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Sample records for fitting uo2 interatomic

  1. Grain growth in UO2

    International Nuclear Information System (INIS)

    Hastings, I.J.; Scoberg, J.A.; Walden, W.

    1979-06-01

    Grain growth studies have been carried out on UO 2 to provide data for the fuel modelling program and to evaluate fuel fabricated in commissioning the Mixed Oxide Fuel Fabrication Laboratory at Chalk River Nuclear Laboratories. Fuel examined includes natural UO 2 commercially fabricated from ADU powder for CANDU reactors; natural UO 2 commercially fabricated from AU powder; natural UO 2 from ADU and AU powder, fabricated in the MOFFL; and commercially fabricated UO 2 enriched 1.7, 4.5, and 9.6 wt. percent U-235 in U. Samples were step-annealed in vacuo at 1870-2070 K for up to 32.5 h. All data fit a (grain size)sup(2.5) versus annealing time relationship. Apparent activation energy for grain growth, Q, depends on fuel type and varies from 150+-10 kJ/mol for early AU powder to 360+-10 kJ/mol for pellets from ADU fabricated in the MOFFL. Grain sizes calculated using the laboratory equation in a fuel performance code tend to be greater than those measured in irradiated natural fuel, suggesting irradiation-induced inhibition of grain growth. However, any inhibition is equivalent to that expected for a systematic 5 percent underpredicition in reactor power. (author)

  2. Sintering of nonstoichiometric UO2

    International Nuclear Information System (INIS)

    Susnik, D.; Holc, J.

    1983-01-01

    Activated sintering of UO 2 pellets at 1100 deg C is described. In CO 2 atmosphere is UO 2 is nonstoichiometric and pellets from active UO 2 powders sinter at 900 deg C to high density. At 1100 deg C the final sintered density is practically achieved at heating on sintering temperature. After reduction and cooling in H 2 atmosphere which is followed sintering in CO 2 the structure is identical to the structured UO 2 pellets sintered at high temperature in H 2 . Density of activated sintered UO 2 pellets is stable, even after additional sintering at 1800 deg C. (author)

  3. New UO2 fuel studies

    International Nuclear Information System (INIS)

    Dehaudt, P.; Lemaignan, C.; Caillot, L.; Mocellin, A.; Eminet, G.

    1998-01-01

    With improved UO 2 fuels, compared with the current PWR, one would enable to: retain the fission products, rise higher burn-ups and deliver the designed power in reactor for longer times, limit the pellet cladding interaction effects by easier deformation at high temperatures. Specific studies are made in each field to understand the basic mechanisms responsible for these improvements. Four programs on new UO 2 fuels are underway in the laboratory: advanced microstructure fuels (doped fuels), fuels containing Er 2 O 3 a burnable absorber, fuels with improved caesium retention, composite fuels. The advanced microstructure UO 2 fuels have special features such as: high grain sizes to lengthen the fission gas diffusion paths, intragranular precipitates as fission gas atoms pinning sites, intergranular silica based viscoplastic phases to improve the creep properties. The grain size growth can be obtained with a long time annealing or with corundum type oxide additives partly soluble in the UO 2 lattice. The amount of doping element compared with its solubility limit and the sintering conditions allows to obtain oxide or metallic precipitates. The fuels containing Er 2 O 3 as a burnable absorber are under irradiation in the TANOX device at the present time. Specific sintering conditions are required to improve the erbium solubility in UO 2 and to reach standard or large grain sizes. The improved caesium retention fuels are doped with SiO 2 +A1 2 O 3 or SiO 2 +ZrO 2 additives which may form stable compounds with the Cs element in accidental conditions. The composite fuels are made of UO 2 particles of about 100 μm in size dispersed in a molybdenum metallic (CERMET) or MgA1 2 O 4 ceramic (CERCER) matrix. The CERMET has a considerably higher thermal conductivity and remains ''cold'' during irradiation. The concept of double barrier (matrix+fuel) against fission products is verified for the CERMET fuel. A thermal analysis of all the irradiated rods shows that the thermal

  4. Magnetoelastic interactions in UO2

    International Nuclear Information System (INIS)

    Faber, J. Jr.; Lander, G.H.; Cooper, B.R.

    1975-12-01

    Neutron diffraction measurements of the elastic magnetic scattering cross section from antiferromagnetic UO 2 show additional nuclear intensity below T/sub N/ = 30.8 0 K. An examination was made of the possibility of analyzing the additional scattering in terms of homogeneous distortions, which involve shifts of the oxygen atoms from their fluorite lattice sites. The behavior arising from the presence of these homogeneous distortion modes formed the basis for Allen's theory of a cooperative Jahn--Teller effect in UO 2 . However, an analysis in terms of these homogeneous distortions cannot explain the neutron data. But, by extending Allen's concepts to include inhomogeneous deformations, corresponding to a zone boundary q = (π/a) (1,0,0) phonon, excellent agreement is obtained between theory and experiment. The oxygen displacement is 0.014(1) A from the fluorite lattice positions and, in addition, the inhomogeneous deformation (T/sub 2g/(Q 1 )--T/sub 1g/) does not require a reduction in the overall symmetry of the unit cell. The essential features of Allen's theory for UO 2 can still be maintained

  5. Study of UO2 radioinduced densification

    International Nuclear Information System (INIS)

    Stora, J.P.; Bruet, M.

    1975-01-01

    Measurements of radioinduced densification were performed on UO 2 DCN (intergranular fine porosity) and UO 2 DCI (interaggregate coarse porosity) in the Anemone device. The densification kinetics was followed by measuring the shrinkage of the oxide column on neutron radiographic plates. UO 2 DCI was found stable in regard to densification. At power near 450Wcm -1 , densification is hitten by restructuring phenomena [fr

  6. Ceramic UO2 powder production at Cameco Corporation

    International Nuclear Information System (INIS)

    Kwong, A.K.; Kuchurean, S.M.

    1997-01-01

    This presentation covers the various aspects of ceramic grade uranium dioxide (UO 2 ) powder production at Cameco Corporation and its use as fuel and blanket fuel for heavy-water and light-water reactors, respectively. In addition, it discusses the significant production variables that affect production and product quality. It also provides an insight into how various support groups such as Quality Assurance, Analytical Services, and Technology Development fit into the quality cycle and contribute to a successful operation. The ability of Cameco to identify, measure and control the physical and chemical properties of ceramic grade UO 2 has resulted in the production of uniform quality powder. This has meant that 100% of Cameco's ceramic grade UO 2 powder produced since mid-1989 has been accepted by the fuel manufacturers. (author)

  7. Sintering diagrams of UO2

    International Nuclear Information System (INIS)

    Mohan, A.; Soni, N.C.; Moorthy, V.K.

    1979-01-01

    Ashby's method (see Acta Met., vol. 22, p. 275, 1974) of constructing sintering diagrams has been modified to obtain contribution diagrams directly from the computer. The interplay of sintering variables and mechanisms are studied and the factors that affect the participation of mechanisms in UO 2 are determined. By studying the physical properties, it emerges that the order of inaccuracies is small in most cases and do not affect the diagrams. On the other hand, even a 10% error in activation energies, which is quite plausible, would make a significant difference to the diagram. The main criticism of Ashby's approach is that the numerous properties and equations used, communicate their inaccuracies to the diagrams and make them unreliable. The present study has considerably reduced the number of factors that need to be refined to make the sintering diagrams more meaningful. (Auth.)

  8. Use of UO 2 films for electrochemical studies

    Science.gov (United States)

    Miserque, F.; Gouder, T.; Wegen, D. H.; Bottomley, P. D. W.

    2001-10-01

    UO 2 films have been prepared by dc reactive sputtering of a uranium metal target in an Ar/O 2 atmosphere. We have used the films deposited on gold substrates as working electrodes for electrochemical investigations as simulating the surfaces of fuel pellets. Film composition was determined by photoelectron spectroscopy (XPS and UPS) and X-ray diffraction (XRD). The oxide stoichiometry as a function of deposition conditions was determined and the appropriate conditions for UO 2.0 formation established. AC impedance and cyclic voltammetry measurements were performed. A double RC electrical equivalent circuit was used to fit the data from impedance measurements, similar to those used in unirradiated UO 2 or spent fuel pellets. However due to the porosity or adhesion defects on the thin films that permitted a direct contact between the solution and the gold substrate, we were obliged to add a contribution simulating the water-gold system. Cyclic voltammetry measurements show the influence of pH on the dissolution mechanism. Alkaline solutions permit the formation of an oxidised layer (UO 2.33) which is not present in the acidic solutions. In both pH=2 and pH=6 solutions, a U VI species layer is formed.

  9. Phonon density of states and anharmonicity of UO2

    Science.gov (United States)

    Pang, Judy W. L.; Chernatynskiy, Aleksandr; Larson, Bennett C.; Buyers, William J. L.; Abernathy, Douglas L.; McClellan, Kenneth J.; Phillpot, Simon R.

    2014-03-01

    Phonon density of states (PDOS) measurements have been performed on polycrystalline UO2 at 295 and 1200 K using time-of-flight inelastic neutron scattering to investigate the impact of anharmonicity on the vibrational spectra and to benchmark ab initio PDOS simulations performed on this strongly correlated Mott insulator. Time-of-flight PDOS measurements include anharmonic linewidth broadening, inherently, and the factor of ˜7 enhancement of the oxygen spectrum relative to the uranium component by the increased neutron sensitivity to the oxygen-dominated optical phonon modes. The first-principles simulations of quasiharmonic PDOS spectra were neutron weighted and anharmonicity was introduced in an approximate way by convolution with wave-vector-weighted averages over our previously measured phonon linewidths for UO2, which are provided in numerical form. Comparisons between the PDOS measurements and the simulations show reasonable agreement overall, but they also reveal important areas of disagreement for both high and low temperatures. The discrepancies stem largely from a ˜10 meV compression in the overall bandwidth (energy range) of the oxygen-dominated optical phonons in the simulations. A similar linewidth-convoluted comparison performed with the PDOS spectrum of Dolling et al. obtained by shell-model fitting to their historical phonon dispersion measurements shows excellent agreement with the time-of-flight PDOS measurements reported here. In contrast, we show by comparisons of spectra in linewidth-convoluted form that recent first-principles simulations for UO2 fail to account for the PDOS spectrum determined from the measurements of Dolling et al. These results demonstrate PDOS measurements to be stringent tests for ab inito simulations of phonon physics in UO2 and they indicate further the need for advances in theory to address the lattice dynamics of UO2.

  10. Characterization of UO2 by infrared spectroscopy

    International Nuclear Information System (INIS)

    Faeda, Kelly C.M.; Machado, Geraldo C.; Lameiras, Fernando S.

    2011-01-01

    The characterization of nuclear fuel is of great importance to minimize the effects related to burnup and temperature and to achieve stability during in-core operation. The understanding the U-O system and its thermodynamic properties has fundamental importance in nuclear industry. Many physical properties of UO 2±x depend on the ratio O / U, such as the electrical conductivity and thermal properties, as well as the diffusivities of its constituents and solutes. The U-O system presents various oxides such as UO 2±x , U 4 O 9 , U 3 O 8 , and UO 3 . The control of the O/U relation is critical to the manufacturing process of UO 2 . In this work, the infrared spectroscopy was used to identify the presence of phases in UO 2 powder samples that cannot be identified by thermogravimetry and X-ray diffraction. (author)

  11. UO2 pellet and manufacturing method

    International Nuclear Information System (INIS)

    Komada, Kiichi; Nishinaka, Keiji; Adachi, Kazunori; Fujiwara, Shuji.

    1995-01-01

    The present invention concerns an uranium dioxide pellet having a large crystal grain size. The grain size of the pellet is enlarged to increase the distance of an FP gas generated in the crystal grain to reach the grain boundary and, as a result, decrease the releasing speed of the FP gas. A UO 2 powder having a specific surface area of from 5 to 50m 2 /g is used as a starting powder in a step of forming a molding product, and chlorine or a chlorine compound is added in such an amount that the chlorine content in the UO 2 pellet is from 3 to 25ppm, in one of a production step, a molding step or a sintering step for UO 2 powder. With such procedures, a UO 2 pellet having a large crystal grain size can be prepared with good reproducibility. (T.M.)

  12. Geometrical dimensioning of PWR UO2 pellets

    International Nuclear Information System (INIS)

    Silva, A.T.

    1988-08-01

    The finite element structural program SAP-IV is used to calculate UO 2 pellet strains developed under thermal gradients in pressurized water reactors. The applied procedure allows to analyse the influence of various aspects of pelet geometry on cladding strains and can be utilized for the dimensioning of UO 2 pellets. Pellets purchased with flat ends, with dishes pressed into both ends, shouders, and a 45-deg edge chamfer are analysed. The analyse results are compared with experiemtnal data. (author) [pt

  13. A charge-optimized many-body potential for the U-UO2-O2 system

    Science.gov (United States)

    Li, Yangzhong; Liang, Tao; Sinnott, Susan B.; Phillpot, Simon R.

    2013-12-01

    Building on previous charge-optimized many-body (COMB) potentials for metallic α-U and gaseous O2, we have developed a new potential for UO2, which also allows the simulation of U-UO2-O2 systems. The UO2 lattice parameter, elastic constants and formation energies of stoichiometric and non-stoichiometric intrinsic defects are well reproduced. Moreover, this is the first rigid-ion potential that produces the correct deviation of the Cauchy relation, as well as the first classical interatomic potential that is able to determine the defect energies of non-stoichiometric intrinsic point defects in UO2 with an appropriate reference state. The oxygen molecule interstitial in the α-U structure is shown to decompose, with some U-O bonds approaching the natural bond length of perfect UO2. Finally, we demonstrate the capability of this COMB potential to simulate a complex system by performing a simulation of the α-U + O2 → UO2 phase transformation. We also identify a possible mechanism for uranium oxidation and the orientation of the resulting fluorite UO2 structure relative to the coordinate system of orthorhombic α-U.

  14. Dissolution of UO2 in redox conditions

    International Nuclear Information System (INIS)

    Casas, I.; Pablo de, J.; Rovira, M.

    1998-01-01

    The performance assessment of the final disposal of the spent nuclear fuel in geological formations is strongly dependent on the spent fuel matrix dissolution. Unirradiated uranium (IV) dioxide has shown to be very useful for such purposes. The stability of UO 2 is very dependent on vault redox conditions. At reducing conditions, which are expected in deep groundwaters, the dissolution of the UO 2 -matrix can be explained in terms of solubility, while under oxidizing conditions, the UO 2 is thermodynamically unstable and the dissolution is kinetically controlled. In this report the parameters which affect the uranium solubility under reducing conditions, basically pH and redox potential are discussed. Under oxidizing conditions, UO 2 dissolution rate equations as a function of pH, carbonate concentration and oxidant concentration are reported. Dissolution experiments performed with spent fuel are also reviewed. The experimental equations presented in this work, have been used to model independent dissolution experiments performed with both unirradiated and irradiated UO 2 . (Author)

  15. A Characterization Research of UO2 Powder for UO2 Pellet Fabrication of Candu Type

    International Nuclear Information System (INIS)

    Rachmawati, M.

    1998-01-01

    A characterization research of of UO 2 powder for UO 2 pellet fabrication of Candu type is reported in this paper. The research has been conducted by characterizing sinterability, compactibility, and compressibility of UO 2 (Cameco) without a pre-compacting and UO 2 powder the result of a pre-compacting. The pre-compacting UO 2 powder has been done to have particle size to less than 150 mu (150-800) mu, and more than 800 mu with distribution varied. Sinterability of each group of particle sizes is analyzed using Thermogravimetric-Differential Thermal Analysis (TG-DTA). Then the final compacting to the powder is done using compaction pressure varied from 1 MP to 4 MP to the all groups of the particle sizes to find the optimum pressure by measuring the density and mechanical strength of the UO 2 green pellet. Both measurements are performed using Micrometer and Universal Testing Machine respectively. The result of this investigation shows that the group of UO 2 powder with no pre-compacting with particle size of less than 150 mu with 60% distribution and (150-800) mu size with 40% distribution are the UO 2 pellets which are eligible in terms of their density and mechanical strength

  16. Fabrication of ThO2, UO2, and PuO2-UO2 pellets

    International Nuclear Information System (INIS)

    Rasmussen, D.E.; Jentzen, W.R.; McCord, R.B.

    1978-01-01

    Fabrication of ThO pellets for EBR-II irradiation testing and fabrication of UO 2 and PuO 2 -UO 2 pellets for United Kingdom Prototype Fast Reactor (PFR) irradiation testing is discussed. Effect of process parameters on density and microstructure of pellets fabricated by the cold press and sinter technique is reviewed

  17. FABRIKASI MIKROSFIR UO2 MENGGUNAKAN TEKNIK AERASI

    Directory of Open Access Journals (Sweden)

    Meniek Rachmawati

    2016-10-01

    Full Text Available ABSTRAK FABRIKASI MIKROSFIR UO2 MENGGUNAKAN TEKNIK AERASI. Telah dikembangkan proses fabrikasi mikrosfir UO2 berdensitas rendah untuk umpan langsung proses peletisasi bahan bakar reaktor PHWR maju. Fabrikasi mikrosfir UO2 berdensitas rendah dilakukan dengan cara sol-gel menggunakan teknik aerasi pada sol/broth dengan metode eksternal dan tiga variasi cara gelasi. Pada teknik aerasi, broth disiapkan langsung digelasi tanpa didiamkan selama satu malam. Teknik aerasi merupakan kebalikan dari teknik deaerasi yang digunakan pada fabrikasi mikrosfir bahan bakar HTGR. Broth yang telah disiapkan dengan perbandingan mol NO3−/U antara 1,5 hingga 1,7 dengan pH larutan 1,6 dan viskositas antara 630-660cP langsung digelasi dengan tiga cara gelasi. Proses gelasi cara 1 dan cara 2 dilakukan dengan melewatkan broth pada dispersion nozzle berdiameter 1mm yang digetarkan dengan electromagnetic vibrator pada 150 Hz dengan media untuk droplet jatuh bebas yang berbeda sebelum masuk ke dalam larutan NH4OH, sedangkan gelasi dengan cara 3 dilakukan secara manual. Mikrosfir UO2 basah yang diperoleh dari ketiga cara gelasi di atas mendapat perlakuan panas yang sama yaitu dikeringkan pada temperatur 85 ºC dan 220 ºC masing-masing selama selama 1jam, dilanjutkan dengan proses kalsinasi mikrosfir UO2 selama 1 jam pada temperatur 500 ºC dalam media gas O2 dan direduksi pada temperatur 600 ºC dalam media campuran gas N2 dan H2 selama 1 jam. Mikrosfir UO2 hasil gelasi dengan cara 3 dipilih untuk disortir dan dikarakterisasi. Hasil karakterisasi memberikan data karakteristik mikrosfir UO2 berupa data diameter mikrosfir sebesar 900 µm, tap density 1,90 g/cm3 dan luas muka spesifik sebesar 6 m2/g. Hasil analisis dan hasil karakterisasi kemudian dibandingkan dengan data penelitian lain sehingga dapat disimpulkan bahwa penggunaan teknik aerasi pada broth menghasilkan mikrosfir UO2 berdensitas rendah yang memenuhi kriteria sebagai umpan langsung proses peletisasi bahan

  18. Adsorptive recovery of UO2(2+) from aqueous solutions using collagen-tannin resin.

    Science.gov (United States)

    Sun, Xia; Huang, Xin; Liao, Xue-pin; Shi, Bi

    2010-07-15

    Collagen-tannin resin (CTR), as a novel adsorbent, was prepared via reaction of collagen with black wattle tannin and aldehyde, and its adsorption properties to UO(2)(2+) were investigated in detail, including pH effect, adsorption kinetics, adsorption equilibrium and column adsorption kinetics. The adsorption of UO(2)(2+) on CTR was pH-dependent, and the optimal pH range was 5.0-6.0. CTR exhibited excellent adsorption capacity to UO(2)(2+). For instance, the adsorption capacity obtained at 303 K and pH 6.0 was as high as 0.91 mmol UO(2)(2+)/g when the initial concentration of UO(2)(2+) was 1.0 mmol/L. In kinetics studies, the adsorption equilibrium can be reached within 300 min, and the experimental data were well fitted by the pseudo-second-order rate model, and the equilibrium adsorption capacities calculated by the model were almost the same as those determined by experiments. The adsorption isotherms could be well described by the Freundlich equation with the correlation coefficients (R(2)) higher than 0.99, the adsorption behaviors of UO(2)(2+) on CTR column were investigated as well. Present study suggested that the CTR can be used for the adsorptive recovery of UO(2)(2+) from aqueous solutions. 2010 Elsevier B.V. All rights reserved.

  19. Thermal conductivity and thermal diffusivity of solid UO2

    International Nuclear Information System (INIS)

    Fink, J.K.; Chasanov, M.G.; Leibowitz, L.

    1981-06-01

    New equations for the thermal conductivity of solid UO 2 were derived based upon a nonlinear least squares fit of the data available in the literature. In the development of these equations, consideration was given to their thermodynamic consistency with heat capacity and density and theoretical consistency with enthalpy and heat capacity. Consistent with our previous treatment of enthalpy and heat capacity, 2670 K was selected as the temperature of a phase transition. A nonlinear equation, whose terms represent contributions due to phonons and electrons, was selected for the temperature region below 2670 K. Above 2670 K, the data were fit by a linear equation

  20. Modelling of UO2 oxidation in steam

    International Nuclear Information System (INIS)

    Brito, A.C.; Iglesias, F.C.; Liu, Y.

    1996-01-01

    A computer model has been developed for calculating oxidation of UO 2 at high temperatures in steam oxidising conditions. Several methods to calculate the partial pressure of oxygen in the fuel and in the environment surrounding the fuel are available. The various methodologies have been compared and the best models have been compiled into a computer model which will be implemented into fuel thermal/mechanical behaviour codes such as FACTAR 2.0 (LOECI) and ELESIM/ELOCA. Calculations from the computer model have been compared to experimental results. The calculated oxidation reaction kinetics are in good agreement with the experimental data. (author)

  1. SEM hot stage sintering of UO2

    International Nuclear Information System (INIS)

    Miller, D.J.

    1976-06-01

    The sintering of hyperstoichiometric uranium dioxide powder compacts, in the hot stage of a scanning electron microscope, was continuously monitored using 16 mm time lapse movies. From alumina microspheres placed on the surface of the compacts, shrinkage measurements were obtained. Converting shrinkage measurements into densification profiles indicates that a maximum densification rate is reached at a critical density, independent of the constant heating rates. At temperatures above 1350 0 C, the movement of the reference microspheres made shrinkage measurements impossible. It is believed the evolution of UO 3 gas from hyperstoichiometric UO 2 is the cause of this limitation

  2. UO2 dissolution rates: A review

    International Nuclear Information System (INIS)

    McKenzie, W.F.

    1992-09-01

    This report reviews literature data on UO 2 dissolution kinetics and provides a framework for guiding future experimental studies as well as theoretical modeling studies. Under oxidizing conditions, UO 2 dissolution involves formation of an oxidized surface layer which is then dissolved by formation of aqueous complexes. Higher oxygen pressures or other oxidants are required at higher temperatures to have dissolution rates independent of oxygen pressure. At high oxygen pressures (1-5 atm, 25-70 C), the dissolution rate has a one-half order dependence on oxygen pressure, whereas at oxygen pressures below 0.2 atm, Grandstaff (1976), but nobody else, observed a first-order dependence on dissolution rate. Most people found a first-order dependence on carbonate concentration; Posey-Dowty (1987) found independence of carbonate at pH 7 to 8.2. Dissolution rates increase with temperature except in experiments involving granitic groundwater. Dissolution rates were generally greater under acid or basic conditions than near neutral pH

  3. Modeling of UO2 aqueous dissolution over a wide range of conditions

    International Nuclear Information System (INIS)

    Steward, S.A.; Weed, H.C.

    1993-11-01

    Previously it was not possible to predict reliably the rate at which spent fuel would react with groundwater because of conflicting data in the literature. The dissolution of the UO 2 spent fuel matrix is a necessary step for aqueous release of radioactive fission products. Statistical experimental design was used to plan a set of UO 2 dissolution experiments to examine systematically the effects of temperature (25--75C), dissolved oxygen (0.002--0.2 atm overpressure), pH (8--10) and carbonate (2-200x10 -4 molar) concentrations on UO 2 dissolution. The average uranium dissolution rate was 4.3 mg/m 2 /day. The regression fit of the data indicate an Arrhenius type activation energy of 8750 cal/mol and a half-power dependence on dissolved oxygen in the simulated groundwater

  4. Measurements of thermal disadvantage factors in light-water moderated PuO2-UO2 and UO2 lattices

    International Nuclear Information System (INIS)

    Ohno, Akio; Kobayashi, Iwao; Tsuruta, Harumichi; Hashimoto, Masao; Suzaki, Takenori

    1980-01-01

    The disadvantage factor for thermal neutrons in light-water moderated PuO 2 -UO 2 and UO 2 square lattices were obtained from measurements of thermal neutron density distributions in a unit lattice cell, measured with Dy-Al wire detectors. The lattices consisted of 3.4 w/o PuO 2 .UO 2 and 2.6 w/o UO 2 fuel rods, and the water-to-fuel volume ratio within the unit cell was parametrically changed. The PuO 2 .UO 2 and UO 2 fuel rods were designed to realize equal fissile atomic number density. The disadvantage factors thus measured were 1.36 +- 0.07, 1.37 +- 0.08, 1.40 +- 0.06 and 1.38 +- 0.06 in the PuO 2 .UO 2 fuel lattices, and 1.30 +- 0.06, 1.31 +- 0.08, 1.30 +- 0.08 and 1.33 +- 0.06 in the UO 2 , for water-to-fuel volume ratios, of 1.76, 2.00, 2.38 and 2.95, respectively. This difference in disadvantage factor between PuO 2 .UO 2 and UO 2 fuel lattices corresponds to about 8%. Calculated results obtained by multigroup transport code LASER agreed well with the measured ones. (author)

  5. Spent fuel UO2 matrix corrosion behaviour studies through alpha-doped UO2 pellets leaching

    International Nuclear Information System (INIS)

    Muzeau, B.; Jegou, C.; Broudic, V.

    2005-01-01

    Full text of publication follows: The option of direct disposal of spent nuclear fuel in a deep geological formation raises the need to investigate the long-term behaviour of the UO 2 matrix in aqueous media subjected to α-β-γ radiations. The β-γ emitters account for the most of the activity of spent fuel at the moment it is removed from the reactor, but diminish within a millennial time frame by over three orders of magnitude to less than the long-term activity. The latter persist over much longer time periods and must therefore be taken into account over geological disposal scale. In the present investigation the UO 2 matrix corrosion under alpha radiation is studied as a function of different parameters such as: the alpha activity, the carbonates and hydrogen concentrations,.. In order to study the effect of alpha radiolysis of water on the UO 2 matrix, 238/239 Pu doped UO 2 pellets (0.22 %wt. Pu total) were fabricated with different 238 Pu/ 239 Pu ratio to reproduce the alpha activity of a 47 GWd.t HMi -1 UOX spent fuel at different milestones in time (15, 50, 1500, 10000 and 40000 years). Undoped UO 2 pellets were also available as reference sample. Leaching experiments were conducted in deionized or carbonated water (NaHCO 3 1 mM), under Argon (O 2 2 30% gas mixture. Previous experiments conducted in deionized water under argon atmosphere, have shown a good correlation between alpha activity and uranium release for the 15-, 1500- and 40000-years alpha doped UO 2 batches. Besides, uranium release in the leachate is controlled either by the kinetics, or by the thermodynamics. Provided the solubility limit of uranium is not achieved, uranium concentration increases and is only limited by the kinetics, unless precipitation occurs and the uranium concentration remains constant over time. These controls are highly dependant on the solution chemistry (HCO 3 - , pH, Eh,..), the atmosphere (Ar, Ar/H 2 ,..), and the radiolysis strength. The experimental matrix

  6. Thermal expansion of UO2 and simulated DUPIC fuel

    International Nuclear Information System (INIS)

    Ho Kang, Kweon; Jin Ryu, Ho; Chan Song, Kee; Seung Yang, Myung

    2002-01-01

    The lattice parameters of simulated DUPIC fuel and UO 2 were measured from room temperature to 1273 K using neutron diffraction to investigate the thermal expansion and density variation with temperature. The lattice parameter of simulated DUPIC fuel is lower than that of UO 2 , and the linear thermal expansion of simulated DUPIC fuel is higher than that of UO 2 . For the temperature range from 298 to 1273 K, the average linear thermal expansion coefficients for UO 2 and simulated DUPIC fuel are 10.471x10 -6 and 10.751x10 -6 K -1 , respectively

  7. Measurements of the viscosity of sodium tetraborate (borax)-UO2 and of sodium metaborate-UO2 liquid solutions

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.; Roth, A.

    1983-01-01

    Adding UO 2 produces an increase of viscosity of borax and sodium metaborate. For temperatures below 920 0 C the measurements with the borax-UO 2 solution show a phase separation. Contrary to borax the sodium metaborate solutions indicate a well defined melting point. At temperatures slightly below the melting point a solid phase is formed. The tested sodium-borates-UO 2 mixtures are in liquid form. (DG)

  8. Dissolution of unirradiated UO2-pellets in nitric acid

    International Nuclear Information System (INIS)

    Herrmann, B.

    1984-02-01

    Cinetics of dissolution of UO 2 -pellets in nitric acid and the gaseous reaction products, N 2 O, NO, NO 2 are determined for different temperatures and acid concentrations. NO 2 :NO ratio increases with temperature and nitrate concentration. The amount of N 2 O formed increases with temperature and acid concentration. At 90 0 C and dissolution in 12 m nitric acid 1l weight-% of UO 2 are dissolved forming N 2 O. The oxidation of UO 2 takes place on the crystal surface or at the interface UO 2 /HNO 3 . U(IV)-ions cannot be detected in the solution. The nitrous acid resulting from reduction of HNO 3 or the species which is in equilibrium with nitrous acid e.g. the nitrosyl-ion is responsible for UO 2 -oxidation. (orig./PW) [de

  9. Development of ceramics based fuel, Phase I, Kinetics of UO2 sintering by vibration compacting of UO2 powder (Introductory report)

    International Nuclear Information System (INIS)

    Ristic, M.M.

    1962-10-01

    After completing the Phase I of the task related to development of ceramics nuclear fuel the following reports are presented: Kinetics of UO 2 sintering; Vibrational compacting and sintering of UO 2 ; Characterisation of of UO 2 powder by DDK and TGA methods; Separation of UO 2 powder

  10. Pressure induced by Na entrapped in molten UO2

    International Nuclear Information System (INIS)

    Clerici, G.; Schins, H.

    1978-01-01

    In a first approach the constraint supplied by the solidifying UO 2 -shell is evaluated. The mass of the injected sodium is assumed to have a spherical form. Its dimensions are negligible in respect to the extension of the UO 2 . Using the Von Karman Pohlhausen method for solving the Fourier equation, the temperature distributions in UO 2 and sodium are determined. The physical properties are taken to be independent of temperature. Once these temperature profiles are obtained, the pressure induced into the heated sodium by the hypothesised mechanical constraint of the rigid shell and the tangential stress produced in this shell, can be calculated. In a second approach then, a liquid-liquid contact between UO 2 and Na is considered. The interface temperature, however, is calculated by means of an adjusted initial temperature of UO 2 . Following an idea of Cho and Wright, to the actual temperature of UO 2 is added a value obtained by dividing its latent heat of fusion by its heat capacity. The thermal expansion of the sodium drop is initially delayed by the inertial constraint of the surrounding heavy UO 2 . The expansion of the liquid drop of sodium continues up to the moment where the average temperature of the entrapped sodium becomes equal to the homogeneous nucleation temperature. At this instant vaporisation starts and the process goes on described by the formation of a two-phase mixture for the sodium. In this way the interaction of an entrapped sodium drop is calculated as a superheat limited explosion

  11. Fabrication of nano-structured UO2 fuel pellets

    International Nuclear Information System (INIS)

    Yang, Jae Ho; Kang, Ki Won; Rhee, Young Woo; Kim, Dong Joo; Kim, Jong Heon; Kim, Keon Sik; Song, Kun Woo

    2007-01-01

    Nano-structured materials have received much attention for their possibility for various functional materials. Ceramics with a nano-structured grain have some special properties such as super plasticity and a low sintering temperature. To reduce the fuel cycle costs and the total mass of spent LWR fuels, it is necessary to extend the fuel discharged burn-up. In order to increase the fuel burn-up, it is important to understand the fuel property of a highly irradiated fuel pellet. Especially, research has focused on the formation of a porous and small grained microstructure in the rim area of the fuel, called High Burn-up Structure (HBS). The average grain size of HBS is about 300nm. This paper deals with the feasibility study on the fabrication of nano-structured UO 2 pellets. The nano sized UO 2 particles are prepared by a combined process of a oxidation-reducing and a mechanical milling of UO 2 powder. Nano-structured UO 2 pellets (∼300nm) with a density of ∼93%TD can be obtained by sintering nano-sized UO 2 compacts. The SEM study reveals that the microstructure of the fabricated nano-structure UO 2 pellet is similar to that of HBS. Therefore, this bulk nano-structured UO 2 pellet can be used as a reference pellet for a measurement of the physical properties of HBS

  12. Ceramic UO2 powder production at Cameco Corporation

    International Nuclear Information System (INIS)

    Mulligan, J.J.

    2005-01-01

    This paper describes the various aspects of ceramic grade UO 2 powder production at Cameco Corporation's Port Hope conversion facility. It discusses the significant safety systems, production processes and plant monitoring and control systems. It also provides an insight into how various support groups such as Quality Assurance, Analytical Services, and Technology Development contribute to the consistent production of high quality UO 2 powder. The ability of Cameco to identify, measure and control the physical and chemical properties of ceramic grade UO 2 has resulted in the production of uniform quality powder that has consistently met customer requirements. (author)

  13. Utilization of dilatometer for characterization of UO2 powders

    International Nuclear Information System (INIS)

    Bhattacharjee, S.; Broczkowski, M.E.; Kryst, K.; Ioffe, M.S.; Murchie, M.P.

    2010-01-01

    Particle size and specific surface area measurements are conventional characterization techniques for ceramic powders to predict sintering behaviour. Dilatometry is one of the most popular techniques in ceramics to study the sintering characteristics for optimization of firing processes. Uranium dioxide (UO 2 ) powders from different sources were characterized by conventional techniques and these characteristics were correlated with sintering behaviour, as measured with a dilatometer. The suitability of dilatometry to correlate the sintering characteristics of UO 2 powder with variations in the O/U ratio, particle size and green density was discussed. Explanations of features in the sintering profiles for UO 2 powders have been presented. (author)

  14. Thermal properties of UO2 - Gd2O3 fuel

    International Nuclear Information System (INIS)

    Kim, G. S.; Yang, J. H.; Kang, K. W.; Kim, Y. M.; Song, G. W.

    2000-01-01

    The thermal properties (thermal conductivity, oxygen potential and thermal expansion) of UO 2 -Gd 2 O 3 fuels were measured by the laser-flash, TGA and dilatometry method. The thermal conductivity decreased with Gd content, but the oxygen potential and thermal expansion increased with Gd content. Substitution of Gd +3 ion in UO 2 structure increases the scattering site for thermal phonon propagation and thereby decreases the thermal conductivity. The oxygen potential of Gd-doped UO 2 increase mainly because the Gd +3 ions, which are inert to oxidation, make it difficult for oxygen interstitials to access just near them

  15. Determination of UO2F2, UO2 and UF4 in tetrafluoride of uranium samples

    International Nuclear Information System (INIS)

    Contreras Guzman, Ariel; Arlegui Hormazabal, Oscar

    2003-01-01

    The combustible elements for investigation reactors that at the present are manufacturing by the Chilean Nuclear Energy Commission (CCHEN) they are based on aluminum and silicide uranium powdered which is obtained from metallic uranium. At the present the Conversion Units, is developing the technology of transformation UF 6 in metallic Uranium, reason for which is necessary that the Chemical Analysis Laboratory have a methodology that allows to quantify the presence of UO 2 F 2 , UO 2 and UF 4 in the samples obtained in this transformation process. For this reason we are implements the methodology of sequential analysis that had been developed previously, for the Institute of Energy and Nuclear Investigations, IPEN Brasil, and to adapt it to the present conditions in the Laboratory of Chemical Analysis of the CCHEN. This method is based on the different solubilities that present those sample in front of solvents as ethanol and solutions of ammonium oxalate, what allows the separation of these compounds for a later analysis by means of the method of Davies and Gray. This method is based on the reduction of the uranium (VI) to uranium (IV) with ferrous ion amid phosphoric acid, quantifying the present uranium in the samples by means of titration with potassium dicromate. With the purpose of checking the efficiency of the method, the sum of all values of uranium coming from each compound and compares it with the total uranium of the sample (author)

  16. UO2/magnetite concrete interaction and penetration study

    International Nuclear Information System (INIS)

    Farhadieh, R.; Purviance, R.; Carlson, N.

    1983-01-01

    The concrete structure represents a line of defense in safety assessment of containment integrity and possible minimization of radiological releases following a reactor accident. The penetration study of hot UO 2 particles into limestone concrete and basalt concrete highlighted some major differences between the two concretes. These included penetration rate, melting and dissolution phenomena, released gases, pressurization of the UO 2 chamber, and characteristics of post-test concrete. The present study focuses on the phenomena associated with core debris interaction with and penetration into magnetite type concrete. The real material experiment was carried out with UO 2 particles and magnetite concrete in a test apparatus similar to the one utilized in the UO 2 /limestone experiment

  17. Densification Behavior of BN-added UO2

    International Nuclear Information System (INIS)

    Rhee, Young Woo; Kim, Keonsik; Kim, Dong Joo; Kim, Jong Hun; Oh, Jang Soo; Yang, Jae Ho

    2013-01-01

    Local wall thinning in pipelines affects the structural integrity of industries like nuclear power plants (NPPs). In the present study a pulsed eddy current (PEC) technology to detect the wall thing of carbon steel pipe covered with insulation is developed. Boron is commercially used as a neutron absorber fuel. A neutron absorber fuel is burned out or depleted during reactor operation. Westinghouse have been produced the Integral Fuel Burnable Absorber (IFBA) which is enriched UO 2 fuel pellets with a thin coating of zirconium diboride (ZrB 2 ) on the outer surface. Standard sintered fuel pellets are sputter coated with ZrB 2 . It is known that IFBA fuel can incur 20% to 30% additional fabrication costs. Boron-dispersed UO 2 fuel pellet made by the conventional pressing and sintering process of a powder mixture of UO 2 and B compound might be more cost-effective than IFBAs. M. G. Andrew et al. tried to sinter boron-dispersed UO 2 green pellet. However, they reported that boron-dispersed UO 2 fuel pellet is very difficult to be fabricated with a sufficient level of boron retention and high sintered density (greater than 90 % of theoretical density) because of the volatilization of boron oxide. We have investigated the densification behavior of mixtures of UO 2 and various boron compounds, such as B 4 C, BN, TiB 2 , ZrB 2 , SiB 6 , and HfB 2 . Boron compounds seemed to act as a sintering additive for UO 2 at a certain low temperature range. In this study, the densification behavior of BN-added UO 2 pellet has been investigated by sintering green pellets of a mixture of UO 2 powder and BN powder in H 2 atmosphere. A high density BN-added UO 2 pellet can be fabricated after sintering at 1200 .deg. C for more than 1 h in a H 2 atmosphere. The sintered density of BN-added UO 2 pellet can be increased up to about 95 %TD

  18. Comparison of neutronic behavior of UO2, (Th-233UO2 and (Th-235UO2 fuels in a typical heavy water reactor

    Directory of Open Access Journals (Sweden)

    Seyed Mohammad Mirvakili

    2015-04-01

    The obtained computational data showed both thorium-based fuels caused less negative temperature reactivity coefficients for the modeled research reactor in comparison with UO2 fuel loading. By contrast, 233U-containing thorium-based fuel and 235U-containing thorium-based fuel loadings in the thermal core did not drastically reduce the effective delayed neutron fractions and delayed neutron fractions compared to UO2 fuel. A provided higher conversion factor and lower transuranic production in the research core fed by the thorium-based fuels make the fuel favorable in achieving higher cycle length and less dangerous and costly nuclear disposals.

  19. Overview of interatomic potentials

    International Nuclear Information System (INIS)

    Bonny, G.; Malerba, L.

    2005-12-01

    In this report an overview on interatomic potentials is given. This overview is by no means complete and it has merely the intention to give the reader an idea of where interatomic potentials come from, as well as to provide the basic ideas behind some commonly used methods for deriving interatomic potentials for molecular dynamics applications. We start by giving a short introduction about the concept of interatomic potential in the framework of quantum mechanics, followed by a short description of commonly used methods for deriving semi-empirical interatomic potentials. After some short theoretical notions on each method, some practical parameterizations of commonly used potentials are given, including very recent ones. An effort has been made to classify existing approaches within a rational and consequent scheme, which is believed to be of use for a thorough comprehension of the topic. Although these approaches can be used in a variety of different materials, we will only discuss the practical cases of metals. Following this, some widespread ad hoc modification of the general methods are discussed. The report is concluded by a generalization of the methods to multi-component materials, in particular metallic alloys. (author)

  20. Interatomic inelastic current

    DEFF Research Database (Denmark)

    Hansen, Tim; Solomon, Gemma C.; Hansen, Thorsten

    2017-01-01

    is evaluated between atoms that are coupled by the electron-phonon interaction, the 1st and 2nd order terms must be added together to form a meaningful transmission. Within the limited scope of the models considered here, the 1st order term appears to be the signature of the inelastic event.......In order to identify the location of an inelastic event and to distinguish between situations that are before or after this event, we derive equations for the interatomic inelastic transmission as a perturbation series in the electron-phonon interaction. This series contains both even and odd...... ordered corrections, and while the even ordered corrections can be thought as a Dyson’s expansion of the interatomic elastic transmission in the electron-phonon self-energy, the odd ordered corrections represent something new. We explicitly derive expressions for the interatomic inelastic transmission up...

  1. UO2 fuel pellet characterization: density and porosity measurement methods

    International Nuclear Information System (INIS)

    Kopuz, B.; Bayram, Y.; Colak, L. and others

    1997-01-01

    The most commonly used fuel in nuclear power plants is UO 2 . UO 2 is a ceramic material and is produced by powder metallurgy techniques. The densities of the material produced can never reach the theoretical densities because of the production technology. The porosity allows the gas fission products, generated under power plant working conditions, to escape and therefor is required. Direct measurement of density which is an application of the Archimedes principle, is based on replacement of liquids. Replacement fluid is m-xylene. Density measurement are made by weighing the dry pellets in air, then weighing the m-xylene impregnated pellets in air and m-xylene impregnated pellets in air and m-xylene. UO 2 pellets densities, total porosities and open porosities can be calculated from the collection data

  2. Onset conditions for flash sintering of UO2

    Science.gov (United States)

    Raftery, Alicia M.; Pereira da Silva, João Gustavo; Byler, Darrin D.; Andersson, David A.; Uberuaga, Blas P.; Stanek, Christopher R.; McClellan, Kenneth J.

    2017-09-01

    In this work, flash sintering was demonstrated on stoichiometric and non-stoichiometric uranium dioxide pellets at temperatures ranging from room temperature (26 °C) up to 600 °C . The onset conditions for flash sintering were determined for three stoichiometries (UO2.00, UO2.08, and UO2.16) and analyzed against an established thermal runaway model. The presence of excess oxygen was found to enhance the flash sintering onset behavior of uranium dioxide, lowering the field required to flash and shortening the time required for a flash to occur. The results from this study highlight the effect of stoichiometry on the flash sintering behavior of uranium dioxide and will serve as the foundation for future studies on this material.

  3. Heat transfer coefficient between UO2 and Zircaloy-2

    International Nuclear Information System (INIS)

    Ross, A.M.; Stoute, R.L.

    1962-06-01

    This paper provides some experimental values of the heat-transfer coefficient between UO 2 and Zircaloy-2 surfaces in contact under conditions of interfacial pressure, temperature, surface roughness and interface atmosphere, that are relevant to UO 2 /Zircaloy-2 fuel elements operating in pressurized-water power reactors. Coefficients were obtained from eight UO 2 / Zircaloy-2 pairs in atmospheres of helium, argon, krypton or xenon, at atmosphere pressure and in vacuum. Interfacial pressures were varied from 50 to 550 kgf/cm 2 while surface roughness heights were in the range 0.2 x 10 -4 to 3.5 x 10 -4 cm. The effect on the coefficients of cycling the interfacial pressure, of interface gas pressure and of temperature were examined. The experimental values of the coefficients were used to test the predictions of expressions for the heat-transfer between two solids in contact. For the particular UO 2 / Zircaloy-2 pairs examined, numerical values were assigned to several parameters that related the surface roughnesses to either the radius of solid/solid contact spots or to the mean thickness of the interface voids and that accounted for the imperfect accommodation of the void gas on the test surfaces. (author)

  4. Electrochemical Reduction of solid UO2 in Molten Fluoride Salts

    International Nuclear Information System (INIS)

    Gibilaro, Mathieu; Cassayre, Laurent; Massot, Laurent; Chamelot, Pierre; Malmbeck, Rikard; Dugne, Olivier; Allegri, Patrick

    2010-01-01

    The direct electrochemical reduction of UO 2 solid pellets was carried out in LiF-CaF 2 (+ 2wt % Li 2 O) at 850 deg. C. An inert gold anode was used instead of the usual reactive sacrificial carbon anode. In this case, reduction of oxide ions yields O 2 gas evolution on the anode. Electrochemical characterisations of UO 2 pellets have been performed by linear sweep voltammetry at 10 mV/s and reduction waves associated to its direct reduction have been observed at a potential 150 mV more positive in comparison with the solvent reduction. Then, galvano-static electrolyses runs have been realised and products were characterised by SEM-EDX, EPMA/WDS and XRD. In one of the runs, uranium oxide was partially reduced and three phases were observed: non reduced 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. (authors)

  5. Characterization of Compaction Process on UO2 Powder Pelletisation

    International Nuclear Information System (INIS)

    Rachmawati, M; Langenati, R; Saputra, T.T; Mahpudin, A; Histori; Sutarya, D; Zahedi

    1998-01-01

    Determination of compaction pressure of pelletization which is based on density characterization in conjunction with satisfactory green strength of the UO 2 pellet, is carried out in this experiment. Cameco UO 2 powder has been mixed up with Zn-stearate lubricant prior to compaction process. The compaction pressure is varied from the range of 2 Mp up to 6 Mp. The mechanical strength is determined using diametral compression strength with the speed of loading of 0.1 mm.min 1 . The density measurement and compression strength test are performed on each of the applied pressure. The result shows that compaction at 5 Mp gives the maximum green strength of UO 2 pellet, while the maximum density is achieved at 5.7 Mp. The maximum green strength and green density of UO 2 (+ TiO 2 ) pellets is achieved at the addition of 0.25% and 0.125% TiO 2 respectively. The compaction pressure which is showing the maximum pellet green strength but still having the required density, is chosen to be the determinant compaction pressure in condition of pelletization

  6. Solid state reactions of monovalent sulphates with UO2

    International Nuclear Information System (INIS)

    Khandakar, R.R.; Krishnan, K.; Singh Mudher, K.D.; Jayadevan, N.C.

    1986-01-01

    Solid state reactions of sulphates of Na + , K + , Rb + , Cs + and Tl + with UO 2 in presence of (NH 4 ) 2 SO 4 lead to the formation of double sulphates at 400degC. The double sulphates decompose at higher temperatures to give metal uranates. Thermogravimetric, x-ray diffraction and chemical analysis have been used to characterise the compounds. (author). 5 refs

  7. Method for fluoride ion depletion of UO2 powders

    International Nuclear Information System (INIS)

    Beutner, R.; Ploeger, F.

    1978-01-01

    The method described consists in removing the hydrogen still present from the reduction during the preparation of UO 2 as completely as possible and in performing a pyrohydrolysis at temperatures above 650 0 C for at least 45 minutes. The removal of fluorine is necessary in order to avoid cladding tube damaging. (UA) [de

  8. High-precision molecular dynamics simulation of UO2–PuO2: Anion self-diffusion in UO2

    International Nuclear Information System (INIS)

    Potashnikov, S.I.; Boyarchenkov, A.S.; Nekrasov, K.A.; Kupryazhkin, A.Ya.

    2013-01-01

    Highlights: ► We perform MD simulation of oxygen diffusion in UO2 (up to 50 000 ions and 1 μs time). ► We reached 1400 K and 10 −12 cm 2 /sec, which allowed direct comparison to experiments. ► S-shaped T-dependence of activation energy and λ-peak of its derivative were obtained. ► Continual superionic phase transition (rather than first or second order) was proved. ► Activation energy of exchange diffusion equals anti-Frenkel defect formation energy. -- Abstract: Our series of articles is devoted to high-precision molecular dynamics simulation of mixed actinide-oxide (MOX) fuel in the approximation of rigid ions and pair interactions (RIPI) using high-performance graphics processors (GPU). In this article we study self-diffusion mechanisms of oxygen anions in uranium dioxide (UO 2 ) with the 10 recent and widely used sets of interatomic pair potentials (SPP) under periodic (PBC) and isolated (IBC) boundary conditions. Wide range of measured diffusion coefficients (from 10 −3 cm 2 /s at melting point down to 10 −12 cm 2 /s at 1400 K) made possible a direct comparison (without extrapolation) of the simulation results with the experimental data, which have been known only at low temperatures (T < 1500 K). A highly detailed (with the temperature step of 1 K) calculation of the diffusion coefficient allowed us to plot temperature dependences of the diffusion activation energy and its derivative, both of which show a wide (∼1000 K) superionic transition region confirming the broad λ-peaks of heat capacity obtained by us earlier. It is shown that regardless of SPP the anion self-diffusion in model crystals without surface or artificially embedded defects goes on via exchange mechanism, rather than interstitial or vacancy mechanisms suggested by the previous works. The activation energy of exchange diffusion turned out to coincide with the anti-Frenkel defect formation energy calculated by the lattice statics

  9. The Effect of Ion Irradiation on the Dissolution of UO2 and UO2-based Simulant Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Popel, Aleksej; Wietsma, Thomas W.; Engelhard, Mark H.; Lea, A Scott; Qafoku, Odeta; Grygiel, Calara; Monnet, I.; Ilton, Eugene S.; Bowden, Mark E.; Farnan, Ian E.

    2018-02-25

    The aim of this work was to study the effect of fission fragment damage on the dissolution of UO2 fuel matrix in water. Plain and doped (43 GWd/tU simulated burn-up) UO2 samples were produced and irradiated by 92 MeV 129Xe23+ ions to a fluence of 4.8 × 1015 ions/cm2 to simulate the fission damage that occurs within nuclear fuels. Dissolution experiments were conducted in single-pass flow-through mode under anoxic conditions (< 0.1 O2 ppm in Ar) to study the effect of the induced irradiation damage on the dissolution of the UO2 matrix. The ion irradiation caused smoothening of the surface features and formation of hollow blisters. Local overheating during the ion irradiation was suggested to cause these features. The dissolution studies showed that the irradiated samples generally showed a higher initial release of uranium than unirradiated ones and then the uranium concentrations converged towards a value of the same order (~10-9 mol/l) with time. No noticeable difference in dissolution behaviour was observed between the plain and doped samples in this work. Secondary phase formations were observed on the surface of UO2 samples after the dissolution experiment.

  10. Densification behaviour of UO2 in six different atmospheres

    International Nuclear Information System (INIS)

    Kutty, T.R.G.; Hegde, P.V.; Khan, K.B.; Basak, U.; Pillai, S.N.; Sengupta, A.K.; Jain, G.C.; Majumdar, S.; Kamath, H.S.; Purushotham, D.S.C.

    2002-01-01

    The shrinkage behaviour of UO 2 has been studied using a dilatometer in various atmospheres of Ar, Ar-8%H 2 , vacuum, CO 2 , commercial N 2 and N 2 +1000 ppm of O 2 . The onset of shrinkage occurs at around 300-400 deg. C lower in oxidizing atmospheres such as CO 2 , commercial N 2 and N 2 +1000 ppm O 2 compared to that in reducing or inert atmospheres. Shrinkage behaviour of UO 2 is almost identical in Ar, Ar-8%H 2 and vacuum. The shrinkage in N 2 +1000 ppm O 2 begins at a lower temperature than that in the commercial N 2 . The mechanism of sintering in the reducing, inert and vacuum atmospheres is explained by diffusion of uranium vacancies and that in the oxidizing atmospheres by cluster formation

  11. Studies on sintering kinetics of ThO2-UO2 pellets using master sintering curve approach

    Science.gov (United States)

    Banerjee, Joydipta; Ray, Aditi; Kumar, Arun; Banerjee, Srikumar

    2013-11-01

    Three different compositions of thoria-urania pellets, namely, ThO2-4%UO2, ThO2-10%UO2 and ThO2-20%UO2 (all compositions are in wt% containing natural uranium) were fabricated by Coated Agglomerate Pelletization (CAP) process. The compositions studied in the current paper are the proposed fuels for the forthcoming Indian Advanced Heavy water Reactor (AHWR) and its variant based on low enriched uranium. Sintering kinetics of ThO2-x%UO2 (x = 4, 10, 20) green pellets, thus fabricated, were evaluated using constant heating rate experiments in a vertical dilatometer. Activation energies of sintering (Q) were estimated using Arrhenius plot as proposed by Wang and Raj. Master Sintering Curves (MSC) for the above three compositions were constructed using shrinkage data. A FORTRAN program, employing optimization based numerical algorithm for fitting relative density vs. work of sintering data with sigmoid function, was used for this purpose. The apparent activation energies, evaluated using MSC principle, appear to be consistent with the values obtained by Arrhenius plot.

  12. Handbook of interatomic potentials

    International Nuclear Information System (INIS)

    Stoneham, A.M.; Taylor, R.

    1981-08-01

    This Handbook collects together interatomic potentials for a large number of metals. Most of the potentials describe the interactions of host metal atoms with each other, and these, in some cases, may be applied to solid and liquid metals. In addition, there are potentials (a) for a metallic impurity alloyed with the host, (b) for a small number of chemical impurities in the metal (eg H, O), and (c) for rare-gas impurities, notably He. The Handbook is intended to be a convenient source of potentials for bulk, surface and defect calculations, both static and dynamic. (author)

  13. Technological aspects of UO2 sintering at low temperature

    International Nuclear Information System (INIS)

    Thern, Gerardo G.; Dominguez, Carlos A.; Benitez, Ana M.; Marajofsky, Adolfo

    1999-01-01

    Within the Fuel Cycle Program of CNEA, the knowledge that plant personnel has on sintering at low temperature was evaluated, because this process could decrease costs for UO 2 and (U,Gd)O 2 pellets production, simplify the furnace maintenance and facilitate the automation of the production process, specially convenient for uranium recovery. By applying this technology, some companies have achieved production at pilot-scale and irradiated a significant number of pellets. (author)

  14. Oxidation of UO2 at 150 to 3500C

    International Nuclear Information System (INIS)

    Gilbert, E.R.; White, G.D.; Knox, C.A.

    1985-02-01

    Tests were performed on nonirradiated UO 2 pellets from 150 to 350 0 C in atmospheric air and controlled environments and on spent light-water reactor (LWR) fuel fragments at 200 and 230 0 C in atmospheric air to determine the variables that affect oxidation behavior under dry storage conditions. The weight of spent fragments increased 50 to 100 times faster than the weight of nonirradiated UO 2 pellets at 230 0 C. Non-irradiated pellet fragments gained weight 5 to 7 times faster than nonirradiated pellets. The fragments simulated fuel fragmented by thermal gradients during reactor power changes. Low-density powder (U 3 O 8 ) formed at 0.05 and 0.3% weight gain for nonirradiated pellets and fragments, respectively, but had not formed at 3% weight gain for spent fuel fragments with a burnup of 29,000 MWd/MTU. Canadian investigators had found that powder formed at intermediate levels of weight gain in CANDU spent fuel fragments with an approximate burnup of 8000 MWd/MTU. The combined effects of the high rate of weight gain in spent fuel and the burnup dependence of weight gain at powder formation resulted in a minimum in a plot of the time for the onset of powder formation versus burnup. The minimum in powder induction time occurs at or below burnup levels typical of CANDU spent fuel and spent fuel at the ends of some LWR rods. The results are described in terms of thermal and neutron irradiation-induced changes in UO 2 pellet structure and chemical composition. Other tests were performed at up to 275 0 C with spent fuel fragments and nonirradiated UO 2 pellets in moist nitrogen to determine the suitability of nitrogen as a cover gas. No measurable weight gain or visible physical changes occurred during the first 2 months of testing. 22 figures, 7 tables

  15. Binding energy and formation heat of UO2

    International Nuclear Information System (INIS)

    Almeida, M.R. de; Veado, J.T.; Siqueira, M.L. de

    The Born-Haber cycle is utilized for the calculation of the heat of formation of UO 2 , on the assumption that the binding energy is predominantly ionic in character. The ionization potentials of U and the repulsion energy are two critical values that influence calculations. Calculations of the ionization potentials with non-relativistic Hartree-Fock-Gaspar-Kohn-Sham approximation are presented [pt

  16. Automation system for production of UO2 granules

    International Nuclear Information System (INIS)

    Swaminathan, N.; Setty, C.R.P.; Banerjee, P.K.; Husnain, G.; Rao, K.C.M.; Satyanarayana, A.

    1990-01-01

    Precompaction of UO 2 powder into slugs and granulation of the slugs were used to be carried out in two different work centres involving manual loading/handling of powder and compacts which resulted in a very high level of air-borne activity. This has been simplified by integrating both the operations into one work centre on both the precompaction presses. In the present system, UO 2 powder is transferred to feed hopper through the use of high vac. feeder. The powder in metered quantities is fed into the shoe by deploying screw feeder driven by a compact hydraulic motor. The die cavity is filled with just the right quantity of powder to prevent spillage. The compacts are pushed on to the granulator through a set of guides mounted on the die platform. The granulated powder is made to pass through Vibro screen for separating the fines before collecting in a replaceable S.S. Container. This container is mounted on the final compacting press by using job crane installed on the press. The replaceable container handling facility drastically cuts down the manual handling of UO 2 granules and also eliminates spillage, air borne activity. The development and fabrication of hydraulically operated screw feeder, feed shoe, replaceable container and the job crane structure etc., were completely carried out at Nuclear Fuel Complex, Hyderabad. Paper deals in detail the design of the system developed, present operational experiences and further improvements planned. (author). 6 figs

  17. Fabrication of metallic channel-containing UO2 fuels

    International Nuclear Information System (INIS)

    Yang, Jae Ho; Song, Kun Woo; Kim, Keon Sik; Jung, Youn Ho

    2004-01-01

    The uranium dioxide is widely used as a fuel material in the nuclear industry, owing to many advantages. But it has a disadvantage of having the lowest thermal conductivity of all kinds of nuclear fuels; metal, carbide, nitride. It is well known that the thermal conductivity of UO 2 fuel is enhanced by making, so called, the CERMET (ceramic-metal) composite which consists of both continuous body of highly thermal-conducting metal and UO 2 islands. The CERMET fuel fabrication technique needs metal phase of at least 30%, mostly more than 50%, of the volume of the pellet in order to keep the metal phase interconnected. This high volume fraction of metal requires such a high enrichment of U that the parasitic effect of metal should be compensated. Therefore, it is attractive to develop an innovative composite fuel that can form continuous metal phase with a small amount of metal. In this investigation, a feasibility study was made on how to make such an innovative fuel. Candidate metals (W, Mo, Cr) were selected, and fabrication process was conceptually designed from thermodynamic calculations. We have experimentally found that a metal phase envelops perfectly UO 2 grains, forming continuous channel throughout the pellet, and improving the thermal conductivity of pellet

  18. Radiolysis and corrosion of Pu-doped UO2 pellets in chloride brine

    Indian Academy of Sciences (India)

    Unknown

    processes of aged spent fuel. 238Pu doped UO2 pellets were frequently used in the past9–11 for experiments in dilute or aerated solutions. This work describes corrosion experiments with 238Pu doped UO2 pellets in deaerated 5 M NaCl solution and compares them with experiments using undoped UO2 pellets in ...

  19. Advanced doped UO2 pellets in LWR applications

    International Nuclear Information System (INIS)

    Arborelius, Jakob; Backman, Karin; Hallstadius, Lars

    2006-01-01

    The nuclear industry strives to reduce the fuel cycle cost, enhance flexibility and improve the reliability of operation. This can be done by both increasing the fuel weight and optimizing rod internal properties that affect operational margins. Further, there is focus on reducing the consequences of fuel failures. To meet these demands Westinghouse has developed ADOPT (Advanced Doped Pellet Technology) UO 2 fuel containing additions of chromium and aluminium oxides. This paper presents results from the extensive investigation program which covered examinations of doped and reference standard pellets both in the manufactured and irradiated states. The additives facilitate pellet densification during sintering and enlarge the pellet grain size. The final manufactured doped pellets reach about 0.5% higher density within a shorter sintering time and a five fold larger grain size compared with standard UO 2 fuel pellets. The physical properties of the pellets, including heat capacity, thermal expansion coefficient, melting temperature, thermal diffusivity, have been investigated and differences between the doped and standard UO 2 pellets are small. The in-reactor performance of the ADOPT pellets has been investigated in pool-side and hotcell Post Irradiation Examinations (PIEs), as well as in the Studsvik R2 test reactor. The investigations have identified three areas of improved operational behaviour: Reduced fission gas release, improved PCI performance thanks to increased pellet plasticity and higher resistance against post-failure degradation. Fuel segments have been exposed to ramp tests and enhanced power steady-state operation in the Studsvik R2 reactor after base-irradiation to above 30MWd/kgU in a commercial BWR. ADOPT reveals up to 50% lower fission gas release than standard UO 2 pellets. The fuel degradation behaviour has been studied in two oxidizing tests, a thermal-microbalance test and an erosion test under irradiation. The tests show that ADOPT pellets

  20. The treatment of large quantities of high fluorin contents UO2 by ammonium double uranate (ADU) techniques

    International Nuclear Information System (INIS)

    Wang Bangwu; Chen Ying

    2010-01-01

    The paper has discussed the sinter action of UO 2 in low temperature. The study indicates the over hot part of UO 2 by the deoxidization hot of oxidation uranate mostly results in the sinter in the process of trans form ADU into UO 2 . The UO 2 settling times in kiln little influences the sinter performance of UO 2 in the same condition of high fluorin contents UO 2 returning kiln, and high fluorin contents UO 2 returning kiln does not sinter UO 2 again. Experiment on large quantities of high fluorin contents UO 2 by Ammonium Double Uranate (ADU) techniques direct returning kiln, the result shows the sinter performance of UO 2 doesn't drop in the process of high fluor in contents UO 2 direct returning kiln, and the performance of UO 2 can meet specification. (authors)

  1. PERBANDINGAN DENSITAS PELET UO2 HASIL PELETISASI MENGGUNAKAN SERBUK DAN MIKROSPIR

    Directory of Open Access Journals (Sweden)

    Etty Mutiara

    2016-06-01

    Full Text Available ABSTRAK PERBANDINGAN DENSITAS PELET UO2 HASIL PELETISASI MENGGUNAKAN SERBUK DAN MIKROSPIR UO2. Telah dilakukan pengembangan proses peletisasi menggunakan mikrospir UO2sebagai pengganti serbuk UO2. Mikrospir bersifat speris, free flowing, porus dengan kekerasan tertentu (soft particle. Keunggulan penggunaan mikrospir pada proses peletisasi adalah tidak menimbulkan debu saat kompaksi dan lebih efektif dalam pengepakan sehingga tidak membutuhkan proses granulasi dan pelumas padat. Dihipotesakan bahwa penggunaan mikrospir UO2 dalam proses peletisasi akan memberikan densitas pelet sinter yang lebih tinggi dibandingkan dengan penggunaan serbuk UO2 pada parameter proses peletisasi yang sama. Mikrospir UO2yang digunakan pada peletisasi ini berukuran 900 µm dan crushing strength 2,0 N/partikel , sedangkan serbuk UO2 yang digunakan berukuran antara 150-850 µm. Proses peletisasi mikrospir UO2 dan serbuk UO2 dilakukan dengan memvariasikan tekanan kompaksi antara 200 Mpa hingga 500 MPa dan disinter pada temperatur 1100 °C selama 6 jam dalam suasana campuran gas hidrogen dan nitrogen. Karakterisasi dilakukan pada pelet mentah dan pelet sinter mikrospir UO2 dan serbuk UO2 yang meliputi pengukuran dimensi, penimbangan berat dan pengukuran densitas. Pada variasi tekanan kompaksi diperoleh pelet mentah dan pelet sinter mikrospir UO2 dengan densitas lebih tinggi dibandingkan hasil peletisasi serbuk UO2. Diperoleh hasil bahwa densitas pelet mentah baik hasil kompaksi serbuk UO2 maupun mikrospir UO2meningkat dengan bertambahnya tekanan kompaksi. Densitas pelet mentah mikrospir UO2berkisar antara 82,1 - 84,2 %TD. Pada kondisi penyinteran yang sama, baik kompakan serbuk UO2 maupun kompakan mikrospir UO2 memperlihatkan densitas meningkat dengan semakin besar tekanan proses kompaksi. Dari penelitian ini belum diperoleh pelet sinter UO2 dengan densitas sesuai persyaratan reaktor pengguna sehingga diperlukan penelitian lanjutan terkait parameter proses peletisasi dan

  2. Determination of UO2 little quantity in UF4 by X-rays diffraction

    International Nuclear Information System (INIS)

    Costa, M.I.; Sato, I.M.; Imakuma, K.

    1977-01-01

    In the fluorination process, the final product UF 4 contain different levels of UO 2 as a contaminant. A routine method for quantitative analysis by x-ray diffraction has been developed. Standard curves have been plotted using mixtures of UO 2 /UF 4 with measures of intensity of (III) peak of UO 2 by the step scanning process. The integrated intensity versus UO 2 concentration curves present a linear behavior in the range from 0 to 4%. A good reprodutibility of measuring process has been observed through statistical analysis which permits to determine low fractions of UO 2 in UF 4 with +- 0,08% of accuracy [pt

  3. Production and release of the fission gas in (Th U)O2 fuel rods

    International Nuclear Information System (INIS)

    Dias, Marcio S.

    1982-06-01

    The volume, composition and release of the fission gas products were caculated for (Th, U)O 2 fuel rods. The theorectical calculations were compared with experimental results available on the literature. In ThO 2 + 5% UO 2 fuel rods it will be produced approximated 5% more fission gas as compared to UO 2 fuel rods. The fission gas composition or Xe to Kr ratio has showed a decreasing fuel brunup dependence, in opposition to that of UO 2 . Under the same fuel rod operational conditions, the (Th, U)O 2 fission gas release will be smaller as compared to UO 2 . This behaviour of (Th, U)O 2 fuel comes from smallest gas atom difusivity and higher activation energies of the processes that increase the fission gas release. (Author) [pt

  4. 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.)

  5. Sintering of Kernel UO2 for High Temperature Reactor Fuel

    International Nuclear Information System (INIS)

    Sukarsono; Dwi-Heru-Sucahyo; Hidayati; Evi-Hertiviana; Bambang-Sugeng

    2000-01-01

    Sintering investigation of UO 2 gel has been done. The gel was preparedthrough two ways. The first, gel was produced using PVA as additive agent.The second gel was produced using HMTA and Urea as additive agent. From thepreparation of gel, the PVA method better than the urea - HMTA method,because was not necessary the cold temperature for sol preparation and alsowas not necessary the hot temperature for gelation process. After nextprocessing, the sintered gel of gel through PVA, also better than HMTAprocess. (author)

  6. Simulation of microstructure of rim region in UO2 fuel

    International Nuclear Information System (INIS)

    Oh, J. Y.; Koo, Y. H.; Lee, B. H.; Cheon, J. S.; Son, D. S.

    2004-01-01

    The rim region in the periphery of high burnup UO 2 pellet has a large number of pores and very small recrystallized grains. If the microstructure of the rim region is modeled more refinedly, it is possible to simulate the behavior of pores in rim region more accurately. In this paper, the microstructure of rim region was simulated through proper assumptions, and it was compared with the observed microstructure of the rim region. The validity of assumptions used in the simulation was verified qualitatively through this comparison

  7. Contribution to the study of UO2 pellet fabrication

    International Nuclear Information System (INIS)

    Fogaca Filho, N.; Gentile, E.F.; Mourao, M.B.; Souza Santos, T.D. de; Haydt, H.M.

    1977-01-01

    The establishment of a set of parametric comparisons related to UO 2 powders of two different origins as the ammonium diuranate and the ammonium uranyl carbonate is presented. It is emphasized the importance due to the pressing capability of the powders and the requirement for homogeneous microstructure for both, the pore distribution and the grain size. In order to establish the parameters of comparison, all the required normal tests for the in-process control of fabrication of fuel elements for nuclear power reactors were performed, particularly to the re-sintering test, in view of the evaluation of dimensional stability of the pellets [pt

  8. UO2 microspheres obtainment through the internal gelation methods

    International Nuclear Information System (INIS)

    Sterba, M.E.; Gomez Constenla, A.

    1987-01-01

    UO 2 microspheres obtainment process through the internal gelation method which allows the spheres' obtainment of uniform size is detailed herein, varying the same among 0.3 and 1.7 mm of diameter. The sintered density reaches 10.78 g/cm 3 , permitting the fuels fabrication dispersed and vibro-compacted fuels. The trichloroethylene use implementation as gelation agent is described, thus reducing the number of stages in the microspheres fabrication. At the same time, the uranium sun composition has been modified so as to be compatible with the use solvent. (Author)

  9. Acoustic emission during the compaction of brittle UO2 particles

    International Nuclear Information System (INIS)

    Hegron, Lise

    2014-01-01

    One of the options considered for recycling minor actinides is to incorporate about 10% to UO 2 matrix. The presence of open pores interconnected within this fuel should allow the evacuation of helium and fission gases to prevent swelling of the pellet and ultimately its interaction with the fuel clad surrounding it. Implementation of minor actinides requires working in shielded cell, reducing their retention and outlawing additions of organic products. The use of fragmentable particles of several hundred micrometers seems a good solution to control the microstructure of the green compacts and thus control the open porosity after sintering. The goal of this study is to monitor the compaction of brittle UO 2 particles by acoustic emission and to link the particle characteristics to the open porosity obtained after the compact sintering. The signals acquired during tensile strength tests on individual granules and compacts show that the acoustic emission allows the detection of the mechanism of fragmentation and enables identification of a characteristic waveform of this fragmentation. The influences of compaction stress, of the initial particle size distribution and of the internal cohesion of the granules, on the mechanical strength of the compact and on the microstructure and open porosity of the sintered pellets, are analyzed. By its ability to identify the range of fragmentation of the granules during compaction, acoustic emission appears as a promising technique for monitoring the compaction of brittle particles in the manufacture of a controlled porosity fuel. (author) [fr

  10. Modelling the high burnup UO2 structure in LWR fuel

    International Nuclear Information System (INIS)

    Lassmann, K.; Walker, C.T.; Laar, J. van de; Lindstroem, F.

    1995-01-01

    The concept of a burnup threshold for the formation of the high burnup UO 2 structure (HBS) is supported by experimental data, which also reveal that a transition zone exists between the normal UO 2 structure and the fully developed HBS. From the analysis of radial xenon profiles measured by EPMA a threshold burnup is obtained in the range 60-75 GW d/t U. The lower value is considered to be the threshold for the onset of the HBS and the higher value the threshold for the fully developed HBS. Xenon depletion in the transition zone and the fully developed HBS can be described by a simple model. At local burnups above 120 GW d/t U the xenon generated is in equilibrium with the xenon lost to the fission gas pores and the concentration does not fall below 0.25 wt%. The TRANSURANUS burnup model TUBRNP predicts reasonably well the penetration of the HBS and the associated xenon depletion up to a cross section average burnup of approximately 70 GW d/t U. (orig.)

  11. Irradiation effects in UO2 and CeO2

    International Nuclear Information System (INIS)

    Ye, Bei; Oaks, Aaron; Kirk, Mark; Yun, Di; Chen, Wei-Ying; Holtzman, Benjamin; Stubbins, James F.

    2013-01-01

    Single crystal CeO 2 , as a surrogate material to UO 2 , was irradiated with 500 keV xenon ions at 800 °C while being observed using in situ transmission electron microscopy (TEM). Experimental results show the formation and growth of defect clusters including dislocation loops and cavities as a function of increasing atomic displacement dose. At high dose, the dislocation loop structure evolves into an extended dislocation line structure, which appears to remain stable to the high dose levels examined in this study. A high concentration of cavities was also present in the microstructure. Despite high atomic displacement doses, the specimen remained crystalline to a cumulated dose of 5 × 10 15 ions/cm 2 , which is consistent with the known stability of the fluorite structure under high dose irradiation. Kinetic Monte Carlo calculations show that oxygen mobility is substantially higher in hypo-stoichiometric UO 2 /CeO 2 than hyper-stoichiometric systems. This result is consistent with the ability of irradiation damage to recover even at intermediate irradiation temperatures

  12. Multi-scale modeling of inter-granular fracture in UO2

    Energy Technology Data Exchange (ETDEWEB)

    Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tonks, Michael R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Biner, S. Bulent [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    A hierarchical multi-scale approach is pursued in this work to investigate the influence of porosity, pore and grain size on the intergranular brittle fracture in UO2. In this approach, molecular dynamics simulations are performed to obtain the fracture properties for different grain boundary types. A phase-field model is then utilized to perform intergranular fracture simulations of representative microstructures with different porosities, pore and grain sizes. In these simulations the grain boundary fracture properties obtained from molecular dynamics simulations are used. The responses from the phase-field fracture simulations are then fitted with a stress-based brittle fracture model usable at the engineering scale. This approach encapsulates three different length and time scales, and allows the development of microstructurally informed engineering scale model from properties evaluated at the atomistic scale.

  13. Sorption of Np by UO2 under repository conditions

    Science.gov (United States)

    Kazakovskaya, T. V.; Zakharova, E. V.; Haire, M. J.

    2010-03-01

    This work is a part of the joint Russian - American Program on Beneficial Use of Depleted Uranium. The production of nuclear fuels results in the accumulation of large quantities of depleted uranium (DU) in the form of uranium hexafluoride (UF6), which is converted to uranium oxides. Depleted uranium dioxide (DUO2) can be used as a component of radiation shielding and as an absorbent for migrating radionuclides that may emerge from casks containing spent nuclear fuel (SNF) that are stored for hundreds of thousands of years in high-level wastes (HLW) and SNF repositories (e.g. Yucca Mountain Project). In this case DU oxides serve as an additional engineered chemical barrier. It is known that the primary radioisotope contributor to the calculated long-term radiation dose to the public at the Yucca Mountain SNF repository site boundary is neptunium-237 (237Np). This paper describes the sorption of 237Np in various media (deionized water and J-13 solution) by DUO2. Samples of DUO2 used in this work originated from the treatment of UF6 in a reducing media to form UO2(DUO2-1 at 600°C, DUO2-2 at 700°C, and DUO2-3 at 800°C). All species of DUO2 sorb Np(V) and Np(IV) from aqueous media. Equilibrium was achieved in 24 hours for Np(V) and in 2 hours for Np(IV). Np(V) sorption is accompanied with partial reduction of Np(V) to Np(IV) and vice versa. The sorption of Np(V) onto DUO2 surfaces is irreversible. The investigations on DUO2 transformations were performed under dynamic and static conditions. Under static conditions the solubility of the DUO2 samples in J-13 solution is considerably higher than in DW. When the pre-treatment temperature is decreased, the solubility of DUO2 samples raises regardless of the media. The experiments on interaction between DUO2 and aqueous media (DW and J-13 solution) under dynamic conditions demonstrated that during 30-40 days the penetration/filtration rate of DW and J-13 solution through a thin DUO2 layer decreased dramatically, and then

  14. Cracking and relocation of UO2 fuel during nuclear operation

    International Nuclear Information System (INIS)

    Appelhans, A.D.; Dagbjartsson, S.J.

    1981-01-01

    Cracking and relocation of light water reactor (LWR) fuel pellets affect the axial gas flow path within nuclear reactor fuel rods and the thermal performance of the fuel. As part of the Nuclear Regulatory Commission's Water Reactor Safety Research Fuel Behavior Program, the Thermal Fuels Behavior Program of EG and G Idaho, Inc., is conducting fuel rod behavior studies in the Heavy Boiling Water Reactor in Halden, Norway. The Instrumental Fuel Assembly-430 (IFA-430) operated in that facility is a multipurpose assembly designed to provide information on fuel cracking and relocation, the long-term thermal response of LWR fuel rods subjected to various internal pressures and gas compositions, and the release of fission gases. This report presents the results of an analysis of fuel cracking and relocation phenomena as deduced from fuel rod axial gas flow and fuel temperature data from the first 6.5 GWd/tUO 2 burnup of the IFA-430

  15. Photochemical assessment of UO2+2 complexation in Triton X-100 micellar system

    International Nuclear Information System (INIS)

    Das, S.K.; Ganguly, B.N.

    1994-01-01

    This is a report on the spectral characteristics of UO 2 +2 in the excited state in the Triton X-100 micellar medium. The downward curving of the Stern-Volmer plot explains the two kinds of populations of UO 2 +2 upon micellization. A blue shift of the quenched emission is ascribed due to the collisional encounter of UO 2 +2 with the head groups of Triton X-100. (author). 5 refs., 2 figs

  16. Interim results from UO2 fuel oxidation tests in air

    International Nuclear Information System (INIS)

    Campbell, T.K.; Gilbert, E.R.; Thornhill, C.K.; White, G.D.; Piepel, G.F.; Griffin, C.W.j.

    1987-08-01

    An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to extend the characterization of spent fuel oxidation in air. To characterize oxidation behavior of irradiated UO 2 , fuel oxidation tests were performed on declad light-water reactor spent fuel and nonirradited UO 2 pellets in the temperature range of 135 to 250 0 C. These tests were designed to determine the important independent variables that might affect spent fuel oxidation behavior. The data from this program, when combined with the test results from other programs, will be used to develop recommended spent fuel dry-storage temperature limits in air. This report describes interim test results. The initial PNL investigations of nonirradiated and spent fuels identified the important testing variables as temperature, fuel burnup, radiolysis of the air, fuel microstructure, and moisture in the air. Based on these initial results, a more extensive statistically designed test matrix was developed to study the effects of temperature, burnup, and moisture on the oxidation behavior of spent fuel. Oxidation tests were initiated using both boiling-water reactor and pressurized-water reactor fuels from several different reactors with burnups from 8 to 34 GWd/MTU. A 10 5 R/h gamma field was applied to the test ovens to simulate dry storage cask conditions. Nonirradiated fuel was included as a control. This report describes experimental results from the initial tests on both the spent and nonirradiated fuels and results to date on the tests in a 10 5 R/h gamma field. 33 refs., 51 figs., 6 tabs

  17. Pressing-sintering process of UO2 pellet of controllable microstructure

    International Nuclear Information System (INIS)

    Wu Zhiming; Peng Qingshan

    1999-11-01

    The authors present the pressability and sinterability of two different matrix UO 2 powders from Ammonium Diuranate (ADU) process and the improvements by adding different quantities of additives. A focal point is made in the effect of the additives on the density, thermal stability and microstructure of UO 2 pellet. It is indicated by the results that a UO 2 pellet being of proper density, good thermal stability and better microstructure can be produced using of high sinterable UO 2 powder ex ADU, adding a certain quantity of pore former through calculation under conditions of high green density and high sintering temperature

  18. Preliminary study of determination of UO2 grain size using X-ray diffraction method

    International Nuclear Information System (INIS)

    Mulyana, T.; Sambodo, G. D.; Juanda, D.; Fatchatul, B.

    1998-01-01

    The determination of UO 2 grain size has accomplished using x-ray diffraction method. The UO 2 powder is obtained from sol-gel process. A copper target as radiation source in the x-ray diffractometer was used in this experiment with CμKα characteristic wavelength 1.54433 Angstrom. The result indicate that the UO 2 mean grain size on presintered (temperature 800 o C) has the value 456.8500 Angstrom and the UO 2 mean grain size on sintered (temperature 1700 o C) has value 651.4934 Angstrom

  19. Cation interdiffusion in the UO2 - (U, Pu)O2 and UO2 - PuO2 systems

    International Nuclear Information System (INIS)

    Leme, D.G.

    1985-01-01

    The interdiffusion of U and Pu ions in UO sub(2 +- x) - (U sub(0,83) Pu sub(0,17))O sub(2 + - x) and UO sub(2 + - x) -PuO sub(2 - x) sintered pellets and UO sub(2 +- x) -(U sub(0,82) Pu sub(0,18))O sub(2 + - x) single crystals has been studied as a function of the oxygen potential ΔG sup(-) (O 2 ) or the stoichiometric ratio O/M. The diffusion profiles of UO 2 /(U,Pu)O 2 and UO 2 /PuO 2 couples of different O/M ratios have been measured using high resolution α-spectrometer and microprobe. Thermal annealing of the specimens was performed in controlled atmospheres using either CO-CO 2 gas mixtures for constant O/M ratios or purified argon. The interdiffusion profiles have been analysed by means of the Boltzmann-Matano and Hall methods. The interdiffusion coefficient D sus(approx.) increases with increasing Pu content in sintered pellets (up to 17 wt. %PuO 2 ) showing a strong dependence of D sup(approx.) on the O/M ratio. The micropobe results show that the interdiffusion along grain boundaries is the main diffusion mechanism in the pellets. Experiments have also been carried out in single cristals to measure just the bulk-interdiffusion and avoiding effects due to grain boundaries. A marked dependence of D sup(approx.) on O/M ratio or on oxygen potential ΔG sup(-) (O 2 ), similar to the dependence already reported for self diffusion by means of radioactive tracers, has also been observed. (Author) [pt

  20. KARAKTERISASI LAPISAN PENYERAP DAPAT BAKAR PADA PERMUKAAN PELET UO2 + DOPAN TiO2

    Directory of Open Access Journals (Sweden)

    Sungkono Sungkono

    2017-01-01

    Full Text Available ABSTRAK KARAKTERISASI LAPISAN PENYERAP DAPAT BAKAR PADA PERMUKAAN PELET UO2 + DOPAN TiO2. Lapisan penyerap dapat bakar pada permukaan pelet UO2 + dopan TiO2 telah berhasil dibuat dengan menggunakan mertoda RF sputtering. Penelitian ini bertujuan untuk mendapatkan karakter mikrostruktur pelet UO2 + dopan, ketebalan, kekerasan mikro, komposisi kimia dan struktur kristal lapisan penyerap dapat bakar pada permukaan pelet UO2.  Penentuan mikrostruktur dan ketebalan lapisan dilakukan dengan menggunakan mikroskop optik, kekerasan lapisan dengan metode kekerasan mikro Vickers, komposisi kimia dengan spektrometri XRF dan struktur kristal dengan difraksi sinar-X. Hasil penelitian menunjukan bahwa semakin besar kandungan TiO2 dalam pelet maka semakin besar ukuran butir dalam mikrostruktur pelet dan semakin tebal lapisan yang terbentuk pada permukaan pelet UO2. Kekerasan lapisan permukaan pelet UO2 + dopan TiO2 sinter relatif sama dan tidak bergantung pada konsentrasi dopan TiO2. Lapisan permukaan pelet UO2 + 0,3 % TiO2, pelet UO2 + 0,5 % TiO2 dan pelet UO2 + 0,7 % TiO2 sinter mengandung unsur zirkonium masing-masing 1,97 mg, 2,47 mg dan 4,81 mg. Lapisan penyerap dapat bakar pada permukaan pelet UO2 + dopan TiO2 sinter mempunyai fasa ZrB2 dengan struktur kristal heksagonal. Kata Kunci: lapisan permukaan, penyerap dapat bakar, pelet UO2, mikrostruktur, kekerasan, komposisi kimia, struktur kristal. ABSTRACT CHARACTERIZATION OF BURNABLE ABSORBER LAYER ON THE SURFACE OF UO2 + DOPED TiO2 PELLETS. Burnable absorber layer on the surface of UO2 + doped TiO2 pellets have successfully created using RF sputtering methods. The objective of this research is to obtain of microstructure characters of UO2 + doped TiO2 pellets, thickness, micro hardness, chemical composition and crystal structure of burnable absorber layer on the surface of UO2 pellets. The methods used are the microstructure and layer thickness using optical microscopy, layer hardness with micro Vickers

  1. Thermal expansion of UO2-Gd2O3 fuel pellets

    International Nuclear Information System (INIS)

    Une, Katsumi

    1986-01-01

    In recent years, more consideration has been given to the application of UO 2 -Gd 2 O 3 burnable poison fuel to LWRs in order to improve the core physics and to extend the burnup. It has been known that UO 2 forms a single phase cubic fluorite type solid solution with Gd 2 O 3 up to 20 - 30 wt.% above 1300 K. The addition of Gd 2 O 3 to UO 2 lattices changes the properties of the fuel pellets. The limited data on the thermal expansion of UO 2 -Gd 2 O 3 fuel exist, but those are inconsistent. UO 2 -Gd 2 O 3 fuel pellets were fabricated, and the linear thermal expansion of UO 2 and UO 2 -(5, 8 and 10 wt.%)Gd 2 O 3 fuel pellets was measured with a differential dilatometer over the temperature range of 298 - 1973 K. A sapphire rod of 6 mm diameter and 15.5 mm length was used as the reference material. After the preheating cycle, the measurement was performed in argon atmosphere. The results for UO 2 pellets showed excellent agreement with the data in literatures. The linear thermal expansion of UO 2 -Gd 2 O 3 fuel pellets showed the increase with increasing the Gd 2 O 3 content. Consideration must be given to this excessive expansion in the fuel design of UO 2 -Gd 2 O 3 pellets. The equations for the linear thermal expansion and density of UO 2 -Gd 2 O 3 fuel pellets were derived by the method of least squares. (Kako, I.)

  2. Determination of U3O8 in UO2 by infrared spectroscopy

    Directory of Open Access Journals (Sweden)

    Liliane Aparecida Silva

    Full Text Available Abstract The oxygen-uranium (O-U system has various oxides, such as UO2, U4O9, U3O8, and UO3. Uranium dioxide is the most important one because it is used as nuclear fuel in nuclear power plants. UO2 can have a wide stoichiometric variation due to excess or deficiency of oxygen in its crystal lattice, which can cause significant modifications of its proprieties. O/U relation determination by gravimetry cannot differentiate a stoichiometric deviation from contents of other uranium oxides in UO2. The presence of other oxides in the manufacturing of UO2 powder or sintered pellets is a critical factor. Fourier Transform Infrared Spectroscopy (FTIR was used to identify U3O8 in samples of UO2 powder. UO2 can be identified by bands at 340 cm-1 and 470 cm-1, and U3O8 and UO3 by bands at 735 cm-1, 910 cm-1, respectively. The methodology for sample preparation for FTIR spectra acquisition is presented, as well as the calibration for quantitative measurement of U3O8 in UO2. The content of U3O8 in partially calcined samples of UO2 powder was measured by FTIR with good agreement with X-rays diffractometry (XRD.

  3. Radiolysis and corrosion of Pu-doped UO2 pellets in chloride brine

    Indian Academy of Sciences (India)

    Unknown

    Radiolysis and corrosion of. 238. Pu-doped UO2 pellets in chloride brine. M KELM* and E BOHNERT. Forschungszentrum Karlsruhe, Institut für Nukleare Entsorgung, Postfach. 3640, 76021 Karlsruhe, Germany e-mail: kelm@ine.f3k.de. Abstract. Deaerated 5 M NaCl solution is irradiated in the presence of UO2 pellets.

  4. Measurement of the friction coefficient between UO2 and cladding tube

    International Nuclear Information System (INIS)

    Tachibana, Toshimichi; Narita, Daisuke; Kaneko, Hiromitsu; Honda, Yutaka

    1978-01-01

    Most of fuel rods used for light water reactors or fast reactors consist of the cladding tubes filled with UO 2 -PuO 2 pellets. The measurement was made on the coefficient of static friction and the coefficient of dynamic friction in helium under high contact load on UO 2 /Zry-2 and UO 2 /SUS 316 combined samples at the temperature ranging from room temperature to 400 deg. C and from room temperature to 600 deg. C, respectively. The coefficient of static friction for Zry-2 tube and UO 2 pellets was 0.32 +- 0.08 at room temperature and 0.47 +- 0.07 at 400 deg. C, and increased with temperature rise in this temperature range. The coefficient of static friction between 316 stainless steel tube and UO 2 pellets was 0.29 +- 0.04 at room temperature and 1.2 +- 0.2 at 600 deg. C, and increased with temperature rise in this temperature range. The coefficient of dynamic friction for both UO 2 /Zry-2 and UO 2 /SUS 316 combinations seems to be equal to or about 10% excess of the coefficient of static friction. The coefficient of static friction for UO 2 /SUS 316 combination decreased with the increasing number of repetition, when repeating slip several times on the same contact surfaces. (Kobatake, H.)

  5. Physicochemical study of UO2SO3·3H2O

    International Nuclear Information System (INIS)

    Blatov, V.A.; Serezhkina, L.B.; Serezhkin, V.N.

    1988-01-01

    Thermal stability of UO 2 SO 3 x3H 2 O is studied and its crystallographical characteristics are determined. The UO 2 SO 3 xnH 2 O IR spectra (n=3 or 4, 5) are discussed. An assumption concerning the structure of uranyl sulfite hydrates and their affiliation to the crystal chemical group AT 3 M 2 1 is stated

  6. Characterisation and compaction behaviour of UO2 powder prepared from ADU and AUC

    International Nuclear Information System (INIS)

    Rachmawati, M.

    2000-01-01

    UO 2 powder prepared from ADU and AUC route process are characterised primarily in terms of compaction and sintering behaviour. Scientific understanding of the phenomena will give useful information leading to processing and product improvement. The investigation has been done by characterising the particle size/shape distribution using SEM and compacting the powder at 4 and 5.4 tons/cm 2 . The behaviour of the powder under compaction is observed by characterizing the pellet length, green density, microstructure, and the compression strength using micrometer SEM, and Universal Testing Machine. The results of the experiment show that the UO 2 powder ex-AUC has particles of spherical type and separate individually which provide the flowable characteristic, important for the die filling aspect during compaction step. The UO 2 powder ex-ADU is more or less agglomerated and contains very fine particles causing the difficulty in pressing. Therefore the green density resulted from UO 2 ex-AUC (6.415 g/cm 3 ) is higher than UO 2 powder of UO 2 ex-ADU (6.117 g/cm 3 . UO 2 at lower pressure (4 tons/cm 3 ) the compression strength ex-AUC green pellet (47.144 kgf) is lower than UO 2 ex-ADU (63,364 kgf), and at higher temperature the compression strength of ex-AUC (92.86 kgf) is higher than UO 2 ex-ADU (82.664 kgf). It is suggested that UO 2 ex-ADU has to be precompacted and granulated in order to increase its flowability so that the pellet length can easily be controlled during pressing (improve reproducibility). (author)

  7. The Surface Reactions of Ethanol over UO2(100) Thin Film

    KAUST Repository

    Senanayake, Sanjaya D.

    2015-10-08

    The study of the reactions of oxygenates on well-defined oxide surfaces is important for the fundamental understanding of heterogeneous chemical pathways that are influenced by atomic geometry, electronic structure and chemical composition. In this work, an ordered uranium oxide thin film surface terminated in the (100) orientation is prepared on a LaAlO3 substrate and studied for its reactivity with a C-2 oxygenate, ethanol (CH3CH2OH). With the use of synchrotron X-ray photoelectron spectroscopy (XPS), we have probed the adsorption and desorption processes observed in the valence band, C1s, O1s and U4f to investigate the bonding mode, surface composition, electronic structure and probable chemical changes to the stoichiometric-UO2(100) [smooth-UO2(100)] and Ar+-sputtered UO2(100) [rough-UO2(100)] surfaces. Unlike UO2(111) single crystal and UO2 thin film, Ar-ion sputtering of this UO2(100) did not result in noticeable reduction of U cations. The ethanol molecule has C-C, C-H, C-O and O-H bonds, and readily donates the hydroxyl H while interacting strongly with the UO2 surfaces. Upon ethanol adsorption (saturation occurred at 0.5 ML), only ethoxy (CH3CH2O-) species is formed on smooth-UO2(100) whereas initially formed ethoxy species are partially oxidized to surface acetate (CH3COO-) on the Ar+-sputtered UO2(100) surface. All ethoxy and acetate species are removed from the surface between 600 and 700 K.

  8. PELAPISAN PERMUKAAN PELET UO2 DENGAN ZIRKONIUM DIBORIDA MENGGUNAKAN METODA SPUTTERING

    Directory of Open Access Journals (Sweden)

    Sungkono Sungkono

    2016-06-01

    Full Text Available ABSTRAK PELAPISAN PERMUKAAN PELET UO2 DENGAN ZIRKONIUM DIBORIDA MENGGUNAKAN METODA SPUTTERING. Pengembangan teknologi bahan bakar nuklir bertujuan untuk meningkatkan efisiensi pengoperasian Pembangkit Listrik Tenaga Nuklir (PLTN. Salah satu solusi yang diajukan adalah penggunaan bahan bakar dengan fraksi bakar (burn up tinggi. Hal ini menyebabkan terjadinya peningkatan gas hasil fisi dan reaktivitas teras reaktor nuklir. Untuk mengendalikan kelebihan reaktivitas teras reaktor digunakan bahan bakar terintegrasi penyerap mampu bakar. Sehubungan dengan hal tersebut telah dibuat pelet UO2 berlapis tipis penyerap mampu bakar. Tujuan penelitian adalah untuk mendapatkan karakter lapisan zirkonium diborida pada permukaan pelet UO2 yaitu mikrostruktur, struktur kristal dan komposisi kimia. Pelapisan permukaan pelet UO2 dilakukan dengan bahan pelapis ZrB2 menggunakan metoda sputtering. Hasil penelitian menunjukkan bahwa mikrostruktur pelet UO2 + 0,4% Cr2O3 berupa butir-butir campuran ekuiaksial dan acicular dengan diameter 2,44 mm, sedangkan pelet UO2 + 0,3% Nb2O5 mempunyai struktur butir berupa ekuiaksial dan batang pipih dengan diameter 2,47 mm. Lapisan zirkonium diborida pada permukaan pelet UO2 + 0,4% Cr2O3 dan pelet UO2 + 0,3% Nb2O5 serupa yaitu tipis dan kompak dengan ketebalan 2,71 mm dan 2,82 mm. Identifikasi terhadap pola difraksi sinar-X pada pelet UO2 + 0,4% Cr2O3 dan pelet UO2 + 0,3% Nb2O5 menunjukkan adanya fasa UO2 dengan struktur kristal kubus dan fasa ZrB2 dengan struktur kristal heksagonal. Sementara itu, konsentrasi zirconium dalam lapisan pelet UO2 + 0,4% Cr2O3 dan pelet UO2 + 0,3%Nb2O5 diperoleh masing-masing sebesar 1,82 mg dan 1,90 mg. Adanya unsur zirkonium membuktikan bahwa lapisan ZrB2 terbentuk pada permukaan pelet UO2. Kata kunci: Pelet UO2, lapisan ZrB2, sputtering, mikrostruktur, ketebalan, struktur kristal, komposisi kimia. ABSTRACT COATING ON SURFACE OF UO2 PELLET WITH ZIRCONIUM DIBORIDE USING THE METHOD OF SPUTTERING

  9. A study on improvement of UO2 powder production process for high sintered density

    International Nuclear Information System (INIS)

    Park, Jin Hoh; Hwang, Sung Tae; Jun, Kwan Sik; Choi, Yoon Dong; Choi, Jong Hyun; Lee, Kyoo Il; Kim, Tae Joon; Jung, Kyung Chae; Kim, Kwang Lak; Kwon, Sang Woon; Kim, Byung Hoh; Hong, Soon Bok

    1995-01-01

    Various conversion processes were reviewed from the viewpoint of manufacturing cost, product quality and liquid waste. The MDD process was selected a suitable target process for the good quality of UO 2 powder and the recycling availability of nitric acid. The MDD process consists of two steps, double salt preparation [(NH 4 ) 2 UO 2 (NO 3 ) 4 ] from uranyl nitrate solution and thermal decomposition/reduction to UO 2 powder. The reaction mechanism and properties for the intermediates were analyzed to define the proposed operational conditions of the process. The conceptual process was proposed and experimental facility was designed and installed. 12 figs, 7 tabs, 7 refs. (Author)

  10. High-temperature studies of UO2 and ThO2 using neutron scattering techniques

    International Nuclear Information System (INIS)

    Hutchings, M.T.

    1987-01-01

    A review of the results of recent neutron diffraction, coherent diffuse and inelastic scattering experiments on UO 2 and ThO 2 at temperatures between 293 K and 2930 K is given. These provide direct evidence for thermally induced Frenkel oxygen lattice disorder at temperatures > 2000 K and give data on the lattice expansion, ionic potentials and elastic stiffness constants to 2930 K for the first time. The results are important for an understanding of the thermophysical behaviour of UO 2 and contribute to the data base of thermodynamic properties of UO 2 . (author)

  11. Interatomic Potentials via the Effective Action Formalism

    Science.gov (United States)

    Rasamny, M.; Valiev, M.; Fernando, G. W.

    1998-03-01

    We present a method for the generation of interatomic potentials from first principles calculations using the marat>effective action formalism which leads to a systematic definition for effective two-body interatomic potentials. We do this by reducing the fully interacting system to an auxilliary system which interacts via a two-body interatomic potential. This definition can be trivially extended to higher order interatomic potentials. Unlike other approaches, our interatomic potentials are obtained from a sampling of configuration space pertaining to the thermodynamic environment of interest.

  12. Modern x-ray spectral methods in the study of the electronic structure of actinide compounds: Uranium oxide UO2 as an example

    Directory of Open Access Journals (Sweden)

    Teterin Yury A.

    2004-01-01

    Full Text Available Fine X-ray photo electron spectral (XPS structure of uranium dioxide UO2 in the binding energy (BE range 0-~č40 eV was associated mostly with the electrons of the outer (OVMO (0-15 eV BE and inner (IVMO (15-40 eV BE valence molecular orbitals formed from the incompletely U5f,6d,7s and O2p and completely filled U6p and O2s shells of neighboring uranium and oxygen ions. It agrees with the relativistic calculation results of the electronic structure for the UO812–(Oh cluster reflecting uranium close environment in UO2, and was confirmed by the X-ray (conversion electron, non-resonance and resonance O4,5(U emission, near O4,5(U edge absorption, resonance photoelectron, Auger spectroscopy data. The fine OVMO and IVMO related XPS structure was established to yield conclusions on the degree of participation of the U6p,5f electrons in the chemical bond, uranium close environment structure and interatomic distances in oxides. Total contribution of the IVMO electrons to the covalent part of the chemical bond can be comparable with that of the OVMO electrons. It has to be noted that the IVMO formation can take place in compounds of any elements from the periodic table. It is a novel scientific fact in solid-state chemistry and physics.

  13. Measurements of density and of thermal expansion coefficient of sodium tetraborate (borax)-UO2 and of sodium metaborate-UO2 solutions

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.

    1980-12-01

    Measurements have been performed of the density and volumetric thermal expansion coefficient of liquid sodium tetraborate (borax) and of sodium metaborate both pure and with two different amounts of UO 2 dissolved in each. These data are required for the design of core-catchers based on sodium borates. The measurements have been performed with the buoyancy method in the temperature range from 850 0 C to 1325 0 C. The data for the pure borax and for the sodium metaborate agree reasonably well with the data from the literature, giving confidence that the measurements are correct and the new data for the salts with UO 2 are reliable. (orig.) [de

  14. Crystal structure of [UO2(NH35]NO3·NH3

    Directory of Open Access Journals (Sweden)

    Patrick Woidy

    2016-12-01

    Full Text Available Pentaammine dioxide uranium(V nitrate ammonia (1/1, [UO2(NH35]NO3·NH3, was obtained in the form of yellow crystals from the reaction of caesium uranyl nitrate, Cs[UO2(NO33], and uranium tetrafluoride, UF4, in dry liquid ammonia. The [UO2]+ cation is coordinated by five ammine ligands. The resulting [UO2(NH35] coordination polyhedron is best described as a pentagonal bipyramid with the O atoms forming the apices. In the crystal, numerous N—H...N and N—H...O hydrogen bonds are present between the cation, anion and solvent molecules, leading to a three-dimensional network.

  15. SPS Fabrication of Nuclear CERMET Fuel Materials using W Powder Coated UO2 Feedstocks

    Data.gov (United States)

    National Aeronautics and Space Administration — To overcome the NTP propellant feedstock challenges, MSFC developed a new powder coating technique that uses a polymer binder to coat UO2 particles with W prior to...

  16. Thermal and Mechanical Properties of UO2 and PuO2

    International Nuclear Information System (INIS)

    Kato, M.; Matsumoto, T.

    2015-01-01

    It is important to evaluate basic properties of UO 2 and PuO 2 as fundamental aspects of MA-bearing MOX fuel development. In this work, mechanical properties of UO 2 and PuO 2 were investigated by an ultrasound pulse-echo method. Longitudinal and transversal wave velocities were measured in UO 2 and PuO 2 pellets, and Young's modulus and shear modulus were evaluated, which were 219 MPa and 89 MPa for PuO 2 , and 249 MPa and 95 MPa for UO 2 , respectively. Poisson's ratio was 0.32 in both materials. The relationship between mechanical and thermal properties was described by using thermal expansion data which had been reported previously, and the heat capacity and thermal conductivity were analysed. (authors)

  17. Effect of alpha irradiation on UO2 surface reactivity in aqueous media

    International Nuclear Information System (INIS)

    Jegou, C.; Muzeau, B.; Broudic, V.; Poulesquen, A.; Roudil, D.; Jorion, F.; Corbel, C.

    2005-01-01

    The option of direct disposal of spent nuclear fuel in a deep geological formation raises the need to investigate the long-term behavior of the UO 2 matrix in aqueous media subjected to α-β-γ radiation. The β-γ emitters account for most of the activity of spent fuel at the moment it is removed from the reactor, but diminish within a millennial time frame by over three orders of magnitude to less than the long-term activity. The latter persists over much longer time periods and must therefore be taken into account over a geological disposal time scale. Leaching experiments with solution renewal were carried out on UO 2 pellets doped with alpha emitters ( 238 Pu and 239 Pu) to quantify the impact of alpha irradiation on UO 2 matrix alteration. Three batches of doped UO 2 pellets with different alpha flux levels (3.30 x 10 4 , 3.30 x 10 5 , and 3.2 x 10 6 α cm -2 s -1 ) were studied. The results obtained in aerated and deaerated media immediately after sample annealing or interim storage in air provide a better understanding of the UO 2 matrix alteration mechanisms under alpha irradiation. Interim storage in air of UO 2 pellets doped with alpha emitters results in variations of the UO 2 surface reactivity, which depends on the alpha particle flux at the interface and on the interim storage duration. The variation in the surface reactivity and the greater uranium release following interim storage cannot be attributed to the effect of alpha radiolysis in aerated media since the uranium release tends toward the same value after several leaching cycles for the doped UO 2 pellet batches and spent fuel. Oxygen diffusion enhanced by alpha irradiation of the extreme surface layer and/or radiolysis of the air could account for the oxidation of the surface UO 2 to UO 2+x . However, leaching experiments performed in deaerated media after annealing the samples and preleaching the surface suggest that alpha radiolysis does indeed affect the dissolution, which varies with the

  18. Effect of water α radiolysis on the spent nuclear fuel UO2 matrix alteration

    International Nuclear Information System (INIS)

    Lucchini, J.F.

    2001-01-01

    In the option of long term storage or direct disposal of nuclear spent fuel, it is essential to study the long-term behaviour of the spent fuel matrix (UO 2 ) in water, in presence of ionizing radiations. This work gives some knowledge elements about the impact of aerated water alpha radiolysis on UO 2 alteration. An original experiment method was used in this study. UO 2 /water interfaces were irradiated by an external He 2+ ions beam. The sequential batch dissolution tests on UO 2 samples were performed in aerated deionized water, before, during and after a-irradiation under high fluxes. A corrosion product, identified as hydrated uranium peroxide, was formed on the UO 2 surface. The uranium release was 3 to 4 orders of magnitude higher under irradiation than out of irradiation. The concentrations of the radiolysis products H 2 O 2 and H 3 O + were affected by the uranium oxide surface. They could not only explain the whole uranium release reached during irradiation in water. Leaching experiments on UO X spent fuel samples (with or without the Zircaloy clad) were also performed, in hot cells. The uranium release was relatively small, and H 2 O 2 was not detected in solution. The rates of uranium release in aerated water during one hour were calculated. They were about mg -1 .m -2 .d -1 for spent fuel and for UO 2 , and about g -1 .m -2 .d -1 for UO 2 irradiated by He 2+ ions. The comparison of the results between the two kinds of experiment shows a difference of the behaviour in water between UO 2 irradiated by He 2+ ions and spent fuel. Some hypothesis are given to explain this difference. (author)

  19. Electron probe micro-analysis of irradiated Triso-coated UO2 particles, (1)

    International Nuclear Information System (INIS)

    Ogawa, Toru; Minato, Kazuo; Fukuda, Kosaku; Ikawa, Katsuichi

    1983-11-01

    The Triso-coated low-enriched UO 2 particles were subjected to the post-irradiation electron probe micro-analysis. Observations and analyses on the amoeba effect, inclusions and solutes in the UO 2 matrix were made. In the cooler side of the particle which suffered extensive kernel migration, two significant features were observed: (1) the wake of minute particles, presumably UO 2 , left by the moving kernel in the carbon phase and (2) carbon precipitation in the pores and along the grain boundaries of the UO 2 kernel. Both features could be hardly explained by the gas-phase mechanism of carbon transport and rather suggest the solid state mechanism. Two-types of 4d-transition metal inclusions were observed: the one which was predominantly Mo with a fraction of Tc and another which was enriched with Ru and containing significant amount of Si. The Mo and Si were also found in the UO 2 matrix; the observation led to the discussion of the oxygen potential in the irradiated Triso-coated UO 2 particle. (author)

  20. Contribution of the study of a nuclear reactor accident: residual power aspects and thermodynamic of U-UO2 and UO2-ZrO2 systems

    International Nuclear Information System (INIS)

    Baichi, Mehdi

    2001-01-01

    This work is a contribution to the study of early delocalization and fission product releases during the formation of corium coming from a nuclear reactor accident. The first part deals with an analysis of corium cooling. The contribution to the power of each corium element has been calculated with time. The main elements are represented but the addition of Pu, Mo and Nb has been proposed. The last release experimental data taken into account result in a loss of residual power of 25% exclusive of corium between the emergency stop and ten days. The second part deals with the early delocalization observed during Vercors experiments. A critical selection on the U-UO 2 and UO 2 -ZrO 2 systems has been carried out. In order to complete the small and inconsistent data, thermodynamic activity measurements have been performed by mass spectrometry. The UO 2 activity on UO 2 -ZrO 2 presents a positive deviation from ideality at 2200 K and approximates ideality at 2400 K. All the data have been used for optimizing the systems with Thermo-Calc. This work has allowed to calculate the ternary systems and to define the required approach to analyze the metallic phase and corium oxides densities. (author) [fr

  1. Complete reduction of high-density UO2 to metallic U in molten Li2O-LiCl

    Science.gov (United States)

    Choi, Eun-Young; Lee, Jeong

    2017-10-01

    The large size and high density of spent fuel pellets make it difficult to use the pellets directly in electrolytic reduction (also called as oxide reduction, OR) for pyroprocessing owing to the slow diffusion of molten Li2O-LiCl salt electrolyte into the pellets. In this study, we investigated complete OR of high-density UO2 to metallic U without any remaining UO2. Only partial reductions near the surface of high-density UO2 pellets were observed under operation conditions employing fast electrolysis rate that allowed previously complete reduction of low-density UO2 pellets. Complete reduction of high-density UO2 pellets was observed at fast electrolysis rate when the pellet size was reduced. The complete reduction of high-density UO2 pellets without size reduction was achieved at slow electrolysis rate, which allowed sufficient chemical reduction of UO2 with the lithium metal generated by the cathode reaction.

  2. Valence XPS structure and chemical bond in Cs2UO2Cl4

    Directory of Open Access Journals (Sweden)

    Teterin Yury A.

    2016-01-01

    Full Text Available Quantitative analysis was done of the valence electrons X-ray photoelectron spectra structure in the binding energy (BE range of 0 eV to ~35 eV for crystalline dicaesium tetrachloro-dioxouranium (VI (Cs2UO2Cl4. This compound contains the uranyl group UO2. The BE and structure of the core electronic shells (~35 eV-1250 eV, as well as the relativistic discrete variation calculation results for the UO2Cl4(D4h cluster reflecting U close environment in Cs2UO2Cl4 were taken into account. The experimental data show that many-body effects due to the presence of cesium and chlorine contribute to the outer valence (0-~15 eV BE spectral structure much less than to the inner valence (~15 eV-~35 eV BE one. The filled U5f electronic states were theoretically calculated and experimentally confirmed to be present in the valence band of Cs2UO2Cl4. It corroborates the suggestion on the direct participation of the U5f electrons in the chemical bond. Electrons of the U6p atomic orbitals participate in formation of both the inner (IVMO and the outer (OVMO valence molecular orbitals (bands. The filled U6p and the O2s, Cl3s electronic shells were found to make the largest contributions to the IVMO formation. The molecular orbitals composition and the sequence order in the binding energy range 0 eV-~35 eV in the UO2Cl4 cluster were established. The experimental and theoretical data allowed a quantitative molecular orbitals scheme for the UO2Cl4 cluster in the BE range 0-~35 eV, which is fundamental for both understanding the chemical bond nature in Cs2UO2Cl4 and the interpretation of other X-ray spectra of Cs2UO2Cl4. The contributions to the chemical binding for the UO2Cl4 cluster were evaluated to be: the OVMO contribution - 76%, and the IVMO contribution - 24 %.

  3. Near Surface Stoichiometry in UO2: A Density Functional Theory Study

    Directory of Open Access Journals (Sweden)

    Jianguo Yu

    2015-01-01

    Full Text Available The mechanisms of oxygen stoichiometry variation in UO2 at different temperature and oxygen partial pressure are important for understanding the dynamics of microstructure in these crystals. However, very limited experimental studies have been performed to understand the atomic structure of UO2 near surface and defect effects of near surface on stoichiometry in which the system can exchange atoms with the external reservoir. In this study, the near (110 surface relaxation and stoichiometry in UO2 have been studied with density functional theory (DFT calculations. On the basis of the point-defect model (PDM, a general expression for the near surface stoichiometric variation is derived by using DFT total-energy calculations and atomistic thermodynamics, in an attempt to pin down the mechanisms of oxygen exchange between the gas environment and defected UO2. By using the derived expression, it is observed that, under poor oxygen conditions, the stoichiometry of near surface is switched from hyperstoichiometric at 300 K with a depth around 3 nm to near-stoichiometric at 1000 K and hypostoichiometric at 2000 K. Furthermore, at very poor oxygen concentrations and high temperatures, our results also suggest that the bulk of the UO2 prefers to be hypostoichiometric, although the surface is near-stoichiometric.

  4. An improved UO2 thermal conductivity model in the ELESTRES computer code

    International Nuclear Information System (INIS)

    Chassie, G.G.; Tochaie, M.; Xu, Z.

    2010-01-01

    This paper describes the improved UO 2 thermal conductivity model for use in the ELESTRES (ELEment Simulation and sTRESses) computer code. The ELESTRES computer code models the thermal, mechanical and microstructural behaviour of a CANDU® fuel element under normal operating conditions. The main purpose of the code is to calculate fuel temperatures, fission gas release, internal gas pressure, fuel pellet deformation, and fuel sheath strains for fuel element design and assessment. It is also used to provide initial conditions for evaluating fuel behaviour during high temperature transients. The thermal conductivity of UO 2 fuel is one of the key parameters that affect ELESTRES calculations. The existing ELESTRES thermal conductivity model has been assessed and improved based on a large amount of thermal conductivity data from measurements of irradiated and un-irradiated UO 2 fuel with different densities. The UO 2 thermal conductivity data cover 90% to 99% theoretical density of UO 2 , temperature up to 3027 K, and burnup up to 1224 MW·h/kg U. The improved thermal conductivity model, which is recommended for a full implementation in the ELESTRES computer code, has reduced the ELESTRES code prediction biases of temperature, fission gas release, and fuel sheath strains when compared with the available experimental data. This improved thermal conductivity model has also been checked with a test version of ELESTRES over the full ranges of fuel temperature, fuel burnup, and fuel density expected in CANDU fuel. (author)

  5. Analysis of UO2-BeO fuel performance using FRAPCON

    International Nuclear Information System (INIS)

    Revankar, Shripad T.; Chandramouli, Deepthi

    2015-01-01

    Enhanced thermal properties of nuclear fuel UO2 have been achieved by addition of compounds like BeO and SiC. This study focuses on studying the performance of UO2-10%vol BeO using FRAPCON code and examines steady state fuel behavior at high burnup. New correlations were developed to calculate thermo-physical and thermo-mechanical properties such as thermal conductivity, thermal expansion, specific heat, specific enthalpy, emissivity, vapor pressure and creep for UO2-BeO. FRAPCON code resulting from modifications done to property subroutines was used to study fission gas release behavior. With combined effect of enhanced thermal conductivity and lower thermal expansion, fission gas release predicted by FRAPCON for UO2-BeO was found to be less than that for UO2, but results from data interpolation did not exactly match for few cases. For lower burnups and lower pressure systems, FRAPCON predictions arising out of modified properties match decently well with those obtained from data interpolations and for higher levels of burnup. Enhancement of the temperature term by a factor 1.27, in the calculation of diffusion constant resulted in a good agreement between the two values. (author)

  6. XPS study of the surface chemistry of UO2 (111) single crystal film

    Science.gov (United States)

    Maslakov, Konstantin I.; Teterin, Yury A.; Popel, Aleksej J.; Teterin, Anton Yu.; Ivanov, Kirill E.; Kalmykov, Stepan N.; Petrov, Vladimir G.; Springell, Ross; Scott, Thomas B.; Farnan, Ian

    2018-03-01

    A (111) air-exposed surface of UO2 thin film (150 nm) on (111) YSZ (yttria-stabilized zirconia) before and after the Ar+ etching and subsequent in situ annealing in the spectrometer analytic chamber was studied by XPS technique. The U 5f, U 4f and O 1s electron peak intensities were employed for determining the oxygen coefficient kO = 2 + x of a UO2+x oxide on the surface. It was found that initial surface (several nm) had kO = 2.20. A 20 s Ar+ etching led to formation of oxide UO2.12, whose composition does not depend significantly on the etching time (up to 180 s). Ar+ etching and subsequent annealing at temperatures 100-380 °C in vacuum was established to result in formation of stable well-organized structure UO2.12 reflected in the U 4f XPS spectra as high intensity (∼28% of the basic peak) shake-up satellites 6.9 eV away from the basic peaks, and virtually did not change the oxygen coefficient of the sample surface. This agrees with the suggestion that a stable (self-assembling) phase with the oxygen coefficient kO ≈ 2.12 forms on the UO2 surface.

  7. Density, thermal expansion coefficient and viscosity of sodium tetraborate (borax)-UO2 and of sodium metaborate-UO2 solutions at high temperatures

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.; Roth, A.

    1983-01-01

    Measurements have been performed of the density, of the volumetric thermal expansion coefficient and of the viscosity of liquid sodium tetraborate (borax) and of sodium metaborate both pure and with two different amounts of UO 2 dissolved in each. The viscosity measurements have been performed for the solution of sodium tetraborate with UO 2 and CeO 2 , and with CeO 2 only as well. These data are required for the design of core-catchers based on sodium borates. The density measurements have been performed with the buoyancy method in the temperature range from 825 0 C to 1300 0 C, the viscosity measurements in the temperature range 700-1250 0 C with a modified Haake viscosity balance. The balance was previously calibrated at ambient temperature with a standard calibration liquid and at high temperatures, with data for pure borax available from the literature. (orig.)

  8. Thermal reactions of uranium metal, UO 2, U 3O 8, UF 4, and UO 2F 2 with NF 3 to produce UF 6

    Science.gov (United States)

    McNamara, Bruce; Scheele, Randall; Kozelisky, Anne; Edwards, Matthew

    2009-11-01

    This paper demonstrates that NF 3 fluorinates uranium metal, UO 2, UF 4, UO 3, U 3O 8, and UO 2F 2·2H 2O to produce the volatile UF 6 at temperatures between 100 and 550 °C. Thermogravimetric and differential thermal analysis reaction profiles are described that reflect changes in the uranium fluorination/oxidation state, physiochemical effects, and instances of discrete chemical speciation. Large differences in the onset temperatures for each system investigated implicate changes in mode of the NF 3 gas-solid surface interaction. These studies also demonstrate that NF 3 is a potential replacement fluorinating agent in the existing nuclear fuel cycle and in actinide volatility reprocessing.

  9. The Effect of the O/U Ratio on the Sintered Density of the UO2 Pellet

    International Nuclear Information System (INIS)

    Na, S. H.; Kang, K. H.; Kim, Y. H.; Park, C. J.; Song, K. C.; Yoo, M. J.

    2008-01-01

    The sintered density of the UO 2 pellet is an important factor to assure a stable nuclear reactor control. There are some methods to control the sintered density of the UO 2 pellet, that is, a sintering temperature and its time, a green density, an addition of pore-former or U 3 O 8 , etc. In general, it is well known that the sintered density of UO 2 pellet increases as the sintering temperature and its time and the green density increases. However the addition of a pore-former or U 3 O 8 decreases the sintered density of the UO 2 pellet, due to the leave various sizes of pore in the UO 2 matrix during sintering. In this work, the effect of the O/U ratio on the sintered density of the UO 2 pellet are investigated

  10. Influence of porosity formation on irradiated UO2 fuel thermal conductivity at high burnup

    Science.gov (United States)

    Roostaii, B.; Kazeminejad, H.; Khakshournia, S.

    2016-10-01

    Based on the existing low temperature high burnup gaseous swelling model for UO2 fuel, the matrix swelling terms are calculated and the formation of total volume porosity up to burnup of 120 MWd/KgU is computed. The irradiated UO2 thermal conductivity model based on the Maxwell-Eucken correlation for porosity factor is selected as a case study and the calculation of porosity evolution with burnup is carried out. It is shown that taking into account the formation of porosity with burnup compared to the case with constant porosity equal to as-fabricated value leads to a decrease in the UO2 fuel thermal conductivity up to 15% at high burnup values of 120 MWd/kgU. Results of the calculations are also compared with the available experimental data and good agreement was found. The conducted parametric study clearly demonstrated the impact of the key parameters on the results of the present investigation.

  11. Reducing the stoichiometric excess of HF in the hydrofluorination of UO2

    International Nuclear Information System (INIS)

    Zhao Jun; Qiu Lufu; Zhong Xing; Xu Heqing

    1989-11-01

    In a fluidized bed, UO 2 obtained from the decomposition-reduction of AUC (Ammonium Uranyl Carbonate) was fed to absorb HF remaining in the exhaust gas of UF 4 production process. In the case of 60% conversion of UO 2 and the reaction temperature in the region of 300 deg C, HF remaining in the exhaust gas in absorbing fluidized bed was less than 7 ∼ 8% (w/w), i.e. apparent stoichiometric excess of HF had reduced to 0% more or less. Hence, with the high hydrofluorination reactivity of UO 2 obtained from the decomposition-reduction of AUC, it is possible to reduce evidently the stoichiometric excess of HF in the hydrofluorination process by two fluidized beds in series in which solids move against the gas flow

  12. Thermal expansion of ThO2-2 wt% UO2 by HT-XRD

    International Nuclear Information System (INIS)

    Tyagi, A.K.; Mathews, M.D.

    2000-01-01

    The linear thermal expansion of polycrystalline ThO 2 -2 wt% UO 2 has been investigated from room temperature to 1473 K in flowing helium atmosphere using high temperature X-ray diffractometry. ThO 2 -2 wt% UO 2 shows a marginally higher linear thermal expansion as compared to pure ThO 2 . The average linear and volume thermal expansion coefficients of ThO 2 -2 wt% UO 2 are found to be α-bar a =9.74x10 -6 K -1 and α-bar v =29.52x10 -6 K -1 (298-1473 K). This study will be useful in designing the nuclear reactor fuel assembly based on ThO 2

  13. Achieving higher productivity of UO2 fuel at NUOFP through improved in-plant quality surveillance

    International Nuclear Information System (INIS)

    Meena, R.; Pramanik, D.; Sairam, S.; Rajkumar, J.V.; Rao, R.V.R.L.V.; Sinha, T.K.; Santra, N.; Rao, G.V.S.H.; Jayaraj, R.N.

    2009-01-01

    At Nuclear Fuel Complex (NFC), in the production of UO 2 fuel for PHWRs, a standard set of process parameters are monitored regularly for every lot of powder and pellet. Quality of intermediate products in the production process like UNP, ADU(dry), U 3 O 8 , UO 2+x , UO 2 granules, green pellets, sintered pellets are also regularly analysed/monitored apart from the final finished pellet and ensured to be within specified range. This range is decided by final product specifications and sometimes also based on the feed requirement in the next process in the downstream of the flow sheet. Vast experience gained over the years, behavior of various equipment under given set of conditions, feed back from the customer plants etc; have been primary criteria hither to, for defining the process conditions and chemical/physical properties of intermediate products

  14. Experimental evaluation of thermal ratcheting behavior in UO2 fuel elements

    Science.gov (United States)

    Phillips, W. M.

    1973-01-01

    The effects of thermal cycling of UO2 at high temperatures has been experimentally evaluated to determine the rates of distortion of UO2/clad fuel elements. Two capsules were rested in the 1500 C range, one with a 50 C thermal cycle, the other with a 100 C thermal cycle. It was observed that eight hours at the lower cycle temperature produced sufficient UO2 redistribution to cause clad distortion. The amount of distortion produced by the 100 C cycle was less than double that produced by the 50 C, indicating smaller thermal cycles would result in clad distortion. An incubation period was observed to occur before the onset of distortion with cycling similar to fuel swelling observed in-pile at these temperatures.

  15. Methods of modification and investigations of properties of fuel UO2

    International Nuclear Information System (INIS)

    Kurina, I.; Popov, V.; Rogov, S.; Dvoryashin, A.; Serebrennikova, O.

    2009-01-01

    In the SSC RF-IPPE the researches are carried out directed towards the uranium dioxide fuel pellets modification with the purpose of improvement of their performance parameters (increase of thermal conductivity, growth of grain for decrease gas release, decrease of interaction with coolant). The following technological methods of manufacturing of modified pellets UO 2 were used: 1) The water method including precipitation of Ammonium Polyuranate (APU) with manufacturing of simultaneously coarse and super dispersed particles, and also coprecipitation APU with additives (Cr, Ti, etc.), with the after calcination of powders, reduction to UO 2 pressing and sintering of pellets; 2) A method including addition of chemical reagent containing ammonia to the powder UO 2 manufactured under the dry or water technology; mechanical mixture; pressing and sintering of pellets. Application of the specified up methods makes manufacturing the UO 2 fuel pellets having the properties differing from pellets manufactured by industrial technology. Conclusions: 1) Properties of powders and the pellets manufactured by different technologies considerably differ; 2) Precipitate manufactured by water industrial technology, consists of phase NH 3 ·3UO 3 ·5H 2 O whereas the precipitate manufactured by nanotechnology contains in addition phase NH 3 ·2UO 3 ·3H 2 O; 3) Powders of U 3 O 8 manufactured by water nanotechnology have particles size 300-500 nm and ultra dispersive particles size ∼70-75 nm; 4) Powder UO 2 obtained by water nanotechnology differs by greater activity because all phase changes under oxidation result at lower temperatures; 5) Basic differences of properties of modified UO 2 pellets was established: decreasing of defects inside and on grains boundaries, minor porosity (pore size 0,05-0,5 μm), presence of pore of spherical form, presence of additional chemical bond U-U (presence of metal clusters), polyvalence of U; 6) Methods including addition of Cr and Ti under

  16. Effects of two types of dryer on ADU and UO2 pellet manufacture

    International Nuclear Information System (INIS)

    Wu Zhiming; He Zhengjie

    1995-05-01

    The concepts of spray drying process and pebble-bed fluidized drying process for ADU slurry is presented. And the effects of ADU powder and UO 2 powder/pellet by these processes using the statistic results from series production are discussed. It is believed that these drying methods have no influence on structure and shape of ADU particle, and thereby no difference will be made to the properties of UO 2 powder and pellet. Thus, spray drying process can really be replaced by pebble-bed fluidized drying process. (10 figs., 6 tabs.)

  17. Out-of-pile UO2/Zircaloy-4 experiments under severe fuel damage conditions

    International Nuclear Information System (INIS)

    Hofmann, P.

    1983-01-01

    Chemical interactions between UO 2 fuel and Zircaloy-4 cladding up to the melting point of zircaloy (Zry) are described. Out-of-pile UO 2 /zircaloy reaction experiments have been performed to investigate the chemical interaction behavior under possible severe fuel damage conditions (very high temperatures and external overpressure). The tests have been conducted in inert gas (1 to 80 bar) with 10-cm-long zircaloy cladding specimens filled with UO 2 pellets. The annealing temperature varied between 1000 and 1700 deg. C and the annealing period between 1 and 150 min. The extent of the chemical reaction depends decisively on whether or not good contact between UO 2 and zircaloy has been established. If solid contact exists, zircaloy reduces the UO 2 to form oxygen-stabilized α-Zr(O) and uranium metal. The uranium reacts with zircaloy to form a (U,Zr) alloy rich in uranium. The (U,Zr) alloy, which is liquid above approx. 1150 deg. C, lies between two α-Zr(O) layers. The UO 2 /zircaloy reaction obeys a parabolic rate law. The degree of chemical interaction is determined by the extent of oxygen diffusion into the cladding, and hence by the time and temperature. The affinity of zirconium for oxygen, which results in an oxygen gradient across the cladding, is the driving force for the reaction. The growth of the reaction layers can be represented in an Arrhenius diagram. The UO 2 /Zry-4 reaction occurs as rapidly as the steam/Zry-4 reaction above about 1100 deg. C. The extent of the interaction is independent of external pressure above about 10 bar at 1400 deg. C and 5 bar at 1700 deg. C. The maximum measured oxygen content of the cladding is approx. 6wt.%. Up to approx. 9 volume % of the UO 2 can be chemically dissolved by the zircaloy. In an actual fuel rod, complete release of the fission products in this region of the fuel must therefore be assumed. (author)

  18. Structure changes in UO2(hfa)2 NH3 near the melting point

    International Nuclear Information System (INIS)

    Johnson, D.A.; Taylor, J.C.; Waugh, A.B.

    1982-01-01

    A reversible colour change from lemon-yellow to orange was observed for polycrystalline UO 2 (hfa) 2 NH 3 at 100 0 C (hfa = (1,1,1,5,5,5-hexafluoro-2,4-pentanedione)). The changes in the X-ray powder pattern between -196 0 and 130 0 C were followed with a Guinier-Simon focussing camera. The colour change was not due to an α - β type structural transition as found earlier in α and β-UO 2 (hfa) 2 tmp (tmp = trimethyl phosphate), but was considered to be due to changes in the hydrogen bonding of the ammonia molecules. (author)

  19. Thermally induced frenkel disorder in UO//2 and ThO//2

    DEFF Research Database (Denmark)

    Macdonald, J. Emyr; Clausen, Kurt Nørgaard; Garrard, Barry

    1985-01-01

    Frenkel defect formation in the oxygen sublattice and the excitation of electronic defects in the form of small polarons are considered. A brief summary is given of the results of a recent investigation into possible oxygen lattice disorder in UO//2 and ThO//2 at temperatures up to 2930 K, using ...... neutron scattering techniques.......Frenkel defect formation in the oxygen sublattice and the excitation of electronic defects in the form of small polarons are considered. A brief summary is given of the results of a recent investigation into possible oxygen lattice disorder in UO//2 and ThO//2 at temperatures up to 2930 K, using...

  20. Fabrication of Tungsten-UO 2 Hexagonal-Celled Fuel-Element Configurations

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, R.R.; Cover, P.W.; Gripshover, P.J.; Wilson, W.J.

    1964-12-04

    The gas-pressure-bonding process is being evaluated as a means of fabricating tungsten-UO 2 hexagonal-celled fuel geometries. A two-part study was initiated to optimize the fuel materials and to develp the required fixturing and loading techniques. Production of fueled tungsten-coated UO 2 particles in in progress so that geometries embodying coated particles or coated particles plus fine tungsten powder can be evaluated. Tests to data have shown the rquirement for a pretreatment in which a gaseous oxide phase is removed. Initial loading and fixturing procedures were proven satisfactory by the fabrication of a 19-cylindrical-hole hexagonal-type composite.

  1. Radiolysis and corrosion of 238 Pu-doped UO2 pellets in chloride ...

    Indian Academy of Sciences (India)

    radiation from 238Pu. Experiments are conducted with 238Pu doped pellets and others with 238Pu dissolved in the brine. The radiolysis products and yields of mobilized U and Pu from the oxidative dissolution of UO2 are determined. Results ...

  2. Statistical model for grain boundary and grain volume oxidation kinetics in UO2 spent fuel

    International Nuclear Information System (INIS)

    Stout, R.B.; Shaw, H.F.; Einziger, R.E.

    1989-09-01

    This paper addresses statistical characteristics for the simplest case of grain boundary/grain volume oxidation kinetics of UO 2 to U 3 O 7 for a fragment of a spent fuel pellet. It also presents a limited discussion of future extensions to this simple case to represent the more complex cases of oxidation kinetics in spent fuels. 17 refs., 1 fig

  3. Pulsed irradiation of enriched UO2 in the Annular Core Pulse Reactor (ACPR)

    International Nuclear Information System (INIS)

    Schmidt, T.R.; Lucoff, D.M.; Reil, K.O.; Croucher, D.W.

    1974-01-01

    A series of experiments have been conducted in the Annular Core Pulse Reactor (ACPR) to determine the energy deposition and behavior of enriched UO 2 under pulse conditions. In the experiment single unirradiated pellets with enrichments up to 25 percent were pulse heated to melt temperatures. Temperature and fission product inventory measurements were made and compared with neutron transport calculations. (author)

  4. UO2 fuel pellets fabrication via Spark Plasma Sintering using non-standard molybdenum die

    Science.gov (United States)

    Papynov, E. K.; Shichalin, O. O.; Mironenko, A. Yu; Tananaev, I. G.; Avramenko, V. A.; Sergienko, V. I.

    2018-02-01

    The article investigates spark plasma sintering (SPS) of commercial uranium dioxide (UO2) powder of ceramic origin into highly dense fuel pellets using non-standard die instead of usual graphite die. An alternative and formerly unknown method has been suggested to fabricate UO2 fuel pellets by SPS for excluding of typical problems related to undesirable carbon diffusion. Influence of SPS parameters on chemical composition and quality of UO2 pellets has been studied. Also main advantages and drawbacks have been revealed for SPS consolidation of UO2 in non-standard molybdenum die. The method is very promising due to high quality of the final product (density 97.5-98.4% from theoretical, absence of carbon traces, mean grain size below 3 μm) and mild sintering conditions (temperature 1100 ºC, pressure 141.5 MPa, sintering time 25 min). The results are interesting for development and probable application of SPS in large-scale production of nuclear ceramic fuel.

  5. Specification of PWR UO2 pellet design parameters with the fuel performance code FRAPCON-1

    International Nuclear Information System (INIS)

    Silva, A.T.; Marra Neto, A.

    1988-08-01

    UO 2 pellet design parameters are analysed to verify their influence in the fuel basic properties and in its performance under irradiation in pressurized water reactors. Three groups of parameters are discussed: 1) content of fissionable and impurity materials; 2) stoichiometry; 3) density pore morpholoy, and microstructure. A methodology is applied with the fuel performance program FRAPCON-1 to specify these parameters. (author [pt

  6. Dissolution of UO2 pellets by electrography for isotope ratio analysis

    International Nuclear Information System (INIS)

    Guereli, L.; Uzmen, R.; Colak, L.; Can, F.

    2002-01-01

    Full text: A method for dissolving UO 2 pellets by using electrography is investigated. Electrography, a method used for rapid identification and qualitative analysis of certain materials was widely used in 1950s. However due to the limitations of the method it is rarely used today. The principle of the method involves anodic dissolution of metallic samples and transfer of the ions to a carrier medium usually a filter paper. Color development is used for identification. Sintered UO 2 pellets are difficult to dissolve and the procedure is time consuming. In case of illicit trafficking incidents, the amount of material may be limited and there usually is a deadline for the results, therefore dissolving the whole pellet is not preferred in most cases. In this study, a method rapid dissolving of UO 2 pellets by an electrograph is investigated. A sheet of qualitative filter paper is soaked in a suitable electrolyte and partially dried. It is placed on the cathode of the instrument. The pellet is placed between the anode and the cathode, a DC current is applied. The dissolved ions are transferred to the filter paper which is washed and diluted to the volume for analysis. Two different electrolytes, dissolving time and DC current were tried. The method is rapid, uses very little sample for analysis. Although it is usually applied to metals, acceptable dissolved amounts for analysis were obtained for UO 2 . 30-50 mA current for 2-5 minutes were sufficient for this application. (author)

  7. Behavior of UO2 and FISSIUM in sodium vapor atmosphere at temperatures up to 28000C

    International Nuclear Information System (INIS)

    Feuerstein, H.; Oschinski, J.

    1986-11-01

    In case of a HCDA a rubble bed of fuel debris may form under a sodium pool and reach high temperatures. An experimental technique was developed to study the behavior of fuel and fission products in out-of-pile tests in a sodium vapor atmosphere. Evaporation rates of UO 2 were measured up to 2800 0 C. The evaporation was found to be a complex process, depending on temperature and the 'active' surface. Evaporation restructures the surface of the samples, however no new 'active' surface is formed. UO 2 forms sometimes well shaped crystals and curious erosion products. The efficiency of the used condenser/filter lines was higher than 99.99%. In case of a HCDA all the evaporated substances will condense in the soidum pool. Thermal reduction of the UO 2 reduces the oxygen potential of the system. The final composition at 2500 0 C was found to be UO 1.95 . The only influence of the sodium vapor was found for the diffusion of UO 2 into the thoria of the crucible. Compared with experiments in an atmosphere of pure argon, the diffusion rate was reduced. (orig.) [de

  8. Methodology for Producing a Uniform Distribution of UO2 in a Tungsten Matrix

    Science.gov (United States)

    Tucker, Dennis S.; O'Conner, Andrew; Hickman, Rickman; Broadway, Jeramie; Belancik, Grace

    2015-01-01

    Current work at NASA's Marshall Space Flight Center (MSFC) is focused on the development CERMET fuel materials for Nuclear Thermal Propulsion (NTP). The CERMETs consist of uranium dioxide (UO2) fuel particles embedded in a tungsten (W) metal matrix. Initial testing of W-UO2 samples fabricated from fine angular powders performed reasonably well, but suffered from significant fuel loss during repeated thermal cycling due to agglomeration of the UO2 (1). The blended powder mixtures resulted in a non-uniform dispersion of the UO2 particles in the tungsten matrix, which allows rapid vaporization of the interconnected UO2 from the sample edges into the bulk material. Also, the angular powders create areas of stress concentrations due to thermal expansion mismatch, which eventually cracks the tungsten matrix. Evenly coating spherical UO2 particles with chemical vapor deposited (CVD) tungsten prior to consolidation was previously demonstrated to provide improved performance. However, the CVD processing technology is expensive and not currently available. In order to reduce cost and enhance performance, a powder coating process has been developed at MSFC to produce a uniform distribution of the spherical UO2 particles in a tungsten matrix. The method involves utilization of a polyethylene binder during mixing which leads to fine tungsten powders clinging to the larger UO2 spherical particles. This process was developed using HfO2 as a surrogate for UO2. Enough powder was mixed to make 8 discs (2cm diameter x 8mm thickness) using spark plasma sintering. A uniaxial pressure of 50 MPa was used at four different temperatures (2 samples at each temperature). The first two samples were heated to 1400C and 1500C respectively for 5 minutes. Densities for these samples were less than 85% of theoretical, so the time at temperature was increased to 20 minutes for the remaining samples. The highest densities were achieved for the two samples sintered at 1700C (approx. 92% of

  9. Interactions with Small and Large Sodium to UO2 Mass Ratios

    International Nuclear Information System (INIS)

    Clerici, G.; Holtbecker, H.; Schins, H.; Schlittenbardt, P.

    1976-01-01

    This paper is divided into the following three parts: - Presentation of final results of the Ispra dropping experiments; - Discussion of preliminary Na entrapment tests; - Presentation of the Press I and II codes. The experiments for which the Ispra UO 2 dropping facility was originally designed were completed in 1975. The experimental facility which initially had had difficulties in reaching the predefined working conditions gave in the last year a series of results. For this reason Ispra decided to built a similar plant for dropping experiments into water which started working in 1975. Concerning the entrapment tests it was originally foreseen to built in collaboration with GfK Karlsruhe a test section having subassembly geometry and in which the UO 2 would have been violently dispersed into the surrounding Na by the expansion of a small quantity of superheated sodium. Preliminary tests and the design work for the facility could be completed. The Press I + II codes were developed to support the above mentioned experiment - al activity. A 1-D analysis is made to investigate phenomena like UO 2 crust formation and calculate delay times between the time of the Na injection into UO 2 and the violent expansion of superheated Na. An estimate was also made of the available mechanical work in such a process which should allow to get an idea of possible energy release in a reactor core. First conclusions can be drawn from this estimate concerning the mechanical energy release in a WCA due to SPI. The result is that considerably lower energies are calculated from Na entrapment in a reactor core due to the limited amount of molten UO 2 present in the core

  10. Cs2SeO4-UO2SeO4-H2O system at 25 deg C

    International Nuclear Information System (INIS)

    Serezhkina, L.B.; Serezhkin, V.N.

    1987-01-01

    Using the method of isothermal solubility at 25 deg C the interaction of cesium and uranyl selenates in aqueous solution is studied. Formation of congruently soluble Cs 2 UO 2 (SeO 4 ) 2 x2H 2 O and Cs 2 (UO 2 ) 2 x(SeO 4 ) 3 is ascertained, their crystallographic characteristics being determined

  11. Microscopic appearance analysis of raw material used for the production of sintered UO2 by scanning electron microscope

    International Nuclear Information System (INIS)

    Liu feiming

    1992-01-01

    The paper describes the microscopic appearance of UO 2 , U 3 O 8 , ADU and AUC powders used for the production of sintered UO 2 slug of nuclear fuel component of PWR. The characteristic analysis of the microscopic appearance observed by scanning electron microscope shows that the quality and finished product rate of sintered UO 2 depend on the appearance characteristic of the active Uo 2 powder, such as grade size and its distribution, spherulitized extent, surface condition and heap model etc.. The addition of U 3 O 8 to the UO 2 powder improves significantly the quality and the finished product rate. The mechanism of this effect is discussed on the basis of the microscopic appearance characteristic for two kinds of powder

  12. Influence of radiolysis on UO2 fuel matrix dissolution under disposal conditions. Literature Study

    International Nuclear Information System (INIS)

    Ollila, K.

    2011-05-01

    The objective of this study was to examine the recent published literature on the influence of water radiolysis on UO 2 fuel matrix dissolution under the disposal conditions. The α radiation is considered to be dominating over the other types of radiations at times longer than 1000 years. The presence of the anaerobic corrosion products of iron, especially of hydrogen, has been observed to play an important role under radiolysis conditions. It is not possible to exclude gamma/beta radiolysis effects in the experiments with spent fuel, since there is not available a fuel over 100 years old. More direct measurements of α radiolysis effects have been conducted with α doped UO 2 materials. On the basis of the results of these experiments, a specific activity threshold to observe α radiolysis effects has been presented. The threshold is 1.8 x 10 7 to 3.3 x 10 7 Bq/g in anoxic 10 -3 M carbonate solution. It is dependent on the environmental conditions, such as the reducing buffer capacity of the conditions. The results of dissolution rate measurements at VTT with 233 U-doped UO 2 samples in 0.01 to 0.1 M NaCl solutions under anoxic conditions did not show any effect of α radiolysis with doping levels of 5 and 10% 233 U (3.2 x 10 7 and 6.3 x 10 7 Bq/g). Both Fe 2+ and hydrogen can act as reducing species and could react with oxidizing radiolytic species. Fe 2+ concentrations of the order of 10 -5 M can decrease the rate of H 2 O 2 production. Low dissolution rates, 2 x 10 -8 to 2 x 10 -7 /yr, have been measured in the presence of metallic Fe with 5 and 10% 233 U-doped UO 2 in 0.01 to 1 M NaCl solutions. The tests with isotope dilution method showed precipitation phenomena of U to occur during dissolution process. The concentrations of dissolved U were extremely low (≤ 8.4 x 10 -11 M). No effects of -radiolysis could be seen. It is difficult to distinguish the effects of metallic Fe, Fe 2+ or hydrogen in these tests. Hydrogen could also act as a reducing agent

  13. BURNY-SQUID, 2-D Burnup of UO2 and Mix UO2 PuO2 Fuel in X-Y or R-Z Geometry

    International Nuclear Information System (INIS)

    Rosa, I.; Zara, G.; Guidotti, R.

    1974-01-01

    1 - Nature of physical problem solved: - Multigroup neutron diffusion and burnup equations for two- to five- energy groups over a rectangular region of the x-y or r-z plane. - For a given geometry and initial enrichment, it calculates the two- to five- group flux distributions, the nuclides burnt in a time step t, and then the flux distribution again. This process is repeated until the maximum burn-up is reached. - Criticality search by uniform variation of a control isotope. - Solution of problems with fuel having different geometrical parameters, by means of super-compositions. - Recycle and restart options are available. - UO 2 and PUO 2 -UO 2 fuel can be handled. 2 - Method of solution: The zero-dimension burn-up program RIBOT-5 is coupled with the two-dimension program SQUID and alternately executed. The differential equations are solved by the difference method. 3 - Restrictions on the complexity of the problem: 200 maximum number of compositions 10,000 maximum number of mesh points 5 maximum Number of groups. 4 maximum number of super-compositions. Diagonal symmetry allowed

  14. Criticality experiments with low enriched UO2 fuel rods in water containing dissolved gadolinium

    International Nuclear Information System (INIS)

    Bierman, S.R.; Murphy, E.S.; Clayton, E.D.; Keay, R.T.

    1984-02-01

    The results obtained in a criticality experiments program performed for British Nuclear Fuels, Ltd. (BNFL) under contract with the United States Department of Energy (USDOE) are presented in this report along with a complete description of the experiments. The experiments involved low enriched UO 2 and PuO 2 -UO 2 fuel rods in water containing dissolved gadolinium, and are in direct support of BNFL plans to use soluble compounds of the neutron poison gadolinium as a primary criticality safeguard in the reprocessing of low enriched nuclear fuels. The experiments were designed primarily to provide data for validating a calculation method being developed for BNFL design and safety assessments, and to obtain data for the use of gadolinium as a neutron poison in nuclear chemical plant operations - particularly fuel dissolution. The experiments program covers a wide range of neutron moderation (near optimum to very under-moderated) and a wide range of gadolinium concentration (zero to about 2.5 g Gd/l). The measurements provide critical and subcritical k/sub eff/ data (1 greater than or equal to k/sub eff/ greater than or equal to 0.87) on fuel-water assemblies of UO 2 rods at two enrichments (2.35 wt % and 4.31 wt % 235 U) and on mixed fuel-water assemblies of UO 2 and PuO 2 -UO 2 rods containing 4.31 wt % 235 U and 2 wt % PuO 2 in natural UO 2 respectively. Critical size of the lattices was determined with water containing no gadolinium and with water containing dissolved gadolinium nitrate. Pulsed neutron source measurements were performed to determine subcritical k/sub eff/ values as additional amounts of gadolinium were successively dissolved in the water of each critical assembly. Fission rate measurements in 235 U using solid state track recorders were made in each of the three unpoisoned critical assemblies, and in the near-optimum moderated and the close-packed poisoned assemblies of this fuel

  15. Electrochemical studies of the effect of H2 on UO2 dissolution

    International Nuclear Information System (INIS)

    King, F.; Shoesmith, D.W.

    2004-09-01

    This report summarises evidence for the effect of H 2 on the oxidation and dissolution of UO 2 derived from electrochemical studies. In the presence of γ-radiation or with SIMFUEL electrodes containing ε-particles, the corrosion potential (E CORR ) of UO 2 is observed to be suppressed in the presence of H 2 by up to several hundred milli volts. This effect has been observed at room temperature with 5 MPa H 2 (in the case of γ-irradiated solutions) and at 60 deg C with a H 2 partial pressure of only 0.002-0.014 MPa H 2 with the SIMFUEL electrode. The suppression of E CORR in the presence of H 2 indicates that the degree of surface oxidation and the rate of dissolution of UO 2 is lower in the presence of H 2 .The precise mechanism of the effect of H 2 is unclear at this time. The mechanism appears to involve a surface heterogeneous process, rather than a homogeneous solution process. Under some circumstances the value of E CORR approaches the equilibrium potential for the H 2 /H + couple, suggesting galvanic coupling between sites on which this electrochemical process is catalysed and the rest of the UO 2 surface. It is also possible that H* radical species, either produced radiolytically from H 2 O or by dissociation of H 2 on ε-particles or surface-active UO 2+x sites, reduce oxidised U(V)/U(VI) surface states to U(IV). The effect of H 2 on reducing the degree of surface oxidation is only partially reversible, since surfaces reduced in H 2 atmospheres (re-)oxidise more slowly and to a lesser degree than surfaces not previously exposed to H 2 . Homogeneous reactions between dissolved H 2 and either oxidants or dissolved U(VI) cannot explain the observed effects.Regardless of the precise mechanism, the suppression of the degree of surface oxidation results in lower UO 2 dissolution rates in the presence of H 2 . Application of an electro-chemical dissolution model to the observed E CORR values suggests that the fractional dissolution rate of used fuel in the

  16. Preparation of UO2 dense spherical particles by sol-gel technique

    International Nuclear Information System (INIS)

    Urbanek, V.; Dolezal, J.

    1977-01-01

    The results of the basic research and development of processes of preparation of dense UO 2 spherical particles by sol-gel technique are presented. Attention was paid to the study of chemistry of internal gelation step in the uranylnitrate-urea-hexamethylentetramine system. The existence regions of several stable gels with different properties were established in connection with variable ratio of basic gel's components and the appropriate ''Phase diagrams'' were drawn. From these diagrams, two of the most interesting types of uranyl gels were chosen for the subsequent thermal processing which included drying, reduction and sintering. The detailed studies of each step of the whole process enabled preparation of UO 2 dense spheres with well defined microstructure

  17. Studies of the role of molten materials in interactions with UO2 and graphite

    International Nuclear Information System (INIS)

    Fink, J.K.; Heiberger, J.J.; Leibowitz, L.

    1979-01-01

    Graphite, which is being considered as a lower reactor shield in gas-cooled fast reactors, would be contacted by core debris during a core disruptive accident. Information on the interaction of graphite, UO 2 , and stainless steel is needed in assessing the safety of the GCFR. In an ongoing study of the interaction of graphite, UO 2 , and stainless steel, the effects of the steel components have been investigated by electron microprobe scans, x-ray diffraction, and reaction-rate measurements. Experiments to study the role of the reaction product, FeUC 2 , in the interaction suggested that FeUC 2 promotes the interaction by acting as a carrier to bring graphite to the reaction site. Additional experiments using pyrolytic graphite show that while the reaction rate is decreased at 2400 K, at higher temperatures the rate is similar to that using other grades of graphite

  18. Development of AUC-based process at BARC for production of free-flowing and sinterable UO2 powder

    International Nuclear Information System (INIS)

    Keni, V.S.; Ghosh, S.K.; Ganguly, C.; Majumdar, S.

    1994-01-01

    Ammonium uranium carbonate (AUC) process has been developed and industrially used in Germany for preparation of free-flowing and sinterable UO 2 powder for fabrication of UO 2 fuel pellets for light water reactors (LWR). Efforts are underway at Bhabha Atomic Research Centre (BARC) for developing AUC-based process which would yield free-flowing UO 2 powder suitable for direct pelletisation and sintering to very high density (> 96% T.D.) UO 2 fuel pellets for pressurised heavy water reactors (PHWRs) in India. The first phase of this work has been completed jointly by Chemical Engineering Division (ChED) and Radiometallurgy Division (RMD) in batches of 1.5 kg. It was possible to fabricate UO 2 pellets of density 93-95% T.D. on a reproducible basis. At ChED, process parameters have been optimised for fabrication of AUC with suitable physical properties in batches of 1.5 kg (U), starting with nuclear pure uranyl nitrate solution. At RMD calcination parameters of AUC was optimised in batches of 500 g for obtaining free-flowing UO 2 powder, suitable for direct pelletisation and sintering. The pelletisation and sintering have been carried out at Radiometallurgy Division in batches of 1-1.5 kg. The maximum achievable density of UO 2 pellets has been in the range of 95.5-96% T.D. (author). 11 refs

  19. In-pile vapor pressure measurements on UO2 and (U,Pu)O2

    International Nuclear Information System (INIS)

    Breitung, W.; Reil, K.O.

    1985-08-01

    The Effective-Equation-of-State (EEOS) experiments investigated the saturation vapor pressures of ultra pure UO 2 , reactor grade UO 2 , and reactor grade (Usub(.77)Pusub(.23))O2 using newly developed in-pile heating techniques. For enthalpies between 2150 and 3700 kJ/kg (about 4700 to 8500 K) vapor pressures from 1.3 to 54 MPa were measured. The p-h curves of all three fuel types were identical within the experimental uncertainties. An assessment of all published p-h measurements showed that the p-h saturation curve of UO 2 appears now well established by the EEOS and the CEA in-pile data. Using an estimate for the heat capacity of liquid UO 2 , the in-pile results were also compared to earlier p-T measurements. The assessments lead to proposal of two equations. Equation I, which includes a factor-of-2 uncertainty band, covers all p-T equilibrium evaporation measurements. Equation I yields 3817 K for the normal boiling point, 415.4 kJ/mol for the corresponding heat of vaporization, and 1.90 MPa for the vapor pressure at 5000 K. Equations I and II, which represent a parametric form of the p-h curve (T=parameter), also give a good description of the EEOS and CEA in-pile data. Thus the proposed equations allow a consistent representation of both p-T and p-h measurements, they are sufficiently precise for CDA analyses and cover the whole range of interest (3120-8500 K, 1400-3700 kJ/kg). (orig./HP) [de

  20. The UO2 pellets plant experimental background of the established process

    International Nuclear Information System (INIS)

    Aparicio Arroyo, E.; Alonso Folgueras, J. A.

    1969-01-01

    An account of the UO 2 research and development carried out at the JEN is first given. This includes a 10 tons/year pellet Plant construction. Experimental background of the process is established, pointing out both milling advantages and risks, granulation devices, automatic press selection, binder removing and sintering furnaces. Origin, surface area, grain size and 0/U rate are considered as raw material reception parameters, although this process shows a wide scope. (Author) 13 refs

  1. Determination of fluoride content in UO2F2 and ADUF solution by ion selective electrode

    International Nuclear Information System (INIS)

    Samanta, Papu; Kumar, Pradeep; Bagchi, A.C.

    2017-01-01

    During production of uranium metal powder, liquid solution UO 2 F 2 and ADUF containing high content of fluoride gets generated. Fluoride being corrosive in nature, fluorides concentration needs to determined. Ion selective electrode, LaF 3 (Eu) crystal, has been used. Uranium was found to interfere with fluoride analysis. Study was carried out to selectively remove uranium by solvent extraction employing D2EHPA+Cyanex 923 and TBP in dodecane. The TBP was found effective to remove uranium. (author)

  2. Physical and chemical characterization of the (Th, U)O2 mixed oxide fuel

    International Nuclear Information System (INIS)

    Santos, A.M.M. dos; Avelar, M.M.; Palmieri, H.E.L.; Lameiras, F.S.; Ferreira, R.A.N.

    1986-01-01

    The NUCLEBRAS R and D Center (Centro de Desenvolvimento da Tecnologia Nuclear - CDTN) has been performing, together with german institutions (Kernforschungsanlage Julich GmbH - KFA, Krafwerk Union A.G. - KWU and NUKEM GmbH), a program for utilization of thorium in pressurized water reactors. In this paper are presented the physical and chemical characterizations necessary to quality the (Th, U)O 2 fuel and the respective methods. (Author) [pt

  3. New fabrication method of UO2-Gd2O3 pellet

    International Nuclear Information System (INIS)

    Yoo, M. J.; Yang, C. M.; Kim, Y. R.; Na, S. H.; Kim, S. Y.; Kim, Y. K.; Lee, S. C.; Lee, Y. W.

    2003-01-01

    UO 2 -8wt%Gd 2 O 3 pellets were fabricated by a new method. Two processes - milling by a continous-type attrition mill and spherodizing- were introduced in the fabrication method. The microstructure of sintered pellet appeared homogenous and showed larger grain size than that of conventional method which generally involves a mechanical mxing. And it appears that both precompacting process and granulating process can be avoided owing to good flow ability of the milled powder with the spherodizing treatment

  4. Analysis of enriched UO2 light water moderated lattices using CAROL code

    International Nuclear Information System (INIS)

    Bhatia, H.K.; Thaker, K.

    1978-01-01

    An analysis of uniform lattice experiments for enriched UO 2 rods in light water moderator using CAROL code is presented. The experiments selected have four different enrichments and moderator to fuel volume ratio varying between 1 to 4. The deviation of ksub(eff) has been observed to lie between +- 0.01 over the important range of moderator to fuel volume ratio. In addition to reactivity prediction, the calculated and measured relative reaction rates have also been compared. (author)

  5. Physical and chemical characterization of the (Th, U)O2 mixed oxide fuel

    International Nuclear Information System (INIS)

    Santos, A.M.M. dos; Avelar, M.M.; Palmieri, H.E.L.; Lameiras, F.S.; Ferreira, R.A.N.

    1986-01-01

    The Nuclebras R and D Center (Centro de Desenvolvimento da Tecnologia Nuclear - CDTN) has been performing, together with german institutions (Kernforschungsanlage Juelich GmbH - KFA, Kraftwerk Union A.G. KWU NUKEM Gmbh), a program for utilization of thorium in pressurized water reactors. In this paper are presented the physical and chemical chacterizations necessary to qualify the (Tn, U)O 2 and the respective methods. (Author) [pt

  6. Neutron scattering investigation of disorder in UO2 and ThO2 at high temperatures

    International Nuclear Information System (INIS)

    Clausen, K.; Hayes, W.; Macdonald, J.E.; Schnabel, P.

    1983-01-01

    A brief account of the use of neutron scattering techniques to investigate anion lattice disorder and electronic disorder in UO 2 is given. Preliminary data on diffraction, diffuse scattering, and phonon mode behaviour at temperatures up to 2673 K are presented. The results show no direct evidence of Frenkel defect formation below 2400 K, but there are indications that some defect formation occurs above this temperature. (author)

  7. Physical characteristics of Gd2O3-UO2 fuel in LWR

    International Nuclear Information System (INIS)

    Matsuura, Shojiro; Kobayashi, Iwao; Furuta, Toshiro; Toba, Masao; Tsuda, Katsuhiro.

    1981-12-01

    A series of critical experiments in light water lattice were carried out on five kinds of Gadolinia-Uranium dioxide (Gd 2 O 3 -UO 2 ) test fuel rods containing 0.0, 0.05, 0.25, 1.50, 3.00 weight % of Gd 2 O 3 in Gd 2 O 3 -UO 2 . Reactivity effect, power distribution, neutron flux distribution, and temperature coefficient were measured for three types of lattices which were in shapes of annular, rectangular parallele-piped, and JPDR mockup core. The theoretical values corresponding to the measured ones were obtained by means of the design method for the FTA which is the test fuel assembly with Gd 2 O 3 -UO 2 rods for JPDR, and the accuracy was checked. In general, the calculated values were in good agreement with the measured ones. Besides, the following characteristics of Gd 2 O 3 -UO 2 rods are recognized both in measurement and calculation, i.e. (1) the effect due to gadolinia on reactivity, power distribution, and thermal neutron flux distribution are steeply saturating; the gadolinia content of only 1.50 weight % is enough to reach the almost saturated condition, (2) the relative power becomes 20% to that of normal fuel under the saturated condition, (3) the relation between the negative reactivity and the power depression effect due to gadolinia is almost linear, and (4) the interference on power depression between the adjacent gadolinia loaded rods is almost negligible, and that on reactivity effect is 15% at most. (author)

  8. Thermal performance prediction of UO2 pellet partly containing 9%w tungsten network

    International Nuclear Information System (INIS)

    Suwardi

    2008-01-01

    Sintered UO 2 exhibits very stable in reactor core compared to UC, UN, U metal and its alloys. However, its thermal conductivity is very low (2.about.5 W/m K), that limits its performance. UO 2 pellet containing Tungsten network invented by Song improves considerably its conductivity. The paper reports an analysis of thermal performance for UO 2 pellet that contains partly or wholly with 9% b. of Tungsten. The tungsten network having a high melting point and excellent thermal conductivity is continuously formed around UO 2 grains. Since the presence of network decreases the amount of fissile material and the burn up of fissile material is higher in the near surface zone of pellet but high temperature zone that releases low conductivity fission gas to the gap located in inner part of pellet, the analysis has been done for different outer radial-portion of tungsten-free pellet. The analysis takes into account the correction factor for pellet conductivity related to both pore and temperature distribution and high burn up effect. The gap conductance has been considered invariable since decrease caused by wider gap size related to lower pellet expansion is compensated by increase caused by fewer of refractory fission gas released. The results (47 kw/m, 40% burnup) show temperature decrease in all of pellet position containing W network. Pellet containing 9%b. tungsten network lower consecutively its center line temperature from 1578 to 1406, 1292, 1231, 1192, 1111, and 1038 deg C for 0, 50, 67, 75, 80, 90, and 100 % portion of network. An 80 to 90 % portion of inner pellet containing tungsten network can be considered a best fuel design. This preliminary analysis is prospective and more realistic one is recommended. (author)

  9. Cyclic process for re-use of waste water generated during the production of UO2

    International Nuclear Information System (INIS)

    Crossley, T.J.

    1976-01-01

    The process is described whereby waste water produced during the hydrolysis and ammonium hydroxide treatment of UF 6 to produce ammonium diuranate is recycled for reuse. The solution containing large amounts of ammonia and fluorides and trace amounts of uranium is first treated with lime to precipitate the fluoride. The ammonia is distilled off and recycled to UO 2 F 2 treatment vessel. The CaF 2 precipitate is separated by centrifugation and the aqueous portion is passed through cationic exchange beds

  10. Modelling the radiolytic corrosion of α-doped UO2 and spent nuclear fuel

    Science.gov (United States)

    Liu, Nazhen; Qin, Zack; Noël, James J.; Shoesmith, David W.

    2017-10-01

    A model previously developed to predict the corrosion rate of spent fuel (UO2) inside a failed waste container has been adapted to simulate the rates measured on a wide range of α-doped UO2 and spent fuel specimens. This simulation confirms the validity of the model and demonstrates that the steady-state corrosion rate is controlled by the radiolytic production of H2O2 (which has been shown to be the primary oxidant driving fuel corrosion), irrespective of the reactivity of the UO2 matrix. The model was then used to determine the consequences of corrosion inside a failed container resealed by steel corrosion products. The possible accumulation of O2, produced by H2O2 decomposition, was found to accelerate the corrosion rate in a closed system. However, the simultaneous accumulation of radiolytic H2, which is activated as a reductant on the noble metal (ε) particles in the spent fuel, rapidly overcame this acceleration leading to the eventual suppression of the corrosion rate to insignificant values. Calculations also showed that, while the radiation dose rate, the H2O2 decomposition ratio, and the surface coverage of ε particles all influenced the short term corrosion rate, the influence of the radiolytically produced H2 was the overwhelming influence in reducing the rate to negligible level (i.e., <10-20 mol m-2 s-1).

  11. Model for evolution of grain size in the rim region of high burnup UO2 fuel

    Science.gov (United States)

    Xiao, Hongxing; Long, Chongsheng; Chen, Hongsheng

    2016-04-01

    The restructuring process of the high burnup structure (HBS) formation in UO2 fuel results in sub-micron size grains that accelerate the fission gas swelling, which will raise some concern over the safety of extended the nuclear fuel operation life in the reactor. A mechanistic and engineering model for evolution of grain size in the rim region of high burnup UO2 fuel based on the experimental observations of the HBS in the literature is presented. The model takes into account dislocations evolution under irradiation and the grain subdivision occur successively at increasing local burnup. It is assumed that the original driving force for subdivision of grain in the HBS of UO2 fuel is the production and accumulation of dislocation loops during irradiation. The dislocation loops can also be annealed through thermal diffusion when the temperature is high enough. The capability of this model is validated by the comparison with the experimental data of temperature threshold of subdivision, dislocation density and sub-grain size as a function of local burnup. It is shown that the calculated results of the dislocation density and subdivided grain size as a function of local burnup are in good agreement with the experimental results.

  12. Thermal diffusivity of high burn-up UO2 pellet irradiated at HBWR

    International Nuclear Information System (INIS)

    Nakamura, J.

    1998-01-01

    Thermal diffusivity of high burn-up UO 2 (63 MWd/kgU) irradiated at HBWR was measured from 290 to 1794 K by laser flash method. The thermal diffusivity of high burn-up UO 2 was lower than half that of unirradiated UO 2 at room temperature and the difference between them decreased as the measurement temperature increased. The measurements were repeated three or four times on the same sample, with increasing the maximum measurement temperature. Then, thermal diffusivity gradually increased at low temperature region. It was estimated that this increase of thermal diffusivity was mainly caused by the recovery of radiation damage. The thermal diffusivity data of the samples were separated into two groups. The difference of the thermal diffusivity of these groups was mostly explained by the effect of density difference. The present results on the samples measured after annealing at temperature between 700 and 1300 K were a little smaller than those of SIMFUEL, which chemically simulated the effects of burn-up by adding solid FPs. The relative degradation of thermal conductivity with burn-up estimated from the present data agreed well with that derived from fuel centre temperature measurement by expansion thermometer at HBWR. (author)

  13. A microstructure-dependent model for fission product gas release and swelling in UO2 fuel

    International Nuclear Information System (INIS)

    Notley, M.J.F.; Hastings, I.J.

    1979-06-01

    A model for the release of fission gas from irradiated UO2 fuel is presented. It incorporates fission gas diffusion bubble and grain boundary movement,intergranular bubble formation and interlinkage. In addition, the model allows estimates of the extent of structural change and fuel swelling. In the latter, contributions of thermal expansion, densification, solid fission products, and gas bubbles are considered. When included in the ELESIM fuel performance code, the model yields predictions which are in good agreement with data from UO2 fuel elements irradiated over a range of water-cooled reactor conditions: linear power outputs between 40 and 120 kW/m, burnups between 10 and 300 MW.h/kg U and power histories including constant, high-to-low and low-to-high power periods. The predictions of the model are shown to be most sensitive to fuel power (temperature), the selection of diffusion coefficient for fission gas in UO2 and burnup. The predictions are less sensitive to variables such as fuel restraint, initial grain size and the rate of grain growth. (author)

  14. Effects of hyperstoichiometry and fission products on the electrochemical reactivity of UO2 nuclear fuel

    International Nuclear Information System (INIS)

    Betteridge, J.S.; Scott, N.A.M.; Shoesmith, D.W.; Bahen, L.E.; Hocking, W.H.; Lucuta, P.G.

    1997-03-01

    The effects of hyperstoichiometry and fission products on the electrochemical reactivity Of UO 2 nuclear fuel have been systematically investigated using cyclic voltammetry and the O 2 reduction reaction. Significant constraints are placed on the active-site model for O 2 reduction by the modest impact of bulk hyperstoichiometry. Formation of the U 4 O 9 derivative phase was associated with a marked increase in transient surface oxidation/reduction processes, which probably involve localized attack and might be fostered by tensile stresses induced during oxidation. Electrocatalytic reduction Of O 2 on simulated nuclear fuel (SIMFUEL) has been determined to increase progressively with nominal burnup and pronounced enhancement of H 2 O reduction has been observed as well. Substitution of uranium by lower-valence (simulated) fission products, which was formerly considered the probable cause for this behaviour, has now been shown to merely provide good electrical conductivity. Instead, the enhanced reduction kinetics for O 2 and H 2 O on SIMFUEL can be fully accounted for by noble metals, which segregate to the UO 2 grain boundaries as micron-sized particles, despite their low effective surface area. Apparent convergence of the electrochemical properties Of UO 2 and SIMFUEL through natural corrosion likely reflects evolution toward a common active surface. (author)

  15. Modeling of Pore Coarsening in the Rim Region of High Burn-up UO2 Fuel

    Directory of Open Access Journals (Sweden)

    Hongxing Xiao

    2016-08-01

    Full Text Available An understanding of the coarsening process of the large fission gas pores in the high burn-up structure (HBS of irradiated UO2 fuel is very necessary for analyzing the safety and reliability of fuel rods in a reactor. A numerical model for the description of pore coarsening in the HBS based on the Ostwald ripening mechanism, which has successfully explained the coarsening process of precipitates in solids is developed. In this model, the fission gas atoms are treated as the special precipitates in the irradiated UO2 fuel matrix. The calculated results indicate that the significant pore coarsening and mean pore density decrease in the HBS occur upon surpassing a local burn-up of 100 GWd/tM. The capability of this model is successfully validated against irradiation experiments of UO2 fuel, in which the average pore radius, pore density, and porosity are directly measured as functions of local burn-up. Comparisons with experimental data show that, when the local burn-up exceeds 100 GWd/tM, the calculated results agree well with the measured data.

  16. TiO2 doped UO2 fuels sintered by spark plasma sintering

    Science.gov (United States)

    Yao, Tiankai; Scott, Spencer M.; Xin, Guoqing; Lian, Jie

    2016-02-01

    UO2 fuels doped with oxide additives Cr2O3 and TiO2 display larger grain size, improving fission product retention capability and thus accident tolerance. Spark plasma sintering (SPS) was applied to consolidate TiO2-doped UO2 fuel pellets with 0.5 wt % dopant concentration, above its solubility, in order to induce eutectic phase formation and promote sintering kinetics. The grain size can reach 80 μm by sintering at 1700 °C for 20 min, and liquid U-Ti-O eutectic phase occurs at the triple junction of grain boundaries and significantly improves grain growth during sintering. The oxide additive also impedes the reduction of the initial hyperstoichiometric fuel powders to more stoichiometric fuel pellets upon SPS process. Thermal-mechanical properties of the sintered doped fuel pellets including thermal conductivity and hardness are measured and compared with undoped fuel pellets. The enlarged grain size (80 μm) and densification within short sintering duration highlight the immense possibility of SPS in fabricating large grained UO2 fuel pellets to improve fuel performance.

  17. The effect of U3O-8 addition on the UO2 pellet

    International Nuclear Information System (INIS)

    Indrati, Y.T.; Syarif, D. G.; Handayani, A.

    1998-01-01

    The purpose of varied U 3 O 8 addition on the UO 2 pellet fabrication is to from 1-3 mu pores. The green pellets, compacted with 3 ton/cm 2 , are a mixture powder of UO 2 , TiO 2 (0.1% weight) and varied U 3 O 8 (0-12.5% weight). The green pellets were presintered by H2 atmosphere. The presintered pellets were put on the ceramic crucibles and than those were put on the SS 316 tube with argon atmosphere. The 1400 o C sintering was hold with the soaking time 3 hr and the same rate of heating and cooling 150 o C/hr. The UO 2 pellet with 5% (weight) U 3 O 8 addition has 95.17% of theoretic density and 548.4 ±6.57 VH. Based on the identification of microstructure of pellet, it is not acceptable for nuclear fuel although pellet has 10.02 mu on grain size and 1.3 mu on closed pore size. By the diffractometer X-ray, crystal structure of pellet is face centered cubic (FCC) with the O/U ratio is 2.08

  18. A Design Study on Experimental Power Reactor Core Fueled with UO2 CFP

    International Nuclear Information System (INIS)

    Aziz, Ferhat; Rivai, Abu Khalid

    2003-01-01

    A neutronic study on core design of a 300 MWt EPR was performed. In this study the use of 4.8% enriched UO 2 coated fuel particle was analyzed. The design was then compared to 5% enriched UO 2 pin fueled EPR based on existing PWRs. Both reactors are operated with single batch refueling system with a cycle length of 3 years. The core physics parameters analyzed were : effective multiplication factor in a cycle, flux distributions and cycle burnup. The results of calculation showed that the core effective multiplication factor for reactor with fuel compact can be maintained at 1.2841 at beginning of cycle (BOC) and 1.0060 at end of cycle (EOC). As for the UO 2 pin fueled reactor, the effective multiplication factor was 1.1927 at BOC and 1.0514 at EOC. The size of active core for the CFP fueled reactor were 320 cm in height and 320 cm in diameter. As for pin fueled reactor, the height was 200 cm and diameter was 180 cm. The results of calculations showed that neutron flux distribution was quite flat for both types of reactor designs, although the volume of CFP fueled reactor was 5 times as big as the pin fueled reactor

  19. The influence of porosity on the thermal conductivity of irradiated UO2 fuel

    International Nuclear Information System (INIS)

    Bakker, K.; Kwast, H.; Cordfunke, E.H.P.

    1994-12-01

    The influence of porosity on the thermal conductivity of irradiated UO 2 fuel has been determined with the Finite Element Method (FEM). Light-microscopy photographs were made of the fuel. The pore shape and the pore distribution are entered in the FEM program from these photographs. The two dimensional (2D) thermal conductivity in the plane of the photograph is obtained from the FEM calculations. The 2D thermal conductivity, that has no physical meaning itself, is the lower limit of the three dimensional (3D) thermal conductivity. For three well defined pore shapes the relation is determined between the 2D thermal conductivity and the 3D thermal conductivity. From these computations a simple relation is obtained that transfers the 2D thermal conductivity into the 3D thermal conductivity, independent of the pore shape. The influence of porosity on the 3D thermal conductivity of irradiated UO 2 fuel and UO 2 fuel doped with Nb 2 O 5 was computed with the FEM. (orig.)

  20. Improving the Thermal Conductivity of UO2 Fuel with the Addition of Graphene

    International Nuclear Information System (INIS)

    Cho, Byoung Jin; Kim, Young Jin; Sohn, Dong Seong

    2012-01-01

    Improvement of fuel performances by increasing the fuel thermal conductivity using the BeO or W were reported elsewhere. In this paper, some major fuel performances of improved thermal conductivity oxide (ICO) nuclear fuel with the addition of 10 v/o graphene have been compared to those of standard UO 2 fuel. The fuel thermal conductivity affects many performance parameters and thus is an important parameter to determine the fuel performance. Furthermore, it also affects the performance of the fuel during reactor accidents. The improved thermal conductivity of the fuel would reduce the fuel temperature at the same power condition and would improve the fission gas release, rod internal pressure and fuel stored energy. Graphene is well known for its excellent electrical conductivity, strength and thermal conductivity. The addition of graphene to the UO 2 fuel could increase the thermal conductivity of the ICO fuel. Although the graphene material is extensively studied recently, the characteristics of the graphene material, especially the thermal properties, are not well-known yet. In this study, we used the Light Water Reactor fuel performance analysis code FRAPCON-3.2 to analyze the performance of standard UO 2 and ICO fuel

  1. Ab initio calculation of oxygen self-diffusion coefficient in uranium dioxide UO2

    Science.gov (United States)

    Dorado, Boris; Garcia, Philippe; Torrent, Marc

    Uranium dioxide UO2 is the most widely used nuclear fuel worldwide and its atomic transport properties are relevant to practically all engineering aspects of the material. Although transport properties have already been studied in UO2 by means of first-principles calculations, the ab initio determination of self-diffusion coefficients has up to now remained unreachable because the relevant computational tools were neither available or adapted. The present work reports our results related to the ab initio calculation of the oxygen self-diffusion coefficient in UO2. We first determine the Gibbs free energies of formation of oxygen charged defects by calculating both the electronic and vibrational (hence entropic) contributions. Then, we use the transition state theory in order to compute the effective jump frequency of the defects, which in turn provides us with the value of the pre-exponential factor. The results are compared to self-diffusion data obtained experimentally with a careful monitoring of the relevant thermodynamic conditions (oxygen partial pressure, temperature, impurity content).

  2. The credit analysis of recycling beryllium and uranium in BeO-UO2 nuclear fuel

    International Nuclear Information System (INIS)

    Kim, Sungki; Ko, Wonil; Zhou, W.; Revankar, Shripad T.; Chung, Yanghon; Bang, Sungsig

    2012-01-01

    This study quantifies the credits of beryllium and uranium which are used as the raw materials for BeO-UO 2 nuclear fuel by analyzing the influence of their credits on the nuclear fuel cycle cost was analyzed, where the credit was defined as the value of raw materials recovered from spent fuel and the raw materials that were re-cycled. The credits of beryllium and uranium at 60 MWD/kg burn-up were -0.22 Mills/kWh and -0.14 Mills/kWh, respectively. These findings were based on the assumption that the optimal mixing proportion of beryllium in the BeO-UO 2 nuclear fuel is 4.8 wt%. In sum, the present study verified that the credits of beryllium and uranium in relation to BeO-UO 2 nuclear fuel are significant cost drivers in the cost of the nuclear fuel cycle and in estimating the nuclear fuel cycle of the reprocessing option for spent nuclear fuels. (author)

  3. Dopant solubility and lattice contraction in gadolinia and gadolinia-chromia doped UO2 fuels

    Science.gov (United States)

    Cardinaels, T.; Hertog, J.; Vos, B.; de Tollenaere, L.; Delafoy, C.; Verwerft, M.

    2012-05-01

    Gadolinia doped UO2 fuel is widely used as burnable neutron absorber in Light Water Reactors to reduce power peaking and excess reactivity during the first reactor cycle of fresh fuel assemblies. The thermal conductivity of gadolinia doped fuel is substantially lower than that of standard UO2. To maintain safety margins later in life, some design or operating restrictions can be defined, for example to compensate higher fission gas release levels. Development of large grain U/Gd fuel by suitable doping, e.g. Cr2O3, could offer a solution to such restrictions, but solid state information about the double doped (U1-x-yGdxCry)O2 system is very scarce. In the present paper, we present X-ray diffraction and microstructure results of standard U/Gd fuel and chromia doped U/Gd fuel manufactured by powder metallurgy. The dissolution of chromium in (U1-xGdx)O2 as a function of Gd content, the role of free UO2 and the lattice contraction at different Gd and Cr doping levels of (U1-x-yGdxCry)O2 is studied both for the single doped and double doped system. On the basis of lattice contraction and precise measurements of the composition of the solid solution phases, the evolution of theoretical density with dopant concentration is derived.

  4. High temperature thermal conductivity measurements of UO2 by Direct Electrical Heating. Final report

    International Nuclear Information System (INIS)

    Bassett, B.

    1980-10-01

    High temperature properties of reactor type UO 2 pellets were measured using a Direct Electrical Heating (DEH) Facility. Modifications to the experimental apparatus have been made so that successful and reproducible DEH runs may be carried out while protecting the pellets from oxidation at high temperature. X-ray diffraction measurements on the UO 2 pellets have been made before and after runs to assure that sample oxidation has not occurred. A computer code has been developed that will model the experiment using equations that describe physical properties of the material. This code allows these equations to be checked by comparing the model results to collected data. The thermal conductivity equation for UO 2 proposed by Weilbacher has been used for this analysis. By adjusting the empirical parameters in Weilbacher's equation, experimental data can be matched by the code. From the several runs analyzed, the resulting thermal conductivity equation is lambda = 1/4.79 + 0.0247T/ + 1.06 x 10 -3 exp[-1.62/kT/] - 4410. exp[-3.71/kT/] where lambda is in w/cm K, k is the Boltzman constant, and T is the temperature in Kelvin

  5. Neutron diffraction study of the in situ oxidation of UO(2).

    Science.gov (United States)

    Desgranges, Lionel; Baldinozzi, Gianguido; Rousseau, Gurvan; Nièpce, Jean-Claude; Calvarin, Gilbert

    2009-08-17

    This paper discusses uranium oxide crystal structure modifications that are observed during the low-temperature oxidation which transforms UO(2) into U(3)O(8). The symmetries and the structural parameters of UO(2), beta-U(4)O(9), beta-U(3)O(7), and U(3)O(8) were determined by refining neutron diffraction patterns on pure single-phase samples. Neutron diffraction patterns were also collected during the in situ oxidation of powder samples at 483 K. The lattice parameters and relative ratios of the four pure phases were measured during the progression of the isothermal oxidation. The transformation of UO(2) into U(3)O(8) involves a complex modification of the oxygen sublattice and the onset of complex superstructures for U(4)O(9) and U(3)O(7), associated with regular stacks of complex defects known as cuboctahedra, which consist of 13 oxygen atoms. The kinetics of the oxidation process are discussed on the basis of the results of the structural analysis.

  6. Neutron Diffraction Study of the in Situ Oxidation of UO2

    International Nuclear Information System (INIS)

    Desgranges, L; Rousseau, G.; Baldinozzi, G.; Calvarin, G.; Baldinozzi, G.; Calvarin, G.; Niepce, J.C.

    2009-01-01

    This paper discusses uranium oxide crystal structure modifications that are observed during the low-temperature oxidation which transforms UO 2 into U 3 O 8 . The symmetries and the structural parameters of UO 2 , β-U 4 O 9 , β-U 3 O 7 , and U 3 O 8 were determined by refining neutron diffraction patterns on pure single-phase samples. Neutron diffraction patterns were also collected during the in situ oxidation of powder samples at 483 K. The lattice parameters and relative ratios of the four pure phases were measured during the progression of the isothermal oxidation. The transformation of UO 2 into U 3 O 8 involves a complex modification of the oxygen sublattice and the onset of complex superstructures for U 4 O 9 and U 3 O 7 , associated with regular stacks of complex defects known as cub-octahedra, which consist of 13 oxygen atoms. The kinetics of the oxidation process are discussed on the basis of the results of the structural analysis. (authors)

  7. Neutronics characteristics of micro-heterogeneous ThO2-UO2 PWR cores

    International Nuclear Information System (INIS)

    Zhao, X.; Driscoll, M.J.; Kazimi, S.

    2001-01-01

    A new fuel concept, axially-micro-heterogeneous ThO 2 -UO 2 fuel, where ThO 2 fuel pellets and UO 2 fuel pellets are stacked in separate layers in the fuel rods, is being studied at MIT as an option to reduce plutonium production in LWR fuel. Very interesting neutronic behavior is observed: (1) A reactivity increase of 3% to 4% at EOL for a given 235 U inventory which results in a 20-30% increase in average core discharge burnup; (2) For certain configurations, a ''burnable poison'' effect is observed. Analysis shows that these effects are achieved due to a combination of changes in self-shielding, local fissile worth, and conversion ratio, among which self-shielding is the dominant effect at the end of a reactivity-limited burnup. Other variations of micro-heterogeneous UO 2 -ThO 2 fuel including duplex pellets, checkerboard pin distribution, and checkerboard-axial combinations have also been investigated, and their neutronic performance compared. It is concluded that the axial fuel micro-heterogeneity provides the largest gain in reactivity-limited burnup. (author)

  8. Thermodynamic stability of the UO2 surfaces: Interplay between over-stoichiometry and polarity compensation

    Science.gov (United States)

    Bottin, François; Geneste, Grégory; Jomard, Gérald

    2016-03-01

    The thermodynamic stability of UO2 surfaces is investigated using ab initio calculations. We employ the GGA+U framework to properly model the strong electronic correlations of the uranium 5 f electrons. Among the seven terminations of the (100), (110), and (111) orientations studied in this paper, we predict that the stoichiometric O-(111) is the most stable one under oxygen-poor or -intermediary environments. At odds with other fluorite surfaces, the overstoichiometric and polar O2-(100) and O2-(111) terminations become the most stable in oxygen-rich environments. For the latter, strong modifications of the electronic structure appear within the upper layers, in order to fulfill the polarity compensation criterion. Some U-5 f states are emptied, leading to higher oxidation 5 + and 6 + states for uranium in the outermost layers, but leaving the surface insulating. This unexpected polarity compensation mechanism is not observed for other charge transfer compounds (such as PuO2) and can be related to the f -f Mott-Hubbard band gap of the UO2 material. By considering the most stable stoichiometric and overstoichiometric terminations, the Castell's ratio can be fulfilled, explaining the Wulff shape of nanovoids in UO2 crystals.

  9. Theoretical comparative study of the industrial fabrication routes for UO2 powder

    International Nuclear Information System (INIS)

    Gonzaga, Reinaldo; Goncalves, Joao da Silva

    2008-01-01

    UO 2 powder is produced in an industrial scale by different fabrication routes, divided into dry and wet routes, or a combination of both. The wet processes most often used industrially are the ADU and AUC processes, whose names originate in the intermediate precipitate obtained during powder fabrication, Ammonium Diuranate and Ammonium Uranil Carbonate. Considering the dry processes, the most widely used ones are the DC (Dry Conversion) and IDR (Integrate Dry Route) process. As to the differences and peculiarities among the fabrication routes, each has marked advantages and disadvantages that are of extreme importance when it comes to selecting and establishing a UO 2 powder production plant based on a particular fabrication route. Among the important factors of comparison to be considered are the product quality characteristics, production capability, quantity of waste, operating costs of each process with raw material, labor, etc. This paper is intended to make a theoretical comparison between wet and dry processes for UO 2 powder fabrication, taking as the basis the previously mentioned factors of comparison. (author)

  10. Complexing in K2SeO4-UO2SeO4-H2O system

    International Nuclear Information System (INIS)

    Serezhkina, L.B.; Kuchumova, N.V.; Serezhkin, V.N.

    1993-01-01

    Complexing in K 2 SeO 4 -UO 2 SeO 4 -H 2 O system was studied by the method of isothermal solubility at 25 deg C. Congruently soluble K 2 UO 2 (SeO 4 )2·4H 2 O (1) and incongruently soluble K 2 (UO 2 ) 2 (SeO 4 )3·6H 2 O (2) compounds were revealed in the system. It is shown that (1) and (2) complexes are the representatives of crystallochemical AB 2 2 M 1 and A 2 T 3 3 M 1 groups of uranyl complexes respectively

  11. Complexing in (NH4)2SeO4-UO2SeO4 H2O system

    International Nuclear Information System (INIS)

    Serezhkina, L.B.

    1994-01-01

    Isotherm of solubility in the (NH 4 ) 2 SeO 4 -UO 2 SeO 4 -H 2 O system has been constructed at 25 deg C. (NH 4 ) 2 (UO 2 ) 2 (SeO 4 ) 3 x6H 2 O formation is established for the first time and certain its physicochemical properties are determined. Regularities of complexing in the R 2 Se) 4 -UO 2 SeO 4 -H 2 O systems, where R-univalent cation are under discussion. 6 refs.; 3 tabs

  12. Efficient hybrid evolutionary optimization of interatomic potential models.

    Science.gov (United States)

    Brown, W Michael; Thompson, Aidan P; Schultz, Peter A

    2010-01-14

    The lack of adequately predictive atomistic empirical models precludes meaningful simulations for many materials systems. We describe advances in the development of a hybrid, population based optimization strategy intended for the automated development of material specific interatomic potentials. We compare two strategies for parallel genetic programming and show that the Hierarchical Fair Competition algorithm produces better results in terms of transferability, despite a lower training set accuracy. We evaluate the use of hybrid local search and several fitness models using system energies and/or particle forces. We demonstrate a drastic reduction in the computation time with the use of a correlation-based fitness statistic. We show that the problem difficulty increases with the number of atoms present in the systems used for model development and demonstrate that vectorization can help to address this issue. Finally, we show that with the use of this method, we are able to "rediscover" the exact model for simple known two- and three-body interatomic potentials using only the system energies and particle forces from the supplied atomic configurations.

  13. The MgSeO4-UO2SeO4-H2O system at 25 deg C

    International Nuclear Information System (INIS)

    Serezhkina, L.B.; Serezhkin, V.N.

    1984-01-01

    The method of isothermal solubility at 25 deg C has been used to study MgSeO 4 -UO 2 SeO 4 -H 2 O system. Formation of the new compound Mg 2 (UO 2 ) 3 (SeO 4 ) 5 X32H 2 O, congruently soluble in water is stated. Thermographic and X-ray diffraction investigations of the prepared magnesium selenato-uranylate and products of its dehydration are conducted

  14. Determination of the cationic self-diffusion coefficient in ThO2-5%UO2 nuclear fuel

    International Nuclear Information System (INIS)

    Sabioni, A.C.S.

    1984-01-01

    The cation self-diffusion coefficient for the ThO 2 -5%UO 2 by means of the densification model developed by Assmann and Stehle was determined. The experimental data of the fuel densification, used in the calculations, were obtained from thermal resinter tests. Our result is comparable to previously published values for U and Th diffusion in polycrystalline ThO 2 and (Th, U)O 2 . (Author) [pt

  15. Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

    International Nuclear Information System (INIS)

    Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Pal, S.; Mishra, G.P.

    2012-01-01

    CERMET fuel with either PuO 2 or enriched UO 2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR’s). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO 2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20 wt%UO 2 , U–25 wt%UO 2 and U–30 wt%UO 2 . It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U–UO 2 compositions.

  16. High-pressure high-temperature equations of state of UO2 and ThO2

    Science.gov (United States)

    Chidester, B.; Campbell, A. J.; Fischer, R. A.; Reaman, D. M.; Heinz, D. L.; Prakapenka, V.

    2013-12-01

    The actinide elements uranium and thorium are important from the standpoint of heat production in the deep Earth. However, the host mineral phases and distribution of these elements in the mantle are not well constrained. Here we investigate the crystal chemistry and coordination preferences of these elements in simple oxides. Room-temperature high-pressure equations of state of uranium dioxide (UO2) and thorium dioxide (ThO2) have been reported [1-5], but no in situ high-pressure, high-temperature (high P-T) data are available for these compounds. We present results from a high P-T synchrotron x-ray diffraction study of the equations of state and observed phase relations of UO2 and ThO2. High-pressure, high-temperature in situ X-ray diffraction data were obtained at beamline 13-ID-D of the Advanced Photon Source, and room temperature compression data were obtained at beamline 12.2.2 of the Advanced Light Source. We observed that UO2 exists only in the cubic fluorite structure (space group Fm3m) up to 32 GPa and 2300 K. By fitting these P-V-T data to a Birch-Murnaghan equation of state (EOS), we obtain the thermodynamic parameters K0 = 222 × 3.9 GPa and the thermal contribution to pressure, αK = 0.00254 × 0.00044 GPa/K (V0 = 24.51 cm3/mol and K0' = 5, fixed). At ~46 GPa and up to 2400 K, the cubic structure was found to coexist with a high-pressure phase, which we indexed as the orthorhombic Pnma space group (Z=8). Above this pressure, only the orthorhombic structure was observed up to 61 GPa and 2400 K. The EOS parameters for this phase are V0 = 23.84 × 0.20 cm3/mol, K0 = 187 × 10 GPa and αK = 0.00308 × 0.00042 GPa/K (K0' = 4, fixed). Similarly, ThO2 has the fluorite (Fm3m) structure up to ~23 GPa. The EOS parameters for this phase are K0 = 199 × 10 GPa, K0' = 7.1 × 2.0 and αK = 0.00656 × 0.00092 GPa/K (V0 = 26.38 cm3/mol, fixed). The cubic phase was observed to coexist with an orthorhombic phase (Pnma, Z=4) between 27 and 31 GPa and up to 1900 K. Above

  17. Fitness

    Science.gov (United States)

    ... gov home http://www.girlshealth.gov/ Home Fitness Fitness Want to look and feel your best? Physical ... are? Check out this info: What is physical fitness? top Physical fitness means you can do everyday ...

  18. U3O8 and UO2 obtained from ADU (ammonium diuranate) ultrasonically treated

    International Nuclear Information System (INIS)

    Boero, N.; Sassone, A.; Mendez de Leo, L.; Novara, O.; Ramella, J.

    1996-01-01

    At present, obtention of U 3 O 8 used in the manufacturing of MTR plates nuclear fuels, is performed by hydrolysis of UF6 to obtain uranyl fluoride. Uranyl fluoride is precipitated with ammonium hydroxide to get ammonium polyuranate (ADU). Afterwards ADU is calcinated to U 3 O 8 and mechanical and thermally treated in order to obtain a powder in a determined specification. In the present work, ultrasound has been applied in the stage of precipitation of ADU and for different times in the stage of digestion in order to fasten the stages of ADU filtering and eliminate the U 3 O 8 milling and sieving. Experiences on UO 2 have also been performed. The aspect of ADU changes considerably when they have been ultrasonically treated, its filtering rate is faster and it is easier to dry as it contains less humidity. U 3 O 8 obtained after 800degreeC calcination of treated ADU results in an easy to desagregate powder. Only a soft mechanical treatment is needed to be performed on it before starting thermal treatment at 1400degreeC. After thermal treatment at 1400degreeC treated U 3 O 8 has shown adequate characteristics of size, shape and density (8.2 g/cm 3 ). Regarding UO 2 , the shape of the agglomerates is almost spherical, leading to a free-flowing powder, whose apparent and TAP density showed to be adequate. The characteristics of the different compounds were followed by electron scanning micrographies, X-Rays, specific area measurements and differential thermal analysis. The great advantage of ultrasound appliance is that hard mechanical treatment is avoided in the obtention of U 3 O 8 , saving time and effort. Furthermore, UO 2 proves to be adequate to make pellets, the same precursor could be used in the obtention of both uranium oxides. (author). 5 refs., 6 figs

  19. A model for evolution of oxygen potential and stoichiometry deviation in irradiated UO 2 fuel

    Science.gov (United States)

    Ozrin, V. D.

    2011-12-01

    A model for radial redistribution of oxygen in irradiated UO 2 fuel under conditions of temperature and fission rate gradients has been developed. The oxygen transport in irradiated fuel is considered as a two-scale problem. On the local scale defined by the grain size, irradiated fuel is considered as a multi-phase system including solid solution of fission products in UO 2 matrix, solid precipitates (metal phase, grey phase of complex ternary compounds, the phase of condensed CsI) formed at the gas/solid interface and the gas phase in the intergranular bubbles. Intraganular transport of fission products is described by a set of diffusion equations which are supplemented by the condition of partial thermochemical equilibrium in the subsystem "precipitates & gas phase". The boundary conditions are formulated basing on thermochemical equilibrium on the interface of subsystems "solid solution" and "precipitates & gas phase". Calculation of the partial thermochemical equilibrium yields local values of the oxygen chemical potential and the deviation from fuel stoichiometry. On the global scale defined by the fuel pellet size, spatial variations of the oxygen potential caused by the temperature gradients or the presence of sources/sinks at the pellet boundary determine thermal diffusion fluxes resulting in redistribution of oxygen. The whole set of equations describing local equilibration and the transport in the local and global scales is solved in a self-consistent manner. The model results for radial distribution of oxygen potential of UO 2 calculated for typical reactor operating conditions and the fuel burnup up ˜100 MW d/kg HM are in satisfactory agreement with experimental data.

  20. Transient fission product release within operating UO2 fuel elements during power cycles

    International Nuclear Information System (INIS)

    Lipsett, J.J.; Hunt, C.E.L.; Hastings, I.J.

    1983-05-01

    We have measured short-lived fission product release during shutdown and startup transients for intact UO 2 fuel elements normally operating at linear powers of 45-62 kW/m. The magnitudes of the transient releases are dependent on the steady state operating power and severity of the transient. It is inferred that the inventory of short-lived species at the fuel-to-sheath gap, and thus the accident source term, could be augmented by a series of normal operation transients

  1. Slave Boson Theory of Orbital Differentiation with Crystal Field Effects: Application to UO2

    Science.gov (United States)

    Lanatà, Nicola; Yao, Yongxin; Deng, Xiaoyu; Dobrosavljević, Vladimir; Kotliar, Gabriel

    2017-03-01

    We derive an exact operatorial reformulation of the rotational invariant slave boson method, and we apply it to describe the orbital differentiation in strongly correlated electron systems starting from first principles. The approach enables us to treat strong electron correlations, spin-orbit coupling, and crystal field splittings on the same footing by exploiting the gauge invariance of the mean-field equations. We apply our theory to the archetypical nuclear fuel UO2 and show that the ground state of this system displays a pronounced orbital differentiation within the 5 f manifold, with Mott-localized Γ8 and extended Γ7 electrons.

  2. Slave Boson Theory of Orbital Differentiation with Crystal Field Effects: Application to UO_{2}.

    Science.gov (United States)

    Lanatà, Nicola; Yao, Yongxin; Deng, Xiaoyu; Dobrosavljević, Vladimir; Kotliar, Gabriel

    2017-03-24

    We derive an exact operatorial reformulation of the rotational invariant slave boson method, and we apply it to describe the orbital differentiation in strongly correlated electron systems starting from first principles. The approach enables us to treat strong electron correlations, spin-orbit coupling, and crystal field splittings on the same footing by exploiting the gauge invariance of the mean-field equations. We apply our theory to the archetypical nuclear fuel UO_{2} and show that the ground state of this system displays a pronounced orbital differentiation within the 5f manifold, with Mott-localized Γ_{8} and extended Γ_{7} electrons.

  3. Hydromechanics calculation for micro sphere UO2 fuel produced by sol-gelation method

    International Nuclear Information System (INIS)

    Jin Xin; Liang Tongxiang; Guo Wenli; Zhao Xingyu; Hao Shaochang

    2009-01-01

    Relation between the jet steam velocity in nozzle and height of glue solution level and relation between the jet steam velocity in nozzle and the pressure of glue solution level in pressure kettle are established with Bernoulli equation. The result calculated from this relations shows that the flow of gelation solution is of laminar, the effect of the height of solution level on the the jet steam velocity in nozzle is little and the maximum error for diameter of micro global UO 2 , resulting from the height of solution level, is far more less than the control error. (authors)

  4. Simulation of LOF accidents with directly electrical heated UO2 pins

    International Nuclear Information System (INIS)

    Alexas, A.

    1976-01-01

    The behavior of directly electrical heated UO 2 pins has been investigated under loss of coolant conditions. Two types of hypothetical accidents have been simulated, first, a LOF accident without power excursion (LOF accident) and second, a LOF accident with subsequent power excursion (LOF-TOP accident). A high-speed film shows the sequence of events for two characteristic experiments. In consequence of the high-speed film analysis as well as the metallographical evaluation statements are given in respect to the cladding meltdown process, the fuel melt fraction and the energy input from the beginning of a power transient to the beginning of the molten fuel ejections

  5. Electrochemical system for the control of oxigen atmospheres in UO2 sintering

    International Nuclear Information System (INIS)

    Caneiro, Alberto; Abriata, J.P.

    1980-01-01

    The behaviour of an electrochemical pump and of an oxygen sensor, allowing a precise control of the UO 2 stoichiometry in the preparation and analysis of gaseous mixtures of low oxygen contents is described. The correct functioning of the system can be tested by applying Faraday's law. The oxygen partial pressures can be continuously controlled by the sole varation of the current applied to the electrochemical pump. The partial pressure of the system is within the range between x 10 -1 atm and 10 -27 atm at 800 deg C. This system may be utilized for sintering experiments at a laboratory scale. (M.E.L) [es

  6. Revisiting the diffusion mechanism of helium in UO 2 : A DFT+ U study

    International Nuclear Information System (INIS)

    Liu, X.-Y.; Andersson, D. A.

    2017-01-01

    The understanding of migration properties of helium atoms after their generation through α-decay of actinides in spent nuclear fuels is important for the safety of nuclear fuel storage and disposal. The diffusion of helium in UO 2 is revisited by using the DFT+U simulation methodology employing the “U-ramping” method to address the issue of metastable energy states. A novel diffusion mechanism by helium interstitials, the “asymmetric hop” mechanism, is reported and compared to other diffusion mechanisms including an oxygen vacancy mediated mechanism and available experimental diffusion data. We show that the new mechanism is the dominant one over a wide temperature range.

  7. Optimization of a Wcl6 CVD System to Coat UO2 Powder with Tungsten

    Science.gov (United States)

    Belancik, Grace A.; Barnes, Marvin W.; Mireles, Omar; Hickman, Robert

    2015-01-01

    In order to achieve deep space exploration via Nuclear Thermal Propulsion (NTP), Marshall Space Flight Center (MSFC) is developing W-UO2 CERMET fuel elements, with focus on fabrication, testing, and process optimization. A risk of fuel loss is present due to the CTE mismatch between tungsten and UO2 in the W-60vol%UO2 fuel element, leading to high thermal stresses. This fuel loss can be reduced by coating the spherical UO2 particles with tungsten via H2/WCl6 reduction in a fluidized bed CVD system. Since the latest incarnation of the inverted reactor was completed, various minor modifications to the system design were completed, including an inverted frit sublimer. In order to optimize the parameters to achieve the desired tungsten coating thickness, a number of trials using surrogate HfO2 powder were performed. The furnace temperature was varied between 930 C and 1000degC, and the sublimer temperature was varied between 140 C and 200 C. Each trial lasted 73-82 minutes, with one lasting 205 minutes. A total of 13 trials were performed over the course of three months, two of which were re-coatings of previous trials. The powder samples were weighed before and after coating to roughly determine mass gain, and Scanning Electron Microscope (SEM) data was also obtained. Initial mass results indicated that the rate of layer deposition was lower than desired in all of the trials. SEM confirmed that while a uniform coating was obtained, the average coating thickness was 9.1% of the goal. The two re-coating trials did increase the thickness of the tungsten layer, but only to an average 14.3% of the goal. Therefore, the number of CVD runs required to fully coat one batch of material with the current configuration is not feasible for high production rates. Therefore, the system will be modified to operate with a negative pressure environment. This will allow for better gas mixing and more efficient heating of the substrate material, yielding greater tungsten coating per trial.

  8. Release of tellurium and cesium from UO2 in LWR fuel rods during irradiation

    International Nuclear Information System (INIS)

    Malen, K.A.

    1983-01-01

    In this paper the release of tellurium (Te-132) and cesium (Cs-134 and Cs-137) from UO 2 -fuel is analyzed. The basis for the analysis is the experimental results from the S176 series of experiments performed at Studsvik. It seems that the model developed earlier for release of iodine applies also to tellurium and cesium. This model assumes sweeping up of the species in question by moving grain boundaries and subsequent release through grain boundary porosity. An interesting extra feature is deposition of tellurium at temperatures in the range 1500-2000 K believed to be due to condensation. (author)

  9. Revisiting the diffusion mechanism of helium in UO2: A DFT+U study

    Science.gov (United States)

    Liu, X.-Y.; Andersson, D. A.

    2018-01-01

    The understanding of migration properties of helium atoms after their generation through α-decay of actinides in spent nuclear fuels is important for the safety of nuclear fuel storage and disposal. The diffusion of helium in UO2 is revisited by using the DFT+U simulation methodology employing the "U-ramping" method to address the issue of metastable energy states. A novel diffusion mechanism by helium interstitials, the "asymmetric hop" mechanism, is reported and compared to other diffusion mechanisms including an oxygen vacancy mediated mechanism and available experimental diffusion data. The new mechanism is shown to be the dominant one over a wide temperature range.

  10. An Overview of Current and Past W-UO[2] CERMET Fuel Fabrication Technology

    International Nuclear Information System (INIS)

    Douglas E. Burkes; Daniel M. Wachs; James E. Werner; Steven D. Howe

    2007-01-01

    Studies dating back to the late 1940s performed by a number of different organizations and laboratories have established the major advantages of Nuclear Thermal Propulsion (NTP) systems, particularly for manned missions. A number of NTP projects have been initiated since this time; none have had any sustained fuel development work that appreciably contributed to fuel fabrication or performance data from this era. As interest in these missions returns and previous space nuclear power researchers begin to retire, fuel fabrication technologies must be revisited, so that established technologies can be transferred to young researchers seamlessly and updated, more advanced processes can be employed to develop successful NTP fuels. CERMET fuels, specifically W-UO2, are of particular interest to the next generation NTP plans since these fuels have shown significant advantages over other fuel types, such as relatively high burnup, no significant failures under severe transient conditions, capability of accommodating a large fission product inventory during irradiation and compatibility with flowing hot hydrogen. Examples of previous fabrication routes involved with CERMET fuels include hot isostatic pressing (HIPing) and press and sinter, whereas newer technologies, such as spark plasma sintering, combustion synthesis and microsphere fabrication might be well suited to produce high quality, effective fuel elements. These advanced technologies may address common issues with CERMET fuels, such as grain growth, ductile to brittle transition temperature and UO2 stoichiometry, more effectively than the commonly accepted 'traditional' fabrication routes. Bonding of fuel elements, especially if the fabrication process demands production of smaller element segments, must be investigated. Advanced brazing techniques and compounds are now available that could produce a higher quality bond segment with increased ease in joining. This paper will briefly address the history of CERMET

  11. Safety and licensing of MOX versus UO2 for BWRs and PWRs: Aspects applicable for civilian and weapons grade Pu

    International Nuclear Information System (INIS)

    Goldstein, L.; Malone, J.

    2000-01-01

    This paper reviews the safety and licensing differences between MOX and UO 2 BWR and PWR cores. MOX produced from the normal recycle route and from weapons grade material are considered. Reload quantities of recycle MOX assemblies have been licensed and continue to operate safely in European LWRs. In general, the European MOX assemblies in a reload are 2 . These studies indicated that no important technical or safety related issues have evolved from these studies. The general specifications used by fuel vendors for recycled MOX fuel and core designs are as follows: MOX assemblies should be designed to minimize or eliminate local power peaking mismatches with co-resident and adjacently loaded UO 2 assemblies. Power peaking at the interfaces arises from different neutronic behavior between UO 2 and MOX assemblies. A MOX core (MOX and UO 2 or all-MOX assemblies) should provide cycle energy equivalent to that of an all-UO 2 core. This applies, in particular, to recycle MOX applications. An important consideration when burning weapons grade material is rapid disposition which may not necessarily allow for cycle energy equivalence. The reactivity coefficients, kinetics data, power peaking, and the worth of shutdown systems with MOX fuel and cores must be such to meet the design criteria and fulfill requirements for safe reactor operation. Both recycle and weapons grade plutonium are considered, and positive and negative impacts are given. The paper contrasts MOX versus UO 2 with respect to safety evaluations. The consequences of some transients/accidents are compared for both types of MOX and UO 2 fuel. (author)

  12. Chiroptical luminescence spectra of UO22+ in cubic Na[UO2(CH3COO)3] crystals

    International Nuclear Information System (INIS)

    Moran, D.M.; Metcalf, D.H.; Richardson, F.S.

    1992-01-01

    Steady-state chiroptical luminescence measurements are reported for cubic crystals of Na[UO 2 (CH 3 COO) 3 ]. These crystals belong to the enantiomorphic space group P2 1 3, with four molecules per unit cell, and each UO 2 (CH 3 COO) 3 - coordination unit has a chiral tris-bidentate chelate structure of C 3 symmetry. The UO 2 O' 6 coordination clusters (where O' denotes an acetate oxygen donor atom) also have chiral structures of C 3 point-group symmetry, but they deviate only slightly from an achiral D 3h symmetry. The luminescence observed for Na[UO 2 (CH 3 COO) 3 ] is assigned to transitions that originate from the lowest electronic excited state (II g ) of UO 2 2+ and terminate on the ground electronic state (Σ g + ). At least two types of UO 2 2+ species contribute to this luminescence, but the luminescence spectra can be analyzed in terms of separate majority species (or bulk site) contributions and minority species (or defect site) contributions. The luminescence spectra show zero-phonon origin lines, one-phonon false-origin lines, and vibronic progressions in the symmetric stretching mode (ν s ) of UO 2 2+ . The false-origin lines and the progressions based on these lines are essentially unpolarized. However, the origin lines and their progressions exhibit a very large degree of circular polarization, with emission dissymmetry factors of g em = 1.31 (majority species) and g em = 0.96 (minority species). The circularly polarized luminescence results for Σ g + left-arrow II g emission are compared to circular dichroism results for Σ g + → II g absorption, and the distribution of Σ g + ↔ II g electronic rotary strength among origin and vibronic lines is discussed within the context of vibronic optical activity theory

  13. Studies on the Sintering Behaviour of UO2-Gd2O3 Nuclear Fuel

    International Nuclear Information System (INIS)

    Durazzo, Michelangelo; Gracher Riella, Humberto

    2008-01-01

    The incorporation of gadolinium directly into nuclear power reactor fuel is important from the point of reactivity compensation and adjustment of power distribution enabling thus longer fuel cycles and optimized fuel utilization. The incorporation of Gd 2 O 3 powder directly into the UO 2 powder by dry mechanical blending is the most attractive process because of its simplicity. Nevertheless, processing by this method leads to difficulties while obtaining sintered pellets with the minimum required density. This is due to blockages during the sintering process. There is little information in published literature about the possible mechanism for this blockage and this is restricted to the hypothesis based on formation of a low diffusivity Gd rich (U,Gd)O 2 phase. Experimental evidences indicated the existence of phases in the (U,Gd)O 2 system with structure different from the fluorite type structure of UO 2 . The apparition of these new phases coincides with the lowering of the density after sintering and with the lowering of the interdiffusion coefficient. However, it has been shown experimentally that the sintering blockage phenomena cannot be explained on the basis of the formation of low diffusivity Gd rich (U,Gd)O 2 phases. The work was continued to investigate other possible blocking mechanism. (authors)

  14. Positron annihilation method for α self radiation effect studies in doped actinide UO2 samples

    International Nuclear Information System (INIS)

    Roudil, D.; Vella, F.; Bonnal, M.; Broudic, V.; Barthe, M.F.; Gentils, A.; Moineau, V.; Jolly, L.

    2008-01-01

    Towards disposal problematic, fine understanding of the α aging of UO 2 and (U, Pu)O 2 remains a fundamental challenge for the prediction of the potential increase of the radionuclide source terms with presence of water. The intrinsic evolution of the matrix is closely related to the behavior of radiogenic helium produced by actinide decay. Interactions between helium atoms and vacancy defects are involved in these mechanisms. Positron Annihilation Spectroscopy is also an appropriated method owing to its sensitivity to the vacancy type defects in solid materials. It is a non destructive technique with a remote acquiring data possibility. Because positron implanted in the material is sensitive to the electronic density, the positron lifetime method allows the characterization of the vacancy defects, namely size and concentration. Such equipment has been implemented in the L30 laboratory of the DHA facility in Atalante and will be applied on doped actinides samples, simulating α aging. This article presents, the analytical protocols and validation results on depleted UO 2 samples and highlights the perspectives on (U, Pu)O 2 for the investigation of different stages of self irradiation matrices and helium behavior. (authors)

  15. Irradiation of defected SAP clad UO2 fuel in the X-7 organic loop

    International Nuclear Information System (INIS)

    Robertson, R.F.S.; Cracknell, A.G.; MacDonald, R.D.

    1961-10-01

    This report describes an experiment designed to test the behaviour under irradiation of a UO 2 fuel specimen clad in a defected SAP sheath and cooled by recirculating organic liquid. The specimen containing the defect was irradiated in the X-7 loop in the NRX reactor from the 25th of November until the 13th of December 1960. Up to the 13th of December the behaviour was analogous to that seen with defected UO 2 specimens clad in zircaloy which were irradiated in water loops. Reactor power transients resulted in peaking of gamma ray activities in the loop, but on steady operation these activities tended to fall to a steady state level, Over this period the pressure drop across the fuel increased by a factor of two, the increases occurring after reactor shut downs and start ups. On 13th December the pressure drop increased rapidly, after a reactor shut down and start up, to over five times its original value and the activities in the loop rose to a high level. The specimen was removed and examination showed that the sheath was very badly split and that the volume between the fuel and the sheath was filled with a hard black organic substance. This report gives full details of the irradiation and of the post -irradiation examination. Correlation of the observed phenomenon is attempted and a preliminary assessment of the problems which would be associated with defect fuel in an organic reactor is given. (author)

  16. Determination of organic phosphorus in UO2C2O4·TRPO complex

    International Nuclear Information System (INIS)

    Guo Yifei; Yuan Jianhua; Liang Junfu; Jiao Rongzhou; Liu Xiuqin

    2001-01-01

    Organic phosphorus in UO 2 C 2 O 4 ·TRPO complex is converted to inorganic phosphorous with H 2 SO 4 -HNO 3 -H 2 O 2 wet cinefaction method. In 0.14 mol/L H 2 SO 4 solution containing water soluble poly vinylalcohol as stabilizing agent, the highly sensitive ion-associates are formed by the reaction of basic dye ethyl violet with heteropoly molybdophosphoric blue. Spectrophotometric method is used for determination of phosphorus with these ion-associates. The absorbance maximum is at 620 nm. Determination of phosphorus is not affected with mass ratios R(UO 2 2+ /P) ≤ 1.4 x 10 3 , R(C 2 O 4 2- /P) ≤ 8.8 x 10 2 and R(C 2 O 4 2- /P ≤ 3.6 x 10 4 (one time wet cinefaction must be carried out). In aqueous phase, phosphorus can be directly developed and determined. This method is contrasted with poly vinylalcohol-Rodamine B-heteropoly molybdophosphoric blue, analytical results are in good coincidence. Conversion ratio of phosphorus is 99.8% - 101.1%. The minimum detection limit is 0.02 mg/L. The relative standard deviation is 3%. The recovery ratio is 97% - 103%

  17. Origin of the second length scale found above TN in UO2

    International Nuclear Information System (INIS)

    Watson, G.M.; Gaulin, B.D.; Gibbs, D.; Thurston, T.R.; Simpson, P.J.; Shapiro, S.M.; Lander, G.H.; Matzke, H.; Wang, S.; Dudley, M.

    1996-01-01

    We present the results of x-ray- and neutron-scattering studies of the temperature dependence of the magnetic scattering exhibited by the type-I, triple-Q antiferromagnet UO 2 . Our neutron-scattering results are consistent with those of earlier studies, including the observation of short-ranged magnetic correlations at temperatures near and above T N . However, it is found by x-ray diffraction that a second, longer length scale is induced near T N when the near-surface volume of the sample is mechanically roughened. The longitudinal and transverse widths of the additional scattering increase continuously with increasing temperature above T N , similar to that which has been observed near the magnetic ordering transitions of Ho, Tb, and NpAs and near the tetragonal-to-cubic transitions of various perovskites. Another unusual feature of the present results for UO 2 involves the apparent shift with temperature of the magnetic scattering along the surface normal direction at the (1,1,0) reflection, but not at the (2,1,0) reflection. To our knowledge, this is the first observation of a second length scale near a first-order transition. copyright 1996 The American Physical Society

  18. Behaviour in air at 175-400 degrees C of irradiated UO2 fuel

    International Nuclear Information System (INIS)

    Hastings, I.J.; McCracken, D.

    1984-09-01

    The authors extended their study of irradiated, defected UO 2 fuel elements to 200 and 400 degrees C. At 200 degrees C there was no diametral change, but at 400 degrees C we observed swelling and severe sheath splitting. Neither short-lived fission products, nor Cs-134, Cs-137 or Ru-106 above background, were detected. Maximum Kr-85 release was 4 Bq ( -6 Ci). Discharge time was 2.5 years. UO 2 fragment studies were extended to 400 degrees C. The oxidation process for unirradiated and irradiated fuel up to 300 degrees C was characterized by activation energies of 140 +- 10 and 120 +- 10 kJ/mol, respectively; enhancement of oxidation rate was confirmed in the irradiated samples. There is an apparent reduction of activation energy above about 300 degrees C. Fuel elements with artificial and natural defects showed similar oxidation and dimensional response at 250 degrees C. Behaviour of fuel fragments from the defect area of a naturally-defected element is consistent with that for fragments from intact elements when prior oxidation during the defect period is considered

  19. Kinetic Monte Carlo Potts Model for Simulating a High Burnup Structure in UO2

    International Nuclear Information System (INIS)

    Oh, Jae-Yong; Koo, Yang-Hyun; Lee, Byung-Ho

    2008-01-01

    A Potts model, based on the kinetic Monte Carlo method, was originally developed for magnetic domain evolutions, but it was also proposed as a model for a grain growth in polycrystals due to similarities between Potts domain structures and grain structures. It has modeled various microstructural phenomena such as grain growths, a recrystallization, a sintering, and so on. A high burnup structure (HBS) is observed in the periphery of a high burnup UO 2 fuel. Although its formation mechanism is not clearly understood yet, its characteristics are well recognized: The HBS microstructure consists of very small grains and large bubbles instead of original as-sintered grains. A threshold burnup for the HBS is observed at a local burnup 60-80 Gwd/tM, and the threshold temperature is 1000-1200 .deg. C. Concerning a energy stability, the HBS can be created if the system energy of the HBS is lower than that of the original structure in an irradiated UO 2 . In this paper, a Potts model was implemented for simulating the HBS by calculating system energies, and the simulation results were compared with the HBS characteristics mentioned above

  20. Simulation of High Burnup Structure in UO2 Using Potts Model

    International Nuclear Information System (INIS)

    Oh, Jae Yong; Koo, Yang Hyun; Lee, Byung Ho

    2009-01-01

    The evolution of a high burnup structure (HBS) in a light water reactor (LWR) UO 2 fuel was simulated using the Potts model. A simulation system for the Potts model was defined as a two-dimensional triangular lattice, for which the stored energy was calculated from both the irradiation damage of the UO 2 matrix and the formation of a grain boundary in the newly recrystallized small HBS grains. In the simulation, the evolution probability of the HBS is calculated by the system energy difference between before and after the Monte Carlo simulation step. The simulated local threshold burnup for the HBS formation was 62 MWd/kgU, consistent with the observed threshold burnup range of 60-80 MWd/kgU. The simulation revealed that the HBS was heterogeneously nucleated on the intergranular bubbles in the proximity of the threshold burnup and then additionally on the intragranular bubbles for a burnup above 86 MWd/kgU. In addition, the simulation carried out under a condition of no bubbles indicated that the bubbles played an important role in lowering the threshold burnup for the HBS formation, thereby enabling the HBS to be observed in the burnup range of conventional high burnup fuels

  1. The anodic dissolution of SIMFUEL (UO2) in slightly alkaline sodium carbonate/bicarbonate solutions

    International Nuclear Information System (INIS)

    Keech, P.G.; Goldik, J.S.; Qin, Z.; Shoesmith, D.W.

    2011-01-01

    The corrosion of nuclear fuel under waste disposal conditions is likely to be influenced by the bicarbonate/carbonate content of the groundwater since it increases the solubility of the U VI corrosion product, [UO 2 ] 2+ . As one of the half reactions involved in the corrosion process, the anodic dissolution of SIMFUEL (UO 2 ) has been studied in bicarbonate/carbonate solutions (pH 9.8) using voltammetric and potentiostatic techniques and electrochemical impedance spectroscopy. The reaction proceeds by two consecutive one electron transfer reactions (U IV → U V → U VI ). At low potentials (≤250 mV (vs. SCE) the rate of the first electron transfer reaction is rate determining irrespective of the total carbonate concentration. At potentials >250 mV (vs. SCE) the formation of a U VI O 2 CO 3 surface layer begins to inhibit the dissolution rate and the current becomes independent of potential indicating rate control by the chemical dissolution of this layer.

  2. Modelling of fission gas swelling in the high burnup UO2 fuel

    International Nuclear Information System (INIS)

    Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho

    1999-06-01

    Discharge burnup of the fuel in LWR has been increased to improve the fuel economy, and currently the high burnup fuel of over 70 MWd/kg U-rod avg. is being developed by the fuel vendors worldwide. At high burnup, thermal / mechanical properties of the fuel is known to change and new phenomenon could arise. This report describes the model development on fission gas swelling in high burnup UO 2 fuel. For the low burnup fuel, swelling only by the solid fission products has been considered in the fuel performance analysis. However, at high burnup fuel, swelling by fission gas bubbles can not be neglected anymore. Therefore, fission gas swelling model which can predict bubble swelling of the high burnup UO 2 fuel during the steady-state and the transient conditions in LWR was developed. Based on the bubble growth model, the empirical fission gas swelling model was developed as function of burnup, time and temperature. The model showed that fuel bubble swelling would be proportional to the burnup by the power of 1.157 and to the time by the power of 0.157. Comparison of the model prediction with the measured fission gas swelling data under the various burnup and temperature conditions showed that the model would predict the measured data reasonably well. (author). 20 refs., 8 tabs., 17 figs

  3. Hydrothermal synthesis, structure, and catalytic properties of UO2Sb2O4

    International Nuclear Information System (INIS)

    Sykora, Richard E.; King, Joseph E.; Illies, Andreas J.; Albrecht-Schmitt, Thomas E.

    2004-01-01

    A new uranyl antimonite, UO 2 Sb 2 O 4 (1), has been prepared from the hydrothermal reaction of UO 3 with Sb 2 O 3 and KCl. The structure of 1 consists of neutral two-dimensional ∞ 2 [UO 2 Sb 2 O 4 ] layers. The U(VI) centers are ligated by two trans oxo ligands and four square pyramidal antimonite anions. In addition, the U(VI) also forms long contacts with two additional oxygen atoms that are distorted by 12.7(2) degree sign out of the equatorial plane perpendicular to the uranyl unit. These long interactions are significant owing to evidence supplied by bond valence sum calculations. The two-dimensional layers found in 1 are built from one-dimensional chains formed from edge-sharing UO 6 octahedra that run along the b-axis, and are linked together by [Sb 2 O 4 ] 2- chains. A flow microreactor system has been used to study the catalytic activity of 1, and these results show that it can be used as a catalyst in the conversion of propene and O 2 to acrolein. Crystallographic data: 1, monoclinic, space group C2/m, a=13.490(2) A, b=4.0034(6) A, c=5.1419(8) A, β=104.165(3) deg., Z=2, MoKα, λ=0.71073, R(F)=1.74% for 30 parameters with 365 reflections with I>2σ(I)

  4. Pressure analysis in the fabrication process of TRISO UO2-coated fuel particle

    International Nuclear Information System (INIS)

    Liu Malin; Shao Youlin; Liu Bing

    2012-01-01

    Highlights: ► The pressure signals during the real TRISO UO2-coated fuel particle fabrication process. ► A new relationship about the pressure drop change and the coated fuel particles properties. ► The proposed relationship is validated by experimental results during successive coating. ► A convenient method for monitoring the fluidized state during coating process. - Abstract: The pressure signals in the coating furnace are obtained experimentally from the TRISO UO 2 -coated fuel particle fabrication process. The pressure signals during the coating process are analyzed and a simplified relationship about the pressure drop change due to the coated layer is proposed based on the spouted bed hydrodynamics. The change of pressure drop is found to be consistent with the change of the combination factor about particle density, bed density, particle diameter and static bed height, during the successive coating process of the buffer PyC, IPyC, SiC and OPyC layer. The newly proposed relationship is validated by the experimental values. Based on this relationship, a convenient method is proposed for real-time monitoring the fluidized state of the particles in a high-temperature coating process in the spouted bed. It can be found that the pressure signals analysis is an effective method to monitor the fluidized state on-line in the coating process at high temperature up to 1600 °C.

  5. Gaseous swelling of B4C and UO2 fuel: similarities and differences

    International Nuclear Information System (INIS)

    Evdokimov, I.; Khoruzhii, O.; Kourtchatov, S.; Likhanskii, V.; Matweev, L.

    2001-01-01

    A major factor limiting the resource of control rods (CRs) for WWER-1000 reactors is their radiation damage. Radiation induced embrittlement of the CRs cladding, core swelling and gaseous internal pressure in CRs result in mechanical core-cladding interaction. This work is devoted to the physical analysis of processes that control the structural changes in neutron absorber elements with B 4 C under irradiation in water reactors. Particularly, the analysis of mechanisms of the helium porosity formation in B 4 C is undertaken. In view of the deficiency of experimental data on the subject, a fruitful approach to the problem is a comparative analysis of the swelling mechanisms in B 4 C absorber and UO 2 fuel. Using this similarity a phenomenological model of fission gas behavior in boron carbide is proposed. The model predictions for radial profile of 10 B burnup under influence of thermal and epithermal neutrons are compared with experimental results. The main results show that despite the external similarity of the process of fission gas accumulation in UO 2 and in B 4 C, phenomenology of gaseous swelling is much different for the fuel and the CR core. The reason for that difference is the distinction of physical conditions in irradiated fuel and CR core

  6. 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

  7. Influence of environment on the alteration of the UO2 matrix of spent fuel in storage condition

    International Nuclear Information System (INIS)

    Gaulard, C.

    2012-01-01

    Within the framework of the geological disposal of spent nuclear fuel, research on the long term behavior of spent fuel is undertaken and in particular the study of mechanisms of UO 2 oxidation and dissolution in water-saturated host rock. Under the law program on the sustainable management of radioactive materials and waste of June 28, 2006, France was chose as the reference solution the retreatment of spent fuel and disposal in deep geological repository of vitrified final waste. Nevertheless, studies on a direct disposal of spent fuel will continue for safety. The disposal concept provides for conditioning spent fuel in a steel container whose seal is guaranteed for a period specified in the order of 10,000 years. It is also reasonable to assume that the groundwater comes into contact with the fuel after the deterioration of container and lead to the UO 2 matrix degradation and the release of radionuclides. The oxidation/dissolution of UO 2 has been studied by means electrochemical methods coupled to XPS and ICP-MS measurements.A thermodynamic and bibliographic study of U(VI)/UO 2 (s) system allowed to show the effect of the physical and chemical conditions of the solution on the system, and to show the different mechanisms proposed to describe the oxidation and the dissolution of the uranium dioxide in different media (non-complexing, carbonate and clay). The study of the oxidation/dissolution of UO 2 in acidic and non-complexing media (0.1 mol/L NaCF 3 SO 3 , pH = 3), where UO 2 2+ /UO 2 (s) predominates and the formation of precipitates is limited or even avoided, showed a mechanism with two electrochemical steps and a model characteristic of UO 2 oxidation in acidic non-complexing media. Then, the study in neutral non-complexing media (0.05 mol/L NaCl, pH = 7.5) showed a mechanism with two electrochemical steps and one chemical step (EEC) in which both electrochemical steps are similar to those proposed in acidic media. Finally, a first approach of the UO 2

  8. Interatomic potentials for materials of nuclear interest

    International Nuclear Information System (INIS)

    Fernandez, Julian R.; Monti, Ana M.; Pasianot, Roberto C.; Simonelli, G.

    2007-01-01

    Procedures to develop embedded atom method (EAM) interatomic potentials are described, with foreseeable applications in nuclear materials. Their reliability is shown by evaluating relevant properties. The studied materials are Nb, Zr and U. The first two were then used to develop an inter species potential for the Zr-Nb binary system. In this sense, the Fe-Cu system was also studied starting from Fe and Cu potentials extracted from the literature. (author) [es

  9. Comparison between Experimentally Measured and Thermodynamically Calculated Solubilities of UO2 and ThO2 in KURT Ground Water

    International Nuclear Information System (INIS)

    Kim, Seung Soo; Baik, Min Hoon; Kang, Kwang Cheol; Choi, Jong Won

    2009-01-01

    Solubility of a radionuclide is important for defining the release source term of a radioactive waste in the safety and performance assessments of a radioactive waste repository. When the pH and redox potential of the KURT groundwater were changed by an electrical method, the concentrations of uranium and thorium released from UO 2 (cr) and ThO 2 (cr) at alkali pH (8.1 ∼ 11.4) and reducing potential (Eh -7 mole/L. Unexpectedly, the concentration of tetravalent thorium is slightly higher than that of uranium at pH = 8.1 and Eh= -0.2 V conditions, and this difference may be due to the formation of hydroxide-carbonate complex ions. When UO 2 (s) and UO 2 (am, hyd.), and ThO 2 (s) and Th(OH) 4 (am) were assumed as solubility limiting solid phases, the concentrations of uranium and thorium in the KURT groundwater calculated by the PHREEQC code were comparable to the experimental results. The dominating aqueous species of uranium and thorium were presumed as UO 2 (CO 3 ) 3 4- and Th(OH) 3 CO 3 - at pH = 8.1 ∼ 9.8, and UO 2 (OH) 3 - and Th(OH) 4 (aq) at pH = 11.4

  10. UO2 leaching and radionuclide release modelling under high and low ionic strength solution and oxidation conditions

    International Nuclear Information System (INIS)

    1995-01-01

    In this work, the UO 2 dissolution under oxidizing conditions has been studied in order to compare these results to those obtained with spent fuel. Two different leaching solutions have been used, one with a high ionic strength trying to simulate the conditions expected in a saline repository and the other at low ionic strength much appropriate to granitic environments. In both cases, the dissolution has been studied studied as a function of pH, redox potential, oxidants, complexing agents, particle size as well as the experimental methodology. Results can be summarized as follows: a) The UO 2 dissolution is rather independent on ionic strength. b) Dissolution rates can be explained in general independent on the oxidant as: Log R=3DK [oxidant] Surface solid evolution is very important to understand the dissolution/oxidation mechanism of UO 2 . d) Under oxidizing conditions, the dissolution is H+ and HCO 3 promoted. e) In carbonate medium, both UO 2 and spent fuel dissolution rates are very similar, while in a non-complexing medium, spent fuel dissolution rate is much higher than the UO 2 one. This fact seems to indicate that radiolysis is much important non-complexing media. (Author)

  11. 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

  12. Interpolation effects in tabulated interatomic potentials

    Science.gov (United States)

    Wen, M.; Whalen, S. M.; Elliott, R. S.; Tadmor, E. B.

    2015-10-01

    Empirical interatomic potentials are widely used in atomistic simulations due to their ability to compute the total energy and interatomic forces quickly relative to more accurate quantum calculations. The functional forms in these potentials are sometimes stored in a tabulated format, as a collection of data points (argument-value pairs), and a suitable interpolation (often spline-based) is used to obtain the function value at an arbitrary point. We explore the effect of these interpolations on the potential predictions by calculating the quasi-harmonic thermal expansion and finite-temperature elastic constant of a one-dimensional chain compared with molecular dynamics simulations. Our results show that some predictions are affected by the choice of interpolation regardless of the number of tabulated data points. Our results clearly indicate that the interpolation must be considered part of the potential definition, especially for lattice dynamics properties that depend on higher-order derivatives of the potential. This is facilitated by the Knowledgebase of Interatomic Models (KIM) project, in which both the tabulated data (‘parameterized model’) and the code that interpolates them to compute energy and forces (‘model driver’) are stored and given unique citeable identifiers. We have developed cubic and quintic spline model drivers for pair functional type models (EAM, FS, EMT) and uploaded them to the OpenKIM repository (https://openkim.org).

  13. Traceable Determination of the Absolute Neutron Emission Yields of UO2F2 Working Reference Materials

    International Nuclear Information System (INIS)

    LaFleur, A.M.; Swinhoe, M.T.; Mayo, D.R.; Sapp, B.A.; Croft, S.; Mayer, R.L.

    2013-06-01

    The nuclear material contained in the process equipment of a uranium enrichment plant (referred to as holdup) is an important component of the overall nuclear material inventory for the plant. Accurate quantification and verification of holdup is needed to improve international safeguards and nuclear material accountancy. This is also needed for criticality safety and waste disposition. Passive neutron and gamma-ray nondestructive assay (NDA) methods are used to measure the holdup in process equipment. A key advantage of neutron measurements is that neutrons are highly penetrating and can be measured through thick walled equipment. The dominant source of neutrons in the UO 2 F 2 holdup is from the 19 F(α, n) 22 Na reaction resulting from 234 U alpha decay when uranium is enriched. There is a considerable spread between different historic determinations of the 19 F(α, n) yield from uranium which limits the accuracy of modeling and the calibration of NDA instruments. Furthermore, the compound form and presence of water also significantly affects the neutron emission rate from the holdup. This paper describes a series of experimental measurements performed at Los Alamos National Laboratory (LANL) to determine the absolute neutron emission yield from 10 different UO 2 F 2 working reference materials (WRMs) fabricated at the Portsmouth Gaseous Diffusion Plant (PGDP). The Mini Epithermal Neutron Multiplicity Counter (Mini ENMC) and a NIST certified 252 Cf neutron source were used for these measurements. The high efficiency and short die-away time of the Mini ENMC provides the high measurement precision needed to certify the neutron emission yield. The experiment was designed to achieve sub 1% accuracy in the net counting rate on each item and to provide assurance that important factors such as instrument stability, item placement and background were well understood. The traceable neutron yields measured from the WRMs were used to determine a more accurate neutron yield

  14. UO2-can thermal transfer. Application to the case of the first EL 4 batch of fuel elements

    International Nuclear Information System (INIS)

    Ringot, C.; Faussat, A.

    1964-01-01

    The UO 2 -can thermal transfer is one of the most important factors affecting the operational working of fuel elements. A systematic study of the elements influencing the heat transfer coefficient has been undertaken: in particular the effects of the contact pressure and of the gas filling pressure have been studied. Tests have been carried out using planar and cylindrical geometrical shapes. Using the values obtained and the integral conductivity curve for UO 2 , a calculation has been made of the evolution of the EL-4 fuel element during its life-time, if it is assumed that the fusion gas occupies a free volume of 2,5 per cent at the end of the fuel element and of zero or 2 per cent in the UO 2 . (authors) [fr

  15. Adsorption of UO2+2 by polyethylene adsorbents with amidoxime, carboxyl, and amidoxime/carboxyl group

    International Nuclear Information System (INIS)

    Choi, Seong-Ho; Nho, Young Chang

    2000-01-01

    The polyethylene (PE) adsorbents were prepared by a radiation-induced grafting of acrylonitrile (AN), acrylic acid (AA), and the mixture of AN/AA onto PE film, and by subsequent amidoximation of cyano groups of poly-AN graft chains. With an increase of AA composition in AN/AA monomer mixture, the water uptake of the grafted polyethylene film increased. In AN/AA mixture, the maximum adsorption of UO 2+ 2 was observed in the adsorbent with a ratio of AN/AA (50/50, mol%) in copolymer. The amidoxime, carboxyl, and amidoxime/carboxyl groups onto PE acted as a chelating site for the selected UO 2+ 2 . The complex structure of polyethylene with three functional groups and UO 2+ 2 was confirmed by Fourier Transform Infrared (FTIR) spectroscopy. (author)

  16. Synthesis and crystal structure of [(NH4)(CH3H6)[UO2(SeO3)2

    International Nuclear Information System (INIS)

    Marukhnov, A.V.; Pushkin, D.V.; Serezhkina, L.B.; Peresypkina, E.V.; Virovets, A.V.

    2009-01-01

    Synthesis and X-ray structural analysis of monocrystals (NH 4 )(CN 3 H 6 ) [UO 2 (SeO 3 ) 2 ] (I) have been conducted. The compound is crystallized in triclinic lattice, a=7.0051(2), b=9.4234(3), c=9.5408(3) A, α=88.727(1), β=70.565(1), γ=77.034(1) Deg, sp. gr. P1-bar, Z=2, R=0.0224. Basic structural units of crystals I form chains of the [UO 2 (SeO 3 ) 2 ] 2- composition related to the crystallochemical group AB 2 B 11 (A=UO 2 2+ , B 2 =SeO 3 2- , B 11 =SeO 3 2- ) of uranyl complexes. Uranium-containing complexes is combined in three-dimensional frame by ammonium and guanidinium ions as well as by the systems of hydrogen bonds

  17. The fabrication process of ceramic grade UO2 powder via fluorid system AUC and the treatment on AUC precipitation filtrate

    International Nuclear Information System (INIS)

    Liu Jinhong; Xu Kui; Li Zhiwan; Yi Wei; Tang Yueming; Li Guangrong; Lei Maolin; Cui Chuanjiang

    2006-10-01

    It is described about the technology of fabricating AUC powder by Circum-fluence Precipitation Reactor with Gas (CPRG) from UF 6 hydrolyzed liquid, manufacturing nuclear pure ceramic grade UO 2 powder via fluorid system AUC process with fluidized bed method, recovering U(VI) with ion exchange resin, depositing fluorin in an outflow of effusion wastewater from the ion exchange using calces. The primary control parameters on the fabricating AUC powder is study, it is discussed to character difference of AUC powder between fluorid system and nitrate. Result show that the composing the manufacture AUC powder is invariable by CORG, and that the AUC quality is consistent, and that by decomposition and reduction of AUC and stabilization of UO 2 powder with fluidized bed, through optimum technological parameters, the excellent UO 2 powder is obtained on the quality. (authors)

  18. TRU transmutation using ThO2-UO2 and fully ceramic micro-encapsulated fuels in LWR fuel assemblies

    International Nuclear Information System (INIS)

    Bae, Gonghoon; Hong, Sergi

    2012-01-01

    The objective of this work is to design new LWR fuel assemblies which are able to efficiently destroy TRU (transuranics) nuclide without degradation of safety aspects by using ThO 2 -UO 2 fuel pins and FCM (Fully Ceramic Micro-encapsulated) fuel pins containing TRU fuel particles. Thorium was mixed to UO 2 in order to reduce the generation of plutonium nuclides and to save the uranium resources in the UO 2 pins. Additionally, the use of thorium contributes to the extension of the fuel cycle length. All calculations were performed by using DeCART (Deterministic Core Analysis based on Ray Tracing) code. The results show that the new concept of fuel assembly has the TRU destruction rates of ∼40% and ∼25% per 1200 EFPD (Effective Full Power Day) over the TRU FCM pins and the overall fuel assembly, respectively, without degradation of FTC and MTC

  19. The contribution of thermal radiation to the thermal conductivity of porous UO2

    International Nuclear Information System (INIS)

    Bakker, K.; Kwast, H.; Cordfunke, E.H.P.

    1994-09-01

    The influence of cylindrical, spherical and ellipsoidal inclusions on the overall thermal conductivity was computed with the finite element technique. The results of these calculations were compared with equations that describe the effect of inclusions on the overall thermal conductivity. The analytical equation of Schulz that describes the effect of inclusions on the overall thermal conductivity is in good agreement with the results of the finite element computations. This good agreement shows that among a variety of porosity correction formulas, the equation of Schulz gives the best description of the effect of inclusions on the overall thermal conductivity. This equation and the results of finite element calculations allow us to compute the contribution of radiation to the overall thermal conductivity of UO 2 with oblate ellipsoidal porosity. The present radiation calculations show that Hayes and Peddicord overestimated the contribution of thermal radiation to the thermal conductivity. (orig.)

  20. Development of UO2-30 WT per cent PuO2 fuel for FBTR

    International Nuclear Information System (INIS)

    Majumdar, S.; Kumar, Arun; Kamath, H.S.; Ramachandran, R.; Purushotham, D.S.C.; Roy, P.R.

    1983-01-01

    The specifications on Fast Breeder Reactor (FBTR) fuel pellets have two apparently contradictory requirements viz. (1) formation of homogeneous solid between UO 2 and PuO 2 which can only be achieved by high temperature sintering and (2) density of sintered pellets in the range of 92 ± 1 per cent T.D. which is normally achieved by low temperature sintering. Deactivation of starting powders under CO 2 or addition of volatile pore formers to the powders are the two methods which have been developed for lowering the denity of the pellets without reducing the sintering temperature. Two alternative fabrication routes utilizing these processes for manufacturing of FBTR pellets are described in this report. (author)

  1. Conversion of ammonium uranyl carbonate to UO2 in a fluidized bed

    International Nuclear Information System (INIS)

    Zhao Jun; Qiu Lufu; Zhong Xing; Xu Heqing

    1989-11-01

    The conversion of AUC (Ammonium Uranyl Carbonate) to UO 2 was studied in a fluidized bed of 60 mm inner diameter based on the thermodynamics and kinetics data of decomposition-reduction of AUC. The influence of the reaction temperature, composition of fluidization gas and fluidization velocity on conversion were investigated by using N 2 , Ar and circulation gas (mixing gas of H 2 and CO obtained from the exhaust gas of the decomposition of AUC by catalyst crack-conversion) as the fluidization gas. The throughput is up to the high levels (3.32 kg(wet)/h·L) by using circulation gas or mixing of circulation gas and Ar (< 21%) as the fluidization gas when the reaction temperature exceeds 570 deg C

  2. Burn-up credit applications for UO2 and MOX fuel assemblies in AREVA/COGEMA

    International Nuclear Information System (INIS)

    Toubon, H.; Riffard, C.; Batifol, M.; Pelletier, S.

    2003-01-01

    For the last seven years, AREVA/COGEMA has been implementing the second phase of its burn-up credit program (the incorporation of fission products). Since the early nineties, major actinides have been taken into account in criticality analyses first for reprocessing applications, then for transport and storage of fuel assemblies Next year (2004) COGEMA will take into account the six main fission products (Rh103, Cs133, Nd143, Sm149, Sm152 and Gd155) that make up 50% of the anti-reactivity of all fission products. The experimental program will soon be finished. The new burn-up credit methodology is in progress. After a brief overview of BUC R and D program and COGEMA's application of the BUC, this paper will focus on the new burn-up measurement for UO2 and MOX fuel assemblies. It details the measurement instrumentation and the measurement experiments on MOX fuels performed at La Hague in January 2003. (author)

  3. Behaviour of short-lived iodines in operating UO2 fuel elements

    International Nuclear Information System (INIS)

    Lipsett, J.J.; Hastings, I.J.; Hunt, C.E.L.

    1984-11-01

    Sweep gas experiments have been done to determine the behaviour of short-lived fission products within operating UO 2 fuel elements at linear powers of 45, 54, and 60 KW/m, and to burnups of 70, 80, and 50 MWh/kgU respectively. Although radioiodine transport was not observed directly during normal operation, equilibrium gap inventories for I-131 were deduced from the shutdown decay behaviour of the fission gases. These inventories were a strong function of fuel power and ranged from 10 GBq (0.27 Ci) to 100 GBq (2.7 Ci) over the range tested. We conclude that the iodine inventory was adsorbed onto the fuel and/or sheath surfaces with a volatile fraction of less than 10 -2 and a charcoal-filter-penetrating fraction of less than 2x10 -4

  4. TEM characterization of UO2-Gd2O3 nuclear fuels synthesized by coprecipitation method

    International Nuclear Information System (INIS)

    Soldati, A.; Gana Watkins, I.; Menghini, J.; Prado, M.

    2013-01-01

    We present a micro and nano structural characterization of 4% weight doped Gd 2 O 3 -UO 2 pellet using Transmission Electron Microscopy (TEM). Agglomerate morphology and crystallite sizes were determined using light/dark field and high resolution (HR-TEM) images. Convergent beam Energy Dispersive Spectroscopy (EDS) and Electron Diffraction (ED) were used to evaluate sample composition and homogeneity, even at the nanometer scale. We obtained an average crystallite size of 90±20 nm. Moreover, from TEM-EDS analyses we determined the presence of Gadolinium in all the analyzed crystallites but with 25% variation among their concentrations. These results show the capability of TEM analysis to characterize a nuclear fuel pellet with burnable poisons nano structure and homogeneity.(author)

  5. Pore pressure and swelling in the rim region of LWR high burnup UO2 fuel

    International Nuclear Information System (INIS)

    Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Sohn, Dong-Seong

    2001-01-01

    Based on measured rim characteristics of LWR high burnup UO 2 fuel, the pressure of rim pores and the additional pellet swelling due to rim formation have been modeled. Using the assumption that the number of Xe atoms retained in the rim pores is the same as that which is depleted from the rim matrix, excessive pore pressure is derived as a function of temperature, pellet average burnup and pore radius. The rim pores with small radii are calculated to be highly overpressurized at high burnup. Comparison with experimental data shows that, while the pellet swelling obtained with best-estimate rim width is underpredicted, the one calculated with conservative rim width agrees well with the measured data for rim burnups between 50 and 65 GWd/tU. On the other hand, the measured swelling at 85 GWd/tU is about in-between the two calculations

  6. Behaviour of (Th, U)O2 microspheres under compression tests and pelletization

    International Nuclear Information System (INIS)

    Ferreira, R.A.N.

    1982-12-01

    The interrelation between the behaviour of isolated microspheres in compression tests and the microstructure of sintered pellets obtained with these microspheres, was investigated. Various batches of (Th, 5 w/o U)O 2 microspheres were produced applying the so-called gel process. The production parameters were diversified both as to the composition and to the heat treatments. The resulting products underwent compression tests in an universal tension and compression machine as single microspheres and, as bulk material, were compacted and sintered. The results of the compression tests revealed the existence of two distinct classes of fragmentation behaviour. Each of these classes causes a distinct behaviour during the pelletization, too, resulting in fuel pellets with quite different microstructures. It was evidenced that there is a relationship between these differences in the microstructure and the behaviour of the single microspheres in the compression test. (Author) [pt

  7. Radiolytic syntheses of hollow UO2 nanospheres in Triton X-100-based lyotropic liquid crystals

    International Nuclear Information System (INIS)

    Wang, Yongming; Chen, Qingde; Shen, Xinghai

    2017-01-01

    Hollow nanospheres (φ: 60-80 nm, wall thickness: 10-20 nm), consisted of UO 2 nanoparticles (φ: 3-5 nm), were successfully prepared in a Triton X-100-water (50:50, w/w) hexagonal lyotropic liquid crystal (LLC) by γ-irradiation, where water soluble ammonium uranyl tricarbonate was added as precursor. The product was stable at least up to 300 C. Furthermore, whether the nanospheres were hollow or not, and the wall thickness of the hollow nanospheres could be easily controlled via adjusting dose rate. While in the Triton X-100 based micellar systems, only solid nanospheres were obtained. At last, a possible combination mechanism containing adsorption, aggregation and fracturing processes was proposed.

  8. A new technique to measure fission-product diffusion coefficients in UO2 fuel

    International Nuclear Information System (INIS)

    Hocking, W.H.; Verrall, R.A.; Bushby, S.J.

    1999-01-01

    This paper describes a new out-reactor technique for the measurement of fission-product diffusion rates in UO 2 . The technique accurately simulates in-reactor fission-fragment effects: a thermal diffusion that is due to localized mixing in the fission track, radiation-enhanced diffusion that is due to point-defect creation by fission fragments, and bubble resolution. The technique utilizes heavy-ion accelerators - low energy (40 keV to 1 MeV) for fission-product implantation, high energy (72 MeV) to create fission-fragment damage effects, and secondary ion mass spectrometry (SIMS) for measuring the depth profile of the implanted species. Preliminary results are presented from annealing tests (not in the 72 MeV ion flux) at 1465 deg. C and 1650 deg. C at low and high concentrations of fission products. (author)

  9. Completion of UO2 pellets production and fuel rods load for the RA-8 critical facility

    International Nuclear Information System (INIS)

    Marajofsky, Adolfo; Perez, Lidia E.; Thern, Gerardo G.; Altamirano, Jorge S.; Benitez, Ana M.; Cardenas, Hugo R.; Becerra, Fabian A.; Perez, Aldo E.; Fuente, Mariano de la

    1999-01-01

    The Advanced Fuels Division produced fuel pellets of 235 U with 1.8% and 3.6% enrichment and Zry-4 cladding loads for the RA-8 reactor at Pilcaniyeu Technological Unit. For economical and availability reasons, the powder acquired was initially UO 2 with 3.4% enrichment in 235 U, therefore the 235 U powder with 1.8% enrichment was produced by mechanical mixture. The production of fuel pellets for both enrichments was carried out by cold pressing and sintering processes in reducing atmosphere. The load of Zry-4 claddings was performed manually. The production stages can be divided into setup, qualification and production. This production allows not only to fulfill satisfactorily the new fuel rods supply for the RA-8 reactor but also to count with a new equipment and skilled personnel as well as to meet quality and assurance control methods for future pilot-scale production and even new fuel elements production. (author)

  10. Mechanistic modelling of gaseous fission product behaviour in UO2 fuel by Rtop code

    International Nuclear Information System (INIS)

    Kanukova, V.D.; Khoruzhii, O.V.; Kourtchatov, S.Y.; Likhanskii, V.V.; Matveew, L.V.

    2002-01-01

    The current status of a mechanistic modelling by the RTOP code of the fission product behaviour in polycrystalline UO 2 fuel is described. An outline of the code and implemented physical models is presented. The general approach to code validation is discussed. It is exemplified by the results of validation of the models of fuel oxidation and grain growth. The different models of intragranular and intergranular gas bubble behaviour have been tested and the sensitivity of the code in the framework of these models has been analysed. An analysis of available models of the resolution of grain face bubbles is also presented. The possibilities of the RTOP code are presented through the example of modelling behaviour of WWER fuel over the course of a comparative WWER-PWR experiment performed at Halden and by comparison with Yanagisawa experiments. (author)

  11. Performance of LMFBR fuel pins with (Pu,Th)O/sub 2-x/ and UO2

    International Nuclear Information System (INIS)

    Lawrence, L.A.

    1983-09-01

    The irradiation performance of (Pu,Th)O/sub 2-x/ and UO 2 fueled pins for breeder reactor application were compared to the extensive performance data base for the (U,Pu)O/sub 2-x/ fuel system. Th-Pu and 238 U- 233 U based fuel systems were candidate fuel fertile/fissile isotopic combinations for development of alternatives to the current LMFBR fuel cycle. Initial screening tests were conducted in the EBR-II to obtain comparative performance data because of the limited experience with these fuel systems. In some cases, 235 U was used as a substitute for 233 U because of the difficulties in fabrication of available 233 U due to its high gamma ray emission rate

  12. Effect of technological parameters and microstructure on mechanical strength of UO2 fuel pellets

    International Nuclear Information System (INIS)

    Radford, K.

    1980-01-01

    The effect of various peculiarities of tablet microstructure namely, sammury porosity (tablet density), grain size and pore distribution over sizes on technological parameters, is studied. It is shown that density decrease leads to a fast reduction of UO 2 tablet strength. The maximum effect on strength is produced by pore distribution over sizes, characterized by a median size, and not by the grain size, though a combined effect of those two factors is also observed. The important role of the technology of tablet production manifests itself in the fact that all operations bringing about the increase of pore or grain sizes leads to a reduction of strength. Such factors as powder origin, granule sizes, U 3 O 8 content and the amount of additions do not cause any considerable changes in the strength of tablets. Bend tests under conditions of biaxial loading should be considered as an ideal method of determining fuel tablets strength [ru

  13. Study of secular equilibrium reinstatement on UO2 pellets manufactured by AUC route

    International Nuclear Information System (INIS)

    Carnaval, João Paulo R.; Beltran, Dalton J.M.C.; Oliveira, Carlos A.

    2017-01-01

    The fuel assemblies manufactured by INB for Angra-1 power plant has axial blanket fuel rods which must be inspected due the columns formed by different enrichment pellets. The equipment used for inspection is built with a group of BGO scintillators detectors which measurement principle is based on the absorption of gamma rays emitted from Uranium decay. The commercial grade UF 6 used by INB is stored into cylinders type 30B. The uranium inside these cylinders is in secular equilibrium before the processing. It has been found that the AUC route causes the loss of that equilibrium because the UF 6 is volatilized from the cylinder and the uranium daughters remain in the container. As AUC is converted to powder and pellets, the secular equilibrium is restored through time. The purpose of this work is to present a study of the secular equilibrium reinstatement on UO 2 pellets manufactured by AUC route before being inspected on Rod Scanner. (author)

  14. Coupled thermochemical, isotopic evolution and heat transfer simulations in highly irradiated UO2 nuclear fuel

    Science.gov (United States)

    Piro, M. H. A.; Banfield, J.; Clarno, K. T.; Simunovic, S.; Besmann, T. M.; Lewis, B. J.; Thompson, W. T.

    2013-10-01

    Predictive capabilities for simulating irradiated nuclear fuel behavior are enhanced in the current work by coupling thermochemistry, isotopic evolution and heat transfer. Thermodynamic models that are incorporated into this framework not only predict the departure from stoichiometry of UO2, but also consider dissolved fission and activation products in the fluorite oxide phase, noble metal inclusions, secondary oxides including uranates, zirconates, molybdates and the gas phase. Thermochemical computations utilize the spatial and temporal evolution of the fission and activation product inventory in the pellet, which is typically neglected in nuclear fuel performance simulations. Isotopic computations encompass the depletion, decay and transmutation of more than 2000 isotopes that are calculated at every point in space and time. These computations take into consideration neutron flux depression and the increased production of fissile plutonium near the fuel pellet periphery (i.e., the so-called “rim effect”). Thermochemical and isotopic predictions are in very good agreement with reported experimental measurements of highly irradiated UO2 fuel with an average burnup of 102 GW d t(U)-1. Simulation results demonstrate that predictions are considerably enhanced when coupling thermochemical and isotopic computations in comparison to empirical correlations. Notice: This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.

  15. Inhibition of radiation induced dissolution of UO2 by sulfide - A comparison with the hydrogen effect

    Science.gov (United States)

    Yang, Miao; Barreiro Fidalgo, Alexandre; Sundin, Sara; Jonsson, Mats

    2013-03-01

    In this work we have studied the influence of H2S on radiation induced dissolution of spent nuclear fuel using simple model systems. The reaction between H2O2 and H2S/HS- has been studied experimentally as well as the effect of H2S/HS- on γ-radiation induced dissolution of a UO2 pellet. The experiments clearly show that the reaction of H2O2 and H2S/HS- is fairly rapid and that H2O2 and H2S/HS- stoichiometry is favorable for inhibition. Radiolysis experiments show that H2S/HS- can effectively protect UO2 from oxidative dissolution. The effect depends on sulfide concentration in combination with dose rate. Autoclave experiments were also conducted to study the role of H2S/HS- in the reduction of U(VI) in the presence and absence of H2 and Pd particles in anoxic aqueous solution. The aqueous solutions were pressurized with H2 or N2 and two different concentrations of H2S/HS- were used in the presence and absence of Pd. No catalytic effect of Pd on the U(VI) reduction by H2S/HS- could be found in N2 atmosphere. U(VI) reduction was found to be proportional to H2S/HS- concentration in H2 and N2 atmosphere. It is clearly shown the Pd catalyzed H2 effect is more powerful than the effect of H2S/HS-. H2S/HS- poisoning of the Pd catalyst is not observed under the present conditions.

  16. Rim characteristics and their effects on the thermal conductivity in high burnup UO2 fuel

    International Nuclear Information System (INIS)

    Lee, Byung-Ho; Koo, Yang-Hyun; Sohn, Dong-Seong

    2001-01-01

    Characteristics of high burnup UO 2 fuel such as threshold burnup for the formation of high burnup microstructure (rim), rim average burnup and rim width were estimated and then the thermal conductivity degradation due to the porous rim region was investigated. The threshold burnup for rim formation was estimated as a function of temperature and fission rate using Rest's model. The calculated threshold burnup, which shows a particular dependence on temperature, ranges from 40 to 50 MWd/kgU at typical fuel periphery temperatures of 400 to 600degC. In addition, the rim average burnup and the rim width were obtained by statistical analysis of the data available in open literature. To consider the additional degradation of thermal conductivity in the rim region, a formula for rim porosity was presented with the assumption that rim pores are overpressurized and that all the produced fission gases are retained in the rim pores. To estimate the thermal conductivity in the porous rim using the general correction method applicable to two-phase structure, it was assumed that the rim region consists of pores and fully dense materials composed of UO 2 matrix and solid fission products. Then by combining the general model for two-phase with the rim porosity developed in the present paper and HALDEN's thermal conductivity model, a thermal conductivity model for the porous rim region was developed. The predicted thermal conductivity shows an additional reduction of ∼20% due to the porous rim structure which would cause to increase the fuel temperature of high burnup fuel during steady-state operation and transient irradiation. (author)

  17. Development of a computer code for the analysis of MOX and UO2 fuel

    International Nuclear Information System (INIS)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong

    1997-01-01

    A computer code called COSMOS has been developed for the thermal analysis of MOX and UO 2 fuel rod during steady-state and transient conditions. The main purpose of COSMOS is to calculate the temperature distribution in the fuel and cladding and the fission gas release. Based on a computer code developed for the analysis of UO 2 fuel, following features have been taken into account to analyze the MOX fuel : 1) changes in thermo-mechanical properties such as thermal conductivity and thermal expansion coefficient, 2) change in radial power depression as a function of Pu fissile content, and 3) change in the mechanism of fission gas release resulting form the heterogeneity of microstructure of MOX fuel. In addition, recent experimental findings such as rim effect and thermal conductivity degradation with burnup are taken into account to analyze high burnup fuel. A mechanistic fission gas release model developed based on physical processes is applied to steady-state operation and an empirical model developed based on the amount of fission gas stored at grain boundary is used for transient operation. Another important feature of COSMOS is that it can analyze the fuel segment refabricated from the base irradiated fuel rods. This feature makes it possible to utilize database obtained from international projects such as HALDEN and RISO, many of which were collected from refabricated fuel segments. The capacity of COSMOS has been tested with a number of experimental results from some international fuel irradiation programs. This paper provides a general description of the models contained in COSMOS and some results of comparison between calculation and measurement. (author). 15 refs., 1 tab., 9 figs

  18. Fabrication, characteristics, and in-pile performance of UO2 pellets prepared from dry route powder

    International Nuclear Information System (INIS)

    Chotard, A.; Ledac, A.; Bernardin, M.

    1991-01-01

    The dry route conversion process of UF 6 to sinterable UO 2 powder has been used in France on a large scale for more than 10 years for the fabrication of PWR fuels. Thus, our fabrication and irradiation experience relates to more than 10,000 tons of fuel. As everyone knows, the dry route conversion process only involves gas-gas and gas-solid reactions which present the advantage of producing very little contaminated wastes and no liquid effluents. Powders obtained by this process are characterized by: - a very high purity, - a low specific surface area (around 2 m 2 /g), therefore a high resistance to spontaneous oxidation, - a good compressibility, - a very high sinterability (.98% T.D.), - a very high reproducibility. This powder also shows a high fineness which leads to very homogeneous blends with additives like pore former, U 3 O 8 or Gd 2 O 3 . On the other hand this fineness requires a granulation step which is actually not a disadvantage since it allows to adjust the granulate size to optimize the filling of press dies and so as to guarantee a good stability of the pellet dimensions and density. This pelletizing process leads to pellets characterized by: - a good thermal stability (0.5% T.D. after 34 hours at 1700degC), - no open porosity, - low H 2 content (0,3 ppm), - an homogeneous microstructure (grain size and porosity). Such characteristics mean that the UO 2 pellets from dry route conversion present an excellent in pile behaviour for high burnup up to 58,000 MWd/MtU in commercial plant, with: - low fission gas release, - good dimensional stability (densification, swelling), of which examples and results of PIE are described in the paper. The qualities of the dry route conversion powder and its flexibility of use make it possible to consider adjustment of the pellet characteristics, mainly: density, grain size and pore size distribution for specific uses or performance upgrade. (orig.)

  19. Studies on the kinetics of UO2 dissolution in carbonate-bicarbonate medium using sodium hypochlorite as oxidant

    International Nuclear Information System (INIS)

    Sharma, J.N.; Bhattacharya, K.; Swami, R.G.; Tangri, S.K.; Mukherjee, T.K.

    1996-01-01

    The dissolution of UO 2 in carbonate-bicarbonate solutions containing sodium hypochlorite as an oxidant has been investigated. The effect of temperature, sodium hypochlorite concentration and stirring speed was examined. In the temperature range of 303 to 318 K, the leaching reaction displayed linear kinetics. Apparent activation energy obtained from the differential approach was found to be 57 kJ mol -1 . This relatively high activation energy value indicates a chemically controlled behavior of UO 2 dissolution. The order of reaction with respect to sodium hypochlorite concentration was found to be unity. (author). 18 refs., 6 figs

  20. Complexing in the system Rb2SeO4-UO2SeO4-H2O

    International Nuclear Information System (INIS)

    Kuchumova, N.V.; Shtokova, I.P.; Serezhkina, L.B.; Serezhkin, V.N.

    1989-01-01

    Method of isothermal solubility at 25 deg C is used to study interaction of rubidium and uranyl selenates in aqueous solution. Formation of congruently soluble Rb 2 UO 2 (SeO 4 ) 2 x2H 2 O and Rb 2 (UO 2 ) 2 x(SeO 4 ) 3 x6H 2 O is stated. For the last compound crystallographic characteristics (a=10.668; b=14.935(9); c=13.891(7) A; β=104.94(1); Z=4, sp.gr. P2 1 /c) are determined. Thermal decomposition of a compound results in formation of Rb 2 U 2 O 7

  1. Estimation of optimum experimental parameters in chlorination of UO2 with Cl2 gas and carbon for UCl4

    International Nuclear Information System (INIS)

    Yang, Y.S.; Kang, Y.H.; Lee, H.K.

    1997-01-01

    For the preparation of uranium tetrachloride, the chlorination of UO 2 was carried out and an appropriate reaction system was confirmed. The effects of reaction temperature, time, injection ratio of N 2 gas and appropriate amount of carbon using a reductant on the conversion ratio and volatilization were evaluated. The optimum reaction time and temperature in chlorination of UO 2 for the preparation of UCl 4 were 2 h and 500-700 C, respectively. Also 50% of N 2 gas in chlorine gas proved to be the appropriate injection ratio. (orig.)

  2. Progress in the dry route conversion process of UF-6 to UO-2: new equipment and theoretical approach

    International Nuclear Information System (INIS)

    Perrais, C.; Ablitzer, C.

    1999-01-01

    The dry route conversion process of UF 6 to UO 2 is used on a large scale to produce powder for UO 2 fuel pellets. However, this powder is not very suitable for other kinds of fuels, such as for instance, Mixed Oxide (Mox) fuel. Thus, CEA and COGEMA have developed a programme to study and model the process in order to identify the parameters which lead to a better quality powder. For this purpose, specific equipment was built at the CEA/Cadarache. The first results of experiments and modelling have shown parameters which clearly modify the powder quality. (authors)

  3. Crystal chemistry of uranyl halides containing mixed(UO2)(XmOn)5 bipyramids (X = Cl,Br). Synthesis and crystal structure of Cs2(UO2)(NO3)Cl3

    International Nuclear Information System (INIS)

    Nazarchuk, Evgeny V.; Siidra, Oleg I.; Krivovichev, Sergey V.

    2011-01-01

    Single crystals of Cs 2 (UO 2 )(NO 3 )Cl 3 were prepared by a hydrothermal method at 205 C. The crystal structure has been solved by Direct Methods: monoclinic, P2 1 /n, a = 10.3748(13), b = 9.4683(13), c = 12.5535(16) A β, = 110.280(2) , V = 1156.7(3) A 3 , R 1 = 0.029. In the structure, strongly bonded linear uranyl cations UO 2 2+ are equatorially coordinated by two O and three Cl atoms to form (UO 2 )Cl 3 O 2 pentagonal bipyramids. Each bipyramid shares its O.O edge with an adjacent (NO 3 ) - anion to form finite clusters with the chemical composition [(UO 2 )(NO 3 )Cl 3 ] 2- . The Cs + cations provide three-dimensional connectivity of the structure by forming Cs-O and Cs-Cl contacts to the uranyl nitrate chloride complexes. Related structures of mixed-ligand uranyl halides are compared. (orig.)

  4. Orientational anharmonicity of interatomic interaction in cubic monocrystals

    International Nuclear Information System (INIS)

    Belomestnykh, Vladimir N.; Tesleva, Elena P.

    2010-01-01

    Anharmonicity of interatomic interaction from a position of physical acoustics under the standard conditions is investigated. It is shown that the measure of anharmonicity of interatomic interaction (Grilneisen parameter) is explicitly expressed through velocities of sound. Calculation results of orientation anharmonicity are shown on the example of 116 cubic monocrystals with different lattice structural type and type of chemical bond. Two types of anharmonicity interatomic interaction anisotropy are determined. Keywords: acoustics, orientational anharmonicity, Gruneisen parameter, velocity of sound

  5. INTERATOM experience of cleaning sodium-wetted components

    International Nuclear Information System (INIS)

    Haubold, W.

    1978-01-01

    INTERATOM has been concerned since 1967 with the development, testing, and application of methods to clean sodium wetted components by moist nitrogen, vacuum distillation or alcohol. The activities of INTERATOM in this area have been reported at the IAEA Specialists Meeting on 'Decontamination of Plant Components from Sodium and Radioactivity' in Dounreay, April 9-12, 1973. The three cleaning methods mentioned above are practised at present, too - with minor modifications - by INTERATOM and in the facilities of the SNR project. This note summarizes the experiences of INTERATOM with methods of sodium removal since 1973

  6. Effect of additives in sintering UO2-7wt%Gd2O3 fuel pellets

    International Nuclear Information System (INIS)

    Santos, L.R.; Riella, H.G.

    2009-01-01

    Gadolinium has been used as burnable poison for reactivity control in modern PWRs. The incorporation of Gd 2 O 3 powder directly into the UO 2 powder enables longer fuel cycles and optimized fuel utilization. Nevertheless, processing by this method leads to difficulties while obtaining sintered pellets with the minimum required density. The process for manufacturing UO 2 - Gd 2 O 3 generates scraps that should be reused. The main scraps are green and sintered pellets, which must be calcined to U 3 O 8 to return to the fabrication process. Also, the incorporation of Gd 2 O 3 in UO 2 requires the use of an additive to improve the sintering process, in order to achieve the physical properties specified for the mixed fuel, mainly density and microstructure. This paper describes the effect of the addition of fabrication scraps on the properties of the UO 2 -Gd 2 O 3 fuel. Aluminum hydroxide Al(OH) 3 was also incorporated to the fuel as a sintering aid. The results shown that the use of 2000 ppm of Al(OH) 3 as additive allow to fabricate good pellets with up to 10 wt% of recycled scraps. (author)

  7. Analysis of neutron parameters in light water moderated lattices of ThO2 and UO2 fuel rods

    International Nuclear Information System (INIS)

    Onusic Junior, J.; Oosterkamp, W.J.

    1977-01-01

    A large number of light water moderated lattices of UO 2 and ThO 2 fuel rods were analyzed with the code HAMMER. The purpose of the study was to compare experimental results with computer calculated values. The model employed is described and some modification were introduced in the resonance parameters of Th-232 to increase the agreement with the experimental value [pt

  8. Solid state reactions of UO2.17 and U4O9 with (NH4)2SO4

    International Nuclear Information System (INIS)

    Singh Mudher, K.D.; Keskar, Meera; Jayadevan, N.C.

    1993-01-01

    The present study on the solid state reactions of (NH 4 ) 2 SO 4 with UO 2.17 and U 4 O 9 is to investigate any differences in the nature of oxidation state of U in the U-O system in the O/U range of 2.00 to 2.25. (author). 5 refs., 2 figs., 1 tab

  9. Precipitation of UO2 in sodium carbonate solutions by electrolytic hydrogen and catalyzed by Ni-Raney - Bibliography

    International Nuclear Information System (INIS)

    Pottier, P.

    1958-01-01

    This report proposes abstracts and short versions of a set of documents (studies, patents) dealing with the precipitation of uranium (notably in its oxide form, UO 2 ) in solutions of sodium carbonate. The main objective is to identify the interest of a chemical reduction by electrolytic hydrogen. The author makes a distinction between the most relevant documents and those relatively relevant ones [fr

  10. (Alpha, gamma) irradiation effect on the alteration of high-level radioactive wastes matrices (UO2, hollandite, glass SON68)

    International Nuclear Information System (INIS)

    Suzuki, T.

    2007-06-01

    The aim of this work is to determine the effect of irradiation on the alteration of high level nuclear waste forms matrices. The matrices investigated are UO 2 to simulate the spent fuel, the hollandite for the specific conditioning of Cs, and the inactive glass SON68 representing the nuclear glass R7T7) The alpha irradiation experiments on UO 2 colloids in aqueous carbonate media have enabled to distinguish between the oxidation of UO 2 matrix as initial and dissolution as subsequent step. The simultaneous presence of carbonate and H 2 O 2 (product resulting from water radiolysis) increased the dissolution rate of UO 2 to its maximum value governed by the oxidation rate. ii) The study of hollandite alteration under gamma irradiation confirmed the good retention capacity for Cs and Ba. Gamma irradiation had brought only a little influence on releasing of Cs and Ba in solution. Electronic irradiation had conducted to the amorphization of the hollandite only for a dose 1000 times higher than the auto-induced dose of Ba over millions of years. iii) The experiences of glass irradiation under alpha beam and of helium implantation in the glass SON68 were analyzed by positon annihilation spectroscopy. No effect has been observed on the solid surface for an irradiation dose equal to 1000 years of storage. (author)

  11. Complexes of biuret with Cr(III), Ag(II), Cu(III) mixed ligand and UO2(II)

    International Nuclear Information System (INIS)

    Sanyal, R.M.; Ansari, B.J.; Srivastava, P.C.; Banerjee, B.K.; Chakraburtty

    1979-01-01

    The methods of preparation of some new complexes of Cr(III), Ag(III), Cu(II) mixed ligand and UO 2 (II) and their characterisation from their electronic, infrared and ESR spectra as well as from their magnetic susceptibility values are described. Crystal field parameters of some of the complexes have been calculated. (auth.)

  12. Overall models and experimental database for UO2 and MOX fuel increasing performance

    International Nuclear Information System (INIS)

    Bernard, L.C.; Blanpain, P.

    2001-01-01

    COPERNIC is an advanced fuel rod performance code developed by Framatome. It is based on the TRANSURANUS code that contains a clear and flexible architecture, and offers many modeling possibilities. The main objectives of COPERNIC are to accurately predict steady-state and transient fuel operations at high burnups and to incorporate advanced materials such as the Framatome M5-alloy cladding. An extensive development program was undertaken to benchmark the code to very high burnups and to new M5-alloy cladding data. New models were developed for the M5-alloy cladding and the COPERNIC thermal models were upgraded and improved to extend the predictions to burnups over 100 GWd/tM. Since key phenomena, like fission gas release, are strongly temperature dependent, many other models were upgraded also. The COPERNIC qualification range extends to 67, 55, 53 GWd/tM respectively for UO 2 , UO 2 -Gd 2 O 3 , and MOX fuels with Zircaloy-4 claddings. The range extends to 63 GWd/tM with UO 2 fuel and the advanced M5-alloy cladding. The paper focuses on thermal and fission gas release models, and on MOX fuel modeling. The COPERNIC thermal model consists of several submodels: gap conductance, gap closure, fuel thermal conductivity, radial power profile, and fuel rim. The fuel thermal conductivity and the gap closure models, in particular, have been significantly improved. The model was benchmarked with 3400 fuel centerline temperature data from many French and international programs. There are no measured to predicted statistical biases with respect to linear heat generation rate or burnup. The overall quality of the model is state-of-the-art as the model uncertainty is below 10 %. The fission gas release takes into account athermal and thermally activated mechanisms. The model was adapted to MOX and Gadolinia fuels. For the heterogeneous MOX MIMAS fuels, an effective burnup is used for the incubation threshold. For gadolinia fuels, a scaled temperature effect is used. The

  13. Experimental simulation of irradiation effects on thermomechanical behaviour of UO2 fuel: Impact of solid and gaseous fission products

    International Nuclear Information System (INIS)

    Balland, J.

    2007-12-01

    Predictive simulation of thermomechanical behaviour of nuclear fuel has to take into account irradiation effects. Fission Products (FP) can modify the thermomechanical behaviour of UO 2 . During this thesis, differentiation was made between fission products which create a solid solution with UO 2 and gaseous products, generating pressurized bubbles. SIMFUELS containing gadolinium oxide and pressurized argon bubbles were manufactured, respectively by conventional process and by Gas Pressure Sintering. Brittle and ductile behaviour of UO 2 was investigated, under experimental conditions representative of Pellet-Cladding Interaction (PCI), respectively with 3 points bending tests and compressive creep tests. Investigation of brittle behaviour of UO 2 showed that fracture is mainly controlled by natural defects, like porosities, acting like starting points for cracks propagation. Addition of simulates fission products increase the brittle-to-ductile transition temperature of UO 2 , up to 400-500 C regarding FP in solid solution, and up to 200 C for gaseous products. Fission products although reduce fracture stresses, by a factor between 1.5 and 4, respectively for gas bubbles and solid solutions. Decrease of fracture stress is linked to an increase of microstructural defects due the solid solution and to pressurized bubbles located at grain boundaries. Pellets were tested under compressive solicitation at high temperatures. Experimental results of creep tests are well represented by Norton laws. Creep controlling mechanisms are evidenced by microstructural analysis performed on pellets at different strains. On the basis of calculations made for fuels having the same microstructures than the SIMFUELs, a creep factor is determined. It revealed a strong hardening effect of the solid solution, due to the fact that the added elements anchor the dislocations, whereas pressurized bubbles showed a coupling between hardening and softening effects. (author)

  14. The Influences of Uranium Concentration and Polyvinyl Alcohol on the Quality UO2 Microsphere for Fuel of High Temperature Reactor

    International Nuclear Information System (INIS)

    Damunir; Sukarsono; Bangun-Wasito; Endang Nawangsih

    2000-01-01

    The influences of uranium concentration and PVA on the quality of UO 2 microspheres for fuel of high temperature reactor have been investigated. The UO 2 particles were prepared by gel precipitation using internal gelation process. Uranyl nitrate solution containing uranium of 100 g/l was neutralized using NH 4 OH 1 M. The solution was changed into sol by adding 60 g PVA/l solution while stirred and heated up to 80 o C for 20 minutes. In order to find gels in spherical shape, the sol solution was dropped into 5 M NH 4 OH medium. The formed gels were small spheres, was washed, screened and heated up to 120 o C. After that, the gels were calcined at 800 o C for 4 hours, resulting in U 3 O 8 spheres. The U 3 O 8 particles were reduced using H 2 gas in a N 2 media at 800 o C for 4 hours, yielded in UO 2 spheres. Using a similar procedure, the influence of uranium concentration of 150-250 g/l and PVA 40-80 g/l were studied. The qualities of UO 2 particles were obtained by their physical properties, i.e. density, specific surface area, total volume of pores and pore radius using surface area meter and N 2 gas used as absorbent, and the particle size was observed using optical microscope. The result showed that the changing of uranium and PVA concentrations on the internal gelation affected the density, specific surface area, total volume of pores and pore radius of UO 2 particles. (author)

  15. Contribution to the identification and the evaluation of a doped UO2 fuel with controlled oxygen potential

    International Nuclear Information System (INIS)

    Pennisi, Vanessa

    2015-01-01

    Temperature and oxygen partial pressure (PO 2 ) of nuclear oxide fuels are the main parameters governing both their thermochemical evolution in reactor and the speciation of volatile fission products such as Cs, I or Te. An innovative way to limit the risk of cladding rupture by corrosion under irradiation consists in buffering the oxygen partial pressure of the fuel under operation in a PO 2 domain where the fission gas are harmless towards Zr clad, by using solid redox buffers as additives. Niobium, with its NbO 2 /NbO and Nb 2 O 5 /NbO 2 redox couples has been found to be a promising candidate to this end. A manufacturing process of a buffered UO 2 fuel, doped with niobium has been optimized, in order to fulfill usual specifications (density, microstructure). The experimental study of the UO 2 -NbO x system has shown the existence of a liquid phase between UO 2 and NbO x at 810 C, which was not reported in the literature. The characterization of Nb containing phases present in UO 2 both in solid solution and as precipitates has lead us to propose a solubility thermodynamic model of niobium in UO 2 at 1700 C. An extensive study of the niobium precipitates shows the co-existence in the fuel of NbO 2 and NbO as major phases, together with small amounts of metallic Nb. The coexistence of niobium under two oxidation states inside the fuel is a key element of demonstration of a possible in-situ buffering effect, which is likely to impact some properties of the material that are dependent upon PO 2 , such as densification. These results confirm the promising potential of oxygen buffered fuels as regard to their performance in reactor. (author) [fr

  16. The growth of intra-granular bubbles in post-irradiation annealed UO2 fuel

    International Nuclear Information System (INIS)

    White, R.J.

    2001-01-01

    Post-irradiation examinations of low temperature irradiated UO 2 reveal large numbers of very small intra-granular bubbles, typically of around 1 nm diameter. During high temperature reactor transients these bubbles act as sinks for fission gas atoms and vacancies and can give rise to large volumetric swellings, sometimes of the order of 10%. Under irradiation conditions, the nucleation and growth of these bubbles is determined by a balance between irradiation-induced nucleation, diffusional growth and an irradiation induced re-solution mechanism. This conceptual picture is, however, incomplete because in the absence of irradiation the model predicts that the bubble population present from the pre-irradiation would act as the dominant sink for fission gas atoms resulting in large intra-granular swellings and little or no fission gas release. In practice, large fission gas releases are observed from post-irradiation annealed fuel. A recent series of experiments addressed the issue of fission gas release and swelling in post-irradiation annealed UO 2 originating from Advanced Gas Cooled Reactor (AGR) fuel which had been ramp tested in the Halden Test reactor. Specimens of fuel were subjected to transient heating at ramp rates of 0.5 deg. C/s and 20 deg. C/s to target temperatures between 1600 deg. C and 1900 deg. C. The release of fission gas was monitored during the tests. Subsequently, the fuel was subjected to post-irradiation examination involving detailed Scanning Electron Microscopy (SEM) analysis. Bubble-size distributions were obtained from seventeen specimens, which entailed the measurement of nearly 26,000 intra-granular bubbles. The analysis reveals that the bubble densities remain approximately invariant during the anneals and the bubble-size distributions exhibit long exponential tails in which the largest bubbles are present in concentrations of 10 4 or 10 5 lower than the concentrations of the average sized bubbles. Detailed modelling of the bubble

  17. Effect of PCMI restraint on bubble size distribution in the rim structure of UO2 fuel

    International Nuclear Information System (INIS)

    Oh, Je-Yong; Koo, Yang-Hyun; Cheon, Jin-Sik; Lee, Byung-Ho; Sohn, Dong-Seong

    2005-01-01

    Generally, the bubble size in the rim structure of UO 2 is not dependent on the fuel burnup and the bubble pressure is higher than that in the equilibrium condition. However it was also observed that if the fuel pellet is not restrained, the size of the bubbles in the rim structure could be larger than that in the restraint condition. Although the wide variety of rim bubble sizes and porosities possibly result from an external restrain effect, the quantitative method to analyze the effect of PCMI restraint on bubble distribution in the rim is not available at the moment. In this paper, a method is developed which can be used to analyze the effect of PCMI restraint on the bubble distribution in the rim structure of UO 2 fuel based on the data in the literatures. The total number of Xe atoms in the rim bubbles per unit rim volume could be derived by a summation of the number of Xe atoms of each rim bubble in a unit rim volume. The number of Xe atoms of each rim bubble could be calculated by the Van der Waals equation of state and the pressure expressed by p=σ+C/r, where C is an unknown constant to be determined as a function of the temperature and the burnup. On the other hand, the total number of Xe atoms in the rim bubbles per unit rim volume can also be calculated by Xe depression data. If the fuel pellet is not restrained, the uniform hydrostatic stress, σ is zero. Hence if the data of the fuel disk without a restraint is used, a constant C can be obtained at 823K and a local burnup of 90 GWd/t. Although the local burnup of PCMI restraint case is slightly different from that without PCMI restraint, the value derived above is used for the analysis of PCMI restraint case. The calculated bubble distribution with PCMI restraint was similar to the measured one. Because the effect of PCMI restraint on bubble size increased with the bubble size, the development of a large bubble was suppressed. Hence, the PCMI restraint caused a typical bubble size in the rim and

  18. Study of the compounds M3UO2F5 (M = K, Rb, Cs, NH4) by i.r. absorption and Raman diffusion spectrophotometry

    International Nuclear Information System (INIS)

    Dao, N.Q.; Knidiri, M.

    1976-01-01

    Infrared spectra of the compound M 3 UO 2 F 5 (M = K, Rb, Cs, NH 4 ) were recorded in the region 4000 to 140 cm -1 . Factor group analysis was applied to the interpretation of the spectra. Site and correlation splitting of the internal modes of the UO 2 F 5 3- ion were discussed. (author)

  19. Effect of the UO2 form on the electrochemical reduction rate in a LiCl-Li2O molten salt

    Science.gov (United States)

    Choi, Eun-Young; Kim, Jong-Kook; Im, Hun-Suk; Choi, In-Kyu; Na, Sang-Ho; Lee, Jae Won; Jeong, Sang Mun; Hur, Jin-Mok

    2013-06-01

    Electrochemical reductions of various UO2 forms were investigated in a molten LiCl-Li2O electrolyte. The study focused on the influence of their sizes and densities on the reduction rate by running experiments with eight UO2 forms. They can be classified into porous and dense forms. The porous forms are one granule and four porous pellets with different densities (55%, 60%, 70% and 80%), all of which were fabricated in our laboratory. The dense forms were prepared by crushing dense pellets, and these were similar in size to the porous pellets and the granules. Systematic comparisons of the reduction rate among the tested UO2 forms revealed that a lower density and smaller size of UO2 led to a faster reduction rate. Particularly, attention was focused on the observation that the size of the UO2 provides more dominant effect on the reduction rate than the density.

  20. Influence of oxygen partial pressure on defect concentrations and on oxygen diffusion in UO2+x

    International Nuclear Information System (INIS)

    Pizzi, Elisabetta

    2013-01-01

    The hyper-stoichiometric uranium dioxide (UO 2+x ) is stable over a wide range of temperature and compositions. Such variations of composition and the eventual presence of doping elements or impurities lead to a variation of anionic and electronic defect concentrations. Moreover, many properties of this material are affected by its composition modifications, in particular their atomic transport properties. Firstly we developed a point defect model to evaluate the dependence of the electronic and oxygen defect concentrations upon temperature, equilibrium oxygen partial pressure and impurity content. The physical constants of the model, in particular the equilibrium constants of the defect formation reactions were determined from deviation from stoichiometry and electrical conductivity measurements of literature. This work enabled us to interpret our measures of conductivity, oxygen chemical and self- diffusion coefficients. From a quantitative standpoint, the analysis of our experimental results allows to evaluate the oxygen interstitial diffusion coefficient but also its formation energy. Moreover, an estimate of oxygen di-interstitial formation energy is also provided. Presence of oxygen clusters leads oxygen self- and chemical diffusion to decrease. X-ray Absorption Spectroscopy characterization shows the presence of the same defect in the entire deviation from stoichiometry studied, confirming the approach used to develop the model. (author) [fr

  1. On the role of H2 as an inhibitor of UO2 matrix dissolution

    International Nuclear Information System (INIS)

    Merino, Juan; Gaona, Xavier; Duro, Lara; Bruno, Jordi; Martinez-Esparza, Aurora

    2007-01-01

    The study of spent fuel behaviour under disposal conditions is usually based on conservative approaches assuming oxidising conditions produced by water radiolysis at the fuel/water interface. However, the presence of H 2 from container corrosion can inhibit the dissolution of the UO 2 matrix and enhance its long-term stability. Several studies have confirmed the decrease in dissolution rates when H 2 is present in the system, although the exact mechanisms of interaction have not been fully established. This paper deals with a radiolytic modelling exercise to explore the consequences of the interaction of H 2 with radicals generated by radiolysis in the homogeneous phase. The main conclusion is that in all the modelled cases the presence of H 2 in the system leads to a decrease in matrix dissolution. The extent of the inhibition, and the threshold partial pressure for the inhibition to take place, both depend in a complex way on the chemical composition of the water and the type of radiation present in the system. (authors)

  2. Results of the irradiation of mixed UO2 - PuO2 oxide fuel elements

    International Nuclear Information System (INIS)

    Mikailoff, H.; Mustelier, J.P.; Bloch, J.; Ezran, L.; Hayet, L.

    1966-01-01

    In order to study the behaviour of fuel elements used for the first charge of the reactor Rapsodie, a first batch of eleven needles was irradiated in the reactor EL3 and then examined. These needles (having a shape very similar lo that of the actual needles to be used) were made up of a stack of sintered mixed-oxide pellets: UO 2 containing about 10 per cent of PuO 2 . The density was 85 to 97 per cent of the theoretical, value. The diametral gap between the oxide and the stainless steel can was between 0,06 and 0,27 mm. The specific powers varied from 1230 to 2700 W/cm 3 and the can temperature was between 450 and 630 C. The maximum burn-up attained was 22000 MW days/tonne. Examination of the needles (metrology, radiography and γ-spectrography) revealed certain macroscopic changes, and the evolution of the fuel was shown by micrographic studies. These observations were used, together with flux measurements results, to calculate the temperature distribution inside the fuel. The volume of the fission gas produced was measured in some of the samples; the results are interpreted taking into account the temperature distribution in the oxide and the burn-up attained. Finally a study was made both of the behaviour of a fuel element whose central part was molten during irradiation, and of the effect of sodium which had penetrated into some of the samples following can rupture. (author) [fr

  3. Subcriticality determination of low-enriched UO2 lattices in water by exponential experiment

    International Nuclear Information System (INIS)

    Suzaki, Takenori

    1991-01-01

    To determine the static k (effective neutron multiplication factor) ranging from the critical to an extremely subcritical states, the exponential experiments were performed using various sizes of light-water moderated and reflected low-enriched UO 2 lattice cores. For comparison, the pulsed neutron source experiments were also carried out. In the manner of the Gozani's bracketing method applied to the pulsed source experiment, a formula to obtain k from the measured spatial-decay constant was derived on the basis of diffusion theory. Parameters in the formulas needed to obtain k from the respective experiments were evaluated by 4-group neutron diffusion calculations. The results of the exponential experiments agreed well with those of the pulsed source experiments, the 4-group diffusion calculations and the 137-group Monte Carlo calculations. Therefore, the present data-processing method developed for the exponential experiment was demonstrated to be valid. Besides, through the examination on the parameters used in the data processing, it was found that the dependence of parameter value upon k is weak in the exponential experiment compared with that in the pulsed source experiment. This indicates the superiority of the exponential experiment over the pulsed source experiment for the subcriticality determination of a wide range. (author)

  4. Estimation of pore pressure in the rim region of high burnup UO2 fuel

    International Nuclear Information System (INIS)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong

    1999-01-01

    An attempt has been made to estimate the pore pressure in the rim region of high burnup UO 2 fuel as a function of rim burnup using the measured rim width, average porosity and pore density in the rim region. First, a linear relationship is developed based on measured rim burnup and rim width. Second, fraction of fission gas retained in the grain boundary of rim region is estimated. Third, total pores in the rim is calculated from the measured pore density in the rim region. Finally using the assumption that all the pores in the rim have the same size of 1.2μm, pore pressure is calculated from the equation of state for ideal gas. An estimated pore pressure of about 60 to 80 MPa for the rim burnup of 90 GWd/tU appears to be in reasonable agreement with other value given in a literature that pore pressure at 800 K become 90-210 MPa for pellet average burnup of 80 GWd/tU

  5. Simulation of pore interlinkage in the rim region of high burnup UO2 fuel

    International Nuclear Information System (INIS)

    Koo, Yang Hyun; Oh, Je Yong; Lee, Byung Ho; Cheon, Jin Sik; Joo, Hyung Kook; Sohn, Dong Seong

    2003-01-01

    Threshold porosity above which fission gas release channels would be formed in the rim region of high burnup UO 2 fuel was estimated by the Monte Carlo method and Hoshen-Kopelman algorithm. With the assumption that both rim pore and rim grain can be represented by cube, pore distribution in the rim was simulated 3-dimensionally by the Monte Carlo method according to porosity and pore size distribution. Then, using the Hoshen-Kopelman algorithm, the fraction of open rim pores interlinked to the outer surface of a fuel pellet was derived as a function of rim porosity. The simulation showed that porosity of 24-25% is the threshold above which the number of rim pores forming release channels increases very rapidly. On the other hand, channels would not be formed if the porosity is less than about 23.5%. This is consistent with the observation that, for porosity less than 23.5%, almost no fission gas is released in the rim. However, once the rim porosity reaches beyond 25%, extensive open paths would be developed and considerable fission gas release would start in the rim

  6. Development and research of the modified nanostructured fuel UO2 with improved performance

    International Nuclear Information System (INIS)

    Kurina, I.; Popov, V.; Rumyantsev, V.; Rogov, S.

    2011-01-01

    Activities of SSC RF IPPE aimed at improvement of fast reactor fuels include the research in the following areas: (1) enhancement of thermal conductivity and crack resistance, (2) structure improvement for reduction of fission-product-gas release, and (3) softening the fuel-cladding interaction. The developed technologies allow us to fabricate UO 2 fuel pellets, which demonstrated improved properties as compared to the standard pellets. The improved properties of the pellets have been attained with the help of a nanotechnology (precipitation with no additives and co-precipitation with additives) and a technology based on addition of ammonium-containing reactants to the standard powders. In the developed, so called modified, fuel, the U-U chemical bonding, attributed to metal nanoclusters, and the dominating fraction of nanosized pores have been found. A lower swelling of the modified fuel under irradiation is expected because of a compensation effect of the surface tension on the fission-product-gas pressure. The enhanced thermal conductivity and thermal stability allow the modified fuel to attain a deeper burn-up and to provide more safety during reactor power maneuvering. (authors)

  7. Fluage en pile de l'oxyde mixte uo 2-puo 2

    Science.gov (United States)

    Milet, Claude; Piconi, Corrado

    1983-06-01

    Le fluage en pile et sous compression d'un oxyde mixte UO 2-PuO 2 a été étudié pour les contraintes inférieures ou égales à 26.5 MPa dans un domaine de température s'étendant de 500 à 1100°C. Nos résultats ont montré que la vitesse de fluage est proportionnelle à la contrainte appliquée et aux taux de fission dans le domaine exploré et jusqu'au taux de combustion de 4 at%. D'autre part on a constaté l'existence d'une zone de transition entre un régime athermique et un régime thermiquement activé vers 900°C et on a mesuré des vitesses de fluage du même ordre de grandeur pour un combustible stoechiométrique (U, Pu)O 2.00 et un combustible hypostoechiométrique (U, Pu)O 1.97 à une température voisine de 800°C.

  8. Is UO2HPO4,4H2O a proton conductor

    International Nuclear Information System (INIS)

    Skou, E.; Andersen, I.G.K.; Simonsen, K.E.; Andersen, E.K.

    1983-01-01

    HUP (UO 2 HPO 4 ,4H 2 O) was washed with water until decomposition. The composition was followed by X-ray diffraction. The experiments show that HUP can be washed free of mother liquor without destruction. The washing time necessary is several days. Washing with water for several weeks converts HUP to a new phase. The ac-conductivity of discs of HUP washed free of mother liquor was 1.3x10 -4 ohm -1 cm -1 , an order of magnitude lower than values reported in literature for discs of unwashed HUP. The ac-conductivity of a cell containing the washing solution after it was equilibrated with HUP was measured. Replacement of some of the liquid with HUP taken from the same washing experiment diminished the conductivity of the cell. The conductivity of HUP is therefore lower than the conductivity of the liquid (1.3x10 -3 ohm -1 cm -1 ). The authors conclude that the high conductivities of HUP reported in literature are caused by adhering liquid, and do not reflect intrinsic conductivity of the material. (Auth.)

  9. Quality assurance and control in the manufacture of metalclad UO2 reactor fuels

    International Nuclear Information System (INIS)

    1976-01-01

    The International Atomic Energy Agency has carried out a programme since its earliest days that includes the collection and dissemination of information on nuclear fuels. Since the 1960 symposium on Fuel Element Fabrication with Special Emphasis on Cladding Materials there has been an average of one meeting a year reviewing some aspect of fuel fabrication technology. A recent meeting dealing with the fabrication of UO 2 fuels was the Study Group on the Facilities and Technology needed for Nuclear Fuel Manufacture, held in Grenoble in 1972 (Rep. IAEA-158). After that meeting it became apparent that the quality of fuel production was an important aspect that had received inadequate coverage so far, and the Panel on Quality Assurance and Control in Nuclear Fuel Manufacture was convened by the Agency in Vienna in November 1974. In the working papers and discussions at the Panel meeting the viewpoints of different countries and of various interested parties, such as manufacturers, reactor operators and government authorities, were presented

  10. Molecular dynamics studies on the structural effects of displacement cascades in UO2 matrix

    International Nuclear Information System (INIS)

    Brutzel, L. Van; Rarivomanantsoa, M.; Ghaleb, D.

    2004-01-01

    A set of molecular dynamics simulations have been carried out in order to study, at the atomic scale, the ballistic damages undergo by the UO 2 matrix. The morphologies of the displacement cascades simulations initiated by an uranium atoms with a Primary Knout on Atom (PKA) energy ranges from 1 keV to 20 keV are analysed. In agreement with all the experimental results no amorphization has been found even at small scales. For the cascade initiated with a PKA energy of 20 keV several sub-cascade branches appear in many directions from the cascade core. It seems that these sub-cascades arise from a quasi channeling of uranium atoms in specific direction over long distances. However, in average the atoms are displaced no more than 2 to 3 crystallographic sites. The evolution of the Frenkel pairs with the initial energy of the PKA exhibits a power law behavior with an exponent close to 0.9 showing a discrepancy with the linear NRT law. No significant clustering of local defects such as vacancies and interstitials have been found, nevertheless vacancies are preferentially created near the core of the cascade whereas the atoms in interstitial positions are mainly located at the periphery of the sub-cascade branches. (authors)

  11. The Width of High Burnup Structure in LWR UO2 Fuel

    International Nuclear Information System (INIS)

    Koo, Yang-Hyun; Lee, Byung-Ho; Oh, Jae-Yong; Sohn, Dong-Seong

    2007-01-01

    The measured data available in the open literature on the width of high burnup structure (HBS) in LWR UO 2 fuel were analyzed in terms of pellet average burnup, enrichment, and grain size. Dependence of the HBS width on pellet average burnup was shown to be divided into three regions; while the HBS width is governed by accumulation of fission damage (i.e., burnup) for burnup below 60 GWd/tU, it seems to be restricted to some limiting value of around 1.5 mm for burnup above 75 GWd/tU due to high temperature which might have caused extensive annealing of irradiation damage. As for intermediate burnup between 60 and 75 GWd/tU, although temperature would not have been so high as to induce extensive annealing, the microstructural damage could have been partly annealed, resulting in the reduction of the HBS width. It was found that both enrichment and grain size also affects the HBS width. However, as long as the pellet average burnup is lower than about 75 GWd/tU, the effect does not appear to be significant for the enrichment and grain size that are typically used in current LWR fuel. (authors)

  12. Study on Reactor Physics Characteristic of the PWR Core Using UO2

    International Nuclear Information System (INIS)

    Tukiran Surbakti

    2009-01-01

    Study on reactor physics characteristic of the PWR core using UO 2 fuel it is necessary to be done to know the characteristic of geometry, condition and configuration of pin cell in the fuel assembly Because the geometry, configuration and condition of the pin cell in fuel core determine the loading strategy of in-core fuel management Calculation of k e ff is a part of the neutronic core parameter calculation to know the reactor physics characteristic. Generally, core calculation is done using computer code starts from modelling one unit fuel lattice cell, fuel assembly, reflector, irradiation facility and until core reactor. In this research, the modelling of pin cell and fuel assembly of the PWR 17 ×17 is done homogeneously. Calculation of the k-eff is done with variation of the fuel volume fraction, fuel pin diameter, fuel enrichment. The calculation is using by NITAWL and CENTRM, and then the results will be compared to KENOVI code. The result showed that the value of k e ff for pin cell and fuel assembly PWR 17 ×17 is not different significantly with homogenous and heterogenous models. The results for fuel volume fraction of 0.5; rod pitch 1.26 cm and fuel pin diameter of 9.6 mm is critical with burn up of 35,0 GWd/t. The modeling and calculation method accurately is needed to calculation the core physic parameter, but sometimes, it is needed along time to calculate one model. (author)

  13. Studies on some mixed ligand complexes of UO2(VI) in solution

    International Nuclear Information System (INIS)

    Katkar, V.S.; Munshi, K.N.

    1987-01-01

    Potentiometric studies on some mixed ligand complexes of UO 2 (VI) with ethylenediamine N,N' diacetic acid as primary ligand and succinic acid, malic acid, maleic acid, fumaric acid, malonic acid, adipic acid, itaconic acid, phthalic acid and mandelic acid as secondary ligands have been carried out employing modified form of Irving-Rossotti's pH titration technique. The study was carried out at three different temperatures, viz. 5deg, 25deg and 45degC and at fixed ionic strength of 0.2M KNO 3 . Thermodynamic parameters, e.g. ΔG, ΔH and ΔS have been evaluated and further ΔG and ΔH values have been separated into their electrostatic components, ΔGe and ΔHe and cratic components, ΔGsub(c) and ΔHsub(c). The effect of change in dielectric constant and change in ionic strength of the medium have also been investigated. The sequence of stability constants has been correlated with the properties of secondary ligands. (author). 21 refs., 5 tabs

  14. Ion chromatographic determination of fluoride and chloride in UO2 using microbore anion exchange columns

    International Nuclear Information System (INIS)

    Kelkar, Anoop; Meena, D.L.; Das, D.K.; Behere, P.G.; Mohd Afzal

    2015-01-01

    Chemical characterization of nuclear fuels is required to ensure that nuclear fuel meets the technical specifications of the fuel. Trace non- metallic impurities like Cl and F is important as they affect clad corrosion. Their effect is more severe in presence of moisture. Chlorine and Fluorine is routinely analysed by ion selective electrode or conventional ion chromatography after pyrohydrolyzing the sample in moist O 2 atmosphere at 950°. Both the technique generates large quantity of liquid waste. Generally 1 ml/min flow rate required for the separation of F - and Cl - in conventional ion-chromatographic separation of F - and Cl - on 4.6- 4.0 mm id analytical column. The waste produced per sample injection is ∼ 30-40 ml with suppressed conductivity detection in ion chromatography. There is a need to reduce this analytical waste in analyzing the radioactive samples for the determination of F - and Cl - . Waste generation could be effectively reduced by using microbore anion exchange analytical column. Present paper describe the use of Metrosep A Supp 16 - 100/2.0 column with Na 2 CO 3 +NaOH mobile phase for the determination of F - and Cl - in UO 2 samples using suppressed conductivity detection

  15. Application of the Cold Crucible for Melting of UO2/ZrO2 Mixtures

    International Nuclear Information System (INIS)

    Hong, S.W.; Min, B.T.; Shin, Y.S.; Park, I.K.; Kim, J.H.; Song, J.H.; Kim, H.D.

    2002-01-01

    The melting and discharge technique of UO 2 /ZrO 2 mixtures using the cold crucible melting method that does not need a separate crucible such as tungsten one with high melting point is developed and applied to the KAERI FCI test called TROI. To discharge the melt from a cold crucible into a fuel-coolant interaction chamber after melting, a plug is specially designed using the concept for electro-magnetic field characteristics so as to as thin as possible the crust that is formed between the melt and plug. Its function keeps the melt in the crucible during melting period and provides the melt discharge path. About 8.5 kg melt is discharged from the cold crucible to the melt-water interaction chamber through the punched hole with 8 cm in diameter. The melt temperature is also measured and analyzed from observation of the melt surface. The power balance using the operating parameters such as current, voltage and coupling factor of R.F generator is analyzed. (authors)

  16. Comparative Studies on UO2 Fueled HTTR Several Nuclear Data Libraries

    Science.gov (United States)

    Hidayati, Anni N.; Prastyo, Puguh A.; Waris, Abdul; Irwanto, Dwi

    2017-07-01

    HTTR (High Temperature Engineering Test Reactor) is one of Generation IV nuclear reactors that has been developed by JAERI (former name of JAEA, JAPAN). HTTR uses graphite moderator, helium gas coolant with UO2 fuel and outlet coolant temperature of 900°C or higher than that. Several studies regarding HTTR have been performed by employing JENDL 3.2 nuclear data libraries. In this paper, comparative evaluation of HTTR with several nuclear data libraries (JENDL 3.3, JENDL 4.0, and JEF 3.1) have been conducted.. The 3-D calculation was performed by using CITATION module of SRAC 2006 code. The result shows some differences between those nuclear data libraries result. K-eff or core effective multiplication factor results are about 1.17, 1,18 and 1,19 (JENDL 3.3, JENDL 4.0, and JEF 3.1) at Begin of Life, also at the End of Life (after two years operation) are 1.16, 1.17 and 1.17 for each nuclear data libraries. There are some different result of K-eff but for neutron spectra results, those nuclear data libraries show the same result.

  17. Factors governing microstructure development of Cr2O3-doped UO2 during sintering

    International Nuclear Information System (INIS)

    Bourgeois, L.; Dehaudt, Ph.; Lemaignan, C.; Hammou, A.

    2001-01-01

    Sintering and grain growth of compacted uranium dioxide powder pellets doped with Cr 2 O 3 were investigated at constant heating rates ranging from 75 to 500 K h -1 . The influence of parameters such as the oxygen potential of the sintering atmosphere and pellet green density on the final microstructure was studied. Dilatometric analysis and monitoring of microstructural development revealed a phenomenon of abnormal grain growth promoting densification. The existence of a eutectic between Cr and Cr 2 O 3 is also discussed. Grain growth does not appear to be widely affected by small differences in residual porosity, which is a function of green density, so that it is possible to propose a solubility limit for Cr 2 O 3 in stoichiometric UO 2 at 1700 deg. C. Examination of microstructural changes during annealing, with or without pore formers, showed the existence of limiting grain sizes for doped samples above the solubility limit. Lastly, experimental sintering conditions need to be checked in order to obtain reproducible results [fr

  18. Atomistic simulations of void migration under thermal gradient in UO2

    International Nuclear Information System (INIS)

    Desai, Tapan G.; Millett, Paul; Tonks, Michael; Wolf, Dieter

    2010-01-01

    It is well known that within a few hours after startup of a nuclear reactor, the temperature gradient within a fuel element causes migration of voids/bubbles radially inwards to form a central hole. To understand the atomic processes that control this migration of voids, we performed molecular dynamics (MD) simulations on single crystal UO 2 with voids of diameter 2.2 nm. An external temperature gradient was applied across the simulation cell. At the end of the simulation run, it was observed that the voids had moved towards the hot end of the simulation cell. The void migration velocity obtained from the simulations was compared with the available phenomenological equations for void migration due to different transport mechanisms. Surface diffusion of the slowest moving specie, i.e. uranium, was found to be the dominant mechanism for void migration. The contribution from lattice diffusion and the thermal stress gradient to the void migration was analyzed and found to be negligible. By extrapolation, a crossover from the surface-diffusion-controlled mechanism to the lattice-diffusion-controlled mechanism was found to occur for voids with sizes in the μm range.

  19. Intergranular fracture in UO2: derivation of traction-separation law from atomistic simulations

    Energy Technology Data Exchange (ETDEWEB)

    Yongfeng Zhang; Paul C Millett; Michael R Tonks; Xian-Ming Bai; S Bulent Biner

    2013-10-01

    In this study, the intergranular fracture behavior of UO2 was studied by molecular dynamics simulations using the Basak potential. In addition, the constitutive traction-separation law was derived from atomistic data using the cohesive-zone model. In the simulations a bicrystal model with the (100) symmetric tilt E5 grain boundaries was utilized. Uniaxial tension along the grain boundary normal was applied to simulate Mode-I fracture. The fracture was observed to propagate along the grain boundary by micro-pore nucleation and coalescence, giving an overall intergranular fracture behavior. Phase transformations from the Fluorite to the Rutile and Scrutinyite phases were identified at the propagating crack tips. These new phases are metastable and they transformed back to the Fluorite phase at the wake of crack tips as the local stress concentration was relieved by complete cracking. Such transient behavior observed at atomistic scale was found to substantially increase the energy release rate for fracture. Insertion of Xe gas into the initial notch showed minor effect on the overall fracture behavior.

  20. Development of ultrasonic technique for measure of porosity of UO2 pellets

    International Nuclear Information System (INIS)

    Baroni, Douglas Brandao

    2008-01-01

    The characterization of nuclear fuel is of great importance to guarantee the efficiency and even the safety in the power stations. At present, the techniques used implicate elevated costs with equipment, materials and installations of radiological protection. Besides, because of being destructive techniques, they impose that the checking of the characteristics of this material is done by sampling. In this work a not destructive technique was developed for measures of porosity in ceramic materials with efficiency and precision. The objective of this work is to this technique will be able to be used in laboratory practice for measures in UO 2 pellets, so it would become viable the inspection of up to 100% of the nuclear fuel, guaranteeing bigger control of the characteristics of the used material, turning in increasing safety, efficiency and economy. The innovation of the technique is due to the fact of analysing the specter of frequency of the ultrasonic wrist, and not his time of course in the material, frequently used. In this work 40 ceramic pellets of alumina were used with values of porosity between 5,09% and 37,30%. A system of recognition of signs using artificial neural networks made possible to distinguish pellets with differences of porosity of 0,04%. It was observed that this technique can be used for several others aims, for example, in the determination of the void fraction in regimen of two-phase flow, what is very important to guarantee the efficiency and safety of nuclear reactors. (author)

  1. Sintering kinetics in (Th, 5%U)O2 in the initial stage

    International Nuclear Information System (INIS)

    Ferraz, W.B.; Cardoso, P.E.; Lameiras, F.S.

    1990-01-01

    The initial sintering kinetics of (Th,5%U)O 2 pellets, in air and Ar-4%H 2 atmosphere in the temperature range 950-1175 0 C, was examined by dilatometric analysis. It was observed that the sintering in air occurs at a lower temperature. Based on Johnson's model, the volume diffusion and the grain boundary diffusion were assumed to cause the linear shrinkage. From the general equation of this model Δl/l=kt n the values n approx. 0,49 and n≥ 0,33, for sintering in air and Ar-4%H 2 atmosphere respectively, were observed. These results revealed that the initial sintering kinetics in air is controlled by volume diffusion and in Ar-4%H 2 atmosphere is weakly predominated by the grain boundary diffusion. The estimated volumetric diffusion coefficient for the sintering in air is about three orders of magnitude greater than in Ar-4%H 2 atmosphere, indicating a strong influence of the stoichiometrie on the sintering kinetics. The grain boundary diffusion coefficient appears to be independent of the sintering atmosphere. (author) [pt

  2. Analysis of burnup and isotopic compositions of BWR 9 x 9 UO2 fuel assemblies

    International Nuclear Information System (INIS)

    Suzuki, M.; Yamamoto, T.; Ando, Y.; Nakajima, T.

    2012-01-01

    In order to extend isotopic composition data focusing on fission product nuclides, measurements are progressing using facilities of JAEA for five samples taken from high burnup BWR 9 x 9 UO 2 fuel assemblies. Neutronics analysis with an infinite assembly model was applied to the preliminary measurement data using a continuous-energy Monte Carlo burnup calculation code MVP-BURN with nuclear libraries based on JENDL-3.3 and JENDL-4.0. The burnups of the samples were determined to be 28.0, 39.3, 56.6, 68.1, and 64.0 GWd/t by the Nd-148 method. They were compared with those calculated using node-average irradiation histories of power and in-channel void fractions which were taken from the plant data. The comparison results showed that the deviations of the calculated burnups from the measurements were -4 to 3%. It was confirmed that adopting the nuclear data library based on JENDL-4.0 reduced the deviations of the calculated isotopic compositions from the measurements for 238 Pu, 144 Nd, 145 Nd, 146 Nd, 148 Nd, 134 Cs, 154 Eu, 152 Sm, 154 Gd, and 157 Gd. On the other hand, the effect of the revision in the nuclear. data library on the neutronics analysis was not significant for major U and Pu isotopes. (authors)

  3. Design, Fabrication, and Testing of an External-Fuel [UO2] Full-Length Thermionic Converter

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred; Raab, B; Giorgio, F.

    1971-09-01

    The development of a double-ended full-core-length external-fuel converter, a prototypical fuel module for a 200- to 300-ekw thermionic reactor, is described. The converter design is based on a revolver-shaped tungsten emitter body, with six peripheral fuel chambers loaded with enriched UO2 pellets. The columbium collector is water-cooled through a sub-atmospheric adjustable-pressure helium gap. The converter employs graded metal-ceramic seals, and its double-ended construction is made possible by bellows to compensate for differential axial expansion. Fission gases are vented from the fuel chambers and collected in an accumulator designed for continuous monitoring of the pressure buildup. Component fabrication, assembly sequence, and joining methods are described; also the test procedures, and the converter load control. All tests are performed in vacuum. During inpile testing, the fuel is triply contained, with thermal insulation between the secondary and tertiary containers. Before insertion inpile, the fully fueled converter is qualification-tested by rf-induction heating using a specially developed high vacuum rf-feedthrough.

  4. Phonon optimized interatomic potential for aluminum

    Science.gov (United States)

    Muraleedharan, Murali Gopal; Rohskopf, Andrew; Yang, Vigor; Henry, Asegun

    2017-12-01

    We address the problem of generating a phonon optimized interatomic potential (POP) for aluminum. The POP methodology, which has already been shown to work for semiconductors such as silicon and germanium, uses an evolutionary strategy based on a genetic algorithm (GA) to optimize the free parameters in an empirical interatomic potential (EIP). For aluminum, we used the Vashishta functional form. The training data set was generated ab initio, consisting of forces, energy vs. volume, stresses, and harmonic and cubic force constants obtained from density functional theory (DFT) calculations. Existing potentials for aluminum, such as the embedded atom method (EAM) and charge-optimized many-body (COMB3) potential, show larger errors when the EIP forces are compared with those predicted by DFT, and thus they are not particularly well suited for reproducing phonon properties. Using a comprehensive Vashishta functional form, which involves short and long-ranged interactions, as well as three-body terms, we were able to better capture interactions that reproduce phonon properties accurately. Furthermore, the Vashishta potential is flexible enough to be extended to Al2O3 and the interface between Al-Al2O3, which is technologically important for combustion of solid Al nano powders. The POP developed here is tested for accuracy by comparing phonon thermal conductivity accumulation plots, density of states, and dispersion relations with DFT results. It is shown to perform well in molecular dynamics (MD) simulations as well, where the phonon thermal conductivity is calculated via the Green-Kubo relation. The results are within 10% of the values obtained by solving the Boltzmann transport equation (BTE), employing Fermi's Golden Rule to predict the phonon-phonon relaxation times.

  5. Phonon optimized interatomic potential for aluminum

    Directory of Open Access Journals (Sweden)

    Murali Gopal Muraleedharan

    2017-12-01

    Full Text Available We address the problem of generating a phonon optimized interatomic potential (POP for aluminum. The POP methodology, which has already been shown to work for semiconductors such as silicon and germanium, uses an evolutionary strategy based on a genetic algorithm (GA to optimize the free parameters in an empirical interatomic potential (EIP. For aluminum, we used the Vashishta functional form. The training data set was generated ab initio, consisting of forces, energy vs. volume, stresses, and harmonic and cubic force constants obtained from density functional theory (DFT calculations. Existing potentials for aluminum, such as the embedded atom method (EAM and charge-optimized many-body (COMB3 potential, show larger errors when the EIP forces are compared with those predicted by DFT, and thus they are not particularly well suited for reproducing phonon properties. Using a comprehensive Vashishta functional form, which involves short and long-ranged interactions, as well as three-body terms, we were able to better capture interactions that reproduce phonon properties accurately. Furthermore, the Vashishta potential is flexible enough to be extended to Al2O3 and the interface between Al-Al2O3, which is technologically important for combustion of solid Al nano powders. The POP developed here is tested for accuracy by comparing phonon thermal conductivity accumulation plots, density of states, and dispersion relations with DFT results. It is shown to perform well in molecular dynamics (MD simulations as well, where the phonon thermal conductivity is calculated via the Green-Kubo relation. The results are within 10% of the values obtained by solving the Boltzmann transport equation (BTE, employing Fermi’s Golden Rule to predict the phonon-phonon relaxation times.

  6. Developing a second nearest-neighbor modified embedded atom method interatomic potential for lithium

    International Nuclear Information System (INIS)

    Cui, Zhiwei; Gao, Feng; Qu, Jianmin; Cui, Zhihua

    2012-01-01

    This paper reports the development of a second nearest-neighbor modified embedded atom method (2NN MEAM) interatomic potential for lithium (Li). The 2NN MEAM potential contains 14 adjustable parameters. For a given set of these parameters, a number of physical properties of Li were predicted by molecular dynamics (MD) simulations. By fitting these MD predictions to their corresponding values from either experimental measurements or ab initio simulations, these adjustable parameters in the potential were optimized to yield an accurate and robust interatomic potential. The parameter optimization was carried out using the particle swarm optimization technique. Finally, the newly developed potential was validated by calculating a wide range of material properties of Li, such as thermal expansion, melting temperature, radial distribution function of liquid Li and the structural stability at finite temperature by simulating the disordered–ordered transition

  7. Crystal structure of [UO2(NO3)2(C6H5NO2)2

    International Nuclear Information System (INIS)

    Mit'kovskaya, E.V.; Serezhkina, L.B.; Serezhkin, V.N.; Mikhajlov, Yu.N.; Gorbunova, Yu.E.

    2004-01-01

    Single crystals of uranyl nitrato complex with nicotinic acid of the composition [UO 2 (NO 3 ) 2 (C 6 H 5 NO 2 ) 2 ] (I) were studied by X-ray diffraction. The compound crystallizes in triclinic system, sp. gr. P1-bar, Z = 1, unit cell parameters being: a = 7.033 A, b = 7.351 A, c = 9.091 A, α = 91.85 Deg, β = 105.78 Deg, γ = 111.83 Deg. Basic structural units in the crystals of I are mononuclear centrosymmetrical groups of the same composition as the molecule. The crystal chemical formula of the complex is AB 2 01 M 2 1 (A = UO 2 2+ ) [ru

  8. Solubility in the UO2SeO4-(CH3)2NCONH2-H2O system

    International Nuclear Information System (INIS)

    Serezhkina, L.B.; Serezhkin, V.N.

    1994-01-01

    Complexing in the system UO 2 SeO 4 -(CH 3 ) 2 NCONH 2 -H 2 O at 25 deg C has been studied by the method of isothermal solubility. Existence of congruently soluble UO 2 SeO 4 x2(CH 3 ) 2 NCONH 2 x2H 2 O (1) has been detected. As a result of X-ray diffraction study of the monocrystals unit cell parameters of compound (1) have been ascertained: a = 14.251, b = 13.896, c = 17.789 A; β = 104.74, Z = 8, sp.gr. P2 1 /c. The assumption is made that compound (1) is a representative of the AT 3 M 2 1 crystallochemical group of uranyl complexes. 4 refs.; 1 fig.; 2 tabs

  9. Analysis of the effect of UO2 high burnup microstructure on fission gas release

    International Nuclear Information System (INIS)

    Jernkvist, Lars Olof; Massih, Ali

    2002-10-01

    This report deals with high-burnup phenomena with relevance to fission gas release from UO 2 nuclear fuel. In particular, we study how the fission gas release is affected by local buildup of fissile plutonium isotopes and fission products at the fuel pellet periphery, with subsequent formation of a characteristic high-burnup rim zone micro-structure. An important aspect of these high-burnup effects is the degradation of fuel thermal conductivity, for which prevalent models are analysed and compared with respect to their theoretical bases and supporting experimental data. Moreover, the Halden IFA-429/519.9 high-burnup experiment is analysed by use of the FRAPCON3 computer code, into which modified and extended models for fission gas release are introduced. These models account for the change in Xe/Kr-ratio of produced and released fission gas with respect to time and space. In addition, several alternative correlations for fuel thermal conductivity are implemented, and their impact on calculated fission gas release is studied. The calculated fission gas release fraction in IFA-429/519.9 strongly depends on what correlation is used for the fuel thermal conductivity, since thermal release dominates over athermal release in this particular experiment. The conducted calculations show that athermal release processes account for less than 10% of the total gas release. However, athermal release from the fuel pellet rim zone is presumably underestimated by our models. This conclusion is corroborated by comparisons between measured and calculated Xe/Kr-ratios of the released fission gas

  10. Fission gas release behavior in high burnup UO2 fuels with developed rim-structure

    International Nuclear Information System (INIS)

    Une, Katsumi; Kashibe, Shinji; Hayashi, Kimio

    2002-01-01

    The effect of rim structure formation and external restraint pressure on fission gas release at transient conditions has been examined by using an out-of-pile high pressure heating technique for high burnup UO 2 fuels (60, 74 and 90 GWd/t), which had been irradiated in test reactors. The latter two fuels bore a developed rim structure. The maximum heating temperature was 1500 degC, and the external pressures were independently controlled in the range of 10-150 MPa. The present high burnup fuel data were compared with those of previously studied BWR fuels of 37 and 54 GWd/t with almost no rim structure. The fission gas release and bubble swelling due to the growth of grain boundary bubbles and coarsened rim bubbles were effectively suppressed by the strong restraint pressure of 150 MPa for all the fuels; however the fission gas release remarkably increased for the two high burnup fuels with the developed rim structure, even at the strong restraint conditions. From the stepwise de-pressurization tests at an isothermal condition of 1500degC, the critical external pressure, below which a large burst release due to the rapid growth and interlinkage of the bubbles abruptly begins, was increased from a 40-60 MPa level for the middle burnup fuels to a high level of 120-140 MPa for the rim-structured high burnup fuels. The high potential for transient fission gas release and bubble swelling in the rim-structured fuels was attributed to highly over-pressurized fission gases in the rim bubbles. (author)

  11. Solubility and Solubility Product at 22C of UO2(c) Precipitated From Aqueous U(IV) Solutions

    International Nuclear Information System (INIS)

    Rai, Dhanpat; Yui, Mikazu; Moore, Dean A.

    2003-01-01

    Solubility studies were conducted under rigidly controlled redox conditions maintained by EuCl2 as a function of pH and from the oversaturation direction where UO2(c) was precipitated at 90 degrees C from degrees ow-pH U(IV) solutions. Samples were equilibrated for 24 days at 90 degrees C and then for day at 22 degrees C

  12. A physical and chemical analysis of fast quenched particles of UO 2 and ZrO 2 mixture

    Science.gov (United States)

    Min, Beong Tae; Song, Jin Ho; Park, Yang Soon; Kim, Jong Gu

    2006-11-01

    An interaction between molten fuel of a nuclear reactor, which is called corium and mainly consisted of UO 2 and ZrO 2, and sub-cooled water may result in a steam explosion. It is one of the outstanding reactor safety issues. To investigate the fundamental mechanism behind the recent experimental observation that the composition of the material highly affected the strength of the steam explosion, a physical and chemical analysis for the fast quenched particles of UO 2 and ZrO 2 mixture at different compositions was performed. Six cases were selected for the study, in which the melt composition was changed, while other initial and boundary conditions of the molten fuel and water interaction tests were maintained the same. It was observed that the cases at eutectic composition resulted in a spontaneous steam explosion, while the cases at non-eutectic composition did not result in a spontaneous steam explosion. Electron probe microanalysis (EPMA) was performed for fast quenched particles along a cross-section. Results demonstrated that the UO 2 and ZrO 2 mixtures formed a solid solution of U 1- xZr xO 2. The mechanism for the hydrogen generation during the molten material and water interaction was examined by thermogravity analysis (TGA), X-ray diffraction (XRD) and hydrogen reduction analysis. It was demonstrated that the hydrogen generation was not directly related to the oxidation of UO 2. Morphologies observed by scanning electron microscopy (SEM) indicated that the particles from the eutectic mixture had many holes, while the particles at non-eutectic mixture did not. The existence of mush phase for the non-eutectic mixture is suggested to be the reason for the non-explosive nature.

  13. Long term irradiation of SAP-clad, UO2 fuelled, trefoil bundles in the X-7 organic cooled loop

    International Nuclear Information System (INIS)

    Robertson, R.F.S.; Thexton, H.E.; Lew, D.E.; MacDonald, R.D.; Heal, K.G.

    1963-05-01

    An irradiation of experimental fuel bundles of UO 2 clad in Sintered Aluminium Product (SAP) sheaths was performed in the X-7 loop in the NRX Reactor from August to November 1961. The twofold objective of this irradiation was to gain confidence in the use of such a fuel in an organic coolant and to assess the fouling problem associated with a coolant which was as 'clean' as the existing technology could produce. The trefoil bundles successfully underwent an irradiation of 2400 MWD/Tonne U (max.). The maximum sheath temperature was 460 o C in a coolant temperature of about 310 o C. Surface heat fluxes were roughly 100 w/cm 2 . After irradiation a film of about 80 μm thickness was found covering the sheath over fuelled sections. The film was by weight 40% polymerized organic and 60% Fe 3 O 4 . The net effect of the film was to increase the pressure drop across the fuel by 50% and increase the sheath temperature by 60 o C. The SAP sheath showed no effects of irradiation except for small apparent increases in diameter at two out of eighteen planes measured. The appearance of the UO 2 was similar to that of UO 2 irradiated in pressurized water loops at similar heat ratings. (author)

  14. Complexation in the system K2SeO4-UO2SeO4-H2O

    International Nuclear Information System (INIS)

    Serezhkina, L.B.; Kuchumova, N.V.; Serezhkin, V.N.

    1994-01-01

    Complexation in the system K 2 SeO 4 -UO 2 SeO 4 -H 2 O at 25 degrees C is studied by isothermal solubility. Congruently soluble K 2 UO 2 (SeO 4 ) 2 ·4H 2 O (I) and incongruently soluble K 2 (UO 2 ) 2 (SeO 4 ) 3 ·6H 2 O (II) are observed. The unit-cell constants of I and II are determined from an X-ray diffraction investigation. For I, a = 12,969, b = 11.588, c = 8.533 angstrom, Z = 4, space group Pmmb. For II, a = 23.36, b = 6.784, c = 13.699 angstrom, β = 104.42 degrees, Z = 4, space group P2/m, P2, or Pm. Complexes I and II are representatives of the crystal-chemical groups AB 2 2 M 1 and A 2 T 3 3 M 1 , respectively, of uranyl complexes

  15. Fission product release from UO2 during irradiation. Diffusion data and their application to reactor fuel pins

    International Nuclear Information System (INIS)

    Findlay, J.R.; Johnson, F.A.; Turnbull, J.A.; Friskney, C.A.

    1980-01-01

    Release of fission product species from UO 2 , and to a limited extent from (U, Pu)0 2 was studied using small scale in-reactor experiments in which these interacting variables may be separated, as far as is possible, and their influences assessed. Experiments were at fuel ratings appropriate to water reactor fuel elements and both single crystal and poly-crystalline specimens were used. They employed highly enriched uranium such that the relative number of fissions occurring in plutonium formed by neutron capture was small. The surface to volume ratio (S/V) of the specimens was well defined thus reducing the uncertainties in the derivation of diffusion coefficients. These experiments demonstrate many of the important characteristics of fission product behaviour in UO 2 during irradiation. The samples used for these experiments were small being always less than 1g with a fissile content usually between 2 and 5mg. Polycrystalline materials were taken from batches of production fuel prepared by conventional pressing and sintering techniques. The enriched single crystals were grown from a melt of sodium and potassium chloride doped with UO 2 powder 20% 235 U content. The irradiations were performed in the DIDO reactor at Harwell. The neutron flux at the specimen was 4x10 16 neutrons m -2 s -1 providing a heat rating within the samples of 34.5 MW/teU

  16. Oxidation of UO2 at 400 to 1000 degrees C in air and its relevance to fission product release

    International Nuclear Information System (INIS)

    McCracken, D.R.

    1985-07-01

    Currently there is great interest in the behaviour of UO 2 under oxidizing conditions because irradiated uranium dioxide fuel can conceivably be exposed to a hot oxidizing atmosphere as a result of accidents. The temperature range covered in this paper is 400 to 1000 degrees C. At these high temperatures, UO 2 in air can oxidize rapidly to U 3 O 8 via U 3 O 7 and/or U 4 O 9 . The accompanying volume increase and corresponding stresses lead to fragmentation of the fuel pellets. The purpose of this work was to investigate the dependence of UO 2 oxidation on temperature, rate of air supply and residence time at temperature; to determine the rate controlling steps and rate of oxygen penetration; and to characterize the oxidation products and size of fragments. In addition, detailed metallography was related to X-ray diffraction studies of the oxidized UO 2 to facilitate future study of irradiated fuel, which is easier to do by metallography in hot-cells than by X-ray diffraction. Samples were heated in argon, then once at temperature they were exposed to air at a controlled flow-rate. Studies of the oxidation of unirradiated UO 2 pellets in air show two distinct types of oxidation with a change in mechanism at 600-700 degrees C. At temperatures ≤ 600 degrees C fragmentation accompanies the formation of U 3 O 8 while at T ≥ 800 degrees C, rapid grain growth occurs. In the first temperature region, volatile fission product releases are small, while in the second region, 100% release can be correlated with U 3 O 8 formation. In the first region, only the grain boundary inventory is released while in the other, 100% of the Xe, Kr, Ru, Sb, Cs and I are released. It appears that, within the error of present measurements, burnup does not affect rates of fission product release and oxidation in air at 400 to 1000 degrees C, so that oxidation rate data gathered using unirradiated pellets can be applied to irradiated fuel. 33 refs

  17. Monte Carlo analysis of experiments on the reactivity temperature coefficient for UO2 and MOX light water moderated lattices

    International Nuclear Information System (INIS)

    Erradi, L.; Chetaine, A.; Chakir, E.; Kharchaf, A.; Elbardouni, T.; Elkhoukhi, T.

    2005-01-01

    In a previous work, we have analysed the main French experiments available on the reactivity temperature coefficient (RTC): CREOLE and MISTRAL experiments. In these experiments, the RTC has been measured in both UO 2 and UO 2 -PuO 2 PWR type lattices. Our calculations, using APOLLO2 code with CEA93 library based on JEF2.2 evaluation, have shown that the calculation error in UO 2 lattices is less than 1 pcm/C degrees which is considered as the target accuracy. On the other hand the calculation error in the MOX lattices is more significant in both low and high temperature ranges: an average error of -2 ± 0.5 pcm/C degrees is observed in low temperatures and an error of +3 ± 2 pcm/C degrees is obtained for temperatures higher than 250 C degrees. In the present work, we analysed additional experimental benchmarks on the RTC of UO 2 and MOX light water moderated lattices. To analyze these benchmarks and with the aim of minimizing uncertainties related to modelling of the experimental set up, we chose the Monte Carlo method which has the advantage of taking into account in the most exact manner the geometry of the experimental configurations. This analysis shows for the UO 2 lattices, a maximum experiment-calculation deviation of about 0,7 pcm/C degrees, which is below the target accuracy for this type of lattices. For the KAMINI experiment, which relates to the measurement of the RTC in a light water moderated lattice using U-233 as fuel our analysis shows that the ENDF/B6 library gives the best result, with an experiment-calculation deviation of the order of -0,16 pcm/C degrees. The analysis of the benchmarks using MOX fuel made it possible to highlight a discrepancy between experiment and calculation on the RTC of about -0.7 pcm/C degrees (for a range of temperatures going from 20 to 248 C degrees) and -1,2 pcm/C degrees (for a range of temperatures going from 20 to 80 C degrees). This result, in particular the tendency which has the error to decrease when the

  18. Analysis of gas flow measurements from the IFA-633 UO2/MOX comparison test

    International Nuclear Information System (INIS)

    Rossiter, Glyn

    2005-01-01

    The release rate to birth rate ratio (R/B) results from the gas flow measurements performed during the joint programme irradiation of the IFA-633 UO 2 /MOX comparison test have been analysed using both classical and fractal methodologies. Possible calculational procedures for precursor enhancement factors and rod average diffusion coefficients were considered and suitable procedures were then implemented. The surface area to volume ratio (S/V) and recoil R/B values generated using the two methodologies have been compared to each other and to results obtained for other Halden Project gas flow rigs (IFAs 504, 558, 563, 569 and 655). The merits of the methodologies have then been discussed. It was found that the trends in the classical recoil R/B and in the fractal S/V for the shortest lived isotopes were in better agreement with the expected S/V behaviour than the trends in the classical S/V and in the fractal S/V for the longer lived isotopes. The beginning of life (BOL) S/V versus temperature behaviour for both IFA-633 and IFA-655 has been investigated and has been found to be more consistent with expectation when the fractal methodology is used. The peak fuel temperature versus rod average burnup behaviour of the IFA-633 fuel rods has been examined in order to investigate whether there is any correlation between the S/V results and the extent of periods during which the Halden (or Vitanza) threshold for significant fission gas release was exceeded. The behaviour was more consistent with the trends in the classical recoil R/B and fractal S/V for the shortest lived isotopes than with the trends in the classical S/V and the fractal S/V for the longer lived isotopes. The analysis of the through-life and BOL S/V and recoil R/B results generated using the classical and fractal methodologies has shown that the behaviour of the classical recoil R/B is difficult to explain. This is evidence that the classical recoil R/B results contain a diffusional release component

  19. Mixed-metal uranium(VI) iodates: hydrothermal syntheses, structures, and reactivity of Rb[UO(2)(CrO(4))(IO(3))(H(2)O)], A(2)[UO(2)(CrO(4))(IO(3))(2)] (A = K, Rb, Cs), and K(2)[UO(2)(MoO(4))(IO(3))(2)].

    Science.gov (United States)

    Sykora, Richard E; McDaniel, Steven M; Wells, Daniel M; Albrecht-Schmitt, Thomas E

    2002-10-07

    The reactions of the molecular transition metal iodates A[CrO(3)(IO(3))] (A = K, Rb, Cs) with UO(3) under mild hydrothermal conditions provide access to four new, one-dimensional, uranyl chromatoiodates, Rb[UO(2)(CrO(4))(IO(3))(H(2)O)] (1) and A(2)[UO(2)(CrO(4))(IO(3))(2)] (A = K (2), Rb (3), Cs (4)). Under basic conditions, MoO(3), UO(3), and KIO(4) can be reacted to form K(2)[UO(2)(MoO(4))(IO(3))(2)] (5), which is isostructural with 2 and 3. The structure of 1 consists of one-dimensional[UO(2)(CrO(4))(IO(3))(H(2)O)](-) ribbons that contain uranyl moieties bound by bridging chromate and iodate anions as well as a terminal water molecule to create [UO(7)] pentagonal bipyramidal environments around the U(VI) centers. These ribbons are separated from one another by Rb(+) cations. When the iodate content is increased in the hydrothermal reactions, the terminal water molecule is replaced by a monodentate iodate anion to yield 2-4. These ribbons can be further modified by replacing tetrahedral chromate anions with MoO(4)(2)(-) anions to yield isostructural, one-dimensional [UO(2)(MoO(4))(IO(3))(2)](2)(-) ribbons. Crystallographic data: 1, triclinic, space group P(-)1, a = 7.3133(5) A, b = 8.0561(6) A, c = 8.4870(6) A, alpha = 88.740(1) degrees, beta = 87.075(1) degrees, gamma = 71.672(1) degrees, Z = 2; 2, monoclinic, space group P2(1)/c, a = 11.1337(5) A, b = 7.2884(4) A, c = 15.5661(7) A, beta = 107.977(1) degrees, Z = 4; 3, monoclinic, space group P2(1)/c, a = 11.3463(6) A, b = 7.3263(4) A, c = 15.9332(8) A, beta = 108.173(1) degrees, Z = 4; 4, monoclinic, space group P2(1)/n, a = 7.3929(5) A, b = 8.1346(6) A, c = 22.126(2) A, beta = 90.647(1) degrees, Z = 4; 5, monoclinic, space group P2(1)/c, a = 11.3717(6) A, b = 7.2903(4) A, c = 15.7122(8) A, beta = 108.167(1) degrees, Z = 4.

  20. Thermodynamic state, specific heat, and enthalpy function of saturated UO2 vapor between 3,000 K and 5,000 K

    International Nuclear Information System (INIS)

    Karow, H.U.

    1977-02-01

    The properties have been determined by means of statistical mechanics. The discussion of the thermodynamic state includes the evaluation of the plasma state and its contribution to the caloric variables-of-state of saturated oxide fuel vapor. Because of the extremely high ion and electron density due to thermal ionization, the ionized component of the fuel vapor does no more represent a perfect kinetic plasma. At temperatures around 5,000 K, UO 2 vapor reaches the collective plasma state and becomes increasingly 'metallic'. - Moreover, the nonuniform molecular equilibrium composition of UO 2 vapor has been taken into account in calculating its caloric functions-of-state. The contribution to specific heat and enthalpy of thermally excited electronic states of the vapor molecules has been derived by means of a Rydberg orbital model of the UO 2 molecule. The resulting enthalpy functions and specific heats for saturated UO 2 vapor of equilibrium composition and that for pure UO 2 gas are compared with the enthalpy and specific heat data of gaseous UO 2 at lower temperatures known from literature. (orig./HP) [de

  1. Determination of Uranium In UO2 And U3O8 Powder Using UV-VIS Spectrophotometry

    International Nuclear Information System (INIS)

    Natalia Adventini; Diah Dwiana Lestiani; Muhayatun; Endah Damastuti

    2009-01-01

    Lab. TAR PTNBR BATAN - Bandung has been accredited by National Accreditation Committee on May 2 nd , 2006 as a test laboratory with number LP-311-ID, has to maintain its laboratory performance by participating in a proficiency test. In this activity, the determination of uranium in 2 samples of UO 2 with A1 and A2 codes and other 2 samples of U 3 O 8 with B1 and B2 codes using UV-Vis spectrophotometry was carried out. Colouring method was used by reacting thiocyanate ion with the uranyl ion in acidic solution to develop a stable yellow colour of uranyl thiocyanate complex solution and measured at wavelength of 380 nm. The result gave that concentration of uranium in A1, A2, B1 and B2 samples were 77.95; 75.29; 64.58 and 63.69% respectively. The Z-score value for A samples was - 1.99, meanwhile for B samples the Z score value of between laboratory was −1.29 with intra laboratory was -1,09. It meant that Z-score values for both samples were in good category. From this result, it showed that UV-Vis spectrophotometry is one of the several methods that can be used to determine uranium in UO 2 and U 3 O 8 powder. The Lab. TAR’s proficiency test for determination of uranium in UO 2 and U 3 O 8 gave a good result and it was hoped to support BATAN's program in the nuclear fuel field. (author)

  2. Role of Microstructure and Surface Defects on the Dissolution Kinetics of CeO2, a UO2 Fuel Analogue.

    Science.gov (United States)

    Corkhill, Claire L; Bailey, Daniel J; Tocino, Florent Y; Stennett, Martin C; Miller, James A; Provis, John L; Travis, Karl P; Hyatt, Neil C

    2016-04-27

    The release of radionuclides from spent fuel in a geological disposal facility is controlled by the surface mediated dissolution of UO2 in groundwater. In this study we investigate the influence of reactive surface sites on the dissolution of a synthesized CeO2 analogue for UO2 fuel. Dissolution was performed on the following: CeO2 annealed at high temperature, which eliminated intrinsic surface defects (point defects and dislocations); CeO2-x annealed in inert and reducing atmospheres to induce oxygen vacancy defects and on crushed CeO2 particles of different size fractions. BET surface area measurements were used as an indicator of reactive surface site concentration. Cerium stoichiometry, determined using X-ray Photoelectron Spectroscopy (XPS) and supported by X-ray Diffraction (XRD) analysis, was used to determine oxygen vacancy concentration. Upon dissolution in nitric acid medium at 90 °C, a quantifiable relationship was established between the concentration of high energy surface sites and CeO2 dissolution rate; the greater the proportion of intrinsic defects and oxygen vacancies, the higher the dissolution rate. Dissolution of oxygen vacancy-containing CeO2-x gave rise to rates that were an order of magnitude greater than for CeO2 with fewer oxygen vacancies. While enhanced solubility of Ce(3+) influenced the dissolution, it was shown that replacement of vacancy sites by oxygen significantly affected the dissolution mechanism due to changes in the lattice volume and strain upon dissolution and concurrent grain boundary decohesion. These results highlight the significant influence of defect sites and grain boundaries on the dissolution kinetics of UO2 fuel analogues and reduce uncertainty in the long term performance of spent fuel in geological disposal.

  3. Design and control of the oxygen partial pressure of UO2 in TGA using the humidification system

    International Nuclear Information System (INIS)

    Lee, S.; Knight, T.W.; Roberts, E.

    2015-01-01

    Highlights: • We focus on measurement of oxygen partial pressure and change of O/M ratio under specific conditions produced by the humidification system. • This shows that the humidification system is stable, accurate, and reliable enough to be used for experiments of the oxygen partial pressure measurement for the oxide fuels. • The humidification system has benefits of easy control and flexibility for producing various oxygen partial pressures with fixed hydrogen gas flow rate. - Abstract: The oxygen to uranium (O/U) ratio of UO 2±x is determined by the oxygen content of the sample and is affected by oxygen partial pressure (pO 2 ) of the surrounding gas. Oxygen partial pressure is controllable by several methods. A common method to produce different oxygen partial pressures is the use of equilibria of different reaction gases. There are two common methods: H 2 O/H 2 reaction and CO 2 /CO reaction. In this work, H 2 O/H 2 reaction using a humidifier was employed and investigated to ensure that this humidification system for oxygen partial pressure is stable and accurate for use in Thermogravimetric Analyzer (TGA) experiments with UO 2 . This approach has the further advantage of flexibility to make a wide range of oxygen partial pressure with fixed hydrogen gas flow rate only by varying temperature of water in the humidifier. The whole system for experiments was constructed and includes the humidification system, TGA, oxygen analyzer, and gas flow controller. Uranium dioxide (UO 2 ) samples were used for experiments and oxygen partial pressure was measured at the equilibrium state of stoichiometric UO 2.0 . Oxygen partial pressures produced by humidification (wet gas) system were compared to the approach using mixed dry gases (without humidification system) to demonstrate that the humidification system provides for more stable and accurate oxygen partial pressure control. This work provides the design, method, and analysis of a humidification system for

  4. Design of springs in non-instrumented capsule for the HANARO irradiation test of advanced UO2 fuel

    International Nuclear Information System (INIS)

    Kim, D. H.; Lee, C. B.; Kang, H. S.

    2001-01-01

    Non-instrumented capsule was designed to irradiate the large grain UO 2 pellet developed for the high burn-up LWR fuel in the HANARO reactor. The non-instrumented capsule will be irradiated for about 30 months in HANARO Outside-core (OR) region. In the non-instrumented capsule, there are four different springs such as top guide spring, bottom spring, plenum spring and fuel rod hol-down spring. To ensure the mechanical integrity of non-instrumented capsule during the long term operation, those springs were designed after the spring characteristic tests

  5. Dependence of rim pore radius on rim porosity and temperature behavior in the high burnup UO2 fuel

    International Nuclear Information System (INIS)

    Lee, Byung Ho; Koo, Yang Hyun; Sohn, Dong Seong

    1997-01-01

    The rim porosity at the high burnup UO 2 fuel is obtained at the various rim pore radius ranging from 0.25 to 1.0 μm. The rim pore radius of 1.0 μm gives the best estimation for the rim porosity. With increasing the rim pore radius, thermal conductivity degrades with pellet average burnup because the rim pore acts as the thermal barrier. And by using the NEA database, the culculated fuel centerline temperature considering the rim effect is compared with the experimentally measured NEA database. The calculated temperature predicts reasonably well the temperature behavior of irradiated fuel

  6. Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-UO2 configuration.

    Energy Technology Data Exchange (ETDEWEB)

    Klann, R. T.; Perret, G.; Nuclear Engineering Division

    2007-10-03

    An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R1-UO2 core configuration were completed. The reactor model was generated using the REBUS code developed at Argonne National Laboratory. The calculations are based on the specifications for fabrication, so they are considered preliminary until sampling and analysis have been completed on the fabricated samples. The estimates indicate a range of reactivity effect from -22 pcm to +25 pcm compared to the natural U sample.

  7. Evaluation of B&W UO2/ThO2 VIII experimental core: criticality and thermal disadvantage factor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Carlo Parisi; Emanuele Negrenti

    2017-02-01

    In the framework of the OECD/NEA International Reactor Physics Experiment (IRPHE) Project, an evaluation of core VIII of the Babcock & Wilcox (B&W) Spectral Shift Control Reactor (SSCR) critical experiment program was performed. The SSCR concept, moderated and cooled by a variable mixture of heavy and light water, envisaged changing of the thermal neutron spectrum during the operation to encourage breeding and to sustain the core criticality. Core VIII contained 2188 fuel rods with 93% enriched UO2-ThO2 fuel in a moderator mixture of heavy and light water. The criticality experiment and measurements of the thermal disadvantage factor were evaluated.

  8. Solubility of unirradiated UO2 fuel in aqueous solutions. Comparison between experimental and calculated (EQ3/6) data

    International Nuclear Information System (INIS)

    Ollila, K.

    1995-11-01

    The solubility behaviour of unirradiated UO 2 pellets was studied under oxic (air-saturated) and anoxic (N 2 ) conditions in deionized water, in sodium bicarbonate solutions with varying bicarbonate content (60 - 600 ppm), in Allard groundwater simulating granitic fresh groundwater conditions, and in bentonite water simulating the effects of bentonite on granitic fresh groundwater (25 deg C). The release of uranium was measured during static batch dissolution experiments of long duration (2-6 years). A comparison was made with the theoretical solubility data calculated with the geochemical code EQ3/6 in order to evaluate solubility (steady state) limiting factors. (orig.) (26 refs., 32 figs., 13 tabs.)

  9. Microstructure changes and thermal conductivity reduction in UO2 following 3.9 MeV He2+ ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Janne Pakrinen; Marat Khafizov; Lingfeng He; Chris Wetland; Jian Gan; Andrew T. Nelson; David H Hurley; Anter El-Azab; Todd R Allen

    2014-11-01

    The microstructural changes and associated effects on thermal conductivity were examined in UO2 after irradiation using 3.9 MeV He2+ ions. Lattice expansion of UO2 was observed in x-ray diffraction after ion irradiation up to 5×1016 He2+/cm2 at low-temperature (< 200 °C). Transmission electron microscopy (TEM) showed homogenous irradiation damage across an 8 µm thick plateau region, which consisted of small dislocation loops accompanied by dislocation segments. Dome-shaped blisters were observed at the peak damage region (depth around 8.5 µm) in the sample subjected to 5×1016 He2+/cm2, the highest fluence reached, while similar features were not detected at 9×1015 He2+/cm2. Laser-based thermo-reflectance measurements showed that the thermal conductivity for the irradiated layer decreased about 55 % for the high fluence sample and 35% for the low fluence sample as compared to an un-irradiated reference sample. Detailed analysis for the thermal conductivity indicated that the conductivity reduction was caused by the irradiation induced point defects.

  10. Manufacturing of 380/220 V 63 A power supply panel on UO2 kernel coating unit

    International Nuclear Information System (INIS)

    Triyono; Mudjiman, Supardjono; Hidayat, Nur

    2013-01-01

    The panel 380/220 Volts voltage source current 63 Amperes relay AC output on UO 2 kernel coating unit has been manufactured. The activities include: installation and function panel test. The electrical installation of load includes : 220 Volts temperature monitor/controller, 220 Volts scraber pump, 220 Volts vacuum pump, 220 Volts solenoid valve gas and 380 Volts induction furnace. The component of installation includes: 380 Volts earth leakage circuit breaker, 380 Volts relay AC, 220 Volts magnetic circuit breaker, 100/5 Amperes current transformator ratio, voltmeter, ampermeter and push button normally open-close and temperature monitor/control in the box size 70x50x20 cm. The testing of voltage source has been unload and full load to determine the performance of the tool. The result of manufacture and function test voltage source 380/220 Volts at the UO 2 kernel coating unit showed that: the voltage source can work without load and full load safely. The load includes: voltage 380-382 Volts current 4-4,1 Amperes of induction furnace, voltage 223 Volts current 0,5 Amperes of vacuum pump, voltage 223 Volts current 4 Amperes of scraber pump, voltage 223 Volts current 0,3 Amperes of solenoid valve gas and 222 Volts current 0,03 Amperes temperature monitor/control. (author)

  11. Measurement of UO2 surface oxidation using grazing-incidence x-ray diffraction: Implications for nuclear forensics

    Science.gov (United States)

    Tracy, Cameron L.; Chen, Chien-Hung; Park, Sulgiye; Davisson, M. Lee; Ewing, Rodney C.

    2018-04-01

    Nuclear forensics involves determination of the origin and history of interdicted nuclear materials based on the detection of signatures associated with their production and trafficking. The surface oxidation undergone by UO2 when exposed to air is a potential signature of its atmospheric exposure during handling and transport. To assess the sensitivity of this oxidation to atmospheric parameters, surface sensitive grazing-incidence x-ray diffraction (GIXRD) measurements were performed on UO2 samples exposed to air of varying relative humidity (34%, 56%, and 95% RH) and temperature (room temperature, 50 °C, and 100 °C). Near-surface unit cell contraction was observed following exposure, indicating oxidation of the surface and accompanying reduction of the uranium cation ionic radii. The extent of unit cell contraction provides a measure of the extent of oxidation, allowing for comparison of the effects of various exposure conditions. No clear influence of relative humidity on the extent of oxidation was observed, with samples exhibiting similar degrees of unit cell contraction at all relative humidities investigated. In contrast, the thickness of the oxidized layers increased substantially with increasing temperature, such that differences on the order of 10 °C yielded readily observable crystallographic signatures of the exposure conditions.

  12. Manufacturing at industrial level of UO2 pellets for the fuel elements of the Atucha I Nuclear Power Plant

    International Nuclear Information System (INIS)

    Dyment, I.G.; Noguera Rojas, Francisco

    1982-01-01

    The interest to produce fuel elements within a policy of self sufficiency arose with the installation of Atucha I. The first steps towards this goal consisted in processing the uranium oxide, transforming it into fuel pellets of high density. The developments towards the fabrication of said pellets, performed by CNEA since 1968, first at a laboratory level and afterwards on an industrial scale, allowed CNEA to obtain its own technological capability to produce 400 kg of UO 2 per day. The fuel pellets manufacturing method developed by CNEA is a powder-metallurgical process, which, besides conventional equipment, involves the use of special equipment that required the performance of systematic testing programmes, as well as special training at operational level. The developed processes respond to a modern and advanced technology. A general scheme of the process, starting with a directly sinterable UO 2 powder, is described, including compacting of the powder into pellets, sintering, control of the temperature in the sintering and reduction zones and of the time of permanence in both zones, and cylindric rectifying of the pellets. During the whole process, specialized personnel controls the operations, after which the material is released by the Quality Control Department. The national contribution to the manufacturing technology of the pellets for fuel elements of power and research reactors was of 100%. (M.E.L.) [es

  13. The oxidative dissolution of unirradiated UO2 by hydrogen peroxide as a function of pH

    International Nuclear Information System (INIS)

    Clarens, F.; Pablo, J. de; Casas, I.; Gimenez, J.; Rovira, M.; Merino, J.; Cera, E.; Bruno, J.; Quinones, J.; Martinez-Esparza, A.

    2005-01-01

    The dissolution of non-irradiated UO 2 was studied as a function of both pH and hydrogen peroxide concentration (simulating radiolytic generated product). At acidic pH and a relatively low hydrogen peroxide concentration (10 -5 mol dm -3 ), the UO 2 dissolution rate decreases linearly with pH while at alkaline pH the dissolution rate increases linearly with pH. At higher H 2 O 2 concentrations (10 -3 mol dm -3 ) the dissolution rates are lower than the ones at 10 -5 mol dm -3 H 2 O 2 , which has been attributed to the precipitation at these conditions of studtite (UO 4 . 4H 2 O, which was identified by X-ray diffraction), together with the possibility of hydrogen peroxide decomposition. In the literature, spent fuel dissolution rates determined in the absence of carbonate fall in the H 2 O 2 concentration range 5 x 10 -7 - 5 x 10 -5 mol dm -3 according to our results, which is in agreement with H 2 O 2 concentrations determined in spent fuel leaching experiments

  14. The sintering blocking mechanism on the UO2-GD2O3 system. Part 1: the hypothesis of diffusion barrier

    International Nuclear Information System (INIS)

    Durazzo, M.; Frajndlich, E.U.C.; Riella, H.G.; Leal Neto, R.M.

    2004-01-01

    The direct incorporation of gadolinium into nuclear power reactor fuel is important to the reactivity compensation and adjustment of power distribution thus enabling longer fuel cycles and optimized fuel utilization. Dry mechanical blending of Gd 2 O 3 and UO 2 powders is commercially the most attractive process route due to its simplicity. Nevertheless, processing by this route leads to difficulties in getting sintered pellets with the minimum required density due to a sintering blocking mechanism. Regarding this, there s little published information and the explanations are focused on the formation of a low diffusivity Gd-rich (U, Gd)O 2 phase during sintering process which decreases pellets density. An attempt to understand the mechanism for this effect was done in this work. Experimental evidences indicated the existence of phases in the (U, Gd)O 2 system with structure different from the fluorite-type UO 2 structure. These new phases were found for Gd molar fractions higher than 0,5, which coincide with the lowering of both the sintered density and the interdiffusion coefficient. However, it has been also shown that these new phases cannot be itself the cause for the density decrease observed. (author)

  15. Highly porous acrylonitrile-based submicron particles for UO2(2+) absorption in an immunosensor assay.

    Science.gov (United States)

    Sahiner, Nurettin; Yu, Haini; Tan, Grace; He, Jibao; John, Vijay T; Blake, Diane A

    2012-01-01

    Our laboratory has previously reported an antibody-based assay for hexavalent uranium (UO(2)(2+)) that could be used on-site to rapidly assess uranium contamination in environmental water samples (Melton, S. J.; et al. Environ. Sci. Technol. 2009, 43, 6703-6709). To extend the utility of this assay to less-characterized sites of uranium contamination, we required a uranium-specific adsorbent that would rapidly remove the uranium from groundwater samples, while leaving the concentrations of other ions in the groundwater relatively unaltered. This study describes the development of hydrogel particles containing amidoxime groups that can rapidly and selectively facilitate the uptake of uranyl ions. A miniemulsion polymerization technique using SDS micelles was employed for the preparation of the hydrogel as linked submicrometer particles. In polymerization, acrylonitrile was used as the initial monomer, ethylene glycol dimethacrylate as the crosslinker and 2-hydroxymethacrylate, 1-vinyl-2-pyrrolidone, acrylic acid, or methacrylic acid were added as co-monomers after the initial seed polymerization of acrylonitrle. The particles were characterized by transmission electron spectroscopy, scanning electron microscopy (SEM) and cryo-SEM. The amidoximated particles were superior to a commercially available resin in their ability to rapidly remove dissolved UO(2)(2+) from spiked groundwater samples. © 2011 American Chemical Society

  16. Monte Carlo analysis of experiments on the reactivity temperature coefficient for UO2 and MOX light water moderated lattices

    International Nuclear Information System (INIS)

    Chakir, E.; Erradi, L.; Bardouni, T El.; Khoukhi, T El.; Boukhal, H.; Meroun, O.; Bakkari, B El

    2007-01-01

    Full text: In a previous work, we have analysed the main french experiments available on the reactivity temperature coefficient (RTC) : CREAOLE and Mistral experiments. In these experiments, the RTC has been measured in both UO2 and UO2-PuO2 PWR type lattices. Our calculations, using APPOLO2 code with CEA93 library based on JEF2.2 evaluation, have shown that the calculation error in UO2 lattices is less than 1 pcm/Deg C which is considered as the target accuracy. On the other hand the calculation error in the MOX lattices is more significant in both low and high temperature ranges : an average error of -2 ± 0.5 pcm/Deg C is observed in low temperatures and an error of +3±2 pcm/Deg C is obtained for temperature higher than 250Deg C. In the present work, we analysed additional experimental benchmarks on the RTC of UO2 and MOX light water moderated lattices. To analyze these benchmarks and with the aim of minimizing uncertainties related to modelling of the experimental set up, we chose the Monte Carlo Method which has the advantage of taking into account in the most exact manner the geometry of the experimental configurations. Thus we have used the code MCNP5, for its recognized power and its availability. This analysis shows for the UO2 lattices, an average experiment-calculation deviation of about 0,5 pcm/Deg C, which is largely below the target accuracy for this type of lattices, that we estimate at approximately 1 pcm/Deg C. For the KAMINI experiment, which relates to the measurement of the RTC in light water moderated lattice using U-233 as fuel our analysis shows that the Endf/B6 library gives the best result, with an experiment -calculation deviation of the order of -0,16 pcm/Deg C. The analysis of the benchmarks using MOX fuel made it possible to highlight a discrepancy between experiment and calculation on the RTC of about -0.7pcm/Deg C ( for a range of temperature going from 20 to 248 Deg C) and -1.2 pcm/Deg C ( for a range of temperature going from 20 to

  17. OpenKIM - Building a Knowledgebase of Interatomic Models

    Science.gov (United States)

    Bierbaum, Matthew; Tadmor, Ellad; Elliott, Ryan; Wennblom, Trevor; Alemi, Alexander; Chen, Yan-Jiun; Karls, Daniel; Ludvik, Adam; Sethna, James

    2014-03-01

    The Knowledgebase of Interatomic Models (KIM) is an effort by the computational materials community to provide a standard interface for the development, characterization, and use of interatomic potentials. The KIM project has developed an API between simulation codes and interatomic models written in several different languages including C, Fortran, and Python. This interface is already supported in popular simulation environments such as LAMMPS and ASE, giving quick access to over a hundred compatible potentials that have been contributed so far. To compare and characterize models, we have developed a computational processing pipeline which automatically runs a series of tests for each model in the system, such as phonon dispersion relations and elastic constant calculations. To view the data from these tests, we created a rich set of interactive visualization tools located online. Finally, we created a Web repository to store and share these potentials, tests, and visualizations which can be found at https://openkim.org along with futher information.

  18. Structural studies coupling X-ray diffraction and high-energy X-ray scattering in the UO2(2+)-HBr(aq) system.

    Science.gov (United States)

    Wilson, Richard E; Skanthakumar, S; Cahill, C L; Soderholm, L

    2011-11-07

    The structural chemistry of uranium(VI) in concentrated aqueous hydrobromic acid solutions was investigated using both single crystal X-ray diffraction and synchrotron-based high-energy X-ray scattering (HEXS) to reveal the structure of the uranium(VI) complexes in solution prior to crystallization. The crystal structures of a series of uranyl tetrabromide salts are reported, including Cs(2)UO(2)Br(4), Rb(2)UO(2)Br(4)·2H(2)O, K(2)UO(2)Br(4)·2H(2)O, and (NH(4))(2)UO(2)Br(4)·2H(2)O, as well as a molecular dimer of uranium(VI), (UO(2))(2)(OH)(2)Br(2)(H(2)O)(4). Limited correspondence exists between the structures observed in the solid state and those in solution. Quantitative analysis of the HEXS data show an average U-Br coordination number of 1.9(2) in solution, in contrast to the U-Br coordination number of 4 in the solid salts. © 2011 American Chemical Society

  19. [UO2(NH3)5]Br2·NH3: synthesis, crystal structure, and speciation in liquid ammonia solution by first-principles molecular dynamics simulations.

    Science.gov (United States)

    Woidy, Patrick; Bühl, Michael; Kraus, Florian

    2015-04-28

    Pentaammine dioxido uranium(VI) dibromide ammonia (1/1), [UO2(NH3)5]Br2·NH3, was synthesized in the form of yellow crystals by the reaction of uranyl bromide, UO2Br2, with dry liquid ammonia. The compound crystallizes orthorhombic in space group Cmcm and is isotypic to [UO2(NH3)5]Cl2·NH3 with a = 13.2499(2), b = 10.5536(1), c = 8.9126(1) Å, V = 1246.29(3) Å(3) and Z = 4 at 123 K. The UO2(2+) cation is coordinated by five ammine ligands and the coordination polyhedron can be best described as pentagonal bipyramid. Car-Parrinello molecular dynamics simulations are reported for [UO2(NH3)5](2+) in the gas phase and in liquid NH3 solution (using the BLYP density functional). According to free-energy simulations, solvation by ammonia has only a small effect on the uranyl-NH3 bond strength.

  20. The spatial and temporal distribution of inhaled UO2 particles in the respiratory tract of the rat. II. The relative concentration of UO2 between the intrapulmonary airways and the pulmonary tissue

    International Nuclear Information System (INIS)

    Gore, D.J.

    1983-01-01

    In this experiment the relative masses of UO 2 particles per unit mass of tissue are determined between intrapulmonary airways and pulmonary tissue in rat lungs. A mean value of 0.57 +/- 0.06 (SE) was determined in tissue sections taken at 3-mm or 5-mm intervals throughout the lobes of animals killed from 2 to 35 days after inhalation. The experimental results are compared with data derived from the ICRP Lung Model for man and indicate that in the rate approximately 4% of the lung burden as opposed to about 0.02% derived for man is associated with the intrapulmonary airways. The 4% figure, if also applicable to man, would have extremely important implications in radiological protection, since the basal cells of the bronchial epithelium are thought to be sensitive to the induction of bronchogenic carcinoma

  1. The spatial and temporal distribution of inhaled UO2 particles in the respiratory tract of the rat: II. the relative concentration of UO2 between the intrapulmonary airways and the pulmonary tissue

    International Nuclear Information System (INIS)

    Gore, D.J.

    1983-01-01

    In this experiment the relative masses of UO 2 particles per unit mass of tissue are determined between intrapulmonary airways and pulmonary tissue in rat lungs. A mean value of 0.57 +/-0.06 (SE) was determined in tissue sections taken at 3-mm or 5-mm intervals throughout the lobes of animals killed from 2 to 35 days after inhalation. The experimental results are compared with data derived from the ICRP Lung Model for man and indicate that in the rat approximately 4% of the lung burden as opposed to about 0.02% derived for man is associated with the intrapulmonary airways. The 4% figure, if also applicable to man, would have extremely important implications in radiological protection, since the basal cells of the bronchial epithelium are thought to be sensitive to the induction of bronchogenic carcinoma

  2. Mesures en continu de la redistribution de l'oxygène sous gradient thermique dans UO2+ x

    Science.gov (United States)

    Ducroux, R.; Fromont, M.; Baptiste, Ph. Jean; Pattoret, A.

    1980-09-01

    RésuméL'objet de ce travail est l'ètude hors-pile sous gradient thermique de la redistribution de l'oxygène dans UO2+x, de façon à expliquer ultèrieurement les profils O/U + Pu obtenus sur du combustible irradiè. Les gradients thermiques ont ètè obtenus à l'aide d'un four à image. La tache focale d'une surface de 0.5 cm2 est amenèe au sommet de l'èchantillon cylindrique d'oxyde dont la partie froide est au contact d'un four à environ 900°C; celui-ci permet de maintenir en conditions isothermes une minijauge à electrolyte solide (Th02-Y203) contenant une rèfèrence chimique Fe/FeO. La minijauge donne à tout instant l'activitè de l'oxygène à la partie froide des èchantillons. Elle a ètè testèe en mesurant les potentiels d'oxygène de diffèrents systèmes chimiques. Les expèriences ont ètè rèalisèes sous atmosphère d'argon purifiè par des pompes èlectrochimiques. Les èchantillons d'UO2+x sont soumis à un gradient thermique modeste de 300°C/cm environ pour limiter en partie chaude les phènomènes d'èvaporation. Après le suivi en continu de la fern les èchantillons sont trempés puis tronçonnès, le rapport O/U de chaque tranche ètant dèterminè. L'oxygène remonte le gradient thermique, de façon importante, bien que la durèe des recuits soit relativement courte et les tempèratures basses. La confrontation des rèsultats obtenus dans UO2+x, avec un calcul èlaborè sur la base de la thermodiffusion en phase solide pour l'oxygène, montre un accord très raisonnable.

  3. NDA (Non Destructive Assay) measurements for isotopic homogeneity of UO2 powder form mechanical blending and results comparison between gamma and mass spectrometry

    International Nuclear Information System (INIS)

    Rojas, Carlos A.; Rojo, Marcelo

    2005-01-01

    Eight batches of 0.95% UO2 powder, obtained by mechanical blending of 3.5% and 0.711 % UO2 powders, were sampled. From each batch, samples at the top and the bottom from four drums were taken. Each sample was analysed using different measurement systems, two with NaI(Tl) detectors and another two with HPGe detectors. The Mini Multichannel Analyser (MMCA), model GBS 166, and the calculation codes NaIGEM and MEGAU-EM for peak area analysis and enrichment determination were used. For all cases the WinSPEC acquisition code was used. From the statistical analysis of the measurement results it arises that it is possible to determine the homogeneity grade of UO2 powder samples with a lower error than 0.5% for both types of detectors. The performance of the HPGe measurement system is only slightly more precise than the NaI system. (author)

  4. The system Na2SeO4-UO2SeO4-H2O at 25 deg C

    International Nuclear Information System (INIS)

    Tatarinova, E.Eh.; Serezhkina, L.B.; Serezhkin, V.N.

    1990-01-01

    By the method of isothermal solubility at 25 deg C interaction in system Na 2 SeO 4 -UO 2 SeO 4 -H 2 O has been studied. Two new crystal compounds are detected: Na 2 UO 2 (SeO 4 ) 2 ·4H 2 O and Na 2 (UO 2 ) 2 (SeO 4 ) 3 ·10H 2 O. X-ray diffraction studies and thermal analysis of sodium selenatouranylates isolated are carried out, monoclinic unit cell parameters are determined: a=8.637(4), 19.745(7); b=11.006(5), 10.783(4); c=13.887(5), 21.261(7) A; β=108.11(4), 103.48(2) deg respectively

  5. Investigation of oxygen disorder, thermal parameters, lattice vibrations and elastic constants of UO2 and ThO2 at temperatures up to 2 930 K

    DEFF Research Database (Denmark)

    Clausen, Kurt Nørgaard; Hayes, W; Hutchings, M.T.

    1984-01-01

    A knowledge of the thermodynamic properties of UO2 at temperatures in the region 1 500-3 100 K is of importance in reactor safety calculations, yet there are relatively few detailed experimental data available. In particular the major question of whether Frenkel disorder occurs in UO2 at high...... temperatures has been unanswered until now. A new high temperature furnace has been purchased by Harwell for work at temperatures in this region, and a series of experiments has been carried out involving diffraction, quasielastic diffuse and inelastic neutron scattering from single crystals of UO2 and ThO2....... These have been backed by experiments in the lower temperature range to 2 500 K at I.L.L. Details of the Harwell furnace, and methods used for temperature measurement and encapsulation of the crystal samples are given, together with some examples of the principal results. These results show unambiguously...

  6. Physics study on recycling of ThO2/UO2 fuel in CANDU reactors through dry reprocess technology

    International Nuclear Information System (INIS)

    Choi, Hang Bok; Park, Chang Je; Jeong, Chang Joon

    2003-06-01

    The dry process fuel technology has high proliferation-resistance which is one of important goals of Generation-IV (Gen-IV) reactor development. It is expected that the dry process fuel technology can be applied not only to existing nuclear systems but also to future nuclear systems. In this report, the homogeneous ThO 2 -UO 2 fuel cycle option of a CANDU reactor has been studied, including the physics analysis of recycling spent fuel. Reactivity swing and variation of isotopic content with irradiation are reported for various cases of initial uranium loadings. It was found that natural uranium saving increases significantly by recycling thorium/uranium fuel and it is feasible to recycle thorium with the dry process technology in a CANDU reactor. It is, however, required to further investigate the dry process that can be applied to the thorium-abundant dioxide fuel

  7. Technical evaluation of the direct denitration process to obtain ceramic-grade UO2 powders using microwaves

    International Nuclear Information System (INIS)

    Lorenzo, Viviana J.; Marchi, Daniel E.; Menghini, Jorge E.

    1999-01-01

    The direct denitration process to obtain ceramic-grade UO 2 powders using microwaves has been studied and developed at laboratory scale. Conditions were given to obtain powders apt for fuel pellets fabrication within the required specifications, where mechanical treatments before pressing are not necessary. This work describes the equipment used in the process, evaluates the necessary supply and waste generation and describes the characteristics of the product obtained, as well as the conditions for its fabrication. Results show that this method allows to reduce the volume of liquid wastes generated due to their partial re-utilization, simplifying their final disposal treatment, which, in addition to their operational advantages, make this method attractive from the economical point of view. (author)

  8. A semi-empirical model for the formation and depletion of the high burnup structure in UO2

    Science.gov (United States)

    Pizzocri, D.; Cappia, F.; Luzzi, L.; Pastore, G.; Rondinella, V. V.; Van Uffelen, P.

    2017-04-01

    In the rim zone of UO2 nuclear fuel pellets, the combination of high burnup and low temperature drives a microstructural change, leading to the formation of the high burnup structure (HBS). In this work, we propose a semi-empirical model to describe the formation of the HBS, which embraces the polygonisation/recrystallization process and the depletion of intra-granular fission gas, describing them as inherently related. For this purpose, we performed grain-size measurements on samples at radial positions in which the restructuring was incomplete. Based on these new experimental data, we infer an exponential reduction of the average grain size with local effective burnup, paired with a simultaneous depletion of intra-granular fission gas driven by diffusion. The comparison with currently used models indicates the applicability of the herein developed model within integral fuel performance codes.

  9. Post-irradiation behaviour of defected UO2 fuel elements in air at 220-250 degrees C

    International Nuclear Information System (INIS)

    Novak, J.; Hastings, I.J.

    1983-08-01

    We have heated in air irradiated CANDU UO 2 fuel elements with and without deliberately-induced defects. The temperature range was 220-250 degrees C for times up to 685 h. Pre-test burnup was about 190 MW.h/kg U (8000 MW.d/TeU) at a maximum linear power of 45 kW/m. Elements with single and multiple defects maintained reasonable dimensional stability up to 685 h at 220 and 230 degrees C, consistent with fuel oxidation primarily to U 3 O 7 . In this temperature range, there was no significant difference in response of defected elements with different power histories and cooling times. Irradiated fuel showed rates of weight increases up to 50 times greater than those for unirradiated fuel. Elements with single and multiple defects showed significant diametral increases and severe sheath splitting after about 200 h at 250 degrees C, consistent with fuel oxidation primarily to U 3 O 8

  10. [Non-empirical interatomic potentials for transition metals

    International Nuclear Information System (INIS)

    1993-01-01

    The report is divided into the following sections: potential-energy functions for d-band metals, potential-energy functions for aluminides and quasicrystals, electronic structure of complex structures and quasicrystals, potential-energy functions in transition-metal oxides, applications to defect structure and mechanical properties, and basic theory of interatomic potentials

  11. Framatome-ANP France UO2 fuel fabrication - criticality safety analysis in the light of the 1999' Tokay Mura accident

    International Nuclear Information System (INIS)

    Doucet, M.; Zheng, S.; Mouton, J.; Porte, R.

    2004-01-01

    In France the 1999' Tokai Mura criticality accident in Japan had a big impact on the nuclear fuel manufacturing facility community. Moreover this accident led to a large public discussion about all the nuclear facilities. The French Safety Authorities made strong requirements to the industrials to revisit completely their safety analysis files mainly those concerning nuclear fuels treatments. The Framatome-ANP production of its French low enriched (5 w/o) UO 2 fuel fabrication plant (FBFC/Romans) exceeds 1000 metric tons a year. Special attention was given to the emergency evacuation plan that should be followed in case of a criticality accident. If a criticality accident happens, site internal and external radioprotection requirements need to have an emergency evacuation plan showing the different routes where the absorbed doses will be as low as possible for people. The French Safety Authorities require also an update of the old based neutron source term accounting for state of the art methodology. UO 2 blenders units contain a large amount of dry powder strictly controlled by moderation; a hypothetical water leakage inside one of these apparatus is simulated by increasing the water content of the powder. The resulted reactivity insertion is performed by several static calculations. The French IRSN/CEA CRISTAL codes are used to perform these static calculations. The kinetic criticality code POWDER simulates the power excursion versus time and determines the consequent total energy source term. MNCP4B performs the source term propagation (including neutrons and gamma) used to determine the isodose curves needed to define the emergency evacuation plant. This paper deals with the approach Framatome-ANP has taken to assess Safety Authorities demands using the more up to date calculation tools and methodology. (authors)

  12. Effect of uranyl nitrate and free acid concentration in feed solution of gelation on UO2 kernel quality

    International Nuclear Information System (INIS)

    Masduki, B.; Wardaya; Widarmoko, A.

    1996-01-01

    An investigation on the effect of uranium and free nitric acid concentration of uranyl nitrate as feed of gelation process on quality of UO 2 kernel was done.The investigation is to look for some concentration of uranyl nitrate solutions those are optimum as feed for preparation of gelled UO 3 . Uranyl nitrate solution of various concentration of uranium (450; 500; 550; 600; 650; 700 g/l) and free nitric acid of (0.9; 1.0; 1.1 N) was made into feed solutions by adding urea and HMTA with mole ratio of urea/uranium and HMTA/uranium 2.1 and 2.0. The feed solutions were changed into spherical gelled UO 3 by dropping was done to get the optimum concentrations of uranyl nitrate solutions. The gelled UO 3 was soaked and washed with 2.5% ammonia solution for 17 hours, dried at 70 o C, calcined at 350 o C for 3 hours then reduced at 850 o C for 3 hours. At every step of the steps process the colour and percentage of well product of gelled UO 3 were noticed. The density and O/U ratio of end product (UO 2 kernel) was determined, the percentage of well product of all steps process was also determined. The three factor were used to chose the optimum concentration of uranyl nitrate solution. From this investigation it was concluded that the optimum concentration of uranyl nitrate was 600 g/l uranium with free nitric acid 0,9 - 1,0 N, the percentage of well product was 97% density of 6.12 - 4.8 g/cc and O/U ratio of 2.15 - 2.06. (author)

  13. Making of U3O8 Microsphere as a Preliminary Material for Manufacturing UO2 Kernel of HTR

    International Nuclear Information System (INIS)

    Hidayati; Triyono; Endang Nawangsih

    2007-01-01

    The making of U 3 O 8 microsphere as a preliminary material for manufacturing UO 2 kernel of HTR on various feed solution with internal gelation method use of paraffin gelation medium has been done. The aim of this research is to make U 3 O 8 microsphere as preliminary material for making UO 2 kernel which has good characteristic and for knowing to some extent the feed solution influence on U 3 O 8 microsphere. Uranyl nitrate solution was used as a feed solution with acidity 1 M and some various of ADUN solution. ADUN solution was made by adding various of ammonia solution on the solution of uranyl nitrate. Each of the feed solution was added urea + HMTA solution and then it was dropped to a column containing hot paraffin solution at the temperature 95°Celsius in order to get UO 3 gel. UO 3 gel was dipped and washed with NH 4 OH, dried and calcined at the temperature of 800°Celsius . The obtained product was analyzed its surface area, radius of pore, total volume of pore and distribution of pore size of Surface Area Analyzer NOVA-1000. The density was analyzed with pycnometer and the form of microsphere was analyzed with SEM. The obtained product shows that U 3 O 8 microsphere with less ammonium nitrate gave U 3 O 8 much better and vice versa. The best U 3 O 8 obtained from the with ratio mole nitrate/uranium = 1.9, namely uranyl nitrate solution with the feed acidity of 1 N which was added by the lest amount of NH 4 OH. U 3 O 8 microsphere has density 7.06 g/cc (85.62% theoretical density), specific surface area = 6.77 m 2 /g, mean pore radius 20.52 Å, and also total pore volume 6.91x10 -3 cc/g. (author)

  14. Framatome-ANP France UO2 fuel fabrication. Criticality safety analysis in the light of the JCO accident

    International Nuclear Information System (INIS)

    Doucet, M.; Zheng, S.; Mouton, J.; Porte, R.

    2003-01-01

    In France the 1999' Tokai Mura criticality accident in Japan had a big impact on the nuclear fuel manufacturing facility community. Moreover this accident led to a large public discussion about all the nuclear facilities. The French Safety Authorities made strong requirements to the industrials to revisit completely their safety analysis files mainly those concerning nuclear fuels treatments. The FRAMATOME-ANP production of its French low enriched (5 w/o) UO2 fuel fabrication plant (FBFC/Romans) exceeds 1000 metric tons a year. Special attention was given to the emergency evacuation plan that should be followed in case of a criticality accident. If a criticality accident happens, site internal and external radioprotection requirements need to have an emergency evacuation plan showing the different routes where the absorbed doses will be as low as possible for people. The French Safety Authorities require also an update of the old based neutron source term accounting for state of the art methodology. UO2 blenders units contain a large amount of dry powder strictly controlled by moderation; a hypothetical water leakage inside one of these apparatus is simulated by increasing the water content of the powder. The resulted reactivity insertion is performed by several static calculations. The French IRSN/CEA CRISTAL codes are used to perform these static calculations. The kinetic criticality code POWDER simulates the power excursion versus time and determines the consequent total energy source term. MNCP4B performs the source term propagation (including neutrons and gamma) used to determine the isodose curves needed to define the emergency evacuation plant. This paper deals with the approach FRAMATOME-ANP has taken to assess Safety Authorities demands using the more up to date calculation tools and methodology. (author)

  15. Interaction between U/UO2 bilayers and hydrogen studied by in-situ X-ray diffraction

    Science.gov (United States)

    Darnbrough, J. E.; Harker, R. M.; Griffiths, I.; Wermeille, D.; Lander, G. H.; Springell, R.

    2018-04-01

    This paper reports experiments investigating the reaction of H2 with uranium metal-oxide bilayers. The bilayers consist of ≤ 100 nm of epitaxial α-U (grown on a Nb buffer deposited on sapphire) with a UO2 overlayer of thicknesses of between 20 and 80 nm. The oxides were made either by depositing via reactive magnetron sputtering, or allowing the uranium metal to oxidise in air at room temperature. The bilayers were exposed to hydrogen, with sample temperatures between 80 and 200 C, and monitored via in-situ x-ray diffraction and complimentary experiments conducted using Scanning Transmission Electron Microscopy - Electron Energy Loss Spectroscopy (STEM-EELS). Small partial pressures of H2 caused rapid consumption of the U metal and lead to changes in the intensity and position of the diffraction peaks from both the UO2 overlayers and the U metal. There is an orientational dependence in the rate of U consumption. From changes in the lattice parameter we deduce that hydrogen enters both the oxide and metal layers, contracting the oxide and expanding the metal. The air-grown oxide overlayers appear to hinder the H2-reaction up to a threshold dose, but then on heating from 80 to 140 C the consumption is more rapid than for the as-deposited overlayers. STEM-EELS establishes that the U-hydride layer lies at the oxide-metal interface, and that the initial formation is at defects or grain boundaries, and involves the formation of amorphous and/or nanocrystalline UH3. This explains why no diffraction peaks from UH3 are observed.

  16. Contact corrosion measurements on the pair UO 2+ x and carbon steel 1.0330 in brines and bentonite porewater with respect to direct waste disposal

    Science.gov (United States)

    Engelhardt, J.; Marx, G.

    1999-01-01

    Contact corrosion between carbon steel and UO 2 was studied in the MgCl 2 rich Q-brine, in bentonite porewater and in saturated NaCl solution by use of contact potential and contact current measurements. In all solutions the carbon steel dominates the contact potential, so that this potential is near to the rest potential of the carbon steel. Only in solutions without precipitation of iron corrosion products, the presence of metallic iron slightly reduces the UO 2 corrosion rate. If iron corrosion products precipitate, the relevant adsorption of the uranium species will be more effective than any direct cathodic corrosion protection.

  17. Measurements of the effective range of fission fragments in UO2 and the disintegration constant for spontaneous fission of 238U

    International Nuclear Information System (INIS)

    Spaggiari, E.R.V.

    1978-01-01

    The results of measuments of the disintegration constant for spontaneous fission in 238 U are presented, with a discussion on the method used for the detection of fission tracks in muscovite mica. Samples of muscovite mica sandwiched between two natural uranium dioxide cylinders were irradiated with fragments of spontaneous fission and the etched tracks counted with projetion optical microscope. The effective thickness of the UO 2 layer which contributed to the observed tracks was measured through irradiation of mica samples, in contact with the UO 2 cylinder with 14,0 MeV neutrons from a (d,t) reaction. (Author) [pt

  18. Thermodynamics of solvent extraction on (C8H17)3N-C6H5CH3-UO2Cl2-HCl system

    International Nuclear Information System (INIS)

    Yigui Li; Jiufang Lu; Xunan Zhou; Teng Teng

    1988-01-01

    Solvent extraction thermodynamics in the system n-trioctylamine-toluene-UO 2 CL 2 -HCl-water was considered. Pitzer equation and improved Frank-Thompson equation were used to calculate coefficients of electrolyte activity in aqueous phase. Activity coefficients of all components in organic phase were measured or calculated. Thermodynamic equilibrium constants of studied system were obtained

  19. Spectral-luminescence properties and energy transfer in Cs4Sm2UO2(P2O7)3 crystals

    International Nuclear Information System (INIS)

    Syt'ko, V.V.; Aleshkevich, N.A.; Pershina, M.Yu.

    1995-01-01

    At the present time a considerable amount of attention has been devoted to the development of new materials with nontraditional properties capable of acting as optical transformers or of expanding the range of the generation characteristics of lasers. In this regard, interest is drawn to double self-activated crystals in which a pair of active ions that implement the transfer of energy of electronic excitation occur in the chemical formula of the compound. Here, it is promising to use materials based on phosphates coactivated by ions of uranyl UO 2 2+ and samarium Sm 3+ . Generation sensitized by Tb 3+ ions and obtained on the transitions 4 G 5/2 →H 7/2 Sm 3+ in TbF 3 -Sm crystals was reported. In this work, the authors investigated vibrational and electronic spectra of uranylpyrophosphates of cesium-samarium Cs 4 Sm 2 UO 2 (P 2 O 7 ) 3 and the transfer of the energy of the electronic excitation *UO 2 2+ + Sm 3+ → UO 2 2+ + *Sm 3+ . The crystals were obtained by solid-phase synthesis and were investigated at temperatures of 77 and 300 K

  20. Steady-state and transient temperature measurements on BWR-type fuel up to 68 MWd/kgUO2 (IFA-533.2)

    International Nuclear Information System (INIS)

    Alvarez, M.T.

    1996-03-01

    The demonstration test IFA-533.2, for re-instrumentation of irradiated rods with fuel thermocouples, has been irradiated in HBWR from March 1992. The objectives of the irradiation include in-pile testing of the re-instrumentation technique and the generation of temperature data, in order to study the fuel thermal behaviour at high burn-up. With this technique, changes in UO 2 thermal conductivity, rim effects and temperature response to fission gas release can be investigated at high irradiation levels. On the other hand, data obtained from this rig enlarge the Halden data base on the thermal behaviour of UO 2 fuel at high burn-up. The rig contains two BWR-type fuel rods, pre-irradiated in IFA-409 to a bum-up of around 44 MWd/kgUO 2 , and re-instrumented with fresh fuel thermocouples. Reliable data could be obtained from one of the thermocouples (TF2), while the other was affected by a bad connection in the in-core plug. This report presents an evaluation of the results in steady-state and transient operation to a burn-up of 68 MWd/kgUO 2 . The data can be interpreted in terms of fuel conductivity degradation and fission gas release. The results are consistent with fuel temperature data obtained in other IFAs. (author)

  1. A comparison of interatomic potentials for modeling tungsten nanocluster structures

    International Nuclear Information System (INIS)

    Hao, Jiannan; Shu, Xiaolin; Jin, Shuo; Zhang, Xuesong; Zhang, Ying; Lu, Guang-Hong

    2017-01-01

    Molecular dynamic simulation is utilized to study the nanocluster and the fuzz structure on the PFM surface of tungsten. The polyhedral and linear cluster structures based on the icosahedron, cuboctahedron and rhombic dodecahedron are built up. Three interatomic potentials are used in calculating the relationship between the cluster energy and the number of atoms. The results are compared with first-principles calculation to show each potential’s best application scale. Furthermore, the transition between the icosahedral and the cuboctahedral clusters is observed in molecular dynamic simulation at different temperatures, which follows a critical curve for different numbers of atoms. The linear structures are proved to be stable at experimental temperatures by thermodynamics. The work presents a selection of interatomic potentials in simulating tungsten cluster systems and helps researchers understand the growth and evolution laws of clusters and the fuzz-like structure formation process in fusion devices.

  2. A comparison of interatomic potentials for modeling tungsten nanocluster structures

    Energy Technology Data Exchange (ETDEWEB)

    Hao, Jiannan; Shu, Xiaolin, E-mail: shuxlin@buaa.edu.cn; Jin, Shuo; Zhang, Xuesong; Zhang, Ying; Lu, Guang-Hong

    2017-02-15

    Molecular dynamic simulation is utilized to study the nanocluster and the fuzz structure on the PFM surface of tungsten. The polyhedral and linear cluster structures based on the icosahedron, cuboctahedron and rhombic dodecahedron are built up. Three interatomic potentials are used in calculating the relationship between the cluster energy and the number of atoms. The results are compared with first-principles calculation to show each potential’s best application scale. Furthermore, the transition between the icosahedral and the cuboctahedral clusters is observed in molecular dynamic simulation at different temperatures, which follows a critical curve for different numbers of atoms. The linear structures are proved to be stable at experimental temperatures by thermodynamics. The work presents a selection of interatomic potentials in simulating tungsten cluster systems and helps researchers understand the growth and evolution laws of clusters and the fuzz-like structure formation process in fusion devices.

  3. Characterization of hydrogen, nitrogen, oxygen, carbon and sulfur in nuclear fuel (UO2) and cladding nuclear rod materials

    International Nuclear Information System (INIS)

    Crewe, Maria Teresa I.; Lopes, Paula Corain; Moura, Sergio C.; Sampaio, Jessica A.G.; Bustillos, Oscar V.

    2011-01-01

    The importance of Hydrogen, Nitrogen, Oxygen, Carbon and Sulfur gases analysis in nuclear fuels such as UO 2 , U 3 O 8 , U 3 Si 2 and in the fuel cladding such as Zircaloy, is a well known as a quality control in nuclear industry. In UO 2 pellets, the Hydrogen molecule fragilizes the metal lattice causing the material cracking. In Zircaloy material the H2 molecules cause the boiling of the cladding. Other gases like Nitrogen, Oxygen, Carbon and Sulfur affect in the lattice structure change. In this way these chemical compounds have to be measure within specify parameters, these measurement are part of the quality control of the nuclear industry. The analytical procedure has to be well established by a convention of the quality assurance. Therefore, the Oxygen, Carbon, Sulfur and Hydrogen are measured by infrared absorption (IR) and the nitrogen will be measured by thermal conductivity (TC). The gas/metal analyzer made by LECO Co. model TCHEN-600 is Hydrogen, Oxygen and Nitrogen analyzer in a variety of metals, refractory and other inorganic materials, using the principle of fusion by inert gas, infrared and thermo-coupled detector. The Carbon and Sulfur compounds are measure by LECO Co. model CS-400. A sample is first weighed and placed in a high purity graphite crucible and is casted on a stream of helium gas, enough to release the oxygen, nitrogen and hydrogen. During the fusion, the oxygen present in the sample combines with the carbon crucible to form carbon monoxide. Then, the nitrogen present in the sample is analyzed and released as molecular nitrogen and the hydrogen is released as gas. The hydrogen gas is measured by infrared absorption, and the sample gases pass through a trap of copper oxide which converts CO to CO 2 and hydrogen into water. The gases enter the cell where infrared water content is then converted making the measurement of total hydrogen present in the sample. The Hydrogen detection limits for the nuclear fuel is 1 μg/g for the Nitrogen

  4. Development of an interatomic EAM type potential for Zr

    International Nuclear Information System (INIS)

    Pasianot, R.C.; Monti, A.M.

    1996-01-01

    In the present work are developed interatomic potentials of the embedded atom type (EAM) adequate for computer simulation of microstructural defects in the Zr lattice. It is observed that the less repulsive potential agrees better with the experimental data of the self-interstitial relaxation volume and predicts the basal crowdion as the stable configuration, the basal dumbbell having a formation energy slightly higher (0.01 eV). (author). 9 refs., 1 fig., 3 tabs

  5. Distribution of interatomic distances in large metallic clusters

    International Nuclear Information System (INIS)

    Glossman, M.D.; Iniguez, M.P.; Alonso, J.A.

    1992-01-01

    Spherically averaged pseudopotential (SAPS) calculations have been done for Mg n clusters, with n up to 250 within the framework of density functional theory. The electronic structure is computed resorting to the Thomas-Fermi-Dirac-Weizsaecker (TFDW) approximation for the kinetic energy. The equilibrium geometries have been obtained by minimizing the total cluster energy with respect to the atomic positions using the steepest-descent method. The ground state geometries obtained in this way are formed by spherical atomic shells, the number of them increasing with cluster size, up to a number of four for the biggest sizes considered here. An analysis of the distribution of the interatomic distances shows that the more internal is the shell, the more contracted are the interatomic distances. This effect diminishes progressively with increasing cluster size. For the purpose of comparison, similar calculations have been done with Cs n clusters in the same size range, allowing us to reproduce previous results obtained using a more elaborated density functional technique (Kohn-Sham method). The inhomogeneous contraction of interatomic distances then appears as a general fact for simple metallic clusters and not only for alkaline ones. (orig.)

  6. Empirical potential and elasticity theory modelling of interstitial dislocation loops in UO2 for cluster dynamics application

    International Nuclear Information System (INIS)

    Le-Prioux, Arno

    2017-01-01

    During irradiation in reactor, the microstructure of UO 2 changes and deteriorates, causing modifications of its physical and mechanical properties. The kinetic models used to describe these changes such as cluster dynamics (CRESCENDO calculation code) consider the main microstructural elements that are cavities and interstitial dislocation loops, and provide a rather rough description of the loop thermodynamics. In order to tackle this issue, this work has led to the development of a thermodynamic model of interstitial dislocation loops based on empirical potential calculations. The model considers two types of interstitial dislocation loops on two different size domains: Type 1: Dislocation loops similar to Frank partials in F.C.C. materials which are stable in the smaller size domain. Type 2: Perfect dislocation loops of Burgers vector (a/2)(110) stable in the larger size domain. The analytical formula used to compute the interstitial dislocation loop formation energies is the one for circular loops which has been modified in order to take into account the effects of the dislocation core, which are significant at smaller sizes. The parameters have been determined by empirical potential calculations of the formation energies of prismatic pure edge dislocation loops. The effect of the habit plane reorientation on the formation energies of perfect dislocation loops has been taken into account by a simple interpolation method. All the different types of loops seen during TEM observations are thus accounted for by the model. (author) [fr

  7. Advanced fuel cycle for the LWR on a basis of UF6 pyrohydrolysis up to UO2 and vibropack technology

    International Nuclear Information System (INIS)

    Ivanov, V.B.; Mayorshin, A.A.; Sokolovsky, Y.S.; Skiba, O.V.; Porodnov, P.T.; Rybin, D.G.; Chernyshov, V.A.

    1997-01-01

    The traditional circuit of a fuel cycle for thermal neutrons reactors provides conversion of enriched uranium hexafluoride in a press-powder uranium dioxide, using it for manufacturing pellet fuel and subsequently pins. It is known that, each of these stages contains rather plenty of technological and control operations. In SSC RF RIAR the large cycle of studies for improving and simplifying fuel cycle of power reactors is executed. One of studies is devoted to the development of one-stage way of granulated uranium dioxide obtaining by hexafluoride pyrohydrolysis in UO 2 particles boiling layer in a combination with vibropack technology for pins manufacture of fast and thermal neutrons reactors. Reduction of time that conversion of uranium hexafluoride into uranium dioxide takes in a combination with potential advantages of vibropacking: 1) minimum quantity of technological and control operations; 2) possibility of introducing various component (getter, burning out absorber) at a stage of preparation of fuel portion; 3) possibility of using fuel on the basis of mechanical mixes and, if it is necessary, distribution of components profiled along length of the fuel column. (J.P.N.)

  8. Control rod effects on reaction rate distributions in tight pitched PuO2-UO2 fuel assembly

    International Nuclear Information System (INIS)

    Gil, Choong-Sup; Okumura, Keisuke; Ishiguro, Yukio

    1991-11-01

    Investigations were made for the heterogeneity effects caused by insertion or withdrawal of a B 4 C control rod on fine structure of reaction rates distributions in a tight pitched PuO 2 -UO 2 fuel assembly. Analysis was carried out by using the VIM and SRAC codes with the libraries based on JENDL-2 for the hexagonal fuel assembly basically corresponding to the PROTEUS-LWHCR experimental core. The reaction rates are affected more remarkably by the withdrawal of the control rod rather than its insertion. The changes of the reaction rates were decomposed into three terms of spectrum shifts, the changes of effective cross sections with fine groups, and their higher order components. From the analysis, it is concluded that most changes of reaction rates are caused by spectral shifts. The SRAC code with fine group constants can predict the distribution of reaction rates and their ratios with the accuracy of about 5 % except for the values related to Pu-242 capture rate, as compared with the VIM results. To increase the accuracy, it is necessary to generate the effective cross sections of the fuel near control rods with consideration of the heterogeneities in the fuel assembly. (author)

  9. Characterization of selenium in UO2 spent nuclear fuel by micro X-ray absorption spectroscopy and its thermodynamic stability.

    Science.gov (United States)

    Curti, E; Puranen, A; Grolimund, D; Jädernas, D; Sheptyakov, D; Mesbah, A

    2015-10-01

    Direct disposal of spent nuclear fuel (SNF) in deep geological formations is the preferred option for the final storage of nuclear waste in many countries. In order to assess to which extent radionuclides could be released to the environment, it is of great importance to understand how they are chemically bound in the waste matrix. This is particularly important for long-lived radionuclides such as (79)Se, (129)I, (14)C or (36)Cl, which form poorly sorbing anionic species in water and therefore migrate without significant retardation through argillaceous repository materials and host rocks. We present here X-ray absorption spectroscopic data providing evidence that in the investigated SNF samples selenium is directly bound to U atoms as Se(-II) (selenide) ion, probably replacing oxygen in the cubic UO2 lattice. This result is corroborated by a simple thermodynamic analysis, showing that selenide is the stable form of Se under reactor operation conditions. Because selenide is almost insoluble in water, our data indirectly explain the unexpectedly low release of Se in short-term aqueous leaching experiments, compared to iodine or cesium. These results have a direct impact on safety analyses for potential nuclear waste repository sites, as they justify assuming a small fractional release of selenium in performance assessment calculations.

  10. Feasibility to convert an advanced PWR from UO2 to a mixed (U,Th)O2 core

    International Nuclear Information System (INIS)

    Stefani, Giovanni Laranjo de; Maiorino, José Rubens; Moreira, João Manoel de Losada; Santos, Thiago Augusto dos; Rossi, Pedro Carlos Russo

    2017-01-01

    This work presents the neutronics and thermal hydraulics feasibility to convert the UO2 core of the Westinghouse AP1000 in a (U-Th)O 2 core, rather than the traditional uranium dioxide, for the purpose of reducing long-lived actinides, especially plutonium, and generates a stock pile of 233 U, which could in the future be used in advanced fuel cycles, in a more sustainable process and taking advantage of the large stock of thorium available on the planet and especially in Brazil. The reactor chosen as reference was the AP1000, which is considered to be one of the most reliable and modern reactor of the current Generation III, and its similarity to the reactors already consolidated and used in Brazil for electric power generation. The results show the feasibility and potentiality of the concept, without the necessity of changes in the core of the AP1000, and even with advantages over this. The neutron calculations were made by the SERPENT code. The results provided a maximum linear power density lower than the AP1000, favoring safety. In addition, the delayed neutron fraction and the reactivity coefficients proved to be adequate to ensure the safety of the concept. The results show that a production of about 260 Kg of 233 U per cycle is possible, with a minimum production of fissile plutonium that favors the use of the concept in U-Th cycles. (author)

  11. Study of UO2 mechanical behaviour implanted with helium ions using X-ray micro-diffraction and mechanical modeling

    International Nuclear Information System (INIS)

    Ibrahim, Marcelle

    2015-01-01

    In order to study the mechanical behavior of nuclear fuel during direct long term storage, UO 2 polycrystals were implanted with Helium ions at a thin surface layer (1 μm approximately), which leads to stress and strain fields in the layer. Strains were measured, at the grains scale, by X-ray micro-diffraction, using synchrotron radiation (ESRF). Image analysis methods were developed for an automatic analysis of the large number of diffraction patterns. Applying statistical tools to Laue patterns allows an automatic detection of low quality images, and enhances the measurement precision. At low layer thickness, the mechanical interaction between grains can be neglected. At higher thickness, experimental results showed a higher mechanical interaction near grain boundaries that can be modeled using finite elements method. Geostatistical tools were used to quantify these interactions. The swelling and the elastic constants in the implanted layer can be estimated through the measured strains on a large number of grains with different orientations. This work allows the determination of the swelling of nuclear fuel in irradiation conditions, as well as the modification of its elastic properties. (author) [fr

  12. Sequence/structure selective thermal and photochemical cleavage of yeast-tRNA(Phe) by UO(2)2+

    DEFF Research Database (Denmark)

    Nielsen, Peter E.; Møllegaard, N E

    1997-01-01

    The uranyl(VI) ion, UO(2)2+, cleaves yeast tRNA(Phe) both thermally and photochemically. Photochemical cleavage takes place at all positions but exhibits maxima at G10, G18, G30, A38, C49 and A62. Furthermore, in the presence of stoichiometric concentrations of citrate, the cleavage is generally...... suppressed except that strong cleavage at positions G10 and C48-U50 persists, indicating the presence of a high-affinity metal-ion binding site. It is proposed that these photocleavage sites reflect the tertiary structure of the yeast tRNA(Phe) molecule in terms of D-loop/T-loop interaction and anticodon...... loop conformation and that uranyl-mediated photocleavage of RNA may be used as a probe of RNA tertiary structure, and in particular for identifying binding sites for divalent metal ions. Thus a high-affinity metal-ion binding site is inferred in the "central pocket" formed by the D...

  13. Chlorination of UO2, PuO2, and rare-earth oxides using ZrCl4

    International Nuclear Information System (INIS)

    Sakamura, Yoshiharu; Inoue, Tadashi; Iwai, Takashi; Moriyama, Hirotake

    2001-01-01

    A new chlorination method using ZrCl 4 , which has a high reactivity with oxygen, has been investigated for more efficient oxide treatment. After actinide oxides are chlorinated and dissolved in a molten salt bath, actinide metals can be selectively collected using the electrorefining process. This process is well suited for pyrochemical reprocessing of metallic fuels. In LiCl-KCI eutectic melts, rare-earth oxides (Y 2 O 3 , La 2 O 3 , CeO 2 , and Nd 2 O 3 ) and actinide oxides (UO 2 and PuO 2 ) were chlorinated by adding ZrCl 4 . As a result, rare-earth and actinide elements were dissolved into the salt as trivalent ions and ZrO 2 was precipitated. When an excess of ZrCI 4 was added, oxides in powder form were completely chlorinated in five hours. It was demonstrated that the ZrCI 4 chlorination method, free from corrosive gas such as chlorine, was very simple and useful. (author)

  14. An interatomic potential for studying CuZr bulk metallic glasses

    International Nuclear Information System (INIS)

    Paduraru, A.; Kenoufi, A.; Bailey, N.P.; Schioetz, J.

    2007-01-01

    Glass forming ability has been found in only a small number of binary alloys, one being CuZr. In order to simulate this glass, we fitted an interatomic potential within Effective Medium Theory (EMT). For this purpose we use basic properties of the B2 crystal structure as calculated from Density Functional Theory (DFT) or obtained from experiments. We then performed Molecular Dynamics (MD) simulations of the cooling process and studied the thermodynamics and structure of CuZr glass. We find that the potential gives a good description of the CuZr glass, with a glass transition temperature and elastic constants close to the experimental values. The local atomic order, as witnessed by the radial distribution function, is also consistent with similar experimental data. (Abstract Copyright [2007], Wiley Periodicals, Inc.)

  15. Retrieval of interatomic separations of molecules from laser-induced high-order harmonic spectra

    International Nuclear Information System (INIS)

    Le, Van-Hoang; Nguyen, Ngoc-Ty; Jin, C; Le, Anh-Thu; Lin, C D

    2008-01-01

    We illustrate an iterative method for retrieving the internuclear separations of N 2 , O 2 and CO 2 molecules using the high-order harmonics generated from these molecules by intense infrared laser pulses. We show that accurate results can be retrieved with a small set of harmonics and with one or few alignment angles of the molecules. For linear molecules the internuclear separations can also be retrieved from harmonics generated using isotropically distributed molecules. By extracting the transition dipole moment from the high-order harmonic spectra, we further demonstrated that it is preferable to retrieve the interatomic separation iteratively by fitting the extracted dipole moment. Our results show that time-resolved chemical imaging of molecules using infrared laser pulses with femtosecond temporal resolutions is possible

  16. Retrieval of interatomic separations of molecules from laser-induced high-order harmonic spectra

    Energy Technology Data Exchange (ETDEWEB)

    Le, Van-Hoang; Nguyen, Ngoc-Ty [Department of Physics, University of Pedagogy, 280 An Duong Vuong, Ward 5, Ho Chi Minh City (Viet Nam); Jin, C; Le, Anh-Thu; Lin, C D [J. R. Macdonald Laboratory, Department of Physics, Kansas State University, Manhattan, KS 66506 (United States)

    2008-04-28

    We illustrate an iterative method for retrieving the internuclear separations of N{sub 2}, O{sub 2} and CO{sub 2} molecules using the high-order harmonics generated from these molecules by intense infrared laser pulses. We show that accurate results can be retrieved with a small set of harmonics and with one or few alignment angles of the molecules. For linear molecules the internuclear separations can also be retrieved from harmonics generated using isotropically distributed molecules. By extracting the transition dipole moment from the high-order harmonic spectra, we further demonstrated that it is preferable to retrieve the interatomic separation iteratively by fitting the extracted dipole moment. Our results show that time-resolved chemical imaging of molecules using infrared laser pulses with femtosecond temporal resolutions is possible.

  17. Interatomic Potential to Simulate Radiation Damage in Fe-Cr Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Bonny, G.; Pasianot, R.; Terentyev, D.; Malerba, L.

    2011-03-15

    The report presents an Fe-Cr interatomic potential to model high-Cr ferritic alloys. The potential is fitted to thermodynamic and point-defect properties obtained from density functional theory (DFT) calculations and experiments. The developed potential is also benchmarked against other potentials available in literature. It shows particularly good agreement with the DFT obtained mixing enthalpy of the random alloy, the formation energy of intermetallics and experimental excess vibrational entropy and phase diagram. In addition, DFT calculated point-defect properties, both interstitial and substitutional, are well reproduced, as is the screw dislocation core structure. As a first validation of the potential, we study the precipitation hardening of Fe-Cr alloys via static simulations of the interaction between Cr precipitates and screw dislocations. It is concluded that the description of the dislocation core modification near a precipitate might have a significant influence on the interaction mechanisms observed in dynamic simulations.

  18. Materials specific work at Forschungszentrum Karlsruhe and in cooperation with the industrial partners ALKEM and Interatom for the development of nuclear oxide fuels for fission reactors; Materialspezifische Arbeiten im Forschungszentrum Karlsruhe und in Kooperation mit den Industriepartnern ALKEM und Interatom zur Entwicklung oxidischer Kernbrennstoffe fuer Spaltungsreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Kleykamp, H.; Muehling, G.

    2005-09-15

    The fabrication of uranium-plutonium oxide fuel started in Forschungszentrum Karlsruhe and at ALKEM company to begin for the criticality experiments in the SNEAK reactor and subsequently for stationary fuel pin irradiations in the FR2, BR2, DFR, Rapsodie, Phenix and KNK II reactors. The production methods comprised first the mechanical blending of UO2 and PuO2 followed by direct pressing and sintering of the pellets, later the advanced methods such as optimized comilling and ammonium uranyl plutonyl coprecititation. The fabrication of pellets was described in the main, further the alternative fuel pin manufacturing processes by vibrational compaction and hot-impact densification were discussed. The first capsule and pin irradiations in the FR2 and BR2 reactors contributed to the assessment of the maximum operation parameters within the fuel pin development such as linear heat rating, cladding temperature and burnup. Subsequently, small-bundle and largebundle irradiations were made in fast reactors in cooperation with Interatom company in order to verify the specifications for the commercial fast reactor SNR 300. Milestones were the maximum burnup of 175 GWd/t metal, corresponding 18.6 % of the heavy atoms, obtained in one of the KNK II fuel pin assemblies, and the displacement rates in the cladding materials of 140 dpa NRT attained in the Phenix reactor. Higher implications gained later the stationary irradiations of defected mixed-oxide pins, the mild fuel pin transient operations, the local blockage experiments and the severe hypothetic accidents in the respective Siloe, HFR, BR2 and CABRI reactors. These experiments were made solely in international partnership. Further activities were the chemical analyses of solid residues and coprecipitations of irradiated mixed-oxide fuels in the head-end of the reprocessing. All these actions were coordinated in the then fast breeder project. Furthermore, irradiated fuels and fuel pins of other reactor types were

  19. A modified Embedded-Atom Method interatomic potential for uranium-silicide

    Science.gov (United States)

    Beeler, Benjamin; Baskes, Michael; Andersson, David; Cooper, Michael W. D.; Zhang, Yongfeng

    2017-11-01

    Uranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). This uranium alloy fuel benefits from higher thermal conductivity and higher fissile density compared to uranium dioxide (UO2). In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multiscale modeling efforts are underway to address this gap in knowledge. In this study, a semi-empirical modified Embedded-Atom Method (MEAM) potential is presented for the description of the U-Si system. The potential is fitted to the formation energy, defect energies and structural properties of U3Si2. The primary phase of interest (U3Si2) is accurately described over a wide temperature range and displays good behavior under irradiation and with free surfaces. The potential can also describe a variety of U-Si phases across the composition spectrum.

  20. HTGR Fuel Recycle Development Program (189a OHO45). Fuel refabrication, Task 500. Rate-controlling factors in the carbothermic preparation of UO2--UC2--C microspheres

    International Nuclear Information System (INIS)

    Stinton, D.P.; Tiegs, S.M.; Lackey, W.J.; Lindemer, T.B.

    1979-01-01

    Rate controlling factors in the conversion of UO 2 + C microspheres to UC 2 + C were investigated using a 13-cm-dia fluidized bed furnace. X-ray diffraction, ion microprobe, and microstructural examination revealed that the conversion of UO 2 to UC 2 began at the surface of the microsphere and progressed toward the central unreacted core. Kinetic models for solid state reactions in spheres were evaluated by using quantitative mass spectrometric data on the rae of evolution of carbon monoxide during conversion. This analysis revealed that the rate of conversion was controlled by reaction at the outer surface of the microsphere. Also, decreased partial pressures of carbon monoxide were found to accelerate the rate of reaction

  1. Thermodynamics of substituted Rhodanine II: binary complexes of Th(IV), UO2(II), Ce(III), and La(III) with 3-benzamidorhodanine and its derivatives

    International Nuclear Information System (INIS)

    El-Bindary, A.A.; Shehatta, I.

    1994-01-01

    Th(IV), UO 2 (II), Ce(III) and La(III) chelates with 3-benzamidorhodanine and its derivatives have been investigated potentiometrically in 0.1 M KCl and 20% (v/v) ethanol-water medium. The stability of the formed complexes increases in the order Th(IV) > UO 2 (II) > Ce(III) > La(III). For the same metal ion, the stability of the chelates is found to increase with decreasing temperature, ionic strength, dielectric constant of the medium and by increasing the electron repelling property of the substituent. The thermodynamic parameters (ΔG, ΔH and ΔS) for complexation are evaluated and discussed. The formation of the complexes has been found to be spontaneous, exothermic and entropically favourable. (author)

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

    Science.gov (United States)

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

    2017-12-01

    In the hypothesis of direct disposal of spent fuel in a geological nuclear waste repository, interactions between the fuel mainly composed of UO2 and its environment must be understood. The dissolution rate of the UO2 matrix, which depends on the redox conditions on the fuel surface, will have a major impact on the release of radionuclides into the environment. The reducing conditions expected for a geological disposal situation would appear to be favorable as regards the solubility and stability of the UO2 matrix, but may be disturbed on the surface of irradiated fuel. In particular, the local redox conditions will result from a competition between the radiolysis effects of water under alpha irradiation (simultaneously producing oxidizing species like H2O2, hydrogen peroxide, and reducing species like H2, hydrogen) and those of redox active species from the environment. In particular, Fe2+, a strongly reducing aqueous species coming from the corrosion of the iron canister or from the host rock, could influence the dissolution of the fuel matrix. The effect of iron on the oxidative dissolution of UO2 was thus investigated under the conditions of the French disposal site, a Callovian-Oxfordian clay formation chosen by the French National Radioactive Waste Management Agency (Andra), here tested under alpha irradiation. For this study, UO2 fuel pellets doped with a radioactive alpha emitter (238/239Pu) were leached in synthetic Callovian-Oxfordian groundwater (representative of the French waste disposal site groundwater) in the presence of a metallic iron foil to simulate the steel canister. The pellets had varying levels of alpha activity, in order to modulate the concentrations of species produced by water radiolysis on the surface and to simulate the activity of aged spent fuel after 50 and 10,000 years of alpha radioactivity decay. The experimental data showed that whatever the sample alpha radioactivity, the presence of iron inhibits the oxidizing dissolution of

  3. Radial distribution of UO2 and Gd2O3 in fuel cells of a BWR Reactor

    International Nuclear Information System (INIS)

    Montes, J.L.; Ortiz, J.J.; Perusquia del C, R.; Francois, J.L.; Martin del Campo M, C.

    2008-01-01

    The fuel system that is used at the moment in a power plant based on power reactors BWR, includes as much like the one of its substantial parts to the distribution of the fissile materials like a distribution of burnt poisons within each one of the cells which they constitute the fuel assemblies, used for the energy generation. Reason why at the beginning of a new operation cycle in a reactor of this type, the reactivity of the nucleus should be compensated by the exhaustion of the assemblies that it moves away of the nucleus for their final disposition. This compensation is given by means of the introduction of the recharge fuel, starting from the UO 2 enriched in U 2 35, and of the Gadolinium (Gd 2 O 3 ). The distribution of these materials not only defines the requirements of energy generation, but in certain measures also the form in that the margins will behave to the limit them thermal during the operation of the reactor. These margins must be taken into account for the safe and efficient extraction of the energy of the fuel. In this work typical fuel cells appear that are obtained by means of the use of a emulation model of an ants colony. This model allows generating from a possible inventory of values of enrichment of U 2 35, as well as of concentration of Gadolinium a typical fuel cell, which consists of an arrangement of lOxlO rods, of which 92 contain U 2 35, some of these rods contain a concentration of Gd 2 O 3 and 8 of the total contain only water. The search of each cell finishes when the value of the Local Peak Power Factor (LPPF) in the cell reaches a minimal value, or when a pre established value of iterations is reached. The cell parameters are obtained from the results of the execution of the code HELIOS, which incorporates like a part integral of the search algorithm. (Author)

  4. Chemical speciation of uranium(VI) in marine environments: complexation of calcium and magnesium ions with [(UO2)(CO3)3]4- and the effect on the extraction of uranium from seawater

    International Nuclear Information System (INIS)

    Endrizzi, Francesco; Rao, Linfeng

    2014-01-01

    The interactions of Ca 2+ and Mg 2+ with [UO 2 (CO 3 ) 3 ] 4- were studied by calcium ion selective electrode potentiometry and spectrophotometry. The stability constants of ternary Ca-UO 2 -CO 3 and Mg-UO 2 -CO 3 complexes were determined with calcium ion selective electrode potentiometry and optical absorption spectrophotometry, respectively. The enthalpies of complexation for two successive complexes, [CaUO 2 (CO 3 ) 3 ] 2- and [Ca 2 UO 2 (CO 3 ) 3 ](aq), were determined for the first time by microcalorimetry. The data help to revise the speciation of uranium(VI) species under seawater conditions. In contrast to the previously accepted assumption that the highly negatively charged [UO 2 (CO 3 ) 3 ] 4- is the dominant species, the revised speciation indicates that the dominant aqueous uranium(VI) species under seawater conditions is the neutral [Ca 2 UO 2 (CO 3 ) 3 ](aq). The results have a significant impact on the strategies for developing efficient sorption processes to extract uranium from seawater. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  5. Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-MOX, R2-UO2 and MORGANE/R configurations.

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Z.; Klann, R. T.; Nuclear Engineering Division

    2007-08-03

    An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R2-UO2 and MORGANE/R core configuration were completed. The calculation model was generated using the lattice physics code DRAGON. In addition, an initial comparison of calculated values to experimental measurements was performed based on preliminary results for the R1-MOX configuration.

  6. Thermal ionization and plasma state of high temperature vapor of UO2, Cs, and Na: Effect on the heat and radiation transport properties of the vapor phase

    International Nuclear Information System (INIS)

    Karow, H.U.

    1979-01-01

    The paper deals with the question how far the thermophysical state and the convective and radiative heat transport properties of vaporized reactor core materials are affected by the thermal ionization existing in the actual vapor state. The materials under consideration here are: nuclear oxide fuel (UO 2 ), Na (as the LMFBR coolant material), and Cs (alkaline fission product, partly retained in the fuel of the core zone). (orig./RW) [de

  7. Study by electronic structure calculations of the radiation damage in the UO2 nuclear fuel: behaviour of the point defects and fission gases

    International Nuclear Information System (INIS)

    Vathonne, Emerson

    2014-01-01

    Uranium dioxide (UO 2 ) is worldwide the most widely used fuel in nuclear plants in the world and in particular in pressurized water reactors (PWR). In-pile the fission of uranium nuclei creates fission products and point defects in the fuel. The understanding of the evolution of these radiation damages requires a multi-scale modelling approach of the nuclear fuel, from the scale of the pellet to the atomic scale. We used an electronic structure calculation method based on the density functional theory (DFT) to model radiation damage in UO 2 at the atomic scale. A Hubbard-type Coulomb interaction term is added to the standard DFT formalism to take into account the strong correlations of the 5f electrons in UO 2 . This method is used to study point defects with various charge states and the incorporation and diffusion of krypton in uranium dioxide. This study allowed us to obtain essential data for higher scale models but also to interpret experimental results. In parallel of this study, three ways to improve the state of the art of electronic structure calculations of UO 2 have been explored: the consideration of the spin-orbit coupling neglected in current point defect calculations, the application of functionals allowing one to take into account the non-local interactions such as van der Waals interactions important for rare gases and the use of the Dynamical Mean Field Theory combined to the DFT method in order to take into account the dynamical effects in the 5f electron correlations. (author) [fr

  8. Neutron flux depression in the UO2-PuO2 (15 to 30%) fuel rods from IVO-FR2-Vg7-Irradiation experiment

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.; Fernandez Marron, J.L.

    1983-01-01

    The thermal-neutron flux depression within a fuel rod has a great influence on the radial temperature profile of the rod, especially for high enrichment fuel. For this reason, a study was made about the UO 2 -PuO 2 (15 to 30% PuO 2 ) fuel pins for the KfK-JEN joint irradiation program IVO, in the FR2 reactor. Different methods (diffusion, Bonalumi, successive generations) were compared and a new approach (parabolic approximation) was developed. (author)

  9. Neutron Flux Depression in the UO2-PuO2 (15 to 30%) Fuel Rods from IVO-FR2-Vg7-Irradiation Experiment

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.; Fernandez Marron, J.L.

    1983-01-01

    The thermal-neutron flux depression within a fuel rod has a great influence in the radial temperature profile of the rod, especially for high enrichment fuel. For this reason, a study was made about the UO 2 -PUO 2 (15 to 30% PUO 2 ) fuel pins for the KfK-JEN joint irradiation program IVO, in the FR2 reactor. Different methods (diffusion, Bonalumi, successive generations) were compared and a new approach (parabolic approximation) was developed. (Author) 22 refs

  10. Synthesis and crystal structure of Na6[(UO2)3O(OH)3(SeO4)2]2·10H2O

    International Nuclear Information System (INIS)

    Baeva, E.Eh.; Serezhkina, L.B.; Virovets, A.V.; Peresypkina, E.V.

    2006-01-01

    The complex Na 6 [(UO 2 ) 3 O(OH) 3 (SeO 4 ) 2 ] 2 ·10H 2 O (I) is synthesized and studied by monocrystal X-ray diffraction. The compound crystallizes in the orthorhombic crystal system with the unit cell parameters: a=14.2225(7) A, b=18.3601(7) A, c=16.5406(6) A, V=4319.2(3) A 3, Z=4, space group Cmcm, R 1 =0.0406. Compound I is found to be a representative of the crystal-chemical group A 3 M 3 M 3 2 T 2 3 (A=UO 2 2+ , M 3 =O 2- , M 2 =OH - , T 3 =SeO 4 2- ) of the uranyl complexes; it contains layer uranium-containing groups [(UO 2 ) 3 O(OH) 3 (SeO 4 ) 2 ] 3- . These layers are linked to form a three-dimensional cage through bonds formed by the sodium atoms with the oxygen atoms of the uranyl ions and SeO 4 groups that belong to different layers [ru

  11. Electrochemical characterisation of CaCl2 deficient LiCl-KCl-CaCl2 eutectic melt and electro-deoxidation of solid UO2

    Science.gov (United States)

    Sri Maha Vishnu, D.; Sanil, N.; Mohandas, K. S.; Nagarajan, K.

    2016-03-01

    The CaCl2 deficient ternary eutectic melt LiCl-KCl-CaCl2 (50.5: 44.2: 5.3 mol %) was electrochemically characterised by cyclic voltammetry and polarization techniques in the context of its probable use as the electrolyte in the electrochemical reduction of solid UO2 to uranium metal. Tungsten (cathodic polarization) and graphite (anodic polarization) working electrodes were used in these studies carried out in the temperature range 623 K-923 K. The cathodic limit of the melt was observed to be set by the deposition of Ca2+ ions followed by Li+ ions on the tungsten electrode and the anodic limit by oxidation of chloride ions on the graphite electrode (chlorine evolution). The difference between the onset potential of deposition of Ca2+ and Li+ was found to be 0.241 V at a scan rate of 20 mV/s at 623 K and the difference decreased with increase in temperature and vanished at 923 K. Polarization measurements with stainless steel (SS) cathode and graphite anode at 673 K showed the possibility of low-energy reactions occurring on the UO2 electrode in the melt. UO2 pellets were cathodically polarized at 3.9 V for 25 h to test the feasibility of electro-reduction to uranium in the melt. The surface of the pellets was found reduced to U metal.

  12. An X-ray photoelectron spectroscopy study of the products of the interaction of gaseous IrF6 with fine UO2F2

    Directory of Open Access Journals (Sweden)

    Prusakov Vladimir N.

    2007-01-01

    Full Text Available Nuclear fuel reprocessing by fluorination, a dry method of regeneration of spent nuclear fuel, uses UO2F2 for the separation of plutonium from gaseous mixtures. Since plutonium requires special treatment, IrF6 was used as a thermodynamic model of PuF6. The model reaction of the interaction of gaseous IrF6 with fine UO2F2 in the sorption column revealed a change of color of the sorption column contents from pale-yellow to gray and black, indicating the formation of products of such an interaction. The X-ray photoelectron spectroscopy study showed that the interaction of gaseous IrF6 with fine UO2F2 at 125 °C results in the formation of stable iridium compounds where the iridium oxidation state is close to Ir3+. The dependence of the elemental compositions of the layers in the sorption column on the penetration depth of IrF6 was established.

  13. Topologically identical, but geometrically isomeric layers in hydrous α-, β-Rb[UO2(AsO3OH)(AsO2(OH)2)]·H2O and anhydrous Rb[UO2(AsO3OH)(AsO2(OH)2)

    International Nuclear Information System (INIS)

    Yu, Na; Klepov, Vladislav V.; Villa, Eric M.; Bosbach, Dirk; Suleimanov, Evgeny V.; Depmeier, Wulf; Albrecht-Schmitt, Thomas E.; Alekseev, Evgeny V.

    2014-01-01

    The hydrothermal reaction of uranyl nitrate with rubidium nitrate and arsenic (III) oxide results in the formation of polymorphic α- and β-Rb[UO 2 (AsO 3 OH)(AsO 2 (OH) 2 )]·H 2 O (α-, β-RbUAs) and the anhydrous phase Rb[UO 2 (AsO 3 OH)(AsO 2 (OH) 2 )] (RbUAs). These phases were structurally, chemically and spectroscopically characterized. The structures of all three compounds are based upon topologically identical, but geometrically isomeric layers. The layers are linked with each other by means of the Rb cations and hydrogen bonding. Dehydration experiments demonstrate that water deintercalation from hydrous α- and β-RbUAs yields anhydrous RbUAs via topotactic reactions. - Graphical abstract: Three different layer geometries observed in the structures of Rb[UO 2 (AsO 3 OH)(AsO 2 (OH) 2 )] and α- and β- Rb[UO 2 (AsO 3 OH)(AsO 2 (OH) 2 )]·H 2 O. Two different coordination environments of uranium polyhedra (types I and II) are shown schematically on the top of the figure. - Highlights: • Three new uranyl arsenates were synthesized from the hydrothermal reactions. • The phases consist of the topologically identical but geometrically different layers. • Topotactic transitions were observed in the processes of mono-hyrates dehydration

  14. UO2 Fuel pellet impurities, pellet surface roughness and n(18O)/n(16O) ratios, applied to nuclear forensic science

    International Nuclear Information System (INIS)

    Pajo, L.

    2001-01-01

    In the last decade, law enforcement has faced the problem of illicit trafficking of nuclear materials. Nuclear forensic science is a new branch of science that enables the identification of seized nuclear material. The identification is not based on a fixed scheme, but further identification parameters are decided based on previous identification results. The analysis is carried out by using traditional analysis methods and applying modern measurement technology. The parameters are generally not unambiguous and not self-explanatory. In order to have a full picture about the origin of seized samples, several identification parameters should be used together and the measured data should be compared to corresponding data from known sources. A nuclear material database containing data from several fabrication plants is installed for the purpose. In this thesis the use of UO 2 fabrication plant specific parameters, fuel impurities, fuel pellet surface roughness and oxygen isotopic ratio in UO 2 were investigated for identification purposes in nuclear forensic science. The potential use of these parameters as 'fingerprints' is discussed for identification purposes of seized nuclear materials. Impurities of the fuel material vary slightly according to the fabrication method employed and a plant environment. Here the impurities of the seized UO 2 were used in order to have some clues about the origin of the fuel material by comparing a measured data to nuclear database information. More certainty in the identification was gained by surface roughness of the UO 2 fuel pellets, measured by mechanical surface profilometry. Categories in surface roughness between a different fuel element type and a producer were observed. For the time oxygen isotopic ratios were determined by Thermal Ionisation Mass Speckometry (TIMS). Thus a TIMS measurement method, using U 16 O + and U 18 0 + ions, was developed and optimised to achieve precise oxygen isotope ratio measurements for the

  15. Electrochemical characterisation of CaCl2 deficient LiCl–KCl–CaCl2 eutectic melt and electro-deoxidation of solid UO2

    International Nuclear Information System (INIS)

    Sri Maha Vishnu, D.; Sanil, N.; Mohandas, K.S.; Nagarajan, K.

    2016-01-01

    The CaCl 2 deficient ternary eutectic melt LiCl–KCl–CaCl 2 (50.5: 44.2: 5.3 mol %) was electrochemically characterised by cyclic voltammetry and polarization techniques in the context of its probable use as the electrolyte in the electrochemical reduction of solid UO 2 to uranium metal. Tungsten (cathodic polarization) and graphite (anodic polarization) working electrodes were used in these studies carried out in the temperature range 623 K–923 K. The cathodic limit of the melt was observed to be set by the deposition of Ca 2+ ions followed by Li + ions on the tungsten electrode and the anodic limit by oxidation of chloride ions on the graphite electrode (chlorine evolution). The difference between the onset potential of deposition of Ca 2+ and Li + was found to be 0.241 V at a scan rate of 20 mV/s at 623 K and the difference decreased with increase in temperature and vanished at 923 K. Polarization measurements with stainless steel (SS) cathode and graphite anode at 673 K showed the possibility of low–energy reactions occurring on the UO 2 electrode in the melt. UO 2 pellets were cathodically polarized at 3.9 V for 25 h to test the feasibility of electro-reduction to uranium in the melt. The surface of the pellets was found reduced to U metal. - Highlights: • Electrochemically characterized LiCl–KCl–CaCl 2 (50.5: 44.2: 5.3 mol %) melt by CV, LSV and polarization techniques. • Ca 2+ deposits first on tungsten working electrode followed by Li + . Cl − discharges on graphite to liberate chlorine gas. • Surface of UO 2 pellet reduced to U in the melt with low carbon contamination of melt. • Slow reduction of UO 2 due to slow kinetics and low solubility of oxide ions in the low temperature melt.

  16. Measurement of the in-pile core temperature of an EL-4 pencil element, first charge (can of type-347 stainless steel, 0.4 mm thick, UO2 fuel, 11 mm diameter). Determination of the apparent thermal conductivity integral of in-pile UO2

    International Nuclear Information System (INIS)

    Lavaud, B.; Ringot, C.; Vignesoult, N.

    1966-11-01

    The core temperature of a pencil fuel element depends on the thermal conductivity of the UO 2 , and on the UO 2 -can contact. This temperature may be known accurately only if in-pile tests using the actual geometry are carried out. The test described concerns the measurement of the core- temperature of an EL-4 fuel element, first charge, having a stainless steel can. This temperature is measured at the center of the in-pile pencil element using a high-temperature thermocouple (W-Re with Ta sheath). The element is subjected to operating conditions similar to those of EL-4, both for the specific power and the can temperature and for the pressure acting on the can. The specific power is obtained in the EL-3 reactor using a slightly higher enrichment for the UO 2 than that planned for EL-4. The required can temperature and pressure are obtained using a Zircaloy-2 irradiation container filled with NaK, adapted for use in the EL-3 reactor. The core temperatures of the UO 2 , and that of the can surface are measured. The power is calculated from the heat exchanges in the container calibrated in the laboratory. The temperature drop at the UO 2 -can interface is deduced from laboratory measurements carried out under comparable heat flux conditions, and in a gas atmosphere corresponding to the beginning of the life-time of the fuel element. It is possible to draw an integral conductivity curve. It is also possible to check the temperature distribution in the oxide, as deduced from the thermal conductivity integral, by micro-graphic examination of the oxide structure. (authors) [fr

  17. Study and modelling of the in-pile densification of the UO2 and MOx nuclear oxides

    International Nuclear Information System (INIS)

    Boulore, A.

    2001-03-01

    Amongst the many phenomena which take place in the course of the irradiation of UO 2 or (U, Pu)O 2 nuclear fuels, one of them involves the elimination of a fraction of the as-fabricated porosity. In-pile densification or sintering can reach 2.5%, i.e. approximately half the initial volume of pores is likely to disappear. Our literature survey indicates that the amplitude and kinetics of the phenomenon are both heavily dependent on the initial fuel microstructure. Micro-structural characterisation techniques of oxide fuels have therefore been developed in conjunction with quantitative image analysis methods. The ensuing methodology enables a quantitative comparison of micro-structural features in different fuels and has been applied to ascertaining the influence of the local fission rate and temperature on in-pile densification. It is thus revealed that in-pile operation eliminates a significant fraction of pores smaller than 3 microns in diameter. The experimental data generated has been used to set up a semi-empirical and a mechanistic model. The former is based on experimental results and is not essentially predictive. The inability of this model to predict the in-pile densification of oxide fuels is illustrated by the fact that the maximum fraction of pores that disappears is proportional to an empirical function of fission rate, and temperature. The proportionality factor appears to be difficult to correlate quantitatively to any given micro-structural feature. The model has however been applied to the interpretation of an in-pile densification experiment carried out in the Halden reactor (Norway). The latter model is mechanistic, i.e. it is based on the solution to a set of equations that describe the coupled temperature and radiation induced phenomena which occur in-pile. These can broadly be broken down into three categories: the fission fragment-pore interaction, the creation of point defects as the fission fragments slow down, and the diffusion of these

  18. Multipole expansion of the retarded interatomic dispersion energy: the long and the short range behaviour

    NARCIS (Netherlands)

    Michels, M.A.J.; Suttorp, L.G.

    1972-01-01

    The long-range asymptotic expression for the multipole expansion of the retarded interatomic dispersion energy is shown to consist of contributions from electric dipole-dipole, dipole-quadrupole and quadrupole-quadrupole interactions, all varying as the inverse seventh power of the interatomic

  19. Effects of temperature and irradiation on the mobility of Xenon in UO2: Profilometric and microstructural study

    International Nuclear Information System (INIS)

    Marchand, B.

    2012-01-01

    In France, electricity is mainly produced (78%) through the operation of 58 PWRs (Pressurized Water Reactors). During reactor operation, many fission products (FP) are generated in the fuel which is, in most cases, UO 2 enriched to about 4% in 235 U. Among FPs, gaseous fission products as Xenon and Krypton, are abundantly produced (around 15% stable fission products). Because of their chemical nature, those two gases have a very low solubility in the fuel and therefore tend to form bubbles (to minimize surface tension) and can cause pellets swelling. The formed gas can also be released out of the pellet, and lead to a substantial increase in the pressure within the fuel cladding, thereby limiting the energy production. However, migration mechanisms, traditionally studied indirectly by measuring the amount of gas released after irradiation, are not yet fully understood. It is frequently assumed that atomic diffusion is the only mechanism that can lead to a migration of xenon. The objective of this thesis is to provide direct evidence of the different mechanisms controlling the behavior of Xenon during thermal annealing and irradiation. Therefore, we used ion implantation to introduce Xenon in uranium dioxide samples. After implantation, the Xenon distribution follows a quasi-Gaussian concentration profile (variation of the concentration regard to the depth) located in the first 300 nanometers of the sample. We have performed post-implantation annealing at 1400 C and 1600 C in order to study the impact of the temperature, and irradiation with ions to simulate the impact of fission products in the fuel. Subsequently, concentration depth profiles were measured by ion microprobe (SIMS). Although the feasibility of Xenon measurement has been demonstrated in several articles, no concentration profile had so far been presented in the literature because a classical data processing of SIMS data is not suitable in uranium dioxide. Therefore a new data processing software has

  20. Investigating microstructural evolution during the electroreduction of UO2 to U in LiCl-KCl eutectic using focused ion beam tomography

    Science.gov (United States)

    Brown, L. D.; Abdulaziz, R.; Tjaden, B.; Inman, D.; Brett, D. J. L.; Shearing, P. R.

    2016-11-01

    Reprocessing of spent nuclear fuels using molten salt media is an attractive alternative to liquid-liquid extraction techniques. Pyroelectrochemical processing utilizes direct, selective, electrochemical reduction of uranium dioxide, followed by selective electroplating of a uranium metal. Thermodynamic prediction of the electrochemical reduction of UO2 to U in LiCl-KCl eutectic has shown to be a function of the oxide ion activity. The pO2- of the salt may be affected by the microstructure of the UO2 electrode. A uranium dioxide filled "micro-bucket" electrode has been partially electroreduced to uranium metal in molten lithium chloride-potassium chloride eutectic. This partial electroreduction resulted in two distinct microstructures: a dense UO2 and a porous U metal structure were characterised by energy dispersive X-ray spectroscopy. Focused ion beam tomography was performed on five regions of this electrode which revealed an overall porosity ranging from 17.36% at the outer edge to 3.91% towards the centre, commensurate with the expected extent of reaction in each location. The pore connectivity was also seen to reduce from 88.32% to 17.86% in the same regions and the tortuosity through the sample was modelled along the axis of propagation of the electroreduction, which was seen to increase from a value of 4.42 to a value of infinity (disconnected pores). These microstructural characteristics could impede the transport of O2- ions resulting in a change in the local pO2- which could result in the inability to perform the electroreduction.

  1. UO2-7%Gd2O3 fuel process development by mechanical blending with reprocessing of waste products and usage of densification additive

    International Nuclear Information System (INIS)

    Santos, Lauro Roberto dos

    2009-01-01

    In the nuclear fuel cycle, reprocessing and storage of 'burned' fuels, either temporary or permanent, demand high investments and, in addition, can potentially generate environmental problems. A strategy to decrease these problems is to adopt measures to reduce the amount of waste generated. The usage of integrated burnable poison based on gadolinium is a measure that contributes to achieve this goal. The reason to use burnable poison is to control the neutron population in the reactor during the early life of the fresh reactor core or the beginning of each recharging fuel cycle, extending its cycle duration. Another advantage of using burnable poison is to be able to operate the reactor with higher burning rate, optimizing the usage of the fuel. The process of manufacturing UO 2 -Gd 2 O 3 integrated burnable fuel poison generates waste that, as much as possible, needs to be recycled. Blending of Gd 2 O 3 in UO 2 powder requires the usage of a special additive to achieve the final fuel pellet specified density. The objective of this work is to develop the process of obtaining UO 2 - 7% Gd 2 O 3 integrated burnable poison using densification additives, aluminum hydroxide (Al(OH)3), and reprocessing manufacturing waste products by mechanical blending. The content of 7%- Gd 2 O 3 is based on commercial PWR reactor fuels - Type Angra 2. The results show that the usage of Al(OH) 3 as an additive is a very effective choice that promotes the densification of fuel pellets with recycle up to 10%. Concentrations of 0,20 % of Al(OH) 3 were found to be the indicated amount on an 7 industrial scale, specially when the recycled products come from U 3 O 8 obtained by calcination of sintered pellets. This is particularly interesting because it is following the steps of sintering and rectifying of the pellets, which is generating the largest amounts of recycled material. (author)

  2. A study on etching of UO2, Co, and Mo surface with R.F. plasma using CF4 and O2

    International Nuclear Information System (INIS)

    Kim, Yong Soo; Seo, Yong Dae

    2003-01-01

    Recently dry decontamination/surface-cleaning technology using plasma etching has been focused in the nuclear industry. In this study, the applicability of this new dry processing technique are experimentally investigated by examining the etching reaction of UO 2 , Co, and Mo in r.f. plasma with the etchant gas of CF 4 /O 2 mixture. UO 2 is chosen as a representing material for uranium and TRU (TRans-Uranic) compounds while metallic Co and Mo are selected because they are the principal contaminants in the used metallic nuclear components such as valves and pipes made of stainless steel or Inconel. Results show that in all cases maximum etching rate is achieved when the mole fraction of O 2 in CF 4 /O 2 mixture gas is 20%, regardless of temperature and r.f. power. In case of UO 2 , the highest etching reaction rate is greater than 1000 monolayers/min. at 370 .deg. C under 150 W r.f. power which is equivalent to 0.4 μm/min. As for Co, etching reaction begins to take place significantly when the temperature exceeds 350 .deg. C. Maximum etching rate achieved at 380 .deg. C is 0.06 μm/min. Mo etching reaction takes place vigorously even at relatively low temperature and the reaction rate increases drastically with increasing temperature. Highest etching rate at 380 .deg. C is 1.9 μm /min. According to OES (Optical Emission Spectroscopy) and AES (Auger Electron Spectroscopy) analysis, primary reaction seems to be a fluorination reaction, but carbonyl compound formation reaction may assist the dominant reaction, especially in case of Co and Mo. Through this basic study, the feasibility and the applicability of plasma decontamination technique are demonstrated

  3. High temperature investigation of the solid/liquid transition in the PuO2-UO2-ZrO2 system

    Science.gov (United States)

    Quaini, A.; Guéneau, C.; Gossé, S.; Sundman, B.; Manara, D.; Smith, A. L.; Bottomley, D.; Lajarge, P.; Ernstberger, M.; Hodaj, F.

    2015-12-01

    The solid/liquid transitions in the quaternary U-Pu-Zr-O system are of great interest for the analysis of core meltdown accidents in Pressurised Water Reactors (PWR) fuelled with uranium-dioxide and MOX. During a severe accident the Zr-based cladding can become completely oxidised due to the interaction with the oxide fuel and the water coolant. In this framework, the present analysis is focused on the pseudo-ternary system UO2-PuO2-ZrO2. The melting/solidification behaviour of five pseudo-ternary and one pseudo-binary ((PuO2)0.50(ZrO2)0.50) compositions have been investigated experimentally by a laser heating method under pre-set atmospheres. The effects of an oxidising or reducing atmosphere on the observed melting/freezing temperatures, as well as the amount of UO2 in the sample, have been clearly identified for the different compositions. The oxygen-to-metal ratio is a key parameter affecting the melting/freezing temperature because of incongruent vaporisation effects. In parallel, a detailed thermodynamic model for the UO2-PuO2-ZrO2 system has been developed using the CALPHAD method, and thermodynamic calculations have been performed to interpret the present laser heating results, as well as the high temperature behaviour of the cubic (Pu,U,Zr)O2±x-c mixed oxide phase. A good agreement was obtained between the calculated and experimental data points. This work enables an improved understanding of the major factors relevant to severe accident in nuclear reactors.

  4. UO2-7%Gd2O3 fuel process development by mechanical blending with reprocessing of waste products and usage of densification additive

    International Nuclear Information System (INIS)

    Santos, Lauro Roberto dos

    2009-01-01

    In the nuclear fuel cycle, reprocessing and storage of 'burned' fuels, either temporary or permanent, demand high investments and, in addition, can potentially generate environmental problems. A strategy to decrease these problems is to adopt measures to reduce the amount of waste generated. The usage of integrated burnable poison based on gadolinium is a measure that contributes to achieve this goal. The reason to use burnable poison is to control the neutron population in the reactor during the early life of the fresh reactor core or the beginning of each recharging fuel cycle, extending its cycle duration. Another advantage of using burnable poison is to be able to operate the reactor with higher burning rate, optimizing the usage of the fuel. The process of manufacturing UO 2 -Gd 2 O 3 integrated burnable fuel poison generates waste that, as much as possible, needs to be recycled. Blending of Gd 2 O 3 in UO 2 powder requires the usage of a special additive to achieve the final fuel pellet specified density. The objective of this work is to develop the process of obtaining UO 2 - 7% Gd 2 O 3 integrated burnable poison using densification additives, aluminum hydroxide (Al(OH) 3 ), and reprocessing manufacturing waste products by mechanical blending. The content of 7%- Gd 2 O 3 is based on commercial PWR reactor fuels - Type Angra 2. The results show that the usage of Al(OH) 3 as an additive is a very effective choice that promotes the densification of fuel pellets with recycle up to 10%. Concentrations of 0,20 % of Al(OH) 3 were found to be the indicated amount on an industrial scale, specially when the recycled products come from U 3 O 8 obtained by calcination of sintered pellets. This is particularly interesting because it is following the steps of sintering and rectifying of the pellets, which is generating the largest amounts of recycled material. (author)

  5. Characterization and property evaluation of (Th–3.75U)O2+x fuel pellets fabricated by impregnated agglomerate pelletization (IAP) process

    International Nuclear Information System (INIS)

    Baghra, C.; Sathe, D.B.; Prakash, A.; Mishra, A.K.; Afzal, Mohd; Panakkal, J.P.; Kamath, H.S.

    2013-01-01

    Highlights: ► Characterization of (Th–3.75%U)O 2+x pellet fabricated by a new technique i.e., impregnated agglomerate pelletization (IAP). ► Comparison with pellets fabricated by powder to pellet (POP) and coated agglomerate pelletization processes. ► Solid solution was more complete in POP and IAP pellets. ► Uranium distribution was less uniform in the pellets fabricated by the CAP route. -- Abstract: Impregnated agglomerate pelletization (IAP) process has been developed by Advanced Fuel Fabrication Facility (AFFF), Tarapur for the fabrication of (Th–3.75U)O 2+x mixed oxide fuel for Indian advanced heavy water reactors (AHWR). In this process, ThO 2 spheroids were impregnated with uranyl nitrate solution and the resultant mixture was compacted to form green pellets which were sintered in the oxidizing atmosphere to obtain high density (Th–3.75%U)O 2+x pellets. An attempt has been made in this paper to characterize the pellets fabricated by IAP route. The characterization of the sintered IAP pellets was made by X-ray diffraction, EPMA, alpha autoradiography, chemical analysis, O/M ratio, immersion density and optical microscopy. The characteristics of (Th–U)O 2 pellets fabricated by impregnated agglomerate pelletization (IAP) were also compared with those fabricated by coated agglomerate pelletization (CAP), and conventional powder pellet (POP) processes. In case of pellets fabricated by IAP route, XRD data showed the presence of single fluorite phase. The uranium concentration and grain size distribution were found to be uniform throughout the pellet. These phase characteristics were also found to be more uniform in pellets made by both IAP and POP processes as compared to pellets fabricated using the CAP process

  6. Analysis of the heat and mass transfer processes of a UO2 bubble in sodium for the Fuel Aerosol Simulant Test (FAST)

    International Nuclear Information System (INIS)

    Tobias, M.L.

    1979-01-01

    The anticipated behavior of uranium oxide vapor bubbles produced by the capacitor discharge vaporization (CDV) method in the Fuel Aerosol Simulant Test (FAST) Facility is discussed on the basis of relatively simple physical models. Results of calculations for the rate of bubble rise and for heat and mass transfer rates are presented. Parametric studies indicate that future analysis efforts should emphasize the diffusion condensation process and the loss of heat from the bubble by radiation. Transfer of heat in the surrounding sodium is rapid enough that simplified models should be adequate. No important effects were noted in connection with bubble depth, initial quantity of UO 2 , or initial superheat

  7. Physiochemical studies on the composition and stability of the complexes of VO(II) and UO2(II) with L-lysinemono-hydrochloride

    International Nuclear Information System (INIS)

    Saxena, R.S.; Dhawan, S.K.

    1981-01-01

    The VO(II) and UO 2 (II) form 1:1 and 1:2 complexes with L-lysine monohydrochloride [NH 2 (CH 2 ) 4 CH(NH 2 ) COOH HCl]. Calvin and Melchior's extension of Bjerrum method as modified by Irving and Rossotti have been used for the determination of stability constant at 30 degC and 40 degC at a constant ionic strength (μ=0.1M NaClO 4 ). These values were further refined by 'Least Square Method' and 'Schroder's Convergence Formula'. The thermodynamic parameters DELTAG, DELTAH and DELTAS have been also evaluated and their importance in complexation has been discussed. (author)

  8. Effect of Al(OH)3 on the sintering of UO2-Gd2O3 fuel pellets with addition of U3O8 from recycle

    Science.gov (United States)

    dos Santos, Lauro Roberto; Durazzo, Michelangelo; Urano de Carvalho, Elita Fontenele; Riella, Humberto Gracher

    2017-09-01

    The incorporation of gadolinium as burnable poison directly into nuclear fuel is important for reactivity compensation, which enables longer fuel cycles. The function of the burnable poison fuel is to control the neutron population in the reactor core during its startup and the beginning of the fuel burning cycle to extend the use of the fuel. The implementation of UO2-Gd2O3 poisoned fuel in Brazil has been proposed according to the future requirements established for the Angra-2 nuclear power plant. The UO2 powder used is produced from the Ammonium Uranyl Carbonate (AUC). The incorporation of Gd2O3 powder directly into the AUC-derived UO2 powder by dry mechanical blending is the most attractive process, because of its simplicity. Nevertheless, processing by this method leads to difficulties while obtaining sintered pellets with the minimum required density. The cause of the low densities is the bad sintering behavior of the UO2-Gd2O3 mixed fuel, which shows a blockage in the sintering process that hinders the densification. This effect has been overcome by microdoping of the fuel with small quantities of aluminum. The process for manufacturing the fuel inevitably generates uranium-rich scraps from various sources. This residue is reincorporated into the production process in the form of U3O8 powder additions. The addition of U3O8 also hinders densification in sintering. This study was carried out to investigate the influence of both aluminum and U3O8 additives on the density of fuel pellets after sintering. As the effects of these additives are counterposed, this work studied the combined effect thereof, seeking to find an applicable composition for the production process. The experimental results demonstrated the effectiveness of aluminum, in the form of Al(OH)3, as an additive to promote increase in the densification of the (U,Gd)O2 pellets during sintering, even with high additions of U3O8 recycled from the manufacturing process.

  9. Materials specific work at Forschungszentrum Karlsruhe and in cooperation with the industrial partners ALKEM and Interatom for the development of nuclear oxide fuels for fission reactors

    International Nuclear Information System (INIS)

    Kleykamp, H.; Muehling, G.

    2005-09-01

    The fabrication of uranium-plutonium oxide fuel started in Forschungszentrum Karlsruhe and at ALKEM company to begin for the criticality experiments in the SNEAK reactor and subsequently for stationary fuel pin irradiations in the FR2, BR2, DFR, Rapsodie, Phenix and KNK II reactors. The production methods comprised first the mechanical blending of UO2 and PuO2 followed by direct pressing and sintering of the pellets, later the advanced methods such as optimized comilling and ammonium uranyl plutonyl coprecititation. The fabrication of pellets was described in the main, further the alternative fuel pin manufacturing processes by vibrational compaction and hot-impact densification were discussed. The first capsule and pin irradiations in the FR2 and BR2 reactors contributed to the assessment of the maximum operation parameters within the fuel pin development such as linear heat rating, cladding temperature and burnup. Subsequently, small-bundle and largebundle irradiations were made in fast reactors in cooperation with Interatom company in order to verify the specifications for the commercial fast reactor SNR 300. Milestones were the maximum burnup of 175 GWd/t metal, corresponding 18.6 % of the heavy atoms, obtained in one of the KNK II fuel pin assemblies, and the displacement rates in the cladding materials of 140 dpa NRT attained in the Phenix reactor. Higher implications gained later the stationary irradiations of defected mixed-oxide pins, the mild fuel pin transient operations, the local blockage experiments and the severe hypothetic accidents in the respective Siloe, HFR, BR2 and CABRI reactors. These experiments were made solely in international partnership. Further activities were the chemical analyses of solid residues and coprecipitations of irradiated mixed-oxide fuels in the head-end of the reprocessing. All these actions were coordinated in the then fast breeder project. Furthermore, irradiated fuels and fuel pins of other reactor types were

  10. Topologically identical, but geometrically isomeric layers in hydrous α-, β-Rb[UO2(AsO3OH)(AsO2(OH)2)]·H2O and anhydrous Rb[UO2(AsO3OH)(AsO2(OH)2)

    Science.gov (United States)

    Yu, Na; Klepov, Vladislav V.; Villa, Eric M.; Bosbach, Dirk; Suleimanov, Evgeny V.; Depmeier, Wulf; Albrecht-Schmitt, Thomas E.; Alekseev, Evgeny V.

    2014-07-01

    The hydrothermal reaction of uranyl nitrate with rubidium nitrate and arsenic (III) oxide results in the formation of polymorphic α- and β-Rb[UO2(AsO3OH)(AsO2(OH)2)]·H2O (α-, β-RbUAs) and the anhydrous phase Rb[UO2(AsO3OH)(AsO2(OH)2)] (RbUAs). These phases were structurally, chemically and spectroscopically characterized. The structures of all three compounds are based upon topologically identical, but geometrically isomeric layers. The layers are linked with each other by means of the Rb cations and hydrogen bonding. Dehydration experiments demonstrate that water deintercalation from hydrous α- and β-RbUAs yields anhydrous RbUAs via topotactic reactions.

  11. Substitution of IO3-, IO4-, SeO32-, and SeO42- for CO32- in Na4[UO2(CO3)3

    International Nuclear Information System (INIS)

    Wu, S.; Notre Dame Univ., IN; Chen, F.; Simonetti, A.; Albrecht-Schmitt, T.E.

    2013-01-01

    Trigonal sodium uranyl carbonate, Na 4 [UO 2 (CO 3 ) 3 ], has been synthesized under hydrothermal conditions, and its incorporation of IO 3 - , IO 4 - , SeO 3 2- , and SeO 4 2- has been investigated. LA-ICP-MS was used to detect the presence and concentration of iodine, selenium, and uranium in single crystals of Na 4 [UO 2 (CO 3 ) 3 ], and these in-situ analyses indicate that IO 3 - , IO 4 - , SeO 3 2- , and SeO 4 2- have been incorporated into its structure. The proposed mechanisms are the substitution of IO 3 - , IO 4 - , SeO 3 2- , and SeO 4 2- for CO 3 2- . The incorporation of iodine oxoanions results in the loss of Na + cations so as to maintain charge balance; the substitution schemes may be expressed as follows: □ + IO 3 - Na + + CO 3 2- and □ + IO 4 - Na + + CO 3 2- (□ = vacancy). (orig.)

  12. Simulation of technetium extraction behavior in UO2 (NO3)2-TcO4--HNO3-H2O/TBP-kerosene system

    International Nuclear Information System (INIS)

    Zhang Chunlong; He Hui; Chen Yanxin; Tang Hongbin

    2012-01-01

    By comparing and analyzing lots of reported data of technetium with the computing results, a modification function P(c 0 (U), t) was introduced to the existing distribution coefficient model of technetium, and a new mathematical model for simulating technetium extraction behavior in the system of UO 2 (NO 3 ) 2 -TcO 4 -HNO 3 -H 2 O/TBP- kerosene was established, as well as a computer program. The reliability of the program was verified by 179 sets of distribution coefficient data, and the results were found to agree well with experimental data. By comparing the reported data of technetium with the computing results, an evaluation was made to test the performance of the revised model. It turned out that the calculation results of the new model were more reliable than that of the one reported previously. The revised model and program can be the foundation to simulating technetium extraction behavior in the system of UO 2 (NO 3 ) 2 - TcO 4 - -HNO 3 -H 2 O/TBP-kerosene with the temperature scope from 10 to 60℃, U concentration from 0 to 280 g/L, and nitric acid concentration from 0.1 to 5 mol/L. (authors)

  13. Spectral neighbor analysis method for automated generation of quantum-accurate interatomic potentials

    International Nuclear Information System (INIS)

    Thompson, A.P.; Swiler, L.P.; Trott, C.R.; Foiles, S.M.; Tucker, G.J.

    2015-01-01

    We present a new interatomic potential for solids and liquids called Spectral Neighbor Analysis Potential (SNAP). The SNAP potential has a very general form and uses machine-learning techniques to reproduce the energies, forces, and stress tensors of a large set of small configurations of atoms, which are obtained using high-accuracy quantum electronic structure (QM) calculations. The local environment of each atom is characterized by a set of bispectrum components of the local neighbor density projected onto a basis of hyperspherical harmonics in four dimensions. The bispectrum components are the same bond-orientational order parameters employed by the GAP potential [1]. The SNAP potential, unlike GAP, assumes a linear relationship between atom energy and bispectrum components. The linear SNAP coefficients are determined using weighted least-squares linear regression against the full QM training set. This allows the SNAP potential to be fit in a robust, automated manner to large QM data sets using many bispectrum components. The calculation of the bispectrum components and the SNAP potential are implemented in the LAMMPS parallel molecular dynamics code. We demonstrate that a previously unnoticed symmetry property can be exploited to reduce the computational cost of the force calculations by more than one order of magnitude. We present results for a SNAP potential for tantalum, showing that it accurately reproduces a range of commonly calculated properties of both the crystalline solid and the liquid phases. In addition, unlike simpler existing potentials, SNAP correctly predicts the energy barrier for screw dislocation migration in BCC tantalum

  14. New interatomic potentials of W, Re and W-Re alloy for radiation defects

    Science.gov (United States)

    Chen, Yangchun; Li, Yu-Hao; Gao, Ning; Zhou, Hong-Bo; Hu, Wangyu; Lu, Guang-Hong; Gao, Fei; Deng, Huiqiu

    2018-04-01

    Tungsten (W) and W-based alloys have been considered as promising candidates for plasma-facing materials (PFMs) in future fusion reactors. The formation of rhenium (Re)-rich clusters and intermetallic phases due to high energy neutron irradiation and transmutations significantly induces the hardening and embrittlement of W. In order to better understand these phenomena, in the present work, new interatomic potentials of W-W, Re-Re and W-Re, suitable for description of radiation defects in such alloys, have been developed. The fitted potentials not only reproduce the results of the formation energy, binding energy and migration energy of various radiation defects and the physical properties from the extended database obtained from DFT calculations, but also predict well the relative stability of different interstitial dislocation loops in W, as reported in experiments. These potentials are applicable for describing the evolution of defects in W and W-Re alloys, thus providing a possibility for the detailed understanding of the precipitation mechanism of Re in W under irradiation.

  15. On macromolecular refinement at subatomic resolution with interatomic scatterers

    Energy Technology Data Exchange (ETDEWEB)

    Afonine, Pavel V., E-mail: pafonine@lbl.gov; Grosse-Kunstleve, Ralf W.; Adams, Paul D. [Lawrence Berkeley National Laboratory, One Cyclotron Road, BLDG 64R0121, Berkeley, CA 94720 (United States); Lunin, Vladimir Y. [Institute of Mathematical Problems of Biology, Russian Academy of Sciences, Pushchino 142290 (Russian Federation); Urzhumtsev, Alexandre [IGMBC, 1 Rue L. Fries, 67404 Illkirch and IBMC, 15 Rue R. Descartes, 67084 Strasbourg (France); Faculty of Sciences, Nancy University, 54506 Vandoeuvre-lès-Nancy (France); Lawrence Berkeley National Laboratory, One Cyclotron Road, BLDG 64R0121, Berkeley, CA 94720 (United States)

    2007-11-01

    Modelling deformation electron density using interatomic scatters is simpler than multipolar methods, produces comparable results at subatomic resolution and can easily be applied to macromolecules. A study of the accurate electron-density distribution in molecular crystals at subatomic resolution (better than ∼1.0 Å) requires more detailed models than those based on independent spherical atoms. A tool that is conventionally used in small-molecule crystallography is the multipolar model. Even at upper resolution limits of 0.8–1.0 Å, the number of experimental data is insufficient for full multipolar model refinement. As an alternative, a simpler model composed of conventional independent spherical atoms augmented by additional scatterers to model bonding effects has been proposed. Refinement of these mixed models for several benchmark data sets gave results that were comparable in quality with the results of multipolar refinement and superior to those for conventional models. Applications to several data sets of both small molecules and macromolecules are shown. These refinements were performed using the general-purpose macromolecular refinement module phenix.refine of the PHENIX package.

  16. Effective Interatomic Potentials Based on The First-Principles Material Database

    OpenAIRE

    Yamamoto, T; Ohnishi, S; Chen, Y; Iwata, S

    2009-01-01

    Effective interatomic potentials are frequently utilized for large-scale simulations of materials. In this work, we generate an effective interatomic potential, with Niobium as an example, using the force-matching method derived from a material database which is created by the first-principle molecular dynamics. It is found that the potentials constructed in the present work are more transferable than other existing potential models. We further discuss how the first-principles material databa...

  17. Simulation of reactivity-initiated accident transients on UO2-M5® fuel rods with ALCYONE V1.4 fuel performance code

    Directory of Open Access Journals (Sweden)

    Isabelle Guénot-Delahaie

    2018-03-01

    Full Text Available The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs, power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs. As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on PWR-UO2 fuel rods with advanced claddings such as M5® under “low pressure–low temperature” or “high pressure–high temperature” water coolant conditions.This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on UO2-M5® fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE—starting from base irradiation conditions it itself computes—is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur.Areas of improvement are finally discussed with a view to simulating and

  18. Boundarylike behaviors of the resonance interatomic energy in a cosmic string spacetime

    Science.gov (United States)

    Zhou, Wenting; Yu, Hongwei

    2018-02-01

    By generalizing the formalism proposed by Dalibard, Dupont-Roc and Cohen Tannoudji, we study the resonance interatomic energy of two identical atoms coupled to quantum massless scalar fields in a symmetric/antisymmetric entangled state in the Minkowski and cosmic string spacetimes. We find that in both spacetimes, the resonance interatomic energy has nothing to do with the field fluctuations but is attributed to the radiation reaction of the atoms only. We then concretely calculate the resonance interatomic energy of two static atoms near a perfectly reflecting boundary in the Minkowski spacetime and near an infinite and straight cosmic string, respectively. We show that the resonance interatomic energy in both cases can be enhanced or suppressed and even nullified as compared with that in an unbounded Minkowski spacetime, because of the presence of the boundary in the Minkowski spacetime or the nontrivial spacetime topological structure of the cosmic string. Besides, we also discover that the resonance interatomic energy in the cosmic string spacetime exhibits some peculiar properties, making it in principle possible to sense different cosmic string spacetimes via the resonance interatomic energy.

  19. Shock-induced plasticity in tantalum single crystals: Interatomic potentials and large-scale molecular-dynamics simulations

    Science.gov (United States)

    Ravelo, R.; Germann, T. C.; Guerrero, O.; An, Q.; Holian, B. L.

    2013-10-01

    We report on large-scale nonequilibrium molecular dynamics simulations of shock wave compression in tantalum single crystals. Two new embedded atom method interatomic potentials of Ta have been developed and optimized by fitting to experimental and density functional theory data. The potentials reproduce the isothermal equation of state of Ta up to 300 GPa. We examined the nature of the plastic deformation and elastic limits as functions of crystal orientation. Shock waves along (100), (110), and (111) exhibit elastic-plastic two-wave structures. Plastic deformation in shock compression along (110) is due primarily to the formation of twins that nucleate at the shock front. The strain-rate dependence of the flow stress is found to be orientation dependent, with (110) shocks exhibiting the weaker dependence. Premelting at a temperature much below that of thermodynamic melting at the shock front is observed in all three directions for shock pressures above about 180 GPa.

  20. Evaluation of natural radionuclide occurrence (Rn-222) in a nuclear facility that processes UF6 for the UO2 powder and pellet productions

    International Nuclear Information System (INIS)

    Mouco, Charles Dickens do Carmo Lacerda; Matta, Luiz Ernesto Santos Carvalho

    2001-01-01

    This paper presents the early data from the concentration evaluation of Rn-222 in a nuclear facility that processes UF 6 for the UO 2 powder and tablet productions. Measurements were accomplished in the same points and under the same climatic conditions with the factory working and not working. The results showed that no significant alteration was detected in the obtained values in both situations. The average value was 15 and 18 Bq/m3 for operating and not operating respectively. There is no significant difference between data. The concentrations values of Rn-222 are lower than the internationally established limits for this radionuclide. The Rn-222 concentration values obtained, are probably originated from the building material, and the Rn-222 concentration levels are due to the great air exchange inside the factory. (author)

  1. Study of the experimental parameters for the determination of Ca, Cr, Cu, Fe, Mn and Ni on nuclear grade UO2 by X-ray fluorescence technique

    International Nuclear Information System (INIS)

    Salvador, V.L.R.

    1982-01-01

    An analytical method for the simultaneous determinations of low concentrations of Ca, Cr, Cu, Fe, Mn and Ni on the nuclear grade UO 2 by X-ray fluorescence technique, without the use of chemical treatment, is described. The optimization of the experimental conditions was established on the X-ray fluorescence spectrometer and a low limit of detection (4 - 7 μg/gU) was achieved which satisfies the requirement in the nuclear fuel specification. The samples were prepared in the form of double layer pressed pellets using boric acid as a binding agent. The characteristic first order K sub(α) line intensity of each minor component was measured and the values of its concentrations were deduced using respective standard calibrations curves. The precision, accuracy and acceptability of the method were determined for all elements. The values of the precision are in the range of 2 - 10% and the accuracy are lower than 7%. (Author) [pt

  2. Computational simulation of the microstructure of irradiation damaged regions for the plate type fuel of UO2 microspheres dispersed in stainless steel matrix

    International Nuclear Information System (INIS)

    Reis, S.C. dos; Lage, A.F.; Braga, D.; Ferraz, W.B.

    2006-01-01

    Plate type fuel elements have high efficiency of thermal transference what benefits the heat flux with high rates of power output. In reactor cores, fuel elements, in general, are subject to a high neutrons flux, high working temperatures, severe corrosion conditions, direct interference of fission products that result from nuclear reactions and radiation interaction-matter. For plate type fuels composed of ceramic particles dispersed in metallic matrix, one can observe the damage regions that arise due to the interaction fission products in the metallic matrix. Aiming at evaluating the extension of the damage regions in function of the particles and its diameters, in this paper, computational geometric simulations structure of plate type fuel cores, composed of UO 2 microspheres dispersed in stainless steel in several fractions of volume and diameters were carried out. The results of the simulations were exported to AutoCAD R where it was possible its visualization and analysis. (author)

  3. Ionophoretic method in the study of mixed ligand ternary chelates of UO2(II), Ni(II) and Zn(II) involving nitrilotriacetate and cytosine as ligands

    International Nuclear Information System (INIS)

    Mishra, A.P.; Mishra, S.K.; Yadava, K.L.

    1987-01-01

    A novel electrophoretic technique is described for the assessment of the equilibria in mixed-ligand complex system in solution. It is based on the movement of spot of the metal ion under an electric field with the complexants added in the background electrolyte at fixed pH. The concentration of primary ligand nitrilotriacetate was constant while that of secondary ligand (cytosine) was varied. The plot of log (cytosine) against mobility was used to obtain information on the formation of the mixed complexes and to calculate its stability constants. Experimentally obtained logK values are as 5.62, 4.55 and 4.42 for mixed complexes of UO 2 (II), Ni(II) and Zn(II) respectively at μ=0.1 and temp.=35 +- 01.degC. (author). 10 refs

  4. Methodology for analysing powder and UO2 pellets samples of nuclear purity, using gravimetry and potentiometric tritation of Davies and Gray/NBL

    International Nuclear Information System (INIS)

    Araujo, R.M.S. de; Almeida, S.G. de; Bezerra, J.H.B.; Silva, S.P. da

    1986-01-01

    It is showed the methodology in use in the CNEN's Safeguards Laboratory to the implantation of the modified Davies and Gray and gravimetric methods applied to UO 2 powder and pellets. The stability's control of the sample is indispensable because its expontaineous oxidation. For each type and quality of sample an appropriated analytical scheme is applied. The precision and accuracy of the measurements are established, and periodically observed through /the standardization and internal quality control of the titration system, and the participation in intercomparison programs with the International Atomic Energy Agency. Several samples have been analysed, and the statistical evaluation of the results has showed /precision of the orders of 0.08% to 0.2% and accuracy of 0.062% U, comparables to international laboratories. (Author) [pt

  5. The distribution of Th(NO3)4, UO2(NO3)2 and HNO3 between an aqueous phase and an organic tributyl phosphate phase

    International Nuclear Information System (INIS)

    Nakashima, T.; Zimmer, E.

    1984-05-01

    The distribution of Th(NO 3 ) 4 , UO 2 (NO 3 ) 2 and HNO 3 between an aqueous phase and an organic phase, consisting of 30 Vol.% tributyl phosphate in dodecane, has been experimentally investigated. About 120 distribution data have been determined in the concentration ranges that can be seen in the THOREX process for reprocessing spent thorium bearing fuel. Based on the experimental data, two computer programs have been developed which make possible interpolations and, to some extent, extrapolations. With model 1, concentrations in the organic phase can be calculated if that in the aqueous phase are known. With model 2, concentrations in the aqueous phase can be calculated vice versa. Besides the description of the calculation models, a large body of calculated data can be found in this report. In a addition, a calculation mode is presented that makes possible the calculation of distribution data for very low thorium concentrations. (orig.) [de

  6. Elaborated studies for the ligitional behavior of thiouracil derivative towards Ni(II), Pd(II), Pt(IV), Cu(II) and UO2 ² ions.

    Science.gov (United States)

    Abou-Melha, Khlood Saad

    2012-11-01

    A synthesis of new thiouracil derivative was carried out and deliberately investigated. A new series of complexes was prepared using Ni(II), Pd(II), Pt(IV), Cu(II) and UO(2)(+2) ions. IR spectral data proposed the coordination mod of the ligand towards each metal ion and displays the binegative pentadentate mod as the maximum mod of coordination obtained with Ni(II) and Cu(II) complexes. (1)HNMR spectrum of UO(2)(+2) complex in comparing with the free ligand spectrum supports the binegative appearance of the coordinated ligand through the ionization of CO and CS groups. The electronic spectral data as well as the magnetic moment measurements are coincide with each others to propose the square-planar geometry with Ni(II), Pd(II) and Cu(II) complexes and octahedral geometry with the others. ESR spectrum of Cu(II) complex displays axially symmetric g tensor parameters with g(11)>g(⊥)>2.0023 indicating that the [Formula: see text] orbital as a ground state with the square-planar geometry. The TG analysis for all isolated complexes were carried out to assert about the presence of water molecules physically or chemically attached with the central atom. The biological study was carried out against different microorganisms as gram negative, gram positive and fungi. The comparable data display the relative priority of Ni(II) complex in comparing with others against all organisms but, the other complexes display activity by the same with the free ligand. Copyright © 2012 Elsevier B.V. All rights reserved.

  7. Reaction kinetics aspect of U3O8 kernel with gas H2 on the characteristics of activation energy, reaction rate constant and O/U ratio of UO2 kernel

    International Nuclear Information System (INIS)

    Damunir

    2007-01-01

    The reaction kinetics aspect of U 3 O 8 kernel with gas H 2 on the characteristics of activation energy, reaction rate constant and O/U ratio of UO 2 kernel had been studied. U 3 O 8 kernel was reacted with gas H 2 in a reduction furnace at varied reaction time and temperature. The reaction temperature was varied at 600, 700, 750 and 850 °C with a pressure of 50 mmHg for 3 hours in gas N 2 atmosphere. The reation time was varied at 1, 2, 3 and 4 hours at a temperature of 750 °C using similar conditions. The reaction product was UO 2 kernel. The reaction kinetic aspect between U 3 O 8 and gas H 2 comprised the minimum activation energy (ΔE), the reaction rate constant and the O/U ratio of UO 2 kernel. The minimum activation energy was determined from a straight line slope of equation ln [{D b . R o {(1 - (1 - X b ) ⅓ } / (b.t.Cg)] = -3.9406 x 10 3 / T + 4.044. By multiplying with the straight line slope -3.9406 x 10 3 , the ideal gas constant (R) 1.985 cal/mol and the molarity difference of reaction coefficient 2, a minimum activation energy of 15.644 kcal/mol was obtained. The reaction rate constant was determined from first-order chemical reaction control and Arrhenius equation. The O/U ratio of UO 2 kernel was obtained using gravimetric method. The analysis result of reaction rate constant with chemical reaction control equation yielded reaction rate constants of 0.745 - 1.671 s -1 and the Arrhenius equation at temperatures of 650 - 850 °C yielded reaction rate constants of 0.637 - 2.914 s -1 . The O/U ratios of UO 2 kernel at the respective reaction rate constants were 2.013 - 2.014 and the O/U ratios at reaction time 1 - 4 hours were 2.04 - 2.011. The experiment results indicated that the minimum activation energy influenced the rate constant of first-order reaction and the O/U ratio of UO 2 kernel. The optimum condition was obtained at reaction rate constant of 1.43 s -1 , O/U ratio of UO 2 kernel of 2.01 at temperature of 750 °C and reaction time of 3

  8. Mechanical Behavior of UO2 at Sub-grain Length Scales: Quantification of Elastic, Plastic and Creep Properties via Microscale Testing

    Energy Technology Data Exchange (ETDEWEB)

    Peralta, Pedro

    2018-04-16

    Techniques were developed to measure properties at sub-grain scales using depleted Uranium Oxide (d-UO2) samples heat-treated to obtain different grain sizes and oxygen stoichiometries, through three main tasks: 1) sample processing and characterization, 2) microscale and conventional testing and 3) modeling. Grain size and crystallography were characterized using Scanning Electron Microscopy (SEM), in conjunction with Electron Backscattering Diffraction (EBSD) and Electron Channeling Contrast Imaging (ECCI). Grains were then carefully selected based on their crystallographic orientations to perform ex-situ micromechanical tests with samples machined via Focused Ion Beam (FIB), with emphasis on micro-cantilever bending. These experiments were performed under controlled atmospheres, to insure stoichiometry control, at temperatures up to 700 °C and allowed measurements involving elastic (effective Young’s modulus), plastic (critical resolved shear stresses) and creep (creep strain rates) behavior. Conventional compression experiments were performed simultaneously to compare with the ex-situ measurements and study potential size effects. Modeling was implemented using anisotropic elasticity and inelastic constitutive relations for plasticity and creep based on kinematics and kinetics of dislocation glide that account for the effects of crystal orientation, and stress. The models will be calibrated and validated using the experimental data. This project provided insight on correlations among stoichiometry, crystallography and mechanical behavior in advanced oxide fuels, provided valuable experimental data to validate and calibrate mesoscale fuel performance codes and also a framework to measure sub-grain scale mechanical properties that should be suitable for use with irradiated samples due to small volumes required. The goals and metrics of the ongoing study of thermo-mechanical behavior in depleted uranium dioxide (d-UO2) outlined in this project have been

  9. Nitric acid titration in the presence of UO2(NO3)2, Th(NO3)4, U(NO3)4 or Zr(NO3)4

    International Nuclear Information System (INIS)

    Nakashima, T.; Lieser, K.H.

    1986-01-01

    Procedures are described for titration of HNO 3 in presence of UO 2 2+ , Th 4+ , U 4+ and Zr 4+ without formation of interfering precipitates. In the first step the hydrolysable ions are masked by addition of complexing agents and in the second step the acid is titrated by NaOH as usual. (orig.)

  10. Variable dimensionality and new uranium oxide topologies in the alkaline-earth metal uranyl selenites AE[UO2)(SeO3)2] (AE=Ca, Ba) and Sr[UO2)(SeO3)2] · 2H2O

    International Nuclear Information System (INIS)

    Almond, Philip M.; Peper, Shane M.; Bakker, Eric; Albrecht-Schmitt, Thomas E.

    2002-01-01

    Three new alkaline-earth metal uranyl selenites, Ca[UO 2 )(SeO 3 ) 2 ] (1), Sr[UO 2 )(SeO 3 ) 2 ] · 2H 2 O (2), and Ba[UO 2 )(SeO 3 ) 2 ] (3), have been prepared from the reactions of CaCO 3 and Ca(OH) 2 , SrCl 2 and Sr(OH) 2 , or BaCl 2 and Ba(OH) 2 with UO 3 and SeO 2 under mild hydrothermal conditions. Single-crystal X-ray diffraction experiments reveal that the structures of 1-3 differ in both connectivity and dimensionality even though all contain the same fundamental building unit, namely [UO 2 (SeO 3 ) 4 ]. This polyhedron consists of a linear uranyl unit that is bound by one chelating and three bridging selenite anions creating a pentagonal bipyramidal environment around the U(VI) center. The crystal structure of 1 contains one-dimensional ribbons where the edges are terminated by monodentate selenite anions. The interior of the ribbons are constructed from edge-sharing pentagonal bipyramidal UO 7 units. The structure of 2 is also one-dimensional; however, here there are chains of edge-sharing pentagonal bipyramidal UO 7 dimers that are connected by bridging selenite anions. Ba[(UO 2 )(SeO 3 ) 2 ] (3) is two-dimensional, and the highly ruffled anionic sheets present in this structure are formed from both bridging and chelating/bridging selenite anions bound to uranyl moieties. The anionic substructures in 1-3 are separated by Ca 2+ , Sr 2+ , or Ba 2+ cations. Crystallographic data (193 K, MoKα, λ=0.71073): 1, triclinic, space group P1-bar, a=5.5502(6) A, b=6.6415(7) A, c=11.013(1) A, α=104.055(2) deg., β=93.342(2) deg., γ=110.589(2) deg. , Z=2, R(F)=4.56% for 100 parameters with 1530 reflections with I>2σ(I); 2, triclinic, space group P1-bar, a=7.0545(5) A, b=7.4656(5) A, c=10.0484(6) A, α=106.995(1) deg., β=108.028(1) deg., γ=98.875(1) deg., Z=2, R(F)= 2.43% for 128 parameters with 2187 reflections with I>2σ(I); 3, monoclinic, space group P2 1 /c, a=7.3067(6) A, b=8.1239(7) A, c=13.651(1) A, β=100.375(2) deg., Z=4, R(F)=4.31% for 105 parameters

  11. Attractive short-range interatomic potential in the lattice dynamics of niobium and tantalum

    International Nuclear Information System (INIS)

    Onwuagba, B.N.; Pal, S.

    1987-01-01

    It is shown in the framework of the pseudopotential approach that there is a sizable attractive short-range component of the interatomic potential due to the s-d interaction which has the same functional form in real space as the Born-Mayer repulsion due to the overlap of core electron wave functions centred on neighbouring ions. The magnitude of this attractive component is such as to completely cancel the conventional Born-Mayer repulsion, making the resultant short-range interatomic potential attractive rather than repulsive. Numerical calculations show that the attractive interatomics potential, which represents the local-field correction, leads to a better understanding of the occurrence of the soft modes in the phonon dispersion curves of niobium and tantalum

  12. Room and high temperature interactions in sodium and rubidium rich ternary nitrate mixtures of UO2(NO3)2.6H2O - NaNO3 - RbNO3

    International Nuclear Information System (INIS)

    Kalekar, Bhupesh B.; Reddy, A.V.R.; Raje, Naina

    2016-01-01

    High temperature interaction behavior of nitrates is important for characterizing different intermediate products and their thermal stabilities during the calcination of nuclear waste before their immobilization in the stable glass matrix. Mixtures of UO 2 (NO 3 ) 2 .6H 2 O (UNH) with NaNO 3 (NaN) and RbNO 3 (RbN) were prepared by mixing the weighed amounts of component nitrates and grinding gently in a mortar and pestle. The mixing and grinding of individual nitrate components in a mortar with pestle showed the agglomeration of solid particles and subsequent dissolution probably in the water of crystallization of UNH. The continued grinding and mixing showed the reappearance of the solid powder. The original yellow color of the mixture was changed to greenish yellow color. The mixtures were subjected to thermal measurements using Netzsch Thermobalance (Model No.: STA 409 PC Luxx) coupled to Bruker FTIR system (Model No.: Tensor 27) via a heated Teflon capillary (1 m long, 2 mm i.d.). TG - DTG curves of equimolar mixture are displayed. The plateau was observed on TG curve in the temperature region of 31- 250 °C. It is reported that Na(UO 2 (NO 3 ) 3 ).H 2 O and Rb(UO 2 (NO 3 ) 3 ) formed around 250 °C in the equimolar nitrate mixtures of UNH-NaN and UNH-RbN. Thermal and XRD results indicated the formation of Na(UO 2 (NO 3 ) 3 ).H 2 O and Rb(UO 2 (NO) 3 ) 3 ) even by mixing the UNH, NaN and RbN in equimolar ratios at room temperature

  13. A novel proof of the DFT formula for the interatomic force field of Molecular Dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Morante, S., E-mail: morante@roma2.infn.it [Dipartimento di Fisica, Università di Roma, “ Tor Vergata ”, INFN, Sezione di Roma 2, Via della Ricerca Scientifica - 00133 Roma (Italy); Rossi, G.C., E-mail: rossig@roma2.infn.it [Dipartimento di Fisica, Università di Roma, “ Tor Vergata ”, INFN, Sezione di Roma 2, Via della Ricerca Scientifica - 00133 Roma (Italy); Centro Fermi-Museo Storico della Fisica e Centro Studi e Ricerche E. Fermi, Compendio del Viminale, Piazza del Viminale 1, I-00184 Rome (Italy)

    2017-02-15

    We give a novel and simple proof of the DFT expression for the interatomic force field that drives the motion of atoms in classical Molecular Dynamics, based on the observation that the ground state electronic energy, seen as a functional of the external potential, is the Legendre transform of the Hohenberg–Kohn functional, which in turn is a functional of the electronic density. We show in this way that the so-called Hellmann–Feynman analytical formula, currently used in numerical simulations, actually provides the exact expression of the interatomic force.

  14. A novel proof of the DFT formula for the interatomic force field of Molecular Dynamics

    Science.gov (United States)

    Morante, S.; Rossi, G. C.

    2017-02-01

    We give a novel and simple proof of the DFT expression for the interatomic force field that drives the motion of atoms in classical Molecular Dynamics, based on the observation that the ground state electronic energy, seen as a functional of the external potential, is the Legendre transform of the Hohenberg-Kohn functional, which in turn is a functional of the electronic density. We show in this way that the so-called Hellmann-Feynman analytical formula, currently used in numerical simulations, actually provides the exact expression of the interatomic force.

  15. Hydrolytic and radiolytic degradation of TBP in TBP.30% V/V-dodecane/UO2(NO3)2.HNO3.H2O systems

    International Nuclear Information System (INIS)

    Barreta, L.G.

    1980-01-01

    The hydrolytic and radiolytic degradation of TBP is investigated in systems of TBP 30% V/V-dodecane/H 2 O . HNO 3 . UO 2 (NO 3 ) 2 by gas chromatographic determination of HDBP. No direct relation between the concentration of HDBP formed and the quantity of HNO 3 extracted by the organic phase is observed in the studies of hydrolysis of TBP. The HDBP concentration is seen to increase non-linearly with the concentration of HNO 3 extracted by the organic phase. Radiolytic studies show that for doses greater than 1 Wh/l, the concentration of HDBP formed increases with the dose absorbed by the system. Whith doses smaller than 1 Wh/l and acid concentration greater than 2 M, two distinct patterns of behavior are observed. The concentration of HDBP as a function of the radiation dose absorbed by the system presents a minimum for uranyl nitrate concentrations smaller than 0.9 M; for uranyl nitrate concentrations greater than 1.3 M the concentration of radiolytic HDBP cannot be calculated because the concentration of the hydrolytic HDBP determined is greater than the sum of the experimental concentrations of hydrolytic and radiolytic HDBP. It is known that the dose absorbed by the process solutions during the reprocessing of light water reactor fuel elements is smaller than one Wh/l. Thus, dose rates between zero and one Wh/l should be studied for this system. (Author) [pt

  16. Molten salt flux synthesis and structure of the new layered uranyl tellurite, K 4[(UO 2) 5(TeO 3) 2O 5

    Science.gov (United States)

    Woodward, Jonathan D.; Albrecht-Schmitt, Thomas E.

    2005-09-01

    The reaction of UO 3 and TeO 3 with a KCl flux at 800 °C for 3 days yields single crystals of K 4[(UO 2) 5(TeO 3) 2O 5]. The structure of the title compound consists of layered, two-dimensional ∞2[(UO)5(TeO)2O] sheets arranged in a stair-like topology separated by potassium cations. Contained within these sheets are one-dimensional uranium oxide ribbons consisting of UO 7 pentagonal bipyramids and UO 6 tetragonal bipyramids. The ribbons are in turn linked by corner-sharing with trigonal pyramidal TeO 3 units to form sheets. The lone-pair of electrons from the TeO 3 groups are oriented in opposite directions with respect to one another on each side of the sheets rendering each individual sheet nonpolar. The potassium cations form contacts with nearby tellurite units and axial uranyl oxygen atoms. Crystallographic data (193 K, Mo Kα, λ=0.71073 Å): triclinic, space group P 1¯, a=6.8514(5) Å, b=7.1064(5) Å, c=11.3135(8) Å, α=99.642(1)°, β=93.591(1)°, γ=100.506(1)°, V=531.48(7) Å3, Z=1,R(F)=4.19% for 149 parameters and 2583 reflections with I>2σ(I).

  17. Fabrication and post-irradiation examination of a zircaloy-2 clad UO2-1.5 wt% PuO2 fuel pin irradiated in PWL, CIRUS

    International Nuclear Information System (INIS)

    Sah, D.N.; Sahoo, K.C.; Chatterjee, S.; Majumdar, S.; Kamath, H.S.; Ramachandran, R.; Bahl, J.K.; Purushottam, D.S.C.; Ramakumar, M.S.; Sivaramakrishnan, K.S.; Roy, P.R.

    1977-01-01

    A zircaloy-2 clad UO 2 -1.5 wt% PuO 2 fuel pin was fabricated at the Radiometallurgy Section of the Bhabha Atomic Research Centre, Bombay, for irradiation in the pressurised water loop in CIRUS. Requisite development work related to powder conditioning, blending, pressing and sintering parameters was carried out to meet the exacting fuel pellet specifications of CANDU fuel. The fuel pin ruptured while being irradiated in the pressurised water loop in CIRUS, after experiencing a low burn-up of 507 MWD/MTM and was subsequently examined at the Radiometallurgy Hot Cells Facility. The results showed that internal clad hydriding led to primary failure of the fuel pin. Subsequent ingress of the coolant water caused excessive swelling of the thermal insulating magnesia pellets located at the ends of the fuel column. The swelling of magnesia pellets caused severe rupturing of the fuel pin at the two ends. The delayed rupturing of the fuel pin at the upper end, caused the fuel column to be displaced downwards by 5.85mm. (author)

  18. Co(II), Ce(III) and UO 2(VI) bis-salicylatothiosemicarbazide complexes . Binary and ternary complexes, thermal studies and antimicrobial activity

    Science.gov (United States)

    El-Wahab, Z. H. Abd; Mashaly, Mahmoud M.; Salman, A. A.; El-Shetary, B. A.; Faheim, A. A.

    2004-10-01

    A series of new metal complexes of Co(II), Ce(III) and UO 2(VI), with the Schiff base ligand, H 2L, bis-salicylatothiosemicarbazide have been prepared in presence of different molar ratios of LiOH·H 2O as a deprotonating agent. Also, the ternary complexes were prepared by using 2-aminopyridine (2-Ampy) or oxalic acid (Ox) as a secondary ligand. All synthesized compounds were identified and confirmed by elemental analyses, molar conductivities, spectral (UV-Vis, IR, 1H NMR, mass) and magnetic moment measurements as well as TG-DSC technique. The changes in the selected vibrational absorption bands in IR and NMR spectra of the Schiff base ligand upon coordination indicate that, the ligand behaves as a neutral, monoanionic and/or dianionic tetradentate manner with ONNO donor sites. Conductance measurements suggest the non-electrolytic and 1:1 electrolytic nature of the metal complexes. Thermal studies suggest a mechanism for degradation of the metal complexes as function of temperature supporting the chelation modes, moreover, show the possibility of obtaining new complexes pyrolytically in the solid state which cannot be synthesized from solution. Antimicrobial screening of the free ligand and its binary complexes showed that, the free ligand and some metal complexes possess antimicrobial activities towards four type of bacteria and five types of fungi and these results were compared with eleven type of known antibiotics.

  19. Comparison study Of H2O determination in UO2 powder by using MEA (Moisture Evaluation Analysis) and KFT (Karl Fischer Titration)

    International Nuclear Information System (INIS)

    Farida; Yudhi, N.; Lilis, W.; Putro, P.K.

    2000-01-01

    To find out an analytical method to determine H 2 O content in UO 2 powder as fuel elements of power reactors which is simple, economical, precise, and accurate, it is necessary to do comparison study of H 2 O content determination using MEA method which is based on electrolysis process with two helically wound electrodes which contains P 2 O 5 that has function to absorb water steam. The platinum electrodes have a 67 Volt potential on them. The quantity of charge required to electrolyse 0.1 μgr of H 2 O is a constant which is the basic of the electronic measurement. In KFT method is based on volumetric titration using the one component reagent hydronol composite contain all reactants i.e. iodine, sulfur dioxide and imidazole as the base, dissolved in a suitable alcohol. The t- (student) test show that there is no different result significantly between those method. The H 2 O contain obtained is 0.956±0.0095 %, for MEA method and 0.953±0.023 % for KFT method. (author)

  20. Reactivity prediction of uniform PuO2-UO2 fuelled lattices and Pu(NO3)4 solutions in light water

    International Nuclear Information System (INIS)

    Mohankrishnan, P.; Huria, H.C.

    A theoretical analysis of the reactivities of the experimentally measured uniform light water moderated and reflected PuO 2 in UO 2 lattices and Pu(NO 3 ) 4 solutions is presented here. The mixed oxide single rod lattices are homogenised by the use of multigroup integral transport theory and diffusion theory is used for the cylindrical core calculations. The cross-sections are derived from the WTIS library. The homogeneous spherical Pu(NO 3 ) 4 solutions are analysed by discrete ordinate transport theory. Due to the small size of these criticals, it is necessary that one dimensional core calculations also be performed with a cross-section energy group structure which can represent neutron slowing down and thermalisation at the core reflector interface accurately. Due to the absence of such core calculation in the BNWL analyses of the mixed oxide lattices, the agreement of or predictions for these lattices with measurement is considered to be more satisfactory. These reactivity predictions are found to agree generally within +- 0.6% of measurements for the mixed oxide lattices and within 1% for the solution system. (author)

  1. An Analysis of the Thermal and Structure Behaviour of the UO2-PuO2-Fuel in the Irradiation Experiment of the UO2-PuO2-Fuel in the Irradiation Experiment FR2 Capsule Test Series 5a

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.; Helmut, E.

    1981-01-01

    In the Karlsruhe research reactor FR2 nine fuel pins were irradiated within three irradiation capsules in the course of the test series 5a. The pins contained UO 2 -PuO 2 fuel pellets. They reached bump values of about 6, 17 and 47 Mwd/Kg Me with linear rod powers of 400 to 600 W/cm and clad surface temperature between 500 and 700 degree centigree. A detailed analysis of the fuel structuration data (columnar-grain and equiaxed- -grain growth regions) have allowed to determine, with the help of physic-mathematical models, the radii of these regions and the heat transfer through the contact zone between fuel and clad depending on the bump. The results of the analysis showed that the fuel surface temperature rose with increasing burnup. (Author) 16 refs

  2. Multipole expansion of the retarded interatomic potential energy: induction energy for degenerate ground-state atoms

    NARCIS (Netherlands)

    Michels, M.A.J.; Suttorp, L.G.

    1972-01-01

    The inductive contribution to the retarded interatomic potential energy of two atoms in degenerate ground states is calculated up to all multipole orders on the basis of quantum electrodynamics. The result, which is found to have nonretarded character, is written in such a way as to show the

  3. Lattice Stability and Interatomic Potential of Non-equilibrium Warm Dense Gold

    Science.gov (United States)

    Chen, Z.; Mo, M.; Soulard, L.; Recoules, V.; Hering, P.; Tsui, Y. Y.; Ng, A.; Glenzer, S. H.

    2017-10-01

    Interatomic potential is central to the calculation and understanding of the properties of matter. A manifestation of interatomic potential is lattice stability in the solid-liquid transition. Recently, we have used frequency domain interferometry (FDI) to study the disassembly of ultrafast laser heated warm dense gold nanofoils. The FDI measurement is implemented by a spatial chirped single-shot technique. The disassembly of the sample is characterized by the change in phase shift of the reflected probe resulted from hydrodynamic expansion. The experimental data is compared with the results of two-temperature molecular dynamic simulations based on a highly optimized embedded-atom-method (EAM) interatomic potential. Good agreement is found for absorbed energy densities of 0.9 to 4.3MJ/kg. This provides the first demonstration of the applicability of an EAM interatomic potential in the non-equilibrium warm dense matter regime. The MD simulations also reveal the critical role of pressure waves in solid-liquid transition in ultrafast laser heated nanofoils. This work is supported by DOE Office of Science, Fusion Energy Science under FWP 100182, and SLAC LDRD program.

  4. Cambridge Crystallographic Data Centre. IV. Preparation of "Interatomic Distances 1960-65"

    Science.gov (United States)

    Allen, F. H.; And Others

    1973-01-01

    The Cambridge Crystallographic Data Centre is concerned with the retrieval, evaluation, synthesis, and dissemination of structural data obtained by diffraction methods. This paper describes the use of a computer-based file system of both bibliographic information and numeric data to produce a compendium of interatomic distances. (10 references)…

  5. Development of a thermo-kinetic diffusion model for UO2 and (U,Pu)O2 oxide fuels using the DICTRA code

    International Nuclear Information System (INIS)

    Moore, Emily Elaine

    2013-01-01

    Uranium dioxide is the most widely used nuclear fuel for light water reactors, while some countries including France make use of the uranium-plutonium (U,Pu)O 2±x mixed oxide (MOX). The MOX is also considered for future use in the Gen IV reactors, of which the sodium cooled fast reactor (SFR) is of current research interest. Both oxides exhibit a large range of non-stoichiometry due to various oxidative states of uranium and plutonium metal. Thermo-physical properties of the fuel strongly depend on deviations in composition and temperature. Extreme temperature gradients (800 K) between the center (2300 K)and periphery of the MOX fuel pellet expose a central void due to the migration and subsequent redistribution of the fuel-elements. To gain insight into the restructuring, which occurs during the fuel lifetime as well as possible accident scenarios the thermodynamic and kinetic behavior, is crucial. A comprehensive evaluation of these properties can be incorporated in computational models to describe fuel behavior over large temperature and compositions ranges, providing a predictive tool that is applicable to other parts of the fuel cycle, such as optimizing the sintering conditions for manufacturing. Atomic transport especially in UO 2 is widely treated in the experimental and computational materials communities. The current understanding of diffusion properties is limited by the stoichiometric deviations inherent to the fuel. The difficulty is apparent in experimental settings as controlling the oxygen content is problematic. Defects (interstitial and vacancy) associated with the stoichiometric deviations of the oxides facilitate the diffusion process and is of interest in regards to the restructuring of the fuel. Experimental data is widely available; however, coherence between the evaluated diffusion coefficients is not always evident. Existing computational models based on the migration of defects are often based on atomistic level simulations. A complete

  6. An interatomic potential model for carbonates allowing for polarization effects

    International Nuclear Information System (INIS)

    Birse, S.E.A.; Archer, T.D.; Dove, Martin T.; Cygan, Randall Timothy; Gale, Julian D.; Redern, Simon A.T.

    2003-01-01

    An empirical model for investigating the behavior of CaCO 3 polymorphs incorporating a shell model for oxygen has been created. The model was constructed by fitting to: the structure of aragonite and calcite; their elastic, static and high-frequency dielectric constants; phonon frequencies at the wave vectors (1/2 0 2) and (0 0 0) of calcite; and vibrational frequencies of the carbonate deformation modes of calcite. The high-pressure phase transition between calcite I and II is observed. The potentials for the CO 3 group were transferred to other carbonates, by refitting the interaction between CO 3 and the cation to both the experimental structures and their bulk modulus, creating a set of potentials for calculating the properties of a wide range of carbonate materials. Defect energies of substitutional cation defects were analyzed for calcite and aragonite phases. The results were rationalized by studying the structure of calcite and aragonite in greater detail.

  7. Effect of densification additive (Al (OH)3) and U3O8 recycle in sintering UO2-7wt% Gd2O3 fuel pellets

    International Nuclear Information System (INIS)

    Santos, L.R.; Riella, H.G.

    2009-01-01

    The nuclear fuels are the consumable parts of nuclear reactors, and this has several consequences. From an economic point of view, it is important to keep the fuel into reactor for long time. In this context the use of burnable poison, as advanced fuel based in gadolinium oxide dispersed in a uranium oxide matrix, is a technological solution adopted worldwide. The function of the burnable poison fuels is to control the neutrons population in the nuclear reactors cores during its start up and the beginning of the fuel burning cycle to extending their use. In consequence of the use of this advanced fuel, the nuclear reactors can operate with higher rate of power, optimizing the use of the nuclear fuels. The objective of the present work is to show the development of UO 2 -7wt% Gd 2 O 3 burnable absorber containing pellets by using mechanical blending of (Al(OH) 3 ) densification additive and U 3 O 8 of the recycling of nuclear fuel scrap. In the procedures, the gadolinium content of 7 wt% was established as a consequence of the P and D Cooperation Programmer firmed by the CTMSP and the INB, looking for the nationalization of this type of nuclear fuel used in the Nuclear Facility of Angra 2. The experimental results permit to observe the effectiveness action of the compound Al(OH) 3 as a additive to promote the increasing in the densification of the (U-Gd)O 2 pellets during its sintering, when amounts of recycle are recycled to the production processing up to 10 wt%, and when 0,20 wt% of Al(OH) 3 is used as additive. (author)

  8. Mesoscale Benchmark Demonstration Problem 1: Mesoscale Simulations of Intra-granular Fission Gas Bubbles in UO2 under Post-irradiation Thermal Annealing

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yulan; Hu, Shenyang Y.; Montgomery, Robert; Gao, Fei; Sun, Xin; Tonks, Michael; Biner, Bullent; Millet, Paul; Tikare, Veena; Radhakrishnan, Balasubramaniam; Andersson , David

    2012-04-11

    A study was conducted to evaluate the capabilities of different numerical methods used to represent microstructure behavior at the mesoscale for irradiated material using an idealized benchmark problem. The purpose of the mesoscale benchmark problem was to provide a common basis to assess several mesoscale methods with the objective of identifying the strengths and areas of improvement in the predictive modeling of microstructure evolution. In this work, mesoscale models (phase-field, Potts, and kinetic Monte Carlo) developed by PNNL, INL, SNL, and ORNL were used to calculate the evolution kinetics of intra-granular fission gas bubbles in UO2 fuel under post-irradiation thermal annealing conditions. The benchmark problem was constructed to include important microstructural evolution mechanisms on the kinetics of intra-granular fission gas bubble behavior such as the atomic diffusion of Xe atoms, U vacancies, and O vacancies, the effect of vacancy capture and emission from defects, and the elastic interaction of non-equilibrium gas bubbles. An idealized set of assumptions was imposed on the benchmark problem to simplify the mechanisms considered. The capability and numerical efficiency of different models are compared against selected experimental and simulation results. These comparisons find that the phase-field methods, by the nature of the free energy formulation, are able to represent a larger subset of the mechanisms influencing the intra-granular bubble growth and coarsening mechanisms in the idealized benchmark problem as compared to the Potts and kinetic Monte Carlo methods. It is recognized that the mesoscale benchmark problem as formulated does not specifically highlight the strengths of the discrete particle modeling used in the Potts and kinetic Monte Carlo methods. Future efforts are recommended to construct increasingly more complex mesoscale benchmark problems to further verify and validate the predictive capabilities of the mesoscale modeling

  9. Effets de la température et de l'irradiation sur la mobilité du xénon dans UO$_2$ : étude profilométrique et microstructurale

    OpenAIRE

    Marchand, Benoît

    2012-01-01

    In France, electricity is mainly produced (78%) through the operation of 58 PWRs (Pressurized Water Reactors). During reactor operation, many fission products (FP) are generated in the fuel which is, in most cases, UO2 enriched to about 4% in 235U. Among FPs, gaseous fission products as Xenon and Krypton, are abundantly produced (around 15% stable fission products). Because of their chemical nature, those two gases have a very low solubility in the fuel and therefore tend to form bubbles (to ...

  10. Chemical and spectrochemical production analysis of ThO2 and 233UO2-ThO2 pellets for the light water breeder reactor core for Shippingport (LWBR development program)

    International Nuclear Information System (INIS)

    Bukowski, J.F.; Hollis, E.D.

    1975-06-01

    The Bettis Atomic Power Laboratory has utilized wet chemical, emission spectrochemical, and mass spectrometric analytical techniques for the production analysis of the ThO 2 and 233 UO 2 -ThO 2 (1 to 6 wt percent 233 UO 2 ) pellets for the Light Water Breeder Reactor (LWBR) core for Shippingport. Proof of the fuel breeding concept necessitates measurement of precise and accurate chemical characterization of all fuel pellets before core life. Chemistry's efforts toward this goal are presented in three main sections: (1) general discussions relating the chemical requirements for ThO 2 and 233 UO 2 -ThO 2 core materials to the analytical capabilities, (2) technical discussions of the chemical and instrumental technology applied for the analysis of aluminum, boron, calcium, carbon, chloride plus bromide, chromium, cobalt, copper, dysprosium, europium, fluoride, gadolinium, iron, magnesium, manganese, mercury, molybdenum, nickel, nitrogen, samarium, silicon, titanium, vanadium, thorium, and uranium (total, trace, and uranium VI), and (3) a formal presentation of the analytical procedures as applied to the LWBR Development Program. (U.S.)

  11. A small long-cycle PWR core design concept using fully ceramic micro-encapsulated (FCM) and UO2–ThO2 fuels for burning of TRU

    International Nuclear Information System (INIS)

    Bae, Gonghoon; Hong, Ser Gi

    2015-01-01

    In this paper, a new small pressurized water reactor (PWR) core design concept using fully ceramic micro-encapsulated (FCM) particle fuels and UO 2 –ThO 2 fuels was studied for effective burning of transuranics from a view point of core neutronics. The core of this concept rate is 100 MWe. The core designs use the current PWR-proven technologies except for a mixed use of the FCM and UO 2 –ThO 2 fuel pins of low-enriched uranium. The significant burning of TRU is achieved with tri-isotropic particle fuels of FCM fuel pins, and the ThO 2 –UO 2 fuel pins are employed to achieve long-cycle length of ∼4 EFPYs (effective full-power year). Also, the effects of several candidate materials for reflector are analyzed in terms of core neutronics because the small core size leads to high sensitivity of reflector material on the cycle length. The final cores having 10 w/o SS303 and 90 w/o graphite reflector are shown to have high TRU burning rates of 33%–35% in FCM pins and significant net burning rates of 24%–25% in the total core with negative reactivity coefficients, low power peaking factors, and sufficient shutdown margins of control rods. (author)

  12. Is there a contraction of the interatomic distance in small metal particles?

    DEFF Research Database (Denmark)

    Hansen, Lars Bruno; Stoltze, Per; Nørskov, Jens Kehlet

    1990-01-01

    A theoretical analysis is made of the bond lengths of small (100–1000 atoms) Cu particles at various temperatures. The interatomic interactions are calculated using the effective-medium theory and the finite-temperature properties obtained from a molecular-dynamics simulation. We find only very s...... small changes in bond length with particle size, but the motion in the small particles is very anharmonic. We use this observation to resolve the current experimental controversy about the existence of bond contraction for small metal particles.......A theoretical analysis is made of the bond lengths of small (100–1000 atoms) Cu particles at various temperatures. The interatomic interactions are calculated using the effective-medium theory and the finite-temperature properties obtained from a molecular-dynamics simulation. We find only very...

  13. High Pressure phase transition in some alkali halides using interatomic potential model

    International Nuclear Information System (INIS)

    Yazar, H.R.

    2002-01-01

    We have predicted the phase transition pressure in some alkali halides using an interatomic potential approach based on rigid ion model.The phase transition pressures(28.69 and 2.4 GPa) obtained by us for two alkali halides (NaCl and KCl ) are in closer agreement with their corresponding experimental data(29.0 and 2.0 GPa).This potential is promising with respect to prediction of the phase transition pressure of other alkali halides as well

  14. Determination of interatomic distance of cadmium by X-ray spectroscopy

    International Nuclear Information System (INIS)

    Sharma, S.K.; Prasad, Ram; Sharma, M.N.

    1988-01-01

    The single potential model of Lytle (1966) is employed to determine the interatomic distance in metallic Cd using experimental data (Kawata and Maeda, 1971) on X-ray Lsub(III) absorption fine structure. It has also been clarified that the linear phase shift method (Lytle et al, 1975; Stern et al, 1975) is applicable only in the case of compounds whereas the present method (Lytle, 1966) is applicable in the case of metallic crystals. (author). 8 refs

  15. Influence of Parameters of a Reactive Interatomic Potential on the Properties of Saturated Hydrocarbons

    Science.gov (United States)

    2017-01-01

    methanol, and formic acid using a reactive force field. J Mater Res. 2013;28(03):513–520. 37. Liang T, Devine B, Phillpot SR, Sinnott SB. Variable...hypercube; hydrocarbon; interatomic potential; molecular dynamics; optimization; carbon ; hydrogen 62 Mark A Tschopp 410-306-0855Unclassified Unclassified...various MEAM-based po- tentials. For example, Xiao et al.20 calculated the interaction of carbon nanotubes with Ni nanoparticles, and Uddin et al.21

  16. Development of an inter-atomic potential for the Pd-H binary system.

    Energy Technology Data Exchange (ETDEWEB)

    Zimmerman, Jonathan A.; Hoyt, Jeffrey John (McMaster University, Hamilton, Ontario, Canada); Leonard, Francois Leonard; Griffin, Joshua D.; Zhou, Xiao Wang

    2007-09-01

    Ongoing research at Sandia National Laboratories has been in the area of developing models and simulation methods that can be used to uncover and illuminate the material defects created during He bubble growth in aging bulk metal tritides. Previous efforts have used molecular dynamics calculations to examine the physical mechanisms by which growing He bubbles in a Pd metal lattice create material defects. However, these efforts focused only on the growth of He bubbles in pure Pd and not on bubble growth in the material of interest, palladium tritide (PdT), or its non-radioactive isotope palladium hydride (PdH). The reason for this is that existing inter-atomic potentials do not adequately describe the thermodynamics of the Pd-H system, which includes a miscibility gap that leads to phase separation of the dilute (alpha) and concentrated (beta) alloys of H in Pd at room temperature. This document will report the results of research to either find or develop inter-atomic potentials for the Pd-H and Pd-T systems, including our efforts to use experimental data and density functional theory calculations to create an inter-atomic potential for this unique metal alloy system.

  17. Interatomic methods for the dispersion energy derived from the adiabatic connection fluctuation-dissipation theorem

    Science.gov (United States)

    Tkatchenko, Alexandre; Ambrosetti, Alberto; DiStasio, Robert A.

    2013-02-01

    Interatomic pairwise methods are currently among the most popular and accurate ways to include dispersion energy in density functional theory calculations. However, when applied to more than two atoms, these methods are still frequently perceived to be based on ad hoc assumptions, rather than a rigorous derivation from quantum mechanics. Starting from the adiabatic connection fluctuation-dissipation (ACFD) theorem, an exact expression for the electronic exchange-correlation energy, we demonstrate that the pairwise interatomic dispersion energy for an arbitrary collection of isotropic polarizable dipoles emerges from the second-order expansion of the ACFD formula upon invoking the random-phase approximation (RPA) or the full-potential approximation. Moreover, for a system of quantum harmonic oscillators coupled through a dipole-dipole potential, we prove the equivalence between the full interaction energy obtained from the Hamiltonian diagonalization and the ACFD-RPA correlation energy. This property makes the Hamiltonian diagonalization an efficient method for the calculation of the many-body dispersion energy. In addition, we show that the switching function used to damp the dispersion interaction at short distances arises from a short-range screened Coulomb potential, whose role is to account for the spatial spread of the individual atomic dipole moments. By using the ACFD formula, we gain a deeper understanding of the approximations made in the interatomic pairwise approaches, providing a powerful formalism for further development of accurate and efficient methods for the calculation of the dispersion energy.

  18. Facilitating the selection and creation of accurate interatomic potentials with robust tools and characterization

    International Nuclear Information System (INIS)

    Trautt, Zachary T; Tavazza, Francesca; Becker, Chandler A

    2015-01-01

    The Materials Genome Initiative seeks to significantly decrease the cost and time of development and integration of new materials. Within the domain of atomistic simulations, several roadblocks stand in the way of reaching this goal. While the NIST Interatomic Potentials Repository hosts numerous interatomic potentials (force fields), researchers cannot immediately determine the best choice(s) for their use case. Researchers developing new potentials, specifically those in restricted environments, lack a comprehensive portfolio of efficient tools capable of calculating and archiving the properties of their potentials. This paper elucidates one solution to these problems, which uses Python-based scripts that are suitable for rapid property evaluation and human knowledge transfer. Calculation results are visible on the repository website, which reduces the time required to select an interatomic potential for a specific use case. Furthermore, property evaluation scripts are being integrated with modern platforms to improve discoverability and access of materials property data. To demonstrate these scripts and features, we will discuss the automation of stacking fault energy calculations and their application to additional elements. While the calculation methodology was developed previously, we are using it here as a case study in simulation automation and property calculations. We demonstrate how the use of Python scripts allows for rapid calculation in a more easily managed way where the calculations can be modified, and the results presented in user-friendly and concise ways. Additionally, the methods can be incorporated into other efforts, such as openKIM. (paper)

  19. Interactions in Zircaloy/UO2 fuel rod bundles with Inconel spacers at temperatures above 1200deg C (posttest results of severe fuel damage experiments CORA-2 and CORA-3)

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Schanz, G.; Sepold, L.

    1990-09-01

    In the CORA experiments test bundles of usually 16 electrically heated fuel rod simulators and nine unheated rods are subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence is simulated, which may develop from a small-break loss-of-coolant accident of an LWR. An aim of CORA-2, as a first test of its kind, was also to gain experience in the test conduct and posttest handling of UO 2 specimens. CORA-3 was performed as a high-temperature test. The transient phases of CORA-2 and CORA-3 were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy(Zry)-steam reaction started at about 1000deg C, leading the bundles to maximum temperatures of 2000deg C and 2400deg C for tests CORA-2 and CORA-3, respectively. The test bundles resulted in severe oxidation and partial melting of the cladding, fuel dissolution by Zry/UO 2 interaction, complete Inconel spacer destruction, and relocation of melts and fragments to lower elevations in the bundle, where extended blockages have formed. In both tests the fuel rod destruction set in together with the formation of initial melts from the Inconel/Zry interaction. The lower Zry spacer acted as a catcher for relocated material. In test CORA-2 the UO 2 pellets partially disintegrated into fine particles. This powdering occurred during cooldown. There was no physical disintegration of fuel in test CORA-3. (orig./MM) [de

  20. Étude par calcul de structure électronique des dégâts d'irradiation dans le combustible nucléaire UO2 : comportement des défauts ponctuels et gaz de fission

    OpenAIRE

    VATHONNE, EMERSON

    2014-01-01

    Uranium dioxide (UO2) is worldwide the most widely used fuel in nuclear plants in the world and in particular in pressurized water reactors (PWR). In-pile the fission of uranium nuclei creates fission products and point defects in the fuel. The understanding of the evolution of these radiation damages requires a multi-scale modelling approach of the nuclear fuel, from the scale of the pellet to the atomic scale. We used an electronic structure calculation method based on the density functiona...

  1. Interatomic interaction of additive elements and their influence on the processes in the double metal solutions

    Directory of Open Access Journals (Sweden)

    Марина Анатоліівна Рябікіна

    2016-07-01

    Full Text Available Modern industry uses a lot of elements as additives to improve the service characteristics of metal products that are to be used for various purposes. These elements can be divided into two groups: the first group includes the elements interacting with iron and improving its characteristics (alloying elements, and the second group includes the elements, that modify the characteristics of the structure and properties in an undesirable direction. These are trace elements: S, P, O, As, and others in steel. The negative impact of these elements shows itself as banding, the formation of non-metallic inclusions, flakes, grain boundary segregations et al. The influence of the elements of the both groups on the properties of steel depends on the nature and level of interatomic interaction in the alloy. Computational and analytical study of the major impurity elements in steel impact on the interatomic bond strength and the probability of forming complexes, clusters, and chemical compounds with the basic alloying elements in the steel has been carried out in the work. The theoretical parameter which defines the strength of the ion-covalent bond of two atoms: non-metallic – metallic is the electronegativity of elements. The electronegativity difference of the metal and non-metallic elements increasing, the ionic bonding and thermodynamic stability of these compounds  increase. On the other hand, concentration of valent electrons is a universal characteristic of an atomic element which determines many of its properties, and especially the energy of interatomic interaction. Energy calculations of pairwise interatomic impurity elements: H, C, N, S, P, As interaction with Fe and major alloying elements in steel: Mn, Cr, Si, V, Al, Ti, W, Cu, Mo, Nb were made. It has been stated that all the impurity elements except phosphorus, hydrogen and arsenic have sufficient high adhesion with the majority of the metal elements in the modern steels. Phosphorus does

  2. Phonon density of states for solid uranium: Accuracy of the embedded atom model classical interatomic potential

    Science.gov (United States)

    Antropov, A. S.; Fidanyan, K. S.; Stegailov, V. V.

    2018-01-01

    An accurate computation of the vibrational properties of a crystal lattice, such as phonon density of states and dispersion curves, is necessary for the description of thermodynamic properties of the solid state as well as defect migration rates. In this work, we use a simple embedded atom model classical interatomic potential. The phonon density of states for the α and γ phases of uranium at different temperatures was calculated by three methods: the lattice dynamics approach, the Fourier transformation of the velocity autocorrelation function and the Green’s function method for lattice dynamics.

  3. The utilization of thorium in light water reactors - a comparison between a PWR fuel element using UO2 with a seed blanket fuel element using (Th-U)O2

    International Nuclear Information System (INIS)

    Fernandes, Ana C.A.A.; Maiorino, Jose R.; Carluccio, Thiago; Russo, Pedro C.R.

    2011-01-01

    In this work a preliminary study on the utilization of thorium in light water reactor is made. The work compares the performance of a typical PWR fuel element, 17X17, Pitch = 1.43 cm, using UO 2 , 3 w/0 enriched(reference) with one seed blanket fuel element, consisting of the (Th 0.9 U 0.1 ) O 2 in the blanket, and UO 2 in the seed, with 20 w/0 of uranium enrichment and the same reference pitch. The parameters of the comparison were, 1) the production of Plutonium, minor actinides (MA) and long lived fission products(LLFP) after 300 days of burn up in order to assess the radio toxicity of these two fuel elements, 2) the conversion ratio, 3) k eff versus time. All the calculations were made using TRITON/NEWT modules of SCALE 5.1, with a library 238groupndf from ENDFB-VI. The results shows that the seed-blanket (Th-U)O 2 produces less plutonium, MA and LLPP than the reference fuel element, demonstrating that it has a lower radio toxicity and therefore more attractive for spent fuel storage. Also the neutronic performance shows advantages for the (Th-U)O 2 . (author)

  4. Physical characterization and reactivity of the uranyl peroxide [UO2(η(2)-O2)(H2O)2]·2H2O: implications for storage of spent nuclear fuels.

    Science.gov (United States)

    Mallon, Colm; Walshe, Aurora; Forster, Robert J; Keyes, Tia E; Baker, Robert J

    2012-08-06

    The unusual uranyl peroxide studtite, [UO(2)(η(2)-O(2))(H(2)O)(2)]·2H(2)O, is a phase alteration product of spent nuclear fuel and has been characterized by solid-state cyclic voltammetry. The voltammogram exhibits two reduction waves that have been assigned to the U(VI/V) redox couple at -0.74 V and to the U(V/IV) redox couple at -1.10 V. This potential shows some dependence upon the identity of the cation of the supporting electrolyte, where cations with larger ionic radii exhibit more cathodic reduction potentials. Raman spectroelectrochemistry indicated that exhaustive reduction at either potential result in a product that does not contain peroxide linkers and is likely to be UO(2). On the basis of the reduction potentials, the unusual behavior of neptunium in the presence of studtite can be rationalized. Furthermore, the oxidation of other species relevant to the long-term storage of nuclear fuel, namely, iodine and iodide, has been explored. The phase altered product should therefore be considered as electrochemically noninnocent. Radiotracer studies with (241)Am show that it does not interact with studtite so mobility will not be retarded in repositories. Finally, a large difference in band gap energies between studtite and its dehydrated congener metastudtite has been determined from the electronic absorption spectra.

  5. Study of interatomic potential and thermal structural properties of β-Zn4Sb3

    International Nuclear Information System (INIS)

    Li, Guodong; Li, Yao; Liu, Lisheng; Zhang, Qingjie; Zhai, Pengcheng

    2012-01-01

    Highlights: ► The multi-body interatomic potentials of various models of β-Zn 4 Sb 3 have been developed to describe atomic interactions. ► The radial distribution function shows that the 10% vacancy of Zn site leads to the disorder of β-Zn 4 Sb 3 . ► The 10% vacancy of Zn site is the main cause of the exceptional low thermal conductivity. -- Abstract: Previous experimental research shows that the disordered Zn atoms in β-Zn 4 Sb 3 may have an important influence on its exceptionally low thermal conductivity and easily occurred phase transition. So the present work aims to study the influence of disordered Zn atoms on thermodynamics properties of β-Zn 4 Sb 3 by using molecular dynamics (MD) method. Firstly, based on first principles calculation and experimental results, the interatomic potentials of β-Zn 4 Sb 3 and MD analysis method are established, and the feasibility is verified. Then, the influence of disordered Zn atoms on thermal conductivity of β-Zn 4 Sb 3 is studied in detail. The simulation results indicate that the 10% vacant Zn atoms is the main reason for the exceptionally low thermal conductivity of β-Zn 4 Sb 3 , and it seems that the interstitial Zn atoms have little effect on its thermal conductivity.

  6. Non-empirical interatomic potentials for transition metals, alloys, and semiconductors

    Science.gov (United States)

    Progress has been made on several fronts in the development and application of simplified energy and force functionals. These elucidate the basic features of bulk and defect structures, and are being coded in a form which can be used in atomistic simulations of materials properties. The main categories of materials which we have treated are transition metals, semiconductors, and aluminum alloys. We have analyzed the basic form of the angular dependence of the interatomic forces in these materials. We have then applied this understanding to the structures of polytetrahedrally packed transition metal compounds, icosahedral phase in the Ti-Mn system, and complex phases in Al-transition metal alloys. A force code for use in atomistic simulations of Si has also been developed. The Principal Investigator has completed a major review article on interatomic potentials for Solid State Physics: Advances in Research and Applications. The significance of the research accomplishment has also been recognized by several invited lectures, as well as solicitation to write an article entitled Cohesion (physics) for the upcoming new edition of the McGraw-Hill Encyclopedia of Science and Technology.

  7. The effect of interatomic potential in molecular dynamics simulation of low energy ion implantation

    International Nuclear Information System (INIS)

    Chan, H.Y.; Nordlund, K.; Peltola, J.; Gossmann, H.-J.L.; Ma, N.L.; Srinivasan, M.P.; Benistant, F.; Chan, Lap

    2005-01-01

    Being able to accurately predict dopant profiles at sub-keV implant energies is critical for the microelectronic industry. Molecular Dynamics (MD), with its capability to account for multiple interactions as energy lowers, is an increasingly popular simulation method. We report our work on sub-keV implantation using MD and investigate the effect of different interatomic potentials on the range profiles. As an approximation, only pair potentials are considered in this work. Density Functional Theory (DFT) is used to calculate the pair potentials for a wide range of dopants (B, C, N, F, Si, P, Ga, Ge, As, In and Sb) in single crystalline silicon. A commonly used repulsive potential is also included in the study. Importance of the repulsive and attractive regions of the potential has been investigated with different elements and we show that a potential depicting the right attractive forces is especially important for heavy elements at low energies

  8. Total scattering cross sections and interatomic potentials for neutral hydrogen and helium on some noble gases

    International Nuclear Information System (INIS)

    Ruzic, D.N.; Cohen, S.A.

    1985-04-01

    Measurements of energy-dependent scattering cross sections for 30 to 1800 eV D incident on He, Ne, Ar, and Kr, and for 40 to 850 eV He incident on He, Ar, and Kr are presented. They are determined by using the charge-exchange efflux from the Princeton Large Torus tokamak as a source of D or He. These neutrals are passed through a gas-filled scattering cell and detected by a time-of-flight spectrometer. The cross section for scattering greater than the effective angle of the apparatus (approx. =20 mrad) is found by measuring the energy-dependent attenuation of D or He as a function of pressure in the scattering cell. The interatomic potential is extracted from the data

  9. Hypothetical planar and nanotubular crystalline structures with five interatomic bonds of Kepler nets type

    Directory of Open Access Journals (Sweden)

    Aleksey I. Kochaev

    2017-02-01

    Full Text Available The possibility of metastable existence of planar and non-chiral nanotubular crystalline lattices in the form of Kepler nets of 34324, 3342, and 346 types (the notations are given in Schläfly symbols, using ab initio calculations, has researched. Atoms of P, As, Sb, Bi from 15th group and atoms of S, Se, Te from 16th group of the periodic table were taken into consideration. The lengths of interatomic bonds corresponding to the steadiest states for such were determined. We found that among these new composed structures crystals encountered strong elastic properties. Besides, some of them can possess pyroelectric and piezoelectric properties. Our results can be used for nanoelectronics and nanoelectromechanical devices designing.

  10. Evidence of interatomic Coulombic decay in ArKr after Ar 2p Auger decay

    International Nuclear Information System (INIS)

    Morishita, Y; Saito, N; Suzuki, I H; Fukuzawa, H; Liu, X-J; Sakai, K; Pruemper, G; Ueda, K; Iwayama, H; Nagaya, K; Yao, M; Kreidi, K; Schoeffler, M; Jahnke, T; Schoessler, S; Doerner, R; Weber, T; Harries, J; Tamenori, Y

    2008-01-01

    We have identified interatomic Coulombic decay (ICD) processes in the ArKr dimer following Ar 2p Auger decay, using momentum-resolved electron-ion-ion coincidence spectroscopy and simultaneously determining the kinetic energy of the ICD electron and the KER between Ar 2+ and Kr + . We find that the spin-conserved ICD processes in which Ar 2+ (3p -3 3d) 1 P and 3 P decay to Ar 2+ (3p -2 ) 1 D and 3 P, respectively, ionizing the Kr atom, are significantly stronger than the spin-flip ICD processes in which Ar 2+ (3p -3 3d) 1 P and 3 P decay to Ar 2+ (3p -2 ) 3 P and 1 D, respectively

  11. Collective excitation frequencies and vortices of a Bose-Einstein condensed state with gravitylike interatomic attraction

    International Nuclear Information System (INIS)

    Ghosh, Tarun Kanti

    2002-01-01

    We study the collective excitations of a neutral atomic Bose-Einstein condensate with gravitylike 1/r interatomic attraction induced by an electromagnetic wave. Using the time-dependent variational approach, we derive an analytical spectrum for monopole and quadrupole mode frequencies of a gravitylike self-bound Bose condensed state at zero temperature. We also analyze the excitation frequencies of the Thomas-Fermi-gravity (TF-G) and gravity (G) regimes. Our result agrees excellently with that of Giovanazzi et al. [Europhysics Lett., 56, 1 (2001)], which is obtained within the sum-rule approach. We also consider the vortex state. We estimate the superfluid coherence length and the critical angular frequencies to create a vortex around the z axis. We find that the TF-G regime can exhibit the superfluid properties more prominently than the G regime. We find that the monopole mode frequency of the condensate decreases due to the presence of a vortex

  12. [H2bipy]2[(UO2)6Zn2(PO3OH)4(PO4)4].H2O: an open-framework uranyl zinc phosphate templated by diprotonated 4,4'-bipyridyl.

    Science.gov (United States)

    Yu, Yaqin; Zhan, Wei; Albrecht-Schmitt, Thomas E

    2008-10-06

    Under mild hydrothermal conditions, a new organically templated uranyl zinc phosphate, [H 2bipy] 2[(UO 2) 6Zn 2(PO 3OH) 4(PO 4) 4].H 2O ( UZnP-2), has been synthesized. Structural analysis reveals that UZnP-2 is constructed from UO 7 pentagonal bipyramids that are linked into edge-sharing dimers that are in turn joined together by ZnO 4 and PO 4 tetrahedra to form a three-dimensional network. Intersecting channels occur along the a, b, and c axes. These channels house the diprotonated 4,4'-bipyridyl cations and water molecules. Ion-exchange experiments demonstrate that replacement of the 4,4'-bipyridyl cations by alkali and alkaline-earth metal cations results in a rearrangement of the framework. Further characterization of UZnP-2 is provided by Raman and fluorescence spectroscopy. The latter method reveals strong emission from the uranyl moieties with characteristic fine structure.

  13. Molten salt flux synthesis and crystal structure of a new open-framework uranyl phosphate Cs3(UO2)2(PO4)O2: Spectroscopic characterization and cationic mobility studies

    Science.gov (United States)

    Yagoubi, S.; Renard, C.; Abraham, F.; Obbade, S.

    2013-04-01

    The reaction of triuranyl diphosphate tetrahydrate precursor (UO2)3(PO4)2(H2O)4 with a CsI flux at 750 °C yields a yellow single crystals of new compound Cs3(UO2)2(PO4)O2. The crystal structure (monoclinic, space group C2/c, a=13.6261 (13) Å, b=8.1081(8) Å, c=12.3983(12) Å, β=114.61(12)°, V=1245.41(20) Å3 with Z=4) has been solved using direct methods and Fourier difference techniques. A full-matrix least-squares refinement on the basis of F2 yielded R1=0.028 and wR2=0.071 for 79 parameters and 1352 independent reflections with I≥2σ(I) collected on a BRUKER AXS diffractometer with MoKα radiation and a charge-coupled device detector. The crystal structure is built by two independent uranium atoms in square bipyramidal coordination, connected by two opposite corners to form infinite chains [UO5]∞1 and by one phosphorus atom in a tetrahedral environment PO4. The two last entities [UO5]∞1 and PO4 are linked by sharing corners to form a three-dimensional structure presenting different types of channels occupied by Cs+ alkaline cations. Their mobility within the tunnels were studied between 280 and 800 °C and compared with other tunneled uranyl minerals. The infrared spectrum shows a good agreement with the values inferred from the single crystal structure analysis of uranyl phosphate compound.

  14. Interatomic Coulombic Decay Effects in Theoretical DNA Recombination Systems Involving Protein Interaction Sites

    Science.gov (United States)

    Vargas, E. L.; Rivas, D. A.; Duot, A. C.; Hovey, R. T.; Andrianarijaona, V. M.

    2015-03-01

    DNA replication is the basis for all biological reproduction. A strand of DNA will ``unzip'' and bind with a complimentary strand, creating two identical strands. In this study, we are considering how this process is affected by Interatomic Coulombic Decay (ICD), specifically how ICD affects the individual coding proteins' ability to hold together. ICD mainly deals with how the electron returns to its original state after excitation and how this affects its immediate atomic environment, sometimes affecting the connectivity between interaction sites on proteins involved in the DNA coding process. Biological heredity is fundamentally controlled by DNA and its replication therefore it affects every living thing. The small nature of the proteins (within the range of nanometers) makes it a good candidate for research of this scale. Understanding how ICD affects DNA molecules can give us invaluable insight into the human genetic code and the processes behind cell mutations that can lead to cancer. Authors wish to give special thanks to Pacific Union College Student Senate in Angwin, California, for their financial support.

  15. Simulated carbon irradiation of carbon nanotubes – A comparative study of interatomic potentials

    Energy Technology Data Exchange (ETDEWEB)

    Heredia-Avalos, Santiago, E-mail: sheredia@ua.es [Departament de Física, Enginyeria de Sistemes i Teoria de la Senyal, Universitat d’Alacant, Apartat 99, E-03080 Alacant (Spain); Moreno-Marín, Juan Carlos [Departament de Física, Enginyeria de Sistemes i Teoria de la Senyal, Universitat d’Alacant, Apartat 99, E-03080 Alacant (Spain); Denton, Cristian D. [Departament de Física Aplicada, Universitat d’Alacant, Apartat 99, E-03080 Alacant (Spain)

    2014-05-01

    We simulate the irradiation of carbon nanotubes (CNT) with carbon ions using a molecular dynamics code. In order to describe the interaction between carbon ions we use the Tersoff or Brenner potential, both joined smoothly to the Universal ZBL potential at short distances. We have analyzed the defects produced after irradiation, the subsequent modification of the CNT structure, and their dependence on the used interatomic potential, the projectile energy (from 10 eV to 5 keV) and the dose. For single projectile irradiation, we have obtained that the coordination defect number increases with the projectile energy, although a saturation value is achieved at high projectile energies (∼3 keV). For continuous projectile irradiation, we have observed that for low energies (∼10 eV) the accumulation of adatoms produces a bump in the irradiated region. However, at intermediate energies (∼100 eV) the irradiation produces vacancies which are healed through non-hexagonal rings. This gives rise to a shrinking of the CNT diameter in the irradiated region. Finally, if the projectile energy is high enough (∼1 keV) the continuous irradiation produces the breaking of the CNT.

  16. Improved parameterization of interatomic potentials for rare gas dimers with density-based energy decomposition analysis.

    Science.gov (United States)

    Zhou, Nengjie; Lu, Zhenyu; Wu, Qin; Zhang, Yingkai

    2014-06-07

    We examine interatomic interactions for rare gas dimers using the density-based energy decomposition analysis (DEDA) in conjunction with computational results from CCSD(T) at the complete basis set (CBS) limit. The unique DEDA capability of separating frozen density interactions from density relaxation contributions is employed to yield clean interaction components, and the results are found to be consistent with the typical physical picture that density relaxations play a very minimal role in rare gas interactions. Equipped with each interaction component as reference, we develop a new three-term molecular mechanical force field to describe rare gas dimers: a smeared charge multipole model for electrostatics with charge penetration effects, a B3LYP-D3 dispersion term for asymptotically correct long-range attractions that is screened at short-range, and a Born-Mayer exponential function for the repulsion. The resulted force field not only reproduces rare gas interaction energies calculated at the CCSD(T)/CBS level, but also yields each interaction component (electrostatic or van der Waals) which agrees very well with its corresponding reference value.

  17. Interatomic electron transport by semiempirical and ab initio tight-binding approaches

    Science.gov (United States)

    Turek, I.; Kudrnovský, J.; Drchal, V.; Szunyogh, L.; Weinberger, P.

    2002-03-01

    A unified approach to interatomic electron transport within Kubo linear-response theory is sketched that is applicable both in semiempirical (matrix-element-based) and ab initio (wave-function-based) tight-binding (TB) techniques. This approach is based on a systematic neglect of the electron motion inside the atomic (Wigner-Seitz) cells leading thus to velocity operators describing pure intersite hopping. This is achieved by using piecewise constant coordinates, i.e., coordinates that are constant inside the cells. The formalism is presented within the simple semiempirical TB method, the TB linear muffin-tin orbital (LMTO) method, and the screened Korringa-Kohn-Rostoker (KKR) method with emphasis on the formal analogy of the derived formulas. The results provide a justification of current assumptions used in semiempirical TB schemes, an assessment of properties of recent TB-LMTO approaches, and an alternative formulation of electron transport within the screened KKR method. The formalism is illustrated by a calculation of residual resistivity of substitutionally disordered fcc Ag-Pd alloys.

  18. Fitness club

    CERN Multimedia

    Fitness club

    2011-01-01

    General fitness Classes Enrolments are open for general fitness classes at CERN taking place on Monday, Wednesday, and Friday lunchtimes in the Pump Hall (building 216). There are shower facilities for both men and women. It is possible to pay for 1, 2 or 3 classes per week for a minimum of 1 month and up to 6 months. Check out our rates and enrol at: http://cern.ch/club-fitness Hope to see you among us! CERN Fitness Club fitness.club@cern.ch  

  19. Burn-up Credit Criticality Safety Benchmark-Phase II-E. Impact of Isotopic Inventory Changes due to Control Rod Insertions on Reactivity and the End Effect in PWR UO2 Fuel Assemblies

    International Nuclear Information System (INIS)

    Neuber, Jens Christian; Tippl, Wolfgang; Hemptinne, Gwendoline de; Maes, Philippe; Ranta-aho, Anssu; Peneliau, Yannick; Jutier, Ludyvine; Tardy, Marcel; Reiche, Ingo; Kroeger, Helge; Nakata, Tetsuo; Armishaw, Malcom; Miller, Thomas M.

    2015-01-01

    The report describes the final results of the Phase II-E Burn-up Credit Criticality Benchmark conducted by the Expert Group on Burn-up Credit Criticality Safety. The objective of Phase II of the Burn-up Credit Criticality Safety programme is to study the impact of axial burn-up profiles of PWR UO 2 spent fuel assemblies on the reactivity of PWR UO 2 spent fuel assembly configurations. The objective of the Phase II-E benchmark was to study the impact of changes on the spent nuclear fuel isotopic composition due to control rod insertion during depletion on the reactivity and the end effect of spent fuel assemblies with realistic axial burn-up profiles for different control rod insertion depths ranging from 0 cm (no insertion) to full insertion (i.e. to the case that the fuel assemblies were exposed to control rod insertion over their full active length). For this purpose two axial burn-up profiles have been extracted from an AREVA-NP-GmbH-owned 17x17-(24+1) PWR UO 2 spent fuel assembly burn-up profile database. One profile has an average burn-up of 30 MWd/kg U, the other profile is related to an average burn-up of 50 MWd/kg U. Two profiles with different average burn-up values were selected because the shape of the burn-up profile is affected by the average burn-up and the end effect depends on the average burn-up of the fuel. The Phase II-E benchmark exercise complements the Phase II-C and Phase II-D benchmark exercises. In Phase II-D different irradiation histories were analysed using different control rod insertion histories during depletion as well as irradiation histories without control rod insertion. But in all the histories analysed a uniform distribution of the burn-up and hence a uniform distribution of the isotopic composition were assumed; and in all the histories including any usage of control rods full insertion of the control rods was assumed. In Phase II-C the impact of the asymmetry of axial burn-up profiles on the reactivity and the end effect of

  20. Repulsive interatomic potentials for noble gas bombardment of Cu and Ni targets

    Energy Technology Data Exchange (ETDEWEB)

    Karolewski, M.A. [Department of Chemistry, University of Brunei Darussalam, Jalan Tungku Link, Gadong BE 1410 (Brunei Darussalam)]. E-mail: mkarol@fos.ubd.edu.bn

    2006-01-15

    Interatomic potentials that are relevant for noble gas bombardment of Cu and Ni targets have been calculated in the energy region below 10 keV. Potentials are calculated for the diatomic species: NeCu, ArCu, KrCu, Cu{sub 2}, ArNi, Ni{sub 2} and NiCu. The calculations primarily employ density functional theory (with the B3LYP exchange-correlation functional). Potential curves derived from Hartree-Fock theory calculations are also discussed. Scalar relativistic effects have been included via the second-order Douglas-Kroll-Hess (DKH2) method. On the basis of a variational argument, it can be shown that the predicted potential curves represent an upper limit to the true potential curves. The potentials provide a basis for assessing corrections required to the ZBL and Moliere screened Coulombic potentials, which are typically found to be too repulsive below 1-2 keV. These corrections significantly improve the accuracy of the sputter yield predicted by molecular dynamics for Ni(1 0 0), whereas the sputter yield predicted for Cu(1 0 0) is negligibly affected. The validity of the pair potential approximation in the repulsive region of the potential is tested by direct calculation of the potentials arising from the interaction of either an Ar or Cu atom with a Cu{sub 3} cluster. The pairwise approximation represents the Ar-Cu{sub 3} potential energy function with an error <3 eV at all Ar-Cu{sub 3} separations. For Cu-Cu{sub 3}, the pairwise approximation underestimates the potential by ca. 10 eV when the interstitial atom is located near the centre of the cluster.

  1. Zircaloy-sheathed element rods fitted with thermo-couples

    International Nuclear Information System (INIS)

    Bernardy de Sigoyer, B.; Jacques, F.; Thome, P.

    1963-01-01

    In order to carry out thermal conductivity measurements on UO 2 in conditions similar to those under which fuel rods are used, it was necessary to measure the temperature at the interior of a fuel element sheathed in zircaloy. The temperatures are taken with Thermocoax type thermocouples, that is to say fitted with a very thin sheath of stainless steel or Inconel. It is known also that fusion welding of zircaloy onto stainless steel is impossible and that high temperature welded joints are very difficult because of their aggressiveness. The technique used consists in brazing the thermocouples to relatively large stainless steel parts and then joining these plugs by electron bombardment welding to diffused stainless steel-zircaloy couplings. The properties of these diffused couplings and of the brazed joints were studied; the various stages in the fabrication of the containers are also described. (authors) [fr

  2. Fitness Club

    CERN Multimedia

    Fitness Club

    2011-01-01

    The CERN Fitness Club is organising Zumba Classes on the first Wednesday of each month, starting 7 September (19.00 – 20.00). What is Zumba®? It’s an exhilarating, effective, easy-to-follow, Latin-inspired, calorie-burning dance fitness-party™ that’s moving millions of people toward joy and health. Above all it’s great fun and an excellent work out. Price: 22 CHF/person Sign-up via the following form: https://espace.cern.ch/club-fitness/Lists/Zumba%20Subscription/NewForm.aspx For more info: fitness.club@cern.ch

  3. Fodbold Fitness

    DEFF Research Database (Denmark)

    Bennike, Søren

    Samfundet forandrer sig og ligeså gør danskernes idrætsmønstre. Fodbold Fitness, der er afhandlingens omdrejningspunkt, kan iagttages som en reaktion på disse forandringer. Afhandlingen ser nærmere på Fodbold Fitness og implementeringen af dette, der ingenlunde er nogen let opgave. Bennike bidrager...

  4. Crystal fields in UO2 - revisited

    Energy Technology Data Exchange (ETDEWEB)

    Nakotte, Heinz [Los Alamos National Laboratory; Rajatram, R [NMSU/UNIV OF N.C.; Kern, S [COLORADO STATE UNIV; Mcqueeney, R J [AMES LAB; Lander, G H [EUROPEAN COMMISIONS, JRC; Robinson, R A [BRAGG INSTITUTE

    2009-01-01

    We performed inelastic neutron scattering (INS) in order to re-investigate the crystal-field ground state and the level splitting in UO{sub 2}. Previous INS studies on UO{sub 2} by Amorelli et al. [Physical Review B 15, 1989, 1856] uncovered four excitations at low temperatures in the 150-180 meV range. Considering the dipole-allowed transitions, only three of these transitions could be explained by the published crystal-field model. Our INS results on a different UO{sub 2} sample revealed that the unaccounted peak at about 180 meV is a spurious one, and thus not intrinsic to UO{sub 2}. In good agreement with Amoretti's results, we corroborated that the ground-state of UO{sub 2} is the {Lambda}{sub 5} triplet, and we computed that the fourth- and six-order crystal field parameters are V{sub 4} = -116 meV and V{sub 6} = 26 meV, respectively. We also studied the INS response of the non-magnetic U{sub 0.4}Th{sub 0.6}O{sub 2}. The splitting for this thorium-doped compound is similar to the one of UO{sub 2}, which orders antiferromagnetically at low temperatures. Therefore, we can conclude that magnetic interactions only weakly perturb the energy level splitting, which is dominated by strong crystal fields.

  5. Fitness cost

    DEFF Research Database (Denmark)

    Nielsen, Karen L.; Pedersen, Thomas M.; Udekwu, Klas I.

    2012-01-01

    phage types, predominantly only penicillin resistant. We investigated whether isolates of this epidemic were associated with a fitness cost, and we employed a mathematical model to ask whether these fitness costs could have led to the observed reduction in frequency. Bacteraemia isolates of S. aureus...... from Denmark have been stored since 1957. We chose 40 S. aureus isolates belonging to phage complex 83A, clonal complex 8 based on spa type, ranging in time of isolation from 1957 to 1980 and with varyous antibiograms, including both methicillin-resistant and -susceptible isolates. The relative fitness...... of each isolate was determined in a growth competition assay with a reference isolate. Significant fitness costs of 215 were determined for the MRSA isolates studied. There was a significant negative correlation between number of antibiotic resistances and relative fitness. Multiple regression analysis...

  6. Synthesis and crystal structure of new uranyl selenite(IV)-selenate(VI) [C5H14N][(UO2)3(SeO4)4(HSeO3)(H2O)](H2SeO3)(HSeO4)

    International Nuclear Information System (INIS)

    Krivovichev, S.V.; Tananaev, I.G.; Myasoedov, B.F.; Kalenberg, V.

    2006-01-01

    Crystals of new uranyl selenite(IV)-selenate(VI) [C 5 H 14 N][(UO 2 ) 3 (SeO 4 ) 4 (HSeO 3 )(H 2 O)](H 2 SeO 3 )(HSeO 4 ) are obtained by the method of evaporation from aqueous solutions. Compound has triclinic lattice, space group P1-bar, a=11.7068(9), b=14.8165(12), c=16.9766(15), α=73.899(6), β=76.221(7), γ=89.361(6) Deg, V=2743.0(4) A 3 , Z=2. Laminated complexes (UO 2 ) 3 (SeO 4 ) 4 (HSeO 3 )(H 2 O)] 3- are the basis of the structure. [HSe(VI)O 4 ] - , [H 2 Se(IV)O 3 ] complexes and protonated methylbutylamine cations are disposed between layers [ru

  7. Quantum dynamics of a BEC interacting with a single-mode quantized field in the presence of interatom collisions

    Energy Technology Data Exchange (ETDEWEB)

    Ghasemian, E. [Atomic and Molecular Group, Faculty of Physics, Yazd University, Yazd (Iran, Islamic Republic of); Tavassoly, M.K., E-mail: mktavassoly@yazd.ac.ir [Atomic and Molecular Group, Faculty of Physics, Yazd University, Yazd (Iran, Islamic Republic of); Photonics Research Group, Engineering Research Center, Yazd University, Yazd (Iran, Islamic Republic of); The Laboratory of Quantum Information Processing, Yazd University, Yazd (Iran, Islamic Republic of)

    2016-09-23

    In this paper, we consider a model in which N two-level atoms in a Bose–Einstein condensate (BEC) interact with a single-mode quantized laser field. Our goal is to investigate the quantum dynamics of atoms in the BEC in the presence of interatom interactions. To achieve the purpose, at first, using the collective angular momentum operators, we try to reduce the dynamical Hamiltonian of the system to a well-known Jaynes–Cummings like model (JCM). We also use the Dicke model to construct the state of atomic subsystem, by which the analytical solution of the system may be obtained. Then, we analyze the atomic population inversion, the degree of entanglement between the “atoms in BEC” and the “field” as well as the Mandel parameter. Numerical results show that, the atomic population inversion, atom-field entanglement and quantum statistics of photons are very sensitive to the evolved parameters in the model (and so can be well-adjusted), such as the number of atoms in BEC, the intensity of initial field, the interatom coupling constant and detuning. To investigate the entanglement properties, we pay attention to the entropy and linear entropy. It is shown that, oscillations in the two entropy criteria may be seen, with some maxima of entanglement at some moments of time. Finally, looking for the quantum statistics, we evaluate the Mandel parameter, by which we demonstrate the sub-Poissonian statistics and so the nonclassical characteristics of the field state of system. Collapse-revival phenomenon, which is a distinguishable nonclassical characteristic of the system, can be apparently observed in the atomic population inversion and the Mandel parameter. - Highlights: • N two-level atoms in a BEC interacting with a laser field in the presence of interatom interactions is considered. • The atomic population inversion, degree of entanglement between the “atoms in BEC” and the “field” and the Mandel parameter are investigated. • Collapse

  8. Conceptual and practical bases for the high accuracy of machine learning interatomic potentials: Application to elemental titanium

    Science.gov (United States)

    Takahashi, Akira; Seko, Atsuto; Tanaka, Isao

    2017-11-01

    Machine learning interatomic potentials (MLIPs) based on a large data set obtained by density functional theory calculation have been developed recently. This study gives both conceptual and practical bases for the high accuracy of MLIPs, although MLIPs have been considered to be simply an accurate black-box description of atomic energy. We also construct the most accurate MLIP of elemental Ti ever reported using a linearized MLIP framework and many angular-dependent descriptors, which also corresponds to a generalization of the modified embedded atom method potential.

  9. Temperature effects on the occurrence of long interatomic distances in atomic chains formed from stretched gold nanowires

    Energy Technology Data Exchange (ETDEWEB)

    Lagos, M J [Laboratorio Nacional de Luz SIncrotron, CP 6192, 13084-971 Campinas, SP (Brazil); Autreto, P A S; Rodrigues, V; Galvao, D S; Ugarte, D [Instituto de Fisica Gleb Wataghin, UNICAMP, CP 6165, 13083-970 Campinas, SP (Brazil); Legoas, S B [Departamento de Fisica, CCT, Universidade Federal de Roraima, 69304-000 Boa Vista, RR (Brazil); Sato, F, E-mail: dmugarte@ifi.unicamp.br [Departamento de Fisica, ICE, Universidade Federal de Juiz de Fora, 36036-330 Juiz de Fora, MG (Brazil)

    2011-03-04

    The origin of long interatomic distances in suspended gold atomic chains formed from stretched nanowires remains the object of debate despite the large amount of theoretical and experimental work. Here, we report new atomic resolution electron microscopy observations acquired at room and liquid-nitrogen temperatures and theoretical results from ab initio quantum molecular dynamics on chain formation and stability. These new data are suggestive that the long distances are due to contamination by carbon atoms originating from the decomposition of adsorbed hydrocarbon molecules.

  10. Fitness Basics

    Science.gov (United States)

    ... on staying active , playing sports , and special fitness gear . Focus on fun. Pick activities you enjoy so ... 27, 2015 Page last updated June 22, 2015 top About this site Mission Statement Privacy Policy For ...

  11. Fitness Club

    CERN Multimedia

    Fitness Club

    2012-01-01

    Open to All: http://cern.ch/club-fitness  fitness.club@cern.ch Boxing Your supervisor makes your life too tough ! You really need to release the pressure you've been building up ! Come and join the fit-boxers. We train three times a week in Bd 216, classes for beginners and advanced available. Visit our website cern.ch/Boxing General Fitness Escape from your desk with our general fitness classes, to strengthen your heart, muscles and bones, improve you stamina, balance and flexibility, achieve new goals, be more productive and experience a sense of well-being, every Monday, Wednesday and Friday lunchtime, Tuesday mornings before work and Thursday evenings after work – join us for one of our monthly fitness workshops. Nordic Walking Enjoy the great outdoors; Nordic Walking is a great way to get your whole body moving and to significantly improve the condition of your muscles, heart and lungs. It will boost your energy levels no end. Pilates A body-conditioning technique de...

  12. Fitness Club

    CERN Multimedia

    Fitness Club

    2012-01-01

      The CERN Fitness Club is pleased to announce its new early morning class which will be taking place on: Tuesdays from 24th April 07:30 to 08:15 216 (Pump Hall, close to entrance C) – Facilities include changing rooms and showers. The Classes: The early morning classes will focus on workouts which will help you build not only strength and stamina, but will also improve your balance, and coordination. Our qualified instructor Germana will accompany you throughout the workout  to ensure you stay motivated so you achieve the best results. Sign up and discover the best way to start your working day full of energy! How to subscribe? We invite you along to a FREE trial session, if you enjoy the activity, please sign up via our website: https://espace.cern.ch/club-fitness/Activities/SUBSCRIBE.aspx. * * * * * * * * Saturday 28th April Get in shape for the summer at our fitness workshop and zumba dance party: Fitness workshop with Germana 13:00 to 14:30 - 216 (Pump Hall) Price...

  13. Fitness club

    CERN Multimedia

    Fitness club

    2013-01-01

      Nordic Walking Classes Come join the Nordic walking classes and outings offered by the CERN Fitness Club starting September 2013. Our licensed instructor Christine offers classes for people who’ve never tried Nordic Walking and who would like to learn the technique, and outings for people who have completed the classes and enjoy going out as a group. Course 1: Tuesdays 12:30 - 13:30 24 September, 1 October, 8 October, 15 October Course 2: Tuesdays 12:30 - 13:30 5 November, 12 November, 19 November, 26 November Outings will take place on Thursdays (12:30 to 13:30) from 12 September 2013. We meet at the CERN Club Barracks car park (close to Entrance A) 10 minutes before departure. Prices: 50 CHF for 4 classes, including the 10 CHF Club membership. Payments made directly to instructor. Renting Poles: Poles can be rented from Christine at 5 CHF / hour. Subscription: Please subscribe at: http://cern.ch/club-fitness Looking forward to seeing you among us! Fitness Club FitnessClub@c...

  14. Fitness Club

    CERN Multimedia

    Fitness Club

    2010-01-01

    Nordic Walking Please note that the subscriptions for the general fitness classes from July to December are open: Subscriptions general fitness classes Jul-Dec 2010 Sign-up to the Fitness Club mailing list here Nordic Walking: Sign-up to the Nordic Walking mailing list here Beginners Nordic walking lessons Monday Lunchtimes (rdv 12:20 for 12:30 departure) 13.09/20.09/27.09/04.10 11.10/18.10/08.11/15.11 22.11/29.11/06.12/20.12 Nordic walking lessons Tuesday evenings (rdv 17:50 for 18:00 departure) 07.09/14.09/21.09/28.09 05.10/12.10/19.10/26.10 Intermediate/Advanced Nordic walking outings (follow the nordic walking lessons before signing up for the outings) every Thursday from 16.09 - 16.12, excluding 28.10 and 09.12 Subscriptions and info: fitness.club@cern.ch  

  15. Fitness Club

    CERN Multimedia

    Fitness Club

    2012-01-01

    Get in Shape for Summer with the CERN Fitness Club Saturday 23 June 2012 from 14:30 to 16.30 (doors open at 14.00) Germana’s Fitness Workshop. Build strength and stamina, sculpt and tone your body and get your heart pumping with Germana’s workout mixture of Cardio Attack, Power Pump, Power Step, Cardio Combat and Cross-Training. Where: 216 (Pump room – equipped with changing rooms and showers). What to wear: comfortable clothes and indoor sports shoes + bring a drink! How much: 15 chf Sign up here: https://espace.cern.ch/club-fitness/Lists/Test_Subscription/NewForm.aspx? Join the Party and dance yourself into shape at Marco + Marials Zumba Masterclass. Saturday 30 June 2012 from 15:00 to 16:30 Marco + Mariel’s Zumba Masterclass Where: 216 (Pump room – equipped with changing rooms and showers). What to wear: comfortable clothes and indoor sports shoes + bring a drink! How much: 25 chf Sign up here: https://espace.cern.ch/club-fitness/Lists/Zumba%20...

  16. Cognitive fitness.

    Science.gov (United States)

    Gilkey, Roderick; Kilts, Clint

    2007-11-01

    Recent neuroscientific research shows that the health of your brain isn't, as experts once thought, just the product of childhood experiences and genetics; it reflects your adult choices and experiences as well. Professors Gilkey and Kilts of Emory University's medical and business schools explain how you can strengthen your brain's anatomy, neural networks, and cognitive abilities, and prevent functions such as memory from deteriorating as you age. The brain's alertness is the result of what the authors call cognitive fitness -a state of optimized ability to reason, remember, learn, plan, and adapt. Certain attitudes, lifestyle choices, and exercises enhance cognitive fitness. Mental workouts are the key. Brain-imaging studies indicate that acquiring expertise in areas as diverse as playing a cello, juggling, speaking a foreign language, and driving a taxicab expands your neural systems and makes them more communicative. In other words, you can alter the physical makeup of your brain by learning new skills. The more cognitively fit you are, the better equipped you are to make decisions, solve problems, and deal with stress and change. Cognitive fitness will help you be more open to new ideas and alternative perspectives. It will give you the capacity to change your behavior and realize your goals. You can delay senescence for years and even enjoy a second career. Drawing from the rapidly expanding body of neuroscience research as well as from well-established research in psychology and other mental health fields, the authors have identified four steps you can take to become cognitively fit: understand how experience makes the brain grow, work hard at play, search for patterns, and seek novelty and innovation. Together these steps capture some of the key opportunities for maintaining an engaged, creative brain.

  17. An atomic string model for a screw dislocation in iron: Implications for the development of interatomic potentials

    International Nuclear Information System (INIS)

    Gilbert, M.R.; Dudarev, S.L.; Chiesa, S.; Derlet, P.M.

    2009-01-01

    Thermally activated motion of screw dislocations is the rate-determining mechanism for plastic deformation and fracture of body centred cubic (bcc) metals and alloys. Recent experimental observations by S.G. Roberts' group at Oxford showed that ductile-brittle behaviour of bcc vanadium, tungsten, pure iron, and iron-chromium alloys is controlled by an Arrhenius process in which the energy for thermal activation is proportional to the formation energy for a double kink on a b= 1/2 screw dislocation, where b is the Burgers vector of the dislocation. Interpreting these experimental observations and extending the analysis to the case of irradiated materials requires developing a full quantitative treatment for perfect and kinked screw dislocations. Modelling screw dislocations also presents a challenge for the development of interatomic potentials. Recent density functional theory (DFT) calculations have revealed that the ground-state structure of the core of screw dislocations in all the bcc transition metals is non-degenerate and symmetric, whereas inter-atomic potentials used in molecular dynamics simulations for these metals often predict a degenerate, symmetry-broken core-structure. In this work we show how, by treating the structure of a screw dislocation within a multistring Frenkel-Kontorova model, we can develop a criterion that guarantees the correct symmetric core of the dislocation. Extending this treatment, we find a systematic recipe for constructing Finnis-Sinclair-type potentials that are able, as a matter of routine, produce non-degenerate core structures of 1/2 screw dislocations. Modelling thermally activated mobility of screw dislocations also requires that the transition pathway between stable core positions of a dislocation is accurately reproduced. DFT data indicates that the shape of the 'Peierls energy barrier' is a single-hump curve, including transitional configurations close to the so-called 'hard' structure. Interatomic potentials have, up

  18. Fitness Club

    CERN Multimedia

    Fitness Club

    2012-01-01

    Nordic Walking Classes Sessions of four classes of one hour each are held on Tuesdays. RDV barracks parking at Entrance A, 10 minutes before class time. Session 1 =  11.09 / 18.09 / 25.09 / 02.10, 18:15 - 19:15 Session 2 = 25.09 / 02.10 / 09.10 / 16.10, 12:30 - 13:30 Session 3 = 23.10 / 30.10 / 06.11 / 13.11, 12:30 - 13:30 Session 4 = 20.11 / 27.11 / 04.12 / 11.12, 12:30 - 13:30 Prices 40 CHF per session + 10 CHF club membership 5 CHF/hour pole rental Check out our schedule and enroll at http://cern.ch/club-fitness   Hope to see you among us!  fitness.club@cern.ch In spring 2012 there was a long-awaited progress in CERN Fitness club. We have officially opened a Powerlifting @ CERN, and the number of members of the new section has been increasing since then reaching 70+ people in less than 4 months. Powerlifting is a strength sport, which is simple as 1-2-3 and efficient. The "1-2-3" are the three basic lifts (bench press...

  19. Expanding the Chemistry of Actinide Metallocene Bromides. Synthesis, Properties and Molecular Structures of the Tetravalent and Trivalent Uranium Bromide Complexes: (C5Me4R2UBr2, (C5Me4R2U(O-2,6-iPr2C6H3(Br, and [K(THF][(C5Me4R2UBr2] (R = Me, Et

    Directory of Open Access Journals (Sweden)

    Alejandro G. Lichtscheidl

    2016-01-01

    Full Text Available The organometallic uranium species (C5Me4R2UBr2 (R = Me, Et were obtained by treating their chloride analogues (C5Me4R2UCl2 (R = Me, Et with Me3SiBr. Treatment of (C5Me4R2UCl2 and (C5Me4R2UBr2 (R = Me, Et with K(O-2,6-iPr2C6H3 afforded the halide aryloxide mixed-ligand complexes (C5Me4R2U(O-2,6-iPr2C6H3(X (R = Me, Et; X = Cl, Br. Complexes (C5Me4R2U(O-2,6-iPr2C6H3(Br (R = Me, Et can also be synthesized by treating (C5Me4R2U(O-2,6-iPr2C6H3(Cl (R = Me, Et with Me3SiBr, respectively. Reduction of (C5Me4R2UCl2 and (C5Me4R2UBr2 (R = Me, Et with KC8 led to isolation of uranium(III “ate” species [K(THF][(C5Me52UX2] (X = Cl, Br and [K(THF0.5][(C5Me4Et2UX2] (X = Cl, Br, which can be converted to the neutral complexes (C5Me4R2U[N(SiMe32] (R = Me, Et. Analyses by nuclear magnetic resonance spectroscopy, X-ray crystallography, and elemental analysis are also presented.

  20. Searching for globally optimal functional forms for interatomic potentials using genetic programming with parallel tempering.

    Science.gov (United States)

    Slepoy, A; Peters, M D; Thompson, A P

    2007-11-30

    Molecular dynamics and other molecular simulation methods rely on a potential energy function, based only on the relative coordinates of the atomic nuclei. Such a function, called a force field, approximately represents the electronic structure interactions of a condensed matter system. Developing such approximate functions and fitting their parameters remains an arduous, time-consuming process, relying on expert physical intuition. To address this problem, a functional programming methodology was developed that may enable automated discovery of entirely new force-field functional forms, while simultaneously fitting parameter values. The method uses a combination of genetic programming, Metropolis Monte Carlo importance sampling and parallel tempering, to efficiently search a large space of candidate functional forms and parameters. The methodology was tested using a nontrivial problem with a well-defined globally optimal solution: a small set of atomic configurations was generated and the energy of each configuration was calculated using the Lennard-Jones pair potential. Starting with a population of random functions, our fully automated, massively parallel implementation of the method reproducibly discovered the original Lennard-Jones pair potential by searching for several hours on 100 processors, sampling only a minuscule portion of the total search space. This result indicates that, with further improvement, the method may be suitable for unsupervised development of more accurate force fields with completely new functional forms. Copyright (c) 2007 Wiley Periodicals, Inc.

  1. Fitness club

    CERN Multimedia

    Fitness club

    2013-01-01

    Nordic Walking Classes New session of 4 classes of 1 hour each will be held on Tuesdays in May 2013. Meet at the CERN barracks parking at Entrance A, 10 minutes before class time. Dates and time: 07.05, 14.05, 21.05 and 28.05, fom  12 h 30 to 13 h 30 Prices: 40 CHF per session + 10 CHF club membership – 5 CHF / hour pole rental Check out our schedule and enroll at http://cern.ch/club-fitness Hope to see you among us! 

  2. NSUSY fits

    CERN Document Server

    Espinosa, José R; Sanz, Verónica; Trott, Michael

    2012-01-01

    We perform a global fit to Higgs signal-strength data in the context of light stops in Natural SUSY. In this case, the Wilson coefficients of the higher dimensional operators mediating g g -> h and h -> \\gamma \\gamma, given by c_g, c_\\gamma, are related by c_g = 3 (1 + 3 \\alpha_s/(2 \\pi)) c_\\gamma/8. We examine this predictive scenario in detail, combining Higgs signal-strength constraints with recent precision measurements of m_W, b-> s \\gamma constraints and direct collider bounds on weak scale SUSY, finding regions of parameter space that are consistent with all of these constraints. However it is challenging for the allowed parameter space to reproduce the observed Higgs mass value with sub-TeV stops. We discuss some of the direct stop discovery prospects and show how global Higgs fits can be used to exclude light stop parameter space difficult to probe by direct collider searches. We determine the current status of such indirect exclusions and estimate their reach by the end of the 8 TeV LHC run.

  3. Polyoxometal cations within polyoxometalate anions. Seven-coordinate uranium and zirconium heteroatom groups in [(UO2)12(μ3-O)4(μ2-H2O)12(P2W15O56)4]32- and [Zr4(μ3-O)2(μ2-OH)2(H2O)4 (P2W16O59)2]14-

    Science.gov (United States)

    Gaunt, Andrew J.; May, Iain; Collison, David; Travis Holman, K.; Pope, Michael T.

    2003-08-01

    Two new composite polyoxotungstate anions with unprecedented structural features, [(UO2)12(μ3-O)4(μ2-H2O)12(P2W15O56)4]32- (1) and [Zr4(μ3-O)2(μ2-OH)2(H2O)4 (P2W16O59)2]14- (2) contain polyoxo-uranium and -zirconium clusters as bridging units. The anions are synthesized by reaction of Na12[P2W15O56] with solutions of UO2(NO3)2 and ZrCl4. The structure of 1 in the sodium salt contains four [P2W15O56]12- anions assembled into an overall tetrahedral cluster by means of trigonal bridging groups formed by three equatorial-edge-shared UO7 pentagonal bipyramids. The structure of anion 2 consists of a centrosymmetric assembly of two [P2W16O59]12- anions linked by a {Zr4O2(OH)2(H2O)4}10+ cluster. Both complexes in solution yield the expected two-line 31P-NMR spectra with chemical shifts of -2.95, -13.58 and -6.45, -13.69 ppm, respectively.

  4. Interatomic potential to predict the favored and optimized compositions for ternary Cu-Zr-Hf metallic glasses

    International Nuclear Information System (INIS)

    Luo, S. Y.; Cui, Y. Y.; Dai, Y.; Li, J. H.; Liu, B. X.

    2012-01-01

    Under the framework of smoothed and long range second-moment approximation of tight-binding, a realistic interatomic potential was first constructed for the Cu-Zr-Hf ternary metal system. Applying the constructed potential, Monte Carlo simulations were carried out to compare the relative stability of crystalline solid solution versus its disordered counterpart over the entire composition triangle of the system (as a function of alloy composition). Simulations not only reveal that the origin of metallic glass formation but also determine, in the composition triangle, a quadrilateral region, within which metallic glass formation is energetically favored. It is proposed to define the energy differences between the crystalline solid solutions and disordered states as the driving force for amorphization and the corresponding calculations pinpoint an optimized composition locating at an composition of Cu 55 Zr 10 Hf 35 , around which the driving force for metallic glass formation reaches its maximum, suggesting that the ternary Cu-Zr-Hf metallic glasses designed to have the compositions around Cu 55 Zr 10 Hf 35 could be more stable than other alloys in the system. Moreover, for the Cu 55 Zr 10 Hf 35 metallic glass, the Voronoi tessellation calculations reveal some interesting features of its atomic configurations and coordination polyhedra distribution.

  5. Activation enthalpy of self-diffusion in pure metals interpreted as a measure of their interatomic force constant

    International Nuclear Information System (INIS)

    Turkdogan, E.T.

    2002-01-01

    The present study of self-diffusion data for pure solid elements has revealed that the diffusivities of body centered cubic (b.c.c.), face centered cubic (f.c.c.) and hexagonal close packed (h.c.p.) metals at their melting point temperatures (T m ) do relate in a systematic manner to the atomic number of the elements and their Period numbers. Also, the activation enthalpy of self-diffusion varies in a regular manner with the atomic number and Period number of the elements. It is surmised from these regularities that the activation enthalpy (Q) of self-diffusion may be considered as a direct measure of the interatomic force constant ε/k = (φ/R)T m where φ = Q/T m is the enthalpy coefficient. For solid metals ε/k = (15 to 19) T m , K; for non-metals the T m coefficient is (20 to 35), for rare-earth metals (10 to 13) and for liquid metals (3.32 ± 0.24). A comparison is made of these ε/k values with those derived from the standard enthalpies of dissociation of inorganic compounds to the constituent elements. A critical assessment is made of the experimental self-diffusion data for pure elements which conform to the observed regularities in the self-diffusion parameters in relation to the atomic number and Period number of the elements. (author)

  6. Interplay of intra-atomic and interatomic effects: An investigation of the 2p core level spectra of atomic Fe and molecular FeCl2

    International Nuclear Information System (INIS)

    Richter, T.; Wolff, T.; Zimmermann, P.; Godehusen, K.; Martins, M.

    2004-01-01

    The 2p photoabsorption and photoelectron spectra of atomic Fe and molecular FeCl 2 were studied by photoion and photoelectron spectroscopy using monochromatized synchrotron radiation and atomic or molecular beam technique. The atomic spectra were analyzed with configuration interaction calculations yielding excellent agreement between experiment and theory. For the analysis of the molecular photoelectron spectrum which shows pronounced interatomic effects, a charge transfer model was used, introducing an additional 3d 7 configuration. The resulting good agreement between the experimental and theoretical spectrum and the remarkable similarity of the molecular with the corresponding spectrum in the solid phase opens a way to a better understanding of the interplay of the interatomic and intra-atomic interactions in the 2p core level spectra of the 3d metal compounds

  7. Comparative analysis of termoscale effects, isomerization and stability of TM-nanoclusters (Pd,Ni,Fe and Si in dependence on interatomic potentials. MD-simulations

    Directory of Open Access Journals (Sweden)

    Galashev А.Е.

    2011-05-01

    Full Text Available Basing on the MD-simulated data the comparison of physicochemical properties of TM-nanoclusters (Pd,Ni,Fe, and Si-nanoparticles has been carried on in the purpose to understand the specificity of structure changes in depending on nature of interatomic bonds and initial structures (fcc, bcc, icosahedral – Ih. MDsimulation of thermic evolution including melting of TM- and Si- clusters was carried on up to 2000K.

  8. FITNESS USERS’ KNOWLEDGE AND ATTITUDE TOWARDS FITNESS

    Directory of Open Access Journals (Sweden)

    Đorđe Nićin

    2009-11-01

    Full Text Available Today, Fitness has become a phenomenon. It is a modern, cultivating movement that involves a lot of people of both genders,various ages, proffesions and affinities. The basic purpose of this research is the information gathering of Fitness practi- tioners’ knowledge and attitude towards Fitness. Using the Likert scale, an anonymous survey was conducted on the exampler of 91 fitness users in order to get the information on their knowledge about fitness. Based on the knowledge questionnare, next step was to analyse the attitude of users as well as to understand the relationship between the know- ledge and attitude of fitness users towards fitness.

  9. FITNESS USERS’ KNOWLEDGE AND ATTITUDE TOWARDS FITNESS

    OpenAIRE

    Đorđe Nićin; Velimir Vukajlović; Nataša Trivić

    2009-01-01

    Today, Fitness has become a phenomenon. It is a modern, cultivating movement that involves a lot of people of both genders,various ages, proffesions and affinities. The basic purpose of this research is the information gathering of Fitness practi- tioners’ knowledge and attitude towards Fitness. Using the Likert scale, an anonymous survey was conducted on the exampler of 91 fitness users in order to get the information on their knowledge about fitness. Based on the knowledge questionnare, nex...

  10. Nitroxoline Molecule: Planar or Not? A Story of Battle between π-π Conjugation and Interatomic Repulsion.

    Science.gov (United States)

    Tikhonov, Denis S; Sharapa, Dmitry I; Otlyotov, Arseniy A; Solyankin, Peter M; Rykov, Anatolii N; Shkurinov, Alexander P; Grikina, Olga E; Khaikin, Leonid S

    2018-02-15

    The conformational properties of the nitro group in nitroxoline (8-hydroxy-5-nitroquinoline, NXN) were investigated in the gas phase by means of gas electron diffraction (GED) and quantum chemical calculations, and also with solid-state analysis performed using terahertz time-domain spectroscopy (THz-TDS). The results of the GED refinement show that in the equilibrium structure the NO 2 group is twisted by angle ϕ = 8 ± 3° with respect to the 8-hydroxyoquinoline plane. This is the result of interatomic repulsion of oxygen in the NO 2 group from the closest hydrogen, which overcomes the energy gain from the π-π conjugation of the nitro group and aromatic system of 8-hydroxyoquinoline. The computation of equilibrium geometry using MP2/cc-pVXZ (X = T, Q) shows a large overestimation of the ϕ value, while DFT with the cc-pVTZ basis set performs reasonably well. On the other hand, DFT computations with double-ζ basis sets yield a planar structure of NXN. The refined potential energy surface of the torsion vibration the of nitro group in the condensed phase derived from the THz-TDS data indicates the NXN molecule to be planar. This result stays in good agreement with the previous X-ray structure determination. The strength of the π-system conjugation for the NO 2 group and 8-hydroxyoquinoline is discussed using NBO analysis, being further supported by comparison of the refined semiexperimental gas-phase structure of NXN from GED with other nitrocompounds.

  11. Theoretical studies of UO2(H2O)n(2+), NpO2(H2O)n(+), and PuO2(H2O)n(2+) complexes (n=4-6) in aqueous solution and gas phase.

    Science.gov (United States)

    Cao, Zhiji; Balasubramanian, K

    2005-09-15

    Extensive ab initio calculations both in gas phase and solution have been carried out to study the equilibrium structure, vibrational frequencies, and bonding characteristics of various actinyl (UO2(2+), NpO2(+), and PuO2(2+)) and their hydrated forms, AnO2(H2O)n(z+) (n=4, 5, and 6). Bulk solvent effects were studied using a continuum method. The geometries were fully optimized at the coupled-cluster singles + doubles (CCSD), density-functional theory (DFT), and Møller-Plesset (MP2) level of theories. In addition vibrational frequencies have been obtained at the CCSD as well as MP2/DFT levels. The results show that both the short-range and long-range solvent effects are important. The combined discrete-continuum model, in which the ionic solute and the solvent molecules in the first and second solvation shells are treated quantum mechanically while the solvent is simulated by a continuum model, can predict accurately the bonding characteristics. Moreover, our values of solvation free energies suggest that five- and six-coordinations are equally preferred for UO2(2+), and five-coordinated species are preferred for NpO2(+) and PuO2(2+). On the basis of combined quantum-chemical and continuum treatments of the hydrated complexes, we are able to determine the optimal cavity radii for the solvation models. The coupled-cluster computations with large basis sets were employed for the vibrational spectra and equilibrium geometries both of which compare quite favorably with experiment. Our most accurate computations reveal that both five- and six-coordination complexes are important for these species.

  12. Effect of a core-softened O-O interatomic interaction on the shock compression of fused silica

    Science.gov (United States)

    Izvekov, Sergei; Weingarten, N. Scott; Byrd, Edward F. C.

    2018-03-01

    Isotropic soft-core potentials have attracted considerable attention due to their ability to reproduce thermodynamic, dynamic, and structural anomalies observed in tetrahedral network-forming compounds such as water and silica. The aim of the present work is to assess the relevance of effective core-softening pertinent to the oxygen-oxygen interaction in silica to the thermodynamics and phase change mechanisms that occur in shock compressed fused silica. We utilize the MD simulation method with a recently published numerical interatomic potential derived from an ab initio MD simulation of liquid silica via force-matching. The resulting potential indicates an effective shoulder-like core-softening of the oxygen-oxygen repulsion. To better understand the role of the core-softening we analyze two derivative force-matching potentials in which the soft-core is replaced with a repulsive core either in the three-body potential term or in all the potential terms. Our analysis is further augmented by a comparison with several popular empirical models for silica that lack an explicit core-softening. The first outstanding feature of shock compressed glass reproduced with the soft-core models but not with the other models is that the shock compression values at pressures above 20 GPa are larger than those observed under hydrostatic compression (an anomalous shock Hugoniot densification). Our calculations indicate the occurrence of a phase transformation along the shock Hugoniot that we link to the O-O repulsion core-softening. The phase transformation is associated with a Hugoniot temperature reversal similar to that observed experimentally. With the soft-core models, the phase change is an isostructural transformation between amorphous polymorphs with no associated melting event. We further examine the nature of the structural transformation by comparing it to the Hugoniot calculations for stishovite. For stishovite, the Hugoniot exhibits temperature reversal and associated

  13. Unge, sundhed og fitness

    DEFF Research Database (Denmark)

    Jensen, Jens-Ole

    2003-01-01

    Artiklen redegør for udbredelsen af fitness blandt unge og diskuterer, hvor det er blevet så populært at dyrke fitness.......Artiklen redegør for udbredelsen af fitness blandt unge og diskuterer, hvor det er blevet så populært at dyrke fitness....

  14. Evaluation of interatomic potentials for rainbow scattering under axial channeling at KCl(0 0 1) surface by three-dimensional computer simulations based on binary collision approximation

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, Wataru, E-mail: take@sp.ous.ac.jp

    2017-05-01

    The rainbow angles corresponding to prominent peaks in the angular distributions of scattered projectiles with small angle, attributed to rainbow scattering (RS), under axial surface channeling conditions are strongly influenced by the interatomic potentials between projectiles and target atoms. The dependence of rainbow angles on normal energy of projectile energy to the target surface, being experimentally obtained by Specht et al. for RS of He, N, Ne and Ar atoms under 〈1 0 0〉 and 〈1 1 0〉 axial channeling conditions at a KCl(0 0 1) surface with projectile energies of 1–60 keV, was evaluated by the three-dimensional computer simulations using the ACOCT code based on the binary collision approximation with interatomic pair potentials. Good agreement between the ACOCT results using the ZBL pair potential and the individual pair potentials calculated from Hartree-Fock (HF) wave functions and the experimental ones was found for RS of He, N and Ne atoms from the atomic rows along 〈1 0 0〉 direction. For 〈1 1 0〉 direction, the ACOCT results employing the Moliere pair potential with adjustable screening length of O’Connor-Biersack (OB) formula, the ZBL pair potential and the individual HF pair potentials except for Ar → KCl using the OB pair potential are nearly in agreement with the experimental ones.

  15. Evaluation of interatomic potentials for rainbow scattering under axial channeling at KCl(0 0 1) surface by three-dimensional computer simulations based on binary collision approximation

    Science.gov (United States)

    Takeuchi, Wataru

    2017-05-01

    The rainbow angles corresponding to prominent peaks in the angular distributions of scattered projectiles with small angle, attributed to rainbow scattering (RS), under axial surface channeling conditions are strongly influenced by the interatomic potentials between projectiles and target atoms. The dependence of rainbow angles on normal energy of projectile energy to the target surface, being experimentally obtained by Specht et al. for RS of He, N, Ne and Ar atoms under and axial channeling conditions at a KCl(0 0 1) surface with projectile energies of 1-60 keV, was evaluated by the three-dimensional computer simulations using the ACOCT code based on the binary collision approximation with interatomic pair potentials. Good agreement between the ACOCT results using the ZBL pair potential and the individual pair potentials calculated from Hartree-Fock (HF) wave functions and the experimental ones was found for RS of He, N and Ne atoms from the atomic rows along direction. For direction, the ACOCT results employing the Moliere pair potential with adjustable screening length of O'Connor-Biersack (OB) formula, the ZBL pair potential and the individual HF pair potentials except for Ar → KCl using the OB pair potential are nearly in agreement with the experimental ones.

  16. Solution (31)P NMR Study of the Acid-Catalyzed Formation of a Highly Charged {U24Pp12} Nanocluster, [(UO2)24(O2)24(P2O7)12](48-), and Its Structural Characterization in the Solid State Using Single-Crystal Neutron Diffraction.

    Science.gov (United States)

    Dembowski, Mateusz; Olds, Travis A; Pellegrini, Kristi L; Hoffmann, Christina; Wang, Xiaoping; Hickam, Sarah; He, Junhong; Oliver, Allen G; Burns, Peter C

    2016-07-13

    The first neutron diffraction study of a single crystal containing uranyl peroxide nanoclusters is reported for pyrophosphate-functionalized Na44K6[(UO2)24(O2)24(P2O7)12][IO3]2·140H2O (1). Relative to earlier X-ray studies, neutron diffraction provides superior information concerning the positions of H atoms and lighter counterions. Hydrogen positions have been assigned and reveal an extensive network of H-bonds; notably, most O atoms present in the anionic cluster accept H-bonds from surrounding H2O molecules, and none of the surface-bound O atoms are protonated. The D4h symmetry of the cage is consistent with the presence of six encapsulated K cations, which appear to stabilize the lower symmetry variant of this cluster. (31)P NMR measurements demonstrate retention of this symmetry in solution, while in situ (31)P NMR studies suggest an acid-catalyzed mechanism for the assembly of 1 across a wide range of pH values.

  17. Finding Time for Fitness

    Science.gov (United States)

    ... ahead. Bring your jump-rope or choose a hotel that has fitness facilities. If you're stuck ... in-depth/fitness/art-20044531 . Mayo Clinic Footer Legal Conditions and Terms Any use of this site ...

  18. Outdoor fitness routine

    Science.gov (United States)

    ... page: //medlineplus.gov/ency/patientinstructions/000891.htm Outdoor fitness routine To use the sharing features on this ... you and is right for your level of fitness. Here are some ideas: Warm up first. Get ...

  19. Family Activities for Fitness

    Science.gov (United States)

    Grosse, Susan J.

    2009-01-01

    This article discusses how families can increase family togetherness and improve physical fitness. The author provides easy ways to implement family friendly activities for improving and maintaining physical health. These activities include: walking, backyard games, and fitness challenges.

  20. FITS: a function-fitting program

    Energy Technology Data Exchange (ETDEWEB)

    Balestrini, S.J.; Chezem, C.G.

    1982-01-01

    FITS is an iterating computer program that adjusts the parameters of a function to fit a set of data points according to the least squares criterion and then lists and plots the results. The function can be programmed or chosen from a library that is provided. The library can be expanded to include up to 99 functions. A general plotting routine, contained in the program but useful in its own right, is described separately in an Appendix.