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Sample records for becquerelite

  1. An investigation of the interactions of Eu(3+) and Am(3+) with uranyl minerals: implications for the storage of spent nuclear fuel.

    Science.gov (United States)

    Biswas, Saptarshi; Steudtner, Robin; Schmidt, Moritz; McKenna, Cora; León Vintró, Luis; Twamley, Brendan; Baker, Robert J

    2016-04-12

    The reaction of a number of uranyl minerals of the (oxy)hydroxide, phosphate and carbonate types with Eu(iii), as a surrogate for Am(iii), have been investigated. A photoluminescence study shows that Eu(iii) can interact with the uranyl minerals Ca[(UO2)6(O)4(OH)6]·8H2O (becquerelite) and A[UO2(CO3)3]·xH2O (A/x = K3Na/1, grimselite; CaNa2/6, andersonite; and Ca2/11, liebigite). For the minerals [(UO2)8(O)2(OH)12]·12H2O (schoepite), K2[(UO2)6(O)4(OH)6]·7H2O (compreignacite), A[(UO2)2(PO4)2]·8H2O (A = Ca, meta-autunite; Cu, meta-torbernite) and Cu[(UO2)2(SiO3OH)2]·6H2O (cuprosklodowskite) no Eu(iii) emission was observed, indicating no incorporation into, or sorption onto the structure. In the examples with Eu(3+) incorporation, sensitized emission is seen and the lifetimes, hydration numbers and quantum yields have been determined. Time Resolved Laser Induced Fluroescence Spectroscpoy (TRLFS) at 10 K have also been measured and the resolution enhancements at these temperatures allow further information to be derived on the sites of Eu(iii) incorporation. Infrared and Raman spectra are recorded, and SEM analysis show significant morphology changes and the substitution of particularly Ca(2+) by Eu(3+) ions. Therefore, Eu(3+) can substitute Ca(2+) in the interlayers of becquerelite and liebigite and in the structure of andersonite, whilst in grimselite only sodium is exchanged. These results have guided an investigation into the reactions with (241)Am on a tracer scale and results from gamma-spectrometry show that becquerelite, andersonite, grimselite, liebigite and compreignacite can include americium in the structure. Shifts in the U[double bond, length as m-dash]O and C-O Raman active bands are similar to that observed in the Eu(iii) analogues and Am(iii) photoluminescence measurements are also reported on these phases; the Am(3+) ion quenches the emission from the uranyl ion. PMID:27028717

  2. New french uranium mineral species; Nouvelles especes uraniferes francaises

    Energy Technology Data Exchange (ETDEWEB)

    Branche, G.; Chervet, J.; Guillemin, C. [Commissariat a l' Energie Atomique, Lab. du Fort de Chatillon, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1952-07-01

    In this work, the authors study the french new uranium minerals: parsonsite and renardite, hydrated phosphates of lead and uranium; kasolite: silicate hydrated of uranium and lead uranopilite: sulphate of uranium hydrated; bayleyite: carbonate of uranium and of hydrated magnesium; {beta} uranolite: silicate of uranium and of calcium hydrated. For all these minerals, the authors give the crystallographic, optic characters, and the quantitative chemical analyses. On the other hand, the following species, very rare in the french lodgings, didn't permit to do quantitative analyses. These are: the lanthinite: hydrated uranate oxide; the {alpha} uranotile: silicate of uranium and of calcium hydrated; the bassetite: uranium phosphate and of hydrated iron; the hosphuranylite: hydrated uranium phosphate; the becquerelite: hydrated uranium oxide; the curite: oxide of uranium and lead hydrated. Finally, the authors present at the end of this survey a primary mineral: the brannerite, complex of uranium titanate. (author) [French] Dans ce travail, les auteurs etudient les nouveaux mineraux uraniferes francais: parsonsite et renardite, phosphates hydrates de plomb et d'uranium; kasolite: silicate hydrate d'uranium et de plomb uranopilite: sulfate d'uranium hydrate; bayleyite: carbonate d'uranium et de magnesium hydrate; {beta} uranolite: silicate d'uranium et de calcium hydrate. Pour tous ces mineraux, les auteurs donnent les caracteres cristallographiques, optiques, et les analyses chimiques quantitatives. Par contre, les especes suivantes, tres rares dans les gites francais, n'ont pas permis d'effectuer d'analyses quantitatives. Ce sont: l'ianthinite: oxyde uraneux hydrate; l'{alpha} uranotile: silicate d'uranium et de calcium hydrate; le bassetite: phosphate d'uranium et de fer hydrate; la hosphuranylite: phosphate duranium hydrate; la becquerelite: oxyde d'uranium hydrate; la curite: oxyde d

  3. Kinetic and thermodynamic studies of uranium minerals. Assessment of the long-term evolution of spent nuclear fuel

    International Nuclear Information System (INIS)

    We have studied the dissolution behavior of uraninite, becquerelite, schoepite and uranophane. The information obtained under a variety of experimental conditions has been combined with extensive solid phase characterizations, performed in both leached and unleached samples. The overall objective is to construct a thermodynamic and kinetic model for the long-term oxidation alteration of UO2(s), as an analogy of the spent nuclear fuel matrix. We have determined the solubility product for becquerelite (logKs0 32.7±1.3) and uranophane (logKs0 = 7.8±0.8). In some experiments, the reaction progress has shown initial dissolution of uranophane followed by precipitation of a secondary solid phase, characterized as soddyite. The solubility production for this phase has been determined (logKs0 = 3.0±2.9). We have studied the kinetics of dissolution of uraninite, uranophane and schoepite under oxidizing conditions in synthetic granitic groundwater. BET measurements have been performed for uraninite and uranophane. For schoepite, the measurement has not been performed due to lack of sufficient amount of sample. The normalized rates of dissolution of uraninite and uranophane have been calculated referred to the uranium release, as 1.97x10-8 moles h-1 m-2 and 4.0x 10-9 moles h-1 m-2, respectively. For schoepite, the dissolution process has shown two different rates, with a relatively fast initial dissolution rate of 1.97x10-8 moles h-1 followed, after approximately 1000 hours, by a slower one of 1.4x10-9 moles h-1. No formation of secondary phases has been observed in those experiments, although final uranium concentrations have in all cases exceeded the solubility of uranophane, the thermodynamically more stable phase under the experimental conditions. 24 refs, 45 figs

  4. Thermodynamics of Uranyl Minerals: Enthalpies of Formation of Uranyl Oxide Hydrates

    Energy Technology Data Exchange (ETDEWEB)

    K. Kubatko; K. Helean; A. Navrotsky; P.C. Burns

    2005-05-11

    The enthalpies of formation of seven uranyl oxide hydrate phases and one uranate have been determined using high-temperature oxide melt solution calorimetry: [(UO{sub 2}){sub 4}O(OH){sub 6}](H{sub 2}O){sub 5}, metaschoepite; {beta}-UO{sub 2}(OH){sub 2}; CaUO{sub 4}; Ca(UO{sub 2}){sub 6}O{sub 4}(OH){sub 6}(H{sub 2}O){sub 8}, becquerelite; Ca(UO{sub 2}){sub 4}O{sub 3}(OH){sub 4}(H{sub 2}O){sub 2}; Na(UO{sub 2})O(OH), clarkeite; Na{sub 2}(UO{sub 2}){sub 6}O{sub 4}(OH){sub 6}(H{sub 2}O){sub 7}, the sodium analogue of compreignacite and Pb{sub 3}(UO{sub 2}){sub 8}O{sub 8}(OH){sub 6}(H{sub 2}O){sub 2}, curite. The enthalpy of formation from the binary oxides, {Delta}H{sub f-ox}, at 298 K was calculated for each compound from the respective drop solution enthalpy, {Delta}H{sub ds}. The standard enthalpies of formation from the elements, {Delta}H{sub f}{sup o}, at 298 K are -1791.0 {+-} 3.2, -1536.2 {+-} 2.8, -2002.0 {+-} 3.2, -11389.2 {+-} 13.5, -6653.1 {+-} 13.8, -1724.7 {+-} 5.1, -10936.4 {+-} 14.5 and -13163.2 {+-} 34.4 kJ mol{sup -1}, respectively. These values are useful in exploring the stability of uranyl oxide hydrates in auxiliary chemical systems, such as those expected in U-contaminated environments.

  5. Dissolution of unirradiated UO{sub 2} fuel in synthetic groundwater. Final report (1996-1998)

    Energy Technology Data Exchange (ETDEWEB)

    Ollila, K. [VTT Chemical Technology, Espoo (Finland)

    1999-05-01

    This study was a part of the EU R and D programme 1994-1998: Nuclear Fission Safety, entitled `Source term for performance assessment of spent fuel as a waste form`. The research carried out at VTT Chemical Technology was focused on the effects of granitic groundwater composition and redox conditions on UO{sub 2} solubility and dissolution mechanisms. The synthetic groundwater compositions simulated deep granitic fresh and saline groundwaters, and the effects of the near-field material, bentonite, on very saline groundwater. Additionally, the Spanish granite/bentonite water was used. The redox conditions (Eh), which are obviously the most important factors that influence on UO{sub 2} solubility under the disposal conditions of spent fuel, varied from strongly oxidising (air-saturated), anaerobic (N{sub 2}, O{sub 2} < l ppm) to reducing (N{sub 2}, low Eh). The objective of the air-saturated dissolution experiments was to yield the maximum solution concentrations of U, and information on the formation of secondary phases that control the concentrations, with different groundwater compositions. The static batch solubility experiments of long duration (up to 1-2 years) were performed using unirradiated UO{sub 2} pellets and powder. Under anaerobic and reducing conditions, the solubilities were also approached from oversaturation. The results of the oxic, air-saturated dissolution experiments with UO{sub 2} powder showed that the increase in the salinity (< 1.7 M) had a minor effect on the measured steady-state concentrations of U. The concentrations, (1.2 ...2.5) x 10{sup -5} M, were at the level of the theoretical solubility of schoepite or another uranyl oxide hydrate, e.g. becquerelite (possibly Na-polyuranate). The higher alkalinity of the fresh (Allard) composition increased the aqueous U concentration. Only some kind of oxidised U-phase (U{sub 3}O{sub 8}-UO{sub 3}) was identified with XRD when studying possible secondary phases after the contact time of one year

  6. Dissolution of unirradiated UO2 fuel in synthetic groundwater. Final report (1996-1998)

    International Nuclear Information System (INIS)

    This study was a part of the EU R and D programme 1994-1998: Nuclear Fission Safety, entitled 'Source term for performance assessment of spent fuel as a waste form'. The research carried out at VTT Chemical Technology was focused on the effects of granitic groundwater composition and redox conditions on UO2 solubility and dissolution mechanisms. The synthetic groundwater compositions simulated deep granitic fresh and saline groundwaters, and the effects of the near-field material, bentonite, on very saline groundwater. Additionally, the Spanish granite/bentonite water was used. The redox conditions (Eh), which are obviously the most important factors that influence on UO2 solubility under the disposal conditions of spent fuel, varied from strongly oxidising (air-saturated), anaerobic (N2, O2 2, low Eh). The objective of the air-saturated dissolution experiments was to yield the maximum solution concentrations of U, and information on the formation of secondary phases that control the concentrations, with different groundwater compositions. The static batch solubility experiments of long duration (up to 1-2 years) were performed using unirradiated UO2 pellets and powder. Under anaerobic and reducing conditions, the solubilities were also approached from oversaturation. The results of the oxic, air-saturated dissolution experiments with UO2 powder showed that the increase in the salinity (-5 M, were at the level of the theoretical solubility of schoepite or another uranyl oxide hydrate, e.g. becquerelite (possibly Na-polyuranate). The higher alkalinity of the fresh (Allard) composition increased the aqueous U concentration. Only some kind of oxidised U-phase (U3O8-UO3) was identified with XRD when studying possible secondary phases after the contact time of one year with all groundwater compositions. Longer contact times are needed to identify secondary phases predicted by modelling (EQ3/6). In the anoxic dissolution experiments with UO2 pellets, the solubilities of