous weld that ran across the center of the coupon. Cracking was not observed in coupons exposed to the headspace gas, nor in coupons exposed to other mixtures with either 0.92 wt% CaCl{sub 2} or no CaCl{sub 2}. The corrosion results point to the significance of the interaction between water loading and the concentration of the hydrating salt CaCl{sub 2} in the susceptibility of austenitic stainless steels to corrosion.

  11. Use of Gadolinium as a Primary Criticality Control in Disposing Waste Containing Plutonium at SRS

    SciTech Connect

    Andrew, Vincent

    2005-01-04

    Use of gadolinium as a neutron poison has been proposed for disposing of several metric tons of excess plutonium at the Savannah River Site (SRS). The plutonium will first be dissolved in nitric acid in small batches. Gadolinium nitrate will then be added to the plutonium nitrate solution. The Gd-poisoned plutonium solution will be neutralized and transferred to large under-ground tanks. The pH of small batches of neutralized plutonium solution will be adjusted prior to addition of the glass frit for eventual production as glass logs. The use of gadolinium as the neutron poison would minimize the number of glass logs generated from this disposition. The primary criticality safety concerns regarding the disposal process are: (1) maintaining neutron moderation under all processing conditions since gadolinium has a very large absorption cross section for thermal neutrons, (2) characteristics of plutonium and gadolinium precipitation during the neutralization process, (3) mixing characteristics of the precipitate to ensure that plutonium would remain homogeneously mixed with gadolinium, and (4) potential separation of plutonium and gadolinium during nitric and formic acids addition. A number of experiments were conducted at the Savannah River National Laboratory to study the behavior of plutonium and gadolinium at various stages of the disposition process.

  12. Influence of sources on plutonium mobility and oxidation state transformations in vadose zone sediments.

    PubMed

    Kaplan, Daniel I; Powell, Brian A; Duff, Martine C; Demirkanli, Deniz I; Denham, Miles; Fjeld, Robert A; Molz, Fred J

    2007-11-01

    Well-defined solid sources of Pu(III) (PuCl3), Pu(IV) (Pu (NO3)4 and Pu (C2O4)2), and Pu(VI) (Pu02(NO3)2) were placed in lysimeters containing vadose zone sediments and exposed to natural weather conditions for 2 or 11 years. The objective of this study was to measure the release rate of Pu and the changes in the Pu oxidation states from these Pu sources with the intent to develop a reactive transport model source-term. Pu(III) and Pu(IV) sources had identical Pu concentration depth profiles and similar Pu release rates. Source release data indicate that PuIV(C2O4)2 was the least mobile, whereas Pu(VI)O2(NO3)2 was the most mobile. Synchrotron X-ray fluorescence (SXRF) revealed that Pu was very unevenly distributed on the sediment and Mn concentrations were too low (630 mg kg(-1)) and perhaps of the wrong mineralogy to influence Pu distribution. The high stability of sorbed Pu(IV) is proposed to be due to the formation of a stable hydrolyzed Pu(IV) surface species. Plutonium X-ray absorption near-edge spectroscopy (XANES) analysis conducted on sediment recovered at the end of the studyfrom the Pu(IV)(NO3)4- and Pu(III)(III)Cl3-amended lysimeters contained essentially identical Pu distributions: approximately 37% Pu(III), 67% Pu(IV), 0% Pu(V), and 0% Pu(VI). These results were similar to those using a wet chemistry Pu oxidation state assay, except the latter method did not detect any Pu(III) present on the sediment but instead indicated that 93-98% of the Pu existed as Pu(IV). This discrepancy was likely attributable to incomplete extraction of sediment Pu(III) by the wet chemistry method. Although Pu has been known to exist in the +3 oxidation state under microbially induced reducing conditions for decades, to our knowledge, this is the first observation of steady-state Pu(III) in association with natural sediments. On the basis of thermodynamic considerations, Pu(III) has a wide potential distribution, especially in acidic environments, and as such may warrant

  13. Influence of Sources on Plutonium Mobility and Oxidation State Transformations in Vadose Zone Sediments

    SciTech Connect

    Kaplan,D.; Powell, B.; Duff, M.; Demirkanli, D.; Denham, M.; Fjeld, R.; Molz, F.

    2007-01-01

    Well-defined solid sources of Pu(III) (PuCl3), Pu(IV) (Pu (NO3)4 and Pu (C2O4)2), and Pu(VI) (PuO2(NO3)2) were placed in lysimeters containing vadose zone sediments and exposed to natural weather conditions for 2 or 11 years. The objective of this study was to measure the release rate of Pu and the changes in the Pu oxidation states from these Pu sources with the intent to develop a reactive transport model source-term. Pu(III) and Pu(IV) sources had identical Pu concentration depth profiles and similar Pu release rates. Source release data indicate that PuIV(C2O4)2 was the least mobile, whereas PuVIO2(NO3)2 was the most mobile. Synchrotron X-ray fluorescence (SXRF) revealed that Pu was very unevenly distributed on the sediment and Mn concentrations were too low (630 mg kg-1) and perhaps of the wrong mineralogy to influence Pu distribution. The high stability of sorbed Pu(IV) is proposed to be due to the formation of a stable hydrolyzed Pu(IV) surface species. Plutonium X-ray absorption near-edge spectroscopy (XANES) analysis conducted on sediment recovered at the end of the study from the PuIV(NO3)4- and PuIIICl3-amended lysimeters contained essentially identical Pu distributions: approximately 37% Pu(III), 67% Pu(IV), 0% Pu(V), and 0% Pu(VI). These results were similar to those using a wet chemistry Pu oxidation state assay, except the latter method did not detect any Pu(III) present on the sediment but instead indicated that 93-98% of the Pu existed as Pu(IV). This discrepancy was likely attributable to incomplete extraction of sediment Pu(III) by the wet chemistry method. Although Pu has been known to exist in the +3 oxidation state under microbially induced reducing conditions for decades, to our knowledge, this is the first observation of steady-state Pu(III) in association with natural sediments. On the basis of thermodynamic considerations, Pu(III) has a wide potential distribution, especially in acidic environments, and as such may warrant further

  14. Spectroscopic investigations of neptunium`s and plutonium`s oxidation states in sol-gel glasses as a function of initial valance and thermal history

    SciTech Connect

    Stump, N.A.; Haire, R.G.; Dai, S.

    1996-12-01

    Several oxidation states of neptunium and plutonium, Pu(III),Pu (IV), PU(VI), Np(IV), Np(V) and Np (VI), were studied in glasses prepared by a sol-gel technology. The oxidation state of these actinides was determined primarily by absorption spectrometry and followed as a function of the solidification process, subsequent aging and thermal treatments. It was determined that the initial oxidation state of the actinides in the starting solutions was essentially maintained through the solidification process to form the glasses. However, during densification and removal of residual solvents at elevated temperatures, both actinides in the different sol-gel products converted completely to their tetravalent states. These results are discussed in terms of our findings in comparable studies that only the tetravalent states of plutonium and neptunium are formed in glasses prepared by dissolving their dioxides in different molten- glass formulations.

  15. CHARACTERIZATION OF EMISSIONS FROM PLUTONIUM-URANIUM OXIDE FUEL FABRICATION

    EPA Science Inventory

    To develop accurate monitoring techniques for the radioactive emissions from new types of nuclear facilities, it is necessary to characterize those emissions as completely as possible. The first facility selected was a mixed-oxide fuel fabrication plant. In-stack, standard hi-vol...

  16. Potential hydrogen and oxygen partial pressures in legacy plutonium oxide packages at Oak Ridge

    SciTech Connect

    Veirs, Douglas K.

    2014-07-07

    An approach to estimate the maximum hydrogen and oxygen partial pressures within sealed containers is described and applied to a set of packages containing high-purity plutonium dioxide. The approach uses experimentally determined maximum hydrogen and oxygen partial pressures and scales the experimentally determined pressures to the relevant packaged material properties. The important material properties are the specific wattage and specific surface area (SSA). Important results from the experimental determination of maximum partial pressures are (1) the ratio of hydrogen to oxygen is stoichiometric, and (2) the maximum pressures increase with increasing initial rates of production. The material properties that influence the rates are the material specific wattage and the SSA. The unusual properties of these materials, high specific wattage and high SSA, result in higher predicted maximum pressures than typical plutonium dioxide in storage. The pressures are well within the deflagration range for mixtures of hydrogen and oxygen.

  17. The underwater coincidence counter for plutonium measurements in mixed-oxide fuel assemblies manual

    SciTech Connect

    G. W. Eccleston; H. O. Menlove; M. Abhold; M. Baker; J. Pecos

    1999-05-01

    This manual describes the Underwater Coincidence Counter (UWCC) that has been designed for the measurement of plutonium in mixed-oxide (MOX) fuel assemblies prior to irradiation. The UWCC uses high-efficiency {sup 3}He neutron detectors to measure the spontaneous-fission and induced-fission rates in the fuel assembly. Measurements can be made on MOX fuel assemblies in air or underwater. The neutron counting rate is analyzed for singles, doubles, and triples time correlations to determine the {sup 240}Pu effective mass per unit length of the fuel assembly. The system can verify the plutonium loading per unit length to a precision of less than 1% in a measurement time of 2 to 3 minutes. System design, components, performance tests, and operational characteristics are described in this manual.

  18. Direct oxide reduction (DOR) solvent salt recycle in pyrochemical plutonium recovery operations

    SciTech Connect

    Fife, K.W.; Bowersox, D.F.; Davis, C.C.; McCormick, E.D.

    1987-02-01

    One method used at Los Alamos for producing plutonium metal is to reduce the oxide with calcium metal in molten CaCl/sub 2/ at 850/sup 0/C. The solvent CaCl/sub 2/ from this reduction step is currently discarded as low-level radioactive waste because it is saturated with the reaction by-product, CaO. We have developed and demonstrated a molten salt technique for rechlorinating the CaO, thereby regenerating the CaCl/sub 2/ and incorporating solvent recycle into the batch PuO/sub 2/ reduction process. We discuss results from the process development experiments and present our plans for incorporating the technique into an advanced design for semicontinuous plutonium metal production.

  19. Method for dissolving plutonium dioxide

    DOEpatents

    Tallent, Othar K.

    1978-01-01

    The fluoride-catalyzed, non-oxidative dissolution of plutonium dioxide in HNO.sub.3 is significantly enhanced in rate by oxidizing dissolved plutonium ions. It is believed that the oxidation of dissolved plutonium releases fluoride ions from a soluble plutonium-fluoride complex for further catalytic action.

  20. Health physics manual of good practices for plutonium facilities. [Contains glossary

    SciTech Connect

    Brackenbush, L.W.; Heid, K.R.; Herrington, W.N.; Kenoyer, J.L.; Munson, L.F.; Munson, L.H.; Selby, J.M.; Soldat, K.L.; Stoetzel, G.A.; Traub, R.J.

    1988-05-01

    This manual consists of six sections: Properties of Plutonium, Siting of Plutonium Facilities, Facility Design, Radiation Protection, Emergency Preparedness, and Decontamination and Decommissioning. While not the final authority, the manual is an assemblage of information, rules of thumb, regulations, and good practices to assist those who are intimately involved in plutonium operations. An in-depth understanding of the nuclear, physical, chemical, and biological properties of plutonium is important in establishing a viable radiation protection and control program at a plutonium facility. These properties of plutonium provide the basis and perspective necessary for appreciating the quality of control needed in handling and processing the material. Guidance in selecting the location of a new plutonium facility may not be directly useful to most readers. However, it provides a perspective for the development and implementation of the environmental surveillance program and the in-plant controls required to ensure that the facility is and remains a good neighbor. The criteria, guidance, and good practices for the design of a plutonium facility are also applicable to the operation and modification of existing facilities. The design activity provides many opportunities for implementation of features to promote more effective protection and control. The application of ''as low as reasonably achievable'' (ALARA) principles and optimization analyses are generally most cost-effective during the design phase. 335 refs., 8 figs., 20 tabs.

  1. Development program to recycle and purify plutonium-238 oxide fuel from scrap

    SciTech Connect

    Schulte, Louis D.; Silver, Gary L.; Avens, Larry R.; Jarvinen, Gordon D.; Espinoza, Jacob; Foltyn, Elizabeth M.; Rinehart, Gary H.

    1997-01-10

    Nuclear Materials Technology (NMT) Division of Los Alamos National Laboratory (LANL) has initiated a development program to recover and purify plutonium-238 oxide from impure sources. A glove box line has been designed and a process flowsheet developed to perform this task on a large scale. Our initial effort has focused on purification of {sup 238}PuO{sub 2} fuel that fails to meet General Purpose Heat Source (GPHS) specifications because of impurities. The most notable non-actinide impurity was silicon, but aluminum, chromium, iron and nickel were also near or in excess of limits specified by GPHS fuel powder specifications. {sup 234}U was by far the largest actinide impurity observed in the feed material because it is the daughter product of {sup 238}Pu by alpha decay. An aqueous method based on nitric acid was selected for purification of the {sup 238}PuO{sub 2} fuel. All aqueous processing used high purity reagents, and was performed in PTFE apparatus to minimize introduction of new contaminants. Impure {sup 238}PuO{sub 2} was finely milled, then dissolved in refluxing HNO{sub 3}/HF and the solution filtered. The dissolved {sup 238}Pu was adjusted to the trivalent state by an excess of reducing reagents to compensate for radiolytic effects, precipitated as plutonium(III) oxalate, and recovered by filtration. The plutonium(III) oxalate was subsequently calcined to convert the plutonium to the oxide. Decontamination factors for silicon, phosphorus and uranium were excellent. Decontamination factors for aluminum, chromium, iron and nickel were very good. The purity of the {sup 238}PuO{sub 2} recovered from this operation was significantly better than specifications. Efforts continue to develop the capability for efficient, safe, cost-effective, and environmentally acceptable methods to recover and purify {sup 238}PuO{sub 2} fuel in a glove box environment. Plutonium-238 materials targeted for recovery includes impure oxide and scrap items that are lean in {sup

  2. Plutonium-uranium mixed oxide characterization by coupling micro-X-ray diffraction and absorption investigations

    NASA Astrophysics Data System (ADS)

    Degueldre, C.; Martin, M.; Kuri, G.; Grolimund, D.; Borca, C.

    2011-09-01

    Plutonium-uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The potential differences of metal redox state and microstructural developments of the matrix before and after irradiation are commonly analysed by electron probe microanalysis. In this work the structure and next-neighbor atomic environments of Pu and U oxide features within unirradiated homogeneous MOX and irradiated (60 MW d kg -1) MOX samples was analysed by micro-X-ray fluorescence (μ-XRF), micro-X-ray diffraction (μ-XRD) and micro-X-ray absorption fine structure (μ-XAFS) spectroscopy. The grain properties, chemical bonding, valences and stoichiometry of Pu and U are determined from the experimental data gained for the unirradiated as well as for irradiated fuel material examined in the center of the fuel as well as in its peripheral zone (rim). The formation of sub-grains is observed as well as their development from the center to the rim (polygonization). In the irradiated sample Pu remains tetravalent (>95%) and no (<5%) Pu(V) or Pu(VI) can be detected while the fuel could undergo slight oxidation in the rim zone. Any slight potential plutonium oxidation is buffered by the uranium dioxide matrix while locally fuel cladding interaction could also affect the redox of the fuel.

  3. Analysis of pressurization of plutonium oxide storage vials during a postulated fire

    SciTech Connect

    Laurinat, J.; Kesterson, M.; Hensel, S.

    2015-02-10

    The documented safety analysis for the Savannah River Site evaluates the consequences of a postulated 1000 °C fire in a glovebox. The radiological dose consequences for a pressurized release of plutonium oxide powder during such a fire depend on the maximum pressure that is attained inside the oxide storage vial. To enable evaluation of the dose consequences, pressure transients and venting flow rates have been calculated for exposure of the storage vial to the fire. A standard B vial with a capacity of approximately 8 cc was selected for analysis. The analysis compares the pressurization rate from heating and evaporation of moisture adsorbed onto the plutonium oxide contents of the vial with the pressure loss due to venting of gas through the threaded connection between the vial cap and body. Tabulated results from the analysis include maximum pressures, maximum venting velocities, and cumulative vial volumes vented during the first 10 minutes of the fire transient. Results are obtained for various amounts of oxide in the vial, various amounts of adsorbed moisture, different vial orientations, and different surface fire exposures.

  4. Loading Capacities for Uranium, Plutonium and Neptunium in High Caustic Nuclear Waste Storage Tanks Containing Selected Sorbents

    SciTech Connect

    OJI, LAWRENCE

    2004-11-16

    In this study the loading capacities of selected actinides onto some of the most common sorbent materials which are present in caustic nuclear waste storage tanks have been determined. Some of these transition metal oxides and activated carbons easily absorb or precipitate plutonium, neptunium and even uranium, which if care is not taken may lead to unwanted accumulation of some of these fissile materials in nuclear waste tanks during waste processing. Based on a caustic synthetic salt solution simulant bearing plutonium, uranium and neptunium and ''real'' nuclear waste supernate solution, the loading capacities of these actinides onto iron oxide (hematite), activated carbon and anhydrous sodium phosphate have been determined. The loading capacities for plutonium onto granular activated carbon and iron oxide (hematite) in a caustic synthetic salt solution were, respectively, 3.4 0.22 plus or minus and 5.5 plus or minus 0.38 microgram per gram of sorbent. The loading capacity for plutonium onto a typical nuclear waste storage tank sludge solids was 2.01 microgram per gram of sludge solids. The loading capacities for neptunium onto granular activated carbon and iron oxide (hematite) in a caustic synthetic salt solution were, respectively, 7.9 plus or minus 0.52 and greater than 10 microgram per gram of sorbent. The loading capacity for neptunium onto a typical nuclear waste storage tank sludge solids was 4.48 microgram per gram of sludge solids. A typical nuclear waste storage tank solid material did not show any significant affinity for uranium. Sodium phosphate showed significant affinity for both neptunium and uranium, with loading capacities of 6.8 and 184.6 plus or minus 18.5 microgram per gram of sorbent, respectively.

  5. Containment system for supercritical water oxidation reactor

    SciTech Connect

    Chastagner, P.

    1991-12-31

    This invention is comprised of a system for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary.

  6. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, Philippe

    1994-01-01

    A system for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary.

  7. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, P.

    1994-07-05

    A system is described for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary. 2 figures.

  8. Radiolysis of hexavalent plutonium in solutions of uranyl nitrate containing fission product simulants

    NASA Astrophysics Data System (ADS)

    Rance, Peter J. W.; Zilberman, B. Ya.; Akopov, G. A.

    2000-07-01

    The effect of the inherent radioactivity on the chemical state of plutonium ions in solution was recognized very shortly after the first macroscopic amounts of plutonium became available and early studies were conducted as part of the Manhattan Project. However, the behavior of plutonium ions, in nitric acid especially, has been found to be somewhat complex, so much so that a relatively modern summary paper included the comment that, "The vast amount of work carried out in nitric acid solutions can not be adequately summarized. Suffice it to say results in these solutions are plagued with irreproducibility and induction periods…" Needless to say, the presence of other ions in solution, as occurs when irradiated nuclear fuel is dissolved, further complicates matters. The purpose of the work described below was to add to the rather small amount of qualitative data available relating to the radiolytic behavior of plutonium in solutions of irradiated nuclear fuel.

  9. The In-Vitro Transport of (238)PLUTONIUM Oxide and (239)PLUTONIUM Oxide Through a Membrane Filter and its Importance for Internal Radiation Dosimetry.

    NASA Astrophysics Data System (ADS)

    Ryan, Michael Terrance

    These experiments were designed to determine if ('238)PuO(,2), due to its higher specific activity and attendant aggregate recoil, undergoes higher transfer through a membrane filter into an interstitial human alveolar lung fluid simulant than ('239)PuO(,2). The rate at which such transfer occurs was determined in an in-vitro chamber designed to simulate residence characteristics of particles of insoluble plutonium oxides in human alveolar interstitium. The ratio of the rate of ('238)Pu/('239)Pu transfer was 138 (+OR -) 76%. Calculations were performed to assess the importance of this finding in terms of the internal dosimetry of insoluble ('238)Pu using methods and models recommended by the International Commission on Radiological Protection. Three cases were evaluated, namely integral 50-year dose commitment, urinary excretion after single acute intake and urinary excretion rate during chronic constant intake. It was found that integral 50-year dose commitments were not influenced by the rate of plutonium transfer from the pulmonary compartment to blood. The evaluation of calculated urinary excretion data after a single acute inhalation intake showed that in the early period, up to about 30 days post exposure, urinary excretion of ('238)PuO(,2) may be 2 to 10 times higher than the urinary excretion rate for ('238)PuO(,2) predicted by the ICRP reference model. From about 50 days to approximately 1000 days the calculated urinary excretion rate for ('238)PuO(,2) may be lower than that predicted by the reference model by a factor of 2 to 10. In the case of chronic constant intake the calculated urinary excretion rate for ('238)PuO(,2) may be up to a factor of 2 higher than that predicted by the reference ICRP Model.

  10. Plutonium incorporation in phosphate and titanate ceramics for minor actinide containment

    NASA Astrophysics Data System (ADS)

    Deschanels, X.; Picot, V.; Glorieux, B.; Jorion, F.; Peuget, S.; Roudil, D.; Jégou, C.; Broudic, V.; Cachia, J. N.; Advocat, T.; Den Auwer, C.; Fillet, C.; Coutures, J. P.; Hennig, C.; Scheinost, A.

    2006-06-01

    Two ceramics, zirconolite and a monazite-brabantite solid solution (MBss) were studied for the immobilization of minor actinides (Np, Am, Cm) produced by reprocessing spent fuel. Monoclinic zirconolite (CaZrTi2O7) is a fluorite derivative structure and is the primary actinide host phase in Synroc (a titanate composite). Monazite (LnPO4, where Ln = La, Ce, Nd, Gd, etc.) is a monoclinic orthophosphate containing trivalent cations, and brabantite (Ca0.5An0.5PO4) is an isostructural monazite compound containing tetravalent cations (An = Th and U). The nominal composition of the ceramics studied in this work is (Ca0.87Pu0.13)Zr(Al0. 26Ti1.74)O7 for zirconolite and (Ca0.09Pu0.09La0.73Th0.09)PO4 for the monazite-brabantite solid solution. These formulas correspond to 10 wt% PuO2 loading in each material. XANES spectroscopy showed that the plutonium is tetravalent in zirconolite and trivalent in MBss. Thorium, another tetravalent cation, can be incorporated at 10 wt% ThO2 in MBss. Aluminum and calcium balance the excess cationic charge resulting from the incorporation of Pu(IV) in zirconolite and Th(IV) in brabantite, respectively. The relative density of the pellets exceeded 90% of theoretical density. The samples exhibited a homogeneous microstructure even if some minor phases, representing less than 2% of the surface area, were detected. The two ceramics are compared in terms of actinide loading, and preliminary results on their long-term behavior are discussed.

  11. Disposition of Uranium -233 (sup 233U) in Plutonium Metal and Oxide at the Rocky Flats Environmental Technology Site

    SciTech Connect

    Freiboth, Cameron J.; Gibbs, Frank E.

    2000-03-01

    This report documents the position that the concentration of Uranium-233 ({sup 233}U) in plutonium metal and oxide currently stored at the DOE Rocky Flats Environmental Technology Site (RFETS) is well below the maximum permissible stabilization, packaging, shipping and storage limits. The {sup 233}U stabilization, packaging and storage limit is 0.5 weight percent (wt%), which is also the shipping limit maximum. These two plutonium products (metal and oxide) are scheduled for processing through the Building 371 Plutonium Stabilization and Packaging System (PuSPS). This justification is supported by written technical reports, personnel interviews, and nuclear material inventories, as compiled in the ''History of Uranium-233 ({sup 233}U) Processing at the Rocky Flats Plant In Support of the RFETS Acceptable Knowledge Program'' RS-090-056, April 1, 1999. Relevant data from this report is summarized for application to the PuSPS metal and oxide processing campaigns.

  12. Reduction of Worldwide Plutonium Inventories Using Conventional Reactors and Advanced Fuels: A Systems Study

    SciTech Connect

    Krakowski, R.A., Bathke, C.G.

    1997-12-31

    The potential for reducing plutonium inventories in the civilian nuclear fuel cycle through recycle in LWRs of a variety of mixed oxide forms is examined by means of a cost based plutonium flow systems model. This model emphasizes: (1) the minimization of separated plutonium; (2) the long term reduction of spent fuel plutonium; (3) the optimum utilization of uranium resources; and (4) the reduction of (relative) proliferation risks. This parametric systems study utilizes a globally aggregated, long term (approx. 100 years) nuclear energy model that interprets scenario consequences in terms of material inventories, energy costs, and relative proliferation risks associated with the civilian fuel cycle. The impact of introducing nonfertile fuels (NFF,e.g., plutonium oxide in an oxide matrix that contains no uranium) into conventional (LWR) reactors to reduce net plutonium generation, to increase plutonium burnup, and to reduce exo- reactor plutonium inventories also is examined.

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

    SciTech Connect

    Judge, Elizabeth J.; Berg, John M.; Le, Loan A.; Lopez, Leon N.; Barefield, James E.

    2012-06-18

    Laser-induced breakdown spectroscopy (LIBS) was used to analyze a mixed actinide fuel pellet containing 75% UO{sub 2}/20% PuO{sub 2}/3% AmO{sub 2}/2% NpO{sub 2}. The preliminary data shown here is the first report of LIBS analysis of a mixed actinide fuel pellet, to the authors knowledge. The LIBS spectral data was acquired in a plutonium facility at Los Alamos National Laboratory where the sample was contained within a glove box. The initial installation of the glove box was not intended for complete ultraviolet (UV), visible (VIS) and near infrared (NIR) transmission, therefore the LIBS spectrum is truncated in the UV and NIR regions due to the optical transmission of the window port and filters that were installed. The optical collection of the emission from the LIBS plasma will be optimized in the future. However, the preliminary LIBS data acquired is worth reporting due to the uniqueness of the sample and spectral data. The analysis of several actinides in the presence of each other is an important feature of this analysis since traditional methods must chemically separate uranium, plutonium, neptunium, and americium prior to analysis. Due to the historic nature of the sample fuel pellet analyzed, the provided sample composition of 75% UO{sub 2}/20% PuO{sub 2}/3% AmO{sub 2}/2% NpO{sub 2} cannot be confirm without further analytical processing. Uranium, plutonium, and americium emission lines were abundant and easily assigned while neptunium was more difficult to identify. There may be several reasons for this observation, other than knowing the exact sample composition of the fuel pellet. First, the atomic emission wavelength resources for neptunium are limited and such techniques as hollow cathode discharge lamp have different dynamics than the plasma used in LIBS which results in different emission spectra. Secondly, due to the complex sample of four actinide elements, which all have very dense electronic energy levels, there may be reactions and

  14. Aqueous biphasic plutonium oxide extraction process with pH and particle control

    DOEpatents

    Chaiko, D.J.; Mensah-Biney, R.

    1997-04-29

    A method is described for simultaneously partitioning a metal oxide and silica from a material containing silica and the metal oxide, using a biphasic aqueous medium having immiscible salt and polymer phases. 2 figs.

  15. Dissolution of Neptunium and Plutonium Oxides Using a Catalyzed Electrolytic Process

    SciTech Connect

    Hylton, TD

    2004-10-25

    This report discusses the scoping study performed to evaluate the use of a catalyzed electrolytic process for dissolving {sup 237}Np oxide targets that had been irradiated to produce {sup 238}Pu oxide. Historically, these compounds have been difficult to dissolve, and complete dissolution was obtained only by adding hydrofluoric acid to the nitric acid solvent. The presence of fluoride in the mixture is undesired because the fluoride ions are corrosive to tank and piping systems and the fluoride ions cause interferences in the spectrophotometric analyses. The goal is to find a dissolution method that will eliminate these issues and that can be incorporated into a processing system to support the domestic production and purification of {sup 238}Pu. This study evaluated the potential of cerium(IV) ions, a strong oxidant, to attack and dissolve the oxide compounds. In the dissolution process, the cerium(IV) ions are reduced to cerium(III) ions, which are not oxidants. Therefore, an electrolytic process was incorporated to continuously convert cerium(III) ions back to cerium(IV) ions so that they can dissolve more of the oxide compounds. This study showed that the neptunium and plutonium oxides were successfully dissolved and that more development work should be performed to optimize the procedure.

  16. LAB-SCALE DEMONSTRATION OF PLUTONIUM PURIFICATION BY ANION EXCHANGE, PLUTONIUM (IV) OXALATE PRECIPITATION, AND CALCINATION TO PLUTONIUM OXIDE TO SUPPORT THE MOX FEED MISSION

    SciTech Connect

    Crowder, M.; Pierce, R.

    2012-08-22

    H-Canyon and HB-Line are tasked with the production of PuO{sub 2} from a feed of plutonium metal. The PuO{sub 2} will provide feed material for the MOX Fuel Fabrication Facility. After dissolution of the Pu metal in H-Canyon, the solution will be transferred to HB-Line for purification by anion exchange. Subsequent unit operations include Pu(IV) oxalate precipitation, filtration and calcination to form PuO{sub 2}. This report details the results from SRNL anion exchange, precipitation, filtration, calcination, and characterization tests, as requested by HB-Line1 and described in the task plan. This study involved an 80-g batch of Pu and employed test conditions prototypical of HB-Line conditions, wherever feasible. In addition, this study integrated lessons learned from earlier anion exchange and precipitation and calcination studies. H-Area Engineering selected direct strike Pu(IV) oxalate precipitation to produce a more dense PuO{sub 2} product than expected from Pu(III) oxalate precipitation. One benefit of the Pu(IV) approach is that it eliminates the need for reduction by ascorbic acid. The proposed HB-Line precipitation process involves a digestion time of 5 minutes after the time (44 min) required for oxalic acid addition. These were the conditions during HB-line production of neptunium oxide (NpO{sub 2}). In addition, a series of small Pu(IV) oxalate precipitation tests with different digestion times were conducted to better understand the effect of digestion time on particle size, filtration efficiency and other factors. To test the recommended process conditions, researchers performed two nearly-identical larger-scale precipitation and calcination tests. The calcined batches of PuO{sub 2} were characterized for density, specific surface area (SSA), particle size, moisture content, and impurities. Because the 3013 Standard requires that the calcination (or stabilization) process eliminate organics, characterization of PuO{sub 2} batches monitored the

  17. Solvent extraction system for plutonium colloids and other oxide nano-particles

    DOEpatents

    Soderholm, Lynda; Wilson, Richard E; Chiarizia, Renato; Skanthakumar, Suntharalingam

    2014-06-03

    The invention provides a method for extracting plutonium from spent nuclear fuel, the method comprising supplying plutonium in a first aqueous phase; contacting the plutonium aqueous phase with a mixture of a dielectric and a moiety having a first acidity so as to allow the plutonium to substantially extract into the mixture; and contacting the extracted plutonium with second a aqueous phase, wherein the second aqueous phase has a second acidity higher than the first acidity, so as to allow the extracted plutonium to extract into the second aqueous phase. The invented method facilitates isolation of plutonium polymer without the formation of crud or unwanted emulsions.

  18. Self-Lubricating Composite Containing Chromium Oxide

    NASA Technical Reports Server (NTRS)

    Dellacorte, Christopher (Inventor); Edmonds, Brian J. (Inventor)

    1999-01-01

    A self lubricating. friction and wear reducing composite material useful over a wide temperature range of from cryogenic temperature up to about 900 C. contains 60 80 wt. % of particulate Cr2O3, dispersed in a metal binder of a metal alloy containing Cr and at least 50 wt. % of Ni, Cr or a mature of Ni and Cr. It also contains 5-20 wt. % of a fluoride of at least one Group I, Group II or rare earth metal and. optionally, 5-20 wt. % of a low temperature lubricant metal, such as Ag. Au, Pt, Pd, Rh and Cu. This composite exhibits less oxidation instability and less abrasiveness than composites containing chromium carbide, is readily applied using plasma spray and can be ground and polished with a silicon carbide abrasive.

  19. New Fecal Method for Plutonium and Americium

    SciTech Connect

    Maxwell, S.L. III

    2000-06-27

    A new fecal analysis method that dissolves plutonium oxide was developed at the Westinghouse Savannah River Site. Diphonix Resin (Eichrom Industries), is used to pre-concentrate the actinides from digested fecal samples. A rapid microwave digestion technique is used to remove the actinides from the Diphonix Resin, which effectively extracts plutonium and americium from acidic solutions containing hydrofluoric acid. After resin digestion, the plutonium and americium are recovered in a small volume of nitric acid that is loaded onto small extraction chromatography columns, TEVA Resin and TRU Resin (Eichrom Industries). The method enables complete dissolution of plutonium oxide and provides high recovery of plutonium and americium with good removal of thorium isotopes such as thorium-228.

  20. PROCESS FOR THE RECOVERY OF PLUTONIUM

    DOEpatents

    Ritter, D.M.

    1959-01-13

    An improvement is presented in the process for recovery and decontamination of plutonium. The carrier precipitate containing plutonium is dissolved and treated with an oxidizing agent to place the plutonium in a hexavalent oxidation state. A lanthanum fluoride precipitate is then formed in and removed from the solution to carry undesired fission products. The fluoride ions in the reniaining solution are complexed by addition of a borate sueh as boric acid, sodium metaborate or the like. The plutonium is then reduced and carried from the solution by the formation of a bismuth phosphate precipitate. This process effects a better separation from unwanted flssion products along with conccntration of the plutonium by using a smaller amount of carrier.

  1. Isotope exchange in oxide-containing catalyst

    NASA Technical Reports Server (NTRS)

    Brown, Kenneth G. (Inventor); Upchurch, Billy T. (Inventor); Hess, Robert V. (Inventor); Miller, Irvin M. (Inventor); Schryer, David R. (Inventor); Sidney, Barry D. (Inventor); Wood, George M. (Inventor); Hoyt, Ronald F. (Inventor)

    1989-01-01

    A method of exchanging rare-isotope oxygen for common-isotope oxygen in the top several layers of an oxide-containing catalyst is disclosed. A sample of an oxide-containing catalyst is exposed to a flowing stream of reducing gas in an inert carrier gas at a temperature suitable for the removal of the reactive common-isotope oxygen atoms from the surface layer or layers of the catalyst without damaging the catalyst structure. The reduction temperature must be higher than any at which the catalyst will subsequently operate. Sufficient reducing gas is used to allow removal of all the reactive common-isotope oxygen atoms in the top several layers of the catalyst. The catalyst is then reoxidized with the desired rare-isotope oxygen in sufficient quantity to replace all of the common-isotope oxygen that was removed.

  2. Structure Determination of Plutonium Oxide Precipitates Formed from Aqueous Plutonium IV and V Solutions and in the Presence of Goethite

    NASA Astrophysics Data System (ADS)

    Dai, Z.; Zavarin, M.; Zhao, P.; Begg, J.; Kersting, A. B.

    2012-12-01

    A series of aqueous Pu(IV) and Pu(V) batch sorption experiments with goethite (α-FeOOH) in a pH 8 ± 0.5 buffer solution (5mM NaCl + 0.7 mM NaHCO3) at room temperature (25 °C) were performed. Intrinsic Pu colloids were synthesized in alkaline solution (pH 8, 25 °C) and acidic solution (0.1 M HNO3, ~80 °C for 10-20 min), respectively, for comparison. Morphology, distribution and crystal structure of Pu oxide precipitates, as well as interaction between the Pu precipitates and goethite, were investigated using transmission electron microscopy (TEM). The Pu oxide precipitates formed from the sorption experiments consist of 3-5 nm primary crystalline particles (nanocrystals) irrespective of the initial form of Pu. The Pu oxide nanocrystals adopt two different crystal structures, either fcc PuO2 or bcc Pu4O7. The relative abundance of one form over the other depends on the initial form of Pu, Pu concentration, and the presence of goethite. For the high Pu concentration sorption cases (>9,000 nmol/m2 goethite), fcc PuO2 is the predominant phase occurring in both aqueous Pu(IV) and Pu(V) samples. In the Pu(IV) samples, the fcc PuO2 nanocrystals form mainly as a product of hydrolysis in solution. In the Pu(V) samples, the fcc PuO2 nanocrystals form by redox reactions dominantly occurring on goethite surface following the sorption of Pu(V). At lower Pu concentrations, the bcc Pu4O7 becomes dominant in the presence of goethite. The bcc Pu4O7 forms directly on the goethite surface as a 3-5 nm isolated nanocrystal in both Pu(IV) and Pu(V) samples and has specific crystallographic orientation relationships to goethite. Nucleation of the bcc Pu4O7 may occur by substitution of Pu(III) at the Fe(III) position on the goethite surface. In the absence of goethite, the intrinsic Pu colloids formed in alkaline solution (pH 8, 25 °C) are also comprised of 3-5 nm fcc PuO2 nanocrystals. As for the intrinsic Pu colloids precipitated from the acidic solution (0.1 M HNO3) at an elevated

  3. Neutralization of Plutonium and Enriched Uranium Solutions Containing Gadolinium as a Neutron Poison

    SciTech Connect

    BRONIKOWSKI, MG.

    2004-04-01

    Materials currently being dissolved in the HB-Line Facility will result in an accumulated solution containing an estimated uranium:plutonium (U:Pu) ratio of 4.3:1 and an 235U enrichment estimated at 30 per cent The U:Pu ratio and the enrichment are outside the evaluated concentration range for disposition to high level waste (HLW) using gadolinium (Gd) as a neutron poison. To confirm that the solution generated during the current HB-Line dissolving campaign can be poisoned with Gd, neutralized and discarded to the Savannah River Site (SRS) high level waste (HLW) system without undue nuclear safety concerns the caustic precipitation of surrogate solutions was examined. Experiments were performed with a U/Pu/Gd solution representative of the HB-Line estimated concentration ratio and also a U/Gd solution. Depleted U was used in the experiments as the enrichment of the U will not affect the chemical behavior during neutralization, but will affect the amount of Gd added to the solution. Settling behavior of the neutralized solutions was found to be comparable to previous studies. The neutralized solutions mixed easily and had expected densities of typical neutralized waste. The neutralized solids were found to be homogeneous and less than 20 microns in size. Partially neutralized solids were more amorphous than the fully neutralized solids. Based on the results of these experiments, Gd was found to be a viable poison for neutralizing a U/Pu/Gd solution with a U:Pu mass ratio of 4.3:1 thus extending the U:Pu mass ratio from the previously investigated 0-3:1 to 4.3:1. However, further work is needed to allow higher U concentrations or U:Pu ratios greater than investigated in this work.

  4. Plutonium isotopic analysis system for plutonium samples enriched in sup 238 Pu in EP 60/61 containers

    SciTech Connect

    Ruhter, W.D.; DeWitt, K.W.; Pederson, K.B.; Watkins, J.A.

    1990-06-01

    This hardware manual is addressed to Savannah River Site supervisors and to authorized mechanical and electrical maintenance and service personnel. Each chapter begins with a table of contents that lists the section titles, illustrations, and tabular data presented in that chapter. The first chapter in this manual is an introduction to the Pu-238 isotopic analysis system. Chapter 2 lists major components of the system and includes electronic block and wiring diagrams. Commercial nuclear instrument modules and their required settings are described in Chapter 3. Chapter 4 contains service and preventive maintenance procedures, as well as a troubleshooting log. In Chapter 5 is a list of spare parts and all relevant mechanical drawings for the system. Chapter 6 contains reference articles. All commercial items mentioned in this manual are assumed to be functioning correctly for the purposes of system operation. Users are referred to individual equipment manufacturers' manuals for details of operation, troubleshooting, and maintenance of this commercial equipment. 31 figs., 1 tab.

  5. CSER 00-006 Storage of Plutonium Residue Containers in 55 Gallon Drums at the PFP

    SciTech Connect

    DOBBIN, K.D.

    2000-05-24

    This criticality safety evaluation report (CSER) provides the required limit set and controls for safe transit and storage of these drums in the 234-5Z Building at the PFP. A mass limit of 200 g of plutonium or fissile equivalent per drum is acceptable

  6. An isotopic analysis system for plutonium samples enriched in sup 238 Pu

    SciTech Connect

    Ruhter, W.D.; Camp, D.C.

    1991-08-01

    We have designed and built a gamma-ray spectrometer system that measures the relative plutonium isotopic abundances of plutonium oxide enriched in {sup 238}Pu. The first system installed at Westinghouse Savannah River Company was tested and evaluated on plutonium oxide in stainless steel EP60/61 containers. {sup 238}Pu enrichments ranged from 20% to 85%. Results show that 200 grams of plutonium oxide in an EP60.61 container can be measured with {plus minus}0.3% precision and better than {plus minus}1.0% accuracy in the specific power using a counting time of 50 minutes. 3 refs., 2 figs.

  7. A summary of volatile impurity measurements and gas generation studies on MISSTD-1, a high-purity plutonium oxide produced by low-temperature calcination of plutonium oxalate

    SciTech Connect

    Berg, John M.; Narlesky, Joshua E.; Veirs, Douglas K.

    2012-06-08

    Plutonium dioxide of high specific surface area was subjected to long-term tests of gas generation in sealed containers. The material preparation and the storage conditions were outside the bounds of acceptable parameters defined by DOE-STD-3013-2012 in that the material was stabilized to a lower temperature than required and had higher moisture content than allowed. The data provide useful information for better defining the bounding conditions for safe storage. Net increases in internal pressure and transient increases in H{sub 2} and O{sub 2} were observed, but were well within the bounds of gas compositions previously shown to not threaten integrity of 3013 containers.

  8. Sampling and Analysis of the Headspace Gas in 3013 Type Plutonium Storage Containers at Los Alamos National Laboratory

    SciTech Connect

    Jackson, Jay M.; Berg, John M.; Hill, Dallas D.; Worl, Laura A.; Veirs, Douglas K.

    2012-07-11

    Department of Energy (DOE) sites have packaged approximately 5200 3013 containers to date. One of the requirements specified in DOESTD-3013, which specifies requirements for packaging plutonium bearing materials, is that the material be no greater than 0.5 weight percent moisture. The containers are robust, nested, welded vessels. A shelf life surveillance program was established to monitor these cans over their 50 year design life. In the event pressurization is detected by radiography, it will be necessary to obtain a head space gas sample from the pressurized container. This technique is also useful to study the head space gas in cans selected for random destructive evaluation. The atmosphere is sampled and the hydrogen to oxygen ratio is measured to determine the effects of radiolysis on the moisture in the container. A system capable of penetrating all layers of a 3013 container assembly and obtaining a viable sample of the enclosed gas and an estimate of internal pressure was designed.

  9. PROCESS FOR PURIFYING PLUTONIUM

    DOEpatents

    Mastick, D.F.; Wigner, E.P.

    1958-05-01

    A method is described of separating plutonium from small amounts of uranium and other contaminants. An acidic aqueous solution of higher valent plutonium and hexavalent uranium is treated with a soluble iodide to obtain the plutonium in the plus three oxidation state while leaving the uranium in the hexavalent state, adding a soluble oxalate such as oxalic acid, and then separating the insoluble plus the plutonium trioxalate from the solution.

  10. Use of a Shielded High Resolution Gamma Spectrometry System to Segregate LLW from Contact Handleable ILW Containing Plutonium - 13046

    SciTech Connect

    Lester, Rosemary; Wilkins, Colin; Chard, Patrick; Jaederstroem, Henrik; LeBlanc, Paul; Mowry, Rick; MacDonald, Sanders; Gunn, William

    2013-07-01

    Dounreay Site Restoration Limited (DSRL) have a number of drums of solid waste that may contain Plutonium Contaminated Material. These are currently categorised as Contact Handleable Intermediate Level Waste (CHILW). A significant fraction of these drums potentially contain waste that is in the Low Level Waste (LLW) category. A Canberra Q2 shielded high resolution gamma spectrometry system is being used to quantify the total activity of drums that are potentially in the LLW category in order to segregate those that do contain LLW from CHILW drums and thus to minimise the total volume of waste in the higher category. Am-241 is being used as an indicator of the presence of plutonium in the waste from its strong 59.54 keV gamma-ray; a knowledge of the different waste streams from which the material originates allows a pessimistic waste 'fingerprint' to be used in order to determine an upper limit to the activities of the weak and non-gamma-emitting plutonium and associated radionuclides. This paper describes the main features of the high resolution gamma spectrometry system being used by DSRL to perform the segregation of CHILW and LLW and how it was configured and calibrated using the Canberra In-Situ Object Counting System (ISOCS). It also describes how potential LLW drums are selected for assay and how the system uses the existing waste stream fingerprint information to determine a reliable upper limit for the total activity present in each measured drum. Results from the initial on-site commissioning trials and the first measurements of waste drums using the new monitor are presented. (authors)

  11. Phosphate monazite- and NaZr2(PO4)3 (NZP)-like ceramics containing uranium and plutonium

    NASA Astrophysics Data System (ADS)

    Orlova, A. I.; Kitaev, D. B.; Lukinich, A. N.; Tomilin, S. B.; Lizin, A. A.; Kulikov, I. A.; Samsonov, V. E.

    2003-01-01

    The synthesis and results of the X-ray diffraction examination of complex orthophosphates involving the cations Ca2+, Cd2+, Gd3+, Ti4+, Hf4+, Ce4+, U4+, and Pu4+ are described. Crystalline products containing either cerium or plutonium or these two elements together were prepared. Cerium was used as plutonium simulator. Phosphates containing cations Ti4+ and Hf4+ were found to form two-phase products. One of these phases was assumed to be of the NaZr2(PO4)3 (NZP) type, the other one of the CePO4 type (monazite). Formation of single-phase products with a monazite structure was found in the specimens that did not contain Ti and Hf. X-ray diffraction characteristics of synthesised phases were identified. An effect of the cation composition on the phase formation and also the effect of the cation size on the crystal lattice parameters of the formed phases were established.

  12. PLUTONIUM CLEANING PROCESS

    DOEpatents

    Kolodney, M.

    1959-12-01

    A method is described for rapidly removing iron, nickel, and zinc coatings from plutonium objects while simultaneously rendering the plutonium object passive. The method consists of immersing the coated plutonium object in an aqueous acid solution containing a substantial concentration of nitrate ions, such as fuming nitric acid.

  13. PREPARATION OF PLUTONIUM HALIDES

    DOEpatents

    Davidson, N.R.; Katz, J.J.

    1958-11-01

    A process ls presented for the preparation of plutonium trihalides. Plutonium oxide or a compound which may be readily converted to plutonlum oxide, for example, a plutonium hydroxide or plutonlum oxalate is contacted with a suitable halogenating agent. Speciflc agents mentioned are carbon tetrachloride, carbon tetrabromide, sulfur dioxide, and phosphorus pentachloride. The reaction is carried out under superatmospberic pressure at about 300 icient laborato C.

  14. Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./ Russian Progress Report for Fiscal Year 1997, Volume 4, Part 8 - Neutron Poison Plates in Assemblies Containing Homogeneous Mixtures of Polystyrene-Moderated Plutonium and Uranium Oxides

    SciTech Connect

    Yavuz, M.

    1999-05-01

    In the 1970s at the Battelle Pacific Northwest Laboratory (PNL), a series of critical experiments using a remotely operated Split-Table Machine was performed with homogeneous mixtures of (Pu-U)O{sub 2}-polystyrene fuels in the form of square compacts having different heights. The experiments determined the critical geometric configurations of MOX fuel assemblies with and without neutron poison plates. With respect to PuO{sub 2} content and moderation [H/(Pu+U)atomic] ratio (MR), two different homogeneous (Pu-U) O{sub 2}-polystyrene mixtures were considered: Mixture (1) 14.62 wt% PuO{sub 2} with 30.6 MR, and Mixture (2) 30.3 wt% PuO{sub 2} with 2.8 MR. In all mixtures, the uranium was depleted to about O.151 wt% U{sup 235}. Assemblies contained copper, copper-cadmium or aluminum neutron poison plates having thicknesses up to {approximately}2.5 cm. This evaluation contains 22 experiments for Mixture 1, and 10 for Mixture 2 compacts. For Mixture 1, there are 10 configurations with copper plates, 6 with aluminum, and 5 with copper-cadmium. One experiment contained no poison plate. For Mixture 2 compacts, there are 3 configurations with copper, 3 with aluminum, and 3 with copper-cadmium poison plates. One experiment contained no poison plate.

  15. USING POLYMERIC HYDROGEN GETTERS TO PREVENT COMBUSTIBLE ATMOSPHERES DURING INTERIM SAFE STORAGE OF PLUTONIUM OXIDE

    SciTech Connect

    Woodsmall, T

    2007-05-24

    Nuclear Materials Management (NMM) of WSRC has recently installed the capability to perform both non-destructive and destructive examination of 3013 containers of Pu oxide in accordance with DOE-STD-3013. The containers will be opened and the oxide will be sampled for analysis. The remaining bulk oxide must then be safely stored in a non-3013-compliant configuration. Available processing equipment and controls cannot prevent the oxide from adsorbing moisture during this process. Subsequent radiolysis of moisture during storage may generate combustible quantities of gases while waiting final processing, and satisfying DOE Interim Safe Storage Criteria (ISSC) would require that storage containers be vented at impractical frequencies. With support from an independent National Laboratory, WSRC/NMM has demonstrated that a commercial hydrogen getter material will effectively prevent the accumulation of combustible gas concentrations. A project overview, including storage requirements and strategies, as well as getter technology, current test results, and anticipated future developments will be addressed.

  16. Investigation Of In-Line Monitoring Options At H Canyon/HB Line For Plutonium Oxide Production

    SciTech Connect

    Sexton, L.

    2015-10-14

    H Canyon and HB Line have a production goal of 1 MT per year of plutonium oxide feedstock for the MOX facility by FY17 (AFS-2 mission). In order to meet this goal, steps will need to be taken to improve processing efficiency. One concept for achieving this goal is to implement in-line process monitoring at key measurement points within the facilities. In-line monitoring during operations has the potential to increase throughput and efficiency while reducing costs associated with laboratory sample analysis. In the work reported here, we mapped the plutonium oxide process, identified key measurement points, investigated alternate technologies that could be used for in-line analysis, and initiated a throughput benefit analysis.

  17. METHOD OF SEPARATING PLUTONIUM

    DOEpatents

    Brown, H.S.; Hill, O.F.

    1958-02-01

    Plutonium hexafluoride is a satisfactory fluorinating agent and may be reacted with various materials capable of forming fluorides, such as copper, iron, zinc, etc., with consequent formation of the metal fluoride and reduction of the plutonium to the form of a lower fluoride. In accordance with the present invention, it has been found that the reactivity of plutonium hexafluoride with other fluoridizable materials is so great that the process may be used as a method of separating plutonium from mixures containing plutonium hexafluoride and other vaporized fluorides even though the plutonium is present in but minute quantities. This process may be carried out by treating a mixture of fluoride vapors comprising plutonium hexafluoride and fluoride of uranium to selectively reduce the plutonium hexafluoride and convert it to a less volatile fluoride, and then recovering said less volatile fluoride from the vapor by condensation.

  18. Cleaning up the Legacy of the Cold War: Plutonium Oxides and the Role of Synchrotron Radiation Research

    SciTech Connect

    Clark, David Lewis

    2015-01-21

    The deceptively simple binary formula of AnO2 belies an incredibly complex structural nature, and propensity to form mixed-valent, nonstoichiometric phases of composition AnO2±x. For plutonium, the very formation of PuO2+x has challenged a long-established dogma, and raised fundamental questions for long-term storage and environmental migration. This presentation covers two aspects of Los Alamos synchrotron radiation studies of plutonium oxides: (1) the structural chemistry of laboratory-prepared AnO2+x systems (An = U, Pu; 0 ≤ x ≤ 0.25) determined through a combination of x-ray absorption fine structure spectroscopy (XAFS) and x-ray scattering of laboratory prepared samples; and (2) the application of synchrotron radiation towards the decontamination and decommissioning of the Rocky Flats Environmental Technology Site. Making the case for particle transport mechanisms as the basis of plutonium and americium mobility, rather than aqueous sorption-desorption processes, established a successful scientific basis for the dominance of physical transport processes by wind and water. The scientific basis was successful because it was in agreement with general theory on insolubility of PuO2 in oxidation state IV, results of ultrafiltration analyses of field water/sediment samples, XAFS analyses of soil, sediment, and concrete samples, and was also in general agreement with on-site monitoring data. This understanding allowed Site contractors to rapidly move to application of soil erosion and sediment transport models as the means of predicting plutonium and americium transport, which led to design and application of site-wide soil erosion control technology to help control downstream concentrations of plutonium and americium in streamflow.

  19. Performance of Thorium-Based Mixed Oxide Fuels for the Consumption of Plutonium in Current and Advanced Reactors

    SciTech Connect

    Weaver, Kevan Dean; Herring, James Stephen

    2003-07-01

    A renewed interest in thorium-based fuels has arisen lately based on the need for proliferation resistance, longer fuel cycles, higher burnup, and improved waste form characteristics. Recent studies have been directed toward homogeneously mixed, heterogeneously mixed, and seed-and-blanket thorium-uranium oxide fuel cycles that rely on "in situ" use of the bred-in 233U. However, due to the higher initial enrichment required to achieve acceptable burnups, these fuels are encountering economic constraints. Thorium can nevertheless play a large role in the nuclear fuel cycle, particularly in the reduction of plutonium inventories. While uranium-based mixed-oxide (MOX) fuel will decrease the amount of plutonium in discharged fuel, the reduction is limited due to the breeding of more plutonium (and higher actinides) from the 238U. Here, we present calculational results and a comparison of the potential burnup of a thorium-based and uranium-based mixed-oxide fuel in a light water reactor. Although the uranium-based fuels outperformed the thorium-based fuels in achievable burnup, a depletion comparison of the initially charged plutonium (both reactor and weapons grade) showed that the thorium-based fuels outperformed the uranium-based fuels by more that a factor of 2, where >70% of the total plutonium in the thorium-based fuel is consumed during the cycle. This is significant considering that the achievable burnups of the thorium-based fuels were 1.4 to 4.6 times less than the uranium-based fuels for similar plutonium enrichments. For equal specific burnups of ~60 MWd/kg (i.e., using variable plutonium weight percentages to give the desired burnup), the thorium-based fuels still outperform the uranium-based fuels by more than a factor of 2, where the total plutonium consumption in a three-batch, 18-month cycle was 60 to 70%. This is fairly significant considering that 10 to 15% (by weight) more plutonium is needed in the thorium-based fuels as compared to the uranium

  20. Source-dependent and source-independent controls on plutonium oxidation state and colloid associations in groundwater.

    PubMed

    Buesseler, Ken O; Kaplan, Daniel I; Dai, Minhan; Pike, Steven

    2009-03-01

    Plutonium (Pu) was characterized for its isotopic composition, oxidation states, and association with colloids in groundwater samples near disposal basins in F-Area of the Savannah River Site and compared to similar samples collected six years earlier. Two sources of Pu were identified, the disposal basins, which contained a 24Pu/l39Pu isotopic signature consistent with weapons grade Pu, and 244Cm, a cocontaminant that is a progenitor radionuclide of 24Pu. 24Pu that originated primarily from 244Cm tended to be appreciably more oxidized (Pu(V/VI)), less associated with colloids (approximately 1 kDa - 0.2 microm), and more mobile than 239Pu, as suggested by our prior studies at this site. This is not evidence of isotope fractionation but rather "source-dependent" controls on 240Pu speciation which are processes that are not at equilibrium, i.e., processes that appear kinetically hindered. There were also "source-independent" controls on 239Pu speciation, which are those processes that follow thermodynamic equilibrium with their surroundings. For example, a groundwater pH increase in one well from 4.1 in 1998 to 6.1 in 2004 resulted in an order of magnitude decrease in groundwater 239Pu concentrations. Similarly, the fraction of 239Pu in the reduced Pu(III/IV) and colloidal forms increased systematically with decreases in redox condition in 2004 vs 1998. This research demonstrates the importance of source-dependent and source-independent controls on Pu speciation which would impact Pu mobility during changes in hydrological, chemical, or biological conditions on both seasonal and decadal time scales, and over short spatial scales. This implies more dynamic shifts in Pu speciation, colloids association, and transport in groundwater than commonly believed. PMID:19350898

  1. Plutonium disproportionation: the ambiguity phenomenon.

    PubMed

    Silver, G L

    2003-05-01

    Plutonium oxidation-state studies may yield ambiguous results if the parameters are not carefully chosen. The effect can be related to environmental plutonium as illustrated by an example. PMID:12735968

  2. Chemical species of plutonium in Hanford radioactive tank waste

    SciTech Connect

    Barney, G.S.

    1997-10-22

    Large quantities of radioactive wastes have been generated at the Hanford Site over its operating life. The wastes with the highest activities are stored underground in 177 large (mostly one million gallon volume) concrete tanks with steel liners. The wastes contain processing chemicals, cladding chemicals, fission products, and actinides that were neutralized to a basic pH before addition to the tanks to prevent corrosion of the steel liners. Because the mission of the Hanford Site was to provide plutonium for defense purposes, the amount of plutonium lost to the wastes was relatively small. The best estimate of the amount of plutonium lost to all the waste tanks is about 500 kg. Given uncertainties in the measurements, some estimates are as high as 1,000 kg (Roetman et al. 1994). The wastes generally consist of (1) a sludge layer generated by precipitation of dissolved metals from aqueous wastes solutions during neutralization with sodium hydroxide, (2) a salt cake layer formed by crystallization of salts after evaporation of the supernate solution, and (3) an aqueous supernate solution that exists as a separate layer or as liquid contained in cavities between sludge or salt cake particles. The identity of chemical species of plutonium in these wastes will allow a better understanding of the behavior of the plutonium during storage in tanks, retrieval of the wastes, and processing of the wastes. Plutonium chemistry in the wastes is important to criticality and environmental concerns, and in processing the wastes for final disposal. Plutonium has been found to exist mainly in the sludge layers of the tanks along with other precipitated metal hydrous oxides. This is expected due to its low solubility in basic aqueous solutions. Tank supernate solutions do not contain high concentrations of plutonium even though some tanks contain high concentrations of complexing agents. The solutions also contain significant concentrations of hydroxide which competes with other

  3. Method for inhibiting oxidation of metal sulfide-containing material

    DOEpatents

    Elsetinow, Alicia; Borda, Michael J.; Schoonen, Martin A.; Strongin, Daniel R.

    2006-12-26

    The present invention provides means for inhibiting the oxidation of a metal sulfide-containing material, such as ore mine waste rock or metal sulfide taiulings, by coating the metal sulfide-containing material with an oxidation-inhibiting two-tail lipid coating (12) thereon, thereby inhibiting oxidation of the metal sulfide-containing material in acid mine drainage conditions. The lipids may be selected from phospholipids, sphingolipids, glycolipids and combinations thereof.

  4. Oxidation state and local structure of plutonium reacted with magnetite, mackinawite, and chukanovite.

    PubMed

    Kirsch, Regina; Fellhauer, David; Altmaier, Marcus; Neck, Volker; Rossberg, Andre; Fanghänel, Thomas; Charlet, Laurent; Scheinost, Andreas C

    2011-09-01

    Due to their redox reactivity, surface sorption characteristics, and ubiquity as corrosion products or as minerals in natural sediments, iron(II)-bearing minerals control to a large extent the environmental fate of actinides. Pu-L(III)-edge XANES and EXAFS spectra were used to investigate reaction products of aqueous (242)Pu(III) and (242)Pu(V) reacted with magnetite, mackinawite, and chukanovite under anoxic conditions. As Pu concentrations in the liquid phase were rapidly below detection limit, oxidation state and local structure of Pu were determined for Pu associated with the solid mineral phase. Pu(V) was reduced in the presence of all three minerals. A newly identified, highly specific Pu(III)-sorption complex formed with magnetite. Solid PuO(2) phases formed in the presence of mackinawite and chukanovite; in the case of chukanovite, up to one-third of plutonium was also present as Pu(III). This highlights the necessity to consider, under reducing anoxic conditions, Pu(III) species in addition to tetravalent PuO(2) for environmental risk assessment. Our results also demonstrate the necessity to support thermodynamic calculations with spectroscopic data. PMID:21755920

  5. Americium characterization by X-ray fluorescence and absorption spectroscopy in plutonium uranium mixed oxide

    NASA Astrophysics Data System (ADS)

    Degueldre, Claude; Cozzo, Cedric; Martin, Matthias; Grolimund, Daniel; Mieszczynski, Cyprian

    2013-06-01

    Plutonium uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The actinides in these fuels need to be analyzed after irradiation for assessing their behaviour with regard to their environment and the coolant. In this work the study of the atomic structure and next-neighbour environment of Am in the (Pu,U)O2 lattice in an irradiated (60 MW d kg-1) MOX sample was performed employing micro-X-ray fluorescence (µ-XRF) and micro-X-ray absorption fine structure (µ-XAFS) spectroscopy. The chemical bonds, valences and stoichiometry of Am (˜0.66 wt%) are determined from the experimental data gained for the irradiated fuel material examined in its peripheral zone (rim) of the fuel. In the irradiated sample Am builds up as Am3+ species within an [AmO8]13- coordination environment (e.g. >90%) and no (<10%) Am(IV) or (V) can be detected in the rim zone. The occurrence of americium dioxide is avoided by the redox buffering activity of the uranium dioxide matrix.

  6. Safe disposal of surplus plutonium

    NASA Astrophysics Data System (ADS)

    Gong, W. L.; Naz, S.; Lutze, W.; Busch, R.; Prinja, A.; Stoll, W.

    2001-06-01

    About 150 tons of weapons grade and weapons usable plutonium (metal, oxide, and in residues) have been declared surplus in the USA and Russia. Both countries plan to convert the metal and oxide into mixed oxide fuel for nuclear power reactors. Russia has not yet decided what to do with the residues. The US will convert residues into a ceramic, which will then be over-poured with highly radioactive borosilicate glass. The radioactive glass is meant to provide a deterrent to recovery of plutonium, as required by a US standard. Here we show a waste form for plutonium residues, zirconia/boron carbide (ZrO 2/B 4C), with an unprecedented combination of properties: a single, radiation-resistant, and chemically durable phase contains the residues; billion-year-old natural analogs are available; and criticality safety is given under all conceivable disposal conditions. ZrO 2/B 4C can be disposed of directly, without further processing, making it attractive to all countries facing the task of plutonium disposal. The US standard for protection against recovery can be met by disposal of the waste form together with used reactor fuel.

  7. In Vitro Dissolution Tests of Plutonium and Americium Containing Contamination Originating From ZPPR Fuel Plates

    SciTech Connect

    William F. Bauer; Brian K. Schuetz; Gary M. Huestis; Thomas B. Lints; Brian K. Harris; R. Duane Ball; Gracy Elias

    2012-09-01

    Assessing the extent of internal dose is of concern whenever workers are exposed to airborne radionuclides or other contaminants. Internal dose determinations depend upon a reasonable estimate of the expected biological half-life of the contaminants in the respiratory tract. One issue with refractory elements is determining the dissolution rate of the element. Actinides such as plutonium (Pu) and Americium (Am) tend to be very refractory and can have biological half-lives of tens of years. In the event of an exposure, the dissolution rates of the radionuclides of interest needs to be assessed in order to assign the proper internal dose estimates. During the November 2011 incident at the Idaho National Laboratory (INL) involving a ZPPR fuel plate, air filters in a constant air monitor (CAM) and a giraffe filter apparatus captured airborne particulate matter. These filters were used in dissolution rate experiments to determine the apparent dissolution half-life of Pu and Am in simulated biological fluids. This report describes these experiments and the results. The dissolution rates were found to follow a three term exponential decay equation. Differences were noted depending upon the nature of the biological fluid simulant. Overall, greater than 95% of the Pu and 93% of the Am were in a very slow dissolving component with dissolution half-lives of over 10 years.

  8. Container Prevents Oxidation Of Metal Powder

    NASA Technical Reports Server (NTRS)

    Woodford, William H.; Power, Christopher A.; Mckechnie, Timothy N.; Burns, David H.

    1992-01-01

    Sealed high-vacuum container holds metal powder required free of contamination by oxygen from point of manufacture to point of use at vacuum-plasma-spraying machine. Container protects powder from air during filling, storage, and loading of spraying machine. Eliminates unnecessary handling and transfer of powder from one container to another. Stainless-steel container sits on powder feeder of vacuum-plasma-spraying machine.

  9. Relative Humidity and the Susceptibility of Austenitic Stainless Steel to Stress Corrosion Cracking in an impure Plutonium Oxide Environment

    SciTech Connect

    Zapp, P.; Duffey, J.; Lam, P.; Dunn, K.

    2010-05-05

    Laboratory tests to investigate the corrosivity of moist plutonium oxide/chloride salt mixtures on 304L and 316L stainless steel coupons showed that corrosion occurred in selected samples. The tests exposed flat coupons for pitting evaluation and 'teardrop' stressed coupons for stress corrosion cracking (SCC) evaluation at room temperature to various mixtures of PuO{sub 2} and chloride-bearing salts for periods up to 500 days. The exposures were conducted in sealed containers in which the oxide-salt mixtures were loaded with about 0.6 wt % water from a humidified helium atmosphere. Observations of corrosion ranged from superficial staining to pitting and SCC. The extent of corrosion depended on the total salt concentration, the composition of the salt and the moisture present in the test environment. The most significant corrosion was found in coupons that were exposed to 98 wt % PuO{sub 2}, 2 wt % chloride salt mixtures that contained calcium chloride and 0.6 wt% water. SCC was observed in two 304L stainless steel teardrop coupons exposed in solid contact to a mixture of 98 wt % PuO{sub 2}, 0.9 wt % NaCl, 0.9 wt % KCl, and 0.2 wt % CaCl{sub 2}. The cracking was associated with the heat-affected zone of an autogenous weld that ran across the center of the coupon. Cracking was not observed in coupons exposed to the headspace gas above the solid mixture, or in coupons exposed to other mixtures with either no CaCl{sub 2} or 0.92 wt% CaCl{sub 2}. SCC was present where the 0.6 wt % water content exceeded the value needed to fully hydrate the available CaCl{sub 2}, but was absent where the water content was insufficient. These results reveal the significance of the relative humidity in the austenitic stainless steels environment to their susceptibility to corrosion. The relative humidity in the test environment was controlled by the water loading and the concentration of the hydrating salts such as CaCl{sub 2}. For each salt or salt mixture there is a threshold relative

  10. Electrode electrolyte interlayers containing cerium oxide for electrochemical fuel cells

    DOEpatents

    Borglum, Brian P.; Bessette, Norman F.

    2000-01-01

    An electrochemical cell is made having a porous fuel electrode (16) and a porous air electrode (13), with solid oxide electrolyte (15) therebetween, where the air electrode surface opposing the electrolyte has a separate, attached, dense, continuous layer (14) of a material containing cerium oxide, and where electrolyte (16) contacts the continuous oxide layer (14), without contacting the air electrode (13).

  11. Zone refining of plutonium metal

    SciTech Connect

    1997-05-01

    The purpose of this study was to investigate zone refining techniques for the purification of plutonium metal. The redistribution of 10 impurity elements from zone melting was examined. Four tantalum boats were loaded with plutonium impurity alloy, placed in a vacuum furnace, heated to 700{degrees}C, and held at temperature for one hour. Ten passes were made with each boat. Metallographic and chemical analyses performed on the plutonium rods showed that, after 10 passes, moderate movement of certain elements were achieved. Molten zone speeds of 1 or 2 inches per hour had no effect on impurity element movement. Likewise, the application of constant or variable power had no effect on impurity movement. The study implies that development of a zone refining process to purify plutonium is feasible. Development of a process will be hampered by two factors: (1) the effect on impurity element redistribution of the oxide layer formed on the exposed surface of the material is not understood, and (2) the tantalum container material is not inert in the presence of plutonium. Cold boat studies are planned, with higher temperature and vacuum levels, to determine the effect on these factors. 5 refs., 1 tab., 5 figs.

  12. Polyacrylonitrile fibers containing graphene oxide nanoribbons.

    PubMed

    Chien, An-Ting; Liu, H Clive; Newcomb, Bradley A; Xiang, Changsheng; Tour, James M; Kumar, Satish

    2015-03-11

    Graphene oxide nanoribbon (GONR) made by the oxidative unzipping of multiwalled carbon nanotube was dispersed in dimethylformamide and mixed with polyacrylonitrile (PAN) to fabricate continuous PAN/GONR composite fibers by gel spinning. Subsequently, PAN/GONR composite fibers were stabilized and carbonized in a batch process to fabricate composite carbon fibers. Structure, processing, and properties of the composite precursor and carbon fibers have been studied. This study shows that GONR can be used to make porous precursor and carbon fibers. In addition, GONR also shows the potential to make higher mechanical property carbon fibers than that achieved from PAN precursor only. PMID:25671488

  13. Functional Design Criteria plutonium stabilization and handling (PUSH) project W-460

    SciTech Connect

    NELSON, D.W.

    1999-09-02

    This Functional Design Criteria (FDC) contains information to guide the design of the Stabilization and Packaging Equipment necessary to oxidize and package the remaining plutonium-bearing Special Nuclear Materials (SNM) currently in the Plutonium Finishing Plant (PFP) inventory. The FDC also guides the design of vault modifications to allow storage of 3013 packages of stabilized SNM for up to 50 years.

  14. SEPARATION OF FISSION PRODUCT VALUES FROM THE HEXAVALENT PLUTONIUM BY CARRIER PRECIPITATION

    DOEpatents

    Davies, T.H.

    1959-12-15

    An improved precipitation of fission products on bismuth phosphate from an aqueous mineral acid solution also containing hexavalent plutonium by incorporating, prior to bismuth phosphate precipitation, from 0.05 to 2.5 grams/ liter of zirconium phosphate, niobium oxide. and/or lanthanum fluoride is described. The plutonium remains in solution.

  15. Oxidation of Pu(IV) and Pu(V) with sodium hypochlorite

    NASA Astrophysics Data System (ADS)

    Choppin, G. R.; Morgenstern, A.

    2000-07-01

    The hypochlorite ion (OCl-), a major product of the radiolysis of water in solutions containing high concentrations of chloride, can be expected to influence the oxidation state distribution of plutonium in these solutions. Since plutonium has significantly different chemistry in each of its oxidation states, knowledge of the oxidation state distribution of plutonium and of the kinetics of transfer between these oxidation states is essential for modeling the behavior of plutonium in aqueous systems and for design of efficient remediation procedures for plutonium containing wastes.

  16. PREPARATION OF PLUTONIUM

    DOEpatents

    Kolodney, M.

    1959-07-01

    Methods are presented for the electro-deposition of plutonium from fused mixtures of plutonium halides and halides of the alkali metals and alkaline earth metals. Th salts, preferably chlorides and with the plutonium prefer ably in the trivalent state, are placed in a refractory crucible such as tantalum or molybdenam and heated in a non-oxidizing atmosphere to 600 to 850 deg C, the higher temperatatures being used to obtain massive plutonium and the lower for the powder form. Electrodes of graphite or non reactive refractory metals are used, the crucible serving the cathode in one apparatus described in the patent.

  17. Extended testing of a modified 18B plutonium nitrate shipping container

    SciTech Connect

    Yoshimura, H.R.; Pope, R.B.; Leisher, W.B.; Joseph, B.J.; Schulz-Forberg, B.; Hubner, H.W.

    1980-01-01

    The container damage observed as the result of the high-speed pulldown impact test was more severe than that of either the 185-m free-fall drop of a prototype container onto a semirigid surface or the crush environment produced by a 9-m drop of a 2-tonne block onto a modified container resting on an unyielding surface. In comparison to the extended tests, the 9-m regulatory drop test onto an unyielding surface of the prototype packaging in its most damaging orientation produced the least amount of damage. Very little deformation in the overpack was observed, and there was no influence on the fire resistivity and leaktightness of the containment vessel. The 128 m/s impact test produced a leak in the container. It appears that the 18B packaging, designed to withstand the environments specified in IAEA Safety Series No. 6, can withstand extended environments including longer duration fires and higher velocity impacts on yielding targets. When modified with ring stiffeners, the packaging withstood a dynamic crush test, but did not survive the high speed impact onto an unyielding surface as specified in NUREG 0360.

  18. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    SciTech Connect

    Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W.

    1995-08-01

    Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program.

  19. Nitrogen oxides storage catalysts containing cobalt

    DOEpatents

    Lauterbach, Jochen; Snively, Christopher M.; Vijay, Rohit; Hendershot, Reed; Feist, Ben

    2010-10-12

    Nitrogen oxides (NO.sub.x) storage catalysts comprising cobalt and barium with a lean NO.sub.x storage ratio of 1.3 or greater. The NO.sub.x storage catalysts can be used to reduce NO.sub.x emissions from diesel or gas combustion engines by contacting the catalysts with the exhaust gas from the engines. The NO.sub.x storage catalysts can be one of the active components of a catalytic converter, which is used to treat exhaust gas from such engines.

  20. Glycerol oxidation using gold-containing catalysts.

    PubMed

    Villa, Alberto; Dimitratos, Nikolaos; Chan-Thaw, Carine E; Hammond, Ceri; Prati, Laura; Hutchings, Graham J

    2015-05-19

    Glycerol is an important byproduct of biodiesel production, and it is produced in significant amounts by transesterification of triglycerides with methanol. Due to the highly functionalized nature of glycerol, it is an important biochemical that can be utilized as a platform chemical for the production of high-added-value products. At present, research groups in academia and industry are exploring potential direct processes for the synthesis of useful potential chemicals using catalytic processes. Over the last 10 years, there has been huge development of potential catalytic processes using glycerol as the platform chemical. One of the most common processes investigated so far is the catalytic oxidation of glycerol at mild conditions for the formation of valuable oxygenated compounds used in the chemical and pharmaceutical industry. The major challenges associated with the selective oxidation of glycerol are (i) the control of selectivity to the desired products, (ii) high activity and resistance to poisoning, and (iii) minimizing the usage of alkaline conditions. To address these challenges, the most common catalysts used for the oxidation of glycerol are based on supported metal nanoparticles. The first significant breakthrough was the successful utilization of supported gold nanoparticles for improving the selectivity to specific products, and the second was the utilization of supported bimetallic nanoparticles based on gold, palladium, and platinum for improving activity and controlling the selectivity to the desired products. Moreover, the utilization of base-free reaction conditions for the catalytic oxidation of glycerol has unlocked new pathways for the production of free-base products, which facilitates potential industrial application. The advantages of using gold-based catalysts are the improvement of the catalyst lifetime, stability, and reusability, which are key factors for potential commercialization. In this Account, we discuss the advantages of the

  1. Decontamination of matrices containing actinide oxides

    SciTech Connect

    Villarreal, Robert

    1997-12-01

    There is provided a method for removing actinides and actinide oxides, particularly fired actinides, from soil and other contaminated matrices, comprising: (a) contacting a contaminated material with a solution of at least one inhibited fluoride and an acid to form a mixture; (b) heating the mixture of contaminated material and solution to a temperature in the range from about 30 C to about 90 C while stirring; (c) separating the solution from any undissolved matrix material in the mixture; (d) washing the undissolved matrix material to remove any residual materials; and (e) drying and returning the treated matrix material to the environment.

  2. MCNP analysis of PNL split-table critical experiments containing mixed-oxide fuels

    SciTech Connect

    Abdurrahman, N.M.; Yavuz, M.; Radulescu, G.

    1997-12-01

    Pacific Northwest Laboratory (PNL) Split-Table Critical experiments containing mixed-oxide (MOX) fuels for various core configurations are studied using MCNP4A with the ENDF/B-VI continuous-energy library. These experiments were performed to provide necessary technical information and experimental criticality data that would serve as benchmark data in support of the liquid-metal fast breeder reactor program. Because of the current interest in the utilization of weapons-grade plutonium in the form of MOX fuel in light water reactors, such experimental data are extremely important for checking the performance of the modem computational tools. The {sup 239}Pu content in plutonium of the PNL MOX fuels is {approximately}91 wt%, which is very close to that of the weapons-grade {sup 239}Pu. The MOX fuels used in these critical experiments consist of 30.0, 14.62, and 7.89 wt% Pu and N{sub H}/(N{sub Pu} + Nu) moderation ratios (MRs) of 47.4, 30.6, and 51.8, respectively.

  3. Shielding Evaluation of Plutonium and Uranium Contents in 9975 Shipping Containers

    SciTech Connect

    Vincent, A.

    2003-01-27

    Shielding evaluations were performed in support of developing the Safety Analysis Report for Packagings (SARP) for the 9975 Shipping container. The objective of these evaluations was to demonstrate compliance with the performance requirements specified in federal regulations for each content envelope.

  4. Simulation and analysis of the plutonium shipping container subject to 30-foot drops

    SciTech Connect

    Gong, C.; Gupta, N.K.; Gromada, R.J.

    1995-12-31

    The shipping container 5320 is a shipping package for radioactive materials. In order to maintain the component in this packaging within the sub-critical state when subjected to any kind of Hypothetical Accident conditions (HAC), this Type B packaging is designed with various impact limiters. The present study is to examine the energy absorbing capacity of the impact limiter design of this container subjected to a 30-foot drop onto a flat unyielding horizontal surface in each of the three critical dropping orientations. This paper presents the results of a three dimensional nonlinear dynamic impact analysis. This analysis shows the deformed configuration of the container caused by the impact and also determines the effects of different stress wave paths in three distinct drops on the stress states in the critical component. The solution to the problem was obtained using the ABAQUS (explicit) finite element computer code. The nonlinearity of this analysis involves large structural deformation, elasto-plastic materials with strain hardening as well as multiple contact interfaces. Three drop orientations were studied, namely, top down impact, bottom down impact and side impact. Results will be compared against actual drop test data.

  5. METHOD FOR DISSOLVING LANTHANUM FLUORIDE CARRIER FOR PLUTONIUM

    DOEpatents

    Koshland, D.E. Jr.; Willard, J.E.

    1961-08-01

    A method is described for dissolving lanthanum fluoride precipitates which is applicable to lanthanum fluoride carrier precipitation processes for recovery of plutonium values from aqueous solutions. The lanthanum fluoride precipitate is contacted with an aqueous acidic solution containing dissolved zirconium in the tetravalent oxidation state. The presence of the zirconium increases the lanthanum fluoride dissolved and makes any tetravalent plutonium present more readily oxidizable to the hexavalent state. (AEC)

  6. Plutonium storage criteria

    SciTech Connect

    Chung, D.; Ascanio, X.

    1996-05-01

    The Department of Energy has issued a technical standard for long-term (>50 years) storage and will soon issue a criteria document for interim (<20 years) storage of plutonium materials. The long-term technical standard, {open_quotes}Criteria for Safe Storage of Plutonium Metals and Oxides,{close_quotes} addresses the requirements for storing metals and oxides with greater than 50 wt % plutonium. It calls for a standardized package that meets both off-site transportation requirements, as well as remote handling requirements from future storage facilities. The interim criteria document, {open_quotes}Criteria for Interim Safe Storage of Plutonium-Bearing Solid Materials{close_quotes}, addresses requirements for storing materials with less than 50 wt% plutonium. The interim criteria document assumes the materials will be stored on existing sites, and existing facilities and equipment will be used for repackaging to improve the margin of safety.

  7. 5. View, oxidizer waste tanks and containment basin in foreground ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    5. View, oxidizer waste tanks and containment basin in foreground with Systems Integration Laboratory (T-28) uphill in background, looking northeast. - Air Force Plant PJKS, Systems Integration Laboratory, Waterton Canyon Road & Colorado Highway 121, Lakewood, Jefferson County, CO

  8. 9. View, oxidizer waste tanks and containment basin associated with ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    9. View, oxidizer waste tanks and containment basin associated with Components Test Laboratory (T-27) located directly uphill, looking north. Located uphill in the upper left portion of the photograph (from right to left) are the Oxidizer Conditioning Structure (T-28D), Long-Term Oxidizer Silo (T-28B), and Systems Integration Laboratory (T-28). - Air Force Plant PJKS, Systems Integration Laboratory, Components Test Laboratory, Waterton Canyon Road & Colorado Highway 121, Lakewood, Jefferson County, CO

  9. Demonstrating a Glass Ceramic route for the Immobilisation of Plutonium containing Wastes and Residues on the Sellafield Site

    SciTech Connect

    Scales, C.R.; Maddrell, E.R.; Gawthorpe, N.; Begg, B.D.; Moricca, S.; Day, R.A.; Stewart, M.A.

    2006-07-01

    A wide range of plutonium containing wastes and residues are currently stored on the Sellafield site. These wastes and residues arising from early fuel development activities at Sellafield are diverse in nature and in the absence of a long-term disposition strategy, ongoing storage represents a major financial commitment. An immobilisation technology is being developed, producing a glass ceramic wasteform that will be suitable for extended periods of storage and ultimate disposal. Hot isostatic pressing [HIP] has been chosen as the preferred consolidation route. Following demonstration of the chemical feasibility of the approach at ANSTO's Lucas Heights facility, the project will demonstrate the technology by designing and building a full-scale inactive facility at Nexia Solutions' Workington site supported by a small scale active facility in the Technology Centre (BTC) at Sellafield. Delivery of this technology will allow the project sponsors, British Nuclear Group, the option of engineering a full scale plant in order to immobilise the actual waste arisings and discharge the liability to site. (authors)

  10. DISPOSAL OF TRU WASTE FROM THE PLUTONIUM FINISHING PLANT IN PIPE OVERPACK CONTAINERS TO WIPP INCLUDING NEW SECURITY REQUIREMENTS

    SciTech Connect

    Hopkins, A.M.; Sutter, C.; Hulse, G.; Teal, J.

    2003-02-27

    The Department of Energy is responsible for the safe management and cleanup of the DOE complex. As part of the cleanup and closure of the Plutonium Finishing Plant (PFP) located on the Hanford site, the nuclear material inventory was reviewed to determine the appropriate disposition path. Based on the nuclear material characteristics, the material was designated for stabilization and packaging for long term storage and transfer to the Savannah River Site or, a decision for discard was made. The discarded material was designated as waste material and slated for disposal to the Waste Isolation Pilot Plant (WIPP). Prior to preparing any residue wastes for disposal at the WIPP, several major activities need to be completed. As detailed a processing history as possible of the material including origin of the waste must be researched and documented. A technical basis for termination of safeguards on the material must be prepared and approved. Utilizing process knowledge and processing history, the material must be characterized, sampling requirements determined, acceptable knowledge package and waste designation completed prior to disposal. All of these activities involve several organizations including the contractor, DOE, state representatives and other regulators such as EPA. At PFP, a process has been developed for meeting the many, varied requirements and successfully used to prepare several residue waste streams including Rocky Flats incinerator ash, Hanford incinerator ash and Sand, Slag and Crucible (SS&C) material for disposal. These waste residues are packed into Pipe Overpack Containers for shipment to the WIPP.

  11. Americium characterization by X-ray fluorescence and absorption spectroscopy in plutonium uranium mixed oxide

    SciTech Connect

    Degueldre, Claude Cozzo, Cedric; Martin, Matthias; Grolimund, Daniel; Mieszczynski, Cyprian

    2013-06-01

    Plutonium uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The actinides in these fuels need to be analyzed after irradiation for assessing their behaviour with regard to their environment and the coolant. In this work the study of the atomic structure and next-neighbour environment of Am in the (Pu,U)O₂ lattice in an irradiated (60 MW d kg⁻¹) MOX sample was performed employing micro-X-ray fluorescence (µ-XRF) and micro-X-ray absorption fine structure (µ-XAFS) spectroscopy. The chemical bonds, valences and stoichiometry of Am (~0.66 wt%) are determined from the experimental data gained for the irradiated fuel material examined in its peripheral zone (rim) of the fuel. In the irradiated sample Am builds up as Am³⁺ species within an [AmO₈]¹³⁻ coordination environment (e.g. >90%) and no (<10%) Am(IV) or (V) can be detected in the rim zone. The occurrence of americium dioxide is avoided by the redox buffering activity of the uranium dioxide matrix. - Graphical abstract: Americium LIII XAFS spectra recorded for the irradiated MOX sub-sample in the rim zone for a 300 μm×300 μm beam size area investigated over six scans of 4 h. The records remain constant during multi-scan. The analysis of the XAFS signal shows that Am is found as trivalent in the UO₂ matrix. This analytical work shall open the door of very challenging analysis (speciation of fission product and actinides) in irradiated nuclear fuels. - Highlights: • Americium was characterized by microX-ray absorption spectroscopy in irradiated MOX fuel. • The americium redox state as determined from XAS data of irradiated fuel material was Am(III). • In the sample, the Am³⁺ face an AmO₈¹³⁻coordination environment in the (Pu,U)O₂ matrix. • The americium dioxide is reduced by the uranium dioxide matrix.

  12. Room-temperature oxidation of hypostoichiometric uranium-plutonium mixed oxides U1-yPuyO2-x - A depth-selective approach

    NASA Astrophysics Data System (ADS)

    Vauchy, Romain; Robisson, Anne-Charlotte; Belin, Renaud C.; Martin, Philippe M.; Scheinost, Andreas C.; Hodaj, Fiqiri

    2015-10-01

    In the present work, TGA, XAS and XRD were used to evidence the spontaneous oxidation of biphasic U1-yPuyO2-x samples, with y = 0.28 and 0.45, at room temperature and upon exposure to low moisture and oxygen contents. The oxidation occurs within very short timescales (e.g. O/M ratio increasing from 1.94 to 1.98 within ∼1 μm surface layer in ∼50 h). The combined use of these three complementary methods offered a depth-selective approach from the sample's bulk to its surface and allowed a thorough understanding of the underlying processes involved during the formation of the oxidized layer and of its thickening with time. We believe our results to be of interest in the prospect of fabricating hypo-stoichiometric uranium-plutonium mixed oxides since mastering the oxygen content is a crucial point for many of the fuel properties.

  13. Purification of crude hexafluoroacetone containing nitrogen oxides and sulfur dioxide

    SciTech Connect

    Bonfield, J. H.; Karsay, B. I.

    1984-09-25

    Crude hexafluoroacetone containing as impurities nitrogen oxides and sulfur dioxide is purified by admixing with water to form an aqueous solution, admixing the aqueous solution with concentrated sulfuric acid or oleum to form a vapor and scrubbing the vapor with liquid concentrated sulfuric acid to produce purified anhydrous hexafluoroacetone. The sulfur dioxide and nitrogen oxides interact with the aqueous solution and conc

  14. Automotive body panel containing thermally exfoliated graphite oxide

    NASA Technical Reports Server (NTRS)

    Prud'Homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor); Adamson, Douglas (Inventor); Abdala, Ahmed (Inventor)

    2011-01-01

    An automotive body panel containing a polymer composite formed of at least one polymer and a modified graphite oxide material, which is a thermally exfoliated graphite oxide with a surface area of from about 300 m.sup.2/g to 2600 m.sup.2/g.

  15. Plutonium Immobilization Project -- Can loading

    SciTech Connect

    Kriikku, E.

    2000-01-18

    The Savannah River Site (SRS) will immobilize excess plutonium in the proposed Plutonium Immobilization Project (PIP). The PIP scope includes unloading transportation containers, preparing the feed streams, converting the metal feed to an oxide, adding the ceramic precursors, pressing the pucks, inspecting pucks, and sintering pucks. The PIP scope also includes loading the pucks into metal cans, sealing the cans, inspecting the cans, loading the cans into magazines, loading magazines into Defense Waste Processing Facility (DWPF) canisters, and transporting the canisters to the DWPF. The DWPF fills the canister with a mixture of high level radioactive waste and glass for permanent storage. Due to the radiation, remote equipment must perform PIP operations in a contained environment.

  16. Treatment of aqueous streams containing strong oxidants using bituminous coal

    SciTech Connect

    Doyle, F.M.; Bodine, D.L.

    1995-12-31

    Certain oxidizing contaminants, notably Cr(VI) and Mn(VII), are attenuated by reduction and sorption on organic matter in soils. Coals have some chemical similarity with this organic matter, and might be used on an industrial scale to treat effluents. We have studied the ability of acidic KMnO{sub 4} to oxidize Upper Freeport, bituminous coal with concurrent sorption of the resulting Mn(IV) and Mn(II). The oxidizing ability of Cr(VI) was briefly investigated. The ability of the oxidized coal to sorb Cu{sup 2+} and Cd{sup 2+} was then studied, and compared with coal oxidized by hydrogen peroxide. The effect of oxidation treatment, metal ion concentration, and solution pH on metal uptake kinetics and coal loading was investigated. Potential applications for treating effluents containing oxidizing ions are discussed.

  17. EXAFS and XANES analysis of plutonium and cerium edges from titanate ceramics for fissile materials disposal.

    SciTech Connect

    Fortner, J. A.; Kropf, A. J.; Bakel, A. J.; Hash, M. C.; Aase, S. B.; Buck, E. C.; Chamerlain, D. B.

    1999-11-16

    We report x-ray absorption near edge structure (XANES) and extended x-ray absorption fine structure (EXAFS) spectra from the plutonium L{sub III} edge and XANES from the cerium L{sub II} edge in prototype titanate ceramic hosts. The titanate ceramics studied are based upon the hafnium-pyrochlore and zirconolite mineral structures and will serve as an immobilization host for surplus fissile materials, containing as much as 10 weight % fissile plutonium and 20 weight % (natural or depleted) uranium. Three ceramic formulations were studied: one employed cerium as a ''surrogate'' element, replacing both plutonium and uranium in the ceramic matrix, another formulation contained plutonium in a ''baseline'' ceramic formulation, and a third contained plutonium in a formulation representing a high-impurity plutonium stream. The cerium XANES from the surrogate ceramic clearly indicates a mixed III-IV oxidation state for the cerium. In contrast, XANES analysis of the two plutonium-bearing ceramics shows that the plutonium is present almost entirely as Pu(IV) and occupies the calcium site in the zirconolite and pyrochlore phases. The plutonium EXAFS real-space structure shows a strong second-shell peak, clearly distinct from that of PuO{sub 2}, with remarkably little difference in the plutonium crystal chemistry indicated between the baseline and high-impurity formulations.

  18. The role of troublesome components in plutonium vitrification

    SciTech Connect

    Li, Hong; Vienna, J.D.; Peeler, D.K.; Hrma, P.; Schweiger, M.J.

    1996-05-01

    One option for immobilizing surplus plutonium is vitrification in a borosilicate glass. Two advantages of the glass form are (1) high tolerance to feed variability and, (2) high solubility of some impurity components. The types of plutonium-containing materials in the United States inventory include: pits, metals, oxides, residues, scrap, compounds, and fuel. Many of them also contain high concentrations of carbon, chloride, fluoride, phosphate, sulfate, and chromium oxide. To vitrify plutonium-containing scrap and residues, it is critical to understand the impact of each component on glass processing and chemical durability of the final product. This paper addresses glass processing issues associated with these troublesome components. It covers solubility limits of chlorine, fluorine, phosphate, sulfate, and chromium oxide in several borosilicate based glasses, and the effect of each component on vitrification (volatility, phase segregation, crystallization, and melt viscosity). Techniques (formulation, pretreatment, removal, and/or dilution) to mitigate the effect of these troublesome components are suggested.

  19. ELECTRODEPOSITION OF PLUTONIUM

    DOEpatents

    Wolter, F.J.

    1957-09-10

    A process of electrolytically recovering plutonium from dilute aqueous solutions containing plutonium ions comprises electrolyzing the solution at a current density of about 0.44 ampere per square centimeter in the presence of an acetate-sulfate buffer while maintaining the pH of the solution at substantially 5 and using a stirred mercury cathode.

  20. APPLICATION OF COLUMN EXTRACTION METHOD FOR IMPURITIES ANALYSIS ON HB-LINE PLUTONIUM OXIDE IN SUPPORT OF MOX FEED PRODUCT SPECIFICATIONS

    SciTech Connect

    Jones, M.; Diprete, D.; Wiedenman, B.

    2012-03-20

    The current mission at H-Canyon involves the dissolution of an Alternate Feedstocks 2 (AFS-2) inventory that contains plutonium metal. Once dissolved, HB-Line is tasked with purifying the plutonium solution via anion exchange, precipitating the Pu as oxalate, and calcining to form plutonium oxide (PuO{sub 2}). The PuO{sub 2} will provide feed product for the Mixed Oxide (MOX) Fuel Fabrication Facility, and the anion exchange raffinate will be transferred to H-Canyon. The results presented in this report document the potential success of the RE resin column extraction application on highly concentrated Pu samples to meet MOX feed product specifications. The original 'Hearts Cut' sample required a 10000x dilution to limit instrument drift on the ICP-MS method. The instrument dilution factors improved to 125x and 250x for the sample raffinate and sample eluent, respectively. As noted in the introduction, the significantly lower dilutions help to drop the total MRL for the analyte. Although the spike recoveries were half of expected in the eluent for several key elements, they were between 94-98% after Nd tracer correction. It is seen that the lower ICD limit requirements for the rare earths are attainable because of less dilution. Especially important is the extremely low Ga limit at 0.12 {mu}g/g Pu; an ICP-MS method is now available to accomplish this task on the sample raffinate. While B and V meet the column A limits, further development is needed to meet the column B limits. Even though V remained on the RE resin column, an analysis method is ready for investigation on the ICP-MS, but it does not mean that V cannot be measured on the ICP-ES at a low dilution to meet the column B limits. Furthermore, this column method can be applicable for ICP-ES as shown in Table 3-2, in that it trims the sample of Pu, decreasing and sometimes eliminating Pu spectral interferences.

  1. In-Drift Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

    SciTech Connect

    H.W> Stockman; S. LeStrange

    2000-09-28

    reactive components in the stone (such as calcium and silica); the contrasting chemistry of water trapped in the pores of the invert; and the possible presence of reducing materials in the support structure. This calculation estimates the amounts of Pu and U that may accumulate in the invert as a consequence of chemical precipitation. The degradation scenario is consistent with the overall degradation analysis methodology outlined in Section 3.1 of the Disposal Criticality Analysis Methodology Topical Report (Ref. 47). Specifically, the scenario NF-1 b of Figure 3-2a of that document prescribes the processes analyzed in this calculation. The only difference is that the present calculation extends the possibilities of in-drift interacting materials to include incompletely oxidized iron, which was not a major constituent of the drift at the time the Topical Report was written.

  2. EXAFS/XANES studies of plutonium-loaded sodalite/glass waste forms

    NASA Astrophysics Data System (ADS)

    Richmann, Michael K.; Reed, Donald T.; Kropf, A. Jeremy; Aase, Scott B.; Lewis, Michele A.

    2001-09-01

    A sodalite/glass ceramic waste form is being developed to immobilize highly radioactive nuclear wastes in chloride form, as part of an electrochemical cleanup process. Two types of simulated waste forms were studied: where the plutonium was alone in an LiCl/KCl matrix and where simulated fission-product elements were added representative of the electrometallurgical treatment process used to recover uranium from spent nuclear fuel also containing plutonium and a variety of fission products. Extended X-ray absorption fine structure spectroscopy (EXAFS) and X-ray absorption near-edge spectroscopy (XANES) studies were performed to determine the location, oxidation state, and particle size of the plutonium within these waste form samples. Plutonium was found to segregate as plutonium(IV) oxide with a crystallite size of at least 4.8 nm in the non-fission-element case and 1.3 nm with fission elements present. No plutonium was observed within the sodalite in the waste form made from the plutonium-loaded LiCl/KCl eutectic salt. Up to 35% of the plutonium in the waste form made from the plutonium-loaded simulated fission-product salt may be segregated with a heavy-element nearest neighbor other than plutonium or occluded internally within the sodalite lattice.

  3. Characterization of uranium and plutonium containing particles originating from the nuclear weapons accident in Thule, Greenland, 1968.

    PubMed

    Lind, O C; Salbu, B; Janssens, K; Proost, K; Dahlgaard, H

    2005-01-01

    To improve long-term radioecological impact assessment for the contaminated ecosystem of Bylot Sound, Greenland, U and Pu containing particles have been characterized with respect to particle size, elemental distribution, morphology and oxidation states. Based on scanning electron microscopy with XRMA, particles ranging from about 20 to 40 microm were isolated. XRMA and mu-XRF mapping demonstrated that U and Pu were homogeneously distributed throughout the particles, indicating that U and Pu have been fused. Furthermore, mu-XANES showed that U and Pu in the particles were present as mixed oxides. U was found to be in oxidation state IV whereas Pu apparently is a mixture of Pu(III) and Pu(IV). As previous assessments are based on PuO2 only, revisions should be made, taking Pu(III) into account. PMID:15748658

  4. PRECIPITATION METHOD FOR THE SEPARATION OF PLUTONIUM AND RARE EARTHS

    DOEpatents

    Thompson, S.G.

    1960-04-26

    A method of purifying plutonium is given. Tetravalent plutonium is precipitated with thorium pyrophosphate, the plutonium is oxidized to the tetravalent state, and then impurities are precipitated with thorium pyrophosphate.

  5. Synthesis and oxidation of some azole-containing thioethers

    PubMed Central

    Chernova, Nina P; Ogorodnikov, Vladimir D; Petrenko, Tatiana V

    2011-01-01

    Summary Pyrazole and benzotriazole-containing thioethers, namely 1,5-bis(3,5-dimethylpyrazol-1-yl)-3-thiapentane, 1,8-bis(3,5-dimethylpyrazol-1-yl)-3,6-dithiaoctane and 1,3-bis(1,2,3-benzotriazol-1-yl)-2-thiapropane were prepared and fully characterized. Oxidation of the pyrazole-containing thioether by hydrogen peroxide proceeds selectively to provide a sulfoxide or sulfone, depending on the amount of oxidant used. Oxidation of the benzotriazole derivative by hydrogen peroxide is not selective, and sulfoxide and sulfone form concurrently. Selenium dioxide-catalyzed oxidation of benzotriazole thioether by H2O2, however, proceeds selectively and yields sulfoxide only. PMID:22238528

  6. The feasibility of using molten carbonate corrosion for separating a nuclear surrogate for plutonium oxide from silicon carbide inert matrix

    NASA Astrophysics Data System (ADS)

    Cheng, Ting; Baney, Ronald H.; Tulenko, James

    2010-10-01

    Silicon carbide is one of the prime candidates as a matrix material in inert matrix fuels (IMF) being designed to reduce the plutonium inventories. Since complete fission and transmutation is not practical in a single in-core run, it is necessary to separate the non-transmuted actinide materials from the silicon carbide matrix for recycling. In this work, SiC was corroded in sodium carbonate (Na 2CO 3) and potassium carbonate (K 2CO 3), to form water soluble sodium or potassium silicate. Separation of the transuranics was achieved by dissolving the SiC corrosion product in boiling water. Ceria (CeO 2), which was used as a surrogate for plutonium oxide (PuO 2), was not corroded in these molten salt environments. The molten salt depth, which is a distance between the salt/air interface to the upper surface of SiC pellets, significantly affected the rate of corrosion. The corrosion was faster in K 2CO 3 than in Na 2CO 3 molten salt at 1050 °C, when the initial molten salt depths were kept the same for both salts.

  7. Dissolution of plutonium metal in HNO/sub 3/-N/sub 2/H/sub 4/-KF

    SciTech Connect

    Karraker, D G

    1983-07-01

    Plutonium metal dissolves in HNO/sub 3/-N/sub 2/H/sub 4/.HNO/sub 3/-KF solution to yield a Pu/sup 3 +/ solution without an accompanying precipitation of plutonium oxide solids. The reaction evolves less than 0.2 mole of gas per mole of plutonium dissolved; the gas contains only 3% H/sub 2/. About 10/sup -3/ moles of HN/sub 3/ are produced per mole of plutonium dissolved. Optimum conditions for dissolving both alpha-phase and delta-phase plutonium metal were developed. Possible applications are to the recovery of plutonium metal or the processing of irradiated plutonium metal and alloys.

  8. The quantification of mixture stoichiometry when fuel molecules contain oxidizer elements or oxidizer molecules contain fuel elements.

    SciTech Connect

    Mueller, Charles J.

    2005-05-01

    The accurate quantification and control of mixture stoichiometry is critical in many applications using new combustion strategies and fuels (e.g., homogeneous charge compression ignition, gasoline direct injection, and oxygenated fuels). The parameter typically used to quantify mixture stoichiometry (i.e., the proximity of a reactant mixture to its stoichiometric condition) is the equivalence ratio, /gf. The traditional definition of /gf is based on the relative amounts of fuel and oxidizer molecules in a mixture. This definition provides an accurate measure of mixture stoichiometry when the fuel molecule does not contain oxidizer elements and when the oxidizer molecule does not contain fuel elements. However, the traditional definition of /gf leads to problems when the fuel molecule contains an oxidizer element, as is the case when an oxygenated fuel is used, or once reactions have started and the fuel has begun to oxidize. The problems arise because an oxidizer element in a fuel molecule is counted as part of the fuel, even though it acts as an oxidizer. Similarly, if an oxidizer molecule contains fuel elements, the fuel elements in the oxidizer molecule are misleadingly lumped in with the oxidizer in the traditional definition of /gf. In either case, use of the traditional definition of /gf to quantify the mixture stoichiometry can lead to significant errors. This paper introduces the oxygen equivalence ratio, /gf/gV, a parameter that properly characterizes the instantaneous mixture stoichiometry for a broader class of reactant mixtures than does /gf. Because it is an instantaneous measure of mixture stoichiometry,/gf/gV can be used to track the time-evolution of stoichiometry as a reaction progresses. The relationship between /gf/gV and /gf is shown. Errors are involved when the traditional definition of /gf is used as a measure of mixture stoichiometry with fuels that contain oxidizer elements or oxidizers that contain fuel elements; /gf/gV is used to quantify

  9. Container effect in nanocasting synthesis of mesoporous metal oxides.

    PubMed

    Sun, Xiaohong; Shi, Yifeng; Zhang, Peng; Zheng, Chunming; Zheng, Xinyue; Zhang, Fan; Zhang, Yichi; Guan, Naijia; Zhao, Dongyuan; Stucky, Galen D

    2011-09-21

    We report a general reaction container effect in the nanocasting synthesis of mesoporous metal oxides. The size and shape of the container body in conjunction with simply modifying the container opening accessibility can be used to control the escape rate of water and other gas-phase byproducts in the calcination process, and subsequently affect the nanocrystal growth of the materials inside the mesopore space of the template. In this way, the particle size, mesostructure ordering, and crystallinity of the final product can be systemically controlled. The container effect also explain some of the problems with reproducibility in previously reported results. PMID:21861449

  10. Container Effect in Nanocasting Synthesis of Mesoporous Metal Oxides

    SciTech Connect

    Sun, Xiaohong; Shi, Yifeng; Zhang, Peng; Zheng, Chunming; Zheng, Xinyue; Zhang, Fan; Zhang, Yichi; Guan, Naijia; Zhao, Dongyuan; Stucky, Galen D.

    2011-08-23

    We report a general reaction container effect in the nanocasting synthesis of mesoporous metal oxides. The size and shape of the container body in conjunction with simply modifying the container opening accessibility can be used to control the escape rate of water and other gas-phase byproducts in the calcination process, and subsequently affect the nanocrystal growth of the materials inside the mesopore space of the template. In this way, the particle size, mesostructure ordering, and crystallinity of the final product can be systemically controlled. The container effect also explain some of the problems with reproducibility in previously reported results.

  11. Plutonium recovery from organic materials

    DOEpatents

    Deaton, R.L.; Silver, G.L.

    1973-12-11

    A method is described for removing plutonium or the like from organic material wherein the organic material is leached with a solution containing a strong reducing agent such as titanium (III) (Ti/sup +3None)/, chromium (II) (Cr/ sup +2/), vanadium (II) (V/sup +2/) ions, or ferrous ethylenediaminetetraacetate (EDTA), the leaching yielding a plutonium-containing solution that is further processed to recover plutonium. The leach solution may also contain citrate or tartrate ion. (Official Gazette)

  12. Process for the separation of sulfur oxides from a gaseous mixture containing sulfur oxides and oxygen

    SciTech Connect

    Derosset, A.J.; Ginger, E.A.

    1980-12-23

    An improved process for the separation of sulfur oxides from a gaseous mixture containing sulfur oxides and oxygen is disclosed. The gaseous mixture is contacted with a solid sulfur oxide acceptor comprising copper, copper oxide, or a mixture thereof dispersed on a carrier material in combination with a platinum group metal component and a component selected from the group consisting of rhenium, germanium and tin.

  13. Durable zinc oxide-containing sorbents for coal gas desulfurization

    DOEpatents

    Siriwardane, Ranjani V.

    1996-01-01

    Durable zinc-oxide containing sorbent pellets for removing hydrogen sulfide from a gas stream at an elevated temperature are made up to contain titania as a diluent, high-surface-area silica gel, and a binder. These materials are mixed, moistened, and formed into pellets, which are then dried and calcined. The resulting pellets undergo repeated cycles of sulfidation and regeneration without loss of reactivity and without mechanical degradation. Regeneration of the pellets is carried out by contacting the bed with an oxidizing gas mixture.

  14. Controlling the oxygen potential to improve the densification and the solid solution formation of uranium-plutonium mixed oxides

    NASA Astrophysics Data System (ADS)

    Berzati, Ségolène; Vaudez, Stéphane; Belin, Renaud C.; Léchelle, Jacques; Marc, Yves; Richaud, Jean-Christophe; Heintz, Jean-Marc

    2014-04-01

    Diffusion mechanisms occurring during the sintering of oxide ceramics are affected by the oxygen content of the atmosphere, as it imposes the nature and the concentration of structural defects in the material. Thus, the oxygen partial pressure, p(O2), of the sintering gas has to be precisely controlled, otherwise a large dispersion in various parameters, critical for the manufacturing of ceramics such as nuclear oxides fuels, is likely to occur. In the present work, the densification behaviour and the solid solution formation of a mixed uranium-plutonium oxide (MOX) were investigated. The initial mixture, composed of 70% UO2 + 30% PuO2, was studied at p(O2) ranging from 10-15 to 10-4 atm up to 1873 K both with dilatometry and in situ high temperature X-ray diffraction. This study has shown that the initial oxides UO2+x and PuO2-x first densify during heating and then the solid solution formation starts at about 200 K higher. The densification and the formation of the solid solution both occur at a lower temperature when p(O2) increases. Based on this result, it is possible to better define the sintering atmosphere, eventually leading to optimized parameters such as density, oxygen stoichiometry and cations homogenization of nuclear ceramics and of a wide range of industrial ceramic materials.

  15. Evaluation of the Magnesium Hydroxide Treatment Process for Stabilizing PFP Plutonium/Nitric Acid Solutions

    SciTech Connect

    Gerber, Mark A.; Schmidt, Andrew J.; Delegard, Calvin H.; Silvers, Kurt L.; Baker, Aaron B.; Gano, Susan R.; Thornton, Brenda M.

    2000-09-28

    This document summarizes an evaluation of the magnesium hydroxide [Mg(OH)2] process to be used at the Hanford Plutonium Finishing Plant (PFP) for stabilizing plutonium/nitric acid solutions to meet the goal of stabilizing the plutonium in an oxide form suitable for storage under DOE-STD-3013-99. During the treatment process, nitric acid solutions bearing plutonium nitrate are neutralized with Mg(OH)2 in an air sparge reactor. The resulting slurry, containing plutonium hydroxide, is filtered and calcined. The process evaluation included a literature review and extensive laboratory- and bench-scale testing. The testing was conducted using cerium as a surrogate for plutonium to identify and quantify the effects of key processing variables on processing time (primarily neutralization and filtration time) and calcined product properties.

  16. Plutonium radiation surrogate

    DOEpatents

    Frank, Michael I.

    2010-02-02

    A self-contained source of gamma-ray and neutron radiation suitable for use as a radiation surrogate for weapons-grade plutonium is described. The source generates a radiation spectrum similar to that of weapons-grade plutonium at 5% energy resolution between 59 and 2614 keV, but contains no special nuclear material and emits little .alpha.-particle radiation. The weapons-grade plutonium radiation surrogate also emits neutrons having fluxes commensurate with the gamma-radiation intensities employed.

  17. PLUTONIUM ELECTROREFINING CELLS

    DOEpatents

    Mullins, L.J. Jr.; Leary, J.A.; Bjorklund, C.W.; Maraman, W.J.

    1963-07-16

    Electrorefining cells for obtaining 99.98% plutonium are described. The cells consist of an impure liquid plutonium anode, a molten PuCl/sub 3/-- alkali or alkaline earth metal chloanode, a molten PuCl/sub 3/-alkali or alkaline earth metal chloride electrolyte, and a nonreactive cathode, all being contained in nonreactive ceramic containers which separate anode from cathode by a short distance and define a gap for the collection of the purified liquid plutonium deposited on the cathode. Important features of these cells are the addition of stirrer blades on the anode lead and a large cathode surface to insure a low current density. (AEC)

  18. Iron aluminide alloy container for solid oxide fuel cells

    DOEpatents

    Judkins, Roddie Reagan; Singh, Prabhakar; Sikka, Vinod Kumar

    2000-01-01

    A container for fuel cells is made from an iron aluminide alloy. The container alloy preferably includes from about 13 to about 22 weight percent Al, from about 2 to about 8 weight percent Cr, from about 0.1 to about 4 weight percent M selected from Zr and Hf, from about 0.005 to about 0.5 weight percent B or from about 0.001 to about 1 weight percent C, and the balance Fe and incidental impurities. The iron aluminide container alloy is extremely resistant to corrosion and metal loss when exposed to dual reducing and oxidizing atmospheres at elevated temperatures. The alloy is particularly useful for containment vessels for solid oxide fuel cells, as a replacement for stainless steel alloys which are currently used.

  19. Evaluation of existing United States` facilities for use as a mixed-oxide (MOX) fuel fabrication facility for plutonium disposition

    SciTech Connect

    Beard, C.A.; Buksa, J.J.; Chidester, K.; Eaton, S.L.; Motley, F.E.; Siebe, D.A.

    1995-12-31

    A number of existing US facilities were evaluated for use as a mixed-oxide fuel fabrication facility for plutonium disposition. These facilities include the Fuels Material Examination Facility (FMEF) at Hanford, the Washington Power Supply Unit 1 (WNP-1) facility at Hanford, the Barnwell Nuclear Fuel Plant (BNFP) at Barnwell, SC, the Fuel Processing Facility (FPF) at Idaho National Engineering Laboratory (INEL), the Device Assembly Facility (DAF) at the Nevada Test Site (NTS), and the P-reactor at the Savannah River Site (SRS). The study consisted of evaluating each facility in terms of available process space, available building support systems (i.e., HVAC, security systems, existing process equipment, etc.), available regional infrastructure (i.e., emergency response teams, protective force teams, available transportation routes, etc.), and ability to integrate the MOX fabrication process into the facility in an operationally-sound manner that requires a minimum amount of structural modifications.

  20. PROCESS FOR EXTRACTING NEPTUNIUM AND PLUTONIUM FROM NITRIC ACID SOLUTIONS OF SAME CONTAINING URANYL NITRATE WITH A TERTIARY AMINE

    DOEpatents

    Sheppard, J.C.

    1962-07-31

    A process of selectively extracting plutonium nitrate and neptunium nitrate with an organic solution of a tertiary amine, away from uranyl nitrate present in an aqueous solution in a maximum concentration of 1M is described. The nitric acid concentration is adjusted to about 4M and nitrous acid is added prior to extraction. (AEC)

  1. Plutonium Immobilization Canister Loading

    SciTech Connect

    Hamilton, E.L.

    1999-01-26

    This disposition of excess plutonium is determined by the Surplus Plutonium Disposition Environmental Impact Statement (SPD-EIS) being prepared by the Department of Energy. The disposition method (Known as ''can in canister'') combines cans of immobilized plutonium-ceramic disks (pucks) with vitrified high-level waste produced at the SRS Defense Waste Processing Facility (DWPF). This is intended to deter proliferation by making the plutonium unattractive for recovery or theft. The envisioned process remotely installs cans containing plutonium-ceramic pucks into storage magazines. Magazines are then remotely loaded into the DWPF canister through the canister neck with a robotic arm and locked into a storage rack inside the canister, which holds seven magazines. Finally, the canister is processed through DWPF and filled with high-level waste glass, thereby surrounding the product cans. This paper covers magazine and rack development and canister loading concepts.

  2. Divalent silver oxide cell containing cadmium and tellurium components

    SciTech Connect

    Ohya, Y.; Shimizu, T.

    1982-07-06

    A divalent silver oxide cell of the type having a positive electrode formed mainly of divalent silver oxide has its stability greatly improved by adding a cadmium component and a tellurium component and, optionally, at least one auxiliary component containing an element selected from the group consisting of thallium, mercury, lead, germanium, yttrium, tin, tungsten, lanthanum, rare earth elements, zinc, aluminum and selenium, each to either the divalent silver oxide, the alkaline electrolyte, or both. The cadmium component is contained in a concentration of not less than 0.03% by weight when in the divalent silver oxide and not less than 0.01 g/liter when in the alkaline electrolyte; and tellurium component in a concentration of not less than 0.01% by weight when in the divalent silver oxide and not less than 0.01 g/liter when in the alkaline electrolyte; and the at least one auxiliary component in a total concentration of not less than 0.01% by weight when in the divalent silver oxide and not less than 0.01 g/liter when in the alkaline electrolyte.

  3. XAF/XANES studies of plutonium-loaded sodalite/glass composite waste forms.

    SciTech Connect

    Aase, S. B.; Kropf, A. J.; Lewis, M. A.; Reed, D. T.; Richmann, M. K.

    1999-07-14

    A sodalite/glass ceramic waste form has been developed to immobilize highly radioactive nuclear wastes in chloride form, as part of an electrochemical cleanup process. Simulated waste forms have been fabricated which contain plutonium and are representative of the salt from the electrometallurgical process to recover uranium from spent nuclear fuel. X-ray absorption fine structure spectroscopy (XAFS) and x-ray absorption near-edge spectroscopy (XANES) studies were performed to determine the location, oxidation state and form of the plutonium within these waste forms. Plutonium, in the non-fission-element case, was found to segregate as plutonium(IV) oxide with a crystallite size of at least 20 nm. With fission elements present, the crystallite size was about 2 nm. No plutonium was observed within the sodalite or glass in the waste form.

  4. Zone refining of plutonium metal

    SciTech Connect

    Blau, M.S.

    1994-08-01

    The zone refining process was applied to Pu metal containing known amounts of impurities. Rod specimens of plutonium metal were melted into and contained in tantalum boats, each of which was passed horizontally through a three-turn, high-frequency coil in such a manner as to cause a narrow molten zone to pass through the Pu metal rod 10 times. The impurity elements Co, Cr, Fe, Ni, Np, U were found to move in the same direction as the molten zone as predicted by binary phase diagrams. The elements Al, Am, and Ga moved in the opposite direction of the molten zone as predicted by binary phase diagrams. As the impurity alloy was zone refined, {delta}-phase plutonium metal crystals were produced. The first few zone refining passes were more effective than each later pass because an oxide layer formed on the rod surface. There was no clear evidence of better impurity movement at the slower zone refining speed. Also, constant or variable coil power appeared to have no effect on impurity movement during a single run (10 passes). This experiment was the first step to developing a zone refining process for plutonium metal.

  5. Plutonium controversy

    SciTech Connect

    Richmond, C.R.

    1980-01-01

    The toxicity of plutonium is discussed, particularly in relation to controversies surrounding the setting of radiation protection standards. The sources, amounts of, and exposure pathways of plutonium are given and the public risk estimated. (ACR)

  6. Purification and neutron emission reduction of 238Plutonium oxide by nitrate anion exchange processing

    NASA Astrophysics Data System (ADS)

    Pansoy-Hjelvik, M. E.; Brock, J.; Nixon, J. Z.; Moniz, P.; Silver, G.; Ramsey, K. B.

    2001-02-01

    The use of ion exchange during the aqueous purification of 238Pu oxide results in low levels of uranium, thorium, and americium in the product oxide. Neutron emission rates are also reduced in the product oxide. Fluorine introduced during the dissolution of impure fuel increases the neutron emission rate of the product oxide due to the 238Pu-19F alpha/n reaction. Treating the 238Pu solution with aluminum nitrate prior to ion exchange reduces the neutron emission rate in the product oxide. Data are presented to show that neutron emission rates and concentrations of uranium, thorium, and americium are reduced by ion exchange processing. .

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

    SciTech Connect

    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.

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

    NASA Astrophysics Data System (ADS)

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

    2010-06-01

    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 233U concentration for criticality of reactor augments with the increasing of MFR for all heavy nuclides confinement and thorium & uranium confinement in PWR.

  9. Polyimides Containing Pendent Phosphine Oxide Groups for Space Applications

    NASA Technical Reports Server (NTRS)

    Thompson, C. M.; Smith, J. G., Jr.; Watson, K. A.; Connell, J. W.

    2002-01-01

    As part of an ongoing materials development activity to produce high performance polymers that are durable to the space environment, phosphine oxide containing polyimides have been under investigation. A novel dianhydride was prepared from 2,5-dihydroxyphenyldiphenylphosphine oxide in good yield. The dianhydride was reacted with commercially available diamines, and a previously reported diamine was reacted with commercially available dianhydrides to prepare isomeric polyimides. The physical and mechanical properties, particularly thermal and optical properties, of the polymers were determined. One material exhibited a high glass transition temperature, high tensile properties, and low solar absorptivity. The chemistry, physical, and mechanical properties of these resins will be discussed.

  10. PROCESS OF PRODUCING SHAPED PLUTONIUM

    DOEpatents

    Anicetti, R.J.

    1959-08-11

    A process is presented for producing and casting high purity plutonium metal in one step from plutonium tetrafluoride. The process comprises heating a mixture of the plutonium tetrafluoride with calcium while the mixture is in contact with and defined as to shape by a material obtained by firing a mixture consisting of calcium oxide and from 2 to 10% by its weight of calcium fluoride at from 1260 to 1370 deg C.

  11. Plutonium Vulnerability Management Plan

    SciTech Connect

    1995-03-01

    This Plutonium Vulnerability Management Plan describes the Department of Energy`s response to the vulnerabilities identified in the Plutonium Working Group Report which are a result of the cessation of nuclear weapons production. The responses contained in this document are only part of an overall, coordinated approach designed to enable the Department to accelerate conversion of all nuclear materials, including plutonium, to forms suitable for safe, interim storage. The overall actions being taken are discussed in detail in the Department`s Implementation Plan in response to the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 94-1. This is included as Attachment B.

  12. PLUTONIUM SEPARATION METHOD

    DOEpatents

    Beaufait, L.J. Jr.; Stevenson, F.R.; Rollefson, G.K.

    1958-11-18

    The recovery of plutonium ions from neutron irradiated uranium can be accomplished by bufferlng an aqueous solutlon of the irradiated materials containing tetravalent plutonium to a pH of 4 to 7, adding sufficient acetate to the solution to complex the uranyl present, adding ferric nitrate to form a colloid of ferric hydroxide, plutonlum, and associated fission products, removing and dissolving the colloid in aqueous nitric acid, oxldizlng the plutonium to the hexavalent state by adding permanganate or dichromate, treating the resultant solution with ferric nitrate to form a colloid of ferric hydroxide and associated fission products, and separating the colloid from the plutonlum left in solution.

  13. Remote material handling in the Plutonium Immobilization Project. Revision 1

    SciTech Connect

    Brault, J.R.

    2000-03-13

    With the downsizing of the US and Russian nuclear stockpiles, large quantities of weapons-usable plutonium in the US are being declared excess and will be disposed of by the Department of Energy Fissile Materials Disposition Program. To implement this program, DOE has selected the Savannah River Site (SRS) for the construction and operation of three new facilities: pit disassembly and conversion; mixed oxide fuel fabrication; and plutonium immobilization. The Plutonium Immobilization Project (PIP) will immobilize a portion of the excess plutonium in a hybrid ceramic and glass form containing high level waste for eventual disposal in a geologic repository. The PIP is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. Processing technology for the PIP is being developed jointly by the Lawrence Livermore National Laboratory and Westinghouse Savannah River Company. This paper will discuss development of the automated unpacking and sorting operations in the conversion area, and the automated puck and tray handling operations in the first stage immobilization area. Due to the high radiation levels and toxicity of the materials to be disposed of, the PIP will utilize automated equipment in a contained (glovebox) facility. Most operations involving plutonium-bearing materials will be performed remotely, separating personnel from the radiation source. Source term materials will be removed from the operations during maintenance. Maintenance will then be performed hands on within the containment using glove ports.

  14. Curium analysis in plutonium uranium mixed oxide by x-ray fluorescence and absorption fine structure spectroscopy.

    PubMed

    Degueldre, C; Borca, C; Cozzo, C

    2013-10-15

    Plutonium uranium mixed oxide (MOX) fuels are being used in commercial nuclear reactors. The actinides in these fuels need to be analyzed after irradiation for assessing their behaviour with regards to their environment and the coolant. In this work the study of the local occurrence, speciation and next-neighbour environment of curium (Cm) in the (Pu,U)O2 lattice within an irradiated (60 MW d kg(-1) average burn-up) MOX sample was performed employing micro-x-ray fluorescence (µ-XRF) and micro-x-ray absorption fine structure (µ-XAFS) spectroscopy. The chemical bonds, valences and stoichiometry of Cm (≈ 0.7 wt% in the rim and ≈ 0.03 wt% in the centre) are determined from the experimental data gained for the irradiated fuel material examined in its centre and peripheral (rim) zones of the fuel. Curium occurrence is also reduced from the centre (hot) to the periphery (colder) because of the condensation of these volatile oxides. In the irradiated sample Cm builds up as Cm(3+) species (>90%) within a [CmO8](13-) or [CmO7](11-) coordination environment and no (<10%) Cm(IV) can be detected in the rim zone. Curium dioxide is reduced because of the redox buffering activity of the uranium dioxide matrix and of its thermodynamic instability. PMID:24054692

  15. Oxygen plasma resistant phosphine oxide containing imide/arylene copolymers

    NASA Technical Reports Server (NTRS)

    Jensen, Brian J.

    1993-01-01

    A series of oxygen plasma resistant imide/arylene ether copolymers were prepared by reacting anhydride-terminated poly(amide acids) and amine-terminated polyarylene ethers containing phosphine oxide units. Inherent viscosities for these copolymers ranged from 0.42 to 0.80 dL/g. After curing, the resulting copolymers had glass transition temperatures ranging from 224 C to 228 C. Solution cast films of the block copolymers were tough and flexible with tensile strength, tensile moduli, and elongation at break up to 16.1 ksi, 439 ksi, and 23 percent, respectively at 25 C and 9.1 ksi, 308 ksi and 97 percent, respectively at 150 C. The copolymers show a significant improvement in resistance to oxygen plasma when compared to the commercial polyimide Kapton. The imide/arylene ether copolymers containing phosphine oxide units are suitable as coatings, films, adhesives, and composite matrices.

  16. Plutonium isotopic analysis system for plutonium samples enriched in sup 238 Pu in EP 60/61 and fuel-clad containers

    SciTech Connect

    Ruhter, W.D.

    1991-07-01

    This two-part manual describes and provides instructions for installing software for Lawrence Livermore National Laboratory's Pu-238 isotopic analysis system built for Westinghouse Hanford's Radioisotope Power Systems Facility. Part 1 contains descriptions of all the subroutines found in the main software program, WHC.ASY238. Also provided in this part are general instructions for modifying a subroutine and specific directions for relinking the WHC.ASY238 program, as well as information on the supporting program PU238.CHNG. Part 2 contains listings of the Pu-238 isotopic analysis system codes. The system uses a large (20% rel. efficiency), coaxial, n-type germanium detector (COAX). Parameter files for the detector have filenames with IS8 extensions. Spectral data files also have WH8 and I01, I02, etc. filename extensions.

  17. Wet oxidation of real coke wastewater containing high thiocyanate concentration.

    PubMed

    Oulego, Paula; Collado, Sergio; Garrido, Laura; Laca, Adriana; Rendueles, Manuel; Díaz, Mario

    2014-01-01

    Coke wastewaters, in particular those with high thiocyanate concentrations, represent an important environmental problem because of their very low biodegradability. In this work, the treatment by wet oxidation of real coke wastewaters containing concentrations of thiocyanate above 17 mM has been studied in a 1-L semi-batch reactor at temperatures between 453 and 493 K, with total oxygen pressures in the range of 2.0-8.0 MPa. A positive effect of the matrix of real coke wastewater was observed, resulting in faster thiocyanate degradation than was obtained with synthetic wastewaters. Besides, the effect of oxygen concentration and temperature on thiocyanate wet oxidation was more noticeable in real effluents than in synthetic wastewaters containing only thiocyanate. It was also observed that the degree of mineralization of the matrix organic compounds was higher when the initial thiocyanate concentration increased. Taking into account the experimental data, kinetic models were obtained, and a mechanism implying free radicals was proposed for thiocyanate oxidation in the matrix considered. In all cases, sulphate, carbonates and ammonium were identified as the main reaction products of thiocyanate wet oxidation. PMID:24269931

  18. Plutonium Finishing Plant (PFP) Final Safety Analysis Report (FSAR) [SEC 1 THRU 11

    SciTech Connect

    ULLAH, M K

    2001-02-26

    The Plutonium Finishing Plant (PFP) is located on the US Department of Energy (DOE) Hanford Site in south central Washington State. The DOE Richland Operations (DOE-RL) Project Hanford Management Contract (PHMC) is with Fluor Hanford Inc. (FH). Westinghouse Safety Management Systems (WSMS) provides management support to the PFP facility. Since 1991, the mission of the PFP has changed from plutonium material processing to preparation for decontamination and decommissioning (D and D). The PFP is in transition between its previous mission and the proposed D and D mission. The objective of the transition is to place the facility into a stable state for long-term storage of plutonium materials before final disposition of the facility. Accordingly, this update of the Final Safety Analysis Report (FSAR) reflects the current status of the buildings, equipment, and operations during this transition. The primary product of the PFP was plutonium metal in the form of 2.2-kg, cylindrical ingots called buttoms. Plutonium nitrate was one of several chemical compounds containing plutonium that were produced as an intermediate processing product. Plutonium recovery was performed at the Plutonium Reclamation Facility (PRF) and plutonium conversion (from a nitrate form to a metal form) was performed at the Remote Mechanical C (RMC) Line as the primary processes. Plutonium oxide was also produced at the Remote Mechanical A (RMA) Line. Plutonium processed at the PFP contained both weapons-grade and fuels-grade plutonium materials. The capability existed to process both weapons-grade and fuels-grade material through the PRF and only weapons-grade material through the RMC Line although fuels-grade material was processed through the line before 1984. Amounts of these materials exist in storage throughout the facility in various residual forms left from previous years of operations.

  19. Actinide Oxidation State and O/M Ratio in Hypostoichiometric Uranium-Plutonium-Americium U0.750Pu0.246Am0.004O2-x Mixed Oxides.

    PubMed

    Vauchy, Romain; Belin, Renaud C; Robisson, Anne-Charlotte; Lebreton, Florent; Aufore, Laurence; Scheinost, Andreas C; Martin, Philippe M

    2016-03-01

    Innovative americium-bearing uranium-plutonium mixed oxides U1-yPuyO2-x are envisioned as nuclear fuel for sodium-cooled fast neutron reactors (SFRs). The oxygen-to-metal (O/M) ratio, directly related to the oxidation state of cations, affects many of the fuel properties. Thus, a thorough knowledge of its variation with the sintering conditions is essential. The aim of this work is to follow the oxidation state of uranium, plutonium, and americium, and so the O/M ratio, in U0.750Pu0.246Am0.004O2-x samples sintered for 4 h at 2023 K in various Ar + 5% H2 + z vpm H2O (z = ∼ 15, ∼ 90, and ∼ 200) gas mixtures. The O/M ratios were determined by gravimetry, XAS, and XRD and evidenced a partial oxidation of the samples at room temperature. Finally, by comparing XANES and EXAFS results to that of a previous study, we demonstrate that the presence of uranium does not influence the interactions between americium and plutonium and that the differences in the O/M ratio between the investigated conditions is controlled by the reduction of plutonium. We also discuss the role of the homogeneity of cation distribution, as determined by EPMA, on the mechanisms involved in the reduction process. PMID:26907589

  20. PREPARATION OF HALIDES OF PLUTONIUM

    DOEpatents

    Garner, C.S.; Johns, I.B.

    1958-09-01

    A dry chemical method is described for preparing plutonium halides, which consists in contacting plutonyl nitrate with dry gaseous HCl or HF at an elevated temperature. The addition to the reaction gas of a small quantity of an oxidizing gas or a reducing gas will cause formation of the tetra- or tri-halide of plutonium as desired.

  1. Plutonium and americium separation from salts

    DOEpatents

    Hagan, Paul G.; Miner, Frend J.

    1976-01-01

    Salts or materials containing plutonium and americium are dissolved in hydrochloric acid, heated, and contacted with an alkali metal carbonate solution to precipitate plutonium and americium carbonates which are thereafter readily separable from the solution.

  2. Technical report for the generic site add-on facility for plutonium polishing

    SciTech Connect

    Collins, E. D.

    1998-06-01

    The purpose of this report is to provide environmental data and reference process information associated with incorporating plutonium polishing steps (dissolution, impurity removal, and conversion to oxide powder) into the genetic-site Mixed-Oxide Fuel Fabrication Facility (MOXFF). The incorporation of the plutonium polishing steps will enable the removal of undesirable impurities, such as gallium and americium, known to be associated with the plutonium. Moreover, unanticipated impurities can be removed, including those that may be contained in (1) poorly characterized feed materials, (2) corrosion products added from processing equipment, and (3) miscellaneous materials contained in scrap recycle streams. These impurities will be removed to the extent necessary to meet plutonium product purity specifications for MOX fuels. Incorporation of the plutonium polishing steps will mean that the Pit Disassembly and Conversion Facility (PDCF) will need to produce a plutonium product that can b e dissolved at the MOXFF in nitric acid at a suitable rate (sufficient to meet overall production requirements) with the minimal usage of hydrofluoric acid, and its complexing agent, aluminum nitrate. This function will require that if the PDCF product is plutonium oxide powder, that powder must be produced, stored, and shipped without exceeding a temperature of 600 C.

  3. Conductive Circuit Containing a Polymer Composition Containing Thermally Exfoliated Graphite Oxide and Method of Making the Same

    NASA Technical Reports Server (NTRS)

    Prud'Homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor)

    2014-01-01

    A conductive circuit containing a polymer composite, which contains at least one polymer and a modified graphite oxide material, containing thermally exfoliated graphite oxide having a surface area of from about 300 m(sup.2)/g to 2600 m(sup.2)/g, and a method of making the same.

  4. A comparison of long-term retention of plutonium oxide in lung based on excretion data with observed lung burdens at autopsy

    SciTech Connect

    Heid, K.R.; Jech, J.J.; Kathren, R.L.; Sula, M.J.

    1985-05-01

    Experience in the nuclear industry has shown that the most frequent route for significant plutonium deposition in man is by inhalation. After an accidental inhalation exposure to plutonium oxide has occurred, an attempt is usually made to determine the lung burden using in-vivo measurement techniques. However, this will not be possible if the amount deposited is less than the detection capability of the counter. Likewise, it may not be possible for even larger intakes if they are first discovered as a results of a routine examination after much of the plutonium has transferred out of the lungs. Under these conditions, the initial lung burden and dose assessment may, of necessity, be based entirely on excretion date. This paper discusses three plutonium inhalation cases for which autopsy data are available to demonstrate the difficulties and uncertainties of such assessments and to emphasize the need for enhanced routine surveillance of the worker and work place to assure that intakes are detected at the time they occur. An in-depth review of the findings for these three cases suggests that the use of urine excretion data applied to ICRP lung model assumption will result in large uncertainties in estimates of the pulmonary burden and lung dose. These estimates will be improved if the material inhaled can be better characterized and by using clearance half times developed for the individual rather than the default values for Class Y material recommended by the ICRP.

  5. Thermal-mechanical performance modeling of thorium-plutonium oxide fuel and comparison with on-line irradiation data

    NASA Astrophysics Data System (ADS)

    Insulander Björk, Klara; Kekkonen, Laura

    2015-12-01

    Thorium-plutonium Mixed OXide (Th-MOX) fuel is considered for use in light water reactors fuel due to some inherent benefits over conventional fuel types in terms of neutronic properties. The good material properties of ThO2 also suggest benefits in terms of thermal-mechanical fuel performance, but the use of Th-MOX fuel for commercial power production demands that its thermal-mechanical behavior can be accurately predicted using a well validated fuel performance code. Given the scant operational experience with Th-MOX fuel, no such code is available today. This article describes the first phase of the development of such a code, based on the well-established code FRAPCON 3.4, and in particular the correlations reviewed and chosen for the fuel material properties. The results of fuel temperature calculations with the code in its current state of development are shown and compared with data from a Th-MOX test irradiation campaign which is underway in the Halden research reactor. The results are good for fresh fuel, whereas experimental complications make it difficult to judge the adequacy of the code for simulations of irradiated fuel.

  6. A Calibration to Predict the Concentrations of Impurities in Plutonium Oxide by Prompt Gamma Analysis Revision 2

    SciTech Connect

    Narlesky, Joshua Edward; Kelly, Elizabeth J.

    2015-09-10

    This report documents the new PG calibration regression equation. These calibration equations incorporate new data that have become available since revision 1 of “A Calibration to Predict the Concentrations of Impurities in Plutonium Oxide by Prompt Gamma Analysis” was issued [3] The calibration equations are based on a weighted least squares (WLS) approach for the regression. The WLS method gives each data point its proper amount of influence over the parameter estimates. This gives two big advantages, more precise parameter estimates and better and more defensible estimates of uncertainties. The WLS approach makes sense both statistically and experimentally because the variances increase with concentration, and there are physical reasons that the higher measurements are less reliable and should be less influential. The new magnesium calibration includes a correction for sodium and separate calibration equation for items with and without chlorine. These additional calibration equations allow for better predictions and smaller uncertainties for sodium in materials with and without chlorine. Chlorine and sodium have separate equations for RICH materials. Again, these equations give better predictions and smaller uncertainties chlorine and sodium for RICH materials.

  7. Interspecies comparison of the metabolism and dosimetry of inhaled mixed oxides of plutonium and uranium

    SciTech Connect

    Boecker, B.B.; Mewhinney, J.A.; Eidson, A.F.

    1997-12-01

    Three studies were conducted to provide information on the biological fate, distribution of radiation doses among tissues, and implications for potential health consequences of an inhalation exposure to mixed-oxide nuclear fuel materials. In each study, Fischer-344 rats, beagle dogs, and cynomolgus monkeys inhaled one of three aerosols: 750{degrees}C calcined mixed oxides of UO{sub 2} and PuO{sub 2}, 1750{degrees}C sintered (U,Pu)O{sub 2}, or 850{degrees}C calcined {open_quotes}pure{close_quotes} PuO{sub 2}. These materials were collected from glove-box enclosures immediately after industrial processing of mixed-oxide fuel materials. Lung retention, tissue distribution, and mode of excretion of {sup 238-240}Pu, {sup 241}Am, and uranium (when present) were quantified by radiochemical analysis of tissue and excreta samples from animals sacrificed at selected times to 6.5 yr after inhalation exposure.

  8. Surprising Coordination for Plutonium in the First Plutonium (III) Borate

    SciTech Connect

    Wang, Shuao; Alekseev, Evgeny V.; Depmeier, Wulf; Albrecht-Schmitt, Thomas E.

    2011-02-22

    The first plutonium(III) borate, Pu2[B12O18(OH)4Br2(H2O)3]·0.5H2O, has been prepared by reacting plutonium(III) with molten boric acid under strictly anaerobic conditions. This compound contains a three-dimensional polyborate network with triangular holes that house the plutonium(III) sites. The plutonium sites in this compound are 9- and 10-coordinate and display atypical geometries.

  9. Reactivation of a tin oxide-containing catalyst

    NASA Technical Reports Server (NTRS)

    Upchurch, Billy T. (Inventor); Miller, Irvin M. (Inventor); Brown, Kenneth G. (Inventor); Hess, Robert V. (Inventor); Schryer, David R. (Inventor); Sidney, Barry D. (Inventor); Wood, George M. (Inventor); Paulin, Patricia A. (Inventor)

    1989-01-01

    A method for the reactivation of a tin oxide-containing catalyst of a CO.sub.2 laser is provided. First, the catalyst is pretreated by a standard procedure. When the catalyst experiences diminished activity during usage, the heated zone surrounding the catalyst is raised to a temperature which is the operating temperature of the laser and 400.degree. C. for approximately one hour. The catalyst is exposed to the same laser gas mixture during this period. The temperature of the heated zone is then lowered to the operating temperature of the CO.sub.2 laser.

  10. All-alkoxide synthesis of strontium-containing metal oxides

    DOEpatents

    Boyle, Timothy J.

    2001-01-01

    A method for making strontium-containing metal-oxide ceramic thin films from a precursor liquid by mixing a strontium neo-pentoxide dissolved in an amine solvent and at least one metal alkoxide dissolved in a solvent, said at least one metal alkoxide selected from the group consisting of alkoxides of calcium, barium, bismuth, cadmium, lead, titanium, tantalum, hafnium, tungsten, niobium, zirconium, yttrium, lanthanum, antimony, chromium and thallium, depositing a thin film of the precursor liquid on a substrate, and heating the thin film in the presence of oxygen at between 550 and 700.degree. C.

  11. Plutonium Finishing Plant. Interim plutonium stabilization engineering study

    SciTech Connect

    Sevigny, G.J.; Gallucci, R.H.; Garrett, S.M.K.; Geeting, J.G.H.; Goheen, R.S.; Molton, P.M.; Templeton, K.J.; Villegas, A.J.; Nass, R.

    1995-08-01

    This report provides the results of an engineering study that evaluated the available technologies for stabilizing the plutonium stored at the Plutonium Finishing Plant located at the hanford Site in southeastern Washington. Further processing of the plutonium may be required to prepare the plutonium for interim (<50 years) storage. Specifically this document provides the current plutonium inventory and characterization, the initial screening process, and the process descriptions and flowsheets of the technologies that passed the initial screening. The conclusions and recommendations also are provided. The information contained in this report will be used to assist in the preparation of the environmental impact statement and to help decision makers determine which is the preferred technology to process the plutonium for interim storage.

  12. ADSORPTION-BISMUTH PHOSPHATE METHOD FOR SEPARATING PLUTONIUM

    DOEpatents

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

    1960-06-28

    A process is given for separating plutonium from uranium and fission products. Plutonium and uranium are adsorbed by a cation exchange resin, plutonium is eluted from the adsorbent, and then, after oxidation to the hexavalent state, the plutonium is contacted with a bismuth phosphate carrier precipitate.

  13. In vitro dissolution of respirable aerosols of industrial uranium and plutonium mixed-oxide nuclear fuels.

    PubMed

    Eidson, A F; Mewhinney, J A

    1983-12-01

    Dissolution characteristics of mixed-oxide nuclear fuels are important considerations for prediction of biological behavior of inhaled particles. Four representative industrial mixed-oxide powders were obtained from fuel fabrication enclosures. Studies of the dissolution of Pu, Am and U from aerosol particles of these materials in a serum simulant solution and in 0.1M HCl showed: (1) dissolution occurred at a rapid rate initially and slowed at longer times, (2) greater percentages of U dissolved than Pu or Am: with the dissolution rates of U and Pu generally reflecting the physical nature of the UO2-PuO2 matrix, (3) the temperature history of industrial mixed-oxides could not be reliably related to Pu dissolution except for a 3-5% increase when incorporated into a solid solution by sintering at 1750 degrees C, and (4) dissolution in the serum simulant agreed with the in vivo UO2 dissolution rate and suggested the dominant role of mechanical processes in PuO2 clearance from the lung. The rapid initial dissolution rate was shown to be related, in part, to an altered surface layer. The advantages and uses of in vitro solubility data for estimation of biological behavior of inhaled industrial mixed oxides, such as assessing the use of chelation therapy and interpretation of urinary excretion data, are discussed. It was concluded that in vitro solubility tests were useful, simple and easily applied to individual materials potentially inhaled by humans. PMID:6643070

  14. Oxide Film Aging on Alloy 22 in Halide Containing Solutions

    SciTech Connect

    Rodriguez, Martin A.; Carranza, Ricardo M.; Rebak, Raul B.

    2007-07-01

    Passive and corrosion behaviors of Alloy 22 in chloride and fluoride containing solutions, changing the heat treatment of the alloy, the halide concentration and the pH of the solutions at 90 deg. C, was investigated. The study was implemented using electrochemical techniques, which included open circuit potential monitoring over time, potentiodynamic polarization and electrochemical impedance spectroscopy (EIS) measurements carried out at open circuit and at passivity potentials. Corrosion rates obtained by EIS measurements after 24 h immersion in naturally aerated solutions were below 0.5 {mu}m/year. The corrosion rates were practically independent of solution pH, alloy heat treatment and halide ion nature and concentration. EIS low frequency resistance values increased with applied potential in the passive domain and with polarization time in pH 6 - 1 M NaCl at 90 deg. C. This effect was attributed to an increase in the oxide film thickness and oxide film aging. High frequency capacitance measurements indicated that passive oxide on Alloy 22 presented a double n-type/p-type semiconductor behavior in the passive potential range. (authors)

  15. Long-term criticality concerns associated with disposition of weapons plutonium

    SciTech Connect

    Choi, J.S.

    1997-06-01

    This paper presents a very brief description of criticality concerns resulting from dismantlement of nuclear weapons. Two plutonium disposal options, and associated criticality considerations, are described: (1) irradiating it into reactor-grade spent fuel, and (2) immobilization and burial, either in a geologic repository or in deep, sealed boreholes. Mixed oxide spent fuel could contain 3 to 4 wt% of reactor-grade plutonium. For the immobilization and the deep borehole options to be economically viable, a plutonium content of 3 to 7 wt% would be required. A study is proposed to evaluate the long-term criticality safety concerns for disposition of fissionable material in a geologic setting. 2 refs.

  16. SOLVENT EXTRACTION PROCESS FOR PLUTONIUM

    DOEpatents

    Anderson, H.H.; Asprey, L.B.

    1960-02-01

    A process of separating plutonium in at least the tetravalent state from fission products contained in an aqueous acidic solution by extraction with alkyl phosphate is reported. The plutonium can then be back-extracted from the organic phase by contact with an aqueous solution of sulfuric, phosphoric, or oxalic acid as a complexing agent.

  17. PLUTONIUM-CERIUM-COPPER ALLOYS

    DOEpatents

    Coffinberry, A.S.

    1959-05-12

    A low melting point plutonium alloy useful as fuel is a homogeneous liquid metal fueled nuclear reactor is described. Vessels of tungsten or tantalum are useful to contain the alloy which consists essentially of from 10 to 30 atomic per cent copper and the balance plutonium and cerium. with the plutontum not in excess of 50 atomic per cent.

  18. PLUTONIUM-URANIUM-TITANIUM ALLOYS

    DOEpatents

    Coffinberry, A.S.

    1959-07-28

    A plutonium-uranium alloy suitable for use as the fuel element in a fast breeder reactor is described. The alloy contains from 15 to 60 at.% titanium with the remainder uranium and plutonium in a specific ratio, thereby limiting the undesirable zeta phase and rendering the alloy relatively resistant to corrosion and giving it the essential characteristic of good mechanical workability.

  19. Process development testing in support of the plutonium immobilization program

    SciTech Connect

    Herman, C; Ebbinghaus, B

    2000-02-11

    As an integral part of the plutonium disposition program, formulation and process development is being performed for the immobilization of surplus plutonium in a titanate-based ceramic. Small-scale process prototypic and lab-scale functionally prototypic equipment have been tested to help define the immobilization process. The testing has included non-radioactive surrogates and actual actinide oxides contained in the immobilized form. A summary of the process development studies, as well as the formulation studies relevant to the process, will be provided.

  20. Plutonium in Concentrated Solutions

    SciTech Connect

    Clark, Sue B.; Delegard, Calvin H.

    2002-08-01

    Complex, high ionic strength media are used throughout the plutonium cycle, from its processing and purification in nitric acid, to waste storage and processing in alkaline solutions of concentrated electrolytes, to geologic disposal in brines. Plutonium oxidation/reduction, stability, radiolysis, solution and solid phase chemistry have been studied in such systems. In some cases, predictive models for describing Pu chemistry under such non-ideal conditions have been developed, which are usually based on empirical databases describing specific ion interactions. In Chapter 11, Non-Ideal Systems, studies on the behavior of Pu in various complex media and available model descriptions are reviewed.

  1. Plutonium Immobilization Task 5.6 Metal Conversion: Milestone Report - Perform Feasibility Demonstrations on Pu-Al Alloys

    SciTech Connect

    Zundelevich, Y; Kerns, J; Bannochie, C

    2001-04-12

    The Plutonium Conversion Task within the Plutonium Immobilization Program (PIP) transforms incoming plutonium (Pu) feed materials into an oxide acceptable for blending with ceramic precursors. One of the feed materials originally planned for PIP was unirradiated fuel, which consisted mainly of the Zero Power Plutonium Reactor (ZPPR) fuel. Approximately 3.5 metric tons of Pu is in ZPPR fuel. The ZPPR fuel is currently stored at the Argonne National Laboratory-West as stainless steel clad metal plates and oxide pellets, with the vast majority of the Pu in the metal plates. The metal plates consist of a Pu-U-Mo alloy (containing 90% of the ZPPR plutonium metal) and a Pu-Al alloy (containing 10% of the ZPPR plutonium metal). The Department of Energy (DOE) decided that ZPPR fuel is a national asset and, therefore, not subject to disposition. This report documents work done prior to that decision. The Hydnde-Oxidation (HYDOX) Process was selected as the method for Metal Conversion in PIP because it provides a universal means for preparing oxide from all feed materials. HYDOX incorporates both the hydride process, originally developed to separate Pu from other pit materials, as well as the oxide formation step. Plutonium hydride is very reactive and is readily converted to either the nitride or the oxide. A previous feasibility study demonstrated that the Pu-U-Mo alloy could be successfully converted to oxide via the HYDOX Process. Another Metal Conversion milestone was to demonstrate the feasibility of the HYDOX Process for converting plutonium-aluminum (Pu-Al) alloy in ZPPR fuel plates to an acceptable oxide. This report documents the results of the latter feasibility study which was performed before the DOE decision to retain ZPPR fuel rather than immobilize it.

  2. Work and disproportionation for aqueous plutonium.

    PubMed

    Silver, G L

    2003-10-01

    The relation of two plutonium work integrals has recently been illustrated. One of the integrals applies to the work of disproportionation of tetravalent plutonium in 1 M acid and the other to the work of oxidation of plutonium from the trivalent to a higher oxidation state. This paper generalizes the disproportionation work integral so that it can be applied to tetravalent plutonium at any acid concentration. An equation is provided that can be used to verify work estimations obtained by integration. It applies to oxidation and disproportionation processes and it is easy to use. PMID:14522227

  3. Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

    DOEpatents

    Lloyd, Milton H.

    1983-01-01

    Method for direct coprocessing of nuclear fuels derived from a product stream of a fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

  4. Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

    DOEpatents

    Lloyd, M.H.

    1981-01-09

    Method for direct coprocessing of nuclear fuels derived from a product stream of fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

  5. Influence of oxidation states on plutonium mobility during long-term transport through an unsaturated subsurface environment.

    PubMed

    Kaplan, Daniel I; Powell, Brian A; Demirkanli, Deniz I; Fjeld, Robert A; Molz, Fred J; Serkiz, Steven M; Coates, John T

    2004-10-01

    Lysimeter and laboratory studies were conducted to identify the controlling chemical processes influencing Pu(IV) mobility through the vadose zone. A 52-L lysimeter containing sediment from the Savannah River Site, South Carolina and solid PuIV(NO3)4 was left exposed to natural wetting and drying cycles for 11 years before the lysimeter sediment was sampled. Pu had traveled 10 cm, with >95% of the Pu remaining within 1.25 cm of the source. Laboratory studies showed that the sediment quickly reduced Pu(V) to Pu(IV) (the pseudo-first-order reduction rate constant, Kobs, was 0.11 h(-1)). Of particular interest was that this same sediment could be induced to release very low concentrations of sorbed Pu under oxidizing conditions, presumably by oxidation of sorbed Pu(IV) to the more mobile Pu(V) species. Transport modeling supported the postulation that Pu oxidation occurred in the lysimeter sediment; the inclusion of an oxidation term in the model produced simulations that capture the Pu depth profile data. By not including the oxidation process in the model, Pu mobility was grossly underestimated by a factor of 3.5. It is concluded that both oxidation and reduction mechanisms can play an important role in Pu transportthrough the vadose zone and should be considered when evaluating disposal of Pu-bearing wastes. PMID:15506198

  6. Americium and plutonium release behavior from irradiated mixed oxide fuel during heating

    NASA Astrophysics Data System (ADS)

    Sato, I.; Suto, M.; Miwa, S.; Hirosawa, T.; Koyama, S.

    2013-06-01

    The release behavior of Pu and Am was investigated under the reducing atmosphere expected in sodium cooled fast reactor severe accidents. Irradiated Pu and U mixed oxide fuels were heated at maximum temperatures of 2773 K and 3273 K. EPMA, γ-ray spectrometry and α-ray spectrometry for released and residual materials revealed that Pu and Am can be released more easily than U under the reducing atmosphere. The respective release rate coefficients for Pu and Am were obtained as 3.11 × 10-4 min-1 and 1.60 × 10-4 min-1 at 2773 K under the reducing atmosphere with oxygen partial pressure less than 0.02 Pa. Results of thermochemical calculations indicated that the main released chemical forms would likely be PuO for Pu and Am for Am under quite low oxygen partial pressure.

  7. Neutronic Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997 - Volume 4, Part 2--Saxton Plutonium Program Critical Experiments

    SciTech Connect

    Abdurrahman, NM

    2000-10-12

    Critical experiments with water-moderated, single-region PuO{sub 2}-UO{sub 2} or UO{sub 2}, and multiple-region PuO{sub 2}-UO{sub 2}- and UO{sub 2}-fueled cores were performed at the CRX reactor critical facility at the Westinghouse Reactor Evaluation Center (WREC) at Waltz Mill, Pennsylvania in 1965 [1]. These critical experiments were part of the Saxton Plutonium Program. The mixed oxide (MOX) fuel used in these critical experiments and then loaded in the Saxton reactor contained 6.6 wt% PuO{sub 2} in a mixture of PuO{sub 2} and natural UO{sub 2}. The Pu metal had the following isotopic mass percentages: 90.50% {sup 239}Pu; 8.57% {sup 239}Pu; 0.89% {sup 240}Pu; and 0.04% {sup 241}Pu. The purpose of these critical experiments was to verify the nuclear design of Saxton partial plutonium cores while obtaining parameters of fundamental significance such as buckling, control rod worth, soluble poison worth, flux, power peaking, relative pin power, and power sharing factors of MOX and UO{sub 2} lattices. For comparison purposes, the core was also loaded with uranium dioxide fuel rods only. This series is covered by experiments beginning with the designation SX.

  8. The hydriding resistance of plutonium oxides and mononitride: A view from ab initio molecular dynamics

    NASA Astrophysics Data System (ADS)

    Sun, Bo; Liu, Haifeng; Song, Haifeng

    2015-06-01

    Based on the non-local van der Waals density functional (vdW-DF)+ U scheme, we carry out the ab initio molecular dynamics study of the interaction dynamics for H2 molecules impingement against Pu-oxides and mononitride surfaces. We show that except for the weak physisorption, both PuO2 and PuN surfaces are so difficult of access that almost all of H2 molecules will bounce back to the vacuum when their initial kinetic energies are not sufficient. Although the dissociative adsorption of H2 on PuO2 surfaces is found to be exothermic, the collision-induced dissociation barriers of H2 are very high (up to 2.2 eV). However, PuO2 overlayer on Pu-metal can be reduced to α-Pu2O3 drived by oxygen-lean conditions, and H2 can penetrate and diffuse in α-Pu2O3 easily. As a result, α-Pu2O3 can promote the hydriding process of Pu. Unlike PuO2, PuN is found to be one kind of stable and uniform passivation layer against Pu-hydriding. Specifically, the incorporation of PuN and H-atom is proven to be thermodynamically unstable. Overall, our current study reveals the mechanical and chemical resistances of Pu-oxide and Pu-nitride to hydrogen corrosion, which have strong implications to the understanding of the surface corrosion and passivation of Pu metal. This work was supported by the FDST of CAEP under Grant No. 9090707.

  9. Sorption Properties of Halogen Containing Graphene Oxide Frameworks

    NASA Astrophysics Data System (ADS)

    Burress, Jacob; Baker, Elizabeth; Bethea, Donald; Frangos, Katherine

    Physisorption of gases has applications in gas storage (e.g. methane, hydrogen for vehicles) and gas separation (carbon dioxide from flue gas). The van der Waals force in narrow pores is strong enough to condense even supercritical gases to much higher densities. Additionally, differences in the binding energy between different gases and the sorbent surface are sufficient to for gas separations. Beyond adsorption interactions, simple steric (size, shape) effects also play a role in gas separations. One class of materials currently being investigated for numerous gas storage/separation applications is graphene oxide frameworks (GOFs). GOFs consist of layers of graphene/graphene oxide separated by chemical linkers covalently bonded on both sides. This presentation will give results from boronic acid-based GOFs that contain halogen group elements. Effects of different linkers on pore shape will be presented. Physical behavior of the gases investigated (hydrogen, methane, carbon dioxide, nitrogen), including binding energies and steric effects for gas separation will also be presented. The physics mechanism behind pore breathing (expansion and contraction of pore volume) in these materials will be discussed.

  10. Plutonium inventory characterization technical evaluation report

    SciTech Connect

    Wittman, G.R., Westinghouse Hanford

    1996-07-10

    This is a technical report on the data, gathered to date, under WHC- SD-CP-TP-086, Rev. 1, on the integrity of the food pack cans currently being used to store plutonium or plutonium compounds at the Plutonium Finishing Plant. Workplan PFP-96-VO-009, `Inspection of Special Nuclear Material Using X-ray`, was used to gather data on material and containment conditions using real time radiography. Some of those images are included herein. A matrix found in the `Plutonium Inventory Characterization Implementation Plan` was used to categorize different plutonium items based upon the type of material being stored and the life expectancy of the containers.

  11. Air transport of plutonium metal: content expansion initiative for the plutonium air transportable (PAT01) packaging

    SciTech Connect

    Caviness, Michael L; Mann, Paul T

    2010-01-01

    The National Nuclear Security Administration (NNSA) has submitted an application to the Nuclear Regulatory Commission (NRC) for the air shipment of plutonium metal within the Plutonium Air Transportable (PAT-1) packaging. The PAT-1 packaging is currently authorized for the air transport of plutonium oxide in solid form only. The INMM presentation will provide a limited overview of the scope of the plutonium metal initiative and provide a status of the NNSA application to the NRC.

  12. Air transport of plutonium metal : content expansion initiative for the Plutonium Air Transportable (PAT-1) packaging.

    SciTech Connect

    Mann, Paul T.; Caviness, Michael L.; Yoshimura, Richard Hiroyuki

    2010-06-01

    The National Nuclear Security Administration (NNSA) has submitted an application to the Nuclear Regulatory Commission (NRC) for the air shipment of plutonium metal within the Plutonium Air Transportable (PAT-1) packaging. The PAT-1 packaging is currently authorized for the air transport of plutonium oxide in solid form only. The INMM presentation will provide a limited overview of the scope of the plutonium metal initiative and provide a status of the NNSA application to the NRC.

  13. Performance of Thorium-Based Mixed Oxide Fuels for the Consumption of Plutonium and Minor Actinides in Current and Advanced Reactors

    SciTech Connect

    Weaver, Kevan Dean; Herring, James Stephen

    2002-06-01

    A renewed interest in thorium-based fuels has arisen lately based on the need for proliferation resistance, longer fuel cycles, higher burnup and improved wasteform characteristics. Recent studies have been directed toward homogeneously mixed, heterogeneously mixed, and seed-and-blanket thorium-uranium fuel cycles that rely on "in situ" use of the bred-in U-233. However, due to the higher initial enrichment required to achieve acceptable burnups, these fuels are encountering economic constraints. Thorium can nevertheless play a large role in the nuclear fuel cycle; particularly in the reduction of plutonium. While uranium-based mixedoxide (MOX) fuel will decrease the amount of plutonium, the reduction is limited due to the breeding of more plutonium (and higher actinides) from the U-238. Here we present calculational results and a comparison of the potential burnup of a thorium-based and uranium-based mixed oxide fuel in a light water reactor (LWR). Although the uranium-based fuels outperformed the thorium-based fuels in achievable burnup, a depletion comparison of the initially charged plutonium (both reactor and weapons grade) showed that the thorium-based fuels outperformed the uranium-based fuels by more that a factor of 2; where more than 70% of the total plutonium in the thorium-based fuel is consumed during the cycle. This is significant considering that the achievable burnup of the thorium-based fuels were 1.4 to 4.6 times less than the uranium-based fuels. Furthermore, use of a thorium-based fuel could also be used as a strategy for reducing the amount of long-lived nuclides (including the minor actinides), and thus the radiotoxicity in spent nuclear fuel. Although the breeding of U-233 is a concern, the presence of U-232 and its daughter products can aid in making this fuel self-protecting, and/or enough U-238 can be added to denature the fissile uranium. From these calculations, it appears that thorium-based fuel for plutonium incineration is superior as

  14. Plutonium immobilization feed batching system concept report

    SciTech Connect

    Erickson, S.

    2000-07-19

    The Plutonium Immobilization Facility will encapsulate plutonium in ceramic pucks and seal the pucks inside welded cans. Remote equipment will place these cans in magazines and the magazines in a Defense Waste Processing Facility (DWPF) canister. The DWPF will fill the canister with high level waste glass for permanent storage. Feed batching is one of the first process steps involved with first stage plutonium immobilization. It will blend plutonium oxide powder before it is combined with other materials to make pucks. This report discusses the Plutonium Immobilization feed batching process preliminary concept, batch splitting concepts, and includes a process block diagram, concept descriptions, a preliminary equipment list, and feed batching development areas.

  15. First-principles local density approximation + U and generalized gradient approximation + U study of plutonium oxides.

    PubMed

    Sun, Bo; Zhang, Ping; Zhao, Xian-Geng

    2008-02-28

    The electronic structure and properties of PuO2 and Pu2O3 have been studied from first principles by the all-electron projector-augmented-wave method. The local density approximation+U and the generalized gradient approximation+U formalisms have been used to account for the strong on-site Coulomb repulsion among the localized Pu 5f electrons. We discuss how the properties of PuO2 and Pu2O3 are affected by the choice of U as well as the choice of exchange-correlation potential. Also, oxidation reaction of Pu2O3, leading to formation of PuO2, and its dependence on U and exchange-correlation potential have been studied. Our results show that by choosing an appropriate U, it is promising to correctly and consistently describe structural, electronic, and thermodynamic properties of PuO2 and Pu2O3, which enable the modeling of redox process involving Pu-based materials possible. PMID:18315070

  16. Understanding the interactions of neptunium and plutonium ions with graphene oxide: scalar-relativistic DFT investigations.

    PubMed

    Wu, Qun-Yan; Lan, Jian-Hui; Wang, Cong-Zhi; Zhao, Yu-Liang; Chai, Zhi-Fang; Shi, Wei-Qun

    2014-11-01

    Due to the vast application potential of graphene oxide (GO)-based materials in nuclear waste processing, it is of pivotal importance to investigate the interaction mechanisms between actinide cations such as Np(V) and Pu(IV, VI) ions and GO. In this work, we have considered four types of GOs modified by hydroxyl, carboxyl, and carbonyl groups at the edge and epoxy group on the surface, respectively. The structures, bonding nature, and binding energies of Np(V) and Pu(IV, VI) complexes with GOs have been investigated systematically using scalar-relativistic density functional theory (DFT). Geometries and harmonic frequencies suggest that Pu(IV) ions coordinate more easily with GOs compared to Np(V) and Pu(VI) ions. NBO and electron density analyses reveal that the coordination bond between Pu(IV) ions and GO possesses more covalency, whereas for Np(V) and Pu(VI) ions electrostatic interaction dominates the An-OG bond. The binding energies in aqueous solution reveal that the adsorption abilities of all GOs for actinide ions follow the order of Pu(IV) > Pu(VI) > Np(V), which is in excellent agreement with experimental observations. It is expected that this study can provide useful information for developing more efficient GO-based materials for radioactive wastewater treatment. PMID:25302669

  17. Plutonium disproportionation: the relation of work integrals.

    PubMed

    Silver, G L

    2003-04-01

    Two plutonium work integrals have been demonstrated in recent years. One of them applies to the work of disproportionation and the other to the work of oxidation of plutonium from the trivalent to a higher oxidation state. This paper illustrates the connection of the integrals by an example and a diagram. PMID:12672623

  18. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    SciTech Connect

    Wishau, R.

    1998-05-01

    Molten salt oxidation (MSO) is proposed as a {sup 238}Pu waste treatment technology that should be developed for volume reduction and recovery of {sup 238}Pu and as an alternative to the transport and permanent disposal of {sup 238}Pu waste to the WIPP repository. In MSO technology, molten sodium carbonate salt at 800--900 C in a reaction vessel acts as a reaction media for wastes. The waste material is destroyed when injected into the molten salt, creating harmless carbon dioxide and steam and a small amount of ash in the spent salt. The spent salt can be treated using aqueous separation methods to reuse the salt and to recover 99.9% of the precious {sup 238}Pu that was in the waste. Tests of MSO technology have shown that the volume of combustible TRU waste can be reduced by a factor of at least twenty. Using this factor the present inventory of 574 TRU drums of {sup 238}Pu contaminated wastes is reduced to 30 drums. Further {sup 238}Pu waste costs of $22 million are avoided from not having to repackage 312 of the 574 drums to a drum total of more than 4,600 drums. MSO combined with aqueous processing of salts will recover approximately 1.7 kilograms of precious {sup 238}Pu valued at 4 million dollars (at $2,500/gram). Thus, installation and use of MSO technology at LANL will result in significant cost savings compared to present plans to transport and dispose {sup 238}Pu TRU waste to the WIPP site. Using a total net present value cost for the MSO project as $4.09 million over a five-year lifetime, the project can pay for itself after either recovery of 1.6 kg of Pu or through volume reduction of 818 drums or a combination of the two. These savings show a positive return on investment.

  19. Computation Results from a Parametric Study to Determine Bounding Critical Systems of Homogeneously Water-Moderated Mixed Plutonium--Uranium Oxides

    SciTech Connect

    Shimizu, Y.

    2001-01-11

    This report provides computational results of an extensive study to examine the following: (1) infinite media neutron-multiplication factors; (2) material bucklings; (3) bounding infinite media critical concentrations; (4) bounding finite critical dimensions of water-reflected and homogeneously water-moderated one-dimensional systems (i.e., spheres, cylinders of infinite length, and slabs that are infinite in two dimensions) that were comprised of various proportions and densities of plutonium oxides and uranium oxides, each having various isotopic compositions; and (5) sensitivity coefficients of delta k-eff with respect to critical geometry delta dimensions were determined for each of the three geometries that were studied. The study was undertaken to support the development of a standard that is sponsored by the International Standards Organization (ISO) under Technical Committee 85, Nuclear Energy (TC 85)--Subcommittee 5, Nuclear Fuel Technology (SC 5)--Working Group 8, Standardization of Calculations, Procedures and Practices Related to Criticality Safety (WG 8). The designation and title of the ISO TC 85/SC 5/WG 8 standard working draft is WD 14941, ''Nuclear energy--Fissile materials--Nuclear criticality control and safety of plutonium-uranium oxide fuel mixtures outside of reactors.'' Various ISO member participants performed similar computational studies using their indigenous computational codes to provide comparative results for analysis in the development of the standard.

  20. Plutonium stabilization and handling (PuSH)

    SciTech Connect

    Weiss, E.V.

    1997-01-23

    This Functional Design Criteria (FDC) addresses construction of a Stabilization and Packaging System (SPS) to oxidize and package for long term storage remaining plutonium-bearing special nuclear materials currently in inventory at the Plutonium Finishing Plant (PFP), and modification of vault equipment to allow storage of resulting packages of stabilized SNM for up to fifty years. The major sections of the project are: site preparation; SPS Procurement, Installation, and Testing; storage vault modification; and characterization equipment additions. The SPS will be procured as part of a Department of Energy nationwide common procurement. Specific design crit1460eria for the SPS have been extracted from that contract and are contained in an appendix to this document.

  1. Plutonium aging

    SciTech Connect

    Olivas, J.D.

    1999-03-01

    The author describes the plutonium aging program at the Los Alamos National Laboratory. The aging of plutonium components in the US nuclear weapons stockpile has become a concern due to several events: the end of the cold war, the cessation of full scale underground nuclear testing as a result of the Comprehensive Test Ban Treaty (CTBT) and the closure of the Rocky Flats Plant--the site where the plutonium components were manufactured. As a result, service lifetimes for nuclear weapons have been lengthened. Dr. Olivas will present a brief primer on the metallurgy of plutonium, and will then describe the technical approach to ascertaining the long-term changes that may be attributable to self-radiation damage. Facilities and experimental techniques which are in use to study aging will be described. Some preliminary results will also be presented.

  2. Processing of the MCC K-26 Plutonium-bearing Sludges to Recover Weapons-grade Plutonium That is Not Under Any Treaty or Monitoring Agreement

    SciTech Connect

    Jardin, L J; Kudinov, K G; Tretyakov, A A; Bondin, V V; Sorokin, Y P; Manakova, L F; Shvedov, A A; Aloy, A S; Borisov, G B; Gupalo, T A

    2001-12-12

    Russian Federation (RF) and United States (US) collaborations from July 1998 through July 2001 conducted investigations of the Pu-bearing sludges in storage at the Mining Chemical Combine (MCC) K-26 site in order to dispose of weapons-grade plutonium and decommission the radiochemical plant. This RF work resulted in the recovery of approximately 20 kg of weapons-grade plutonium (and {approx}19 MT of uranium) from the sludges which was stored as oxide. Another method investigated and partially developed as joint collaborative efforts during this time period was direct immobilization of plutonium with no recovery of plutonium. This method melts the untreated recovered sludges by microwave ultrahigh frequency (UHF) heating with glass formers. After cooling, melter-crucibles of vitrified sludge are stored on site in underground cavities for eventual disposal in a geologic repository. Cost and technical feasibility studies of the two methods show that direct immobilization (i.e., vitrification) of the plutonium-containing sludge is the preferred alternative. It is also preferred from the ecological point of view. However, RF funding alone is insufficient to continue this work, and US funding has been suspended. It appears unlikely that development of full scale vitrification technologies for the plutonium-bearing sludges can be undertaken without continuing support from the US or from others. Thus, the only demonstrated technology for the MCC for removing weapons-grade plutonium in sludges will remain recovery and extraction of plutonium for storage and reuse for the indefinite future. It is estimated the about 1200 to 1800 kg of weapons plutonium are in the sludges that must be removed and treated as part of the MCC facility decommissioning. This specific plutonium is not covered under any current monitoring or treaty agreement between the RF and the US.

  3. Processing of the MCC K26 Plutonium-Bearing Sludges to Recover Weapons-Grade Plutonium That is Not Under any Treaty or Monitoring Agreement

    SciTech Connect

    Jardine, L. J.; Kudinov, K. G.; Tretyakov, A. A.; Bondin, V. V.; Sorokin, Y. P.; Manakova, L. F.; Shvedov, A. A.; Aloy, A. S.; Borisov, G. B.; Gupalo, T. A.

    2002-02-26

    Russian Federation (RF) and United States (US) collaborations from July 1998 through July 2001 conducted investigations of the Pu-bearing sludges in storage at the Mining Chemical Combine (MCC) K-26 site in order to dispose of weapons-grade plutonium and decommission the radiochemical plant. This RF work resulted in the recovery of approximately 20 kg of weapons-grade plutonium (and {approx}19 MT of uranium) from the sludges which was stored as oxide. Another method investigated and partially developed as joint collaborative efforts during this time period was direct immobilization of plutonium with no recovery of plutonium. This method melts the untreated recovered sludges by microwave ultrahigh frequency (UHF) heating with glass formers. After cooling, melter-crucibles of vitrified sludge are stored on site in underground cavities for eventual disposal in a geologic repository. Cost and technical feasibility studies of the two methods show that direct immobilization (i.e., vitrification)of the plutonium-containing sludge is the preferred alternative. It is also preferred from the ecological point of view. However, RF funding alone is insufficient to continue this work, and US funding has been suspended. It appears unlikely that development of full scale vitrification technologies for the plutonium-bearing sludges can be undertaken without continuing support from the US or from others. Thus, the only demonstrated technology for the MCC for removing weapons-grade plutonium in sludges will remain recovery and extraction of plutonium for storage and reuse for the indefinite future. It is estimated the about 1200 to 1800 kg of weapons plutonium are in the sludges that must be removed an d treated as part of the MCC facility decommissioning. This specific plutonium is not covered under any current monitoring or treaty agreement between the RF and the US.

  4. Plutonium pyrophoricity

    SciTech Connect

    Stakebake, J.L.

    1992-06-02

    A review of the published literature on ignition and burning of plutonium metal was conducted in order to better define the characteristic of pyrophoric plutonium. The major parameter affecting ignition is the surface area/mass ratio of the sample. Based on this parameter, plutonium metal can be classified into four categories: (1) bulk metal, (2) film and foils, (3) chips and turnings, and (4) powder. Other parameters that can alter the ignition of the metal include experimental, chemical, physical, and environmental effects. These effects are reviewed in this report. It was concluded from this review that pyrophoric plutonium can be conservatively defined as: Plutonium metal that will ignite spontaneously in air at a temperature of 150{degrees}C or below in the absence of external heat, shock, or friction. The 150{degrees}C temperature was used to compensate for the self-heating of plutonium metal. For a practical definition of whether any given metal is pyrophoric, all of the factors affecting ignition must be considered.

  5. ION EXCHANGE ADSORPTION PROCESS FOR PLUTONIUM SEPARATION

    DOEpatents

    Boyd, G.E.; Russell, E.R.; Taylor, M.D.

    1961-07-11

    Ion exchange processes for the separation of plutonium from fission products are described. In accordance with these processes an aqueous solution containing plutonium and fission products is contacted with a cation exchange resin under conditions favoring adsorption of plutonium and fission products on the resin. A portion of the fission product is then eluted with a solution containing 0.05 to 1% by weight of a carboxylic acid. Plutonium is next eluted with a solution containing 2 to 8 per cent by weight of the same carboxylic acid, and the remaining fission products on the resin are eluted with an aqueous solution containing over 10 per cent by weight of sodium bisulfate.

  6. Design-only conceptual design report: Plutonium Immobilization Plant

    SciTech Connect

    DiSabatino, A A

    2000-05-01

    This design-only conceptual design report was prepared to support a funding request by the Department of Energy Office of Fissile Materials Disposition for engineering and design of the Plutonium Immobilization Plant, which will be used to immobilize up to 50 tonnes of surplus plutonium. The Plutonium Immobilization Plant will be located at the Savannah River Site pursuant to the Surplus Plutonium Disposition Final Environmental Impact Statement Record of Decision, January 4, 2000. This document reflects a new facility using the ceramic immobilization technology and the can-in-canister approach. The Plutonium Immobilization Plant accepts plutonium oxide from pit conversion and plutonium and plutonium oxide from non-pit sources and, through a ceramic immobilization process, converts the plutonium into mineral-like forms that are subsequently encapsulated within a large canister of high-level waste glass. The final immobilized product must make the plutonium as inherently unattractive and inaccessible for use in nuclear weapons as the plutonium in spent fuel from commercial reactors; it must also be suitable for geologic disposal. Plutonium immobilization at the Savannah River Site uses a new building, the Plutonium Immobilization Plant, which will receive and store feed materials, convert non-pit surplus plutonium to an oxide form suitable for the immobilization process, immobilize the plutonium oxide in a titanate-based ceramic form, place cans of the plutonium-ceramic forms into magazines, and load the magazines into a canister. The existing Defense Waste Processing Facility is used for the pouring of high-level waste glass into the canisters. The Plutonium Immobilization Plant uses existing Savannah River Site infrastructure for analytical laboratory services, waste handling, fire protection, training, and other support utilities and services. This design-only conceptual design report also provides the cost for a Plutonium Immobilization Plant which would process

  7. Purification of aqueous plutonium chloride solutions via precipitation and washing.

    SciTech Connect

    Stroud, M. A.; Salazar, R. R.; Abney, Kent David; Bluhm, E. A.; Danis, J. A.

    2003-01-01

    Pyrochemical operations at Los Alamos Plutonium Facility (TA-55) use high temperature melt s of calcium chloride for the reduction of plutonium oxide to plutonium metal and hi gh temperature combined melts of sodium chloride and potassium chloride mixtures for the electrorefining purification of plutonium metal . The remaining plutonium and americium are recovered from thes e salts by dissolution in concentrated hydrochloric acid followed by either solvent extraction or io n exchange for isolation and ultimately converted to oxide after precipitation with oxalic acid . Figur e 1 illustrates the current aqueous chloride flow sheet used for plutonium processing at TA-55 .

  8. SOLVENT EXTRACTION PROCESS FOR PLUTONIUM

    DOEpatents

    Seaborg, G.T.

    1959-04-14

    The separation of plutonium from aqueous inorganic acid solutions by the use of a water immiscible organic extractant liquid is described. The plutonium must be in the oxidized state, and the solvents covered by the patent include nitromethane, nitroethane, nitropropane, and nitrobenzene. The use of a salting out agents such as ammonium nitrate in the case of an aqueous nitric acid solution is advantageous. After contacting the aqueous solution with the organic extractant, the resulting extract and raffinate phases are separated. The plutonium may be recovered by any suitable method.

  9. Long term plutonium solubility and speciation studies in a synthetic brine

    SciTech Connect

    Nitsche, Heino; Roberts, K.; Xi, Ruihua

    1993-12-31

    The rate at which elements can be transported in groundwater systems is governed in part by the solubility of the element in the groundwater. This report documents plutonium solubility experiments in a brine simulant relevant to the Waste Isolation Pilot Plant. Approximately 1 to 2.5 mL of five stock solutions containing single oxidation states of plutonium (Pu(IV)-polymer, Pu{sup 3+}, Pu{sup 4+}, PuO{sub 2}{sup +}, and PuO{sub 2}{sup 2+}) were added to {approximately}75 mL of synthetic H-17 Brine in five reaction vessels. Initial plutonium concentrations ranged from 1.3 {times} l0{sup {minus}4} to 5.l {times} l0{sup {minus}4} M (moles per liter) total plutonium. Because these initial concentrations were far above the plutonium solubility limit in H-17 Brine, plutonium-containing solids precipitated. Aqueous plutonium concentrations were measured over time until steady-state was reached, requiring over 300 days in H-17 Brine.

  10. Hydrogen Reduction of Zinc and Iron Oxides Containing Mixtures

    NASA Astrophysics Data System (ADS)

    de Siqueira, Rogério Navarro C.; de Albuquerque Brocchi, Eduardo; de Oliveira, Pamela Fernandes; Motta, Marcelo Senna

    2013-10-01

    Zinc is a metal of significant technological importance and its production from secondary sources has motivated the development of alternative processes, such as the chemical treatment of electrical arc furnace (EAF) dust. Currently, the extraction of zinc from the mentioned residue using a carbon-containing reducing agent is in the process of being established commercially and technically. In the current study, the possibility of reducing zinc from an EAF dust sample through a H2 constant flux in a horizontal oven is studied. The reduction of a synthetic oxide mixture of analogous composition is also investigated. The results indicated that the reduction process is thermodynamically viable for temperatures higher than 1123 K (850 °C), and all zinc metal produced is transferred to the gas stream, enabling its complete separation from iron. The same reaction in the presence of zinc crystals was considered for synthesizing FeZn alloys. However, for the experimental conditions employed, although ZnO reduction was indeed thermodynamically hindered because of the presence of zinc crystals (the metal's partial pressure was enhanced), the zinc metal's escape within the gaseous phase could not be effectively avoided.

  11. The formation of crystals in glasses containing rare earth oxides

    NASA Astrophysics Data System (ADS)

    Fadzil, Syazwani Mohd; Hrma, Pavel; Crum, Jarrod; Siong, Khoo Kok; Ngatiman, Mohammad Fadzlee; Said, Riduan Mt

    2014-02-01

    Korean spent nuclear fuel will reach the capacity of the available temporary storage by 2016. Pyroprocessing and direct disposal seems to be an alternative way to manage and reuse spent nuclear fuel while avoiding the wet reprocessing technology. Pyroprocessing produces several wastes streams, including metals, salts, and rare earths, which must be converted into stabilized form. A suitable form for rare earth immobilization is borosilicate glass. The borosilicate glass form exhibits excellent durability, allows a high waste loading, and is easy to process. In this work, we combined the rare earths waste of composition (in wt%) 39.2Nd2O3-22.7CeO2-11.7La2O3-10.9PrO2-1.3Eu2O3-1.3Gd2O3-8.1Sm2O3-4.8Y2O3 with a baseline glass of composition 60.2SiO2-16.0B2O3-12.6Na2O-3.8Al2O3-5.7CaO-1.7ZrO2. Crystallization in waste glasses occurs as the waste loading increases. It may produce complicate glass processing and affect the product quality. To study crystal formation, we initially made glasses containing 5%, 10% and 15% of La2O3 and then glasses with 5%, 10% and 15% of the complete rare earth mix. Samples were heat-treated for 24 hours at temperatures 800°C to 1150°C in 50°C increments. Quenched samples were analyzed using an optical microscope, scanning electron microscope with energy dispersive spectroscopy, and x-ray diffraction. Stillwellite (LaBSiO5) and oxyapatite (Ca2La8Si6O26) were found in glasses containing La2O3, while oxyapatite (Ca2La8Si6O26 and NaNd9Si6O26) precipitated in glasses with additions of mixed rare earths. The liquidus temperature (TL) of the glasses containing 5%, 10% and 15% La2O3 were 800°C, 959°C and 986°C, respectively; while TL was 825°C, 1059°C and 1267°C for glasses with 5%, 10% and 15% addition of mixed rare earth oxides. The component coefficients TB2O3, TSiO2, TCaO, and TRE2O3 were also evaluated using a recently published study.

  12. Hydride-catalyzed corrosion of plutonium by air: Initiation by plutonium monoxide monohydride

    SciTech Connect

    Allen, T.H.; Haschke, J.M.

    1998-06-01

    Chemistry and kinetics of air reactions with plutonium monoxide monohydride (PuOH) and with mixtures of the oxide hydride and plutonium metal are defined by results of pressure-volume-temperature (PVT) measurements. Test with specimens prepared by total and partial corrosion of plutonium in 0.05 M sodium chloride solution show that reaction of residual water continues to generate H{sub 2} after liquid water is removed by evacuation. Rapid exposure of PuOH to air at room temperature does not produce a detectable reaction, but similar exposure of a partially corroded metal sample containing Pu and PuOH results in hydride (PuH{sub x})-catalyzed corrosion of the residual Pu. Kinetics of he first-order reaction resulting in formation of the PuH{sub x} catalyst and of the indiscriminate reaction of N{sub 2} and O{sub 2} with plutonium metal are defined. The rate of the catalyzed Pu+air reaction is independent of temperature (E{sub a} = 0), varies as the square of air pressure, and equals 0.78 {+-} 0.03 g Pu/cm{sup 2} min in air at one atmosphere. The absence of pyrophoric behavior for PuOH and differences in the reactivities of PuOH and PuOH + Pu mixtures are attributed to kinetic control by gaseous reaction products. Thermodynamic properties of the oxide hydride are estimated, particle size distributions of corrosion products are presented, and potential hazards associated with products formed by aqueous corrosion of plutonium are discussed.

  13. SEPARATION OF PLUTONIUM IONS FROM SOLUTION BY ADSORPTION ON ZIRCONIUM PYROPHOSPHATE

    DOEpatents

    Stoughton, R.W.

    1961-01-31

    A method is given for separating plutonium in its reduced, phosphate- insoluble state from other substances. It involves contacting a solution containing the plutonium with granular zirconium pyrophosphate.

  14. Oxidative dissolution of actinide oxides in H 2O 2 containing aqueous solution - A preliminary study

    NASA Astrophysics Data System (ADS)

    Pehrman, Reijo; Amme, Marcus; Roth, Olivia; Ekeroth, Ella; Jonsson, Mats

    2010-02-01

    Oxidative dissolution of spent nuclear fuel is an important issue in the safety assessment of a future geological repository for spent nuclear fuel. Although UO 2 constitutes, in terms of mass, the majority of the spent fuel material, its main radiotoxicity is (after extended storage times) contained in actinides with half lives shorter than that of 238-uranium, such as isotopes of Np and Pu. Relatively little information is available on the dissolution behavior of Np and Pu in comparable environments. This work investigates the oxidative dissolution of NpO 2 and PuO 2 in non-complexing aqueous solutions containing H 2O 2 and compares their behavior with that of UO 2. We have found that oxidative dissolution takes place for all three actinides in the presence of H 2O 2. Based on the obtained dissolution rates, we would not expect the dissolution of the actinides to be congruent. Instead, in a system without complexing agent, the release rates of Np and Pu are expected to be lower than the U release rate.

  15. SUPPORTING SAFE STORAGE OF PLUTONIUM-BEARING MATERIALS THROUGH SCIENCE, ENGINEERING AND SURVEILLANCE

    SciTech Connect

    Dunn, K.; Chandler, G.; Gardner, C.; Louthan, M.; Mcclard, J.

    2009-11-10

    Reductions in the size of the U. S. nuclear weapons arsenal resulted in the need to store large quantities of plutonium-bearing metals and oxides for prolonged periods of time. To assure that the excess plutonium from the U. S. Department of Energy (DOE) sites was stored in a safe and environmentally friendly manner the plutonium-bearing materials are stabilized and packaged according to well developed criteria published as a DOE Standard. The packaged materials are stored in secure facilities and regular surveillance activities are conducted to assure continuing package integrity. The stabilization, packaging, storage and surveillance requirements were developed through extensive science and engineering activities including those related to: plutonium-environment interactions and container pressurization, corrosion and stress corrosion cracking, plutonium-container material interactions, loss of sealing capability and changes in heat transfer characteristics. This paper summarizes some of those activities and outlines ongoing science and engineering programs that assure continued safe and secure storage of the plutonium-bearing metals and oxides.

  16. TECHNICAL BASIS FOR DOE STANDARD 3013 EQUIVALENCY SUPPORTING REDUCED TEMPERATURE STABILIZATION OF OXALATE-DERIVED PLUTONIUM OXIDE PRODUCED BY THE HB-LINE FACILITY AT SAVANNAH RIVER SITE

    SciTech Connect

    Duffey, J.; Livingston, R.; Berg, J.; Veirs, D.

    2012-07-02

    The HB-Line (HBL) facility at the Savannah River Site (SRS) is designed to produce high-purity plutonium dioxide (PuO{sub 2}) which is suitable for future use in production of Mixed Oxide (MOX) fuel. The MOX Fuel Fabrication Facility (MFFF) requires PuO{sub 2} feed to be packaged per the U.S. Department of Energy (DOE) Standard 3013 (DOE-STD-3013) to comply with the facility's safety basis. The stabilization conditions imposed by DOE-STD-3013 for PuO{sub 2} (i.e., 950 C for 2 hours) preclude use of the HBL PuO{sub 2} in direct fuel fabrication and reduce the value of the HBL product as MFFF feedstock. Consequently, HBL initiated a technical evaluation to define acceptable operating conditions for production of high-purity PuO{sub 2} that fulfills the DOE-STD-3013 criteria for safe storage. The purpose of this document is to demonstrate that within the defined operating conditions, the HBL process will be equivalent for meeting the requirements of the DOE-STD-3013 stabilization process for plutonium-bearing materials from the DOE complex. The proposed 3013 equivalency reduces the prescribed stabilization temperature for high-purity PuO{sub 2} from oxalate precipitation processes from 950 C to 640 C and places a limit of 60% on the relative humidity (RH) at the lowest material temperature. The equivalency is limited to material produced using the HBL established flow sheet, for example, nitric acid anion exchange and Pu(IV) direct strike oxalate precipitation with stabilization at a minimum temperature of 640 C for four hours (h). The product purity must meet the MFFF acceptance criteria of 23,600 {micro}g/g Pu (i.e., 2.1 wt %) total impurities and chloride content less than 250 {micro}g/g of Pu. All other stabilization and packaging criteria identified by DOE-STD-3013-2012 or earlier revisions of the standard apply. Based on the evaluation of test data discussed in this document, the expert judgment of the authors supports packaging the HBL product under a 3013

  17. Packaging material and flexible medical tubing containing thermally exfoliated graphite oxide

    NASA Technical Reports Server (NTRS)

    Prud'homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor)

    2011-01-01

    A packaging material or flexible medical tubing containing a modified graphite oxide material, which is a thermally exfoliated graphite oxide with a surface area of from about 300 m.sup.2/g to 2600 m.sup.2/g.

  18. Weapons-grade plutonium dispositioning. Volume 4. Plutonium dispositioning in light water reactors

    SciTech Connect

    Sterbentz, J.W.; Olsen, C.S.; Sinha, U.P.

    1993-06-01

    This study is in response to a request by the Reactor Panel Subcommittee of the National Academy of Sciences (NAS) Committee on International Security and Arms Control (CISAC) to evaluate the feasibility of using plutonium fuels (without uranium) for disposal in existing conventional or advanced light water reactor (LWR) designs and in low temperature/pressure LWR designs that might be developed for plutonium disposal. Three plutonium-based fuel forms (oxides, aluminum metallics, and carbides) are evaluated for neutronic performance, fabrication technology, and material and compatibility issues. For the carbides, only the fabrication technologies are addressed. Viable plutonium oxide fuels for conventional or advanced LWRs include plutonium-zirconium-calcium oxide (PuO{sub 2}-ZrO{sub 2}-CaO) with the addition of thorium oxide (ThO{sub 2}) or a burnable poison such as erbium oxide (Er{sub 2}O{sub 3}) or europium oxide (Eu{sub 2}O{sub 3}) to achieve acceptable neutronic performance. Thorium will breed fissile uranium that may be unacceptable from a proliferation standpoint. Fabrication of uranium and mixed uranium-plutonium oxide fuels is well established; however, fabrication of plutonium-based oxide fuels will require further development. Viable aluminum-plutonium metallic fuels for a low temperature/pressure LWR include plutonium aluminide in an aluminum matrix (PuAl{sub 4}-Al) with the addition of a burnable poison such as erbium (Er) or europium (Eu). Fabrication of low-enriched plutonium in aluminum-plutonium metallic fuel rods was initially established 30 years ago and will require development to recapture and adapt the technology to meet current environmental and safety regulations. Fabrication of high-enriched uranium plate fuel by the picture-frame process is a well established process, but the use of plutonium would require the process to be upgraded in the United States to conform with current regulations and minimize the waste streams.

  19. Oxygen self-diffusion in polycrystalline uranium-plutonium mixed oxide U0.55Pu0.45O2

    NASA Astrophysics Data System (ADS)

    Vauchy, Romain; Robisson, Anne-Charlotte; Bienvenu, Philippe; Roure, Ingrid; Hodaj, Fiqiri; Garcia, Philippe

    2015-12-01

    Atomic transport properties in uranium-plutonium mixed oxides U1-yPuyO2 are of essential concern because they impact numerous aspects of their physicochemical behavior at all stages of the fuel cycle. In this paper, we report oxygen tracer diffusion coefficients in homogeneous U0.55Pu0.45O2 mixed oxide. The study is based on tracer diffusion coefficient measurements obtained using Secondary Ion Mass Spectrometry (SIMS) following diffusion annealing involving gas-solid 18O/16O isotopic exchange. As for other fundamental material properties governed by the nature and behavior of point defects, we show that a careful study of tracer diffusion coefficients as a function of oxygen partial pressure and temperature is liable to provide insight into prevailing diffusion mechanisms. Under the conditions studied in this work, it appears that oxygen diffusion is vacancy mediated.

  20. Plutonium 238 facilities at Los Alamos

    NASA Astrophysics Data System (ADS)

    Rinehart, Gary H.

    1991-01-01

    Plutonium 238 operations at Los Alamos are performed at the Plutonium Facility (TA-55), the Chemistry and Metallurgy Research (CMR) Building, and the Radioisotope Fuels Impact Test Facility. The plutonium 238 facilities at Los Alamos support a wide variety of heat source activities including development of new fuel forms and containment materials, research on the high temperature properties of containment materials, investigation of the high temperature compatibility of fuels with potential container materials, processing plutonium 238 fuel forms, manufacture of heat sources under quality assurance surveillance, and performing safety testing on heat sources and radioisotope thermoelectric generators.

  1. Plutonium-238 facilities at Los Alamos

    NASA Astrophysics Data System (ADS)

    Rinehart, Gary H.

    Plutonium-238 operations at Los Alamos are performed at the Plutonium Facility (TA-55), the Chemistry and Metallurgy Research (CMR) Building, and the Radioisotope Fuels Impact Test Facility. The plutonium-238 facilities at Los Alamos support a wide variety of heat source activities including development of new fuel forms and containment materials, research on the high temperature properties of containment materials, investigation of the high temperature compatibility of fuels with potential container materials, processing plutonium-238 fuel forms, manufacture of heat sources under quality assurance surveillance, and performing safety testing on heat sources and radioisotope thermoelectric generators.

  2. METHOD FOR RECOVERING PLUTONIUM VALUES FROM SOLUTION USING A BISMUTH HYDROXIDE CARRIER PRECIPITATE

    DOEpatents

    Faris, B.F.

    1961-04-25

    Carrier precipitation processes for separating plutonium values from aqueous solutions are described. In accordance with the invention a bismuth hydroxide precipitate is formed in the plutonium-containing solution, thereby carrying plutonium values from the solution.

  3. Method of forming supported doped palladium containing oxidation catalysts

    SciTech Connect

    Mohajeri, Nahid

    2014-04-22

    A method of forming a supported oxidation catalyst includes providing a support comprising a metal oxide or a metal salt, and depositing first palladium compound particles and second precious metal group (PMG) metal particles on the support while in a liquid phase including at least one solvent to form mixed metal comprising particles on the support. The PMG metal is not palladium. The mixed metal particles on the support are separated from the liquid phase to provide the supported oxidation catalyst.

  4. Oxide-Nanoparticle Containing Coatings for High Temperature Alloys

    SciTech Connect

    2009-04-01

    This factsheet describes a study whose objective is to examine the feasibility of using Electromagnetic Stirring (EMS) techniques in dispersing the oxide nanoparticles uniformly within the liquid steel.

  5. Phagocytosis of hybrid molecular nanosomal compositions containing oxidized dextrans conjugated with isonicotinic acid hydrazide by macrophages.

    PubMed

    Shkurupy, V A; Arkhipov, S A; Troitsky, A V; Luzgina, N G; Zaikovskaja, M V; Ufimceva, E G; Iljine, D A; Akhramenko, E S; Gulyaeva, E P; Bistrova, T N

    2009-12-01

    We studied phagocytic activity of macrophages towards hybrid molecular nanosomal compositions consisting of 150-800-nm nanoliposomes containing oxidized dextrans with a molecular weight of 35 and 60 kDa obtained by chemical ("permanganate") and radiochemical oxidation of dextran conjugated with isonicotinic acid hydrazide (dextrazides, intracellular prolonged antituberculous drugs). Phagocytic activity of macrophages towards hybrid molecular nanosomal compositions containing dextrazides obtained by chemical oxidation of dextrans is higher than activity towards hybrid molecular nanosomal compositions containing dextrazides prepared by radiochemical oxidation and depends on the size of hybrid molecular nanosomal compositions and molecular weight of oxidized dextrans. PMID:21116494

  6. WET METHOD OF PREPARING PLUTONIUM TRIBROMIDE

    DOEpatents

    Davidson, N.R.; Hyde, E.K.

    1958-11-11

    S> The preparation of anhydrous plutonium tribromide from an aqueous acid solution of plutonium tetrabromide is described, consisting of adding a water-soluble volatile bromide to the tetrabromide to provide additional bromide ions sufficient to furnish an oxidation-reduction potential substantially more positive than --0.966 volt, evaporating the resultant plutonium tribromides to dryness in the presence of HBr, and dehydrating at an elevated temperature also in the presence of HBr.

  7. EVALUATION OF FIRE HAZARDS WHILE REPACKAGING PLUTONIUM-CONTAMINATED SCRAP IN HB-LINE

    SciTech Connect

    Hallman, D

    2003-12-18

    The potential for a fire while repackaging plutonium-contaminated scrap was evaluated. The surface-to-mass ratio indicates the metal alone will not spontaneously ignite. Uranium hydride can form when uranium metal is exposed to water vapor or hydrogen; uranium hydride reacts rapidly and energetically with atmospheric oxygen. The plutonium-contaminated scrap has been inside containers qualified for shipping, and these containers are leak-tight. The rate of diffusion of water vapor through the seals is small, and the radiolytic hydrogen generation rate is low. Radiography of samples of the storage containers indicates no loose oxide/hydride powder has collected in the storage container to date. The frequently of a fire while repackaging the plutonium-contaminated scrap is extremely unlikely.

  8. The formation of crystals in glasses containing rare earth oxides

    SciTech Connect

    Fadzil, Syazwani Mohd; Hrma, Pavel; Crum, Jarrod; Siong, Khoo Kok; Ngatiman, Mohammad Fadzlee; Said, Riduan Mt

    2014-02-12

    Korean spent nuclear fuel will reach the capacity of the available temporary storage by 2016. Pyroprocessing and direct disposal seems to be an alternative way to manage and reuse spent nuclear fuel while avoiding the wet reprocessing technology. Pyroprocessing produces several wastes streams, including metals, salts, and rare earths, which must be converted into stabilized form. A suitable form for rare earth immobilization is borosilicate glass. The borosilicate glass form exhibits excellent durability, allows a high waste loading, and is easy to process. In this work, we combined the rare earths waste of composition (in wt%) 39.2Nd{sub 2}O{sub 3}–22.7CeO{sub 2}–11.7La{sub 2}O{sub 3}–10.9PrO{sub 2}–1.3Eu{sub 2}O{sub 3}–1.3Gd{sub 2}O{sub 3}–8.1Sm{sub 2}O{sub 3}–4.8Y{sub 2}O{sub 3} with a baseline glass of composition 60.2SiO{sub 2}–16.0B{sub 2}O{sub 3}–12.6Na{sub 2}O–3.8Al{sub 2}O{sub 3}–5.7CaO–1.7ZrO{sub 2}. Crystallization in waste glasses occurs as the waste loading increases. It may produce complicate glass processing and affect the product quality. To study crystal formation, we initially made glasses containing 5%, 10% and 15% of La{sub 2}O{sub 3} and then glasses with 5%, 10% and 15% of the complete rare earth mix. Samples were heat-treated for 24 hours at temperatures 800°C to 1150°C in 50°C increments. Quenched samples were analyzed using an optical microscope, scanning electron microscope with energy dispersive spectroscopy, and x-ray diffraction. Stillwellite (LaBSiO{sub 5}) and oxyapatite (Ca{sub 2}La{sub 8}Si{sub 6}O{sub 26}) were found in glasses containing La{sub 2}O{sub 3}, while oxyapatite (Ca{sub 2}La{sub 8}Si{sub 6}O{sub 26} and NaNd{sub 9}Si{sub 6}O{sub 26}) precipitated in glasses with additions of mixed rare earths. The liquidus temperature (T{sub L}) of the glasses containing 5%, 10% and 15% La{sub 2}O{sub 3} were 800°C, 959°C and 986°C, respectively; while T{sub L} was 825°C, 1059°C and 1267°C for glasses

  9. Sodium sulfur container with chromium/chromium oxide coating

    DOEpatents

    Ludwig, Frank A.; Higley, Lin R.

    1981-01-01

    A coating of chromium/chromium oxide is disclosed for coating the surfaces of electrically conducting components of a sodium sulfur battery. This chromium/chromium oxide coating is placed on the surfaces of the electrically conducting components of the battery which are in contact with molten polysulfide and sulfur reactants during battery operation.

  10. Structure-Property Relationships in Oxides Containing Tellurium

    NASA Astrophysics Data System (ADS)

    Siritanon, Theeranun

    Oxides of post-transition metals often show unique structures and properties due to the presence of lone pair electrons and the diffused s orbitals. The present work focuses on synthesis and characterizations of oxides containing Te, a heavy post transition metal. New series of pyrochlore oxides of the formula Cs(M,Te)2O 6 (M = Al, Ga, Cr, Fe, Co, In, Ho, Lu, Yb, Er, Ge, Rh, Ti, Zn, Ni, and Mg) have been prepared. The samples were highly colored (ranging from black to dark green) indicating a possible mixed valency for Te with appreciable charge transfer between them in the octahedral sites. Electronic conductivity was observed in some phases and could be as high as 2S/cm (M=Ge). Seebeck coefficients of conducting samples show negative values which suggest that electrons are the major charge carriers. Temperature dependence of conductivity indicates that the samples are semiconductors with, in some cases, degenerate semiconducting behavior. Detailed studies on the conduction mechanism indicate the mixed valency of tellurium which leads to semiconducting behavior and the color of the compounds. Systematic studies of cesium tellurate with CsTe2O6-x where x = 0, 0.15, 0.25, 1.5 have been investigated. On heating at slightly above 600ºC, CsTe2O6 loses oxygen resulting in cubic structure with disordered Te4+/Te6+ and oxygen vacancies. Two novel phases of CsTe2O6-x were prepared with orthorhombic structure. The first phase with x value of about 0.2-0.3 crystallizes in Pnma symmetry. At higher values of x, a new compound was discovered with a structure related to Rb4Te 8O23. Optical properties of the compounds are consistent with their colors. CsTe2O6 belongs to class II mixed valency according to Robin and Day classification. However, structures and properties of CsTe 2O6-x phases indicate that they are class I mixed valence compounds. Series of compounds with formula CsTe2-xWxO 6 with x=0.2-0.5 have been made which can be considered as solid solution of CsTe2O6 and Cs

  11. IDENTIFYING IMPURITIES IN SURPLUS NON PIT PLUTONIUM FEEDS FOR MOX OR ALTERNATIVE DISPOSITION

    SciTech Connect

    Allender, J; Moore, E

    2010-07-14

    This report provides a technical basis for estimating the level of corrosion products in materials stored in DOE-STD-3013 containers based on extrapolating available chemical sample results. The primary focus is to estimate the levels of nickel, iron, and chromium impurities in plutonium-bearing materials identified for disposition in the United States Mixed Oxide fuel process.

  12. Sludge stabilization at the Plutonium Finishing Plant, Hanford Site, Richland, Washington

    SciTech Connect

    Not Available

    1994-10-01

    This Environmental Assessment evaluates the proposed action to operate two laboratory-size muffle furnaces in glovebox HC-21C, located in the Plutonium Finishing Plant (PFP), Hanford Site, Richland, Washington. The muffle furnaces would be used to stabilize chemically reactive sludges that contain approximately 25 kilograms (55 pounds) of plutonium by heating to approximately 500 to 1000{degrees}C (900 to 1800{degrees}F). The resulting stable powder, mostly plutonium oxide with impurities, would be stored in the PFP vaults. The presence of chemically reactive plutonium-bearing sludges in the process gloveboxes poses a risk to workers from radiation exposure and limits the availability of storage space for future plant cleanup. Therefore, there is a need to stabilize the material into a form suitable for long-term storage. This proposed action would be an interim action, which would take place prior to completion of an Environmental Impact Statement for the PFP which would evaluate stabilization of all plutonium-bearing materials and cleanout of the facility. However, only 10 percent of the total quantity of plutonium in reactive materials is in the sludges, so this action will not limit the choice of reasonable alternatives or prejudice the Record of Decision of the Plutonium Finishing Plant Environmental Impact Statement.

  13. Design and Dynamic Performance of a Small Water Cooled Reactor Fuelled with Plutonium in Rock-Like Oxide (ROX) Form

    SciTech Connect

    Gaultier, M.; Danguy, G.; Ritchie, D.; Williams, A.; Thompson, A.; Brushwood, J.; Beeley, P.A.; Greenlees, L.; Perry, A.

    2006-07-01

    The results of a design study for a small water-cooled reactor with plutonium fuel in a rock like oxide (ROX) form are reported. A summary is given of the five study areas, Physics, Thermal Hydraulics, Materials, Navalisation and Dynamics, and Shielding and Decommissioning. The dynamics simulation for the whole plant is then described in more detail. The physics of the fuel module is studied using the WIMS suite of deterministic codes with selected computations checked with the Monte-Carlo code MONK. Whole core calculations are undertaken with the WIMS/SNAP code. Essential parameters are provided to the other study areas including reactivity feedback coefficients for the Dynamics. The Thermal Hydraulic design aims to remove the required maximum power using pumped flow and also to provide significant power removal using natural circulation. The major components of the primary circuit are sized and flow rates in pumped and natural circulation calculated by hand and by using the TRACPFQ code. This information is also used in the dynamics study. Further details of the Physics and Thermal hydraulics studies will be given at PHYSOR 2006. The materials study is being published elsewhere, but a brief description of the temperature and stress calculations for the fuel pellet performed with the ABACUS finite element code is given. Navalisation and dynamics of the plant are examined. The power requirements for the plant are estimated and a suitable electric propulsion system is proposed and sized. A whole plant model is built using the AcslXtreme computer package in which a block diagram of the system is constructed via a graphical interface and simulations of the system transients are produced. The block diagram for the whole system is described followed by the describing equations for the major blocks representing neutron kinetics, fuel element heat transfer, thermal hydraulics of the primary circuit and of the steam generators. Also included are describing equations for

  14. Preparation and properties of a compound containing nickel in highest oxidation states

    SciTech Connect

    Nikol'skii, V.A.; Bekreneva, L.A.; Garmash, L.A.; Makashev, Yu.A.; Mikhailova, K.A.; Raikhel'son, L.B.; Shchelkunova, L.I.

    1986-04-10

    The purpose of this paper is to obtain nickel hydroxide oxide of the optimal composition both by chemical and by electrochemical oxidation for use as the cathode material in alkaline batteries. The authors selected the oxidation conditions (oxidant, component ratio) for production of gamma-NiOOH; hydrogen peroxide, potassium permanganate, and potassium persulfate were used as oxidants. Gamma-NiOOH was also obtained by electrochemical oxidation of nickel (II) oxide hydrate in alkaline solutions. In samples obtained by electrochemical oxidation virtually 100% of the nickel is in the highly oxidized state. It is shown that gamma-NiOOH obtained by both chemical and electrochemical oxidation is a complex individual compound containing nickel and in the highest oxidation states, Ni/sup 3 +/ and Ni/sup 4 +/, together with salt-forming potassium ions and water molecules. The gamma-NiOOH obtained may be used with success in galvanic cells.

  15. IODATE METHOD FOR PURIFYING PLUTONIUM

    DOEpatents

    Stoughton, R.W.; Duffield, R.B.

    1958-10-14

    A method is presented for removing radioactive fission products from aqueous solutions containing such fission products together with plutonium. This is accomplished by incorporating into such solutions a metal iodate precipitate to remove fission products which form insoluble iodates. Suitable metal iodates are those of thorium and cerium. The plutonium must be in the hexavalent state and the pH of the solution must be manintained at less than 2.

  16. MOLTEN PLUTONIUM FUELED FAST BREEDER REACTOR

    DOEpatents

    Kiehn, R.M.; King, L.D.P.; Peterson, R.E.; Swickard, E.O. Jr.

    1962-06-26

    A description is given of a nuclear fast reactor fueled with molten plutonium containing about 20 kg of plutonium in a tantalum container, cooled by circulating liquid sodium at about 600 to 650 deg C, having a large negative temperature coefficient of reactivity, and control rods and movable reflector for criticality control. (AEC)

  17. 10 CFR 71.63 - Special requirement for plutonium shipments.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Special requirement for plutonium shipments. 71.63 Section... MATERIAL Package Approval Standards § 71.63 Special requirement for plutonium shipments. Shipments containing plutonium must be made with the contents in solid form, if the contents contain greater than...

  18. 10 CFR 71.63 - Special requirement for plutonium shipments.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Special requirement for plutonium shipments. 71.63 Section... MATERIAL Package Approval Standards § 71.63 Special requirement for plutonium shipments. Shipments containing plutonium must be made with the contents in solid form, if the contents contain greater than...

  19. 10 CFR 71.63 - Special requirement for plutonium shipments.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Special requirement for plutonium shipments. 71.63 Section... MATERIAL Package Approval Standards § 71.63 Special requirement for plutonium shipments. Shipments containing plutonium must be made with the contents in solid form, if the contents contain greater than...

  20. 10 CFR 71.63 - Special requirement for plutonium shipments.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Special requirement for plutonium shipments. 71.63 Section... MATERIAL Package Approval Standards § 71.63 Special requirement for plutonium shipments. Shipments containing plutonium must be made with the contents in solid form, if the contents contain greater than...

  1. 10 CFR 71.63 - Special requirement for plutonium shipments.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Special requirement for plutonium shipments. 71.63 Section... MATERIAL Package Approval Standards § 71.63 Special requirement for plutonium shipments. Shipments containing plutonium must be made with the contents in solid form, if the contents contain greater than...

  2. PROCESS OF FORMING PLUOTONIUM SALTS FROM PLUTONIUM EXALATES

    DOEpatents

    Garner, C.S.

    1959-02-24

    A process is presented for converting plutonium oxalate to other plutonium compounds by a dry conversion method. According to the process, lower valence plutonium oxalate is heated in the presence of a vapor of a volatile non- oxygenated monobasic acid, such as HCl or HF. For example, in order to produce plutonium chloride, the pure plutonium oxalate is heated to about 700 deg C in a slow stream of hydrogen plus HCl. By the proper selection of an oxidizing or reducing atmosphere, the plutonium halide product can be obtained in either the plus 3 or plus 4 valence state.

  3. Plutonium Surveillance Destructive Examination Requirements at Savannah River National Laboratory for K-Area Interim Surveillance

    SciTech Connect

    Stefek, T. M.

    2005-09-29

    The DOE 3013 storage standard requires nested, welded 300 series stainless steel containers to store plutonium-bearing materials for up to 50 years. Packaged contents include stabilized plutonium-bearing residues that contain chloride salts and a low (< 0.5 weight %) water content. The DOE 3013 STD requires surveillance of the packages over the 50 year lifetime. These surveillance requirements have been further defined by the Integrated Surveillance Program to include both non-destructive examination (NDE) and destructive examination (DE) of the 3013 container. The DE portion of surveillance involves examining the 3013 nested containers, analyzing the head space gas, and evaluating the plutonium oxide chemistry. At SRS, the stored 3013 containers will undergo preparation for the DE surveillance activities in facilities located in K-Area. The actual DE surveillance will be performed in SRNL. This report provides preliminary functional requirements for the destructive examination (DE) of plutonium-bearing oxide materials and containers in support of K-Area Interim Surveillance (KIS). The KIS project will install interim facilities to prepare the samples for analysis in SRNL. This document covers the requirements for the interim period beginning in 2007, and lasting until the Container Storage and Surveillance Capability (CSSC) project provides the permanent facilities in K-Area to perform sampling and repackaging operations associated with the 3013 container storage and surveillance program. Initial requirements for the CSSC project have been previously defined in WSRC-TR-2004-00584 ''Plutonium Surveillance Destructive Examination Requirements at Savannah River National Laboratory''. As part of the Plutonium Surveillance Program of 3013 Containers at the Savannah River Site (SRS), the Savannah River National Laboratory (SRNL) will receive the emptied 3013 container components, plutonium oxide samples and headspace gas samples from K-Area. The DE program scope

  4. URANOUS IODATE AS A CARRIER FOR PLUTONIUM

    DOEpatents

    Miller, D.R.; Seaborg, G.T.; Thompson, S.G.

    1959-12-15

    A process is described for precipitating plutonium on a uranous iodate carrier from an aqueous acid solution conA plutonium solution more concentrated than the original solution can then be obtained by oxidizing the uranium to the hexavalent state and dissolving the precipitate, after separating the latter from the original solution, by means of warm nitric acid.

  5. Conductive ink containing thermally exfoliated graphite oxide and method a conductive circuit using the same

    NASA Technical Reports Server (NTRS)

    Prud'Homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor)

    2011-01-01

    A conductive ink containing a conductive polymer, wherein the conductive polymer contains at least one polymer and a modified graphite oxide material, which is a thermally exfoliated graphite oxide with a surface area of from about 300 sq m/g to 2600 sq m/g, and it use in a method for making a conductive circuit.

  6. Electrospray painted article containing thermally exfoliated graphite oxide and method for their manufacture

    NASA Technical Reports Server (NTRS)

    Korkut, Sibel (Inventor); Prud'Homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor)

    2011-01-01

    A painted polymer part containing a conductive polymer composition containing at least one polymer and a modified graphite oxide material, which is a thermally exfoliated graphite oxide with a surface area of from about 300 sq m/g to 2600 sq m/g, wherein the painted polymer part has been electrospray painted.

  7. EVIDENCE OF CORROSIVE GAS FORMED BY RADIOLYSIS OF CHLORIDE SALTS IN PLUTONIUM-BEARING MATERIALS

    SciTech Connect

    Dunn, K.; Louthan, M.

    2010-02-01

    Corrosion and pitting have been observed in headspace regions of stainless steel containers enclosing plutonium oxide/salt mixtures. These observations are consistent with the formation of a corrosive gas, probably HCl, and transport of that gas to the headspace regions of sealed containers. The NH{sub 4}Cl films found on the walls of the sealed containers is also indicative of the presence of HCl gas. Radiolysis of hydrated alkaline earth salts is the probable source of HCl.

  8. OXIDATION OF TRANSURANIC ELEMENTS

    DOEpatents

    Moore, R.L.

    1959-02-17

    A method is reported for oxidizing neptunium or plutonium in the presence of cerous values without also oxidizing the cerous values. The method consists in treating an aqueous 1N nitric acid solution, containing such cerous values together with the trivalent transuranic elements, with a quantity of hydrogen peroxide stoichiometrically sufficient to oxidize the transuranic values to the hexavalent state, and digesting the solution at room temperature.

  9. On Barium Oxide Solubility in Barium-Containing Chloride Melts

    NASA Astrophysics Data System (ADS)

    Nikolaeva, Elena V.; Zakiryanova, Irina D.; Bovet, Andrey L.; Korzun, Iraida V.

    2016-08-01

    Oxide solubility in chloride melts depends on temperature and composition of molten solvent. The solubility of barium oxide in the solvents with barium chloride content is essentially higher than that in molten alkali chlorides. Spectral data demonstrate the existence of oxychloride ionic groupings in such melts. This work presents the results of the BaO solubility in two molten BaCl2-NaCl systems with different barium chloride content. The received data together with earlier published results revealed the main regularities of BaO solubility in molten BaO-BaCl2-MCl systems.

  10. A Structural Molar Volume Model for Oxide Melts Part III: Fe Oxide-Containing Melts

    NASA Astrophysics Data System (ADS)

    Thibodeau, Eric; Gheribi, Aimen E.; Jung, In-Ho

    2016-04-01

    As part III of this series, the model is extended to iron oxide-containing melts. All available experimental data in the FeO-Fe2O3-Na2O-K2O-MgO-CaO-MnO-Al2O3-SiO2 system were critically evaluated based on the experimental condition. The variations of FeO and Fe2O3 in the melts were taken into account by using FactSage to calculate the Fe2+/Fe3+ distribution. The molar volume model with unary and binary model parameters can be used to predict the molar volume of the molten oxide of the Li2O-Na2O-K2O-MgO-CaO-MnO-PbO-FeO-Fe2O3-Al2O3-SiO2 system in the entire range of compositions, temperatures, and oxygen partial pressures from Fe saturation to 1 atm pressure.

  11. PROCESS OF SEPARATING PLUTONIUM VALUES BY ELECTRODEPOSITION

    DOEpatents

    Whal, A.C.

    1958-04-15

    A process is described of separating plutonium values from an aqueous solution by electrodeposition. The process consists of subjecting an aqueous 0.1 to 1.0 N nitric acid solution containing plutonium ions to electrolysis between inert metallic electrodes. A current density of one milliampere io one ampere per square centimeter of cathode surface and a temperature between 10 and 60 d C are maintained. Plutonium is electrodeposited on the cathode surface and recovered.

  12. PROCESS FOR THE RECOVERY OF PLUTONIUM

    DOEpatents

    Potratz, H.A.

    1958-12-16

    A process for the separation of plutonium from uranlum and other associated radioactlve fission products ls descrlbed conslstlng of contacting an acid solution containing plutonium in the tetravalent state and uranium in the hexavalent state with enough ammonium carbonate to form an alkaline solution, adding cupferron to selectlvely form plutonlum cupferrlde, then recoverlng the plutonium cupferride by extraction with a water lmmiscible organic solvent such as chloroform.

  13. PLUTONIUM METALLOGRAPHY AT LOS ALAMOS

    SciTech Connect

    PEREYRA, RAMIRO A.; LOVATO, DARRYL

    2007-01-08

    From early days of the Manhattan program to today, scientists and engineers have continued to investigate the metallurgical properties of plutonium (Pu). Although issues like aging was not a concern to the early pioneers, today the reliability of our aging stockpile is of major focus. And as the country moves toward a new generation of weapons similar problems that the early pioneers faced such as compatibility, homogeneity and malleability have come to the forefront. And metallography will continue to be a principle tool for the resolution of old and new issues. Standard metallographic techniques are used for the preparation of plutonium samples. The samples are first cut with a slow speed idamond saw. After mounting in Epon 815 epoxy resin, the samples are ground through 600 grit silicon carbide paper. PF 5070 (a Freon substitute) is used as a coolant, lubricant, and solvent for most operations. Rough mechanical polished is done with 9-{mu} diamond using a nap less cloth, for example nylon or cotton. Final polish is done with 1-{mu} diamond on a nappy cloth such as sylvet. Ethyl alcohol is then used ultrasonically to clean the samples before electro polishing. The sample is then electro-polished and etched in an electrolyte containing 10% nitric acid, and 90% dimethyleneformalmide. Ethyl alcohol is used as a final cleaning agent. Although standard metallographic preparation techniques are used, there are several reasons why metallography of Pu is difficult and challenging. Firstly, because of the health hazards associated with its radioactive properties, sample preparation is conducted in glove boxes. Figure 1 shows the metallography line, in an R and D facility. Since they are designed to be negative in pressure to the laboratory, cross-contamination of abrasives is a major problem. In addition, because of safety concerns and waste issues, there is a limit to the amount of solvent that can be used. Secondly, Pu will readily hydride or oxidize when in contact

  14. Plutonium story

    SciTech Connect

    Seaborg, G T

    1981-09-01

    The first nuclear synthesis and identification (i.e., the discovery) of the synthetic transuranium element plutonium (isotope /sup 238/Pu) and the demonstration of its fissionability with slow neutrons (isotope /sup 239/Pu) took place at the University of California, Berkeley, through the use of the 60-inch and 37-inch cyclotrons, in late 1940 and early 1941. This led to the development of industrial scale methods in secret work centered at the University of Chicago's Metallurgical Laboratory and the application of these methods to industrial scale production, at manufacturing plants in Tennessee and Washington, during the World War II years 1942 to 1945. The chemical properties of plutonium, needed to devise the procedures for its industrial scale production, were studied by tracer and ultramicrochemical methods during this period on an extraordinarily urgent basis. This work, and subsequent investigations on a worldwide basis, have made the properties of plutonium very well known. Its well studied electronic structure and chemical properties give it a very interesting position in the actinide series of inner transition elements.

  15. Plutonium Story

    DOE R&D Accomplishments Database

    Seaborg, G. T.

    1981-09-01

    The first nuclear synthesis and identification (i.e., the discovery) of the synthetic transuranium element plutonium (isotope /sup 238/Pu) and the demonstration of its fissionability with slow neutrons (isotope /sup 239/Pu) took place at the University of California, Berkeley, through the use of the 60-inch and 37-inch cyclotrons, in late 1940 and early 1941. This led to the development of industrial scale methods in secret work centered at the University of Chicago's Metallurgical Laboratory and the application of these methods to industrial scale production, at manufacturing plants in Tennessee and Washington, during the World War II years 1942 to 1945. The chemical properties of plutonium, needed to devise the procedures for its industrial scale production, were studied by tracer and ultramicrochemical methods during this period on an extraordinarily urgent basis. This work, and subsequent investigations on a worldwide basis, have made the properties of plutonium very well known. Its well studied electronic structure and chemical properties give it a very interesting position in the actinide series of inner transition elements.

  16. Arginine-Containing Ligands Enhance H-2 Oxidation Catalyst Performance

    SciTech Connect

    Dutta, Arnab; Roberts, John A.; Shaw, Wendy J.

    2014-06-16

    In H2 fuel cells, performance depends on factors controlling turnover frequency and energy efficiency in the electrocatalytic oxidation of H2. Nature uses the hydrogenase enzymes to oxidize H2 at high turnover frequencies (up to 20,000 s-1) and low overpotentials (<100 mV), while the fastest synthetic catalyst reported to date only oxidizes H2 at 50 s-1 under 1 atm H2. Here we report a water-soluble complex incorporating the amino acid arginine, [NiII(PCy2NArg2)2]6+, that operates at 210 s-1 (180 mV overpotential) under 1 atm H2 and 144,000 s-1 (460 mV overpotential) under 133 atm H2. The complex functions from pH 0-14 with rates increasing at lower pH values. The arginine groups impart water solubility and play a critical role in enhancing turnover frequency, most consistent with an intramolecular Arg-Arg interaction that controls the structure of the catalyst active site. This work was funded by the Office of Science Early Career Research Program through the US DOE, BES (AD, WJS), and the Center for Molecular Electrocatalysis, an Energy Frontier Research Center funded by the US DOE, BES (JASR). PNNL is operated by Battelle for the US DOE.

  17. Laboratory directed research and development on disposal of plutonium recovered from weapons. FY1994 final report

    SciTech Connect

    Pitts, J.H.; Choi, J.S.

    1994-11-14

    This research project was conceived as a multi-year plan to study the use of mixed plutonium oxide-uranium oxide (MOX) fuel in existing nuclear reactors. Four areas of investigation were originally proposed: (1) study reactor physics including evaluation of control rod worth and power distribution during normal operation and transients; (2) evaluate accidents focusing upon the reduced control rod worth and reduced physical properties of PuO{sub 2}; (3) assess the safeguards required during fabrication and use of plutonium bearing fuel assemblies; and (4) study public acceptance issues associated with using material recovered from weapons to fuel a nuclear reactor. First year accomplishments are described. Appendices contain 2 reports entitled: development and validation of advanced computational capability for MOX fueled ALWR assembly designs; and long-term criticality safety concerns associated with weapons plutonium disposition.

  18. SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS

    DOEpatents

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

    1958-10-01

    A chromatographic adsorption process is presented for the separation of plutonium from other fission products formed by the irradiation of uranium. The plutonium and the lighter element fission products are adsorbed on a sulfonated phenol-formaldehyde resin bed from a nitric acid solution containing the dissolved uranium. Successive washes of sulfuric, phosphoric, and nitric acids remove the bulk of the fission products, then an eluate of dilute phosphoric and nitric acids removes the remaining plutonium and fission products. The plutonium is selectively removed by passing this solution through zirconium phosphate, from which the plutonium is dissolved with nitric acid. This process provides a convenient and efficient means for isolating plutonium.

  19. Inert electrode containing metal oxides, copper and noble metal

    DOEpatents

    Ray, Siba P.; Woods, Robert W.; Dawless, Robert K.; Hosler, Robert B.

    2000-01-01

    A cermet composite material is made by treating at an elevated temperature a mixture comprising a compound of iron and a compound of at least one other metal, together with an alloy or mixture of copper and a noble metal. The alloy or mixture preferably comprises particles having an interior portion containing more copper than noble metal and an exterior portion containing more noble metal than copper. The noble metal is preferably silver. The cermet composite material preferably includes alloy phase portions and a ceramic phase portion. At least part of the ceramic phase portion preferably has a spinel structure.

  20. Inert electrode containing metal oxides, copper and noble metal

    DOEpatents

    Ray, Siba P.; Woods, Robert W.; Dawless, Robert K.; Hosler, Robert B.

    2001-01-01

    A cermet composite material is made by treating at an elevated temperature a mixture comprising a compound of iron and a compound of at least one other metal, together with an alloy or mixture of copper and a noble metal. The alloy or mixture preferably comprises particles having an interior portion containing more copper than noble metal and an exterior portion containing more noble metal than copper. The noble metal is preferably silver. The cermet composite material preferably includes alloy phase portions and a ceramic phase portion. At least part of the ceramic phase portion preferably has a spinel structure.

  1. Structural evolution and adhesion of titanium oxide film containing phosphorus and calcium on titanium by anodic oxidation.

    PubMed

    Lin, C S; Chen, M T; Liu, J H

    2008-05-01

    This study investigated the microstructure evolution and defects of the titanium oxide layer containing calcium (Ca) and phosphorus (P) formed by anodic oxidation in a solution containing Ca and P compounds. Results show that the anodic film exhibited a two-layer structure: a pore-containing amorphous titanium oxide layer dispersed with nano-sized crystallites formed prior to sparking, and a porous overlay dotted with craters formed after sparking. Ca and P were predominantly incorporated in the porous overlay, in which the amorphous region contained more Ca and P than the crystalline region regardless of the anodizing voltages. Moreover, the ratio of amorphous to crystalline regions in the porous overlay changed insignificantly with anodizing voltage. Increasing anodizing voltage enhanced the incorporation of Ca and P in the anodic film, but deteriorated the adhesion of the anodic film to the substrate. This deterioration was related to two inherent adhesive weaknesses: the aligned pores in the titanium oxide layer and the craters in the major overlay, signifying that a new anodic oxidation process that can produce high Ca- and P-containing oxide film at relatively-low anodizing voltages, i.e. approximately 200 V, is a necessity. PMID:17688247

  2. Enhancing collagen stability through nanostructures containing chromium(III) oxide.

    PubMed

    Sangeetha, Selvam; Ramamoorthy, Usha; Sreeram, Kalarical Janardhanan; Nair, Balachandran Unni

    2012-12-01

    Stabilization of collagen for various applications employs chemicals such as aldehydes, metal ions, polyphenols, etc. Stability against enzymatic, thermal and mechanical degradation is required for a range of biomedical applications. The premise of this research is to explore the use of nanoparticles with suitable functionalization/encapsulation to crosslink with collagen, such that the three dimensional architecture had the desired stability. Collagen solution prepared as per standard protocols is treated with chromium(III) oxide nanoparticules encapsulated within a polymeric matrix (polystyrene-block-polyacrylic acid copolymer). Selectivity towards encapsulation was ensured by the reaction in dimethyl sulfoxide, where the PS groups popped out and encapsulated the Cr(2)O(3). Subsequently when immersed in aqueous solution, PAA units popped up to react with functional groups of collagen. The interaction with collagen was monitored through techniques such as CD, FTIR, viscosity measurements, stress analysis. CD studies and FTIR showed no degradation of collagen. Thermal stability was enhanced upon interaction of nanostructures with collagen. Self-assembly of collagen was delayed but not inhibited, indicating a compete binding of the metal oxide encapsulated polymer to collagen. Metal oxide nanoparticles encapsulated within a polymeric matrix could provide thermal and mechanical stability to collagen. The formed fibrils of collagen could serve as ideal material for various smart applications such as slow/sustained drug release. The study is also relevant to the leather industry in that the nanostructures can diffuse through the highly networked collagen fibre bundles in skin matrix easily, thus overcoming the rate limiting step of diffusion. PMID:22766281

  3. Elaboration of aluminum oxide-based graphite containing castables

    NASA Astrophysics Data System (ADS)

    Zhou, Ningsheng

    The aim of this work was set to develop effective and practicable new methods to incorporate natural flake graphite (FG) into the Al2O 3 based castables for iron and steel making applications. Three approaches, viz. micro-pelletized graphite (PG), crushed briquette of Al2O3-graphite (BAG) and TiO2 coated graphite (CFG), have been developed to insert flake graphite into Al2O 3 rich Al2O3-SiC based and Al2O 3-MgO based castables. These approaches were put into effect as countermeasures against the problems caused by FG in order: (1) to agglomerate the FG powders so as to decrease the specific surface area; (2) to diminish the density difference by using crushed carbon bonded compact of oxide-FG mixture; (3) to modify the surface of the flake graphite by forming hydrophilic coating; (4) to control the dispersion state of the graphite in the castable to maintain enough bonding strength; and (5) to use appropriate antioxidants to inhibit the oxidation of FG. The whole work was divided into two stages. In stage one, Al2O 3-SiC-C castables were dealt with to compare 4 modes of inserting graphite, i.e., by PG, BAG, CFG and FG. Overall properties were measured, all in correlation with graphite amount and incorporating mode. In stage two, efforts were made to reduce water demand in the Al2O3-MgO castables system. For this purpose, the matrix portion of the castable mixes was extracted and a coaxial double cylinder viscometer was adopted to investigate rheological characteristics of the matrix slurries vs. 4 kinds of deflocculants, through which the best deflocculant and its appropriate amount were found. Efforts were then made to add up to 30% MgO into the castables, using a limited amount of powders (<0.3 mm), the rest being increased in size gradually up to the top size of 4.76 mm. Into the optimized Al2O3-MgO castables graphite was incorporated by PG and BAG, and 4 kinds of antioxidants, Si, SiC, B4C and ZrB2, were added respectively or in combination. Overall properties

  4. Interaction between Barium Oxide and Barium Containing Chloride Melt

    NASA Astrophysics Data System (ADS)

    Nikolaeva, Elena V.; Zakiryanova, Irina D.; Korzun, Iraida V.; Bovet, Andrey L.; Antonov, Boris D.

    2015-05-01

    Thermal analysis was applied to determine the liquidus temperatures in the NaCl-KCl-BaCl2-BaO system, with BaO concentration varied from 0 to 6 mole%. The temperature dependence of the BaO solubility in the NaCl-KCl-BaCl2 eutectic melt was investigated; the thermodynamic parameters of BaO dissolution were calculated. The caloric effects of melting of the NaCl-KCl-BaCl2 eutectic with barium oxide and barium oxychloride additions were studied. The type, morphology, and composition of oxychloride ionic groupings in the melt were determined in situ using Raman spectroscopy.

  5. METHOD OF REDUCING PLUTONIUM WITH FERROUS IONS

    DOEpatents

    Dreher, J.L.; Koshland, D.E.; Thompson, S.G.; Willard, J.E.

    1959-10-01

    A process is presented for separating hexavalent plutonium from fission product values. To a nitric acid solution containing the values, ferrous ions are added and the solution is heated and held at elevated temperature to convert the plutonium to the tetravalent state via the trivalent state and the plutonium is then selectively precipitated on a BiPO/sub 4/ or LaF/sub 3/ carrier. The tetravalent plutonium formed is optionally complexed with fluoride, oxalate, or phosphate anion prior to carrier precipitation.

  6. Influence of Iron Redox Transformations on Plutonium Sorption to Sediments

    SciTech Connect

    Hixon, Amy E.; Hu, Yung-Jin; Kaplan, Daniel I.; Kukkadapu, Ravi K.; Nitsche, Heino; Qafoku, Odeta; Powell, Brian A.

    2010-10-01

    Plutonium subsurface mobility is primarily controlled by its oxidation state, which in turn is loosely coupled to the oxidation state of iron in the system. Experiments were conducted to examine the effect of sediment iron mineral composition and oxidation state on plutonium sorption and oxidation state. A pH 6.3 vadose zone sediment containing iron oxides and iron-containing phyllosilicates was treated with various complexants (ammonium oxalate) and reductants (dithionite-citrate-bicarbonate) to selectively leach and/or reduce iron oxide and phyllosilicate phases. Mössbauer spectroscopy was used to identify initial iron mineral composition of the sediment and monitor dissolution and reduction of iron oxides. Sorption of Pu(V) was monitored over one week for each of six treated sediment fractions. Plutonium oxidation state speciation in the aqueous and solid phases was monitored using solvent extraction, coprecipitation, and XANES. Mössbauer spectroscopy showed that the sediment contained 25-30% hematite, 60-65% Al-goethite, and <10%Fe(III) in phyllosilicate; there was no detectable Fe(II). Upon reduction with a strong chemical reductant (dithionite-citrate buffer, DCB), much of the hematite and goethite disappeared and the Fe in the phyllosilicate reduced to Fe(II). The rate of sorption was found to correlate with the 1 fraction of Fe(II) remaining within each treated sediment phase. Pu(V) was the only oxidation state measured in the aqueous phase, irrespective of treatment, whereas Pu(IV) and much smaller amounts of Pu(V) and Pu(VI) were measured in the solid phase. Surface-mediated reduction of Pu(V) to Pu(IV) occurred in treated and untreated sediment samples; Pu(V) remained on untreated sediment surface for two days before reducing to Pu(IV). Similar to the sorption kinetics, the reduction rate was correlated with sediment Fe(II) concentration. The correlation between Fe(II) concentrations and Pu(V) reduction demonstrates the potential impact of changing

  7. Porous desulfurization sorbent pellets containing a reactive metal oxide and an inert zirconium compound

    SciTech Connect

    Gardner, Todd H.; Gasper-Galvin, Lee D.

    1996-12-01

    Sorbent pellets for removing hydrogen sulfide from coal gas are prepared by combining a reactive oxide, in particular zinc oxide, with a zirconium compound such as an oxide, silicate, or aluminate of zirconium, and an inorganic binder and pelletizing and calcining the mixture. Alternately, the zinc oxide may be replaced by copper oxide or a combination of copper, molybdenum, and manganese oxides. The pellet components may be mixed in dry form, moistened to produce a paste, and converted to pellets by forming an aqueous slurry of the components and spray drying the slurry, or the reactive oxide may be formed on existing zirconium-containing catalyst-carrier pellets by infusing a solution of a salt of the active metal onto the existing pellets and firing at a high temperature to produce the oxide. Pellets made according to this invention show a high reactivity with hydrogen sulfide and durability such as to be useful over repeated cycles of sorption and regeneration.

  8. RECOVERY OF PLUTONIUM BY CARRIER PRECIPITATION

    DOEpatents

    Goeckermann, R.H.

    1961-04-01

    A process is given for recovering plutonium from an aqueous nitric acid zirconium-containing solution of an acidity between 0.2 and 1 N by adding fluoride anions (1.5 to 5 mg/l) and precipitating the plutonium with an excess of hydrogen peroxide at from 53 to 65 deg C.

  9. FABRICATION AND CHARACTERIZATION OF BOROSILICATE GLASSES CONTAINING ALPHA-RADIONUCLIDES AND SILVER FROM CONVERSION AND MIXED-OXIDE FACILITIES PROPOSED FOR RUSSIA

    SciTech Connect

    Aloy, A; Trofimenko, V; Uspensky, A; Jardine, L

    2005-10-25

    Liquid and solid radioactive wastes are formed during conversion of plutonium metal to oxide and during fabrication of weapons-grade plutonium into mixed-oxide (MOX) fuel. In Russia, these wastes are to be processed for disposition by immobilization in either borosilicate glass or cement matrices depending upon the waste stream-specific radionuclide contents. Vitrification is planned for the liquid high-level waste raffinate stream containing the bulk of the Am-241 produced from Pu-241 decay. Previous work on the Russian MOX Fuel Fabrication Facility (R-MFFF) by the Public Joint Stock Corporation (TVEL) [1] showed that this waste stream may contain significant amounts of silver derived from the electrochemical dissolution of PuO2 using a Ag(II) catalyst. The work reported here further investigated silver solubility limits, which, if exceeded in a production glass melter, allow discrete silver grains to form in the glass and also deposit over time on the bottom of a joule-heated ceramic melter. In melters with immersed electrodes, such as the Russian EP-100 for phosphate glasses or the US Duratek DP-100 type melters for borosilicate glasses that are being considered for use at the Siberian Chemical Combine (SCC) Tomsk site, the undissolved silver could cause a short circuit and an unacceptable production melter failure. The silver solubility limit of 3.85 wt% Ag{sub 2}O in liquid, alpha-bearing wastes determined in this work will guide the production scale use of borosilicate glass compositions, and effectively increase the capacity of the ceramic melters and reduce the total volume of solidified vitrified wastes at SCC Tomsk that require storage prior to geologic disposal.

  10. Photoelastic constants of germanate glasses containing lead and bismuth oxides

    SciTech Connect

    Rabukhin, A.I.

    1995-07-01

    Regression equations which accurately approximate the concentration curves of the photoelastic constants of lead bismuth germanate glasses were obtained and the isolines of the photoelastic constants were plotted and graphically illustrate the change in the properties of the glasses in almost the entire glass-formation region of the PbO-Bi{sub 2}O{sub 3}-GeO{sub 2} system. The partial values of the photoelastic constants of the oxides, components of these glasses, were determined and are in agreement with the values established for glasses of other systems. The data obtained can be used in planning the compositions of effective optical media for fabrication of light and acoustic lines for acousto-optic instruments and glasses with a zero optical stress coefficient.

  11. Structural silicon nitride materials containing rare earth oxides

    DOEpatents

    Andersson, Clarence A.

    1980-01-01

    A ceramic composition suitable for use as a high-temperature structural material, particularly for use in apparatus exposed to oxidizing atmospheres at temperatures of 400 to 1600.degree. C., is found within the triangular area ABCA of the Si.sub.3 N.sub.4 --SiO.sub.2 --M.sub.2 O.sub.3 ternary diagram depicted in FIG. 1. M is selected from the group of Yb, Dy, Er, Sc, and alloys having Yb, Y, Er, or Dy as one component and Sc, Al, Cr, Ti, (Mg +Zr) or (Ni+Zr) as a second component, said alloy having an effective ionic radius less than 0.89 A.

  12. Vanadium oxide bronzes containing rare-earth elements

    SciTech Connect

    Volkov, V.L.; Zubkov, V.G.; Fedyukov, A.S.; Zainulin, Yu.G.

    1988-05-01

    We attempted to make phases having the general formula Ln/sub x/V/sub 2/O/sub 5/ (Ln = La, Eu, Yb) without success; the specimens usually consisted of three phases: the rare-earth orthovanadate LnVO/sub 4/, vanadium(V) oxide, and VO/sub 2/. To shift the process to give Ln/sub x/V/sub 2/O/sub 5/, heat treatment was applied to mixtures of the initial high-purity substances. The x-ray patterns were recorded with a DRON-UM1 apparatus with Cr K..cap alpha.. radiation and were processed by the Poroshok program. The IR spectra were recorded with UR-20 spectrometer with oil mulls.

  13. 15. VIEW OF THE SAFE GEOMETRY PLUTONIUM METAL STORAGE PALLETS ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    15. VIEW OF THE SAFE GEOMETRY PLUTONIUM METAL STORAGE PALLETS FROM THE INSIDE OF AN INPUT-OUTPUT STATION. INDIVIDUAL CONTAINERS OF PLUTONIUM ARE STORED IN THE WATER-FILLED, DOUBLE-WALLED STAINLESS STEEL TUBES THAT ARE WELDED ONTO THE PALLETS. (12/3/88) - Rocky Flats Plant, Plutonium Recovery Facility, Northwest portion of Rocky Flats Plant, Golden, Jefferson County, CO

  14. VIEW OF THE INTERIOR OF THE PLUTONIUM LABORATORY IN BUILDING ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    VIEW OF THE INTERIOR OF THE PLUTONIUM LABORATORY IN BUILDING 559. THE LABORATORY WAS USED TO ANALYZE THE PURITY OF PLUTONIUM. PLUTONIUM SAMPLES WERE CONTAINED WITHIN GLOVE BOXES - Rocky Flats Plant, Chemical Analytical Laboratory, North-central section of Plant, Golden, Jefferson County, CO

  15. PLUTONIUM ALLOYS

    DOEpatents

    Chynoweth, W.

    1959-06-16

    The preparation of low-melting-point plutonium alloys is described. In a MgO crucible Pu is placed on top of the lighter alloying metal (Fe, Co, or Ni) and the temperature raised to 1000 or 1200 deg C. Upon cooling, the alloy slug is broke out of the crucible. With 14 at. % Ni the m.p. is 465 deg C; with 9.5 at. % Fe the m.p. is 410 deg C; and with 12.0 at. % Co the m.p. is 405 deg C. (T.R.H.) l6262 l6263 ((((((((Abstract unscannable))))))))

  16. Pressurized solid oxide fuel cell integral air accumular containment

    DOEpatents

    Gillett, James E.; Zafred, Paolo R.; Basel, Richard A.

    2004-02-10

    A fuel cell generator apparatus contains at least one fuel cell subassembly module in a module housing, where the housing is surrounded by a pressure vessel such that there is an air accumulator space, where the apparatus is associated with an air compressor of a turbine/generator/air compressor system, where pressurized air from the compressor passes into the space and occupies the space and then flows to the fuel cells in the subassembly module, where the air accumulation space provides an accumulator to control any unreacted fuel gas that might flow from the module.

  17. Copper-containing protein from the membranes of methane oxidizing bacterium Methylococcus capsulatus (strain M) containing methane monooxygenase

    SciTech Connect

    Burbaev, D.Sh.; Moroz, I.A.; Gvozdev, R.I. |

    1994-12-31

    The goal of the present work was to study copper-containing center of the membrane of methane oxidizing bacterium, M. capsulatus, strain M, by ESR spectroscopy. The bacteria were grown and the membrane preparation particles, fraction O{sub 1} were isolated as described earlier. The results reveal that the fraction of particles mediating oxidation of CH{sub 4} includes a Cu-protein with a minimal molecular mass of 49 kD. This protein has the type 2 ESR signal characteristic of copper with nitrogen-containing ligands. Histidine residues are most probable ligands. The protein is likely to be incorporated into pMMO, although its function (electron transfer, activation of {sub 2}) is far for clear.

  18. Assessment of plutonium storage safety issues at Department of Energy facilities

    SciTech Connect

    Not Available

    1994-01-01

    The Department of Energy (DOE) mission for utilization and storage of nuclear materials has recently changed as a result of the end of the ``Cold War`` era. Past and current plutonium storage practices largely reflect a temporary, in-process, or in-use storage condition which must now be changed to accommodate longer-term storage. This report summarizes information concerning current plutonium metal and oxide storage practices which was presented at the Office of Defense programs (DP) workshop in Albuquerque, New Mexico on May 26-27, 1993 and contained in responses to questions by DP-62 from the field organizations.

  19. Plutonium stabilization and packaging system

    SciTech Connect

    1996-05-01

    This document describes the functional design of the Plutonium Stabilization and Packaging System (Pu SPS). The objective of this system is to stabilize and package plutonium metals and oxides of greater than 50% wt, as well as other selected isotopes, in accordance with the requirements of the DOE standard for safe storage of these materials for 50 years. This system will support completion of stabilization and packaging campaigns of the inventory at a number of affected sites before the year 2002. The package will be standard for all sites and will provide a minimum of two uncontaminated, organics free confinement barriers for the packaged material.

  20. The radiolysis of aqueous solutions containing dithiothreitol and oxidized dithiothreitol

    NASA Astrophysics Data System (ADS)

    John Elliot, A.; Sopchyshyn, Frederick C.

    The radiation chemistry of aqueous solutions (pH 1-9) of dithiothreitol (DTT) has been investigated by pulsed and steady-state techniques. From G( H2S) in N 2- and N 2O-saturated solutions it was concluded for pH 5-9 that the majority of the solvated electrons and 25-40% of the hydroxyl radicals reacted with DTT to form H 2S. Oxidized dithiothreitol (DTT- ox) was the other major product observed for pH 5-9. However, DTT- ox was not formed when DTT solutions were irradiated at pH 1 and 2. When nitrate ions were added as an electron scavenger over the pH range 5-9, G( DTT- ox) was not affected; at pH 9.0 G( H2S) decreased slightly but at pH 5-6, G( H2S) was reduced to that arising from the reaction of the hydroxyl radical. The reaction of CO 2- with DTT- ox was studied at pH 4.6 and pH 9.0. At pH 9.0 the G( DTT)=3.3-0.2 indicated a simple two electron reduction of DTT- ox while at pH 4.6 the G( DTT) >25 implied a chain reaction had occurred.

  1. NON-CORROSIVE PLUTONIUM FUEL SYSTEMS

    DOEpatents

    Coffinberry, A.S.; Waber, J.T.

    1962-10-23

    An improved plutonium reactor liquid fuel is described for utilization in a nuclear reactor having a tantalum fuel containment vessel. The fuel consists of plutonium and a diluent such as iron, cobalt, nickel, cerium, cerium-- iron, cerium--cobalt, cerium--nickel, and cerium--copper, and an additive of carbon and silicon. The carbon and silicon react with the tantalum container surface to form a coating that is self-healing and prevents the corrosive action of liquid plutonium on the said tantalum container. (AEC)

  2. Optical properties of borotellurite glasses containing metal oxides

    NASA Astrophysics Data System (ADS)

    Kaur, Nirmal; Khanna, Atul; Krishna, P. S. R.; González, Fernando

    2015-06-01

    Glass samples of the system: 5MxOy-20B2O3-75TeO2 : MxOy = WO3, Nb2O5, PbO, Nd2O3, Y2O3, Eu2O3 were prepared by melt quenching and characterized by X-ray diffraction, density, Differential Scanning Calorimetry, UV-visible and FTIR spectroscopy. XRD patterns confirmed the amorphous structure of all samples. Glass transition temperature was maximum in borotellurite glass containing Y2O3. Refractive index, atomic polarizability and basicity increased in the following order of ions: Y3+ < Eu3+ < Pb2+ < Nd3+ < Nb3+ < W6+. FTIR studies showed that PbO is outstanding in enhancing the concentration of tetrahedral borons in the borotellurite network.

  3. PRODUCTION OF PLUTONIUM METAL

    DOEpatents

    Lyon, W.L.; Moore, R.H.

    1961-01-17

    A process is given for producing plutonium metal by the reduction of plutonium chloride, dissolved in alkali metal chloride plus or minus aluminum chloride, with magnesium or a magnesium-aluminum alloy at between 700 and 800 deg C and separating the plutonium or plutonium-aluminum alloy formed from the salt.

  4. SEPARATION OF PLUTONIUM

    DOEpatents

    Maddock, A.G.; Smith, F.

    1959-08-25

    A method is described for separating plutonium from uranium and fission products by treating a nitrate solution of fission products, uranium, and hexavalent plutonium with a relatively water-insoluble fluoride to adsorb fission products on the fluoride, treating the residual solution with a reducing agent for plutonium to reduce its valence to four and less, treating the reduced plutonium solution with a relatively insoluble fluoride to adsorb the plutonium on the fluoride, removing the solution, and subsequently treating the fluoride with its adsorbed plutonium with a concentrated aqueous solution of at least one of a group consisting of aluminum nitrate, ferric nitrate, and manganous nitrate to remove the plutonium from the fluoride.

  5. Reactivation of a Tin-Oxide-Containing Catalyst

    NASA Technical Reports Server (NTRS)

    Hess, Robert; Sidney, Barry; Schryer, David; Miller, Irvin; Miller, George; Upchurch, Bill; Davis, Patricia; Brown, Kenneth

    2010-01-01

    The electrons in electric-discharge CO2 lasers cause dissociation of some CO2 into O2 and CO, and attach themselves to electronegative molecules such as O2, forming negative O2 ions, as well as larger negative ion clusters by collisions with CO or other molecules. The decrease in CO2 concentration due to dissociation into CO and O2 will reduce the average repetitively pulsed or continuous wave laser power, even if no disruptive negative ion instabilities occur. Accordingly, it is the primary object of this invention to extend the lifetime of a catalyst used to combine the CO and O2 products formed in a laser discharge. A promising low-temperature catalyst for combining CO and O2 is platinum on tin oxide (Pt/SnO2). First, the catalyst is pretreated by a standard procedure. The pretreatment is considered complete when no measurable quantity of CO2 is given off by the catalyst. After this standard pretreatment, the catalyst is ready for its low-temperature use in the sealed, high-energy, pulsed CO2 laser. However, after about 3,000 minutes of operation, the activity of the catalyst begins to slowly diminish. When the catalyst experiences diminished activity during exposure to the circulating gas stream inside or external to the laser, the heated zone surrounding the catalyst is raised to a temperature between 100 and 400 C. A temperature of 225 C was experimentally found to provide an adequate temperature for reactivation. During this period, the catalyst is still exposed to the circulating gas inside or external to the laser. This constant heating and exposing the catalyst to the laser gas mixture is maintained for an hour. After heating and exposing for an appropriate amount of time, the heated zone around the catalyst is allowed to return to the nominal operating temperature of the CO2 laser. This temperature normally resides in the range of 23 to 100 C. Catalyst activity can be measured as the percentage conversion of CO to CO2. In the specific embodiment

  6. Complex Phosphates, Containing Elements with Oxidation Degrees +1 and +4

    NASA Astrophysics Data System (ADS)

    Pet'kov, V. I.; Asabina, E. A.

    Crystal-chemical approach provided the scientific basis for the design of new compounds with a desirable structure and properties. In the present investigation we generalized and systematized our and literature data on synthesis, structures, thermal and physical characteristics of double and more complex anhydrous phosphates containing M+ and M4+ cations (M+ - H, NH4, Li, Na, K, Rb, Cs, Cu, Ag, Tl; M4+ - Ti, Zr, Hf, U, Th, Np, Pu). In order to study the influence of M+ and M4+ cationic radii on the stability regions of different structural types, we investigated the phase formation in the M1-x+Mx'+M24+(PO4)3 systems (M+, M'+ - Li, Na, K, Rb, Cs; M4+ - Ti, Zr, Hf) with smoothly varying compositions. The compounds and solid solutions were studied using an X-ray diffraction technique (using Rietveld method of structure refinement), electron microprobe analysis, IR-spectroscopy, adiabatic calorimetry (in the range 6-320 K) and differential scanning calorimetry (in the range 300-650 K).

  7. Optical properties of borotellurite glasses containing metal oxides

    SciTech Connect

    Kaur, Nirmal; Khanna, Atul; Krishna, P. S. R.; González, Fernando

    2015-06-24

    Glass samples of the system: 5M{sub x}O{sub y}-20B{sub 2}O{sub 3}-75TeO{sub 2} : M{sub x}O{sub y} = WO{sub 3}, Nb{sub 2}O{sub 5}, PbO, Nd{sub 2}O{sub 3}, Y{sub 2}O{sub 3}, Eu{sub 2}O{sub 3} were prepared by melt quenching and characterized by X-ray diffraction, density, Differential Scanning Calorimetry, UV-visible and FTIR spectroscopy. XRD patterns confirmed the amorphous structure of all samples. Glass transition temperature was maximum in borotellurite glass containing Y{sub 2}O{sub 3}. Refractive index, atomic polarizability and basicity increased in the following order of ions: Y{sup 3+} < Eu{sup 3+} < Pb{sup 2+} < Nd{sup 3+} < Nb{sup 3+} < W{sup 6+}. FTIR studies showed that PbO is outstanding in enhancing the concentration of tetrahedral borons in the borotellurite network.

  8. PLUTONIUM LOADING CAPACITY OF REILLEX HPQ ANION EXCHANGE COLUMN - AFS-2 PLUTONIUM FLOWSHEET FOR MOX

    SciTech Connect

    Kyser, E.; King, W.; O'Rourke, P.

    2012-07-26

    Radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the dependence of column loading performance on the feed composition in the H-Canyon dissolution process for plutonium oxide (PuO{sub 2}) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). These loading experiments show that a representative feed solution containing {approx}5 g Pu/L can be loaded onto Reillex{trademark} HPQ resin from solutions containing 8 M total nitrate and 0.1 M KF provided that the F is complexed with Al to an [Al]/[F] molar ratio range of 1.5-2.0. Lower concentrations of total nitrate and [Al]/[F] molar ratios may still have acceptable performance but were not tested in this study. Loading and washing Pu losses should be relatively low (<1%) for resin loading of up to 60 g Pu/L. Loading above 60 g Pu/L resin is possible, but Pu wash losses will increase such that 10-20% of the additional Pu fed may not be retained by the resin as the resin loading approaches 80 g Pu/L resin.

  9. STRIPPING PROCESS FOR PLUTONIUM

    DOEpatents

    Kolodney, M.

    1959-10-01

    A method for removing silver, nickel, cadmium, zinc, and indium coatings from plutonium objects while simultaneously rendering the plutonium object passive is described. The coated plutonium object is immersed as the anode in an electrolyte in which the plutonium is passive and the coating metal is not passive, using as a cathode a metal which does not dissolve rapidly in the electrolyte. and passing an electrical current through the electrolyte until the coating metal is removed from the plutonium body.

  10. Interaction between stainless steel and plutonium metal

    SciTech Connect

    Dunwoody, John T; Mason, Richard E; Freibert, Franz J; Willson, Stephen P; Veirs, Douglas K; Worl, Laura A; Archuleta, Alonso; Conger, Donald J

    2010-01-01

    Long-term storage of excess plutonium is of great concern in the U.S. as well as abroad. The current accepted configuration involves intimate contact between the stored material and an iron-bearing container such as stainless steel. While many safety scenario studies have been conducted and used in the acceptance of stainless steel containers, little information is available on the physical interaction at elevated temperatures between certain forms of stored material and the container itself. The bulk of the safety studies has focused on the ability of a package to keep the primary stainless steel containment below the plutonium-iron eutectic temperature of approximately 410 C. However, the interactions of plutonium metal with stainless steel have been of continuing interest. This paper reports on a scoping study investigating the interaction between stainless steel and plutonium metal in a pseudo diffusion couple at temperatures above the eutectic melt-point.

  11. Oxidized In-containing III-V(100) surfaces: Formation of crystalline oxide films and semiconductor-oxide interfaces

    NASA Astrophysics Data System (ADS)

    Punkkinen, M. P. J.; Laukkanen, P.; Lång, J.; Kuzmin, M.; Tuominen, M.; Tuominen, V.; Dahl, J.; Pessa, M.; Guina, M.; Kokko, K.; Sadowski, J.; Johansson, B.; Väyrynen, I. J.; Vitos, L.

    2011-05-01

    Previously found oxidized III-V semiconductor surfaces have been generally structurally disordered and useless for applications. We disclose a family of well-ordered oxidized InAs, InGaAs, InP, and InSb surfaces found by experiments. The found epitaxial oxide-III-V interface is insulating and free of defects related to the harmful Fermi-level pinning, which opens up new possibilities to develop long-sought III-V metal-oxide-semiconductor transistors. Calculations reveal that the early stages in the oxidation process include only O-III bonds due to the geometry of the III-V(100)c(8×2) substrate, which is responsible for the formation of the ordered interface. The found surfaces provide a different platform to study the oxidation and properties of oxides, e.g., the origins of the photoemission shifts and electronic structures, using surface science methods.

  12. Structural Properties and Charge Distribution of the Sodium Uranium, Neptunium, and Plutonium Ternary Oxides: A Combined X-ray Diffraction and XANES Study.

    PubMed

    Smith, Anna L; Martin, Philippe; Prieur, Damien; Scheinost, Andreas C; Raison, Philippe E; Cheetham, Anthony K; Konings, Rudy J M

    2016-02-15

    The charge distributions in α-Na2UO4, Na3NpO4, α-Na2NpO4, Na4NpO5, Na5NpO6, Na2PuO3, Na4PuO5, and Na5PuO6 are investigated in this work using X-ray absorption near-edge structure (XANES) spectroscopy at the U-L3, Np-L3, and Pu-L3 edges. In addition, a Rietveld refinement of monoclinic Na2PuO3, in space group C2/c, is reported for the first time, and the existence of the isostructural Na2NpO3 phase is revealed. In contrast to measurements in solution, the number of published XANES data for neptunium and plutonium solid phases with a valence state higher than IV is very limited. The present results cover a wide range of oxidation states, namely, IV to VII, and can serve as reference for future investigations. The sodium actinide series show a variety of local coordination geometries, and correlations between the shape of the XANES spectra and the local structural environments are discussed herein. PMID:26835549

  13. Development of quality control data base ``DANTE'' and relational analysis between fission gas release behavior and plutonium homogeneity in mixed oxide fuel by DANTE

    NASA Astrophysics Data System (ADS)

    Takahashi, K.; Kamimura, K.; Yamaguchi, T.; Masuda, S.

    1988-04-01

    The PNC Plutonium Fabrication Facility (PPFF) of Tokai Works of the "Power Reactor and Nuclear Fuel Development Corporation" (PNC) has produced several types of mixed oxide fuels which have been used as the fuels for the Advanced Thermal Reactor FUGEN, and the Experimental Fast Reactor JOYO. Accumulated production of MOX fuel is about 58000 pins as of December, 1986. PPFF production technique is so reliable that no failure of driver fuel pins has been found in the FUGEN and JOYO. This is based on the adequate process control and quality control. In this paper the outline of quality control method applied to MOX fuels in our facility will be reviewed. The discussion is done mainly on the viewpoint of the software system including data gathering, data analysis and data storage, etc. The code system, Data Analysing Technology for Engineers, DANTE, has been developed for this purpose. The DANTE code system is consisted of real time system having relational data base. It is possible to refer the data base relationally and to make tables and figures of the analyzed results on real time by using the DANTE code system. The DANTE code system contributes to quality assurance in the MOX Fuel production facility by evaluating overall data which come from each process such as production and inspection processes.

  14. Plutonium waste incineration using pyrohydrolysis

    SciTech Connect

    Meyer, M.L.

    1991-01-01

    Waste generated by Savannah River Site (SRS) plutonium operations includes a contaminated organic waste stream. A conventional method for disposing of the organic waste stream and recovering the nuclear material is by incineration. When the organic material is burned, the plutonium remains in the incinerator ash. Plutonium recovery from incinerator ash is highly dependent on the maximum temperature to which the oxide is exposed. Recovery via acid leaching is reduced for a high fired ash (>800{degree}C), while plutonium oxides fired at lower decomposition temperatures (400--800{degrees}C) are more soluble at any given acid concentration. To determine the feasibility of using a lower temperature process, tests were conducted using an electrically heated, controlled-air incinerator. Nine nonradioactive, solid, waste materials were batch-fed and processed in a top-heated cylindrical furnace. Waste material processing was completed using a 19-liter batch over a nominal 8-hour cycle. A processing cycle consisted of 1 hour for heating, 4 hours for reacting, and 3 hours for chamber cooling. The water gas shift reaction was used to hydrolyze waste materials in an atmosphere of 336% steam and 4.4% oxygen. Throughput ranged from 0.14 to 0.27 kg/hr depending on the variability in the waste material composition and density.

  15. Plutonium waste incineration using pyrohydrolysis

    SciTech Connect

    Meyer, M.L.

    1991-12-31

    Waste generated by Savannah River Site (SRS) plutonium operations includes a contaminated organic waste stream. A conventional method for disposing of the organic waste stream and recovering the nuclear material is by incineration. When the organic material is burned, the plutonium remains in the incinerator ash. Plutonium recovery from incinerator ash is highly dependent on the maximum temperature to which the oxide is exposed. Recovery via acid leaching is reduced for a high fired ash (>800{degree}C), while plutonium oxides fired at lower decomposition temperatures (400--800{degrees}C) are more soluble at any given acid concentration. To determine the feasibility of using a lower temperature process, tests were conducted using an electrically heated, controlled-air incinerator. Nine nonradioactive, solid, waste materials were batch-fed and processed in a top-heated cylindrical furnace. Waste material processing was completed using a 19-liter batch over a nominal 8-hour cycle. A processing cycle consisted of 1 hour for heating, 4 hours for reacting, and 3 hours for chamber cooling. The water gas shift reaction was used to hydrolyze waste materials in an atmosphere of 336% steam and 4.4% oxygen. Throughput ranged from 0.14 to 0.27 kg/hr depending on the variability in the waste material composition and density.

  16. Calibration of the Lawrence Livermore National Laboratory Passive-Active Neutron Drum Shuffler for Measurement of Highly Enriched Uranium in Oxides within DOE-STD-3013-2000 Containers

    SciTech Connect

    Mount, M E; O'Connell, W J

    2005-06-03

    Lawrence Livermore National Laboratory (LLNL) uses the LLNL passive-active neutron drum (PAN) shuffler (Canberra Model JCC-92) for accountability measurement of highly enriched uranium (HEU) oxide and HEU in mixed uranium-plutonium (U-Pu) oxide. In June 2002, at the 43rd Annual Meeting of the Institute of Nuclear Material Management, LLNL reported on an extensive effort to calibrate this shuffler, based on standards measurements and extensive simulations, for HEU oxides and mixed U-Pu oxides in thin-walled primary and secondary containers. In August 2002, LLNL began to also use DOE-STD-3013-2000 containers for HEU oxide and mixed U-Pu oxide. These DOE-STD-3013-2000 containers are comprised of a stainless steel convenience can enclosed in welded stainless steel primary and secondary containers. Compared to the double thin-walled containers, the DOE-STD-3013-2000 containers have substantially thicker walls, and the density of materials in these containers was found to extend over a greater range (1.35 g/cm{sup 3} to 4.62 g/cm{sup 3}) than foreseen for the double thin-walled containers. Further, the DOE-STD-3013-2000 Standard allows for oxides containing at least 30 wt% Pu plus U whereas the calibration algorithms for thin-walled containers were derived for virtually pure HEU or mixed U-Pu oxides. An initial series of Monte Carlo simulations of the PAN shuffler response to given quantities of HEU oxide and mixed U-Pu oxide in DOE-STD-3013-2000 containers was generated and compared with the response predicted by the calibration algorithms for thin-walled containers. Results showed a decrease on the order of 10% in the count rate, and hence a decrease in the calculated U mass for measured unknowns, with some varying trends versus U mass. Therefore a decision was made to develop a calibration algorithm for the PAN shuffler unique to the DOE-STD-3013-2000 container. This paper describes that effort and selected unknown item measurement results.

  17. Selecting a plutonium vitrification process

    SciTech Connect

    Jouan, A.

    1996-05-01

    Vitrification of plutonium is one means of mitigating its potential danger. This option is technically feasible, even if it is not the solution advocated in France. Two situations are possible, depending on whether or not the glass matrix also contains fission products; concentrations of up to 15% should be achievable for plutonium alone, whereas the upper limit is 3% in the presence of fission products. The French continuous vitrification process appears to be particularly suitable for plutonium vitrification: its capacity is compatible with the required throughout, and the compact dimensions of the process equipment prevent a criticality hazard. Preprocessing of plutonium metal, to convert it to PuO{sub 2} or to a nitric acid solution, may prove advantageous or even necessary depending on whether a dry or wet process is adopted. The process may involve a single step (vitrification of Pu or PuO{sub 2} mixed with glass frit) or may include a prior calcination step - notably if the plutonium is to be incorporated into a fission product glass. It is important to weigh the advantages and drawbacks of all the possible options in terms of feasibility, safety and cost-effectiveness.

  18. International shipment of plutonium by air

    SciTech Connect

    Mercado, J.E.; McGrogan, J.P.

    1995-05-01

    In support of the United States (US) Government`s decision to place excess plutonium oxide at the US Department of Energy`s (DOE) Hanford Site under International Atomic Energy Agency (IAEA) safeguards, the Department of State notified the Congress that a plutonium storage vault at the Plutonium Finishing Plant at the Hanford Site would be added to the eligible facilities list. As part of the preparations to transfer the plutonium oxide under IAEA safeguards, samples of the powder were taken from the inventory to be shipped to the IAEA headquarters in Vienna, Austria, for laboratory analysis. The analysis of these samples was of high priority, and the IAEA requested that the material be shipped by aircraft, the most expeditious method.

  19. SEPARATION OF URANIUM, PLUTONIUM, AND FISSION PRODUCTS

    DOEpatents

    Spence, R.; Lister, M.W.

    1958-12-16

    Uranium and plutonium can be separated from neutron-lrradiated uranium by a process consisting of dissolvlng the lrradiated material in nitric acid, saturating the solution with a nitrate salt such as ammonium nitrate, rendering the solution substantially neutral with a base such as ammonia, adding a reducing agent such as hydroxylamine to change plutonium to the trivalent state, treating the solution with a substantially water immiscible organic solvent such as dibutoxy diethylether to selectively extract the uranium, maklng the residual aqueous solutlon acid with nitric acid, adding an oxidizing agent such as ammonlum bromate to oxidize the plutonium to the hexavalent state, and selectlvely extracting the plutonium by means of an immlscible solvent, such as dibutoxy dlethyletber.

  20. PLUTONIUM CARRIER METATHESIS WITH ORGANIC REAGENT

    DOEpatents

    Thompson, S.G.

    1958-07-01

    A method is described for converting a plutonium containing bismuth phosphate carrier precipitate Into a compositton more readily soluble in acid. The method consists of dissolving the bismuth phosphate precipitate in an aqueous solution of alkali metal hydroxide, and adding one of a certaia group of organic compounds, e.g., polyhydric alcohols or a-hydrorycarboxylic acids. The mixture is then heated causiing formation of a bismuth hydroxide precipitate containing plutonium which may be readily dissolved in nitric acid for further processing.

  1. Plutonium-238 processing at Savannah River Plant

    SciTech Connect

    Burney, G.A.

    1983-01-01

    Plutonium-238 is produced by irradiating NpO/sub 2/-Al cermet slugs or tubes with neutrons. The neptunium-237 is produced as a by-product when natural or enriched uranium is irradiated with neutrons. The neptunium is separated by solvent extraction and ion exchange and precipitated as neptunium oxalate. Neptunium oxalate is calcined to neptunium oxide and fabricated into targets for irradiation. The irradiation conditions are controlled to produce plutonium with 80 to 90 wt % /sup 238/Pu.

  2. FORM AND AGING OF PLUTONIUM IN SAVANNAH RIVER SITE WASTE TANK 18

    SciTech Connect

    Hobbs, D.

    2012-02-24

    This report provides a summary of the effects of aging on and the expected forms of plutonium in Tank 18 waste residues. The findings are based on available information on the operational history of Tank 18, reported analytical results for samples taken from Tank 18, and the available scientific literature for plutonium under alkaline conditions. These findings should apply in general to residues in other waste tanks. However, the operational history of other waste tanks should be evaluated for specific conditions and unique operations (e.g., acid cleaning with oxalic acid) that could alter the form of plutonium in heel residues. Based on the operational history of other tanks, characterization of samples from the heel residues in those tanks would be appropriate to confirm the form of plutonium. During the operational period and continuing with the residual heel removal periods, Pu(IV) is the dominant oxidation state of the plutonium. Small fractions of Pu(V) and Pu(VI) could be present as the result of the presence of water and the result of reactions with oxygen in air and products from the radiolysis of water. However, the presence of Pu(V) would be transitory as it is not stable at the dilute alkaline conditions that currently exists in Tank 18. Most of the plutonium that enters Savannah River Site (SRS) high-level waste (HLW) tanks is freshly precipitated as amorphous plutonium hydroxide, Pu(OH){sub 4(am)} or hydrous plutonium oxide, PuO{sub 2(am,hyd)} and coprecipitated within a mixture of hydrous metal oxide phases containing metals such as iron, aluminum, manganese and uranium. The coprecipitated plutonium would include Pu{sup 4+} that has been substituted for other metal ions in crystal lattice sites, Pu{sup 4+} occluded within hydrous metal oxide particles and Pu{sup 4+} adsorbed onto the surface of hydrous metal oxide particles. The adsorbed plutonium could include both inner sphere coordination and outer sphere coordination of the plutonium. PuO{sub 2

  3. SEPARATION OF PLUTONIUM FROM LANTHANUM BY CHELATION-EXTRACTION

    DOEpatents

    James, R.A.; Thompson, S.G.

    1958-12-01

    Plutonium can be separated from a mixture of plutonlum and lanthanum in which the lanthanum to plutonium molal ratio ls at least five by adding the ammonium salt of N-nitrosoarylhydroxylamine to an aqueous solution having a pH between about 3 and 0.2 and containing the plutonium in a valence state of at least +3, to form a plutonium chelate compound of N-nitrosoarylhydroxylamine. The plutonium chelate compound may be recovered from the solution by extracting with an immiscible organic solvent such as chloroform.

  4. Synthesis and characterization of hybrid materials containing iron oxide for removal of sulfides from water.

    PubMed

    Jacukowicz-Sobala, Irena; Wilk, Łukasz J; Drabent, Krzysztof; Kociołek-Balawejder, Elżbieta

    2015-12-15

    Hybrid materials containing iron oxides based on macroporous and gel-type sulfonic and carboxylic cation exchangers as supporting materials were obtained. Multiple factors, including the kind of functional groups, ion exchange capacity, and polymer matrix type (chemical constitution and porous structure), affected the amount of iron oxides introduced into their matrix (7.8-35.2% Fe). Products containing the highest iron content were obtained using carboxylic cation exchangers, with their inorganic deposit being mostly a mixture of iron(III) oxides, including maghemite. Obtained hybrid polymers were used for removal of sulfides from anoxic aqueous solutions (50-200mgS(2-)/dm(3)). The research showed that the form (Na(+) or H(+)) of ionic groups of hybrid materials had a crucial impact on the sulfide removal process. Due to high iron oxide content (35% Fe), advantageous chemical constitution and porous structure, the highest removal efficiency (60mgS(2-)/g) was exhibited by a hybrid polymer obtained using a macroporous carboxylic cation exchanger as the host material. The process of sulfide removal was very complex and proceeded with heterogeneous oxidation, iron(III) oxide reductive dissolution and formation of sulfide oxidation and precipitation products such as iron(II) sulfides, thiosulfates and polysulfides. PMID:26319332

  5. [Oxidation of sulfur-containing substrates by aboriginal and experimentally designed microbial communities].

    PubMed

    Pivovarova, T A; Bulaev, A G; Roshchupko, P V; Belyĭ, A V; Kondrat'eva, T F

    2012-01-01

    Aboriginal and experimental (constructed of pure microbial cultures) communities of acidophilic chemolithotrophs have been studied. The oxidation of elemental sulfur, sodium thiosulfate, and potassium tetrathionate as sole sources of energy has been monitored. The oxidation rate of the experimental community is higher as compared to the aboriginal community isolated from a flotation concentrate of pyrrhotine-containing pyrite-arsenopyrite gold-arsenic sulfide ore. The degree of oxidation of the mentioned S substrates amounts to 17.91, 68.30, and 93.94% for the experimental microbial community and to 10.71, 56.03, and 79.50% for the aboriginal community, respectively. The degree of oxidation of sulfur sulfide forms in the ore flotation concentrate is 59.15% by the aboriginal microbial community and 49.40% by the experimental microbial community. Despite a higher rate of oxidation of S substrates as a sole source of energy by the experimental microbial community, the aboriginal community oxidizes S substrates at a higher rate in the flotation concentrate of pyrrhotine-containing pyrite-arsenopyrite gold-arsenic sulfide ore, from which it was isolated. Bacterial-chemical oxidation of the flotation concentrate by the aboriginal microbial community allows for the extraction of an additional 32.3% of gold from sulfide minerals, which is by 5.7% larger compared to the yield obtained by the experimental microbial community. PMID:23330391

  6. Mechanisms of plutonium sorption to mineral oxide surfaces: new insights with implications for colloid-enhanced migration

    SciTech Connect

    Schwantes, J. M.; Santschi, Peter H.

    2010-11-01

    New equilibrium and kinetic models have been developed to describe rate-limited sorption and desorption of Pu onto and off of mineral oxide surfaces using a generic approach to estimate sorption constants that require minimal laboratory calibrations. Equilibrium reactions describing a total of six surface species were derived from a combination of empirical relationships previously described in the literature and generated as part of this work.

  7. Microstructure, mechanical and oxidation behavior of RE-containing lead-free solders

    NASA Astrophysics Data System (ADS)

    Dudek, Martha A.

    Pb-free solders pose new challenges associated with their incorporation and reliability during service of electronic components. Recently, a new class of alloys containing rare-earth (RE) elements has been discovered. In this study, solder alloys containing lanthanum (La), cerium (Ce) and yttrium (Y) were developed and characterized. It was found that small additions of La and Ce to Sn-Ag-Cu alloys significantly improved their ductility. This has direct implications for mechanical shock and drop reliability. Microstructure characterization of solder and solder/Cu joints containing Sn-3.9Ag-0.7Cu-XRE (X = 0, 0.1, 0.5 and 2 weight percent) was conducted using optical microscopy, scanning electron microscopy and transmission electron microscopy. It was found that RE elements refined the solder microstructure. A serial-sectioning 3D reconstruction process was used to visualize the RE-containing intermetallics. Solidification of these alloys was studied using differential scanning calorimetry. The melting point of Sn-3.9Ag-0.7Cu did not change with the incorporation of RE elements. Additionally, the effect of RE content on shear and creep of lap-shear joints was studied. It was found that additions of La and Ce up to 0.5 weight percent improved the elongation of Sn-3.9Ag-0.7Cu. Y-containing alloys did not show an improvement. Creep tests were conducted at 60, 95 and 120 degrees Celsius. RE content did not markedly alter the creep behavior. Due to RE's high affinity for oxygen, oxidation of RE-containing alloys may affect their mechanical performance. Thus, the effect of 2 weight percent Ce, La or Y on the oxidation behavior was studied at 60, 95 and 130 degrees Celsius. All alloys exhibited parabolic oxidation kinetics. La and Y-containing alloys oxidized significantly faster than the Ce-containing alloy. Sn whiskering was observed to take place during oxidation, likely due to the compressive stresses developed during oxidation. Serial-sectioning with a focused ion beam

  8. PREPARATION OF PLUTONIUM TRIFLUORIDE

    DOEpatents

    Burger, L.L.; Roake, W.E.

    1961-07-11

    A process of producing plutonium trifluoride by reacting dry plutonium(IV) oxalate with chlorofluorinated methane or ethane at 400 to 450 deg C and cooling the product in the absence of oxygen is described.

  9. PLUTONIUM RECOVERY FROM NEUTRON-BOMBARDED URANIUM FUEL

    DOEpatents

    Moore, R.H.

    1962-04-10

    A process of recovering plutonium from neutronbombarded uranium fuel by dissolving the fuel in equimolar aluminum chloride-potassium chloride; heating the mass to above 700 deg C for decomposition of plutonium tetrachloride to the trichloride; extracting the plutonium trichloride into a molten salt containing from 40 to 60 mole % of lithium chloride, from 15 to 40 mole % of sodium chloride, and from 0 to 40 mole % of potassium chloride or calcium chloride; and separating the layer of equimolar chlorides containing the uranium from the layer formed of the plutonium-containing salt is described. (AEC)

  10. Magnetic iron oxides in the cementation technology of the boron-containing radioactive waste

    NASA Astrophysics Data System (ADS)

    Fedotov, M. A.; Gorbunova, O. A.; Fedorova, O. V.; Folmanis, G. E.; Kovalenko, L. V.

    2015-04-01

    Two ways of synthesis of non-detachable dispersed particles of magnetic materials useful for the boron-containing waste cementation process regulation were developed. Powder XRD showed that the method of carbothermic recovery of nanoscale iron hydroxide allows obtaining a mixture of iron oxides with content of the magnetic phase up to 70%. Method of low-temperature hydrogen reduction of the raw materials allows obtaining various compositions of a-iron and iron oxides with the possibility to change the size of the final particles in a wide range. The possibility of using composites of magnetic iron oxides and metal oxide compositions instead of ferromagnetic rods with VEP of boron-containing liquid radioactive waste in the fluidized field was studied. It was shown that the use of fine and nano particles of the iron oxides in the pre-treatment of the boron-containing LRW increases the strength of the final compounds and accelerates the cement setting compounds from 13 to 5-9 days.

  11. Thermal behavior of Ni- and Cu-containing plasma electrolytic oxide coatings on titanium

    NASA Astrophysics Data System (ADS)

    Rudnev, V. S.; Wybornov, S.; Lukiyanchuk, I. V.; Staedler, T.; Jiang, X.; Ustinov, A. Yu.; Vasilyeva, M. S.

    2012-09-01

    In this work the effect of thermal annealing on the surface composition, structure and catalytic activity in CO oxidation of NiO + CuO/TiO2/Ti composites is studied. The composites have been obtained by a plasma electrolytic oxidation (PEO) technique, followed by impregnation in a solution of nickel (II) and copper (II) salts and air annealing. The structures contain ˜20 at% Ni and ˜12 at% Cu. It has been shown that the additional air annealing of such structures at temperature above 750 °C results in phosphate crystallization in the coatings and decreasing of Cu concentration in the surface layers. A growth of filiform nanocrystals containing mainly oxygen compounds of nickel and titanium on the coating surface takes place at the temperatures above 700 °C. The nanocrystals have a diameter of 50-200 nm and lengths below 10 μm. Such changes result in decreasing of catalytic activity of the composites in CO oxidation. At the same time the ascertained regularities may be of interest for obtaining the Ni-containing oxide catalysts with an extended surface, perspective for usage in organic catalysis or for preparing oxide nanofibers.

  12. Plutonium immobilization -- Can loading

    SciTech Connect

    Kriikku, E.

    2000-02-17

    The Savannah River Site (SRS) will immobilize excess plutonium in the proposed Plutonium Immobilization Project (PIP). The PIP adds the excess plutonium to ceramic pucks, loads the pucks into cans, and places the cans into DWPF canisters. This paper discusses the PIP process steps, the can loading conceptual design, can loading equipment design, and can loading work completed.

  13. PLUTONIUM-THORIUM ALLOYS

    DOEpatents

    Schonfeld, F.W.

    1959-09-15

    New plutonium-base binary alloys useful as liquid reactor fuel are described. The alloys consist of 50 to 98 at.% thorium with the remainder plutonium. The stated advantages of these alloys over unalloyed plutonium for reactor fuel use are easy fabrication, phase stability, and the accompanying advantuge of providing a means for converting Th/sup 232/ into U/sup 233/.

  14. Oxidation states, geometries, and electronic structures of plutonium tetroxide PuO4 isomers: is octavalent Pu viable?

    PubMed

    Huang, Wei; Xu, Wen-Hua; Su, Jing; Schwarz, W H E; Li, Jun

    2013-12-16

    In neutral chemical compounds, the highest known oxidation state of all elements in the Periodic Table is +VIII. While PuO4 is viewed as an exotic Pu(+VIII) complex, we have shown here that no stable electronic homologue of octavalent RuO4 and OsO4 exists for PuO4, even though Pu has the same number of eight valence electrons as Ru and Os. Using quantum chemical approaches at the levels of quasi-relativistic DFT, MP2, CCSD(T), and CASPT2, we find the ground state of PuO4 as a quintet (5)C2v-(PuO2)(+)(O2)(-) complex with the leading valence configuration of an (f(3))plutonyl(V) unit, loosely coupled to a superoxido (π*(3))O2(-) ligand. This stable isomer is likely detectable as a transient species, while the previously suggested planar (1)D4h-Pu(VIII)O4 isomer is only metastable. Through electronic structure analyses, the bonding and the oxidation states are explained and rationalized. We have predicted the characteristics of the electronic and vibrational spectra to assist future experimental identification of (PuO2)(+)(O2)(-) by IR, UV-vis, and ionization spectroscopy. PMID:24274785

  15. Syntheses of hydroxamic acid-containing bicyclic β-lactams via palladium-catalyzed oxidative amidation of alkenes.

    PubMed

    Jobbins, Maria O; Miller, Marvin J

    2014-02-21

    Palladium-catalyzed oxidative amidation has been used to synthesize hydroxamic acid-containing bicyclic β-lactam cores. Oxidative cleavage of the pendant alkene provides access to the carboxylic acid in one step. PMID:24483144

  16. Preparation of oxygen-containing organic products from bed-oxidized brown coal by ozonation

    SciTech Connect

    Semenova, S.A.; Patrakov, Y.F.; Batina, M.V.

    2009-01-15

    The possibility of modifying the functional composition of humic acids by gas-phase ozonation of bed-oxidized brown coal was examined. About 90% of the organic matter of brown coal was converted to low-molecular weight soluble oxygen-containing products by stepwise liquid-phase ozonation (in chloroform and acetic acid).

  17. Lignin Peroxidase Oxidation of Aromatic Compounds in Systems Containing Organic Solvents

    PubMed Central

    Vazquez-Duhalt, Rafael; Westlake, Donald W. S.; Fedorak, Phillip M.

    1994-01-01

    Lignin peroxidase from Phanerochaete chrysosporium was used to study the oxidation of aromatic compounds, including polycyclic aromatic hydrocarbons and heterocyclic compounds, that are models of moieties of asphaltene molecules. The oxidations were done in systems containing water-miscible organic solvents, including methanol, isopropanol, N, N-dimethylformamide, acetonitrile, and tetrahydrofuran. Of the 20 aromatic compounds tested, 9 were oxidized by lignin peroxidase in the presence of hydrogen peroxide. These included anthracene, 1-, 2-, and 9-methylanthracenes, acenaphthene, fluoranthene, pyrene, carbazole, and dibenzothiophene. Of the compounds studied, lignin peroxidase was able to oxidize those with ionization potentials of <8 eV (measured by electron impact). The reaction products contain hydroxyl and keto groups. In one case, carbon-carbon bond cleavage, yielding anthraquinone from 9-methylanthracene, was detected. Kinetic constants and stability characteristics of lignin peroxidase were determined by using pyrene as the substrate in systems containing different amounts of organic solvent. Benzyl alkylation of lignin peroxidase improved its activity in a system containing water-miscible organic solvent but did not increase its resistance to inactivation at high solvent concentrations. PMID:16349176

  18. Lignin peroxidase oxidation of aromatic compounds in systems containing organic solvents.

    PubMed

    Vazquez-Duhalt, R; Westlake, D W; Fedorak, P M

    1994-02-01

    Lignin peroxidase from Phanerochaete chrysosporium was used to study the oxidation of aromatic compounds, including polycyclic aromatic hydrocarbons and heterocyclic compounds, that are models of moieties of asphaltene molecules. The oxidations were done in systems containing water-miscible organic solvents, including methanol, isopropanol, N, N-dimethylformamide, acetonitrile, and tetrahydrofuran. Of the 20 aromatic compounds tested, 9 were oxidized by lignin peroxidase in the presence of hydrogen peroxide. These included anthracene, 1-, 2-, and 9-methylanthracenes, acenaphthene, fluoranthene, pyrene, carbazole, and dibenzothiophene. Of the compounds studied, lignin peroxidase was able to oxidize those with ionization potentials of <8 eV (measured by electron impact). The reaction products contain hydroxyl and keto groups. In one case, carbon-carbon bond cleavage, yielding anthraquinone from 9-methylanthracene, was detected. Kinetic constants and stability characteristics of lignin peroxidase were determined by using pyrene as the substrate in systems containing different amounts of organic solvent. Benzyl alkylation of lignin peroxidase improved its activity in a system containing water-miscible organic solvent but did not increase its resistance to inactivation at high solvent concentrations. PMID:16349176

  19. Direct detection of nitrotyrosine-containing proteins using an aniline-based oxidative coupling strategy.

    PubMed

    Sangsuwan, Rapeepat; Obermeyer, Allie C; Tachachartvanich, Phum; Palaniappan, Krishnan K; Francis, Matthew B

    2016-08-21

    A convenient two-step method is described for the detection of nitrotyrosine-containing proteins. First, nitrotyrosines are reduced to aminophenols using sodium dithionite. Following this, an oxidative coupling reaction is used to attach anilines bearing fluorescence reporters or affinity probes. Features of this approach include fast reaction times, pmol-level sensitivity, and excellent chemoselectivity. PMID:27447346

  20. DEVELOPMENT OF GLASS AND CRYSTALLINE CERAMIC FORMS FOR DISPOSITION OF EXCESS PLUTONIUM

    SciTech Connect

    Marra, James; Cozzi, A; Crawford, C.; Herman, C.; Marra, John; Peeler, D.

    2009-09-10

    In the aftermath of the Cold War, the United States Department of Energy (DOE) has identified up to 50 metric tons of excess plutonium that needs to be dispositioned. The bulk of the material is slated to be blended with uranium and fabricated into a Mixed Oxide (MOX) fuel for subsequent burning in commercial nuclear reactors. Excess plutonium-containing impurity materials making it unsuitable for fabrication into MOX fuel will need to be dispositioned via other means. Glass and crystalline ceramics have been developed and studied as candidate forms to immobilize these impure plutonium feeds. A titanate-based ceramic was identified as an excellent actinide material host. This composition was based on Synroc compositions previously developed for nuclear waste immobilization. These titanate ceramics were found to be able to accommodate extremely high quantities of fissile material and exhibit excellent aqueous durability. A lanthanide borosilicate (LaBS) glass was developed to accommodate high concentrations of plutonium and to be very tolerant of impurities yet still maintain good aqueous durability. Recent testing of alkali borosilicate compositions showed promise of using these compositions to disposition lower concentrations of plutonium using existing high level waste vitrification processes. The developed waste forms all appear to be suitable for Pu disposition. Depending on the actual types and concentrations of the Pu residue streams slated for disposition, each waste form offers unique advantages.

  1. Plutonium: The first 50 years. United States plutonium production, acquisition, and utilization from 1944 through 1994

    SciTech Connect

    1996-02-01

    The report contains important newly declassified information regarding the US production, acquisition, and removals of plutonium. This new information, when combined with previously declassified data, has allowed the DOE to issue, for the first time, a truly comprehensive report on the total DOE plutonium inventory. At the December 7, 1993, Openness Press Conference, the DOE declassified the plutonium inventories at eight locations totaling 33.5 metric tons (MT). This report declassifies the remainder of the DOE plutonium inventory. Newly declassified in this report is the quantity of plutonium at the Pantex Site, near Amarillo, Texas, and in the US nuclear weapons stockpile of 66.1 MT, which, when added to the previously released inventory of 33.5 MT, yields a total plutonium inventory of 99.5 MT. This report will document the sources which built up the plutonium inventory as well as the transactions which have removed plutonium from that inventory. This report identifies four sources that add plutonium to the DOE/DoD inventory, and seven types of transactions which remove plutonium from the DOE/DoD inventory. This report also discusses the nuclear material control and accountability system which records all nuclear material transactions, compares records with inventory and calculates material balances, and analyzes differences to verify that nuclear materials are in quantities as reported. The DOE believes that this report will aid in discussions in plutonium storage, safety, and security with stakeholders as well as encourage other nations to declassify and release similar data. These data will also be available for formulating policies with respect to disposition of excess nuclear materials. The information in this report is based on the evaluation of available records. The information contained in this report may be updated or revised in the future should additional or more detailed data become available.

  2. Design-Only Conceptual Design Report: Plutonium Immobilization Plant

    SciTech Connect

    DiSabatino, A.; Loftus, D.

    1999-01-01

    This design-only conceptual design report was prepared to support a funding request by the Department of Energy Office of Fissile Materials Disposition for engineering and design of the Plutonium Immobilization Plant, which will be used to immobilize up to 50 tonnes of surplus plutonium. The siting for the Plutonium Immobilization Plant will be determined pursuant to the site-specific Surplus Plutonium Disposition Environmental Impact Statement in a Plutonium Deposition Record of Decision in early 1999. This document reflects a new facility using the preferred technology (ceramic immobilization using the can-in-canister approach) and the preferred site (at Savannah River). The Plutonium Immobilization Plant accepts plutonium from pit conversion and from non-pit sources and, through a ceramic immobilization process, converts the plutonium into mineral-like forms that are subsequently encapsulated within a large canister of high-level waste glass. The final immobilized product must make the plutonium as inherently unattractive and inaccessible for use in nuclear weapons as the plutonium in spent fuel from commercial reactors and must be suitable for geologic disposal. Plutonium immobilization at the Savannah River Site uses: (1) A new building, the Plutonium Immobilization Plant, which will convert non-pit surplus plutonium to an oxide form suitable for the immobilization process, immobilize plutonium in a titanate-based ceramic form, place cans of the plutonium-ceramic forms into magazines, and load the magazines into a canister; (2) The existing Defense Waste Processing Facility for the pouring of high-level waste glass into the canisters; and (3) The Actinide Packaging and Storage Facility to receive and store feed materials. The Plutonium Immobilization Plant uses existing Savannah River Site infra-structure for analytical laboratory services, waste handling, fire protection, training, and other support utilities and services. The Plutonium Immobilization Plant

  3. REMOVAL OF LEGACY PLUTONIUM MATERIALS FROM SWEDEN

    SciTech Connect

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

    2013-08-18

    U.S. Department of Energy’s National Nuclear Security Administration (NNSA) Office of Global Threat Reduction (GTRI) recently removed legacy plutonium materials from Sweden in collaboration with AB SVAFO, Sweden. This paper details the activities undertaken through the U.S. receiving site (Savannah River Site (SRS)) to support the characterization, stabilization, packaging and removal of legacy plutonium materials from Sweden in 2012. This effort was undertaken as part of GTRI’s Gap Materials Program and culminated with the successful removal of plutonium from Sweden as announced at the 2012 Nuclear Security Summit. The removal and shipment of plutonium materials to the United States was the first of its kind under NNSA’s Global Threat Reduction Initiative. The Environmental Assessment for the U.S. receipt of gap plutonium material was approved in May 2010. Since then, the multi-year process yielded many first time accomplishments associated with plutonium packaging and transport activities including the application of the of DOE-STD-3013 stabilization requirements to treat plutonium materials outside the U.S., the development of an acceptance criteria for receipt of plutonium from a foreign country, the development and application of a versatile process flow sheet for the packaging of legacy plutonium materials, the identification of a plutonium container configuration, the first international certificate validation of the 9975 shipping package and the first intercontinental shipment using the 9975 shipping package. This paper will detail the technical considerations in developing the packaging process flow sheet, defining the key elements of the flow sheet and its implementation, determining the criteria used in the selection of the transport package, developing the technical basis for the package certificate amendment and the reviews with multiple licensing authorities and most importantly integrating the technical activities with the Swedish partners.

  4. Microwave calcination for plutonium immobilization and residue stabilization

    SciTech Connect

    Harris, M.J.; Rising, T.L.; Roushey, W.J.; Sprenger, G.S.

    1995-12-01

    In the late 1980`s development was begun on a process using microwave energy to vitrify low level mixed waste sludge and transuranic mixed waste sludge generated in Building 374 at Rocky Flats. This process was shown to produce a dense, highly durable waste form. With the cessation of weapons production at Rocky Flats, the emphasis has changed from treatment of low level and TRU wastes to stabilizaiton of plutonium oxide and residues. This equipment is versatile and can be used as a heat source to calcine, react or vitrify many types of residues and oxides. It has natural economies in that it heats only the material to be treated, significantly reducing cycle times over conventional furnaces. It is inexpensive to operate in that most of the working components remain outside of any necessary contamination enclosure and therefore can easily be maintained. Limited testing has been successfully performed on cerium oxide (as a surrogate for plutonium oxide), surrogate electrorefining salts, surrogate residue sludge and residue ash. Future plans also include tests on ion exchange resins. In an attempt to further the usefullness of this technology, a mobile, self-contained microwave melting system is currently under development and expected to be operational at Rocky Flats Enviromental Technology Site by the 4th quarter of FY96.

  5. Influence of a reaction medium on the oxidation of aromatic nitrogen-containing compounds by peroxyacids

    NASA Astrophysics Data System (ADS)

    Dutka, V. S.; Matsyuk, N. V.; Dutka, Yu. V.

    2011-01-01

    The influence of different solvents on the oxidation reaction rate of pyridine (Py), quinoline (QN), acridine (AN), α-oxyquinoline (OQN) and α-picolinic acid (APA) by peroxydecanoic acid (PDA) was studied. It was found that the oxidation rate grows in the series Py < QN < AN, and the rate of the oxidation reaction of compounds containing a substituent in the α position from a reactive center is significantly lower than for unsubstituted analogues. The effective energies of activation of the oxidation reaction were found. It was shown that in the first stage, the reaction mechanism includes the rapid formation of an intermediate complex nitrogen-containing compound, peroxyacid, which forms products upon decomposing in the second stage. A kinetic equation that describes the studied process is offered. The constants of equilibrium of the intermediate complex formation ( K eq) and its decomposition constant ( k 2) in acetone and benzene were calculated. It was shown that the nature of the solvent influences the numerical values of both K p and k 2. It was established that introduction of acetic acid (which is able to form compounds with Py) into the reaction medium slows the rate of the oxidation process drastically. Correlation equations linking the polarity, polarizability, electrophilicity, and basicity of solvents with the constant of the PDA oxidation reaction rate for Py were found. It was concluded that the basicity and polarity of the solvent have a decisive influence on the oxidation reaction rate, while the polarizability and electrophilicity of the reaction medium do not influence the oxidation reaction rate.

  6. Plutonium oxalate precipitation for trace elemental determination in plutonium materials

    DOE PAGESBeta

    Xu, Ning; Gallimore, David; Lujan, Elmer; Garduno, Katherine; Walker, Laurie; Taylor, Fiona; Thompson, Pam; Tandon, Lav

    2015-05-26

    In this study, an analytical chemistry method has been developed that removes the plutonium (Pu) matrix from the dissolved Pu metal or oxide solution prior to the determination of trace impurities that are present in the metal or oxide. In this study, a Pu oxalate approach was employed to separate Pu from trace impurities. After Pu(III) was precipitated with oxalic acid and separated by centrifugation, trace elemental constituents in the supernatant were analyzed by inductively coupled plasma-optical emission spectroscopy with minimized spectral interferences from the sample matrix.

  7. Preparation of hydroxyapatite-containing titania coating on titanium substrate by micro-arc oxidation

    SciTech Connect

    Ni Jiahua; Shi Yulong Yan Fengying; Chen Jianzhi; Wang Lei

    2008-01-08

    Hydroxyapatite-containing titania coatings on titanium substrates were formed by micro-arc oxidation (MAO) in electrolyte containing calcium acetate monohydrate (CH{sub 3}COO){sub 2}Ca.H{sub 2}O) and sodium phosphate monobasic dihydrate (NaH{sub 2}PO{sub 4}.2H{sub 2}O) using a pulse power supply. Scanning electron microscopy (SEM) with Energy dispersive X-ray spectrometer (EDX) and X-ray diffraction (XRD) were employed to characterize the microstructure, elemental composition and phase components of the coatings. The coatings were rough and porous, without apparent interface to the titanium substrates. All the oxidized coatings contained Ca and P as well as Ti and O, and the porous coatings were made up of anatase, rutile and hydroxyapatite. Such MAO films are expected to have significant applications as artificial bone joints and dental implants.

  8. Ceramification: A plutonium immobilization process

    SciTech Connect

    Rask, W.C.; Phillips, A.G.

    1996-05-01

    This paper describes a low temperature technique for stabilizing and immobilizing actinide compounds using a combination process/storage vessel of stainless steel, in which measured amounts of actinide nitrate solutions and actinide oxides (and/or residues) are systematically treated to yield a solid article. The chemical ceramic process is based on a coating technology that produces rare earth oxide coatings for defense applications involving plutonium. The final product of this application is a solid, coherent actinide oxide with process-generated encapsulation that has long-term environmental stability. Actinide compounds can be stabilized as pure materials for ease of re-use or as intimate mixtures with additives such as rare earth oxides to increase their degree of proliferation resistance. Starting materials for the process can include nitrate solutions, powders, aggregates, sludges, incinerator ashes, and others. Agents such as cerium oxide or zirconium oxide may be added as powders or precursors to enhance the properties of the resulting solid product. Additives may be included to produce a final product suitable for use in nuclear fuel pellet production. The process is simple and reduces the time and expense for stabilizing plutonium compounds. It requires a very low equipment expenditure and can be readily implemented into existing gloveboxes. The process is easily conducted with less associated risk than proposed alternative technologies.

  9. First-principles local density approximation+U and generalized gradient approximation+U study of plutonium oxides

    NASA Astrophysics Data System (ADS)

    Sun, Bo; Zhang, Ping; Zhao, Xian-Geng

    2008-02-01

    The electronic structure and properties of PuO2 and Pu2O3 have been studied from first principles by the all-electron projector-augmented-wave method. The local density approximation+U and the generalized gradient approximation+U formalisms have been used to account for the strong on-site Coulomb repulsion among the localized Pu 5f electrons. We discuss how the properties of PuO2 and Pu2O3 are affected by the choice of U as well as the choice of exchange-correlation potential. Also, oxidation reaction of Pu2O3, leading to formation of PuO2, and its dependence on U and exchange-correlation potential have been studied. Our results show that by choosing an appropriate U, it is promising to correctly and consistently describe structural, electronic, and thermodynamic properties of PuO2 and Pu2O3, which enable the modeling of redox process involving Pu-based materials possible.

  10. Impact of Fission Products Impurity on the Plutonium Content of Metal- and Oxide- Fuels in Sodium Cooled Fast Reactors

    SciTech Connect

    Hikaru Hiruta; Gilles Youinou

    2013-09-01

    This short report presents the neutronic analysis to evaluate the impact of fission product impurity on the Pu content of Sodium-cooled Fast Reactor (SFR) metal- and oxide- fuel fabrication. The similar work has been previously done for PWR MOX fuel [1]. The analysis will be performed based on the assumption that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate SFR fuels. Only non-gaseous FPs have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1 of Reference 1). Throughout of this report, we define the mixture of Pu and FPs as PuFP. The main objective of this analysis is to quantify the increase of the Pu content of SFR fuels necessary to maintain the same average burnup at discharge independently of the amount of FP in the Pu stream, i.e. independently of the PuFP composition. The FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  11. Identification of an iridium-containing compound with a formal oxidation state of IX.

    PubMed

    Wang, Guanjun; Zhou, Mingfei; Goettel, James T; Schrobilgen, Gary J; Su, Jing; Li, Jun; Schlöder, Tobias; Riedel, Sebastian

    2014-10-23

    One of the most important classifications in chemistry and within the periodic table is the concept of formal oxidation states. The preparation and characterization of compounds containing elements with unusual oxidation states is of great interest to chemists. The highest experimentally known formal oxidation state of any chemical element is at present VIII, although higher oxidation states have been postulated. Compounds with oxidation state VIII include several xenon compounds (for example XeO4 and XeO3F2) and the well-characterized species RuO4 and OsO4 (refs 2-4). Iridium, which has nine valence electrons, is predicted to have the greatest chance of being oxidized beyond the VIII oxidation state. In recent matrix-isolation experiments, the IrO4 molecule was characterized as an isolated molecule in rare-gas matrices. The valence electron configuration of iridium in IrO4 is 5d(1), with a formal oxidation state of VIII. Removal of the remaining d electron from IrO4 would lead to the iridium tetroxide cation ([IrO4](+)), which was recently predicted to be stable and in which iridium is in a formal oxidation state of IX. There has been some speculation about the formation of [IrO4](+) species, but these experimental observations have not been structurally confirmed. Here we report the formation of [IrO4](+) and its identification by infrared photodissociation spectroscopy. Quantum-chemical calculations were carried out at the highest level of theory that is available today, and predict that the iridium tetroxide cation, with a Td-symmetrical structure and a d(0) electron configuration, is the most stable of all possible [IrO4](+) isomers. PMID:25341786

  12. Identification of an iridium-containing compound with a formal oxidation state of IX

    NASA Astrophysics Data System (ADS)

    Wang, Guanjun; Zhou, Mingfei; Goettel, James T.; Schrobilgen, Gary J.; Su, Jing; Li, Jun; Schlöder, Tobias; Riedel, Sebastian

    2014-10-01

    One of the most important classifications in chemistry and within the periodic table is the concept of formal oxidation states. The preparation and characterization of compounds containing elements with unusual oxidation states is of great interest to chemists. The highest experimentally known formal oxidation state of any chemical element is at present VIII, although higher oxidation states have been postulated. Compounds with oxidation state VIII include several xenon compounds (for example XeO4 and XeO3F2) and the well-characterized species RuO4 and OsO4 (refs 2, 3, 4). Iridium, which has nine valence electrons, is predicted to have the greatest chance of being oxidized beyond the VIII oxidation state. In recent matrix-isolation experiments, the IrO4 molecule was characterized as an isolated molecule in rare-gas matrices. The valence electron configuration of iridium in IrO4 is 5d1, with a formal oxidation state of VIII. Removal of the remaining d electron from IrO4 would lead to the iridium tetroxide cation ([IrO4]+), which was recently predicted to be stable and in which iridium is in a formal oxidation state of IX. There has been some speculation about the formation of [IrO4]+ species, but these experimental observations have not been structurally confirmed. Here we report the formation of [IrO4]+ and its identification by infrared photodissociation spectroscopy. Quantum-chemical calculations were carried out at the highest level of theory that is available today, and predict that the iridium tetroxide cation, with a Td-symmetrical structure and a d0 electron configuration, is the most stable of all possible [IrO4]+ isomers.

  13. Zinc-Containing Magnetic Oxides Stabilized by a Polymer: One Phase or Two?

    PubMed

    Baird, Nicholas; Losovyj, Yaroslav; Yuzik-Klimova, Ekaterina Yu; Kuchkina, Nina V; Shifrina, Zinaida B; Pink, Maren; Stein, Barry D; Morgan, David Gene; Wang, Tianhao; Rubin, Mikhail A; Sidorov, Alexander I; Sulman, Esther M; Bronstein, Lyudmila M

    2016-01-13

    Here we developed a new family of Zn-containing magnetic oxides of different structures by thermal decomposition of Zn(acac)2 in the reaction solution of preformed magnetite nanoparticles (NPs) stabilized by polyphenylquinoxaline. Upon an increase of the Zn(acac)2 loading from 0.15 to 0.40 mmol (vs 1 mmol of Fe(acac)3), the Zn content increases, and the Zn-containing magnetic oxide NPs preserve a spinel structure of magnetite and an initial, predominantly multicore NP morphology. X-ray photoelectron spectroscopy (XPS) of these samples revealed that the surface of iron oxide NPs is enriched with Zn, although Zn species were also found deep under the iron oxide NP surface. For all the samples, XPS also demonstrates the atom ratio of Fe(3+)/Fe(2+) = 2:1, perfectly matching Fe3O4, but not ZnFe2O4, where Fe(2+) ions are replaced with Zn(2+). The combination of XPS with other physicochemical methods allowed us to propose that ZnO forms an ultrathin amorphous layer on the surface of iron oxide NPs and also diffuses inside the magnetite crystals. At higher Zn(acac)2 loading, cubic ZnO nanocrystals coexist with magnetite NPs, indicating a homogeneous nucleation of the former. The catalytic testing in syngas conversion to methanol demonstrated outstanding catalytic properties of Zn-containing magnetic oxides, whose activities are dependent on the Zn loading. Repeat experiments carried out with the best catalyst after magnetic separation showed remarkable catalyst stability even after five consecutive catalytic runs. PMID:26673012

  14. Mössbauer and magnetic studies of nanocomposites containing iron oxides and humic acids

    NASA Astrophysics Data System (ADS)

    Chistyakova, N. I.; Shapkin, A. A.; Gubaidulina, T. V.; Matsnev, M. E.; Sirazhdinov, R. R.; Rusakov, V. S.

    2014-04-01

    Nanocomposites containing iron oxides and humic acids were studied by Mössbauer and magnetic measurements. The concentrations of humic acids as the precursor in nanocomposites were varied. Mössbauer investigations were carried out at temperature range from room temperature to 5 K. The magnetization M( T, H) was measured in the temperature interval 80-300 K and magnetic field up to 10 kOe. It was found that particles of investigated nanocomposites exhibit superparamagnetic properties. The core of the nanocomposite was a mixture of non-stoichiometric magnetite and maghemite. The "iron-polymer" interface was formed on the surface of the iron oxide particles.

  15. Fuel electrode containing pre-sintered nickel/zirconia for a solid oxide fuel cell

    DOEpatents

    Ruka, Roswell J.; Vora, Shailesh D.

    2001-01-01

    A fuel cell structure (2) is provided, having a pre-sintered nickel-zirconia fuel electrode (6) and an air electrode (4), with a ceramic electrolyte (5) disposed between the electrodes, where the pre-sintered fuel electrode (6) contains particles selected from the group consisting of nickel oxide, cobalt and cerium dioxide particles and mixtures thereof, and titanium dioxide particles, within a matrix of yttria-stabilized zirconia and spaced-apart filamentary nickel strings having a chain structure, and where the fuel electrode can be sintered to provide an active solid oxide fuel cell.

  16. Reduction of worldwide plutonium inventories using conventional reactors and advanced fuels: A systems study

    SciTech Connect

    Krakowski, R.A.; Bathke, C.G.; Chodak, P. III

    1997-09-01

    The potential for reducing plutonium inventories in the civilian nuclear fuel cycle through recycle in LWRs of a variety of mixed-oxide forms is examined by means of a cost-based plutonium-flow systems model that includes an approximate measure of proliferation risk. The impact of plutonium recycle in a number of forms is examined, including the introduction of nonfertile fuels into conventional (LWR) reactors to reduce net plutonium generation, to increase plutonium burnup, and to reduce exo-reactor plutonium inventories.

  17. The Oxidation of Sulfur-Containing Compounds Using Heterogeneous Catalysts of Transition Metal Oxides Deposited on the Polymeric Matrix

    NASA Astrophysics Data System (ADS)

    Dinh Vu, Ngo; Dinh Bui, Nhi; Thi Minh, Thao; Thi Thanh Dam, Huong; Thi Tran, Hang

    2016-05-01

    We investigate the activity of heterogeneous catalysts of transition metal oxides deposited on the polymeric matrix in the oxidation of sulfur-containing compounds. It is shown that MnO2-10/CuO-10 has the highest catalytic activity. The physicomechanical properties of polymeric heterogeneous catalysts of transition-metal oxides, including the specific surface area, elongation at break and breaking strength, specific electrical resistance, and volume resistivity were studied by using an Inspekt mini 3 kN universal tensile machine in accordance with TCVN 4509:2006 at a temperature of 20 ± 2°C. Results show that heterogeneous polymeric catalysts were stable under severe reaction conditions. Scanning electron microscopy, and energy-dispersive analysis are used to study the surfaces of the catalysts. Microstructural characterization of the catalysts is performed by using x-ray computed tomography. We demonstrate the potential application of polymeric heterogeneous catalysts of transition-metal oxides in industrial wastewater treatment.

  18. The interfacial orientation relationship of oxide nanoparticles in a hafnium-containing oxide dispersion-strengthened austenitic stainless steel

    DOE PAGESBeta

    Miao, Yinbin; Mo, Kun; Cui, Bai; Chen, Wei-Ying; Miller, Michael K.; Powers, Kathy A.; McCreary, Virginia; Gross, David; Almer, Jonathan; Robertson, Ian M.; et al

    2015-01-26

    We report comprehensive investigations on the orientation relationship of the oxide nanoparticles in a hafnium-containing austenitic oxide dispersion-strengthened 316 stainless steel. The phases of the oxide nanoparticles were determined by a combination of scanning transmission electron microscopy–electron dispersive X-ray spectroscopy, atom probe tomography and synchrotron X-ray diffraction to be complex Y–Ti–Hf–O compounds with similar crystal structures, including bixbyite Y2O3, fluorite Y2O3–HfO2 solid solution and pyrochlore (or fluorite) Y2(Ti,Hf)2 - xO7 - x. High resolution transmission electron microscopy was used to characterize the particle–matrix interfaces. Moreover, two different coherency relationships along with one axis-parallel relation between the oxide nanoparticles and themore » steel matrix were found. The size of the nanoparticles significantly influences the orientation relationship. Our results provide insight into the relationship of these nanoparticles with the matrix, which has implications for interpreting material properties as well as responses to radiation.« less

  19. The interfacial orientation relationship of oxide nanoparticles in a hafnium-containing oxide dispersion-strengthened austenitic stainless steel

    SciTech Connect

    Miao, Yinbin; Mo, Kun; Cui, Bai; Chen, Wei-Ying; Miller, Michael K.; Powers, Kathy A.; McCreary, Virginia; Gross, David; Almer, Jonathan; Robertson, Ian M.; Stubbins, James F.

    2015-01-26

    We report comprehensive investigations on the orientation relationship of the oxide nanoparticles in a hafnium-containing austenitic oxide dispersion-strengthened 316 stainless steel. The phases of the oxide nanoparticles were determined by a combination of scanning transmission electron microscopy–electron dispersive X-ray spectroscopy, atom probe tomography and synchrotron X-ray diffraction to be complex Y–Ti–Hf–O compounds with similar crystal structures, including bixbyite Y2O3, fluorite Y2O3–HfO2 solid solution and pyrochlore (or fluorite) Y2(Ti,Hf)2 - xO7 - x. High resolution transmission electron microscopy was used to characterize the particle–matrix interfaces. Moreover, two different coherency relationships along with one axis-parallel relation between the oxide nanoparticles and the steel matrix were found. The size of the nanoparticles significantly influences the orientation relationship. Our results provide insight into the relationship of these nanoparticles with the matrix, which has implications for interpreting material properties as well as responses to radiation.

  20. ESADA Plutonium Program Critical Experiments: Power Distribution Measurements

    SciTech Connect

    Akkurt, H.

    2001-06-12

    In 1967, a series of critical experiments were conducted at the Westinghouse Reactor Evaluation Center (WREC) using mixed-oxide (MOX) PuO{sub 2}-UO{sub 2} and/or UO{sub 2} fuels in various lattices and configurations. These experiments were performed under the joint sponsorship of Empire State Atomic Development Associates (ESADA) plutonium program and Westinghouse. The purpose of these experiments was to develop experimental data useful in validating analytical methods used in the design of plutonium-bearing replacement fuel for water reactors. Three different fuel types were used during the experimental program: two MOX fuels and a low-enriched UO{sub 2} fuel. The MOX fuels were distinguished by their {sup 240}Pu content: 8 wt % {sup 240}Pu and 24 wt % {sup 240}Pu. Both MOX fuels contained 2.0 wt % PuO{sub 2} in natural UO{sub 2}. The UO{sub 2} fuel with 2.72 wt % enrichment was used for comparison with the plutonium data and for use in multiregion experiments.

  1. PROCESS USING POTASSIUM LANTHANUM SULFATE FOR FORMING A CARRIER PRECIPITATE FOR PLUTONIUM VALUES

    DOEpatents

    Angerman, A.A.

    1958-10-21

    A process is presented for recovering plutonium values in an oxidation state not greater than +4 from fluoride-soluble fission products. The process consists of adding to an aqueous acidic solution of such plutonium values a crystalline potassium lanthanum sulfate precipitate which carries the plutonium values from the solution.

  2. Crystalline plutonium hosts derived from high-level waste formulations.

    SciTech Connect

    O'Holleran, T. P.

    1998-04-24

    The Department of Energy has selected immobilization for disposal in a repository as one approach for disposing of excess plutonium (1). Materials for immobilizing weapons-grade plutonium for repository disposal must meet the ''spent fuel standard'' by providing a radiation field similar to spent fuel (2). Such a radiation field can be provided by incorporating fission products from high-level waste into the waste form. Experiments were performed to evaluate the feasibility of incorporating high-level waste (HLW) stored at the Idaho Chemical Processing Plant (ICPP) into plutonium dispositioning materials to meet the spent fuel standard. A variety of materials and preparation techniques were evaluated based on prior experience developing waste forms for immobilizing HLW. These included crystalline ceramic compositions prepared by conventional sintering and hot isostatic pressing (HIP), and glass formulations prepared by conventional melting. Because plutonium solubility in silicate melts is limited, glass formulations were intentionally devitrified to partition plutonium into crystalline host phases, thereby allowing increased overall plutonium loading. Samarium, added as a representative rare earth neutron absorber, also tended to partition into the plutonium host phases. Because the crystalline plutonium host phases are chemically more inert, the plutonium is more effectively isolated from the environment, and its attractiveness for proliferation is reduced. In the initial phase of evaluating each material and preparation method, cerium was used as a surrogate for plutonium. For promising materials, additional preparation experiments were performed using plutonium to verify the behavior of cerium as a surrogate. These experiments demonstrated that cerium performed well as a surrogate for plutonium. For the most part, cerium and plutonium partitioned onto the same crystalline phases, and no anomalous changes in oxidation state were observed. The only observed

  3. Excess Weapons Plutonium Immobilization in Russia

    SciTech Connect

    Jardine, L.; Borisov, G.B.

    2000-04-15

    The joint goal of the Russian work is to establish a full-scale plutonium immobilization facility at a Russian industrial site by 2005. To achieve this requires that the necessary engineering and technical basis be developed in these Russian projects and the needed Russian approvals be obtained to conduct industrial-scale immobilization of plutonium-containing materials at a Russian industrial site by the 2005 date. This meeting and future work will provide the basis for joint decisions. Supporting R&D projects are being carried out at Russian Institutes that directly support the technical needs of Russian industrial sites to immobilize plutonium-containing materials. Special R&D on plutonium materials is also being carried out to support excess weapons disposition in Russia and the US, including nonproliferation studies of plutonium recovery from immobilization forms and accelerated radiation damage studies of the US-specified plutonium ceramic for immobilizing plutonium. This intriguing and extraordinary cooperation on certain aspects of the weapons plutonium problem is now progressing well and much work with plutonium has been completed in the past two years. Because much excellent and unique scientific and engineering technical work has now been completed in Russia in many aspects of plutonium immobilization, this meeting in St. Petersburg was both timely and necessary to summarize, review, and discuss these efforts among those who performed the actual work. The results of this meeting will help the US and Russia jointly define the future direction of the Russian plutonium immobilization program, and make it an even stronger and more integrated Russian program. The two objectives for the meeting were to: (1) Bring together the Russian organizations, experts, and managers performing the work into one place for four days to review and discuss their work with each other; and (2) Publish a meeting summary and a proceedings to compile reports of all the excellent

  4. Highly Efficient Elimination of Carbon Monoxide with Binary Copper-Manganese Oxide Contained Ordered Nanoporous Silicas

    NASA Astrophysics Data System (ADS)

    Lee, Jiho; Kim, Hwayoun; Lee, Hyesun; Jang, Seojun; Chang, Jeong Ho

    2016-01-01

    Ordered nanoporous silicas containing various binary copper-manganese oxides were prepared as catalytic systems for effective carbon monoxide elimination. The carbon monoxide elimination efficiency was demonstrated as a function of the [Mn]/[Cu] ratio and reaction time. The prepared catalysts were characterized by Brunauer-Emmett-Teller (BET) method, small- and wide-angle X-ray diffraction (XRD), and high-resolution transmission electron microscopy (HR-TEM) for structural analysis. Moreover, quantitative analysis of the binary metal oxides within the nanoporous silica was achieved by inductively coupled plasma (ICP). The binary metal oxide-loaded nanoporous silica showed high room temperature catalytic efficiency with over 98 % elimination of carbon monoxide at higher concentration ratio of [Mn]/[Cu].

  5. Highly Efficient Elimination of Carbon Monoxide with Binary Copper-Manganese Oxide Contained Ordered Nanoporous Silicas.

    PubMed

    Lee, Jiho; Kim, Hwayoun; Lee, Hyesun; Jang, Seojun; Chang, Jeong Ho

    2016-12-01

    Ordered nanoporous silicas containing various binary copper-manganese oxides were prepared as catalytic systems for effective carbon monoxide elimination. The carbon monoxide elimination efficiency was demonstrated as a function of the [Mn]/[Cu] ratio and reaction time. The prepared catalysts were characterized by Brunauer-Emmett-Teller (BET) method, small- and wide-angle X-ray diffraction (XRD), and high-resolution transmission electron microscopy (HR-TEM) for structural analysis. Moreover, quantitative analysis of the binary metal oxides within the nanoporous silica was achieved by inductively coupled plasma (ICP). The binary metal oxide-loaded nanoporous silica showed high room temperature catalytic efficiency with over 98 % elimination of carbon monoxide at higher concentration ratio of [Mn]/[Cu]. PMID:26744146

  6. A Novel Anti-Influenza Copper Oxide Containing Respiratory Face Mask

    PubMed Central

    Borkow, Gadi; Zhou, Steve S.; Page, Tom; Gabbay, Jeffrey

    2010-01-01

    Background Protective respiratory face masks protect the nose and mouth of the wearer from vapor drops carrying viruses or other infectious pathogens. However, incorrect use and disposal may actually increase the risk of pathogen transmission, rather than reduce it, especially when masks are used by non-professionals such as the lay public. Copper oxide displays potent antiviral properties. A platform technology has been developed that permanently introduces copper oxide into polymeric materials, conferring them with potent biocidal properties. Methodology/Principal Findings We demonstrate that impregnation of copper oxide into respiratory protective face masks endows them with potent biocidal properties in addition to their inherent filtration properties. Both control and copper oxide impregnated masks filtered above 99.85% of aerosolized viruses when challenged with 5.66±0.51 and 6.17±0.37 log10TCID50 of human influenza A virus (H1N1) and avian influenza virus (H9N2), respectively, under simulated breathing conditions (28.3 L/min). Importantly, no infectious human influenza A viral titers were recovered from the copper oxide containing masks within 30 minutes (≤0.88 log10TCID50), while 4.67±1.35 log10TCID50 were recovered from the control masks. Similarly, the infectious avian influenza titers recovered from the copper oxide containing masks were ≤0.97±0.01 log10TCID50 and from the control masks 5.03±0.54 log10TCID50. The copper oxide containing masks successfully passed Bacterial Filtration Efficacy, Differential Pressure, Latex Particle Challenge, and Resistance to Penetration by Synthetic Blood tests designed to test the filtration properties of face masks in accordance with the European EN 14683:2005 and NIOSH N95 standards. Conclusions/Significance Impregnation of copper oxide into respiratory protective face masks endows them with potent anti-influenza biocidal properties without altering their physical barrier properties. The use of biocidal masks

  7. Modeling anti-Trypanosoma cruzi activity of N-oxide containing heterocycles.

    PubMed

    Boiani, Mariana; Cerecetto, Hugo; Gonzalez, Mercedes; Gasteiger, Johann

    2008-01-01

    In the present study a systematic approach was used to model the anti-T. cruzi activity of a series of N-oxide containing heterocycles belonging to four chemical families with a wide structural diversity. The proposed mode of action implies the reduction of the N-oxide moiety; however, the biochemical mechanism underlying the anti-T. cruzi activity is still unkown. For structural representation two types of descriptors were analyzed: quantum chemical (AM1) global descriptors and properties coded by radial distribution function (RDF). Both types of descriptors point to the relevance of electronic properties. The local-RDF (LRDF) identified an electrophilic center at 4.1-4.9 A from the oxygen atom of the N-oxide moiety, although other properties are required to explain the biological activity. While the mode of action of N-oxide containing heterocycles is still unknown, the results obtained here strengthen the importance of the electrophilic character of the molecule and the possible participation of the heterocycle in a reduction process. The ability of these descriptors to distinguish among activity classes was assessed using Kohonen neural networks, and the best clustering descriptors were later used for model building. Different learning algorithms were used for model development, and stratified 10-fold cross-validation was used to evaluate the performance of each classifier. The best results were obtained using k-nearest neighbors (k-NN) and decision tree (J48) methods combined with global descriptors. Since tree-based methods are easily translated into classification rules, the J48 model is a useful tool in the de novo construction of new N-oxide containing heterocycle lead structures. PMID:18163603

  8. Synthesis of aliphatic sulfur pentafluorides by oxidation of SF₅-containing anisole, phenols, and anilines.

    PubMed

    Vida, Norbert; Pastýříková, Tereza; Klepetářová, Blanka; Beier, Petr

    2014-09-19

    4-(Pentafluorosulfanyl)catechol, 2-amino-4-(pentafluorosulfanyl)phenol, and 2-amino-5-(pentafluorosulfanyl)phenol undergo oxidation by lead tetraacetate at ambient temperature leading to dearomatization and the formation of SF5-substituted nitriles and esters of cis,cis-hexa-2,4-dienedioic (muconic) acid in good yields. 4-(Pentafluorosulfanyl)phenol and 4-(pentafluorosulfanyl)anisole are oxidized by 30% aqueous hydrogen peroxide in concentrated sulfuric acid to provide 2-(5-oxo-3-(pentafluorosulfanyl)-2,5-dihydrofuran-2-yl)acetic acid [3-(pentafluorosulfanyl)muconolactone] and small amounts of side products--SF5-containing maleic and succinic acids. The methods presented are the first examples of the practical synthesis of aliphatic SF5-containing compounds from readily available aromatic ones. PMID:25137015

  9. Effect of LEO Exposure on Aromatic Polymers Containing Phenylphosphine Oxide Groups

    NASA Technical Reports Server (NTRS)

    Lillehei, P. T.; Smith, J. G., Jr.; Connell, J. W.

    2008-01-01

    As part of the Materials on The International Space Station Experiment (MISSE), aromatic polymers containing phenylphosphine oxide groups were exposed to low Earth orbit for approx.4 years. All of the aromatic polymers containing phenylphosphine oxide groups survived the exposure despite the high fluence of atomic oxygen that completely eroded other polymer films such as Kapton(TradeMark) and Mylar(Trademark) of comparable or greater thickness. The samples were characterized for changes in physical properties, thermal/optical properties surface chemistry, and surface topography. The data from the polymer samples on MISSE were compared to samples from the same batch of material stored under ambient conditions on Earth. In addition, comparisons were made between the MISSE samples and those subjected to shorter term space flight exposures. The results of these analyses will be presented.

  10. Oxidation and biodegradation of polyethylene films containing pro-oxidantadditives: Synergistic effects of sunlight exposure, thermal aging and fungal biodegradation

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Synergistic effects of sunlight exposure, thermal aging and fungal biodegradation on the oxidation and biodegradation of linear low density poly (ethylene) PE-LLD films containing pro-oxidant were examined. To achieve oxidation and degradation, films were first exposed to the sunlight for 93 days du...

  11. Reduction of aqueous transition metal species on the surfaces of Fe(II)-containing oxides

    USGS Publications Warehouse

    White, A.F.; Peterson, M.L.

    1996-01-01

    Experimental studies demonstrate that structural Fe(II) in magnetite and ilmenite heterogeneously reduce aqueous ferric, cupric, vanadate, and chromate ions at the oxide surfaces over a pH range of 1-7 at 25??C. For an aqueous transition metal m, such reactions are 3[Fe2+Fe3+2]O4(magnetite) + 2/nmz ??? 4[Fe3+2]O3(maghemite) + Fe2+ + 2/nmz-n and 3[Fe2+Ti]O3(ilmenite) + 2/nmz ??? Fe3+2Ti3O9(pseudorutile) + Fe2+ + 2/nmz-n, where z is the valance state and n is the charge transfer number. The half cell potential range for solid state oxidation [Fe(II)] ??? [Fe(III)] is -0.34 to -0.65 V, making structural Fe(II) a stronger reducing agent than aqueous Fe2+ (-0.77 V). Reduction rates for aqueous metal species are linear with time (up to 36 h), decrease with pH, and have rate constants between 0.1 and 3.3 ?? 10-10 mol m-2 s-1. Iron is released to solution both from the above reactions and from dissolution of the oxide surface. In the presence of chromate, Fe2+ is oxidized homogeneously in solution to Fe3+. X-ray photoelectron spectroscopy (XPS) denotes a Fe(III) oxide surface containing reduced Cr(III) and V(IV) species. Magnetite and ilmenite electrode potentials are insensitive to increases in divalent transition metals including Zn(II), Co(II), Mn(II), and Ni(II) and reduced V(IV) and Cr(III) but exhibit a log-linear concentration-potential response to Fe(III) and Cu(II). Complex positive electrode responses occur with increasing Cr(VI) and V(V) concentrations. Potential dynamic scans indicate that the high oxidation potential of dichromate is capable of suppressing the cathodic reductive dissolution of magnetite. Oxide electrode potentials are determined by the Fe(II)/Fe(III) composition of the oxide surface and respond to aqueous ion potentials which accelerate this oxidation process. Natural magnetite sands weathered under anoxic conditions are electrochemically reactive as demonstrated by rapid chromate reduction and the release of aqueous Fe(III) to experimental

  12. Viscosity of sodium-borate melts containing mechanically activated REM oxide additions

    NASA Astrophysics Data System (ADS)

    Ryabov, V. V.; Istomin, S. A.; Khokhryakov, A. A.; Ivanov, A. V.; Paivin, A. S.

    2015-08-01

    The viscosity of sodium-borate melts containing 1 wt % mechanically activated lanthanide group REM oxide is measured in the temperature range 950-1650 K using an oscillating viscometer. The dependences of the viscosity on temperature and composition are determined. The logarithmic dependences have high- and low-temperature portions with different viscous flow activation energies. It is found that the melt structure changes as a function of composition and temperature.

  13. Plutonium speciation in water from Mono Lake, California

    USGS Publications Warehouse

    Cleveland, J.M.; Rees, T.F.; Nash, K.L.

    1983-01-01

    The solubility of plutonium in Mono Lake water is enhanced by the presence of large concentrations of indigenous carbonate ions and moderate concentrations of fluoride ions. In spite of the complex chemical composition of this water, only a few ions govern the behavior of plutonium, as demonstrated by the fact that it was possible to duplicate plutonium speciation in a synthetic water containing only the principal components of Mono Lake water.

  14. 14. END VIEW OF THE PLUTONIUM STORAGE VAULT FROM THE ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    14. END VIEW OF THE PLUTONIUM STORAGE VAULT FROM THE REMOTE CONTROL STATION. THE STACKER-RETRIEVER, A REMOTELY-OPERATED, MECHANIZED TRANSPORT SYSTEM, RETRIEVES CONTAINERS OF PLUTONIUM FROM SAFE GEOMETRY PALLETS STORED ALONG THE LENGTH OF THE VAULT. THE STACKER-RETRIEVER RUNS ALONG THE AISLE BETWEEN THE PALLETS OF THE STORAGE CHAMBER. (3/2/86) - Rocky Flats Plant, Plutonium Recovery Facility, Northwest portion of Rocky Flats Plant, Golden, Jefferson County, CO

  15. SEPARATION OF PLUTONIUM FROM AQUEOUS SOLUTIONS BY ION-EXCHANGE

    DOEpatents

    Schubert, J.

    1958-06-01

    A process is described for the separation of plutonium from an aqueous solution of a plutonium salt, which comprises adding to the solution an acid of the group consisting of sulfuric acid, phosphoric acid, and oxalic acid, and mixtures thereof to provide an acid concentration between 0.0001 and 1 M, contacting the resultant solution with a synthetic organic anion exchange resin, and separating the aqueous phase and the resin which contains the plutonium.

  16. Thrombin-activated human platelets acutely generate oxidized docosahexaenoic-acid-containing phospholipids via 12-lipoxygenase.

    PubMed

    Morgan, Lloyd T; Thomas, Christopher P; Kühn, Hartmut; O'Donnell, Valerie B

    2010-10-01

    Arachidonate-containing oxidized phospholipids are acutely generated by 12-LOX (12-lipoxygenase) in agonist-activated platelets. In the present study, formation of structurally related lipids by oxidation of DHA (docosahexaenoic acid)-containing phospholipids is demonstrated using lipidomic approaches. Precursor scanning reverse-phase LC (liquid chromatography)-MS/MS (tandem MS) identified a new family of lipids that comprise phospholipid-esterified HDOHE (hydroxydocosahexaenoic acid). Two diacyl and two plasmalogen PEs (phosphatidylethanolamines) containing predominantly the 14-HDOHE positional isomer (18:0p/14-HDOHE-PE, 18:0a/14-HDOHE-PE, 16:0a/14-HDOHE-PE and 16:0p/14-HDOHE-PE) were structurally characterized using MS/MS and by comparison with biogenic standards. An involvement of 12-LOX was indicated as purified recombinant human 12-LOX also generated the 14-HDOHE isomer from DHA. Pharmacological studies using inhibitors and recombinant platelet 12-LOX indicate that they form via esterification of newly formed non-esterified HDOHE. HDOHE-PEs formed at significant rates (2-4 ng/4×10(7) cells) within 2-180 min of thrombin stimulation, and their formation was blocked by calcium chelation. In summary, a new family of oxidized phospholipid was identified in thrombin-activated human platelets. PMID:20653566

  17. IMPROVED PROCESS OF PLUTONIUM CARRIER PRECIPITATION

    DOEpatents

    Faris, B.F.

    1959-06-30

    This patent relates to an improvement in the bismuth phosphate process for separating and recovering plutonium from neutron irradiated uranium, resulting in improved decontamination even without the use of scavenging precipitates in the by-product precipitation step and subsequently more complete recovery of the plutonium in the product precipitation step. This improvement is achieved by addition of fluomolybdic acid, or a water soluble fluomolybdate, such as the ammonium, sodium, or potassium salt thereof, to the aqueous nitric acid solution containing tetravalent plutonium ions and contaminating fission products, so as to establish a fluomolybdate ion concentration of about 0.05 M. The solution is then treated to form the bismuth phosphate plutonium carrying precipitate.

  18. Response to comments received from the State of Colorado and the public on the Environmental Assessment for resumption of thermal stabilization of plutonium oxide in Building 707

    SciTech Connect

    Not Available

    1994-02-01

    The Department of Energy (DOE) prepared this document to respond to comments from the State of Colorado and the public on the draft Environmental Assessment (EA) for the Resumption of Thermal Stabilization of Plutonium Oxide in Building 707 at the Rocky Flats Plant (RFP) in Golden, Colorado. The draft EA was provided to the State of Colorado and the public on September 8, 1993, for a comment period of 60 days. The Department`s National Environmental Policy Act (NEPA) Implementing Procedures (10 Code of Federal Regulations [CFR] 1021.301) requires that prior to approval of the EA, DOE is to allow the host State and Indian Tribe a period of from 14 to 30 days to review and comment on the EA. The Department established a comment period of 60 days for this EA in response to requests by the public during the first public meeting on July 7, 1993, before preparation of the EA. Other issues raised at the July 7 meeting included the range of alternatives to be considered, the time period for preparation of the EA, and the amount of material to be thermally stabilized. These and other comments made by the public at that meeting were carefully considered in preparation of the EA. In addition to providing the preapproval draft EA to the State of Colorado, DOE distributed the EA to all persons and groups on the RFP public information mailing list and placed the EA and reference documents in the DOE Public Reading Rooms in the RFP area. A public meeting was held on October 6, 1993, to hear public comments on the draft EA. All comments on the draft EA, those received both at the October 6 public meeting and through correspondence, have been reproduced in their entirety in this Response to Comments document. Responses to the commenters` questions and concerns are provided, and changes made to the body of the EA are indicated in the responses. All comments received have been considered in the revision of the EA.

  19. Experimental studies and thermodynamic modelling of volatilities of uranium, plutonium, and americium from their oxides and from their oxides interacted with ash

    SciTech Connect

    Krikorian, O.H.; Ebbinghaus, B.B.; Adamson, M.G.; Fontes, A.S. Jr.; Fleming, D.L.

    1993-09-15

    The purpose of this study is to identify the types and amounts of volatile gaseous species of U, Pu, and Am that are produced in the combustion chamber offgases of mixed waste oxidation processors. Primary emphasis is on the Rocky Flats Plant Fluidized Bed Incinerator. Transpiration experiments have been carried out on U{sub 3}O{sub 8}(s), U{sub 3}O{sub 8} interacted with various ash materials, PuO{sub 2}(s), PuO{sub 2} interacted with ash materials, and a 3%PuO{sub 2}/0.06%AmO{sub 2}/ash material, all in the presence of steam and oxygen, and at temperatures in the vicinity of 1,300 K. UO{sub 3}(g) and UO{sub 2}(OH){sub 2}(g) have been identified as the uranium volatile species and thermodynamic data established for them. Pu and Am are found to have very low volatilities, and carryover of Pu and Am as fine dust particulates is found to dominate over vapor transport. The authors are able to set upper limits on Pu and Am volatilities. Very little lowering of U volatility is found for U{sub 3}O{sub 8} interacted with typical ashes. However, ashes high in Na{sub 2}O (6.4 wt %) or in CaO (25 wt %) showed about an order of magnitude reduction in U volatility. Thermodynamic modeling studies were carried out that show that for aluminosilicate ash materials, it is the presence of group I and group II oxides that reduces the activity of the actinide oxides. K{sub 2}O is the most effective followed by Na{sub 2}O and CaO for common ash constituents. A more major effect in actinide activity lowering could be achieved by adding excess group I or group II oxides to exceed their interaction with the ash and lead to direct formation of alkali or alkaline earth uranates, plutonates, and americates.

  20. Formation of iodinated organic compounds by oxidation of iodide-containing waters with manganese dioxide.

    PubMed

    Gallard, Hervé; Allard, Sébastien; Nicolau, Rudy; von Gunten, Urs; Croué, Jean Philippe

    2009-09-15

    This study shows that iodinated organic compounds can be produced when iodide-containing waters are in contact with manganese oxide birnessite (delta-MnO2) in the pH range of 5-7. In the absence of natural organic matter (NOM), iodide is oxidized to iodate that is also adsorbed onto delta-MnO2. In the presence of iodide and NOM, adsordable organic iodine compounds (AOI) are formed at pH < 7 because of the oxidation of iodide to iodine by delta-MnO2 and the reactions of iodine with NOM. In addition, iodoacetic acid and iodoform have been identified as specific iodinated byproducts. Formation of iodoform is not observed for high NOM/delta-MnO2 ratios due to inhibition of the catalytic effect of delta-MnO2 by NOM poisoning. Experiments with model compounds such as resorcinol and 3,5-heptanedione confirmed that the delta-MnO2/l(-) system is very effective for the formation of iodinated organic compounds. These results suggest that birnessite acts as a catalyst through the oxidation of iodide to iodine and the polarization of the iodine molecule, which then reacts with NOM moieties. Furthermore, our results indicate that during water treatment in the presence of manganese oxide, iodinated organic compounds may be formed, which may lead to taste and odor or toxicological problems. PMID:19806734

  1. Observed changes in the mechanism and rates of Pu(V) reduction on hematite as a function of total plutonium concentration.

    PubMed

    Hixon, Amy E; Powell, Brian A

    2014-08-19

    Changes in aqueous- and solid-phase plutonium oxidation states were monitored as a function of time and plutonium concentration in hematite (α-Fe2O3) suspensions containing initially Pu(V). Batch kinetic experiments were conducted at plutonium concentrations between 10(-8) and 10(-6) M at pH 5 and 0.3 g/L (9.3 m(2)/L) hematite. Surface-mediated reduction of Pu(V) was observed under all conditions studied. However, differences in the reaction kinetics demonstrate that the mechanism of Pu(V) reduction changes as a function of plutonium concentration. Adsorption of Pu(V) was found to be the rate-limiting step at plutonium concentrations less than approximately 10(-7) M Pu(V). Plutonium reduction in these systems was attributed to trace amounts of Fe(II) in the hematite structure. Reduction of Pu(V) was found to be the rate-limiting step at concentrations higher than approximately 10(-6) M Pu(V) and is attributed to the formation of PuO(2+x)·nH2O nanoparticles and the Nernstian favorability of Pu(IV) surface complexes. The reaction order with respect to plutonium concentration was found to be -0.68 ± 0.09, indicating that there is a concentration dependence in these systems. This work strongly suggests that the kinetics of experiments carried out under high plutonium concentrations (i.e., >10(-7) M Pu) cannot be directly extrapolated to environmental concentrations of plutonium. PMID:25003955

  2. Method for dissolving delta-phase plutonium

    DOEpatents

    Karraker, David G.

    1992-01-01

    A process for dissolving plutonium, and in particular, delta-phase plutonium. The process includes heating a mixture of nitric acid, hydroxylammonium nitrate (HAN) and potassium fluoride to a temperature between 40.degree. and 70.degree. C., then immersing the metal in the mixture. Preferably, the nitric acid has a concentration of not more than 2M, the HAN approximately 0.66M, and the potassium fluoride 0.1M. Additionally, a small amount of sulfamic acid, such as 0.1M can be added to assure stability of the HAN in the presence of nitric acid. The oxide layer that forms on plutonium metal may be removed with a non-oxidizing acid as a pre-treatment step.

  3. Interaction of divalent plutonium and curium

    SciTech Connect

    Mikheev, N.B.; Kazakevich, M.Z.; Rumer, I.A.

    1988-11-01

    It has been established that at plutonium concentrations ranging from 10/sup -5/ to 10/sup -4/ mole % the oxidation potentials of the Pu/sup 3 +//Pu/sup 2 +/ and Cm/sup 3 +//Cm/sup 2 +/ pairs increased by 0.15-0.2 V due to the dimerization of Pu/sup 2 +/ and the formation of mixed dimers of plutonium and curium. Promethium(2+) does not have a similar ability to form mixed dimers owing to the fact that Pm/sup 2 +/ does not have a free d electron. The oxidation potential of the Pm/sup 3 +//Pm/sup 2 +/ pair does not vary in the presence of massive quantities of plutonium

  4. Plutonium Finishing Plant safety evaluation report

    SciTech Connect

    Not Available

    1995-01-01

    The Plutonium Finishing Plant (PFP) previously known as the Plutonium Process and Storage Facility, or Z-Plant, was built and put into operation in 1949. Since 1949 PFP has been used for various processing missions, including plutonium purification, oxide production, metal production, parts fabrication, plutonium recovery, and the recovery of americium (Am-241). The PFP has also been used for receipt and large scale storage of plutonium scrap and product materials. The PFP Final Safety Analysis Report (FSAR) was prepared by WHC to document the hazards associated with the facility, present safety analyses of potential accident scenarios, and demonstrate the adequacy of safety class structures, systems, and components (SSCs) and operational safety requirements (OSRs) necessary to eliminate, control, or mitigate the identified hazards. Documented in this Safety Evaluation Report (SER) is DOE`s independent review and evaluation of the PFP FSAR and the basis for approval of the PFP FSAR. The evaluation is presented in a format that parallels the format of the PFP FSAR. As an aid to the reactor, a list of acronyms has been included at the beginning of this report. The DOE review concluded that the risks associated with conducting plutonium handling, processing, and storage operations within PFP facilities, as described in the PFP FSAR, are acceptable, since the accident safety analyses associated with these activities meet the WHC risk acceptance guidelines and DOE safety goals in SEN-35-91.

  5. PLUTONIUM-ZIRCONIUM ALLOYS

    DOEpatents

    Schonfeld, F.W.; Waber, J.T.

    1960-08-30

    A series of nuclear reactor fuel alloys consisting of from about 5 to about 50 at.% zirconium (or higher zirconium alloys such as Zircaloy), balance plutonium, and having the structural composition of a plutonium are described. Zirconium is a satisfactory diluent because it alloys readily with plutonium and has desirable nuclear properties. Additional advantages are corrosion resistance, excellent fabrication propenties, an isotropie structure, and initial softness.

  6. METHOD OF SEPARATING PLUTONIUM

    DOEpatents

    Heal, H.G.

    1960-02-16

    BS>A method of separating plutonium from aqueous nitrate solutions of plutonium, uranium. and high beta activity fission products is given. The pH of the aqueous solution is adjusted between 3.0 to 6.0 with ammonium acetate, ferric nitrate is added, and the solution is heated to 80 to 100 deg C to selectively form a basic ferric plutonium-carrying precipitate.

  7. Continuous plutonium dissolution apparatus

    DOEpatents

    Meyer, F.G.; Tesitor, C.N.

    1974-02-26

    This invention is concerned with continuous dissolution of metals such as plutonium. A high normality acid mixture is fed into a boiler vessel, vaporized, and subsequently condensed as a low normality acid mixture. The mixture is then conveyed to a dissolution vessel and contacted with the plutonium metal to dissolve the plutonium in the dissolution vessel, reacting therewith forming plutonium nitrate. The reaction products are then conveyed to the mixing vessel and maintained soluble by the high normality acid, with separation and removal of the desired constituent. (Official Gazette)

  8. Trawsfynydd Plutonium Estimate

    SciTech Connect

    Reid, Bruce D.; Gerlach, David C.; Heasler, Patrick G.; Livingston, J.

    2009-11-20

    Report serves to document an estimate of the cumulative plutonium production of the Trawsfynydd Unit II reactor (Traws II) over its operating life made using the Graphite Isotope Ratio Method (GIRM). The estimate of the plutonium production in Traws II provided in this report has been generated under blind conditions. In other words, the estimate ofthe Traws II plutonium production has been generated without the knowledge of the plutonium production declared by the reactor operator (Nuclear Electric). The objective of this report is to demonstrate that the GIRM can be employed to serve as an accurate tool to verify weapons materials production declarations.

  9. GLASS FABRICATION AND PRODUCT CONSISTENCY TESTING OF LANTHANIDE BOROSILICATE FRIT B COMPOSITION FOR PLUTONIUM DISPOSITION

    SciTech Connect

    Marra, J

    2006-01-19

    leachate solutions were ultrafiltered to quantify colloid formation. The leached solids from select PCTs were examined in an attempt to evaluate the Pu and neutron absorber release behavior from the glass and to identify the formation of alteration phases on the glass surface. Characterization of the glass prior to testing revealed that some undissolved plutonium oxide was present in the glass. The undissolved particles had a disk-like morphology and likely formed via coarsening of particles in areas compositionally enriched in plutonium. Similar disk-like PuO{sub 2} phases were observed in previous LaBS glass testing at PNNL. In that work, researchers concluded that plutonium formed with this morphology as a result of the leaching process. It was more likely that the presence of the plutonium oxide crystals in the PNNL testing was a result of glass fabrication. A series of PCTs were conducted at 90 C in ASTM Type 1 water. The PCT-Method A (PCT-A) was conducted to compare the Pu LaBS Frit B glass durability to current requirements for High Level Waste (HLW) glass in a geologic repository. The PCT-A test has a strict protocol and is designed to specifically be used to evaluate whether the chemical durability and elemental release characteristics of a nuclear waste glass have been consistently controlled during production and, thus, meet the repository acceptance requirements. The PCT-A results on the Pu containing LaBS Frit B glass showed that the glass was very durable with a normalized elemental release value for boron of approximately 0.02 g/L. This boron release value was better than two orders of magnitude better from a boron release standpoint than the current Environmental Assessment (EA) glass used for repository acceptance. The boron release value for EA glass is 16.7 g/L.

  10. Relaxation of intrapulmonary artery and vein by nitrogen oxide-containing vasodilators and cyclic GMP

    SciTech Connect

    Edwards, J.C.; Ignarro, L.J.; Hyman, A.L.; Kadowitz, P.J.

    1984-01-01

    The present study examines the relationship between tissue cyclic nucleotide levels and relaxation of bovine intrapulmonary arterial and venous smooth muscle in response to nitroglycerin, nitroprusside, S-nitroso-N-acetylpenicillamine and isoproterenol. Recent studies have suggested that cyclic GMP may be involved in the relaxation of vascular smooth muscle produced by nitrogen oxide-containing vasodilators and that S-nitrosothiols may act as intermediates of the latter agents. In the present study, nitroglycerin, nitroprusside and S-nitroso-N-acetylpenicillamine were more potent as relaxants of venous than arterial segments. Each of these agents elevated tissue cyclic GMP levels, but not cyclic AMP levels, before relaxation. These nitrogen oxide-containing agents were more potent as elevators of cyclic GMP levels in venous than arterial tissue and this correlated generally with their effects on vascular smooth muscle tone. Methylene blue antagonized both relaxation and increased cyclic GMP levels elicited by nitroglycerin, nitroprusside and S-nitroso-N-acetylpenicillamine. In contrast to the nitrogen oxide vasodilators, 8-bromo-cyclic GMP was equally effective in reducing induced tone in arterial or venous segments. Similarly, isoproterenol relaxed arterial and venous segments with equivalent sensitivities. Relaxation by isoproterenol was preceded by or occurred concomitantly with increased levels of cyclic AMP but not cyclic GMP and both effects were antagonized by propranolol. These findings are consistent with the hypothesis that vascular smooth muscle relaxation in response to nitrogen oxide-containing vasodilators or isoproterenol may be mediated or modulated by the intracellular accumulation of cyclic GMP or cyclic AMP, respectively.

  11. Criticality Evaluation of Plutonium-239 Moderated by High-Density Polyethylene in Stainless Steel and Aluminum Containers Suitable for Non-Exclusive Use Transport

    SciTech Connect

    Watson, T T

    2007-08-10

    Research is conducted at the Joint Actinide Shock Physics Experimental Facility (JASPER) on the effects of high pressure and temperature environments on plutonium-239, in support of the stockpile stewardship program. Once an experiment has been completed, it is necessary to transport the end products for interim storage or final disposition. Federal shipping regulations for nonexclusive use transportation require that no more than 180 grams of fissile material are present in at least 360 kilograms of contiguous non-fissile material. To evaluate the conservatism of these regulatory requirements, a worst-case scenario of 180g {sup 239}Pu and a more realistic scenario of 100g {sup 239}Pu were modeled using one of Lawrence Livermore National Laboratory's Monte Carlo transport codes known as COG 10. The geometry consisted of {sup 239}Pu spheres homogeneously mixed with high-density polyethylene surrounded by a cube of either stainless steel 304 or aluminum. An optimized geometry for both cube materials and hydrogen-to-fissile isotope (H/X) ratio were determined for a single unit. Infinite and finite 3D arrays of these optimized units were then simulated to determine if the systems would exceed criticality. Completion of these simulations showed that the optimal H/X ratio for the most reactive units ranged from 800 to 1600. A single unit of either cube type for either scenario would not reach criticality. An infinite array was determined to reach criticality only for the 180g case. The offsetting of spheres in their respective cubes was also considered and showed a considerable decrease in the number of close-packed units needed to reach criticality. These results call into question the current regulations for fissile material transport, which under certain circumstances may not be sufficient in preventing the development of a critical system. However, a conservative, theoretical approach was taken in all assumptions and such idealized configurations may not be likely to

  12. Modelling the behaviour of oxide fuels containing minor actinides with urania, thoria and zirconia matrices in an accelerator-driven system

    NASA Astrophysics Data System (ADS)

    Sobolev, V.; Lemehov, S.; Messaoudi, N.; Van Uffelen, P.; Aı̈t Abderrahim, H.

    2003-06-01

    The Belgian Nuclear Research Centre, SCK • CEN, is currently working on the pre-design of the multipurpose accelerator-driven system (ADS) MYRRHA. A demonstration of the possibility of transmutation of minor actinides and long-lived fission products with a realistic design of experimental fuel targets and prognosis of their behaviour under typical ADS conditions is an important task in the MYRRHA project. In the present article, the irradiation behaviour of three different oxide fuel mixtures, containing americium and plutonium - (Am,Pu,U)O 2- x with urania matrix, (Am,Pu,Th)O 2- x with thoria matrix and (Am,Y,Pu,Zr)O 2- x with inert zirconia matrix stabilised by yttria - were simulated with the new fuel performance code MACROS, which is under development and testing at the SCK • CEN. All the fuel rods were considered to be of the same design and sizes: annular fuel pellets, helium bounded with the stainless steel cladding, and a large gas plenum. The liquid lead-bismuth eutectic was used as coolant. Typical irradiation conditions of the hottest fuel assembly of the MYRRHA subcritical core were pre-calculated with the MCNPX code and used in the following calculations as the input data. The results of prediction of the thermo-mechanical behaviour of the designed rods with the considered fuels during three irradiation cycles of 90 EFPD are presented and discussed.

  13. Accident resistant transport container

    DOEpatents

    Andersen, John A.; Cole, James K.

    1980-01-01

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident.

  14. Accident resistant transport container

    DOEpatents

    Anderson, J.A.; Cole, K.K.

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident.

  15. Ambiguous and forbidden parameter combinations for aqueous plutonium.

    PubMed

    Silver, G L

    2009-01-01

    The concepts of forbidden and ambiguous oxidation-state distributions for plutonium are easier to understand when presented graphically. This note describes two diagrams that illustrate the phenomena. PMID:18986812

  16. Fine Astrocyte Processes Contain Very Small Mitochondria: Glial Oxidative Capability May Fuel Transmitter Metabolism.

    PubMed

    Derouiche, Amin; Haseleu, Julia; Korf, Horst-Werner

    2015-12-01

    The peripheral astrocyte process (PAP) is the glial compartment largely handling inactivation of transmitter glutamate, and supplying glutamate to the axon terminal. It is not clear how these energy demanding processes are fueled, and whether the PAP exhibits oxidative capability. Whereas the GFAP-positive perinuclear cytoplasm and stem process are rich in mitochondria, the PAP is often considered too narrow to contain mitochondria and might thus not rely on oxidative metabolism. Applying high resolution light microscopy, we investigate here the presence of mitochondria in the PAPs of freshly dissociated, isolated astrocytes. We provide an overview of the subcellular distribution and the approximate size of astrocytic mitochondria. A substantial proportion of the astrocyte's mitochondria are contained in the PAPs and, on the average, they are smaller there than in the stem processes. The majority of mitochondria in the stem and peripheral processes are surprisingly small (0.2-0.4 µm), spherical and not elongate, or tubular, which is supported by electron microscopy. The density of mitochondria is two to several times lower in the PAPs than in the stem processes. Thus, PAPs do not constitute a mitochondria free glial compartment but contain mitochondria in large numbers. No juxtaposition of mitochondria-containing PAPs and glutamatergic synapses has been reported. However, the issue of sufficient ATP concentrations in perisynaptic PAPs can be seen in the light of (1) the rapid, activity dependent PAP motility, and (2) the recently reported activity-dependent mitochondrial transport and immobilization leading to spatial, subcellular organisation of glutamate uptake and oxidative metabolism. PMID:25894677

  17. Effect Of Oxidation On Chromium Leaching And Redox Capacity Of Slag-Containing Waste Forms

    SciTech Connect

    Almond, P. M.; Stefanko, D. B.; Langton, C. A.

    2013-03-01

    The rate of oxidation is important to the long-term performance of reducing salt waste forms because the solubility of some contaminants, e.g., technetium, is a function of oxidation state. TcO4- in the salt solution is reduced to Tc(IV) and has been shown to react with ingredients in the waste form to precipitate low solubility sulfide and/or oxide phases [Shuh, et al., 1994, Shuh, et al., 2000, Shuh, et al., 2003]. Upon exposure to oxygen, the compounds containing Tc(IV) oxidize to the pertechnetate ion, Tc(VII)O4-, which is very soluble. Consequently the rate of technetium oxidation front advancement into a monolith and the technetium leaching profile as a function of depth from an exposed surface are important to waste form performance and ground water concentration predictions. An approach for measuring contaminant oxidation rate (effective contaminant specific oxidation rate) based on leaching of select contaminants of concern is described in this report. In addition, the relationship between reduction capacity and contaminant oxidation is addressed. Chromate was used as a non-radioactive surrogate for pertechnetate in simulated waste form samples. Depth discrete subsamples were cut from material exposed to Savannah River Site (SRS) field cured conditions. The subsamples were prepared and analyzed for both reduction capacity and chromium leachability. Results from field-cured samples indicate that the depth at which leachable chromium was detected advanced further into the sample exposed for 302 days compared to the sample exposed to air for 118 days (at least 50 mm compared to at least 20 mm). Data for only two exposure time intervals is currently available. Data for additional exposure times are required to develop an equation for the oxidation front progression. Reduction capacity measurements (per the Angus-Glasser method, which is a measurement of the ability of a material to chemically reduce Ce(IV) to Ce

  18. A Note on the Reaction of Hydrogen and Plutonium

    SciTech Connect

    Noone, Bailey C

    2012-08-15

    Plutonium hydride has many practical and experimental purposes. The reaction of plutonium and hydrogen has interesting characteristics, which will be explored in the following analysis. Plutonium is a radioactive actinide metal that emits alpha particles. When plutonium metal is exposed to air, the plutonium oxides and hydrides, and the volume increases. PuH{sub 2} and Pu{sub 2}O{sub 3} are the products. Hydrogen is a catalyst for plutonium's corrosion in air. The reaction can take place at room temperature because it is fairly insensitive to temperature. Plutonium hydride, or PuH{sub 2}, is black and metallic. After PuH{sub 2} is formed, it quickly flakes off and burns. The reaction of hydrogen and plutonium is described as pyrophoric because the product will spontaneously ignite when oxygen is present. This tendency must be considered in the storage of metal plutonium. The reaction is characterized as reversible and nonstoichiometric. The reaction goes as such: Pu + H{sub 2} {yields} PuH{sub 2}. When PuH{sub 2} is formed, the hydrogen/plutonium ratio is between 2 and 2.75 (approximately). As more hydrogen is added to the system, the ratio increases. When the ratio exceeds 2.75, PuH{sub 3} begins to form along with PuH{sub 2}. Once the ratio surpasses 2.9, only PuH{sub 3} remains. The volume of the plutonium sample increases because of the added hydrogen and the change in crystal structure which the sample undergoes. As more hydrogen is added to a system of metal plutonium, the crystal structure evolves. Plutonium has a crystal structure classified as monoclinic. A monoclinic crystal structure appears to be a rectangular prism. When plutonium reacts with hydrogen, the product PuH{sub 2}, becomes a fluorite structure. It can also be described as a face centered cubic structure. PuH{sub 3} forms a hexagonal crystal structure. As plutonium evolves from metal plutonium to plutonium hydride to plutonium trihydride, the crystal structure evolves from monoclinic to

  19. Synthesis of ligands containing N-oxide donor atoms and their assembly into metallosupramolecular structures.

    PubMed

    Bullock, Samantha J; Harding, Lindsay P; Moore, Michelle P; Mills, Andrew; Piela, Sean A F; Rice, Craig R; Towns-Andrews, Liz; Whitehead, Martina

    2013-04-28

    A series of ligands that contain both N-donor and N-oxide donor atoms have been synthesised and ligands L5, L6, L7 and L8 contain, 4, 6, 5, and 6 donor atoms respectively. The smallest ligand L5 forms a mononuclear complex with Cu2+ ([Cu(L5)(ClO4)2(MeCN)]) whereas L6 and L7 form dinuclear double helicates with Ni2+ and Cu2+ respectively ([Ni2(L6)2]4+ and [Cu2(L7)2]4+). L8 forms a tetranuclear cyclic helicate upon reaction with Co2+ ([Co4(L8)4]8+) and in all cases the complexes are characterised by single-crystal X-ray diffraction and ESI-MS. The N-oxide units imparts flexibility in the ligand strand and where the unoxidised ligand strand forms a cyclic helicate, the incorporation of an N-oxide unit allows the formation of the dinuclear double helicate. PMID:23462743

  20. 30. VIEW OF A GLOVEBOX LINE USED IN PLUTONIUM OPERATIONS. ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    30. VIEW OF A GLOVEBOX LINE USED IN PLUTONIUM OPERATIONS. SAFETY AND HEALTH CONCERNS WERE OF MAJOR IMPORTANCE AT THE PLANT, BECAUSE OF THE RADIOACTIVE NATURE OF THE MATERIALS USED. PLUTONIUM GIVES OFF ALPHA AND BETA PARTICLES, GAMMA PROTONS, NEUTRONS, AND IS ALSO PYROPHORIC. AS A RESULT, PLUTONIUM OPERATIONS ARE PERFORMED UNDER CONTROLLED CONDITIONS THAT INCLUDE CONTAINMENT, FILTERING, SHIELDING, AND CREATING AN INERT ATMOSPHERE. PLUTONIUM WAS HANDLED WITHIN GLOVEBOXES THAT WERE INTERCONNECTED AND RAN SEVERAL HUNDRED FEET IN LENGTH (5/5/70). - Rocky Flats Plant, Bounded by Indiana Street & Routes 93, 128 & 72, Golden, Jefferson County, CO