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Sample records for grade uranium oxide

  1. Analysis of nuclear grade uranium oxides by atomic absorption spectrometry with electrothermal atomization

    International Nuclear Information System (INIS)

    Batistoni, D.A.; Erlijman, L.H.; Pazos, A.L.

    1986-01-01

    The application of atomic absorption spectrometry for the determination of five trace impurities in nuclear grade uranium oxides is described. The elements were separated from the uranium matrix by extraction chromatography and determined in 5.5 M nitric acid by electrothermal atomization using pyrolytic graphite coated tubes. Two elements, cadmium and chromium, with different volatility characteristics were employed to investigate the operating conditions. Drying and ashing conditions were studied for both elements. Ramp and constant potential (step) heating modes have also been studied and compared. Good reproducibility and a longer life of graphite tubes were obtained with ramp atomization. Detection limits (in micrograms per gram of uranium) were: Cd 0.01; Cr 0.1; Cu 0.4; Mn 0.04 and Ni 0.2. (author) [es

  2. Standard specification for uranium oxides with a 235U content of less than 5 % for dissolution prior to conversion to nuclear-grade uranium dioxide

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2005-01-01

    1.1 This specification covers uranium oxides, including processed byproducts or scrap material (powder, pellets, or pieces), that are intended for dissolution into uranyl nitrate solution meeting the requirements of Specification C788 prior to conversion into nuclear grade UO2 powder with a 235U content of less than 5 %. This specification defines the impurity and uranium isotope limits for such urania powders that are to be dissolved prior to processing to nuclear grade UO2 as defined in Specification C753. 1.2 This specification provides the nuclear industry with a general standard for such uranium oxide powders. It recognizes the diversity of conversion processes and the processes to which such powders are subsequently to be subjected (for instance, by solvent extraction). It is therefore anticipated that it may be necessary to include supplementary specification limits by agreement between the buyer and seller. 1.3 The scope of this specification does not comprehensively cover all provisions for prevent...

  3. Development of a pneumatic transport system for bulk transfer of metal grade uranium oxide powder

    International Nuclear Information System (INIS)

    Manna, S.; Satpati, S.K.; Roy, S.B.

    2010-01-01

    Uranium oxide powder is a commonly handled ceramic powder in nuclear industries. Design of the powder transfer system is an important aspect because of some of its typical characteristics. Pneumatic transport system has been widely used in transferring powder from one place to another. A pneumatic transport system using vacuum has been presented in the paper. This is used for bulk transfer of UO 3 powder. The system consists of a cyclone separator and filter cloth at the top of the cyclone separator. The pneumatic transfer system provides high efficiency with sustainable performance and it is a compact, robust, handy and moveable unit. No degradation of the powder quality has been observed during transfer. The system provides highly efficient, easy and safe transfer of radioactive powder, better working environment for the operator. (author)

  4. Determination of chlorine in nuclear-grade uranium compounds

    International Nuclear Information System (INIS)

    Yang Chunqing; Liu Fuyun; Huang Dianfan

    1988-01-01

    The determination of chlorine in nuclear-grade uranium compounds is discribed. Chlorine is separated from uranium oxide pyrohydrolytically with stream of wet oxygen in a furnace at 800 ∼ 900 deg C. Chlorine is volatilized as hydrochloric acid, absorbed in a dilute alkaline solution and measured with chlorine-selective electrode. This method covers the concentration range of 10 ∼ 500 pm chlorine in uranium oxide. Precision of at least ± 10% and recovery of 85 ∼ 108% have been reported

  5. Direct determination by atomic absorption of calcium, cobalt and zinc in nuclear grade uranium oxide

    International Nuclear Information System (INIS)

    Guido, O.O.; Amaya, Carlos.

    1975-05-01

    A study has been made of the effect of flame composition (fuel: C 2 H 2 , comburent: air or N 2 O) and location of the burner on the three analytes in a nitric medium, in presence and in absence of uranium. For calcium it was necessary to use N 2 O, while for zinc and cobalt the use of air was found more adequate. The standard additions method for the quantitative determination was adopted. The absorption at the analytical wavelength not corresponding to the elements studied was determined by comparison between this method and another indirect one, using extraction with TBP, and the results were expressed as equivalent concentrations. Confidence intervals of the analytical results were evaluated statistically using a scheme of calculation adapted to the proposed method. This evaluation allowed an estimation of the detection limits (calcium: 5 ppm, cobalt: 3 ppm, zinc: 1 ppm). (author)

  6. Standard specification for blended uranium oxides with 235U content of less than 5 % for direct hydrogen reduction to nuclear grade uranium dioxide

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2001-01-01

    1.1 This specification covers blended uranium trioxide (UO3), U3O8, or mixtures of the two, powders that are intended for conversion into a sinterable uranium dioxide (UO2) powder by means of a direct reduction process. The UO2 powder product of the reduction process must meet the requirements of Specification C 753 and be suitable for subsequent UO2 pellet fabrication by pressing and sintering methods. This specification applies to uranium oxides with a 235U enrichment less than 5 %. 1.2 This specification includes chemical, physical, and test method requirements for uranium oxide powders as they relate to the suitability of the powder for storage, transportation, and direct reduction to UO2 powder. This specification is applicable to uranium oxide powders for such use from any source. 1.3 The scope of this specification does not comprehensively cover all provisions for preventing criticality accidents, for health and safety, or for shipping. Observance of this specification does not relieve the user of th...

  7. Uranium oxide recovering method

    International Nuclear Information System (INIS)

    Ota, Kazuaki; Takazawa, Hiroshi; Teramae, Naoki; Onoue, Takeshi.

    1997-01-01

    Nitrates containing uranium nitrate are charged in a molten salt electrolytic vessel, and a heat treatment is applied to prepare molten salts. An anode and a cathode each made of a graphite rod are disposed in the molten salts. AC voltage is applied between the anode and the cathode to conduct electrolysis of the molten salts. Uranium oxides are deposited as a recovered product of uranium, on the surface of the anode. The nitrates containing uranium nitrate are preferably a mixture of one or more nitrates selected from sodium nitrate, potassium nitrate, calcium nitrate and magnesium nitrate with uranium nitrate. The nitrates may be liquid wastes of nitrates. The temperature for the electrolysis of the molten salts is preferably from 150 to 300degC. The voltage for the electrolysis of the molten salts is preferably an AC voltage of from 2 to 6V, more preferably from 4 to 6V. (I.N.)

  8. Estimation grade of uranium from drill hole gamma logs

    International Nuclear Information System (INIS)

    Juliao, B.

    1986-01-01

    Radiometric grade of uranium deposits can be determined from drill hole gamma logs. The calculation of uranium oxide content can be obtained with good precision when the uranium ore is in radioactive equilibrium, containing only a small amount of thorium and no interference of potassium. This is the case of uranium ore from the Lagoa Real Uranium Province presented in this paper. The radioactive disequilibrium study in this province were made working over nine hundred samples analised with this special purpose in the CDTN-NUCLEBRAS laboratories. The data obtained indicated that the uranium in the ore is in perfect equilibrium with their daughter gamma emitters. Futhermore, the amount of Th and K is of no significance, so that the gamma counting represents exactly the uranium content of the ore. (author) [pt

  9. Uranium production from low grade Swedish shale

    International Nuclear Information System (INIS)

    Carlsson, O.

    1977-01-01

    In view of the present nuclear programmes a steep increase in uranium demand is foreseen which will pose serious problems for the uranium industry. The annual additions to uranium ore reserves must almost triple within the next 15 years in order to support the required production rates. Although there are good prospects for the discovery of further conventional deposits of uranium there is a growing interest in low grade uranium deposits. Large quantities of uranium exist in black shales, phosphates, granites, sea water and other unconventional sources. There are however factors which limit the utilization of these low grade materials. These factors include the extraction costs, the environmental constrains on mining and milling of huge amounts of ore, the development of technologies for the beneficiation of uranium and, in the case of very low grade materials, the energy balance. The availability of by-product uranium is limited by the production rate of the main product. The limitations differ very much according to types of ores, mining and milling methods and the surroundings. As an illustration a description is given of the Swedish Ranstad uranium shale project, its potential, constraints and technical solutions

  10. Method for converting uranium oxides to uranium metal

    International Nuclear Information System (INIS)

    Duerksen, W.K.

    1988-01-01

    A method for converting uranium oxide to uranium metal is described comprising the steps of heating uranium oxide in the presence of a reducing agent to a temperature sufficient to reduce the uranium oxide to uranium metal and form a heterogeneous mixture of a uranium metal product and oxide by-products, heating the mixture in a hydrogen atmosphere at a temperature sufficient to convert uranium metal in the mixture to uranium hydride, cooling the resulting uranium hydride-containing mixture to a temperature sufficient to produce a ferromagnetic transition in the uranium hydride, magnetically separating the cooled uranium hydride from the mixture, and thereafter heating the separated uranium hydride in an inert atmosphere to a temperature sufficient to convert the uranium hydride to uranium metal

  11. Electrolytic recovery of uranium oxides

    International Nuclear Information System (INIS)

    Gurr, W.R.

    1979-01-01

    A method is described for extracting uranium oxide from a solution of one or more uranium compounds, e.g. leach liquors, comprising subjecting the solution to electrolysis utilizing a high current density, e.g. 500 to 4000 amp/m 2 , whereby uranium oxide is formed at the cathode and is recovered. The method is particularly suited to a continuous process using a rotating cathode cell. (author)

  12. Method for converting uranium oxides to uranium metal

    Science.gov (United States)

    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.

  13. Contribution to characteristics of uranium oxides

    International Nuclear Information System (INIS)

    Fritsche, R.; Dahlkamp, F.J.

    2001-01-01

    Uranium oxides from pegmatitic, metamorphic and metasomatic uranium occurrences were investigated with the objective to check for differences in their physico-chemical properties and, whether such properties are sufficiently distinct to be applied as an exploration tool. Research methods included microscopy, electron microprobe and X-ray diffractometry amended by determinations of reflectance, Vickers hardness, unit-cell dimension and oxidation grade. Tentative research results are as follows: (a) U-oxides (uraninites) of pegmatites always contain significant amounts of Th (1,5-10 wt.% ThO2). (b) U-oxides from metasomatic environments have high, but variable contents of Fe, Ca, Ti, Si and Th (around 10 wt.%), Th being low. (c) U-oxides crystallised during metamorphism contain minor impurities of the above listed elements (total of oxides < 2 wt.%). (d) Redistributed U-oxides have elevated amounts of these elements. (e) Unit-cell dimensions of U-oxides tend to reflect a complex function of formation temperature, oxidation grade and the influence of incorporated elements caused by their radius and electro-negativity. (f) A global negative correlation of unit-cell dimension and oxidation grade of uranium oxides is indicated but based on widely varying ratios of the two parameters. (g) Colloform U-oxide (pitchblende) is characterised by elevated Ca-contents (1-5 wt.% CaO) and an almost complete lack of Th (< 1 wt.% ThO2). (h) Idiomorphic U-oxide (uraninite) is commonly low in Ca (< 1.5 wt.% CaO) but contains relatively high Th values. (i) The reflectance of U-oxides generally correlates positively with Vickers hardness and unit-cell dimension, but the incorporation of other elements in the lattice of U-oxides may cause strong interference. (author)

  14. Uranium plutonium oxide fuels

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  15. Comparison of oxidants in alkaline leaching of uranium ore

    International Nuclear Information System (INIS)

    Sreenivas, T.; Rajan, K.C.; Srinivas, K.; Anand Rao, K.; Manmadha Rao, M.; Venkatakrishnan, R.R.; Padmanabhan, N.P.H.

    2007-01-01

    The uranium minerals occurring in various ore deposits consists of predominantly uranous ion (U +4 ), necessitating use of an oxidant and other lixiviants for efficient dissolution during leaching. Unlike acid leaching route, where uranium minerals dissolution could be achieved efficiently with cheaper lixiviants, processing of ores by alkaline leaching route involve expensive lixiviants and drastic leaching conditions. Alkaline leaching of uranium ores becomes economical only upon using cheaper and efficient oxidants and conservation of other reagents by their recycle. The present paper gives efficacy of various oxidants - KMnO 4 , NaOCl, Cu - NH 3 , air and oxygen, in the leaching of uranium from a low-grade dolostone hosted uranium ore of India. A comparison based on technical merits and cost of the oxidant chemicals is discussed. (author)

  16. Oxygen isotope fractionation in uranium oxides

    International Nuclear Information System (INIS)

    Zheng Yongfei

    1995-01-01

    Thermodynamic oxygen isotope factors for uranium oxides have been calculated by means of the modified increment method. The sequence of 18 O-enrichment in the uranium oxides with respect to the common rock-forming minerals is predicted as follows: spinel 3 < illite. Two sets of self-consistent fractionation factors between the uranium oxides and water and between the uranium oxides and the other minerals have been obtained for 0∼1200 degree C. The theoretical results are applicable to the isotopic geothermometry of uranium ores when pairing with other gangue minerals in hydrothermal uranium deposits

  17. Feasibility studies on electrochemical separation and recovery of uranium by using domestic low grade uranium resources

    International Nuclear Information System (INIS)

    Oh, Won Zin; Jung, Chong Hun; Lee, Kune Woo; Won, Hui Jun; Choi, Wang Kyu; Kim, Gye Nam; Lee, Yu Ri; Lee, Joong Moung

    2005-12-01

    The up-to-date electrochemical uranium separation technology has been developed for uranium sludge waste treatment funded by a long term national nuclear technology development program. The objective of the studies is to examine applicability of the uranium separation technology to making use of the low grade uranium resources in the country. State of the arts of uranium separation and recovery from the low grade national uranium resources. - The amount of the high grade uranium resources(0.1 % U 3 O 8 contents) in the world is 1,750,000MTU and that of the low grade uranium resources(0.04 % U 3 O 8 contents) in the country is 340,000MTU. - The world uranium price will be increase to more than 30$/l0b in 10 years, so that the low grade uranium in the country become worth while to recover. - The conventional uranium recovery technologies are based on both acidic - The ACF electrochemical uranium separation technology is the state of the art technology in the world and the adsorption capability of 690 mgU/g is several ten times higher than that of a conventional zeolite and the uranium stripping efficiency by desorption is more than 99%. So, this technology is expected to replace the existing solvent extraction technology. Feasibility of the ACF electrochemical uranium separation technology as an uranium recovery method. Lab scale demonstration of uranium separation and recovery technologies have been carried out by using an ACF electrochemical method

  18. Status Report from the United Kingdom [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    North, A A [Warren Spring Laboratory, Stevenage, Herts. (United Kingdom)

    1967-06-15

    The invitation to present this status report could have been taken literally as a request for information on experience gained in the actual processing of low-grade uranium ores in the United Kingdom, in which case there would have been very little to report; however, the invitation naturally was considered to be a request for a report on the experience gained by the United Kingdom of the processing of uranium ores. Lowgrade uranium ores are not treated in the United Kingdom simply because the country does not possess any known significant deposits of uranium ore. It is of interest to record the fact that during the nineteenth century mesothermal vein deposits associated with Hercynian granite were worked at South Terras, Cornwall, and ore that contained approximately 100 tons of uranium oxide was exported to Germany. Now only some 20 tons of contained uranium oxide remain at South Terras; also in Cornwall there is a small number of other vein deposits that each hold about five tons of uranium. Small lodes of uranium ore have been located in the southern uplands of Scotland; in North Wales lower palaeozoic black shales have only as much as 50 to 80 parts per million of uranium oxide, and a slightly lower grade carbonaceous shale is found near the base of the millstone grit that occurs in the north of England. Thus the experience gained by the United Kingdom has been of the treatment of uranium ores that occur abroad.

  19. Influence of uranium hydride oxidation on uranium metal behaviour

    International Nuclear Information System (INIS)

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

    2013-01-01

    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, if 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)

  20. Influence of uranium hydride oxidation on uranium metal behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Patel, N.; Hambley, D. [National Nuclear Laboratory (United Kingdom); Clarke, S.A. [Sellafield Ltd (United Kingdom); Simpson, K.

    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, if 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)

  1. Uranium mobilization from low-grade ore by cyanobacteria

    International Nuclear Information System (INIS)

    Lorenz, M.G.; Krumbein, W.E.

    1985-01-01

    Three cyanobacterial isolates (two LPP-B forms and one Anabaena or Nostoc species) from different environments could mobilize uranium from low-grade ores. After 80 days, up to 18% uranium had been extracted from coal and 51% from a carbonate rock by the filamentous cyanobacterium OL3, a LPP-B form. Low growth requirements with regard to light and temperature optima make this strain a possible candidate for leaching neutral and alkaline low-grade uranium ores. (orig.)

  2. Uranium mobilization from low-grade ore by cyanobacteria

    Energy Technology Data Exchange (ETDEWEB)

    Lorenz, M.G.; Krumbein, W.E.

    1985-04-01

    Three cyanobacterial isolates (two LPP-B forms and one Anabaena or Nostoc species) from different environments could mobilize uranium from low-grade ores. After 80 days, up to 18% uranium had been extracted from coal and 51% from a carbonate rock by the filamentous cyanobacterium OL3, a LPP-B form. Low growth requirements with regard to light and temperature optima make this strain a possible candidate for leaching neutral and alkaline low-grade uranium ores.

  3. Production of sized particles of uranium oxides and uranium oxyfluorides

    International Nuclear Information System (INIS)

    Knudsen, I.E.; Randall, C.C.

    1976-01-01

    A process is claimed for converting uranium hexafluoride (UF 6 ) to uranium dioxide (UO 2 ) of a relatively large particle size in a fluidized bed reactor by mixing uranium hexafluoride with a mixture of steam and hydrogen and by preliminary reacting in an ejector gaseous uranium hexafluoride with steam and hydrogen to form a mixture of uranium and oxide and uranium oxyfluoride seed particles of varying sizes, separating the larger particles from the smaller particles in a cyclone separator, recycling the smaller seed particles through the ejector to increase their size, and introducing the larger seed particles from the cyclone separator into a fluidized bed reactor where the seed particles serve as nuclei on which coarser particles of uranium dioxide are formed. 9 claims, 2 drawing figures

  4. Standard specification for sintered gadolinium oxide-uranium dioxide pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This specification is for finished sintered gadolinium oxide-uranium dioxide pellets for use in light-water reactors. It applies to gadolinium oxide-uranium dioxide pellets containing uranium of any 235U concentration and any concentration of gadolinium oxide. 1.2 This specification recognizes the presence of reprocessed uranium in the fuel cycle and consequently defines isotopic limits for gadolinium oxide-uranium dioxide pellets made from commercial grade UO2. Such commercial grade UO2 is defined so that, regarding fuel design and manufacture, the product is essentially equivalent to that made from unirradiated uranium. UO2 falling outside these limits cannot necessarily be regarded as equivalent and may thus need special provisions at the fuel fabrication plant or in the fuel design. 1.3 This specification does not include (1) provisions for preventing criticality accidents or (2) requirements for health and safety. Observance of this specification does not relieve the user of the obligation to be aw...

  5. Electrochemical preparation of new uranium oxide phases

    International Nuclear Information System (INIS)

    Smolenskij, V.V.; Lyalyushkin, N.V.; Bove, A.L.; Komarov, V.K.; Kapshukov, I.I.

    1992-01-01

    Behaviour of uranium ions in oxidation states 3+ and 4+ in molten chlorides of alkali metals in the temperature range of 700-900 degC in the atmosphere of an inert gas was studied by the method of cyclic voltametry. It is shown that as a result of introduction of crystal uranium dioxide into the salt melt formation of uranium oxide ions of the composition UO + and UO 2+ occurs, the ions participating in electrode reactions and bringing about formation of the following uranium oxides on the cathode: UO and, presumably, U 3 O 4 . Oxides UO and U 3 O 4 are thermodynamically unstable at low temperatures and decompose into uranium oxide of the composition UO 2-x , where x varies from 0 to 0.05, and metal uranium

  6. Determination of chlorine in nuclear-grade uranium compounds by ion-selective electrode

    International Nuclear Information System (INIS)

    Yang Chunqing; Liu Fuyun; Huang Dianfan.

    1989-01-01

    The determination of microamount chlorine in nuclear-grade uranium compounds is described. Chlorine is separated from uranium oxide pyrohydrolytically with stream of wet oxygen in a furnace at 800-900 deg C. Chlorine is volatilized as hydrochloric acid, which then is absorbed in a dilute alkaline solution and measured with chlorine selective electrode. This method covers the concentration range of 10-500 ppm chlorine in uranium oxide. The relative standard diviation is better than 10% and recovery of 85-108% has been reported

  7. Electrical impedance studies of uranium oxide

    International Nuclear Information System (INIS)

    Hampton, R.N.

    1986-11-01

    The thesis presents data on the electrical properties of uranium oxide at temperatures from 1700K to 4.2K, and pressures between 25 K bar and 70 K bar. The impedance data were analysed using the technique of complex plane representation to establish the conductivity and dielectric constant of uranium dioxide. The thermophysical data were compared with previously reported experimental and theoretical work on uranium dioxide and other fluorite structured oxides. (U.K.)

  8. Process for producing uranium oxide rich compositions from uranium hexafluoride

    International Nuclear Information System (INIS)

    DeHollander, W.R.; Fenimore, C.P.

    1978-01-01

    Conversion of gaseous uranium hexafluoride to a uranium dioxide rich composition in the presence of an active flame in a reactor defining a reaction zone is achieved by separately introducing a first gaseous reactant comprising a mixture of uranium hexafluoride and a reducing carrier gas, and a second gaseous reactant comprising an oxygen-containing gas. The reactants are separated by a shielding gas as they are introduced to the reaction zone. The shielding gas temporarily separates the gaseous reactants and temporarily prevents substantial mixing and reacting of the gaseous reactants. The flame occurring in the reaction zone is maintained away from contact with the inlet introducing the mixture to the reaction zone. After suitable treatment, the uranium dioxide rich composition is capable of being fabricated into bodies of desired configuration for loading into nuclear fuel rods. Alternatively, an oxygen-containing gas as a third gaseous reactant is introduced when the uranium hexafluoride conversion to the uranium dioxide rich composition is substantially complete. This results in oxidizing the uranium dioxide rich composition to a higher oxide of uranium with conversion of any residual reducing gas to its oxidized form

  9. Oxidation of uranium and uranium alloys

    International Nuclear Information System (INIS)

    Orman, S.

    1976-01-01

    The corrosion behaviour of uranium in oxygen, water and water + oxygen mixtures is compared and contrasted. A considerable amount of work, much of it conflicting, has been published on the U + H 2 O and U + H 2 O + O 2 systems. An attempt has been made to summarise this data and to explain the reasons for the lack of agreement between the experimental results. The evidence for the mechanism involving OH - ion diffusion as the reacting entity in both the U + H 2 O and U + O 2 + H 2 O reactions is advanced. The more limited corrosion data on some lean uranium alloys and on some higher addition alloys referred to as stainless materials is summarised together with some previously unreported results obtained with these materials at AWRE. The data indicates that in the absence of oxygen the lean alloys behave in a similar manner to uranium and evolve hydrogen in approximately theoretical quantities. But the stainless alloys absorb most of the product hydrogen and assessments of reactivity based on hydrogen evolution would be very inaccurate. The direction that future corrosion work on these materials should take is recommended

  10. Low grade uranium ores as potential sources of raw material

    International Nuclear Information System (INIS)

    Venzlaff, H.

    1976-01-01

    Reports on the uranium requirement and the uranium reserves show that, even if the demand were to be stretched out slightly, the rate of new discoveries of uranium would have to be doubled or even tripled within a few years in order to ensure supply. Despite some spectacular discoveries of large scale deposits in Australia it must be said that only very few truly new uranium provinces have been discovered over the past twenty years. In this situation more attention is now being devoted to low grade uranium depositis, to findings whose concentration does not exceed 1,000 ppm. These deposits contain quantities of uranium many times larger than the deposits that can now be mined at prices up to 30/lb of U 3 O 8 . Even now low grade uranium ore is being mined as a byproduct, with the actual valuable mineral producing most of the income from mining activities. However, if one strikes a balance in this situation, one finds that only part of the requirement can be met in this way. Hence, all possibilities must be exhausted to mine uranium as a byproduct, new techniques of uranium production from low grade ores must be developed, and also conventional prospection must be intensified, if the continuity of supply of the nuclear power stations in the eighties and nineties is to be guaranteed. (orig.) [de

  11. PROCESSES OF CHLORINATION OF URANIUM OXIDES

    Science.gov (United States)

    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.

  12. Some characteristics of uranium oxides in China

    International Nuclear Information System (INIS)

    Xu, Guoqing; Wang, Aizhen; Gu, Qifang; Zhang, Jingyi; Zhang, Zhaoming; Huang, Yuzhu

    1981-01-01

    According to the analytical data of seventy-seven samples from several tens of uranium ore deposits and occurrences in China, chemical properties, cell dimensions and reflectance of uranium oxides are studied. Chemical properties of uranium oxides from different types of uranium ore deposits and the influence of various mineralization ages and hosts on the compositions of uranium oxides are presented. The influence of these factor such as mineralization temperatures, the compositions of hosts and geochemical background on the compositions of uranium oxides are evident. Lead in proterozoic uranium oxides is relatively enriched by the decay of radio-active elements. Cell dimensions have positive correlation with mineralization ages, formation temperatures and concentration of rare earths and Pb, and negative correlation with the oxidation coefficient. The cell size is an exponential function of the content in CaO. It is suggested that among the factors of influence the most important is the mineralization temperature. The size of ionic radius of elements substituted U 4 + and autooxidation of U 4 + during the process of the decay of radioactive elements are of secondary importance. The reflectance is independent of the content of CaO and SiO 2 . The reflectance is positively correlative with the cell size and negatively correlative with oxidation coefficient. The relation between the reflectance and the content of PbO is logarithmic

  13. Role of oxidizing agent in the chemistry of in-situ uranium leaching

    International Nuclear Information System (INIS)

    Carlson, R.H.; Norris, R.D.; Schellinger, R.

    1982-01-01

    Synthetic two-component mixtures (uraninite and iron sulfide) as well as native uranium ores obtained from Texas and Wyoming have been examined. Physical/chemical ore properties are correlated with observed laboratory leach response. Data show a large inherent selectivity of oxidant for uranium in the early stages of a leach period. Uranium head grade was found to increase in a nearly linear fashion with hydrogen peroxide concentration in the leach solution. As uranium in the ore is depleted, uranium response decreases and the oxidant serves mainly to leach iron sulfide gangue material. 6 refs

  14. Uranium oxidation: characterization of oxides formed by reaction with water

    International Nuclear Information System (INIS)

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

    1983-01-01

    Three different uranium oxide samples have been characterized with respect to the different preparation techniques. 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. 27 figures

  15. Review of experience gained in fabricating nuclear grade uranium and thorium compounds and their analytical quality control at the Instituto de Energia Atomica, Sao Paulo, Brazil

    International Nuclear Information System (INIS)

    Abrao, A.; Franca Junior, J.M.; Ikuta, A.

    1977-01-01

    The main activities developed at 'Instituto de Energia Atomica' Sao Paulo, Brazil, on the recovery of uranium from ores, the purification of uranium and thorium raw concentrates and their transformation in nuclear grade compounds, are reviewed. The design and assemble of pilot facilities for ammonium diuranate (ADV) uranium tetrafluoride, uranium trioxide, uranium oxide microspheres, uranyl nitrate denitration, uranim hexafluoride and thorium compounds are discussed. The establishment of analytical procedures are emphasized [pt

  16. A discussion on several problems in determination of uranium ore grade criteria

    International Nuclear Information System (INIS)

    Zhu Zhixiang.

    1991-01-01

    The course of determination of uranium ore grade criteria in China is briefly introduced. The cut-off grade minimum industrial grade and allowable minimum average grade uranium ore bodies used in China are reviewed. The meanings and role of various grade criteria and their economic basis for determination in uranium exploration, mining and sorting are discussed and the author's ideas are given

  17. Oxidation and crystal field effects in uranium

    Energy Technology Data Exchange (ETDEWEB)

    Tobin, J. G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Booth, C. H. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Shuh, D. K. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); van der Laan, G. [Diamond Light Source, Didcot (United Kingdom); Sokaras, D. [Stanford Synchrotron Radiation Lightsource, Stanford, CA (United States); Weng, T. -C. [Stanford Synchrotron Radiation Lightsource, Stanford, CA (United States); Yu, S. W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bagus, P. S. [Univ. of North Texas, Denton, TX (United States); Tyliszczak, T. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Nordlund, D. [Stanford Synchrotron Radiation Lightsource, Stanford, CA (United States)

    2015-07-06

    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 (UO2), uranium trioxide (UO3), and uranium tetrafluoride (UF4). As a result, a discussion of the role of non-spherical perturbations, i.e., crystal or ligand field effects, will be presented.

  18. Continued Multicolumns Bioleaching for Low Grade Uranium Ore at a Certain Uranium Deposit

    OpenAIRE

    Gongxin Chen; Zhanxue Sun; Yajie Liu

    2016-01-01

    Bioleaching has lots of advantages compared with traditional heap leaching. In industry, bioleaching of uranium is still facing many problems such as site space, high cost of production, and limited industrial facilities. In this paper, a continued column bioleaching system has been established for leaching a certain uranium ore which contains high fluoride. The analysis of chemical composition of ore shows that the grade of uranium is 0.208%, which is lower than that of other deposits. Howev...

  19. Conversion and Blending Facility highly enriched uranium to low enriched uranium as oxide. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-05

    This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials into pure HEU oxide and (2) blend the pure HEU oxide with depleted and natural uranium oxide to produce an LWR grade LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU will be produced as a waste suitable for storage or disposal.

  20. Development of Pneumatic Transport System (PTS) for safe handling of uranium oxide powder in UMP/UED

    International Nuclear Information System (INIS)

    Manna, S.; Satpati, S.K.; Roy, S.B.

    2009-01-01

    Tonnage quantity radioactive uranium oxide powder of particle size sub micron to 100 micron is handled in Uranium Metal Plant (UMP), UED/BARC for production of nuclear grade uranium metal, required for fuelling research reactors - Dhruva and Cirus. A Pneumatic Transfer System (PTS) using vacuum has been introduced and is being used for handling radioactive powder to improve radiation protection

  1. Low grade uranium deposits of India - a bane or boon

    International Nuclear Information System (INIS)

    Chaki, Anjan

    2010-01-01

    Uranium resources of the world is estimated to be 5.5 million tonnes and the proven resources in India forms 3% of the world resources. The biggest uranium deposit is the Olympic dam deposit in Australia, which contains nearly one million tonnes of 0.04% U 3 O 8 , while the highest grade of nearly 20% is established in the McArthur river deposit, Canada. Another very high grade deposit, the Cigar lake deposit, is established in Canada with an average grade of nearly 18%. Most of the uranium deposits established in India so far falls under the category of low grade. These low grade uranium deposits are distributed mainly in Singhbhum Shear Zone, eastern India; in parts of Chhattisgarh; Southern parts of Meghalaya; Cuddapah Basin, Andhra Pradesh; in parts of Karnataka and Aravalli- and Delhi Supergroups, Rajasthan and Haryana. These deposits are mainly hydrothermal vein type, stratabound type and unconformity related. The Singhbhum Shear Zone, Jharkhand hosts a seventeen low grade uranium deposits, aggregating about 30% of Indian uranium resources. The uranium mineralisation hosted by Vempalle dolostone extends over 160 km belt along southwestern margin of Cuddapah Basin in Andhra Pradesh and accounts 23% of the Indian resources. Though the dolostone hosted Tummalapalle uranium deposit was established in the early nineties, because of techno-economic constraints, the deposit remained dormant. As a consequence of the development of an innovative pressure alkali beneficiation process, the deposit became economically viable and a mine and mill are being constructed here. Recent exploration inputs are leading to prove a number of low grade uranium deposits in the extension areas of Tummalapalle. Nearly 10 blocks have been identified within a 30 km belt which are being actively explored and a large uranium deposit has already been proved in this province. The deposit at Tummalapalle and adjoining areas is likely to become the second biggest deposit in the world. The

  2. Underground Milling of High-Grade Uranium Ore

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, C., E-mail: chuck.edwards@amec.com [AMEC Americas Limited, Saskatoon, Saskatchewan (Canada)

    2014-05-15

    There are many safety and technical issues involved in the mining and progressing of high grade uranium ores such as those exploited in Northern Canada at present. With more of this type of mine due to commence production in the near future, operators have been looking at ways to better manage the situation. The paper describes underground milling of high-grade uranium ore as a means of optimising production costs and managing safety issues. In addition the paper presents some examples of possible process flowsheets and plant layouts that could be applicable to such operations. Finally an assessment of potential benefits from underground milling from a variety of viewpoints is provided. (author)

  3. Processing of low-grade uranium ores

    International Nuclear Information System (INIS)

    Michel, P.

    1975-01-01

    Four types of low grade ores are studied. Low grade ores which must be extracted because they are enclosed in a normal grade deposit. Heap leaching is the processing method which is largely used. It allows to obtain solutions or preconcentrates which may be delivered at the nearest plant. Normal grade ores contained in a low amplitude deposit which can be processed using leaching as far as the operation does not need any large expensive equipment. Medium grade ores in medium amplitude deposits to which a simplified conventional process can be applied using fast heap leaching. Low grade ores in large deposits. The processing possibilities leading to use in place leaching are explained. The operating conditions of the method are studied (leaching agent, preparation of the ore deposit to obtain a good tightness with regard to the hydrological system and to have a good contact between ore and reagent) [fr

  4. Processing of low grade uranium ores

    International Nuclear Information System (INIS)

    Michel, P.

    1978-10-01

    Four types of low-grade ores are studied: (1) Low-grade ores that must be extracted because they are enclosed in a normal-grade deposit. Heap leaching is the processing method which is largely used. (2) Normal-grade ores contained in low-amplitude deposits. They can be processed using in-place leaching as far as the operation does not need any large and expensive equipment. (3) Medium-grade ores in medium-amplitude deposits. A simplified conventional process can be applied using fast heap leaching. (4) Low-grade ores in large deposits. The report explains processing possibilities leading in most cases to the use of in-place leaching. The operating conditions of this method are laid out, especially the selection of the leaching agents and the preparation of the ore deposit

  5. Formation conditions of uranium minerals in oxidation zone of uranium deposits

    International Nuclear Information System (INIS)

    Li Youzhu

    2005-01-01

    The paper concerns about the summary and classification of hydrothermal uranium deposit with oxidation zone. Based on the summary of observation results of forty uranium deposits located in CIS and Bulgaria which are of different sizes and industrial-genetic types, analysis on available published information concerning oxidation and uranium mineral enrichment in supergenic zone, oxidation zone classification of hydrothermal uranium had been put forward according to the general system of the exogenetic uranium concentration. (authors)

  6. Copper Mountain, Wyoming, intermediate-grade uranium resource assessment project. Final report. National Uranium Resource Evaluation

    International Nuclear Information System (INIS)

    Madson, M.E.; Ludlam, J.R.; Fukui, L.M.

    1982-11-01

    Intermediate-grade uranium resources were delineated and estimated for Eocene and Precambrian host rock environments in the 39.64 mi 2 Copper Mountain, Wyoming, assessment area. Geologic reconnaissance and geochemical, geophysical, petrologic, borehole, and structural data were interpreted and used to develop a genetic model for uranium mineralization in these environments. Development of a structural scoring system and application of computer graphics in a high-confidence control area established the basis for estimations of uranium resources in the total assessment area. 8 figures, 5 tables

  7. Estimation of intermediate grade uranium resources. Final report

    International Nuclear Information System (INIS)

    Lambie, F.W.; Kendall, G.R.; Klahn, L.J.; Davis, J.C.; Harbaugh, J.W.

    1980-12-01

    The purpose of this project is to analyze the technique currently used by DOE to estimate intermediate grade uranium (0.01 to 0.05% U 3 O 8 ) and, if possible, suggest alternatives to improve the accuracy and precision of the estimate. There are three principal conclusions resulting from this study. They relate to the quantity, distribution and sampling of intermediate grade uranium. While the results of this study must be validated further, they indicate that DOE may be underestimating intermediate level reserves by 20 to 30%. Plots of grade of U 3 O 8 versus tonnage of ore and tonnage U 3 O 8 indicate grade-tonnage relationships that are essentially log-linear, at least down to 0.01% U 3 O 8 . Though this is not an unexpected finding, it may provide a technique for reducing the uncertainty of intermediate grade endowment. The results of this study indicate that a much lower drill hole density is necessary for DOE to estimate uranium resources than for a mining company to calculate ore resources. Though errors in local estimates will occur, they will tend to cancel over the entire deposit

  8. Studies on O/M ratio determination in uranium oxide, plutonium oxide and uranium-plutonium mixed oxide

    International Nuclear Information System (INIS)

    Sampath, S.; Chawla, K.L.

    1975-01-01

    Thermogravimetric studies were carried out in unsintered and sintered samples of uranium oxide, plutonium oxide and uranium-plutonium mixed oxide under different atmospheric conditions (air, argon and moist argon/hydrogen). Moisture loss was found to occur below 200 0 C for uranium dioxide samples, upto 700 0 C for sintered plutonium dioxide and negligible for sintered samples. The O/M ratios for non-stoichiometric uranium dioxide (sintered and unsintered), plutonium dioxide and mixed uranium and plutonium oxides (sintered) could be obtained with a precision of +- 0.002. Two reference states UOsub(2.000) and UOsub(2.656) were obtained for uranium dioxide and the reference state MOsub(2.000) was used for other cases. For unsintered plutonium dioxide samples, accurate O/M ratios could not be obtained of overlap of moisture loss with oxygen loss/gain. (author)

  9. Complex defects in the oxidation of uranium

    International Nuclear Information System (INIS)

    MacCrone, R.K.; Sankaran, S.; Shatynski, S.R.; Colmenares, C.A.

    1986-01-01

    We are reporting EPR results obtained with uranium powder samples fully oxidized in dry air, water vapor, and air/water vapor mixtures. The results reported previously are confirmed and additional paramagnetic centers, associated with chemisorbed species, have been identified. The temperature dependence of the g-value for these centers from room temperature to 10K is also reported

  10. Uranium bed oxidation vacuum process system

    International Nuclear Information System (INIS)

    McLeland, H.L.

    1977-01-01

    Deuterium and tritium gases are occluded in uranium powder for release into neutron generator tubes. The uranium powder is contained in stainless steel bottles, termed ''beds.'' If these beds become damaged, the gases must be removed and the uranium oxidized in order not to be flammable before shipment to ERDA disposal grounds. This paper describes the system and methods designed for the controlled degassing and oxidation process. The system utilizes sputter-ion, cryo-sorption and bellows pumps for removing the gases from the heated source bed. Removing the tritium gas is complicated by the shielding effect of helium-3, a byproduct of tritium decay. This effect is minimized by incremental pressure changes, or ''batch'' processing. To prevent runaway exothermic reaction, oxidation of the uranium bed is also done incrementally, or by ''batch'' processing, rather than by continuous flow. The paper discusses in detail the helium-3 shielding effect, leak checks that must be made during processing, bed oxidation, degree of gas depletion, purity of gases sorbed from beds, radioactivity of beds, bed disposal and system renovation

  11. Plutonium oxides and uranium and plutonium mixed oxides. Carbon determination

    International Nuclear Information System (INIS)

    Anon.

    Determination of carbon in plutonium oxides and uranium plutonium mixed oxides, suitable for a carbon content between 20 to 3000 ppm. The sample is roasted in oxygen at 1200 0 C, the carbon dioxide produced by combustion is neutralized by barium hydroxide generated automatically by coulometry [fr

  12. Uranium

    International Nuclear Information System (INIS)

    Hamdoun, N.A.

    2007-01-01

    The article includes a historical preface about uranium, discovery of portability of sequential fission of uranium, uranium existence, basic raw materials, secondary raw materials, uranium's physical and chemical properties, uranium extraction, nuclear fuel cycle, logistics and estimation of the amount of uranium reserves, producing countries of concentrated uranium oxides and percentage of the world's total production, civilian and military uses of uranium. The use of depleted uranium in the Gulf War, the Balkans and Iraq has caused political and environmental effects which are complex, raising problems and questions about the effects that nuclear compounds left on human health and environment.

  13. Uranium Processing Research in Australia [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, J R [Australian Atomic Energy Commission, Coogee, N.S.W. (Australia)

    1967-06-15

    Uranium processing research in Australia has included studies of flotation, magnetic separation, gravity separation, heavy medium separation, atmospheric leaching, multi-stage leaching, alkali leaching, solar heating of leach pulps, jigged-bed resin-in-pulp and solvent-in-pulp extraction. Brief details of the results obtained are given. In general, it can be said that gravity, magnetic and flotation methods are of limited usefulness in the treatment of Australian uranium ores. Alkali leaching seldom gives satisfactory recoveries and multi-stage leaching is expensive. Jigged-bed resin-in-pulp and packed tower solvent-in-pulp extraction systems both show promise, but plant-scale development work is required. Bacterial leaching may be useful in the case of certain low-grade ores. The main difficulties to be overcome, either singly or in combination, in the case of Australian uranium ores not currently considered economically exploitable, are the extremely finely divided state of the uranium mineral, the refractory nature of the uranium mineral and adverse effects due to the gangue minerals present. With respect to known low-grade ores, it would be possible in only a few cases to achieve satisfactory recovery of uranium at reasonable cost by standard treatment methods. (author)

  14. RECOVERY OF URANIUM FROM LOW GRADE URANIUM BEARING ORES

    Science.gov (United States)

    Rhodes, H.B.; Pesold, W.F.; Hirshon, J.M.

    1959-06-01

    Recovery of U, Fe, and Al from Chattanooga shale is described. Ground shale (-4 to +325 mesh) is roasted to remove organic and volatile matter. The heated shale is then reacted with a chlorinating agent (CCl/sub 4/, COCl/sub 2/, Cl, and SCl) at 600 to 1000 C. The metal chloride vapor is separated from entrained solids and then contacted with a liquid alkali metal chloride which removes U. The U is reeovered by cooling and dissolving the bath followed by acidification and solvent extraction. A condensed phase of Al, Fe, and K chlorides is treated to separate Al as alumina by passing through a Fe/sub 2/O/ sub 3/ bed. The remaining FeCl/sub 3/ is oxidized by O/sub 2/ at 1000 C to form Fe/sub 2/O/sub 3/ and Cl/sub 2/. Alternatively, vapor from the U separation step may be passed to a liquid KCl bath at 500 to 650 C. The resulting mixture is oxidized to form Cl/sub 2/ and Fe/sub 2/O/sub 3/ + Al/sub 2/O/sub 3/. The Al and Fe are separated by reaction with NaOH at high temperatures and pressures. (T.R.H.)

  15. Continued Multicolumns Bioleaching for Low Grade Uranium Ore at a Certain Uranium Deposit

    Directory of Open Access Journals (Sweden)

    Gongxin Chen

    2016-01-01

    Full Text Available Bioleaching has lots of advantages compared with traditional heap leaching. In industry, bioleaching of uranium is still facing many problems such as site space, high cost of production, and limited industrial facilities. In this paper, a continued column bioleaching system has been established for leaching a certain uranium ore which contains high fluoride. The analysis of chemical composition of ore shows that the grade of uranium is 0.208%, which is lower than that of other deposits. However, the fluoride content (1.8% of weight is greater than that of other deposits. This can be toxic for bacteria growth in bioleaching progress. In our continued multicolumns bioleaching experiment, the uranium recovery (89.5% of 4th column is greater than those of other columns in 120 days, as well as the acid consumption (33.6 g/kg. These results indicate that continued multicolumns bioleaching technology is suitable for leaching this type of ore. The uranium concentration of PLS can be effectively improved, where uranium recovery can be enhanced by the iron exchange system. Furthermore, this continued multicolumns bioleaching system can effectively utilize the remaining acid of PLS, which can reduce the sulfuric acid consumption. The cost of production of uranium can be reduced and this benefits the environment too.

  16. Determination of uranium in uranium metal, uranium oxides, and uranyl nitrate solutions by potentiometric titration

    International Nuclear Information System (INIS)

    Tucker, H.L.; McElhaney, R.J.

    1983-01-01

    A simple, fast method for the determination of uranium in uranium metal, uranium oxides, and uranyl nitrate solutions has been adapted from the Davies-Gray volumetric method to meet the needs of Y-12. One-gram duplicate aliquots of uranium metal or uranium oxide are dissolved in 1:1 HNO 3 and concentrated H 2 SO 4 to sulfur trioxide fumes, and then diluted to 100-mL volume. Duplicate aliquots are then weighed for analysis. For uranyl nitrate samples, duplicate aliquots containing between 50 and 150 mg of U are weighed and analyzed directly. The weighed aliquot is transferred to a Berzelius beaker; 1.5 M sulfamic acid is added, followed in order by concentrated phosphoric acid, 1 M ferrous sulfate, and (after a 30-second interval) the oxidizing reagent. After a timed 3-minute waiting period, 100 mL of the 0.1% vanadyl sulfate-sulfuric acid mixture is added. The sample is then titrated past its endpoint with standard potassium dichromate, and the endpoint is determined by second derivative techniques on a mV/weight basis

  17. Uranium Oxide Aerosol Transport in Porous Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, Jeremy; Gerlach, David C.; Scheele, Randall D.; Stewart, Mark L.; Reid, Bruce D.; Gauglitz, Phillip A.; Bagaasen, Larry M.; Brown, Charles C.; Iovin, Cristian; Delegard, Calvin H.; Zelenyuk, Alla; Buck, Edgar C.; Riley, Brian J.; Burns, Carolyn A.

    2012-01-23

    The objective of this paper is to investigate the transport of uranium oxide particles that may be present in carbon dioxide (CO2) gas coolant, into the graphite blocks of gas-cooled, graphite moderated reactors. The transport of uranium oxide in the coolant system, and subsequent deposition of this material in the graphite, of such reactors is of interest because it has the potential to influence the application of the Graphite Isotope Ratio Method (GIRM). The GIRM is a technology that has been developed to validate the declared operation of graphite moderated reactors. GIRM exploits isotopic ratio changes that occur in the impurity elements present in the graphite to infer cumulative exposure and hence the reactor’s lifetime cumulative plutonium production. Reference Gesh, et. al., for a more complete discussion on the GIRM technology.

  18. Aqueous dissolution rates of uranium oxides

    International Nuclear Information System (INIS)

    Steward, S.A.; Mones, E.T.

    1994-10-01

    An understanding of the long-term dissolution of waste forms in groundwater is required for the safe disposal of high level nuclear waste in an underground repository. The main routes by which radionuclides could be released from a geological repository are the dissolution and transport processes in groundwater flow. Because uranium dioxide is the primary constituent of spent nuclear fuel, the dissolution of its matrix in spent fuel is considered the rate-limiting step for release of radioactive fission products. The purpose of our work has been to measure the intrinsic dissolution rates of uranium oxides under a variety of well-controlled conditions that are relevant to a repository and allow for modeling. The intermediate oxide phase U 3 O 8 , triuranium octaoxide, is quite stable and known to be present in oxidized spent fuel. The trioxide, UO 3 , has been shown to exist in drip tests on spent fuel. Here we compare the results of essentially identical dissolution experiments performed on depleted U 3 O 8 and dehyrated schoepite or uranium trioxide monohydrate (UO 3 ·H 2 O). These are compared with earlier work on spent fuel and UO 2 under similar conditions

  19. Extraction of uranium low-grade ores from Great Divide Basin, Wyoming. National Uranium Resource Evaluation

    International Nuclear Information System (INIS)

    Judd, J.C.; Nichols, I.L.; Huiatt, J.L.

    1983-04-01

    The US Bureau of Mines is investigating the leachability of carbonaceous uranium ore samples submitted by the DOE under an Interagency Agreement. Studies on eight samples from the Great Divide Basin, Wyoming, are the basis of this report. The uranium content of the eight ore samples ranged from 0.003 to 0.03% U 3 O 8 and contained 0.7 to 45% organic carbon. Experiments were performed to determine the feasibility of extracting uranium using acid leaching, roast-acid leaching and pressure leaching techniques. Acid leaching with 600 lb/ton H 2 SO 4 plus 10 lb/ton NaClO 3 for 18 h at 70 0 C extracted 65 to 83% of the uranium. One sample responded best to a roast-leach treatment. When roasting for 4 h at 500 0 C followed by acid leaching of the calcine using 600 lb/ton H 2 SO 4 , the uranium extraction was 82%. Two of the samples responded best to an oxidative pressure leach for 3 h at 200 0 C under a total pressure of 260 psig; uranium extractions were 78 and 82%

  20. Midwest Joint Venture high-grade uranium mining

    International Nuclear Information System (INIS)

    Fredrickson, H.K.

    1992-01-01

    Midwest Joint Venture (MJV) owns a high-grade uranium deposit in northern Saskatchewan. The deposit is located too deep below surface to be mined economically by open pit methods, and as a consequence, present plans are that it will be mined by underground methods. High-grade uranium ore of the type at MJV, encased in weak, highly altered ground and with radon-rich water inflows, has not before been mined by underground methods. The test mining phase of the project, completed in 1989, had three objectives: To evaluate radiation protection requirements associated with the handling of large quantities of radon-rich water and mining high-grade uranium ore in an underground environment; to investigate the quantity and quality of water inflows into the mine; and, to investigate ground conditions in and around the ore zone as an aid in determining the production mining method to be used. With information gained from the test mining project, a mining method for the production mine has been devised. Level plans have been drawn up, ventilation system designed, pumping arrangements made and methods of ore handling considered. All this is to be done in a manner that will be safe for those doing the work underground. Some of the mining methods planned are felt to be unique in that they are designed to cope with mining problems not known to have been encountered before. New problems underground have required new methods to handle them. Remote drilling, blasting, mucking and backfilling form the basis of the planned mining method

  1. A study on the oxidation behavior of uranium

    International Nuclear Information System (INIS)

    Kim, Tae Kook; Kang, Kweon Ho; Kim, Kil Jeang; Kang, Il Sik; Jung, Kyung Whan

    1998-03-01

    When storing depleted uranium wastes, careful handling is required due to their very high oxidation rates. To determine the oxidation mechanism and oxidation rate of depleted uranium wastes, the most important factors to be considered in their treatment, an experiment was carried out by varying the heating rates of the Air-Controlled Oxidizer. The experiment, showed that depleted uranium wastes are pulverized after complete oxidation because of the density difference and then converted to UO 2 , U 3 O 7 , U 3 O 8 . The grain size of pulverized powder decrease with increased temperature. (author). 30 refs., 5 tabs., 28 figs

  2. Mining the high grade McArthur River uranium deposit

    International Nuclear Information System (INIS)

    Jamieson, B.W.

    2002-01-01

    The McArthur River deposit, discovered in 1988, is recognized as the world's largest, highest grade uranium deposit, with current mineable reserves containing 255 million lb U 3 O 8 at an average grade of 17.33% U 3 O 8 . In addition the project has resources of 228 million pounds U 3 O 8 averaging 12.02% U 3 O 8 . Mining this high-grade ore body presents serious challenges in controlling radiation and in dealing with high water pressures. Experience from the underground exploration programme has provided the information needed to plan the safe mining of the massive Pelite ore zone, which represents the most significant source of ore discovered during the underground drilling programme, with 220 million pounds of U 3 O 8 at an average grade in excess of 17%. Non-entry mining will be used in the high-grade ore zones. Raise boring will be the primary method to safely extract the ore, with all underground development in waste rock to provide radiation shielding. Water will be controlled by grouting and perimeter freezing. The ore cuttings from the raise boring will be ground underground and pumped to surface as slurry, at an average daily production of 150 tonnes. The slurry will be transported to the Key Lake mill and diluted to 4% before processing. The annual production is projected to be 18 million lb U 3 O 8 . The paper focuses on the activities undertaken since discovery, including the initiation of the raise bore mining method utilized to safely mine this high grade ore body. Radiation protection, environmental protection and worker health and safety are discussed in terms of both design and practical implementation. (author)

  3. Oxidative lixiviation of pitchblende and precipitation of uranium with hydrogen

    International Nuclear Information System (INIS)

    Mouret, P.

    1958-01-01

    Earlier work on the preparation of uranium by F.A. Forward and his colleagues has shown the possibilities presented by oxidative lixiviation of ores in a carbonate medium, and the catalytic reduction of uranyl carbonate solutions by hydrogen. The carbonate attack is of considerable interest because of the selectivity of the uranium dissolution, which means it can be applied particularly to the treatment of low grade ores with a reduced consumption of cheap reagents. The subsequent reduction with hydrogen is of the same nature, and not only enables relatively dilute uranyl carbonate solutions to be treated, but also avoids any significant alteration of the attacking solution which can therefore be used again in the lixiviation stage. The experimental work, undertaken at the request of the Commissariat a I'Energie Atomique, was aimed at determining the quantitative characteristics of each of the two stages in order to ascertain their possibilities for industrial application to the principal low grade ores found in France. (author) [fr

  4. Discussion on the interlayer oxidation and uranium metallogenesis in Qianjiadian uranium deposit, Songliao Basin

    International Nuclear Information System (INIS)

    Pang Yaqing; Chen Xiaolin; Fang Xiheng; Sun Ye

    2010-01-01

    Through systematic drill core observation, section contrast and analysis,it is proved that the ore-controlling interlayer oxidation zone of Qianjiadian uranium deposit is mainly composed by the red oxidized sandstone and locally distributed yellow and off-white sandstones. The red sandstone contains charcoal fragments, pyrite, ilmenite, siderite, which have been oxidized intensively, and it can be deduced that their original color was gray and became red due to the oxidization. The distribution of the oxidation zone is mainly controlled by the sedimentary facies,which also controll uranium metallization. The uranium orebodies mainly developed in the thinning or pinch parts of the red oxidation zone in section. On the plans, the uranium mineralization distributes near the front of the red interlayer oxidation zone. (authors)

  5. Development of metallic uranium recovery technology from uranium oxide by Li reduction and electrorefining

    International Nuclear Information System (INIS)

    Tokiwai, Moriyasu; Kawabe, Akihiro; Yuda, Ryouichi; Usami, Tsuyoshi; Fujita, Reiko; Nakamura, Hitoshi; Yahata, Hidetsugu

    2002-01-01

    The purpose of the study is to develop technology for pre-treatment of oxide fuel reprocessing through pyroprocess. In the pre-treatment process, it is necessary to reduce actinide oxide to metallic form. This paper outlines some experimental results of uranium oxide reduction and recovery of refined metallic uranium in electrorefining. Both uranium oxide granules and pellets were used for the experiments. Uranium oxide granules was completely reduced by lithium in several hours at 650degC. Reduced uranium pellets by about 70% provided a simulation of partial reduction for the process flow design. Almost all adherent residues of Li and Li 2 O were successfully washed out with fresh LiCl salt. During electrorefining, metallic uranium deposited on the iron cathode as expected. The recovery efficiencies of metallic uranium from reduced uranium oxide granules and from pellets were about 90% and 50%, respectively. The mass balance data provided the technical bases of Li reduction and refining process flow for design. (author)

  6. Heap bioleaching of uranium from low-grade granite-type ore by mixed acidophilic microbes

    International Nuclear Information System (INIS)

    Xuegang Wang; Zhongkui Zhou

    2017-01-01

    We evaluated uranium bioleaching from low-grade, granite-type uranium ore using mixed acidophilic microbes from uranium mine leachate. A 4854-ton plant-scale heap bioleaching process achieved sustained leaching with a uranium leaching efficiency of 88.3% using a pH of 1.0-2.0 and an Fe"3"+ dosage of 3.0-5.5 g/L. Acid consumption amounted to 25.8 g H_2SO_4 kg"-"1 ore. Uranium bioleaching follows a diffusion-controlled kinetic model with a correlation coefficient of 0.9136. Almost all uranium was dissolved in aqueous solution, except those encapsulated in quartz particles. Therefore, heap bioleaching by mixed acidophilic microbes enables efficient, economical, large-scale recovery of uranium from low-grade ores. (author)

  7. Ultrafiltration evaluation with depleted uranium oxide

    International Nuclear Information System (INIS)

    Weisbrod, K.R.; Schake, A.R.; Morgan, A.N.; Purdy, G.M.; Martinez, H.E.; Nelson, T.O.

    1998-03-01

    Scientists at the Los Alamos National Laboratory Plutonium Facility are using electrodissolution in neutral to alkaline solutions to decontaminate oralloy parts that have surface plutonium contamination. Ultrafiltration of the electrolyte stream removes precipitate so that the electrolyte stream to the decontamination fixture is precipitate free. This report describes small-scale laboratory ultrafiltration experiments that the authors performed to determine conditions necessary for full-scale operation of an ultrafiltration module. Performance was similar to what they observed in the ferric hydroxide system. At 12 psi transmembrane pressure, a shear rate of 12,000 sec -1 was sufficient to sustain membrane permeability. Ultrafiltration of uranium(VI) oxide appears to occur as easily as ultrafiltration of ferric hydroxide. Considering the success reported in this study, the authors plan to add ultrafiltration to the next decontamination system for oralloy parts

  8. Oxidation-extraction of uranium from wet-process phosphoric acid

    International Nuclear Information System (INIS)

    Lawes, B.C.

    1985-01-01

    The invention involves an improvement to the reductive stripping process for recovering uranium values from wet-process phosphoric acid solution, where uranium in the solution is oxidized to uranium (VI) oxidation state and then extracted from the solution by contact with a water immiscible organic solvent, by adding sufficient oxidant, hydrogen peroxide, to obtain greater than 90 percent conversion of the uranium to the uranium (VI) oxidation state to the phosphoric acid solution and simultaneously extracting the uranium (VI)

  9. Sintering uranium oxide using a preheating step

    International Nuclear Information System (INIS)

    Jensen, N.J.; Nivas, Y.; Packard, D.R.

    1977-01-01

    Compacted pellets of uranium oxide or uranium oxide with one or more additives are heated in a kiln in a process having a preheating step, a sintering step, a reduction step, and a cooling step in a controlled atmosphere. The process is practiced to give a range of temperature and atmosphere conditions for obtaining optimum fluoride removal from the compacted pellets along with optimum sintering in a single process. The preheating step of this process is conducted in a temperature range of about 600 0 to about 900 0 C and the pellets are held for at least twenty min, and preferably about 60 min, in an atmosphere having a composition in the range of about 10 to about 75 vol % hydrogen with the balance being carbon dioxide. The sintering step is conducted at a temperature in the range of about 900 0 C to 1500 0 C in the presence of an atmosphere having a composition in the range of about 0.5 to about 90 vol % hydrogen with the balance being carbon dioxide. The reduction step reduces the oxygen to metal ratio of the pellets to a range of about 1.98 to 2.10:1 and this is accomplished by gradually cooling the pellets for about 30 to about 120 min from the temperature of the sintering step to about 1100 0 C in an atmosphere of about 10 to 90 vol % hydrogen with the balance being carbon dioxide. Thereafter the pellets are cooled to about 100 0 C under a protective atmosphere, and in one preferred practice the same atmosphere used in the reduction step is used in the cooling step. The preheating, sintering and reduction steps may also be conducted with their respective atmospheres having an initial additional component of water vapor and the water vapor can comprise up to about 20 vol %

  10. Ferrite grade iron oxides from ore rejects

    Indian Academy of Sciences (India)

    Iron oxyhydroxides and hydroxides were synthesized from chemically beneficiated high SiO2/Al2O3 low-grade iron ore (57.49% Fe2O3) rejects and heated to get iron oxides of 96–99.73% purity. The infrared band positions, isothermal weight loss and thermogravimetric and chemical analysis established the chemical ...

  11. Treatment of uranium turning with the controllable oxidizing process

    International Nuclear Information System (INIS)

    Shen Bingyi; Zhang Yonggang; Zhen Huikuan

    1989-02-01

    The concept, procedure and safety measures of the controllable oxidizing for uranium turning is described. The feasibility study on technological process has been made. The process provided several advantages such as: simplicity of operation, no pollution environment, safety, high efficiency and low energy consumption. The process can yield nuclear pure uranium dioxide under making no use of a great number of chemical reagent. It may supply raw material for fluoration and provide a simply method of treatment for safe store of uranium turning

  12. Effect of CO on surface oxidation of uranium metal

    International Nuclear Information System (INIS)

    Wang, X.; Fu, Y.; Xie, R.

    1997-01-01

    The surface reactions of uranium metal with carbon monoxide at 25 and 200 deg C have been studied by X-ray photoelectron spectroscopy (XPS);respectively. Adsorption of carbon monoxide on the surface layer of uranium metal leads to partial reduction of surface oxide and results in U4f photoelectron peak shifting to the lower binding energy. The content of oxygen in the surface oxide is decreased and O1s/O4f ratio decreases with increasing the exposure of carbon monoxide. The investigation indicates the surface layer of uranium metal has resistance to further oxidation in the atmosphere of carbon monoxide. (author)

  13. Oxidizing attack process of uranium ore by a carbonated liquor

    International Nuclear Information System (INIS)

    Maurel, Pierre; Nicolas, Francois.

    1981-01-01

    A continuous process for digesting a uraniferous ore by oxidation with a recycling aqueous liquor containing alkaline carbonates and bicarbonates in solution as well as uranium in a concentration close to its solubility limit at digestion temperature, and of recuperation of the precipitated uranium within the solid phase remaining after digestion. The digestion is carried out by spraying oxygen into the hot reactional medium in order not only to permit oxidation of the uranium and its solubilization but also to ensure that the sulphides of impurities and organic substances present in the ore are oxidized [fr

  14. Uranium induces oxidative stress in lung epithelial cells

    International Nuclear Information System (INIS)

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

    2007-01-01

    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. (orig.)

  15. Determination of uranium in uranyl nitrate solutions of nuclear grade quality - Gravimetric method

    International Nuclear Information System (INIS)

    1990-01-01

    This international Standard specifies a precise and accurate gravimetric method for determining the uranium content in uranyl nitrate product solutions of nuclear grade quality at concentrations above 100 g/l of uranium. Non-volatile impurities influence the accuracy of the method. Uranyl nitrate is converted into uranium octoxide (U 3 O 8 ) by ignition in air to constant mass at 900 deg. C ± 10 deg. C. Calculation of the uranium content in the sample using a gravimetric conversion factor which depends on the isotopic composition of the uranium. The isotopic composition is determined by mass spectrometry

  16. Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Ross, Kyle W. (Los Alamos National Laboratory, Los Alamos, NM); Smith, James Dean; Longmire, Pamela

    2010-04-01

    The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.

  17. Production and analysis of ultradispersed uranium oxide powders

    Science.gov (United States)

    Zajogin, A. P.; Komyak, A. I.; Umreiko, D. S.; Umreiko, S. D.

    2010-05-01

    Spectroscopic studies are made of the laser plasma formed near the surface of a porous body containing nanoquantities of uranium compounds which is irradiated by two successive laser pulses. The feasibility of using laser chemical methods for obtaining nanoclusters of uranium oxide particles in the volume of a porous body and the simultaneous possibility of determining the uranium content with good sensitivity are demonstrated. The thermochemical and spectral characteristics of the analogs of their compounds with chlorine are determined and studied. The possibility of producing uranium dioxides under ordinary conditions and their analysis in the reaction products is demonstrated.

  18. Nuclear fuel technology - Determination of uranium in uranyl nitrate solutions of nuclear grade quality - Gravimetric method

    International Nuclear Information System (INIS)

    2003-01-01

    This International Standard specifies a precise and accurate gravimetric method for determining the mass fraction of uranium in uranyl nitrate solutions of nuclear grade quality containing more than 100 g/kg of uranium. Non-volatile impurities influence the accuracy of the method

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

    Science.gov (United States)

    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.

  20. Precision determination of uranium in uranium oxide by constant-current coulometry

    International Nuclear Information System (INIS)

    Xu Laili; Wang Chunhuan.

    1990-01-01

    A method of constant-current coulometric titration for determination of uranium in uranium oxide is described. This method involves preliminary reduction of U (VI) in H 2 SO 4 -H 3 PO 4 medium by Cr (II) as a reductant, followed by air oxidation of excess of Cr (II), addition of solid K 2 Cr 2 O 7 in quantity slightly more than that of the required for quantitative oxidation of U (IV) and final titration of excess of K 2 Cr 2 O 7 with the electrogenerated Fe (II). The endpoint is determined amperometrically. The effect of various factors on the sample treatment and reduction-oxidation processes has been examined. The precision of the method as indicated by the standard deviation of an individual observation is less than 0.01% for l gram uranium oxide

  1. Titrimetric determination of uranium in low-grade ores by the ferrous ion-phosphoric acid reduction method

    International Nuclear Information System (INIS)

    Hitchen, A.; Zechanowitsch, G.

    1980-01-01

    The modification and extension of the U.S.A.E.C. ferrous ion-phosphoric acid reduction method for the determination of uranium in high-grade or relatively pure material to a method for the determination of uranium with a high accuracy and precision, in ores containing 0.004 to 7% U is described. It is simple, rapid and requires no prior separations from elements that, in other methods, frequently interfere. For sample materials having very high concentrations of interfering elements, a prior concentration step using extraction with tri-n-octylphosphine oxide is described, but it is shown that, for most low-grade ores, this step is unnecessary. (author)

  2. Formation of corrosion-resistant oxide film on uranium

    International Nuclear Information System (INIS)

    Petit, G.S.

    1976-01-01

    A vacuum heat-treatment method was developed for coating metallic uranium with an adherent protective film of uranium oxide. The film is prepared by vacuum heat-treating the metallic uranium at 625 0 C for 1 h while controlling the amount of oxygen being metered into the furnace. Uranium coupons with the protective film were exposed for several hundred hours in a corrosion test bath at 95 0 C and 100 percent RH without corroding. Film thicknesses ranging from 5 to 25 μm (0.0002 to 0.001 in.) were prepared and corrosion tested; the film thickness can be controlled to less than +-2.5 μm (+-0.0001 in.). The oxide film is hard, nonwetting, and very adherent. The resulting surface finish of the metal is equivalent to that of the original finish. The advantages of the oxide films over other protective coatings are given. 12 fig

  3. Using NJOY99 and MCNP4B2 to Estimate the Radiation Damage Displacements per Atom per Second in Steel Within the Boiling Water Reactor Core Shroud and Vessel Wall from Reactor-Grade Mixed-Oxide/Uranium Oxide Fuel for the Nuclear Power Plant at Laguna Verde, Veracruz, Mexico

    International Nuclear Information System (INIS)

    Vickers, Lisa

    2003-01-01

    The government of Mexico has expressed interest in utilizing the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18 to 30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons.There is concern that a core with a fraction of MOX fuel (i.e., increased 239 Pu wt%) would increase the radiation damage displacements per atom per second (dpa-s -1 ) in steel within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation damage within the core shroud and vessel wall is a concern because of the potentially adverse affect to personnel and public safety, environment, and operating life of the reactor.The primary uniqueness of this paper is the computation of radiation damage (dpa-s -1 ) using NJOY99-processed cross sections for steel within the core shroud and vessel wall. Specifically, the unique radiation damage results are several orders of magnitude greater than results of previous works. In addition, the conclusion of this paper was that the addition of the maximum fraction of one-third MOX fuel to the LV1 BWR core did significantly increase the radiation damage in steel within the core shroud and vessel wall such that without mitigation of radiation damage by periodic thermal annealing or reduction in operating parameters such as neutron fluence, core temperature, and pressure, it posed a potentially adverse affect to the personnel and public safety, environment, and operating life of the reactor

  4. Study of reactions between uranium-plutonium mixed oxide and uranium nitride and between uranium oxide and uranium nitride; Etude des reactions entre l`oxyde mixte d`uranium-plutonium et le nitrure d`uranium et entre l`oxyde d`uranium et le nitrure d`uranium

    Energy Technology Data Exchange (ETDEWEB)

    Lecraz, C

    1993-06-11

    A new type of combustible elements which is a mixture of uranium nitride and uranium-plutonium oxide could be used for Quick Neutrons Reactors. Three different studies have been made on the one hand on the reactions between uranium nitride (UN) and uranium-plutonium mixed oxide (U,Pu)O{sub 2}, on the other hand on these between UN and uranium oxide UO{sub 2}. They show a sizeable reaction between nitride and oxide for the studied temperatures range (1573 K to 1973 K). This reaction forms a oxynitride compound, MO{sub x} N{sub y} with M=U or M=(U,Pu), whose crystalline structure is similar to oxide`s. Solubility of nitride in both oxides is studied, as the reaction kinetics. (TEC). 32 refs., 48 figs., 22 tabs.

  5. Kinetics of the reduction of uranium oxide catalysts

    International Nuclear Information System (INIS)

    Heynen, H.W.G.; Camp-van Berkel, M.M.; Bann, H.S. van der

    1977-01-01

    The reduction of uranium oxide and uranium oxide on alumina catalysts by ethylbenzene and by hydrogen has been studied in a thermobalance. Ethylbenzene mole fractions between 0.0026 and 0.052 and hydrogen mole fractions between 0.1 and 0.6 were applied at temperatures of 425--530 0 C. During the reduction the uranium oxides are converted into UO 2 . The rate of reduction of pure uranium oxide appears to be constant in the composition region UO/sub 2.6/-UO/sub 2.25/. The extent of this region is independent of the concentration of the reducing agents and of the reaction temperature. The constant rate is explained in terms of a constant oxygen pressure which is in equilibrium with the two solid phases, U 3 O/sub 8-x/ and U 4 O 9 . The reduction rate is first order in hydrogen and zero order in ethylbenzene with activation energies of 120 and 190 kJ mol -1 , respectively. Oxygen diffusion through the lattice is probably not rate limiting. The reduction behavior of uranium oxide on alumina is different from that of pure uranium oxide; the rate of reduction continuously decreases with increasing degree of reduction. An explanation for this behavior has been given by visualizing this catalyst as a set of isolated uranium oxide crystallites with a relative wide variation of diameters, in an alumina matrix. At the beginning of the reduction, carbon dioxide and water are the only reaction products. Thereafter, benzene is found as well and, finally, at U/O ratios below 2.25, styrene also appears in the reactor outlet

  6. Heat-induced redistribution of surface oxide in uranium

    International Nuclear Information System (INIS)

    Swissa, E.; Shamir, N.; Bloch, J.; Mintz, M.H.; Israel Atomic Energy Commission, Beersheba. Nuclear Research Center-Negev)

    1990-01-01

    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 450deg 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 800deg C. The activation energy obtained was E a =15.4±1.9 kcal/mole and the pre-exponential factor, D 0 =1.1x10 -8 cm 2 /s. An internal oxidation mechanism is proposed to explain the results. (orig.)

  7. Heat-induced redistribution of surface oxide in uranium

    Science.gov (United States)

    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.

  8. Standard test methods for chemical and mass spectrometric analysis of nuclear-grade gadolinium oxide (Gd2O3) powder

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2006-01-01

    1.1 These test methods cover procedures for the chemical and mass spectrometric analysis of nuclear-grade gadolinium oxide powders to determine compliance with specifications. 1.2 The analytical procedures appear in the following order: Sections Carbon by Direct CombustionThermal Conductivity C1408 Test Method for Carbon (Total) in Uranium Oxide Powders and Pellets By Direct Combustion-Infrared Detection Method Total Chlorine and Fluorine by Pyrohydrolysis Ion Selective Electrode C1502 Test Method for Determination of Total Chlorine and Fluorine in Uranium Dioxide and Gadolinium Oxide Loss of Weight on Ignition 7-13 Sulfur by CombustionIodometric Titration Impurity Elements by a Spark-Source Mass Spectrographic C761 Test Methods for Chemical, Mass Spectrometric, Spectrochemical,Nuclear, and Radiochemical Analysis of Uranium Hexafluoride C1287 Test Method for Determination of Impurities In Uranium Dioxide By Inductively Coupled Plasma Mass Spectrometry Gadolinium Content in Gadolinium Oxid...

  9. Extraction separation studies of uranium(VI) by amine oxides

    International Nuclear Information System (INIS)

    Ejaz, M.

    1975-01-01

    The extraction of uranium(VI) by two amine oxides, 4-(5-nonyl)pyridine oxide and trioctylamine oxide has been studied. The extraction behavior of these two N-oxides is compared. The dependence of extraction on the type of amine oxide and acid, nature of organic diluent, and amine oxide concentration has been investigated. The influence of the concentration of the metal and salting-out agents is described. The possible mechanism of extraction is discussed in the light of the results of extraction isotherms, loading radiodata, and log-log plots of amine oxide concentration vs distribution ratio. The separation factors for a number of metal ions are reported, and the separation of uranium from some fission elements has also been achieved

  10. Simulation of uranium and plutonium oxides compounds obtained in plasma

    Science.gov (United States)

    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.

  11. Report on the feasibility of the in situ radiometric determination of uranium grade in Witwatersrand gold and uranium mines

    International Nuclear Information System (INIS)

    Smit, C.J.B.; Wesolinski, E.S.; Corner, B.

    1982-08-01

    The chip-sampling technique currently employed by the South African gold and uranium-mining industry, for the prediction of face grade, has several drawbacks, namely: 1) it is labour-intensive; 2) sample volumes are often unrepresentative and prone to human error; and 3) the uranium mineralisation may be very erratic along the reef. In situ radiometric assaying for uranium along the reef, on the other hand, is a rapid, essentially one-man operation, enabling a much larger and hence a more representative sample volume to be measured. The high radiometric background inherent in any uranium mine necessitates some form of high-density shielding in order to facilitate quantitative in situ assaying. This report, therefore, briefly outlines the origin, nature, detection and shielding of gamma rays. Results obtained with a frontally shielded total-count instrument showed that radiometric estimates of uranium grade are comparable to those obtained by batch mining and can be used for the prediction of face grades, provided that the ore is in radiometric equilibrium and that thorium and potassium are either not present, or vary sympathetically with the uranium grade. Spectral analysis showed, however, that these circumstances will also permit the use of a collimated (side-shielded) detector of acceptable weight, provided that only the low-energy portion of the spectrum is measured. The advantages of a collimated detector over a frontally shielded detector are also noteworthy, viz.: 1) only one reading is taken per sample point rather than two, as is the case with the frontally shielded system, thus improving counting statistics; and 2) the shielding is permanently fixed to the detector. Comprehensive design considerations for a compact, portable instrument are suggested and methods for determining background radiation as applicable to a collimated detector are described

  12. Studies on the fluorination of tri uranium octa oxide to Uranium tetrafluoride

    Energy Technology Data Exchange (ETDEWEB)

    Rofail, N H; Elfekey, S A [Nuclear chemistry department, hot laboratories centre, atomic energy authority, Cairo, (Egypt)

    1995-10-01

    Uranium tetrafluoride suitable for both uranium metal and hexafluoride preparations, was prepared by fluorination of U{sub 3} O{sub 8} with C F{sub 2} Cl{sub 2}. It was found that the oct oxide must have certain physical and chemical specifications to satisfy the specifications needed for subsequent operations. X-ray diffraction analysis, infra red investigations and chemical analysis confirm that the obtained uranium tetrafluoride contains more than 97% of U F{sub 4} with tap density equals to 3.5 g/cc. 3 FIGS., 2 TABS.

  13. The ignitability potential of uranium {open_quotes}roaster oxide{close_quotes}

    Energy Technology Data Exchange (ETDEWEB)

    Stakebake, J.L.

    1994-11-01

    The oxidation of uranium to form Uranium `roaster oxide` was investigated with respect to concerns of unreacted metal remaining in the roaster oxide matrix. It was found that ignition of unreacted uranium chips in the roaster oxide as synthesized is unlikely under normal storage conditions.

  14. Studies on yttrium oxide coatings for corrosion protection against molten uranium

    International Nuclear Information System (INIS)

    Chakravarthy, Y.; Bhandari, Subhankar; Pragatheeswaran; Thiyagarajan, T.K.; Ananthapadmanabhan, P.V.; Das, A.K.; Kumar, Jay; Kutty, T.R.G.

    2012-01-01

    Yttrium oxide is resistant to corrosion by molten uranium and its alloys. Yttrium oxide is recommended as a protective oxide layer on graphite and metal components used for melting and processing uranium and its alloys. This paper presents studies on the efficacy of plasma sprayed yttrium oxide coatings for barrier applications against molten uranium

  15. Grade control and the determination of ore reserves at a low-grade uranium mine

    International Nuclear Information System (INIS)

    Reid, J.A.F.; Robertson, B.

    1982-01-01

    In 1966 an intensive exploration programme was conducted by Rio Tinto South Africa Ltd, involving airborne and ground radiometric surveys, percussion and diamond drilling, and pilot-plant tests. This investigation established the presence of a large deposit of low-grade uranium that could be mined on a large scale by open-pitting. Soon after production started in June 1974, it was realized that the original presentation of the ore reserves did not give a true reflection of the orebody. A comparison between the estimates and the ore mined from reserve blocks over the period 1975 to 1977 showed that the reserve grades were over-optimistic. In 1977 Rio Tinto Zinc Consultants were commissioned to re-evaluate the ore reserves from exploration diamond-drill data with the help of The Centre de Geostatistique at Fontainebleau, France. A new global reserve using disjunctive kriging techniques was developed, and is now the long-term reserve used for planning purposes at the Roessing Mine. For short-term planning, ore-reserve data are replaced by information from an intermediate drilling programme and blast-hole assays if these are available. A computerized short-term planning system has been established that reconciles the ore mined against the reserve blocks, as well as a system by which composited blast-hole assays are analysed for grade-control purposes. Owing to the complexity of the orebody, additional monitoring is achieved by the scanning of each truckload of ore with a group of scintillation crystal heads that evaluate the material mined

  16. Growth characteristics of a strain of iron-oxidizing bacterium and its application in bioleaching of uranium ores

    International Nuclear Information System (INIS)

    Zhang Rui; Liu Yajie; Gao Feng; Xu Lingling

    2008-01-01

    05B is a strain of iron-oxidizing bacterium which separated from a uranium ore. The effect of temperature, initial pH, inoculation amount and initial total iron concentration on the strain's growth and activities in bioleaching of uranium ores are studied. The results show that the optimum growth temperature is 40-45 degree C, the optimum inoculation pH value being 1.5-1.7, the optimum initial inoculation amount being 10%-20%, and the initial total iron concentration being not more than 5 g/L. 05B is fit for leaching of low grade uranium ores. (authors)

  17. Pentavalent Uranium Chemistry - Synthetic Pursuit Of A Rare Oxidation State

    Energy Technology Data Exchange (ETDEWEB)

    Graves, Christopher R [Los Alamos National Laboratory; Kiplinger, Jaqueline L [Los Alamos National Laboratory

    2009-01-01

    This feature article presents a comprehensive overview of pentavalent uranium systems in non-aqueous solution with a focus on the various synthetic avenues employed to access this unusual and very important oxidation state. Selected characterization data and theoretical aspects are also included. The purpose is to provide a perspective on this rapidly evolving field and identify new possibilities for future developments in pentavalent uranium chemistry.

  18. Fundamental study on recovery uranium oxide from HEPA filters

    International Nuclear Information System (INIS)

    Izumida, T.; Noguchi, Y.

    1993-01-01

    Large numbers of spent HEPA filters are produced at uranium fuel fabrication facilities. Uranium oxide particles have been collected on these filters. Then, a spent HEPA filter treatment system was developed from the viewpoint of recovering the UO 2 and minimizing the volume. The system consists of a mechanical separation process and a chemical dissolution process. This paper describes the results of fundamental experiments on recovering UO 2 from HEPA filters

  19. Metallization of uranium oxide powders by lithium reduction

    International Nuclear Information System (INIS)

    Kim, I. S.; Seo, J. S.; Oh, S. C.; Hong, S. S.; Lee, W. K.

    2002-01-01

    Laboratory scale experiments on the reduction of uranium oxide powders into metal by lithium were performed in order to determine the equipment setup and optimum operation conditions. The method of filtration using the porous magnesia filter was introduced to recover uranium metal powders produced. Based on the laboratory scale experimental results, mock-up scale (20 kg U/batch) metallizer was designed and made. The applicability to the metallization process was estimated with respect to the thermal stability of the porous magnesia filter in the high temperature molten salt, the filtration of the fine uranium metal powders, and the operability of the equipment

  20. The use of radiometric-logging techniques to determine uranium grade in certain mineralised Karoo boreholes

    International Nuclear Information System (INIS)

    Corner, B.; De Beer, G.P.

    1976-05-01

    During the period September-October 1975, 22 mineralised boreholes in nine different Karoo uranium occurrences were logged radiometrically with the aim of determining to what accuracy the actual uranium grade could be predicted from the gamma logs. The true uranium grades of the mineralised zones logged were known from existing chemical analyses. The results showed that the uranium grades could be predicted to an accuracy of better than 10% through the use of gamma-logging equipment calibrated at Pelindaba, provided that the ore was in equilibrium and that little or no thorium was present. Disequilibrium is, however, prevalent in the Karoo, and in the holes logged it occurred by depletion of uranium relative to its gamma-emitting daughter products. Such effects were mostly confined to the zone above the water table, and it is concluded that for Karoo-type occurrences, the high radiometric background levels observed over extended distances in some boreholes were indicative of radon-gas buildup, and hence of disequilibrium. It is further concluded that radiometric borehole logging can largely replace chemical analyses in the determination of uranium grade for ore-reserve calculations, although chemical checks for disequilibrium would always be necessary [af

  1. Sustainability of Water Cooled Reactors - Energy Balance for Low Grade Uranium Resources

    International Nuclear Information System (INIS)

    Strupczewski, A.

    2011-01-01

    The opponents of nuclear power claim that as uranium resources get exhausted the energy needed to mine low grade uranium ore will be larger than the energy that can be obtained from fission in a nuclear power plant. This would result in loss of sustainability of nuclear power, with the negative energy balance expected within the next 40-60 years. Since the opponents state clearly that the ore containing less than 0.013% U 3 O 8 cannot yield positive energy balance, the study of the Institute of Atomic Energy in Poland referenced three mines of decreasing ore grade: Ranger 0.234% U 3 O 8 , Rossing 0.028% U 3 O 8 and Trekkopje 0.00126% U 3 O 8 , that is with ore grade below the postulated cut off value. The study considered total energy needs for uranium mining, including not only electricity needed for mining and milling, for water treatment and delivery, but also fuel for transportation and ore crushing, explosives for rock blasting, chemicals for uranium leaching and the energy needed for mine reclamation after completed exploitation. It has been shown that the energy estimates of nuclear opponents are wrong for Ranger mine and go off much further for the mines with lower uranium ore grades. The reasons for erroneous reasoning of nuclear opponents have been found. Their errors arise from treating the uranium ore deposits as if their layout and properties were the same as those of uranium ore mined in the US in the 70-ies. This results in an oversimplified formula, which yields large errors when the thickness of the overlayer is less than it was in the US. In addition the energy needs claimed for mine reclamation are much too high. The study showed that the energy needed for very low grade uranium ore mining and milling increases but the overall energy balance of the nuclear fuel cycle remains strongly positive. (author)

  2. The Resonance Absorption of Uranium Metal and Oxide

    Energy Technology Data Exchange (ETDEWEB)

    Hellstrand, E; Lundgren, G

    1962-06-15

    The resonance integrals for uranium metal and uranium oxide have been determined for a 1/E flux. The following results were obtained Metal RI 2.95 + 25.8{radical}(S/M); Oxide RI = 4.15 + 26.6{radical}(S/M). The oxide value agrees with the expression found earlier at this laboratory. But the result for the metal is 4. 5 % larger than the earlier one. In addition, the resonance absorption in a R1 fuel rod has been compared with that for a cadmium-covered rod placed in an approximate cell boundary flux. The former came out 3 % larger than the latter. A comparison of the fuel rod absorption with that for a 1/E flux yields a corresponding figure of 7 %. The neutron flux was monitored below the lowest resonance in uranium.

  3. Analysis of impurities in uranium oxide by ICP-MS

    International Nuclear Information System (INIS)

    Paul, M.; Hager, J.

    1989-01-01

    The method offers detection limits of less than 0.1 μg/g in solid samples and a wide linear dynamic working range and allows a fast sample throughput. The poster summarizes recent authors' experience with the analysis of high purity uranium oxide. The samples are analyzed using both conventional solution analysis and solid sampling method, using a new powerful laser sampling system attached to the ICP mass spectrometer. The total analysis time required for one uranium oxide sample is about 4 minutes including sample handling, data acquisition and processing. The results achieved from semiquantitative and quantitative analysis applying both techniques are in good agreement with the reference values. (author)

  4. Standard test method for analysis of isotopic composition of uranium in nuclear-grade fuel material by quadrupole inductively coupled plasma-mass spectrometry

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2000-01-01

    1.1 This test method is applicable to the determination of the isotopic composition of uranium (U) in nuclear-grade fuel material. The following isotopic weight percentages are determined using a quadrupole inductively coupled plasma-mass spectrometer (Q-ICP-MS): 233U, 234U, 235U, 236U, and 238U. The analysis can be performed on various material matrices after acid dissolution and sample dilution into water or dilute nitric (HNO3) acid. These materials include: fuel product, uranium oxide, uranium oxide alloys, uranyl nitrate (UNH) crystals, and solutions. The sample preparation discussed in this test method focuses on fuel product material but may be used for uranium oxide or a uranium oxide alloy. Other preparation techniques may be used and some references are given. Purification of the uranium by anion-exchange extraction is not required for this test method, as it is required by other test methods such as radiochemistry and thermal ionization mass spectroscopy (TIMS). This test method is also described i...

  5. Comparative study of the oxidation of various qualities of uranium in carbon dioxide at high temperatures

    International Nuclear Information System (INIS)

    Desrues, R.; Paidassi, J.

    1965-01-01

    Uranium samples of six different qualities were subjected, in the temperature range 400 - 1000 C, to the action of carbon dioxide carefully purified to eliminate oxygen and water vapour; the resulting oxidation was followed micro-graphically and also (but only in the range 400 - 700 C) gravimetrically using an Ugine-Eyraud microbalance. A comparison of the results leads to the following 3 observations. First, the oxidation of the six uraniums studied obeys a linear law, (followed at 700 C by an accelerating law). The rates of reaction differ by a maximum of 100 per cent, the higher purity grades being oxidized more slowly except at 700 C when the reverse is true. Secondly, simultaneously with the growth, of an approximately uniform film of uranium dioxide on the metal, there occurs a localized attack in the form of blisters in the immediate neighbourhood of the monocarbide inclusions in the uranium. The relative importance of this attack is greater for lower oxidation temperatures and for a larger size, number and inequality of distribution of the inclusions, that is to say for higher carbon concentrations in the uranium (which have values from 7 to 1000 ppm in our tests). Thirdly, for oxidation temperatures above 600 C blistering is much less pronounced, but at 700 C the beginning of a general deformation of the sample occurs, which, above 750 C, becomes much greater; this leads to an acceleration of the reaction rate with respect to the linear law. In view of the over-heating, the sample must already be in the γ-phase which is particularly easily deformed; furthermore this expansion phenomenon is more pronounced when the sample is more plastic and therefore purer. (authors) [fr

  6. Uranium Metal to Oxide Conversion by Air Oxidation –Process Development

    Energy Technology Data Exchange (ETDEWEB)

    Duncan, A

    2001-12-31

    Published technical information for the process of metal-to-oxide conversion of uranium components has been reviewed and summarized for the purpose of supporting critical decisions for new processes and facilities for the Y-12 National Security Complex. The science of uranium oxidation under low, intermediate, and high temperature conditions is reviewed. A process and system concept is outlined and process parameters identified for uranium oxide production rates. Recommendations for additional investigations to support a conceptual design of a new facility are outlined.

  7. The impact of new technology on the economics of uranium production from low-grade ores

    International Nuclear Information System (INIS)

    Simonsen, H.A.; Boydell, D.W.; James, H.E.

    1981-01-01

    The subject is discussed under the following headings: influence of a depressed market on uranium supply from low-grade ores; potential areas for a reduction in uranium ore processing costs; in-situ leaching (solution mining; heap leaching; resin-in-pulp; solvent-in-pulp; belt filtration; continuous ion exchange; solvent extraction); preconcentration (upgrading of coarse rock; upgrading in the mill; wet high-intensity magnetic separation; flotation); summary and conclusions. (U.K.)

  8. Transformations of highly enriched uranium into metal or oxide

    International Nuclear Information System (INIS)

    Nollet, P.; Sarrat, P.

    1964-01-01

    The enriched uranium workshops in Cadarache have a double purpose on the one hand to convert uranium hexafluoride into metal or oxide, and on the other hand to recover the uranium contained in scrap materials produced in the different metallurgical transformations. The principles that have been adopted for the design and safety of these workshops are reported. The nuclear safety is based on the geometrical limitations of the processing vessels. To establish the processes and the technology of these workshops, many studies have been made since 1960, some of which have led to original achievements. The uranium hexafluoride of high isotopic enrichment is converted either by injection of the gas into ammonia or by an original process of direct hydrogen reduction to uranium tetrafluoride. The uranium contained m uranium-zirconium metal scrap can be recovered by combustion with hydrogen chloride followed treatment of the uranium chloride by fluorine in order to obtain the uranium in the hexafluoride state. Recovery of the uranium contained m various scrap materials is obtained by a conventional refining process combustion of metallic scrap, nitric acid dissolution of the oxide, solvent purification by tributyl phosphate, ammonium diuranate precipitation, calcining, reduction and hydro fluorination into uranium tetrafluoride, bomb reduction by calcium and slag treatment. Two separate workshops operate along these lines one takes care of the uranium with an isotopic enrichment of up to 3 p. 100, the other handles the high enrichments. The handling of each step of this process, bearing in mind the necessity for nuclear safety, has raised some special technological problems and has led to the conception of new apparatus, in particular the roasting furnace for metal turnings, the nitric acid dissolution unit, the continuous precipitator and ever safe filter and dryer for ammonium diuranate, the reduction and hydro fluorination furnace and the slag recovery apparatus These are

  9. Amenability of low-grade uranium towards column bioleaching by acidithiobacillus ferrooxidans

    International Nuclear Information System (INIS)

    Abhilash; Mehta, K.D.; Kumar, V.; Pandey, B.D.; Tamrakar, P.K.

    2007-01-01

    R and D studies were carried out at NML using Acidithiobacillus ferrooxidans (Ac.Tf) in column for the bio-recovery of uranium from the low-grade uranium ore containing 0.024% U 3 O 8 of Turamdih mines, Singhbhum. A recovery of 55.48% uranium was obtained in bio-leaching as against ∼ 44.9% in sterile control in 30 days at 1.7 pH in a column containing 2.5kg ore of particle size mainly in the range 5-1mm. In the large scale column, leaching with 80kg ore of particle size ∼ 0.5cm, uranium bio-recovery was found to be 69.8% in comparison to a recovery of 55% in control set at 1.7 pH in 50 days. The uranium recoveries followed indirect leaching mechanism. (author)

  10. Remote sensing technology prospecting methods of interlayer oxidation zone type sandstone uranium deposit in Yili basin

    International Nuclear Information System (INIS)

    Huang Xianfang; Huang Shutao; Pan Wei; Feng Jie; Liu Dechang; Zhang Jingbo; Xuan Yanxiu; Rui Benshan

    1998-12-01

    Taking Yili Basin as an example, remote sensing technology and method of interlayer oxidation zone type sandstone uranium deposit have systematically been summarized. Firstly, principle, methods and procedures of the second development of scientific experimental satellite photograph have been elaborated in detail. Three dimensional stereo simulation, display, and multi-parameters extraction have been recommended. Secondarily, the research is focused on prospective section image features in different type images and their geological implications and on establishing recognition keys of promising areas. Finally, based on above research results, three graded predictions, i.e. regional prospect, promising sections and favourable location in the deposit have been made step by step and reconnaissance and prospecting range are gradually reduced. The practice has indicated that breakthrough progress has been made in application to prospect prognosis of interlayer oxidation zone type sandstone uranium deposit and good verified results have been obtained

  11. Reaction of hydrogen peroxide with uranium zirconium oxide solid solution - Zirconium hinders oxidative uranium dissolution

    Science.gov (United States)

    Kumagai, Yuta; Takano, Masahide; Watanabe, Masayuki

    2017-12-01

    We studied oxidative dissolution of uranium and zirconium oxide [(U,Zr)O2] in aqueous H2O2 solution to estimate (U,Zr)O2 stability to interfacial reactions with H2O2. Studies on the interfacial reactions are essential for anticipating how a (U,Zr)O2-based molten fuel may chemically degrade after a severe accident. The fuel's high radioactivity induces water radiolysis and continuous H2O2 generation. Subsequent reaction of the fuel with H2O2 may oxidize the fuel surface and facilitate U dissolution. We conducted our experiments with (U,Zr)O2 powder (comprising Zr:U mole ratios of 25:75, 40:60, and 50:50) and quantitated the H2O2 reaction via dissolved U and H2O2 concentrations. Although (U,Zr)O2 reacted more quickly than UO2, the dissolution yield relative to H2O2 consumption was far less for (U,Zr)O2 compared to that of UO2. The reaction kinetics indicates that most of the H2O2 catalytically decomposed to O2 at the surface of (U,Zr)O2. We confirmed the H2O2 catalytic decomposition via O2 production (quantitative stoichiometric agreement). In addition, post-reaction Raman scattering spectra of the undissolved (U,Zr)O2 showed no additional peaks (indicating a lack of secondary phase formation). The (U,Zr)O2 matrix is much more stable than UO2 against H2O2-induced oxidative dissolution. Our findings will improve understanding on the molten fuels and provide an insight into decommissioning activities after a severe accident.

  12. An instrument for rapid delineation of grade boundaries in selective mining of uranium ore

    International Nuclear Information System (INIS)

    Clark, G.J.; Dickson, B.L.; Meakins, R.L.; Kenny, D.; Talaska, A.

    1982-01-01

    A vehicle-mounted radiation detector interfaced to a microprocessor called PRAM (programmable radioactive analyser mobile) has been developed to provide grade control for selective mining of a soft rock uranium ore. The grade of ore over which the vehicle passes is determined and indicated by several coloured lights to a pegman who walks behind the vehicle. Coloured pegs are then laid out to mark the uranium grade ranges on the floor of the mine pit. Comparison between grade ranges determined by the PRAM and by prior drilling and downhole logging at the Yeelirrie deposit, Western Australia indicate good agreement. Use of the PRAM decreases the cost, manpower and time required to grade extensive areas of a mine pit floor

  13. Sensitivity analysis of uranium solubility under strongly oxidizing conditions

    International Nuclear Information System (INIS)

    Liu, L.; Neretnieks, I.

    1999-01-01

    To evaluate the effect of geochemical conditions in the repository on the solubility of uranium under strongly oxidizing conditions, a mathematical model has been developed to determine the solubility, by utilizing a set of nonlinear algebraic equations to describe the chemical equilibria in the groundwater environment. The model takes into account the predominant precipitation-dissolution reactions, hydrolysis reactions and complexation reactions that may occur under strongly oxidizing conditions. The model also includes the solubility-limiting solids induced by the presence of carbonate, phosphate, silicate, calcium, and sodium in the groundwater. The thermodynamic equilibrium constants used in the solubility calculations are essentially taken from the NEA Thermochemical Data Base of Uranium, with some modification and some uranium minerals added, such as soddyite, rutherfordite, uranophane, uranyl orthophosphate, and becquerelite. By applying this model, the sensitivities of uranium solubility to variations in the concentrations of various groundwater component species are systematically investigated. The results show that the total analytical concentrations of carbonate, phosphate, silicate, and calcium in deep groundwater play the most important role in determining the solubility of uranium under strongly oxidizing conditions

  14. Selective Oxidation of Soft Grade Carbon

    Directory of Open Access Journals (Sweden)

    Zecevic, N.

    2007-12-01

    Full Text Available Oil-furnace carbon black is produced by pyrolysis of gaseous or liquid hydrocarbons or their mixtures. The oil feedstock for the production of oil-furnace carbon black is mainly composed of high-boiling aromatic hydrocarbons, which are residues of petroleum cracking, while the gaseous raw material is commonly natural gas. Most of the oil-furnace carbon black production (> 99 % is used as a reinforcing agent in rubber compounds. Occasionally, oil-furnace carbon blacks are used in contact with other rubber compounds and fillers that have different pigments, particularly with the color white. It has been observed that frequently a migrating rubber soluble colorant would enter the white or light colored rubber composition from the adjacent carbon black filled rubber, resulting in a highly undesirable staining effect. Methods for determining non-oxidized residue on the surface of the oil-furnace carbon black include extraction of carbon black with the appropriate organic solvent, and measuring the color of the organic solvent by means of a colorimeter on 425 nm (ASTM D 1618-99. Transmittance values of 85 % or more are indicative of a practically non-staining carbon black, while transmittance values below 50 % generally lead to a carbon black with pronounced staining characteristics. Many oil-furnace carbon blacks, particularly those with a larger particle size (dp > 50 nm which are produced by pyrolysis, have strongly adsorbed non-reacted oil on their surfaces. Upon incorporation in a rubber compound, the colored materials are gradually dissolved by the rubber matrix and migrate freely into adjacent light colored rubber compounds, causing a highly objectionable staining effect. Adjusting furnace parameters in the industrial process of producing specific soft grades of carbon black cannot obtain minimal values of toluene discoloration. The minimal value of toluene discoloration is very important in special applications. Therefore, after-treatment of

  15. Recovery of uranium from low-grade sandstone ores and phosphate rock

    Energy Technology Data Exchange (ETDEWEB)

    Kennedy, R H [United States Atomic Energy Commission, Washington, D. C. (United States)

    1967-06-15

    This paper is concerned principally with commercial-scale experience in the United States in the recovery of uranium from low-grade sources. Most of these operations have been conducted by the operators of uranium mills as an alternative to processing normal-grade ores. The operations have been generally limited, therefore, to the treatment of low-grade materials generated in the course of mining normal-grade ores. In some circumstances such materials can be treated by simplified procedures as an attractive source of additional production. The experience gained in uranium recovery from phosphate rock will be treated in some detail. The land pebble phosphate rock of central Florida generally contains about 0.01 to 0.02% U{sub 3}O{sub 8}. While no uranium is being recovered from this source at the present time, it does represent a significant potential source of by-product uranium production because of the large tonnages being mined. (author)

  16. Bioleaching of low-grade uranium ore using Acidithiobacillus ferrooxidans

    International Nuclear Information System (INIS)

    Pal, S.; Das, T.; Sukla, L.B.; Roy Chaudhury, G.; Pradhan, D.

    2010-01-01

    Bioleaching of uranium was carried out with Turamdih ore sample procured from Uranium Corporation of India Limited, Jaduguda. The bacterial strain that was used in the leaching experiments was isolated from the Jaduguda mine water sample. Efficiency of bioleaching was studied by varying parameters like pulp density and initial ferrous concentration as source of energy. It is observed that the efficiency of bioleaching was 49% at 10% pulp density (w/v) and initial pH 2.0. Addition of external has no effect on efficiency of bioleaching showing domination of direct leaching mechanism over indirect. (author)

  17. Oxidation experiment of metal uranium waste for the treatment of depleted uranium waste

    International Nuclear Information System (INIS)

    Kang, K. H.; Kwac, K. I.; Kim, K. J.

    2001-01-01

    A study was conducted on the oxidation behavior of U-Ti chips(Depleted Uranium, DU chips) using an XRD and a thermogravimetric analyzer in the temperature range from 250 to 500 .deg. C in air. At the temperature lower than 400 .deg. C, DU chips were converted to UO 2 , U 3 O 7 and U 3 O 8 whereas at the temperature higher than 400 .deg. C, DU chips were completely converted to U 3 O 8 , the most stable form of uranium oxide. The activation energy for the oxidation of U-Ti chips is found, 44.9 kJ/mol and the oxidation rate in terms of weight gain (%) can be expressed as ; dW/dt=8.4 x 10 2 e(-44.9 kJ/mol /RT) wt %/min (250≤T(deg. C)≤500) where W=weight gain (%), t=time and T=temperature

  18. Voltametric determination of O:U relation in uranium oxide

    International Nuclear Information System (INIS)

    Carvalho, F.M.S. de; Abrao, A.

    1988-07-01

    Uranium oxide samples are dissolved in hot concentrated H 3 PO 4 - H 2 SO 4 mixture and the solution diluted with 1M H 2 SO 4 . One aliquot of such solution (A) is used to record the first voltamogram which gives the U(VI) content. To a second aliquot HNO 3 and H 2 O 2 is added to oxidise uranium to the hexavalent state (B) and the second voltamogram is recorded from 0.0 to 0.4 V X SCE. The O:U ratio in the original sample is calculated by the expression: O/U = 2.000 + [U (VI) soln.A/% U(VI) soln. B]. The method provides an accurate means for determining O to U ratios in high-purity uranium dioxide, fuel pellets and a variety of oxides prepared for developmental work on ceramic fuel materials. (author) [pt

  19. Practical considerations of pyrite oxidation control in uranium tailings

    International Nuclear Information System (INIS)

    1984-05-01

    The problems posed by the oxidation of pyrite in uranium tailings include the generation of sulfuric acid and acid sulfate metal salts. These have substantial negative impacts on watercourse biota by themselves, and the lowered pH levels tend to mobilize heavy metals present in the tailings the rate of oxidation of pyrite at lower pH levels is catalyzed by sulfur and iron oxidizing bacteria present in soils. No single clear solution to the problems came from this study. Exclusion of air is a most important preventative of bacterial catalysis of oxidation. Bactericides, chemically breaking the chain of integrated oxidation reactions, maintaining anaerobic conditions, or maintaining a neutral or alkaline pH all reduce the oxidation rate. Removal of pyrite by flotation will reduce but not eliminate the impact of pyrite oxidation. Controlled oxidation of the remaining sulfide in the flotation tails would provide an innocuous tailing so far as acidity generation is concerned

  20. Study on the solid state chemistry of ternary uranium oxides

    International Nuclear Information System (INIS)

    Yamashita, Toshiyuki

    1988-03-01

    With the increase of burnup of uranium oxide fuels, various kinds of fission products are formed, and the oxygen atoms combined with the consumed heavy atoms are freed. The solid state chemical and/or thermodynamic properties of these elements at high temperatures are complex, and have not been well clarified. In the present report, an approach was taken that the chemical interactions between UO 2 and these fission products can be regarded as causing overlapped effects of composing ternary uranium oxides, and formation reactions and phase behavior were studied for several ternary uranium oxides with typical fission product elements such as alkaline earth metals and rare earth elements. Precise determination methods for the composition of ternary uranium oxides were developed. The estimated accuracies for x and y values in M y U 1-y O 2+x were ± 0.006 and ± 0.004, respectively. The thermodynamic properties and the lattice parameters of the phases in the Ca-U-O and Pr-U-O systems were discussed in relation to the composition determined by the methods. Crystal structure analyses of cadmium monouranates were made with X-ray diffraction method. (author) 197 refs

  1. Resuspension of uranium-plutonium oxide particles from burning Plexiglas

    International Nuclear Information System (INIS)

    Pickering, S.

    1987-01-01

    Nuclear fuel materials such as Uranium-Plutonium oxide must be handled remotely in gloveboxes because of their radiotoxicity. These gloveboxes are frequently constructed largely of combustible Plexiglas sheet. To estimate the potential airborne spread of radioactive contamination in the event of a glovebox fire, the resuspension of particles from burning Plexiglas was investigated. (author)

  2. Controlled synthesis of thorium and uranium oxide nano-crystals

    International Nuclear Information System (INIS)

    Hudry, Damien; Apostolidis, Christos; Walter, Olaf; Gouder, Thomas; Courtois, Eglantine; Kubel, Christian; Meyer, Daniel

    2013-01-01

    Very little is known about the size and shape effects on the properties of actinide compounds. As a consequence, the controlled synthesis of well-defined actinide-based nano-crystals constitutes a fundamental step before studying their corresponding properties. In this paper, we report on the non-aqueous surfactant-assisted synthesis of thorium and uranium oxide nano-crystals. The final characteristics of thorium and uranium oxide nano-crystals can be easily tuned by controlling a few experimental parameters such as the nature of the actinide precursor and the composition of the organic system (e.g., the chemical nature of the surfactants and their relative concentrations). Additionally, the influence of these parameters on the outcome of the synthesis is highly dependent on the nature of the actinide element (thorium versus uranium). By using optimised experimental conditions, monodisperse isotropic uranium oxide nano-crystals with different sizes (4.5 and 10.7 nm) as well as branched nano-crystals (overall size ca. 5 nm), nano-dots (ca. 4 nm) and nano-rods (with ultra-small diameters of 1 nm) of thorium oxide were synthesised. (authors)

  3. Advantage of uranium contained in low grade dolomite ore

    International Nuclear Information System (INIS)

    Carneiro, A.L.M.

    1988-01-01

    The purpose of this work is to investigate a technological route to recover uranium from a lean mineral ore. The experimental work includes studies concerning calcination, carbonate leaching, settling, filtration and resin-ion-exchange. Experimental data confirm the technological feasibility of the proposed process and two different preliminary flowsheets of a pilot plant were suggested. (author) [pt

  4. Uranium R and D directed to low-grade ores

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The treatment of depleted uranium ores by in-situ leaching and by counterflow ion exchange in the USA is described. In-situ leaching is mainly suitable for sandstone deposits. The research was originally focused on leaching with an acid and with carbonates. Phosphoric acid appears to be a promising leaching agent. The equipment for continuous ion exchange may be used for sludge processing but the application depends on the type of equipment and mineralogy of processed ores. The method is advantageous for lower capital costs and for smooth operation. Ion exchange is also used for uranium extraction from mine waters in the USA as well as in Canada. For example, in Grants, New Mexico, a yield exceeding 90% was reached in mine waters only containing 5 to 7 ppm U 3 O 8 . In the future, the treatment of ores with a low uranium content will require more selective extraction methods in view of the more stringent technical conditions of uranium concentrate processing. (J.P.)

  5. PHASE ANALYSES OF URANIUM BEARING MINERALS FROM THE HIGH GRADE ORE, NOPAL I, PENA BLANCA, MEXICO

    International Nuclear Information System (INIS)

    Ren, M.; Goodell, P.; Kelts, A.; Anthony, E.Y.; Fayek, M.; Fan, C.; Beshears, C.

    2005-01-01

    The Nopal I uranium deposit is located in the Pena Blanca district, approximately 40 miles north of Chihuahua City, Mexico. The deposit was formed by hydrothermal processes within the fracture zone of welded silicic volcanic tuff. The ages of volcanic formations are between 35 to 44 m.y. and there was secondary silicification of most of the formations. After the formation of at least part of the uranium deposit, the ore body was uplifted above the water table and is presently exposed at the surface. Detailed petrographic characterization, electron microprobe backscatter electron (BSE) imagery, and selected x-ray maps for the samples from Nopal I high-grade ore document different uranium phases in the ore. There are at least two stages of uranium precipitation. A small amount of uraninite is encapsulated in silica. Hexavalent uranium may also have been a primary precipitant. The uranium phases were precipitated along cleavages of feldspars, and along fractures in the tuff. Energy dispersive spectrometer data and x-ray maps suggest that the major uranium phases are uranophane and weeksite. Substitutions of Ca and K occur in both phases, implying that conditions were variable during the mineralization/alteration process, and that compositions of the original minerals have a major influence on later stage alteration. Continued study is needed to fully characterize uranium behavior in these semi-arid to arid conditions

  6. On the stability of sub-stoichiometric uranium oxides

    Science.gov (United States)

    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.

  7. Characterization and uranium bioleaching performance of mixed iron- and sulfur-oxidizers versus iron-oxidizers

    International Nuclear Information System (INIS)

    Qian Li; Jing Sun; Dexin Ding; Qingliang Wang; Wenge Shi; Eming Hu; Xiaoyu Jiang; University of South China, Hengyang; Xingxing Wang

    2017-01-01

    In order to develop and apply mixed iron- and sulfur-oxidizers in uranium bioleaching, the characteristics of a mixed iron- and sulfur-oxidizing consortium (Consortium ISO) were comparatively investigated versus an iron-oxidizing consortium (Consortium IO). The results showed, the Consortium ISO exerted stronger oxidative ability and acid-producing ability than Consortium IO did. The synergy of sulfur-oxidizers and iron-oxidizers could change the structure and properties of the passivation substance, and work positively for eliminating the accumulation of passivation substance. In the bioleaching process, the uranium bioleaching experiments showed the recovery percentage of uranium reached 99.5% with Consortium ISO, 6.3% more than that of Consortium IO. (author)

  8. Reaction of uranium oxides with chlorine and carbon or carbon monoxide to prepare uranium chlorides

    Energy Technology Data Exchange (ETDEWEB)

    Haas, P.A.; Lee, D.D.; Mailen, J.C.

    1991-11-01

    The preferred preparation concept of uranium metal for feed to an AVLIS uranium enrichment process requires preparation of uranium tetrachloride (UCI{sub 4}) by reacting uranium oxides (UO{sub 2}/UO{sub 3}) and chlorine (Cl{sub 2}) in a molten chloride salt medium. UO{sub 2} is a very stable metal oxide; thus, the chemical conversion requires both a chlorinating agent and a reducing agent that gives an oxide product which is much more stable than the corresponding chloride. Experimental studies in a quartz reactor of 4-cm ID have demonstrated the practically of some chemical flow sheets. Experimentation has illustrated a sequence of results concerning the chemical flow sheets. Tests with a graphite block at 850{degrees}C demonstrated rapid reactions of Cl{sub 2} and evolution of carbon dioxide (CO{sub 2}) as a product. Use of carbon monoxide (CO) as the reducing agent also gave rapid reactions of Cl{sub 2} and formation of CO{sub 2} at lower temperatures, but the reduction reactions were slower than the chlorinations. Carbon powder in the molten salt melt gave higher rates of reduction and better steady state utilization of Cl{sub 2}. Addition of UO{sub 2} feed while chlorination was in progress greatly improved the operation by avoiding the plugging effects from high UO{sub 2} concentrations and the poor Cl{sub 2} utilizations from low UO{sub 2} concentrations. An UO{sub 3} feed gave undesirable effects while a feed of UO{sub 2}-C spheres was excellent. The UO{sub 2}-C spheres also gave good rates of reaction as a fixed bed without any molten chloride salt. Results with a larger reactor and a bottom condenser for volatilized uranium show collection of condensed uranium chlorides as a loose powder and chlorine utilizations of 95--98% at high feed rates. 14 refs., 7 figs., 14 tabs.

  9. Preparation and characterization of uranium alkoxides through oxidation of uranium metal

    International Nuclear Information System (INIS)

    Gordon, P.L.; Sauer, N.N.; Burns, C.J.; Watkin, J.G.; Van Der Sluys, W.G.

    1993-01-01

    Currently the authors are investigating the preparation of halide-containing uranium alkoxides by simultaneous halogen and alcohol oxidation of uranium metal. They recently reported the formation of U 2 I 4 (O-i-Pr) 4 (HO-i-Pr) 2 which upon addition of excess isopropanol forms UI 2 (O-i-Pr) 2 (HO-i-Pr) 2 . They report further characterization and reactivity for this monomeric species. Attempts to prepare similar complexes are being made using chlorine gas in the presence of other alcohols. They describe this ongoing research

  10. Sintering of uranium oxide of high specific surface area

    International Nuclear Information System (INIS)

    Bel, Alain; Francois, Bernard; Delmas, Roger; Caillat, Roger

    1959-01-01

    The extent to which a uranium oxide powder deriving from ammonium uranate or uranium peroxide lends itself to the sintering process depends largely on its specific surface area. When this is greater than 5 m 2 / g there is an optimum temperature for sintering in hydrogen. This temperature becomes less as the specific area of the powder is greater. Reprint of a paper published in Comptes rendus des seances de l'Academie des Sciences, t. 249, p. 1045-1047, sitting of 21 September 1959 [fr

  11. Technology for down-blending weapons grade uranium into commercial reactor-usable uranium

    International Nuclear Information System (INIS)

    Arbital, J.G.; Snider, J.D.

    1996-01-01

    The US Department of Energy (DOE) is evaluating options for rendering surplus inventories of highly enriched uranium (HEU) incapable of being used in nuclear weapons. Weapons-capable HEU was earlier produced by enriching the uranium isotope 235 U from its natural occurring 0.71 percent isotopic concentration to at least 20 percent isotopic concentration. Now, by permanently diluting the concentration of the 235 U isotope, the weapons capability of HEU can be eliminated in a manner that is reversible only through isotope re-enrichment, and therefore, highly resistant to proliferation. To the extent that can be economically and technically justified, the down-blended, low-enriched uranium product will be made suitable for use as commercial reactor fuel. Such down-blended uranium product can also be disposed of as waste if chemical or isotopic impurities preclude its use as reactor fuel. The DOE has evaluated three candidate processes for down blending surplus HEU. These candidate processes are: (1) uranium hexafluoride blending; (2) molten uranium metal blending; and (3) uranyl nitrate solution blending. This paper describes each of these candidate processes. It also compares the relative advantages and disadvantages of each process with respect to: (1) the various forms and compounds of HEU comprising the surplus inventory, (2) the use of down-blended product as commercial reactor fuel, or (3) its disposal as waste

  12. Status Report from Czechoslovakia [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    Civin, V; Belsky, M [Research and Development Laboratory No.3 of the Uranium Industry, Prague, Czechoslovakia (Czech Republic)

    1967-06-15

    The present paper deals with the fundamental problems and the main routes followed in processing low-grade uranium ores in CSSR. In this connection it may be useful to discuss the definition of low-grade ore. In our country this term is applied to uraniferous material with a very low content of uranium (of the order of 0.01%) whose treatment causes no particular difficulty. However, the same term is also used to designate those materials whose processibility lies on the verge of economic profitability. In our view, this classification, of an ore using two independent criteria (i.e. uranium content and processing economy) is useful from the standpoint of technology. The treatment of both such ore types is as a rule carried out by specific technological processes. Consequently, low-grade uranium ores can be divided into two groups: (1) Ores with a low uranium content. To this category belong in our country uraniferous materials which originate as a by-product of technological processes used in processing other materials. This is primarily gangue and tailings of various physical or physico-chemical pretreatment operations to which the ore is subjected at the mining site. Mention should be made in this connection of mine waters, which represent a useful complementary source of uranium despite their low uranium content (of the order of milligrams per litre). (2) Ores whose economical treatment is problematic. To this category belong deposits of conventional ore types with a uranium content on the limit of profitable treatment. Also, those deposits containing atypical materials possessing such properties which impair the economy of their treatment. This includes ores with a considerable amount of components which are difficult to separate and which at the same time consume the leaching agents. Finally, it covers uranium-bearing materials in refractory forms which are difficult to dissolve and also some special materials, such as lignites, uranium-bearing shales, loams

  13. Review of experience gained in fabricating nuclear grade uranium and thorium compounds and their analytical quality control at the Instituto de Energia Atomica, Sao Paulo, Brazil

    International Nuclear Information System (INIS)

    Abrao, A.; Franca, J.M. Jr.; Ikuta, A.; Pueschel, C.R.; Federgruen, L.; Lordello, A.R.; Tomida, E.K.; Moraes, S.; Brito, J. de; Gomes, R.P.; Araujo, J.A.; Floh, B.; Matsuda, H.T.

    1977-01-01

    This paper summarizes the main activities dealing with the fabrication of nuclear grade uranium and thorium compounds at the Instituto de Energia Atomica, Sao Paulo. Identification of problems and their resolutions, the experience gained in plant operation, the performance characteristics of an ion-exchange facility and a solvent extraction unit (a demonstration plant based on pulsed columns for purification of uranium and production of ammonium diuranate) are described. A moving-bed facility for UF 4 preparation and its operation is discussed. A pilot plant for uranium and thorium oxide microsphere preparation based on internal gelation for HTGR fuel type is also described. A solvent extraction pilot plant for thorium purification based on a compound extraction-scrubbing column and a mixer-settler battery and the involved technology for thorium purification are commented. The main products, namely ammonium diuranate, uranyl amonium tricarbonate, uranium trioxide, uranium tetrafluoride, thorium nitrate and thorium oxalate and their quality are commented. The development of necessary analytical procedures for the quality control of the mentioned nuclear grade products is summarized. A great majority of such procedures was particularly suitable for analyzing traces impurities. Designed for installation are the units for denitration of uranyl nitrate solutions and pilot plants for elemental fluorine and UF 6 . The installation of a laboratory-scale plant designed for reprocessing irradiated uranium and an experimental unit for the recovery of protactinium from irradiated thorium is in progress

  14. Preliminary study on weapon grade uranium utilization in molten salt reactor miniFUJI

    International Nuclear Information System (INIS)

    Aji, Indarta Kuncoro; Waris, A.

    2014-01-01

    Preliminary study on weapon grade uranium utilization in 25MWth and 50MWth of miniFUJI MSR (molten salt reactor) has been carried out. In this study, a very high enriched uranium that we called weapon grade uranium has been employed in UF 4 composition. The 235 U enrichment is 90 - 95 %. The results show that the 25MWth miniFUJI MSR can get its criticality condition for 1.56 %, 1.76%, and 1.96% of UF 4 with 235 U enrichment of at least 93%, 90%, and 90%, respectively. In contrast, the 50 MWth miniFUJI reactor can be critical for 1.96% of UF 4 with 235 U enrichment of at smallest amount 95%. The neutron spectra are almost similar for each power output

  15. Preliminary study on weapon grade uranium utilization in molten salt reactor miniFUJI

    Energy Technology Data Exchange (ETDEWEB)

    Aji, Indarta Kuncoro [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Waris, A., E-mail: awaris@fi.itb.ac.id [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa No. 10 Bandung 40132 (Indonesia)

    2014-09-30

    Preliminary study on weapon grade uranium utilization in 25MWth and 50MWth of miniFUJI MSR (molten salt reactor) has been carried out. In this study, a very high enriched uranium that we called weapon grade uranium has been employed in UF{sub 4} composition. The {sup 235}U enrichment is 90 - 95 %. The results show that the 25MWth miniFUJI MSR can get its criticality condition for 1.56 %, 1.76%, and 1.96% of UF{sub 4} with {sup 235}U enrichment of at least 93%, 90%, and 90%, respectively. In contrast, the 50 MWth miniFUJI reactor can be critical for 1.96% of UF{sub 4} with {sup 235}U enrichment of at smallest amount 95%. The neutron spectra are almost similar for each power output.

  16. Technologies for processing low-grade uranium ores and their relevance to the Indian situation

    International Nuclear Information System (INIS)

    Murthy, T.K.S.

    1991-01-01

    The technology for uranium ore processing is well established. Various estimates have shown that on a global basis uranium resources are adequate to meet the forseeable demand. The Indian resources are estimated to be about 60,000 t U. The grade of the ores is low and the individual deposits are small. The nature of the deposits, precarious resources position and relatively small capacity of the mines do not permit the country to take advantage of large throughputs in the mill to achieve substantial cost reduction. However by resorting to as high a scale of milling as the mines would permit, by reducing the loss of solubilised uranium after leaching and by undertaking production of nuclear grade final product at the mill site, significant though not a major, economic benefit can be derived. (author). 2 figs., 3 tabs

  17. Fabrication of Cerium Oxide and Uranium Oxide Microspheres for Space Nuclear Power Applications

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey A. Katalenich; Michael R. Hartman; Robert C. O' Brien

    2013-02-01

    Cerium oxide and uranium oxide microspheres are being produced via an internal gelation sol-gel method to investigate alternative fabrication routes for space nuclear fuels. Depleted uranium and non-radioactive cerium are being utilized as surrogates for plutonium-238 (Pu-238) used in radioisotope thermoelectric generators and for enriched uranium required by nuclear thermal rockets. While current methods used to produce Pu-238 fuels at Los Alamos National Laboratory (LANL) involve the generation of fine powders that pose a respiratory hazard and have a propensity to contaminate glove boxes, the sol-gel route allows for the generation of oxide microsphere fuels through an aqueous route. The sol-gel method does not generate fine powders and may require fewer processing steps than the LANL method with less operator handling. High-quality cerium dioxide microspheres have been fabricated in the desired size range and equipment is being prepared to establish a uranium dioxide microsphere production capability.

  18. Uranium in 50 years? - Deeper, lower grade and more metallurgically complex? - 5312

    International Nuclear Information System (INIS)

    Polak, C.

    2015-01-01

    The economic exploitation of uranium deposits in the next 50 years will benefit from advances in mining and processing technology. The 'easiness' to find uranium deposits are a relic of the past. Exploration will need to make use of new technologies to help find blind or deep deposits. These issues are already being addressed by the uranium industry. Another issue will be to obtain social and environmental acceptation of the industry. To summarize the uranium mining industry is faced with 3 main challenges that are not necessarily mutually exclusive: deeper deposits, lower grades and chemically complex ores. The deposits of the next half of 21. century are likely to face at least one but potentially a combination of two or three of these challenges

  19. Standard test method for the determination of uranium by ignition and the oxygen to uranium (O/U) atomic ratio of nuclear grade uranium dioxide powders and pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2000-01-01

    1.1 This test method covers the determination of uranium and the oxygen to uranium atomic ratio in nuclear grade uranium dioxide powder and pellets. 1.2 This test method does not include provisions for preventing criticality accidents or requirements for health and safety. Observance of this test method does not relieve the user of the obligation to be aware of and conform to all international, national, or federal, state and local regulations pertaining to possessing, shipping, processing, or using source or special nuclear material. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. 1.4 This test method also is applicable to UO3 and U3O8 powder.

  20. A new type on line fast analysis instrument of uranium ore grade

    International Nuclear Information System (INIS)

    Guo Maojin.

    1992-01-01

    The instrument is used to analyse the average grade of uranium ore on the belt. Its average analysis speed is about 300 t/h. The physical mechanism of measurement, characteristics, principle of operation and applications in several years are described. The CMOS LSI IC are adopted. The stability, reliability and anti-interference ability are very good

  1. Preparation of uranium-based oxide catalysts; Preparation de catalyseurs oxydes a base d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Bressat, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1965-07-01

    We have studied the thermal decomposition of uranyl and uranium IV oxalates as a mean of producing uranium dioxide. We have isolated the main intermediate phases of the decompositions and have indexed the lines of their X-ray diffraction patterns. The oxides produced by the decomposition are ill-defined and unstable: they strongly absorb atmospheric oxygen with modification of the composition and, in certain cases, of the structure (pyrophoric oxide). With a view to obtaining stable oxides, we have prepared mixed uranium-thorium oxalates. In order to prepare an oxalate having a homogeneous composition, it is necessary to adopt a well-defined preparation method: the addition of solutions of thorium and uranium IV nitrates to a continually saturated oxalic acid solution. The mixed oxide obtained from the thermal decomposition of an oxalate U{sub x}Th{sub 1-x}(C{sub 2}O{sub 4}){sub 2}, 2 H{sub 2}O at 500 C for 24 hours in a current of oxygen leads to a cubic structure which is well-defined both in the bulk and superficially when x is less than 0.35. Above this atomic concentration of uranium, some uranium moves out of the lattice in the form of UO{sub 3} or U{sub 3}O{sub 8} according to the temperature. The mixed oxide is not stoichiometric,(U{sub x}Th{sub 1-x}O{sub 2+y}) and the average degree of oxidation of the uranium varies with the temperature and partial oxygen pressure. The oxides thus formed have a high surface area. By dissolving the mixed oxalates in a concentrated solution of ammonium oxalate, it is possible to deposit the catalyst on a support, but the differences in the solubilities of the thorium and uranium IV oxalates in the ammonium oxalate make it impossible to prepare double salts formed either of thorium and uranium and of ammonium. (author) [French] Nous avons etudie la decomposition thermique des oxalates d'uranyle et d'uranium IV en vue d'aboutir au dioxide d'uranium. Nous avons pu isoler les principales phases intermediaires des decompositions

  2. Optimization of operating parameters and rate of uranium bioleaching from a low-grade ore

    International Nuclear Information System (INIS)

    Rashidi, A.; Roosta-Azad, R.; Safdari, S.J.

    2014-01-01

    In this study the bioleaching of a low-grade uranium ore containing 480 ppm uranium has been reported. The studies involved extraction of uranium using Acidithiobacillus ferrooxidans derived from the uranium mine samples. The maximum specific growth rate (μ max ) and doubling time (t d ) were obtained 0.08 h -1 and 8.66 h, respectively. Parameters such as Fe 2+ concentration, particle size, temperature and pH were optimized. The effect of pulp density (PD) was also studied. Maximum uranium bio-dissolution of 100 ± 5 % was achieved under the conditions of pH 2.0, 5 % PD and 35 deg C in 48 h with the particles of d 80 = 100 μm. The optimum concentration of supplementary Fe 2+ was dependent to the PD. This value was 0 and 10 g of FeSO 4 ·7H 2 O/l at the PD of 5 and 15 %, respectively. The effects of time, pH and PD on the bioleaching process were studied using central composite design. New rate equation was improved for the uranium leaching rate. The rate of leaching is controlled with the concentrations of ferric and ferrous ions in solution. This study shows that uranium bioleaching may be an important process for the Saghand U mine at Yazd (Iran). (author)

  3. Oxidation kinetics of reaction products formed in uranium metal corrosion

    International Nuclear Information System (INIS)

    Totemeier, T. C.

    1998-01-01

    The oxidation behavior of uranium metal ZPPR fuel corrosion products in environments of Ar-4%O 2 and Ar-20%O 2 were studied using thermo-gravimetric analysis (TGA). These tests were performed to extend earlier work in this area specifically, to assess plate-to-plate variations in corrosion product properties and the effect of oxygen concentration on oxidation behavior. The corrosion products from two relatively severely corroded plates were similar, while the products from a relatively intact plate were not reactive. Oxygen concentration strongly affected the burning rate of reactive products, but had little effect on low-temperature oxidation rates

  4. Oxidation kinetics of reaction products formed in uranium metal corrosion.

    Energy Technology Data Exchange (ETDEWEB)

    Totemeier, T. C.

    1998-04-22

    The oxidation behavior of uranium metal ZPPR fuel corrosion products in environments of Ar-4%O{sub 2} and Ar-20%O{sub 2} were studied using thermo-gravimetric analysis (TGA). These tests were performed to extend earlier work in this area specifically, to assess plate-to-plate variations in corrosion product properties and the effect of oxygen concentration on oxidation behavior. The corrosion products from two relatively severely corroded plates were similar, while the products from a relatively intact plate were not reactive. Oxygen concentration strongly affected the burning rate of reactive products, but had little effect on low-temperature oxidation rates.

  5. Oxidation of uranium monocarbide in dry or moist oxygen

    International Nuclear Information System (INIS)

    Herrmann, B.; Herrmann, F.J.

    1968-01-01

    The kinetics of oxidation of uranium monocarbide either in dry or moist air or in oxygen-argon mixtures, has been studied thermogravimetrically, between 500 and 800 C in a circulating atmosphere. In all cases the oxidation leads to the formation of U 3 O 8 . Between 500 and 700 C, the activation energy is about 21 +3 kcal/mole. It seems to decrease between 700 and 800 C, but the reaction follows always a linear rate law. In moist air, the oxidation proceeds more swiftly, due to an increase in the reactional interface. An evaluation of the over-temperature has been made at 800 C. (author) [fr

  6. Comparison of open cycles of uranium and mixed oxides of thorium-uranium using advanced reactors

    International Nuclear Information System (INIS)

    Gonçalves, Letícia C.; Maiorino, José R.

    2017-01-01

    A comparative study of the mass balance and production costs of uranium oxide fuels was carried out for an AP1000 reactor and thorium-uranium mixed oxide in a reactor proposal using thorium called AP-Th1000. Assuming the input mass values for a fuel load the average enrichment for both reactors as well as their feed mass was determined. With these parameters, the costs were calculated in each fuel preparation process, assuming the prices provided by the World Nuclear Association. The total fuel costs for the two reactors were quantitatively compared with 18-month open cycle. Considering enrichment of 20% for the open cycle of mixed U-Th oxide fuel, the total uranium consumption of this option was 50% higher and the cost due to the enrichment was 70% higher. The results show that the use of U-Th mixed oxide fuels can be advantageous considering sustainability issues. In this case other parameters and conditions should be investigated, especially those related to fuel recycling, spent fuel storage and reduction of the amount of transuranic radioactive waste

  7. Water Sorption and Gamma Radiolysis Studies for Uranium Oxides

    Energy Technology Data Exchange (ETDEWEB)

    Icenhour, A.S.

    2002-02-27

    During the development of a standard for the safe, long-term storage of {sup 233}U-containing materials, several areas were identified that needed additional experimental studies. These studies were related to the perceived potential for the radiolytic generation of large pressures or explosive concentrations of gases in storage containers. This report documents the results of studies on the sorption of water by various uranium oxides and on the gamma radiolysis of uranium oxides containing various amounts of sorbed moisture. In all of the experiments, {sup 238}U was used as a surrogate for the {sup 233}U. For the water sorption experiments, uranium oxide samples were prepared and exposed to known levels of humidity to establish the water uptake rate. Subsequently, the amount of water removed was studied by heating samples in a oven at fixed temperatures and by thermogravimetric analysis (TGA)/differential thermal analysis (DTA). It was demonstrated that heating at 650 C adequately removes all moisture from the samples. Uranium-238 oxides were irradiated in a {sup 60}Co source and in the high-gamma-radiation fields provided by spent nuclear fuel elements of the High Flux Isotope Reactor. For hydrated samples of UO{sub 3}, H{sub 2} was the primary gas produced; but the total gas pressure increase reached steady value of about 10 psi. This production appears to be a function of the dose and the amount of water present. Oxygen in the hydrated UO{sub 3} sample atmosphere was typically depleted, and no significant pressure rise was observed. Heat treatment of the UO{sub 3} {center_dot} xH{sub 2}O at 650 C would result in conversion to U{sub 3}O{sub 8} and eliminate the H{sub 2} production. For all of the U{sub 3}O{sub 8} samples loaded in air and irradiated with gamma radiation, a pressure decrease was seen and little, if any, H{sub 2} was produced--even for samples with up to 9 wt % moisture content. Hence, these results demonstrated that the efforts to remove trace

  8. Standard test methods for analysis of sintered gadolinium oxide-uranium dioxide pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2006-01-01

    1.1 These test methods cover procedures for the analysis of sintered gadolinium oxide-uranium dioxide pellets to determine compliance with specifications. 1.2 The analytical procedures appear in the following order: Section Carbon (Total) by Direct CombustionThermal Conductivity Method C1408 Test Method for Carbon (Total) in Uranium Oxide Powders and Pellets By Direct Combustion-Infrared Detection Method Chlorine and Fluorine by Pyrohydrolysis Ion-Selective Electrode Method C1502 Test Method for Determination of Total Chlorine and Fluorine in Uranium Dioxide and Gadolinium Oxide Gadolinia Content by Energy-Dispersive X-Ray Spectrometry C1456 Test Method for Determination of Uranium or Gadolinium, or Both, in Gadolinium Oxide-Uranium Oxide Pellets or by X-Ray Fluorescence (XRF) Hydrogen by Inert Gas Fusion C1457 Test Method for Determination of Total Hydrogen Content of Uranium Oxide Powders and Pellets by Carrier Gas Extraction Isotopic Uranium Composition by Multiple-Filament Surface-Ioni...

  9. Literature information applicable to the reaction of uranium oxides with chlorine to prepare uranium tetrachloride

    Energy Technology Data Exchange (ETDEWEB)

    Haas, P.A.

    1992-02-01

    The reaction of uranium oxides and chlorine to prepare anhydrous uranium tetrachloride (UCl{sub 4}) are important to more economical preparation of uranium metal. The most practical reactions require carbon or carbon monoxide (CO) to give CO or carbon dioxide (CO{sub 2}) as waste gases. The chemistry of U-O-Cl compounds is very complex with valances of 3, 4, 5, and 6 and with stable oxychlorides. Literature was reviewed to collect thermochemical data, phase equilibrium information, and results of experimental studies. Calculations using thermodynamic data can identify the probable reactions, but the results are uncertain. All the U-O-Cl compounds have large free energies of formation and the calculations give uncertain small differences of large numbers. The phase diagram for UCl{sub 4}-UO{sub 2} shows a reaction to form uranium oxychloride (UOCl{sub 2}) that has a good solubility in molten UCl{sub 4}. This appears more favorable to good rates of reaction than reaction of solids and gases. There is limited information on U-O-Cl salt properties. Information on the preparation of titanium, zirconium, silicon, and thorium tetrachlorides (TiCl{sub 4}, ZrCl{sub 4}, SiCl{sub 4}, ThCl{sub 4}) by reaction of oxides with chlorine (Cl{sub 2}) and carbon has application to the preparation of UCl{sub 4}.

  10. Literature information applicable to the reaction of uranium oxides with chlorine to prepare uranium tetrachloride

    International Nuclear Information System (INIS)

    Haas, P.A.

    1992-02-01

    The reaction of uranium oxides and chlorine to prepare anhydrous uranium tetrachloride (UCl 4 ) are important to more economical preparation of uranium metal. The most practical reactions require carbon or carbon monoxide (CO) to give CO or carbon dioxide (CO 2 ) as waste gases. The chemistry of U-O-Cl compounds is very complex with valances of 3, 4, 5, and 6 and with stable oxychlorides. Literature was reviewed to collect thermochemical data, phase equilibrium information, and results of experimental studies. Calculations using thermodynamic data can identify the probable reactions, but the results are uncertain. All the U-O-Cl compounds have large free energies of formation and the calculations give uncertain small differences of large numbers. The phase diagram for UCl 4 -UO 2 shows a reaction to form uranium oxychloride (UOCl 2 ) that has a good solubility in molten UCl 4 . This appears more favorable to good rates of reaction than reaction of solids and gases. There is limited information on U-O-Cl salt properties. Information on the preparation of titanium, zirconium, silicon, and thorium tetrachlorides (TiCl 4 , ZrCl 4 , SiCl 4 , ThCl 4 ) by reaction of oxides with chlorine (Cl 2 ) and carbon has application to the preparation of UCl 4

  11. Dissolution of uranium oxide materials in simulated lung fluid

    International Nuclear Information System (INIS)

    Scripsick, R.C.; Soderholm, S.C.

    1985-01-01

    Depleted uranium (DU) oxide aerosols prepared in the laboratory and collected in the field were tested to characterize their dissolution in simulated lung fluid and to determine how dissolution is affected by aerosol preparation. DU, a by-product of the uranium fuel cycle, has been selected by the US military for use in several types of munitions. During development, manufacture, testing, and use of these munitions, opportunities exist for inhalation exposure to various (usually oxide) aerosol forms of DU. The hazard potential associated with such exposures is closely related to the chemical form, the size of the DU aerosol material, and its dissolution properties. Five DU sample materials produced by exposing uranium alloy penetrators to certain controlled oxidation atmospheres were studied (oxidation temperatures ranged from 500 to 900 0 C). In addition, two DU sample materials collected in the field were provided by the US Air Force. All sample materials were generated as aerosols and the respirable fraction was separated and collected. Data suggest that under some conditions a rapidly dissolving U 3 O 8 fraction may be formed concurrent with the production of UO 2

  12. Preparation of Ceramic-Grade Thorium-Uranium Oxide; Preparation d'un melange d'oxydes de thorium et d'uranium propre a la fabrication de combustible ceramique; Izgotovlenie keramicheskogo torievo-uranovogo okisla; Preparacion de mezclas de oxidos de uranio y torio, de tipo ceramico

    Energy Technology Data Exchange (ETDEWEB)

    Cogliati, G.; De Leone, R.; Ferrari, S.; Gabaglio, M.; Liscia, A. [Centro Studi Nucleari della Casaccia, Rome (Italy)

    1963-11-15

    A method for the preparation of sintered bodies of thorium-uranium mixed oxide starting from a solution of thorium nitrate and uranyl nitrate was investigated. This method can be useful both in the fabrication of fuel elements and in the reprocessing of such type of materials. In the first step of the method, uranyl nitrate is reduced to uranium (IV) nitrate. As reducing agent, both gaseous hydrogen and formic acid are employed; urea is added to prevent the formation of nitrous acid, which catalyses the reoxidation of uranium (IV). As catalyst, both platinum and palladium can be employed. Data are given for a continuous process, in which formic acid and urea are added to the solution, which is then pre-heated and passed in a column packed with 1/8 in x 1/8 in alumina pellets, carrying 0.5 wt.% of platinum. The influence of flow rate, temperature, formic acid and urea concentration, as well as catalyst life and poisoning are studied. The second step in the method is the precipitation of an oxalate of thorium and uranium (IV). The influence of oxalic acid to thorium-uranium ratio, temperature, aging time on settling and filtering characteristics of the precipitate and on the ceramic properties of the obtained powders is reported. Firing was carried out both in reducing and oxidizing atmosphere. After preliminary tests, two standard procedures were set up for the fabrication of ceramic bodies, namely by cold pressing and sintering and by extrusion and sintering. The ability of the different powders to sinter was tested by both of the two standard methods. With some of the powders, densities higher than 95% of theoretical density were obtained; reproducibility tests were successfully carried out. (author) [French] Les auteurs etudient une methode de preparation de melanges frittes d'oxydes de thorium et d'uranium a partir d'une solution de nitrate de thorium et de nitrate d'uranyle. Cette methode peut etre utile auessi bien pour la fabrication d'elements combustibles

  13. Experimental investigation and thermodynamic simulation of the uranium oxide-zirconium oxide-iron oxide system in air

    Czech Academy of Sciences Publication Activity Database

    Petrov, Y. B.; Udalov, Y. P.; Šubrt, Jan; Bakardjieva, Snejana; Sázavský, P.; Kiselová, M.; Selucký, P.; Bezdička, Petr; Joumeau, C.; Piluso, P.

    2011-01-01

    Roč. 37, č. 2 (2011), s. 212-229 ISSN 1087-6596 Institutional research plan: CEZ:AV0Z40320502 Keywords : uranium oxide * zirconium oxide * iron oxide * fusibility curve * oxygen partial pressure * crystallization * phase composition Subject RIV: CA - Inorganic Chemistry Impact factor: 0.492, year: 2011

  14. Failure mechanisms for compacted uranium oxide fuel cores

    International Nuclear Information System (INIS)

    Berghaus, D.G.; Peacock, H.B.

    1980-01-01

    Tension, compression, and shear tests were performed on test specimens of aluminum-clad, compacted powder fuel cores to determine failure mechanisms of the core material. The core, which consists of 70% uranium oxide in an aluminum matrix, frequently fails during post-extrusion drawing. Tests were conducted to various strain levels up to failure of the core. Sections were made of tested specimens to microscopically study initiation of failure. Two failure modes wee observed. Tensile failure mode is initiated by prior tensile failure of uranium oxide particles with the separation path strongly influenced by the arrangement of particles. Delamination mode consists of the separation of laminae formed during extrusion of tubes. Separation proceeds from fine cracks formed parallel to the laminae. Tensile failure mode was experienced in tension and shear tests. Delamination mode was produced in compression tests

  15. Ionization and Coulomb explosion of small uranium oxide clusters

    International Nuclear Information System (INIS)

    Ross, Matt W; Castleman, A W Jr

    2012-01-01

    Femtosecond pulses are used to study the strong-field ionization and subsequent Coulomb explosion of small uranium oxide clusters. The resulting high atomic charge states are explored as a function of laser intensity and compared to ionization rates calculated using semi-classical tunneling theory with sequential ionization potential values. The gap in laser intensity between saturation intensity values for the 7s, 6d, and 5f orbitals are identified and quantified. Extreme charge states of oxygen up to O 4+ are observed indicating multiple ionization enhancement processes occurring within the clusters. The peak splittings of the atomic charge states are explored and compared to previous results on transition metal oxide species. Participation of the 5f orbitals in bonding is clearly identified based on the saturation intensity dependence of oxygen to uranium metal.

  16. Characteristics of uranium oxide cathode for neutron streak camera

    International Nuclear Information System (INIS)

    Niki, H.; Itoga, K.; Yamanaka, M.; Yamanaka, T.; Yamanaka, C.

    1986-01-01

    In laser fusion research, time-resolved neutron measurements require 20ps resolution in order to obtain the time history of the D-T burn. Uranium oxide was expected to be a sensitive material as a cathode of a neutron streak camera because of its large fission cross section. The authors report their measurements of some characteristics of the uranium oxide cathode connected to a conventional streak tube. 14 MeV neutron signal were observed as the bright spots on a TV monitor using a focus mode opration. Detection efficiency was ∼ 1 x 10 -6 for 1 μm thick cathode. Each signal consisted of more than several tens of components, which were corresponding to the secondary electrons dragged out from the cathode by a fission fragment. Time resolution is thought to be limited mainly by the transit time spread of the secondary electrons. 14ps resolution was obtained by a streak mode operation for a single fission event

  17. Uranium tetracyclopentadienyl interaction with carbon oxide and dioxide

    International Nuclear Information System (INIS)

    Leonov, M.R.; Solov'eva, G.V.; Kozina, I.Z.; Bolotova, G.T.

    1983-01-01

    Using the methods of gas-liquid chromatography, IR and UV spectroscopy and element analysis, the reactions of tetracyclogentadienyluranium with carbon oxide and dioxide have been studied. It is shown that complete uranium cyclopentadienyl π-complex-tetracyclopentadienyluranium - in pentane under normal conditions for 100 hr reacts with carbon oxide and dioxide with the formation of polymeric complex ([(etasup(5)-Csub(5)Hsub(5))x(-CO-)U(etasup(5)-Csub(5)Hsub(4))(-CO-)]sub(2)]sub(n), in which two uranium atoms are bonded with two bridge fragments (eta 5 -C 5 H 4 -CO-), and dimeric complex [(eta 5 -C 5 H 5 ) 2 UH 2 xCO 2 ] 2 respectively

  18. Thermal diffusivity and conductivity of thorium- uranium mixed oxides

    Science.gov (United States)

    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.

  19. Role of Some Isolated Fungi in The Biological Leaching of Uranium From Low Grade Cretaceous Sandstone

    International Nuclear Information System (INIS)

    Ibrahim, H.A.; Morsy, A.; El-Sheikh, E.M.

    2012-01-01

    Microbiological leaching has been used as an alternative approach to conventional hydrometallurgical methods of uranium extraction. In this investigation, the biological leaching of uranium by isolated fungi from low grade sandstone was studied. Five isolates of fungi were obtained from sandstone sample. Cladosporium oxysporum and Penicilluim stoloniferum exhibited high potential in generating a variety of organic acids effective for uranium extraction. The percentages of organic acid produced by fungi were determined. By-product such as molasses was tested. The maximum dissolution of uranium was achieved at the following conditions; incubation period 6 days, pulp density 1:3 g/L, ph 3.5 and at 30 degree C. Maximum solubilization of uranium with values of 54% and 67% were achieved by Cladosporium oxysporum and Penicilluim stoloniferum, respectively. From properly prepared pregnant bio-leach liquor, the leached uranium was recovered in the form of marketable products (3UO 3 NH 3 .5H 2 O) using classical chemical technique and the product was confirmed using XRD techniques

  20. Microbial leaching of low grade sandstone uranium ores: column leaching studies

    International Nuclear Information System (INIS)

    Bhatti, T.M.; Malik, K.A.; Khalid, A.M.

    1991-01-01

    Microbial leaching studies on a low-grade sandstone uranium ore from Baghalchur Ore Deposits, D. G. Khan, Pakistan, containing 0.027 % U/sub 3/O/sub 8/ for extraction of uranium, were conducted in columns. Baghalchur sandstone uranium ore which is alkaline in nature, contained 5.0% calcite [CaCo/sub 3/], 2-3 % Fe/sub 2/O/sub 3/ and pyrite [FeS/sub 2/] less than 0.1 %. The ore amended with sulfur and/or sulfur slag as external energy source was found to leach with indigenous microflora mostly belonging to the genus Thiobacillus which are present in the uranium mine water. Column leaching studies revealed that when the ore was amended with elemental sulfur and irrigated with mine water (pH 3.5) 53 % U/sub 3/O/sub 8/ could be solubilized from it. However, when the natural mine water was used as such (pH 7.4) the solubilization of uranium was decreased to 41 % U/sub 3/O/sub 8/ in 90 days under similar conditions of percolation rate and temperature. The addition of (NH/sub 4/)/sub 2/SO/sub 4/ (3.0 g/L) in mine water was found to enhance the uranium leaching to 70 % U/sub 3/O/sub 8/ from the columns containing ore amended with sulfur slag. (author)

  1. Preconcentration of low-grade uranium ores with environmentally acceptable tailings, part I

    International Nuclear Information System (INIS)

    Raicevic, D.; Raicevic, M.; McCarthy, D.R.

    1979-08-01

    The low-grade ore sample used for this investigation originated from Agnew Lake Mines Limited, Espanola, Ontario. It contained about 1% pyrite and 0.057% uranium, mainly as uranothorite with a small amount of brannerite. Both of these minerals occur in the quartz-sericite matrix of a conglomerate. A preconcentration process has been developed to give a high uranium recovery, reject pyrite, radium and thorium from the ore and produce environmentally acceptable tailings. This process applies flotation in combination with high intensity magnetic separation and gravity concentration

  2. Chlorination separation of uranium, thorium, and radium from low-grade ores

    International Nuclear Information System (INIS)

    Sastri, V.S.; Perumareddi, J.R.

    1995-01-01

    Low-temperature chlorination of low-grade uranium ores containing uranium in the 0.02 to 0.06% range, thorium in the 0.036 to 0.12% range, and radium in the 70 to 200 pci/g range resulted in the extraction of >90% of the constituents. The residue left after chlorination was found to be innocuous and suitable for disposal as a waste acceptable to the environment. Use of sodium chloride in the charge was useful in reducing the chlorination temperature and in the formation of nonvolatile anionic chloro complexes of the metal ions in the ore

  3. Status Report from the United States of America [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    Kennedy, R H [United States Atomic Energy Commission, Washington, D.C. (United States)

    1967-06-15

    The US uranium production rate has been dropping gradually from a high of 17 760 tons in fiscal year 1961 to a level of about 10 400 tons in fiscal year 1966. As of 1 January 1966, there were 17 uranium mills in operation in the USA compared with a maximum of 26 during 1961, the peak production year. Uranium procurement contracts between the USAEC and companies operating 11 mills have been extended through calendar year 1970. The USAEC contracts for the other six mills are scheduled to expire 31 December 1966. Some of these mills, however, have substantial private orders for production of uranium for nuclear power plants and will continue to operate after completion of deliveries under USAEC contracts. No new uranium mills have been brought into production since 1962. Under these circumstances the emphasis in process development activities in recent years has tended toward improvements that could be incorporated within the general framework of the existing plants. Some major flowsheet changes have been made, however. For example, two of the ore-processing plants have shifted from acid leaching to sodium carbonate leach in order to provide the flexibility to process an increasing proportion of ores of high limestone content in the tributary areas. Several mills employing ion exchange as the primary step for recovery of uranium from solution have added an 'Eluex' solvent extraction step on the ion exchange eluate. This process not only results in a highgrade final product, but also eliminates several metallurgical problems formerly caused by the chloride and nitrate eluants. Such changes together with numerous minor improvements have gradually reduced production cost and increased recoveries. The domestic uranium milling companies have generally had reserves of normal-grade ores well in excess of the amounts required to fulfil the requirements for their contracts with the USAEC. Therefore, there has been little incentive to undertake the processing of lower grade

  4. Electrical conductivity of uranium-antimony oxide catalysts

    International Nuclear Information System (INIS)

    Golunski, S.E.; Nevell, T.G.; Hucknall, D.J.

    1985-01-01

    The relative ionic and electronic contributions to the electrical conductivity of a uranium-antimony oxide catalyst and of USbO 5 have been determined from measurements of a.c. and d.c. conductance. Under inert atmospheres (390 to 775 K) conduction in the catalyst (predominantly USb 3 O 10 together with small proportions of Sb 2 O 4 and USbO 5 ) is associated with both electronic and effectively charged atomic point defects. Only electronic conduction occurs in USbO 5 . Under oxygen (10 to 70 kPa, 493 to 682 K) both materials are n-type semiconductors at higher temperatures, but at lower temperatures semiconducting behaviour varies with the pressure of oxygen. Heating USbO 5 in oxygen induces an ionic contribution to conductivity. Ionic conduction in the catalyst is eliminated by heating in hydrogen or propene at 470 K but is restored by heating in oxygen. It is suggested that both charged oxygen vacancies and interstitial oxide ions are involved in interactions of gaseous components with uranium-antimony oxides. With alkenes, interstitial oxide ions give rise to the products of selective partial oxidation. (author)

  5. Electrical conductivity of uranium-antimony oxide catalysts

    Energy Technology Data Exchange (ETDEWEB)

    Golunski, S.E.; Nevell, T.G. (Portsmouth Polytechnic (UK)); Hucknall, D.J. (Southampton Univ. (UK). Dept. of Chemistry)

    1985-05-01

    The relative ionic and electronic contributions to the electrical conductivity of a uranium-antimony oxide catalyst and of USbO/sub 5/ have been determined from measurements of a.c. and d.c. conductance. Under inert atmospheres (390 to 775 K) conduction in the catalyst (predominantly USb/sub 3/O/sub 10/ together with small proportions of Sb/sub 2/O/sub 4/ and USbO/sub 5/) is associated with both electronic and effectively charged atomic point defects. Only electronic conduction occurs in USbO/sub 5/. Under oxygen (10 to 70 kPa, 493 to 682 K) both materials are n-type semiconductors at higher temperatures, but at lower temperatures semiconducting behaviour varies with the pressure of oxygen. Heating USbO/sub 5/ in oxygen induces an ionic contribution to conductivity. Ionic conduction in the catalyst is eliminated by heating in hydrogen or propene at 470 K but is restored by heating in oxygen. It is suggested that both charged oxygen vacancies and interstitial oxide ions are involved in interactions of gaseous components with uranium-antimony oxides. With alkenes, interstitial oxide ions give rise to the products of selective partial oxidation.

  6. Alpha Radiolysis of Sorbed Water on Uranium Oxides and Uranium Oxyfluorides

    Energy Technology Data Exchange (ETDEWEB)

    Icenhour, A.S.

    2003-09-10

    The radiolysis of sorbed water and other impurities contained in actinide oxides has been the focus of a number of studies related to the establishment of criteria for the safe storage and transport of these materials. Gamma radiolysis studies have previously been performed on uranium oxides and oxyfluorides (UO{sub 3}, U{sub 3}O{sub 8}, and UO{sub 2}F{sub 2}) to evaluate the long-term storage characteristics of {sup 233}U. This report describes a similar study for alpha radiolysis. Uranium oxides and oxyfluorides (with {sup 238}U as the surrogate for {sup 233}U) were subjected to relatively high alpha radiation doses (235 to 634 MGy) by doping with {sup 244}Cm. The typical irradiation time for these samples was about 1.5 years, which would be equivalent to more than 50 years irradiation by a {sup 233}U sample. Both dry and wet (up to 10 wt % water) samples were examined in an effort to identify the gas pressure and composition changes that occurred as a result of radiolysis. This study shows that several competing reactions occur during radiolysis, with the net effect that only very low pressures of hydrogen, nitrogen, and carbon dioxide are generated from the water, nitrate, and carbon impurities, respectively, associated with the oxides. In the absence of nitrate impurities, no pressures greater than 1000 torr are generated. Usually, however, the oxygen in the air atmosphere over the oxides is consumed with the corresponding oxidation of the uranium oxide. In the presence of up to 10 wt % water, the oxides first show a small pressure rise followed by a net decrease due to the oxygen consumption and the attainment of a steady-state pressure where the rate of generation of gaseous components is balanced by their recombination and/or consumption in the oxide phase. These results clearly demonstrate that alpha radiolysis of either wet or dry {sup 233}U oxides will not produce deleterious pressures or gaseous components that could compromise the long-term storage of

  7. Uranium accumulation in valentinite within the oxidation zone of an antimony occurrence

    Energy Technology Data Exchange (ETDEWEB)

    Sergeev, I.P.; Kurilo, M.V.

    1985-01-01

    As a result of mineralogic-radiogeochemical study of real composition of oxidation zone of antimony occurrence represented by quartz - antimonite vien in silicificated sandstones the previousy undescribed phenomenon of uranium concentration in valentinite Sb/sub 2/O/sub 3/ one of antimonite oxidation products has been found. According to the data of fission radiography the enrichment of valentinite with uranium is clearly seen, particularly of its concentrically zonal aggregates. The valentinite is the basic uranium mineral-concentrator in the oxidation zone, whereas in the initial mineral - antimonite as well as in the product of its more complete oxidation - stibiconite - uranium is lacking. Probably the crystallochemical properties of anhydrous antimony oxide (valentinite) by analogy with those for iron (goethite hematite) and titanium (leucoxene) are the most favourable for uranium adsorption as compared with high-water antimony oxides (stibiconite) or iron (limonite), which do not usually sorb uranium.

  8. Uranium accumulation in valentinite within the oxidation zone of an antimony occurrence

    International Nuclear Information System (INIS)

    Sergeev, I.P.; Kurilo, M.V.

    1985-01-01

    As a result of mineralogic-radiogeochemical study of real composition of oxidation zone of antimony occurrence represented by quartz - antimonite vien in silicificated sandstones the previousy undescribed phenomenon of uranium concentration in valentinite Sb 2 O 3 one of antimonite oxidation products has been found. According to the data of fission radiography the enrichment of valentinite with uranium is clearly seen, particularly of its concentrically zonal aggregates. The valentinite is the basic uranium mineral-concentrator in the oxidation zone, whereas in the initial mineral - antimonite as well as in the product of its more complete oxidation - stibiconite - uranium is lacking. Probably the crystallochemical properties of anhydrous antimony oxide (valentinite) by analogy with those for iron (goethite hematite) and titanium (leucoxene) are the most favourable for uranium adsorption as compared with high-water antimony oxides (stibiconite) or iron (limonite), which do not usually sorb uranium

  9. Status Report from Sweden [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, A [AB Atomenergi, Stockholm (Sweden)

    1967-06-15

    The Ministry of Education was authorized in November 1945 to appoint a commission to study the organization of nuclear energy research. In April 1947 this commission, the Swedish Atomic Energy Commission, proposed the formation of a semi-state-owned company to be a central body for applied research work and development in the nuclear energy field in Sweden. In November 1947 the Atomic Energy Company (AB Atomenergi) had its statutory meeting. The State owns 4/7 of the share capital and the remaining 3/7 is owned by 71 private and municipal share-holders. Except for a part of the stock capital, all investments and running costs of the company have been financed by the Government. The company is in practice answerable to the Department of Commerce which has an advisory body, the Atomic Energy Board. AB Atomenergi is responsible for Government-financed research on the industrial applications of nuclear energy, the milling of uranium ores and refining of uranium. The total number of employees is at present about 1400, 800 of which work at the company's research establishment Studsvik about 120 km south of Stockholm. As early as 1945 the Research Institute of the Swedish National Defence started work in the field of uranium processing. Similar work was also started quite early by the Boliden Mining Company, the Swedish Shale Oil Company and Wargons AB. After the establishment of AB Atomenergi, all work in the uranium processing field was transferred to this company. In fact one of the main reasons for the formation of AB Atomenergi was the need for Swedish uranium production as there was no possibility of importing uranium at that time. As a result of research and development in uranium processing a pilot plant at Kvarntorp near Orebro in central Sweden started milling a low-grade uranium ore (shale) in 1953. The capacity of this plant was 5-10 tons of uranium a year. A uranium mill at Ranstad in south-west Sweden, near Skovde, with a capacity of 120 tons of uranium a

  10. A case study of shrinkage-in place leaching of low grade uranium ore deposit

    International Nuclear Information System (INIS)

    Ding Dexin; Zhou Guohe

    1998-09-01

    A case study of shrinkage-in place leaching of low grade uranium ore deposit is dealt with. A test block was selected, and the shrinkage mining method was employed to construct the in place heap for leaching. Blast parameters and operations were carefully tried in order to make sure that the fragment size composition was adequate for leaching. A leaching system was planned and the corresponding leaching parameters were tried, too. The results show that the shrinkage method and the parameters for blasting and leaching are all adequate for the in-situ leaching of the blasted ore. This shrinkage-in place leaching system combines the mining and metallurgy processes into one and produces a lot of profits and could be applicable to many low grade uranium ore deposits which are so hard and compact that they have to be fragmented before being leached

  11. Surface Characterization and Electrochemical Oxidation of Metal Doped Uranium Dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeongmook; Kim, Jandee; Youn, Young-Sang; Kim, Jong-Goo; Ha, Yeong-Keong; Kim, Jong-Yun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Trivalent element in UO{sub 2} matrix makes the oxygen vacancy from loss of oxygen for charge compensation. Tetravalent element alters lattice parameter of UO{sub 2} due to diameter difference between the tetravalent element and replaced U. These structural changes have significant effect on not only relevant fuel performance but also the kinetics of fuel oxidation. Park and Olander explained the stabilization of Ln (III)-doped UO{sub 2} against oxidation based on oxygen potential calculations. In this work, we have been investigated the effect of Gd{sup 3+} and Th{sup 4+} doping on the UO{sub 2} structure with Raman spectroscopy and X-ray diffraction to characterize the surface structure of nuclear fuel material. For Gd doped UO{sub 2}, its electrochemical oxidation behaviors are also investigated. The Gd and Th doped uranium dioxide solid solution pellets with various doping level were investigated by XRD, Raman spectroscopy, SEM, electrochemical experiments to investigate surface structure and electro chemical oxidation behaviors. The lattice parameter evaluated from XRD spectra indicated the formation of solid solutions. Raman spectra showed the existence of the oxygen vacancy. SEM images showed the grain structure on the surface of Gd doped uranium dioxide depending on doping level and oxygen-to-metal ratio.

  12. An innovative jet boring mining method available for the high grade uranium ore underground deposits

    International Nuclear Information System (INIS)

    Narcy, J.L.

    1996-01-01

    An innovative mining method, based on the capability of a high pressure water jet to desaggregate rock, has been conceived and tested with success at the highest grade uranium ore deposit in the world, the Cigar Lake deposit in Saskatchewan, Canada. 113 tonnes of ore at 13% U were mined out by a new jet-boring mining method operated on a semi-industrial basis, in 1992 during the test mining program of Cigar Lake Project. (author). 9 figs

  13. Spectrophotometric study of bio-sorption of uranium on glass grade spodumene shell powder

    International Nuclear Information System (INIS)

    Parakudyil, A.S.; Pillai, A.K.; Reddy, A.V.R.; Singal, R.K.; Sharma, P.K.; Michael, K.M.

    2012-01-01

    Separation of uranium found in iron ore leachates was done by extraction chromatography using glass grade spodumene shellpowder (GSS) in nitric acid medium and analyzed spectrophotometrically. The influences of metal ion concentration, pH and adsorption capacity of biomass were investigated. Biosorption is a potential method of separation of heavy and trace metals from waste water and effluents from various sources. The adsorption capacities of biomass were investigated by batch experiments and column experiments. In the present study, glass grade spodumene shell powder (GSS) in acidic medium is being used as a biosorbent

  14. Radiation dose estimates from a mining plan for a high-grade uranium deposit

    International Nuclear Information System (INIS)

    Scott, L.M.

    1981-01-01

    The significance of gamma exposure to uranium miners has been recognized only in the last few years. Most ore deposits which have been underground mined, were 1% or less U 3 O 8 . Full-time mining of this grade ore can result in exposure exceeding 1 Rem per year. Several companies in Saskatchewan are planning to mine recently discovered ore bodies which contain ore pods in excess of 10% U 3 O 8 . The purpose of this paper is to present dose data which can be used to estimate gamma exposure from high-grade ore deposits, and to present mining techniques which will minimize miner exposure

  15. Biomineral processing of high apatite containing low-grade indian uranium ore

    Energy Technology Data Exchange (ETDEWEB)

    Abhilash; Mehta, K.D.; Pandey, B.D., E-mail: biometnml@gmail.com [National Metallurgical Laboratory (CSIR), Jamshedpur (India); Ray, L. [Jadavpur Univ., FTBE Dept., Kolkata (India); Tamrakar, P.K. [Uranium Corp. of India Limited, CR& D Dept., Jaduguda (India)

    2010-07-01

    Microbial species isolated from source mine water, primarily an enriched culture of Acidithiobacillus ferrooxidans was employed for bio-leaching of uranium from a low-grade apatite rich uranium ore of Narwapahar Mines, India while varying pH, pulp density (PD), particle size, etc. The ore (0.047% U{sub 3}O{sub 8}), though of Singhbhum area (richest deposit of uranium ores in India), due to presence of some refractory minerals and high apatite (5%) causes a maximum 78% recovery through conventional processing. Bioleaching experiments were carried out by varying pH at 35{sup o}C using 20%(w/v) PD and <76μm size particles resulting in 83.5% and 78% uranium bio-recovery at 1.7 and 2.0 pH in 40 days as against maximum recovery of 46% and 41% metal in control experiments respectively. Finer size (<45μm) ore fractions exhibited higher uranium dissolution (96%) in 40 days at 10% (w/v) pulp density (PD), 1.7 pH and 35{sup o}C. On increasing the pulp density from 10% to 20% under the same conditions, the biorecovery of uranium fell down from 96% to 82%. The higher uranium dissolution during bioleaching at 1.7 pH with the fine size particles (<45μm) can be correlated with increase in redox potential from 598 mV to 708 mV and the corresponding variation of Fe(III) ion concentration in 40 days. (author)

  16. Biomineral processing of high apatite containing low-grade indian uranium ore

    International Nuclear Information System (INIS)

    Abhilash; Mehta, K.D.; Pandey, B.D.; Ray, L.; Tamrakar, P.K.

    2010-01-01

    Microbial species isolated from source mine water, primarily an enriched culture of Acidithiobacillus ferrooxidans was employed for bio-leaching of uranium from a low-grade apatite rich uranium ore of Narwapahar Mines, India while varying pH, pulp density (PD), particle size, etc. The ore (0.047% U_3O_8), though of Singhbhum area (richest deposit of uranium ores in India), due to presence of some refractory minerals and high apatite (5%) causes a maximum 78% recovery through conventional processing. Bioleaching experiments were carried out by varying pH at 35"oC using 20%(w/v) PD and <76μm size particles resulting in 83.5% and 78% uranium bio-recovery at 1.7 and 2.0 pH in 40 days as against maximum recovery of 46% and 41% metal in control experiments respectively. Finer size (<45μm) ore fractions exhibited higher uranium dissolution (96%) in 40 days at 10% (w/v) pulp density (PD), 1.7 pH and 35"oC. On increasing the pulp density from 10% to 20% under the same conditions, the biorecovery of uranium fell down from 96% to 82%. The higher uranium dissolution during bioleaching at 1.7 pH with the fine size particles (<45μm) can be correlated with increase in redox potential from 598 mV to 708 mV and the corresponding variation of Fe(III) ion concentration in 40 days. (author)

  17. Estimation of intermediate-grade uranium resources II. Proposed method for estimating intermediate-grade uranium resources in roll-front deposits. Final report

    International Nuclear Information System (INIS)

    Lambie, F.W.; Yee, S.N.

    1981-09-01

    The purpose of this and a previous project was to examine the feasibility of estimating intermediate grade uranium (0.01 to 0.05% U 3 O 8 ) on the basis of existing, sparsely drilled holes. All data are from the Powder River Basin in Wyoming. DOE makes preliminary estimates of endowment by calculating an Average Area of Influence (AAI) based on densely drilled areas, multiplying that by the thickness of the mineralization and then dividing by a tonnage factor. The resulting tonnage of ore is then multiplied by the average grade of the interval to obtain the estimate of U 3 O 8 tonnage. Total endowment is the sum of these values over all mineralized intervals in all wells in the area. In regions where wells are densely drilled and approximately regularly spaced this technique approaches the classical polygonal estimation technique used to estimate ore reserves and should be fairly reliable. The method is conservative because: (1) in sparsely drilled regions a large fraction of the area is not considered to contribute to endowment; (2) there is a bias created by the different distributions of point grades and mining block grades. A conservative approach may be justified for purposes of ore reserve estimation, where large investments may hinge on local forecasts. But for estimates of endowment over areas as large as 1 0 by 2 0 quadrangles, or the nation as a whole, errors in local predictions are not critical as long as they tend to cancel and a less conservative estimation approach may be justified.One candidate, developed for this study and described is called the contoured thickness technique. A comparison of estimates based on the contoured thickness approach with DOE calculations for five areas of Wyoming roll-fronts in the Powder River Basin is presented. The sensitivity of the technique to well density is examined and the question of predicting intermediate grade endowment from data on higher grades is discussed

  18. National Uranium Resource Evaluation: intermediate-grade uranium resource assessment project for part of the Maybell District, Sand Wash Basin, Colorado

    International Nuclear Information System (INIS)

    Goodknight, C.S.

    1983-04-01

    Intermediate-grade uranium resources in the Miocene Browns Park Formation were assessed for part of the Maybell district in the Sand Wash Basin, Colorado, as part of the National Uranium Resource Evaluation program conducted by Bendix Field Engineering Corporation for the US Department of Energy. Two sites, each 2 mi 2 (5 km 2 ) in size, in the district were selected to be assessed. Site selection was based on evaluation of geologic, geophysical, and geochemical data that were collected from a larger project area known to contain uranium enrichment. The assessment of the sites was accomplished primarily by drilling 19 holes through the Browns Park Formation and by using the geophysical and geochemical data from those holes and from a larger number of industry-drilled holes. Analytical results of samples from uranium prospects, mainly along faults in the sites, were also used for the assessment. Data from surface samples and from drill-hole samples and logs of the site south of Lay Creek indicate that no intermediate-grade uranium resources are present. However, similar data from the site north of Lay Creek verify that approximately 25 million lb (11.2 million kg) of intermediate-grade uranium resources may be present. This assessment assumes that an average uranium-enriched thickness of 10 ft (3 m) at a grade of 0.017% U 3 O 8 is present in at least two thirds of the northern site. Uranium enrichment in this site occurs mainly in the lower 150 ft (45 m) of the Browns Park Formation in fine- to medium-grained sandstone that contains abundant clay in its matrix. Facies variations within the Browns Park preclude correlation of individual beds or zones of uranium enrichment between closely spaced drill holes

  19. Behavior of Colorado Plateau uranium minerals during oxidation

    Science.gov (United States)

    Garrels, Robert Minard; Christ, C.L.

    1956-01-01

    Uranium occurs as U(VI) and U(IV) in minerals of the Colorado Plateau ores. The number of species containing U(VI) is large, but only two U(IV) minerals are known from the Plateau: uraninite, and oxide, and coffinite, a hydroxy-silicate. These oxidize to yield U(VI) before reacting significantly with other mineral constituents. Crystal-structure analysis has shown that U(VI) invariable occurs as uranyl ion, UO2+2. Uranyl ion may form complex carbonate or sulfate ions with resulting soluble compounds, but only in the absence of quinquevalent vanadium, arsenic, or phosphorous. In the presence of these elements in the +5 valence state, the uranyl ion is fixed in insoluble layer compounds formed by union of uranyl ion with orthovanadate, orthophosphate, or orthoarsenate. Under favorable conditions UO2+2 may react to form the relatively insoluble rutherfordine, UO2CO3, or hydrated uranyl hydroxides. These are rarely found on the Colorado Plateau as opposed to their excellent development in other uraniferous areas, a condition which is apparently related to the semiarid climate and low water table of the Plateau. Uranium may also be fixed as uranyl silicate, but little is known about minerals of this kind. In the present study emphasis has been placed on a detailing of the chemical and crystal structural changes which occur in the oxidation paragenetic sequence.

  20. Small cell experiments for electrolytic reduction of uranium oxides to uranium metal using fluoride salts

    International Nuclear Information System (INIS)

    Haas, P.A.; Adcock, P.W.; Coroneos, A.C.; Hendrix, D.E.

    1994-01-01

    Electrolytic reduction of uranium oxide was proposed for the preparation of uranium metal feed for the atomic vapor laser isotope separation (AVLIS) process. A laboratory cell of 25-cm ID was operated to obtain additional information in areas important to design and operation of a pilot plant cell. Reproducible test results and useful operating and control procedures were demonstrated. About 20 kg of uranium metal of acceptable purity were prepared. A good supply of dissolved UO 2 feed at the anode is the most important controlling requirement for efficient cell operation. A large fraction of the cell current is nonproductive in that it does not produce a metal product nor consume carbon anodes. All useful test conditions gave some reduction of UF 4 to produce CF 4 in addition to the reduction of UO 2 , but the fraction of metal from the reduction of UF 4 can be decreased by increasing the concentration of dissolved UO 2 . Operation of large continuous cells would probably be limited to current efficiencies of less than 60 pct, and more than 20 pct of the metal would result from the reduction of UF 4

  1. Sol-gel preparation of uranium oxide spheres

    International Nuclear Information System (INIS)

    Dolezal, J.; Urbanek, V.

    1978-01-01

    Information is presented on problems of preparing nuclear fuel by the sol-gel method. Basic data on different process types are given. A more detailed description of the method of preparation of spherical particles of uranium oxide gel developed and used at the Nuclear Research Institute at Rez is given. Advantages and disadvantages of sol-gel materials are discussed in comparison with fuel materials prepared by classical precipitation methods. The feasibility of the sol-gel methods for preparing other materials is shortly mentioned and their application outlined. (author)

  2. Uranium reduction by carbon oxide during ore formation

    International Nuclear Information System (INIS)

    Matyash, I.V.; Gavrusevich, I.B.; Pasal'skaya, L.F.; Shcherba, D.I.

    1981-01-01

    Using the method of gas chromatography the gas content in Pre-Cambrian granitoils of various types and in natrometasomatites associted with them is studied. It is established that granites associated with ore-bearing albitites have sharply elevated amounts of CO as compared with granites, which do not include mineralization. Simultaneously in ore samples the absence or sharply low amounts of CO as compared with ore-free samples is observed, that is reverse dependence of CO and ore components. Carbon oxide is the reducing agent of uranium mineralization and alongside with other reducing agents can be a geochemical barrier in the process of ore formation [ru

  3. Chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade mixed oxides [(U,Pu)O2

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    Mixed oxide, a mixture of uranium and plutonium oxides, is used as a nuclear-reactor fuel in the form of pellets. The plutonium content may be up to 10 wt %, and the diluent uranium may be of any U-235 enrichment. In order to be suitable for use as a nuclear fuel, the material must meet certain criteria for combined uranium and plutonium content, effective fissile content, and impurity content. Analytical procedures used to determine if mixed oxides comply with specifications are: uranium by controlled-potential coulometry; plutonium by controlled-potential coulometry; plutonium by amperometric titration with iron (II); nitrogen by distillation spectrophotometry using Nessler reagent; carbon (total) by direct combustion-thermal-conductivity; total chlorine and fluorine by pyrohydrolysis; sulfur by distillation-spectrophotometry; moisture by the coulometric, electrolytic moisture analyzer; isotopic composition by mass spectrometry; rare earths by copper spark spectroscopy; trace impurities by carrier distillation spectroscopy; impurities by spark-source mass spectrography; total gas in reactor-grade mixed dioxide pellets; tungsten by dithiol-spectrophotometry; rare earth elements by spectroscopy; plutonium-238 isotopic abundance by alpha spectrometry; uranium and plutonium isotopic analysis by mass spectrometry; oxygen-to-metal atom ratio by gravimetry

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

    Science.gov (United States)

    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.

  5. Method of converting uranium fluoride to intermediate product for uranium oxide manufacture with recycling or reusing valuable materials

    International Nuclear Information System (INIS)

    Baran, V.; Moltasova, J.

    1982-01-01

    Uranium fluoride is acted upon by water with nitrate containing a cation capable of binding fluoride ions. The uranium is extracted, for instance, with tributyl phosphate with the generated organic phase containing the prevalent proportion of uranium and representing the required intermediate product and the aqueous phase from which is isolated the fluorine component which may be used within the fuel cycle. The nitrate component of the aqueous phase is recycled following treatment. It is also possible to act on uranium fluoride directly with an aqueous solution. Here the cations of nitrate form with the fluorides soluble nondissociated complexes and reduce the concentration of free fluoride ions. The nitrate +s mostly used in an amount corresponding to its solubility in the system prior to the introduction of UF 6 . The uranium from the solution with the reduced concentration of free fluoride ions is extracted into the reaction system under such conditions as to make the prevalent majority of fluorides and an amount of uranium smaller than 5x10 -2 mol/l remain in the aqueous phase and that such an amount of fluorides should remain in the organic phase which is smaller than corresponds to the fluorine/uranium molar ratio in the organic phase. Uranium contained in the organic phase is processed into uranium oxide, with advantage into UO 2 . From the isolated compounds of fluorine and the cation of the nitrate gaseous HF is released which is used either inside or outside of the fuel cycle. (J.P.)

  6. Status Report from Canada [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    Thunaes, A [Eldorado Mining and Refining Ltd., Ottawa (Canada)

    1967-06-15

    The Canadian production of uranium increased in a spectacular manner during the period 1955-1959 from 1000 to 15 500 tons U{sub 3}O{sub 8} per year. Since 1959 the production has declined to the 1966 level of 3900 tons U{sub 3}O{sub 8} per year; stretch-out of contracts and government stockpiling programmes has made the decline gradual, and is maintaining the current rate of production until 1970. Nineteen mills were in operation during the period of peak production but only three are operating today. Ten mills were shut down and dismantled because of exhaustion of ore bodies or because the operation was uneconomical; six mills are maintained in stand-by condition. The total daily capacity of mills in operation or standing by is about 28 000 tons ore, but some of these mills would not be reopened unless an appreciable increase in uranium price occurs. The tide of uranium demand is about ready to turn and prospecting for uranium is very active this year, particularly in the Elliot Lake and Beaverlodge areas. The estimates for uranium demand in 1975-1980 are such that new ore will have to be found and developed, and new treatment plants must be built. The new ore that is found will likely be of lower grade or more expensive to mine than most of the current proven reserves in Canada and the most efficient methods of treatment will be needed to avoid excessive increases in production costs. This seems an opportune time to review Canadian milling of uranium ore, the improvements that have been made and development work towards further improvements.

  7. Standard test methods for chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade uranium dioxide powders and pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1999-01-01

    1.1 These test methods cover procedures for the chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade uranium dioxide powders and pellets to determine compliance with specifications. 1.2 This test method covers the determination of uranium and the oxygen to uranium atomic ratio in nuclear-grade uranium dioxide powder and pellets. 1.4 This test method covers the determination of chlorine and fluorine in nuclear-grade uranium dioxide. With a 1 to 10-g sample, concentrations of 5 to 200 g/g of chlorine and 1 to 200 μg/g of fluorine are determined without interference. 1.5 This test method covers the determination of moisture in uranium dioxide samples. Detection limits are as low as 10 μg. 1.6 This test method covers the determination of nitride nitrogen in uranium dioxide in the range from 10 to 250 μg. 1.7 This test method covers the spectrographic analysis of nuclear-grade UO2 for the 26 elements in the ranges indicated in Table 2. 1.8 For simultaneous determination of trace ele...

  8. Recent Pilot Plant Experience on Alkaline Leaching of Low Grade Uranium Ore in India

    Energy Technology Data Exchange (ETDEWEB)

    Suri, A. K; Ghosh, S. K.; Padmanabhan, N. P.H., [Bhabha Atomic Research Centre, Mumbai (India)

    2014-05-15

    Uranium deposits in India are low grade and are relatively smaller in extent as compared to present worldwide commercial practice. So far, the vein type deposits of Singhbhum Thrust Belt (STB) are being exploited for meeting the Indian requirements of uranium. The deposits are currently processed by acid leaching in the mills located at Jaduguda and Turamdih near Jamshedpur in Jharkhand State of India. The deposits at Jaduguda and Narwapahar are being mined by underground mining and are processed in Jaduguda mill using airagitated Pachucas. The deposits at Banduhurang and Turamdih are being mined by open cast and underground mining respectively and are processed at Turamdih by acid leaching in mechanically agitated reactors. The occurrences of uranium in North East and Northern part of Kadapa basin are relatively moderate in size and are expected to be processed in the near future by acid leaching. Uranium is also found to occur near Tummalapalle in granitic and limestone host rocks in Southern part of Kadapa basin (Andhra Pradesh) and in Gogi in Bhima basin (Karnataka). The deposit in Tummalapalle is relatively lower in grade (≈ 0.042% U{sub 3}O{sub 8}) but is a reasonably large reserve, whereas that in Gogi is rich in uranium content (≈0.18% U{sub 3}O{sub 8}) but is relatively small reserve. Laboratory tests based on alkaline leaching have been carried out on both types of deposits. Studies for Tummalapalle deposits have been extended to pilot plant level and a complete flow sheet has been established with the regeneration and recirculation of lixiviants and recovery of sodium sulphate as a by-product. The process involves alkaline leaching under oxygen pressure in batch type and/or continuous leach reactor using sodium carbonate/bicarbonate as a leaching media and uranium is recovered as sodium diuranate. Based on the techno-economic evaluation of the process, an industrial scale mill (3 000 tonnes ore/day) is being set up at Tummalapalle in Andhra Pradesh

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

    OpenAIRE

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

    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) interfacia...

  10. Comparison of Oxidation Characteristics of Selected Nuclear Graphite Grades

    International Nuclear Information System (INIS)

    Chi, Se Hwan; Kim, Gen Chan

    2010-02-01

    The oxidation behavior of some selected nuclear graphite grades (i.e., IG-110, IG-430, NBG-18, NBG-25) were compared in view of their filler coke type and the physical property of the grades. Oxidation rates were determined at six temperatures between 600 ∼ 960 .deg. C in air by using a three-zone vertical tube furnace at a 10 L/min air flow rate. The specimens were a cylinder with a 25.4 mm diameter and a 25.4 mm length. Results showed that, even though the four examined nuclear graphite grades showed a highly temperature-sensitive oxidation behavior through out the test temperature range of 600 ∼ 950 .deg. C, the differences between the grades were not significant. The oxidation rates determined for a 5∼10 % weight loss at the six temperatures were nearly the same except for 702 and 808 .deg. C, where the pitch coke graphites showed an apparent decrease in their oxidation rate, more so than the petroleum coke graphites. These effects of the coke type reduced or nearly disappeared with an increasing temperature. The average activation energy determined for 608 ∼ 808 .deg. C was 161.5 ± 7.3 kJ/mol, showing that the dominant oxidation reaction occurred by a chemical control

  11. On bacteria oxidizing enlargement scale test for uranium in-situ leaching at. 381 mine

    International Nuclear Information System (INIS)

    Hu Kaiguang; Wang Qingliang; Liu Yingjiu; Shi Wenge; Hu Shihe; Hu Yincai; Fang Qiu

    1999-01-01

    The results of enlarged scale test of bacteria as oxidizer for uranium in-situ leaching at No 381 mine showed that redox potential of the oxidized absorbed tailing water by bacteria is more than 510 mV, without any effects on after treatments by using bacteria as oxidizer and reduce oxidizer costs 70% compared with H 2 O 2 as oxidizer

  12. Determination of the impurities Al, Mn, Fe, Ni, Cu and Zn in nuclear grade uranium by ICP-OES

    Energy Technology Data Exchange (ETDEWEB)

    Kakazu, Mauricio H.; Cotrim, Marycel E.B.; Silva, Douglas B. da; Pires, Maria Aparecida F., E-mail: mhkakazu@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    Uranium compounds are specially produced for use in nuclear reactors and must meet strict physical and chemicals specifications. The current study discusses the separation of Al, Mn, Fe, Ni, Cu and Zn from uranium compounds and their quantitative determination using inductively coupled plasma optical emission spectrometry. To avoid interference effect caused by uranium in the spectroscopic emission lines of elements of interest, the chromatographic behavior of TBP impregnated macroporous Amberlite XAD-4 column was investigated. The break through curves of uranium obtained showed maximum retention of up to 11 grams of uranium and the elution curves for the elements showed reproducible recovery rate of 90% on 50 mL elution. Synthetic samples of nuclear grade uranium prepared with the addition of 250 {mu}g/g of U of the above individual elements were used to verify the performance of the method. The method showed a recovery rate of 80 to 90% for the elements under study. (author)

  13. Determination of the impurities Al, Mn, Fe, Ni, Cu and Zn in nuclear grade uranium by ICP-OES

    International Nuclear Information System (INIS)

    Kakazu, Mauricio H.; Cotrim, Marycel E.B.; Silva, Douglas B. da; Pires, Maria Aparecida F.

    2011-01-01

    Uranium compounds are specially produced for use in nuclear reactors and must meet strict physical and chemicals specifications. The current study discusses the separation of Al, Mn, Fe, Ni, Cu and Zn from uranium compounds and their quantitative determination using inductively coupled plasma optical emission spectrometry. To avoid interference effect caused by uranium in the spectroscopic emission lines of elements of interest, the chromatographic behavior of TBP impregnated macroporous Amberlite XAD-4 column was investigated. The break through curves of uranium obtained showed maximum retention of up to 11 grams of uranium and the elution curves for the elements showed reproducible recovery rate of 90% on 50 mL elution. Synthetic samples of nuclear grade uranium prepared with the addition of 250 μg/g of U of the above individual elements were used to verify the performance of the method. The method showed a recovery rate of 80 to 90% for the elements under study. (author)

  14. Grain growth kinetics in uranium-plutonium mixed oxides

    International Nuclear Information System (INIS)

    Sari, C.

    1986-01-01

    Grain growth rates were investigated in uranium-plutonium mixed oxide specimens with oxygen-to-metal ratios 1.97 and 2.0. The specimens in the form of cylindrical pellets were heated in a temperature gradient similar to that existing in a fast reactor. The results are in agreement with the cubic rate law. The mean grain size D(μm) after annealing for time t (min) is represented by D 3 -D 0 3 =1.11x10 12 . exp(-445870/RT).t and D 3 -D 0 3 =2.55x10 9 .exp(-319240/RT).t for specimens with overall oxygen-to-metal ratios 1.97 and 2.0, respectively (activation energies expressed in J/mol). An example for the influence of the oxygen-to-metal ratio on the grain growth in mixed oxide fuel during operation in a fast reactor is also given. (orig.)

  15. Characterization of depleted uranium oxides fabricated using different processing methods

    International Nuclear Information System (INIS)

    Hastings, E.P.; Lewis, C.; FitzPatrick, J.; Rademacher, D.; Tandon, L.

    2008-01-01

    Identifying both physical and chemical characteristics of Special Nuclear Material (SNM) production processes is the corner stone of nuclear forensics. Typically, processing markers are based on measuring an interdicted sample's bulk chemical properties, such as the elemental or isotopic composition, or focusing on the chemical and physical morphology of only a few particles. Therefore, it is imperative that known SNM processes be fully characterized from bulk to trace level for each particle size range. This report outlines a series of particle size measurements and fractionation techniques that can be applied to a bulk SNM powders, categorizing both chemical and physical properties in discrete particle size fractions. This will be demonstrated by characterizing the process signatures of a series of different depleted uranium oxides prepared at increasing firing temperatures (350-1100 deg C). Results will demonstrate how each oxides' material density, particle size distribution, and morphology varies. (author)

  16. Study of uranium oxidation states in geological material.

    Science.gov (United States)

    Pidchenko, I; Salminen-Paatero, S; Rothe, J; Suksi, J

    2013-10-01

    A wet chemical method to determine uranium (U) oxidation states in geological material has been developed and tested. The problem faced in oxidation state determinations with wet chemical methods is that U redox state may change when extracted from the sample material, thereby leading to erroneous results. In order to quantify and monitor U redox behavior during the acidic extraction in the procedure, an analysis of added isotopic redox tracers, (236)U(VI) and (232)U(IV), and of variations in natural uranium isotope ratio ((234)U/(238)U) of indigenous U(IV) and U(VI) fractions was performed. Two sample materials with varying redox activity, U bearing rock and U-rich clayey lignite sediment, were used for the tests. The Fe(II)/Fe(III) redox-pair of the mineral phases was postulated as a potentially disturbing redox agent. The impact of Fe(III) on U was studied by reducing Fe(III) with ascorbic acid, which was added to the extraction solution. We observed that ascorbic acid protected most of the U from oxidation. The measured (234)U/(238)U ratio in U(IV) and U(VI) fractions in the sediment samples provided a unique tool to quantify U oxidation caused by Fe(III). Annealing (sample heating) to temperatures above 500 °C was supposed to heal ionizing radiation induced defects in the material that can disturb U redox state during extraction. Good agreement between two independent methods was obtained for DL-1a material: an average 38% of U(IV) determined by redox tracer corrected wet chemistry and 45% for XANES. Copyright © 2013 Elsevier Ltd. All rights reserved.

  17. The cathodic reduction of dioxygen on uranium oxide in dilute alkaline aqueous solution

    International Nuclear Information System (INIS)

    Hocking, W.H.; Betteridge, J.S.; Shoesmith, D.W.

    1991-09-01

    The cathodic reduction of dioxygen on uranium oxide in dilute alkaline aqueous solutions has been investigated within the context of a program to develop a comprehensive model to predict the behaviour of used CANDU (Canada Deuterium Uranium) nuclear fuel under disposal-vault conditions. Two different kinds of ceramic UO 2 were studied: reactor-grade CANDU fuel with normal p-type electrical conductivity and low-resistance material that exhibits n-type photoelectrochemical behaviour. The transport of electroactive species in solution was controlled by varying the rotation rate of rotating disc electrodes (RDE) and rotating ring-disc electrodes (RRDE). Steady-state polarization measurements were made using the current-interrupt method to compensate for the potential drop caused by ohmic resistance. Any release of peroxide to solution from the UO 2 (disc) surface could be monitored by oxidizing it at the Au ring of an RRDE. The existing theory for the cathodic 0 2 -reduction process as applied to RDE and RRDE experiments has been reviewed as a starting point for the interpretation of the results obtained in our work. (37 figs., 2 tabs., 170 refs.)

  18. Production of uranium metal via electrolytic reduction of uranium oxide in molten LiCl and salt distillation

    International Nuclear Information System (INIS)

    Eun-Young Choi; Chan Yeon Won; Dae-Seung Kang; Sung-Wook Kim; Ju-Sun Cha; Sung-Jai Lee; Wooshin Park; Hun Suk Im; Jin-Mok Hur

    2015-01-01

    Recovery of metallic uranium has been achieved by electrolytic reduction of uranium oxide in a molten LiCl-Li 2 O electrolyte at 650 deg C, followed by the removal of the residual salt by vacuum distillation at 850 deg C. Four types of stainless steel mesh baskets, with various mesh sizes (325, 1,400 and 2,300 meshes) and either three or five ply layers, were used both as cathodes and to contain the reduced product in the distillation stage. The recovered uranium had a metal fraction greater than 98.8 % and contained no residual salt. (author)

  19. Determination of the oxygen-metal-ratio of uranium-americium mixed oxides

    International Nuclear Information System (INIS)

    Bartscher, W.

    1982-01-01

    During the dissolution of uranium-americium mixed oxides in phosphoric acid under nitrogen tetravalent uranium is oxidized by tetravalent americium. The obtained hexavalent uranium is determined by constant potential coulometry. The coulombs measured are equivalent to the oxygen in excess of the minimum composition of UO 2 x AmO 1 . 5 . The total uranium content of the sample is determined in a subsequent coulometric titration. The oxygen-metal ratio of the sample can be calculated for a given uranium-americium ratio. An excess of uranium dioxide is necessary in order to suppress the oxidation of water by tetravalent americium. The standard deviation of the method is 0.0017 O/M units. (orig.) [de

  20. Status Report from Yugoslavia [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    Bunji, B [Institute for Technology of Nuclear and Other Raw Materials, Belgrade, Yugoslavia (Serbia)

    1967-06-15

    Full text: The greater part of our activities is connected with the problem of extracting uranium from low-grade ores. In this paper, a brief review of the most important recent developments will be presented. In this connection, it may be useful to determine the definition of low-grade ores. This term can be applied to ore from which the uranium content cannot be extracted under normal economic conditions. Thus this term can be applied to uranium-bearing material with a uranium content of no more than 0. 05%. But, in general, it could be said that there is a very large range of uranium content where uranium extraction may not be economic for such different reasons as; (a) the size or other facts in connection with the orebodies themselves; (b) refractory ore; or (c) other local conditions. During research on the treatment of low-grade ore from the deposit at Gabrovnica (Stara Planina, Yugoslavia) it became apparent that an alkaline leaching process would have to be carried out. The treatment of this granitic type of ore causes no particular difficulties. The required temperature is about 90{sup o}C. The retention time in the leaching stage is from 4 to 12 hours. Sodium carbonate consumption is not higher than 15 kg/t of ore. Pachuca-type leaching shows satisfactory maintenance and processing costs. At Kalna uranium precipitation by means of hydrogen pressure reduction has been developed, and is being developed and investigated in full-scale operation. Details of the process were published in Geneva in 1963. On the basis of the experience gained from full-scale operation, many refinements and cost-saving changes have been made. A normal steel wire screen used as a catalyst carrier shows a very good improvement over free-moving UO{sub 2} as catalyst. In large-scale operation (200 t/d), after the precipitation of uranium the barren solution content is about 1 g U/m{sup 3}. The content of the pregnant solution is of the order of 300-600 g/m{sup 3}. Recycling the

  1. Studies on the recovery of uranium from low-grade ores in India

    International Nuclear Information System (INIS)

    Jayaram, K.M.V.; Dwivedy, K.K.; Deshpande, A.S.; Ramachar, T.M.

    1976-01-01

    Investigations were carried out to utilize the available para-marginal and low-grade ores - chlorite schists, amphibolites, carbonate ores, clays and quartzites - analysing between 0.027 and 0.08% U 3 O 8 . In addition, tests were undertaken on the technical and economic feasibility of recovering uranium as a byproduct from the copper flotation tailings and phosphorites. Heap and bacterial leaching tests were conducted on quartz-chlorite schists from the Singhbhum district, Bihar, analysing about 0.03% U 3 O 8 . Studies also showed that the ores harbour active Ferrobacillus ferrooxidans. Studies on 10-mesh samples of amphibolites from Inderwa, Bihar, (0.08% U 3 O 8 ) showed that only 32.8% recovery could be obtained by wet tabling and 85% by agitation leaching, while static leaching tests yielded 81% recovery in 24 hours of contact time. Similar tests on calcareous phyllites (0.05% U 3 O 8 ) with 30 kg/t Na 2 CO 3 and 8 kg/t NaHCO 3 yielded 86% uranium leachability at ambient temperature. Biogenic uraniferous clay from Udaisagar (0.029% U 3 O 8 ) yielded 43.3% uranium recovery using 1000 l/t of neutral water for 6 h. Percolation leaching tests were conducted with hard quartzites (0.06% U 3 O 8 ), and the results showed that 81% uranium could be recovered in 24 days. Although preliminary ore dressing studies on tailings obtained from the copper flotation (0.013% U 3 O 8 ) at Surda yielded a concentrate analysing 0.063% U 3 O 8 at 66% recovery, recent tests on the tailings from the copper concentrator indicated only 48% recovery at a grade of 0.112% owing to decrease in the feed grade. Studies on the utilization of large-capacity gravity machines and selective mining of uranium-rich copper lodes may render this source economic. Preliminary studies on a phosphorite sample containing 22.0% P 2 O 5 and 0.04% U 3 O 8 from the Mussorie area in Uttar Pradesh on calcination followed by scrubbing yielded a sand enriched in P 2 O 5 values (33.7% P 2 O 5 at 92.5% recovery) but

  2. A gravimetric method for the determination of oxygen in uranium oxides and ternary uranium oxides by addition of alkaline earth compounds

    International Nuclear Information System (INIS)

    Fujino, Takeo; Tagawa, Hiroaki; Adachi, Takeo; Hashitani, Hiroshi

    1978-01-01

    A simple gravimetric determination of oxygen in uranium oxides and ternary uranium oxides is described. In alkaline earth uranates which are formed by heating in air at 800-1100 0 C, uranium is in the hexavalent state over certain continuous ranges of alkaline earth-to-uranium ratios. Thus, if an alkaline earth uranate or a compound containing an alkaline earth element, e.g. MgO, is mixed with the oxide sample and heated in air under suitable conditions, oxygen can be determined from the weight change before and after the reaction. The standard deviation of the O:U ratio for a UOsub(2+x) test sample is +-0.0008-0.001, if a correction is applied for atmospheric moisture absorbed during mixing. (Auth.)

  3. Reoxidation of uranium in electrolytically reduced simulated oxide fuel during residual salt distillation

    International Nuclear Information System (INIS)

    Eun-Young Choi; Jin-Mok Hur; Min Ku Jeon; University of Science and Technology, Yuseong-gu, Daejeon

    2017-01-01

    We report that residual salt removal by high-temperature distillation causes partial reoxidation of uranium metal to uranium oxide in electrolytically reduced simulated oxide fuel. Specifically, the content of uranium metal in the above product decreases with increasing distillation temperatures, which can be attributed to reoxidation by Li 2 O contained in residual salt (LiCl). Additionally, we estimate the fractions of Li 2 O reacted with uranium metal under these conditions, showing that they decrease with decreasing temperature, and calculate some thermodynamic parameters of the above reoxidation. (author)

  4. Preparation and Characterization of Uranium Oxides in Support of the K Basin Sludge Treatment Project

    Energy Technology Data Exchange (ETDEWEB)

    Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2008-07-08

    Uraninite (UO2) and metaschoepite (UO3·2H2O) are the uranium phases most frequently observed in K Basin sludge. Uraninite arises from the oxidation of uranium metal by anoxic water and metaschoepite arises from oxidation of uraninite by atmospheric or radiolytic oxygen. Studies of the oxidation of uraninite by oxygen to form metaschoepite were performed at 21°C and 50°C. A uranium oxide oxidation state characterization method based on spectrophotometry of the solution formed by dissolving aqueous slurries in phosphoric acid was developed to follow the extent of reaction. This method may be applied to determine uranium oxide oxidation state distribution in K Basin sludge. The uraninite produced by anoxic corrosion of uranium metal has exceedingly fine particle size (6 nm diameter), forms agglomerates, and has the formula UO2.004±0.007; i.e., is practically stoichiometric UO2. The metaschoepite particles are flatter and wider when prepared at 21°C than the particles prepared at 50°C. These particles are much smaller than the metaschoepite observed in prolonged exposure of actual K Basin sludge to warm moist oxidizing conditions. The uraninite produced by anoxic uranium metal corrosion and the metaschoepite produced by reaction of uraninite aqueous slurries with oxygen may be used in engineering and process development testing. A rapid alternative method to determine uranium metal concentrations in sludge also was identified.

  5. Preparation and Characterization of Uranium Oxides in Support of the K Basin Sludge Treatment Project

    International Nuclear Information System (INIS)

    Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2008-01-01

    Uraninite (UO2) and metaschoepite (UO3-2H2O) are the uranium phases most frequently observed in K Basin sludge. Uraninite arises from the oxidation of uranium metal by anoxic water and metaschoepite arises from oxidation of uraninite by atmospheric or radiolytic oxygen. Studies of the oxidation of uraninite by oxygen to form metaschoepite were performed at 21 C and 50 C. A uranium oxide oxidation state characterization method based on spectrophotometry of the solution formed by dissolving aqueous slurries in phosphoric acid was developed to follow the extent of reaction. This method may be applied to determine uranium oxide oxidation state distribution in K Basin sludge. The uraninite produced by anoxic corrosion of uranium metal has exceedingly fine particle size (6 nm diameter), forms agglomerates, and has the formula UO2.004 ± 0.007; i.e., is practically stoichiometric UO2. The metaschoepite particles are flatter and wider when prepared at 21 C than the particles prepared at 50 C. These particles are much smaller than the metaschoepite observed in prolonged exposure of actual K Basin sludge to warm moist oxidizing conditions. The uraninite produced by anoxic uranium metal corrosion and the metaschoepite produced by reaction of uraninite aqueous slurries with oxygen may be used in engineering and process development testing. A rapid alternative method to determine uranium metal concentrations in sludge also was identified.

  6. Operating experience in processing of differently sourced deeply depleted uranium oxide and production of deeply depleted uranium metal ingots

    International Nuclear Information System (INIS)

    Manna, S.; Ladola, Y.S.; Sharma, S.; Chowdhury, S.; Satpati, S.K.; Roy, S.B.

    2009-01-01

    Uranium Metal Plant (UMP) of BARC had first time experience on production of three Depleted Uranium Metal (DUM) ingots of 76kg, 152kg and 163kg during March 1991. These ingots were produced by processing depleted uranyl nitrate solution produced at Plutonium Plant (PP), Trombay. In recent past Uranium Metal Plant (UMP), Uranium Extraction Division (UED), has been assigned to produce tonnage quantity of Deeply DUM (DDUM) from its oxide obtained from PP, PREFRE and RMP, BARC. This is required for shielding the high radioactive source of BHABHATRON Tele-cobalt machine, which is used for cancer therapy. The experience obtained in processing of various DDU oxides is being utilized for design of large scale DDU-metal plant under XIth plan project. The physico- chemical characteristics like morphology, density, flowability, reactivity, particle size distribution, which are having direct effect on reactivity of the powders of the DDU oxide powder, were studied and the shop-floor operational experience in processing of different oxide powder were obtained and recorded. During campaign trials utmost care was taken to standardized all operating conditions using the same equipment which are in use for natural uranium materials processing including safety aspects both with respect to radiological safety and industrial safety. Necessary attention and close monitoring were specially arranged and maintained for the safety aspects during the trial period. In-house developed pneumatic transport system was used for powder transfer and suitable dust arresting system was used for reduction of powder carry over

  7. Practice of the counter-current trickle leaching of uranium ore by refreshed liquor of bacterial oxidation

    International Nuclear Information System (INIS)

    Chen Shian; Huang Xiangfu; Fan Baotuan

    1995-01-01

    The uranium ore of the Mine No. 753 is a high-silicate type primary one, in which the tetravalent uranium accounts for 85%, and the uranium grade is in the range of 0.36% to 0.442%. To reduce the engineering investment and the operating cost a four-stage counter-current trickle leaching pilot-plant test was carried out with the leaching time 50 days and acid consumption 38 kg per ton of ore, and the recovery of more than 95% was obtained. Using the counter-current trickle leaching mode and controlling the limit concentration of the harmful matters in the bacterial leaching liquor, the latter can be effectively oxidized by the synchronical regeneration. A trickle leaching comparative test of 25 ton ore single heap also gave a good result of more than 95% in extraction rate, and 30% acid consumption was saved and the 2.0% pyrolusite (containing MnO 2 40%) was eliminated. This process is feasible in technology and worth-while in economy for treating the uranium ore of Mine No. 753, and provides a new method of uranium ore trickle leaching

  8. Thermogravimetric study of the reaction of uranium oxides with fluorine

    International Nuclear Information System (INIS)

    Komura, Motohiro; Sato, Nobuaki; Kirishima, Akira; Tochiyama, Osamu

    2008-01-01

    Thermogravimetric study of the reaction of uranium oxides with fluorine was conducted by TG-DTA method using anti-corrosion type differential thermobalance. When UO 2 was heated from R.T. to 500 deg. C in 5% F 2 /He atmosphere, the weight increase appeared at ca. 250 deg. C with an exothermic peak. Then the weight decreased slightly with a small exothermic peak followed by the complete volatilization with a large exothermic peak at ca. 350 deg. C. At a flow rate of 15, 30, 60 ml min -1 , there seemed to be no significant change for the fluorination of UO 2 . With the different heating rates of 1, 2, 5 and 10 deg. C min -1 , the fluorination peak shifted to higher temperature with increasing heating rates. For the comparison with thermogravimetric results, phase analysis by XRD method was conducted for the products obtained at different temperatures. At 260 deg. C, the product was UO 2 with a small amount of the intermediate compound, UO 2 F. The amount of this compound increased with increasing temperature up to 320 deg. C. Then another phase of UO 2 F 2 appeared at 340 deg. C but it was immediately fluorinated to the volatile fluoride. When U 3 O 8 was used as a starting material, it was found that the steep weight decrease peak appeared at ca. 350 deg. C and the uranium volatilized completely. This result suggests that fluorination of U 3 O 8 occurs at this temperature forming UF 6 . Uranium trioxide showed the similar fluorination behavior to that of U 3 O 8

  9. The Dissolution of Uranium Oxides in HB-Line Phase 1 Dissolvers

    International Nuclear Information System (INIS)

    Gray, J.H.

    2003-01-01

    A series of characterization and dissolution studies has been performed to define flowsheet conditions for the dissolution of uranium oxide materials in dissolvers. The samples selected for analysis were uranium oxide materials. The selection of these uranium oxide materials for characterization and dissolution studies was based on high enriched uranium content and trace levels of plutonium. Test results from the characterization study identified ferric oxide (Fe2O3) and iron/chromium/nickel (Fe/Cr/Ni) particles as impurities along with the tri-uranium oxide (U3O8) and uranium trioxide (UO3). The weight percent uranium in this material was found to vary depending on the impurity content. The trace impurity plutonium appears to be associated with the Fe/Cr/Ni particles. A small amount of absorbed moisture and waters of hydration is present. Most of the uranium oxides easily dissolved in low-molar nitric acid solutions without fluoride within one to two hours at solution temperature s between 60-80 degrees C. A small amount of residue remained following this dissolution step. To assure complete dissolution of uranium from these oxide materials, an additional dissolution step at 90 degrees C to boiling for at least one to two hours has been suggested. Only trace amounts of iron associated with Fe2O3 and Fe/Cr/Ni particles will dissolve during the dissolution steps. Neither hydrogen nor heat will be generated during the dissolution of these uranium oxide materials in nitric acid solutions. Some brown nitrogen dioxide (NO2) fumes will be generated during the dissolution of U3O8

  10. Fire hazards analysis for the uranium oxide (UO3) facility

    International Nuclear Information System (INIS)

    Wyatt, D.M.

    1994-01-01

    The Fire Hazards Analysis (FHA) documents the deactivation end-point status of the UO 3 complex fire hazards, fire protection and life safety systems. This FHA has been prepared for the Uranium Oxide Facility by Westinghouse Hanford Company in accordance with the criteria established in DOE 5480.7A, Fire Protection and RLID 5480.7, Fire Protection. The purpose of the Fire Hazards Analysis is to comprehensively and quantitatively assess the risk from a fire within individual fire areas in a Department of Energy facility so as to ascertain whether the objectives stated in DOE Order 5480.7, paragraph 4 are met. Particular attention has been paid to RLID 5480.7, Section 8.3, which specifies the criteria for deactivating fire protection in decommission and demolition facilities

  11. Dosage of boron traces in graphite, uranium and beryllium oxide

    International Nuclear Information System (INIS)

    Coursier, J.; Hure, J.; Platzer, R.

    1955-01-01

    The problem of the dosage of the boron in the materials serving to the construction of nuclear reactors arises of the following way: to determine to about 0,1 ppm close to the quantities of boron of the order of tenth ppm. We have chosen the colorimetric analysis with curcumin as method of dosage. To reach the indicated contents, it is necessary to do a previous separation of the boron and the materials of basis, either by extraction of tetraphenylarsonium fluoborate in the case of the boron dosage in uranium and the beryllium oxide, either by the use of a cations exchanger resin of in the case of graphite. (M.B.) [fr

  12. Caramel, uranium oxide fuel plates for water cooled reactors

    International Nuclear Information System (INIS)

    Bussy, Pierre; Delafosse, Jacques; Lestiboudois, Guy; Cerles, J.-M.; Schwartz, J.-P.

    1979-01-01

    The fuel is composed of thin plates assembled parallel to each other to form bundles or assemblies. Each plate is composed of a pavement of uranium oxide pellets, insulated from each other by a zircaloy cladding. The 235 U enrichment does not exceed 8%. The range of uses for this fuel extends from electric power generating reactors to irradiation reactors for research work. A parametric study in test loops has made it possible to determine the operating limits of this thick fuel, without bursting. The resulting diagram gives the permissible power densities, with and without cycling for specific burn-ups beyond 50,000 MWd/t. The thinnest plates were also irradiated in total in the form of advance assemblies irradiated in the core of the OSIRIS pile prior to its transformation. This transformation and the operation of this reactor with a core of 'Caramel' elements is the main trial experiment of this fuel [fr

  13. Oxidation kinetics of hydride-bearing uranium metal corrosion products

    Science.gov (United States)

    Totemeier, Terry C.; Pahl, Robert G.; Frank, Steven M.

    The oxidation behavior of hydride-bearing uranium metal corrosion products from Zero Power Physics Reactor (ZPPR) fuel plates was studied using thermo-gravimetric analysis (TGA) in environments of Ar-4%O 2, Ar-9%O 2, and Ar-20%O 2. Ignition of corrosion product samples from two moderately corroded plates was observed between 125°C and 150°C in all environments. The rate of oxidation above the ignition temperature was found to be dependent only on the net flow rate of oxygen in the reacting gas. Due to the higher net oxygen flow rate, burning rates increased with increasing oxygen concentration. Oxidation rates below the ignition temperature were much slower and decreased with increasing test time. The hydride contents of the TGA samples from the two moderately corroded plates, determined from the total weight gain achieved during burning, were 47-61 wt% and 29-39 wt%. Samples from a lightly corroded plate were not reactive; X-ray diffraction (XRD) confirmed that they contained little hydride.

  14. Oxidation kinetics of hydride-bearing uranium metal corrosion products

    International Nuclear Information System (INIS)

    Totemeier, T.C.; Pahl, R.G.; Frank, S.M.

    1998-01-01

    The oxidation behavior of hydride-bearing uranium metal corrosion products from zero power physics reactor (ZPPR) fuel plates was studied using thermo-gravimetric analysis (TGA) in environments of Ar-4%O 2 , Ar-9%O 2 , and Ar-20%O 2 . Ignition of corrosion product samples from two moderately corroded plates was observed between 125 C and 150 C in all environments. The rate of oxidation above the ignition temperature was found to be dependent only on the net flow rate of oxygen in the reacting gas. Due to the higher net oxygen flow rate, burning rates increased with increasing oxygen concentration. Oxidation rates below the ignition temperature were much slower and decreased with increasing test time. The hydride contents of the TGA samples from the two moderately corroded plates, determined from the total weight gain achieved during burning, were 47-61 wt% and 29-39 wt%. Samples from a lightly corroded plate were not reactive; X-ray diffraction (XRD) confirmed that they contained little hydride. (orig.)

  15. Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-09-20

    The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project.

  16. Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

    International Nuclear Information System (INIS)

    PAJUNEN, A.L.

    2000-01-01

    The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project

  17. Possibilities for recycling of weapon-grade uranium and plutonium and its peaceful use as reactor fuel

    International Nuclear Information System (INIS)

    Floeter, W.

    2000-01-01

    At present 90% of the energy production is based on fossil fuels. Since March 1999, however, the peaceful use of weapon-grade uranium as reactor fuel is being discussed politically. Partners of this discussion is a group of some private western companies on one side and a state-owned company of the Russian Federation (GUS) on the other. Main topic of the deal besides the winning of electrical energy is the useful disposal of the surplus on weapon-grade material of both leading nations. According to the deal, about 160,000 t of Russian uranium, expressed as natural uranium U 3 O 8 , would be processed during the next 15 years. Proven processes would be applied. Those methods are being already used in Russian facilities at low capacity rates. There are shortages in the production of low enriched uranium (LEU), because of the low capacity rates in the old facilities. The capacity should be increased by a factor of ten, but there is not enough money available in Russia for financing the remodeling of the plants. Financing should therefore probably be provided by the western clients of this deal. The limited amount of uranium produced could be furnised to the uranium market without major difficulties for the present suppliers of natural uranium. The discussions regarding the security of the details of the deal - however - are not yet finalized. (orig.) [de

  18. Dissolution of uranium oxide TBP-HNO3 complex

    International Nuclear Information System (INIS)

    Mizuno, Mineo; Kosaka, Yuji; Mori, Yukihide; Shimada, Takashi

    2002-12-01

    As a head end process for the pulverization of the spent fuel, the mechanical method (the shredder method) and the pyro-chemical method (oxidisation heat-treatment) have been examined. UO 2 is a main ingredient of Uranium oxide powder by the mechanical method, and U 3 O 8 is that by the pyro-chemical method. Moreover, the particle size of the pulverized powder depend on the conditions of the pulverizing process. As it was considered that the difference of dissolution rates of samples was caused by the difference of sample chemical forms and dissolution temperature, parametric surveys on chemical form and particle size of powder and dissolution temperature were carried out, and the following results were obtained. 1) The remarkable difference of dissolution rate between U 3 O 8 powder (average particle size 3.7 μm) and UO 2 powder (average particle size 2.4 μm) which have comparatively similar particle size was not observed. 2) It was confirmed that the dissolution rate became lower according to the particle size increase (average particle size 2.4 μm-1 mm). And it was considered that dissolution rate had strong dependency on particle size, according to the results that the powder with 1 mm particle size did not dissolute completely after 5 hours test. 3) The temperature dependency of the dissolution rate was confirmed by dissolution test with UO 2 powder (average particle size 2.4 μm-1 mm). The higher dissolution rate was obtained in the higher dissolution temperature, and 11 kcal/mol was obtained as activation energy of dissolution. 4) In the dissolution test of UO 2 powder, the nitric acid concentration started to change earlier than that of U 3 O 8 powder and concentration change range became larger compared with that in the dissolution test of U 3 O 8 powder. It was considered that those differences were caused by difference in mole ratio of Uranium and nitric acid which are consumed in the dissolution reaction (3:7 for U 3 O 8 , 3:8 for UO 2 ). 5) In case

  19. Microcontroller based, ore grade measuring portable instruments for uranium mining industry

    International Nuclear Information System (INIS)

    Dheeraj Reddy, J.; Narender Reddy, J.

    2004-01-01

    Ore Face Scanning and Bore Hole Logging are important essential activities which are required to be carried out in any Uranium mining industry. Microcontroller based, portable instruments with built-in powerful embedded code for data acquisition (of Radiation counts) and Ore Grade calculations will become a handy measuring tool for miners. Nucleonix Systems has recently developed and made these two portable instruments available to UCIL, which are under use at Jaduguda and Narvapahar mines. Some of the important features of these systems are compact, light weight, portable, hand held, battery powered. Modes of Data Acquisition: CPS, CPM and ORE GRADE. Detector: Sensitive GM Tube. Choice of Adj. TC (Time Constant) in 'ORE GRADE', acquisition mode. Built-in automatic BG (Background) recording and subtraction provided to indicate net CPS, CPM or ore GRADE in PPM. Can store 1000 readings at users choice. Built-in RS232 serial port facilitates data downloading into PC. This paper focuses on design concepts and technical details for the above two products. (author)

  20. Study on the surface oxidation resistance of uranium metal in the atmosphere of carbon monoxide

    International Nuclear Information System (INIS)

    Wang Xiaolin; Fu Yibei; Xie Renshou

    1999-01-01

    The surface reactions of different layers on uranium metal with carbon monoxide at 25, 80 and 200 degree C are studied by X-ray photoelectron spectroscopy (XPS). The experimental results show that the carbon monoxide is adsorbed on the surface oxide layer of uranium and interacted each other. The content of oxygen in the surface oxide and O/U ratio are decreased with increasing the exposure of carbon monoxide to the surface layer. The effect of reduction on the metal surface is more obviously with a higher temperature and increasing of layer thickness. The investigation indicates the uranium metal has resistance to further oxidation in the atmosphere of carbon monoxide

  1. Influence of carbon monoxide to the surface layer of uranium metal and its oxides

    International Nuclear Information System (INIS)

    Wang Xiaoling; Fu Yibei; Xie Renshou; Huang Ruiliang

    1996-09-01

    The surface structures of uranium metal and triuranium octaoxide (U 3 O 8 ) and the influence of carbon monoxide to the surface layers have been studied by X-ray photoelectron spectroscopy (XPS). After exposure to carbon monoxide, contents of oxygen in the surface oxides of uranium metal and U 3 O 8 are decreased and O/U ratios decrease 7.2%, 8.0% respectively. The investigation indicated the surface layers of uranium metal and its oxides were forbidden to further oxidation in the atmosphere of carbon monoxide. (11 refs., 9 figs., 2 tabs.)

  2. Uranium

    International Nuclear Information System (INIS)

    Cuney, M.; Pagel, M.; Leroy, J.

    1992-01-01

    First, this book presents the physico-chemical properties of Uranium and the consequences which can be deduced from the study of numerous geological process. The authors describe natural distribution of Uranium at different scales and on different supports, and main Uranium minerals. A great place in the book is assigned to description and classification of uranium deposits. The book gives also notions on prospection and exploitation of uranium deposits. Historical aspects of Uranium economical development (Uranium resources, production, supply and demand, operating costs) are given in the last chapter. 7 refs., 17 figs

  3. Status report from India [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    Fareeduddin, S [Atomic Energy Establishment, Trombay, Bombay (India)

    1967-06-15

    The Energy Survey Committee of India, in its report to the Government, has estimated that the energy requirements in the year 1985/86 would be 290X10{sup 9} kWh, i. e. eight times the present requirement, and in the year 2000 it would be 820X10{sup 9} kWh, which is about 22 times the present requirement. The hydropotential that can be developed during the next 20 years is estimated to be of the order of 150X10{sup 9} kWh and hence the difference of about 140X10{sup 9} kWh will have to be obtained from either fossil or nuclear fuel. This would mean installating a generation capacity of about 26 000 MW in the next 20 years. To conserve the limited fossil fuel reserves, it has been estimated that about 70% of this capacity, i. e. about 18 000 MW, should form the nuclear component. This will be about 25% of the total energy requirements by 1985/86. The uranium requirements to meet this growth will be about 10 000 tonnes by 1985/86 which, from the point of view of our resources, is a substantial quantity. The most important uranium deposits are located in South Bihar in the Singhbhum Thrust belt, which is well known for its copper, apatite magnetite and kyanite deposits. On the basis of their uranium contents, these ores can be classified into two broad categories - one with low copper and high uranium contents and the other with high copper and low uranium contents. Another source of uranium in India is monazite. Some particulars about these deposits are given. Facilities for the recovery of byproduct uranium from monazite already exist in the country. But its production from this source, conditioned as it is by the limited demand for thorium, cannot be very large. Both the categories of the ores from the Singhbhum belt can be considered as low grade. Uranium from the ores in category (B) can be recovered, in the present state of knowledge, only as a byproduct of the copper industry. In the case of ores in the category (A), attempts have been made to recover uranium

  4. The spectrographic analysis of plutonium oxide or mixed plutonium oxide/uranium oxide fuel pellets by the dried residue technique

    International Nuclear Information System (INIS)

    Jarbo, G.J.; Faught, P.; Hildebrandt, B.

    1980-05-01

    An emission spectrographic method for the quantitative determination of metallic impurities in plutonium oxide and mixed plutonium oxide/uranium oxide is described. The fuel is dissolved in nitric acid and the plutonium and/or uranium extracted with tributyl phosphate. A small aliquot of the aqueous residue is dried on a 'mini' pyrolitic graphite plate and excited by high voltage AC spark in an oxygen atmosphere. Spectra are recorded in a region which has been specially selected to record simultaneously lines of boron and cadmium in the 2nd order and all the other elements of interest in the 1st order. Indium is used as an internal standard. The excitation of very small quantities of the uraniumm/plutonium free residue by high voltage spark, together with three separate levels of containment reduce the hazards to personnel and the environment to a minimum with limited effect on sensitivity and accuracy of the results. (auth)

  5. The development and application of quantitative methods for the determination of in-situ radiometric uranium grade on the Witwatersrand gold and uranium mines

    International Nuclear Information System (INIS)

    Symons, G.

    1985-12-01

    A detailed investigation of background radiation levels near the reef zone in the uranium section of the Western Areas Mine was conducted using a collimated radiometric face scanner. This study demonstrated that these radiation levels can be high; 25% or more of the counts measured when sampling a reef face may originate from a background source, especially from uranium ore rubble on the footwall close to the reef face. A method using a 20mm frontal shield was devised to obtain an accurate background correction. Three calibration schemes, the Area method, the Gamlog method, and the Deconvolution method were implemented for the production of accurate in-situ radiometric uranium grades. This involved the construction of a step-response calibration pad at Pelindaba together with the establisment of appropriate software and underground radiometric sampling procedures. Radiometric grades generated by these calibration procedures from 60 channel sections were on average 10% below those procured from conventional chip sampling. A correlation between gold and uranium grades was also evident. Crushed rock samples were collected to investigate the thorium problem and are still undergoing analysis at the time of writing. Refinements in the design of the collimated face scanner are also described

  6. PHASE ANALYSES OF URANIUM-BEARING MINERALS FROM THE HIGH GRADE ORE, NOPAL I, PENA BLANCA, MEXICO

    Energy Technology Data Exchange (ETDEWEB)

    M. Ren; P. Goodell; A. Kelts; E.Y. Anthony; M. Fayek; C. Fan; C. Beshears

    2005-07-11

    The Nopal I uranium deposit is located in the Pena Blanca district, approximately 40 miles north of Chihuahua City, Mexico. The deposit was formed by hydrothermal processes within the fracture zone of welded silicic volcanic tuff. The ages of volcanic formations are between 35 to 44 m.y. and there was secondary silicification of most of the formations. After the formation of at least part of the uranium deposit, the ore body was uplifted above the water table and is presently exposed at the surface. Detailed petrographic characterization, electron microprobe backscatter electron (BSE) imagery, and selected x-ray maps for the samples from Nopal I high-grade ore document different uranium phases in the ore. There are at least two stages of uranium precipitation. A small amount of uraninite is encapsulated in silica. Hexavalent uranium may also have been a primary precipitant. The uranium phases were precipitated along cleavages of feldspars, and along fractures in the tuff. Energy dispersive spectrometer data and x-ray maps suggest that the major uranium phases are uranophane and weeksite. Substitutions of Ca and K occur in both phases, implying that conditions were variable during the mineralization/alteration process, and that compositions of the original minerals have a major influence on later stage alteration. Continued study is needed to fully characterize uranium behavior in these semi-arid to arid conditions.

  7. Study on uranium metallization yield of spent Pressurized Water Reactor fuels and oxidation behavior of fission products in uranium metals

    International Nuclear Information System (INIS)

    Choi, Ke Chon; Lee, Chang Heon; Kim, Won Ho

    2003-01-01

    Metallization yield of uranium oxide to uranium metal from lithium reduction process of spent Pressurized Water Reactor (PWR) fuels was measured using thermogravimetric analyzer. A reduced metal produced in the process was divided into a solid and a powder part, and each metallization yield was measured. Metallization yield of the solid part was 90.7∼95.9 wt%, and the powder being 77.8∼71.5 wt% individually. Oxidation behaviour of the quarternary alloy was investigated to take data on the thermal oxidation stability necessary for the study on dry storage of the reduced metal. At 600∼700 .deg. C, weight increments of allow of No, Ru, Rh and Pd was 0.40∼0.55 wt%. Phase change on the surface of the allow was started at 750 .deg. C. In particular, Mo was rapidly oxidized and then the alloy lost 0.76∼25.22 wt% in weight

  8. Oxidation state analyses of uranium with emphasis on chemical speciation in geological media

    International Nuclear Information System (INIS)

    Ervanne, H.

    2004-01-01

    This thesis focuses on chemical methods suitable for the determination of uranium redox species in geological materials. Nd-coprecipitation method was studied for the determination of uranium oxidation states in ground waters. This method is ideally suited for the separation of uranium oxidation states in the field, which means that problems associated with the instability of U(IV) during transport are avoided. An alternative method, such as ion exchange, is recommended for the analysis of high saline and calcium- and iron-rich ground waters. U(IV)/Utot was 2.8-7.2% in ground waters in oxidizing conditions and 60-93% in anoxic conditions. From thermodynamic model calculations applied to results from anoxic ground waters it was concluded that uranium can act as redox buffer in granitic ground waters. An ion exchange method was developed for the analysis of uranium oxidation states in different solid materials of geological origin. These included uranium minerals, uraniumbearing minerals, fracture coatings and bulk rock. U(IV)/Utot was 50-70% in uraninites, 5.8-8.7% in secondary uranium minerals, 15-49% in different fracture coatings and 64- 77% in samples from deep bedrock. In the uranium-bearing minerals, U(IV)/Utot was 33-43% (allanites), 5.9% (fergusonite) and 93% (monazite). Although the ion exchange method gave reliable results, there is a risk for the conversion of uranium oxidation states during analysis of heterogeneous samples due to the redox reactions that take place in the presence of some iron compounds. This risk was investigated in a study of several common iron-bearing minerals. The risk for conversion of uranium oxidation states can be screened by sample selection and minimized with use of a redox buffer compound such as polyacrylic acid (PAA). In studies of several carboxylic acids, PAA was found to be the most suitable for extending the application of the method. The stability of uranium oxidation states during analysis and the selectivity

  9. Metallographic preparation of sintered oxides, carbides and nitrides of uranium and plutonium

    International Nuclear Information System (INIS)

    Martin, A.; Arles, L.

    1967-12-01

    We describe the methods of polishing, attack and coloring used at the section of plutonium base ceramics studies. These methods have stood the test of experience on the uranium and plutonium carbides, nitrides and carbonitrides as well on the mixed uranium and plutonium oxides. These methods have been particularly adapted to fit to the low dense and sintered samples [fr

  10. Study on the surface oxidation of uranium in different gaseous atmospheres

    International Nuclear Information System (INIS)

    Wang Xiaoling; Fu Yibei; Xie Renshou

    1996-03-01

    The studying for the surface oxidation of uranium and oxide by X-ray photoelectron spectroscopy (XPS), Auger electron spectroscopy (AES), secondary ion mass spectroscopy (SIMS), and the surface oxidation of uranium in different gaseous atmospheres such as O 2 , H 2 , CO, CO 2 , H 2 O(v) and air were reviewed. The surface oxidation of uranium is greatly influenced by a number of parameters including atmospheric temperature, pressure, diffusion of adsorbed gas atoms through the oxide layer, surface and interface chemical component, and defect structure and electron nature of the oxide layer. The initial oxidation mechanism and kinetics have been discussed. Suggestions for future work have also been presented. (32 refs., 7 figs., 5 tabs.)

  11. Organic matter in uranium concentration during ancient bed oxidation of carboniferons sediments

    International Nuclear Information System (INIS)

    Kruglova, V.G.; Uspenskij, V.A.; Dement'ev, P.K.; Kochenov, A.V.

    1984-01-01

    Changes in the organic matter accompanying the process of epigenetic ore formation are studied using the example of a deposit localized in carboniferous molasse strata of the Cretaceous period. Peculiarities of the organic matter as the main mineralization agent are studied by a complex of physical and themical methods. A distinct relationship between the uranium concentration and the degree of organic matter oxigenation is a most characteristic feature of the ore localization, however, there is no direct correlation between the contents of uranium and organic matter in ores. Uranium minerallzation was accumulated during infiltration of acid uraniferous.waters into grey stratum in the process of the bed oxidation zone formation oxidizing. Brown coal matter possessing a maximum adsorbability, as compared to other sedimentary rocks, apprared to be the uranium precipitator. The adsorption was accompanie by the formation of proper uranium minerals (coffinite, pitchblende) due to uranium reduction by oxidizing organic matter. Thus, the oxidative epigenesis was an are-forming process with the uranium concentration on organic matter proportionally to oxidation of the latter

  12. Criticality of moderated and undermoderated low-enriched uranium oxide systems

    International Nuclear Information System (INIS)

    Goebel, G.R.

    1980-06-01

    Uranium oxide was enriched to 4.46 wt % 235 U compacted to a density of 4.68 g/cm 3 . The uranium oxide was packed into cubical aluminum cans and water added to the oxide until an H/U atomic ratio of 0.77 was achieved. A 5 x 5 x 5 array of uranium oxide cans for the experiments were used when no plastic moderator material was placed between cans. High enriched uranium drivers were used to achieve criticality. Criticality was achieved for smaller arrays without a driver when 24.5 mm plastic moderator material was placed between the cans. Twelve critical experiments are reported, six in each reflector

  13. Chlorination of antimony and its volatilization treatment of waste antimony-uranium composite oxide catalyst

    International Nuclear Information System (INIS)

    Sawada, K.; Enokida, Y.

    2011-01-01

    For the waste antimony-uranium composite oxide catalyst, the chlorination of antimony and its volatilization treatment were proposed, and evaluated using hydrogen chloride gas at 873-1173 K. During the treatment, the weight loss of the composite oxide sample, which resulted from the volatilization of antimony, was confirmed. An X-ray diffraction analysis showed that uranium oxide, U 3 O 8 , was formed during the reaction. After the treatment at 1173 K for 1 h, almost all the uranium contained in the waste catalyst was dissolved by a 3 M nitric acid solution at 353 K within 10 min, although that of the non-treated catalyst was less than 0.1%. It was found that the chlorination and volatilization treatment was effective to separate antimony from the composite oxide catalyst and change uranium into its removable form. (orig.)

  14. Standard specification for nuclear-grade beryllium oxide powder

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    This specification defines the physical and chemical requirements of nuclear-grade beryllium oxide (BeO) powder to be used in fabricating nuclear components. This specification does not include requirements for health and safety. It recognizes the material as a Class B poison and suggests that producers and users become thoroughly familiar with and comply to applicable federal, state and local regulations and handling guidelines. Special tests and procedures are given

  15. Characterizing uranium oxide reference particles for isotopic abundances and uranium mass by single particle isotope dilution mass spectrometry

    International Nuclear Information System (INIS)

    Kraiem, M.; Richter, S.; Erdmann, N.; Kühn, H.; Hedberg, M.; Aregbe, Y.

    2012-01-01

    Highlights: ► A method to quantify the U mass in single micron particles by ID-TIMS was developed. ► Well-characterized monodisperse U-oxide particles produced by an aerosol generator were used. ► A linear correlation between the mass of U and the volume of particle(s) was found. ► The method developed is suitable for determining the amount of U in a particulate reference material. - Abstract: Uranium and plutonium particulate test materials are becoming increasingly important as the reliability of measurement results has to be demonstrated to regulatory bodies responsible for maintaining effective nuclear safeguards. In order to address this issue, the Institute for Reference Materials and Measurements (IRMM) in collaboration with the Institute for Transuranium Elements (ITU) has initiated a study to investigate the feasibility of preparing and characterizing a uranium particle reference material for nuclear safeguards, which is finally certified for isotopic abundances and for the uranium mass per particle. Such control particles are specifically required to evaluate responses of instruments based on mass spectrometric detection (e.g. SIMS, TIMS, LA-ICPMS) and to help ensuring the reliability and comparability of measurement results worldwide. In this paper, a methodology is described which allows quantifying the uranium mass in single micron particles by isotope dilution thermal ionization mass spectrometry (ID-TIMS). This methodology is characterized by substantial improvements recently achieved at IRMM in terms of sensitivity and measurement accuracy in the field of uranium particle analysis by TIMS. The use of monodisperse uranium oxide particles prepared using an aerosol generation technique developed at ITU, which is capable of producing particles of well-characterized size and isotopic composition was exploited. The evidence of a straightforward correlation between the particle volume and the mass of uranium was demonstrated in this study

  16. Evidence for Single Metal Two Electron Oxidative Addition and Reductive Elimination at Uranium

    OpenAIRE

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

    2017-01-01

    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 compl...

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    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.

  19. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    The article briefly discusses the Australian government policy and the attitude of political party factions towards the mining and exporting of the uranium resources in Australia. Australia has a third of the Western World's low-cost uranium resources

  20. Identifying high-grade uranium deposits in the Proterozoic basins of India- a challenge to exploration

    International Nuclear Information System (INIS)

    Mahadevan, T.M.

    1995-01-01

    The favorability factors that bestow on the Proterozoic formation of India, a potential to host high grade uranium deposits are discussed in the light of the known features of the new class of unconformity- related and strata bound uranium deposits. The need to reorient several past approaches is emphasised and it is suggested that future programmes must avail of the constraining benefits of a spectrum of geophysical, geochemical, and sedimentological studies in the choice of target areas for detailed exploration and development. A synthesis of geological and geochemical data with such geophysical features as magnetic and gravity anomalies, velocity structure, seismic reflectivity, electrical conductivity, and radioactivity can effectively lead to relatively more favourable exploration targets. Such efforts may lead to the generation of more than one model of the deep basinal features, which then provide wider options for drilling and proving of ore bodies. The alternative to the above approach is saturation drilling, which is a costly and time-consuming process and, therefore, very often self-defeating. (author). 28 refs., 2 figs., 2 tabs

  1. Processing of Low-Grade Uranium Ores. Proceedings of a Panel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1967-06-15

    Proceedings of a panel convened by the IAEA in Vienna, 27 June - 1 July 1966. The 22 specialists from 15 countries and one international organization who attended the meeting were asked to give an appraisal of the current situation with regard to the processing of low-grade uranium ores and make recommendations for a possible IAEA programme of activities. This publication covers the work of the panel. Contents: Status reports (13 reports); Technical reports (13 reports); Summaries of discussions; Recommendations of the panel. Each report is in its original language (16 English, 4 French, 2 Russian and 4 Spanish) and each technical report is preceded by an abstract in English and one in the original language if this is not English. The summaries of discussions and the panel recommendations are in English. (author)

  2. Processing of Low-Grade Uranium Ores. Proceedings of a Panel

    International Nuclear Information System (INIS)

    1967-01-01

    The 22 specialists from 15 countries and one international organization who attended the meeting were asked to give an appraisal of the current situation with regard to the processing of low-grade uranium ores and make recommendations for a possible IAEA programme of activities. This publication covers the work of the panel. Contents: Status reports (13 reports); Technical reports (13 reports); Summaries of discussions; Recommendations of the panel. Each report is in its original language (16 English, 4 French, 2 Russian and 4 Spanish) and each technical report is preceded by an abstract in English and one in the original language if this is not English. The summaries of discussions and the panel recommendations are in English. (author)

  3. Processing of Low-Grade Uranium Ores. Proceedings of a Panel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1967-06-15

    The 22 specialists from 15 countries and one international organization who attended the meeting were asked to give an appraisal of the current situation with regard to the processing of low-grade uranium ores and make recommendations for a possible IAEA programme of activities. This publication covers the work of the panel. Contents: Status reports (13 reports); Technical reports (13 reports); Summaries of discussions; Recommendations of the panel. Each report is in its original language (16 English, 4 French, 2 Russian and 4 Spanish) and each technical report is preceded by an abstract in English and one in the original language if this is not English. The summaries of discussions and the panel recommendations are in English. (author)

  4. Determination of trace metals in nuclear-grade uranium dioxide by X-ray fluorescence spectrometry

    International Nuclear Information System (INIS)

    Salvador, V.L.R.; Imakuma, K.

    1988-04-01

    A method is described for the simultaneous determination of low concentrations of Ca, Cr, Cu, Fe, Mn and Ni in nuclear-grade uranium dioxide by X-ray fluorescence spectrometry, without the use of chemical treatment. The lower limits of detection range from 2 μg g -1 for nickel and manganese to 5 μg g -1 for copper. Samples are prepared in the form of double-layer pellets with boric acid as a binding agent. Standards are prepared in a U 3 O 8 matrix, which is more chemically stable than UO 2 and has similar matrix behaviour. The correlation coefficients for calibration curves are better than 0.999. Erros range from 2.4 % for chromium to 6.8 % for nickel. (author) [pt

  5. On a Bayesian estimation procedure for determining the average ore grade of a uranium deposit

    International Nuclear Information System (INIS)

    Heising, C.D.; Zamora-Reyes, J.A.

    1996-01-01

    A Bayesian procedure is applied to estimate the average ore grade of a specific uranium deposit (the Morrison formation in New Mexico). Experimental data taken from drilling tests for this formation constitute deposit specific information, E 2 . This information is combined, through a single stage application of Bayes' theorem, with the more extensive and well established information on all similar formations in the region, E 1 . It is assumed that the best estimate for the deposit specific case should include the relevant experimental evidence collected from other like formations giving incomplete information on the specific deposit. This follows traditional methods for resource estimation, which presume that previous collective experience obtained from similar formations in the geological region can be used to infer the geologic characteristics of a less well characterized formation. (Author)

  6. Uranium

    International Nuclear Information System (INIS)

    Poty, B.; Cuney, M.; Bruneton, P.; Virlogeux, D.; Capus, G.

    2010-01-01

    With the worldwide revival of nuclear energy comes the question of uranium reserves. For more than 20 years, nuclear energy has been neglected and uranium prospecting has been practically abandoned. Therefore, present day production covers only 70% of needs and stocks are decreasing. Production is to double by 2030 which represents a huge industrial challenge. The FBR-type reactors technology, which allows to consume the whole uranium content of the fuel, is developing in several countries and will ensure the long-term development of nuclear fission. However, the implementation of these reactors (the generation 4) will be progressive during the second half of the 21. century. For this reason an active search for uranium ores will be necessary during the whole 21. century to ensure the fueling of light water reactors which are huge uranium consumers. This dossier covers all the aspects of natural uranium production: mineralogy, geochemistry, types of deposits, world distribution of deposits with a particular attention given to French deposits, the exploitation of which is abandoned today. Finally, exploitation, ore processing and the economical aspects are presented. Contents: 1 - the uranium element and its minerals: from uranium discovery to its industrial utilization, the main uranium minerals (minerals with tetravalent uranium, minerals with hexavalent uranium); 2 - uranium in the Earth's crust and its geochemical properties: distribution (in sedimentary rocks, in magmatic rocks, in metamorphic rocks, in soils and vegetation), geochemistry (uranium solubility and valence in magmas, uranium speciation in aqueous solution, solubility of the main uranium minerals in aqueous solution, uranium mobilization and precipitation); 3 - geology of the main types of uranium deposits: economical criteria for a deposit, structural diversity of deposits, classification, world distribution of deposits, distribution of deposits with time, superficial deposits, uranium

  7. Uranium

    International Nuclear Information System (INIS)

    Mackay, G.A.

    1978-01-01

    The author discusses the contribution made by various energy sources in the production of electricity. Estimates are made of the future nuclear contribution, the future demand for uranium and future sales of Australian uranium. Nuclear power growth in the United States, Japan and Western Europe is discussed. The present status of the six major Australian uranium deposits (Ranger, Jabiluka, Nabarlek, Koongarra, Yeelerrie and Beverley) is given. Australian legislation relevant to the uranium mining industry is also outlined

  8. Uranium

    International Nuclear Information System (INIS)

    1982-01-01

    The development, prospecting, research, processing and marketing of South Africa's uranium industry and the national policies surrounding this industry form the headlines of this work. The geology of South Africa's uranium occurences and their positions, the processes used in the extraction of South Africa's uranium and the utilisation of uranium for power production as represented by the Koeberg nuclear power station near Cape Town are included in this publication

  9. The oxidation behaviour of uranium in air at 348-765 K

    International Nuclear Information System (INIS)

    Bennett, M.J.; Price, J.B.

    1981-01-01

    The oxidation behaviour of adjusted uranium has been examined in air, at atmospheric pressure, at 348-765 K. Particular emphasis has been directed to the role of swelling resulting from prior irradiation of the metal to a burn-up of 5600-9100 MWd/t and an addition of 2.5 x 10 4 vpm water vapour to the air. Pre-irradiation of uranium enhanced its attack by air at 348-523 K, the enhancement increasing progressively with percentage swelling. This effect resulted primarily from the break-up of the uranium surface during oxidation with the generation of a greater reaction surface area and was independent of the temperature of oxidation in dry air and also above 423 K in wet air. At lower temperatures, however, the water vapour addition increased the effective reaction rate, possibly by the transistory involvement of uranium hybride. The influence of the water vapour increased with swelling of the irradiated uranium and was greater than that exerted on the oxidation of unirradiated uranium at comparable temperatures. With increasing temperature above 623 K, swelling had a progressively decreasing influence upon the attack of irradiated uranium in both environments. (orig.)

  10. Uranium

    International Nuclear Information System (INIS)

    Stewart, E.D.J.

    1974-01-01

    A discussion is given of uranium as an energy source in The Australian economy. Figures and predictions are presented on the world supply-demand position and also figures are given on the added value that can be achieved by the processing of uranium. Conclusions are drawn about Australia's future policy with regard to uranium (R.L.)

  11. Uranium

    International Nuclear Information System (INIS)

    Toens, P.D.

    1981-03-01

    The geological setting of uranium resources in the world can be divided in two basic categories of resources and are defined as reasonably assured resources, estimated additional resources and speculative resources. Tables are given to illustrate these definitions. The increasing world production of uranium despite the cutback in the nuclear industry and the uranium requirements of the future concluded these lecture notes

  12. Study on the influence of carbon monoxide to the surface oxide layer of uranium metal

    International Nuclear Information System (INIS)

    Wang Xiaolin; Duan Rongliang; Fu Yibei; Xie Renshou; Zuo Changming; Zhao Chunpei; Chen Hong

    1997-01-01

    The influence of carbon monoxide to the surface oxide layer of uranium metal has been studied by X-ray photoelectron spectroscopy (XPS) and gas chromatography (GC). Carbon monoxide adsorption on the oxide layer resulted in U4f peak shifting to the lower binding energy. The content of oxygen in the oxide is decreased and the atomic ratio (O/U) is decreased by 7.2%. The amount of carbon dioxide in the atmosphere after the surface reaction is increased by 11.0%. The investigation indicates that the surface layer can prevent the further oxidation uranium metal in the atmosphere of carbon monoxide

  13. Laboratory studies on leaching of low grade uranium ores and treatment of low level liquid waste generated by leaching experiments

    International Nuclear Information System (INIS)

    Palabrica, O.T.; Antonino, E.J.; Caluag, L.A.; Villamater, D.

    1980-07-01

    Acid leaching experiments of preconcentrated uranium ore were carried out at a pulp density of 50% solids, using sulfuric acid with sodium chlorate as oxidant. The different leaching parameters considered in this work were temperature, oxidant level and leaching time. In the experimental procedure, the concentration of oxidant and the temperature were varied to determine how they affect the leaching process. Experimental results are illustrated in tabulated form for better interpretation. Uranium analyses were done by fluorimetric and delayed-neutron activation analysis. An anion exchange method using Dowex 1 x 8, 200-400 mesh (Cl - ) was used in treating the low-level liquid waste generated by leaching experiments. The purpose of this treatment was to minimize radioactive contamination in the waste materials and also to recover some of the uranium left in the liquid waste. (author)

  14. Preparation of uranium-based oxide catalysts; Preparation de catalyseurs oxydes a base d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Bressat, R. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1965-07-01

    We have studied the thermal decomposition of uranyl and uranium IV oxalates as a mean of producing uranium dioxide. We have isolated the main intermediate phases of the decompositions and have indexed the lines of their X-ray diffraction patterns. The oxides produced by the decomposition are ill-defined and unstable: they strongly absorb atmospheric oxygen with modification of the composition and, in certain cases, of the structure (pyrophoric oxide). With a view to obtaining stable oxides, we have prepared mixed uranium-thorium oxalates. In order to prepare an oxalate having a homogeneous composition, it is necessary to adopt a well-defined preparation method: the addition of solutions of thorium and uranium IV nitrates to a continually saturated oxalic acid solution. The mixed oxide obtained from the thermal decomposition of an oxalate U{sub x}Th{sub 1-x}(C{sub 2}O{sub 4}){sub 2}, 2 H{sub 2}O at 500 C for 24 hours in a current of oxygen leads to a cubic structure which is well-defined both in the bulk and superficially when x is less than 0.35. Above this atomic concentration of uranium, some uranium moves out of the lattice in the form of UO{sub 3} or U{sub 3}O{sub 8} according to the temperature. The mixed oxide is not stoichiometric,(U{sub x}Th{sub 1-x}O{sub 2+y}) and the average degree of oxidation of the uranium varies with the temperature and partial oxygen pressure. The oxides thus formed have a high surface area. By dissolving the mixed oxalates in a concentrated solution of ammonium oxalate, it is possible to deposit the catalyst on a support, but the differences in the solubilities of the thorium and uranium IV oxalates in the ammonium oxalate make it impossible to prepare double salts formed either of thorium and uranium and of ammonium. (author) [French] Nous avons etudie la decomposition thermique des oxalates d'uranyle et d'uranium IV en vue d'aboutir au dioxide d'uranium. Nous avons pu isoler les principales phases

  15. Melting temperature of uranium - plutonium mixed oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, Tetsuya; Hirosawa, Takashi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-08-01

    Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960`s and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960`s and that some of the 1960`s data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO{sub 2} - PuO{sub 2} - PuO{sub 1.61} ideal solution model, and then formulized. (J.P.N.)

  16. Melting temperature of uranium - plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    Ishii, Tetsuya; Hirosawa, Takashi

    1997-08-01

    Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960's and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960's and that some of the 1960's data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO 2 - PuO 2 - PuO 1.61 ideal solution model, and then formulized. (J.P.N.)

  17. An autoclave system for uranium oxide dissolution experiments

    International Nuclear Information System (INIS)

    Nykyri, Mikko

    1985-05-01

    According to the decision in principle of the Council of State of Finland the nuclear energy producers must provide preparedness for carrying out the final disposal of spent nuclear fuel in Finland. By the present principal concept the spent fuel will be disposed deep into the granitic bedrock. A parameter needed by risk analysis models is the dissolution rate of the uranium oxide matrix in the fuel pellets. In order to approach conditions prevailing deep in the groundwater, and autoclave system for dissolution experiments was developed at the Technical Research Centre of Finland. The low oxygen content and high pressure at elevated temperatures are simulated in the system. 20 MPa and 100 deg C are the upper operation limits of pressure and temperature. Water can be changed in the experiment autoclave without remarkable pressure and temperature variations. This has been arranged by using three pressure vessels: a supply vessel, a dissolution vessel and a depletion vessel. The extreme vessels serve pressure balancing purposes during water exchange. The water is deoxygenated during a preparation phase in the supply vessel by flushing it with nitrogen gas. Polytetrafluoroethylene is the principal material in contact with the water. A redox electrode couple was developed for potential measurements inside the dissolution vessel. The reference electrode is of Ag/AgCl-type with saturated KC1 electrolyte. A platinum wire operates as a measuring electrode

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

    International Nuclear Information System (INIS)

    Swanson, G.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 crucible at 1200 0 C and oxidizing with moist He at 250 0 C. The individual and solid solution oxides were isothermally equilibrated with controlled oxygen potentials between 800 and 1300 0 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. (auth)

  19. Study of the oxidation state of arsenic and uranium in individual particles from uranium mine tailings, Hungary

    International Nuclear Information System (INIS)

    Alsecz, A.; Osan, J.; Palfalvi, J.; Torok, Sz.; Sajo, I.; Mathe, Z.; Simon, R.; Falkenberg, G.

    2007-01-01

    Uranium ore mining and milling have been terminated in the Mecsek Mountains (southwest Hungary) in 1997. Mine tailings ponds are located between two important water bases, which are resources of the drinking water of the city of Pecs and the neighbouring villages. The average U concentration of the tailings material is 71.73 μg/g, but it is inhomogeneous. Some microscopic particles contain orders of magnitude more U than the rest of the tailings material. Other potentially toxic elements are As and Pb of which chemical state is important to estimate mobility, because in mobile form they can risk the water basis and the public health. Individual U-rich particles were selected with solid state nuclear track detector (SSNTD) and after localisation the particles were investigated by synchrotron radiation based microanalytical techniques. The distribution of elements over the particles was studied by micro beam X-ray fluorescence (μ-XRF) and the oxidation state of uranium and arsenic was determined by micro X-ray absorption near edge structure (μ-XANES) spectroscopy. Some of the measured U-rich particles were chosen for studying the heterogeneity with μ-XRF tomography. Arsenic was present mainly in As(V) and uranium in U(VI) form in the original uranium ore particles, but in the mine tailings samples uranium was present mainly in the less mobile U(IV) form. Correlation was found between the oxidation state of As and U in the same analyzed particles. These results suggest that dissolution of uranium is not expected in short term period. (authors)

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    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.

  2. Distribution characteristics of interlayer oxidation zone and its relationship with sedimentary facies and uranium mineralization in QJD uranium deposit, Songliao basin, NE China

    International Nuclear Information System (INIS)

    Chen Xiaolin; Xiang Weidong; Li Tiangang; Fang Xiheng; Xia Yuliang; Pang Yaqing; Zheng Jiwei; Zhang Mingyu; Zhang Zegui; Tian Shifeng

    2006-01-01

    QJD uranium deposit is a sandstone-type uranium deposit with the special shape of the interlayer oxidation zone. After studying the palaeoclimate condition and sedimentation of host sandstones, contrasting drilling cross sections, this paper suggests that primary colors of host sandstones are mostly in gray, red and yellow colors of sandstones are the result of oxidation alteration. According to the positions of drill holes with red and yellow alteration sandstones, the plane distribution of oxidation roll fronts of the Upper and Lower Member of Yaojia Formation are delineated. By contrasting the plane distribution of oxidation fronts and sedimentary facies, it can be found that oxidation fronts are obviously controlled by interchannel fine-grained sediments. The movement of uranium-bearing oxidizing groundwater can be obstructed by interchannel sediments and pelitic interbeds of host sandstones. As a result, redox interface will be developed and uranium can be concentrated in neighbouring sandstones. (authors)

  3. Assessing the Renal Toxicity of Capstone Depleted Uranium Oxides and Other Uranium Compounds

    International Nuclear Information System (INIS)

    Roszell, Laurie E.; Hahn, Fletcher; Lee, Robyn B.; Parkhurst, MaryAnn

    2009-01-01

    The primary target for uranium toxicity is the kidney. The most frequently used guideline for uranium kidney burdens is the International Commission on Radiation Protection (ICRP) value of 3 (micro)g U/g kidney, a value that is based largely upon chronic studies in animals. In the present effort, we have developed a risk model equation to assess potential outcomes of acute uranium exposure. Twenty-seven previously published case studies in which workers were acutely exposed to soluble compounds of uranium (as a result of workplace accidents) were analyzed. Kidney burdens of uranium for these individuals were determined based on uranium in the urine, and correlated with health effects observed over a period of up to 38 years. Based upon the severity of health effects, each individual was assigned a score (- to +++) and then placed into an Effect Group. A discriminant analysis was used to build a model equation to predict the Effect Group based on the amount of uranium in the kidneys. The model equation was able to predict the Effect Group with 85% accuracy. The risk model was used to predict the Effect Group for Soldiers exposed to DU as a result of friendly fire incidents during the 1991 Gulf War. This model equation can also be used to predict the Effect Group of new cases in which acute exposures to uranium have occurred

  4. Argentinian uranium production

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    A profit-making process for the exploitation of low grade uranium is presented. The process of lixiviation will be used, which will make it possible to obtain a final product whose humidity level will not exceed 10% and whose uranium oxide content will be no less than 68%. The operations of the plant are described. The plant can produce between 100 and 150 t of U 3 O 8 /yr in the form of yellow cake

  5. The influence of surface morphology and oxide microstructure on the nucleation and growth of uranium hydride on alpha uranium

    International Nuclear Information System (INIS)

    Hanrahan, R.J. Jr.; Hawley, M.E.; Brown, G.W.

    1998-01-01

    While the bulk kinetics of the uranium-hydrogen reaction are well understood, the mechanisms underlying the initial nucleation of uranium hydride on uranium remain controversial. In this study, the authors have employed environmental cell optical microscopy, Scanning Electron Microscopy (SEM) and Atomic Force Microscopy, (AFM) in an attempt to relate the structure of the surface and the microstructure of the substrate with the susceptibility and site of hydride nucleation. Samples have been investigated with varying grain size, inclusion (carbide) concentration, and thermal history. There is a clear correlation to heat treatment immediately prior to hydrogen exposure. Susceptibility to hydride formation also appears to be related to impurities in the uranium. The oxidized surface is very complex, exhibiting wide variations in thickness and topography between samples, between grains in the same sample, and within individual grains. It is, however, very difficult to relate this fine scale variability to the relatively sparse hydride initiation sites. Therefore, the surface oxide layer itself does not appear to control the sites where hydride attack is initiated, although it must play a role in the induction period prior to hydride initiation

  6. The influence of surface morphology and oxide microstructure on the nucleation and growth of uranium hydride on alpha uranium

    Energy Technology Data Exchange (ETDEWEB)

    Hanrahan, R.J. Jr.; Hawley, M.E.; Brown, G.W. [Los Alamos National Lab., NM (United States). Materials Science and Technology Div.

    1998-12-31

    While the bulk kinetics of the uranium-hydrogen reaction are well understood, the mechanisms underlying the initial nucleation of uranium hydride on uranium remain controversial. In this study the authors have employed environmental cell optical microscopy, Scanning Electron Microscopy (SEM) and Atomic Force Microscopy, (AFM) in an attempt to relate the structure of the surface and the microstructure of the substrate with the susceptibility and site of hydride nucleation. Samples have been investigated with varying grain size, inclusion (carbide) concentration, and thermal history. There is a clear correlation to heat treatment immediately prior to hydrogen exposure. Susceptibility to hydride formation also appears to be related to impurities in the uranium. The oxidized surface is very complex, exhibiting wide variations in thickness and topography between samples, between grains in the same sample, and within individual grains. It is, however, very difficult to relate this fine scale variability to the relatively sparse hydride initiation sites. Therefore, the surface oxide layer itself does not appear to control the sites where hydride attack is initiated, although it must play a role in the induction period prior to hydride initiation.

  7. The influence of surface morphology and oxide microstructure on the nucleation and growth of uranium hydride on alpha uranium

    Energy Technology Data Exchange (ETDEWEB)

    Hanrahan, R.J. Jr.; Hawley, M.E.; Brown, G.W.

    1998-12-31

    While the bulk kinetics of the uranium-hydrogen reaction are well understood, the mechanisms underlying the initial nucleation of uranium hydride on uranium remain controversial. In this study, the authors have employed environmental cell optical microscopy, Scanning Electron Microscopy (SEM) and Atomic Force Microscopy, (AFM) in an attempt to relate the structure of the surface and the microstructure of the substrate with the susceptibility and site of hydride nucleation. Samples have been investigated with varying grain size, inclusion (carbide) concentration, and thermal history. There is a clear correlation to heat treatment immediately prior to hydrogen exposure. Susceptibility to hydride formation also appears to be related to impurities in the uranium. The oxidized surface is very complex, exhibiting wide variations in thickness and topography between samples, between grains in the same sample, and within individual grains. It is, however, very difficult to relate this fine scale variability to the relatively sparse hydride initiation sites. Therefore, the surface oxide layer itself does not appear to control the sites where hydride attack is initiated, although it must play a role in the induction period prior to hydride initiation.

  8. Preconcentration of a low-grade uranium ore yielding tailings of greatly reduced environmental concerns. Part V

    International Nuclear Information System (INIS)

    Raicevic, D.; Raicevic, M.

    1980-11-01

    The low-grade ore sample used for this investigation contained 0.057 percent uranium with uranothorite as the major uranium-bearing mineral and a small amount of brannerite, occurring in the quartz-sericite matrix of a conglomerate. The preconcentration procedures, consisting of pyrite flotation with or without flotation of radioactive minerals, followed by high intensity wet magnetic treatment of the sized flotation tailings, produced pyrite and radioactive concentrates of acceptable uranium grades ranging from 0.1 to 0.135 percent uranium. The combined concentrates comprised 37 to 49 percent of the ore by weight with the following combined recoveries: 95.6 to 97.9 percent of the uranium; 94.7 to 96.3 percent of the radium; 97.8 to 99.3 percent of the thorium over 98 percent of the pyrite. The preconcentration tailings produced comprised between 51 and 63 percent of the ore by weight and contained from: 0.0022 to 0.0037 percent U; 12 to 17 pCi/g Ra; 0.002 to 0.004 percent Th less than 0.03 percent S. Because these tailings are practically pyrite-free, they should not generate acidic conditions. Due to their low radium content, their radionuclide hazards are greatly reduced. These preconcentration tailings therefore, could be suitable for surface disposal, mine backfill, revegetation or other uses

  9. Influence of attrition scrubbing, ultrasonic treatment, and oxidant additions on uranium removal from contaminated soils

    International Nuclear Information System (INIS)

    Timpson, M.E.; Elless, M.P.; Francis, C.W.

    1994-01-01

    As part of the Uranium in Soils Integrated Demonstration Project being conducted by the US Department of Energy, bench-scale investigations of selective leaching of uranium from soils at the Fernald Environmental Management Project site in Ohio were conducted at Oak Ridge National Laboratory. Two soils (storage pad soil and incinerator soil), representing the major contaminant sources at the site, were extracted using carbonate- and citric acid-based lixiviants. Physical and chemical processes were used in combination with the two extractants to increase the rate of uranium release from these soils. Attrition scrubbing and ultrasonic dispersion were the two physical processes utilized. Potassium permanganate was used as an oxidizing agent to transform tetravalent uranium to the hexavalent state. Hexavalent uranium is easily complexed in solution by the carbonate radical. Attrition scrubbing increased the rate of uranium release from both soils when compared with rotary shaking. At equivalent extraction times and solids loadings, however, attrition scrubbing proved effective only on the incinerator soil. Ultrasonic treatments on the incinerator soil removed 71% of the uranium contamination in a single extraction. Multiple extractions of the same sample removed up to 90% of the uranium. Additions of potassium permanganate to the carbonate extractant resulted in significant changes in the extractability of uranium from the incinerator soil but had no effect on the storage pad soil

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

    Science.gov (United States)

    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

  11. Possible domestication of uranium oxides using biological assistance reduction

    Directory of Open Access Journals (Sweden)

    Slah Hidouri

    2017-01-01

    Full Text Available Uranium has been defined in material research engineering field as one of the most energetic radioactive elements in the entire Mendeleev periodic table. The manipulation of uranium needs higher theories and sophisticated apparatus even in nuclear energy extraction or in many other chemical applications. Above the nuclear exploitation level, the chemical conventional approaches used, require a higher temperature and pressure to control the destination of ionic form. However, it has been discovered later that at biological scale, the manipulation of this actinide is possible under friendly conditions. The review summarizes the relevant properties of uranium element and a brief characterization of nanoparticles, based on some structural techniques. These techniques reveal the common link between chemical approaches and biological assistance in nanoparticles. Also, those biological entities have been able to get it after reduction. Uranium is known for its ability to destroy ductile materials. So, if biological cell can really reduce uranium, then how does it work?

  12. Thermochemistry of uranium(VI), arsenic, and alkali metal triple oxides

    International Nuclear Information System (INIS)

    Karyakin, N.V.; Chernorukov, G.N.

    1994-01-01

    The standard enthalpies of reactions of stoichiometric mixtures of potassium dyhydrogen orthoarsenate, uranium(VI) oxide, alkali metal nitrates, and of mixtures of triple oxides with the general formula M I AsUO 6 (M I =Li, Na, K, Rb, and Cs) and potassium nitrate with aqueous solution of hydrofluoric acid were determined an an adiabatic calorimeter at 298.15 K. The standard enthalpies of formation of uranium(VI), arsenic, and alkali metal triple oxides at 298.15 K were calculated form the data obtained. 8 refs., 1 tab

  13. Determination of uranium in plutonium--238 metal and oxide by differential pulse polarography

    International Nuclear Information System (INIS)

    Fawcett, N.C.

    1976-01-01

    A differential pulse polarographic method was developed for the determination of total uranium in 238 Pu metal and oxides. A supporting electrolyte of 0.5 M ascorbic acid in 0.15 N H 2 SO 4 was found satisfactory for the determination of 500 ppM or more of uranium in 10 mg or less of plutonium. A relative standard deviation of 0.27 to 4.3 percent was obtained in the analysis of samples ranging in uranium content from 0.65 to 2.79 percent. The limit of detection was 0.18 μg ml -1 . Peak current was a linear function of uranium concentration up to at least 100 μg ml -1 . Amounts of neptunium equal to the uranium content were tolerated. The possible interference of a number of other cations and anions were investigated

  14. Electrochemical behavior of uranium oxide in a LiCl-Li2O molten salt with the integrated cathode assembly

    International Nuclear Information System (INIS)

    Park, Sung Bin; Park, Byung Heung; Kang, Dae Seoung; Kwon, Seon Gil; Seo, Chung Seok; Park, Seong Won

    2005-01-01

    Electrochemical reduction of uranium oxide to uranium metal was studied in a LiCl-Li 2 O molten salt system. By means of a cyclic voltammetry and a chronopotentiometry, the electrolytic reduction of uranium oxide has been studied to establish the reduction mechanisms and the effects of the thickness of the uranium oxide on the overpotential of the cathode and anode were investigated. From the voltamograms, the reduction potentials of the uranium oxide and Li 2 O were obtained and the two mechanisms of the electrolytic reduction were considered with regards to the applied cathode potential. In the chronopotentiograms, the exchange current, the transfer coefficient and the maximum allowable current based on the Tafel behavior were obtained according to the thickness of the uranium oxide which is loaded into the porous MgO membrane. (author)

  15. Measurement of the oxidation-extraction of uranium from wet-process phosphoric acid

    International Nuclear Information System (INIS)

    Lawes, B.C.

    1985-01-01

    The present invention relates to processes for the recovery of uranium from wet-process phosphoric acid and more particularly to the oxidation-extraction steps in the DEPA-TOPO process for such recovery. A more efficient use of oxidant is obtained by monitoring the redox potential during the extraction step

  16. Biomembrane oxidizing tank used in the process of bacterial heap leaching of uranium ore

    International Nuclear Information System (INIS)

    Meng Yunsheng; Fan Baotuan; Liu Jian; Zheng Ying; Liu Chao

    2004-01-01

    The construction characteristic of biomembrane oxidizing tank and specialty of packing material used in the process of bacterial heap leaching of uranium ore are introduced in this paper. Method for designing biomembrane oxidizing tank, layout principle of aeration system and measurements on running management are summarized

  17. Simulation of uranium oxides reduction kinetics by hydrogen. Reactivities of germination and growth

    International Nuclear Information System (INIS)

    Brun, C.

    1997-01-01

    The aim of this work is to simulate the reduction by hydrogen of the tri-uranium octo-oxide U 3 O 8 (obtained by uranium trioxide calcination) into uranium dioxide. The kinetics curves have been obtained by thermal gravimetric analysis, the hydrogen and steam pressures being defined. The geometrical modeling which has allowed to explain the trend of the kinetics curves and of the velocity curves is an anisotropic germination-growth modeling. The powder is supposed to be formed of spherical grains with the same radius. The germs of the new UO 2 phase appear at the surface of the U 3 O 8 grains with a specific germination frequency. The growth reactivity is anisotropic and is very large in the tangential direction to the grains surface. Then, the uranium dioxide growths inside the grain and the limiting step is the grain surface. The variations of the growth reactivity and of the germination specific frequency in terms of the gases partial pressures and of the temperature have been explained by two different mechanisms. The limiting step of the growth mechanism is the desorption of water in the uranium dioxide surface. Concerning the germination mechanism the limiting step is a water desorption too but in the tri-uranium octo-oxide surface. The same geometrical modeling and the same germination and growth mechanisms have been applied to the reduction of a tri-uranium octo-oxide obtained by calcination of hydrated uranium trioxide. The values of the germination specific frequency of this solid are nevertheless weaker than those of the solid obtained by direct calcination of the uranium trioxide. (O.M.)

  18. Oxidative lixiviation of pitchblende and precipitation of uranium with hydrogen; Lixiviation oxydante des pechblendes et precipitation de l'uranium par l'hydrogene

    Energy Technology Data Exchange (ETDEWEB)

    Mouret, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Balaceanu, J C; Coussemant, F [Institut Francais du Petrole (IFP), 92 - Rueil-Malmaison (France)

    1958-07-01

    Earlier work on the preparation of uranium by F.A. Forward and his colleagues has shown the possibilities presented by oxidative lixiviation of ores in a carbonate medium, and the catalytic reduction of uranyl carbonate solutions by hydrogen. The carbonate attack is of considerable interest because of the selectivity of the uranium dissolution, which means it can be applied particularly to the treatment of low grade ores with a reduced consumption of cheap reagents. The subsequent reduction with hydrogen is of the same nature, and not only enables relatively dilute uranyl carbonate solutions to be treated, but also avoids any significant alteration of the attacking solution which can therefore be used again in the lixiviation stage. The experimental work, undertaken at the request of the Commissariat a I'Energie Atomique, was aimed at determining the quantitative characteristics of each of the two stages in order to ascertain their possibilities for industrial application to the principal low grade ores found in France. (author) [French] Les travaux anterieurs de F.A. FORWARD et de ses collaborateurs ont mis en evidence les possibilites que presentent, dans la preparation de l'uranium, la lixiviation oxydante des minerais en milieu carbonate, et la reduction catalytique des solutions d'uranyl carbonate par l'hydrogene. L'attaque carbonatee presente, en effet, un interet considerable du fait de la selectivite de la dissolution de l'uranium qui permet de l'appliquer en particulier au traitement des minerais pauvres avec une consommation reduite de reactifs peu couteux. La reduction subsequente par l'hydrogene presente les memes caracteres et permet non seulement de traiter des solutions relativement diluees d'uranyl carbonate, mais encore evite toute modification significative de la solution d'attaque qui peut donc etre reemployee dans l'etape de lixiviation. L'experimentation, entreprise a la demande du Commissariat a l'Energie atomique, avait pour but de determiner

  19. Ore controlling oxidized zonation epigenetic uranium-coal deposits and regularities in lignite transformations

    International Nuclear Information System (INIS)

    Uspenskij, V.A.; Kulakova, Ya.M.

    1982-01-01

    Complex of analytical methods was used to study epigenetic transformations in uranium-coal ore manifestation. To clarify the principle scheme of oxidized zonation in coals the materials, related to three similar objects were used. When comparing obtained epigenetic column with columns of similar ore objects the principle scheme of oxidized epigenetic zonation for ancient infiltration uranium-coal deposits was specified; general regularities of eignite transformations and characteristics of profile distribution of uranium and accessory metal zonations were revealed. Infiltration processes, proceeded in coal measureses, formed the steady epigenetic oxidized zonation: O - zone of barren unoxidized coals, 1 - zone of ore-bearing unoxidized coals, 2 - zone of weakly ore-bearing oxidized coals, 3 - zone of oxidized terrigenous rocks with zonules of development of yellow and red iron hydroxides. Capacities of some zones and zonules reflect the intensity and duration of ore-forming processes. Distribution of U and accessory elements obeys completely epigenetic zonation. It is assumed, that ancient infiltration uranium-coal deposits formed due to weakly uranium-bearing oxygen-containing waters

  20. Treatment of an isolated high-grade, low-tonnage uranium orebody

    International Nuclear Information System (INIS)

    Ajuria-Garza, S.

    1976-01-01

    A completely integrated process leading to a commercial-grade uranium concentrate has been developed for the El Nopal orebody which contains 115000t of ore with an average grade of 0.283%, equivalent to 325t of U 3 O 8 . The process consists of crushing (from -12 in to -1.5in), heap leaching with recirculation, a special type of countercurrent washing (also in the heap), solvent extraction to give an exceptionally high uranium concentration, re-extraction, precipitation with ammonium hydroxide, filtration and calcining. The main factors that influence heap leaching are analysed in detail: heap geometry, crushed ore size distribution, base design and construction, method of heaping, acid feeding method, flow of liquors through the heap and washing procedures. Leaching efficiencies range from 80 to 85% and washing efficiencies from 96 to almost 100% with an overall extraction efficiency of 77 to 85%. Acid consumption is usually less than 25kg/t of ore. The leaching and washing processes described are designed to use an overall solid-to-liquid ratio of about 3:1 thus producing very concentrated liquors with a U 3 O 8 content of about 7g/l. These liquors are filtered and fed directly to a solvent extraction system using Alamine 336 (tri-capryl amine) and isodecanol in kerosene. The high feed concentration makes it possible to use a high concentration of amine. The organic solvent contains 120g/l of Alamine and 98g/l of isodecanol. Re-extraction can be with a saline solution containing 100g/lNH 4 Cl and 250g/l(NH 4 )SO 4 . The strong liquor obtained from re-extraction contains 70-75g/lU 3 O 8 . The remaining steps of precipitation, filtration and calcining can then follow general practice and lead to a product which meets currently accepted commercial specifications. As an alternative, nitrate re-extraction has been shown to produce a solution of about 70g/l which will feed directly to tributyl phosphate (TBP) refining without the necessity of ever producing a

  1. The collection of uranium from sea water with hydrous metal oxide, 3

    International Nuclear Information System (INIS)

    Yamashita, Hisao; Nakajima, Fumito; Ozawa, Yoshihiro; Murata, Toshifumi.

    1980-01-01

    The effect of diverse ions present in sea water on the uranium adsorption is elucidated in the present paper. The uranium-adsorption experiments were conducted using sea water and a solution containing 0.72 mol dm -3 NaCl and 2.3 x 10 -3 mol dm -3 NaHCO 3 . The uranium uptake was about ten times larger from the NaCl-NaHCO 3 solution than from sea water. The ions which depressed the uranium uptake were the calcium, magnesium, and fluoride present in sea water. Among these ions, calcium had the largest effect on the uranium uptake. The analysis of calcium and carbonate in the adsorbent after the adsorption experiment has revealed that the molar ratio between calcium and carbonate was about one. It was considered that calcium carbonate was deposited on the adsorbent during the uranium adsorption. The specific surface area and the pore volume decreased after the deposition of calcium carbonate. It was supposed that the decrease in the uranium uptake was caused by the coverage of the surface of hydrous titanium(IV) oxide with calcium carbonate. Magnesium ions depressed the uranium uptake in the same manner as calcium ions. The effect of the magnesium ions, however, was relatively small compared with that of the calcium ions. (author)

  2. Uranium

    International Nuclear Information System (INIS)

    Whillans, R.T.

    1981-01-01

    Events in the Canadian uranium industry during 1980 are reviewed. Mine and mill expansions and exploration activity are described, as well as changes in governmental policy. Although demand for uranium is weak at the moment, the industry feels optimistic about the future. (LL)

  3. The effect of hydrogen peroxide on uranium oxide films on 316L stainless steel

    Science.gov (United States)

    Wilbraham, Richard J.; Boxall, Colin; Goddard, David T.; Taylor, Robin J.; Woodbury, Simon E.

    2015-09-01

    For the first time the effect of hydrogen peroxide on the dissolution of electrodeposited uranium oxide films on 316L stainless steel planchets (acting as simulant uranium-contaminated metal surfaces) has been studied. Analysis of the H2O2-mediated film dissolution processes via open circuit potentiometry, alpha counting and SEM/EDX imaging has shown that in near-neutral solutions of pH 6.1 and at [H2O2] ⩽ 100 μmol dm-3 the electrodeposited uranium oxide layer is freely dissolving, the associated rate of film dissolution being significantly increased over leaching of similar films in pH 6.1 peroxide-free water. At H2O2 concentrations between 1 mmol dm-3 and 0.1 mol dm-3, formation of an insoluble studtite product layer occurs at the surface of the uranium oxide film. In analogy to corrosion processes on common metal substrates such as steel, the studtite layer effectively passivates the underlying uranium oxide layer against subsequent dissolution. Finally, at [H2O2] > 0.1 mol dm-3 the uranium oxide film, again in analogy to common corrosion processes, behaves as if in a transpassive state and begins to dissolve. This transition from passive to transpassive behaviour in the effect of peroxide concentration on UO2 films has not hitherto been observed or explored, either in terms of corrosion processes or otherwise. Through consideration of thermodynamic solubility product and complex formation constant data, we attribute the transition to the formation of soluble uranyl-peroxide complexes under mildly alkaline, high [H2O2] conditions - a conclusion that has implications for the design of both acid minimal, metal ion oxidant-free decontamination strategies with low secondary waste arisings, and single step processes for spent nuclear fuel dissolution such as the Carbonate-based Oxidative Leaching (COL) process.

  4. The crystallographic structure of the air-grown oxide on depleted uranium metal

    International Nuclear Information System (INIS)

    Jones, Christopher P.; Petherbridge, James R.; Davis, Sean A.; Jones, Jonathon A.; Scott, Thomas B.

    2016-01-01

    Highlights: • Oxidation of depleted uranium coupons under ambient conditions and 150 °C. • Oxide characterised using SEM, TEM and electron backscatter diffraction analysis, • Layer comprises of UO 2 crystallites 12 nm in diameter. • Preferred [110] growth direction normal to the surface of the metal. • Oxide growth direction is independent of the underlying crystal orientation. - Abstract: Oxide formation on depleted uranium metal was investigated using a combination of electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM) characterisation. Diffraction analysis of the oxide revealed an FCC crystalline formation of UO 2 crystallites whilst TEM data indicated an average grain size of 12 nm with a standard deviation of 3.8 nm. EBSD analysis revealed a preferential texture of [110] normal to the surface of the metal. This data implied that lattice matching between the oxide and the underlying metal did not occur, therefore, the observed preferential growth direction is independent of the underlying crystal orientation.

  5. Effects of pH on uranium uptake and oxidative stress responses induced in Arabidopsis thaliana

    OpenAIRE

    Saenen, Eline; Horemans, Nele; Vanhoudt, Nathalie; Vandenhove, Hildegarde; Biermans, Geert; Van Hees, May; Wannijn, Jean; Vangronsveld, Jaco; Cuypers, Ann

    2013-01-01

    Uranium (U) causes oxidative stress in Arabidopsis thaliana plants grown at pH 5.5. However, U speciation and its toxicity strongly depend on environmental parameters, for example pH. It is unknown how different U species determine U uptake and translocation within plants and how they might affect the oxidative defense mechanisms of these plants. The present study analyzed U uptake and oxidative stress-related responses in A. thaliana (Columbia ecotype) under contrasted U chemical speciation ...

  6. Exploration on trickle leaching of uranium ore by refreshed liquor of bacterial oxidation

    International Nuclear Information System (INIS)

    Chen Shian; Huang Xiangfu; Fan Baotuan

    1995-01-01

    The paper describes the adaptation of the domesticated thiobacillus ferroxidans to the trickle leaching conditions of uranium ore. When the bacterial leaching liquor through multiple cycles of oxidation and regeneration was used to return to the trickle leaching, the following results were obtained: the extraction rate was more than 95%, the acid consumption was saved by 30%, and the consumed 2.0% pyrolusite (MnO 2 40%) was eliminated. The following problems are discussed: the basic principle, process and some factors influencing the process of the trickle leaching of uranium ore using regenerated liquor of bacterial oxidation, counter-current trickle leaching mode, oxidation and regeneration techniques of bacterial leaching liquor and other technological problems on the process of uranium extraction by thiobacillus ferroxidans

  7. XPS characterization of the anodic oxide film formed on uranium metal in sodium hydroxide solution

    International Nuclear Information System (INIS)

    Fu Xiaoguo; Wang Xiaolin; Guo Huanjun; Wang Qingfu; Zhao Zhengping; Zhong Yongqiang

    2002-01-01

    X-ray photoelectron spectroscopy (XPS) is used to examine the anodic oxide film formed on uranium metal in 0.8 mol/L NaOH solution. The U4f 7/2 fitting spectra suggests that the anodic oxide film is composed of uranium trioxide and a small amount of UO 2+x . Under UHV condition, the U4f peak shifts to the lower binding energy, while a gradual increase in the intensity of U5f peak and the broad of U4f peak are also observed. All of these changes are due to reduction of uranium trioxide in the anodic oxide film. XPS quantitative analysis confirms the occurrence of reduction reaction

  8. Standard test method for determination of impurities in nuclear grade uranium compounds by inductively coupled plasma mass spectrometry

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This test method covers the determination of 67 elements in uranium dioxide samples and nuclear grade uranium compounds and solutions without matrix separation by inductively coupled plasma mass spectrometry (ICP-MS). The elements are listed in Table 1. These elements can also be determined in uranyl nitrate hexahydrate (UNH), uranium hexafluoride (UF6), triuranium octoxide (U3O8) and uranium trioxide (UO3) if these compounds are treated and converted to the same uranium concentration solution. 1.2 The elements boron, sodium, silicon, phosphorus, potassium, calcium and iron can be determined using different techniques. The analyst's instrumentation will determine which procedure is chosen for the analysis. 1.3 The test method for technetium-99 is given in Annex A1. 1.4 The values stated in SI units are to be regarded as standard. 1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish ...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-02-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  11. Quantification of trace level of fluoride content in uranium oxide produced by deconversion of HEX gas by ion chromatography

    International Nuclear Information System (INIS)

    Unnikrishnan, E.K.; Padmakumar, P.R.; Shanmugavelu, P.; Sudhakar, T.M.; Bhowmik, A.

    2015-01-01

    Fluoride content in nuclear fuel is detrimental due to its corrosion behavior with cladding material. It is essential to monitor and control the fluoride concentration in nuclear material at various processing stages. Deconversion of upgraded HEX gas is carried out to produce uranium oxide. The performance of the deconversion process of HEX gas is evaluated for which trace level of fluoride concentration accompanying uranium oxide is considered as a marker. An analytical method has been developed for testing the uranium oxide produced from deconversion process of HEX gas. The method involves sample pretreatment followed by analysis using ion chromatography. The test method was validated for its performance using in house synthetic uranyl fluoride (UO 2 F 2 ) standard solutions prepared with different level of fluoride content. The results are in agreement with the expected values with the recovery in the range of 80-95%. This method has been successfully implemented for routine analysis of samples at our lab. Since UO 2 F 2 reference material is not available to validate this method, in house UO 2 F 2 standards were prepared from U 3 O 8 prepared from nuclear grade uranyl nitrate solution. UO 2 F 2 standards were prepared by converting U 3 O 8 to UO 2 F 2 by the addition of HF followed by H 2 O 2 at 200°C on a hot plate. The entire yellow colored UO 2 F 2 was dissolved in nano pure water and recrystallised several times to ensure that all free HF is removed. The crystals dried in air oven at 120° for three hours. Samples containing 1000 mg kg -1 fluoride prepared from this UO 2 F 2 , and subsequently from this sample containing 5 mg kg -1 to 35 mg kg -1 fluoride samples were prepared and analysed against fluoride CRM and the fluoride concentration obtained was analysed

  12. Pilot-scale demonstration of the modified direct denitration process to prepare uranium oxide for fuel fabrication evaluation

    International Nuclear Information System (INIS)

    Kitts, F.G.

    1994-04-01

    The Uranium-Atomic Vapor Laser Isotope Separation (U-AVLIS) Program has the objective of developing a cost-competitive enrichment process that will ultimately replace the gaseous diffusion process used in the United States. Current nuclear fuel fabricators are set up to process only the UF 6 product from gaseous diffusion enrichment. Enriched uranium-iron alloy from the U-AVLIS separator system must be chemically converted into an oxide form acceptable to these fabricators to make fuel pellets that meet American Society for Testing and Materials (ASTM) and utility company specifications. A critical step in this conversion is the modified direct denitration (MDD) that has been selected and presented in the AVLIS Conceptual Design for converting purified uranyl nitrate to UO 3 to be shipped to fabricators for making UO 2 pellets for power reactor fuel. This report describes the MDD process, the equipment used, and the experimental work done to demonstrate the conversion of AVLIS product to ceramic-grade UO 3 suitable for making reactor-grade fuel pellets

  13. Heap leaching for uranium

    International Nuclear Information System (INIS)

    1988-01-01

    Denison Mines Ltd. is using two bacterial leaching processes to combat the high cost of extracting uranium from low grade ore in thin reefs. Both processes use thiobacillus ferro-oxidans, a bacterium that employs the oxidation of ferrous iron and sulphur as its source of energy for growth. The first method is flood leaching, in which ore is subjected to successive flood, drain and rest cycles. The second, trickle leaching, uses sprinklers to douse the broken muck continuously with leaching solution. In areas where grades are too low to justify the expense of hauling the ore to the surface, the company is using this biological process underground to recover uranium. In 1987 Denison recovered 840 000 lb of uranium through bacterial heap leaching. It plans to have biological in-place leaching contribute 25% of the total uranium production by 1990. (fig.)

  14. White paper on possible inclusion of mixed plutonium-uranium oxides in DOE-STD-3013-96

    International Nuclear Information System (INIS)

    Haschke, J.M.; Venetz, T.; Szempruch, R.; McClard, J.W.

    1997-11-01

    This report assesses stabilization issues concerning mixed plutonium-uranium oxides containing 50 mass % Pu. Possible consequences of uranium substitution on thermal stabilization, specific surface areas, moisture readsorption behavior, loss-on-ignition analysis, and criticality safety of the oxide are examined and discussed

  15. Polarography applied to the determination of uranium oxide composition; Application de la polarographie a la determination de la composition d'oxydes d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Nens, C; Canton, C; Molina, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires, Departement de Chimie, Services d' Etudes Chimiques et d' Analyse, Service d' Etudes Analytiques, Section de Chimie Analytique du Plutonium et d' Electroanalyse

    1967-03-01

    An analytical method based on conventional polarography has been developed, for the determination of the O/U ratio in uranium oxides. The dissolution of the samples is effected by means of molten ammonium bifluoride. After a transfer to aqueous solution, polarography is used to determine the oxide composition by measurement of both the hexavalent and the total uranium. (author) [French] Une methode d'analyse utilisamt la polarographie conventionnelle a ete mise au point pour la determination du rapport O/U dans les oxydes d'uranium. La mise en solution des echgantillons est realisee en milieu de bifluorure d'ammonium fondu. Apres passage en salution aqueuse, la polarographie permet d'atteindre la composition de l'oxyde par determination de l'uranium hexavalent et de l'uranium total. (auteur)

  16. The relations between hydrodynamic characteristics and interbedding oxidation zone type uranium mineralization

    International Nuclear Information System (INIS)

    Bai Jingping

    2001-01-01

    Infiltrating type hydrodynamic way controls the formation of interbedding oxidation zone type uranium deposit. The author analyzes hydrodynamic condition of Songliao basin and concludes that during evolution and development of Songliao basin, Water-bearing petrofabric of Mingshui Formation and above inherit completely infiltrating hydrodynamic way as they were deposited and that Sifangtai Formation inherit the way to some extent, that below Sifangtai Formation water bearing petrofabric were completely reformed in northern part of Songliao watershed. The contact line between infiltrating and out filtrating type hydrodynamic way, e.g. underground water dividing lines formed in different geological period, restricts development of interbedding oxidation zone in this period and controls uranium mineralization

  17. Reaction of Antimony-Uranium Composite Oxide in the Chlorination Treatment of Waste Catalyst - 13521

    International Nuclear Information System (INIS)

    Sawada, Kayo; Hirabayashi, Daisuke; Enokida, Youichi

    2013-01-01

    The effect of oxygen gas concentration on the chlorination treatment of antimony-uranium composite oxide catalyst waste was investigated by adding different concentrations of oxygen at 0-6 vol% to its chlorination agent of 0.6 or 6 vol% hydrogen chloride gas at 1173 K. The addition of oxygen tended to prevent the chlorination of antimony in the oxide. When 6 vol% hydrogen chloride gas was used, the addition of oxygen up to 0.1 vol% could convert the uranium contained in the catalyst to U 3 O 8 without any significant decrease in the reaction rate compared to that of the treatment without oxygen. (authors)

  18. Ammonium carbonate and/or bicarbonate plus alkaline chlorate oxidant for recovery of uranium values

    International Nuclear Information System (INIS)

    Stapp, P.R.

    1983-01-01

    In accordance with the present invention, uranium values are extracted from materials containing uranium in valence states lower than its hexavalent state by contacting the materials containing uranium with an aqueous alkaline leach solution containing an alkaline chlorate in an amount sufficient to oxidize at least a portion of the uranium in valence states lower than its hexavalent state to its hexavalent state. In a further embodiment of the present invention, the alkaline leach solution is an aqueous solution of a carbonate selected from the group consisting of ammonium carbonate, ammonium bicarbonate and mixtures thereof. In yet another embodiment of the present invention, at least one catalytic compound of a metal selected from the group consisting of copper, cobalt, iron, nickel, chromium and mixtures thereof adapted to assure the presence of the ionic species Cu ++ , Co ++ , Fe +++ , Ni ++ , Cr +++ and mixtures thereof, respectively, during the contacting of the material containing uranium with the alkaline leach solution and in an amount sufficient to catalyze the oxidation of at least a portion of the uranium in its lower valence states to its hexavalent state, is present

  19. Specification for nuclear-grade beryllium oxide powder

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This specification defines the physical and chemical requirements of nuclear-grade beryllium oxide (BeO) powder to be used in fabricating nuclear components. 1.2 This specification does not include requirements for health and safety. , , It recognizes the material as a Class B poison and suggests that producers and users become thoroughly familiar with and comply to applicable federal, state, and local regulations and handling guidelines. 1.3 Special tests and procedures are given in Annex A1 and Annex A2. 1.4 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

  20. Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R M

    1976-01-01

    Evidence of expanding markets, improved prices and the short supply of uranium became abundantly clear in 1975, providing the much needed impetus for widespread activity in all phases of uranium operations. Exploration activity that had been at low levels in recent years in Canada was evident in most provinces as well as the Northwest Territories. All producers were in the process of expanding their uranium-producing facilities. Canada's Atomic Energy Control Board (AECB) by year-end had authorized the export of over 73,000 tons of U/sub 3/0/sub 8/ all since September 1974, when the federal government announced its new uranium export guidelines. World production, which had been in the order of 25,000 tons of U/sub 3/0/sub 8/ annually, was expected to reach about 28,000 tons in 1975, principally from increased output in the United States.

  1. The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

    International Nuclear Information System (INIS)

    Townes, B.M.; Hilborn, J.W.

    1985-06-01

    A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

  2. The collection of uranium from sea water with hydrous metal oxide, 2

    International Nuclear Information System (INIS)

    Yamashita, Hisao; Nakajima, Fumito; Ozawa, Yoshihiro; Murata, Toshifumi.

    1980-01-01

    The mechanism of uranium adsorption from sea water by hydrous titanium(IV) oxide was investigated. The uranium adsorption experiments were conducted using a solution containing NaCl, NaHCO 3 , and uranium. Thermochemical calculation showed that the tris(carbonato)dioxouranate(VI) ion (UO 2 (CO 3 ) 3 ) 4- found in sea water existed in the solution at pH 8 and 25 0 C. The uranium uptake varied with the pH of the solution, exhibiting a minimum value at pH 8. The enthalpy change, delta H, and the activation energy, E, of the uranium adsorption were found to be 23.6 kJ mol -1 and 52.7 kJ mol -1 respectively. The analysis of carbonate in the adsorbent showed that the carbonate ion in(UO 2 (CO 3 ) 3 ) 4- was released into the solution during the uranium adsorption. On the basis of the present experimental results, the mechanism of uranium adsorption was discussed. (author)

  3. Uranium dioxide and beryllium oxide enhanced thermal conductivity nuclear fuel development

    International Nuclear Information System (INIS)

    Andrade, Antonio Santos; Ferreira, Ricardo Alberto Neto

    2007-01-01

    The uranium dioxide is the most used substance as nuclear reactor fuel for presenting many advantages such as: high stability even when it is in contact with water in high temperatures, high fusion point, and high capacity to retain fission products. The conventional fuel is made with ceramic sintered pellets of uranium dioxide stacked inside fuel rods, and presents disadvantages because its low thermal conductivity causes large and dangerous temperature gradients. Besides, the thermal conductivity decreases further as the fuel burns, what limits a pellet operational lifetime. This research developed a new kind of fuel pellets fabricated with uranium dioxide kernels and beryllium oxide filling the empty spaces between them. This fuel has a great advantage because of its higher thermal conductivity in relation to the conventional fuel. Pellets of this kind were produced, and had their thermophysical properties measured by the flash laser method, to compare with the thermal conductivity of the conventional uranium dioxide nuclear fuel. (author) (author)

  4. Uranium and plutonium distribution in unirradiated mixed oxide fuel from industrial fabrication

    International Nuclear Information System (INIS)

    Hanus, D.; Kleykamp, H.

    1982-01-01

    Different process variants developed in the last few years by the firm ALKEM to manufacture FBR and LWR mixed oxide fuel are given. The uranium and plutonium distribution is determined on the pellets manufactured with the help of the electron beam microprobe. The stepwise improvement of the uranium-plutonium homogeneity in the short-term developed granulate variants and in the long-term developed new processes are illustrated starting with early standard processes for FBR fuel. An almost uniform uranium-plutonium distribution could be achieved for the long-term developed new processes (OKOM, AuPuC). The uranium-plutonium homogeneity are quantified in the pellets manufactured according to the considered process variants with a newly defined quality number. (orig.)

  5. Study the oxidation kinetics of uranium using XRD and Rietveld method

    Energy Technology Data Exchange (ETDEWEB)

    Zhang Yanzhi; Guan Weijun; Wang Qinguo; Wang Xiaolin; Lai Xinchun; Shuai Maobing, E-mail: yanzhizh@163.com [China Academy of Engineering Physics, PO Box 919-71, Mianyang, Sichuan, 621900 (China)

    2010-03-15

    The surface oxidation of uranium metal has been studied by X-ray diffraction (XRD) and Rietveld method in the range of 50{approx}300deg. 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 UO{sub 2} 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 UO{sub 2} 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.

  6. Study the oxidation kinetics of uranium using XRD and Rietveld method

    Science.gov (United States)

    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.

  7. Critical experiments on low-enriched uranium oxide system with H/U=1.25

    International Nuclear Information System (INIS)

    Oh, I.; Rothe, R.E.; Tuck, G.

    1982-01-01

    Fifteen (15) critical experiments were performed on a horizontal split table machine using 4.48%-enriched sup(235)U uranium oxide(U 3 O 8 ). The oxide was compacted to a density of 4.68g/cm 3 and placed in 152 mm cubical aluminum cans. Water was added to achive an H/U of 1.25. Various arrays of oxide cans were distributed on each half of the split table, and the separation between halves reduced until criticality occurred. The critical table separation varied from 3.59 mm to 18.40 mm. Twelve (12) experiments required the addition of a high-enriched(-93 %sup(235)U) metal or solution driver to achieve criticality. These experiments were performed in a plastic, concrete, or thin steel reflector. Three additional experiments in the plastic reflector contained either 9.3-mm- or 24.3-mm-thick plastic moderator material between the oxide cans and did not require a driver to achieve criticality. Critical uranium driver masses ranged from 9.999 kg to 14.000 kg (solution driver), and from 25.378 kg to 29.278 kg (metal driver) for 5X5X5 arrays of uranium oxide cans. Always, one or four of these 125 cans had to be removed to make room for the drivers. Therefore, the uranium oxide masses used were 1823.8 kg and 1863.5 kg. For the moderated experiments, the uranium oxide mass ranged between 574.4 kg and 1210.0 kg. (Author)

  8. Study of interaction of uranium, plutonium and rare earth fluorides with some metal oxides in fluoric salt melts

    International Nuclear Information System (INIS)

    Gorbunov, V.F.; Novoselov, G.P.; Ulanov, S.A.

    1976-01-01

    Interaction of plutonium, uranium, and rare-earth elements (REE) fluorides with aluminium and calcium oxides in melts of eutectic mixture LiF-NaF has been studied at 800 deg C by X-ray diffraction method. It has been shown that tetravalent uranium and plutonium are coprecipitated by oxides as a solid solution UO 2 -PuO 2 . Trivalent plutonium in fluorides melts in not precipitated in the presence of tetravalent uranium which can be used for their separation. REE are precipitated from a salt melt by calcium oxide and are not precipitated by aluminium oxide. Thus, aluminium oxide in a selective precipitator for uranium and plutonium in presence of REE. Addition of aluminium fluoride retains trivalent plutonium and REE in a salt melt in presence of Ca and Al oxides. The mechanism of interacting plutonium and REE trifluorides with metal oxides in fluoride melts has been considered

  9. Metallogenic characteristics, model and exploration prospect for the paleo-interlayer-oxidation type sandstone-hosted uranium deposits in China

    International Nuclear Information System (INIS)

    Huang Jingbai; Li Shengxiang

    2007-01-01

    In this paper, the paleo-interlayer-oxidation type sandstone-hosted uranium deposits occurred in the Meso-Cenozoic continental basins in China are divided into 3 subtype, they are stratum over lapping buried subtype, structure-uplifting destroy subtype and faulted-folding conserved subtype. The metallogenic characteristics, metallogenic model and exploration prospect for these 3 subtypes uranium deposits are discussed. It is proposed that the paleo-interlayer-oxidation type sandstone-hosted uranium deposits, besides the recent interlayer oxidation type sandstone-hosted uranium deposits, are of great prospecting potential in the Meso-Cenozoic continental basins in China. Therefore, the metallogenic theory of these types uranium deposits should be conscientiously summarized and replenished continuously so as to propel forward the exploration of the sandstone-hosted uranium deposits in China. (authors)

  10. The study on microb and organic metallogenetic process of the interlayer oxidized zone uranium deposit. A case study of the Shihongtan uranium deposit in Turpan-Hami basin

    International Nuclear Information System (INIS)

    Qiao Haiming; Shang Gaofeng

    2010-01-01

    Microbial and organic process internationally leads the field in the study of metallogenetic process presently. Focusing on Shi Hongtan uranium deposit, a typical interlayer oxidized zone sandstone-type deposit, this paper analyzes the geochemical characteristics of microb and organic matter in the deposit, and explores the interaction of microb and organic matter. It considers that the anaerobic bacterium actively takes part in the formation of the interlayer oxidized zone, as well as the mobilization and migration of uranium. In the redox (oxidation-reduction) transition zone, sulphate-reducing bacteria reduced sulphate to stink damp, lowing Eh and acidifying pH in the groundwater, which leads to reducing and absorbing of uranium, by using light hydrocarbon which is the product of the biochemical process of organism and the soluble organic matter as the source of carbon. The interaction of microb and organic matter controls the metallogenetic process of uranium in the deposit. (authors)

  11. Electrodeposition of uranium metal by reduction of uranium oxides in molten Lif-KF=NaF-CaF 2-UF4

    International Nuclear Information System (INIS)

    Pao, D.S.; Burris, L.; Steunenberg, R.K.; Tomczuk, Z.

    1990-01-01

    Although electrolytic reduction of uranium oxides was shown to be feasible in the early 1960's it is recognized that considerable improvement in the electrolytic reduction technology must be achieved for practical applications. This exploratory work on electrolytic reduction of uranium oxide was undertaken to investigate potential improvements in the technology. The approach taken was to deposit solid uranium metal directly on a solid cathode at temperatures below the melting point of uranium (1132 degrees C). The lower temperature electrolytic reduction process has several advantages over the existing chemical reduction processes. It lessens materials problems and special heating and insulating requirements associated with high-temperature operations. It removes most impurities. It does not produce the large quantities of byproduct oxides wastes typical of chemical reduction processes

  12. Study on thermo-oxide layers of uranium-niobium alloy

    International Nuclear Information System (INIS)

    Luo Lizhu; Yang Jiangrong; Zhou Ping

    2010-01-01

    Surface oxides structure of uranium-niobium alloys which were annealed under different temperatures (room temperature, 100, 200, 300 degree C, respectively)in air were studied by X-ray photoelectron spectroscopy (XPS) analysis and depth profile. Thickness of thermo-oxide layers enhance with the increasing oxide temperature, and obvious changes to oxides structure are observed. Under different delt temperatures, Nb 2 O 5 are detected on the initial surface of U-Nb alloys, and a layer of NbO mixed with some NbO x (0 2 O 5 and Nb metal. Dealing samples in air from room temperature to 200 degree C, non-stoichiometric UO 2+x (UO 2 + interstitial oxygen, P-type semiconductor) are found on initial surface of U-Nb alloys, which has 0.7 eV shift to lower binding energy of U 4f 7/2 characteristics comparing to that of UO 2 . Under room temperature, UO 2 are commonly detected in the oxides layer, while under temperature of 100 and 200 degree C, some P-type UO 2+x are found in the oxide layers,which has a satellite at binding energy of 396.6 eV. When annealing at 300 degree C, higher valence oxides, such as U 3 O 8 or UO x (2 5/2 and U 4f 7/2 peaks are 392.2 and 381.8 eV, respectively. UO 2 mixed uranium metal are the main compositions in the oxide layers. From the results, influence of temperature to oxidation of uranium is more visible than to niobium in uranium-niobium alloys. (authors)

  13. A comparative study on determination of composition of uranium thorium mixed oxides by tube and radioisotope excited EDXRF

    International Nuclear Information System (INIS)

    Dhara, Sangita; Sanjay Kumar, S.; Misra, N.L.; Aggarwal, S.K.; Singh, Ajit Kumar; Lodha, G.S.

    2009-01-01

    Energy Dispersive X-ray Fluorescence (EDXRF) methods for determination of uranium and thorium in their mixed oxide matrices using tube and radioisotope excitation sources have been developed. The methodology involves preparation of mixed oxide calibration/sample mixtures of uranium and thorium oxides, mixing of fixed amount of internal standard Yttrium in form of Yttrium oxide, pelletizing these mixtures after thorough mixing and recording their EDXRF spectra using Rh target as well as 109 Cd radioisotope source. The samples were analysed for uranium and thorium on the basis of calibration plots

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

    Science.gov (United States)

    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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

    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 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 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

  16. The creation of a uranium oxide industry, from the laboratory stage to a pilot plant (1961)

    International Nuclear Information System (INIS)

    Caillat, R.; Delange, M.; Sauteron, J.

    1961-01-01

    The qualities of uranium oxide, in particular its good in-pile characteristics and its resistance to corrosion by the usual heat-exchange fluids, have led to this material being chose at the present time as a nuclear fuel in many power reactors, either planned or under construction. A great effort has been made these last few years in France in studying processes for transforming powdered uranium oxide into a dense material with satisfactory behaviour in a neutron flux. The laboratories at Saclay have studied the physico-chemical features of the phenomena accompanying the calcination of uranium peroxide or ammonium uranate to give uranium trioxide, and the subsequent reduction of the latter to dioxide as well as the sintering of the powders obtained. This work has made it possible on one hand to prepare powder of known specific surface area, and on the other to show the overriding influence of this factor, all other things being equal, on the behaviour of powders during sintering in a hydrogen atmosphere. The work has led to defining two methods for sintering stoichiometric uranium oxide of high density. The technological study of the preparation of the powder and its industrial production are carried out at the plant of Le Bouchet which produces at the moment powders of known characteristics suitable for sintering in hydrogen at 1650 deg. C without prior grinding. The industrial sintering is carried out by the Compagnie industrielle des Combustibles Atomiques Frittes who has set up a pilot plant having a capacity of 25 metric tons/year, for the Commissariat l'Energie Atomique and has been operating this plant since May 1958. This plant is presented by a film entitled 'uranium oxide'. (author) [fr

  17. Thermochemistry of the complex oxides of uranium, vanadium, and alkali metals

    International Nuclear Information System (INIS)

    Karyakin, N.V.; Chernorukov, N.G.; Suleimanov, E.V.; Kharyushina, E.A.

    1992-01-01

    The standard enthalpies of the formation at T 298.15 K of complex oxides of uranium(VI), vanadium(V) and alkali metals with the general formula M 1 VUO 6 where M 1 = Na, K, Rb, and Cs, were calculated from the results of calorimetric experiments and from published data. 8 refs., 1 tab

  18. A system for the synthesis of uranium hexafluoride by high pressure fluorination of uranium oxides

    International Nuclear Information System (INIS)

    Elizalde T, J.; Saniger B, J.M.; Nava S, R.

    1986-01-01

    An equipment for the synthesis of uranium hexafluoride by a direct fluorination method is reported. The equipment is composed by a gaseous fluorine supply, a gas burette, a reactor tube inside a protective shield, a soda-lime chemical trap and a vacuum system. The fluorination is accomplished at a pressure of about 70 kg/cm 2 (1000 lb in 2 ), using gaseous fluorine. (Author). 5 refs, 4 figs, 2 tabs

  19. Uranium

    International Nuclear Information System (INIS)

    Perkin, D.J.

    1982-01-01

    Developments in the Australian uranium industry during 1980 are reviewed. Mine production increased markedly to 1841 t U 3 O 8 because of output from the new concentrator at Nabarlek and 1131 t of U 3 O 8 were exported at a nominal value of $37.19/lb. Several new contracts were signed for the sale of yellowcake from Ranger and Nabarlek Mines. Other developments include the decision by the joint venturers in the Olympic Dam Project to sink an exploration shaft and the release of an environmental impact statement for the Honeymoon deposit. Uranium exploration expenditure increased in 1980 and additions were made to Australia's demonstrated economic uranium resources. A world review is included

  20. Uranium

    International Nuclear Information System (INIS)

    Gabelman, J.W.; Chenoweth, W.L.; Ingerson, E.

    1981-01-01

    The uranium production industry is well into its third recession during the nuclear era (since 1945). Exploration is drastically curtailed, and many staffs are being reduced. Historical market price production trends are discussed. A total of 3.07 million acres of land was acquired for exploration; drastic decrease. Surface drilling footage was reduced sharply; an estimated 250 drill rigs were used by the uranium industry during 1980. Land acquisition costs increased 8%. The domestic reserve changes are detailed by cause: exploration, re-evaluation, or production. Two significant discoveries of deposits were made in Mohave County, Arizona. Uranium production during 1980 was 21,850 short tons U 3 O 8 ; an increase of 17% from 1979. Domestic and foreign exploration highlights were given. Major producing areas for the US are San Juan basin, Wyoming basins, Texas coastal plain, Paradox basin, northeastern Washington, Henry Mountains, Utah, central Colorado, and the McDermitt caldera in Nevada and Oregon. 3 figures, 8 tables

  1. Plasma decontamination of uranium oxide from stainless steel surfaces

    International Nuclear Information System (INIS)

    Veilleux, J.M.; El-Genk, M.S.; Chamberlin, E.P.

    1997-01-01

    The U.S. Department of Energy (DOE) is expected to have 845000 m 3 of transuranic (TRU) waste by the year 2000 that has accumulated during the development and assembly of the nation's nuclear stockpile. The TRU disposal costs alone range up to $28000/m 3 , which could be reduced to $1800/m 3 or less by treating and converting the material to low-level waste. Plasma-based processes have been shown to remove plutonium and uranium surface contaminants from metallic components and could be used to treat TRU with significant cost avoidance, estimated at over $1.0 billion. Martz and Hess conducted the initial work of plutonium etching in low-power radio-frequency (rf) plasma with etch rates ranging from 0.007 to 0.025 kg/m 2 ·h. Veilleux et al. reported that plasma decontamination of uranium from the interior of aluminum objects results in etch rates an order of magnitude greater. The current work reports on removal rates of uranium from stainless steel surfaces and includes estimates of the etch rates and characteristic times for removal

  2. Induction of oxidative stress related responses in Arabidopsis thaliana after uranium exposure

    International Nuclear Information System (INIS)

    Vanhoudt, N.

    2009-01-01

    Uranium contamination in the environment has resulted from releases linked with nuclear fuel cycle activities and from industries extracting and processing materials containing naturally occurring radionuclides (for example phosphate industry). Uranium toxicity effects are predominantly studied on man and animal species, but little information is available for plants. If phytomanagement of uranium contaminated soils is considered, biological effects on the vegetation have to be investigated. Information on the contamination impact can also be used for risk assessment and derivation of clean-up standards. Plants can experience oxidative stress when they are exposed to environmental stress situations (for example exposure to heavy metals). Reactive oxygen species (ROS) are produced in both stressed and unstressed cells potentially leading to cellular damage. Consequently, plants have developed an antioxidative defence system comprising ROS-scavenging enzymes (e.g. SOD (superoxide dismutase), CAT (catalase) and metabolites (e.g. ascorbate, glutathione). Previous results showed that uranium exposure can cause an imbalance between the oxidative and antioxidative capacities of the plant cells. The present study aimed to analyse biological effects induced in Arabidopsis thaliana after bioaccumulation of uranium and to define possible dose-effect relationships. Subtle effects on the antioxidative defence system (enzymes, metabolites viewed as early responses for individual disturbances (growth, nutrient profile) were analysed

  3. Uranium metal oxidation, grinding, and encapsulation in BorobondR: TRU waste management - 59279

    International Nuclear Information System (INIS)

    Cook, Kevin S.; Addington, Larry A.; Utley, Beth

    2012-01-01

    Hydrogen generation mitigation for K Basin sludge was examined by encapsulation of uranium metal in BoroBond R , pre-oxidation of uranium metal with Fenton's reagent and grinding of Densalloy SD170, an irradiated uranium metal surrogate. Encapsulation in BoroBond R resulted in pressure increase rates at 60 deg. C ranging from 0.116 torr/h to 0.186 torr/h compared to 0.240 torr/h for a uranium metal in water standard. Samples cast with higher water content led to increased rates. A Fenton's reagent system consisting of a simple reagent mix of FeSO 4 .7H 2 O, H 2 O 2 and HCl effectively oxidized 1/4'' cubes of uranium metal in under four days at room temperature. Increased peroxide addition rate, increased FeSO 4 .7H 2 O concentration and low pH all increase the corrosion rate. Densalloy SD170 with an average particle size of 581 μm with 7.63 % of particles less than 90 μm was milled so that over 90 % of the Densalloy mass measured less than 90 μm in 6 hours of milling. Acceptable wear rates were seen on wear components that were from standard materials (Nitronic SS and 440SS). (authors)

  4. A study on the reduction of uranium oxide to uranium metal in LiCl molten salt

    International Nuclear Information System (INIS)

    Seo, J. S.; Hur, J. M.; Lee, W. K.; Hong, S. S.; Kang, D. S.; Park, S. W.

    2002-01-01

    Research for the analysis on a metallization process of uranium oxide in LiCl-Li molten salt was carried out. Effect of a concentration of Li 2 O on the metallization process was also studied. The new concept, electrochemical reduction of uranium oxide in LiCl-Li 2 O molten salt was proposed. The concept is based on the integrated process of metallization of UO 2 with simultaneous electrochemical reduction of Li 2 O which is recycled in a closed system. In a LiCl-Li molten salt system, U 3 O 8 whose conversion ratio to U turns out to be 97.1%, showed a better metallization characteristic than UO 2 . It is verified that electrochemically reduced Li is well deposited on the UO 2 powder cathode through a porous magnesia filter in LiCl-Li 2 O molten salt. In that process Li 2 O was from by the reduction process of UO 2 to U. This electrochemical reduction process showed good results to covert UO 2 to U

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

    International Nuclear Information System (INIS)

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

    2003-01-01

    Lawrence Livermore National Laboratory (LLNL) has completed an extensive effort to calibrate the LLNL passive-active neutron drum (PAN) shuffler (Canberra Model JCC-92) for accountability measurement of highly enriched uranium (HEU) oxide and HEU in mixed uranium-plutonium (U-Pu) oxide. Earlier papers described the PAN shuffler calibration over a range of item properties by standards measurements and an extensive series of detailed simulation calculations. With a single normalization factor, the simulations agree with the HEU oxide standards measurements to within ±1.2% at one standard deviation. Measurement errors on mixed U-Pu oxide samples are in the ±2% to ±10% range, or ±20 g for the smaller items. The purpose of this paper is to facilitate transfer of the LLNL procedure and calibration algorithms to external users who possess an identical, or equivalent, PAN shuffler. Steps include (1) measurement of HEU standards or working reference materials (WRMs); (2) MCNP simulation calculations for the standards or WRMs and a range of possible masses in the same containers; (3) a normalization of the calibration algorithms using the standard or WRM measurements to account for differences in the 252 Cf source strength, the delayed-neutron nuclear data, effects of the irradiation protocol, and detector efficiency; and (4) a verification of the simulation series trends against like LLNL results. Tools include EXCEL/Visual Basic programs which pre- and post-process the simulations, control the normalization, and embody the calibration algorithms

  6. Impact Of Low Grade Uranium Ores On The Echo System and the Workers of Phosphate Industry

    International Nuclear Information System (INIS)

    Anwar, S.M.

    1999-01-01

    The present study aims to investigate the influence of uranium present in phosphate rocks as an environmental factor in the ccho system and on the workers of Abu-Zaabal Phosphate Company subjected to the inhalation of big quantities of rock phosphate dust during the benefication of the ore and the production of the fertilizers. Besides. extra amount of uranium reach the workers also through two path ways.The first is direct through eating contaminated planted grown in the near by area.The second is indirect through eating animals fed with contaminated plants. The uranium content is estimated in the soil samples at different depths, water (irrigation and drainage), air samples and plant samples (shoot and root) in Berseem from the four directions, urine samples from twenty workers in charge of the processing of phosphate compared to twenty volunteers far from the contaminated area.The results showed an elevated values for phosphorus and uranium in the air, water. soil and plant (Berseem) around Abu Zaabal Factory and extending to about 2 km from all directions. Urine may be considered as a biological indicator medium for the uptake of uranium in uranium miners and the workers in charge of ore processing and can represent the major route of excretion for the absorbed metal. Significant differences were shown between the uranium level in the urine of workers group and the control group

  7. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    Recent decisions by the Australian Government will ensure a significant expansion of the uranium industry. Development at Roxby Downs may proceed and Ranger may fulfil two new contracts but the decision specifies that apart from Roxby Downs, no new mines should be approved. The ACTU maintains an anti-uranium policy but reaction to the decision from the trade union movement has been muted. The Australian Science and Technology Council (ASTEC) has been asked by the Government to conduct an inquiry into a number of issues relating to Australia's role in the nuclear fuel cycle. The inquiry will examine in particular Australia's nuclear safeguards arrangements and the adequacy of existing waste management technology. In two additional decisions the Government has dissociated itself from a study into the feasibility of establishing an enrichment operation and has abolished the Uranium Advisory Council. Although Australian reserves account for 20% of the total in the Western World, Australia accounts for a relatively minor proportion of the world's uranium production

  8. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The French Government has decided to freeze a substantial part of its nuclear power programme. Work has been halted on 18 reactors. This power programme is discussed, as well as the effect it has on the supply of uranium by South Africa

  9. Bioleaching of low grade uranium ore containing pyrite using A. ferrooxidans and A. thiooxidans

    International Nuclear Information System (INIS)

    Alexey Borisovich Umanskii; Anton Mihaylovich Klyushnikov

    2013-01-01

    A process of uranium extraction from ore containing 3.1 % pyrite by bacterial leaching was investigated in shaken flasks during 90 days. The highest uranium recovery amounting to 85.1 % was obtained using binary mixture of Acidithiobacillus ferrooxidans and Acidithiobacillus thiooxidans that was exceeding results obtained by traditional acid leaching technique up to 27 %. High uranium recovery was founded to be due to the high degree of pyrite dissolution that can be readily achieved by bacterial leaching (up to 98.0 %). (author)

  10. Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids - Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method

    International Nuclear Information System (INIS)

    2004-01-01

    This first edition of ISO 7097-1 together with ISO 7097-2:2004 cancels and replaces ISO 7097:1983, which has been technically revised, and ISO 9989:1996. ISO 7097 consists of the following parts, under the general title Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids: Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method; Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method. This part 2. of ISO 7097 describes procedures for determination of uranium in solutions, uranium hexafluoride and solids. The procedures described in the two independent parts of this International Standard are similar: this part uses a titration with cerium(IV) and ISO 7097-1 uses a titration with potassium dichromate

  11. Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids - Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method

    International Nuclear Information System (INIS)

    2004-01-01

    This first edition of ISO 7097-1 together with ISO 7097-2:2004 cancels and replaces ISO 7097:1983, which has been technically revised, and ISO 9989:1996. ISO 7097 consists of the following parts, under the general title Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids: Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method; Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method. This part 1. of ISO 7097 describes procedures for the determination of uranium in solutions, uranium hexafluoride and solids. The procedures described in the two independent parts of this International Standard are similar: this part uses a titration with potassium dichromate and ISO 7097-2 uses a titration with cerium(IV)

  12. Oxidation of 1-butene over uranium oxide (UO3)-antimony oxide (Sb2O3) catalysts

    NARCIS (Netherlands)

    Simons, T.; Houtman, P.N.; Schuit, G.C.A.

    1971-01-01

    The oxidative dehydrogenation of butene to butadiene over U-Sb catalysts was investigated. The presence of two compds., (UO2)Sb3O7 and Sb3U3O14, reported by Grasselli and Callahan (1969), was confirmed with (UO2)Sb3O7 being the actual catalyst. The reaction is first order in butene and zero order in

  13. Process for recovery of plutonium from fabrication residues of mixed fuels consisting of uranium oxide and plutonium oxide

    International Nuclear Information System (INIS)

    Heremanns, R.H.; Vandersteene, J.J.

    1983-01-01

    The invention concerns a process for recovery of plutonium from fabrication residues of mixed fuels consisting of uranium oxide and plutonium oxide in the form of PuO 2 . Mixed fuels consisting of uranium oxide and plutonium oxide are being used more and more. The plants which prepare these mixed fuels have around 5% of the total mass of fuels as fabrication residue, either as waste or scrap. In view of the high cost of plutonium, it has been attempted to recover this plutonium from the fabrication residues by a process having a purchase price lower than the price of plutonium. The problem is essentially to separate the plutonium, the uranium and the impurities. The residues are fluorinated, the UF 6 and PuF 6 obtained are separated by selective absorption of the PuF 6 on NaF at a temperature of at least 400 0 C, the complex obtained by this absorption is dissolved in nitric acid solution, the plutonium is precipitated in the form of plutonium oxalate by adding oxalic acid, and the precipitated plutonium oxalate is calcined

  14. High-temperature, Knudsen cell-mass spectroscopic studies on lanthanum oxide/uranium dioxide solid solutions

    International Nuclear Information System (INIS)

    Sunder, S.; McEachern, R.; LeBlanc, J.C.

    2001-01-01

    Knudsen cell-mass spectroscopic experiments were carried out with lanthanum oxide/uranium oxide solid solutions (1%, 2% and 5% (metal at.% basis)) to assess the volatilization characteristics of rare earths present in irradiated nuclear fuel. The oxidation state of each sample used was conditioned to the 'uranium dioxide stage' by heating in the Knudsen cell under an atmosphere of 10% CO 2 in CO. The mass spectra were analyzed to obtain the vapour pressures of the lanthanum and uranium species. It was found that the vapour pressure of lanthanum oxide follows Henry's law, i.e., its value is directly proportional to its concentration in the solid phase. Also, the vapour pressure of lanthanum oxide over the solid solution, after correction for its concentration in the solid phase, is similar to that of uranium dioxide. (authors)

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

    Science.gov (United States)

    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.

  16. Formation conditions and prospecting criteria for sandstone uranium deposit of interlayer oxidation type

    International Nuclear Information System (INIS)

    Huang Shijie

    1994-01-01

    This paper comprehensively analyses the geotectonic setting and favourable conditions, such as structure of the basin, sedimentary facies and paleogeography, geomorphology and climate, hydrodynamics and hydrogeochemistry, the development of interlayered oxidation etc, necessary for the formation of sandstone uranium deposit of interlayered oxidation type. The following prospecting criteria is proposed, namely: abundant uranium source, arid climate, stable big basin, flat-lying sandstone bed, big alluvial fan, little change in sedimentary facies, intercalation of sandstone and mudstone beds, shallow burying of sandstone bed, well-aquiferous sandstone bed, high permeability of sandstone bed, development of interlayered oxidation, and high content of reductant in sandstone. In addition, the 6 in 1 hydrogenic genetic model is proposed

  17. Uranium oxide nanocrystals by microwave-assisted thermal decomposition. Electronic and structural properties

    International Nuclear Information System (INIS)

    Leduc, Jennifer; Mathur, Sanjay; Pacold, Joseph I.; Shuh, David K.; Dong, Chung-Li

    2018-01-01

    Uranium oxides have attracted much attention not only in the context of nuclear energy generation but also for their application as pristine catalysts or as supports for other (transition metal) oxides and (precious) metals. Their propensity to adopt high coordination numbers and manifest multiple oxidation states (from +II to +VI) makes them attractive candidates for catalyzed transformation reactions. Herein, we report a new synthesis route to phase-pure, crystalline UO 2 nanoparticles via microwave-assisted decomposition of a molecular uranium(IV) precursor. The electronic structure and optical absorption properties of these nanocrystals were investigated using spectroscopic methods to evaluate their suitability for photo(electro)catalytic applications. (copyright 2018 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  18. Uranium oxide nanocrystals by microwave-assisted thermal decomposition. Electronic and structural properties

    Energy Technology Data Exchange (ETDEWEB)

    Leduc, Jennifer; Mathur, Sanjay [Institute of Inorganic Chemistry, University of Cologne (Germany); Pacold, Joseph I.; Shuh, David K. [Chemical Sciences Division, The Glenn T. Seaborg Center, Lawrence Berkeley National Laboratory, Berkeley, CA (United States); Dong, Chung-Li [Department of Physics, Tamkang University, Tamsui, Taiwan (China)

    2018-01-17

    Uranium oxides have attracted much attention not only in the context of nuclear energy generation but also for their application as pristine catalysts or as supports for other (transition metal) oxides and (precious) metals. Their propensity to adopt high coordination numbers and manifest multiple oxidation states (from +II to +VI) makes them attractive candidates for catalyzed transformation reactions. Herein, we report a new synthesis route to phase-pure, crystalline UO{sub 2} nanoparticles via microwave-assisted decomposition of a molecular uranium(IV) precursor. The electronic structure and optical absorption properties of these nanocrystals were investigated using spectroscopic methods to evaluate their suitability for photo(electro)catalytic applications. (copyright 2018 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  19. First principle active neutron coincidence counting measurements of uranium oxide

    Energy Technology Data Exchange (ETDEWEB)

    Goddard, Braden, E-mail: goddard.braden@gmail.com [Nuclear Security Science and Policy Institute, Texas A and M University, College Station, Texas 77843 (United States); Charlton, William [Nuclear Security Science and Policy Institute, Texas A and M University, College Station, Texas 77843 (United States); Peerani, Paolo [European Commission, EC-JRC-ITU, Ispra (Italy)

    2014-03-01

    Uranium is present in most nuclear fuel cycle facilities ranging from uranium mines, enrichment plants, fuel fabrication facilities, nuclear reactors, and reprocessing plants. The isotopic, chemical, and geometric composition of uranium can vary significantly between these facilities, depending on the application and type of facility. Examples of this variation are: enrichments varying from depleted (∼0.2 wt% {sup 235}U) to high enriched (>20 wt% {sup 235}U); compositions consisting of U{sub 3}O{sub 8}, UO{sub 2}, UF{sub 6}, metallic, and ceramic forms; geometries ranging from plates, cans, and rods; and masses which can range from a 500 kg fuel assembly down to a few grams fuel pellet. Since {sup 235}U is a fissile material, it is routinely safeguarded in these facilities. Current techniques for quantifying the {sup 235}U mass in a sample include neutron coincidence counting. One of the main disadvantages of this technique is that it requires a known standard of representative geometry and composition for calibration, which opens up a pathway for potential erroneous declarations by the State and reduces the effectiveness of safeguards. In order to address this weakness, the authors have developed a neutron coincidence counting technique which uses the first principle point-model developed by Boehnel instead of the “known standard” method. This technique was primarily tested through simulations of 1000 g U{sub 3}O{sub 8} samples using the Monte Carlo N-Particle eXtended (MCNPX) code. The results of these simulations showed good agreement between the simulated and exact {sup 235}U sample masses.

  20. Status Report from Australia [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, J R [Australian Atomic Energy Commission, Coogee, NSW (Australia)

    1967-06-15

    During the period 1954-59, five plants for the production of uranium chemical concentrates (yellowcake) were constructed in Australia. Only one of these is in operation at the present time, but two more have been kept on a 'care-and-maintenance' basis pending revival of the market for uranium. The other two were disbanded after completion of contract production. Brief details of each of the plants are given.

  1. Radiometric determination in situ of the face grades in Witwatersrand gold and uranium mines

    International Nuclear Information System (INIS)

    Smit, C.J.B.

    1985-01-01

    A prototype collimated radiometric face scanner was tested in the Harmony Gold Mine. The results obtained during the pilot study indicate that in situ radiometric uranium assays are statistically indistinguishable from those obtained conventionally from channel chip samples. In addition, the study demonstrated that reasonably reliable gold estimates can be deduced from the radiometric measurements, by use of the ratio of gold to uranium within a mine. The instrumentation, calibration procedures, and background determination are described briefly

  2. Uranium exploration target selection for proterozoic iron oxide/breccia complex type deposits in India

    International Nuclear Information System (INIS)

    Dwivedy, K.K.; Sinha, K.K.

    1997-01-01

    Multimetal iron oxide/breccia complex (IOBC) type deposits exemplified by Olympic Dam in Australia, fall under low grade, large tonnage deposits. A multidisciplinary integrated exploration programme consisting of airborne surveys, ground geological surveys, geophysical and geochemical investigations and exploratory drilling, supported adequately by the state of the art analytical facilities, data processing using various software and digital image processing has shown moderate success in the identification of target areas for this type of deposits in the Proterozoic terrains of India. Intracratonic, anorogenic, continental rift to continental margin environment have been identified in a very wide spectrum of rock associations. The genesis and evolution of such associations during the Middle Proterozoic period have been reviewed and applied for target selection in the (i) Son-Narmada rift valley zone; (ii) areas covered by Dongargarh Supergroup of rocks in Madhya Pradesh; (iii) areas exposing ferruginous breccia in the western part of the Singhbhum Shear Zone (SSZ) around Lotapahar; (iv) Siang Group of rocks in Arunachal Pradesh; (v) Crystalline rocks of Garo Hills around Anek; and (vi) Chhotanagpur Gneissic complex in the Bahia-Ulatutoli tract of Ranchi Plateau. Of theses six areas, the Son-Narmada rift area appears to be the most promising area for IOBC type deposits. Considering occurrences of the uranium anomalies near Meraraich, Kundabhati, Naktu and Kudar and positive favourability criteria observed in a wide variety of rocks spatially related to the rifts and shears, certain sectors in Son-Narmada rift zone have been identified as promising for intense subsurface exploration. 20 refs, 4 figs, 1 tab

  3. Comparative study of the oxidation of various qualities of uranium in carbon dioxide at high temperatures; Etude comparative de l'oxydation de diverses qualites d'uranium dans l'anhydride carbonique aux temperatures elevees

    Energy Technology Data Exchange (ETDEWEB)

    Desrues, R; Paidassi, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    Uranium samples of six different qualities were subjected, in the temperature range 400 - 1000 C, to the action of carbon dioxide carefully purified to eliminate oxygen and water vapour; the resulting oxidation was followed micro-graphically and also (but only in the range 400 - 700 C) gravimetrically using an Ugine-Eyraud microbalance. A comparison of the results leads to the following 3 observations. First, the oxidation of the six uraniums studied obeys a linear law, (followed at 700 C by an accelerating law). The rates of reaction differ by a maximum of 100 per cent, the higher purity grades being oxidized more slowly except at 700 C when the reverse is true. Secondly, simultaneously with the growth, of an approximately uniform film of uranium dioxide on the metal, there occurs a localized attack in the form of blisters in the immediate neighbourhood of the monocarbide inclusions in the uranium. The relative importance of this attack is greater for lower oxidation temperatures and for a larger size, number and inequality of distribution of the inclusions, that is to say for higher carbon concentrations in the uranium (which have values from 7 to 1000 ppm in our tests). Thirdly, for oxidation temperatures above 600 C blistering is much less pronounced, but at 700 C the beginning of a general deformation of the sample occurs, which, above 750 C, becomes much greater; this leads to an acceleration of the reaction rate with respect to the linear law. In view of the over-heating, the sample must already be in the {gamma}-phase which is particularly easily deformed; furthermore this expansion phenomenon is more pronounced when the sample is more plastic and therefore purer. (authors) [French] Des echantillons de six qualites d'uranium ont ete soumis, dans l'intervalle 400-1000 C, a l'action de l'anhydride carbonique tres soigneusement purifie en oxygene et en vapeur d'eau, et leur oxydation a ete suivie par voie micrographique et egalement (mais seulement entre 400

  4. Contribution to the study of the sintering of uranium oxide; Contribution a l'etude du frittage de l'oxyde d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Bel, A; Carteret, Y [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The sintering ofnium oxide has been considered and the following factors have been particularly taken in consideration: - the particle size and the particles in shape of the initial powder, - the specific area of the initial powder, - the chemical composition of the oxide, - and the medium in which the sintering was carried out. A method of sintering uranium oxide on semi-industrial scale is presented. (author)Fren. [French] On xamine l'influence de differents facteurs sur le frittage de l'oxyde d'uranium. Sont particulierement prises en consideration: - la taille et la forme des grains de la poudre initiale, - la surface specifique de la poudre initiale, - la composition chimique de l'oxyde, - ainsi que la nature de l'atmosphere durant le frittage. D'autre part, une technique de frittage de l'oxyde d'uranium a l'echelle semi-industrielle est presentee. (auteur)

  5. Study of the sulfuric acid leaching and bacterial leaching of low grade uranium ore by orbital shaker experiment

    International Nuclear Information System (INIS)

    Li Guangyue; Liu Yulong; Wang Yongdong; Ding Dexin

    2009-01-01

    The sulphuric acid leaching and bacteria leaching by orbital shaker experiments were conducted for the low grade uranium ore from a uranium mine in Guangdong Province. The results show that, when the concentration of sulphuric acid and that of slurry were 30 g/L and 25%, respectively, the conditions were most favourable for sulphuric acid leaching and the rate of leaching reached 92.92%, that, when pH value was 1.5, inoculation amount, 10%, concentration of slurry, 10%, the conditions were most favourable for bacteria leaching and the rate of leaching reached 95.93%, that, compared with sulphuric acid leaching, bacteria leaching decreased sulphuric acid consumption by 17.2% and increased the rate of leaching by 3%, and that, under the most favourable conditions for suphuric acid leaching, if 1% natrium chlorate was added, the rate of leaching increased to 96.46%, but 10 kg of natrium chlorate was consumed for 1 ton of uranium ore. (authors)

  6. Inhalation toxicology of industrial plutonium and uranium oxide aerosols I. Physical chemical characterization

    International Nuclear Information System (INIS)

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

    1978-01-01

    In the fabrication of mixed plutonium and uranium oxide fuel, large quantities of dry powders are processed, causing dusty conditions in glove box enclosures. Inadvertent loss of glove box integrity or failure of air filter systems can lead to human inhalation exposure. Powdered samples and aerosol samples of these materials obtained during two fuel fabrication process steps have been obtained. A regimen of physical chemical tests of properties of these materials has been employed to identify physical chemical properties which may influence their biological behavior and dosimetry. Materials to be discussed are 750 deg. C heat-treated, mixed uranium and plutonium oxides obtained from the ball milling operation and 1750 deg. C heat-treated, mixed uranium and plutonium oxides obtained from the centerless grinding of fuel pellets. Results of x-ray diffraction studies have shown that the powder generated by the centerless grinding of fuel pellets is best described as a solid solution of UO x and PuO x consistent with its temperature history. In vitro dissolution studies of both mixed oxide materials indicate a generally similar dissolution rate for both materials. In one solvent, the material with the higher temperature history dissolves more rapidly. The x-ray diffraction and in vitro dissolution results as well as preliminary results of x-ray photoelectron spectroscopic analyses will be compared and the implications for the associated biological studies will be discussed. (author)

  7. Studies on conversion of crude sodium diuranate to high purity uranium oxide by chemical route

    International Nuclear Information System (INIS)

    Rajan, K.C.; Ramadevi, G.; Giri, Nitai; Chakravartty, J.K.; Sreenivas, T.

    2016-01-01

    Uranium values are recovered from ores and mine rejects by employing chemical processing methods. The process chemistry could be either acidic (H 2 SO 4 medium) or alkaline (Na 2 CO 3 - NaHCO 3 medium) based depending on the nature of host rock. The end product is termed as 'yellow cake' which is sent to refining plants for final purification and fuel manufacture. Conventionally the yellow cake product is magnesium (MDU) or ammonium diuranate (ADU) in U-plants following the acidic processing route while it is sodium diuranate (SDU) in the alkaline route. The preference for SDU in plants operating with alkaline route is mainly driven by the requirement of conserving and reusing the expensive chemicals. Though alkaline processing gives relatively purer diuranate product, the practice of direct precipitation of dissolved uranium from pregnant leach liquors invariably leads to presence of impurities detrimental for downstream unit operations, mostly in solvent extraction based refining stages. In this score uranium peroxide is superior to other uranates in many respects. Two alternatives are practiced for the conversion of SDU to uranium peroxide - chemical and fluid bed precipitation. Though generic flowsheet for the conversion stage is well documented, the process conditions and complexities vary with the chemistry of the pregnant leach liquor which is specific for a given ore or a deposit. This paper discusses the process scheme tailored, optimized and tested on a kilo-gram scale for chemical conversion of SDU produced by alkaline processing of a low-grade carbonate hosted uranium ore. The entire process scheme is first of its kind in Indian uranium ore processing arena

  8. Does uranium exposure induce oxidative stress and genotoxicity in the teleostean Danio rerio? first experimental results

    International Nuclear Information System (INIS)

    Barillet, S.; Devaux, A.; Simon, O.; Buet, A.; Pradines, C.

    2004-01-01

    Within the Envirhom research program, key advances have been obtained in uranium bioaccumulation and underlying mechanisms understanding in various biological models at the individual level. However, considering different scales of biological effects (from early to delayed ones, from low to high level of organization) is crucial to provide ecologically relevant indicators. Organisms counteract stress induced by pollutant exposure through a wide range of physiological responses being both dose and time dependent. Effects at higher hierarchical levels are always preceded by early changes in biological processes, from subtle biochemical disturbances to impaired physiological functions, increased susceptibility to other stresses, reduced life-span Within this global context, preliminary experiments were carried out on adult zebra fish (Danio rerio), to assess early changes after short-term uranium exposure. Among the subsequent primary subcellular damages oxidative stress and genotoxicity (characterizing both chemo-toxicity and radiotoxicity) are relevant endpoints, thus requiring the knowledge of dose-effects relationships as a first operational approach to provide useful tool in predicting possible effects of U exposure. Zebra fish has been selected due to its small size (facilitating its maintenance) and its extended use in eco-toxicological studies. Moreover, its short life-cycle will allow to carry out chronic exposure experiments (along the whole life-cycle). Four uranium concentrations (0, 20, 100 and 500μg.L -1 ) and five sampling times (0, 0.5, 1, 5 and 10 days) were selected. Catalase, glutathione peroxidase (GPx) and superoxide dismutase (SOD) activities were measured as oxidative stress bio-markers. DNA damage level was assessed in zebra fish erythrocytes using the comet assay. Uranium bioaccumulation was concurrently studied to understand observed bio-marker responses. Further experiments, dedicated to the assessment of the impact of chronic uranium

  9. Solvent extraction of uranium from lean grade acidic sulfate leach liquor with alamine 336 reagent

    International Nuclear Information System (INIS)

    Ramadevi, G.; Sreenivas, T.; Navale, A.S.; Padmanabhan, N.P.H.

    2012-01-01

    This paper describes the solvent extraction studies carried out on an acidic low assay uranium bearing leach liquor generated during sulfuric acid leaching of a refractory uranium ore using alamine 336-isodecenol-kerosene reagent combine. The leach liquor has a U 3 O 8 content of about 270 mg/L, free acidity 2.4 N H 2 SO 4 and total dissolved solids concentration of 260 g/L. Process parameteric variation studies indicated strong influence of free acidity of the leach liquor, alamine 336 concentration and aqueous to organic phase ratio on the extraction efficiency of uranium. An extraction efficiency of about 95% was achieved when the free acidity of leach liquor was 1 N H 2 SO 4 or lower, using 2% (v/v) alamine 336 at ambient temperature with an aqueous to organic phase ratio of 1:1. The loading capacity under these conditions was 1.2 g/L of U 3 O 8 . About 98% of the uranium values could be stripped from the loaded organic using 1 N NaCl in 0.2 N H 2 SO 4 . The solvent extraction studies aided in developing a suitable process flowsheet for treating refractory uranium ores which need high acidity during leaching and relatively lower acidity for purification by solvent extraction. (author)

  10. Ianthinite: A rare hydrous uranium oxide mineral from Akkavaram ...

    Indian Academy of Sciences (India)

    In the area of ianthinite occurrence (figure 1), the basement ... Akkavaram area, from where ianthinite is reported. (figure 1), falls in ... feldspar grains from separated yellowish material .... oxidizing environment and its relevance to the disposal.

  11. Dissolution of synthetic uranium dibutyl phosphate deposits in oxidizing and reducing chemical formulations

    International Nuclear Information System (INIS)

    Rufus, A.L.; Sathyaseelan, V.S.; Narasimhan, S.V.; Velmurugan, S.

    2013-01-01

    Graphical abstract: SEM of the U-DBP coated stainless steel coupon before and after exposure to chemical formulation containing acid permanganate at 80 °C. -- Highlights: •Combination of oxidation and reduction processes efficiently dissolves U-DBP deposits. •NP and NAC formulations are compatible with SS-304. •Dissolved uranium and added chemicals are effectively removed via ion exchangers. -- Abstract: Permanganate and nitrilotriacetic acid (NTA) based dilute chemical formulations were evaluated for the dissolution of uranium dibutyl phosphate (U-DBP), a compound that deposits over the surfaces of nuclear reprocessing plants and waste storage tanks. A combination of an acidic, oxidizing treatment (nitric acid with permanganate) followed by reducing treatment (NTA based formulation) efficiently dissolved the U-DBP deposits. The dissolution isotherm of U-DBP in its as precipitated form followed a logarithmic fit. The same chemical treatment was also effective in dissolving U-DBP coated on the surface of 304-stainless steel, while resulting in minimal corrosion of the stainless steel substrate material. Investigation of uranium recovery from the resulting decontamination solutions by ion exchange with a bed of mixed anion and cation resins showed quantitative removal of uranium

  12. Dissolution of synthetic uranium dibutyl phosphate deposits in oxidizing and reducing chemical formulations

    Energy Technology Data Exchange (ETDEWEB)

    Rufus, A.L.; Sathyaseelan, V.S.; Narasimhan, S.V.; Velmurugan, S., E-mail: svelu@igcar.gov.in

    2013-06-15

    Graphical abstract: SEM of the U-DBP coated stainless steel coupon before and after exposure to chemical formulation containing acid permanganate at 80 °C. -- Highlights: •Combination of oxidation and reduction processes efficiently dissolves U-DBP deposits. •NP and NAC formulations are compatible with SS-304. •Dissolved uranium and added chemicals are effectively removed via ion exchangers. -- Abstract: Permanganate and nitrilotriacetic acid (NTA) based dilute chemical formulations were evaluated for the dissolution of uranium dibutyl phosphate (U-DBP), a compound that deposits over the surfaces of nuclear reprocessing plants and waste storage tanks. A combination of an acidic, oxidizing treatment (nitric acid with permanganate) followed by reducing treatment (NTA based formulation) efficiently dissolved the U-DBP deposits. The dissolution isotherm of U-DBP in its as precipitated form followed a logarithmic fit. The same chemical treatment was also effective in dissolving U-DBP coated on the surface of 304-stainless steel, while resulting in minimal corrosion of the stainless steel substrate material. Investigation of uranium recovery from the resulting decontamination solutions by ion exchange with a bed of mixed anion and cation resins showed quantitative removal of uranium.

  13. Identification and quantitative grade estimation of Uranium mineralization based on gross-count gamma ray log at Lemajung sector West Kalimantan

    International Nuclear Information System (INIS)

    Adi Gunawan Muhammad

    2014-01-01

    Lemajung sector, is one of uranium potential sector in Kalan Area, West Kalimantan. Uranium mineralization is found in metasiltstone and schistose metapelite rock with general direction of mineralization east - west tilted ± 70° to the north parallel with schistocity pattern (S1). Drilling evaluation has been implemented in 2013 in Lemajung sector at R-05 (LEML-(S1). Drilling evaluation has been implemented in 2013 in Lemajung sector at R-05 (LEML-gamma ray. The purpose of this activity is to determine uranium mineralization grade with quantitatively methode in the rocks and also determine the geological conditions in sorounding of drilling area. The methodology involves determining the value of k-factor, geological mapping for the sorounding of drill hole, determination of the thickness and grade estimation of uranium mineralization with gross-count gamma ray. Quantitatively from grade estimation of uranium using gross-count gamma ray log can be known that the highest % eU_3O_8 in the hole R-05 (LEML-40) reaches 0.7493≈6354 ppm eU found at depth interval from 30.1 to 34.96 m. Uranium mineralization is present as fracture filling (vein) or tectonic breccia matrix filling in metasiltstone with thickness from 0.10 to 2.40 m associated with sulphide (pyrite) and characterized by high ratio of U/Th. (author)

  14. Sodium oxide and uranium oxide aerosol experiments: NSPP Tests 106-108 and Tests 204-207, data record report

    Energy Technology Data Exchange (ETDEWEB)

    Adams, R.E.; Kress, T.S.; Tobias, M.L.

    1981-03-01

    This data record report describes three sodium oxide aerosol tests and four uranium oxide aerosol tests conducted in the Nuclear Safety Pilot Plant project at Oak Ridge National Laboratory. The goal of this project is to establish the validity (or level of conservatism) of the aerosol behavioral code, HAARM-3, and follow-on codes under development at the Battelle Columbus Laboratories for the US Nuclear Regulatory Commission. Descriptions of the seven tests with tables and graphs summarizing the results are included. 92 figs.

  15. A study of geochemical prospecting for uranium-bearing low grade coal beds in Korea

    International Nuclear Information System (INIS)

    Kim, O.B.

    1980-01-01

    Trend surface analysis was applied in order to find the criteria for geochemcial prospecting of uranium bearing narrow coal bed in Ogcheon Group. Soil samples were taken from the Mogso-ri area, the Deogpyeong-ri area, and the Jeogum-ri area and were analyzed for U, V, Mo, Pb, Zn, Cu, Cd, and Cr by colorimetry and atomic absorption. All data were processed statistically by HP 3000 computer. The results were as follows: Molybdenium could be used as the best competent indicator element for uranium. Lead, Copper, Vanadium could be used as assistant indicator. The trend surface analysis and the residual map were very useful for statistical interpretation of analyzed data. Second or third degree trend surface analysis was sufficient for this work. The trend map revealed that the origin of uranium in these area was the same. (Author)

  16. Status report from South Africa [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, R E [Atomic Energy Board, Pretoria (South Africa)

    1967-06-15

    Most of the research work on the processing of uranium ores in South Africa is being conducted by the Extraction Metallurgy Division of the S.A. Atomic Energy Board. Nevertheless, a considerable amount of applied research has been done by the different mining groups concerned with the operation of uranium plants, and also by the Transvaal and Orange Free State Chamber of Mines research laboratories. There is, however, very close collaboration between the various research groups and the Atomic Energy Board and the main research described is conducted on a collaborative basis.

  17. Comparison of two lung clearance models based on the dissolution rates of oxidized depleted uranium

    International Nuclear Information System (INIS)

    Crist, K.C.

    1984-10-01

    An in-vitro dissolution study was conducted on two respirable oxidized depleted uranium samples. The dissolution rates generated from this study were then utilized in the International Commission on Radiological Protection Task Group lung clearance model and a lung clearance model proposed by Cuddihy. Predictions from both models based on the dissolution rates of the amount of oxidized depleted uranium that would be cleared to blood from the pulmonary region following an inhalation exposure were compared. It was found that the predictions made by both models differed considerably. The difference between the predictions was attributed to the differences in the way each model perceives the clearance from the pulmonary region. 33 references, 11 figures, 9 tables

  18. Study about uranium oxides at high temperature by X-ray diffraction

    International Nuclear Information System (INIS)

    Costa, M.I.

    1978-01-01

    In this work a technique to study the lattice parameters in the crystalline substances at hight temperature by X-rays diffraction is developed. The results obtained agree very well with the experimental data found in the literature. The crystalline structure of uranium oxide at different temperature is studied in detail by this technique. At the range of the temperature investigated, i.e., 20 0 C to 640 0 C, the following forms for uranium oxide: U 3 O 8 in its hexagonal modification, cubic UO 2 , cubic U 4 O 9 and tetragonal U 3 O 7 is observed. The appearance of two hexagonal units observed in this work is identified by Milne. A good reproducibillity is observed for measurements at the same temperature [pt

  19. Study on metallogenetic prospect of interlayer oxidation zone sandstone type uranium deposit in Shanganning basin

    International Nuclear Information System (INIS)

    Wang Jinping

    1998-01-01

    As Compared with orogenic zone basin, which the interlayer oxidation zone sandstone type uranium deposits are found, the Shanganning basin a continental platform type basin is distinct either in the geodynamic background and the post-basin hydrogeological evolution or in the appearance of the metallogenetic dynamics-orogenesis. The prediction criteria summarized for interlayer oxidation zone type U-deposits in Middle Asia therefore can not be completely applied in such a basin. Based on analysis of the typical regional geological setting, the hydrogeology of the Meso-Cenozoic cover is studied in detail. Three hydrogeological cycles have been divided, and prospects of uranium deposits have been clarified and the most promising target have been proposed

  20. Comparison of two lung clearance models based on the dissolution rates of oxidized depleted uranium

    Energy Technology Data Exchange (ETDEWEB)

    Crist, K.C.

    1984-10-01

    An in-vitro dissolution study was conducted on two respirable oxidized depleted uranium samples. The dissolution rates generated from this study were then utilized in the International Commission on Radiological Protection Task Group lung clearance model and a lung clearance model proposed by Cuddihy. Predictions from both models based on the dissolution rates of the amount of oxidized depleted uranium that would be cleared to blood from the pulmonary region following an inhalation exposure were compared. It was found that the predictions made by both models differed considerably. The difference between the predictions was attributed to the differences in the way each model perceives the clearance from the pulmonary region. 33 references, 11 figures, 9 tables.

  1. Uranium and main oxides in soil in the Northeast part of Parana basil

    International Nuclear Information System (INIS)

    Fagundes, I.C.; Bonotto, D.M.; Jimenez-Rueda, J.R.

    2015-01-01

    Uranium is a litophile element that migrates to crust together with other light silicates. This work evaluated the relationships among the concentrations of uranium, major oxides (SiO_2, Al_2O_3, Na_2O, K_2O, CaO, MgO, Fe_2O_3, MnO, TiO_2, P_2O_5) and organic matter in different horizons of a soil profile located over siltstone from Tatui Formation, Piracicaba River sub-basin, Sao Paulo State, Brazil. Different techniques were utilized for data acquisition, for instance, alpha spectrometry, X-ray fluorescence spectrometry and smectrophotometry. The major U concentrations were found in horizons enriched in Fe_2O_3, indicating its tending to be retained in iron oxides. (author)

  2. Reaction of Antimony-Uranium Composite Oxide in the Chlorination Treatment of Waste Catalyst - 13521

    Energy Technology Data Exchange (ETDEWEB)

    Sawada, Kayo [EcoTopia Science Institute (Japan); Hirabayashi, Daisuke; Enokida, Youichi [Department of Materials, Physics and Energy Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya, Aichi 464-8603 (Japan)

    2013-07-01

    The effect of oxygen gas concentration on the chlorination treatment of antimony-uranium composite oxide catalyst waste was investigated by adding different concentrations of oxygen at 0-6 vol% to its chlorination agent of 0.6 or 6 vol% hydrogen chloride gas at 1173 K. The addition of oxygen tended to prevent the chlorination of antimony in the oxide. When 6 vol% hydrogen chloride gas was used, the addition of oxygen up to 0.1 vol% could convert the uranium contained in the catalyst to U{sub 3}O{sub 8} without any significant decrease in the reaction rate compared to that of the treatment without oxygen. (authors)

  3. Method of decontamination for uranium oxide particles floating in liquid waste

    International Nuclear Information System (INIS)

    Terakado, Tsutomu; Ebara, Tsuneo; Sato, Kuniaki.

    1981-01-01

    Purpose: To rapidly treat liquid waste containing uranium oxide particles floating in it and to enable substantially complete decontamination. Method: An iron salt such as ferrous sulfate or the like is added to liquid waste with floating uranium oxide particles, an alkaline solution such as caustic soda or the like is then added to the liquid waste while feeding compressed air at 0.1 to 0.02 l/sec. per ton of liquid waste, and the pH of the liquid waste is made to from 6.5 to 7.5. Thereafter, the feed of compressed air is stopped, the liquid waste is allowed to stand, and is then filtered. (Aizawa, K.)

  4. The controlling role of positive structures over the metallogenesis and emplacement of inter layer oxidation sandstone type uranium deposits

    International Nuclear Information System (INIS)

    Gu Kangheng; Chen Zuyi

    2010-01-01

    The positive structures in this paper mean the geological structures related to the occurrence of U-metallogenic zones or U-deposit such as anticlines, uplifts and uplifted fault-blocks. Occurrence features of interlayer oxidation sandstone type deposit at the southern margin of Yili basin and southwestern margin of Turpan-Hami basin, the northeastern margin of Jiudong basin illustrate that the sandstone-hosted uranium deposits, the U-mineralized sections and the uranium occurrences are always selectively emplaced on/in positive structures. The reasons for this lie in the formation mechanism of sandstone-hosted U-deposits. The positive structures raised the elevation of ore-hosting sandstone horizon and make it close to ground surface or exposed at the ground surface, which result in the infiltration of uranium and oxygen bearing groundwater from recharge area into host sandstone horizon, and the interlayer oxidation of host sandstone, as well as the dissolution and the migration of uranium in host sandstone, and the reduction mineralization at the oxidation-reduction interface. Sufficient attention should be paid to the controlling role of positive structures over the metallogenesis and emplacement of sandstone-hosted uranium deposits. They could act as an important criterion for recognizing and prognosticating potential uranium mineralized areas in uranium metallogenic zones or uranium-productive sedimentary basins. (authors)

  5. Influences of structures on the interlayer oxidation zone sandstone-type uranium deposits on the southern margin of Yili basin

    International Nuclear Information System (INIS)

    Wang Mou; Li Shengfu

    2006-01-01

    Based on geology and the theory of hydromorphic origin uranium deposit, structural conditions of uranium formation on the southern margin of Yili Basin are analyzed from two aspects of structural movements and deformation. It is suggested that the subsidiary structures caused by the neotectonic movement are the major factor that control and reform the interlayer oxidation zone sandstone-type uranium deposit, and the differences lie in the tectonics at the eastern and western section on the southern margin of Yili Basin. At the western section, because Mesozoic and Cenozoic strata are tilted by the subsidiary structures, some strata on the margin of the basin outcrop at the surface and suffer from the weathering and erosion, which is favorable for the formation of large size uranium deposits. But at the eastern section, the fault and fold are predominant, outcropping at the surface, cause the redistribution of the uranium, which is favorable for the formation of small size uranium deposits. (authors)

  6. Studies involving direct heating of uranium and plutonium oxides by microwaves

    Energy Technology Data Exchange (ETDEWEB)

    Mallik, G K; Malav, R K; Karande, A P; Bhargava, V K; Kamath, H S [Bhabha Atomic Research Centre, Tarapur (India). Advanced Fuel Fabrication Facility

    1997-08-01

    Nuclear fuel fabrication and recovery of nuclear materials from scraps generated during fabrication involve heating steps like dewaxing, sintering, roasting of scraps, calcination, etc. The dielectric properties of uranium and plutonium oxides place them in the category of materials which are susceptible to absorption of microwaves. The studies were carried out to explore the microwave heating technique for these steps required in nuclear fuel fabrication and scrap recovery laboratories. (author). 1 ref.

  7. Investigation of the kinetics of the reactions of oxidation, nitration, and hydrogenation of uranium

    International Nuclear Information System (INIS)

    Adda, Y.

    1955-06-01

    Various physico-chemical methods have been used to investigate the kinetics of the oxidation hydridation and nitridation of uranium. The experimental results show that the kinetics of these reactions are influenced by many factors also the Pilling and Bedworth rule is valid only under very limited conditions. The disagreement between this rule and the experimental results could be explained by the existence of numerous mechanical faults in the compounds obtained by the dry corrosion of the metal. (author) [fr

  8. Depleted uranium oxides as spent-nuclear-fuel waste-package invert and backfill materials

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Haire, M.J.

    1997-01-01

    A new technology has been proposed in which depleted uranium, in the form of oxides or silicates, is placed around the outside of the spent nuclear fuel waste packages in the geological repository. This concept may (1) reduce the potential for repository nuclear criticality events and (2) reduce long-term release of radionuclides from the repository. As a new concept, there are significant uncertainties

  9. Thermal expansion studies on uranium-neodymium mixed oxide solid solutions

    International Nuclear Information System (INIS)

    Panneerselvam, G.; Venkata Krishnan, R.; Antony, M.P.; Nagarajan, K.

    2012-01-01

    Uranium-Neodymium mixed oxides solid solutions (U 1-y Nd y ) O 2 (y=0.2-0.95) were prepared by combustion synthesis using citric acid as fuel. Structural characterization and computation of lattice parameter was carried out from room temperature X-ray diffraction measurements. Single-phase fluorite structure was observed up to y=0.80. For solid solutions with y>0.80 additional Nd 2 O 3 lines were visible

  10. A study on oxidation treatment of uranium metal chip under controlling atmosphere for safe storage

    International Nuclear Information System (INIS)

    Kim, Chang Kyu; Ji, Chul Goo; Bae, Sang Oh; Woo, Yoon Myeoung; Kim, Jong Goo; Ha, Yeong Keong

    2011-01-01

    The U metal chips generated in developing nuclear fuel and a gamma radioisotope shield have been stored under immersion of water in KAERI. When the water of the storing vessels vaporizes or drains due to unexpected leaking, the U metal chips are able to open to air. A new oxidation treatment process was raised for a long time safe storage with concepts of drying under vacuum, evaporating the containing water and organic material with elevating temperature, and oxidizing the uranium metal chips at an appropriate high temperature under conditions of controlling the feeding rate of oxygen gas. In order to optimize the oxidation process the uranium metal chips were completely dried at higher temperature than 300 .deg. C and tested for oxidation at various temperatures, which are 300 .deg. C, 400 .deg. C, and 500 .deg. C. When the oxidation temperature was 400 .deg. C, the oxidized sample for 7 hours showed a temperature rise of 60 .deg. C in the self-ignition test. But the oxidized sample for 14 hours revealed a slight temperature rise of 7 .deg. C representing a stable behavior in the self-ignition test. When the temperature was 500 .deg. C, the shorter oxidation for 7 hours appeared to be enough because the self-ignition test represented no temperature rise. By using several chemical analyses such as carbon content determination, X-ray deflection (XRD), Infrared spectra (IR) and Thermal gravimetric analysis (TGA) on the oxidation treated samples, the results of self-ignition test of new oxidation treatment process for U metal chip were interpreted and supported

  11. Sintering uranium oxide in the reaction product of hydrogen-carbon dioxide mixtures

    International Nuclear Information System (INIS)

    De Hollander, W.R.; Nivas, Y.

    1975-01-01

    Compacted pellets of uranium oxide alone or containing one or more additives such as plutonium dioxide, gadolinium oxide, titanium dioxide, silica, and alumina are heated to 900 to 1599 0 C in the presence of a mixture of hydrogen and carbon dioxide, either alone or with an inert carrier gas and held at the desired temperature in this atmosphere to sinter the pellets. The sintered pellets are then cooled in an atmosphere having an oxygen partial pressure of 10 -4 to 10 -18 atm of oxygen such as dry hydrogen, wet hydrogen, dry carbon monoxide, wet carbon monoxide, inert gases such as nitrogen, argon, helium, and neon and mixtures of ayny of the foregoing including a mixture of hydrogen and carbon dioxide. The ratio of hydrogen to carbon dioxide in the gas mixture fed to the furnace is controlled to give a ratio of oxygen to uranium atoms in the sintered particles within the range of 1.98:1 to about 2.10:1. The water vapor present in the reaction products in the furnace atmosphere acts as a hydrolysis agent to aid removal of fluoride should such impurity be present in the uranium oxide. (U.S.)

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

    Science.gov (United States)

    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.

  13. Chlorination of uranium oxides with CCl4 using a mechanochemical method

    Science.gov (United States)

    Kitawaki, Shinichi; Nagai, Takayuki; Sato, Nobuaki

    2013-08-01

    A chlorination method for uranium oxides at low temperature was investigated by using a mechanochemical method. In particular, the possibility of the chlorination of uranium oxides, such as UO2 and U3O8, via mechanochemical reaction with CCl4 was studied using a planetary ball mill. Mechanochemical experiments were conducted to evaluate the effect of milling time, CCl4/uranium oxide molar ratio, and revolution speed on the reaction. The synthesized products were then subjected to X-ray diffraction analysis, and it was found that the chlorination of U3O8 with CCl4 to UOCl2, UCl4, and U2O2Cl5 proceeded. However, the chlorination reaction could not be observed when using UO2 powder as the raw material. The chlorination reaction could not be observed when using UO2 powder as the raw material. The chlorination of U3O8 with CCl4 to form UOCl2, UCl4, and U2O2Cl5 via mechanochemical reaction occurs at room temperature. The ratio of chlorination increases with milling time when the appropriate amount of CCl4 is employed. However, the use of excess liquid CCl4 decreases the mechanochemical effect.

  14. Chlorination of uranium oxides with CCl4 using a mechanochemical method

    International Nuclear Information System (INIS)

    Kitawaki, Shinichi; Nagai, Takayuki; Sato, Nobuaki

    2013-01-01

    Highlights: • UCl 4 or UOCl 2 could be synthesized from U 3 O 8 with CCl 4 by using a planetary ball mill. • The chlorination could not be observed when using UO 2 powder as the starting material. • Extension of milling time was effective for chlorinating U 3 O 8 with the appropriate amount of CCl 4 . -- Abstract: A chlorination method for uranium oxides at low temperature was investigated by using a mechanochemical method. In particular, the possibility of the chlorination of uranium oxides, such as UO 2 and U 3 O 8 , via mechanochemical reaction with CCl 4 was studied using a planetary ball mill. Mechanochemical experiments were conducted to evaluate the effect of milling time, CCl 4 /uranium oxide molar ratio, and revolution speed on the reaction. The synthesized products were then subjected to X-ray diffraction analysis, and it was found that the chlorination of U 3 O 8 with CCl 4 to UOCl 2 , UCl 4 , and U 2 O 2 Cl 5 proceeded. However, the chlorination reaction could not be observed when using UO 2 powder as the raw material

  15. Chlorination of uranium oxides with CCl{sub 4} using a mechanochemical method

    Energy Technology Data Exchange (ETDEWEB)

    Kitawaki, Shinichi, E-mail: kitawaki.shinichi@jaea.go.jp [Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai, Naka, Ibaraki 319-1194 (Japan); Institute of Multidisciplinary Research for Advanced Materials, Tohoku University, 2-1-1 Katahira, Aoba, Sendai, Miyagi 980-8577 (Japan); Nagai, Takayuki [Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai, Naka, Ibaraki 319-1194 (Japan); Sato, Nobuaki [Institute of Multidisciplinary Research for Advanced Materials, Tohoku University, 2-1-1 Katahira, Aoba, Sendai, Miyagi 980-8577 (Japan)

    2013-08-15

    Highlights: • UCl{sub 4} or UOCl{sub 2} could be synthesized from U{sub 3}O{sub 8} with CCl{sub 4} by using a planetary ball mill. • The chlorination could not be observed when using UO{sub 2} powder as the starting material. • Extension of milling time was effective for chlorinating U{sub 3}O{sub 8} with the appropriate amount of CCl{sub 4}. -- Abstract: A chlorination method for uranium oxides at low temperature was investigated by using a mechanochemical method. In particular, the possibility of the chlorination of uranium oxides, such as UO{sub 2} and U{sub 3}O{sub 8}, via mechanochemical reaction with CCl{sub 4} was studied using a planetary ball mill. Mechanochemical experiments were conducted to evaluate the effect of milling time, CCl{sub 4}/uranium oxide molar ratio, and revolution speed on the reaction. The synthesized products were then subjected to X-ray diffraction analysis, and it was found that the chlorination of U{sub 3}O{sub 8} with CCl{sub 4} to UOCl{sub 2}, UCl{sub 4}, and U{sub 2}O{sub 2}Cl{sub 5} proceeded. However, the chlorination reaction could not be observed when using UO{sub 2} powder as the raw material.

  16. Environmental control technology for mining and milling low-grade uranium resources

    International Nuclear Information System (INIS)

    Weakley, S.A.; Blahnik, D.E.; Long, L.W.; Bloomster, C.H.

    1981-04-01

    This study examined the type and level of wastes that would be generated in the mining and milling of U 3 O 8 from four potential domestic sources of uranium. The estimated costs of the technology to control these wastes to different degrees of stringency are presented

  17. Bacteriological lixiviation of low-grade uranium ores at low temperatures, by phiobacillus ferrooxidaus

    International Nuclear Information System (INIS)

    Lobato Filho, A.N.S.

    1976-12-01

    Laboratory experiments are described that, using selective and mutagenic agents, allowed the isolation of a strain of thiobacillus ferrooxidams capable of developing at 8 0 C, and keeping its oxidesing characteristics tests showed that the isoled sample is capable of solubilizing 95% of the uranium content in samples with U 3 O 8 content below 1000ppm [pt

  18. Insertion compounds of transition-metal and uranium oxides

    International Nuclear Information System (INIS)

    Chippindale, A.M.; Dickens, P.G.; Powell, A.V.

    1991-01-01

    Several transition-metal and actinide oxides, in which the metal occurs in a high oxidation state, have open covalent structures and are capable of incorporating alkali and other electropositive metals under mild conditions to form insertion compounds A x MO n . These are solids which have several features in common: Over a range of compositions, A x MO n exists as one or more stable or metastable phases in which the structure of the parent oxide MO n is largely retained and the insertion element A is accommodated interstitially. Insertion is accompanied by a redox process A=A i . + e - M in which M is reduced and the electronic properties of the parent oxide change to those typical of a mixed-valence compound. The insertion process xA + MO n = A x MO n can be reversed, at least to some extent, by chemical or electrochemical reaction, with retention of structure (topotactic reaction). This review concentrates on methods of synthesis, characterisation, crystal structure and thermochemistry of these insertion compounds. It updates and extends previous work. (author)

  19. The temperature coefficient of the resonance integral for uranium metal and oxide

    Energy Technology Data Exchange (ETDEWEB)

    Blomberg, P; Hellstrand, E; Homer, S

    1960-06-15

    The temperature coefficient of the resonance integral in uranium metal and oxide has been measured over a wide temperature range for rods with three different diameters. The results for metal agree with most earlier results from activation measurements but differ as much as a factor of two from results obtained with reactivity methods. For oxide only one measurement has been reported recently. Our value is considerably lower than the result of that measurement. The experiments will continue in order to find the reason for the large discrepancy mentioned above.

  20. The temperature coefficient of the resonance integral for uranium metal and oxide

    International Nuclear Information System (INIS)

    Blomberg, P.; Hellstrand, E.; Homer, S.

    1960-06-01

    The temperature coefficient of the resonance integral in uranium metal and oxide has been measured over a wide temperature range for rods with three different diameters. The results for metal agree with most earlier results from activation measurements but differ as much as a factor of two from results obtained with reactivity methods. For oxide only one measurement has been reported recently. Our value is considerably lower than the result of that measurement. The experiments will continue in order to find the reason for the large discrepancy mentioned above

  1. TSUNAMI analysis of the applicability of proposed experiments to reactor-grade and weapons-grade mixed-oxide systems

    International Nuclear Information System (INIS)

    Rearden, Bradley T.; Hopper, Calvin M.; Elam, Karla R.

    2005-01-01

    The applicability of proposed critical experiments for the criticality code validation of a series of prototypic reactor-grade and weapons-grade mixed-oxide systems has been assessed with the TSUNAMI methodology from SCALE 5. The application systems were proposed by the Nuclear Energy Agency (NEA) Organization for Economic Cooperation and Development (OECD) Working Party on Nuclear Criticality Safety MOX Experimental Needs Working Group. Forty-eight application systems were conceived to envelope the range of conditions in processing and fabrication of reactor-grade and weapons-grade MOX fuel. The applicability of 303 existing critical benchmarks to each of the 48 applications was assessed, and validation coverage was found to be lacking for certain applications. Two series of proposed critical experiments were also considered in this analysis. The TSUNAMI analysis has revealed that both series of proposed experiments are applicable to numerous configurations of the reactor-grade and weapons-grade systems. A detailed assessment of which experiments were revealed by TSUNAMI to be most applicable to specific prototypic fuel processing systems has been performed. (author)

  2. Anaerobic U(IV) Bio-oxidation and the Resultant Remobilization of Uranium in Contaminated Sediments

    International Nuclear Information System (INIS)

    Coates, John D.

    2005-01-01

    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, 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

  3. A Study on the Electrolytic Reduction Mechanism of Uranium Oxide in a LiCl-Li2O Molten Salt

    International Nuclear Information System (INIS)

    Oh, Seung Chul; Hur, Jin Mok; Seo, Chung Seok; Park, Seong Won

    2003-01-01

    This study proposed a new electrolytic reduction technology that is based on the integration of simultaneous uranium oxide metallization and Li 2 O electrowinning. In this electrolytic reduction reaction, electrolytically reduced Li deposits on cathode and simultaneously reacts with uranium oxides to produce uranium metal showing more than 99% conversion. For the verification of process feasibility, the experiments to obtain basic data on the metallization of uranium oxide, investigation of reaction mechanism, the characteristics of closed recycle of Li 2 O and mass transfer were carried out. This evolutionary electrolytic reduction technology would give benefits over the conventional Li-reduction process improving economic viability such as: avoidance of handling of chemically active Li-LiCl molten salt increase of metallization yield, and simplification of process.

  4. A study on the electrolytic reduction of uranium oxide in a LiCl-Li2O molten salt

    International Nuclear Information System (INIS)

    Su, J. S.; Hu, J. M.; Hong, S. S.; Jang, D. S.; Park, S. W.

    2003-01-01

    New electrolytic reduction technology was proposed that is based on the integration of metallization of uranium oxide and Li 2 O electrowinning. In this electrolytic reduction reaction, electrolytically reduced Li deposits on cathode and simultaneously reacts with uranium oxides to produce uranium metal showing more than 99% conversion. For the verification of process feasibility, the experiments to obtain basic data on the metallization of uranium oxide, investigation of reaction mechanism, the characteristics of closed recycle of Li 2 O and mass transfer were carried out. This evolutionary electrolytic reduction technology would give benefits over the conventional Li-reduction process improving economic viability such as: avoidance of handling of chemically active Li-LiCl molten salt, increase of metallization yield, and simplification of process

  5. Direct electrochemical reduction of solid uranium oxide in molten fluoride salts

    Science.gov (United States)

    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.

  6. Enhanced bioremediation of oil contaminated soil by graded modified Fenton oxidation.

    Science.gov (United States)

    Xu, Jinlan; Xin, Lei; Huang, Tinglin; Chang, Kun

    2011-01-01

    Graded modified Fenton's (MF) oxidation is a strategy in which H2O2 is added intermittently to prevent a sharp temperature increase and undesired soil sterilization at soil circumneutral pH versus adding the same amount of H2O2 continuously. The primary objective of the present study was to investigate whether a mild MF pre-oxidation such as a stepwise addition of H2O2 can prevent sterilization and achieve a maximum degradation of tank oil in soil. Optimization experiments of graded MF oxidation were conducted using citric acid, oxalic acid and SOLV-X as iron chelators under different frequencies of H2O2 addition. The results indicated that the activity order of iron chelates decreased as: citric acid (51%) > SOLV-X (44%) > oxalic acid (9%), and citric acid was found to be an optimized iron chelating agent of graded MF oxidation. Three-time addition of H2O2 was found to be favorable and economical due to decreasing total petroleum hydrocarbon removal from three time addition (51%) to five time addition (59%). Biological experiments were conducted after graded MF oxidation of tank oil completed under optimum conditions mentioned above. After graded oxidation, substantially higher increase (31%) in microbial activity was observed with excessive H2O2 (1470 mmol/L, the mol ratio of H2O2:Fe2+ was 210:1) than that of non-oxidized soil. Removal efficiency of tank oil was up to 93% after four weeks. Especially, the oil fraction (C10-C40) became more biodagradable after graded MF oxidation than its absence. Therefore, graded MF oxidation is a mild pretreatment to achieve an effective bioremediation of oil contaminated soil.

  7. The effect of hydrogen peroxide on uranium oxide films on 316L stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Wilbraham, Richard J., E-mail: r.wilbraham@lancaster.ac.uk [The Lloyd’s Register Foundation Centre for Nuclear Engineering, Engineering Department, Lancaster University, Bailrigg, Lancashire LA1 4YR (United Kingdom); Boxall, Colin, E-mail: c.boxall@lancaster.ac.uk [The Lloyd’s Register Foundation Centre for Nuclear Engineering, Engineering Department, Lancaster University, Bailrigg, Lancashire LA1 4YR (United Kingdom); Goddard, David T., E-mail: dave.t.goddard@nnl.co.uk [National Nuclear Laboratory, Preston Laboratory, Springfields, Preston, Lancashire PR4 0XJ (United Kingdom); Taylor, Robin J., E-mail: robin.j.taylor@nnl.co.uk [National Nuclear Laboratory, Central Laboratory, Seascale, Cumbria CA20 1PG (United Kingdom); Woodbury, Simon E., E-mail: simon.woodbury@nnl.co.uk [National Nuclear Laboratory, Central Laboratory, Seascale, Cumbria CA20 1PG (United Kingdom)

    2015-09-15

    Highlights: • The first report of the presence of both UO{sub 2} and polymeric UO{sub 2}{sup 2+} in the same electrodeposited U oxide sample. • The action of H{sub 2}O{sub 2} on electrodeposited U oxides is described using corrosion based concepts. • Electrodeposited U oxide freely dissolves at hydrogen peroxide concentrations <100 μmol dm{sup −3}. • At [H{sub 2}O{sub 2}] > 0.1 mmol dm{sup −3} dissolution is inhibited by formation of a studtite passivation layer. • At [H{sub 2}O{sub 2}] ⩾ 1 mol dm{sup −3} studtite formation competes with uranyl–peroxide complex formation. - Abstract: For the first time the effect of hydrogen peroxide on the dissolution of electrodeposited uranium oxide films on 316L stainless steel planchets (acting as simulant uranium-contaminated metal surfaces) has been studied. Analysis of the H{sub 2}O{sub 2}-mediated film dissolution processes via open circuit potentiometry, alpha counting and SEM/EDX imaging has shown that in near-neutral solutions of pH 6.1 and at [H{sub 2}O{sub 2}] ⩽ 100 μmol dm{sup −3} the electrodeposited uranium oxide layer is freely dissolving, the associated rate of film dissolution being significantly increased over leaching of similar films in pH 6.1 peroxide-free water. At H{sub 2}O{sub 2} concentrations between 1 mmol dm{sup −3} and 0.1 mol dm{sup −3}, formation of an insoluble studtite product layer occurs at the surface of the uranium oxide film. In analogy to corrosion processes on common metal substrates such as steel, the studtite layer effectively passivates the underlying uranium oxide layer against subsequent dissolution. Finally, at [H{sub 2}O{sub 2}] > 0.1 mol dm{sup −3} the uranium oxide film, again in analogy to common corrosion processes, behaves as if in a transpassive state and begins to dissolve. This transition from passive to transpassive behaviour in the effect of peroxide concentration on UO{sub 2} films has not hitherto been observed or explored, either in terms

  8. Uranium

    International Nuclear Information System (INIS)

    Battey, G.C.; McKay, A.D.

    1988-01-01

    Production for 1986 was 4899 t U 3 O 8 (4154 t U), 30% greater than in 1985, mainly because of a 39% increase in production at Ranger. Exports for 1986 were 4166 t U 3 O 8 at an average f.o.b. unit value of $40.57/lb U 3 O 8 . Private exploration expenditure for uranium in Australia during the 1985-86 fiscal year was $50.2 million. Plans were announced to increase the nominal capacity of the processing plant at Ranger from 3000 t/year U 3 O 8 to 4500 t and later to 6000 t/year. Construction and initial mine development at Olympic Dam began in March. Production is planned for mid 1988 at an annual rate of 2000 t U 3 O 8 , 30 000 t Cu, and 90 000 oz (2800 kg) Au. The first long-term sales agreement was concluded in September 1986. At the Manyingee deposit, testing of the alkaline solution mining method was completed, and the treatment plant was dismantled. Spot market prices (in US$/lb U 3 O 8 ) quoted by Nuexco were generally stable. From January-October the exchange value fluctuated from US$17.00-US$17.25; for November and December it was US$16.75. Australia's Reasonably Assured Resources of uranium recoverable at less than US$80/kg U at December 1986 were estimated as 462 000 t U, 3000 t U less than in 1985. This represents 30% of the total low-cost RAR in the WOCA (World Outside the Centrally Planned Economy Areas) countries. Australia also has 257 000 t U in the low-cost Estimated Additional Resources Category I, 29% of the WOCA countries' total resources in this category

  9. Influence of uranyl speciation and iron oxides on uranium biogeochemical redox reactions

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, B.D.; Amos, R.T.; Nico, P.S.; Fendorf, 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 the 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

  10. Thorium/uranium mixed oxide nano-crystals: Synthesis, structural characterization and magnetic properties

    International Nuclear Information System (INIS)

    Hudry, Damien; Griveau, Jean-Christophe; Apostolidis, Christos; Colineau, Eric; Rasmussen, Gert; Walter, Olaf; Wang, Di; Venkata Sai Kiran Chakravadhaluna; Courtois, Eglantine; Kubel, Christian

    2014-01-01

    One of the primary aims of the actinide community within nano-science is to develop a good understanding similar to what is currently the case for stable elements. As a consequence, efficient, reliable and versatile synthesis techniques dedicated to the formation of new actinide-based nano-objects (e.g., nano-crystals) are necessary. Hence, a 'library' dedicated to the preparation of various actinide based nano-scale building blocks is currently being developed. Nano-scale building blocks with tunable sizes, shapes and compositions are of prime importance. So far, the non-aqueous synthesis method in highly coordinating organic media is the only approach which has demonstrated the capability to provide size and shape control of actinide-based nano-crystals (both for thorium and uranium, and recently extended to neptunium and plutonium). In this paper, we demonstrate that the non-aqueous approach is also well adapted to control the chemical composition of the nano-crystals obtained when mixing two different actinides. Indeed, the controlled hot co-injection of thorium acetylacetonate and uranyl acetate (together with additional capping agents) into benzyl ether can be used to synthesize thorium/uranium mixed oxide nano-crystals covering the full compositional spectrum. Additionally, we found that both size and shape are modified as a function of the thorium/uranium ratio. Finally, the magnetic properties of the different thorium/uranium mixed oxide nano-crystals were investigated. Contrary to several reports, we did not observe any ferromagnetic behavior. As a consequence, ferromagnetism cannot be described as a universal feature of nano-crystals of non-magnetic oxides as recently claimed in the literature. (authors)

  11. Reduction of uranium and plutonium oxides by aluminum. Application to the recycling of plutonium

    International Nuclear Information System (INIS)

    Gallay, J.

    1968-01-01

    A process for treating plutonium oxide calcined at high temperatures (1000 to 2000 deg. C) with a view to recovering the metal consists in the reduction of this oxide dissolved in a mixture of aluminium, sodium and calcium fluorides by aluminium at about 1180 deg. C. The first part of the report presents the results of reduction tests carried out on the uranium oxides UO 2 and U 3 O 8 ; these are in agreement with the thermodynamic calculations of the exchange reaction at equilibrium. The second part describes the application of this method to plutonium oxides. The Pu-Al alloy obtained (60 per cent Pu) is then recycled in an aqueous medium. (author) [fr

  12. Description and classification of uranium oxide hydrate sheet topologies

    International Nuclear Information System (INIS)

    Miller, M.L.; Burns, P.C.; Ewing, R.C.; Finch, R.J.

    1996-01-01

    The uranyl oxide hydrates (UOH) are important corrosion products of uraninite and UO 2 in spent nuclear fuel under oxidizing conditions. However, the systematics of the crystal chemistry, thermodynamic parameters, and solubilities of this mineral group are poorly understood. With the exception of the synthetic UO 2 (OH) 2 polymorphs, all UOH crystal structures are based on sheets of edge-sharing 5 and 4-coordinated uranyl dipyramids. This structural similarity suggests that it is possible to develop a model by which to estimate the thermodynamic behavior of UOHs from data on structural endmember phases. Toward this end, a method of quantitatively describing all known UOH sheets has been developed. Only four structural unit chains are required to construct the uranyl oxide hydrate sheets (as well as the structurally similar U 3 O 8 sheets). The H-chain is restricted to α-UO 2 (OH) 2 and is made up of hexagonally coordinated uranyl ions sharing opposing edges. The arrowhead chain composed of pentagonal dipyramids sharing edges and alternating with trigonal vacancies is present in all other UOH sheets. These arrowhead chains are directed and can occur in both an Up-arrow and Down-arrow sense within a single sheet. The P-chain consists of edge-sharing pentagonal dipyramids forming a zigzag chain. The P-chain is flanked on both sides by arrowhead chains of the same sense. The remaining structural unit is a discontinuous chain of rhombic dipyramids. This R-chain is produced when nested adjacent Up-arrow and Down-arrow arrowhead chains are translated by a diagonal shift. This chain occurs in sheets which contain only 4-coordinate uranyl ion and those containing both 4- and 5-coordinate uranyl ions

  13. Subsurface Conditions Controlling Uranium Incorporation in Iron Oxides: A Redox Stable Sink

    International Nuclear Information System (INIS)

    Fendorf, Scott

    2016-01-01

    Toxic metals and radionuclides throughout the U.S. Department of Energy Complex pose a serious threat to ecosystems and to human health. Of particular concern is the redox-sensitive radionuclide uranium, which is classified as a priority pollutant in soils and groundwaters at most DOE sites owing to its large inventory, its health risks, and its mobility with respect to primary waste sources. The goal of this research was to contribute to the long-term mission of the Subsurface Biogeochemistry Program by determining reactions of uranium with iron (hydr)oxides that lead to long-term stabilization of this pervasive contaminant. The research objectives of this project were thus to (1) identify the (bio)geochemical conditions, including those of the solid-phase, promoting uranium incorporation in Fe (hydr)oxides, (2) determine the magnitude of uranium incorporation under a variety of relevant subsurface conditions in order to quantify the importance of this pathway when in competition with reduction or adsorption; (3) identify the mechanism(s) of U(VI/V) incorporation in Fe (hydr)oxides; and (4) determine the stability of these phases under different biogeochemical (inclusive of redox) conditions. Our research demonstrates that redox transformations are capable of achieving U incorporation into goethite at ambient temperatures, and that this transformation occurs within days at U and Fe(II) concentrations that are common in subsurface geochemical environments with natural ferrihydrites - inclusive of those with natural impurities. Increasing Fe(II) or U concentration, or initial pH, made U(VI) reduction to U(IV) a more competitive sequestration pathway in this system, presumably by increasing the relative rate of U reduction. Uranium concentrations commonly found in contaminated subsurface environments are often on the order of 1-10 μM, and groundwater Fe(II) concentrations can reach exceed 1 mM in reduced zones of the subsurface. The redox-driven U(V) incorporation

  14. Subsurface Conditions Controlling Uranium Incorporation in Iron Oxides: A Redox Stable Sink

    Energy Technology Data Exchange (ETDEWEB)

    Fendorf, Scott [Stanford Univ., CA (United States)

    2016-04-05

    Toxic metals and radionuclides throughout the U.S. Department of Energy Complex pose a serious threat to ecosystems and to human health. Of particular concern is the redox-sensitive radionuclide uranium, which is classified as a priority pollutant in soils and groundwaters at most DOE sites owing to its large inventory, its health risks, and its mobility with respect to primary waste sources. The goal of this research was to contribute to the long-term mission of the Subsurface Biogeochemistry Program by determining reactions of uranium with iron (hydr)oxides that lead to long-term stabilization of this pervasive contaminant. The research objectives of this project were thus to (1) identify the (bio)geochemical conditions, including those of the solid-phase, promoting uranium incorporation in Fe (hydr)oxides, (2) determine the magnitude of uranium incorporation under a variety of relevant subsurface conditions in order to quantify the importance of this pathway when in competition with reduction or adsorption; (3) identify the mechanism(s) of U(VI/V) incorporation in Fe (hydr)oxides; and (4) determine the stability of these phases under different biogeochemical (inclusive of redox) conditions. Our research demonstrates that redox transformations are capable of achieving U incorporation into goethite at ambient temperatures, and that this transformation occurs within days at U and Fe(II) concentrations that are common in subsurface geochemical environments with natural ferrihydrites—inclusive of those with natural impurities. Increasing Fe(II) or U concentration, or initial pH, made U(VI) reduction to U(IV) a more competitive sequestration pathway in this system, presumably by increasing the relative rate of U reduction. Uranium concentrations commonly found in contaminated subsurface environments are often on the order of 1-10 μM, and groundwater Fe(II) concentrations can reach exceed 1 mM in reduced zones of the subsurface. The redox-driven U(V) incorporation

  15. Oxidative Uranium Release from Anoxic Sediments under Diffusion-Limited Conditions.

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    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 (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.

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

    Science.gov (United States)

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

    2010-09-01

    Accidental exposure by inhalation to alpha-emitting particles from mixed oxide (MOX: uranium and plutonium oxide) fuels is a potential long-term health risk to workers in nuclear fuel fabrication plants. For MOX fuels, the risk of lung cancer development may be different from that assigned to individual components (plutonium, uranium) given different physico-chemical characteristics. The objective of this study was to investigate late effects in rat lungs following inhalation of MOX aerosols of similar particle size containing 2.5 or 7.1% plutonium. Conscious rats were exposed to MOX aerosols and kept for their entire lifespan. Different initial lung burdens (ILBs) were obtained using different amounts of MOX. Lung total alpha activity was determined by external counting and at autopsy for total lung dose calculation. Fixed lung tissue was used for anatomopathological, autoradiographical, and immunohistochemical analyses. Inhalation of MOX at ILBs ranging from 1-20 kBq resulted in lung pathologies (90% of rats) including fibrosis (70%) and malignant lung tumors (45%). High ILBs (4-20 kBq) resulted in reduced survival time (N = 102; p inhalation result in similar risk for development of lung tumors as compared with industrial plutonium oxide.

  18. Role of thermal analysis in uranium oxide fuel fabrication process

    International Nuclear Information System (INIS)

    Balaji Rao, Y.; Yadav, R.B.

    2006-01-01

    The present paper discusses the application of thermal analysis, particularly, differential thermal analysis (Dta) at various stages of fuel fabrication process. The useful role of Dta in knowing the decomposition pattern and calcination temperature of Adu along with de-nitration temperature is explained. The decomposition pattern depends upon the type of drying process adopted for wet ADU cake (ADU C). Also, the paper highlights the utility of DTA in determining the APS and SSA of UO 2+x and U 3 O 8 powders as an alternate technique. Further, the temperature difference (ΔT max ) between the two exothermic peaks obtained in UO 2+x powder oxidation is related to sintered density of UO 2 pellets. (author)

  19. Irradiation behavior of uranium oxide - Aluminum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Rest, Jeffrey; Snelgrove, James L.

    1996-01-01

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products and as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show that, with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 g U/cm 3 ) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼63% 235 U burnup). (author)

  20. Irradiation behavior of uranium oxide-aluminum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, G.L.; Rest, J.; Snelgrove, J.L.

    1996-01-01

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products, as well as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show, that with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 gm/cc) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼ 63% 235 U burnup)

  1. Study of the discontinuous dissolution of uranium oxides

    International Nuclear Information System (INIS)

    Bueno, L.A.O.; Floh, B.; Araujo, J.A. de.

    1978-01-01

    Dissolution studies of UO 2 and U 3 O 8 particles and pellets in nitric acid were carried out to find the best flowsheet conditions for treatment of irradiated materials. All experiments were accomplished with unirradiated oxides at room-and boiling point temperature of the nitric acid solutions, the acid molarity ranging from 1 up to 12M in stoichiometric-and (100% up to 300%) excess conditions. The UO 2 (10g) and U 3 O 8 (10g) are easily dissolved (10 and 50s), respectively, at boiling point of 6M nitric acid solution. At the same conditions compacted pellets are dissolved in 29 min (U 3 O 8 =5g) and in 330 min (UO 2 =20g) [pt

  2. Individual monitoring of internal exposure to uranium oxides in two fuel fabrication plants

    International Nuclear Information System (INIS)

    Bourdeix, F.; Achiary, J.; Berard, P.

    1989-01-01

    Individual monitoring of personal exposure to inhalation of uranium oxides throughout the manufacture of fuel for pressurized water reactor (PWR) includes lung gamma-spectrometry, fecal analysis and urine analysis. Examination of the results shows the following: internal exposure is the consequence of repeated intake incidents as revealed by early peaks of urinary and particularly fecal elimination; a shift is often observed with the results of aerosol concentration measured through air collectors; the measured variations of uranium lung incorporations are relatively fast (apparent mean period 165 d). Correct evaluation of the effective dose equivalent from inhalation requires further information concerning the aerosol size distribution at work stations, the physico-chemical characteristics of the product leading to an estimate of its actual biological solubility, and the measurement of the fraction of aerosol liable to intake with an individual portable collector [fr

  3. Selective oxidation of propene on bismuth molybdate and mixed oxides of tin and antimony and of uranium and antimony

    International Nuclear Information System (INIS)

    Pendleton, P.; Taylor, D.

    1976-01-01

    Propene + 18 0 2 reactions have been studied in a static reaction system on bismuth molybdate and mixed oxides of tin and antimony and of uranium and antimony. The [ 16 0] acrolein content of the total acrolein formed and the proportion of 16 0 in the oxygen of the carbon dioxide by-product have been determined. The results indicate that for each catalyst the lattice is the only direct source of the oxygen in the aldehyde, and that lattice and/or gas phase oxygen is used in carbon dioxide formation. Oxygen anion mobility appears to be greater in the molybdate catalyst than in the other two. (author)

  4. Oxidation-reduction phenomena in tabular uranium-vanadium bearing sandstone from the Salt Wash deposits (Upper Jurassic) of the Cottonwood Wash district (Utah, USA)

    International Nuclear Information System (INIS)

    Meunier, J.D.

    1984-02-01

    A braided to meandering fluvial environment has been postulated for this area after a sedimentological study. The mineralization is spatially related with conifer derived organic matter and wood is preserved in these sediments because of the reducing environment of deposition. The degree of maturation of the organic matter has been estimated from chemical analyses. Results show the presence of variable diagenetic oxidation depending on the environment. The organic matter which was least affected by this oxidation have attained a thermal maturation characteristic of the end stage of diagenesis. The high grade ore is situated at the edges of or within the trunks of trees (which remained permeable during diagenesis) and at the boundaries of the carbonaceous beds. Geochemical study shows there to be good correlation between uranium and vanadium. Uranium occurs as pitchblende, coffinite or as impregnations in the vanadiferous clay cement. A detailed study of clays shows an association of chlorite and roscoelite which most probably contain V 3+ . Fluid inclusion study suggests burying temperatures of >= 100 0 C and shows the existance of brines before the mineralization. The following genetical model is proposed. Low Eh uraniferous solutions move through a reduced pyritised environment. The low degree of oxidation of the pyrites propagates the destabilization of the clastic iron-titanium oxides which release vanadium and the dissociation of uranylcarbonates. Then, the deposit of pitchblende, coffinite, montroseite and vanadiferous clays took place in association with a secondary pyrite. When the rocks were uplifted to the subsurface, uranium (IV) and vanadium (III) were remobilised in an oxidising environment to form a secondary mineralization essentially represented by tyuyamunite [fr

  5. A procedure for oxidation during the acid leaching of non-ferrous ores, particularly uranium ores

    International Nuclear Information System (INIS)

    Zubcek, L.; Baloun, S.; Martinek, K.; Vebr, Z.; Krepelka, J.; Lasica, S.

    1989-01-01

    It is suggested that dust from the production of ferroalloys of manganese, particularly ferrosilicomanganese and ferromanganese, can be conveniently used for oxidation during the acid leaching of non-ferrous ores, particularly uranium ores. This dust contains 30 to 40% oxides of manganese, about one-half of this is MnO 2 . Iron in the dust is present in the trivalent form, and the dust is pefectly dry. The conventional grinding of oxidants for the ore processing is eliminated, the dust being available in particle size below 0.2 mm. The dust is added in amounts of 5 to 100 kg per ton of the ore, and the suspension is typically heated at 115 degC for 3.5 hr. (P.A.)

  6. Natural uranium concentrations of native plants over a low-grade ore body

    International Nuclear Information System (INIS)

    Sheppard, M.I.; Thibault, D.H.

    1984-01-01

    Plant uranium concentrations generally reflect soil or rock substrate concentrations in upland areas, but they may not in lowland areas where the rhizoids of Sphagnum spp. and rocks of Ledum groenlandicum may be in direct contact either continuously or on a seasonal basis with the groundwater. This study points out the importance of selecting plant species and collection sites where the true substrate can be well defined and sampled. Sphagnum spp. and Ledum groenlandicum best reflect the substrate uranium concentrations in lowland areas, Umbilicaria spp. and Cladonia spp. in rock outcrop, and Picea mariana and Betula papyrifera in upland locations. The study shows the best plant part to sample is the older tissue such as the stems, twigs, and wood. Since no systematic changes in plant tissue concentrations were found throughout the season, sampling can be carried out anytime. Expression of soil concentrations on an ash weight basis gave a considerably different result than those on a dry weight basis, particularly when comparisons were made between litter-enriched mineral soil and true organic soils. The amount of ash varied among plant organs, species, and taxonomic divisions, and a constant value cannot be used to convert plant ash concentrations on a dry weight basis

  7. Evaluating chemical extraction techniques for the determination of uranium oxidation state in reduced aquifer sediments

    Science.gov (United States)

    Stoliker, Deborah L.; Campbell, Kate M.; Fox, Patricia M.; Singer, David M.; Kaviani, Nazila; Carey, Minna; Peck, Nicole E.; Barger, John R.; Kent, Douglas B.; Davis, James A.

    2013-01-01

    Extraction techniques utilizing high pH and (bi)carbonate concentrations were evaluated for their efficacy in determining the oxidation state of uranium (U) in reduced sediments collected from Rifle, CO. Differences in dissolved concentrations between oxic and anoxic extractions have been proposed as a means to quantify the U(VI) and U(IV) content of sediments. An additional step was added to anoxic extractions using a strong anion exchange resin to separate dissolved U(IV) and U(VI). X-ray spectroscopy showed that U(IV) in the sediments was present as polymerized precipitates similar to uraninite and/or less ordered U(IV), referred to as non-uraninite U(IV) species associated with biomass (NUSAB). Extractions of sediment containing both uraninite and NUSAB displayed higher dissolved uranium concentrations under oxic than anoxic conditions while extractions of sediment dominated by NUSAB resulted in identical dissolved U concentrations. Dissolved U(IV) was rapidly oxidized under anoxic conditions in all experiments. Uraninite reacted minimally under anoxic conditions but thermodynamic calculations show that its propensity to oxidize is sensitive to solution chemistry and sediment mineralogy. A universal method for quantification of U(IV) and U(VI) in sediments has not yet been developed but the chemical extractions, when combined with solid-phase characterization, have a narrow range of applicability for sediments without U(VI).

  8. Determination of oxygen in ternary uranium oxides by a gravimetric alkaline earth addition method

    International Nuclear Information System (INIS)

    Fujino, T.; Tagawa, H.

    1979-01-01

    The applicability of a gravimetric method based on alkaline earth metal addition for the determination of oxygen in ternary uranium oxides of the tupe M-U-O (M=La, Ce and Th) is described. The oxide sample is mixed with MgO or Basub(2.8)UOsub(5.8) and heated in air under suitable conditions. Because uranium is completely oxidized to the hexavalent state during the reaction, oxygen can be determined from the weight change. Oxygen in Lasub(y)Usub(1-y)Osub(2+x) is determined up to y = 0.8 with a standard deviation for x of +- 0.006 with MgO. For Thsub(y)Usub(1-y)Osub(2+x) the value of x is determined with Basub(2.8)UOsub(5.8) with a standard deviation of +- 0.01 at y = 0.8. For Cesub(y)Usub(1-y)Osub(2+x), the method can be applied only for low cerium concentrations where y = 0-0.2; the value for x with Basub(2.8)UOsub(5.8) at y = 0.2 showed a standard deviation of +- 0.002. (Auth.)

  9. Study for electrochemical behavior of uranium oxide in a molten LiCl-Li2O system

    International Nuclear Information System (INIS)

    Park, Sung Bin; Park, Byung Heung; Seo, Chung Seok; Jung, Ki Jung; Park, Seong Won

    2005-01-01

    Interest in the electrolytic reduction of uranium oxide is increasing in the treatment of spent fuel oxides. With complicated and expensive procedures many reactive metals can be prepared in a pure metal form, the electrochemical reduction of a metal oxide has been recently proposed in metallurgy. The electrochemical reduction process is simple and rapid when compared to the conventional processes. The process can reduce the production costs and be applicable to a wide range of metal oxides. Chen et al. proposed the direct electrochemical reduction of titanium dioxide to titanium in a molten calcium chloride. Argonne National Laboratory (ANL) has reported the experimental results of an electrochemical reduction of the uranium oxide fuel in a bench-scale apparatus with a cyclic voltammetry, and has designed high-capacity reduction (HCR) cells and conducted three kg-scale UO 2 reduction runs. Gourishankar et al. classified the mechanisms of the electrolytic reduction of the metal oxides in a LiCl-Li 2 O molten salt system into two types; the simultaneous reduction and the direct electrochemical reduction. The uranium oxide in LiCl-Li 2 O molten salt was converted to uranium metal according to two mechanisms. Korea Atomic Energy Research Institute (KAERI) has developed the Advanced Spent Fuel Conditioning Process (ACP) to be an innovative technology in handling the PWR spent fuel. As part of ACP, the electrolytic reduction process (ER process) is the electrochemical reduction process of uranium oxide to uranium metal in molten salt. The ER process has advantages in a technical stability, an economic potential and a good proliferation resistance. KAERI has reported on the good experimental results of an electrochemical reduction of the uranium oxide in a 20 kg HM/batch lab-scale. In this work, cyclic voltammograms for a LiCl-3 wt% Li 2 O system and an U 3 O 8 -LiCl-3 wt% Li 2 O system with the integrated cathode assembly have been obtained. From the cyclic

  10. Weapons-grade plutonium dispositioning. Volume 3: A new reactor concept without uranium or thorium for burning weapons-grade plutonium

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Schnitzler, B.G.; Fletcher, C.D.

    1993-06-01

    The National Academy of Sciences (NAS) requested that the Idaho National Engineering Laboratory (INEL) examine concepts that focus only on the destruction of 50,000 kg of weapons-grade plutonium. A concept has been developed by the INEL for a low-temperature, low-pressure, low-power density, low-coolant-flow-rate light water reactor that destroys plutonium quickly without using uranium or thorium. This concept is very safe and could be designed, constructed, and operated in a reasonable time frame. This concept does not produce electricity. Not considering other missions frees the design from the paradigms and constraints used by proponents of other dispositioning concepts. The plutonium destruction design goal is most easily achievable with a large, moderate power reactor that operates at a significantly lower thermal power density than is appropriate for reactors with multiple design goals. This volume presents the assumptions and requirements, a reactor concept overview, and a list of recommendations. The appendices contain detailed discussions on plutonium dispositioning, self-protection, fuel types, neutronics, thermal hydraulics, off-site radiation releases, and economics

  11. Transformations of highly enriched uranium into metal or oxide; Etudes des procedes de transformation des composes d'uranium a fort enrichissement isotopique

    Energy Technology Data Exchange (ETDEWEB)

    Nollet, P; Sarrat, P [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-07-01

    The enriched uranium workshops in Cadarache have a double purpose on the one hand to convert uranium hexafluoride into metal or oxide, and on the other hand to recover the uranium contained in scrap materials produced in the different metallurgical transformations. The principles that have been adopted for the design and safety of these workshops are reported. The nuclear safety is based on the geometrical limitations of the processing vessels. To establish the processes and the technology of these workshops, many studies have been made since 1960, some of which have led to original achievements. The uranium hexafluoride of high isotopic enrichment is converted either by injection of the gas into ammonia or by an original process of direct hydrogen reduction to uranium tetrafluoride. The uranium contained m uranium-zirconium metal scrap can be recovered by combustion with hydrogen chloride followed treatment of the uranium chloride by fluorine in order to obtain the uranium in the hexafluoride state. Recovery of the uranium contained m various scrap materials is obtained by a conventional refining process combustion of metallic scrap, nitric acid dissolution of the oxide, solvent purification by tributyl phosphate, ammonium diuranate precipitation, calcining, reduction and hydro fluorination into uranium tetrafluoride, bomb reduction by calcium and slag treatment. Two separate workshops operate along these lines one takes care of the uranium with an isotopic enrichment of up to 3 p. 100, the other handles the high enrichments. The handling of each step of this process, bearing in mind the necessity for nuclear safety, has raised some special technological problems and has led to the conception of new apparatus, in particular the roasting furnace for metal turnings, the nitric acid dissolution unit, the continuous precipitator and ever safe filter and dryer for ammonium diuranate, the reduction and hydro fluorination furnace and the slag recovery apparatus These are

  12. Calculation of oxygen distribution in uranium-plutonium oxide fuels during irradiation (programme CODIF)

    International Nuclear Information System (INIS)

    Moreno, A.; Sari, C.

    1978-01-01

    Radial gradients of oxygen to metal ratio, O/M, in uranium-plutonium oxide fuel pins, during irradiation and at the end of life, have been calculated on the basis of solid-state thermal diffusion using measured values of the heat of transport. A detailed computer model which includes the calculation of temperature profiles and the variation of the average O/M ratio as a function of burn-up is given. Calculations show that oxygen profiles are affected by the isotopic composition of the fuel, by the temperature profiles and by fuel-cladding interactions

  13. Oxidative destruction of ammonia for restoration of uranium solution mining sites

    International Nuclear Information System (INIS)

    Humenick, M.J.; Garwacka, K.

    1984-01-01

    A laboratory experimental research project was conducted to evaluate the use of chlorine for the oxidative destruction of residual ammonia that may remain in ground water after in-situ uranium solution mining operations. The work tested the idea of injecting high strength calcium hypochlorite solution into the mining zone to convert ammonia to nitrogen gas as a final cleanup process for ammonia removal from the ground water system. This paper details ammonia removal efficiency as a function of chlorine dose, reactant, and product material balances, and how the concept may be used as a final ground water restoration process

  14. Adsorption of uranium composites onto saltrock oxides - experimental and theoretical study.

    Science.gov (United States)

    Ivanova, Bojidarka; Spiteller, Michael

    2014-09-01

    The study encompassed experimental mass spectrometric and theoretical quantum chemical studies on adsorption of uranium species in different oxidation states of the metal ion, and oxides of UxOy(n+) type, where x = 1 or 3, y = 2 or 8, and n = 0, 1 or 2 onto nanosize-particles of saltrock oxides MO (M = Mg(II), Ca(II), Ni(II), Co(II), Sr(II) or Ba(II)), M2Oy (M = Au(III) or Ag(I), y = 3 or 1) silicates 3Al2O3.2SiO2, natural kaolinite (Al2O2·2SiO2·2H2O), illite (K0.78Ca0.02Na0.02(Mg0.34Al1.69Fe(III)0.02)[Si3.35Al0.65]O10(OH)2·nH2O), CaSiO3, 3MgO·4SiO2,H2O, and M(1)M(2)(SiO4)X2 (M(1) = M(2) = Al or M(1) = K, M(2) = Al, X = F or Cl), respectively. The UV-MALDI-Orbitrap mass spectrometry was utilized in solid-state and semi-liquid colloidal state, involving the laser ablation at λex = 337.2 nm. The theoretical modeling and experimental design was based on chemical-, physico-chemical, physical and biological processes involving uranium species under environmental conditions. Therefore, the results reported are crucial for quality control and monitoring programs for assessment of radionuclide migration. They impact significantly the methodology for evaluation of human health risk from radioactive contamination. The study has importance for understanding the coordination and red-ox chemistry of uranium compounds as well. Due to the double nature of uranium between rare element and superconductivity like materials as well as variety of oxidation states ∈ (+1)-(+6), the there remain challenging areas for theoretical and experimental research, which are of significant importance for management of nuclear fuel cycles and waste storage. Copyright © 2014 Elsevier Ltd. All rights reserved.

  15. Preliminary testing of a planar converter with uranium oxide pellets in the emitter

    International Nuclear Information System (INIS)

    Miskolczy, G.; Lieb, D.P.; Hatch, G.L.

    1992-01-01

    Nuclear reactor thermionic space power systems incorporating thermionic fuel element generally use refractory metal emitters, which contain the nuclear fuel. The purpose of the current work is to determine the effect, if any, of the diffusion of uranium oxide fuel through chemically vapor deposited (CVD) tungsten on converter performance. This paper describes the preliminary testing of the converter to assess the converter performance before any significant diffusion takes place. In testing, the emitter temperature was 1800 K and the collector temperature was varied from 1000 K to 1070 K. Experiments also examined pressure versus loading characteristics of the graphite

  16. Chlorination of uranium oxides in melts of alkali metal chlorides and their mixtures

    International Nuclear Information System (INIS)

    Vorobej, M.P.; Bevz, A.S.; Skiba, O.V.

    1978-01-01

    Chlorination of UO 2 , U 3 O 8 , and UO 3 in melts of chlorides of alkali metals and of their mixtures has been studied by thermogravimetric, X-ray phase, and spectrophotometric methods. The thermogravimetric method has been proposed for evaluating the state of uranylcation in the melt; the effect of the composition of the oxide being chlorinated and of the salt-solvent on the composition of the chlorination products has been studied. The effect of the composition of the chlorination products on the stoichiometry of the electrolytic uranium dioxide has been shown

  17. Oxidative destruction of ammonia for restoration of uranium solution mining sites

    Energy Technology Data Exchange (ETDEWEB)

    Humenick, M.J.; Garwacka, K.

    1984-02-01

    A laboratory experimental research project was conducted to evaluate the use of chlorine for the oxidative destruction of residual ammonia that may remain in ground water after in-situ uranium solution mining operations. The work tested the idea of injecting high strength calcium hypochlorite solution into the mining zone to convert ammonia to nitrogen gas as a final cleanup process for ammonia removal from the ground water system. This paper details ammonia removal efficiency as a function of chlorine dose, reactant, and product material balances, and how the concept may be used as a final ground water restoration process.

  18. Oxidative destruction of ammonia for restoration of uranium solution mining sites

    Energy Technology Data Exchange (ETDEWEB)

    Humenick, M.J.; Garwacka, K.

    1984-01-01

    A laboratory experimental research project was conducted to evaluate the use of chlorine for the oxidative destruction of residual ammonia that may remain in ground water after in-situ uranium solution mining operations. The work tested the idea of injecting high strength calcium hypochlorite solution into the mining zone to convert ammonia to nitrogen gas as a final cleanup process for ammonia removal from the ground water system. This paper details ammonia removal efficiency as a function of chlorine dose, reactant, and product material balances, and how the concept may be used as a final ground water restoration process.

  19. Standard guide for the determination of iodine-129 In uranium oxide

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2006-01-01

    1.1 This method covers the determination of iodine-129 (129I) in uranium oxide by gamma-ray spectrometry. The method could also be applicable to the determination of 129I in aqueous matrices. 1.2 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and to determine the applicability of regulatory limitations prior to use.

  20. Standard test methods for chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade mixed oxides ((U, Pu)O2)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 These test methods cover procedures for the chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade mixed oxides, (U, Pu)O2, powders and pellets to determine compliance with specifications. 1.2 The analytical procedures appear in the following order: Sections Uranium in the Presence of Pu by Potentiometric Titration Plutonium by Controlled-Potential Coulometry Plutonium by Amperometric Titration with Iron (II) Nitrogen by Distillation Spectrophotometry Using Nessler Reagent 7 to 14 Carbon (Total) by Direct Combustion-Thermal Conductivity 15 to 26 Total Chlorine and Fluorine by Pyrohydrolysis 27 to 34 Sulfur by Distillation-Spectrophotometry 35 to 43 Moisture by the Coulometric, Electrolytic Moisture Analyzer 44 to 51 Isotopic Composition by Mass Spectrometry Rare Earths by Copper Spark Spectroscopy 52 to 59 Trace Impurities by Carrier Distillation Spectroscopy 60 to 69 Impurities by Spark-Source Mass Spectrography 70 to 76 Total Gas in Reactor-Grade Mixed Dioxide P...

  1. Uranium oxidation kinetics monitored by in-situ X-ray diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Zalkind, S., E-mail: shimonzl@nrcn.org.il; Rafailov, G.; Halevy, I.; Livneh, T.; Rubin, A.; Maimon, H.; Schweke, D.

    2017-03-15

    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 UO{sub 2}(111) is the prominent one. As the oxide thickens, the growth rate of UO{sub 2}(220) plane increases and both planes grow concurrently. The activation energies obtained for the oxide growth are Q{sub parabolic} = 17.5 kcal/mol and Q{sub linear} = 19 kcal/mol. Enhanced oxidation around uranium carbide (UC) inclusions is clearly observed by scanning electron microscopy (SEM).

  2. A review of the environmental corrosion, fate and bioavailability of munitions grade depleted uranium

    Energy Technology Data Exchange (ETDEWEB)

    Handley-Sidhu, Stephanie, E-mail: s.handley-sidhu@bham.ac.uk [Water Sciences Research Group, School of Geography, Earth, Environmental Sciences, University of Birmingham, Edgbaston, Birmingham B15 2TT (United Kingdom); Keith-Roach, Miranda J. [Biogeochemistry and Environmental Analytical Chemistry Research Group, and School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom); Lloyd, Jonathan R.; Vaughan, David J. [Williamson Research Centre for Molecular Environmental Science, and School of Earth, Atmospheric and Environmental Sciences, University of Manchester, Manchester M13 9PL (United Kingdom)

    2010-11-01

    Depleted uranium (DU) is a by-product of nuclear fuel enrichment and is used in antitank penetrators due to its high density, self-sharpening, and pyrophoric properties. Military activities have left a legacy of DU waste in terrestrial and marine environments, and there have been only limited attempts to clean up affected environments. Ten years ago, very little information was available on the dispersion of DU as penetrators hit their targets or the fate of DU penetrators left behind in environmental systems. However, the marked increase in research since then has improved our knowledge of the environmental impact of firing DU and the factors that control the corrosion of DU and its subsequent migration through the environment. In this paper, the literature is reviewed and consolidated to provide a detailed overview of the current understanding of the environmental behaviour of DU and to highlight areas that need further consideration.

  3. Uranium Immobilization through Fe(II) bio-oxidation: A Column study

    Energy Technology Data Exchange (ETDEWEB)

    Coates, John D.

    2009-09-14

    Current research on the bioremediation of heavy metals and radionuclides is focused on the ability of reducing organisms to use these metals as alternative electron acceptors in the absence of oxygen and thus precipitate them out of solution. However, many aspects of this proposed scheme need to be resolved, not the least of which is the time frame of the treatment process. Once treatment is complete and the electron donor addition is halted, the system will ultimately revert back to an oxic state and potentially result in the abiotic reoxidation and remobilization of the immobilized metals. In addition, the possibility exists that the presence of more electropositive electron acceptors such as nitrate or oxygen will also stimulate the biological oxidation and remobilization of these contaminants. The selective nitrate-dependent biooxidation of added Fe(II) may offer an effective means of “capping off” and completing the attenuation of these contaminants in a reducing environment making the contaminants less accessible to abiotic and biotic reactions and allowing the system to naturally revert to an oxic state. Our previous DOE-NABIR funded studies demonstrated that radionuclides such as uranium and cobalt are rapidly removed from solution during the biogenic formation of Fe(III)-oxides. In the case of uranium, X-ray spectroscopy analysis indicated that the uranium was in the hexavalent form (normally soluble) and was bound to the precipitated Fe(III)-oxides thus demonstrating the bioremediative potential of this process. We also demonstrated that nitrate-dependent Fe(II)- oxidizing bacteria are prevalent in the sediment and groundwater samples collected from sites 1 and 2 and the background site of the NABIR FRC in Oakridge, TN. However, all of these studies were performed in batch experiments in the laboratory with pure cultures and although a significant amount was learned about the microbiology of nitrate-dependent bio-oxidation of Fe(II), the effects of

  4. Heap-leaching of low-grade uranium ore at SOMAIR: from laboratory tests to production of 700 tonnes U per year

    Energy Technology Data Exchange (ETDEWEB)

    Durupt, N.; Blanvillain, J.J., E-mail: nicolas.durupt@areva.com [AREVA NC, Service d' Etudes de Procedes et d' Analyses (SEPA), Bessines sur Gartempe (France)

    2010-07-01

    In 2006, SOMAIR decided to increase the uranium production by 50% using heap leaching for the treatment of low grade ores. These ores, which come from different ores with various properties, have been studied in four steps: Lab tests: to compare the ores (characterization, acid consumptions, recovery); Column tests on an average sample: to define significant parameters for a feasibility study; Column tests on specific samples: to optimize recovery for each ore and identify problems of percolation due to the clays; and, Pilot tests in large boxes (stalls): to validate process parameters. Uranium production by heap leaching started commercially in July 2009. (author)

  5. Whole-body fat oxidation determined by graded exercise and indirect calorimetry: a role for muscle oxidative capacity?

    DEFF Research Database (Denmark)

    Nordby, P; Saltin, B; Helge, J W

    2006-01-01

    During whole-body exercise, peak fat oxidation occurs at a moderate intensity. This study investigated whole-body peak fat oxidation in untrained and trained subjects, and the presence of a relation between skeletal muscle oxidative enzyme activity and whole-body peak fat oxidation. Healthy male...... subjects were recruited and categorized into an untrained (N=8, VO(2max) 3.5+/-0.1 L/min) and a trained (N=8, VO(2max) 4.6+/-0.2 L/min) group. Subjects performed a graded exercise test commencing at 60 W for 8 min followed by 35 W increments every 3 min. On a separate day, muscle biopsies were obtained...... oxidation was determined. The body composition was determined by DEXA. Whole-body peak fat oxidation (250+/-25 and 462+/-33 mg/min) was higher (Ptrained compared with untrained subjects, respectively. Muscle...

  6. Standard Specification for Nuclear Grade Zirconium Oxide Pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This specification applies to pellets of stabilized zirconium oxide used in nuclear reactors. 1.2 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.

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

    Energy Technology Data Exchange (ETDEWEB)

    Vanhoudt, Nathalie, E-mail: nvanhoud@sckcen.be [Belgian Nuclear Research Center (SCK-CEN), Biosphere Impact Studies, Boeretang 200, 2400 Mol (Belgium); Hasselt University, Environmental Biology, Centre for Environmental Sciences, Agoralaan Building D, 3590 Diepenbeek (Belgium); Cuypers, Ann [Hasselt University, Environmental Biology, Centre for Environmental Sciences, Agoralaan Building D, 3590 Diepenbeek (Belgium); Horemans, Nele [Belgian Nuclear Research Center (SCK-CEN), Biosphere Impact Studies, Boeretang 200, 2400 Mol (Belgium); Remans, Tony; Opdenakker, Kelly; Smeets, Karen [Hasselt University, Environmental Biology, Centre for Environmental Sciences, Agoralaan Building D, 3590 Diepenbeek (Belgium); Bello, Daniel Martinez [Hasselt University, Interuniversity Institute for Biostatistics and Statistical Bioinformatics, Agoralaan Building D, 3590 Diepenbeek (Belgium); Havaux, Michel [Commissariat a l' Energie Atomique (CEA)/Cadarache, Direction des Sciences du Vivant, Departement d' Ecophysiologie Vegetale et de Microbiologie, Laboratoire d' Ecophysiologie de la Photosynthese, 13108 Saint-Paul-lez-Durance (France); Wannijn, Jean; Van Hees, May [Belgian Nuclear Research Center (SCK-CEN), Biosphere Impact Studies, Boeretang 200, 2400 Mol (Belgium); Vangronsveld, Jaco [Hasselt University, Environmental Biology, Centre for Environmental Sciences, Agoralaan Building D, 3590 Diepenbeek (Belgium); Vandenhove, Hildegarde [Belgian Nuclear Research Center (SCK-CEN), Biosphere Impact Studies, Boeretang 200, 2400 Mol (Belgium)

    2011-06-15

    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 {mu}M 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 {mu}M 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

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

    Science.gov (United States)

    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

  9. Recovery of uranium low grade ores by froth flotation: study of the texture and synergetic effects of flotation reagents

    International Nuclear Information System (INIS)

    Duverger, Agathe

    2013-01-01

    Due to the energy growing demand, uranium low grade ores may be those exploited in the future. Uranium ores conventional treatment does not often use mineral processing such as concentration methods for reducing leaching reagent consumption. The aim of this work is to develop an upgrading process to improve the operating process (alkaline heap leaching) taking into account the mineralogical and textural variability of the ore. The Trekkopje deposit is composed of calcrete and a gypscrete. The uranium bearing mineral is carnotite (K 2 (UO 2 ) 2 [VO 4 ] 2 .3H 2 O). The gangue minerals are composed by silicates, such as quartz, feldspars, micas and Ca-minerals, calcite and gypsum (XRD and ICP-MS analysis). A SEM image processing was used to study the textural properties and the exposed free surface of mineral inclusions in clay clusters. In calcrete milled to -200 μm, 50 % of all carnotite is associated with clay clusters, which are composed by 98 % of palygorskite, 2 % of illite, montmorillonite, and interbedded clays (XRD and microprobe analysis). The carnotite grain size is 95 % less than 70 μm. Calcite is the main inclusion in clay clusters. Indeed, the calcite inclusions average rate in the clay clusters is 12 % and 5 % for carnotite inclusion. And the free exposed surface percentage of these minerals in clay clusters is 3 % and 6 %, thus indicating that the inclusions should not affect the behavior of mixed clay particles. However, ore flotation essays did not verify this hypothesis. Three minerals separation have been proposed based on the mineral ability to consume leaching reagents: separating Ca-minerals from silicates, palygorskite from gangue minerals and carnotite from gangue minerals. A study of silicates and Ca-minerals electrokinetic properties (electrophoresis) was carried out to select the collectors and the optimum pH range for selective flotation. Basic pH near neutral was proved to be optimal for the separation of gangue minerals with cationic

  10. Influent of Carbonization of Sol Solution at the External Gelation Process on the Quality of Uranium Oxide Kernel

    International Nuclear Information System (INIS)

    Damunir; Sukarsono

    2007-01-01

    The influent of carbonization of sol solution at the external gelation process on the quality of uranium oxide kernel was done. Variables observed are the influent of carbon, temperature and time of reduction process of U 3 O 8 kernel resulted from carbonization of sol solution. First of all, uranyl nitrate was reacted with 1 M NH 4 OH solution, producing the colloid of UO 3 . Then by mixing and heating up to the temperature of 60-80 °C, the colloid solution was reacted with PVA, mono sorbitol oleate and paraffin producing of uranium-PVA sol. Then sol solution was carbonized with carbon black of mol ratio of carbon to uranium =2.32-6.62, produce of carbide gel. Gel then washed, dried and calcined at 800 °C for 4 hours to produce of U 3 O 8 kernel containing carbon. Then the kernel was reduced by H 2 gas in the medium of N 2 gas at 500-800 °C, 50 mmHg pressure for 3 hours. The process was repeated at 700 °C, 50 mmHg pressure for 1-4 hours. The characterization of chemical properties of the gel grains and uranium oxide kernel using FTIR covering the analysis of absorption band of infra red spectrum of UO 3 , C-OH, NH 3 , C-C, C-H and OH functional group. The physical properties of uranium oxide covering specific surface area, void volume, mean diameter using surface area meter Nova-1000 and as N 2 gas an absorbent. And O/U ratio of uranium dioxide kernel by gravimetry method. The result of experiment showed that carbonization of sol solution at the external gelation process give influencing the quality of uranium oxide kernel. (author)

  11. Standard specification for nuclear-grade zirconium oxide powder

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This specification defines the physical and chemical requirements for zirconium oxide powder intended for fabrication into shapes, either entirely or partially of zirconia, for use in a nuclear reactor core. 1.2 The material described herein shall be particulate in nature. 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

  12. Characteristics of interlayer oxidation zone and uranium metallogenetic prospect of Zhiluo formation in Daliuta area, Ordos Basin

    International Nuclear Information System (INIS)

    Yang Jianxin; Li Xide; Zhang Zhaolin

    2006-01-01

    Ordos Basin is a large down-warping basin in the northwest of North-China Platform; Daliuta area is located in the northeast of Ordos Basin. In this area, sand bodies of fluvial facies developed well in the submember of the lower member of the target Zhiluo Formation of Middle Jurassic and several sand belt of large scale occurred. Yellow interlayer oxidation zone have been discovered in belt I and belt III by the drilling and it is of a certain scale. Due to the young age of interlayer oxidation and unsatisfied uranium sources, uranium metallogenic prospect of this area need more research and exploration. (authors)

  13. Towards uranium dose-effects relationships for bio markers of oxidative stress in fish

    International Nuclear Information System (INIS)

    Buet, A.; Camilleri, V.; Simon, O.; Garnier-Laplace, J.

    2004-01-01

    Within the Envirhom program, the bioavailability of uranium, a widely spread metal in fresh waters, has been studied in various organisms to gain understanding of metal-organisms interactions. Experiments are still in progress to establish a comprehensive basis of early and delayed involved toxicological mechanisms. Uranium is known to be both radio-toxic and chemo-toxic. The cellular damage of radiation, but also of heavy metal exposure, is mainly associated with an oxidative injury due to the formation of reactive oxygen species. Moreover, the intensity of oxidative damages is dependant on the efficiency of antioxidant defense systems of the organism. In this context, short-term experiments were performed with juvenile rainbow trout (Onchorynchus mykiss) in order to (1) assess the response of some antioxidant parameters and other potential bio-markers and (2) to better characterize the kinetics of the responses in relation with U concentrations and exposure duration. Trout were exposed by direct pathway to a range of U concentrations in water (low, medium and high: 20, 100 and 500 μg U.L -1 respectively) during 10 days. Several antioxidant parameters were measured: the rate of reduced glutathione (GSH) that plays a major role in cellular detoxification and antioxidant defense, and the activities of superoxide dismutases (SOD) and catalase that are involved in the detoxification of oxygen reactive species. The activity of glutathione reductase (GR), that restores the pool of GSH was measured, as well as the activity of acetylcholinesterase (AChE), a marker of neurotoxicity. In parallel, U analysis were performed in gill, skin, muscle, skeleton, intestine, liver, kidneys and body residues of exposed trout in order to assess the dependence of biological responses with a potential uranium bioaccumulation in fish tissues. (author)

  14. URANIUM OXIDE-CONTAINING FUEL ELEMENT COMPOSITION AND METHOD OF MAKING SAME

    Science.gov (United States)

    Handwerk, J.H.; Noland, R.A.; Walker, D.E.

    1957-09-10

    In the past, bodies formed of a mixture of uranium dioxide and aluminum powder have been used in fuel elements; however, these mixtures were found not to be suitable when exposed to temperatures of about 600 deg C, because at such high temperatures the fuel elements were distorted. If uranosic oxide, U/sub 3/O/sub 8/, is substituted for UO/sub 2/, the mechanical properties are not impaired when these materials are used at about 600 deg C and no distortion takes place. The uranosic oxide and aluminum, both in powder form, are first mixed, and after a homogeneous mixture has been obtained, are shaped into fuel elements by extrusion at elevated temperature. Magnesium powder may be used in place of the aluminum.

  15. Tables of thermodynamic functions for gaseous thorium, uranium, and plutonium oxides

    International Nuclear Information System (INIS)

    Green, D.W.

    1980-03-01

    Measured and estimated spectroscopic data for thorium, uranium, and plutonium oxide vapor species have been used with the methods of statistical mechanics to calculate thermodynamic functions. Some inconsistencies between spectroscopic data and some thermodynamic data have been resolved by recalculating ΔH 0 /sub f/ (298.15 0 K) values for the vapor species of these oxides. Evaluation of the uncertainties in data, methods of estimating molecular parameters, and effects of assumptions have been discussed elsewhere. The tables of thermodynamic functions that were reported earlier have been revised principally because the low-frequency vibrational modes of UO 2 and UO 3 have now been measured. These new empirical data resulted in changes in the electronic contributions to the calculated thermodynamic functions of UO 2 and the estimated vibrational contributions for PuO 2 . In addition, some minor changes have been made in the methods of calculation of the electronic contributions for all molecules

  16. Implication of volume changes in uranium oxides: A density functional study

    International Nuclear Information System (INIS)

    Szpunar, B.; Szpunar, J.A.; Milman, V.; Goldberg, A.

    2013-01-01

    In severe nuclear accident scenarios (in air environments and high temperatures) UO 2 fuel pellets oxidise to produce uranium oxides with higher oxygen content, e.g., U 4 O 9 or U 3 O 8 . As a first step in investigating the microstructural changes following UO 2 oxidation to hexagonal high temperature phase of U 3 O 8 , density functional quantum mechanical calculations of the structure, elastic properties and electronic structure of U 3 O 8 have been performed. The calculated properties of hexagonal phase of U 3 O 8 are compared to those of the orthorhombic pseudo-hexagonal phase which is stable at room temperature. The total energy technique based on the local density approximation plus Hubbard U as implemented in the CASTEP code is used to investigate changes in the lattice constants. The first-principles calculations predict a 35-42% increase in volume per uranium atom as a result of the transformation from UO 2 to U 3 O 8 , in agreement with experimental data. The implications of this prediction on the linear expansion and fragmentation of fuel are discussed. (authors)

  17. Inorganic oxides as alternative in the separation of non fissioned residual uranium

    International Nuclear Information System (INIS)

    Baca G, A.

    1997-01-01

    The Al 2 O 3 , SiO 2 and SnO 2 as well as vegetable carbon have been studied for its possible use as sorbent in the concentration and separation of non fissioned residual uranium of some fission products such as: 141 Ce, 134 Cs, 125 Sb, 103 Ru, 95 Zr, 95 Nb of alkaline aqueous systems. The separation efficiency has been evaluated using natural uranium and radionuclides in static and dynamic processes, through liquid scintillation and gamma spectrometry. Therefore Al 2 O 3 , SiO 2 , SnO 2 and carbon were pre-treated thermic and chemically and characterized through the technique of Nitrogen absorption analysis, X-ray diffraction and IR spectroscopy. By means of the p H determination and the aqueous system potential the present hydrolysis products were determined. The inorganic oxides show structural and surface changes due to the treatment. The adsorption process is realized by different mechanism depending of the sorbent. The results show that the retention capacity is a dependence of the oxides pre-treatment and of the hydrolysis products in the aqueous system, as well as of the experimental conditions. Not in this way for carbon in which the results show the treatment and the experimental conditions significantly have not influence in its adsorption capacity. (Author)

  18. Mineralogical-geochemical specificity of the uranium mineralization superposed on the oxidized rocks

    International Nuclear Information System (INIS)

    Bulatov, S.G.; Shchetochkin, V.N.

    1975-01-01

    Taking as an example a uranium deposit connected with oxidation zones developing along the strata, the author examines the mineralogical and geochemical features of a pitchblende-sooty uraninite mineralization superimposed on limonitized sandstones. The typical relations between ore mineralization with new formations of the infiltration oxidation process and the changes caused by the action of rising thermal solutions on the rocks are given. Based on these relations, two generations of different ages of rich pitchblende-sooty uraninite ores are distinguished, separated by the time of development of the oxidation processes. The typical change around the ore is a reduction of limonitized rocks, accompanied by their pyritization, clarification and hematitization. The ore concentrations were formed as a result of the action of rising thermal solutions that had interacted with oxidized rocks. The development of late oxidation processes caused the redistribution of these ore concentrations and their downward shift along the stratum slope following the limonitization boundary. On the basis of the data presented, comments of a forecasting and prospecting nature are made. (author)

  19. A study of integrated cathode assembly for electrolytic reduction of uranium oxide in LiCl-Li2O molten salt

    International Nuclear Information System (INIS)

    Park, Sung Bin; Seo, Jung Seok; Kang, Dae Seung; Kwon, Sun Kil; Park, Seong Won

    2004-01-01

    Interest of electrolytic reduction of uranium oxide is increasing in treatment of spent metal fuels. Argonne National Laboratory (ANL) has reported the experimental results of electrochemical reduction of uranium oxide fuel in bench-scale apparatus with cyclic voltammetry, and has designed high-capacity reduction (HCR) cells and conducted three kg-scale UO 2 reduction runs. From the cyclic voltammograms, the mechanism of electrolytic reduction of metal oxides is analyzed. The uranium oxide in LiCl-Li 2 O is converted to uranium metal according to the two mechanism; direct and indirect electrolytic reduction. In this study, cyclic voltammograms for LiCl-3wt% Li 2 O system and U 3 O 8 -LiCl-3wt% Li 2 O system using the 325-mesh stainless steel screen in cathode assembly have been obtained. Direct electrolytic reduction of uranium oxide in LiCl-3wt% Li 2 O molten salt has been conducted

  20. Characteristics of an integrated cathode assembly for the electrolytic reduction of uranium oxide in a LiCl-Li2O molten salt

    International Nuclear Information System (INIS)

    Sung Bin Park; Byung Heung Park; Sang Mun Jeong; Jin Mok Hur; Chung Seok Seo; Seong Won Park; Seung-Hoon Choi

    2006-01-01

    Electrochemical behavior of the reduction of uranium oxide was studied in a LiCl-Li 2 O molten salt system with an integrated cathode assembly. The mechanism for the electrolytic reduction of uranium oxide was studied through cyclic voltammetry. By means of a chronopotentiometry, the effects of the thickness of the uranium oxide, the thickness of the MgO membrane and the material of the conductor of an integrated cathode assembly on the overpotential of the cathode were investigated. From the voltamograms, the reduction potential of the uranium oxide and Li 2 O was obtained and the two mechanisms of the electrolytic reduction were considered with regard to the applied cathode potential. From the chronopotentiograms, the exchange current, the transfer coefficient and the maximum allowable current based on the Tafel behavior were obtained with regard to the thickness of the uranium oxide, and of the MgO membrane and the material of the conductor of an integrated cathode assembly. (author)

  1. The association between low-grade inflammation, iron status and nucleic acid oxidation in the elderly

    DEFF Research Database (Denmark)

    Broedbaek, Kasper; Siersma, Volkert Dirk; Andersen, Jon T

    2011-01-01

    This study applied a case-control approach to investigate the association between low-grade inflammation, defined by high values within the normal range of C-reactive protein (CRP) and interleukin-6 (IL-6), and urinary markers of nucleic acid oxidation. No differences in excretion of urinary...... markers of nucleic acid oxidation between cases and controls were found and multivariable linear regression analysis showed no association between urinary markers of nucleic acid oxidation and inflammatory markers. Post-hoc multivariable linear regression analysis showed significant associations between...... suggest that low-grade inflammation only has a negligible impact on whole body nucleic acid oxidation, whereas iron status seems to be of great importance....

  2. The combined effect of uranium and gamma radiation on biological responses and oxidative stress induced in Arabidopsis thaliana

    International Nuclear Information System (INIS)

    Vanhoudt, Nathalie; Vandenhove, Hildegarde; Horemans, Nele; Wannijn, Jean; Van Hees, May; Vangronsveld, Jaco; Cuypers, Ann

    2010-01-01

    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 μM 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.

  3. Gamma-induced radiation polymerization of kaolin composite for sorption of lanthanum, europium and uranium ions from low-grade monazite leachate

    International Nuclear Information System (INIS)

    Metwally, S.S.; Hassan, R.S.; El-Masry, E.H.; Borai, E.H.

    2018-01-01

    Gamma radiation polymerization method was used for the modification of kaolin to produce (poly acrylamide-acrylic acid)-Kaolin (PAM-AA-K). Monazite ore is one of the main resources of uranium and lanthanide elements, therefore, this work focused on sorption of uranium, lanthanum and europium ions from low grade monazite leachate. The removal percent for Eu 3+ , La 3+ and UO 2 2+ are 94.6, 91.6 and 73.4%, respectively. Monolayer capacity of Eu 3+ , La 3+ and UO 2 2+ were found to be 54.64, 45.87 and 37.59 mg/g, respectively. The sorption mechanism of lanthanum and europium ions on PAM-AA-K composite mainly takes place as Ln(OH) 2+ , and for uranium as uranyl ion, UO 2 2+ . (author)

  4. Standard specification for nuclear-grade aluminum oxide pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This specification applies to pellets of aluminum oxide that may be ultimately used in a reactor core, for example, as filler or spacers within fuel, burnable poison, or control rods. In order to distinguish between the subject pellets and “burnable poison” pellets, it is established that the subject pellets are not intended to be used as neutron-absorbing material. 1.2 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard.

  5. Red blood cells sensitivity to oxidative stress in the presence of low concentrations of uranium compound

    Energy Technology Data Exchange (ETDEWEB)

    Shevchenko, O.G. [Institute of Biology, Komi Scientific Centre, Ural Branch of Russian Academy of Sciences, 167982, Syktyvkar (Russian Federation)

    2014-07-01

    Uranium is a natural radioactive element widespread in biosphere. There are a few works that examined cellular and molecular mechanisms of uranium toxicity. Red blood cells are classical model to investigate toxicity mechanisms on cell membrane system. The aim of present work is to study the effect of uranyl ion in nano-molar concentrations on erythrocytes sensitivity (in vitro) to factors provoking acute oxidative stress. Uranyl ions were added to suspension of mice red blood cells in PBS as UO{sub 2}Cl{sub 2} solution. Samples were incubated in a thermostatic shaker at 37 deg. C during 3-5 hours. Than acute oxidative stress was induced by H{sub 2}O{sub 2} (0.9 mM) or AAPH (5 mM) solutions. Destabilization of the membrane was induced by nonionic detergent Triton X-100. The hemolysis degree and the content of LPO secondary products reacting with 2-thiobarbituric acid in the incubation mixture were determined spectrophotometrically. The ratio of hemoglobin various forms (oxyHb, metHb and ferrylHb) was calculated taking into account extinction coefficients. It was shown that uranyl chloride enhances cell sensitivity to nonionic detergent Triton X-100 effects, indicating alterations of membrane acyl chain order due to contact with the radionuclide ions. Uranium exposure also caused an increase in the cell sensitivity to the AAPH effects, resulted in a decrease in red cell survival rate, a sharp increase in accumulation of hemoglobin oxidation products and a slight increase in the concentration of LPO secondary products. Thus, uranyl ions change physicochemical properties of the erythrocyte membranes that resulted in increased sensitivity to effects of peroxyl radicals formed by thermal decomposition of AAPH. On the contrary, use of another source of free radicals - H{sub 2}O{sub 2} - after uranyl ions exposure resulted in marked decrease of oxidative hemolysis, inhibition of LPO and hemoglobin oxidation. Since the uranium chemical properties similar to properties of

  6. Electrochemical behavior for a reduction of uranium oxide in a LiCl-Li2O molten salt with an integrated cathode assembly

    International Nuclear Information System (INIS)

    Park, Sung Bin; Park, Byung Heung; Seo, Chung Seok; Jung, Ki Jung; Park, Seong Won

    2005-01-01

    Electrolytic reduction of uranium oxide to uranium metal was studied in a LiCl-Li 2 O molten salt system. The reduction mechanism of the uranium oxide to a uranium metal has been studied by means of a cyclic voltammetry. Effects of the layer thickness of the uranium oxide and the thickness of the MgO on the overpotential of the cathode and the anode were investigated by means of a chronopotentiometry. From the cyclic voltamograms, the decomposition potentials of the metal oxides are the determining factors for the mechanism of the reduction of the uranium oxide in a Li Cl-3 wt% Li 2 O molten salt and the two mechanisms of the electrolytic reduction were considered with regards to the applied cathode potential. In the chronopotentiograms, the exchange current and the transfer coefficient based on the Tafel behavior were obtained with regard to the layer thickness of the uranium oxide which is loaded into the porous MgO membrane and the thickness of the porous MgO membrane. The maximum allowable currents for the changes of the layer thickness of the uranium oxide and the thickness of the MgO membrane were also obtained from the limiting potential which is the decomposition potential of LiCl

  7. Radiation dose control in the mining of high grade uranium ores

    International Nuclear Information System (INIS)

    Webster, S.T.; Brown, L.D.

    1996-01-01

    The control of radiation doses received by uranium miners is an unusually complex procedure, as three separate components of their total effective dose may be significant and may have to be evaluated separately. Apart from external and internal doses evaluated in the usual way, it is also necessary to evaluate the inhalation dose from radon progeny separately. Although this essentially forms part of the internal dose received, it is not evaluated in the conventional way since the associated dose equivalent must be derived from conversion factors based on epidemiological studies, instead of by the usual approach of calculating the dose to tissue from the inhaled activity and multiplying this by a recognized conversion factor to derive a whole body effective dose. Historically the traditional unit used for monitoring the concentration of radon progeny in a workplace is the Working Level (WL), this is now defined as a concentration such that the potential alpha emission from all the short lived progeny present in the sample will total 1.3 x 108 MeV per m 3 . The corresponding unit of exposure is the Working Level Month (WLM) and is the exposure that would be received by a reference man working in such an atmosphere for a standard working month lasting 170 hours. Unfortunately the relationship between exposures, measured in WLM, and the conventional radiation dose to the target tissues is complex and calculated values depend greatly upon the assumptions made in the lung model that must be used. Risks are therefore still controlled by limiting exposures in WLM on the basis of epidemiological studies of lung cancer incidence among miners employed at a time when the magnitude of the risk was not fully appreciated, and cancer incidence was high enough to permit reasonably accurate risk estimates to be derived directly from exposures in WLM. (author)

  8. Status report from USSR [Processing of Low-Grade Uranium Ores]; Doklad o sostoyanii voprosa v SSSR

    Energy Technology Data Exchange (ETDEWEB)

    Zefirov, A P [Gosudarstvennyj Komitet Po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moskva, Union of Soviet Socialist Republics (Russian Federation)

    1967-06-15

    The uranium industry for processing poor uranium ores in the USSR was established in recent years. As a result of research work institutions and enterprises in the development of this industry was provided by rapid technological advances that allowed dramatically increased productivity, reduced consumption of reagents, simplified process flow diagrams, and reduced production costs. At present, the basis for uranium industry, including and poor uranium ore deposits in the USSR are with different content valuable components (uranium, phosphorus, molybdenum, rare earth elements, thorium, iron, .. .)

  9. Development of functionally graded anti-oxidation coatings for carbon/carbon composites

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, J.H. [Dept. of Materials Technology, Korea Inst. of Machinery and Materials, Changwon (Korea); Fang Hai-Tao; Lai Zhong-Hong; Yin Zhong-Da [Materials Science and Engineering School, Harbin Inst. of Tech., Harbin (China)

    2005-07-01

    The concept of functionally graded materials (FGMs) was originated in the research field of thermal barrier coatings. Continuous changes in the composition, grain size, porosity, etc., of these materials result in gradients in such properties as mechanical strength and thermal conductivity. In recent years, functionally graded structural composite materials have received increased attention as promising candidate materials to exhibit better mechanical and functional properties than homogeneous materials or simple composite materials. Therefore the research area of FGMs has been expending in the development of various structural and functional materials, such as cutting tools, photonic crystals, dielectric and piezoelectric ceramics, thermoelectric semiconductors, and biomaterials. We have developed functionally graded structural ceramic/metal composite materials for relaxation of thermal stress, functionally graded anti-oxidation coatings for carbon/carbon composites, and functionally graded dielectric ceramic composites to develop advanced dielectric ceramics with flat characteristics of dielectric constant in a wide temperature range. This paper introduces functionally graded coatings for C/C composites with superior oxidation resistance at high temperatures. (orig.)

  10. Process development study on production of uranium metal from monazite sourced crude uranium tetra-fluoride

    International Nuclear Information System (INIS)

    Chowdhury, S; Satpati, S.K.; Hareendran, K.N.; Roy, S.B.

    2014-01-01

    Development of an economic process for recovery, process flow sheet development, purification and further conversion to nuclear grade uranium metal from the crude UF 4 has been a technological challenge and the present paper, discusses the same.The developed flow-sheet is a combination of hydrometallurgical and pyrometallurgical processes. Crude UF 4 is converted to uranium di-oxide (UO 2 ) by chemical conversion route and UO 2 produced is made fluoride-free by repeated repulping, followed by solid liquid separation. Uranium di-oxide is then purified by two stages of dissolution and suitable solvent extraction methods to get uranium nitrate pure solution (UNPS). UNPS is then precipitated with air diluted ammonia in a leak tight stirred vessel under controlled operational conditions to obtain ammonium di-uranate (ADU). The ADU is then calcined and reduced to produce metal grade UO 2 followed by hydro-fluorination using anhydrous hydrofluoric acid to obtain metal grade UF 4 with ammonium oxalate insoluble (AOI) content of 4 is essential for critical upstream conversion process. Nuclear grade uranium metal ingot is finally produced by metallothermic reduction process at 650℃ in a closed vessel, called bomb reactor. In the process, metal-slag separation plays an important role for attaining metal purity as well as process yield. Technological as well economic feasibility of indigenously developed process for large scale production of uranium metal from the crude UF 4 has been established in Bhabha Atomic Research Centre (BARC), India

  11. Erosion-Oxidation Response of Boiler Grade Steels: A Mathematical Investigation

    Directory of Open Access Journals (Sweden)

    S. K. Das

    2008-01-01

    Full Text Available A ductile erosion model embodying the mechanisms of erosion involving cutting wear and repeated plastic deformation has been developed to predict erosion rates of boiler grade steels. The issue of erosion-oxidation interaction has also been addressed to further predict the mass loss resulted from this composite mechanism. A deterministic formalism for the kinetics of oxide-scale growth and a probabilistic approach to characterize the material loss are employed to describe simultaneous actions of high-temperature oxidation and mechanical erosion. The model predictions are in good agreement with the published data.

  12. Analysis of metallogenic conditions of sandstone type uranium deposits in interlayer oxidation zone in the northwest of Junggar basin, Xinjiang

    International Nuclear Information System (INIS)

    Lin Shuangxing

    1997-01-01

    From various aspects such as the basin structures, structure features of the sedimentary cover, lithofacies features of basin sediments, geochemical characteristics of sedimentary rocks, regional hydrogeologic conditions and epigenetic reworking of the sedimentary cover and so on, the author analyzes the metallogenic conditions of sandstone type uranium deposits in interlayer oxidation zone in the northwest of Junggar basin. The author proposes that the area has abundant uranium sources, and possesses favourable stratigraphic combinations for the development of interlayer oxidation zone. Secondarily oxidized sandstone bodies present universally at places from the source area to stream channel facies. Reducing beds and secondarily reducing barrier are developed at the front of the secondarily oxidized sandstone body. The tectonic features of the sedimentary cover indicate that the area belongs to a relatively-stable suborogenic region and possesses secondary mobilization and reworking conditions during the uplifting and contraction stage of the basin. Epigenetic metallization is evident in the sedimentary cover of the basin

  13. A study on uranium metallogenetic prospects of ground water oxidation zone type in the lower cretaceous, north Shanganning basin

    International Nuclear Information System (INIS)

    Wang Jinping

    2000-01-01

    Lower Cretaceous is developed well in the north part of Shanganning basin. The area was widely uplifting vertically after their deposited. Based on the features of lithology, lithophase and Neotectonic forms, two main periods of oxidation-erosion of K2-E1 and N1-present can be distinguished. During these two periods, large scale horizontal oxidation were occurred. It is significant that the ground water oxidation related to the uranium mineralization and has been proved by the field investigation and the data of γ-logging in drill hole for oil. Meanwhile, according to the hydrodynamic features of present Shanganning plateau type artesian basin, it seems that uranium mineralization main related to the ground water oxidation the upper parts of the Lower Cretaceous

  14. Study of the Electrolytic Reduction of Uranium Oxide in LiCl-Li2O Molten Salts with an Integrated Cathode Assembly

    International Nuclear Information System (INIS)

    Park, Sung Bin; Seo, Chung Seok; Kang, Dae Seung; Kwon, Seon Gil; Park, Seong Won

    2005-01-01

    The electrolytic reduction of uranium oxide in a LiCl-Li 2 O molten salt system has been studied in a 10 g U 3 O 3 /batch-scale experimental apparatus with an integrated cathode assembly at 650 .deg. C. The integrated cathode assembly consists of an electric conductor, the uranium oxide to be reduced and the membrane for loading the uranium oxide. From the cyclic voltammograms for the LiCl-3 wt% Li 3 O system and the U 3 O 3 -LiCl-3 wt% Li 2 O system according to the materials of the membrane in the cathode assembly, the mechanisms of the predominant reduction reactions in the electrolytic reactor cell were to be understood; direct and indirect electrolytic reduction of uranium oxide. Direct and indirect electrolytic reductions have been performed with the integrated cathode assembly. Using the 325-mesh stainless steel screen the uranium oxide failed to be reduced to uranium metal by a direct and indirect electrolytic reduction because of a low current efficiency and with the porous magnesia membrane the uranium oxide was reduced successfully to uranium metal by an indirect electrolytic reduction because of a high current efficiency.

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

    Science.gov (United States)

    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

  16. Role of oxidative stress in liver and kidney in uranium toxicity after chronic exposure

    International Nuclear Information System (INIS)

    Poisson - Moreau-De-Lizoreux, C.

    2013-01-01

    Uranium is a radioactive heavy metal found in the environment. Due to its natural presence and to civil and militaries activities, general population can be exposed to U throughout drinking water or contaminated food. The pro/anti-oxidative system is a defense system which is often implicated in case of acute exposure to U. The aim of this thesis is to study the role of the pro/anti-oxidative system after chronic exposure to U in the liver and the kidney. After chronic exposure of rats to different U concentrations, this radionuclide accumulated in the organs in proportion to U intake; until 6 μg.g -1 of kidney tissues. U is localized in nucleus of the proximal tubular cells of the kidney. No nephrotoxicity was described even for the higher U level in drinking water and a reinforcement of the pro/anti-oxidative system with an increase in glutathione is observed. The study of U internal contamination in Nrf2 deficient mice, a cytoprotective transcription factor involved in the anti-oxidative defense has been realized. U accumulate more in Nrf2 mice than in WT mice but the biologic effects of U on the pro/anti-oxidative system did not seem to implicate Nrf2. At the cell level, a correlation between U distribution in HepG2 cells and the biological effects on this system is observed after U exposure at low concentrations. Soluble distribution of U is observed in cell nucleus. The apparition of U precipitates is correlated to the establishment of the adaptive mechanisms overtime which are overwhelmed and lead to a cellular toxicity at higher U level. In conclusion, these results suggest that the reinforcement of pro/anti-oxidative system could be an adaptive mechanism after chronic exposure at low U concentration. (author) [fr

  17. Assembly of tantalum porous films with graded oxidation profile from size-selected nanoparticles

    Science.gov (United States)

    Singh, Vidyadhar; Grammatikopoulos, Panagiotis; Cassidy, Cathal; Benelmekki, Maria; Bohra, Murtaza; Hawash, Zafer; Baughman, Kenneth W.; Sowwan, Mukhles

    2014-05-01

    Functionally graded materials offer a way to improve the physical and chemical properties of thin films and coatings for different applications in the nanotechnology and biomedical fields. In this work, design and assembly of nanoporous tantalum films with a graded oxidation profile perpendicular to the substrate surface are reported. These nanoporous films are composed of size-selected, amorphous tantalum nanoparticles, deposited using a gas-aggregated magnetron sputtering system, and oxidized after coalescence, as samples evolve from mono- to multi-layered structures. Molecular dynamics computer simulations shed light on atomistic mechanisms of nanoparticle coalescence, which govern the films porosity. Aberration-corrected (S) TEM, GIXRD, AFM, SEM, and XPS were employed to study the morphology, phase and oxidation profiles of the tantalum nanoparticles, and the resultant films.

  18. The effects of uranium oxide high-level waste on the structure of iron phosphate glasses

    International Nuclear Information System (INIS)

    Badyal, Y.

    1998-01-01

    Because of their unusually good chemical durability, iron phosphate glasses are a natural candidate for a nuclear waste disposal glass. We have studied the effects of UO 2 high-level waste on the structure of iron phosphate glasses with both neutron and high-energy x-ray diffraction using the GLAD instrument of the Intense Pulsed Neutron Source and the 1-BM bending magnet beamline of the Advanced Photon Source, respectively. The results of neutron scattering, which is mostly sensitive to correlations involving light atoms i.e. O-O, Fe-O and P-O, suggest the main structural features of the base glass are largely unaffected by the addition of UO 2 . The nearest-neighbor P-O, Fe-O and O-O peaks remain at the same position in real space and their intensities scale approximately with concentration. These findings are consistent with the earlier results of Raman scattering and EXAFS on the Fe-K edge wherein both cases the spectra remain similar to the base glass. High-energy x-ray scattering which is sensitive to correlations involving the heavier atoms and thus complements the neutron measurements, is also consistent with uranium occupying interstitial sites in the relatively undisturbed base glass structure. However, important questions remain as to the precise local structure and oxidation state of uranium in these glasses

  19. Retention and chemical speciation of uranium in an oxidized wetland sediment from the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    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 mechanism onto Savannah River Site (SRS) wetland sediments was studied using batch (ad)sorption experiments, sequential extraction desorption tests and U L{sub 3}-edge X-ray absorption near-edge structure (XANES) spectroscopy of contaminated wetland sediments. 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 tests indicated that the U(VI) species were primarily associated with the acid soluble fraction (weak acetic acid extractable) and NOM fraction (Na-pyrophosphate extractable). Uranium L3- edge XANES spectra of the U-retained sediments were nearly identical to that of uranyl acetate. The primary oxidation state of U in these sediments was as 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. The molecular mechanism responsible for the high U retention in the SRS wetland sediments is likely related to the chemical bonding of U to organic carbon.

  20. Preparation of porous magnesium oxide foam and study on its enrichment of uranium

    Science.gov (United States)

    Wang, Yunsheng; Chen, Yuantao; Liu, Chen; Yu, Fang

    2018-06-01

    Herein, we prepared porous magnesium oxide (MP-MgO) by in situ carbonization and using Mg(NO3)2 as precursor along with P123 as soft template. The BET surface area, and total pore volume of MP-MgO were found to be 14.76 m2 g, 0.15 cm3 g-1, respectively. The adsorption behavior of U (Ⅵ) by the porous MgO was studied by static adsorption experiments, and also the effects of adsorption time, pH of wastewater and U (Ⅵ) concentration were discussed. The results indicated that the optimum pH for MP-MgO was 3.0-4.5, the removal of uranium in this case was mainly due to surface complexation. Through the study, we found that the adsorption capacity of MP-MgO for uranium was more than 2500 mg g-1, the adsorption equilibrium time was about 150 min. Moreover, the kinetic study showed that the process followed the pseudo-first-order model, and the adsorption process was spontaneous and endothermic.

  1. Kinetics of uranium (4) oxidation with nitrous acid in the presence of oxalate-ions in nitric and perchloric acid solutions

    International Nuclear Information System (INIS)

    Nikitenko, S.I.; Astafurova, L.N.

    1991-01-01

    Kinetics of uranium (4) oxidation with nitrous acid in the presence of oxalate-ions are studied spectrometrically. It is shown that even at small oxalate concentrations a notable effect of tetravalent uranium stabilization is observed relatively to the oxidation with nitrous acid. In case of a significant excess of oxalate-ions the oxidation rate will be considerably slower as a result of the formation of U(4) bisoxalate complex

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

    Science.gov (United States)

    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.

  3. Uranium oxide fuel cycle analysis in VVER-1000 with VISTA simulation code

    Science.gov (United States)

    Mirekhtiary, Seyedeh Fatemeh; Abbasi, Akbar

    2018-02-01

    The VVER-1000 Nuclear power plant generates about 20-25 tons of spent fuel per year. In this research, the fuel transmutation of Uranium Oxide (UOX) fuel was calculated by using of nuclear fuel cycle simulation system (VISTA) code. In this simulation, we evaluated the back end components fuel cycle. The back end component calculations are Spent Fuel (SF), Actinide Inventory (AI) and Fission Product (FP) radioisotopes. The SF, AI and FP values were obtained 23.792178 ton/y, 22.811139 ton/y, 0.981039 ton/y, respectively. The obtained value of spent fuel, major actinide, and minor actinide and fission products were 23.8 ton/year, 22.795 ton/year, 0.024 ton/year and 0.981 ton/year, respectively.

  4. Environmental report of Purex Plant and Uranium Oxide Plant - Hanford reservation

    International Nuclear Information System (INIS)

    1979-04-01

    A description of the site, program, and facilities is given. The data and calculations indicate that there will be no significant adverse environmental impact from the resumption of full-scale operations of the Purex and Uranium Oxide Plants. All significant pathways of radionuclides in Purex Plant effluents are evaluated. This includes submersion in the airborne effluent plumes, consumption of drinking water and foodstuffs irrigated with Columbia River water, ingestion of radioactive iodine through the cow-to-milk pathway, consumption of fish, and other less significant pathways. A summary of research and surveillance programs designed to assess the possible changes in the terresstrial and aquatic environments on or near the Hanford Reservation is presented. The nonradiological discharges to the environment of prinicpal interest are chemicals, sewage, and solid waste. These discharges will not lead to any significant adverse effects on the environment

  5. Crystal and molecular structure of dichlorodioxobis (triphenylphosphine oxide)uranium(VI)

    Energy Technology Data Exchange (ETDEWEB)

    Bombieri, G; Forsellini, E [Consiglio Nazionale delle Ricerche, Padua (Italy). Lab. di Chimica e Tecnologia dei Radioelementi; Day, J P; Azeez, W I

    1978-01-01

    The crystal structure of (UO/sub 2/Cl/sub 2/(PPh/sub 3/O)/sub 2/) has been determined from three-dimensional X-ray diffraction data. Crystals are triclinic, space group P1, with a = 10.0101(6), b = 10.2589(9), c = 9.2347(8) A, ..cap alpha.. = 110.093(6). ..beta.. = 92.129(6), and ..gamma.. = 78.384(6), and Z = 1. The structure has been solved by the heavy-atom method from counter data, and refined by least squares to a final R of 0.054. The coordination polyhedron around uranium is a distorted octahedron, with a linear uranyl group (U-O 1.764A) perpendicular to a plane in which the two chloride and two oxide ions trans to each other occupy the corners of a rectangle (U-O 2.300; U-Cl 2.645 A).

  6. Chemical states of fission products in irradiated uranium-plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    Kurosaki, Ken; Uno, Masayoshi; Yamanaka, Shinsuke

    1999-01-01

    The chemical states of fission products (FPs) in irradiated uranium-plutonium mixed oxide (MOX) fuel for the light water reactor (LWR) were estimated by thermodynamic equilibrium calculations on system of fuel and FPs by using ChemSage program. A stoichiometric MOX containing 6.1 wt. percent PuO 2 was taken as a loading fuel. The variation of chemical states of FPs was calculated as a function of oxygen potential. Some pieces of information obtained by the calculation were compared with the results of the post-irradiation examination (PIE) of UO 2 fuel. It was confirmed that the multicomponent and multiphase thermodynamic equilibrium calculation between fuel and FPs system was an effective tool for understanding the behavior of FPs in fuel. (author)

  7. Thermodynamic functions and vapor pressures of uranium and plutonium oxides at high temperatures

    International Nuclear Information System (INIS)

    Green, D.W.; Reedy, G.T.; Leibowitz, L.

    1977-01-01

    The total energy release in a hypothetical reactor accident is sensitive to the total vapor pressure of the fuel. Thermodynamic functions which are accurate at high temperature can be calculated with the methods of statistical mechanics provided that needed spectroscopic data are available. This method of obtaining high-temperature vapor pressures should be greatly superior to the extrapolation of experimental vapor pressure measurements beyond the temperature range studied. Spectroscopic data needed for these calculations are obtained from infrared spectroscopy of matrix-isolated uranium and plutonium oxides. These data allow the assignments of the observed spectra to specific molecular species as well as the calculation of anharmonicities for monoxides, bond angles for dioxides, and molecular geometries for trioxides. These data are then employed, in combination with data on rotational and electronic molecular energy levels, to determine thermodynamic functions that are suitable for the calculation of high-temperature vapor pressures

  8. Fluorine and chlorine determination in mixed uranium-plutonium oxide fuel and plutonium dioxide

    International Nuclear Information System (INIS)

    Elinson, S.V.; Zemlyanukhina, N.A.; Pavlova, I.V.; Filatkina, V.P.; Tsvetkova, V.T.

    1981-01-01

    A technique of fluorine and chlorine determination in the mixed uranium-plutonium oxide fuel and plutonium dioxide, based on their simultaneous separation by means of pyrohydrolysis, is developed. Subsequently, fluorine is determined by photometry with alizarincomplexonate of lanthanum or according to the weakening of zirconium colouring with zylenol orange. Chlorine is determined using the photonephelometric method according to the reaction of chloride-ion interaction with silver nitrate or by spectrophotometric method according to the reaction with mercury rhodanide. The lower limit of fluorine determination is -6x10 -5 %, of chlorine- 1x10 -4 % in the sample of 1g. The relative mean quadratic deviation of the determination result (Ssub(r)), depends on the character of the material analyzed and at the content of nx10 -4 - nx10 -3 mass % is equal to from 0.05 to 0.32 for fluorine and from 0.11 to 0.35 for chlorine [ru

  9. Precipitation of uranium oxide by reduction in alkaline solution; Precipitation d'oxyde d'uranium par reduction en milieu alcalin

    Energy Technology Data Exchange (ETDEWEB)

    Pottier, P; Claus, J [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-07-01

    In the first part of the report the authors study the reaction mechanism for this reduction which makes it possible to precipitate a hydrated uranium oxide from alkaline uranyl carbonate solutions. The research into the effects of different variables on numerous cycles are then summarized. Optical, X-ray and thermogravimetric examinations then make it possible to predict the properties of this oxide. In the second part the authors carry out calculations for the continuous operation of single cells and cells in series. These calculations give the data required for the construction of 2 cells having capacities of 0.3 and 10 litres. Results obtained from the continuous operation of this latter cell lead to certain conclusions concerning the applicability of this method to the hydrometallurgy of uranium. (authors) [French] Dans une premiere partie, les auteurs etudient le mecanisme de reaction de cette reduction qui permet la precipitation d'un oxyde d'uranium hydrate dans les solutions d'uranyle-carbonates alcalins. Les etudes de diverses variables sur de nombreux cycles sont ensuite resumees. Puis des examens optiques, aux rayons X et par thermogravimetrie, permettent de proposer une hypothese sur les proprietes de l'oxyde obtenu. Dans la deuxieme partie, les auteurs developpent un calcul prevoyant la marche continue de cellules uniques et en cascades. De ces calculs on tire les elements permettant la realisation de deux cellules de 0,3 et 10 litres. Des resultats de marche continue sur cette derniere cellule, on peut conclure a l'applicabilite de cette methode a l'hydrometallurgie de l'uranium. (auteurs)

  10. U(IV) chalcogenolates synthesized via oxidation of uranium metal by dichalcogenides.

    Science.gov (United States)

    Gaunt, Andrew J; Scott, Brian L; Neu, Mary P

    2006-09-04

    Treatment of uranium metal with dichalcogenides in the presence of a catalytic amount of iodine in pyridine affords molecular U(IV) chalcogenolates that do not require stabilizing ancillary ligands. Oxidation of U(0) by PhEEPh yields monomeric seven-coordinate U(EPh)4(py)3 (E = S(1), Se(2)). The dimeric eight-coordinate complexes [U(EPh)2(mu2-EPh)2(CH3CN)2]2 (E = S(3), Se(4)) are obtained by crystallization from solutions of 1 and 2 dissolved in acetonitrile. Oxidation of U(0) by pySSpy and crystallization from thf yields nine-coordinate U(Spy)4(thf) (5). Incorporation of elemental selenium into the oxidation of U(0) by PhSeSePh results in the isolation of [U(py)2(SePh)(mu3-Se)(mu2-SePh)]4.4py (6), a tetrameric cluster in which each U(IV) ion is eight-coordinate and the U4Se4 core forms a distorted cube. The compounds were analyzed spectroscopically and the single-crystal X-ray structures of 1 and 3-6 were determined. The isolation of 1-6 represents six new examples of actinide chalcogenolates and allows insight into the nature of "hard" actinide ion-"soft" chalcogen donor interactions.

  11. Assessment of uranium dioxide fuel performance with the addition of beryllium oxide

    Energy Technology Data Exchange (ETDEWEB)

    Muniz, Rafael O.R.; Abe, Alfredo; Gomes, Daniel S.; Silva, Antonio T., E-mail: romuniz@usp.br, E-mail: ayabe@ipen.br, E-mail: danieldesouza@gmail.com, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energética s e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LabRisco/USP), Sao Paulo, SP (Brazil). Lab. de Análise, Avaliação e Gerenciamento de Risco; Aguiar, Amanda A., E-mail: amanda.abati.aguiar@gmail.com [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil)

    2017-07-01

    The Fukushima Daiichi accident in 2011 pointed the problem related to the hydrogen generation under accident scenarios due to the oxidation of zirconium-based alloys widely used as fuel rod cladding in water-cooled reactors. This problem promoted research programs aiming the development of accident tolerant fuels (ATF) which are fuels that under accident conditions could keep longer its integrity enabling the mitigation of the accident effects. In the framework of the ATF program, different materials have been studied to be applied as cladding to replace zirconium-based alloy; also efforts have been made to improve the uranium dioxide thermal conductivity doping the fuel pellet. This paper evaluates the addition of beryllium oxide (BeO) to the uranium dioxide in order to enhance the thermal conductivity of the fuel pellet. Investigations performed in this area considering the addition of 10% in volume of BeO, resulting in the UO{sub 2}-BeO fuel, have shown good results with the improvement of the fuel thermal conductivity and the consequent reduction of the fuel temperatures under irradiation. In this paper, two models obtained from open literature for the thermal conductivity of UO{sub 2}- BeO fuel were implemented in the FRAPCON 3.5 code and the results obtained using the modified code versions were compared. The simulations were carried out using a case available in the code documentation related to a typical pressurized water reactor (PWR) fuel rod irradiated under steady state condition. The results show that the fuel centerline temperatures decrease with the addition of BeO, when compared to the conventional UO{sub 2} pellet, independent of the model applied. (author)

  12. Uranium and cadmium provoke different oxidative stress responses in Lemna minor L.

    Science.gov (United States)

    Horemans, N; Van Hees, M; Van Hoeck, A; Saenen, E; De Meutter, T; Nauts, R; Blust, R; Vandenhove, H

    2015-01-01

    Common duckweed (Lemna minor L.) is ideally suited to test the impact of metals on freshwater vascular plants. Literature on cadmium (Cd) and uranium (U) oxidative responses in L. minor are sparse or, for U, non-existent. It was hypothesised that both metals impose concentration-dependent oxidative stress and growth retardation on L. minor. Using a standardised 7-day growth inhibition test, the adverse impact of these metals on L. minor growth was confirmed, with EC50 values for Cd and U of 24.1 ± 2.8 and 29.5 ± 1.9 μm, respectively, and EC10 values of 1.5 ± 0.2 and 6.5 ± 0.9 μm, respectively. The metal-induced oxidative stress response was compared through assessing the activity of different antioxidative enzymes [catalase, glutathione reductase, superoxide dismutase (SOD), ascorbate peroxidase (APOD), guaiacol peroxidase (GPOD) and syringaldizyne peroxidase (SPOD)]. Significant changes in almost all antioxidative enzymes indicated their importance in counteracting the U- and Cd-imposed oxidative burden. However, some striking differences were also observed. For activity of APODs and SODs, a biphasic but opposite response at low Cd compared to U concentrations was found. In addition, Cd (0.5-20 μm) strongly enhanced plant GPOD activity, whereas U inhibited it. Finally, in contrast to Cd, U up to 10 μm increased the level of chlorophyll a and b and carotenoids. In conclusion, although U and Cd induce similar growth arrest in L. minor, the U-induced oxidative stress responses, studied here for the first time, differ greatly from those of Cd. © 2014 German Botanical Society and The Royal Botanical Society of the Netherlands.

  13. A study on the upgrading by physical means of low grade uranium ore from Phu Wiang District

    International Nuclear Information System (INIS)

    Wihokratna, W.

    1980-01-01

    The purpose of this thesis is to upgrade the uranium ore at Phu Wiang district. Because of the fine grains and high degree of dissemination of uranium in ores, resulting practically complete envelopment of the uranium minerals by the gangue minerals, the ore must first undergo digestion in order to reveal the uranium minerals. After digestion, sodium hydroxide of 0.05 normal was added to the ore and the mixture was fed into the agitator provided with baffles and two specially designed propellers. Due to the 'Push - Pull' motion of the propellers a zone of specially high turbulence was created between them. Also in this region higher concentration of uranium is revealed and the high concentrated uranium ore was regularly stripped off for further analysis. It was found that by using mineral of grain size of 100 mesh and 0.0187% of uranium content a concentration up to 0.063% uranium content (an upgrading better than by a factor of three times) was achieved with the above method. The uranium content was analyzed with 3'' x 3'' NaI (Tl) detector and 1024 channels MCA

  14. Uranium - the world picture

    International Nuclear Information System (INIS)

    Silver, J.M.; Wright, W.J.

    1976-01-01

    The world resources of uranium and the future demand for uranium are discussed. The amount of uranium available depends on the price which users are prepared to pay for its recovery. As the price is increased, there is an incentive to recover uranium from lower grade or more difficult deposits. In view of this, attention is drawn to the development of the uranium industry in Australias

  15. The Complete Burning of Weapons Grade Plutonium and Highly Enriched Uranium with (Laser Inertial Fusion-Fission Energy) LIFE Engine

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J C; Diaz de la Rubia, T; Moses, E

    2008-12-23

    The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission

  16. Determination of the O/M ratios of polynary uranium oxides by Ce(IV)-Fe(II) back titration after dissolution in mixed sulphuric and phosphoric acids

    International Nuclear Information System (INIS)

    Fujino, T.; Sato, N.; Yamada, K.

    1996-01-01

    Uranium (IV) in polynary uranium oxides is determined after the solid has been dissolved in a warm mixed solution of sulphuric and phosphoric acids containing excess Ce(IV). The latter is titrated with a Fe(II) standard solution using ferroin as indicator. This method is especially effective for (mixed) uranium oxides which are difficult to dissolve in hot Ce(IV) sulphuric acid. The standard deviation of the determined x value in polynary oxides is estimated to be below ± 0.004 for samples of 10-30 mg. (orig.)

  17. Determination of the O/M ratios of polynary uranium oxides by Ce(IV)-Fe(II) back titration after dissolution in mixed sulphuric and phosphoric acids.

    Science.gov (United States)

    Fujino, T; Sato, N; Yamada, K

    1996-01-01

    Uranium (IV) in polynary uranium oxides is determined after the solid has been dissolved in a warm mixed solution of sulphuric and phosphoric acids containing excess Ce(IV). The latter is titrated with a Fe(II) standard solution using ferroin as indicator. This method is especially effective for (mixed) uranium oxides which are difficult to dissolve in hot Ce(IV) sulphuric acid. The standard deviation of the determined x value in polynary oxides is estimated to be below +/- 0.004 for samples of 10-30 mg.

  18. Process for electrolytically preparing uranium metal

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    Freiboth, Cameron J.; Gibbs, Frank E.

    2000-01-01

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

  20. Microstructure and properties of aluminium-aluminium oxide graded composite materials

    Science.gov (United States)

    Kamaruzaman, F. F.; Nuruzzaman, D. M.; Ismail, N. M.; Hamedon, Z.; Iqbal, A. K. M. A.; Azhari, A.

    2018-03-01

    In this research works, four-layered aluminium-aluminium oxide (Al-Al2O3) graded composite materials were fabricated using powder metallurgy (PM) method. In processing, metal-ceramic graded composite materials of 0%, 10%, 20% and 30% weight percentage of ceramic concentration were prepared under 30 ton compaction load using a cylindrical die-punch set made of steel. After that, two-step pressureless sintering was carried out at sintering temperature and time 600°C and 3 hours respectively. It was observed that the sintered cylindrical specimens of 30 mm diameter were prepared successfully. The graded composite specimens were analysed and the properties such as density, microstructure and hardness were measured. It was found that after sintering process, the diameter of the graded cylindrical structure was decreased. Using both Archimedes method and rule of mixture (ROM), he density of structure was measured. The obtained results revealed that the microvickers hardness was increased as the ceramic component increases in the graded layer. Moreover, it was observed that the interface of the graded structure is clearly distinguished within the multilayer stack and the ceramic particles are almost uniformly distributed in the Al matrix.

  1. Discrimination and Nitric Oxide Inhibitory Activity Correlation of Ajwa Dates from Different Grades and Origin

    Directory of Open Access Journals (Sweden)

    Nur Ashikin Abdul-Hamid

    2016-10-01

    Full Text Available This study was aimed at examining the variations in the metabolite constituents of the different Ajwa grades and farm origins. It is also targeted at establishing the correlations between the metabolite contents and the grades and further to the nitric oxide (NO inhibitory activity. Identification of the metabolites was generated using 1H-NMR spectroscopy metabolomics analyses utilizing multivariate methods. The NO inhibitory activity was determined using a Griess assay. Multivariate data analysis, for both supervised and unsupervised approaches, showed clusters among different grades of Ajwa dates obtained from different farms. The compounds that contribute towards the observed separation between Ajwa samples were suggested to be phenolic compounds, ascorbic acid and phenylalanine. Ajwa dates were shown to have different metabolite compositions and exhibited a wide range of NO inhibitory activity. It is also revealed that Ajwa Grade 1 from the al-Aliah farm exhibited more than 90% NO inhibitory activity compared to the other grades and origins. Phenolic compounds were among the compounds that played a role towards the greater capacity of NO inhibitory activity shown by Ajwa Grade 1 from the al-Aliah farm.

  2. Development of the heap leaching of low-grade uranium ores for conditions of OJSC Priargunsky Mining and Chemical plant (PPGKhO)

    International Nuclear Information System (INIS)

    Morozov, A.; Litvinenko, V.

    2014-01-01

    The treatment of low-grade commercial uranium ores by heap leaching has been carried out at the enterprise since 1996. During the initial stage of development, the ore piles were formed of the raw ore having the run-of-mine coarseness with uranium content around 0.08%. Under such conditions, recovery of the metal to the solution is 60-65% in case of a pile treatment lasting 2 years. To intensify the process and to provide a stable concentration of uranium in the productive solutions transferred to sorption, the enterprise developed and implemented a method of percolation leaching of low-grade ores with re-circulation of productive solutions through the re-treated ore bulk (RF patent No. 2226564). The main peculiarity of such leaching is simultaneous moistening of the ore by productive solutions and by barren solutions that are sharpened with sulphuric acid; that gives the possibility to wet far bigger areas of piles under constant volume of productive solutions outputting to the sorption treatment. Such scheme enables to treat successively first the piles at the “re-treatment” (where the metal is mainly extracted), and then the piles at the “active leaching” stage (where the metal is mainly inside the ore bulk). The technical and economic indexes of the heap leaching of low-grade uranium ores were significantly increased in 2006, when the X-ray-radiometric treatment plant was commissioned. The technological scheme of ore treatment at the processing plant includes mould and grating of the raw material with delivery of undersized products enriched with uranium: -5 mm are transferred to the pulp process; fractions (-200+40) mm to the X-ray-radiometric separation; the material of size (-40+5) mm, washed-out from clayey and fine particles, are sent to the uranium heap leaching in piles. Delivery of the ore material having size (-40+5) mm to treatment by the acid leaching method excluded colmatage and creation of zones impermeable for water, and in combination

  3. Preconcentration of a low grade uranium ore in CPDU and laboratory investigation to optimize the dewatering conditions of the preconcentration products

    International Nuclear Information System (INIS)

    Cristovici, M.A.; Berry, T.F.; Raicevic, M.M.; Brady, E.L.; Bredin, E.L.; Leigh, G.W.; Rouleau, J.P.

    1982-04-01

    A process consisting of pyrite flotation and magnetic concentration of radionuclides was developed by CANMET over several years, to preconcentrate low grade uranium ores prior to leaching. When the economics of the preconcentration-leaching technology was compared with the leaching of the entire ore after pyrite flotation (Base Case variant), the preconcentration method appeared to be economically less advantageous than expected, due to the high cost of dewatering the preconcentration products. Further investigations examined in-depth the metallurgy and dewatering of the two variants: preconcentration and base case. A typical low grade uranium ore from Elliot Lake area was used. The metallurgy was compared based on data from continuous operation (CPDU). In the preconcentration variant the amount of ore directed to leaching was reduced to more than one third of that processed in the base case, while the radionuclide concentration became more than three times higher. However, by preconcentration 7% of the uranium was lost before leaching. Systematic laboratory-scale settling and filter tests optimized the dewatering conditions of the preconcentration technology to the extent that rates similar to those of the base case were obtained

  4. Highly transparent vanadium oxide-graded indium zinc oxide electrodes for flexible organic solar cells

    International Nuclear Information System (INIS)

    Ko, Eun-Hye; Kim, Han-Ki

    2016-01-01

    We investigated characteristics of amorphous V_2O_5-graded InZnO (IZO) films to use as a flexible anode for flexible organic solar cells (FOSCs). Graded sputtering of the V_2O_5 layer on the IZO layer produced V_2O_5-graded IZO anodes (VGIZO) with a sheet resistance of 42.14 Ω/square, a resistivity of 6.32 × 10"−"4 Ω cm, and an optical transmittance of 82.15%, as well as good mechanical flexibility. In addition, the VGIZO electrode showed a greater work function of 5.2 eV than that (4.9 eV) of an IZO anode, which is beneficial for hole extraction from an organic active layer. Due to the higher work function of the VGIZO electrodes, FOSCs fabricated on the flexible VGIZO anode exhibited a higher power conversion efficiency 2.753% than that of FOSCs on the IZO anode. This indicates that the V_2O_5 graded sputtering is a promising technique to increase the work function of the IZO anode without change in sheet resistance and transmittance. - Highlights: • Transparent and flexible V_2O_5 graded IZO (VGIZO) electrodes. • High work function of VGIZO electrodes • The VGIZO film is a promising flexible anode for flexible organic solar cells.

  5. Identification of Mn(II)-oxidizing bacteria from a low-pH contaminated former uranium mine.

    Science.gov (United States)

    Akob, Denise M; Bohu, Tsing; Beyer, Andrea; Schäffner, Franziska; Händel, Matthias; Johnson, Carol A; Merten, Dirk; Büchel, Georg; Totsche, Kai Uwe; Küsel, Kirsten

    2014-08-01

    Biological Mn oxidation is responsible for producing highly reactive and abundant Mn oxide phases in the environment that can mitigate metal contamination. However, little is known about Mn oxidation in low-pH environments, where metal contamination often is a problem as the result of mining activities. We isolated two Mn(II)-oxidizing bacteria (MOB) at pH 5.5 (Duganella isolate AB_14 and Albidiferax isolate TB-2) and nine strains at pH 7 from a former uranium mining site. Isolate TB-2 may contribute to Mn oxidation in the acidic Mn-rich subsoil, as a closely related clone represented 16% of the total community. All isolates oxidized Mn over a small pH range, and isolates from low-pH samples only oxidized Mn below pH 6. Two strains with different pH optima differed in their Fe requirements for Mn oxidation, suggesting that Mn oxidation by the strain found at neutral pH was linked to Fe oxidation. Isolates tolerated Ni, Cu, and Cd and produced Mn oxides with similarities to todorokite and birnessite, with the latter being present in subsurface layers where metal enrichment was associated with Mn oxides. This demonstrates that MOB can be involved in the formation of biogenic Mn oxides in both moderately acidic and neutral pH environments. Copyright © 2014, American Society for Microbiology. All Rights Reserved.

  6. Identification of Mn(II)-oxidizing bacteria from a low-pH contaminated former uranium mine

    Science.gov (United States)

    Akob, Denise M.; Bohu, Tsing; Beyer, Andrea; Schäffner, Franziska; Händel, Matthias; Johnson, Carol A.; Merten, Dirk; Büchel, Georg; Totsche, Kai Uwe; Küsel, Kirsten

    2014-01-01

    Biological Mn oxidation is responsible for producing highly reactive and abundant Mn oxide phases in the environment that can mitigate metal contamination. However, little is known about Mn oxidation in low-pH environments, where metal contamination often is a problem as the result of mining activities. We isolated two Mn(II)-oxidizing bacteria (MOB) at pH 5.5 (Duganella isolate AB_14 and Albidiferax isolate TB-2) and nine strains at pH 7 from a former uranium mining site. Isolate TB-2 may contribute to Mn oxidation in the acidic Mn-rich subsoil, as a closely related clone represented 16% of the total community. All isolates oxidized Mn over a small pH range, and isolates from low-pH samples only oxidized Mn below pH 6. Two strains with different pH optima differed in their Fe requirements for Mn oxidation, suggesting that Mn oxidation by the strain found at neutral pH was linked to Fe oxidation. Isolates tolerated Ni, Cu, and Cd and produced Mn oxides with similarities to todorokite and birnessite, with the latter being present in subsurface layers where metal enrichment was associated with Mn oxides. This demonstrates that MOB can be involved in the formation of biogenic Mn oxides in both moderately acidic and neutral pH environments.

  7. Pro-oxidant effects in the brain of rats concurrently exposed to uranium and stress

    International Nuclear Information System (INIS)

    Linares, Victoria; Sanchez, Domenec J.; Belles, Montserrat; Albina, Luisa; Gomez, Mercedes; Domingo, Jose L.

    2007-01-01

    Metal toxicity may be associated with increased rates of reactive oxygen species (ROS) generation within the central nervous system (CNS). Although the kidney is the main target organ for uranium (U) toxicity, this metal can also accumulate in brain. In this study, we investigated the modifications on endogenous antioxidant capacity and oxidative damage in several areas of the brain of U-exposed rats. Eight groups of adult male rats received uranyl acetate dihydrate (UAD) in the drinking water at 0, 10, 20, and 40 mg/kg/day for 3 months. Animals in four groups were concurrently subjected to restraint stress during 2 h/day throughout the study. At the end of the experimental period, cortex, hippocampus and cerebellum were removed and processed to examine the following stress markers: reduced glutathione (GSH), oxidized glutathione (GSSG), glutathione reductase (GR), glutathione peroxidase (GPx), superoxide dismutase (SOD), catalase (CAT), thiobarbituric acid reactive substances (TBARS), as well as U concentrations. The results show that U significantly accumulated in hippocampus, cerebellum and cortex after 3 months of exposure. Moreover, UAD exposure promoted oxidative stress in these cerebral tissues. In cortex and cerebellum, TBARS levels were positively correlated with the U content, while in cerebellum GSSG and GSH levels were positively and negatively correlated, respectively, with U concentrations. In hippocampus, CAT and SOD activities were positively correlated with U concentration. The present results suggest that chronic oral exposure to UAD can cause progressive perturbations on physiological brain levels of oxidative stress markers. Although at the current UAD doses restraint scarcely showed additional adverse effects, its potential influence should not be underrated

  8. Modeling of solid oxide fuel cells with particle size and porosity grading in anode electrode

    Energy Technology Data Exchange (ETDEWEB)

    Liu, L.; Flesner, R.; Kim, G.Y.; Chandra, A. [Department of Mechanical Engineering, Iowa State University, Ames, Iowa (United States)

    2012-02-15

    Solid oxide fuel cells (SOFCs) have the potential to meet the critical energy needs of our modern civilization and minimize the adverse environmental impacts from excessive energy consumption. They are highly efficient, clean, and can run on variety of fuel gases. However, little investigative focus has been put on optimal power output based on electrode microstructure. In this work, a complete electrode polarization model of SOFCs has been developed and utilized to analyze the performance of functionally graded anode with different particle size and porosity profiles. The model helps to understand the implications of varying the electrode microstructure from the polarization standpoint. The work identified conditions when grading can improve the cell performance and showed that grading is not always beneficial or necessary. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  9. Uranium industry framework

    International Nuclear Information System (INIS)

    Riley, K.

    2008-01-01

    The global uranium market is undergoing a major expansion due to an increase in global demand for uranium, the highest uranium prices in the last 20 years and recognition of the potential greenhouse benefits of nuclear power. Australia holds approximately 27% of the world's uranium resources (recoverable at under US$80/kg U), so is well placed to benefit from the expansion in the global uranium market. Increasing exploration activity due to these factors is resulting in the discovery and delineation of further high grade uranium deposits and extending Australia's strategic position as a reliable and safe supplier of low cost uranium.

  10. Some characterisation profiles of uranium oxides pertaining to-ceramic fuel preparation

    International Nuclear Information System (INIS)

    El-Fekey, S.A.; Bishay, A.F.; Farah, M.Y.; Hammad, F.H.

    1978-01-01

    The effect of several variables during various stages of preparation of UO 2 powder such as precipitation, calcination and reduction on some physico-chemical properties of intermediate oxides and UO 2 powders were investigated. The methods used for characterisation were powder density, specific surface area by permeability, X-ray analysis and O/U ratio by T.G.A. The powder density of uranates was found to increase with the decrease of uranium content in uranyl nitrate solution, the increase of pH and duration of precipitation. The surface area of uranates was found to increase with the increase of pH, temperature of precipitation and the decrease of duration of precipitation. The effect of calcination for 18 hours showed that as the calcination temperature was increased for 100-800 0 C, the powder density increses and the surface area decreases. The variaton in powder density of uranates were eliminated when calcination was carried out at 700 0 C for 18 hours. The progress of the formation of intermediate oxide was followed by X-rays analysis. Two steps reduction methods, two hours at 400 0 C followed by two hours at 700 0 C, eliminated the effect of preparatory history on the 0/U ratio and produced powders having more or less the same 0/U ratio

  11. Study of uranium oxide milling in order to obtain nanostructured UC{sub x} target

    Energy Technology Data Exchange (ETDEWEB)

    Guillot, Julien, E-mail: guillotjulien@ipno.in2p3.fr [Institut de Physique Nucléaire CNRS/IN2P3 UMR 8608 – Université Paris Sud, F-91406 ORSAY Cedex (France); Tusseau-Nenez, Sandrine; Roussière, Brigitte; Barré-Boscher, Nicole [Institut de Physique Nucléaire CNRS/IN2P3 UMR 8608 – Université Paris Sud, F-91406 ORSAY Cedex (France); Brisset, François [ICMMO UMR 8182 CNRS – Université Paris Sud, F-91405 Orsay Cedex (France); Mhamed, Maher Cheikh; Lau, Christophe [Institut de Physique Nucléaire CNRS/IN2P3 UMR 8608 – Université Paris Sud, F-91406 ORSAY Cedex (France); Nowak, Sophie [ITODYS, Université Paris Diderot, Sorbonne Paris Cité, CNRS UMR-7086, F-75013 Paris (France); IPSL CNRS UMR 7583 Universités Paris Est Créteil et Paris Diderot, F-94010 Créteil Cedex (France)

    2016-05-01

    A R&D program is developed at the ALTO facility to provide new beams of exotic neutron-rich nuclei, as intense as possible. In the framework of European projects, it has been shown that the use of refractory targets with nanometric structure allows us to obtain beams of nuclei unreachable until now. The first parameter to be controlled in the processing to obtain targets with a homogeneous nanostructure is the grinding of uranium dioxide, down to 100 nm grain size. In this study, dry and wet grinding routes are studied and the powders are analyzed in terms of phase stabilization, specific surface area and grain morphology. It appears that the grinding, as well dry as wet, leads to the decrease of the particle size. The oxidation of UO{sub 2} is observed whatever the grinding. However, the dry grinding is the most efficient and leads to the oxidation of UO{sub 2} into U{sub 4}O{sub 9} and U{sub 3}O{sub 7} whose quantities increase with the grinding time while crystallite sizes decrease.

  12. Dissolution study of thorium-uranium oxides in aqueous triflic acid solutions

    Energy Technology Data Exchange (ETDEWEB)

    Bulemela, E.; Bergeron, A.; Stoddard, T. [Canadian Nuclear Laboratories - CNL, 286 Plant Rd., Chalk River, Ontario, K0J 1J0 (Canada)

    2016-07-01

    The dissolution of sintered mixed oxides of thorium with uranium in various concentrations of trifluoromethanesulfonic (triflic) acid solutions was investigated under reflux conditions to evaluate the suitability of the method. Various fragment sizes (1.00 mm < x < 7.30 mm) of sintered (Th,U)O{sub 2} and simulated high-burnup nuclear fuel (SIMFUEL) were almost completely dissolved in a few hours, which implies that triflic acid could be used as an alternative to the common dissolution method, involving nitric acid-hydrofluoric acid mixture. The influence of acid concentration, composition of the solids, and reaction time on the dissolution yield of Th and U ions was studied using Inductively Coupled Plasma - Mass Spectrometry (ICP-MS). The dissolution rate was found to depend upon the triflic acid concentration and size of the solid fragments, with near complete dissolution for the smallest fragments occurring in boiling 87% w/w triflic acid. The formation of Th and U ions in solution appears to occur at the same rate as the triflic acid simultaneously reacts with the constituent oxides as evidenced by the results of a constant U/Th concentration ratio with the progress of the dissolution. (authors)

  13. Uranium-oxide-based catalysts for the destruction of volatile chloro-organic compounds

    International Nuclear Information System (INIS)

    Hutchings, G.; Heneghan, C.S.; Taylor, S.H.

    1996-01-01

    The industrial release of hydrocarbons and chlorine-containing organic molecules into the environment continues to attract considerable public concern, which in turn has led to governmental attempts to control such emissions. The challenge is to reduce pollution without stifling economic growth. Chlorine-containing pollutants are known to be particularly stable, and at present the main industrial process for their destruction involves thermal oxidation at 1,000 o C, an expensive process that can lead to the formation of highly toxic by-products such as dioxins and dibenzofurans. Catalytic combustion at lower temperatures could potentially destroy pollutants more efficiently (in terms of energy requirements) and without forming toxic by-products. Current industrial catalysts are based on precious metals that are deactivated rapidly by organochlorine compounds. Here we report that catalysts based on uranium oxide efficiently destroy a range of hydrocarbon and chlorine-containing pollutants, and that these catalysts are resistant to deactivation. We show that benzene, toluene, chlorobutane and chlorobenzene can be destroyed at moderate temperatures ( o C) and industrially relevant flow rates. (Author)

  14. The uranium and thorium separation in the chemical reprocessing of the irradiated fuel of thorium and uranium mixed oxides

    International Nuclear Information System (INIS)

    Oliveira, E.F. de.

    1984-09-01

    A bibliographic research has been carried out for reprocessing techniques of irradiated thorium fuel from nuclear reactors. The Thorex/Hoechst process has been specially considered to establish a method for reprocessing thorium-uranium fuel from PWR. After a series of cold tests performed in laboratory it was possible to set the behavior of several parameters affecting the Thorex/Hoechst process. Some comments and suggestions are presented for modifications in the process flosheet conditions. A discussion is carried out for operational conditions such as the aqueous to organic flow ratio the acidity of strip and scrub solutions in the process steps for thorium and uranium recovery. The operation diagrams have been constructed using equilibrium experimental data which correspond to conditions observed in laboratory. (Author) [pt

  15. Application of secondary ion mass spectrometry to the study of a corrosion process: oxidation of uranium by water

    International Nuclear Information System (INIS)

    Cristy, S.S.; Condon, J.B.

    1985-01-01

    Corrosion of metals is an extremely important field with great economic and engineering implications at the Oak Ridge Y-12 Plant. To effectively combat corrosion, one must understand the processes occurring. This paper shows the utility of Secondary Ion Mass Spectrometry (SIMS) data for elucidating the processes occurring in one particular corrosion process - the oxidation of uranium by water - and for validating a theoretical model. It had long been known that the oxidation of uranium by water is retarded by the presence of oxygen gas and the retardation has been assumed to occur by site blocking at the surface. However, when alternate isotopic exposures were made, followed by exposure to a mixture of 16 O 2 and 18 OH 2 , the rapid exchange of 16 O and 18 O occurred in the oxide layer, but the further oxidation by water in this and subsequent exposures was retarded for up to 21 hours. This shows graphically that OH 2 is not held up at the surface and that the retarding mechanism is effective at the oxide/metal interface rather than at the surface. The effectiveness of the O 2 to retard the further water oxidation was much reduced if no water-formed oxide layer were present. The effectiveness was also crystallite related. 12 refs., 5 figs

  16. Application of a chronoamperometric measurement to the on-line monitoring of a lithium metal reduction for uranium oxide

    International Nuclear Information System (INIS)

    Kim, Tack-Jin; Cho, Young-Hwan; Choi, In-Kyu; Kang, Jun-Gill; Song, Kyuseok; Jee, Kwang-Yong

    2008-01-01

    Both a potentiometric and a chronoamperometric electrochemical technique have been applied in an attempt to develop an efficient method for an on-line monitoring of a lithium metal reduction process of uranium oxides at a high-temperature in a molten salt medium. As a result of this study, it was concluded that the chronoamperometric method provided a simple and effective way for a direct on-line monitoring measurement of a lithium metal reduction process of uranium oxides at 650 o C by the measuring electrical currents dependency on a variation of the reduction time for the reaction. A potentiometric method, by adopting a homemade oxide ion selective electrode made of ZrO 2 stabilized by a Y 2 O 3 doping, however, was found to be inappropriate for an on-line monitoring of the reduction reaction of uranium oxide in the presence of lithium metal due to an abnormal behavior of the adopted electrodes. The observed experimental results were discussed in detail by comparing them with previously published experimental data

  17. Synthesis of Uranium-di-Oxide nano-particles by pulsed laser ablation in ethanol and their characterisation

    International Nuclear Information System (INIS)

    Kumar, Aniruddha; Prasad, Manisha; Shail, Shailini

    2015-01-01

    The importance of actinide based nano-structures is well known in the area of biology, nuclear medicine, and nuclear industry. Pulsed laser ablation in liquid is recognised as an attractive technique for production of nano-structures of different metals and metal oxides with high purity. In this paper, we report synthesis of uranium-di-oxide nano particles by pulsed laser ablation in ethanol. The second harmonic emission of an electro- optically Q-switched nano-second Nd-YAG laser was used as the coherent source here. The structural and optical properties of the fabricated Uranium-di-oxide nano- particles were investigated by XRD, SEM, TEM, EDX and UV- Vis-NIR spectrophotometry. The mean size of the particles was found to be dependent on the laser ablation parameters. XRD and TEM analysis confirmed the phase of the synthesised material as pure crystalline Uranium-di- oxide with Face Centred Cubic structure. UV- Vis- NIR absorption spectra of the colloidal solution show high absorption in the UV regime. (author)

  18. TYPE AF CERTIFICATE FOR TRANSPORTATION OF LOW ENRICHED URANIUM OXIDE (LEUO) FOR DISPOSAL

    International Nuclear Information System (INIS)

    Opperman, E; Kenneth Yates, K

    2007-01-01

    Washington Savannah River Company (WSRC) operates the Savannah River Site (SRS) in Aiken, SC under contract with the U.S. Department of Energy (DOE). SRS had the need to ship 227 drums of low enriched uranium oxide (LEUO) to a disposal site. The LEUO had been packaged nearly 25 years ago in U.S. Department of Transportation (DOT) 17C 55-gallon drums and stored in a warehouse. Since the 235U enrichment was just above 1 percent by weight (wt%) the material did not qualify for the fissile material exceptions in 49 CFR 173.453, and therefore was categorized as 'fissile material' for shipping purposes. WSRC evaluated all existing Type AF packages and did not identify any feasible packaging. Applying for a new Type AF certificate of compliance was considered too costly for a one-time/one-way shipment for disposal. Down-blending the material with depleted uranium (to reduce enrichment below 1 wt% and enable shipment as low specific activity (LSA) radioactive material) was considered, but appropriate blending facilities do not exist at SRS. After reviewing all options, WSRC concluded that seeking a DOT Special Permit was the best option to enable shipment of the material for permanent disposal. WSRC submitted the Special Permit application to the DOT, and after one request-for-additional-information (RAI) the permit was considered acceptable. However, in an interesting development that resulted from the DOT Special Permit application process, it was determined that it was more appropriate for the DOE to issue a Type AF certificate [Ref. 1] for this shipping campaign. This paper will outline the DOT Special Permit application and Type AF considerations, and will discuss the issuance of the new DOE Type AF certificate of compliance

  19. Study on treatment of radioactive liquid waste from uranium ore processing by the use of nano oxide ferromagnetic

    International Nuclear Information System (INIS)

    Vuong Huu Anh; Nguyen Van Chinh; Nguyen Ba Tien; Doan Thi Thu Hien; Luu Cao Nguyen

    2015-01-01

    Nano oxide ferromagnetic Fe_3O_4 KT which was produced by the Military Institute of Science and Technology were used to adsorbed heavy metal elements in liquid waste. In this report, the nano oxide ferromagnetic Fe_3O_4 KT with the particle size of 80-100 nm and the specific surface area of 50-70 m"2/g was applied to study the adsorption of radioactive elements in the liquid waste of uranium ores processing. The effective parameters on adsorption process included temperature, stirring rate, stirring time, the pH value of the solution, the initial concentration of uranium in solution were investigated. The results showed that the maximum adsorption capacity for uranium of the nano Fe_3O_4 KT was 53.5 mgU/g with conditions such as: room temperature, stirring speed 120 rounds/minute, the pH value of solution was 8, stirring time about 2 hours . From the results obtained, nano Fe_3O_4 KT was tested to treatment real liquid waste of uranium ore processing after removing almost heavy metals and a part of radioactive elements by preliminary precipitation at pH 8. The results were analyzed on the ICP-MS and α, β total activity equipment, the solution concentration after treatment suitable for Vietnamese Technical Regulation on industrial wastewater QCVN 40: 2011 (concentrations of heavy metals; total activity of α and β). (author)

  20. Heat capacity measurements and XPS studies on uranium-lanthanum mixed oxides

    International Nuclear Information System (INIS)

    Venkata Krishnan, R.; Mittal, V.K.; Babu, R.; Senapati, Abhiram; Bera, Santanu; Nagarajan, K.

    2011-01-01

    Research highlights: → Heat capacity measurements were carried out on (U 1-y La y )O 2±x (y = 0.2, 0.4, 0.6, and 0.8) using differential scanning calorimeter (DSC) in the temperature range 298-800 K. → Enthalpy increment measurements were carried out on the above solid solution using high temperature drop calorimetry in the temperature range 800-1800 K. → Chemical states of U and La in the solid solutions of mixed oxides were determined using X-ray photoelectron spectroscopy (XPS). → The anomalous increase in the heat capacity is attributed to certain thermal excitation process namely Frenkel pair defect of oxygen. → From the XPS investigation, it is observed that the O/M ratio at the surface is higher than that to the bulk. → In uranium rich mixed oxide samples, the surface O/M is greater than 2 whereas that in La rich mixed oxides, it is less than 2, though the bulk O/M in all the samples are less than 2. - Abstract: Heat capacity measurements were carried out on (U 1-y La y )O 2±x (y = 0.2, 0.4, 0.6, and 0.8) using differential scanning calorimeter (DSC) in the temperature range 298-800 K. Enthalpy increment measurements were carried out on the above solid solutions using high temperature drop calorimetry in the temperature range 800-1800 K. Chemical states of U and La in the solid solutions of mixed oxides were determined using X-ray photoelectron spectroscopy (XPS). Oxygen to metal ratios of (U 1-y La y )O 2±x were estimated from the ratios of different chemical states of U present in the sample. Anomalous increase in the heat capacity is observed for (U 1-y La y )O 2±x (y = 0.4, 0.6 and 0.8) with onset temperatures in the range of 1000-1200 K. The anomalous increase in the heat capacity is attributed to certain thermal excitation process, namely, Frenkel pair defect of oxygen. The heat capacity value of (U 1-y La y )O 2±x (y = 0.2, 0.4, 0.6, and 0.8) at 298 K are 65.3, 64.1, 57.7, 51.9 J K -1 mol -1 , respectively. From the XPS investigations