Extraction of uranium from tailings by sulfuric acid leaching with oxidants

NASA Astrophysics Data System (ADS)

Huang, Jing; Li, Mi; Zhang, Xiaowen; Huang, Chunmei; Wu, Xiaoyan

2017-06-01

Recovery of uranium have been performed by leaching uranium-containing tailings in sulfuric acid system with the assistance of HF, HClO4, H2O2 and MnO2. The effect of reagent dosage, sulfuric acid concentration, Liquid/solid ratio, reaction temperature and particle size on the leaching of uranium were investigated. The results show that addiction of HF, HClO4, H2O2 and MnO2 significantly increased the extraction of uranium under 1M sulphuric acid condition and under the optimum reaction conditions a dissolution fraction of 85% by HClO4, 90% by HF, 95% by H2O2 can be reached respectively. The variation of technological mineralogy properites of tailings during leaching process show that the assistants can break gangue effectively. These observations suggest that optimum oxidants could potentially influence the extraction of uranium from tailings even under dilute acid condition.

Capstone Depleted Uranium Aerosols: Generation and Characterization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parkhurst, MaryAnn; Szrom, Fran; Guilmette, Ray

2004-10-19

In a study designed to provide an improved scientific basis for assessing possible health effects from inhaling depleted uranium (DU) aerosols, a series of DU penetrators was fired at an Abrams tank and a Bradley fighting vehicle. A robust sampling system was designed to collect aerosols in this difficult environment and continuously monitor the sampler flow rates. Aerosols collected were analyzed for uranium concentration and particle size distribution as a function of time. They were also analyzed for uranium oxide phases, particle morphology, and dissolution in vitro. The resulting data provide input useful in human health risk assessments.

FUSED SALT PROCESS FOR RECOVERY OF VALUES FROM USED NUCLEAR REACTOR FUELS

DOEpatents

Moore, R.H.

1960-08-01

A process is given for recovering plutonium from a neutron-irradiated uranium mass (oxide or alloy) by dissolving the mass in an about equimolar alkali metalaluminum double chloride, adding aluminum metal to the mixture obtained at a temperature of between 260 and 860 deg C, and separating a uranium-containing metal phase and a plutonium-chloride- and fission-product chloridecontaining salt phase. Dissolution can be expedited by passing carbon tetrachloride vapors through the double salt. Separation without reduction of plutonium from neutron- bombarded uranium and that of cerium from uranium are also discussed.

Dissolution of Used Nuclear Fuel Using a TBP/N-Paraffin Solvent

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rudisill, T. S.; Shehee, T. C.; Jones, D. H.

2017-10-02

The dissolution of unirradiated used nuclear fuel (UNF) pellets pretreated for tritium removal was demonstrated using a tributly phosphate (TBP) solvent. Dissolution of pretreated fuel in TBP could potentially combine dissolution with two cycle of solvent extraction required for separating the actinides and lanthanides from other fission products. Dissolutions were performed using UNF surrogates prepared from both uranyl nitrate and uranium trioxide produced from the pretreatment process by adding selected actinide and stable fission product elements. In laboratory-scale experiments, the U dissolution efficiency ranged from 80-99+% for both the nitrate and oxide surrogate fuels. On average, 80% of the Pumore » and 50% of the Np and Am in the nitrate surrogate dissolved; however, little of the transuranic elements dissolved in the oxide form. The majority of the 3+ lanthanide elements dissolved. Only small amounts of Sr (0-1.6%) and Mo (0.1-1.7%) and essentially no Cs, Ru, Zr, or Pd dissolved.« less

KINETICS OF THE DISSOLUTION OF URANIUM DIOXIDE IN CARBONATE-BICARBONATE SOLUTIONS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schortmann, W.E.; DeSesa, M.A.

The kinetics of the dissolution of uranium dioxide in sodium carbonate- sodium bicarbonate solutions were determined. The study was undertaken in order to obtain fundamental information about the commercial carbonate process for leaching uranium from its ores. A rate equation incorporating the effects of surface area oxygen partial pressure, temperature, and reagent concentrations was empirically developed. A mechanism consisting essentially of two consecutive reactions at steady state is proposed. These reactions are the oxidation of U/ sup 4+/ to U/sup 6+/ and the subsequent formation of the uranyl dicarbonate complexion. Depending on the conditions, either or both of these reactionsmore » can determine the over-all rate. The conversion of uranyl dicarbonate to the uranyl tricarbonate complexion is postulated to be very rapid. In the suggested mechanism, the rate-determining phase of the oxidation is the dissociation of adsorbed molecular oxygen. and both the carbonate and bicarbonate ions play equivalent roles in the formation of the uranyl dicarbonate. As indicated by their high activation energies of about 13 and 14 kcal per mole uranium, both reactions are chemical rather than diffusional processes. A mathematical examination of the proposed mechanism produced a rate equation consistent with the experimental information. The credibility of the mechanism was thereby strengthened. (auth)« less

Influence of acidic and alkaline waste solution properties on uranium migration in subsurface sediments.

PubMed

Szecsody, Jim E; Truex, Mike J; Qafoku, Nikolla P; Wellman, Dawn M; Resch, Tom; Zhong, Lirong

2013-08-01

This study shows that acidic and alkaline wastes co-disposed with uranium into subsurface sediments have significant impact on changes in uranium retardation, concentration, and mass during downward migration. For uranium co-disposal with acidic wastes, significant rapid (i.e., hours) carbonate and slow (i.e., 100 s of hours) clay dissolution resulted, releasing significant sediment-associated uranium, but the extent of uranium release and mobility change was controlled by the acid mass added relative to the sediment proton adsorption capacity. Mineral dissolution in acidic solutions (pH2) resulted in a rapid (<10 h) increase in aqueous carbonate (with Ca(2+), Mg(2+)) and phosphate and a slow (100 s of hours) increase in silica, Al(3+), and K(+), likely from 2:1 clay dissolution. Infiltration of uranium with a strong acid resulted in significant shallow uranium mineral dissolution and deeper uranium precipitation (likely as phosphates and carbonates) with downward uranium migration of three times greater mass at a faster velocity relative to uranium infiltration in pH neutral groundwater. In contrast, mineral dissolution in an alkaline environment (pH13) resulted in a rapid (<10h) increase in carbonate, followed by a slow (10 s to 100 s of hours) increase in silica concentration, likely from montmorillonite, muscovite, and kaolinite dissolution. Infiltration of uranium with a strong base resulted in not only uranium-silicate precipitation (presumed Na-boltwoodite) but also desorption of natural uranium on the sediment due to the high ionic strength solution, or 60% greater mass with greater retardation compared with groundwater. Overall, these results show that acidic or alkaline co-contaminant disposal with uranium can result in complex depth- and time-dependent changes in uranium dissolution/precipitation reactions and uranium sorption, which alter the uranium migration mass, concentration, and velocity. Copyright © 2013 Elsevier B.V. All rights reserved.

In situ spectroscopy and spectroelectrochemistry of uranium in high-temperature alkali chloride molten salts.

PubMed

Polovov, Ilya B; Volkovich, Vladimir A; Charnock, John M; Kralj, Brett; Lewin, Robert G; Kinoshita, Hajime; May, Iain; Sharrad, Clint A

2008-09-01

Soluble uranium chloride species, in the oxidation states of III+, IV+, V+, and VI+, have been chemically generated in high-temperature alkali chloride melts. These reactions were monitored by in situ electronic absorption spectroscopy. In situ X-ray absorption spectroscopy of uranium(VI) in a molten LiCl-KCl eutectic was used to determine the immediate coordination environment about the uranium. The dominant species in the melt was [UO 2Cl 4] (2-). Further analysis of the extended X-ray absorption fine structure data and Raman spectroscopy of the melts quenched back to room temperature indicated the possibility of ordering beyond the first coordination sphere of [UO 2Cl 4] (2-). The electrolytic generation of uranium(III) in a molten LiCl-KCl eutectic was also investigated. Anodic dissolution of uranium metal was found to be more efficient at producing uranium(III) in high-temperature melts than the cathodic reduction of uranium(IV). These high-temperature electrolytic processes were studied by in situ electronic absorption spectroelectrochemistry, and we have also developed in situ X-ray absorption spectroelectrochemistry techniques to probe both the uranium oxidation state and the uranium coordination environment in these melts.

A multiphase interfacial model for the dissolution of spent nuclear fuel

NASA Astrophysics Data System (ADS)

Jerden, James L.; Frey, Kurt; Ebert, William

2015-07-01

The Fuel Matrix Dissolution Model (FMDM) is an electrochemical reaction/diffusion model for the dissolution of spent uranium oxide fuel. The model was developed to provide radionuclide source terms for use in performance assessment calculations for various types of geologic repositories. It is based on mixed potential theory and consists of a two-phase fuel surface made up of UO2 and a noble metal bearing fission product phase in contact with groundwater. The corrosion potential at the surface of the dissolving fuel is calculated by balancing cathodic and anodic reactions occurring at the solution interfaces with UO2 and NMP surfaces. Dissolved oxygen and hydrogen peroxide generated by radiolysis of the groundwater are the major oxidizing agents that promote fuel dissolution. Several reactions occurring on noble metal alloy surfaces are electrically coupled to the UO2 and can catalyze or inhibit oxidative dissolution of the fuel. The most important of these is the oxidation of hydrogen, which counteracts the effects of oxidants (primarily H2O2 and O2). Inclusion of this reaction greatly decreases the oxidation of U(IV) and slows fuel dissolution significantly. In addition to radiolytic hydrogen, large quantities of hydrogen can be produced by the anoxic corrosion of steel structures within and near the fuel waste package. The model accurately predicts key experimental trends seen in literature data, the most important being the dramatic depression of the fuel dissolution rate by the presence of dissolved hydrogen at even relatively low concentrations (e.g., less than 1 mM). This hydrogen effect counteracts oxidation reactions and can limit fuel degradation to chemical dissolution, which results in radionuclide source term values that are four or five orders of magnitude lower than when oxidative dissolution processes are operative. This paper presents the scientific basis of the model, the approach for modeling used fuel in a disposal system, and preliminary calculations to demonstrate the application and value of the model.

Development of a mathematical model for the dissolution of uranium dioxide. II. Statistical model for the dissolution of uranium dioxide tablets in nitric acid

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhukovskii, Yu.M.; Luksha, O.P.; Nenarokomov, E.A.

1988-03-01

We have derived a statistical model for the dissolution of uranium dioxide tablets for the 6 to 12 M concentration range and temperatures from 80/sup 0/C to the boiling point. The model differs qualitatively from the dissolution model for ground uranium dioxide. In the indicated range of experimental conditions, the mean-square deviation of the curves for the model from the experimental curves is not greater than 6%.

Effects of nitrate on the stability of uranium in a bioreduced region of the subsurface.

PubMed

Wu, Wei-Min; Carley, Jack; Green, Stefan J; Luo, Jian; Kelly, Shelly D; Van Nostrand, Joy; Lowe, Kenneth; Mehlhorn, Tonia; Carroll, Sue; Boonchayanant, Benjaporn; Löfller, Frank E; Watson, David; Kemner, Kenneth M; Zhou, Jizhong; Kitanidis, Peter K; Kostka, Joel E; Jardine, Philip M; Criddle, Craig S

2010-07-01

The effects of nitrate on the stability of reduced, immobilized uranium were evaluated in field experiments at a U.S. Department of Energy site in Oak Ridge, TN. Nitrate (2.0 mM) was injected into a reduced region of the subsurface containing high levels of previously immobilized U(IV). The nitrate was reduced to nitrite, ammonium, and nitrogen gas; sulfide levels decreased; and Fe(II) levels increased then deceased. Uranium remobilization occurred concomitant with nitrite formation, suggesting nitrate-dependent, iron-accelerated oxidation of U(IV). Bromide tracer results indicated changes in subsurface flowpaths likely due to gas formation and/or precipitate. Desorption-adsorption of uranium by the iron-rich sediment impacted uranium mobilization and sequestration. After rereduction of the subsurface through ethanol additions, background groundwater containing high levels of nitrate was allowed to enter the reduced test zone. Aqueous uranium concentrations increased then decreased. Clone library analyses of sediment samples revealed the presence of denitrifying bacteria that can oxidize elemental sulfur, H(2)S, Fe(II), and U(IV) (e.g., Thiobacillus spp.), and a decrease in relative abundance of bacteria that can reduce Fe(III) and sulfate. XANES analyses of sediment samples confirmed changes in uranium oxidation state. Addition of ethanol restored reduced conditions and triggered a short-term increase in Fe(II) and aqueous uranium, likely due to reductive dissolution of Fe(III) oxides and release of sorbed U(VI). After two months of intermittent ethanol addition, sulfide levels increased, and aqueous uranium concentrations gradually decreased to <0.1 microM.

Tandem dissolution of UO 3 in amide-based acidic ionic liquid and in situ electrodeposition of UO 2 with regeneration of the ionic liquid: a closed cycle

DOE PAGES

Wanigasekara, Eranda; Freiderich, John W.; Sun, Xiao-Guang; ...

2016-05-19

A closed cycle is demonstrated for the tandem dissolution and electroreduction of UO 3 to UO 2 with regeneration of the acidic ionic liquid. The dissolution is achieved by use of the acidic ionic liquid N,N-dimethylacetimidium bis(trifluoromethanesulfonimide) in 1-ethyl-3-methylimidazolium bis(trifluoromethanesulfonimide) serving as the diluent. Bulk electrolysis performed at 1.0 V vs. Ag reference yields a dark brown-black uranium deposit (UO 2) on the cathode. Anodic oxidation of water in the presence of dimethylacetamide regenerates the acidic ionic liquid. We have demonstrated the individual steps in the cycle together with a sequential dissolution, electroreduction, and regeneration cycle.

Tandem dissolution of UO 3 in amide-based acidic ionic liquid and in situ electrodeposition of UO 2 with regeneration of the ionic liquid: a closed cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wanigasekara, Eranda; Freiderich, John W.; Sun, Xiao-Guang

A closed cycle is demonstrated for the tandem dissolution and electroreduction of UO 3 to UO 2 with regeneration of the acidic ionic liquid. The dissolution is achieved by use of the acidic ionic liquid N,N-dimethylacetimidium bis(trifluoromethanesulfonimide) in 1-ethyl-3-methylimidazolium bis(trifluoromethanesulfonimide) serving as the diluent. Bulk electrolysis performed at 1.0 V vs. Ag reference yields a dark brown-black uranium deposit (UO 2) on the cathode. Anodic oxidation of water in the presence of dimethylacetamide regenerates the acidic ionic liquid. We have demonstrated the individual steps in the cycle together with a sequential dissolution, electroreduction, and regeneration cycle.

Characterisation and dissolution of depleted uranium aerosols produced during impacts of kinetic energy penetrators against a tank.

PubMed

Chazel, V; Gerasimo, P; Dabouis, V; Laroche, P; Paquet, F

2003-01-01

Aerosols produced during impacts of depleted uranium (DU) penetrators against the glacis (sloping armour) and the turret of a tank were sampled. The concentration and size distribution were determined. Activity median aerodynamic diameters were 1 microm (geometric standard deviation, sigma(g) = 3.7) and 2 microm (sigma(g) = 2.5), respectively, for glacis and turret. The mean air concentration was 120 Bq m(-3), i.e. 8.5 mg m(-3) of DU. Filters analysed by scanning electron microscopy (SEM) and X ray diffraction showed two types of particles (fine particles and large molten particles) composed mainly of a mixture of uranium and aluminium. The uranium oxides were mostly U3O8, UO2.25 and probably UO3.01 and a mixed compound of U and Al. The kinetics of dissolution in three media (HCO3-, HCl and Gamble's solution) were determined using in-vitro tests. The slow dissolution rates were respectively slow, and intermediate between slow and moderate, and the rapid dissolution fractions were mostly intermediate between moderate and fast. According to the in-vitro results for Gamble's solution, and based on a hypothetical single acute inhalation of 90 Bq, effective doses integrated up to 1 y after incorporation were 0.54 and 0.56 mSv, respectively, for aerosols from glacis and turret. In comparison, the ICRP limits are 20 mSv y(-1) for workers and 1 mSv y(-1) for members of the public. A kidney concentration of approximately 0.1 microg U g(-1) was predicted and should not, in this case, lead to kidney damage.

The roles of organic matter in the formation of uranium deposits in sedimentary rocks

USGS Publications Warehouse

Spirakis, C.S.

1996-01-01

Because reduced uranium species have a much smaller solubility than oxidized uranium species and because of the strong association of organic matter (a powerful reductant) with many uranium ores, reduction has long been considered to be the precipitation mechanism for many types of uranium deposits. Organic matter may also be involved in the alterations in and around tabular uranium deposits, including dolomite precipitation, formation of silicified layers, iron-titanium oxide destruction, dissolution of quartz grains, and precipitation of clay minerals. The diagenetic processes that produced these alterations also consumed organic matter. Consequently, those tabular deposits that underwent the more advanced stages of diagenesis, including methanogenesis and organic acid generation, display the greatest range of alterations and contain the smallest amount of organic matter. Because of certain similarities between tabular uranium deposits and Precambrian unconformity-related deposits, some of the same processes might have been involved in the genesis of Precambrian unconformity-related deposits. Hydrologic studies place important constraints on genetic models of various types of uranium deposits. In roll-front deposits, oxidized waters carried uranium to reductants (organic matter and pyrite derived from sulfate reduction by organic matter). After these reductants were oxidized at any point in the host sandstone, uranium minerals were reoxidized and transported further down the flow path to react with additional reductants. In this manner, the uranium ore migrated through the sandstone at a rate slower than the mineralizing ground water. In the case of tabular uranium deposits, the recharge of surface water into the ground water during flooding of lakes carried soluble humic material to the water table or to an interface where humate precipitated in tabular layers. These humate layers then established the chemical conditions for mineralization and related alterations. In the case of Precambrian unconformity-related deposits, free thermal convection in the thick sandstones overlying the basement rocks carried uranium to concentrations of organic matter in the basement rocks.

Origin of the Mariano Lake uranium deposit, McKinley County, New Mexico

USGS Publications Warehouse

Fishman, Neil S.; Reynolds, Richard L.

1982-01-01

The Mariano Lake uranium deposit, hosted by the Brushy Basin Member of the Jurassic Morrison Formation, occurs in the trough of an east-west trending syncline at the western end of the Smith Lake-Mariano Lake group of uranium deposits near Crownpoint, New Mexico. The orebody, which contains abundant amorphous organic material, is situated on the reduced side of a regional reduction-oxidation (redox) interface. The presence of amorphous organic material suggests the orebody may represent a tabular (primary) deposit, whereas the close proximity of the orebody to the redox interface is suggestive that uranium was secondarily redistributed by oxidative processes from pre-existing tabular orebodies. Uranium contents correlate positively with both organic carbon and vanadium contents. Petrographic evidence and scanning electron microscope-energy dispersive analyses point to uranium residence in the epigentically introduced amorphous organic material, which coats detrital grains and fills voids. Uranium mineralization was preceded by the following diagenetic alterations: precipitation of pyrite (d34S values ranging from-11.0 to-38.2 per mil); precipitation of mixed-layer smectite-illite clays; partial dissolution of some of the detrital feldspar population; and precipitation of quartz and adularia overgrowths. Alterations associated with uranium mineralization include emplacement of amorphous organic material (possibly uranium bearing); destruction of detrital iron-titanium oxide grains; coprecipitation of chlorite and microcrystalline quartz, and precipitation of pyrite and marcasite (d34S values for these sulfides ranging from -29.4 to -41.6 per mil). After mineralization, calcite, dolomite, barite, and kaolinite precipitated, and authigenic iron disulfides were replaced by ferric oxides and hydroxides. Geochemical data (primarily the positive correlation of uranium content to both organic carbon and vanadium contents) and petrographic observations (epigentically introduced amorphous organic matter and uranium residence in this organic matter) indicate that the Mariano Lake orebody is a tabular-type uranium deposit. Oxidative processes have not noticeably redistributed and reconcentrated primary uranium in the immediate vicinity of the deposit nor have they greatly modified geochemical characteristics in the ore. Preservation of the Mariano Lake deposit may not only be related to its position along the synclinal trough, where oxidative destruction of the orebody has been inhibited by stagnation of oxidizing ground waters by the structure, but also due to the deflection of ground waters (resulting from low orebody porosity) around the orebody.

XPS studies of UO 2 oxidation by alpha radiolysis of water at 100°C

NASA Astrophysics Data System (ADS)

Sunder, S.; Boyer, G. D.; Miller, N. H.

1990-12-01

The effect of alpha radiolysis of water on the oxidation and dissolution of UO 2 was studied at 100°C as a function of alpha-field strength and water chemistry using X-ray photoelectron spectroscopy. In N 2-purged solutions the oxidation of UO 2 increases with the strength of the alpha flux; an alpha flux greater than or equal to that from a 250-μ Ci americium-241 source leads to oxidation of UO 2 beyond the UO 2.33 (U 3O 7) stage, and an alpha flux equal to that from a 5-μ Ci source does not result in UO 2 oxidation beyond the UO 2.33 stage. The presence of dissolved H 2 in water, at a concentration ≥ 1.6 × 10 -4moldm-3, reduces the oxidation and dissolution of UO 2 due to alpha radiolysis at temperatures ≥ 100° C. It is concluded that the radiolysis of groundwater at ~ 100°C, due to the alpha flux associated with used CANDU fuel, is unlikely to make a significant contribution to its oxidative dissolution in the geological disposal vault planned in the Canadian Nuclear Fuel Waste Management Program. CANada Deuterium Uranium. Registered trademark.

Solubility classification of airborne uranium products collected at the perimeter of the Allied Chemical Plant, Metropolis, Illinois

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kalkwarf, D.R.

1980-05-01

Airborne uranium products were collected at the perimeter of the uranium-conversion plant operated by the Allied Chemical Corporation at Metropolis, Illinois, and the dissolution rates of these products were classified in terms of the ICRP Task Group Lung Model. Assignments were based on measurements of the dissolution half-times exhibited by uranium components of the dust samples as they dissolved in simulated lung fluid at 37/sup 0/C. Based on three trials, the dissolution behavior of dust with aerodynamic equivalent diameter (AED) less than 5.5 ..mu..m and collected nearest the closest residence to the plant was classified 0.40 D, 0.60 Y. Basedmore » on two trials, the dissolution behavior of dust with AED greater than 5.5 ..mu..m and collected at this location was classified 0.37 D, 0.63 Y. Based on one trial, the dissolution behavior of dust with AED less than 5.5 ..mu..m and collected at a location on the opposite side of the plant was classified 0.68 D, 0.32 Y. There was some evidence for adsorption of dissolved uranium onto other dust components during dissolution, and preliminary dissolution trials are recommended for future samples in order to optimize the fluid replacement schedule.« less

Behavior of S.A.P. in the Mercury Catalyzed Nitric Acid Dissolution; COMPORTAMENTO DEL S.A.P. ALL'ATTACCO DI SOLUZIONI DI ACIDO NITRICO E NITRATO MERCURICO

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beone, G.

1963-10-01

Plates of S.A.P. (sintered Aluminum Powder) were dissolved under different conditions in a nitric acid solution containing mercuric nitrate as a catalyst. These experiments nim at establishing a head-end dissolution process for S.A.P. cladded uranium oxide fuels. The results of preliminary dissolution experiments on simulated fuel rods are also described. The behavior of S.A.P. in the mercury catalyzed nitric acid dissolutions differs strongly from the behavior of aluminum: reaction rates are very low for S.A.P. and the dissolution time borders on being unacceptable in an industrial process. Settling rates of suspended alumina are however favorable. A tentative head end flowsheetmore » lay out for PRO second core fuel elements is included. (auth)« less

Effects of ammonium on uranium partitioning and kaolinite mineral dissolution.

PubMed

Emerson, Hilary P; Di Pietro, Silvina; Katsenovich, Yelena; Szecsody, Jim

2017-02-01

Ammonia gas injection is a promising technique for the remediation of uranium within the vadose zone. It can be used to manipulate the pH of a system and cause co-precipitation processes that are expected to remove uranium from the aqueous phase and decrease leaching from the solid phase. The work presented in this paper explores the effects of ammonium and sodium hydroxide on the partitioning of uranium and dissolution of the kaolinite mineral in simplified synthetic groundwaters using equilibrium batch sorption and sequential extraction experiments. It shows that there is a significant increase in uranium removal in systems with divalent cations present in the aqueous phase but not in sodium chloride synthetic groundwaters. Further, the initial conditions of the aqueous phase do not affect the dissolution of kaolinite. However, the type of base treatment does have an effect on mineral dissolution. Published by Elsevier Ltd.

10 CFR Appendix I to Part 110 - Illustrative List of Reprocessing Plant Components Under NRC Export Licensing Authority

Code of Federal Regulations, 2014 CFR

2014-01-01

... dissolution, solvent extraction, and process liquor storage. There may also be equipment for thermal denitration of uranium nitrate, conversion of plutonium nitrate to oxide metal, and treatment of fission product waste liquor to a form suitable for long term storage or disposal. However, the specific type and...

10 CFR Appendix I to Part 110 - Illustrative List of Reprocessing Plant Components Under NRC Export Licensing Authority

Code of Federal Regulations, 2013 CFR

2013-01-01

... dissolution, solvent extraction, and process liquor storage. There may also be equipment for thermal denitration of uranium nitrate, conversion of plutonium nitrate to oxide metal, and treatment of fission product waste liquor to a form suitable for long term storage or disposal. However, the specific type and...

10 CFR Appendix I to Part 110 - Illustrative List of Reprocessing Plant Components Under NRC Export Licensing Authority

Code of Federal Regulations, 2012 CFR

2012-01-01

... dissolution, solvent extraction, and process liquor storage. There may also be equipment for thermal denitration of uranium nitrate, conversion of plutonium nitrate to oxide metal, and treatment of fission product waste liquor to a form suitable for long term storage or disposal. However, the specific type and...

Molybdenum isotope fractionation during acid leaching of a granitic uranium ore

NASA Astrophysics Data System (ADS)

Migeon, Valérie; Bourdon, Bernard; Pili, Eric; Fitoussi, Caroline

2018-06-01

As an attempt to prevent illicit trafficking of nuclear materials, it is critical to identify the origin and transformation of uranium materials from the nuclear fuel cycle based on chemical and isotope tracers. The potential of molybdenum (Mo) isotopes as tracers is considered in this study. We focused on leaching, the first industrial process used to release uranium from ores, which is also known to extract Mo depending on chemical conditions. Batch experiments were performed in the laboratory with pH ranging from 0.3 to 5.5 in sulfuric acid. In order to span a large range in uranium and molybdenum yields, oxidizers such as nitric acid, hydrogen peroxide and manganese dioxide were also added. An enrichment in heavy Mo isotopes is produced in the solution during leaching of a granitic uranium ore, when Mo recovery is not quantitative. At least two Mo reservoirs were identified in the ore: ∼40% as Mo oxides soluble in water or sulfuric acid, and ∼40% of Mo hosted in sulfides soluble in nitric acid or hydrogen peroxide. At pH > 1.8, adsorption and/or precipitation processes induce a decrease in Mo yields with time correlated with large Mo isotope fractionations. Quantitative models were used to evaluate the relative importance of the processes involved in Mo isotope fractionation: dissolution, adsorption, desorption, precipitation, polymerization and depolymerization. Model best fits are obtained when combining the effects of dissolution/precipitation, and adsorption/desorption onto secondary minerals. These processes are inferred to produce an equilibrium isotope fractionation, with an enrichment in heavy Mo isotopes in the liquid phase and in light isotopes in the solid phase. Quantification of Mo isotope fractionation resulting from uranium leaching is thus a promising tool to trace the origin and transformation of nuclear materials. Our observations of Mo leaching are also consistent with observations of natural Mo isotope fractionation taking place during chemical weathering in terrestrial environments where the role of secondary processes such as adsorption is significant.

Geochemistry of vanadium in an epigenetic, sandstone-hosted vanadium- uranium deposit, Henry Basin, Utah

USGS Publications Warehouse

Wanty, R.B.; Goldhaber, M.B.; Northrop, H.R.

1990-01-01

The epigenetic Tony M vanadium-uranium orebody in south-central Utah is hosted in fluvial sandstones of the Morrison Formation (Upper Jurassic). Measurements of the relative amounts of V+3 and V +4 in ore minerals show that V+3 is more abundant. Thermodynamic calculations show that vanadium was more likely transported to the site of mineralization as V+4. The ore formed as V+4 was reduced by hydrogen sulfide, followed by hydrolysis and precipitation of V+3 in oxide minerals or chlorite. Uranium was transported as uranyl ion (U+6), or some complex thereof, and reduced by hydrogen sulfide, forming coffinite. Detrital organic matter in the rocks served as the carbon source for sulfate-reducing bacteria. Vanadium most likely was derived from the dissolution of iron-titanium oxides. Uranium probably was derived from the overlying Brushy Basin Member of the Morrison Formation. Previous studies have shown that the ore formed at the density-stratified interface between a basinal brine and dilute meteoric water. The mineralization processes described above occurred within the mixing zone between these two fluids. -from Authors

Method for providing uranium articles with a corrosion resistant anodized coating

DOEpatents

Waldrop, Forrest B.; Washington, Charles A.

1982-01-01

Uranium articles are provided with anodized oxide coatings in an aqueous solution of an electrolyte selected from the group consisting of potassium phosphate, potassium hydroxide, ammonium hydroxide, and a mixture of potassium tetraborate and boric acid. The uranium articles are anodized at a temperature greater than about 75.degree. C. with a current flow of less than about 0.036 A/cm.sup.2 of surface area while the pH of the solution is maintained in a range of about 2 to 11.5. The pH values of the aqueous solution and the low current density utilized during the electrolysis prevent excessive dissolution of the uranium and porosity in the film or watering. The relatively high temperature of the electrolyte bath inhibits hydration and the attendant deleterious pitting so as to enhance corrosion resistance of the anodized coating.

Composition, stability, and measurement of reduced uranium phases for groundwater bioremediation at Old Rifle, CO

DOE Office of Scientific and Technical Information (OSTI.GOV)

Campbell, K. M.; Davis, J. A.; Bargar, J.

2011-10-15

Reductive biostimulation is currently being explored as a possible remediation strategy for uranium (U) contaminated groundwater, and is currently being investigated at a field site in Rifle, CO, USA. The long-term stability of the resulting U(IV) phases is a key component of the overall performance and depends upon a variety of factors, including rate and mechanism of reduction, mineral associations in the subsurface, and propensity for oxidation. To address these factors, several approaches were used to evaluate the redox sensitivity of U: measurement of the rate of oxidative dissolution of biogenic uraninite (UO{sub 2(s)}) deployed in groundwater at Rifle, characterizationmore » of a zone of natural bioreduction exhibiting relevant reduced mineral phases, and laboratory studies of the oxidative capacity of Fe(III) and reductive capacity of Fe(II) with regard to U(IV) and U(VI), respectively.« less

METHOD OF DISSOLVING URANIUM METAL

DOEpatents

Slotin, L.A.

1958-02-18

This patent relates to an economicai means of dissolving metallic uranium. It has been found that the addition of a small amount of perchloric acid to the concentrated nitric acid in which the uranium is being dissolved greatly shortens the time necessary for dissolution of the metal. Thus the use of about 1 or 2 percent of perchioric acid based on the weight of the nitric acid used, reduces the time of dissolution of uranium by a factor of about 100.

Influence of microwaves on the leaching kinetics of uraninite from a low grade ore in dilute sulfuric acid.

PubMed

Madakkaruppan, V; Pius, Anitha; T, Sreenivas; Giri, Nitai; Sarbajna, Chanchal

2016-08-05

This paper describes a study on microwave assisted leaching of uranium from a low-grade ore of Indian origin. The host rock for uranium mineralization is chlorite-biotite-muscovite-quartzo-feldspathic schist. The dominant presence of siliceous minerals determined leaching of uranium values in sulfuric acid medium under oxidizing conditions. Process parametric studies like the effect of sulfuric acid concentration (0.12-0.50M), redox potential (400-500mV), particle size (600-300μm) and temperature (35°-95°C) indicated that microwave assisted leaching is more efficient in terms of overall uranium dissolution, kinetics and provide relatively less impurities (Si, Al, Mg and Fe) in the leach liquor compared to conventional conductive leaching. The kinetics of leaching followed shrinking core model with product layer diffusion as controlling mechanism. Copyright © 2016 Elsevier B.V. All rights reserved.

Method for providing uranium articles with a corrosion-resistant anodized coating

DOEpatents

Waldrop, F.B.; Washington, C.A.

1981-01-07

Uranium articles are provided with anodized oxide coatings in an aqueous solution of an electrolyte selected from the group consisting of potassium phosphate, potassium hydroxide, ammonium hydroxide, and a mixture of potassium tetraborate and boric acid. The uranium articles are anodized at a temperature greater than about 75/sup 0/C with a current flow of less than about 0.036 A/cm/sup 2/ of surface area while the pH of the solution is maintained in a range of about 2 to 11.5. The pH values of the aqueous solution and the low current density utilized during the electrolysis prevent excessive dissolution of the uranium and porosity in the film or watering. The relatively high temperature of the electrolyte bath inhibits hydration and the attendant deleterious pitting so as to enhance corrosion resistance of the anodized coating.

Rapid fusion method for the determination of refractory thorium and uranium isotopes in soil samples

DOE PAGES

Maxwell, Sherrod L.; Hutchison, Jay B.; McAlister, Daniel R.

2015-02-14

Recently, approximately 80% of participating laboratories failed to accurately determine uranium isotopes in soil samples in the U.S Department of Energy Mixed Analyte Performance Evaluation Program (MAPEP) Session 30, due to incomplete dissolution of refractory particles in the samples. Failing laboratories employed acid dissolution methods, including hydrofluoric acid, to recover uranium from the soil matrix. The failures illustrate the importance of rugged soil dissolution methods for the accurate measurement of analytes in the sample matrix. A new rapid fusion method has been developed by the Savannah River National Laboratory (SRNL) to prepare 1-2 g soil sample aliquots very quickly, withmore » total dissolution of refractory particles. Soil samples are fused with sodium hydroxide at 600 ºC in zirconium crucibles to enable complete dissolution of the sample. Uranium and thorium are separated on stacked TEVA and TRU extraction chromatographic resin cartridges, prior to isotopic measurements by alpha spectrometry on cerium fluoride microprecipitation sources. Plutonium can also be separated and measured using this method. Batches of 12 samples can be prepared for measurement in <5 hours.« less

Kinetics of dissolution of UO2 in nitric acid solutions: A multiparametric study of the non-catalysed reaction

NASA Astrophysics Data System (ADS)

Cordara, T.; Szenknect, S.; Claparede, L.; Podor, R.; Mesbah, A.; Lavalette, C.; Dacheux, N.

2017-12-01

UO2 pellets were prepared by densification of oxides obtained from the conversion of the oxalate precursor. Then characterized in order to perform a multiparametric study of the dissolution in nitric acid medium. In this frame, for each sample, the densification rate, the grain size and the specific surface area of the prepared pellets were determined prior to the final dissolution experiments. By varying the concentration of the nitric acid solution and temperature, three different and successive steps were identified during the dissolution. Under the less aggressive conditions considered, a first transient step corresponding to the dissolution of the most reactive phases was observed at the solid/solution interface. Then, for all the tested conditions, a steady state step was established during which the normalised dissolution rate was found to be constant. It was followed by a third step characterized by a strong and continuous increase of the normalised dissolution rate. The duration of the steady state, also called "induction period", was found to vary drastically as a function of the HNO3 concentration and temperature. However, independently of the conditions, this steady state step stopped at almost similar dissolved material weight loss and dissolved uranium concentration. During the induction period, no important evolution of the topology of the solid/liquid interface was evidenced authorizing the use of the starting reactive specific surface area to evaluate the normalised dissolution rates thus the chemical durability of the sintered pellets. From the multiparametric study of UO2 dissolution proposed, oxidation of U(IV) to U(VI) by nitrate ions at the solid/liquid interface constitutes the limiting step in the overall dissolution mechanism associated to this induction period.

Hydrogen suppresses UO 2 corrosion

NASA Astrophysics Data System (ADS)

Carbol, Paul; Fors, Patrik; Gouder, Thomas; Spahiu, Kastriot

2009-08-01

Release of long-lived radionuclides such as plutonium and caesium from spent nuclear fuel in deep geological repositories will depend mainly on the dissolution rate of the UO 2 fuel matrix. This dissolution rate will, in turn, depend on the redox conditions at the fuel surface. Under oxidative conditions UO 2 will be oxidised to the 1000 times more soluble UO 2.67. This may occur in a repository as the reducing deep groundwater becomes locally oxidative at the fuel surface under the effect of α-radiolysis, the process by which α-particles emitted from the fuel split water molecules. On the other hand, the groundwater corrodes canister iron generating large amounts of hydrogen. The role of molecular hydrogen as reductant in a deep bedrock repository is questioned. Here we show evidence of a surface-catalysed reaction, taking place in the H 2-UO 2-H 2O system where molecular hydrogen is able to reduce oxidants originating from α-radiolysis. In our experiment the UO 2 surface remained stoichiometric proving that the expected oxidation of UO 2.00 to UO 2.67 due to radiolytic oxidants was absent. As a consequence, the dissolution of UO 2 stopped when equilibrium was reached between the solid phase and U 4+ species in the aqueous phase. The steady-state concentration of uranium in solution was determined to be 9 × 10 -12 M, about 30 times lower than previously reported for reducing conditions. Our findings show that fuel dissolution is suppressed by H 2. Consequently, radiotoxic nuclides in spent nuclear fuel will remain immobilised in the UO 2 matrix. A mechanism for the surface-catalysed reaction between molecular hydrogen and radiolytic oxidants is proposed.

Determination of actinides in urine and fecal samples

DOEpatents

McKibbin, Terry T.

1993-01-01

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

Determination of actinides in urine and fecal samples

DOEpatents

McKibbin, T.T.

1993-03-02

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

U redox fronts and kaolinisation in basement-hosted unconformity-related U ores of the Athabasca Basin (Canada): late U remobilisation by meteoric fluids

NASA Astrophysics Data System (ADS)

Mercadier, Julien; Cuney, Michel; Cathelineau, Michel; Lacorde, Mathieu

2011-02-01

Proterozoic basement-hosted unconformity-related uranium deposits of the Athabasca Basin (Saskatchewan, Canada) were affected by significant uranium redistribution along oxidation-reduction redox fronts related to cold and late meteoric fluid infiltration. These redox fronts exhibit the same mineralogical and geochemical features as the well-studied uranium roll-front deposits in siliclastic rocks. The primary hydrothermal uranium mineralisation (1.6-1.3 Ga) of basement-hosted deposits is strongly reworked to new disseminated ores comprising three distinctly coloured zones: a white-green zone corresponding to the previous clay-rich alteration halo contemporaneous with hydrothermal ores, a uranium front corresponding to the uranium deposition zone of the redox front (brownish zone, rich in goethite) and a hematite-rich red zone marking the front progression. The three zones directly reflect the mineralogical zonation related to uranium oxides (pitchblende), sulphides, iron minerals (hematite and goethite) and alumino-phosphate-sulphate (APS) minerals. The zoning can be explained by processes of dissolution-precipitation along a redox interface and was produced by the infiltration of cold (<50°C) meteoric fluids to the hydrothermally altered areas. U, Fe, Ca, Pb, S, REE, V, Y, W, Mo and Se were the main mobile elements in this process, and their distribution within the three zones was, for most of them, directly dependent on their redox potential. The elements concentrated in the redox fronts were sourced by the alteration of previously crystallised hydrothermal minerals, such as uranium oxides and light rare earth element (LREE)-rich APS. The uranium oxides from the redox front are characterised by LREE-enriched patterns, which differ from those of unconformity-related ores and clearly demonstrate their distinct conditions of formation. Uranium redox front formation is thought to be linked to fluid circulation episodes initiated during the 400-300 Ma period during uplift and erosion of the Athabasca Basin when it was near the Equator and to have been still active during the last million years. A major kaolinisation event was caused by changes in the fluid circulation regime, reworking the primary uranium redox fronts and causing the redistribution of elements originally concentrated in the uranium-enriched meteoric-related redox fronts.

Rapid Dissolution of Soluble Uranyl Phases in Arid, Mine-Impacted Catchments Near Church Rock, NM

DOE Office of Scientific and Technical Information (OSTI.GOV)

deLemos, J.L.; Bostick, B.C.; Quicksall, A.N.

2009-05-14

We tested the hypothesis that runoff of uranium-bearing particles from mining waste disposal areas was a significant mechanism for redistribution of uranium in the northeastern part of the Upper Puerco River watershed (New Mexico). However, our results were not consistent with this hypothesis. Analysis of >100 sediment and suspended sediment samples collected adjacent to and downstream from uranium source areas indicated that uranium levels in the majority of the samples were not elevated above background. Samples collected within 50 m of a known waste disposal site were subjected to detailed geochemical characterization. Uranium in these samples was found to bemore » highly soluble; treatment with synthetic pore water for 24 h caused dissolution of 10-50% of total uranium in the samples. Equilibrium uranium concentrations in pore water were >4.0 mg/L and were sustained in repeated wetting events, effectively depleting soluble uranium from the solid phase. The dissolution rate of uranium appeared to be controlled by solid-phase diffusion of uranium from within uranium-bearing mineral particles. X-ray adsorption spectroscopy indicated the presence of a soluble uranyl silicate, and possibly a uranyl phosphate. These phases were exhausted in transported sediment suggesting that uranium was readily mobilized from sediments in the Upper Puerco watershed and transported in the dissolved load. These results could have significance for uranium risk assessment as well as mining waste management and cleanup efforts.« less

Rapid Dissolution of Soluble Uranyl Phases in Arid, Mine-Impacted Catchments near Church Rock, NM

PubMed Central

DELEMOS, JAMIE L.; BOSTICK, BENJAMIN C.; QUICKSALL, ANDREW N.; LANDIS, JOSHUA D.; GEORGE, CHRISTINE C.; SLAGOWSKI, NAOMI L.; ROCK, TOMMY; BRUGGE, DOUG; LEWIS, JOHNNYE; DURANT, JOHN L.

2008-01-01

We tested the hypothesis that runoff of uranium-bearing particles from mining waste disposal areas was a significant mechanism for redistribution of uranium in the northeastern part of the Upper Puerco River watershed (New Mexico). However, our results were not consistent with this hypothesis. Analysis of >100 sediment and suspended sediment samples collected adjacent to and downstream from uranium source areas indicated that uranium levels in the majority of the samples were not elevated above background. Samples collected within 50 m of a known waste disposal site were subjected to detailed geochemical characterization. Uranium in these samples was found to be highly soluble; treatment with synthetic pore water for 24 h caused dissolution of 10–50% of total uranium in the samples. Equilibrium uranium concentrations in pore water were >4.0 mg/L and were sustained in repeated wetting events, effectively depleting soluble uranium from the solid phase. The dissolution rate of uranium appeared to be controlled by solid-phase diffusion of uranium from within uranium-bearing mineral particles. X-ray adsorption spectroscopy indicated the presence of a soluble uranyl silicate, and possibly a uranyl phosphate. These phases were exhausted in transported sediment suggesting that uranium was readily mobilized from sediments in the Upper Puerco watershed and transported in the dissolved load. These results could have significance for uranium risk assessment as well as mining waste management and cleanup efforts. PMID:18589950

Uranium carbide dissolution in nitric solution: Sonication vs. silent conditions

NASA Astrophysics Data System (ADS)

Virot, Matthieu; Szenknect, Stéphanie; Chave, Tony; Dacheux, Nicolas; Moisy, Philippe; Nikitenko, Sergey I.

2013-10-01

The dissolution of uranium carbide (UC) in nitric acid media is considered by means of power ultrasound (sonication) or magnetic stirring. The induction period required to initiate UC dissolution was found to be dramatically shortened when sonicating a 3 M nitric solution (Ar, 20 kHz, 18 W cm-2, 20 °C). At higher acidity, magnetic stirring offers faster dissolution kinetics compared to sonication. Ultrasound-assisted UC dissolution is found to be passivated after ∼60% dissolution and remains incomplete whatever the acidity which is confirmed by ICP-AES, LECO and SEM-EDX analyses. In general, the kinetics of UC dissolution is linked to the in situ generation of nitrous acid in agreement with the general mechanism of UC dissolution; the nitrous acid formation is reported to be faster under ultrasound at low acidity due to the nitric acid sonolysis. The carbon balance shared between the gaseous, liquid, and solid phases is strongly influenced by the applied dissolution procedure and HNO3 concentration.

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.

Investigation on Microbial Dissolution of Uranium (VI) from Autunite Mineral - 13421

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sepulveda, Paola; Katsenovich, Yelena; Lagos, Leonel

2013-07-01

Precipitating autunite minerals by polyphosphate injection was identified as a feasible remediation strategy for sequestering uranium in contaminated groundwater and soil in situ at the Hanford Site. Autunite stability under vadose and saturated zone environmental conditions can help to determine the long-term effectiveness of this remediation strategy. The Arthrobacter bacteria are one of the most common groups in soils and are found in large numbers in Hanford soil as well as other subsurface environments contaminated with radionuclides. Ubiquitous in subsurface microbial communities, these bacteria can play a significant role in the dissolution of minerals and the formation of secondary minerals.more » The main objective of this investigation was to study the bacterial interactions under oxidizing conditions with uranium (VI); study the potential role of bicarbonate, which is an integral complexing ligand for U(VI) and a major ion in groundwater compositions; and present data from autunite dissolution experiments using Arthrobacter strain G968, a less U(VI)-tolerant strain. Sterile 100 mL glass mixed reactors served as the major bioreactor for initial experimentation. These autunite-containing bioreactors were injected with bacterial cells after the autunite equilibrated with the media solution amended with 0 mM, 3 mM 5 mM and 10 mM concentrations of bicarbonate. G968 Arthrobacter cells in the amount of 10{sup 6} cells/mL were injected into the reactors after 27 days, giving time for the autunite to reach steady state. Abiotic non-carbonate controls were kept without bacterial inoculation to provide a control for the biotic samples. Samples of the solution were analyzed for dissolved U(VI) by means of kinetic phosphorescence analyzer KPA-11 (Chemcheck Instruments, Richland, WA). Analysis showed that as [HCO{sub 3}{sup -}] increases, a diminishing trend on the effect of bacteria on autunite leaching is observed. Viability of cells was conducted after 24 hours of cell incubation with the appropriate uranium and bicarbonate concentration treatment. As expected, the cells started to reduce after day 41 due to the nutritional exhaustion of the media. Moreover, viable bacteria accounted for more than 94% in the presence of 10 mM bicarbonate. Experiments showed that despite differences between the G975 and the G968 bacterial strains resistance to U(VI), in the presence of bicarbonate ions they are able to dissolute uranium from autunite mineral at the same capacity. The effect of both bacterial strains on autunite dissolution is reduced as the concentration of bicarbonate increases while the increase in soluble U(VI) concentration induced by G968 and G975 is dwarfed, for larger [HCO{sub 3}{sup -}]. (authors)« less

Effect of mineral constituents in the bioleaching of uranium from uraniferous sedimentary rock samples, Southwestern Sinai, Egypt.

PubMed

Amin, Maisa M; Elaassy, Ibrahim E; El-Feky, Mohamed G; Sallam, Abdel Sattar M; Talaat, Mona S; Kawady, Nilly A

2014-08-01

Bioleaching, like Biotechnology uses microorganisms to extract metals from their ore materials, whereas microbial activity has an appreciable effect on the dissolution of toxic metals and radionuclides. Bioleaching of uranium was carried out with isolated fungi from uraniferous sedimentary rocks from Southwestern Sinai, Egypt. Eight fungal species were isolated from different grades of uraniferous samples. The bio-dissolution experiments showed that Aspergillus niger and Aspergillus terreus exhibited the highest leaching efficiencies of uranium from the studied samples. Through monitoring the bio-dissolution process, the uranium grade and mineralogic constituents of the ore material proved to play an important role in the bioleaching process. The tested samples asserted that the optimum conditions of uranium leaching are: 7 days incubation time, 3% pulp density, 30 °C incubation temperature and pH 3. Both fungi produced the organic acids, namely; oxalic, acetic, citric, formic, malonic, galic and ascorbic in the culture filtrate, indicating an important role in the bioleaching processes. Copyright © 2014 Elsevier Ltd. All rights reserved.

Potential aquifer vulnerability in regions down-gradient from uranium in situ recovery (ISR) sites.

PubMed

Saunders, James A; Pivetz, Bruce E; Voorhies, Nathan; Wilkin, Richard T

2016-12-01

Sandstone-hosted roll-front uranium ore deposits originate when U(VI) dissolved in groundwater is reduced and precipitated as insoluble U(IV) minerals. Groundwater redox geochemistry, aqueous complexation, and solute migration are important in leaching uranium from source rocks and transporting it in low concentrations to a chemical redox interface where it is deposited in an ore zone typically containing the uranium minerals uraninite, pitchblende, and/or coffinite; various iron sulfides; native selenium; clays; and calcite. In situ recovery (ISR) of uranium ores is a process of contacting the uranium mineral deposit with leaching and oxidizing (lixiviant) fluids via injection of the lixiviant into wells drilled into the subsurface aquifer that hosts uranium ore, while other extraction wells pump the dissolved uranium after dissolution of the uranium minerals. Environmental concerns during and after ISR include water quality degradation from: 1) potential excursions of leaching solutions away from the injection zone into down-gradient, underlying, or overlying aquifers; 2) potential migration of uranium and its decay products (e.g., Ra, Rn, Pb); and, 3) potential mobilization and migration of redox-sensitive trace metals (e.g., Fe, Mn, Mo, Se, V), metalloids (e.g., As), and anions (e.g., sulfate). This review describes the geochemical processes that control roll-front uranium transport and fate in groundwater systems, identifies potential aquifer vulnerabilities to ISR operations, identifies data gaps in mitigating these vulnerabilities, and discusses the hydrogeological characterization involved in developing a monitoring program. Published by Elsevier Ltd.

The application of zero-valent iron nanoparticles for the remediation of a uranium-contaminated waste effluent.

PubMed

Dickinson, Michelle; Scott, Thomas B

2010-06-15

Zero-valent iron nanoparticles (INP) were investigated as a remediation strategy for a uranium-contaminated waste effluent from AWE, Aldermaston. Nanoparticles were introduced to the effluent, under both oxic and anoxic conditions, and allowed to react for a 28-d period during which the liquid and nanoparticle solids were periodically sampled. Analysis of the solution indicated that under both conditions U was removed to <1.5% of its initial concentration within 1h of introduction and remained at similar concentrations until approximately 48 h. A rapid release of Fe into solution was also recorded during this initial period; attributed to the limited partial dissolution of the INP. XPS analyses of the reacted nanoparticulate solids between 1 and 48 h showed an increased Fe(III):Fe(II) ratio, consistent with the detection of iron oxidation products (akaganeite and magnetite) by XRD and FIB. XPS analysis also recorded uranium on the recovered particulates indicating the chemical reduction of U(VI) to U(IV) within 1h. Following the initial retention period U-dissolution of U was recorded from 48 h, and attributed to reoxidation. The efficient uptake and retention of U on the INP for periods up to 48 h provide proof that INP may be effectively used for the remediation of complex U-contaminated effluents. Copyright 2010 Elsevier B.V. All rights reserved.

DISSOLUTION OF ZIRCONIUM-CONTAINING FUEL ELEMENTS

DOEpatents

Horn, F.L.

1961-12-12

Uranium is recovered from spent uranium fuel elements containing or clad with zirconium. These fuel elements are placed in an anhydrous solution of hydrogen fluoride and nitrogen dioxide. Within this system uranium forms a soluble complex and zirconium forms an insoluble complex. The uranium can then be separated, treated, and removed from solution as uranium hexafluoride. (AEC)

13. VIEW OF A BBOX, WHICH WAS USED IN THE ...

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

13. VIEW OF A B-BOX, WHICH WAS USED IN THE FAST RECOVERY PROCESS. URANIUM OXIDE WAS TRANSFERRED FOR DISSOLUTION IN A ROOM WHICH HOUSED 3 ROWS OF B-BOXES. B-BOXES ARE CONTROLLED HOODS, SIMILAR TO LAB HOODS THAT OPERATED WITH HIGH AIR VELOCITIES AT THEIR OPENINGS TO ENSURE THAT THE VAPORS WERE CONTAINED WITHIN THE HOOD. (2/14/79) - Rocky Flats Plant, General Manufacturing, Support, Records-Central Computing, Southern portion of Plant, Golden, Jefferson County, CO

A method for phenomenological and chemical kinetics study of autocatalytic reactive dissolution by optical microscopy. The case of uranium dioxide dissolution in nitric acid media

NASA Astrophysics Data System (ADS)

Marc, Philippe; Magnaldo, Alastair; Godard, Jérémy; Schaer, Éric

2018-03-01

Dissolution is a milestone of the head-end of hydrometallurgical processes, as the stabilization rates of the chemical elements determine the process performance and hold-up. This study aims at better understanding the chemical and physico-chemical phenomena of uranium dioxide dissolution reactions in nitric acid media in the Purex process, which separates the reusable materials and the final wastes of the spent nuclear fuels. It has been documented that the attack of sintering-manufactured uranium dioxide solids occurs through preferential attack sites, which leads to the development of cracks in the solids. Optical microscopy observations show that in some cases, the development of these cracks leads to the solid cleavage. It is shown here that the dissolution of the detached fragments is much slower than the process of the complete cleavage of the solid, and occurs with no disturbing phenomena, like gas bubbling. This fact has motivated the measurement of dissolution kinetics using optical microscopy and image processing. By further discriminating between external resistance and chemical reaction, the "true" chemical kinetics of the reaction have been measured, and the highly autocatalytic nature of the reaction confirmed. Based on these results, the constants of the chemical reactions kinetic laws have also been evaluated.

Rates and mechanisms of uranyl oxyhydroxide mineral dissolution

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reinoso-Maset, Estela; Steefel, Carl I.; Um, Wooyong

Uranyl oxyhydroxide minerals are important weathering products in uranium-contaminated surface and subsurface environments that regulate dissolved uranium concentrations. However, dissolution rates for this class of minerals and associated dissolution mechanisms have not been previously reported for circumneutral pH conditions, particularly for the case of flow through porous media. In this paper, the dissolution rates of K- and Na-compreignacite (K 2(UO 2) 6O 4(OH) 6·8H 2O and Na 2(UO 2) 6O 4(OH) 6·8H 2O respectively) were measured using flow-through columns reacted with two simulated background porewater (BPW) solutions of low and high dissolved total carbonate content (ca. 0.2 and 2.8 mmolmore » L -1). Column materials were characterized before and after reaction with electron microscopy, bulk chemistry, and EXAFS to identify structural and chemical changes during dissolution and to obtain insight into molecular-scale processes. The reactive transport code CrunchFlow was used to calculate overall dissolution rates while accounting for fluid transport and changes in mineral volume and reactive surface area and results were compared to steady-state dissolution rate calculations. In low carbonate BPW systems, interlayer K and Na were initially leached from both minerals, and in Na-compreignacite, K and minor divalent cations from the input solution were incorporated into the mineral structure. Results of characterization analyses suggested that after reaction both K- and Na-compreignacite resembled a disordered K-compreignacite with altered surfaces. A 10-fold increase in dissolved carbonate concentration and corresponding increase in pH (from 6.65 to 8.40) resulted in a net removal of 58-87% of total uranium mass from the columns, compared to <1% net loss in low carbonate BPW systems. Steady-state release of dissolved uranium was not observed with high carbonate solutions and post-reaction characterizations indicated a lack of development of leached or altered surfaces. Dissolution rates (normalized to specific surface area) were about 2.5-3 orders-of-magnitude faster in high versus low carbonate BPW systems, with Na-compreignacite dissolving more rapidly than K-compreignacite under both BPW conditions, possibly due to greater ion exchange (1.57·10 -10 vs. 1.28·10 -13 mol m -2 s -1 [log R = -9.81 and -12.89] and 5.79·10 -10 vs. 3.71·10 -13 mol m -2 s -1 [log R = -9.24 and -12.43] for K- and Na-compreignacite respectively). Experimental and spectroscopic results suggest that the dissolution rate is controlled by bond breaking of a uranyl group and detachment from polyhedral layers of the mineral structure. With higher dissolved carbonate concentrations, this rate-determining step is accelerated by the formation of Ca-uranyl carbonate complexes (dominant species under these conditions), which resulted in an increase of the dissolution rates. Optimization of both dissolution rate and mineral volume fraction in the reactive transport model to account for uranium mass removal during dissolution more accurately reproduced effluent data in high carbonate systems, and resulted in faster overall rates compared with a steady-state dissolution assumption. Finally, this study highlights the importance of coupling reaction and transport processes during the quantification of mineral dissolution rates to accurately predict the fate of contaminants such as uranium in porous geomedia.« less

Rates and mechanisms of uranyl oxyhydroxide mineral dissolution

DOE PAGES

Reinoso-Maset, Estela; Steefel, Carl I.; Um, Wooyong; ...

2017-06-01

Uranyl oxyhydroxide minerals are important weathering products in uranium-contaminated surface and subsurface environments that regulate dissolved uranium concentrations. However, dissolution rates for this class of minerals and associated dissolution mechanisms have not been previously reported for circumneutral pH conditions, particularly for the case of flow through porous media. In this paper, the dissolution rates of K- and Na-compreignacite (K 2(UO 2) 6O 4(OH) 6·8H 2O and Na 2(UO 2) 6O 4(OH) 6·8H 2O respectively) were measured using flow-through columns reacted with two simulated background porewater (BPW) solutions of low and high dissolved total carbonate content (ca. 0.2 and 2.8 mmolmore » L -1). Column materials were characterized before and after reaction with electron microscopy, bulk chemistry, and EXAFS to identify structural and chemical changes during dissolution and to obtain insight into molecular-scale processes. The reactive transport code CrunchFlow was used to calculate overall dissolution rates while accounting for fluid transport and changes in mineral volume and reactive surface area and results were compared to steady-state dissolution rate calculations. In low carbonate BPW systems, interlayer K and Na were initially leached from both minerals, and in Na-compreignacite, K and minor divalent cations from the input solution were incorporated into the mineral structure. Results of characterization analyses suggested that after reaction both K- and Na-compreignacite resembled a disordered K-compreignacite with altered surfaces. A 10-fold increase in dissolved carbonate concentration and corresponding increase in pH (from 6.65 to 8.40) resulted in a net removal of 58-87% of total uranium mass from the columns, compared to <1% net loss in low carbonate BPW systems. Steady-state release of dissolved uranium was not observed with high carbonate solutions and post-reaction characterizations indicated a lack of development of leached or altered surfaces. Dissolution rates (normalized to specific surface area) were about 2.5-3 orders-of-magnitude faster in high versus low carbonate BPW systems, with Na-compreignacite dissolving more rapidly than K-compreignacite under both BPW conditions, possibly due to greater ion exchange (1.57·10 -10 vs. 1.28·10 -13 mol m -2 s -1 [log R = -9.81 and -12.89] and 5.79·10 -10 vs. 3.71·10 -13 mol m -2 s -1 [log R = -9.24 and -12.43] for K- and Na-compreignacite respectively). Experimental and spectroscopic results suggest that the dissolution rate is controlled by bond breaking of a uranyl group and detachment from polyhedral layers of the mineral structure. With higher dissolved carbonate concentrations, this rate-determining step is accelerated by the formation of Ca-uranyl carbonate complexes (dominant species under these conditions), which resulted in an increase of the dissolution rates. Optimization of both dissolution rate and mineral volume fraction in the reactive transport model to account for uranium mass removal during dissolution more accurately reproduced effluent data in high carbonate systems, and resulted in faster overall rates compared with a steady-state dissolution assumption. Finally, this study highlights the importance of coupling reaction and transport processes during the quantification of mineral dissolution rates to accurately predict the fate of contaminants such as uranium in porous geomedia.« less

Measuring Aerosols Generated Inside Armoured Vehicles Perforated by Depleted Uranium Ammunition

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parkhurst, MaryAnn

2003-01-01

In response to questions raised after the Gulf War about the health significance of exposure to depleted uranium (DU), the U.S. Department of Defense initiated a study designed to provide an improved scientific basis for assessment of possible health effects of soldiers in vehicles struck by these munitions. As part of this study, a series of DU penetrators were fired at an Abrams tank and a Bradley fighting vehicle, and the aerosols generated by vehicle perforation were collected and characterized. A robust sampling system was designed to collect aerosols in this difficult environment and to monitor continuously the sampler flowmore » rates. Interior aerosols collected were analyzed for uranium concentration and particle size distribution as a function of time. They were also analyzed for uranium oxide phases, particle morphology, and dissolution in vitro. These data will provide input for future prospective and retrospective dose and health risk assessments of inhaled or ingested DU aerosols. This paper briefly discusses the target vehicles, firing trajectories, aerosol samplers and instrumentation control systems, and the types of analyses conducted on the samples.« less

Formation and Geological Sequestration of Uranium Nanoparticles in Deep Granitic Aquifer

PubMed Central

Suzuki, Yohey; Mukai, Hiroki; Ishimura, Toyoho; Yokoyama, Takaomi D.; Sakata, Shuhei; Hirata, Takafumi; Iwatsuki, Teruki; Mizuno, Takashi

2016-01-01

The stimulation of bacterial activities that convert hexavalent uranium, U(VI), to tetravalent uranium, U(IV), appears to be feasible for cost-effective remediation of contaminated aquifers. However, U(VI) reduction typically results in the precipitation of U(IV) particles less than 5 nanometers in diameter, except for environmental conditions enriched with iron. Because these tiny particles are mobile and susceptible to oxidative dissolution after the termination of nutrient injection, in situ bioremediation remains to be impractical. Here we show that U(IV) nanoparticles of coffinite (U(SiO4)1−x(OH)4x) formed in fracture-filling calcium carbonate in a granitic aquifer. In situ U-Pb isotope dating demonstrates that U(IV) nanoparticles have been sequestered in the calcium carbonate for at least 1 million years. As the microbiologically induced precipitation of calcium carbonate in aquifer systems worldwide is extremely common, we anticipate simultaneous stimulation of microbial activities for precipitation reactions of calcium carbonate and U(IV) nanoparticles, which leads to long-term sequestration of uranium and other radionuclides in contaminated aquifers and deep geological repositories. PMID:26948389

Formation and Geological Sequestration of Uranium Nanoparticles in Deep Granitic Aquifer.

PubMed

Suzuki, Yohey; Mukai, Hiroki; Ishimura, Toyoho; Yokoyama, Takaomi D; Sakata, Shuhei; Hirata, Takafumi; Iwatsuki, Teruki; Mizuno, Takashi

2016-03-07

The stimulation of bacterial activities that convert hexavalent uranium, U(VI), to tetravalent uranium, U(IV), appears to be feasible for cost-effective remediation of contaminated aquifers. However, U(VI) reduction typically results in the precipitation of U(IV) particles less than 5 nanometers in diameter, except for environmental conditions enriched with iron. Because these tiny particles are mobile and susceptible to oxidative dissolution after the termination of nutrient injection, in situ bioremediation remains to be impractical. Here we show that U(IV) nanoparticles of coffinite (U(SiO4)1-x(OH)4x) formed in fracture-filling calcium carbonate in a granitic aquifer. In situ U-Pb isotope dating demonstrates that U(IV) nanoparticles have been sequestered in the calcium carbonate for at least 1 million years. As the microbiologically induced precipitation of calcium carbonate in aquifer systems worldwide is extremely common, we anticipate simultaneous stimulation of microbial activities for precipitation reactions of calcium carbonate and U(IV) nanoparticles, which leads to long-term sequestration of uranium and other radionuclides in contaminated aquifers and deep geological repositories.

Oxidizing dissolution mechanism of an irradiated MOX fuel in underwater aerated conditions at slightly acidic pH

NASA Astrophysics Data System (ADS)

Magnin, M.; Jégou, C.; Caraballo, R.; Broudic, V.; Tribet, M.; Peuget, S.; Talip, Z.

2015-07-01

The (U,Pu)O2 matrix behavior of an irradiated MIMAS-type (MIcronized MASter blend) MOX fuel, under radiolytic oxidation in aerated pure water at pH 5-5.5 was studied by combining chemical and radiochemical analyses of the alteration solution with Raman spectroscopy characterizations of the surface state. Two leaching experiments were performed on segments of irradiated fuel under different conditions: with or without an external γ irradiation field, over long periods (222 and 604 days, respectively). The gamma irradiation field was intended to be representative of the irradiation conditions for a fuel assembly in an underwater interim storage situation. The data acquired enabled an alteration mechanism to be established, characterized by uranium (UO22+) release mainly controlled by solubility of studtite over the long-term. The massive precipitation of this phase was observed for the two experiments based on high uranium oversaturation indexes of the solution and the kinetics involved depended on the irradiation conditions. External gamma irradiation accelerated the precipitation kinetics and the uranium concentrations (2.9 × 10-7 mol/l) were lower than for the non-irradiated reference experiment (1.4 × 10-5 mol/l), as the quantity of hydrogen peroxide was higher. Under slightly acidic pH conditions, the formation of an oxidized UO2+x phase was not observed on the surface and did not occur in the radiolysis dissolution mechanism of the fuel matrix. The Raman spectroscopy performed on the heterogeneous MOX fuel matrix surface, showed that the fluorite structure of the mainly UO2 phase surrounding the Pu-enriched aggregates had not been particularly impacted by any major structural change compared to the data obtained prior to leaching. For the plutonium, its behavior in solution involved a continuous release up to concentrations of approximately 3 × 10-6 mol L-1 with negligible colloid formation. This data appears to support a predominance of the +V oxidation state for plutonium in solution under highly oxidizing conditions. Furthermore, the Raman spectroscopy monitoring of the sample surface oxidation states did not point to any significant effect from the high Pu content of the aggregates (10-15%) and therefore did not indicate a better aggregate stability under radiolysis compared to the mainly UO2 matrix. This is because acidic pH conditions do not favor the development of oxidized layers on a fuel surface, with the exception of secondary phases.

DISSOLUTION OF URANIUM FUELS BY MONOOR DIFLUOROPHOSPHORIC ACID

DOEpatents

Johnson, R.; Horn, F.L.; Strickland, G.

1963-05-01

A method of dissolving and separating uranium from a uranium matrix fuel element by dissolving the uraniumcontaining matrix in monofluorophosphoric acid and/or difluorophosphoric acid at temperatures ranging from 150 to 275 un. Concent 85% C, thereafter neutralizing the solution to precipitate uranium solids, and converting the solids to uranium hexafluoride by treatment with a halogen trifluoride is presented. (AEC)

Uranium comminution age tested by the eolian deposits on the Chinese Loess Plateau

NASA Astrophysics Data System (ADS)

Li, Le; Liu, Xiangjun; Li, Tao; Li, Laifeng; Zhao, Liang; Ji, Junfeng; Chen, Jun; Li, Gaojun

2017-06-01

The 234U/238U ratio of fine particles can record the time since their separation from bed rock because of the disruption of uranium series equilibrium introduced by the recoil of daughter 234Th nuclei (precursor of 234U) out of particle surfaces during the decay of 238U. Application of the uranium comminution age method, which has great potential in tracing production and transportation of sediments is however complicated by the weathering dissolution of 234U depleted particle surfaces, the difficulty in determining the fraction of recoiled nuclei, and the precipitation of exogenetic 234U. Here we minimize these complications by using a newly developed precise size separation using electroformed sieve, and a chemical protocol that involves reductive and oxidative leaching. Eolian deposits collected from the Chinese Loess Plateau (CLP) were used to test the validity of our method. Possible effects of weathering dissolution were also evaluated by comparing samples with different weathering intensities. The results show decreasing 234U/238U ratios in fine eolian particles with increasing sedimentation age, agreeing well with the theoretical prediction of the comminution age model. This successful application of the uranium comminution age approach to the eolian deposits on the CLP is also aided by a stable dust source, the low weathering intensity, the lack of consolidation, and the well-defined age model of the deposits. A transportation time of 242 ± 18 ka was calculated for the eolian deposits, which indicates a long residence time, and thus extensive mixing, of the dust particles in source regions, partly explaining the stable and homogeneous composition of the eolian dust over glacial-interglacial cycles.

Uranium Anodic Dissolution under Slightly Alkaline Conditions Progress Report Full-Scale Demonstration with DU Foil

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gelis, A.; Brown, M. A.; Wiedmeyer, S.

2014-02-18

Argonne National Laboratory (Argonne) is developing an alternative method for digesting irradiated low enriched uranium (LEU) foil targets to produce 99Mo in neutral/alkaline media. This method consists of the electrolytic dissolution of irradiated uranium foil in sodium bicarbonate solution, followed by precipitation of base-insoluble fission and activation products, and uranyl-carbonate species with CaO. The addition of CaO is vital for the effective anion exchange separation of 99MoO 4 2- from the fission products, since most of the interfering anions (e.g., CO 3 2-) are removed from the solution, while molybdate remains in solution. An anion exchange is used to retainmore » and to purify the 99Mo from the filtrate. The electrochemical dissolver has been designed and fabricated in 304 stainless-steel (SS), and tested for the dissolution of a full-size depleted uranium (DU) target, wrapped in Al foil. Future work will include testing with low-burn-up DU foil at Argonne and later with high-burn-up LEU foils at Oak Ridge National Laboratory.« less

US Transuranium and Uranium Registries case study on accidental exposure to uranium hexafluoride.

PubMed

Avtandilashvili, Maia; Puncher, Matthew; McComish, Stacey L; Tolmachev, Sergei Y

2015-03-01

The United States Transuranium and Uranium Registries' (USTUR) whole-body donor (Case 1031) was exposed to an acute inhalation of uranium hexafluoride (UF6) produced from an explosion at a uranium processing plant 65 years prior to his death. The USTUR measurements of tissue samples collected at the autopsy indicated long-term retention of inhaled slightly enriched uranium material (0.85% (235)U) in the deep lungs and thoracic lymph nodes. In the present study, the authors combined the tissue measurement results with historical bioassay data, and analysed them with International Commission on Radiological Protection (ICRP) respiratory tract models and the ICRP Publication 69 systemic model for uranium using maximum likelihood and Bayesian statistical methods. The purpose of the analysis was to estimate intakes and model parameter values that best describe the data, and evaluate their effect on dose assessment. The maximum likelihood analysis, which used the ICRP Publication 66 human respiratory tract model, resulted in a point estimate of 79 mg of uranium for the occupational intake composed of 86% soluble, type F material and 14% insoluble, type S material. For the Bayesian approach, the authors applied the Markov Chain Monte Carlo method, but this time used the revised human respiratory tract model, which is currently being used by ICRP to calculate new dose coefficients for workers. The Bayesian analysis estimated that the mean uranium intake was 160 mg, and calculated the case-specific lung dissolution parameters with their associated uncertainties. The parameters were consistent with the inhaled uranium material being predominantly soluble with a small but significant insoluble component. The 95% posterior range of the rapid dissolution fraction (the fraction of deposited material that is absorbed to blood rapidly) was 0.12 to 0.91 with a median of 0.37. The remaining fraction was absorbed slowly, with a 95% range of 0.000 22 d(-1) to 0.000 36 d(-1) and a median of 0.000 31 d(-1). The effective dose per unit intake calculated using the dissolution parameters derived from the maximum likelihood and the Bayesian analyses was higher than the current ICRP dose coefficient for type F uranium by a factor of 2 or 7, respectively; the higher value of the latter was due to use of the revised respiratory tract model. The dissolution parameter values obtained here may be more appropriate to use for radiation protection purposes when individuals are exposed to a UF6 mixture that contains an insoluble uranium component.

Electrorefining cell with parallel electrode/concentric cylinder cathode

DOEpatents

Gay, Eddie C.; Miller, William E.; Laidler, James J.

1997-01-01

A cathode-anode arrangement for use in an electrolytic cell is adapted for electrochemically refining spent nuclear fuel from a nuclear reactor and recovering purified uranium for further treatment and possible recycling as a fresh blanket or core fuel in a nuclear reactor. The arrangement includes a plurality of inner anodic dissolution baskets that are each attached to a respective support rod, are submerged in a molten lithium halide salt, and are rotationally displaced. An inner hollow cylindrical-shaped cathode is concentrically disposed about the inner anodic dissolution baskets. Concentrically disposed about the inner cathode in a spaced manner are a plurality of outer anodic dissolution baskets, while an outer hollow cylindrical-shaped is disposed about the outer anodic dissolution baskets. Uranium is transported from the anode baskets and deposited in a uniform cylindrical shape on the inner and outer cathode cylinders by rotating the anode baskets within the molten lithium halide salt. Scrapers located on each anode basket abrade and remove the spent fuel deposits on the surfaces of the inner and outer cathode cylinders, with the spent fuel falling to the bottom of the cell for removal. Cell resistance is reduced and uranium deposition rate enhanced by increasing the electrode area and reducing the anode-cathode spacing. Collection efficiency is enhanced by trapping and recovery of uranium dendrites scrapped off of the cylindrical cathodes which may be greater in number than two.

Electrorefining cell with parallel electrode/concentric cylinder cathode

DOEpatents

Gay, E.C.; Miller, W.E.; Laidler, J.J.

1997-07-22

A cathode-anode arrangement for use in an electrolytic cell is adapted for electrochemically refining spent nuclear fuel from a nuclear reactor and recovering purified uranium for further treatment and possible recycling as a fresh blanket or core fuel in a nuclear reactor. The arrangement includes a plurality of inner anodic dissolution baskets that are each attached to a respective support rod, are submerged in a molten lithium halide salt, and are rotationally displaced. An inner hollow cylindrical-shaped cathode is concentrically disposed about the inner anodic dissolution baskets. Concentrically disposed about the inner cathode in a spaced manner are a plurality of outer anodic dissolution baskets, while an outer hollow cylindrical-shaped is disposed about the outer anodic dissolution baskets. Uranium is transported from the anode baskets and deposited in a uniform cylindrical shape on the inner and outer cathode cylinders by rotating the anode baskets within the molten lithium halide salt. Scrapers located on each anode basket abrade and remove the spent fuel deposits on the surfaces of the inner and outer cathode cylinders, with the spent fuel falling to the bottom of the cell for removal. Cell resistance is reduced and uranium deposition rate enhanced by increasing the electrode area and reducing the anode-cathode spacing. Collection efficiency is enhanced by trapping and recovery of uranium dendrites scrapped off of the cylindrical cathodes which may be greater in number than two. 12 figs.

Technological pretreatment of the synchysite non-oxidized ore

NASA Astrophysics Data System (ADS)

Munkhtsetseg, B.; Burmaa, G.

2013-06-01

Mongolia has rich deposits of rare, precious, and poly-metallic ores. Nowadays, it is important to research separation of rare earth elements oxides concentrates from the ores, analyze their unique physical chemical characteristics, and purified it. Our investigation on raw materials focuses on rare earth non-oxidized ores. Main mineral in this rock sample is Synchysite (LnCa(CO3)2F. We did technological and thermal pretreatment: direct sulphurization (H2SO4), sulphurization with subsequent roasting (800°C+H2SO4), sulphurization prior to roasting (H2SO4+650°C). Sulphurization method based on dissolution of rare earth mineral into sulfuric acid (93%) according to the reaction. The amount of rare earth element oxides is almost 10 times greater (29.16%) after direct sulphurization process, almost 8 times greater (21.14%) after sulphurization with subsequent roasting, and almost 20 times greater (44.62%) after sulphurization prior to roasting process. After those technological pretreatment raw material's micro elements Thorium and Uranium contents are reduced as follows: H2SO4>800°C+H2SO4>H2SO4+650°C. These results show that cerium group rare earth elements have very good solubility in water at +2°C temperature and decreasing micro elements content uranium and thorium good pretreatment condition is prior to roasting (H2SO4+650°C) of synchysite non-oxidized ore.

METHOD OF RECOVERING TRANSURANIC ELEMENTS OF AN ATOMIC NUMBER BELOW 95

DOEpatents

Seaborg, G.T.; James, R.A.

1959-12-15

The concentration of neptanium or plutonium by two carrier precipitation steps with identical carriers but using (after dissolution of the first carrier in nitric acid) a reduced quantity of carrier for the second precipitation is discussed. Carriers suitable are uranium(IV) hypophosphate, uranium(IV) pyrophosphate, uranium(IV) oxalate, thorium oxalate, thorium citrate, thorium tartrate, thorium sulfide, and uranium(IV) sulfide.

Comparison of the Kinetic Rate Law Parameters for the Dissolution of Natural and Synthetic Autunite in the Presence of Aqueous Bicarbonate Ions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gudavalli, Ravi; Katsenovich, Yelena; Wellman, Dawn M.

2013-08-02

Bicarbonate is one of the most significant components within the uranium geochemical cycle. In aqueous solutions, bicarbonate forms strong complexes with uranium. As such, aqueous bicarbonate may significantly increase the rate of uranium release from uranium minerals. Quantifying the relationship of aqueous bicarbonate concentration to the rate of uranium release during dissolution is critical to understanding the long-term fate of uranium within the environment. Single-pass flow-through (SPTF) experiments were conducted to estimate the rate of uranium release from Na meta-autunite as a function of bicarbonate (0.0005-0.003 M) under the pH range of 6-11 and a temperature range of 5-60oC. Consistentmore » with the results of previous investigation, the rate of uranium release exhibited minimal dependency on temperature; but were strongly dependent on pH. Increasing aqueous bicarbonate concentrations afforded comparable increases in the rate of release of uranium. Most notably under low pH conditions the aqueous bicarbonate resulted in up to 370 fold increases in the rate of uranium release in relative to the rate of uranium release in the absence of bicarbonate. However, the effect of aqueous bicarbonate on the release of uranium was significantly less under higher pH conditions. It is postulated that at high pH values, surface sites are saturated with carbonate, thus the addition of more bicarbonate would have less effect on uranium release.« less

Modelling of the dissolution and reprecipitation of uranium under oxidising conditions in the zone of shallow groundwater circulation.

PubMed

Dutova, Ekaterina M; Nikitenkov, Aleksei N; Pokrovskiy, Vitaly D; Banks, David; Frengstad, Bjørn S; Parnachev, Valerii P

2017-11-01

Generic hydrochemical modelling of a grantoid-groundwater system, using the Russian software "HydroGeo", has been carried out with an emphasis on simulating the accumulation of uranium in the aqueous phase. The baseline model run simulates shallow granitoid aquifers (U content 5 ppm) under conditions broadly representative of southern Norway and southwestern Siberia: i.e. temperature 10 °C, equilibrated with a soil gas partial CO 2 pressure (P CO2 , open system) of 10 -2.5 atm. and a mildly oxidising redox environment (Eh = +50 mV). Modelling indicates that aqueous uranium accumulates in parallel with total dissolved solids (or groundwater mineralisation M - regarded as an indicator of degree of hydrochemical evolution), accumulating most rapidly when M = 550-1000 mg L -1 . Accumulation slows at the onset of saturation and precipitation of secondary uranium minerals at M = c. 1000 mg L -1 (which, under baseline modelling conditions, also corresponds approximately to calcite saturation and transition to Na-HCO 3 hydrofacies). The secondary minerals are typically "black" uranium oxides of mixed oxidation state (e.g. U 3 O 7 and U 4 O 9 ). For rock U content of 5-50 ppm, it is possible to generate a wide variety of aqueous uranium concentrations, up to a maximum of just over 1 mg L -1 , but with typical concentrations of up to 10 μg L -1 for modest degrees of hydrochemical maturity (as indicated by M). These observations correspond extremely well with real groundwater analyses from the Altai-Sayan region of Russia and Norwegian crystalline bedrock aquifers. The timing (with respect to M) and degree of aqueous uranium accumulation are also sensitive to Eh (greater mobilisation at higher Eh), uranium content of rocks (aqueous concentration increases as rock content increases) and P CO2 (low P CO2 favours higher pH, rapid accumulation of aqueous U and earlier saturation with respect to uranium minerals). Copyright © 2017 Elsevier Ltd. All rights reserved.

The effect of ion irradiation on the dissolution of UO 2 and UO 2 -based simulant fuel

DOE PAGES

Popel, Aleksej J.; Wietsma, Thomas W.; Engelhard, Mark H.; ...

2017-11-21

Our aim is to study the separate effect of fission fragment damage on the dissolution of simulant UK advanced gas-cooled reactor nuclear fuel in water. Plain UO 2 and UO 2 samples, doped with inactive fission products to simulate 43 GWd/tU of burn-up, were fabricated. A set of these samples were then irradiated with 92 MeV 129Xe 23+ ions to a fluence of 4.8 × 10 15 ions/cm 2 to simulate the fission damage that occurs within nuclear fuels. The primary effect of the irradiation on the UO 2 samples, observed by scanning electron microscopy, was to induce a smootheningmore » of the surface features and formation of hollow blisters, which was attributed to multiple overlap of ion tracks. Dissolution experiments were conducted in single-pass flow-through (SPFT) mode under anoxic conditions (<0.1 O 2 ppm in Ar) to study the effect of the induced irradiation damage on the dissolution of the UO 2 matrix with data collection capturing six minute intervals for several hours. These time-resolved data showed that the irradiated samples showed a higher initial release of uranium than unirradiated samples, but that the uranium concentrations converged towards ~10 -9 mol/l after a few hours. And apart from the initial spike in uranium concentration, attributed to irradiation induced surficial micro-structural changes, no noticeable difference in uranium chemistry as measured by X-ray electron spectroscopy or ‘effective solubility’ was observed between the irradiated, doped and undoped samples in this work. Some secondary phase formation was observed on the surface of UO 2 samples after the dissolution experiment.« less

The effect of ion irradiation on the dissolution of UO 2 and UO 2 -based simulant fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Popel, Aleksej J.; Wietsma, Thomas W.; Engelhard, Mark H.

Our aim is to study the separate effect of fission fragment damage on the dissolution of simulant UK advanced gas-cooled reactor nuclear fuel in water. Plain UO 2 and UO 2 samples, doped with inactive fission products to simulate 43 GWd/tU of burn-up, were fabricated. A set of these samples were then irradiated with 92 MeV 129Xe 23+ ions to a fluence of 4.8 × 10 15 ions/cm 2 to simulate the fission damage that occurs within nuclear fuels. The primary effect of the irradiation on the UO 2 samples, observed by scanning electron microscopy, was to induce a smootheningmore » of the surface features and formation of hollow blisters, which was attributed to multiple overlap of ion tracks. Dissolution experiments were conducted in single-pass flow-through (SPFT) mode under anoxic conditions (<0.1 O 2 ppm in Ar) to study the effect of the induced irradiation damage on the dissolution of the UO 2 matrix with data collection capturing six minute intervals for several hours. These time-resolved data showed that the irradiated samples showed a higher initial release of uranium than unirradiated samples, but that the uranium concentrations converged towards ~10 -9 mol/l after a few hours. And apart from the initial spike in uranium concentration, attributed to irradiation induced surficial micro-structural changes, no noticeable difference in uranium chemistry as measured by X-ray electron spectroscopy or ‘effective solubility’ was observed between the irradiated, doped and undoped samples in this work. Some secondary phase formation was observed on the surface of UO 2 samples after the dissolution experiment.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

WANG,YIFENG; XU,HUIFANG

Correctly identifying the possible alteration products and accurately predicting their occurrence in a repository-relevant environment are the key for the source-term calculation in a repository performance assessment. Uraninite in uranium deposits has long been used as a natural analog to spent fuel in a repository because of their chemical and structural similarity. In this paper, a SEM/AEM investigation has been conducted on a partially alternated uraninite sample from a uranium ore deposit of Shinkolobwe of Congo. The mineral formation sequences were identified: uraninite {yields} uranyl hydrates {yields} uranyl silicates {yields} Ca-uranyl silicates or uraninite {yields} uranyl silicates {yields} Ca-uranyl silicates.more » Reaction-path calculations were conducted for the oxidative dissolution of spent fuel in a representative Yucca Mountain groundwater. The predicted sequence is in general consistent with the SEM observations. The calculations also show that uranium carbonate minerals are unlikely to become major solubility-controlling mineral phases in a Yucca Mountain environment. Some discrepancies between model predictions and field observations are observed. Those discrepancies may result from poorly constrained thermodynamic data for uranyl silicate minerals.« less

SELECTIVE SEPARATION OF URANIUM FROM THORIUM, PROTACTINIUM AND FISSION PRODUCTS BY PEROXIDE DISSOLUTION METHOD

DOEpatents

Seaborg, G.T.; Gofman, J.W.; Stoughton, R.W.

1959-08-18

A method is described for separating U/sup 233/ from thorium and fission products. The separation is effected by forming a thorium-nitric acid solution of about 3 pH, adding hydrogen peroxide to precipitate uranium and thorium peroxide, treating the peroxides with sodium hydroxide to selectively precipitate the uranium peroxide, and reacting the separated solution with nitric acid to re- precipitate the uranium peroxide.

ELECTROCHEMICAL DECONTAMINATION AND RECOVERY OF URANIUM VALUES

DOEpatents

McLaren, J.A.; Goode, J.H.

1958-05-13

An electrochemical process is described for separating uranium from fission products. The method comprises subjecting the mass of uranium to anodic dissolution in an electrolytic cell containing aqueous alkali bicarbonate solution as its electrolyte, thereby promoting a settling from the solution of a solid sludge from about the electrodes and separating the resulting electrolyte solution containing the anodically dissolved uranium from the sludge which contains the rare earth fission products.

REMOVAL OF URANIUM FROM ORGANIC LIQUIDS

DOEpatents

Vavalides, S.P.

1959-08-25

A process is described for recovering small quantities of uranium from organic liquids such as hydrocarbon oils. halogen-substituted hydrocarbons, and alcohols. The organic liquid is contacted with a comminuted alkaline earth hydroxide, calcium hydroxide particularly, and the resulting uranium-bearing solid is separated from the liquid by filtration. Uranium may then be recovered from the solid by means of dissolution in nitric acid and conventional extraction with an organic solvent such as tributyl phosphate.

Uranium in granitic magmas: Part 1. Experimental determination of uranium solubility and fluid-melt partition coefficients in the uranium oxide-haplogranite-H 2O-Na 2CO 3 system at 720-770°C, 2 kbar

NASA Astrophysics Data System (ADS)

Peiffert, Chantal; Cuney, Michel; Nguyen-Trung, Chinh

1994-06-01

The solubility of uranium was investigated in both carbonated aqueous fluid and granitic melt in equilibrium in the system haplogranite-uranium oxide-H 2O-Na 2CO 3 (0.5-1 molal) at 720-770°C, 2 kbar, andƒo 2 fixed by Ni-NiO, Fe 3O 4-Fe 2O 3, and Cu 2O-CuO buffers. As complete solid solution exists between UO 2.00 and UO 2.25 (i.e., 75 mol% UO 2 + 25 mol% UO 3), three distinct uranium oxides: UO (2.01 ± 0.01), UO (2.1.0 ± 0.02), and UO (2.25 ± 0.02) were, respectively, obtained at equilibrium, under the three ƒo 2 conditions cited above. Thus, the percentage of U (VI) in uranium oxide increased with increasing log ƒo 2. The thermal decomposition of Na 2CO 3 to CO 2 and Na 2O led to the decrease of the sodium carbonate concentration from 0.5-1 molal to ~10 -2 molal in all aqueous fluids and to the dissolution of Na in the silicate melts. Crystal-free silicate glasses with four agpaitic coefficients, α = ( (Na+K)/Al) = 1.1, 1.3, 1.5, and 1.7 were obtained. The uranium solubility in 10 -2 m aqueous carbonated fluid ((8.1 ± 0.1) ≤ quench pH ≤ (8.9 ± 0.1)) was in the range 1-17 ppm and increased linearly with increasing ƒo 2 according to the expression: log (U) (ppm) = 0.09 ·log ƒo 2 (bar) + 1.47 . This equation is valid for the temperature range 720-770°C and 2 kbar. U(IV) carbonate possibly were major species in aqueous solutions under reducing conditions (Ni-NiO buffer) whereas U(VI) carbonate complexes dominated under higher oxidation conditions (Fe 3O 4-Fe 2O 3, Cu 2O-CuO buffers). The uranium content in silicate glasses varied in a large range (10 2-2 × 10 5 ppm) and log (U) (ppm) increases linearly with both ƒo 2, and α in the range 1.1-1.5 according to the equation log (U) (ppm) = 0.04 log ƒo 2 (bar) + 3.80α -1.34 . This equation is valid for (1)ƒ o 2 ranging from Ni-NiO to Cu 2O-CuO, and (2) the temperature range 720-770°C at 2 kbar. The effect of ƒo 2 on the uranium solubility in silicate melt slightly decreased with increasing α from 1.1 to 1.5. For α in the range 1.5-1.7, the effect of both ƒo 2 and agpaicity index on the uranium solubility was considerably reduced. The temperature variation in the range 720-770°C had no significant effect on the uranium solubility in either aqueous fluid or silicate melt. The partition coefficient (D fluid/melt) of uranium was in the range 10 -4.0-10 -1.5 and depended on both ƒo 2 and α according to the equation log D fluid/melt = 0.05 log ƒo 2 (bar) - 3.78α + 2.84 . The validity conditions of this equation are similar to those of the preceding one. Results obtained in the present study could be used to predict the geochemical behaviour of uranium during magma fractionation and to further understanding of the formation of uranium ore deposits related to partial melting or fractional crystallization of felsic magmas. The genesis of the Kvanefjeld (Ilimaussaq, Greenland) uranium deposit is discussed.

Method for converting uranium oxides to uranium metal

DOEpatents

Duerksen, Walter K.

1988-01-01

A process is described for converting scrap and waste uranium oxide to uranium metal. The uranium oxide is sequentially reduced with a suitable reducing agent to a mixture of uranium metal and oxide products. The uranium metal is then converted to uranium hydride and the uranium hydride-containing mixture is then cooled to a temperature less than -100.degree. C. in an inert liquid which renders the uranium hydride ferromagnetic. The uranium hydride is then magnetically separated from the cooled mixture. The separated uranium hydride is readily converted to uranium metal by heating in an inert atmosphere. This process is environmentally acceptable and eliminates the use of hydrogen fluoride as well as the explosive conditions encountered in the previously employed bomb-reduction processes utilized for converting uranium oxides to uranium metal.

Estimation of the reactive mineral surface area during CO2-rich fluid-rock interaction: the influence of neogenic phases

NASA Astrophysics Data System (ADS)

Scislewski, A.; Zuddas, P.

2010-12-01

Mineral dissolution and precipitation reactions actively participate to control fluid chemistry during water-rock interaction. It is however, difficult to estimate and well normalize bulk reaction rates if the mineral surface area exposed to the aqueous solution and effectively participating on the reactions is unknown. We evaluated the changing of the reactive mineral surface area during the interaction between CO2-rich fluids and Albitite/Granitoid rocks (similar mineralogy but different abundances), reacting under flow-through conditions. Our methodology, adopting an inverse modeling approach, is based on the estimation of dissolution rate and reactive surface area of the different minerals participating in the reactions by the reconstruction the chemical evolution of the interacting fluids. The irreversible mass-transfer processes is defined by a fractional degree of advancement, while calculations were carried out for Albite, Microcline, Biotite and Calcite assuming that the ion activity of dissolved silica and aluminium ions was limited by the equilibrium with quartz and kaolinite. Irrespective of the mineral abundance in granite and albitite, we found that mineral dissolution rates did not change significantly in the investigated range of time where output solution’s pH remained in the range between 6 and 8, indicating that the observed variation in fluid composition depends not on pH but rather on the variation of the parent mineral’s reactive surface area. We found that the reactive surface area of Albite varied by more than 2 orders of magnitude, while Microcline, Calcite and Biotite surface areas changed by 1-2 orders of magnitude. We propose that parent mineral chemical heterogeneity and, particularly, the stability of secondary mineral phases may explain the observed variation of the reactive surface area of the minerals. Formation of coatings at the dissolving parent mineral surfaces significantly reduced the amount of surface available to react with CO2-rich fluids, decreasing the effective reactive surface area. Predictive models of CO2 sequestration under geological conditions should take into account the inhibiting role of surface coating formation. The CO2 rich fluid-rock interactions may also have significant consequences on metal mobilization. Our results indicated that the formation of stable carbonate complexes enhances the solubility of uranium minerals of both albitite and granite, facilitating the U(IV) oxidation, and limiting the extent of uranium adsorption onto particles in oxidized waters. This clearly produces an increase of the uranium mobility with significant consequences for the environment.

Quantification of Kinetic Rate Law Parameters of Uranium Release from Sodium Autunite as a Function of Aqueous Bicarbonate Concentrations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gudavalli, Ravi; Katsenovich, Yelena; Wellman, Dawn M.

2013-09-05

ABSTRACT: Hydrogen carbonate is one of the most significant components within the uranium geochemical cycle. In aqueous solutions, hydrogen carbonate forms strong complexes with uranium. As such, aqueous bicarbonate may significantly increase the rate of uranium release from uranium minerals. Quantifying the relationship of aqueous hydrogen carbonate solutions to the rate of uranium release during dissolution is critical to understanding the long-term fate of uranium within the environment. Single-pass flow-through (SPTF) experiments were conducted to estimate the rate of uranium release from Na meta-autunite as a function of bicarbonate solutions (0.0005-0.003 M) under the pH range of 6-11 and temperaturesmore » of 5-60oC. Consistent with the results of previous investigation, the rate of uranium release from sodium autunite exhibited minimal dependency on temperature; but were strongly dependent on pH and increasing concentrations of bicarbonate solutions. Most notably at pH 7, the rate of uranium release exhibited 370 fold increases relative to the rate of uranium release in the absence of bicarbonate. However, the effect of increasing concentrations of bicarbonate solutions on the release of uranium was significantly less under higher pH conditions. It is postulated that at high pH values, surface sites are saturated with carbonate, thus the addition of more bicarbonate would have less effect on uranium release. Results indicate the activation energies were unaffected by temperature and bicarbonate concentration variations, but were strongly dependent on pH conditions. As pH increased from 6 to 11, activation energy values were observed to decrease from 29.94 kJ mol-1 to 13.07 kJ mol-1. The calculated activation energies suggest a surface controlled dissolution mechanism.« less

PROCESS OF PRODUCING REFRACTORY URANIUM OXIDE ARTICLES

DOEpatents

Hamilton, N.E.

1957-12-01

A method is presented for fabricating uranium oxide into a shaped refractory article by introducing a uranium halide fluxing reagent into the uranium oxide, and then mixing and compressing the materials into a shaped composite mass. The shaped mass of uranium oxide and uranium halide is then fired at an elevated temperature so as to form a refractory sintered article. It was found in the present invention that the introduction of a uraninm halide fluxing agent afforded a fluxing action with the uranium oxide particles and that excellent cohesion between these oxide particles was obtained. Approximately 90% of uranium dioxide and 10% of uranium tetrafluoride represent a preferred composition.

ANODIC TREATMENT OF URANIUM

DOEpatents

Kolodney, M.

1959-02-01

A method is presented for effecting eloctrolytic dissolution of a metallic uranium article at a uniform rate. The uranium is made the anode in an aqueous phosphoric acid solution containing nitrate ions furnished by either ammonium nitrate, lithium nitrate, sodium nitrate, or potassium nitrate. A stainless steel cathode is employed and electrolysls carried out at a current density of about 0.1 to 1 ampere per square inch.

The separation of lanthanides and actinides in supercritical fluid carbon dioxide

DOE PAGES

Mincher, Bruce J.; Wai, Chien M.; Fox, Robert V.; ...

2015-10-28

Supercritical fluid carbon dioxide presents an attractive alternative to conventional solvents for recovery of the actinides and lanthanides. Carbon dioxide is a good solvent for fluorine and phosphate-containing ligands, including the traditional tributylphosphate ligand used in process-scale uranium separations. Actinide and lanthanide oxides may even be directly dissolved in carbon dioxide containing the complexes formed between these ligands and mineral acids, obviating the need for large volumes of acids for leaching and dissolution, and the corresponding organic liquid–liquid solvent extraction solutions. As a result, examples of the application of this novel technology for actinide and lanthanide separations are presented.

Diel Sampling of Groundwater and Surface Water for Trace Elements and Select Water-Quality Constituents at a Former Zinc Smelter Site near Hegeler, Illinois, August 1-3, 2007

USGS Publications Warehouse

Kay, Robert T.; Groschen, George E.; Dupre, David H.; Drexler, Timothy D.; Thingvold, Karen L.; Rosenfeld, Heather J.

2009-01-01

Surface water can exhibit substantial diel variations in the concentration of a number of constituents. Sampling regimens that do not characterize diel variations in water quality can result in an inaccurate understanding of site conditions and of the threat posed by the site to human health and the environment. Surface- and groundwater affected by acid drainage were sampled every 60 to 90 minutes over a 48-hour period at a former zinc smelter known as the Hegeler Zinc Superfund Site, in Hegeler, Ill. Groundwater-quality data from a well at the site indicate stable, low pH, weakly oxidizing geochemical conditions in the aquifer. With the exceptions of temperature and pH, no constituents exhibited diel variations in groundwater. Variations in temperature and pH likely were not representative of conditions in the aquifer. Surface water was sampled at a site on Grape Creek. Diel variations were observed in temperature, dissolved oxygen, pH, and specific conductance, and in the concentrations of nitrite, barium, iron, lead, vanadium, and possibly uranium. Concentrations during the diel cycles varied by about an order of magnitude for nitrite and varied by about a factor of two for barium, iron, lead, vanadium, and uranium. Temperature, dissolved oxygen, specific conductance, nitrite, barium, lead, and uranium generally reached maximum values during the afternoon and minimum values during the night. Iron, vanadium, and pH generally reached minimum values during the afternoon and maximum values during the night. These variations would need to be accounted for during sampling of surface-water quality in similar hydrologic settings. The temperature variations in surface water were affected by variations in air temperature. Concentrations of dissolved oxygen were affected by variations in the intensity of photosynthetic activity and respiration. Nitrite likely was formed by the oxidation of ammonium by dissolved oxygen and degraded by its anaerobic oxidation by ammonium or as part of the decomposition of organic matter. Variations in pH were affected by the photoreduction of Fe3+ to Fe2+ and the precipitation of iron oxyhydroxides. Diel variations in concentrations of iron and vanadium were likely caused by variations in the dissolution and precipitation of iron oxyhydroxides, oxyhydroxysulfates, and hydrous sulfates, which may have been affected by in the intensity of insolation, iron photoreduction, and the concentration of dissolved oxygen. The concentrations of lead, uranium, and perhaps barium in Grape Creek may have been affected by competition for sorption sites on iron oxyhydroxides. Competition for sorption sites was likely affected by variations in pH and the concentration of Fe2+. Constituent concentrations likely also were affected by precipitation and dissolution of minerals that are sensitive to changes in pH, temperature, oxidation-reduction conditions, and biologic activity. The chemical and biologic processes that resulted in the diel variations observed in Grape Creek occurred within the surface-water column or in the underlying sediments.

Ground-water quality and geochemistry, Carson Desert, western Nevada

USGS Publications Warehouse

Lico, Michael S.; Seiler, R.L.

1994-01-01

Aquifers in the Carson Desert are the primary source of drinking water, which is highly variable in chemical composition. In the shallow basin-fill aquifers, water chemistyr varies from a dilute calcium bicarbonate-dominated water beneath the irrigated areas to a saline sodium chloride- dominated water beneath unirrigated areas. Water samples from the shallow aquifers commonly have dissolved solids, chloride, magnesium, sulfate, arsenic, and manganese concentrations that exceed State of Nevada drinking-water standards. Water in the intermediante basin-fill aquifers is a dilute sodium bicarbonate type in the Fallon area and a distinctly more saline sodium chloride type in the Soda Lake-Upsal Hogback area. Dissolved solids, chloride, arsenic, fluoride, and manganese concen- trations commonly exceed drinking-water standards. The basalt aquifer contains a dilute sodium bicarbonate chloride water. Arsenic concentrations exceed standards in all sampled wells. The concen- trations of major constituents in ground water beneath the southern Carson Desert are the result of evapotranspiration and natural geochemical reactions with minerals derived mostly from igneous rocks. Water with higher concentrations of iron and manganese is near thermodynamic equilibrium with siderite and rhodochrosite and indicates that these elements may be limited by the solubility of their respective carbonate minerals. Naturally occurring radionuclides (uranium and radon-222) are present in ground water from the Carson Desert in concen- tratons higher than proposed drinking-water standards. High uranium concentrations in the shallow aquifers may be caused by evaporative concentration and the release of uranium during dissolution of iron and manganese oxides or the oxidation of sedimentary organic matter that typically has elevated uranium concentrations. Ground water in the Carson Desert does not appear to have be contaminated by synthetic organic chemicals.

Microbial mobilization of plutonium and other actinides from contaminated soil

DOE PAGES

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

2015-12-01

Here we examined the dissolution of Pu, U, and Am in contaminated soil from the Nevada Test Site (NTS) due to indigenous microbial activity. Scanning transmission x-ray microscopy (STXM) analysis of the soil showed that Pu was present in its polymeric form and associated with Fe- and Mn- oxides and aluminosilicates. Uranium analysis by x-ray diffraction (μ-XRD) revealed discrete U-containing mineral phases, viz., schoepite, sharpite, and liebigite; synchrotron x-ray fluorescence (μ-XRF) mapping showed its association with Fe- and Ca-phases; and μ-x-ray absorption near edge structure (μ-XANES) confirmed U(IV) and U(VI) oxidation states. Addition of citric acid or glucose to themore » soil and incubated under aerobic or anaerobic conditions enhanced indigenous microbial activity and the dissolution of Pu. Detectable amount of Am and no U was observed in solution. In the citric acid-amended sample, Pu concentration increased with time and decreased to below detection levels when the citric acid was completely consumed. In contrast, with glucose amendment, Pu remained in solution. Pu speciation studies suggest that it exists in mixed oxidation states (III/IV) in a polymeric form as colloids. Although Pu(IV) is the most prevalent and generally considered to be more stable chemical form in the environment, our findings suggest that under the appropriate conditions, microbial activity could affect its solubility and long-term stability in contaminated environments.« less

Microbial mobilization of plutonium and other actinides from contaminated soil

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

Here we examined the dissolution of Pu, U, and Am in contaminated soil from the Nevada Test Site (NTS) due to indigenous microbial activity. Scanning transmission x-ray microscopy (STXM) analysis of the soil showed that Pu was present in its polymeric form and associated with Fe- and Mn- oxides and aluminosilicates. Uranium analysis by x-ray diffraction (μ-XRD) revealed discrete U-containing mineral phases, viz., schoepite, sharpite, and liebigite; synchrotron x-ray fluorescence (μ-XRF) mapping showed its association with Fe- and Ca-phases; and μ-x-ray absorption near edge structure (μ-XANES) confirmed U(IV) and U(VI) oxidation states. Addition of citric acid or glucose to themore » soil and incubated under aerobic or anaerobic conditions enhanced indigenous microbial activity and the dissolution of Pu. Detectable amount of Am and no U was observed in solution. In the citric acid-amended sample, Pu concentration increased with time and decreased to below detection levels when the citric acid was completely consumed. In contrast, with glucose amendment, Pu remained in solution. Pu speciation studies suggest that it exists in mixed oxidation states (III/IV) in a polymeric form as colloids. Although Pu(IV) is the most prevalent and generally considered to be more stable chemical form in the environment, our findings suggest that under the appropriate conditions, microbial activity could affect its solubility and long-term stability in contaminated environments.« less

Influence of oxalic acid on the dissolution kinetics of manganese oxide

NASA Astrophysics Data System (ADS)

Godunov, E. B.; Artamonova, I. V.; Gorichev, I. G.; Lainer, Yu. A.

2012-11-01

The kinetics and electrochemical processes of the dissolution of manganese oxides with various oxidation states in sulfuric acid solutions containing oxalate ion additives is studied under variable conditions (concentration, pH, temperature). The parameters favoring a higher degree of the dissolution of manganese oxides in acidic media are determined. The optimal conditions are found for the dissolution of manganese oxides in acidic media in the presence of oxalate ions. The mechanism proposed for the dissolution of manganese oxides in sulfuric acid solutions containing oxalic acid is based on the results of kinetic and electrochemical studies. The steps of the dissolution mechanism are discussed.

Behavior of uranium under conditions of interaction of rocks and ores with subsurface water

NASA Astrophysics Data System (ADS)

Omel'Yanenko, B. I.; Petrov, V. A.; Poluektov, V. V.

2007-10-01

The behavior of uranium during interaction of subsurface water with crystalline rocks and uranium ores is considered in connection with the problem of safe underground insulation of spent nuclear fuel (SNF). Since subsurface water interacts with crystalline rocks formed at a high temperature, the mineral composition of these rocks and uranium species therein are thermodynamically unstable. Therefore, reactions directed toward the establishment of equilibrium proceed in the water-rock system. At great depths that are characterized by hindered water exchange, where subsurface water acquires near-neutral and reducing properties, the interaction is extremely sluggish and is expressed in the formation of micro- and nanoparticles of secondary minerals. Under such conditions, the slow diffusion redistribution of uranium with enrichment in absorbed forms relative to all other uranium species is realized as well. The products of secondary alteration of Fe- and Ti-bearing minerals serve as the main sorbents of uranium. The rate of alteration of minerals and conversion of uranium species into absorbed forms is slow, and the results of these processes are insignificant, so that the rocks and uranium species therein may be regarded as unaltered. Under reducing conditions, subsurface water is always saturated with uranium. Whether water interacts with rock or uranium ore, the equilibrium uranium concentration in water is only ≤10-8 mol/l. Uraninite ore under such conditions always remains stable irrespective of its age. The stability conditions of uranium ore are quite suitable for safe insulation of SNF, which consists of 95% uraninite (UO2) and is a confinement matrix for all other radionuclides. The disposal of SNF in massifs of crystalline rocks at depths below 500 m, where reducing conditions are predominant, is a reliable guarantee of high SNF stability. Under oxidizing conditions of the upper hydrodynamic zone, the rate of interaction of rocks with subsurface water increases by orders of magnitude and subsurface water is commonly undersaturated with uranium. Uranium absorbed by secondary minerals, particularly by iron hydroxides and leucoxene, is its single stable species under oxidizing conditions. The impact of oxygen-bearing water leads to destruction of uranium ore. This process is realized simultaneously at different hypsometric levels even if the permeability of the medium is variable in both the lateral and vertical directions. As a result, intervals containing uranyl minerals and relics of primary uranium ore are combined in ore-bearing zones with intervals of completely dissolved uranium minerals. A wide halo of elevated uranium contents caused by sorption is always retained at the location of uranium ore entirely destroyed by weathering. Uranium ore commonly finds itself in the aeration zone due to technogenic subsidence of the groundwater table caused by open-pit mining or pumping out of water from underground mines. The capillary and film waters that interact with rocks and ores in this zone are supplemented by free water filtering along fractures when rain falls or snow is thawing. The interaction of uranium ore with capillary water results in oxidation of uraninite, accompanied by loosening of the mineral surface, formation of microfractures, and an increase in solubility with enrichment of capillary water in uranium up to 10-4 mol/l. Secondary U(VI) minerals, first of all, uranyl hydroxides and silicates, replace uraninite, and uranium undergoes local diffusion redistribution with its sorption by secondary minerals of host rocks. The influx of free water facilitates the complete dissolution of primary and secondary uranium minerals, the removal of uranium at the sites of groundwater discharge, and its redeposition under reducing conditions at a greater depth. It is evident that the conditions of the upper hydrodynamic zone and the aeration zone are unfit for long-term insulation of SNF and high-level wastes because, after the failure of containers, the leakage of radionuclides into the environment becomes inevitable.

Potential for U sequestration with select minerals and sediments via base treatment.

PubMed

Emerson, Hilary P; Di Pietro, Silvina; Katsenovich, Yelena; Szecsody, Jim

2018-06-13

Temporary base treatment is a potential remediation technique for heavy metals through adsorption, precipitation, and co-precipitation with minerals. Manipulation of pH with ammonia gas injection may be especially useful for vadose zone environments as it does not require addition of liquids that would increase the flux towards groundwater. In this research, we conducted laboratory batch experiments to evaluate the changes in uranium mobility and mineral dissolution with base treatments including sodium hydroxide, ammonium hydroxide, and ammonia gas. Our data show that partitioning of uranium to the solid phase increases by several orders of magnitude following base treatment in the presence of different minerals and natural sediments from the Hanford site. The presence of dissolved calcium and carbonate play an important role in precipitation and co-precipitation of uranium at elevated pH. In addition, significant incongruent dissolution of bulk mineral phases occurs and likely leads to precipitation of secondary mineral phases. These secondary phases may remove uranium via adsorption, precipitation, and co-precipitation processes and may coat uranium phases with low solubility minerals as the pH returns to natural conditions. Copyright © 2018 Elsevier Ltd. All rights reserved.

Modifying Surface Chemistry of Metal Oxides for Boosting Dissolution Kinetics in Water by Liquid Cell Electron Microscopy.

PubMed

Lu, Yue; Geng, Jiguo; Wang, Kuan; Zhang, Wei; Ding, Wenqiang; Zhang, Zhenhua; Xie, Shaohua; Dai, Hongxing; Chen, Fu-Rong; Sui, Manling

2017-08-22

Dissolution of metal oxides is fundamentally important for understanding mineral evolution and micromachining oxide functional materials. In general, dissolution of metal oxides is a slow and inefficient chemical reaction. Here, by introducing oxygen deficiencies to modify the surface chemistry of oxides, we can boost the dissolution kinetics of metal oxides in water, as in situ demonstrated in a liquid environmental transmission electron microscope (LETEM). The dissolution rate constant significantly increases by 16-19 orders of magnitude, equivalent to a reduction of 0.97-1.11 eV in activation energy, as compared with the normal dissolution in acid. It is evidenced from the high-resolution TEM imaging, electron energy loss spectra, and first-principle calculations where the dissolution route of metal oxides is dynamically changed by local interoperability between altered water chemistry and surface oxygen deficiencies via electron radiolysis. This discovery inspires the development of a highly efficient electron lithography method for metal oxide films in ecofriendly water, which offers an advanced technique for nanodevice fabrication.

Measuring aerosols generated inside armoured vehicles perforated by depleted uranium ammunition.

PubMed

Parkhurst, M A

2003-01-01

In response to questions raised after the Gulf War about the health significance of exposure to depleted uranium (DU), the US Department of Defense initiated a study designed to provide an improved scientific basis for assessment of possible health effects on soldiers in vehicles struck by these munitions. As part of this study, a series of DU penetrators were fired at an Abrams tank and a Bradley fighting vehicle, and the aerosols generated by vehicle perforation were collected and characterised. A robust sampling system was designed to collect aerosols in this difficult environment and monitor continuously the sampler flow rates. The aerosol samplers selected for these tests included filter cassettes, cascade impactors, a five-stage cyclone and a moving filter. Sampler redundancy was an integral part of the sampling system to offset losses from fragment damage. Wipe surveys and deposition trays collected removable deposited particulate matter. Interior aerosols were analysed for uranium concentration and particle size distribution as a function of time. They were also analysed for uranium oxide phases, particle morphology and dissolution in vitro. These data, currently under independent peer review, will provide input for future prospective and retrospective dose and health risk assessments of inhaled or ingested DU aerosols. This paper briefly discusses the target vehicles, firing trajectories, aerosol samplers and instrumentation control systems, and the types of analyses conducted on the samples.

Dissolution Kinetics of Meta-Torbernite under Circum-neutral to Alkaline Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wellman, Dawn M.; McNamara, Bruce K.; Bacon, Diana H.

2009-12-21

Autunite group minerals have been identified in contaminated sediments as the long-term controlling phase of uranium. Meta-torbernite, has been identified in subsurface environments which were subjected to co-contaminant disposal practices from past nuclear weapons and fuel operations. Under these conditions the mobility of uranium in subsurface pore waters is limited by the rate of meta-torbernite dissolution; however, there are no known investigations which report the dissolution behavior of meta-torbernite. The purpose of this investigation was to conduct a series of single-pass flow-through (SPFT) tests to 1) quantify the effect of temperature (23 - 90oC) and pH (6 -10) on meta-torbernitemore » dissolution, 2) compare the dissolution of meta-torbernite to other autunite-group minerals, and 3) evaluate the effect of aqueous phosphate on the dissolution kinetics of meta-torbernite. Results presented here illustrate meta-torbernite dissolution rates increase by ~100X over the pH interval of 6 to 10 (eta = 0.59 ± 0.07), irrespective of temperature. The power law coefficient for meta-torbernite, eta = 0.59 ± 0.07, is greater than that quantified for Ca-meta-autunite, eta = 0.42 ± 0.12. This suggests the stability of meta-torbernite is greater than that of meta-autunite, which is reflected in the predicted stability constants. The rate equation for the dissolution of meta-torbernite as a function of aqueous phosphate concentration is log rdissol (mol m-2 sec-1) = -4.7 x 10-13 + 4.1 x 10-10 [PO43-].« less

METHOD FOR DISSOLVING ZIRCONIUM-URANIUM COMPOSITIONS

DOEpatents

Gens, T.A.

1961-07-18

A method is descrioed for treating a zirconium-- uranium composition to form a stable solution from which uranium and other values may be extracted by contacting the composition with at least a 4 molar aqueous solution of ammonium fluoride at a temperature of about 100 deg C, adding a peroxide, in incremental amounts, to the heated solution throughout the period of dissolution until all of the uranium is converted to soluble uranyl salt, adding nitric acid to the resultant solution to form a solvent extraction feed solution to convert the uranyl salt to a solvent extractable state, and thereafter recovering the uranium and other desired values from the feed solution by solvent extraction.

DECONTAMINATION OF URANIUM

DOEpatents

Feder, H.M.; Chellew, N.R.

1958-02-01

This patent deals with the separation of rare earth and other fission products from neutron bombarded uranium. This is accomplished by melting the uranium in contact with either thorium oxide, maguesium oxide, alumnum oxide, beryllium oxide, or uranium dioxide. The melting is preferably carried out at from 1150 deg to 1400 deg C in an inert atmosphere, such as argon or helium. During this treatment a scale of uranium dioxide forms on the uranium whtch contains most of the fission products.

Uranium mobility during interaction of rhyolitic glass with alkaline solutions: dissolution of glass

USGS Publications Warehouse

Zielinski, Robert A.

1977-01-01

This report concerns investigations designed to identify the important physical and chemical parameters influencing the rate of release of uranium from glass shards of rhyolitic air-fall ash. Oxidizing, silica undersaturated, alkaline solutions are eluted through a column of rhyolitic glass shards at a carefully controlled temperature, pressure, and flow rate. The solutions are monitored for the concentration of uranium and selected additional elements (Si, K, Li, F), and the glass is recovered and examined for physical and/or chemical evidence of attack. The flushing mode is designed to mimic leaching of glass shards by intermittent, near-surface waters with which the glass is not in equilibrium. Reported rates are applicable only to the experimental conditions (120?C, 7,000 psi), but it is assumed that the reaction mechanisms and the relative importance of rate-influencing parameters remain unchanged, at reduced temperature and pressure. Results of the above experiment indicate that silica and uranium are released from glass shards at comparable rates, while lithium and potassium are released faster and fluorine slower than either Si or U. Rates of release of silica and uranium correlate positively with the surface area of the shards. Rhyolitic shards release uranium at faster rates than rhyodacitic shards of comparable surface area. Changes in the shards resulting from experimental treatment and observed in the original glass separates from an Oligocene ash (compared to a Pleistocene ash) include; surface pitting, increased surface area, devitrification rinds (<1l micron wide) and reduced lithium contents. Future investigations will study the effect of temperature, pressure, solution composition, and flow rate on the relative mobility of U, Si, Li, F, and K.

Dissolution process for ZrO.sub.2 -UO.sub.2 -CaO fuels

DOEpatents

Paige, Bernice E.

1976-06-22

The present invention provides an improved dissolution process for ZrO.sub.2 -UO.sub.2 -CaO-type pressurized water reactor fuels. The zirconium cladding is dissolved with hydrofluoric acid, immersing the ZrO.sub.2 -UO.sub.2 -CaO fuel wafers in the resulting zirconium-dissolver-product in the dissolver vessel, and nitric acid is added to the dissolver vessel to facilitate dissolution of the uranium from the ZrO.sub.2 -UO.sub.2 -CaO fuel wafers.

Leachability of uranium and other elements from freshly erupted volcanic ash

USGS Publications Warehouse

Smith, D.B.; Zielinski, R.A.; Rose, W.I.

1982-01-01

A study of leaching of freshly erupted basaltic and dacitic air-fall ash and bomb fragment samples, unaffected by rain, shows that glass dissolution is the dominant process by which uranium is initially mobilized from air-fall volcanic ash. Si, Li, and V are also preferentially mobilized by glass dissolution. Gaseous transfer followed by fixation of soluble uranium species on volcanic-ash particles is not an important process affecting uranium mobility. Gaseous transfer, however, may be important in forming water-soluble phases, adsorbed to ash surfaces, enriched in the economically and environmentally important elements Zn, Cu, Cd, Pb, B, F, and Ba. Quick removal of these adsorbed elements by the first exposure of freshly erupted ash to rain and surface water may pose short-term hazards to certain forms of aquatic and terrestrial life. Such rapid release of material may also represent the first step in transportation of economically important elements to environments favorable for precipitation into deposits of commercial interest. Ash samples collected from the active Guatemalan volcanoes Fuego and Pacaya (high-Al basalts) and Santiaguito (hornblende-hypersthene dacite); bomb fragments from Augustine volcano (andesite-dacite), Alaska, and Heimaey (basalt), Vestmann Islands, Iceland; and fragments of "rhyolitic" pumice from various historic eruptions were subjected to three successive leaches with a constant water-to-ash weight ratio of 4:1. The volcanic material was successively leached by: (1) distilled-deionized water (pH = 5.0-5.5) at room temperature for 24 h, which removes water-soluble gases and salts adsorbed on ash surfaces during eruption; (2) dilute HCl solution (pH = 3.5-4.0) at room temperature for 24 h, which continues the attack initiated by the water and also attacks acid-soluble sulfides and oxides; (3) a solution 0.05 M in both Na,CO, and NaHCO, (pH = 9.9) at 80°C for one week, which preferentially dissolves volcanic glass. The first two leaches mimic interaction of ash with rain produced in the vicinity of an active eruption. The third leach accelerates the effect of prolonged contact of volcanic ash with alkaline ground water present during ash diagenesis.

High temperature dissolution of chromium substituted nickel ferrite in nitrilotriacetic acid medium

NASA Astrophysics Data System (ADS)

Sathyaseelan, V. S.; Chandramohan, P.; Velmurugan, S.

2016-12-01

High temperature (HT) dissolution of chromium substituted nickel ferrite was carried out with relevance to the decontamination of nuclear reactors by way of chemical dissolution of contaminated corrosion product oxides present on stainless steel coolant circuit surfaces. Chromium substituted nickel ferrites of composition, NiFe(2-x)CrxO4 (x ≤ 1), was synthetically prepared and characterized. HT dissolution of these oxides was carried out in nitrilotriacetic acid medium at 160 °C. Dissolution was remarkably increased at 160 °C when compared to at 85 °C in a reducing decontamination formulation. Complete dissolution could be achieved for the oxides with chromium content 0 and 0.2. Increasing the chromium content brought about a marked reduction in the dissolution rate. About 40 fold decrease in rate of dissolution was observed when chromium was increased from 0 to 1. The rate of dissolution was not very significantly reduced in the presence of N2H4. Dissolution of oxide was found to be stoichiometric.

Absorption of Thermal Neutrons in Uranium

DOE R&D Accomplishments Database

Creutz, E. C.; Wilson, R. R.; Wigner, E. P.

1941-09-26

A knowledge of the absorption processes for neutrons in uranium is important for planning a chain reaction experiment. The absorption of thermal neutrons in uranium and uranium oxide has been studied. Neutrons from the cyclotron were slowed down by passage through a graphite block. A uranium or uranium oxide sphere was placed at various positions in the block. The neutron intensity at different points in the sphere and in the graphite was measured by observing the activity induced in detectors or uranium oxide or manganese. It was found that both the fission activity in the uranium oxide and the activity induced in manganese was affected by non-thermal neutrons. An experimental correction for such effects was made by making measurements with the detectors surrounded by cadmium. After such corrections the results from three methods of procedure with the uranium oxide detectors and from the manganese detectors were consistent to within a few per cent.

Evolution of uranium distribution and speciation in mill tailings, COMINAK Mine, Niger.

PubMed

Déjeant, Adrien; Galoisy, Laurence; Roy, Régis; Calas, Georges; Boekhout, Flora; Phrommavanh, Vannapha; Descostes, Michael

2016-03-01

This study investigated the evolution of uranium distribution and speciation in mill tailings from the COMINAK mine (Niger), in production since 1978. A multi-scale approach was used, which combined high resolution remote sensing imagery, ICP-MS bulk rock analyses, powder X-ray diffraction, Scanning Electron Microscopy, Focused Ion Beam--Transmission Electron Microscopy and X-ray Absorption Near Edge Spectroscopy. Mineralogical analyses showed that some ore minerals, including residual uraninite and coffinite, undergo alteration and dissolution during tailings storage. The migration of uranium and other contaminants depends on (i) the chemical stability of secondary phases and sorbed species (dissolution and desorption processes), and (ii) the mechanical transport of fine particles bearing these elements. Uranium is stabilized after formation of secondary uranyl sulfates and phosphates, and adsorbed complexes on mineral surfaces (e.g. clay minerals). In particular, the stock of insoluble uranyl phosphates increases with time, thus contributing to the long-term stabilization of uranium. At the surface, a sulfate-cemented duricrust is formed after evaporation of pore water. This duricrust limits water infiltration and dust aerial dispersion, though it is enriched in uranium and many other elements, because of pore water rising from underlying levels by capillary action. Satellite images provided a detailed description of the tailings pile over time and allow monitoring of the chronology of successive tailings deposits. Satellite images suggest that uranium anomalies that occur at deep levels in the pile are most likely former surface duricrusts that have been buried under more recent tailings. Copyright © 2015 Elsevier B.V. All rights reserved.

Separation of uranium from (Th,U)O.sub.2 solid solutions

DOEpatents

Chiotti, Premo; Jha, Mahesh Chandra

1976-09-28

Uranium is separated from mixed oxides of thorium and uranium by a pyrometallurgical process in which the oxides are mixed with a molten chloride salt containing thorium tetrachloride and thorium metal which reduces the uranium oxide to uranium metal which can then be recovered from the molten salt. The process is particularly useful for the recovery of uranium from generally insoluble high-density sol-gel thoria-urania nuclear reactor fuel pellets.

Method of preparation of uranium nitride

DOEpatents

Kiplinger, Jaqueline Loetsch; Thomson, Robert Kenneth James

2013-07-09

Method for producing terminal uranium nitride complexes comprising providing a suitable starting material comprising uranium; oxidizing the starting material with a suitable oxidant to produce one or more uranium(IV)-azide complexes; and, sufficiently irradiating the uranium(IV)-azide complexes to produce the terminal uranium nitride complexes.

Process for electrolytically preparing uranium metal

DOEpatents

Haas, Paul A.

1989-01-01

A process for making uranium metal from uranium oxide by first fluorinating uranium oxide to form uranium tetrafluoride and next electrolytically reducing the uranium tetrafluoride with a carbon anode to form uranium metal and CF.sub.4. The CF.sub.4 is reused in the fluorination reaction rather than being disposed of as a hazardous waste.

Process for electrolytically preparing uranium metal

DOEpatents

Haas, Paul A.

1989-08-01

A process for making uranium metal from uranium oxide by first fluorinating uranium oxide to form uranium tetrafluoride and next electrolytically reducing the uranium tetrafluoride with a carbon anode to form uranium metal and CF.sub.4. The CF.sub.4 is reused in the fluorination reaction rather than being disposed of as a hazardous waste.

PLUTONIUM RECOVERY FROM NEUTRON-BOMBARDED URANIUM FUEL

DOEpatents

Moore, R.H.

1964-03-24

A process of recovering plutonium from fuel by dissolution in molten KAlCl/sub 4/ double salt is described. Molten lithium chloride plus stannous chloride is added to reduce plutonium tetrachloride to the trichloride, which is dissolved in a lithium chloride phase while the uranium, as the tetrachloride, is dissolved in a double-salt phase. Separation of the two phases is discussed. (AEC)

Arsenic release and speciation during the oxidative dissolution of arsenopyrite by O2 in the absence and presence of EDTA.

PubMed

Wang, Shaofeng; Jiao, BeiBei; Zhang, Mingmei; Zhang, Guoqing; Wang, Xin; Jia, Yongfeng

2018-03-15

The oxidative decomposition of arsenopyrite is an important source of As in surface environment. This study investigated the oxidative dissolution of arsenopyrite by O 2 and aqueous arsenic transformation at different pHs, dissolved oxygen (DO) contents, and temperatures in the absence and presence of EDTA. The oxidative dissolution was greatly inhibited at neutral and alkaline pH in the absence of EDTA. However, in the presence of EDTA, the oxidative dissolution rate increased linearly from pH 4 to 7. The highest dissolution rate was 3-4 times higher than that at pH 4 and 1-2 orders of magnitude higher than that at pH 7 in the absence of EDTA. This is possibly due to the lack of Fe oxyhydroxides on the surface of arsenopyrite. In the pH range of 7-10, the oxidative dissolution rate decreased linearly, possibly due to the formation of goethite and/or hematite coating. The oxidation of released arsenite (As III ) to arsenate (As V ) took place simultaneously during the oxidative dissolution of arsenopyrite in the presence of dissolved Fe without EDTA, while no obvious aqueous As III oxidation was observed in the presence of EDTA, indicating that aqueous Fe species play an important role in As III oxidation. Copyright © 2017 Elsevier B.V. All rights reserved.

Microbial transformations of uranium in wastes and implication on its mobility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Suzuki,Y.; Nankawa, T.; Ozaki, T.

2008-09-14

Uranium exists in several chemical forms in mining and mill tailings and in nuclear and weapons production wastes. Under appropriate conditions, microorganisms can affect the stability and mobility of U in wastes by altering the chemical speciation, solubility and sorption properties and thus could increase or decrease the concentrations of U in solution and the bioavailability. Dissolution or immobilization of U is brought about by direct enzymatic action or indirect nonenzymatic action of microorganisms. Although the physical, chemical, and geochemical processes affecting dissolution, precipitation, and mobilization of U have been extensively investigated, we have only limited information on the mechanismsmore » of microbial transformations of various chemical forms of U in the presence of electron donors and acceptors.« less

Decontamination of uranium-contaminated waste oil using supercritical fluid and nitric acid.

PubMed

Sung, Jinhyun; Kim, Jungsoo; Lee, Youngbae; Seol, Jeunggun; Ryu, Jaebong; Park, Kwangheon

2011-07-01

The waste oil used in nuclear fuel processing is contaminated with uranium because of its contact with materials or environments containing uranium. Under current law, waste oil that has been contaminated with uranium is very difficult to dispose of at a radioactive waste disposal site. To dispose of the uranium-contaminated waste oil, the uranium was separated from the contaminated waste oil. Supercritical R-22 is an excellent solvent for extracting clean oil from uranium-contaminated waste oil. The critical temperature of R-22 is 96.15 °C and the critical pressure is 49.9 bar. In this study, a process to remove uranium from the uranium-contaminated waste oil using supercritical R-22 was developed. The waste oil has a small amount of additives containing N, S or P, such as amines, dithiocarbamates and dialkyldithiophosphates. It seems that these organic additives form uranium-combined compounds. For this reason, dissolution of uranium from the uranium-combined compounds using nitric acid was needed. The efficiency of the removal of uranium from the uranium-contaminated waste oil using supercritical R-22 extraction and nitric acid treatment was determined.

Femtosecond Laser Ablation Multicollector ICPMS Analysis of Uranium Isotopes in NIST Glass

DOE Office of Scientific and Technical Information (OSTI.GOV)

Duffin, Andrew M.; Springer, Kellen WE; Ward, Jesse D.

We have utilized femtosecond laser ablation coupled to multi-collector inductively couple plasma mass spectrometry to measure the uranium isotopic content of NIST 61x (x=0,2,4,6) glasses. The uranium content of these glasses is a linear two-component mixing between isotopically natural uranium and the isotopically depleted spike used in preparing the glasses. Laser ablation results match extremely well, generally within a few ppm, with solution analysis following sample dissolution and chemical separation. In addition to isotopic data, sample utilization efficiency measurements indicate that over 1% of ablated uranium atoms reach a mass spectrometer detector, making this technique extremely efficient. Laser sampling alsomore » allows for spatial analysis and our data indicate that rare uranium concentration inhomogeneities exist in NIST 616 glass.« less

DNAPL remediation with in situ chemical oxidation using potassium permanganate - Part I. Mineralogy of Mn oxide and its dissolution in organic acids

NASA Astrophysics Data System (ADS)

Li, X. David; Schwartz, Franklin W.

2004-01-01

Previous studies on in situ chemical oxidation of trichloroethylene (TCE) with potassium permanganate indicated that the solid reaction product, Mn oxide, could reduce the permeability of the porous medium and impact the success of dense non-aqueous phase liquid (DNAPL) removal. In order to address the issue of permeability reduction caused by precipitation, this study investigated the mineralogy of Mn oxides and the possibilities of removing the solid precipitates by dissolution. The solid reaction product from the oxidation of TCE by permanganate is semi-amorphous potassium-rich birnessite, which has a layered mineral structure with an interlayer spacing of 7.3 Å. The chemical formula is K 0.854Mn 1.786O 4·1.55H 2O. It has a relatively small specific surface area at 23.6±0.82 m 2/g. Its point of zero charge (pzc) was measured as 3.7±0.4. This birnessite is a relatively active species and could participate in various reactions with existing organic and inorganic matter. The dissolution kinetics of Mn oxide was evaluated in batch experiments using solutions of citric acid, oxalic acid, and ethylenediaminetetraacetic acid (EDTA). Initial dissolution rates were determined to be 0.126 mM/m 2/h for citric acid, 1.35 mM/m 2/h for oxalic acid, and 5.176 mM/m 2/h for EDTA. These rates compare with 0.0025 mM/m 2/h for nitric acid at pH=2. Organic acids dissolve Mn oxide quickly. Reaction rates increase with acid concentration, as tested with citric acid. The dissolution mechanism likely involves proton and ligand-promoted dissolution and reductive dissolution. Citric and oxalic acid can induce ligand-promoted dissolution, while EDTA can induce ligand-promoted and reductive dissolutions. At low pH, proton-promoted dissolution seems to occur with all the acids tested, but this process is not dominant. Reductive dissolution appears to be the most effective process in dissolving the solid, followed by ligand-promoted dissolution. These experiments indicate the significant potential in using these organic acids to remove precipitates formed during the oxidation reaction.

Uranium oxidation: Characterization of oxides formed by reaction with water by infrared and sorption analyses

NASA Astrophysics Data System (ADS)

Fuller, E. L.; Smyrl, N. R.; Condon, J. B.; Eager, M. H.

1984-04-01

Three different uranium oxide samples have been characterized with respect to the different preparation techniques. The results show that the water reaction with uranium metal occurs cyclically forming laminar layers of oxide which spall off due to the strain at the oxide/metal interface. Single laminae are released if liquid water is present due to the prizing penetration at the reaction zone. The rate of reaction of water with uranium is directly proportional to the amount of adsorbed water on the oxide product. Rapid transport is effected through the open hydrous oxide product. Dehydration of the hydrous oxide irreversibly forms a more inert oxide which cannot be rehydrated to the degree that prevails in the original hydrous product of uranium oxidation with water. Inert gas sorption analyses and diffuse reflectance infrared studies combined with electron microscopy prove valuable in defining the chemistry and morphology of the oxidic products and hydrated intermediates.

Uranium oxide catalysts: environmental applications for treatment of chlorinated organic waste from nuclear industry.

PubMed

Lazareva, Svetlana; Ismagilov, Zinfer; Kuznetsov, Vadim; Shikina, Nadezhda; Kerzhentsev, Mikhail

2018-02-05

Huge amounts of nuclear waste, including depleted uranium, significantly contribute to the adverse environmental situation throughout the world. An approach to the effective use of uranium oxides in catalysts for the deep oxidation of chlorine-containing hydrocarbons is suggested. Investigation of the catalytic activity of the synthesized supported uranium oxide catalysts doped with Cr, Mn and Co transition metals in the chlorobenzene oxidation showed that these catalysts are comparable with conventional commercial ones. Physicochemical properties of the catalysts were studied by X-ray diffraction, temperature-programmed reduction with hydrogen (H 2 -TPR), and Fourier transform infrared spectroscopy. The higher activity of Mn- and Co-containing uranium oxide catalysts in the H 2 -TPR and oxidation of chlorobenzene in comparison with non-uranium catalysts may be related to the formation of a new disperse phase represented by uranates. The study of chlorobenzene adsorption revealed that the surface oxygen is involved in the catalytic process.

Estimated Marine Residence Times for Drowned Barbadian Paleoreefs

NASA Astrophysics Data System (ADS)

Mey, J. L.

2008-12-01

Fossil corals are used to estimate past sea level and also to calibrate 14C ages with the aid of U-Th and U-Pa dating methods. These coral fossils have often been subaerially exposed and thus are affected by diagenesis during their initial interaction with fresh water. In an effort to understand when such disequilibria in fossil coral reefs occurred, we have quantified our 'dissolution-cum-adsorption' model (Mey, 2008) for the uranium series disequilibria using a geometrical construction, based on the evolution of the activities in a 230Th/238U versus 234U/238U diagram for closed versus open systems. The traditional age equations for the uranium-series with excess daughters have been used to construct a relationship between (i) the angles of the equal age lines in the 230Th/238U versus 234U/238U activity diagrams, and (ii) the quantified angles of the regressed lines of several uranium series disequilibria trends from Barbados. Our results indicate that the severity of the Barbados uranium series disequilibria is not only explained by 234U and 230Th addition, but may also reflect a loss of 238U through dissolution of coral skeletal structure. The net effect is 238U removal, whereas 234U and 230Th remain; thus, the disequilibria for the extant coral increase the excess daughters' ratio. Our results further indicate that the activity of 234U is reduced (compared to 230Th), as would be expected in regard to the lower mobility of trapped 230Th. It is proposed that the major dissolution that caused the uranium series disequilibria occurred during one relatively short-lived event when the paleoreefs experienced the very first freshwater exposure. During this event, the diagenetic potential was at its maximum for redistribution of the uranium series; this then caused the 234U and the 230Th to behave in a systematic way, resulting in linear trends. The linear trends in the open system uranium series were set early, as shown in the 230Th/238U versus 234U/238U activity diagrams. The timing of the first exposure of the freshwater in the reefs is calculated based on the results of our new model. From the relationship between, (i) dissolution, (ii) in-grown 230Th, and (iii) excess 234U, we derived that the 60,000 old Marine Isotope stage 3 (MIS 3) reef was exposed to freshwater 36-38,000 years after growth in the marine environment. We have calculated these 'marine residence times' for the MIS 3 5a, 5c, 5e, 6.0, 7a and 7c reefs; our results correspond with the duration of the sea level high stand in each of the stages. References: Mey, J. L., (2008) The Uranium Series Diagenesis and the Morphology of Drowned Barbadian Paleoreefs, PhD dissertation, 325pp: Graduate Center, City University of New York, New York.

Uranium mobility during interaction of rhyolitic obsidian, perlite and felsite with alkaline carbonate solution: T = 120° C, P = 210 kg/cm2

USGS Publications Warehouse

Zielinski, Robert A.

1979-01-01

Well-characterized samples of rhyolitic obsidian, perlite and felsite from a single lava flow are leached of U by alkaline oxidizing solutions under open-system conditions. Pressure, temperature, flow rate and solution composition are held constant in order to evaluate the relative importance of differences in surface area and crystallinity. Under the experimental conditions U removal from crushed glassy samples proceeds by a mechanism of glass dissolution in which U and silica are dissolved in approximately equal weight fractions. The rate of U removal from crushed glassy samples increases with decreasing average grain size (surface area). Initial rapid loss of a small component (≈ 2.5%) of the total U from crushed felsite. followed by much slower U loss, reflects variable rates of attack of numerous uranium sites. The fractions of U removed during the experiment ranged from 3.2% (felsite) to 27% (perlite). An empirical method for evaluating the relative rate of U loss from contemporaneous volcanic rocks is presented which incorporates leaching results and rock permeability data.

Uranium dioxide electrolysis

DOEpatents

Willit, James L [Batavia, IL; Ackerman, John P [Prescott, AZ; Williamson, Mark A [Naperville, IL

2009-12-29

This is a single stage process for treating spent nuclear fuel from light water reactors. The spent nuclear fuel, uranium oxide, UO.sub.2, is added to a solution of UCl.sub.4 dissolved in molten LiCl. A carbon anode and a metallic cathode is positioned in the molten salt bath. A power source is connected to the electrodes and a voltage greater than or equal to 1.3 volts is applied to the bath. At the anode, the carbon is oxidized to form carbon dioxide and uranium chloride. At the cathode, uranium is electroplated. The uranium chloride at the cathode reacts with more uranium oxide to continue the reaction. The process may also be used with other transuranic oxides and rare earth metal oxides.

Atomic-scale Studies of Uranium Oxidation and Corrosion by Water Vapour.

PubMed

Martin, T L; Coe, C; Bagot, P A J; Morrall, P; Smith, G D W; Scott, T; Moody, M P

2016-07-12

Understanding the corrosion of uranium is important for its safe, long-term storage. Uranium metal corrodes rapidly in air, but the exact mechanism remains subject to debate. Atom Probe Tomography was used to investigate the surface microstructure of metallic depleted uranium specimens following polishing and exposure to moist air. A complex, corrugated metal-oxide interface was observed, with approximately 60 at.% oxygen content within the oxide. Interestingly, a very thin (~5 nm) interfacial layer of uranium hydride was observed at the oxide-metal interface. Exposure to deuterated water vapour produced an equivalent deuteride signal at the metal-oxide interface, confirming the hydride as originating via the water vapour oxidation mechanism. Hydroxide ions were detected uniformly throughout the oxide, yet showed reduced prominence at the metal interface. These results support a proposed mechanism for the oxidation of uranium in water vapour environments where the transport of hydroxyl species and the formation of hydride are key to understanding the observed behaviour.

Atomic-scale Studies of Uranium Oxidation and Corrosion by Water Vapour

NASA Astrophysics Data System (ADS)

Martin, T. L.; Coe, C.; Bagot, P. A. J.; Morrall, P.; Smith, G. D. W.; Scott, T.; Moody, M. P.

2016-07-01

Understanding the corrosion of uranium is important for its safe, long-term storage. Uranium metal corrodes rapidly in air, but the exact mechanism remains subject to debate. Atom Probe Tomography was used to investigate the surface microstructure of metallic depleted uranium specimens following polishing and exposure to moist air. A complex, corrugated metal-oxide interface was observed, with approximately 60 at.% oxygen content within the oxide. Interestingly, a very thin (~5 nm) interfacial layer of uranium hydride was observed at the oxide-metal interface. Exposure to deuterated water vapour produced an equivalent deuteride signal at the metal-oxide interface, confirming the hydride as originating via the water vapour oxidation mechanism. Hydroxide ions were detected uniformly throughout the oxide, yet showed reduced prominence at the metal interface. These results support a proposed mechanism for the oxidation of uranium in water vapour environments where the transport of hydroxyl species and the formation of hydride are key to understanding the observed behaviour.

Atomic-scale Studies of Uranium Oxidation and Corrosion by Water Vapour

PubMed Central

Martin, T. L.; Coe, C.; Bagot, P. A. J.; Morrall, P.; Smith, G. D. W; Scott, T.; Moody, M. P.

2016-01-01

Understanding the corrosion of uranium is important for its safe, long-term storage. Uranium metal corrodes rapidly in air, but the exact mechanism remains subject to debate. Atom Probe Tomography was used to investigate the surface microstructure of metallic depleted uranium specimens following polishing and exposure to moist air. A complex, corrugated metal-oxide interface was observed, with approximately 60 at.% oxygen content within the oxide. Interestingly, a very thin (~5 nm) interfacial layer of uranium hydride was observed at the oxide-metal interface. Exposure to deuterated water vapour produced an equivalent deuteride signal at the metal-oxide interface, confirming the hydride as originating via the water vapour oxidation mechanism. Hydroxide ions were detected uniformly throughout the oxide, yet showed reduced prominence at the metal interface. These results support a proposed mechanism for the oxidation of uranium in water vapour environments where the transport of hydroxyl species and the formation of hydride are key to understanding the observed behaviour. PMID:27403638

Nanostructured Metal Oxide Sorbents for the Collection and Recovery of Uranium from Seawater

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chouyyok, Wilaiwan; Warner, Cynthia L.; Mackie, Katherine E.

2016-02-07

The ability to collect uranium from seawater offers the potential for a long-term green fuel supply for nuclear energy. However, extraction of uranium, and other trace minerals, is challenging due to the high ionic strength and low mineral concentrations in seawater. Herein we evaluate the use of nanostructured metal oxide sorbents for the collection and recovery of uranium from seawater. Chemical affinity, chemical adsorption capacity and kinetics of preferred sorbent materials were evaluated. High surface area manganese and iron oxide nanomaterials showed excellent performance for uranium collection from seawater. Inexpensive nontoxic carbonate solutions were demonstrated to be an effective andmore » environmental benign method of stripping the uranium from the metal oxide sorbents. Various formats for the utilization of the nanostructured metals oxide sorbent materials are discussed including traditional and nontraditional methods such as magnetic separation. Keywords: Uranium, nano, manganese, iron, sorbent, seawater, magnetic, separations, nuclear energy« less

High Useful Yield and Isotopic Analysis of Uranium by Resonance Ionization Mass Spectrometry

DOE PAGES

Savina, Michael R.; Isselhardt, Brett H.; Kucher, Andrew; ...

2017-05-09

Useful yields from resonance ionization mass spectrometry can be extremely high compared to other mass spectrometry techniques, but uranium analysis shows strong matrix effects arising from the tendency of uranium to form strongly bound oxide molecules that do not dissociate appreciably on energetic ion bombardment. Here, we demonstrate a useful yield of 24% for metallic uranium. Modeling the laser ionization and ion transmission processes shows that the high useful yield is attributable to a high ion fraction achieved by resonance ionization. We quantify the reduction of uranium oxide surface layers by Ar + and Ga + sputtering. The useful yieldmore » for uranium atoms from a uranium dioxide matrix is 0.4% and rises to 2% when the surface is in sputter equilibrium with the ion beam. The lower useful yield from the oxide is almost entirely due to uranium oxide molecules reducing the neutral atom content of the sputtered flux. We also demonstrate rapid isotopic analysis of solid uranium oxide at a precision of <0.5% relative standard deviation using relatively broadband lasers to mitigate spectroscopic fractionation.« less

High Useful Yield and Isotopic Analysis of Uranium by Resonance Ionization Mass Spectrometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Savina, Michael R.; Isselhardt, Brett H.; Kucher, Andrew

Useful yields from resonance ionization mass spectrometry can be extremely high compared to other mass spectrometry techniques, but uranium analysis shows strong matrix effects arising from the tendency of uranium to form strongly bound oxide molecules that do not dissociate appreciably on energetic ion bombardment. Here, we demonstrate a useful yield of 24% for metallic uranium. Modeling the laser ionization and ion transmission processes shows that the high useful yield is attributable to a high ion fraction achieved by resonance ionization. We quantify the reduction of uranium oxide surface layers by Ar + and Ga + sputtering. The useful yieldmore » for uranium atoms from a uranium dioxide matrix is 0.4% and rises to 2% when the surface is in sputter equilibrium with the ion beam. The lower useful yield from the oxide is almost entirely due to uranium oxide molecules reducing the neutral atom content of the sputtered flux. We also demonstrate rapid isotopic analysis of solid uranium oxide at a precision of <0.5% relative standard deviation using relatively broadband lasers to mitigate spectroscopic fractionation.« less

Effect of metallic iron on the oxidative dissolution of UO2 doped with a radioactive alpha emitter in synthetic Callovian-Oxfordian groundwater

NASA Astrophysics Data System (ADS)

Odorowski, Mélina; Jegou, Christophe; De Windt, Laurent; Broudic, Véronique; Jouan, Gauthier; Peuget, Sylvain; Martin, Christelle

2017-12-01

In the hypothesis of direct disposal of spent fuel in a geological nuclear waste repository, interactions between the fuel mainly composed of UO2 and its environment must be understood. The dissolution rate of the UO2 matrix, which depends on the redox conditions on the fuel surface, will have a major impact on the release of radionuclides into the environment. The reducing conditions expected for a geological disposal situation would appear to be favorable as regards the solubility and stability of the UO2 matrix, but may be disturbed on the surface of irradiated fuel. In particular, the local redox conditions will result from a competition between the radiolysis effects of water under alpha irradiation (simultaneously producing oxidizing species like H2O2, hydrogen peroxide, and reducing species like H2, hydrogen) and those of redox active species from the environment. In particular, Fe2+, a strongly reducing aqueous species coming from the corrosion of the iron canister or from the host rock, could influence the dissolution of the fuel matrix. The effect of iron on the oxidative dissolution of UO2 was thus investigated under the conditions of the French disposal site, a Callovian-Oxfordian clay formation chosen by the French National Radioactive Waste Management Agency (Andra), here tested under alpha irradiation. For this study, UO2 fuel pellets doped with a radioactive alpha emitter (238/239Pu) were leached in synthetic Callovian-Oxfordian groundwater (representative of the French waste disposal site groundwater) in the presence of a metallic iron foil to simulate the steel canister. The pellets had varying levels of alpha activity, in order to modulate the concentrations of species produced by water radiolysis on the surface and to simulate the activity of aged spent fuel after 50 and 10,000 years of alpha radioactivity decay. The experimental data showed that whatever the sample alpha radioactivity, the presence of iron inhibits the oxidizing dissolution of UO2 and leads to low uranium concentrations (between 4 × 10-10 and 4 × 10-9 M), through a reactional mechanism located in the very first microns of the UO2/water reactional interface. The mechanism involves consumption of oxidizing species, in particular of H2O2 by Fe2+ at the precise place where these species are produced, and is accompanied by the precipitation of an akaganeite-type Fe3+ hydroxide on the surface. The higher the radioactivity of the samples, the greater the precipitation induced. Modeling has been developed, coupling chemistry with transport and based on the main reactional mechanisms identified, which enables accurate reproduction of the mineralogy of the system under study, giving the nature of the phases under observation as well as the location of their precipitation. Obviously without excluding a potential contribution from the hydrogen produced by the anoxic corrosion of the iron foil, this study has shown that iron plays a major role in this oxidizing dissolution inhibition process for the system investigated (localized alpha radiolysis). This inhibitor effect associated with iron is therefore strongly dependent on the location of the redox front, which is found on the surface in the case of alpha irradiation UO2/water reactional interface.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jackson, Jay M.; Lopez, Jacquelyn C.; Wayne, David M.

2012-07-05

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

Heat-induced redistribution of surface oxide in uranium

NASA Astrophysics Data System (ADS)

Swissa, Eli; Shamir, Noah; Mintz, Moshe H.; Bloch, Joseph

1990-09-01

The redistribution of oxygen and uranium metal at the vicinity of the metal-oxide interface of native and grown oxides due to vacuum thermal annealing was studied for uranium and uranium-chromium alloy using Auger depth profiling and metallographic techniques. It was found that uranium metal is segregating out through the uranium oxide layer for annealing temperatures above 450°C. At the same time the oxide is redistributed in the metal below the oxide-metal interface in a diffusion like process. By applying a diffusion equation of a finite source, the diffusion coefficients for the process were obtained from the oxygen depth profiles measured for different annealing times. An Arrhenius like behavior was found for the diffusion coefficient between 400 and 800°C. The activation energy obtained was Ea = 15.4 ± 1.9 kcal/mole and the pre-exponential factor, D0 = 1.1 × 10 -8cm2/ s. An internal oxidation mechanism is proposed to explain the results.

Dissolution behaviour of 238U, 234U and 230Th deposited on filters from personal dosemeters.

PubMed

Becková, Vera; Malátová, Irena

2008-01-01

Kinetics of dissolution of (238)U, (234)U and (230)Th dust deposited on filters from personal alpha dosemeters was studied by means of a 26-d in vitro dissolution test with a serum ultrafiltrate simulant. Dosemeters had been used by miners at the uranium mine 'Dolní Rozínka' at Rozná, Czech Republic. The sampling flow-rate as declared by the producer is 4 l h(-1) and the sampling period is typically 1 month. Studied filters contained 125 +/- 6 mBq (238)U in equilibrium with (234)U and (230)Th; no (232)Th series nuclides were found. Half-time of rapid dissolution of 1.4 d for (238)U and (234)U and slow dissolution half-times of 173 and 116 d were found for (238)U and (234)U, respectively. No detectable dissolution of (230)Th was found.

Dissolution of Fe(III) (hydr) oxides by metal-EDTA complexes

NASA Astrophysics Data System (ADS)

Ngwack, Bernd; Sigg, Laura

1997-03-01

The dissolution of Fe(III)(hydr)oxides (goethite and hydrous ferric oxide) by metal-EDTA complexes occurs by ligand-promoted dissolution. The process is initiated by the adsorption of metal-EDTA complexes to the surface and is followed by the dissociation of the complex at the surface and the release of Fe(III)EDTA into solution. The dissolution rate is decreased to a great extent if EDTA is complexed by metals in comparison to the uncomplexed EDTA. The rate decreases in the order EDTA CaEDTA ≫ PbEDTA > ZnEDTA > CuEDTA > Co(II)EDTA > NiEDTA. Two different rate-limiting steps determine the dissolution process: (1) detachment of Fe(III) from the oxide-structure and (2) dissociation of the metal-EDTA complexes. In the case of goethite, step 1 is slower than step 2 and the dissolution rates by various metals are similar. In the case of hydrous ferric oxide, step 2 is rate-limiting and the effect of the complexed metal is very pronounced.

Anodic behavior of uranium in AlCl3-1-ethyl-3-methyl-imidazolium chloride ionic liquid

NASA Astrophysics Data System (ADS)

Jiang, Yidong; Luo, Lizhu; Wang, Shaofei; Bin, Ren; Zhang, Guikai; Wang, Xiaolin

2018-01-01

The oxidation state of metals unambiguously affects its anodic behavior in ionic liquid. We systematically investigated the anodic behavior of uranium with different surface oxidation states by electrochemical measurements, spectroscopic methods and surface analysis techniques. In the anodic process, metal uranium can be oxidized to U3+. The corresponding products accumulated on the metal/ILs interface will form a viscous layer. The anodic behavior of uranium is also strongly dependent upon the surface oxide states including thickness and homogeneity of the oxide film. With an increase in the thickness of oxide film, it will be breached at potentials in excess of a critical value. A uniform oxide on uranium surface can be breached evenly, and then the underlying metal starts to dissolve forming a viscous layer which can facilitate uniformly stripping of oxide, thus giving an oxide-free surface. Otherwise, a nonuniform oxide can result in a severe pitted surface with residue oxygen.

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.

High temperature dissolution of oxides in complexing media

NASA Astrophysics Data System (ADS)

Sathyaseelan, Valil S.; Rufus, Appadurai L.; Subramanian, Hariharan; Bhaskarapillai, Anupkumar; Wilson, Shiny; Narasimhan, Sevilimedu V.; Velmurugan, Sankaralingam

2011-12-01

Dissolution of transition metal oxides such as magnetite (Fe 3O 4), mixed ferrites (NiFe 2O 4, ZnFe 2O 4, MgFe 2O 4), bonaccordite (Ni 2FeBO 5) and chromium oxide (Cr 2O 3) in organic complexing media was attempted at higher temperatures (80-180 °C). On increasing the temperature from 80 to 180 °C, the dissolution rate of magnetite in nitrilo triacetic acid (NTA) medium increased six folds. The trend obtained for the dissolution of other oxides was ZnFe 2O 4 > NiFe 2O 4 > MgFe 2O 4 > Cr 2O 3, which followed the same trend as the lability of their metal-oxo bonds. Other complexing agents such as ethylene diamine tetra acetic acid (EDTA), pyridine dicarboxylic acid (PDCA), citric acid and reducing agents viz., oxalic acid and ascorbic acid were also evaluated for their oxide dissolution efficiency at 160 °C. EDTA showed maximum dissolution rate of 21.4 μm/h for magnetite. Addition of oxalic acid/ascorbic acid to complexing media (NTA/EDTA) showed identical effect on the dissolution of magnetite. Addition of hydrazine, another reducing agent, to NTA decreased the rate of dissolution of magnetite by 50%.

URANIUM SEPARATION PROCESS

DOEpatents

Lyon, W.L.

1962-04-17

A method of separating uranium oxides from PuO/sub 2/, ThO/sub 2/, and other actinide oxides is described. The oxide mixture is suspended in a fused salt melt and a chlorinating agent such as chlorine gas or phosgene is sparged through the suspension. Uranium oxides are selectively chlorinated and dissolve in the melt, which may then be filtered to remove the unchlorinated oxides of the other actinides. (AEC)

PRODUCTION OF URANIUM METAL BY CARBON REDUCTION

DOEpatents

Holden, R.B.; Powers, R.M.; Blaber, O.J.

1959-09-22

The preparation of uranium metal by the carbon reduction of an oxide of uranium is described. In a preferred embodiment of the invention a charge composed of carbon and uranium oxide is heated to a solid mass after which it is further heated under vacuum to a temperature of about 2000 deg C to produce a fused uranium metal. Slowly ccoling the fused mass produces a dendritic structure of uranium carbide in uranium metal. Reacting the solidified charge with deionized water hydrolyzes the uranium carbide to finely divide uranium dioxide which can be separated from the coarser uranium metal by ordinary filtration methods.

Oxidative dissolution of silver nanoparticles: A new theoretical approach.

PubMed

Adamczyk, Zbigniew; Oćwieja, Magdalena; Mrowiec, Halina; Walas, Stanisław; Lupa, Dawid

2016-05-01

A general model of an oxidative dissolution of silver particle suspensions was developed that rigorously considers the bulk and surface solute transport. A two-step surface reaction scheme was proposed that comprises the formation of the silver oxide phase by direct oxidation and the acidic dissolution of this phase leading to silver ion release. By considering this, a complete set of equations is formulated describing oxygen and silver ion transport to and from particles' surfaces. These equations are solved in some limiting cases of nanoparticle dissolution in dilute suspensions. The obtained kinetic equations were used for the interpretation of experimental data pertinent to the dissolution kinetics of citrate-stabilized silver nanoparticles. In these kinetic measurements the role of pH and bulk suspension concentration was quantitatively evaluated by using the atomic absorption spectrometry (AAS). It was shown that the theoretical model adequately reflects the main features of the experimental results, especially the significant increase in the dissolution rate for lower pH. Also the presence of two kinetic regimes was quantitatively explained in terms of the decrease in the coverage of the fast dissolving oxide layer. The overall silver dissolution rate constants characterizing these two regimes were determined. Copyright © 2015 Elsevier Inc. All rights reserved.

METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Handwerk, J.H.; BAch, R.A.

1959-08-18

A method is described for preparing a reactor fuel element by forming a mixture of thorium dioxide and an oxide of uranium, the uranium being present. In an oxidation state at least as high as it is in U/sub 3/O/sub 8/, into a desired shape and firing in air at a temperature siifficiently high to reduce the higher uranium oxide to uranium dioxide.

Improving the Estimates of Waste from the Recycling of Used Nuclear Fuel - 13410

DOE Office of Scientific and Technical Information (OSTI.GOV)

Phillips, Chris; Willis, William; Carter, Robert

2013-07-01

Estimates are presented of wastes arising from the reprocessing of 50 GWD/tonne, 5 year and 50 year cooled used nuclear fuel (UNF) from Light Water Reactors (LWRs), using the 'NUEX' solvent extraction process. NUEX is a fourth generation aqueous based reprocessing system, comprising shearing and dissolution in nitric acid of the UNF, separation of uranium and mixed uranium-plutonium using solvent extraction in a development of the PUREX process using tri-n-butyl phosphate in a kerosene diluent, purification of the plutonium and uranium-plutonium products, and conversion of them to uranium trioxide and mixed uranium-plutonium dioxides respectively. These products are suitable for usemore » as new LWR uranium oxide and mixed oxide fuel, respectively. Each unit process is described and the wastes that it produces are identified and quantified. Quantification of the process wastes was achieved by use of a detailed process model developed using the Aspen Custom Modeler suite of software and based on both first principles equilibrium and rate data, plus practical experience and data from the industrial scale Thermal Oxide Reprocessing Plant (THORP) at the Sellafield nuclear site in the United Kingdom. By feeding this model with the known concentrations of all species in the incoming UNF, the species and their concentrations in all product and waste streams were produced as the output. By using these data, along with a defined set of assumptions, including regulatory requirements, it was possible to calculate the waste forms, their radioactivities, volumes and quantities. Quantification of secondary wastes, such as plant maintenance, housekeeping and clean-up wastes, was achieved by reviewing actual operating experience from THORP during its hot operation from 1994 to the present time. This work was carried out under a contract from the United States Department of Energy (DOE) and, so as to enable DOE to make valid comparisons with other similar work, a number of assumptions were agreed. These include an assumed reprocessing capacity of 800 tonnes per year, the requirement to remove as waste forms the volatile fission products carbon-14, iodine-129, krypton-85, tritium and ruthenium-106, the restriction of discharge of any water from the facility unless it meets US Environmental Protection Agency drinking water standards, no intentional blending of wastes to lower their classification, and the requirement for the recovered uranium to be sufficiently free from fission products and neutron-absorbing species to allow it to be re-enriched and recycled as nuclear fuel. The results from this work showed that over 99.9% of the radioactivity in the UNF can be concentrated via reprocessing into a fission-product-containing vitrified product, bottles of compressed krypton storage and a cement grout containing the tritium, that together have a volume of only about one eighth the volume of the original UNF. The other waste forms have larger volumes than the original UNF but contain only the remaining 0.1% of the radioactivity. (authors)« less

Uranium induces oxidative stress in lung epithelial cells

PubMed Central

Periyakaruppan, Adaikkappan; Kumar, Felix; Sarkar, Shubhashish; Sharma, Chidananda S.

2009-01-01

Uranium compounds are widely used in the nuclear fuel cycle, antitank weapons, tank armor, and also as a pigment to color ceramics and glass. Effective management of waste uranium compounds is necessary to prevent exposure to avoid adverse health effects on the population. Health risks associated with uranium exposure includes kidney disease and respiratory disorders. In addition, several published results have shown uranium or depleted uranium causes DNA damage, mutagenicity, cancer and neurological defects. In the current study, uranium toxicity was evaluated in rat lung epithelial cells. The study shows uranium induces significant oxidative stress in rat lung epithelial cells followed by concomitant decrease in the antioxidant potential of the cells. Treatment with uranium to rat lung epithelial cells also decreased cell proliferation after 72 h in culture. The decrease in cell proliferation was attributed to loss of total glutathione and superoxide dismutase in the presence of uranium. Thus the results indicate the ineffectiveness of antioxidant system’s response to the oxidative stress induced by uranium in the cells. PMID:17124605

METHOD OF SEPARATING URANIUM VALUES, PLUTONIUM VALUES AND FISSION PRODUCTS BY CHLORINATION

DOEpatents

Brown, H.S.; Seaborg, G.T.

1959-02-24

The separation of plutonium and uranium from each other and from other substances is described. In general, the method comprises the steps of contacting the uranium with chlorine in the presence of a holdback material selected from the group consisting of lanthanum oxide and thorium oxide to form a uranium chloride higher than uranium tetrachloride, and thereafter heating the uranium chloride thus formed to a temperature at which the uranium chloride is volatilized off but below the volatilizalion temperature of plutonium chloride.

URANIUM LEACHING AND RECOVERY PROCESS

DOEpatents

McClaine, L.A.

1959-08-18

A process is described for recovering uranium from carbonate leach solutions by precipitating uranium as a mixed oxidation state compound. Uranium is recovered by adding a quadrivalent uranium carbon;te solution to the carbonate solution, adjusting the pH to 13 or greater, and precipitating the uranium as a filterable mixed oxidation state compound. In the event vanadium occurs with the uranium, the vanadium is unaffected by the uranium precipitation step and remains in the carbonate solution. The uranium-free solution is electrolyzed in the cathode compartment of a mercury cathode diaphragm cell to reduce and precipitate the vanadium.

Potential Dependence of Pt and Co Dissolution from Platinum-Cobalt Alloy PEFC Catalysts Using Time-Resolved Measurements

DOE PAGES

Ahluwalia, Rajesh K.; Papadias, Dionissios D.; Kariuki, Nancy N.; ...

2018-02-09

An electrochemical flow cell system with catalyst-ionomer ink deposited on glassy carbon is used to investigate the aqueous stability of commercial PtCo alloys under cyclic potentials. An on-line inductively coupled plasma-mass spectrometer, capable of real-time measurements, is used to resolve the anodic and cathodic dissolution of Pt and Co during square-wave and triangle-wave potential cycles. We observe Co dissolution at all potentials, distinct peaks in anodic and cathodic Pt dissolution rates above 0.9 V, and potential-dependent Pt and Co dissolution rates. The amount of Pt that dissolves cathodically is smaller than the amount that dissolves anodically if the upper potentialmore » limit (UPL) is lower than 0.9 V. At the highest UPL investigated, 1.0 V, the cathodic dissolution greatly exceeds the anodic dissolution. A non-ideal solid solution model indicates that the anodic dissolution can be associated with the electrochemical oxidation of Pt and PtOH to Pt 2+, and the cathodic dissolution to electrochemical reduction of a higher Pt oxide, PtO x (x > 1), to Pt 2+. Pt also dissolves oxidatively during the cathodic scans but in smaller amounts than due to the reductive dissolution of PtO x. The relative amounts Pt dissolving oxidatively as Pt and PtOH depend on the potential cycle and UPL.« less

Potential Dependence of Pt and Co Dissolution from Platinum-Cobalt Alloy PEFC Catalysts Using Time-Resolved Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahluwalia, Rajesh K.; Papadias, Dionissios D.; Kariuki, Nancy N.

An electrochemical flow cell system with catalyst-ionomer ink deposited on glassy carbon is used to investigate the aqueous stability of commercial PtCo alloys under cyclic potentials. An on-line inductively coupled plasma-mass spectrometer, capable of real-time measurements, is used to resolve the anodic and cathodic dissolution of Pt and Co during square-wave and triangle-wave potential cycles. We observe Co dissolution at all potentials, distinct peaks in anodic and cathodic Pt dissolution rates above 0.9 V, and potential-dependent Pt and Co dissolution rates. The amount of Pt that dissolves cathodically is smaller than the amount that dissolves anodically if the upper potentialmore » limit (UPL) is lower than 0.9 V. At the highest UPL investigated, 1.0 V, the cathodic dissolution greatly exceeds the anodic dissolution. A non-ideal solid solution model indicates that the anodic dissolution can be associated with the electrochemical oxidation of Pt and PtOH to Pt 2+, and the cathodic dissolution to electrochemical reduction of a higher Pt oxide, PtO x (x > 1), to Pt 2+. Pt also dissolves oxidatively during the cathodic scans but in smaller amounts than due to the reductive dissolution of PtO x. The relative amounts Pt dissolving oxidatively as Pt and PtOH depend on the potential cycle and UPL.« less

Phase discrimination of uranium oxides using laser-induced breakdown spectroscopy

NASA Astrophysics Data System (ADS)

Campbell, Keri R.; Wozniak, Nicholas R.; Colgan, James P.; Judge, Elizabeth J.; Barefield, James E.; Kilcrease, David P.; Wilkerson, Marianne P.; Czerwinski, Ken R.; Clegg, Samuel M.

2017-08-01

Nuclear forensics goals for characterizing samples of interest include qualitative and quantitative analysis of major and trace elements, isotopic analysis, phase identification, and physical analysis. These samples may include uranium oxides UO2, U3O8, and UO3, which play an important role in the front end of the nuclear fuel cycle, from mining to fuel fabrication. The focus of this study is to compare the ratios of the intensities of uranium and oxygen emission lines which can be used to distinguish between different uranium oxide materials using Laser-Induced Breakdown Spectroscopy (LIBS). Measurements at varying laser powers were made under an argon atmosphere at 585 Torr to ensure the oxygen emission intensity was originating from the sample, and not from the atmosphere. Fifteen uranium emission lines were used to compare experimental results with theoretical calculations in order to determine the plasma conditions. Using a laser energy of 26 mJ, the uranium lines 591.539 and 682.692 nm provide the highest degree of discrimination between the uranium oxides. The study presented here suggests that LIBS is useful for discriminating uranium oxide phases, UO2, U3O8, and UO3.

Kinetic dissolution of carbonates and Mn oxides in acidic water: Measurement of in situ field rates and reactive transport modeling

USGS Publications Warehouse

Brown, J.G.; Glynn, P.D.

2003-01-01

The kinetics of carbonate and Mn oxide dissolution under acidic conditions were examined through the in situ exposure of pure phase samples to acidic ground water in Pinal Creek Basin, Arizona. The average long-term calculated in situ dissolution rates for calcite and dolomite were 1.65??10-7 and 3.64??10-10 mmol/(cm2 s), respectively, which were about 3 orders of magnitude slower than rates derived in laboratory experiments by other investigators. Application of both in situ and lab-derived calcite and dolomite dissolution rates to equilibrium reactive transport simulations of a column experiment did not improve the fit to measured outflow chemistry: at the spatial and temporal scales of the column experiment, the use of an equilibrium model adequately simulated carbonate dissolution in the column. Pyrolusite (MnO2) exposed to acidic ground water for 595 days increased slightly in weight despite thermodynamic conditions that favored dissolution. This result might be related to a recent finding by another investigator that the reductive dissolution of pyrolusite is accompanied by the precipitation of a mixed Mn-Fe oxide species. In PHREEQC reactive transport simulations, the incorporation of Mn kinetics improved the fit between observed and simulated behavior at the column and field scales, although the column-fitted rate for Mn-oxide dissolution was about 4 orders of magnitude greater than the field-fitted rate. Remaining differences between observed and simulated contaminant transport trends at the Pinal Creek site were likely related to factors other than the Mn oxide dissolution rate, such as the concentration of Fe oxide surface sites available for adsorption, the effects of competition among dissolved species for available surface sites, or reactions not included in the model.

Flowsheet Analysis of U-Pu Co-Crystallization Process as a New Reprocessing System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shunji Homma; Jun-ichi Ishii; Jiro Koga

2006-07-01

A new fuel reprocessing system by U-Pu co-crystallization process is proposed and examined by flowsheet analysis. This reprocessing system is based on the fact that hexavalent plutonium in nitric acid solution is co-crystallized with uranyl nitrate, whereas it is not crystallized when uranyl nitrate does not exist in the solution. The system consists of five steps: dissolution of spent fuel, plutonium oxidation, U-Pu co-crystallization as a co-decontamination, re-dissolution of the crystals, and U re-crystallization as a U-Pu separation. The system requires a recycling of the mother liquor from the U-Pu co-crystallization step and the appropriate recycle ratio is determined bymore » flowsheet analysis such that the satisfactory decontamination is achieved. Further flowsheet study using four different compositions of LWR spent fuels demonstrates that the constant ratio of plutonium to uranium in mother liquor from the re-crystallization step is achieved for every composition by controlling the temperature. It is also demonstrated by comparing to the Purex process that the size of the plant based on the proposed system is significantly reduced. (authors)« less

A preliminary report on the geology of the Dennison-Bunn uranium claim, Sandoval County, New Mexico

USGS Publications Warehouse

Ridgley, Jennie L.

1978-01-01

Uranium at the Dennison-Bunn claim, south of Cuba, N. Mex., along the east margin of the San Juan Basin, occurs in unoxidized gray, fluvial channel sandstone of the Westwater Canyon Member of the Upper Jurassic Morrison Formation. The uranium-bearing sandstone is bounded on the north and south by a variable zone of buff and orange sandstone. Within the mineralized zone, the uranium has been remobilized and reconcentrated along the margins of numerous smaller tongues of oxidized rock in a configuration similar to that found in roll-type uranium deposits. In cross section, these small-scale features are zoned; they have an inner, pale orange, oxidized core, a mineralized redox rim cemented with hematite(?), and an outer-shell of -gray, slightly to moderately mineralized rock. The uranium content in the mineralized rock ranges from 0.001 to 0.07 percent U3O8. The uranium, at this locality, is believed to have originated within the Westwater Canyon Member or to have been derived from the overlying Brushy Basin Member. Based on observed outcrop relations, two hypotheses are proposed for explaining the origin of the occurrence. Briefly these hypotheses are: (1) the mineralized zone represents the remnant of an original roll-type uranium deposit, formed during early Eocene time, which has undergone subsequent oxidation with remobilization and redeposition of uranium around the margins of smaller tongues of oxidized rock; and (2) the mineralized zone represents the remnant of an original tabular deposit which has undergone subsequent oxidation with remobilization and redeposition of uranium around the margins of smaller tongues of oxidized rock.

Radiation induced dissolution of UO 2 based nuclear fuel - A critical review of predictive modelling approaches

NASA Astrophysics Data System (ADS)

Eriksen, Trygve E.; Shoesmith, David W.; Jonsson, Mats

2012-01-01

Radiation induced dissolution of uranium dioxide (UO 2) nuclear fuel and the consequent release of radionuclides to intruding groundwater are key-processes in the safety analysis of future deep geological repositories for spent nuclear fuel. For several decades, these processes have been studied experimentally using both spent fuel and various types of simulated spent fuels. The latter have been employed since it is difficult to draw mechanistic conclusions from real spent nuclear fuel experiments. Several predictive modelling approaches have been developed over the last two decades. These models are largely based on experimental observations. In this work we have performed a critical review of the modelling approaches developed based on the large body of chemical and electrochemical experimental data. The main conclusions are: (1) the use of measured interfacial rate constants give results in generally good agreement with experimental results compared to simulations where homogeneous rate constants are used; (2) the use of spatial dose rate distributions is particularly important when simulating the behaviour over short time periods; and (3) the steady-state approach (the rate of oxidant consumption is equal to the rate of oxidant production) provides a simple but fairly accurate alternative, but errors in the reaction mechanism and in the kinetic parameters used may not be revealed by simple benchmarking. It is essential to use experimentally determined rate constants and verified reaction mechanisms, irrespective of whether the approach is chemical or electrochemical.

Visualizing different uranium oxidation states during the surface alteration of uraninite and uranium tetrachloride.

PubMed

Grossmann, Kay; Arnold, Thuro; Steudtner, Robin; Weiss, Stefan; Bernhard, Gert

2009-08-01

Low-temperature alteration reactions on uranium phases may lead to the mobilization of uranium and thereby poses a potential threat to humans living close to uranium-contaminated sites. In this study, the surface alteration of uraninite (UO(2)) and uranium tetrachloride (UCl(4)) in air atmosphere was studied by confocal laser scanning microscopy (CLSM) and laser-induced fluorescence spectroscopy using an excitation wavelength of 408 nm. It was found that within minutes the oxidation state on the surface of the uraninite and the uranium tetrachloride changed. During the surface alteration process U(IV) atoms on the uraninite and uranium tetrachloride surface became stepwise oxidized by a one-electron step at first to U(V) and then further to U(VI). These observed changes in the oxidation states of the uraninite surface were microscopically visualized and spectroscopically identified on the basis of their fluorescence emission signal. A fluorescence signal in the wavelength range of 415-475 nm was indicative for metastable uranium(V), and a fluorescence signal in the range of 480-560 nm was identified as uranium(VI). In addition, the oxidation process of tetravalent uranium in aqueous solution at pH 0.3 was visualized by CLSM and U(V) was fluorescence spectroscopically identified. The combination of microscopy and fluorescence spectroscopy provided a very convincing visualization of the brief presence of U(V) as a metastable reaction intermediate and of the simultaneous coexistence of the three states U(IV), U(V), and U(VI). These results have a significant importance for fundamental uranium redox chemistry and should contribute to a better understanding of the geochemical behavior of uranium in nature.

DISSOLUTION OF ZIRCONIUM AND ALLOYS THEREFOR

DOEpatents

Swanson, J.L.

1961-07-11

The dissolution of zirconium cladding in a water solution of ammonium fluoride and ammonium nitrate is described. The method finds particular utility in processing spent fuel elements for nuclear reactors. The zirconium cladding is first dissolved in a water solution of ammonium fluoride and ammonium nitrate; insoluble uranium and plutonium fiuorides formed by attack of the solvent on the fuel materiai of the fuel element are then separated from the solution, and the fuel materiai is dissolved in another solution.

DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daniel, G.; Rudisill, T.; Almond, P.

The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Sitemore » (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.« less

Uranium(VI) interactions with mackinawite in the presence and absence of bicarbonate and oxygen.

PubMed

Gallegos, Tanya J; Fuller, Christopher C; Webb, Samuel M; Betterton, William

2013-07-02

Mackinawite, Fe(II)S, samples loaded with uranium (10(-5), 10(-4), and 10(-3) mol U/g FeS) at pH 5, 7, and 9, were characterized using X-ray absorption spectroscopy and X-ray diffraction to determine the effects of pH, bicarbonate, and oxidation on uptake. Under anoxic conditions, a 5 g/L suspension of mackinawite lowered 5 × 10(-5) M uranium(VI) to below 30 ppb (1.26 × 10(-7) M) U. Between 82 and 88% of the uranium removed from solution by mackinawite was U(IV) and was nearly completely reduced to U(IV) when 0.012 M bicarbonate was added. Near-neighbor coordination consisting of uranium-oxygen and uranium-uranium distances indicates the formation of uraninite in the presence and absence of bicarbonate, suggesting reductive precipitation as the dominant removal mechanism. Following equilibration in air, mackinawite was oxidized to mainly goethite and sulfur and about 76% of U(IV) was reoxidized to U(VI) with coordination of uranium to axial and equatorial oxygen, similar to uranyl. Additionally, uranium-iron distances, typical of coprecipitation of uranium with iron oxides, and uranium-sulfur distances indicating bidentate coordination of U(VI) to sulfate were evident. The affinity of mackinawite and its oxidation products for U(VI) provides impetus for further study of mackinawite as a potential reactive medium for remediation of uranium-contaminated water.

Reconnaissance for uranium in the coal of Sao Paulo, Santa Catarina, and Rio Grande do Sul, Brazil

USGS Publications Warehouse

Haynes, Donald D.; Pierson, Charles T.; White, Max G.

1958-01-01

Uranium-bearing coal and carbonaceous shale of the Rio Bonito formation of Pennsylvanian age have been found in the States of Sao Paulo, Santa Catarlna and Rio Grande do Sul, Brazil. The uranium oxide content of the samples collected in the State of Sao Paulo ranges from 0.001 percent to 0.082 percent. The samples collected in Santa Catarina averaged about 0.002 percent uranium oxide; those collected in Rio Grande do Sul, about 0.003 percent uranium oxide. Since the field and laboratory investigations are still in their initial stages, only raw data on the radioactivity and uranium content of Brazilian coals are given in this report.

PROCESS FOR PRODUCTION OF URANIUM HEXAFLUORIDE

DOEpatents

Fowler, R.D.

1958-11-01

A process is described for the manufacture of uranium bexafluoride which consists in contacting an oxide of uranium simultaneously with elemental carbon and elemental fluorine at an elevated temperature, using a proportion of the carbon to the oxide about 50% in excess of that theoretically required to combine with f the oxygen as C0/.sub 2/. The process has the advantage that the uranium oxide is reduced by tbe carbon aad converted to the hexafluoride in a single operation.

Chemical aspects of uranium behavior in soils: A review

NASA Astrophysics Data System (ADS)

Vodyanitskii, Yu. N.

2011-08-01

Uranium has varying degrees of oxidation (+4 and +6) and is responsive to changes in the redox potential of the environment. It is deposited at the reduction barrier with the participation of biota and at the sorption barrier under oxidative conditions. Iron (hydr)oxides are the strongest sorbents of uranium. Uranium, being an element of medium biological absorption, can accumulate (relative to thorium) in the humus horizons of some soils. The high content of uranium in uncontaminated soils is most frequently inherited from the parent rocks in the regions of positive U anomalies: in the soils developed on oil shales and in the marginal zone of bogs at the reduction barrier. The development of nuclear and coal-fired power engineering resulted in the environmental contamination with uranium. The immobilization of anthropogenic uranium at artificial geochemical barriers is based on two preconditions: the stimulation of on-site metal-reducing bacteria or the introduction of strong mineral reducers, e.g., Fe at low degrees of oxidation.

PROCESSES OF CHLORINATION OF URANIUM OXIDES

DOEpatents

Rosenfeld, S.

1958-09-16

An improvement is described in the process fur making UCl/sub 4/ from uranium oxide and carbon tetrachloride. In that process, oxides of uranium are contacted with carbon tetrachloride vapor at an elevated temperature. It has been fuund that the reaction product and yield are improved if the uranlum oxide charge is disposed in flat trays in the reaction zone, to a depth of not more than 1/2 centimeter.

SOLVENT EXTRACTION PROCESS FOR URANIUM FROM CHLORIDE SOLUTIONS

DOEpatents

Blake, C.A. Jr.; Brown, K.B.; Horner, D.E.

1960-05-24

An improvement was made in a uranium extraction process wherein the organic extractant is a phosphine oxide. An aqueous solution containing phosphate ions or sulfate ions together with uranium is provided with a source of chloride ions during the extraction step. The presence of the chloride ions enables a phosphine oxide to extract uranium in the presence of strong uranium- complexing ions such as phosphate or sulfate ions.

Molten salt extraction of transuranic and reactive fission products from used uranium oxide fuel

DOEpatents

Herrmann, Steven Douglas

2014-05-27

Used uranium oxide fuel is detoxified by extracting transuranic and reactive fission products into molten salt. By contacting declad and crushed used uranium oxide fuel with a molten halide salt containing a minor fraction of the respective uranium trihalide, transuranic and reactive fission products partition from the fuel to the molten salt phase, while uranium oxide and non-reactive, or noble metal, fission products remain in an insoluble solid phase. The salt is then separated from the fuel via draining and distillation. By this method, the bulk of the decay heat, fission poisoning capacity, and radiotoxicity are removed from the used fuel. The remaining radioactivity from the noble metal fission products in the detoxified fuel is primarily limited to soft beta emitters. The extracted transuranic and reactive fission products are amenable to existing technologies for group uranium/transuranic product recovery and fission product immobilization in engineered waste forms.

Uranium(VI) interactions with mackinawite in the presence and absence of bicarbonate and oxygen

USGS Publications Warehouse

Gallegos, Tanya J.; Fuller, Christopher C.; Webb, Samuel M.; Betterton, William J.

2013-01-01

Mackinawite, Fe(II)S, samples loaded with uranium (10-5, 10-4, and 10-3 mol U/g FeS) at pH 5, 7, and 9, were characterized using X-ray absorption spectroscopy and X-ray diffraction to determine the effects of pH, bicarbonate, and oxidation on uptake. Under anoxic conditions, a 5 g/L suspension of mackinawite lowered 5 × 10-5 M uranium(VI) to below 30 ppb (1.26 × 10-7 M) U. Between 82 and 88% of the uranium removed from solution by mackinawite was U(IV) and was nearly completely reduced to U(IV) when 0.012 M bicarbonate was added. Near-neighbor coordination consisting of uranium–oxygen and uranium-uranium distances indicates the formation of uraninite in the presence and absence of bicarbonate, suggesting reductive precipitation as the dominant removal mechanism. Following equilibration in air, mackinawite was oxidized to mainly goethite and sulfur and about 76% of U(IV) was reoxidized to U(VI) with coordination of uranium to axial and equatorial oxygen, similar to uranyl. Additionally, uranium-iron distances, typical of coprecipitation of uranium with iron oxides, and uranium-sulfur distances indicating bidentate coordination of U(VI) to sulfate were evident. The affinity of mackinawite and its oxidation products for U(VI) provides impetus for further study of mackinawite as a potential reactive medium for remediation of uranium-contaminated water.

Depleted uranium dust from fired munitions: physical, chemical and biological properties.

PubMed

Mitchel, R E J; Sunder, S

2004-07-01

This paper reports physical, chemical and biological analyses of samples of dust resulting from munitions containing depleted uranium (DU) that had been live-fired and had impacted an armored target. Mass spectroscopic analysis indicated that the average atom% of U was 0.198 +/- 0.10, consistent with depleted uranium. Other major elements present were iron, aluminum, and silicon. About 47% of the total mass was particles with diameters <300 microm, of which about 14% was <10 microm. X-ray diffraction analysis indicated that the uranium was present in the sample as uranium oxides-mainly U3O7 (47%), U3O8 (44%) and UO2 (9%). Depleted uranium dust, instilled into the lungs or implanted into the muscle of rats, contained a rapidly soluble uranium component and a more slowly soluble uranium component. The fraction that underwent dissolution in 7 d declined exponentially with increasing initial burden. At the lower lung burdens tested (<15 microg DU dust/lung) about 14% of the uranium appeared in urine within 7 d. At the higher lung burdens tested (~80-200 microg DU dust/lung) about 5% of the DU appeared in urine within 7 d. In both cases about 50% of that total appeared in urine within the first day. DU implanted in muscle similarly showed that about half of the total excreted within 7 d appeared in the first day. At the lower muscle burdens tested (<15 microg DU dust/injection site) about 9% was solubilized within 7 d. At muscle burdens >35 microg DU dust/injection site about 2% appeared in urine within 7 d. Natural uranium (NU) ore dust was instilled into rat lungs for comparison. The fraction dissolving in lung showed a pattern of exponential decline with increasing initial burden similar to DU. However, the decline was less steep, with about 14% appearing in urine for lung burdens up to about 200 microg NU dust/lung and 5% at lung burdens >1,100 microg NU dust/lung. NU also showed both a fast and a more slowly dissolving component. At the higher lung burdens of both DU and NU that showed lowered urine excretion rates, histological evidence of kidney damage was seen. Kidney damage was not seen with the muscle burdens tested. DU dust produced kidney damage at lower lung burdens and lower urine uranium levels than NU dust, suggesting that other toxic metals in DU dust may contribute to the damage.

Thermochemistry of rare earth doped uranium oxides LnxU1-xO2-0.5x+y (Ln = La, Y, Nd)

NASA Astrophysics Data System (ADS)

Zhang, Lei; Navrotsky, Alexandra

2015-10-01

Lanthanum, yttrium, and neodymium doped uranium dioxide samples in the fluorite structure have been synthesized, characterized in terms of metal ratio and oxygen content, and their enthalpies of formation measured by high temperature oxide melt solution calorimetry. For oxides doped with 10-50 mol % rare earth (Ln) cations, the formation enthalpies from constituent oxides (LnO1.5, UO2 and UO3 in a reaction not involving oxidation or reduction) become increasingly exothermic with increasing rare earth content, while showing no significant dependence on the varying uranium oxidation state. The oxidation enthalpy of LnxU1-xO2-0.5x+y is similar to that of UO2 to UO3 for all three rare earth doped systems. Though this may suggest that the oxidized uranium in these systems is energetically similar to that in the hexavalent state, thermochemical data alone can not constrain whether the uranium is present as U5+, U6+, or a mixture of oxidation states. The formation enthalpies from elements calculated from the calorimetric data are generally consistent with those from free energy measurements.

Oxygen potential of uranium--plutonium oxide as determined by controlled- atmosphere thermogravimetry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swanson, Gerald C.

1975-10-01

The oxygen-to-metal atom ratio, or O/M, of solid solution uranium- plutonium oxide reactor fuel is a measure of the concentration of crystal defects in the oxide which affect many fuel properties, particularly, fuel oxygen potential. Fabrication of a high-temperature oxygen electrode, employing an electro-active tip of oxygen-deficient solid-state electrolyte, intended to confirm gaseous oxygen potentials is described. Uranium oxide and plutonium oxide O/M reference materials were prepared by in situ oxidation of high purity metals in the thermobalance. A solid solution uranium-plutonium oxide O/M reference material was prepared by alloying the uranium and plutonium metals in a yttrium oxide cruciblemore » at 1200°C and oxidizing with moist He at 250°C. The individual and solid solution oxides were isothermally equilibrated with controlled oxygen potentials between 800 and 1300°C and the equilibrated O/ M ratios calculated with corrections for impurities and buoyancy effects. Use of a reference oxygen potential of -100 kcal/mol to produce an O/M of 2.000 is confirmed by these results. However, because of the lengthy equilibration times required for all oxides, use of the O/M reference materials rather than a reference oxygen potential is recommended for O/M analysis methods calibrations.« less

RECONDITIONING FUEL ELEMENTS

DOEpatents

Brandt, H.L.

1962-02-20

A process is given for decanning fuel elements that consist of a uranium core, an intermediate section either of bronze, silicon, Al-Si, and uranium silicide layers or of lead, Al-Si, and uranium silicide layers around said core, and an aluminum can bonded to said intermediate section. The aluminum can is dissolved in a solution of sodium hydroxide (9 to 20 wt%) and sodium nitrate (35 to 12 wt %), and the layers of the intermediate section are dissolved in a boiling sodium hydroxide solution of a minimum concentration of 50 wt%. (AEC) A method of selectively reducing plutonium oxides and the rare earth oxides but not uranium oxides is described which comprises placing the oxides in a molten solvent of zinc or cadmium and then adding metallic uranium as a reducing agent. (AEC)

Water table dynamics and biogeochemical cycling in a shallow, variably-saturated floodplain

DOE PAGES

Yabusaki, Steven B.; Wilkins, Michael J.; Fang, Yilin; ...

2017-02-20

Three-dimensional variably saturated flow and multicomponent biogeochemical reactive transport modeling, based on published and newly generated data, is used to better understand the interplay of hydrology, geochemistry, and biology controlling the cycling of carbon, nitrogen, oxygen, iron, sulfur, and uranium in a shallow floodplain. In this system, aerobic respiration generally maintains anoxic groundwater below an oxic vadose zone until seasonal snowmelt-driven water table peaking transports dissolved oxygen (DO) and nitrate from the vadose zone into the alluvial aquifer. The response to this perturbation is localized due to distinct physico-biogeochemical environments and relatively long time scales for transport through the floodplainmore » aquifer and vadose zone. Naturally reduced zones (NRZs) containing sediments higher in organic matter, iron sulfides, and non-crystalline U(IV) rapidly consume DO and nitrate to maintain anoxic conditions, yielding Fe(II) from FeS oxidative dissolution, nitrite from denitrification, and U(VI) from nitrite-promoted U(IV) oxidation. Redox cycling is a key factor for sustaining the observed aquifer behaviors despite continuous oxygen influx and the annual hydrologically induced oxidation event. Furthermore, depth-dependent activity of fermenters, aerobes, nitrate reducers, sulfate reducers, and chemolithoautotrophs (e.g., oxidizing Fe(II), S compounds, and ammonium) is linked to the presence of DO, which has higher concentrations near the water table.« less

Electrochemical and spectroscopic evidence on the one-electron reduction of U(VI) to U(V) on magnetite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yuan, Ke; Ilton, Eugene S.; Antonio, Mark R.

2015-05-19

Reduction of U(VI) to U(IV) on mineral surfaces has been considered as a one-step two electron process. However, stabilized U(V), with no evidence of U(IV), found in recent studies indicates U(VI) can undergo a one electron reduction to U(V) without further progression to U(IV). We investigated the mechanisms of uranium reduction by reducing U(VI) electrochemically on a magnetite electrode at pH 3.4 . The one electron reduction of U(VI) was first confirmed using the cyclic voltammetry method. Formation of nano-size uranium precipitates on the surface of magnetite at reducing potentials and dissolution of the solids at oxidizing potentials were observedmore » by in situ electrochemical AFM. XPS analysis of the magnetite electrodes polarized in uranium solutions at voltages from 0.1 ~ 0.9 V (vs. Ag/AgCl) showed the presence of only U(V) and U(VI). The highest amount of U(V) relative to U(VI) was prepared at 0.7 V, where the longest average U–Oaxial distance of 2.05 ± 0.01 Å was evident in the same sample revealed by EXAFS analysis. The results demonstrate that the electrochemical reduction of U(VI) on magnetite only yields U(V), even at a potential of 0.9 V, which favors the one-electron reduction mechanism. U(V) did not disproportionate but stabilized on magnetite through precipitation of mixed-valence state U(VI)/U(V) solids.« less

Pentavalent uranium trans-dihalides and -pseudohalides.

PubMed

Lewis, Andrew J; Nakamaru-Ogiso, Eiko; Kikkawa, James M; Carroll, Patrick J; Schelter, Eric J

2012-05-21

Pentavalent uranium complexes of the formula U(V)X(2)[N(SiMe(3))(2)](3) (X = F(-), Cl(-), Br(-), N(3)(-), NCS(-)) are accessible from the oxidation of U(III)[N(SiMe(3))(2)](3) through two sequential, one-electron oxidation reactions (halides) and substitution through salt metathesis (pseudohalides). Uranium(v) mixed-halides are also synthesized by successive one-electron oxidation reactions.

PREPARATION OF REFRACTORY OXIDE CRYSTALS

DOEpatents

Grimes, W.R.; Shaffer, J.H.; Watson, G.M.

1962-11-13

A method is given for preparing uranium dioxide, thorium oxide, and beryllium oxide in the form of enlarged individual crystals. The surface of a fused alkali metal halide melt containing dissolved uranium, thorium, or beryllium values is contacted with a water-vapor-bearing inert gas stream at a rate of 5 to 10 cubic centimeters per minute per square centimeter of melt surface area. Growth of individual crystals is obtained by prolonged contact. Beryllium oxide-coated uranium dioxide crystals are prepared by disposing uranium dioxide crystals 5 to 20 microns in diameter in a beryllium-containing melt and contacting the melt with a water-vapor-bearing inert gas stream in the same manner. (AEC)

Method for fluorination of uranium oxide

DOEpatents

Petit, George S.

1987-01-01

Highly pure uranium hexafluoride is made from uranium oxide and fluorine. The uranium oxide, which includes UO.sub.3, UO.sub.2, U.sub.3 O.sub.8 and mixtures thereof, is introduced together with a small amount of a fluorine-reactive substance, selected from alkali chlorides, silicon dioxide, silicic acid, ferric oxide, and bromine, into a constant volume reaction zone. Sufficient fluorine is charged into the zone at a temperature below approximately 0.degree. C. to provide an initial pressure of at least approximately 600 lbs/sq. in. at the ambient atmospheric temperature. The temperature is then allowed to rise in the reaction zone until reaction occurs.

METHOD OF SEPARATING Pu FROM METATHESIZED BiPO$sub 4$ CARRIER

DOEpatents

Knox, W.J.; Thompson, S.G.

1960-05-31

A process is given for separating uranium, neptunium, and/or plutonium from a bismuth hydroxide carrier by selective dissolution of these actinides with nitric acid of a concentration of from 0.05 to 0.5N.

Influence of uranium hydride oxidation on uranium metal behaviour

DOE Office of Scientific and Technical Information (OSTI.GOV)

Patel, N.; Hambley, D.; Clarke, S.A.

2013-07-01

This work addresses concerns that the rapid, exothermic oxidation of active uranium hydride in air could stimulate an exothermic reaction (burning) involving any adjacent uranium metal, so as to increase the potential hazard arising from a hydride reaction. The effect of the thermal reaction of active uranium hydride, especially in contact with uranium metal, does not increase in proportion with hydride mass, particularly when considering large quantities of hydride. Whether uranium metal continues to burn in the long term is a function of the uranium metal and its surroundings. The source of the initial heat input to the uranium, ifmore » sufficient to cause ignition, is not important. Sustained burning of uranium requires the rate of heat generation to be sufficient to offset the total rate of heat loss so as to maintain an elevated temperature. For dense uranium, this is very difficult to achieve in naturally occurring circumstances. Areas of the uranium surface can lose heat but not generate heat. Heat can be lost by conduction, through contact with other materials, and by convection and radiation, e.g. from areas where the uranium surface is covered with a layer of oxidised material, such as burned-out hydride or from fuel cladding. These rates of heat loss are highly significant in relation to the rate of heat generation by sustained oxidation of uranium in air. Finite volume modelling has been used to examine the behaviour of a magnesium-clad uranium metal fuel element within a bottle surrounded by other un-bottled fuel elements. In the event that the bottle is breached, suddenly, in air, it can be concluded that the bulk uranium metal oxidation reaction will not reach a self-sustaining level and the mass of uranium oxidised will likely to be small in relation to mass of uranium hydride oxidised. (authors)« less

Release behavior of uranium in uranium mill tailings under environmental conditions.

PubMed

Liu, Bo; Peng, Tongjiang; Sun, Hongjuan; Yue, Huanjuan

2017-05-01

Uranium contamination is observed in sedimentary geochemical environments, but the geochemical and mineralogical processes that control uranium release from sediment are not fully appreciated. Identification of how sediments and water influence the release and migration of uranium is critical to improve the prevention of uranium contamination in soil and groundwater. To understand the process of uranium release and migration from uranium mill tailings under water chemistry conditions, uranium mill tailing samples from northwest China were investigated with batch leaching experiments. Results showed that water played an important role in uranium release from the tailing minerals. The uranium release was clearly influenced by contact time, liquid-solid ratio, particle size, and pH under water chemistry conditions. Longer contact time, higher liquid content, and extreme pH were all not conducive to the stabilization of uranium and accelerated the uranium release from the tailing mineral to the solution. The values of pH were found to significantly influence the extent and mechanisms of uranium release from minerals to water. Uranium release was monitored by a number of interactive processes, including dissolution of uranium-bearing minerals, uranium desorption from mineral surfaces, and formation of aqueous uranium complexes. Considering the impact of contact time, liquid-solid ratio, particle size, and pH on uranium release from uranium mill tailings, reducing the water content, decreasing the porosity of tailing dumps and controlling the pH of tailings were the key factors for prevention and management of environmental pollution in areas near uranium mines. Copyright © 2017 Elsevier Ltd. All rights reserved.

High Temperature Reactions of Uranium Dioxide with Various Metal Oxides

DTIC Science & Technology

1956-02-20

manganese, nickel , lead, and tin. Subtracting the total of these impurities from the oxygen remainder would give a more nearly 1:2 uranium -oxygen ratio. The...Astin, Dire~ctor High -Temperature Reactions of Uranium Dioxide With Various Metal Oxides Acceson For NTIS CRAWI DTfC TAB Unannounced D JustifiCation...1 2. The uranium -oxygen system ------------------------------------- 1 3. Binary systems containing

Actinide metal processing

DOEpatents

Sauer, Nancy N.; Watkin, John G.

1992-01-01

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

Actinide metal processing

DOEpatents

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

1992-03-24

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

Quantification of kinetic rate law parameters for the dissolution of natural autunite in the presence of aqueous bicarbonate ions at high concentrations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gudavalli, Ravi; Katsenovich, Yelena; Wellman, Dawn

Uranium is a key contaminant of concern in the groundwater at 91 waste sites at 18 U.S. Department of Energy (DOE) facilities within the United States and is a potential source of groundwater contamination and a risk to human health and the environment through discharges to surface water. Dissolved inorganic carbon (bicarbonate/carbonate) has a high affinity for complexing with uranium that is present as sorbed or unique uranium-bearing mineral phases within the sedimentary matrix. This process can result in the formation of soluble uranyl carbonate aqueous species, which are mobile under circumneutral pH conditions. This study was conducted to quantifymore » the rate of release of uranium from the autunite mineral, (Ca[(UO 2)(PO 4)] 2∙3H 2O), that was formed during polyphosphate injection to remediate uranium; the dissolution of uranium was studied as a function of the aqueous bicarbonate concentration, ranging from 25 to 100 mM. Experiments were carried out in the pH range from 7 to 11 in the temperature range of 23-90°C via single-pass flow-through testing. Consistent with the results of previous studies (Gudavalli et al., 2013 a, b), the rate of uranium release from autunite exhibited minimal dependency on temperature, but was strongly dependent on pH and increasing concentrations of bicarbonate in the solution. Data obtained during these experiments were compared with results of previous experiments conducted using a low-concentration range of bicarbonate solutions (0.5-3.0 mM). An 8- to 30 fold increase in the rate of uranium release was observed in the presence of high bicarbonate concentrations at pH 7-8 compared to low bicarbonate values, while at pH 9-11, there was only a 5-fold increase in uranium rate of release with an increase in bicarbonate concentrations. The rate of uranium release was calculated to be between 5.18 x 10 -8 and 1.69 x 10 -7 mol m -2 s -1. The activation energy values at high and low bicarbonate concentrations were similar, with ratio values in the range of 0.6-1.0.« less

Quantification of kinetic rate law parameters for the dissolution of natural autunite in the presence of aqueous bicarbonate ions at high concentrations.

PubMed

Gudavalli, Ravi; Katsenovich, Yelena; Wellman, Dawn

2018-05-02

Uranium is a key contaminant of concern in the groundwater at U.S. Department of Energy (DOE) facilities within the United States and is a potential source of groundwater contamination and a risk to human health and the environment through discharges to surface water. Dissolved inorganic carbon (bicarbonate/carbonate) has a high affinity for complexing with uranium that is present as sorbed or unique uranium-bearing mineral phases within the sedimentary matrix. This process can result in the formation of soluble uranyl carbonate aqueous species, which are mobile under circumneutral pH conditions. This study was conducted to quantify the rate of release of uranium from the autunite mineral, (Ca[(UO 2 )(PO 4 )] 2 •3H 2 O), that was formed during polyphosphate injection to remediate uranium; the dissolution of uranium was studied as a function of the aqueous bicarbonate concentration, ranging from 25 to 100 mM. Experiments were carried out in the pH range from 7 to 11 in the temperature range of 23-90 °C via single-pass flow-through testing. Consistent with the results of previous studies (Gudavalli et al., 2013a, 2013b), the rate of uranium release from autunite exhibited minimal dependency on temperature, but was strongly dependent on pH and increasing concentrations of bicarbonate in the solution. Data obtained during these experiments were compared with results of previous experiments conducted using a low-concentration range of bicarbonate solutions (0.5-3.0 mM). An 8- to 30-fold increase in the rate of uranium release was observed in the presence of high bicarbonate concentrations at pH 7-8 compared to low bicarbonate values, while at pH 9-11, there was only a 5-fold increase in uranium rate of release with an increase in bicarbonate concentrations. The rate of uranium release was calculated to be between 5.18 × 10 -8 and 1.69 × 10 -7  mol m -2 s -1 . The activation energy values at high and low bicarbonate concentrations were similar, with ratio values in the range of 0.6-1.0. Copyright © 2018 Elsevier Ltd. All rights reserved.

Process for continuous production of metallic uranium and uranium alloys

DOEpatents

Hayden, H.W. Jr.; Horton, J.A.; Elliott, G.R.B.

1995-06-06

A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO{sub 3}), or any other substantially stable uranium oxide, to form the uranium dioxide (UO{sub 2}). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl{sub 4}), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation. 4 figs.

Process for continuous production of metallic uranium and uranium alloys

DOEpatents

Hayden, Jr., Howard W.; Horton, James A.; Elliott, Guy R. B.

1995-01-01

A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO.sub.3), or any other substantially stable uranium oxide, to form the uranium dioxide (UO.sub.2). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl.sub.4), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation.

First-principles study on oxidation effects in uranium oxides and high-pressure high-temperature behavior of point defects in uranium dioxide

NASA Astrophysics Data System (ADS)

Geng, Hua Y.; Song, Hong X.; Jin, K.; Xiang, S. K.; Wu, Q.

2011-11-01

Formation Gibbs free energy of point defects and oxygen clusters in uranium dioxide at high-pressure high-temperature conditions are calculated from first principles, using the LSDA+U approach for the electronic structure and the Debye model for the lattice vibrations. The phonon contribution on Frenkel pairs is found to be notable, whereas it is negligible for the Schottky defect. Hydrostatic compression changes the formation energies drastically, making defect concentrations depend more sensitively on pressure. Calculations show that, if no oxygen clusters are considered, uranium vacancy becomes predominant in overstoichiometric UO2 with the aid of the contribution from lattice vibrations, while compression favors oxygen defects and suppresses uranium vacancy greatly. At ambient pressure, however, the experimental observation of predominant oxygen defects in this regime can be reproduced only in a form of cuboctahedral clusters, underlining the importance of defect clustering in UO2+x. Making use of the point defect model, an equation of state for nonstoichiometric oxides is established, which is then applied to describe the shock Hugoniot of UO2+x. Furthermore, the oxidization and compression behavior of uranium monoxide, triuranium octoxide, uranium trioxide, and a series of defective UO2 at 0 K are investigated. The evolution of mechanical properties and electronic structures with an increase of the oxidation degree are analyzed, revealing the transition of the ground state of uranium oxides from metallic to Mott insulator and then to charge-transfer insulator due to the interplay of strongly correlated effects of 5f orbitals and the shift of electrons from uranium to oxygen atoms.

Ligand-controlled Fe mobilization catalyzed by adsorbed Fe(II) on Fe(hydr)oxides

NASA Astrophysics Data System (ADS)

Kang, Kyounglim; Biswakarma, Jagannath; Borowski, Susan C.; Hug, Stephan J.; Hering, Janet G.; Schenkeveld, Walter D. C.; Kraemer, Stephan M.

2017-04-01

Dissolution of Fe(hydr)oxides is a key process in biological iron acquisition. Due to the low solubility of iron oxides in environments with a circumneutral pH, organisms may exude organic compounds catalyzing iron mobilization by reductive and ligand controlled dissolution mechanisms. Recently, we have shown synergistic effects between reductive dissolution and ligand-controlled dissolution that may operate in biological iron acquisition. The synergistic effects were observed in Fe mobilization from single goethite suspensions as well as in suspensions containing calcareous soil[1],[2]. However, how the redox reaction accelerates Fe(hydr)oxide dissolution by ligands is not studied intensively. In our study, we hypothesized that electron transfer to structural Fe(III) labilizes the Fe(hydr)oxide structure, and that this can accelerate ligand controlled dissolution. Systematical batch dissolution experiments were carried out under anoxic conditions at environmentally relevant pH values in which various Fe(hydr)oxides (goethite, hematite, lepidocrocite) interacted with two different types of ligand (desferrioxamine B (DFOB) and N,N'-Di(2-hydroxybenzyl)ethylenediamine-N,N'-diacetic acid monohydrochloride (HBED)). Electron transfer to the structure was induced by adsorbing Fe(II) to the mineral surface at various Fe(II) concentrations. Our results show a distinct catalytic effect of adsorbed Fe(II) on ligand controlled dissolution, even at submicromolar Fe(II) concentrations. We observed the effect for a range of iron oxides, but it was strongest in lepidocrocite, most likely due to anisotropy in conductivity leading to higher near-surface concentration of reduced iron. Our results demonstrate that the catalytic effect of reductive processes on ligand controlled dissolution require a very low degree of reduction making this an efficient process for biological iron acquisition and a potentially important effect in natural iron cycling. References 1. Wang, Z. M.; Schenkeveld, W. D. C.; Kraemer, S. M.; Giammar, D. E. Environ. Sci. Technol. 2015, 49, (12), 7236-7244. 2. Schenkeveld, W. D. C.; Wang, Z. M.; Giammar, D. E.; Kraemer, S. M. Environ. Sci. Technol. 2016, 50, (12), 6381-6388.

Developing uranium dicarbide-graphite porous materials for the SPES project

NASA Astrophysics Data System (ADS)

Biasetto, L.; Zanonato, P.; Carturan, S.; Di Bernardo, P.; Colombo, P.; Andrighetto, A.; Prete, G.

2010-09-01

Uranium carbide dispersed in graphite was produced under vacuum by means of carbothermic reduction of different uranium oxides (UO 2, U 3O 8 and UO 3), using graphite as the source of carbon. The thermal process was monitored by mass spectrometry and the gas evolution confirmed the reduction of the U 3O 8 and UO 3 oxides to UO 2 before the carbothermic reaction, that started to occur at T > 1000 °C. XRD analysis confirmed the formation of α-UC 2 and of a minor amount of UC. The morphology of the produced uranium carbide was not affected by the oxides employed as the source of uranium.

REFRACTORY ARTICLE AND PROCESS OF MANUFACTURING SAME

DOEpatents

Hamilton, N.E.

1957-12-10

A method is described for fabricating improved uranium oxide crucibles. In the past, such crucibles have lacked mechanical strength due to the poor cohesion of the uranium oxide particles. This difficulty has now been overcome by admixing with the uranium oxide a quantity of a refractory oxide binder, and dry pressing and sintering the resulting mixture into the desired shape. Suitable as binders are BeO, CaO, Al/sub 2/C/sub 3/, and ThO/sub 2/ among others.

METHOD OF DISSOLVING REFRACTORY ALLOYS

DOEpatents

Helton, D.M.; Savolainen, J.K.

1963-04-23

This patent relates to the dissolution of alloys of uranium with zirconium, thorium, molybdenum, or niobium. The alloy is contacted with an anhydrous solution of mercuric chloride in a low-molecular-weight monohydric alcohol to produce a mercury-containing alcohol slurry. The slurry is then converted to an aqueous system by adding water and driving off the alcohol. The resulting aqueous slurry is electrolyzed in the presence of a mercury cathode to remove the mercury and produce a uranium-bearing aqueous solution. This process is useful for dissolving irradiated nuclear reactor fuels for radiochemical reprocessing by solvent extraction. In addition, zirconium-alloy cladding is selectively removed from uranium dioxide fuel compacts by this means. (AEC)

Oxidation and crystal field effects in uranium

NASA Astrophysics Data System (ADS)

Tobin, J. G.; Yu, S.-W.; Booth, C. H.; Tyliszczak, T.; Shuh, D. K.; van der Laan, G.; Sokaras, D.; Nordlund, D.; Weng, T.-C.; Bagus, P. S.

2015-07-01

An extensive investigation of oxidation in uranium has been pursued. This includes the utilization of soft x-ray absorption spectroscopy, hard x-ray absorption near-edge structure, resonant (hard) x-ray emission spectroscopy, cluster calculations, and a branching ratio analysis founded on atomic theory. The samples utilized were uranium dioxide (U O2) , uranium trioxide (U O3) , and uranium tetrafluoride (U F4) . A discussion of the role of nonspherical perturbations, i.e., crystal or ligand field effects, will be presented.

Potential Aquifer Vulnerability in Regions Down-Gradient from ...

EPA Pesticide Factsheets

Sandstone-hosted roll-front uranium ore deposits originate when U(VI) dissolved in groundwater is reduced and precipitated as insoluble U(IV) minerals. Groundwater redox geochemistry, aqueous complexation, and solute migration are instrumental in leaching uranium from source rocks and transporting it in low concentrations to a chemical redox interface where it is deposited in an ore zone typically containing the uranium minerals uraninite, pitchblende, and/or coffinite; various iron sulfides; native selenium; clays; and calcite. In situ recovery (ISR) of these uranium ores is a process of contacting the uranium mineral deposit with leaching (lixiviant) fluids via injection of the lixiviant into wells drilled into the subsurface aquifer that hosts uranium ore, while other extraction wells pump the dissolved uranium after dissolution of the uranium minerals. Environmental concerns during and after ISR include water quality impacts from: 1) potential excursions of leaching solutions away from the injection zone into down-dip, underlying, or overlying aquifers; 2) potential migration of uranium and its decay products (e.g., Ra, Rn, Pb); and, 3) potential migration of redox-sensitive trace metals (e.g., Fe, Mn, Mo, Se, V), metalloids (e.g., As), and anions (e.g., sulfate). This review describes the geochemical processes that control roll-front uranium transport and fate in groundwater systems, identifies potential aquifer vulnerabilities to ISR operations, identifies

Separation of uranium from (U, Th)O 2 and (U, Pu)O 2 by solid state reactions route

NASA Astrophysics Data System (ADS)

Keskar, Meera; Mudher, K. D. Singh; Venugopal, V.

2005-01-01

Solid state reactions of UO 2, ThO 2, PuO 2 and their mixed oxides (U, Th)O 2 and (U, Pu)O 2 were carried out with sodium nitrate upto 900 °C, to study the formation of various phases at different temperatures, which are amenable for easy dissolution and separation of the actinide elements in dilute acid. Products formed by reacting unsintered as well as sintered UO 2 with NaNO 3 above 500 °C were readily soluble in 2 M HNO 3, whereas ThO 2 and PuO 2 did not react with NaNO 3 to form any soluble products. Thus reactions of mixed oxides (U, Th)O 2 and (U, Pu)O 2 with NaNO 3 were carried out to study the quantitative separation of U from (U, Th)O 2 and (U, Pu)O 2. X-ray diffraction, X-ray fluorescence, thermal analysis and chemical analysis techniques were used for the characterization of the products formed during the reactions.

Study the oxidation kinetics of uranium using XRD and Rietveld method

NASA Astrophysics Data System (ADS)

Zhang, Yanzhi; Guan, Weijun; Wang, Qinguo; Wang, Xiaolin; Lai, Xinchun; Shuai, Maobing

2010-03-01

The surface oxidation of uranium metal has been studied by X-ray diffraction (XRD) and Rietveld method in the range of 50~300°C in air. The oxidation processes are analyzed by XRD to determine the extent of surface oxidation and the oxide structure. The dynamics expression for the formation of UO2 was derived. At the beginning, the dynamic expression was nonlinear, but switched to linear subsequently for uranium in air and humid oxygen. That is, the growth kinetics of UO2 can be divided into two stages: nonlinear portion and linear portion. Using the kinetic data of linear portion, the activation energy of reaction between uranium and air was calculated about 46.0 kJ/mol. However the content of oxide as a function of time was linear in humid helium ambience. Contrast the dynamics results, it prove that the absence of oxygen would accelerate the corrosion rate of uranium in the humid gas. We can find that the XRD and Rietveld method are a useful convenient method to estimate the kinetics and thermodynamics of solid-gas reaction.

Conceptual Model of Uranium in the Vadose Zone for Acidic and Alkaline Wastes Discharged at the Hanford Site Central Plateau

DOE Office of Scientific and Technical Information (OSTI.GOV)

Truex, Michael J.; Szecsody, James E.; Qafoku, Nikolla

2014-09-01

Historically, uranium was disposed in waste solutions of varying waste chemistry at the Hanford Site Central Plateau. The character of how uranium was distributed in the vadose zone during disposal, how it has continued to migrate through the vadose zone, and the magnitude of potential impacts on groundwater are strongly influenced by geochemical reactions in the vadose zone. These geochemical reactions can be significantly influenced by the disposed-waste chemistry near the disposal location. This report provides conceptual models and supporting information to describe uranium fate and transport in the vadose zone for both acidic and alkaline wastes discharged at amore » substantial number of waste sites in the Hanford Site Central Plateau. The conceptual models include consideration of how co-disposed acidic or alkaline fluids influence uranium mobility in terms of induced dissolution/precipitation reactions and changes in uranium sorption with a focus on the conditions near the disposal site. This information, when combined with the extensive information describing uranium fate and transport at near background pH conditions, enables focused characterization to support effective fate and transport estimates for uranium in the subsurface.« less

Covalency in oxidized uranium

NASA Astrophysics Data System (ADS)

Tobin, J. G.; Yu, S.-W.; Qiao, R.; Yang, W. L.; Booth, C. H.; Shuh, D. K.; Duffin, A. M.; Sokaras, D.; Nordlund, D.; Weng, T.-C.

2015-07-01

Using x-ray emission spectroscopy and absorption spectroscopy, it has been possible to directly access the states in the unoccupied conduction bands that are involved with 5 f and 6 d covalency in oxidized uranium. By varying the oxidizing agent, the degree of 5 f covalency can be manipulated and monitored, clearly and irrevocably establishing the importance of 5 f covalency in the electronic structure of the key nuclear fuel, uranium dioxide.

Covalency in oxidized uranium

DOE PAGES

Tobin, J. G.; Yu, S. -W.; Qiao, R.; ...

2015-07-01

Here, using x-ray emission spectroscopy and absorption spectroscopy, it has been possible to directly access the states in the unoccupied conduction bands that are involved with 5f and 6d covalency in oxidized uranium. By varying the oxidizing agent, the degree of 5f covalency can be manipulated and monitored, clearly and irrevocably establishing the importance of 5f covalency in the electronic structure of the key nuclear fuel, uranium dioxide.

DISSOLUTION METHOD OF REMOVING BONDING AGENTS

DOEpatents

Hyman, H.H.

1960-04-19

A method is given for removing residual aluminumsilicon bonding agents from uranium slugs after the removal of aluminum coatings. To accomplish this the slug is immersed in an aqueous solution about 0.75 N in hydrofluoric acid and about 7 N in nitric acid.

Analytical Capability of Plasma Spectrometry Team

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gallimore, David L.

2012-07-19

Samples analyzed were: (1) Pu and U metal; (2) Pu oxide for nuclear fuel; (3) {sup 238}Pu oxide for heat source; and (4) Nuclear forensic samples - filters, swipes. Sample preparations that we did were: metal dissolution, marple filter dissolution, Pu oxide closed vessel acid digestion, and column separation to remove Pu.

Carbon in oxides and silicates - Dissolution versus exsolution

NASA Technical Reports Server (NTRS)

Freund, F.

1986-01-01

A theory of CO2 dissolution in the solid state is developed, using the idea proposed by Freund (1983) concerning dissolution of CO/CO2 in MgO on the basis of their experimental results obtained with an MgO-containing carbon impurity. It is shown that the dissolution mechanism may be linked to an internal redox reaction by which a certain number of lattice oxygens change their formal oxidation state from -2 to -1, while the carbon becomes reduced. The similarities between the mechanisms of CO and/or CO2 dissolution and that of H2O dissolution are pointed out. A hypothesis is proposed concerning the exsolution of reduced carbon from supersaturated solid solutions under conditions which permit C-C bond formation.

Simultaneous determination of uranium carbide dissolution products by capillary zone electrophoresis.

PubMed

Sladkov, Vladimir; Fourest, Blandine

2009-03-20

Separation and simultaneous determination of a number of organic acid anions (oxalate, mellitate, trimellitate and benzoate) and U(VI) with direct UV detection is developed for analysis of uranium carbide (UC) dissolution products by capillary zone electrophoresis (CZE). Reverse polarity mode is used. It is found that complex formation of U(VI) with carbonate, used as a carrier electrolyte, allows U(VI) to be determined, as negatively charged species, in a single run with organic acid anions. Some parameters such as pH value, composition of electrolyte and detection wavelength are optimized. Under the chosen conditions (carbonate buffer (ionic strength of 100 mM), pH 9.8, 0.15 mM tetradecyltrimethylammonium bromide (TTAB)) a complete separation is achieved. Calibration plots are linear in two ranges of concentration for U(VI) ( approximately 1 x 10(-5) to 1 x 10(-3)), mellitate and trimellitate ( approximately 5 x 10(-6) to 5 x 10(-4)), and about one range ( approximately 1 x 10(-4) to 5 x 10(-3)) for oxalate and benzoate. Accuracy of the procedure is checked by the "added-found" method in standard mixture solutions. Relative standard deviation is within the range of 2-10% and the recovery is in the range of 90-110%. This method is applied for the analysis of real UC dissolution samples.

K-Basin gel formation studies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beck, M.A.

1998-07-23

A key part of the proposed waste treatment for K Basin sludge is the elimination of reactive uranium metal by dissolution in nitric acid (Fkirnent, 1998). It has been found (Delegard, 1998a) that upon nitric acid dissolution of the sludge, a gel sometimes forms. Gels are known to sometimes impair solid/liquid separation and/or material transfer. The purpose of the work reported here is to determine the cause(s) of the gel formation and to determine operating parameters for the sludge dissolution that avoid formation of gel. This work and related work were planned in (Fkunent, 1998), (Jewett, 1998) and (Beck, 1998a).more » This report describes the results of the tests in (Beck, 1998a) with non-radioactive surrogates.« less

Laser fluorometric analysis of plants for uranium exploration

USGS Publications Warehouse

Harms, T.F.; Ward, F.N.; Erdman, J.A.

1981-01-01

A preliminary test of biogeochemical exploration for locating uranium occurrences in the Marfa Basin, Texas, was conducted in 1978. Only 6 of 74 plant samples (mostly catclaw mimosa, Mimosa biuncifera) contained uranium in amounts above the detection limit (0.4 ppm in the ash) of the conventional fluorometric method. The samples were then analyzed using a Scintrex UA-3 uranium analyzer* * Use of trade names in this paper is for descriptive purposes only and does not constitute endorsement by the U.S. Geological Survey. - an instrument designed for direct analysis of uranium in water, and which can be conveniently used in a mobile field laboratory. The detection limit for uranium in plant ash (0.05 ppm) by this method is almost an order of magnitude lower than with the fluorometric conventional method. Only 1 of the 74 samples contained uranium below the detection limit of the new method. Accuracy and precision were determined to be satisfactory. Samples of plants growing on mineralized soils and nonmineralized soils show a 15-fold difference in uranium content; whereas the soils themselves (analyzed by delayed neutron activation analysis) show only a 4-fold difference. The method involves acid digestion of ashed tissue, extraction of uranium into ethyl acetate, destruction of the ethyl acetate, dissolution of the residue in 0.005% nitric acid, and measurement. ?? 1981.

Freezing-Enhanced Dissolution of Iron Oxides: Effects of Inorganic Acid Anions.

PubMed

Jeong, Daun; Kim, Kitae; Min, Dae Wi; Choi, Wonyong

2015-11-03

Dissolution of iron from mineral dust particles greatly depends upon the type and amount of copresent inorganic anions. In this study, we investigated the roles of sulfate, chloride, nitrate, and perchlorate on the dissolution of maghemite and lepidocrocite in ice under both dark and UV irradiation and compared the results with those of their aqueous counterparts. After 96 h of reaction, the total dissolved iron in ice (pH 3 before freezing) was higher than that in the aqueous phase (pH 3) by 6-28 times and 10-20 times under dark and UV irradiation, respectively. Sulfuric acid was the most efficient in producing labile iron under dark condition, whereas hydrochloric acid induced the most dissolution of the total and ferrous iron in the presence of light. This ice-induced dissolution result was also confirmed with Arizona Test Dust (AZTD). In the freeze-thaw cycling test, the iron oxide samples containing chloride, nitrate, or perchlorate showed a similar extent of total dissolved iron after each cycling while the sulfate-containing sample rapidly lost its dissolution activity with repeating the cycle. This unique phenomenon observed in ice might be related to the freeze concentration of protons, iron oxides, and inorganic anions in the liquid-like ice grain boundary region. These results suggest that the ice-enhanced dissolution of iron oxides can be a potential source of bioavailable iron, and the acid anions critically influence this process.

Effects of pretreatment processes for Zr electrorefining of oxidized Zircaloy-4 cladding tubes

NASA Astrophysics Data System (ADS)

Hwa Lee, Chang; Lee, Yoo Lee; Jeon, Min Ku; Choi, Yong Taek; Kang, Kweon Ho; Park, Geun Il

2014-06-01

The effect of pretreatment processes for the Zr electrorefining of oxidized Zircaloy-4 cladding tubes is examined in LiCl-KCl-ZrCl4 molten salts at 500 °C. The cyclic voltammetries reveal that the Zr dissolution kinetics is highly dependent on the thickness of a Zr oxide layer formed at 500 °C under air atmosphere. For the Zircaloy-4 tube covered with a 1 μm thick oxide layer, the Zr dissolution process is initiated from a non-stoichiometric Zr oxide surface through salt treatment at an open circuit potential in the molten salt electrolyte. The Zr dissolution of the samples in the middle range of oxide layer thickness appears to be more effectively derived by the salt treatment coupled with an anodic potential application at an oxidation potential of Zr. A modification of the process scheme offers an applicability of Zr electrorefining for the treatment of oxidized cladding hull wastes.

Summary of the mineralogy of the Colorado Plateau uranium ores

USGS Publications Warehouse

Weeks, Alice D.; Coleman, Robert Griffin; Thompson, Mary E.

1956-01-01

In the Colorado Plateau uranium has been produced chiefly from very shallow mines in carnotite ores (oxidized vanadiferous uranium ores) until recent deeper mining penetrated black unoxidized ores in water-saturated rocks and extensive exploration has discovered many deposits of low to nonvanadiferous ores. The uranium ores include a wide range from highly vanadiferous and from as much as one percent to a trace of copper, and contain a small amount of iron and traces of lead, zinc, molybdenum, cobalt, nickel, silver, manganese, and other metals. Recent investigation indicates that the carnotite ores have been derived by progressive oxidation of primary (unoxidized) black ores that contain low-valent uranium and vanadium oxides and silicates. The uranium minerals, uraninite and coffinite, are associated with coalified wood or other carbonaceous material. The vanadium minerals, chiefly montroseite, roscoelite, and other vanadium silicates, occur in the interstices of the sandstone and in siltstone and clay pellets as well as associated with fossil wood. Calcite, dolomite, barite and minor amounts of sulfides, arsenides, and selenides occur in the unoxidized ore. Partially oxidized vanadiferous ore is blue black, purplish brown, or greenish black in contrast to the black or dark gray unoxidized ore. Vanadium combines with uranium to form rauvite. The excess vanadium is present in corvusite, fernandinite, melanovanadite and many other quadrivalent and quinquevalent vanadium minerals as well as in vanadium silicates. Pyrite and part or all of the calcite are replaced by iron oxides and gypsum. In oxidized vanadiferous uranium ores the uranium is fixed in the relatively insoluble minerals carnotite and tyuyamunite, and the excess vanadium commonly combines with one or more of the following: calcium, sodium, potassium, magnesium, aluminum, iron, copper, manganese, or barium, or rarely it forms the hydrated pentoxide. The relatively stable vanadium silicates are little affected by oxidation. The unoxidized nonvanadiferous ores contain uraninite and coffinite in close association with coalified wood and iron and copper sulfides, and traces of many other sulfides, arsenides and selenides. The oxidized nonvanadiferous ores differ from the vanadiferous ores because, in the absence of vanadium to complex the uranium, a great variety of secondary yellow and greenish-yellow uranyl minerals are formed. The uranyl sulfates and carbonates are more common than the oxides, phosphates, arsenates, and silicates. Because the sulfates and carbonates are much less stable that carnotite, the oxidized nonvanadiferous ores occure only as halos around cores of unoxidized ore and do not form large oxidized deposits close to the surface of the ground as carnotite ores. Oxidation has taken place since the lowering of the water table in the present erosion cycle. Because of local structures and the highly lenticular character of the fluviatile host rocks perched water tables and water-saturated lenses of sandstone are common high above the regional water table. Unoxidized ore has been preserved in these water-saturated rocks and the boundary between oxidized and unoxidized ore is very irregular.

Carbothermic synthesis of 820 μm uranium nitride kernels: Literature review, thermodynamics, analysis, and related experiments

NASA Astrophysics Data System (ADS)

Lindemer, T. B.; Voit, S. L.; Silva, C. M.; Besmann, T. M.; Hunt, R. D.

2014-05-01

The US Department of Energy is developing a new nuclear fuel that would be less susceptible to ruptures during a loss-of-coolant accident. The fuel would consist of tristructural isotropic coated particles with uranium nitride (UN) kernels with diameters near 825 μm. This effort explores factors involved in the conversion of uranium oxide-carbon microspheres into UN kernels. An analysis of previous studies with sufficient experimental details is provided. Thermodynamic calculations were made to predict pressures of carbon monoxide and other relevant gases for several reactions that can be involved in the conversion of uranium oxides and carbides into UN. Uranium oxide-carbon microspheres were heated in a microbalance with an attached mass spectrometer to determine details of calcining and carbothermic conversion in argon, nitrogen, and vacuum. A model was derived from experiments on the vacuum conversion to uranium oxide-carbide kernels. UN-containing kernels were fabricated using this vacuum conversion as part of the overall process. Carbonitride kernels of ∼89% of theoretical density were produced along with several observations concerning the different stages of the process.

Impact of biostimulated redox processes on metal dynamics in an iron-rich creek soil of a former uranium mining area.

PubMed

Burkhardt, Eva-Maria; Akob, Denise M; Bischoff, Sebastian; Sitte, Jana; Kostka, Joel E; Banerjee, Dipanjan; Scheinost, Andreas C; Küsel, Kirsten

2010-01-01

Understanding the dynamics of metals and radionuclides in soil environments is necessary for evaluating risks to pristine sites. An iron-rich creek soil of a former uranium-mining district (Ronneburg, Germany) showed high porewater concentrations of heavy metals and radionuclides. Thus, this study aims to (i) evaluate metal dynamics during terminal electron accepting processes (TEAPs) and (ii) characterize active microbial populations in biostimulated soil microcosms using a stable isotope probing (SIP) approach. In biostimulated soil slurries, concentrations of soluble Co, Ni, Zn, As, and unexpectedly U increased during Fe(III)-reduction. This suggests that there was a release of sorbed metals and As during reductive dissolution of Fe(III)-oxides. Subsequent sulfate-reduction was concurrent with a decrease of U, Co, Ni, and Zn concentrations. The relative contribution of U(IV) in the solid phase changed from 18.5 to 88.7% after incubation. The active Fe(III)-reducing population was dominated by delta-Proteobacteria (Geobacter) in (13)C-ethanol amended microcosms. A more diverse community was present in (13)C-lactate amended microcosms including taxa related to Acidobacteria, Firmicutes, delta-Proteobacteria, and beta-Proteobacteria. Our results suggested that biostimulated Fe(III)-reducing communities facilitated the release of metals including U to groundwater which is in contrast to other studies.

PROCESSES FOR SEPARATING AND RECOVERING CONSTITUENTS OF NEUTRON IRRADIATED URANIUM

DOEpatents

Connick, R.E.; Gofman, J.W.; Pimentel, G.C.

1959-11-10

Processes are described for preparing plutonium, particularly processes of separating plutonium from uranium and fission products in neutron-irradiated uraniumcontaining matter. Specifically, plutonium solutions containing uranium, fission products and other impurities are contacted with reducing agents such as sulfur dioxide, uranous ion, hydroxyl ammonium chloride, hydrogen peroxide, and ferrous ion whereby the plutoninm is reduced to its fluoride-insoluble state. The reduced plutonium is then carried out of solution by precipitating niobic oxide therein. Uranium and certain fission products remain behind in the solution. Certain other fission products precipitate along with the plutonium. Subsequently, the plutonium and fission product precipitates are redissolved, and the solution is oxidized with oxidizing agents such as chlorine, peroxydisulfate ion in the presence of silver ion, permanganate ion, dichromate ion, ceric ion, and a bromate ion, whereby plutonium is oxidized to the fluoride-soluble state. The oxidized solution is once again treated with niobic oxide, thus precipitating the contamirant fission products along with the niobic oxide while the oxidized plutonium remains in solution. Plutonium is then recovered from the decontaminated solution.

Photoreductive dissolution of iron oxides trapped in ice and its environmental implications.

PubMed

Kim, Kitae; Choi, Wonyong; Hoffmann, Michael R; Yoon, Ho-Il; Park, Byong-Kwon

2010-06-01

The availability of iron has been thought to be a main limiting factor for the productivity of phytoplankton and related with the uptake of atmospheric CO(2) and algal blooms in fresh and sea waters. In this work, the formation of bioavailable iron (Fe(II)(aq)) from the dissolution of iron oxide particles was investigated in the ice phase under both UV and visible light irradiation. The photoreductive dissolution of iron oxides proceeded slowly in aqueous solution (pH 3.5) but was significantly accelerated in polycrystalline ice, subsequently releasing more bioavailable ferrous iron upon thawing. The enhanced photogeneration of Fe(II)(aq) in ice was confirmed regardless of the type of iron oxides [hematite, maghemite (gamma-Fe(2)O(3)), goethite (alpha-FeOOH)] and the kind of electron donors. The ice-enhanced dissolution of iron oxides was also observed under visible light irradiation, although the dissolution rate was much slower compared with the case of UV radiation. The iron oxide particles and organic electron donors (if any) in ice are concentrated and aggregated in the liquid-like grain boundary region (freeze concentration effect) where protons are also highly concentrated (lower pH). The enhanced photodissolution of iron oxides should occur in this confined boundary region. We hypothesized that electron hopping through the interconnected grain boundaries of iron oxide particles facilitates the separation of photoinduced charge pairs. The outdoor experiments carried out under ambient solar radiation of Ny-Alesund (Svalbard, 78 degrees 55'N) also showed that the generation of dissolved Fe(II)(aq) via photoreductive dissolution is enhanced when iron oxides are trapped in ice. Our results imply that the ice(snow)-covered surfaces and ice-cloud particles containing iron-rich mineral dusts in the polar and cold environments provide a source of bioavailable iron when they thaw.

Simulation of uranium and plutonium oxides compounds obtained in plasma

NASA Astrophysics Data System (ADS)

Novoselov, Ivan Yu.; Karengin, Alexander G.; Babaev, Renat G.

2018-03-01

The aim of this paper is to carry out thermodynamic simulation of mixed plutonium and uranium oxides compounds obtained after plasma treatment of plutonium and uranium nitrates and to determine optimal water-salt-organic mixture composition as well as conditions for their plasma treatment (temperature, air mass fraction). Authors conclude that it needs to complete the treatment of nitric solutions in form of water-salt-organic mixtures to guarantee energy saving obtainment of oxide compounds for mixed-oxide fuel and explain the choice of chemical composition of water-salt-organic mixture. It has been confirmed that temperature of 1200 °C is optimal to practice the process. Authors have demonstrated that condensed products after plasma treatment of water-salt-organic mixture contains targeted products (uranium and plutonium oxides) and gaseous products are environmental friendly. In conclusion basic operational modes for practicing the process are showed.

Depleted uranium as a backfill for nuclear fuel waste package

DOEpatents

Forsberg, Charles W.

1998-01-01

A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

Depleted uranium as a backfill for nuclear fuel waste package

DOEpatents

Forsberg, C.W.

1998-11-03

A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

A Multireference Density Functional Approach to the Calculation of the Excited States of Uranium Ions

DTIC Science & Technology

2007-03-01

approach. xiv A MULTIREFERENCE DENSITY FUNCTIONAL APPROACH TO THE CALCULATION OF THE EXCITED STATES OF URANIUM IONS I. Introduction Actinide chemistry, in...oxidation state of the uranium atom. Uranium, like most early actinides , can possess a wide range of oxidation states, ranging from +3 to +6, due in part...in predicting the electronic spectra for heavy element compounds. The first difficulty is that relativistic effects for actinides are significant

Study of oxidative stress related responses induced in Arabidopsis thaliana following mixed exposure to uranium and cadmium.

PubMed

Vanhoudt, Nathalie; Vandenhove, Hildegarde; Horemans, Nele; Wannijn, Jean; Bujanic, Andelko; Vangronsveld, Jaco; Cuypers, Ann

2010-01-01

In this study, toxicity effects in plants of uranium in a binary pollution condition were investigated by studying biological responses and unraveling oxidative stress related mechanisms in Arabidopsis thaliana seedlings, grown on hydroponics and exposed for 3 days to 10 μM uranium in combination with 5 μM cadmium. While uranium mostly accumulated in the roots with very low root-to-shoot transport, cadmium was taken up less by the roots but showed higher translocation to the shoots. Under mixed exposure, cadmium influenced uranium uptake highly but not the other way round resulting in a doubled uranium concentration in the roots. Under our mixed exposure conditions, it is clear that micronutrient concentrations in the roots are strongly influenced by addition of cadmium as a second stressor, while leaf macronutrient concentrations are mostly influenced by uranium. Oxidative stress related responses are highly affected by cadmium while uranium influence is more limited. Hereby, an important role was attributed to the ascorbate redox balance together with glutathione as both metabolites, but more explicitly for ascorbate, increased their reduced form, indicating an important defense and regulatory function. While for roots, based on an increase in FSD1 gene expression, oxidative stress was suggested to be superoxide induced, in leaves on the other hand, hydrogen peroxide related genes were mostly altered. Copyright © 2010 Elsevier Masson SAS. All rights reserved.

Electrochemical and Spectroscopic Evidence on the One-Electron Reduction of U(VI) to U(V) on Magnetite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yuan, Ke; Ilton, Eugene S.; Antonio, Mark R.

2015-05-19

Reduction of U(VI) to U(VI) on mineral surfaces is often considered a one-step two-electron process. However, stabilized U(V), with no evidence of U(IV), found in recent studies Indicates U(VI) can undergo a one-electron reduction to U(V) without further progression to U(VI),. We investigated reduction pathways of uranium by reducing U(VI) electrochemically on a, magnetite electrode at,pH 3.4. Cyclic voltammetry confirms the one-electron reduction of U(VI) . Formation of nanosize uranium precipitates on the magnetite surface at reducing potentials and dissolution of the solids at oxidizing potentials are observed by in situ electrochemical atomic force microscopy. XPS, analysis Of the magnetitemore » electrodes polarized in uranium solutions at voltages - from -0.1 to -0.9 V (E-U(VI)/U(V)(0)= -0.135 V vs Ag/AgCl) show the presence of, only U(V) and U(VI). The sample with the highest U(V)/U(VI) ratio was prepared at -0.7 V, where the longest average U-O-axial distance of 2.05 + 0.01 A was evident in the same sample revealed by extended X-ray absorption fine structure analysis. The results demonstrate that the electrochemical reduction of U(VI) On magnetite only yields,U(V), even at a potential of -0.9 V, which favors the one-electron reduction mechanism, U(V) does not disproportionate but stabilizes on magnetite through precipitation Of mixed-valence state -U(V)/U(VI) solids.« less

Formulations for iron oxides dissolution

DOEpatents

Horwitz, Earl P.; Chiarizia, Renato

1992-01-01

A mixture of a di- or polyphosphonic acid and a reductant wherein each is present in a sufficient amount to provide a synergistic effect with respect to the dissolution of metal oxides and optionally containing corrosion inhibitors and pH adjusting agents.

An unusual temperature dependence in the oxidation of oxycarbide layers on uranium

NASA Astrophysics Data System (ADS)

Ellis, Walton P.

1981-09-01

An anomalous temperature dependence has been observed for the oxidation kinetics of outermost oxycarbide layers on polycrystalline uranium metal. Normally, oxidation or corrosion reactions are expected to proceed more rapidly as the temperature is elevated. Thus, it came as a surprise when we observed that the removal of the outermost atomic layers of carbon from uranium oxycarbide by O 2 reproducibly proceeds at a much faster rate at 25°C than at 280°C.

An unusual temperature dependence in the oxidation of oxycarbide layers on uranium

NASA Astrophysics Data System (ADS)

Ellis, Walton P.

An anomalous temperature dependence has been observed for the oxidation kinetics of outermost oxycarbide layers on polycrystalline uranium metal. Normally, oxidation or corrosion reactions are expected to proceed more rapidly as the temperature is elevated. Thus, it came as a surprise when we observed that the removal of the outermost atomic layers of carbon from uranium oxycarbide by O 2 reproducibly proceeds at a much faster rate at 25°C than at 280°C.

Radium and uranium concentrations and associated hydrogeochemistry in ground water in southwestern Pueblo County, Colorado

USGS Publications Warehouse

Felmlee, J. Karen; Cadigan, Robert Allen

1979-01-01

Radium and uranium concentrations in water from 37 wells tapping the aquifer system of the Dakota Sandstone and Purgatoire Formation in southwestern Pueblo County, Colorado, have a wide range of values and define several areas of high radioactivity in the ground water. Radium ranges from 0.3 to 420 picocuries per liter and has a median value of 8.8, and uranium ranges from 0.02 to 180 micrograms per liter and has a median value of 2.4. Radon concentrations, measured in 32 of the 37 wells, range from less than 100 picocuries per liter to as much as 27,000 and have a median value of 580. Relationships among the radioactive elements and 28 other geochemical parameters were studied by using correlation coefficients and R-mode factor analysis. Five factor groups were determined to represent major influences on water chemistry: (1) short-term solution reactions, (2) oxidation reactions, (3) hydrolysis reactions, (4) uranium distribution, and (5) long-term solution reactions. Uranium concentrations are most strongly influenced by oxidation reactions but also are affected by solution reactions and distribution of uranium in the rocks of the aquifer system. Radon and radium concentrations are mostly controlled by uranium distribution; radium also shows a moderate negative relationship with oxidation. To explain the statistical and spatial relationships among the parameters, a model was developed involving the selective leaching of uranium-bearing phases and metal sulfides which occur in discontinuous zones in sandstone and shale. When reducing conditions prevail, uranium is immobile, but radium can be taken into solution. When faults and associated fractured rocks allow oxidizing conditions to dominate, uranium can be taken into solution; radium can also be taken into solution, or it may become immobilized by coprecipitation with iron and manganese oxides or with barite. Several areas within the study area are discussed in terms of the model.

Pyrophoric behaviour of uranium hydride and uranium powders

NASA Astrophysics Data System (ADS)

Le Guyadec, F.; Génin, X.; Bayle, J. P.; Dugne, O.; Duhart-Barone, A.; Ablitzer, C.

2010-01-01

Thermal stability and spontaneous ignition conditions of uranium hydride and uranium metal fine powders have been studied and observed in an original and dedicated experimental device placed inside a glove box under flowing pure argon. Pure uranium hydride powder with low amount of oxide (<0.5 wt.%) was obtained by heat treatment at low temperature in flowing Ar/5%H2. Pure uranium powder was obtained by dehydration in flowing pure argon. Those fine powders showed spontaneous ignition at room temperature in air. An in situ CCD-camera displayed ignition associated with powder temperature measurement. Characterization of powders before and after ignition was performed by XRD measurements and SEM observations. Oxidation mechanisms are proposed.

In situ mobility of uranium in the presence of nitrate following sulfate-reducing conditions.

PubMed

Paradis, Charles J; Jagadamma, Sindhu; Watson, David B; McKay, Larry D; Hazen, Terry C; Park, Melora; Istok, Jonathan D

2016-04-01

Reoxidation and mobilization of previously reduced and immobilized uranium by dissolved-phase oxidants poses a significant challenge for remediating uranium-contaminated groundwater. Preferential oxidation of reduced sulfur-bearing species, as opposed to reduced uranium-bearing species, has been demonstrated to limit the mobility of uranium at the laboratory scale yet field-scale investigations are lacking. In this study, the mobility of uranium in the presence of nitrate oxidant was investigated in a shallow groundwater system after establishing conditions conducive to uranium reduction and the formation of reduced sulfur-bearing species. A series of three injections of groundwater (200 L) containing U(VI) (5 μM) and amended with ethanol (40 mM) and sulfate (20 mM) were conducted in ten test wells in order to stimulate microbial-mediated reduction of uranium and the formation of reduced sulfur-bearing species. Simultaneous push-pull tests were then conducted in triplicate well clusters to investigate the mobility of U(VI) under three conditions: 1) high nitrate (120 mM), 2) high nitrate (120 mM) with ethanol (30 mM), and 3) low nitrate (2 mM) with ethanol (30 mM). Dilution-adjusted breakthrough curves of ethanol, nitrate, nitrite, sulfate, and U(VI) suggested that nitrate reduction was predominantly coupled to the oxidation of reduced-sulfur bearing species, as opposed to the reoxidation of U(IV), under all three conditions for the duration of the 36-day tests. The amount of sulfate, but not U(VI), recovered during the push-pull tests was substantially more than injected, relative to bromide tracer, under all three conditions and further suggested that reduced sulfur-bearing species were preferentially oxidized under nitrate-reducing conditions. However, some reoxidation of U(IV) was observed under nitrate-reducing conditions and in the absence of detectable nitrate and/or nitrite. This suggested that reduced sulfur-bearing species may not be fully effective at limiting the mobility of uranium in the presence of dissolved and/or solid-phase oxidants. The results of this field study confirmed those of previous laboratory studies which suggested that reoxidation of uranium under nitrate-reducing conditions can be substantially limited by preferential oxidation of reduced sulfur-bearing species. Copyright © 2016 The Authors. Published by Elsevier B.V. All rights reserved.

In situ mobility of uranium in the presence of nitrate following sulfate-reducing conditions

DOE PAGES

Paradis, Charles J.; Jagadamma, Sindhu; Watson, David B.; ...

2016-02-11

Reoxidation and mobilization of previously reduced and immobilized uranium by dissolved phase oxidants poses a significant challenge for remediating uranium-contaminated groundwater. Preferential oxidation of reduced sulfur-bearing species, as opposed to reduced uranium bearing species, has been demonstrated to limit the mobility of uranium at the laboratory scale yet field-scale investigations are lacking. Here in this study, the mobility of uranium in the presence of nitrate oxidant was investigated in a shallow groundwater system after establishing conditions conducive to uranium reduction and the formation of reduced sulfur-bearing species. A series of three injections of groundwater (200 L) containing U(VI) (5 μM)more » and amended with ethanol (40 mM) and sulfate (20 mM) were conducted in ten test wells in order to stimulate microbial mediated reduction of uranium and the formation of reduced sulfur-bearing species. Simultaneous push-pull tests were then conducted in triplicate well clusters to investigate the mobility of U(VI) under three conditions: 1) high nitrate (120 mM), 2) high nitrate (120 mM) with ethanol (30 mM), and 3) low nitrate (2 mM) with ethanol (30 mM). Dilution-adjusted breakthrough curves of ethanol, nitrate, nitrite, sulfate, and U(VI) suggested that nitrate reduction was predominantly coupled to the oxidation of reduced-sulfur bearing species, as opposed to the reoxidation of U(IV), under all three conditions for the duration of the 36-day tests. The amount of sulfate, but not U(VI), recovered during the push-pull tests was substantially more than injected, relative to bromide tracer, under all three conditions and further suggested that reduced sulfur-bearing species were preferentially oxidized under nitrate-reducing conditions. However, some reoxidation of U(IV) was observed under nitrate-reducing conditions and in the absence of detectable nitrate and/or nitrite. This suggested that reduced sulfur-bearing species may not be fully effective at limiting the mobility of uranium in the presence of dissolved and/or solid-phase oxidants. Lastly, the results of this field study confirmed those of previous laboratory studies which suggested that reoxidation of uranium under nitrate-reducing conditions can be substantially limited by preferential oxidation of reduced sulfur-bearing species.« less

Evaluation of Non-Oxide Fuel for Fission-based Nuclear Reactors on Spacecraft

DTIC Science & Technology

smaller and potentially lighter core, whichis a significant advantage. The results of this study indicate that use of both UC and UN may result in significant weight savings due tohigher uranium loading density....The goal of this project was to study the performance of atypical uranium-based fuels in a nuclear reactor capable of producing 1 megawattof thermal...UN), or uranium carbide (UC) and compared their performance to uranium oxide (UO2) which is thefuel form used in the vast majority of commercial

METHOD OF PRODUCING URANIUM METAL BY ELECTROLYSIS

DOEpatents

Piper, R.D.

1962-09-01

A process is given for making uranium metal from oxidic material by electrolytic deposition on the cathode. The oxidic material admixed with two moles of carbon per one mole of uranium dioxide forms the anode, and the electrolyte is a mixture of from 40 to 75% of calcium fluoride or barium fluoride, 15 to 45% of uranium tetrafluoride, and from 10 to 20% of lithium fluoride or magnesium fluoride; the temperature of the electrolyte is between 1150 and 1175 deg C. (AEC)

Process for electroslag refining of uranium and uranium alloys

DOEpatents

Lewis, P.S. Jr.; Agee, W.A.; Bullock, J.S. IV; Condon, J.B.

1975-07-22

A process is described for electroslag refining of uranium and uranium alloys wherein molten uranium and uranium alloys are melted in a molten layer of a fluoride slag containing up to about 8 weight percent calcium metal. The calcium metal reduces oxides in the uranium and uranium alloys to provide them with an oxygen content of less than 100 parts per million. (auth)

Proteomic analysis reveals contrasting stress response to uranium in two nitrogen-fixing Anabaena strains, differentially tolerant to uranium.

PubMed

Panda, Bandita; Basu, Bhakti; Acharya, Celin; Rajaram, Hema; Apte, Shree Kumar

2017-01-01

Two strains of the nitrogen-fixing cyanobacterium Anabaena, native to Indian paddy fields, displayed differential sensitivity to exposure to uranyl carbonate at neutral pH. Anabaena sp. strain PCC 7120 and Anabaena sp. strain L-31 displayed 50% reduction in survival (LD 50 dose), following 3h exposure to 75μM and 200μM uranyl carbonate, respectively. Uranium responsive proteome alterations were visualized by 2D gel electrophoresis, followed by protein identification by MALDI-ToF mass spectrometry. The two strains displayed significant differences in levels of proteins associated with photosynthesis, carbon metabolism, and oxidative stress alleviation, commensurate with their uranium tolerance. Higher uranium tolerance of Anabaena sp. strain L-31 could be attributed to sustained photosynthesis and carbon metabolism and superior oxidative stress defense, as compared to the uranium sensitive Anabaena sp. strain PCC 7120. Uranium responsive proteome modulations in two nitrogen-fixing strains of Anabaena, native to Indian paddy fields, revealed that rapid adaptation to better oxidative stress management, and maintenance of metabolic and energy homeostasis underlies superior uranium tolerance of Anabaena sp. strain L-31 compared to Anabaena sp. strain PCC 7120. Copyright © 2016 Elsevier B.V. All rights reserved.

An Investigation on the Persistence of Uranium Hydride during Storage of Simulant Nuclear Waste Packages.

PubMed

Stitt, C A; Harker, N J; Hallam, K R; Paraskevoulakos, C; Banos, A; Rennie, S; Jowsey, J; Scott, T B

2015-01-01

Synchrotron X-rays have been used to study the oxidation of uranium and uranium hydride when encapsulated in grout and stored in de-ionised water for 10 months. Periodic synchrotron X-ray tomography and X-ray powder diffraction have allowed measurement and identification of the arising corrosion products and the rates of corrosion. The oxidation rates of the uranium metal and uranium hydride were slower than empirically derived rates previously reported for each reactant in an anoxic water system, but without encapsulation in grout. This was attributed to the grout acting as a physical barrier limiting the access of oxidising species to the uranium surface. Uranium hydride was observed to persist throughout the 10 month storage period and industrial consequences of this observed persistence are discussed.

An Investigation on the Persistence of Uranium Hydride during Storage of Simulant Nuclear Waste Packages

PubMed Central

Harker, N. J.; Hallam, K. R.; Paraskevoulakos, C.; Banos, A.; Rennie, S.; Jowsey, J.

2015-01-01

Synchrotron X-rays have been used to study the oxidation of uranium and uranium hydride when encapsulated in grout and stored in de-ionised water for 10 months. Periodic synchrotron X-ray tomography and X-ray powder diffraction have allowed measurement and identification of the arising corrosion products and the rates of corrosion. The oxidation rates of the uranium metal and uranium hydride were slower than empirically derived rates previously reported for each reactant in an anoxic water system, but without encapsulation in grout. This was attributed to the grout acting as a physical barrier limiting the access of oxidising species to the uranium surface. Uranium hydride was observed to persist throughout the 10 month storage period and industrial consequences of this observed persistence are discussed. PMID:26176551

Standoff Detection of Uranium and its Isotopes by Femtosecond Filament Laser Ablation Molecular Isotopic Spectrometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hartig, Kyle C.; Ghebregziabher, Isaac; Jovanovic, Igor

The ability to perform not only elementally but also isotopically sensitive detection and analysis at standoff distances is important for remote sensing applications in diverse ares, such as nuclear nonproliferation, environmental monitoring, geophysics, and planetary science. We demonstrate isotopically sensitive real-time standoff detection of uranium by the use of femtosecond filament-induced laser ablation molecular isotopic spectrometry. A uranium oxide molecular emission isotope shift of 0.05 ± 0.007 nm is reported at 593.6 nm. We implement both spectroscopic and acoustic diagnostics to characterize the properties of uranium plasma generated at different filament- uranium interaction points. The resulting uranium oxide emission exhibitsmore » a nearly constant signal-to-background ratio over the length of the filament, unlike the uranium atomic and ionic emission, for which the signal-to-background ratio varies significantly along the filament propagation. This is explained by the different rates of increase of plasma density and uranium oxide density along the filament length resulting from spectral and temporal evolution of the filament along its propagation. Lastly, the results provide a basis for the optimal use of filaments for standoff detection and analysis of uranium isotopes and indicate the potential of the technique for a wider range of remote sensing applications that require isotopic sensitivity.« less

Standoff Detection of Uranium and its Isotopes by Femtosecond Filament Laser Ablation Molecular Isotopic Spectrometry

PubMed Central

Hartig, Kyle C.; Ghebregziabher, Isaac; Jovanovic, Igor

2017-01-01

The ability to perform not only elementally but also isotopically sensitive detection and analysis at standoff distances is impor-tant for remote sensing applications in diverse ares, such as nuclear nonproliferation, environmental monitoring, geophysics, and planetary science. We demonstrate isotopically sensitive real-time standoff detection of uranium by the use of femtosecond filament-induced laser ablation molecular isotopic spectrometry. A uranium oxide molecular emission isotope shift of 0.05 ± 0.007 nm is reported at 593.6 nm. We implement both spectroscopic and acoustic diagnostics to characterize the properties of uranium plasma generated at different filament-uranium interaction points. The resulting uranium oxide emis-sion exhibits a nearly constant signal-to-background ratio over the length of the filament, unlike the uranium atomic and ionic emission, for which the signal-to-background ratio varies significantly along the filament propagation. This is explained by the different rates of increase of plasma density and uranium oxide density along the filament length resulting from spectral and temporal evolution of the filament along its propagation. The results provide a basis for the optimal use of filaments for standoff detection and analysis of uranium isotopes and indicate the potential of the technique for a wider range of remote sensing applications that require isotopic sensitivity. PMID:28272450

Standoff Detection of Uranium and its Isotopes by Femtosecond Filament Laser Ablation Molecular Isotopic Spectrometry

DOE PAGES

Hartig, Kyle C.; Ghebregziabher, Isaac; Jovanovic, Igor

2017-03-08

The ability to perform not only elementally but also isotopically sensitive detection and analysis at standoff distances is important for remote sensing applications in diverse ares, such as nuclear nonproliferation, environmental monitoring, geophysics, and planetary science. We demonstrate isotopically sensitive real-time standoff detection of uranium by the use of femtosecond filament-induced laser ablation molecular isotopic spectrometry. A uranium oxide molecular emission isotope shift of 0.05 ± 0.007 nm is reported at 593.6 nm. We implement both spectroscopic and acoustic diagnostics to characterize the properties of uranium plasma generated at different filament- uranium interaction points. The resulting uranium oxide emission exhibitsmore » a nearly constant signal-to-background ratio over the length of the filament, unlike the uranium atomic and ionic emission, for which the signal-to-background ratio varies significantly along the filament propagation. This is explained by the different rates of increase of plasma density and uranium oxide density along the filament length resulting from spectral and temporal evolution of the filament along its propagation. Lastly, the results provide a basis for the optimal use of filaments for standoff detection and analysis of uranium isotopes and indicate the potential of the technique for a wider range of remote sensing applications that require isotopic sensitivity.« less

Standoff Detection of Uranium and its Isotopes by Femtosecond Filament Laser Ablation Molecular Isotopic Spectrometry

NASA Astrophysics Data System (ADS)

Hartig, Kyle C.; Ghebregziabher, Isaac; Jovanovic, Igor

2017-03-01

The ability to perform not only elementally but also isotopically sensitive detection and analysis at standoff distances is impor-tant for remote sensing applications in diverse ares, such as nuclear nonproliferation, environmental monitoring, geophysics, and planetary science. We demonstrate isotopically sensitive real-time standoff detection of uranium by the use of femtosecond filament-induced laser ablation molecular isotopic spectrometry. A uranium oxide molecular emission isotope shift of 0.05 ± 0.007 nm is reported at 593.6 nm. We implement both spectroscopic and acoustic diagnostics to characterize the properties of uranium plasma generated at different filament-uranium interaction points. The resulting uranium oxide emis-sion exhibits a nearly constant signal-to-background ratio over the length of the filament, unlike the uranium atomic and ionic emission, for which the signal-to-background ratio varies significantly along the filament propagation. This is explained by the different rates of increase of plasma density and uranium oxide density along the filament length resulting from spectral and temporal evolution of the filament along its propagation. The results provide a basis for the optimal use of filaments for standoff detection and analysis of uranium isotopes and indicate the potential of the technique for a wider range of remote sensing applications that require isotopic sensitivity.

PROCESS FOR REMOVING NOBLE METALS FROM URANIUM

DOEpatents

Knighton, J.B.

1961-01-31

A pyrometallurgical method is given for purifying uranium containing ruthenium and palladium. The uranium is disintegrated and oxidized by exposure to air and then the ruthenium and palladium are extracted from the uranium with molten zinc.

Direct observation of pure pentavalent uranium in U2O5 thin films by high resolution photoemission spectroscopy.

PubMed

Gouder, T; Eloirdi, R; Caciuffo, R

2018-05-29

Thin films of the elusive intermediate uranium oxide U 2 O 5 have been prepared by exposing UO 3 precursor multilayers to atomic hydrogen. Electron photoemission spectra measured about the uranium 4f core-level doublet contain sharp satellites separated by 7.9(1) eV from the 4f main lines, whilst satellites characteristics of the U(IV) and U(VI) oxidation states, expected respectively at 6.9(1) and 9.7(1) eV from the main 4f lines, are absent. This shows that uranium ions in the films are in a pure pentavalent oxidation state, in contrast to previous investigations of binary oxides claiming that U(V) occurs only as a metastable intermediate state coexisting with U(IV) and U(VI) species. The ratio between the 5f valence band and 4f core-level uranium photoemission intensities decreases by about 50% from UO 2 to U 2 O 5 , which is consistent with the 5f  2 (UO 2 ) and 5f  1 (U 2 O 5 ) electronic configurations of the initial state. Our studies conclusively establish the stability of uranium pentoxide.

Dissolution flowsheet for high flux isotope reactor fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Foster, T.

2016-09-27

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U 3O 8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H 2. The HFIR fuel cores will be dissolved and the recovered U will be down-blendedmore » into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H 2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H 2 and other permanent gases in the dissolution offgas allowing the development of H 2 generation rate profiles. The H 2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the calculated lower flammability limit (LFL) for H 2 at a given Hg concentration.« less

Different pattern of brain pro-/anti-oxidant activity between depleted and enriched uranium in chronically exposed rats.

PubMed

Lestaevel, P; Romero, E; Dhieux, B; Ben Soussan, H; Berradi, H; Dublineau, I; Voisin, P; Gourmelon, P

2009-04-05

Uranium is not only a heavy metal but also an alpha particle emitter. The main toxicity of uranium is expected to be due to chemiotoxicity rather than to radiotoxicity. Some studies have demonstrated that uranium induced some neurological disturbances, but without clear explanations. A possible mechanism of this neurotoxicity could be the oxidative stress induced by reactive oxygen species imbalance. The aim of the present study was to determine whether a chronic ingestion of uranium induced anti-oxidative defence mechanisms in the brain of rats. Rats received depleted (DU) or 4% enriched (EU) uranyl nitrate in the drinking water at 2mg(-1)kg(-1)day(-1) for 9 months. Cerebral cortex analyses were made by measuring mRNA and protein levels and enzymatic activities. Lipid peroxidation, an oxidative stress marker, was significantly enhanced after EU exposure, but not after DU. The gene expression or activity of the main antioxidant enzymes, i.e. superoxide dismutase (SOD), catalase (CAT) and glutathione peroxidase (GPx), increased significantly after chronic exposure to DU. On the contrary, oral EU administration induced a decrease of these antioxidant enzymes. The NO-ergic pathway was almost not perturbed by DU or EU exposure. Finally, DU exposure increased significantly the transporters (Divalent-Metal-Transporter1; DMT1), the storage molecule (ferritin) and the ferroxidase enzyme (ceruloplasmin), but not EU. These results illustrate that oxidative stress plays a key role in the mechanism of uranium neurotoxicity. They showed that chronic exposure to DU, but not EU, seems to induce an increase of several antioxidant agents in order to counteract the oxidative stress. Finally, these results demonstrate the importance of the double toxicity, chemical and radiological, of uranium.

Oxidation and Hydration of U 3O 8 Materials Following Controlled Exposure to Temperature and Humidity

DOE PAGES

Tamasi, Alison L.; Boland, Kevin S.; Czerwinski, Kenneth; ...

2015-03-18

Chemical signatures correlated with uranium oxide processing are of interest to forensic science for inferring sample provenance. Identification of temporal changes in chemical structures of process uranium materials as a function of controlled temperatures and relative humidities may provide additional information regarding sample history. In our study, a high-purity α-U 3O 8 sample and three other uranium oxide samples synthesized from reaction routes used in nuclear conversion processes were stored under controlled conditions over 2–3.5 years, and powder X-ray diffraction analysis and X-ray absorption spectroscopy were employed to characterize chemical speciation. We measured signatures from the α-U 3O 8 samplemore » indicated that the material oxidized and hydrated after storage under high humidity conditions over time. Impurities, such as uranyl fluoride or schoepites, were initially detectable in the other uranium oxide samples. After storage under controlled conditions, the analyses of the samples revealed oxidation over time, although the signature of the uranyl fluoride impurity diminished. The presence of schoepite phases in older uranium oxide material is likely indicative of storage under high humidity and should be taken into account for assessing sample history. Finally, the absence of a signature from a chemical impurity, such as uranyl fluoride hydrate, in an older material may not preclude its presence at the initial time of production.« less

Oxidation and Hydration of U 3 O 8 Materials Following Controlled Exposure to Temperature and Humidity

DOE PAGES

Tamasi, Alison L.; Boland, Kevin S.; Czerwinski, Kenneth; ...

2015-03-18

Chemical signatures correlated with uranium oxide processing are of interest to forensic science for inferring sample provenance. Identification of temporal changes in chemical structures of process uranium materials as a function of controlled temperatures and relative humidities may provide additional information regarding sample history. In our study, a high-purity α-U 3O 8 sample and three other uranium oxide samples synthesized from reaction routes used in nuclear conversion processes were stored under controlled conditions over 2–3.5 years, and powder X-ray diffraction analysis and X-ray absorption spectroscopy were employed to characterize chemical speciation. We measured signatures from the α-U 3O 8 samplemore » indicated that the material oxidized and hydrated after storage under high humidity conditions over time. Impurities, such as uranyl fluoride or schoepites, were initially detectable in the other uranium oxide samples. After storage under controlled conditions, the analyses of the samples revealed oxidation over time, although the signature of the uranyl fluoride impurity diminished. The presence of schoepite phases in older uranium oxide material is likely indicative of storage under high humidity and should be taken into account for assessing sample history. Finally, the absence of a signature from a chemical impurity, such as uranyl fluoride hydrate, in an older material may not preclude its presence at the initial time of production. LA-UR-15-21495.« less

A graphene oxide/amidoxime hydrogel for enhanced uranium capture

PubMed Central

Wang, Feihong; Li, Hongpeng; Liu, Qi; Li, Zhanshuang; Li, Rumin; Zhang, Hongsen; Liu, Lianhe; Emelchenko, G. A.; Wang, Jun

2016-01-01

The efficient development of selective materials for the recovery of uranium from nuclear waste and seawater is necessary for their potential application in nuclear fuel and the mitigation of nuclear pollution. In this work, a graphene oxide/amidoxime hydrogel (AGH) exhibits a promising adsorption performance for uranium from various aqueous solutions, including simulated seawater. We show high adsorption capacities (Qm = 398.4 mg g−1) and high % removals at ppm or ppb levels in aqueous solutions for uranium species. In the presence of high concentrations of competitive ions such as Mg2+, Ca2+, Ba2+ and Sr2+, AGH displays an enhanced selectivity for uranium. For low uranium concentrations in simulated seawater, AGH binds uranium efficiently and selectively. The results presented here reveal that the AGH is a potential adsorbent for remediating nuclear industrial effluent and adsorbing uranium from seawater. PMID:26758649

A two-dimensional, finite-difference model of the oxidation of a uranium carbide fuel pellet

NASA Astrophysics Data System (ADS)

Shepherd, James; Fairweather, Michael; Hanson, Bruce C.; Heggs, Peter J.

2015-12-01

The oxidation of spent uranium carbide fuel, a candidate fuel for Generation IV nuclear reactors, is an important process in its potential reprocessing cycle. However, the oxidation of uranium carbide in air is highly exothermic. A model has therefore been developed to predict the temperature rise, as well as other useful information such as reaction completion times, under different reaction conditions in order to help in deriving safe oxidation conditions. Finite difference-methods are used to model the heat and mass transfer processes occurring during the reaction in two dimensions and are coupled to kinetics found in the literature.

Method for fabricating laminated uranium composites

DOEpatents

Chapman, L.R.

1983-08-03

The present invention is directed to a process for fabricating laminated composites of uranium or uranium alloys and at least one other metal or alloy. The laminated composites are fabricated by forming a casting of the molten uranium with the other metal or alloy which is selectively positioned in the casting and then hot-rolling the casting into a laminated plate in or around which the casting components are metallurgically bonded to one another to form the composite. The process of the present invention provides strong metallurgical bonds between the laminate components primarily since the bond disrupting surface oxides on the uranium or uranium alloy float to the surface of the casting to effectively remove the oxides from the bonding surfaces of the components.

Actinide removal from spent salts

DOEpatents

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

2002-01-01

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

FORMING TUBES AND RODS OF URANIUM METAL BY EXTRUSION

DOEpatents

Creutz, E.C.

1959-01-27

A method and apparatus are presented for the extrusion of uranium metal. Since uranium is very brittle if worked in the beta phase, it is desirable to extrude it in the gamma phase. However, in the gamma temperature range thc uranium will alloy with the metal of the extrusion dic, and is readily oxidized to a great degree. According to this patent, uranium extrusion in thc ganmma phase may be safely carried out by preheating a billet of uranium in an inert atmosphere to a trmperature between 780 C and 1100 C. The heated billet is then placed in an extrusion apparatus having dies which have been maintained at an elevated temperature for a sufficient length of time to produce an oxide film, and placing a copper disc between the uranium billet and the die.

The measurement of U(VI) and Np(IV) mass transfer in a single stage centrifugal contactor

NASA Astrophysics Data System (ADS)

May, I.; Birkett, E. J.; Denniss, I. S.; Gaubert, E. T.; Jobson, M.

2000-07-01

BNFL currently operates two reprocessing plants for the conversion of spent nuclear fuel into uranium and plutonium products for fabrication into uranium oxide and mixed uranium and plutonium oxide (MOX) fuels. To safeguard the future commercial viability of this process, BNFL is developing novel single cycle flowsheets that can be operated in conjunction with intensified centrifugal contactors.

Calyptogena-cemented rocks and concretions from the eastern part of Nankai accretionary prism: Age and geochemistry of uranium

NASA Astrophysics Data System (ADS)

Lalou, Claude; Fontugne, Michel; Lallemand, Serge E.; Lauriat-Rage, Agnès

1992-04-01

Calyptogena valves included in a carbonate-rich cement, and fragments of a carbonate-rich chimney, have been examined for their stable isotopic (C and O) composition,14C activity and uranium series disequilibrium. The fossil shells were formed essentially with seawater carbon and a negligible contribution of cold seepage organic carbon, as shown by theirδ13C values. This allows the14C method to be used to determine their age. A fairly good concordance between the14C and230Th234U ages of the youngest shells gives confidence in the dating of the older samples using the latter technique. Thus, theCalyptogena are dated at ca. 150,000 and 20,000 yrs B.P. They have been preserved from dissolution by rapid cementation by a supersaturated carbonate solution. The cement is especially rich in uranium (as high as 75 ppm), whose source is seawater; the enrichment is due to local reducing conditions brought about by the bacterial decomposition of the soft tissues of the bivalves shortly after death. TheCalyptogena that probably developed between these two events (the events of ca. 20,000 and 150,000 yrs) have not been preserved from dissolution because, as is presently the case, the cold seepages were undersaturated with calcium carbonate. The two events probably represent periods of intense fluid venting connected with tectonic activity.

Analysis of long-term bacterial vs. chemical Fe(III) oxide reduction kinetics

NASA Astrophysics Data System (ADS)

Roden, Eric E.

2004-08-01

Data from studies of dissimilatory bacterial (10 8 cells mL -1 of Shewanella putrefaciens strain CN32, pH 6.8) and ascorbate (10 mM, pH 3.0) reduction of two synthetic Fe(III) oxide coated sands and three natural Fe(III) oxide-bearing subsurface materials (all at ca. 10 mmol Fe(III) L -1) were analyzed in relation to a generalized rate law for mineral dissolution (J t/m 0 = k'(m/m 0) γ, where J t is the rate of dissolution and/or reduction at time t, m 0 is the initial mass of oxide, and m/m 0 is the unreduced or undissolved mineral fraction) in order to evaluate changes in the apparent reactivity of Fe(III) oxides during long-term biological vs. chemical reduction. The natural Fe(III) oxide assemblages demonstrated larger changes in reactivity (higher γ values in the generalized rate law) compared to the synthetic oxides during long-term abiotic reductive dissolution. No such relationship was evident in the bacterial reduction experiments, in which temporal changes in the apparent reactivity of the natural and synthetic oxides were far greater (5-10 fold higher γ values) than in the abiotic reduction experiments. Kinetic and thermodynamic considerations indicated that neither the abundance of electron donor (lactate) nor the accumulation of aqueous end-products of oxide reduction (Fe(II), acetate, dissolved inorganic carbon) are likely to have posed significant limitations on the long-term kinetics of oxide reduction. Rather, accumulation of biogenic Fe(II) on residual oxide surfaces appeared to play a dominant role in governing the long-term kinetics of bacterial crystalline Fe(III) oxide reduction. The experimental findings together with numerical simulations support a conceptual model of bacterial Fe(III) oxide reduction kinetics that differs fundamentally from established models of abiotic Fe(III) oxide reductive dissolution, and indicate that information on Fe(III) oxide reactivity gained through abiotic reductive dissolution techniques cannot be used to predict long-term patterns of reactivity toward enzymatic reduction at circumneutral pH.

Uranium Release from Acidic Weathered Hanford Sediments: Single-Pass Flow-Through and Column Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Guohui; Um, Wooyong; Wang, Zheming

The reaction of acidic radioactive waste with sediments can induce mineral transformation reactions that, in turn, control contaminant fate. Here, sediment weathering by synthetic uranium-containing acid solutions was investigated using bench-scale experiments to simulate waste disposal conditions at Hanford’s cribs, USA. During acid weathering, the presence of phosphate exerted a strong influence over uranium mineralogy and a rapidly precipitated, crystalline uranium phosphate phase (meta-ankoleite [K(UO2)(PO4)·3H2O]) was identified using spectroscopic and diffraction-based techniques. In phosphate-free system, uranium oxyhydroxide minerals such as K-compreignacite [K2(UO2)6O4(OH)6·7H2O] were formed. Single-pass flow-through (SPFT) and column leaching experiments using synthetic Hanford pore water showed that uranium precipitatedmore » as meta-ankoleite during acid weathering was strongly retained in the sediments, with an average release rate of 2.67E-12 mol g-1 s-1. In the absence of phosphate, uranium release was controlled by dissolution of uranium oxyhydroxide (compreignacite-type) mineral with a release rate of 1.05-2.42E-10 mol g-1 s-1. The uranium mineralogy and release rates determined for both systems in this study support the development of accurate U-release models for prediction of contaminant transport. These results suggest that phosphate minerals may be a good candidate for uranium remediation approaches at contaminated sites.« less

Distribution and potential health risk of groundwater uranium in Korea.

PubMed

Shin, Woosik; Oh, Jungsun; Choung, Sungwook; Cho, Byong-Wook; Lee, Kwang-Sik; Yun, Uk; Woo, Nam-Chil; Kim, Hyun Koo

2016-11-01

Chronic exposure even to extremely low specific radioactivity of natural uranium in groundwater results in kidney problems and potential toxicity in bones. This study was conducted to assess the potential health risk via intake of the groundwater containing uranium, based on the determination of the uranium occurrence in groundwater. The groundwater was investigated from a total of 4140 wells in Korea. Most of the groundwater samples showed neutral pH and (sub-)oxic condition that was influenced by the mixing with shallow groundwater due to long-screened (open) wells. High uranium contents exceeding the WHO guideline level of 30 μg L(-1) were observed in the 160 wells located mainly in the plutonic bedrock regions. The statistical analysis suggested that the uranium component was present in groundwater by desorption and re-dissolution processes. Predominant uranium phases were estimated to uranyl carbonates under the Korean groundwater circumstances. These mobile forms of uranium and oxic condition facilitate the increase of potential health risk downgradient. In particular, long-term intake of groundwater containing >200 μg U L(-1) may induce internal exposure to radiation as well as the effects of chemical toxicity. These high uranium concentrations were found in twenty four sampling wells of rural areas in this study, and they were mainly used for drinking. Therefore, the high-level uranium wells and neighboring areas must be properly managed and monitored to reduce the exposure risk for the residents by drinking groundwater. Copyright © 2016 Elsevier Ltd. All rights reserved.

Uranium Release from Acidic Weathered Hanford Sediments: Single-Pass Flow-Through and Column Experiments.

PubMed

Wang, Guohui; Um, Wooyong; Wang, Zheming; Reinoso-Maset, Estela; Washton, Nancy M; Mueller, Karl T; Perdrial, Nicolas; O'Day, Peggy A; Chorover, Jon

2017-10-03

The reaction of acidic radioactive waste with sediments can induce mineral transformation reactions that, in turn, control contaminant fate. Here, sediment weathering by synthetic uranium-containing acid solutions was investigated using bench-scale experiments to simulate waste disposal conditions at Hanford's cribs (Hanford, WA). During acid weathering, the presence of phosphate exerted a strong influence over uranium mineralogy and a rapidly precipitated, crystalline uranium phosphate phase (meta-ankoleite [K(UO 2 )(PO 4 )·3H 2 O]) was identified using spectroscopic and diffraction-based techniques. In phosphate-free system, uranium oxyhydroxide minerals such as K-compreignacite [K 2 (UO 2 ) 6 O 4 (OH) 6 ·7H 2 O] were formed. Single-pass flow-through (SPFT) and column leaching experiments using synthetic Hanford pore water showed that uranium precipitated as meta-ankoleite during acid weathering was strongly retained in the sediments, with an average release rate of 2.67 × 10 -12 mol g -1 s -1 . In the absence of phosphate, uranium release was controlled by dissolution of uranium oxyhydroxide (compreignacite-type) mineral with a release rate of 1.05-2.42 × 10 -10 mol g -1 s -1 . The uranium mineralogy and release rates determined for both systems in this study support the development of accurate U-release models for the prediction of contaminant transport. These results suggest that phosphate minerals may be a good candidate for uranium remediation approaches at contaminated sites.

Recovery of protactinium from molten fluoride nuclear fuel compositions

DOEpatents

Baes, C.F. Jr.; Bamberger, C.; Ross, R.G.

1973-12-25

A method is provided for separating protactinium from a molten fluonlde salt composition consisting essentially of at least one alkali and alkaline earth metal fluoride and at least one soluble fluoride of uranium or thorium which comprises oxidizing the protactinium in said composition to the + 5 oxidation state and contacting said composition with an oxide selected from the group consisting of an alkali metal oxide, an alkaline earth oxide, thorium oxide, and uranium oxide, and thereafter isolating the resultant insoluble protactinium oxide product from said composition. (Official Gazette)

Crucible cast from beryllium oxide and refractory cement is impervious to flux and molten metal

NASA Technical Reports Server (NTRS)

Jastrzebski, Z. D.

1966-01-01

Crucible from a mixture of a beryllium oxide aggregate and hydraulic refractory cement, and coated with an impervious refractory oxide will not deteriorate in the presence of fused salt- molten metal mixtures such as uranium- magnesium-zinc-halide salt systems. Vessels cast by this process are used in the flux reduction of oxides of thorium and uranium.

PROGRESS REPORT ON GEOLOGIC STUDIES OF THE RANGER OREBODIES, NORTHERN TERRITORY, AUSTRALIA.

USGS Publications Warehouse

Nash, J. Thomas; Frishman, David; ,

1985-01-01

The Ranger No. 1 and No. 3 orebodies contain about 124,000 tonnes U//3O//8 in highly chloritized metasediments of the lower Proterozoic Cahill Formation within about 500 m of the projected sub-Kombolgie Formation unconformity. In both orebodies, oxidized and reduced uranium minerals occur chiefly in quartzose schists that have highly variable amounts of muscovite, sericite, and chlorite. The effects of several periods of alteration are pervasive in the vicinity of orebodies where biotite and garnet are altered to chlorite, and feldspars to white mica or chlorite. Oxidized uranium minerals, associated with earthy iron oxides, occur from the surface to a depth of about 60 m. Below the oxidized zone, uranium occurs chiefly as uraninite and pitchblende disseminated through thick sections of quartz-chlorite-muscovite schist and has no apparent association with graphite or sulphides. The geologic age(s) of uranium emplacement are obscure because there are few age criteria. Reduced uranium minerals are younger than 1. 8-b. y. -old granite dykes, and some occur locally in 1. 65-b. y. -old Kombolgie Formation.

Compact reaction cell for homogenizing and down-blending highly enriched uranium metal

DOEpatents

McLean, W. II; Miller, P.E.; Horton, J.A.

1995-05-02

The invention is a specialized reaction cell for converting uranium metal to uranium oxide. In a preferred form, the reaction cell comprises a reaction chamber with increasing diameter along its length (e.g. a cylindrical chamber having a diameter of about 2 inches in a lower portion and having a diameter of from about 4 to about 12 inches in an upper portion). Such dimensions are important to achieve the necessary conversion while at the same time affording criticality control and transportability of the cell and product. The reaction chamber further comprises an upper port and a lower port, the lower port allowing for the entry of reactant gases into the reaction chamber, the upper port allowing for the exit of gases from the reaction chamber. A diffuser plate is attached to the lower port of the reaction chamber and serves to shape the flow of gas into the reaction chamber. The reaction cell further comprises means for introducing gases into the reaction chamber and a heating means capable of heating the contents of the reaction chamber. The present invention also relates to a method for converting uranium metal to uranium oxide in the reaction cell of the present invention. The invention is useful for down-blending highly enriched uranium metal by the simultaneous conversion of highly enriched uranium metal and natural or depleted uranium metal to uranium oxide within the reaction cell. 4 figs.

Compact reaction cell for homogenizing and down-blanding highly enriched uranium metal

DOEpatents

McLean, II, William; Miller, Philip E.; Horton, James A.

1995-01-01

The invention is a specialized reaction cell for converting uranium metal to uranium oxide. In a preferred form, the reaction cell comprises a reaction chamber with increasing diameter along its length (e.g. a cylindrical chamber having a diameter of about 2 inches in a lower portion and having a diameter of from about 4 to about 12 inches in an upper portion). Such dimensions are important to achieve the necessary conversion while at the same time affording criticality control and transportability of the cell and product. The reaction chamber further comprises an upper port and a lower port, the lower port allowing for the entry of reactant gasses into the reaction chamber, the upper port allowing for the exit of gasses from the reaction chamber. A diffuser plate is attached to the lower port of the reaction chamber and serves to shape the flow of gas into the reaction chamber. The reaction cell further comprises means for introducing gasses into the reaction chamber and a heating means capable of heating the contents of the reaction chamber. The present invention also relates to a method for converting uranium metal to uranium oxide in the reaction cell of the present invention. The invention is useful for down-blending highly enriched uranium metal by the simultaneous conversion of highly enriched uranium metal and natural or depleted uranium metal to uranium oxide within the reaction cell.

Precipitation and Dissolution of Uranyl Phosphates in a Microfluidic Pore Structure

NASA Astrophysics Data System (ADS)

Werth, C. J.; Fanizza, M.; Strathmann, T.; Finneran, K.; Oostrom, M.; Zhang, C.; Wietsma, T. W.; Hess, N. J.

2011-12-01

The abiotic precipitation of uranium (U(VI)) was evaluated in a microfluidic pore structure (i.e. micromodel) to assess the efficacy of using a phosphate amendment to immobilize uranium in groundwater and mitigate the risk of this contaminant to potential down-gradient receptor sites. U(VI) was mixed transverse to the direction of flow with hydrogen phosphate (HPO42-), in the presence or absence of calcium (Ca2+) or sulfate (SO42-), in order to identify precipitation rates, the morphology and types of minerals formed, and the stability of these minerals to dissolution with and without bicarbonate (HCO3-) present. Raman backscattering spectroscopy and micro X-ray diffraction (μ-XRD) results both showed that the only mineral precipitated was chernikovite (also known as hydrogen uranyl phosphate; UO2HPO4), even though the formation of other minerals were thermodynamically favored depending on the experimental conditions. Precipitation and dissolution rates varied with influent conditions. Relative to when only U(VI) and HPO42- were present, precipitation rates were 2.3 times slower when SO42- was present, and 1.4 times faster when Ca2+ was present. These rates were inversely related to the size of crystals formed during precipitation. Dissolution rates for chernikovite increased with increasing HCO3- concentrations, consistent with formation of uranyl carbonate complexes in aqueous solution, and they were the fastest for chernikovite formed in the presence of SO42-, and slowest for the chernikovite formed in the presence of Ca2+. These rates are related to the ratios of mineral-water interfacial area to mineral volume. Fluorescent tracer studies and laser confocal microscopy images showed that densely aggregated precipitates blocked pores and reduced permeability. The results suggest that changes in the solute conditions evaluated affect precipitation rates, crystal morphology, and crystal stability, but not mineral type.

SURFACE TREATMENT OF METALLIC URANIUM

DOEpatents

Gray, A.G.; Schweikher, E.W.

1958-05-27

The treatment of metallic uranium to provide a surface to which adherent electroplates can be applied is described. Metallic uranium is subjected to an etchant treatment in aqueous concentrated hydrochloric acid, and the etched metal is then treated to dissolve the resulting black oxide and/or chloride film without destroying the etched metal surface. The oxide or chloride removal is effected by means of moderately concentrated nitric acid in 3 to 20 seconds.

PROCESS FOR THE RECOVERY AND PURIFICATION OF URANIUM DEPOSITS

DOEpatents

Carter, J.M.; Kamen, M.D.

1958-10-14

A process is presented for recovering uranium values from UCl/sub 4/ deposits formed on calutrons. Such deposits are removed from the calutron parts by an aqueous wash solution which then contains the uranium values in addition to the following impurities: Ni, Cu, Fe, and Cr. This impurity bearing wash solution is treated with an oxidizing agent, and the oxidized solution is then treated with ammonia in order to precipitate the uranium as ammonium diuranate. The metal impurities of iron and chromium, which form insoluble hydroxides, are precipitated along with the uranium values. The precipitate is separated from the solution, dissolved in acid, and the solution again treated with ammonia and ammonium carbonate, which results in the precipitation of the metal impurities as hydroxides while the uranium values remain in solution.

Physicochemical characterization of Capstone depleted uranium aerosols IV: in vitro solubility analysis.

PubMed

Guilmette, Raymond A; Cheng, Yung Sung

2009-03-01

As part of the Capstone Depleted Uranium (DU) Aerosol Study, the solubility of selected aerosol samples was measured using an accepted in vitro dissolution test system. This static system was employed along with a SUF (synthetic ultrafiltrate) solvent, which is designed to mimic the physiological chemistry of extracellular fluid. Using sequentially obtained solvent samples, the dissolution behavior over a 46-d test period was evaluated by fitting the measurement data to two- or three-component negative exponential functions. These functions were then compared with Type M and S absorption taken from the International Commission on Radiological Protection Publication 66 Human Respiratory Tract Model. The results indicated that there was a substantial variability in solubility of the aerosols, which in part depended on the type of armor being impacted by the DU penetrator and the particle size fraction being tested. Although some trends were suggested, the variability noted leads to uncertainties in predicting the solubility of other DU-based aerosols. Nevertheless, these data provide a useful experimental basis for modeling the intake-dose relationships for inhaled DU aerosols arising from penetrator impact on armored vehicles.

Composition and method for brazing graphite to graphite

DOEpatents

Taylor, A.J.; Dykes, N.L.

1982-08-10

A brazing material is described for joining graphite structures that can be used up to 2800/sup 0/C. The brazing material is formed of a paste-like composition of hafnium carbide and uranium oxide with a thermosetting resin. The uranium oxide is converted to uranium dicarbide during the brazing operation and then the hafnium carbide and uranium dicarbide form a liquid phase at a temperature about 2600/sup 0/C with the uranium diffusing and vaporizing from the joint area as the temperature is increased to about 2800/sup 0/C so as to provide a brazed joint consisting essentially of hafnium carbide. The resulting brazed joint is chemically and thermally compatible with the graphite structures.

URANIUM RECOVERY FROM COMPOSITE UF$sub 4$ REDUCTION BOMB WASTES

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, E R; Doyle, R L; Coleman, J R

1954-01-28

A number of techniques have been investigated on a laboratory-scale for separating uranium from fluorides during the recovery of uranium from UF4 reduction bomb wastes (C-oxide) by an HCl leach - NH4OH precipitation process. Among these are included adsorption of fluorides from filtered leach liquors, fractional precipitation of fluorides and uranium, complexing of fluorides into forms soluble in slightly acid solutions, and fluoride volatilization from the uranium concentrate. Solubility studies of CaF2 and MgF2 in aqueous hydrochloric acid at various acidities and temperatures were also conducted. A description of the production-scale processing of C-oxide in the FMPC scrap plant hasmore » been included.« less

Preliminary reconnaissance survey for thorium, uranium, and rare-earth oxides, Bear Lodge Mountains, Crook County, Wyoming

USGS Publications Warehouse

Wilmarth, V.R.; Johnson, D.H.

1953-01-01

An area about 6 miles north of Sundance, in the Bear Lodge Mountains, in Crook County, Wyo., was examined during August 1950 for thorium, uranium, and rare-earth oxides and samples were collected. Uranium is known to occur in fluorite veins and iron-manganese veins and in the igneous rocks of Tertiary age that compose the core of the Bear Lodge Mountains. The uranium content of the samples ranges from 0.001 to 0.015 percent in those from the fluorite veins, from 0.005 to 0.018 percent in those from the iron-manganese veins, and from 0.001 to 0.017 percent in those from the igneous rocks. The radioactivity of the samples is more than that expected from the uranium content. Thorium accounts for most of this discrepancy. The thorium oxide content of samples ranges from 0.07 to 0.25 percent in those from the iron-manganese veins and from 0.07 to 0.39 percent in those from the sedimentary rocks, and from0.04 to 0.30 in those from the igneous rocks. Rare-earth oxides occur in iron-manganese veins and in zones of altered igneous rocks. The veins contain from 0.16 to 12.99 percent rare-earth oxides, and the igneous rocks, except for two localities, contain from 0.01 to 0.42 percent rare-earth oxides. Inclusions of metamorphosed sedimentary rocks in the intrusive rocks contain from 0.07 to 2.01 percent rare-earth oxides.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bakel, Allen J.; Conner, Cliff; Quigley, Kevin

One of the missions of the Reduced Enrichment for Research and Test Reactors (RERTR) program (and now the National Nuclear Security Administrations Material Management and Minimization program) is to facilitate the use of low enriched uranium (LEU) targets for 99Mo production. The conversion from highly enriched uranium (HEU) to LEU targets will require five to six times more uranium to produce an equivalent amount of 99Mo. The work discussed here addresses the technical challenges encountered in the treatment of uranyl nitrate hexahydrate (UNH)/nitric acid solutions remaining after the dissolution of LEU targets. Specifically, the focus of this work is themore » calcination of the uranium waste from 99Mo production using LEU foil targets and the Modified Cintichem Process. Work with our calciner system showed that high furnace temperature, a large vent tube, and a mechanical shield are beneficial for calciner operation. One- and two-step direct calcination processes were evaluated. The high-temperature one-step process led to contamination of the calciner system. The two-step direct calcination process operated stably and resulted in a relatively large amount of material in the calciner cup. Chemically assisted calcination using peroxide was rejected for further work due to the difficulty in handling the products. Chemically assisted calcination using formic acid was rejected due to unstable operation. Chemically assisted calcination using oxalic acid was recommended, although a better understanding of its chemistry is needed. Overall, this work showed that the two-step direct calcination and the in-cup oxalic acid processes are the best approaches for the treatment of the UNH/nitric acid waste solutions remaining from dissolution of LEU targets for 99Mo production.« less

Porous membrane electrochemical cell for uranium and transuranic recovery from molten salt electrolyte

DOEpatents

Willit, James L [Ratavia, IL

2007-09-11

An improved process and device for the recovery of the minor actinides and the transuranic elements (TRU's) from a molten salt electrolyte. The process involves placing the device, an electrically non-conducting barrier between an anode salt and a cathode salt. The porous barrier allows uranium to diffuse between the anode and cathode, yet slows the diffusion of uranium ions so as to cause depletion of uranium ions in the catholyte. This allows for the eventual preferential deposition of transuranics present in spent nuclear fuel such as Np, Pu, Am, Cm. The device also comprises an uranium oxidation anode. The oxidation anode is solid uranium metal in the form of spent nuclear fuel. The spent fuel is placed in a ferric metal anode basket which serves as the electrical lead or contact between the molten electrolyte and the anodic uranium metal.

Porous membrane electrochemical cell for uranium and transuranic recovery from molten salt electrolyte

DOEpatents

Willit, James L [Batavia, IL

2010-09-21

An improved process and device for the recovery of the minor actinides and the transuranic elements (TRU's) from a molten salt electrolyte. The process involves placing the device, an electrically non-conducting barrier between an anode salt and a cathode salt. The porous barrier allows uranium to diffuse between the anode and cathode, yet slows the diffusion of uranium ions so as to cause depletion of uranium ions in the catholyte. This allows for the eventual preferential deposition of transuranics present in spent nuclear fuel such as Np, Pu, Am, Cm. The device also comprises an uranium oxidation anode. The oxidation anode is solid uranium metal in the form of spent nuclear fuel. The spent fuel is placed in a ferric metal anode basket which serves as the electrical lead or contact between the molten electrolyte and the anodic uranium metal.

Ferrihydrite dissolution by pyridine-2,6-bis(monothiocarboxylic acid) and hydrolysis products

NASA Astrophysics Data System (ADS)

Dhungana, Suraj; Anthony, Charles R.; Hersman, Larry E.

2007-12-01

Pyridine-2,6-bis(monothiocarboxylate) (pdtc), a metabolic product of microorganisms, including Pseudomonas putida and Pseudomonas stutzeri was investigated for its ability of dissolve Fe(III)(hydr)oxides at pH 7.5. Concentration dependent dissolution of ferrihydrite under anaerobic environment showed saturation of the dissolution rate at the higher concentration of pdtc. The surface controlled ferrihydrite dissolution rate was determined to be 1.2 × 10 -6 mol m -2 h -1. Anaerobic dissolution of ferrihydrite by pyridine-2,6-dicarboxylic acid or dipicolinic acid (dpa), a hydrolysis product of pdtc, was investigated to study the mechanism(s) involved in the pdtc facilitated ferrihydrite dissolution. These studies suggest that pdtc dissolved ferrihydrite using a reduction step, where dpa chelates the Fe reduced by a second hydrolysis product, H 2S. Dpa facilitated dissolution of ferrihydrite showed very small increase in the Fe dissolution when the concentration of external reductant, ascorbate, was doubled, suggesting the surface dynamics being dominated by the interactions between dpa and ferrihydrite. Greater than stoichiometric amounts of Fe were mobilized during dpa dissolution of ferrihydrite assisted by ascorbate and cysteine. This is attributed to the catalytic dissolution of Fe(III)(hydr)oxides by the in situ generated Fe(II) in the presence of a complex former, dpa.

Influence of low concentration V and Co oxide doping on the dissolution behaviors of simplified nuclear waste glasses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lu, Xiaonan; Neeway, James J.; Ryan, Joseph V.

Transition metal oxides are commonly present in nuclear waste and they can alter the structure, property and especially dissolution behaviors of the glasses used for waste immobilization. In this paper, we investigated vanadium and cobalt oxide induced structural and properties changes, especially dissolution behaviors, of International Simple Glass (ISG), a model nuclear waste glass system. Static chemical durability tests were performed at 90 °C with a pH value of 7 and a surface-area-to-solution-volume of 200 m-1 for 112 days on three glasses: ISG, ISG doped with 0.5 mol% Co2O3, and ISG doped with 2.0 mol% V2O5. ICP-MS was used tomore » analyze the dissolved ion concentrations. It was found that doping with vanadium and cobalt oxide, even at the low doping concentration, significantly reduced the extent of the ISG glass dissolution. Differential Scanning Calorimetry (DSC) analysis showed that vanadium oxide doping reduced the glass transition temperature (Tg) while cobalt oxide did not significantly change the Tg of ISG. X-ray diffraction (XRD), Raman spectrometry and scanning electron microscopy (SEM) were used to analyze the glass samples before and after corrosion to understand the phase and microstructure changes.« less

[Solidification of volatile oil with graphene oxide].

PubMed

Yan, Hong-Mei; Jia, Xiao-Bin; Zhang, Zhen-Hai; Sun, E; Xu, Yi-Hao

2015-02-01

To evaluate the properties of solidifying volatile oil with graphene oxide, clove oil and zedoary turmeric oil were solidified by graphene oxide. The amount of graphene oxide was optimized with the eugenol yield and curcumol yield as criteria. Curing powder was characterized by differential scanning calorimetry (DSC) and scanning electron microscopy (SEM). The effects of graphene oxide on dissolution in vitro and thermal stability of active components were studied. The optimum solidification ratio of graphene oxide to volatile oil was 1:1. Dissolution rate of active components had rare influence while their thermal stability improved after volatile oil was solidified. Solidifying herbal volatile oil with graphene oxide deserves further study.

Electronic structure and fine structural features of the air-grown UNxOy on nitrogen-rich uranium nitride

NASA Astrophysics Data System (ADS)

Long, Zhong; Zeng, Rongguang; Hu, Yin; Liu, Jing; Wang, Wenyuan; Zhao, Yawen; Luo, Zhipeng; Bai, Bin; Wang, Xiaofang; Liu, Kezhao

2018-06-01

Oxide formation on surface of nitrogen-rich uranium nitride film/particles was investigated using X-ray photoelectron spectroscopy (XPS), auger electron spectroscopy (AES), aberration-corrected transmission electron microscopy (TEM), and high-angle annular dark-field scanning transmission electron microscopy (HAADF-STEM) coupled with electron energy-loss spectroscopy (EELS). XPS and AES studies indicated that the oxidized layer on UN2-x film is ternary compound uranium oxynitride (UNxOy) in 5-10 nm thickness. TEM/HAADF-STEM and EELS studies revealed the UNxOy crystallizes in the FCC CaF2-type structure with the lattice parameter close to the CaF2-type UN2-x matrix. The work can provide further information to the oxidation mechanism of uranium nitride.

Uranium-series constraints on radionuclide transport and groundwater flow at the Nopal I uranium deposit, Sierra Pena Blanca, Mexico.

PubMed

Goldstein, Steven J; Abdel-Fattah, Amr I; Murrell, Michael T; Dobson, Patrick F; Norman, Deborah E; Amato, Ronald S; Nunn, Andrew J

2010-03-01

Uranium-series data for groundwater samples from the Nopal I uranium ore deposit were obtained to place constraints on radionuclide transport and hydrologic processes for a nuclear waste repository located in fractured, unsaturated volcanic tuff. Decreasing uranium concentrations for wells drilled in 2003 are consistent with a simple physical mixing model that indicates that groundwater velocities are low ( approximately 10 m/y). Uranium isotopic constraints, well productivities, and radon systematics also suggest limited groundwater mixing and slow flow in the saturated zone. Uranium isotopic systematics for seepage water collected in the mine adit show a spatial dependence which is consistent with longer water-rock interaction times and higher uranium dissolution inputs at the front adit where the deposit is located. Uranium-series disequilibria measurements for mostly unsaturated zone samples indicate that (230)Th/(238)U activity ratios range from 0.005 to 0.48 and (226)Ra/(238)U activity ratios range from 0.006 to 113. (239)Pu/(238)U mass ratios for the saturated zone are <2 x 10(-14), and Pu mobility in the saturated zone is >1000 times lower than the U mobility. Saturated zone mobility decreases in the order (238)U approximately (226)Ra > (230)Th approximately (239)Pu. Radium and thorium appear to have higher mobility in the unsaturated zone based on U-series data from fractures and seepage water near the deposit.

A method for preparation and cleaning of uniformly sized arsenopyrite particles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parthasarathy, Hariprasad; Baltrus, John P; Dzombak, David A

The oxidative dissolution of sulfide minerals, such as arsenopyrite (FeAsS), is of critical importance in many geochemical systems. A comprehensive understanding of their dissolution rates entails careful preparation of the mineral surface. Measurements of dissolution rates of arsenic from arsenopyrite are dependent on the size and degree of oxidation of its particles, among other factors. In this work, a method was developed for preparation and cleaning of arsenopyrite particles with size range of 150–250 μm. Four different cleaning methods were evaluated for effectiveness based on the removal of oxidized species of iron (Fe), arsenic (As) and sulfur (S) from themore » surface. The percentage oxidation of the surface was determined using X-ray photoelectron spectroscopy (XPS), and surface stoichiometry was measured using scanning electron microscopy – energy dispersive X-ray spectroscopy (SEM-EDS). Results indicate that sonicating the arsenopyrite particles and then cleaning them with 12N HCl followed by 50% ethanol, and drying in nitrogen was the most effective method. This method was successful in greatly reducing the oxide species of Fe while completely removing oxides of As and S from the arsenopyrite surface. Although sonication and acid cleaning have been widely used for mineral preparation, the method described in this study can significantly reduce grain size heterogeneity as well as surface oxidation, which enables greater control in surface and dissolution experiments.« less

Effect of surface chemistries and characteristics of Ti6Al4V on the Ca and P adsorption and ion dissolution in Hank's ethylene diamine tetra-acetic acid solution.

PubMed

Chang, E; Lee, T M

2002-07-01

This study examined the influence of chemistries and surface characteristics of Ti6Al4V on the adsorption of Ca and P species and ion dissolution behavior of the material exposed in Hank's solution with 8.0 mM ethylene diamine tetra-acetic acid at 37 degrees C. The variation of chemistries of the alloy and nano-surface characteristics (chemistries of nano-surface oxides, amphoteric OH group adsorbed on oxides, and oxide thickness) was effected by surface modification and three passivation methods (34% nitric acid passivation. 400 degrees C heated in air, and aged in 100 degrees C water). X-ray photoelectron spectroscopy and Auger electron spectroscopy were used for surface analyses. The chemistries of nano-surface oxides in a range studied should not change the capability of Ca and P adsorption. Nor is the capability affected significantly by amphoteric OH group and oxide thickness. However, passivations influence the surface oxide thickness and the early stage ion dissolution rate of the alloy. The rate-limiting step of the rate can be best explained by metal-ion transport through the oxide film, rather than hydrolysis of the film. Variation of the chemistries of titanium alloy alters the electromotive force potential of the metal, thereby affecting the corrosion and ion dissolution rate.

A method for preparation and cleaning of uniformly sized arsenopyrite particles

DOE PAGES

Parthasarathy, Hariprasad; Baltrus, John P; Dzombak, David A; ...

2014-10-11

The oxidative dissolution of sulfide minerals, such as arsenopyrite (FeAsS), is of critical importance in many geochemical systems. A comprehensive understanding of their dissolution rates entails careful preparation of the mineral surface. Measurements of dissolution rates of arsenic from arsenopyrite are dependent on the size and degree of oxidation of its particles, among other factors. In this work, a method was developed for preparation and cleaning of arsenopyrite particles with size range of 150–250 μm. Four different cleaning methods were evaluated for effectiveness based on the removal of oxidized species of iron (Fe), arsenic (As) and sulfur (S) from themore » surface. The percentage oxidation of the surface was determined using X-ray photoelectron spectroscopy (XPS), and surface stoichiometry was measured using scanning electron microscopy – energy dispersive X-ray spectroscopy (SEM-EDS). Results indicate that sonicating the arsenopyrite particles and then cleaning them with 12N HCl followed by 50% ethanol, and drying in nitrogen was the most effective method. This method was successful in greatly reducing the oxide species of Fe while completely removing oxides of As and S from the arsenopyrite surface. Although sonication and acid cleaning have been widely used for mineral preparation, the method described in this study can significantly reduce grain size heterogeneity as well as surface oxidation, which enables greater control in surface and dissolution experiments.« less

PRODUCTION OF URANIUM TETRACHLORIDE

DOEpatents

Calkins, V.P.

1958-12-16

A process is descrlbed for the production of uranium tetrachloride by contacting uranlum values such as uranium hexafluoride, uranlum tetrafluoride, or uranium oxides with either aluminum chloride, boron chloride, or sodium alumlnum chloride under substantially anhydrous condltlons at such a temperature and pressure that the chlorldes are maintained in the molten form and until the uranium values are completely converted to uranlum tetrachloride.

The scavenging of silver by manganese and iron oxides in stream sediments collected from two drainage areas of Colorado

USGS Publications Warehouse

Chao, T.T.; Anderson, B.J.

1974-01-01

Stream sediments of two well-weathered and aerated drainage areas of Colorado containing anomalous amounts of silver were allowed to react by shaking with nitric acid of different concentrations (1-10M). Silver, manganese, and iron simultaneously dissolved were determined by atomic absorption. The relationship between silver dissolution and the dissolution of manganese and/or iron was evaluated by linear and multiple regression analyses. The highly significant correlation coefficient (r = 0.913) between silver and manganese dissolution suggests that manganese oxides are the major control on the scavenging of silver in these stream sediments, whereas iron oxides only play a secondary role in this regard. ?? 1974.

VOLATILE CHLORIDE PROCESS FOR THE RECOVERY OF METAL VALUES

DOEpatents

Hanley, W.R.

1959-01-01

A process is presented for recovering uranium, iron, and aluminum from centain shale type ores which contain uranium in minute quantities. The ore is heated wiih a chlorinating agent. such as chlorine, to form a volatilized stream of metal chlorides. The chloride stream is then passed through granular alumina which preferentially absorbs the volatile uranium chloride and from which the uranium may later be recovered. The remaining volatilized chlorides, chiefly those of iron and aluminum, are further treated to recover chlorine gas for recycle, and to recover ferric oxide and aluminum oxide as valuable by-products.

The oxidative dissolution of arsenopyrite (FeAsS) and enargite (Cu 3AsS 4) by Leptospirillum ferrooxidans

NASA Astrophysics Data System (ADS)

Corkhill, C. L.; Wincott, P. L.; Lloyd, J. R.; Vaughan, D. J.

2008-12-01

Arsenopyrite (FeAsS) and enargite (Cu 3AsS 4) fractured in a nitrogen atmosphere were characterised after acidic (pH 1.8), oxidative dissolution in both the presence and absence of the acidophilic microorganism Leptospirillum ferrooxidans. Dissolution was monitored through analysis of the coexisting aqueous solution using inductively coupled plasma atomic emission spectroscopy and coupled ion chromatography-inductively coupled plasma mass spectrometry, and chemical changes at the mineral surface observed using X-ray photoelectron spectroscopy and environmental scanning electron microscopy (ESEM). Biologically mediated oxidation of arsenopyrite and enargite (2.5 g in 25 ml) was seen to proceed to a greater extent than abiotic oxidation, although arsenopyrite oxidation was significantly greater than enargite oxidation. These dissolution reactions were associated with the release of ˜917 and ˜180 ppm of arsenic into solution. The formation of Fe(III)-oxyhydroxides, ferric sulphate and arsenate was observed for arsenopyrite, thiosulphate and an unknown arsenic oxide for enargite. ESEM revealed an extensive coating of an extracellular polymeric substance associated with the L. ferrooxidans cells on the arsenopyrite surface and bacterial leach pits suggest a direct biological oxidation mechanism, although a combination of indirect and direct bioleaching cannot be ruled out. Although the relative oxidation rates of enargite were greater in the presence of L. ferrooxidans, cells were not in contact with the surface suggesting an indirect biological oxidation mechanism. Cells of L. ferrooxidans appear able to withstand several hundreds of ppm of As(III) and As(V).

Preliminary Numerical Simulation of IR Structure Development in a Hypothetical Uranium Release.

DTIC Science & Technology

1981-11-16

art Identify by block nAsb.’) IR Structure Power spectrum Uranium release Parallax effects Numerical simulation PHARO code Isophots LWIR 20. _PSTRACT...release at 200 km altitude. Of interest is the LWIR emission from uranium oxide ions, induced by sunlight and earthshine. Assuming a one-level fluid...defense systems of long wave infrared ( LWIR ) emissions from metallic oxides in the debris from a high altitude nuclear explosion (HANE) is an

Energy dependence of the trapping of uranium atoms by aluminum oxide surfaces

NASA Technical Reports Server (NTRS)

Librecht, K. G.

1979-01-01

The energy dependence of the trapping probability for sputtered U-235 atoms striking an oxidized aluminum collector surface at energies between 1 eV and 184 eV was measured. At the lowest energies, approximately 10% of the uranium atoms are not trapped, while above 10 eV essentially all of them stick. Trapping probabilities averaged over the sputtered energy distribution for uranium incident on gold and mica are also presented.

MELTING AND PURIFICATION OF URANIUM

DOEpatents

Spedding, F.H.; Gray, C.F.

1958-09-16

A process is described for treating uranium ingots having inner metal portions and an outer oxide skin. The method consists in partially supporting such an ingot on the surface of a grid or pierced plate. A sufficient weight of uranium is provided so that when the mass becomes molten, the oxide skin bursts at the unsupported portions of its bottom surface, allowing molten urantum to flow through the burst skin and into a container provided below.

Oxidative Remobilization of Technetium Sequestered by Sulfide-Transformed Nano Zerovalent Iron

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fan, Dimin; Anitori, Roberto; Tebo, Bradley M.

2014-06-02

The dissolution of Tc(IV) sulfide and concurrent transformation of sulfidated ZVI during 2 oxidation were examined. Kinetic data obtained with 10 mL batch reactors showed that Tc(VII) 3 reduced by sulfidated nZVI has significantly slower reoxidation rates than Tc(VII) reduced by 4 nZVI only. In a 50 mL batch reactor, initial inhibition of Tc(IV) dissolution was apparent and 5 lasted until 120 hours at S/Fe = 0.112, presumably due to the redox buffer capacity of FeS. This 6 is evidenced by the parallel trends in oxidation-reduction potentials (ORP) and Tc dissolution 7 kinetics. Mӧssbauer spectra and micro X-ray diffraction ofmore » S/Fe = 0.112 suggested the 8 persistence of FeS after 24-h oxidation although X-ray photoelectron spectroscopy indicated 9 substantial surface oxidation. After 120-h oxidation, all characterizations showed complete 10 oxidation of FeS, which further indicates that FeS inhibits Tc oxidation. X-ray absorption 11 spectroscopy for S/Fe = 0.011 showed significantly increasing percentage of TcS2 in the solid 12 phase after 24-h oxidation, indicating TcS2 is more resistant to oxidation than TcO2. At S/Fe = 13 0.112, the XAS results revealed significant transformation of Tc speciation from TcS2 to TcO2 14 after 120-h oxidation at S/Fe = 0.112. Given that no apparent Tc dissolution occurred during this 15 period, the speciation transformation might play a secondary role in hindering Tc oxidation, 16 especially as redox buffer capacity approached depletion.« less

Studies on separation and purification of fission (99)Mo from neutron activated uranium aluminum alloy.

PubMed

Rao, Ankita; Kumar Sharma, Abhishek; Kumar, Pradeep; Charyulu, M M; Tomar, B S; Ramakumar, K L

2014-07-01

A new method has been developed for separation and purification of fission (99)Mo from neutron activated uranium-aluminum alloy. Alkali dissolution of the irradiated target (100mg) results in aluminum along with (99)Mo and a few fission products passing into solution, while most of the fission products, activation products and uranium remain undissolved. Subsequent purification steps involve precipitation of aluminum as Al(OH)3, iodine as AgI/AgIO3 and molybdenum as Mo-α-benzoin oxime. Ruthenium is separated by volatilization as RuO4 and final purification of (99)Mo was carried out using anion exchange method. The radiochemical yield of fission (99)Mo was found to be >80% and the purity of the product was in conformity with the international pharmacopoeia standards. Copyright © 2014 Elsevier Ltd. All rights reserved.

Chalcopyrite dissolution: Scanning photoelectron microscopy examination of the evolution of sulfur species with and without added iron or pyrite

NASA Astrophysics Data System (ADS)

Li, Yubiao; Qian, Gujie; Brown, Paul L.; Gerson, Andrea R.

2017-09-01

Dissolution and oxidation of sulfide minerals play key roles in both acid and metalliferous rock drainage and supergene enrichment. Surface speciation heterogeneity, critical to understanding mechanisms of mineral sulfide dissolution, has to date largely not been considered. To this end synchrotron scanning photoelectron microscopy (SPEM) was employed to examine freshly fractured and partially dissolved chalcopyrite (CuFeS2) surfaces (pH 1.0 HClO4 solution, redox potential 650 mV relative to a standard hydrogen electrode, 75 °C). S2- (bulk), S22- and Sn2- were found to be present on all samples at varying concentrations. Oxidation was observed to take place heterogeneously at the sub-micron scale. As compared to chalcopyrite partially dissolved for 5 days, extended dissolution to 10 days did not show appreciably enhanced oxidation of surface species; however surface roughness increased markedly due to the growth/overlap of oxidised sulfur species. On addition of 4 mM iron both S0 and SO42- were observed but not SO32-, indicating that the greater Fe3+ activity/concentration promotes heterogeneous sulfur oxidation. On contact of pyrite (FeS2) with chalcopyrite, significantly greater chalcopyrite surface oxidation was observed than for the other systems examined, with S0, SO32- and SO42- being identified heterogeneously across the surface. It is proposed that chalcopyrite oxidative dissolution is enhanced by increasing its cathodic area, e.g. contacting with pyrite, while increased Fe3+ activity/concentration also contributes to increased dissolution rates. The high degree of surface heterogeneity of these surface products indicates that these surfaces are not passivated by their formation. These results suggest that chalcopyrite dissolution will be accelerated when in contact with pyrite at solution redox potential intermediate between the rest potentials of chalcopyrite and pyrite (560 mV and 660 mV, respectively) and/or iron rich acidic waters with resulting enhanced formation of secondary sulfur containing species and release of copper and iron. This in turn suggests accelerated supergene formation and enhanced metalliferous drainage under these conditions.

Analysis of plutonium isotope ratios including 238Pu/239Pu in individual U-Pu mixed oxide particles by means of a combination of alpha spectrometry and ICP-MS.

PubMed

Esaka, Fumitaka; Yasuda, Kenichiro; Suzuki, Daisuke; Miyamoto, Yutaka; Magara, Masaaki

2017-04-01

Isotope ratio analysis of individual uranium-plutonium (U-Pu) mixed oxide particles contained within environmental samples taken from nuclear facilities is proving to be increasingly important in the field of nuclear safeguards. However, isobaric interferences, such as 238 U with 238 Pu and 241 Am with 241 Pu, make it difficult to determine plutonium isotope ratios in mass spectrometric measurements. In the present study, the isotope ratios of 238 Pu/ 239 Pu, 240 Pu/ 239 Pu, 241 Pu/ 239 Pu, and 242 Pu/ 239 Pu were measured for individual Pu and U-Pu mixed oxide particles by a combination of alpha spectrometry and inductively coupled plasma mass spectrometry (ICP-MS). As a consequence, we were able to determine the 240 Pu/ 239 Pu, 241 Pu/ 239 Pu, and 242 Pu/ 239 Pu isotope ratios with ICP-MS after particle dissolution and chemical separation of plutonium with UTEVA resins. Furthermore, 238 Pu/ 239 Pu isotope ratios were able to be calculated by using both the 238 Pu/( 239 Pu+ 240 Pu) activity ratios that had been measured through alpha spectrometry and the 240 Pu/ 239 Pu isotope ratios determined through ICP-MS. Therefore, the combined use of alpha spectrometry and ICP-MS is useful in determining plutonium isotope ratios, including 238 Pu/ 239 Pu, in individual U-Pu mixed oxide particles. Copyright © 2016 Elsevier B.V. All rights reserved.

METHOD OF MAKING UO$sub 2$-Bi SLURRIES

DOEpatents

Hahn, H.T.

1960-05-24

A process is given of preparing an easily dispersible slurry of uranium dioxide in bismuth. A mixture of bismuth oxide, uranium, and bismuth are heated in a capsule to a temperature over the melting point of bismuth oxide. The amount of bismuth oxide used is less than that stoichiometrically required because the oxygen in the capsule also enters into the reaction.

METHOD FOR RECOVERING URANIUM FROM OILS

DOEpatents

Gooch, L.H.

1959-07-14

A method is presented for recovering uranium from hydrocarbon oils, wherein the uranium is principally present as UF/sub 4/. According to the invention, substantially complete removal of the uranium from the hydrocarbon oil may be effected by intimately mixing one part of acetone to about 2 to 12 parts of the hydrocarbon oil containing uranium and separating the resulting cake of uranium from the resulting mixture. The uranium in the cake may be readily recovered by burning to the oxide.

Rates and mechanisms of uranyl oxyhydroxide mineral dissolution

NASA Astrophysics Data System (ADS)

Reinoso-Maset, Estela; Steefel, Carl I.; Um, Wooyong; Chorover, Jon; O'Day, Peggy A.

2017-06-01

Uranyl oxyhydroxide minerals are important weathering products in uranium-contaminated surface and subsurface environments that regulate dissolved uranium (U) concentrations. However, dissolution rates for this class of minerals and associated dissolution mechanisms have not been previously reported for circumneutral pH conditions, particularly for the case of flow through porous media. In this work, the dissolution rates of K- and Na-compreignacite (K2(UO2)6O4(OH)6·8H2O and Na2(UO2)6O4(OH)6·8H2O, respectively) were measured using flow-through columns reacted with two simulated background porewater (BPW) solutions of low and high dissolved carbonate concentration (ca. 0.2 and 2.8 mmol L-1). Column materials were characterized before and after reaction with electron microscopy, bulk chemistry, and EXAFS to identify structural and chemical changes during dissolution and to obtain insight into molecular-scale processes. The reactive transport code CrunchFlow was used to calculate overall dissolution rates while accounting for fluid transport and changes in mineral volume and reactive surface area, and results were compared to steady-state dissolution rate calculations. In low carbonate BPW systems, interlayer K and Na were initially leached from both minerals, and in Na-compreignacite, K and minor divalent cations from the input solution were incorporated into the mineral structure. Results of characterization analyses suggested that after reaction both K- and Na-compreignacite resembled a disordered K-compreignacite with altered surfaces. A 10-fold increase in dissolved carbonate concentration and corresponding increase in pH (from 6.65 to 8.40) resulted in a net removal of 58-87% of total U mass from the columns, compared to <1% net loss in low carbonate BPW systems. Steady-state release of dissolved U was not observed with high carbonate solutions and post-reaction characterizations indicated a lack of development of leached or altered surfaces. Dissolution rates (normalized to specific surface area) were 2.5-3 orders-of-magnitude faster in high versus low carbonate BPW systems, with Na-compreignacite dissolving more rapidly than K-compreignacite under both BPW conditions, possibly due to greater ion exchange (1.57 · 10-10 vs. 1.28 · 10-13 mol m-2 s-1 [log R = -9.81 and -12.89] and 5.79 · 10-10 vs. 3.71 · 10-13 mol m-2 s-1 [log R = -9.24 and -12.43] for K- and Na-compreignacite, respectively). Experimental and spectroscopic results suggest that the dissolution rate is controlled by bond breaking of a uranyl group and detachment from polyhedral layers of the mineral structure. With higher dissolved carbonate concentrations, this rate-determining step is accelerated by the formation of Ca-uranyl carbonate complexes (dominant species under these conditions), which resulted in an increase of the dissolution rates. Optimization of both dissolution rate and mineral volume fraction in the reactive transport model to account for U mass removal during dissolution more accurately reproduced effluent data in high carbonate systems, and resulted in faster overall rates compared with a steady-state dissolution assumption. This study highlights the importance of coupling reaction and transport processes during the quantification of mineral dissolution rates to accurately predict the fate of contaminants such as U in porous geomedia.

SUMMARY TECHNICAL REPORT ON FEED MATERIALS FOR THE PERIOD APRIL 1, 1959 TO JUNE 30, 1959

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simmons, J.W. ed.

1959-07-20

Anaconda Acld, Kermac, Moab, Rifle, and Texas Zinc uranium concentrates were evaluated (the laboratory portlon of feed material evaluation). Laboratory equilibrium tests and Pilot Plant 2-inch-column extraction tests demonstrated effective distribution of uranium into a TBPkerosene solvent from aqueous phases containing as little as 0.5N HNO/sub 3/ and varying amounts of added metal nitrates (NaNO/sub 3/). The concentration of assoclated nitric acid in dilute aqueous nitric acld solutions was determined after values were obtained for the equillbrium constant for the reaction of tri-n-butyl phosphate with associated nitric acid and for the equilibrium distribution constant for the partition of associated nitricmore » acld into tri-n-butyl phosphate. Optimum partition of uranium into tri-n-butyl phosphate was realized in the laboratory by using an aqueous uranyl nitrate solution containing sufficient hydrogen ions to promote extraction and a low concentration of associated nitric acid. An Ohmart system for controlling the uranium profile in the A'' extractlon column was installed on Refinery pulse columns. Use of this system improved control but did not stop all column upsets. The effect of 13 to l89 ppm sodium contaminatlon upon hydrofluorination conversion of teraperature at the site of the reaction. Uranyl sulfate was shown to undergo an enantiotroplc transitlon at 755 deg C and to decompose to U/sub 3/O/sub 8/ in an atmosphere of oxygen sulfur dioxide, which gases are evolved during decoraposition. Decontamination of sodium, calcium, nickel, magnesium, gadolinium, and dysprosium was achieved in a laboratory investigatlon of the ADU process. UO/sub 2/ produced by reductions programmed from 700 to ll00 deg F was hydrofluorinated at programmed temperatures of 550 to 1100 deg F and isothermally at ll00 deg F. Good conversion was obtained for material whose source was ADU calcined at 1200 deg F. Uranium derbles were classified by the present method of derby grading and were then examined for slag coverage, slag volume, and slag weight. There was a high degree of overlap of these parameters for adjacent grades. A hydraulic separator for separatlng uranlum from magnesium and magnesium fluorlde was fabrlcated. Excellent separatlon was obtained for +l6 mesh material. A hydrochloric acid dissolution- UF/sub 4/ precipitation process for routing scrap materials to the reductlon-to- metal step was examined. The purification obtained was noted, and process conditions were varied to determine their effect upon UF/sub 4/ density, UF/sub 4/ purity and precipitation time. Three types of uranium scrap were subjected to the HCl dissolution-aqueous precipitation Winlo process to determine the purification achieved. Green salt made from dolomitlc bomb liner residues was found to be grossly contaminated. Acceptable green salt was raade from pickle liquor treated with formaldehyde and from pickle liquor plus black oxide. Nominal 80% yields were obtained in the recovery of magnesium metal by reaction of calcium carblde with magnesium fluoride slag and in the recovery of HF by the reactlon of sulfuric acid wlth magnesium fluoride slag. A sample holder for use in quantitative preferred orientation studies was fabricated. The holder, designed to fit a North American Philips Gonionweter, will accommodate specimens up to l 13/16 inches in diameter and incorporates a precision ball bearing. A satisfactory technique was developed for the analysis of uranium metal for traces of fluoride. A direct flame photometric method is glven for the determination of magnesium in uranium ore concentrates. No chemical separation step is required, except for high-iron-content ores. (auth)« less

Structural and silver/vanadium ratio effects on silver vanadium phosphorous oxide solution formation kinetics: impact on battery electrochemistry.

PubMed

Bock, David C; Takeuchi, Kenneth J; Marschilok, Amy C; Takeuchi, Esther S

2015-01-21

The detailed understanding of non-faradaic parasitic reactions which diminish battery calendar life is essential to the development of effective batteries for use in long life applications. The dissolution of cathode materials including manganese, cobalt and vanadium oxides in battery systems has been identified as a battery failure mechanism, yet detailed dissolution studies including kinetic analysis are absent from the literature. The results presented here provide a framework for the quantitative and kinetic analyses of the dissolution of cathode materials which will aid the broader community in more fully understanding this battery failure mechanism. In this study, the dissolution of silver vanadium oxide, representing the primary battery powering implantable cardioverter defibrillators (ICD), is compared with the dissolution of silver vanadium phosphorous oxide (Ag(w)VxPyOz) materials which were targeted as alternatives to minimize solubility. This study contains the first kinetic analyses of silver and vanadium solution formation from Ag0.48VOPO4·1.9H2O and Ag2VP2O8, in a non-aqueous battery electrolyte. The kinetic results are compared with those of Ag2VO2PO4 and Ag2V4O11 to probe the relationships among crystal structure, stoichiometry, and solubility. For vanadium, significant dissolution was observed for Ag2V4O11 as well as for the phosphate oxide Ag0.49VOPO4·1.9H2O, which may involve structural water or the existence of multiple vanadium oxidation states. Notably, the materials from the SVPO family with the lowest vanadium solubility are Ag2VO2PO4 and Ag2VP2O8. The low concentrations and solution rates coupled with their electrochemical performance make these materials interesting alternatives to Ag2V4O11 for the ICD application.

Evidence for single metal two electron oxidative addition and reductive elimination at uranium.

PubMed

Gardner, Benedict M; Kefalidis, Christos E; Lu, Erli; Patel, Dipti; McInnes, Eric J L; Tuna, Floriana; Wooles, Ashley J; Maron, Laurent; Liddle, Stephen T

2017-12-01

Reversible single-metal two-electron oxidative addition and reductive elimination are common fundamental reactions for transition metals that underpin major catalytic transformations. However, these reactions have never been observed together in the f-block because these metals exhibit irreversible one- or multi-electron oxidation or reduction reactions. Here we report that azobenzene oxidises sterically and electronically unsaturated uranium(III) complexes to afford a uranium(V)-imido complex in a reaction that satisfies all criteria of a single-metal two-electron oxidative addition. Thermolysis of this complex promotes extrusion of azobenzene, where H-/D-isotopic labelling finds no isotopomer cross-over and the non-reactivity of a nitrene-trap suggests that nitrenes are not generated and thus a reductive elimination has occurred. Though not optimally balanced in this case, this work presents evidence that classical d-block redox chemistry can be performed reversibly by f-block metals, and that uranium can thus mimic elementary transition metal reactivity, which may lead to the discovery of new f-block catalysis.

High-temperature electrically conductive ceramic composite and method for making same

DOEpatents

Beck, David E.; Gooch, Jack G.; Holcombe, Jr., Cressie E.; Masters, David R.

1983-01-01

The present invention relates to a metal-oxide ceramic composition useful in induction heating applications for treating uranium and uranium alloys. The ceramic composition is electrically conductive at room temperature and is nonreactive with molten uranium. The composition is prepared from a particulate admixture of 20 to 50 vol. % niobium and zirconium oxide which may be stabilized with an addition of a further oxide such as magnesium oxide, calcium oxide, or yttria. The composition is prepared by blending the powders, pressing or casting the blend into the desired product configuration, and then sintering the casting or compact in an inert atmosphere. In the casting operation, calcium aluminate is preferably added to the admixture in place of a like quantity of zirconia for providing a cement to help maintain the integrity of the sintered product.

Uranium metallogenesis of the peraluminous leucogranite from the Pontivy-Rostrenen magmatic complex (French Armorican Variscan belt): the result of long-term oxidized hydrothermal alteration during strike-slip deformation

NASA Astrophysics Data System (ADS)

Ballouard, C.; Poujol, M.; Mercadier, J.; Deloule, E.; Boulvais, P.; Baele, J. M.; Cuney, M.; Cathelineau, M.

2018-06-01

In the French Armorican Variscan belt, most of the economically significant hydrothermal U deposits are spatially associated with peraluminous leucogranites emplaced along the south Armorican shear zone (SASZ), a dextral lithospheric scale wrench fault that recorded ductile deformation from ca. 315 to 300 Ma. In the Pontivy-Rostrenen complex, a composite intrusion, the U mineralization is spatially associated with brittle structures related to deformation along the SASZ. In contrast to monzogranite and quartz monzodiorite (3 < U < 9 ppm; Th/U > 3), the leucogranite samples are characterized by highly variable U contents ( 3 to 27 ppm) and Th/U ratios ( 0.1 to 5) suggesting that the crystallization of magmatic uranium oxide in the more evolved facies was followed by uranium oxide leaching during hydrothermal alteration and/or surface weathering. U-Pb dating of uranium oxides from the deposits reveals that they mostly formed between ca. 300 and 270 Ma. In monzogranite and quartz monzodiorite, apatite grains display magmatic textures and provide U-Pb ages of ca. 315 Ma reflecting the time of emplacement of the intrusions. In contrast, apatite grains from the leucogranite display textural, geochemical, and geochronological evidences for interaction with U-rich oxidized hydrothermal fluids contemporaneously with U mineralizing events. From 300 to 270 Ma, infiltration of surface-derived oxidized fluids leached magmatic uranium oxide from fertile leucogranite and formed U deposits. This phenomenon was sustained by brittle deformation and by the persistence of thermal anomalies associated with U-rich granitic bodies.

Thermal properties of nonstoichiometry uranium dioxide

NASA Astrophysics Data System (ADS)

Kavazauri, R.; Pokrovskiy, S. A.; Baranov, V. G.; Tenishev, A. V.

2016-04-01

In this paper, was developed a method of oxidation pure uranium dioxide to a predetermined deviation from the stoichiometry. Oxidation was carried out using the thermogravimetric method on NETZSCH STA 409 CD with a solid electrolyte galvanic cell for controlling the oxygen potential of the environment. 4 samples uranium oxide were obtained with a different ratio of oxygen-to-metal: O / U = 2.002, O / U = 2.005, O / U = 2.015, O / U = 2.033. For the obtained samples were determined basic thermal characteristics of the heat capacity, thermal diffusivity, thermal conductivity. The error of heat capacity determination is equal to 5%. Thermal diffusivity and thermal conductivity of the samples decreased with increasing deviation from stoichiometry. For the sample with O / M = 2.033, difference of both values with those of stoichiometric uranium dioxide is close to 50%.

Synergistic effect of reductive and ligand-promoted dissolution of goethite.

PubMed

Wang, Zimeng; Schenkeveld, Walter D C; Kraemer, Stephan M; Giammar, Daniel E

2015-06-16

Ligand-promoted dissolution and reductive dissolution of iron (hydr)oxide minerals control the bioavailability of iron in many environmental systems and have been recognized as biological iron acquisition strategies. This study investigated the potential synergism between ligands (desferrioxamine B (DFOB) or N,N'-Di(2-hydroxybenzyl)ethylenediamine-N,N'-diacetic acid (HBED)) and a reductant (ascorbate) in goethite dissolution. Batch experiments were performed at pH 6 with ligand or reductant alone and in combination, and under both oxic and anoxic conditions. Goethite dissolution in the presence of reductant or ligand alone followed classic surface-controlled dissolution kinetics. Ascorbate alone does not promote goethite dissolution under oxic conditions due to rapid reoxidation of Fe(II). The rate coefficients for goethite dissolution by ligands are closely correlated with the stability constants of the aqueous Fe(III)-ligand complexes. A synergistic effect of DFOB and ascorbate on the rate of goethite dissolution was observed (total rates greater than the sum of the individual rates), and this effect was most pronounced under oxic conditions. For HBED, macroscopically the synergistic effect was hidden due to the inhibitory effect of ascorbate on HBED adsorption. After accounting for the concentrations of adsorbed ascorbate and HBED, a synergistic effect could still be identified. The potential synergism between ligand and reductant for iron (hydr)oxide dissolution may have important implications for iron bioavailability in soil environments.

New Technique for Speciation of Uranium in Sediments Following Acetate-Stimulated Bioremediation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

2011-06-22

Acetate-stimulated bioremediation is a promising new technique for sequestering toxic uranium contamination from groundwater. The speciation of uranium in sediments after such bioremediation attempts remains unknown as a result of low uranium concentration, and is important to analyzing the stability of sequestered uranium. A new technique was developed for investigating the oxidation state and local molecular structure of uranium from field site sediments using X-Ray Absorption Spectroscopy (XAS), and was implemented at the site of a former uranium mill in Rifle, CO. Glass columns filled with bioactive Rifle sediments were deployed in wells in the contaminated Rifle aquifer and amendedmore » with a hexavalent uranium (U(VI)) stock solution to increase uranium concentration while maintaining field conditions. This sediment was harvested and XAS was utilized to analyze the oxidation state and local molecular structure of the uranium in sediment samples. Extended X-Ray Absorption Fine Structure (EXAFS) data was collected and compared to known uranium spectra to determine the local molecular structure of the uranium in the sediment. Fitting was used to determine that the field site sediments did not contain uraninite (UO{sub 2}), indicating that models based on bioreduction using pure bacterial cultures are not accurate for bioremediation in the field. Stability tests on the monomeric tetravalent uranium (U(IV)) produced by bioremediation are needed in order to assess the efficacy of acetate-stimulation bioremediation.« less

Module-oriented modeling of reactive transport with HYTEC

NASA Astrophysics Data System (ADS)

van der Lee, Jan; De Windt, Laurent; Lagneau, Vincent; Goblet, Patrick

2003-04-01

The paper introduces HYTEC, a coupled reactive transport code currently used for groundwater pollution studies, safety assessment of nuclear waste disposals, geochemical studies and interpretation of laboratory column experiments. Based on a known permeability field, HYTEC evaluates the groundwater flow paths, and simulates the migration of mobile matter (ions, organics, colloids) subject to geochemical reactions. The code forms part of a module-oriented structure which facilitates maintenance and improves coding flexibility. In particular, using the geochemical module CHESS as a common denominator for several reactive transport models significantly facilitates the development of new geochemical features which become automatically available to all models. A first example shows how the model can be used to assess migration of uranium from a sub-surface source under the effect of an oxidation front. The model also accounts for alteration of hydrodynamic parameters (local porosity, permeability) due to precipitation and dissolution of mineral phases, which potentially modifies the migration properties in general. The second example illustrates this feature.

On the stability of sub-stoichiometric uranium oxides

NASA Astrophysics Data System (ADS)

Winer, K.; Colmenares, C. A.; Smith, R. L.; Wooten, F.

1986-12-01

The oxidation of clean, high-purity polycrystalline uranium metal surfaces for low exposures to dry oxygen was studied with AES and XPS in an attempt to substantiate claims for the formation of a stable UO surface phase at ambient temperatures. We found no evidence for such a surface phase and found instead that grossly sub-stoichiometric surface oxides were formed after sequential oxygen saturation and heating.

Retardation of uranium and thorium by a cementitious backfill developed for radioactive waste disposal.

PubMed

Felipe-Sotelo, M; Hinchliff, J; Field, L P; Milodowski, A E; Preedy, O; Read, D

2017-07-01

The solubility of uranium and thorium has been measured under the conditions anticipated in a cementitious, geological disposal facility for low and intermediate level radioactive waste. Similar solubilities were obtained for thorium in all media, comprising NaOH, Ca(OH) 2 and water equilibrated with a cement designed as repository backfill (NRVB, Nirex Reference Vault Backfill). In contrast, the solubility of U(VI) was one order of magnitude higher in NaOH than in the remaining solutions. The presence of cellulose degradation products (CDP) results in a comparable solubility increase for both elements. Extended X-ray Absorption Fine Structure (EXAFS) data suggest that the solubility-limiting phase for uranium corresponds to a becquerelite-type solid whereas thermodynamic modelling predicts a poorly crystalline, hydrated calcium uranate phase. The solubility-limiting phase for thorium was ThO 2 of intermediate crystallinity. No breakthrough of either uranium or thorium was observed in diffusion experiments involving NRVB after three years. Nevertheless, backscattering electron microscopy and microfocus X-ray fluorescence confirmed that uranium had penetrated about 40 μm into the cement, implying active diffusion governed by slow dissolution-precipitation kinetics. Precise identification of the uranium solid proved difficult, displaying characteristics of both calcium uranate and becquerelite. Copyright © 2017 Elsevier Ltd. All rights reserved.

Physicochemical characterization of Capstone depleted uranium aerosols III: morphologic and chemical oxide analyses.

PubMed

Krupka, Kenneth M; Parkhurst, Mary Ann; Gold, Kenneth; Arey, Bruce W; Jenson, Evan D; Guilmette, Raymond A

2009-03-01

The impact of depleted uranium (DU) penetrators against an armored target causes erosion and fragmentation of the penetrators, the extent of which is dependent on the thickness and material composition of the target. Vigorous oxidation of the DU particles and fragments creates an aerosol of DU oxide particles and DU particle agglomerations combined with target materials. Aerosols from the Capstone DU aerosol study, in which vehicles were perforated by DU penetrators, were evaluated for their oxidation states using x-ray diffraction (XRD), and particle morphologies were examined using scanning electron microscopy/energy dispersive spectroscopy (SEM/EDS). The oxidation state of a DU aerosol is important as it offers a clue to its solubility in lung fluids. The XRD analysis showed that the aerosols evaluated were a combination primarily of U3O8 (insoluble) and UO3 (relatively more soluble) phases, though intermediate phases resembling U4O9 and other oxides were prominent in some samples. Analysis of particle residues in the micrometer-size range by SEM/EDS provided microstructural information such as phase composition and distribution, fracture morphology, size distribution, and material homogeneity. Observations from SEM analysis show a wide variability in the shapes of the DU particles. Some of the larger particles were spherical, occasionally with dendritic or lobed surface structures. Others appear to have fractures that perhaps resulted from abrasion and comminution, or shear bands that developed from plastic deformation of the DU material. Amorphous conglomerates containing metals other than uranium were also common, especially with the smallest particle sizes. A few samples seemed to contain small bits of nearly pure uranium metal, which were verified by EDS to have a higher uranium content exceeding that expected for uranium oxides. Results of the XRD and SEM/EDS analyses were used in other studies described in this issue of Health Physics to interpret the results of lung solubility studies and in selecting input parameters for dose assessments.

Physicochemical Characterization of Capstone Depleted Uranium Aerosols III: Morphologic and Chemical Oxide Analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krupka, Kenneth M.; Parkhurst, MaryAnn; Gold, Kenneth

2009-03-01

The impact of depleted uranium (DU) penetrators against an armored target causes erosion and fragmentation of the penetrators, the extent of which is dependent on the thickness and material composition of the target. Vigorous oxidation of the DU particles and fragments creates an aerosol of DU oxide particles and DU particle agglomerations combined with target materials. Aerosols from the Capstone DU aerosol study, in which vehicles were perforated by DU penetrators, were evaluated for their oxidation states using X-ray diffraction (XRD) and particle morphologies using scanning electron microscopy/energy dispersive spectrometry (SEM/EDS). The oxidation state of a DU aerosol is importantmore » as it offers a clue to its solubility in lung fluids. The XRD analysis showed that the aerosols evaluated were a combination primarily of U3O8 (insoluble) and UO3 (relatively more soluble) phases, though intermediate phases resembling U4O9 and other oxides were prominent in some samples. Analysis of particle residues in the micrometer-size range by SEM/EDS provided microstructural information such as phase composition and distribution, fracture morphology, size distribution, and material homogeneity. Observations from SEM analysis show a wide variability in the shapes of the DU particles. Some of the larger particles appear to have been fractured (perhaps as a result of abrasion and comminution); others were spherical, occasionally with dendritic or lobed surface structures. Amorphous conglomerates containing metals other than uranium were also common, especially with the smallest particle sizes. A few samples seemed to contain small chunks of nearly pure uranium metal, which were verified by EDS to have a higher uranium content exceeding that expected for uranium oxides. Results of the XRD and SEM/EDS analyses were used in other studies described in this issue of The Journal of Health Physics to interpret the results of lung solubility studies and in selecting input parameters for dose assessments.« less

U.sup.+4 generation in HTER

DOEpatents

Miller, William E [Naperville, IL; Gay, Eddie C [Park Forest, IL; Tomczuk, Zygmunt [Homer Glen, IL

2006-03-14

A improved device and process for recycling spent nuclear fuels, in particular uranium metal, that facilitates the refinement and recovery of uranium metal from spent metallic nuclear fuels. The electrorefiner device comprises two anodes in predetermined spatial relation to a cathode. The anodese have separate current and voltage controls. A much higher voltage than normal for the electrorefining process is applied to the second anode, thereby facilitating oxidization of uranium (III), U.sup.+, to uranium (IV), U.sup.+4. The current path from the second anode to the cathode is physically shorter than the similar current path from the second anode to the spent nuclear fuel contained in a first anode shaped as a basket. The resulting U.sup.+4 oxidizes and solubilizes rough uranium deposited on the surface of the cathode. A softer uranium metal surface is left on the cathode and is more readily removed by a scraper.

Isolation and characterization of a uranium(VI)-nitride triple bond

NASA Astrophysics Data System (ADS)

King, David M.; Tuna, Floriana; McInnes, Eric J. L.; McMaster, Jonathan; Lewis, William; Blake, Alexander J.; Liddle, Stephen T.

2013-06-01

The nature and extent of covalency in uranium bonding is still unclear compared with that of transition metals, and there is great interest in studying uranium-ligand multiple bonds. Although U=O and U=NR double bonds (where R is an alkyl group) are well-known analogues to transition-metal oxo and imido complexes, the uranium(VI)-nitride triple bond has long remained a synthetic target in actinide chemistry. Here, we report the preparation of a uranium(VI)-nitride triple bond. We highlight the importance of (1) ancillary ligand design, (2) employing mild redox reactions instead of harsh photochemical methods that decompose transiently formed uranium(VI) nitrides, (3) an electrostatically stabilizing sodium ion during nitride installation, (4) selecting the right sodium sequestering reagent, (5) inner versus outer sphere oxidation and (6) stability with respect to the uranium oxidation state. Computational analyses suggest covalent contributions to U≡N triple bonds that are surprisingly comparable to those of their group 6 transition-metal nitride counterparts.

Significance of a Shelf-wide Dissolution Event during the Paleocene-Eocene Thermal Maximum, Maryland and New Jersey, USA

NASA Astrophysics Data System (ADS)

Bralower, T. J.; Kump, L. R.; Robinson, M. M.; Self-Trail, J. M.; Zachos, J. C.

2016-12-01

Continental-shelf sediments of the US Atlantic margin experienced a brief episode of carbonate dissolution during the onset of the Paleocene-Eocene Thermal Maximum (PETM). Dissolution is represented by reduced percentages of carbonate, and calcareous microfossil distribution and preservation trends, in cores from Maryland and New Jersey. The base and the top of the dissolution zone are abrupt compared to the gradual nature of the onset of the carbon isotope excursion (CIE). The thickness of the dissolution zone varies from 9 cm in the Bass River core (outer paleoshelf) to 1.6 m in the CamDor core (middle paleoshelf). The decrease in %CaCO3 suggests dissolution locally removed 83 to 100% of the initial biogenic carbonate. Shelf-wide dissolution during the onset of the PETM may be a regional event, associated, for example, with eutrophication. Samples from across the paleoshelf contain abundant fine-grained framboidal pyrite, which suggests photic-zone euxinia occurred before, during, and after the dissolution event. Dissolution may also be associated with oxidation of this pyrite during later exposure to oxidizing groundwaters, although the restricted duration of the dissolution interval argues against this. Alternatively, the dissolution event may have global significance related to surface ocean-water acidification or shoaling of the calcite compensation depth (CCD) to shelf depths. The event began near the onset of the CIE on the shelf, whereas dissolution in deep-sea sections began later. Earlier shelf dissolution is consistent with surface ocean acidification while later deep-sea dissolution is thought to be associated with shoaling of the CCD. In our presentation, we weigh evidence for each of these possibilities and test them using the global dataset.

Unraveling uranium induced oxidative stress related responses in Arabidopsis thaliana seedlings. Part II: responses in the leaves and general conclusions.

PubMed

Vanhoudt, Nathalie; Cuypers, Ann; Horemans, Nele; Remans, Tony; Opdenakker, Kelly; Smeets, Karen; Bello, Daniel Martinez; Havaux, Michel; Wannijn, Jean; Van Hees, May; Vangronsveld, Jaco; Vandenhove, Hildegarde

2011-06-01

The cellular redox balance seems an important modulator under heavy metal stress. While for other heavy metals these processes are well studied, oxidative stress related responses are also known to be triggered under uranium stress but information remains limited. This study aimed to further unravel the mechanisms by which plants respond to uranium stress. Seventeen-day-old Arabidopsis thaliana seedlings, grown on a modified Hoagland solution under controlled conditions, were exposed to 0, 0.1, 1, 10 and 100 μM uranium for 1, 3 and 7 days. While in Part I of this study oxidative stress related responses in the roots were discussed, this second Part II discusses oxidative stress related responses in the leaves and general conclusions drawn from the results of the roots and the leaves will be presented. As several responses were already visible following 1 day exposure, when uranium concentrations in the leaves were negligible, a root-to-shoot signaling system was suggested in which plastids could be important sensing sites. While lipid peroxidation, based on the amount of thiobarbituric acid reactive compounds, was observed after exposure to 100 μM uranium, affecting membrane structure and function, a transient concentration dependent response pattern was visible for lipoxygenase initiated lipid peroxidation. This transient character of uranium stress responses in leaves was emphasized by results of lipoxygenase (LOX2) and antioxidative enzyme transcript levels, enzyme capacities and glutathione concentrations both in time as with concentration. The ascorbate redox balance seemed an important modulator of uranium stress responses in the leaves as in addition to the previous transient responses, the total ascorbate concentration and ascorbate/dehydroascorbate redox balance increased in a concentration and time dependent manner. This could represent either a slow transient response or a stable increase with regard to plant acclimation to uranium stress. Copyright © 2011 Elsevier Ltd. All rights reserved.

Investigation of the Dissolution-Reformation Cycle of the Passive Oxide Layer on NiTi Orthodontic Archwires

NASA Astrophysics Data System (ADS)

Uzer, B.; Birer, O.; Canadinc, D.

2017-09-01

Dissolution-reformation cycle of the passive oxide layer on the nickel-titanium (NiTi) orthodontic archwires was investigated, which has recently been recognized as one of the key parameters dictating the biocompatibility of archwires. Specifically, commercially available NiTi orthodontic archwires were immersed in artificial saliva solutions of different pH values (2.3, 3.3, and 4.3) for four different immersion periods: 1, 7, 14, and 30 days. Characterization of the virgin and tested samples revealed that the titanium oxide layer on the NiTi archwire surfaces exhibit a dissolution-reformation cycle within the first 14 days of the immersion period: the largest amount of Ni ion release occurred within the first week of immersion, while it significantly decreased during the reformation period from day 7 to day 14. Furthermore, the oxide layer reformation was catalyzed on the grooves within the peaks and valleys due to relatively larger surface energy of these regions, which eventually decreased the surface roughness significantly within the reformation period. Overall, the current results clearly demonstrate that the analyses of dissolution-reformation cycle of the oxide layer in orthodontic archwires, surface roughness, and ion release behavior constitute utmost importance in order to ensure both the highest degree of biocompatibility and an efficient medical treatment.

Performance of a Steel/Oxide Composite Waste Form for Combined Waste Steams from Advanced Electrochemical Processes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Indacochea, J. E.; Gattu, V. K.; Chen, X.

The results of electrochemical corrosion tests and modeling activities performed collaboratively by researchers at the University of Illinois at Chicago and Argonne National Laboratory as part of workpackage NU-13-IL-UIC-0203-02 are summarized herein. The overall objective of the project was to develop and demonstrate testing and modeling approaches that could be used to evaluate the use of composite alloy/ceramic materials as high-level durable waste forms. Several prototypical composite waste form materials were made from stainless steels representing fuel cladding, reagent metals representing metallic fuel waste streams, and reagent oxides representing oxide fuel waste streams to study the microstructures and corrosion behaviorsmore » of the oxide and alloy phases. Microelectrodes fabricated from small specimens of the composite materials were used in a series of electrochemical tests to assess the corrosion behaviors of the constituent phases and phase boundaries in an aggressive acid brine solution at various imposed surface potentials. The microstructures were characterized in detail before and after the electrochemical tests to relate the electrochemical responses to changes in both the electrode surface and the solution composition. The results of microscopic, electrochemical, and solution analyses were used to develop equivalent circuit and physical models representing the measured corrosion behaviors of the different materials pertinent to long-term corrosion behavior. This report provides details regarding (1) the production of the composite materials, (2) the protocol for the electrochemical measurements and interpretations of the responses of multi-phase alloy and oxide composites, (3) relating corrosion behaviors to microstructures of multi-phase alloys based on 316L stainless steel and HT9 (410 stainless steel was used as a substitute) with added Mo, Ni, and/or Mn, and (4) modeling the corrosion behaviors and rates of several alloy/oxide composite materials made with added lanthanide and uranium oxides. These analyses show the corrosion behaviors of the alloy/ceramic composite materials are very similar to the corrosion behaviors of multi-phase alloy waste forms, and that the presence of oxide inclusions does not impact the corrosion behaviors of the alloy phases. Mixing with metallic waste streams is beneficial to lanthanide and uranium oxides in that they react with Zr in the fuel waste to form highly durable zirconates. The measured corrosion behaviors suggest properly formulated composite materials would be suitable waste forms for combined metallic and oxide waste streams generated during electrometallurgical reprocessing of spent nuclear fuel. Electrochemical methods are suitable for evaluating the durability and modeling long-term behavior of composite waste forms: the degradation model developed for metallic waste forms can be applied to the alloy phases formed in the composite and an affinity-based mineral dissolution model can be applied to the ceramic phases.« less

Rapid extraction and assay of uranium from environmental surface samples

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barrett, Christopher A.; Chouyyok, Wilaiwan; Speakman, Robert J.

Extraction methods enabling faster removal and concentration of uranium compounds for improved trace and low-level assay are demonstrated for standard surface sampling material in support of nuclear safeguards efforts, health monitoring, and other nuclear analysis applications. A key problem with the existing surface sampling swipes is the requirement for complete digestion of sample and sampling matrix. This is a time-consuming and labour-intensive process that limits laboratory throughput, elevates costs, and increases background levels. Various extraction methods are explored for their potential to quickly and efficiently remove different chemical forms of uranium from standard surface sampling material. A combination of carbonatemore » and peroxide solutions is shown to give the most rapid and complete form of uranyl compound extraction and dissolution. This rapid extraction process is demonstrated to be compatible with standard inductive coupled plasma mass spectrometry methods for uranium isotopic assay as well as screening techniques such as x-ray fluorescence. The general approach described has application beyond uranium to other analytes of nuclear forensic interest (e.g., rare earth elements and plutonium) as well as heavy metals for environmental and industrial hygiene monitoring.« less

Sensitivity evaluation of the green alga Chlamydomonas reinhardtii to uranium by pulse amplitude modulated (PAM) fluorometry.

PubMed

Herlory, Olivier; Bonzom, Jean-Marc; Gilbin, Rodolphe

2013-09-15

Although ecotoxicological studies tend to address the toxicity thresholds of uranium in freshwaters, there is a lack of information on the effects of the metal on physiological processes, particularly in aquatic plants. Knowing that uranium alters photosynthesis via impairment of the water photo-oxidation process, we determined whether pulse amplitude modulated (PAM) fluorometry was a relevant tool for assessing the impact of uranium on the green alga Chlamydomonas reinhardtii and investigated how and to what extent uranium hampered photosynthetic performance. Photosynthetic activity and quenching were assessed from fluorescence induction curves generated by PAM fluorometry, after 1 and 5h of uranium exposure in controlled conditions. The oxygen-evolving complex (OEC) of PSII was identified as the primary action site of uranium, through alteration of the water photo-oxidation process as revealed by F0/Fv. Limiting re-oxidation of the plastoquinone pool, uranium impaired the electron flux between the photosystems until almost complete inhibition of the PSII quantum efficiency ( [Formula: see text] , EC50=303 ± 64 μg UL(-1) after 5h of exposure) was observed. Non-photochemical quenching (qN) was identified as the most sensitive fluorescence parameter (EC50=142 ± 98 μg UL(-1) after 5h of exposure), indicating that light energy not used in photochemistry was dissipated in non-radiative processes. It was shown that parameters which stemmed from fluorescence induction kinetics are valuable indicators for evaluating the impact of uranium on PSII in green algae. PAM fluorometry provided a rapid and reasonably sensitive method for assessing stress response to uranium in microalgae. Copyright © 2013 Elsevier B.V. All rights reserved.

A model of early formation of uranium molecular oxides in laser-ablated plasmas

NASA Astrophysics Data System (ADS)

Finko, Mikhail S.; Curreli, Davide; Weisz, David G.; Crowhurst, Jonathan C.; Rose, Timothy P.; Koroglu, Batikan; Radousky, Harry B.; Armstrong, Michael R.

2017-12-01

In this work, we present a newly constructed U x O y reaction mechanism that consists of 30 reaction channels (21 of which are reversible channels) for 11 uranium molecular species (including ions). Both the selection of reaction channels and calculation of corresponding rate coefficients is accomplished via a comprehensive literature review and application of basic reaction rate theory. The reaction mechanism is supplemented by a detailed description of oxygen plasma chemistry (19 species and 142 reaction channels) and is used to model an atmospheric laser ablated uranium plume via a 0D (global) model. The global model is used to analyze the evolution of key uranium molecular species predicted by the reaction mechanism, and the initial stage of formation of uranium oxide species.

Use of carbon paste electrodes for the voltammetric detection of silver leached from the oxidative dissolution of silver nanoparticles

NASA Astrophysics Data System (ADS)

Mullaugh, Katherine M.; Pearce, Olivia M.

2017-04-01

The widespread use of silver nanoparticles (Ag NPs) in consumer goods has raised concerns about the release of silver in environmental waters. Of particular concern is the oxidative dissolution of Ag NPs to release Ag+ ions, which are highly toxic to many aquatic organisms. Here, we have investigated the application of differential pulse stripping voltammetry (DPSV) with carbon paste electrodes (CPEs) in monitoring the oxidation of Ag NPs. Using a commercially available, unmodified carbon paste and 60-s deposition times, a detection limit of 3 nM Ag+ could be achieved. We demonstrate its selectivity for free Ag+ ions over Ag nanoparticles, allowing for analysis of the oxidation of Ag NPs without the need for separation of ions and nanoparticles prior to analysis. We applied this approach to investigate the effect of pH in the oxidative dissolution of Ag NPs, demonstrating the usefulness of CPEs in studies of this type.

Comparison of bio-dissolution of spent Ni-Cd batteries by sewage sludge using ferrous ions and elemental sulfur as substrate.

PubMed

Zhao, Ling; Zhu, Nan-Wen; Wang, Xiao-Hui

2008-01-01

Bioleaching of spent Ni-Cd batteries using acidified sewage sludge was carried out in a continuous flow two-step leaching system including an acidifying reactor and a leaching reactor. Two systems operated about 30d to achieve almost complete dissolution of heavy metals Ni, Cd and Co in four Ni-Cd batteries. Ferrous sulphate and elemental sulfur were used as two different substrates to culture indigenous thiobacilli in sewage sludge. pH and ORP of the acidifying reactor was stabilized around 2.3 and 334mV for the iron-oxidizing system and 1.2 and 390mV for the sulfur-oxidizing system. It was opposite to the acidifying reactor, the pH/ORP in the leaching reactor of the iron-oxidizing system was relatively lower/higher than that of the sulphur-oxidizing system in the first 17d. The metal dissolution, in the first 12-16d, was faster in the iron-oxidizing system than in the sulphur-oxidizing system due to the lower pH. In the iron-oxidizing system, the maximum solubilization of cadmium (2500mg l(-1)) and cobalt (260mg l(-1)) can be reached at day 6-8 and the most of metal nickel was leached in the first 16d. But in the sulphur-oxidizing system there was a lag period of 4-8d to reach the maximum solubilization of cadmium and cobalt. The maximum dissolution of nickel hydroxide (1400mg l(-1)) and metallic nickel (2300mg l(-1)) occurred at about day 12 and day 20, respectively.

Determination of the solubility and size distribution of radioactive aerosols in the uranium processing plant at NRCN.

PubMed

Kravchik, T; Oved, S; Paztal-Levy, O; Pelled, O; Gonen, R; German, U; Tshuva, A

2008-01-01

Inhalation is the main route of internal exposure to radioactive aerosols in the nuclear industry. To assess the radiation dose from the intake of these aerosols, it is necessary to know their physical (aerodynamic diameter distribution) and chemical (dissolution rate in extracellular lung fluid) characteristics. Air samples were taken from the uranium processing plant at the Nuclear Research Center, Negev. Measurements of aerodynamic diameter distribution using a cascade impactor indicated an average activity median aerodynamic diameter value close to 5 microm, in accordance with the recent recommended values of International Commission on Radiological Protection (ICRP) model. Solubility profiles of these aerosols were determined by performing in vitro solubility tests over 100 d in a simultant solution of the extracellular fluid. The tests indicated that the uranium aerosols should be assigned to an absorption between Types M and S (as defined by the ICRP Publication 66 model).

Carbothermic Synthesis of 820 m UN Kernels: Literature Review, Thermodynamics, Analysis, and Related Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lindemer, Terrence; Voit, Stewart L; Silva, Chinthaka M

2014-01-01

The U.S. Department of Energy is considering a new nuclear fuel that would be less susceptible to ruptures during a loss-of-coolant accident. The fuel would consist of tristructural isotropic coated particles with large, dense uranium nitride (UN) kernels. This effort explores many factors involved in using gel-derived uranium oxide-carbon microspheres to make large UN kernels. Analysis of recent studies with sufficient experimental details is provided. Extensive thermodynamic calculations are used to predict carbon monoxide and other pressures for several different reactions that may be involved in conversion of uranium oxides and carbides to UN. Experimentally, the method for making themore » gel-derived microspheres is described. These were used in a microbalance with an attached mass spectrometer to determine details of carbothermic conversion in argon, nitrogen, or vacuum. A quantitative model is derived from experiments for vacuum conversion to an uranium oxide-carbide kernel.« less

Stability of silver nanoparticles: agglomeration and oxidation in biological relevant conditions

NASA Astrophysics Data System (ADS)

Valenti, Laura E.; Giacomelli, Carla E.

2017-05-01

Silver nanoparticles (Ag-NP) are the most used nanomaterial in consumer products due to the intrinsic antimicrobial capacity of silver. However, Ag-NP may be also harmful to algae, aquatic species, mammalian cells, and higher plants because both Ag+ and nanoparticles are responsible of cell damages. The oxidative dissolution of Ag-NP would proceed to completion under oxic conditions, but the rate and extent of the dissolution depend on several factors. This work correlates the effect of the capping agent (albumin and citrate) with the stability of Ag-NP towards agglomeration in simulated body fluid (SBF) and oxidation in the presence of ROS species (H2O2). Capping provides colloidal stability only through electrostatic means, whereas albumin acts as bulky ligands giving steric and electrostatic repulsion, inhibiting the agglomeration in SBF. However, citrate capping protects Ag-NP from dissolution to a major extent than albumin does because of its reducing power. Moreover, citrate in solution minimizes the oxidation of albumin-coated Ag-NP even after long incubation times. H2O2-induced dissolution proceeds to completion with Ag-NP incubated in SBF, while incubation in citrate leads to an incomplete oxidation. In short, albumin is an excellent capping agent to minimize Ag-NP agglomeration whereas citrate provides a mild-reductive medium that prevents dissolution in biological relevant media as well as in the presence of ROS species. These results provide insight into how the surface properties and media composition affect the release of Ag+ from Ag-NP, related to the cell toxicity and relevant to the storage and lifetime of silver-containing nanomaterials.

Dissolution of Biogenic and Synthetic UO2 under Varied Reducing Conditions

PubMed Central

ULRICH, KAI – UWE; SINGH, ABHAS; SCHOFIELD, ELEANOR J.; BARGAR, JOHN R.; VEERAMANI, HARISH; SHARP, JONATHAN O.; LATMANI, RIZLAN BERNIER -; GIAMMAR, DANIEL E.

2008-01-01

The chemical stability of biogenic UO2, a nanoparticulate product of environmental bioremediation, may be impacted by the particles’ surface free energy, structural defects, and compositional variability in analogy to abiotic UO2+x (0 ≤ x ≤ 0.25). This study quantifies and compares intrinsic solubility and dissolution rate constants of biogenic nano-UO2 and synthetic bulk UO2.00, taking molecular-scale structure into account. Rates were determined under anoxic conditions as a function of pH and dissolved inorganic carbon in continuous-flow experiments. The dissolution rates of biogenic and synthetic UO2 solids were lowest at near neutral pH and increased with decreasing pH. Similar surface area-normalized rates of biogenic and synthetic UO2 suggest comparable reactive surface site densities. This finding is consistent with the identified structural homology of biogenic UO2 and stoichiometric UO2.00. Compared to carbonate-free anoxic conditions, dissolved inorganic carbon accelerated the dissolution rate of biogenic UO2 by 3 orders of magnitude. This phenomenon suggests continuous surface oxidation of U(IV) to U(VI), with detachment of U(VI) as the rate-determining step in dissolution. Although reducing conditions were maintained throughout the experiments, the UO2 surface can be oxidized by water and radiogenic oxidants. Even in anoxic aquifers, UO2 dissolution may be controlled by surface U(VI) rather than U(IV) phases. PMID:18754482

On the use of thermal NF3 as the fluorination and oxidation agent in treatment of used nuclear fuels

NASA Astrophysics Data System (ADS)

Scheele, Randall; McNamara, Bruce; Casella, Andrew M.; Kozelisky, Anne

2012-05-01

This paper presents results of our investigation on the use of nitrogen trifluoride as a fluorination or fluorination/oxidation agent for separating valuable constituents from used nuclear fuels by exploiting the different volatilities of the constituent fission product and actinide fluorides. Our thermodynamic calculations show that nitrogen trifluoride has the potential to produce volatile fission product and actinide fluorides from oxides and metals that can form volatile fluorides. Simultaneous thermogravimetric and differential thermal analyses show that the oxides of lanthanum, cerium, rhodium, and plutonium are fluorinated but do not form volatile fluorides when treated with nitrogen trifluoride at temperatures up to 550 °C. However, depending on temperature, volatile fluorides or oxyfluorides can form from nitrogen trifluoride treatment of the oxides of niobium, molybdenum, ruthenium, tellurium, uranium, and neptunium. Thermoanalytical studies demonstrate near-quantitative separation of uranium from plutonium in a mixed 80% uranium and 20% plutonium oxide. Our studies of neat oxides and metals suggest that the reactivity of nitrogen trifluoride may be adjusted by temperature to selectively separate the major volatile fuel constituent uranium from minor volatile constituents, such as Mo, Tc, Ru and from the non-volatile fuel constituents based on differences in their reaction temperatures and kinetic behaviors. This reactivity is novel with respect to that reported for other fluorinating reagents F2, BrF5, ClF3.

Effects of natural organic matter properties on the dissolution kinetics of zinc oxide nanoparticles

USGS Publications Warehouse

Jiang, Chuanjia; Aiken, George R.; Hsu-Kim, Heileen

2015-01-01

The dissolution of zinc oxide (ZnO) nanoparticles (NPs) is a key step of controlling their environmental fate, bioavailability, and toxicity. Rates of dissolution often depend upon factors such as interactions of NPs with natural organic matter (NOM). We examined the effects of 16 different NOM isolates on the dissolution kinetics of ZnO NPs in buffered potassium chloride solution using anodic stripping voltammetry to directly measure dissolved zinc concentrations. The observed dissolution rate constants (kobs) and dissolved zinc concentrations at equilibrium increased linearly with NOM concentration (from 0 to 40 mg C L–1) for Suwannee River humic and fulvic acids and Pony Lake fulvic acid. When dissolution rates were compared for the 16 NOM isolates, kobs was positively correlated with certain properties of NOM, including specific ultraviolet absorbance (SUVA), aromatic and carbonyl carbon contents, and molecular weight. Dissolution rate constants were negatively correlated to hydrogen/carbon ratio and aliphatic carbon content. The observed correlations indicate that aromatic carbon content is a key factor in determining the rate of NOM-promoted dissolution of ZnO NPs. The findings of this study facilitate a better understanding of the fate of ZnO NPs in organic-rich aquatic environments and highlight SUVA as a facile and useful indicator of NOM interactions with metal-based nanoparticles.

Uranium Bio-accumulation and Cycling as revealed by Uranium Isotopes in Naturally Reduced Sediments from the Upper Colorado River Basin

NASA Astrophysics Data System (ADS)

Lefebvre, Pierre; Noël, Vincent; Jemison, Noah; Weaver, Karrie; Bargar, John; Maher, Kate

2016-04-01

Uranium (U) groundwater contamination following oxidized U(VI) releases from weathering of mine tailings is a major concern at numerous sites across the Upper Colorado River Basin (CRB), USA. Uranium(IV)-bearing solids accumulated within naturally reduced zones (NRZs) characterized by elevated organic carbon and iron sulfide compounds. Subsequent re-oxidation of U(IV)solid to U(VI)aqueous then controls the release to groundwater and surface water, resulting in plume persistence and raising public health concerns. Thus, understanding the extent of uranium oxidation and reduction within NRZs is critical for assessing the persistence of the groundwater contamination. In this study, we measured solid-phase uranium isotope fractionation (δ238/235U) of sedimentary core samples from four study sites (Shiprock, NM, Grand Junction, Rifle and Naturita, CO) using a multi-collector inductively coupled plasma mass spectrometer (MC-ICP-MS). We observe a strong correlation between U accumulation and the extent of isotopic fractionation, with Δ238U up to +1.8 ‰ between uranium-enriched and low concentration zones. The enrichment in the heavy isotopes within the NRZs appears to be especially important in the vadose zone, which is subject to variations in water table depth. According to previous studies, this isotopic signature is consistent with biotic reduction processes associated with metal-reducing bacteria. Positive correlations between the amount of iron sulfides and the accumulation of reduced uranium underline the importance of sulfate-reducing conditions for U(IV) retention. Furthermore, the positive fractionation associated with U reduction observed across all sites despite some variations in magnitude due to site characteristics, shows a regional trend across the Colorado River Basin. The maximum extent of 238U enrichment observed in the NRZ proximal to the water table further suggests that the redox cycling of uranium, with net release of U(VI) to the groundwater by non-fractionating oxidation, is occurring within this zone. Thus, release of uranium from the NRZs may play a critical role in the persistence of groundwater contamination at these sites.

Hydrogen sulfide (H2S) attenuates uranium-induced acute nephrotoxicity through oxidative stress and inflammatory response via Nrf2-NF-κB pathways.

PubMed

Zheng, Jifang; Zhao, Tingting; Yuan, Yan; Hu, Nan; Tang, Xiaoqing

2015-12-05

As an endogenous gaseous mediator, H2S exerts anti-oxidative, anti-inflammatory and cytoprotective effects in kidneys. This study was designed to investigate the protective effect of H2S against uranium-induced nephrotoxicity in adult SD male rats after in vivo effect of uranium on endogenous H2S formation was explored in kidneys. The levels of endogenous H2S and H2S-producing enzymes (CBS and CSE) were measured in renal homogenates from rats intoxicated by an intraperitoneally (i.p.) injection of uranyl acetate at a single dose of 2.5, 5 or 10 mg/kg. In rats injected i.p. with uranyl acetate (5 mg/kg) or NaHS (an H2S donor, 28 or 56 μmol/kg) alone or in combination, we determined biochemical parameters and histopathological alteration to assess kidney function, examined oxidative stress markers, and investigated Nrf2 and NF-κB pathways in kidney homogenates. The results suggest that uranium intoxication in rats decreased endogenous H2S generation as well as CBS and CSE protein expression. NaHS administration in uranium-intoxicated rats ameliorated the renal biochemical indices and histopathological effects, lowered MDA accumulation, and restored GSH level and anti-oxidative enzymes activities like SOD, CAT, GPx and GST. NaHS treatment in uranium-intoxicated rats activated uranium-inhibited protein expression and nuclear translocation of transcription factor Nrf2, which increased protein expression of downstream target-Nrf2 genes HO-1, NQO-1, GCLC, and TXNRD-1. NaHS administration in uranium-intoxicated rats inhibited uranium-induced nuclear translocation and phosphorylation of transcription factor κB/p65, which decreased protein expression of target-p65 inflammatory genes TNF-α, iNOS, and COX-2. Taken together, these data implicate that H2S can afford protection to rat kidneys against uranium-induced adverse effects through induction of antioxidant defense by activating Nrf2 pathway and reduction of inflammatory response by suppressing NF-κB pathway. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

Effect of Food Thickener on Dissolution and Laxative Activity of Magnesium Oxide Tablets in Mice.

PubMed

Tomita, Takashi; Goto, Hidekazu; Yoshimura, Yuya; Kato, Kazushige; Yoshida, Tadashi; Tanaka, Katsuya; Sumiya, Kenji; Kohda, Yukinao

2016-01-01

The present study examined the dissolution of magnesium oxide (MgO) from MgO tablets placed in a food thickening agent (food thickener) and its effects on laxative activity. We prepared mixtures of MgO tablets suspended in an aqueous suspension and food thickeners in order to evaluate the dissolution of MgO. The results of the dissolution tests revealed that agar-based food thickeners did not affect the MgO dissolution. In contrast, some xanthan gum-based food-thickener products show dissolution rates with certain mixtures containing disintegrated MgO tablets suspended in a food thickener that decrease over time. However, other xanthan gum-based food-thickener products show dissolution rates that decrease immediately after mixing, regardless of the time they were allowed to stand. In order to investigate the laxative activity of MgO, we orally administered a mixture of MgO suspension and food thickener to mice and observed their bowel movements. The animal experiments showed that when agar-based food thickeners were used, the laxative activity of MgO was not affected, but it decreased when xanthan gum-based food thickeners were used.

Dissolution profile of dolomite in chloric acid solution: The effect of chloric acid concentration and pulp density

NASA Astrophysics Data System (ADS)

Solihin, Indriani, Mubarok, M. Zaki

2018-05-01

Dolomite is one of carbonate minerals that contain magnesium. Magnesium is important element used in many aspects of life such as cofactor of many enzymes in human body, nutrient for plants, and raw material in automotive industry. Dolomite can be processed through low temperature process to obtain magnesium and calcium oxide that is needed in important applications such as base material for making drugs, raw material in the synthesize slow release fertilizer, materials for fire retardant, component for catalyst, etc. One of the important step of this low temperature process is dissolution of dolomite. Optimizing the dissolution process determines the % extraction of magnesium and calcium oxide from dolomite. The dissolution of dolomite from Gresik, East Java Provence Indonesia, in chloric acid solution has been conducted. Chloric acid concentration and pulp density are the variables that were observed. The dissolution of magnesium and calcium from Gresik dolomite was found to be very fast. The stable stage of dissolution can be reached for 5-10 seconds. The % extraction is mainly determined by the molar ratio of chloric acid / dolomite. At molar ratio of chloric acid / dolomite equal or above stoichiometric of dolomite dissolution, % extraction of magnesium is almost 100 %.

Spectroscopic studies of uranium species for environmental decontamination applications

NASA Astrophysics Data System (ADS)

Eng, Charlotte

After the Cold War, Department of Energy began to concentrate its efforts on cleanup of former nuclear material processing facilities, especially uranium-contaminated groundwater and soil. This research aims to study uranium association to both organic and inorganic compounds found in the contaminated environment in the hopes that the information gathered can be applied to the development and optimization of cost-effective remediation techniques. Spectroscopic and electrochemical methods will be employed to examine the behavior of uranium in given conditions to further our understanding of its impact on the environment. Uranium found in groundwater and soil bind with various ligands, especially organic ligands present in the environment due to natural sources (e.g. metabolic by-products or degradation of plants and animals) or man-made sources (e.g. chelating agents used in operating or cleanup of uranium processing facilities). We selected reasonable analogs of naturally occurring matter and studied their structure, chemical and electrochemical behavior and found that the structure of uranyl complexes depends heavily on the nature of the ligand and environmental factors such as pH. Association of uranium-organic complexes with anaerobic bacteria, Clostridium sp. was studied to establish if the bacteria can effectively bioreduce uranium while going through normal bacterial activity. It was found that the nature of the organic ligand affected the bioavailability and toxicity of the uranium on the bacteria. In addition, we have found that the type of iron corrosion products and uranyl species present on the surface of corroded steel depended on various environmental factors, which subsequently affected the removal rate of uranium by a citric acid/hydrogen peroxide/deionized water cleaning process. The method was found to remove uranium from only the topmost corrosion layers and residual uranium could be found (a) deeper in the corrosion layers where it is occluded by the steel corrosion products or (b) in areas where the dissolved uranium/iron species, the products generated by the dissolution power of citric acid, was not properly rinsed away.

Oxidation of naturally reduced uranium in aquifer sediments by dissolved oxygen and its potential significance to uranium plume persistence

NASA Astrophysics Data System (ADS)

Davis, J. A.; Smith, R. L.; Bohlke, J. K.; Jemison, N.; Xiang, H.; Repert, D. A.; Yuan, X.; Williams, K. H.

2015-12-01

The occurrence of naturally reduced zones is common in alluvial aquifers in the western U.S.A. due to the burial of woody debris in flood plains. Such reduced zones are usually heterogeneously dispersed in these aquifers and characterized by high concentrations of organic carbon, reduced mineral phases, and reduced forms of metals, including uranium(IV). The persistence of high concentrations of dissolved uranium(VI) at uranium-contaminated aquifers on the Colorado Plateau has been attributed to slow oxidation of insoluble uranium(IV) mineral phases found in association with these reducing zones, although there is little understanding of the relative importance of various potential oxidants. Four field experiments were conducted within an alluvial aquifer adjacent to the Colorado River near Rifle, CO, wherein groundwater associated with the naturally reduced zones was pumped into a gas-impermeable tank, mixed with a conservative tracer (Br-), bubbled with a gas phase composed of 97% O2 and 3% CO2, and then returned to the subsurface in the same well from which it was withdrawn. Within minutes of re-injection of the oxygenated groundwater, dissolved uranium(VI) concentrations increased from less than 1 μM to greater than 2.5 μM, demonstrating that oxygen can be an important oxidant for uranium in such field systems if supplied to the naturally reduced zones. Dissolved Fe(II) concentrations decreased to the detection limit, but increases in sulfate could not be detected due to high background concentrations. Changes in nitrogen species concentrations were variable. The results contrast with other laboratory and field results in which oxygen was introduced to systems containing high concentrations of mackinawite (FeS), rather than the more crystalline iron sulfides found in aged, naturally reduced zones. The flux of oxygen to the naturally reduced zones in the alluvial aquifers occurs mainly through interactions between groundwater and gas phases at the water table. Seasonal variations of the water table at the Rifle, CO site may play an important role in introducing oxygen into the system. Although oxygen was introduced directly to the naturally reduced zones in these experiments, delivery of oxidants to the system may also be controlled by other oxidative pathways in which oxygen plays an indirect role.

Fungal oxidative dissolution of the Mn(II)-bearing mineral rhodochrosite and the role of metabolites in manganese oxide formation.

PubMed

Tang, Yuanzhi; Zeiner, Carolyn A; Santelli, Cara M; Hansel, Colleen M

2013-04-01

Microbially mediated oxidation of Mn(II) to Mn(III/IV) oxides influences the cycling of metals and remineralization of carbon. Despite the prevalence of Mn(II)-bearing minerals in nature, little is known regarding the ability of microbes to oxidize mineral-hosted Mn(II). Here, we explored oxidation of the Mn(II)-bearing mineral rhodochrosite (MnCO3 ) and characteristics of ensuing Mn oxides by six Mn(II)-oxidizing Ascomycete fungi. All fungal species substantially enhanced rhodochrosite dissolution and surface modification. Mineral-hosted Mn(II) was oxidized resulting in formation of Mn(III/IV) oxides that were all similar to δ-MnO2 but varied in morphology and distribution in relation to cellular structures and the MnCO3 surface. For four fungi, Mn(II) oxidation occurred along hyphae, likely mediated by cell wall-associated proteins. For two species, Mn(II) oxidation occurred via reaction with fungal-derived superoxide produced at hyphal tips. This pathway ultimately resulted in structurally unique Mn oxide clusters formed at substantial distances from any cellular structure. Taken together, findings for these two fungi strongly point to a role for fungal-derived organic molecules in Mn(III) complexation and Mn oxide templation. Overall, this study illustrates the importance of fungi in rhodochrosite dissolution, extends the relevance of biogenic superoxide-based Mn(II) oxidation and highlights the potential role of mycogenic exudates in directing mineral precipitation. © 2012 Society for Applied Microbiology and Blackwell Publishing Ltd.

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.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lombardo, N.J.; Marseille, T.J.; White, M.D.

TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic inmore » form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.« less

The interaction of molecular hydrogen with α-radiolytic oxidants on a (U,Pu)O2 surface

NASA Astrophysics Data System (ADS)

Bauhn, Lovisa; Hansson, Niklas; Ekberg, Christian; Fors, Patrik; Delville, Rémi; Spahiu, Kastriot

2018-07-01

In order to assess the impact of α-radiolysis of water on the oxidative dissolution of spent fuel, an un-irradiated, annealed MOX fuel pellet with high content of Pu (∼24 wt%), and a specific α-activity of 4.96 GBq/gMOX, was leached in carbonate-containing solutions of low ionic strength. The high Pu content in the pellet stabilizes the (U,Pu)O2(s) matrix towards oxidative dissolution, whereas the α-decays emitted from the surface are expected to produce ∼3.6 × 10-7 mol H2O2/day, contributing to the oxidative dissolution of the pellet. Two sets of leaching tests were conducted under different redox conditions: Ar gas atmosphere and deuterium gas atmosphere. A relatively slow increase of the U and Pu concentrations was observed in the Ar case, with U concentrations increasing from 1·10-6 M after 1 h to ∼7 × 10-5 M after 58 days. Leaching under an atmosphere starting at 1 MPa deuterium gas was undertaken in order to evaluate any effect of dissolved hydrogen on the radiolytic dissolution of the pellet, as well as to investigate any potential recombination of the α-radiolytic products with dissolved deuterium. For the latter purpose, isotopic analysis of the D/H content was carried out on solution samples taken during the leaching. Despite the continuous production of radiolytic oxidants, the concentrations of U and Pu remained quite constant at the level of ∼3 × 10-8 M during the first 30 days, i.e. as long as the deuterium pressure remained higher than 0.8 MPa. These data rule out any oxidative dissolution of the pellet during the first month. The un-irradiated MOX fuel does not contain metallic ε-particles, hence it is mainly the interaction of radiolytic oxidants and dissolved deuterium with the surface of the mixed actinide oxide that causes the neutralization of the oxidants. This conclusion is supported by the steadily increasing levels of HDO measured in the leachate samples.

Leaching behavior of U, Mn, Sr, and Pb from different particle-size fractions of uranium mill tailings.

PubMed

Liu, Bo; Peng, Tongjiang; Sun, Hongjuan

2017-06-01

Pollution by the release of heavy metals from tailings constitutes a potential threat to the environment. To characterize the processes governing the release of Mn, Sr, Pb, and U from the uranium mill tailings, a dynamic leaching test was applied for different size of uranium mill tailings samples. Inductively coupled plasma atomic emission spectroscopy (ICP-AES) and inductively coupled plasma mass spectrometry (ICP-MS) were performed to determine the content of Mn, Sr, Pb, and U in the leachates. The release of mobile Mn, Sr, Pb, and U fraction was slow, being faster in the initial stage and then attained a near steady-state condition. The experimental results demonstrate that the release of Mn, Sr, Pb, and U from uranium mill tailings with different size fractions is controlled by a variety of mechanisms. Surface wash-off is the release mechanism for Mn. The main release mechanism of Sr and Pb is the dissolution in the initial leaching stage. For U, a mixed process of wash-off and diffusion is the controlling mechanism.

A model of early formation of uranium molecular oxides in laser-ablated plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Finko, Mikhail S.; Curreli, Davide; Weisz, David G.

Here, in this work, we present a newly constructed U xO y reaction mechanism that consists of 30 reaction channels (21 of which are reversible channels) for 11 uranium molecular species (including ions). Both the selection of reaction channels and calculation of corresponding rate coefficients is accomplished via a comprehensive literature review and application of basic reaction rate theory. The reaction mechanism is supplemented by a detailed description of oxygen plasma chemistry (19 species and 142 reaction channels) and is used to model an atmospheric laser ablated uranium plume via a 0D (global) model. Finally, the global model is usedmore » to analyze the evolution of key uranium molecular species predicted by the reaction mechanism, and the initial stage of formation of uranium oxide species.« less

A model of early formation of uranium molecular oxides in laser-ablated plasmas

DOE PAGES

Finko, Mikhail S.; Curreli, Davide; Weisz, David G.; ...

2017-10-12

Here, in this work, we present a newly constructed U xO y reaction mechanism that consists of 30 reaction channels (21 of which are reversible channels) for 11 uranium molecular species (including ions). Both the selection of reaction channels and calculation of corresponding rate coefficients is accomplished via a comprehensive literature review and application of basic reaction rate theory. The reaction mechanism is supplemented by a detailed description of oxygen plasma chemistry (19 species and 142 reaction channels) and is used to model an atmospheric laser ablated uranium plume via a 0D (global) model. Finally, the global model is usedmore » to analyze the evolution of key uranium molecular species predicted by the reaction mechanism, and the initial stage of formation of uranium oxide species.« less

Anodic Behavior of the Aluminum Current Collector in Imide-Based Electrolytes: Influence of Solvent, Operating Temperature, and Native Oxide-Layer Thickness.

PubMed

Meister, Paul; Qi, Xin; Kloepsch, Richard; Krämer, Elisabeth; Streipert, Benjamin; Winter, Martin; Placke, Tobias

2017-02-22

The inability of imide salts to form a sufficiently effective passivation layer on aluminum current collectors is one of the main obstacles that limit their broad application in electrochemical energy-storage systems. However, under certain circumstances, the use of electrolytes with imide electrolyte salts in combination with the aluminum current collector is possible. In this contribution, the stability of the aluminum current collector in electrolytes containing either lithium bis(trifluoromethanesulfonyl) imide (LiTFSI) or lithium fluorosulfonyl-(trifluoromethanesulfonyl) imide (LiFTFSI) as conductive salt was investigated by electrochemical techniques, that is, cyclic voltammetry (CV) and chronocoulometry (CC) in either room-temperature ionic liquids or in ethyl methyl sulfone. In particular, the influence of the solvent, operating temperature, and thickness of the native oxide layer of aluminum on the pit formation at the aluminum current collector surface was studied by means of scanning electron microscopy. In general, a more pronounced aluminum dissolution and pit formation was found at elevated temperatures as well as in solvents with a high dielectric constant. An enhanced thickness of the native aluminum oxide layer increases the oxidative stability versus dissolution. Furthermore, we found a different reaction rate depending on dwell time at the upper cut-off potential for aluminum dissolution in TFSI- and FTFSI-based electrolytes during the CC measurements; the use of LiFTFSI facilitated the dissolution of aluminum compared to LiTFSI. Overall, the mechanism of anodic aluminum dissolution is based on: i) the attack of the Al 2 O 3 surface by acidic species and ii) the dissolution of bare aluminum into the electrolyte, which, in turn, is influenced by the electrolyte's dielectric constant. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Uranium(iii) complexes supported by hydrobis(mercaptoimidazolyl)borates: synthesis and oxidation chemistry.

PubMed

Maria, Leonor; Santos, Isabel C; Santos, Isabel

2018-05-23

The reaction of [UI3(thf)4] with the sodium or lithium salts of hydrobis(2-mercapto-1-methylimidazolyl)borate ligands ([H(R)B(timMe)2]-) in a 1 : 2 ratio, in tetrahydrofuran, gave the U(iii) complexes [UI{κ3-H,S,S'-H(R)B(timMe)2}2(thf)2] (R = H (1), Ph (2)) in good yields. Crystals of [UI{κ3-H,S,S'-H(Ph)B(timMe)2}2(thf)2] (2) were obtained by recrystallization from a tetrahydrofuran/acetonitrile solution, and the ion-separated uranium complex [U{κ3-H,S,S'-H(Ph)B(timMe)2}2(CH3CN)3][I] (3-I) was obtained by dissolution of 2 in acetonitrile followed by recrystallization. One-electron oxidation of 2 with AgBPh4 or I2 resulted in the formation of the cationic U(iv) complexes [U{κ3-H,S,S'-H(Ph)B(timMe)2}3][X] (X = BPh4 (6-BPh4), I (6-I)), due to a ligand redistribution process. These complexes are the first examples of homoleptic poly(azolyl)borate U(iv) complexes. Treatment of complex 2 with azobenzene led to the isolation of crystals of the U(iv) compound [UI{κ3-H(Ph)B(timMe)2}2(κ2-timMe)] (7). Treatment of 2 with pyridine-N oxide (pyNO) led to the formation of the uranyl complex [UO2{κ2-S,S'-H(Ph)B(timMe)2}2] (8) and of complex 6-I, while from the reaction of [U{κ3-H(Ph)B(timMe)2}2(thf)3][BPh4] (5) with pyNO, the oxo-bridged U(iv) complex [{U{κ3-H(Ph)B(timMe)2}2(pyNO)}2(μ-O)][BPh4]2 (9) was also obtained. In the U(iii) and U(iv) complexes, the bis(azolyl)borate ligands bind to the uranium center in a κ3-H,S,S' coordination mode, while in the U(vi) complex the ligands bind to the metal in a κ2-S,S' mode. The presence of UH-B interactions in the solid-state, for the nine-coordinate complexes 1, 2, 3, 6 and 7 and for the eight-coordinate complex 9, was supported by IR spectroscopy and/or X-ray diffraction analysis.

Unraveling uranium induced oxidative stress related responses in Arabidopsis thaliana seedlings. Part I: responses in the roots.

PubMed

Vanhoudt, Nathalie; Vandenhove, Hildegarde; Horemans, Nele; Remans, Tony; Opdenakker, Kelly; Smeets, Karen; Bello, Daniel Martinez; Wannijn, Jean; Van Hees, May; Vangronsveld, Jaco; Cuypers, Ann

2011-06-01

When aiming to evaluate the environmental impact of uranium contamination, it is important to unravel the mechanisms by which plants respond to uranium stress. As oxidative stress seems an important modulator under other heavy metal stress, this study aimed to investigate oxidative stress related responses in Arabidopsis thaliana exposed to uranium concentrations ranging from 0.1 to 100 μM for 1, 3 and 7 days. Besides analyzing relevant reactive oxygen species-producing and -scavenging enzymes at protein and transcriptional level, the importance of the ascorbate-glutathione cycle under uranium stress was investigated. These results are reported separately for roots and leaves in two papers: Part I dealing with responses in the roots and Part II unraveling responses in the leaves and presenting general conclusions. Results of Part I indicate that oxidative stress related responses in the roots were only triggered following exposure to the highest uranium concentration of 100 μM. A fast oxidative burst was suggested based on the observed enhancement of lipoxygenase (LOX1) and respiratory burst oxydase homolog (RBOHD) transcript levels already after 1 day. The first line of defense was attributed to superoxide dismutase (SOD), also triggered from the first day. The enhanced SOD-capacity observed at protein level corresponded with an enhanced expression of iron SOD (FSD1) located in the plastids. For the detoxification of H(2)O(2), an early increase in catalase (CAT1) transcript levels was observed while peroxidase capacities were enhanced at the later stage of 3 days. Although the ascorbate peroxidase capacity and gene expression (APX1) increased, the ascorbate/dehydroascorbate redox balance was completely disrupted and shifted toward the oxidized form. This disrupted balance could not be inverted by the glutathione part of the cycle although the glutathione redox balance could be maintained. Copyright © 2011 Elsevier Ltd. All rights reserved.

Autotrophic denitrification supported by biotite dissolution in crystalline aquifers (1): New insights from short-term batch experiments.

PubMed

Aquilina, Luc; Roques, Clément; Boisson, Alexandre; Vergnaud-Ayraud, Virginie; Labasque, Thierry; Pauwels, Hélène; Pételet-Giraud, Emmanuelle; Pettenati, Marie; Dufresne, Alexis; Bethencourt, Lorine; Bour, Olivier

2018-04-01

We investigate denitrification mechanisms through batch experiments using crushed rock and groundwater from a granitic aquifer subject to long term pumping (Ploemeur, France). Except for sterilized experiments, extensive denitrification reaction induces NO 3 decreases ranging from 0.3 to 0.6mmol/L. Carbon concentrations, either organic or inorganic, remain relatively stable and do not document potential heterotrophic denitrification. Batch experiments show a clear effect of mineral dissolution which is documented through cation (K, Na, Ca) and Fluoride production. These productions are tightly related to denitrification progress during the experiment. Conversely, limited amounts of SO 4 , systematically lower than autotrophic denitrification coupled to sulfur oxidation stoichiometry, are produced during the experiments which indicates that sulfur oxidation is not likely even when pyrite is added to the experiments. Analysis of cation ratios, both in isolated minerals of the granite and within water of the batch, allow the mineral dissolution during the experiments to be quantified. Using cation ratios, we show that batch experiments are characterized mainly by biotite dissolution. As biotite contains 21 to 30% of Fe and 0.3 to 1.7% of F, it constitutes a potential source for these two elements. Denitrification could be attributed to the oxidation of Fe(II) contained in biotite. We computed the amount of K and F produced through biotite dissolution when entirely attributing denitrification to biotite dissolution. Computed amounts show that this process may account for the observed K and F produced. We interpret these results as the development of microbial activity which induces mineral dissolution in order to uptake Fe(II) which is used for denitrification. Although pyrite is probably available, SO 4 and cation measurements favor a large biotite dissolution reaction which could account for all the observed Fe production. Chemical composition of groundwater produced from the Ploemeur site indicates similar denitrification processes although original composition shows mainly plagioclase dissolution. Copyright © 2017 Elsevier B.V. All rights reserved.

Subsurface Uranium Fate and Transport: Integrated Experiments and Modeling of Coupled Biogeochemical Mechanisms of Nanocrystalline Uraninite Oxidation by Fe(III)-(hydr)oxides - Project Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peyton, Brent M.; Timothy, Ginn R.; Sani, Rajesh K.

2013-08-14

Subsurface bacteria including sulfate reducing bacteria (SRB) reduce soluble U(VI) to insoluble U(IV) with subsequent precipitation of UO 2. We have shown that SRB reduce U(VI) to nanometer-sized UO 2 particles (1-5 nm) which are both intra- and extracellular, with UO 2 inside the cell likely physically shielded from subsequent oxidation processes. We evaluated the UO 2 nanoparticles produced by Desulfovibrio desulfuricans G20 under growth and non-growth conditions in the presence of lactate or pyruvate and sulfate, thiosulfate, or fumarate, using ultrafiltration and HR-TEM. Results showed that a significant mass fraction of bioreduced U (35-60%) existed as a mobile phasemore » when the initial concentration of U(VI) was 160 µM. Further experiments with different initial U(VI) concentrations (25 - 900 M) in MTM with PIPES or bicarbonate buffers indicated that aggregation of uraninite depended on the initial concentrations of U(VI) and type of buffer. It is known that under some conditions SRB-mediated UO 2 nanocrystals can be reoxidized (and thus remobilized) by Fe(III)-(hydr)oxides, common constituents of soils and sediments. To elucidate the mechanism of UO 2 reoxidation by Fe(III) (hydr)oxides, we studied the impact of Fe and U chelating compounds (citrate, NTA, and EDTA) on reoxidation rates. Experiments were conducted in anaerobic batch systems in PIPES buffer. Results showed EDTA significantly accelerated UO 2 reoxidation with an initial rate of 9.5 M day-1 for ferrihydrite. In all cases, bicarbonate increased the rate and extent of UO 2 reoxidation with ferrihydrite. The highest rate of UO 2 reoxidation occurred when the chelator promoted UO 2 and Fe(III) (hydr)oxide dissolution as demonstrated with EDTA. When UO 2 dissolution did not occur, UO 2 reoxidation likely proceeded through an aqueous Fe(III) intermediate as observed for both NTA and citrate. To complement to these laboratory studies, we collected U-bearing samples from a surface seep at the Rifle field site and have measured elevated U concentrations in oxic iron-rich sediments. To translate experimental results into numerical analysis of U fate and transport, a reaction network was developed based on Sani et al. (2004) to simulate U(VI) bioreduction with concomitant UO 2 reoxidation in the presence of hematite or ferrihydrite. The reduction phase considers SRB reduction (using lactate) with the reductive dissolution of Fe(III) solids, which is set to be microbially mediated as well as abiotically driven by sulfide. Model results show the oxidation of HS– by Fe(III) directly competes with UO 2 reoxidation as Fe(III) oxidizes HS– preferentially over UO 2. The majority of Fe reduction is predicted to be abiotic, with ferrihydrite becoming fully consumed by reaction with sulfide. Predicted total dissolved carbonate concentrations from the degradation of lactate are elevated (log(pCO 2) ~ –1) and, in the hematite system, yield close to two orders-of-magnitude higher U(VI) concentrations than under initial carbonate concentrations of 3 mM. Modeling of U(VI) bioreduction with concomitant reoxidation of UO 2 in the presence of ferrihydrite was also extended to a two-dimensional field-scale groundwater flow and biogeochemically reactive transport model for the South Oyster site in eastern Virginia. This model was developed to simulate the field-scale immobilization and subsequent reoxidation of U by a biologically mediated reaction network.« less

Uranium(III)-carbon multiple bonding supported by arene δ-bonding in mixed-valence hexauranium nanometre-scale rings.

PubMed

Wooles, Ashley J; Mills, David P; Tuna, Floriana; McInnes, Eric J L; Law, Gareth T W; Fuller, Adam J; Kremer, Felipe; Ridgway, Mark; Lewis, William; Gagliardi, Laura; Vlaisavljevich, Bess; Liddle, Stephen T

2018-05-29

Despite the fact that non-aqueous uranium chemistry is over 60 years old, most polarised-covalent uranium-element multiple bonds involve formal uranium oxidation states IV, V, and VI. The paucity of uranium(III) congeners is because, in common with metal-ligand multiple bonding generally, such linkages involve strongly donating, charge-loaded ligands that bind best to electron-poor metals and inherently promote disproportionation of uranium(III). Here, we report the synthesis of hexauranium-methanediide nanometre-scale rings. Combined experimental and computational studies suggest overall the presence of formal uranium(III) and (IV) ions, though electron delocalisation in this Kramers system cannot be definitively ruled out, and the resulting polarised-covalent U = C bonds are supported by iodide and δ-bonded arene bridges. The arenes provide reservoirs that accommodate charge, thus avoiding inter-electronic repulsion that would destabilise these low oxidation state metal-ligand multiple bonds. Using arenes as electronic buffers could constitute a general synthetic strategy by which to stabilise otherwise inherently unstable metal-ligand linkages.

Thermodynamic calculations of oxygen self-diffusion in mixed-oxide nuclear fuels

DOE PAGES

Parfitt, David C.; Cooper, Michael William; Rushton, Michael J.D.; ...

2016-07-29

Mixed-oxide fuels containing uranium with thorium and/or plutonium may play an important part in future nuclear fuel cycles. There are, however, significantly less data available for these materials than conventional uranium dioxide fuel. In the present study, we employ molecular dynamics calculations to simulate the elastic properties and thermal expansivity of a range of mixed oxide compositions. These are then used to support equations of state and oxygen self-diffusion models to provide a self-consistent prediction of the behaviour of these mixed oxide fuels at arbitrary compositions.

Reactive transport modeling at uranium in situ recovery sites: uncertainties in uranium sorption on iron hydroxides

USGS Publications Warehouse

Johnson, Raymond H.; Tutu, Hlanganani; Brown, Adrian; Figueroa, Linda; Wolkersdorfer, Christian

2013-01-01

Geochemical changes that can occur down gradient from uranium in situ recovery (ISR) sites are important for various stakeholders to understand when evaluating potential effects on surrounding groundwater quality. If down gradient solid-phase material consists of sandstone with iron hydroxide coatings (no pyrite or organic carbon), sorption of uranium on iron hydroxides can control uranium mobility. Using one-dimensional reactive transport models with PHREEQC, two different geochemical databases, and various geochemical parameters, the uncertainties in uranium sorption on iron hydroxides are evaluated, because these oxidized zones create a greater risk for future uranium transport than fully reduced zones where uranium generally precipitates.

Method for the recovery of uranium values from uranium tetrafluoride

DOEpatents

Kreuzmann, Alvin B.

1983-01-01

The invention is a novel method for the recovery of uranium from dry, particulate uranium tetrafluoride. In one aspect, the invention comprises reacting particulate uranium tetrafluoride and calcium oxide in the presence of gaseous oxygen to effect formation of the corresponding alkaline earth metal uranate and alkaline earth metal fluoride. The product uranate is highly soluble in various acidic solutions wherein the product fluoride is virtually insoluble therein. The product mixture of uranate and alkaline earth metal fluoride is contacted with a suitable acid to provide a uranium-containing solution, from which the uranium is recovered. The invention can achieve quantitative recovery of uranium in highly pure form.

Method for the recovery of uranium values from uranium tetrafluoride

DOEpatents

Kreuzmann, A.B.

1982-10-27

The invention is a novel method for the recovery of uranium from dry, particulate uranium tetrafluoride. In one aspect, the invention comprises reacting particulate uranium tetrafluoride and calcium oxide in the presence of gaseous oxygen to effect formation of the corresponding alkaline earth metal uranate and alkaline earth metal fluoride. The product uranate is highly soluble in various acidic solutions whereas the product fluoride is virtually insoluble therein. The product mixture of uranate and alkaline earth metal fluoride is contacted with a suitable acid to provide a uranium-containing solution, from which the uranium is recovered. The invention can achieve quantitative recovery of uranium in highly pure form.

Nuclear Forensics: Measurements of Uranium Oxides Using Time-of-Flight Secondary Ion Mass Spectrometry (TOF-SIMS)

DTIC Science & Technology

2010-03-01

Isotope Ratio Analysis of Actinides , Fission Products, and Geolocators by High- efficiency Multi-collector Thermal Ionization Mass Spectrometry...Information, 1999. Hou, Xiaolin, and Per Roos. “ Critical Comparison of radiometric and Mass Spectrometric Methods for the Determination of...NUCLEAR FORENSICS: MEASUREMENTS OF URANIUM OXIDES USING TIME-OF-FLIGHT SECONDARY ION MASS

Study of the oxidation of uranium by external and diffuse reflectance FTIR spectroscopy using remote-sensing and evacuable cell techniques

NASA Astrophysics Data System (ADS)

Powell, G. L.; Dobbins, A.; Cristy, S. S.; Cliff, T. L.; Meyer, H. M., III; Lucania, J.; Milosevic, Milan

1994-01-01

This report describes the application of reflectance FTIR spectroscopy to the measurement of the oxidation rate of uranium by environmental gases near room temperature. It also describes very efficient evacuable cells designed for 75 degree(s) external reflectance with polarized light and for diffuse reflectance using mid-infrared FTIR spectroscopy. These cells, along with functionally similar remote sensing accessories, have been applied to the study of the oxidation of uranium metal in air, oxygen, and water vapor by precisely measuring the 575 cm-1 band of UO2 and other properties of the corrosion film such as absorbed water and reflective losses caused by film degradation related to pitting or nucleation phenomena.

Fluorination process using catalyst

DOEpatents

Hochel, Robert C.; Saturday, Kathy A.

1985-01-01

A process for converting an actinide compound selected from the group consisting of uranium oxides, plutonium oxides, uranium tetrafluorides, plutonium tetrafluorides and mixtures of said oxides and tetrafluorides, to the corresponding volatile actinide hexafluoride by fluorination with a stoichiometric excess of fluorine gas. The improvement involves conducting the fluorination of the plutonium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF.sub.3, AgF.sub.2 and NiF.sub.2, whereby the fluorination is significantly enhanced. The improvement also involves conducting the fluorination of one of the uranium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF.sub.3 and AgF.sub.2, whereby the fluorination is significantly enhanced.

Fluorination process using catalysts

DOEpatents

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

1983-08-25

A process is given for converting an actinide compound selected from the group consisting of uranium oxides, plutonium oxides, uranium tetrafluorides, plutonium tetrafluorides and mixtures of said oxides and tetrafluorides, to the corresponding volatile actinide hexafluoride by fluorination with a stoichiometric excess of fluorine gas. The improvement involves conducting the fluorination of the plutonium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF/sub 3/, AgF/sub 2/ and NiF/sub 2/, whereby the fluorination is significantly enhanced. The improvement also involves conducting the fluorination of one of the uranium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF/sub 3/ and AgF/sub 2/, whereby the fluorination is significantly enhanced.

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.

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

2016-03-07

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.

METHOD OF SEPARATING URANIUM SUSPENSIONS

DOEpatents

Wigner, E.P.; McAdams, W.A.

1958-08-26

A process is presented for separating colloidally dissed uranium oxides from the heavy water medium in upwhich they are contained. The method consists in treating such dispersions with hydrogen peroxide, thereby converting the uranium to non-colloidal UO/sub 4/, and separating the UO/sub 4/ sfter its rapid settling.

Oxide Dissolution and Oxygen Diffusion in Solid-State Recycled Ti-6Al-4V: Numerical Modeling, Verification by Nanoindentation, and Effects on Grain Growth and Recrystallization

NASA Astrophysics Data System (ADS)

Lui, E. W.; Palanisamy, S.; Dargusch, M. S.; Xia, K.

2017-12-01

The oxide dissolution and oxygen diffusion during annealing of Ti-6Al-4V solid-state recycled from machining chips by equal-channel angular pressing (ECAP) have been investigated using nanoindentation and numerical modeling. The hardness profile from nanoindentation was converted into the oxygen concentration distribution using the Fleisher and Friedel model. An iterative fitting method was then employed to revise the ideal model proposed previously, leading to correct predictions of the oxide dissolution times and oxygen concentration profiles and verifying nanoindentation as an effective method to measure local oxygen concentrations. Recrystallization started at the prior oxide boundaries where local strains were high from the severe plastic deformation incurred in the ECAP recycling process, forming a band of ultrafine grains whose growth was retarded by solute dragging thanks to high oxygen concentrations. The recrystallized fine-grained region would advance with time to eventually replace the lamellar structure formed during ECAP.

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

NASA Astrophysics Data System (ADS)

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

2016-02-01

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

Characterization of cores from an in-situ recovery mined uranium deposit in Wyoming: Implications for post-mining restoration

DOE Office of Scientific and Technical Information (OSTI.GOV)

WoldeGabriel, G.; Boukhalfa, H.; Ware, S. D.

In-situ recovery (ISR) of uranium (U) from sandstone-type roll-front deposits is a technology that involves the injection of solutions that consist of ground water fortified with oxygen and carbonate to promote the oxidative dissolution of U, which is pumped to recovery facilities located at the surface that capture the dissolved U and recycle the treated water. The ISR process alters the geochemical conditions in the subsurface creating conditions that are more favorable to the migration of uranium and other metals associated with the uranium deposit. There is a lack of clear understanding of the impact of ISR mining on themore » aquifer and host rocks of the post-mined site and the fate of residual U and other metals within the mined ore zone. We performed detailed petrographic, mineralogical, and geochemical analyses of several samples taken from about 7 m of core of the formerly the ISR-mined Smith Ranch–Highland uranium deposit in Wyoming. We show that previously mined cores contain significant residual uranium (U) present as coatings on pyrite and carbonaceous fragments. Coffinite was identified in three samples. Core samples with higher organic (> 1 wt.%) and clay (> 6–17 wt.%) contents yielded higher 234U/ 238U activity ratios (1.0–1.48) than those with lower organic and clay fractions. The ISR mining was inefficient in mobilizing U from the carbonaceous materials, which retained considerable U concentrations (374–11,534 ppm). This is in contrast with the deeper part of the ore zone, which was highly depleted in U and had very low 234U/ 238U activity ratios. This probably is due to greater contact with the lixiviant (leaching solution) during ISR mining. EXAFS analyses performed on grains with the highest U and Fe concentrations reveal that Fe is present in a reduced form as pyrite and U occurs mostly as U(IV) complexed by organic matter or as U(IV) phases of carbonate complexes. Moreover, U–O distances of ~ 2.05 Å were noted, indicating the potential formation of other poorly defined U(IV/VI) species. We also noted a small contribution from Udouble bond; length as m-dashO at 1.79 Å, which indicates that U is partially oxidized. There is no apparent U–S or U–Fe interaction in any of the U spectra analyzed. However, SEM analysis of thin sections prepared from the same core material reveals surficial U associated with pyrite which is probably a minor fraction of the total U present as thin coatings on the surface of pyrite. Our data show the presence of different structurally variable uranium forms associated with the mined cores. U associated with carbonaceous materials is probably from the original U mobilization that accumulated in the organic matter-rich areas under reducing conditions during shallow burial diagenesis. U associated with pyrite represents a small fraction of the total U and was likely deposited as a result of chemical reduction by pyrite. Our data suggest that areas rich in carbonaceous materials had limited exposure to the lixiviant solution, continue to be reducing, and still hold significant U resources. Because of their limited access to fluid flow, these areas might not contribute significantly to post-mining U release or attenuation. Areas with pyrite that are accessible to fluids seem to be more reactive and could act as reductants and facilitate U reduction and accumulation, limiting its migration.« less

Characterization of cores from an in-situ recovery mined uranium deposit in Wyoming: Implications for post-mining restoration

DOE PAGES

WoldeGabriel, G.; Boukhalfa, H.; Ware, S. D.; ...

2014-10-08

In-situ recovery (ISR) of uranium (U) from sandstone-type roll-front deposits is a technology that involves the injection of solutions that consist of ground water fortified with oxygen and carbonate to promote the oxidative dissolution of U, which is pumped to recovery facilities located at the surface that capture the dissolved U and recycle the treated water. The ISR process alters the geochemical conditions in the subsurface creating conditions that are more favorable to the migration of uranium and other metals associated with the uranium deposit. There is a lack of clear understanding of the impact of ISR mining on themore » aquifer and host rocks of the post-mined site and the fate of residual U and other metals within the mined ore zone. We performed detailed petrographic, mineralogical, and geochemical analyses of several samples taken from about 7 m of core of the formerly the ISR-mined Smith Ranch–Highland uranium deposit in Wyoming. We show that previously mined cores contain significant residual uranium (U) present as coatings on pyrite and carbonaceous fragments. Coffinite was identified in three samples. Core samples with higher organic (> 1 wt.%) and clay (> 6–17 wt.%) contents yielded higher 234U/ 238U activity ratios (1.0–1.48) than those with lower organic and clay fractions. The ISR mining was inefficient in mobilizing U from the carbonaceous materials, which retained considerable U concentrations (374–11,534 ppm). This is in contrast with the deeper part of the ore zone, which was highly depleted in U and had very low 234U/ 238U activity ratios. This probably is due to greater contact with the lixiviant (leaching solution) during ISR mining. EXAFS analyses performed on grains with the highest U and Fe concentrations reveal that Fe is present in a reduced form as pyrite and U occurs mostly as U(IV) complexed by organic matter or as U(IV) phases of carbonate complexes. Moreover, U–O distances of ~ 2.05 Å were noted, indicating the potential formation of other poorly defined U(IV/VI) species. We also noted a small contribution from Udouble bond; length as m-dashO at 1.79 Å, which indicates that U is partially oxidized. There is no apparent U–S or U–Fe interaction in any of the U spectra analyzed. However, SEM analysis of thin sections prepared from the same core material reveals surficial U associated with pyrite which is probably a minor fraction of the total U present as thin coatings on the surface of pyrite. Our data show the presence of different structurally variable uranium forms associated with the mined cores. U associated with carbonaceous materials is probably from the original U mobilization that accumulated in the organic matter-rich areas under reducing conditions during shallow burial diagenesis. U associated with pyrite represents a small fraction of the total U and was likely deposited as a result of chemical reduction by pyrite. Our data suggest that areas rich in carbonaceous materials had limited exposure to the lixiviant solution, continue to be reducing, and still hold significant U resources. Because of their limited access to fluid flow, these areas might not contribute significantly to post-mining U release or attenuation. Areas with pyrite that are accessible to fluids seem to be more reactive and could act as reductants and facilitate U reduction and accumulation, limiting its migration.« less

METHOD FOR SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS BY SOLVENT EXTRACTION

DOEpatents

Seaborg, G.T.; Blaedel, W.J.; Walling, M.T. Jr.

1960-08-23

A process is given for separating from each other uranium, plutonium, and fission products in an aqueous nitric acid solution by the so-called Redox process. The plutonium is first oxidized to the hexavalent state, e.g., with a water-soluble dichromate or sodium bismuthate, preferably together with a holding oxidant such as potassium bromate. potassium permanganate, or an excess of the oxidizing agent. The solution is then contacted with a water-immiscible organic solvent, preferably hexone. whereby uranium and plutonium are extracted while the fission products remain in the aqueous solution. The separated organic phase is then contacted with an aqueous solution of a reducing agent, with or without a holding reductant (e.g., with a ferrous salt plus hydrazine or with ferrous sulfamate), whereby plutonium is reduced to the trivalent state and back- extracted into the aqueous solution. The uranium may finally be back-extracted from the organic solvent (e.g., with a 0.1 N nitric acid).

Reductive reactivity of iron(III) oxides in the east china sea sediments: characterization by selective extraction and kinetic dissolution.

PubMed

Chen, Liang-Jin; Zhu, Mao-Xu; Yang, Gui-Peng; Huang, Xiang-Li

2013-01-01

Reactive Fe(III) oxides in gravity-core sediments collected from the East China Sea inner shelf were quantified by using three selective extractions (acidic hydroxylamine, acidic oxalate, bicarbonate-citrate buffered sodium dithionite). Also the reactivity of Fe(III) oxides in the sediments was characterized by kinetic dissolution using ascorbic acid as reductant at pH 3.0 and 7.5 in combination with the reactive continuum model. Three parameters derived from the kinetic method: m 0 (theoretical initial amount of ascorbate-reducible Fe(III) oxides), k' (rate constant) and γ (heterogeneity of reactivity), enable a quantitative characterization of Fe(III) oxide reactivity in a standardized way. Amorphous Fe(III) oxides quantified by acidic hydroxylamine extraction were quickly consumed in the uppermost layer during early diagenesis but were not depleted over the upper 100 cm depth. The total amounts of amorphous and poorly crystalline Fe(III) oxides are highly available for efficient buffering of dissolved sulfide. As indicated by the m 0, k' and γ, the surface sediments always have the maximum content, reactivity and heterogeneity of reactive Fe(III) oxides, while the three parameters simultaneously downcore decrease, much more quickly in the upper layer than at depth. Albeit being within a small range (within one order of magnitude) of the initial rates among sediments at different depths, incongruent dissolution could result in huge discrepancies of the later dissolution rates due to differentiating heterogeneity, which cannot be revealed by selective extraction. A strong linear correlation of the m 0 at pH 3.0 with the dithionite-extractable Fe(III) suggests that the m 0 may represent Fe(III) oxide assemblages spanning amorphous and crystalline Fe(III) oxides. Maximum microbially available Fe(III) predicted by the m 0 at pH 7.5 may include both amorphous and a fraction of other less reactive Fe(III) phases.

Reductive Reactivity of Iron(III) Oxides in the East China Sea Sediments: Characterization by Selective Extraction and Kinetic Dissolution

PubMed Central

Chen, Liang-Jin; Zhu, Mao-Xu; Yang, Gui-Peng; Huang, Xiang-Li

2013-01-01

Reactive Fe(III) oxides in gravity-core sediments collected from the East China Sea inner shelf were quantified by using three selective extractions (acidic hydroxylamine, acidic oxalate, bicarbonate-citrate buffered sodium dithionite). Also the reactivity of Fe(III) oxides in the sediments was characterized by kinetic dissolution using ascorbic acid as reductant at pH 3.0 and 7.5 in combination with the reactive continuum model. Three parameters derived from the kinetic method: m 0 (theoretical initial amount of ascorbate-reducible Fe(III) oxides), k′ (rate constant) and γ (heterogeneity of reactivity), enable a quantitative characterization of Fe(III) oxide reactivity in a standardized way. Amorphous Fe(III) oxides quantified by acidic hydroxylamine extraction were quickly consumed in the uppermost layer during early diagenesis but were not depleted over the upper 100 cm depth. The total amounts of amorphous and poorly crystalline Fe(III) oxides are highly available for efficient buffering of dissolved sulfide. As indicated by the m 0, k′ and γ, the surface sediments always have the maximum content, reactivity and heterogeneity of reactive Fe(III) oxides, while the three parameters simultaneously downcore decrease, much more quickly in the upper layer than at depth. Albeit being within a small range (within one order of magnitude) of the initial rates among sediments at different depths, incongruent dissolution could result in huge discrepancies of the later dissolution rates due to differentiating heterogeneity, which cannot be revealed by selective extraction. A strong linear correlation of the m 0 at pH 3.0 with the dithionite-extractable Fe(III) suggests that the m 0 may represent Fe(III) oxide assemblages spanning amorphous and crystalline Fe(III) oxides. Maximum microbially available Fe(III) predicted by the m 0 at pH 7.5 may include both amorphous and a fraction of other less reactive Fe(III) phases. PMID:24260377

Plutonium Decontamination of Uranium using CO2 Cleaning

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blau, M

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 pitsmore » 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 showed that CO{sub 2} pellet blasting (or CO{sub 2} cleaning) reduced both fixed and smearable contamination on tools. In 1997, LLNL proved that even tritium contamination could be removed from a variety of different matrices using CO{sub 2}cleaning. CO{sub 2} cleaning is a non-toxic, nonconductive, nonabrasive decontamination process whose primary cleaning mechanisms are: (1) Impact of the CO{sub 2} pellets loosens the bond between the contaminant and the substrate. (2) CO{sub 2} pellets shatter and sublimate into a gaseous state with large expansion ({approx}800 times). The expanding CO{sub 2} gas forms a layer between the contaminant and the substrate that acts as a spatula and peels off the contaminant. (3) Cooling of the contaminant assists in breaking its bond with the substrate. Thus, LLNL conducted feasibility testing to determine if CO{sub 2} pellet blasting could remove Pu contamination (e.g., uranium oxide) from uranium metal without abrading the metal matrix. This report contains a summary of events and the results of this test.« less

Purification of uranium alloys by differential solubility of oxides and production of purified fuel precursors

DOEpatents

McLean, II, William; Miller, Philip E.

1997-01-01

A method for purifying metallic alloys of uranium for use as nuclear reactor fuels in which the metal alloy is first converted to an oxide and then dissolved in nitric acid. Initial removal of metal oxide impurities not soluble in nitric acid is accomplished by filtration or other physical means. Further purification can be accomplished by carbonate leaching of uranyl ions from the partially purified solution or using traditional methods such as solvent extraction.

Purification of uranium alloys by differential solubility of oxides and production of purified fuel precursors

DOEpatents

McLean, W. II; Miller, P.E.

1997-12-16

A method is described for purifying metallic alloys of uranium for use as nuclear reactor fuels in which the metal alloy is first converted to an oxide and then dissolved in nitric acid. Initial removal of metal oxide impurities not soluble in nitric acid is accomplished by filtration or other physical means. Further purification can be accomplished by carbonate leaching of uranyl ions from the partially purified solution or using traditional methods such as solvent extraction. 3 figs.

THERMAL FISSION REACTOR COMPOSITIONS AND METHOD OF FABRICATING SAME

DOEpatents

Blainey, A.

1959-10-01

A body is presented for use in a thermal fission reactor comprising a sintered compressed mass of a substance of the group consisting of uranium, thorium, and oxides and carbides of uranium and thorium, enclosed in an envelope of a sintered, compacted, heat-conductive material of the group consisting of beryllium, zirconium, and oxides and carbides of beryllium and zirconium.

[Aluminum dissolution and changes of pH in soil solution during sorption of copper by aggregates of paddy soil].

PubMed

Xu, Hai-Bo; Zhao, Dao-Yuan; Qin, Chao; Li, Yu-Jiao; Dong, Chang-Xun

2014-01-01

Size fractions of soil aggregates in Lake Tai region were collected by the low-energy ultrasonic dispersion and the freeze-desiccation methods. The dissolution of aluminum and changes of pH in soil solution during sorption of Cu2+ and changes of the dissolution of aluminum at different pH in the solution of Cu2+ by aggregates were studied by the equilibrium sorption method. The results showed that in the process of Cu2+ sorption by aggregates, the aluminum was dissoluted and the pH decreased. The elution amount of aluminum and the decrease of pH changed with the sorption of Cu2+, both increasing with the increase of Cu2+ sorption. Under the same conditions, the dissolution of aluminum and the decrease of pH were in the order of coarse silt fraction > silt fraction > sand fraction > clay fraction, which was negatively correlated with the amount of iron oxide, aluminum and organic matter. It suggested that iron oxide, aluminum and organic matters had inhibitory and buffering effect on the aluminum dissolution and the decrease of pH during the sorption of Cu2+.

Fate of Uranium in Wetlands: Impact of Drought Followed by Re-flooding

NASA Astrophysics Data System (ADS)

Gilson, E.; Huang, S.; Koster van Groos, P. G.; Scheckel, K.; Peacock, A. D.; Kaplan, D. I.; Jaffe, P. R.

2014-12-01

Uranium contamination in groundwater can be mitigated in anoxic zones by iron-reducing bacteria that reduce soluble U(VI) to insoluble U(IV) and by uranium immobilization through complexation and sorption. Wetlands often link ground and surface-waters, making them strategic systems for potentially limiting migration of uranium contamination. Little is known about how drought periods that result in the drying of wetland soils, and consequent redox changes, affect uranium fate and transport in wetlands. In order to better understand the fate and stability of immobilized uranium in wetland soils, and how dry periods affect the uranium stability, we dosed saturated wetland mesocosms planted with Scirpus acutus with low levels of uranyl-acetate for 5 months before imposing a 9-day drying period followed by a 13-day rewetting period. Concentrations of uranium in mesocosm effluent increased after rewetting, but the cumulative amount of uranium released in the 13 days following the drying constituted less than 1% of the uranium immobilized in the soil during the 5 months prior to the drought. This low level of remobilization suggests that the uranium immobilized in these soils was not primarily bioreduced U(IV), which could have been oxidized to soluble U(VI) during the drought and released in the effluent during the subsequent flood. XANES analyses confirm that most of the uranium immobilized in the mesocosms was U(VI) sorbed to iron oxides. Compared to mesocosms that did not experience drying or rewetting, mesocosms that were sacrificed immediately after drying and after 13 days of rewetting had less uranium in soil near roots and more uranium on root surfaces. Metal-reducing bacteria only dominated the bacterial community after 13 days of rewetting and not immediately after drying, indicating that these bacteria are not responsible for this redistribution of uranium after the drying and rewetting. Results show that short periods of drought conditions in a wetland may impact uranium distribution, but these conditions may not cause large losses of immobilized uranium from the wetland.

DOE Office of Scientific and Technical Information (OSTI.GOV)

B.R. Westphal; J.C. Price; R.D. Mariani

The pyroprocessing of used nuclear fuel via electrorefining requires the continued addition of uranium trichloride to sustain operations. Uranium trichloride is utilized as an oxidant in the system to allow separation of uranium metal from the minor actinides and fission products. The inventory of uranium trichloride had diminished to a point that production was necessary to continue electrorefiner operations. Following initial experimentation, cupric chloride was chosen as a reactant with uranium metal to synthesize uranium trichloride. Despite the variability in equipment and charge characteristics, uranium trichloride was produced in sufficient quantities to maintain operations in the electrorefiner. The results andmore » conclusions from several experiments are presented along with a set of optimized operating conditions for the synthesis of uranium trichloride.« less

Status of the nuclear measurement stations for the process control of spent fuel reprocessing at AREVA NC/La Hague

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eleon, Cyrille; Passard, Christian; Hupont, Nicolas

2015-07-01

Nuclear measurements are used at AREVA NC/La Hague for the monitoring of spent fuel reprocessing. The process control is based on gamma-ray spectroscopy, passive neutron counting and active neutron interrogation, and gamma transmission measurements. The main objectives are criticality and safety, online process monitoring, and the determination of the residual fissile mass and activities in the metallic waste remained after fuel shearing and dissolution (empty hulls, grids, end pieces), which are put in radioactive waste drums before compaction. The whole monitoring system is composed of eight measurement stations which will be described in this paper. The main measurement stations no.more » 1, 3 and 7 are needed for criticality control. Before fuel element shearing for dissolution, station no. 1 allows determining the burn-up of the irradiated fuel by gamma-ray spectroscopy with HP Ge (high purity germanium) detectors. The burn-up is correlated to the {sup 137}Cs and {sup 134}Cs gamma emission rates. The fuel maximal mass which can be loaded in one bucket of the dissolver is estimated from the lowest burn-up fraction of the fuel element. Station no. 3 is dedicated to the control of the correct fuel dissolution, which is performed with a {sup 137}Cs gamma ray measurement with a HP Ge detector. Station no. 7 allows estimating the residual fissile mass in the drums filled with the metallic residues, especially in the hulls, from passive neutron counting (spontaneous fission and alpha-n reactions) and active interrogation (fission prompt neutrons induced by a pulsed neutron generator) with proportional {sup 3}He detectors. The measurement stations have been validated for the reprocessing of Uranium Oxide (UOX) fuels with a burn-up rate up to 60 GWd/t. This paper presents a brief overview of the current status of the nuclear measurement stations. (authors)« less

Surface spectroscopy studies of the oxidation behavior of uranium

NASA Astrophysics Data System (ADS)

Bloch, J.; Atzmony, U.; Dariel, M. P.; Mintz, M. H.; Shamir, N.

1982-02-01

Auger electron spectroscopy (AES) and X-ray photoelectron spectroscopy (XPS) techniques were utilized to study the oxidation behavior of clean uranium surfaces, at very low pressures of various atmospheres (UHV, H 2, O 2, and CO 2), at room temperature. Both for O 2 and CO 2, a precursor chemisorbed oxygen species has been identified at the very initial stage of the oxidation reaction. This chemisorbed oxygen transforms to the oxide form at a rate which depends on the pressure of the oxidizing atmosphere. Residual gaseous carbon compounds which are present even under UHV conditions result in the simultaneous formation of surface carbide which accompanies the initial stage of oxidation. This carbide however decomposes later as oxidation proceeds. Adventitious hydrocarbon adsorption occurs on the formed oxide layer.

Transportability Class of Americium in K Basin Sludge under Ambient and Hydrothermal Processing Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Delegard, Calvin H.; Schmitt, Bruce E.; Schmidt, Andrew J.

2006-08-01

This report establishes the technical bases for using a ''slow uptake'' instead of a ''moderate uptake'' transportability class for americium-241 (241Am) for the K Basin Sludge Treatment Project (STP) dose consequence analysis. Slow uptake classes are used for most uranium and plutonium oxides. A moderate uptake class has been used in prior STP analyses for 241Am based on the properties of separated 241Am and its associated oxide. However, when 241Am exists as an ingrown progeny (and as a small mass fraction) within plutonium mixtures, it is appropriate to assign transportability factors of the predominant plutonium mixtures (typically slow) to themore » Am241. It is argued that the transportability factor for 241Am in sludge likewise should be slow because it exists as a small mass fraction as the ingrown progeny within the uranium oxide in sludge. In this report, the transportability class assignment for 241Am is underpinned with radiochemical characterization data on K Basin sludge and with studies conducted with other irradiated fuel exposed to elevated temperatures and conditions similar to the STP. Key findings and conclusions from evaluation of the characterization data and published literature are summarized here. Plutonium and 241Am make up very small fractions of the uranium within the K Basin sludge matrix. Plutonium is present at about 1 atom per 500 atoms of uranium and 241Am at about 1 atom per 19000 of uranium. Plutonium and americium are found to remain with uranium in the solid phase in all of the {approx}60 samples taken and analyzed from various sources of K Basin sludge. The uranium-specific concentrations of plutonium and americium also remain approximately constant over a uranium concentration range (in the dry sludge solids) from 0.2 to 94 wt%, a factor of {approx}460. This invariability demonstrates that 241Am does not partition from the uranium or plutonium fraction for any characterized sludge matrix. Most of the K Basin sludge characterization data is derived spent nuclear fuel corroded within the K Basins at 10-15?C. The STP process will place water-laden sludges from the K Basin in process vessels at {approx}150-180 C. Therefore, published studies with other irradiated (uranium oxide) fuel were examined. From these studies, the affinity of plutonium and americium for uranium in irradiated UO2 also was demonstrated at hydrothermal conditions (150 C anoxic liquid water) approaching those proposed for the STP process and even for hydrothermal conditions outside of the STP operating envelope (e.g., 150 C oxic and 100 C oxic and anoxic liquid water). In summary, by demonstrating that the chemical and physical behavior of 241Am in the sludge matrix is similar to that of the predominant species (uranium and for the plutonium from which it originates), a technical basis is provided for using the slow uptake transportability factor for 241Am that is currently used for plutonium and uranium oxides. The change from moderate to slow uptake for 241Am could reduce the overall analyzed dose consequences for the STP by more than 30%.« less

The combined effect of uranium and gamma radiation on biological responses and oxidative stress induced in Arabidopsis thaliana.

PubMed

Vanhoudt, Nathalie; Vandenhove, Hildegarde; Horemans, Nele; Wannijn, Jean; Van Hees, May; Vangronsveld, Jaco; Cuypers, Ann

2010-11-01

Uranium never occurs as a single pollutant in the environment, but always in combination with other stressors such as ionizing radiation. As effects induced by multiple contaminants can differ markedly from the effects induced by the individual stressors, this multiple pollution context should not be neglected. In this study, effects on growth, nutrient uptake and oxidative stress induced by the single stressors uranium and gamma radiation are compared with the effects induced by the combination of both stressors. By doing this, we aim to better understand the effects induced by the combined stressors but also to get more insight in stressor-specific response mechanisms. Eighteen-day-old Arabidopsis thaliana seedlings were exposed for 3 days to 10 muM uranium and 3.5 Gy gamma radiation. Gamma radiation interfered with uranium uptake, resulting in decreased uranium concentrations in the roots, but with higher transport to the leaves. This resulted in a better root growth but increased leaf lipid peroxidation. For the other endpoints studied, effects under combined exposure were mostly determined by uranium presence and only limited influenced by gamma presence. Furthermore, an important role is suggested for CAT1/2/3 gene expression under uranium and mixed stressor conditions in the leaves.

Alternative Anodes for the Electrolytic Reduction of Uranium Dioxide

NASA Astrophysics Data System (ADS)

Merwin, Augustus

Reprocessing of spent nuclear fuel is an essential step in closing the nuclear fuel cycle. In order to consume current stockpiles, ceramic uranium dioxide spent nuclear fuel will be subjected to an electrolytic reduction process. The current reduction process employs a platinum anode and a stainless steel alloy 316 cathode in a molten salt bath consisting of LiCl-2wt% Li 2O and occurs at 700°C. A major shortcoming of the existing process is the degradation of the platinum anode under the severely oxidizing conditions encountered during electrolytic reduction. This work investigates alternative anode materials for the electrolytic reduction of uranium oxide. The high temperature and extreme oxidizing conditions encountered in these studies necessitated a unique set of design constraints on the system. Thus, a customized experimental apparatus was designed and constructed. The electrochemical experiments were performed in an electrochemical reactor placed inside a furnace. This entire setup was housed inside a glove box, in order to maintain an inert atmosphere. This study investigates alternative anode materials through accelerated corrosion testing. Surface morphology was studied using scanning electron microscopy. Surface chemistry was characterized using energy dispersive spectroscopy and Raman spectroscopy. Electrochemical behavior of candidate materials was evaluated using potentiodynamic polarization characteristics. After narrowing the number of candidate electrode materials, ferrous stainless steel alloy 316, nickel based Inconel 718 and elemental tungsten were chosen for further investigation. Of these materials only tungsten was found to be sufficiently stable at the anodic potential required for electrolysis of uranium dioxide in molten salt. The tungsten anode and stainless steel alloy 316 cathode electrode system was studied at the required reduction potential for UO2 with varying lithium oxide concentrations. Electrochemical impedance spectroscopy showed mixed (kinetic and diffusion) control and an overall low impedance due to extreme corrosion. It was observed that tungsten is sufficiently stable in LiCl - 2wt% Li 2O at 700°C at the required anodic potential for the reduction of uranium oxide. This study identifies tungsten to be a superior anode material to platinum for the electrolytic reduction of uranium oxide, both in terms of superior corrosion behavior and reduced cost, and thus recommends that tungsten be further investigated as an alternative anode for the electrolytic reduction of uranium dioxide.

PROCESS FOR PRODUCING URANIUM HEXAFLUORIDE

DOEpatents

Fowler, R.D.

1957-10-22

A process for the production of uranium hexafluoride from the oxides of uranium is reported. In accordance with the method the higher oxides of uranium may be reduced to uranium dioxide (UO/sub 2/), the latter converted into uranium tetrafluoride by reaction with hydrogen fluoride, and the UF/sub 4/ convented to UF/sub 6/ by reaction with a fluorinating agent. The UO/sub 3/ or U/sub 3/O/sub 8/ is placed in a reaction chamber in a copper boat or tray enclosed in a copper oven, and heated to 500 to 650 deg C while hydrogen gas is passed through the oven. The oven is then swept clean of hydrogen and the water vapor formed by means of nitrogen and then while continuing to maintain the temperature between 400 and 600 deg C, anhydrous hydrogen fluoride is passed through. After completion of the conversion to uranium tetrafluoride, the temperature of the reaction chamber is lowered to ahout 400 deg C, and elemental fluorine is used as the fluorinating agent for the conversion of UF/sub 4/ into UF/sub 6/. The fluorine gas is passed into the chamber, and the UF/sub 6/ formed passes out and is delivered to a condenser.

Preparation, characterization and dissolution of passive oxide film on the 400 series stainless steel surfaces

NASA Astrophysics Data System (ADS)

Sathyaseelan, V. S.; Rufus, A. L.; Chandramohan, P.; Subramanian, H.; Velmurugan, S.

2015-12-01

Full system decontamination of Primary Heat Transport (PHT) system of Pressurised Heavy Water Reactors (PHWRs) resulted in low decontamination factors (DF) on stainless steel (SS) surfaces. Hence, studies were carried out with 403 SS and 410 SS that are the material of construction of "End-Fitting body" and "End-Fitting Liner tubes". Three formulations were evaluated for the dissolution of passive films formed over these alloys viz., i) Two-step process consisting of oxidation and reduction reactions, ii) Dilute Chemical Decontamination (DCD) and iii) High Temperature Process. The two-step and high temperature processes could dissolve the oxide completely while the DCD process could remove only 60%. Various techniques like XRD, Raman spectroscopy and SEM-EDX were used for assessing the dissolution process. The two-step process is time consuming, laborious while the high temperature process is less time consuming and is recommended for SS decontamination.

Electron transfer at the cell-uranium interface in Geobacter spp.

PubMed

Reguera, Gemma

2012-12-01

The in situ stimulation of Fe(III) oxide reduction in the subsurface stimulates the growth of Geobacter spp. and the precipitation of U(VI) from groundwater. As with Fe(III) oxide reduction, the reduction of uranium by Geobacter spp. requires the expression of their conductive pili. The pili bind the soluble uranium and catalyse its extracellular reductive precipitation along the pili filaments as a mononuclear U(IV) complexed by carbon-containing ligands. Although most of the uranium is immobilized by the pili, some uranium deposits are also observed in discreet regions of the outer membrane, consistent with the participation of redox-active foci, presumably c-type cytochromes, in the extracellular reduction of uranium. It is unlikely that cytochromes released from the outer membrane could associate with the pili and contribute to the catalysis, because scanning tunnelling microscopy spectroscopy did not reveal any haem-specific electronic features in the pili, but, rather, showed topographic and electronic features intrinsic to the pilus shaft. Pili not only enhance the rate and extent of uranium reduction per cell, but also prevent the uranium from traversing the outer membrane and mineralizing the cell envelope. As a result, pili expression preserves the essential respiratory activities of the cell envelope and the cell's viability. Hence the results support a model in which the conductive pili function as the primary mechanism for the reduction of uranium and cellular protection in Geobacter spp.

Physico-Chemical Heterogeneity of Organic-Rich Sediments in the Rifle Aquifer, CO: Impact on Uranium Biogeochemistry.

PubMed

Janot, Noémie; Lezama Pacheco, Juan S; Pham, Don Q; O'Brien, Timothy M; Hausladen, Debra; Noël, Vincent; Lallier, Florent; Maher, Kate; Fendorf, Scott; Williams, Kenneth H; Long, Philip E; Bargar, John R

2016-01-05

The Rifle alluvial aquifer along the Colorado River in west central Colorado contains fine-grained, diffusion-limited sediment lenses that are substantially enriched in organic carbon and sulfides, as well as uranium, from previous milling operations. These naturally reduced zones (NRZs) coincide spatially with a persistent uranium groundwater plume. There is concern that uranium release from NRZs is contributing to plume persistence or will do so in the future. To better define the physical extent, heterogeneity and biogeochemistry of these NRZs, we investigated sediment cores from five neighboring wells. The main NRZ body exhibited uranium concentrations up to 100 mg/kg U as U(IV) and contains ca. 286 g of U in total. Uranium accumulated only in areas where organic carbon and reduced sulfur (as iron sulfides) were present, emphasizing the importance of sulfate-reducing conditions to uranium retention and the essential role of organic matter. NRZs further exhibited centimeter-scale variations in both redox status and particle size. Mackinawite, greigite, pyrite and sulfate coexist in the sediments, indicating that dynamic redox cycling occurs within NRZs and that their internal portions can be seasonally oxidized. We show that oxidative U(VI) release to the aquifer has the potential to sustain a groundwater contaminant plume for centuries. NRZs, known to exist in other uranium-contaminated aquifers, may be regionally important to uranium persistence.

X-Ray Photoelectron Study of the Oxides Formed on Nickel Metal and Nickel-Chromium 20% Alloy Surfaces Under Reducing and Oxidizing Potentials in Basic, Neutral and Acidic Solutions

NASA Astrophysics Data System (ADS)

Payne, Brad P.; Keech, Peter G.; McIntyre, N. Stewart

The corrosion products produced on polycrystalline Ni metal and Ni-Cr (20%) (NiCr) alloy surfaces exposed to aqueous environments chosen to emulate possible solution conditions in the steam generator (SG) tubing of pressurized water reactors (PWR) were studied using XPS. Additional measurements modelling the distribution of oxidized Ni and Cr species on select alloy specimens were carried out using ToF SIMS. Exposure of Ni metal and NiCr alloy samples to mildly oxidizing potentials in basic solutions resulted in the preferential growth of a β-Ni(OH)2 phase; driven by the dissolution of metallic Ni at both 25°C and 150°C. The presence of β-Ni(OH)2, Cr(OH)3 and small amounts of a Cr6+-containing oxide on NiCr specimens oxidized under mildly oxidizing conditions at 150°C in neutral solutions suggested that the dissolution of both metallic Ni and Cr followed by the back deposition of the corresponding corrosion products was responsible for oxide growth under these conditions. In acidic media oxide nucleation at 150°C under mildly oxidizing potentials was determined to occur via the dissolution of both Ni and Cr species on NiCr specimens as well. The increased stability of Ni2+ in acidic solution led to a limited precipitation of β-Ni(OH)2 resulting in the formation of very thin oxides containing higher levels of Cr(OH)3. Reactions on metallic Ni and NiCr surfaces under highly oxidizing potentials resulted in an increase in the NiO content of these films compared to similar exposures carried out at milder oxidation conditions attributed to accelerated dehydration of the β-Ni(OH)2 phase. In addition, an increase in the Cr(OH)3 contribution on the alloy surface oxidized at a more oxidative potential suggested a more rapid dissolution of Cr under these conditions; overall, uneven films were formed from these conditions. The composition of the corrosion product formed after an exposure to a highly oxidizing potential was found to be unchanged following a subsequent reaction of equivalent length at a much lower oxidizing potential in basic solution.

Webinar on the Removal of Uranium from Drinking Water by Small System Treatment Technology

EPA Science Inventory

Abstract: Radionuclides, such as uranium (U), occur naturally as trace elements in rocks and soils and thus can be found in dissolved forms in ground waters. Uranium has four oxidation states (+3, +4, +5, and +6) and is a very reactive element forming a variety of stable complexe...

Nanoscale Zirconium-(oxyhydr)oxide in Contaminated Sediments From Hanford, WA - A New Host for Uranium

NASA Astrophysics Data System (ADS)

Stubbs, J. E.; Elbert, D. C.; Veblen, L. A.; Zachara, J. M.; Davis, J. A.; Veblen, D. R.

2008-12-01

Zirconium-, uranium-, and copper-bearing wastes have leached from former disposal ponds into vadose zone sediments in the 300 Area at the Department of Energy's Hanford Site. Zirconium is enriched in the shallow portion of the vadose zone, and we have discovered an amorphous Zr-(oxyhydr)oxide that contains 16% of the total uranium budget (84.24 ppm) in one of the shallow samples. We have characterized the oxide using electron microprobe analysis (EMPA), a focused ion beam (FIB) instrument, and transmission electron microscopy (TEM). It occurs in fine-grained coatings found on lithic and mineral fragments in these sediments. The oxide is intimately intergrown with the phyllosilicates and other minerals of the coatings, and in places can be seen coating individual, nano-sized phyllosilicate mineral grains. Electron energy-loss spectroscopy (EELS) shows that the Zr-(oxyhydr)oxide has a P:Zr atomic ratio around 0.2, suggesting it is either intergrown with minor amounts of a Zr-phosphate or has adsorbed a significant amount of phosphate. This material has adsorbed or incorporated a substantial amount of uranium. Thus, understanding its nature is critical to predicting the long-term fate of U in the Hanford vadose zone. While the low-temperature uptake of U by Zr-(oxhydr)oxides and phosphates has been studied for several decades in laboratory settings, to our knowledge ours is the first report of such uptake in the field.

Solubility of nano-zinc oxide in environmentally and biologically important matrices

PubMed Central

Reed, Robert B.; Ladner, David A.; Higgins, Christopher P.; Westerhoff, Paul; Ranville, James F.

2011-01-01

Increasing manufacture and use of engineered nanoparticles (NPs) is leading to a greater probability for release of ENPs into the environment and exposure to organisms. In particular, zinc oxide (ZnO) is toxic, although it is unclear whether this toxicity is due to the zinc oxide nanoparticles (ZnO), dissolution to Zn2+, or some combination thereof. The goal of this study was to determine the relative solubilites of both commercially available and in-house synthesized ZnO in matrices used for environmental fate and transport or biological toxicity studies. Dissolution of ZnO was observed in nanopure water (7.18– 7.40 mg/L dissolved Zn, as measured by filtration) and Roswell Park Memorial Institute medium (RPMI-1640) (~5 mg/L), but much more dissolution was observed in Dulbecco’s Modified Eagle’s Medium (DMEM), where the dissolved Zn concentration exceeded 34 mg/L. Moderately hard water exhibited low zinc solubility, likely due to precipitation of a zinc carbonate solid phase. Precipitation of a zinc-containing solid phase in RPMI also appeared to limit zinc solubility. Equilibrium conditions with respect to ZnO solubility were not apparent in these matrices, even after more than 1,000 h of dissolution. These results suggest that solution chemistry exerts a strong influence on ZnO dissolution and can result in limits on zinc solubility due to precipitation of less soluble solid phases. PMID:21994124

Uranium in foraminiferal calcite as a recorder of seawater uranium concentrations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Russell, A.D.; Emerson, S.; Nelson, B.K.

The authors present results of an investigation of uranium/calcium ratios in cleaned foraminiferal calcite as a recorder of seawater uranium concentrations. For accurate reconstruction of past seawater uranium content, shell calcite must incorporate uranium in proportion to seawater concentration and must preserve its original uranium composition over time. Laboratory culture experiments with live benthic (Amphistegina lobifera) and live planktonic (Globigerinell calida) foraminifera show that the U/Ca ratio of cleaned calcite tests is proportional to the concentration of uranium in solution. After correcting results for the presence of initial calcite, the apparent distribution coefficient D = (U/Ca[sub calcite])/(U/Ca)[sub solution] = 10.6more » [+-] 0.3 (x10[sup [minus]3]) for A. lobifera and D = 7.9 [+-] 0.1 (x10[sup [minus]3]) for G. calida. U/Ca ratios in planktonic foraminifera from core tops collected above 3900 m in the equatorial Atlantic and above 2100 m in the Pacific Ocean show no significant difference among the species analyzed. D estimated form core top samples ranges from 7.6 [+-] 0.4 (x10[sup [minus]3]) for O. universa to 8.4 [+-] 0.5 (x10[sup [minus]3]) for G. ruber. In benthic species C. wuellerstorfi, D = 7.0 [+-] 0.8 (x10[sup [minus]3]). U/Ca and Mg/Ca in G. tumida and G. sacculifer from core tops taken near and below the regional lysocline decrease with water depth. Smaller decreases in U/Ca and Mg/Ca with depth were observed in C. wuellerstorfi. In the planktonic species, the authors believe that U/CA and Mg/Ca are lower in the more dissolution-resistant fraction of calcite, leading to lower U/Ca in more highly dissolved samples.« less

Soil Sample Dissolution Development by Ultrawave Digester, Followed by Isotopic Separation and Analysis

DTIC Science & Technology

2017-01-09

uranium, americium, and thorium were analyzed, along with other transition and rare earth metals, utilizing inductively coupled plasma- mass spectrometry...inductively coupled plasma- mass spectrometry and/or alpha spectrometry, following digestion. For validation of the microwave protocol, radioactive... actinide elements. HF is a hazardous acid to work with and it is highly toxic. In this evaluation and validation, the actinides are of particular

Dissolution and characterization of HEV NiMH batteries.

PubMed

Larsson, Kristian; Ekberg, Christian; Ødegaard-Jensen, Arvid

2013-03-01

Metal recovery is an essential part of the recycling of hybrid electric vehicle battery waste and the first step in a hydrometallurgical treatment is dissolution of the solid material. The properties of separated battery electrode materials were investigated. Focus was put on both the solid waste and then the dissolution behaviour. The cathode contains metallic nickel that remains undissolved when utilizing non-oxidizing conditions such as hydrochloric or sulphuric acid in combination with a low oxygen atmosphere. In these conditions the cathode active electrode material is fully dissolved. Not dissolving the nickel metal saves up to 37% of the acid consumption for the cathode electrode material. In the commonly used case of oxidizing conditions the nickel metal dissolves and a cobalt-rich phase remains undissolved from the cathode active material. For the anode material a complete and rapid dissolution can be achieved at mild conditions with hydrochloric, nitric or sulphuric acid. Optimal parameters for all cases of dissolution was pH 1 with a reaction time of approximately ≥ 20,000 s. Copyright © 2012 Elsevier Ltd. All rights reserved.

Superconcentrated electrolytes for a high-voltage lithium-ion battery

PubMed Central

Wang, Jianhui; Yamada, Yuki; Sodeyama, Keitaro; Chiang, Ching Hua; Tateyama, Yoshitaka; Yamada, Atsuo

2016-01-01

Finding a viable electrolyte for next-generation 5 V-class lithium-ion batteries is of primary importance. A long-standing obstacle has been metal-ion dissolution at high voltages. The LiPF6 salt in conventional electrolytes is chemically unstable, which accelerates transition metal dissolution of the electrode material, yet beneficially suppresses oxidative dissolution of the aluminium current collector; replacing LiPF6 with more stable lithium salts may diminish transition metal dissolution but unfortunately encounters severe aluminium oxidation. Here we report an electrolyte design that can solve this dilemma. By mixing a stable lithium salt LiN(SO2F)2 with dimethyl carbonate solvent at extremely high concentrations, we obtain an unusual liquid showing a three-dimensional network of anions and solvent molecules that coordinate strongly to Li+ ions. This simple formulation of superconcentrated LiN(SO2F)2/dimethyl carbonate electrolyte inhibits the dissolution of both aluminium and transition metal at around 5 V, and realizes a high-voltage LiNi0.5Mn1.5O4/graphite battery that exhibits excellent cycling durability, high rate capability and enhanced safety. PMID:27354162

Spent nuclear fuel recycling with plasma reduction and etching

DOEpatents

Kim, Yong Ho

2012-06-05

A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF₆ and PuF₄. During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF₆ (gas) that is formed to decompose back to PuF₄ (solid), and b) to maintain stability of the UF₆. Uranium (in the form of gaseous UF₆) is easily extracted and separated from the plutonium (in the form of solid PuF₄). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF₆ (as a final product). Use of plasmas provide faster reaction rates, greater control over the individual electron and ion temperatures, and allow the use of CF₄ or NF₃ as the fluorine sources instead of F₂ or HF.

Thermodynamic and experimental study of UC powders ignition

NASA Astrophysics Data System (ADS)

Le Guyadec, F.; Rado, C.; Joffre, S.; Coullomb, S.; Chatillon, C.; Blanquet, E.

2009-09-01

Mixed plutonium and uranium carbide (UPuC) is considered as a possible fuel material for future nuclear reactors. However, UPuC is pyrophoric and fine powders of UPuC are subject to temperature increase due to oxidation with air and possible ignition during conditioning and handling. In a first approach and to allow easier experimental conditions, this study was undertaken on uranium monocarbide (UC) with the aim to determine safe handling conditions for the production and reprocessing of uranium carbide fuels. The reactivity of uranium monocarbide in oxidizing atmosphere was studied in order to analyze the ignition process. Experimental thermogravimetric analysis (TGA) and differential thermal analysis (DTA) revealed that UC powder obtained by arc melting and milling is highly reactive in air at about 200 °C. The phases formed at the various observed stages of the oxidation process were analyzed by X-ray diffraction. At the same time, ignition was analyzed thermodynamically along isothermal sections of the U-C-O ternary diagram and the pressure of the gas produced by the UC + O 2 reaction was calculated. Two possible oxidation schemes were identified on the U-C-O phase diagram and assumptions are proposed concerning the overall oxidation and ignition paths. It is particularly important to understand the mechanisms involved since temperatures as high as 2500 °C could be reached, leading to CO(g) production and possibly to a blast effect.

Uranium oxidation kinetics monitored by in-situ X-ray diffraction

NASA Astrophysics Data System (ADS)

Zalkind, S.; Rafailov, G.; Halevy, I.; Livneh, T.; Rubin, A.; Maimon, H.; Schweke, D.

2017-03-01

The oxidation kinetics of U-0.1 wt%Cr at oxygen pressures of 150 Torr and the temperature range of 90-150 °C was studied by means of in-situ X-ray diffraction (XRD). A "breakaway" in the oxidation kinetics is found at ∼0.25 μm, turning from a parabolic to a linear rate law. At the initial stage of oxidation the growth plane of UO2(111) is the prominent one. As the oxide thickens, the growth rate of UO2(220) plane increases and both planes grow concurrently. The activation energies obtained for the oxide growth are Qparabolic = 17.5 kcal/mol and Qlinear = 19 kcal/mol. Enhanced oxidation around uranium carbide (UC) inclusions is clearly observed by scanning electron microscopy (SEM).

Inherently safe in situ uranium recovery

DOEpatents

Krumhansl, James L; Brady, Patrick V

2014-04-29

An in situ recovery of uranium operation involves circulating reactive fluids through an underground uranium deposit. These fluids contain chemicals that dissolve the uranium ore. Uranium is recovered from the fluids after they are pumped back to the surface. Chemicals used to accomplish this include complexing agents that are organic, readily degradable, and/or have a predictable lifetime in an aquifer. Efficiency is increased through development of organic agents targeted to complexing tetravalent uranium rather than hexavalent uranium. The operation provides for in situ immobilization of some oxy-anion pollutants under oxidizing conditions as well as reducing conditions. The operation also artificially reestablishes reducing conditions on the aquifer after uranium recovery is completed. With the ability to have the impacted aquifer reliably remediated, the uranium recovery operation can be considered inherently safe.

The effect of bicarbonate on the microbial dissolution of autunite mineral in the presence of gram-positive bacteria.

PubMed

Sepulveda-Medina, Paola M; Katsenovich, Yelena P; Wellman, Dawn M; Lagos, Leonel E

2015-06-01

Bacteria are key players in the processes that govern fate and transport of contaminants. The uranium release from Na and Ca-autunite by Arthrobacter oxydans strain G968 was evaluated in the presence of bicarbonate ions. This bacterium was previously isolated from Hanford Site soil and in earlier prescreening tests demonstrated low tolerance to U(VI) toxicity compared to other A. oxydans isolates. Experiments were conducted using glass serum bottles as mixed bioreactors and sterile 6-well cell culture plates with inserts separating bacteria cells from mineral solids. Reactors containing phosphorus-limiting media were amended with bicarbonate ranging between 0 and 10 mM and meta-autunite solids to provide a U(VI) concentration of 4.4 mmol/L. Results showed that in the presence of bicarbonate, A. oxydans G968 was able to enhance the release of U(VI) from Na and Ca autunite at the same capacity as other A. oxydans isolates with relatively high tolerance to U(VI). The effect of bacterial strains on autunite dissolution decreases as the concentration of bicarbonate increases. The results illustrate that direct interaction between the bacteria and the mineral is not necessary to result in U(VI) biorelease from autunite. The formation of secondary calcium-phosphate mineral phases on the surface of the mineral during the dissolution can ultimately reduce the natural autunite mineral contact area, which bacterial cells can access. This thereby reduces the concentration of uranium released into the solution. This study provides a better understanding of the interactions between meta-autunite and microbes in conditions mimicking arid and semiarid subsurface environments of western U.S. Copyright © 2015 Elsevier Ltd. All rights reserved.

The influence of citrate and oxalate on 99TcVII, Cs, NpV and UVI sorption to a Savannah River Site soil.

PubMed

Montgomery, D; Barber, K; Edayilam, N; Oqujiuba, K; Young, S; Biotidara, T; Gathers, A; Danjaji, M; Tharayil, N; Martinez, N; Powell, B

2017-06-01

Batch sorption experiments were conducted with 0.5-50 ppb 99 Tc, 133 Cs, 237 Np and U in the presence and absence of citrate and/or oxalate in a 25 g/L Savannah River Site (SRS) soil suspension. Citrate and oxalate were the ligands of choice due to their relevancy to plant exudates, the nuclides were selected for their wide range of biogeochemical behavior, and the soil from SRS was selected as a model Department of Energy (DOE) site soil. Batch samples were continually mixed on a rotary shaker and maintained at a pH of approximately 5. Analysis via ICP-MS indicated that sorption of 237 Np increased with ligand concentration compared to baseline studies, as did sorption of 99 Tc although to a lesser extent. The increased sorption of 237 Np is proposed to be due to a combination of factors that are dependent on the ligand(s) present in the specific system including, ligand dissolution of the soil by citrate and formation of tertiary soil-oxalate-Np complexes. The increased 99 Tc sorption is attributed to the dissolution of the soil by the ligands, leading to an increase in the number of available sorption sites for 99 Tc. Uranium sorption decreased and dissolution of native uranium was also observed with increasing ligand concentration, thought to be a result of the formation of strong U-ligand complexes remaining in the aqueous phase. The majority of these effects were observed at the highest ligand concentrations of 50 mg C /L. No notable changes were observed for the 133 Cs system which is ascribed to the minimal interaction of Cs + with these organic ligands. Copyright © 2017 Elsevier Ltd. All rights reserved.

The Effect of Bicarbonate on the Microbial Dissolution of Autunite Mineral in the Presence of Gram-Positive Bacteria

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sepulveda-Medina, Paola; Katsenovich, Yelena; Wellman, Dawn M.

Bacteria are key players in the processes that govern fate and transport of contaminants. The uranium release from Na and Ca-autunite by Arthrobacter oxydans strain G968 was evaluated in the presence of bicarbonate ions. This bacterium was previously isolated from Hanford Site soil and in earlier prescreening tests demonstrated low tolerance to U(VI) toxicity compared to other A.oxydans isolates. Experiments were conducted using glass serum bottles as mixed bioreactors and sterile 6-well cell culture plates with inserts separating bacteria cells from mineral solids. Reactors containing phosphorus-limiting media were amended with bicarbonate ranging between 0-10 mM and metaautunite solids to providemore » a U(VI) concentration of 4.4 mmol/L. Results showed that in the presence of bicarbonate, A.oxydans G968 was able to enhance the release of U(VI) from Na and Ca autunite at the same capacity as other A.oxydans isolates with relatively high tolerance to U(VI). The effect of bacterial strains on autunite dissolution decreases as the concentration of bicarbonate increases. The results illustrate that direct interaction between the bacteria and the mineral is not necessary to result in U (VI) biorelease from autunite. The formation of secondary calcium-phosphate mineral phases on the surface of the mineral during the dissolution can ultimately reduce the natural autunite mineral contact area, which bacterial cells can access. This thereby reduces the concentration of uranium released into the solution. This study provides a better understanding of the interactions between meta-autunite and microbes in conditions mimicking arid and semiarid subsurface environments of western U.S.« less

Column Testing and 1D Reactive Transport Modeling to Evaluate Uranium Plume Persistence Processes

NASA Astrophysics Data System (ADS)

Johnson, R. H.; Morrison, S.; Morris, S.; Tigar, A.; Dam, W. L.; Dayvault, J.

2015-12-01

At many U.S. Department of Energy Office of Legacy Management sites, 100 year natural flushing was selected as a remedial option for groundwater uranium plumes. However, current data indicate that natural flushing is not occurring as quickly as expected and solid-phase and aqueous uranium concentrations are persistent. At the Grand Junction, Colorado office site, column testing was completed on core collected below an area where uranium mill tailings have been removed. The total uranium concentration in this core was 13.2 mg/kg and the column was flushed with laboratory-created water with no uranium and chemistry similar to the nearby Gunnison River. The core was flushed for a total of 91 pore volumes producing a maximum effluent uranium concentration of 6,110 μg/L at 2.1 pore volumes and a minimum uranium concentration of 36.2 μg/L at the final pore volume. These results indicate complex geochemical reactions at small pore volumes and a long tailing affect at greater pore volumes. Stop flow data indicate the occurrence of non-equilibrium processes that create uranium concentration rebound. These data confirm the potential for plume persistence, which is occurring at the field scale. 1D reactive transport modeling was completed using PHREEQC (geochemical model) and calibrated to the column test data manually and using PEST (inverse modeling calibration routine). Processes of sorption, dual porosity with diffusion, mineral dissolution, dispersion, and cation exchange were evaluated separately and in combination. The calibration results indicate that sorption and dual porosity are major processes in explaining the column test data. These processes are also supported by fission track photographs that show solid-phase uranium residing in less mobile pore spaces. These procedures provide valuable information on plume persistence and secondary source processes that may be used to better inform and evaluate remedial strategies, including natural flushing.

The effect of iron content and dissolved O2 on dissolution rates of clinopyroxene at pH 5.8 and 25°C: Preliminary results

USGS Publications Warehouse

Hoch, A.R.; Reddy, M.M.; Drever, J.I.

1996-01-01

Dissolution experiments using augite (Mg0.87Ca0.85Fe0.19Na0.09Al0.03Si2O6) and diopside (Mg0.91Ca0.93Fe0.07Na0.03Al0.03Si2O6) were conducted in flow-through reactors (5-ml/h flow rate). A pH of 5.8 was maintained by bubbling pure CO2 through a solution of 0.01 M KHCO3 at 25°C. Two experiments were run for each pyroxene type. In one experiment dissolved O2 concentration in reactors was 0.6 (±0.1) ppm and in the second dissolved O2 was 1.5 (±0.1) ppm. After 60 days, augite dissolution rates (based on Si release) were approximately three times greater in the 1.5 ppm. dissolved O2 experiments than in the sealed experiments. In contrast, diopside dissolution rates were independent of dissolved O2 concentrations. Preliminary results from the augite experiments suggest that dissolution rate is directly related to oxidation of iron. This effect was not observed in experiments performed on iron-poor diopside. Additionally, dissolution rates of diopside were much slower than those of augite, again suggesting a relationship between Fe content, Fe oxidation and dissolution rates.

Effect of electrolyte temperature on the formation of self-organized anodic niobium oxide microcones in hot phosphate-glycerol electrolyte

NASA Astrophysics Data System (ADS)

Yang, S.; Aoki, Y.; Habazaki, H.

2011-07-01

Nanoporous niobium oxide films with microcone-type surface morphology were formed by anodizing at 10 V in glycerol electrolyte containing 0.6 mol dm -3 K 2HPO 4 and 0.2 mol dm -3 K 3PO 4 in a temperature range of 428-453 K. The microcones appeared after prolonged anodizing, but the required time was largely reduced by increasing electrolyte temperature. The anodic oxide was initially amorphous at all temperatures, but crystalline oxide nucleated during anodizing. The anodic oxide microcones, which were crystalline, appeared on surface as a consequence of preferential chemical dissolution of initially formed amorphous oxide. The chemical dissolution of an initially formed amorphous layer was accelerated by increasing the electrolyte temperature, with negligible influence of the temperature on the morphology of microcones up to 448 K.

Thermal diffusivity and conductivity of thorium- uranium mixed oxides

NASA Astrophysics Data System (ADS)

Saoudi, M.; Staicu, D.; Mouris, J.; Bergeron, A.; Hamilton, H.; Naji, M.; Freis, D.; Cologna, M.

2018-03-01

Thorium-uranium oxide pellets with high densities were prepared at the Canadian Nuclear Laboratories (CNL) by co-milling, pressing, and sintering at 2023 K, with UO2 mass contents of 0, 1.5, 3, 8, 13, 30, 60 and 100%. At the Joint Research Centre, Karlsruhe (JRC-Karlsruhe), thorium-uranium oxide pellets were prepared using the spark plasma sintering (SPS) technique with 79 and 93 wt. % UO2. The thermal diffusivity of (Th1-xUx)O2 (0 ≤ x ≤ 1) was measured at CNL and at JRC-Karlsruhe using the laser flash technique. ThO2 and (Th,U)O2 with 1.5, 3, 8 and 13 wt. % UO2 were found to be semi-transparent to the infrared wavelength of the laser and were coated with graphite for the thermal diffusivity measurements. This semi-transparency decreased with the addition of UO2 and was lost at about 30 wt. % of UO2 in ThO2. The thermal conductivity was deduced using the measured density and literature data for the specific heat capacity. The thermal conductivity for ThO2 is significantly higher than for UO2. The thermal conductivity of (Th,U)O2 decreases rapidly with increasing UO2 content, and for UO2 contents of 60% and higher, the conductivity of the thorium-uranium oxide fuel is close to UO2. As the mass difference between the Th and U atoms is small, the thermal conductivity decrease is attributed to the phonon scattering enhanced by lattice strain due to the introduction of uranium in ThO2 lattice. The new results were compared to the data available in the literature and were evaluated using the classical phonon transport model for oxide systems.

METHOD OF ROLLING URANIUM

DOEpatents

Smith, C.S.

1959-08-01

A method is described for rolling uranium metal at relatively low temperatures and under non-oxidizing conditions. The method involves the steps of heating the uranium to 200 deg C in an oil bath, withdrawing the uranium and permitting the oil to drain so that only a thin protective coating remains and rolling the oil coated uranium at a temperature of 200 deg C to give about a 15% reduction in thickness at each pass. The operation may be repeated to accomplish about a 90% reduction without edge cracking, checking or any appreciable increase in brittleness.

Reductive stripping process for the recovery of uranium from wet-process phosphoric acid

DOEpatents

Hurst, Fred J.; Crouse, David J.

1984-01-01

A reductive stripping flow sheet for recovery of uranium from wet-process phosphoric acid is described. Uranium is stripped from a uranium-loaded organic phase by a redox reaction converting the uranyl to uranous ion. The uranous ion is reoxidized to the uranyl oxidation state to form an aqueous feed solution highly concentrated in uranium. Processing of this feed through a second solvent extraction cycle requires far less stripping reagent as compared to a flow sheet which does not include the reductive stripping reaction.

CASTING METHOD AND APPARATUS

DOEpatents

Gray, C.F.; Thompson, R.H.

1958-10-01

An improved apparatus for the melting and casting of uranium is described. A vacuum chamber is positioned over the casting mold and connected thereto, and a rod to pierce the oxide skin of the molten uranium is fitted into the bottom of the melting chamber. The entire apparatus is surrounded by a jacket, and operations are conducted under a vacuum. The improvement in this apparatus lies in the fact that the top of the melting chamber is fitted with a plunger which allows squeezing of the oxide skin to force out any molten uranium remaining after the skin has been broken and the molten charge has been cast.

A review of the high temperature oxidation of uranium oxides in molten salts and in the solid state to form alkali metal uranates, and their composition and properties

NASA Astrophysics Data System (ADS)

Griffiths, Trevor R.; Volkovich, Vladimir A.

An extensive review of the literature on the high temperature reactions (both in melts and in the solid state) of uranium oxides (UO 2, U 3O 8 and UO 3) resulting in the formation of insoluble alkali metal (Li to Cs) uranates is presented. Their uranate(VI) and uranate(V) compounds are examined, together with mixed and oxygen-deficient uranates. The reactions of uranium oxides with carbonates, oxides, per- and superoxides, chlorides, sulfates, nitrates and nitrites under both oxidising and non-oxidising conditions are critically examined and systematised, and the established compositions of a range of uranate(VI) and (V) compounds formed are discussed. Alkali metal uranates(VI) are examined in detail and their structural, physical, thermodynamic and spectroscopic properties considered. Chemical properties of alkali metal uranates(VI), including various methods for their reduction, are also reported. Errors in the current theoretical treatment of uranate(VI) spectra are identified and the need to develop routes for the preparation of single crystals is stressed.

Optimization of Uranium-Doped Americium Oxide Synthesis for Space Application.

PubMed

Vigier, Jean-François; Freis, Daniel; Pöml, Philipp; Prieur, Damien; Lajarge, Patrick; Gardeur, Sébastien; Guiot, Antony; Bouëxière, Daniel; Konings, Rudy J M

2018-04-16

Americium 241 is a potential alternative to plutonium 238 as an energy source for missions into deep space or to the dark side of planetary bodies. In order to use the 241 Am isotope for radioisotope thermoelectric generator or radioisotope heating unit (RHU) production, americium materials need to be developed. This study focuses on the stabilization of a cubic americium oxide phase using uranium as the dopant. After optimization of the material preparation, (Am 0.80 U 0.12 Np 0.06 Pu 0.02 )O 1.8 has been successfully synthesized to prepare a 2.96 g pellet containing 2.13 g of 241 Am for fabrication of a small scale RHU prototype. Compared to the use of pure americium oxide, the use of uranium-doped americium oxide leads to a number of improvements from a material properties and safety point of view, such as good behavior under sintering conditions or under alpha self-irradiation. The mixed oxide is a good host for neptunium (i.e., the 241 Am daughter element), and it has improved safety against radioactive material dispersion in the case of accidental conditions.

THE ANALYSIS OF URANIUM-ZIRCONIUM ALLOYS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Milner, G.W.C.; Skewies, A.F.

1953-03-01

A satisfactory procedure is described for the analysis of uranium-zirconium alloys containing up to 25% zirconium. It is based on the separation of the zirconium from the uranium by dissolving the cupferron complex of the former element into chloroform. After the evaporation of the solvent from the combined organic extracts, the residue is ignited to zirconium oxide. The latter is then re-dissolved and zirconium is separated from other elements co-extracted in the solvent extraction procedure by precipitation with mandelic acid. The zirconium mandelate is finally ignited to oxide at 960 deg C. The uranium is separated from the aqueous solutionmore » remaining from the cupferron extraction by precipitating with tannin at a pH of 8; the precipitate being removed by filtration and then ignited a t 800 deg C. The residue is dissolved in nitric acid and the uranium is finally determined by precipitating as ammonium diuranate and then igniting to U{sub 3}O{sub 8}. (auth)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shepherd, James; Fairweather, Michael; Hanson, Bruce C.

The oxidation of spent uranium carbide fuel, a candidate fuel for Generation IV nuclear reactors, is an important process in its potential reprocessing cycle. However, the oxidation of uranium carbide in air is highly exothermic. A model has therefore been developed to predict the temperature rise, as well as other useful information such as reaction completion times, under different reaction conditions in order to help in deriving safe oxidation conditions. Finite difference-methods are used to model the heat and mass transfer processes occurring during the reaction in two dimensions and are coupled to kinetics found in the literature.

XPS and SIMS study of the surface and interface of aged C + implanted uranium

DOE PAGES

Donald, Scott B.; Siekhaus, Wigbert J.; Nelson, Art J.

2016-09-08

X-ray photoelectron spectroscopy in combination with secondary ion mass spectrometry depth profiling were used to investigate the surface and interfacial chemistry of C + ion implanted polycrystalline uranium subsequently oxidized in air for over 10 years at ambient temperature. The original implantation of 33 keV C + ions into U 238 with a dose of 4.3 × 10 17 cm –3 produced a physically and chemically modified surface layer that was characterized and shown to initially prevent air oxidation and corrosion of the uranium after 1 year in air at ambient temperature. The aging of the surface and interfacial layersmore » were examined by using the chemical shift of the U 4f, C 1s, and O 1s photoelectron lines. In addition, valence band spectra were used to explore the electronic structure of the aged carbide surface and interface layer. Moreover, the time-of-flight secondary ion mass spectrometry depth profiling results for the aged sample confirmed an oxidized uranium carbide layer over the carbide layer/U metal interface.« less

Direct electrochemical reduction of solid uranium oxide in molten fluoride salts

NASA Astrophysics Data System (ADS)

Gibilaro, Mathieu; Cassayre, Laurent; Lemoine, Olivier; Massot, Laurent; Dugne, Olivier; Malmbeck, Rikard; Chamelot, Pierre

2011-07-01

The direct electrochemical reduction of UO 2 solid pellets was carried out in LiF-CaF 2 (+2 mass.% Li 2O) at 850 °C. An inert gold anode was used instead of the usual reactive sacrificial carbon anode. In this case, oxidation of oxide ions present in the melt yields O 2 gas evolution on the anode. Electrochemical characterisations of UO 2 pellets were performed by linear sweep voltammetry at 10 mV/s and reduction waves associated to oxide direct reduction were observed at a potential 150 mV more positive in comparison to the solvent reduction. Subsequent, galvanostatic electrolyses runs were carried out and products were characterised by SEM-EDX, EPMA/WDS, XRD and microhardness measurements. In one of the runs, uranium oxide was partially reduced and three phases were observed: nonreduced UO 2 in the centre, pure metallic uranium on the external layer and an intermediate phase representing the initial stage of reduction taking place at the grain boundaries. In another run, the UO 2 sample was fully reduced. Due to oxygen removal, the U matrix had a typical coral-like structure which is characteristic of the pattern observed after the electroreduction of solid oxides.

TERNARY ALLOYS OF URANIUM, COLUMBIUM, AND ZIRCONIUM

DOEpatents

Foote, F.G.

1960-08-01

Ternary alloys of uranium are described which are useful as neutron- reflecting materials in a fast neutron reactor. They are especially resistant to corrosion caused by oxidative processes of gascous or aqueous origin and comprise uranium as the predominant metal with zirconiunn and niobium wherein the total content of the minor alloying elements is between 2 and 8% by weight.

Assessment of undiscovered sandstone-hosted uranium resources in the Texas Coastal Plain, 2015

USGS Publications Warehouse

Mihalasky, Mark J.; Hall, Susan M.; Hammarstrom, Jane M.; Tureck, Kathleen R.; Hannon, Mark T.; Breit, George N.; Zielinski, Robert A.; Elliott, Brent

2015-12-02

The U.S. Geological Survey estimated a mean of 220 million pounds of recoverable uranium oxide (U3O8 ) remaining as potential undiscovered resources in southern Texas. This estimate used a geology-based assessment method for Tertiary sandstone-hosted uranium deposits in the Texas Coastal Plain sedimentary strata (fig.1).

A physical model for evaluating uranium nitride specific heat

NASA Astrophysics Data System (ADS)

Baranov, V. G.; Devyatko, Yu. N.; Tenishev, A. V.; Khlunov, A. V.; Khomyakov, O. V.

2013-03-01

Nitride fuel is one of perspective materials for the nuclear industry. But unlike the oxide and carbide uranium and mixed uranium-plutonium fuel, the nitride fuel is less studied. The present article is devoted to the development of a model for calculating UN specific heat on the basis of phonon spectrum data within the solid state theory.

Influence of oxidation state on water solubility of Si nanoparticles prepared by laser ablation in water

NASA Astrophysics Data System (ADS)

Ryabchikov, Yu. V.; Al-Kattan, A.; Chirvony, V.; Sanchez-Royo, J. F.; Sentis, M.; Timoshenko, V. Yu.; Kabashin, A. V.

2017-02-01

Femtosecond laser fragmentation from preliminarily prepared water-dispersed Si microcolloids was used to synthesize bare (ligand-free) spherical silicon nanoparticles (Si-NPs) with low size dispersion and controllable mean size from a few nm to several tens of nm. In order to control the oxidation state of Si-NPs, the fragmentation was performed in normal oxygen-saturated water (oxygen-rich conditions) or in water disoxygenated by pumping with noble gases (Ag, He) before and during the experiment (oxygen-free conditions). XPS and TEM studies revealed that Si-NPs were composed of Si nanocrystals with inclusions of silicon oxide species, covered by SiOx (1 < x < 2) shell, while the total oxide content depended whether Si-NPs were prepared in oxygen-rich or oxygen-free conditions. When placed into a dialysis box, waterdispersed Si-NPs rapidly dissolved, which was evidenced by TEM data. In this case, NPs prepared under oxygen-rich conditions demonstrated much faster dissolution kinetics and their complete disappearance after 7-10 days, while the dissolution process of less oxidized counterparts could last much longer (25-30 days). Much fast dissolution kinetics of more oxidized Si-NPs was attributed to more friable structure of nanoparticle core due to the presence of numerous oxidation-induced defects. Laser-synthesized Si-NPs are of paramount importance for biomedical applications.

Nuclear fuel electrorefiner

DOEpatents

Ahluwalia, Rajesh K.; Hua, Thanh Q.

2004-02-10

The present invention relates to a nuclear fuel electrorefiner having a vessel containing a molten electrolyte pool floating on top of a cadmium pool. An anodic fuel dissolution basket and a high-efficiency cathode are suspended in the molten electrolyte pool. A shroud surrounds the fuel dissolution basket and the shroud is positioned so as to separate the electrolyte pool into an isolated electrolyte pool within the shroud and a bulk electrolyte pool outside the shroud. In operation, unwanted noble-metal fission products migrate downward into the cadmium pool and form precipitates where they are removed by a filter and separator assembly. Uranium values are transported by the cadmium pool from the isolated electrolyte pool to the bulk electrolyte pool, and then pass to the high-efficiency cathode where they are electrolytically deposited thereto.

Reduction of arsenic content in a complex galena concentrate by Acidithiobacillus ferrooxidans

PubMed Central

Makita, Mario; Esperón, Margarita; Pereyra, Benito; López, Alejandro; Orrantia, Erasmo

2004-01-01

Background Bioleaching is a process that has been used in the past in mineral pretreatment of refractory sulfides, mainly in the gold, copper and uranium benefit. This technology has been proved to be cheaper, more efficient and environmentally friendly than roasting and high pressure moisture heating processes. So far the most studied microorganism in bioleaching is Acidithiobacillus ferrooxidans. There are a few studies about the benefit of metals of low value through bioleaching. From all of these, there are almost no studies dealing with complex minerals containing arsenopyrite (FeAsS). Reduction and/or elimination of arsenic in these ores increase their value and allows the exploitation of a vast variety of minerals that today are being underexploited. Results Arsenopyrite was totally oxidized. The sum of arsenic remaining in solution and removed by sampling represents from 22 to 33% in weight (yield) of the original content in the mineral. The rest of the biooxidized arsenic form amorphous compounds that precipitate. Galena (PbS) was totally oxidized too, anglesite (PbSO4) formed is virtually insoluble and remains in the solids. The influence of seven factors in a batch process was studied. The maximum rate of arsenic dissolution in the concentrate was found using the following levels of factors: small surface area of particle exposure, low pulp density, injecting air and adding 9 K medium to the system. It was also found that ferric chloride and carbon dioxide decreased the arsenic dissolution rate. Bioleaching kinetic data of arsenic solubilization were used to estimate the dilution rate for a continuous culture. Calculated dilution rates were relatively small (0.088–0.103 day-1). Conclusion Proper conditions of solubilization of arsenic during bioleaching are key features to improve the percentage (22 to 33% in weight) of arsenic removal. Further studies are needed to determine other factors that influence specifically the solubilization of arsenic in the bioleaching system such as: pH, dissolved oxygen concentration, redox potentials, nature of concentrate and temperature among others. At. ferrooxidans was able to completely oxidize the minerals present during the arsenic bioleaching. Other elements present originally in the concentrate such as Zn, Sb, and Cu were also solubilized. The process of bioleaching is expected to be influenced by mechanisms that still need to be established due to the diversity of the minerals involved and by the presence of traces of metals in the concentrate. The increase in pulp density generates a decrease in the dissolved arsenic concentration. This decrease is greater in runs where air was not injected to the system. The maximum rate of arsenic dissolution in the concentrate was found using; small surface area of particle exposure, low pulp density, injecting air and adding 9 K medium to the system. The effect of addition of ferric chloride during the arsenic bioleaching resulted in a decrease of the solubilized arsenic in the system. The presence of CO2 is associated to the decrease in arsenic dissolution. PMID:15482595

Extracellular reduction of uranium via Geobacter conductive pili as a protective cellular mechanism.

PubMed

Cologgi, Dena L; Lampa-Pastirk, Sanela; Speers, Allison M; Kelly, Shelly D; Reguera, Gemma

2011-09-13

The in situ stimulation of Fe(III) oxide reduction by Geobacter bacteria leads to the concomitant precipitation of hexavalent uranium [U(VI)] from groundwater. Despite its promise for the bioremediation of uranium contaminants, the biological mechanism behind this reaction remains elusive. Because Fe(III) oxide reduction requires the expression of Geobacter's conductive pili, we evaluated their contribution to uranium reduction in Geobacter sulfurreducens grown under pili-inducing or noninducing conditions. A pilin-deficient mutant and a genetically complemented strain with reduced outer membrane c-cytochrome content were used as controls. Pili expression significantly enhanced the rate and extent of uranium immobilization per cell and prevented periplasmic mineralization. As a result, pili expression also preserved the vital respiratory activities of the cell envelope and the cell's viability. Uranium preferentially precipitated along the pili and, to a lesser extent, on outer membrane redox-active foci. In contrast, the pilus-defective strains had different degrees of periplasmic mineralization matching well with their outer membrane c-cytochrome content. X-ray absorption spectroscopy analyses demonstrated the extracellular reduction of U(VI) by the pili to mononuclear tetravalent uranium U(IV) complexed by carbon-containing ligands, consistent with a biological reduction. In contrast, the U(IV) in the pilin-deficient mutant cells also required an additional phosphorous ligand, in agreement with the predominantly periplasmic mineralization of uranium observed in this strain. These findings demonstrate a previously unrecognized role for Geobacter conductive pili in the extracellular reduction of uranium, and highlight its essential function as a catalytic and protective cellular mechanism that is of interest for the bioremediation of uranium-contaminated groundwater.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gregoire, D.C.; Goltz, D.M.; Chakrabarti, C.L.

Graphite furnace atomic absorption spectrometry (GFAAS) is an insensitive technique for determination of uranium. Experiments were conducted using electrothermal vaporization inductively coupled plasma mass spectrometry to investigate the atomization and vaporization of atomic and molecular uranium species in the graphite furnace. ETV-ICP-MS signals for uranium were observed at temperatures well below the appearance temperature of uranium atoms suggesting the vaporization of molecular uranium oxide at temperatures below 2000{degrees}C. Examination of individual uranium ETV-ICP-MS signals reveals the vaporization of uranium carbide at temperatures above 2600{degrees}C. Chemical modifiers such as 0.2% HF and 0.1% CHF{sub 3} in the argon carrier gas, weremore » ineffective in preventing the formation of uranium carbide at 2700{degrees}C. Vaporization of uranium from a tungsten surface using tungsten foil inserted into the graphite tube prevented the formation of uranium carbide and eliminated the ETV-ICP-MS signal suppression caused by a sodium chloride matrix.« less

Origin of the Okrouhlá Radouň episyenite-hosted uranium deposit, Bohemian Massif, Czech Republic: fluid inclusion and stable isotope constraints

NASA Astrophysics Data System (ADS)

Dolníček, Zdeněk; René, Miloš; Hermannová, Sylvie; Prochaska, Walter

2014-04-01

The Okrouhlá Radouň shear zone hosted uranium deposit is developed along the contact of Variscan granites and high-grade metasedimentary rocks of the Moldanubian Zone of the Bohemian Massif. The pre-ore pervasive alteration of wall rocks is characterized by chloritization of mafic minerals, followed by albitization of feldspars and dissolution of quartz giving rise to episyenites. The subsequent fluid circulation led to precipitation of disseminated uraninite and coffinite, and later on, post-ore quartz and carbonate mineralization containing base metal sulfides. The fluid inclusion and stable isotope data suggest low homogenization temperatures (˜50-140 °C during pre-ore albitization and post-ore carbonatization, up to 230 °C during pre-ore chloritization), variable fluid salinities (0-25 wt.% NaCl eq.), low fluid δ18O values (-10 to +2 ‰ V-SMOW), low fluid δ13C values (-9 to -15 ‰ V-PDB), and highly variable ionic composition of the aqueous fluids (especially Na/Ca, Br/Cl, I/Cl, SO4/Cl, NO3/Cl ratios). The available data suggest participation of three fluid endmembers of primarily surficial origin during alteration and mineralization at the deposit: (1) local meteoric water, (2) Na-Ca-Cl basinal brines or shield brines, (3) SO4-NO3-Cl-(H)CO3 playa-like fluids. Pre-ore albitization was caused by circulation of alkaline, oxidized, and Na-rich playa fluids, whereas basinal/shield brines and meteoric water were more important during the post-ore stage of alteration.

URANIUM PURIFICATION PROCESS

DOEpatents

Winters, C.E.

1957-11-12

A method for the preparation of a diethyl ether solution of uranyl nitrate is described. Previously the preparation of such ether solutions has been difficult and expensive, since crystalline uranyl nitrate hexahydrate dissolves very slowly in ether. An improved method for effecting such dissolution has been found, and it comprises adding molten uranyl nitrate hexahydrate at a temperature of 65 to 105 deg C to the ether while maintaining the temperature of the ether solvent below its boiling point.

Ground-Water Geochemistry of Kwajalein Island, Republic of the Marshall Islands, 1991

USGS Publications Warehouse

Tribble, Gordon W.

1997-01-01

Ground water on Kwajalein Island is an important source of drinking water, particularly during periods of low rainfall. Fresh ground water is found as a thin lens underlain by saltwater. The concentration of dissolved ions increases with depth below the water table and proximity to the shoreline as high-salinity seawater mixes with fresh ground water. The maximum depth of the freshwater lens is 37 ft. Chloride is assumed to be non-reactive under the range of geochemical conditions on the atoll. The concentration of chloride thus is used as a conservative constituent to evaluate freshwater-saltwater mixing within the aquifer. Concentrations of sodium and for the most part, potassium and sulfate, also appear to be determined by conservative mixing between saltwater and rain. Concentrations of calcium, magnesium, and strontium are higher than expected from conservative mixing; these higher concentrations are a result of the dissolution of carbonate minerals. An excess in dissolved inorganic carbon results from carbonate-mineral dissolution and from the oxidation of organic matter in the aquifer; the stoichiometric difference between excess dissolved inorganic carbon and excess bivalent cations is used as a measure of the amount of organic-matter oxidation. Organic-matter oxidation also is indicated by the low concentration of dissolved oxygen, high concentrations of nutrients, and the presence of hydrogen sulfide in many of the water samples. Low levels of dissolved oxygen indicate oxic respiration, and sulfate reduction is indicated by hydrogen sulfide. The amount of dissolved inorganic carbon released during organic-matter oxidation is nearly equivalent to the amount of carbonate-mineral dissolution. Organic-matter oxidation and carbonate-mineral dissolution seem to be most active either in the unsaturated zone or near the top of the water table. The most plausible explanation is that high amounts of oxic respiration in the unsaturated zone generate carbon dioxide, which causes carbonate minerals to dissolve. Ground water contaminated by petroleum hydrocarbons had the highest levels of mineral dissolution and organic respiration (including sulfate reduction), indicating that bacteria are oxidizing the contaminants.

AFS-2 FLOWSHEET MODIFICATIONS TO ADDRESS THE INGROWTH OF PU(VI) DURING METAL DISSOLUTION

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crapse, K.; Rudisill, T.; O'Rourke, P.

2014-07-02

In support of the Alternate Feed Stock Two (AFS-2) PuO{sub 2} production campaign, Savannah River National Laboratory (SRNL) conducted a series of experiments concluding that dissolving Pu metal at 95°C using a 6–10 M HNO{sub 3} solution containing 0.05–0.2 M KF and 0–2 g/L B could reduce the oxidation of Pu(IV) to Pu(VI) as compared to dissolving Pu metal under the same conditions but at or near the boiling temperature. This flowsheet was demonstrated by conducting Pu metal dissolutions at 95°C to ensure that PuO{sub 2} solids were not formed during the dissolution. These dissolution parameters can be used formore » dissolving both Aqueous Polishing (AP) and MOX Process (MP) specification materials. Preceding the studies reported herein, two batches of Pu metal were dissolved in the H-Canyon 6.1D dissolver to prepare feed solution for the AFS-2 PuO{sub 2} production campaign. While in storage, UV-visible spectra obtained from an at-line spectrophotometer indicated the presence of Pu(VI). Analysis of the solutions also showed the presence of Fe, Ni, and Cr. Oxidation of Pu(IV) produced during metal dissolution to Pu(VI) is a concern for anion exchange purification. Anion exchange requires Pu in the +4 oxidation state for formation of the anionic plutonium(IV) hexanitrato complex which absorbs onto the resin. The presence of Pu(VI) in the anion feed solution would require a valence adjustment step to prevent losses. In addition, the presence of Cr(VI) would result in absorption of chromate ion onto the resin and could limit the purification of Pu from Cr which may challenge the purity specification of the final PuO{sub 2} product. Initial experiments were performed to quantify the rate of oxidation of Pu(IV) to Pu(VI) (presumed to be facilitated by Cr(VI)) as functions of the HNO{sub 3} concentration and temperature in simulated dissolution solutions containing Cr, Fe, and Ni. In these simulated Pu dissolutions studies, lowering the temperature from near boiling to 95 °C reduced the oxidation rate of Pu(IV) to Pu(VI). For 8.1 M HNO{sub 3} simulated dissolution solutions, at near boiling conditions >35% Pu(VI) was present in 50 h while at 95 °C <10% Pu(VI) was present at 50 h. At near boiling temperatures, eliminating the presence of Cr and varying the HNO{sub 3} concentration in the range of 7–8.5 M had little effect on the rate of conversion of Pu(IV) to Pu(VI). HNO{sub 3} oxidation of Pu(IV) to Pu(VI) in a pure solution has been reported previously. Based on simulated dissolution experiments, this study concluded that dissolving Pu metal at 95°C using a 6 to 10 M HNO{sub 3} solution 0.05–0.2 M KF and 0–2 g/L B could reduce the rate of oxidation of Pu(IV) to Pu(VI) as compared to near boiling conditions. To demonstrate this flowsheet, two small-scale experiments were performed dissolving Pu metal up to 6.75 g/L. No Pu-containing residues were observed in the solutions after cooling. Using Pu metal dissolution rates measured during the experiments and a correlation developed by Holcomb, the time required to completely dissolve a batch of Pu metal in an H-Canyon dissolver using this flowsheet was estimated to require nearly 5 days (120 h). This value is reasonably consistent with an estimate based on the Batch 2 and 3 dissolution times in the 6.1D dissolver and Pu metal dissolution rates measured in this study and by Rudisill et al. Data from the present and previous studies show that the Pu metal dissolution rate decreases by a factor of approximately two when the temperature decreased from boiling (112 to 116°C) to 95°C. Therefore, the time required to dissolve a batch of Pu metal in an H-Canyon dissolver at 95°C would likely double (from 36 to 54 h) and require 72 to 108 h depending on the surface area of the Pu metal. Based on the experimental studies, a Pu metal dissolution flowsheet utilizing 6–10 M HNO{sub 3} containing 0.05–0.2 M KF (with 0–2 g/L B) at 95°C is recommended to reduce the oxidation of Pu(IV) to Pu(VI) as compared to near boiling conditions. The time required to completely dissolve a batch of Pu metal will increase, however, by approximately a factor of two as compared to initial dissolutions at near boiling (assuming the KF concentration is maintained at nominally 0.1 M). By lowering the temperature to 95°C under otherwise the same operating parameters as previous dissolutions, the Pu(VI) concentration should not exceed 15% after a 120 h heating cycle. Increasing the HNO{sub 3} concentration and lowering Pu concentration are expected to further limit the amount of Pu(VI) formed.« less

Assessment of undiscovered resources in calcrete uranium deposits, Southern High Plains region of Texas, New Mexico, and Oklahoma, 2017

USGS Publications Warehouse

Hall, Susan M.; Mihalasky, Mark J.; Van Gosen, Bradley S.

2017-11-14

The U.S. Geological Survey estimates a mean of 40 million pounds of in-place uranium oxide (U3O8) remaining as potential undiscovered resources in the Southern High Plains region of Texas, New Mexico, and Oklahoma. This estimate used a geology-based assessment method specific to calcrete uranium deposits.

Retention and chemical speciation of uranium in an oxidized wetland sediment from the Savannah River Site.

PubMed

Li, Dien; Seaman, John C; Chang, Hyun-Shik; Jaffe, Peter R; Koster van Groos, Paul; Jiang, De-Tong; Chen, Ning; Lin, Jinru; Arthur, Zachary; Pan, Yuanming; Scheckel, Kirk G; Newville, Matthew; Lanzirotti, Antonio; Kaplan, Daniel I

2014-05-01

Uranium speciation and retention mechanisms onto Savannah River Site (SRS) wetland sediments was studied using batch (ad)sorption experiments, sequential extraction, U L3-edge X-ray absorption near-edge structure (XANES) spectroscopy, fluorescence mapping and μ-XANES. Under oxidized conditions, U was highly retained by the SRS wetland sediments. In contrast to other similar but much lower natural organic matter (NOM) sediments, significant sorption of U onto the SRS sediments was observed at pH < 4 and pH > 8. Sequential extraction indicated that the U species were primarily associated with the acid soluble fraction (weak acetic acid extractable) and organic fraction (Na-pyrophosphate extractable). Uranium L3-edge XANES spectra of the U-bound sediments were nearly identical to that of uranyl acetate. Based on fluorescence mapping, U and Fe distributions in the sediment were poorly correlated, U was distributed throughout the sample and did not appear as isolated U mineral phases. The primary oxidation state of U in these oxidized sediments was U(VI), and there was little evidence that the high sorptive capacity of the sediments could be ascribed to abiotic or biotic reduction to the less soluble U(IV) species or to secondary mineral formation. Collectively, this study suggests that U may be strongly bound to wetland sediments, not only under reducing conditions by reductive precipitation, but also under oxidizing conditions through NOM-uranium bonding. Published by Elsevier Ltd.

Uranium Hydride Nucleation and Growth Model FY'16 ESC Annual Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hill, Mary Ann; Richards, Andrew Walter; Holby, Edward F.

2016-12-20

Uranium hydride corrosion is of great interest to the nuclear industry. Uranium reacts with water and/or hydrogen to form uranium hydride which adversely affects material performance. Hydride nucleation is influenced by thermal history, mechanical defects, oxide thickness, and chemical defects. Information has been gathered from past hydride experiments to formulate a uranium hydride model to be used in a Canned Subassembly (CSA) lifetime prediction model. This multi-scale computer modeling effort started in FY’13, and the fourth generation model is now complete. Additional high-resolution experiments will be run to further test the model.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buck, E.C.; Dietz, N.L.; Bates, J.K.

Uranium contaminated soils from the Fernald Operation Site, Ohio, have been examined by a combination of optical microscopy, scanning electron microscopy with backscattered electron detection (SEM/BSE), and analytical electron microscopy (AEM). A method is described for preparing of transmission electron microscopy (TEM) thin sections by ultramicrotomy. By using these thin sections, SEM and TEM images can be compared directly. Uranium was found in iron oxides, silicates (soddyite), phosphates (autunites), and fluorite. Little uranium was associated with clays. The distribution of uranium phases was found to be inhomogeneous at the microscopic level.

Uranium Sequestration During Biostimulated Reduction and In Response to the Return of Oxic Conditions In Shallow Aquifers

USGS Publications Warehouse

Fuller, Christopher C.; Johnson, Kelly J.; Akstin, Katherine; Singer, David M.; Yabusaki, Steven B.; Fang, Yilin; Fuhrmann, M.

2015-01-01

A proposed approach for groundwater remediation of uranium contamination is to generate reducing conditions by stimulating the growth of microbial populations through injection of electron donor compounds into the subsurface. Sufficiently reducing conditions will result in reduction of soluble hexavalent uranium, U(VI), and precipitation of the less soluble +4 oxidation state uranium, U(IV). This process is termed biostimulated reduction. A key issue in the remediation of uranium (U) contamination in aquifers by biostimulated reduction is the long term stability of the sequestered uranium. Three flow-through column experiments using aquifer sediment were used to evaluate the remobilization of bioreduced U sequestered under conditions in which biostimulation extended well into sulfate reduction to enhance precipitation of reduced sulfur phases such as iron sulfides. One column received added ferrous iron, Fe(II), increasing production of iron sulfides, to test their effect on remobilization of the sequestered uranium, either by serving as a redox buffer by competing for dissolved oxygen, or by armoring the reduced uranium. During biostimulation of the ambient microbial population with acetate, dissolved uranium was lowered by a factor of 2.5 or more with continued removal for over 110 days of biostimulation, well after the onset of sulfate reduction at ~30 days. Sequestered uranium was essentially all U(IV) resulting from the formation of nano-particulate uraninite that coated sediment grains to a thickness of a few 10’s of microns, sometimes in association with S and Fe. A multicomponent biogeochemical reactive transport model simulation of column effluents during biostimulation was generally able to describe the acetate oxidation, iron, sulfate, and uranium reduction for all three columns using parameters derived from simulations of field scale biostimulation experiments. Columns were eluted with artificial groundwater at equilibrium with atmospheric oxygen to simulate the upper limit of dissolved oxygen in recharge water. Overall about 9% of total uranium removed from solution during biostimulation was remobilized. Release of U during oxic elution was a continuous process over 140 days with dissolved uranium concentrations about 0.2 and 0.8 aM for columns with and without ferrous iron addition, respectively. Uranium remaining on the sediment was in the reduced form. The prolonged period of biostimulation and concomitant sulfate reduction appears to limit the rate of U(IV) oxidative remobilization in contrast to a large release observed for columns in previous studies that did not undergo sulfate reduction. Although continued sulfate reduction may cause decreased permeability from precipitation of iron sulfide, the greater apparent stability of the sequestered U(IV) provided by the sustained biostimulation should be considered in design of field scale remediation efforts. Remobilization of uranium following biostimulated reduction should be tested further at the field scale.

Effect of the microstructure of Ti-5Mo on the anodic dissolution in H/sub 2/SO/sub 4/

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Y.J.; Oriani, R.A.

1987-04-01

The effect of microstructure of the Ti-5Mo alloy on its anodic dissolution rate in sulfuric acid solution at various temperatures has been investigated. TiMo alloys exhibit a region of increased dissolution rate in the vicinity of +0.20 V (saturated calomel electrode (SCE)) in 10% H/sub 2/SO/sub 4/, the same potential region in which pure Mo exhibits a large anodic dissolution rate. Aging of Ti-5Mo at 350 C was found to lead to the formation of ..omega.. phase. Heat treatment caused larger passive currents in 10% H/sub 2/SO/sub 4/, but the critical passivation potentials and corrosion potentials were not significantly affected.more » Molybdenum was enriched in the oxide formed on aged Ti-5Mo at +0.23 V (SCE), in comparison with the Mo concentration found in the oxide on as-received Ti-5Mo.« less

The synthesis, characterization and application of iron oxide nanocrystals in magnetic separations for arsenic and uranium removal

NASA Astrophysics Data System (ADS)

Mayo, John Thomas

Arsenic and uranium in the environment are hazardous to human health and require better methods for detection and remediation. Nanocrystalline iron oxides offer a number of advantages as sorbents for water purification and environmental remediation. First, highly uniform and crystalline iron oxide nanocrystals (nMAG) were prepared using thermal decomposition of iron salts in organic solutions; for the applications of interest in this thesis, a central challenge was the adaptation of these conventional synthetic methods to the needs of low infrastructure and economically disadvantaged settings. We show here that it is possible to form highly uniform and magnetically responsive nanomaterials using starting reagents and equipment that are readily available and economical. The products of this approach, termed the 'Kitchen Synthesis', are of comparable quality and effectiveness to laboratory materials. The narrow size distributions of the iron oxides produced in the laboratory synthesis made it possible to study the size-dependence of the magnetic separation efficiency of nanocrystals; generally as the diameter of particles increased they could be removed under lower applied magnetic fields. In this work we take advantage of this size-dependence to use magnetic separation as a tool to separate broadly distributed populations of magnetic materials. Such work makes it possible to use these materials in multiplexed separation and sensing schemes. With the synthesis and magnetic separation studies of these materials completed, it was possible to optimize their applications in water purification and environmental remediation. These materials removed both uranium and arsenic from contaminated samples, and had remarkably high sorption capacities --- up to 12 wt% for arsenic and 30 wt% for uranium. The contaminated nMAG is removed from the drinking water by either retention in a sand column, filter, or by magnetic separation. The uranium adsorption process was also utilized for the enhanced detection of uranium in environmental matrices. By relying on alpha-particle detection in well-formed and dense nMAG films, it was possible to improve soil detection of uranium by more than ten-thousand-fold. Central for this work was a detailed understanding of the chemistry at the iron oxide interface, and the role of the organic coatings in mediating the sorption process.

METHOD OF MAKING WIRE FUEL ELEMENTS

DOEpatents

Zambrow, J.L.

1960-08-01

A method is given for making a nuclear reactor fuel element in the form of a uranium-bearing wire clad with zirconium. A uranium bar is enclosed in a zirconium sheath which is coated with an oxide of magnesium, beryllium, or zirconium. The sheathed bar is then placed in a steel tube and reduced to the desired diameter by swaging at 800 to 900 deg C, after which the steel and oxide are removed.

Determination of irradiated reactor uranium in soil samples in Belarus using 236U as irradiated uranium tracer.

PubMed

Mironov, Vladislav P; Matusevich, Janna L; Kudrjashov, Vladimir P; Boulyga, Sergei F; Becker, J Sabine

2002-12-01

This work presents experimental results on the distribution of irradiated reactor uranium from fallout after the accident at Chernobyl Nuclear Power Plant (NPP) in comparison to natural uranium distribution in different soil types. Oxidation processes and vertical migration of irradiated uranium in soils typical of the 30 km relocation area around Chernobyl NPP were studied using 236U as the tracer for irradiated reactor uranium and inductively coupled plasma mass spectrometry as the analytical method for uranium isotope ratio measurements. Measurements of natural uranium yielded significant variations of its concentration in upper soil layers from 2 x 10(-7) g g(-1) to 3.4 x 10(-6) g g(-1). Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 5 x 10(-12) g g(-1) to 2 x 10(-6) g g(-1) depending on the distance from Chernobyl NPP. In the majority of investigated soil profiles 78% to 97% of irradiated "Chernobyl" uranium is still contained in the upper 0-10 cm soil layers. The physical and chemical characteristics of the soil do not have any significant influence on processes of fuel particle destruction. Results obtained using carbonate leaching of 236U confirmed that more than 60% of irradiated "Chernobyl" uranium is still in a tetravalent form, ie. it is included in the fuel matrix (non-oxidized fuel UO2). The average value of the destruction rate of fuel particles determined for the Western radioactive trace (k = 0.030 +/- 0.005 yr(-1)) and for the Northern radioactive trace (k = 0.035 + 0.009 yr(-1)) coincide within experimental errors. Use of leaching of fission products in comparison to leaching of uranium for study of the destruction rate of fuel particles yielded poor coincidence due to the fact that use of fission products does not take into account differences in the chemical properties of fission products and fuel matrix (uranium).

Influence of Calcium on Microbial Reduction of Solid Phase Uranium (VI)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, Chongxuan; Jeon, Byong-Hun; Zachara, John M.

2007-06-27

The effect of calcium on microbial reduction of a solid phase U(VI), sodium boltwoodite (NaUO2SiO3OH ∙1.5H2O), was evaluated in a culture of a dissimilatory metal-reducing bacterium (DMRB), Shewanella oneidensis strain MR-1. Batch experiments were performed in a non-growth bicarbonate medium with lactate as electron donor at pH 7 buffered with PIPES. Calcium increased both the rate and extent of Na-boltwoodite dissolution by increasing its solubility through the formation of a ternary aqueous calcium-uranyl-carbonate species. The ternary species, however, decreased the rates of microbial reduction of aqueous U(VI). Laser-induced fluorescence spectroscopy (LIFS) and transmission electron microscopy (TEM) revealed that microbial reductionmore » of solid phase U(VI) is a sequentially coupled process of Na-boltwoodite dissolution, U(VI) aqueous speciation, and microbial reduction of dissolved U(VI) to U(IV) that accumulated on bacterial surfaces/periplasm. The overall rates of microbial reduction of solid phase U(VI) can be described by the coupled rates of dissolution and microbial reduction that were both influenced by calcium. The results demonstrated that dissolved U(VI) concentration during microbial reduction was a complex function of solid phase U(VI) dissolution kinetics, aqueous U(VI) speciation, and microbial activity.« less

Anaerobic microbial dissolution of lead and production of organic acids

DOEpatents

Francis, Arokiasamy J.; Dodge, Cleveland; Chendrayan, Krishnachetty; Quinby, Helen L.

1988-01-01

The present invention relates to an anaerobic bacterial culture of Clostridium sp. ATCC No. 53464 which solubilizes lead oxide under anaerobic conditions in coal and industrial wastes and therefore presents a method of removing lead from such wastes before they are dumped into the environment. The rate of lead dissolution during logarithmic growth of the bacteria in 40 ml medium containing 3.32 .mu.moles of lead as lead oxide was 0.042 .mu.moles ml.sup.-1 hr.sup.-1. Dissolution of lead oxide by the bacterial isolate is due to the production of metabolites and acidity in the culture medium. The major metabolites are acetic, butyric and lactic acid. Clostridium sp. ATCC No. 53464 can be used in the recovery of strategic metals from ores and wastes and also for the production of lactic acid for commercial purposes. The process yields large quantities of lactic acid as well as lead complexed in a stable form with said acids.

Synergistic bioleaching of chalcopyrite and bornite in the presence of Acidithiobacillus ferrooxidans.

PubMed

Zhao, Hongbo; Wang, Jun; Hu, Minghao; Qin, Wenqing; Zhang, Yansheng; Qiu, Guanzhou

2013-12-01

Bioleaching of chalcopyrite and bornite in the presence of Acidithiobacillus ferrooxidans was carried out to investigate the influences between each other during bioleaching. Bioleaching results indicated that bornite accelerated the dissolution of chalcopyrite, and chalcopyrite also accelerated the dissolution of bornite, it could be described as a synergistic effect during bioleaching, this synergistic effect might be attributed to the galvanic effect between chalcopyrite and bornite, and to the relatively low solution potential as the addition of bornite. Significantly amount of elemental sulfur and jarosite formed on the minerals surface might be the main passivation film inhibiting the further dissolution, and the amount of elemental sulfur significantly increased with the addition of bornite. Results of electrochemical measurements indicated that the oxidation and reduction mechanisms of chalcopyrite and bornite were similar, the addition of bornite or chalcopyrite did not change the oxidative and reductive mechanisms, but increased the oxidation rate. Copyright © 2013 Elsevier Ltd. All rights reserved.

Evaluating the risk from depleted uranium after the Boeing 747-258F crash in Amsterdam, 1992.

PubMed

Uijt de Haag, P A; Smetsers, R C; Witlox, H W; Krüs, H W; Eisenga, A H

2000-08-28

On 4 October 1992, a large cargo plane crashed into an apartment building in the Bijlmermeer quarter of Amsterdam. In the years following the accident, an increasing number of people started reporting health complaints, which they attributed to exposure to dangerous substances after the crash. Since the aircraft had been carrying depleted uranium as counterbalance weights and about 150 kg uranium had been found missing after clearance of the crash site, exposure to uranium oxide particles was pointed out as the possible cause of their health complaints. Six years after the accident, a risk analysis was therefore carried out to investigate whether the health complaints could be attributed to exposure to uranium oxide set free during the accident. The scientific challenge was to come up with reliable results, knowing that - considering the late date - virtually no data were available to validate any calculated result. The source term of uranium was estimated using both generic and specific data. Various dispersion models were applied in combination with the local setting and the meteorological conditions at the time of the accident to estimate the exposure of bystanders during the fire caused by the crash. Emphasis was given to analysing the input parameters, inter-comparing the various models and comparing model results with the scarce information available. Uranium oxide formed in the fire has a low solubility, making the chemical toxicity to humans less important than the radiotoxicity. Best-estimate results indicated that bystanders may have been exposed to a radiation dose of less than 1 microSv, whereas a worst-case approach indicated an upper limit of less than 1 mSv. This value is considerably less than the radiation dose for which acute effects are to be expected. It is therefore considered to be improbable that the missing uranium had indeed led to the health complaints reported.

DEHYDRATION OF DEUTERIUM OXIDE SLURRIES

DOEpatents

Hiskey, C.F.

1959-03-10

A method is presented for recovering heavy water from uranium oxide-- heavy water slurries. The method consists in saturating such slurries with a potassium nitrate-sodium nitrate salt mixture and then allowing the self-heat of the slurry to raise its temperature to a point slightly in excess of 100 deg C, thus effecting complete evaporation of the free heavy water from the slurry. The temperature of the slurry is then allowed to reach 300 to 900 deg C causing fusion of the salt mixture and expulsion of the water of hydration. The uranium may be recovered from the fused salt mixture by treatment with water to leach the soluble salts away from the uranium-containing residue.

Release of 226Ra from uranium mill tailings by microbial Fe(III) reduction

USGS Publications Warehouse

Landa, E.R.; Phillips, E.J.P.; Lovley, D.R.

1991-01-01

Uranium mill tailings were anaerobically incubated in the presence of H2 with Alteromonas putrefaciens, a bacterium known to couple the oxidation of H2 and organic compounds to the reduction of Fe(III) oxides. There was a direct correlation between the extent of Fe(III) reduction and the accumulation of dissolved 226Ra. In sterile tailings in which Fe(III) was not reduced, there was negligible leaching of 226Ra. The behavior of Ba was similar to that of Ra in inoculated and sterile systems. These results demonstrate that under anaerobic conditions, microbial reduction of Fe(III) may result in the release of dissolved 226Ra from uranium mill tailings. ?? 1991.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 tomore » 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 for lower purity materials and for fuel-type uranium-plutonium oxides. However, even in these cases testing may show that bias effects are manageable.« less

Efficient Removal of Uranium from Aqueous Solution by Reduced Graphene Oxide-Zn0.5Ni0.5Fe2O4 Ferrite-Polyaniline Nanocomposite

NASA Astrophysics Data System (ADS)

Tran, Dat Quang; Pham, Hung Thanh; Do, Hung Quoc

2017-06-01

Reduced graphene oxide-Zn0.5Ni0.5Fe2O4 ferrite-polyaniline nanocomposite (RGO-ZNF-PANI) was synthesized by a three-step method. The prepared samples were characterized by x-ray diffraction, Raman spectroscopy, scanning electron microscopy and vibrating sample magnetometer. In particular, we found that this material is capable of effectively removing uranium from an aquatic environment. This is confirmed by our experimental results using the method of inductively coupled plasma mass spectrometry. Adsorptive behaviour of uranium from an aqueous solution on the RGO-ZNF-PANI nanocomposite was examined as a function of pH, contact time, and equilibrium. Uranium concentration was carried out by batch techniques. The adsorption isotherm agrees well with the Langmuir model, having a maximum sorption capacity of 1885 mg/g, at pH 5 and 25°C.

TRACE ELEMENT ANALYSES OF URANIUM MATERIALS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beals, D; Charles Shick, C

The Savannah River National Laboratory (SRNL) has developed an analytical method to measure many trace elements in a variety of uranium materials at the high part-per-billion (ppb) to low part-per-million (ppm) levels using matrix removal and analysis by quadrapole ICP-MS. Over 35 elements were measured in uranium oxides, acetate, ore and metal. Replicate analyses of samples did provide precise results however none of the materials was certified for trace element content thus no measure of the accuracy could be made. The DOE New Brunswick Laboratory (NBL) does provide a Certified Reference Material (CRM) that has provisional values for a seriesmore » of trace elements. The NBL CRM were purchased and analyzed to determine the accuracy of the method for the analysis of trace elements in uranium oxide. These results are presented and discussed in the following paper.« less

Suppressing Manganese Dissolution from Lithium Manganese Oxide Spinel Cathodes with Single-Layer Graphene

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jaber-Ansari, Laila; Puntambekar, Kanan P.; Kim, Soo

2015-06-24

Spinel-structured LiMn 2 O 4 (LMO) is a desirable cathode material for Li-ion batteries due to its low cost, abundance, and high power capability. However, LMO suffers from limited cycle life that is triggered by manganese dissolution into the electrolyte during electrochemical cycling. Here, it is shown that single-layer graphene coatings suppress manganese dissolution, thus enhancing the performance and lifetime of LMO cathodes. Relative to lithium cells with uncoated LMO cathodes, cells with graphene-coated LMO cathodes provide improved capacity retention with enhanced cycling stability. X-ray photoelectron spectroscopy reveals that graphene coatings inhibit manganese depletion from the LMO surface. Additionally, transmissionmore » electron microscopy demonstrates that a stable solid electrolyte interphase is formed on graphene, which screens the LMO from direct contact with the electrolyte. Density functional theory calculations provide two mechanisms for the role of graphene in the suppression of manganese dissolution. First, common defects in single-layer graphene are found to allow the transport of lithium while concurrently acting as barriers for manganese diffusion. Second, graphene can chemically interact with Mn 3+ at the LMO electrode surface, promoting an oxidation state change to Mn 4+ , which suppresses dissolution.« less

Dissolution of Platinum in the Operational Range of Fuel Cells

PubMed Central

Keeley, Gareth P.; Geiger, Simon; Zeradjanin, Aleksandar R.; Hodnik, Nejc; Kulyk, Nadiia

2015-01-01

Abstract One of the most important practical issues in low‐temperature fuel‐cell catalyst degradation is platinum dissolution. According to the literature, it initiates at 0.6–0.9 VRHE, whereas previous time‐ and potential‐resolved inductively coupled plasma mass spectrometry (ICP–MS) experiments, however, revealed dissolution onset at only 1.05 VRHE. In this manuscript, the apparent discrepancy is addressed by investigating bulk and nanoparticulated catalysts. It is shown that, given enough time for accumulation, traces of platinum can be detected at potentials as low as 0.85 VRHE. At these low potentials, anodic dissolution is the dominant process, whereas, at more positive potentials, more platinum dissolves during the oxide reduction after accumulation. Interestingly, the potential and time dissolution dependence is similar for both types of electrode. Dissolution processes are discussed with relevance to fuel‐cell operation and plausible dissolution mechanisms are considered. PMID:27525206

Interpretation of aircraft multispectral scanner images for mapping of alteration with uranium mineralization, Copper Mountain, Wyoming

NASA Technical Reports Server (NTRS)

Conel, J. E.

1983-01-01

NS-001 multispectral scanner data (0.45-2.35 micron) combined as principal components were utilized to map distributions of surface oxidation/weathering in Precambrian granitic rocks at Copper Mountain, Wyoming. Intense oxidation is found over granitic outcrops in partly exhumed pediments along the southern margin of the Owl Creek uplift, and along paleodrainages higher in the range. Supergene(?) uranium mineralization in the granites is localized beneath remnant Tertiary sediments covering portions of the pediments. The patterns of mineralization and oxidation are in agreement, but the genetic connections between the two remain in doubt.

Characterization And Dissolution Properties Of Ruthenium Oxides

EPA Science Inventory

Ruthenium oxides (RuO₂•1.10H₂O and RuO₂) have been synthesized by forced hydrolysis and oxidation of ruthenium chloride. The resulting materials were extensively characterized to determine the crystallinity, surface area, and ruthenium oxidation ...

Uranium Isotope Ratios in Modern and Precambrian Soils

NASA Astrophysics Data System (ADS)

DeCorte, B.; Planavsky, N.; Wang, X.; Auerbach, D. J.; Knudsen, A. C.

2015-12-01

Uranium isotopes (δ238U values) are an emerging paleoredox proxy that can help to better understand the redox evolution of Earth's surface environment. Recently, uranium isotopes have been used to reconstruct ocean and atmospheric redox conditions (Montoya-Pino et al., 2010; Brennecka et al., 2011; Kendall et al., 2013; Dahl et al., 2014). However, to date, there have not been studies on paleosols, despite that paleosols are, arguably better suited to directly tracking the redox conditions of the atmosphere. Sedimentary δ238U variability requires the formation of the soluble, oxidized form of U, U(VI). The formation of U(VI) is generally thought to require oxygen levels orders of magnitude higher than prebiotic levels. Without significant U mobility, it would have been impossible to develop isotopically distinct pools of uranium in ancient Earth environments. Conversely, an active U redox cycle leads to significant variability in δ238U values. Here we present a temporally and geographically expansive uranium isotope record from paleosols and modern soils to better constrain atmospheric oxygen levels during the Precambrian. Preliminary U isotope measurements of paleosols are unfractionated (relative to igneous rocks), possibly because of limited fractionation during oxidation (e.g., {Wang, 2015 #478}) or insufficient atmospheric oxygen levels to oxidize U(IV)-bearing minerals in the bedrock. Further U isotope measurements of paleosols with comparison to modern soils will resolve this issue.

A Kinetics and Equilibrium Study of Vanadium Dissolution from Vanadium Oxides and Phosphates in Battery Electrolytes: Possible Impacts on ICD Battery Performance.

PubMed

Bock, David C; Marschilok, Amy C; Takeuchi, Kenneth J; Takeuchi, Esther S

2013-06-01

Silver vanadium oxide (Ag 2 V 4 O 11 , SVO) has enjoyed widespread commercial success over the past 30 years as a cathode material for implantable cardiac defibrillator (ICD) batteries. Recently, silver vanadium phosphorous oxide (Ag 2 VO 2 PO 4 , SVPO) has been studied as possibly combining the desirable thermal stability aspects of LiFePO 4 with the electrical conductivity of SVO. Further, due to the noted insoluble nature of most phosphate salts, a lower material solubility of SVPO relative to SVO is anticipated. Thus, the first vanadium dissolution studies of SVPO in battery electrolyte solutions are described herein. The equilibrium solubility of SVPO was ~5 times less than SVO, with a rate constant of dissolution ~3.5 times less than that of SVO. The vanadium dissolution in SVO and SVPO can be adequately described with a diffusion layer model, as supported by the Noyes-Whitney equation. Cells prepared with vanadium-treated anodes displayed higher AC impedance and DC resistance relative to control anodes. These data support the premise that SVPO cells are likely to exhibit reduced cathode solubility and thus less affected by increased cell resistance due to cathode solubility compared to SVO based cells.

GEOCHEMICAL AND BIOLOGICAL ASPECTS OF SULFIDE MINERAL DISSOLUTION: LESSONS FROM IRON MOUNTAIN, CALIFORNIA. (R826189)

EPA Science Inventory

Abstract

The oxidative dissolution of sulfide minerals leading to acid mine drainage (AMD) involves a complex interplay between microorganisms, solutions, and mineral surfaces. Consequently, models that link molecular level reactions and the microbial communities that ...

Dissolution of Nickel Ferrite in Aqueous Solutions Containing Oxalic Acid and Ferrous Salts.

PubMed

Figueroa, Carlos A.; Sileo, Elsa E.; Morando, Pedro J.; Blesa, Miguel A.

2000-05-15

The dissolution of nickel ferrite in oxalic acid and in ferrous oxalate-oxalic acid aqueous solution was studied. Nickel ferrite was synthesized by thermal decomposition of a mixed tartrate; the particles were shown to be coated with a thin ferric oxide layer. Dissolution takes place in two stages, the first one corresponding to the dissolution of the ferric oxide outer layer and the second one being the dissolution of Ni(1.06)Fe(1.96)O(4). The kinetics of dissolution during this first stage is typical of ferric oxides: in oxalic acid, both a ligand-assisted and a redox mechanism operates, whereas in the presence of ferrous ions, redox catalysis leads to a faster dissolution. The rate dependence on both oxalic acid and on ferrous ion is described by the Langmuir-Hinshelwood equation; the best fitting corresponds to K(1)(ads)=25.6 mol(-1) dm(-3) and k(1)(max)=9.17x10(-7) mol m(-2) s(-1) and K(2)(ads)=37.1x10(3) mol(-1) dm(-3) and k(2)(max)=62.3x10(-7) mol m(-2) s(-1), respectively. In the second stage, Langmuir-Hinshelwood kinetics also describes the dissolution of iron and nickel from nickel ferrite, with K(1)(ads)=20.8 mol(-1) dm(3) and K(2)(ads)=1.16x10(5) mol(-1) dm(3). For iron, k(1)(max)=1.02x10(-7) mol of Fe m(-2) s(-1) and k(2)(max)=2.38x10(-7) mol of Fe m(-2) s(-1); for nickel, the rate constants k(1)(max) and k(2)(max) are 2.4 and 1.79 times smaller, respectively. The factor 1.79 agrees nicely with the stoichiometric ratio, whereas the factor 2.4 implies the accumulation of some nickel in the residual particles. The rate of nickel dissolution in oxalic acid is higher than that in bunsenite by a factor of 8, whereas hematite is more reactive by a factor of 9 (in the absence of Fe(II)) and 27 (in the presence of Fe (II)). It may be concluded that oxalic acid operates to dissolve iron, and the ensuing disruption of the solid framework accelerates the release of nickel. Copyright 2000 Academic Press.

Fast Oxidation Processes in a Naturally Reduced Aquifer Zone Caused by Dissolved Oxygen

NASA Astrophysics Data System (ADS)

Davis, J. A.; Jemison, N. E.; Williams, K. H.; Hobson, C.; Bush, R. P.

2014-12-01

The occurrence of naturally reduced zones is quite common in alluvial aquifers in the western U.S.A. due to the burial of woody debris in flood plains. The naturally reduced zones are heterogeneously dispersed in such aquifers and are characterized by high concentrations of organic carbon and reduced phases, including iron sulfides and reduced forms of metals, including uranium(IV). The persistence of high concentrations of dissolved uranium(VI) at uranium-contaminated aquifers on the Colorado Plateau has been attributed to slow oxidation of insoluble uranium(IV) mineral phases that are found in association with these natural reducing zones, although there is little understanding of the relative importance of various potential oxidants. Three field experiments were conducted within an alluvial aquifer adjacent to the Colorado River near Rifle, CO wherein groundwater associated with naturally reduced zones was pumped into a gas-impermeable tank, mixed with a conservative tracer (Br-), bubbled with a gas phase composed of 97% O2 and 3% CO2, and then returned to the subsurface in the same well from which it was withdrawn. Within minutes of re-injection of the oxygenated groundwater, dissolved uranium(VI) concentrations increased from less than 1 μM to greater than 2.5 μM, demonstrating that oxygen can be an important oxidant for uranium in these field systems if supplied to the naturally reduced zones. Small concentrations of nitrate were also observed in the previously nitrate-free groundwater, and Fe(II) decreased to the detection limit. These results contrast with other laboratory and field results in which oxygen was introduced to systems containing high concentrations of mackinawite (FeS) rather than the more crystalline iron sulfides found in aged, naturally reduced zones. The flux of oxygen to the naturally reduced zones in the alluvial aquifers occurs mainly through interactions between groundwater and gas phases at the water table, and seasonal variations of the water table at the Rifle, CO site may play an important role in introducing oxygen into the system. Although oxygen was introduced directly to the naturally reduced zones in these experiments, delivery of oxidants to the system may normally be controlled by other oxidative pathways in which oxygen plays an indirect role.

Investigation of molybdate melts as an alternative method of reprocessing used nuclear fuel

DOE PAGES

Hames, Amber L.; Tkac, Peter; Paulenova, Alena; ...

2017-01-17

Here, an investigation of molybdate melts containing sodium molybdate (Na 2MoO 4) and molybdenum trioxide (MoO 3) to achieve the separation of uranium from fission products by crystallization has been performed. The separation is based on the difference in solubility of the fission product metal oxides compared to the uranium oxide or molybdate in the molybdate melt. The molybdate melt dissolves uranium dioxide at high temperatures, and upon cooling, uranium precipitates as uranium dioxide or molybdate, whereas the fission product metals remain soluble in the melt. Small-scale experiments using gram quantities of uranium dioxide have been performed to investigate themore » feasibility of UO 2 purification from the fission products. The composition of the uranium precipitate as well as data for partitioning of several fission product surrogates between the uranium precipitate and molybdate melt for various melt compositions are presented and discussed. The fission products Cs, Sr, Ru and Rh all displayed very large distribution ratios. The fission products Zr, Pd, and the lanthanides also displayed good distribution ratios (D > 10). A melt consisting of 20 wt% MoO 3-50 wt% Na 2MoO 4-30 wt% UO 2 heated to 1313 K and cooled to 1123 K for the physical separation of the UO 2 product from the melt, and washed once with Na 2MoO 4 displays optimum conditions for separation of the UO 2 from the fission products.« less

Ionic Liquids as templating agents in formation of uranium-containing nanomaterials

DOEpatents

Visser, Ann E; Bridges, Nicholas J

2014-06-10

A method for forming nanoparticles containing uranium oxide is described. The method includes combining a uranium-containing feedstock with an ionic liquid to form a mixture and holding the mixture at an elevated temperature for a period of time to form the product nanoparticles. The method can be carried out at low temperatures, for instance less than about 300.degree. C.

On the Use of Thermal NF3 as the Fluorination and Oxidation Agent in Treatment of Used Nuclear Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scheele, Randall D.; McNamara, Bruce K.; Casella, Andrew M.

2012-05-01

This paper presents results of our investigation on the use of nitrogen trifluoride as the fluorination or fluorination/oxidation agent for use in a process for separating valuable constituents from used nuclear fuels by employing the volatility of many transition metal and actinide fluorides. Nitrogen trifluoride is less chemically and reactively hazardous than the hazardous and aggressive fluorinating agents used to prepare uranium hexafluoride and considered for fluoride volatility based nuclear fuels reprocessing. In addition, nitrogen trifluoride’s less aggressive character may be used to separate the volatile fluorides from used fuel and from themselves based on the fluorination reaction’s temperature sensitivitymore » (thermal tunability) rather than relying on differences in sublimation/boiling temperature and sorbents. Our thermodynamic calculations found that nitrogen trifluoride has the potential to produce volatile fission product and actinide fluorides from candidate oxides and metals. Our simultaneous thermogravimetric and differential thermal analyses found that the oxides of lanthanum, cerium, rhodium, and plutonium fluorinated but did not form volatile fluorides and that depending on temperature volatile fluorides formed from the oxides of niobium, molybdenum, ruthenium, tellurium, uranium, and neptunium. We also demonstrated near-quantitative removal of uranium from plutonium in a mixed oxide.« less

Tuning the Oxidation State, Nuclearity, and Chemistry of Uranium Hydrides with Phenylsilane and Temperature: The Case of the Classic Uranium(III) Hydride Complex [(C 5 Me 5) 2U(μ-H)] 2

DOE PAGES

Pagano, Justin K.; Dorhout, Jacquelyn M.; Czerwinski, Kenneth R.; ...

2016-03-18

Here, this work demonstrates that the oxidation state and chemistry of uranium hydrides can be tuned with temperature and the stoichiometry of phenylsilane. The trivalent uranium hydride [(C 5Me 5) 2U–H] x (5) was found to be comprised of an equilibrium mixture of U(III) hydrides in solution at ambient temperature. A single U(III) species can be selectively prepared by treating (C 5Me5)2UMe2 (4) with 2 equiv of phenylsilane at 50 °C. The U(III) system is a potent reducing agent and displayed chemistry distinct from the U(IV) system [(C 5Me 5) 2U(H)(μ-H)] 2 (2), which was harnessed to prepare a varietymore » of organometallic complexes, including (C 5Me 5) 2U(dmpe)(H) (6), and the novel uranium(IV) metallacyclopentadiene complex (C 5Me 5) 2U(C 4Me 4) (11).« less

High-Quality Draft Genome Sequence of Desulfovibrio carbinoliphilus FW-101-2B, an Organic Acid-Oxidizing Sulfate-Reducing Bacterium Isolated from Uranium(VI)-Contaminated Groundwater

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ramsay, Bradley D.; Hwang, Chiachi; Woo, Hannah L.

2015-03-12

Desulfovibrio carbinoliphilus subsp. oakridgensis FW-101-2B is an anaerobic, organic acid/alcohol-oxidizing, sulfate-reducing δ-proteobacterium. FW-101-2B was isolated from contaminated groundwater at The Field Research Center at Oak Ridge National Lab after in situ stimulation for heavy metal-reducing conditions. The genome will help elucidate the metabolic potential of sulfate-reducing bacteria during uranium reduction.

Preliminary report on the White Canyon area, San Juan county, Utah

USGS Publications Warehouse

Benson, William E.; Trites, Albert F.; Beroni, Ernest P.; Feeger, John A.

1952-01-01

The White Canyon area, in the central part of San Juan County, Utah, consists of approximately two 15-minute quadrangles. Approximately 75 square miles have been mapped by the Geological Survey on a scale of 1 inch equals 1 mile, using a combined aerial photography-plane table method. Structure contours were drawn on top of the Organ Rock member of the Cutler formation. Parts of the Gonway and North Point claims, 1/4 mile east of the Happy Jack mine, were mapped in detail. The principal objectives of the investigations were: (1) to establish ore guides; (2) to select areas favorable for exploration; and (3) to map the general geology and to determine the regional relationships of the uranium deposits. The White Canyon area is comprised of sedimentary rocks of Carboniferous to Jurassic age, more than 2,000 feet thick, having a regional dip of 1° to 2° SW. The nearest igneous rocks are in the Henry Mountains about 7 miles west of the northern part of the area; The Shinarump conglomerate of the late Triassic age, the principal ore horizon in the White Canyon area, consists of lenticular beds of sandstone, conglomeratic sandstone, conglomerate, clay, and siltstone. The Shinarump conglomerate, absent in places, is as much as 75 feet thick. The sandstones locally contain molds of logs and fragments of altered volcanic ash. Some of the logs have been replaced by copper and uranium minerals and iron oxides. The clay and siltstone underlie and are interbedded with the sandstone, and are most common in channels that cut into the underlying Moenkopi formation. The Shinarump conglomerate contains reworked Moenkopi siltstone fragments, clay balls, carbonized wood, and pebbles of quarts, quartzite, and chert. Jointing is prominent in the Western part of the mapped area. The three most prominent joint trends are due east, N. 65°-75° W., and N. 65°-75° E. All joints have vertical dips. The red beds are bleached along some joints, especially those that trend N. 65°-75° W. All uranium ore produced has been from the lower part of the Shinarump conglomerate, where it commonly occurs with copper as disseminations and fracture coatings in sandstone. Uranium and copper minerals also occur in low-grade disseminated deposits in the lower Chinle and in the Moenkopi formation and in veins cutting these formations. Although some uranium deposits occur in Chinarump channels and scours, copper and uranium minerals along fractures suggest that channel control may be secondary. Logs and clay balls apparently have exerted some chemical influences for deposition. The uranium occurs as the oxide in some deposits, and as secondary hydrous sulfates, phosphates, oxides, and silicates in these and several other deposits. Charcoal, iron and manganese oxides, and veinlets of hydrocarbon are abnormally radioactive in most of the deposits. Base-metal sulfides are commonly found inside the oxidized zone. Secondary copper minerals include the hydrous sulfates and carbonate. Gangue minerals include quarts, clay minerals, and manganese oxides, dickite (?), calcite, gypsum, pyrite, and chalcedony (?). Principal wall-rock alteration appears to have been silicification, clay alteration, and bleaching. Most of the shipped ore has contained more than 0.3 percent uranium. The ore also contains copper, commonly in grades lower than 1.0 percent. Criteria believed to be most useful for prospecting for concealed uranium deposits are (1) visible uranium minerals; (2) sulfide minerals; (3) secondary copper minerals; (4) dickite (?); (5) hydrocarbons; and (6) bleaching and alteration of the Moenkopi formation.

Reconnaissance for radioactive materials in the southern part of Brazil

USGS Publications Warehouse

Pierson, Charles T.; Haynes, Donald D.; Filho, Evaristo Ribeiro

1957-01-01

During 1954-1956 a reconnaissance for radioactive minerals was made with carborne, airborne and handborne scintillation equipment in the southern Brazilian states of Rio de Janeiro, Sao Paulo, Parana, Santa Catarina and Rio Grande do Sul. During the traverse covering more than 5,000 kilometers the authors checked the radioactivity of Precambrian igneous and metamorphic rocks, Paleozoic, Mesozoic and Cenozoic sedimentary rocks, and Mesozoic alkalic intrusive and basaltic extrusive rocks. The 22 samples collected contained from 0.003 to 0.029 percent equivalent uranium oxide and from 0.10 to 0.91 percent equivalent thorimn; two samples were taken from radioactive pegmati tes for mineralogic studies. None of the localities is at present a commercial source of uranium or thorium; however, additional work should be done near the alkalic stock at Lages in the State of Santa Catarina and at the Passo das Tropas fossil plant locality near Santa Maria in the state of Rio Grande do Sul. Near Lages highly altered alkalic rock from a dike contained 0.026 percent uranium oxide. At Passo das Tropas highly altered, limonite-impregnated sandstone from the Rio do Rasto group of sedimentary rocks contained 0.029 percent uranium oxide.

Processing of uranium oxide and silicon carbide based fuel using polymer infiltration and pyrolysis

NASA Astrophysics Data System (ADS)

Singh, Abhishek K.; Zunjarrao, Suraj C.; Singh, Raman P.

2008-09-01

Ceramic composite pellets consisting of uranium oxide, UO 2, contained within a silicon carbide matrix, were fabricated using a novel processing technique based on polymer infiltration and pyrolysis (PIP). In this process, particles of depleted uranium oxide, in the form of U 3O 8, were dispersed in liquid allylhydridopolycarbosilane (AHPCS), and subjected to pyrolysis up to 900 °C under a continuous flow of ultra high purity argon. The pyrolysis of AHPCS, at these temperatures, produced near-stoichiometric amorphous silicon carbide ( a-SiC). Multiple polymer infiltration and pyrolysis (PIP) cycles were performed to minimize open porosity and densify the silicon carbide matrix. Analytical characterization was conducted to investigate chemical interaction between U 3O 8 and SiC. It was observed that U 3O 8 reacted with AHPCS during the very first pyrolysis cycle, and was converted to UO 2. As a result, final composition of the material consisted of UO 2 particles contained in an a-SiC matrix. The physical and mechanical properties were also quantified. It is shown that this processing scheme promotes uniform distribution of uranium fuel source along with a high ceramic yield of the parent matrix.

A procedural manual for measurement of uranium and thorium isotopes utilizing the USGS-Stanford Finnegan Mat 262

USGS Publications Warehouse

Shamp, Donald D.

2001-01-01

Over the past several decades investigators have extensively examined the 238U-234U- 230Th systematics of a variety of geologic materials using alpha spectroscopy. Analytical uncertainty for 230Th by alpha spectroscopy has been limited to about 2% (2σ). The advantage of thermal ionization mass spectroscopy (TIMS), introduced by Edwards and co-workers in the late 1980’s is the increased detectability of these isotopes by a factor of ~200, and decreases in the uncertainty for 230Th to about 5‰ (2σ) error. This report is a procedural manual for using the USGS-Stanford Finnegan-Mat 262 TIMS to collect and isolate Uranium and Thorium isotopic ratio data. Chemical separation of Uranium and Thorium from the sample media is accomplished using acid dissolution and then processed using anion exchange resins. The Finnegan-Mat262 Thermal Ionization Mass Spectrometer (TIMS) utilizes a surface ionization technique in which nitrates of Uranium and Thorium are placed on a source filament. Upon heating, positive ion emission occurs. The ions are then accelerated and focused into a beam which passes through a curved magnetic field dispersing the ions by mass. Faraday cups and/or an ion counter capture the ions and allow for quantitative analysis of the various isotopes.

Dissolution of a metal oxide film during titanium carbide synthesis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bloshenko, V.N.; Bokii, V.A.; Borovinskaya, I.P.

1985-05-01

Oxygen is most difficult to remove during combustion of the mixture Ti + C. Its fundamental mass is in two states in the initial charge: part of the oxygen is dissolved in the titanium particles; the rest is bound in the metal oxide film (an insignificant part of the oxygen is in the adsorbed state in the carbon and titanium particles). On the basis of the results of vacuum annealing of specimens from a Ti + C mixture, the possibility is shown in this paper for dissolution of the intrinsic oxide film by titanium particles during residency of these particlesmore » in the heating zone of the combustion wave.« less

Dissolution Flowsheet for High Flux Isotope Reactor Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daniel, W. E.; Rudisill, T. S.; O'Rourke, P. E.

2016-09-27

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U 3O 8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H 2. The HFIR fuel cores will be dissolved and the recovered U will be down-blendedmore » into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H 2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy, and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H 2 and other permanent gases in the dissolution offgas, allowing the development of H 2 generation rate profiles. The H 2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the calculated lower flammability limit (LFL) for H 2 at a given Hg concentration. Complete dissolution of the Al 1100 and Al 6061 T6 alloys up to a final Al concentration of 2 M was obtained using a 7 M HNO 3 solution containing a 0.002 M Hg catalyst. However, following the dissolutions, solids were observed in the solution. The solids were amorphous, but likely originated from the Si present in the alloys. No crystalline materials, such as Al(NO 3) 3 were observed. During the course of the dissolution experiments, it was determined that delaying the addition of Hg once the HNO 3 solution reached the boiling point can reduce the total offgas and H 2 generation rates. The delay in starting the Hg addition is not necessary for HFIR fuel dissolution, but could be useful in other research reactor dissolution campaigns. The potential to generate flammable concentrations of H 2 in the offgas during a HFIR fuel dissolution was evaluated using the experimental data. The predicted H 2 concentration in the dissolver offgas stream was compared with 60% of the calculated H 2 LFL at 200 °C using several prototypical experiments. The calculations showed that a full HFIR core can be dissolved using nominally 0.002 M Hg to catalyze the dissolution. The margin between the predicted H 2 concentration and the calculated LFL was greater when the solution was allowed to boil for 45 min prior to initiating the Hg addition. When the Hg was increased to 0.004 M, the predicted H 2 concentration exceeded the calculated LFL early in the dissolution. The dissolution experiments also demonstrated that additional Hg (beyond the initial 0.002 M) could be added as the Al concentration increases. The ability to add more Hg during a HFIR fuel dissolution could be beneficial if slow dissolution rates are observed at high Al concentrations. Experimental data were used to demonstrate that the predicted H 2 concentration in a dissolver was below 60% of the calculated LFL at 200 °C when 0.004 M Hg was used to catalyze the dissolution if the Al concentration is conservatively greater than 0.5 M. Data also show that the Hg concentration during a HFIR fuel dissolution can be increased from 0.002 to 0.008 M at an Al concentration of 1.3 M.« less

4. VIEW OF ROOM 103 IN 1980. SIX OF THE ...

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

4. VIEW OF ROOM 103 IN 1980. SIX OF THE NINE URANIUM NITRATE STORAGE TANKS ARE SHOWN. HIGHLY ENRICHED URANIUM WAS INTRODUCED INTO THE BUILDING IN THE SUMMER OF 1965 AND THE FIRST EXPERIMENTS WERE PERFORMED IN SEPTEMBER OF 1965. EXPERIMENTS WERE PERFORMED ON ENRICHED URANIUM METAL AND SOLUTION, PLUTONIUM METAL, LOW ENRICHED URANIUM OXIDE, AND SEVERAL SPECIAL APPLICATIONS. AFTER 1983, EXPERIMENTS WERE CONDUCTED PRIMARILY WITH URANYL NITRATE SOLUTIONS, AND DID NOT INVOLVE SOLID MATERIALS. - Rocky Flats Plant, Critical Mass Laboratory, Intersection of Central Avenue & 86 Drive, Golden, Jefferson County, CO

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.

Method of Making Uranium Dioxide Bodies

DOEpatents

Wilhelm, H. A.; McClusky, J. K.

1973-09-25

Sintered uranium dioxide bodies having controlled density are produced from U.sub.3 O.sub.8 and carbon by varying the mole ratio of carbon to U.sub.3 O.sub.8 in the mixture, which is compressed and sintered in a neutral or slightly oxidizing atmosphere to form dense slightly hyperstoichiometric uranium dioxide bodies. If the bodies are to be used as nuclear reactor fuel, they are subsequently heated in a hydrogen atmosphere to achieve stoichiometry. This method can also be used to produce fuel elements of uranium dioxide -- plutonium dioxide having controlled density.

RECOVERY OF URANIUM VALUES FROM URANIUM BEARING RAW MATERIALS

DOEpatents

Michal, E.J.; Porter, R.R.

1959-06-16

Uranium leaching from ground uranium-bearing raw materials using MnO/sub 2/ in H/sub 2/SO/sub 4/ is described. The MnO/sub 2/ oxidizes U to the leachable hexavalent state. The MnO/sub 2/ does not replace Fe normally added, because the Fe complexes P and catalyzes the MnO/sub 2/ reaction. Three examples of continuous processes are given, but batch operation is also possible. The use of MnO/sub 2/ makes possible recovery of very low U values. (T.R.H.)

Electrolytic process for preparing uranium metal

DOEpatents

Haas, Paul A.

1990-01-01

An electrolytic process for making uranium from uranium oxide using Cl.sub.2 anode product from an electrolytic cell to react with UO.sub.2 to form uranium chlorides. The chlorides are used in low concentrations in a melt comprising fluorides and chlorides of potassium, sodium and barium in the electrolytic cell. The electrolysis produces Cl.sub.2 at the anode that reacts with UO.sub.2 in the feed reactor to form soluble UCl.sub.4, available for a continuous process in the electrolytic cell, rather than having insoluble UO.sub.2 fouling the cell.

A study on recovery of uranium in the anode basket residues delivered from the pyrochemical process of used nuclear fuel

NASA Astrophysics Data System (ADS)

Eun, H. C.; Kim, T. J.; Jang, J. H.; Kim, G. Y.; Park, S. B.; Yoon, D. S.; Kim, S. H.; Paek, S. W.; Lee, S. J.

2018-04-01

In this study, the chlorination of uranium oxide (UO2) using ammonium chloride and zirconium as chemical agents was conducted to recover the uranium in the anode basket residues from the pyrochemical process of used nuclear fuel. The chlorination of UO2 was predicted using thermodynamic equilibrium calculations. The experimental conditions for the chlorination were determined using a chlorination test with cerium oxide (CeO2). In the chlorination test, it was confirmed that UO2 was chlorinated into UCl3 at 320 °C, some UO2 remained without changes in the chemical form, and ZrO2, Zr2O, and ZrCl2 were generated as byproducts.

Continuous process electrorefiner

DOEpatents

Herceg, Joseph E [Naperville, IL; Saiveau, James G [Hickory Hills, IL; Krajtl, Lubomir [Woodridge, IL

2006-08-29

A new device is provided for the electrorefining of uranium in spent metallic nuclear fuels by the separation of unreacted zirconium, noble metal fission products, transuranic elements, and uranium from spent fuel rods. The process comprises an electrorefiner cell. The cell includes a drum-shaped cathode horizontally immersed about half-way into an electrolyte salt bath. A conveyor belt comprising segmented perforated metal plates transports spent fuel into the salt bath. The anode comprises the conveyor belt, the containment vessel, and the spent fuel. Uranium and transuranic elements such as plutonium (Pu) are oxidized at the anode, and, subsequently, the uranium is reduced to uranium metal at the cathode. A mechanical cutter above the surface of the salt bath removes the deposited uranium metal from the cathode.

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.

Microbially catalyzed nitrate-dependent metal/radionuclide oxidation in shallow subsurface sediments

NASA Astrophysics Data System (ADS)

Weber, K.; Healy, O.; Spanbauer, T. L.; Snow, D. D.

2011-12-01

Anaerobic, microbially catalyzed nitrate-dependent metal/radionuclide oxidation has been demonstrated in a variety of sediments, soils, and groundwater. To date, studies evaluating U bio-oxidation and mobilization have primarily focused on anthropogenically U contaminated sites. In the Platte River Basin U originating from weathering of uranium-rich igneous rocks in the Rocky Mountains was deposited in shallow alluvial sediments as insoluble reduced uranium minerals. These reduced U minerals are subject to reoxidation by available oxidants, such nitrate, in situ. Soluble uranium (U) from natural sources is a recognized contaminant in public water supplies throughout the state of Nebraska and Colorado. Here we evaluate the potential of anaerobic, nitrate-dependent microbially catalyzed metal/radionuclide oxidation in subsurface sediments near Alda, NE. Subsurface sediments and groundwater (20-64ft.) were collected from a shallow aquifer containing nitrate (from fertilizer) and natural iron and uranium. The reduction potential revealed a reduced environment and was confirmed by the presence of Fe(II) and U(IV) in sediments. Although sediments were reduced, nitrate persisted in the groundwater. Nitrate concentrations decreased, 38 mg/L to 30 mg/L, with increasing concentrations of Fe(II) and U(IV). Dissolved U, primarily as U(VI), increased with depth, 30.3 μg/L to 302 μg/L. Analysis of sequentially extracted U(VI) and U(IV) revealed that virtually all U in sediments existed as U(IV). The presence of U(IV) is consistent with reduced Fe (Fe(II)) and low reduction potential. The increase in aqueous U concentrations with depth suggests active U cycling may occur at this site. Tetravalent U (U(IV)) phases are stable in reduced environments, however the input of an oxidant such as oxygen or nitrate into these systems would result in oxidation. Thus co-occurrence of nitrate suggests that nitrate could be used by bacteria as a U(IV) oxidant. Most probable number enumeration of nitrate-dependent U(IV) oxidizing microorganisms demonstrated an abundant community ranging from 1.61x104 to 2.74x104 cells g-1 sediment. Enrichments initiated verified microbial U reduction and U oxidation coupled to nitrate reduction. Sediment slurries were serially diluted and incubated over a period of eight weeks and compared to uninoculated controls. Oxidation (0-4,554 μg/L) and reduction (0-55 μg/L) of U exceeded uninoculated controls further providing evidence of a U biogeochemical cycling in these subsurface sediments. The oxidation of U(IV) could contribute to U mobilization in the groundwater and result in decreased water quality. Not only could nitrate serve as an oxidant, but Fe(III) could also contribute to U mobilization. Nitrate-dependent Fe(II) oxidation is an environmentally ubiquitous process facilitated by a diversity of microorganisms. Additional research is necessary in order to establish a role of biogenic Fe(III) oxides in U geochemical cycling at this site. These microbially mediated processes could also have a confounding effect on uranium mobility in subsurface environments.

URANIUM EXTRACTION PROCESS

DOEpatents

Baldwin, W.H.; Higgins, C.E.

1958-12-16

A process is described for recovering uranium values from acidic aqueous solutions containing hexavalent uranium by contacting the solution with an organic solution comprised of a substantially water-immiscible organlc diluent and an organic phosphate to extract the uranlum values into the organic phase. Carbon tetrachloride and a petroleum hydrocarbon fraction, such as kerosene, are sultable diluents to be used in combination with organlc phosphates such as dibutyl butylphosphonate, trlbutyl phosphine oxide, and tributyl phosphate.

Tags to Track Illicit Uranium and Plutonium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haire, M. Jonathan; Forsberg, Charles W.

2007-07-01

With the expansion of nuclear power, it is essential to avoid nuclear materials from falling into the hands of rogue nations, terrorists, and other opportunists. This paper examines the idea of detection and attribution tags for nuclear materials. For a detection tag, it is proposed to add small amounts [about one part per billion (ppb)] of {sup 232}U to enriched uranium to brighten its radioactive signature. Enriched uranium would then be as detectable as plutonium and thus increase the likelihood of intercepting illicit enriched uranium. The use of rare earth oxide elements is proposed as a new type of 'attribution'more » tag for uranium and thorium from mills, uranium and plutonium fuels, and other nuclear materials. Rare earth oxides are chosen because they are chemically compatible with the fuel cycle, can survive high-temperature processing operations in fuel fabrication, and can be chosen to have minimal neutronic impact within the nuclear reactor core. The mixture of rare earths and/or rare earth isotopes provides a unique 'bar code' for each tag. If illicit nuclear materials are recovered, the attribution tag can identify the source and lot of nuclear material, and thus help police reduce the possible number of suspects in the diversion of nuclear materials based on who had access. (authors)« less

PRODUCTION OF URANIUM HEXAFLUORIDE

DOEpatents

Fowler, R.D.

1957-08-27

A process for the production of uranium hexafluoride from the oxides of uranium is reported. In accordance with the method, the higher oxides of uranium may be reduced to uranium dioxide (UO/sub 2/), the latter converted into uranium tetrafluoride by reaction with hydrogen fluoride, and the UF/sub 4/ converted to UF/sub 6/ by reaction with a fluorinating agent, such as CoF/sub 3/. The UO/sub 3/ or U/sub 3/O/sub 8/ is placed in a reac tion chamber in a copper boat or tray enclosed in a copper oven, and heated to 500 to 650 deg C while hydrogen gas is passed through the oven. After nitrogen gas is used to sweep out the hydrogen and the water vapor formed, and while continuing to inaintain the temperature between 400 deg C and 600 deg C, anhydrous hydrogen fluoride is passed through. After completion of the conversion of UO/sub 2/ to UF/sub 4/ the temperature of the reaction chamber is lowered to about 400 deg C or less, the UF/sub 4/ is mixed with the requisite quantity of CoF/sub 3/, and after evacuating the chamber, the mixture is heated to 300 to 400 deg C, and the resulting UF/sub 6/ is led off and delivered to a condenser.

DOUBLE-BAKED, SELF-CHANNELLING ELECTRODE

DOEpatents

Piper, R.D.; Leifield, R.F.

1963-03-12

A method is given for making an electrode for use in the electrolytic reduction of uranium oxides to uranium metal in a fused salt electrolyte. Uranlum oxide such as UO/sub 2/ is mixed with somewhat less than the stoichiometric amount of carbon needed for the reduction, and the mixture is baked and crushed to make a nonspherical material. The latter is then mixed with a carbon binder sufficient to satisfy stoichiometry, pressed into a shape such as a cylinder, and baked. (AEC)

Dissolution mechanism of aluminum hydroxides in acid media

NASA Astrophysics Data System (ADS)

Lainer, Yu. A.; Gorichev, I. G.; Tuzhilin, A. S.; Gololobova, E. G.

2008-08-01

The effects of the concentration, temperature, and potential at the hydroxide/electrolyte interface on the aluminum hydroxide dissolution in sulfuric, hydrochloric, and perchloric acids are studied. The limiting stage of the aluminum hydroxide dissolution in the acids is found to be the transition of the complexes that form on the aluminum hydroxide surface from the solid phase into the solution. The results of the calculation of the acid-base equilibrium constants at the oxide (hydroxide)/solution interface using the experimental data on the potentiometric titration of Al2O3 and AlOOH suspensions are analyzed. A mechanism is proposed for the dissolution of aluminum hydroxides in acid media.

Oxygen migration enthalpy likely limits oxide precipitate dissolution during tabula rasa

NASA Astrophysics Data System (ADS)

Looney, E. E.; Laine, H. S.; Youssef, A.; Jensen, M. A.; LaSalvia, V.; Stradins, P.; Buonassisi, T.

2017-09-01

In industrial silicon solar cells, oxygen-related defects lower device efficiencies by up to 20% (rel.). In order to mitigate these defects, a high-temperature homogenization anneal called tabula rasa (TR) that has been used in the electronics industry is now proposed for use in solar-grade wafers. This work addresses the kinetics of tabula rasa by elucidating the activation energy governing oxide precipitate dissolution, which is found to be 2.6 ± 0.5 eV. This value is consistent within uncertainty to the migration enthalpy of oxygen interstitials in silicon, implying TR to be kinetically limited by oxygen point-defect diffusion. This large activation energy is observed to limit oxygen precipitate dissolution during standard TR conditions, suggesting that more aggressive annealing conditions than conventionally used are required for complete bulk microdefect mitigation.

Oxygen migration enthalpy likely limits oxide precipitate dissolution during tabula rasa

DOE Office of Scientific and Technical Information (OSTI.GOV)

Looney, E. E.; Laine, H. S.; Youssef, A.

In industrial silicon solar cells, oxygen-related defects lower device efficiencies by up to 20% (rel.). In order to mitigate these defects, a high-temperature homogenization anneal called tabula rasa (TR) that has been used in the electronics industry is now proposed for use in solar-grade wafers. This work addresses the kinetics of tabula rasa by elucidating the activation energy governing oxide precipitate dissolution, which is found to be 2.6 +/- 0.5 eV. This value is consistent within uncertainty to the migration enthalpy of oxygen interstitials in silicon, implying TR to be kinetically limited by oxygen point-defect diffusion. This large activation energymore » is observed to limit oxygen precipitate dissolution during standard TR conditions, suggesting that more aggressive annealing conditions than conventionally used are required for complete bulk microdefect mitigation.« less

Determination of the oxidation states of metals and metalloids: An analytical review

NASA Astrophysics Data System (ADS)

Vodyanitskii, Yu. N.

2013-12-01

The hazard of many heavy metals/metalloids in the soil depends on their oxidation state. The problem of determining the oxidation state has been solved due to the use of synchrotron radiation methods with the analysis of the X-ray absorption near-edge structure (XANES). The determination of the oxidation state is of special importance for some hazardous heavy elements (arsenic, antimony, selenium, chromium, uranium, and vanadium). The mobility and hazard of each of these elements depend on its oxidation state. The mobilities are higher at lower oxidation states of As, Cr, V, and Se and at higher oxidation states of Sb and U. The determination of the oxidation state of arsenic has allowed revealing its fixation features in the rhizosphere of hydrophytes. The known oxidation states of chromium and uranium are used for the retention of these elements on geochemical barriers. Different oxidation states have been established for vanadium displacing iron in goethite. The determination of the oxidation state of manganese in the rhizosphere and the photosynthetic apparatus of plants is of special importance for agricultural chemists.

Top Ten Reasons for DEOX as a Front End to Pyroprocessing

DOE Office of Scientific and Technical Information (OSTI.GOV)

B.R. Westphal; K.J. Bateman; S.D. Herrmann

A front end step is being considered to augment chopping during the treatment of spent oxide fuel by pyroprocessing. The front end step, termed DEOX for its emphasis on decladding via oxidation, employs high temperatures to promote the oxidation of UO2 to U3O8 via an oxygen carrier gas. During oxidation, the spent fuel experiences a 30% increase in lattice structure volume resulting in the separation of fuel from cladding with a reduced particle size. A potential added benefit of DEOX is the removal of fission products, either via direct release from the broken fuel structure or via oxidation and volatilizationmore » by the high temperature process. Fuel element chopping is the baseline operation to prepare spent oxide fuel for an electrolytic reduction step. Typical chopping lengths range from 1 to 5 mm for both individual elements and entire assemblies. During electrolytic reduction, uranium oxide is reduced to metallic uranium via a lithium molten salt. An electrorefining step is then performed to separate a majority of the fission products from the recoverable uranium. Although DEOX is based on a low temperature oxidation cycle near 500oC, additional conditions have been tested to distinguish their effects on the process.[1] Both oxygen and air have been utilized during the oxidation portion followed by vacuum conditions to temperatures as high as 1200oC. In addition, the effects of cladding on fission product removal have also been investigated with released fuel to temperatures greater than 500oC.« less

Materials. Section 1 of Symposium on the peaceful uses of atomic energy in Australia, 1958, held in Sydney, in June 1958

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

The environments of the known uranium occurences in South Australia arc described, and the relation of uranium mineralization with sodic granitic rocks is emphasized. The problems in designing equipment for radiometric prospecting are reviewed. The fabrication and properties of BeO, UO/sub 2/, ThO/sub 2/, and mixed oxides are discussed. The use of pulsing in a uranium extraction pilot plant ion exchange column is described. The wetting of metals by liquid metals is reviewed with emphasis on liquid sodium. The geological nature, extent, and future prospects of minerals with atomic energy applications, occurring in New South Wales are outlined. The developmentmore » of a process for uranium recovery from Mary Kathleen ores is described. Techniques and processes involved in locating, mining, and concentrating davidite-type ores at Radium Hill, South Australia are described. The uranium deposits of the Northern Territory, Australia, are classified and described. The flotation behavior of the simple oxide minerals, uraninite and the colloform variety is discussed. The Port Pirie Treatment Plant for uranium recovery from refractory Radium Hill concentrates is described. The plant utilizes the sulfuric acid-ion exchange process. The uranium deposits of Queensland are described. the details of the production of uranium ore concentrates at Rum jungle near Darwin, Australia, are given. A brief account of the use of neutron diffraction analysis in crystallography is given, and the neutron spectrometers installed on the High Flux Australian Research Reactor are described. (T.R.H.)« less

The nucleation and growth of uranium on the basal plane of graphite studied by scanning tunneling microscopy

NASA Astrophysics Data System (ADS)

Tench, R. J.

1992-11-01

For the first time, nanometer scale uranium clusters were created on the basal plane of highly oriented pyrolytic graphite by laser ablation under ultra-high vacuum conditions. The physical and chemical properties of these clusters were investigated by scanning tunneling microscopy (STM) as well as standard surface science techniques. Auger electron and X-ray photoelectron spectroscopies found the uranium deposit to be free of contamination and showed that no carbide had formed with the underlying graphite. Clusters with sizes ranging from 42 to 630 sq A were observed upon initial room temperature deposition. Surface diffusion of uranium was observed after annealing the substrate above 800 K, as evidenced by the decreased number density and the increased size of the clusters. Preferential depletion of clusters on terraces near step edges as a result of annealing was observed. The activation energy for diffusion deduced from these measurements was found to be 15 Kcal/mole. Novel formation of ordered uranium thin films was observed for coverages greater than two monolayers after annealing above 900 K. These ordered films displayed islands with hexagonally faceted edges rising in uniform step heights characteristic of the unit cell of the P-phase of uranium. In addition, atomic resolution STM images of these ordered films indicated the formation of the (beta)-phase of uranium. The chemical properties of these surfaces were investigated and it was shown that these uranium films had a reduced oxidation rate in air as compared to bulk metal and that STM imaging in air induced a polarity-dependent enhancement of the oxidation rate.

In-situ evidence for uranium immobilization and remobilization

USGS Publications Warehouse

Senko, John M.; Istok, Jonathan D.; Suflita, Joseph M.; Krumholz, Lee R.

2002-01-01

The in-situ microbial reduction and immobilization of uranium was assessed as a means of preventing the migration of this element in the terrestrial subsurface. Uranium immobilization (putatively identified as reduction) and microbial respiratory activities were evaluated in the presence of exogenous electron donors and acceptors with field push−pull tests using wells installed in an anoxic aquifer contaminated with landfill leachate. Uranium(VI) amended at 1.5 μM was reduced to less than 1 nM in groundwater in less than 8 d during all field experiments. Amendments of 0.5 mM sulfate or 5 mM nitrate slowed U(VI) immobilization and allowed for the recovery of 10% and 54% of the injected element, respectively, as compared to 4% in the unamended treatment. Laboratory incubations confirmed the field tests and showed that the majority of the U(VI) immobilized was due to microbial reduction. In these tests, nitrate treatment (7.5 mM) inhibited U(VI) reduction, and nitrite was transiently produced. Further push−pull tests were performed in which either 1 or 5 mM nitrate was added with 1.0 μM U(VI) to sediments that already contained immobilized uranium. After an initial loss of the amendments, the concentration of soluble U(VI) increased and eventually exceeded the injected concentration, indicating that previously immobilized uranium was remobilized as nitrate was reduced. Laboratory experiments using heat-inactivated sediment slurries suggested that the intermediates of dissimilatory nitrate reduction (denitrification or dissimilatory nitrate reduction to ammonia), nitrite, nitrous oxide, and nitric oxide were all capable of oxidizing and mobilizing U(IV). These findings indicate that in-situ subsurface U(VI) immobilization can be expected to take place under anaerobic conditions, but the permanence of the approach can be impaired by disimilatory nitrate reduction intermediates that can mobilize previously reduced uranium.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Donald, Scott B.; Siekhaus, Wigbert J.; Nelson, Art J.

X-ray photoelectron spectroscopy in combination with secondary ion mass spectrometry depth profiling were used to investigate the surface and interfacial chemistry of C + ion implanted polycrystalline uranium subsequently oxidized in air for over 10 years at ambient temperature. The original implantation of 33 keV C + ions into U 238 with a dose of 4.3 × 10 17 cm –3 produced a physically and chemically modified surface layer that was characterized and shown to initially prevent air oxidation and corrosion of the uranium after 1 year in air at ambient temperature. The aging of the surface and interfacial layersmore » were examined by using the chemical shift of the U 4f, C 1s, and O 1s photoelectron lines. In addition, valence band spectra were used to explore the electronic structure of the aged carbide surface and interface layer. Moreover, the time-of-flight secondary ion mass spectrometry depth profiling results for the aged sample confirmed an oxidized uranium carbide layer over the carbide layer/U metal interface.« less

Recovery and recycling of uranium from rejected coated particles for compact high temperature reactors

NASA Astrophysics Data System (ADS)

Pai, Rajesh V.; Mollick, P. K.; Kumar, Ashok; Banerjee, J.; Radhakrishna, J.; Chakravartty, J. K.

2016-05-01

UO2 microspheres prepared by internal gelation technique were coated with pyrolytic carbon and silicon carbide using CVD technique. The particles which were not meeting the specifications were rejected. The rejected/failed UO2 based coated particles prepared by CVD technique was used for oxidation and recovery and recycling. The oxidation behaviour of sintered UO2 microspheres coated with different layers of carbon and SiC was studied by thermal techniques to develop a method for recycling and recovery of uranium from the failed/rejected coated particles. It was observed that the complete removal of outer carbon from the spheres is difficult. The crushing of microspheres enabled easier accessibility of oxygen and oxidation of carbon and uranium at 800-1000 °C. With the optimized process of multiple crushing using die & plunger and sieving the broken coated layers, we could recycle around fifty percent of the UO2 microspheres which could be directly recoated. The rest of the particles were recycled using a wet recycling method.

PEP Support: Laboratory Scale Leaching and Permeate Stability Tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Russell, Renee L.; Peterson, Reid A.; Rinehart, Donald E.

2010-05-21

This report documents results from a variety of activities requested by the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The activities related to caustic leaching, oxidative leaching, permeate precipitation behavior of waste as well as chromium (Cr) leaching are: • Model Input Boehmite Leaching Tests • Pretreatment Engineering Platform (PEP) Support Leaching Tests • PEP Parallel Leaching Tests • Precipitation Study Results • Cr Caustic and Oxidative Leaching Tests. Leaching test activities using the PEP simulant provided input to a boehmite dissolution model and determined the effect of temperature on mass loss during caustic leaching, the reaction rate constantmore » for the boehmite dissolution, and the effect of aeration in enhancing the chromium dissolution during caustic leaching. Other tests were performed in parallel with the PEP tests to support the development of scaling factors for caustic and oxidative leaching. Another study determined if precipitate formed in the wash solution after the caustic leach in the PEP. Finally, the leaching characteristics of different chromium compounds under different conditions were examined to determine the best one to use in further testing.« less

Anaerobic microbial dissolution of lead and production of organic acids

DOEpatents

Francis, A.J.; Dodge, C.; Chendrayan, K.; Quinby, H.L.

1987-04-16

The present invention related to an anaerobic bacterial culture of Clostridium sp. ATCC No. 53464 which solubilizes lead oxide under anaerobic conditions in coal and industrial wastes and therefore presents a method of removing lead from such wastes before they are dumped into the environment. The rat of lead dissolution during logarithmic growth of the bacteria in 40 ml medium containing 3.32 ..mu..moles of lead as lead oxide was 0.042 ..mu..moles m1/sup /-/1/ hr/sup /-/1/. Dissolution of lead oxide by the bacterial isolate is due to the production of metabolites and acidity in the culture medium. The major metabolites are acetic, butyric and lactic acid. The major metabolites are acetic, butyric and lactic acid. Clostridium sp. ATCC No. 53464 can be used in the recovery of the strategic metals from ores and wastes and also for the production of lactic acid for commercial purposes. The process yields large quantities of lactic acid as well as lead complexed in a stable form with said acids. 4 figs., 3 tabs.

Extraction of manganese from electrolytic manganese residue by bioleaching.

PubMed

Xin, Baoping; Chen, Bing; Duan, Ning; Zhou, Changbo

2011-01-01

Extraction of manganese from electrolytic manganese residues using bioleaching was investigated in this paper. The maximum extraction efficiency of Mn was 93% by sulfur-oxidizing bacteria at 4.0 g/l sulfur after bioleaching of 9days, while the maximum extraction efficiency of Mn was 81% by pyrite-leaching bacteria at 4.0 g/l pyrite. The series bioleaching first by sulfur-oxidizing bacteria and followed by pyrite-leaching bacteria evidently promoted the extraction of manganese, witnessing the maximum extraction efficiency of 98.1%. In the case of sulfur-oxidizing bacteria, the strong dissolution of bio-generated sulfuric acid resulted in extraction of soluble Mn2+, while both the Fe2+ catalyzed reduction of Mn4+ and weak acidic dissolution of Mn2+ accounted for the extraction of manganese with pyrite-leaching bacteria. The chemical simulation of bioleaching process further confirmed that the acid dissolution of Mn2+ and Fe2+ catalyzed reduction of Mn4+ were the bioleaching mechanisms involved for Mn extraction from electrolytic manganese residues. Copyright © 2010 Elsevier Ltd. All rights reserved.

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.

Redox behavior of uranium at the nanoporous aluminum oxide-water interface: implications for uranium remediation.

PubMed

Jung, Hun Bok; Boyanov, Maxim I; Konishi, Hiromi; Sun, Yubing; Mishra, Bhoopesh; Kemner, Kenneth M; Roden, Eric E; Xu, Huifang

2012-07-03

Sorption-desorption experiments show that the majority (ca. 80-90%) of U(VI) presorbed to mesoporous and nanoporous alumina could not be released by extended (2 week) extraction with 50 mM NaHCO(3) in contrast with non-nanoporous α alumina. The extent of reduction of U(VI) presorbed to aluminum oxides was semiquantitatively estimated by comparing the percentages of uranium desorbed by anoxic sodium bicarbonate between AH(2)DS-reacted and unreacted control samples. X-ray absorption spectroscopy confirmed that U(VI) presorbed to non-nanoporous alumina was rapidly and completely reduced to nanoparticulate uraninite by AH(2)DS, whereas reduction of U(VI) presorbed to nanoporous alumina was slow and incomplete (<5% reduction after 1 week). The observed nanopore size-dependent redox behavior of U has important implications in developing efficient remediation techniques for the subsurface uranium contamination because the efficiency of in situ bioremediation depends on how effectively and rapidly U(VI) bound to sediment or soil can be converted to an immobile phase.

Anaerobic U(IV) Bio-oxidation and the Resultant Remobilization of Uranium in Contaminated Sediments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coates, John D.

2005-06-01

A proposed strategy for the remediation of uranium (U) contaminated sites is based on immobilizing U by reducing the oxidized soluble U, U(VI), to form a reduced insoluble end product, U(IV). Due to the use of nitric acid in the processing of nuclear fuels, nitrate is often a co-contaminant found in many of the environments contaminated with uranium. Recent studies indicate that nitrate inhibits U(VI) reduction in sediment slurries. However, the mechanism responsible for the apparent inhibition of U(VI) reduction is unknown, i.e. preferential utilization of nitrate as an electron acceptor, direct biological oxidation of U(IV) coupled to nitrate reduction,more » and/or abiotic oxidation by intermediates of nitrate reduction. Recent studies indicates that direct biological oxidation of U(IV) coupled to nitrate reduction may exist in situ, however, to date no organisms have been identified that can grow by this metabolism. In an effort to evaluate the potential for nitrate-dependent bio-oxidation of U(IV) in anaerobic sedimentary environments, we have initiated the enumeration of nitrate-dependent U(IV) oxidizing bacteria. Sediments, soils, and groundwater from uranium (U) contaminated sites, including subsurface sediments from the NABIR Field Research Center (FRC), as well as uncontaminated sites, including subsurface sediments from the NABIR FRC and Longhorn Army Ammunition Plant, Texas, lake sediments, and agricultural field soil, sites served as the inoculum source. Enumeration of the nitrate-dependent U(IV) oxidizing microbial population in sedimentary environments by most probable number technique have revealed sedimentary microbial populations ranging from 9.3 x 101 - 2.4 x 103 cells (g sediment)-1 in both contaminated and uncontaminated sites. Interestingly uncontaminated subsurface sediments (NABIR FRC Background core FB618 and Longhorn Texas Core BH2-18) both harbored the most numerous nitrate-dependent U(IV) oxidizing population 2.4 x 103 cells (g sediment)-1. The nitrate-dependent U(IV) oxidizing microbial population in groundwaters is less numerous ranging from 0 cells mL-1 (Well FW300, Uncontaminated Background NABIR FRC) to 4.3 x 102 cells mL-1 (Well TPB16, Contaminated Area 2 NABIR FRC). The presence of nitrate-dependent U(IV) oxidizing bacteria supports our hypothesis that bacteria capable of anaerobic U(IV) oxidation are ubiquitous and indigenous to sedimentary and groundwater environments.« less

SINTERING METAL OXIDES

DOEpatents

Roake, W.E.

1960-09-13

A process is given for producing uranium dioxide material of great density by preparing a compacted mixture of uranium dioxide and from 1 to 3 wt.% of calcium hydride, heating the mixture to at least 675 deg C for decomposition of the hydride and then for sintering, preferably in a vacuum, at from 1550 to 2000 deg C. Calcium metal is formed, some uranium is reduced by the calcium to the metal and a product of high density is obtained.

Dynamic processes occurring at the Cr IIIaq-manganite (γ-MnOOH) interface: simultaneous adsorption, microprecipitation, oxidation/reduction, and dissolution

NASA Astrophysics Data System (ADS)

Weaver, Robert M.; Hochella, Michael F.; Ilton, Eugene S.

2002-12-01

The complex interaction between Cr IIIaq and manganite (γ-MnOOH) was systematically studied at room temperature over a pH range of 3 to 6, and within a concentration range of 10 -4 to 10 -2 M CrOH 2+aq. Solution compositional changes during batch reactions were characterized by inductively coupled plasma spectroscopy and ultraviolet-visible spectrophotometry. The manganites were characterized before and after reaction with X-ray photoelectron spectroscopy, scanning electron microscopy (SEM), high-resolution field-emission SEM, and energy-dispersive spectroscopy analysis. Fluid-cell atomic force microscopy was used to follow these metal-mineral interactions in situ. The reactions are characterized by (1) sorption of Cr III and the surface-catalyzed microprecipitation of Cr III-hydroxy hydrate on manganite surfaces, (2) the acidic dissolution of the manganite, and (3) the simultaneous reductive dissolution of manganite coupled with the oxidation of Cr IIIaq to highly toxic Cr VIaq. Cr III-hydroxy hydrate was shown to precipitate on the manganite surface while still undersaturated in bulk solution. The rate of manganite dissolution increased with decreasing pH due both to acid-promoted and Mn-reduction-promoted dissolution. Cr oxidation also increased in the lower pH range, this as a result of its direct redox coupling with Mn reduction. Neither Mn II nor Cr VI were ever detected on manganite surfaces, even at the maximum rate of their generation. At the highest pHs of this study, Cr IIIaq was effectively removed from solution to form Cr III-hydroxy hydrate on manganite surfaces and in the bulk solution, and manganite dissolution and Cr VIaq generation were minimized. All interface reactions described above were heterogeneous across the manganite surfaces. This heterogeneity is a direct result of the heterogeneous semiconducting nature of natural manganite crystals and is also an expression of the proximity effect, whereby redox processes on semiconducting surfaces are not limited to next nearest neighbor sites.

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

NASA Astrophysics Data System (ADS)

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

2011-09-01

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

Oxidative Uranium Release from Anoxic Sediments under Diffusion-Limited Conditions.

PubMed

Bone, Sharon E; Cahill, Melanie R; Jones, Morris E; Fendorf, Scott; Davis, James; Williams, Kenneth H; Bargar, John R

2017-10-03

Uranium (U) contamination occurs as a result of mining and ore processing; often in alluvial aquifers that contain organic-rich, reduced sediments that accumulate tetravalent U, U(IV). Uranium(IV) is sparingly soluble, but may be mobilized upon exposure to nitrate (NO 3 - ) and oxygen (O 2 ), which become elevated in groundwater due to seasonal fluctuations in the water table. The extent to which oxidative U mobilization can occur depends upon the transport properties of the sediments, the rate of U(IV) oxidation, and the availability of inorganic reductants and organic electron donors that consume oxidants. We investigated the processes governing U release upon exposure of reduced sediments to artificial groundwater containing O 2 or NO 3 - under diffusion-limited conditions. Little U was mobilized during the 85-day reaction, despite rapid diffusion of groundwater within the sediments and the presence of nonuraninite U(IV) species. The production of ferrous iron and sulfide in conjunction with rapid oxidant consumption suggested that the sediments harbored large concentrations of bioavailable organic carbon that fueled anaerobic microbial respiration and stabilized U(IV). Our results suggest that seasonal influxes of O 2 and NO 3 - may cause only localized mobilization of U without leading to export of U from the reducing sediments when ample organic carbon is present.

U(VI) adsorption on aquifer sediments at the Hanford Site.

PubMed

Um, Wooyong; Serne, R Jeffrey; Brown, Christopher F; Last, George V

2007-08-15

Aquifer sediments collected via split-spoon sampling in two new groundwater wells in the 200-UP-1 operable unit at the Hanford Site were characterized and showed typical Ringold Unit E Formation properties dominated by gravel and sand. High iron-oxide content in Fe oxide/clay coatings caused the highest U(VI) adsorption as quantified by batch K(d) values, indicating iron oxides are the key solid adsorbent in the 200-UP-1 sediments that affect U(VI) fate and mobility. Even though U(VI) adsorption on the gravel-sized fraction of the sediments is considered to be negligible, careful characterization should be conducted to determine U(VI) adsorption on gravel, because of presence of Fe oxides coatings and diffusion-controlled adsorption into the gravel particles' interior surfaces. A linear adsorption isotherm was observed up to 10(-6) M (238 microg/L) of total U(VI) concentration in batch U(VI) adsorption tests with varying total U(VI) concentrations in spiked groundwater. U(VI) adsorption decreased with increasing concentrations of dissolved carbonate, because strong anionic aqueous uranium-carbonate complexes formed at high pH and high alkalinity conditions. Noticeable uranium desorption hysteresis was observed in a flow-through column experiment, suggesting that desorption K(d) values for aged uranium-contaminated sediments at the Hanford Site can be larger than adsorption K(d) values determined in short-term laboratory experiments and slow uranium release from contaminated sediments into the groundwater is expected.

Influence of uranyl speciation and iron oxides on uranium biogeochemical redox reactions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stewart, B.D.; Amos, R.T.; Nico, P.S.

2010-03-15

Uranium is a pollutant of concern to both human and ecosystem health. Uranium's redox state often dictates its partitioning between the aqueous- and solid-phases, and thus controls its dissolved concentration and, coupled with groundwater flow, its migration within the environment. In anaerobic environments, the more oxidized and mobile form of uranium (UO{sub 2}{sup 2+} and associated species) may be reduced, directly or indirectly, by microorganisms to U(IV) with subsequent precipitation of UO{sub 2}. However, various factors within soils and sediments may limit biological reduction of U(VI), inclusive of alterations in U(VI) speciation and competitive electron acceptors. Here we elucidate themore » impact of U(VI) speciation on the extent and rate of reduction with specific emphasis on speciation changes induced by dissolved Ca, and we examine the impact of Fe(III) (hydr)oxides (ferrihydrite, goethite and hematite) varying in free energies of formation on U reduction. The amount of uranium removed from solution during 100 h of incubation with S. putrefaciens was 77% with no Ca or ferrihydrite present but only 24% (with ferrihydrite) and 14% (no ferrihydrite) were removed for systems with 0.8 mM Ca. Imparting an important criterion on uranium reduction, goethite and hematite decrease the dissolved concentration of calcium through adsorption and thus tend to diminish the effect of calcium on uranium reduction. Dissimilatory reduction of Fe(III) and U(VI) can proceed through different enzyme pathways, even within a single organism, thus providing a potential second means by which Fe(III) bearing minerals may impact U(VI) reduction. We quantify rate coefficients for simultaneous dissimilatory reduction of Fe(III) and U(VI) in systems varying in Ca concentration (0 to 0.8 mM), and using a mathematical construct implemented with the reactive transport code MIN3P, we reveal the predominant influence of uranyl speciation, specifically the formation of uranyl-calcium-carbonato complexes, and ferrihydrite on the rate and extent of uranium reduction in complex geochemical systems.« less

RATES OF HYDROUS FERRIC OXIDE CRYSTALLIZATION AND THE INFLUENCE ON COPRECIPITATED ARSENATE

EPA Science Inventory

Arsenate coprecipitated with hydrous ferric oxide (HFO) was stabilized against dissolution during transformation of HFO to more crystalline iron (hydr)oxides. The rate of arsenate stabilization approximately coincided with the rate of HFO transformation at pH 6 and 40 ?C. Compa...

METHOD OF ELECTROPLATING ON URANIUM

DOEpatents

Rebol, E.W.; Wehrmann, R.F.

1959-04-28

This patent relates to a preparation of metallic uranium surfaces for receiving coatings, particularly in order to secure adherent electroplated coatings upon uranium metal. In accordance with the invention the uranium surface is pretreated by degreasing in trichloroethylene, followed by immersion in 25 to 50% nitric acid for several minutes, and then rinsed with running water, prior to pickling in trichloroacetic acid. The last treatment is best accomplished by making the uranium the anode in an aqueous solution of 50 per cent by weight trichloroacetic acid until work-distorted crystals or oxide present on the metal surface have been removed and the basic crystalline structure of the base metal has been exposed. Following these initial steps the metallic uranium is rinsed in dilute nitric acid and then electroplated with nickel. Adnerent firmly-bonded coatings of nickel are obtained.

High temperature fuel/emitter system for advanced thermionic fuel elements

NASA Astrophysics Data System (ADS)

Moeller, Helen H.; Bremser, Albert H.; Gontar, Alexander; Fiviesky, Evgeny

1997-01-01

Specialists in space applications are currently focusing on bimodal power systems designed to provide both electric power and thermal propulsion (Kennedy, 1994 and Houts, 1995). Our work showed that thermionics is a viable technology for nuclear bimodal power systems. We demonstrated that materials for a thermionic fuel-emitter combination capable of performing at operating temperatures of 2473 K are not only possible but available. The objective of this work, funded by the US Department of Energy, Office of Space and Defense Power Systems, was to evaluate the compatibility of fuel material consisting of an uranium carbide/tantalum carbide solid solution with an emitter material consisting of a monocrystalline tungsten-niobium alloy. The uranium loading of the fuel material was 70 mole% uranium carbide. The program was successfully accomplished by a B&W/SIA LUTCH team. Its workscope was integrated with tasks being performed at both Babcock & Wilcox, Lynchburg Research Center, Lynchburg, Virginia, and SIA LUTCH, Podolsk, Russia. Samples were fabricated by LUTCH and seven thermal tests were performed in a hydrogen atmosphere. The first preliminary test was performed at 2273 K by LUTCH, and the remaining six tests were performed At B&W. Three tests were performed at 2273 K, two at 2373 K, and the final test at 2473 K. The results showed that the fuel and emitter materials were compatible in the presence of hydrogen. No evidence of liquid formation, dissolution of the uranium carbide from the uranium carbide/tantalum carbide solid solution, or diffusion of the uranium into the monocrystalline tungsten alloy was observed. Among the highlights of the program was the successful export of the fuel samples from Russia and their import into the US by commercial transport. This paper will discuss the technical aspects of this work.

Distribution and mode of occurrence of uranium in bottom ash derived from high-germanium coals.

PubMed

Sun, Yinglong; Qi, Guangxia; Lei, Xuefei; Xu, Hui; Li, Lei; Yuan, Chao; Wang, Yi

2016-05-01

The radioactivity of uranium in radioactive coal bottom ash (CBA) may be a potential danger to the ambient environment and human health. Concerning the limited research on the distribution and mode of occurrence of uranium in CBA, we herein report our investigations into this topic using a number of techniques including a five-step Tessier sequential extraction, hydrogen fluoride (HF) leaching, Siroquant (Rietveld) quantification, magnetic separation, and electron probe microanalysis (EPMA). The Tessier sequential extraction showed that the uranium in the residual and Fe-Mn oxide fractions was dominant (59.1% and 34.9%, respectively). The former was mainly incorporated into aluminosilicates, retained with glass and cristobalite, whereas the latter was especially enriched in the magnetic fraction, of which about 50% was present with magnetite (Fe3O4) and the rest in other iron oxides. In addition, the uranium in the magnetic fraction was 2.6 times that in the non-magnetic fraction. The experimental findings in this work may be important for establishing an effective strategy to reduce radioactivity from CBA for the protection of our local environment. Copyright © 2015. Published by Elsevier B.V.

Effect of metal sulfide pulp density on gene expression of electron transporters in Acidithiobacillus sp. FJ2.

PubMed

Fatemi, Faezeh; Miri, Saba; Jahani, Samaneh

2017-05-01

In Acidithiobacillus ferrooxidans, one of the most important bioleaching bacterial species, the proteins encoded by the rus operon are involved in the electron transfer from Fe 2+ to O 2 . To obtain further knowledge about the mechanism(s) involved in the adaptive responses of the bacteria to growth on the different uranium ore pulp densities, we analyzed the expression of the four genes from the rus operon by real-time PCR, when Acidithiobacillus sp. FJ2 was grown in the presence of different uranium concentrations. The uranium bioleaching results showed the inhibitory effects of the metal pulp densities on the oxidation activity of the bacteria which can affect Eh, pH, Fe oxidation and uranium extractions. Gene expression analysis indicated that Acidithiobacillus sp. FJ2 tries to survive in the stress with increasing in the expression levels of cyc2, cyc1, rus and coxB, but the metal toxicity has a negative effect on the gene expression in different pulp densities. These results indicated that Acidithiobacillus sp. FJ2 could leach the uranium even in high pulp density (50%) by modulation in rus operon gene responses.

Uranium behaviour in an estuary polluted by mining and industrial effluents: the Ría of Huelva (SW of Spain).

PubMed

Hierro, A; Martín, J E; Olías, M; Vaca, F; Bolivar, J P

2013-10-15

This paper describes a comprehensive study of the behaviour of U in the Ría of Huelva estuary, formed by the Tinto and Odiel rivers. This ecosystem is conditioned by two hydrochemical facts: one connected with the acid mining drainage (AMD) generated in the first section of the river basins, and another one related to the fertilizer industry located at the estuary. AMD gives a singular character to these rivers; low pH and high redox potential that keep high amounts of toxic elements and radionuclides in dissolution. Most of the data for dissolved U in estuaries indicate conservative mixing, but there are examples of non-conservative behaviour attributed to oxidation/reduction processes or solubility variations. In the Ría of Huelva estuary the U shows a non-conservative behaviour due to solubility changes produced by variations in the pH. A complete removal of riverine dissolved U is observed in a pH range of 4-6. At higher pH values, U release from suspended matter, and probably also from sediments into the dissolved phase is found. Copyright © 2013 Elsevier Ltd. All rights reserved.

Use of UO 2 films for electrochemical studies

NASA Astrophysics Data System (ADS)

Miserque, F.; Gouder, T.; Wegen, D. H.; Bottomley, P. D. W.

2001-10-01

UO 2 films have been prepared by dc reactive sputtering of a uranium metal target in an Ar/O 2 atmosphere. We have used the films deposited on gold substrates as working electrodes for electrochemical investigations as simulating the surfaces of fuel pellets. Film composition was determined by photoelectron spectroscopy (XPS and UPS) and X-ray diffraction (XRD). The oxide stoichiometry as a function of deposition conditions was determined and the appropriate conditions for UO 2.0 formation established. AC impedance and cyclic voltammetry measurements were performed. A double RC electrical equivalent circuit was used to fit the data from impedance measurements, similar to those used in unirradiated UO 2 or spent fuel pellets. However due to the porosity or adhesion defects on the thin films that permitted a direct contact between the solution and the gold substrate, we were obliged to add a contribution simulating the water-gold system. Cyclic voltammetry measurements show the influence of pH on the dissolution mechanism. Alkaline solutions permit the formation of an oxidised layer (UO 2.33) which is not present in the acidic solutions. In both pH=2 and pH=6 solutions, a U VI species layer is formed.

Studies of the kinetics and mechanism of the oxidation of uranium by dry and moist air A model for determining the oxidation rate over a wide range of temperatures and water vapour pressures

NASA Astrophysics Data System (ADS)

McGillivray, G. W.; Geeson, D. A.; Greenwood, R. C.

1994-01-01

The rate of oxidation of uranium metal by moist air has been measured at temperatures from 115 to 350°C and water vapour pressures from 0 to 47 kPa (350 Torr). From this and from previously reported data, a model has been developed which allows the rate of uranium oxidation to be calculated at any particular combination of temperature and water vapour pressure of interest, in the range 0-350°C and 0-101.3 kPa (760 Torr). The model is based on the assumption that the surface concentration of water determines the rate of reaction and that the adsorption of water onto the oxide follows a Langmuir type isotherm. Theoretical plots of rate as a function of water vapour pressure and Arrhenius plots derived from the model have been shown to be in good agreement with experimental data. The model assumes separate contributions to the overall observed rate from oxygen and water vapour. Surface studies have been carried out using SIMS (secondary ion mass spectrometry). Depth profiling of the oxide produced by isotopically labelled reagents ( 18O 2 and H 218O), has shown that oxygen from both reactants is incorporated into the oxide layer in the ratio predicted by the kinetic model. This supports a mechanism in which oxygen and water vapour produce separate diffusing species (possibly O 2- and OH -).

Influence of calcium on microbial reduction of solid phase uranium(VI).

PubMed

Liu, Chongxuan; Jeon, Byong-Hun; Zachara, John M; Wang, Zheming

2007-08-15

The effect of calcium on the dissolution and microbial reduction of a representative solid phase uranyl [U(VI)], sodium boltwoodite (NaUO(2)SiO(3)OH . 1.5H(2)O), was investigated to evaluate the rate-limiting step of microbial reduction of the solid phase U(VI). Microbial reduction experiments were performed in a culture of a dissimilatory metal-reducing bacterium (DMRB), Shewanella oneidensis strain MR-1, in a bicarbonate medium with lactate as electron donor at pH 6.8 buffered with PIPES. Calcium increased the rate of Na-boltwoodite dissolution and U(VI) bioavailability by increasing its solubility through the formation of a ternary aqueous calcium-uranyl-carbonate species. The ternary species, however, decreased the rates of microbial reduction of aqueous U(VI). Laser-induced fluorescence spectroscopy (LIFS) and transmission electron microscopy (TEM) collectively revealed that microbial reduction of solid phase U(VI) was a sequentially coupled process of Na-boltwoodite dissolution, U(VI) aqueous speciation, and microbial reduction of dissolved U(VI) to U(IV) that accumulated on bacterial surfaces/periplasm. Under studied experimental conditions, the overall rate of microbial reduction of solid phase U(VI) was limited by U(VI) dissolution reactions in solutions without calcium and limited by microbial reduction in solutions with calcium. Generally, the overall rate of microbial reduction of solid phase U(VI) was determined by the coupling of solid phase U(VI) dissolution, U(VI) aqueous speciation, and microbial reduction of dissolved U(VI) that were all affected by calcium. (c) 2007 Wiley Periodicals, Inc.

The role of uranium-arene bonding in H2O reduction catalysis

NASA Astrophysics Data System (ADS)

Halter, Dominik P.; Heinemann, Frank W.; Maron, Laurent; Meyer, Karsten

2018-03-01

The reactivity of uranium compounds towards small molecules typically occurs through stoichiometric rather than catalytic processes. Examples of uranium catalysts reacting with water are particularly scarce, because stable uranyl groups form that preclude the recovery of the uranium compound. Recently, however, an arene-anchored, electron-rich uranium complex has been shown to facilitate the electrocatalytic formation of H2 from H2O. Here, we present the precise role of uranium-arene δ bonding in intermediates of the catalytic cycle, as well as details of the atypical two-electron oxidative addition of H2O to the trivalent uranium catalyst. Both aspects were explored by synthesizing mid- and high-valent uranium-oxo intermediates and by performing comparative studies with a structurally related complex that cannot engage in δ bonding. The redox activity of the arene anchor and a covalent δ-bonding interaction with the uranium ion during H2 formation were supported by density functional theory analysis. Detailed insight into this catalytic system may inspire the design of ligands for new uranium catalysts.

Nitride stabilized core/shell nanoparticles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuttiyiel, Kurian Abraham; Sasaki, Kotaro; Adzic, Radoslav R.

Nitride stabilized metal nanoparticles and methods for their manufacture are disclosed. In one embodiment the metal nanoparticles have a continuous and nonporous noble metal shell with a nitride-stabilized non-noble metal core. The nitride-stabilized core provides a stabilizing effect under high oxidizing conditions suppressing the noble metal dissolution during potential cycling. The nitride stabilized nanoparticles may be fabricated by a process in which a core is coated with a shell layer that encapsulates the entire core. Introduction of nitrogen into the core by annealing produces metal nitride(s) that are less susceptible to dissolution during potential cycling under high oxidizing conditions.

RECOVERY OF URANIUM VALUES FROM RESIDUES

DOEpatents

Schaap, W.B.

1959-08-18

A process is described for the recovery of uranium from insoluble oxide residues resistant to repeated leaching with mineral acids. The residue is treated with gaseous hydrogen fluoride, then with hydrogen and again with hydrogen fluoride, preferably at 500 to 700 deg C, prior to the mineral acid leaching.

Bioleaching mechanism of Co and Li from spent lithium-ion battery by the mixed culture of acidophilic sulfur-oxidizing and iron-oxidizing bacteria.

PubMed

Xin, Baoping; Zhang, Di; Zhang, Xian; Xia, Yunting; Wu, Feng; Chen, Shi; Li, Li

2009-12-01

The bioleaching mechanism of Co and Li from spent lithium-ion batteries by mixed culture of sulfur-oxidizing and iron-oxidizing bacteria was investigated. It was found that the highest release of Li occurred at the lowest pH of 1.54 with elemental sulfur as an energy source, the lowest occurred at the highest pH of 1.69 with FeS(2). In contrast, the highest release of Co occurred at higher pH and varied ORP with S + FeS(2), the lowest occurred at almost unchanged ORP with S. It is suggested that acid dissolution is the main mechanism for Li bioleaching independent of energy matters types, however, apart from acid dissolution, Fe(2+) catalyzed reduction takes part in the bioleaching process as well. Co(2+) was released by acid dissolution after insoluble Co(3+) was reduced into soluble Co(2+) by Fe(2+) in both FeS(2) and FeS(2) + S systems. The proposed bioleaching mechanism mentioned above was confirmed by the further results obtained from the experiments of bioprocess-stimulated chemical leaching and from the changes in structure and component of bioleaching residues characterized by XPS, SEM and EDX.

Activation characteristics of multiphase Zr-based hydrogen storage alloys for Ni/MH rechargeable batteries

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, H.; Lee, S.M.; Lee, J.Y.

1999-10-01

AB{sub 2} type Zr-based Laves phase alloys have been studied for possible use as negative electrodes of Ni/MH batteries with high hydrogen storage capacity. However, these alloys have the serious problem of slow activation owing to the formation of surface oxide films. To overcome this problem, alloys with multiphase microstructures have been developed. These alloys become electrochemically active via the creation of micropores by the dissolution of soluble oxide components such as vanadium oxide. However, this phenomenon has been described based only on changes in the chemical composition of the oxide layer. In the present study, this phenomenon is approachedmore » with respect to interactions between the constituent phases. An electrochemical analysis of constituent phases showed that the second phase, resulting in localized Ni-rich pits on the alloy surface. The presence of microcracks at the periphery of the Ni-rich pits after 30 h exposure to KOH electrolyte implies that hydrogen is absorbed preferentially at Ni-rich pits, thereby forming a large active surface area. However, such multiphase alloys have poor cycle durability due to the persistent dissolution of components in the second phase. Through Cr substitution, the authors have developed a family of durable alloys to prevent this unwanted dissolution from the second phase.« less

Effects of selective handling of pyritic, acid-forming materials on the chemistry of pore gas and ground water at a reclaimed surface coal mine in Clarion County, PA, USA

USGS Publications Warehouse

Cravotta,, Charles A.; Dugas, Diana L.; Brady, Keith; Kovalchuck, Thomas E.

1994-01-01

A change from dragline to “selective handling” mining methods at a reclaimed surface coal mine in western Pennsylvania did not significantly affect concentrations of metals in ground water because oxidation of pyrite and dissolution of siderite were not abated. Throughout the mine, placement of pyritic material near the land surface facilitated the oxidation of pyrite, causing the consumption of oxygen (O2) and release of acid, iron, and sulfate ions. Locally in the unsaturated zone, water sampled within or near pyritic zones was acidic, with concentrations of sulfate exceeding 3,000 milligrams per liter (mg/L). However, acidic conditions generally did not persist below the water table because of neutralization by carbonate minerals. Dissolution of calcite, dolomite, and siderite in unsaturated and saturated zones produced elevated concentrations of carbon dioxide (CO2), alkalinity, calcium, magnesium, iron, and manganese. Alkalinity concentrations of 600 to 800 mg/L as CaCO3 were common in water samples from the unsaturated zone in spoil, and alkalinities of 100 to 400 mg/L as CaCO3 were common in ground-water samples from the underlying saturated zone in spoil and bedrock. Saturation indices indicated that siderite could dissolve in water throughout the spoil, but that calcite dissolution or precipitation could occur locally. Calcite dissolution could be promoted as a result of pyrite oxidation, gypsum precipitation, and calcium ion exchange for sodium. Calcite precipitation could be promoted by evapotranspiration and siderite dissolution, and corresponding increases in concentrations of alkalinity and other solutes. Partial pressures of O2 (Po2) and CO2 (Pco2) in spoil pore gas indicated that oxidation of pyrite and precipitation of ferric hydroxide, coupled with dissolution of calcite, dolomite, and siderite were the primary reactions affecting water quality. Highest vertical gradients in Po2, particularly in the near-surface zone (0-1 m), did not correlate with concentrations of total sulfur in spoil. This lack of correlation could indicate that total sulfur concentrations in spoil do not reflect the amount of reactive pyrite or that oxidation rates can be controlled more by rates of O2 diffusion than the amount of pyrite. Hence, if placed in O2-rich zones near the land surface, even small amounts of disseminated pyritic material can be relatively significant sources of acid and mineralized water.