WorldWideScience

Sample records for uranium mixed dioxide

  1. Fluorine and chlorine determination in mixed uranium-plutonium oxide fuel and plutonium dioxide

    International Nuclear Information System (INIS)

    Elinson, S.V.; Zemlyanukhina, N.A.; Pavlova, I.V.; Filatkina, V.P.; Tsvetkova, V.T.

    1981-01-01

    A technique of fluorine and chlorine determination in the mixed uranium-plutonium oxide fuel and plutonium dioxide, based on their simultaneous separation by means of pyrohydrolysis, is developed. Subsequently, fluorine is determined by photometry with alizarincomplexonate of lanthanum or according to the weakening of zirconium colouring with zylenol orange. Chlorine is determined using the photonephelometric method according to the reaction of chloride-ion interaction with silver nitrate or by spectrophotometric method according to the reaction with mercury rhodanide. The lower limit of fluorine determination is -6x10 -5 %, of chlorine- 1x10 -4 % in the sample of 1g. The relative mean quadratic deviation of the determination result (Ssub(r)), depends on the character of the material analyzed and at the content of nx10 -4 - nx10 -3 mass % is equal to from 0.05 to 0.32 for fluorine and from 0.11 to 0.35 for chlorine [ru

  2. Uranium dioxide pellets

    International Nuclear Information System (INIS)

    Zawidzki, T.W.

    1979-01-01

    Sintered uranium dioxide pellets composed of particles of size > 50 microns suitable for power reactor use are made by incorporating a small amount of sulphur into the uranium dioxide before sintering. The increase in grain size achieved results in an improvement in overall efficiency when such pellets are used in a power reactor. (author)

  3. Uranium dioxide. Sintering test

    International Nuclear Information System (INIS)

    Anon.

    Description of a sintering method and of the equipment devoted to uranium dioxide powder caracterization and comparison between different samples. Determination of the curve giving specific volume versus pressure and micrographic examination of a pellet at medium pressure [fr

  4. Production of uranium dioxide

    International Nuclear Information System (INIS)

    Hart, J.E.; Shuck, D.L.; Lyon, W.L.

    1977-01-01

    A continuous, four stage fluidized bed process for converting uranium hexafluoride (UF 6 ) to ceramic-grade uranium dioxide (UO 2 ) powder suitable for use in the manufacture of fuel pellets for nuclear reactors is disclosed. The process comprises the steps of first reacting UF 6 with steam in a first fluidized bed, preferably at about 550 0 C, to form solid intermediate reaction products UO 2 F 2 , U 3 O 8 and an off-gas including hydrogen fluoride (HF). The solid intermediate reaction products are conveyed to a second fluidized bed reactor at which the mol fraction of HF is controlled at low levels in order to prevent the formation of uranium tetrafluoride (UF 4 ). The first intermediate reaction products are reacted in the second fluidized bed with steam and hydrogen at a temperature of about 630 0 C. The second intermediate reaction product including uranium dioxide (UO 2 ) is conveyed to a third fluidized bed reactor and reacted with additional steam and hydrogen at a temperature of about 650 0 C producing a reaction product consisting essentially of uranium dioxide having an oxygen-uranium ratio of about 2 and a low residual fluoride content. This product is then conveyed to a fourth fluidized bed wherein a mixture of air and preheated nitrogen is introduced in order to further reduce the fluoride content of the UO 2 and increase the oxygen-uranium ratio to about 2.25

  5. Uranium dioxide electrolysis

    Science.gov (United States)

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

    2009-12-29

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

  6. Uranium dioxide pellets

    International Nuclear Information System (INIS)

    Zawidzki, T.W.

    1982-01-01

    A process for the preparation of a sintered, high density, large crystal grain size uranium dioxide pellet is described which involves: (i) reacting a uranyl nitrate of formula UO 2 (NO 3 ) 2 .6H 2 O with a sulphur source, at a temperature of from about 300 deg. C to provide a sulphur-containing uranium trioxide; (ii) reacting the thus-obtained modified uranium trioxide with ammonium nitrate to form an insoluble sulphur-containing ammonium uranate; (iii) neutralizing the thus-formed slurry with ammonium hydroxide to precipitate out as an insoluble ammonium uranate the remaining dissolved uranium; (iv) recovering the thus-formed precipitates in a dry state; (v) reducing the dry precipitate to UO 2 , and forming it into 'green' pellets; and (vi) sintering the pellets in a hydrogen atmosphere at an elevated temperature

  7. Uranium dioxide calcining apparatus

    International Nuclear Information System (INIS)

    Cole, E.A.; Peterson, R.S.

    1978-01-01

    This invention relates to an improved continuous calcining apparatus for consistently and controllably producing from calcinable reactive solid compounds of uranium, such as ammonium diuranate, uranium dioxide (UO 2 ) having an oxygen to uranium ratio of less than 2.2. The apparatus comprises means at the outlet end of a calciner kiln for receiving hot UO 2 , means for cooling the UO 2 to a temperature of below 100 deg C and conveying the cooled UO 2 to storage or to subsequent UO 2 processing apparatus where it finally comes into contact with air, the means for receiving cooling and conveying being sealed to the outlet end of the calciner and being maintained full of UO 2 and so operable as to exclude atmospheric oxygen from coming into contact with any UO 2 which is at elevated temperatures where it would readily oxidize, without the use of extra hydrogen gas in said means. (author)

  8. Thermal conductivity of uranium dioxide

    International Nuclear Information System (INIS)

    Pillai, C.G.S.; George, A.M.

    1993-01-01

    The thermal conductivity of uranium dioxide of composition UO 2.015 was measured from 300 to 1400 K. The phonon component of the conductivity is found to be quantitatively accounted for by the theoretical expression of Slack derived by modifying the Leibfried-Schlomann equation. (orig.)

  9. Manufacture of uranium dioxide powder

    International Nuclear Information System (INIS)

    Becker, M.

    1976-01-01

    Uranium dioxide powder is prepared by the AUC (ammonium uranyl carbonate) method. Supplementing the known process steps, the AUC, after separation from the mother liquor, is washed with an ammonium hydrogen carbonate or an NH 4 OH solution and is subsequently post-treated with a liquid which reduces the surface tension of the residual water in an AUC. Such a liquid is, for instance, alcohol

  10. Internal friction in uranium dioxide

    International Nuclear Information System (INIS)

    Paulin Filho, Pedro Iris

    1979-01-01

    The uranium dioxide inelastic properties were studied measuring internal friction at low frequencies (of the order of 1 Hz). The work was developed in the 160 to 400 deg C temperature range. The effect of stoichiometry variation was studied oxidizing the sample with consequent change of the defect structure originally present in the non-stoichiometric uranium dioxide. The presence of a wide and irregular peak due to oxidation was observed at low temperatures. Activation energy calculations indicated the occurrence of various relaxation processes and assuming the existence of a peak between - 80 and - 70 deg C , the absolute value obtained for the activation energy (0,54 eV) is consistent with the observed values determined at medium and high frequencies for the stress induced reorientation of defects. The microstructure effect on the inelastic properties was studied for stoichiometric uranium dioxide, by varying grain size and porosity. These parameters have influence on the high temperature measurements of internal friction. The internal friction variation for temperatures higher than 340 deg C is thought to be due to grain boundary relaxation phenomena. (author)

  11. A METHOD OF PREPARING URANIUM DIOXIDE

    Science.gov (United States)

    Scott, F.A.; Mudge, L.K.

    1963-12-17

    A process of purifying raw, in particular plutonium- and fission- products-containing, uranium dioxide is described. The uranium dioxide is dissolved in a molten chloride mixture containing potassium chloride plus sodium, lithium, magnesium, or lead chloride under anhydrous conditions; an electric current and a chlorinating gas are passed through the mixture whereby pure uranium dioxide is deposited on and at the same time partially redissolved from the cathode. (AEC)

  12. The cohesive energy of uranium dioxide and thorium dioxide

    International Nuclear Information System (INIS)

    Childs, B.G.

    1958-08-01

    Theoretical values have been calculated of the heats of formation of uranium dioxide and thorium dioxide on the assumption that the atomic binding forces in these solids are predominantly ionic in character. The good agreement found between the theoretical and observed values shows that the ionic model may, with care, be used in calculating the energies of defects in the uranium and thorium dioxide crystal structures. (author)

  13. Yellow cake to ceramic uranium dioxide

    International Nuclear Information System (INIS)

    Zawidzki, T.W.; Itzkovitch, I.J.

    1983-01-01

    This overview article first reviews the processes for converting uranium ore concentrates to ceramic uranium dioxide at the Port Hope Refinery of Eldorado Resources Limited. In addition, some of the problems, solutions, thoughts and research direction with respect to the production and properties of ceramic UO 2 are described

  14. Process for the preparation of uranium dioxide

    International Nuclear Information System (INIS)

    Watt, G.W.; Baugh, D.W. Jr.

    1981-01-01

    A method for the preparation of actinide dioxides using actinide nitrate hexahydrates as starting materials is described. The actinide nitrate hexahydrate is reacted with sodium dithionite, and the product is heated in the absence of oxygen to obtain the dioxide. Preferably, the actinide is uranium, plutonium or neptunium. (LL)

  15. Rapid non-destructive quantitative estimation of urania/ thoria in mixed thorium uranium di-oxide pellets by high-resolution gamma-ray spectrometry

    International Nuclear Information System (INIS)

    Shriwastwa, B.B.; Kumar, Anil; Raghunath, B.; Nair, M.R.; Abani, M.C.; Ramachandran, R.; Majumdar, S.; Ghosh, J.K.

    2001-01-01

    A non-destructive technique using high-resolution gamma-ray spectrometry has been standardised for quantitative estimation of uranium/thorium in mixed (ThO 2 -UO 2 ) fuel pellets of varying composition. Four gamma energies were selected; two each from the uranium and thorium series and the time of counting has been optimised. This technique can be used for rapid estimation of U/Th percentage in a large number of mixed fuel pellets from a production campaign

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

    International Nuclear Information System (INIS)

    Sampath, S.; Chawla, K.L.

    1975-01-01

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

  17. Extraction of Uranium Using Nitrogen Dioxide and Carbon Dioxide for Spent Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Kayo Sawada; Daisuke Hirabayashi; Youichi Enokida [EcoTopia Science Institute, Nagoya University, Nagoya, 464-8603 (Japan)

    2008-07-01

    For the reprocessing of spent nuclear fuels, a new method to extract actinides from spent fuel using highly compressed gases, nitrogen dioxide and carbon dioxide was proposed. Uranium extraction from broken pieces, whose average grain size was 5 mm, of uranium dioxide pellet with nitrogen dioxide and carbon dioxide was demonstrated in the present study. (authors)

  18. Uranium Dioxide Powder Flow ability Improvement Using Sol-Gel

    International Nuclear Information System (INIS)

    Juanda, D.; Sambodo Daru, G.

    1998-01-01

    The improvement of flow ability characteristics of uranium dioxide powder has been done using sol-gel process. To anticipate a pellet mass production with uniform pellet dimension, the uranium dioxide powder must be have a spherical form. Uranium dioxide spherical powder has been diluted in acid transformed into sol colloidal solution. To obtain uranium dioxide spherical form, the uranium sol-colloidal solution has been dropped in a hot paraffin ( at the temperature of 90 0 C) to form gelatinous colloid and then dried at 800 0 C, and sintered at the temperature of 1700 0 C. The flow ability of spherical uranium dioxide powder has been examined by using Flowmeter Hall (ASTM. B. 213-46T). The measurement result reveals that the spherical uranium dioxide powder has a flow ability twice than that of unprocessed uranium dioxide powder

  19. Rapid non-destructive quantitative estimation of urania/ thoria in mixed thorium uranium di-oxide pellets by high-resolution gamma-ray spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Shriwastwa, B.B.; Kumar, Anil; Raghunath, B.; Nair, M.R.; Abani, M.C.; Ramachandran, R.; Majumdar, S.; Ghosh, J.K

    2001-06-01

    A non-destructive technique using high-resolution gamma-ray spectrometry has been standardised for quantitative estimation of uranium/thorium in mixed (ThO{sub 2}-UO{sub 2}) fuel pellets of varying composition. Four gamma energies were selected; two each from the uranium and thorium series and the time of counting has been optimised. This technique can be used for rapid estimation of U/Th percentage in a large number of mixed fuel pellets from a production campaign.

  20. Dissolution experiments of unirradiated uranium dioxide pellets

    International Nuclear Information System (INIS)

    Ollila, K.

    1985-01-01

    The purpose of this study was to measure the dissolution rate of uranium from unirradiated uranium dioxide pellets in deionized water and natural groundwater. Moreover, the solubility limit of uranium in natural groundwater was measured. Two different temperatures, 25 and 60 deg C were used. The low oxygen content of deep groundwater was simulated. The dissolution rate of uranium varied from 10 -7 to 10 -8 g cm -2 d -1 . The rate in reionized water was one order of magnitude lower than in groundwater. No great difference was observed between the natural groundwaters with different composition. Temperature seems to have effect on the dissolution rate. The solubility limit of uranium in natural groundwater in reducing conditions, at 25 deg C, varied from 20 to 600 μg/l and in oxidizing conditions, at 60 deg C, from 4 to 17 mg/l

  1. Improved ionic model of liquid uranium dioxide

    NARCIS (Netherlands)

    Gryaznov, [No Value; Iosilevski, [No Value; Yakub, E; Fortov, [No Value; Hyland, GJ; Ronchi, C

    The paper presents a model for liquid uranium dioxide, obtained by improving a simplified ionic model, previously adopted to describe the equation of state of this substance [1]. A "chemical picture" is used for liquid UO2 of stoichiometric and non-stoichiometric composition. Several ionic species

  2. Thermal properties of nonstoichiometry uranium dioxide

    Science.gov (United States)

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

    2016-04-01

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

  3. Low temperature sintering of hyperstoichiometric uranium dioxide

    International Nuclear Information System (INIS)

    Chevrel, H.

    1991-12-01

    In the lattice of uranium dioxide with hyperstoichiometric oxygen content (UO 2+x ), each additional oxygen atoms is introduced by shifting two anions from normal sites to interstitial ones, thereby creating two oxygen vacancies. The point defects then combine to form complex defects comprising several interstitials and vacancies. The group of anions (3x) in the interstitial position participate in equilibria promoting the creation of uranium vacancies thereby considerably increasing uranium self-diffusion. However, uranium grain boundaries diffusion governs densification during the first two stages of sintering of uranium dioxide with hyperstoichiometric oxygen content, i.e., up to 93% of the theoretical density. Surface diffusion and evaporation-condensation, which are considerably accentuated by the hyperstoichiometric deviation, play an active role during sintering by promoting crystalline growth during the second and third stages of sintering. U 8 O 8 can be added to adjust the stoichiometry and to form a finely porous structure and thus increase the pore area subjected to surface phenomena. The composition with an O/U ratio equal to 2.25 is found to densify the best, despite a linear growth in sintering activation energy with hyperstoichiometric oxygen content, increasing from 300 kj.mol -1 for UO 2.10 to 440 kJ.mol -1 for UO 2.25 . Seeds can be introduced to obtain original microstructures, for example the presence of large grains in small-grain matrix

  4. Uranium dioxide calcining apparatus and method

    International Nuclear Information System (INIS)

    Cole, E.A.; Peterson, R.S.

    1978-01-01

    This invention relates to an improved continuous calcining apparatus for consistently and controllably producing from calcinable reactive solid compounds of uranium, such as ammonium diuranate, uranium dioxide (UO 2 ) having an oxygen to uranium ratio of less than 2.2. The apparatus comprises means at the outlet end of a calciner kiln for receiving hot UO 2 , means for cooling the UO 2 to a temperature of below 100 0 C and conveying the cooled UO 2 to storage or to subsequent UO 2 processing apparatus where it finally comes into contact with air, the means for receiving, cooling and conveying being sealed to the outlet end of the calciner and being maintained full of UO 2 and so operable as to exclude atmospheric oxygen from coming into contact with any UO 2 which is at elevated temperatures where it would readily oxidize, without the use of extra hydrogen gas in said means

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

    International Nuclear Information System (INIS)

    Anon.

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

  6. Process for the preparation of uranium dioxide

    International Nuclear Information System (INIS)

    Watt, G.W.; Baugh, D.W. Jr.

    1977-01-01

    An actinide dioxide, e.g., uranium dioxide, plutonium dioxide, neptunium dioxide, etc., is prepared by reacting the actinide nitrate hexahydrate with sodium dithionite as a first step; the reaction product from this first step is a novel composition of matter comprising the actinide sulfite tetrahydrate. The reaction product resulting from this first step is then converted to the actinide dioxide by heating it in the absence of an oxygen-containing atmosphere (e.g., nitrogen) to a temperature of about 500 0 to about 950 0 C for about 15 to about 135 minutes. If the reaction product resulting from the first step is, prior to carrying out the second heating step, exposed to an oxygen-containing atmosphere such as air, the resultant product is a novel composition of matter comprising the actinide oxysulfite tetrahydrate which can also be readily converted to the actinide dioxide by heating it in the absence of an oxygen-containing atmosphere (e.g., nitrogen) at a temperature of about 400 0 to about 900 0 C for about 30 to about 150 minutes. Further, the actinide oxysulfite tetrahydrate can be partially dehydrated at reduced pressures (and in the presence of a suitable dehydrating agent such as phosphorus pentoxide). The partially dehydrated product may be readily converted to the dioxide form by heating it in the absence of an oxygen-containing atmosphere (e.g., nitrogen) at a temperature of about 500 0 to about 900 0 C for about 30 to about 150 minutes. 16 claims

  7. SULPHUR DIOXIDE LEACHING OF URANIUM CONTAINING MATERIAL

    Science.gov (United States)

    Thunaes, A.; Rabbits, F.T.; Hester, K.D.; Smith, H.W.

    1958-12-01

    A process is described for extracting uranlum from uranium containing material, such as a low grade pitchblende ore, or mill taillngs, where at least part of the uraniunn is in the +4 oxidation state. After comminuting and magnetically removing any entrained lron particles the general material is made up as an aqueous slurry containing added ferric and manganese salts and treated with sulfur dioxide and aeration to an extent sufficient to form a proportion of oxysulfur acids to give a pH of about 1 to 2 but insufficient to cause excessive removal of the sulfur dioxide gas. After separating from the solids, the leach solution is adjusted to a pH of about 1.25, then treated with metallic iron in the presence of a precipitant such as a soluble phosphate, arsonate, or fluoride.

  8. Polarographic determination of uranium dioxide stoichiometry

    International Nuclear Information System (INIS)

    Viguie, J.; Uny, G.

    1966-10-01

    The method described allows the determination of small deviations from stoichiometry for uranium dioxide. It was applied to the study of surface oxidation of bulk samples. The sample is dissolved in phosphoric acid under an argon atmosphere; U(VI) is determined by polarography in PO 4 H 3 4.5 N - H 2 SO 4 4 N. U(IV) is determined by potentiometry. The detection limit is UO 2,0002 . The accuracy for a single determination at the 95% confidence level is ±20 per cent for samples with composition included between UO 2,001 and UO 2,01 . (authors) [fr

  9. Design of a Uranium Dioxide Spheroidization System

    Science.gov (United States)

    Cavender, Daniel P.; Mireles, Omar R.; Frendi, Abdelkader

    2013-01-01

    The plasma spheroidization system (PSS) is the first process in the development of tungsten-uranium dioxide (W-UO2) fuel cermets. The PSS process improves particle spherocity and surface morphology for coating by chemical vapor deposition (CVD) process. Angular fully dense particles melt in an argon-hydrogen plasma jet at between 32-36 kW, and become spherical due to surface tension. Surrogate CeO2 powder was used in place of UO2 for system and process parameter development. Particles range in size from 100 - 50 microns in diameter. Student s t-test and hypothesis testing of two proportions statistical methods were applied to characterize and compare the spherocity of pre and post process powders. Particle spherocity was determined by irregularity parameter. Processed powders show great than 800% increase in the number of spherical particles over the stock powder with the mean spherocity only mildly improved. It is recommended that powders be processed two-three times in order to reach the desired spherocity, and that process parameters be optimized for a more narrow particles size range. Keywords: spherocity, spheroidization, plasma, uranium-dioxide, cermet, nuclear, propulsion

  10. Process for preparing sintered uranium dioxide nuclear fuel

    International Nuclear Information System (INIS)

    Carter, R.E.

    1975-01-01

    Uranium dioxide is prepared for use as fuel in nuclear reactors by sintering it to the desired density at a temperature less than 1300 0 C in a chemically controlled gas atmosphere comprised of at least two gases which in equilibrium provide an oxygen partial pressure sufficient to maintain the uranium dioxide composition at an oxygen/uranium ratio of at least 2.005 at the sintering temperature. 7 Claims, No Drawings

  11. Fluorination reaction uranium dioxide by fluorine

    International Nuclear Information System (INIS)

    Ogata, Shinji; Homma, Shunji; Koga, Jiro; Matsumoto, Shiro; Sasahira, Akira; Kawamura, Fumio

    2004-01-01

    Kinetics of the fluorination reaction of uranium dioxide is studied using un-reacted core model with shrinking particles. The model includes the film mass transfer of fluorine gas and its diffusion in the particle. The rate constants of the model are determined by fitting the experimental data for 370-450degC. The model successfully represents the fluorination in this temperature range. The rate control step is identified by examining the rate constants of the model for 300-1,800degC. For temperature range up to 900degC, the fluorination reaction is rate controlling. For over 900degC, both mechanisms of the mass transfer of fluorine and the fluorination reaction control the rate of the fluorination. With further increase of the temperature, however, the fluorination reaction becomes so fast that the mass transfer of fluorine eventually controls the rate of the fluorination. (author)

  12. Investigation of transformation of uranium hexafluoride into dioxide

    International Nuclear Information System (INIS)

    Galkin, N.P.; Veryatin, U.D.; Yakhonin, I.F.; Logunov, A.F.; Dymkov, Yu.M.

    1982-01-01

    The process of transformation of uranium hexafluoride into dioxide using the method of pyrohydrolysis by steam-hydrogen mixture in a boiling layer using uranium dioxide granules applicable for production of fuel elements is considered. Technological parameters and equipment of the process are described, intermediate stages and process products are considered. Physicochemical and physicomechanical properties of the obtained uranium dioxide granules are given. The results of metallographical investigations into solid products of pyrohydrolysis in phase transformations at certain stages of the process as well as test on vibration packing of the obtained granules in fuel cans are presented

  13. Behaviour of uranium dioxide in liquid nitrogen tetraoxide

    International Nuclear Information System (INIS)

    Kobets, L.V.; Klavsut', G.N.; Dolgov, V.M.

    1983-01-01

    Interaction kinetics of uranium dioxide with liquid nitrogen tetroxide at 25-150 deg C has been studied. It is shown that in the temperature range studied NO[UO 2 (NO 3 ) 3 ] is the final product of the reaction. With the increase of specific surface of uranium dioxide and with the temperature increase the degree of oxide transformation increases. Uranium dioxide-liquid N 2 O 4 interaction proceeds in the diffusion region. Seeming activation energies and rate constants of the mentioned processes are calculated. Effect of nitrogen trioxide additions on transformation kinetics is considered

  14. Certification of a uranium dioxide reference material for chemical analyses

    International Nuclear Information System (INIS)

    Le Duigou, Y.

    1984-01-01

    This report, issued by the Central Bureau for Nuclear Measurements (CBNM), describes the characterization of a uranium dioxide reference material with accurately determined uranium mass fraction for chemical analyses. The preparation, conditioning, homogeneity tests and the analyses performed on this material are described in Annex 1. The evaluation of the individual impurity results, total of impurities and uranium mass fraction are given in Annex 2. Information on a direct determination of uranium by titration is given in Annex 3. The uranium mass fraction (881.34+-0.13) g.kg -1 calculated in Annex 2 is given on the certificate

  15. Study of non stoichiometric uranium dioxide samples (UO2)

    International Nuclear Information System (INIS)

    Moura, Sergio C.; Lima, Nelson B. de; Bustillos, Jose O.V.

    1999-01-01

    The gravimetric and voltammetric methods for determination of non-stoichiometric O/U ratio in uranium dioxide used as nuclear fuel are discussed in this work. The oxidation of uranium oxide is very complex due to many phase changes. gravimetric and voltammetric methods do not detect phase changes. The results of this work shown that, to evaluate both methods is requiring to be done Rietveld methods by x-ray diffraction data to identify the uranium oxide phase changes. (author)

  16. The preparation of uranium tetrafluoride from dioxide by aqueous way

    International Nuclear Information System (INIS)

    Aquino, A.R. de; Abrao, A.

    1990-01-01

    This paper describes the study for the wet way obtention of uranium tetrafluoride by the reaction of hydrofluoric acid and powder uranium dioxide. With the results obtained at laboratory scale a pilot plant was planned and erected. It is presently in operation for experimental data aquisition. Time of reaction, temperature, excess of reagents and the hydrofluoric acid / uranium dioxide ratio were the main parameters studied to obtain a product with the following characteristics: - density greater than 1 g/cm 3 , - conversion rate greater than 96%, -water content equal to 0,2%, that allows its application to hexafluoride convertion or to magnesiothermic process. (authOr) [pt

  17. Uranium tetrafluoride production via dioxide by wet process

    International Nuclear Information System (INIS)

    Aquino, A.R. de.

    1988-01-01

    The study for the wet way obtention of uranium tetrafluoride by the reaction of hydrofluoric acid and powder uranium dioxide, is presented. From the results obtained at laboratory scale a pilot plant was planned and erected. It is presently in operation for experimental data aquisition. Time of reaction, temperature, excess of reagents and the hydrofluoric acid / uranium dioxide ratio were the main parameters studied to obtain a product with the following characteristics: - density greater than 1 g/cm 3 , conversion rate greater than 96%, and water content equal to 0,2% that allows its application to heaxafluoride convertion or to magnesiothermic process. (author) [pt

  18. A density functional theory study of uranium-doped thoria and uranium adatoms on the major surfaces of thorium dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Shields, Ashley E. [Department of Chemistry, University College London, 20 Gordon Street, London WC1H 0AJ (United Kingdom); Santos-Carballal, David [School of Chemistry, Cardiff University, Main Building, Park Place, Cardiff CF10 3AT (United Kingdom); Leeuw, Nora H. de, E-mail: DeLeeuwN@Cardiff.ac.uk [Department of Chemistry, University College London, 20 Gordon Street, London WC1H 0AJ (United Kingdom); School of Chemistry, Cardiff University, Main Building, Park Place, Cardiff CF10 3AT (United Kingdom)

    2016-05-15

    Thorium dioxide is of significant research interest for its use as a nuclear fuel, particularly as part of mixed oxide fuels. We present the results of a density functional theory (DFT) study of uranium-substituted thorium dioxide, where we found that increasing levels of uranium substitution increases the covalent nature of the bonding in the bulk ThO{sub 2} crystal. Three low Miller index surfaces have been simulated and we propose the Wulff morphology for a ThO{sub 2} particle and STM images for the (100), (110), and (111) surfaces studied in this work. We have also calculated the adsorption of a uranium atom and the U adatom is found to absorb strongly on all three surfaces, with particular preference for the less stable (100) and (110) surfaces, thus providing a route to the incorporation of uranium into a growing thoria particle. - Highlights: • Uranium substitution in ThO{sub 2} is found to increase the covalent nature of the ionic bonding. • The (111), (110), and (100) surfaces of ThO{sub 2} are studied and the particle morphology is proposed. • STM images of the (111), (110), and (100) surfaces of ThO{sub 2} are simulated. • Uranium adsorption on the major surfaces of ThO{sub 2} is studied.

  19. Operating conditions of T.B.P. line uranium purification plant, for uranium dioxide production

    International Nuclear Information System (INIS)

    Vardich, R.N.; La Gamma, A.M.; Anasco, R.; Soler, S.M.G. de; Isnardi, E.; Gea, V.; Chiaraviglio, R.; Matyjasczyk, E.; Aramayo, R.

    1992-01-01

    In this contribution are presented the operative conditions and the results obtained step of the Uranium dioxide production plant of Argentina. The refining step involve the Uranium concentrate dissolution, the silica ageing, the filtration and liquid - liquid extraction with n-tributyl phosphate solution in kerosene. The established operative conditions allow to obtain Uranyl nitrate solutions of nuclear purity in industrial scale. (author)

  20. Dissolution of uranium dioxide in supercritical carbon dioxide modified with tri-n-butyl phosphate-hydrogen peroxide

    International Nuclear Information System (INIS)

    Kanekar, A.S.; Pathak, P.N.; Mohapatra, P.K.; Manchanda, V.K.

    2009-01-01

    Direct dissolution of uranium dioxide in supercritical carbon dioxide modified with tri-n-butyl phosphate (TBP) has been attempted. The effects of TBP concentration and pressure on the extraction of uranium have been studied. Addition of hydrogen peroxide in the modifier enhances the dissolution/extraction of uranium. (author)

  1. Method for preparing a sinterable uranium dioxide powder

    International Nuclear Information System (INIS)

    Thornton, T.A.; Holaday, V.D. Jr.

    1985-01-01

    This invention provides an improved method for preparing a sinterable uranium dioxide powder for the preparation of nuclear fuel, using microwave radiation in a microwave induction furnace. The starting compound may be uranyl nitrate hexahydrate, ammonium diuranate or ammonium uranyl carbonate. The starting compound is heated in a microwave induction furnace for a period of time sufficient for compound decomposition. The decomposed compound is heated in a microwave induction furnace in a reducing atmosphere for a period of time sufficient to reduce the decomposed compound to uranium dioxide powder

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

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

  3. Hot pressing of uranium nitride and mixed uranium plutonium nitride

    International Nuclear Information System (INIS)

    Chang, J.Y.

    1975-01-01

    The hot pressing characteristics of uranium nitride and mixed uranium plutonium nitride were studied. The utilization of computer programs together with the experimental technique developed in the present study may serve as a useful purpose of prediction and fabrication of advanced reactor fuel and other high temperature ceramic materials for the future. The densification of nitrides follow closely with a plastic flow theory expressed as: d rho/ dt = A/T(t) (1-rho) [1/1-(1-rho)/sup 2/3/ + B1n (1-rho)] The coefficients, A and B, were obtained from experiment and computer curve fitting. (8 figures) (U.S.)

  4. Evaluation of Hydrothermally Synthesized Uranium Dioxide for Novel Semiconductor Applications

    Science.gov (United States)

    2016-08-29

    Technology Air University Air Education and Training Command In Partial Fulfillment of the Requirements for the Degree of Doctor of Philosophy ...Senanayake, G. Waterhouse, A. Chan, T. Madey, D. Mullins and H. Idriss, "Probing Surface Oxidation of Reduced Uranium Dioxide Thin Film Using

  5. Immobilization of chlorine dioxide modified cells for uranium absorption

    International Nuclear Information System (INIS)

    He, Shengbin; Ruan, Binbiao; Zheng, Yueping; Zhou, Xiaobin; Xu, Xiaoping

    2014-01-01

    There has been a trend towards the use of microorganisms to recover metals from industrial wastewater, for which various methods have been reported to be used to improve microorganism adsorption characteristics such as absorption capacity, tolerance and reusability. In present study, chlorine dioxide(ClO 2 ), a high-efficiency, low toxicity and environment-benign disinfectant, was first reported to be used for microorganism surface modification. The chlorine dioxide modified cells demonstrated a 10.1% higher uranium adsorption capacity than control ones. FTIR analysis indicated that several cell surface groups are involved in the uranium adsorption and cell surface modification. The modified cells were further immobilized on a carboxymethylcellulose (CMC) matrix to improve their reusability. The cell-immobilized adsorbent could be employed either in a high concentration system to move vast UO 2 2+ ions or in a low concentration system to purify UO 2 2+ contaminated water thoroughly, and could be repeatedly used in multiple adsorption-desorption cycles with about 90% adsorption capacity maintained after seven cycles. - Highlights: • Chlorine dioxide was first reported to be used for microorganism surface modification. • The chlorine dioxide modified cells demonstrated a 10.1% higher uranium adsorption capacity than control ones. • The chlorine dioxide modified cells were further immobilized by carboxymethylcellulose to improve their reusability

  6. Contribution to the study of uranium dioxide aqueous corrosion mechanisms

    International Nuclear Information System (INIS)

    Gallien, J.-P.

    1994-01-01

    The corrosion of uranium dioxide by a synthetical ground water has been studied in order to understand the behaviour of nuclear fuels in the hypothesis of a direct storage. An original leaching unit has been carried out in order to control the parameters occurring in the oxidation-dissolution of the uranium dioxide and to condition the leachate (in particular the temperature and the partial pressure of the carbon dioxide). A ground water in equilibrium with the geological enveloping site has been reconstituted from data acquired on the site. The influence of two parameters has been followed: the carbon dioxide carbon pressure and the redox potential. Each experiment has been carried out at 96 C during one month and the time-history of the solutions and of the solids has been studied. In oxidizing conditions, the uranium concentration in solution has been controlled by an U(VI) complex (one oxide, one hydroxide or a carbonate). The possibility of a control by an U(IV) complex (as coffinite, uraninite or uraninite B) has been confirmed in the case of reducing leaching. An original interpretation of the Rutherford backscattering spectra has allowed to describe the decomposition of the samples in a succession of layers of different densities. A very good agreement between the analyses of the solids and those of the solutions has been obtained in the experiments occurring in reducing conditions. Complementary leaching involving solutions containing stable isotopes (deuterium, O 18 ) have revealed the formation of an hydrated layer and the contribution of grain boundaries to the corrosion phenomenon of uranium dioxide. The results of the current hydro-geochemistry study on the uranium Oklo deposit prove the realism of the experiments that have been carried out in the laboratory. (O.M.)

  7. Uranium tetracyclopentadienyl interaction with carbon oxide and dioxide

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  8. Determination of carbon chlorine and fluorine in uranium dioxide

    International Nuclear Information System (INIS)

    Kijko, N.I.; Timofeev, G.A.

    1983-01-01

    Techniques of chlorine and fluorine determination and simultaneous determination of carbon and chlorine in electrolytic uranium dioxide are described. The method of chlorine and fluorine determination is based on their separation during oxide pyrohydrolysis with subsequent spectrophotometric analysis of condensate. Lower determination limits constitute 1 μg for chlorine, 0.5 μg for fluorine. Relative standard deviation when the content of impurities analyzed is 10 -3 % constitutes 0.05-0.07

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

    International Nuclear Information System (INIS)

    Bartscher, W.

    1982-01-01

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

  10. Coarsening-densification transition temperature in sintering of uranium dioxide

    International Nuclear Information System (INIS)

    Balakrishna, Palanki; Narasimha Murty, B.; Chakraborthy, K.P.; Jayaraj, R.N.; Ganguly, C.

    2001-01-01

    The concept of coarsening-densification transition temperature (CDTT) has been proposed to explain the experimental observations of the study of sintering undoped uranium dioxide and niobia-doped uranium dioxide powder compacts in argon atmosphere in a laboratory tubular furnace. The general method for deducing CDTT for a given material under the prevailing conditions of sintering and the likely variables that influence the CDTT are described. Though the present work is specific in nature for uranium dioxide sintering in argon atmosphere, the concept of CDTT is fairly general and must be applicable to sintering of any material and has immense potential to offer advantages in designing and/or optimizing the profile of a sintering furnace, in the diagnosis of the fault in the process conditions of sintering, and so on. The problems of viewing the effect of heating rate only in terms of densification are brought out in the light of observing the undesirable phenomena of coring and bloating and causes were identified and remedial measures suggested

  11. Determination of gas residues in uranium dioxide pellets

    International Nuclear Information System (INIS)

    Riella, H.G.

    1978-01-01

    The measurement of low amounts of residual gases, excluding water, in ceramic grade uranium dioxide pellets, using high temperature vacuum extraction technique, is dealt with. The high temperature extraction gas analysis apparatus was designed and assembled for sequential analysis of up to eight uranium dioxide pellets by run. The system consists of three major units, namely outgassing unit, transfer unit and analytical unit. The whole system is evacuated to a final pressure of less then 10 -5 torr. A weighed pellet is transfered into the outgassing unit for subsequent dropping into a Platinum-Rhodium crucible which is heated inductively up to 1600 0 C during 30 minutes. The released gases are imediately transfered from the outgassing to analytical unit passing through a cold trap at -95 0 C to remove water vapor. The gases are transfered to previously calibrated volumetric bulb where the total pressure and temperature are determined. An estimate of the gas content in the pellets at STP condition is obtained from the measured volume, pressure and temperature of the gas mixture by applying ideal gases equation. Analysis to two lots (fourteen samples) of uranium dioxide pellets by the method described here indicated a mean gas content of 0,060cm 3 /g UO 2 . The lower limit of this technique is 0,03cm 3 /g UO 2 (STP). The time required for the analysis of eight pellets is about 9 hours [pt

  12. Determination of the stoichiometric ratio uranium dioxide samples

    International Nuclear Information System (INIS)

    Moura, Sergio Carvalho

    1999-01-01

    The determination of the O/U stoichiometric ratio in uranium dioxide is an important parameter in order to qualify nuclear fuels. The excess oxygen in the crystallographic structure can cause changes in the physico-chemical properties of this compound such as variation of the thermal conductivity alterations, fuel plasticity and others, affecting the efficiency of this material when it is utilized as nuclear fuel in the reactor core. The purpose of this work is to evaluate methods for the determination of uranium oxide samples from two different production processes, using gravimetric, voltammetric and X-ray diffraction techniques. After the evaluation of these techniques, the main aspect of this work is to define a reliable methodology in order to characterize the behavior of uranium oxide. The methodology used in this work consisted of two different steps: utilization of gravimetric and volumetric methods in order to determine the ratio in uranium dioxide samples; utilization of X-ray diffraction technique in order to determine the lattice parameters using patterns and application of the Rietveld method during refining of the structural data. As a result of the experimental part of this work it was found that the X-ray diffraction analysis performs better and detects the presence of more phases than gravimetric and voltammetric techniques, not sensitive enough in this detection. (author)

  13. New method for conversion of uranium hexafluoride to uranium dioxide

    International Nuclear Information System (INIS)

    Nakabayashi, S.; Suzuki, M.; Tanaka, H.

    1987-01-01

    Five different methods for conversion of UF 6 to ceramic-grade UO 2 powder have been developed to industrial scale. Two of them, the ammonium diuranate (ADU) and AUC processes, are based on precipitation of uranium compounds from aqueous solutions. The other three follow a dry route in which UF 6 is hydrolyzed and reduced by steam and hydrogen using fluidized bed techniques, rotating kilns, or flame chemistry methods. The ADU process has the advantage of flexible product powder characteristics, while disadvantages include a large quantity of waste, low powder fluidity, and a complicated process. On the other hand, the dry process using fluidized-bed techniques has the advantages of hydrofluoric acid recovery, a free-flowing powder, and process simplicity, but the disadvantages of poorer ceramic properties for the product. The new method developed at Mitsubishi Metal Corp. is a semidry process, which has well-balanced merits over the ADU process and the dry process using fluidized-bed techniques. This process is very attractive from powder characteristics, process simplicity, and waste reduction

  14. Determination of Oxygen - to - Uranium Ratio in Hyperstoichio - Metric Uranium Dioxide. RCN Report

    International Nuclear Information System (INIS)

    Tolk, A.; Lingerak, W.A.

    1970-09-01

    For the determination of the O/U ratio in hyperstoichiometric uranium dioxide we prefer the following chemical procedure. The sample is dissolved in concentrated phosphoric acid without change in valence of the uranium. Then the amount of U (VI) present in the solution is titrated with a Fe (II) - standard solution in phosphoric acid. The titrimetric end-point is detected following the ''dead-stop-end-point'' procedure. When special precautions are made the O/U value can be determined with an accuracy and precision of + 0.0001 0/U units when 500 mg sample aliquots are used. (author)

  15. Surface characterization of uranium metal and uranium dioxide using X-ray photoelectron spectroscopy

    International Nuclear Information System (INIS)

    Allen, G.C.; Trickle, I.R.; Tucker, P.M.

    1981-01-01

    X-ray photoelectron spectra of pure uranium metal and stoichiometric uranium dioxide have been obtained using an AEI ES300 spectrometer. Binding energy values for core and valence electrons have been determined using an internally calibrated energy scale and monochromatic Al Kα radiation. Satellite peaks observed accompanying certain principal core ionizations are discussed in relation to the mechanisms by which they arise. Confirmation is obtained that for stoichiometric UOsub(2.00) a single shake-up satellite is observed accompanying the U 4fsub(7/2,5/2) principal core lines, separated by 6.8 eV to higher binding energy. (author)

  16. Uranium metal and uranium dioxide powder and pellets - Determination of nitrogen content - Method using ammonia-sensing electrode. 1. ed.

    International Nuclear Information System (INIS)

    1994-01-01

    This International Standard specifies an analytical method for determining the nitrogen content in uranium metal and uranium dioxide powder and pellets. It is applicable to the determination of nitrogen, present as nitride, in uranium metal and uranium dioxide powder and pellets. The concentration range within which the method can be used is between 9 μg and 600 μg of nitrogen per gram. Interference can occur from metals which form complex ammines, but these are not normally present in significant amounts

  17. Evaluation of uranium dioxide thermal conductivity using molecular dynamics simulations

    International Nuclear Information System (INIS)

    Kim, Woongkee; Kaviany, Massoud; Shim, J. H.

    2014-01-01

    It can be extended to larger space, time scale and even real reactor situation with fission product as multi-scale formalism. Uranium dioxide is a fluorite structure with Fm3m space group. Since it is insulator, dominant heat carrier is phonon, rather than electrons. So, using equilibrium molecular dynamics (MD) simulation, we present the appropriate calculation parameters in MD simulation by calculating thermal conductivity and application of it to the thermal conductivity of polycrystal. In this work, we investigate thermal conductivity of uranium dioxide and optimize the parameters related to its process. In this process, called Green Kubo formula, there are two parameters i.e correlation length and sampling interval, which effect on ensemble integration in order to obtain thermal conductivity. Through several comparisons, long correlation length and short sampling interval give better results. Using this strategy, thermal conductivity of poly crystal is obtained and comparison with that of pure crystal is made. Thermal conductivity of poly crystal show lower value that that of pure crystal. In further study, we broaden the study to transport coefficient of radiation damaged structures using molecular dynamics. Although molecular dynamics is tools for treating microscopic scale, most macroscopic issues related to nuclear materials such as voids in fuel materials and weakened mechanical properties by radiation are based on microscopic basis. Thus, research on microscopic scale would be expanded in this field and many hidden mechanism in atomic scales will be revealed via both atomic scale simulations and experiments

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  19. Thermodynamic and transport properties of uranium dioxide and related phases

    International Nuclear Information System (INIS)

    1965-01-01

    The high melting point of uranium dioxide and its stability under irradiation have led to its use as a fuel in a variety of types of nuclear reactors. A wide range of chemical and physical studies has been stimulated by this circumstances and by the complex nature of the uranium dioxide phase itself. The boundaries of this phase widen as the temperature is increased; at 2000 deg. K a single, homogeneous phase exists from U 2.27 to a hypostoichiometric (UO 2-x ) composition, depending on the oxygen potential of the surroundings. Since there is often an incentive to operate a reactor at the maximum practicable heat rating and, therefore, maximum thermal gradient in the fuel, the determination of the physical properties of the UO 2-x phase becomes a matter of great technological importance. In addition a complex sequence of U-O phases may be formed during the preparation of powder feed material or during the sintering process; these affect the microstructure and properties of the final product and have also received much attention. 184 refs, 33 figs, 15 tabs

  20. On the nature of the phase transition in uranium dioxide

    Science.gov (United States)

    Gofryk, K.; Mast, D.; Antonio, D.; Shrestha, K.; Andersson, D.; Stanek, C.; Jaime, M.

    Uranium dioxide (UO2) is by far the most studied actinide material as it is a primary fuel used in light water nuclear reactors. Its thermal and magnetic properties remain, however, a puzzle resulting from strong couplings between magnetism and lattice vibrations. UO2 crystalizes in the face-centered-cubic fluorite structure and is a Mott-Hubbard insulator with well-localized uranium 5 f-electrons. In addition, below 30 K, a long range antiferromagnetic ordering of the electric-quadrupole of the uranium moments is observed, forming complex non-collinear 3-k magnetic structure. This transition is accompanied by Jahn-Teller distortion of oxygen atoms. It is believed that the first order nature of the transition results from the competition between the exchange interaction and the Jahn-Teller distortion. Here we present results of our extensive thermodynamic investigations on well-characterized and oriented single crystals of UO2+x (x = 0, 0.033, 0.04, and 0.11). By focusing on the transition region under applied magnetic field we are able to study the interplay between different competing interactions (structural, magnetic, and electrical), its dynamics, and relationship to the oxygen content. We will discuss implications of these results. Work supported by the Department of Energy, Office of Basic Energy Sciences, Materials Sciences, and Engineering Division.

  1. Minimization of the fission product waste by using thorium based fuel instead of uranium dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, A. Abdelghafar, E-mail: Agalahom@yahoo.com

    2017-04-01

    This research discusses the neutronic characteristics of VVER-1200 assembly fueled with five different fuel types based on thorium. These types of fuel based on mixing thorium as a fertile material with different fissile materials. The neutronic characteristics of these fuels are investigated by comparing their neutronic characteristics with the conventional uranium dioxide fuel using the MCNPX code. The objective of this study is to reduce the production of long-lived actinides, get rid of plutonium component and to improve the fuel cycle economy while maintaining acceptable values of the neutronic safety parameters such as moderator temperature coefficient, Doppler coefficient and effective delayed neutrons (β). The thorium based fuel has a more negative Doppler coefficient than uranium dioxide fuel. The moderator temperature coefficient (MTC) has been calculated for the different proposed fuels. Also, the fissile inventory ratio has been calculated at different burnup step. The use of Th-232 as a fertile material instead of U-238 in a nuclear fuel is the most promising fuel in VVER-1200 as it is the ideal solution to avoid the production of more plutonium components and long-lived minor actinides. The reactor grade plutonium accumulated in light water reactor with burnup can be recycled by mixing it with Th-232 to fuel the VVER-1200 assembly. The concentrations of Xe-135 and Sm-151 have been investigated, due to their high thermal neutron absorption cross section.

  2. Electronic structure of the actinides and their dioxides. Application to the defect formation energy and krypton solubility in uranium dioxide

    International Nuclear Information System (INIS)

    Petit, T.; CEA Centre d'Etudes de Grenoble, 38

    1996-01-01

    Uranium dioxide is the standard nuclear fuel used in French h power plants. During irradiation, fission products such as krypton and xenon are created inside fuel pellets. So, gas release could become, at very high burnup, a limiting factor in the reactor exploitation. To study this subject, we have realised calculations using the Density Functional Theory (DFT) into the Local Density Approximation (LDA) and the Atomic Sphere Approximation (ASA). First, we have validated our approach by calculating cohesive properties of thorium, protactinium and uranium metals. The good agreement between our results and experimental values implies that 5f electrons are itinerant. Calculated lattice parameter, cohesive energy and bulk modulus for uranium and thorium dioxides are in very good agreement with experiment. We show that binding between uranium and oxygen atoms is not completely ionic but partially covalent. The question of the electrical conductivity still remains an open problem. We have been able to calculate punctual defect formation energies in uranium dioxide. Accordingly to experimental observations, we find that it is easier to create a defect in the oxygen sublattice than in the uranium sublattice. Finally, we have been able to predict a probable site of krypton atoms in nuclear fuel: the Schottky trio. Experiences of Extended X-ray Absorption Fine structure Spectroscopy (EXAFS) and X-ray Photoelectron Spectroscopy (XPS) on uranium dioxide doped by ionic implantation will help us in the comprehension of the studied phenomena and the interpretation of our calculations. (author)

  3. Pyrochemical reduction of uranium dioxide and plutonium dioxide by lithium metal

    International Nuclear Information System (INIS)

    Usami, T.; Kurata, M.; Inoue, T.; Sims, H.E.; Beetham, S.A.; Jenkins, J.A.

    2002-01-01

    The lithium reduction process has been developed to apply a pyrochemical recycle process for oxide fuels. This process uses lithium metal as a reductant to convert oxides of actinide elements to metal. Lithium oxide generated in the reduction would be dissolved in a molten lithium chloride bath to enhance reduction. In this work, the solubility of Li 2 O in LiCl was measured to be 8.8 wt% at 650 deg. C. Uranium dioxide was reduced by Li with no intermediate products and formed porous metal. Plutonium dioxide including 3% of americium dioxide was also reduced and formed molten metal. Reduction of PuO 2 to metal also occurred even when the concentration of lithium oxide was just under saturation. This result indicates that the reduction proceeds more easily than the prediction based on the Gibbs free energy of formation. Americium dioxide was also reduced at 1.8 wt% lithium oxide, but was hardly reduced at 8.8 wt%

  4. Strain fields and line energies of dislocations in uranium dioxide

    International Nuclear Information System (INIS)

    Parfitt, David C; Bishop, Clare L; Wenman, Mark R; Grimes, Robin W

    2010-01-01

    Computer simulations are used to investigate the stability of typical dislocations in uranium dioxide. We explain in detail the methods used to produce the dislocation configurations and calculate the line energy and Peierls barrier for pure edge and screw dislocations with the shortest Burgers vector 1/2 . The easiest slip system is found to be the {100}(110) system for stoichiometric UO 2 , in agreement with experimental observations. We also examine the different strain fields associated with these line defects and the close agreement between the strain field predicted by atomic scale models and the application of elastic theory. Molecular dynamics simulations are used to investigate the processes of slip that may occur for the three different edge dislocation geometries and nudged elastic band calculations are used to establish a value for the Peierls barrier, showing the possible utility of the method in investigating both thermodynamic average behaviour and dynamic processes such as creep and plastic deformation.

  5. XAS characterisation of xenon bubbles in uranium dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Martin, P. [CEA Cadarache, DEN/DEC/SESC/LLCC, Bat. 130, 13108 St. Paul Lez Durance (France)], E-mail: martinp@drncad.cea.fr; Garcia, P.; Carlot, G.; Sabathier, C.; Valot, C. [CEA Cadarache, DEN/DEC/SESC/LLCC, Bat. 130, 13108 St. Paul Lez Durance (France); Nassif, V. [CEA Grenoble, DSM/DRFMC/SP2M/NRS, 17 Avenue des Martyrs, 38054 Grenoble Cedex 9 (France); Proux, O. [Laboratoire de Geophysique Interne et Tectonophysique, UMR CNRS/Universite Joseph Fourier, 1381 rue de la Piscine, Domaine Universitaire, 38400 Saint-Martin-D' Heres (France); Hazemann, J.-L. [Institut Neel, CNRS, 25 Avenue des Martyrs, BP 166, 38042 Grenoble Cedex 9 (France)

    2008-06-15

    X-ray absorption spectroscopy experiments were performed on a set of uranium dioxide samples implanted with 10{sup 17} xenon cm{sup -2} at 800 keV (8 at.% at 140 nm). EXAFS measurements performed at 12 K showed that during implantation the gas forms highly pressurised nanometre size inclusions. Bubble pressures were estimated at 2.8 {+-} 0.3 GPa at low temperature. Following the low energy xenon implantation, samples were annealed between 1073 and 1773 K for several hours. Stability of nanometre size highly pressurized xenon aggregates in UO{sub 2} is demonstrated up to 1073 K as for this temperature almost no modification of the xenon environment was observed. Above this temperature, bubbles will trap migrating vacancies and their inner pressure is seen to decrease substantially.

  6. Predictor of regulation of uranium dioxide powder pressing process

    International Nuclear Information System (INIS)

    Motta, Eduardo Souza; Araujo, Victor Hugo Leal de; Bernardelli, Sergio Henrique

    2007-01-01

    One of the most important steps of the uranium dioxide pellets fabrication used in the nuclear fuel elements is the green pellets pressing. The target density of the pellets after the sintering process determines the density of the green pellet. To meet the same sintered target density the green density may vary according to the powder characteristics. These variations implies in changing the regulation of the press for different powder's patches. The regulation done empirically imply in productivity loss and necessity of reprocessing the pellets pressed during the press regulation and also depends on the operator experience. At this work, was developed an artificial neural network feed forward back propagation to predict the press regulation, depending on the powder characteristics and the green pellet's target density. The results obtained at INB - Industrias Nucleares do Brasil S. A. during the fabrication of the fifth recharge of Angra II nuclear power plant are presented. (author)

  7. The production of sinterable uranium dioxide from ammonium diuranate

    International Nuclear Information System (INIS)

    Fane, A.G.; Le Page, A.H.

    1975-02-01

    The development of a 0.13 m diameter pulsed fluidised bed reactor for the continuous production of sinterable uranium dioxide from ammonium diuranate is described. Calcination-reduction at 670 to 680 0 C produced powders with surface areas of 4 to 6 m 2 g -1 giving pellet densities in excess of 10.6 g cm -3 . Sinterability was relatively insensitive to changes in operating conditions, provided the availability of hydrogen was adequate, for gas flow rates in the range 0.95 to 1.4 l S -1 , pulse frequencies of 0.5 and 0.75 Hz and mean residence times of the solids from 0.6 to 1.4 hours. Sinterability was shown to be improved either by use of higher input concentrations, or by use of a secondary flow of hydrogen (about 5 per cent of input) fed into the powder collection system and flowing countercurrent to the UO 2 product. The maximum throughput of 17 kg UO 2 h -1 (0.6 hours mean residence time) required only 120 per cent of the stoichiometric requirement at an input concentration of 50 vol.per cent with secondary hydrogen flow. Results are given for studies of the kinetics of reduction of calcined ammonia diuranate in hydrogen and the residence time distribution of solids in a pulsed fluidised bed. Estimates based on these data suggested that the overall conversion of ammonium diuranate to uranium dioxide in the continuously operated pulsed fluidised bed reactor was in excess of 99 per cent. Continuous stabilisation of the UO 2 product was demonstrated at 12 kg h -1 or UO 2 , in a 0.15 m diameter glass stabiliser, using 10 vol.per cent air in nitrogen and a temperature of about 50 0 C. (author)

  8. Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

    Science.gov (United States)

    Mcdonald, G. E.

    1973-01-01

    A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.

  9. Uranium dioxide sintering Kinetics and mechanisms under controlled oxygen potentials

    International Nuclear Information System (INIS)

    Freitas, C.T. de.

    1980-06-01

    The initial, intermediate, and final sintering stages of uranium dioxide were investigated as a function of stoichiometry and temperature by following the kinetics of the sintering reaction. Stoichiometry was controlled by means of the oxygen potential of the sintering atmosphere, which was measured continuously by solid-state oxygen sensors. Included in the kinetic study were microspheres originated from UO 2 gels and UO 2 pellets produced by isostatic pressing ceramic grade powders. The microspheres sintering behavior was examined using hot-stage microscopy and a specially designed high-temperature, controlled atmosphere furnace. This same furnace was employed as part of an optical dilatometer, which was utilized in the UO 2 pellet sintering investigations. For controlling the deviations from stoichiometry during heat treatment, the oxygen partial pressure in the sintering atmosphere was varied by passing the gas through a Cu-Ti-Cu oxygen trap. The trap temperature determined the oxygen partial pressure of the outflowing mixture. Dry hydrogen was also used in some of the UO sub(2+x) sintering experiments. The determination of diametrial shrinkages and sintering indices was made utilizing high-speed microcinematography and ultra-microbalance techniques. It was observed that the oxygen potential has a substantial influence on the kinetics of the three sintering stages. The control of the sintering atmosphere oxygen partial pressure led to very fast densification of UO sub(2+x). Values in the interval 95.0 to 99.5% of theoretical density were reached in less than one minute. Uranium volume diffusion is the dominant mechanism in the initial and intermediate sintering stages. For the final stage, uranium grain boundary diffusion was found to be the main sintering mechanism. (Author) [pt

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    Andrade, Antonio Santos; Ferreira, Ricardo Alberto Neto

    2007-01-01

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

  12. Uranium leaching using mixed organic acids produced by Aspergillus niger

    International Nuclear Information System (INIS)

    Yong-dong Wang; Guang-yue Li; De-xin Ding; Zhi-xiang Zhou; Qin-wen Deng; Nan Hu; Yan Tan

    2013-01-01

    Both of culture temperature and pH value had impacts on the degree of uranium extraction through changing types and concentrations of mixed organic acids produced by Aspergillus niger, and significant interactions existed between them though pH value played a leading role. And with the change of pH value of mixed organic acids, the types and contents of mixed organic acids changed and impacted on the degree of uranium extraction, especially oxalic acid, citric acid and malic acid. The mean degree of uranium extraction rose to peak when the culture temperature was 25 deg C (76.14 %) and pH value of mixed organic acids was 2.3 (82.40 %) respectively. And the highest one was 83.09 %. The optimal culture temperature (25 deg C) of A. niger for uranium leaching was different from the most appropriate growing temperature (37 deg C). (author)

  13. Investigation of the dissolution of uranium dioxide in nitric media: a new approach aiming at understanding interface mechanisms

    International Nuclear Information System (INIS)

    Delwaulle, Celine

    2011-01-01

    This research thesis deals with the back-end cycle of the nuclear fuel by improving, modernizing and optimizing the processes used for all types of fuels which are to be re-processed. After a presentation of the industrial context and of the state of the art concerning dissolution kinetic data for uranium dioxide and mixed oxide, the author proposes a model which couples dissolution kinetics and hydrodynamics of a solid in presence of auto-catalytic species, in order to better understand phenomena occurring at the solid-liquid-gas interface. The next part reports dissolution experiments on a non-radioactive material (copper) and out of a nuclear environment. Then, the author identifies steps which are required to transpose this experiment within a nuclear environment. The first results obtained on uranium dioxide are discussed. Recommendations for further studies conclude the report

  14. Fabrication of chamfered uranium-plutonium mixed carbide pellets

    International Nuclear Information System (INIS)

    Arai, Yasuo; Iwai, Takashi; Shiozawa, Kenichi; Handa, Muneo

    1985-10-01

    Chamfered uranium-plutonium mixed carbide pellets for high burnup irradiation test in JMTR were fabricated in glove boxes with purified argon gas. The size of die and punch in a press was decided from pellet densities and dimensions including the angle of chamfered parts. No chip or crack caused by adopting chamfered pellets was found in both pressing and sintering stages. In addition to mixed carbide pellets, uranium carbide pellets used as insulators were also successfully fabricated. (author)

  15. A kinetic study of the reaction of water vapor and carbon dioxide on uranium

    International Nuclear Information System (INIS)

    Santon, J.P.

    1964-09-01

    The kinetic study of the reaction of water vapour and carbon dioxide with uranium has been performed by thermogravimetric methods at temperatures between 160 and 410 deg G in the first case, 350 and 1050 deg C in the second: Three sorts of uranium specimens were used: uranium powder, thin evaporated films, and small spheres obtained from a plasma furnace. The experimental results led in the case of water vapour, to a linear rate of reaction controlled by diffusion at the lower temperatures, and by a surface reaction at the upper ones. In the case of carbon dioxide, a parabolic law has been found, controlled by diffusional processes. (author) [fr

  16. Green strength of zirconium sponge and uranium dioxide powder compacts

    International Nuclear Information System (INIS)

    Balakrishna, Palanki; Murty, B. Narasimha; Sahoo, P.K.; Gopalakrishna, T.

    2008-01-01

    Zirconium metal sponge is compacted into rectangular or cylindrical shapes using hydraulic presses. These shapes are stacked and electron beam welded to form a long electrode suitable for vacuum arc melting and casting into solid ingots. The compact electrodes should be sufficiently strong to prevent breakage in handling as well as during vacuum arc melting. Usually, the welds are strong and the electrode strength is limited by the green strength of the compacts, which constitute the electrode. Green strength is also required in uranium dioxide (UO 2 ) powder compacts, to withstand stresses during de-tensioning after compaction as well as during ejection from the die and for subsequent handling by man and machine. The strengths of zirconium sponge and UO 2 powder compacts have been determined by bending and crushing respectively, and Weibul moduli evaluated. The green density of coarse sponge compact was found to be larger than that from finer sponge. The green density of compacts from lightly attrited UO 2 powder was higher than that from unattrited category, accompanied by an improvement in UO 2 green crushing strength. The factors governing green strength have been examined in the light of published literature and experimental evidence. The methodology and results provide a basis for quality control in metal sponge and ceramic powder compaction in the manufacture of nuclear fuel

  17. Synthesis and preservation of graphene-supported uranium dioxide nanocrystals

    Energy Technology Data Exchange (ETDEWEB)

    Ma, Hanyu [Department of Civil and Environmental Engineering and Earth Sciences, University of Notre Dame, 156 Fitzpatrick Hall, Notre Dame, IN 46556 (United States); Wang, Haitao [Department of Civil and Environmental Engineering and Earth Sciences, University of Notre Dame, 156 Fitzpatrick Hall, Notre Dame, IN 46556 (United States); Department of Civil, Environmental, and Construction Engineering, Texas Tech University, 911 Boston Ave., Lubbock, TX 79409 (United States); Burns, Peter C. [Department of Civil and Environmental Engineering and Earth Sciences, University of Notre Dame, 156 Fitzpatrick Hall, Notre Dame, IN 46556 (United States); Department of Chemistry and Biochemistry, University of Notre Dame, 251 Nieuwland Science Hall, Notre Dame, IN 46556 (United States); McNamara, Bruce K.; Buck, Edgar C. [Nuclear Chemistry & Engineering Group, Pacific Northwest National Laboratory, 902 Battelle Boulevard, Richland, WA 99352 (United States); Na, Chongzheng, E-mail: chongzheng.na@gmail.com [Department of Civil and Environmental Engineering and Earth Sciences, University of Notre Dame, 156 Fitzpatrick Hall, Notre Dame, IN 46556 (United States); Department of Civil, Environmental, and Construction Engineering, Texas Tech University, 911 Boston Ave., Lubbock, TX 79409 (United States)

    2016-07-15

    Graphene-supported uranium dioxide (UO{sub 2}) nanocrystals are potentially important fuel materials. Here, we investigate the possibility of synthesizing graphene-supported UO{sub 2} nanocrystals in polar ethylene glycol compounds by the polyol reduction of uranyl acetylacetone under boiling reflux, thereby enabling the use of an inexpensive graphene precursor graphene oxide into a one-pot process. We show that triethylene glycol is the most suitable solvent with an appropriate reduction potential for producing nanometer-sized UO{sub 2} crystals compared to monoethylene glycol, diethylene glycol, and polyethylene glycol. Graphene-supported UO{sub 2} nanocrystals synthesized with triethylene glycol show evidence of heteroepitaxy, which can be beneficial for facilitating heat transfer in nuclear fuel particles. Furthermore, we show that graphene-supported UO{sub 2} nanocrystals synthesized by polyol reduction can be readily stored in alcohols, impeding oxidation from the prevalent oxygen in air. Together, these methods provide a facile approach for preparing and storing graphene-supported UO{sub 2} nanocrystals for further investigation and development under ambient conditions. - Highlights: • UO{sub 2} nanocrystals are synthesized using polyol reduction method. • Triethylene glycol is the best reducing agent for nano-sized UO{sub 2} crystals. • UO{sub 2} nanocrystals grow on graphene through heteroepitaxy. • Graphene-supported UO{sub 2} nanocrystals can be stored in alcohols to prevent oxidation.

  18. Micromechanical approach of behavior of uranium dioxide nuclear fuel

    International Nuclear Information System (INIS)

    Soulacroix, Julian

    2014-01-01

    Uranium dioxide (UO 2 ) is the reference fuel for pressurized water nuclear reactors. Our study deals with understanding and modeling of mechanical behavior at the microstructure scale at low temperatures (brittle fracture) and high temperature (viscoplastic strain). We have first studied the geometrical properties of polycrystals at large and of UO 2 polycrystal more specifically. As of now, knowledge of this behavior in the brittle fracture range is limited. Consequently, we developed an experimental method which allows better understanding of brittle fracture phenomenon at grain scale. We show that fracture is fully intra-granular and {100} planes seem to be the most preferential cleavage planes. Experimental results are directly used to deduce constitutive equations of intra-granular brittle fracture at crystal scale. This behavior is then used in 3D polycrystal simulation of brittle fracture. The full field calculation gives access to the initiation of fracture and propagation of the crack through the grains. Finally, we developed a mechanical behavior model of UO 2 in the viscoplastic range. We first present constitutive equations at macroscopic scale which accounts for an ageing process caused by migration of defects towards dislocations. Secondly, we have developed a crystal plasticity model which was fitted to UO 2 . This model includes the rotation of the crystal lattice. We present examples of polycrystalline simulations. (author) [fr

  19. A thermal modelling of displacement cascades in uranium dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Martin, G., E-mail: guillaume.martin@cea.fr [CEA – DEN/DEC/SESC/LLCC, Bât. 352, 13108 Saint-Paul-Lez-Durance Cedex (France); Garcia, P.; Sabathier, C. [CEA – DEN/DEC/SESC/LLCC, Bât. 352, 13108 Saint-Paul-Lez-Durance Cedex (France); Devynck, F.; Krack, M. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Maillard, S. [CEA – DEN/DEC/SESC/LLCC, Bât. 352, 13108 Saint-Paul-Lez-Durance Cedex (France)

    2014-05-01

    The space and time dependent temperature distribution was studied in uranium dioxide during displacement cascades simulated by classical molecular dynamics (MD). The energy for each simulated radiation event ranged between 0.2 keV and 20 keV in cells at initial temperatures of 700 K or 1400 K. Spheres into which atomic velocities were rescaled (thermal spikes) have also been simulated by MD to simulate the thermal excitation induced by displacement cascades. Equipartition of energy was shown to occur in displacement cascades, half of the kinetic energy of the primary knock-on atom being converted after a few tenths of picoseconds into potential energy. The kinetic and potential parts of the system energy are however subjected to little variations during dedicated thermal spike simulations. This is probably due to the velocity rescaling process, which impacts a large number of atoms in this case and would drive the system away from a dynamical equilibrium. This result makes questionable MD simulations of thermal spikes carried out up to now (early 2014). The thermal history of cascades was compared to the heat equation solution of a punctual thermal excitation in UO{sub 2}. The maximum volume brought to a temperature above the melting temperature during the simulated cascade events is well reproduced by this simple model. This volume eventually constitutes a relevant estimate of the volume affected by a displacement cascade in UO{sub 2}. This definition of the cascade volume could also make sense in other materials, like iron.

  20. Studies on the sintering behaviour of uranium dioxide powder compacts

    International Nuclear Information System (INIS)

    Das, P.; Chowdhury, R.

    1988-01-01

    Uranium dioxide fuel pellets are normally made from their precursor ammonium diuranate, followed by calcination, subsequent reduction to sinterable grade powders and a post operation treatment of pressing and sintering. The low temperature calcined powders, usually exhibiting non-crystalline behaviour (under X-ray diffraction studies) progressively transforms into a crystalline variety on subsequent heat treatment at higher temperature. It is observed however that powders calcined between 800 to 900 0 C exhibit enhanced densification behaviour when sintered at higher temperatures. The isothermal shrinkage versus time plot of the sintered compacts are well described by a hyperbolic relationship which takes care of the observed shrinkage (λ) as caused due to a cumulative effect from the initial sintering of the powder compacts at zero time (α) and that caused due to the structural transformation from a non-crystalline modification with increased thermal treatment (β). The derived equation is a modification of the sintering mechanism of the viscous flow type proposed by Frenkel, involving sintering of an amorphous phase, the viscosity of the latter is presumed to increase with increasing thermal treatment to assume the final modified form as λ=t/(α+βt), where t = time, λ = shrinkage and α and β are the unknown parameters. (orig.)

  1. Study of the changes of uranium dioxide properties resulting from sintering; Izucavanje procesa sinterovanja urandioksida sa gledista promene karakteristicnih osobina

    Energy Technology Data Exchange (ETDEWEB)

    Ristic, M M [Institute of Nuclear Sciences Vinca, Laboratorija za reaktorske materijale, Beograd (Serbia and Montenegro)

    1962-12-15

    Uranium dioxide powder used for studying the sintering process having grain size 63 {mu}. Sintering was performed in the temperature interval from 1000 - 1300 deg C in argon atmosphere. The O/U ratio of the used uranium dioxide was 2.07. Densities obtained by sintering under the mentioned conditions were not higher than 91% TG (theoretical density). This showed that the mentioned conditions were optimal, but the uranium dioxide obtained could be used for studying the radiation damage of fuel.

  2. Methods for oxygen/uranium ratio determination in substoichiometric uranium dioxide

    International Nuclear Information System (INIS)

    Baranov, V.G.; Godin, Yu.G.; S'edin, Yu.D.; Kosykh, V.G.; Nepryakhin, A.M.; Komarenko, F.F.; Kutyreva, G.A.

    1994-01-01

    Investigations are performed into a possibility to use the methods of thermal gravimetric analysis, gas chromatography, hydration-dehydration, and e.m.f. of high-temperature solid-electrode galvanic cell for determining O-U atomic ratio in UO 2-x . It is shown that the investigated methods have an analysis error of ± 0.001 O/U units. However, the e.m.f. method, which feature a high accuracy near stoichiometry can be applied only within the limits of UO 2-x homogeneity. A possibility is shown to expend the area of e.m.f. method application during the analysis of substoichiometric uranium dioxide. 9 refs.; 1 tab

  3. Preparation, sintering and leaching of optimized uranium thorium dioxides

    International Nuclear Information System (INIS)

    Hingant, N.; Clavier, N.; Dacheux, N.; Barre, N.; Hubert, S.; Obbade, S.; Taborda, F.; Abraham, F.

    2009-01-01

    Mixed actinide dioxides are currently studied as potential fuels for several concepts associated to the fourth generation of nuclear reactors. These solids are generally obtained through dry chemistry processes from powder mixtures but could present some heterogeneity in the distribution of the cations in the solid. In this context, wet chemistry methods were set up for the preparation of U 1-x Th x O 2 solid solutions as model compounds for advanced dioxide fuels. Two chemical routes of preparation, involving the precipitation of crystallized precursor, were investigated: on the one hand, a mixture of acidic solutions containing cations and oxalic acid was introduced in an open vessel, leading to a poorly-crystallized precipitate. On the other hand, the starting mixture was placed in an acid digestion bomb then set in an oven in order to reach hydrothermal conditions. By this way, small single-crystals were obtained then characterized by several techniques including XRD and SEM. The great differences in terms of morphology and crystallization state of the samples were correlated to an important variation of the specific surface area of the oxides prepared after heating, then the microstructure of the sintered pellets prepared at high temperature. Preliminary leaching tests were finally undertaken in dynamic conditions (i.e. with high renewal of the leachate) in order to evaluate the influence of the sample morphology on the chemical durability of the final cohesive materials

  4. Sorption behaviour of uranium and thorium on cryptomelane-type hydrous manganese dioxide from aqueous solution

    International Nuclear Information System (INIS)

    El-Naggar, I.M.; El-Absy, M.A.; Abdel-Hamid, M.M.; Aly, H.F.

    1993-01-01

    The kinetics of sorption of uranium and thorium from aqueous nitrate solutions on cryptomelane-type hydrous manganese dioxide (CRYMO) was studied. The exchange of uranium is particle diffusion controlled while that of thorium is chemical reaction at the exchange sites. Sorption of uranium and thorium by CRYMO has been also studied as a function of metal concentrations and temperature. The sorption of both cations is found to be an endothermic process and increases markedly with temperature between 30 and 60 degree C. The sorption results have been analysed by the langmuir adsorption isotherm over the entire range of uranium and thorium concentrations investigated. 35 refs., 8 figs., 5 tabs

  5. Improvement of cesium retention in uranium dioxide by additional phases

    International Nuclear Information System (INIS)

    Gamaury Dubois, S.

    1995-01-01

    The objective of this study is to improve the cesium retention in nuclear fuel. A bibliographic survey indicates that cesium is rapidly released from uranium dioxide in an accident condition. At temperatures higher than 1500 deg C or in oxidising conditions, our experiments show the difficulty of maintaining cesium inside simulated fuel. Two ternary systems are potentially interesting for the retention of cesium and to reduce the kinetics of release from the fuel: Cs 2 O-Al 2 O 3 -SiO 2 et Cs 2 O-ZrO 2 -SO 2 . The compounds CsAISi 2 O 6 and Cs 2 ZrSi 6 O 15 were studied from 1200 deg C to 2000 deg C by thermogravimetric analysis. The volumetric diffusion coefficients of cesium in these structures, in solid state as well as in liquid one, were measured. A fuel was sintered with (Al 2 O 3 + SiO 2 ) or (ZrO 2 + SiO 2 ) and the intergranular phase was characterized. In the presence of (Al 2 O 3 + SiO 2 ), the sintering is realized at 1610 deg C in H 2 . It is a liquid phase sintering. On the other end, with (ZrO 2 + SiO 2 ), the sintering is a low temperature one in oxidising atmosphere. Finally, cesium containing simulated fuels were produced with these additives. According to the effective diffusion coefficients that were measured, the additives improved the retention of cesium. We have predicted the improvement that could be hoped for in a nuclear reactor, depending on the dispersion of the intergranular additives, the temperature and the degree of oxidation of the UO 2+x . We wait for a factor of 2 for x=0 and more than 8 for x=0.05, up to 2000 deg C. (author). 148 refs., 122 figs., 34 tabs

  6. Micromechanical simulation of Uranium dioxide polycrystalline aggregate behaviour under irradiation

    International Nuclear Information System (INIS)

    Pacull, J.

    2011-02-01

    In pressurized water nuclear power reactor (PWR), the fuel rod is made of dioxide of uranium (UO 2 ) pellet stacked in a metallic cladding. A multi scale and multi-physic approaches are needed for the simulation of fuel behavior under irradiation. The main phenomena to take into account are thermomechanical behavior of the fuel rod and chemical-physic behavior of the fission products. These last years one of the scientific issue to improve the simulation is to take into account the multi-physic coupling problem at the microscopic scale. The objective of this ph-D study is to contribute to this multi-scale approach. The present work concerns the micro-mechanical behavior of a polycrystalline aggregate of UO 2 . Mean field and full field approaches are considered. For the former and the later a self consistent homogenization technique and a periodic Finite Element model base on the 3D Voronoi pattern are respectively used. Fuel visco-plasticity is introduced in the model at the scale of a single grain by taking into account specific dislocation slip systems of UO 2 . A cohesive zone model has also been developed and implemented to simulate grain boundary sliding and intergranular crack opening. The effective homogenous behaviour of a Representative Volume Element (RVE) is fitted with experimental data coming from mechanical tests on a single pellet. Local behavior is also analyzed in order to evaluate the model capacity to assess micro-mechanical state. In particular, intra and inter granular stress gradient are discussed. A first validation of the local behavior assessment is proposed through the simulation of intergranular crack opening measured in a compressive creep test of a single fuel pellet. Concerning the impact of the microstructure on the fuel behavior under irradiation, a RVE simulation with a representative transient loading of a fuel rod during a power ramp test is achieved. The impact of local stress and strain heterogeneities on the multi

  7. Study of uranium dioxide pellets by micro-acoustic techniques

    International Nuclear Information System (INIS)

    Roque, V.

    1999-01-01

    In order to reduce the volume of spent fuel to reprocess and to improve the productivity and the safety of the nuclear reactor, 'Electricite De France' aim to increase the average fuel discharge burn-up. To elaborate the safety reports, EDF develops codes to simulate the thermo-mechanical behaviour of the nuclear fuel element. These numeric simulations need to evaluate accurately and locally the evolution of the material and of its properties. One of the major concern today is the local characterisation of the intrinsic volume fraction porosity and the mechanical properties of the irradiated fuel. The fuel pellet fragmentation, the steep radial gradient in its physical properties evolution and the chemical evolution of the irradiated material make difficult nay the use of the conventional techniques. This leads EDF to pay interest for the use of two complementary techniques: micro-indentation on the one hand and acoustic methods on the other hand (acoustic microscopy and micro-echography), with an additional constrain to perform on active materials. The objective of this work has been to adapt the acoustic methods for an application on uranium dioxide pellets, used as nuclear fuel in Water Pressurised Reactor. Acquisitions protocols have been set to measure accurately the Rayleigh velocity and the longitudinal velocity of the UO 2 . Using these protocols, we have calibrated these acoustic methods by analysing non irradiated nuclear pellet which properties were well known. This process enable to quantify the effects of different physico-chemical parameters of the UO 2 on the ultrasonic velocities measured. Particularly, the large influence of the porosity has been demonstrated and empirical laws to express the evolution of the acoustic velocities as a function of the volume fraction porosity were established. Moreover, we have established a methodology to characterise the intrinsic elastic constants and the volume fraction porosity on irradiated UO 2 fuel pellets

  8. Theoretical study using electronic structure calculations of uranium and cerium dioxides containing defects and impurities

    International Nuclear Information System (INIS)

    Shi, Lei

    2016-01-01

    Uranium dioxide (UO_2) is the most widely used nuclear fuel in existing nuclear reactors around the world. While in service for energy supply, UO_2 is submitted to the neutron flux and undergoes nuclear fission chain reactions, which create large number of fission products and point defects. The study of the behavior of the fission products and point defects is important to understand the fuel properties under irradiation. We conduct electronic structure calculations based on the density functional theory (DFT) to model this radiation damage at the atomic scale. The DFT+U method is used to describe the strong correlation of the 4f electrons of cerium and 5f electrons of uranium in the materials studied (UO_2, CeO_2 and (U, Ce)O_2). (U, Ce)O_2 is studied because it is considered as a low radioactive model material of mixed actinide oxides such as the MOX fuel (U, Pu)O_2 used in light water reactors and fast neutron reactors. Cerium dioxide (CeO_2) is studied to provide reference data of (U, Ce)O_2. We perform a DFT+U study of point defects and gaseous fission products (Xe and Kr) in CeO_2 and compare our results to the existing ones of UO_2. We study the bulk properties as well as the behavior of defects for (U, Ce)O_2, and compare our results to the ones of (U, Pu)O_2. Furthermore, for the study of defects in UO_2, methodological improvements are explored considering the spin-orbit coupling effect and the finite-size effect of the simulation supercell. (author) [fr

  9. Dissolution testing of intermediary products in uranium dioxide production by the sol-gel method

    International Nuclear Information System (INIS)

    Melichar, F.; Landspersky, H.; Urbanek, V.

    1979-01-01

    A method was developed of dissolving polyuranates and uranium dioxides in sulphuric acid and in carbonate solutions for testing intermediate products in the sol-gel process preparation of uranium dioxide. A detailed granulometric analysis of spherical particle dispersion was included as part of the tests. Two different production methods were used for the two types of studied materials. The test results show that the test method is suitable for determining temperature sensitivity of the materials to dissolution reaction. The geometrical distribution of impurities in the spherical particles can be determined from the dissolution kinetics. The method allows the determination of the effect of carbon from impurities on the process of uranium dioxide leaching and is thus applicable for testing materials prepared by the sol-gel method. (Z.M.)

  10. Models for the adsorption of uranium on titanium dioxide

    International Nuclear Information System (INIS)

    Jaffrezic-Renault, N.; Poirier-Andrade, H.; Trang, D.H.

    1980-01-01

    A hydrated titanium oxide whose acid-base properties are well defined has been used to study the retention mechanism of uranium as UO 2 2+ (in acidic media) and as UO 2 (CO 3 ) 3 4- (in carbonate media). The influence of various parameters on the distribution coefficient of uranium (pH, [CO 3 2- ]) and of the adsorption of uranium on the electrophoretic mobilities of the titanium oxide have been investigated. It is shown that, in both media, coordinative TiO-UO 2 bonds are formed. These strong bonds explain the high affinity of the titanium oxide for uranium. (orig.)

  11. Characterisation of electrodeposited polycrystalline uranium dioxide thin films on nickel foil for industrial applications

    International Nuclear Information System (INIS)

    Adamska, A.M.; Bright, E. Lawrence; Sutcliffe, J.; Liu, W.; Payton, O.D.; Picco, L.; Scott, T.B.

    2015-01-01

    Polycrystalline uranium dioxide thin films were grown on nickel substrates via aqueous electrodeposition of a precursor uranyl salt. The arising semiconducting uranium dioxide thin films exhibited a tower-like morphology, which may be suitable for future application in 3D solar cell applications. The thickness of the homogenous, tower-like films reached 350 nm. Longer deposition times led to the formation of thicker (up to 1.5 μm) and highly porous films. - Highlights: • Electrodeposition of polycrystalline UO_2 thin films • Tower-like morphology for 3D solar cell applications • Novel technique for separation of heavy elements from radioactive waste streams

  12. Fuel Retention Improvement at High Temperatures in Tungsten-Uranium Dioxide Dispersion Fuel Elements by Plasma-Spray Cladding

    Science.gov (United States)

    Grisaffe, Salvatore J.; Caves, Robert M.

    1964-01-01

    An investigation was undertaken to determine the feasibility of depositing integrally bonded plasma-sprayed tungsten coatings onto 80-volume-percent tungsten - 20-volume-percent uranium dioxide composites. These composites were face clad with thin tungsten foil to inhibit uranium dioxide loss at elevated temperatures, but loss at the unclad edges was still significant. By preheating the composite substrates to approximately 3700 degrees F in a nitrogen environment, metallurgically bonded tungsten coatings could be obtained directly by plasma spraying. Furthermore, even though these coatings were thin and somewhat porous, they greatly inhibited the loss of uranium dioxide. For example, a specimen that was face clad but had no edge cladding lost 5.8 percent uranium dioxide after 2 hours at 4750 dgrees F in flowing hydrogen. A similar specimen with plasma-spray-coated edges, however, lost only 0.75 percent uranium dioxide under the same testing conditions.

  13. Selective oxidation of propene on bismuth molybdate and mixed oxides of tin and antimony and of uranium and antimony

    International Nuclear Information System (INIS)

    Pendleton, P.; Taylor, D.

    1976-01-01

    Propene + 18 0 2 reactions have been studied in a static reaction system on bismuth molybdate and mixed oxides of tin and antimony and of uranium and antimony. The [ 16 0] acrolein content of the total acrolein formed and the proportion of 16 0 in the oxygen of the carbon dioxide by-product have been determined. The results indicate that for each catalyst the lattice is the only direct source of the oxygen in the aldehyde, and that lattice and/or gas phase oxygen is used in carbon dioxide formation. Oxygen anion mobility appears to be greater in the molybdate catalyst than in the other two. (author)

  14. High temperature behavior of metallic inclusions in uranium dioxide

    International Nuclear Information System (INIS)

    Yang, R.L.

    1980-08-01

    The object of this thesis was to construct a temperature gradient furnace to simulate the thermal conditions in the reactor fuel and to study the migration of metallic inclusions in uranium oxide under the influence of temperature gradient. No thermal migration of molybdenum and tungsten inclusions was observed under the experimental conditions. Ruthenium inclusions, however, dissolved and diffused atomically through grain boundaries in slightly reduced uranium oxide. An intermetallic compound (probably URu 3 ) was formed by reaction of Ru and UO/sub 2-x/. The diffusivity and solubility of ruthenium in uranium oxide were measured

  15. Calculation of the energy of stacking faults in uranium dioxide

    International Nuclear Information System (INIS)

    Lefebvre, J.-M.; Soullard, J.

    1976-01-01

    Energy computations of some (100), (110) and (111), planar defects were performed using an ionic bond model for stoichiometric uranium dioxyde. The repulsive contribution to the fault was estimated in two different ways, i.e. using the Born-Mayer classical treatment, or potentials derived from shell model calculations. The stability of the various defect configurations has been studied; on the basis of the numerical values, it may be concluded that dislocation dissociation is unlikely in stoichiometric uranium dioxyde. (Auth.)

  16. Method and device for the dry preparation of ceramic uranium dioxide nuclear fuel wastes

    International Nuclear Information System (INIS)

    Pirk, H.; Roepenack, H.; Goeldner, U.

    1977-01-01

    Reprocessing of waste, resulting from the production of ceramic sintered bodies from uranium dioxide for use as nuclear fuel, in a dry process into very finely dispersed pure U 3 O 8 powder may be improved by applying vibrating screening during oxidation. An appropriate device is described. (UWI) [de

  17. Qualitative relations between the kinetics of sintering in hydrogen and the observed microstructures of uranium dioxide

    International Nuclear Information System (INIS)

    Francois, B.; Delmas, R.; Caillat, F.; Lacombe, P.

    1975-01-01

    The microscopic study of uranium dioxide sintered in hydrogen, together with density measurements, shows on the one hand that the large scale appearance of pores trapped at the grain boundaries in the course of sintering has the effect of practically stopping densification, and on the other hand that this particular microstructure is stable over a wide range of time and temperature. (author)

  18. Nuclear energy - Determination of chlorine and fluorine in uranium dioxide powder and sintered pellets

    International Nuclear Information System (INIS)

    2008-01-01

    This International Standard describes a method for determining the chlorine and fluorine concentrations in uranium dioxide and in sintered fuel pellets by pyrohydrolysis of samples, followed either by liquid ion-exchange chromatography or by selective electrode measurement of chlorine and fluorine ions. Many ion-exchange chromatography systems and ion-selective electrode measurement systems are available

  19. Nuclear energy - Uranium dioxide powder and sintered pellets - Determination of oxygen/uranium atomic ratio by the amperometric method. 2. ed.

    International Nuclear Information System (INIS)

    2007-01-01

    This International Standard specifies an analytical method for the determination of the oxygen/uranium atomic ratio in uranium dioxide powder and sintered pellets. The method is applicable to reactor grade samples of hyper-stoichiometric uranium dioxide powder and pellets. The presence of reducing agents or residual organic additives invalidates the procedure. The test sample is dissolved in orthophosphoric acid, which does not oxidize the uranium(IV) from UO 2 molecules. Thus, the uranium(VI) that is present in the dissolved solution is from UO 3 and/or U 3 O 8 molecules only, and is proportional to the excess oxygen in these molecules. The uranium(VI) content of the solution is determined by titration with a previously standardized solution of ammonium iron(II) sulfate hexahydrate in orthophosphoric acid. The end-point of the titration is determined amperometrically using a pair of polarized platinum electrodes. The oxygen/uranium ratio is calculated from the uranium(VI) content. A portion, weighing about 1 g, of the test sample is dissolved in orthophosphoric acid. The dissolution is performed in an atmosphere of nitrogen or carbon dioxide when sintered material is being analysed. When highly sintered material is being analysed, the dissolution is performed at a higher temperature in purified phosphoric acid from which the water has been partly removed. The cooled solution is titrated with an orthophosphoric acid solution of ammonium iron(II) sulfate, which has previously been standardized against potassium dichromate. The end-point of the titration is detected by the sudden increase of current between a pair of polarized platinum electrodes on the addition of an excess of ammonium iron(II) sulfate solution. The paper provides information about scope, principle, reactions, reagents, apparatus, preparation of test sample, procedure (uranium dioxide powder, sintered pellets of uranium dioxide, highly sintered pellets of uranium dioxide and determination

  20. Mixing of Al into uranium silicides reactor fuels

    International Nuclear Information System (INIS)

    Ding, F.R.; Birtcher, R.C.; Kestel, B.J.; Baldo, P.M.

    1996-11-01

    SEM observations have shown that irradiation induced interaction of the aluminum cladding with uranium silicide reactor fuels strongly affects both fission gas and fuel swelling behaviors during fuel burn-up. The authors have used ion beam mixing, by 1.5 MeV Kr, to study this phenomena. RBS and the 27 Al(p, γ) 28 Si resonance nuclear reaction were used to measure radiation induced mixing of Al into U 3 Si and U 3 Si 2 after irradiation at 300 C. Initially U mixes into the Al layer and Al mixes into the U 3 Si. At a low dose, the Al layer is converted into UAl 4 type compound while near the interface the phase U(Al .93 Si .07 ) 3 grows. Under irradiation, Al diffuses out of the UAl 4 surface layer, and the lower density ternary, which is stable under irradiation, is the final product. Al mixing into U 3 Si 2 is slower than in U 3 Si, but after high dose irradiation the Al concentration extends much farther into the bulk. In both systems Al mixing and diffusion is controlled by phase formation and growth. The Al mixing rates into the two alloys are similar to that of Al into pure uranium where similar aluminide phases are formed

  1. Sintering uranium oxide in the reaction product of hydrogen-carbon dioxide mixtures

    International Nuclear Information System (INIS)

    De Hollander, W.R.; Nivas, Y.

    1975-01-01

    Compacted pellets of uranium oxide alone or containing one or more additives such as plutonium dioxide, gadolinium oxide, titanium dioxide, silica, and alumina are heated to 900 to 1599 0 C in the presence of a mixture of hydrogen and carbon dioxide, either alone or with an inert carrier gas and held at the desired temperature in this atmosphere to sinter the pellets. The sintered pellets are then cooled in an atmosphere having an oxygen partial pressure of 10 -4 to 10 -18 atm of oxygen such as dry hydrogen, wet hydrogen, dry carbon monoxide, wet carbon monoxide, inert gases such as nitrogen, argon, helium, and neon and mixtures of ayny of the foregoing including a mixture of hydrogen and carbon dioxide. The ratio of hydrogen to carbon dioxide in the gas mixture fed to the furnace is controlled to give a ratio of oxygen to uranium atoms in the sintered particles within the range of 1.98:1 to about 2.10:1. The water vapor present in the reaction products in the furnace atmosphere acts as a hydrolysis agent to aid removal of fluoride should such impurity be present in the uranium oxide. (U.S.)

  2. Fracture toughness of WWER Uranium dioxide fuel pellets with various grain size

    International Nuclear Information System (INIS)

    Sivov, R.; Novikov, V.; Mikheev, E.; Fedotov, A.

    2015-01-01

    Uranium dioxide fuel pellets with grain sizes 13, 26, and 33 μm for WWER were investigated in the present work in order to determine crack formation and the fracture toughness.The investigation of crack formation in uranium oxide fuel pellets of the WWER-types showed that Young’s modulus and the microhardness of polycrystalline samples increase with increasing grain size, while the fracture toughness decreases. Characteristically, radial Palmqvist cracks form on the surface of uranium dioxide pellets for loads up to 1 kg. Transgranular propagation of cracks over distances several-fold larger than the length of the imprint diagonal is observed in pellets with large grains and small intragrain pores. Intergranular propagation of cracks along grain boundaries with branching occurs in pellets with small grains and low pore concentration on the grain boundaries. Blunting on large pores and at breaks in direction does not permit the cracks to reach a significant length

  3. Characterization of transport properties in uranium dioxide: the case of the oxygen auto-diffusion

    International Nuclear Information System (INIS)

    Fraczkiewicz, M.; Baldinozzi, G.

    2008-01-01

    Point defects in uranium dioxide which control the transport phenomena are still badly known. The aim of this work is to show how in carrying out several experimental techniques, it is possible to demonstrate both the existence and to determine the nature (charge and localization) of predominant defects responsible of the transport phenomena in a fluorite-type structure oxide. The oxygen diffusion in the uranium dioxide illustrates this. In the first part of this work, the accent is put on the electric properties of uranium dioxide and more particularly on the variation laws of the electric conductivity in terms of temperature, of oxygen potential and of the impurities amounts present in the material. These evolutions are connected to point and charged complex defects models and the pertinence of these models is discussed. Besides, it is shown how the electric conductivity measurements can allow to define oxygen potential domains in which the concentrations in electronic carriers are controlled. This characterization being made, it is shown that the determination of the oxygen intrinsic diffusion coefficient and particularly its dependence to the oxygen potential and to the amount of impurity, allows to determine the main defect responsible to the atomic diffusion as well as its nature and its charge. In the second part, the experimental techniques to determine the oxygen diffusion coefficient are presented: there are the isotopic exchange technique for introducing the tracer in the material, and two techniques to characterize the diffusion profiles (SIMS and NRA). Examples of preliminary results are given for mono and polycrystalline samples. At last, from this methodology on uranium dioxide, studies considered to quantify the thermal and physicochemical effects are presented. Experiments considered with the aim to characterize the radiation diffusion in uranium dioxide are presented too. (O.M.)

  4. Laboratory sol-gel preparation of fine fraction of sintered uranium dioxide spheres

    International Nuclear Information System (INIS)

    Landspersky, H.; Tympl, M.

    1984-01-01

    The results are summed up of the laboratory investigation of preparing the fine fraction of sintered uranium dioxide particles from uranyl gel using the method of the mixed reactor and the method of the dual-liquid nozzle, processed by leaching, drying, calcination and sintering. None of the two methods provides monodispersion particles under the given conditions but better control of the throughflow of the liquid media may improve results. Leaching of the fine fraction is very quick and the leaching of most components takes no longer than 5 minutes. In view of the fact that leaching of all components does not proceed at the same rate it is recommended that leaching time be doubled, or that leaching take place in two stages. Azeotropic distillation with chlorinated hydrocarbons is a favourable procedure for obtaining quality material; it is, however, necessary to prevent dried particles from comino. into contact with the water phase condensing on the walls of the distillation vessel and running down onto the surface of the distilling mixture. Calcination at a temperature of 500 degC in a thin layer and sintering at temperatures between 1350 and 1550 degC at an adequate rate of inflow of gaseous media and adequate rate of outflow of reaction wastes results in the production of high quality material whose density exceeds 97 to 98% theoretical density. (author)

  5. Simulation of the interaction between uranium dioxide and zircaloy

    International Nuclear Information System (INIS)

    Denis, A.; Garcia, E.A.

    1984-01-01

    The code solves the oxygen diffusion equations of the five phases formed during the UO 2 /Zircaloy interaction, using an implicit finite difference method with parabolic interpolation at the interfaces. Uranium and Zirconium mass conservation are considered. The code gives a good simulation of the experimental results for isothermal conditions. (orig.)

  6. Investigating the structural changes of uranium dioxide dependent on additives, Phase I - Uranium-oxide system from structural-phase aspect

    International Nuclear Information System (INIS)

    Manojlovic, Lj.

    1962-12-01

    Having in mind the complex structure of the system uranium-oxygen, and that experimental studies of this system lead to controversial conclusions, an extensive review and analysis of the papers published on this subject were needed. This review wold be very useful for interpreting the expected structural changes of the uranium dioxide dependent on the additives

  7. X-ray photoelectron and Auger electron spectroscopic study of the adsorption of molecular iodine on uranium metal and uranium dioxide

    International Nuclear Information System (INIS)

    Dillard, J.G.; Moers, H.; Klewe-Nebenius, H.; Kirch, G.; Pfennig, G.; Ache, H.J.

    1984-01-01

    The adsorption of molecular iodine on uranium metal and on uranium dioxide has been investigated at 25 0 C. Clean surfaces were prepared in an ultrahigh vacuum apparatus and were characterized by X-ray photoelectron (XPS) and X-ray and electron-induced Auger electron spectroscopies (AES). Adsorption of I 2 was studied for exposures up to 100 langmuirs (1 langmuir = 10 -6 torr s) on uranium metal and to 75 langmuirs on uranium dioxide. Above about 2-langmuir I 2 exposure on uranium, spectroscopic evidence is obtained to indicate the beginning of UI 3 formation. Saturation coverage for I 2 adsorption on uranium dioxide occurs at approximately 10-15 langmuirs. Analysis of the XPS and AES results as well as studies of spectra as a function of temperature lead to the conclusions that a dissociative chemisorption/reaction process occurs on uranium metal while nondissociative adsorption occurs on uranium dioxide. Variations in the iodine Auger kinetic energy and in the Auger parameter are interpreted in light of extra-atomic relaxation processes. 42 references, 10 figures, 1 table

  8. Uranium removal during low discharge in the Ganges-Brahmaputra mixing zone

    International Nuclear Information System (INIS)

    Carroll, J.; Moore, W.S.

    1993-01-01

    The Ganges-Brahmaputra river system supplies more dissolved uranium to the ocean than any other system in the world (Sarin et al., 1990; Sackett et al., 1973). However, there have been no investigations to determine whether riverine supplies of uranium are altered by geochemical reactions in the river-ocean mixing zone. In this study, uranium and salinity data were collected in the Ganges-Brahmaputra mixing zone during a period of low river discharge. The uranium distribution with salinity shows that in waters <12 ppt salinity, uranium activities are significantly lower than predicted from conservative mixing of river and seawater. This suggests that uranium is being removed within the mixing zone. The behavior of uranium in the Ganges-Brahmaputra is in sharp contrast to its behavior in the Amazon mixing zone where McKee et al. (1978) found uranium activities significantly higher than predicted from conservative mixing. The contrasting behaviors for uranium in these systems are due to the different locations where mixing between river and seawater occurs. For the Amazon, mixing takes place on the continental shelf whereas for the Ganges-Brahmaputra, mixing occurs within shoreline sedimentary environments. The physiochemical processes controlling uranium removal to sediment deposits in the Amazon are partly known. The authors discuss mechanisms which may be removing uranium to suspended and mangrove sediments in the Ganges-Brahmaputra

  9. Development of a reduction process of ammonium uranyl carbonate to uranium dioxide in a fluidized bed

    International Nuclear Information System (INIS)

    Gomes, R.P.; Riella, H.G.

    1990-07-01

    Laboratory development of ammonium uranyl carbonate (AUC) reduction to uranium dioxide (UO 2 ) using fluidized bed furnace technique is described. The reaction is carried out at 500-550 0 C using hydrogen, liberated from cracking of ammonia, as a reducing agent. As the AUC used is obtained from uranium hexafluoride (UF 6 ) it contains considerable amount of fluoride (approx. 500μg/g) as contaminant. The presence of fluoride leads to high corrosion rates and hence the fluoride concentration is reduced by pyrohydrolisis of UO 2 . Physical and Chemical properties of the final product (UO 2 ) obtained were characterized. (author) [pt

  10. Experience with a uranyl nitrate/uranium dioxide conversion pilot plant

    International Nuclear Information System (INIS)

    Arcuri, L.; Pietrelli, L.

    1984-01-01

    A plant for the precipitation of sinterable nuclear grade UO 2 powders is described in this report. The plant has been designed, built and set up by SNIA TECHINT. ENEA has been involved in the job as nuclear consultant. Main process steps are: dissolution of UO 2 powder or sintered UO 2 pellets, adjustment of uranyl nitrate solutions, precipitation of uranium peroxide by means of hydrogen peroxide, centrifugation of the precipitate, drying, calcination and reduction to uranium dioxide. The report is divided in two main section: the process description and the ''hot test'' report. Some laboratory data on precipitation of ammonium diuranate by means of NH 4 OH, are also reported

  11. Development of ammonium uranyl carbonate reduction to uranium dioxide using fluidized bed

    International Nuclear Information System (INIS)

    Gomes, R.P.; Riella, H.G.

    1988-01-01

    Laboratory development of Ammonium Uranyl Carbonate (AUC) reduction to uranium dioxide (UO 2 ) using fluidized bed furnace technique is described. The reaction is carried out at 500-550 0 C using hydrogen, liberated from cracking of ammonia, as a reducing agent. As the AUC used is obtained from uranium hexafluoride (UF 6 ) it contains considerable amounts of fluoride ( - 500μgF - /gTCAU) as contaminant. The presence of fluoride leads to high corrosion rates and hence the fluoride concentrations is reduced by pyrohydrolisis of UO 2 . Physical and Chemical proterties of the final product (UO 2 ) obtained were characterized. (author) [pt

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2000-01-01

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

  13. Following the electroreduction of uranium dioxide to uranium in LiCl–KCl eutectic in situ using synchrotron radiation

    Energy Technology Data Exchange (ETDEWEB)

    Brown, L.D.; Abdulaziz, R.; Jervis, R.; Bharath, V.J. [Electrochemical Innovation Lab, Dept. Chemical Engineering, UCL, London WC1E 7JE (United Kingdom); Atwood, R.C.; Reinhard, C.; Connor, L.D. [Diamond Light Source, Harwell Science and Innovation Campus, Didcot, Oxfordshire OX11 0DE (United Kingdom); Simons, S.J.R.; Inman, D.; Brett, D.J.L. [Electrochemical Innovation Lab, Dept. Chemical Engineering, UCL, London WC1E 7JE (United Kingdom); Shearing, P.R., E-mail: p.shearing@ucl.ac.uk [Electrochemical Innovation Lab, Dept. Chemical Engineering, UCL, London WC1E 7JE (United Kingdom)

    2015-09-15

    Highlights: • We investigated the electroreduction of UO{sub 2} to U in LiCl/KCL eutectic molten salt. • Combined electrochemical measurement and in situ XRD is utilised. • The electroreduction appears to occur in a single, 4-electron-step, process. • No intermediate compounds were observed. - Abstract: The electrochemical reduction of uranium dioxide to metallic uranium has been investigated in lithium chloride–potassium chloride eutectic molten salt. Laboratory based electrochemical studies have been coupled with in situ energy dispersive X-ray diffraction, for the first time, to deduce the reduction pathway. No intermediate phases were identified using the X-ray diffraction before, during or after electroreduction to form α-uranium. This suggests that the electrochemical reduction occurs via a single, 4-electron-step, process. The rate of formation of α-uranium is seen to decrease during electrolysis and could be a result of a build-up of oxygen anions in the molten salt. Slow transport of O{sup 2−} ions away from the UO{sub 2} working electrode could impede the electrochemical reduction.

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

    International Nuclear Information System (INIS)

    Gonçalves, Letícia C.; Maiorino, José R.

    2017-01-01

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

  15. Stability with temperature of mixed uranium plutonium monocarbides

    International Nuclear Information System (INIS)

    Riglet-Martial, Ch.; Dumas, J.C.; Piron, J.P.; Gueneau, Ch.

    2008-01-01

    Full text: Among the different advanced fuel materials of concern for Generation IV systems, the mixed carbide of uranium and plutonium fuel is considered as one of the key materials for Gas Fast Reactors (GFR) systems. For purposes of optimising its fabrication process as well as its performances in various operating conditions, the losses of gaseous plutonium specially at elevated temperatures have to be controlled and minimized. The paper is therefore concerned with a parametric analysis of the stability with temperature of mixed carbides of uranium and plutonium. Previous published experimental studies have shown that mixed (U ,Pu) carbides undergo a highly incongruent sublimation at high temperatures: the vapour phase in equilibrium with the solid is mainly composed of gaseous plutonium (P Pu /P total > 99 % ) while the contribution of gaseous U and C remains very low. The composition of the system U 1-z Pu z C 1+x ' (z =Pu/(U+Pu) and x C/(U+Pu)), the temperature (T) and the expansion volume (V) of the gas are the main parameters in the loss of gaseous Pu. The calculations are carried out using the SAGE (Solgasmix Advanced Gibbs Energy) software, by assuming ideal solid solutions between UC and PuC, as well as between U 2 C 3 and Pu 2 C 3 . The validity of the model is previously tested using published equilibrium vapour pressure data. This work gives rise to a large description of the variations of Pu losses from mixed uranium plutonium carbides and leads to some basic recommendations in connection with the use of this advanced fuel materials

  16. Certification of a uranium-238 dioxide reference material for neutron dosimetry (EC nuclear reference material 501)

    International Nuclear Information System (INIS)

    Pauwels, J.; Lievens, F.; Ingelbrecht, C.

    1989-01-01

    Uranium-238 oxide of 99.999% isotopic and 99.98% chemical purity was transformed into dioxide spheres of nominal 0.5 and 1.0 mm diameter by gel precipitation and subsequent calcination under carbon dioxide and under argon containing 5% hydrogen at 1 125 K. The spheres were analysed by thermal ionization mass spectrometry, including isotope dilution, by gravimetry and by potentiometric titration. On the basis of these analyses, the uranium mass fraction was certified at 879.4 ± 2.8 g.kg -1 , and the 235 U/U - and 238 U/U abundances at 10.4 ± 0.5 mg.kg -1 and 999.9896 ± 0.0005 g.kg -1 , respectively. The material is intended to be used as a reference material in neutron metrology

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1999-01-01

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

  18. Contribution to the study of the microstructure of uranium dioxide (1962)

    International Nuclear Information System (INIS)

    Porneuf, A.

    1960-05-01

    The microstructure of sintered uranium dioxide is studied in relation with several parameters, specially the sintering temperatures and atmospheres. The external surface and the internal microstructure of the sintered are examined, using fractography and ceramography. Various techniques for preparing surfaces (mechanical and electrolytic polishing) and for revealing the structure (chemical and anodic attack, ionic bombardment oxidation) have been experienced and compared. Patterns similar to those revealed in metals and probably related with interactions between dislocations and vacancies have been observed. (author) [fr

  19. Safety analysis report of uranium dioxide fuel laboratory, Nuclear Research Centre Inchas, Egypt

    International Nuclear Information System (INIS)

    Abdel-Azim, M.S.; Abdel-Halim, A.

    1987-07-01

    In the Nuclear Research Center Inchas a uranium dioxide fuel laboratory is planned and built by the AEA Cairo (Atomic Energy Authority). The layout of this fuel lab and the programmatical contents are subject to the bilaterial cooperation between Egypt and the Federal Republic of Germany. In this report the safety analysis as basic items for the approval procedure are started in detail. (orig.) [de

  20. An oxyde mixture fuel containing uranium and plutonium dioxides and process to obtain this oxyde mixture

    International Nuclear Information System (INIS)

    Hannerz, K.

    1976-01-01

    An oxide-mixture fuel containing uranium and plutonium dioxides having the slage of spherical, or nearly spherical, oxide-mixture particles with a diameter within the range of from 0.2 to 2 mn charactarized in that each oxide-mixture particles is provided with an outer layer comprising mainly UO2, the thickness of which is at least 0.05; whereas the inner portion of the oxide-mixture particles comprises mainly PUO 2

  1. Effect of additives on enhanced sintering and grain growth in uranium dioxide

    International Nuclear Information System (INIS)

    Bourgeois, L.

    1992-06-01

    The use of sintering additives has been the most effective way of promoting grain growth of uranium dioxide. We have established a same mechanism for additives which belongs to corundum structure: chromium, aluminium, vanadium and titanium sesquioxides. Study of thermodynamical stabilities of dopants has lead to define suitable sintering atmospheres in order to enhance grain growth. Low solubility limits have been defined at T=1700 deg C for four additives, from variations of final grain size versus initial dopant concentration Identification of second phase after cooling has been done from electronic diffraction patterns. It appears that these solubilities decrease sharply as positive deviation from stoichiometry of uranium dioxide increases. Dilatometric analysis of sintering of doped uranium dioxide has shown in certain cases some enhancement in densification rates, at the point of onset of abnormal grain growth, which is believed to be the source. Nevertheless, the following growth is accompanied with pores coalescence mechanisms and pores entrapment inside grains. Increased thermal stability, during standard annealing, is expected, limiting thereby redensification of nuclear fuel in reactors. Finally, from investigations of additives vaporizations, Al 2 O 3 and Cr 2 O 3 , oxygen exchanges between additives and matrix are believed to occur, which should lead to enhance pore mobility. (Author)., refs., figs., tabs

  2. The industrial application of a uranium dioxide electrode

    International Nuclear Information System (INIS)

    Needes, C.R.S.; Nicol, M.J.; Finkelstein, N.P.; Ormrod, G.T.W.

    1975-01-01

    A correlation between the potential of a UO 2 electrode and the rate of recovery of uranium has been proved in laboratory and plant trials. When the recovery rates change because of variation in the concentrations of Fe(III), Fe(II), SO 2- 4 , and H + , a positive correlation is observed. However, an increase in the concentration of phosphate in solution produces an increase in the UO 2 electrode potential but a decrease in the rate of leaching of UO 2 . The correlation between the UO 2 electrode potential and the rate of leaching of UO 2 is then negative. It is concluded that, as a control device, the electrode cannot compete with the platinum electrode for use on certain plants. Nevertheless, the UO 2 electrode will act as a useful warning device if the total concentration of iron in solution decreases to below a level concomitant with the economic recovery of uranium. Furthermore, because of the positive correlation between the UO 2 electrode potential and the phosphate concentration, the electrode will also be of value in the detection of an increase in the phosphate level in solution. When it was incorporated in a suitable industrial probe, the electrode was found to be able to withstand the rigours of the leaching conditions in a large pilot-plant pachuca, and only failed after six weeks operation [af

  3. Determination of microquantities of zirconium and thorium in uranium dioxide

    International Nuclear Information System (INIS)

    Weber de D'Alessio, Ana; Zucal, Raquel.

    1975-07-01

    A method for the determination of 10 to 50 ppm of zirconium and thorium in uranium IV oxide of nuclear purity is established. Zirconium and thorium are retained in a strong cation-exchange resin Dowex 50 WX8 in 1 M HCl. Zirconium is eluted with 0,5% oxalic acid solution and thorium with 4% ammonium oxalate. The colorimetric determination of zirconium with xilenol orange is done in perchloric acid after destructtion of oxalic acid and thorium is determined with arsenazo III in 5 M HCl. 10 μg of each element were determined with a standard deviation of 2,1% for thorium and 3,4% for zirconium. (author) [es

  4. Soiled-based uranium disequilibrium and mixed uranium-thorium series radionuclide reference materials

    International Nuclear Information System (INIS)

    Donivan, S.; Chessmore, R.

    1988-12-01

    The US Department of Energy (DOE) Office of Remedial Action and Waste Technology has assigned the Technical Measurements Center (TMC), located at the DOE Grand Junction Colorado, Projects Office and operated by UNC Geotech (UNC), the task of supporting ongoing remedial action programs by providing both technical guidance and assistance in making the various measurements required in all phases of remedial action work. Pursuant to this task, the Technical Measurements Center prepared two sets of radionuclide reference materials for use by remedial action contractors and cognizant federal and state agencies. A total of six reference materials, two sets comprising three reference materials each, were prepared with varying concentrations of radionuclides using mill tailings materials, ores, and a river-bottom soil diluent. One set (disequilibrium set) contains varying amounts of uranium with nominal amounts of radium-226. The other set (mixed-nuclide set) contains varying amounts of uranium-238 and thorium-232 decay series nuclides. 14 refs., 10 tabs

  5. The Amster concept: a configuration generating its own uranium with a mixed thorium and uranium support

    International Nuclear Information System (INIS)

    Vergnes, J.; Garzenne, C.; Lecarpentier, D.; Mouney, H.; Delpech, M.

    2001-01-01

    AMSTER is a continuously reloaded, graphite-moderated molten salt critical reactor, using a 238 U or 232 Th fuel support, slightly enriched with 235 U if necessary. Using this concept, one can define a large number of configurations according to the products loaded and recycled. The choice of thorium fuel support leads to two configurations requiring no additional 235 U as fissile material: a configuration with one moderating zone, incinerating Transuranium elements (TRU); a configuration with 2 moderating zones self-consuming TRU and regenerating the fissile uranium ( 233 U). In this configuration, it is even possible to burn 238 U (from depleted uranium) by adding it to the thorium support. These configurations use a minimum amount of fuel (100 kg of 232 Th or 100 kg of a 232 Th- 238 U mix per TWh) and produce very little TRU (a few tens of grams per TWh). (author)

  6. Grain growth kinetics in uranium-plutonium mixed oxides

    International Nuclear Information System (INIS)

    Sari, C.

    1986-01-01

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

  7. Improving the neutronic characteristics of a boiling water reactor by using uranium zirconium hydride fuel instead of uranium dioxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, Ahmed Abdelghafar [Higher Technological Institute, Ramadan (Egypt)

    2016-06-15

    The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO{sub 2}) and uranium zirconium hydride (UZrH{sub 1.6}) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO{sub 2} contains 8 × 8 fuel rods while that fueled with UZrH{sub 1.6} contains 9 × 9 fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. UZrH{sub 1.6} fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

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

    International Nuclear Information System (INIS)

    Desrues, R.; Paidassi, J.

    1965-01-01

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

  9. Penetrate-leach dissolution of zirconium-clad uranium and uranium dioxide fuels

    International Nuclear Information System (INIS)

    Harmon, H.D.

    1975-01-01

    A new decladding-dissolution process was developed for zirconium-clad uranium metal and UO 2 fuels. The proposed penetrate-leach process consists of penetrating the zirconium cladding with Alniflex solution (2M HF--1M HNO 3 --1M Al(NO 3 ) 3 --0.1M K 2 Cr 2 O 7 ) and of leaching the exposed core with 10M HNO 3 . Undissolved cladding pieces are discarded as solid waste. Periodic HF and HNO 3 additions, efficient agitation, and in-line zirconium analyses are required for successful control of ZrF 4 and/or AlF 3 precipitation during the cladding-penetration step. Preliminary solvent extraction studies indicated complete recovery of uranium with 30 vol. percent tributyl phosphate (TBP) from both Alniflex solution and blended Alniflex-HNO 3 leach solutions. With 7.5 vol. percent TBP, high extractant/feed flow ratios and low scrub flows are required for satisfactory uranium recovery from Alniflex solution. Modified waste-handling procedures may be required for Alniflex waste, because it cannot be evaporated before neutralization and large quantities of solids are generated on neutralization. The effect of unstable UZr 3 (epsilon phase of uranium-zirconium system) on the safety of penetrate-leach dissolution was investigated

  10. Plutonium-uranium mixed oxide characterization by coupling micro-X-ray diffraction and absorption investigations

    Science.gov (United States)

    Degueldre, C.; Martin, M.; Kuri, G.; Grolimund, D.; Borca, C.

    2011-09-01

    Plutonium-uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The potential differences of metal redox state and microstructural developments of the matrix before and after irradiation are commonly analysed by electron probe microanalysis. In this work the structure and next-neighbor atomic environments of Pu and U oxide features within unirradiated homogeneous MOX and irradiated (60 MW d kg -1) MOX samples was analysed by micro-X-ray fluorescence (μ-XRF), micro-X-ray diffraction (μ-XRD) and micro-X-ray absorption fine structure (μ-XAFS) spectroscopy. The grain properties, chemical bonding, valences and stoichiometry of Pu and U are determined from the experimental data gained for the unirradiated as well as for irradiated fuel material examined in the center of the fuel as well as in its peripheral zone (rim). The formation of sub-grains is observed as well as their development from the center to the rim (polygonization). In the irradiated sample Pu remains tetravalent (>95%) and no (oxidation in the rim zone. Any slight potential plutonium oxidation is buffered by the uranium dioxide matrix while locally fuel cladding interaction could also affect the redox of the fuel.

  11. Deuterium migration and trapping in uranium and uranium dioxide during D+ implantation

    International Nuclear Information System (INIS)

    Lewis, M.B.

    1980-01-01

    Uranium and UO 2 have been implanted with deuterium ions in the energy range 30-85 keV. Subsequently, the near surface regions (100-90000 Angstroem) of these samples were quantitatively profiled for deuterium oxygen using the method of ion beam microanalysis. Mean ranges and widths of the implanted ions were measured and compared with theoretical predictions. Fully oxidized samples were compared with those having only thin oxide films on their surfaces. While the deuterium appeared to migrate during its implantation in uranium, little or no migration appeared either during or after implantation in UO 2 . Further measurements suggest that thin surface oxide films strongly trap the deuterium migrating beneath the surface. It is suggested that the electronic energy loss of the ion beam lowers the effective activation energy for the formation of OD bonds near the target surface. (orig.)

  12. Bonding xenon and krypton on the surface of uranium dioxide single crystal

    Directory of Open Access Journals (Sweden)

    Dąbrowski Ludwik

    2014-08-01

    Full Text Available We present density functional theory (DFT calculation results of krypton and xenon atoms interaction on the surface of uranium dioxide single crystal. A pseudo-potential approach in the generalised gradient approximation (GGA was applied using the ABINIT program package. To compute the unit cell parameters, the 25 atom super-cell was chosen. It has been revealed that close to the surface of a potential well is formed for xenon and krypton atom due to its interaction with the atoms of oxygen and uranium. Depth and shape of the well is the subject of ab initio calculations in adiabatic approximation. The calculations were performed both for the case of oxygenic and metallic surfaces. It has been shown that the potential well for the oxygenic surface is deeper than for the metallic surface. The thermal stability of immobilising the atoms of krypton and xenon in the potential wells were evaluated. The results are shown in graphs.

  13. Determination of trace elements in ceramic uranium dioxide pellets powders CRMs by ICP-AES

    International Nuclear Information System (INIS)

    Liu Husheng; Li Jun

    1997-01-01

    The 237-quaternary ammonium extraction resin chromatography is used to the separation of 6 trace elements in ceramic uranium dioxide pellets powders, which are used as certified reference materials (CRMs). The sample is dissolved in 6.5 mol/L HNO 3 and uranium is separated by chromatographic column. the 6 trace elements Al, Ba, Co, Ta, Ti and V contained in the elutriant are determined by using ICP directly reading spectrometer. For a 300 mg sample, the lowest determinable concentration of impurities in ceramic UO 2 pellets powders CRMs is (0.016-0.250) x 10 -6 . The relative standard deviation is less than 7.5%. The proposed method provides excellent and accurate analytical data for the ceramic UO 2 pellets powders samples (CRMs)

  14. Energetics of intrinsic point defects in uranium dioxide from electronic-structure calculations

    International Nuclear Information System (INIS)

    Nerikar, Pankaj; Watanabe, Taku; Tulenko, James S.; Phillpot, Simon R.; Sinnott, Susan B.

    2009-01-01

    The stability range of intrinsic point defects in uranium dioxide is determined as a function of temperature, oxygen partial pressure, and non-stoichiometry. The computational approach integrates high accuracy ab initio electronic-structure calculations and thermodynamic analysis supported by experimental data. In particular, the density functional theory calculations are performed at the level of the spin polarized, generalized gradient approximation and includes the Hubbard U term; as a result they predict the correct anti-ferromagnetic insulating ground state of uranium oxide. The thermodynamic calculations enable the effects of system temperature and partial pressure of oxygen on defect formation energy to be determined. The predicted equilibrium properties and defect formation energies for neutral defect complexes match trends in the experimental literature quite well. In contrast, the predicted values for charged complexes are lower than the measured values. The calculations predict that the formation of oxygen interstitials becomes increasingly difficult as higher temperatures and reducing conditions are approached

  15. DISSOLUTION OF METAL OXIDES AND SEPARATION OF URANIUM FROM LANTHANIDES AND ACTINIDES IN SUPERCRITICAL CARBON DIOXIDE

    Energy Technology Data Exchange (ETDEWEB)

    Donna L. Quach; Bruce J. Mincher; Chien M. Wai

    2013-10-01

    This paper investigates the feasibility of extracting and separating uranium from lanthanides and other actinides by using supercritical fluid carbon dioxide (sc-CO2) as a solvent modified with tri-n-butylphosphate (TBP) for the development of a counter current stripping technique, which would be a more efficient and environmentally benign technology for spent nuclear fuel reprocessing compared to traditional solvent extraction. Several actinides (U, Pu, and Np) and europium were extracted in sc-CO2 modified with TBP over a range of nitric acid concentrations and then the actinides were exposed to reducing and complexing agents to suppress their extractability. According to this study, uranium/europium and uranium/plutonium extraction and separation in sc-CO2 modified with TBP is successful at nitric acid concentrations of less than 6 M and at nitric acid concentrations of less than 3 M with acetohydroxamic acid or oxalic acid, respectively. A scheme for recycling uranium from spent nuclear fuel by using sc-CO2 and counter current stripping columns is presented.

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

    International Nuclear Information System (INIS)

    Xuegang Wang; Zhongkui Zhou

    2017-01-01

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

  17. Thermochemical modeling of the plutonium and uranium-plutonium dioxides

    International Nuclear Information System (INIS)

    Besmann, T.M.; Lindemer, T.B.

    1984-01-01

    The chemical thermodynamic properties of the actinide oxides have long been of interest for nuclear fuel design and for predicting fuel behavior under accident conditions. The result of such interest has been the publication of many studies over several decades containing thousands of measurements. The calcium fluorite structure and phases have been intensely studied, with in excess of 1000 data points having been determined. The object of the current work is to develop quantitative models of and which accurately describe the oxygen potential-temperature-composition behavior of the phases. The entire available data base of oxygen potential-temperature-composition values were extracted for use in the development of the models for the plutonia and mixed oxide phases. With perhaps the exception of Babelot et al., little effort has been made to utilize the large existing data base in such analyses. These data were instrumental in developing our models for the oxides, indicating the appropriate oxygen potential-composition relationships and providing for the determination of parametric values. The modeling approach used by us is fundamentally simple, utilizing the assumption that the complex oxides can be described as solutions of oxides with invariant stoichiometries. The chemical thermodynamic models for and described here are among the first to make extensive use of the large oxygen potential-temperature-composition data base which exists for these systems. These relatively simple models should be easily applied to the design of fuel compositions the analysis of behavior during burnup, and the development of codes for accident analysis. 5 references

  18. Welding uranium with a multikilowatt, continuous-wave, carbon dioxide laser welder

    International Nuclear Information System (INIS)

    Turner, P.W.; Townsend, A.B.

    1977-01-01

    A 15-kilowatt, continuous-wave carbon dioxide laser was contracted to make partial-penetration welds in 6.35-and 12.7-mm-thick wrought depleted uranium plates. Welding power and speed ranged from 2.3 to 12.9 kilowatts and from 21 to 127 millimeters per second, respectively. Results show that depth-to-width ratios of at least unity are feasible. The overall characteristics of the process indicate it can produce welds resembling those made by the electron-beam welding process

  19. Influence of uranium dioxide nonstoichiometric oxygen on the work function of Mo(110) single crystal

    International Nuclear Information System (INIS)

    Bekmukhabetov, E.S.; Dzhajmurzin, A.A.; Imanbekov, Zh.Zh.

    1985-01-01

    The influence of the uranium dioxide nonstoichiometric oxygen on the work function of a Mo(110) single crystal has been studied. When the surface diffusion of oxygen on the tested surface takes place, the work function is shown to decrease and, subsequently, to increase until it becomes stable. The dependence of the work function on the temperature of the specimen in the range of 1600-1900 K with a minimum at 1730 K has been found. The minimum is attributed to the dipole layer formation

  20. Supercritical Fluid Extraction (SFE) of uranium and thorium nitrates using carbon dioxide modified with phosphonates

    International Nuclear Information System (INIS)

    Pitchaiah, K.C.; Sujatha, K.; Brahmananda Rao, C.V.S.; Sivaraman, N.; Vasudeva Rao, P.R.

    2014-01-01

    Supercritical Fluid Extraction (SFE) has emerged as a powerful technique for the extraction of metal ions.The liquid like densities and gas like physical properties of supercritical fluids make them unique to act as special solvents. SFE based procedures were developed and demonstrated in our laboratory for the recovery of actinides from various matrices. In the present study, we have examined for the first time, the use of dialkylalkylphosphonates in supercritical carbon dioxide (Sc-CO 2 ) medium to study the extraction behavior of uranium and thorium nitrates. A series of phosphonates were synthesised by Michaelis-Becker reaction in our laboratory and employed for the SFE

  1. Quantification of the effect of in-situ generated uranium metal on the experimentally determined O/U ratio of a sintered uranium dioxide fuel pellet

    International Nuclear Information System (INIS)

    Narasimha Murty, B.; Bharati Misra, U.; Yadav, R.B.; Srivastava, R.K.

    2005-01-01

    This paper describes quantitatively the effect of in-situ generated uranium metal (that could be formed due to the conducive manufacturing conditions) in a sintered uranium dioxide fuel pellet on the experimentally determined O/U ratio using analytical methods involving dissolution of the pellet material. To quantify the effect of in-situ generated uranium metal in the fuel pellet, a mathematical expression is derived for the actual O/U ratio in terms of the O/U ratio as determined by an experiment involving dissolution of the material and the quantity of uranium metal present in the uranium dioxide pellet. The utility of this derived mathematical expression is demonstrated by tabulating the calculated actual O/U ratios for varying amounts of uranium metal (from 5 to 95% in 5% intervals) and different O/U ratio values (from 2.001 to 2.015 in 0.001 intervals). This paper brings out the necessity of care to be exercised while interpreting the experimentally determined O/U ratio and emphasizes the fact that it is always safer to produce the nuclear fuel with oxygen to uranium ratios well below the specified maximum limit of 2.015. (author)

  2. Carbonate effects on hexavalent uranium removal from water by nanocrystalline titanium dioxide

    International Nuclear Information System (INIS)

    Wazne, Mahmoud; Meng, Xiaoguang; Korfiatis, George P.; Christodoulatos, Christos

    2006-01-01

    A novel nanocrystalline titanium dioxide was used to treat depleted uranium (DU)-contaminated water under neutral and alkaline conditions. The novel material had a total surface area of 329 m 2 /g, total surface site density of 11.0 sites/nm 2 , total pore volume of 0.415 cm 3 /g and crystallite size of 6.0 nm. It was used in batch tests to remove U(VI) from synthetic solutions and contaminated water. However, the capacity of the nanocrystalline titanium dioxide to remove U(VI) from water decreased in the presence of inorganic carbonate at pH > 6.0. Adsorption isotherms, Fourier transform infrared (FTIR) spectroscopy, and surface charge measurements were used to investigate the causes of the reduced capacity. The surface charge and the FTIR measurements suggested that the adsorbed U(VI) species was not complexed with carbonate at neutral pH values. The decreased capacity of titanium dioxide to remove U(VI) from water in the presence of carbonate at neutral to alkaline pH values was attributed to the aqueous complexation of U(VI) by inorganic carbonate. The nanocrystalline titanium dioxide had four times the capacity of commercially available titanium dixoide (Degussa P-25) to adsorb U(VI) from water at pH 6 and total inorganic carbonate concentration of 0.01 M. Consequently, the novel material was used to treat DU-contaminated water at a Department of Defense (DOD) site

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  4. The utilization of uranium industry technology and relevant chemistry to leach uranium from mixed-waste solids

    International Nuclear Information System (INIS)

    Mattus, A.J.; Farr, L.L.

    1991-01-01

    Methods for the chemical extraction of uranium from a number of refractory uranium-containing minerals found in nature have been in place and employed by the uranium mining and milling industry for nearly half a century. These same methods, in conjunction with the principles of relevant uranium chemistry, have been employed at the Oak Ridge National Laboratory (ORNL) to chemically leach depleted uranium from mixed-waste sludge and soil. The removal of uranium from what is now classified as mixed waste may result in the reclassification of the waste as hazardous, which may then be delisted. The delisted waste might eventually be disposed of in commercial landfill sites. This paper generally discusses the application of chemical extractive methods to remove depleted uranium from a biodenitrification sludge and a storm sewer soil sediment from the Y-12 weapons plant in Oak Ridge. Some select data obtained from scoping leach tests on these materials are presented along with associated limitations and observations which might be useful to others performing such test work. 6 refs., 2 tabs

  5. The utilization of uranium industry technology and relevant chemistry to leach uranium from mixed-waste solids

    Energy Technology Data Exchange (ETDEWEB)

    Mattus, A.J.; Farr, L.L.

    1991-01-01

    Methods for the chemical extraction of uranium from a number of refractory uranium-containing minerals found in nature have been in place and employed by the uranium mining and milling industry for nearly half a century. These same methods, in conjunction with the principles of relevant uranium chemistry, have been employed at the Oak Ridge National Laboratory (ORNL) to chemically leach depleted uranium from mixed-waste sludge and soil. The removal of uranium from what is now classified as mixed waste may result in the reclassification of the waste as hazardous, which may then be delisted. The delisted waste might eventually be disposed of in commercial landfill sites. This paper generally discusses the application of chemical extractive methods to remove depleted uranium from a biodenitrification sludge and a storm sewer soil sediment from the Y-12 weapons plant in Oak Ridge. Some select data obtained from scoping leach tests on these materials are presented along with associated limitations and observations which might be useful to others performing such test work. 6 refs., 2 tabs.

  6. Observations concerning the particle-size of the oxidation products of uranium formed in air or in carbon dioxide

    International Nuclear Information System (INIS)

    Baque, P.; Leclercq, D.

    1964-01-01

    This report brings together the particle-size analysis results obtained on products formed by the oxidation or the ignition of uranium in moist air or dry carbon dioxide. The results bring out the importance of the nature of the oxidising atmosphere, the combustion in moist air giving rise to the formation of a larger proportion of fine particles than combustion in carbon dioxide under pressure. (authors) [fr

  7. Contribution to the study of sputtering and damage of uranium dioxide by fast heavy ions

    International Nuclear Information System (INIS)

    Schlutig, S.

    2001-03-01

    Swift heavy ion-solid interaction leads in volume to track creation and on the surface to the ejection of particles into the vacuum. To learn more about initial mechanisms of track formation, we are focused on the sputtering of uranium dioxide by fast heavy ions. This present study is exclusively devoted to the influence of the electronic stopping power on the emission of neutral particles and especially on their angular distribution. These measurements are completed by those of the ions emitted from UO 2 targets bombarded with swift heavy ions. The whole experimental results give access to: i) the nature of the sputtered particles; ii) the charge state of the emitted particles; iii) the direction of ejection of the sputtered particles ; iv) the sputtering yields deduced from the angular distributions. These results are compared to the prediction of the sputtering models proposed in the literature and it seems that the supersonic gas flow model is well suited to describe our results. Finally, the sputtering yields are compared with a set of earlier experimental data on uranium dioxide damage obtained by T. Wiss and we observe that only a small fraction of UO 2 monolayers are sputtered. (author)

  8. Anomalous behaviour of thermophysical properties of stoichiometric uranium dioxide by molecular dynamics simulation

    International Nuclear Information System (INIS)

    Lunev, A.V.; Tarasov, B.A.; Nazarov, A.V.

    2011-01-01

    We present a classical molecular dynamics simulation of uranium dioxide in the temperature range of 300-3000 K. Temperature dependences of thermal conductivity, heat capacity and ionic conductivity are investigated. Our study shows the rise of thermal conductivity of uranium dioxide at very high temperatures (above 2500 K), which is not predicted by the former anharmonic theories. Several pair potentials are used in the simulation, and they depict similar effects. Long range forces are accounted by Ewald sums. Static thermal properties are evaluated in NPT ensemble. It is shown that a high-temperature peak on heat capacity is present and is more legible in large systems. To ensure the best reliability, transport properties are evaluated using the theory of autocorrelation functions in NVE ensemble. In order to properly define thermal conductivity in ionic systems with charge fluxes, an expression which accounts the thermoelectric effect is derived from Onsager reciprocal relations. The rise on temperature dependence of thermal conductivity is accompanied by the peak on heat capacity and an anomalous rise of ionic conductivity. However, it is shown that there is no partial melting of the oxygen sublattice, which suggests that the system does not necessarily exhibit a superionic transition. Instead, kick-out diffusion in oxygen sublattice is proposed to be the origin of such anomalous behavior of thermophysical properties. (author)

  9. Contribution to the study of second phases particles dispersion in polycrystalline uranium dioxide

    International Nuclear Information System (INIS)

    Peres, V.

    1994-06-01

    To reduce fission gas release of irradiated polycrystalline uranium dioxide, the dispersion of intragranular nanometric particles of second phase necessary to pin gas bubbles may complete the advantage of a large-grained fuel microstructure. Moreover, intergranular glass films may improve high temperatures mechanical properties of UO 2 . In this study, mixtures of additives composed of ''corindon'' structure oxides that enhance the fuel grain growth and composed of different oxides with variable solid solubilities in the UO 2 matrix were achieved. Additives with a negligible solubility inhibit grain boundaries motion except those, such as silica, that involve a liquid phase at the sintering temperature. Rare earth oxides that form stable solid solutions with UO 2 cannot lead to precipitation, but have no effect on the fuel grain growth doped with ''corindon'' type oxides. A chromium oxide excess allows the creation of a fuel microstructure described by large grains and intragranular spherical Cr 2 O 3 inclusions observed by scanning electron microscopy. Values for the bulk lattice diffusion coefficient of Cr 3+ cations in UO 2 can be deduced from the experimental growth of those dispersed particles by an Ostwald ripening mechanism. The formation of small precipitated metal particles inside the uranium dioxide matrix induced by the internal reduction of a solid solution has not been performed. However, direct reduction of insoluble chromium oxide particles is easy and produces metallic intragranular precipitates. (author). 119 refs., 112 figs., 33 tabs., 5 annexes

  10. Synthesis of uranium and thorium dioxides by Complex Sol-Gel Processes (CSGP). Synthesis of uranium oxides by Complex Sol-Gel Processes (CSGP)

    International Nuclear Information System (INIS)

    Deptula, A.; Brykala, M.; Lada, W.; Olczak, T.; Wawszczak, D.; Chmielewski, A.G.; Modolo, G.; Daniels, H.

    2010-01-01

    In the Institute of Nuclear Chemistry and Technology (INCT), a new method of synthesis of uranium and thorium dioxides by original variant of sol-gel method - Complex Sol-Gel Process (CSGP), has been elaborated. The main modification step is the formation of nitrate-ascorbate sols from components alkalized by aqueous ammonia. Those sols were gelled into: - irregularly agglomerates by evaporation of water; - medium sized microspheres (diameter <150) by IChTJ variant of sol-gel processes by water extraction from drops of emulsion sols in 2-ethylhexanol-1 by this solvent. Uranium dioxide was obtained by a reduction of gels with hydrogen at temperatures >700 deg. C, while thorium dioxide by a simple calcination in the air atmosphere. (authors)

  11. The reaction of sintered aluminium products with uranium dioxide and monocarbide

    DEFF Research Database (Denmark)

    Lauritzen, T.; Knudsen, Per

    1965-01-01

    The compatibility of SAP 930 with uranium dioxide and uranium monocarbide was investigated in the temperature range 450–600° C. The results indicate that a severe reaction occurs between SAP 930 and UO2 within 8000 hours at 600° C, a slight reaction at 600° C for 1000 hours and after 11 900 hours...... at 525° C, and no reaction in 14 300 hours at 450° C. Of the three grades of UC tested (hot pressed, arc cast, cold pressed and sintered) the slightly substoichiometric, hot-pressed UC is judged to be least compatible with SAP 930, reaction occurring after 7300 hours at 450° C. No reaction was observed...... between SAP 930 and the other carbides at this temperature. All SAP−UC combinations are incompatible at 600° C for as little as 100 hours of heat treatment. Tests designed to study the effect of a diffusion barrier on the SAP−UC reaction have shown that anodized SAP 930 and the three uranium carbides...

  12. Electronic structure calculations of atomic transport properties in uranium dioxide: influence of strong correlations

    International Nuclear Information System (INIS)

    Dorado, B.

    2010-09-01

    Uranium dioxide UO 2 is the standard nuclear fuel used in pressurized water reactors. During in-reactor operation, the fission of uranium atoms yields a wide variety of fission products (FP) which create numerous point defects while slowing down in the material. Point defects and FP govern in turn the evolution of the fuel physical properties under irradiation. In this study, we use electronic structure calculations in order to better understand the fuel behavior under irradiation. In particular, we investigate point defect behavior, as well as the stability of three volatile FP: iodine, krypton and xenon. In order to take into account the strong correlations of uranium 5f electrons in UO 2 , we use the DFT+U approximation, based on the density functional theory. This approximation, however, creates numerous metastable states which trap the system and induce discrepancies in the results reported in the literature. To solve this issue and to ensure the ground state is systematically approached as much as possible, we use a method based on electronic occupancy control of the correlated orbitals. We show that the DFT+U approximation, when used with electronic occupancy control, can describe accurately point defect and fission product behavior in UO 2 and provide quantitative information regarding point defect transport properties in the oxide fuel. (author)

  13. Polarographic determination of uranium dioxide stoichiometry; La determination polarographique de la stoechiometrie du dioxyde d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Viguie, J.; Uny, G. [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Grenoble, 38 (France)

    1966-10-01

    The method described allows the determination of small deviations from stoichiometry for uranium dioxide. It was applied to the study of surface oxidation of bulk samples. The sample is dissolved in phosphoric acid under an argon atmosphere; U(VI) is determined by polarography in PO{sub 4}H{sub 3} 4.5 N - H{sub 2}SO{sub 4} 4 N. U(IV) is determined by potentiometry. The detection limit is UO{sub 2,0002}. The accuracy for a single determination at the 95% confidence level is {+-}20 per cent for samples with composition included between UO{sub 2,001} and UO{sub 2,01}. (authors) [French] La methode decrite permet de determiner les faibles ecarts a la stoechiometrie du dioxyde d'uranium. Elle a ete appliquee a l'etude de l'oxydation superficielle des echantillons. La mise en solution s'effectue dans l'acide phosphorique concentre sous atmosphere d'argon; U(VI) est dose par polarographie dans le milieu PO{sub 4}H{sub 3} 4,5 N et H{sub 2}SO{sub 4} 4 N; U(IV) est dose par potentiometrie. La limite de detection est UO{sub 2,0002}. La precision obtenue pour une determination au taux de certitude 0,95 est de l'ordre de 20 pour cent pour des echantillons dont la teneur est comprise entre UO{sub 2,001} et UO{sub 2,01}. (auteurs)

  14. Evaluation of a titanium dioxide-based DGT technique for measuring inorganic uranium species in fresh and marine waters

    DEFF Research Database (Denmark)

    Hutchins, Colin M.; Panther, Jared G.; Teasdale, Peter R.

    2012-01-01

    A new diffusive gradients in a thin film (DGT) technique for measuring dissolved uranium (U) in freshwater is reported. The new method utilises a previously described binding phase, Metsorb (a titanium dioxide based adsorbent). This binding phase was evaluated and compared to the well-established...

  15. The 1/4 technical scale, continuous process of obtaining the ceramic uranium dioxide from ammonium polyuranates containing fluoride

    International Nuclear Information System (INIS)

    Wlodarski, R.

    1977-01-01

    Based on the laboratory results, the 1/4 technical apparatus for the continuous reduction and defluorination of ammonium polyuranate containing fluoride was designed and constructed. The possibility of obtaining the ceramic uranium dioxide in a continuous process has been confirmed. The main part of the apparatus used in this process was the horizontal tubular oven with the extruder transporting material. (author)

  16. Nuclear energy - Uranium dioxide pellets - Determination of density and volume fraction of open and closed porosity. 2. ed. 2. ed.

    International Nuclear Information System (INIS)

    2008-01-01

    This International Standard describes a method for determining the chlorine and fluorine concentrations in uranium dioxide and in sintered fuel pellets by pyrohydrolysis of samples, followed either by liquid ion-exchange chromatography or by selective electrode measurement of chlorine and fluorine ions. Many ion-exchange chromatography systems and ion-selective electrode measurement systems are available

  17. Melting temperature of uranium - plutonium mixed oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  18. Melting temperature of uranium - plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    Ishii, Tetsuya; Hirosawa, Takashi

    1997-08-01

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

  19. Preparation of uranium-plutonium mixed nitride pellets with high purity

    International Nuclear Information System (INIS)

    Arai, Yasuo; Shiozawa, Ken-ichi; Ohmichi, Toshihiko

    1992-01-01

    Uranium-plutonium mixed nitride pellets have been prepared in the gloveboxes with high purity Ar gas atmosphere. Carbothermic reduction of the oxides in N 2 -H 2 mixed gas stream was adopted for synthesizing mixed nitride. Sintering was carried out in various conditions and the effect on the pellet characteristics was investigated. (author)

  20. Mixed Uranium/Refractory Metal Carbide Fuels for High Performance Nuclear Reactors

    International Nuclear Information System (INIS)

    Knight, Travis; Anghaie, Samim

    2002-01-01

    Single phase, solid-solution mixed uranium/refractory metal carbides have been proposed as an advanced nuclear fuel for advanced, high-performance reactors. Earlier studies of mixed carbides focused on uranium and either thorium or plutonium as a fuel for fast breeder reactors enabling shorter doubling owing to the greater fissile atom density. However, the mixed uranium/refractory carbides such as (U, Zr, Nb)C have a lower uranium densities but hold significant promise because of their ultra-high melting points (typically greater than 3700 K), improved material compatibility, and high thermal conductivity approaching that of the metal. Various compositions of (U, Zr, Nb)C were processed with 5% and 10% metal mole fraction of uranium. Stoichiometric samples were processed from the constituent carbide powders, while hypo-stoichiometric samples with carbon-to-metal (C/M) ratios of 0.92 were processed from uranium hydride, graphite, and constituent refractory carbide powders. Processing techniques of cold uniaxial pressing, dynamic magnetic compaction, sintering, and hot pressing were investigated to optimize the processing parameters necessary to produce high density (low porosity), single phase, solid-solution mixed carbide nuclear fuels for testing. This investigation was undertaken to evaluate and characterize the performance of these mixed uranium/refractory metal carbides for high performance, ultra-safe nuclear reactor applications. (authors)

  1. Synthesis and insertion chemistry of mixed tether uranium metallocene complexes

    Energy Technology Data Exchange (ETDEWEB)

    Siladke, Nathan A.; LeDuc, Jennifer; Ziller, Joseph W.; Evans, William J. [Department of Chemistry, University of California, Irvine, CA (United States)

    2012-11-12

    The synthesis of mixed tethered alkyl uranium metallocenes has been investigated by examining the reactivity of the bis(tethered alkyl) metallocene [(η{sup 5}-C{sub 5}Me{sub 4}SiMe{sub 2}CH{sub 2}-κC){sub 2}U] (1) with substrates that react with only one of the U-C linkages. The effect of these mixed tether coordination environments on the reactivity of the remaining U-C bond has been studied by using CO insertion chemistry. One equivalent of azidoadamantane (AdN{sub 3}) reacts with 1 to yield the mixed tethered alkyl triazenido complex [(η{sup 5}-C{sub 5}Me{sub 4}SiMe{sub 2}CH{sub 2}-κC)U(η{sup 5}-C{sub 5}Me{sub 4}SiMe{sub 2}-CH{sub 2}NNN-Ad-κ{sup 2}N{sup 1,3})]. Similarly, a single equivalent of CS{sub 2} reacts with 1 to form the mixed tethered alkyl dithiocarboxylate complex [(η{sup 5}-C{sub 5}Me{sub 4}SiMe{sub 2}CH{sub 2}-κC)U(η{sup 5}-C{sub 5}Me{sub 4}SiMe{sub 2}- CH{sub 2}C(S){sub 2}-κ{sup 2}S,S{sup '})], a reaction that constitutes the first example of CS{sub 2} insertion into a U{sup 4+}-C bond. Complex 1 reacts with one equivalent of pyridine N-oxide by C-H bond activation of the pyridine ring to form a mixed tethered alkyl cyclometalated pyridine N-oxide complex [(η{sup 5}-C{sub 5}Me{sub 4}SiMe{sub 2}CH{sub 2}-κC)(η{sup 5}-C{sub 5}Me{sub 4}SiMe{sub 3})U(C{sub 6}H{sub 4}NO-κ{sup 2} C,O)]. The remaining (η{sup 5}-C{sub 5}Me{sub 4}SiMe{sub 2}CH{sub 2}-κC){sup 2-} ligand in each of these mixed tethered species show reactivity towards CO and tethered enolate ligands form by insertion. Subsequent rearrangement have been identified in [(η{sup 5}-C{sub 5}Me{sub 4}SiMe{sub 3})U(C{sub 5}H{sub 4}NO-κ{sup 2}C,O)(η{sup 5}-C{sub 5}Me{sub 4}SiMe{sub 2}C(=CH{sub 2})O- κO)] and [(η{sup 5}-C{sub 5}Me{sub 4}SiMe{sub 2}CH{sub 2}NNN-Ad-κ{sup 2}N{sup 1,3})U(η{sup 5}-C{sub 5}Me{sub 4}SiMe{sub 2}C(=CH{sub 2})O-κO)]. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  2. Fracture toughness and fracture surface energy of sintered uranium dioxide fuel pellets

    International Nuclear Information System (INIS)

    Kutty, T.R.G.; Chandrasekharan, K.N.; Panakkal, J.P.; Ghosh, J.K.

    1987-01-01

    The paper concerns the variation of fracture toughness Ksub(ic) and fracture surface energy γsub(s) in sintered uranium dioxide pellets in the density range 9.86 to 10.41 g cm -3 , using Vickers indentation technique. A minimum of four indentations were made on each pellet sample and the average crack length of each indentation and the hardness values were determined. The overall average crack-length datra and the data on volume fraction porosity in the pellets fitted a straight line, from which Ksub(ic) and γsub(s) were calculated. The fracture parameters of nonporous polycrystalline UO 2 , calculated from the experimental data, are presented in tabular form. (U.K.)

  3. The migration of intra-granular fission gas bubbles in irradiated uranium dioxide

    International Nuclear Information System (INIS)

    Baker, C.

    1977-05-01

    The mobility of intragranular fission gas bubbles in uranium dioxide irradiated at 1600-1800 0 C has been studied following isothermal annealing at temperatures below 1600 0 C. The intragranular fission gas bubbles, average diameter approximately 2nm, are virtually immobile at temperatures below 1500 0 C. The bubbles have clean surfaces with no solid fission product contamination and are faceted to the highest observed irradiation temperature of 1800 0 C. This bubble faceting is believed to be a major cause of bubble immobility. In fuel operating below 1500 0 C the predominant mechanism allowing the growth of intragranular bubbles and the subsequent gas release must be the diffusion of dissolved gas atoms rather than the movement of entire intragranular bubbles. (author)

  4. Plastic deformation of uranium dioxide: observation of the sub-structures of dislocations

    International Nuclear Information System (INIS)

    Alamo, A.; Lefebvre, J.M.; Soullard, J.

    1978-01-01

    Single crystals of uranium dioxide were deformed in compression at imposed strain rates in the temperature range of 700 0 C to 1400 0 C. The crystals were oriented to promote slip over one or two slip systems of the family [100] and also on the [110] system. Thin films of the deformed specimens were examined by transmission electron microscopy. When [100] single glide system operates, the dislocation substructure consist of numerous dipoles, their edge components lying along directions. For the [100] double glide system the grain boundaries and dislocation hexagonal network are observed, the complexity of which increases with the nominal strain. Dislocation arrangments consisting of extensive cellular networks of tangling dislocations and hexagonal netting were detected for [110] system. The auxillary role of [111] planes on the dislocation cross slip from [100] and [110] system was demonstrated. Weak beam images suggest that dissociation of dislocations can occur. (Auth.)

  5. Vapor pressures and vapor compositions in equilibrium with hypostoichiometric uranium-plutonium dioxide at high temperatures

    International Nuclear Information System (INIS)

    Green, D.W.; Fink, J.K.; Leibowitz, L.

    1982-01-01

    Vapor pressures and vapor compositions in equilibrium with a hypostoichiometric uranium-plutonium dioxide condensed phase (U/sub 1-y/Pu/sub y/)O/sub 2-x/, as functions of T, x, and y, have been calculated for 0.0 less than or equal to x less than or equal to 0.1, 0.0 less than or equal to y less than or equal to 0.3, and for the temperature range 2500 less than or equal to T less than or equal to 6000 K. The range of compositions and temperatures was limited to the region of interest to reactor safety analysis. Thermodynamic functions for the condensed phase and for each of the gaseous species were combined with an oxygen potential model to obtain partial pressures of O, O 2 , Pu, PuO, PuO 2 , U, UO, UO 2 , and UO 3 as functions of T, x, and y

  6. A new mechanistic and engineering fission gas release model for a uranium dioxide fuel

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Yang, Yong Sik; Kim, Dae Ho; Kim, Sun Ki; Bang, Je Geun

    2008-01-01

    A mechanistic and engineering fission gas release model (MEGA) for uranium dioxide (UO 2 ) fuel was developed. It was based upon the diffusional release of fission gases from inside the grain to the grain boundary and the release of fission gases from the grain boundary to the external surface by the interconnection of the fission gas bubbles in the grain boundary. The capability of the MEGA model was validated by a comparison with the fission gas release data base and the sensitivity analyses of the parameters. It was found that the MEGA model correctly predicts the fission gas release in the broad range of fuel burnups up to 98 MWd/kgU. Especially, the enhancement of fission gas release in a high-burnup fuel, and the reduction of fission gas release at a high burnup by increasing the UO 2 grain size were found to be correctly predicted by the MEGA model without using any artificial factor. (author)

  7. The study of Ashby-type sintering diagrams for uranium dioxide

    International Nuclear Information System (INIS)

    Georgeoni, P.

    1980-01-01

    Computer modelling of binary and ternary Ashby-type sintering diagrams for stoechiometric and hyperstoechiometric uranium dioxide (in the range O/U = 2, 0-2, 10). Material data and mass transfer equations, selected from the literature, were used. Sintering isochronous curves were calculated and traced as well. Improvement of a modern dilatometric method by reading and processing experimental curves on a computer and by determining for them a criterion of proximity to the theoretical model equation. It was possible: to develop a reliable method of determination for the dominant mechanism, diffusion coefficient and real process activation energy; to draw up the real sintering diagram; to understand the quantitative and qualitative changes occuring during the actual sintering process of UO 2 , concerning massing and modification of pore shape; to recommend the technological parameters of the thermal regime concerning the elimination of lubricant and binder additives in order to obtain high quality sintered tablets. (author)

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

    International Nuclear Information System (INIS)

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

    1988-04-01

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

  9. Study of process parameters for reducing ammonium uranyl carbonate to uranium dioxide in fluidized bed furnace

    International Nuclear Information System (INIS)

    Leitao Junior, C.B.

    1992-01-01

    This work consists of studying the process parameters of AUC (ammonium uranyl carbonate) to U O 2 (uranium dioxide) reduction, with good physical and chemical characteristics, in fluidized bed. Initially, it was performed U O 2 cold fluidization experiments with an acrylic column. Afterward, it was done AUC to U O 2 reduction experiments, in which the process parameters influence in the granulometry, specific surface area, porosity and fluoride amount on the U O 2 powder produced were studied. As a last step, it was done compacting and sintering tests of U O 2 pellets in order to appreciate the U O 2 powder performance, obtained by fluidized bed, in the fuel pellets fabrication. (author)

  10. Hot deformation of polycrystalline uranium dioxide: from microscopic mechanisms to macroscopic behaviour

    International Nuclear Information System (INIS)

    Dherbey, Francine

    2000-01-01

    The improvement of nuclear fuels performances in PWR requires in particular an enhancement of creep ability of uranium dioxide in order to minimise rupture risks of the cladding material during interactions between pellets and cladding. The aim of this study is to investigate the link between the ceramic macroscopic thermo-mechanical behaviour and the changes in the fuel microstructure during deformation. Stoichiometric UO 2 pellets with various grains sizes from 9 pm to 36 μm have been deformed by compression at intermediate temperatures, i.e. near T M /2, and quenched under stress. The damage is characterised by the presence of cavities at low stresses and cracks at high stresses, both along grain boundaries parallel to the compression axis. Inside grains, dislocations organise themselves into cellular substructures in which sub-boundaries are made of dislocation hexagonal networks. In these conditions, uranium dioxide deformation is described by grain boundary sliding, which is the main origin of material damage, partially accommodated by dislocational creep inside grains. A steady-state creep model is proposed on a physical basis. It accounts for the almost similar contributions of two mechanisms which are grain boundaries sliding and intragranular creep, and takes into account the grain boundary roughness. In contrast with phenomenological descriptions used up to now, this picture leads to a unique creep law on the whole range of stresses explored here, from 10 MPa to 80 MPa. The creep rate controlling mechanism seems to be the migration of sub-boundaries. The deformation at constant strain rate is controlled by the same mechanisms as creep. (author) [fr

  11. Contribution to the study of the creep of uranium dioxide. Role of grain growth promoters

    International Nuclear Information System (INIS)

    Vivant-Duguay, Christelle

    1998-01-01

    Improvement of nuclear fuel performances involves enhancing the plasticity of uranium dioxide UO 2 , in order to reduce the stress applied by the pellet to the cladding during a power ramp. The objective of this work is to identify and to formulate the effects produced by the nature and the concentration of additives of corundum structure, Cr 2 O 3 or Al 2 O 3 , which are grain growth promoters for UO 2 . The review of literature data establishes that oxygen content, grain size or porosity markedly affect the mechanical properties of uranium dioxide. On the other hand, there is relatively little reported work on the influence of doping. Prepared samples have been deformed by uniaxial compression. In the case of standard undoped UO 2 , two distinct preponderant creep mechanisms occur depending on stress level: a grain boundary diffusional creep, as per Coble, for stresses below the transition stress and a dislocation creep above. The doped materials have a large grained microstructure, which allows a dislocation creep only. In the range of temperature and stress investigated here, doping significantly improves the plasticity of standard UO 2 . This common effect of dopants is characterized by a decrease in the flow stress for tests with constant strain rate and by enhanced steady-state creep rates. Cr 2 O 3 doping is the more effective. The apparent benefit of doping results from the gain due to the increased grain size, but it is compensated by the strengthening effect of the additive. The creep law used to describe the behavior of standard UO 2 , has been modified to account for the influence of the dopant, by including either the concentration or the grain size. (author) [fr

  12. Assessment of uranium dioxide fuel performance with the addition of beryllium oxide

    Energy Technology Data Exchange (ETDEWEB)

    Muniz, Rafael O.R.; Abe, Alfredo; Gomes, Daniel S.; Silva, Antonio T., E-mail: romuniz@usp.br, E-mail: ayabe@ipen.br, E-mail: danieldesouza@gmail.com, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energética s e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LabRisco/USP), Sao Paulo, SP (Brazil). Lab. de Análise, Avaliação e Gerenciamento de Risco; Aguiar, Amanda A., E-mail: amanda.abati.aguiar@gmail.com [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil)

    2017-07-01

    The Fukushima Daiichi accident in 2011 pointed the problem related to the hydrogen generation under accident scenarios due to the oxidation of zirconium-based alloys widely used as fuel rod cladding in water-cooled reactors. This problem promoted research programs aiming the development of accident tolerant fuels (ATF) which are fuels that under accident conditions could keep longer its integrity enabling the mitigation of the accident effects. In the framework of the ATF program, different materials have been studied to be applied as cladding to replace zirconium-based alloy; also efforts have been made to improve the uranium dioxide thermal conductivity doping the fuel pellet. This paper evaluates the addition of beryllium oxide (BeO) to the uranium dioxide in order to enhance the thermal conductivity of the fuel pellet. Investigations performed in this area considering the addition of 10% in volume of BeO, resulting in the UO{sub 2}-BeO fuel, have shown good results with the improvement of the fuel thermal conductivity and the consequent reduction of the fuel temperatures under irradiation. In this paper, two models obtained from open literature for the thermal conductivity of UO{sub 2}- BeO fuel were implemented in the FRAPCON 3.5 code and the results obtained using the modified code versions were compared. The simulations were carried out using a case available in the code documentation related to a typical pressurized water reactor (PWR) fuel rod irradiated under steady state condition. The results show that the fuel centerline temperatures decrease with the addition of BeO, when compared to the conventional UO{sub 2} pellet, independent of the model applied. (author)

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  14. Monte Carlo criticality analysis of simple geometries containing tungsten-rhenium alloys engrained with uranium dioxide and uranium mononitride

    International Nuclear Information System (INIS)

    Webb, Jonathan A.; Charit, Indrajit

    2011-01-01

    Highlights: → The addition of rhenium to the tungsten matrix within W-UO 2 and W-UN CERMET materials can help reduce the risk of submersion criticality accidents while increasing the strength and ductility of tungsten based nuclear fuel elements. → The addition of rhenium up to 30 at.% to simple geometries containing W-UO 2 mixtures can increase the critical mass by 65 kg. → The addition of rhenium up to 30 at.% to simple geometries containing W-UN mixtures can increase the critical mass by 22 kg. → The addition of rhenium by up to 30 at.% to simple geometries containing W-UO 2 mixtures can reduce the change in reactivity change due to water submersion by $5.07. → The addition of rhenium by up to 30 at.% to simple geometries containing W-UN mixtures can reduce the change in reactivity due to water submersion by $3.24. - Abstract: The critical mass and dimensions of simple geometries containing highly enriched uranium dioxide (UO 2 ) and uranium mononitride (UN) encapsulated in tungsten-rhenium alloys are determined using MCNP5 criticality calculations. Spheres as well as cylinders with length to radius ratios of 1.82 are computationally built to consist of 60 vol.% fuel and 40 vol.% metal matrix. Within the geometries, the uranium is enriched to 93 wt.% uranium-235 and the rhenium content within the metal alloy was modeled over the range of 0-30 at.%. The spheres containing UO 2 were determined to have a critical radius of 18.29-19.11 cm and a critical mass ranging from 366 kg to 424 kg. The cylinders containing UO 2 were found to have a critical radius ranging from 17.07 cm to 17.84 cm with a corresponding critical mass of 406-471 kg. Spheres engrained with UN were determined to have a critical radius ranging from 14.82 cm to 15.19 cm and a critical mass between 222 kg and 242 kg. Cylinders which were engrained with UN were determined to have a critical radius ranging from 13.81 cm to 14.15 cm and a corresponding critical mass of 245-267 kg. The critical

  15. Results of the analysis of the intercomparison samples of the depleted uranium dioxide SR-20

    International Nuclear Information System (INIS)

    Aigner, H.; Deron, S.; Kuhn, E.; Ronesch, K.; Zoigner, A.

    Samples of a homogeneous powder of depleted uranium dioxide, SR-20, were distributed to 32 laboratories in January 1980 for intercomparison of the precisions and accuracies of wet chemical assay. 11 laboratories reported their results (ANNEX 1). 5 laboratories applied titration procedures, 4 of them applied methods derived from the Davies and Gray procedure (1), 2 laboratories used controlled potential coulometry, 2 laboratories used precipitation procedures, 1 laboratory used fluorimetry and 1 laboratory used activation analysis. An analysis of variance yields for each laboratory the estimates of the measurement errors, the dissolution or treatment errors and the random calibration errors. The measurement errors vary between 0.01% and 1.7% relative. The differences to the reference value vary between -9.1% and +0.92% uranium, but 9 laboratories agree within +-1%U with the reference value. The mean bias of these 9 laboratories is equal to +0.04%U. The standard deviation of the biases of these 9 laboratories is equal to 0.36%.U

  16. Heat processing of gels into sintered uranium dioxide modelled by thermal analysis. I

    International Nuclear Information System (INIS)

    Landspersky, H.; Urbanek, V.

    1979-01-01

    Thermoanalytical methods were used for investigating the processes of air drying and calcination of gels prepared by internal gelation of uranyl nitrate, urea and urotropine solutions at 90 degC. The gels were dried in air at room temperature, at 220 degC in a controlled atmosphere or by azeotropic distillation with CCl 4 . The course of thermal decomposition of the gel depends not only on the drying method used but also on the medium in which the drying process takes place. If the drying is carried out so as to produce a macroporous structure after the elimination of most of the water, ammonia and possibly other gelation by-products and non-reacted gelating agents, the resulting gels can be further processed by calcination, reduction and sintering, thus obtaining compact undamaged spheres of sintered uranium dioxide. Dilatometric analysis generated of uranium trioxide gels showed that the transformation of UO 3 to U 3 O 8 generated another intermediate thermal decomposition product showing a change in dimensions at temperatures of about 520 degC and a change in colour. This phenomenon is analogous to the decomposition of UO 3 prepared by thermal decomposition of α-UO 3 .2H 2 O involving a change in weight producing the UOsub(3-x) compound or a phase transformation with a change in colour; the structural conversion cannot be identified by X-ray structural analysis. (author)

  17. Results of the analysis of the intercomparison samples of the depleted uranium dioxide SR-10

    International Nuclear Information System (INIS)

    Aigner, H.; Deron, S.; Kuhn, E.; Zoigner, A.

    1981-01-01

    Samples of a homogeneous powder of depleted uranium dioxide, SR-10, were distributed to 27 laboratories in February 1979 for intercomparison of the precisions and accuracies of wet chemical assay. 7 laboratories reported their results. 6 laboratories applied titration procedures, 4 of them applied methods derived from the Davies and Gray procedure (1), and one laboratory used controlled potential coulometry. An analysis of variance yields for each laboratory the estimates of the measurement errors, the dissolution or treatment errors and the random calibration errors. The measurement errors vary between 0.01% and 0.10% relative. The differences to the reference value vary between -0.48% and +0.87% uranium, but 5 laboratories agree within +-0.25% U with the reference value. The biases of 5 laboratories are greater than expected from their random errors. The mean bias of the 7 laboratories is equal to +0.03% U. The standard deviation of the laboratory biases is equal to 0.43% U. (author)

  18. Investigation of high burnup structures in uranium dioxide applying cellular automata: algorithms and codes

    International Nuclear Information System (INIS)

    Akishina, E.P.; Kostenko, B.F.; Ivanov, V.V.

    2003-01-01

    A new method of research in spatial structures that result from uranium dioxide burning in nuclear reactors of modern atomic plants is suggested. The method is based on the presentation of images of the mentioned structures in the form of the working field of a cellular automaton (CA). First, it has allowed one to extract some important quantitative characteristics of the structures directly from the micrographs of the uranium fuel surface. Secondly, the CA has been found out to allow one to formulate easily the dynamics of the evolution of the studied structures in terms of such micrograph elements as spots, spots' boundaries, cracks, etc. Relation has been found between the dynamics and some exactly solvable models of the theory of cellular automata, in particular, the Ising model and the vote model. This investigation gives a detailed description of some CA algorithms which allow one to perform the fuel surface image processing and to model its evolution caused by burnup or chemical etching. (author)

  19. Research on calculation of mixing fraction for natural uranium equivalent fuel

    International Nuclear Information System (INIS)

    Huang Shien; Wang Lianjie; Wei Yanqin; Li Qing; Zheng Jiye

    2013-01-01

    Based on the first-order perturbation theory and reasonable approximations, the calculation method of recycled uranium (RU) and depleted uranium (DU) mixing fraction for natural uranium equivalent (NUE) fuel was studied, so the equivalence between NUE fuel and natural uranium (NU) fuel was assured. The adopted calculation method accurately takes the variation of micro cross sections alone with fuel depletion into account. A computer code named ALPHA was programmed to execute the calculation procedure. Then the ALPHA code and the WIMS-AECL code compose a processing system, which is applicable to the mixing fraction calculation for heavy water reactor NUE fuel. The validation shows that the processing system can accurately calculate the mixing fraction for NUE fuel. (authors)

  20. Study of Physical modifications induced by chromium doping of uranium dioxide

    International Nuclear Information System (INIS)

    Fraczkiewicz, M.

    2010-01-01

    Improvement of nuclear fuel performances requires reducing fission gas release. Doping uranium dioxide with chromium is the improvement axis considered in this work. Indeed, chromium fastens crystal growth in UO 2 , and thus enables a significant increase of the grain size. This work aims at the identification of defects produced by chromium addition in UO 2 , and their impact on properties of interest of the material. First, defects existing in doped fuel directly after sintering have been studied. X-ray Absorption Spectroscopy allowed the identification of the environment of solubilised chromium in UO 2 . Chromium atoms are roughly substituting for uranium atoms, but generate a complete reorganisation of neighbouring oxygen atoms, and distortion of uranium sublattice. Characterisation of transport properties (electrical conductivity and oxygen self-diffusion) have shown that because of charge balance, chromium plays a leading role on such properties. A model of point defects in UO 2 has been proposed, showing how complex the involved phenomena are. Observations by Transmission Electron Microscopy of ion-irradiated thin foils have shown that chromium makes the coalescence of irradiation defects easier. This behaviour can be explained by a stabilisation of defect clusters due to precipitation of chromium. Finally, study of thermal diffusion of helium in doped UO 2 , performed by Nuclear Reaction Analysis, has confirmed this interaction between chromium atoms and irradiation defects. Indeed, μ-NRA measures have shown no fast gas diffusion close to grain boundaries, in contrast with standard UO 2 behaviour, which is associated with defects recovery in grain boundaries. (author) [fr

  1. Viscoplastic behavior of uranium dioxide at high temperature; Comportement viscoplastique du dioxyde d'uranium a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Sauter, F

    2001-02-01

    This work is a part of a project led by EDF the purpose of which is the development of more predictive models to describe the thermomechanical behavior of fuel assembly. First, we recall the baselines of the Power Water Reactors then we deal with the viscoplastic behavior of uranium dioxide (UO{sub 2}). This knowledge enables an accurate description of the stress relaxation during Pellet Cladding Interactions. The pellets we have used in the last part are similar to the industrial ones. They exhibit a yield point during strain hardening tests and a sigma creep curve. In order to describe these characteristics, we have adapted different kind of approaches: thermodynamical - the Distribution of Non Linear Relaxations, approaches based on dislocation glide inspired by Alexander and Haasen and introduced in the Pilvin polycrystalline model. We recall the purpose of internal variables in the thermodynamics of system far from equilibrium then in case of a viscoplastic flow controlled by dislocation glide, we establish a link between densities of dislocations and internal variables in the D.N.L.R. approach. As vacancy diffusion in the grain boundary has a contribution to the viscoplastic strain, a similar is presented in appendix. These models are able to reproduce the behavior of UO{sub 2} pellets in strain hardening, stress relaxation and creep tests. Much possible progress has been revealed by the analysis of the tests. Further more, we propose a model for yield point and sigma creep curve. We also have extended these results to the behavior of irradiated pellets and stressed the influence of damage. (author)

  2. Effect of chloride concentration on the solubility of amorphous uranium dioxide at 25deg C under reducing conditions

    International Nuclear Information System (INIS)

    Aguilar, M.; Casas, I.; Pablo, J. de; Torrero, M.E.

    1991-01-01

    The dependence of the solubility of a microcrystalline uranium dioxide on the chloride concentration has been studied at 25deg C under reducing conditions. The concentration of uranium in solution has been found to be some orders of magnitude lower than in perchlorate media. Possible changes of both the morphology and the composition of the solid phase have been investigated by means of Energy Dispersive X-ray Analysis (EDX) and X-ray Powder Difraction (XPD). The formation of a secondary solid phase as a reason for the decrease of the solubility has been postulated. (orig.)

  3. Reactions of plutonium dioxide with water and oxygen-hydrogen mixtures: Mechanisms for corrosion of uranium and plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Haschke, John M.; Allen, Thomas H.; Morales, Luis A.

    1999-06-18

    Investigation of the interactions of plutonium dioxide with water vapor and with an oxygen-hydrogen mixture show that the oxide is both chemically reactive and catalytically active. Correspondence of the chemical behavior with that for oxidation of uranium in moist air suggests that similar catalytic processes participate in the mechanism of moisture-enhanced corrosion of uranium and plutonium. Evaluation of chemical and kinetic data for corrosion of the metals leads to a comprehensive mechanism for corrosion in dry air, water vapor, and moist air. Results are applied in confirming that the corrosion rate of Pu in water vapor decreases sharply between 100 and 200 degrees C.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lecraz, C

    1993-06-11

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

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2001-01-01

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

  6. Assessment of current atomic scale modelling methods for the investigation of nuclear fuels under irradiation: Example of uranium dioxide

    International Nuclear Information System (INIS)

    Bertolus, M.; Freyss, M.; Krack, M.; Devanathan, R.

    2015-01-01

    We focus here on the assessment of the description of interatomic interactions in uranium dioxide using, on the one hand, electronic structure methods, in particular in the Density Functional Theory (DFT) framework, and on the other hand, empirical potential methods. These two types of methods are complementary, the former enabling results to be obtained from a minimal amount of input data and further insight into the electronic and magnetic properties to be achieved, while the latter are irreplaceable for studies where a large number of atoms need to be considered. We consider basic properties as well as specific ones, which are important for the description of nuclear fuel under irradiation. These are especially energies, which are the main data passed on to higher scale models. For this exercise, we limit ourselves to uranium dioxide (UO 2 ) because of the extensive amount of studies available on this system. (authors)

  7. Kinetic of formation for single carbon dioxide and mixed carbon dioxide and tetrahydrofuran hydrates in water and sodium chloride aqueous solution

    NARCIS (Netherlands)

    Sabil, K.M.; Duarte, A.R.C.; Zevenbergen, J.F.; Ahmad, M.M.; Yusup, S.; Omar, A.A.; Peters, C.J.

    2010-01-01

    A laboratory-scale reactor system is built and operated to measure the kinetic of formation for single and mixed carbon dioxide-tetrahydrofuran hydrates. The T-cycle method, which is used to collect the kinetic data, is briefly discussed. For single carbon dioxide hydrate, the induction time

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

    Science.gov (United States)

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

    2011-11-01

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

  9. Fabrication of uranium dioxide of different granulation from uranyl nitrate by ammonia diuranate; Dobijanje urandioksida razlicitih granulacija iz uranilnitrata preko amonijumdiuranata

    Energy Technology Data Exchange (ETDEWEB)

    Vojnovic, J; Stamenkovic, I [Institute of Nuclear Sciences Boris Kidric, Laboratorija za termotehniku reaktora, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    Uranium dioxide is most frequently produced by reduction of higher oxides (UO{sub 3}, U{sub 3}O{sub 8}) or reduction of uranium salts (uranium diuranate, uranium peroxide, uranyl oxalate). Reduction is most frequently done in hydrogen or carbon monoxide atmosphere under temperatures from 500 - 1700 deg C. One of the most frequently methods for producing uranium oxide is certainly reduction of ammonia diuranate by hydrogen (ADU method). Properties of uranium dioxide obtained by ADU method depend on properties of the initial substance. Investigations shown in this report are concerned with determining the properties of UO{sub 2} powders for determining the connection between their properties and conditions of fabrication and reduction of ADU and U{sub 3}O{sub 8}.

  10. Xenon Defects in Uranium Dioxide From First Principles and Interatomic Potentials

    Science.gov (United States)

    Thompson, Alexander

    In this thesis, we examine the defect energetics and migration energies of xenon atoms in uranium dioxide (UO2) from first principles and interatomic potentials. We also parameterize new, accurate interatomic potentials for xenon and uranium dioxide. To achieve accurate energetics and provide a foundation for subsequent calculations, we address difficulties in finding consistent energetics within Hubbard U corrected density functional theory (DFT+U). We propose a method of slowly ramping the U parameter in order to guide the calculation into low energy orbital occupations. We find that this method is successful for a variety of materials. We then examine the defect energetics of several noble gas atoms in UO2 for several different defect sites. We show that the energy to incorporate large noble gas atoms into interstitial sites is so large that it is energetically favorable for a Schottky defect cluster to be created to relieve the strain. We find that, thermodynamically, xenon will rarely ever be in the interstitial site of UO2. To study larger defects associated with the migration of xenon in UO 2, we turn to interatomic potentials. We benchmark several previously published potentials against DFT+U defect energetics and migration barriers. Using a combination of molecular dynamics and nudged elastic band calculations, we find a new, low energy migration pathway for xenon in UO2. We create a new potential for xenon that yields accurate defect energetics. We fit this new potential with a method we call Iterative Potential Refinement that parameterizes potentials to first principles data via a genetic algorithm. The potential finds accurate energetics for defects with relatively low amounts of strain (xenon in defect clusters). It is important to find accurate energetics for these sorts of low-strain defects because they essentially represent small xenon bubbles. Finally, we parameterize a new UO2 potential that simultaneously yields accurate vibrational properties

  11. Method and device to produce pourable, directly pressable uranium dioxide powder. Verfahren und Vorrichtung zur Herstellung von rieselfaehigem, direkt verpressbarem Urandioxid-Pulver

    Energy Technology Data Exchange (ETDEWEB)

    Boerner, P.; Isensee, H.J.; Vietzke, H.

    1978-08-17

    The uranium dioxide powder is produced from uranium peroxide which is obtained by continuous precipitation of uranyl nitrate solutions. By varying the precipitation conditions, one can exactly adjust the desired properties of the UO/sub 2/ powder, there is no 'post sintering'. The individual process steps are shown in detail.

  12. Behaviour of uranium during mixing in the Delaware and Chesapeake estuaries

    International Nuclear Information System (INIS)

    Sarin, M.M.; Church, T.M.

    1994-01-01

    Unequivocal evidence is presented for the removal of uranium in two major estuarine systems of the north-eastern United States: the Delaware and Chesapeake Bays. In both the estuaries, during all seasons but mostly in summer, dissolved uranium shows distinctly non-conservative behaviour at salinities ≤ 5. At salinities above 5, there are no deviations from the ideal dilution line. In these two estuaries as much as 22% of dissolved uranium is removed at low salinities, around salinity 2. This pronounced removal of uranium observed at low salinities has been investigated in terms of other chemical properties measured in the Delaware Estuary. In the zone of uranium removal, dissolved oxygen is significantly depleted and pH goes through a minimum down to 6.8. In the same low salinity regime, total alkalinity shows negative deviation from the linear dilution line and phosphate is removed. Humic acids, dissolved iron and manganese are also rapidly removed during estuarine mixing in this low salinity region. Thus, it appears that removal of uranium is most likely related to those properties of alkalinity and acid-base system of the upper estuary that may destabilize the uranium-carbonate complex. Under these conditions, uranium may associate strongly with phosphates or humic substances and be removed onto particulate phases and deposited within upper estuarine sediments. (author)

  13. Accountability methods for plutonium and uranium: the NRC manuals

    Energy Technology Data Exchange (ETDEWEB)

    Gutmacher, R.G.; Stephens, F.B.

    1977-09-28

    Four manuals containing methods for the accountability of plutonium nitrate solutions, plutonium dioxide, uranium dioxide and mixed uranium-plutonium oxide have been prepared by us and issued by the U.S. Nuclear Regulatory Commission. A similar manual on methods for the accountability of uranium and plutonium in reprocessing plant dissolver solutions is now in preparation. In the present paper, we discuss the contents of the previously issued manuals and give a preview of the manual now being prepared.

  14. Accountability methods for plutonium and uranium: the NRC manuals

    International Nuclear Information System (INIS)

    Gutmacher, R.G.; Stephens, F.B.

    1977-01-01

    Four manuals containing methods for the accountability of plutonium nitrate solutions, plutonium dioxide, uranium dioxide and mixed uranium-plutonium oxide have been prepared by us and issued by the U.S. Nuclear Regulatory Commission. A similar manual on methods for the accountability of uranium and plutonium in reprocessing plant dissolver solutions is now in preparation. In the present paper, we discuss the contents of the previously issued manuals and give a preview of the manual now being prepared

  15. Detection of carbon dioxide in the gases evolved during the hot extraction determination of hydrogen in uranium ingots

    International Nuclear Information System (INIS)

    Jursik, M.L.; Pope, J.D.

    1977-08-01

    The hot extraction method was used at the National Lead Company of Ohio to determine hydrogen in uranium metal at the 2 ppM level. The volume of gas evolved from the heated sample was assumed to be hydrogen. When a liquid nitrogen trap was placed into the system the hydrogen values were reduced 5 to 10%. The gas retained by the nitrogen trap was identified by mass spectrometry as predominantly carbon dioxide. Low hydrogen values were observed only when the nitrogen trap was used in the analysis of high-carbon (300 to 600 ppM) uranium from NLO production ingots. However, hydrogen values for low-carbon (30 to 50 ppM) uranium were unaffected by the nitrogen trap. The formation of carbon dioxide appears to be associated with the carbon content of the uranium metal. Comparisons of hydrogen values obtained with the hot extraction method and with an inert fusion--thermal conductivity method are also presented. 3 tables, 4 figures

  16. Quantitative analysis of occluded gases in uranium dioxide pellets by the mass spectrometry technique

    International Nuclear Information System (INIS)

    Vega Bustillos, J.O.W.; Rodrigues, C.; Iyer, S.S.

    1981-05-01

    A quantitative analysis of different components of occluded gases except water in uranium dioxide pellets is attempted here. A high temperature vacuum extration system is employed for the liberation and the determination of total volume of the occluded gases. A mass spectrometric technique is employed for the qualitative and quantitative analysis of these gases. The UO 2 pellets are placed in a graphite crucible and are subjected to varing temperatures (1000 0 C - 1700 0 C). The liberated gases are dehydrated and transferred to a measuring unit consisting essentially of a Toepler pump and a McLeod gauge. In this system the total volume of the gases liberated at N. T. P. is determined with a sensitivity of 0.002 cm 3 /g of UO 2 . An aliquot of the liberated gas is introduced into a quadrupole mass spectrometer (VGA-100 Varian Corp.) for the determination of the different components of the gas. On the basis of the analysis suggestions are made for the possible sources of these gas components. (Author) [pt

  17. Critical experiments simulating accidental water immersion of highly enriched uranium dioxide fuel elements

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoi, N.N.; Glushkov, L.S.

    2003-01-01

    The paper focuses on experimental analysis of nuclear criticality safety at accidental water immersion of fuel elements of the Russian TOPAZ-2 space nuclear power system reactor. The structure of water-moderated heterogeneous critical assemblies at the NARCISS facility is described in detail, including sizes, compositions, densities of materials of the main assembly components for various core configurations. Critical parameters of the assemblies measured for varying number of fuel elements, height of fuel material in fuel elements and their arrangement in the water moderator with a uniform or variable spacing are presented. It has been found from the experiments that at accidental water immersion of fuel elements involved, the minimum critical mass equal to approximately 20 kg of uranium dioxide is achieved at 31-37 fuel elements. The paper gives an example of a physical model of the water-moderated heterogeneous critical assembly with a detailed characterization of its main components that can be used for calculations using different neutronic codes, including Monte Carlo ones. (author)

  18. Uranium dioxide thermal characterization by the flash laser method from 23 Celsius to 175 Celsius

    International Nuclear Information System (INIS)

    Faeda, K.C.M.; Lameiras, F.S.; Carneiro, L.S.S.; Camarano, D.M.; Ferreira, R.A.N.

    2010-01-01

    The Laser Flash Method has become one of the most common techniques for measuring thermal diffusivity and conductivity in solids and liquids. This method is recognized by INMETRO as standard to be used in Brazil for measuring thermophysical properties of materials, such as metals, carbon composites, ceramics, and also nuclear materials. This article describes the experimental bench of the LMPT-Laboratorio de Medicao de Propriedades Termofisicas de Combustiveis Nucleares e Materiais of the CDTN-Centro de Desenvolvimento da Tecnologia Nuclear, (LMPT), as well as the mathematical model developed based on this method. The obtained results for the thermal diffusivity and for the thermal conductivity of uranium dioxide (U0 2 ) pellets in the temperature range from 25 deg to 175 deg C, are discussed and compared with the literature data. The estimative of the input quantities uncertainty of the mathematical model was determined according to ISO - BIPM-Guide to the Expression of Uncertainty in Measurement and the Monte Carlo Method was used to estimate of the output quantities uncertainty (thermal diffusivity and thermal conductivity). Additionally the results of the x-rays of these pellets are presented. (author)

  19. Cask size and weight reduction through the use of depleted uranium dioxide-concrete material

    International Nuclear Information System (INIS)

    Lobach, S.Yu.; Haire, J.M.

    2007-01-01

    Newly developed depleted uranium (DU) composite materials enable fabrication of spent nuclear fuel (SNF) transport and storage casks that are smaller and lighter in weight than casks made with conventional materials. One such material is DU dioxide (DUO2)-concrete, so-called DUCRETE TM . This paper examines the radiation shielding efficiency of DUCRETE as compared with that of a conventional concrete cask that holds 32 pressurized-water-reactor SNF assemblies. In this analysis, conventional concrete shielding material is replaced with DUCRETE. The thickness of the DUCRETE shielding is adjusted to give the same radiation surface dose, 200 mrem/h (2 mSv/hr), as the conventional concrete cask. It was found that the concrete shielding thickness decreased from 71 to 20 cm and that the cask radial cross-section shielding area was reduced approx 50 %. The weight was reduced approx 21 %, from 154 to approx 127 tons. Should one choose to add an extra outer ring of SNF assemblies, the number of such assemblies would increase from 32 to 52. In this case, the outside cask diameter would still decrease, from 169 to 137 cm. However, the weight would increase somewhat from 156 to 177 tons. Neutron cask surface dose is only approx 10 % of the gamma dose. These reduced sizes and weights will significantly influence the design of next-generation SNF casks

  20. Properties of raw materials and intermediate products in the production of uranium dioxide sintered tablets

    International Nuclear Information System (INIS)

    Landspersky, H.; Vanecek, I.; Podest, M.

    1977-01-01

    The properties are described of ammonium polyuranate and of powder uranium dioxide. Ammonium polyuranate, an intermediate product, is prepared by filtering the precipitate from uranyl nitrate solution precipitation, this either by an ammonia aqueous solution from a uranyl nitrate aqueous solution or by direct U 6+ precipitation from a TBP kerosene solution by aqueous concentrated ammonia. With relation to further processing, the major properties of the intermediate product include grain size, shape and appearance of crystallites, structure and thermal decomposition. These properties affect the properties of UO 2 , the following intermediate product obtained by reduction of ammonium polyuranate. Powder UO 2 is the final intermediate product; high-compacted UO 2 pellets are manufactured from it by compacting and sintering. The final product properties are affected by the following parameters: specific surface, grain size and shape, U/O ratio and compactibility. The effect of and the techniques of determining these parameters are shown. The necessity is emphasised of studying the properties of powder ammonium polyuranate because changes in its production technology affect the properties of further products. (J.P.)

  1. Study of rare gases behavior in uranium dioxide: diffusion and bubble nucleation and growth mechanisms

    International Nuclear Information System (INIS)

    Michel, A.

    2011-01-01

    During in-reactor irradiation of the nuclear fuel, fission gases, mainly xenon and krypton, are generated that are subject to several phenomena: diffusion and precipitation. These phenomena can have adverse consequences on the fuel physical and chemical properties and its in-reactor behavior. The purpose of this work is to better understand the behavior of fission gases by identifying diffusion, bubble nucleation and growth mechanisms. To do this, studies involving separate effects have been established coupling ion irradiations/implantations with fine characterizations on Large Scale Facilities. The influence of several parameters such as gas type, concentration and temperature has been identified separately. Interpretation of the Thermal Desorption Spectrometry (TDS) measurements has enabled us to determine xenon and krypton diffusion coefficients in uranium dioxide. A heterogeneous nucleation mechanism on defects was determined by means of experiments on the JANNuS platform in Orsay that consists of a coupling of an implantor, an accelerator and a Transmission Electron Microscope (TEM). Finally, TEM and X-ray Absorption Spectroscopy characterizations of implanted and annealed samples put in relieve a bubble growth mechanism by atoms and vacancies capture. (author) [fr

  2. Measurement of uranium dioxide thermophysical properties by the laser flash method

    International Nuclear Information System (INIS)

    Grossi, Pablo Andrade; Ferreira, Ricardo Alberto Neto; Camarano, Denise das Merces; Andrade, Roberto Marcio de

    2009-01-01

    The evaluation of the thermophysical properties of uranium dioxide (UO 2 ), including a reliable uncertainty assessment, are required by the nuclear reactor design. These important information are used by thermohydraulic codes to define operational aspects and to assure the safety, when analyzing various potential situations of accident. The laser flash method had become the most popular method to measure the thermophysical properties of materials. Despite its several advantages, some experimental obstacles have been found due to the difficulty to obtain experimentally the ideals initial and boundary conditions required by the original method. An experimental apparatus and a methodology for estimating uncertainties of thermal diffusivity, thermal conductivity and specific heat measurements based on the laser flash method are presented. A stochastic thermal diffusion modeling has been developed and validated by standard samples. Inverse heat conduction problems (IHCPs) solved by finite volumes technique were applied to the measurement process with real initial and boundary conditions, and Monte Carlo Method was used for propagating the uncertainties. The main sources of uncertainty were due to: pulse time, laser power, thermal exchanges, absorptivity, emissivity, sample thickness, specific mass and dynamic influence of temperature measurement system. As results, mean values and uncertainties of thermal diffusivity, thermal conductivity and specific heat of UO 2 are presented. (author)

  3. Design of a uranium-dioxide powder spheroidization system by plasma processing

    Science.gov (United States)

    Cavender, Daniel

    The plasma spheroidization system (PSS) is the first process in the development of a tungsten-uranium dioxide (W-UO2) ceramic-metallic (cermet) fuel for nuclear thermal rocket (NTR) propulsion. For the purposes of fissile fuel retention, UO2 spheroids ranging in size from 50 - 100 micrometers (μm) in diameter will be encapsulated in a tungsten shell. The PSS produces spherical particles by melting angular stock particles in an argon-hydrogen plasma jet where they become spherical due to surface tension. Surrogate CeO 2 powder was used in place of UO2 for system and process parameter development. Stock and spheroidized powders were micrographed using optical and scanning electron microscopy and evaluated by statistical methods to characterize and compare the spherocity of pre and post process powders. Particle spherocity was determined by irregularity parameter. Processed powders showed a statistically significant improvement in spherocity, with greater that 60% of the examined particles having an irregularity parameter of equal to or lower than 1.2, compared to stock powder.

  4. Swelling and gas release of grain-boundary pores in uranium dioxide

    International Nuclear Information System (INIS)

    Schrire, D.I.

    1983-12-01

    The swelling and gas release of overpressured grain boundary pores is sintered unirradiated uranium dioxide were investigated under isothermal conditions. The pores became overpressured when the ambient pressure was reduced, and the excess pressure driving force caused growth and interconnection of the pores, leading to eventual gas release. Swelling was measured continuously by a linear variable differential transformer, and open and closed porosity fractions were determined after the tests by immersion density and quantitative microscopy measurements. The sinter porosity consisted of pores situated on grain faces, grain edges, and grain corners. Isolated pores maintained their equilibrium shape while growing, without any measurable change in dihedral angle. Interconnection occurred predominantly along grain edges, without any evidence of pore sharpening or crack propagation at low driving forces. Extensive open porosity occurred at a threshold density of about 85% TD. There was an almost linear dependence of the initial swelling rate on the driving force, with an activation energy of 200+- 8 kJ/mole, in good agreement with published values of the activation energy for grain boundary diffusion

  5. Computer simulation of structural modifications induced by highly energetic ions in uranium dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Sasajima, Y., E-mail: sasajima@mx.ibaraki.ac.jp [Department of Materials Science and Engineering, Faculty of Engineering, Ibaraki University, 4-12-1 Nakanarusawa, Hitachi 316-8511 (Japan); Frontier Research Center for Applied Atomic Sciences, Ibaraki University, Shirakata 162-4, Tokai 319-1106 (Japan); Osada, T. [Graduate School of Science and Engineering, Ibaraki University, 4-12-1 Nakanarusawa, Hitachi 316-8511 (Japan); Ishikawa, N. [Japan Atomic Energy Agency (JAEA), Shirakata Shirane 2-4, Tokai 319-1195 (Japan); Iwase, A. [Department of Materials Science, Osaka Prefecture University, Gakuen-cho 1-1, Sakai 599-8531 (Japan)

    2013-11-01

    The structural modification caused by the high-energy-ion irradiation of single-crystalline uranium dioxide was simulated by the molecular dynamics method. As the initial condition, high kinetic energy was supplied to the individual atoms within a cylindrical region of nanometer-order radius located in the center of the specimen. The potential proposed by Basak et al. [C.B. Basak, A.K. Sengupta, H.S. Kamath, J. Alloys Compd. 360 (2003) 210–216] was utilized to calculate interaction between atoms. The supplied kinetic energy was first spent to change the crystal structure into an amorphous one within a short period of about 0.3 ps, then it dissipated in the specimen. The amorphous track radius R{sub a} was determined as a function of the effective stopping power gS{sub e}, i.e., the kinetic energy of atoms per unit length created by ion irradiation (S{sub e}: electronic stopping power, g: energy transfer ratio from stopping power to lattice vibration energy). It was found that the relationship between R{sub a} and gS{sub e} follows the relation R{sub a}{sup 2}=aln(gS{sub e})+b. Compared to the case of Si and β-cristobalite single crystals, it was harder to produce amorphous track because of the long range interaction between U atoms.

  6. Computer simulation of structural modifications induced by highly energetic ions in uranium dioxide

    International Nuclear Information System (INIS)

    Sasajima, Y.; Osada, T.; Ishikawa, N.; Iwase, A.

    2013-01-01

    The structural modification caused by the high-energy-ion irradiation of single-crystalline uranium dioxide was simulated by the molecular dynamics method. As the initial condition, high kinetic energy was supplied to the individual atoms within a cylindrical region of nanometer-order radius located in the center of the specimen. The potential proposed by Basak et al. [C.B. Basak, A.K. Sengupta, H.S. Kamath, J. Alloys Compd. 360 (2003) 210–216] was utilized to calculate interaction between atoms. The supplied kinetic energy was first spent to change the crystal structure into an amorphous one within a short period of about 0.3 ps, then it dissipated in the specimen. The amorphous track radius R a was determined as a function of the effective stopping power gS e , i.e., the kinetic energy of atoms per unit length created by ion irradiation (S e : electronic stopping power, g: energy transfer ratio from stopping power to lattice vibration energy). It was found that the relationship between R a and gS e follows the relation R a 2 =aln(gS e )+b. Compared to the case of Si and β-cristobalite single crystals, it was harder to produce amorphous track because of the long range interaction between U atoms

  7. Results of Uranium Dioxide-Tungsten Irradiation Test and Post-Test Examination

    Science.gov (United States)

    Collins, J. F.; Debogdan, C. E.; Diianni, D. C.

    1973-01-01

    A uranium dioxide (UO2) fueled capsule was fabricated and irradiated in the NASA Plum Brook Reactor Facility. The capsule consisted of two bulk UO2 specimens clad with chemically vapor deposited tungsten (CVD W) 0.762 and 0.1016 cm (0.030-and 0.040-in.) thick, respectively. The second specimen with 0.1016-cm (0.040-in.) thick cladding was irradiated at temperature for 2607 hours, corresponding to an average burnup of 1.516 x 10 to the 20th power fissions/cu cm. Postirradiation examination showed distortion in the bottom end cap, failure of the weld joint, and fracture of the central vent tube. Diametral growth was 1.3 percent. No evidence of gross interaction between CVD tungsten or arc-cast tungsten cladding and the UO2 fuel was observed. Some of the fission gases passed from the fuel cavity to the gas surrounding the fuel specimen via the vent tube and possibly the end-cap weld failure. Whether the UO2 loss rates through the vent tube were within acceptable limits could not be determined in view of the end-cap weld failure.

  8. The influence of alkali metal impurities on the uranium dioxide hydrofluorination reaction

    International Nuclear Information System (INIS)

    Ponelis, A.A.

    1989-01-01

    The effect alkali metal impurities (sodium and potassium) in the uranium dioxide (UO 2 ) feed material have on the conversion to uraniumtetrafluoride (UF 4 ) was examined. A direct correlation exists between impurity level and sintering with concomitant reduced conversion. The sintering mechanism is attributable to decreased specific surface area. The typical 'die-off' of reaction or conversion can be explained in terms of increased particle growth rather than an arbitray zero porosity function. Hydrofluorination temperatures varied from 250 to 650 degrees C using pellets varying in size from 0.42 mm to 10 mm. Scanning electron microscope photographs show clearly the particle or grain growth in the pellet as well as the increased size with impurity level. A new dimensionless constant, N KP , is defined to facilitate explanation of the reaction as a function of pellet radius. N KP is defined as the ratio of pellet diffusion resistance to particle diffusion resistance of the reacting HF gas. At high values of this number (N KP >40) the conversion is limited to the outer periphery of the pellet while at low values (N KP KP at higher reaction temperatures which means that the particle diffusion resistance increases with increasing impurity level and results in easier sintering of these materials. 53 refs., 206 figs., 94 tabs

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

    Science.gov (United States)

    Saoudi, M.; Staicu, D.; Mouris, J.; Bergeron, A.; Hamilton, H.; Naji, M.; Freis, D.; Cologna, M.

    2018-03-01

    Thorium-uranium oxide pellets with high densities were prepared at the Canadian Nuclear Laboratories (CNL) by co-milling, pressing, and sintering at 2023 K, with UO2 mass contents of 0, 1.5, 3, 8, 13, 30, 60 and 100%. At the Joint Research Centre, Karlsruhe (JRC-Karlsruhe), thorium-uranium oxide pellets were prepared using the spark plasma sintering (SPS) technique with 79 and 93 wt. % UO2. The thermal diffusivity of (Th1-xUx)O2 (0 ≤ x ≤ 1) was measured at CNL and at JRC-Karlsruhe using the laser flash technique. ThO2 and (Th,U)O2 with 1.5, 3, 8 and 13 wt. % UO2 were found to be semi-transparent to the infrared wavelength of the laser and were coated with graphite for the thermal diffusivity measurements. This semi-transparency decreased with the addition of UO2 and was lost at about 30 wt. % of UO2 in ThO2. The thermal conductivity was deduced using the measured density and literature data for the specific heat capacity. The thermal conductivity for ThO2 is significantly higher than for UO2. The thermal conductivity of (Th,U)O2 decreases rapidly with increasing UO2 content, and for UO2 contents of 60% and higher, the conductivity of the thorium-uranium oxide fuel is close to UO2. As the mass difference between the Th and U atoms is small, the thermal conductivity decrease is attributed to the phonon scattering enhanced by lattice strain due to the introduction of uranium in ThO2 lattice. The new results were compared to the data available in the literature and were evaluated using the classical phonon transport model for oxide systems.

  10. Uranium

    International Nuclear Information System (INIS)

    Hamdoun, N.A.

    2007-01-01

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

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

    International Nuclear Information System (INIS)

    Hanus, D.; Kleykamp, H.

    1982-01-01

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

  12. The uranium dioxide-uranium system at high temperature; Le systeme uranium-dioxyde d'uranium a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Guinet, Ph.; Vaugoyeau, H.; Blum, P. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1966-07-01

    The liquidus curve has been determined by a saturation method in which the thermal gradient was cancelled upon cooling, and the solidus curve by analyzing the deposits in equilibrium with the liquid at each temperature. The diagram, of a displaced eutectic type, presents a liquid immiscibility domain between 47 and 59 mol per cent of dioxide and a substoichiometry range UO{sub 2x}, the minimum O/U ratio being 1,6 at 3470 {+-} 30 C. The monotectic composition was found by chemical analysis to be 59 mol per cent of dioxide and the reaction temperature 2470 {+-} 30 C. (author) [French] La courbe liquidus a ete determinee par une methode de saturation en annulant le gradient thermique au cours du refroidissement, la courbe solidus par analyse des depots en equilibre avec le liquide a chaque temperature. Le diagramme du type a eutectique deporte comporte un domaine d'immiscibilite liquide entre 47 et 59 moles pour cent de dioxyde, ainsi qu'un domaine d'existence du compose sous stoechiometrique UO{sub 2x}, le rapport O/U minimum etant egal a 1,6 a 2470 {+-} 30 C. La composition du monotectique, obtenue par analyse chimique, est de 59 moles pour cent de dioxyde et la temperature de la reaction a ete trouvee egale a 2470 {+-} 30 C. (auteur)

  13. Etching of uranium dioxide in nitrogen trifluoride RF plasma glow discharge

    Science.gov (United States)

    Veilleux, John Mark

    1999-10-01

    A series of room temperature, low pressure (10.8 to 40 Pa), low power (25 to 210 W) RF plasma glow discharge experiments with UO2 were conducted to demonstrate that plasma treatment is a viable method for decontaminating UO2 from stainless steel substrates. Experiments were conducted using NF3 gas to decontaminate depleted uranium dioxide from stainless-steel substrates. Results demonstrated that UO2 can be completely removed from stainless-steel substrates after several minutes processing at under 200 W. At 180 W and 32.7 Pa gas pressure, over 99% of all UO2 in the samples was removed in just 17 minutes. The initial etch rate in the experiments ranged from 0.2 to 7.4 mum/min. Etching increased with the plasma absorbed power and feed gas pressure in the range of 10.8 to 40 Pa. A different pressure effect on UO2 etching was also noted below 50 W in which etching increased up to a maximum pressure, ˜23 Pa, then decreased with further increases in pressure. A computer simulation, CHEMKIN, was applied to predict the NF3 plasma species in the experiments. The code was validated first by comparing its predictions of the NF3 plasma species with mass spectroscopy etching experiments of silicon. The code predictions were within +/-5% of the measured species concentrations. The F atom radicals were identified as the primary etchant species, diffusing from the bulk plasma to the UO2 surface and reacting to form a volatile UF6, which desorbed into the gas phase to be pumped away. Ions created in the plasma were too low in concentration to have a major effect on etching, but can enhance the etch rate by removing non-volatile reaction products blocking the reaction of F with UO2. The composition of these non-volatile products were determined based on thermodynamic analysis and the electronic structure of uranium. Analysis identified possible non-volatile products as the uranium fluorides, UF2-5, and certain uranium oxyfluorides UO2F, UO2F2, UOF3, and UOF 4 which form over the

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  16. The recovery of 99Mo from solutions of irradiated Uranium using a column with nanoparticles of Titanium Dioxide

    International Nuclear Information System (INIS)

    Androne, G. E.; Petre, M.; Lazar, C. G.

    2016-01-01

    Molyibdenum-99 (T½ = 66.02 h) decays by beta emission to 99 Tcm (T½ = 6.02 h). The latter nuclide is used in many nuclear medicine applications. The 99 Mo is produced from irradiated high (HEU) or low (LEU) enriched uranium. In this work a sensitive and selective method for recovering Mo from uranium solution, using a column with titanium dioxide nanoparticles, is developed. The titanium dioxide (TiO 2 ) nanoparticles were synthesized via sol-gel method using titanium tetra-chloride as starting material and urea as a reacting medium. A 40 ml uranium solution containing 450 g/L uranyl nitrate, 1 M HNO 3 , and 4 mg Mo was loaded on a column containing 6 g of TiO 2 sorbent at 75°C. After loading, the column was washed with 1 M HNO 3 and H 2 O. Mo was stripped from the column with 0.1 M NaOH at 25°C. The ICP-MS results indicate that 80-95% of the initial mass of Mo was loaded on the column, and 90-94% of this quantity was recovered in the strip fraction. (authors)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  18. A new characterization approach for studying relationships between microstructure and creep damage mechanisms of uranium dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Iltis, X., E-mail: xaviere.iltis@cea.fr [CEA, DEN, DEC, Cadarache, 13108 Saint-Paul-Lez-Durance (France); Ben Saada, M. [CEA, DEN, DEC, Cadarache, 13108 Saint-Paul-Lez-Durance (France); Laboratoire d' Etudes des Microstructures et de Mécanique des Matériaux (LEM3), CNRS UMR 7239, Université de Lorraine, Ile du Saulcy, 57045 Metz Cedex 1 (France); Mansour, H.; Gey, N.; Hazotte, A.; Maloufi, N. [Laboratoire d' Etudes des Microstructures et de Mécanique des Matériaux (LEM3), CNRS UMR 7239, Université de Lorraine, Ile du Saulcy, 57045 Metz Cedex 1 (France)

    2016-06-15

    Four batches of UO{sub 2} pellets were studied comparatively, before and after creep tests, to evaluate a characterization methodology aimed to determine the links between microstructure and damage mechanisms induced by compressive creep of uranium dioxide at 1500 °C. They were observed by means of scanning electron microscopy (SEM) coupled with image analysis, to quantify their fabrication porosity and the occurrence of inter-granular cavities after creep, and electron back scattered diffraction (EBSD), especially to characterize sub-structures development associated with plastic deformation. Electron channeling contrast imaging (ECCI) was also applied to evidence dislocations, at an exploratory stage, on one of the deformed pellets. This approach helped to identify and quantify microstructural differences between batches. Their as-fabricated microstructures differed in terms of grain size and fabrication porosity distribution. The pellets which had the lowest strain rates were those with the largest number of intra-granular pores, regardless of their grain size. They also exhibited less numerous sub-boundaries within the grains. These first results clearly illustrate the benefit of systematic examinations of crept UO{sub 2} pellets at a mesoscopic scale, by SEM and EBSD, to study their deformation process. In addition, ECCI appears as a powerful tool to evidence local dislocations arrangements, in bulk samples. Even if the sampling was limited, the results of this study also tend to indicate that the intra-granular pores population, resulting from the manufacturing of the samples by powder metallurgy, could have a significant influence on the UO{sub 2} viscoplastic deformation mechanisms. - Highlights: • Four different UO{sub 2} pellets batches are microstructurally compared, before and after compression creep tests. • Development of sub-boundaries within the original grains, in crept samples, is quantified by EBSD. • Links are observed between the intra

  19. Experimental study and kinetic modeling of the hydro-fluorination of uranium dioxide

    International Nuclear Information System (INIS)

    Pages, Simon

    2014-01-01

    A kinetic study of hydro-fluorination of uranium dioxide was performed between 375 and 475 C under partial pressures of HF between 42 and 720 mbar. The reaction was followed by thermogravimetry in isothermal and isobaric conditions. The kinetic data obtained coupled with a characterization of the powder before, during and after reaction by SEM, EDS, BET and XRD showed that the powder grains of UO 2 are transformed according a model of instantaneous germination, anisotropic growth and internal development. The rate limiting step of the growth process is the diffusion of HF in the UF 4 layer. A mechanism of growth of the UF 4 layer has been proposed. In the temperature and pressure range studied, the reaction is of first order with respect to HF and follows an Arrhenius law. A rate equation was determined and used to perform kinetic simulations which have shown a very good correlation with experience. Coupling of this rate equation with heat and mass transport phenomena allowed to perform simulations at the scale of a powder's agglomerate. They have shown that some structures of agglomerates influence the rate of diffusion of the gases in the porous medium and thereby influence the reaction rate. Finally kinetic simulations on powder's beds and pellets were carried out and compared with experimental rates. The experimental and simulated kinetic curves have the same paces, but improvements in the simulations are needed to accurately predict rates: the coupling between the three scales (grain, agglomerate, oven) would be a good example. (author) [fr

  20. Attempts at estimating mixed venous carbon dioxide tension by the single-breath method.

    Science.gov (United States)

    Ohta, H; Takatani, O; Matsuoka, T

    1989-01-01

    The single-breath method was originally proposed by Kim et al. [1] for estimating the blood carbon dioxide tension and cardiac output. Its reliability has not been proven. The present study was undertaken, using dogs, to compare the mixed venous carbon dioxide tension (PVCO2) calculated by the single-breath method with the PVCO2 measured in mixed venous blood, and to evaluate the influence of variations in the exhalation duration and the volume of expired air usually discarded from computations as the deadspace. Among the exhalation durations of 15, 30 and 45 s tested, the 15 s duration was found to be too short to obtain an analyzable O2-CO2 curve, but at either 30 or 45 s, the calculated values of PVCO2 were comparable to the measured PVCO2. A significant agreement between calculated and measured PVCO2 was obtained when the expired gas with PCO2 less than 22 Torr was considered as deadspace gas.

  1. Determination of radium and uranium isotopes in natural waters by sorption on hydrous manganese dioxide followed by alpha-spectrometry

    International Nuclear Information System (INIS)

    Bojanowski, R.; Radecki, Z.; Burns, K.

    2005-01-01

    Water samples, spiked with 133 Ba and 232 U radiotracers, are scavenged for radium and uranium isotopes using hydrous manganese dioxide which is produced in-situ, by reacting manganese (+2) and permanganate ions at pH 8-9. The precipitate is solubilized with ascorbic and acetic acids and the resulting solution filtered through a glass fibre filter GF/F to remove particulate matter. The radium is co-precipitated with barium ions by the addition of a saturated Na 2 SO 4 solution where a small amount of BaSO 4 suspension is introduced to initiate crystallization. The micro precipitate containing the radium is collected on a 0.1 membrane filter and the filtrate saved for follow-up uranium analysis. The 226 Ra on the filter is determined by alpha-spectrometry and its recovery is assessed by measuring the 133 Ba on the same filter using gamma-spectrometry. The filtrate containing uranium is passed through a Dowex AG 1 x 4 ion-exchange resin in the SO 4 2- form which retains uranium while other ions are eluted by dilute (0.25M) sulphuric acid. Uranium is eluted from the column by distilled water, electrodeposited on a silver disc and the uranium isotopes and their recovery are determined by alpha-spectrometry. The method was tested on a variety of natural and spiked water samples with known concentrations of 226 Ra and 238 U and was found to yield accurate results within ±10% RSD of the target values. (author)

  2. Uranium

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  3. Method of manufacturing mixed stock powders for nuclear fuel elements

    International Nuclear Information System (INIS)

    Hirayama, Satoshi.

    1980-01-01

    Purpose: To alleviate the limit of the present reactor operating conditions by uniformly mixing an additive to the main content as an uranium dioxide or mixture of the uranium dioxide with plutonium dioxide. Method: A mixed stock powder is obtained by adding an additive of at least two of aluminium oxide, beryllium oxide, calcium oxide, magnesium oxide, silicon oxide, sodium oxide, potassium oxide, phosphorus oxide, titanium oxide and iron oxide to suspension having ammonia water as dispersion medium to start the deposition of precipitation at a step of precipitating ammonium diuranate or plutionium hydroxide of a main content of uranium dioxide or mixture of uranium dioxide and plutonium dioxide and deposited precipitate is calcinated and reduced. (Yoshihara, H.)

  4. Report on in-situ studies of flash sintering of uranium dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Raftery, Alicia Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-24

    Flash sintering is a novel type of field assisted sintering that uses an electric field and current to provide densification of materials on very short time scales. The potential for field assisted sintering techniques to be used in producing nuclear fuel is gaining recognition due to the potential economic benefits and improvements in material properties. The flash sintering behavior has so far been linked to applied and material parameters, but the underlying mechanisms active during flash sintering have yet to be identified. This report summarizes the efforts to investigate flash sintering of uranium dioxide using dilatometer studies at Los Alamos National Laboratory and two separate sets of in-situ studies at Brookhaven National Laboratory’s NSLS-II XPD-1 beamline. The purpose of the dilatometer studies was to understand individual parameter (applied and material) effects on the flash behavior and the purpose of the in-situ studies was to better understand the mechanisms active during flash sintering. As far as applied parameters, it was found that stoichiometry, or oxygen-to-metal ratio, has a significant effect on the flash behavior (time to flash and speed of flash). Composite systems were found to have degraded sintering behavior relative to pure UO2. The critical field studies are complete for UO2.00 and will be analyzed against an existing model for comparison. The in-situ studies showed that the strength of the field and current are directly related to the sample temperature, with temperature-driven phase changes occurring at high values. The existence of an ‘incubation time’ has been questioned, due to a continuous change in lattice parameter values from the moment that the field is applied. Some results from the in-situ experiments, which should provide evidence regarding ion migration, are still being analyzed. Some preliminary conclusions can be made from these results with regard to using field assisted sintering to

  5. Determination of plutonium and uranium in mixed nuclear fuel by means of potentiostatic and amperostatic coulometry

    International Nuclear Information System (INIS)

    Kuperman, A.Ya.; Moiseev, I.V.; Galkina, V.N.; Yakushina, G.S.; Nikitskaya, V.N.

    1977-01-01

    Product solution occurs in HClO 4 + HNO 3 mixing. In prepared plutonium (6) and uranium (6) perchloric acid solution Cl and Cr (6), Mn (7,6,3) foreign oxidizers are selectively reduced with formic and malonic acids. Potentiostatic variant of method is based on successive reduction of Pu(6) to Pu(3) and U(6) to U(4) in 4.5M HCl, containing 5x10 -4 M bismuth (3). In using amperostatic variant of method plutonium and uranium are determined separately. In sulfur-phosphoric acid media plutonium (6) is titrated to Pu(4) with continuously generated iron (2) ions. Uranium (6) in phosphoric acid media is initially reduced to U(4) with Fe(2), and then after Fe(2) excess reduction with nitric acid it is titrated to uranium (6) with continuously electrogenerated manganese (3) ions or vanadium (5). To obtain equivalent point in plutonium (6) and uranium (4) titration amperometric method is used. Coefficient of variation is 0.2-0.3 % rel

  6. Electronic structure of the actinides and their dioxides. Application to the defect formation energy and krypton solubility in uranium dioxide; Etude de la structure electronique des actinides et de leurs dioxydes. Application aux defauts ponctuels et aux gaz de fission dans le dioxyde d`uranium

    Energy Technology Data Exchange (ETDEWEB)

    Petit, T. [CEA Centre d`Etudes Nucleaires de Grenoble, 38 (France)]|[CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Thermohydraulique et de Physique

    1996-09-28

    Uranium dioxide is the standard nuclear fuel used in French h power plants. During irradiation, fission products such as krypton and xenon are created inside fuel pellets. So, gas release could become, at very high burnup, a limiting factor in the reactor exploitation. To study this subject, we have realised calculations using the Density Functional Theory (DFT) into the Local Density Approximation (LDA) and the Atomic Sphere Approximation (ASA). First, we have validated our approach by calculating cohesive properties of thorium, protactinium and uranium metals. The good agreement between our results and experimental values implies that 5f electrons are itinerant. Calculated lattice parameter, cohesive energy and bulk modulus for uranium and thorium dioxides are in very good agreement with experiment. We show that binding between uranium and oxygen atoms is not completely ionic but partially covalent. The question of the electrical conductivity still remains an open problem. We have been able to calculate punctual defect formation energies in uranium dioxide. Accordingly to experimental observations, we find that it is easier to create a defect in the oxygen sublattice than in the uranium sublattice. Finally, we have been able to predict a probable site of krypton atoms in nuclear fuel: the Schottky trio. Experiences of Extended X-ray Absorption Fine structure Spectroscopy (EXAFS) and X-ray Photoelectron Spectroscopy (XPS) on uranium dioxide doped by ionic implantation will help us in the comprehension of the studied phenomena and the interpretation of our calculations. (author). 256 refs.

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

    International Nuclear Information System (INIS)

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

    1997-11-01

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

  8. Beryllium Project: developing in CDTN of uranium dioxide fuel pellets with addition of beryllium oxide to increase the thermal conductivity

    International Nuclear Information System (INIS)

    Ferreira, Ricardo Alberto Neto; Camarano, Denise das Merces; Miranda, Odair; Grossi, Pablo Andrade; Andrade, Antonio Santos; Queiroz, Carolinne Mol; Gonzaga, Mariana de Carvalho Leal

    2013-01-01

    Although the nuclear fuel currently based on pellets of uranium dioxide be very safe and stable, the biggest problem is that this material is not a good conductor of heat. This results in an elevated temperature gradient between the center and its lateral surface, which leads to a premature degradation of the fuel, which restricts the performance of the reactor, being necessary to change the fuel before its full utilization. An increase of only 5 to 10 percent in its thermal conductivity, would be a significant increase. An increase of 50 percent would be a great improvement. A project entitled 'Beryllium Project' was developed in CDTN - Centro de Desenvolvimento da Tecnologia Nuclear, which aimed to develop fuel pellets made from a mixture of uranium dioxide microspheres and beryllium oxide powder to obtain a better heat conductor phase, filling the voids between the microspheres to increase the thermal conductivity of the pellet. Increases in the thermal conductivity in the range of 8.6% to 125%, depending on the level of addition employed in the range of 1% to 14% by weight of beryllium oxide, were obtained. This type of fuel promises to be safer than current fuels, improving the performance of the reactor, in addition to last longer, resulting in great savings. (author)

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

    Science.gov (United States)

    Vauchy, Romain; Belin, Renaud C.; Robisson, Anne-Charlotte; Hodaj, Fiqiri

    2016-02-01

    In situ X-ray diffraction was used to study the structural changes occurring in uranium-plutonium mixed oxides U1-yPuyO2-x with y = 0.15; 0.28 and 0.45 during cooling from 1773 K to room-temperature under He + 5% H2 atmosphere. We compare the fastest and slowest cooling rates allowed by our apparatus i.e. 2 K s-1 and 0.005 K s-1, respectively. The promptly cooled samples evidenced a phase separation whereas samples cooled slowly did not due to their complete oxidation in contact with the atmosphere during cooling. Besides the composition of the annealing gas mixture, the cooling rate plays a major role on the control of the Oxygen/Metal ratio (O/M) and then on the crystallographic properties of the U1-yPuyO2-x uranium-plutonium mixed oxides.

  10. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

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

  11. Thermal expansion studies on uranium-neodymium mixed oxide solid solutions

    International Nuclear Information System (INIS)

    Panneerselvam, G.; Venkata Krishnan, R.; Antony, M.P.; Nagarajan, K.

    2012-01-01

    Uranium-Neodymium mixed oxides solid solutions (U 1-y Nd y ) O 2 (y=0.2-0.95) were prepared by combustion synthesis using citric acid as fuel. Structural characterization and computation of lattice parameter was carried out from room temperature X-ray diffraction measurements. Single-phase fluorite structure was observed up to y=0.80. For solid solutions with y>0.80 additional Nd 2 O 3 lines were visible

  12. Uranium

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  13. Uranium

    International Nuclear Information System (INIS)

    Mackay, G.A.

    1978-01-01

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

  14. Uranium

    International Nuclear Information System (INIS)

    1982-01-01

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

  15. Uranium

    International Nuclear Information System (INIS)

    Stewart, E.D.J.

    1974-01-01

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

  16. Uranium

    International Nuclear Information System (INIS)

    Toens, P.D.

    1981-03-01

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

  17. Analysis of refabricated fuel: determination of carbon in uranium plutonium mixed carbide

    International Nuclear Information System (INIS)

    Huwyler, S.

    1977-09-01

    In developing uranium plutonium mixed carbide which represents an advanced fuel for breeder reactors carbon analysis is an important means of determining the stoichiometry. Methods of carbon determination are briefly reviewed. The carbon determination using a LECO WR-12 Carbon Determinator is treated in detail and experience of three years operation communicated. Problems arising from operating the LECO-apparatus in a glove box are discussed. It is pointed out that carbon determination with the LECO-apparatus is a very fast method with good precision and well suited for the routine analysis of mixed carbide fuel. The accuracy of the method is checked by means of a standard. (Auth.)

  18. Contribution to the study of the microstructure of uranium dioxide (1962); Contribution a l'etude de la microstructure du dioxyde d'uranium (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Porneuf, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1960-05-15

    The microstructure of sintered uranium dioxide is studied in relation with several parameters, specially the sintering temperatures and atmospheres. The external surface and the internal microstructure of the sintered are examined, using fractography and ceramography. Various techniques for preparing surfaces (mechanical and electrolytic polishing) and for revealing the structure (chemical and anodic attack, ionic bombardment oxidation) have been experienced and compared. Patterns similar to those revealed in metals and probably related with interactions between dislocations and vacancies have been observed. (author) [French] La microstructure de frittes d'oxyde d'uranium est etudiee en fonction de divers parametres, en particulier de la temperature et de l'atmosphere de frittage, par examen de la surface externe des frittes, puis de leur microstructure interne (fractographie, ceramographie). Differentes techniques de preparation des surfaces (polissage mecanique ou electrolytique) et de revelation de la structure (attaque chimique ou anodique, bombardement ionique, oxydation preferentielle) ont ete experimentees et comparees. Des figures comparables a celles revelees dans les metaux et liees probablement a des interactions entre dislocations et lacunes ont ete observees. (auteur)

  19. Contribution to the study of the microstructure of uranium dioxide (1962); Contribution a l'etude de la microstructure du dioxyde d'uranium (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Porneuf, A. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1960-05-15

    The microstructure of sintered uranium dioxide is studied in relation with several parameters, specially the sintering temperatures and atmospheres. The external surface and the internal microstructure of the sintered are examined, using fractography and ceramography. Various techniques for preparing surfaces (mechanical and electrolytic polishing) and for revealing the structure (chemical and anodic attack, ionic bombardment oxidation) have been experienced and compared. Patterns similar to those revealed in metals and probably related with interactions between dislocations and vacancies have been observed. (author) [French] La microstructure de frittes d'oxyde d'uranium est etudiee en fonction de divers parametres, en particulier de la temperature et de l'atmosphere de frittage, par examen de la surface externe des frittes, puis de leur microstructure interne (fractographie, ceramographie). Differentes techniques de preparation des surfaces (polissage mecanique ou electrolytique) et de revelation de la structure (attaque chimique ou anodique, bombardement ionique, oxydation preferentielle) ont ete experimentees et comparees. Des figures comparables a celles revelees dans les metaux et liees probablement a des interactions entre dislocations et lacunes ont ete observees. (auteur)

  20. Thorium/uranium mixed oxide nano-crystals: Synthesis, structural characterization and magnetic properties

    International Nuclear Information System (INIS)

    Hudry, Damien; Griveau, Jean-Christophe; Apostolidis, Christos; Colineau, Eric; Rasmussen, Gert; Walter, Olaf; Wang, Di; Venkata Sai Kiran Chakravadhaluna; Courtois, Eglantine; Kubel, Christian

    2014-01-01

    One of the primary aims of the actinide community within nano-science is to develop a good understanding similar to what is currently the case for stable elements. As a consequence, efficient, reliable and versatile synthesis techniques dedicated to the formation of new actinide-based nano-objects (e.g., nano-crystals) are necessary. Hence, a 'library' dedicated to the preparation of various actinide based nano-scale building blocks is currently being developed. Nano-scale building blocks with tunable sizes, shapes and compositions are of prime importance. So far, the non-aqueous synthesis method in highly coordinating organic media is the only approach which has demonstrated the capability to provide size and shape control of actinide-based nano-crystals (both for thorium and uranium, and recently extended to neptunium and plutonium). In this paper, we demonstrate that the non-aqueous approach is also well adapted to control the chemical composition of the nano-crystals obtained when mixing two different actinides. Indeed, the controlled hot co-injection of thorium acetylacetonate and uranyl acetate (together with additional capping agents) into benzyl ether can be used to synthesize thorium/uranium mixed oxide nano-crystals covering the full compositional spectrum. Additionally, we found that both size and shape are modified as a function of the thorium/uranium ratio. Finally, the magnetic properties of the different thorium/uranium mixed oxide nano-crystals were investigated. Contrary to several reports, we did not observe any ferromagnetic behavior. As a consequence, ferromagnetism cannot be described as a universal feature of nano-crystals of non-magnetic oxides as recently claimed in the literature. (authors)

  1. NARCISS critical stand experiments for studying the nuclear safety in accident water immersion of highly enriched uranium dioxide fuel elements

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoj, N.N.; Glushkov, E.S.; Bubelev, V.G.

    2005-01-01

    A brief description of the Topaz-2 SNPS designed under scientific supervision of RRC KI in Russia, and of the NARCISS critical facility, is given. At the NARCISS critical facility, neutronic peculiarities and nuclear safety issues of the Topaz-2 system reactor were studied experimentally. This work is devoted to a detailed description of experiments on investigation of criticality safety in accident water immersion og highly enriched uranium dioxide fuel elements, performed at the NARCISS facility. The experiments were carried out at water-moderated critical assemblies with varying height, number, and spacing of fuel elements. The results obtained in the critical experiments, computational models of the investigated critical configurations, and comparison of the computational and experimental results are given [ru

  2. Compositional changes at the interface between thorium-doped uranium dioxide and zirconium due to high-temperature annealing

    Science.gov (United States)

    Youn, Young-Sang; Lee, Jeongmook; Kim, Jandee; Kim, Jong-Yun

    2018-06-01

    Compositional changes at the interface between thorium-doped uranium dioxide (U0.97Th0.03O2) and Zr before and after annealing at 1700 °C for 18 h were studied by X-ray photoelectron spectroscopy, X-ray diffraction, and Raman spectroscopy. At room temperature, the U0.97Th0.03O2 pellet consisted of hyperstoichiometric UO2+x with UO2 and ThO2, and the Zr sample contained Zr with ZrO2. After annealing, the former contained stoichiometric UO2 with ThO2 and the latter consisted of ZrO2 along with ZrO2·2H2O.

  3. French experience in the field of internal dosimetry assessment at a nuclear workplace. Methods and results on industrial uranium dioxide

    International Nuclear Information System (INIS)

    Ansoborlo, E.; Henge-Napoli, M.H.; Rannou, A.; Pihet, P.; Dewez, P.

    1995-01-01

    The implementation of the new ICRP recommendations and the diversity of industrial exposure materials make it necessary to modify our approach of assessing internal dosimetry. This paper describes a methodology developed to asses different parameters such as activity concentration and particle size distribution at the workplace; physico-chemical characteristics of industrial dust handled; and in vitro and in vivo solubility in order to determine the absorption rate blood. The determination of such specific parameters will lead to dose calculation in terms of committed effective Dose Per Unit of Intake (DPUI). Results obtained for an industrial uranium dioxide, UO 2 , at a French nuclear facility are presented. (author). 21 refs., 2 figs., 4 tabs

  4. High-temperature vaporization of thorium-uranium mixed monocarbide (Th1-y, Uy)C

    International Nuclear Information System (INIS)

    Koyama, Tadafumi; Yamawaki, Michio

    1989-01-01

    Vaporization thermodynamics of thorium-uranium mixed monocarbide phase (Th 1-y , U y )C was studied by mass spectrometric Knudsen effusion method for the compositions of (Th 0.9 , U 0.1 )C 0.855 , (Th 0.8 , U 0.2 )C 0.973 and (Th 0.6 , U 0.4 )C 0.973 . The partial vapor pressures of Th(g) and U(g) and activities of Th and U of these mixed monocarbides were determined at temperatures ranging from about 2000 to 2200 K. Further, the partial pressures of Th(g) and U(g) and activities of Th and U of the stoichiometric mixed monocarbides (Th 1-y , U y )C 1.00 were evaluated by compensating for the effect of carbon content. The Gibbs energies of formation of stoichiometric (Th 1-y , U y )C 1.00 were also evaluated. (orig.)

  5. Uranium dioxide in Fe(III)-containing ionic liquids with DMSO: Dissolution, separation, and structural characterization

    Energy Technology Data Exchange (ETDEWEB)

    Yao, Aining; Chu, Taiwei, E-mail: twchu@pku.edu.cn

    2016-11-15

    UO{sub 2} can be successfully dissolved in imidazolium-based Fe(III)-containing ionic liquids (ILs) with the help of DMSO. Spectroscopic studies and X-ray diffraction show that UO{sub 2}Cl{sub 4}{sup 2−} is the principal product. The dissolved uranyl species can be easily separated from the ILs via a combination of crystallization and solvent extraction. Moreover, even if 15.2 wt% of the rare-earth elements of Sm, Eu, and Gd, compared with the total amount of uranium and the rare-earth elements, exist in the IL, only uranium-containing crystals would be selectively formed and separated from the system. The solvents of acetone and acetonitrile could be used to separate the rare-earth elements from uranium in the IL with the help of imidazolium chloride. Considering the complete process from the dissolution of UO{sub 2} and some rare-earth oxides to the separation of uranium and rare-earth elements in the IL, the facile approach is promising for the spent nuclear fuel reprocessing. - Graphical abstract: UO{sub 2} can be successfully dissolved in Fe-containing ILs with the help of DMSO to form UO{sub 2}Cl{sub 4}{sup 2−}. The rare earth elements of Sm, Eu, and Gd can be separated from uranium in the IL, and meanwhile, the recovery of dissolved uranyl species and Fe-containing IL can also be achieved. - Highlights: • Dissolution of UO{sub 2} can be successfully achieved in imidazolium-based Fe-containing ILs with the help of DMSO without additional oxidants. • Compared with the total amount of uranium and the rare-earth elements, even if 15.2 wt% of the rare-earth elements of Sm, Eu, and Gd exist in the IL, only uranium-containing crystals would be selectively formed and separated from the system. • The separation of the rare-earth elements from uranium has also been achieved via a combination of crystallization and solvent extraction.

  6. Uranium dioxide in Fe(III)-containing ionic liquids with DMSO: Dissolution, separation, and structural characterization

    International Nuclear Information System (INIS)

    Yao, Aining; Chu, Taiwei

    2016-01-01

    UO_2 can be successfully dissolved in imidazolium-based Fe(III)-containing ionic liquids (ILs) with the help of DMSO. Spectroscopic studies and X-ray diffraction show that UO_2Cl_4"2"− is the principal product. The dissolved uranyl species can be easily separated from the ILs via a combination of crystallization and solvent extraction. Moreover, even if 15.2 wt% of the rare-earth elements of Sm, Eu, and Gd, compared with the total amount of uranium and the rare-earth elements, exist in the IL, only uranium-containing crystals would be selectively formed and separated from the system. The solvents of acetone and acetonitrile could be used to separate the rare-earth elements from uranium in the IL with the help of imidazolium chloride. Considering the complete process from the dissolution of UO_2 and some rare-earth oxides to the separation of uranium and rare-earth elements in the IL, the facile approach is promising for the spent nuclear fuel reprocessing. - Graphical abstract: UO_2 can be successfully dissolved in Fe-containing ILs with the help of DMSO to form UO_2Cl_4"2"−. The rare earth elements of Sm, Eu, and Gd can be separated from uranium in the IL, and meanwhile, the recovery of dissolved uranyl species and Fe-containing IL can also be achieved. - Highlights: • Dissolution of UO_2 can be successfully achieved in imidazolium-based Fe-containing ILs with the help of DMSO without additional oxidants. • Compared with the total amount of uranium and the rare-earth elements, even if 15.2 wt% of the rare-earth elements of Sm, Eu, and Gd exist in the IL, only uranium-containing crystals would be selectively formed and separated from the system. • The separation of the rare-earth elements from uranium has also been achieved via a combination of crystallization and solvent extraction.

  7. Uranium

    International Nuclear Information System (INIS)

    Whillans, R.T.

    1981-01-01

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

  8. Separation and mass spectrometry of nanogram quantities of uranium and thorium from thorium-uranium dioxide fuels

    International Nuclear Information System (INIS)

    Green, L.W.; Elliot, N.L.; Longhurst, T.H.

    1983-01-01

    A microchemical procedure was developed for the separation and isotopic analysis of U and Th from irradiated (Th,U)O 2 fuel. The separation procedure consisted of two stages; in the first a tributyl phosphate impregnated resin bead was equilibrated with the dissolved fuel in 0.08 M HF/6 M HNO 3 solution. The bead sorbed approximately 1.7 μg of U and 4.8μg of Th and provided good separation of these from the fission products. In the second stage, the U and Th were back-extracted into 0.025 M HF/8 M HNO 3 solution, which contained a small anion-exchange membrane disk. The disk adsorbed approximately 14 ng of U and 45 ng of Th, and subsequently was transferred to the ionizing filament of a thermal-ionization mass spectrometer and covered with a starch deposit. Sensitivities were sufficiently high for sequential analysis of these quantities of Th and U from a single disk. Isotopic data obtained for a combined U and Th standard showed excellent agreement with certified values: overall bias and precision were < 0.03% and 0.2% relative standard deviation, respectively, for both elements. The applicability of the procedure to uranium fuels was also demonstrated. 6 figures, 2 tables

  9. Separation and mass spectrometry of nanogram quantities of uranium and thorium from thorium-uranium dioxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Green, L.W.; Elliot, N.L.; Longhurst, T.H

    1983-07-01

    A convenient and sensitive microchemical procedure was developed for the separation and isotopic analysis of U and Th from irradiated (Th,U)O{sub 2} fuel. The separation procedure consisted of two stages; in the first a tributyl phosphate impregnated resin bead was equilibrated with the dissolved fuel in 0.08 M HF/6 M HNO{sub 3} solution. The bead sorbed approximately 1.7 {mu}g of U and 4.8 {mu}g of Th and provided good separation of these from the fission products. In the second stage, the U and Th were back-extracted into 0.025 M HF/8 M HNO{sub 3} solution, which contained a small anion-exchange membrane disk. The disk adsorbed approximately 14 ng of U and 45 ng of Th, and subsequently was transferred to the ionizing filament of a thermal-ionization mass spectrometer and covered with a starch deposit. Sensitivities were sufficiently high for sequential analysis of these quantities of Th and U from a single disk. Isotopic data obtained for a combined U and Th standard showed excellent agreement with certified values: overall bias and precision were < -0.03% and 0.2% relative standard deviation, respectively, for both elements. The applicability of the procedure to uranium fuels was also demonstrated. (author)

  10. On the separation of so-called non-volatile uranium fission products of uranium using the conversion of neutron-irradiated uranium dioxide and graphite

    International Nuclear Information System (INIS)

    Elhardt, W.

    1979-01-01

    The investigations are continued in the following work which arose from the concept of separating uranium fission products from uranium. This is achieved in that due to the lattice conversions occurring during the course of solid chemical reactions, fission products can easily pass from the uranium-contained solid to a second solid. The investigations carried out primarily concern the release behaviour of cerium and neodymium in the temperature region of 1200 to 1700 0 C. UO 2 + graphite, both in powder form, are selected as suitable reaction system having the preconditions needed for the lattice conversion for the release effect. The target aimed at from the practical aspect for the improved release of lanthanoids is achieved by an isobar test course - changing temperature from 1200 to 1500 0 C at constant pressure, with a cerium release of 75-80% and a neodynium release of 80-90% (maximum at 1400 0 C). The concepts on the mechanism of the fission product release are related to transport processes in crystal lattices, as well as chemical solid reactions and evaporation processes on the surface of UC 2 grains. (orig./RB) [de

  11. Carbon dioxide selective adsorption within a highly stable mixed-ligand Zeolitic Imidazolate Framework

    KAUST Repository

    Huang, Lin

    2014-08-01

    A new mixed-ligand Zeolitic Imidazolate Framework Zn4(2-mbIm) 3(bIm)5·4H2O (named JUC-160, 2-mbIm = 2-methylbenzimidazole, bIm = benzimidazole and JUC = Jilin University China) was synthesized with a solvothermal reaction of Zn(NO3) 2·6H2O, bIm and 2-mbIm in DMF solution at 180 °C. Topological analysis indicated that JUC-160 has a zeolite GIS (gismondine) topology. Study of the gas adsorption and thermal and chemical stability of JUC-160 demonstrated its selective adsorption property for carbon dioxide, high thermal stability, and remarkable chemical resistance to boiling alkaline water and organic solvent for up to one week. © 2014 Elsevier B.V.

  12. Ultrafast Dynamics in Vanadium Dioxide: Separating Spatially Segregated Mixed Phase Dynamics in the Time-domain

    Science.gov (United States)

    Hilton, David

    2011-10-01

    In correlated electronic systems, observed electronic and structural behavior results from the complex interplay between multiple, sometimes competing degrees-of- freedom. One such material used to study insulator-to-metal transitions is vanadium dioxide, which undergoes a phase transition from a monoclinic-insulating phase to a rutile-metallic phase when the sample is heated to 340 K. The major open question with this material is the relative influence of this structural phase transition (Peirels transition) and the effects of electronic correlations (Mott transition) on the observed insulator-to-metal transition. Answers to these major questions are complicated by vanadium dioxide's sensitivity to perturbations in the chemical structure in VO2. For example, related VxOy oxides with nearly a 2:1 ratio do not demonstrate the insulator-to- metal transition, while recent work has demonstrated that W:VO2 has demonstrated a tunable transition temperature controllable with tungsten doping. All of these preexisting results suggest that the observed electronic properties are exquisitely sensitive to the sample disorder. Using ultrafast spectroscopic techniques, it is now possible to impulsively excite this transition and investigate the photoinduced counterpart to this thermal phase transition in a strongly nonequilibrium regime. I will discuss our recent results studying the terahertz-frequency conductivity dynamics of this photoinduced phase transition in the poorly understood near threshold temperature range. We find a dramatic softening of the transition near the critical temperature, which results primarily from the mixed phase coexistence near the transition temperature. To directly study this mixed phase behavior, we directly study the nucleation and growth rates of the metallic phase in the parent insulator using non-degenerate optical pump-probe spectroscopy. These experiments measure, in the time- domain, the coexistent phase separation in VO2 (spatially

  13. Advances on reverse strike co-precipitation method of uranium-plutonium mixed solutions

    International Nuclear Information System (INIS)

    Menghini, Jorge E.; Marchi, Daniel E.; Orosco, Edgardo H.; Greco, Luis

    2000-01-01

    The reverse strike coprecipitation of uranium-plutonium mixed solutions, is an alternative way to obtain MOX fuel pellets. Previous tests, carried out in the Alpha Laboratory, included a stabilization step for transforming 100 % of plutonium into Pu +4 . Therefore, the plutonium precipitated as Pu(OH) 4 . In this second step, the stabilization process was suppressed. In this way, besides Pu(OH) 4 , a part of the precipitated is composed of a mixed salt: AD(U,Pu). Then, a homogeneous solid solution is formed in the early steps of the process. The powders showed higher tap density, better performance during the pressing and lower sinterability than the powders obtained in previous tests. The advantageous and disadvantageous effects of the stabilization step are analyzed in this paper. (author)

  14. Preparation of uranium dioxide by thermal decomposition and direct reduction of ammonium uranate

    International Nuclear Information System (INIS)

    Hernandez R, R.

    1995-01-01

    The thermal decomposition of ammonium uranate has been studied by infrared spectroscopy, and X-ray diffraction. It has been show that ammonia remains in the solid until substantially 350 Centigrade degrees, when gaseous nitrogen is released. It is concluded that compounds derived from the calcination of ammonium uranate at atmospheric pressure, produced amorphous U O 3 at about 350-400 Centigrade degrees and transform to U 3 O 8 via α - U O 3 and/or α - U O 3 . The object of this study was to obtain reliable fundamental information regarding the character of the pure carbon monoxide-ammonium uranate-uranium trioxide-uranium octaoxide reaction, in the range of temperatures that has been used in commercial reduction processes. Through the use of high-purity samples and by the proper control of incidental variable, this object was realized. (Author)

  15. AES study of growth process of al thin films on uranium dioxide

    International Nuclear Information System (INIS)

    Zhou Wei; Liu Kezhao; Yang Jiangrong; Xiao Hong

    2009-01-01

    Metallic uranium was exposed to 40 languirs of oxygen at room temperature in order to form UO 2 on the surface of metallic U. And thin layers of aluminum on UO 2 were prepared by sputter deposition under ultra high vacuum conditions. Process of Al thin film growth and its interaction with UO 2 were investigated by auger electron spectroscopy (AES) and electron energy loss spectroscopy (EELS). It was shown that the Al thin film growth underwent via the Volmer-Weber (VW) mode. At room temperature, Al and UO 2 interact with each other, electrons transfer occurres from Al atoms to uranium ions, and a few of Al 2 O 3 exist in the region of UO 2 /Al interface due to O 2 adsorption to the surface. Inter-diffusion between UO 2 and Al is observable. Aluminum diffuses into interface region of UO 2 and U. It results in the formation of a coexistence regime containing uranium oxide, metallic U and Al. (authors)

  16. Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R M

    1976-01-01

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

  17. A conceptual framework to quantify the influence of convective boundary layer development on carbon dioxide mixing ratios

    NARCIS (Netherlands)

    Pino, D.; Vilà-Guerau de Arellano, J.; Peters, W.; Schröter, J.; van Heerwaarden, C. C.; Krol, M. C.

    2012-01-01

    Interpretation of observed diurnal carbon dioxide (CO2) mixing ratios near the surface requires knowledge of the local dynamics of the planetary boundary layer. In this paper, we study the relationship between the boundary layer dynamics and the CO2 budget in convective conditions through a newly

  18. Chloride pyrometallurgy of uranium ore. 2. Selective extraction of uranium using the mixed gas of chlorine and oxygen in the presence of activated carbon

    International Nuclear Information System (INIS)

    Taki, Tomihiro; Komoto, Shigetoshi; Otomura, Keiichiro; Takenaka, Toshihide; Sato, Nobuaki; Fujino, Takeo.

    1996-01-01

    Selective extraction of uranium from a phosphate ore was studied by using the mixed gas of Cl 2 and O 2 . On heating the ore and carbon mixture in Cl 2 , the volatilized chloride of uranium is accompanied by iron, aluminum, phosphorus and silicon chlorides. The addition of O 2 gas effectively lowered the volatilization ratios of aluminum, phosphorus and silicon. The ratio decreased with increasing oxygen flow rate up to 50 ml/min at 1,223 K (Cl 2 : 100 ml/min, O 2 +N 2 : 400 ml/min). The volatilization ratio of uranium was almost unchanged at 90% up to 50 ml/min O 2 (carbon amount: 15 wt%). The effect of the other parameters, i.e. Cl 2 flow rate, carbon amount, reaction temperature and time was examined. (author)

  19. Amine–mixed oxide hybrid materials for carbon dioxide adsorption from CO2/H2 mixture

    Science.gov (United States)

    Ravi, Navin; Aishah Anuar, Siti; Yusuf, Nur Yusra Mt; Isahak, Wan Nor Roslam Wan; Shahbudin Masdar, Mohd

    2018-05-01

    Bio-hydrogen mainly contains hydrogen and high level of carbon dioxide (CO2). High concentration of CO2 lead to a limitation especially in fuel cell application. In this study, the amine-mixed oxide hybrid materials for CO2 separation from bio-hydrogen model (50% CO2:50% H2) have been studied. Fourier-transform infrared spectroscopy (FTIR) and x-ray diffraction (XRD) characterizations showed that the amine–mixed oxide hybrid materials successfully adsorbed CO2 physically with no chemical adsorption evidence. The dry gas of CO2/H2 mixture adsorbed physically on amine–CuO–MgO hybrid material. No carbonates were detected after several times of adsorption, which indicated the good recyclability of adsorbents. The adsorbent system of diethanolamine (DEA)/15% CuO–75% MgO showed the highest CO2 adsorption capacity of 21.2 wt% due to the presence of polar substance on MgO surface, which can adsorb CO2 at ambient condition. The alcohol group of DEA can enhance the CO2 solubility on the adsorbent surface. In the 20% CuO–50% MgO adsorbent system, DEA as amine type showed a high CO2 adsorption of 19.4 wt%. The 10% amine loading system showed that the DEA adsorption system provided high CO2 adsorption. The BET analysis confirmed that a high amine loading contributed to the decrease in CO2 adsorption due to the low surface area of the adsorbent system.

  20. Carbon dioxide capture from reforming gases using acetic acid-mixed chemical absorbents

    Energy Technology Data Exchange (ETDEWEB)

    Rahmanian, Amin; Zaini, Muhammad Abbas ahmad; Abdullah, Tuan Amran Tuan [Faculty of Chemical Engineering, Universiti Teknologi Malaysia, Johor Bahru (Malaysia)

    2015-07-15

    Carbon dioxide (CO{sub 2}) is a major problem in the production of natural gas as it may contribute to the operational problems such as foaming, corrosion, high solution viscosity, and fouling, thereby decreasing the plant life. The presence of acid gas in natural gas reforming may also result in the increase of transported gas volume and the decrease of heating value. Absorption using aqueous solutions of alkanolamines has been a preferred approach in current industry for CO{sub 2} removal. Concentration of ammonia and DEA affects the CO{sub 2} removal; increasing the absorbents concentration increases the CO{sub 2} removal. On molar basis, DEA shows a greater CO{sub 2} absorption than ammonia. Acetic acid-mixed absorbents display a lower CO{sub 2} removal than the nonmixed ones. Decrease in solubility due to the decrease in solution pH has resulted in a lower CO{sub 2} absorption by acetic acid-mixed absorbents. Liquid flow rate offers only small influence on the absorption of CO{sub 2}, while decreasing the gas flow rate increases the CO{sub 2} removal. On the operational point of view, blend of ammonia and DEA absorbent would be beneficial for CO{sub 2} removal from reforming gases as it could partly solve the problems associated with regeneration and corrosion.

  1. Hanford/Rocky Flats collaboration on development of supercritical carbon dioxide extraction to treat mixed waste

    International Nuclear Information System (INIS)

    Hendrickson, D.W.; Biyani, R.K.; Brown, C.M.; Teter, W.L.

    1995-11-01

    Proposals for demonstration work under the Department of Energy's Mixed Waste Focus Area, during the 1996 through 1997 fiscal years included two applications of supercritical carbon dioxide to mixed waste pretreatment. These proposals included task RF15MW58 of Rocky Flats and task RL46MW59 of Hanford. Analysis of compatibilities in wastes and work scopes yielded an expectation of substantial collaboration between sites whereby Hanford waste streams may undergo demonstration testing at Rocky Flats, thereby eliminating the need for test facilities at Hanford. This form of collaboration is premised the continued deployment at Rocky Flats and the capability for Hanford samples to be treated at Rocky Flats. The recent creation of a thermal treatment contract for a facility near Hanford may alleviate the need to conduct organic extraction upon Rocky Flats wastes by providing a cost effective thermal treatment alternative, however, some waste streams at Hanford will continue to require organic extraction. Final site waste stream treatment locations are not within the scope of this document

  2. Hanford/Rocky Flats collaboration on development of supercritical carbon dioxide extraction to treat mixed waste

    Energy Technology Data Exchange (ETDEWEB)

    Hendrickson, D.W.; Biyani, R.K. [Westinghouse Hanford Co., Richland, WA (United States); Brown, C.M.; Teter, W.L. [Kaiser-Hill Co., Golden, CO (United States)

    1995-11-01

    Proposals for demonstration work under the Department of Energy`s Mixed Waste Focus Area, during the 1996 through 1997 fiscal years included two applications of supercritical carbon dioxide to mixed waste pretreatment. These proposals included task RF15MW58 of Rocky Flats and task RL46MW59 of Hanford. Analysis of compatibilities in wastes and work scopes yielded an expectation of substantial collaboration between sites whereby Hanford waste streams may undergo demonstration testing at Rocky Flats, thereby eliminating the need for test facilities at Hanford. This form of collaboration is premised the continued deployment at Rocky Flats and the capability for Hanford samples to be treated at Rocky Flats. The recent creation of a thermal treatment contract for a facility near Hanford may alleviate the need to conduct organic extraction upon Rocky Flats wastes by providing a cost effective thermal treatment alternative, however, some waste streams at Hanford will continue to require organic extraction. Final site waste stream treatment locations are not within the scope of this document.

  3. Study of the behaviour of cesium fission product in uranium dioxide by the ab initio method

    International Nuclear Information System (INIS)

    Gupta, Florence

    2008-01-01

    The knowledge of the behaviour of fission products in the nuclear fuel is very important for safety considerations and for understanding the evolution of the fuel properties under irradiation. In this work, we focussed mainly on the behaviour of caesium in UO 2 through ab initio studies of its solubility at point defects in the matrix, its diffusion and its contribution to the formation of solid phases in the fuel. The role of electronic correlation effects of the f electrons of uranium on these properties and on the description of the defect free crystal, is assessed. The formation energies of the main point defects are calculated and their concentration as a function of fuel stoichiometry and temperature is estimated. The migration barriers and migration paths for the self-diffusion of oxygen and uranium vacancies and oxygen interstitials in UO 2 are discussed. The solubility of Cs is found to be very low in UO 2 in agreement with experimental findings. The most favourable trapping sites are determined as a function of oxygen concentration in the fuel. Our results show that in the hyper-stoichiometric regime, the diffusion of Cs from its most favourable trapping site is limited by the uranium vacancy diffusion mechanism. We also considered the formation of the main solid phases of caesium resulting from its oxidation (Cs 2 O, Cs 2 O 2 , CsO 2 ) and from its interaction with the fuel (Cs 2 UO 4 ), with molybdenum (Cs 2 MoO 4 ) and with the zirconium of the clad (Cs 2 ZrO 3 ), since the formation of such phases, their solubility and their interdependence will affect the release of caesium. (author)

  4. Groundwater contamination from an inactive uranium mill tailings pile. 2. Application of a dynamic mixing model

    International Nuclear Information System (INIS)

    Narashimhan, T.N.; White, A.F.; Tokunaga, T.

    1986-01-01

    At Riverton, Wyoming, low pH process waters from an abandoned uranium mill tailings pile have been infiltrating into and contaminating the shallow water table aquifer. The contamination process has been governed by transient infiltration rates, saturated-unsaturated flow, as well as transient chemical reactions between the many chemical species present in the mixing waters and the sediments. In the first part of this two-part series the authors presented field data as well as an interpretation based on a static mixing models. As an upper bound, the authors estimated that 1.7% of the tailings water had mixed with the native groundwater. In the present work they present the results of numerical investigation of the dynamic mixing process. The model, DYNAMIX (DYNamic MIXing), couples a chemical speciation algorithm, PHREEQE, with a modified form of the transport algorithm, TRUMP, specifically designed to handle the simultaneous migration of several chemical constituents. The overall problem of simulating the evolution and migration of the contaminant plume was divided into three sub problems that were solved in sequential stages. These were the infiltration problem, the reactive mixing problem, and the plume-migration problem. The results of the application agree reasonably with the detailed field data. The methodology developed in the present study demonstrates the feasibility of analyzing the evolution of natural hydrogeochemical systems through a coupled analysis of transient fluid flow as well as chemical reactions. It seems worthwhile to devote further effort toward improving the physicochemical capabilities of the model as well as to enhance its computational efficiency

  5. Uranium

    International Nuclear Information System (INIS)

    Perkin, D.J.

    1982-01-01

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

  6. Uranium

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  7. A kinetic study of the reaction of water vapor and carbon dioxide on uranium; Cinetique de la reaction de la vapeur d'eau et du dioxyde de carbone sur l'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Santon, J P [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1964-09-15

    The kinetic study of the reaction of water vapour and carbon dioxide with uranium has been performed by thermogravimetric methods at temperatures between 160 and 410 deg G in the first case, 350 and 1050 deg C in the second: Three sorts of uranium specimens were used: uranium powder, thin evaporated films, and small spheres obtained from a plasma furnace. The experimental results led in the case of water vapour, to a linear rate of reaction controlled by diffusion at the lower temperatures, and by a surface reaction at the upper ones. In the case of carbon dioxide, a parabolic law has been found, controlled by diffusional processes. (author) [French] L'etude cinetique de la reaction de la vapeur d'eau et du dioxyde de carbone sur l'uranium a ete entreprise au moyen de methodes thermogravimetriques, dans te premier cas entre 160 et 410 deg C et dans le second entre 350 et 1050 deg C. Le materiau utilise se presentait sous trois formes: poudres, couches minces evaporees et billes obtenues par fusion en chalumeau a plasma. Les resultats experimentaux ont permis de mettre en evidence, dans le cas de la vapeur d'eau, une cinetique lineaire controlee par la diffusion a basse temperature et d'interface a haute temperature. Dans le cas du dioxyde de carbone par contre, on trouve une cinetique parabolique controlee par la diffusion. (auteur)

  8. Creep of uranium dioxide: bending test and mechanical behaviour; Etude du fluage du dioxyde d'uranium: caracterisation par essais de flexion et modelisation mecanique

    Energy Technology Data Exchange (ETDEWEB)

    Colin, Ch

    2003-09-01

    These PhD work in the frame of Pellet-Cladding Interactions studies, in the fuel assemblies of nuclear plants. Electricite de France (EDF) must well demonstrate and insure the integrity of the cladding. For that purpose, the viscoplastic behaviour of the nuclear fuel has to be known and, if possible, controlled. This PhD work aimed to characterize the creep of uranium dioxide, in conditions of transient power regime. First, a literature survey on mechanical behaviour of UO{sub 2} revealed that the ceramic was essentially studied with compressive tests, and that its creep behaviour is characterized by two domains, depending on the stress level. To estimate the loadings in a fuel pellet, EDF and CEA developed specific global codes. A simulation during a power ramp allowed the order of magnitude of the loadings in the pellet to be determined (temperature, thermal gradients, strains, strain rate...). The stress calculation using a finite element simulation requires the identification of behaviour laws, able to describe the behaviour under small strains, low strain rates, and under tensile stresses. Starting from this observation, three point bending method has been chosen to test the uranium dioxide. As, for representativeness reasons, testing specimens cut in actual fuel pads was required in our study; a ten millimeters span has been used. For this study, a specific three-point testing device has been developed, that can tests specimens up to 2 000 C in a controlled atmosphere (Ar + 5% H{sub 2}). A special care has been taken for the measurement of the deflexion of the sample, which is measured using a laser beam, that allow an accuracy of {+-}2{mu}m to be reached at high temperature. Specimens with 0,5 to 1 mm thickness have been tested using this jig. A Norton's law describe, with respective stress exponent and activation energy values of 1.73 and 540 kJ.mole-1, provided a good description of the stationary creep rate. Then, the mechanical behaviour of the fuel

  9. Mixed core management: Use of 93% and 72% enriched uranium in the BR2 reactor

    International Nuclear Information System (INIS)

    Ponsard, B.

    2000-01-01

    The BR2 reactor, put into operation in 1963 and refurbished from July 1995 till April 1997, is a 100 MW high-flux Materials Testing Reactor, using 93% 235 U enriched uranium as standard fuel, light water as coolant and beryllium as moderator. The present operating regime consists of five irradiation cycles per year at an operating power between 50 and 70 MW; each cycle is characterized by 21 days operation. In the framework of a 'qualification programme', six 72% 235 U fuel elements fabricated with uranium recovered from the reprocessing of BR2 spent fuel at UKAEA-Dounreay have been successfully irradiated in the period 1994-1995 reaching a maximum mean burnup of 48% without the release of fission products. Since 1998, this type of fuel element is irradiated routinely together with standard 93% 235 U fuel elements in order to optimize the utilization of the available HEU inventory. The purpose of this paper is to present the strategy developed in order to optimize the mixed core management of the BR2 reactor. (author)

  10. Extraction of Uranium from Aqueous Solutions Using Ionic Liquid and Supercritical Carbon Dioxide in Conjunction

    International Nuclear Information System (INIS)

    Wang, Joanna S.; Sheaff, Chrystal N.; Yoon, Byunghoon; Addleman, Raymond S.; Wai, Chien M.

    2009-01-01

    Uranyl ions (UO2)2+ in aqueous nitric acid solutions can be extracted into supercritical CO2 (sc-CO2) via an imidazolium-based ionic liquid using tri-n-butylphosphate (TBP) as a complexing agent. The transfer of uranium from the ionic liquid to the supercritical fluid phase was monitored by UV/Vis spectroscopy using a high-pressure fiberoptic cell. The form of the uranyl complex extracted into the supercritical CO2 phase was found to be UO2(NO3)2(TBP)2. The extraction results were confirmed by UV/Vis spectroscopy and by neutron activation analysis. This technique could potentially be used to extract other actinides for applications in the field of nuclear waste management.

  11. Determination of Krypton Diffusion Coefficients in Uranium Dioxide Using Atomic Scale Calculations.

    Science.gov (United States)

    Vathonne, Emerson; Andersson, David A; Freyss, Michel; Perriot, Romain; Cooper, Michael W D; Stanek, Christopher R; Bertolus, Marjorie

    2017-01-03

    We present a study of the diffusion of krypton in UO 2 using atomic scale calculations combined with diffusion models adapted to the system studied. The migration barriers of the elementary mechanisms for interstitial or vacancy assisted migration are calculated in the DFT+U framework using the nudged elastic band method. The attempt frequencies are obtained from the phonon modes of the defect at the initial and saddle points using empirical potential methods. The diffusion coefficients of Kr in UO 2 are then calculated by combining this data with diffusion models accounting for the concentration of vacancies and the interaction of vacancies with Kr atoms. We determined the preferred mechanism for Kr migration and the corresponding diffusion coefficient as a function of the oxygen chemical potential μ O or nonstoichiometry. For very hypostoichiometric (or U-rich) conditions, the most favorable mechanism is interstitial migration. For hypostoichiometric UO 2 , migration is assisted by the bound Schottky defect and the charged uranium vacancy, V U 4- . Around stoichiometry, migration assisted by the charged uranium-oxygen divacancy (V UO 2- ) and V U 4- is the favored mechanism. Finally, for hyperstoichiometric or O-rich conditions, the migration assisted by two V U 4- dominates. Kr migration is enhanced at higher μ O , and in this regime, the activation energy will be between 4.09 and 0.73 eV depending on nonstoichiometry. The experimental values available are in the latter interval. Since it is very probable that these values were obtained for at least slightly hyperstoichiometric samples, our activation energies are consistent with the experimental data, even if further experiments with precisely controlled stoichiometry are needed to confirm these results. The mechanisms and trends with nonstoichiometry established for Kr are similar to those found in previous studies of Xe.

  12. Study on principle and method of measuring system for external dimensions, geometric density and appearance quality of uranium dioxide pellet

    International Nuclear Information System (INIS)

    Cao Wei; Deng Hua; Wang Tao

    2010-01-01

    To adapt to the need of nuclear power development, and keep in step with the increasingly growing nuclear fuel element production, a special measuring system for integrated measuring, calculation, data processing method of External Dimensions, Tolerance of figure and place, Geometric Density and Appearance Quality of Uranium Dioxide Pellet is studied and discussed. This system is with important guiding significance for the improvement of technologic and frocking level.. The measuring system is primarily applied to sampling test during production and is the same with several types of products.The successful application of this measuring method ensures the accuracy and reliability of measured data, reduces the artificial error and makes the measuring be move convenient and fast, thus achieves high precision and high efficiency of measuring process. The measuring method is approach the advanced world level of measuring method at the same industry. So, based on the product inspection requirement, using special measuring instrument and computer data processing system is an important approach we use for nonce and future. (authors)

  13. Plasmachemical synthesis and evaluation of the thermal conductivity of metal-oxide compounds "Molybdenum-uranium dioxide"

    Science.gov (United States)

    Kotelnikova, Alexandra A.; Karengin, Alexander G.; Mendoza, Orlando

    2018-03-01

    The article represents possibility to apply oxidative and reducing plasma for plasma-chemical synthesis of metal-oxide compounds «Mo‒UO2» from water-salt mixtures «molybdic acid‒uranyl nitrate» and «molybdic acid‒ uranyl acetate». The composition of water-salt mixture was calculated and the conditions ensuring plasma-chemical synthesis of «Mo‒UO2» compounds were determined. Calculations were carried out at atmospheric pressure over a wide range of temperatures (300-4000 K), with the use of various plasma coolants (air, hydrogen). The heat conductivity coefficients of metal-oxide compounds «Mo‒UO2» consisting of continuous component (molybdenum matrix) are calculated. Inclusions from ceramics in the form of uranium dioxide were ordered in the matrix. Particular attention is paid to methods for calculating the coefficients of thermal conductivity of these compounds with the use of different models. Calculated results were compared with the experimental data.

  14. Fracture of Zircaloy cladding by interactions with uranium dioxide pellets in LWR fuel rods. Technical report 10

    International Nuclear Information System (INIS)

    Smith, E.; Ranjan, G.V.; Cipolla, R.C.

    1976-11-01

    Power reactor fuel rod failures can be caused by uranium dioxide fuel pellet-Zircaloy cladding interactions. The report summarizes the current position attained in a detailed theoretical study of Zircaloy cladding fracture caused by the growth of stress corrosion cracks which form near fuel pellet cracks as a consequence of a power increase after a sufficiently high burn-up. It is shown that stress corrosion crack growth in irradiated Zircaloy must be able to proceed at very low stress intensifications if uniform friction effects are operative at the fuel-cladding interface, when the interfacial friction coefficient is less than unity, when a symmetric distribution of fuel cracks exists, and when symmetric interfacial slippage occurs (i.e., ''uniform'' conditions). Otherwise, the observed fuel rod failures must be due to departures from ''uniform'' conditions, and a very high interfacial friction coefficient and particularly fuel-cladding bonding, are means of providing sufficient stess intensification at a cladding crack tip to explain the occurrence of cladding fractures. The results of the investigation focus attention on the necessity for reliable experimental data on the stress corrosion crack growth behavior of irradiated Zircaloy, and for further investigations on the correlation between local fuel-cladding bonding and stress corrosion cracking

  15. The pressure bonding ability of uranium dioxide powders in relation to the evolution of their surface properties

    International Nuclear Information System (INIS)

    Danroc, J.

    1982-09-01

    The long term storage of sinterable uranium dioxide powders generally improves their pressure bonding ability and the strength of the resulting green pellets. Evidence of the gradual evolution of the surface texture and composition of these powders during storage at room temperature and pressure has been provided by infrared spectroscopy, X-ray diffraction and thermogravimetric and microcalorimetric methods. These techniques demonstrated the existence of a thin adherent surface layer of UO 3 2H 2 0. Such a natural evolutionary process can be reproduced and substantially amplified by subjecting the powder to thermal treatments at temperatures up to 90 0 C in a moist air environment. It was shown that powder treated in this manner could be more readily compacted into strong green pellets than could raw material. The tensile strength, commonly regarded as a quality test for such pellets and measured by the brazilian method, was found to be at least twice that of normal pellets. The high density and geometric integrity of these sintered products ensures the extrapolation of these preparation techniques to the mass production of nuclear reactor fuel pellets [fr

  16. Influence of instrument conditions on the evaporation behavior of uranium dioxide with UV laser-assisted atom probe tomography

    International Nuclear Information System (INIS)

    2015-01-01

    Atom probe tomography (APT) provides the ability to detect subnanometer chemical variations spatially with high accuracy. Due to its ability to spatially characterize chemistry in non-conducting materials, such as oxides, provides the opportunity to characterize stoichiometry, which strongly is tied to material performance. However, accuracy has been correlated with instrument run parameters. A systematic study of the effect of laser energy, temperature, and detection rate is performed on the evaporation behavior of a model oxide, uranium dioxide (UO 2 ). Modifying the detection rate and temperature did not affect its evaporation behavior as laser energy. It was discovered that three laser evaporation regimes are present in UO 2 . Very low laser energy produces a behavior similar to DC-field evaporation, moderate laser energy produces the desired laser assisted field evaporation and high laser energy produces thermal effects in the evaporation behavior. Laser energy had the greatest impact on evaporation and the optimal instrument condition for UO 2 was determined to be 50K, 10 pJ laser energy, 0.3% detection rate, and a 100 kHz repetition rate. These conditions provide the best combination of mass resolution, accurate stoichiometry, and evaporation behavior.

  17. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

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

  18. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

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

  19. Uranium conversion

    International Nuclear Information System (INIS)

    Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina

    2006-03-01

    FOI, has performed a study on uranium conversion processes that are of importance in the production of different uranium compounds in the nuclear industry. The same conversion processes are of interest both when production of nuclear fuel and production of fissile material for nuclear weapons are considered. Countries that have nuclear weapons ambitions, with the intention to produce highly enriched uranium for weapons purposes, need some degree of uranium conversion capability depending on the uranium feed material available. This report describes the processes that are needed from uranium mining and milling to the different conversion processes for converting uranium ore concentrate to uranium hexafluoride. Uranium hexafluoride is the uranium compound used in most enrichment facilities. The processes needed to produce uranium dioxide for use in nuclear fuel and the processes needed to convert different uranium compounds to uranium metal - the form of uranium that is used in a nuclear weapon - are also presented. The production of uranium ore concentrate from uranium ore is included since uranium ore concentrate is the feed material required for a uranium conversion facility. Both the chemistry and principles or the different uranium conversion processes and the equipment needed in the processes are described. Since most of the equipment that is used in a uranium conversion facility is similar to that used in conventional chemical industry, it is difficult to determine if certain equipment is considered for uranium conversion or not. However, the chemical conversion processes where UF 6 and UF 4 are present require equipment that is made of corrosion resistant material

  20. Chemical states of fission products in irradiated uranium-plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    Kurosaki, Ken; Uno, Masayoshi; Yamanaka, Shinsuke

    1999-01-01

    The chemical states of fission products (FPs) in irradiated uranium-plutonium mixed oxide (MOX) fuel for the light water reactor (LWR) were estimated by thermodynamic equilibrium calculations on system of fuel and FPs by using ChemSage program. A stoichiometric MOX containing 6.1 wt. percent PuO 2 was taken as a loading fuel. The variation of chemical states of FPs was calculated as a function of oxygen potential. Some pieces of information obtained by the calculation were compared with the results of the post-irradiation examination (PIE) of UO 2 fuel. It was confirmed that the multicomponent and multiphase thermodynamic equilibrium calculation between fuel and FPs system was an effective tool for understanding the behavior of FPs in fuel. (author)

  1. Methods for analysis of uranium-plutonium mixed fuel and transplutonium elements at RIAR (Preprint no. IT-25)

    International Nuclear Information System (INIS)

    Timofeev, G.A.

    1991-02-01

    Different methods for analysis of the uranium-plutonium mixed nuclear fuel and transplutonium elements are briefly discussed in this paper: coulometry, radiometric techniques, emission spectrography, mass-spectrometry, chromatography, spectrophotometry. The main analytical characteristics of the methods developed are given. (author). 30 refs., 2 tabs

  2. Investigating the structural changes of uranium dioxide dependent on additives, Phase I - Uranium-oxide system from structural-phase aspect; Ispitivanje strukturnih promena kod urandioksida u zavisnosti od aditiva, I faza - Sistem Uran-kiseonik sa strukturno-faznog aspekta

    Energy Technology Data Exchange (ETDEWEB)

    Manojlovic, Lj [Institute of Nuclear Sciences Boris Kidric, Laboratorija za reaktorske materijale, Vinca, Beograd (Serbia and Montenegro)

    1962-12-15

    Having in mind the complex structure of the system uranium-oxygen, and that experimental studies of this system lead to controversial conclusions, an extensive review and analysis of the papers published on this subject were needed. This review wold be very useful for interpreting the expected structural changes of the uranium dioxide dependent on the additives.

  3. Application of Radio-Frequency Plasma Glow Discharge to Removal of Uranium Dioxide from Metal Surfaces

    International Nuclear Information System (INIS)

    El-Genk, Mohamed S.; Saber, Hamed H.

    2000-01-01

    Recent experiments have shown that radio-frequency (rf) plasma glow discharge using NF 3 gas is an effective technique for the removal of uranium oxide from metal surfaces. The results of these experiments are analyzed to explain the measured dependence of the UO 2 removal or etch rate on the NF 3 gas pressure and the absorbed power in the plasma. The NF 3 gas pressure in the experiments was varied from 10.8 to 40 Pa, and the deposited power in the plasma was varied from 25 to 210 W. The UO 2 etch rate was strongly dependent on the absorbed power and, to a lesser extent, on the NF 3 pressure and decreased exponentially with immersion time. At 210 W and 17 Pa, all detectable UO 2 in the samples (∼10.6 mg each) was removed at the endpoint, whereas the initial etch rate was ∼3.11 μm/min. When the absorbed power was ≤50 W, however, the etch rate was initially ∼0.5 μg/min and almost zero at the endpoint, with UO 2 only partially etched. This self-limiting etching of UO 2 at low power is attributed to the formation of nonvolatile intermediates UF 2 , UF 3 , UF 4 , UF 5 , UO 2 F, and UO 2 F 2 on the surface. Analysis indicated that the accumulation of UF 6 and, to a lesser extent, O 2 near the surface partially contributed to the exponential decrease in the UO 2 etch rate with immersion time. Unlike fluorination with F 2 gas, etching of UO 2 using rf glow discharge is possible below 663 K. The average etch rates of the amorphous UO 2 in the NF 3 experiments are comparable to the peak values reported in other studies for crystalline UO 2 using CF 4 /O 2 glow discharge performed at ∼150 to 250 K higher sample temperatures

  4. Improvement of cesium retention in uranium dioxide by additional phases; Amelioration de la retention du cesium dans le dioxyde d`uranium au moyen de phases exogenes

    Energy Technology Data Exchange (ETDEWEB)

    Gamaury Dubois, S

    1995-09-19

    The objective of this study is to improve the cesium retention in nuclear fuel. A bibliographic survey indicates that cesium is rapidly released from uranium dioxide in an accident condition. At temperatures higher than 1500 deg C or in oxidising conditions, our experiments show the difficulty of maintaining cesium inside simulated fuel. Two ternary systems are potentially interesting for the retention of cesium and to reduce the kinetics of release from the fuel: Cs{sub 2}O-Al{sub 2}O{sub 3}-SiO{sub 2} et Cs{sub 2}O-ZrO{sub 2}-SO{sub 2}. The compounds CsAISi{sub 2}O{sub 6} and Cs{sub 2}ZrSi{sub 6}O{sub 15} were studied from 1200 deg C to 2000 deg C by thermogravimetric analysis. The volumetric diffusion coefficients of cesium in these structures, in solid state as well as in liquid one, were measured. A fuel was sintered with (Al{sub 2}O{sub 3} + SiO{sub 2}) or (ZrO{sub 2} + SiO{sub 2}) and the intergranular phase was characterized. In the presence of (Al{sub 2}O{sub 3} + SiO{sub 2}), the sintering is realized at 1610 deg C in H{sub 2}. It is a liquid phase sintering. On the other end, with (ZrO{sub 2} + SiO{sub 2}), the sintering is a low temperature one in oxidising atmosphere. Finally, cesium containing simulated fuels were produced with these additives. According to the effective diffusion coefficients that were measured, the additives improved the retention of cesium. We have predicted the improvement that could be hoped for in a nuclear reactor, depending on the dispersion of the intergranular additives, the temperature and the degree of oxidation of the UO{sub 2+x}. We wait for a factor of 2 for x=0 and more than 8 for x=0.05, up to 2000 deg C. (author). 148 refs., 122 figs., 34 tabs.

  5. Heat capacity measurements and XPS studies on uranium-lanthanum mixed oxides

    International Nuclear Information System (INIS)

    Venkata Krishnan, R.; Mittal, V.K.; Babu, R.; Senapati, Abhiram; Bera, Santanu; Nagarajan, K.

    2011-01-01

    Research highlights: → Heat capacity measurements were carried out on (U 1-y La y )O 2±x (y = 0.2, 0.4, 0.6, and 0.8) using differential scanning calorimeter (DSC) in the temperature range 298-800 K. → Enthalpy increment measurements were carried out on the above solid solution using high temperature drop calorimetry in the temperature range 800-1800 K. → Chemical states of U and La in the solid solutions of mixed oxides were determined using X-ray photoelectron spectroscopy (XPS). → The anomalous increase in the heat capacity is attributed to certain thermal excitation process namely Frenkel pair defect of oxygen. → From the XPS investigation, it is observed that the O/M ratio at the surface is higher than that to the bulk. → In uranium rich mixed oxide samples, the surface O/M is greater than 2 whereas that in La rich mixed oxides, it is less than 2, though the bulk O/M in all the samples are less than 2. - Abstract: Heat capacity measurements were carried out on (U 1-y La y )O 2±x (y = 0.2, 0.4, 0.6, and 0.8) using differential scanning calorimeter (DSC) in the temperature range 298-800 K. Enthalpy increment measurements were carried out on the above solid solutions using high temperature drop calorimetry in the temperature range 800-1800 K. Chemical states of U and La in the solid solutions of mixed oxides were determined using X-ray photoelectron spectroscopy (XPS). Oxygen to metal ratios of (U 1-y La y )O 2±x were estimated from the ratios of different chemical states of U present in the sample. Anomalous increase in the heat capacity is observed for (U 1-y La y )O 2±x (y = 0.4, 0.6 and 0.8) with onset temperatures in the range of 1000-1200 K. The anomalous increase in the heat capacity is attributed to certain thermal excitation process, namely, Frenkel pair defect of oxygen. The heat capacity value of (U 1-y La y )O 2±x (y = 0.2, 0.4, 0.6, and 0.8) at 298 K are 65.3, 64.1, 57.7, 51.9 J K -1 mol -1 , respectively. From the XPS investigations

  6. Formalization of the kinetics for autocatalytic dissolutions. Focus on the dissolution of uranium dioxide in nitric medium

    International Nuclear Information System (INIS)

    Charlier, F.; Canion, D.; Gravinese, A.; Magnaldo, A.; Lalleman, S.; Borda, G.; Schaer, E.

    2017-01-01

    Uranium dioxide dissolution in nitric acid is a complex reaction. On the one hand, the dissolution produces nitrous oxides (NOX), which makes it a triphasic reaction. On the other hand, one of the products accelerates the kinetic rate; the reaction is hence called autocatalytic.The kinetics for these kinds of reactions need to be formalized in order to optimize and design innovative dissolution reactors. In this work, the kinetics rates have been measured by optical microscopy using a single particle approach. The advantages of this analytical technique are an easier management of species transport in solution and a precise following of the dissolution rate. The global rate is well described by a mechanism considering two steps: a non-catalyzed reaction, where the catalyst concentration has no influence on the dissolution rate, and a catalyzed reaction. The mass transfer rate of the catalyst was quantified in order to discriminate when the reaction was influenced by catalyst accumulated in the boundary layer or uncatalyzed. This first approximation described well the sigmoid dissolution curve profile. Moreover, experiments showed that solutions filled with catalyst proved to lose reactivity over time. Results pointed out that the higher the liquid-gas exchanges, the faster the kinetic rate decreases with time. Thus, it was demonstrated, for the first time, that there is a link between catalyst and nitrous oxides. The outcome of this study leads to new ways for improving the design of dissolvers. Gas-liquid exchanges are indeed a lever to impact dissolution rates. Temperature and catalyst concentration can be optimized to reduce residence times in dissolvers. (authors)

  7. Mixed Matrix Composite Membranes Containing POSS Molecules for Carbon Dioxide Removal Application

    KAUST Repository

    Rini, Eki Listya

    2011-05-10

    CO2 removal by membrane processes is considerably potential for several applications such as natural gas and synthesis gas purification, enhanced oil recovery application, and carbon dioxide capture in combat against global warming. Dense polymeric membranes are commonly utilized for these type of gas separation applications. Nevertheless, the intrinsic properties of dense polymeric membranes, which commonly characterize by the low gas permeability versus high gas selectivity trade–off or vice versa, is less desirable. In order to meet the increased demand of CO2 removal, a strategy to improve the gas separation performance of a polymeric membrane is investigated in this study. With this regard, mixed matrix membranes in which inorganic non porous fillers are incorporated into a polymeric matrix were prepared to achieve the aforementioned objective. The mixed matrix membranes were prepared from Pebax® block copolymers and PEG POSS® molecules. These hybrid membranes were formed as both dense and multilayer composite membranes. The dense transparent membranes with well–dispersed fillers could be obtained by variation of the solvent mixture. The DSC analyses showed that incorporation of PEG POSS® into Pebax® matrix altered the thermal properties of the matrix. The multilayer composite membranes were then prepared from a PTMSP gutter layer deposited on a PAN porous support and an adjacent hybrid Pebax®/PEG POSS® as the top layer. These hybrid multilayer composite membranes exhibited an enhanced CO2 selectiv4 ity by a factor of two relative to the pure Pebax®. In these hybrid systems, the CO2 separation was presumably enhanced by the high ether oxides content from PEG POSS® that has high affinities for CO2. For particular composition of Pebax® and PEG POSS® concentrations, the PTMSP gutter layer harnessed the CO2 selectivity without losing the CO2 permeation rate. At the same time, these membrane, however, suffered severe adhesion between the gutter layer

  8. The uranium and thorium separation in the chemical reprocessing of the irradiated fuel of thorium and uranium mixed oxides

    International Nuclear Information System (INIS)

    Oliveira, E.F. de.

    1984-09-01

    A bibliographic research has been carried out for reprocessing techniques of irradiated thorium fuel from nuclear reactors. The Thorex/Hoechst process has been specially considered to establish a method for reprocessing thorium-uranium fuel from PWR. After a series of cold tests performed in laboratory it was possible to set the behavior of several parameters affecting the Thorex/Hoechst process. Some comments and suggestions are presented for modifications in the process flosheet conditions. A discussion is carried out for operational conditions such as the aqueous to organic flow ratio the acidity of strip and scrub solutions in the process steps for thorium and uranium recovery. The operation diagrams have been constructed using equilibrium experimental data which correspond to conditions observed in laboratory. (Author) [pt

  9. Uranium

    International Nuclear Information System (INIS)

    Battey, G.C.; McKay, A.D.

    1988-01-01

    Production for 1986 was 4899 t U 3 O 8 (4154 t U), 30% greater than in 1985, mainly because of a 39% increase in production at Ranger. Exports for 1986 were 4166 t U 3 O 8 at an average f.o.b. unit value of $40.57/lb U 3 O 8 . Private exploration expenditure for uranium in Australia during the 1985-86 fiscal year was $50.2 million. Plans were announced to increase the nominal capacity of the processing plant at Ranger from 3000 t/year U 3 O 8 to 4500 t and later to 6000 t/year. Construction and initial mine development at Olympic Dam began in March. Production is planned for mid 1988 at an annual rate of 2000 t U 3 O 8 , 30 000 t Cu, and 90 000 oz (2800 kg) Au. The first long-term sales agreement was concluded in September 1986. At the Manyingee deposit, testing of the alkaline solution mining method was completed, and the treatment plant was dismantled. Spot market prices (in US$/lb U 3 O 8 ) quoted by Nuexco were generally stable. From January-October the exchange value fluctuated from US$17.00-US$17.25; for November and December it was US$16.75. Australia's Reasonably Assured Resources of uranium recoverable at less than US$80/kg U at December 1986 were estimated as 462 000 t U, 3000 t U less than in 1985. This represents 30% of the total low-cost RAR in the WOCA (World Outside the Centrally Planned Economy Areas) countries. Australia also has 257 000 t U in the low-cost Estimated Additional Resources Category I, 29% of the WOCA countries' total resources in this category

  10. Off gas processing device for degreasing furnace for uranium/plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    Ueda, Masaya; Akasaka, Takayuki; Noura, Takeshi.

    1996-01-01

    A low melting ingredient capturing-cooling trap connected to a degreasing sintering furnace by way of sealed pipelines, a burning/decomposing device for decomposing high melting ingredient gases discharged from the cooling trap by burning them and a gas sucking means for forming the flow of off gases are contained in a glovebox, the inside pressure of which is kept negative. Since the degreasing sintering furnace for uranium/plutonium mixed oxide fuels is disposed outside of the glovebox, operation can be performed safely without greatly increasing the scale of the device, and the back flow of gases is prevented easily by keeping the pressure in the inside of the glovebox negative. Further, a heater is disposed at the midway of the sealed pipelines from the degreasing sintering furnace to the cooling trap, the temperature is kept high to prevent deposition of low melting ingredients to prevent clogging of the sealed pipelines. Further, a portion of the pipelines is made extensible in the axial direction to eliminate thermal stresses caused by temperature change thereby enabling to extend the life of the sealed pipelines. (N.H.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-02-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  13. Present status of uranium-plutonium mixed carbide fuel development for LMFBRs

    International Nuclear Information System (INIS)

    Handa, Muneo; Suzuki, Yasufumi

    1984-01-01

    The feature of carbide fuel is that it has the doubling time as short as about 13 years, that is, close to one half as compared with oxide fuel. The development of the carbide fuel in the past 10 years has been started in amazement. Especially in the program of new fuel development in USA started in 1974, He and Na bond fuel attained the burnup of 16 a/o without causing the breaking of cladding tubes. In 1984, the irradiation of the assembly composed of 91 fuel pins in the FFTF is expected. On the other hand in Japan, the fuel research laboratory was constructed in 1974 in the Oarai Laboratory, Japan Atomic Energy Research Institute, to carry out the studies on carbide fuel. In the autumn of 1982, two carbide fuel pins with different chemical composition have been successfully made. Accordingly, the recent status of the development is explained. The uranium-plutonium mixed carbide fuel is suitable to liquid metal-cooled fast breeder reactors because of large heat conductivity and the high density of nuclear fission substances. The thermal and nuclear characteristics of carbide fuel, the features of the reactor core using carbide fuel, the chemical and mechanical interaction of fuel and cladding tubes, the selection of bond materials, the manufacturing techniques for the fuel, the development of the analysis code for fuel behavior, and the research and development of carbide fuel in Japan are described. (Kako, I.)

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

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

  16. Radiation enhanced thermal diffusion of chlorine in uranium dioxide; Diffusion thermique et sous irradiation du chlore dans le dioxyde d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Pipon, Yves [Ecole doctorale de physique et d' astrophysique, Universite Claude Bernard Lyon-I, Lyon (France)

    2006-12-15

    This work concerns the study of the thermal and radiation enhanced diffusion of {sup 36}Cl in uranium dioxide. It is a contribution to PRECCI programme (research programme on the long-term behaviour of the spent nuclear fuel). {sup 36}Cl is a long lived volatile activation product (T = 300 000 years) able to contribute significantly to the instant release fraction in geological disposal conditions. We simulated the presence of {sup 36}Cl by implanting a quantity of {sup 37}Cl comparable to the impurity content of chlorine in UO{sub 2}. In order to evaluate the diffusion properties of chlorine in the fuel and in particular to assess the influence of the irradiation defects, we performed two kinds of experiments: - the influence of the temperature was studied by carrying out thermal annealings in the temperature range 900 - 1300 deg. C; we showed that implanted chlorine was mobile from temperatures as low as 1000 deg. C and determined a thermal diffusion coefficient D{sub 1000} {sub deg.} {sub C} around 10{sup -16} cm{sup 2}s{sup -1} and deduced an activation energy of 4.3 eV. This value is one of lowest compared to that of volatile fission products such as iodine or the xenon. These parameters reflect the very mobile behaviour of chlorine; - the irradiation effects induced by fission products were studied by irradiating the samples with {sup 127}I (energy of 63.5 MeV). We showed that the implanted chlorine diffusion in the temperature range 30 - 250 deg. C is not purely athermal. In these conditions, the diffusion coefficient D{sub 250} {sub deg.} {sub C} for the implanted chlorine is around 10{sup -14} cm{sup 2}s{sup -1} and the activation energy is calculated to be 0.1 eV. Moreover, at 250 deg. C, we observed an important transport of the pristine chlorine from the bulk towards the surface. This chlorine comes from a zone where the defects are mainly produced by the nuclear energy loss process at the end of iodine range. We showed the importance of the

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  18. Determination of the O/M ratios of polynary uranium oxides by Ce(IV)-Fe(II) back titration after dissolution in mixed sulphuric and phosphoric acids

    International Nuclear Information System (INIS)

    Fujino, T.; Sato, N.; Yamada, K.

    1996-01-01

    Uranium (IV) in polynary uranium oxides is determined after the solid has been dissolved in a warm mixed solution of sulphuric and phosphoric acids containing excess Ce(IV). The latter is titrated with a Fe(II) standard solution using ferroin as indicator. This method is especially effective for (mixed) uranium oxides which are difficult to dissolve in hot Ce(IV) sulphuric acid. The standard deviation of the determined x value in polynary oxides is estimated to be below ± 0.004 for samples of 10-30 mg. (orig.)

  19. Determination of the O/M ratios of polynary uranium oxides by Ce(IV)-Fe(II) back titration after dissolution in mixed sulphuric and phosphoric acids.

    Science.gov (United States)

    Fujino, T; Sato, N; Yamada, K

    1996-01-01

    Uranium (IV) in polynary uranium oxides is determined after the solid has been dissolved in a warm mixed solution of sulphuric and phosphoric acids containing excess Ce(IV). The latter is titrated with a Fe(II) standard solution using ferroin as indicator. This method is especially effective for (mixed) uranium oxides which are difficult to dissolve in hot Ce(IV) sulphuric acid. The standard deviation of the determined x value in polynary oxides is estimated to be below +/- 0.004 for samples of 10-30 mg.

  20. Emanation of /sup 232/U daughter products from submicrometer particles of uranium oxide and thorium dioxide by nuclear recoil and inert gas diffusion

    Energy Technology Data Exchange (ETDEWEB)

    Coombs, M.A.; Cuddihy, R.G. (Lovelace Biomedical and Environmental Research Inst., Albuquerque, NM (USA). Inhalation Toxicology Research Inst.)

    1983-01-01

    Emanation of /sup 232/U daughter products by nuclear recoil and inert gas diffusion from spherical, submicrometer particles of uranium oxide and thorium dioxide was studied. Monodisperse samples of particles containing 1% /sup 232/U and having physical diameters between 0.1 and 1 ..mu..m were used for the emanation measurements. Thorium-228 ions recoiling from the particles after alpha-decay of /sup 232/U were collected electrostatically on a recoil cathode. Radon-220 diffusing from the particles was swept by an airstream into a 4 l. chamber where the /sup 220/Rn daughters were collected on a second cathode. Mathematical models of radionuclide emanation from spherical particles were used to calculate the recoil range of /sup 228/Th and the diffusion coefficient of /sup 220/Rn in the particle matrix. A /sup 228/Th recoil range of 0.02 ..mu..m and a /sup 220/Rn diffusion coefficient of 3 x 10/sup -14/ cm/sup 2//sec were obtained in both uranium oxide and thorium dioxide particles.

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

    Directory of Open Access Journals (Sweden)

    Marc Philippe

    2018-01-01

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

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

    Science.gov (United States)

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

    2018-03-01

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

  3. Study and simulation of the behaviour under irradiation of helium in uranium dioxide; Etude et modelisation du comportement sous irradiation de l'helium dans le dioxyde d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Martin, G

    2007-06-15

    Large quantities of helium are produced from {alpha}-decay of actinides in nuclear fuels during its in-pile operating and its storage. It is important to understand the behaviour of helium in these matrix in order to well simulate the evolution and the resistance of the fuel element. During this thesis, we have used nuclear reaction analyses (NRA) to follow the evolution of the helium implanted in polycrystalline and monocrystalline uranium dioxide (UO{sub 2}). An experimental rig was developed to follow the on-line helium release in UO{sub 2} and the evolution of {sup 3}He profiles as a function of annealing temperature. An automated procedure taking into account the evolution of the depth resolution was developed. Analyses performed with a nuclear microprobe allowed to characterise the spatial distribution of helium at the grain scale and to study the influence of the sample microstructure on the helium migration. This work put into evidence the particular role of grain boundaries and irradiation defects in the helium release process. The analyse of experimental results with a diffusion model corroborates these interpretations. It allowed to determine quantitatively physical properties that characterise the helium behaviour in uranium dioxide (diffusion coefficient, activation energy..). (author)

  4. The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

    Science.gov (United States)

    Syarifah, Ratna Dewi; Suud, Zaki

    2015-09-01

    Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

  5. The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

    International Nuclear Information System (INIS)

    Syarifah, Ratna Dewi; Suud, Zaki

    2015-01-01

    Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better

  6. Extraction of uranium from simulated ore by the supercritical carbon dioxide fluid extraction method with nitric acid-TBP complex

    International Nuclear Information System (INIS)

    Dung, Le Thi Kim; Imai, Tomoki; Tomioka, Osamu; Nakashima, Mikio; Takahashi, Kuniaki; Meguro, Yoshihiro

    2006-01-01

    The supercritical fluid extraction (SFE) method using CO 2 as a medium with an extractant of HNO 3 -tri-n-butyl phosphate (TBP) complex was applied to extract uranium from several uranyl phosphate compounds and simulated uranium ores. An extraction method consisting of a static extraction process and a dynamic one was established, and the effects of the experimental conditions, such as pressure, temperature, and extraction time, on the extraction of uranium were ascertained. It was found that uranium could be efficiently extracted from both the uranyl phosphates and simulated ores by the SFE method using CO 2 . It was thus demonstrated that the SFE method using CO 2 is useful as a pretreatment method for the analysis of uranium in ores. (author)

  7. Process for continuous production of metallic uranium and uranium alloys

    Science.gov (United States)

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

    1995-01-01

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

  8. Process for continuous production of metallic uranium and uranium alloys

    Science.gov (United States)

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

    1995-06-06

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

  9. Dissolution of thorium/uranium mixed oxide in nitric acid-hydrofluoric acid solution

    International Nuclear Information System (INIS)

    Filgueiras, S.A.C.

    1984-01-01

    The dissolution process of thorium oxide and mixed uranium-thorium oxide is studied, as a step of the head-end of the fuel reprocessing. An extensive bibliography was analysed, concerning the main aspects of the system, specially the most important process variables. Proposed mechanisms and models for the thorium oxide dissolution are presented. The laboratory tests were performed in two phases: at first, powdered thoria was used as the material to be dissolved. The objective was to know how changes in he concentrations of the dissolvent solution components HNO 3 , HF and Al(NO 3 ) 3 affect the dissolution rate. The tests were planned according to the fractional factorial method. Thes results showed that it is advantageous to work with powdered material, since the reaction occurs rapidly. And, if the Thorex solution (HNO 3 13M, HF 0.05M and Al(NO 3 ) 3 0.10M) is a suitable dissolvent, it was verified that it is possible to reduce the concentration of either nitric or fluoridric acid, without reducing the reaction rate to an undesirable value. It was also observed significant interaction between the components of the dissolvent solution. In the second phase of the tests, (Th, 5%U)O 2 sintered pellets were used. The main goals were to know the pellets dissolution behaviour and to compare the results for different pellets among themselves. It was observed that the metallurgical history of the material strongly influences its dissolution, specially the density and the microstructure. It was also studied how the (Th,U)O 2 mass/Thorex solution volume ratio affects the time needed to obtain an 1 M Th/liter solution. The activation energy for the reaction was obtained. (Author) [pt

  10. Expanding the family of uranium(III) alkyls. Synthesis and characterization of mixed-ligand derivatives

    Energy Technology Data Exchange (ETDEWEB)

    Matson, Ellen M.; Kiernicki, John J.; Fanwick, Phillip E.; Bart, Suzanne C. [Department of Chemistry, Purdue University, West Lafayette, IN (United States)

    2016-06-15

    The generation of uranium(III) alkyls supported by hydrotris(pyrazolyl)borate (Tp) and pentamethylcyclopentadienyl (Cp*) ligands is reported. Mixed ancillary ligand frameworks were synthesized by treating TpUI{sub 2}(THF){sub 3} (1) and Cp*UI{sub 2}(THF){sub 3} with potassium hydrotris(pyrazolyl)borate salts. Addition of one equivalent of potassium hydrotris(3,5-dimethylpyrazolyl)borate (Tp*) generated TpTp*UI (2), while treatment of Cp*UI{sub 2}(THF){sub 3} with either KTp or KTp* resulted in the respective formation of Cp*TpUI(THF) (3) or Cp*Tp*UI(THF) (4). Alkylation of 2 with KCH{sub 2}Ph or NaCH{sub 2}SiMe{sub 3} furnished TpTp*UCH{sub 2}Ph (2-CH{sub 2}Ph) or TpTp*UCH{sub 2}SiMe{sub 3} (2-CH{sub 2}SiMe{sub 3}). Similarly, treatment of 3 with NaCH{sub 2}SiMe{sub 3} formed Cp*TpUCH{sub 2}SiMe{sub 3} (3-CH{sub 2}SiMe{sub 3}), whereas treatment of 4 with KCH{sub 2}Ph generated Cp*Tp*UCH{sub 2}Ph (4-CH{sub 2}Ph). All compounds were characterized by multinuclear NMR, IR, and electronic absorption spectroscopy. Compounds 2-CH{sub 2}Ph, 3, and 3-CH{sub 2}SiMe{sub 3} were structurally characterized using X-ray crystallography as well. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  11. Synthesis, sintering and dissolution of thorium and uranium (IV) mixed oxide solid solutions: influence of the method of precursor preparation; Synthese, frittage et caracterisation de solutions solides d'oxydes mixtes de thorium et d'uranium (IV): influence de la methode de preparation du precurseur

    Energy Technology Data Exchange (ETDEWEB)

    Hingant, N

    2008-12-15

    Mixed actinide dioxides are currently considered as potential fuels for the third and fourth generations of nuclear reactors. In this context, thorium-uranium (IV) dioxide solid solutions were studied as model compounds to underline the influence of the method of preparation on their physico-chemical properties. Two methods of synthesis, both based on the initial precipitation of oxalate precursors have been developed. The first consisted in the direct precipitation ('open' system) while the second involved hydrothermal conditions ('closed' system). The second method led to a significant improvement in the crystallization of the samples especially in the field of the increase of the grain size. In these conditions, the formation of a complete solid solution Th{sub 1-x}U{sub x}(C{sub 2}O{sub 4}){sub 2}.2H{sub 2}O was prepared between both end-members. Its crystal structure was also resolved. Whatever the initial method considered, these compounds led to the final dioxides after heating above 400 C. The various steps associated to this transformation, involving the dehydration of precursors then the decomposition of oxalate groups have been clarified. Moreover, the use of wet chemistry methods allowed to reduce the sintering temperature of the final thorium-uranium (IV) dioxide solid solutions. Whatever the method of preparation considered, dense samples (95% to 97% of the calculated value) were obtained after only 3 hours of heating at 1500 C. Additionally, the use of hydrothermal conditions significantly increased the grain size, leading to the reduction of the occurrence of the grain boundaries and of the global residual porosity. The significant improvement in the homogeneity of cations distribution in the samples was also highlighted. Finally, the chemical durability of thorium-uranium (IV) dioxide solid solutions was evaluated through the development of leaching tests in nitric acid. The optimized homogeneity especially in terms of the

  12. Processing Pa-Lua Uranium ore by Mixing and Curing with Sulfuric Acid on a Scale of 500 kg/Batch to Recover Yellowcake

    International Nuclear Information System (INIS)

    Le Quang Thai; Cao Hung Thai; Le Thi Kim Dung; Phung Vu Phong; Tran Van Son

    2007-01-01

    Uranium ore in Pa-Lua area is sandstone with different levels of weathering. This kind of ore contains calcium and clay that may cause clogs during heap leaching. In this study, a technique of mixing and curing with strong acids is used and followed by washing to recover uranium. This study also focuses on study of ore processing issues such as crushing, regenerating particles in fine ores, mixing, curing and washing. The leach solution is treated by ion-exchange and precipitation of products by NH 4 OH. The experiment results show that regenerating a portion of fine ores, mixing and curing help washing residues in the column more effectively. Flow rate of the input solution can be controllable and stable. Columns do not clog even when washing takes place in the ore column of 5 meters high. Efficiency of uranium recovery can reach to 85-90%. Products of technical uranium are obtained with high quality. (author)

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2005-01-01

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

  14. Sorption behaviour of uranium and thorium on hydrous tin oxide from aqueous and mixed-solvent HNO3 media

    International Nuclear Information System (INIS)

    Misak, N.Z.; Salama, H.N.; El-Naggar, I.M.

    1983-01-01

    In aqueous nitric acid, uranyl and thorium ions seem to be sorbed on hydrous tin oxide mainly by a cation exchange mechanism. In 10 - 3 M aqueous solutions, the hydrous oxide prefers thorium to uranium at the relative low pH values, while the reverse is true at the higher pH values. The exchange of uranium is particle diffusion controlled while that of thorium is chemically controlled, and the isotherms point to the presence of different-energy sites in the hydrous oxide. Except for the solutions containing 80% of methanol, ethanol, or acetone, cation exchange is probably still the main mechanism of sorption of uranium. Anionic sorption of thorium seems to occur in all the mixed-solvent solutions and is perhaps the main mechanism in 80% ethanol. The equilibrium distribution coefficient K sub (d) increases almost in all cases with organic solvent content, probably due to dehydration of sorbed ions and to increasing superposition on anionic sorption. Unlike the aqueous medium, large U/Th separation factors are achieved in many of the mixed-solvent solutions and separation schemes are suggested. (Authors)

  15. Uranium dioxide Caramel fuel

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    The work performed in France on Caramel fuels for research reactors reflects the reality of a program based on non proliferation criteria, as they have already appeared several years ago. This work actually includes the following different aspects: identification of the non proliferation criterion defining this action; determination of the economical and technical goals to be reached; realization of research and development studies finalized in a full scale demonstration; transposition to an industrial and commercial level

  16. The electrical and thermal properties of sodium sulfate mixed with lithium sulfate, yttrium sulfate, and silicon dioxide

    International Nuclear Information System (INIS)

    Imanaka, N.; Yamaguchi, Y.; Adachi, G.; Shiokawa, J.

    1986-01-01

    Sodium sulfate mixed with lithium sulfate, yttrium sulfate, and silicon dioxide was prepared. The thermal and electrical properties of its phases were investigated. The Na 2 SO 4 -Li 2 SO 4 -Y 2 (SO 4 ) 3 SiO 2 samples are similar to the Na 2 SO 4 -I phase (a high temperature phase), which is appreciably effective for Na + ionic conduction. Phase transformation was considerably suppressed by mixing. Electromotive force (EMF) was measured, using Na 2 SO 4 -Li 2 SO 4 -Y 2 (SO 4 ) 3 -SiO 2 as a solid electrolyte, by constructing an SO 2 gas concentration cell. The measured EMF's at 823 and 773 K were in fairly good accordance with the calculated EMF's for inlet SO 2 gas concentration between 30 ppm and 1%, and 500 ppm and 0.5% respectively

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

    International Nuclear Information System (INIS)

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

    1980-05-01

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

  18. The hydrolysis of thorium dicarbide and of mixed uranium-thorium dicarbides

    International Nuclear Information System (INIS)

    Del Litto, B.

    1966-09-01

    The hydrolysis of thorium dicarbide leads to the formation of a complex mixture of gaseous and condensed carbon hydrides. The temperature, between 25 and 100 deg. C, has no influence on the nature and composition of the gas phase. The reaction kinetics, however, are strongly temperature dependent. In a hydrochloric medium, an enrichment in hydrogen of the gas mixture is observed. On the other hand a decrease in hydrogen and an increase in acetylene content take place in an oxidizing medium. The general results can be satisfactorily interpreted through a reaction mechanism involving C-C radical groups. In the same way, the hydrolysis of uranium-thorium-carbon ternary alloys leads to the formation of gaseous and condensed carbon hydrides. The variation of the composition of the gas phase versus uranium content in the alloy suggests an hypothesis about the carbon-carbon distance in the alloy crystal lattice. The variation of methane content, on the other hand, has lead us to discuss the nature of the various phases present in uranium-carbon alloys and carbon-rich uranium-thorium-carbon alloys. We have reached the conclusion that these alloys include a proportion of monocarbide which is dependent upon the ratio. Th/(Th + U). We put forward a diagram of the system uranium-carbon with features proper to explain some phenomena which have been observed in the uranium-thorium-carbon ternary diagram. (author) [fr

  19. Forecasting carbon dioxide emissions based on a hybrid of mixed data sampling regression model and back propagation neural network in the USA.

    Science.gov (United States)

    Zhao, Xin; Han, Meng; Ding, Lili; Calin, Adrian Cantemir

    2018-01-01

    The accurate forecast of carbon dioxide emissions is critical for policy makers to take proper measures to establish a low carbon society. This paper discusses a hybrid of the mixed data sampling (MIDAS) regression model and BP (back propagation) neural network (MIDAS-BP model) to forecast carbon dioxide emissions. Such analysis uses mixed frequency data to study the effects of quarterly economic growth on annual carbon dioxide emissions. The forecasting ability of MIDAS-BP is remarkably better than MIDAS, ordinary least square (OLS), polynomial distributed lags (PDL), autoregressive distributed lags (ADL), and auto-regressive moving average (ARMA) models. The MIDAS-BP model is suitable for forecasting carbon dioxide emissions for both the short and longer term. This research is expected to influence the methodology for forecasting carbon dioxide emissions by improving the forecast accuracy. Empirical results show that economic growth has both negative and positive effects on carbon dioxide emissions that last 15 quarters. Carbon dioxide emissions are also affected by their own change within 3 years. Therefore, there is a need for policy makers to explore an alternative way to develop the economy, especially applying new energy policies to establish a low carbon society.

  20. Haematological malignancies in childhood in Croatia: Investigating the theories of depleted uranium, chemical plant damage and 'population mixing'

    International Nuclear Information System (INIS)

    Labar, B.; Rudan, I.; Ivankovic, D.; Biloglav, Z.; Mrsic, M.; Strnad, M.; Fucic, A.; Znaor, A.; Bradic, T.; Campbell, H.

    2004-01-01

    Some of potential causes proposed to explain the reported increase of haematological malignancies in childhood during or after the war period in several countries include depleted uranium, chemical pollution and population mixing theory. The aim of this study was to define the population of Croatian children aged 0-14 years who were potentially exposed to each of those risks during the war and to investigate any possible association between the exposure and the incidence of haematological malignancies. The authors analyzed the data reported by the Cancer Registry of Croatia during the pre-war period (1986-1990), war period (1991-1995) and post-war period (1996-1999). In the group of 10 counties potentially exposed to depleted uranium and two counties where chemical war damage occurred, no significant difference in incidence of the studied haematological malignancies was noted in comparison to pre-war period. The incidence of lymphatic leukaemia significantly increased in four counties where population mixing had occurred during the war period, supporting the 'mixing theory'. In those counties, the incidence of Hodgkin's lymphoma decreased during and after the war. In Croatia as a whole, decreases in incidence of myeloid leukaemias during war and non-Hodgkin lymphoma after the war were noted

  1. Thermal conductivity of beginning-of-life uranium-plutonium mixed oxide fuel for fast reactor (Interim report)

    International Nuclear Information System (INIS)

    Inoue, Masaki; Mizuno, Tomoyasu; Asaga, Takeo

    1997-11-01

    Thermal conductivity of uranium-plutonium mixed oxide fuel for fast reactor at beginning-of-life was correlated based on the recent results in order to apply to the fuel design and the fuel performance analysis. A number of experimental results of unirradiated fuel specimens were corrected from open literatures and PNC internal reports and examined for the database. In this work two porosity correction factors were needed for high density fuel and low density fuel (around the current Monju specification). The universal porosity correction factor was not determined in this work. In the next step, theoretical and analytical considerations should be taken into account. (J.P.N.)

  2. Fabrication of uranium-plutonium mixed nitride fuel pins (88F-5A) for first irradiation test at JMTR

    International Nuclear Information System (INIS)

    Suzuki, Yasufumi; Iwai, Takashi; Arai, Yasuo; Sasayama, Tatsuo; Shiozawa, Ken-ichi; Ohmichi, Toshihiko; Handa, Muneo

    1990-07-01

    A couple of uranium-plutonium mixed nitride fuel pins was fabricated for the first irradiation tests at JMTR for the purpose of understanding the irradiation behavior and establishing the feasibility of nitride fuels as advanced FBR fuels. The one of the pins was fitted with thermocouples in order to observe the central fuel temperature. In this report, the fabrication procedure of the pins such as pin design, fuel pellet fabrication and characterizations, welding of fuel pins, and inspection of pins are described, together with the outline of the new TIG welder installed recently. (author)

  3. Impact of Sundarban mangrove biosphere on the carbon dioxide and methane mixing ratios at the NE Coast of Bay of Bengal, India

    International Nuclear Information System (INIS)

    Mukhopadhyay, S.K.; Biswas, H.; De, T.K.; Jana, T.K.; Sen, B.K.; Sen, S.

    2002-01-01

    Diurnal and seasonal variations in carbon dioxide and methane fluxes between Sundarban biosphere and atmosphere were measured using micrometeorological method during 1998-2000. Study of the diurnal variation of micrometeorological conditions in the atmosphere was found to be necessary to determine the duration of neutral stability when flux estimation was reliable. Neutral stability of the atmosphere occurred in the limited micrometeorological conditions, when friction velocity ranged between 0.360 and 0.383ms -1 . The value of drag coefficient (1.62-20.6) x 10 3 obtained at variable wind speed could be deemed specific for this particular surface. 58.2% drop of carbon dioxide and 63.4% drop of methane in the atmosphere at 1m height were observed during day time, between dawn and early evening. Diurnal variations in methane and carbon dioxide mixing ratios showed a positive correlation with Richardson's number (Ri). This environment acted as a net source for carbon dioxide and methane. The mixing ratios of methane were found to vary between 1.42 and 2.07ppmv, and that of carbon dioxide, between 324.3 and 528.7ppmv during the study period. The biosphere-atmosphere flux of carbon dioxide ranged between - 3.29 and 34.4mgm -2 s -1 , and that of methane, between - 4.53 and 8.88μgm -2 s -1 . The overall annual estimate of carbon dioxide and methane fluxes from this ecosystem to atmosphere were estimated to be 694Tgyr -1 and 184Ggyr -1 , respectively. Considerable variations in mixing ratios of carbon dioxide and methane at the NE coast of Bay of Bengal were observed due to the seasonal variations of their fluxes from the biosphere to the atmosphere. The composition was inferred by fitting model prediction to measurements. (Author)

  4. Exact Solution of Fractional Diffusion Model with Source Term used in Study of Concentration of Fission Product in Uranium Dioxide Particle

    International Nuclear Information System (INIS)

    Fang Chao; Cao Jianzhu; Sun Lifeng

    2011-01-01

    The exact solution of fractional diffusion model with a location-independent source term used in the study of the concentration of fission product in spherical uranium dioxide (UO 2 ) particle is built. The adsorption effect of the fission product on the surface of the UO 2 particle and the delayed decay effect are also considered. The solution is given in terms of Mittag-Leffler function with finite Hankel integral transformation and Laplace transformation. At last, the reduced forms of the solution under some special physical conditions, which is used in nuclear engineering, are obtained and corresponding remarks are given to provide significant exact results to the concentration analysis of nuclear fission products in nuclear reactor. (nuclear physics)

  5. Ex-reactor determination of thermal gap and contact conductance between uranium dioxide: zircaloy-4 interfaces. Stage I: low gas pressure. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Garnier, J.E.; Begej, S.

    1979-04-01

    A study of thermal gap and contact conductance between depleted uranium dioxide (UO/sub 2/) and Zircaloy-4 (Zr4) has been made utilizing two measurement apparatuses developed as part of this program. The Modified Pulse Design (MPD) apparatus is a transient technique employing a heat pulse (laser) and a signal detector to monitor the thermal energy transmitted through a UO/sub 2//Zr4 sample pair which are either physically separated or in contact. The Modified Longitudinal Design (MLD) apparatus is a steady-state technique based on a modified cylindrical column design with a self-guarding sample geometry. Description of the MPD and MLD apparatus, data acquisition, reduction and error analysis is presented along with information on specimen preparation, thermal property and surface characterization. A technique using an optical height gauge to determine the average mean-plane of separation between the simple pairs is also presented.

  6. Ex-reactor determination of thermal gap and contact conductance between uranium dioxide: zircaloy-4 interfaces. Stage I: low gas pressure

    International Nuclear Information System (INIS)

    Garnier, J.E.; Begej, S.

    1979-04-01

    A study of thermal gap and contact conductance between depleted uranium dioxide (UO 2 ) and Zircaloy-4 (Zr4) has been made utilizing two measurement apparatuses developed as part of this program. The Modified Pulse Design (MPD) apparatus is a transient technique employing a heat pulse (laser) and a signal detector to monitor the thermal energy transmitted through a UO 2 /Zr4 sample pair which are either physically separated or in contact. The Modified Longitudinal Design (MLD) apparatus is a steady-state technique based on a modified cylindrical column design with a self-guarding sample geometry. Description of the MPD and MLD apparatus, data acquisition, reduction and error analysis is presented along with information on specimen preparation, thermal property and surface characterization. A technique using an optical height gauge to determine the average mean-plane of separation between the simple pairs is also presented

  7. Method to determine the thermal conductivity of uranium dioxide and the surface conductance at the cladding-core interface from internal reactions

    Energy Technology Data Exchange (ETDEWEB)

    Tsykanov, V A; Samsonov, B F; Spiridonov, Yu G; Fomin, N A

    1975-01-01

    A method is given for determining the temperature-dependent thermal conductivity of uranium dioxide and the contact conductance of the gas gap between the core and cladding of a fuel element. These quantities should be determined on various samples with different diameters. A method is described for determining the heat-production rate of a fuel element to within 1.5 to 2.5 percent. The method is based on using a calibrated electric heater and a sensor to measure the specific energy evolution from reactor gamma-radiation. The total errors in determining the thermal conductivity and the contact conductance do not exceed 4.5 and 8 percent, respectively.

  8. Determination of thermodynamic properties and stability limit from fluorite phase of uranium and lanthanide mixed oxides, using galvanic cells with solid electrolytes

    International Nuclear Information System (INIS)

    Santiago, T.N.

    1980-10-01

    A method for thermodynamic properties determination for oxygen solubility in oxide systems at temperature interval 973 ≤ T [K] ≤ 1773 is described. A galvanic cell using as solid electrolytes zircon dioxide doped with 15% of calcium oxide is presented. This method was used for determining the phase change, temperature dependent, of uranium-lanthanides-oxygen Ln U O 4 stoichiometric system. (C.G.C.)

  9. Carbon dioxide selective mixed matrix composite membrane containing ZIF-7 nano-fillers

    KAUST Repository

    Li, Tao; Pan, Yichang; Peinemann, Klaus-Viktor; Lai, Zhiping

    2013-01-01

    Mixed matrix materials made from selective inorganic fillers and polymers are very attractive for the manufacturing of gas separation membranes. But only few of these materials could be manufactured into high-performance asymmetric or composite

  10. Thermoeconomic Optimization of Cascade Refrigeration System Using Mixed Carbon Dioxide and Hydrocarbons at Low Temperature Circuit

    Directory of Open Access Journals (Sweden)

    Nasruddin Nasruddin

    2016-12-01

    Full Text Available Many applications and industrial processes require very low cooling temperature, such as cold storage in the biomedical field, requiring temperature below -80 °C. However,single-cycle refrigeration systems can only achieve the effective cooling temperature of -40 °C and, also, the performance of the cycle will decrease drastically for cooling temperatures lower than -35°C. Currently, most of cascade refrigeration systems use refrigerants that have ozone depletion potential (ODP and global warming potential (GWP, therefore, in this study, a cascade system is simulated using a mixture of environmentally friendly refrigerants, namely, carbon dioxide and a hydrocarbon (propane, ethane or ethylene as the refrigerant of the low temperature circuit. A thermodynamic analysis is performed to determine the optimal composition of the mixture of carbon dioxide and hydrocarbons in the scope of certain operating parameters. In addition, an economic analysis was also performed to determine the annual cost to be incurred from the cascade refrigeration system. The multi-objective/thermoeconomic optimization points out optimal operating parameter values of the system, to addressing both exergy efficiency and its relation to the costs to be incurred.

  11. Comparison of physical chemical properties of powders and respirable aerosols of industrial mixed uranium and plutonium oxide fuels

    International Nuclear Information System (INIS)

    Eidson, A.F.

    1982-01-01

    Studies were performed to characterize physical and chemical properties which may be important in determining the metabolism of accidentally released, inhaled aerosols of industrial mixed uranium and plutonium oxide fuels and to compare the properties of bulk powders and the respirable fraction they include. X-ray diffraction measurements showed that analysis of mixed-oxide powders from four process steps served to characterize their respirable fractions. IR spectroscopy was useful as a method to detect organic binders that were not observed by X-ray diffraction methods. Both X-ray diffraction and IR spectroscopy methods can be used in combination to identify the sources of a complex aerosol that might be released from more than one fabrication step. Isotopic distributions in powders and aerosols showed that information important for radiation dose to tissue calculations or Pu lung burden estimates can be obtained by analysis of powders. (U.K.)

  12. Criticality analysis for mixed thorium-uranium fuel in the Angra-2 PWR reactor using KENO-VI

    Energy Technology Data Exchange (ETDEWEB)

    Wichrowski, Caio C.; Gonçalves, Isadora C.; Oliveira, Claudio L.; Vellozo, Sergio O.; Baptista, Camila O., E-mail: wichrowski@ime.eb.br, E-mail: isadora.goncalves@ime.eb.br, E-mail: d7luiz@yahoo.com.br, E-mail: vellozo@ime.eb.br, E-mail: camila.oliv.baptista@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Seção de Engenharia Nuclear

    2017-07-01

    The increasing energy demand associated to the current sustainability challenges have given the thorium nuclear fuel cycle renewed interest in the scientific community. Studies have focused on energy production in different reactor designs through the fission of uranium 233, the product of thorium fertilization by neutrons. In order to make it possible for near future applications a strategy based on the adaptation of current nuclear reactors for the use of thorium fuels is being considered. In this work, bearing in mind these limitations, a code was used to evaluate the effect on criticality (k{sub inf}) of the mixing of thorium and uranium in different proportions in the fuel of a PWR, the German designed Angra-2 Brazilian reactor in order to scrutinise its behaviour and determine the feasibility of an adapted ThO{sub 2}-UO{sub 2} mixed fuel cycle using current PWR technology. The analysis is performed using the KENO-VI module in the SCALE 6.1 nuclear safety analysis simulation code and the information is taken from the Angra-2 FSAR (Final Security Analysis Report). (author)

  13. Mixed-ligand complexes of zirconium (IV) and uranium with salicylaldehyde and some heterocyclic azopyrazolones

    International Nuclear Information System (INIS)

    Amrallahi, A.H.

    1994-01-01

    Mixed-ligand complexes of Zr(IV) and U(IV) with salicylaldehyde (SA) and some heterocyclic azopyrazolenes have been studied spectrophotometrically. All formed chelates have ratio 1:1:1. The stoichiometry and stability of the binary mixed chelates been evaluated. Elemental analysis, molar conductance and IR spectra have been used for identification of solid mixed complexes. (author)

  14. Carbon dioxide selective adsorption within a highly stable mixed-ligand Zeolitic Imidazolate Framework

    KAUST Repository

    Huang, Lin; Xue, Ming; Song, Qingshan; Chen, Siru; Pan, Ying; Qiu, Shilun

    2014-01-01

    A new mixed-ligand Zeolitic Imidazolate Framework Zn4(2-mbIm) 3(bIm)5·4H2O (named JUC-160, 2-mbIm = 2-methylbenzimidazole, bIm = benzimidazole and JUC = Jilin University China) was synthesized with a solvothermal reaction of Zn(NO3) 2·6H2O, b

  15. Irradiation behaviour of mixed uranium-plutonium carbides, nitrides and carbonitrides; Comportement a l'irradiation de carbures, nitrures et carbonitrures mixtes d'uranium et de plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Mikailoff, H; Mustelier, J P; Bloch, J; Leclere, J; Hayet, L [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1967-07-01

    In the framework of the research program of fast reactor fuels two irradiation experiments have been carried out on mixed uranium-plutonium carbides, nitrides and carbo-nitrides. In the first experiment carried out with thermal neutrons, the fuel consisted of sintered pellets sheathed in a stainless steel can with a small gap filled with helium. There were three mixed mono-carbide samples and the maximum linear power was 715 W/cm. After a burn-up slightly lower than 20000 MW day/tonne, a swelling of the fuel which had ruptured the cans was observed. In the second experiment carried out in the BR2 reactor with epithermal neutrons, the samples consisted of sintered pellets sodium bonded in a stainless steel tube. There were three samples containing different fuels and the linear power varies between 1130 and 1820 W/cm. Post-irradiation examination after a maximal burn-up of 1550 MW day/tonne showed that the behaviour of the three fuel elements was satisfactory. (authors) [French] Dans le cadre du programme d'etude des conibustiles pour reacteurs rapides, on a realise deux experiences d'irradiation de carbures, nitrures et carbonitrures mixtes d'uranium et de plutonium. Dans la premiere experience, faite en neutrons thermiques, le combustible etait constitue de,pastilles frittees gainees dans un tube d'acier inoxydable avec un faible jeu rempli d'helium. Il y avait trois echantillons de monocarbures mixtes, et la puissance lineaire maximale etait de 715 W/cm. Apres un taux de combustion legerement inferieur a 20 000 MWj/t, on a observe un gonflement des combustible qui a provoque, la rupture des gaines. Pans la seconde experience, realisee dans le reacteur BR2 en neutrons epithermiques, les echantillons etaient constitues de pastilles frittees gainees dans un tube d'acier avec un joint sodium. Il y avait trois echantillons contenant des combustibles differents, et la puissance lineaire variait de 1130 a 1820 W/cm. Les examens apres irradiation a un taux maximal de

  16. Carbon dioxide selective mixed matrix composite membrane containing ZIF-7 nano-fillers

    KAUST Repository

    Li, Tao

    2013-01-01

    Mixed matrix materials made from selective inorganic fillers and polymers are very attractive for the manufacturing of gas separation membranes. But only few of these materials could be manufactured into high-performance asymmetric or composite membranes. We report here the first mixed matrix composite membrane made of commercially available poly (amide-b-ethylene oxide) (Pebax®1657, Arkema) mixed with the nano-sized zeolitic imidazole framework ZIF-7. This hybrid material has been successfully deposited as a thin layer (less than 1μm) on a porous polyacrylonitrile (PAN) support. An intermediate gutter layer of PTMSP was applied to serve as a flat and smooth surface for coating to avoid polymer penetration into the porous support. Key features of this work are the preparation and use of ultra-small ZIF-7 nano-particles (around 30-35nm) and the membrane processability of Pebax®1657. SEM pictures show that excellent adhesion and almost ideal morphology between the two phases has been obtained simply by mixing the as-synthesized ZIF-7 suspension into the Pebax®1657 dope, and no voids or clusters can be observed. The performance of the composite membrane is characterized by single gas permeation measurement of CO2, N2 and CH4. Both, permeability (PCO2 up to 145barrer) and gas selectivity (CO2/N2 up to 97 and CO2/CH4 up to 30) can be increased at low ZIF- loading. The CO2/CH4 selectivity can be further increased to 44 with the filler loading of 34wt%, but the permeability is reduced compared to the pure Pebax®1657 membrane. Polymer chain rigidification at high filler loading is supposed to be a reason for the reduced permeability. The composite membranes prepared in this work show better performance in terms of permeance and selectivity when compared with asymmetric mixed matrix membranes described in the recent literature. Overall, the ZIF 7/Pebax mixed matrix membranes show a high performance for CO2 separation from methane and other gas streams. They are easy to

  17. Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

    Science.gov (United States)

    Bowles, K. J.; Gluyas, R. E.

    1975-01-01

    The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

  18. Impact of uranium-233/thorium cycle on advanced accountability concepts and fabrication facilities. Addendum 2 to application of advanced accountability concepts in mixed oxide fabrication

    International Nuclear Information System (INIS)

    Bastin, J.J.; Jump, M.J.; Lange, R.A.; Crandall, C.C.

    1977-11-01

    The Phase I study of the application of advanced accountability methods (DYMAC) in a uranium/plutonium mixed oxide facility was extended to cover the possible fabrication of uranium-233/thorium fuels. Revisions to Phase II of the DYMAC plan which would be necessitated by such a process are specified. These revisions include shielding requirements, measurement systems, licensing conditions, and safeguards considerations. The impact of the uranium/thorium cycle on a large-scale fuel fabrication facility was also reviewed; it was concluded that the essentially higher radioactivity of uranium/thorium feeds would lead to increased difficulties which tend to preclude early commercial application of the process. An amended schedule for Phase II is included

  19. Studies on the absorption of uranium and plutonium on macroporous anion-exchange resins from mixed solvent media

    International Nuclear Information System (INIS)

    Chetty, K.V.; Mapara, P.M.; Godbole, A.G.; Swarup, Rajendra

    1995-01-01

    The ion-exchange studies on uranium and plutonium using macroporous anion-exchange resins from an aqueous-organic solvent mixed media were carried out to develop a method for their separation. Out of the several water miscible organic solvents tried, methanol and acetone were found to be best suited. Distribution data for U(VI) and Pu(IV) for three macroporous resins Tulsion A-27(MP) (strong base), Amberlyst A-26(MP) (strong base) and Amberlite XE-270(MP) (weak base) as a function of (i) nitric acid concentration (ii) organic solvent concentration were obtained. Based on the data separation factors for Pu/U were calculated. Column experiments using Tulsion A-27(MP) from a synthetic feed (HNO 3 - methanol and HNO 3 - acetone) containing Pu and U in different ratios were carried out. Plutonium was recovered from the bulk of the actual solution generated during the dissolution of plutonium bearing fuels. The method has the advantage of loading plutonium from as low as 1M nitric acid in presence of methanol or acetone and could be used satisfactorily for its recovery from solutions containing plutonium and uranium. (author). 11 refs., 4 figs., 16 tabs

  20. Late-occurring pulmonary pathologies following inhalation of mixed oxide (uranium + plutonium oxide) aerosol in the rat.

    Science.gov (United States)

    Griffiths, N M; Van der Meeren, A; Fritsch, P; Abram, M-C; Bernaudin, J-F; Poncy, J L

    2010-09-01

    Accidental exposure by inhalation to alpha-emitting particles from mixed oxide (MOX: uranium and plutonium oxide) fuels is a potential long-term health risk to workers in nuclear fuel fabrication plants. For MOX fuels, the risk of lung cancer development may be different from that assigned to individual components (plutonium, uranium) given different physico-chemical characteristics. The objective of this study was to investigate late effects in rat lungs following inhalation of MOX aerosols of similar particle size containing 2.5 or 7.1% plutonium. Conscious rats were exposed to MOX aerosols and kept for their entire lifespan. Different initial lung burdens (ILBs) were obtained using different amounts of MOX. Lung total alpha activity was determined by external counting and at autopsy for total lung dose calculation. Fixed lung tissue was used for anatomopathological, autoradiographical, and immunohistochemical analyses. Inhalation of MOX at ILBs ranging from 1-20 kBq resulted in lung pathologies (90% of rats) including fibrosis (70%) and malignant lung tumors (45%). High ILBs (4-20 kBq) resulted in reduced survival time (N = 102; p inhalation result in similar risk for development of lung tumors as compared with industrial plutonium oxide.

  1. A contribution to the study of the mixed uranium-plutonium mono-carbides containing small quantities of zirconium

    International Nuclear Information System (INIS)

    Bocker, S.

    1970-03-01

    We have studied a mixed monocarbide, type (U,Pu)C, containing small additions of zirconium for the application as a fast neutron reactor fuel. A preliminary study was conducted on the (U,Zr)C monocarbide (Report CEA-R-3765(1). It was found that small additions of zirconium to the uranium-plutonium monocarbide improve a number of properties such as atmospheric corrosion, the hardness, and particularly the compatibility with 316 stainless steel. However, properties such as the coefficient of expansion and the melting point are only slightly changed. The relative percentage of Pu/U+Pu in the monocarbide was fixed at 20 per cent. Two processes of fabrication were employed: casting in an arc furnace, sintering, carried out after having the hydrides of the metals carburized. The metallurgical results indicate, that the above mentioned fuel might be of interest for fast neutron reactor application. (author) [fr

  2. Theoretical studies on the stopping power of deuterium-tritium mixed with uranium plasmas for α particles

    International Nuclear Information System (INIS)

    Wang, Zhigang; Fu, Zhen-Guo; Zhang, Ping

    2014-01-01

    The stopping power of a compressed and highly ionized deuterium-tritium (DT) and uranium (U) plasma for α particles at very high temperatures (T = 5 keV) is examined theoretically with the dimensional continuation method. We show that with increasing density of U, both the magnitude and width of the resonance peak in the stopping power (as a function of the α particle energy), increases because of the ions, while the penetration distance of the α particles decreases. A simple relation of decreasing penetration distance as a function of plasma density is observed, which may be useful for inertial confinement fusion experiments. Moreover, by comparing the results with the case of a DT plasma mixed with beryllium, we find that the effect of a higher Z plasma is stronger, with regard to energy loss as well as the penetration distance of α particles, than that of a lower Z plasma

  3. Post-irradiation examinations of uranium-plutonium mixed nitride fuel irradiated in JMTR (89F-3A capsule)

    International Nuclear Information System (INIS)

    Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Arai, Yasuo; Kimura, Yasuhiko; Nagashima, Hisao; Sekita, Noriaki

    2000-03-01

    Two helium-bonded fuel pins filled with uranium-plutonium mixed nitride pellets were encapsulated in 89F-3A and irradiated in JMTR up to 5.5% FIMA at a maximum linear power of 73 kW/m. The capsule cooled for ∼5 months was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in the irradiated fuel pins. Very low fission gas release rate of about 2 ∼ 3% was observed, while the diametric increase of fuel pin was limited to ∼0.4% at the position of maximum reading. The inner surface of cladding tube did not show any signs of chemical interaction with fuel pellet. (author)

  4. Soil treatment to remove uranium and related mixed radioactive contaminants. Final report September 1992--October 1995

    International Nuclear Information System (INIS)

    1996-07-01

    A research and development project to remove uranium and related radioactive contaminants from soil by an ultrasonically-aided chemical leaching process began in 1993. The project objective was to develop and design, on the basis of bench-scale and pilot-scale experimental studies, a cost-effective soil decontamination process to produce a treated soil containing less than 35 pCi/g. The project, to cover a period of about thirty months, was designed to include bench-scale and pilot-scale studies to remove primarily uranium from the Incinerator Area soil, at Fernald, Ohio, as well as strontium-90, cobalt-60 and cesium-137 from a Chalk River soil, at the Chalk River Laboratories, Ontario. The project goal was to develop, design and cost estimate, on the basis of bench-scale and pilot-scale ex-situ soil treatment studies, a process to remove radionuclides form the soils to a residual level of 35 pCi/g of soil or less, and to provide a dischargeable water effluent as a result of soil leaching and a concentrate that can be recovered for reuse or solidified as a waste for disposal. In addition, a supplementary goal was to test the effectiveness of in-situ soil treatment through a field study using the Chalk River soil

  5. Stabilization of mixed carbides of uranium-plutonium by zirconium. Part 1.: uranium carbide with small additions of zirconium; Etude de la stabilisation des carbures mixtes d'uranium et de plutonium par addition de zirconium. 1. partie: etude des carbures d'uranium avec de faibles additions de zirconium

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-07-01

    Cast carbide samples, being of a high density and purity, are preferable for research purposes, to samples produced by powder metallurgy methods. Samples of uranium carbide with small additions of zirconium (1 to 5 per cent) were cast, as rods, in an arc furnace. A single phase carbide with interesting qualities was produced. As cast, a dendrite structure is observed, which does not disappear, after a treatment at 1900 deg. C during 110 hours. In comparison with uranium monocarbide the compatibility with stainless steel is much improved. The specific heat (between room temperature and 2500 deg. C) is similar to the specific heat of uranium monocarbide. A study of these mixed carbides, but having a part of the uranium replaced by plutonium is under way. (author) [French] Les echantillons de monocarbures obtenus par coulee sont tres interessants pour les recherches experimentales a cause de leur grande purete, de leur densite tres elevee et de la facilite d'obtention des lingots de forme et dimensions variees. On a prepare et coule dans un four a arc des echantillons de carbures d'uranium avec de faibles additions de zirconium (1 a 5 at. pour cent). On obtient ainsi des carbures monophases presentant de meilleures proprietes que le monocarbure d'uranium. A l'etat brut de coulee on observe une structure dendritique qui n'est pas detruite par un traitement thermique de 110 heures a 1900 deg. C. La compatibilite avec l'acier inoxydable 316 (a 925 deg. C pendant 500 heures) est nettement amelioree par rapport a UC. La chaleur specifique (entre la temperature ordinaire et 2500 deg. C) et la densite sont tres peu differentes de celles du monocarbure d'uranium. Une etude concernant les composes U-Pu-Zr-C est actuellement en cours. (auteur)

  6. Removals of aqueous sulfur dioxide and hydrogen sulfide using CeO2-NiAl-LDHs coating activated carbon and its mix with carbon nano-tubes

    KAUST Repository

    Li, Jing; Chen, Fangping; Jin, Guanping; Feng, Xiaoshuang; Li, Xiaoxuan

    2015-01-01

    Ce-doped NiAl/layered double hydroxide was coated at activated carbon by urea hydrolysis method (CeO2-NiAl-LDHs/AC) in one pot, which was characterized by X-ray diffraction, infrared spectra, field emission scanning electron microscope and electrochemical techniques. CeO2-NiAl-LDHs/AC shows good uptake for aqueous sulfur dioxide (483.09mg/g) and hydrogen sulfide (181.15mg/g), respectively at 25°C. Meanwhile, the electrochemical removals of aqueous sulfur dioxide and hydrogen sulfide were respectively investigated at the mix of CeO2-NiAl-LDHs/AC and carbon nano-tubes modified homed paraffin-impregnated electrode. Both sulfur dioxide and hydrogen sulfide could be effectively oxidized to sulfuric acid at 1.0V in alkaline aqueous solution. © 2015 Elsevier B.V.

  7. Removals of aqueous sulfur dioxide and hydrogen sulfide using CeO2-NiAl-LDHs coating activated carbon and its mix with carbon nano-tubes

    KAUST Repository

    Li, Jing

    2015-07-01

    Ce-doped NiAl/layered double hydroxide was coated at activated carbon by urea hydrolysis method (CeO2-NiAl-LDHs/AC) in one pot, which was characterized by X-ray diffraction, infrared spectra, field emission scanning electron microscope and electrochemical techniques. CeO2-NiAl-LDHs/AC shows good uptake for aqueous sulfur dioxide (483.09mg/g) and hydrogen sulfide (181.15mg/g), respectively at 25°C. Meanwhile, the electrochemical removals of aqueous sulfur dioxide and hydrogen sulfide were respectively investigated at the mix of CeO2-NiAl-LDHs/AC and carbon nano-tubes modified homed paraffin-impregnated electrode. Both sulfur dioxide and hydrogen sulfide could be effectively oxidized to sulfuric acid at 1.0V in alkaline aqueous solution. © 2015 Elsevier B.V.

  8. Effect of ascorbate ions in DTPA treatment after contamination by mixed plutonium dioxide-sodium burning products

    International Nuclear Information System (INIS)

    Metivier, H.; Masse, R.; Nolibe, D.; Nenot, J.C.; Lafuma, J.

    1977-01-01

    Plutonium toxicity problems arising from the use of molten sodium cooled fast breeder reactors involving the accidental mixing and ignition of sodium contaminated by plutonium dioxide have been investigated. The possibility of using the strong reducing agent, ascorbate ion, in order to reduce Pu VI to a chemical form more complexable by DTPA, has been tested using male and female rats. Urine, feces, liver and skeletal measurements of 239 Pu daughter products were carried out. It was shown that the transportable fraction was not significantly modified by the associate ascorbate-DTPA treatment. However the skeletal burden which represents 2.2% with DTPA, reached only 1.6% with ascorbate association and 0.3% with preventive ascorbate injections. Whatever the DTPA ascorbate treatment, the liver burden was not modified significantly. Difference observable only with preventive ascorbate treatment cancels out ascorbate association in order to greatly increase DTPA therapy efficiency, but seems to indicate that the Pu-proteins associations in blood is reduced more easily by ascorbate than association of Pu with local proteins. Ascorbate reduction can also be observed in urinary elimination in that preventive ascorbate, followed by DTPA-ascorbate, leads to a quicker urinary elimination than DTPA or DTPA plus ascorbate post treatment. (U.K.)

  9. Uranium conversion; Urankonvertering

    Energy Technology Data Exchange (ETDEWEB)

    Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina [Swedish Defence Research Agency (FOI), Stockholm (Sweden)

    2006-03-15

    FOI, has performed a study on uranium conversion processes that are of importance in the production of different uranium compounds in the nuclear industry. The same conversion processes are of interest both when production of nuclear fuel and production of fissile material for nuclear weapons are considered. Countries that have nuclear weapons ambitions, with the intention to produce highly enriched uranium for weapons purposes, need some degree of uranium conversion capability depending on the uranium feed material available. This report describes the processes that are needed from uranium mining and milling to the different conversion processes for converting uranium ore concentrate to uranium hexafluoride. Uranium hexafluoride is the uranium compound used in most enrichment facilities. The processes needed to produce uranium dioxide for use in nuclear fuel and the processes needed to convert different uranium compounds to uranium metal - the form of uranium that is used in a nuclear weapon - are also presented. The production of uranium ore concentrate from uranium ore is included since uranium ore concentrate is the feed material required for a uranium conversion facility. Both the chemistry and principles or the different uranium conversion processes and the equipment needed in the processes are described. Since most of the equipment that is used in a uranium conversion facility is similar to that used in conventional chemical industry, it is difficult to determine if certain equipment is considered for uranium conversion or not. However, the chemical conversion processes where UF{sub 6} and UF{sub 4} are present require equipment that is made of corrosion resistant material.

  10. Advanced Proliferation Resistant, Lower Cost, Uranium-Thorium Dioxide Fuels for Light Water Reactors (Progress report for work through June 2002, 12th quarterly report)

    International Nuclear Information System (INIS)

    Mac Donald, Philip Elsworth

    2002-01-01

    The overall objective of this NERI project is to evaluate the potential advantages and disadvantages of an optimized thorium-uranium dioxide (ThO2/UO2) fuel design for light water reactors (LWRs). The project is led by the Idaho National Engineering and Environmental Laboratory (INEEL), with the collaboration of three universities, the University of Florida, Massachusetts Institute of Technology (MIT), and Purdue University; Argonne National Laboratory; and all of the Pressurized Water Reactor (PWR) fuel vendors in the United States (Framatome, Siemens, and Westinghouse). In addition, a number of researchers at the Korean Atomic Energy Research Institute and Professor Kwangheon Park at Kyunghee University are active collaborators with Korean Ministry of Science and Technology funding. The project has been organized into five tasks: Task 1 consists of fuel cycle neutronics and economics analysis to determine the economic viability of various ThO2/UO2 fuel designs in PWRs; Task 2 will determine whether or not ThO2/UO2 fuel can be manufactured economically; Task 3 will evaluate the behavior of ThO2/UO2 fuel during normal, off-normal, and accident conditions and compare the results with the results of previous UO2 fuel evaluations and U.S. Nuclear Regulatory Commission (NRC) licensing standards; Task 4 will determine the long-term stability of ThO2/UO2 high-level waste; and Task 5 consists of the Korean work on core design, fuel performance analysis, and xenon diffusivity measurements

  11. URANIUM LEACHING AND RECOVERY PROCESS

    Science.gov (United States)

    McClaine, L.A.

    1959-08-18

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

  12. Synthesis, structure, spectroscopy and redox energetics of a series of uranium(4) mixed-ligand metallocene complexes

    Energy Technology Data Exchange (ETDEWEB)

    Thomson, R.K.; Scott, B.L.; Morris, D.E.; Kiplinger, J.L. [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2010-06-15

    A series of uranium(IV) mixed-ligand amide-halide/pseudo-halide complexes (C{sub 5}Me{sub 5}){sub 2}U[N(SiMe{sub 3}){sub 2}](X) (X = F (1), Cl (2), Br (3), I (4), N{sub 3} (5), NCO (6)), (C{sub 5}Me{sub 5}){sub 2}U(NPh{sub 2})(X) (X = Cl (7), N{sub 3} (8)), and (C{sub 5}Me{sub 5}){sub 2}U[N(Ph)(SiMe{sub 3})](X) (X Cl (9), N{sub 3} (10)) have been prepared by one electron oxidation of the corresponding uranium(III) amide precursors using either copper halides, silver iso-cyanate, or triphenylphosphine gold(I)azide. Agostic U---H-C interactions and {eta}{sub 3}-(N,C,C') coordination are observed for these complexes in both the solid-state and solution. There is a linear correlation between the chemical shift values of the C{sub 5}Me{sub 5} ligand protons in the {sup 1}H NMR spectra and the U(IV)/U(III) reduction potentials of the (C{sub 5}Me{sub 5}){sub 2}U[N(SiMe{sub 3}){sub 2}](X) complexes, suggesting that there is a common origin, that is overall {sigma}-/{pi}-donation from the ancillary (X) ligand to the metal, contributing to both observables. Optical spectroscopy of the series of complexes 1-6 is dominated by the (C{sub 5}Me{sub 5}){sub 2}U[N(SiMe{sub 3}){sub 2}] core, with small variations derived from the identity of the halide/pseudo-halide. The considerable {pi}-donating ability of the fluoride ligand is reflected in both the electrochemistry and UV-visible-NIR spectroscopic behavior of the fluoride complex (C{sub 5}Me{sub 5}){sub 2}U[N(SiMe{sub 3}){sub 2}](F) (1). The syntheses of the new trivalent uranium amide complex, (C{sub 5}Me{sub 5}){sub 2}U[N(Ph)(SiMe{sub 3})](THF), and the two new weakly-coordinating electrolytes, [Pr{sub 4}N][B{l_brace}3,5-(CF{sub 3}){sub 2}C{sub 6}H{sub 3{r_brace}4}] and [Pr{sub 4}N][B(C{sub 6}F{sub 5}){sub 4}], are also reported. (authors)

  13. Preparation of uranium tetrafluoride

    International Nuclear Information System (INIS)

    Wirths, G.

    1981-01-01

    Uranium dioxide is converted to uranium tetrafluoride under stoichiometric excess of hydrogen fluoride. The water formed in the process and the unreacted hydrogen fluoride are cooled and the condensate fractionally distilled into water and approx. 40% hydrofluoric acid. The hydrofluoric acid and water-free hydrogen fluoride are fed back into the process. (WI) [de

  14. Uranium extraction from underground deposits

    International Nuclear Information System (INIS)

    Wolfe, C.R.

    1982-01-01

    Uranium is extracted from underground deposits by passing an aqueous oxidizing solution of carbon dioxide over the ore in the presence of calcium ions. Complex uranium carbonate or bicarbonate ions are formed which enter the solution. The solution is forced to the surface and the uranium removed from it

  15. Preparation and study of the nitrides and mixed carbide-nitrides of uranium and of plutonium

    International Nuclear Information System (INIS)

    Anselin, F.

    1966-06-01

    A detailed description is given of a simple method for preparing uranium and plutonium nitrides by the direct action of nitrogen under pressure at moderate temperatures (about 400 C) on the partially hydrogenated bulk metal. It is shown that there is complete miscibility between the UN and PuN phases. The variations in the reticular parameters of the samples as a function of temperature and in the presence of oxide have been used to detect and evaluate the solubility of oxygen in the different phases. A study has been made of the sintering of these nitrides as a function of the preparation conditions with or without sintering additives. A favorable but non-reproducible, effect has been found for traces of oxide. The best results were obtained for pure UN at 1600 C (96 per cent theoretical density) on condition that a well defined powder, was used. The criterion used is the integral width of the X-ray diffraction lines. The compounds UN and PuN are completely miscible with the corresponding carbides. This makes it possible to prepare carbide-nitrides of the general formula (U,Pu) (C,N) by solid-phase diffusion, at around 1400 C. The sintering of these carbide-nitrides is similar to that of the carbides if the nitrogen content is low; in particular, nickel is an efficient sintering agent. For high contents, the sintering is similar to that of pure nitrides. (author) [fr

  16. Calibration of the Lawrence Livermore National Laboratory Passive-Active Neutron Drum Shuffler for Measurement of Highly Enriched Uranium in Mixed Oxide

    International Nuclear Information System (INIS)

    Mount, M.; O'Connell, W.; Cochran, C.; Rinard, P.; Dearborn, D.; Endres, E.

    2002-01-01

    As a follow-on to the Lawrence Livermore National Laboratory (LLNL) effort to calibrate the LLNL passive-active neutron drum (PAN) shuffler for measurement of highly enriched uranium (HEU) oxide, a method has been developed to extend the use of the PAN shuffler to the measurement of HEU in mixed uranium-plutonium (U-Pu) oxide. This method uses the current LLNL HEU oxide calibration algorithms, appropriately corrected for the mixed U-Pu oxide assay time, and recently developed PuO 2 calibration algorithms to yield the mass of 235 U present via differences between the expected count rate for the PuO 2 and the measured count rate of the mixed U-Pu oxide. This paper describes the LLNL effort to use PAN shuffler measurements of units of certified reference material (CRM) 149 (uranium (93% Enriched) Oxide - U 3 O 8 Standard for Neutron Counting Measurements) and CRM 146 (uranium Isotopic Standard for Gamma Spectrometry Measurements) and a selected set of LLNL PuO 2 -bearing containers in consort with Monte Carlo simulations of the PAN shuffler response to each to (1) establish and validate a correction to the HEU calibration algorithm for the mixed U-Pu oxide assay time, (2) develop a PuO 2 calibration algorithm that includes the effect of PuO 2 density (2.4 g/cm 3 to 4.8 g/cm 3 ) and container size (8.57 cm to 9.88 cm inside diameter and 9.60 cm to 13.29 cm inside height) on the PAN shuffler response, and (3) develop and validate the method for establishing the mass of 235 U present in an unknown of mixed U-Pu oxide.

  17. PRODUCTION OF URANIUM METAL BY CARBON REDUCTION

    Science.gov (United States)

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

    1959-09-22

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

  18. A metallogenetic model of supergene extraction, releasing and enrichment in the mixed zone for granite-type uranium deposits in south China

    International Nuclear Information System (INIS)

    Li Minglian.

    1986-01-01

    The major geological features and their related geological events provide a base for the modelling of granite-type uranium deposits in South China. This paper presents a metallogenetic model to suggest the process of ore fluid circulation. There are two streams of ore fluids moving in the fracture zone: one comes from meteoric water and extracts uranium from wall rocks, flowing from top to bottom which is named uranium-loading fluid; another derives from the depth of the crust flowing from bottom to top and contains reducing matters as H 2 S etc. called uranium-releasing fluid. These two streams of solutions of different genesis, composition and character encountered and mixed at certain depth to precipitate the uranium. During the process the longitudinal circulation of underground thermal water in fracture zone results in the Bernoulli latitudinal circulation of ore fluids, which caused the ore fluids to ceaselessly flow into the minerogenetic location, where mineralization can be formed continuously in a certain period

  19. Impact of receipt of coprocessed uranium/plutonium on advanced accountability concepts and fabrication facilities. Addendum 1 to application of advanced accountability concepts in mixed oxide fabrication

    International Nuclear Information System (INIS)

    Bastin, J.J.; Jump, M.J.; Lange, R.A.; Randall, C.C.

    1977-11-01

    The Phase I study of the application of advanced accountability methods (DYMAC) in a uranium/plutonium mixed oxide facility was extended to assess the effect of coprocessed UO 2 --PuO 2 feed on the observations made in the original Phase I effort and on the proposed Phase II program. The retention of plutonium mixed with uranium throughout the process was also considered. This addendum reports that coprocessed feed would have minimal effect on the DYMAC program, except in the areas of material specifications, starting material delivery schedule, and labor requirements. Each of these areas is addressed, as are the impact of coprocessed feed at a large fuel fabrication facility and the changes needed in the dirty scrap recovery process to maintain the lower plutonium levels which may be required by future nonproliferation philosophy. An amended schedule for Phase II is included

  20. Post irradiation examinations of uranium-plutonium mixed carbide fuels irradiated at low linear power rate

    International Nuclear Information System (INIS)

    Maeda, Atsushi; Sasayama, Tatsuo; Iwai, Takashi; Aizawa, Sakuei; Ohwada, Isao; Aizawa, Masao; Ohmichi, Toshihiko; Handa, Muneo

    1988-11-01

    Two pins containing uranium-plutonium carbide fuels which are different in stoichiometry, i.e. (U,Pu)C 1.0 and (U,Pu)C 1.1 , were constructed into a capsule, ICF-37H, and were irradiated in JRR-2 up to 1.0 at % burnup at the linear heat rate of 420 W/cm. After being cooled for about one year, the irradiated capsule was transferred to the Reactor Fuel Examination Facility where the non-destructive examinations of the fuel pins in the β-γ cells and the destructive ones in two α-γ inert gas atmosphere cells were carried out. The release rates of fission gas were low enough, 0.44 % from (U,Pu)C 1.0 fuel pin and 0.09% from (U,Pu)C 1.1 fuel pin, which is reasonable because of the low central temperature of fuel pellets, about 1000 deg C and is estimated that the release is mainly governed by recoil and knock-out mechanisms. Volume swelling of the fuels was observed to be in the range of 1.3 ∼ 1.6 % for carbide fuels below 1000 deg C. Respective open porosities of (U,Pu)C 1.0 and (U,Pu)C 1.1 fuel were 1.3 % and 0.45 %, being in accordance with the release behavior of fission gas. Metallographic observation of the radial sections of pellets showed the increase of pore size and crystal grain size in the center and middle region of (U,Pu)C 1.0 pellets. The chemical interaction between fuel pellets and claddings in the carbide fuels is the penetration of carbon in the fuels to stainless steel tubes. The depth of corrosion layer in inner sides of cladding tubes ranged 10 ∼ 15 μm in the (U,Pu)C 1.0 fuel and 15 #approx #25 μm in the (U,Pu)C 1.1 fuel, which is correlative with the carbon potential of fuels posibly affecting the amount of carbon penetration. (author)

  1. In situ studies of uranium-plutonium mixed oxides. Influence of composition on phase equilibria and thermodynamic properties

    International Nuclear Information System (INIS)

    Strach, Michal

    2015-01-01

    Due to their physical and chemical properties, mixed uranium-plutonium oxides are considered for fuel in 4. generation nuclear reactors. In this frame, complementary experimental studies are necessary to develop a better understanding of the phenomena that take place during fabrication and operation in the reactor. The focus of this work was to study the U-Pu-O phase diagram in a wide range of compositions and temperatures to ameliorate our knowledge of the phase equilibria in this system. Most of experiments were done using in situ X-ray diffraction at elevated temperatures. The control of the oxygen partial pressure during the treatments made it possible to change the oxygen stoichiometry of the sample, which gave us an opportunity to study rapidly different compositions and the processes involved. The experimental approach was coupled with thermodynamic modeling using the CALPHAD method, to precisely plan the experiments and interpret the obtained results. This approach enabled us to enhance the knowledge of phase equilibria in the U-Pu-O system. (author) [fr

  2. Advanced Proliferation Resistant, Lower Cost, Uranium-Thorium Dioxide Fuels for Light Water Reactors (Progress report for work through June 2002, 12th quarterly report)

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth

    2002-09-01

    The overall objective of this NERI project is to evaluate the potential advantages and disadvantages of an optimized thorium-uranium dioxide (ThO2/UO2) fuel design for light water reactors (LWRs). The project is led by the Idaho National Engineering and Environmental Laboratory (INEEL), with the collaboration of three universities, the University of Florida, Massachusetts Institute of Technology (MIT), and Purdue University; Argonne National Laboratory; and all of the Pressurized Water Reactor (PWR) fuel vendors in the United States (Framatome, Siemens, and Westinghouse). In addition, a number of researchers at the Korean Atomic Energy Research Institute and Professor Kwangheon Park at Kyunghee University are active collaborators with Korean Ministry of Science and Technology funding. The project has been organized into five tasks: · Task 1 consists of fuel cycle neutronics and economics analysis to determine the economic viability of various ThO2/UO2 fuel designs in PWRs, · Task 2 will determine whether or not ThO2/UO2 fuel can be manufactured economically, · Task 3 will evaluate the behavior of ThO2/UO2 fuel during normal, off-normal, and accident conditions and compare the results with the results of previous UO2 fuel evaluations and U.S. Nuclear Regulatory Commission (NRC) licensing standards, · Task 4 will determine the long-term stability of ThO2/UO2 high-level waste, and · Task 5 consists of the Korean work on core design, fuel performance analysis, and xenon diffusivity measurements.

  3. Establishment of THERPRO Database and Estimation of the Effect of Fuel Burn-up on the Thermal Conductivity of Uranium Dioxide

    International Nuclear Information System (INIS)

    Lee, Hyun Seon

    2005-02-01

    Materials property data are an essential part of major disciplines in many engineering fields. To nuclear engineering, fundamental understanding of thermo-physical chemical mechanical properties of nuclear materials is very important. THERPRO data base that is re-designed and re-constructed through this study is a web-based on-line nuclear materials properties data base. For the future upgrade of the data base contemporary information technologies have been incorporated during the construction. Basically THERPRO data base has a hierarchical structure consisting of several levels: home page, element, compound, property, author, report, and bibliography level. All of data sets in each level are interconnected using network structure and thus every data can be easily retrieved including the bibliographical information by an appropriate query action. As a part of THERPRO DB utilization, the effect of fuel burn-up on the thermal conductivity of irradiated uranium dioxide is analyzed with the data contained in the data base as well as recent data published in the relevant journals. Their data are comparatively studied and the effect is estimated using FRAPCON-3 code with two in-pile data sets, BR-3 111i5 and Oconee rod 15309. The results show that the fuel center line temperature can differ 200 .deg. C∼400 .deg. C from thermal conductivity models depending on burn-up, which can significantly influence high burn-up fuel performance. In conclusion, it is demonstrated through this study that THERPRO data base can be a great utility for nuclear engineers and researchers, if appropriately utilized

  4. Experimental study and model development for 'uranium dioxide-epoxy resin' heat treatment

    International Nuclear Information System (INIS)

    Chairat, Aziza

    2015-01-01

    kinetic model to the partial differential equations (mass, energy and momentum balance) to obtain a representative model of the oven in terms of temperature and chemical species composition. The Modeling of the oven is carried out using COMSOL Multiphysics software. The results showed a good agreement with experimental measurements. After pyrolysis, char still contains significant amount of hydrogen. To minimize this quantity, the oxidation of the char is a necessary step. Two treatment types are proposed: An oxidation under a controlled oxygen atmosphere and carbon dioxide gasification. These methods are efficient to eliminate the residual of hydrogen content while keeping the fuel integrity. (author) [fr

  5. Final disposal options for mercury/uranium mixed wastes from the Oak Ridge Reservation

    International Nuclear Information System (INIS)

    Gorin, A.H.; Leckey, J.H.; Nulf, L.E.

    1994-01-01

    Laboratory testing was completed on chemical stabilization and physical encapsulation methods that are applicable (to comply with federal and state regulations) to the final disposal of both hazardous and mixed hazardous elemental mercury waste that is in either of the following categories: (1) waste generated during decontamination and decommissioning (D and D) activities on mercury-contaminated buildings, such as Building 9201-4 at the Oak Ridge Y-12 Plant, or (2) waste stored and regulated under either the Federal Facilities Compliance Agreement or the Federal Facilities Compliance Act. Methods were used that produced copper-mercury, zinc-mercury, and sulfur-mercury materials at room temperature by dry mixing techniques. Toxicity Characteristic Leaching Procedure (TCLP) results for mercury on batches of both the copper-mercury and the sulfur-mercury amalgams consistently produced leachates with less than the 0.2-mg/L Resource Conservation and Recovery Act (RCRA) regulatory limit for mercury. The results clearly showed that the reaction of mercury with sulfur at room temperature produces black mercuric sulfide, a material that is well suited for land disposal. The results also showed that the copper-mercury and zinc-mercury amalgams had major adverse properties that make them undesirable for land disposal. In particular, they reacted readily in air to form oxides and liberate elemental mercury. Another major finding of this study is that sulfur polymer cement is potentially useful as a physical encapsulating agent for mercuric sulfide. This material provides a barrier in addition to the chemical stabilization that further prevents mercury, in the form of mercuric sulfide, from migrating into the environment

  6. Dry recovery test of plutonium-uranium mixed oxide fuel pellets

    International Nuclear Information System (INIS)

    Kinugasa, Manabu; Kawamata, Kazuhiko; Kashima, Sadamitsu

    1981-01-01

    The oxidation conditions for pulverizing directly Pu-U mixed oxide pellets without mechanical crushing were examined to simplify the process and to reduce radiation exposure during the dry recovery of highly enriched Pu pellets. The specimens used were the Pusub(0.3) Usub(0.7) Osub(2-x) pellets with different density, which were sintered at 1650 deg C for 2 hours under an atmosphere of 5 % H 2 - N 2 . The oxidation experiment was carried out under several conditions. The oxidation products were examined by weight gain, X-ray diffraction, appearance pictures, SEM photographs and so on. From these studies, it can be concluded that the oxidation in NO 2 diluted with air was very powerful, but if only the coarse spalling of Pusub(0.3) Usub(0.7) O 2 sintered pellets is required, it is sufficient to oxidize them in air for 1 hr in a temperature range from 400 to 600 deg C. (Asami, T.)

  7. Uranium mining in Australia

    International Nuclear Information System (INIS)

    Mackay, G.A.

    1978-01-01

    Western world requirements for uranium based on increasing energy consumption and a changing energy mix, will warrant the development of Australia's resources. By 1985 Australian mines could be producing 9500 tonnes of uranium oxide yearly and by 1995 the export value from uranium could reach that from wool. In terms of benefit to the community the economic rewards are considerable but, in terms of providing energy to the world, Australias uranium is vital

  8. Inflammability of magnesium and its alloys in carbon dioxide either pure or mixed with water vapour, air or oxygen

    International Nuclear Information System (INIS)

    Baque, P.; Chevilliard, C.; Darras, R.

    1964-01-01

    Whereas low contents of metallic additions have only a small influence on the temperature at which magnesium begins to burn in carbon dioxide, an increase in the specific surface area of the samples is likely to reduce it considerably, the self-heating phenomena being then more pronounced. On the other hand, the exact nature of the surrounding atmosphere is a very important factor; thus the temperature at which ignition begins increases with increasing carbon dioxide pressure, decreases very rapidly when the moisture content of this gas increases up to 3000 v.p.m., and decreases regularly when the partial pressure of air or oxygen increases. (authors) [fr

  9. Micromechanical simulation of Uranium dioxide polycrystalline aggregate behaviour under irradiation; Modele numerique micro-mecanique d'agregat polycristallin pour le comportement des combustibles oxydes

    Energy Technology Data Exchange (ETDEWEB)

    Pacull, J.

    2011-02-15

    In pressurized water nuclear power reactor (PWR), the fuel rod is made of dioxide of uranium (UO{sub 2}) pellet stacked in a metallic cladding. A multi scale and multi-physic approaches are needed for the simulation of fuel behavior under irradiation. The main phenomena to take into account are thermomechanical behavior of the fuel rod and chemical-physic behavior of the fission products. These last years one of the scientific issue to improve the simulation is to take into account the multi-physic coupling problem at the microscopic scale. The objective of this ph-D study is to contribute to this multi-scale approach. The present work concerns the micro-mechanical behavior of a polycrystalline aggregate of UO{sub 2}. Mean field and full field approaches are considered. For the former and the later a self consistent homogenization technique and a periodic Finite Element model base on the 3D Voronoi pattern are respectively used. Fuel visco-plasticity is introduced in the model at the scale of a single grain by taking into account specific dislocation slip systems of UO{sub 2}. A cohesive zone model has also been developed and implemented to simulate grain boundary sliding and intergranular crack opening. The effective homogenous behaviour of a Representative Volume Element (RVE) is fitted with experimental data coming from mechanical tests on a single pellet. Local behavior is also analyzed in order to evaluate the model capacity to assess micro-mechanical state. In particular, intra and inter granular stress gradient are discussed. A first validation of the local behavior assessment is proposed through the simulation of intergranular crack opening measured in a compressive creep test of a single fuel pellet. Concerning the impact of the microstructure on the fuel behavior under irradiation, a RVE simulation with a representative transient loading of a fuel rod during a power ramp test is achieved. The impact of local stress and strain heterogeneities on the multi

  10. Mixed complexes of uranium(IV) and thorium(IV) with N,N'-ethylenebis(salicylideneimine) and N,N'-propylenebis(salicylideneimine)

    Energy Technology Data Exchange (ETDEWEB)

    Doretti, L; Madalosso, F; Sitran, S; Faleschini, S [Consiglio Nazionale delle Ricerche, Padua (Italy). Lab. di Chimica e Tecnologia dei Radioelementi

    1976-01-01

    Recently some uranium tetrachloride adducts with N-(hydroxyphenyl)-salicylaldimine and N-(methoxyphenyl)salicylaldimine have been prepared (Doretti et al., J. Inorg. Nucl. Chem., (in press)): the study has now extended to the complexes of Uranium(IV) and Thorium(IV) with N,N'-Ethylenebis(salicylideneimine), H/sub 2/salen, and N,N'-Propylenebis(salicylideneimine), H/sub 2/salpropen, with the aim to prepare fully substituted complexes, mixed complexes containing both chloride and Schiff base groups and adducts of tetrachlorides with free Schiff bases. This paper reports the preparation and properties of the compounds MCl/sub 4/(H/sub 2/salen), MCl/sub 4/(H/sub 2/salpropen), M(salen)Cl/sub 2/, M(salen)/sub 2/ and M(salpropen)/sub 2/, where M = U or Th. These compounds have been characterized by elemental analyses and their IR spectra are reported and discussed.

  11. A mixed-valent uranium phosphonate framework containing U{sup IV}, U{sup V}, and U{sup VI}

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Lanhua; Zheng, Tao; Wang, Yaxing; Diwu, Juan; Chai, Zhifang; Wang, Shuao [School for Radiological and interdisciplinary Sciences (RAD-X), Soochow University, Suzhou (China); Collaborative Innovation Center of Radiation Medicine, Jiangsu Higher Education Institutions, Suzhou (China); Bao, Songsong; Zheng, Limin [State Key Laboratory of Coordination Chemistry, School of Chemistry and Chemical Engineering, Collaborative Innovation Center of Advanced Microstructures, Nanjing University (China); Zhang, Linjuan; Wang, Jianqiang [Shanghai Institute of Applied Physics and, Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai (China); Liu, Hsin-Kuan [Department of Chemistry, National Central University, Jhongli (China); Albrecht-Schmitt, Thomas E. [Department of Chemistry and Biochemistry, Florida State University, Tallahassee, FL (United States)

    2016-08-16

    It is shown that U{sup V}O{sub 2}{sup +} ions can reside at U{sup VI}O{sub 2}{sup 2+} lattice sites during mild reduction and crystallization process under solvothermal conditions, yielding a complicated and rare mixed-valent uranium phosphonate compound that simultaneously contains U{sup IV}, U{sup V}, and U{sup VI}. The presence of uranium with three oxidation states was confirmed by various characterization techniques, including X-ray crystallography, X-ray photoelectron, electron paramagnetic resonance, FTIR, UV/Vis-NIR absorption, and synchrotron radiation X-ray absorption spectroscopy, and magnetism measurements. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  12. About the elaboration of pure uranium dicarbide

    International Nuclear Information System (INIS)

    Besson, J.; Blum, P.; Guinet, Ph.; Spitz, J.

    1963-01-01

    In order to develop methods for the elaboration of as pure as possible uranium dicarbide, the authors report the study of different elaboration processes based on the reaction between uranium and carbon, or between uranium and hydrocarbon, or between uranium oxide and carbon. They finally choose a method which comprises an arc-induced fusion of a mixture of uranium dioxide and carbon. The fusion process is described. The influence of thermal treatments is discussed as well as the graphite electrode carburization

  13. Research of natural resources saving by design studies of Pressurized Light Water Reactors and High Conversion PWR cores with mixed oxide fuels composed of thorium/uranium/plutonium

    International Nuclear Information System (INIS)

    Vallet, V.

    2012-01-01

    Within the framework of innovative neutronic conception of Pressurized Light Water Reactors (PWR) of 3. generation, saving of natural resources is of paramount importance for sustainable nuclear energy production. This study consists in the one hand to design high Conversion Reactors exploiting mixed oxide fuels composed of thorium/uranium/plutonium, and in the other hand, to elaborate multi-recycling strategies of both plutonium and 233 U, in order to maximize natural resources economy. This study has two main objectives: first the design of High Conversion PWR (HCPWR) with mixed oxide fuels composed of thorium/uranium/plutonium, and secondly the setting up of multi-recycling strategies of both plutonium and 233 U, to better natural resources economy. The approach took place in four stages. Two ways of introducing thorium into PWR have been identified: the first is with low moderator to fuel volume ratios (MR) and ThPuO 2 fuel, and the second is with standard or high MR and ThUO 2 fuel. The first way led to the design of under-moderated HCPWR following the criteria of high 233 U production and low plutonium consumption. This second step came up with two specific concepts, from which multi-recycling strategies have been elaborated. The exclusive production and recycling of 233 U inside HCPWR limits the annual economy of natural uranium to approximately 30%. It was brought to light that the strong need in plutonium in the HCPWR dedicated to 233 U production is the limiting factor. That is why it was eventually proposed to study how the production of 233 U within PWR (with standard MR), from 2020. It was shown that the anticipated production of 233 U in dedicated PWR relaxes the constraint on plutonium inventories and favours the transition toward a symbiotic reactor fleet composed of both PWR and HCPWR loaded with thorium fuel. This strategy is more adapted and leads to an annual economy of natural uranium of about 65%. (author) [fr

  14. Extraction of uranium and lead from mixed waste debris using a variety of metal/ligand complexes

    International Nuclear Information System (INIS)

    Needham, D.A.; Duran, B.L.; Ehler, D.S.; Sauer, N.N.

    1997-01-01

    To ensure the safety of our Nation's nuclear stockpile, Los Alamos National Laboratory is in the process of constructing the DARHT (Dual Axis Radiographic Hydrodynamic Test) facility. DARHT will examine the effects of aging and the stability of our stockpile. Contained testing will be phased in to reduce the impact of these tests, which contain depleted uranium, on our environment. The main focus of this research is to develop a treatment scheme for the recovery of depleted uranium and lead from shot debris that will result from these tests. The goals of this research are to optimize the conditions on a bench scale using a commercially available water soluble polymer to bind the lead and a variety of metal/ligand complexes such as 4,5-dihydroxy-1,3-benzene-disulfonic acid, dithionite, sodium carbonate/bicarbonate, and sodium hypochlorite to bind the uranium. Studies were conducted on a mixture of debris, such as wood, cable, paper towels, and tubing that contained both uranium and lead ranging in concentration from 10-1000's of ppm of contaminants. Experiments were done varying combinations and successions of extractant solutions as well as a number of sequential extractions. Studies show that a mixture of sodium hypochlorite and carbonate removed 90+% of both uranium and lead. We then focused on a separation scheme for the lead and uranium

  15. In situ leach method for recovering uranium and related values

    International Nuclear Information System (INIS)

    Yan, T.Y.

    1981-01-01

    A process is provided for in-situ leaching of uranium from a calcium-containing clay which does not result in contamination of the clay formation by any cations not already present. A lixiviant is prepared by dissolving carbon dioxide into water having essentially the same cationic composition as that of the formation connate water. The solution is injected along with an oxidant, for example oxygen, into the formation. Calcium that has become dissolved in the lixiviant must be removed to control the pH, preferably by the addition of lime in a calcium precipitator. After calcium removal the lixiviant is filtered to remove suspended solids and is passed through an ion exchange resin or other uranium extraction means. The barren solution goes to a mix tank where carbon dioxide is added, and the fresh lixiviant is injected along with additional oxidant into the formation

  16. Catalytic removal of sulfur dioxide from dibenzothiophene sulfone over Mg-Al mixed oxides supported on mesoporous silica.

    Science.gov (United States)

    You, Nansuk; Kim, Min Ji; Jeong, Kwang-Eun; Jeong, Soon-Yong; Park, Young-Kwon; Jeon, Jong-Ki

    2010-05-01

    Dibenzothiophene sulfone (DBTS), one of the products of the oxidative desulfurization of heavy oil, can be removed through extraction as well as by an adsorption process. It is necessary to utilize DBTS in conjunction with catalytic cracking. An object of the present study is to provide an Mg-Al-mesoporous silica catalyst for the removal of sulfur dioxide from DBTS. The characteristics of the Mg-Al-mesoporous silica catalyst were investigated through N2 adsorption, XRD, ICP, and XRF. An Mg-Al-mesoporous silica catalyst formulated in a direct incorporation method showed higher catalytic performance compared to pure MgO during the catalytic removal of sulfur dioxide from DBTS. The higher dispersion of Mg as well as the large surface area of the Mg-Al-mesoporous silica catalyst strongly influenced the catalyst basicity in DBTS cracking.

  17. Thermophysical properties of uranium dioxide

    International Nuclear Information System (INIS)

    Fink, J.K.

    2000-01-01

    Experimental data on thermodynamic and transport properties of solid and liquid UO 2 have been reviewed and analyzed to obtain consistent equations for the thermophysical properties. Thermodynamic properties that have been assessed include enthalpy, heat capacity, enthalpy of fusion, thermal expansion, density, surface tension and vapor pressure. Transport properties that have been assessed are thermal diffusivity, thermal conductivity, viscosity, emissivity and optical constants. The assessments include a review of the experiments and data, review of previous recommendations, analysis of data to obtain new recommendations, determination of uncertainties in the recommended values, and comparisons of new recommendations with data and previous recommendations

  18. A kinetic study of cation release from a mixed mineral assemblage: implications for determination of uranium uptake

    International Nuclear Information System (INIS)

    Fenton, B.R.; Waite, T.D.

    1996-01-01

    The uptake of U(VI) as UO 2+ 2 on a natural complex mineral assemblage has been studied using batch selective chemical extraction techniques and secondary ion mass spectroscopy (SIMS). Sediments used in the study consisted of a quartz/mica schist collected from the locale of the Koongarra Uranium ore body, Alligator Rivers Uranium Province, Northern Territory, Australia. The bulk sediment was gravity separated into four size fractions, with attention focused on the nominally <25 μm and 250-1000 μm fractions of the bulk sample, in order to assess the effects of particle size on uranium uptake. Investigation of the kinetics of elemental release in the presence of selective extractants show that uranium is bound largely within the iron and aluminium oxyhydroxides of the assemblage, with a highly mobile fraction of this associated with aluminol sites. SIMS analysis of the natural substrate confirms that significant quantities of aluminium are present in surface layers. The effect of particle size on the uptake of uranium indicates very little change with respect to particle size. This finding may be attributed to the presence of highly porous surface coatings. (orig.)

  19. Preparation of UO_2 Fine Particle by Hydrolysis of Uranium(IV) Alkoxide

    OpenAIRE

    Satoh, Isamu; Takahashi, Mitsuyuki; Miura, Shigeyuki

    1997-01-01

    Fine particles of uranium(IV) dioxides were obtained by hydrolysis of uranium(IV) ethoxide which was synthesized by reacting uranium tetrachloride with sodium ethoxide. The monodispersed submicrometer particles were confirmed by SEM observation.

  20. Recovery of uranium values

    International Nuclear Information System (INIS)

    Rowden, G.A.

    1982-01-01

    A process is provided for the recovery of uranium from an organic extractant phase containing an amine. The extractant phase is contacted in a number of mixing stages with an acidic aqueous stripping phase containing sulphate ions, and the phases are passed together through a series of mixing stages while maintaining a dispersion of droplets of one phase in the other. Uranium is precipitated from the final stage by raising the pH. An apparatus having several mixing chambers is described

  1. Electrical impedance studies of uranium oxide

    International Nuclear Information System (INIS)

    Hampton, R.N.

    1986-11-01

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

  2. METHOD OF RECOVERING URANIUM COMPOUNDS

    Science.gov (United States)

    Poirier, R.H.

    1957-10-29

    S>The recovery of uranium compounds which have been adsorbed on anion exchange resins is discussed. The uranium and thorium-containing residues from monazite processed by alkali hydroxide are separated from solution, and leached with an alkali metal carbonate solution, whereby the uranium and thorium hydrorides are dissolved. The carbonate solution is then passed over an anion exchange resin causing the uranium to be adsorbed while the thorium remains in solution. The uranium may be recovered by contacting the uranium-holding resin with an aqueous ammonium carbonate solution whereby the uranium values are eluted from the resin and then heating the eluate whereby carbon dioxide and ammonia are given off, the pH value of the solution is lowered, and the uranium is precipitated.

  3. Study of the catalytic activity of mixed non-stoichiometric uranium-thorium oxides in carbon monoxide oxidation; Etude de l'activite catalytique des oxydes mixtes d'uranium et de thorium non stoechiometriques dans l'oxydation du monoxyde de carbone

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-06-01

    The aim of this work has been to study the catalytic properties of non-stoichiometric uranium-thorium oxides having the general formula U{sub x}Th{sub 1-x}O{sub 2+y}, for the oxidation of carbon monoxide. The preparation of pure, homogeneous, isotropic solids having good structural stability and a surface area as high as possible calls for a strict control of the conditions of preparation of these oxides right from the preparation of 'mother salts': the mixed oxalates U{sub x}Th{sub 1-x}(C{sub 2}O{sub 4}){sub 2}, 2H{sub 2}O. A study has been made of their physico-chemical properties (overall and surface chemical constitution, texture, structure, electrical conductivity), as well as of their adsorption properties with respect to gaseous species occurring in the catalytic reaction. This analysis has made it possible to put forward a reaction mechanism based on successive oxidations and reductions of the active surface by the reactants. A study of the reactions kinetics has confirmed the existence of this oxidation-reduction mechanism which only occurs for oxides having a uranium content of above 0.0014. The carbon dioxide produced by the reaction acts as an inhibitor by blocking the sites on which carbon monoxide can be adsorbed. These non-stoichiometric mixed oxides are a particularly clear example of catalysis by oxygen exchange between the solid and the gas phase. (author) [French] Ce travail a pour but l'etude des proprietes catalytiques des oxydes mixtes d'uranium et de thorium non stoechiometriques de formule generale U{sub x}Th{sub 1-x}O{sub 2+y} dans l'oxydation du monoxyde de carbone. L'obtention de solides purs, homogenes, isotropes, de bonne stabilite structurale et d'aire specifique aussi elevee que possible, exige de controler rigoureusement les conditions de preparation de ces oxydes des l'elaboration de leurs 'ascendants': les oxalates mixtes U{sub x}Th{sub 1-x}(C{sub 2}O{sub 4}){sub 2}, 2H{sub 2}O. Leurs proprietes physico-chimiques (composition

  4. Impact of mixed low-molecular-weight organic acids on uranium accumulation and distribution in a variant of mustard (Brassica juncea var. tumida)

    International Nuclear Information System (INIS)

    Fangfang Qi; Dingna Wang; Lijian Ma; Yongdong Jin; Liang Du; Dong Zhang; Chuanqin Xia; Sichuan University, Chengdu

    2014-01-01

    The impact of a mixture of low-molecular-weight organic acids (LMWOAs) composed of CA/MA/OA/LA with a molar ratio of 2.5:2.31:1.15:0.044 on uranium (U) accumulation and distribution in mustard (Brassica juncea var. tumida) was studied in this paper in order to understand the mechanism of rhizosphere-exudation assisted phytoremediation by hydroponic and pot culture experiments. The impact of the mixture of LWMOAs (Mix) on U accumulation showed that in hydroponic conditions Mix could enhance U translocation from root-to-shoot in mustard, but inhibit U uptake in root. In pot experiments, Mix enhanced both root and shoot U accumulation in mustard. The time-dependent kinetics of U uptake in mustard on Mix treatment showed that U content in plant shoots and roots increased with time increasing, and the steady state conditions were obtained at the 8th and 5th day with the U content of 1,528 and 2,300 mg/kg, respectively. Transmission electron microscope and energy dispersive X-ray spectrometry analysis for mustard roots showed that U was mainly observed on cell membrane of mustard roots on Mix treatment. This study would provide new insights for the mixture of LWMOAs-assisted phytoremediation of U-contaminated soil. (author)

  5. Studies relating to construction materials to be used in different options for head end treatment in reprocessing of mixed carbide fuel of plutonium and uranium

    International Nuclear Information System (INIS)

    Rajan, S.K.; Palamalai, A.; Ravi, T.N.; Sampath, M.; Raman, V.R.; Balasubramanian, G.R.

    1993-01-01

    Mixed carbide of uranium and plutonium has been chosen as the fuel for the first core of Fast Breeder Test Reactor, installed in the Indira Gandhi Centre for Atomic Research. Reprocessing of this fuel is one of the vital steps to prove the viability of the fuel cycle. The head end treatment process introduces constraints in the reprocessing of carbide fuel when compared to the commonly used mixed oxide fuel. Three head end processes, namely direct oxidation, pyrohydrolysis and direct dissolution in nitric acid with oxidation of organic acids were considered for study for exercising the choice. The paper briefly describes the three processes. In each process probable material of construction and related problems are discussed. (author). 3 refs, 5 figs, 7 tabs

  6. The possibility of the mixed valence state in the uranium intermetallic compounds: UCoGa5, U2Ru2Sn and U2RuGa8

    International Nuclear Information System (INIS)

    Troc, Robert

    2007-01-01

    The mixed valence (MV) phenomenon has been observed so far in a large number of various compounds but containing only lanthanides. These properties are usually associated with the mixing of the localised f-state and the band states. The usual valence state for magnetic uranium intermetallics is the trivalent state 5f 3 or hybridised 5f 2 6d 1 , both are nearly degenerate in energy and can compete for a stability of the compound. In some cases a gain in an energy minimum may be achieved by very fast fluctuating between these two states with a time of 10 -14 s, which does not allow to yield the ordered state even if the exchange interactions (favourite the U-U distances) would be able for that. The latter cases seem to concern the described here intermetallics: one ternary compound based on Co, UCoGa 5 , and the two uranium ternary compounds based on Ru, namely U 2 Ru 2 Sn and U 2 RuGa 8 which all crystallize in a tetragonal unit cell. All these compounds show a maximum in their temperature dependences of the magnetic susceptibility measured along and perpendicular to the c-axis. Such a behaviour, which is reminiscent of a number of Ce (Sm, Eu) and Yb compounds for which χ(T) has in the past been considered by Sales and Wohlleben (SW) by applying their ICF model or by Lawrance et al. following their scaling procedure. It turned out that these phenomenological models can also be applied to the considered here two Ru-based uranium ternaries from which some reliable energy parameters could be found. In order to further support the mixing valence scenario for the first such cases in uranium compounds presented here, the transport and thermodynamic properties are also discussed. However, some of the most important results confirming the MV state, e.g., in U 2 RuGa 8 , has recently been achieved from the inelastic neutron scattering performed in the Rutherford Appleton Laboratory on the ISIS facility. From these measurements a characteristic gap of 60 meV has been

  7. Spatiotemporal variation of radon and carbon dioxide concentrations in an underground quarry: coupled processes of natural ventilation, barometric pumping and internal mixing.

    Science.gov (United States)

    Perrier, Frédéric; Richon, Patrick

    2010-04-01

    Radon-222 and carbon dioxide concentrations have been measured during several years at several points in the atmosphere of an underground limestone quarry located at a depth of 18 m in Vincennes, near Paris, France. Both concentrations showed a seasonal cycle. Radon concentration varied from 1200 to 2000 Bq m(-3) in summer to about 800-1400 Bq m(-3) in winter, indicating winter ventilation rates varying from 0.6 to 2.5 x 10(-6) s(-1). Carbon dioxide concentration varied from 0.9 to 1.0% in summer, to about 0.1-0.3% in winter. Radon concentration can be corrected for natural ventilation using temperature measurements. The obtained model also accounts for the measured seasonal variation of carbon dioxide. After correction, radon concentrations still exhibit significant temporal variation, mostly associated with the variation of atmospheric pressure, with coupling coefficients varying from -7 to -26 Bq m(-3) hPa(-1). This variation can be accounted for using a barometric pumping model, coupled with natural ventilation in winter, and including internal mixing as well. After correction, radon concentrations exhibit residual temporal variation, poorly correlated between different points, with standard deviations varying from 3 to 6%. This study shows that temporal variation of radon concentrations in underground cavities can be understood to a satisfactory level of detail using non-linear and time-dependent modelling. It is important to understand the temporal variation of radon concentrations and the limitations in their modelling to monitor the properties of natural or artificial underground settings, and to be able to assess the existence of new processes, for example associated with the preparatory phases of volcanic eruptions or earthquakes. Copyright 2009 Elsevier Ltd. All rights reserved.

  8. Spatiotemporal variation of radon and carbon dioxide concentrations in an underground quarry: coupled processes of natural ventilation, barometric pumping and internal mixing

    International Nuclear Information System (INIS)

    Perrier, Frederic; Richon, Patrick

    2010-01-01

    Radon-222 and carbon dioxide concentrations have been measured during several years at several points in the atmosphere of an underground limestone quarry located at a depth of 18 m in Vincennes, near Paris, France. Both concentrations showed a seasonal cycle. Radon concentration varied from 1200 to 2000 Bq m -3 in summer to about 800-1400 Bq m -3 in winter, indicating winter ventilation rates varying from 0.6 to 2.5 x 10 -6 s -1 . Carbon dioxide concentration varied from 0.9 to 1.0% in summer, to about 0.1-0.3% in winter. Radon concentration can be corrected for natural ventilation using temperature measurements. The obtained model also accounts for the measured seasonal variation of carbon dioxide. After correction, radon concentrations still exhibit significant temporal variation, mostly associated with the variation of atmospheric pressure, with coupling coefficients varying from -7 to -26 Bq m -3 hPa -1 . This variation can be accounted for using a barometric pumping model, coupled with natural ventilation in winter, and including internal mixing as well. After correction, radon concentrations exhibit residual temporal variation, poorly correlated between different points, with standard deviations varying from 3 to 6%. This study shows that temporal variation of radon concentrations in underground cavities can be understood to a satisfactory level of detail using non-linear and time-dependent modelling. It is important to understand the temporal variation of radon concentrations and the limitations in their modelling to monitor the properties of natural or artificial underground settings, and to be able to assess the existence of new processes, for example associated with the preparatory phases of volcanic eruptions or earthquakes.

  9. Carbon dioxide dangers demonstration model

    Science.gov (United States)

    Venezky, Dina; Wessells, Stephen

    2010-01-01

    Carbon dioxide is a dangerous volcanic gas. When carbon dioxide seeps from the ground, it normally mixes with the air and dissipates rapidly. However, because carbon dioxide gas is heavier than air, it can collect in snowbanks, depressions, and poorly ventilated enclosures posing a potential danger to people and other living things. In this experiment we show how carbon dioxide gas displaces oxygen as it collects in low-lying areas. When carbon dioxide, created by mixing vinegar and baking soda, is added to a bowl with candles of different heights, the flames are extinguished as if by magic.

  10. Ground-based measurement of column-averaged mixing ratios of methane and carbon dioxide in the Sichuan Basin of China by a desktop optical spectrum analyzer

    Science.gov (United States)

    Qin, Xiu-Chun; Nakayama, Tomoki; Matsumi, Yutaka; Kawasaki, Masahiro; Ono, Akiko; Hayashida, Sachiko; Imasu, Ryoichi; Lei, Li-Ping; Murata, Isao; Kuroki, Takahiro; Ohashi, Masafumi

    2018-01-01

    Remote sensing of the atmospheric greenhouse gases, methane (CH4) and carbon dioxide (CO2), contributes to the understanding of global warming and climate change. A portable ground-based instrument consisting of a commercially available desktop optical spectrum analyzer and a small sun tracker has been applied to measure the column densities of atmospheric CH4 and CO2 at Yanting observation station in a mountainous paddy field of the Sichuan Basin from September to November 2013. The column-averaged dry-air molar mixing ratios, XCH4/XCO2, are compared with those retrieved by satellite observations in the Sichuan Basin and by ground-based network observations in the same latitude zone as the Yanting observation station.

  11. Contribution to the study of sputtering and damage of uranium dioxide by fast heavy ions; Contribution a l'etude de la pulverisation et de l'endommagement du dioxyde d'uranium par les ions lourds rapides

    Energy Technology Data Exchange (ETDEWEB)

    Schlutig, S

    2001-03-01

    Swift heavy ion-solid interaction leads in volume to track creation and on the surface to the ejection of particles into the vacuum. To learn more about initial mechanisms of track formation, we are focused on the sputtering of uranium dioxide by fast heavy ions. This present study is exclusively devoted to the influence of the electronic stopping power on the emission of neutral particles and especially on their angular distribution. These measurements are completed by those of the ions emitted from UO{sub 2} targets bombarded with swift heavy ions. The whole experimental results give access to: i) the nature of the sputtered particles; ii) the charge state of the emitted particles; iii) the direction of ejection of the sputtered particles ; iv) the sputtering yields deduced from the angular distributions. These results are compared to the prediction of the sputtering models proposed in the literature and it seems that the supersonic gas flow model is well suited to describe our results. Finally, the sputtering yields are compared with a set of earlier experimental data on uranium dioxide damage obtained by T. Wiss and we observe that only a small fraction of UO{sub 2} monolayers are sputtered. (author)

  12. Recycling of reprocessed uranium

    International Nuclear Information System (INIS)

    Randl, R.P.

    1987-01-01

    Since nuclear power was first exploited in the Federal Republic of Germany, the philosophy underlying the strategy of the nuclear fuel cycle has been to make optimum use of the resource potential of recovered uranium and plutonium within a closed fuel cycle. Apart from the weighty argument of reprocessing being an important step in the treatment and disposal of radioactive wastes, permitting their optimum ecological conditioning after the reprocessing step and subsequent storage underground, another argument that, no doubt, carried weight was the possibility of reducing the demand of power plants for natural uranium. In recent years, strategies of recycling have emerged for reprocessed uranium. If that energy potential, too, is to be exploited by thermal recycling, it is appropriate to choose a slightly different method of recycling from the one for plutonium. While the first generation of reprocessed uranium fuel recycled in the reactor cuts down natural uranium requirement by some 15%, the recycling of a second generation of reprocessed, once more enriched uranium fuel helps only to save a further three per cent of natural uranium. Uranium of the second generation already carries uranium-232 isotope, causing production disturbances, and uranium-236 isotope, causing disturbances of the neutron balance in the reactor, in such amounts as to make further fabrication of uranium fuel elements inexpedient, even after mixing with natural uranium feed. (orig./UA) [de

  13. Process for in-situ leaching of uranium

    International Nuclear Information System (INIS)

    Espenscheid, W.F.; Yan, F.Y.

    1983-01-01

    The present invention relates to the recovery of uranium from subterranean ore deposits, and more particularly to an in-situ leaching operation employing an aqueous solution of sulfuric acid and carbon dioxide as the lixiviant. Uranium is solubilized in the lixiviant as it traverses the subterranean uranium deposit. The lixiviant is subsequently recovered and treated to remove the uranium

  14. Applications of fluorescence techniques to the study of uranium in homogeneous and heterogeneous environments: hydrolysis and photo-reduction reactions on titanium dioxide

    International Nuclear Information System (INIS)

    Eliet, Veronique

    1996-01-01

    This thesis describes the use of Time-Resolved Fluorescence to characterise the spectroscopy of hydroxo-complexes of hexavalent Uranium, and to study photochemical reactions involving these species at mineral/water interfaces. The instrumentation used comprised of either an excimer laser coupled to an optical multichannel analyser OMA or a Nd-YAG laser coupled to a stroboscopic photomultiplier. The hydrolysis of Uranium at a constant temperature of 25 deg. C, has been studied in the pH ranges 0-5 and 9-12. Deconvolution of spectra and fluorescence decay curves for Uranium yielded individual fluorescence spectra and decay times for uranyl UO 2 2+ and its hydroxo-complexes UO 2 OH + , (UO 2 )2(OH) 2 2+ , (UO 2 ) 3 (OH) 5 + et UO 2 (OH) 3 - . The comparison of fluorescence efficiencies for the various species showed that the complex (UO 2 )2(OH) 2 2+ is up to 85 times more fluorescent than uranyl, depending on the emission wavelength. Further, investigations of fluorescence decays as a function of temperature in the pH range 0-6, yielded activation energies for the various Uranium hydroxo species. The knowledge gained in homogeneous media served in the study of the photochemical behaviour of Uranium in suspensions of the semi-conductor mineral, TiO 2 . After UV-light absorption, charge carriers formed at the mineral surface were found to reduce hexavalent Uranium to the tetravalent oxidation state. Time-Resolved Fluorescence Spectroscopy has been used to monitor the kinetics of the oxidation state change. A reaction mechanism is proposed on the basis of results obtained by studying the kinetics of the process at different values of pH The role of humic substances on the heterogeneous redox reaction has also been examined. (author) [fr

  15. Terminal uranium(V/VI) nitride activation of carbon dioxide and carbon disulfide. Factors governing diverse and well-defined cleavage and redox reactions

    International Nuclear Information System (INIS)

    Cleaves, Peter A.; Gardner, Benedict M.; Liddle, Stephen T.; Kefalidis, Christos E.; Maron, Laurent; Tuna, Floriana; McInnes, Eric J.L.; Lewis, William

    2017-01-01

    The reactivity of terminal uranium(V/VI) nitrides with CE 2 (E=O, S) is presented. Well-defined C=E cleavage followed by zero-, one-, and two-electron redox events is observed. The uranium(V) nitride [U(Tren TIPS )(N)][K(B15C5) 2 ] (1, Tren TIPS =N(CH 2 CH 2 NSiiPr 3 ) 3 ; B15C5=benzo-15-crown-5) reacts with CO 2 to give [U(Tren TIPS )(O)(NCO)][K(B15C5) 2 ] (3), whereas the uranium(VI) nitride [U(Tren TIPS )(N)] (2) reacts with CO 2 to give isolable [U(Tren TIPS )(O)(NCO)] (4); complex 4 rapidly decomposes to known [U(Tren TIPS )(O)] (5) with concomitant formation of N 2 and CO proposed, with the latter trapped as a vanadocene adduct. In contrast, 1 reacts with CS 2 to give [U(Tren TIPS )(κ 2 -CS 3 )][K(B15C5) 2 ] (6), 2, and [K(B15C5) 2 ][NCS] (7), whereas 2 reacts with CS 2 to give [U(Tren TIPS )(NCS)] (8) and ''S'', with the latter trapped as Ph 3 PS. Calculated reaction profiles reveal outer-sphere reactivity for uranium(V) but inner-sphere mechanisms for uranium(VI); despite the wide divergence of products the initial activation of CE 2 follows mechanistically related pathways, providing insight into the factors of uranium oxidation state, chalcogen, and NCE groups that govern the subsequent divergent redox reactions that include common one-electron reactions and a less-common two-electron redox event. Caution, we suggest, is warranted when utilising CS 2 as a reactivity surrogate for CO 2 . (copyright 2017 Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)

  16. Soil Characterization by Large Scale Sampling of Soil Mixed with Buried Construction Debris at a Former Uranium Fuel Fabrication Facility

    International Nuclear Information System (INIS)

    Nardi, A.J.; Lamantia, L.

    2009-01-01

    Recent soil excavation activities on a site identified the presence of buried uranium contaminated building construction debris. The site previously was the location of a low enriched uranium fuel fabrication facility. This resulted in the collection of excavated materials from the two locations where contaminated subsurface debris was identified. The excavated material was temporarily stored in two piles on the site until a determination could be made as to the appropriate disposition of the material. Characterization of the excavated material was undertaken in a manner that involved the collection of large scale samples of the excavated material in 1 cubic meter Super Sacks. Twenty bags were filled with excavated material that consisted of the mixture of both the construction debris and the associated soil. In order to obtain information on the level of activity associated with the construction debris, ten additional bags were filled with construction debris that had been separated, to the extent possible, from the associated soil. Radiological surveys were conducted of the resulting bags of collected materials and the soil associated with the waste mixture. The 30 large samples, collected as bags, were counted using an In-Situ Object Counting System (ISOCS) unit to determine the average concentration of U-235 present in each bag. The soil fraction was sampled by the collection of 40 samples of soil for analysis in an on-site laboratory. A fraction of these samples were also sent to an off-site laboratory for additional analysis. This project provided the necessary soil characterization information to allow consideration of alternate options for disposition of the material. The identified contaminant was verified to be low enriched uranium. Concentrations of uranium in the waste were found to be lower than the calculated site specific derived concentration guideline levels (DCGLs) but higher than the NRC's screening values. The methods and results are presented

  17. Geology and recognition criteria for sandstone uranium deposits in mixed fluvial-shallow marine sedimentary sequences, South Texas. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Adams, S.S.; Smith, R.B.

    1981-01-01

    Uranium deposits in the South Texas Uranium Region are classical roll-type deposits that formed at the margin of tongues of altered sandstone by the encroachment of oxidizing, uraniferous solutions into reduced aquifers containing pyrite and, in a few cases, carbonaceous plant material. Many of the uranium deposits in South Texas are dissimilar from the roll fronts of the Wyoming basins. The host sands for many of the deposits contain essentially no carbonaceous plant material, only abundant disseminated pyrite. Many of the deposits do not occur at the margin of altered (ferric oxide-bearing) sandstone tongues but rather occur entirely within reduced, pyurite-bearing sandstone. The abundance of pyrite within the sands probably reflects the introduction of H/sub 2/S up along faults from hydrocarbon accumulations at depth. Such introductions before ore formation prepared the sands for roll-front development, whereas post-ore introductions produced re-reduction of portions of the altered tongue, leaving the deposit suspended in reduced sandstone. Evidence from three deposits suggests that ore formation was not accompanied by the introduction of significant amounts of H/sub 2/S.

  18. Geology and recognition criteria for sandstone uranium deposits in mixed fluvial-shallow marine sedimentary sequences, South Texas. Final report

    International Nuclear Information System (INIS)

    Adams, S.S.; Smith, R.B.

    1981-01-01

    Uranium deposits in the South Texas Uranium Region are classical roll-type deposits that formed at the margin of tongues of altered sandstone by the encroachment of oxidizing, uraniferous solutions into reduced aquifers containing pyrite and, in a few cases, carbonaceous plant material. Many of the uranium deposits in South Texas are dissimilar from the roll fronts of the Wyoming basins. The host sands for many of the deposits contain essentially no carbonaceous plant material, only abundant disseminated pyrite. Many of the deposits do not occur at the margin of altered (ferric oxide-bearing) sandstone tongues but rather occur entirely within reduced, pyurite-bearing sandstone. The abundance of pyrite within the sands probably reflects the introduction of H 2 S up along faults from hydrocarbon accumulations at depth. Such introductions before ore formation prepared the sands for roll-front development, whereas post-ore introductions produced re-reduction of portions of the altered tongue, leaving the deposit suspended in reduced sandstone. Evidence from three deposits suggests that ore formation was not accompanied by the introduction of significant amounts of H 2 S

  19. Mixed

    Directory of Open Access Journals (Sweden)

    Pau Baya

    2011-05-01

    Full Text Available Remenat (Catalan (Mixed, "revoltillo" (Scrambled in Spanish, is a dish which, in Catalunya, consists of a beaten egg cooked with vegetables or other ingredients, normally prawns or asparagus. It is delicious. Scrambled refers to the action of mixing the beaten egg with other ingredients in a pan, normally using a wooden spoon Thought is frequently an amalgam of past ideas put through a spinner and rhythmically shaken around like a cocktail until a uniform and dense paste is made. This malleable product, rather like a cake mixture can be deformed pulling it out, rolling it around, adapting its shape to the commands of one’s hands or the tool which is being used on it. In the piece Mixed, the contortion of the wood seeks to reproduce the plasticity of this slow heavy movement. Each piece lays itself on the next piece consecutively like a tongue of incandescent lava slowly advancing but with unstoppable inertia.

  20. Preparation of Uranium Dioxide by Electrochemical Reduction in Ammonium Carbonate Solutions and Subsequent Precipitation; Preparation de bioxyde d'uranium par reduction electrochimique dans des solutions de carbonate d'ammonium et precipitation; Prigotovlenie dvuokisi urana metodom ehlektrokhimicheskogo vosstanovleniya v rastvore karbonata ammoniya s posleduyushchim osazhdeniem; Preparacion de dioxido de uranio por reduccion electroquimica en soluciones de carbonato amonico u precipitacion subsiguiente

    Energy Technology Data Exchange (ETDEWEB)

    Pravdic, V.; Branica, M.; Pucar, Z. [Department of Physical Chemistry, Rudjer Boskovic Institute, Zagreb, Yugoslavia (Serbia)

    1963-11-15

    Experiments in a small scale electrolysis cell on cathodic reduction of uranium (VI) to uranium (IV) show the possibility of an efficient way to obtain uranium (IV) in carbonate solutions. From this solution uranium (IV) hydrous oxide precipitates by merely raising the temperature. To obtain larger quantities of material needed for technological testing, a scale-up of the process was attempted. An electrolysis cell of hard PVC (polyvinylchloride) was constructed with a mercury pool cathode of approximately 2.5 dm{sup 2} and platinum anodes. The catholyte was separated from the anolyte by cationexchange membranes. The catholyte was circulated between two 50-1 reservoirs and streamed toward the vigorously stirred mercury cathode. The working potential of mercury was controlled against an Ag/AgCl/KC1 (sat.) reference electrode, the potential being held constant at -1.5 V. The current efficiency is approximately 90%; the power consumed for the reduction process is about 0.8 kWh/kg of uranium dioxide. After the electrolysis was completed the precipitation was initiated only by heating the deeply green clear solution up to 70 deg. C in a separate all-glass vessel of 60-1 volume. From 50, 1 of the catholyte solution 1 kg of a centrifuged product (containing about 20% of water) was obtained. The coulometric analysis of the oxygen-uranium ratio always gave results in the range of 2.04 to 2.09. By the procedure described uranium (IV) hydrous oxide is selectively precipitated, and the oxygen-uranium ratio in the precipitate was found to be independent of the degree of completion of the reduction. The product was identified as the alpha phase of uranium dioxide by the X-ray powder diffraction. Experiments in sintering and characterization of uranium dioxide thus obtained for the ceramic nuclear fuel requirements are under way. (author) [French] Des experiences faites dans une petite cellule d'electrolyse sur la reduction cathodique d'uranium (VI) en uranium (IV) montrent qu

  1. Scaling properties for the first RE-like mixed valence examples in uranium compounds: U2Ru2Sn and U2RuGa8

    International Nuclear Information System (INIS)

    Troc, Robert

    2006-01-01

    The present study was motivated by the scaling characterization of the first example of mixed valence (MV) RE-like behaviour found recently among intermetallic ternary uranium compounds. The χ(T) function for both title compounds has been fitted to the interconfigurational fluctuation (ICF) model of Sales and Wohlleben in order to determine the characteristic fluctuation temperatures T sf and interconfigurational excitation energies E ex . A good scaling, with similar values of T sf like from those derived from the ICF model, has been achieved for both these ternaries by plotting Tχ(T)/C against the reduced T/T sf . Moreover, this scaling follows almost exactly those found earlier in a number of MV- RE compounds

  2. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 2: A survey of the accuracy of the Studsvik of America CMS codes

    International Nuclear Information System (INIS)

    Demaziere, C.

    1999-02-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). The tools that are available to perform a modeling in the Department of Reactor Physics in Chalmers are CASMO-4/TABLES-3/SIMULATE-3 from Studsvik of America. Thus, before performing any kind of calculation with MOx fuels, it is necessary to be able to establish the reliability and the accuracy of these Core Management System (CMS) codes. This report presents a quantitative analysis of the models used in the package. A qualitative presentation is realized in a coming report

  3. Field measurements of mixed exposure of operators to radioactive aerosol, gas and quartz in confinement of mining equipment cabs during open-pit mining of high-grade uranium ores

    International Nuclear Information System (INIS)

    Yoshida, K.; Atiemo, M.A.; Markham, J.W.

    1982-07-01

    A series of field measurements of miners mixed exposure to radon and daughters, uranium ore dust and respirable quartz, was conducted in an open-pit mine in Northern Saskatchewan during 1980-81. Control of radon gas levels in the mining equipment cabs is required. Dust may be reduced by minimizing the resuspension of dust from contaminated surfaces within the cabs

  4. Terminal uranium(V/VI) nitride activation of carbon dioxide and carbon disulfide. Factors governing diverse and well-defined cleavage and redox reactions

    Energy Technology Data Exchange (ETDEWEB)

    Cleaves, Peter A.; Gardner, Benedict M.; Liddle, Stephen T. [School of Chemistry, The University of Manchester (United Kingdom); Kefalidis, Christos E.; Maron, Laurent [LPCNO, CNRS and INSA, Universite Paul Sabatier, Toulouse (France); Tuna, Floriana; McInnes, Eric J.L. [School of Chemistry and Photon Science Institute, The University of Manchester (United Kingdom); Lewis, William [School of Chemistry, The University of Nottingham (United Kingdom)

    2017-02-24

    The reactivity of terminal uranium(V/VI) nitrides with CE{sub 2} (E=O, S) is presented. Well-defined C=E cleavage followed by zero-, one-, and two-electron redox events is observed. The uranium(V) nitride [U(Tren{sup TIPS})(N)][K(B15C5){sub 2}] (1, Tren{sup TIPS}=N(CH{sub 2}CH{sub 2}NSiiPr{sub 3}){sub 3}; B15C5=benzo-15-crown-5) reacts with CO{sub 2} to give [U(Tren{sup TIPS})(O)(NCO)][K(B15C5){sub 2}] (3), whereas the uranium(VI) nitride [U(Tren{sup TIPS})(N)] (2) reacts with CO{sub 2} to give isolable [U(Tren{sup TIPS})(O)(NCO)] (4); complex 4 rapidly decomposes to known [U(Tren{sup TIPS})(O)] (5) with concomitant formation of N{sub 2} and CO proposed, with the latter trapped as a vanadocene adduct. In contrast, 1 reacts with CS{sub 2} to give [U(Tren{sup TIPS})(κ{sup 2}-CS{sub 3})][K(B15C5){sub 2}] (6), 2, and [K(B15C5){sub 2}][NCS] (7), whereas 2 reacts with CS{sub 2} to give [U(Tren{sup TIPS})(NCS)] (8) and ''S'', with the latter trapped as Ph{sub 3}PS. Calculated reaction profiles reveal outer-sphere reactivity for uranium(V) but inner-sphere mechanisms for uranium(VI); despite the wide divergence of products the initial activation of CE{sub 2} follows mechanistically related pathways, providing insight into the factors of uranium oxidation state, chalcogen, and NCE groups that govern the subsequent divergent redox reactions that include common one-electron reactions and a less-common two-electron redox event. Caution, we suggest, is warranted when utilising CS{sub 2} as a reactivity surrogate for CO{sub 2}. (copyright 2017 Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)

  5. Preparation of uranium dioxide by thermal decomposition and direct reduction of ammonium uranate; Preparacion del dioxido de uranio por descomposicion termica y reduccion directa del uranato de amonio

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez R, R

    1996-12-31

    The thermal decomposition of ammonium uranate has been studied by infrared spectroscopy, and X-ray diffraction. It has been show that ammonia remains in the solid until substantially 350 Centigrade degrees, when gaseous nitrogen is released. It is concluded that compounds derived from the calcination of ammonium uranate at atmospheric pressure, produced amorphous U O{sub 3} at about 350-400 Centigrade degrees and transform to U{sub 3} O{sub 8} via {alpha} - U O{sub 3} and/or {alpha} - U O{sub 3}. The object of this study was to obtain reliable fundamental information regarding the character of the pure carbon monoxide-ammonium uranate-uranium trioxide-uranium octaoxide reaction, in the range of temperatures that has been used in commercial reduction processes. Through the use of high-purity samples and by the proper control of incidental variable, this object was realized. (Author).

  6. An 8-year, high-resolution reanalysis of atmospheric carbon dioxide mixing ratios based on OCO-2 and GOSAT-ACOS retrievals

    Science.gov (United States)

    Weir, B.; Chatterjee, A.; Ott, L. E.; Pawson, S.

    2017-12-01

    This talk presents an overview of results from the GEOS-Carb reanalysis of retrievals of average-column carbon dioxide (XCO2) from the Orbiting Carbon Observatory 2 (OCO-2) and Greenhouse Gases Observing Satellite (GOSAT) satellite missions. The reanalysis is a Level 3 (L3) product: a collection of 3D fields of carbon dioxide (CO2) mixing ratios every 6 hours beginning in April 2009 going until the present on a grid with a 0.5 degree horizontal resolution and 72 vertical levels from the surface to 0.01 hPa. Using an assimilation methodology based on the Goddard Earth Observing System (GEOS) atmospheric data assimilation system (ADAS), the L3 fields are weighted averages of the two satellite retrievals and predictions from the GEOS general circulation model driven by assimilated meteorology from the Modern-Era Retrospective analysis for Research and Applications, Version 2 (MERRA-2). In places and times where there are a dense number of soundings, the observations dominate the predicted mixing ratios, while the model is used to fill in locations with fewer soundings, e.g., high latitudes and the Amazon. Blending the satellite observations with model predictions has at least two notable benefits. First, it provides a bridge for evaluating the satellite retrievals and their uncertainties against a heterogeneous collection of observations including those from surface sites, towers, aircraft, and soundings from the Total Carbon Column Observing Network (TCCON). Extensive evaluations of the L3 reanalysis clearly demonstrate both the strength and the deficiency of the satellite retrievals. Second, it is possible to estimate variables from the reanalysis without introducing bias due to spatiotemporal variability in sounding coverage. For example, the assimilated product provides robust estimates of the monthly CO2 global growth rate. These monthly growth rate estimates show significant differences from estimates based on in situ observations, which have sparse coverage

  7. Separation of carbon dioxide from flue gas by mixed matrix membranes using dual phase microporous polymeric constituents.

    Science.gov (United States)

    Sekizkardes, Ali K; Kusuma, Victor A; Dahe, Ganpat; Roth, Elliot A; Hill, Lawrence J; Marti, Anne; Macala, Megan; Venna, Surendar R; Hopkinson, David

    2016-09-27

    This study presents the fabrication of a new mixed matrix membrane using two microporous polymers: a polymer of intrinsic microporosity PIM-1 and a benzimidazole linked polymer, BILP-101, and their CO 2 separation properties from post-combustion flue gas. 17, 30 and 40 wt% loadings of BILP-101 into PIM-1 were tested, resulting in mechanically stable films showing very good interfacial interaction due to the inherent H-bonding capability of the constituent materials. Gas transport studies showed that BILP-101/PIM-1 membranes exhibit high CO 2 permeability (7200 Barrer) and selectivity over N 2 (15). The selected hybrid membrane was further tested for CO 2 separation using actual flue gas from a coal-fired power plant.

  8. A Mixed Matrix Membrane of Poly (4-methyl-1-pentyne Filled with MIL 53 Particles and Its Application in Carbon Dioxide and Nitrogen Separation

    Directory of Open Access Journals (Sweden)

    Reza Abedini

    2016-05-01

    Full Text Available The performance of poly (4-metyl-1-pentyne as mixed matrix membrane (MMM mixed with MIL 53 particles was studied to separate mixtures of carbon dioxide and nitrogen. MIL 53 particles was added to the polymer matrix with 10, 20 and 30 weight percentages. The adsorption of CO2 and N2 gases by MIL 53 was evaluated and the adsorption data was analyzed by Langmuir equation. Structure and thermal/mechanical properties of prepared membranes were characterized by means of FTIR, SEM, TGA and elongation test. Moreover, the gas permeation properties of membranes were studied by measuring the permeation of pure CO2 and N2. Furthermore, for accurate understanding of the gas permeation properties of the membranes, diffusion and solution coefficient of gases in neat membrane and MMMs were calculated using modified time-lag method. The results from TGA analysis showed that the degradation temperature of MMMs was enhanced and increased to 348ºC for membrane containing 30 wt% of MIL 53. The SEM images also illustrated a relatively uniform dispersion of particles with proper polymer/filler interfaces in the polymer matrix. The gas permeation results revealed that the permeability of both gases (especially CO2 increased with increasing MIL 53 loading, in which the permeability of CO2 increased from 98.74 Barrer in neat membrane to 217.65 Barrer in MMM containing 30 wt% filler. Moreover, calculation of CO2/N2 selectivity depicted that the selectivity enhanced from 16.66 to 22.70. Finally, the performance of MMMs was compared with Robeson’s upper bound in CO2/N2 separation and results showed that the MMM having 30 wt% of MIL 53 took over the Robeson bound.

  9. Spectroscopy and DFT studies of uranyl carbonate, rutherfordine, UO2CO3: a model for uranium transport, carbon dioxide sequestration, and seawater species

    Science.gov (United States)

    Kalashnyk, N.; Perry, D. L.; Massuyeau, F.; Faulques, E.

    2017-12-01

    Several optical microprobe experiments of the anhydrous uranium carbonate—rutherfordine—are presented in this work and compared to periodic density functional theory results. Rutherfordine is the simplest uranyl carbonate and constitutes an ideal model system for the study of the rich uranium carbonate family relevant for environmental sustainability. Micro-Raman, micro-reflectance, and micro-photoluminescence (PL) spectroscopy studies have been carried out in situ on native, micrometer-sized crystals. The sensitivity of these techniques is sufficient to analyze minute amounts of samples in natural environments without using x-ray analysis. In addition, very intense micro-PL and micro-reflectance spectra that were not reported before add new results on the ground and excited states of this mineral. The optical gap value determined experimentally is found at about 2.6-2.8 eV. Optimized geometry, band structure, and phonon spectra have been calculated. The main vibrational lines are identified and predicted by this theoretical study. This work is pertinent for optical spectroscopy, for identification of uranyl species in various environmental settings, and for nuclear forensic analysis.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

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

  11. EPR of uranium ions

    International Nuclear Information System (INIS)

    Ursu, I.; Lupei, V.

    1984-02-01

    A review of the electron paramagnetic resonance data on the uranium ions is given. After a general account of the electronic structure of the uranium free atoms and ions, the influence of the external fields (magnetic field, crystal fields) is discussed. The main information obtained from EPR studies on the uranium ions in crystals are emphasized: identification of the valence and of the ground electronic state, determination of the structure of the centers, crystal field effects, role of the intermediate coupling and of the J-mixing, role of the covalency, determination of the nuclear spin, maqnetic dipole moment and electric quadrupole moment of the odd isotopes of uranium. These data emphasize the fact that the actinide group has its own identity and this is accutely manifested at the beginning of the 5fsup(n) series encompassed by the uranium ions. (authors)

  12. Depleted uranium

    International Nuclear Information System (INIS)

    Huffer, E.; Nifenecker, H.

    2001-02-01

    This document deals with the physical, chemical and radiological properties of the depleted uranium. What is the depleted uranium? Why do the military use depleted uranium and what are the risk for the health? (A.L.B.)

  13. METHOD OF PRODUCING URANIUM METAL BY ELECTROLYSIS

    Science.gov (United States)

    Piper, R.D.

    1962-09-01

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

  14. Boiling water reactors with Uranium-Plutonium mixed oxide fuel. Report 1: Accuracy of the nuclide concentrations calculated by CASMO-4

    International Nuclear Information System (INIS)

    Demaziere, C.

    1999-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). The tools that are available to perform a modeling in the Department of Reactor Physics in Chalmers are CASMO-4/TABLES-3/SIMULATE-3 from Studsvik of America. These CMS (Core Management System) programs have been extensively compared with both measurements and reference codes. Nevertheless some data are proprietary in particular the comparison of the calculated nuclide concentrations versus experiments (because of the cost of this kind of experimental study). This is why this report describes such a comparative investigation carried out with a General Electric 7x7 BWR bundle. Unfortunately, since some core history parameters were unknown, a lot of hypotheses have been adopted. This invokes sometimes a significant discrepancy in the results without being able to determine the origin of the differences between calculations and experiments. Yet one can assess that, except for four nuclides - Plutonium-238, Curium-243, Curium-244 and Cesium-135 - for which the approximate power history (history effect) can be invoked, the accuracy of the calculated nuclide concentrations is rather good if one takes the numerous approximations into account

  15. A contribution to the study of the mixed uranium-plutonium mono-carbides containing small quantities of zirconium; Contribution a l'etude du monocarbure d'uranium et de plutonium avec de faibles additions de zirconium

    Energy Technology Data Exchange (ETDEWEB)

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

    1970-03-01

    We have studied a mixed monocarbide, type (U,Pu)C, containing small additions of zirconium for the application as a fast neutron reactor fuel. A preliminary study was conducted on the (U,Zr)C monocarbide (Report CEA-R-3765(1). It was found that small additions of zirconium to the uranium-plutonium monocarbide improve a number of properties such as atmospheric corrosion, the hardness, and particularly the compatibility with 316 stainless steel. However, properties such as the coefficient of expansion and the melting point are only slightly changed. The relative percentage of Pu/U+Pu in the monocarbide was fixed at 20 per cent. Two processes of fabrication were employed: casting in an arc furnace, sintering, carried out after having the hydrides of the metals carburized. The metallurgical results indicate, that the above mentioned fuel might be of interest for fast neutron reactor application. (author) [French] On a etudie un combustible de type carbure (U,Pu)C pour les reacteurs a neutrons rapides. Les recherches preliminaires ont porte sur le carbure (UZr)C (rapport CEA-R-3765(1)). L'addition de faibles quantites de zirconium (3 at. pour cent) au monocarbure (U,Pu)C, ameliore certaines proprietes, commee la tenue a la corrosion atmospherique, la durete et surtout la compatibilite avec l'acier inoxydable X-18 M, Par contre le coefficient de dilatation et la densite sont peu changes. Le rapport Pu/Pu+U etait fixe a 20 pour cent. Deux procedes de fabrication ont ete etudies: l'un par fusion a l'arc, l'autre par frittage a partir de metaux hydrures. Au vu des resultats metallurgiques obtenus le carbure (U,Pu,Zr)C semble presenter un interet certain. (auteur)

  16. A contribution to the study of the mixed uranium-plutonium mono-carbides containing small quantities of zirconium; Contribution a l'etude du monocarbure d'uranium et de plutonium avec de faibles additions de zirconium

    Energy Technology Data Exchange (ETDEWEB)

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

    1970-03-01

    We have studied a mixed monocarbide, type (U,Pu)C, containing small additions of zirconium for the application as a fast neutron reactor fuel. A preliminary study was conducted on the (U,Zr)C monocarbide (Report CEA-R-3765(1). It was found that small additions of zirconium to the uranium-plutonium monocarbide improve a number of properties such as atmospheric corrosion, the hardness, and particularly the compatibility with 316 stainless steel. However, properties such as the coefficient of expansion and the melting point are only slightly changed. The relative percentage of Pu/U+Pu in the monocarbide was fixed at 20 per cent. Two processes of fabrication were employed: casting in an arc furnace, sintering, carried out after having the hydrides of the metals carburized. The metallurgical results indicate, that the above mentioned fuel might be of interest for fast neutron reactor application. (author) [French] On a etudie un combustible de type carbure (U,Pu)C pour les reacteurs a neutrons rapides. Les recherches preliminaires ont porte sur le carbure (UZr)C (rapport CEA-R-3765(1)). L'addition de faibles quantites de zirconium (3 at. pour cent) au monocarbure (U,Pu)C, ameliore certaines proprietes, commee la tenue a la corrosion atmospherique, la durete et surtout la compatibilite avec l'acier inoxydable X-18 M, Par contre le coefficient de dilatation et la densite sont peu changes. Le rapport Pu/Pu+U etait fixe a 20 pour cent. Deux procedes de fabrication ont ete etudies: l'un par fusion a l'arc, l'autre par frittage a partir de metaux hydrures. Au vu des resultats metallurgiques obtenus le carbure (U,Pu,Zr)C semble presenter un interet certain. (auteur)

  17. Separation of thorium and uranium by liquid-liquid extraction from mixed aqueous nitric acidic-methanolic solutions

    International Nuclear Information System (INIS)

    Schmid, E.R.; Kenndler, E.

    1976-01-01

    A method is described for the separation of Th and U from each other and from other elements, usually occuring in minerals, by liquid-liquid extraction with Aliquat Nitrate (tricaprylmethyl ammoniumnitrate, 6 vol%) in benzene from a mixed solution of 2.5 M HNO 3 and methanol (1:1 volume ratio). Permissible upper concentration ratios of interfering elements, such as Li, Na, K, Mg, Ca, Al, Cu, Co(II), Fe(III), Mn(II), Ti(IV), La(III), U(VI), Cl - , ClO 4 - , SO 4 2- , PO 4 3- , have been determined. Following the separation, Th has been determined by spectrophotometry using Thorin, and U by fluorometry. Results for yield under varying conditions, together with elemental concentrations in the ppm range for U and Th in minerals, are given. (B.T.)

  18. Determination of rare earth elements, hafnium and cadmium in sintered pellets of mixed thorium and uranium oxides by neutron activation

    International Nuclear Information System (INIS)

    Cardoso, S.N.M.

    1987-01-01

    This work shows the development of a method for determination of the rare-earth elements (Eu, Sm, Dy and Gd), Hf and Cd contents in sinterized U and Th mixed oxides by neutron activation analysis. The sample is dissolved in nitric/fluoridric (0,1% HF) medium, to dryness and redissolved in 6N HCl solution. The Hf is extracted into organic phase (0,5 M TTA/benzene), irradiated and analysed through 181 Hf isotope energy peak. The aqueous phase is treated with NH 4 OH for the precipitation of hidroxides. Then, these are dissolved in 6N HNO 3 solution. The extraction of U and Th is made in two steps, one with TBP/CCl 4 and another with 0,5 M TTA/C 6 H 6 . Then the rare-earth elements and Cd are irradiated and determined by gamma spectrometry. (author) [pt

  19. Geobacter daltonii sp. nov., an Fe(III)- and uranium(VI)-reducing bacterium isolated from a shallow subsurface exposed to mixed heavy metal and hydrocarbon contamination.

    Science.gov (United States)

    Prakash, Om; Gihring, Thomas M; Dalton, Dava D; Chin, Kuk-Jeong; Green, Stefan J; Akob, Denise M; Wanger, Greg; Kostka, Joel E

    2010-03-01

    An Fe(III)- and uranium(VI)-reducing bacterium, designated strain FRC-32(T), was isolated from a contaminated subsurface of the USA Department of Energy Oak Ridge Field Research Center (ORFRC) in Oak Ridge, Tennessee, where the sediments are exposed to mixed waste contamination of radionuclides and hydrocarbons. Analyses of both 16S rRNA gene and the Geobacteraceae-specific citrate synthase (gltA) mRNA gene sequences retrieved from ORFRC sediments indicated that this strain was abundant and active in ORFRC subsurface sediments undergoing uranium(VI) bioremediation. The organism belonged to the subsurface clade of the genus Geobacter and shared 92-98 % 16S rRNA gene and 75-81 % rpoB gene sequence similarities with other recognized species of the genus. In comparison to its closest relative, Geobacter uraniireducens Rf4(T), according to 16S rRNA gene sequence similarity, strain FRC-32(T) showed a DNA-DNA relatedness value of 21 %. Cells of strain FRC-32(T) were Gram-negative, non-spore-forming, curved rods, 1.0-1.5 microm long and 0.3-0.5 microm in diameter; the cells formed pink colonies in a semisolid cultivation medium, a characteristic feature of the genus Geobacter. The isolate was an obligate anaerobe, had temperature and pH optima for growth at 30 degrees C and pH 6.7-7.3, respectively, and could tolerate up to 0.7 % NaCl although growth was better in the absence of NaCl. Similar to other members of the Geobacter group, strain FRC-32(T) conserved energy for growth from the respiration of Fe(III)-oxyhydroxide coupled with the oxidation of acetate. Strain FRC-32(T) was metabolically versatile and, unlike its closest relative, G. uraniireducens, was capable of utilizing formate, butyrate and butanol as electron donors and soluble ferric iron (as ferric citrate) and elemental sulfur as electron acceptors. Growth on aromatic compounds including benzoate and toluene was predicted from preliminary genomic analyses and was confirmed through successive transfer with

  20. In vitro and in vivo measurements of the dissolution parameters of uranium and plutonium mixed oxides in biological environment

    International Nuclear Information System (INIS)

    Matton, S.

    1999-01-01

    During the mixed-oxide fuel fabrication process, inhalation is potentially the main route of internal contamination. The International Commission on Radiological Protection recommends experimental measurement of parameters such as size and dissolution rate for specific industrial compounds. First, we validated the use of PERALS (Photon Electron Rejecting Alpha Liquid Scintillation) for alpha measurement in biological samples which, in some cases, could improve detection limit. We characterised physical chemical properties in terms of size, specific area and activity of 3 different powders: MOX made according to either the MIMAS process, which showed heterogeneous chemical composition, or the SOLGEL, which showed homogeneous chemical composition and industrial PuO 2 . Their dissolution parameters, f r and s s , as defined in the simplest model proposed by ICRP 66 were measured in vivo, after inhalation in the rat, and in vitro. The statistical variation of these values were expressed as standard deviation. Moreover, in vitro studies demonstrated variation of the s s value depending on the duration of the incubation. We also developed methods to characterise interactions between UO 2 particles and phosphate ions which could be involved in the actinide toxicity. (author) [fr

  1. Study of the oxidation risks during the sintering of uranium dioxide, and characterization of the excess oxygen; Etude du risque d'oxydation lors du frittage du bioxyde d'uranium et caracterisation de l'oxygene excedentaire

    Energy Technology Data Exchange (ETDEWEB)

    Conte, M; Brandela, M

    1966-05-01

    During sintering in reducing atmospheres, UO{sub 2} pellets can be oxidized by gaseous impurities. The effects of temperature cycles, the partial pressure of O{sub 2} and the flow rate of the gas over the pellets were investigated. In these atmospheres, the O{sub 2} partial pressure during sintering is low at high temperatures, as a consequence of the dissociation rate of the combined water, but below 1000 deg C, it can be high enough to result in a noticeable oxidation of the surface of the pellets during cooling. The crystalline phases which can occur have been identified and two methods of detection have been proposed: a micrographic examination after chemical etching and radiocrystallography. (authors) [French] Lors du frittage industriel du bioxyde d'uranium en atmosphere reductrice (hydrogene ou ammoniac dissocie) la presence d'impuretes oxydantes dans l'atmosphere peut provoquer l'oxydation des pastilles d'UO{sub 2}; les auteurs ont etudie les phenomenes en faisant varier le cycle de temperature, la pression partielle d'oxygene introduit dans l'hydrogene, la vitesse de passage du gaz sur les pastilles. Dans les atmospheres considerees la pression partielle d'oxygene au-dessus de l'UO{sub 2} en cours de frittage est faible a temperature elevee car elle resulte de la dissociation de l'eau formee, mais a t < 1000 degrees C elle, peut etre assez importante pour provoquer une oxydation notable de la surface des pastilles lors du refroidissement. Les phases cristallines susceptibles d'etre formees ont ete reperees et deux methodes de detection proposees: la micrographie apres attaque chimique specifique et la radiocristallographie. (auteurs)

  2. Uranium, its impact on the national and global energy mix; and its history, distribution, production, nuclear fuel-cycle, future, and relation to the environment

    Science.gov (United States)

    Finch, Warren Irvin

    1997-01-01

    The many aspects of uranium, a heavy radioactive metal used to generate electricity throughout the world, are briefly described in relatively simple terms intended for the lay reader. An adequate glossary of unfamiliar terms is given. Uranium is a new source of electrical energy developed since 1950, and how we harness energy from it is explained. It competes with the organic coal, oil, and gas fuels as shown graphically. Uranium resources and production for the world are tabulated and discussed by country and for various energy regions in the United States. Locations of major uranium deposits and power reactors in the United States are mapped. The nuclear fuel-cycle of uranium for a typical light-water reactor is illustrated at the front end-beginning with its natural geologic occurrence in rocks through discovery, mining, and milling; separation of the scarce isotope U-235, its enrichment, and manufacture into fuel rods for power reactors to generate electricity-and at the back end-the reprocessing and handling of the spent fuel. Environmental concerns with the entire fuel cycle are addressed. The future of the use of uranium in new, simplified, 'passively safe' reactors for the utility industry is examined. The present resource assessment of uranium in the United States is out of date, and a new assessment could aid the domestic uranium industry.

  3. Uranium exploration

    International Nuclear Information System (INIS)

    De Voto, R.H.

    1984-01-01

    This paper is a review of the methodology and technology that are currently being used in varying degrees in uranium exploration activities worldwide. Since uranium is ubiquitous and occurs in trace amounts (0.2 to 5 ppm) in virtually all rocks of the crust of the earth, exploration for uranium is essentially the search of geologic environments in which geologic processes have produced unusual concentrations of uranium. Since the level of concentration of uranium of economic interest is dependent on the present and future price of uranium, it is appropriate here to review briefly the economic realities of uranium-fueled power generation. (author)

  4. A family of uranium-carboxylic acid hybrid materials: synthesis, structure and mixed-dye selective adsorption

    International Nuclear Information System (INIS)

    Xue Gao; Jian Song; Yong Heng Xing; Feng Ying Bai; Li Xian Sun; Zhan Shi

    2016-01-01

    Four uranyl complexes (UO_2)_2(pht)_2(Hpac)_2(H_2O)_2 (pht = phthalic acid and Hpac = nicotinic acid) (1), (UO_2)(pac)_2(H_2O)_2 (2), [(UO_2)(CMA)_3][H_2N(CH_3)_2] (CMA = cinnamic acid) (3) and (UO_2)_2(C_2O_4)(μ_2- OH)_2(H_2O)_2H_2O (4) were synthesized by the reaction of UO_2(CH_3COO)_2.2H_2O as the metal source with phthalic acid, nicotinic acid, cinnamic acid or oxalic acid as the ligand. They were characterized by elemental analysis, IR, UV-vis, XRD, single crystal X-ray diffraction and thermal gravimetric analysis. The structural analysis showed that complexes 1, 2 and 3 were discrete structures, and by hydrogen bonding interactions, the adjacent molecular units are connected to form a three-dimensional (3D) supramolecular network structure for complex 1 and one-dimensional (1D) chains for complexes 2 and 3. Meanwhile, in the structure of complex 4, a tetrameric SBU (UO_2)_4(μ_2-OH)_4(H_2O)_4 is linked to a 2D layer through a bridging oxalic acid ligand, and furthermore extends the 2D layer into a 3D supramolecular architecture by hydrogen bonding interactions. In order to extend their functional properties, their photoluminescence, surface photovoltage and mixed-dye selective adsorption properties have been studied for the first time. Through experiments, we found that the adsorption performance of complex 3 was better than others, and the amount of adsorbed RhB was 4.22 mg.g"-"1. (authors)

  5. Comparison of open cycles of uranium and mixed oxides of thorium-uranium using advanced reactors; Comparação de ciclos abertos de urânio e óxidos mistos de tório-urânio utilizando reatores avançados

    Energy Technology Data Exchange (ETDEWEB)

    Gonçalves, Letícia C.; Maiorino, José R., E-mail: goncalves.leticiac@gmail.com [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil). Centro de Engenharia, Modelagem e Ciências Sociais Aplicadas

    2017-07-01

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

  6. Process for uranium separation and preparation of UO4.2NH3.2HF

    International Nuclear Information System (INIS)

    Dokuzoguz, H.Z.

    1976-01-01

    A process for treating the aqueous effluents that are produced in converting gaseous UF 6 (uranium hexafluoride) into solid UO 2 (uranium dioxide) by way of an intermediate (NH 4 ) 4 UO 2 (CO 3 ) 3 (''AUC'' Compound) is disclosed. These effluents, which contain large amounts of NH 4 + , CO 3 2- , F - , and a small amount of U are mixed with H 2 SO 4 (sulfuric acid) in order to expel CO 2 (carbon dioxide) and thereby reduce the carbonate concentration. The uranium is precipitated through treatment with H 2 O 2 (hydrogen peroxide) and the fluoride is easily recovered in the form of CaF 2 (calcium fluoride) by contacting the process liquid with CaO (calcium oxide). The presence of SO 4 2- (sulfate) in the process liquid during CaO contacting seems to prevent the development of a difficult-to-filter colloid. The process also provides for NH 3 recovery and recycling. Liquids discharged from the process, moreover, are essentially free of environmental pollutants. The waste treatment products, i.e., CO 2 , NH 3 , and U are economically recovered and recycled back into the UF 6 → UO 2 conversion process. The process, moreover, recovers the uranium as a precipitate in the second stage. This precipitate is a new inorganic chemical compound UO 4 .2NH 3 .2HF [uranyl peroxide-2-ammonia-2-(hydrogen fluoride)

  7. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 5: Analysis of the reactivity coefficients and the stability of a BWR loaded with MOx fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    2000-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). For this purpose, the Core Management System (CMS) codes of Studsvik Scandpower are used. This package is constituted by CASMO-4/TABLES-3/SIMULATE-3. It has been shown in previous reports that these codes are able to accurately represent and model MOx bundles. This report is thus devoted to the study of BWR cores loaded (partially or totally) with MOx bundles. The plutonium quality used is the Pu type 2016 (mostly Pu-239, 56 %, and Pu-240, 26 %), but a variation of the plutonium isotopic vector was also investigated, in case of a partial MOx loading. One notices that the reactivity coefficients do not present significant changes in comparison with a full UOx loading. Nevertheless, two main problems arise: the shutdown margin at BOC is lower than 1 % and the stability to in-phase oscillations is slightly decreased. (The SIMULATE-3 version used for this study does not contain the latest MOx enhancements described in literature, since these code developments have not been provided to the department. Nevertheless, as the nominal average enrichment of the MOx bundles is 5.41 % (total amount of plutonium), which can still be considered as a relatively low enrichment, the accuracy of the CMS codes is acceptable without the use of the MOx improvements for this level of Pu enrichment.

  8. Uranium complex recycling method of purifying uranium liquors

    International Nuclear Information System (INIS)

    Elikan, L.; Lyon, W.L.; Sundar, P.S.

    1976-01-01

    Uranium is separated from contaminating cations in an aqueous liquor containing uranyl ions. The liquor is mixed with sufficient recycled uranium complex to raise the weight ratio of uranium to said cations preferably to at least about three. The liquor is then extracted with at least enough non-interfering, water-immiscible, organic solvent to theoretically extract about all of the uranium in the liquor. The organic solvent contains a reagent which reacts with the uranyl ions to form a complex soluble in the solvent. If the aqueous liquor is acidic, the organic solvent is then scrubbed with water. The organic solvent is stripped with a solution containing at least enough ammonium carbonate to precipitate the uranium complex. A portion of the uranium complex is recycled and the remainder can be collected and calcined to produce U 3 O 8 or UO 2

  9. Oxidation and crystal field effects in uranium

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-06

    An extensive investigation of oxidation in uranium has been pursued. This includes the utilization of soft x-ray absorption spectroscopy, hard x-ray absorption near-edge structure, resonant (hard) x-ray emission spectroscopy, cluster calculations, and a branching ratio analysis founded on atomic theory. The samples utilized were uranium dioxide (UO2), uranium trioxide (UO3), and uranium tetrafluoride (UF4). As a result, a discussion of the role of non-spherical perturbations, i.e., crystal or ligand field effects, will be presented.

  10. The hydrolysis of thorium dicarbide and of mixed uranium-thorium dicarbides; L'hydrolyse du dicarbure de thorium et des dicarbures mixtes d'uranium et de thorium

    Energy Technology Data Exchange (ETDEWEB)

    Del Litto, B [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1966-09-01

    The hydrolysis of thorium dicarbide leads to the formation of a complex mixture of gaseous and condensed carbon hydrides. The temperature, between 25 and 100 deg. C, has no influence on the nature and composition of the gas phase. The reaction kinetics, however, are strongly temperature dependent. In a hydrochloric medium, an enrichment in hydrogen of the gas mixture is observed. On the other hand a decrease in hydrogen and an increase in acetylene content take place in an oxidizing medium. The general results can be satisfactorily interpreted through a reaction mechanism involving C-C radical groups. In the same way, the hydrolysis of uranium-thorium-carbon ternary alloys leads to the formation of gaseous and condensed carbon hydrides. The variation of the composition of the gas phase versus uranium content in the alloy suggests an hypothesis about the carbon-carbon distance in the alloy crystal lattice. The variation of methane content, on the other hand, has lead us to discuss the nature of the various phases present in uranium-carbon alloys and carbon-rich uranium-thorium-carbon alloys. We have reached the conclusion that these alloys include a proportion of monocarbide which is dependent upon the ratio. Th/(Th + U). We put forward a diagram of the system uranium-carbon with features proper to explain some phenomena which have been observed in the uranium-thorium-carbon ternary diagram. (author) [French] L'hydrolyse du dicarbure de thorium conduit a la formation d'un melange complexe d'hydrures de carbone gazeux et condenses. La temperature entre 25 et 100 deg. C n'a pas d'influence sur la nature ef la composition de la phase gazeuse. Par contre la cinetique en depend fortement. En milieu chlorhydrique, on observe un enrichissement en hydrogene du melange gazeux. Au contraire, en milieu oxydant il se produit une diminution du taux d'hydrogene et une augmentation tres nette du taux d'acetylene. L'ensemble des resultats obtenus peut etre interprete d'une maniere

  11. Thorium dioxide: properties and nuclear applications

    International Nuclear Information System (INIS)

    Belle, J.; Berman, R.M.

    1984-01-01

    This is the sixth book on reactor materials published under sponsorship of the Naval Reactors Office of the United States Department of Energy, formerly the United States Atomic Energy Commission. This book presents a comprehensive compilation of the most significant properties of thorium dioxide, much like the book Uranium Dioxide: Properties and Nuclear Applications presented information on the fuel material used in the Shippingport Pressurized Water Reactor core

  12. Thorium dioxide: properties and nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Belle, J.; Berman, R.M. (eds.)

    1984-01-01

    This is the sixth book on reactor materials published under sponsorship of the Naval Reactors Office of the United States Department of Energy, formerly the United States Atomic Energy Commission. This book presents a comprehensive compilation of the most significant properties of thorium dioxide, much like the book Uranium Dioxide: Properties and Nuclear Applications presented information on the fuel material used in the Shippingport Pressurized Water Reactor core.

  13. Uranium recovery from mine water

    International Nuclear Information System (INIS)

    Sarkar, K.M.

    1984-01-01

    In many plant trials it has been proven that very small amounts (10 to 20 ppm) of uranium dissolved in mine water can be effectively recovered by the use of ion exchange resins and this uranium recovery has many advantages. In this paper an economic analysis at different levels of uranium contamination and at different market prices of uranium are described. For this study an operating mine-mill complex with a sulphuric acid leach circuit, followed by solvent extraction (SX) process, is considered, where contaminated mine water is available in excess of process requirements. It is further assumed that the sulphuric acid eluant containing uranium would be mixed with the mill pregnant liquor stream that proceeds to the SX plant for final uranium recovery

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

    Science.gov (United States)

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

    2018-05-29

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

  15. Fabrication of mixed oxide fuel using plutonium from dismantled weapons

    International Nuclear Information System (INIS)

    Blair, H.T.; Chidester, K.; Ramsey, K.B.

    1996-01-01

    A very brief summary is presented of experimental studies performed to support the use of plutonium from dismantled weapons in fabricating mixed oxide (MOX) fuel for commercial power reactors. Thermal treatment tests were performed on plutonium dioxide powder to determine if an effective dry gallium removal process could be devised. Fabrication tests were performed to determine the effects of various processing parameters on pellet quality. Thermal tests results showed that the final gallium content is highly dependent on the treatment temperature. Fabrication tests showed that the milling process, sintering parameters, and uranium feed did effect pellet properties. 1 ref., 1 tab

  16. Long term carbon dioxide exchange above a mixed forest in the Belgian Ardennes: evaluation of different approaches to deduce total ecosystem respiration from Eddy covariance measurements

    Science.gov (United States)

    Jérôme, Elisabeth; Aubinet, Marc; Heinesch, Bernard

    2010-05-01

    The general aim of this research is to analyze inter annual variability of carbon dioxide (CO2) fluxes exchanged by a mixed forest located at the Vielsalm experimental site in Belgium. At this site, CO2 flux measurements started in 1996 and are still going on. Thirteen complete years of measurements are thus available. Net Ecosystem Exchange (NEE) inter annual variability may be driven by gross primary productivity (GPP) or Total Ecosystem Respiration (TER), which should thus be both quantified. Using flux partitioning methods, TER is deduced from NEE measurements. GPP is then obtained by subtracting TER from NEE. Initially, a robust estimation of TER is required. This work seeks to compare two independent approaches to assess TER in order to quantify the implications on inter-annual variability. The comparison was performed on twelve complete years. TER estimates can be deduced by extrapolating to the whole day NEE measurements taken during selected night or day periods. In both case, the extrapolation is performed by using a respiration response to temperature. The first approach, referred as the night-time approach, consisted in calculating TER using a temperature response function derived from night-time data sets (Reichstein et al., 2005). The second approach, referred as the daytime approach, consisted in assessing TER from the intercept of the NEE/Photosynthetically Photon Flux Density (PPFD) response (Wohlfahrt et al., 2005). For each approach, different modalities were compared: the use of long term (annual) or short term (15 days) data sets for the night-time approach and the use of different types of regression for the daytime approach. In addition, the impact of the temperature choice was studied for each of the approaches. For the night-time approach, main results showed that air temperature sensitivity of ecosystem respiration derived from annual data did not reflect the short-term air temperature sensitivity. Vielsalm is a summer active ecosystem

  17. The genesis of surficial uranium deposits

    International Nuclear Information System (INIS)

    Boyle, D.R.

    1984-01-01

    Surficial uranium deposits can form in such diverse environments as calcareous-dolomitic-gypsiferous fluvial and aeolian valley sediments in hot arid and semi-arid regions, oxidizing and reducing alkaline and saline playas, highly organic and/or clay-rich wetland areas, calcareous regoliths in arid terranes, laterites, lake sediments, and highly fractured zones in igneous and metamorphic basement complexes. Formation of ore is governed by the interrelationships between source of ore-forming elements, mechanisms of migration, environment of deposition, climate, preservation, tectonic history and structural framework. The principal factors controlling mobilization of ore-forming elements from source to site of deposition are the availability of elements in source rocks, presence of complexing agents, climate, nature of source rock regolith and structure of source rock terrane. The major processes governing precipitation of uranium in the surficial environment are reduction mechanisms, sorption processes, dissociation of uranyl complexes, change in redox states of ore-forming constituents, evaporation of surface and groundwaters, change in partial pressure of dissolved carbon dioxide, changes in pH, colloidal precipitation, and mixing of two or more surface and groundwaters. One or a number of these processes may be actively involved in ore formation. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

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

  19. In situ leaching process for recording uranium values

    International Nuclear Information System (INIS)

    McKnight, W.M.; Timmins, T.H.; Sherry, H.S.

    1977-01-01

    A method of recovering uranium values from a subterranean deposit comprising: injecting an alkaline carbonate lixiviant into said deposit; flowing said alkaline carbonate lixiviant through said deposit to dissolve said uranium values into said lixiviant; producing said lixiviant and said dissolved uranium values from said deposit; flowing said lixiviant and said dissolved uranium values through an adsorption material to adsorp said uranium values from said lixiviant; eluting said adsorption material with an eluant of ammonium carbonate to desorb said uranium values from said adsorption material into said eluate in a concentration greater than in said lixiviant; heating said eluate and said desorbed uranium values to vaporize off ammonia and carbon dioxide therefrom, thereby causing uranium values to crystallize from the eluate; and recovering said solid uranium values

  20. Actinide Oxidation State and O/M Ratio in Hypostoichiometric Uranium-Plutonium-Americium U0.750Pu0.246Am0.004O2-x Mixed Oxides.

    Science.gov (United States)

    Vauchy, Romain; Belin, Renaud C; Robisson, Anne-Charlotte; Lebreton, Florent; Aufore, Laurence; Scheinost, Andreas C; Martin, Philippe M

    2016-03-07

    Innovative americium-bearing uranium-plutonium mixed oxides U1-yPuyO2-x are envisioned as nuclear fuel for sodium-cooled fast neutron reactors (SFRs). The oxygen-to-metal (O/M) ratio, directly related to the oxidation state of cations, affects many of the fuel properties. Thus, a thorough knowledge of its variation with the sintering conditions is essential. The aim of this work is to follow the oxidation state of uranium, plutonium, and americium, and so the O/M ratio, in U0.750Pu0.246Am0.004O2-x samples sintered for 4 h at 2023 K in various Ar + 5% H2 + z vpm H2O (z = ∼ 15, ∼ 90, and ∼ 200) gas mixtures. The O/M ratios were determined by gravimetry, XAS, and XRD and evidenced a partial oxidation of the samples at room temperature. Finally, by comparing XANES and EXAFS results to that of a previous study, we demonstrate that the presence of uranium does not influence the interactions between americium and plutonium and that the differences in the O/M ratio between the investigated conditions is controlled by the reduction of plutonium. We also discuss the role of the homogeneity of cation distribution, as determined by EPMA, on the mechanisms involved in the reduction process.

  1. Electrochemical preparation of new uranium oxide phases

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  2. Uranium geochemistry of Orca Basin

    International Nuclear Information System (INIS)

    Weber, F.F. Jr.; Sackett, W.M.

    1981-01-01

    Orca Basin, an anoxic, brine-filled depression at a depth of 2200 m in the Northwestern Gulf of Mexico continental slope, has been studied with respect to its uranium geochemistry. Uranium concentration profiles for four cores from within the basin were determined by delayed-neutron counting. Uranium concentrations ranged from 2.1 to 4.1 ppm on a salt-free and carbonate-corrected basis. The highest uranium concentrations were associated with the lowest percentage and delta 13 C organic carbon values. For comparison, cores from the brine-filled Suakin and Atlantis II Deeps, both in the Red Sea, were also analyzed. Uranium concentrations ranged from 1.2 to 2.6 ppm in the Suakin Deep and from 8.0 to 11.0 ppm in the Atlantis II Deep. No significant correlation was found between uranium concentrations and organic carbon concentrations and delta 13 C values for these cores. Although anoxic conditions are necessary for significant uranium uptake by non-carbonate marine sediments, other factors such as dilution by rapidly depositing materials and uranium supply via mixing and diffusion across density gradients may be as important in determining uranium concentrations in hypersaline basin sediments. (author)

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

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

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

  4. [Values of mixed venous oxygen saturation and difference of mixed venous-arterial partial pressure of carbon dioxide in monitoring of oxygen metabolism and treatment after open-heart operation].

    Science.gov (United States)

    Pan, Chuanliang; Zhang, Haiying; Liu, Jianping

    2014-10-01

    To explore the clinic values of early goal directed treatment (EGDT) with the target of mixed venous oxygen saturation (SvO₂) and difference of mixed venous-arterial partial pressure of carbon dioxide (Pv-aCO₂) in monitoring of oxygen metabolism and treatment for patients post open-heart operation. A prospective study was conducted. The adult patients admitted to Third People's Hospital of Chengdu from December 2011 to March 2014 with SvO₂2 mmol/L when admitted in intensive care unit (ICU) were selected on whom elective open-heart operation and pulmonary artery catheter examination were done. All patients received EGDT with the target of SvO₂≥0.65 and Pv-aCO₂<6 mmHg (1 mmHg=0.133 kPa) and were divided into three groups by the values of SvO₂and Pv-aCO₂at 6-hour after ICU admission: A group with SvO₂≥0.65 and Pv-aCO₂<6 mmHg, B group with SvO₂≥0.65 and Pv-aCO₂≥6 mmHg, and C group with SvO₂<0.65. Then the changes and prognosis of the patients in different groups were observed. 103 cases were included, 44 in A group, 31 in B group and 28 in C group. The acute physiology and chronic health evaluation II (APACHEII) score in group A were significantly lower than that in group B or C at 6, 24, 48 and 72 hours (T6, T24, T48, T72) of ICU admission (T6: 11.4 ± 5.8 vs. 13.9 ± 5.4, 13.7 ± 6.4; T24: 8.8 ± 3.7 vs. 10.8 ± 4.8, 11.8 ± 5.4; T48: 8.7 ± 4.1 vs. 9.6 ± 4.2, 10.2 ± 5.1; T72: 7.5 ± 3.4 vs. 8.6 ± 2.9, 9.2 ± 4.2, all P<0.05), and the sequential organ failure assessment (SOFA) showed the same tendency (T6: 6.5 ± 4.3 vs. 8.0 ± 3.8, 9.1 ± 4.5; T24: 6.6 ±3.6 vs. 8.6 ± 3.9, 8.5 ± 3.3; T48: 5.2 ± 3.4 vs. 7.0 ± 3.6, 7.6 ± 5.1; T72: 4.6 ± 2.4 vs. 5.8 ± 2.5, 6.8 ± 3.5, all P<0.05). The values of blood lactic acid (mmol/L) in group A and B were significant lower than that in group C at T6, T24, T48 and T72 (T6: 1.60 ± 0.95, 2.20 ± 1.02 vs. 2.55 ± 1.39; T24: 2.26 ± 1.26, 2.70 ± 1.36 vs. 3.34 ± 2.36; T48: 2.01 ± 1.15, 2.17

  5. Analysis of burnup of Angra 2 PWR nuclear with addition of thorium dioxide fuel using ORIGEN-ARP

    Energy Technology Data Exchange (ETDEWEB)

    Goncalves, Isadora C.; Wichrowski, Caio C.; Oliveira, Claudio L. de; Vellozo, Sergio O.; Baptista, Camila O., E-mail: isadora.goncalves@ime.eb.br, E-mail: wichrowski@ime.eb.br, E-mail: d7luiz@yahoo.com.br, E-mail: vellozo@ime.eb.br, E-mail: camila.oliv.baptista@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Secao de Engenharia Nuclear

    2017-11-01

    It is known that isotope {sup 232}thorium is a fertile nuclide with the ability to convert into {sup 233}uranium, a potentially fissile isotope, after absorbing a neutron. As there is a large stock of available thorium in the world, this element shows great promise in mitigate the world energy crisis, more particularly in the problem of uranium scarcity, besides being an alternative nuclear fuel for those currently used in reactors, and yet presenting advantages as an option for the non-proliferation movement, among others. In this study, the analysis of the remaining nuclides of burnup was carried out for the core configuration of a PWR (pressurized water reactor) reactor, specifically the Angra 2 reactor, using only uranium dioxide, its current configuration, and in different configurations including a mixed oxide of uranium and thorium in three concentrations, allowing a preliminary assessment of the feasibility of the modification of the fuel, the resulting production of {sup 233}uranium, the emergence of {sup 231}protactinium (an isotope that only occurs as a fission product of {sup 232}Th) resulting from burning. The study was carried out using data obtained from FSAR (Final Safety Analysis Report) of Angra 2, using the SCALE 6.1, a modeling and simulation nuclear code, especially its ORIGEN-ARP module, which analyzes the depletion of isotopes presents in a reactor. (author)

  6. Analysis of burnup of Angra 2 PWR nuclear with addition of thorium dioxide fuel using ORIGEN-ARP

    International Nuclear Information System (INIS)

    Goncalves, Isadora C.; Wichrowski, Caio C.; Oliveira, Claudio L. de; Vellozo, Sergio O.; Baptista, Camila O.

    2017-01-01

    It is known that isotope "2"3"2thorium is a fertile nuclide with the ability to convert into "2"3"3uranium, a potentially fissile isotope, after absorbing a neutron. As there is a large stock of available thorium in the world, this element shows great promise in mitigate the world energy crisis, more particularly in the problem of uranium scarcity, besides being an alternative nuclear fuel for those currently used in reactors, and yet presenting advantages as an option for the non-proliferation movement, among others. In this study, the analysis of the remaining nuclides of burnup was carried out for the core configuration of a PWR (pressurized water reactor) reactor, specifically the Angra 2 reactor, using only uranium dioxide, its current configuration, and in different configurations including a mixed oxide of uranium and thorium in three concentrations, allowing a preliminary assessment of the feasibility of the modification of the fuel, the resulting production of "2"3"3uranium, the emergence of "2"3"1protactinium (an isotope that only occurs as a fission product of "2"3"2Th) resulting from burning. The study was carried out using data obtained from FSAR (Final Safety Analysis Report) of Angra 2, using the SCALE 6.1, a modeling and simulation nuclear code, especially its ORIGEN-ARP module, which analyzes the depletion of isotopes presents in a reactor. (author)

  7. Czechoslovak uranium

    International Nuclear Information System (INIS)

    Pluskal, O.

    1992-01-01

    Data and knowledge related to the prospecting, mining, processing and export of uranium ores in Czechoslovakia are presented. In the years between 1945 and January 1, 1991, 98,461.1 t of uranium were extracted. In the period 1965-1990 the uranium industry was subsidized from the state budget to a total of 38.5 billion CSK. The subsidies were put into extraction, investments and geologic prospecting; the latter was at first, ie. till 1960 financed by the former USSR, later on the two parties shared costs on a 1:1 basis. Since 1981 the prospecting has been entirely financed from the Czechoslovak state budget. On Czechoslovak territory uranium has been extracted from deposits which may be classified as vein-type deposits, deposits in uranium-bearing sandstones and deposits connected with weathering processes. The future of mining, however, is almost exclusively being connected with deposits in uranium-bearing sandstones. A brief description and characteristic is given of all uranium deposits on Czechoslovak territory, and the organization of uranium mining in Czechoslovakia is described as is the approach used in the world to evaluate uranium deposits; uranium prices and actual resources are also given. (Z.S.) 3 figs

  8. The behaviour of uranium metal in hydrogen atmospheres

    International Nuclear Information System (INIS)

    Allen, G.C.; Stevens, J.C.H.

    1988-01-01

    The reaction between commercial H 2 and uranium metal leads to the formation of UO 2 due to traces of water vapour or oxygen. When extremely pure H 2 is used uranium hydride may be formed but, even with 99.9999% H 2 , uranium dioxide forms preferentially. The present work identifies the presence of UH 3 in the X-ray photoelectron spectrum of a uranium sample which has been exposed to ca. 10 10 L† H 2 at ca. 200 0 C. This spectrum indicates that the hydride possesses a high degree of covalency, since the oxidation state of uranium in UH 3 appears to be ca. 1.4. (author)

  9. Fission products stability in uranium dioxide

    International Nuclear Information System (INIS)

    Brillant, G.; Gupta, F.; Pasturel, A.

    2011-01-01

    Fission product stability in nuclear fuels is investigated using density functional theory (DFT). In particular, incorporation and solution energies of He, Kr, Xe, I, Te, Ru, Sr and Ce in pre-existing trap sites of UO 2 (vacancies, interstitials, U-O divacancy, and Schottky trio defects) are calculated using the projector-augmented-wave method as implemented in the Vienna ab initio simulation package. Correlation effects are taken into account within the DFT+U approach. The stability of many binary and ternary compounds in comparison to soluted atoms is also explored. Finally the involvement of FP in the formation of metallic and oxide precipitates in oxide fuels is discussed in the light of experimental results.

  10. Methods for the accountability of uranium dioxide

    International Nuclear Information System (INIS)

    Stephens, F.B.; Gutmacher, R.G.; Ernst, K.; Harrar, J.E.; Turel, S.P.

    1975-06-01

    Procedures for the determination of the total U and the amount of 235 U isotope in UO 2 powder and pellets are given. Methods for U determination include coulometry, titration, and gravimetry. Surface-ionization mass spectroscopy is described for 235 U measurement. Spectrometric procedures are described for determining the impurity content in the UO 2 . (U.S.)

  11. Kinetics of sintering of uranium dioxide

    International Nuclear Information System (INIS)

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

    1978-01-01

    The kinetics of sintering of UO 2 powders derived from ADU route and calcined at different temperatures was studied. The activation energy for sintering was found to depend on the calcination temperature, the density chosen and the sintering temperature range. The motive force for sintering is the excess free energy in the particle system. This exists in the powder compact in the form of surface energy and the excess lattice energy due to defects. The defects which can be eliminated at the operating temperature are responsible for the mobility and hence sintering. This concept of the motive force for sintering has been used to explain the difference in the activation energies observed in the present study. This would also explain phenomena such as attainment of limiting density, presence of optimum sintering temperature and the influence of calcination treatments on the sintering behaviour of powders. (author)

  12. Hydrogen retention and release from uranium dioxide

    International Nuclear Information System (INIS)

    Sherman, D.F.

    1987-08-01

    The ceramic samples (UO 2 ) are exposed to high pressure hydrogen gas at a fixed temperature for a time sufficient to achieve equilibrium. After rapid quenching, the hydrogen-saturated sample is transferred to a vacuum-outgassing furnace. The sample is outgassed in a linear temperature ramp and the released hydrogen is detected by an in-situ mass spectrometer. This technique measures the rate of release of hydrogen with a sensitivity level of about 2 ng of hydrogen (as D 2 ) per hour. In this study, experiments were conducted on both polycrystalline and single-crystal UO 2 . Experimental variables included temperature (1000 to 1600 0 C) and infusion pressure (5 to 32 atm D 2 ), and for the polycrystalline specimen, stoichiometry. Dissolution of H 2 in both single-crystal and polycrystalline UO 2 was found to obey Seivert's law. The Sievert's law constant of deuterium in single-crystal UO 2 was determined to be: 3.0 x 10 7 exp(-235 kJ/RT) ppM atomic/√atm and for polycrystalline UO 2 : 5.5 x 10 4 exp(-100 kJ/RT) ppM atomic/√atm. The solubility of hydrogen in hypostoichiometric urania was found to be up to three orders of magnitude greater than in stoichiometric UO 2 depending on the O/U ratios, implying the anion vacancy is the primary solution site in the UO 2 lattice. The release-rate curves for the single crystal and polycrystalline UO 2 specimens exhibited multiple peaks, with most of the deuterium released between 600 and 1200 0 C for the polycrystalline samples, and between 700 and 1800 0 C in the single-crystal specimens. This release of hydrogen from UO 2 could not be adequately modeled as diffusion or diffusion with trapping and resolution. It was determined that release was governed by release from traps in both the polycrystalline and single crystal UO 2 specimens. 40 refs., 72 figs., 6 tabs

  13. Thermal diffusion of chlorine in uranium dioxide

    International Nuclear Information System (INIS)

    Pipon, Y.; Toulhoat, N.; Moncoffre, N.; Jaffrezic, H.; Gavarini, S.; Martin, P.; Raimbault, L.; Scheidegger, A.M.

    2006-01-01

    In a nuclear reactor, isotopes such as 35 Cl present as impurities in the nuclear fuel are activated by thermal neutron capture. During interim storage or geological disposal of nuclear fuel, the activation products such as 36 Cl may be released from the fuel to the geo/biosphere and contribute to the ''instant release fraction'' as they are likely to migrate in defects and grain boundaries. In order to differentiate diffusion mechanisms due to ''athermal'' processes during irradiation from thermally activated diffusion, both irradiation and thermal effects must be assessed. This work concerns the measurement of the thermal diffusion coefficient of chlorine in UO 2 . 37 Cl was implanted at a 10 13 at/cm 2 fluence in depleted UO 2 samples which were then annealed in the 900-1200 C temperature range and finally analyzed by secondary ion mass spectrometry (SIMS) to obtain 37 Cl depth profiles. The migration process appears to be rather complex, involving mechanisms such as atomic, grain boundary, directed diffusion along preferential patterns as well as trapping into sinks before successive effusion. However, using a diffusion model based on general equation of transport, apparent diffusion coefficients could be calculated for 1000 and 1100 C and a mean activation energy of 4.3 eV is proposed. This value is one of the lowest values compared to those found in literature for other radionuclides pointing out a great ability of chlorine to migrate in UO 2 at relatively low temperatures. In order to unequivocally determine the diffusion behaviour of both implanted and pristine chlorine before and after thermal annealing, the structural environment of chlorine in UO 2 was examined using micro X-ray fluorescence (micro-XRF) and micro X-ray absorption spectroscopy (micro-XAS). (orig.)

  14. Uranium ores

    International Nuclear Information System (INIS)

    Poty, B.; Roux, J.

    1998-01-01

    The processing of uranium ores for uranium extraction and concentration is not much different than the processing of other metallic ores. However, thanks to its radioactive property, the prospecting of uranium ores can be performed using geophysical methods. Surface and sub-surface detection methods are a combination of radioactive measurement methods (radium, radon etc..) and classical mining and petroleum prospecting methods. Worldwide uranium prospecting has been more or less active during the last 50 years, but the rise of raw material and energy prices between 1970 and 1980 has incited several countries to develop their nuclear industry in order to diversify their resources and improve their energy independence. The result is a considerable increase of nuclear fuels demand between 1980 and 1990. This paper describes successively: the uranium prospecting methods (direct, indirect and methodology), the uranium deposits (economical definition, uranium ores, and deposits), the exploitation of uranium ores (use of radioactivity, radioprotection, effluents), the worldwide uranium resources (definition of the different categories and present day state of worldwide resources). (J.S.)

  15. Uranium market

    International Nuclear Information System (INIS)

    Rubini, L.A.; Asem, M.A.D.

    1990-01-01

    The historical development of the uranium market is present in two periods: The initial period 1947-1970 and from 1970 onwards, with the establishment of a commercial market. The world uranium requirements are derived from the corresponding forecast of nuclear generating capacity, with, particular emphasis to the brazilian requirements. The forecast of uranium production until the year 2000 is presented considering existing inventories and the already committed demand. The balance between production and requirements is analysed. Finally the types of contracts currently being used and the development of uranium prices in the world market are considered. (author)

  16. Uranium enrichment

    International Nuclear Information System (INIS)

    1990-01-01

    This report looks at the following issues: How much Soviet uranium ore and enriched uranium are imported into the United States and what is the extent to which utilities flag swap to disguise these purchases? What are the U.S.S.R.'s enriched uranium trading practices? To what extent are utilities required to return used fuel to the Soviet Union as part of the enriched uranium sales agreement? Why have U.S. utilities ended their contracts to buy enrichment services from DOE?

  17. Treatment of uranium turning with the controllable oxidizing process

    International Nuclear Information System (INIS)

    Shen Bingyi; Zhang Yonggang; Zhen Huikuan

    1989-02-01

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

  18. Microbial conversion of sulfur dioxide in flue gas to sulfide using bulk drug industry wastewater as an organic source by mixed cultures of sulfate reducing bacteria

    International Nuclear Information System (INIS)

    Rao, A. Gangagni; Ravichandra, P.; Joseph, Johny; Jetty, Annapurna; Sarma, P.N.

    2007-01-01

    Mixed cultures of sulfate reducing bacteria (SRB) were isolated from anaerobic cultures and enriched with SRB media. Studies on batch and continuous reactors for the removal of SO 2 with bulk drug industry wastewater as an organic source using isolated mixed cultures of SRB revealed that isolation and enrichment methodology adopted in the present study were apt to suppress the undesirable growth of anaerobic bacteria other than SRB. Studies on anaerobic reactors showed that process was sustainable at COD/S ratio of 2.2 and above with optimum sulfur loading rate (SLR) of 5.46 kg S/(m 3 day), organic loading rate (OLR) of 12.63 kg COD/(m 3 day) and at hydraulic residence time (HRT) of 8 h. Free sulfide (FS) concentration in the range of 300-390 mg FS/l was found to be inhibitory to mixed cultures of SRB used in the present studies

  19. Sorption behaviour of uranium and thorium on hydrons tin oxide from aqueous and mixed-solvent H2SO4 media

    International Nuclear Information System (INIS)

    Misak, N.Z.; Salema, H.N.; El-Naggar, J.M.

    1983-01-01

    At pH values > about 2 in 10 -3 -10 -2 M aqueous sulphate solutions, uranium seems to be sorbed by hydrous tin oxide mainly as cations, while thorium is sorced as cations and as the neutral complex. At pH values of about 1.1-1.4, both uranium and thorium seem to be mainly sorbed as the neutral complexes. while at lower pH values, sorption of anionic commplexes comes into play. The sorption of uranium generally increased progressively on addition of increasing amounts of methanol, ethanol or acetone. The sorption of thorium decreases a little at 0.01 N H 2 SO 4 and increases a little at 0.5 N H 2 SO 4 on adding the organic solvents. At 0.1 N H 2 SO 4 , the addition of 20percent of the organic solvents brings the sorption of thorium to almost negligible values, which seems to offer an attractive means for U/Th separation. (author)

  20. NURE uranium deposit model studies

    International Nuclear Information System (INIS)

    Crew, M.E.

    1981-01-01

    The National Uranium Resource Evaluation (NURE) Program has sponsored uranium deposit model studies by Bendix Field Engineering Corporation (Bendix), the US Geological Survey (USGS), and numerous subcontractors. This paper deals only with models from the following six reports prepared by Samuel S. Adams and Associates: GJBX-1(81) - Geology and Recognition Criteria for Roll-Type Uranium Deposits in Continental Sandstones; GJBX-2(81) - Geology and Recognition Criteria for Uraniferous Humate Deposits, Grants Uranium Region, New Mexico; GJBX-3(81) - Geology and Recognition Criteria for Uranium Deposits of the Quartz-Pebble Conglomerate Type; GJBX-4(81) - Geology and Recognition Criteria for Sandstone Uranium Deposits in Mixed Fluvial-Shallow Marine Sedimentary Sequences, South Texas; GJBX-5(81) - Geology and Recognition Criteria for Veinlike Uranium Deposits of the Lower to Middle Proterozoic Unconformity and Strata-Related Types; GJBX-6(81) - Geology and Recognition Criteria for Sandstone Uranium Deposits of the Salt Wash Type, Colorado Plateau Province. A unique feature of these models is the development of recognition criteria in a systematic fashion, with a method for quantifying the various items. The recognition-criteria networks are used in this paper to illustrate the various types of deposits

  1. integrated vertical photobioreactor system for carbon dioxide

    African Journals Online (AJOL)

    Astri Nugroho

    2013-07-02

    Jul 2, 2013 ... efficient system for converting carbon dioxide (CO2) into biomass. The use of ... often been thought to achieve the most efficient mixing and the best ... such process a photobioreactor is designed. Photobioreactor is a device ...

  2. Antipollution system to remove nitrogen dioxide gas

    Science.gov (United States)

    Metzler, A. J.; Slough, J. W.

    1971-01-01

    Gas phase reaction system using anhydrous ammonia removes nitrogen dioxide. System consists of ammonia injection and mixing section, reaction section /reactor/, and scrubber section. All sections are contained in system ducting.

  3. Gravimetric determination of uranium in SALE samples

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    As a participant in the Safeguards Analytical Laboratory Evaluation (SALE) program, the Analytical Chemistry Laboratory at General Atomic routinely assays uranium dioxide and uranyl nitrate SALE samples for uranium content. Gravimetric methods are relatively easy and inexpensive to apply when the samples for uranium content. Gravimetric methods are relatively easy and inexpensive to apply when the samples are free from substantial amounts of metallic impurities. Clearly the gravimetric procedure alone is not specific for uranium and must be enhanced by the use of impurity corrections. Emission spectrography is used routinely as the technique of choice for making such corrections. In cases where it is essential to assay specifically for uranium, the modified Davies-Gray titration using a weighed titrant method is applied. In this paper some essential features of these gravimetric and titrimetric procedures are discussed

  4. Feasibility study of the dissolution rates of uranium ore dust, uranium concentrates and uranium compounds in simulated lung fluid

    International Nuclear Information System (INIS)

    Robertson, R.

    1986-01-01

    A flow-through apparatus has been devised to study the dissolution in simulated lung fluid of aerosol materials associated with the Canadian uranium industry. The apparatus has been experimentally applied over 16 day extraction periods to approximately 2g samples of < 38um and 53-75um particle-size fractions of both Elliot Lake and Mid-Western uranium ores. The extraction of uranium-238 was in the range 24-60% for these samples. The corresponding range for radium-226 was 8-26%. Thorium-230, lead-210, polonium-210, and thorium-232 were not significantly extracted. It was incidentally found that the elemental composition of the ores studied varies significantly with particle size, the radionuclide-containing minerals and several extractable stable elements being concentrated in the smaller size fraction. Samples of the refined compounds uranium dioxide and uranium trioxide were submitted to similar 16 day extraction experiments. Approximately 0.5% of the uranium was extracted from a 0.258g sample of unsintered (fluid bed) uranium dioxide of particle size < 38um. The corresponding figure for a 0.292g sample of uranium trioxide was 97%. Two aerosol samples on filters were also studied. Of the 88ug uranium initially measured on stage 2 of a cascade impactor sample collected from the yellow cake packing area of an Elliot Lake mill, essentially 100% was extracted over a 16 day period. The corresponding figure for an open face filter sample collected in a fuel fabrication plant and initially measured at 288ug uranium was approximately 3%. Recommendations are made with regard to further work of a research nature which would be useful in this area. Recommendations are also made on sampling methods, analytical methods and extraction conditions for various aerosols of interest which are to be studied in a work of broader scope designed to yield meaningful data in connection with lung dosimetry calculations

  5. Uranium mining

    International Nuclear Information System (INIS)

    Lange, G.

    1975-01-01

    The winning of uranium ore is the first stage of the fuel cycle. The whole complex of questions to be considered when evaluating the profitability of an ore mine is shortly outlined, and the possible mining techniques are described. Some data on uranium mining in the western world are also given. (RB) [de

  6. Separation of uranium from (Th,U)O2 solid solutions

    International Nuclear Information System (INIS)

    Chiotti, P.; Jha, M.C.

    1976-01-01

    Uranium is separated from mixed oxides of thorium and uranium by a pyrometallurgical process in which the oxides are mixed with a molten chloride salt containing thorium tetrachloride and thorium metal which reduces the uranium oxide to uranium metal which can then be recovered from the molten salt. The process is particularly useful for the recovery of uranium from generally insoluble high-density sol-gel thoria-urania nuclear reactor fuel pellets. 7 claims

  7. Uranium enrichment

    International Nuclear Information System (INIS)

    1989-01-01

    GAO was asked to address several questions concerning a number of proposed uranium enrichment bills introduced during the 100th Congress. The bill would have restructured the Department of Energy's uranium enrichment program as a government corporation to allow it to compete more effectively in the domestic and international markets. Some of GAO's findings discussed are: uranium market experts believe and existing market models show that the proposed DOE purchase of a $750 million of uranium from domestic producers may not significantly increase production because of large producer-held inventories; excess uranium enrichment production capacity exists throughout the world; therefore, foreign producers are expected to compete heavily in the United States throughout the 1990s as utilities' contracts with DOE expire; and according to a 1988 agreement between DOE's Offices of Nuclear Energy and Defense Programs, enrichment decommissioning costs, estimated to total $3.6 billion for planning purposes, will be shared by the commercial enrichment program and the government

  8. Uranium resources

    International Nuclear Information System (INIS)

    1976-01-01

    This is a press release issued by the OECD on 9th March 1976. It is stated that the steep increases in demand for uranium foreseen in and beyond the 1980's, with doubling times of the order of six to seven years, will inevitably create formidable problems for the industry. Further substantial efforts will be needed in prospecting for new uranium reserves. Information is given in tabular or graphical form on the following: reasonably assured resources, country by country; uranium production capacities, country by country; world nuclear power growth; world annual uranium requirements; world annual separative requirements; world annual light water reactor fuel reprocessing requirements; distribution of reactor types (LWR, SGHWR, AGR, HWR, HJR, GG, FBR); and world fuel cycle capital requirements. The information is based on the latest report on Uranium Resources Production and Demand, jointly issued by the OECD's Nuclear Energy Agency (NEA) and the International Atomic Energy Agency. (U.K.)

  9. Uranium supply and demand

    Energy Technology Data Exchange (ETDEWEB)

    Spriggs, M J

    1976-01-01

    Papers were presented on the pattern of uranium production in South Africa; Australian uranium--will it ever become available; North American uranium resources, policies, prospects, and pricing; economic and political environment of the uranium mining industry; alternative sources of uranium supply; whither North American demand for uranium; and uranium demand and security of supply--a consumer's point of view. (LK)

  10. Trends in the specific carbon dioxide emissions of the german fuel mix; Entwicklung der spezifischen Kohlendioxid-Emissionen des deutschen Strommix

    Energy Technology Data Exchange (ETDEWEB)

    Machat, Marcus; Werner, Kathrin

    2007-04-15

    The specific emission factor of the German fuel mix is calculated from the direct CO2 emissions of electric power generation and from the net electric power available for end use consumption. The emission factor was reduced between 1990 and 2005, owing to higher efficiencies of modernized power plants and to a bigger share of renewable energy sources. The brochure, published by the Federal Environmental Office, presents trends and diagrams.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-04-01

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

  12. Uranium, depleted uranium, biological effects; Uranium, uranium appauvri, effets biologiques

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    Physicists, chemists and biologists at the CEA are developing scientific programs on the properties and uses of ionizing radiation. Since the CEA was created in 1945, a great deal of research has been carried out on the properties of natural, enriched and depleted uranium in cooperation with university laboratories and CNRS. There is a great deal of available data about uranium; thousands of analyses have been published in international reviews over more than 40 years. This presentation on uranium is a very brief summary of all these studies. (author)

  13. Uranium toxicology

    International Nuclear Information System (INIS)

    Ferreyra, Mariana D.; Suarez Mendez, Sebastian

    1997-01-01

    In this paper are presented the methods and procedures optimized by the Nuclear Regulatory Authority (ARN) for the determination of: natural uranium mass, activity of enriched uranium in samples of: urine, mucus, filters, filter heads, rinsing waters and Pu in urine, adopted and in some cases adapted, by the Environmental Monitoring and Internal Dosimetry Laboratory. The analyzed material corresponded to biological and environmental samples belonging to the staff professionally exposed that work in plants of the nuclear fuel cycle. For a better comprehension of the activities of this laboratory, it is included a brief description of the uranium radiochemical toxicity and the limits internationally fixed to preserve the workers health

  14. Computation Results from a Parametric Study to Determine Bounding Critical Systems of Homogeneously Water-Moderated Mixed Plutonium--Uranium Oxides

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Y.

    2001-01-11

    This report provides computational results of an extensive study to examine the following: (1) infinite media neutron-multiplication factors; (2) material bucklings; (3) bounding infinite media critical concentrations; (4) bounding finite critical dimensions of water-reflected and homogeneously water-moderated one-dimensional systems (i.e., spheres, cylinders of infinite length, and slabs that are infinite in two dimensions) that were comprised of various proportions and densities of plutonium oxides and uranium oxides, each having various isotopic compositions; and (5) sensitivity coefficients of delta k-eff with respect to critical geometry delta dimensions were determined for each of the three geometries that were studied. The study was undertaken to support the development of a standard that is sponsored by the International Standards Organization (ISO) under Technical Committee 85, Nuclear Energy (TC 85)--Subcommittee 5, Nuclear Fuel Technology (SC 5)--Working Group 8, Standardization of Calculations, Procedures and Practices Related to Criticality Safety (WG 8). The designation and title of the ISO TC 85/SC 5/WG 8 standard working draft is WD 14941, ''Nuclear energy--Fissile materials--Nuclear criticality control and safety of plutonium-uranium oxide fuel mixtures outside of reactors.'' Various ISO member participants performed similar computational studies using their indigenous computational codes to provide comparative results for analysis in the development of the standard.

  15. Influent of Carbonization of Sol Solution at the External Gelation Process on the Quality of Uranium Oxide Kernel

    International Nuclear Information System (INIS)

    Damunir; Sukarsono

    2007-01-01

    The influent of carbonization of sol solution at the external gelation process on the quality of uranium oxide kernel was done. Variables observed are the influent of carbon, temperature and time of reduction process of U 3 O 8 kernel resulted from carbonization of sol solution. First of all, uranyl nitrate was reacted with 1 M NH 4 OH solution, producing the colloid of UO 3 . Then by mixing and heating up to the temperature of 60-80 °C, the colloid solution was reacted with PVA, mono sorbitol oleate and paraffin producing of uranium-PVA sol. Then sol solution was carbonized with carbon black of mol ratio of carbon to uranium =2.32-6.62, produce of carbide gel. Gel then washed, dried and calcined at 800 °C for 4 hours to produce of U 3 O 8 kernel containing carbon. Then the kernel was reduced by H 2 gas in the medium of N 2 gas at 500-800 °C, 50 mmHg pressure for 3 hours. The process was repeated at 700 °C, 50 mmHg pressure for 1-4 hours. The characterization of chemical properties of the gel grains and uranium oxide kernel using FTIR covering the analysis of absorption band of infra red spectrum of UO 3 , C-OH, NH 3 , C-C, C-H and OH functional group. The physical properties of uranium oxide covering specific surface area, void volume, mean diameter using surface area meter Nova-1000 and as N 2 gas an absorbent. And O/U ratio of uranium dioxide kernel by gravimetry method. The result of experiment showed that carbonization of sol solution at the external gelation process give influencing the quality of uranium oxide kernel. (author)

  16. Uranium tailings reference materials

    International Nuclear Information System (INIS)

    Smith, C.W.; Steger, H.F.; Bowman, W.S.

    1984-01-01

    Samples of uranium tailings from Bancroft and Elliot Lake, Ontario, and from Beaverlodge and Rabbit Lake, Saskatchewan, have been prepared as compositional reference materials at the request of the National Uranium Tailings Research Program. The four samples, UTS-1 to UTS-4, were ground to minus 104 μm, each mixed in one lot and bottled in 200-g units for UTS-1 to UTS-3 and in 100-g units for UTS-4. The materials were tested for homogeneity with respect to uranium by neutron activation analysis and to iron by an acid-decomposition atomic absorption procedure. In a free choice analytical program, 18 laboratories contributed results for one or more of total iron, titanium, aluminum, calcium, barium, uranium, thorium, total sulphur, and sulphate for all four samples, and for nickel and arsenic in UTS-4 only. Based on a statistical analysis of the data, recommended values were assigned to all elements/constituents, except for sulphate in UTS-3 and nickel in UTS-4. The radioactivity of thorium-230, radium-226, lead-210, and polonium-210 in UTS-1 to UTS-4 and of thorium-232, radium-228, and thorium-228 in UTS-1 and UTS-2 was determined in a radioanalytical program composed of eight laboratories. Recommended values for the radioactivities and associated parameters were calculated by a statistical treatment of the results

  17. Rossing uranium

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    In this article the geology of the deposits of the Rossing uranium mine in Namibia is discussed. The planning of the open-pit mining, the blasting, drilling, handling and the equipment used for these processes are described

  18. Uranium, depleted uranium, biological effects

    International Nuclear Information System (INIS)

    2001-01-01

    Physicists, chemists and biologists at the CEA are developing scientific programs on the properties and uses of ionizing radiation. Since the CEA was created in 1945, a great deal of research has been carried out on the properties of natural, enriched and depleted uranium in cooperation with university laboratories and CNRS. There is a great deal of available data about uranium; thousands of analyses have been published in international reviews over more than 40 years. This presentation on uranium is a very brief summary of all these studies. (author)

  19. Conversion of highly enriched uranium in thorium-232 based oxide fuel for light water reactors: MOX-T fuel

    Energy Technology Data Exchange (ETDEWEB)

    Vapirev, E; Jordanov, T; Khristoskov, I [Sofia Univ. (Bulgaria). Fizicheski Fakultet

    1996-12-31

    The possibility of using highly enriched uranium available from military inventories for production of mixed oxide fuel (MOX) has been proposed. The fuel is based on U-235 dioxide as fissile isotope and Th-232 dioxide as a non-fissile isotope. It is shown that although the fuel conversion coefficient to U-233 is expected to be less than 1, the proposed fuel has several important advantages resulting in cost reduction of the nuclear fuel cycle. The expected properties of MOX fuel (cross-sections, generated chains, delayed neutrons) are estimated. Due to fuel generation the initial enrichment is expected to be 1% less for production of the same energy. In contrast to traditional fuel no long living actinides are generated which reduces the disposal and reprocessing cost. 7 refs.

  20. Effects of uranium compounds on skin

    International Nuclear Information System (INIS)

    Rey, B.M. de

    1982-12-01

    The following uranium compounds were topically applied to the dorsal skin of 35 day-old Wistar rats (60 g, male): uranium dioxide, uranyl nitrate, uranyl acetate, ammonium uranyl tricarbonate and ammonium diuranate. Percutaneous absorption was mediated with the aid of a vehicle and known quantities of various particle-sized batches of uranium compounds were directly implanted in the subcutaneous tissue. Animals were sacrificed 3, 6, 24 and 48 hours after implantation. Subcutaneous tissue and muscle underneath the implantation site were anlaysed by light and electron microscopy. A Cameca 322 X-ray microanalyzer was used to analyze uranium traces in calcified tissue (bones and teeth) and kidneys. A steady loss in body weight was observed in animals given high concentration of uranyl nitrate and ammonium uranyl tricarbonate. All animals died five days after the onset of the experiment due to renal failure. Slightly soluble compounds, ammonium diuranate and uranyl acetate, caused only a slight decrease in body weight. Uranium dioxide, the most insoluble compound used, induced only a transitory slight body weight decrease. Histopathological study revealed damages to the tissues of topicated skin, hair follicles and adnexal glands. High concentration of uranium was indicated in bone, teeth and kidneys by X-ray scanning

  1. Uranium loans

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    When NUEXCO was organized in 1968, its founders conceived of a business based on uranium loans. The concept was relatively straightforward; those who found themselves with excess supplies of uranium would deposit those excesses in NUEXCO's open-quotes bank,close quotes and those who found themselves temporarily short of uranium could borrow from the bank. The borrower would pay interest based on the quantity of uranium borrowed and the duration of the loan, and the bank would collect the interest, deduct its service fee for arranging the loan, and pay the balance to those whose deposits were borrowed. In fact, the original plan was to call the firm Nuclear Bank Corporation, until it was discovered that using the word open-quotes Bankclose quotes in the name would subject the firm to various US banking regulations. Thus, Nuclear Bank Corporation became Nuclear Exchange Corporation, which was later shortened to NUEXCO. Neither the nuclear fuel market nor NUEXCO's business developed quite as its founders had anticipated. From almost the very beginning, the brokerage of uranium purchases and sales became a more significant activity for NUEXCO than arranging uranium loans. Nevertheless, loan transactions have played an important role in the international nuclear fuel market, requiring the development of special knowledge and commercial techniques

  2. Chemical treatment of ammonium fluoride solution in uranium reconversion plant

    International Nuclear Information System (INIS)

    Carvalho Frajndlich, E.U. de.

    1992-01-01

    A chemical procedure is described for the treatment of the filtrate, produced from the transformation of uranium hexafluoride (U F 6 ) into ammonium uranyl carbonate (AUC). This filtrate is an intermediate product in the U F 6 to uranium dioxide (U O 2 ) reconversion process. The described procedure recovers uranium as ammonium peroxide fluoro uranate (APOFU) by precipitation with hydrogen peroxide (H 2 O 2 ), and as later step, its calcium fluoride (CaF 2 ) co-precipitation. The recovered uranium is recycled to the AUC production plant. (author)

  3. Production and analysis of ultradispersed uranium oxide powders

    Science.gov (United States)

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

    2010-05-01

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

  4. Method to manufacture a nuclear fuel from uranium-plutonium monocarbide or uranium-plutonium mononitride

    International Nuclear Information System (INIS)

    Krauth, A.; Mueller, N.

    1977-01-01

    Pure uranium carbide or nitride is converted with plutonium oxide and carbon (all in powder form) to uranium-plutonium monocarbide or mononitride by cold pressing and sintering at about 1600 0 C. Pure uranium carbide or uranium nitride powder is firstly prepared without extensive safety measures. The pure uranium carbide or nitride powder can also be inactivated by using chemical substances (e.g. stearic acid) and be handled in air. The sinterable uranium carbide or nitride powder (or also granulate) is then introduced into the plutonium line and mixed with a nonstoichiometrically adjusted, prereacted mixture of plutonium oxide and carbon, pressed to pellets and reaction sintered. The surface of the uranium-plutonium carbide (higher metal content) can be nitrated towards the end of the sinter process in a stream of nitrogen. The protective layer stabilizes the carbide against the water and oxygen content in air. (IHOE) [de

  5. Synthesis of uranium metal using laser-initiated reduction of uranium tetrafluoride by calcium metal

    International Nuclear Information System (INIS)

    West, M.H.; Martinez, M.M.; Nielsen, J.B.; Court, D.C.; Appert, Q.D.

    1995-09-01

    Uranium metal has numerous uses in conventional weapons (armor penetrators) and nuclear weapons. It also has application to nuclear reactor designs utilizing metallic fuels--for example, the former Integral Fast Reactor program at Argonne National Laboratory. Uranium metal also has promise as a material of construction for spent-nuclear-fuel storage casks. A new avenue for the production of uranium metal is presented that offers several advantages over existing technology. A carbon dioxide (CO 2 ) laser is used to initiate the reaction between uranium tetrafluoride (UF 4 ) and calcium metal. The new method does not require induction heating of a closed system (a pressure vessel) nor does it utilize iodine (I 2 ) as a chemical booster. The results of five reductions of UF 4 , spanning 100 to 200 g of uranium, are evaluated, and suggestions are made for future work in this area

  6. Preparation and study of the nitrides and mixed carbide-nitrides of uranium and of plutonium; Preparation et etude des nitrures et carbonitrures d'uranium et de plutonium

    Energy Technology Data Exchange (ETDEWEB)

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

    1966-06-01

    A detailed description is given of a simple method for preparing uranium and plutonium nitrides by the direct action of nitrogen under pressure at moderate temperatures (about 400 C) on the partially hydrogenated bulk metal. It is shown that there is complete miscibility between the UN and PuN phases. The variations in the reticular parameters of the samples as a function of temperature and in the presence of oxide have been used to detect and evaluate the solubility of oxygen in the different phases. A study has been made of the sintering of these nitrides as a function of the preparation conditions with or without sintering additives. A favorable but non-reproducible, effect has been found for traces of oxide. The best results were obtained for pure UN at 1600 C (96 per cent theoretical density) on condition that a well defined powder, was used. The criterion used is the integral width of the X-ray diffraction lines. The compounds UN and PuN are completely miscible with the corresponding carbides. This makes it possible to prepare carbide-nitrides of the general formula (U,Pu) (C,N) by solid-phase diffusion, at around 1400 C. The sintering of these carbide-nitrides is similar to that of the carbides if the nitrogen content is low; in particular, nickel is an efficient sintering agent. For high contents, the sintering is similar to that of pure nitrides. (author) [French] On decrit en detail une methode simple de preparation des nitrures d'uranium et de plutonium par action directe de l'azote sous pression, a temperature moyenne (vers 400 C), sur les metaux massifs partiellement hydrures. On montre que la miscibilite est complete entre les phases UN et PuN. L'evolution des parametres reticulaires des echantillons en fonction de la temperature et en presence d'oxyde a ete utilisee pour detecter et estimer la solubilite de l'oxygene dans les diverses phases. On a etudie le frittage de ces nitrures en fonction des conditions de preparation, avec ou sans additif de

  7. Uranium extraction from gold-uranium ores

    Energy Technology Data Exchange (ETDEWEB)

    Laskorin, B.N.; Golynko, Z.Sh.

    1981-01-01

    The process of uranium extraction from gold-uranium ores in the South Africa is considered. Flowsheets of reprocessing gold-uranium conglomerates, pile processing and uranium extraction from the ores are presented. Continuous counter flow ion-exchange process of uranium extraction using strong-active or weak-active resins is noted to be the most perspective and economical one. The ion-exchange uranium separation with the succeeding extraction is also the perspective one.

  8. Uranium mining

    International Nuclear Information System (INIS)

    2008-01-01

    Full text: The economic and environmental sustainability of uranium mining has been analysed by Monash University researcher Dr Gavin Mudd in a paper that challenges the perception that uranium mining is an 'infinite quality source' that provides solutions to the world's demand for energy. Dr Mudd says information on the uranium industry touted by politicians and mining companies is not necessarily inaccurate, but it does not tell the whole story, being often just an average snapshot of the costs of uranium mining today without reflecting the escalating costs associated with the process in years to come. 'From a sustainability perspective, it is critical to evaluate accurately the true lifecycle costs of all forms of electricity production, especially with respect to greenhouse emissions, ' he says. 'For nuclear power, a significant proportion of greenhouse emissions are derived from the fuel supply, including uranium mining, milling, enrichment and fuel manufacture.' Dr Mudd found that financial and environmental costs escalate dramatically as the uranium ore is used. The deeper the mining process required to extract the ore, the higher the cost for mining companies, the greater the impact on the environment and the more resources needed to obtain the product. I t is clear that there is a strong sensitivity of energy and water consumption and greenhouse emissions to ore grade, and that ore grades are likely to continue to decline gradually in the medium to long term. These issues are critical to the current debate over nuclear power and greenhouse emissions, especially with respect to ascribing sustainability to such activities as uranium mining and milling. For example, mining at Roxby Downs is responsible for the emission of over one million tonnes of greenhouse gases per year and this could increase to four million tonnes if the mine is expanded.'

  9. Final Project Report, DE-SC0001280, Characterizing the Combined Roles of Iron and Transverse Mixing on Uranium Bioremediation in Groundwater using Microfluidic Devices

    Energy Technology Data Exchange (ETDEWEB)

    Finneran, Kevin [Clemson Univ., SC (United States); Werth, Charles [Univ. of Texas, Austin, TX (United States); Strathmann, Timothy [Univ. of Illinois, Urbana-Champaign, IL (United States)

    2015-01-10

    In situ bioremediation of U(VI) involves amending groundwater with an appropriate electron donor and limiting nutrients to promote biological reduction to the less soluble and mobile U(IV) oxidation state. Groundwater flow is laminar; mixing is controlled by hydrodynamic dispersion. Recent studies indicate that transverse dispersion along plume margins can limit mixing of the amended electron donor and accepter (such as U(VI) in remediation applications). As a result, microbial growth, and subsequently contaminant reaction, may be limited to these transverse mixing zones during bioremediation. The primary objective of this work was to characterize the combined effects of hydrology, geochemistry, and biology on the (bio)remediation of U(VI). Our underlying hypothesis was that U(VI) reaction in groundwater is controlled by transverse mixing with an electron donor along plume margins, and that iron bioavailability in these zones affects U(VI) reduction kinetics and U(IV) re-oxidation. Our specific objectives were to a) quantify reaction kinetics mediated by biological versus geochemical reactions leading to U(VI) reduction and U(IV) re-oxidation, b) understand the influence of bioavailable iron on U(VI) reduction and U(IV) re-oxidation along the transverse mixing zones, c) determine how transverse mixing limitations and the presence of biomass in pores affects these reactions, and d) identify how microbial populations that develop along transverse mixing zones are influenced by the presence of iron and the concentration of electron donor. In the completed work, transverse mixing zones along plume margins were re-created in microfluidic pore networks, referred to as micromodels. We conducted a series of experiments that allowed us to distinguish among the hydraulic, biological, and geochemical mechanisms that contribute to U(VI) reduction, U(IV) re-oxidation, and U(VI) abiotic reaction with the limiting biological nutrient HP042-. This systematic approach may lead to a

  10. Uranium enrichment

    International Nuclear Information System (INIS)

    Rae, H.K.; Melvin, J.G.

    1988-06-01

    Canada is the world's largest producer and exporter of uranium, most of which is enriched elsewhere for use as fuel in LWRs. The feasibility of a Canadian uranium-enrichment enterprise is therefore a perennial question. Recent developments in uranium-enrichment technology, and their likely impacts on separative work supply and demand, suggest an opportunity window for Canadian entry into this international market. The Canadian opportunity results from three particular impacts of the new technologies: 1) the bulk of the world's uranium-enrichment capacity is in gaseous diffusion plants which, because of their large requirements for electricity (more than 2000 kW·h per SWU), are vulnerable to competition from the new processes; 2) the decline in enrichment costs increases the economic incentive for the use of slightly-enriched uranium (SEU) fuel in CANDU reactors, thus creating a potential Canadian market; and 3) the new processes allow economic operation on a much smaller scale, which drastically reduces the investment required for market entry and is comparable with the potential Canadian SEU requirement. The opportunity is not open-ended. By the end of the century the enrichment supply industry will have adapted to the new processes and long-term customer/supplier relationships will have been established. In order to seize the opportunity, Canada must become a credible supplier during this century

  11. Boron nitride coated uranium dioxide and uranium dioxide-gadolinium oxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Gunduz, G [Department of Chemical Engineering, Middle East Technical Univ., Ankara (Turkey); Uslu, I; Tore, C; Tanker, E [Turkiye Atom Enerjisi Kurumu, Ankara (Turkey)

    1997-08-01

    Pure Urania and Urania-gadolinia (5 and 10%) fuels were produced by sol-gel technique. The sintered fuel pellets were then coated with boron nitride (BN). This is achieved through chemical vapor deposition (CVD) using boron trichloride and ammonia. The coated samples were sintered at 1600 K. The analyses under scanning electron microscope (SEM) showed a variety of BN structures, mainly platelike and rodlike structures were observed. Burnup calculations by using WIMSD4 showed that BN coated and gadolinia containing fuels have larger burnups than other fuels. The calculations were repeated at different pitch distances. The change of the radius of the fuel pellet or the moderator/fuel ratio showed that BN coated fuel gives the highest burnups at the present design values of a PWR. Key words: burnable absorber, boron nitride, gadolinia, CVT, nuclear fuel. (author). 32 refs, 14 figs.

  12. Boron nitride coated uranium dioxide and uranium dioxide-gadolinium oxide fuels

    International Nuclear Information System (INIS)

    Gunduz, G.; Uslu, I.; Tore, C.; Tanker, E.

    1997-01-01

    Pure Urania and Urania-gadolinia (5 and 10%) fuels were produced by sol-gel technique. The sintered fuel pellets were then coated with boron nitride (BN). This is achieved through chemical vapor deposition (CVD) using boron trichloride and ammonia. The coated samples were sintered at 1600 K. The analyses under scanning electron microscope (SEM) showed a variety of BN structures, mainly platelike and rodlike structures were observed. Burnup calculations by using WIMSD4 showed that BN coated and gadolinia containing fuels have larger burnups than other fuels. The calculations were repeated at different pitch distances. The change of the radius of the fuel pellet or the moderator/fuel ratio showed that BN coated fuel gives the highest burnups at the present design values of a PWR. Key words: burnable absorber, boron nitride, gadolinia, CVT, nuclear fuel. (author). 32 refs, 14 figs

  13. A spectroscopic study of uranium species formed in chloride melts

    International Nuclear Information System (INIS)

    Volkovich, Vladimir A.; Bhatt, Anand I.; May, Iain; Griffiths, Trevor R.; Thied, Robert C.

    2002-01-01

    The chlorination of uranium metal or uranium oxides in chloride melts offers an acceptable process for the head-end of pyrochemical reprocessing of spent nuclear fuels. The reactions of uranium metal and ceramic uranium dioxide with chlorine and with hydrogen chloride were studied in the alkali metal chloride melts, NaCl-KCl at 973K, NaCl-CsCl between 873 and 923K and LiCl-KCl at 873K. The uranium species formed therein were characterized from their electronic absorption spectra measured in situ. The kinetic parameters of the reactions depend on melt composition, temperature and chlorinating agent used. The reaction of uranium dioxide with oxygen in the presence of alkali metal chlorides results in the formation of alkali metal uranates. A spectroscopic study, between 723 and 973K, on their formation and their solutions was undertaken in LiCl, LiCl-KCl eutectic and NaCl-CsCl eutectic melts. The dissolution of uranium dioxide in LiCl-KCl eutectic at 923K containing added aluminium trichloride in the presence of oxygen has also been investigated. In this case, the reaction leads to the formation of uranyl chloride species. (author)

  14. Soil treatment to remove uranium and related mixed radioactive heavy metal contaminants. Ninth quarterly technical and financial progress report, January 1, 1995--March 31, 1995

    International Nuclear Information System (INIS)

    1995-05-01

    The objective of this project is to design and develop a physico-chemical treatment process for the removal of uranium and heavy metals from contaminated soil to achieve target contamination levels below 35 pCi/g of soil and a target for non-radioactive heavy metals below concentration levels permissible for release of the soil. The work will involve bench-scale and pilot-scale tests, using chelation-flotation, chemical leaching and ultrasonic leaching techniques, in conjunction with cross-flow microfiltration and filter-press operations. The effectiveness of an integrated process to treat leachates generated from soil processing will be demonstrated. Process flow-sheets suitable for in-situ and ex-situ applications will be developed and preliminary costs will be provided for the soil and leachate treatment technologies. In accordance with 10CFR 600.31 (d)(i), an extension of the project period including final report submission to 31 July 1995 was made in anticipation of potential delays in receiving Fernald soil samples at Chalk River Laboratories for the planned pilot-scale verification tests. Ex-situ pilot-scale soil decontamination and leachate treatment tests using Chalk River Chemical Pit soil are nearing completion. Soil decontamination tests using Fernald Incinerator Area soil originally scheduled for February 1995 was postponed to May 1995 as result of unexpected delays in the preparation of two drums of soils (∼416 kg) by FERMCO and paperwork required to arrange for export/import licenses

  15. Uranium update

    International Nuclear Information System (INIS)

    Steane, R.

    1997-01-01

    This paper is about the current uranium mining situation, especially that in Saskatchewan. Canada has a unique advantage with the Saskatchewan uranium deposits. Making the most of this opportunity is important to Canada. The following is reviewed: project development and the time and capital it takes to bring a new project into production; the supply and demand situation to show where the future production fits into the world market; and our foreign competition and how we have to be careful not to lose our opportunity. (author)

  16. Machining of uranium and uranium alloys

    International Nuclear Information System (INIS)

    Morris, T.O.

    1981-01-01

    Uranium and uranium alloys can be readily machined by conventional methods in the standard machine shop when proper safety and operating techniques are used. Material properties that affect machining processes and recommended machining parameters are discussed. Safety procedures and precautions necessary in machining uranium and uranium alloys are also covered. 30 figures

  17. Microbial uptake of uranium, cesium, and radium

    Energy Technology Data Exchange (ETDEWEB)

    Strandberg, G.W.; Shumate, S.E. II; Parrott, J.R. Jr.; McWhirter, D.A.

    1980-01-01

    The ability of diverse microbial species to concentrate uranium, cesium, and radium was examined. Saccharomyces cerevisiae, Pseudomonas aeruginosa, and a mixed culture of denitrifying bacteria accumulated uranium to 10 to 15% of the dry cell weight. Only a fraction of the cells in a given population had visible uranium deposits in electron micrographs. While metabolism was not required for uranium uptake, mechanistic differences in the metal uptake process were indicated. Uranium accumulated slowly (hours) on the surface of S. cerevisiae and was subject to environmental factors (i.e., temperature, pH, interfering cations and anions). In contrast, P. aeruginosa and the mixed culture of denitrifying bacteria accumulated uranium rapidly (minutes) as dense, apparently random, intracellular deposits. This very rapid accumulation has prevented us from determining whether the uptake rate during the transient between the initial and equilibrium distribution of uranium is affected by environmental conditions. However, the final equilibrium distributions are not affected by those conditions which affect uptake by S. cerevisiae. Cesium and radium were concentrated to a considerably lesser extent than uranium by the several microbial species tested. The potential utility of microorganisms for the removal and concentration of these metals from nuclear processing wastes and several bioreactor designs for contacting microorganisms with contaminated waste streams will be discussed.

  18. Method of precipitating uranium from an aqueous solution and/or sediment

    Science.gov (United States)

    Tokunaga, Tetsu K; Kim, Yongman; Wan, Jiamin

    2013-08-20

    A method for precipitating uranium from an aqueous solution and/or sediment comprising uranium and/or vanadium is presented. The method includes precipitating uranium as a uranyl vanadate through mixing an aqueous solution and/or sediment comprising uranium and/or vanadium and a solution comprising a monovalent or divalent cation to form the corresponding cation uranyl vanadate precipitate. The method also provides a pathway for extraction of uranium and vanadium from an aqueous solution and/or sediment.

  19. Process development study on production of uranium metal from monazite sourced crude uranium tetra-fluoride

    International Nuclear Information System (INIS)

    Chowdhury, S; Satpati, S.K.; Hareendran, K.N.; Roy, S.B.

    2014-01-01

    Development of an economic process for recovery, process flow sheet development, purification and further conversion to nuclear grade uranium metal from the crude UF 4 has been a technological challenge and the present paper, discusses the same.The developed flow-sheet is a combination of hydrometallurgical and pyrometallurgical processes. Crude UF 4 is converted to uranium di-oxide (UO 2 ) by chemical conversion route and UO 2 produced is made fluoride-free by repeated repulping, followed by solid liquid separation. Uranium di-oxide is then purified by two stages of dissolution and suitable solvent extraction methods to get uranium nitrate pure solution (UNPS). UNPS is then precipitated with air diluted ammonia in a leak tight stirred vessel under controlled operational conditions to obtain ammonium di-uranate (ADU). The ADU is then calcined and reduced to produce metal grade UO 2 followed by hydro-fluorination using anhydrous hydrofluoric acid to obtain metal grade UF 4 with ammonium oxalate insoluble (AOI) content of 4 is essential for critical upstream conversion process. Nuclear grade uranium metal ingot is finally produced by metallothermic reduction process at 650℃ in a closed vessel, called bomb reactor. In the process, metal-slag separation plays an important role for attaining metal purity as well as process yield. Technological as well economic feasibility of indigenously developed process for large scale production of uranium metal from the crude UF 4 has been established in Bhabha Atomic Research Centre (BARC), India

  20. Manufacturing sector carbon dioxide emissions in nine OECD countries 1973--87: A Divisia index decomposition to changes in fuel mix, emission coefficients, industry structure, energy intensities, and international structure

    International Nuclear Information System (INIS)

    Torvanger, A.

    1990-11-01

    In this paper the reduction in energy-related manufacturing carbon dioxide emissions for nine OECD countries in the period 1973 to 1987 is analyzed. Carbon dioxide emissions are estimated from energy use data. The emphasis is on carbon dioxide intensities, defined as emissions divided by value added. The overall manufacturing carbon dioxide intensity for the nine OECD countries was reduced by 42% in the period 1973--1987. Five fuels are specified together with six subsectors of manufacturing. Carbon dioxide emissions are estimated from fossil fuel consumption, employing emissions coefficients for gas, oil and solids. In addition, electricity consumption is specified. For electricity use an emission coefficient index is calculated from the shares of fossil fuels, nuclear power and hydro power used to generate electricity, and the efficiency in electricity generation from these energy sources. A Divisia index approach is used to sort out the contribution to reduced carbon dioxide intensity from different components. The major finding is that the main contribution to reduced carbon dioxide intensity is from the general reduction in manufacturing energy intensity, most likely driven by economic growth and increased energy prices, giving incentives to invest in new technology and new industrial processes. There is also a significant contribution from reduced production in the most carbon dioxide intensive subsectors, and a contribution from higher efficiency in electricity generation together with a larger nuclear power share at the expense of oil. 19 refs., 5 figs., 11 tabs