WorldWideScience

Sample records for schoepite

  1. Weathering of natural uranyl oxide hydrates: Schoepite polytypes and dehydration effects

    International Nuclear Information System (INIS)

    Finch, R.J.; Miller, M.L.; Ewing, R.C.

    1992-01-01

    Partial dehydration of schoepite, UO 3 x2H 2 O, is reported to produce three discrete schoepite polytypes with characteristic unit cell parameters, but this has not been confirmed. The loss of structural water from the schoepite interlayer results in progressive modification to the structure; expansion parallel to schoepite cleavage planes, and extensive fracturing. Dehydration of schoepite commences at grain boundaries and progresses inward until the entire grain is converted to dehydrated schoepite, UO 3 x0.8H 2 O. The volume decrease associated with dehydration results in expanded grain boundaries. These gaps can provide pathways for the access of groundwater, and uranyl silicates and uranyl carbonates have precipitated within these gaps, replacing both schoepite and dehydrated schoepite. Schoepite, however, is not observed to re-precipitate where in contact with dehydrated schoepite. Thus, while the formation of schoepite early during the corrosion of uraninite may be favored, schoepite is not a long-term solubility limiting phase for oxidized uranium in natural ground waters containing dissolved silica or carbonate. (orig.)

  2. Periodic Density Functional Theory Study of the Structure, Raman Spectrum, and Mechanical Properties of Schoepite Mineral.

    Science.gov (United States)

    Colmenero, Francisco; Cobos, Joaquín; Timón, Vicente

    2018-04-16

    The structure and Raman spectrum of schoepite mineral, [(UO 2 ) 8 O 2 (OH) 12 ]·12H 2 O, was studied by means of theoretical calculations. The computations were carried out by using density functional theory with plane waves and pseudopotentials. A norm-conserving pseudopotential specific for the U atom developed in a previous work was employed. Because it was not possible to locate H atoms directly from X-ray diffraction (XRD) data by structure refinement in previous experimental studies, all of the positions of the H atoms in the full unit cell were determined theoretically. The structural results, including the lattice parameters, bond lengths, bond angles, and powder XRD pattern, were found to be in good agreement with their experimental counterparts. However, the calculations performed using the unit cell designed by Ostanin and Zeller in 2007, involving half of the atoms of the full unit cell, led to significant errors in the computed powder XRD pattern. Furthermore, Ostanin and Zeller's unit cell contains hydronium ions, H 3 O + , which are incompatible with the experimental information. Therefore, while the use of this schoepite model may be a very useful approximation requiring a much smaller amount of computational effort, the full unit cell should be used to study this mineral accurately. The Raman spectrum was also computed by means of density functional perturbation theory and compared with the experimental spectrum. The results were also in agreement with the experimental data. A normal-mode analysis of the theoretical spectra was performed to assign the main bands of the Raman spectrum. This assignment significantly improved the current empirical assignment of the bands of the Raman spectrum of schoepite mineral. In addition, the equation of state and elastic properties of this mineral were determined. The crystal structure of schoepite was found to be stable mechanically and dynamically. Schoepite can be described as a brittle material exhibiting

  3. Leesite, K(H.sub.2./sub.O).sub.2./sub.[(UO.sub.2./sub.).sub.4./sub.O.sub.2./sub.(OH).sub.5./sub.]∙3H.sub.2./sub.O, a new K-bearing schoepite-family mineral from the Jomac mine, San Juan County, Utah, USA

    Czech Academy of Sciences Publication Activity Database

    Olds, T.A.; Plášil, Jakub; Kampf, A.R.; Spano, T.; Haynes, P.; Carlson, S.M.; Burns, P.C.; Simonetti, A.; Mills, O.P.

    2018-01-01

    Roč. 103, č. 1 (2018), s. 143-150 ISSN 0003-004X R&D Projects: GA ČR(CZ) GA15-12653S Institutional support: RVO:68378271 Keywords : Leesite * sheet anion topology * schoepite * uranium * uraninite * crystal structure Subject RIV: DB - Geology ; Mineralogy OBOR OECD: Geology Impact factor: 2.021, year: 2016

  4. Dissolution kinetics of UO2: Flow-through tests on UO2.00 pellets and polycrystalline schoepite samples in oxygenated, carbonate/bicarbonate buffer solutions at 25 degree C

    International Nuclear Information System (INIS)

    Nguyen, S.N.; Weed, H.C.; Leider, H.R.; Stout, R.B.

    1991-10-01

    The modelling of radionuclide release from waste forms is an important part of the performance assessment of a potential, high-level radioactive waste repository. Since spent fuel consists of UO 2 containing actinide elements and other fission products, it is necessary to determine the principal parameters affecting UO 2 dissolution and quantify their effects on the dissolution rate before any prediction of long term release rates of radionuclides from the spent fuel can be made. As part of a complex matrix to determine the dissolution kinetics of UO 2 as a function of time, pH, carbonate/bicarbonate concentration and oxygen activity, we have measured the dissolution rates at 25 degrees C of: (1) UO 2 pellets; (2) UO 2.00 powder and (3) synthetic dehydrated schoepite, UO 3 .H 2 O using a single-pass flow through system in an argon-atmosphere glove box. Carbonate, carbonate/bicarbonate, and bicarbonate buffers with concentrations ranging from 0.0002 M to 0.02 M and pH values form 8 to 11 have been used. Argon gas mixtures containing oxygen (from 0.002 to 0.2 atm) and carbon dioxide (from 0 to 0.011 atm) were bubbled through the buffers to stabilize their pH values. 12 refs., 2 tabs

  5. Alternate source term models for Yucca Mountain performance assessment based on natural analog data and secondary mineral solubility

    International Nuclear Information System (INIS)

    Murphy, W.M.; Codell, R.B.

    1999-01-01

    Performance assessment calculations for the proposed high level radioactive waste repository at Yucca Mountain, Nevada, were conducted using the Nuclear Regulatory Commission Total-System Performance Assessment (TPA 3.2) code to test conceptual models and parameter values for the source term based on data from the Pena Blanca, Mexico, natural analog site and based on a model for coprecipitation and solubility of secondary schoepite. In previous studies the value for the maximum constant oxidative alteration rate of uraninite at the Nopal I uranium body at Pena Blanca was estimated. Scaling this rate to the mass of uranium for the proposed Yucca Mountain repository yields an oxidative alteration rate of 22 kg/y, which was assumed to be an upper limit on the release rate from the proposed repository. A second model was developed assuming releases of radionuclides are based on the solubility of secondary schoepite as a function of temperature and solution chemistry. Releases of uranium are given by the product of uranium concentrations at equilibrium with schoepite and the flow of water through the waste packages. For both models, radionuclides other than uranium and those in the cladding and gap fraction were modeled to be released at a rate proportional to the uranium release rate, with additional elemental solubility limits applied. Performance assessment results using the Pena Blanca oxidation rate and schoepite solubility models for Yucca Mountain were compared to the TPA 3.2 base case model, in which release was based on laboratory studies of spent fuel dissolution, cladding and gap release, and solubility limits. Doses calculated using the release rate based on natural analog data and the schoepite solubility models were smaller than doses generated using the base case model. These results provide a degree of confidence in safety predictions using the base case model and an indication of how conservatism in the base case model may be reduced in future analyses

  6. Alternate source term models for Yucca Mountain performance assessment based on natural analog data and secondary mineral solubility

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, W.M.; Codell, R.B.

    1999-07-01

    Performance assessment calculations for the proposed high level radioactive waste repository at Yucca Mountain, Nevada, were conducted using the Nuclear Regulatory Commission Total-System Performance Assessment (TPA 3.2) code to test conceptual models and parameter values for the source term based on data from the Pena Blanca, Mexico, natural analog site and based on a model for coprecipitation and solubility of secondary schoepite. In previous studies the value for the maximum constant oxidative alteration rate of uraninite at the Nopal I uranium body at Pena Blanca was estimated. Scaling this rate to the mass of uranium for the proposed Yucca Mountain repository yields an oxidative alteration rate of 22 kg/y, which was assumed to be an upper limit on the release rate from the proposed repository. A second model was developed assuming releases of radionuclides are based on the solubility of secondary schoepite as a function of temperature and solution chemistry. Releases of uranium are given by the product of uranium concentrations at equilibrium with schoepite and the flow of water through the waste packages. For both models, radionuclides other than uranium and those in the cladding and gap fraction were modeled to be released at a rate proportional to the uranium release rate, with additional elemental solubility limits applied. Performance assessment results using the Pena Blanca oxidation rate and schoepite solubility models for Yucca Mountain were compared to the TPA 3.2 base case model, in which release was based on laboratory studies of spent fuel dissolution, cladding and gap release, and solubility limits. Doses calculated using the release rate based on natural analog data and the schoepite solubility models were smaller than doses generated using the base case model. These results provide a degree of confidence in safety predictions using the base case model and an indication of how conservatism in the base case model may be reduced in future analyses.

  7. Depleted Uranium Toxicity, Accumulation, and Uptake in Cynodon dactylon (Bermuda) and Aristida purpurea (Purple Threeawn).

    Science.gov (United States)

    Butler, Afrachanna D; Wynter, Michelle; Medina, Victor F; Bednar, Anthony J

    2016-06-01

    Yuma Proving Grounds (YPG) in western Arizona is a testing range where Depleted uranium (DU) penetrators have been historically fired. A portion of the fired DU penetrators are being managed under controlled conditions by leaving them in place. The widespread use of DU in armor-penetrating weapons has raised environmental and human health concerns. The present study is focused on the onsite management approach and on the potential interactions with plants local to YPG. A 30 day study was conducted to assess the toxicity of DU corrosion products (e.g., schoepite and meta-schoepite) in two grass species that are native to YPG, Bermuda (Cynodon dactylon) and Purple Threeawn (Aristida purpurea). In addition, the ability for plants to uptake DU was studied. The results of this study show a much lower threshold for biomass toxicity and higher plant concentrations, particularly in the roots than shoots, compared to previous studies.

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

    Energy Technology Data Exchange (ETDEWEB)

    Casas, I; Bruno, J; Cera, E [MBT Tecnologia Ambiental, Cerdanyola (Spain); Finch, R J; Ewing, R C [Department of Earth and Planetary Sciences, University of New Mexico, Albuquerque, NM (United States)

    1994-10-01

    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 UO{sub 2}(s), as an analogy of the spent nuclear fuel matrix. We have determined the solubility product for becquerelite (logK{sub s0} 32.7{+-}1.3) and uranophane (logK{sub s0} = 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 (logK{sub s0} = 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{sup -8} moles h{sup -1} m{sup -2} and 4.0x 10{sup -9} moles h{sup -1} m{sup -2}, respectively. For schoepite, the dissolution process has shown two different rates, with a relatively fast initial dissolution rate of 1.97x10{sup -8} moles h{sup -1} followed, after approximately 1000 hours, by a slower one of 1.4x10{sup -9} moles h{sup -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.

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

    International Nuclear Information System (INIS)

    Casas, I.; Bruno, J.; Cera, E.; Finch, R.J.; Ewing, R.C.

    1994-10-01

    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 UO 2 (s), as an analogy of the spent nuclear fuel matrix. We have determined the solubility product for becquerelite (logK s0 32.7±1.3) and uranophane (logK s0 = 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 (logK s0 = 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

  10. Leaching action of EJ-13 water on unirradiated UO2 surfaces under unsaturated conditions at 90 degree C: Interim report

    International Nuclear Information System (INIS)

    Wronkiewicz, D.J.; Bates, J.K.; Gerding, T.J.; Veleckis, E.; Tani, B.S.

    1991-07-01

    A set of experiments, based on the application of the Unsaturated Test method to the reaction of UO 2 with EJ-13 water, has been conducted over a period of 182.5 weeks. One half of the experiments have been terminated, while one half are still ongoing. Solutions that have dripped from UO 2 specimens have been analyzed for all experiments, while the reacted UO 2 surfaces have been examined for only the terminated experiments. A pulse of uranium release from the UO 2 solid, in conjunction with the formation of dehydrated schoepite on the surface of the UO 2 , was observed during the 39- to 96-week period. Thereafter, the uranium release decreased and a second set of secondary phases was observed. The latter phases incorporate cations from the EJ-13 water and include boltwoodite, uranophane, sklodowskite, compreignacite, and schoepite. The experiments are being continued to monitor for additional changes in solution composition and secondary phase formation, and have now reached the 319-week period. 9 refs., 17 figs., 25 tabs

  11. Identification of secondary phases formed during unsaturated reaction of UO2 with EJ-13 water

    International Nuclear Information System (INIS)

    Bates, J.K.; Tani, B.S.; Veleckis, E.

    1989-01-01

    A set of experiments, wherein UO 2 has been contacted by dripping water, has been conducted over a period of 182.5 weeks. The experiments are being conducted to develop procedures to study spent fuel reaction under unsaturated conditions that are expected to exist over the lifetime of the proposed Yucca Mountain repository site. One half of the experiments have been terminated, while one half are ongoing. Analyses of solutions that have dripped from the reacted UO 2 have been performed for all experiments, while the reacted UO 2 surfaces have been examined for the terminated experiments. A pulse of uranium release from the UO 2 solid, combined with the formation of schoepite on the surface of the UO 2 , was observed between 39 and 96 weeks of reaction. Thereafter, the uranium release decreased and a second set of secondary phases was observed. The latter phases incorporated cations from the EJ-13 water and included boltwoodite, uranophane, sklodowskite, compreignacite, and schoepite. The experiments are continuing to monitor whether additional changes in solution chemistry or secondary phase formation occurs. 6 refs., 2 figs., 2 tabs

  12. Effect of temperature on studtite stability: Thermogravimetry and differential scanning calorimetry investigations

    International Nuclear Information System (INIS)

    Rey, A.; Casas, I.; Gimenez, J.; Quinones, J.; Pablo, J. de

    2009-01-01

    The main objective of this work is the study of the influence of temperature on the stability of the uranyl peroxide tetrahydrate (UO 2 O 2 . 4H 2 O) studtite, which may form on the spent nuclear fuel surface as a secondary solid phase. Preliminary results on the synthesis of studtite in the laboratory at different temperatures have shown that the solid phases formed when mixing hydrogen peroxide and uranyl nitrate depends on temperature. Studtite is obtained at 298 K, meta-studtite (UO 2 O 2 . 2H 2 O) at 373 K, and meta-schoepite (UO 3 . nH 2 O, with n 3 O 8 . By means of the differential scanning calorimetry the molar enthalpies of the transformations occurring at 403 and 504 K have been determined to be -42 ± 10 and -46 ± 2 kJ mol -1 , respectively

  13. Unsaturated zone waters from the Nopal I natural analog, Chihuahua, Mexico -- Implications for radionuclide mobility at Yucca Mountain

    Energy Technology Data Exchange (ETDEWEB)

    Pickett, D.A.; Murphy, W.M.

    1999-07-01

    Chemical and U-Th isotopic data on unsaturated zone waters from the Nopal I natural analog reveal effects of water-rock interaction and help constrain models of radionuclide release and transport at the site and, by analogy, at the proposed nuclear waste repository at Yucca Mountain. Geochemical reaction-path modeling indicates that, under oxidizing conditions, dissolution of uraninite (spent fuel analog) by these waters will lead to eventual schoepite precipitation regardless of initial silica concentration provided that groundwater is not continuously replenished. Thus, less soluble uranyl silicates may not dominate the initial alteration assemblage and keep dissolved U concentrations low. Uranium-series activity ratios are consistent with models of U transport at the site and display varying degrees of leaching versus recoil mobilization. Thorium concentrations may reflect the importance of colloidal transport of low-solubility radionuclides in the unsaturated zone.

  14. Unsaturated zone waters from the Nopal I natural analog, Chihuahua, Mexico -- Implications for radionuclide mobility at Yucca Mountain

    International Nuclear Information System (INIS)

    Pickett, D.A.; Murphy, W.M.

    1999-01-01

    Chemical and U-Th isotopic data on unsaturated zone waters from the Nopal I natural analog reveal effects of water-rock interaction and help constrain models of radionuclide release and transport at the site and, by analogy, at the proposed nuclear waste repository at Yucca Mountain. Geochemical reaction-path modeling indicates that, under oxidizing conditions, dissolution of uraninite (spent fuel analog) by these waters will lead to eventual schoepite precipitation regardless of initial silica concentration provided that groundwater is not continuously replenished. Thus, less soluble uranyl silicates may not dominate the initial alteration assemblage and keep dissolved U concentrations low. Uranium-series activity ratios are consistent with models of U transport at the site and display varying degrees of leaching versus recoil mobilization. Thorium concentrations may reflect the importance of colloidal transport of low-solubility radionuclides in the unsaturated zone

  15. A spectroscopic study of uranium(VI) interaction with magnetite

    International Nuclear Information System (INIS)

    El Aamrani, S.; Gimenez, J.; Rovira, M.; Seco, F.; Grive, M.; Bruno, J.; Duro, L.; Pablo, J. de

    2007-01-01

    The uranium sorbed onto commercial magnetite has been characterized by using two different spectroscopic techniques such as X-ray photoelectron spectroscopy (XPS), and extended X-ray absorption fine structure (EXAFS). Magnetite samples have been put in contact with uranium(VI) solutions in conditions in which a high uranium uptake is expected. After several days, the magnetite surface has been analysed by XPS and EXAFS. The XPS results obtained are not conclusive regarding the uranium oxidation state in the magnetite surface. On the other hand, the results obtained with the EXAFS technique show that the uranium-magnetite sample spectrum has characteristics from both the UO 2 and schoepite spectra, e.g. a relatively high coordination number of equatorial oxygens and two axial oxygens, respectively. These results would indicate that the uranium sorbed onto magnetite would be a mixture of uranium(IV) and uranium(VI)

  16. Leaching patterns and secondary phase formation during unsaturated leaching of UO2 at 90 degrees C

    International Nuclear Information System (INIS)

    Wronkiewicz, D.J.; Bates, J.K.; Gerding, T.J.; Veleckis, E.; Tani, B.S.

    1991-11-01

    Experiments are being conducted that examine the reaction of UO 2 with dripping oxygenated ground water at 90 degrees C. The experiments are designed to identify secondary phases formed during UO 2 alteration, evaluate parameters controlling U release, and act as scoping tests for studies with spent fuel. This study is the first of its kind that examines the alteration of UO 2 under unsaturated conditions expected to exist at the proposed Yucca Mountain repository site. Results suggest the UO 2 matrix will readily react within a few months after being exposed to simulated Yucca Mountain conditions. A pulse of rapid U release, combined with the formation of dehydrated schoepite on the UO 2 surface, characterizes the reaction between one to two years. Rapid dissolution of intergrain boundaries and spallation of UO 2 granules appears to be responsible for much of the U released. Differential release of the UO 2 granules may be responsible for much of the variation observed between duplicate experiments. Less than 5 wt % of the released U remains in solution or in a suspended form, while the remaining settles out of solution as fine particles or is reprecipitated as secondary phases. Subsequent to the pulse period, U release rates decline and a more stable assemblage of uranyl silicate phases are formed by incorporating cations from the ground water leachant. Uranophane, boltwoodite, and sklodowskite appear as the final solubility limiting phases that form in these tests. This observed paragenetic sequence (from uraninite to schoepite-type phases to uranyl silicates) is identical to those observed in weathered zones of natural uraninite occurrences. The combined results indicate that the release of radionuclides from spent fuel may not be limited by U solubility constraints, but that spallation of particulate matter may be an important, if not the dominant release mechanism affecting release

  17. Identification of Uranium Minerals in Natural U-Bearing Rocks Using Infrared Reflectance Spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Beiswenger, Toya N. [Pacific Northwest National Laboratory, Richland, WA, USA; Gallagher, Neal B. [Eigenvector Research, Inc., Manson, WA, USA; Myers, Tanya L. [Pacific Northwest National Laboratory, Richland, WA, USA; Szecsody, James E. [Pacific Northwest National Laboratory, Richland, WA, USA; Tonkyn, Russell G. [Pacific Northwest National Laboratory, Richland, WA, USA; Su, Yin-Fong [Pacific Northwest National Laboratory, Richland, WA, USA; Sweet, Lucas E. [Pacific Northwest National Laboratory, Richland, WA, USA; Lewallen, Tricia A. [Pacific Northwest National Laboratory, Richland, WA, USA; Johnson, Timothy J. [Pacific Northwest National Laboratory, Richland, WA, USA

    2017-10-24

    The identification of minerals, including uranium-bearing minerals, is traditionally a labor-intensive-process using x-ray diffraction (XRD), fluorescence, or other solid-phase and wet chemical techniques. While handheld XRD and fluorescence instruments can aid in field identification, handheld infrared reflectance spectrometers can also be used in industrial or field environments, with rapid, non-destructive identification possible via spectral analysis of the solid’s reflectance spectrum. We have recently developed standard laboratory measurement methods for the infrared (IR) reflectance of solids and have investigated using these techniques for the identification of uranium-bearing minerals, using XRD methods for ground-truth. Due to the rich colors of such species, including distinctive spectroscopic signatures in the infrared, identification is facile and specific, both for samples that are pure or are partially composed of uranium (e.g. boltwoodite, schoepite, tyuyamunite, carnotite, etc.) or non-uranium minerals. The method can be used to detect not only pure and partial minerals, but is quite sensitive to chemical change such as hydration (e.g. schoepite). We have further applied statistical methods, in particular classical least squares (CLS) and multivariate curve resolution (MCR) for discrimination of such uranium minerals and two uranium pure chemicals (U3O8 and UO2) against common background materials (e.g. silica sand, asphalt, calcite, K-feldspar) with good success. Each mineral contains unique infrared spectral features; some of the IR features are similar or common to entire classes of minerals, typically arising from similar chemical moieties or functional groups in the minerals: phosphates, sulfates, carbonates, etc. These characteristic 2 infrared bands generate the unique (or class-specific) bands that distinguish the mineral from the interferents or backgrounds. We have observed several cases where the chemical moieties that provide the

  18. Solid state speciation and potential bioavailability of depleted uranium particles from Kosovo and Kuwait

    Energy Technology Data Exchange (ETDEWEB)

    Lind, O.C. [Isotope Laboratory, Department of Plant and Environmental Sciences, Norwegian University of Life Sciences, P.O. Box 5003, N-1432 As (Norway)], E-mail: ole-christian.lind@umb.no; Salbu, B.; Skipperud, L. [Isotope Laboratory, Department of Plant and Environmental Sciences, Norwegian University of Life Sciences, P.O. Box 5003, N-1432 As (Norway); Janssens, K.; Jaroszewicz, J.; De Nolf, W. [Department of Chemistry, University of Antwerp, Universiteitsplein 1, Antwerp (Belgium)

    2009-04-15

    A combination of synchrotron radiation based X-ray microscopic techniques ({mu}-XRF, {mu}-XANES, {mu}-XRD) applied on single depleted uranium (DU) particles and semi-bulk leaching experiments has been employed to link the potential bioavailability of DU particles to site-specific particle characteristics. The oxidation states and crystallographic forms of U in DU particles have been determined for individual particles isolated from selected samples collected at different sites in Kosovo and Kuwait that were contaminated by DU ammunition during the 1999 Balkan conflict and the 1991 Gulf war. Furthermore, small soil or sand samples heavily contaminated with DU particles were subjected to simulated gastrointestinal fluid (0.16 M HCl) extractions. Characteristics of DU particles in Kosovo soils collected in 2000 and in Kuwait soils collected in 2002 varied significantly depending on the release scenario and to some extent on weathering conditions. Oxidized U (+6) was determined in large, fragile and bright yellow DU particles released during fire at a DU ammunition storage facility and crystalline phases such as schoepite (UO{sub 3}.2.25H{sub 2}O), dehydrated schoepite (UO{sub 3}.0.75H{sub 2}O) and metaschoepite (UO{sub 3}.2.0H{sub 2}O) were identified. As expected, these DU particles were rapidly dissolved in 0.16 M HCl (84 {+-} 3% extracted after 2 h) indicating a high degree of potential mobility and bioavailability. In contrast, the 2 h extraction of samples contaminated with DU particles originating either from corrosion of unspent DU penetrators or from impacted DU ammunition appeared to be much slower (20-30%) as uranium was less oxidized (+4 to +6). Crystalline phases such as UO{sub 2}, UC and metallic U or U-Ti alloy were determined in impacted DU particles from Kosovo and Kuwait, while the UO{sub 2,34} phase, only determined in particles from Kosovo, could reflect a more corrosive environment. Although the results are based on a limited number of DU particles

  19. Characterization of Nuclear Materials in extreme conditions: the Raman spectroscopy approach

    International Nuclear Information System (INIS)

    Guimbretiere, G.; Canizares, A.; Simon, P.; Raimboux, N.; Ammar, M.R.; Duval, F.; Omnee, R.; Maslova, O.; Barthe, M.F.; Desgranges, L.; Caraballo, R.; Jegou, C.

    2013-06-01

    In this article, we review our last advances on the Raman spectroscopy characterization of irradiated and/or leached UO 2 , PuO 2 and (U, Pu)O 2 samples. For this, three original Raman setups dedicated to the study of nuclear materials were involved. In all cases, irradiation was observed inducing the appearance of Raman defect bands in the 500-750 cm -1 range. For UO 2 , annealing temperature experiment suggests that these defect bands may be the signatures of a medium range structured defect. Besides, the leaching experiments show different behavior between the in situ probed UO 2 ceramic and post mortem probed [UO 2 + (U, Pu)O 2 ] sample: In the first case we observed the growth of an altered layer made of studtite and schoepite phases. In the second we observed, in one hand the creation of U 3 O 8 and studtite phases at the surface of the UO 2 grains, in second hand no Raman signatures of any layer at the surface of the (U, Pu)O 2 grains. (authors)

  20. Colloidal products and actinide species in leachate from spent nuclear fuel

    International Nuclear Information System (INIS)

    Finn, P.A.; Buck, E.C.; Gong, M.; Hoh, J.C.; Emery, J.W.; Hafenrichter, L.D.; Bates, J.K.

    1993-01-01

    Two well-characterized types of spent nuclear fuel (ATM-103 and ATM-106) were subjected to unsaturated leach tests with simulated groundwater at 90 degrees C. The actinides present in the leachate were determined at the end of two successive periods of ∼60 days and after an acid strip done at the end of the second period. Both colloidal and soluble actinide species were detected in the leachates which had pHs ranging from 4 to 7. The uranium phases identified in the colloids were schoepite and soddyite. In addition, the actinide release behavior of the two fuels appeared to be different for both the total amount of material released and the relative amount of each isotope released. This paper will focus on the detection and identification of the colloidal species observed in the leachate that was collected after each of the first two successive testing periods of approximately 60 days each. In addition, preliminary values for the total actinide release for these two periods are reported

  1. 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

  2. Analysis of a Uranium Oxide Sample Interdicted in Slovakia (FSC 12-3-1)

    Energy Technology Data Exchange (ETDEWEB)

    Borg, Lars E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dai, Zurong [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Eppich, Gary R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Gaffney, Amy M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Genetti, Victoria G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Grant, Patrick M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Gray, Leonard W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Holiday, Kiel S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hutcheon, Ian D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kayzar, Theresa M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Klunder, Gregory L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Knight, Kimberly B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kristo, Michael J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Lindvall, Rachel E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marks, Naomi E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ramon, Christina E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ramon, Erick C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Robel, Martin [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Roberts, Sarah K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Schorzman, Kerri C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sharp, Michael A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Singleton, Michael J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Williams, Ross W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2014-01-17

    We provide a concise summary of analyses of a natural uranium sample seized in Slovakia in November 2007. Results are presented for compound identification, water content, U assay, trace element abundances, trace organic compounds, isotope compositions for U, Pb, Sr and O, and age determination using the 234U – 230Th and 235U – 231Pa chronometers. The sample is a mixture of two common uranium compounds - schoepite and uraninite. The uranium isotope composition is indistinguishable from natural; 236U was not detected. The O, Sr and Pb isotope compositions and trace element abundances are unremarkable. The 234U – 230Th chronometer gives an age of 15.5 years relative to the date of analysis, indicating the sample was produced in January 1997. A comparison of the data for this sample with data in the Uranium Sourcing database failed to find a match, indicating the sample was not produced at a facility represented in the database.

  3. Contaminant Release from Residual Waste in Single Shell Tanks at the Hanford Site, Washington, USA - 9276

    International Nuclear Information System (INIS)

    Cantrell, Kirk J.; Krupka, Kenneth M.; Deutsch, William J.; Lindberg, Michael J.

    2009-01-01

    concentrations of U were likely controlled by the solubility of schoepite (UO 3 2 H 2 O). Therefore, a reactive transport model based upon solubility of schoepite and the expected composition and infiltration rates of pore water could be used to simulate future release of U from this residual tank waste. In addition to the development of release models, the residual tank waste studies completed so far have provided a number of new insights that have changed our understanding of residual tank waste. For example, the release of contaminants from different tanks, although governed by the same general chemical principles, can be very different. It has also been found that significant fractions of Tc-99 and other typically highly mobile contaminants are frequently not readily released from tank residuals and occur in recalcitrant phases that are resistant to aqueous dissolution. As these studies progress, such key cross-cutting geochemical processes and solid phase characteristics important to contaminant release from residual tank waste are becoming apparent. This may allow the grouping of tanks into general categories with certain common chemical features and contaminant release characteristics - an important goal because complete characterization of residual wastes from all 149 single-shell storage tanks is not practical

  4. Behaviour of high purity UO2/H2O interfaces under helium beam irradiation in deaerated conditions

    International Nuclear Information System (INIS)

    Mendes, E.

    2005-11-01

    A question put within the framework of the nuclear fuel storage worn in geological site is what become to them in the presence of water. The aim of a fundamental program, of PRECCI project (ECA), is to highlight the behaviour of interfaces which can be used as models for the interfaces nuclear spent fuel/water if the fuel is uranium UO 2 dioxide. This doctorate is interested in the effect of the alpha activity which is the only one that exist in the spent fuel after long periods. The aim is to identify the mechanisms of alteration and of leaching of surfaces under alpha irradiation. A method is developed to irradiate UO 2 /H 2 O interfaces in deaerated conditions with the beam of He 2+ produced by a cyclotron. The He 2+ ions cross an UO 2 disc and emerge in water with an energy of 5 MeV. Leachings under irradiation are carried with a large range of particles flux. The post-irradiation characterization of the surface of the discs realised by micro-Raman spectroscopy allowed to identify the alteration layer. It is made up of studtite UO 2 (O 2 ),4H 2 O, and of schoepite UO 3 ,xH 2 O. The analysis of the solutions shows that the uranium release strongly increases. The electrochemical properties of the interfaces under irradiation strongly differ from those before irradiation. This work allows to propose that the radiolytic species seen by the interface are it during the heterogeneous phase of evolution of the traces and are species of short lives. Modeling show that the radiolytic radicals species can migrate toward the interface and react with the UO 2 surface. (author)

  5. Behaviour of the UO2/clayey water. A spectroscopic approach

    International Nuclear Information System (INIS)

    Guilbert, S.

    2000-05-01

    This work deals with the disposal of spent nuclear fuels in deep geological layers. After three years of irradiation, these fuels are constituted of 95 % UO 2 . It is then indispensable to know the leaching behaviour of this solid because ground waters are the main agents of dispersion to biosphere of the radioelements contained in these fuels. This work includes alteration tests carried out with a device allowing to synthesize a clayey water equilibrated with a partial pressure in CO 2 in oxidizing or reducing conditions. After the tests, the solid and the solution have been characterized in order to establish a balance of the alteration. The UO 2 matrix has been characterized by XPS. The uranium in solution has been titrated by ICP-MS. In oxidizing conditions, after some weeks, the dissolution velocity of UO 2 has stabilized around 3*10 11 mol/m 2 .s. This velocity is of 4*10 12 mol/m 2 .s in a reducing medium. The uranium concentrations in the oxidized water are of about 2*10 4 mol/l after two years of leaching. After 33 days of alteration in a reducing medium, the uranium amount is of 3*10 6 mol/l. The XPS technique has revealed a superficial and progressive oxidation of the uranium(IV) and the formation of U-OH bonds in the oxidizing medium. A U(VI)/U(IV) ratio has been determined by this technique. It has stabilized around 2 in some weeks. In reducing conditions, this ratio is stable and is of about 0.5. Modeling tools have allowed to propose a class of solids potentially able to control the uranium solubility. In oxidizing conditions, the uranyl hydrates (schoepite) evolve towards uranyl silicates which are thermodynamically more stable. In reducing conditions, a control of the uranium concentration in solution by U 4 O 9 is probable. (O.M.)

  6. Geochemical modelling of the weathering zone of the 'Mina Fe' U deposit (Spain): A natural analogue for nuclear spent fuel alteration and stability processes in radwaste disposal

    International Nuclear Information System (INIS)

    Arcos, D.; Perez del Villar, L.; Bruno, J.; Domenech, C.

    2008-01-01

    The 'Mina Fe' U deposit (Salamanca, Spain) has been studied in the context of Enresa's programme for U-mine sites restoration and also as a natural analogue for processes in high-level nuclear waste (HLNW) geological disposal. The investigations encompassed an array of geoscience disciplines, such as structural geology, mineralogy, hydrogeology and elemental and isotopic geochemistry and hydrogeochemistry of the site. Based on the obtained results, a conceptual mineralogical and geochemical model was performed integrating the main geochemical processes occurring at the site: the interaction between oxidised and slightly acidic water with pyrite, pitchblende, calcite and dolomite, as essential minerals of the U fracture-filling mineralisation, and hydroxyapatite from the host rock, as the main source of P. This conceptual model has been tested in a systematic numerical model, which includes the main kinetic (pyrite and pitchblende dissolution) and equilibrium processes (carbonate mineral dissolution, and goethite, schoepite and autunite secondary precipitation). The results obtained from the reactive-transport model satisfactorily agree with the conceptual model previously established. The assumption of the precipitation of coffinite as a secondary mineral in the system cannot be correctly evaluated due to the lack of hydrochemical data from the reducing zone of the site and valid thermodynamic and kinetic data for this hydrated U(IV)-silicate. This precipitation can also be hampered by the probable existence of dissolved U(IV)-organic matter and/or uranyl carbonate complexes, which are thermodynamically stable under the alkaline and reducing conditions that prevail in the reducing zone of the system. Finally, the intense downwards oxic and acidic alteration in the upper part of the system is of no relevance for the performance assessment of a HLNW disposal. However, the acidic and oxidised conditions are quickly buffered to neutral-alkaline and reducing at very

  7. Identification of chemical processes influencing constituent mobility during in-situ uranium leaching

    International Nuclear Information System (INIS)

    Sherwood, D.R.; Hostetler, C.J.; Deutsch, W.J.

    1984-07-01

    In-situ leaching of uranium has become a widely accepted method for production of uranium concentrate from ore zones that are too small, too deep, and/or too low in grade to be mined by conventional techniques. One major environmental concern that exists with in-situ leaching of uranium is the possible adverse effects mining might have on regional ground water quality. The leaching solution (lixiviant), which extracts uranium from the ore zone, might also mobilize other potential contaminants (As, Se, Mo, and SO 4 ) associated with uranium ore. Column experiments were performed to investigate the geochemical interactions between a lixiviant and a uranium ore during in-situ leaching and to identify chemical processes that might influence contaminant mobility. The analytical composition data for selected column effluents were used with the MINTEQ code to develop a computerized geochemical model of the system. MINTEQ was used to calculate saturation indices for solid phases based on the composition of the solution. A potential constraint on uranium leaching efficiency appears to be the solubility control of schoepite. Gypsum and powellite solubilities may limit the mobilities of sulfate and molybdenum, respectively. In contrast, the mobilities of arsenic and selenium were not limited by solubility constraints, but were influenced by other chemical interaction between the solution and sediment, perhaps adsorption. Bulk chemical and mineralogical analyses were performed on both the original and leached ores. Using these analyses together with the column effluent data, mass balance calculations were performed on five constituents based on solution chemical analysis and bulk chemical and γ-spectroscopy analysis for the sediment. 6 references, 10 figures, 10 tables

  8. Behaviour of high purity UO{sub 2}/H{sub 2}O interfaces under helium beam irradiation in deaerated conditions; Comportement des interfaces UO{sub 2}/H{sub 2}O de haute purete sous faisceau d'ions He{sup 2+} en milieu desaere

    Energy Technology Data Exchange (ETDEWEB)

    Mendes, E

    2005-11-15

    A question put within the framework of the nuclear fuel storage worn in geological site is what become to them in the presence of water. The aim of a fundamental program, of PRECCI project (ECA), is to highlight the behaviour of interfaces which can be used as models for the interfaces nuclear spent fuel/water if the fuel is uranium UO{sub 2} dioxide. This doctorate is interested in the effect of the alpha activity which is the only one that exist in the spent fuel after long periods. The aim is to identify the mechanisms of alteration and of leaching of surfaces under alpha irradiation. A method is developed to irradiate UO{sub 2}/H{sub 2}O interfaces in deaerated conditions with the beam of He{sup 2+} produced by a cyclotron. The He{sup 2+} ions cross an UO{sub 2} disc and emerge in water with an energy of 5 MeV. Leachings under irradiation are carried with a large range of particles flux. The post-irradiation characterization of the surface of the discs realised by micro-Raman spectroscopy allowed to identify the alteration layer. It is made up of studtite UO{sub 2}(O{sub 2}),4H{sub 2}O, and of schoepite UO{sub 3},xH{sub 2}O. The analysis of the solutions shows that the uranium release strongly increases. The electrochemical properties of the interfaces under irradiation strongly differ from those before irradiation. This work allows to propose that the radiolytic species seen by the interface are it during the heterogeneous phase of evolution of the traces and are species of short lives. Modeling show that the radiolytic radicals species can migrate toward the interface and react with the UO{sub 2} surface. (author)

  9. Validation of the WATEQ4 geochemical model for uranium

    International Nuclear Information System (INIS)

    Krupka, K.M.; Jenne, E.A.; Deutsch, W.J.

    1983-09-01

    As part of the Geochemical Modeling and Nuclide/Rock/Groundwater Interactions Studies Program, a study was conducted to partially validate the WATEQ4 aqueous speciation-solubility geochemical model for uranium. The solubility controls determined with the WATEQ4 geochemical model were in excellent agreement with those laboratory studies in which the solids schoepite [UO 2 (OH) 2 . H 2 O], UO 2 (OH) 2 , and rutherfordine ((UO 2 CO 3 ) were identified as actual solubility controls for uranium. The results of modeling solution analyses from laboratory studies of uranyl phosphate solids, however, identified possible errors in the characterization of solids in the original solubility experiments. As part of this study, significant deficiencies in the WATEQ4 thermodynamic data base for uranium solutes and solids were corrected. Revisions included recalculation of selected uranium reactions. Additionally, thermodynamic data for the hydroxyl complexes of U(VI), including anionic (VI) species, were evaluated (to the extent permitted by the available data). Vanadium reactions were also added to the thermodynamic data base because uranium-vanadium solids can exist in natural ground-water systems. This study is only a partial validation of the WATEQ4 geochemical model because the available laboratory solubility studies do not cover the range of solid phases, alkaline pH values, and concentrations of inorganic complexing ligands needed to evaluate the potential solubility of uranium in ground waters associated with various proposed nuclear waste repositories. Further validation of this or other geochemical models for uranium will require careful determinations of uraninite solubility over the pH range of 7 to 10 under highly reducing conditions and of uranyl hydroxide and phosphate solubilities over the pH range of 7 to 10 under oxygenated conditions

  10. Trace metal distribution and mobility in drill cuttings and produced waters from Marcellus Shale gas extraction: Uranium, arsenic, barium

    International Nuclear Information System (INIS)

    Phan, Thai T.; Capo, Rosemary C.; Stewart, Brian W.; Graney, Joseph R.; Johnson, Jason D.; Sharma, Shikha; Toro, Jaime

    2015-01-01

    -bearing minerals in drill cuttings would likely be followed by immobilization of U in secondary minerals such as schoepite, uranophane, and soddyite, or uraninite as conditions become more reducing. Oxidative dissolution of arsenic containing sulfides could release soluble As in arsenate form under oxic acidic conditions. The degree to which the As is subsequently immobilized depends on the redox conditions along the landfill flow path. The results suggest that proper management of drill cuttings can minimize mobilization of these metals by monitoring and controlling Eh, pH and dissolved constituents in landfill leachates

  11. Removal of uranium(VI) from the aqueous phase by iron(II) minerals in presence of bicarbonate

    Energy Technology Data Exchange (ETDEWEB)

    Regenspurg, Simona, E-mail: regens@gfz-potsdam.de [Industrial Ecology, Royal Institute of Technology (KTH), SE 10044 Stockholm (Sweden); Schild, Dieter; Schaefer, Thorsten; Huber, Florian [Institut fuer Nukleare Entsorgung (INE), Forschungszentrum Karlsruhe, 76344 Eggenstein-Leopoldshafen (Germany); Malmstroem, Maria E. [Industrial Ecology, Royal Institute of Technology (KTH), SE 10044 Stockholm (Sweden)

    2009-09-15

    Uranium(VI) mobility in groundwater is strongly affected by sorption of mobile U(VI) species (e.g. uranyl, UO{sub 2}{sup 2+}) to mineral surfaces, precipitation of U(VI) compounds, such as schoepite (UO{sub 2}){sub 4}O(OH){sub 6}.6H{sub 2}O), and by reduction to U(IV), forming sparingly soluble phases (uraninite; UO{sub 2}). The latter pathway, in particular, would be very efficient for long-term immobilization of U. In nature, Fe(II) is an important reducing agent for U(VI) because it frequently occurs either dissolved in natural waters, sorbed to matrix minerals, or structurally bound in many minerals. Redox reactions between U(VI) and Fe(II) depend not only on the availability of Fe(II) in the environment, but also on the chemical conditions in the aqueous solution. Under natural groundwater condition U(VI) forms complexes with many anionic ligands, which strongly affect its speciation. Carbonate, in particular, is known to form stable complexes with U, raising the question, if U(VI), when complexed by carbonate, can be reduced to UO{sub 2}. The goal of this study was to find out if Fe(II) when structurally bound in a mineral (as magnetite, Fe{sub 3}O{sub 4}) or sorbed to a mineral surface (as corundum, Al{sub 2}O{sub 3}) can reduce U(VI) to U(IV) in the presence of HCO{sub 3}{sup -}. Batch experiments were conducted under anaerobic conditions to observe U removal from the aqueous phase by the two minerals depending on HCO{sub 3}{sup -} addition (1 mM), U concentration (0.01-30 {mu}M) and pH value (6-10). Immediately after the experiments, the mineral surfaces were analyzed by X-ray photoelectron spectroscopy (XPS) to obtain information on the redox state of U bound to the solid surfaces. XPS results gave evidence that U(VI) can be reduced both by magnetite and by corundum amended with Fe(II). In the presence of HCO{sub 3}{sup -} the amount of reduced U on the mineral surfaces increased compared to carbonate-free solutions. This can be explained by the formation

  12. Hydrothermal interactions of cesium and strontium phases from spent unreprocessed fuel with basalt phases and basalts

    International Nuclear Information System (INIS)

    Komarneni, S.; Scheetz, B.E.; McCarthy, G.J.; Coons, W.E.

    1980-03-01

    This investigation is a segment of an extensive research program aimed at investigating the feasibility of long-term, subsurface storage of commercial nuclear waste. Specifically, it is anticipated that the waste will be housed in a repository mined from the basalt formations which lie beneath the Hanford Site. The elements monitored during the present experiments were Cs and Sr. These two elements represent significant biohazards if released from a repository and are the major heat producing radionuclides present in commercial radioactive waste. Several Cs phases and/or solutions were reacted with either isolated basalt phases or bulk-rock basalt, and the resulting solids and solutions were analyzed. The hydrothermal reactivity of SrZrO 3 , which is believed to be a probable host for Sr in SFE was investigated. While so far no evidence exists which indicates that Sr is present in a water soluble phase in spent fuel elements (SFE), detailed investigation of a potential hazard is warranted. This investigation has determined that some Cs compounds likely to be stable components of spent fuel (i.e., CsOH, Cs 2 MoO 4 , Cs 2 U 2 O 7 ) have significant hydrothermal solubilities. These solubilities are greatly decreased in the presence of basalt and/or basalt minerals. The decrease in the amount of Cs in solution results from reactions which form pollucite and/or CsAlSiO 4 , with the production of pollucite exceeding that of CsAlSiO 4 . Dissolution of β-Cs 2 U 2 O 7 implies solubilizing a uranium species to an undetermined extent. The production of schoepite (UO 3 .3H 2 O) during some experiments containing basalt phases, indicates a tendency to oxidize U 4+ to U 6+ . When diopside (nominally CaMgSi 2 O 6 ) and β-Cs 2 U 2 O 7 were hydrothermally reacted, at 300 0 C both UO 2 and UO 3 .3H 2 O were produced. Experiments on SrZrO 3 show it to be an unreactive phase

  13. Exploration for sandstone- type uranium mineralisation in the Siwaliks of northwestern Himalaya, India

    International Nuclear Information System (INIS)

    Swarnkar, B.M.; Kothari, P.K.; Umamaheswar, K.; Srinivasan, S.

    2002-01-01

    The Siwalik Group with a thickness of about 6000m of fluvial sediments of middle Miocene to Pleistocene age has been explored extensively over two decades for U, using various types of exploration techniques involving air-borne gamma-ray spectrometry, radiation jeep survey, hydrogeochemical survey, ground radiometric survey, radon survey, exploratory drilling and mining, Exploration effort by the Atomic Minerals Directorate for Exploration and Research (AMD) has helped in identifying numerous uranium occurrences spread over the entire Siwalik belt between Poonch (Jammu and Kashmir) in the west and Tanakpur (Uttar Pradesh) in the east, in the northwest Himalaya. Eight significant zones were delineated, mostly confining to distinct stratigraphic horizons of the transition zone between Middle and Upper Siwaliks, and occasionally the transition zone between Lower and Middle Siwaliks. These mineralised zones have a considerable lateral extent of up to 12 km and are associated with sandstones and rarely conglomerates. Uranium mineralisation occurs in the form of peneconcordant lensoidal bodies with individual lenses traceable from a few tens of metres to 700m, sub-parallel to strike or dip, with average grades varying from 0.020 - 0.060% U 3 O 8 and thickness less than a metre to 4m. The host rock of uranium mineralisation is predominantly sandstone containing carbonaceous matter, pyrite and clay pellets. The sandstone is often arkosic and micaceous, and termed as lithic wacke and arkosic wacke. The uranium minerals present are uraninite, pitchblende, coffinite and secondary minerals such as tyuyamunite, metatyuyamunite, uranophane, bayleyite, andersonite, schoepite, liebegite, swartzite, schroekingerite, wulfenite, billictite, betauranophane, autunite and torbernite. Relatively higher concentrations or Se, Mo, Cu, Co, V and Au have been noted in a few uranirerous zones. Concentration or uranium in the Siwalik clastic sediments is controlled by the redox interface

  14. Hydrothermal interactions of cesium and strontium phases from spent unreprocessed fuel with basalt phases and basalts

    Energy Technology Data Exchange (ETDEWEB)

    Komarneni, S.; Scheetz, B.E.; McCarthy, G.J.; Coons, W.E.

    1980-03-01

    This investigation is a segment of an extensive research program aimed at investigating the feasibility of long-term, subsurface storage of commercial nuclear waste. Specifically, it is anticipated that the waste will be housed in a repository mined from the basalt formations which lie beneath the Hanford Site. The elements monitored during the present experiments were Cs and Sr. These two elements represent significant biohazards if released from a repository and are the major heat producing radionuclides present in commercial radioactive waste. Several Cs phases and/or solutions were reacted with either isolated basalt phases or bulk-rock basalt, and the resulting solids and solutions were analyzed. The hydrothermal reactivity of SrZrO/sub 3/, which is believed to be a probable host for Sr in SFE was investigated. While so far no evidence exists which indicates that Sr is present in a water soluble phase in spent fuel elements (SFE), detailed investigation of a potential hazard is warranted. This investigation has determined that some Cs compounds likely to be stable components of spent fuel (i.e., CsOH, Cs/sub 2/MoO/sub 4/, Cs/sub 2/U/sub 2/O/sub 7/) have significant hydrothermal solubilities. These solubilities are greatly decreased in the presence of basalt and/or basalt minerals. The decrease in the amount of Cs in solution results from reactions which form pollucite and/or CsAlSiO/sub 4/, with the production of pollucite exceeding that of CsAlSiO/sub 4/. Dissolution of ..beta..-Cs/sub 2/U/sub 2/O/sub 7/ implies solubilizing a uranium species to an undetermined extent. The production of schoepite (UO/sub 3/.3H/sub 2/O) during some experiments containing basalt phases, indicates a tendency to oxidize U/sup 4 +/ to U/sup 6 +/. When diopside (nominally CaMgSi/sub 2/O/sub 6/) and ..beta..-Cs/sub 2/U/sub 2/O/sub 7/ were hydrothermally reacted, at 300/sup 0/C both UO/sub 2/ and UO/sub 3/.3H/sub 2/O were produced. Results of experiments on SrZrO/sub 3/ show it to be

  15. Classification of distribution coefficient data by mineral components and chemical forms

    International Nuclear Information System (INIS)

    Takeda, Seiji; Kimura, Hideo; Matsuzuru, Hideo

    1996-01-01

    The use of distribution coefficient (Kd) in radionuclide transport model has been reported in a number of papers. However, Kd data cover a wide range even for a specific element. In this study the Kd data of neptunium, uranium and selenium, which are included in sorption database (SDB, OECD/NEA) of radionuclides, were classified by a solid phase and a dominant species in a solution. The aqueous species of these elements were estimated by a geochemical model. The Kd data classified by the analyzed speciation were tested by a nonparametric statistical method. The results of tests proved that the Kd data of neptunium or uranium, which covered a wide range, were influenced by the supersaturation of Np(OH) 4 (s) or schoepite. The Kd data of neptunium could be classified by the dominant aqueous species, NpO 2 + , NpO 2 CO 3 - , NpO 2 OH(aq) and Np(OH) 4 (aq). The Kd data of these four dominant species which are not equilibrated with supersaturated Np(OH) 4 (s) are less than 100 ml/g. The analyzed aqueous species of uranium were UO 2 (OH) 2 (aq) and UO 2 (CO 3 ) n 2-2n (n=2,3) in hexavalent state. It is suggested that the distribution coefficient of neptunium and uranium depends on dominant aqueous species or charged species, i.e., cationic, anionic and nonionic forms. The dominant aqueous species of selenium are HSe - , HSeO 3 - , SeO 3 2- and SeO 4 2- . The result of the nonparametric statistical test shows that the Kd value of HSeO 3 - is higher than of other anionic forms. However, the influence of the species, HSe - , SeO 3 2- and SeO 4 2- , on Kd values is not clearly identified. Considering the dominant species, the Kd of elements are in ranges of 1 to 2 orders of magnitude being in general narrower than those classified by mineral and rock types. (author)

  16. Spent fuel. Dissolution and oxidation

    International Nuclear Information System (INIS)

    Grambow, B.

    1989-03-01

    Data from studies of the low temperature air oxidation of spent fuel were retrieved in order to provide a basis for comparison between the mechanism of oxidation in air and corrosion in water. U 3 O 7 is formed by diffusion of oxygen into the UO 2 lattice. A diffusion coefficient of oxygen in the fuel matric was calculated for 25 degree C to be in the range of 10 -23 to 10 -25 m 2 /s. The initial rates of U release from spent fuel and from UO 2 appear to be similar. The lowest rates (at 25 degree c >10 -4 g/(m 2 d)) were observed under reducing conditions. Under oxidizing conditions the rates depend mainly of the nature and concentraion of the oxidant and/or on corbonate. In contact with air, typical initial rates at room temperature were in the range between 0.001 and 0.1 g/(m 2 d). A study of apparent U solubility under oxidizing conditions was performed and it was suggested that the controlling factor is the redox potential at the UO 2 surface rather than the E h of the bulk solution. Electrochemical arguments were used to predict that at saturation, the surface potential will eventually reach a value given by the boundaries at either the U 3 O 7 /U 3 O 8 or the U 3 O 7 /schoepite stability field, and a comparison with spent fuel leach data showed that the solution concentration of uranium is close to the calculated U solubility at the U 3 O 7 /U 3 O 8 boundary. The difference in the cumulative Sr and U release was calculated from data from Studsvik laboratory. The results reveal that the rate of Sr release decreases with the square root of time under U-saturated conditions. This time dependence may be rationalized either by grain boundary diffusion or by diffusion into the fuel matrix. Hence, there seems to be a possibility of an agreement between the Sr release data, structural information and data for oxygen diffusion in UO 2 . (G.B.)

  17. Geochemical modelling of the weathering zone of the 'Mina Fe' U deposit (Spain): A natural analogue for nuclear spent fuel alteration and stability processes in radwaste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Arcos, D. [AMPHOS XXI Consulting S.L., Passeig de Rubi, 29-31, 08197 Valldoreix, Barcelona (Spain)], E-mail: david.arcos@amphos21.com; Perez del Villar, L. [CIEMAT, Dpto.de Medio Ambiente, Avda, Complutense 22, 28040 Madrid (Spain); Bruno, J.; Domenech, C. [AMPHOS XXI Consulting S.L., Passeig de Rubi, 29-31, 08197 Valldoreix, Barcelona (Spain)

    2008-04-15

    The 'Mina Fe' U deposit (Salamanca, Spain) has been studied in the context of Enresa's programme for U-mine sites restoration and also as a natural analogue for processes in high-level nuclear waste (HLNW) geological disposal. The investigations encompassed an array of geoscience disciplines, such as structural geology, mineralogy, hydrogeology and elemental and isotopic geochemistry and hydrogeochemistry of the site. Based on the obtained results, a conceptual mineralogical and geochemical model was performed integrating the main geochemical processes occurring at the site: the interaction between oxidised and slightly acidic water with pyrite, pitchblende, calcite and dolomite, as essential minerals of the U fracture-filling mineralisation, and hydroxyapatite from the host rock, as the main source of P. This conceptual model has been tested in a systematic numerical model, which includes the main kinetic (pyrite and pitchblende dissolution) and equilibrium processes (carbonate mineral dissolution, and goethite, schoepite and autunite secondary precipitation). The results obtained from the reactive-transport model satisfactorily agree with the conceptual model previously established. The assumption of the precipitation of coffinite as a secondary mineral in the system cannot be correctly evaluated due to the lack of hydrochemical data from the reducing zone of the site and valid thermodynamic and kinetic data for this hydrated U(IV)-silicate. This precipitation can also be hampered by the probable existence of dissolved U(IV)-organic matter and/or uranyl carbonate complexes, which are thermodynamically stable under the alkaline and reducing conditions that prevail in the reducing zone of the system. Finally, the intense downwards oxic and acidic alteration in the upper part of the system is of no relevance for the performance assessment of a HLNW disposal. However, the acidic and oxidised conditions are quickly buffered to neutral-alkaline and

  18. 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

  19. Production and characterization of monodisperse uranium particles for nuclear safeguards applications

    International Nuclear Information System (INIS)

    Knott, Alexander

    2016-01-01

    controlled primarily by the aerosol precursor solution and the production parameters during the aerosol generation - in particular the liquid feed rate and the frequency of the orifice. The final particle morphology is controlled by the precipitation conditions during the conversion from aerosol droplets to solid entities. Small changes to these parameters have a significant influence on the final geometry, size and morphology. The second part of this thesis deals with the characterization of microparticles. A selection of particles was chosen to present the developments over a period of 12 months. Scanning Electron Microscopy coupled with Energy Dispersive X-Ray Spectroscopy (SEM-EDX) was used for various applications, e.g. to verify the elemental content and to assess the size and geometry of the particles. Combined Focused Ion Beam (FIB-SEM) studies revealed the presence of a porous inner structure for all solid particles. Hence, the resulting overall density was less than expected. Time of Flight Secondary Ionization Mass Spectrometry (TOF-SIMS) studies evaluated the elemental content and demonstrated the need for cleanliness since minute quantities of contaminations could be found in single particles. Micro Raman investigations were used to determine the crystallinity, crystal orientation and uranium species. The measurements showed that particles primarily consist of U_3O_8. Parts consist of Meta-schoepite and U(IV)-hydroxide which indicates residual water inside the crystal lattice. Micro Raman investigations were performed at CEA (Ile de France) and at the TU-Vienna. SIMS measurements were performed at Safeguards Analytical Services - Environmental Sample Laboratory (SGAS-ESL) on the Large Geometry-SIMS (LG-SIMS) with the scope to assess their performance as a QC material. Particles produced at Juelich were also compared directly against existing QC- and reference materials. Investigations and characterization assays on monodisperse microparticles indicate

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

    International Nuclear Information System (INIS)

    Ollila, K.

    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 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 2 solubility under the disposal conditions of spent fuel, varied from strongly oxidising (air-saturated), anaerobic (N 2 , O 2 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 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 2 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 (U 3 O 8 -UO 3 ) 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 UO 2 pellets, the

  1. Production and characterization of monodisperse uranium particles for nuclear safeguards applications

    Energy Technology Data Exchange (ETDEWEB)

    Knott, Alexander

    2016-07-01

    demonstrated that the particle size can be controlled primarily by the aerosol precursor solution and the production parameters during the aerosol generation - in particular the liquid feed rate and the frequency of the orifice. The final particle morphology is controlled by the precipitation conditions during the conversion from aerosol droplets to solid entities. Small changes to these parameters have a significant influence on the final geometry, size and morphology. The second part of this thesis deals with the characterization of microparticles. A selection of particles was chosen to present the developments over a period of 12 months. Scanning Electron Microscopy coupled with Energy Dispersive X-Ray Spectroscopy (SEM-EDX) was used for various applications, e.g. to verify the elemental content and to assess the size and geometry of the particles. Combined Focused Ion Beam (FIB-SEM) studies revealed the presence of a porous inner structure for all solid particles. Hence, the resulting overall density was less than expected. Time of Flight Secondary Ionization Mass Spectrometry (TOF-SIMS) studies evaluated the elemental content and demonstrated the need for cleanliness since minute quantities of contaminations could be found in single particles. Micro Raman investigations were used to determine the crystallinity, crystal orientation and uranium species. The measurements showed that particles primarily consist of U{sub 3}O{sub 8}. Parts consist of Meta-schoepite and U(IV)-hydroxide which indicates residual water inside the crystal lattice. Micro Raman investigations were performed at CEA (Ile de France) and at the TU-Vienna. SIMS measurements were performed at Safeguards Analytical Services - Environmental Sample Laboratory (SGAS-ESL) on the Large Geometry-SIMS (LG-SIMS) with the scope to assess their performance as a QC material. Particles produced at Juelich were also compared directly against existing QC- and reference materials. Investigations and characterization assays