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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. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. Precipitation of plutonium from acidic solutions using magnesium oxide

    SciTech Connect

    Jones, S.A.

    1994-12-05

    Magnesium oxide will be used as a neutralizing agent for acidic plutonium-containing solutions. It is expected that as the magnesium oxide dissolves, the pH of the solution will rise, and plutonium will precipitate. The resulting solid will be tested for suitability to storage. The liquid is expected to contain plutonium levels that meet disposal limit requirements.

  11. 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}.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. Uranium plutonium oxide fuels. [LMFBR

    SciTech Connect

    Cox, C.M.; Leggett, R.D.; Weber, E.T.

    1981-01-01

    Uranium plutonium oxide is the principal fuel material for liquid metal fast breeder reactors (LMFBR's) throughout the world. Development of this material has been a reasonably straightforward evolution from the UO/sub 2/ used routinely in the light water reactor (LWR's); but, because of the lower neutron capture cross sections and much lower coolant pressures in the sodium cooled LMFBR's, the fuel is operated to much higher discharge exposures than that of a LWR. A typical LMFBR fuel assembly is shown. Depending on the required power output and the configuration of the reactor, some 70 to 400 such fuel assemblies are clustered to form the core. There is a wide variation in cross section and length of the assemblies where the increasing size reflects a chronological increase in plant size and power output as well as considerations of decreasing the net fuel cycle cost. Design and performance characteristics are described.

  6. 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.

  7. Cleanup of plutonium oxide reduction black salts

    SciTech Connect

    Giebel, R.E.; Wing, R.O.

    1986-12-17

    This work describes pyrochemical processes employed to convert direc oxide reduction (DOR) black salts into discardable white salt and plutonium metal. The DOR process utilizes calcium metal as the reductant in a molten calcium chloride solvent salt to convert plutonium oxide to plutonium metal. An insoluble plutonium-rich dispersion called black salt sometimes forms between the metal phase and the salt phase. Black salts accumulated for processing were treated by one of two methods. One method utilized a scrub alloy of 70 wt % magnesium/30 wt % zinc. The other method utilized a pool of plutonium metal to agglomerate the metal phase. The two processes were similar in that calcium metal reductant and calcium chloride solvent salt were used in both cases. Four runs were performed by each method, and each method produced greater than 93% conversion of the black salt.

  8. Development of characterization of plutonium storage containers

    SciTech Connect

    James, D.; Stevkovski, S.

    1999-02-01

    As a result of the end of the Cold War, at least 11,000 (possibly 20,000 or more) plutonium pits are projected to be stored at Pantex for up to fifty years. The current pit container, the ALR8 was not designed for this length of storage duration. As a result, Pantex officials have searched for alternative container options. The objective of this research is to develop and validate a model to predict the temperature distribution within the stored components and the internal structure of the proposed ALR8(SI) container, and to consider and analyze the safety features of the ALR8(SI) container as seen from the thermal performance view. Due to the time scale involved with the current simulations, the radioactive decay of the plutonium may be assumed to provide a uniform rate of heat generation. This heat is conducted to the surroundings through the solid structures of the assembly. In addition to conduction, the inert gas that fills the volume within the steel container convects a fraction of the generated heat from the plutonium to the colder steel surfaces. Radiation must also be accounted for as natural convection and limited conduction paths are present within the container. The research efforts in this project have been directed into two paths, numerical and experimental. First, the temperature distribution within the stored components are being determined experimentally as a function of fill gases, energy generation rate, and boundary conditions. Second, a finite element model of the ALR8 container has been developed so that the temperature distribution can be predicted as a function of the same experimental parameters. This paper presents the experimental method and data that have been obtained thus far, as well as the finite element model created using SDRC I-DEAS.

  9. Criteria for safe storage of plutonium metals and oxides

    SciTech Connect

    Not Available

    1994-12-01

    This standard establishes safety criteria for safe storage of plutonium metals and plutonium oxides at DOE facilities; materials packaged to meet these criteria should not need subsequent repackaging to ensure safe storage for at least 50 years or until final disposition. The standard applied to Pu metals, selected alloys (eg., Ga and Al alloys), and stabilized oxides containing at least 50 wt % Pu; it does not apply to Pu-bearing liquids, process residues, waste, sealed weapon components, or material containing more than 3 wt % {sup 238}Pu. Requirements for a Pu storage facility and safeguards and security considerations are not stressed as they are addressed in detail by other DOE orders.

  10. 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

  11. 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.

  12. 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.

  13. Criteria for Preparing and Packaging Plutonium Metals and Oxides for Long-Term Storage

    SciTech Connect

    1994-12-01

    This Standard provides criteria for packaging of plutonium metals and stabilized oxides for storage periods of at least 50 years. To meet the criteria, plutonium-bearing materials must be in stable forms and be packaged in containers designed to maintain their integrity both under normal storage conditions and during anticipated handling accidents.

  14. Development of weld closure stations for plutonium long-term storage containers

    SciTech Connect

    Fernandez, R.; Martinez, D.A.; Martinez, H.E.; Nelson, T.O.; Ortega, R.E.; Rofer, C.K.; Romero, W.; Stewart, J.; Trujillo, V.L.

    1998-12-31

    Weld closure stations for plutonium long-term storage containers have been designed, fabricated, and tested for the Advanced Recovery and Integrated Extraction System (ARIES) at the TA-55 Plutonium Facility of the Los Alamos National Laboratory. ARIES is a processing system used for the dismantlement of the plutonium pits from nuclear weapons. ARIES prepares the extracted-plutonium in a form which is compatible with long-term storage and disposition options and meets international inspection requirements. The processed plutonium is delivered to the canning module of the ARIES line, where it is packaged in a stainless steel container. This container is then packaged in a secondary container for long-term storage. Each of the containers is hermetically sealed with a full penetration weld closure that meets the requirements of the ASME Section IX Boiler and Pressure Vessel Code. Welding is performed with a gas tungsten arc process in an inert atmosphere of helium. The encapsulated helium in the nested containers allows for leak testing the weld closure and container. The storage package was designed to meet packaging requirements of DOE Standard 3013-96 for long-term storage of plutonium metal and oxides. Development of the process parameters, weld fixture, weld qualification, and the welding chambers is discussed in this paper.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. A portable concentrator for processing plutonium containing solutions

    SciTech Connect

    Chamberlain, D.B.; Conner, C.; Chen, L.

    1995-12-01

    This report describes a horizontal, compact agitated-film concentrator called a Rototherm, manufactured by Artisan Industries, Inc. which can be used to process aqueous solutions of radioactive wastes containing plutonium. The unit is designed to concentrate liquid streams to a high-solid content slurry.

  1. 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.

  2. 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

  3. 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.

  4. Introducing equipment and plutonium glove box modifications for monitoring gas generation over plutonium oxide materials.

    SciTech Connect

    Padilla, D. D.; Berg, J. M.; Carrillo, A. G.; Montoya, A. R.; Morris, J. S.; Veirs, D. K.; Martinez, M. A.; Worl, L. A.; Harradine, D. M.; Hill, D. D.

    2002-01-01

    DOE is embarking on a program to store large quantities of Pu-bearing materials for up to fifty years. Materials for long-term storage are metals and oxides that are stabilized and packaged according to the DOE storage standard. Experience with PuO, materials has shown that gases generated by catalytic and/or radiolytic processes may accumulate. Of concern are the generation of H, gas from adsorbed water and the generation of HCI or CI, gases from the radiolysis of chloride-containing salts. We have designed instrumented storage containers that mimic the inner storage can specified in the standard. The containers and surveillance equipment are interfaced with a plutonium glovebox and are designed to allow the gas composition and pressure to be monitored over time. The surveillance activities and glovebox interfaces include Raman fiber optic probes, a gas analysis sampling port, corrosion monitors, and pressure and temperature feedthrus. Data collection for these containers is automated in order to reduce worker exposure. The equipment design and glovebox modifications are presented.

  5. A simple model for neutron radiography of uranium-plutonium mixed oxide fuel pins

    NASA Astrophysics Data System (ADS)

    Panakkal, J. P.; Ghosh, J. K.

    1988-04-01

    Neutron radiography has been used for monitoring plutonium enrichment in uranium-plutonium mixed oxide fuel pellets inside welded nuclear fuel pins by correlating the optical density of radiographs at the centre of the pellets and plutonium enrichment. Optical density data corresponding to different thickness of the pellets starting from the centre towards the periphery was generated by microdensitometer scanning of neutron radiographs of the experimental fuel pins. An attempt has been made to correlate the optical density at points corresponding to different thickness segments of the pellets and thermal neutron interaction probability (product of the total macroscopic neutron cross section and the distance traversed by the neutrons). Based on the experimental data generated, a simple model for transmission of neutrons through nuclear fuel pins has been evolved. Using this model, it is possible to predict the optical density of plutonium bearing fuel pins containing pellets of different composition or diameter in neutron radiographic investigations.

  6. 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.

  7. 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.

  8. 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

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. Ultra-small plutonium oxide nanocrystals: an innovative material in plutonium science.

    PubMed

    Hudry, Damien; Apostolidis, Christos; Walter, Olaf; Janssen, Arne; Manara, Dario; Griveau, Jean-Christophe; Colineau, Eric; Vitova, Tonya; Prüssmann, Tim; Wang, Di; Kübel, Christian; Meyer, Daniel

    2014-08-11

    Apart from its technological importance, plutonium (Pu) is also one of the most intriguing elements because of its non-conventional physical properties and fascinating chemistry. Those fundamental aspects are particularly interesting when dealing with the challenging study of plutonium-based nanomaterials. Here we show that ultra-small (3.2±0.9 nm) and highly crystalline plutonium oxide (PuO2 ) nanocrystals (NCs) can be synthesized by the thermal decomposition of plutonyl nitrate ([PuO2 (NO3 )2 ]⋅3 H2 O) in a highly coordinating organic medium. This is the first example reporting on the preparation of significant quantities (several tens of milligrams) of PuO2 NCs, in a controllable and reproducible manner. The structure and magnetic properties of PuO2 NCs have been characterized by a wide variety of techniques (powder X-ray diffraction (PXRD), X-ray absorption fine structure (XAFS), X-ray absorption near edge structure (XANES), TEM, IR, Raman, UV/Vis spectroscopies, and superconducting quantum interference device (SQUID) magnetometry). The current PuO2 NCs constitute an innovative material for the study of challenging problems as diverse as the transport behavior of plutonium in the environment or size and shape effects on the physics of transuranium elements.

  15. A Calibration to Predict Concentrations of Impurities in Plutonium Oxide by Prompt Gamma Analysis

    SciTech Connect

    J.E. Narlesky; E.J. Kelly; L.A. Foster

    2005-12-13

    Prompt gamma (PG) analysis has been used to identify the presence of certain impurities in plutonium oxide, which has been stored in 3013 containers. A regression analysis was used to evaluate the trends between the count rates obtained from PG analysis and the concentration of the impurities in plutonium oxide samples measured by analytical chemistry techniques. The results of the analysis were used to obtain calibration curves, which may be used to predict the concentration of Al, Be, Cl, F, Mg, and Na in the 3013 containers. The scatter observed in the data resulted from several factors including sample geometry, error in sampling for chemical assay, statistical counting error, and intimacy of mixing of impurities and plutonium. Standards prepared by mixing plutonium oxide with CaF{sub 2}, NaCl, and KCl show that intimacy mixing and sampling error have the largest influence on the results. Although these factors are difficult to control, the calibrations are expected to yield semiquantitative results that are sufficient for the purpose of ordering or ranking.

  16. 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

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. 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)

  5. 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}.

  6. 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.

  7. 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

  8. 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.

  9. 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

  10. 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).

  11. 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.

  12. 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.

  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. 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.

  15. 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.

  16. 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

  17. 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.

  18. Effects on the long term storage container by thermal cycling alpha plutonium

    SciTech Connect

    Flamm, B.F.; Prenger, F.C.; Veirs, D.K.; Hill, D.D.; Isom, G.M.

    1998-03-01

    Experiments were conducted to determine the validity of the steady state temperature limit of 100 C established by the DOE-STD-3013-96 for storing alpha plutonium metal. Studies with an alpha plutonium ingot combined with strain gauge measurements indicate that the stainless steel storage container, yields very little (0.005 in.) to the expanding plutonium metal as it undergoes alpha beta phase transformation at temperatures above 112 C. Another experiment using an alpha plutonium rod for point loading of the container wall showed no measured deformation of the container. The results of strain measurements for alpha beta and beta alpha transformations for twenty five thermal cycles are reported. Finite element modeling using the measured data predicts that the compressive yield strength is 3,500 psi versus the literature value of 13,000 psi.

  19. 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.

  20. Precipitation of plutonium from acidic solutions using magnesium oxide

    SciTech Connect

    Jones, S.A.

    1994-09-06

    Plutonium (IV) is only marginally soluble in alkaline solution. Precipitation of plutonium using sodium or potassium hydroxide to neutralize acidic solutions produces a gelatinous solid that is difficult to filter and an endpoint that is difficult to control. If the pH of the solution is too high, additional species precipitate producing an increased volume of solids separated. The use of magnesium oxide as a reagent has advantages. It is added as a solid (volume of liquid waste produced is minimized), the pH is self-limiting (pH does not exceed about 8.5), and the solids precipitated are more granular (larger particle size) than those produced using KOH or NaOH. Following precipitation, the raffinate is expected to meet criteria for disposal to tank farms. The solid will be heated in a furnace to dry it and convert any hydroxide salts to the oxide form. The material will be cooled in a desiccator. The material is expected to meet vault storage criteria.

  1. 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.

  2. 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.

  3. Evaluation of PFP Furnace Systems for Thermal Stabilization of Washed High Chloride Plutonium Oxide Items

    SciTech Connect

    Fischer, Christopher M.; Elmore, Monte R.; Schmidt, Andrew J.; Gerber, Mark A.; Muzatko, Danielle S.; Gano, Susan R.; Thornton, Brenda M.

    2002-12-17

    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. A high-temperature furnace (same make and model as used at the RMC at Plutonium Finishing Plant) and the associated offgas system were set up at PNNL to identify system vulnerabilities and to investigate alternative materials and operating conditions that would mitigate any corrosion and plugging of furnace and offgas components. The key areas of interest for this testing were the furnace heating elements, the offgas line located inside the furnace, the offgas line between the furnace and the filter/knockout pot, the filter/knockout pot, the sample boat, and corrosion coupons to evaluate alternative materials of construction. The evaluation was conducted by charging the furnace with CeO2 that had been impregnated with a mixture of chloride salts (selected to represent the expected residual chloride salt level in washed high chloride items) and heated in the furnace in accordance with the temperature ramp rates and hold times used at PFP.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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

  9. 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.

  10. 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

  11. 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.

  12. 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.

  13. Analysis of optical density data generated from neutron radiographs of uranium-plutonium mixed oxide fuel pellets inside sealed nuclear fuel pins

    NASA Astrophysics Data System (ADS)

    Panakkal, J. P.; Ghosh, J. K.; Roy, P. R.

    1986-03-01

    A quantitative analysis of neutron radiographs of welded nuclear fuel pins containing uranium-plutonium mixed oxide fuel pellets has been carried out to obtain a simple model for the transmission of neutrons through fuel pins during neutron radiography. The optical density data obtained by detailed microdensitometer scanning across the image of pellets of varying plutonium enrichment has been correlated to the product of the macroscopic neutron cross section and the distance traversed by the neutrons. Based on the experimental data, a simple model which can be applied to fuel pins of different dimensions and plutonium enrichment has been derived.

  14. Efficient powder blending in support of plutonium conversion for mixed oxide fuel

    SciTech Connect

    Dennison, D.K.; Brucker, J.P.; Martinez, H.E.

    1999-06-07

    This paper describes a unique system that is used to mix and blend multiple batches of plutonium oxide powder of various consistencies into an equivalent number of identical and homogeneously mixed batches. This system is being designed and built to support the Advanced Recovery and Integrated Extraction System (ARIES) at the Los Alamos TA-55 Plutonium Facility. The ARIES program demonstrates dismantlement of nuclear pits, retrieval of the plutonium components, and conversion of the plutonium into an oxide for eventual use in mixed oxide (MOX) fuel for nuclear reactors. The purpose of this powder blending work is to assure that ARIES oxide is converted into an unclassified homogeneous mixture and that consistent feed material is available for MOX fuel assembly. This blending system is being assembled in a selected glovebox a TA-55 using an LANL designed split/combine apparatus, a commercial Turbula blending unit, and several additional supporting hardware components.

  15. Catalyzed Electrolytic Plutonium Oxide Dissolution (CEPOD): The past seventeen years and future potential

    SciTech Connect

    Ryan, J.L.; Bray, L.A.; Wheelwright, E.J.; Bryan, G.H.

    1990-07-01

    Catalyzed Electrolytic Plutonium Oxide Dissolution (CEPOD) was first demonstrated at PNL in early 1974 in work funded by EXXON Corporation. That work was aimed at dissolution of Pu-containing residues remaining in mixed-oxide reactor fuels dissolution and was first publicly disclosed in 1981. The process dissolves PuO{sub 2} in an anolyte containing small (catalytic) amounts of elements that form kinetically fast, strongly oxidizing ions. These are continuously regenerated at the anode. Catalysts used, in their oxidized form, include Ag{sup 2+}, Ce{sup 4+}, Co{sup 3+}, and AmO{sub 2}{sup 2+}. This paper reviews the chemistry involved in CEPOD and the results of its application to the dissolution of the Pu content of a variety of PuO{sub 2}-containing materials such as off-standard oxide, fuels dissolution residues, incinerator ash, contaminated soils, and other scrapes or wastes. Results are presented for both laboratory-scale and plant-scale dissolvers. Spin-off applications such as decontamination of metallic surfaces and destruction of organics are discussed. 27 refs., 14 figs.

  16. 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.

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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).

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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

  8. 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

  9. 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.

  10. 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.

  11. 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

  12. The underwater coincidence counter (UWCC) for plutonium measurements in mixed oxide fuels

    SciTech Connect

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

    1998-12-31

    The use of fresh uranium-plutonium mixed oxide (MOX) fuel in light-water reactors (LWR) is increasing in Europe and Japan and it is necessary to verify the plutonium content in the fuel for international safeguards purposes. The UWCC is a new instrument that has been designed to operate underwater and nondestructively measure the plutonium in unirradiated MOX fuel assemblies. The UWCC can be quickly configured to measure either boiling-water reactor (BWR) or pressurized-water reactor (PWR) fuel assemblies. The plutonium loading per unit length is measured using the UWCC to precisions of less than 1% in a measurement time of 2 to 3 minutes. Initial calibrations of the UWCC were completed on measurements of MOX fuel in Mol, Belgium. The MCNP-REN Monte Carlo simulation code is being benchmarked to the calibration measurements to allow accurate simulations for extended calibrations of the UWCC.

  13. 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.

  14. Effects of pressure on thermal transport in plutonium oxide powder.

    SciTech Connect

    Bielenberg, P.; Prenger, F. C.; Veirs, D. K.; Jones, J.

    2004-01-01

    Radial temperature profiles in plutonium oxide (PuO{sub 2}) powder were measured in a cylindrical vessel over a pressure range of 0.055 to 334.4 kPa with two different fill gases, helium and argon. The fine PuO{sub 2} powder provides a very uniform self-heating medium amenable to relatively simple mathematical descriptions. At low pressures (<0.1 kPa), the effective thermal conductivity of the powder bed was approximately the same with either helium or argon since the dominant mechanisms are thermal radiation between particles and solid-solid conduction pathways. At high pressures, the effective thermal conductivity of the powder bed is typically assumed in the literature to be dominated by the gas thermal conductivities. However, from experimental measurements at high pressures, the effective thermal conductivity of the powder bed with argon as a fill gas is approximately three times higher than would be predicted from the gas thermal conductivities. Additionally, a significant pressure dependence was measured at pressures greater than atmospheric where the gas thermal conductivity would typically be assumed to be in the continuum limit and independent of pressure. An analytical model was developed for heat conduction in the fine ceramic powder with conduction pathways in parallel and in series through the gaseous and solid components. Many analytical models in the literature were unsuitable for this system because they make limiting assumptions about the particle dimensions and shape and are developed for packed beds with higher packing fractions. PuO{sub 2} powder has small particle sizes (on the order of 1 to 10 {mu}m), random particle shapes, and high porosity so a more general model was required for this system. The model correctly predicts the temperature profiles of the powder over the wide pressure range for both argon and helium as fill gases. The effective thermal conductivity of the powder bed exhibits a pressure dependence at higher pressures because

  15. 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.

  16. Test plan for demonstrating plutonium extraction from 10-L solutions using EIChrom extraction chromatographic resins

    SciTech Connect

    Barney, G.S.

    1994-08-15

    Corrosive plutonium solutions stored in 10-L containers at the Plutonium Finishing Plant must be treated to convert the plutonium to a safe, solid form for storage and to remove the americium so that radiation exposure can be reduced. Extraction chromatographic resins will be tested for separating plutonium from these solutions in the laboratory. Separation parameters will be developed during the testing for large scale processing of the 10-L solutions and solutions of similar composition. Use of chromatographic resins will allow plutonium separation with minimum of chemical addition to the feed and without the need for plutonium valence adjustment. The separated plutonium will be calcined to plutonium oxide by direct solution calcination.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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

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

    SciTech Connect

    Schulte, L.D.; Silver, G.L.; Avens, L.R.; Jarvinen, G.D.; Espinoza, J.; Foltyn, E.M.; Rinehart, G.H.

    1996-12-31

    Nuclear Materials Technology (NMT) Division 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. The 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 notable non-actinide impurities were silicon and phosphorus, but aluminum, chromium, iron and nickel were also near or in excess of limits specified by GPHS fuel powder specifications. Among actinide impurities, uranium is of paramount concern because {sup 234}U is the daughter of {sup 2238}Pu by alpha decay, and is the largest actinide impurity. 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 first dissolved in refluxing HNO{sub 3}/HF and then the solution was 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 glovebox environment. Plutonium-238 materials targeted for recovery includes impure oxide and scrap items that are lean in {sup 238}Pu values.

  10. 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.

  11. 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.

  12. A calibration to predict the concentrations of impurities in plutonium oxide by prompt gamma analysis: Revision 1

    SciTech Connect

    Narlesky, Joshua E.; Foster, Lynn A.; Kelly, Elizabeth J.; Murray, Roy E., IV

    2009-12-01

    Over 5,500 containers of excess plutonium-bearing materials have been packaged for long-term storage following the requirements of DOE-STD- 3013. Knowledge of the chemical impurities in the packaged materials is important because certain impurities, such as chloride salts, affect the behavior of the material in storage leading to gas generation and corrosion when sufficient moisture also is present. In most cases, the packaged materials are not well characterized, and information about the chemical impurities is limited to knowledge of the material’s processing history. The alpha-particle activity from the plutonium and americium isotopes provides a method of nondestructive self-interrogation to identify certain light elements through the characteristic, prompt gamma rays that are emitted from alpha-particle-induced reactions with these elements. Gamma-ray spectra are obtained for each 3013 container using a highresolution, coaxial high-purity germanium detector. These gamma-ray spectra are scanned from 800 to 5,000 keV for characteristic, prompt gamma rays from the detectable elements, which include lithium, beryllium, boron, nitrogen, oxygen, fluorine, sodium, magnesium, aluminum, silicon, phosphorus, chlorine, and potassium. The lower limits of detection for these elements in a plutonium-oxide matrix increase with atomic number and range from 100 or 200 ppm for the lightest elements such as lithium and beryllium, to 19,000 ppm for potassium. The peak areas from the characteristic, prompt gamma rays can be used to estimate the concentration of the light-element impurities detected in the material on a semiquantitative basis. The use of prompt gamma analysis to assess impurity concentrations avoids the expense and the risks generally associated with performing chemical analysis on radioactive materials. The analyzed containers are grouped by impurity content, which helps to identify high-risk containers for surveillance and in sorting materials before packaging.

  13. 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

  14. The 871 keV gamma ray from 17O and the identification of plutonium oxide

    NASA Astrophysics Data System (ADS)

    Peurrung, Anthony; Arthur, Richard; Elovich, Robert; Geelhood, Bruce; Kouzes, Richard; Pratt, Sharon; Scheele, Randy; Sell, Richard

    2001-12-01

    Disarmament agreements and discussions between the United States and the Russian Federation for reducing the number of stockpiled nuclear weapons require verification of the origin of materials as having come from disassembled weapons. This has resulted in the identification of measurable "attributes" that characterize such materials. It has been proposed that the 871 keV gamma ray of 17O can be observed as an indicator of the unexpected presence of plutonium oxide, as opposed to plutonium metal, in such materials. We have shown that the observation of the 871 keV gamma ray is not a specific indicator of the presence of the oxide, but rather indicates the presence of nitrogen.

  15. 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)

  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. 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.

  18. 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.

  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. 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.

  1. 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.

  2. Test Results Using a Bell Jar to Measure Containment Vessel Pressurization

    SciTech Connect

    Hensel, S.J.

    2002-05-10

    A bell jar is used to determine containment vessel pressurization due to outgassing of plutonium materials. Fifteen food cans containing plutonium bearing materials, including plutonium packaged in direct contact with plastic and plutonium contaminated enriched oxide have been tested to date.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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

  9. 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 user's manual is addressed to the Hanford 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 titles, 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 in Chapter 7. Chapter 8 contains reference articles on measuring relative plutonium isotopics in solid samples.

  10. 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.

  11. 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.

  12. 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.

  13. 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

  14. 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.

  15. 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.

  16. Oxidation behaviour of plutonium rich (U, Pu)C and (U, Pu)O2

    NASA Astrophysics Data System (ADS)

    Sali, S. K.; Kulkarni, N. K.; Phatak, Rohan; Agarwal, Renu

    2016-10-01

    Oxidation behaviour of (U0.3Pu0.7)C1.06 was investigated in air by heating samples up to 1073 K and 1273 K. Thermogravimetry (TG) of the samples and X-ray powder diffraction (XRD) of the intermediate products were used to understand the phenomenon taking place during this process. Theoretical calculations were carried out to understand the multiple phase changes taking place during oxidation of carbide. Theoretical results were validated by XRD analysis of the products obtained at different stages of oxidation. The final oxidized products were found to be a single FCC phase with O/M = 2.15 (M = U + Pu). Oxidation kinetic studies of (U0.3Pu0.7)O2 and (U0.47Pu0.53)O2 were carried out in dry air, using thermogravimetry, under non-isothermal conditions. The activation energy of oxidation was found to be 49 and 70 kJ/mol, respectively. Lattice parameter dependence on Pu/M and O/M of plutonium rich mixed oxide (MOX) was established using combined results of XRD and TG analysis of (U0.3Pu0.7)O2+x and (U0.47Pu0.53)O2+x.

  17. 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.

  18. 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.

  19. 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.

  20. Assay of scrap plutonium oxide by thermal neutron multiplicity counting for IAEA verification of excess materials from nuclear weapons production

    SciTech Connect

    Stewart, J.E.; Krick, M.S.; Xiao, J.; LeMaire, R.J.; Fotin, V.; McRae, L.; Scott, D.; Westsik, G.

    1996-09-01

    The US Nonproliferation and Export Control Policy commits the US to placing under International Atomic Energy Agency (IAEA) safeguards excess nuclear materials no longer needed for the US nuclear deterrent. As of January 1,1996, the IAEA has completed Initial Physical Inventory Verification (IPIV) at the Oak Ridge Y-12 plant, the Hanford Plutonium Finishing Plant, and a plutonium storage vault at Rock Flats. Two IPIVs were performed at Hanford . This paper reports the results of thermal neutron multiplicity assay of plutonium residues during the second IPIV at Hanford. Using the Three Ring Multiplicity Counter (3RMC), measurements were performed on 69 individual cans of plutonium residues, each containing approximately 1 kg of material. Of the 69 items, 67 passed the IAEA acceptance criteria and two were selected for destructive analysis.

  1. 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.

  2. Criticality experiments with planar arrays of three-liter bottles containing plutonium nitrate solution

    SciTech Connect

    Durst, B.M.; Clayton, E.D.; Smith, J.H.

    1985-01-01

    The objective of these experiments was to provide benchmark data to validate calculational codes used in critically safety assessments of plant configurations. Arrays containing up to as many as sixteen three-liter bottles filled with plutonium nitrate were used in the experiments. A split-table device was used in the final assembly of the arrays. Ths planar arrays were reflected with close fitting plexiglas on each side and on the bottom but not the top surface. The experiments addressed a number of factors effecting criticality: the critical air gap between bottles in an array of fixed number of bottles, the number of bottles required for criticality if the bottles were touching, and the effect on critical array spacing and critical bottle number due to the insertion of an hydrogeneous substance into the air gap between bottles. Each bottle contained about 2.4l of Pu(NO{sub 3}){sub 4} solution at a Pu concentration of 105g Pu/l, with the {sup 240}Pu content being 2.9 wt% at a free acid molarity H{sup +} of 5.1. After the initial series of experiments were performed with bottles separated by air gaps, plexiglas shells of varying thicknesses were placed around each bottle to investigate how moderation between bottles affects both the number of bottles required for criticality and the critical spacing between each bottle. The minimum of bottles required for criticality was found to be 10.9 bottles, occurring for a square array with bottles in contact. As the bottles were spaced apart, the critical number increased. For sixteen bottles in a square array, the critical separation between surfaces in both x and y direction was 0.96 cm. The addition of plexiglas around each bottle decreased the critical bottle number, compared to those separated in air, but the critical bottle number, even with interstitial plastic in place was always greater than 10.9 bottles. The most reactive configuration was a tightly packed array of bottles with no intervening material.

  3. 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.

  4. Calibration Tools for Measurement of Highly Enriched Uranium in Oxide and Mixed Uranium-Plutonium Oxide with a Passive-Active Neutron Drum Shuffler

    SciTech Connect

    Mount, M; O'Connell, W; Cochran, C; Rinard, P

    2003-06-13

    Lawrence Livermore National Laboratory (LLNL) has completed an extensive effort to calibrate 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. Earlier papers described the PAN shuffler calibration over a range of item properties by standards measurements and an extensive series of detailed simulation calculations. With a single normalization factor, the simulations agree with the HEU oxide standards measurements to within {+-}1.2% at one standard deviation. Measurement errors on mixed U-Pu oxide samples are in the {+-}2% to {+-}10% range, or {+-}20 g for the smaller items. The purpose of this paper is to facilitate transfer of the LLNL procedure and calibration algorithms to external users who possess an identical, or equivalent, PAN shuffler. Steps include (1) measurement of HEU standards or working reference materials (WRMs); (2) MCNP simulation calculations for the standards or WRMs and a range of possible masses in the same containers; (3) a normalization of the calibration algorithms using the standard or WRM measurements to account for differences in the {sup 252}Cf source strength, the delayed-neutron nuclear data, effects of the irradiation protocol, and detector efficiency; and (4) a verification of the simulation series trends against like LLNL results. Tools include EXCEL/Visual Basic programs which pre- and post-process the simulations, control the normalization, and embody the calibration algorithms.

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

    SciTech Connect

    Wishau, R.; Ramsey, K.B.; Montoya, A.

    1998-12-31

    This paper presents the technical and economic feasibility of molten salt oxidation technology as a volume reduction and recovery process for {sup 238}Pu contaminated waste. Combustible low-level waste material contaminated with {sup 238}Pu residue is destroyed by oxidation in a 900 C molten salt reaction vessel. The combustible waste is destroyed creating carbon dioxide and steam and a small amount of ash and insoluble {sup 2328}Pu in the spent salt. The valuable {sup 238}Pu is recycled using aqueous recovery techniques. Experimental test results for this technology indicate a plutonium recovery efficiency of 99%. Molten salt oxidation stabilizes the waste converting it to a non-combustible waste. Thus installation and use of molten salt oxidation technology will substantially reduce the volume of {sup 238}Pu contaminated waste. Cost-effectiveness evaluations of molten salt oxidation indicate a significant cost savings when compared to the present plans to package, or re-package, certify and transport these wastes to the Waste Isolation Pilot Plant for permanent disposal. Clear and distinct cost advantages exist for MSO when the monetary value of the recovered {sup 238}Pu is considered.

  6. 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.

  7. 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

  8. Properties of Plutonium-Containing Colloids Released from Glass-Bonded Sodalite Nuclear Waste Form

    SciTech Connect

    Morss, L.R.; Mertz, C.J.; Kropf, A.J.; Holly, J.L.

    2004-10-11

    In glass-bonded sodalite, which is the ceramic waste form (CWF) to immobilize radioactive electrorefiner salt from spent metallic reactor fuel, uranium and plutonium are found as 20-50 nm (U,Pu)O{sub 2} particles encapsulated in glass near glass-sodalite phase boundaries. In order to determine whether the (U,Pu)O{sub 2} affects the durability of the CWF, and to determine release behavior of uranium and plutonium during CWF corrosion, tests were conducted to measure the release of matrix and radioactive elements from crushed CWF samples into water and the properties of released plutonium. Released colloids have been characterized by sequential filtration of test solutions followed by elemental analysis, dynamic light scattering, transmission electron microscopy (TEM), and X-ray absorption spectroscopy. This paper reports the composition, size, and agglomeration of these colloids. Significant amounts of colloidal, amorphous aluminosilicates and smaller amounts of colloidal crystalline (U,Pu)O{sub 2} were identified in test solutions. The normalized releases of uranium and plutonium were significantly less than the normalized releases of matrix elements.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. Conceptual studies for pressurised water reactor cores employing plutonium erbium zirconium oxide inert matrix fuel assemblies

    NASA Astrophysics Data System (ADS)

    Stanculescu, A.; Kasemeyer, U.; Paratte, J.-M.; Chawla, R.

    1999-08-01

    The most efficient way to enhance plutonium consumption in light water reactors is to eliminate the production of plutonium all together. This requirement leads to fuel concepts in which the uranium is replaced by an inert matrix. At PSI, studies have focused on employing ZrO 2 as inert matrix. Adding a burnable poison to such a fuel proves to be necessary. As a result of scoping studies, Er 2O 3 was identified as the most suitable burnable poison material. The results of whole-core three-dimensional neutronics analyses indicated, for a present-day 1000 MW e pressurised water reactor (PWR), the feasibility of an asymptotic equilibrium four-batch cycle fuelled solely with the proposed PuO 2-Er 2O 3-ZrO 2 inert matrix fuel (IMF). The present paper presents the results of more recent investigations related to `real-life' situations, which call for transition configurations in which mixed IMF and UO 2 assembly loadings must be considered. To determine the influence of the introduction of IMF assemblies on the characteristics of a UO 2-fuelled core, three-dimensional full-core calculations have been performed for a present-day 1000 MW e PWR containing up to 12 optimised IMF assemblies.

  15. 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)

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  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. 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.

  4. 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

  5. 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.

  6. 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.

  7. Isotopic and elemental composition of plutonium/americium oxides influence pulmonary and extra-pulmonary distribution after inhalation in rats.

    PubMed

    Van der Meeren, A; Grémy, O

    2010-09-01

    The biodistribution of plutonium and americium has been studied in a rat model after inhalation of two PuO(2) powders in lungs and extra-pulmonary organs from 3 d to 3 mo. The main difference between the two powders was the content of americium (approximately 46% and 4.5% of total alpha activity). The PuO(2) with a higher proportion of americium shows an accelerated transfer of activity from lungs to blood as compared to PuO(2) with the lower americium content, illustrated by increased urinary excretion and higher bone and liver actinide retention. The total alpha activity measured reflects mostly the americium biological behavior. The activity contained in epithelial lining fluid, recovered in the acellular phase of broncho-alveolar lavages, mainly contains americium, whereas plutonium remains trapped in macrophages. Epithelial lining fluid could represent a transitional pulmonary compartment prior to translocation of actinides to the blood and subsequent deposition in extra-pulmonary retention organs. In addition, differential behaviors of plutonium and americium are also observed between the PuO(2) powders with a higher dissolution rate for both plutonium and americium being obtained for the PuO(2) with the highest americium content. Our results indicate that the biological behavior of plutonium and americium after translocation into blood differ two-fold: (1) for the two actinides for the same PuO(2) aerosol, and (2) for the same actinide from the two different aerosols. These results highlight the importance of considering the specific behavior of each contaminant after accidental pulmonary intake when assessing extra-pulmonary deposits from the level of activity excreted in urine or for therapeutic strategy decisions.

  8. 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.

  9. Oxidation of delta-phase plutonium alloy: Corrosion kinetics in dry and humid air at 35 {degree}C

    SciTech Connect

    Haschke, J.M.

    1997-06-01

    Kinetic data for oxidation of delta-phase plutonium alloy are evaluated to provide a technical basis for assessing the merit of an existing time limitation on air exposure of components during process operations. Data describing the effects of humidity and oxygen pressure on the oxidation rate of the Pu-1.0 wt% Ga alloy at elevated temperatures are obtained from literature sources and used to predict the oxidation behavior of the alloy in air at 35 C and 0 to 100% relative humidity. A mandated six-hour limit on air exposure is inconsistent with a predicted thirty-day period required for formation of a 1-{micro}m-thick oxide layer in moisture-saturated air at 35 C. Relationships are defined for predicting kinetic behavior of the alloy at other conditions, and recommendations for addressing oxidation-related concerns in production are presented.

  10. 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.

  11. 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

  12. 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.

  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. A glass-encapsulated calcium phosphate wasteform for the immobilization of actinide-, fluoride-, and chloride-containing radioactive wastes from the pyrochemical reprocessing of plutonium metal

    SciTech Connect

    Donald, Ian W.; Metcalfe, Brian; Fong, Shirley K.; Gerrard, Lee A.; Strachan, Denis M.; Scheele, Randall D.

    2007-03-31

    The presence of halide anions in four types of wastes arising from the pyrochemical reprocessing of plutonium required an immobilization process to be developed in which not only the actinide cations but also the halide anions were immobilized in a durable waste form. At AWE, we have developed such a process using Ca3(PO4)2 as the host material. Successful trials of the process with actinide- and Cl-bearing Type I waste were carried out at PNNL where the immobilization of the waste in a form resistant to aqueous leaching was confirmed. Normalized mass losses determined at 40°C and 28 days were 12 x 10-6 g∙m-2 and 2.7 x 10-3 g∙m-2 for Pu and Cl, respectively. Accelerated radiation-induced damage effects are being determined with specimens containing 238Pu. No changes in the crystalline lattice have been detected with XRD after the 239Pu equivalent of 400 years ageing. Confirmation of the process for Type II waste (a oxyhydroxide-based waste) is currently underway at PNNL. Differences in the ionic state of Pu in the four types of waste have required different surrogates to be used. Samarium chloride was used successfully as a surrogate for both Pu(III) and Am(III) chlorides. Initial investigations into the use of HfO2 as the surrogate for Pu(IV) oxide in Type II waste indicated no significant differences.

  16. 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.

  17. 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.

  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. 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.

  20. 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)

  1. Mechanics of plutonium metal aerosolization

    SciTech Connect

    Alvis, J.M.

    1996-06-01

    Reliable estimates of hazards posed by a plutonium release are contingent on the availability of technical data to define the source term for aerosolization of plutonium oxide particles and the resulting size distribution. The release of aerosols from the oxidation of plutonium metal depends partly on the forces acting on the particles while they remain attached to the bulk material and partly on the ability of the airstream around the metal ingot to transport the particles when they detach. The forces that attach or detach the plutonium oxide particles can be described as binding of the particle to the metal or oxide layer around it and expansion and contraction stresses and external vibration. Experimental data forms the basis for defining size distributions and release fractions for plutonium oxide. The relevance of the data must be evaluated in the light of the chemical and physical properties of plutonium metal, plutonium oxide, and intermediate Plutonium compounds. The effects of temperature on reaction kinetics must also be understood when evaluating experimental data. Size distribution functions are remarkably similar for products of all Pu+gas reactions. The distributions are all bimodal. Marked differences are seen in the sizes of large particles depending on reaction temperature and reaction rate. However, the size distributions of small particles are very similar. The bimodal distribution of small particles vanishes as the sizes of the large particles decrease to the point of equal dimensions with the small particles. This is the situation realized for the fine plutonium oxide powder produced by air oxidation at room temperature. This report addresses important factors for defining the formation of an aerosol from the oxidation of plutonium metal. These factors are oxidation kinetics of plutonium metal and plutonium hydride, the particle distribution of products formed by the reactions, and the kinetics of processes limiting entrainment of particles.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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)

  8. 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.

  9. 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).

  10. 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.

  11. 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.

  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. 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.

  14. 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.

  15. 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.

  16. The oxidation behavior of Co-15 wt % Cr alloy containing dispersed oxides formed by internal oxidation

    SciTech Connect

    Hou, P.Y.; Shui, Z.R.; Stringer, J.

    1991-12-01

    Internal oxidation pretreatments of Co-15wt%Cr and Co-15wt%Cr-1wt%Ti were carried out using a Rhines pack in quartz, in mullite and in alumina. A dispersion of titanium oxide particles formed in the Ti-containing alloy as a result of the internal oxidation. However, silicon also diffused into all treated specimens when the pretreatments were carried out in quartz or in mullite. The effect of various pretreatments on the subsequent oxidation of these alloys was studied at 1000{degree}C, and compared with that of Co-15wt%Cr-1wt%Si alloy. The main purpose of this study was to determine the relative effectiveness of the dispersed oxide particles and the contaminated silicon on the selective oxidation of chromium. It was found that the oxidation behavior of both treated alloys were strongly affected by the degree of silicon contamination. Selective oxidation of chromium to form a nearly continuous protective Cr{sub 2}O{sub 3} scale was achieved with greater than 0.4wt% silicon. The presence of dispersed particles reduced initial oxidation rate, but was ineffective in promoting Cr{sub 2}O{sub 3} scale formation.

  17. 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.

  18. 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.

  19. 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)

  20. 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.

  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. 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)

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  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. 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.

  10. 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.

  11. 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)

  12. 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.

  13. 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.

  14. 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)

  15. 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.

  16. Recent plutonium science and technology at ORNL

    SciTech Connect

    Bell, J.T.

    1985-01-01

    Plutonium research and development (R and D) at ORNL has generally followed development of the nuclear fuel cycle. Basic plutonium chemistry studies have diminished since the mid-1970s; however, significant efforts have been made recently to determine fundamental characteristics of the aqueous plutonium polymer and to develop thermodynamic representations of plutonium oxides. Some studies have also been made on plutonium phosphates related to waste isolation and on definition of the oxidation states of environmental plutonium. The remaining work has been supported by the Consolidated Fuel Reprocessing Program (CFRP) and includes: (1) establishment of boundary limits for polymer formation in Purex systems; (2) preparation of mixed uranium-plutonium oxide microspheres by internal gelation sol-gel techniques; (3) direct thermal denitration of aqueous systems; and (4) plutonium/uranium extraction from spent fast reactor fuels.

  17. 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.

  18. Heme-containing dioxygenases involved in tryptophan oxidation.

    PubMed

    Millett, Elizabeth S; Efimov, Igor; Basran, Jaswir; Handa, Sandeep; Mowat, Christopher G; Raven, Emma Lloyd

    2012-04-01

    Heme iron is often used in biology for activation of oxygen. The mechanisms of oxygen activation by heme-containing monooxygenases (the cytochrome P450s) are well known, and involve formation of a Compound I species, but information on the heme-containing dioxygenase enzymes involved in tryptophan oxidation lags far behind. In this review, we gather together information emerging recently from structural, mechanistic, spectroscopic, and computational approaches on the heme dioxygenase enzymes involved in tryptophan oxidation. We explore the subtleties that differentiate various heme enzymes from each other, and use this to piece together a developing picture for oxygen activation in this particular class of heme-containing dioxygenases.

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. 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

  5. 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.

  6. 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.

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

    SciTech Connect

    Siriwardane, R.V.

    1994-12-31

    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 as a matrix material, 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Plutonium and surrogate fission products in a composite ceramic waste form.

    SciTech Connect

    Esh, D. W.; Frank, S. M.; Goff, K. M.; Johnson, S. G.; Moschetti, T. L.; O'Holleran, T.

    1999-05-19

    Argonne National Laboratory is developing a ceramic waste form to immobilize salt containing fission products and transuranic elements. Preliminary results have been presented for ceramic waste forms containing surrogate fission products such as cesium and the lanthanides. In this work results from scanning electron microscopy/energy dispersive spectroscopy and x-ray diffraction are presented in greater detail for ceramic waste forms containing surrogate fission products. Additionally, results for waste forms containing plutonium and surrogate fission products are presented. Most of the surrogate fission products appear to be silicates or aluminosilicates whereas the plutonium is usually found in an oxide form. There is also evidence for the presence of plutonium within the sodalite phase although the chemical speciation of the plutonium is not known.

  13. 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.

  14. 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

  15. Mycotoxin-Containing Diet Causes Oxidative Stress in the Mouse

    PubMed Central

    Hou, Yan-Jun; Zhao, Yong-Yan; Xiong, Bo; Cui, Xiang-Shun; Kim, Nam-Hyung; Xu, Yin-Xue; Sun, Shao-Chen

    2013-01-01

    Mycotoxins which mainly consist of Aflatoxin (AF), Zearalenone (ZEN) and Deoxynivalenol (DON) are commonly found in many food commodities. Although each component has been shown to cause liver toxicity and oxidative stress in several species, there is no evidence regarding the effect of naturally contained multiple mycotoxins on tissue toxicity and oxidative stress in vivo. In the present study, mycotoxins-contaminated maize (AF 597 µg/kg, ZEN 729 µg/kg, DON 3.1 mg/kg maize) was incorporated into the diet at three different doses (0, 5 and 20%) to feed the mice, and blood and tissue samples were collected to examine the oxidative stress related indexes. The results showed that the indexes of liver, kidney and spleen were all increased and the liver and kidney morphologies changed in the mycotoxin-treated mice. Also, the treatment resulted in the elevated glutathione peroxidase (GPx) activity and malondialdehyde (MDA) level in the serum and liver, indicating the presence of the oxidative stress. Moreover, the decrease of catalase (CAT) activity in the serum, liver and kidney as well as superoxide dismutase (SOD) activity in the liver and kidney tissue further confirmed the occurrence of oxidative stress. In conclusion, our data indicate that the naturally contained mycotoxins are toxic in vivo and able to induce the oxidant stress in the mouse. PMID:23555961

  16. 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.

  17. 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.

  18. Zirconia ceramics for excess weapons plutonium waste

    NASA Astrophysics Data System (ADS)

    Gong, W. L.; Lutze, W.; Ewing, R. C.

    2000-01-01

    We synthesized a zirconia (ZrO 2)-based single-phase ceramic containing simulated excess weapons plutonium waste. ZrO 2 has large solubility for other metallic oxides. More than 20 binary systems A xO y-ZrO 2 have been reported in the literature, including PuO 2, rare-earth oxides, and oxides of metals contained in weapons plutonium wastes. We show that significant amounts of gadolinium (neutron absorber) and yttrium (additional stabilizer of the cubic modification) can be dissolved in ZrO 2, together with plutonium (simulated by Ce 4+, U 4+ or Th 4+) and impurities (e.g., Ca, Mg, Fe, Si). Sol-gel and powder methods were applied to make homogeneous, single-phase zirconia solid solutions. Pu waste impurities were completely dissolved in the solid solutions. In contrast to other phases, e.g., zirconolite and pyrochlore, zirconia is extremely radiation resistant and does not undergo amorphization. Baddeleyite (ZrO 2) is suggested as the natural analogue to study long-term radiation resistance and chemical durability of zirconia-based waste forms.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Preliminary safety evaluation for the plutonium stabilization and packaging system

    SciTech Connect

    Shapley, J.E., Fluor Daniel Hanford

    1997-03-14

    This Preliminary Safety Evaluation (PSE) describes and analyzes the installation and operation of the Plutonium Stabilization and Packaging System (SPS) at the Plutonium Finishing Plant (PFP). The SPS is a combination of components required to expedite the safe and timely storage of Plutonium (Pu) oxide. The SPS program will receive site Pu packages, process the Pu for storage, package the Pu into metallic containers, and safely store the containers in a specially modified storage vault. The location of the SPS will be in the 2736- ZB building and the storage vaults will be in the 2736-Z building of the PFP, as shown in Figure 1-1. The SPS will produce storage canisters that are larger than those currently used for Pu storage at the PFP. Therefore, the existing storage areas within the PFP secure vaults will require modification. Other modifications will be performed on the 2736-ZB building complex to facilitate the installation and operation of the SPS.

  8. 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.

  9. Gas Analysis of Plutonium Materials Tested in a Bell Jar

    SciTech Connect

    Kies, J.

    2003-04-15

    Seventy cans containing plutonium bearing materials have been tested for gas generation in bell jars at the Savannah River Site (SRS). Gas samples from five of the tests were taken and analyzed for hydrogen, oxygen, nitrogen, methane, carbon dioxide, carbon monoxide, and nitrous oxide. Gas samples from all five cans showed significant oxygen consumption. Hydrogen and carbon dioxide were found in non-negligible quantities in all gas samples. Hydrogen and carbon dioxide generation rates are presented along with oxygen depletion rates. A description of the plutonium bearing materials is also provided.

  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 (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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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

  17. 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.

  18. Preparation of a glovebox for casting enriched plutonium.

    SciTech Connect

    Ronquillo, R. D.; Trujillo, C. M.; Trujillo, C. C.

    2002-01-01

    Objectives: Prepare existing glovebox for casting, heat treating and storing enriched plutonium, Upgrade seismic systems to reduce dispersion hazard, Upgrade atmospheric systems to reduce oxidation of plutonium, Upgrade vacuum system to prevent oxidation, InstalI/upgrade induction heating systems to melt plutonium and heat mold

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. Study of the formation process of titanium oxides containing micro arc oxidation film on Mg alloys

    NASA Astrophysics Data System (ADS)

    Song, Yingwei; Dong, Kaihui; Shan, Dayong; Han, En-Hou

    2014-09-01

    A novel micro arc oxidation (MAO) film is developed to protect Mg alloys from corrosion. This film contains plenty of titanium oxides, which exhibits higher chemical stability than traditional MAO films. Especially, the micropores on the surface of the film are in situ sealed during the film formation process instead of the sealing pores post treatment. The film formation process is investigated by the observation of surface and cross-section morphologies and analysis of chemical composition. It is found that the sizes of micropores increase but the quantities decrease with increasing oxidation voltages. The micropores are open in the initial stage of oxidation and then they are sealed gradually. The growth direction of the film takes place change at different oxidation voltages. The elements of F and Ti play a significant role in the film growth process.

  6. 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.

  7. 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.

  8. 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.

  9. 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 .

  10. 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.

  11. 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.

  12. Preparation of Small Well Characterized Plutonium Oxide Reference Materials and Demonstration of the Usefulness of Such Materials for Nondestructive Analysis

    SciTech Connect

    B.A. Guillen; S.T. Hsue; J.Y Huang; P.A. Hypes; S.M. Long; C.R. Rudy; P.A. Russo; J.E. Stewart; D.J. Temer

    2003-01-01

    Calibration of neutron coincidence and multiplicity counters for passive nondestructive analysis (NDA) of plutonium requires knowledge of the detector efficiency parameters. These are most often determined empirically. Bias from multiplication and unknown impurities may be incurred even with small plutonium metal samples. Five sets of small, pure plutonium metal standards prepared with well-known geometry and very low levels of impurities now contribute to determining accurate multiplication corrections. Recent measurements of these metal standards, with small but well-defined multiplication and negligible yield of other than fission neutrons, demonstrate an improved characterization and calibration of neutron coincidence/multiplicity counters. The precise knowledge of the mass and isotopic composition of each standard also contributes significantly to verifying the accuracy of the most precise calorimetry and gamma-ray spectroscopy measurements.

  13. Surprising coordination for plutonium in the first plutonium(III) borate.

    PubMed

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

    2011-03-21

    The first plutonium(III) borate, Pu(2)[B(12)O(18)(OH)(4)Br(2)(H(2)O)(3)]·0.5H(2)O, 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.

  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. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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)

  9. 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.

  10. 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.

  11. 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...

  12. 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...

  13. 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...

  14. 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...

  15. 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...

  16. Photocatalytic degradation of organochlorine compounds over titanium oxide and titanium oxide containing zinc oxide decatungstate and hydrogen peroxide

    SciTech Connect

    Sattari, D.

    1995-12-01

    Ten organochlorine solvents, commonly used in industry, which are di- tri- and tetra- chloro derivatives of methane, ethane and ethylene were photocatalytically degraded over titanium oxide and titanium oxide containing zinc oxide: decatungstate and hydrogen peroxide. The study has been carried out by determining the yield of organochlorine degradation at different irradiation time. It was found that for derivatives of the three compounds the degradation rates were in the order of di>tri>tetra and for those containing the same number of chlorine substituents the order was ethylene>ethane. During the degradation of each organochlorine solvent chloride was liberated. Stoichiometry for photocatalytic degradation of organochlorine solvents are C{sub x}H{sub y-z}Cl{sub z} + xO{sub 2}{yields}xCO{sub 2} + (y-z)H{sup +} + zCl{sup -}.

  17. 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

  18. 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

  19. 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.

  20. 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.

  1. Facile synthesis of hydroxymethylcytosine-containing oligonucleotides and their reactivity upon osmium oxidation.

    PubMed

    Sugizaki, Kaori; Ikeda, Shuji; Yanagisawa, Hiroyuki; Okamoto, Akimitsu

    2011-06-01

    DNA strands containing a 5-hydroxymethylcytosine ((hm)C), which have recently been found in neuron cells and embryonic stem cells, were synthesized through a facile synthetic technique. The (hm)C-containing strands were efficiently oxidized at (hm)C using an osmium oxidation assay. The (hm)C was oxidized as easily as 5-methylcytosine, which can be distinguished from unmethylated cytosine.

  2. Polymeric beads containing Cyanex 923 for actinide uptake from nitric acid medium: Studies with uranium and plutonium.

    PubMed

    Gujar, R B; Lakshmi, D Shanthana; Figoli, A; Mohapatra, P K

    2013-08-30

    Conventional phase inversion technique has been successfully applied for the preparation of the solid phase extractant (SPE), Cyanex 923 loaded polymer beads. Two types of polymer beads prepared by blending Polyetherether ketone with card (PEEKWC)/DMF with 5% Cyanex 923 (SPE-I, av bead size: 900μm) and 10% Cyanex 923(SPE-II, av. bead size: 1100μm) were evaluated for the uptake of actinide ions. The polymer beads were characterized by various physical methods such as thermal analysis, surface morphology analysis by SEM, EDAX techniques, etc. The polymer beads were used for the experiments involving the uptake of both U(VI) and Pu(IV) at tracer scale and U(VI) at milli molar concentrations from nitric acid feeds. The actinide ion uptake studies involved kinetics of metal ion sorption, adsorption isotherms, and column studies. The metal sorption capacities for U(VI) at 3M HNO3 were found to be 38.8±1.9mg and 54.5±1.7mg per g of SPE-I and SPE-II, respectively. The sorption isotherm analysis with Langmuir, D-R and Freundlisch isotherms indicated chemisorption monolayer mechanism. Column studies were also carried out using 4.5mL bed volume columns containing about 0.4 and 0.45g of SPE-I and SPE-II, respectively. The breakthrough profiles were obtained for U(VI) and the elution profiles were obtained using 1M Na2CO3 as the eluent.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  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. Plutonium worker dosimetry.

    PubMed

    Birchall, Alan; Puncher, M; Harrison, J; Riddell, A; Bailey, M R; Khokryakov, V; Romanov, S

    2010-05-01

    Epidemiological studies of the relationship between risk and internal exposure to plutonium are clearly reliant on the dose estimates used. The International Commission on Radiological Protection (ICRP) is currently reviewing the latest scientific information available on biokinetic models and dosimetry, and it is likely that a number of changes to the existing models will be recommended. The effect of certain changes, particularly to the ICRP model of the respiratory tract, has been investigated for inhaled forms of (239)Pu and uncertainties have also been assessed. Notable effects of possible changes to respiratory tract model assumptions are (1) a reduction in the absorbed dose to target cells in the airways, if changes under consideration are made to the slow clearing fraction and (2) a doubling of absorbed dose to the alveolar region for insoluble forms, if evidence of longer retention times is taken into account. An important factor influencing doses for moderately soluble forms of (239)Pu is the extent of binding of dissolved plutonium to lung tissues and assumptions regarding the extent of binding in the airways. Uncertainty analyses have been performed with prior distributions chosen for application in epidemiological studies. The resulting distributions for dose per unit intake were lognormal with geometric standard deviations of 2.3 and 2.6 for nitrates and oxides, respectively. The wide ranges were due largely to consideration of results for a range of experimental data for the solubility of different forms of nitrate and oxides. The medians of these distributions were a factor of three times higher than calculated using current default ICRP parameter values. For nitrates, this was due to the assumption of a bound fraction, and for oxides due mainly to the assumption of slower alveolar clearance. This study highlights areas where more research is needed to reduce biokinetic uncertainties, including more accurate determination of particle transport rates

  9. 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

  10. 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.

  11. 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.

  12. [Preparation of polyelectrolyte microcapsules containing ferrosoferric oxide nanoparticles].

    PubMed

    Liu, Xiao-Qing; Zheng, Chun-Li; Zhu, Jia-Bi

    2011-01-01

    In this study, polyelectrolyte microcapsules have been fabricated by biocompatible ferrosoferric oxide nanoparticles (Fe3O4 NPs) and poly allyamine hydrochloride (PAH) using layer by layer assembly technique. The Fe3O4 NPs were prepared by chemical co-precipitation, and characterized by transmission electron microscopy (TEM) and infrared spectrum (IR). Quartz cell also was used as a substrate for building multilayer films to evaluate the capability of forming planar film. The result showed that Fe3O4 NPs were selectively deposited on the surface of quartz cell. Microcapsules containing Fe3O4 NPs were fabricated by Fe3O4 NPs and PAH alternately self-assembly on calcium carbonate microparticles firstly, then 0.2 molL(-1) EDTA was used to remove the calcium carbonate. Scanning electron microscopy (SEM), Zetasizer and vibrating sample magnetometer (VSM) were used to characterize the microcapsule's morphology, size and magnetic properties. The result revealed that Fe3O4 NPs and PAH were successfully deposited on the surface of CaCO3 microparticles, the microcapsule manifested superparamagnetism, size and saturation magnetization were 4.9 +/- 1.2 microm and 8.94 emu x g(-1), respectively. As a model drug, Rhodamin B isothiocyanate labeled bovine serum albumin (RBITC-BSA) was encapsulated in microcapsule depended on pH sensitive of the microcapsule film. When pH 5.0, drug add in was 2 mg, the encapsulation efficiency was (86.08 +/- 3.36) % and the drug loading was 8.01 +/- 0.30 mg x m(L-1).

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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

  18. 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.

  19. 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.

  20. 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))))))))

  1. 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.

  2. 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

  3. 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

  4. Diesel fuel containing wax oxidates to reduce particulate emissions

    SciTech Connect

    Sprague, H.G.; Sweeney, W.M.

    1980-09-16

    Addition of 0.1 to 1.5 percent by weight of wax oxidates to a diesel fuel is found to reduce the amount of soot and invisible particles produced when the fuel is used in a diesel engine. The wax oxidates act synergistically with fuel-soluble organometallic compounds such as alkyl cyclopentadienyl manganese tricarbonyl complex salts in reducing particulates. The wax oxidates used have a ratio of neutralization number to saponification number below about 0.40 and a saybolt universal viscosity at 210* F. Higher than 1600.

  5. 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

  6. Spectrophotometric determination of plutonium-239 based on the spectrum of plutonium(III) chloride

    SciTech Connect

    Temer, D.J.; Walker, L.F.

    1994-07-01

    This report describes a spectrophotometric method for determining plutonium-239 (Pu-239) based on the spectrum of Pu(III) chloride. The authors used the sealed-reflux technique for the dissolution of plutonium oxide with hydrochloric acid (HCl) and small amounts of nitric and hydrofluoric acids. To complex the fluoride, they added zirconium, and to reduce plutonium to Pu(III), they added ascorbic acid. They then adjusted the solution to a concentration of 2 M HCl and measured the absorbances at five wavelengths of the Pu(III) chloride spectrum. This spectrophotometric determination can also be applied to samples of plutonium metal dissolved in HCl.

  7. 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.

  8. 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.

  9. 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.

  10. 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)

  11. 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.

  12. 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.

  13. 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.

  14. 46 CFR 153.1011 - Changing containment systems and hoses to and from alkylene oxide service.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 5 2012-10-01 2012-10-01 false Changing containment systems and hoses to and from... HAZARDOUS MATERIALS Operations Special Cargo Procedures § 153.1011 Changing containment systems and hoses to... approve alkylene oxide tank cleaning; and (4) No hose marked “For Alkylene Oxide Transfer Only” is...

  15. 46 CFR 153.1011 - Changing containment systems and hoses to and from alkylene oxide service.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Changing containment systems and hoses to and from... HAZARDOUS MATERIALS Operations Special Cargo Procedures § 153.1011 Changing containment systems and hoses to... approve alkylene oxide tank cleaning; and (4) No hose marked “For Alkylene Oxide Transfer Only” is...

  16. 46 CFR 153.1011 - Changing containment systems and hoses to and from alkylene oxide service.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 5 2013-10-01 2013-10-01 false Changing containment systems and hoses to and from... HAZARDOUS MATERIALS Operations Special Cargo Procedures § 153.1011 Changing containment systems and hoses to... approve alkylene oxide tank cleaning; and (4) No hose marked “For Alkylene Oxide Transfer Only” is...

  17. 46 CFR 153.1011 - Changing containment systems and hoses to and from alkylene oxide service.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 5 2011-10-01 2011-10-01 false Changing containment systems and hoses to and from... HAZARDOUS MATERIALS Operations Special Cargo Procedures § 153.1011 Changing containment systems and hoses to... approve alkylene oxide tank cleaning; and (4) No hose marked “For Alkylene Oxide Transfer Only” is...

  18. 46 CFR 153.1011 - Changing containment systems and hoses to and from alkylene oxide service.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 5 2014-10-01 2014-10-01 false Changing containment systems and hoses to and from... HAZARDOUS MATERIALS Operations Special Cargo Procedures § 153.1011 Changing containment systems and hoses to... approve alkylene oxide tank cleaning; and (4) No hose marked “For Alkylene Oxide Transfer Only” is...

  19. 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.

  20. Treatment of waste containing EDTA by chemical oxidation

    SciTech Connect

    Tucker, M.D.; Barton, L.L.; Thomson, B.M.; Wagener, B.M.; Aragon, A.

    1999-07-01

    Ethylenediaminetetraacetic acid (EDTA) is a chelating agent that has been extensively used to enhance the solubilization of heavy metal cations and release of EDTA contributes to environmental problems. EDTA is recalcitrant to microbial metabolism and chemical oxidation is considered a possible method of remedial treatment. The use of the commercially available process of MIOX Corporation generates mixed oxidants on site and this solution is markedly effective in the destruction of the chelating characteristic and the decarboxylation of EDTA. When measuring the release of C-14 from carboxyl labeled EDTA, the mixed oxidant solution was comparable to the Fenton's reaction over a broad pH range. The presence of Mn{sup 2+}, Cr{sup 3+}, or Fe{sup 3+} at levels equal to that of EDTA stimulated the rate of EDTA decomposition; however, the rate of EDTA breakdown was inhibited when the concentration of Cr{sup 3+} or Mn{sup 2+} exceeded the concentration of EDTA. The treatment of Co{sup 2+} -EDTA or Cu{sup 2+} -EDTA with mixed oxidants in the presence of ultra violet light resulted in the loss of chelation ability of EDTA. In the absence of chelated metals, over 75% of the chelation property of a 70 nM EDTA solution was destroyed in 45 min. The reaction products resulting from the use of mixed oxidants added to EDTA were non-toxic to bacteria and should not contribute to additional environmental problems.

  1. 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.

  2. 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.

  3. Investigation of the oxidative ammonolysis of propylene on oxide catalysts containing molybdenum and using the response method

    SciTech Connect

    Gadzhiev, K.N.; Adzhamov, K.Y.; Alkhazov, T.G.; Khanmamedova, A.K.

    1985-07-01

    The response method has been used to study the oxidative ammonolysis of propylene on MoO/sub 3/ and molybdenum oxide systems containing bismuth, silicon, and phosphorous ions. The response curves obtained for ammonia, propylene, CO/sub 2/, acrolein, acrylonitrile in these systems are discussed and compared with individual molybdenum trioxide. It has been shown that the modifying action of ammonia on the catalyst surfaces determines the direction of the oxidative conversion of the propylene.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. Accident Analysis for the Plutonium Finishing Plant Polycube Stabilization Process

    SciTech Connect

    NELSON-MAKI, B.B.

    2001-05-14

    The Polycube Stabilization Project involves low temperature oxidation, without combustion, of polystyrene cubes using the production muffle furnaces in Glovebox HC-21C located in the Remote Mechanical ''C'' (RMC) Line in Room 230A in the 234-52 Facility. Polycubes are polystyrene cubes containing various concentrations of plutonium and uranium oxides. Hundreds of these cubes were manufactured for criticality experiments, and currently exist as unstabilized storage forms at the Plutonium Finishing Plant (PFP). This project is designed to stabilize and prepare the polycube material for stable storage using a process very similar to the earlier processing of sludges in these furnaces. The significant difference is the quantity of hydrogenous material present, and the need to place additional controls on the heating rate of the material. This calculation note documents the analyses of the Representative Accidents identified in Section 2.4.4 of Hazards Analysis for the Plutonium Finishing Plant Polycube Stabilization Process, HNF-7278 (HNF 2000). These two accidents, ''Deflagration in Glovebox HC-21C due to Loss of Power'' and ''Seismic Failure of Glovebox HC-21C'', will be further assessed in this accident analysis.

  17. 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.

  18. 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

  19. 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.

  20. 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

  1. 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.

  2. Application of solar photocatalytic oxidation to VOC-containing airstreams

    SciTech Connect

    Magrini, K.A.; Watt, A.S.; Boyd, L.C.; Wolfrum, E.J.; Larson, S.A.; Roth, C.; Glatzmaier, G.C.

    1999-07-01

    Researchers from the National Renewable Energy Laboratory (NREL) recently conducted two pilot-scale field tests located at McClellan Air Force Base (AFB) in Sacramento, California and at the Fort Carson US Army Installation in Colorado Springs, Colorado. The objective of the tests was to determine the effectiveness of solar-powered photocatalytic oxidation (PCO) treatment units for destroying emissions of chlorinated organic compounds (trichloroethylene and dichloroethylenes) from an air stripper at ambient temperature and destroying paint solvent emissions (toluene and MEK) from a painting facility at higher temperatures. Their goals for field testing these solar-driven systems was to gather real-world treatability data and establish that the systems maintained performance during the duration of the testing. This paper reports the results of these field tests.

  3. 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.

  4. 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.

  5. 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.

  6. Fused salt processing of impure plutonium dioxide to high-purity plutonium metal

    SciTech Connect

    Mullins, L.J.; Christensen, D.C.; Babcock, B.R.

    1982-01-01

    A process for converting impure plutonium dioxide (approx. 96% pure) to high-purity plutonium metal (>99.9%) was developed. The process consists of reducing the oxide to an impure plutonium metal intermediate with calcium metal in molten calcium chloride. The impure intermediate metal is cast into an anode and electrorefined to produce high-purity plutonium metal. The oxide reduction step is being done now on a 0.6-kg scale with the resulting yield being >99.5%. The electrorefining is being done on a 4.0-kg scale with the resulting yield being 80 to 85%. The purity of the product, which averages 99.98%, is essentially insensitive to the purity of the feed metal. The yield, however, is directly dependent on the chemical composition of the feed. To date, approximately 250 kg of impure oxide has been converted to pure metal by this processing sequence. The availability of impure plutonium dioxide, together with the need for pure plutonium metal, makes this sequence a valuable plutonium processing tool.

  7. Hot gas desulfurization with sorbents containing mixed metal oxides

    SciTech Connect

    Akyurtlu, J.F.; Akyurtlu, A.

    1992-12-31

    Advanced power generation systems such as the integrated gasification combined cycle power generators and the molten carbonate fuel cells have stringent fuel gas desulfurization requirements and process economics dictates that this desulfurization be performed near the temperature of the gasification off-gas. The most advanced hot gas desulfurization technology today is based on the zinc ferrite sorbent which has several shortcomings such as zinc loss by evaporation, and incomplete regeneration due to sulfate formation. The objective of this study is to develop an improved sorbent which can reduce H{sub 2}S levels to 1 ppmv or less, which can stabilize zinc, and produce economically recoverable amounts of elemental sulfur during regeneration. For this purpose, the desulfurization performance.of sorbents prepared by the addition of various amounts of V{sub 2}0{sub 5} to the zinc ferrite sorbent is investigated. Preliminary experiments show that the sorbent containing about 4.8 mass % vanadium shows a superior desulfurization performance compared to zinc ferrite. Addition of vanadium suppresses residual sulfate formation and possibly zinc evaporation. significant quantities of elemental sulfur were observed after the regeneration of vanadium containing sorbents.

  8. Thermal Properties of Phase Change Composites Containing Ferric Oxide Nanoparticles.

    PubMed

    Wang, Jifen; Xie, Huaqing; Li, Yang

    2015-04-01

    We prepared a series of homogeneous nanocomposites by suspending Fe2O3 nanoparticles into paraffin wax (PW) matrix. Fe2O3/PW nanocomposites have reduced both solid-solid phase change latent heat capacity (Ls-s) and solid-liquid phase change latent heat capacity (Ls-l) with an increase in the mass fraction of Fe2O3 nanoparticles. There is almost equable solid-solid phase change temperature (Ts-s) between PW and Fe2O3/PW composites, as well as melting temperature (Ts-l). Fe2O3 nanoparticle addition leads to substantial enhancement in the thermal conductivity of Fe2O3/PW and the enhancement ratio increases with the nanoparticle loading. Thermal conductivity of Fe2O3/PW composite with 3.0 wt% nanoparticles is about 0.27 W/(m · K) at 15 °C, which close to that of γ-Al2O3/PW with 5.0 wt% nanoparticles but higher than that of ZnO/PW containing 5.0 wt% nanoparticles. At 60 °C, Fe2O3/PW has higher thermal conductivity than γ-A12O3/PW and ZnO/PW contained with same fraction of nanoparticles. PMID:26353577

  9. Hole doping in Al-containing nickel oxide materials to improve electrochromic performance.

    PubMed

    Lin, Feng; Nordlund, Dennis; Weng, Tsu-Chien; Moore, Rob G; Gillaspie, Dane T; Dillon, Anne C; Richards, Ryan M; Engtrakul, Chaiwat

    2013-01-23

    Electrochromic materials exhibit switchable optical properties that can find applications in various fields, including smart windows, nonemissive displays, and semiconductors. High-performing nickel oxide electrochromic materials have been realized by controlling the material composition and tuning the nanostructural morphology. Post-treatment techniques could represent efficient and cost-effective approaches for performance enhancement. Herein, we report on a post-processing ozone technique that improves the electrochromic performance of an aluminum-containing nickel oxide material in lithium-ion electrolytes. The resulting materials were studied using X-ray diffraction (XRD), scanning electron microscopy (SEM), X-ray photoelectron spectroscopy (XPS), ultraviolet-visible-near-infrared (UV-vis-NIR) spectroscopy, and X-ray absorption spectroscopy (XAS). It was observed that ozone exposure increased the Ni oxidation state by introducing hole states in the NiO(6) octahedral unit. In addition, ozone exposure gives rise to higher-performing aluminum-containing nickel oxide films, relative to nickel oxide containing both Al and Li, in terms of switching kinetics, bleached-state transparency, and optical modulation. The improved performance is attributed to the decreased crystallinity and increased nickel oxidation state in aluminum-containing nickel oxide electrochromic films. The present study provides an alternative route to improve electrochromic performance for nickel oxide materials. PMID:23249159

  10. 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)

  11. 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/.

  12. 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.

  13. Thorium plutonium (TREX) fuel for weapons-grade plutonium disposition in pressurized water reactors

    SciTech Connect

    Comfort, C.; Ferguson, C.; Klima, S.; Lilly, D.E.; Rahnema, F.

    1996-12-31

    The goal of this study was to create a pressurized water reactor (PWR) reactor assembly (17 x 17) that would burn weapons-grade plutonium (WGP). Current designs of mixed-oxide (MOX) fuels combine WGP with uranium as the fuel. MOX fuel assemblies will destroy plutonium, but only 40 to 50% of the plutonium present in the fuel. This percentage is limited by the presence of {sup 238}U in the core, which becomes {sup 239}Pu by absorption and decay. The production of plutonium counteracts the disposition of WGP in current MOX fuel designs. This problem can be overcome by replacing the uranium in a MOX design with thorium. This loss of uranium (primarily {sup 238}U) halts the production of {sup 239}Pu in the thorium plutonium (TREX) fuel. The absence of {sup 239}Pu production allows the TREX design to burn up to 85 wt% of the {sup 239}Pu, originally loaded in the fuel.

  14. 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.

  15. 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

  16. 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.

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. Removal and recovery of nitrogen and sulfur oxides from gaseous mixtures containing them

    SciTech Connect

    Cooper, H.B.H.

    1984-01-10

    A cyclic process for removing lower valence nitrogen oxides from gaseous mixtures includes treating the mixtures with an aqueous media including alkali metal carbonate and alkali metal bicarbonate and a preoxygen oxidant to form higher valence nitrogen oxides and to capture these oxides as alkali metal salts, expecially nitrites and nitrates, in a carbonate/bicarbonate-containing product aqueous media. Highly selective recovery of nitrates in high purity and yield may then follow, as by crystallization, with the carbonate and bicarbonate alkali metal salts strongly increasing the selectivity and yield of nitrates. The product nitrites are converted to nitrates by oxidation after lowering the product aqueous media pH to below about 9. A cyclic process for removing sulfur oxides from gas mixtures includes treating these mixtures includes treating these mixtures with aqueous media including alkali metal carbonate and alkali metal bicarbonate where the ratio of alkali metal to sulfur dioxide is not less than 2. The sulfur values may be recovered from the resulting carbonate/bicarbonate/-sulfite containing product aqueous media as alkali metal sulfate or sulfite salts which are removed by crystallization from the carbonate-containing product aqueous media. As with the nitrates, the carbonate/bicarbonate system strongly increases yield of sulfate or sulfite during crystallization. Where the gas mixtures include both sulfur dioxide and lower valence nitrogen oxides, the processes for removing lower valence nitrogen oxides and sulfur dioxide may be combined into a single removal/recovery system, or may be effected in sequence.

  2. Plutonium oxalate precipitation for trace elemental determination in plutonium materials

    DOE PAGES

    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.

  3. 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.

  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. 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.

  6. 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

  7. Calibration of a neutron coincidence counter for measurement of the plutonium content of wet oxalate cakes

    SciTech Connect

    Crane, T.W.

    1983-04-01

    A novel calibration procedure has been applied to the dual-ring thermal neutron coincidence counter used at the Los Alamos Plutonium Processing Facility to measure the plutonium content of wet oxalate cakes. The calibration uses the measured content of the plutonium oxide product and the coincidence counter response to estimate the plutonium content in up to four wet plutonium oxalate cakes. These estimated mass values are then used to calibrate the counter. The calibration calculation is iterative in determining the calibration function coefficients and plutonium masses for oxalate material.

  8. CSER 96-023: CSER for PFP glovebox HC-21A with 4.4 kilogram plutonium cans

    SciTech Connect

    Wittekind, W.D., Westinghouse Hanford

    1996-12-17

    This criticality safety evaluation report addresses the criticality impact of increasing plutonium oxide content from 2.5 kg oxide storage cans to 5.0 kg oxide Pu storage cans. Glovebox HC-21A is used to move plutonium metal buttons from cans into furnace boats prior to transferring them to the muffle furnace gloveboxes. Glovebox HC-21A supports muffle furnace operations where plutonium buttons are burned to form paw, (H/Pu < 2). The paw, is returned to glovebox HC-21A and sieved and packed into the 4.,f kg Pu cans. The plutonium mass limit is set at 7.5 kg plutonium when plutonium metal is present. The plutonium mass limit is set at 15. kg plutonium when no plutonium metal is present. Additionally, there are other requirements to assure criticality safety during this operation.

  9. 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

  10. 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.

  11. 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

  12. 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.

  13. Remote Oxidation of Aliphatic C-H Bonds in Nitrogen-Containing Molecules.

    PubMed

    Howell, Jennifer M; Feng, Kaibo; Clark, Joseph R; Trzepkowski, Louis J; White, M Christina

    2015-11-25

    Nitrogen heterocycles are ubiquitous in natural products and pharmaceuticals. Herein, we disclose a nitrogen complexation strategy that employs a strong Brønsted acid (HBF4) or an azaphilic Lewis acid (BF3) to enable remote, non-directed C(sp(3))-H oxidations of tertiary, secondary, and primary amine- and pyridine-containing molecules with tunable iron catalysts. Imides resist oxidation and promote remote functionalization. PMID:26536374

  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. 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.

  17. Acidic properties of oxides containing niobia on silica and niobia in silica

    SciTech Connect

    Burke, P.A.; Ko E.I. )

    1991-05-01

    A series of composite oxides containing niobia (Nb{sub 2}O{sub 5}) and silica (SiO{sub 2}) was prepared by either incipient wetness impregnation to form surface phase oxides or by coprecipitation to form mixed oxides. At low Nb{sub 2}O{sub 5} concentrations, both the surface phase and mixed oxides showed strong Lewis acidity as determined by thermogravimetric and infrared studies of adsorbed pyridine. A surface phase oxide containing 0.25 monolayer of Nb{sub 2}O{sub 5} on SiO{sub 2} further showed strong Broensted acidity. With increasing Nb{sub 2}O{sub 5} concentration and/or treatment temperature, Nb{sub 2}O{sub 5} in these composite oxides became more bulk-like and showed a corresponding decrease in acid strength. These acidic properties are discussed in terms of proposed structural models which contain tetrahedral niobia, octahedral niobia, and Nb=O bonds as basic building blocks.

  18. Transformation and destabilization of graphene oxide in reducing aqueous solutions containing sulfide.

    PubMed

    Fu, Heyun; Qu, Xiaolei; Chen, Wei; Zhu, Dongqiang

    2014-12-01

    The colloidal stability of carbon nanomaterials is a key factor controlling their fate and bioavailability in natural aquatic systems. The authors report that graphene oxide nanoparticles could be destabilized in reducing aqueous solutions containing a low concentration (0.5 mM) of sulfide, a naturally occurring reductant. Spectroscopic characterization using combined X-ray photoelectron, Fourier-transform infrared, X-ray diffraction, and Raman analyses revealed that the surface oxygen-containing groups (mainly epoxy groups) of graphene oxide were significantly reduced after reacting with sodium sulfide. The destabilization of graphene oxide was likely caused by the enhanced surface hydrophobicity of the reduced graphene oxide, whereas electrostatic repulsion played a minimal role. Solution pH was found to affect both the deoxygenation process and the aggregation behavior of graphene oxide. Coexisting humic acid reduced the reaction efficiency and stabilized graphene oxide through steric hindrance. These findings suggest for the first time that the colloidal behavior of carbon nanomaterials might change drastically when they enter natural reducing environments containing sulfide such as anaerobic aquifers and sediments.

  19. 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.

  20. Absorption of inorganic halides produced from Freon 12 by calcium carbonate containing iron(III) oxide

    SciTech Connect

    Imamura, Seiichiro; Matsuba, Yoichi; Yamada, Etsu; Takai, Kenji; Utani, Kazunori

    1997-09-01

    Inorganic halides produced by the catalytic decomposition of Freon 12 were fixed by calcium carbonate, which is the main component of limestone. Iron(III) oxide, which is present as a contaminant in limestone, promoted the absorption of the halides by calcium carbonate at low temperatures. The supposed action of iron(III) oxide was to first react with inorganic halides, forming iron halides, and, then, transfer them to calcium carbonate to replace carbonate ion in a catalytic way. Thus, calcium carbonate containing iron oxides (limestone) can be used as an effective absorbent for the inorganic halogens produced during the decomposition of Freons.

  1. 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.

  2. 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.

  3. 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.

  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. 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

  6. 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.

  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. 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.

  9. 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

  10. Validation efforts for the neutronics of a plutonium erbium zirconium oxide inert matrix light water reactor fuel

    NASA Astrophysics Data System (ADS)

    Paratte, J. M.; Chawla, R.; Früh, R.; Joneja, O. P.; Pelloni, S.; Pralong, C.

    1999-08-01

    Light water reactor (LWR) neutronics codes and cross-section libraries need further qualification when used for the calculation of inert matrix fuel (IMF) cells. Three types of validation efforts have been undertaken for the PuO 2-Er 2O 3-ZrO 2 IMF concept under development at the Paul Scherrer Institute (PSI). Firstly, the PSI calculational scheme, based on the BOXER code and its data library, has been applied to the analysis of a range of LWR experiments with PuO 2-UO 2 fuel, conducted earlier at PSI's PROTEUS facility. The generally good agreement obtained between calculated and measured parameters gives confidence in the ability of the employed calculational scheme to correctly modelize Pu-containing fuel cells. Secondly, reactivity effects of various burnable poisons in a ZrO 2 matrix were measured in the CROCUS reactor of the Swiss Federal Institute of Technology at Lausanne. Modelling these experiments with BOXER resulted in satisfactory prediction of measured reactivity ratios (relative to a soluble-boron standard) for most of the experimental rods employed. This was particularly the case for experiments with erbium, as well as with mixtures of erbium and europium (the latter being used to simulate the effects of overlapping resonances, as would be expected in the case of a Pu-Er IMF). Finally, as there are no experimental results available from power reactors employing IMFs, the validation of burnup calculations (at the cell level) has been based on results obtained in the framework of an international benchmark exercise on the physics of LWRs employing IMFs. Certain discrepancies in calculated parameters have been observed in this context, several of which can be attributed to specific differences in cross-section libraries.

  11. Study of the formation, prevention, and recovery of plutonium from plutonium esters in the Purex process

    SciTech Connect

    Gray, L. W.; Burney, G. A.

    1981-01-01

    The Savannah River Plant uses the basic Purex process to separate /sup 239/Pu from /sup 238/U and fission products. Dark-brown, dense solids containing up to 30% Pu have previously occurred in rotameters in the plutonium finishing operations. The kinetics of formation of this mixture of DBP- and MBP-Pu esters suggest two methods to prevent the formation of the solids. A selective dissolution method using NaOH metathesis has been developed to separate the phosphate ester from the plutonium before dissolution of the residual plutonium hydroxide in a HNO/sub 3/-HF medium.

  12. 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

  13. 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.

  14. 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.

    2015-03-15

    This work reports 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 Y{sub 2}O{sub 3}, fluorite Y{sub 2}O{sub 3}–HfO{sub 2} solid solution and pyrochlore (or fluorite) Y{sub 2}(Ti,Hf){sub 2−x}O{sub 7−x}. High resolution transmission electron microscopy was used to characterize the particle–matrix interfaces. 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. The 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. - Highlights: • The oxide nanoparticles in a hafnium-containing austenitic ODS were characterized. • The nanoparticles are Y–Hf–Ti–O enriched phases according to APT and STEM–EDS. • Two coherency and an axis-parallel orientation relationships were found by HR-TEM. • Particle size has a prominent effect on the orientation relationship (OR). • Formation mechanism of the oxide nanoparticles was discussed based on the ORs.

  15. 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.

  16. 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

  17. 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

  18. Femtosecond third-order nonlinear spectra of lead-germanium oxide glasses containing silver nanoparticles.

    PubMed

    De Boni, Leonardo; Barbano, Emerson C; de Assumpção, Thiago A; Misoguti, Lino; Kassab, Luciana R P; Zilio, Sergio C

    2012-03-12

    This work reports on the spectral dependence of both nonlinear refraction and absorption in lead-germanium oxide glasses (PbO-GeO₂) containing silver nanoparticles. We have found that this material is suitable for all-optical switching at telecom wavelengths but at the visible range it behaves either as a saturable absorber or as an optical limiter.

  19. Plutonium disposition via immobilization in ceramic or glass

    SciTech Connect

    Gray, L.W.; Kan, T.; Shaw, H.F.; Armantrout, A.

    1997-03-05

    The management of surplus weapons plutonium is an important and urgent task with profound environmental, national, and international security implications. In the aftermath of the Cold War, Presidential Policy Directive 13, and various analyses by renown scientific, technical, and international policy organizations have brought about a focused effort within the Department of Energy to identify and implement paths for the long term disposition of surplus weapons- usable plutonium. The central goal of this effort is to render surplus weapons plutonium as inaccessible and unattractive for reuse in nuclear weapons as the much larger and growing stock of plutonium contained in spent fuel from civilian reactors. One disposition option being considered for surplus plutonium is immobilization, in which the plutonium would be incorporated into a glass or ceramic material that would ultimately be entombed permanently in a geologic repository for high-level waste.

  20. Nonlinear optical properties of sol-gel-type glasses containing oxidized Cu2Se nanoparticles

    NASA Astrophysics Data System (ADS)

    Yumashev, K. V.

    2000-03-01

    Nonlinear absorption of sol-gel-type glasses containing oxidized Cu2Se nanoparticles (25 nm in radius) is investigated with the aid of the picosecond excitation-probing technique. The oxidation of Cu2Se nanoparticles leads to the appearance of a wide absorption band, which has a maximum roughly at 1 µm and which is bleached on exposure to laser pulses. The time of relaxation of the bleached state approaches 300 psec, and the cross section for absorption at 1.06 µm is equal to about 1.3·10-16 cm2, as evaluated from the measured absorption saturation curve. The regime of Q-switching is obtained for a number of solid-state lasers with generation wavelengths 1.06, 1.34, and 1.54 µm, when sol-gel-type glasses containing oxidized Cu2Se nanoparticles were used as passive shutters.

  1. Parahalogenated phenols accelerate the photochemical release of nitrogen oxides from frozen solutions containing nitrate.

    PubMed

    Abida, Otman; Osthoff, Hans D

    2012-06-21

    The photolysis of nitrate anion (NO(3)(-)) contained in surface ice and snow can be a regionally significant source of gas-phase nitrogen oxides and affect the composition of the planetary boundary layer. In this study, the photochemical release of nitrogen oxides from frozen solutions containing NO(3)(-) in the presence of organic compounds was investigated. Gas-phase nitrogen oxides were quantified primarily by NO-O(3) chemiluminescence detection of NO and NO(y) (=NO + NO(2) + HONO + HNO(3) + ∑PAN + ∑AN ...) and cavity ring-down spectroscopy of NO(2) and total alkyl nitrates (∑AN). The photochemical production of gas-phase NO(y) was suppressed by the presence of formate, methanesulfonate, toluene, or phenol. In contrast, para-halogenated phenols (in the order of Cl > Br > F) promoted the conversion of NO(3)(-) to gas-phase NO(y), rationalized by acidification of the ice surface.

  2. Synthesis and catalytic water oxidation activities of ruthenium complexes containing neutral ligands.

    PubMed

    Xu, Yunhua; Duan, Lele; Åkermark, Torbjörn; Tong, Lianpeng; Lee, Bao-Lin; Zhang, Rong; Åkermark, Björn; Sun, Licheng

    2011-08-16

    Two dinuclear and one mononuclear ruthenium complexes containing neutral polypyridyl ligands have been synthesised as pre-water oxidation catalysts and characterised by (1)H and (13)C NMR spectroscopy and ESI-MS. Their catalytic water oxidation properties in the presence of [Ce(NH(4))(2)(NO(3))(6)] (Ce(IV)) as oxidant at pH 1.0 have been investigated. At low concentrations of Ce(IV) (5 mM), high turnover numbers of up to 4500 have been achieved. An (18)O-labelling experiment established that both O atoms in the evolved O(2) originate from water. Combined electrochemical study and electrospray ionisation mass spectrometric analysis suggest that ligand exchange between coordinated 4-picoline and free water produces Ru aquo species as the real water oxidation catalysts.

  3. Excess Weapons Plutonium Disposition: Plutonium Packaging, Storage and Transportation and Waste Treatment, Storage and Disposal Activities

    SciTech Connect

    Jardine, L J; Borisov, G B

    2004-07-21

    A fifth annual Excess Weapons Plutonium Disposition meeting organized by Lawrence Livermore National Laboratory (LLNL) was held February 16-18, 2004, at the State Education Center (SEC), 4 Aerodromnya Drive, St. Petersburg, Russia. The meeting discussed Excess Weapons Plutonium Disposition topics for which LLNL has the US Technical Lead Organization responsibilities. The technical areas discussed included Radioactive Waste Treatment, Storage, and Disposal, Plutonium Oxide and Plutonium Metal Packaging, Storage and Transportation and Spent Fuel Packaging, Storage and Transportation. The meeting was conducted with a conference format using technical presentations of papers with simultaneous translation into English and Russian. There were 46 Russian attendees from 14 different Russian organizations and six non-Russian attendees, four from the US and two from France. Forty technical presentations were made. The meeting agenda is given in Appendix B and the attendance list is in Appendix C.

  4. PLUTONIUM-CERIUM ALLOY

    DOEpatents

    Coffinberry, A.S.

    1959-01-01

    An alloy is presented for use as a reactor fuel. The binary alloy consists essentially of from about 5 to 90 atomic per cent cerium and the balance being plutonium. A complete phase diagram for the cerium--plutonium system is given.

  5. DELTA PHASE PLUTONIUM ALLOYS

    DOEpatents

    Cramer, E.M.; Ellinger, F.H.; Land. C.C.

    1960-03-22

    Delta-phase plutonium alloys were developed suitable for use as reactor fuels. The alloys consist of from 1 to 4 at.% zinc and the balance plutonium. The alloys have good neutronic, corrosion, and fabrication characteristics snd possess good dimensional characteristics throughout an operating temperature range from 300 to 490 deg C.

  6. 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.

  7. 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.

  8. Magnetic separation as a plutonium residue enrichment process

    SciTech Connect

    Avens, L.R.; McFarlan, J.T.; Gallegos, U.F.

    1989-01-01

    We have subjected several plutonium contaminated residues to Open Gradient Magnetic Separation (OGMS) on an experimental scale. Separation of graphite, bomb reduction sand, and bomb reduction sand, and bomb reduction sand, slag, and crucible, resulted in a plutonium rich fraction and a plutonium lean fraction. The lean fraction varied between about 20% to 85% of the feed bulk. The plutonium content of the lean fraction can be reduced from about 2% in the feed to the 0.1% to 0.5% range dependent on the portion of the feed rejected to this lean fraction. These values are low enough in plutonium to meet economic discard limits and be considered for direct discard. Magnetic separation of direct oxide reduction and electrorefining pyrochemical salts gave less favorable results. While a fraction very rich in plutonium could be obtained, the plutonium content of the lean fraction was to high for direct discard. This may still have chemical processing applications. OGMS experiments at low magnetic field strength on incinerator ash did give two fractions but the plutonium content of each fraction was essentially identical. Thus, no chemical processing advantage was identified for magnetic separation of this residue. The detailed results of these experiments and the implications for OGMS use in recycle plutonium processing are discussed. 4 refs., 3 figs., 9 tabs.

  9. SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS BY ADSORPTION

    DOEpatents

    Seaborg, G.T.; Willard, J.E.

    1958-01-01

    A method is presented for the separation of plutonium from solutions containing that element in a valence state not higher than 41 together with uranium ions and fission products. This separation is accomplished by contacting the solutions with diatomaceous earth which preferentially adsorbs the plutonium present. Also mentioned as effective for this adsorbtive separation are silica gel, filler's earth and alumina.

  10. Removal of plutonium and americium from alkaline waste solutions

    DOEpatents

    Schulz, Wallace W.

    1979-01-01

    High salt content, alkaline waste solutions containing plutonium and americium are contacted with a sodium titanate compound to effect removal of the plutonium and americium from the alkaline waste solution onto the sodium titanate and provide an effluent having a radiation level of less than 10 nCi per gram alpha emitters.

  11. 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.

  12. Plutonium bioaccumulation in seabirds.

    PubMed

    Strumińska-Parulska, Dagmara I; Skwarzec, Bogdan; Fabisiak, Jacek

    2011-12-01

    The aim of the paper was plutonium (²³⁸Pu and ²³⁹⁺²⁴⁰Pu) determination in seabirds, permanently or temporarily living in northern Poland at the Baltic Sea coast. Together 11 marine birds species were examined: 3 species permanently residing in the southern Baltic, 4 species of wintering birds and 3 species of migrating birds. The obtained results indicated plutonium is non-uniformly distributed in organs and tissues of analyzed seabirds. The highest plutonium content was found in the digestion organs and feathers, the smallest in skin and muscles. The plutonium concentration was lower in analyzed species which feed on fish and much higher in herbivorous species. The main source of plutonium in analyzed marine birds was global atmospheric fallout.

  13. 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

  14. 31. VIEW OF A WORKER HOLDING A PLUTONIUM 'BUTTON.' PLUTONIUM, ...

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

    31. VIEW OF A WORKER HOLDING A PLUTONIUM 'BUTTON.' PLUTONIUM, A MAN-MADE SUBSTANCE, WAS RARE. SCRAPS RESULTING FROM PRODUCTION AND PLUTONIUM RECOVERED FROM RETIRED NUCLEAR WEAPONS WERE REPROCESSED INTO VALUABLE PURE-PLUTONIUM METAL (9/19/73). - Rocky Flats Plant, Bounded by Indiana Street & Routes 93, 128 & 72, Golden, Jefferson County, CO

  15. Recommended plutonium release fractions from postulated fires. Final report

    SciTech Connect

    Kogan, V.; Schumacher, P.M.

    1993-12-01

    This report was written at the request of EG&G Rocky Flats, Inc. in support of joint emergency planning for the Rocky Flats Plant (RFP) by EG&G and the State of Colorado. The intent of the report is to provide the State of Colorado with an independent assessment of any respirable plutonium releases that might occur in the event of a severe fire at the plant. Fire releases of plutonium are of interest because they have been used by EG&G to determine the RFP emergency planning zones. These zones are based on the maximum credible accident (MCA) described in the RFP Final Environmental Impact Statement (FEIS) of 1980, that MCA is assumed to be a large airplane crashing into a RFP plutonium building.The objective of this report was first, to perform a worldwide literature review of relevant release experiments from 1960 to the present and to summarize those findings, and second, to provide recommendations for application of the experimental data to fire release analyses at Rocky Flats. The latter step requires translation between experimental and expected RFP accident parameters, or ``scaling.`` The parameters of particular concern are: quantities of material, environmental parameters such as the intensity of a fire, and the physico-chemical forms of the plutonium. The latter include plutonium metal, bulk plutonium oxide powder, combustible and noncombustible wastes contaminated with plutonium oxide powder, and residues from plutonium extraction processes.

  16. PLUTONIUM-CERIUM-COBALT AND PLUTONIUM-CERIUM-NICKEL ALLOYS

    DOEpatents

    Coffinberry, A.S.

    1959-08-25

    >New plutonium-base teroary alloys useful as liquid reactor fuels are described. The alloys consist of 10 to 20 atomic percent cobalt with the remainder plutonium and cerium in any desired proportion, with the plutonium not in excess of 88 atomic percent; or, of from 10 to 25 atomic percent nickel (or mixture of nickel and cobalt) with the remainder plutonium and cerium in any desired proportion, with the plutonium not in excess of 86 atomic percent. The stated advantages of these alloys over unalloyed plutonium for reactor fuel use are a lower melting point and a wide range of permissible plutonium dilution.

  17. 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

  18. 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.

  19. 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.

  20. Sonochemical Digestion of High-Fired Plutonium Dioxide Samples

    SciTech Connect

    Sinkov, Sergei I.; Lumetta, Gregg J.

    2006-10-12

    This work was performed as part of a broader effort to automate analytical methods for determining plutonium and other radioisotopes in environmental samples. The work described here represented a screening study to evaluate the effect of applying ultrasonic irradiation to dissolve high-fired plutonium oxide. The major findings of this work can be summarized as follows: (1) High-fired plutonium oxide does not undergo measurable dissolution when sonicated in nitric acid solutions, even at a high concentration range of nitric acid where the calculated thermodynamic solubility of plutonium oxide exceeds the ?g/mL level. (2) Applying organic complexants (nitrilotriacetic acid) and reductants (hydroxyurea) in 1.5 M nitric acid does not significantly increase the dissolution compared with digestion in nitric acid alone. Nearly all (99.5%) of the plutonium oxide remains undissolved under these conditions. (3) The action of a strong inorganic reductant, titanium trichloride in 25 wt% HCl, results in 40% dissolution of the plutonium oxide when the titanium trichloride concentration is ?1 wt% under sonication. (4) Oxidative treatment of plutonium oxide by freshly dissolved AgO ({approx}20 mg/mL) in 1.5 M nitric acid with sonication resulted in 95% plutonium oxide dissolution. However, the same treatment of plutonium oxide mechanically mixed with 50 mg of Columbia River sediment (CRS) results in a significant decrease of dissolution yield of plutonium oxide (<20% dissolved at the same AgO loading) because of parasitic consumption of AG(II) by oxidizable components of the CRS. (5) Digesting plutonium oxide in HF resulted in dissolution yields slightly higher than 80% for HF concentration from 6 M to 14 M. Sonication did not result in any improvement in dissolution efficiency in HF. (6) Mixed nitric acid/HF solutions result in a higher dissolution yield of plutonium oxide compared with digestion in HF alone (at the same HF concentrations). Practically quantitative dissolution

  1. Dehydration of plutonium or neptunium trichloride hydrate

    DOEpatents

    Foropoulos, Jr., Jerry; Avens, Larry R.; Trujillo, Eddie A.

    1992-01-01

    A process of preparing anhydrous actinide metal trichlorides of plutonium or neptunium by reacting an aqueous solution of an actinide metal trichloride selected from the group consisting of plutonium trichloride or neptunium trichloride with a reducing agent capable of converting the actinide metal from an oxidation state of +4 to +3 in a resultant solution, evaporating essentially all the solvent from the resultant solution to yield an actinide trichloride hydrate material, dehydrating the actinide trichloride hydrate material by heating the material in admixture with excess thionyl chloride, and recovering anhydrous actinide trichloride is provided.

  2. Dehydration of plutonium or neptunium trichloride hydrate

    DOEpatents

    Foropoulos, J. Jr.; Avens, L.R.; Trujillo, E.A.

    1992-03-24

    A process is described for preparing anhydrous actinide metal trichlorides of plutonium or neptunium by reacting an aqueous solution of an actinide metal trichloride selected from the group consisting of plutonium trichloride or neptunium trichloride with a reducing agent capable of converting the actinide metal from an oxidation state of +4 to +3 in a resultant solution, evaporating essentially all the solvent from the resultant solution to yield an actinide trichloride hydrate material, dehydrating the actinide trichloride hydrate material by heating the material in admixture with excess thionyl chloride, and recovering anhydrous actinide trichloride.

  3. PROCESS USING BISMUTH PHOSPHATE AS A CARRIER PRECIPITATE FOR FISSION PRODUCTS AND PLUTONIUM VALUES

    DOEpatents

    Finzel, T.G.

    1959-03-10

    A process is described for separating plutonium from fission products carried therewith when plutonium in the reduced oxidation state is removed from a nitric acid solution of irradiated uranium by means of bismuth phosphate as a carrier precipitate. The bismuth phosphate carrier precipitate is dissolved by treatment with nitric acid and the plutonium therein is oxidized to the hexavalent oxidation state by means of potassium dichromate. Separation of the plutonium from the fission products is accomplished by again precipitating bismuth phosphate and removing the precipitate which now carries the fission products and a small percentage of the plutonium present. The amount of plutonium carried in this last step may be minimized by addition of sodium fluoride, so as to make the solution 0.03N in NaF, prior to the oxidation and prccipitation step.

  4. 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.

  5. Low oxidation state aluminum-containing cluster anions: Cp∗AlnH-, n = 1-3

    NASA Astrophysics Data System (ADS)

    Zhang, Xinxing; Ganteför, Gerd; Eichhorn, Bryan; Mayo, Dennis; Sawyer, William H.; Gill, Ann F.; Kandalam, Anil K.; Schnöckel, Hansgeorg; Bowen, Kit

    2016-08-01

    Three new, low oxidation state, aluminum-containing cluster anions, Cp*AlnH-, n = 1-3, were prepared via reactions between aluminum hydride cluster anions, AlnHm-, and Cp*H ligands. These were characterized by mass spectrometry, anion photoelectron spectroscopy, and density functional theory based calculations. Agreement between the experimentally and theoretically determined vertical detachment energies and adiabatic detachment energies validated the computed geometrical structures. Reactions between aluminum hydride cluster anions and ligands provide a new avenue for discovering low oxidation state, ligated aluminum clusters.

  6. 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.

  7. 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.

  8. The interfacial orientation relationship of oxide nanoparticles in a hafnium-containing oxide dispersion-strengthened austenitic stainless steel

    DOE PAGES

    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

  9. Monomers, Dimers, and Helices: Complexities of Cerium and Plutonium Phenanthrolinecarboxylates.

    PubMed

    Cary, Samantha K; Ferrier, Maryline G; Baumbach, Ryan E; Silver, Mark A; Lezama Pacheco, Juan; Kozimor, Stosh A; La Pierre, Henry S; Stein, Benjamin W; Arico, Alexandra A; Gray, Danielle L; Albrecht-Schmitt, Thomas E

    2016-05-01

    The reaction of Ce(III) or Pu(III) with 1,10-phenanthroline-2,9-dicarboxylic acid (PDAH2) results in the formation of new f-element coordination complexes. In the case of cerium, Ce(PDA)(H2O)2Cl·H2O (1) or [Ce(PDAH)(PDA)]2[Ce(PDAH)(PDA)] (2) was isolated depending on the Ce/ligand ratio in the reaction. The structure of 2 is composed of two distinct substructures that are constructed from the same monomer. This monomer is composed of a Ce(III) cation bound by one PDA(2-) dianionic ligand and one PDAH(-) monoanionic ligand, both of which are tetradentate. Bridging by the carboxylate moieties leads to either [Ce(PDAH)(PDA)]2 dimers or [Ce(PDAH)(PDA)]1∞ helical chains. For plutonium, Pu(PDA)2 (3) was the only product isolated regardless of the Pu/ligand ratio employed in the reaction. During the reaction of plutonium with PDAH2, Pu(III) is oxidized to Pu(IV), generating 3. This assignment is consistent with structural metrics and the optical absorption spectrum. Ambiguity in the assignment of the oxidation state of cerium in 1 and 2 from UV-vis-near-IR spectra invoked the use of Ce L3,2-edge X-ray absorption near-edge spectroscopy, magnetic susceptibility, and heat capacity measurements. These experiments support the assignment of Ce(III) in both compounds. The bond distances and coordination numbers are also consistent with these assignments. 3 contains 8-coordinate Pu(IV), whereas the cerium centers in 1 and 2 are 9- and/or 10-coordinate, which correlates with the increased size of Ce(III) versus Pu(IV). Taken together, these data provide an example of a system where the differences in the redox behavior between these f elements creates more complex chemistry with cerium than with plutonium. PMID:27070401

  10. Modeling of diffusion of plutonium in other metals and of gaseous species in plutonium-based systems. 1998 annual progress report

    SciTech Connect

    Cooper, B.R.; Fernando, G.W.

    1998-06-01

    'The problem being addressed is to establish standards for storage conditions for containers (e.g. stainless steel containers) meant to hold nuclear waste for long periods of time, e.g. fifty years, such as to ensure the integrity of the containers without being excessively conservative, i.e., being unnecessarily costly. Allowable storage temperature is an important parameter for the standards. Diffusion processes are central to certain processes that initiate corrosion of the containers or that lead to the release of gases increasing pressure that can rupture the container. The scientific goals of this project are: (1) to predict diffusion constants on an ab initio basis, i.e. diffusion distances in specified time at specified temperature, for atomic species from plutonium-based waste materials into storage container materials such as stainless steel or vice versa, (2) to predict the time development at specified temperature of complex diffusion processes in plutonium-based oxide type waste materials in the presence of water vapor such that gases may be released within a container. This predictive ability will help to provide information relevant to setting standards for waste container vessels and storage conditions. This project was initiated on September 15, 1997, so that this summary reports progress after about 2/3 of the first year of a three-year project.'

  11. 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.

  12. 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.

  13. Technical considerations and policy requirements for plutonium management

    SciTech Connect

    Christensen, D.C.; Dinehart, S.M.; Yarbro, S.L.

    1995-12-31

    The goals for plutonium management have changed dramatically over the past few years. Today, the challenge is focused on isolating plutonium from the environment and preparing it for permanent disposition. In parallel, the requirements for managing plutonium are rapidly changing. For example, there is a significant increase in public awareness on how facilities operate, increased attention to environmental safety and health (ES and H) concerns, greater interest in minimizing waste, more emphasis on protecting material from theft, providing materials for international inspection, and a resurgence of interest in using plutonium as an energy source. Of highest concern, in the immediate future, is protecting plutonium from theft or diversion, while the national policy on disposition is debated. These expanded requirements are causing a broadening of responsibilities within the Department of Energy (DOE) to include at least seven organizations. An unavoidable consequence is the divergence in approach and short-term goals for managing similar materials within each organization. The technology base does exist, properly, safely, and cost effectively to extract plutonium from excess weapons, residues, waste, and contaminated equipment and facilities, and to properly stabilize it. Extracting the plutonium enables it to be easily inventoried, packaged, and managed to minimize the risk of theft and diversion. Discarding excess plutonium does not sufficiently reduce the risk of diversion, and as a result, long-term containment of plutonium from the environment may not be able to be proven to the satisfaction of the public.

  14. PROCESS FOR SEPARATING PLUTONIUM BY REPEATED PRECIPITATION WITH AMPHOTERIC HYDROXIDE CARRIERS

    DOEpatents

    Faris, B.F.

    1960-04-01

    A multiple carrier precipitation method is described for separating and recovering plutonium from an aqueous solution. The hydroxide of an amphoteric metal is precipitated in an aqueous plutonium-containing solution. This precipitate, which carries plutonium, is then separated from the supernatant liquid and dissolved in an aqueous hydroxide solution, forming a second plutonium- containing solution. lons of an amphoteric metal which forms an insoluble hydroxide under the conditions existing in this second solution are added to the second solution. The precipitate which forms and which carries plutonium is separated from the supernatant liquid. Amphoteric metals which may be employed are aluminum, bibmuth, copper, cobalt, iron, lanthanum, nickel, and zirconium.

  15. ESTIMATING IMPURITIES IN SURPLUS PLUTONIUM FOR DISPOSITION

    SciTech Connect

    Allender, J.; Moore, E.

    2013-07-17

    The United States holds at least 61.5 metric tons (MT) of plutonium that is permanently excess to use in nuclear weapons programs, including 47.2 MT of weapons-grade plutonium. Surplus inventories will be stored safely by the Department of Energy (DOE) and then transferred to facilities that will prepare the plutonium for permanent disposition. The Savannah River National Laboratory (SRNL) operates a Feed Characterization program for the Office of Fissile Materials Disposition of the National Nuclear Security Administration and the DOE Office of Environmental Management. Many of the items that require disposition are only partially characterized, and SRNL uses a variety of techniques to predict the isotopic and chemical properties that are important for processing through the Mixed Oxide Fuel Fabrication Facility and alternative disposition paths. Recent advances in laboratory tools, including Prompt Gamma Analysis and Peroxide Fusion treatment, provide data on the existing inventories that will enable disposition without additional, costly sampling and destructive analysis.

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

    SciTech Connect

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

    2007-07-01

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

  17. 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].

  18. Progress on plutonium stabilization

    SciTech Connect

    Hurt, D.

    1996-05-01

    The Defense Nuclear Facilities Safety Board has safety oversight responsibility for most of the facilities where unstable forms of plutonium are being processed and packaged for interim storage. The Board has issued recommendations on plutonium stabilization and has has a considerable influence on DOE`s stabilization schedules and priorities. The Board has not made any recommendations on long-term plutonium disposition, although it may get more involved in the future if DOE develops plans to use defense nuclear facilities for disposition activities.

  19. 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.

  20. Suppressive ability of defatted rice bran against lipid oxidation in cookies containing iron.

    PubMed

    Watanabe, Yoshiyuki; Hashimoto, Kazue; Omori, Ayako; Uda, Yasutaka; Nomura, Masato

    2010-01-01

    Cookies containing iron, defatted rice bran, and several oils were prepared, and their oxidative stability evaluated. Oxidation was suppressed by the defatted rice bran, but a limit to the suppressive effect was observed. The maximum peroxide values obtained with defatted rice bran were low, and similar, regardless of the degree of unsaturation of the oils. A chemical analysis suggested that proteins and polyphenols in the defatted rice bran contributed to the suppressive effect. Cookies without the defatted rice bran showed decreasing maximum peroxide value as the relative humidity increased. No such dependence was observed for the cookies containing defatted rice bran. The water sorption isotherm of defatted rice bran indicated that the weak dependence was due to low water sorption.

  1. Long-term plutonium storage: Design concepts

    SciTech Connect

    Wilkey, D.D.; Wood, W.T.; Guenther, C.D.

    1994-08-01

    An important part of the Department of Energy (DOE) Weapons Complex Reconfiguration (WCR) Program is the development of facilities for long-term storage of plutonium. The WCR design goals are to provide storage for metals, oxides, pits, and fuel-grade plutonium, including material being held as part of the Strategic Reserve and excess material. Major activities associated with plutonium storage are sorting the plutonium inventory, material handling and storage support, shipping and receiving, and surveillance of material in storage for both safety evaluations and safeguards and security. A variety of methods for plutonium storage have been used, both within the DOE weapons complex and by external organizations. This paper discusses the advantages and disadvantages of proposed storage concepts based upon functional criteria. The concepts discussed include floor wells, vertical and horizontal sleeves, warehouse storage on vertical racks, and modular storage units. Issues/factors considered in determining a preferred design include operational efficiency, maintenance and repair, environmental impact, radiation and criticality safety, safeguards and security, heat removal, waste minimization, international inspection requirements, and construction and operational costs.

  2. Plutonium Chemistry in the UREX+ Separation Processes

    SciTech Connect

    ALena Paulenova; George F. Vandegrift, III; Kenneth R. Czerwinski

    2009-10-01

    The project "Plutonium Chemistry in the UREX+ Separation Processes” is led by Dr. Alena Paulenova of Oregon State University under collaboration with Dr. George Vandegrift of ANL and Dr. Ken Czerwinski of the University of Nevada at Las Vegas. The objective of the project is to examine the chemical speciation of plutonium in UREX+ (uranium/tributylphosphate) extraction processes for advanced fuel technology. Researchers will analyze the change in speciation using existing thermodynamics and kinetic computer codes to examine the speciation of plutonium in aqueous and organic phases. They will examine the different oxidation states of plutonium to find the relative distribution between the aqueous and organic phases under various conditions such as different concentrations of nitric acid, total nitrates, or actinide ions. They will also utilize techniques such as X-ray absorbance spectroscopy and small-angle neutron scattering for determining plutonium and uranium speciation in all separation stages. The project started in April 2005 and is scheduled for completion in March 2008.

  3. Plutonium immobilization ceramic feed batching component test report

    SciTech Connect

    Erickson, S.A.

    1999-10-04

    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. Ceramic feed batching (CFB) is one of the first process steps involved with first stage plutonium immobilization. The CFB step will blend plutonium oxide powder before it is combined with other materials to make pucks. This report discusses the Plutonium Immobilization CFB process preliminary concept (including a process block diagram), batch splitting component test results, CFB development areas, and FY 1999 and 2000 CFB program milestones.

  4. Disposing of the world`s excess plutonium

    SciTech Connect

    McCormick, J.M.; Bullen, D.B.

    1998-12-31

    The authors undertake three key objectives in addressing the issue of plutonium disposition at the end of the Cold War. First, the authors estimate the total global inventory of plutonium both from weapons dismantlement and civil nuclear power reactors. Second, they review past and current policy toward handling this metal by the US, Russia, and other key countries. Third, they evaluate the feasibility of several options (but especially the vitrification and mixed oxide fuel options announced by the Clinton administration) for disposing of the increasing amounts of plutonium available today. To undertake this analysis, the authors consider both the political and scientific problems confronting policymakers in dealing with this global plutonium issue. Interview data with political and technical officials in Washington and at the International Atomic Energy Agency in Vienna, Austria, and empirical inventory data on plutonium from a variety of sources form the basis of their analysis.

  5. Pyrochemical recovery of plutonium from calcium fluoride reduction slag

    DOEpatents

    Christensen, D.C.

    A pyrochemical method of recovering finely dispersed plutonium metal from calcium fluoride reduction slag is claimed. The plutonium-bearing slag is crushed and melted in the presence of at least an equimolar amount of calcium chloride and a few percent metallic calcium. The calcium chloride reduces the melting point and thereby decreases the viscosity of the molten mixture. The calcium reduces any oxidized plutonium in the mixture and also causes the dispersed plutonium metal to coalesce and settle out as a separate metallic phase at the bottom of the reaction vessel. Upon cooling the mixture to room temperature, the solid plutonium can be cleanly separated from the overlying solid slag, with an average recovery yield on the order of 96 percent.

  6. SEPARATION OF PLUTONIUM VALUES FROM URANIUM AND FISSION PRODUCT VALUES

    DOEpatents

    Maddock, A.G.; Booth, A.H.

    1960-09-13

    Separation of plutonium present in small amounts from neutron irradiated uranium by making use of the phenomenon of chemisorption is described. Plutonium in the tetravalent state is chemically absorbed on a fluoride in solid form. The steps for the separation comprise dissolving the irradiated uranium in nitric acid, oxidizing the plutonium in the resulting solution to the hexavalent state, adding to the solution a soluble calcium salt which by the common ion effect inhibits dissolution of the fluoride by the solution, passing the solution through a bed or column of subdivided calcium fluoride which has been sintered to about 8OO deg C to remove the chemisorbable fission products, reducing the plutonium in the solution thus obtained to the tetravalent state, and again passing the solution through a similar bed or column of calcium fluoride to selectively absorb the plutonium, which may then be recovered by treating the calcium fluoride with a solution of ammonium oxalate.

  7. Minutes of the 28th Annual Plutonium Sample Exchange Meeting. Part II: metal sample exchange

    SciTech Connect

    Not Available

    1984-01-01

    Contents of this publication include the following list of participating laboratories; agenda; attendees; minutes of October 25 and 26 meeting; and handout materials supplied by speakers. The handout materials cover the following: statistics and reporting; plutonium - chemical assay 100% minus impurities; americium neptunium, uranium, carbon and iron data; emission spectroscopy data; plutonium metal sample exchange; the calorimetry sample exchange; chlorine determination in plutonium metal using phyrohydrolysis; spectrophotometric determination of 238-plutonium in oxide; plutonium measurement capabilities at the Savannah River Plant; and robotics in radiochemical laboratory.

  8. 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...

  9. 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

  10. 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.

  11. Plutonium Decontamination of Uranium using CO2 Cleaning

    SciTech Connect

    Blau, M

    2002-12-01

    A concern of the Department of Energy (DOE) Environmental Management (EM) and Defense Programs (DP), and of the Los Alamos National Laboratory (LANL) and the Lawrence Livermore National Laboratory (LLNL), is the disposition of thousands of legacy and recently generated plutonium (Pu)-contaminated, highly enriched uranium (HEU) parts. These parts take up needed vault space. This presents a serious problem for LLNL, as site limit could result in the stoppage of future weapons work. The Office of Fissile Materials Disposition (NN-60) will also face a similar problem as thousands of HEU parts will be created with the disassembly of site-return pits for plutonium recovery when the Pit Disassembly and Conversion Facility (PDCF) at the Savannah River Site (SRS) becomes operational. To send HEU to the Oak Ridge National Laboratory and the Y-12 Plant for disposition, the contamination for metal must be less than 20 disintegrations per minute (dpm) of swipable transuranic per 100 cm{sup 2} of surface area or the Pu bulk contamination for oxide must be less than 210 parts per billion (ppb). LANL has used the electrolytic process on Pu-contaminated HEU weapon parts with some success. However, this process requires that a different fixture be used for every configuration; each fixture cost approximately $10K. Moreover, electrolytic decontamination leaches the uranium metal substrate (no uranium or plutonium oxide) from the HEU part. The leaching rate at the uranium metal grain boundaries is higher than that of the grains and depends on the thickness of the uranium oxide layer. As the leaching liquid flows past the HEU part, it carries away plutonium oxide contamination and uranium oxide. The uneven uranium metal surface created by the leaching becomes a trap for plutonium oxide contamination. In addition, other DOE sites have used CO{sub 2} cleaning for Pu decontamination successfully. In the 1990's, the Idaho National Engineering Laboratory investigated this technology and

  12. New nuclear safe plutonium ceramic compositions with neutron poisons for plutonium storage

    NASA Astrophysics Data System (ADS)

    Nadykto, B. A.; Timofeeva, L. F.

    2000-07-01

    A complex of works is conducted to study the possibility of reprocessing surplus weapon-grade plutonium to a critical-mass-free composition with neutron poison. Nuclear safe ceramic compositions of PuO2 with four most efficient neutron poisons, Hf, Gd, Li, and B, are fabricated in the laboratory. Various methods for fabrication of the compositions with PuO2 depending on neutron poison element are used and studied: a — by sintering initial component powders; b — by impregnation of a porous skeleton made of neutron poison oxide with plutonium sol-gel; c — by sintering microspheres made of plutonium oxide with neutron poison (B4C), with the microspheres having a coating completely absorbing alpha particles.

  13. Automated amperometric plutonium assay system

    SciTech Connect

    Burt, M.C.

    1985-01-01

    The amperometric titration for plutonium assay has been used in the nuclear industry for over twenty years and has been in routine use at the Hanford Engineering Development Laboratory since 1976 for the analysis of plutonium oxide and mixed oxide fuel material for the Fast Flux Test Facility. It has proven itself to be an accurate and reliable method. The method may be used as a direct end point titration or an excess of titrant may be added and a back titration performed to aid in determination of the end point. Due to the slowness of the PuVI-FeII reaction it is difficult to recognize when the end point is being approached and is very time consuming if the current is allowed to decay to the residual value after each titrant addition. For this reason the back titration in which the rapid FeII-CrVI reaction occurs is used by most laboratories. The back titration is performed by the addition of excess ferrous solution followed by two measured aliquots of standard dichromate with measurement of cell current after each addition.

  14. Expected behavior of plutonium in the IFR fuel cycle

    NASA Astrophysics Data System (ADS)

    Steunenberg, R. K.; Johnson, I.

    The Integral Fast Reactor (IFR) is a metal-fueled, sodium-cooled reactor that will consist initially of a U-Zr alloy core in which the enriched uranium will be replaced gradually by plutonium bred in a uranium blanket. The plutonium is concentrated to the required level by extraction from the molten blanket material with a CaCl2-BaCl2 salt containing MgCl2 as an oxidant (halide slagging). The CaCl2-BaCl2 salt containing dissolved PuCl3 and UCl3 is added to the core process where fission products are removed by electrorefining, using a liquid cadmium anode, a metal cathode, and a LiCl-NaCl-CaCl2-BaCl2 molten salt electrolyte. The product is recovered as a metallic deposit on the cathode. The Halide slagging step is operated at about 1250 deg and the electrorefining step at about 450 C. These processes are expected to give low fission-product decontamination factors of the order of 100.

  15. Effects of Oxide Film on the Corrosion Resistance of Titanium Grade 7 in Fluoride-Containing NaCl Brines

    SciTech Connect

    Lian, T; Whalen, M T; Wong, L

    2004-11-30

    The effects of oxide film on the corrosion behavior of Titanium Grade 7 (0.12-0.25% Pd) in fluoride-containing NaCl brines have been investigated. With the presence of a 0.6 {micro}m thick oxide layer, the annealed Ti grade 7 exhibited a significant improvement on the anodic polarization behavior. However, the oxide film did not demonstrate sustainable corrosion resistance in fluoride-containing solutions.

  16. Plutonium: Requiem or reprieve

    SciTech Connect

    Pillay, K.K.S.

    1996-01-01

    Many scientific discoveries have had profound effects on humanity and its future. However, the discovery of fissionable characteristics of a man-made element, plutonium, discovered in 1941 by Glenn Seaborg and associates, has probably had the greatest impact on world affairs. Although about 20 new elements have been synthesized since 1940, element 94 unarguably had the most dramatic impact when it was introduced to the world as the core of the nuclear bomb dropped on Nagasaki. Ever since, large quantities of this element have been produced, and it has had a major role in maintaining peace during the past 50 years. in addition, the rapid spread of nuclear power technology worldwide contributed to major growth in the production of plutonium as a by-product. This article discusses the following issues related to plutonium: plutonium from Nuclear Power Generation; environmental safety and health issues; health effects; safeguards issues; extended storage; disposal options.

  17. Plutonium solution storage in plastic bottles: Operational experience and safety issues

    SciTech Connect

    Conner, W.V.

    1995-03-15

    Computer spread sheet models were developed to gain a better understanding of the factors that lead to pressurization and failure of plastic bottles containing plutonium solutions. These models were developed using data obtained from the literature on gas generation rates for plutonium solutions. Leak rates from sealed plastic bottles were obtained from bottle leak tests conducted at Rocky Flats. Results from these bottle leak tests showed that narrow mouth four liter bottles will seal much better than wide mouth four liter bottles. The gas generation rate and leak rate data were used to develop models for predicting the rate of pressurization and maximum pressures expected in sealed bottles of plutonium solution containing various plutonium and acid concentrations. The computer models were used to develop proposed time limits for storing or transporting plutonium solutions in sealed plastic bottles. For plutonium solutions containing < 1.5 g/l, maximum safe storage times from 4 weeks to 12 months are proposed. The maximum safe storage times vary depending upon the plutonium concentration in the solution. Low concentration plutonium solutions can be stored safely for longer periods of time than high concentration plutonium solutions. For solutions containing > 1.5 g/l plutonium, storage in sealed bottles should not be allowed. However, transportation of higher concentration plutonium solution in sealed bottles is required, and safe transportation times of 1 shift to 6 days are proposed.

  18. Characterization of calcium containing plasma electrolytic oxidation coatings on AM50 magnesium alloy

    NASA Astrophysics Data System (ADS)

    Srinivasan, P. Bala; Liang, J.; Blawert, C.; Störmer, M.; Dietzel, W.

    2010-04-01

    An attempt was made to produce calcium containing plasma electrolytic oxidation (PEO) coatings on AM50 magnesium alloy using an alkaline electrolyte. This study was performed in three alkaline electrolytes containing calcium hydroxide and sodium phosphate with three different mass ratios viz., 1:2.5, 1:5 and 1:7.5. All the three coatings produced were found to contain Ca and P in appreciable amounts. The concentration of P was found to be higher in the coatings obtained in the electrolytes with higher concentration of phosphate ions. Even though all the three coatings were found to be constituted with magnesium oxide and magnesium phosphate phases, X-ray diffraction analyses revealed that the phase composition was influenced by the phosphate ion concentration/conductivity of the electrolyte. Further, the PEO coating obtained in the 1:7.5 ratio electrolyte was found to contain di-calcium phosphate (monetite) and calcium peroxide phases, which were absent in the other two coatings. Potentiodynamic polarization studies performed in 0.1 M NaCl solution showed that the coatings obtained from the 1:5 ratio electrolyte possessed a superior corrosion resistance, which is attributed to the combined effect of thickness, compactness and phase/chemical composition of this coating.

  19. Urease gene-containing Archaea dominate autotrophic ammonia oxidation in two acid soils.

    PubMed

    Lu, Lu; Jia, Zhongjun

    2013-06-01

    The metabolic traits of ammonia-oxidizing archaea (AOA) and bacteria (AOB) interacting with their environment determine the nitrogen cycle at the global scale. Ureolytic metabolism has long been proposed as a mechanism for AOB to cope with substrate paucity in acid soil, but it remains unclear whether urea hydrolysis could afford AOA greater ecological advantages. By combining DNA-based stable isotope probing (SIP) and high-throughput pyrosequencing, here we show that autotrophic ammonia oxidation in two acid soils was predominately driven by AOA that contain ureC genes encoding the alpha subunit of a putative archaeal urease. In urea-amended SIP microcosms of forest soil (pH 5.40) and tea orchard soil (pH 3.75), nitrification activity was stimulated significantly by urea fertilization when compared with water-amended soils in which nitrification resulted solely from the oxidation of ammonia generated through mineralization of soil organic nitrogen. The stimulated activity was paralleled by changes in abundance and composition of archaeal amoA genes. Time-course incubations indicated that archaeal amoA genes were increasingly labelled by (13) CO2 in both microcosms amended with water and urea. Pyrosequencing revealed that archaeal populations were labelled to a much greater extent in soils amended with urea than water. Furthermore, archaeal ureC genes were successfully amplified in the (13) C-DNA, and acetylene inhibition suggests that autotrophic growth of urease-containing AOA depended on energy generation through ammonia oxidation. The sequences of AOB were not detected, and active AOA were affiliated with the marine Group 1.1a-associated lineage. The results suggest that ureolytic N metabolism could afford AOA greater advantages for autotrophic ammonia oxidation in acid soil, but the mechanism of how urea activates AOA cells remains unclear.

  20. ANL-W MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    SciTech Connect

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1997-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement (EIS). This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. The paper describes the following: Site map and the LA facility; process descriptions; resource needs; employment requirements; wastes, emissions, and exposures; accident analysis; transportation; qualitative decontamination and decommissioning; post-irradiation examination; LA fuel bundle fabrication; LA EIS data report assumptions; and LA EIS data report supplement.