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Sample records for uranium nitrate solution

  1. Removal of uranium from ammonium nitrate solution by nanofiltration

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Runci; Yuan, Zhongwei; Yan, Taihong; Zheng, Weifang [China Institute of Atomic Energy, Beijing (China). Dept. of Radiochemistry

    2017-07-01

    Two types of nanofiltration membranes were tested to remove uranium dissolved in ammonium nitrate solution, and the influence of operating parameters as transmembrane pressure, tangential velocity and feed temperature was investigated. Experimental results showed NF270 membrane can reject more than 96% uranium and allow most (90% min) ammonium nitrate solution passed by, and with a permeate flux of 60 L/(m{sup 2}.h). Nanofiltration seems to be a promising technology for the removal of uranium and recovery of ammonium nitrate simultaneously.

  2. Study on removing nitrate from uranium solution by ion-exchange method

    International Nuclear Information System (INIS)

    Zhou Genmao

    2004-01-01

    Nitrate of low concentration can interfere with adsorption of uranyl sulfate anion on anion-exchange resins because the anion-exchange resins have a stronger affinity for nitrate in uranium solution. Nitrate can be adsorbed with a high efficiency resin, then desorbed by sodium hydroxide. The nitrate concentration is about 60 g/L in eluate. The research results show that nitrate can be recovered from uranium solution with N-3 anion-exchange resin

  3. Plans and equipment for criticality measurements on plutonium-uranium nitrate solutions

    International Nuclear Information System (INIS)

    Lloyd, R.C.; Clayton, E.D.; Durst, B.M.

    1982-01-01

    Data from critical experiments are required on the criticality of plutonium-uranium nitrate solutions to accurately establish criticality control limits for use in processing and handling of breeder type fuels. Since the fuel must be processed both safely and economically, it is necessary that criticality considerations be based on accurate experimental data. Previous experiments have been reported on plutonium-uranium solutions with Pu weight ratios extending up to some 38 wt %. No data have been presented, however, for plutonium-uranium nitrate solutions beyond this Pu weight ratio. The current research emphasis is on the procurement of criticality data for plutonium-uranium mixtures up to 60 wt % Pu that will serve as the basis for handling criticality problems subsequently encountered in the development of technology for the breeder community. Such data also will provide necessary benchmarks for data testing and analysis on integral criticality experiments for verification of the analytical techniques used in support of criticality control. Experiments are currently being performed with plutonium-uranium nitrate solutions in stainless steel cylindrical vessels and an expandable slab tank system. A schematic of the experimental systems is presented

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

    1990-01-01

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

  6. Corrosion behavior of 321 stainless steel in low-acidity uranium nitrate solution

    International Nuclear Information System (INIS)

    Liao Junsheng; Sun Ying; Zhang Wanglin; Ding Ping; Yang Jiangrong; Wu Lunqiang

    2003-01-01

    Weighing and electrochemical methods have been used to investigate the high-temperature uniform corrosion and electrochemical corrosion behavior of lCr18Ni9Ti (321) stainless steel in uranium nitrate solution at different concentrations and pH values. The uniform corrosion results showed that the corrosion rate of 321 stainless steel was less than 0.04 g/m 2 .h, and the visible change of surface smoothness was not observed through 960 h. It was perfect corrosion-resisting in obtained conditions. The electro-chemical corrosion behavior study has been performed to investigate 321 stainless steel in uranium nitrate solutions of the dissolved and saturated oxygen. The corrosion potential and corrosion current density were obtained. Auger photoelectron spectroscopy for measurement of uranium in specimen was used to indicate that uranium is in corrosion product. The corrosion film was measured by Ar ion gun sputter, and the thickness is 10-15 nm. (authors)

  7. Rapid determination of fluoride in uranyl nitrate solution obtained in conversion process of uranium tetrafluoride

    International Nuclear Information System (INIS)

    Levin, R.; Feldman, R.; Sahar, E.

    1976-01-01

    In uranium production the conversion of impure uranium tetrafluoride by sodium hydroxide was chosen as a current process. A rapid method for determination of fluoride in uranyl-nitrate solution was developed. The method includes precipitation of uranium as diuranate, separation by centrifugation, and subsequent determination of fluoride in supernate by titration with thorium nitrate. Fluoride can be measured over the range 0.15-2.5 gr/gr U, with accuracy of +-5%, within 15 minutes. (author)

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

    International Nuclear Information System (INIS)

    2003-01-01

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

  9. Standard test method for isotopic analysis of hydrolyzed uranium hexafluoride and uranyl nitrate solutions by thermal ionization mass spectrometry

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2005-01-01

    1.1 This method applies to the determination of isotopic composition in hydrolyzed nuclear grade uranium hexafluoride. It covers isotopic abundance of 235U between 0.1 and 5.0 % mass fraction, abundance of 234U between 0.0055 and 0.05 % mass fraction, and abundance of 236U between 0.0003 and 0.5 % mass fraction. This test method may be applicable to other isotopic abundance providing that corresponding standards are available. 1.2 This test method can apply to uranyl nitrate solutions. This can be achieved either by transforming the uranyl nitrate solution to a uranyl fluoride solution prior to the deposition on the filaments or directly by depositing the uranyl nitrate solution on the filaments. In the latter case, a calibration with uranyl nitrate standards must be performed. 1.3 This test method can also apply to other nuclear grade matrices (for example, uranium oxides) by providing a chemical transformation to uranyl fluoride or uranyl nitrate solution. 1.4 This standard does not purport to address al...

  10. Derivation of an empirical formula for determining water content of mixed uranyl nitrate-thorium nitrate solutions

    International Nuclear Information System (INIS)

    Min, Duck Kee; Choi, Byung Il; Ro, Seung Gy; Eom, Tae Yoon; Kim, Zong Goo

    1986-01-01

    Densities of a large number of mixed uranyl nitrate-thorium nitrate solutions were measured with pycnometer. By the least squares analysis of the experimental result, an empirical formula for determining water content of mixed uranyl nitrate-thorium nitrate solutions as functions of uranium concentration, thorium concentration and nitric acid normality is derived; W=1.0-0.3580 C u -0.4538 C Th -0.0307H + where W, C u , C Th , and H + stand for water content(g/cc), uranium concentration (g/cc), thorium concentration(g/cc), and nitric acid normality, respectively. Water contents of the mixed uranyl nitrate-thorium nitrate solutions are calculated by using the empirical formular, and compared with the values calculated by Bouly's equation in which an additional data, solution density, is required. The two results show good agreements within 2.7%. (Author)

  11. A study of precipitation from pure solutions of uranyl nitrate; Etude de la precipitaion de solutions pures de nitrate d'uranyle

    Energy Technology Data Exchange (ETDEWEB)

    Decrop, J; Holder, J; Sauteron, J [Commissariat a l' Energie Atomique, Usine du Bouchet, Service des Lab. de Recherches et de Controle, Saclay (France). Centre d' Etudes Nucleaires

    1961-07-01

    After its purification by extraction of the uranyl nitrate from the organic solvent, uranium has to be converted into solid form again: uranium trioxide (UO{sub 3}). It can be done either by thermal decomposition of uranyl nitrate or by precipitation of uranium, followed by filtration and calcination. Only the second method has been studied for now at the Bouchet plant. This paper reports the bench-scale and pilot-scale experiments of the studies of the precipitation of pure solutions of uranyl nitrate using ammonia (gaseous or in solution) or ammonium carbonate. These have been carried out at the Bouchet plant. It investigates the chemical aspect (pH, precipitates chemical composition) and the technical aspect of the different ways of precipitation (conditions of precipitation, decantation and filtration of precipitates). (M.P.)

  12. Recovery of uranium from uranyl nitrate raffinate. Contributed Paper PE-06

    International Nuclear Information System (INIS)

    Anilkumar Reddy, A.M.; Shiva Kumar, M.; Varadan, K.M.K.; Babaji, P.; Sairam, S. Sheela; Saibaba, N.

    2014-01-01

    At New Uranium Oxide Fuel Plant, NUOFP(O) of Nuclear Fuel Complex (NFC), the Uranyl Nitrate Raffinate (UNR) generated during solvent extraction process is washed with Treated Lean Solvent(TLS) to recover residual U. Earlier this UNR consisting of 0.5-1 gm/l and 2.5 FA was neutralised with vapour ammonia. The slurry was then filtered over pre coat drum filter and the resultant Uranyl Nitrate Raffinate cake (UNRC) was stored in polyethylene lined MS drums. The valuable U was thus being locked up in UNRC. Also, the storage of UNRC drums required lot of floor space which have to be repacked frequently to contain the radioactivity. Hence the need has come to avoid the generation of UNRC and the recovery of U from the already generated UNRC. The generation of UNRC was avoided by developing alternate process of UNR treatment with Treated Lean Solvent for the removal of residual U and the resulting Acidic Raffinate Slurry (ARS) is disposed. The Uranium recovery from UNRC is done by dissolving the cake in Uranyl Nitrate Raffinate solution to leach the hexavalent Uranium by utilizing the free acidity in UNR. The leaching time is about six hours and the uranium forms uranyl nitrate. The resulting leach solutions are relatively dilute but complex acidic nitrate solutions containing wide variety of ions. Metallic ions commonly present include uranium, iron, magnesium, aluminium, sodium, calcium etc. The uranium concentration is normally 1-1.5 g/L. This uranium is separated by solvent extraction. The active agent in solvent extraction is Tri Butyl Phosphate in kerosene that can selectively extract uranium into an organic complex which is insoluble in aqueous. The organic used for extraction is Treated Lean Solvent in the quality of freshly prepared solvent and the resulting Acidic Raffinate Slurry is disposed by sale. The leaching of Uranium from UNRC was done in plant scale and about 1200 kgs of UNRC was successfully processed in trial batch. The paper deals with details of

  13. Standard test method for gamma energy emission from fission products in uranium hexafluoride and uranyl nitrate solution

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2005-01-01

    1.1 This test method covers the measurement of gamma energy emitted from fission products in uranium hexafluoride (UF6) and uranyl nitrate solution. It is intended to provide a method for demonstrating compliance with UF6 specifications C 787 and C 996 and uranyl nitrate specification C 788. 1.2 The lower limit of detection is 5000 MeV Bq/kg (MeV/kg per second) of uranium and is the square root of the sum of the squares of the individual reporting limits of the nuclides to be measured. The limit of detection was determined on a pure, aged natural uranium (ANU) solution. The value is dependent upon detector efficiency and background. 1.3 The nuclides to be measured are106Ru/ 106Rh, 103Ru,137Cs, 144Ce, 144Pr, 141Ce, 95Zr, 95Nb, and 125Sb. Other gamma energy-emitting fission nuclides present in the spectrum at detectable levels should be identified and quantified as required by the data quality objectives. 1.4 This standard does not purport to address all of the safety concerns, if any, associated with its us...

  14. Study of electrolytic reduction of uranium VI to uranium IV in nitrate systems

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, B.F. de; Almeida, S.G. de; Forbicini, S; Matsuda, H T; Araujo, J.A. de [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil). Centro de Engenharia Quimica

    1980-01-01

    Experimental parameters are optimized in order to obtain uranium (IV) nitrate solutions at maximum yield, using hydrazine as stabilizer. Uranium (VI) electrolytic reduction was chosen because: there is no increase in the volume of radioactive effluents; there are no secondary reactions; there is no need for further separations; all reagents used are not inflammable. The method is, therefore, efficient and of low cost.

  15. Critical experiment study on uranyl nitrate solution experiment facility

    International Nuclear Information System (INIS)

    Zhu Qingfu; Shi Yongqian; Wang Jinrong

    2005-01-01

    The Uranyl Nitrate Solution Experiment Facility was constructed for the research on nuclear criticality safety. In this paper, the configuration of the facility is introduced; a series of critical experiments on uranyl nitrate solution is described later, which were performed for various uranium concentrations under different conditions, i.e. with or without neutron absorbers in the core and with or without water-reflector outside the core. Critical volume and the minimum 235U critical mass for different uranium concentrations are presented. Finally, theoretical analysis is made on the experimental results. (authors)

  16. A study of precipitation from pure solutions of uranyl nitrate

    International Nuclear Information System (INIS)

    Decrop, J.; Holder, J.; Sauteron, J.

    1961-01-01

    After its purification by extraction of the uranyl nitrate from the organic solvent, uranium has to be converted into solid form again: uranium trioxide (UO 3 ). It can be done either by thermal decomposition of uranyl nitrate or by precipitation of uranium, followed by filtration and calcination. Only the second method has been studied for now at the Bouchet plant. This paper reports the bench-scale and pilot-scale experiments of the studies of the precipitation of pure solutions of uranyl nitrate using ammonia (gaseous or in solution) or ammonium carbonate. These have been carried out at the Bouchet plant. It investigates the chemical aspect (pH, precipitates chemical composition) and the technical aspect of the different ways of precipitation (conditions of precipitation, decantation and filtration of precipitates). (M.P.)

  17. Simultaneous analysis of uranium and nitrate

    Energy Technology Data Exchange (ETDEWEB)

    Bostick, D.T.

    1978-04-01

    A direct spectrophotometric method has been developed for the determination of 20 to 200 g/l of uranium in the presence of 3 to 5 M nitric acid. A dual-wavelength analysis is used to eliminate the enhancing effect of nitrate ion on the uranium visible spectra. The precision and accuracy of the simultaneous analysis of uranium and nitrate were compared using combinations of the four uranium wavelength maxima, occurring at 426, 416, 403 and 359 nm. Calculations based on 426 and 416 nm data yielded the most accurate results. The calculated relative standard deviation of uranium and nitrate concentrations was 5.4 percent and 15.5 percent, respectively. The photometric procedure is slightly affected by temperature; an increase of one degree centigrade results in a 0.2 g/l overestimation of uranium concentration. Because the method is non-destructive, it is directly applicable to the continuous in-line analysis of dissolved uranium in aqueous fuel reprocessing streams.

  18. Distribution of iron during full loading of amberlite IRC-72 resin with uranium from nitrate solutions at 300C

    International Nuclear Information System (INIS)

    Shaffer, J.H.; Greene, C.W.

    1979-01-01

    The integrity of resin-based fuel kernels used in the fabrication of fuel elements for a high-temperature gas-cooled reactor will depend, in part, on the concentration of iron incorporated in the resin particles during their loading with uranium. Consequently, assessment of chemical specifications for iron as an impurity in uranyl nitrate solution should be based on its distribution during the resin loading operation. For this purpose, the behavior of iron, as an impurity in uranyl nitrate solutions, was investigated under equilibrium conditions at 30 0 C during full loading of Amberlite IRC-72 cation exchange reaction were derived from calculations based on complex coordination of ferric ion with the resin over the nitrate ion concentration range of approx. 0.5 to 2 N

  19. Determination of the total nitrate content of thorium nitrate solution with a selective electrode

    International Nuclear Information System (INIS)

    Wirkner, F.M.

    1979-01-01

    The nitrate content of thorium nitrate solutions is determined with a liquid membrane nitrate selective electrode utilizing the known addition method in 0.1 M potassium fluoride medium as ionic strength adjustor. It is studied the influence of pH and the presence of chloride, sulphate, phosphate, meta-silicate, thorium, rare earths, iron, titanium, uranium and zirconium at the same concentrations as for the aqueous feed solutions in the thorium purification process. The method is tested in synthetic samples and in samples proceeding from nitric dissolutions of thorium hidroxide and thorium oxicarbonate utilized as thorium concentrates to be purified [pt

  20. Fixation and separation of the elements thorium and uranium using anion exchange resins in nitrate solution

    International Nuclear Information System (INIS)

    Korgaonkar, V.

    1967-10-01

    The exchange of thorium and uranium between a strong base anion resin and a mixed water + ethanol solvent containing nitrate ions is studied. It is assumed that in the resin the thorium and uranium are fixed in the form of the complexes Th(NO 3 ) 6 2- and UO 2 (NO 3 ) 4 2- in solution these elements are present in the form of complexes having the general formula: Th(NO 3 ) 6-n n-2 and UO 2 (NO 3 ) 4-n n-2 It has been possible to deduce a law for the changes in the partition functions of thorium and uranium as a function of the concentrations of the various species in solution and of the complexing ion NO 3 . From this has been deduced the optimum operational conditions for separating a mixture of these two elements. Finally, in these conditions, the influence of a few interfering ions has been studied: Ba, Bi, Ce, La, Mo, Pb, Zr. The method proposed can be used either as a preparation, or for the dosage of thorium by a quantitative separation. (author) [fr

  1. An investigation to compare the performance of methods for the determination of free acid in highly concentrated solutions of plutonium and uranium nitrate

    International Nuclear Information System (INIS)

    Crossley, D.

    1980-08-01

    An investigation has been carried out to compare the performance of the direct titration method and the indirect mass balance method, for the determination of free acid in highly concentrated solutions of uranium nitrate and plutonium nitrate. The direct titration of free acid with alkali is carried out in a fluoride medium to avoid interference from the hydrolysis of uranium or plutonium, while free acid concentration by the mass balance method is obtained by calculation from the metal concentration, metal valency state, and total nitrate concentration in a sample. The Gran plot end-point prediction technique has been used extensively in the investigation to gain information concerning the hydrolysis of uranium and plutonium in fluoride media and in other complexing media. The use of the Gran plot technique has improved the detection of the end-point of the free acid titration which gives an improvement in the precision of the determination. The experimental results obtained show that there is good agreement between the two methods for the determination of free acidity, and that the precision of the direct titration method in a fluoride medium using the Gran plot technique to detect the end-point is 0.75% (coefficient of variation), for a typical separation plant plutonium nitrate solution. The performance of alternative complexing agents in the direct titration method has been studied and is discussed. (author)

  2. Corrosion evaluation of uranyl nitrate solution evaporator and denitrator in Tokai reprocessing plant

    International Nuclear Information System (INIS)

    Yamanaka, Atsushi; Hashimoto, Kowa; Uchida, Toyomi; Shirato, Yoji; Isozaki, Toshihiko; Nakamura, Yoshinobu

    2011-01-01

    The Tokai reprocessing plant (TRP) adopted the PUREX method in 1977 and has reprocessed spent nuclear fuel of 1140 tHM (tons of heavy metals) since then. The reprocessing equipment suffers from various corrosion phenomena because of high nitric acidity, solution ion concentrations, such as uranium, plutonium, and fission products, and temperature. Therefore, considering corrosion performance in such a severe environment, stainless steels, titanium steel, and so forth were employed as corrosion resistant materials. The severity of the corrosive environment depends on the nitric acid concentration and the temperature of the solution, and uranium in the solution reportedly does not significantly affect the corrosion of stainless steels and controls the corrosion rates of titanium steel. The TRP equipment that handles uranyl nitrate solution operates at a low nitric acid concentration and has not experienced corrosion problems until now. However, there is a report that corrosion rates of some stainless steels increase in proportion to rising uranium concentrations. The equipment that handles the uranyl nitrate solution in the TRP includes the evaporators, which concentrate uranyl nitrate to a maximum concentration of about 1000 gU/L (grams of uranium per liter), and the denitrator, where uranyl nitrate is converted to UO 3 powder at about 320degC. These equipments are therefore required to grasp the degree of the progress of corrosion to handle high-temperature and high-concentration uranyl nitrate. The evaluation of this equipment on the basis of thickness measurement confirmed only minor corrosion and indicated that the equipment would be fully adequate for future operation. (author)

  3. Design of one evaporation system for uranyl nitrate solution

    International Nuclear Information System (INIS)

    Mancilla Romero, R.J.

    1975-01-01

    The authors propose an instant evaporation system with recirculation of the concentrated solution to raise the concentration from 50 to 1500 g of uranium per litre of solution. The capacity of the plant is to be 14.1 kg of uranium per hour. The main equipment used in the system is as follows: 1. Ring-type heat exchanger, for increasing the temperature of the mixture of fresh and recirculated solution from 80 to 115 0 C; 2. Separation tank, in which instant evaporation is carried out. The absolute pressure inside the tank will be 500 mmHg, with steam separation from a concentrated (78.5 wt.%) uranyl nitrate solution; 3. Desuperheater-condenser of horizontal tubular type for condensing water vapour and recovering any uranyl nitrate that may have been entrained; 4. Storage tank for the concentrate, with a capacity for one day's normal operation, and a heating coil to prevent crystallization of the concentrated solution; 5. Two storage tanks for feed and condensate with capacity for one day's normal operation; 6. Supporting structure for the above components. Virtually all equipment in contact with the uranyl nitrate solution will be made of 304 stainless steel. Saturated steam at 143.3 0 C will be required. The cost of the proposed system is $543 030.00. (author)

  4. Contribution to the study of the evaporation of aqueous uranyl nitrate solutions; Contribution a l'etude de l'evaporation des solutions aqueuses de nitrate d'uranyle

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-05-15

    This work was carried out with a view to define the conditions under which is affected the concentration of aqueous uranyl nitrate solutions one of the steps in uranium extraction metallurgy. The first port is devoted to the experimental determination of the physical characteristics of aqueous uranyl nitrate solutions, from dilute to concentrated solutions. The second part of this work is devoted to the isothermal evaporation of solution a west ted-wall column; this chemical engineering study has been more particularly devoted to the definition of the influence of the dynamics of the liquid phase on the exchange of matter between the two phases in contact. (author) [French] La concentration par evaporation des solutions aqueuses de nitrate d'uranyle constitue une etape de la metallurgie de l'uranium dont ce travail a voulu preciser la connaissance. Dans ce but, une premiere partie a ete consacree a la determination experimentale de caracteristiques physiques des solutions aqueuses de nitrate d'uranyle, des solutions diluees aux solutions saturees. Dans une deuxieme partie, ce travail a porte sur l'evaporation isotherme des solutions dans une colonne a paroi mouillee; cette etude de genie chimique a ete plus particulierement orientee de facon a preciser l'influence de la dynamique de la phase liquide sur l'echange de matiere entre les deux phases en contact. (auteur)

  5. Fixation and separation of the elements thorium and uranium using anion exchange resins in nitrate solution; Fixation et separation des elements thorium et uranium par les resines echangeuses d'anions en milieu nitrate

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-10-01

    The exchange of thorium and uranium between a strong base anion resin and a mixed water + ethanol solvent containing nitrate ions is studied. It is assumed that in the resin the thorium and uranium are fixed in the form of the complexes Th(NO{sub 3}){sub 6}{sup 2-} and UO{sub 2}(NO{sub 3}){sub 4}{sup 2-} in solution these elements are present in the form of complexes having the general formula: Th(NO{sub 3}){sub 6-n}{sup n-2} and UO{sub 2}(NO{sub 3}){sub 4-n}{sup n-2} It has been possible to deduce a law for the changes in the partition functions of thorium and uranium as a function of the concentrations of the various species in solution and of the complexing ion NO{sub 3}. From this has been deduced the optimum operational conditions for separating a mixture of these two elements. Finally, in these conditions, the influence of a few interfering ions has been studied: Ba, Bi, Ce, La, Mo, Pb, Zr. The method proposed can be used either as a preparation, or for the dosage of thorium by a quantitative separation. (author) [French] On etudie l'echange du thorium et de l'uranium entre une resine anion base forte et un solvant mixte eau + ethanol charge en ions nitrates. On a suppose que, dans la resine, le thorium et l'uranium sont fixes sous forme de complexes Th(NO{sub 3}){sub 6}{sup 2-} et UO{sub 2}(NO{sub 3}){sub 4}{sup 2-} en solution, ces elements sont engages dans des complexes de formule generale: Th(NO{sub 3}){sub 6-n}{sup n-2} and UO{sub 2}(NO{sub 3}){sub 4-n}{sup n-2} On a pu degager une loi de variation des coefficients de partage du thorium et de l'uranium en fonction des concentrations des diverses especes en solution et de l'anion complexant NO{sub 3}{sup -}. On en a deduit les conditions operatoires optimales necessaires pour separer les deux elements a partir de leurs melanges. Enfin, dans ces conditions, on a etudie l'influence de quelques elements genants: Ba, Bi, Ce, La, Mo, Pb, Zr. La methode preconisee peut etre

  6. Ion exchange separation of nitrate from uranium compounds and its determination by spectrophotometry and ion chromatography

    International Nuclear Information System (INIS)

    Pires, M.A.F.; Atalla, L.T.; Abrao, A.

    1985-11-01

    A procedure for the separation of nitrate from uranium compounds by retaintion of uranyl ion on a cationic ion exchanger and its determination in the effluent is described. Nitrate is analysed by the spectrometric method with 1-phenol-2,4-dissulphonic acid. This determination covers the 1 to 10 μg NO - 3 /mL range and requires an amount of 10 to 100 μg NO - 3 . The main interference is uranium (VI) due its own intense yellow color. This difficulty is overcome by the complete separation of UO 2 ++ with the cationic resin. Alternatively, the ion chromatography technique is used for the determination of nitrate in the effluent of the cationic resin. The determination was easily made by the comparison of the nitrate peak hights of the analyte and the standard solutions. The ion chromatography method is very sensitive (0,3 μg NO - 3 /mL), reproducible and suitable for routine analysis and permits the determination of fraction of part per million of nitrate in uranium. The results of nitrate determination using both spectrophotometric and ion chromatography techniques are compared. The method is being routinely applied for the quality control of uranium compounds in the fuel cycle, specially uranium oxide, ammonium diuranate, uranium peroxide and ammonium uranyl tricarbonate. (Author) [pt

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

    International Nuclear Information System (INIS)

    Masduki, B.; Wardaya; Widarmoko, A.

    1996-01-01

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

  8. Study on the electrolytic reduction of Uranium-VI to Uranium-IV in a nitrate system

    International Nuclear Information System (INIS)

    Araujo, B.F. de; Almeida, S.G. de; Forbicini, S.; Matsuda, H.T.; Araujo, J.A. de.

    1981-05-01

    The determination of the best conditions to prepare hydrazine stabilized uranium (IV) nitrate solutions for utilization in Purex flowsheets is dealt with. Electrolytic reduction of U(VI) has been selected as the basic method, using an open electrolytic cell with titanum and platinum electrodes. The hydrazine concentration, the current density, acidity, U(VI) concentration and reduction time were the parameters studied and U(IV)/U(VI) ratio was used to evaluate the degree of reduction. From the results it could be concluded that the technique is reliable. The U(IV) solutions remains constant for at least two weeks and can be used in the chemical processing of irradiated uranium fuels. (Author) [pt

  9. EFFECT OF CURRENT, TIME, FEED AND CATHODE TYPE ON ELECTROPLATING PROCESS OF URANIUM SOLUTION

    Directory of Open Access Journals (Sweden)

    Sigit Sigit

    2017-02-01

    Full Text Available ABSTRACT   EFFECT OF CURRENT, TIME, FEED AND CATHODE TYPE ON ELECTROPLATING PROCESS OF URANIUM SOLUTION. Electroplating process of uranyl nitrate and effluent process has been carried out in order to collect uranium contained therein using electrode Pt / Pt and Pt / SS at various currents and times. Material used for electrode were Pt (platinum and SS (Stainlees Steel. Feed solution of 250 mL was entered into a beaker glass equipped with Pt anode - Pt cathode or Pt anode - SS cathode, then fogged direct current from DC power supply with specific current and time so that precipitation of uranium sticking to the cathode. After the processes completed, the cathode was removed and weighed to determine weight of precipitates, while the solution was analyzed to determine the uranium concentration decreasing after and before electroplating process. The experiments showed that a relatively good time to acquire uranium deposits at the cathode was 1 hour by current 7 ampere, uranyl nitrate as feed, and Pt (platinum as cathode. In these conditions, uranium deposits attached to the cathode amounted to 74.96% of the original weight of uranium oxide in the feed or 206.5 mg weight. The use of Pt cathode for  uranyl nitrate, SS and Pt cathode for effluent process feed gave uranium specific weight at the cathode of 12.99 mg/cm2, 2.4 mg/cm2 and 5.37 mg/cm2 respectively for current 7 ampere and electroplating time 1 hour. Keywords: Electroplating, uranyl nitrate, effluent process, Pt/Pt electrode, Pt/SS electrode

  10. Investigation into sorption of uranium fron its high-concentrated nitric acid solutions on resin AMP

    International Nuclear Information System (INIS)

    Savel'eva, V.I.; Sudarikov, B.N.; Kireeva, G.N.; Ryzhkova, V.N.; Kandaryuk, V.V.

    1976-01-01

    Sorption of uranium has been studied on strongly basic anion-exchange resin from nitric acid solutions with concentration in metal 10-150 g/l in presence of sodium, calcium, and aluminium nitrates. Sorption of uranium from solutions has been performed by the static method with the aid of contacting the initial solution with airdry resin for 4 hours, resin to solution ratio being 1:12.5. It has been established that sorption of uranium increases with a rise in concentration of salting out agents in the following order: Al(NO 3 ) 3 > Ca(NO 3 ) 2 > Na(NO 3 ). It has been shown spectrophotometricatly that in solutions of nitrates and HNO 3 with a concentration 3 exceeds 6 mol/l

  11. Establishing the traceability of a uranyl nitrate solution to a standard reference material

    International Nuclear Information System (INIS)

    Jackson, C.H.; Clark, J.P.

    1978-01-01

    A uranyl nitrate solution for use as a Working Calibration and Test Material (WCTM) was characterized, using a statistically designed procedure to document traceability to National Bureau of Standards Reference Material (SPM-960). A Reference Calibration and Test Material (PCTM) was prepared from SRM-960 uranium metal to approximate the acid and uranium concentration of the WCTM. This solution was used in the characterization procedure. Details of preparing, handling, and packaging these solutions are covered. Two outside laboratories, each having measurement expertise using a different analytical method, were selected to measure both solutions according to the procedure for characterizing the WCTM. Two different methods were also used for the in-house characterization work. All analytical results were tested for statistical agreement before the WCTM concentration and limit of error values were calculated. A concentration value was determined with a relative limit of error (RLE) of approximately 0.03% which was better than the target RLE of 0.08%. The use of this working material eliminates the expense of using SRMs to fulfill traceability requirements for uranium measurements on this type material. Several years' supply of uranyl nitrate solution with NBS traceability was produced. The cost of this material was less than 10% of an equal quantity of SRM-960 uranium metal

  12. Determination of halogens, silicon, phosphorus, carbon, sulfur, tributyl phosphate and of free acid in uranyl nitrate solutions

    International Nuclear Information System (INIS)

    Chu Van Vinh

    2003-01-01

    High-purity uranium compounds are widely used in nuclear field in the form of uranyl nitrate or uranium oxides. In production of uranium material the estimation and the control of products quality is necessary and very important. Halogens was separated from uranium compounds by steam distillation and they were later determined by high performance liquid chromatography (HPLC) for Cl - , Br - , I - ions. Br - was also determined by spectrophotometric and iodide by the individual pulse polarography. Silicon and phosphorus in uranyl nitrate solutions were determined by the photometric method. Sulfur was determined as sulfate form by the measurement of turbidity by the titrimetry. TBP in kerosene and free acid in aqueous solution were determined by the titration. (author)

  13. Determination of free nitric acid in uranyl nitrate solution

    International Nuclear Information System (INIS)

    Mayankutty, P.C.; Ravi, S.; Nadkarni, M.N.

    1981-01-01

    Potentiometric titration of uranyl nitrate solution with sodium hydroxide exhibits two peaks. The first peak characterises the following reaction, UO 2 (C 2 O 4 )+NaOH Na[UO 2 (C 2 O 4 )(OH)]. This reaction, indicating the partial hydrolysis of uranyl oxalate complex, appears to be complete at pH9. If the titration is carried out to this end-point pH, the total alkali consumed can be equated to the sum of uranium content and the free acidity present in the sample volume. Based on this, a method was standardised to determine the free acidity in uranyl nitrate solution. The sample, taken in a solution of potassium oxalate previously adjusted to pH9, is titrated to this pH with standard sodium hydroxide. The free acidity in the sample can be computed by subtracting the alkali reacted with uranium from the total alkali consumed. Analyses of several synthetic samples containing uranium and nitric acid in a wide range of combinations indicate that the free acidity can be accurately determined by this method, if uranium concentration in the sample is known. The results are compared to those obtained by two other widely used methods, viz., (i) titration of pH7 in the presence of neutral potassium oxalate to suppress hydrolysis and (ii) separation of hydrolyzable ions on a cationic resin and alkali titration of the free acid released. The advantages of and the precision obtained with the present method over the above two methods are discussed. (author)

  14. Effect of nitrate on corrosion of austenitic stainless steel in boiling nitric acid solution containing chromium ions

    International Nuclear Information System (INIS)

    Hasegawa, Satoshi; Kim, Seong-Yun; Ebina, Tetsunari; Ito, Tatsuya; Nagano, Nobumichi; Hitomi, Keitaro; Ishii, Keizo; Tokuda, Haruaki

    2016-01-01

    The oxidation behavior of chromium and the corrosion behavior of austenitic stainless steel in boiling nitric acid solution containing highly concentrated nitrates were investigated using UV-visible spectroscopic measurements, Raman spectral measurements, immersion tests, and potentiodynamic polarization measurements. The oxidation rate measurement of chromium from Cr(III) to Cr(VI) was performed by 1 M boiling nitric acid solution containing each highly concentrated nitrates: Al(NO_3)_3, Nd(NO_3)_3, Ca(NO_3)_2, Mg(NO_3)_2, and NaNO_3 as a simulant of uranium nitrate in uranium concentrator in reprocessing plants. As a result, the rate of chromium oxidation was different depending on the added nitrates even at the same nitric acid concentration. In addition, the oxidation rate of chromium was increased with increasing the calculated partial pressure of nitric acid in consideration of the hydration of cation of nitrates. Furthermore, the corrosion rate of type 310 stainless steel was accelerated by the solution having a high chromium oxidation rate containing nitrates. These results indicated that the acceleration of the corrosion rate in the solutions depending on the oxidation rate of chromium, and the rate is affected by the salt-effect of nitrates. (author)

  15. SEPARATION OF BARIUM VALUES FROM URANYL NITRATE SOLUTIONS

    Science.gov (United States)

    Tompkins, E.R.

    1959-02-24

    The separation of radioactive barium values from a uranyl nitrate solution of neutron-irradiated uranium is described. The 10 to 20% uranyl nitrate solution is passed through a flrst column of a cation exchange resin under conditions favoring the adsorption of barium and certain other cations. The loaded resin is first washed with dilute sulfuric acid to remove a portion of the other cations, and then wash with a citric acid solution at pH of 5 to 7 to recover the barium along with a lesser amount of the other cations. The PH of the resulting eluate is adjusted to about 2.3 to 3.5 and diluted prior to passing through a smaller second column of exchange resin. The loaded resin is first washed with a citric acid solution at a pH of 3 to elute undesired cations and then with citric acid solution at a pH of 6 to eluts the barium, which is substantially free of undesired cations.

  16. HTGR fuel development: loading of uranium on carboxylic acid cation-exchange resins using solvent extraction of nitrate

    International Nuclear Information System (INIS)

    Haas, P.A.

    1975-09-01

    The reference fuel kernel for recycle of 233 U to HTGR's (High-Temperature Gas-Cooled Reactors) is prepared by loading carboxylic acid cation-exchange resins with uranium and carbonizing at controlled conditions. The purified 233 UO 2 (NO 3 ) 2 solution from a fuel reprocessing plant contains excess HNO 3 (NO 3 - /U ratio of approximately 2.2). The reference flowsheet for a 233 U recycle fuel facility at Oak Ridge uses solvent extraction of nitrate by a 0.3 M secondary amine in a hydrocarbon diluent to prepare acid-deficient uranyl nitrate. This nitrate extraction, along with resin loading and amine regeneration steps, was demonstrated in 14 runs. No significant operating difficulties were encountered. The process is controlled via in-line pH measurements for the acid-deficient uranyl nitrate solutions. Information was developed on pH values for uranyl nitrate solution vs NO 3 - /U mole ratios, resin loading kinetics, resin drying requirements, and other resin loading process parameters. Calculations made to estimate the capacities of equipment that is geometrically safe with respect to control of nuclear criticality indicate 100 kg/day or more of uranium for single nitrate extraction lines with one continuous resin loading contactor or four batch loading contactors. (auth)

  17. 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

  18. Uranium removal from organic solutions of PUREX process

    International Nuclear Information System (INIS)

    Dell'Occhio, L.A.; Dupetit, G.A.; Pascale, A.A.; Vicens, H.E.

    1987-01-01

    During the uranium extraction process with tributyl phosphate (TBP) in nitric medium, a bi solvated, non hydrated complex is formed, of formula UO2(NO3)2TBP, which is soluble in the diluent, a paraffin hydrocarbon. As it is known that some uranium salts, for instance the nitrate, when dissolved in organic solvents, like isopropanol, can be discharged as complex molecules at the cathode of an electrodeposition cell, it was decided to apply this technique to uranium loaded TBP solutions. From preliminary experiments resulted a practical possibility for the analytical control through the alpha measurement of electro deposits. This technique could be applied as well to the treatment of depleted organic streams carrying undesirable alpha activity, because the so treated solutions become deprived of uranium. This work presents the curves obtained working at constant voltage with uranium-loaded TBP solutions, the determination of the optimal operation voltage in these conditions, the electrodeposition yield for electro polished copper and stainless steel cathodes and the tests of reproducibility of deposits. A summary of the results obtained operating the high voltage supply at constant power is also presented. (Author)

  19. Contribution to the characterization of the ideality deviation of concentrated solutions of electrolytes: application to the case plutonium and uranium (IV) nitrates

    International Nuclear Information System (INIS)

    Charrin, N.

    1999-01-01

    The purpose of this work is to establish a base of binary data referring to the plutonium and uranium nitrates (IV) activity coefficients, which will permit to take account the medium effects in the process of liquid-liquid extraction set in action during the reprocessing of irradiated combustibles in a more scrupulous way. The first chapter sticks to establish the problematic of acquisition of actinides binary data at an oxidation state (IV) linked to two characteristics of this type of electrolyte its radioactive properties and its chemical properties. Its chemical properties bring us to define the fictitious binary data and to use an approach based on the thermodynamic concept of simple solutions, on the measurements of water activity of ternary or quaternary mixtures of the actinide, in nitric acid medium and on the binary data of nitric acid. The second chapter intended to propose reliable binary data concerning nitric acid. The validation of acquisition of fictitious binary data method suggested is undertaken. The electrolyte test is the thorium nitrate (IV). The very encouraging results has determined the carrying out of this work of research in that way. The third chapter is based on the experimental acquisition of uranium and plutonium nitrates (IV) binary data. It emphasises the importance given to the preparation of the studied mixtures which characteristics, very high actinide concentrations and low acidities, make them atypical solutions and without any referenced equivalents. The last chapter describes the exploitation which was made of the established binary data. The characteristic parameters of Pu(NO 3 ) 4 and U(NO 3 ) 4 of Pitzer model and of the specific interaction theory has been appraised. Then the application of' the concept of simple solutions to the calculation of the density or quaternary mixtures like Pu(NO 3 ) 4 / UO 2 (NO 3 ) 2 /HNO 3 / H 2 O was proposed. (author)

  20. Nitrification and in-situ uranium solution mining

    International Nuclear Information System (INIS)

    Johnson, D.; Humenick, M.J.

    1980-01-01

    The objective of this research was to determine the potential for conversion of ammonia to nitrate as a result of uranium solution mining operations. The work included literature evaluation and laboratory experimentation in both batch and continuous systems. Results indicate that a potential for nitrification could exist for some portions of the solution mining operating cycle. However, inhibition of nitrification was observed due to high ammonia and peroxide concentrations. Nitrification of ammonia also was observed to occur due to chemical oxidation by peroxide. 28 refs

  1. Development of a chronocoulometric method for uranium traces determination with basis on nitrate catalytic reduction

    International Nuclear Information System (INIS)

    Cantagallo, M.I.C.; Gutz, I.G.R.

    1990-01-01

    The application of chronocoulometric technique with catalytic reduction of uranium/nitrate with catalytic reduction of uranium/nitrate system is described to give a detection limits on the sub-nanomolar region of uranium. (author)

  2. In situ mobility of uranium in the presence of nitrate following sulfate-reducing conditions.

    Science.gov (United States)

    Paradis, Charles J; Jagadamma, Sindhu; Watson, David B; McKay, Larry D; Hazen, Terry C; Park, Melora; Istok, Jonathan D

    2016-04-01

    Reoxidation and mobilization of previously reduced and immobilized uranium by dissolved-phase oxidants poses a significant challenge for remediating uranium-contaminated groundwater. Preferential oxidation of reduced sulfur-bearing species, as opposed to reduced uranium-bearing species, has been demonstrated to limit the mobility of uranium at the laboratory scale yet field-scale investigations are lacking. In this study, the mobility of uranium in the presence of nitrate oxidant was investigated in a shallow groundwater system after establishing conditions conducive to uranium reduction and the formation of reduced sulfur-bearing species. A series of three injections of groundwater (200 L) containing U(VI) (5 μM) and amended with ethanol (40 mM) and sulfate (20 mM) were conducted in ten test wells in order to stimulate microbial-mediated reduction of uranium and the formation of reduced sulfur-bearing species. Simultaneous push-pull tests were then conducted in triplicate well clusters to investigate the mobility of U(VI) under three conditions: 1) high nitrate (120 mM), 2) high nitrate (120 mM) with ethanol (30 mM), and 3) low nitrate (2 mM) with ethanol (30 mM). Dilution-adjusted breakthrough curves of ethanol, nitrate, nitrite, sulfate, and U(VI) suggested that nitrate reduction was predominantly coupled to the oxidation of reduced-sulfur bearing species, as opposed to the reoxidation of U(IV), under all three conditions for the duration of the 36-day tests. The amount of sulfate, but not U(VI), recovered during the push-pull tests was substantially more than injected, relative to bromide tracer, under all three conditions and further suggested that reduced sulfur-bearing species were preferentially oxidized under nitrate-reducing conditions. However, some reoxidation of U(IV) was observed under nitrate-reducing conditions and in the absence of detectable nitrate and/or nitrite. This suggested that reduced sulfur-bearing species may not be fully effective at

  3. Preparation of uranium standard solutions for x-ray fluorescence analysis

    International Nuclear Information System (INIS)

    Wong, C.M.; Cate, J.L.; Pickles, W.L.

    1978-03-01

    A method has been developed for gravimetrically preparing uranium nitrate standards with an estimated mean error of 0.1% (1 sigma) and a maximum error of 0.2% (1 sigma) for the total uranium weight. Two source materials, depleted uranium dioxide powder and NBS Standard Reference Material 960 uranium metal, were used to prepare stock solutions. The NBS metal proved to be superior because of the small but inherent uncertainty in the stoichiometry of the uranium oxide. These solutions were used to prepare standards in a freeze-dried configuration suitable for x-ray fluorescence analysis. Both gravimetric and freeze-drying techniques are presented. Volumetric preparation was found to be unsatisfactory for 0.1% precision for the sample size of interest. One of the primary considerations in preparing uranium standards for x-ray fluorescence analysis is the development of a technique for dispensing a 50-μl aliquot of a standard solution with a precision of 0.1% and an accuracy of 0.1%. The method developed corrects for variation in aliquoting and for evaporation loss during weighing. Two sets, each containing 50 standards have been produced. One set has been retained by LLL and one set retained by the Savannah River project

  4. Influence of Acidity on Uranyl Nitrate Association in Aqueous Solutions: A Molecular Dynamics Simulation Study

    International Nuclear Information System (INIS)

    de Almeida, Valmor F.; Cui, Shengting; Khomami, Bamin; Ye, Xianggui; Smith, Rodney Bryan

    2010-01-01

    Uranyl ion complexation with water and nitrate is a key aspect of the uranium/plutonium extraction process. We have carried out a molecular dynamics simulation study to investigate this complexation process, including the molecular composition of the various complex species, the corresponding structure, and the equilibrium distribution of the complexes. The observed structures of the complexes suggest that in aqueous solution, uranyls are generally hydrated by 5 water molecules in the equatorial plane. When associating with nitrate ions, a water molecule is replaced by a nitrate ion, preserving the five-fold coordination and planar symmetry. Analysis of the pair correlation function between uranyl and nitrate suggests that nitrates bind to uranyl in aqueous solution mainly in a monodentate mode, although a small portion of bidentates occur. Dynamic association and dissociation between uranyls and nitrates take place in aqueous solution with a substantial amount of fluctuation in the number of various uranyl nitrate species. The average number of the uranyl mononitrate complexes shows a dependence on acid concentration consistent with equilibrium-constant analysis, namely, the concentration of [UO2NO3]+ increases with nitric acid concentration.

  5. Molybdenum solubility in aluminium nitrate solutions

    Energy Technology Data Exchange (ETDEWEB)

    Heres, X.; Sans, D.; Bertrand, M.; Eysseric, C. [CEA, Centre de Marcoule, Nuclear Energy Division, DRCP, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France); Brackx, E.; Domenger, R.; Excoffier, E. [CEA, Centre de Marcoule, Nuclear Energy Division, DTEC, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France); Valery, J.F. [AREVA-NC, DOR/RDP, Paris - La Defense (France)

    2016-07-01

    For over 60 years, research reactors (RR or RTR for research testing reactors) have been used as neutron sources for research, radioisotope production ({sup 99}Mo/{sup 99m}Tc), nuclear medicine, materials characterization, etc... Currently, over 240 of these reactors are in operation in 56 countries. They are simpler than power reactors and operate at lower temperature (cooled to below 100 C. degrees). The fuel assemblies are typically plates or cylinders of uranium alloy and aluminium (U-Al) coated with pure aluminium. These fuels can be processed in AREVA La Hague plant after batch dissolution in concentrated nitric acid and mixing with UOX fuel streams. The aim of this study is to accurately measure the solubility of molybdenum in nitric acid solution containing high concentrations of aluminium. The higher the molybdenum solubility is, the more flexible reprocessing operations are, especially when the spent fuels contain high amounts of molybdenum. To be most representative of the dissolution process, uranium-molybdenum alloy and molybdenum metal powder were dissolved in solutions of aluminium nitrate at the nominal dissolution temperature. The experiments showed complete dissolution of metallic elements after 30 minutes long stirring, even if molybdenum metal was added in excess. After an induction period, a slow precipitation of molybdic acid occurs for about 15 hours. The data obtained show the molybdenum solubility decreases with increasing aluminium concentration. The solubility law follows an exponential relation around 40 g/L of aluminium with a high determination coefficient. Molybdenum solubility is not impacted by the presence of gadolinium, or by an increasing concentration of uranium. (authors)

  6. Determination of nitrate in effluents from Uranium Extraction Plant

    International Nuclear Information System (INIS)

    Dudwadkar, Ayushi; Kumar, Sangita D.; Reddy, A.V.R.

    2014-01-01

    Determination of nitrate concentration in the effluent samples from Uranium Extraction Plant is required before its safe discharge. As the different streams are diluted with sea water these samples contain high concentration of chloride. The large concentration of chloride poses a challenge in the determination of nitrate; hence, matrix elimination is accomplished by adopting a sample pretreatment technique. The present study was carried out to develop a simple, accurate and rapid analytical methodology for the determination of nitrate in the above matrices. The quantitative determination of nitrate was accomplished using anion exchange chromatography with conductometric detection. (author)

  7. Preparation of acid deficient solutions of uranyl nitrate and thorium nitrate by steam denitration

    International Nuclear Information System (INIS)

    Yamagishi, Shigeru; Takahashi, Yoshihisa

    1996-01-01

    Acid deficient heavy metal (HM) nitrate solutions are often required in the internal gelation processes for nuclear fuel fabrication. The stoichiometric HM-nitrate solutions are needed in a sol-gel process for fuel fabrication. A method for preparing such nitrate solutions with a controlled molar ratio of nitrate/metal by denitration of acid-excess nitrate solutions was developed. The denitration was conducted by bubbling a nitrate solution with a mixture of steam+Ar. It was found that steam was more effective for the denitration than Ar. The acid deficient uranyl nitrate solution with nitrate/U=1.55 was yielded by steam bubbling, while not by only Ar bubbling. As for thorium nitrate, acid deficient solutions of nitrate/Th≥3.1 were obtained by steam bubbling. (author)

  8. Effects of nitrate on the stability of uranium in a bioreduced region of the subsurface

    International Nuclear Information System (INIS)

    Wu, Weimin; Carley, Jack M.; Green, Stefan; Luo, Jian; Kelly, Shelly D.; Van Nostrand, Joy; Lowe, Kenneth Alan; Mehlhorn, Tonia L.; Carroll, Sue L.; Boonchayanant, Benjaporn; Loeffler, Frank E.; Jardine, Philip M.; Criddle, Craig

    2010-01-01

    The effects of nitrate on the stability of reduced, immobilized uranium were evaluated in field experiments at a U.S. Department of Energy site in Oak Ridge, TN. Nitrate (2.0 mM) was injected into a reduced region of the subsurface containing high levels of previously immobilized U(IV). The nitrate was reduced to nitrite, ammonium, and nitrogen gas; sulfide levels decreased; and Fe(II) levels increased then deceased. Uranium remobilization occurred concomitant with nitrite formation, suggesting nitrate-dependent, iron-accelerated oxidation of U(IV). Bromide tracer results indicated changes in subsurface flowpaths likely due to gas formation and/or precipitate. Desorption-adsorption of uranium by the iron-rich sediment impacted uranium mobilization and sequestration. After rereduction of the subsurface through ethanol additions, background groundwater containing high levels of nitrate was allowed to enter the reduced test zone. Aqueous uranium concentrations increased then decreased. Clone library analyses of sediment samples revealed the presence of denitrifying bacteria that can oxidize elemental sulfur, H 2 S, Fe(II), and U(IV) (e.g., Thiobacillus spp.), and a decrease in relative abundance of bacteria that can reduce Fe(III) and sulfate. XANES analyses of sediment samples confirmed changes in uranium oxidation state. Addition of ethanol restored reduced conditions and triggered a short-term increase in Fe(II) and aqueous uranium, likely due to reductive dissolution of Fe(III) oxides and release of sorbed U(VI). After two months of intermittent ethanol addition, sulfide levels increased, and aqueous uranium concentrations gradually decreased to <0.1 μM.

  9. Effect of pH and uranium concentration on interaction of uranium(VI) and uranium(IV) with organic ligands in aqueous solutions

    International Nuclear Information System (INIS)

    Li, W.C.; Victor, D.M.; Chakrabarti, C.L.

    1980-01-01

    The effect of pH and uranium concentration on the interactions of uranium(VI) and uranium(IV) with organic ligands was studied by employing dialysis and ultrafiltration techniques. The interactions of U(VI) and U(IV) with organic ligands in nitrate or chloride aqueous solution have been found to be pH-dependent. The stability constants of uranium-organic complexes decrease in the order: fulvic acid>humic acid>tannic acid for U(VI) and humic acid>tannic acid>fulvic acid for U(IV). Scatchard plots for the uranium-organic acid systems indicate two types of binding sites with a difference in stability constants of about 10 2 . Ultrafiltration of uranium-humic acid complexes indicates that U(VI) and U(IV) ions are concentrated in larger molecular size fractions (>5.1 nm) at pH less than or equal to 3 and in smaller molecular size fractions (in the range 5.1 to 3.1 nm and 2.4 to 1.9 nm) at pH greater than or equal to 5. 7 figures, 4 tables

  10. Studies on supercritical fluid extraction behaviour of uranium and thorium nitrates using amides

    International Nuclear Information System (INIS)

    Sujatha, K.; Kumar, R.; Sivaraman, N.; Srinivasan, T.G.; Vasudeva Rao, P.R.

    2007-01-01

    Supercritical fluid extraction studies of uranyl nitrate and thorium nitrate in mixture were carried out using various amides such as N,N-di(2-ethylhexyl) isobutyramide (D2EHIBA),N,N-dihexyl octanamide (DHOA) and Diisooctyl Butanamide (DiOBA). These studies established a preferential extraction of uranium over thorium. Among the various amides studied, D2EHIBA offered the best rate of preferential extraction of uranium over thorium. (author)

  11. Effect of Co-Contaminants Uranium and Nitrate on Iodine Remediation

    Energy Technology Data Exchange (ETDEWEB)

    Szecsody, James E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lee, Brady D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lawter, Amanda R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Qafoku, Nikolla [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Resch, Charles T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Baum, Steven R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Leavy, Ian I. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Freedman, Vicky L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2017-09-01

    The objective of this study is to evaluate the significance of co-contaminants on the migration and transformation of iodine species in the Hanford subsurface environment. These impacts are relevant because remedies that target individual contaminants like iodine, may not only impact the fate and transport of other contaminants in the subsurface, but also inhibit the effectiveness of a targeted remedy. For example, iodine (as iodate) co-precipitates with calcite, and has been identified as a potential remedy because it immobilizes iodine. Since uranium also co-precipitates with calcite in field sediments, the presence of uranium may also inhibit iodine co-precipitation. Another potentially significant impact from co-existing contaminants is iodine and nitrate. The presence of nitrate has been shown to promote biogeochemical reduction of iodate to iodide, thereby increasing iodine species subsurface mobility (as iodide exhibits less sorption). Hence, this study reports on both laboratory batch and column experiments that investigated a) the change in iodate uptake mass and rate of uptake into precipitating calcite due to the presence of differing amounts of uranium, b) the amount of change of the iodate bio-reduction rate due to the presence of differing nitrate concentrations, and c) whether nitrite can reduce iodate in the presence of microbes and/or minerals acting as catalysts.

  12. A bibliographical review on the radiolysis of uranyl nitrate solutions in nitric acid medium

    International Nuclear Information System (INIS)

    Siri, Sandra; Mondino, Angel V.

    2004-01-01

    A bibliographical study on the effects of ionizing radiation on uranyl nitrate solutions in nitric acid medium was performed, and the state of knowledge on this subject is presented. The main experimental and theoretical results on water, nitric acid and uranium solutions radiolysis are reviewed and critically evaluated. This paper provides a collection of references as an aid to the development of practical applications, and to stimulate new research on fundamental processes in these systems. (author) [es

  13. Role of ion transfer membrane in the production of uranous nitrate

    International Nuclear Information System (INIS)

    Nair, M.K.T.; Singh, R.K.; Bajpai, D.D.; Venugopalan, A.K.; Singh, R.R.; Gurba, P.B.; Thomas, Mathew

    1992-01-01

    In Purex process, plutonium and uranium are co-extracted into organic phase and these are partitioned by reducing Pu(IV) to Pu(III) using hydrazine stabilized uranous nitrate solution. Usually, uranous nitrate is added in much higher quantity than the stoichiometric requirement to effect complete reduction of plutonium. In conventional electrolytic cells only 60 to 70% of uranyl to uranous conversion is achieved. Use of this solution results in dilution of plutonium product. In addition to this, each externally fed uranous nitrate batch increases uranium processing load and affects the plant throughput. In order to keep the additional uranium processing load to a minimum, it is necessary to increase the uranous content to near cent percent level in the externally fed uranous nitrate solution. The studies carried out at PREFRE (Power Reactor Fuel Reprocessing) laboratory have shown that it is possible to produce concentrated uranous nitrate solution, nearly free from uranyl nitrate, by using a cation exchange membrane. This paper describes the development work carried out at PREFRE plant, Tarapur for production of cent percent uranous nitrate solution. Development of electrolytic cells for uranous production, from laboratory scale to pilot plant scale, has been explained. (author). 24 refs., 8 figs., 8 tabs

  14. Catalyzed reduction of nitrate in aqueous solutions

    International Nuclear Information System (INIS)

    Haas, P.A.

    1994-08-01

    Sodium nitrate and other nitrate salts in wastes is a major source of difficulty for permanent disposal. Reduction of nitrate using aluminum metal has been demonstrated, but NH 3 , hydrazine, or organic compounds containing oxygen would be advantageous for reduction of nitrate in sodium nitrate solutions. Objective of this seed money study was to determine minimum conditions for reduction. Proposed procedure was batchwise heating of aqueous solutions in closed vessels with monitoring of temperatures and pressures. A simple, convenient apparatus and procedure were demonstrated for observing formation of gaseous products and collecting samples for analyses. The test conditions were 250 degree C and 1000 psi max. Any useful reduction of sodium nitrate to sodium hydroxide as the primary product was not found. The nitrate present at pHs 3 or NH 4 NO 3 is easily decomposed, and the effect of nitromethane at these low pHs was confirmed. When acetic acid or formic acid was added, 21 to 56% of the nitrate in sodium nitrate solutions was reduced by methanol or formaldehyde. With hydrazine and acetic acid, 73 % of the nitrate was decomposed to convert NaNO 3 to sodium acetate. With hydrazine and formic acid, 36% of the nitrate was decomposed. If these products are more acceptable for final disposal than sodium nitrate, the reagents are cheap and the conversion conditions would be practical for easy use. Ammonium acetate or formate salts did not significantly reduce nitrate in sodium nitrate solutions

  15. Sub-critical pulsed neutron experiments with uranyl nitrate solutions in spherical geometry

    International Nuclear Information System (INIS)

    Gurin, Victor N.; Ryazanov, Boris G.; Sviridov, Victor I.; Volnistov, Vladimir V.

    2003-01-01

    The pulse source method is used to study homogeneous solution assemblies. Three sets of sub-critical pulse experiments with spherical tanks filled with water solution of uranyl nitrate (90% enrichment) were carried out at the RF-GS facility, Obninsk, Russia. Seven spherical tanks with the volume within the range of 1.29 L to 19.8 L were used in the experiments. Three uranium concentrations were studied, i.e. 20.7, 29.6 and 37.5 g/L. The sub-critical experiments were analyzed with the MCNP 4A code based on the Monte-Carlo method, and with ENDF/B-V library. (author)

  16. Denitrification of nitrate waste solutions

    International Nuclear Information System (INIS)

    Michaels, S.L.; Michel, R.C.; Terpandjian, P.D.; Vora, J.N.

    1976-01-01

    Bacterial denitrification by Pseudomonas Stutzeri has been chosen as the method for removing nitrate from the effluent stream of the Y-12 uranium purification process. A model was developed to predict bacterial growth and carbon and nitrate depletion during the induction period and steady state operation. Modification of analytical procedures and automatic control of the pH in the reactor are recommended to improve agreement between the prediction of the model and experimental data. An initial carbon-to-nitrogen (C/N) mass ratio of 1.4-1.5 insures adequate population growth during the induction period. Further experiments in batch reactors and in steady state flow reactors are recommended to obtain more reliable kinetic rate constants

  17. Contribution to the study of uranium migration and some trace elements in solution from Pocos de Caldas uranium mining

    International Nuclear Information System (INIS)

    Zenaro, R.

    1989-01-01

    It was studied the chemical composition of ground water from four boreholes as a contribution to the hydrogeochemical studies in the Pocos de Caldas uranium mining. Methods for water analyses were selected and optimized in order to determine the main anions, specially the ones which form stable complexes with uranium ions. Fluoride and chloride were determined by potentiometry; phosphate, nitrate and silicate by spectrophotometry. Cations were determined by atomic absorption spectrophotometry flame emission and argon plasma emission excited by continuous current arch (DCP). Uranium was determined by fluorimetry with a concentration range from 3 to 7 ppb and its distribution calculated among the different species into solution through the measures of pH, Eh, anion amounts and stability of their respective complexes. (author) [pt

  18. 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

  19. Radiation-induced nitration of organic compounds in aqueous solutions

    International Nuclear Information System (INIS)

    Ershov, B.G.; Gordeev, A.V.; Bykov, G.L.; Moisy, P.

    2012-01-01

    Radiolysis in aqueous nitrate and acetic acid solutions and nitrate/nitric acid and phenol was studied. The radiolysis of these solutes occurs with circle NO 2 radical, which is the active nitrating agent. Accumulation of nitromethane and nitrite was determined during γ-irradiation of aqueous solutions containing acetic and nitrate solutions. Irradiation of aqueous phenol-nitrate/nitric acid solutions results in the formation of 2- and 4-nitrophenols.

  20. Radiation-induced nitration of organic compounds in aqueous solutions

    Energy Technology Data Exchange (ETDEWEB)

    Ershov, B.G.; Gordeev, A.V.; Bykov, G.L. [Russian Academy of Sciences, Moscow (Russian Federation). Frumkin Inst. of Physical Chemistry and Electrochemistry; Moisy, P. [CEA, Bagnols sur Ceze (France). Nuclear Energy Div.

    2012-07-01

    Radiolysis in aqueous nitrate and acetic acid solutions and nitrate/nitric acid and phenol was studied. The radiolysis of these solutes occurs with {sup circle} NO{sup 2} radical, which is the active nitrating agent. Accumulation of nitromethane and nitrite was determined during {gamma}-irradiation of aqueous solutions containing acetic and nitrate solutions. Irradiation of aqueous phenol-nitrate/nitric acid solutions results in the formation of 2- and 4-nitrophenols.

  1. Characterization of the deviation of the ideality of concentrated electrolytic solutions: plutonium 4 and uranium 4 nitrate salts study; Contribution a la caracterisation de l'ecart a l'idealite des solutions concentrees d'electrolytes: application aux cas de nitrates de plutonium (4) et d'uranium (4)

    Energy Technology Data Exchange (ETDEWEB)

    Charrin, N

    2000-07-01

    The purpose of this work was to establish a new binary data base by compiling the activity coefficients of plutonium and uranium at oxidation state +IV to better account for media effects in the liquid-liquid extraction operations implemented to reprocess spent nuclear fuel. Chapter 1: first reviews the basic thermodynamic concepts before describing the issues involved in acquiring binary data for the tetravalent actinides. The difficulties arise from two characteristics of this type of electrolyte: its radioactive properties (high specific activity requiring nuclearization of the experimental instrumentation) and its physicochemical properties (strong hydrolysis). After defining the notion of fictive binary data, an approach based on the thermodynamic concept of simple solutions is described in which the activity coefficient of an aqueous phase constituent is dependent on two parameters: the water activity of the system and the total concentration of dissolved constituents. The method of acquiring fictive binary electrolyte data is based on water activity measurements for ternary or quaternary actinide mixtures in nitric acid media, and binary data for nitric acid. The experimental value is then correlated with the characteristics of the fictive binary solution of the relevant electrolyte. Chapter 2: proposes more reliable binary data for nitric acid than the published equivalents, the disparities of which are discussed. The validation of the method described in Chapter 1 for acquiring fictive binary data is then addressed. The test electrolyte, for which binary data are available in the literature, is thorium(IV) nitrate. The method is validated by comparing the published binary data obtained experimentally for binary solutions with the data determined for the ternary Th(NO{sub 3}){sub 4}/HNO{sub 3}/H{sub 2}O system investigated in this study. The very encouraging results of this comparison led us to undertake further research in this area. Chapter 3 discusses

  2. Development of processes for pilot plant production of purified uranyl nitrate solutions

    Energy Technology Data Exchange (ETDEWEB)

    Alfredson, P. G.; Charlton, B. G.; Ryan, R. K.; Vilkaitis, V. K.

    1975-01-15

    Nuclear purity uranyl nitrate solutions were produced from Rum Jungle yellow cake by dissolution in nitric acid and purification by solvent extraction with 20 vol. per cent tributyl phosphate in kerosene using pump - mix mixer-settler contactors. The design of the equipment, experimental studies and operating experience are described. Dissolution of yellow cake and recycled uranium oxide materials was readily carried out in a 100 l dissolver to give solutions containing 300 gU l{sup -1} and 0.5 to 4 M nitric acid. Filtration of silica from this solution prior to solvent extraction was not necessary in this work for yellow cake containing 0.25 per cent silica. A low acid flowsheet for uranium purification was developed in which the nitric acid consumption was reduced by 76 per cent and the throughput of the mixer-settler units was increased by 67 per cent compared with the initial design flowsheet. Nine extraction and seven scrubbing stages were used with a feed solution containing 300 gU l{sup -1} and 1.0 M nitric acid and with a portion of the product recycled as scrub solution. The loaded organic phase was stripped in 16 stages with 0.05 M nitric acid heated to 60 deg C to give a 120 gU l{sup -1} product. The uranium concentration in the raffinate was < 0.04 g l{sup -1}, corresponding to approximately 0.01 per cent of the feed. (author)

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

    International Nuclear Information System (INIS)

    Baran, V.; Moltasova, J.

    1982-01-01

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

  4. Investigation of disposal of nitrate-bearing effluent from in-situ leaching process by natural evaporation in Yining uranium mine

    International Nuclear Information System (INIS)

    Huang Chongyuan; Li Weicai; Zhang Yutai; Gao Xizhen

    2000-01-01

    Experiments indicated, after lime neutralization and precipitation of nitrate-bearing effluent from in-situ leaching process, uranium concentration increase with the increasing of nitrate concentration. Only when nitrate concentration is <0.5 mg/L, uranium concentration can drop from 1.5-2.0 mg/L to about 1.0 mg/L. The permeability coefficient of soil is about 1.0-1.1 m/d in the place which is scheduled for building natural evaporation pool. After lime neutralization of nitrate-bearing effluent, it can drop to 0.03-0.01 m/d. Setting up water-proof layer in natural evaporation pool can reduce pollution of underground water by uranium, nitrate and ammonium

  5. Neptunium separation in trace levels from uranium solutions by extraction chromatography

    International Nuclear Information System (INIS)

    Figols, M.E.B.

    1991-01-01

    Neptunium and uranium behavior in extraction chromatography system, aiming the separation of microquantities of neptunium from uranyl nitrate solutions is described. Tri-n-octylamina (TOA), tri-n-butylphosphate (TBP), thenoyltrifluoroacetone (TTA) as stationary phase, alumina, Voltalef-UF-300, silica as support material were verified. The impregnation conditions as well as the best stationary phase/support material ratio were established. TBP/alumina, TBP/Voltalef and TOA/alumina system were selected to uranium and neptunium separation studies. In the system using TBP as extractant agent uranium and neptunium separation was reached by selective elution after the retention of both elements on the column. U-Np separation by selective retention of Np was possible with TOA system. The capacity of the column was the 66.6 mg U/mL and 191.6mg U/mL for the TBP/alumina and TBP/Voltalef systems, respectively. An application of extraction chromatography system in the final phase of irradiated uranium treatment process is proposed. (author)

  6. Effects of solution pH and complexing reagents on uranium and thorium desorption under saturated equilibrium conditions

    International Nuclear Information System (INIS)

    Wang, Yug-Yea; Yu, C.

    1992-01-01

    Three contaminated bulk surface soils were used for investigating the effect of solution pH and complexing reagents on uranium and thorium desorption. At a low solution pH, the major chemical species of uranium and thorium, uranyl UO 2 +2 , thorium dihydroxide Th(OH) 2 +2 , and thorium hydroxide Th(OH) +3 , tend to form complexes with acetates in the solution phase, which increases the fractions of uranium and thorium desorbed into this phase. At a high solution pH, important uranium and thorium species such as uranyl tricarbonate complex UO 2 (CO) 33 -4 and thorium tetrahydroxide complex Th(OH) 4 tend to resist complexation with acetates. The presence of complexing reagents in solution can release radionuclides such as uranium and/or thorium from the soil to the solution by forming soluble complexes. Sodium bicarbonate (NaHCO 3 ) and diethylenetriaminepentaacetic acid (DTPA) are strong complex formers that released 38% to 62% of total uranium activity and 78% to 86% of total thorium activity, respectively, from the soil samples investigated. Solutions of 0.1 molar sodium nitrate (NaNO 3 ) and 0.1 molar sodium sulfate (Na 2 SO 4 ) were not effective complex formers with uranium and thorium under the experimental conditions. Fractions of uranium and thorium desorbed by 0.15g/200ml humic acid ranged from 4.62% to 6.17% and 1.59% to 7.09%, respectively. This work demonstrates the importance of a knowledge of solution chemistry in investigating the desorption of radionuclides

  7. Studies of the conversion-chemistry of plutonium and uranium in the nitrate- and carbonate-systems

    International Nuclear Information System (INIS)

    Hoffmann, G.; Steinhauser, M.; Boehm, M.

    1988-01-01

    A novel type construction of an autoclave for dissolving of plutonium dioxide in concentrated nitric acid (without any admixtures) has been developed. This process allows the dissolving of batches with high oxide/acid ratio and yields plutonium-solutions of high concentration. The tests for separation of plutonium- and, respectively, uranium-process-solutions from Am-241 and other interfering impurities are described. The time-factor for the oxidation-reaction of plutonium in nitric acid with ozone has been optimized. Important data on the solubility-behavior of plutonyl(VI)- and of pure Pu(IV)-nitrates have been gained. The majority of the precipitates, occuring in theses reactions, were characterized. (orig.) [de

  8. SEPARATION OF URANIUM, PLUTONIUM, AND FISSION PRODUCTS

    Science.gov (United States)

    Spence, R.; Lister, M.W.

    1958-12-16

    Uranium and plutonium can be separated from neutron-lrradiated uranium by a process consisting of dissolvlng the lrradiated material in nitric acid, saturating the solution with a nitrate salt such as ammonium nitrate, rendering the solution substantially neutral with a base such as ammonia, adding a reducing agent such as hydroxylamine to change plutonium to the trivalent state, treating the solution with a substantially water immiscible organic solvent such as dibutoxy diethylether to selectively extract the uranium, maklng the residual aqueous solutlon acid with nitric acid, adding an oxidizing agent such as ammonlum bromate to oxidize the plutonium to the hexavalent state, and selectlvely extracting the plutonium by means of an immlscible solvent, such as dibutoxy dlethyletber.

  9. Solvent extraction of uranium(VI) and thorium(IV) from nitrate media by carboxylic acid amides

    International Nuclear Information System (INIS)

    Preston, J.S.; Preez, A.C. du

    1995-01-01

    A series of nineteen N-alkyl carboxylic acid amides (R.CO.NHR') has been prepared, in which the alkyl groups R and R' have been varied in order to introduce different degrees of steric complexity into the compounds. A smaller number of N,N-dialkyl amides (R.CO.NR 2 ') and non-substituted amides (R.CO.NH 2 ) has also been prepared for comparison purposes. These amides were characterized by measurement of their boiling points, melting points, refractive indices and densities. The solvent extraction of uranium(VI) and thorium(IV) from sodium nitrate media by solutions of the amides in toluene was studied. Increasing steric bulk of the alkyl groups R and R' was found to cause a marked decrease in the extraction of thorium, with a much smaller effect on the extraction of uranium, thus considerably enhancing the separation between these metals. Vapour pressure osmometry studies indicate that the N-alkyl amides are self-associated in toluene solution, with aggregation numbers up to about 2.5 for 0.6 M solutions at 35 degree C. In contrast, the N,N-dialkyl amides behave as monomers under these conditions. The distribution ratios for the extraction of uranium and thorium show second- and third-order dependences, respectively, on the extractant concentration for both the N-alkyl and N,N-dialkyl amides. 15 refs., 8 figs., 8 tabs

  10. The role of Eh and pH in leaching Saskatchewan uranium ores with chloride and nitrate leaching systems

    International Nuclear Information System (INIS)

    Nirdosh, I.; Muthuswami, S.V.

    1992-01-01

    The effects of solution E h and pH on the extractions of U, 230 Th, 226 Ra, As and Ni from two typical uranium ores from the province of Saskatchewan in Canada are discussed for the leachants ferric chloride, ferric nitrate, nitric acid and hydrochloric acid. It is concluded that E h > 700 mV and pH 230 Th extraction is more sensitive to solution pH than to E h whereas Ni extraction is sensitive mainly to the solution E h . Arsenic extraction is very sensitive to solution E h , and for a given E h , is high at pH 1.3. (orig.) [de

  11. Spectrographic determination of impurities in enriched uranium solutions

    International Nuclear Information System (INIS)

    Capdevila, C.; Roca, M.

    1980-01-01

    A spectrographic procedure for the determination of trace amounts of Al, B, Ba, Be, Bi, Ca, Cd, Co, Cr, Cu, Fe, K, L i , Hg, Mn, Mo, Na, Nb, Ni, P, Pb, Ru, Sb, Sn, Sr, Ti, V, Zn, and Zr in enriched uranyl nitrate solutions from the reprocessing of spent nuclear fuels is described. After removal of uranium by either TBP or TNOA solvent extraction, the aqueous phase Is analysed by the graphite spark technique. TBP is adequate for all impurities, excepting boron and phosphorus; both of these elements can sat is factory be determined by using TNOA after the addition of mannitol to avoid boron losses. (Author) 4 refs

  12. PROCESS OF RECOVERING URANIUM FROM ITS ORES

    Science.gov (United States)

    Galvanek, P. Jr.

    1959-02-24

    A process is presented for recovering uranium from its ores. The crushed ore is mixed with 5 to 10% of sulfuric acid and added water to about 5 to 30% of the weight of the ore. This pugged material is cured for 2 to 3 hours at 100 to 110 deg C and then cooled. The cooled mass is nitrate-conditioned by mixing with a solution equivalent to 35 pounds of ammunium nitrate and 300 pounds of water per ton of ore. The resulting pulp containing 70% or more solids is treated by upflow percolation with a 5% solution of tributyl phosphate in kerosene at a rate equivalent to a residence time of about one hour to extract the solubilized uranium. The uranium is recovered from the pregnant organic liquid by counter-current washing with water. The organic extractant may be recycled. The uranium is removed from the water solution by treating with ammonia to precipitate ammonium diuranate. The filtrate from the last step may be recycled for the nitrate-conditioning treatment.

  13. Photochemical reduction of uranyl nitrate

    Energy Technology Data Exchange (ETDEWEB)

    Duerksen, W.K.

    1993-10-20

    The photochemical reduction of uranyl nitrate solutions to tetravalent uranium was investigated as a means of producing uranium dioxide feed for the saltless direct oxide reduction (SDOR) process. At high uranium concentrations, reoxidation of U{sup +4} occurs rapidly. The kinetics of the nitric oxidation of tetravalent uranium depend on the concentrations of hydrogen ion, nitrate ion, nitrous acid, and tetravalent uranium in the same manner as was reported elsewhere for the nitrate oxidation of PU{sup +3}. Reaction rate data were successfully correlated with a mechanism in which nitrogen dioxide is the reactive intermediate. Addition of a nitrous acid scavenger suppresses the reoxidation reaction. An immersion reactor employing a mercury vapor lamp gave reduction times fast enough for routine production usage. Precipitation techniques for conversion of aqueous U(NO{sub 3}){sub 4} to hydrous UO{sub 2} were evaluated. Prolonged dewatering times tended to make the process time consuming. Use of 3- to 4-M aqueous NaOH gave the best dewatering times observed. Reoxidation of the UO{sub 2} by water of hydration was encountered, which required the drying process to be carried out under a reducing atmosphere.

  14. Decomposition of metal nitrate solutions

    International Nuclear Information System (INIS)

    Haas, P.A.; Stines, W.B.

    1982-01-01

    Oxides in powder form are obtained from aqueous solutions of one or more heavy metal nitrates (e.g. U, Pu, Th, Ce) by thermal decomposition at 300 to 800 deg C in the presence of about 50 to 500% molar concentration of ammonium nitrate to total metal. (author)

  15. TREATMENT TESTS FOR EX SITU REMOVAL OF CHROMATE & NITRATE & URANIUM (VI) FROM HANFORD (100-HR-3) GROUNDWATER FINAL REPORT

    Energy Technology Data Exchange (ETDEWEB)

    BECK MA; DUNCAN JB

    1994-01-03

    This report describes batch and ion exchange column laboratory scale studies investigating ex situ methods to remove chromate (chromium [VI]), nitrate (NO{sub 3}{sup -}) and uranium (present as uranium [VI]) from contaminated Hanford site groundwaters. The technologies investigated include: chemical precipitation or coprecipitation to remove chromate and uranium; and anion exchange to remove chromate, uranium and nitrate. The technologies investigated were specified in the 100-HR-3 Groundwater Treatability Test Plan. The method suggested for future study is anion exchange.

  16. Vibrational spectra of monouranates and uranium hydroxides as reaction products of alkali with uranyl nitrate

    International Nuclear Information System (INIS)

    Komyak, A.I.; Umrejko, D.S.; Posledovich, M.R.

    2013-01-01

    Vibrational (IR absorption and Raman scattering) spectra for the reaction products of uranyl nitrate hexahydrate with NaOH and KOH have been studied. As a result of exchange reactions, the uranyl-ion coordinated nitrate groups are completely replaced by hydroxyl ions and various uranium and uranyl oxides or hydrates are formed. An analysis of the vibrations has been performed in terms of the frequencies of a free or coordinated nitrate group; comparison with the vibrations of the well-known uranium oxides and of the uranyl group UO 2 2+ has been carried out. Vibrational spectra of a free nitrate group are characterized by four vibrational frequencies 1050, 724, 850, and 1380 cm -1 , among which the frequencies at 724 and 1380 cm -1 are doubly degenerate and attributed to E’ symmetry of the point group D 3h . When this group is uranium coordinated, its symmetry level is lowered to C 2v , all vibrations of this group being active both in Raman and IR spectra. The doubly degenerate vibrations are exhibited as two bands and a frequency of the out-of-plane vibration is lowered to 815 cm -1 . (authors)

  17. Selective Removal of Uranium from the Washing Solution of Uranium-Contaminated Soil

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Soo; Han, G. S.; Kim, G. N.; Koo, D. S.; Jeong, J. W.; Choi, J. W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This study examined selective removal methods of uranium from the waste solution by ion exchange resins or solvent extraction methods to reduce amount of the 2{sup nd} waste. Alamine-336, known as an excellent extraction reagent of uranium from the leaching solution of uranium ore, did not remove uranium from the acidic washing solution of soil. Uranyl ions in the acidic waste solution were sorbed on ampholyte resin with a high sorption efficiency, and desorbed from the resin by a washing with 0.5 M Na{sub 2}CO{sub 3} solution at 60 .deg. C. However, the uranium dissolved in the sulfuric acid solution was not sorbed onto the strong anion exchanger resins. A great amount of uranium-contaminated (U-contaminated) soil had been generated from the decommissioning of a uranium conversion plant. Our group has developed a decontamination process with washing and electrokinetic methods to decrease the amount of waste to be disposed of. However, this process generates a large amount of waste solution containing various metal ions.

  18. 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)

  19. Precise coulometric titration of uranium in a high-purity uranium metal and in uranium compounds

    International Nuclear Information System (INIS)

    Tanaka, Tatsuhiko; Yoshimori, Takayoshi

    1975-01-01

    Uranium in uranyl nitrate, uranium trioxide and a high-purity uranium metal was assayed by the coulometric titration with biamperometric end-point detection. Uranium (VI) was reduced to uranium (IV) by solid bismuth amalgam in 5M sulfuric acid solution. The reduced uranium was reoxidized to uranium (VI) with a large excess of ferric ion at a room temperature, and the ferrous ion produced was titrated with the electrogenerated manganese(III) fluoride. In the analyses of uranium nitrate and uranium trioxide, the results were precise enough when the error from uncertainty in water content in the samples was considered. The standard sample of pure uranium metal (JAERI-U4) was assayed by the proposed method. The sample was cut into small chips of about 0.2g. Oxides on the metal surface were removed by the procedure shown by National Bureau of Standards just before weighing. The mean assay value of eleven determinations corrected for 3ppm of iron was (99.998+-0.012) % (the 95% confidence interval for the mean), with a standard deviation of 0.018%. The proposed coulometric method is simple and permits accurate and precise determination of uranium which is matrix constituent in a sample. (auth.)

  20. Studies on the reverse osmosis treatment of uranyl nitrate solution

    International Nuclear Information System (INIS)

    Prabhakar, S.; Panicker, S.T.; Misra, B.M.; Ramani, P.S.

    1992-01-01

    The aqueous effluent generated in uranium processing, particularly in the nuclear fuel fabrication step, contains mainly uranium nitrate. This requires treatment before discharge into the environment to meet stringent standards. This paper presents the performance of cellulose acetate membranes with regard to rejection of uranium under reverse osmotic conditions for feed concentrations up to 200 mg/l of uranium, which corresponds to the levels normally prevalent in the effluents. The use of additives like the disodium salt of ethylenediaminetetraacetic acid and sodium sulfate for the improvement of reverse osmosis performance of the above membranes was also investigated. In the light of the experimental results, the suitability of reverse osmosis for the decontamination of uranium effluents is discussed

  1. Development of a low cost, low environmental impact process for disposal of nitrate wastes

    International Nuclear Information System (INIS)

    Napier, J.M.

    1975-01-01

    A uranium recycle process in the Y-12 Plant generates nitrate ions which must be discarded. Scrap enriched uranium is dissolved in nitric acid and solvent extracted to remove impurities from the uranium. Aluminum nitrate is also used in the process to remove the purified uranium from the solvent extraction process. Dilute nitric acid, aluminum nitrate, and metallic impurities must be discarded from this process. A program was started to develop a low cost, low environmental impact process for disposal of these nitrate wastes. Several disposal methods were considered. A process was selected which included: distillation and recycle of nitric acid; crystallization and recycle of aluminum nitrate; and biodegradation of the remaining nitrate waste solutions. For this presentation, only the biodegradation process will be discussed. A colony of Pseudomonas stutzeri, which is capable of using the nitrate ion as the oxygen supply, was used. An excess of organic material was used to insure that the maximum amount of nitrate was destroyed

  2. Extraction behavior of uranium(VI) with polyurethane foam

    International Nuclear Information System (INIS)

    Tingchia Huang; Donghwang Chen; Muchang Shieh; Chingtsven Huang

    1992-01-01

    The extraction of uranium(VI) from aqueous solution with polyether-based polyurethane (PU) foam was studied. The effects of the kinds and concentrations of nitrate salts, uranium(VI) concentration, temperature, nitric acid concentration, pH, the content of poly(ethylene oxide) in the polyurethane foam, and the ratio of PU foam weight and solution volume on the extraction of uranium(VI) were investigated. The interferences of fluoride and carbonate ions on the extraction of uranium(VI) were also examined, and methods to overcome both interferences were suggested. It was found that no uranium was extracted in the absence of a nitrate salting-out agent, and the extraction behaviors of uranium(IV) with polyurethane foam could be explained in terms of an etherlike solvent extraction mechanism. In addition, the percentage extraction of a multiple stage was also estimated theoretically

  3. Extraction kinetics of uranium (VI) with polyurethane foam

    International Nuclear Information System (INIS)

    Huang, Ting-Chia; Chen, Dong-Hwang; Huang, Shius-Dong; Huang, Ching-Tsven; Shieh, Mu-Chang.

    1993-01-01

    The extraction kinetics of uranium(VI) from aqueous nitrate solution with polyether-based polyurethane foam was investigated in a batch reactor with automatic squeezing. The extraction curves of uranium(VI) concentration in solution vs. extraction time exhibited a rather rapid exponential decay within the first few minutes, followed by a slower exponential decay during the remaining period. This phenomenon can be attributed to the presence of two-phase structure, hard segment domains and soft segment matrix in the polyurethane foam. A two-stage rate model expressed by a superposition of two exponential curves was proposed, according to which the experimental data were fitted by an optimization method. The extraction rate of uranium (VI) was also found to increase with increasing temperature, nitrate concentration, and hydration of the cation of nitrate salt. (author)

  4. Experimental critical parameters of enriched uranium solution in annular tank geometries

    International Nuclear Information System (INIS)

    Rothe, R.E.

    1996-04-01

    A total of 61 critical configurations are reported for experiments involving various combinations of annular tanks into which enriched uranium solution was pumped. These experiments were performed at two widely separated times in the 1980s under two programs at the Rocky Flats Plant's Critical Mass Laboratory. The uranyl nitrate solution contained about 370 g of uranium per liter, but this concentration varied a little over the duration of the studies. The uranium was enriched to about 93% [sup 235]U. All tanks were typical of sizes commonly found in nuclear production plants. They were about 2 m tall and ranged in diameter from 0.6 m to 1.5 m. Annular thicknesses and conditions of neutron reflection, moderation, and absorption were such that criticality would be achieved with these dimensions. Only 13 of the entire set of 74 experiments proved to be subcritical when tanks were completely filled with solution. Single tanks of several radial thicknesses were studied as well as small line arrays (1 x 2 and 1 x 3) of annular tanks. Many systems were reflected on four sides and the bottom by concrete, but none were reflected from above. Many experiments also contained materials within and outside the annular regions that contained strong neutron absorbers. One program had such a thick external moderator/absorber combination that no reflector was used at all

  5. Experimental critical parameters of enriched uranium solution in annular tank geometries

    Energy Technology Data Exchange (ETDEWEB)

    Rothe, R.E.

    1996-04-01

    A total of 61 critical configurations are reported for experiments involving various combinations of annular tanks into which enriched uranium solution was pumped. These experiments were performed at two widely separated times in the 1980s under two programs at the Rocky Flats Plant`s Critical Mass Laboratory. The uranyl nitrate solution contained about 370 g of uranium per liter, but this concentration varied a little over the duration of the studies. The uranium was enriched to about 93% [sup 235]U. All tanks were typical of sizes commonly found in nuclear production plants. They were about 2 m tall and ranged in diameter from 0.6 m to 1.5 m. Annular thicknesses and conditions of neutron reflection, moderation, and absorption were such that criticality would be achieved with these dimensions. Only 13 of the entire set of 74 experiments proved to be subcritical when tanks were completely filled with solution. Single tanks of several radial thicknesses were studied as well as small line arrays (1 x 2 and 1 x 3) of annular tanks. Many systems were reflected on four sides and the bottom by concrete, but none were reflected from above. Many experiments also contained materials within and outside the annular regions that contained strong neutron absorbers. One program had such a thick external moderator/absorber combination that no reflector was used at all.

  6. EPR study of the production of OH radicals in aqueous solutions of uranium irradiated by ultraviolet light

    Directory of Open Access Journals (Sweden)

    MARKO DAKOVIĆ

    2009-06-01

    Full Text Available The aim of the study was to establish whether hydroxyl radicals (•OH were produced in UV-irradiated aqueous solutions of uranyl salts. The production of •OH was studied in uranyl acetate and nitrate solutions by an EPR spin trap method over a wide pH range, with variation of the uranium concentrations. The production of •OH in uranyl solutions irradiated with UV was unequivocally demonstrated for the first time using the EPR spin-trapping method. The production of •OH can be connected to speciation of uranium species in aqueous solutions, showing a complex dependence on the solution pH. When compared with the results of radiative de-excitation of excited uranyl (*UO22+ by the quenching of its fluorescence, the present results indicate that the generation of hydroxyl radicals plays a major role in the fluorescence decay of *UO22+. The role of the presence of carbonates and counter ions pertinent to environmental conditions in biological systems on the production of hydroxyl radicals was also assessed in an attempt to reveal the mechanism of *UO22+ de-excitation. Various mechanisms, including •OH production, are inferred but the main point is that the generation of •OH in uranium containing solutions must be considered when assessing uranium toxicity.

  7. The impact of ammonium and nitrate impurities on the formation of uranium oxides, in thecomposition range UO3-U3O8-z during thermal decomposition of ammonium uranates

    International Nuclear Information System (INIS)

    El-Mamoon Yahai, M.; El-Fekey, S.A.; Abd El-Razek, A.M.

    1996-01-01

    Ammonium uranates (AU) were precipitated from a nuclear-pure uranyl nitrate solution using ammonia liquor. Unwashed and washed uranate samples were heated at temperatures varying between 400 and 800 C and analysed thermally and by X-ray diffraction analysis. The results indicated that amorphous uranium trioxide (A-UO 3 ) is mainly formed in absence of carried ammonium and nitrate ions, whereas deamination of the retained ammonia leads to β-UO 3 formation. The retained ammonium ions in the AU structure increased with the pH of precipitation and these ions lead to formation of two polymorphs of uranium octoxides. (orig.)

  8. Critical experiments on low enriched uranyl nitrate solution with STACY

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori

    1996-01-01

    As the STACY started steady operations, systematic criticality data on low enriched uranyl nitrate solution system could be accumulated. Main experimental parameters for the cylindrical tank of 60 cm in diameter were uranium concentration and the reflector condition. Basic data on a simple geometry will be helpful for the validation of the standard criticality safety codes, and for evaluating the safety margin included in the criticality designs. Experiments on the reactivity effects of structural materials such as borated concrete and polyethylene are on schedule next year as the second series of experiments using 10 wt% enriched uranyl solution. Furthermore, neutron interacting experiments with two slab tanks will be performed to investigate the fundamental properties of neutron interaction effects between core tanks. These data will be useful for making more reasonable calculation models and for evaluating the safety margin in the criticality designs for the multiple unit system. (J.P.N.)

  9. EXTRACTION OF URANYL NITRATE FROM AQUEOUS SOLUTIONS

    Science.gov (United States)

    Furman, N.H.; Mundy, R.J.

    1957-12-10

    An improvement in the process is described for extracting aqueous uranyl nitrate solutions with an organic solvent such as ether. It has been found that the organic phase will extract a larger quantity of uranyl nitrate if the aqueous phase contains in addition to the uranyl nitrate, a quantity of some other soluble nitrate to act as a salting out agent. Mentioned as suitable are the nitrates of lithium, calcium, zinc, bivalent copper, and trivalent iron.

  10. Optimisation of parameters for co-precipitation of uranium and plutonium - results of simulation studies

    International Nuclear Information System (INIS)

    Pandey, N.K.; Velvandan, P.V.; Murugesan, S.; Ahmed, M.K.; Koganti, S.B.

    1999-01-01

    Preparation of plutonium oxide from plutonium nitrate solution generally proceeds via oxalate precipitation route. In a nuclear fuel reprocessing scheme this step succeeds the partitioning step (separation of uranium and plutonium). Results of present studies confirm that it is possible to avoid partitioning step and recover plutonium and uranium as co-precipitated product. This also helps in minimising the risk of proliferation of fissile material. In this procedure, the solubility of uranium oxalate in nitric acid is effectively used. Co-precipitation parameters are optimised with simulated solutions of uranium nitrate and thorium nitrate (in place of plutonium). On the basis of obtained results a reconversion flow-sheet is designed and reported here. (author)

  11. Calibration of uranium 232 solution

    International Nuclear Information System (INIS)

    Galan, M.P.; Acena, M.L.

    1988-01-01

    A method for acertainning the activity by alpha spectroscopy with semiconductor detectors, of a solution of Uranium-232 is presented. It consists of the comparison with a Uranium-233 solution activity previously measured in a gridded ionization chamber of 2 π geometry. The total measurement uncertainty is about +- 0,02. (Author)

  12. Lime, agent to uranium concentration; La chaux comme agent de concentration de l'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Mouret, P; Le Bris, J; Kremer, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Gautier, R [Etablissement Kuhlmann, Service d' Etudes et de Pilotages Industriels (France)

    1958-07-01

    Choice of the process according to health requirements. Description of the process: dissolution of uranium by sulfuric leaching of ores, precipitation of uranium by lime, re-dissolution of the concentrate with nitric ions, purification by T.B.P. finally resulting in pure uranyl nitrate solution containing 400 g/litre. (author)Fren. [French] Les raisons du choix du procede en fonction des imperatifs d'hygiene, sont exposees ainsi que le procede qui consiste en une dissolution de l'uranium des minerais par lixiviation sulfurique, precipitation de l'uranium par la chaux et redissolution du concentre en presence d'ions nitriques, purification par le T.B.P. et obtention d'un concentre final de nitrate d'uranyle pur a 400 g/litre. (auteur)

  13. Comparative study involving the uranium determination through catalytic reduction of nitrates and nitrides by using decoupled plasma nitridation (DPN)

    International Nuclear Information System (INIS)

    Aguiar, Marco Antonio Souza; Gutz, Ivano G. Rolf

    1999-01-01

    This paper reports a comparative study on the determination of uranium through the catalytic reduction of nitrate and nitride using the decoupled plasma nitridation. The uranyl ions are a good catalyst for the reduction of NO - 3 and NO - 2 ions on the surface of a hanging drop mercury electrode (HDME). The presence of NO - in a solution with p H = 3 presented a catalytic signal more intense than the signal obtained with NO - 3 (concentration ten times higher). A detection limit of 1x10 9 M was obtained using the technique of decoupled plasma nitridation (DPN), suggesting the development of a sensitive way for the determination of uranium in different matrixes

  14. Recovery of uranium from an irradiated solid target after removal of molybdenum-99 produced from the irradiated target

    Science.gov (United States)

    Reilly, Sean Douglas; May, Iain; Copping, Roy; Dale, Gregory Edward

    2017-10-17

    A process for minimizing waste and maximizing utilization of uranium involves recovering uranium from an irradiated solid target after separating the medical isotope product, molybdenum-99, produced from the irradiated target. The process includes irradiating a solid target comprising uranium to produce fission products comprising molybdenum-99, and thereafter dissolving the target and conditioning the solution to prepare an aqueous nitric acid solution containing irradiated uranium. The acidic solution is then contacted with a solid sorbent whereby molybdenum-99 remains adsorbed to the sorbent for subsequent recovery. The uranium passes through the sorbent. The concentrations of acid and uranium are then adjusted to concentrations suitable for crystallization of uranyl nitrate hydrates. After inducing the crystallization, the uranyl nitrate hydrates are separated from a supernatant. The process results in the purification of uranyl nitrate hydrates from fission products and other contaminants. The uranium is therefore available for reuse, storage, or disposal.

  15. The clearance of uranium after deposition of the nitrate and bicarbonate in different regions of the rat lung

    International Nuclear Information System (INIS)

    Ellender, M.

    1987-01-01

    This study investigated the tissue distribution and excretion of uranium after its deposition as either the nitrate or bicarbonate in the three regions of the respiratory system of the rat. Results confirm the recommendations of ICRP that uranyl nitrate and bicarbonate should be treated as class D compounds; but imply that some of the parameters used in the ICRP lung model are not applicable to soluble uranium compounds. (author)

  16. Complexes of pentavalent plutonium in lithium nitrate solutions

    International Nuclear Information System (INIS)

    Mekhail, F.M.; Zaki, M.R.

    1977-01-01

    Pu 0 2 ion can form nitrate complexes in concentrated solution of lithium nitrate of PH 3.5. Spectrophotometric and ion exchange studies revealed the existence of two complexes, presumably the mono-and the dinitro. The rate of adsorption of the dinitrato complex, formed in 4 to 6 M-lithium nitrate solutions, on De-Acidite FF has been investigated and suggested to be diffusion controlled. The adsorption isotherm found to obey satisfactorily Freundlich equation

  17. Uranium refining by solvent extraction

    International Nuclear Information System (INIS)

    Kraikaew, J.; Srinuttrakul, W.

    2014-01-01

    The solvent extraction process to produce higher purity uranium from yellowcake was studied in laboratory scale. Yellowcake, which the uranium purity is around 70% and the main impurity is thorium, was obtained from monazite processing pilot plant of Rare Earth Research and Development Center in Thailand. For uranium re-extraction process, the extractant chosen was Tributylphosphate (TBP) in kerosene. It was found that the optimum concentration of TBP was 10% in kerosene and the optimum nitric acid concentration in uranyl nitrate feed solution was 4 N. An increase in concentrations of uranium and thorium in feed solution resulted in a decrease in the distribution of both components in the extractant. However, the distribution of uranium into the extractant was found to be more than that of thorium. The equilibration study of the extraction system, UO_2(NO_3)/4N HNO_3 – 10%TBP/Kerosene, was also investigated. Two extraction stages were calculated graphically from 100,000 ppm uranium concentration in feed solution input with 90% extraction efficiency and the flow ratio of aqueous phase to organic phase was adjusted to 1.0. For thorium impurity scrubbing process, 10% TBP in kerosene was loaded with uranium and minor thorium from uranyl nitrate solution prepared from yellowcake and was scrubbed with different low concentration nitric acid. The results showed that at nitric acid normality was lower than 1 N, uranium distributed well to aqueous phase. As conclusion, optimum nitric acid concentration for scrubbing process should not less than 1 N and diluted nitric acid or de-ionized water should be applied to strip uranium from organic phase in the final refining process. (author)

  18. Biological treatment of nitrate bearing wastewater from a uranium production plant

    International Nuclear Information System (INIS)

    Benear, A.K.; Kneip, R.W.

    1988-01-01

    The Feed Materials Production Center (FMPC) produces uranium metal products used for DOE defense programs resulting in the generation of nitrate-bearing wastewaters. To treat these wastewaters, a two-column fluidized bed biodenitrification facility (BDN) was constructed at the FMPC. The operation of the BDN resulted in substantial compliance with the design criteria limits for nitrate from July through November, 1987. Since the BDN surge lagoon (BSL) proved inadequate for providing nitrate concentration equalization, the BDN feed nitrate concentration fluctuated widely throughout this period of operation. BDN effluent caused a doubling of the hydraulic loading and a tripling of the organic loading on the FMPC sewage treatment plant (STP). Better control of the methanol feed to the BDN, coupled with reduced throughput and improved preaeration, caused a significant improvement in the operation of the STP. The overloading of the STP prompted a decision to add a stand-alone effluent treatment system to the BDN

  19. Uranium in aqueous solutions by colorimetry

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The method covers the quantitative determination of uranium in known volumes of aqueous solutions that contain radioactive nuclides. These solutions arise from processing of irradiated nuclear fuel and from laboratory studies on irradiated uranium. The method is applicable to solutions containing a minimum of 30 μg of uranium per sample although as little as 0.5 μg can be detected but with lower precision. Highest precision is obtained with 50 to 75 μg of uranium in the test sample. Dilutions must be made at concentrations above 750 μg/ml. The method includes a discussion of photometers and photometric practice, apparatus, reagents, cell matching, preparation of standard curves, calibration by the method of internal standards, procedure, calculation, and precision

  20. Photodegradation of Paracetamol in Nitrate Solution

    Science.gov (United States)

    Meng, Cui; Qu, Ruijuan; Liang, Jinyan; Yang, Xi

    2010-11-01

    The photodegradation of paracetamol in nitrate solution under simulated solar irradiation has been investigated. The degradation rates were compared by varying environmental parameters including concentrations of nitrate ion, humic substance and pH values. The quantifications of paracetamol were conducted by HPLC method. The results demonstrate that the photodegradation of paracetamol followed first-order kinetics. The photoproducts and intermediates of paracetamol in the presence of nitrate ions were identified by extensive GC-MS method. The photodegradation pathways involving. OH radicals as reactive species were proposed.

  1. Photodegradation of Paracetamol in Nitrate Solution

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Meng; Ruijuan, Qu; Jinyan, Liang; Xi, Yang [State Key Laboratory of Pollution Control and Resource Reuse, School of the Environment, Nanjing University, Nanjing 210093 (China)

    2010-11-24

    The photodegradation of paracetamol in nitrate solution under simulated solar irradiation has been investigated. The degradation rates were compared by varying environmental parameters including concentrations of nitrate ion, humic substance and pH values. The quantifications of paracetamol were conducted by HPLC method. The results demonstrate that the photodegradation of paracetamol followed first-order kinetics. The photoproducts and intermediates of paracetamol in the presence of nitrate ions were identified by extensive GC-MS method. The photodegradation pathways involving. OH radicals as reactive species were proposed.

  2. Photodegradation of Paracetamol in Nitrate Solution

    International Nuclear Information System (INIS)

    Meng Cui; Qu Ruijuan; Liang Jinyan; Yang Xi

    2010-01-01

    The photodegradation of paracetamol in nitrate solution under simulated solar irradiation has been investigated. The degradation rates were compared by varying environmental parameters including concentrations of nitrate ion, humic substance and pH values. The quantifications of paracetamol were conducted by HPLC method. The results demonstrate that the photodegradation of paracetamol followed first-order kinetics. The photoproducts and intermediates of paracetamol in the presence of nitrate ions were identified by extensive GC-MS method. The photodegradation pathways involving. OH radicals as reactive species were proposed.

  3. Investigation of complexing of trivalent lanthanoids in aqueous nitrate solutions

    International Nuclear Information System (INIS)

    Kopyrin, A.A.; Proyaev, V.V.; Edinakova, V.

    1985-01-01

    Complexing of trivalent lanthanoids (Ce, Eu) with nitrate-ions in concentrated solutions of lithium and sodium nitrates has been studied in a wide range of ionic forces (1.0-7.0), using the extractional, densimetric and solubility methods. Nitrate complexes registered by the extraction and solubility methods mainly are of second sphere character. During rare earth extraction from concentrated nitrate solutions in the range of nitrate-ion concentrations <= 5 mol/l second sphere neutral nitrate complexes take part in distribution, at higher values of nitrate-ion concentration formation of intrasphere monoligand complexes of lanthanoids should be taken into account

  4. Recovering and recycling uranium used for production of molybdenum-99

    Science.gov (United States)

    Reilly, Sean Douglas; May, Iain; Copping, Roy; Dale, Gregory Edward

    2017-12-12

    A processes for recycling uranium that has been used for the production of molybdenum-99 involves irradiating a solution of uranium suitable for forming fission products including molybdenum-99, conditioning the irradiated solution to one suitable for inducing the formation of crystals of uranyl nitrate hydrates, then forming the crystals and a supernatant and then separating the crystals from the supernatant, thus using the crystals as a source of uranium for recycle. Molybdenum-99 is recovered from the supernatant using an adsorbent such as alumina. Another process involves irradiation of a solid target comprising uranium, forming an acidic solution from the irradiated target suitable for inducing the formation of crystals of uranyl nitrate hydrates, then forming the crystals and a supernatant and then separating the crystals from the supernatant, thus using the crystals as a source of uranium for recycle. Molybdenum-99 is recovered from the supernatant using an adsorbent such as alumina.

  5. Stripping study of U(VI) from loaded TBP/n-paraffin using ammonium nitrate bearing waste as strippant

    International Nuclear Information System (INIS)

    Shrishma Paik; Biswas, S.; Bhattacharya, S.; Roy, S.B.

    2013-01-01

    Stripping studies of U(VI) from loaded solvent TBP/n-paraffin was carried out using ammonium nitrate solution as strippant. Effects of various stripping parameters such as concentration of ammonium nitrate solution, U(VI) concentration in organic phase, initial pH of strippant, temperature etc. have been investigated in detail. Kinetics of the stripping process by ammonium nitrate was found to be slower than that of stripping with water. It was observed that with the increase in ammonium nitrate concentration in aqueous solution, stripping of U(VI) decreased. With the increase in U(VI) loading in the organic phase, there was an increase in uranium stripping for ammonium nitrate whereas for distilled water it becomes reverse. With the increase in pH of the aqueous ammonium nitrate solution, stripping increased up to a certain pH of 8.5 and after that precipitation of uranium started. Increase in temperature of the biphasic system shows an enhancing effect of U(VI) stripping. Evaluation of thermodynamic data such as ΔH indicated that the process is endothermic. Based on the optimized conditions, McCabe-Thiele diagram was constructed for U(VI) stripping using ammonium nitrate solution at room temperature. (author)

  6. Uranium oxide recovering method

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  7. 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)

  8. Coupled jump rotational dynamics in aqueous nitrate solutions.

    Science.gov (United States)

    Banerjee, Puja; Yashonath, Subramanian; Bagchi, Biman

    2016-12-21

    A nitrate ion (NO 3 - ) with its trigonal planar geometry and charges distributed among nitrogen and oxygen atoms can couple to the extensive hydrogen bond network of water to give rise to unique dynamical characteristics. We carry out detailed atomistic simulations and theoretical analyses to investigate these aspects and report certain interesting findings. We find that the nitrate ions in aqueous potassium nitrate solution exhibit large amplitude rotational jump motions that are coupled to the hydrogen bond rearrangement dynamics of the surrounding water molecules. The jump motion of nitrate ions bears certain similarities to the Laage-Hynes mechanism of rotational jump motions of tagged water molecules in neat liquid water. We perform a detailed atomic-level investigation of hydrogen bond rearrangement dynamics of water in aqueous KNO 3 solution to unearth two distinct mechanisms of hydrogen bond exchange that are instrumental to promote these jump motions of nitrate ions. As observed in an earlier study by Xie et al., in the first mechanism, after breaking a hydrogen bond with nitrate ion, water forms a new hydrogen bond with a water molecule, whereas the second mechanism involves just a switching of hydrogen bond between the two oxygen atoms of the same nitrate ion (W. J. Xie et al., J. Chem. Phys. 143, 224504 (2015)). The magnitude as well as nature of the reorientational jump of nitrate ion for the two mechanisms is different. In the first mechanism, nitrate ion predominantly undergoes out-of-plane rotation, while in the second mechanism, in-plane reorientation of NO 3 - is favourable. These have been deduced by computing the torque on the nitrate ion during the hydrogen bond switching event. We have defined and computed the time correlation function for coupled reorientational jump of nitrate and water and obtained the associated relaxation time which is also different for the two mechanisms. These results provide insight into the relation between the

  9. The uranium (VI) determination method in the technological products

    International Nuclear Information System (INIS)

    Fursa, L.I.; Belykevich, N.A.

    1994-01-01

    The method includes an extraction of the uranyl nitrate by solution' hexa butyl tries amide of phosphoric acid in the hexane, a reextraction of uranium by arsenazo III solution under pH 2 and following the solution' photometry. With the purpose of excluding the emulsion' formation under uranium reextraction has been added chloroform in volume ratio to hexane 1:(3.0-7.5). tabs. 2

  10. Research of calcium oxide hydration in calcium nitrate solutions

    Directory of Open Access Journals (Sweden)

    M.A. Oliynyk

    2016-09-01

    Full Text Available Mineral fertilizers are one of the important factors of agriculture intensification and increasing of food products quantity. The volume of fertilizers production and its domestic consumption in Ukraine indicate that nitrogen fertilizer using only comes nearer to the required number of science-based. One of the most widespread artificial fertilizers is the calcium nitrate. Aim: The aim is to study and theoretically substantiate the processes occurring in the preparation of suspensions of calcium hydroxide Са(ОН2 in solution of calcium nitrate Ca(NО32. Materials and Methods: The technical calcium oxide (quicklime DSTU BV.2.7-90-99, solutions of calcium nitrate of 15, 20, 25, 30, 35 and 40% Ca(NО32 concentrations were used in the work. The content of lime in the preparation of a suspension in the solution changed (in terms of calcium oxide CaO from 150 g/dm3 to the maximum possible. Each of these solutions saturated at 40°С in lime to maximum concentration. Suitable for use in these experiments and in the technology of calcium nitrate obtaining are considered the solutions (suspensions that within 12 hours did not lose their mobility (transportability. Results: The experimental results show that increasing of the concentration of calcium nitrate in solution within the range 15...40%, the amount of lime that you can put into the solution without loss of transportability decreases. Further increasing of lime quantity in solutions concentrations causes to its solidifying, loss of mobility (transportability. Calculations showed that in the presence of calcium nitrate the solubility of Са(ОН2 is reduced nearly by order that can lead to the formation of calcium oxide CaO the solid phase Са(ОН2 on the surface, which also can form hydrogen bonds with the components of the solution. As the probability of formation of hydrogen bonds in solutions is high, there is a possibility of formation of clusters.

  11. Heterogeneous-catalytic redox reactions in nitrate - formate systems

    International Nuclear Information System (INIS)

    Ananiev, A.V.; Shilov, V.P.; Tananaev, I.G.; Brossard, Ph.; Broudic, J.Ch.

    2000-01-01

    It was found that an intensive destruction of various organic and mineral substances - usual components of aqueous waste solutions (oxalic acid, complexones, urea, hydrazine, ammonium nitrate, etc.) takes place under the conditions of catalytic denitration. Kinetics and mechanisms of urea and ammonium nitrate decomposition in the system HNO 3 - HCOOH - Pt/SiO 2 are comprehensively investigated. The behaviour of uranium, neptunium and plutonium under the conditions of catalytic denitration is studied. It is shown, that under the certain conditions the formic acid is an effective reducer of the uranium (VI), neptunium (VI, V) and plutonium (VI, IV) ions. Kinetics of heterogeneous-catalytic red-ox reactions of uranium (VI), neptunium (VI, V) and plutonium (VI, IV) with formic acid are investigated. The mechanisms of the appropriate reactions are evaluated. (authors)

  12. Safety aspects of solvent nitration in HTGR fuel reprocessing

    International Nuclear Information System (INIS)

    Wilbourn, R.G.

    1977-06-01

    Reprocessing of HTGR fuels requires evaporative concentration of uranium and thorium nitrate solutions. The results of a bench-scale test program conducted to assess the safety aspects of planned concentrator operations are reported

  13. Loading ion exchange resins with uranium for HTGR fuel kernels

    International Nuclear Information System (INIS)

    Notz, K.J.; Greene, C.W.

    1976-12-01

    Uranium-loaded ion exchange beads provide an excellent starting material in the production of uranium carbide microspheres for nuclear fuel applications. Both strong-acid (sulfonate) and weak-acid (carboxylate) resins can be fully loaded with uranium from a uranyl nitrate solution utilizing either a batch method or a continuous column technique

  14. LITERATURE SURVEY FOR GROUNDWATER TREATMENT OPTIONS FOR NITRATE IODINE-129 AND URANIUM 200-ZP-1 OPERABLE UNIT HANFORD SITE

    Energy Technology Data Exchange (ETDEWEB)

    BYRNES ME

    2008-06-05

    This literature review presents treatment options for nitrate, iodine-129, and uranium, which are present in groundwater at the 200-ZP-I Groundwater Operable Unit (OU) within the 200 West Area of the Hanford Site. The objective of this review is to determine available methods to treat or sequester these contaminants in place (i.e., in situ) or to pump-and-treat the groundwater aboveground (i.e., ex situ). This review has been conducted with emphasis on commercially available or field-tested technologies, but theoretical studies have, in some cases, been considered when no published field data exist. The initial scope of this literature review included only nitrate and iodine-I 29, but it was later expanded to include uranium. The focus of the literature review was weighted toward researching methods for treatment of nitrate and iodine-129 over uranium because of the relatively greater impact of those compounds identified at the 200-ZP-I OU.

  15. LITERATURE SURVEY FOR GROUNDWATER TREATMENT OPTIONS FOR NITRATE, IODINE-129 AND URANIUM 200-ZP-1 OPERABLE UNIT, HANFORD SITE

    International Nuclear Information System (INIS)

    BYRNES ME

    2008-01-01

    This literature review presents treatment options for nitrate, iodine-129, and uranium, which are present in groundwater at the 200-ZP-I Groundwater Operable Unit (OU) within the 200 West Area of the Hanford Site. The objective of this review is to determine available methods to treat or sequester these contaminants in place (i.e., in situ) or to pump-and-treat the groundwater aboveground (i.e., ex situ). This review has been conducted with emphasis on commercially available or field-tested technologies, but theoretical studies have, in some cases, been considered when no published field data exist. The initial scope of this literature review included only nitrate and iodine-I 29, but it was later expanded to include uranium. The focus of the literature review was weighted toward researching methods for treatment of nitrate and iodine-129 over uranium because of the relatively greater impact of those compounds identified at the 200-ZP-I OU

  16. 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.

  17. 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

  18. Recovery of uranium from sulphate solutions containing molybdenum

    International Nuclear Information System (INIS)

    Weir, D.R.; Genik-Sas-Berezowsky, R.M.

    1983-01-01

    A process for recovering uranium from a sulphate solution containing dissolved uranium and molybdenum includes reacting the solution with ammonia (pH 8 to 10), the pH of the original solution must not exceed 5.5 and after the addition of ammonia the pH must not be in the vicinity of 7 for a significant time. The resultant uranium precipitate is relatively uncontaminated by molybdenum. The precipitate is then separated from the remaining solution while the pH is maintained within the stated range

  19. Method of processing plutonium and uranium solution

    International Nuclear Information System (INIS)

    Otsuka, Katsuyuki; Kondo, Isao; Suzuki, Toru.

    1989-01-01

    Solutions of plutonium nitrate solutions and uranyl nitrate recovered in the solvent extraction step in reprocessing plants and nuclear fuel production plants are applied with low temperature treatment by means of freeze-drying under vacuum into residues containing nitrates, which are denitrated under heating and calcined under reduction into powders. That is, since complicate processes of heating, concentration and dinitration conducted so far for the plutonium solution and uranyl solution are replaced with one step of freeze-drying under vacuum, the process can be simplified significantly. In addition, since the treatment is applied at low temperature, occurrence of corrosion for the material of evaporation, etc. can be prevented. Further, the number of operators can be saved by dividing the operations into recovery of solidification products, supply and sintering of the solutions and vacuum sublimation. Further, since nitrates processed at a low temperature are powderized by heating dinitration, the powderization step can be simplified. The specific surface area and the grain size distribution of the powder is made appropriate and it is possible to obtain oxide powders of physical property easily to be prepared into pellets. (N.H.)

  20. Radiation-induced nitration of organic compounds in aqueous solutions

    International Nuclear Information System (INIS)

    Ershov, B.G.; Gordeev, A.V.; Bykov, G.L.

    2009-01-01

    Radiation-induced nitration of organic compounds in aqueous solutions was studied. It was found that γ-irradiation of solutions containing acetic and nitric acid and/or their salts gives nitromethane. Dependences of the product yield on the absorbed dose and the contents of components were established. The mechanism of radiation nitration involving radicals is discussed. (author)

  1. Determination of small amounts of nitric acid in the presence of large amounts of uranium (VI) and extraction of nitric acid into TBP solutions highly loaded with uranyl nitrate

    International Nuclear Information System (INIS)

    Kolarik, Z.; Schuler, R.

    1982-10-01

    A new method for the determination of small amounts of nitric acid in the presence of large amounts of uranium(VI) was elaborated. The method is based on the precipitation of uranium(VI) as iodate and subsequent alkalimetric titration of the acid in the supernatant. The extraction of nitric acid and uranium(VI) with 30% TBP in dodecane was studied at high loading of the organic phase with uranyl nitrate and at 25, 40 and 60 0 C. The results are compared with available published data on the extraction of nitric acid under similar conditions. (orig.) [de

  2. Densities concentrations of aqueous of uranyl nitrate solutions

    International Nuclear Information System (INIS)

    Rodrigo Otero, A.; Rodriguez Hernandez, B.; Fernandez Rodriguez, L.

    1966-01-01

    The ratio density-concentration of aqueous uranyl nitrate solutions expressed as U 3 O 8 grams/liter, U grams/liter and hexahydrate uranyl nitrate weight percent at different temperatures, are established. Experimental values are graphically correlated and compared whit some published data. (Author) 2 refs

  3. Extraction of Uranium (VI) Nitrate Complexes By Adogen 464. Vol. 3

    International Nuclear Information System (INIS)

    El-Yamani, I.S.; Abd El-Messieh, E.N.

    1996-01-01

    Long-chain amines are frequently used for the extraction of actinides and offer several advantages for their use in the reprocessing of high burn-up nuclear fuels. The present investigation was undertaken to obtain some information on the extraction of Uranium (VI)by adogen 464 from nitrate medium. Extraction parameters studied include: acidity, salting agent, metal and extractant concentrations, diluent type, and temperature. Extraction mechanism was proposed on the basis of results obtained. Best results were attained at 6 M H N O 3 and sodium nitrate was found to increase appreciably the extraction. It was infrared that extraction was dominated by solvation, and ion exchange reaction mechanisms at lower (<6 M) and higher acidities, respectively. As far as diluents are concerned, the dielectric constant, solvation power and donor characteristics were used to explain the variation in the extraction efficiency of the diluents. Kerosene, the most economic particularly when used on industrial scale, was recommended as optimal diluent. Extractability decreases markedly with rise of temperature, suggesting that the extraction behaviour is exothermic; the thermodynamic functions were also calculated and discussed. A method for the separation of uranium (VI) from some fission products existing in high liquid waste was outlined. 3 figs., 2 tabs

  4. Intrinsic neutron source strengths in uranium solutions

    International Nuclear Information System (INIS)

    Anderson, R.E.; Robba, A.A.; Seale, R.L.; Rutherford, D.A.; Butterfield, K.B.; Brunson, G.S.

    1991-01-01

    Neutron production rates for 5% enriched uranyl fluoride and 93% uranyl nitrate solutions have been measured using a high-efficiency neutron well counter. Measurements were made for both solution types as a function of sample volume. These results were extrapolated to zero sample volume to eliminate sample size effects, such as multiplication and absorption. For the 5% enriched uranyl fluoride solution, a neutron production rate of 0.0414 ± 0.0041 n/s/ml was measured; for the 93% enriched uranyl nitrate solution, a neutron production rate of 0.0232 ± 0.0023 n/s/ml was measured. The biggest uncertainty is in measuring the detector efficiency, and further work on this aspect of the experiment is planned. Calculations for the neutron production rates based on measured thick-target (alpha, n) production rates and shown alpha stopping powers are in reasonable agreement with the data for the uranyl nitrate solution, but are in poor agreement with the data for the uranyl fluoride solution. 8 refs., 7 figs., 5 tabs

  5. Development and performance of on-line uranium analyzers

    International Nuclear Information System (INIS)

    Ofalt, A.E.; O'Rourke, P.E.

    1985-10-01

    A diode-array spectrophotometer and and x-ray fluorescence analyzer were installed online in a full-scale prototype facility to monitor uranium loading and breakthrough of ion exchange columns. Uranium concentrations of 10 ppM in uranyl nitrate solutions can be detected online to improve process control and material accountability. 9 figs

  6. Uranium nanoparticle synthesis from leaching solution

    International Nuclear Information System (INIS)

    Sadowski, Z.; Sklodowska, A.

    2014-01-01

    The removal of uranium from leaching and bioleaching solutions is of great significance for an environment protection. In comparison with conventional separation techniques, synthesis of uranium nanoparticles has a number of benefits. It has been demonstrated that the uranium nanoparticles show high catalytic activity. In the present studies a variety of synthesis systems have been used for reduction of uranium from bioleaching solution. Among various catalytical templates the hematite Fe_2O_3 nanoparticles are most interest It was presented the report on development of synthesis method to produce nano structured Fe_2O_3 particles. The efficiency of hematite nanoparticles for adsorption of uranium ions from bioleaching solutions was investigated. Bacterial leaching is alternate technique used to extract uranium from mining wastes. The bioleaching process is environment friendly and gives the extraction yield of over 90%. The bioleaching solutions were obtained from bioleaching experiments using waste materials from different places at Lower Silesia (Kowary, Grzmiaca, Kopaniec, Radoniow). Chemoautotrophic bacteria were used for bioleaching tests. The significant adsorption capacity of U(VI) onto iron oxide and hydroxides (goethite, hematite, and magnetite) was observed. The sorption of U(VI) onto the hematite surface was connected with the chemical reduction of U(VI) to U(IV) by Fe"2"+ ions. The initial reaction system contained excess of Fe"2"+ ions which were used to reduce of U(VI). The reduction of U(VI) occurred at pH at the vicinity of pH=2.4. The colloid particles of hematite with UO_2 nanoparticles were obtained. The results of zeta potential measurements of hematite nanoparticles showed that at the ionic strength equals 10"-"3M NaCl, the average zeta potential was +32.4±3.5 mV at pH = 2.6. The interaction of hematite nanoparticles with the bioleaching solutions led to decrease of positive zeta potential to the value of 6.4± 2.7 mV. (author)

  7. Method for improved decomposition of metal nitrate solutions

    Science.gov (United States)

    Haas, Paul A.; Stines, William B.

    1983-10-11

    A method for co-conversion of aqueous solutions of one or more heavy metal nitrates wherein thermal decomposition within a temperature range of about 300.degree. to 800.degree. C. is carried out in the presence of about 50 to 500% molar concentration of ammonium nitrate to total metal.

  8. Resin-based preparation of HTGR fuels: operation of an engineering-scale uranium loading system

    International Nuclear Information System (INIS)

    Haas, P.A.

    1977-10-01

    The fuel particles for recycle of 233 U to High-Temperature Gas-Cooled Reactors are prepared from uranium-loaded carboxylic acid ion exchange resins which are subsequently carbonized, converted, and refabricated. The development and operation of individual items of equipment and of an integrated system are described for the resin-loading part of the process. This engineering-scale system was full scale with respect to a hot demonstration facility, but was operated with natural uranium. The feed uranium, which consisted of uranyl nitrate solution containing excess nitric acid, was loaded by exchange with resin in the hydrogen form. In order to obtain high loadings, the uranyl nitrate must be acid deficient; therefore, nitric acid was extracted by a liquid organic amine which was regenerated to discharge a NaNO 3 or NH 4 NO 3 solution waste. Water was removed from the uranyl nitrate solution by an evaporator that yielded condensate containing less than 0.5 ppM of uranium. The uranium-loaded resin was washed with condensate and dried to a controlled water content via microwave heating. The loading process was controlled via in-line measurements of the pH and density of the uranyl nitrate. The demonstrated capacity was 1 kg of uranium per hour for either batch loading contractors or a continuous column as the resin loading contractor. Fifty-four batch loading runs were made without a single failure of the process outlined in the chemical flowsheet or any evidence of inability to control the conditions dictated by the flowsheet

  9. Non destructive testing of uranium in solution using a portable optical fiber photometer

    International Nuclear Information System (INIS)

    Boisde, Gilbert; Guillot, Philippe; Monier, Jean; Perez, J.J.

    1983-01-01

    The portable instrument, called TELEPHOT 3 N, has be following main characteristics: - regulated light source, - optical fibers used as a light vector, - an optical probe fitted with a mirror immersed in the solution, - optical system with interference filters to balance the three measurement channels, - analog and digital absorbance measurement electronic, - associated microcomputer for automatic data acquisition and processing. The parameters and computer programs were determined by a series of measurements taken on reference solutions covering the target ranges. A mathematical model showing uranium complexes in solution leads to the solution of second degree equation. A first degree equation is sufficient for low acidities. This unit -hardware and programs- has been qualified on different uranyl nitrate solutions subject to nuclear materials control. The small minimum quantities of product required, about 25 ml, and the short response time of around 1 minute, wake at a highly practical instrument for check measurements during inspection proceedings. Similar remote measurements is planned during inspection for the glove box control of plutonium solutions. This spectrophotometry technique can be adapted for the process control of industrial solutions [fr

  10. Determination of microamounts of uranium in waste solutions

    International Nuclear Information System (INIS)

    Birringer, K.J.; Netzer, S.; Kuhn, E.; Groll, P.

    1975-07-01

    A method for the determination of microamounts of uranium in presence of high amounts of fission and corrosion products is described. Uranium is separated by reversed-phase chromatography on a small column, packed with Voltalef micro and impregnated with TOPO. For the direct photometric determination uranium is eluted by TAM dissolved in ethanol/pyridine. The efficiency of the separation, using a suitable scrub-solution, was tested with solutions of simulated inactive fission and corrosion products. The reproducibility of the method, with 24 μg of uranium, is +- 2,5%. (orig.) [de

  11. Development of a choronocoulometric method for determining traces of uranium using the catalytic nitrate reduction

    International Nuclear Information System (INIS)

    Cantagallo, M.I.C.; Gutz, I.G.R.

    1990-01-01

    With the aim of improving the sensitivity of the electroanalytical determination of uranium at trace levels. The uranium catalyzed reduction of nitrate on mercury electrode and the technique of chronocoulometry were used. Several experimental parameters were investigated (electrolyte composition, potential program, integration time, blank correction, temperature, previous separation) and adequate conditions were selected for the analytical determination. Under these conditions it was possible to exceed the best reported sensitivity for the catalytic determination, extending the detection limit to 3.10 -10 M. Exploratory study of the combination of this procedure with pre-concentration of uranium ions on the electrode revealed a detection limit ten limes lower. (author) [pt

  12. 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.

  13. Anaerobic bacterial systems result in the removal of soluble uranium

    International Nuclear Information System (INIS)

    Thomson, B.M.; Barton, L.L.; Steenhoudt, K.; Tucker, M.D.

    1994-01-01

    Sulfate-reducing bacteria, nitrate-reducing bacteria and bacteria present in sewage sludge were examined for their ability to reduce the level of soluble U(VI) in enriched media. Cultures of Desulfovibrio desulfuricans, D. gigas, and D. vulgaris were grown in sulfate-containing media while Pseudomonas putida and P. denitrificans were cultivated in nitrate media. The amount of U(VI) removed from solution was dependent on metabolism because greater levels of uranium were removed when U(VI) was added to a growing culture than when added to a culture in stationary phase. The presence of vanadate, arsenate, selenate or molybdate at 0.1 and 0.01 M levels in sulfate-reducing cultures, nitrate-respiring cultures or in sludge cultures did not have an effect on the amount of uranium removed. In all cultures the amount of uranium in solution was markedly reduced after 10 to 20 days and reduced uranium, as U(IV), was detected in several cultures. Present in the cultures of D. desulfuricans were crystals of uranium. Examination of these cultures by electron microscopy indicates that the uranium (IV) is deposited outside of the cell and these needle-like crystals are associated with cellular material. X-ray probe analysis with the electron microscope gave an image that was in close agreement with U(IV). With D. desulfuricans in a continuous stirred tank reactor, kinetic parameters have been calculated for uranium reduction. Over a period of 20 to 60 hours, the amount of soluble uranium removed from the bioreactor was proportional to residence time over a period of 20 to 60 hours

  14. Separation of uranium by extraction with foamed plastics

    International Nuclear Information System (INIS)

    Korkisch, J.

    1983-07-01

    Polyurethane foams are frequently used for the extraction and separation of inorganic and organic species. The attraction of the materials lies in their favourable hydrodynamic properties obviating the need for the forced-flow conditions associated with conventional chromatographic-type column packing of small particles. The research work described has been directed to providing information on the extraction and separation of uranium (and thorium) by an open-cell polyurethane foam from media containing nitrates and from hydrochloric acid systems. The influence of many different experimental parameters (concentrations, acidity, impregnation of the foam with organic extractants) on the extraction was investigated. Based on the results of these investigations two methods were developed to separate uranium from nitric acid solution and from hydrochloric acid solution, respectively. The first uses calcium or aluminium nitrate salting and foam impregnated with Aliquat 336, the second ascorbic acid addition and TOPO-impregnated foam. The methods separate uranium and thorium from each other and from most other elements and can be used analytically or in the purification of uranium from impure plant products such as yellow cake

  15. Radiolysis studies of uranyl nitrate solution in nitric acid medium

    International Nuclear Information System (INIS)

    Siri, Sandra; Mondino, Angel V.

    2005-01-01

    The radiolysis of acidic uranyl nitrate solutions was investigated using Co-60 gamma radiation. Hydrogen peroxide was determined as a function of increasing dose. The UV-vis absorption spectra of the irradiated solutions were measured and the spectral changes were analyzed. The increasing dose increases the absorbance intensities, possibly by an increment in nitrate concentration produced by radiolysis, which can originate the formation of different uranyl complexes in solution. (author)

  16. A discussion about maximum uranium concentration in digestion solution of U3O8 type uranium ore concentrate

    International Nuclear Information System (INIS)

    Xia Dechang; Liu Chao

    2012-01-01

    On the basis of discussing the influence of single factor on maximum uranium concentration in digestion solution,the influence degree of some factors such as U content, H 2 O content, mass ratio of P and U was compared and analyzed. The results indicate that the relationship between U content and maximum uranium concentration in digestion solution was direct ratio, while the U content increases by 1%, the maximum uranium concentration in digestion solution increases by 4.8%-5.7%. The relationship between H 2 O content and maximum uranium concentration in digestion solution was inverse ratio, the maximum uranium concentration in digestion solution decreases by 46.1-55.2 g/L while H 2 O content increases by 1%. The relationship between mass ratio of P and U and maximum uranium concentration in digestion solution was inverse ratio, the maximum uranium concentration in digestion solution decreases by 116.0-181.0 g/L while the mass ratio of P and U increase 0.1%. When U content equals 62.5% and the influence of mass ratio of P and U is no considered, the maximum uranium concentration in digestion solution equals 1 578 g/L; while mass ratio of P and U equals 0.35%, the maximum uranium concentration decreases to 716 g/L, the decreased rate is 54.6%, so the mass ratio of P and U in U 3 O 8 type uranium ore concentrate is the main controlling factor. (authors)

  17. Chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade uranyl nitrate solutions

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The standard covers analytical procedures to determine compliance of nuclear-grade uranyl nitrate solution to specifications. The following methods are described in detail: uranium by ferrous sulfate reduction-potassium dichromate titrimetry and by ignition gravimetry; specific gravity by pycnometry; free acid by oxalate complexation; thorium by the Arsenazo(III) (photometric) method; chromium by the diphenylcarbazide (photometric) method; molybdenum by the thiocyanate (photometric) method; halogens separation by steam distillation; fluorine by specific ion electrode; halogen distillate analysis: chloride, bromide and iodide by amperometric microtitrimetry; bromine by the fluorescein (photometric) method; sulfate sulfur by (photometric) turbidimetry; phosphorus by the molybdenum blue (photometric) method; silicon by the molybdenum blue (photometric) method; carbon by persulfate oxidation-acid titrimetry; nonvolatile impurities by spectrography; volatile impurities by rotating-disk spark spectrography; boron by emission spectrography; impurity elements by spark source mass spectrography; isotopic composition by multiple filament surface-ionization mass spectrometry; uranium-232 by alpha spectrometry; total alpha activity by direct alpha counting; fission product activity by beta and gamma counting; entrained organic matter by infrared spectrophotometry

  18. Chemical aspects of the precise and accurate determination of uranium and plutonium from nuclear fuel solutions

    International Nuclear Information System (INIS)

    Heinonen, O.J.

    1981-01-01

    A method for the simultaneous or separate determination of uranium and plutonium has been developed. The method is based on the sorption of uranium and plutonium as their chloro complexes on Dowex 1x10 column. When separate uranium and plutonium fractions are desired, plutonium ions are reduced to Pu (III) and eluted, after which the uranium ions are eluted with dilute HCl. Simultaneous stripping of a mass ratio U/Pu approximately 1 fraction for mass spectrometric measurements is achieved by proper choice of eluant HC1 concentration. Special attention was paid to the obtaining of americium free plutonium fractions. The distribution coefficient measurements showed that at 12.5-M HCl at least 30 % of americium ions formed anionic chloro complexes. The chemical aspects of isotopic fractionation in a multiple filament thermal ionization source were also investigated. Samples of uranium were loaded as nitrates, chlorides, and sulphates and the dependence of the measured uranium isotopic ratios on the chemical form of the loading solution as well as on the filament material was studied. Likewise the dependence of the formation of uranium and its oxide ions on various chemical and instrumental conditions was investigated using tungsten and rhenium filaments. Systematic errors arising from the chemical conditions are compared with errors arising from the automatic evaluation of of spectra. (author)

  19. Development and optimisation of process parameters for recovery of uranium from calcia slag and lining material (SLM) by leaching process and subsequent recovery of uranium from the leach liquor generated

    International Nuclear Information System (INIS)

    Verma, Dinesh Kumar; Srivastava, Praveen Kumar; Das, Santanu; Kumar, Raj; Roy, S.B.

    2014-01-01

    Presently uranium value is recovered by nitric acid dissolution of the SLM, to get uranyl nitrate solution (UNS) and subsequent solvent extraction process. UNS generated After SLM dissolution is very lean in uranium content and create difficulty in solvent extraction. Moreover, NO X is also generated during SLM dissolution in nitric acid. An alternate process was developed where nitric acid is not being used and uranium is being recovered by leaching out the SLM using acetic acid. The process was also optimised for recovery and overall economics of the process by using process effluent AALL (Acetic Acid Leach Liquor) as a leaching agent. The uranium value in the leach liquor was precipitated by using sodium hydroxide. The precipitate was dissolved in nitric acid and the Uranyl Nitrate Solution generated was having Uranium concentration of 15-30 g/l. The alternate process developed will have less effluent generation, less NO X generation and will produce more concentrated UNS in comparison to the nitric acid dissolution process

  20. Removal of Nitrate From Aqueous Solution Using Rice Chaff

    Directory of Open Access Journals (Sweden)

    Dehghani

    2015-09-01

    Full Text Available Background Nitrate is largely dissolved in the surface and ground water, due to its high solubility. Continual uptake of nitrite through drinking water can lead to problems and diseases (such as blue baby for humans, especially children. Objectives The aim of this study was to develop a new and inexpensive method for the removal of nitrate from water. In this regard, the possibility of using chaff for removal of nitrate from aqueous solutions was studied and the optimum operating conditions of nitrate removal was determined. Materials and Methods This is a cross-sectional study conducted in laboratory scale. The UV spectrophotometer at a wavelength of maximum absorbance (220 nm was used to determine the nitrate concentration. The effect of pH, amount of chaff, temperature, and contact time were investigated. Results The result of this study revealed that chaff as an absorbent could remove nitrate from solutions, and the efficiency of adsorption increased as contact time increased from 5 to 30 minutes, amount of chaff increased from 1 to 3 g, temperature increased in a range of 300 - 400°C and the amount of pH decreased from 10 to 3. The maximum adsorption rate was around pH 3 (53.14%. Conclusions It was shown that the removal efficiency of nitrate was directly proportional to the amount of chaff, temperature, and contact time but inversely to the pH. This study showed that nitrate removal by chaff is a promising technique.

  1. Uranium extraction from sulfuric acid solution using anion exchange resin

    International Nuclear Information System (INIS)

    Sheta, M. E.; Abdel Aal, M. M.; Kandil, A. T.

    2012-12-01

    Uranium is currently recovered from sulfuric acid leach liquor using anion exchange resin as Amberlite IRA 402 (CT). This technology is based on fact that, uranium exists as anionic complexes. This takes place by controlling the pH of the solution, agitation time, temperature and resin to solution ratio (R/S). In this work, batch stirrer tank used for uranium extraction from sulfate medium and after extraction, elution process was done using 1M NaCl solution. After extraction and elution process, the resin was separated from the system and uranium was determined in the solution. (Author)

  2. Methods of Uranium Determination in solutions of Tributyl Phosphate and Kerosene

    International Nuclear Information System (INIS)

    Petrement Eguiluz, J.; Palomares Delgado, F.

    1962-01-01

    A new analytical method for the determination of uranium in organic solutions of tributyl phosphate and kerosene is proposed. In this method the uranium is reextracted from the aqueous phase by reduction with cadmium in acid solution. The uranium can be determined in this solution by the usual methods. In case of very diluted solutions, a direct spectrophtometrical determination of uranium in the organic phase with dibenzoylmethane is proposed. (Author) 21 refs

  3. Nitrate concentrations in soil solutions below Danish forests

    DEFF Research Database (Denmark)

    Callesen, Ingeborg; Raulund-Rasmussen, Karsten; Gundersen, Per

    1999-01-01

    leaching in relation to land-use, a national monitoring programme has established sampling routines in a 7x7 km grid including 111 points in forests. During winters of 1986-1993, soil samples were obtained from a depth of 0-25, 25-50, 50-75 and 75-100 cm. Nitrate concentrations in soil solutions were...... species. A few sites deviated radically from the general pattern of low concentrations. The elevated concentrations recorded there were probably caused by high levels of N deposition due to emission from local sources or temporal disruptions of the N cycle. The nitrate concentration in the soil solution...

  4. Recovery of uranium from uranium mine waters and copper ore leaching solutions

    Energy Technology Data Exchange (ETDEWEB)

    George, D R; Ross, J R [Salt Lake City Metallurgy Research Center, Salt Lake City, UT (United States)

    1967-06-15

    Waters pumped from uranium mines in New Mexico are processed by ion exchange to recover uranium. Production is approximately 200 lb U{sub 3}O{sub 8}/d from waters containing 5 to 15 ppm U{sub 3}O{sub 8}. Recoveries range from 80 to 90%. Processing plants are described. Uranium has been found in the solutions resulting from the leaching of copper-bearing waste rock at most of the major copper mines in western United States. These solutions, which are processed on a very large scale for recovery of copper, contain 2 to 12 ppm U{sub 3}O{sub 8}. Currently, uranium is not being recovered, but a potential production of up to 6000 lb U{sub 3}O{sub 8}/d is indicated. Ion exchange and solvent extraction research studies are described. (author)

  5. Removal of phosphate and nitrate from aqueous solution using ...

    African Journals Online (AJOL)

    The aim of the present study was the removal of phosphate and nitrate by sodium alginate seagrass (Cymodocea rotundata) beads from aqueous solutions. The adsorption characteristics of phosphate and nitrate on the seagrass beads were optimized under different operational parameters like adsorbent dosage, initial ...

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  7. Restoration of uranium solution mining deposits

    International Nuclear Information System (INIS)

    Devries, F.W.; Lawes, B.C.

    1982-01-01

    A process is provided for restoring an ore deposit after uranium solution mining using ammonium carbonate leaching solutions has ceased. The process involves flushing the deposit with an aqueous solution of a potassium salt during which potassium ions exchange with ammonium ions remaining in the deposit. The ammonium containing flushing solution is withdrawn from the deposit for disposal

  8. Restoration of uranium solution mining deposits

    Energy Technology Data Exchange (ETDEWEB)

    Devries, F.W.; Lawes, B.C.

    1982-01-19

    A process is provided for restoring an ore deposit after uranium solution mining using ammonium carbonate leaching solutions has ceased. The process involves flushing the deposit with an aqueous solution of a potassium salt during which potassium ions exchange with ammonium ions remaining in the deposit. The ammonium containing flushing solution is withdrawn from the deposit for disposal.

  9. Preparation of working calibration and test materials: uranyl nitrate solution

    International Nuclear Information System (INIS)

    Yamamura, S.S.; Spraktes, F.W.; Baldwin, J.M.; Hand, R.L.; Lash, R.P.

    1977-05-01

    Reliable working calibration and test materials (WCTMs) are essential to a meaningful analytical measurements quality assurance program. This report describes recommended methods for the preparation of uranyl nitrate solution WCTMs for testing analytical methods, for calibrating methods, and for testing personnel. Uranyl nitrate solution WCTMs can be synthesized from characterized starting materials or prepared from typical plant materials by thorough characterization with reference to primary or secondary reference calibration and test materials (PRCTMs or SRCTMs). Recommended starting materials are described along with detailed procedures for (a) preparing several widely-used types of uranyl nitrate solution WCTMs, (b) packaging the WCTMs, (c) analyzing the WCTMs to establish the reference values or to confirm the synthesis, and (d) statistically evaluating the analytical data to assign reference values and to assess the accuracy of the WCTMs

  10. 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)

  11. The effect of surfactants and complexing agents on the fluorimetric determination of uranium

    International Nuclear Information System (INIS)

    Pakalns, P.; Lane, H.T.

    1979-01-01

    The effect of surfactants, condensed phosphates, soap, EDTA, NTA, citric acid, oil and Alamine-336 on the fluorimetric determination of uranium in a 5-ml aqueous sample extracted with methyl isobutyl ketone after addition of saturated calcium nitrate or 1 N acid-deficient saturated aluminium nitrate salting-out solutions has been established. The error produced by these compounds, even at concentrations up to 500 mg/l, does not exceed = 10%, except for anionic LAS-type detergent, which at above the 50 mg/1 level forms stable emulsions after extraction from solutions containing the acid-deficient aluminium nitrate salting-out reagent. Up to 1 hour of standing time after extraction may be required for the emulsions to break. The detection limit for uranium is 1.1 μg (2s). (author)

  12. Removal of Uranium by Exchanger Resins from Soil Washing Solution

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Soo; Han, G. S.; Kim, G. N.; Koo, D. S.; Jeong, J. W.; Moon, J. K. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Uranyl ions in the acidic waste solution were sorbed on AM-resin resin with a high sorption efficiency, and desorbed from the resin by a batch-type washing with a 60 .deg. C heated 0.5 M Na{sub 2}CO{sub 3} solution. However, the uranium dissolved in the sulfuric acid solution was not sorbed onto the strong anion exchanger resins. Our group has developed a decontamination process with washing and electrokinetic methods for uranium-contaminated (U-contaminated) soil. However, this process generates a large amount of waste solution containing various metal ions. If the uranium selectively removed from the waste solution, a very small amount of the 2nd waste would be generated. Thus, selective sorption of uranium by ion exchange resins was examined in this study.

  13. Diffusive gradient in thin FILMS (DGT) compared with soil solution and labile uranium fraction for predicting uranium bioavailability to ryegrass.

    Science.gov (United States)

    Duquène, L; Vandenhove, H; Tack, F; Van Hees, M; Wannijn, J

    2010-02-01

    The usefulness of uranium concentration in soil solution or recovered by selective extraction as unequivocal bioavailability indices for uranium uptake by plants is still unclear. The aim of the present study was to test if the uranium concentration measured by the diffusive gradient in thin films (DGT) technique is a relevant substitute for plant uranium availability in comparison to uranium concentration in the soil solution or uranium recovered by ammonium acetate. Ryegrass (Lolium perenne L. var. Melvina) is grown in greenhouse on a range of uranium spiked soils. The DGT-recovered uranium concentration (C(DGT)) was correlated with uranium concentration in the soil solution or with uranium recovered by ammonium acetate extraction. Plant uptake was better predicted by the summed soil solution concentrations of UO(2)(2+), uranyl carbonate complexes and UO(2)PO(4)(-). The DGT technique did not provide significant advantages over conventional methods to predict uranium uptake by plants. Copyright 2009 Elsevier Ltd. All rights reserved.

  14. Diffusive gradient in thin FILMS (DGT) compared with soil solution and labile uranium fraction for predicting uranium bioavailability to ryegrass

    Energy Technology Data Exchange (ETDEWEB)

    Duquene, L. [SCK-CEN, Biosphere Impact Studies, Boeretang 200, B-2400 Mol (Belgium); Vandenhove, H., E-mail: hvandenh@sckcen.b [SCK-CEN, Biosphere Impact Studies, Boeretang 200, B-2400 Mol (Belgium); Tack, F. [Ghent University, Laboratory for Analytical Chemistry and Applied Ecochemistry, Coupure Links 653, B-9000 Gent (Belgium); Van Hees, M.; Wannijn, J. [SCK-CEN, Biosphere Impact Studies, Boeretang 200, B-2400 Mol (Belgium)

    2010-02-15

    The usefulness of uranium concentration in soil solution or recovered by selective extraction as unequivocal bioavailability indices for uranium uptake by plants is still unclear. The aim of the present study was to test if the uranium concentration measured by the diffusive gradient in thin films (DGT) technique is a relevant substitute for plant uranium availability in comparison to uranium concentration in the soil solution or uranium recovered by ammonium acetate. Ryegrass (Lolium perenne L. var. Melvina) is grown in greenhouse on a range of uranium spiked soils. The DGT-recovered uranium concentration (C{sub DGT}) was correlated with uranium concentration in the soil solution or with uranium recovered by ammonium acetate extraction. Plant uptake was better predicted by the summed soil solution concentrations of UO{sub 2}{sup 2+}, uranyl carbonate complexes and UO{sub 2}PO{sub 4}{sup -}. The DGT technique did not provide significant advantages over conventional methods to predict uranium uptake by plants.

  15. Diffusive gradient in thin FILMS (DGT) compared with soil solution and labile uranium fraction for predicting uranium bioavailability to ryegrass

    International Nuclear Information System (INIS)

    Duquene, L.; Vandenhove, H.; Tack, F.; Van Hees, M.; Wannijn, J.

    2010-01-01

    The usefulness of uranium concentration in soil solution or recovered by selective extraction as unequivocal bioavailability indices for uranium uptake by plants is still unclear. The aim of the present study was to test if the uranium concentration measured by the diffusive gradient in thin films (DGT) technique is a relevant substitute for plant uranium availability in comparison to uranium concentration in the soil solution or uranium recovered by ammonium acetate. Ryegrass (Lolium perenne L. var. Melvina) is grown in greenhouse on a range of uranium spiked soils. The DGT-recovered uranium concentration (C DGT ) was correlated with uranium concentration in the soil solution or with uranium recovered by ammonium acetate extraction. Plant uptake was better predicted by the summed soil solution concentrations of UO 2 2+ , uranyl carbonate complexes and UO 2 PO 4 - . The DGT technique did not provide significant advantages over conventional methods to predict uranium uptake by plants.

  16. Uranium recovery from phosphonitric solutions

    International Nuclear Information System (INIS)

    Bunus, F.T.; Miu, I.

    1997-01-01

    A new technology for uranium and rare earth recovery applied in a semi-industrial plant processing 5 m 3 /h phosphoric acid has been extended to phosphonitric solution, resulting in the process of nitric acid attack of phosphate rock for complex fertilizer production. In this process uranium and rare earths are obtained at larger quantities due to the complete dissolution of elements involved. The method is based on a one cycle extraction-stripping process using as extractants: di(2-ethylhexyl) phosphate (DEPA) in mixture either with tri-n-butylphosphate (TBP) or tri-n-octylphosphine oxide (TOPO) in view of obtaining a synergic effect for U (VI). A mixer-settler extractor in four steps was used. Two stripping steps are involved for the elements mentioned. Before uranium stripping a scrubbing with urea was introduced to eliminate nitric acid extracted. Uranium was obtained as green cake (hydrated uranium tetrafluoride) which can be easily transformed in hexfluoride or converted to a diuranate. At the same time the radium is also eliminated leading to a non-radioactive fertilizer product. (author),. 8 refs, 4 figs

  17. Corrosion performance of several metals in plutonium nitrate solution

    International Nuclear Information System (INIS)

    Takeda, Seiichiro; Nagai, Takayuki; Yasu, Shozo; Koizumi, Tsutomu

    1995-01-01

    Corrosion behavior of several metals exposed in plutonium nitrate solution was studied. Plutonium nitrate solution with the plutonium concentration ranging from 0.01 to 300 g/l was used as a corrosive medium. Specimens tested were type 304 ULC (304 ULC) stainless steel, type 310 Nb (310 Nb) stainless steel, titanium (Ti), titanium-5% tantalum alloy (Ti-5Ta), and zirconium (Zr). Corrosion behavior of these metals in plutonium nitrate solution was evaluated through examining electrochemical characteristics and corrosion rates obtained by weight loss measurement. From the results of the corrosion tests, it was found that the corrosion rate of stainless steels i.e. 304 ULC and 310 Nb, increases by the presence of plutonium in nitric acid solution. The corrosion potential of the stainless steels shifted linearly towards the noble direction as the concentration of plutonium increases. It is thought that the shifts in corrosion potential of the stainless steels to the noble direction results an increase in anodic current and, hence, corrosion rate. Valve metals, i.e. Ti, Ti-5Ta and Zr, showed good corrosion resistance over the whole range of plutonium concentration examined here. (author)

  18. Novel precipitation technique for uranium recovery from carbonate leach solutions

    International Nuclear Information System (INIS)

    Sujoy Biswas; Rupawate, V.H.; Hareendran, K.N.; Roy, S.B.; Chakravartty, J.K.

    2015-01-01

    The recovery of uranium from carbonate ore leach solution was studied using novel precipitation method. The uranium from leach liquor was recovered as magnesium diuranate with NaOH in presence of trace amount of Mg 2+ . Effects of various parameters such as addition of H 2 SO 4 , MgO, MgSO 4 as well as NaOH were investigated for maximum uranium recovery. Overall uranium recovery of the process was 97 % with improved particle size (∼57 µm). Based on the experimental findings, a process flow-sheet was developed for uranium recovery from carbonate ore leach solution with a uranium concentration of <1 g/L. (author)

  19. Recovery of uranium from different acidic solutions by di-nonyl phenyl phosphoric acid (DNPPA) and TOPO

    International Nuclear Information System (INIS)

    Mishra, S.L.; Vijayalakshmi, R.; Singh, H.

    2004-01-01

    The extraction mechanism of uranium with DNPPA in combination with TOPO has been established from different acidic media. The extraction order of uranium from these media has been found as perchlorate>nitrate>chloride>sulphate>phosphate. Based on these results extraction of uranium from samples of leach liquors generated in the plant has been carried out and >99% recovery was obtained

  20. Uranium. Suppl. Vol. C7

    International Nuclear Information System (INIS)

    Keim, R.; Keller, C.

    1982-01-01

    In this supplement volume C7 the nitrogen compounds of uranium-anides, imides, nitrides, nitrites, nitrates are dealt with. Whereas amides, imides and nitrates have only been of scientific interest up to now, uranium nitride and uranylnitrate are of great technological importance. Therefore the description of the chemical and physical characteristics of UN as a potential fuel for future reactors already comprises about 1/4 of this volume. Also the description of uranyl nitrate - as one of the most important commercial forms of uranium and because of its importance in the chemistry of nuclear fuel element reprocessing - comprises many pages. This is supplemented by further uranium nitrides, ternary and polynary nitrides, oxide nitrides, double nitrides of the various valence steps as well as nitrate complexes and ternary and quarternary systems containing uranyl nitrate. The radiation behaviour of UN, and its distribution (liquid/liquid, liquid solid) as well as the complex formation of the uranyl ion with nitrate are described in other volumes of the uranium series. (RB) [de

  1. Application of solution-mineral equilibrium chemistry to solution mining of uranium ores

    International Nuclear Information System (INIS)

    Riese, A.C.; Propp, C.J.

    1980-01-01

    Modern methods of uranium solution mining are typically accompanied by gains and losses of mass through reagent consumption by rock-forming minerals, with subsequent formation of clay minerals, gypsum, carbonates, and iron oxyhydroxides. A systematic approach to alleviate such problems involves the application of leach solutions that are in equilibrium with the host-rock minerals but in disequilibrium with the ore-forming minerals. This partial equilibrium can be approximated by solution-composition adjustments within the systems K 2 O-Al 2 O 3 SiO 2 -H 2 O and Na 2 O 3 -Al 2 O 3 SiO 2 -H 2 O. Uranium ore containing 0.15 percent U 3 O 8 from the Gulf Mineral Resources Corporation's Mariano Lake mine, the Smith Lake district of the Grants mineral belt, was collected for investigation. Presented are a theoretical evaluation of leachate data and an experimental treatment of the ore, which contained mainly K-feldspar, plagioclase feldspar, and quartz (with lesser amounts of micas, clay minerals, and organic carbonaceous material). Small-scale (less than or equal to 1 kg) column-leaching experiments were conducted to model the results of conventional leaching operations and to provide leachate solutions that could be compared with solutions calculated to be in equilibrium with the matrix minerals. Leach solutions employed include: 1) sulfuric acid, 2) sodium bicarbonate, and 3) sulfuric acid with 1.0 molal potassium chloride. The uranium concentrations in the sodium-bicarbonate leach solution and the acid-leach solution were about a gram per liter at the termination of the tests. However, the permeability of the ore in the acid leach was greatly reduced, owing to the formation of clay minerals. Uranium solubility in the leach column stabilized with the potassium-chloride solution was calculated from leachate compositions to be limited by the solubility of carnotite

  2. Transformation of thorium sulfate in thorium nitrate by ion exchange resin

    International Nuclear Information System (INIS)

    Pereira, W.

    1991-01-01

    A procedure for transforming thorium sulfate into thorium nitrate by means of a strong cationic ion exchanger is presented. The thorium sulfate solution (approximately 15 g/L Th (SO 4 ) 2 ) is percolate through the resin and the column is washed first with water, with a 0,2 M N H 4 OH solution and then with a 0.2 M N H 4 NO 3 solution in order to eliminate sulfate ion. Thorium is eluted with a 2 M solution of (N H 4 ) 2 CO 3 . This eluate is treated with a solution of nitric acid in order to obtain the complete transformation into Th (NO 3 ) 4 . The proposed procedure leads to good quality thorium nitrate with high uranium decontamination. (author)

  3. Nitrate Removal from Aqueous Solutions Using Almond Charcoal Activated with Zinc Chloride

    Directory of Open Access Journals (Sweden)

    Mohsen Arbabi

    2017-10-01

    Full Text Available Background & Aims of the Study: Nitrate is one of the most important contaminants in aquatic environments that can leached to water resources from various sources such as sewage, fertilizers and decomposition of organic waste. Reduction of nitrate to nitrite in infant’s blood stream can cause “blue baby” disease in infants. The aim of this study was to evaluate the nitrate removal from aqueous solutions using modified almond charcoal with zinc chloride. Materials &Methods: This study is an experimental survey. At the first charcoal almond skins were prepared in 5500C and then modified with ZnCl2. Morphologies and characterization of almond shell charcoal were evaluated by using FTIR, EDX, BET and FESEM. Adsorption experiments were conducted with 500 ml sample in Becker. The nitrate concentration removal, contact time, pH and charcoal dosage were investigated. The central composite design method was used to optimizing the nitrate removal process. The results analyzed with ANOVA test. Results: The best condition founded in 48 min, 1250 ppm, 125 mg/l and 3 for retention time, primary nitrate concentration, charcoal dosage and pH respectively. The results showed that the nitrate removal decreases with increasing pH. Modification of skin charcoal is show increasing of nitrate removal from aquatic solution. Conclusion: In this study, the maximum nitrate removal efficiency for raw charcoal and modified charcoal was determined 15.47% and 62.78%, respectively. The results showed that this method can be used as an effective method for removing nitrate from aqueous solutions.

  4. The use of a digital density meter for reprocessing plant analysis of aqueous uranyl nitrate in nitric acid

    International Nuclear Information System (INIS)

    Brown, N.L.; Coubrough, A.; Allan, C.G.

    1980-11-01

    The application of a commercial digital density meter, to control analysis of uranyl nitrate process streams, is described. Its operation under high α and high βγ active conditions is considered. Sources of error inherent in the equipment and in the recommended operating procedure are discussed. Density equations for the uranyl nitrate - nitric acid - water system, and its component sub-systems, are reported. These can be used to measure the acidity of pure HNO 3 solutions, with a precision comparable to that achieved in a high precision acidimetric titration. They also enable uranium concentrations in acidic solutions to be estimated, with a precision of better than 4 gram U per litre, provided that the solution acidity is known to within p.14 molar. The densimeter technique is therefore applicable to process control analysis, with less than 5% coefficient of variation in uranium estimate, at uranium concentrations above 40 grams litre -1 . (U.K.)

  5. Bases for DOT exemption uranyl nitrate solution shipments

    International Nuclear Information System (INIS)

    Moyer, R.A.

    1982-07-01

    Uranyl nitrate solutions from a Savannah River Plant reprocessing facility have been transported in cargo tank trailers for more than 20 years without incident during transit. The solution is shipped to Oak Ridge for further processing and returned to SRP in a solid metal form for recycle. This solution, called uranyl nitrate hexahydrate (UNH) solution in Department of Transportation (DOT) regulations, is currently diluted about 2-fold to comply with DOT concentration limits (10% of low specific activity levels) specified for bulk low specific activity (LSA) liquid shipments. Dilution of the process solution increases the number of shipments, the cost of transportation, the cost of shipper preparations, the cost of further reprocessing in the receiving facility to first evaporate the added water, and the total risk to the population along the route of travel. However, the radiological risk remains about the same. Therefore, obtaining an exemption from DOT regulations to permit shipment of undiluted UNH solution, which is normally about two times the present limit, is prudent and more economical. The radiological and nonradiological risks from shipping a unit load of undiluted solution are summarized for the probable route. Data and calculations are presented on a per load or per shipment basis throughout this memorandum to keep it unclassified

  6. Extractive spectrophotometric determination of uranium with malachite green

    International Nuclear Information System (INIS)

    Dubey, S.C.; Nadkarni, M.N.

    1977-01-01

    A sensitive spectrophotometric method based on the extraction of a uranium-benzoate-Malachite Green complex by chlorobenzene is described. The absorption maximum is at 635 nm and the molar absorptivity is 8.3 x 10 4 1. mole -1 cm -1 . A preliminary separation of uranium by extraction with methyl isobutyl ketone from acid-deficient aluminium nitrate solution is used to avoid interferences. An aliquot of the extract in then diluted with chlorobenzene and shaken with benzoate buffer containing Malachite Green (MG). The method has been applied for the determination of uranium in a synthetic leach solution. The complex extracted is probably [U0 2 (C 6 H 5 C00 3 - ][MG + ]. (author)

  7. Extraction of lanthanide(III) nitrates from water-salt solutions with n.-octanol

    International Nuclear Information System (INIS)

    Keskinov, V.A.; Kudrova, A.V.; Valueva, O.V.; Pyartman, A.K.

    2004-01-01

    Extraction of lanthanide(III) nitrates (Ln=La-Nd, Sm-Gd) from aqueous-salt solutions at 298.15 K was studied using solution of n.-octanol, its concentration 6.31 mol/l. It was ascertained that at Ln(NO 3 ) 3 concentration in aqueous phase below 0.6 mol/l, there is actually no extraction. At higher concentrations of nitrates in aqueous phase the content of lanthanides(III) in organic phase increases in the series La-Gd. Isotherms of extraction were ascertained, its phase equilibria being described mathematically. It is shown that extraction of lanthanide(III) nitrates with n.-octanol should be realized from concentrated aqueous solutions [ru

  8. Uranium solution mining: comparison of New Mexico with South Texas

    International Nuclear Information System (INIS)

    Conine, W.D.

    1980-01-01

    In-situ uranium-leaching or solution-mining operations are currently underway in both south Texas and Wyoming. Mobil Oil Corporation is in the process of applying solution-mining technology, such as that developed at the O'Hern facility in south Texas, to uranium orebodies located near Crownpoint, New Mexico. The O'Hern facility uses an alkaline-leach process to bring the uranium to the surface, where it is removed from solution using ion-exchange resin and chemical precipitation. Line-drive and five-spot well field patterns are used to inject and recover the leach solutions. Although details of ore occurrence in New Mexico differ from those in south Texas, laboratory, engineering-design, and field-hydrology tests indicate that solution mining of uranium should be feasible in New Mexico. To determine the commercial feasibility, Mobil is proceeding with the construction of pilot-plant facilities for a 75-gallon-perminute (gpm) test at an orebody near Crownpoint. The pilot test will use five-spot patterns at various spacings for production of uranium-bearing leachate. Initial surface processing will be the same as that used in south Texas

  9. The study of kinetics of uranyl nitrate extraction and reextraction, di-n-butylphosphoric acid reextraction in the flow mixer in the system aqueous solutions - tri-n-butyl phosphate in diluent

    International Nuclear Information System (INIS)

    Shchepetil'nikov, N.N.; Timofeev, A.N.; Kharitonov, V.V.

    1992-01-01

    Kinetics of uranyl nitrate and HNO 3 extraction and reextraction in a flow-type mixer for the system 30 vol.% TBP in extractant dearomatized diluent (EDD) was studied. Kinetics of dibutylphosphate acid reextraction was considered and the influence of uranium and zirconium presence on the process was investigated. It is shown that in laboratory mixer of continuous action in case of phase contact duration of 1 min. in system 30 % solution of TBP in EDD extraction sufficiently similar to equilibrium extraction of uranyl nitrate and nitric acid during their extraction and reextraction is achieved

  10. Process for extracting uranium from phosphoric acid solutions

    International Nuclear Information System (INIS)

    1977-01-01

    The description is given of a method for extracting uranium from phosphoric acid solutions whereby the previously oxided acid is treated with an organic solvent constituted by a mixture of dialkylphosphoric acid and trialkylphosphine oxide in solution in a non-reactive inert solvent so as to obtain de-uraniated phosphoric acid and an organic extract constituted by the solvent containing most of the uranium. The uranium is then separated from the extract as uranyl ammonium tricarbonate by reaction with ammonia and ammonium carbonate and the extract de-uraniated at the extraction stage is recycled. The extract is treated in a re-extraction apparatus comprising not less than two stages. The extract to be treated is injected at the top of the first stage. At the bottom of the first stage, ammonia is introduced counter current as gas or as an aqueous solution whilst controlling the pH of the first stage so as to keep it to 8.0 or 8.5 and at the bottom of the last stage an ammonium carbonate aqueous solution is injected in a quantity representing 50 to 80% of the stoichiometric quantity required to neutralize the dialkylphosphoric acid contained in the solvent and transform the uranium into uranyl ammonium tricarbonate [fr

  11. Recovery of uranium from analytical waste solution

    International Nuclear Information System (INIS)

    Kumar, Pradeep; Anitha, M.; Singh, D.K.

    2016-01-01

    Dispersion fuels are considered as advance fuel for the nuclear reactor. Liquid waste containing significant quantity of uranium gets generated during chemical characterization of dispersion fuel. The present paper highlights the effort in devising a counter current solvent extraction process based on the synergistic mixture of D2EHPA and Cyanex 923 to recover uranium from such waste solutions. A typical analytical waste solution was found to have the following composition: U 3 O 8 (∼3 g/L), Al: 0.3 g/L, V: 15 ppm, Phosphoric acid: 3M, sulphuric acid : 1M and nitric acid : 1M. The aqueous solution is composed of mixture of either 3M phosphoric acid and 1M sulphuric acid or 1M sulphuric acid and 1M nitric acid, keeping metallic concentrations in the above mentioned range. Different organic solvents were tested. Based on the higher extraction of uranium with synergistic mixture of 0.5M D2EHPA + 0.125M Cyanex 923, it was selected for further investigation in the present work

  12. Rirang uranium ore processing: continuous solvent extraction of uranium from Rirang ore acid digestion solution

    International Nuclear Information System (INIS)

    Riza, F.; Nuri, H. L.; Waluya, S.; Subijanto, A.; Sarono, B.

    1998-01-01

    Separation of uranium from Rirang ore acid digestion solution by means of continuous solvent extraction using mixer-settlers has been studied and a mixture of 0.3 M D2EHPA and 0.075 M TOPO extracting agent and kerosene diluent is employed to recover and separate uranium from Th, RE, phosphate containing solution. The experiments have been conducted batch-wise and several parameters have been studied including the aqueous to organic phase ratio, A/O, the extraction and the stripping times, and the operation temperature. The optimum conditions for extraction have been found to be A/O = 2 ratio, five minute extraction time per stage at room temperature. The uranium recovery of 99.07% has been achieved at those conditions whilst U can be stripped from the organic phase by 85% H 3 PO 4 solution with an O/A = 1 for 5 minutes stripping time per stage, and in a there stage operation at room temperature yielding a 100% uranium recovery from the stripping process

  13. Preferential adsorption of uranium ions in aqueous solutions by polymers

    International Nuclear Information System (INIS)

    Sakuragi, Masako; Ichimura, Kunihiro; Fujishige, Shoei; Kato, Masao

    1981-01-01

    Amidoxime fiber and triazine fiber were prepared by chemical modification of commercially available polyacrylonitril fiber. It was found that the Amidoxime fiber is efficient to adsorb uranium ions in the artificial sea water. The efficiency of the preferential adsorption decreases by treatment the material with an acid-or an alkaline-solution. The triazine fiber adsorbs uranium ions only in aqueous solutions of such uranyl acetate, in the absence of other ions. In the artificial sea water, it adsorbs other ions instead of uranium. The preferential adsorption of uranium ions was further investigated with a series of polystyrenesulfonamides. Among the polystyrene derivatives, those having carboxyl groups, derived from imino diacetic acid (PSt-Imi), β-alanine (PSt-Ala), glycine (PSt-Gly), and sarcosine (PSt-Sar) were qualified for further discussion. However, it was found that the amount of adsorption of uranium ions by PSt-Imi decreases with increasing the volume of the artificial sea water and/or the duration of the treatment. Taking into account the facts, the preferential adsorption of uranium ions by PSt-Imi in aqueous solution was discussed in detail. (author)

  14. Spectroscopic Characterization of HAN-Based Liquid Gun Propellants and Nitrate Salt Solutions

    Science.gov (United States)

    1989-01-15

    spectra were recorded of bubbles of a concentrated aqueous nitrate solution, mineral oil, and an aqueous surfactant solution. Polymethacrylic acid ...FTIR spectra of droplets of a concentrated aqueous nitrate salt based solution (LGP1845), of solid particles cf polymethacrylic acid packing IO, 3... polymethacrylic acid low density packing foam cut to a 3x4 mnn rectangle was levitated with a low acoustic power. The sample was easily I positioned in the

  15. Precipitation behavior of uranium in multicomponent solution by oxalic acid

    International Nuclear Information System (INIS)

    Shin, Y.J.; Kim, I.S.; Lee, W.K.; Shin, H.S.; Ro, S.G.

    1996-01-01

    A study on the precipitation of uranium by oxalic acid was carried out in a multicomponent solution. The precipitation method is usually applied to the treatment of radioactive waste and the recovery of uranium from a uranium-scrap contaminated with impurities. In these cases, the problem is how to increase the precipitation yield of target element and to prevent impurities from coprecipitation. The multicomponent solution in the present experiment was prepared by dissolving U, Nd, Cs and Sr in nitric acid. The effects of concentrations of oxalic acid and ascorbic acid on the precipitation yield and purity of uranium were observed. As results of the study, the maximum precipitation yield of uranium is revealed to be about 96.5% and the relative precipitation ratio of Nd, Cs and Sr versus uranium are discussed at the condition of the maximum precipitation yield of uranium, respectively. (author). 11 refs., 5 figs., 1 tab

  16. Determination of plutonium in pure plutonium nitrate solutions - Gravimetric method

    International Nuclear Information System (INIS)

    1987-01-01

    This International Standard specifies a precise and accurate gravimetric method for determining the concentration of plutonium in pure plutonium nitrate solutions and reference solutions, containing between 100 and 300 g of plutonium per litre, in a nitric acid medium. The weighed portion of the plutonium nitrate is treated with sulfuric acid and evaporated to dryness. The plutonium sulfate is decomposed and formed to oxide by heating in air. The oxide is ignited in air at 1200 to 1250 deg. C and weighed as stoichiometric plutonium dioxide, which is stable and non-hygroscopic

  17. Electrocnecical behaviour of zirconium during its anodic polarization in nitrate solutions

    International Nuclear Information System (INIS)

    Stabrovskij, A.I.; Karasev, A.F.

    1983-01-01

    Electrochemical behaviour of zirconium during its anodic polarization in nitrate solutions is investigated in detail to find the method of its complete dissolution. A study has been made of the influence of varioUs factors: current density electric potential, composition and temperature of the solution, anodic polarization duration on the Zr anodic polarization in nitric acid, on the maximum permissible current density and on the zirconium yield to the solution. The zirconium polarization decreases with an acid concentration and temperature increase and increases with the current density. Iron nitrate additions to nitric acid decrease, while ammonium fluoride additions increase zirconium yield into the solution

  18. Cementation feasibility of a uranium-thorium based solution by physical and mechanical characterization

    International Nuclear Information System (INIS)

    Carpentiero, R.; Luce, A.; Troiani, F.

    2002-01-01

    By reprocessing Elk River nuclear fuel, at the ENEA ITREC Plant (South of Italy), about 3 m 3 of Uranium-Thorium based solution were produced. Previously considered an intermediate product to be further treated to recover U and Th, it is now being considered a waste, due to considerable content of fission products and to phasing out of the Italian nuclear industry. Together with other treatment options, a conditioning process in cement matrix is being evaluated, supported by some chemical, physical and mechanical tests on samples prepared with simulated waste. The main components selected to simulate the real solution were thorium nitrate (at two different concentrations), ferrous nitrate and nitric acid. This solution has been neutralized with sodium carbonate (at two different concentration) and cemented by means of a properly defined formulation. Pozzolanic blend cement, at different water to cement ratio, with and without a silica type additive, has been investigated. Cubic samples were subjected to compression tests and repeated freeze-thaw cycles followed by compression tests. Cylindrical samples were subjected to a leach test (according. to the tn ANSI/ANS-16.1 standard). The obtained results are above the minimum acceptance values established by the Italian authority. The evaluated properties are the first important elements to estimate the long term-instability of conditioned radioactive waste. Meanwhile a preliminary theoretical study has been done to evaluate the gas evolution from the matrix due to radiolysis effect. The reached conclusions encourage the development of further analysis to implement a cementation facility. (Author)

  19. XPS study of the passive layers formed on lead in aqueous nitrate solutions

    International Nuclear Information System (INIS)

    Uchida, Miho; Okuwaki, Akitsugu

    1997-01-01

    The analysis of the lead surface immersed in aqueous nitrate solutions by X-ray photoelectron spectroscopy (XPS) shows the formation of passive oxide layer containing nitrogen compound. The oxide layer formed on the lead surface in aqueous ammonium nitrate solution was hydrolyzed and cracked. (author)

  20. Application of chronocoulomentry for trace levels uranium determination using catalytic nitrate reduction on mercury electrode

    International Nuclear Information System (INIS)

    Cantagallo, M.I.C.

    1988-01-01

    With the aim of improving the sensitivity of the electro-analytical determination of uranium at trace levels, the uranium catalyzed reduction of nitrate on mercury electrodes was used and the technique of chronocoulometry was compared with other voltammetric techniques. The catalytic process offers high sensitivity in comparison with uranyl reduction in absence of nitrate. The chronocoulometry, virtually unexplored for analytical applications, was found to be specially well suited for determinations based on this kind of electrode process, when using current integration times in the range of several seconds. Under these conditions the interference from diffusion controlled faradaic processes is reduced to a minimum. Several experimental parameters were investigated (eletrolyte composition, potential program, integration time, blank correction, temperature, previous separation) and adequate conditions were selected for the analytical determination of pure and real samples. The proposed method was applied and evaluated with real and, when necessary, an adapted liquid-liquid extraction procedure was used. Reference materials with complex matrices like rocks were first solubilized by hot digestion under pressure. The obtained results are in good agreement with the values obtained with other techniques such as X-ray fluorescence, mass spectrometry-isotope dilution and epithermal netron activation analysis. (author) [pt

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

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

  2. Bioaccumulation of uranium and thorium from the solution containing both elements using various microorganisms

    International Nuclear Information System (INIS)

    Tsuruta, T.

    2006-01-01

    The effects of proton, thorium and uranium on the bioaccumulation of thorium and uranium from the solution (pH 3.5) containing uranium and thorium using Streptomyces levoris cells were examined. The amount of thorium accumulated using the cells decreased by the pre-contact between the cells and the solution (pH 3.5) containing no metals, whereas that of uranium was almost unaffected by the treatment. The amount of thorium was almost unaffected by the existence of uranium. On the other hand, the amount of uranium accumulated was strongly affected by the thorium, especially thorium addition after uranium accumulation. The decrease of uranium accumulated by the addition of thorium after the accumulation of uranium was higher than that from the solution containing both elements. Therefore, the contribution of uranium-thorium exchange reaction was higher than that of competition reaction. Accordingly, proton-uranium-thorium exchange reaction was occurred in the accumulation of thorium from the solution containing thorium and uranium. The gram-positive bacteria, such as Micrococcus luteus, Arthrobacter nicotianae, Bacillus subtilis and B. megaterium, has a much higher separation factor as thorium/uranium than that of actinomycetes. These gram-positive bacterial strains can be used for the accumulation of thorium from the solution containing uranium and thorium

  3. Extractive spectrophotometric determination of uranium with malachite green

    Energy Technology Data Exchange (ETDEWEB)

    Dubey, S C; Nadkarni, M N [Bhabha Atomic Research Centre, Bombay (India). Fuel Reprocessing Div.

    1977-04-01

    A sensitive spectrophotometric method based on the extraction of a uranium-benzoate-Malachite Green complex by chlorobenzene is described. The absorption maximum is at 635 nm and the molar absorptivity is 8.3 x 10/sup 4/1. mole /sup -1/ cm/sup -1/. A preliminary separation of uranium by extraction with methyl isobutyl ketone from acid-deficient aluminium nitrate solution is used to avoid interferences. An aliquot of the extract in then diluted with chlorobenzene and shaken with benzoate buffer containing Malachite Green (MG). The method has been applied for the determination of uranium in a synthetic leach solution. The complex extracted is probably (U0/sub 2/(C/sub 6/H/sub 5/C00/sub 3//sup -/)(MG/sup +/).

  4. Maintenance of a labour safety and ecological safety in chinks at extraction of uranium

    International Nuclear Information System (INIS)

    Svambaev, Z.A.; Svambaev, E.A.; Sultanbekov, G.A; Tusupbekova, S.T.; Svambaev, A.S.

    2010-01-01

    then an overload of the sorbent sated with uranium in column disobeyed are named. Uterus sobered the rest after clearing by pitch, moves in sediment bowls of sand uterus sobered. At the given stage of processing it is made disrobers that is removal of uranium from pitch with the help nitrate sobered solutions and reception eluded in other words commodity disobeyed. Pitch passes procedure de nitrate a solution of a sulfuric acid with the purpose reduction the maintenance nitrate of ions, and then procedure of washing from a sulfuric acid, further pitch is sent back in column sobered. The following stage of processing it of sedimentation. Here there is a sedimentation of uranium from commodity disobeyed then a filtration of a pulp to reception of an intermediate product of a concentrate of natural uranium 'yellow concentrate'. After processing by organic chemistry yellow concentrate receive natural uranium, so-called sour - oxide uranium. At all stages of a work cycle of work described above in the text is radiating, toxicological and ecological dangerous demanding the special approach. Many experts refer frequently to parameters of measurement of capacity of an exposition doze, and thus in attention activity of a source ionization radiations not always undertakes. Practical value of relative biological efficiency the radiations established for the control of a degree of radiating danger at a chronic irradiation, a so-called weight multiplier or the weighing factor is not taken into account regulated. Therefore the works connected to processing of productive solutions in particular, and extraction of uranium underground chink in the way alkali as a whole is radiating, toxicological and ecological dangerous process.

  5. Restoration of uranium solution mining deposits

    International Nuclear Information System (INIS)

    DeVries, F.W.; Lawes, B.C.

    1981-01-01

    Ammonium carbonates are commonly used as the lixiviant for in-situ leaching of uranium ores. However this leads to the deposition of ammonium ions in the uranium ore formation and the problem of ammonia contamination of ground water which may find its way into the drinking water supply. The ammonia contamination of the ore deposit may be reduced by injecting an aqueous solution of a potassium salt (carbonate, bicarbonate, halide, sulfate, bisulfate, persulfate, or monopersulfate) into the deposit after mining has ceased

  6. Validation of gamma-ray detection techniques for safeguards monitoring at natural uranium conversion facilities

    Energy Technology Data Exchange (ETDEWEB)

    Dewji, S.A., E-mail: dewjisa@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Road, MS-6335, Oak Ridge, TN 37831-6335 (United States); Lee, D.L.; Croft, S. [Oak Ridge National Laboratory, 1 Bethel Valley Road, MS-6335, Oak Ridge, TN 37831-6335 (United States); Hertel, N.E. [Oak Ridge National Laboratory, 1 Bethel Valley Road, MS-6335, Oak Ridge, TN 37831-6335 (United States); Nuclear and Radiological Engineering Program, Georgia Institute of Technology, 770 State Street, Atlanta, GA 30332-0745 (United States); Chapman, J.A.; McElroy, R.D.; Cleveland, S. [Oak Ridge National Laboratory, 1 Bethel Valley Road, MS-6335, Oak Ridge, TN 37831-6335 (United States)

    2016-07-01

    Recent IAEA circulars and policy papers have sought to implement safeguards when any purified aqueous uranium solution or uranium oxides suitable for isotopic enrichment or fuel fabrication exists. Under the revised policy, IAEA Policy Paper 18, the starting point for nuclear material under safeguards was reinterpreted, suggesting that purified uranium compounds should be subject to safeguards procedures no later than the first point in the conversion process. In response to this technical need, a combination of simulation models and experimental measurements were employed to develop and validate concepts of nondestructive assay monitoring systems in a natural uranium conversion plant (NUCP). In particular, uranyl nitrate (UO{sub 2}(NO{sub 3}){sub 2}) solution exiting solvent extraction was identified as a key measurement point (KMP), where gamma-ray spectroscopy was selected as the process monitoring tool. The Uranyl Nitrate Calibration Loop Equipment (UNCLE) facility at Oak Ridge National Laboratory was employed to simulate the full-scale operating conditions of a purified uranium-bearing aqueous stream exiting the solvent extraction process in an NUCP. Nondestructive assay techniques using gamma-ray spectroscopy were evaluated to determine their viability as a technical means for drawing safeguards conclusions at NUCPs, and if the IAEA detection requirements of 1 significant quantity (SQ) can be met in a timely way. This work investigated gamma-ray signatures of uranyl nitrate circulating in the UNCLE facility and evaluated various gamma-ray detector sensitivities to uranyl nitrate. These detector validation activities include assessing detector responses to the uranyl nitrate gamma-ray signatures for spectrometers based on sodium iodide, lanthanum bromide, and high-purity germanium detectors. The results of measurements under static and dynamic operating conditions at concentrations ranging from 10–90 g U/L of natural uranyl nitrate are presented. A range of

  7. Automated assay of uranium solution concentration and enrichment

    International Nuclear Information System (INIS)

    Horley, E.C.; Gainer, K.; Hansen, W.J.; Kelley, T.A.; Parker, J.L.; Sampson, T.E.; Walton, G.; Jones, S.A.

    1992-01-01

    For the first time, the concentration and enrichment of uranium solutions can be measured in one step. We have developed a new instrument to automatically measure the concentration and enrichment of uranium solutions through the adaptation of a commercial robot. Two identical solution enrichment systems are being installed in the Portsmouth Gaseous Diffusion Plant. These automated systems will reduce radiation exposure to personnel and increase the reliability and repeatability of the measurements. Each robotic system can process up to 40 batch and 8 priority samples in an unattended mode. Both passive gamma-ray and x-ray fluorescence (XRF) analyses are performed to determine total uranium concentration and 235 U enrichment. Coded samples are read by a bar-code reader to determine measurement requirements, then assayed by either or both of the gamma-ray and XRF instruments. The robot moves the sample containers and operates all shield doors and shutters, reducing hardware complexity. If the robots is out of service, an operator can manually perform all operations

  8. Molybdenum from uranium solutions

    International Nuclear Information System (INIS)

    Gardner, H.E.

    1981-01-01

    A method of removing molybdenum from a uranium bearing solution is claimed. It comprises adding sufficient reactive lead compound to supply at least 90 percent of the stoichiometric quantity of lead ion required to fully react with the molybdenum present to form insoluble lead molybdate and continuing the reaction with agitation until the desired percentage of the molybdenum present has reacted with the lead ion

  9. On line spectrophotometry with optical fibers. Application to uranium-plutonium separation in a spent fuel reprocessing plant

    International Nuclear Information System (INIS)

    Boisde, G.; Mus, G.; Tachon, M.

    1985-06-01

    Optimization of mixer-settler operation for uranium-plutonium separation in the Purex process can be obtained by remote spectrophotometry with optical fibers. Data acquisition on uranium VI, uranium IV and plutonium III is examined in function of acidity and nitrate content of the solution. Principles for on line multicomponent monitoring and mathematical modelization of the measurements are described [fr

  10. Determination of uranium and plutonium by sequential potentiometric titration

    International Nuclear Information System (INIS)

    Kato, Yoshiharu; Takahashi, Masao

    1976-01-01

    The determination of uranium and plutonium in mixed oxide fuels has been developed by sequential potentiometric titration. A weighed sample of uranium and plutonium oxides is dissolved in a mixture of nitric and hydrofluoric acids and the solution is fumed with sulfuric acid. After the reduction of uranium and plutonium to uranium(IV) and plutonium(III) by chromium(II) sulfate, 5 ml of buffer solution (KCl-HCl, pH 1.0) is added to the solution. Then the solution is diluted to 30 ml with water and the pH of the solution is adjusted to 1.0 -- 1.5 with 1 M sodium hydroxide. The solution is stirred until the oxidation of the excess of chromium(II) by air is completed. After the removal of dissolved oxygen by bubbling nitrogen through the solution for 10 minutes, uranium (IV) is titrated with 0.1 N cerium(IV) sulfate. Then, plutonium is titrated by 0.02 N cerium(IV) sulfate. When a mixture of uranium and plutonium is titrated with 0.1 N potassium dichromate potentiometrically, the potential change at the end point of plutonium is very small and the end point of uranium is also unclear when 0.1 N potassium permanganate is used as a titrant. In the present method, nitrate, fluoride and copper(II) interfere with the determination of plutonium and uranium. Iron interferes quantitatively with the determination of plutonium but not of uranium. Results obtained in applying the proposed method to 50 mg of mixtures of plutonium and uranium ((7.5 -- 50))% Pu were accurate to within 0.15 mg of each element. (auth.)

  11. Criticality safety calculations of 'poison tube tank' compared with annular tanks for storing fissile solutions

    International Nuclear Information System (INIS)

    Gopalakrishnan, C.R.; Joseph, G.

    1995-01-01

    A comparative study of the shielded area space required for storing fissile solution by the conventional annular tank and by poison tube tank is made. Poison tube tank is similar to commercial heat exchanger. The neutron poisons studied are gadolinium oxide and borax. Variation of multiplication factor for an array of annular tanks containing uranium nitrate or plutonium nitrate solutions are presented for annular widths of 10, 7.5 and 5 cm. It is concluded that for the given concentration, 5 cm annular width tanks are safe at a pitch distance of 120 and 90 cm for uranium and plutonium solutions respectively. Using these, as reference values, it is found that the shielded area saving for the poison tube tank is a factor of 12 and 8 for the given concentration of uranium and plutonium solutions respectively. (author)

  12. Electrochemical processing of nitrate waste solutions

    International Nuclear Information System (INIS)

    Genders, D.; Weinberg, N.; Hartsough, D.

    1992-01-01

    The second phase of research performed at The Electrosynthesis Co., Inc. has demonstrated the successful removal of nitrite and nitrate from a synthetic effluent stream via a direct electrochemical reduction at a cathode. It was shown that direct reduction occurs at good current efficiencies in 1,000 hour studies. The membrane separation process is not readily achievable for the removal of nitrites and nitrates due to poor current efficiencies and membrane stability problems. A direct reduction process was studied at various cathode materials in a flow cell using the complete synthetic mix. Lead was found to be the cathode material of choice, displaying good current efficiencies and stability in short and long term tests under conditions of high temperature and high current density. Several anode materials were studied in both undivided and divided cell configurations. A divided cell configuration was preferable because it would prevent re-oxidation of nitrite by the anode. The technical objective of eliminating electrode fouling and solids formation was achieved although anode materials which had demonstrated good stability in short term divided cell tests corroded in 1,000 hour experiments. The cause for corrosion is thought to be F - ions from the synthetic mix migrating across the cation exchange membrane and forming HF in the acid anolyte. Other possibilities for anode materials were explored. A membrane separation process was investigated which employs an anion and cation exchange membrane to remove nitrite and nitrate, recovering caustic and nitric acid. Present research has shown poor current efficiencies for nitrite and nitrate transport across the anion exchange membrane due to co-migration of hydroxide anions. Precipitates form within the anion exchange membranes which would eventually result in the failure of the membranes. Electrochemical processing offers a highly promising and viable method for the treatment of nitrate waste solutions

  13. Extraction of rare earth metals (3) from aqueous solutions containing thorium and uranyl nitrates

    International Nuclear Information System (INIS)

    Pyartman, A.K.; Kopyrin, A.A.; Berinskij, A.E.; Keskinov, V.A.

    2000-01-01

    Isotherms of extraction of rare earth metals (3) from aqueous solutions containing thorium and uranyl nitrates by solutions of tributylphosphate (TBP) and diisooctylmethylphosphonate (DIOMP) in kerosene at 298.15 Deg C and pH 1 are presented. Equations for description of interphase distribution of components of the systems considered are suggested. These equations describe distribution of components adequately in the systems of thorium nitrate (uranyl nitrate) - rare earth nitrates - (TBP, DIOMP) in the case of wide variation of phase compositions. Dependences of separation factors on composition of aqueous phase are considered [ru

  14. A solvent proceed for the extraction of the irradiate uranium and plutonium in the reactor core; Un procede par solvant pour l'extraction du plutonium de l'uranium irradie dans les piles

    Energy Technology Data Exchange (ETDEWEB)

    Goldschmidt, B; Regnaut, P; Prevot, I [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    Description of the conditions of plutonium, fission products and of uranium separation by selective extraction of the nitrates by organic solvent, containing a simultaneous extraction of plutonium and uranium, followed by a plutonium re-extraction after reduction, and an uranium re-extraction. The rates of decontamination being insufficient in this first stage, we also describes the processes of decontamination permitting separately to get the rates wanted for uranium and plutonium. Finally, we describes the beginning of the operation that consists in a nitric dissolution of the active uranium while capturing the products of gaseous fission, as well as the final concentration of the products of fission in a concentrated solution. (authors) [French] Description des conditions de separation du plutonium, des produits de fission et de l'uranium au moyen d'une extraction selective des nitrates par solvant organique, comprenant une extraction simultanee du plutonium et de l'uranium, suivie d'une reextraction du plutonium apres reduction, et d'une reextraction de l'uranium. Les taux de decontamination etant insuffisants dans ce premier stade, on decrit egalement les processus de decontamination permettant separement d'obtenir les taux desires pour l'uranium et le plutonium. Enfin, on decrit aussi le debut de l'operation qui consiste en une dissolution nitrique de l'uranium actif en captant les produits de fission gazeux, ainsi que la concentration finale des produits de fission sous forme de solution concentree. (auteurs)

  15. The determination of phosphorus in uranium minerals and resulting solutions

    International Nuclear Information System (INIS)

    Petrement Eguiluz, J. C.; Parellada Bellod, R.; Fernandez Cellini, R.

    1964-01-01

    Interferences of several elements present in Spanish uranium minerals in the phosphorus determination by the spectrophotometrical method of the molibdovanada te phosphoric acid are studied. A method is described with a previous separation of these element by a cationic resin. This method is successfully applied to the phosphorus determination in acid or alkaline lixiviation solutions of uranium minerals, as well as in the evaluates of ion exchange resins used used technically for the concentration of solutions with a low uranium content. (Author) 11 refs

  16. The direct effect in the gamma radiolysis of frozen aqueous solutions of nitrates

    International Nuclear Information System (INIS)

    Kalecinski, J.

    1974-01-01

    The gamma radiolysis of frozen at 77 and 195 K solutions of lithium, sodium, potassium, magnesium, strontium and silver nitrates was examined. The yields of the direct effect G(NO 2 - ) were shown to depend on the type of the nitrate and to correlate with the free volumes of the solutions. (author)

  17. Interlaboratory comparison exercise for the determination of uranium by potentiometric titration (first phase)

    International Nuclear Information System (INIS)

    Verdingh, V.; Le Duigou, Y.

    1991-01-01

    Upon request of the Esarda working group on low-enriched uranium conversion and fuel fabrication plants an interlaboratory comparison was organized, to assess the precision and accuracy concerning the determination of uranium by the potentiometric titration method. This report presents the results of the first phase of this exercise (pure uranyl-nitrate solutions). The solutions used in this intercomparison have been certified for their uranium content by the CBNM, Geel. Comparison of the laboratory results with the certified values shows excellent, good and fairly good agreement for many of the participating laboratories. 10 tabs., 5 figs., 10 refs

  18. Radiation chemistry of the aqueous aluminium nitrate solution (Preprint no. RC-26)

    International Nuclear Information System (INIS)

    Kalkar, C.D.; Date, D.B.

    1991-01-01

    Radiolysis of aqueous aluminium nitrate solution is studied as a function of concentration in the range 10 -4 M to 10 -1 M. The stable radiolytic product of nitrate radiolysis is nitrite. The yield of nitrite linearly increases with absorbed dose. The G(NO 2 ) values are determined at various concentrations of aluminium nitrate. A suitable mechanism is proposed to explain the observed G-value for the reduction of nitrate to nitrite. (author). 6 refs., 1 tab

  19. Process for recovering uranium using an alkyl pyrophosphoric acid and alkaline stripping solution

    International Nuclear Information System (INIS)

    Worthington, R.E.; Magdics, A.

    1987-01-01

    A process is described for stripping uranium for a pregnant organic extractant comprising an alkyl pyrophosphoric acid dissolved in a substantially water-immiscible organic diluent. The organic extractant contains tetravalent uranium and an alcohol or phenol modifier in a quantity sufficient to retain substantially all the unhydrolyzed alkyl pyrophosphoric acid in solution in the diluent during stripping. The process comprises adding an oxidizing agent to the organic extractant and thereby oxidizing the tetravalent uranium to the +6 state in the organic extractant, and contacting the organic extractant containing the uranium in the +6 state with a stripping solution comprising an aqueous solution of an alkali metal or ammonium carbonate or hydroxide thereby stripping uranium from the organic extractant into the stripping solution. The resulting barren organic extractant containing substantially all of the unhydrolyzed alkyl pyrophosphoric acid dissolved in the diluent is separated from the stripping solution containing the stripped uranium, the barren extractant being suitable for recycle

  20. Formation of the second organic phase during uranyl nitrate extraction from aqueous solution by 30% tributylphosphate solution in paraffin

    International Nuclear Information System (INIS)

    Yhrkin, V.G.

    1996-01-01

    For extraction systems aqueous solution of uranyl nitrate-30% solution of tributylphosphate in individual paraffins from C 13 to C 17 the influence of the second organic phase of uranyl nitrate concentration in aqueous and organic phases, the length of hydrocarbon chain of paraffin hydrocarbon and temperature from 25 to 50 deg C on formation conditions has been defected. A special method of achieving the conditions of organic phase stratification from three-phase region, involving definition of equilibrium phases composition by density and refractive index, has been elaborated for more precise definition of organic phase homogeneity region. It has been revealed that without addition of nitric acid to uranyl nitrate solution the organic phase homogeneity limits can be achieved solely on paraffins C 15 , C 16 and C 17 and only under conditions similar to equeous phase saturation in terms of uranyl nitrate. 16 refs., 2 figs

  1. Potentiometric titration of free acid in uranium solutions

    International Nuclear Information System (INIS)

    Suh, M. Y.; Kim, W. H.; Kim, J. S.; Sohn, S. C.; Eom, T. Y.; Lee, C. H.; Jeon, Y. S.; Han, S. H.

    1998-02-01

    Hydrolysis properties of metal cations and fundamental principles of the potentiometric titration of free acid in aqueous solutions containing metal cations were described. The published papers and reports for the alkalimetric and acidimetric titration of free acid were surveyed, and the applicability of these titration methods to the uranium and/or plutonium solutions were discussed. This technical report also includes the various results obtained from the authors' researches to establish the alkalimetric and acidimetric titration methods for the determination of free acid in nitric acid solutions containing uranium and/or oxalic acid, and aluminum. The procedure manuals used in chemical processes and the newly prepared manuals based on the authors' researches are appended. (author). 26 refs., 54 figs

  2. Potentiometric titration of free acid in uranium solutions

    Energy Technology Data Exchange (ETDEWEB)

    Suh, M. Y.; Kim, W. H.; Kim, J. S.; Sohn, S. C.; Eom, T. Y.; Lee, C. H.; Jeon, Y. S.; Han, S. H.

    1998-02-01

    Hydrolysis properties of metal cations and fundamental principles of the potentiometric titration of free acid in aqueous solutions containing metal cations were described. The published papers and reports for the alkalimetric and acidimetric titration of free acid were surveyed, and the applicability of these titration methods to the uranium and/or plutonium solutions were discussed. This technical report also includes the various results obtained from the authors` researches to establish the alkalimetric and acidimetric titration methods for the determination of free acid in nitric acid solutions containing uranium and/or oxalic acid, and aluminum. The procedure manuals used in chemical processes and the newly prepared manuals based on the authors` researches are appended. (author). 26 refs., 54 figs.

  3. Precipitation-filtering technology for uranium waste solution generated on washing-electrokinetic decontamination

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Gye-Nam, E-mail: kimsum@kaeri.re.kr; Park, Uk-Ryang; Kim, Seung-Soo; Moon, Jei-Kwon

    2015-05-15

    Graphical abstract: A recycling process diagram for the volume reduction of waste solution generated from washing-electrokinetic decontamination. - Highlights: • A process for recycling a waste solution generated was developed. • The total metal precipitation rate by NaOH in a supernatant after precipitation was the highest at pH 9. • The uranium radioactivity in the treated solution upon injection of 0.2 g of alum was lower. • After drying, the volume of sludge was reduced to 35% of the initial sludge volume. - Abstract: Large volumes of uranium waste solution are generated during the operation of washing-electrokinetic decontamination equipment used to remove uranium from radioactive soil. A treatment technology for uranium waste solution generated upon washing-electrokinetic decontamination for soil contaminated with uranium has been developed. The results of laboratory-size precipitation experiments were as follows. The total amount of metal precipitation by NaOH for waste solution was highest at pH 11. Ca(II), K(I), and Al(III) ions in the supernatant partially remained after precipitation, whereas the concentration of uranium in the supernatant was below 0.2 ppm. Also, when NaOH was used as a precipitant, the majority of the K(I) ions in the treated solution remained. The problem of CaO is to need a long dissolution time in the precipitation tank, while Ca(OH){sub 2} can save a dissolution time. However, the volume of the waste solution generated when using Ca(OH){sub 2} increased by 8 mL/100 mL (waste solution) compared to that generated when using CaO. NaOH precipitant required lower an injection volume lower than that required for Ca(OH){sub 2} or CaO. When CaO was used as a precipitant, the uranium radioactivity in the treated solution at pH 11 reached its lowest value, compared to values of uranium radioactivity at pH 9 and pH 5. Also, the uranium radioactivity in the treated solution upon injection of 0.2 g of alum with CaO or Ca(OH){sub 2} was

  4. Effect of ion concentrations on uranium absorption from sodium carbonate solutions

    International Nuclear Information System (INIS)

    Traut, D.E.; El Hazek, N.M.T.; Palmer, G.R.; Nichols, I.L.

    1979-01-01

    The effect of various ion concentrations on uranium absorption from a sodium carbonate solution by a strong-base, anion resin was investigated in order to help assure an adequate uranium supply for future needs. The studies were conducted to improve the recovery of uranium from in situ leach solutions by ion exchange. The effects of carbonate, bicarbonate, chloride, and sulfate ions were examined. Relatively low (less than 5 g/l) concentrations of chloride, sulfate, and bicarbonate were found to be detrimental to the absorption of uranium. High (greater than 10 g/l) carbonate concentrations also adversely affected the uranium absorption. In addition, the effect of initial resin form was investigated in tests of the chloride, carbonate, and bicarbonate forms; resin form was shown to have no effect on the absorption of uranium

  5. Spectrographic determination of impurities in enriched uranium solutions; Determinacion espectrografica de impurezas en soluciones de uranio enriquecido

    Energy Technology Data Exchange (ETDEWEB)

    Capdevila, C; Roca, M

    1980-07-01

    A spectrographic procedure for the determination of trace amounts of Al, B, Ba, Be, Bi, Ca, Cd, Co, Cr, Cu, Fe, K, L i , Hg, Mn, Mo, Na, Nb, Ni, P, Pb, Ru, Sb, Sn, Sr, Ti, V, Zn, and Zr in enriched uranyl nitrate solutions from the reprocessing of spent nuclear fuels is described. After removal of uranium by either TBP or TNOA solvent extraction, the aqueous phase Is analysed by the graphite spark technique. TBP is adequate for all impurities, excepting boron and phosphorus; both of these elements can sat is factory be determined by using TNOA after the addition of mannitol to avoid boron losses. (Author) 4 refs.

  6. Study of the Changes in Composition of Ammonium Diuranate with Progress of Precipitation, and Study of the Properties of Ammonium Diuranate and its Subsequent Products Produced from both Uranyl Nitrate and Uranyl Fluoride Solutions

    Directory of Open Access Journals (Sweden)

    Subhankar Manna

    2017-04-01

    Full Text Available Uranium metal used for fabrication of fuel for research reactors in India is generally produced by magnesio-thermic reduction of UF4. Performance of magnesio-thermic reaction and recovery and quality of uranium largely depends on properties of UF4. As ammonium diuranate (ADU is first product in powder form in the process flow-sheet, properties of UF4 depend on properties of ADU. ADU is generally produced from uranyl nitrate solution (UNS for natural uranium metal production and from uranyl fluoride solution (UFS for low enriched uranium metal production. In present paper, ADU has been produced via both the routes. Variation of uranium recovery and crystal structure and composition of ADU with progress in precipitation reaction has been studied with special attention on first appearance of the precipitate Further, ADU produced by two routes have been calcined to UO3, then reduced to UO2 and hydroflorinated to UF4. Effect of two different process routes of ADU precipitation on the characteristics of ADU, UO3, UO2 and UF4 were studied here.

  7. Study of the changes in composition of ammonium diuranate with progress of precipitation, and study of the properties of ammonium diuranate and its subsequent products produced from both uranyl nitrate and uranyl fluoride solutions

    International Nuclear Information System (INIS)

    Manna, Subhankar; Kumar, Raj; Satpati, Santosh K.; Roy, Saswati B.; Joshi, Jyeshtharaj B.

    2017-01-01

    Uranium metal used for fabrication of fuel for research reactors in India is generally produced by magnesio-thermic reduction of UF 4 . Performance of magnesio-thermic reaction and recovery and quality of uranium largely depends on properties of UF 4 . As ammonium diuranate (ADU) is first product in powder form in the process flow-sheet, properties of UF 4 depend on properties of ADU. ADU is generally produced from uranyl nitrate solution (UNS) for natural uranium metal production and from uranyl fluoride solution (UFS) for low enriched uranium metal production. In present paper, ADU has been produced via both the routes. Variation of uranium recovery and crystal structure and composition of ADU with progress in precipitation reaction has been studied with special attention on first appearance of the precipitate Further, ADU produced by two routes have been calcined to UO 3 , then reduced to UO 2 and hydroflorinated to UF 4 . Effect of two different process routes of ADU precipitation on the characteristics of ADU, UO 3 , UO 2 and UF 4 were studied here

  8. Criticality safety benchmark experiment on 10% enriched uranyl nitrate solution using a 28-cm-thickness slab core

    International Nuclear Information System (INIS)

    Yamamoto, Toshihiro; Miyoshi, Yoshinori; Kikuchi, Tsukasa; Watanabe, Shouichi

    2002-01-01

    The second series of critical experiments with 10% enriched uranyl nitrate solution using 28-cm-thick slab core have been performed with the Static Experiment Critical Facility of the Japan Atomic Energy Research Institute. Systematic critical data were obtained by changing the uranium concentration of the fuel solution from 464 to 300 gU/l under various reflector conditions. In this paper, the thirteen critical configurations for water-reflected cores and unreflected cores are identified and evaluated. The effects of uncertainties in the experimental data on k eff are quantified by sensitivity studies. Benchmark model specifications that are necessary to construct a calculational model are given. The uncertainties of k eff 's included in the benchmark model specifications are approximately 0.1%Δk eff . The thirteen critical configurations are judged to be acceptable benchmark data. Using the benchmark model specifications, sample calculation results are provided with several sets of standard codes and cross section data. (author)

  9. Standard test methods for chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade plutonium nitrate solutions

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 These test methods cover procedures for the chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade plutonium nitrate solutions to determine compliance with specifications. 1.2 The analytical procedures appear in the following order: Sections Plutonium by Controlled-Potential Coulometry Plutonium by Amperometric Titration with Iron(II) Plutonium by Diode Array Spectrophotometry Free Acid by Titration in an Oxalate Solution 8 to 15 Free Acid by Iodate Precipitation-Potentiometric Titration Test Method 16 to 22 Uranium by Arsenazo I Spectrophotometric Test Method 23 to 33 Thorium by Thorin Spectrophotometric Test Method 34 to 42 Iron by 1,10-Phenanthroline Spectrophotometric Test Method 43 to 50 Impurities by ICP-AES Chloride by Thiocyanate Spectrophotometric Test Method 51 to 58 Fluoride by Distillation-Spectrophotometric Test Method 59 to 66 Sulfate by Barium Sulfate Turbidimetric Test Method 67 to 74 Isotopic Composition by Mass Spectrom...

  10. Coprecipitation of thorium and uranium peroxides from acid solutions

    Energy Technology Data Exchange (ETDEWEB)

    McTaggart, D.R.; Mailen, J.C.

    1981-01-01

    The factors affecting successful coprecipitation of thorium and uranium peroxides from acid media were studied. Variables considered in this work were H/sup +/ concentration, H/sub 2/O/sub 2/ concentration, duration of contact, and rate of feed solution addition. In all experiments, stock solutions of Th(NO/sub 3/)/sub 4/ and UO/sub 2/(NO/sub 3/)/sub 2/ were fed at a controlled rate into H/sub 2/O/sub 2/ solutions with constant stirring. Samples were taken as a function of time to follow the H/sup +/ concentration of the solution, uranium precipitation, thorium precipitation, precipitant weight/volume of solution, and crystalline structure and growth. The optimum conditions for maximum coprecipitation are low H/sup +/ concentration, high H/sub 2/O/sub 2/ concentration, and extended contact time between the solutions.

  11. Evaluation of neutralization treatment processes and their use for uranium tailings solutions

    International Nuclear Information System (INIS)

    Sherwood, D.R.; Opitz, B.E.; Serne, R.J.

    1985-01-01

    The potential for groundwater contamination from the typically acidic mill wastes that are disposed of in tailings impoundments is of primary concern at uranium mill sites in the US. Solution-treatment processes provide a system for limiting the environmental impact from acidic seepage. Treatment of uranium tailings solutions from evaporation ponds, underdrains, and surface seeps could aid in decommissioning active sites or be used as an emergency measure to avert possible uncontrolled discharges. At present, neutralization processes appear to be best suited for treating uranium mill tailings solution because they can, at a reasonable cost, limit the solution concentration of many contaminants and thus reduce the potential for groundwater contamination. However, the effectiveness of the neutralization process depends on the reagent used as well as the chemistry of the waste stream. This article provides a description of neutralization processes, an assessment of their performance on acidic uranium tailings leachates, and recommendations for their use at US uranium mill sites

  12. Ruthenium release from thermally overheated nitric acid solution containing ruthenium nitrosyl nitrate and sodium nitrate to solidify

    Energy Technology Data Exchange (ETDEWEB)

    Sawada, Kayo; Ueda, Yasuyuki; Enokida, Youichi [Nuclear Chemical Engineering Laboratory, Nagoya University, Nagoya 4648603 (Japan)

    2016-07-01

    Radioactive ruthenium (Ru) is one of the dominant elemental species released into the environment from a fuel reprocessing plant in a hypothetical design accident due to its relatively higher fission yield and longer half-life. After the hypothetical accident assuming the loss of all electric power and cooling functions, high-level liquid waste (HLLW) may be overheated by the energetic decays of many fission products in it, and Ru may be oxidized to the volatile tetroxide, RuO{sub 4}, which is released through the off-gas pathway. At a reprocessing plant in Japan, alkaline solution from the solvent scrubbing stream is sometimes mixed with the HLLW followed by vitrification, which can be influenced by the addition of sodium nitrate to a simulated HLLW containing ruthenium nitrosyl nitrate that was experimentally evaluated on a small scale using the overheated nitric acid solution of 2 mol/dm{sup 3}, which was kept at 180 Celsius degrees in a glass evaporator placed in a thermostatic bath. The release fraction of Ru increased by approximately 30% by the addition of sodium nitrate. This may be partially explained by the existence of relatively highly concentrated nitrate ions in the liquid phase that oxidize the ruthenium species to RuO{sub 4} during the drying process. (authors)

  13. Anion analysis in uranium more concentrates by ion chromatography

    International Nuclear Information System (INIS)

    Badaut, V.

    2009-01-01

    In the present exploratory study, the applicability of anionic impurities or attributing nuclear material to a certain chemical process or origin has been investigated. Anions (e.g., nitrate, sulphate, fluoride, chloride) originate from acids or salt solutions that are used for processing of solutions containing uranium or plutonium. The study focuses on uranium ore concentrates ('yellow cakes') originating from different mines. Uranium is mined from different types of ore body and depending on the type of rock, different chemical processes for leaching, dissolving and precipitating the uranium need to be applied. Consequently, the anionic patterns observed in he products of these processes (the 'ore concentrates') are different. The concentrations of different anionic species were measured by ion chromatography using conductivity detection. The results show clear differences of anion concentrations and patterns between samples from different uranium mines. Besides this, differences between sampling campaigns n a same mine were also observed indicating that the uranium ore is not homogeneous in a mine. These within-mine variations, however, were smaller than the between-mine variations. (author)

  14. Electroanalytical studies of uranium, neptunium, and plutonium ions in solutions

    International Nuclear Information System (INIS)

    Yoshida, Zenko; Aoyagi, Hisao; Kihara, Sorin

    1989-01-01

    Redox behavior of uranium, neptunium, and plutonium ions, whose oxidation states in acid solutions are between (VI) and (III), were investigated by flow-coulometry with a column electrode of glassy carbon fibers as well as ordinary voltammetry with a rotating disc electrode. Based on current-potential curves the electrode processes were elucidated taking their disproportionation and/or complexation reactions into account. The flow-coulometry, which provides rapid and quantitative electrolysis, was applied to such analytical purposes as follows; the determination of uranium and plutonium in the solution or the solid with discerning their oxidation states, the preparation of species in a desired oxidation state, even in an unstable state which cannot be prepared by ordinary procedure, and the separation of trace amount of uranium in solutions by the electrodeposition of its hydroxide

  15. The investigation of wood hydrolysis lignin ability for uranium sorption

    International Nuclear Information System (INIS)

    Rachkova, N.G.; Shuktomova, I.I.; Taskaev, A.I.

    2001-01-01

    The uranium are sorbed in wood hydrolysis lignin efficacious and very strong both in uranyl nitrate solutions and in podsolic soil. It may well be that formation of complexes are possible mechanism of irreversible sorption. The static capacity of lignin are 2.7 mg/g. (author)

  16. Non-Invasive Acoustic-Based Monitoring of Heavy Water and Uranium Process Solutions

    Energy Technology Data Exchange (ETDEWEB)

    Pantea, Cristian [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sinha, Dipen N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lakis, Rollin Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Beedle, Christopher Craig [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Davis, Eric Sean [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-10-20

    This presentation includes slides on Project Goals; Heavy Water Production Monitoring: A New Challenge for the IAEA; Noninvasive Measurements in SFAI Cell; Large Scatter in Literature Values; Large Scatter in Literature Values; Highest Precision Sound Speed Data Available: New Standard in H/D; ~400 pts of data; Noninvasive Measurements in SFAI Cell; New funding from NA241 SGTech; Uranium Solution Monitoring: Inspired by IAEA Challenge in Kazakhstan; Non-Invasive Acoustic-Based Monitoring of Uranium in Solutions; Non-Invasive Acoustic-Based Monitoring of Uranium in Solutions; and finally a summary.

  17. Recent studies of uranium and plutonium chemistry in alkaline radioactive waste solutions

    International Nuclear Information System (INIS)

    King, William D.; Wilmarth, William R.; Hobbs, David T.; Edwards, Thomas B.

    2008-01-01

    Solubility studies of uranium and plutonium in a caustic, radioactive Savannah River Site tank waste solution revealed the existence of uranium supersaturation in the as-received sample. Comparison of the results to predictions generated from previously published models for solubility in these waste types revealed that the U model poorly predicts solubility while Pu model predictions are quite consistent with experimental observations. Separate studies using simulated Savannah River Site evaporator feed solution revealed that the known formation of sodium aluminosilicate solids in waste evaporators can promote rapid precipitation of uranium from supersaturated solutions

  18. Recovery and treatment of uranium from uranium-containing solution by liquid membrane emulsion technology

    International Nuclear Information System (INIS)

    Xia Liangshu; Zhou Yantong; Xiao Yiqun; Peng Anguo; Xiao Jingshui; Chen Wei

    2014-01-01

    The recovery and treatment of uranium from uranium-containing solution using liquid membrane emulsion (LME) technology were studied in this paper, which contained the best volume ratio of membrane materials, stirring speed during emulsion process, the conditions of extracting, such as temperature, pH, initial concentration of uranium. Moreover, the mechanism for extracting uranium was also discussed. The best experimental conditions of emulsifying were acquired. The volume fractions of P 204 and liquid paraffin are 0.1 and 0.05, the volume ratios of Span80 and sulphonated kerosene to P 204 are 0.06 and 0.79 respectively, stirring speed is controlled in 2 000 r/min, and the concentration of inner phase is 4 mol/L. The recovery rate of uranium is up to 99% through the LME extracted uranium for 0.5 h at pH 2.5 and room temperature when the initial concentration is less than 400 mg/L and the volume ratio is 5 between the uranium-containing waste water and LME. The calculation results of Gibbs free energy show that the reaction process is spontaneous. (authors)

  19. Potentiometric determination of free nitric-acid in trilaurylamine solutions containing plutonium nitrate

    International Nuclear Information System (INIS)

    Perez, J.J.; Saey, J.C.

    1965-01-01

    A potentiometric method of determination of the free nitric acid in trilaurylamine solutions containing plutonium or thorium nitrates is described. The potentiometric titration is carried out in a mixture of benzene and 1,2-dichloro ethane with a standard solution of trilaurylamine as the titrant. When thorium nitrate is present the metal complex is not dissociated then the titration has a single end-point. In the case of plutonium nitrate the partial dissociation of the plutonium complex corresponds to a second point. The experimental error in duplicate analyses of 50 samples is about 1 per cent for free acid concentrations in the range of 0,03 to 0,1 N and plutonium concentrations between 1 to 5 g/l. (authors) [fr

  20. Potentiometric determination of uranium in simulated Purex Process solutions by acidiometry

    International Nuclear Information System (INIS)

    Cohen, V.H.; Matsuda, H.T.; Araujo, B.F. de; Araujo, J.A. de

    1983-01-01

    A potentiometric methods for sequential free acidity and uranium determination in simulated Purex Process solutions is described. An oxalate solution or a mixture of fluoride-oxalate pellets were used as complexing agent for free titration. Following this first equivalent point, uranium is determined-by indirect titration of H + liberated in the peruanate reaction. Some elements present in the standard fuel elements with a burn-up of 33.000 Mwd/t, neutron flux of 3,2 x 10 13 n.cm -2 .s -1 and cooling time of two years were considered as interfering elements in uranium analyses. As a substitute of Pu-IV, Th(NO 3 ) 4 solution was used. The method can be applied to aqueous and organic (TBP/diluent) solutions with 2% precision and 2% accuracy. (Autor) [pt

  1. Potentiometric determination of uranium in simulated Purex Process solutions by acidiometry

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, V H; Matsuda, H T; Araujo, B.F. de; Araujo, J.A. de

    1984-01-01

    A potentiometric methods for sequential free acidity and uranium determination in simulated Purex Process solutions is described. An oxalate solution or a mixture of fluoride-oxalate pellets were used as complexing agent for free titration. Following this first equivalent point, uranium is determined-by indirect titration of H/sup +/ liberated in the peruanate reaction. Some elements present in the standard fuel elements with a burn-up of 33.000 Mwd/t, neutron flux of 3,2 x 10/sup 13/n.cm/sup -2/.s/sup -1/ and cooling time of two years were considered as interfering elements in uranium analyses. As a substitute of Pu-IV, Th(NO/sub 3/)/sub 4/ solution was used. The method can be applied to aqueous and organic (TBP/diluent) solutions with 2% precision and 2% accuracy. (Autor).

  2. Adsorptive Removal of Nitrate from Aqueous Solution Using Nitrogen Doped Activated Carbon.

    Science.gov (United States)

    Machida, Motoi; Goto, Tatsuru; Amano, Yoshimasa; Iida, Tatsuya

    2016-01-01

    Activated carbon (AC) has been widely applied for adsorptive removal of organic contaminants from aqueous phase, but not for ionic pollutants. In this study, nitrogen doped AC was prepared to increase the adsorption capacity of nitrate from water. AC was oxidized with (NH 4 ) 2 S 2 O 8 solution to maximize oxygen content for the first step, and then NH 3 gas treatment was carried out at 950°C to aim at forming quaternary nitrogen (N-Q) species on AC surface (Ox-9.5AG). Influence of solution pH was examined so as to elucidate the relationship between surface charge and adsorption amounts of nitrate. The results showed that Ox-9.5AG exhibited about twice higher adsorption capacity than non-treatment AC at any initial nitrate concentration and any equilibrium solution pH (pH e ) investigated. The more decrease in pH e value, the more adsorption amount of negatively charged nitrate ion, because the surface charge of AC and Ox-9.5AG could become more positive in acidic solution. The oxidation and consecutive ammonia treatments lead to increase in nitrogen content from 0.35 to 6.4% and decrease in the pH of the point of zero charge (pH pzc ) from 7.1 to 4.0 implying that positively charged N-Q of a Lewis acid was created on the surface of Ox-9.5AG. Based on a Langmuir data analysis, maximum adsorption capacity attained 0.5-0.6 mmol/g of nitrate and adsorption affinity was 3.5-4.0 L/mmol at pH e 2.5 for Ox-9.5AG.

  3. Potential for rhizofiltration of uranium using hairy root cultures of Brassica juncea and Chenopodium amaranticolor

    International Nuclear Information System (INIS)

    Eapen, Susan; Suseelan, K.N.; Tivarekar, Suchita; Kotwal, S.A.; Mitra, R.

    2003-01-01

    Hairy root cultures of Brassica juncea and Chenopodium amaranticolor were developed by genetic transformation using Agrobacterium rhizogenes. The stable, transformed root systems demonstrated a high growth rate of 1.5-3. g/g dry weight/day in Murashige and Skoog medium. In the present study, hairy root system was used for removal of uranium from the solution of concentration up to 5000 μM. The results indicated that the hairy roots could remove uranium from the aqueous solution within a short period of incubation. B. juncea could take up 20-23% of uranium from the solution containing up to 5000 μM, when calculated on g/g dry weight basis. C. amaranticolor showed a slow and steady trend in taking up uranium, with 13 uptake from the solution of 5000 μM concentration. Root growth was not affected up to 500 μM of uranium nitrate over a period of 10 days

  4. Mitigation of Hydrogen Gas Generation from the Reaction of Uranium Metal with Water in K Basin Sludge and Sludge Waste Forms

    International Nuclear Information System (INIS)

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

    2011-01-01

    Prior laboratory testing identified sodium nitrate and nitrite to be the most promising agents to minimize hydrogen generation from uranium metal aqueous corrosion in Hanford Site K Basin sludge. Of the two, nitrate was determined to be better because of higher chemical capacity, lower toxicity, more reliable efficacy, and fewer side reactions than nitrite. The present lab tests were run to determine if nitrate's beneficial effects to lower H2 generation in simulated and genuine sludge continued for simulated sludge mixed with agents to immobilize water to help meet the Waste Isolation Pilot Plant (WIPP) waste acceptance drainable liquid criterion. Tests were run at ∼60 C, 80 C, and 95 C using near spherical high-purity uranium metal beads and simulated sludge to emulate uranium-rich KW containerized sludge currently residing in engineered containers KW-210 and KW-220. Immobilization agents tested were Portland cement (PC), a commercial blend of PC with sepiolite clay (Aquaset II H), granulated sepiolite clay (Aquaset II G), and sepiolite clay powder (Aquaset II). In all cases except tests with Aquaset II G, the simulated sludge was mixed intimately with the immobilization agent before testing commenced. For the granulated Aquaset II G clay was added to the top of the settled sludge/solution mixture according to manufacturer application directions. The gas volumes and compositions, uranium metal corrosion mass losses, and nitrite, ammonia, and hydroxide concentrations in the interstitial solutions were measured. Uranium metal corrosion rates were compared with rates forecast from the known uranium metal anoxic water corrosion rate law. The ratios of the forecast to the observed rates were calculated to find the corrosion rate attenuation factors. Hydrogen quantities also were measured and compared with quantities expected based on non-attenuated H2 generation at the full forecast anoxic corrosion rate to arrive at H2 attenuation factors. The uranium metal

  5. Applicability of hydroxylamine nitrate reductant in pulse-column contactors

    International Nuclear Information System (INIS)

    Reif, D.J.

    1983-05-01

    Uranium and plutonium separations were made from simulated breeder reactor spent fuel dissolver solution with laboratory-sized pulse column contactors. Hydroxylamine nitrate (HAN) was used for reduction of plutonium (1V). An integrated extraction-partition system, simulating a breeder fuel reprocessing flowsheet, carried out a partial partition of uranium and plutonium in the second contactor. Tests have shown that acceptable coprocessing can be ontained using HAN as a plutonium reductant. Pulse column performance was stable even though gaseous HAN oxidation products were present in the column. Gas evolution rates up to 0.27 cfm/ft 2 of column cross section were tested and found acceptable

  6. Terbium nitrate luminescence quenching by eosin in he presence of lithium perchlorate in sulfolane solutions

    International Nuclear Information System (INIS)

    Ostakhov, S.S.; Kolosnitsyn, V.S.; Krasnogorskaya, N.N.; Kazakov, V.P.

    2000-01-01

    Quenching of terbium nitrate luminescence by anionic dye, eosin, in the presence of lithium perchlorate in sulfolane solutions was studied. Temperature dependence of terbium nitrate luminescence in sulfolane solutions in the presence of perchlorate anions were considered. The values of energy required for water molecular substitution in Tb 3+ ion coordination sphere for solvent molecule in electrolyte solution were ascertained [ru

  7. 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

  8. Occurrence forms of uranium in the production solutions in the areas of underground leaching of epigenetic uranium deposits

    International Nuclear Information System (INIS)

    Serebrennikov, V.S.; Dorofeeva, V.A.

    1980-01-01

    Redox, acid-basic features of solutions (Eh changes from + 50 to 650 mV, pH from 7.5 to 1.5) and their chemical composition are studied in the process of hydrogeochemical investigations at the areas of underground leaching (UL) of epigenetic uranium deposits. It is shown that at studied areas of UL under neutral and weakly acidic conditions up to (pH 6.0-5.8), carbonate complexes of uranyl are the prevailing form of uranium existence in the solution, and sulfate complexes prevail under more acidic conditions. A supposition is made that it is expedient to process separate ore blocks with increased carbonate contents, particularly with oxidant additions under near-neutral acid-basic conditions (pH 7.2-6.8) with the use of weakly acid pumping solutions, which act (at the expense of their interaction with carbonates of ore-containing rocks) for enrichment of working solutions with HCO 3 - and CO 3 2- ions, promoting uranium transfer into solution

  9. Uranium uptake of Vetiveria zizanioides (L.) Nash

    International Nuclear Information System (INIS)

    Luu Viet Hung; Maslov, O.D.; Trinh Thi Thu My; Phung Khac Nam Ho; Dang Duc Nhan

    2010-01-01

    Uranium uptake of vetiver grass (Vetiveria zizanioides (L.) Nash) from Eutric Fluvisols (AK), Albic Acrisols (BG), Dystric Fluvisols (HP) and Ferralic Acrisols (TC) in northern Vietnam is assessed. The soils were mixed with aqueous solution of uranyl nitrate to make soils contaminated with uranium at 0, 50, 100, 250 mg/kg before planting the grass. The efficiency of uranium uptake by the grass was assessed based on the soil-to-plant transfer factor (TF U , kg·kg -1 ). It was found that the TF U values are dependent upon the soils properties. CEC facilitates the uptake and the increased soil pH could reduce the uptake and translocation of uranium in the plant. Organic matter content, as well as iron and potassium, inhibits the uranium uptake of the grass. It was revealed that the lower fertile soil, the higher uranium uptake. The translocation of uranium in root for all the soil types studied is almost higher than that in its shoot. It seems that vetiver grass could potentially be used for the purpose of phytoremediation of soils contaminated with uranium

  10. Criticality of mixtures of plutonium and high enriched uranium

    International Nuclear Information System (INIS)

    Grolleau, E.; Lein, M.; Leka, G.; Maidou, B.; Klenov, P.

    2003-01-01

    This paper presents a criticality evaluation of moderated homogeneous plutonium-uranium mixtures. The fissile media studied are homogeneous mixtures of plutonium and high enriched uranium in two chemical forms: aqueous mixtures of metal and mixtures of nitrate solutions. The enrichment of uranium considered are 93.2wt.% 235 U and 100wt.% 235 U. The 240 Pu content in plutonium varies from 0wt.% 240 Pu to 12wt.% 240 Pu. The critical parameters (radii and masses of a 20 cm water reflected sphere) are calculated with the French criticality safety package CRISTAL V0. The comparison of the calculated critical parameters as a function of the moderator-to-fuel atomic ratio shows significant ranges in which high enriched uranium systems, as well as plutonium-uranium mixtures, are more reactive than plutonium systems. (author)

  11. Method of removing uranium and its compounds from mine wastewaters and from aqueous solutions discharged in hydrometallurgical uranium ore treatment

    International Nuclear Information System (INIS)

    Jilek, R.; Prochazka, H.; Kuhr, I.; Fuska, J.; Nemec, P.; Katzer, J.

    1974-01-01

    The separation of uranium and its compounds from mine wastewaters and from water solutions discharged from uranium ore hydrometallurgical treatment, and its eventual simultaneous concentration in the biomass during uranium ore technological processing are described. The solutions are replenished with nutrients necessary for the growth of microorganisms, mainly with nitrogen, carbon and phosphorus and inoculated with fungi. During submersion cultivation, uranium incorporates in the mycelium, or is bound physico-chemically to the mycelium components. Together with these components, uranium is mechanically separated, i.e., by filtration, centrifugation or sedimentation. Organisms of the Fungi imperfecti class, mainly the Aspergillus and Penicillium genera are used for cultivation which may be continuous or semicontinuous. (B.S.)

  12. Study of the changes in composition of ammonium diuranate with progress of precipitation, and study of the properties of ammonium diuranate and its subsequent products produced from both uranyl nitrate and uranyl fluoride solutions

    Energy Technology Data Exchange (ETDEWEB)

    Manna, Subhankar; Kumar, Raj; Satpati, Santosh K.; Roy, Saswati B. [Bhabha Atomic Research Centre, Trombay, Mumbai (India); Joshi, Jyeshtharaj B. [Dept. of Chemical Engineering, Institute of Chemical Technology, Mumbai (India)

    2017-04-15

    Uranium metal used for fabrication of fuel for research reactors in India is generally produced by magnesio-thermic reduction of UF{sub 4}. Performance of magnesio-thermic reaction and recovery and quality of uranium largely depends on properties of UF{sub 4}. As ammonium diuranate (ADU) is first product in powder form in the process flow-sheet, properties of UF{sub 4} depend on properties of ADU. ADU is generally produced from uranyl nitrate solution (UNS) for natural uranium metal production and from uranyl fluoride solution (UFS) for low enriched uranium metal production. In present paper, ADU has been produced via both the routes. Variation of uranium recovery and crystal structure and composition of ADU with progress in precipitation reaction has been studied with special attention on first appearance of the precipitate Further, ADU produced by two routes have been calcined to UO{sub 3}, then reduced to UO{sub 2} and hydroflorinated to UF{sub 4}. Effect of two different process routes of ADU precipitation on the characteristics of ADU, UO{sub 3}, UO{sub 2} and UF{sub 4} were studied here.

  13. Mitigation of Hydrogen Gas Generation from the Reaction of Uranium Metal with Water in K Basin Sludge and Sludge Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-06-08

    Prior laboratory testing identified sodium nitrate and nitrite to be the most promising agents to minimize hydrogen generation from uranium metal aqueous corrosion in Hanford Site K Basin sludge. Of the two, nitrate was determined to be better because of higher chemical capacity, lower toxicity, more reliable efficacy, and fewer side reactions than nitrite. The present lab tests were run to determine if nitrate’s beneficial effects to lower H2 generation in simulated and genuine sludge continued for simulated sludge mixed with agents to immobilize water to help meet the Waste Isolation Pilot Plant (WIPP) waste acceptance drainable liquid criterion. Tests were run at ~60°C, 80°C, and 95°C using near spherical high-purity uranium metal beads and simulated sludge to emulate uranium-rich KW containerized sludge currently residing in engineered containers KW-210 and KW-220. Immobilization agents tested were Portland cement (PC), a commercial blend of PC with sepiolite clay (Aquaset II H), granulated sepiolite clay (Aquaset II G), and sepiolite clay powder (Aquaset II). In all cases except tests with Aquaset II G, the simulated sludge was mixed intimately with the immobilization agent before testing commenced. For the granulated Aquaset II G clay was added to the top of the settled sludge/solution mixture according to manufacturer application directions. The gas volumes and compositions, uranium metal corrosion mass losses, and nitrite, ammonia, and hydroxide concentrations in the interstitial solutions were measured. Uranium metal corrosion rates were compared with rates forecast from the known uranium metal anoxic water corrosion rate law. The ratios of the forecast to the observed rates were calculated to find the corrosion rate attenuation factors. Hydrogen quantities also were measured and compared with quantities expected based on non-attenuated H2 generation at the full forecast anoxic corrosion rate to arrive at H2 attenuation factors. The uranium metal

  14. Remotely operated facility for in situ solidification of fissile uranium

    International Nuclear Information System (INIS)

    McGinnis, C.P.; Collins, E.D.; Patton, B.D.

    1986-01-01

    A heavily shielded, remotely operated facility, located within the Radiochemical processing Plant at Oak Ridge National Laboratory (ORNL), has been designed and is being operated to convert approx.1000 kg of fissile uranium (containing approx.75% 235 U, approx.10% 233 U, and approx.140 ppM 232 U) from a nitrate solution (130 g of uranium per L) to a solid oxide form. This project, the Consolidated Edison Uranium Solidification Program (CEUSP), is being carried out in order to prepare a stable uranium form for longterm storage. This paper describes the solidification process selected, the equipment and facilities required, the experimental work performed to ensure successful operation, some problems that were solved, and the initial operations

  15. Spectrophotometric titrations: Application to the determination of some elements in uranium solutions

    International Nuclear Information System (INIS)

    L'Her, M.

    1967-01-01

    The aim of this work is the application of spectrophotometric titrations to the analysis of uranium-containing solutions. We have been led to examine the general principles involved in these titrations, and we give a brief outline of these principles. In the first part we deal therefore with spectrophotometric titrations from a general point of view, examining their fundamental principle, their practical execution as well as the various possibilities of the method. The advantage of the titration are examined, in particular that of lending itself simultaneous determination of two species. The possibility of applying these spectrophotometric titrations to the analysis of uranium-containing solutions is the subject of the second part of this report: the dosage of a few species in uranium (VI) solutions is described. To this second part is added an experimental appendix consisting of a description of the apparatus, as well as of the operational techniques used for certain titrations, in particular those involving solutions containing uranium. (author) [fr

  16. 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

  17. Process for denitrating waste solutions containing nitrates and actinides with simultaneous separation of the actinides

    International Nuclear Information System (INIS)

    Gompper, K.

    1986-01-01

    The invention is intended to reduce the acid and nitrate content of nitrate waste solutions, to reduce the total salt content of the waste solution, to remove the actinides contained in it by precipitation, without any danger of violent reactions or an increase in the volume of the waste solution. The invention achieves this by mixing the waste solution with diethyl oxalate at room temperature and heating the mixture to at least 80 0 C. (orig./PW) [de

  18. Development of system on predicting uranium concentration from pregnant solution

    International Nuclear Information System (INIS)

    Yi Weiping

    2004-01-01

    Uranium concentration from pregnant solution is primary index of process for in-situ leaching of uranium, and the suitable method with which to predicate this index and effective means to solve it with were continuously studied hard. SPUC-system on predicting uranium concentration based on GM model of gray system theory is developed, and the mathematical model, constitution, function and theory foundation of this system are introduced. (authors)

  19. Conversion and Blending Facility highly enriched uranium to low enriched uranium as uranyl nitrate hexahydrate. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-05

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

  20. Corrosion electrochemical behaviors of silane coating coated magnesium alloy in NaCl solution containing cerium nitrate

    Energy Technology Data Exchange (ETDEWEB)

    Luo, F.; Li, Q.; Zhong, X.K.; Gao, H.; Dai, Y.; Chen, F.N. [School of Chemistry and Chemical Engineering, Southwest University Chongqing (China)

    2012-02-15

    Sol-gel coatings cannot provide adequate corrosion protection for metal/alloys in the corrosive environments due to their high crack-forming potential. This paper demonstrates the possibility to employ cerium nitrate as inhibitor to decrease the corrosion development of sol-gel-based silane coating on the magnesium alloy in NaCl solution. Cerium nitrate was added into the NaCl solution where the silane coating coated magnesium alloy was immersed. Scanning electron microscopy (SEM) was used to examine surface morphology of the silane coating coated magnesium alloy immersed in NaCl solutions doped and undoped with cerium nitrate. The corrosion electrochemical behaviors were investigated using potentiodynamic polarization and electrochemical impedance spectroscopy (EIS) tests. The results showed that the introduction of cerium nitrate into NaCl solution could effectively inhibit the corrosion of the silane coating coated magnesium alloy. Moreover, the influence of concentration of cerium nitrate on the corrosion inhibition and the possible inhibiting mechanism were also discussed in detail. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  1. Can we predict uranium bioavailability based on soil parameters? Part 1: Effect of soil parameters on soil solution uranium concentration

    International Nuclear Information System (INIS)

    Vandenhove, H.; Hees, M. van; Wouters, K.; Wannijn, J.

    2007-01-01

    Present study aims to quantify the influence of soil parameters on soil solution uranium concentration for 238 U spiked soils. Eighteen soils collected under pasture were selected such that they covered a wide range for those parameters hypothesised as being potentially important in determining U sorption. Maximum soil solution uranium concentrations were observed at alkaline pH, high inorganic carbon content and low cation exchange capacity, organic matter content, clay content, amorphous Fe and phosphate levels. Except for the significant correlation between the solid-liquid distribution coefficients (K d , L kg -1 ) and the organic matter content (R 2 = 0.70) and amorphous Fe content (R 2 = 0.63), there was no single soil parameter significantly explaining the soil solution uranium concentration (which varied 100-fold). Above pH = 6, log(K d ) was linearly related with pH [log(K d ) = - 1.18 pH + 10.8, R 2 = 0.65]. Multiple linear regression analysis did result in improved predictions of the soil solution uranium concentration but the model was complex. - Uranium solubility in soil can be predicted from organic matter or amorphous iron content and pH or with complex multilinear models considering several soil parameters

  2. Improved fluorimetric measurement of uranium uptake and distribution in spring wheat (Triticum aestivum L.)

    Energy Technology Data Exchange (ETDEWEB)

    Borcia, Catalin [' ' Alexandru Ioan Cuza' ' Univ., Iasi (Romania). Dept. of Physics; Popa, Karin; Cecal, Alexandru [' ' Alexandru Ioan Cuza' ' Univ., Iasi (Romania). Dept. of Chemistry; Murariu, Manuela [' ' Petru Poni' ' Institute of Macromolecular Chemistry, Iasi (Romania)

    2016-08-01

    Uranium uptake and (radio)toxicity was tested on spring wheat (Triticum aestivum L.) in a laboratory study using differently concentrated uranium nitrate solutions. Within these experiments, two analytical assays of uranium were comparatively tested: a fast and improved fluorimetric assay and the classical colorimetric (U(IV)-arsenazo(III) complexation) one. During the germination, the wheat seeds and plantlets supported well the uranium solutions of treatment within the entire concentration range (1 x 10{sup -4} -5 x 10{sup -3} M). Uranium proved to be non (radio)toxic to wheat as compared with other natural and anthropogenic radiocations, probably because its uptake by spring wheat during the germination is low. Indeed, only a small fraction of uranium administered was located within the roots, whereas the uranium content of the stems was negligible. A high correlation between the results obtained by two analytical methods was found. However, the fluorimetric assay proved to be more reliable and fast, and accurate.

  3. Uranium and radium content in the soil solutions of the south-western part of Belarus

    International Nuclear Information System (INIS)

    Sokolik, G.A.; Ovsyannikova, S.V.; Vojnikova, E.V.; Popenya, M.V.

    2008-01-01

    The contents of uranium and radium in the pore soil solutions, which are the main chain in the geochemical and biological migration of the chemical elements, has been determined for the first time in Belarus. The control sites have been located outside the zone of Chernobyl fallout radionuclide contamination, that allowed evaluating the current background level of uranium and radium content in the soil solutions. The data on accumulation of the radioactive elements in the pore solutions give the opportunity to estimate the reserve of the radioactive elements in the migratory active forms in the soils. In the majority of soils studied, uranium content in the pore solution is higher than radium content, that points to the higher migratory ability of uranium. The direct correlation between content of fulvic acids' components in the soil solutions and accumulation of uranium in such solutions has been established. (authors)

  4. Electron spectra and mechanism of complexing of uranyl nitrate in water-acetone solutions

    International Nuclear Information System (INIS)

    Zazhogin, A.A.; Zazhogin, A.P.; Komyak, A.I.; Serafimovich, A.I.

    2003-01-01

    Based on the analysis of the luminescence and electronic absorption spectra, the processes of complexing in an aqueous solution of UO 2 (NO 3 ) 2 ·6H 2 O with small additions of acetone have been studied. In a pure aqueous solution, uranyl exists as the complex UO 2 ·5H 2 O. It is shown that the addition of acetone to the solution leads to the displacement of some water molecules out of the first coordination sphere of uranyl and the formation of the uranyl nitrate dihydrate complexes UO 2 (NO 3 ) 2 ·2H 2 O. It has been established that the stability of these complexes is determined by the decrease in the water activity and in the degree of hydration of uranyl and nitrate, which is the result of the local increase in the concentration of acetone molecules (due to their hydrophobicity) in the regions of the solution where uranyl and nitrate ions are found. The experimental facts supported the mechanism proposed are presented. (authors)

  5. Evaluation of adsorption of uranium from aqueous solution using biochar materials

    Energy Technology Data Exchange (ETDEWEB)

    Correa, Wagner Clayton; Guilhen, Sabine Neusatz; Ortiz, Nilce; Fungaro, Denise Alves, E-mail: wcorrea@ipen.br, E-mail: snguilhen@ipen.br, E-mail: notriz@ipen.br, E-mail: dfungaro@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Uranium is present in the environment as a result of leaching from natural deposits and activities associated with nuclear fuel, copper mining, uranium mining, milling industry, etc. For the purpose of protecting ecosystem stability and public health, it is crucial to eliminate uranium from aqueous solutions before they are discharged into the environment. Various technologies have been used for removing U(VI) ions from aqueous systems. Among these methods, adsorption has been applied in wastewater because of simple operation procedure and high removal efficiency. Brazil is the largest producer of charcoal in the world, with nearly half of the woody biomass harvested for energy in Brazil being transformed into charcoal. Biochar exhibits a great potential as an adsorbent because of favorable physical/chemical surface characteristics. The objective of this work was to evaluate the adsorption potential of biochar materials prepared from pyrolysis of Bamboo (CBM), Eucalyptus (CEM) and Macauba (CMA) nuts for the removal of uranium from solutions. Adsorption experiments were carried out by a batch technique. Equilibrium adsorption experiments were performed by shaking a known amount of biochar material with 100 mL of U(VI) solution in Erlenmeyer flasks in a shaker at 120 rpm and room temperature (25 deg C) for 24 h. The adsorbent was separated by centrifugation from the solution. The U(VI) concentration remaining in the supernatant solution was determined using inductively coupled plasma optical emission spectrometry (ICP-OES). The influences of different experimental parameters such as solution pH and bioadsorbent dose on adsorption were investigated. The highest uranium adsorption capacity were obtained at pH 3.0 and 16 g/L biomass dosage for CMA, pH 3.0 and 12 g/L biomass dosage for CBM and pH 2.0 and 10 g/L biomass dosage for CEM. The results demonstrated that the biomass derived char can be used as a low-cost adsorbent for removal of uranium from wastewater. (author)

  6. Evaluation of adsorption of uranium from aqueous solution using biochar materials

    International Nuclear Information System (INIS)

    Correa, Wagner Clayton; Guilhen, Sabine Neusatz; Ortiz, Nilce; Fungaro, Denise Alves

    2015-01-01

    Uranium is present in the environment as a result of leaching from natural deposits and activities associated with nuclear fuel, copper mining, uranium mining, milling industry, etc. For the purpose of protecting ecosystem stability and public health, it is crucial to eliminate uranium from aqueous solutions before they are discharged into the environment. Various technologies have been used for removing U(VI) ions from aqueous systems. Among these methods, adsorption has been applied in wastewater because of simple operation procedure and high removal efficiency. Brazil is the largest producer of charcoal in the world, with nearly half of the woody biomass harvested for energy in Brazil being transformed into charcoal. Biochar exhibits a great potential as an adsorbent because of favorable physical/chemical surface characteristics. The objective of this work was to evaluate the adsorption potential of biochar materials prepared from pyrolysis of Bamboo (CBM), Eucalyptus (CEM) and Macauba (CMA) nuts for the removal of uranium from solutions. Adsorption experiments were carried out by a batch technique. Equilibrium adsorption experiments were performed by shaking a known amount of biochar material with 100 mL of U(VI) solution in Erlenmeyer flasks in a shaker at 120 rpm and room temperature (25 deg C) for 24 h. The adsorbent was separated by centrifugation from the solution. The U(VI) concentration remaining in the supernatant solution was determined using inductively coupled plasma optical emission spectrometry (ICP-OES). The influences of different experimental parameters such as solution pH and bioadsorbent dose on adsorption were investigated. The highest uranium adsorption capacity were obtained at pH 3.0 and 16 g/L biomass dosage for CMA, pH 3.0 and 12 g/L biomass dosage for CBM and pH 2.0 and 10 g/L biomass dosage for CEM. The results demonstrated that the biomass derived char can be used as a low-cost adsorbent for removal of uranium from wastewater. (author)

  7. The research of technology and equipment for a microwave denitration process of the uranyl nitrate solution

    International Nuclear Information System (INIS)

    Bao Weimin; Wang Xuejun; Ma Xuquan; Shi Miaoyi; Zhang Zhicheng; Bao Zhu Tian.

    1991-01-01

    In order to improve the present process of converting the plutonium nitrate into oxide powder in the nuclear fuel cycle, a new conversion process for the direct denitration using microwave heating has been developed. Microwave denitration is based on intramolecular polarization of a material in electric field and has no need of a process of heat transfer during microwave heating, so that the whole material can be heated quickly and uniformly. The thermal decomposition reactions of Pu, U, Th and RE nitrate have been analyzed and compared. The uranyl nitrate solution was chosen as imitative plutonium nitrate solution. The performance parameters ε r tanδ of U, Th and RE nitrate and oxide in microwave field were measured. The data obtained show that all of them could absorb microwave energy well and cause heating decomposition reactions. The microwave denitration test unit was designed and made. Denitration tests for rare-earths nitrate and uranyl nitrate solutions were performed. It could be completed in one step that the uranyl nitrate solution was evaporated, dryed and denitrated in a vessel. The denitrated products are a porous lump and easy to scrape off from the denitration vessel. The main forms of the products UO 3 ·0.8H 2 O and U 3 O 8 which have excellent powder properties. The capacity of the denitration unit is 1.3 kg UO 3 /h. According to the experimental results the simplicity, feasibility and good repeatability of the process have been fully proved. The unit operates easily and is adaptable to conversion of nitrate in nuclear fuel cycle. (author)

  8. Extraction and Separation of Uranium (VI) and Thorium (IV) Using Tri-n-dodecylamine Impregnated Resins

    International Nuclear Information System (INIS)

    Metwally, E.; Saleh, A.Sh.; El-Naggar, H.A.

    2005-01-01

    Extraction of U(VI) and Th(IV) from chloride and nitrate solutions with tri-n- dodecylamine impregnated on Amberlite XAD4, was investigated. The distribution of U(VI) and Th(IV) was studied at different concentrations of acid, salting-out agent, extractant, aqueous metal ion and other parameters. Absorption spectral studies have been investigated for uranium species in both aqueous HCl solution and the resin phase. From these studies, it is suggested that the tetrachloro complex of U(VI) is formed in the extraction of uranium (VI) from hydrochloric acid solutions by TDA impregnated resin. Stripping of the extracted U(VI) and Th(IV) was assayed with HCl and HNO 3 . Finally, the separation of uranium from thorium and fission products in HCl media was achieved

  9. PROCESS FOR RECOVERY OF URANIUM VALUES FROM IMPURE SOLUTIONS THEREOF

    Science.gov (United States)

    Kilner, S.B.

    1959-11-01

    A process is presented for the recovery of uraninm values from impure solutions which are obtained, for example, by washing residual uranium salt or uranium metal deposits from stainless steel surfaces using an aqueous or certain acidic aqueous solutions. The solutions include uranyl and oxidized iron, chromium, nickel, and copper ions and may contain manganese, zinc, and silver ions. In accordance with one procedure. the uranyl ions are reduced to the uranous state, and the impurity ions are complexed with cyanide under acidic conditions. The solution is then treated with ammonium hydroxide or alkali metal hydroxide to precipitate uranous hydroxide away from the complexed impurity ions in the solution. Alternatively, an excess of alkali metal cyanide is added to the reduced solution until the solution becomes sufficiently alkaline for the uranons hydroxide to precipitate. An essential feature in operating the process is in maintaining the pH of the solution sufficiently acid during the complexing operation to prevent the precipitation of the impurity metal hydroxides.

  10. Measurement of uranium concentration by molecular absorption spectrophotometry by means optical fibers

    International Nuclear Information System (INIS)

    Gauna, Alberto C.; Pascale, Ariel A.

    1996-01-01

    An on-line method for measuring the concentration of uranium in uranyl nitrate-nitric acid aqueous solutions is described. The method is based on molecular absorption spectrophotometry with transmission of light by means of optical fibers. It is ideally suited for control and processes development applications. (author)

  11. Composition and method for solution mining of uranium ores

    International Nuclear Information System (INIS)

    Lawes, B.C.; Watts, J.C.

    1981-01-01

    It has been found that, in the solution mining of uranium ores using ammonium carbonate solutions containing hydrogen peroxide or ozone as an oxidant, the tendency of the formation being treated to become less permeable during the leaching process can be overcome by including in the leaching solution a very small concentration of sodium silicate

  12. Uranium chemistry in stack solutions and leachates of phosphogypsum disposed at a coastal area in Cyprus.

    Science.gov (United States)

    Lysandrou, M; Pashalidis, I

    2008-02-01

    The effect of the matrix composition (main constituents) on the concentration and chemical behavior of uranium in phosphogypsum stack solutions and leachates has been investigated. Solid and aqueous samples were taken from three different sub-areas of a phosphogypsum stack at a coastal area in Vasilikos (Cyprus). The sub-areas are characterized whether by their acidity (e.g. "aged" and "non-aged" phosphogypsum) or by their salt content, originating from pulping water during wet stacking or (after deposition) from the adjacent sea. Measurements in stack solutions and leachates showed that phosphogypsum characteristics affect both, the concentration and the chemical behavior of uranium in solution. Uranium concentration in solutions of increased salinity is up to three orders of magnitude higher than in solutions of low salinity and this is attributed to the effect of ionic strength on the solubility of phosphogypsum. Modelling showed that uranium in stack solutions is predominantly present in the form of uranium(VI) phosphate complexes (e.g. UO(2)(H(2)PO(4))(2), UO(2)HPO(4)), whereas in leachates uranium(VI) fluoro complexes (e.g. UO(2)F(2), UO(2)F(3)(-)) are predominant in solution. The latter indicates that elution of uranium from phosphogypsum takes places most probably in the form of fluoro complexes. Both, effective elution by saline water and direct migration of uranium to the sea, where it forms very stable uranium(VI) carbonato complexes, indicate that the adjacent sea will be the final receptor of uranium released from Vasilikos phosphogypsum.

  13. Nitrate adsorption from aqueous solution using granular chitosan-Fe{sup 3+} complex

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Qili [School of Water Resources and Environment, China University of Geosciences (Beijing), Beijing, 100083 (China); Key Laboratory of Groundwater Cycle and Environment Evolution,China University of Geosciences (Beijing), Ministry of Education, Beijing, 100083 (China); Chen, Nan, E-mail: chennan@cugb.edu.cn [School of Water Resources and Environment, China University of Geosciences (Beijing), Beijing, 100083 (China); Key Laboratory of Groundwater Cycle and Environment Evolution, China University of Geosciences (Beijing), Ministry of Education, Beijing, 100083 (China); Feng, Chuanping [School of Water Resources and Environment, China University of Geosciences (Beijing), Beijing, 100083 (China); Key Laboratory of Groundwater Cycle and Environment Evolution, China University of Geosciences (Beijing), Ministry of Education, Beijing, 100083 (China); Hu, WeiWu [The Journal Center, China University of Geosciences (Beijing), Beijing, 100083 (China)

    2015-08-30

    Highlights: • Granular chitosan-Fe{sup 3+} complex had high performance for nitrate adsorption. • Granular chitosan-Fe{sup 3+} complex had shorter equilibrium time (1.5 h). • Nitrate adsorption was ascribed to ion exchange and electrostatic attraction. • Granular chitosan-Fe{sup 3+} complex could be regenerated using NaCl solution. - Abstract: In the present study, In order to efficiently remove nitrate, granular chitosan-Fe{sup 3+} complex with high chemical stability and good environmental adaptation was synthesized through precipitation method and characterized using SEM, XRD, BET and FTIR. The nitrate adsorption performance was evaluated by batch experiments. The results indicated that granular chitosan-Fe{sup 3+} complex was an amorphous and mesoporous material. The BET specific surface area and average pore size were 8.98 m{sup 2} g{sup −1} and 56.94 Å, respectively. The point of zero charge was obtained at pH 5. The maximum adsorption capacity reached 8.35 mg NO{sub 3}{sup −}-N g{sup −1} based on Langmuir–Freundlich model. Moreover, no significant change in the nitrate removal efficiency was observed in the pH range of 3.0–10.0. The adverse influence of sulphate on nitrate removal was the most significant, followed by bicarbonate and fluoride, whereas chloride had slightly adverse effect. Adsorption process followed the pseudo-second-order kinetic model, and the experimental equilibrium data were fitted well with the Langmuir–Freundlich and D–R isotherm models. Thermodynamic parameters revealed that nitrate adsorption was a spontaneous and exothermic process. Granular chitosan-Fe{sup 3+} complex could be effectively regenerated by NaCl solution.

  14. The separation of uranium ions by natural and modified diatomite from aqueous solution

    Energy Technology Data Exchange (ETDEWEB)

    Sprynskyy, Myroslav, E-mail: sprynsky@yahoo.com [Department of Environmental Chemistry and Bioanalytics, Faculty of Chemistry, Nicolaus Copernicus University, 7 Gagarina Str., 87-100 Torun (Poland); Kovalchuk, Iryna [Department of Environmental Chemistry and Bioanalytics, Faculty of Chemistry, Nicolaus Copernicus University, 7 Gagarina Str., 87-100 Torun (Poland); Institute of Adsorption and Problem of Endoecology, National Academy of Sciences of Ukraine, 13 General Naumov Str., 03164 Kyiv (Ukraine); Buszewski, Boguslaw [Department of Environmental Chemistry and Bioanalytics, Faculty of Chemistry, Nicolaus Copernicus University, 7 Gagarina Str., 87-100 Torun (Poland)

    2010-09-15

    In this work the natural and the surfactant modified diatomite has been tested for ability to remove uranium ions from aqueous solutions. Such controlling factors of the adsorption process as initial uranium concentration, pH, contact time and ionic strength have been investigated. Effect of ionic strength of solution has been examined using the solutions of NaCl, Na{sub 2}CO{sub 3} and K{sub 2}SO{sub 4}. The pseudo-first order and the pseudo-second order models have been used to analyze the adsorption kinetic results, whereas the Langmuir and the Freundlich isotherms have been used to the equilibrium adsorption data. The effects of the adsorbent modification as well as uranium adsorption on the diatomite surface have been studied using X-ray powder diffraction, scanning electron microscopy and FTIR spectroscopy. The maximum adsorption capacities of the natural and the modified diatomite towards uranium were 25.63 {mu}mol/g and 667.40 {mu}mol/g, respectively. The desorptive solutions of HCl, NaOH, Na{sub 2}CO{sub 3}, K{sub 2}SO{sub 4}, CaCO{sub 3}, humic acid, cool and hot water have been tested to recover uranium from the adsorbent. The highest values of uranium desorption (86%) have been reached using 0.1 M HCl.

  15. The separation of uranium ions by natural and modified diatomite from aqueous solution.

    Science.gov (United States)

    Sprynskyy, Myroslav; Kovalchuk, Iryna; Buszewski, Bogusław

    2010-09-15

    In this work the natural and the surfactant modified diatomite has been tested for ability to remove uranium ions from aqueous solutions. Such controlling factors of the adsorption process as initial uranium concentration, pH, contact time and ionic strength have been investigated. Effect of ionic strength of solution has been examined using the solutions of NaCl, Na(2)CO(3) and K(2)SO(4). The pseudo-first order and the pseudo-second order models have been used to analyze the adsorption kinetic results, whereas the Langmuir and the Freundlich isotherms have been used to the equilibrium adsorption data. The effects of the adsorbent modification as well as uranium adsorption on the diatomite surface have been studied using X-ray powder diffraction, scanning electron microscopy and FTIR spectroscopy. The maximum adsorption capacities of the natural and the modified diatomite towards uranium were 25.63 micromol/g and 667.40 micromol/g, respectively. The desorptive solutions of HCl, NaOH, Na(2)CO(3), K(2)SO(4), CaCO(3), humic acid, cool and hot water have been tested to recover uranium from the adsorbent. The highest values of uranium desorption (86%) have been reached using 0.1M HCl. Copyright 2010 Elsevier B.V. All rights reserved.

  16. Physicochemical analysis of cryocrystallization processes of aqueous solutions of yttrium, barium, copper nitrates and their mixtures

    International Nuclear Information System (INIS)

    Kulakov, A.B.; Mozhaev, A.P.; Tesker, A.M.; Churagulov, B.R.

    1992-01-01

    Products of fast hardening of aqueous solutions of different concentration of yttrium, barium copper nitrates and their mixtures including mixture of three nitrates with molar ratio equal to 1:2:3 used for synthesis of YBa 2 Cu 3 O 7-x HTSC by cryochemical technique, in liquid nitrogen, are studied using low-temperature, differential thermal and X-ray phase analyses. Aqueous solutions of barium, copper, yttrium nitrates are shown to belong to three different classes which differ in behaviour at fast cooling and subsequent slow heating. Cryogranulate at YBa 2 Cu 2 O 7-x synthesis using cryochemical technique represents mixture of X-ray amorphous Ba(NO 3 ) 2 , crystalline Cu(NO 3 ) 2 ·6H 2 O and ice, as well as, supercooled aqueous solution of yttrium and copper nitrates

  17. Spectrophotometric titrations: Application to the determination of some elements in uranium solutions; Les titrages spectrophotometriques: Application a la determination de quelques elements dans les solutions d'uranium

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-01-01

    The aim of this work is the application of spectrophotometric titrations to the analysis of uranium-containing solutions. We have been led to examine the general principles involved in these titrations, and we give a brief outline of these principles. In the first part we deal therefore with spectrophotometric titrations from a general point of view, examining their fundamental principle, their practical execution as well as the various possibilities of the method. The advantage of the titration are examined, in particular that of lending itself simultaneous determination of two species. The possibility of applying these spectrophotometric titrations to the analysis of uranium-containing solutions is the subject of the second part of this report: the dosage of a few species in uranium (VI) solutions is described. To this second part is added an experimental appendix consisting of a description of the apparatus, as well as of the operational techniques used for certain titrations, in particular those involving solutions containing uranium. (author) [French] Le but de ce travail est l'application des titrages spectrophotometriques a l'analyse des solutions uraniferes. Nous avons ete amenes a examiner les principes generaux de ces titrages, principes qu'il nous est apparu necessaire de rappeler. Dans une premiere partie nous traiterons donc d'une facon generale des titrages spectrophotometriques, en examinant leur principe fondamental, leur mise en oeuvre ainsi que les possibilites diverses de dosage. Nous examinerons aussi les avantages de la methode de titrage, en insistant notamment sur la possibilite de faire des dosages successifs. La possibilite d'application de ces titrages spectrophotometriques a l'analyse des solutions uraniferes sera le sujet de la deuxieme partie: nous y decrivons le dosage de quelques especes, dans les solutions d'uranium (VI). A cette deuxieme partie nous joindrons une annexe experimentale comportant une description de l'appareillage que

  18. Solubility Limits of Dibutyl Phosphoric Acid in Uranium Solutions at SRS

    International Nuclear Information System (INIS)

    Thompson, M.C.; Pierce, R.A.; Ray, R.J.

    1998-06-01

    The Savannah River Site has enriched uranium (EU) solution which has been stored for almost 10 years since being purified in the second uranium cycle of the H area solvent extraction process. The concentrations in solution are 6 g/L U and about 0.1 M nitric acid. Residual tributylphosphate in the solutions has slowly hydrolyzed to form dibutyl phosphoric acid (HDBP) at concentrations averaging 50 mg/L. Uranium is known to form compounds with DBP which have limited solubility. The potential to form uranium-DBP solids raises a nuclear criticality safety issue. SRTC tests have shown that U-DBP solids will precipitate at concentrations potentially attainable during storage of enriched uranium solutions. Evaporation of the existing EUS solution without additional acidification could result in the precipitation of U-DBP solids if DBP concentration in the resulting solution exceeds 110 ppm at ambient temperature. The same potential exists for evaporation of unwashed 1CU solutions. The most important variables of interest for present plant operations are HNO 3 and DBP concentrations. Temperature is also an important variable controlling precipitation. The data obtained in these tests can be used to set operating and safety limits for the plant. It is recommended that the data for 0 degrees C with 0.5 M HNO 3 be used for setting the limits. The limit would be 80 mg/L which is 3 standard deviations below the average of 86 observed in the tests. The data shows that super-saturation can occur when the DBP concentration is as much as 50 percent above the solubility limit. However, super-saturation cannot be relied on for maintaining nuclear criticality safety. The analytical method for determining DBP concentration in U solutions was improved so that analyses for a solution are accurate to within 10 percent. However, the overall uncertainty of results for periodic samples of the existing EUS solutions was only reduced slightly. Thus, sampling appears to be the largest portion

  19. Supercritical fluid extraction of uranium and neodymium nitrates

    International Nuclear Information System (INIS)

    Sujatha, K.; Sivaraman, N.; Srinivasan, T.G.; Vasudeva Rao, P.R.

    2011-01-01

    Supercritical fluid extraction (SFE) of uranyl nitrate and neodymium nitrate salts from a mixture was investigated in the present study using Sc-CO 2 modified with various ligands such as organophosphorous compounds, amides, and diketones. Preferential extraction of uranyl nitrate over neodymium nitrate was demonstrated using Sc-CO 2 modified with amide, di-(2ethylhexyl) isobutyramide (D2EHIBA). (author)

  20. Can we predict uranium bioavailability based on soil parameters? Part 1: effect of soil parameters on soil solution uranium concentration.

    Science.gov (United States)

    Vandenhove, H; Van Hees, M; Wouters, K; Wannijn, J

    2007-01-01

    Present study aims to quantify the influence of soil parameters on soil solution uranium concentration for (238)U spiked soils. Eighteen soils collected under pasture were selected such that they covered a wide range for those parameters hypothesised as being potentially important in determining U sorption. Maximum soil solution uranium concentrations were observed at alkaline pH, high inorganic carbon content and low cation exchange capacity, organic matter content, clay content, amorphous Fe and phosphate levels. Except for the significant correlation between the solid-liquid distribution coefficients (K(d), L kg(-1)) and the organic matter content (R(2)=0.70) and amorphous Fe content (R(2)=0.63), there was no single soil parameter significantly explaining the soil solution uranium concentration (which varied 100-fold). Above pH=6, log(K(d)) was linearly related with pH [log(K(d))=-1.18 pH+10.8, R(2)=0.65]. Multiple linear regression analysis did result in improved predictions of the soil solution uranium concentration but the model was complex.

  1. Can we predict uranium bioavailability based on soil parameters? Part 2: soil solution uranium concentration is not a good bioavailability index.

    Science.gov (United States)

    Vandenhove, H; Van Hees, M; Wannijn, J; Wouters, K; Wang, L

    2007-01-01

    The present study aimed to quantify the influence of soil parameters on uranium uptake by ryegrass. Ryegrass was established on eighteen distinct soils, spiked with (238)U. Uranium soil-to-plant transfer factors (TF) ranged from 0.0003 to 0.0340kgkg(-1). There was no significant relation between the U soil-to-plant transfer (or total U uptake or flux) and the uranium concentration in the soil solution or any other soil factor measured, nor with the U recovered following selective soil extractions. Multiple linear regression analysis resulted in a significant though complex model explaining up to 99% of variation in TF. The influence of uranium speciation on uranium uptake observed was featured: UO(2)(+2), uranyl carbonate complexes and UO(2)PO(4)(-) seem the U species being preferentially taken up by the roots and transferred to the shoots. Improved correlations were obtained when relating the uranium TF with the summed soil solution concentrations of mentioned uranium species.

  2. Removal of nitrate from ammonium hydroxide solution containing organics by ion exchange method

    International Nuclear Information System (INIS)

    Venugopal Chetty, K.; Gamare, Jayashree S.; Vaidya, V.N.

    2004-01-01

    Removal of nitrate from ammonium hydroxide solution containing HMTA (hexamethyltetramine) and Urea was studied using indigenously available anion exchange resins. This type of waste is produced during nuclear fuel preparation by internal gelation process. The resins used are Tulsion A-27(MP) and Duolite A. 102D. The time of equilibration and capacity of the resins were determined from distribution ratios obtained by equilibrating resin with nitrate solution. The loading, washing and elution behavior of nitrate on these resins were studied using synthetic mixture having similar composition of the waste produced. Elution studies were carried out using sodium hydroxide, hydrochloric acid and ammonium chloride. The studies were also carried out at higher temperature of around 60 degC. The data was compared with that obtained using Dowex 1x4 for the same purpose. (author)

  3. Field evaporation test of uranium tailings solution

    International Nuclear Information System (INIS)

    Chandler, B.L.; Shepard, T.A.; Stewart, T.A.

    1985-01-01

    A field experiment was performed to observe the effect on evaporation rate of a uranium tailings impoundment pond water as salt concentration of the water increased. The duration of the experiment was long enough to cause maximum salt concentration of the water to be attained. The solution used in the experiment was tailings pond water from an inactive uranium tailings disposal site in the initial stages of reclamation. The solution was not neutralized. The initial pH was about 1.0 decreasing to a salt gel at the end of the test. The results of the field experiment show a gradual and slight decrease in evaporation efficiency. This resulted as salt concentrations increased and verified the practical effectiveness of evaporation as a water removal method. In addition, the physical and chemical nature of the residual salts suggest that no long-term stability problem would likely result due to their presence in the impoundment during or after reclamation

  4. Linking specific heterotrophic bacterial populations to bioreduction of uranium and nitrate using stable isotope probing in contaminated subsurface sediments

    International Nuclear Information System (INIS)

    Akob, Denise M.; Kerkhof, Lee; Kusel, Kirsten; Watson, David B.; Palumbo, Anthony Vito; Kostka, Joel

    2011-01-01

    Shifts in terminal electron-accepting processes during biostimulation of uranium-contaminated sediments were linked to the composition of stimulated microbial populations using DNA-based stable isotope probing. Nitrate reduction preceded U(VI) and Fe(III) reduction in [ 13 C]ethanol-amended microcosms. The predominant, active denitrifying microbial groups were identified as members of the Betaproteobacteria, whereas Actinobacteria dominated under metal-reducing conditions.

  5. First start-up of nuclear criticality safety experiment facility for uranyl nitrate solution

    International Nuclear Information System (INIS)

    Zhu Qingfu; Shi Yongqian; Shen Leisheng; Hu Dingsheng; Zhao Shouzhi; He Tao; Sun Zheng; Lin Shenghuo; Yao Shigui

    2005-01-01

    The uranyl nitrate solution experiment facility for the research on nuclear criticality safety is described. The nuclear fuel loading steps in the first start-up for water-reflected core are presented. During the experiments, the critical volume of uranyl nitrate solution was determined as 20479.62 mL with count rate inverse extrapolation method, reactivity interpolation method, and steady power method. By calculation, critical mass of 235 U was derived as 1579.184 g from experimental data. The worth of control rods was also calibrated in the first start-up of the facility. (authors)

  6. Laboratory evaluation of the hydrogen sulfide gas treatment approach for remediation of chromate-, uranium(VI)-, and nitrate-contaminated soils

    International Nuclear Information System (INIS)

    Thornton, E.C.; Baechler, M.A.; Beck, M.A.; Amonette, J.E.

    1994-08-01

    Bench-scale soil treatment tests were conducted as part of an effort to develop and implement an in situ chemical treatment approach to the remediation of metal and radionuclide contaminated soils through the use of reactive gases. In general, > 90% immobilization of chromium and > 50% immobilization of uranium was achieved. Leach test results indicate that the treatment process is irreversible for chromium but partially reversible for uranium indicates that immobilization for this contaminant is more readily achieved in organic rich soils. This observation is ascribed to the reducing nature of organic matter. Additional tests were also conducted with soils contaminated to the 5,000 ppm level with nitrate. Nitrate was not found to interfere significantly with treatment of the contaminants. Nitrite was observed in the leachate samples obtained from tests with an organic-rich soil containing clay, however. Leachate chemistries suggested that no other significantly hazardous byproducts were generated by the treatment process and that soil alteration effects were minimal. Test results also suggest that treatment effectiveness is somewhat lower in very dry soils but still able to immobilize chromium and uranium to an acceptable degree. Results of these testing activities indicate that the concentration of hydrogen sulfide in the gas mixture is not a limited factor in treatment as long as a sufficient volume of the mixture is delivered to the soil to achieve a mole ratio of hydrogen sulfide to contaminant of at least 10

  7. Preferential solvation, ion pairing, and dynamics of concentrated aqueous solutions of divalent metal nitrate salts

    Science.gov (United States)

    Yadav, Sushma; Chandra, Amalendu

    2017-12-01

    We have investigated the characteristics of preferential solvation of ions, structure of solvation shells, ion pairing, and dynamics of aqueous solutions of divalent alkaline-earth metal nitrate salts at varying concentration by means of molecular dynamics simulations. Hydration shell structures and the extent of preferential solvation of the metal and nitrate ions in the solutions are investigated through calculations of radial distribution functions, tetrahedral ordering, and also spatial distribution functions. The Mg2+ ions are found to form solvent separated ion-pairs while the Ca2+ and Sr2+ ions form contact ion pairs with the nitrate ions. These findings are further corroborated by excess coordination numbers calculated through Kirkwood-Buff G factors for different ion-ion and ion-water pairs. The ion-pairing propensity is found to be in the order of Mg(NO3) 2 lead to the presence of substantial dynamical heterogeneity in these solutions of strongly interacting ions. The current study helps us to understand the molecular details of hydration structure, ion pairing, and dynamics of water in the solvation shells and also of ion diffusion in aqueous solutions of divalent metal nitrate salts.

  8. Optimization of the recycling process of precipitation barren solution in a uranium mine

    International Nuclear Information System (INIS)

    Long Qing; Yu Suqin; Zhao Wucheng; Han Wei; Zhang Hui; Chen Shuangxi

    2014-01-01

    Alkaline leaching process was adopted to recover uranium from ores in a uranium mine, and high concentration uranium solution, which would be later used in precipitation, was obtained after ion-exchange and elution steps. The eluting agent consisted of NaCl and NaHCO 3 . Though precipitation barren solution contained as high as 80 g/L Na 2 CO 3 , it still can not be recycled due to presence of high Cl - concentration So, both elution and precipitation processes were optimized in order to control the Cl - concentration in the precipitation barren solution to the recyclable concentration range. Because the precipitation barren solution can be recycled by optimization, the agent consumption was lowered and the discharge of waste water was reduced. (authors)

  9. Critical experiments for large scale enriched uranium solution handling

    International Nuclear Information System (INIS)

    Tanner, J.E.; Forehand, H.M.

    1985-01-01

    The authors have performed 17 critical experiments with a concentrated aqueous uranyl nitrate solution contained in an annular cylindrical tank, with annular cylindrical absorbers of stainless steel and/or polyethylene inside. k/sub eff/ calculated by KENO IV, employing 16-group Hansen-Roach cross sections, average 0.977. There is a variation of the calculational bias among the separate experiments, but it is too small to allow assigning it to specific components of the equipment. They are now performing critical experiments with a more concentrated uranyl nitrate solution in pairs of very squat cylindrical tanks with disc shaped absorbers and reflectors of carbon steel, stainless steel, nitronic-50, plain and borated polyethylene. These experiments are in support of upgrading fuel reprocessing at the Idaho Chemical Processing Plant

  10. Investigation of uranium sorption from carbonate solutions by different ion exchange materials

    International Nuclear Information System (INIS)

    Nekrasova, N.A.; Kudryavtseva, S.P.; Milyutin, V.V.; Chuveleva, Eh.A.; Firsova, L.A.; Gelis, V.M.

    2008-01-01

    One studied the uranium sorption from the reference carbonate solutions based on the ion-exchange resins varying in the rank. The PFA-300, the A-560, the AB-17x8 highly basic anionites and the ampholytes (S-930, S-922, S-957, ANKB-35) were shown to manifest the best sorption characteristics as to U. One determined the dependences of the static exchange capacity of the PFA-300, the A-560 and the S-922 resins as to the uranium on the carbonate solution pH, as well as the absorbed uranium desorption conditions [ru

  11. Activation of Graphene Oxide with Hydrochloric Acid for Nitrate Removal from Aqueous Solutions

    Directory of Open Access Journals (Sweden)

    Abolghasem Alighardashi

    2017-11-01

    Full Text Available Long-term drinking of nitrate-contaminated water poses a serious risk to human health. The present study explores the possibility of enhancing the adsorption capacity of graphene oxide via activation with hydrochloric acid for nitrate removal from aqueous solutions. Experiments were performed in a batch reactor in which such major factors as pH, reaction time, and concentrations of both graphene oxide (GO and activated graphene oxide (AGO were used as variables. Nitrate removal efficiency was investigated using the One-Way ANOVA statistical test and SPSS-16 software. The chemical composition and solid structure of the synthesized AGO were analyzed using FE-SEM coupled with energy dispersive spectrometry (EDS. The micropore volumes of the samples were determined using the BET and BJH. The predominant composition (52% of the synthesized AGO was C and its mean pore diameter was 26.896 nm. The maximum adsorption capacity of AGO was estimated at 3333.33 mg/g. Based on the results, the AGO nano-structure may be recomended as a new means for nitrate removal from aqueous solutions.

  12. Application of solution-mineral equilibrium chemistry to solution mining of uranium ores

    International Nuclear Information System (INIS)

    Riese, A.C.; Popp, C.J.

    1979-01-01

    The tests described were undertaken to determine the extent to leach solution-rock interactions with uranium-bearing ore obtained from the Mariano Lake mine. Leach solutions of an acidic (H/sub 2/O/sub 4/-sulfuric acid) and basic (NaHCO/sub 3/-sodium bicarbonate) nature were tested, in addition to a leach solution containing potassium chloride and sulfuric acid (KCl/H/sub 2/SO/sub 4/). The latter solution was chosen in an attempt to equilibrate the aqueous phase with the rock-forming silicate minerals and minimize adverse effects such as clay formation, porosity loss, and lixiviant loss. 29 refs

  13. Electronic spectra of plutonium ions in nitric acid and in lithium nitrate solutions

    International Nuclear Information System (INIS)

    Mekhail, F.M.

    1987-01-01

    The absorption spectra of plutonium ions in nitric acid have been described. There is a characteristic change in the absorption spectra of Pu v in lithium nitrate solutions. In 2 M-lithium nitrate a new peak at 969 nm and high absorption at 1200 nm are noticed. A decrease in the absorption by about 20% and the appearance of a new shoulder at 1120 nm in 6 M-lithium nitrate are found. There is no change in the spectrum in 4 M-lithium nitrate. The absorption spectra of plutonium ions in the spectral range 200 - 400 nm are interesting. All plutonium ions have an intense band in the region 250 - 260 nm as well as a less intense and rather diffuse band at 320 - 330 nm in lithium nitrate solutions the sharp band at 250 - 260 nm has disappeared. This suggests that this band is very sensitive to the environmental field. The band is probably produced by 5 F q → 5 f q-1 6 d transition as well as electron transfer. It is believed that the spectrum of Pu V at pH 6.5 represents the hydrolysis product Pu O 2 (O H). 9 fig., 4 tab

  14. 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.

  15. Synthesis of microspheres of triuranium octaoxide by simultaneous water and nitrate extraction from ascorbate-uranyl sols.

    Science.gov (United States)

    Brykala, M; Deptula, A; Rogowski, M; Lada, W; Olczak, T; Wawszczak, D; Smolinski, T; Wojtowicz, P; Modolo, G

    A new method for synthesis of uranium oxide microspheres (diameter nitrate-ascorbate sols were prepared by addition of ascorbic acid to uranyl nitrate hexahydrate solution and alkalizing by aqueous ammonium hydroxide and then emulsified in 2-ethylhexanol-1 containing 1v/o SPAN-80. Drops of emulsion were firstly gelled by extraction of water by the solvent. Destruction of the microspheres during thermal treatment, owing to highly reactive components in the gels, requires modification of the gelation step by Double Extraction Process-simultaneously extraction of water and nitrates using Primene JMT, which completely eliminates these problem. Final step was calcination in air of obtained microspheres of gels to triuranium octaoxide.

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  17. stripping of uranium from DEHPA/TOPO solvent by ammonium carbonate solutions

    International Nuclear Information System (INIS)

    Khorfan, S.; Shino, O.; Wahood, A.; Dahdouh, A.

    2002-01-01

    Uranium is recovered from phosphoric acid by the DEHPA/TOPO process. In this process uranium is stripped from the loaded DEHPA/TOPO solvent in the second cycle by an ammonium carbonate solution. This paper studied stripping of uranium from 0.3 Mol DEHPA/0.075 Mol TOPO in kerosene by different ammonium carbonate solutions. The ammonium carbonate solutions tested were either made locally from ammonia and carbon dioxide gases or commercial and laboratory grades available on the market. A comparison was made between these carbonate solutions in terms of purity, stripping efficiency and phase separation. Both stripping and phase separation were carried out under different conditions of phase ratio and concentrations. The results obtained showed that ammonium carbonate prepared from direct synthesis of ammonia and carbon dioxide gases had a high purity and gave the same stripping yield as the laboratory grade. The phase separation was also slightly improved using a pure synthesized ammonium carbonate solution. the phase separation was found to be best at concentration of 0.5 Mol/L ammonium carbonate solution and at a phase A/O of 1/1 and a temperature of 50 degree centigrade. It was possible to obtain >99% yield by operating 2 stripping stages counter currently under these conditions. (authors)

  18. Removal of phosphate and nitrate from aqueous solution using ...

    African Journals Online (AJOL)

    sunny t

    water, 3.5 g of NaCl were dissolved to obtain 3.5 g/l salinity final solution. When the ... The nitrate adsorption was highly pH dependent, which affects the ... adsorption mechanism that the optimum pH for phosphate removal by .... Biosorption of copper(ii) from aqueous ... Accumulation and detoxification of toxic elements by ...

  19. Highland Uranium Solution Mining Project. Draft environmental statement

    International Nuclear Information System (INIS)

    1978-05-01

    Exxon Minerals Co. proposes to conduct production-scale solution mining (in situ leaching) of uranium within the existing Highland Operation area in Converse County, Wyoming. The project would result in the temporary removal of 200 acres from grazing and the actual disturbance of 50 acres. About 4500 acre-ft of water will be withdrawn from the ore zone aquifer over the 10-year life of the project. There will be no discharge of liquid effluents from the project; atmospheric effluents will be within acceptable limits. Radiation doses at the nearest ranch resulting from solution mining activities were estimated. The project proposes the production and utilization of 1 to 3 million lb of uranium resources. It will not produce any significant socioeconomic impact on the local area. Alternatives to the project were considered, and conditions for issuing the source material license are listed

  20. Hydrolysis-precipitation studies of aluminum (III) solutions. I. Titration of acidified aluminum nitrate solutions

    NARCIS (Netherlands)

    Vermeulen, A.C.; Geus, John W.; Stol, R.J.; Bruyn, P.L. de

    Acidified aluminum nitrate solutions were titrated with alkali (NaOH or KOH) over a temperature range of 24°C to 90°C. A homogeneous distribution of added base was achieved by: (i) in situ decomposition of urea (90°C); and (ii) a novel method involving injection through a capillary submerged in the

  1. Electrical conductivity of solutions of copper(II) nitrate crystalohydrate in dimethyl sulfoxide

    Science.gov (United States)

    Mamyrbekova, Aigul K.; Mamitova, A. D.; Mamyrbekova, Aizhan K.

    2016-06-01

    Conductometry is used to investigate the electric conductivity of Cu(NO3)2 ṡ 3H2O solutions in dimethyl sulfoxide in the 0.01-2.82 M range of concentrations and at temperatures of 288-318 K. The limiting molar conductivity of the electrolyte and the mobility of Cu2+ and NO 3 - ions, the effective coefficients of diffusion of copper(II) ions and nitrate ions, and the degree and constant of electrolytic dissociation are calculated for different temperatures from the experimental results. It is established that solutions containing 0.1-0.6 M copper nitrate trihydrate in DMSO having low viscosity and high electrical conductivity can be used in electrochemical deposition.

  2. Seasonal dynamics of nitrate and ammonium ion concentrations in soil solutions collected using MacroRhizon suction cups.

    Science.gov (United States)

    Kabala, Cezary; Karczewska, Anna; Gałka, Bernard; Cuske, Mateusz; Sowiński, Józef

    2017-07-01

    The aims of the study were to analyse the concentration of nitrate and ammonium ions in soil solutions obtained using MacroRhizon miniaturized composite suction cups under field conditions and to determine potential nitrogen leaching from soil fertilized with three types of fertilizers (standard urea, slow-release urea, and ammonium nitrate) at the doses of 90 and 180 kg ha -1 , applied once or divided into two rates. During a 3-year growing experiment with sugar sorghum, the concentration of nitrate and ammonium ions in soil solutions was the highest with standard urea fertilization and the lowest in variants fertilized with slow-release urea for most of the months of the growing season. Higher concentrations of both nitrogen forms were noted at the fertilizer dose of 180 kg ha -1 . One-time fertilization, at both doses, resulted in higher nitrate concentrations in June and July, while dividing the dose into two rates resulted in higher nitrate concentrations between August and November. The highest potential for nitrate leaching during the growing season was in July. The tests confirmed that the miniaturized suction cups MacroRhizon are highly useful for routine monitoring the concentration of nitrate and ammonium ions in soil solutions under field conditions.

  3. 21 CFR 172.167 - Silver nitrate and hydrogen peroxide solution.

    Science.gov (United States)

    2010-04-01

    ... SERVICES (CONTINUED) FOOD FOR HUMAN CONSUMPTION (CONTINUED) FOOD ADDITIVES PERMITTED FOR DIRECT ADDITION TO FOOD FOR HUMAN CONSUMPTION Food Preservatives § 172.167 Silver nitrate and hydrogen peroxide solution... intended to be filtered through a silver-containing water filter. (g) Bottled water must meet the quality...

  4. Contribution to the study of uranyl salts in butyl phosphate solutions; Contribution a l'etude des solutions de sels d'uranyle dans les phosphates butyliques

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-06-01

    A spectroscopic study in the normal infrared region and involving the following associations: tri-alkyl phosphates (tri-butyl, tri-ethyl, tri-methyl), uranyl salts (nitrate, chloride, acetate) has confirmed the existence of an interaction between the phosphoryl group and the uranium atom, as shown by a movement of absorption band for the valency P = 0 from {approx} 1270 cm{sup -1} to {approx} 1180 cm{sup -1}. A study of the preparation, analysis and spectroscopy of the solids obtained by the precipitation of uranyl salts by acid butyl phosphates has been carried out. By infrared spectrophotometry it has been shown that the tri-butyl and di-butyl phosphates are associated in non-polar diluents even before the uranium is introduced. The extraction of uranyl salts from acid aqueous solutions by a diluted mixture of tri-butyl and di-butyl phosphates proceeds by different mechanisms according to the nature of the ion (nitrate or chloride). (author) [French] Une etude spectroscopique dans l'infrarouge moyen portant sur les associations: - phosphates trialcoyliques (tributylique - triethylique - trimethylique) - sels d'uranyle (nitrate, chlorure, acetate) a confirme l'existence d'une interaction entre le groupement phosphoryle et l'atome d'uranium, se manifestant par un deplacement de la bande d'absorption de la vibration de valence P = 0 de {approx} 1270 cm{sup -1} a {approx} 1180 cm{sup -1}. Une etude preparative, analytique et spectroscopique des solides obtenus par precipitation de sels d'uranyle par les phosphates butyliques acides a ete effectuee. La spectrophotomerie infrarouge met en evidence l'association, anterieure a toute introduction d'uranium, des phosphates tributylique et dibutylique dans des diluants non polaires. L'extraction de sels d'uranyle, d'une solution aqueuse acide par un melange dilue de phosphates tributylique et dibutylique, s'effectue suivant des processus differents a la nature de l'anion (nitrate ou chlorure). (auteur)

  5. Study of uranium (VI) in carbonate solution by potentiometric titrations and ion-exchange

    International Nuclear Information System (INIS)

    Billon, A.

    1968-04-01

    The present work is devoted to the fixation of uranium (VI) on the conventional anion-exchange resin Dowex 2 X 8 in carbonate and hydrogen-carbonate media. Both media were successfully used for the recuperation of uranium (VI) from very dilute solutions. Equilibrium constant of the exchange [UO 2 (CO 3 ) 3 4+ ] S + 2 [CO 3 2- ] R ↔ [UO 2 (CO 3 ) 3 4- ] R + 2[CO 3 2- ] S is determined for carbonate concentration range 0.1 M to 0.6 M from partition curves. A markedly increase in the relative fixation of uranium results with: - increasing free carbonate concentration of the solution, - decreasing uranium concentration. A study in the same conditions of the fixation of molybdenum has made it possible to separate the latter from uranium by elution, the carbonate concentration being molar. It is suggested a possibility of separation on a larger scale, based upon molybdenum displacement by uranium in hydrogen-carbonate medium. (author) [fr

  6. Study of uranium (VI) in carbonate solution by potentiometric titrations and ion-exchange; Etude des solutions d'uranium (VI) en milieu carbonate par titrages potentiometriques et echange d'ions

    Energy Technology Data Exchange (ETDEWEB)

    Billon, A [Commissariat a l' Energie Atomique, 92 - Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1968-04-01

    The present work is devoted to the fixation of uranium (VI) on the conventional anion-exchange resin Dowex 2 X 8 in carbonate and hydrogen-carbonate media. Both media were successfully used for the recuperation of uranium (VI) from very dilute solutions. Equilibrium constant of the exchange [UO{sub 2}(CO{sub 3}){sub 3}{sup 4+}]{sub S} + 2 [CO{sub 3}{sup 2-}]{sub R} {r_reversible} [UO{sub 2}(CO{sub 3}){sub 3}{sup 4-}]{sub R} + 2[CO{sub 3}{sup 2-}]{sub S} is determined for carbonate concentration range 0.1 M to 0.6 M from partition curves. A markedly increase in the relative fixation of uranium results with: - increasing free carbonate concentration of the solution, - decreasing uranium concentration. A study in the same conditions of the fixation of molybdenum has made it possible to separate the latter from uranium by elution, the carbonate concentration being molar. It is suggested a possibility of separation on a larger scale, based upon molybdenum displacement by uranium in hydrogen-carbonate medium. (author) [French] Le present travail precise la fixation de l'uranium (VI) sur la resine echangeuse d'anions Dowex 2 X 8, en milieu carbonate et hydrogeno-carbonate. Nous en avons deduit que ces deux milieux sont egalement favorables a la recuperation de l'uranium a partir de solutions tres diluees. La constante d'equilibre de la reaction d'echange [UO{sub 2}(CO{sub 3}){sub 3}{sup 4+}]{sub S} + 2 [CO{sub 3}{sup 2-}]{sub R} {r_reversible} [UO{sub 2}(CO{sub 3}){sub 3}{sup 4-}]{sub R} + 2[CO{sub 3}{sup 2-}]{sub S} a ete determinee pour le milieu carbonate 0.1 M a 0.6 M, a partir deb courbes de partage. La fixation relative de l'uranium augmente considerablement lorsque: - la concentration du carbonate libre (respectivement hydrogenocarbonate) diminue, - la concentration de l'uranium en solution diminue. Le comportement du molybdene a ete etudie en vue de la separation uranium-molybdene. L'ion fixe sur la resine est l'ion molybdate MoO{sub 4}{sup 2-}. La separation est

  7. Estimation of rare earth elements in uranium matrix after solvent extraction of uranium as uranium-antipyrine-anion complex using chloroform as solvent

    International Nuclear Information System (INIS)

    Bose, Roopa; Murthy, D.S.R.; Malhotra, R.K.

    1999-01-01

    The neutron economy in a nuclear reactor dictates the occurrence of neutron absorbers at very low levels. Hence the determination of lanthanides especially Sm, Eu, Gd and Dy is one of the most difficult and complicated analytical tasks particularly in high uranium matrix. Solvent extraction is a potent and versatile technique for the separation of lanthanides. The systems generally used for lanthanide extraction are TBP-nitrate, TBP-chloride, TBP-thiocyanate TOPO, DEHPA-nitrate etc. However, these methods of extraction of lanthanides fail to give a clear cut separation for their determination from uranium matrix. Hence analytical procedures have been standardised for extraction of uranium matrix into the organic phase leaving lanthanides unextracted in the aqueous phase. In this direction Cyanex-923 a mixture of 4 trialkyl phosphine oxides, TBP- TOPO and trioctylamine in xylene have been used for extraction of uranium and consequent determination of lanthanides by ICP-AES in the aqueous phase. In this paper the authors have investigated uranium -antipyrine -anion, a different combination other than the well known phosphine oxides and tertiary amines for extraction of uranium

  8. Accumulation of uranium in plant roots absorbed from aqueous solutions

    International Nuclear Information System (INIS)

    Dohi, Terumi; Haga, Nobuhiko; Nakashima, Satoru; Tagai, Tokuhei

    2007-01-01

    In order to study accumulation mechanisms of uranium (U) in terrestrial plants, uptake experiments for U have been carried out by using Indian mustard (Brassica juncea). This plant is edible and known as a heavy metal accumulator, especially for cadmium (Cd). About 30 rootsstocks of Indian mustard grown hydroponically in laboratory dishes were kept in uranyl (UO 2 2+ ) nitrate solutions (initially 0.5 mmol/l) at 25degC for 24, 48 and 72 hours (h). The average U concentrations in leaves increased until 48 h up to about 0.6 mg/g and then decreased slightly. Those in roots showed similar trends, but with much higher maximum U concentrations of about 30 mg/g. Backscattered electron images under SEM of the roots showed that U was accumulated on the cell edges. EPMA elemental mapping indicated that phosphorus (P) distribution had a very strong correlation with that of U. The distribution of sulfur (S) appeared to be somewhat different form these U and P distributions. These results suggest that U can be absorbed into plant roots as uranyl (UO 2 2+ ) and might be fixed at the phospholipid rich cell membranes. This U accumulation mechanism appeared to be different from that for Cd which has a close association with S. (author)

  9. Recovery of uranium and lining material from magnesium fluoride slag at UMP

    International Nuclear Information System (INIS)

    Bandyopadhyay, P.K.; Singh, H.; Shadakshari, B.M.; Meghal, A.M.

    1991-01-01

    At Uranium Metal Plant, uranium metal is produced by reduction of UF 4 with magnesium metal, in a closed reactor lined with refractory MgF 2 lining material. During this reduction, more MgF 2 is produced as the slag. This slag generally contains 2-4% uranium and hence is processed to recover these values and part of the slag, free from uranium is reused for lining the reactor. This paper describes the process parameters finalised for crushing and grinding of the slag and for leaching uranium with nitric acid. The leach liquor contains appreciable amount of fluoride and hence is processed through a separate solvent extraction cycle with tributyl phosphate. The resultant purified uranyl nitrate solution is mixed with the main stream crude solution for final purification. The conditions optimised for the solvent extraction step, the problems faced during the regular operation over the last few years and the experience gained are described. An outline of the scheme to treat larger quantities of the slag on a regular basis is presented. (author). 6 refs., 1 fig., 1 tab

  10. Standard test methods for chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade uranyl nitrate solutions

    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, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade uranyl nitrate solution to determine compliance with specifications. 1.2 The analytical procedures appear in the following order: Sections Determination of Uranium 7 Specific Gravity by Pycnometry 15-20 Free Acid by Oxalate Complexation 21-27 Determination of Thorium 28 Determination of Chromium 29 Determination of Molybdenum 30 Halogens Separation by Steam Distillation 31-35 Fluoride by Specific Ion Electrode 36-42 Halogen Distillate Analysis: Chloride, Bromide, and Iodide by Amperometric Microtitrimetry 43 Determination of Chloride and Bromide 44 Determination of Sulfur by X-Ray Fluorescence 45 Sulfate Sulfur by (Photometric) Turbidimetry 46 Phosphorus by the Molybdenum Blue (Photometric) Method 54-61 Silicon by the Molybdenum Blue (Photometric) Method 62-69 Carbon by Persulfate Oxidation-Acid Titrimetry 70 Conversion to U3O8 71-74 Boron by ...

  11. Vapor-liquid equilibria for nitric acid-water and plutonium nitrate-nitric acid-water solutions

    International Nuclear Information System (INIS)

    Maimoni, A.

    1980-01-01

    The liquid-vapor equilibrium data for nitric acid and nitric acid-plutnonium nitrate-water solutions were examined to develop correlations covering the range of conditions encountered in nuclear fuel reprocessing. The scanty available data for plutonium nitrate solutions are of poor quality but allow an order of magnitude estimate to be made. A formal thermodynamic analysis was attempted initially but was not successful due to the poor quality of the data as well as the complex chemical equilibria involved in the nitric acid and in the plutonium nitrate solutions. Thus, while there was no difficulty in correlating activity coefficients for nitric acid solutions over relatively narrow temperature ranges, attempts to extend the correlations over the range 25 0 C to the boiling point were not successful. The available data were then analyzed using empirical correlations from which normal boiling points and relative volatilities can be obtained over the concentration ranges 0 to 700 g/l Pu, 0 to 13 M nitric acid. Activity coefficients are required, however, if estimates of individual component vapor pressures are needed. The required ternary activity coefficients can be approximated from the correlations

  12. Comparision of Chitosan Function as Adsorbent for Nitrate Removal Using Synthetic Aqueous Solution and Drinking Water

    Directory of Open Access Journals (Sweden)

    Mohammad Norisepehr

    2013-12-01

    Full Text Available Background & Objectives: Nitrate and nitrite compounds pollution of groundwater resources in recent years which recently their mean concentration due to enhancement of different kind of municipal, industrial and agriculture waste water, were increased. The most common source of nitrates entering the water include chemical fertilizers and animal manure in agriculture, septic tank effluent, wastewater, wastewater treatment plants, animal and plant residue analysis on the ground of non-sanitary disposal of solid waste and the use of absorbing wells for sewage disposal. Materials and methods: This experimental study is applied to the nitrate removal using chitosan in laboratory scale at ambient temperature and the design of the system was Batch. Effects of parameters such as pH, contact time, initial concentration and adsorbent concentration of nitrate on nitrate removal from aqueous solution was studied. Results: Function of chitosan in synthetic aqueous solution and drinking water according to the slurry system results, the optimum condition was obtained at pH=4, 20 min contact time and increasing the initial concentration of nitrate enhance the adsorption capacity of chitosan. Also optimum dosage of adsorbent was obtained at 0.5 g/l. The data obtained from the experiments of adsorbent isotherm were analyzed using Langmuir and Freundlich isotherm models. The Langmuir equation was found to be the best fitness with the experimental data (R2>0.93. Conclusion: Although efficiency of Nitrate removal in synthetic aqueous solution was better than drinking water, adsorption process using chitosan as an option for the design and selection nitrate removal should be considered in order to achieve environmental standards.

  13. Spectrophotometric determination of uranium and plutonium in nitric acid solutions at their co-presence

    International Nuclear Information System (INIS)

    Levakov, B.I.; Mishenev, V.B.; Nezgovorov, N.Yu.; Ryazanova, G.K.; Timofeev, G.A.

    1986-01-01

    The method of spectrophotometric determination of uranium (6) and plutonium (4) in nitric acid solutions is described. Uranium is determined by light absorption of the complex with arsenazo 3 in 0.05 mol/l nitric acid at λ=654 nm, plutonium - by light absorption of the complex with xylenol orange in 0.1 mol/l nitric acid at λ=540 nm. To disguise plutonium, tetravalent and certain trivalent elements DTPA is introduced into photometered solution for uranium determination. The relative root-mean square deviation of determination results does not exceed 0.03 in uranium concenration ranges 0.5-5 μg/ml, of plutonium -1-3 μg/ml

  14. Bioprecipitation of uranium from alkaline waste solutions using recombinant Deinococcus radiodurans

    Energy Technology Data Exchange (ETDEWEB)

    Kulkarni, Sayali; Ballal, Anand; Apte, Shree Kumar, E-mail: aptesk@barc.gov.in

    2013-11-15

    Highlights: • Deinococcus radiodurans was genetically engineered to overexpress alkaline phosphatase (PhoK). • Deino-PhoK bioprecipitated U efficiently over a wide range of input U concentration. • A maximal loading of 10.7 g U/g of biomass at 10 mM input U was observed. • Radioresistance and U precipitation by Deino-PhoK remained unaffected by γ radiation. • Immobilization of Deino-PhoK facilitated easy separation of precipitated U. -- Abstract: Bioremediation of uranium (U) from alkaline waste solutions remains inadequately explored. We engineered the phoK gene (encoding a novel alkaline phosphatase, PhoK) from Sphingomonas sp. for overexpression in the radioresistant bacterium Deinococcus radiodurans. The recombinant strain thus obtained (Deino-PhoK) exhibited remarkably high alkaline phosphatase activity as evidenced by zymographic and enzyme activity assays. Deino-PhoK cells could efficiently precipitate uranium over a wide range of input U concentrations. At low uranyl concentrations (1 mM), the strain precipitated >90% of uranium within 2 h while a high loading capacity of around 10.7 g U/g of dry weight of cells was achieved at 10 mM U concentration. Uranium bioprecipitation by Deino-PhoK cells was not affected in the presence of Cs and Sr, commonly present in intermediate and low level liquid radioactive waste, or after exposure to very high doses of ionizing radiation. Transmission electron micrographs revealed the extracellular nature of bioprecipitated U, while X-ray diffraction and fluorescence analysis identified the precipitated uranyl phosphate species as chernikovite. When immobilized into calcium alginate beads, Deino-PhoK cells efficiently removed uranium, which remained trapped in beads, thus accomplishing physical separation of precipitated uranyl phosphate from solutions. The data demonstrate superior ability of Deino-PhoK, over earlier reported strains, in removal of uranium from alkaline solutions and its potential use in

  15. Removal of nitrate from aqueous solution using cetylpyridinium bromide (CPB) modified zeolite as adsorbent

    International Nuclear Information System (INIS)

    Zhan Yanhui; Lin Jianwei; Zhu Zhiliang

    2011-01-01

    Surfactant modified zeolites (SMZ) with different coverage types were prepared by loading the cetylpyridinium bromide (CPB) onto the surface of the natural zeolites. The adsorption behavior of nitrate on SMZ was investigated. Natural zeolite and SMZ with monolayer CPB coverage were inefficient for the removal of nitrate from aqueous solution. However, SMZ with patchy bilayer or bilayer CPB coverage was efficient in nitrate removal, and the nitrate adsorption capacity of SMZ increased with its CPB loading. For typical SMZ with bilayer CPB coverage, the nitrate adsorption process was well described by the pseudo-second-order kinetic model, and the experimental isotherm data fitted well with the Langmuir, Freundlich and Dubinin-Redushkevich isotherm models. Thermodynamic parameters such as Gibbs free energy change, enthalpy change and entropy change were calculated and the results showed that the adsorption of nitrate on SMZ was spontaneous and exothermic in nature. The presence of competing anions such as chloride, sulfate and bicarbonate ions slightly reduced the nitrate adsorption efficiency. Anionic exchange and electrostatic interaction were proven to be the main mechanisms that govern the adsorption of nitrate on SMZ.

  16. Effects of over-winter green cover on soil solution nitrate concentrations beneath tillage land.

    Science.gov (United States)

    Premrov, Alina; Coxon, Catherine E; Hackett, Richard; Kirwan, Laura; Richards, Karl G

    2014-02-01

    There is a growing need to reduce nitrogen losses from agricultural systems to increase food production while reducing negative environmental impacts. The efficacy of vegetation cover for reducing nitrate leaching in tillage systems during fallow periods has been widely investigated. Nitrate leaching reductions by natural regeneration (i.e. growth of weeds and crop volunteers) have been investigated to a lesser extent than reductions by planted cover crops. This study compares the efficacy of natural regeneration and a sown cover crop (mustard) relative to no vegetative cover under both a reduced tillage system and conventional plough-based system as potential mitigation measures for reducing over-winter soil solution nitrate concentrations. The study was conducted over three winter fallow seasons on well drained soil, highly susceptible to leaching, under temperate maritime climatic conditions. Mustard cover crop under both reduced tillage and conventional ploughing was observed to be an effective measure for significantly reducing nitrate concentrations. Natural regeneration under reduced tillage was found to significantly reduce the soil solution nitrate concentrations. This was not the case for the natural regeneration under conventional ploughing. The improved efficacy of natural regeneration under reduced tillage could be a consequence of potential stimulation of seedling germination by the autumn reduced tillage practices and improved over-winter plant growth. There was no significant effect of tillage practices on nitrate concentrations. This study shows that over winter covers of mustard and natural regeneration, under reduced tillage, are effective measures for reducing nitrate concentrations in free draining temperate soils. © 2013.

  17. Chemical thermodynamics of uranium

    International Nuclear Information System (INIS)

    Grenthe, I.; Fuger, J.; Lemire, R.J.; Muller, A.B.; Nguyen-Trung Cregu, C.; Wanner, H.

    1992-01-01

    A comprehensive overview on the chemical thermodynamics of those elements that are of particular importance in the safety assessment of radioactive waste disposal systems is provided. This is the first volume in a series of critical reviews to be published on this subject. The book provides an extensive compilation of chemical thermodynamic data for uranium. A description of procedures for activity corrections and uncertainty estimates is given. A critical discussion of data needed for nuclear waste management assessments, including areas where significant gaps of knowledge exist is presented. A detailed inventory of chemical thermodynamic data for inorganic compounds and complexes of uranium is listed. Data and their uncertainty limits are recommended for 74 aqueous complexes and 199 solid and 31 gaseous compounds containing uranium, and on 52 aqueous and 17 solid auxiliary species containing no uranium. The data are internally consistent and compatible with the CODATA Key Values. The book contains a detailed discussion of procedures used for activity factor corrections in aqueous solution, as well as including methods for making uncertainty estimates. The recommended data have been prepared for use in environmental geochemistry. Containing contributions written by experts the chapters cover various subject areas such a s: oxide and hydroxide compounds and complexes, the uranium nitrides, the solid uranium nitrates and the arsenic-containing uranium compounds, uranates, procedures for consistent estimation of entropies, gaseous and solid uranium halides, gaseous uranium oxides, solid phosphorous-containing uranium compounds, alkali metal uranates, uncertainties, standards and conventions, aqueous complexes, uranium minerals dealing with solubility products and ionic strength corrections. The book is intended for nuclear research establishments and consulting firms dealing with uranium mining and nuclear waste disposal, as well as academic and research institutes

  18. Computer assisted optimization of sensitivity of the catalytic polarography wave of uranly/nitrate system

    International Nuclear Information System (INIS)

    Betteridge, D.; Wade, A.P.; Neves, E.A.; Gutz, I.

    1982-01-01

    The utilization of 'Simplex' optimization program, writes in Basic language, used to the study of catalytic polarographic wave procedure the reaction of uranium (III) with nitrate in acidic medium at the mercury electrode/solution interface is described. This 'Simplex' optimization requires a much smaller number of experiments to reach an improved condition without increase of the number of trial and error experiments. (C.G.C.) [pt

  19. Purification of alkali metal nitrates

    Science.gov (United States)

    Fiorucci, Louis C.; Gregory, Kevin M.

    1985-05-14

    A process is disclosed for removing heavy metal contaminants from impure alkali metal nitrates containing them. The process comprises mixing the impure nitrates with sufficient water to form a concentrated aqueous solution of the impure nitrates, adjusting the pH of the resulting solution to within the range of between about 2 and about 7, adding sufficient reducing agent to react with heavy metal contaminants within said solution, adjusting the pH of the solution containing reducing agent to effect precipitation of heavy metal impurities and separating the solid impurities from the resulting purified aqueous solution of alkali metal nitrates. The resulting purified solution of alkali metal nitrates may be heated to evaporate water therefrom to produce purified molten alkali metal nitrate suitable for use as a heat transfer medium. If desired, the purified molten form may be granulated and cooled to form discrete solid particles of alkali metal nitrates.

  20. Preparation standardisation and use of plutonium nitrate reference solutions

    International Nuclear Information System (INIS)

    Brown, M.L.; Drummond, J.L.

    1981-07-01

    A procedure is described for the purification of a plutonium nitrate solution in nitric acid for use as a plutonium master standard. Anion exchange chromatography followed by oxalate precipitation is used to purify the plutonium and the residual cationic impurities are analysed by emission spectroscopy. The plutonium content is accurately and precisely measured by two independent methods, namely by gravimetry as PuO 2 at 1250 0 C and by ceric oxidation, ferrous reduction and dichromate titration. Full details of the purification procedure are given, with recommended methods for storing and using the standard solution. It is concluded that such a solution is the most satisfactory reference material, available for plutonium analysis for reprocessing plants, and is adequately related to other, internationally accepted, standard reference materials. (author)

  1. Uranium extraction from high content chlorine leach liquor

    International Nuclear Information System (INIS)

    Fatemi, K.

    1998-01-01

    In this work uranium solution has been leached out by leaching process of uranium ores from Bandar-Ab bass port using sea water, since fresh water could not be available when it is processed in large scale. Two samples of different batches containing 11 and 20 gr./lit chlorine underwent two stages of precipitation by lead nitrate. As the result of this treatment the chlorine removed and its final concentration reduced to 530 p.p.m. which is well below allowances. Then, the uranium of this recent dechlorinated solu ton has been extracted by T.B.P. Uranium in organic phase was stripped out into inorganic phase by sodium carbonate and precipitated in a form of yellow cake and converted to U3o8. The total recovery of U, was well above 90% and the purity of the conc. U was better than 94%. The lead used at the beginning of the process was recovered for next use

  2. 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

  3. Uranium Immobilization through Fe(II) bio-oxidation: A Column study

    Energy Technology Data Exchange (ETDEWEB)

    Coates, John D.

    2009-09-14

    Current research on the bioremediation of heavy metals and radionuclides is focused on the ability of reducing organisms to use these metals as alternative electron acceptors in the absence of oxygen and thus precipitate them out of solution. However, many aspects of this proposed scheme need to be resolved, not the least of which is the time frame of the treatment process. Once treatment is complete and the electron donor addition is halted, the system will ultimately revert back to an oxic state and potentially result in the abiotic reoxidation and remobilization of the immobilized metals. In addition, the possibility exists that the presence of more electropositive electron acceptors such as nitrate or oxygen will also stimulate the biological oxidation and remobilization of these contaminants. The selective nitrate-dependent biooxidation of added Fe(II) may offer an effective means of “capping off” and completing the attenuation of these contaminants in a reducing environment making the contaminants less accessible to abiotic and biotic reactions and allowing the system to naturally revert to an oxic state. Our previous DOE-NABIR funded studies demonstrated that radionuclides such as uranium and cobalt are rapidly removed from solution during the biogenic formation of Fe(III)-oxides. In the case of uranium, X-ray spectroscopy analysis indicated that the uranium was in the hexavalent form (normally soluble) and was bound to the precipitated Fe(III)-oxides thus demonstrating the bioremediative potential of this process. We also demonstrated that nitrate-dependent Fe(II)- oxidizing bacteria are prevalent in the sediment and groundwater samples collected from sites 1 and 2 and the background site of the NABIR FRC in Oakridge, TN. However, all of these studies were performed in batch experiments in the laboratory with pure cultures and although a significant amount was learned about the microbiology of nitrate-dependent bio-oxidation of Fe(II), the effects of

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

    International Nuclear Information System (INIS)

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    2004-01-01

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

  6. Root uptake of uranium by a higher plant model (Phaseolus vulgaris) bioavailability from soil solution

    International Nuclear Information System (INIS)

    Laroche, L.; Henner, P.; Camilleri, V.; Garnier-Laplace, J.

    2004-01-01

    Uranium behaviour in soils is controlled by actions and interactions between physicochemical and biological processes that also determine its bioavailability. In soil solution, uranium(+VI) aqueous speciation undergoes tremendous changes mainly depending on pH, carbonates, phosphates and organic matter. In a first approach to identify bioavailable species of U to plants, cultures were performed using hydroponics, to allow an easy control of the composition of the exposure media. The latter, here an artificial soil solution, was designed to control the uranium species in solution. The geochemical speciation code JCHESS using a database compiled from the OECD/NEA thermochemical database project and verified was used to perform the solution speciation calculations. On this theoretical basis, three domains were defined for short-duration well-defined laboratory experiments in simplified conditions: pH 4.9, 5.8 and 7 where predicted dominant species are uranyl ions, hydroxyl complexes and carbonates respectively. For these domains, biokinetics and characterization of transmembrane transport according to a classical Michaelis Menten approach were investigated. The Free Ion Model (or its derived Biotic Ligand Model) was tested to determine if U uptake is governed by the free uranyl species or if other metal complexes can be assimilated. The effect of different variables on root assimilation efficiency and phyto-toxicity was explored: presence of ligands such as phosphates or carbonates and competitive ions such as Ca 2+ at the 3 pH. According to previous experiments, uranium was principally located in roots whatever the pH and no difference in uranium uptake was evidenced between the main growth stages of the plant. Within the 3 studied chemical domains, results from short-term kinetics evidenced a linear correlation between total uranium concentration in bean roots and that in exposure media, suggesting that total uranium in soil solution could be a good predictor for

  7. Root uptake of uranium by a higher plant model (Phaseolus vulgaris) bioavailability from soil solution

    Energy Technology Data Exchange (ETDEWEB)

    Laroche, L.; Henner, P.; Camilleri, V.; Garnier-Laplace, J. [CEA Cadarache (DEI/SECRE/LRE), Laboratory of Radioecology and Ecotoxicology, Institute for Radioprotection and Nuclear Safety, 13 - Saint-Paul-lez-Durance (France)

    2004-07-01

    Uranium behaviour in soils is controlled by actions and interactions between physicochemical and biological processes that also determine its bioavailability. In soil solution, uranium(+VI) aqueous speciation undergoes tremendous changes mainly depending on pH, carbonates, phosphates and organic matter. In a first approach to identify bioavailable species of U to plants, cultures were performed using hydroponics, to allow an easy control of the composition of the exposure media. The latter, here an artificial soil solution, was designed to control the uranium species in solution. The geochemical speciation code JCHESS using a database compiled from the OECD/NEA thermochemical database project and verified was used to perform the solution speciation calculations. On this theoretical basis, three domains were defined for short-duration well-defined laboratory experiments in simplified conditions: pH 4.9, 5.8 and 7 where predicted dominant species are uranyl ions, hydroxyl complexes and carbonates respectively. For these domains, biokinetics and characterization of transmembrane transport according to a classical Michaelis Menten approach were investigated. The Free Ion Model (or its derived Biotic Ligand Model) was tested to determine if U uptake is governed by the free uranyl species or if other metal complexes can be assimilated. The effect of different variables on root assimilation efficiency and phyto-toxicity was explored: presence of ligands such as phosphates or carbonates and competitive ions such as Ca{sup 2+} at the 3 pH. According to previous experiments, uranium was principally located in roots whatever the pH and no difference in uranium uptake was evidenced between the main growth stages of the plant. Within the 3 studied chemical domains, results from short-term kinetics evidenced a linear correlation between total uranium concentration in bean roots and that in exposure media, suggesting that total uranium in soil solution could be a good predictor

  8. A Solution-Based Approach for Mo-99 Production: Considerations for Nitrate versus Sulfate Media

    Directory of Open Access Journals (Sweden)

    Amanda J. Youker

    2013-01-01

    Full Text Available Molybdenum-99 is the parent of Technetium-99m, which is used in nearly 80% of all nuclear medicine procedures. The medical community has been plagued by Mo-99 shortages due to aging reactors, such as the NRU (National Research Universal reactor in Canada. There are currently no US producers of Mo-99, and NRU is scheduled for shutdown in 2016, which means that another Mo-99 shortage is imminent unless a potential domestic Mo-99 producer fills the void. Argonne National Laboratory is assisting two potential domestic suppliers of Mo-99 by examining the effects of a uranyl nitrate versus a uranyl sulfate target solution configuration on Mo-99 production. Uranyl nitrate solutions are easier to prepare and do not generate detectable amounts of peroxide upon irradiation, but a high radiation field can lead to a large increase in pH, which can lead to the precipitation of fission products and uranyl hydroxides. Uranyl sulfate solutions are more difficult to prepare, and enough peroxide is generated during irradiation to cause precipitation of uranyl peroxide, but this can be prevented by adding a catalyst to the solution. A titania sorbent can be used to recover Mo-99 from a highly concentrated uranyl nitrate or uranyl sulfate solution; however, different approaches must be taken to prevent precipitation during Mo-99 production.

  9. Transpassive electrodissolution of depleted uranium in alkaline electrolytes

    International Nuclear Information System (INIS)

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

    1998-03-01

    To aid in removal of oralloy from the nuclear weapons stockpile, scientists at the Los Alamos National Laboratory Plutonium Facility are decontaminating oralloy parts by electrodissolution in neutral to alkaline electrolytes composed of sodium nitrate and sodium sulfate. To improve the process, electrodissolution experiments were performed with depleted uranium to understand the effects of various operating parameters. Sufficient precipitate was also produced to evaluate the feasibility of using ultrafiltration to separate the uranium oxide precipitates from the electrolyte before it enters the decontamination fixture. In preparation for the experiments, a potential-pH diagram for uranium was constructed from thermodynamic data for fully hydrated species. Electrodissolution in unstirred solutions showed that uranium dissolution forms two layers, an acidic bottom layer rich in uranium and an alkaline upper layer. Under stirred conditions results are consistent with the formation of a yellow precipitate of composition UO 3 ·2H 2 O, a six electron process. Amperometric experiments showed that current efficiency remained near 100% over a wide range of electrolytes, electrolyte concentrations, pH, and stirring conditions

  10. Process for plutonium rextraction in aqueous solution from an organic solvent, especially for uranium plutonium partition

    International Nuclear Information System (INIS)

    Germain, M.; Gillet, B.; Pasquiou, J.Y.

    1989-01-01

    The organic solvent containing plutonium is contacted with an aqueous solution of a uranous salt, for instance uranous nitrate, and a hydroxylamine salt, for instance the nitrate. In these conditions uranous nitrate is a reducing agent of Pu III and hydroxylamine nitrate stabilizes Pu III and U IV in the aqueous phase. Performances are similar to these of the U IV-hydrazine nitrate without interference of hydrazine nitrate degradation products [fr

  11. Nitrates of rare earths

    International Nuclear Information System (INIS)

    Komissarova, L.N.; Pushkina, L.Ya.

    1984-01-01

    The systematization of experimental data with account of the last achievements in the field of studying the RE nitrate properties is realized. The methods of production, solubility in aqueous solutions structure, thermodynamic characteristics and thermal stability of nitrate hydrates, RE anhydrous and basic nitrates are considered. The data on RE nirtrate complexing in aqueous solutions are given. Binary nitrates, nitrate solvates and RE nitrate adducts with organic compounds are described. The use of RE nitrates in the course of RE production, in the processes of separation and fine cleaning of RE preparations is considered

  12. Study of uranium (VI) in carbonate solution by potentiometric titrations and ion-exchange; Etude des solutions d'uranium (VI) en milieu carbonate par titrages potentiometriques et echange d'ions

    Energy Technology Data Exchange (ETDEWEB)

    Billon, A. [Commissariat a l' Energie Atomique, 92 - Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1968-04-01

    The present work is devoted to the fixation of uranium (VI) on the conventional anion-exchange resin Dowex 2 X 8 in carbonate and hydrogen-carbonate media. Both media were successfully used for the recuperation of uranium (VI) from very dilute solutions. Equilibrium constant of the exchange [UO{sub 2}(CO{sub 3}){sub 3}{sup 4+}]{sub S} + 2 [CO{sub 3}{sup 2-}]{sub R} {r_reversible} [UO{sub 2}(CO{sub 3}){sub 3}{sup 4-}]{sub R} + 2[CO{sub 3}{sup 2-}]{sub S} is determined for carbonate concentration range 0.1 M to 0.6 M from partition curves. A markedly increase in the relative fixation of uranium results with: - increasing free carbonate concentration of the solution, - decreasing uranium concentration. A study in the same conditions of the fixation of molybdenum has made it possible to separate the latter from uranium by elution, the carbonate concentration being molar. It is suggested a possibility of separation on a larger scale, based upon molybdenum displacement by uranium in hydrogen-carbonate medium. (author) [French] Le present travail precise la fixation de l'uranium (VI) sur la resine echangeuse d'anions Dowex 2 X 8, en milieu carbonate et hydrogeno-carbonate. Nous en avons deduit que ces deux milieux sont egalement favorables a la recuperation de l'uranium a partir de solutions tres diluees. La constante d'equilibre de la reaction d'echange [UO{sub 2}(CO{sub 3}){sub 3}{sup 4+}]{sub S} + 2 [CO{sub 3}{sup 2-}]{sub R} {r_reversible} [UO{sub 2}(CO{sub 3}){sub 3}{sup 4-}]{sub R} + 2[CO{sub 3}{sup 2-}]{sub S} a ete determinee pour le milieu carbonate 0.1 M a 0.6 M, a partir deb courbes de partage. La fixation relative de l'uranium augmente considerablement lorsque: - la concentration du carbonate libre (respectivement hydrogenocarbonate) diminue, - la concentration de l'uranium en solution diminue. Le comportement du molybdene a ete etudie en vue de la separation uranium-molybdene. L'ion fixe sur la resine est l

  13. Crystallization of sodium nitrate from radioactive waste

    International Nuclear Information System (INIS)

    Krapukhin, V.B.; Krasavina, E.P.; Pikaev, A.K.

    1997-07-01

    From the 1940s to the 1980s, the Institute of Physical Chemistry of the Russian Academy of Sciences (IPC/RAS) conducted research and development on processes to separate acetate and nitrate salts and acetic acid from radioactive wastes by crystallization. The research objective was to decrease waste volumes and produce the separated decontaminated materials for recycle. This report presents an account of the IPC/RAS experience in this field. Details on operating conditions, waste and product compositions, decontamination factors, and process equipment are described. The research and development was generally related to the management of intermediate-level radioactive wastes. The waste solutions resulted from recovery and processing of uranium, plutonium, and other products from irradiated nuclear fuel, neutralization of nuclear process solutions after extractant recovery, regeneration of process nitric acid, equipment decontamination, and other radiochemical processes. Waste components include nitric acid, metal nitrate and acetate salts, organic impurities, and surfactants. Waste management operations generally consist of two stages: volume reduction and processing of the concentrates for storage, solidification, and disposal. Filtration, coprecipitation, coagulation, evaporation, and sorption were used to reduce waste volume. 28 figs., 40 tabs

  14. Preparation of the pur uranium-metal; La preparation de l'uranium-metal pur

    Energy Technology Data Exchange (ETDEWEB)

    Goldschmidt, B; Vertes, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    A detailed description of the chemical processes used to prepare in the factory of Bouchet of the CEA (Seine-Et-Oise) pur metal uranium with either relatively rich ores, or extracts coming of physical or chemical treatment of poor ores. The nitric treatment of ores succeeds to the production of uranate of impure sodium carbonate. This last last product is dissolved in nitric acid and the uranyl nitrate is extracted by tributyl-phosphate diluted in an inert solvent. The uranyl nitrate pure is re-extracted and successively transformed in uranium peroxide, in orange oxide then in brown oxide which is transformed in fluoride by the anhydrous hydrofluoric acid. Uranate fluoride is then reduced in metal by the pure calcium with an yield superior to 99%. (authors) [French] Description detaillee des procedes chimiques mis en jeu pour preparer a l'Usine du Bouchet du Commissariat a l'Energie Atomique (Seine-et-Oise) l'uranium metal pur a partir soit de minerais relativement riches, soit de concentres provenant de traitement physique ou chimique de minerais pauvres. Le traitement nitrique des minerais aboutit a la production d'uranate de soude impur. Ce dernier est a son tour dissous dans l'acide nitrique et le nitrate d'uranyle est extrait par du tributyl-phosphate dilue par un solvant inerte. Le nitrate d'uranyle pur reextrait est transforme successivement en peroxyde d'uranium, en oxyde orange puis en oxyde brun qui est transforme en fluorure par l'acide fluorhydrique anhydre. Le fluorure uraneux est reduit en metal par le calcium pur avec un rendement superieur a 99 %. (auteurs)

  15. The extraction of lanthanides and americium by benzyldiakylamines and benzyltrialkylammonium nitrates from the nitrate solutions; structure and aggregation of their salts

    International Nuclear Information System (INIS)

    Jedinakova, V.; Zilkova, J.; Dvorak, Z.; Vojtiskova, M.

    1982-01-01

    Benzyldialkylamine and benzyltrialkylammonium nitrates were used for the extraction of lanthanides and americium from aqueous nitrate solutions. The dependence of the extraction performance for Ln(III) and Am(III) on the concentration of nitric acid, the kind and concentration of salting-out agents in the aqueous phase, and the kind of solvent were investigated. The extraction of Am(III) is compared with the extraction of lanthanides. The difference in distribution coefficients for lanthanides and americium can be utilized for the separation of lanthanides and americium. Using vapor phase osmometry and cryoscopy the association of these compounds was measured at 5.5deg, 25deg and 37deg C, allowing rough estimates of ΔH and ΔS for the formation of the aggregates, monomers in the case of benzyldiethylamine, benzyldibutylamine, benzyldihexylamine and benzyldioctylamine, tetramers for the benzyldibutylamine nitrate and tetramers for benzyldimethyldodecylammonium nitrate. (author)

  16. Evaluation of neutronic characteristics of STACY 80-cm-diameter cylindrical core fueled with 6% enriched uranyl nitrate solution

    International Nuclear Information System (INIS)

    Yanagisawa, Hiroshi; Sono, Hiroki

    2003-06-01

    For the examination of neutronic safety design of forthcoming experimental core configurations in the Static Experiment Critical Facility (STACY), neutronic characteristics of 80-cm-diameter cylindrical cores fueled with 6% enriched uranyl nitrate solution have been evaluated by computational analyses. In the analyses, the latest nuclear data library, JENDL-3.3, was used as neutron cross section data. The neutron diffusion and transport calculations were performed using a diffusion code, CITATION, in the SRAC code system and a continuous-energy Monte Carlo code, MVP. Critical level heights of the cores were obtained using such parameters as uranium concentration (up to 500 gU/l), free nitric acid concentration (up to 8 mol/l), and concentration of soluble neutron poisons, gadolinium and boron. It has been confirmed from the evaluation that all critical cores comply with safety criteria required in the STACY operation concerning excess reactivity, reactivity addition rates and shutdown margins by safety rods. (author)

  17. Process for iron separation from an organic solution containing uranium

    International Nuclear Information System (INIS)

    Textoris, A.; Lyaudet, G.; Bathelier, A.

    1987-01-01

    Iron is separated from an organic solution of U and Fe in a phosphine oxide and an acid organic phosphorus compound by reaction on oxalic acid or a mixture of sulfuric and phosphoric acid or phosphoric acid. Uranium stays in the initial organic solution and iron is transferred to the aqueous phase [fr

  18. Determination of humic acid in alkali leaching solution of uranium by spectophotrometry-COD method

    International Nuclear Information System (INIS)

    Feng Yu; An Wei; Chen Shusen

    2014-01-01

    It is one of the main causes of extraction emulsification or resin toxicosis during alkali leaching process in uranium metallurgy which organic matters including humic acid exist in lixiviums. In order to study the effect of humic acid in uranium metallurgy, a method for determination of content of humic acid in aqueous solution need to be established. Spectrophotometry is a simple and convenient method in humic acid analysis. However, accuracy of spectrophotometry can be reduced greatly because of interference of uranium and other elements in the humic acid solutions. Although chemical oxygen demand (COD) method is a common analysis way of organic matters in aqueous solutions, the concentration of humic acid cannot be directly measured. In this paper, COD method is related with spectrophotometry to avoid the interference of uranium and ensure the accurate analysis of humic acid. The results showed that the detection limit of the method was 1.78 mg/L and the recovery rate was 101.2%. (authors)

  19. TG/DTA and X ray Diffraction Studies on Ammonium Uranyl Nitrate

    International Nuclear Information System (INIS)

    Kim, Byung Ho; Lee, Young Bum; Jeong, Ji Young; Choi, Jong Hyun; Kim, Tae Joon; Nam, Ho Yun; Kim, Jong Man

    2011-01-01

    Ammonium uranyl nitrate (AUN) is an important intermediate product during conversion of a uranyl nitrate[UO 2 (NO 3 ) 2 ] solution to UO 2 powder for the fabrication of nuclear fuels, the so-called modified direct denitration (MDD) process. The MDD process involves the thermal decomposition of AUN double salts, which are prepared from a mixture consisting of a UO 2 (NO 3 ) 2 solution and NH 4 NO 3 . The physical and chemical properties of an oxide powder depend upon its thermal treatment. Three double salts are known for the UO 2 (NO 3 ) 2 - NH 4 NO 3 -H 2 O system, but there have been only a few studies done on thermal decomposition of these salts. Therefore, the objective of this study is to investigate the reaction pathways during a thermal decomposition and reduction of AUN to achieve a better knowledge of the influence of an AUN preparation process and thermal decomposition procedures on uranium oxides under a nitrogen, air, or hydrogen atmosphere

  20. Comparison of numerical and physico-chemical models for on-line spectrophotometric control of uranium

    International Nuclear Information System (INIS)

    Corriou, J.P.; Boisde, G.

    1986-04-01

    In view of on-line spectrophotometric control of fuel reprocessing streams, a physico-chemical model able to predict uranium and nitric acid concentrations in an uranyl nitrate-nitric acid system has been searched. Thus the influences of the following parameters: uranium, nitrate, hydrogen ion concentrations, ionic strength, on the equilibria of complexation of uranium (VI) nitrate have been evaluated. Extinction coefficients for the uranium mononitrate and uranium dinitrate complexes are given between 410 and 440 nm. The apparent equilibrium constants are determined as a function of the ionic strength. The limitations of this predictive model are emphasized and comparisons with numerical models are discussed. (16 refs)

  1. Resolving superimposed ground-water contaminant plumes characterized by chromium, nitrate, uranium, and technetium--99

    International Nuclear Information System (INIS)

    Hall, S.H.

    1990-02-01

    Leakage from a liquid waste storage and solar evaporation basin at the Hanford Site in southeastern Washington State has resulted in a ground-water contaminant plume characterized by nitrate, hexavalent chromium, uranium, and technetium-99. The plume is superimposed on a larger, pre-existing plume extending from upgradient sites and having the same suite of contaminants. However, the relative abundance of contaminant species is quite different for each plume source. Thus, characteristic concentration ratios, rather than concentrations of individual species, are used as geochemical tracers, with emphasis on graphical analysis. Accordingly, it has been possible to resolve the boundaries of the smaller plume and to estimate the contribution of each plume to the observed contamination downgradient from the storage basin. 11 refs., 7 figs

  2. Characterization and thermal stability of uranium peroxides by thermogravimetry and differential scanning calorimetry

    International Nuclear Information System (INIS)

    Souza Junior, P.T. de; Abrao, A.

    1982-01-01

    The characterization and thermal behaviour of uranium peroxide samples prepared by precipitation with hydrogen peroxide from uranyl nitrate solution is described. The latter was obtained by dissolution of ammonium diuranate and ammonium uranyl tricarbonate. TG and DTG curves were recorded in the temperature range from room temperature to 700 0 C. DSC curves were recorded from room temperature to 600 0 C. The heating rate was 5 0 C/min. Based upon the DTG curves of a great number of samples, the uranium peroxides were classified in five groups. The collected information was used to recognise the temperature at which the peroxide decomposes and to know its sequential conversion to UO 3 and U 3 O 8 , the evolution of molecules of water of crystallization and absorption, and the elimination of occluded nitrate ions. The results allowed to conclude that no NH + 4 nor H 2 O 2 molecules were occluded by uranium peroxide. The stoichiometric composition of representative samples for the five groups is indicated. (Author) [pt

  3. The uranium recovery from UO{sub 2} kernel production effluent

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Xiaotong, E-mail: chenxiaotong@tsinghua.edu.cn; He, Linfeng; Liu, Bing; Tang, Yaping; Tang, Chunhe

    2016-12-15

    Graphical abstract: In this study, a flow sheet including evaporation, flocculation, filtration, adsorption, and reverse osmosis was established for the UO{sub 2} kernel production effluent of HTR spherical fuel elements. The uranium recovery could reach 99.9% after the treatment, with almost no secondary pollution produced. Based on the above experimental results, the treating flow process in this study would be feasible for laboratory- and engineering-scale treatment of UO{sub 2} kernel production effluent of HTR spherical fuel elements. - Highlights: • A flow sheet including evaporation, flocculation, filtration, adsorption, and reverse osmosis was established for the UO{sub 2} kernel production effluent. • The uranium recovery could reach 99.9% after the treatment, with almost no secondary pollution produced. • The treating flow process would be feasible for laboratory- and engineering-scale treatment of UO{sub 2} kernel production effluent. - Abstract: For the fabrication of coated particle fuel elements of high temperature gas cooled reactors, the ceramic UO{sub 2} kernels are prepared through chemical gelation of uranyl nitrate solution droplets, which produces radioactive effluent with components of ammonia, uranium, organic compounds and ammonium nitrate. In this study, a flow sheet including evaporation, flocculation, filtration, adsorption, and reverse osmosis was established for the effluent treating. The uranium recovery could reach 99.9% after the treatment, with almost no secondary pollution produced.

  4. Synthesis of YBa2Cu3O7-y from nitrate solutions with urea additions

    International Nuclear Information System (INIS)

    Pershin, V.I.; Naumov, V.S.; Mozhaev, A.P.; Lyashchenko, A.K.; Pobedina, A.B.; Khajlova, E.G.

    1994-01-01

    Solubility of bariun nitrate is studied in the Y(NO 3 ) 3 -Ba(NO 3 ) 2 -Cu(NO 3 )-CO(NH 2 ) 2 -H 2 O system at the ratio Y:B:Cu-1:2:3 and variable concentration of urea in the solution. Mentioned aqua-salt compositions are used in cryochemical synthesis of HTSC. Solutions of yttrium, barium and copper nitrates with urea additions were demonstrated to be recommended to improvements in the process during development of alternative synthesis from aqua-salt compositions. 15 refs., 3 figs., 2 tabs

  5. Retention and translocation of inhaled uranyl nitrate (233U and 232U) in rats

    International Nuclear Information System (INIS)

    Ballou, J.E.; Gies, R.A.; Wogman, N.A.

    1978-01-01

    The uranium-thorium breeder reactors proposed for nuclear power production, and other thorium fuel systems in conventional reactors, utilize fuels and fuel recycle process solutions that have not been evaluated for biological hazard. This project emphasizes studies of the metabolism of the oxide fuels and the nitrate solutions of the major radionuclides, following inhalation, ingestion, or cutaneous application in rodents. Preliminary data are reported for the clearance of inhaled 233 UO 2 (NO 3 ) 2 and 232 UO 2 (NO 3 ) 2 from the lung and their translocation to skeleton

  6. Characterization of aqueous silver nitrate solutions for leakage tests

    Directory of Open Access Journals (Sweden)

    José Ferreira Costa

    2011-06-01

    Full Text Available OBJECTIVE: To determine the pH over a period of 168 h and the ionic silver content in various concentrations and post-preparation times of aqueous silver nitrate solutions. Also, the possible effects of these factors on microleakage test in adhesive/resin restorations in primary and permanent teeth were evaluated. MATERIAL AND METHODS: A digital pHmeter was used for measuring the pH of the solutions prepared with three types of water (purified, deionized or distilled and three brands of silver nitrate salt (Merck, Synth or Cennabras at 0, 1, 2, 24, 48, 72, 96 and 168 h after preparation, and storage in transparent or dark bottles. Ionic silver was assayed according to the post-preparation times (2, 24, 48, 72, 96, 168 h and concentrations (1, 5, 25, 50% of solutions by atomic emission spectrometry. For each sample of each condition, three readings were obtained for calculating the mean value. Class V cavities were prepared with enamel margins on primary and permanent teeth and restored with the adhesive systems OptiBond FL or OptiBond SOLO Plus SE and the composite resin Filtek Z-250. After nail polish coverage, the permanent teeth were immersed in 25% or 50% AgNO3 solution and the primary teeth in 5% or 50% AgNO3 solutions for microleakage evaluation. ANOVA and the Tukey's test were used for data analyses (α=5%. RESULTS: The mean pH of the solutions ranged from neutral to alkaline (7.9±2.2 to 11.8±0.9. Mean ionic silver content differed depending on the concentration of the solution (4.75±0.5 to 293±15.3 ppm. In the microleakage test, significant difference was only observed for the adhesive system factor (p=0.000. CONCLUSIONS: Under the tested experimental conditions and based on the obtained results, it may be concluded that the aqueous AgNO3 solutions: have neutral/alkaline pH and service life of up to 168 h; the level of ionic silver is proportional to the concentration of the solution; even at 5% concentration, the solutions were

  7. Fabrication of ceramic grade UO2 by direct conversion of uranyl nitrate hexahydrate

    International Nuclear Information System (INIS)

    Lainetti, P.E.O.; Riella, H.G.

    1992-01-01

    A method of direct conversion of uranyl nitrate hexahydrate (UNH) solution to ceramic grade uranium dioxide powders by thermal denitration in a furnace that combines atomization nozzle and a gas stirred bed is described. The main purpose of this work is to show that this alternative process is technically viable, specially if the recovery of the scrap generated in the nuclear fuel pellet production is required, without further generation of new liquid wastes. (author)

  8. Method of chemical reprocessing of irradiated nuclear fuels (especially fuels containing uranium)

    International Nuclear Information System (INIS)

    Koch, G.

    1975-01-01

    The invention deals with a method for the extraction especially of fast breeder fuels of high burn-up. A quaternary ammonium nitrate of high molecular weight is put into an organic diluting medium as extraction agent, corresponding to the general formula NRR'R''R'''NO 3 where R,R' and R'' are aliphatic radicals, R''' a methyl radical and the sum of the C atoms is greater than 16. After the extraction of the aqueous nitric acid containing nuclear fuel solution with this extracting agent, uranium, plutonium (or also thorium) can be found to a very high percentage in the organic phase and can be practically quantitatively back-extracted by means of diluted nitric acid, sulphuric acid or acetic acid. By using 30 volume percent tricapryl methyl ammonium nitrate in diethyl benzene for example, a distribution coefficient of 10.3 is obtained for uranium. (RB/LH) [de

  9. Quantitative determination of uranium in organic solution by X-ray fluorescence

    International Nuclear Information System (INIS)

    Leyt, D.V. de; Colangelo, C.H.

    1987-01-01

    An X-ray fluorescent method for the determination of uranium in tributilphosphate-kerosene-nitriacid solution has been developed. Chemical properties of the matrix elements were studied in order to select a convenient procedure to determine samples and standards on the same way. The method avoids the destruction of the organic material and has proved to be very useful for the fast control of uranium concentration. (Author) [es

  10. Subcritical measurements with a cylindrical tank of Pu-U nitrate

    International Nuclear Information System (INIS)

    Mihalczo, J.T.; Valentine, T.E.; King, W.T.

    1997-01-01

    This series of measurements with a mixed Pu-U nitrate solution (280 g Pu/liter, 180 g U/liter) in a 35.54-cm-diam cylindrical tank provides a wide variety of experimental data for subcritical configurations that can be used to verify calculational methods and nuclear data. The Pu contained 7.85 wt% 240 Pu and the uranium was natural uranium. The measurements performed were: inverse count rate, prompt neutron decay constants, inverse kinetics, and frequency analysis by the 252 Cf source driven method. These data are presented in sufficient detail that the results of the experiments can be calculated directly. For purposes of extrapolating to the delayed critical height the ratio of spectral densities was linear with height and thus provided the best estimate of critical height

  11. Water-Reflected 233U Uranyl Nitrate Solutions in Simple Geometry

    International Nuclear Information System (INIS)

    Elam, K.R.

    2001-01-01

    A number of critical experiments involving 233 U were performed in the Oak Ridge National Laboratory Building 9213 Critical Experiments Facility during the years 1952 and 1953. These experiments, reported in Reference 1, were directed toward determining bounding values for the minimum critical mass, minimum critical volume, and maximum safe pipe size of water-moderated solutions of 233 U. Additional information on the critical experiments was found in the experimental logbooks. Two experiments utilizing uranyl nitrate (UO 2 (NO 3 ) 2 ) solutions in simple geometry are evaluated in this report. Experiment 37 is in a 10.4-inch diameter sphere, and Experiment 39 is in a 10-inch diameter cylinder. The 233 U concentration ranges from 49 to 62 g 233 U/l. Both experiments were reflected by at least 6 inches of water in all directions. Paraffin-reflected uranyl nitrate experiments, also reported in Reference 1, are evaluated elsewhere. Experiments with smaller paraffin reflected 5-, 6-, and 7.5-inch diameter cylinders are evaluated in U233-SOL-THERM-004. Experiments with paraffin reflected 8-, 8.5-, 9-, 10-, and 12-inch diameter cylinders are evaluated in U233-SOL-THERM-002. Later experiments with highly-enriched 235 U uranyl fluoride solution in the same 10.4-inch diameter sphere are reported in HEU-SOL-THERM-010. Both experiments were judged acceptable for use as criticality-safety benchmark experiments

  12. Analysis of cadmium in high alpha solutions

    International Nuclear Information System (INIS)

    Gray, L.W.; Overman, L.A.; Hodgens, H.F.

    1977-07-01

    Cadmium nitrate is occasionally used as a neutron poison for convenience in the separation of uranium, neptunium, and plutonium. As the classical methods of analysis for cadmium are very time-consuming, a method to isolate it in solution using solvent extraction of uranium, neptunium, and plutonium with TBP in an n-paraffin hydrocarbon was investigated. After removal of the radionuclides, the cadmium is determined by atomic absorption spectroscopy. Precision of the method at the 95 percent confidence level is +-2.4 percent. Alpha content of the solutions was typically reduced from 1-10 x 10 11 dis/(min ml) 238 Pu to 1-15 x 10 4 dis/(min ml). Analysis time was typically reduced from approximately 24 hours per sample to less than 1 hour

  13. Development of a technique for the on line determination of uranium in solution by gamma ray spectroscopy

    International Nuclear Information System (INIS)

    Singh, Sarabjit; Ramaswami, A.; Gill, Jatinder Singh

    2005-02-01

    A technique based on gamma ray spectrometry has been developed for the continuous monitoring of uranium in the solution form. Simulated container and support system was designed and fabricated for the development of an efficiency calibration curve and to find the detection limit for the estimation of uranium using 185.7 keV ( 235 U) gamma ray. The system was calibrated for its counting efficiency using HPGe detector system, in a standard source mount to detector geometry. The sensitivity of the detection system and counting time for low-level estimation of uranium has also been established. The detection limit of the monitor is ∼10 mg of uranium per litre of the solution. In order to correct for the density variation of the solution experiment was carried to study the variation of count rate of 185.7 ke V gamma ray of 235 U as a function of the density of the solution. This report gives the details of the development of a continuous monitor for the determination of uranium in the solution streams. (author)

  14. Coordinated safeguards for materials management in a uranium--plutonium nitrate-to-oxide coconversion facility: Coprecal

    International Nuclear Information System (INIS)

    Dayem, H.A.; Cobb, D.D.; Dietz, R.J.; Hakkila, E.A.; Kern, E.A.; Schelonka, E.P.; Shipley, J.P.; Smith, D.B.

    1979-02-01

    This report describes the conceptual design of an advanced materials-management system for safeguarding special nuclear materials in a uranium--plutonium nitrate-to-oxide coconversion facility based on the Coprecal process. Design concepts are presented for near real-time (dynamic) accountability by forming dynamic materials balances from information provided by chemical and nondestructive analyses and from process-control instrumentation. Modeling and simulation techniques are used to compare the sensitivities of proposed dynamic materials accounting strategies to both abrupt and protracted diversion. The safeguards implications of coconversion as well as some unique features of the reference process are discussed and design criteria are identified to improve the safeguardability of the Coprecal coconversion process

  15. Chemical denitration of aqueous nitrate solutions

    International Nuclear Information System (INIS)

    Burrill, K.A.

    1987-11-01

    The Plant for Active Waste Liquids (PAWL) at CRNL will immobilize in glass the fission products in waste from Mo-99 production. The nitrate ions in the waste can be destroyed by heating, but also by chemical reaction with formic acid (HCOOH). Since chemical denitration has several advantages over thermal denitration it was studied in the course of vitrification process development. Two free radical mechanisms are examined here to explain kinetic data on chemical denitration of nitric acid solutions with formic acid. One mechanism is applicable at > 1 mol/L HNO 3 and involves the formate radical (HCOO . ). The second mechanism holds at 3 and involves the hyponitrous radical (HNO . ). Mass balances for various species were written based on the law of mass action applied to the equations describing the reaction mechanism. Analytical and numerical solutions were obtained and compared. Literature data on batch denitration were used to determine some of the rate constants while others were set arbitrarily. Observed stoichiometry and trends in reactant concentrations are predicted accurately for batch data. There are no literature data to compare with the prediction of negligible induction time

  16. Chemical effects induced by dissolving γ-irradiated alkali halides in aqueous nitrate, permanganate and chromate solutions

    International Nuclear Information System (INIS)

    Phansalkar, V.K.; Bapat, L.; Ravishankar, D.

    1982-01-01

    Dissolution of γ-irradiated alkali halides in aqueous solutions of sodium nitrate, potassium permanganate and potassium chromate at neutral pH induces chemical changes leading to the formation of NO 2 - in nitrate, Mn(IV) and Cr(III) species in permanganate and chromate solutions, respectively. Further, the studies on nitrate and permanganate systems show that the amount of NO 2 - and Mn(IV) formed grows by increasing the dose of γ-irradiation of the salt and the amount of irradiated salt. Moreover, the extent of chemical changes effected by irradiated chlorides has been found to be more than that of bromides. The mesh size of the irradiated salt and the presence of scavengers like I - and methanol in the system, affects the yield of NO 2 - . (author)

  17. recovery of enriched uranium from waste solution obtained from fuel fabrication laboratories

    International Nuclear Information System (INIS)

    Othman, S.H.A.

    2003-01-01

    reversed-phase partition chromatography is shown to be a convenient and applicable method for the quantitative recovery of uranium (19.7% enriched with 235 U) from highly impure solution . the processing of uranium compounds for atomic energy project especially in FMPP(Egyptian fuel manufacture pilot plant) gives rise to a variety of wastes in which the uranium content is of considerable importance. the recovery of uranium from concentrated mother liquors produced from ADU (ammonium diuranate ) precipitation, as well as those due to ADU washing is studied in this work. column of poly-trifluoro-monochloro-ethilene (Kel-F) supporting tri-n-butyl-phosphate (TBP) retains uranium .impurities are eluted with 6.5 M HCl, and the uranium is eluted with water and the recovery of uranium is better than 94%. A mathematical model was suggested to stimulate the sorption process of uranium ions (or any other ion ) by column of solvent impregnated resin containing organic extractant (the same as the previous column) . An excellent agreement was founded between the experimental results and the mathematical model

  18. 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)

  19. Uranous nitrate production for purex process applications using PtO2 catalyst and H2/H2-gas mixtures

    International Nuclear Information System (INIS)

    Sreenivasa Rao, K.; Shyamali, R.; Narayan, C.V.; Patil, A.R.; Jambunathan, U.; Ramanujam, A.; Kansara, V.P.

    2003-04-01

    In the Purex process of spent fuel reprocessing. the twin objectives- decontamination and partitioning are achieved by extracting uranium (VI) and plutonium (IV) together in the solvent 30% TBP-dodecane and then selectively reducing Pu (IV) to Pu (III) in which valency it is least extractable in the solvent. Uranous nitrate stabilized with hydrazine nitrate is the widely employed partitioning agent. The conventional method of producing U(IV) is by the electrolytic reduction of uranyl nitrate with hydrazine nitrate as uranous ion stabilizer. Tre percentage conversion of U(VI) to U(IV) obtained in this method is 50 -60 %. The use of this solution as partitioning agent leads not only to the dilution of the plutonium product but also to increase in uranium processing load by each externally fed uranous nitrate batch. Also the oxide coating of the anode, TSIA (Titanium Substrate Insoluble Anode) wears out after a certain period of operation. This necessitates recoating which is quite cumbersome considering the amount of the decontamination involved. An alternative to the conventional electrolytic method of reduction of uranyl nitrate to uranous nitrate was explored at FRD laboratory .The studies have revealed that near 100% uranous nitrate can be produced by reducing uranyl nitrate with H 2 gas or H 2 (8%)- Ar/N 2 gas mixture in presence of PtO 2 catalyst. This report describes the laboratory scale studies carried out to optimize the various parameters. Based on these studies reduction of uranyl nitrate on a pilot plant scale was carried out. The design and operation of the reductor column and also the various studies carried out in the pilot plant studies are discussed. Near 100% conversion of uranyl nitrate to uranous nitrate and also the redundancy of supply of electrical energy make this process a viable alternative to the existing electrolytic method. (author)

  20. High temperature interaction studies on equimolar nitrate mixture of uranyl nitrate hexahydrate and gadolinium nitrate hexahydrate

    International Nuclear Information System (INIS)

    Kalekar, Bhupesh B.; Raje, Naina; Reddy, A.V.R.

    2015-01-01

    Rare earths including gadolinium form a sizeable fraction of the fission products in the nuclear fission of fissile material in the reactor. These fission products can interact with uranium dioxide fuel and can form various compounds which can alter the thermal behavior of the fuel. The mixed oxide formed due to the high temperature interactions of mixture of uranyl nitrate hexahydrate (UNH) and gadolinium nitrate hexahydrate (GdNH) has been studied using thermal and X- ray diffraction techniques. The equimolar mixture of UNH and GdNH was prepared by mixing the weighed amount of individual nitrates and grinding gently with mortar and pestle. Thermogravimetry (TG) measurements were carried out by separately heating 100 mg of mixture and individual nitrates at heating rate of 10°C min -1 using Netzsch thermal analyzer (Model No.: STA 409 PC Luxx) in high purity nitrogen atmosphere with a flow rate of 120 mL min -1 . The XRD measurement was carried out on a Philips X-ray diffractometer (Model PW1710) using nickel-filtered Cu-Kα radiation

  1. Improvement of INVS Measurement Uncertainty for Pu and U-Pu Nitrate Solution

    Energy Technology Data Exchange (ETDEWEB)

    Swinhoe, Martyn Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Menlove, Howard Olsen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Marlow, Johnna Boulds [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Makino, Risa [Japan Atomic Energy Agency (JAEA), Tokai (Japan); Nakamura, Hironbu [Japan Atomic Energy Agency (JAEA), Tokai (Japan)

    2017-04-27

    In the Tokai Reprocessing Plant (TRP) and the Plutonium Conversion Development Facility (PCDF), a large amount of plutonium nitrate solution which is recovered from light water reactor (LWR) and advanced thermal reactor (ATR), FUGEN are being stored. Since the solution is designated as a direct use material, the periodical inventory verification and flow verification are being conducted by Japan Safeguard Government Office (JSGO) and International Atomic Agency (IAEA).

  2. Energy dispersive X-ray fluorescence determination of cadmium in uranium matrix using Cd Kα line excited by continuum

    International Nuclear Information System (INIS)

    Dhara, Sangita; Misra, N.L.; Aggarwal, S.K.; Venugopal, V.

    2010-01-01

    An energy dispersive X-ray fluorescence method for determination of cadmium (Cd) in uranium (U) matrix using continuum source of excitation was developed. Calibration and sample solutions of cadmium, with and without uranium were prepared by mixing different volumes of standard solutions of cadmium and uranyl nitrate, both prepared in suprapure nitric acid. The concentration of Cd in calibration solutions and samples was in the range of 6 to 90 μg/mL whereas the concentration of Cd with respect to U ranged from 90 to 700 μg/g of U. From the calibration solutions and samples containing uranium, the major matrix uranium was selectively extracted using 30% tri-n-butyl phosphate in dodecane. Fixed volumes (1.5 mL) of aqueous phases thus obtained were taken directly in specially designed in-house fabricated leak proof Perspex sample cells for the energy dispersive X-ray fluorescence measurements and calibration plots were made by plotting Cd Kα intensity against respective Cd concentration. For the calibration solutions not having uranium, the energy dispersive X-ray fluorescence spectra were measured without any extraction and Cd calibration plots were made accordingly. The results obtained showed a precision of 2% (1σ) and the results deviated from the expected values by < 4% on average.

  3. Optimizing Electrocoagulation Process for the Removal of Nitrate From Aqueous Solution

    Directory of Open Access Journals (Sweden)

    Dehghani

    2016-01-01

    Full Text Available Background High levels of nitrate anion are frequently detected in many groundwater resources in Fars province. Objectives The present study aimed to determine the removal efficiency of nitrate from aqueous solutions by electrocoagulation process using aluminum and iron electrodes. Materials and Methods A laboratory-scale batch reactor was conducted to determine nitrate removal efficiency using the electrocoagulation method. The removal of nitrate was determined at pH levels of 3, 7, and 11, different voltages (15, 20, and 30 V, and operation times of 30, 60, and 75 min, respectively. Data were analyzed using the SPSS software version 16 (Chicago, Illinois, USA and Pearson’s correlation coefficient was used to analyze the relationship between the parameters. Results Results of the present study showed that the removal efficiency was increased from 27% to 86% as pH increased from 3 to 11 at the optimal condition of 30 V and 75 min operation time. Moreover, by increasing the reaction time from 30 V to 75 min the removal efficiency was increased from 63% to 86%, respectively (30 V and pH = 11. Pearson’s correlation analysis showed that there was a significant relationship between removal efficiency and voltage and reaction time as well (P < 0.01. Conclusions In conclusion, the electrocoagulation process can be used for removing nitrate from water resources because of high efficiency, simplicity, and relatively low cost.

  4. Transport and fate of ammonium and its impact on uranium and other trace elements at a former uranium mill tailing site

    International Nuclear Information System (INIS)

    Miao, Ziheng; Akyol, Hakan N.; McMillan, Andrew L.; Brusseau, Mark L.

    2013-01-01

    Highlights: • Nitrification of ammonium evidenced by stable isotopes of nitrate at a mining site. • Concentrations of uranium and other trace elements related to ammonium conc. • Observed impact of ammonium on redox, pH, and possibly complexation. • Proposed impact of transformation of NO 3 and NH 4 on trace elements. - Abstract: The remediation of ammonium-containing groundwater discharged from uranium mill tailing sites is a difficult problem facing the mining industry. The Monument Valley site is a former uranium mining site in the southwest US with both ammonium and nitrate contamination of groundwater. In this study, samples collected from 14 selected wells were analyzed for major cations and anions, trace elements, and isotopic composition of ammonium and nitrate. In addition, geochemical data from the U.S. Department of Energy (DOE) database were analyzed. Results showing oxic redox conditions and correspondence of isotopic compositions of ammonium and nitrate confirmed the natural attenuation of ammonium via nitrification. Moreover, it was observed that ammonium concentration within the plume area is closely related to concentrations of uranium and a series of other trace elements including chromium, selenium, vanadium, iron, and manganese. It is hypothesized that ammonium–nitrate transformation processes influence the disposition of the trace elements through mediation of redox potential, pH, and possibly aqueous complexation and solid-phase sorption. Despite the generally relatively low concentrations of trace elements present in groundwater, their transport and fate may be influenced by remediation of ammonium or nitrate at the site

  5. Determination of uranium in aqueous solutions by HPLC post column UV/VIS detection

    International Nuclear Information System (INIS)

    Abbasi, W.A.; Street, M.

    1995-01-01

    Analysis of uranium in different nitric acid concentration solutions is interest in nuclear industry where it is present in various process and waste streams. A method for the determination of uranium(VI) in nitric acid solution in described using ion chromatography and post-column derivatization with 4-(2-pyridylazo) resorcinol (PAR). The method does not require any sample treatment for acid removal or neutralization and can be used for routine laboratory analysis. The effluent used was ammonium sulphate / sulphuric acid and PAR was present in ammonium hydroxide/acetic acid buffer. The column effluents is reacted with PAR in a reaction coil and U(VI)-PAR complex is detector at 528 nm. The lowest uranium concentration detected was 100 ppb (0.1 mg/sup -1/) and linearity of up to 8 mgl -1 concentration was verified. The standard deviation in reproducibility at uranium concentration of 0.1 mgl/sup -1/ was 0.82%. (author)

  6. Separation of neptunium from uranium and plutonium in the Purex process

    International Nuclear Information System (INIS)

    Kolarik, Z.; Schuler, R.

    1984-01-01

    The possibility of removing neptunium from the Purex process in the first extraction cycle was investigated. Butyraldehyde was found to reduce Np(VI) to Np(V), but not Pu(IV) to Pu(III). Up to 99.7% Np can be separated from uranium and plutonium in the 1A extractor or, much more favourably, in an additional partitioning extractor. Hydroxylamine nitrate can be used for reducing Np(VI) to Np(V) in a uranium purification cycle at a high U concentration in the feed solution. Here the decontamination factor for Np can be as high as 2300 and is lowered if iron is present in the feed. (author)

  7. Bois-Noirs ore. Recovery of uranium of solutions from acid treatment. Results of industrial tests at the Gueugnon plant; Minerai des Bois-Noirs. Recuperation de l'uranium des solutions d'attaques acides. Resultats des essais industriels effectues a l'usine de Gueugnon

    Energy Technology Data Exchange (ETDEWEB)

    Le Bris, J

    1959-04-01

    Industrial-scale tests are reported of the efficiency of two recovery processes for the separation of uranium from sulfuric acid pickling solutions used on ore from Bois-Noirs, at the Gueugnon works. The final stage of each process is sodium uranate. The earlier part of the report deals with tests of the separation of uranium from foreign metals by fractional precipitation. The second part deals with the separation of uranium from these metals by carbonation of the solutions. (author) [French] Le present rapport concerne les essais industriels de deux procedes de recuperation de l'uranium de solutions d'attaque sulfurique du minerai des Bois-Noirs a l'usine de Gueugnon. Le stade final pour ces deux procedes etant l'uranate de sodium, une premiere partie est consacree aux essais de separation de l'uranium des metaux etrangers par precipitation fractionnee; une deuxieme partie est consacree aux essais de separation de l'uranium des metaux etrangers par carbonatation des solutions d'attaque du minerai. (auteur)

  8. Reverse osmosis performance of cellulose acetate membranes in the separation of uranium from dilute solutions

    International Nuclear Information System (INIS)

    Sastri, V.S.; Ashbrook, A.W.

    1976-01-01

    Batch 316-type cellulose acetate membranes were characterized in terms of pure water permeability constant, solute transport parameter, and mass transfer coefficient with a reference system of aqueous sodium chloride solution. These membranes were used in the determination of reverse osmosis characteristics such as product rate and solute separation in the case of uranium sulfate solutions of different concentrations (100 to 8000 ppM) in the feed solutions. A long-term test extending over a week has been carried out with dilute uranium solutions. Reverse osmosis treatment of synthetic mine water sample showed satisfactory performance of the membranes in the separation of metal ions

  9. 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

  10. Root uptake of uranium (6) in solution by a higher plant: speciation in hydroponic solution, bioavailability, micro-localisation and biological effects induced

    International Nuclear Information System (INIS)

    Laroche, L.

    2005-01-01

    Uranium exists naturally in the environment, usually present in trace quantities. In soil solution and oxic conditions, uranium is present in the +VI oxidation state and forms a large number of inorganic and organic complexes. The exposure medium, an artificial soil solution, was designed in such a way as to control the uranium species in solution. The geochemical speciation code JCHESS was used to calculate the uranium aqueous species concentration and to define the domains of interest, each of them characterized by a limited number of dominant U species. These domains were defined as follows: pH 4.9 with uranyl ions as dominant species, pH 5.8 with hydroxyl complexes and pH 7 where carbonates play a major role. For each pH, short-duration (5 hours of exposure) well-defined laboratory experiments were carried out with Phaseolus vulgaris as plant model. The effect of competitive ions such as Ca 2+ or the presence of ligands such as phosphate or citrate on root assimilation efficiency was explored. Results have shown that uranium transfer was not affected by the presence of calcium, phosphate or citrate (but was decreased of 60% with citrate (10 μM) at pH 5.8) in our experimental conditions. Moreover, observation in Transmission Electronic Microscopy (TEM), equipped with an EDAX probe, have shown that uranium was associated with granules rich in phosphorus and that there were some chloroplast anomalies. Finally, the presence of uranium affects root CEC by reducing it and stimulates root elongation at low uranium concentrations (100 nM, 400 nM and 2 μM at pHs 4.9, 5.8 and 7 respectively) and inhibits it at high uranium concentrations. (author)

  11. Selective extraction of metals from acidic uranium(VI) solutions using neo-tridecano-hydroxamic acid

    International Nuclear Information System (INIS)

    Bardoncelli, F.; Grossi, G.

    1975-01-01

    According to this invention neo-alkyl-hydroxamic acids are employed as ion-exchanging agents in processes for liquid-liquid extraction with the aim of separating, purifying dissolved metals and of converting a metal salt solution into a solution of a salt of the same metal but with different anion. In particular it is an objective of this invention to provide a method whereby a molecular pure uranium solution is obtained by selective extraction from a uranium solution delivered by irradiated fuel reprocessing plants and containing plutonium, fission products and other unwanted metals, in which method neo-tridecane-hydroxamic acid is employed as ion exchanger. (Official Gazette)

  12. Preparation of the pur uranium-metal; La preparation de l'uranium-metal pur

    Energy Technology Data Exchange (ETDEWEB)

    Goldschmidt, B.; Vertes, P. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    A detailed description of the chemical processes used to prepare in the factory of Bouchet of the CEA (Seine-Et-Oise) pur metal uranium with either relatively rich ores, or extracts coming of physical or chemical treatment of poor ores. The nitric treatment of ores succeeds to the production of uranate of impure sodium carbonate. This last last product is dissolved in nitric acid and the uranyl nitrate is extracted by tributyl-phosphate diluted in an inert solvent. The uranyl nitrate pure is re-extracted and successively transformed in uranium peroxide, in orange oxide then in brown oxide which is transformed in fluoride by the anhydrous hydrofluoric acid. Uranate fluoride is then reduced in metal by the pure calcium with an yield superior to 99%. (authors) [French] Description detaillee des procedes chimiques mis en jeu pour preparer a l'Usine du Bouchet du Commissariat a l'Energie Atomique (Seine-et-Oise) l'uranium metal pur a partir soit de minerais relativement riches, soit de concentres provenant de traitement physique ou chimique de minerais pauvres. Le traitement nitrique des minerais aboutit a la production d'uranate de soude impur. Ce dernier est a son tour dissous dans l'acide nitrique et le nitrate d'uranyle est extrait par du tributyl-phosphate dilue par un solvant inerte. Le nitrate d'uranyle pur reextrait est transforme successivement en peroxyde d'uranium, en oxyde orange puis en oxyde brun qui est transforme en fluorure par l'acide fluorhydrique anhydre. Le fluorure uraneux est reduit en metal par le calcium pur avec un rendement superieur a 99 %. (auteurs)

  13. Sensitivity analysis of high resolution gamma-ray detection for safeguards monitoring at natural uranium conversion facilities

    Energy Technology Data Exchange (ETDEWEB)

    Dewji, S.A., E-mail: dewjisa@ornl.gov [Oak Ridge National Laboratory, PO Box 2008 MS-6335, Oak Ridge TN 37831 (United States); Georgia Institute of Technology, 770 State Street, Atlanta, GA 30332-0745 (United States); Croft, S. [Oak Ridge National Laboratory, PO Box 2008 MS-6335, Oak Ridge TN 37831 (United States); Hertel, N.E. [Oak Ridge National Laboratory, PO Box 2008 MS-6335, Oak Ridge TN 37831 (United States); Georgia Institute of Technology, 770 State Street, Atlanta, GA 30332-0745 (United States)

    2017-03-11

    Under the policies proposed by recent International Atomic Energy Agency (IAEA) circulars and policy papers, implementation of safeguards exists when any purified aqueous uranium solution or uranium oxides suitable for isotopic enrichment or fuel fabrication exists. Under IAEA Policy Paper 18, the starting point for nuclear material under safeguards was reinterpreted, suggesting that purified uranium compounds should be subject to safeguards procedures no later than the first point in the conversion process. In response to this technical need, a combination of simulation models and experimental measurements were employed in previous work to develop and validate gamma-ray nondestructive assay monitoring systems in a natural uranium conversion plant (NUCP). In particular, uranyl nitrate (UO{sub 2}(NO{sub 3}){sub 2}) solution exiting solvent extraction was identified as a key measurement point (KMP). Passive nondestructive assay techniques using high resolution gamma-ray spectroscopy were evaluated to determine their viability as a technical means for drawing safeguards conclusions at NUCPs, and if the IAEA detection requirements of 1 significant quantity (SQ) can be met in a timely manner. Building upon the aforementioned previous validation work on detector sensitivity to varying concentrations of uranyl nitrate via a series of dilution measurements, this work investigates detector response parameter sensitivities to gamma-ray signatures of uranyl nitrate. The full energy peak efficiency of a detection system is dependent upon the sample, geometry, absorption, and intrinsic efficiency parameters. Perturbation of these parameters translates into corresponding variations of the 185.7 keV peak area of the {sup 235}U in uranyl nitrate. Such perturbations in the assayed signature impact the quality or versatility of the safeguards conclusions drawn. Given the potentially high throughput of uranyl nitrate in NUCPs, the ability to assay 1 SQ of material requires

  14. Contribution to the study of uranyl salts in butyl phosphate solutions; Contribution a l'etude des solutions de sels d'uranyle dans les phosphates butyliques

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-06-01

    A spectroscopic study in the normal infrared region and involving the following associations: tri-alkyl phosphates (tri-butyl, tri-ethyl, tri-methyl), uranyl salts (nitrate, chloride, acetate) has confirmed the existence of an interaction between the phosphoryl group and the uranium atom, as shown by a movement of absorption band for the valency P = 0 from {approx} 1270 cm{sup -1} to {approx} 1180 cm{sup -1}. A study of the preparation, analysis and spectroscopy of the solids obtained by the precipitation of uranyl salts by acid butyl phosphates has been carried out. By infrared spectrophotometry it has been shown that the tri-butyl and di-butyl phosphates are associated in non-polar diluents even before the uranium is introduced. The extraction of uranyl salts from acid aqueous solutions by a diluted mixture of tri-butyl and di-butyl phosphates proceeds by different mechanisms according to the nature of the ion (nitrate or chloride). (author) [French] Une etude spectroscopique dans l'infrarouge moyen portant sur les associations: - phosphates trialcoyliques (tributylique - triethylique - trimethylique) - sels d'uranyle (nitrate, chlorure, acetate) a confirme l'existence d'une interaction entre le groupement phosphoryle et l'atome d'uranium, se manifestant par un deplacement de la bande d'absorption de la vibration de valence P = 0 de {approx} 1270 cm{sup -1} a {approx} 1180 cm{sup -1}. Une etude preparative, analytique et spectroscopique des solides obtenus par precipitation de sels d'uranyle par les phosphates butyliques acides a ete effectuee. La spectrophotomerie infrarouge met en evidence l'association, anterieure a toute introduction d'uranium, des phosphates tributylique et dibutylique dans des diluants non polaires. L'extraction de sels d'uranyle, d'une solution aqueuse acide par un melange dilue de phosphates tributylique et dibutylique, s'effectue suivant des processus differents a la

  15. Environmental impact of solution mining for uranium

    International Nuclear Information System (INIS)

    Hunkin, G.G.

    1975-01-01

    Compared with most other mining systems, uranium solution mining has a negligible effect on such environmental factors as surface disturbance, interference with natural groundwater quality and distribution, and aerial discharge of radionuclides. The following topics are discussed: the process, personnel safety and health, tailings disposal, impact on groundwater, operating licenses and controls, legislation, and economics. It is concluded that engineered well systems and controlled input/production flow rates, combined with full recirculation systems that maintain constant fluid volumes in the mineralized formations, result in containment of leach solutions within the operating area. The very dilute leach solutions, compatible with natural groundwaters, ensure that no environmental damage results, even if a loss of control occurred. Reduction in the number of processing steps and virtual elimination of operator hazards, waste disposal and land rehabilitation costs help in reducing overall costs

  16. On the influence of molecular structure on the conductivity of electrolyte solutions - sodium nitrate in water

    Directory of Open Access Journals (Sweden)

    H. Krienke

    2013-01-01

    Full Text Available Theoretical calculations of the conductivity of sodium nitrate in water are presented and compared with experimental measurements. The method of direct correlation force in the framework of the interionic theory is used for the calculation of transport properties in connection with the associative mean spherical approximation (AMSA. The effective interactions between ions in solutions are derived with the help of Monte Carlo and Molecular Dynamics calculations on the Born-Oppenheimer level. This work is based on earlier theoretical and experimental studies of the structure of concentrated aqueous sodium nitrate solutions.

  17. Modeling of critical experiments employing Raschig rings in uranyl nitrate solution

    International Nuclear Information System (INIS)

    Tanner, J.E.

    1989-01-01

    Four critical experiments employing borated glass rings in concentrated uranyl nitrate solution yielded k eff higher by 0. 04 when modeled with a flux-weighted, homogenized cross section set than when modeled with discrete rings. k eff varied by 0.014 for a 10% boron uncertainty and by up to 0.04 for a 10% packing fraction uncertainty

  18. A study on prediction of uranium concentration in pregnant solution from in-situ leaching

    International Nuclear Information System (INIS)

    Yi Weiping; Zhou Quan; Yu Yunzhen; Wang Shude; Yang Yihan; Lei Qifeng

    2005-01-01

    The modeling course on prediction of uranium concentration in pregnant solution from in-situ leaching of uranium is described, a mathematical model based on grey system theory is put forward, and a set of computer application software is correspondingly developed. (authors)

  19. Electrochemical processing of nitrate waste solutions. Phase 2, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Genders, D.; Weinberg, N.; Hartsough, D. [Electrosynthesis Co., Inc., Cheektowaga, NY (US)

    1992-10-07

    The second phase of research performed at The Electrosynthesis Co., Inc. has demonstrated the successful removal of nitrite and nitrate from a synthetic effluent stream via a direct electrochemical reduction at a cathode. It was shown that direct reduction occurs at good current efficiencies in 1,000 hour studies. The membrane separation process is not readily achievable for the removal of nitrites and nitrates due to poor current efficiencies and membrane stability problems. A direct reduction process was studied at various cathode materials in a flow cell using the complete synthetic mix. Lead was found to be the cathode material of choice, displaying good current efficiencies and stability in short and long term tests under conditions of high temperature and high current density. Several anode materials were studied in both undivided and divided cell configurations. A divided cell configuration was preferable because it would prevent re-oxidation of nitrite by the anode. The technical objective of eliminating electrode fouling and solids formation was achieved although anode materials which had demonstrated good stability in short term divided cell tests corroded in 1,000 hour experiments. The cause for corrosion is thought to be F{sup {minus}} ions from the synthetic mix migrating across the cation exchange membrane and forming HF in the acid anolyte. Other possibilities for anode materials were explored. A membrane separation process was investigated which employs an anion and cation exchange membrane to remove nitrite and nitrate, recovering caustic and nitric acid. Present research has shown poor current efficiencies for nitrite and nitrate transport across the anion exchange membrane due to co-migration of hydroxide anions. Precipitates form within the anion exchange membranes which would eventually result in the failure of the membranes. Electrochemical processing offers a highly promising and viable method for the treatment of nitrate waste solutions.

  20. Nitrate conversion and supercritical fluid extraction of UO2-CeO2 solid solution prepared by an electrolytic reduction-coprecipitation method

    International Nuclear Information System (INIS)

    Zhu, L.Y.; Duan, W.H.; Wen, M.F.; Xu, J.M.; Zhu, Y.J.

    2014-01-01

    A low-waste technology for the reprocessing of spent nuclear fuel (SNF) has been developed recently, which involves the conversion of actinide and lanthanide oxides with liquid N 2 O 4 into their nitrates followed by supercritical fluid extraction of the nitrates. The possibility of the reprocessing of SNF from high-temperature gas-cooled reactors (HTGRs) with nitrate conversion and supercritical fluid extraction is a current area of research in China. Here, a UO 2 -CeO 2 solid solution was prepared as a surrogate for a UO 2 -PuO 2 solid solution, and the recovery of U and Ce from the UO 2 -CeO 2 solid solution with liquid N 2 O 4 and supercritical CO 2 containing tri-n-butyl phosphate (TBP) was investigated. The UO 2 -CeO 2 solid solution prepared by electrolytic reduction-coprecipitation method had square plate microstructures. The solid solution after heat treatment was completely converted into nitrates with liquid N 2 O 4 . The XRD pattern of the nitrates was similar to that of UO 2 (NO 3 ) 2 . 3H 2 O. After 120 min of online extraction at 25 MPa and 50 , 99.98% of the U and 98.74% of the Ce were recovered from the nitrates with supercritical CO 2 containing TBP. The results suggest a promising potential technology for the reprocessing of SNF from HTGRs. (orig.)

  1. Microbially catalyzed nitrate-dependent metal/radionuclide oxidation in shallow subsurface sediments

    Science.gov (United States)

    Weber, K.; Healy, O.; Spanbauer, T. L.; Snow, D. D.

    2011-12-01

    Anaerobic, microbially catalyzed nitrate-dependent metal/radionuclide oxidation has been demonstrated in a variety of sediments, soils, and groundwater. To date, studies evaluating U bio-oxidation and mobilization have primarily focused on anthropogenically U contaminated sites. In the Platte River Basin U originating from weathering of uranium-rich igneous rocks in the Rocky Mountains was deposited in shallow alluvial sediments as insoluble reduced uranium minerals. These reduced U minerals are subject to reoxidation by available oxidants, such nitrate, in situ. Soluble uranium (U) from natural sources is a recognized contaminant in public water supplies throughout the state of Nebraska and Colorado. Here we evaluate the potential of anaerobic, nitrate-dependent microbially catalyzed metal/radionuclide oxidation in subsurface sediments near Alda, NE. Subsurface sediments and groundwater (20-64ft.) were collected from a shallow aquifer containing nitrate (from fertilizer) and natural iron and uranium. The reduction potential revealed a reduced environment and was confirmed by the presence of Fe(II) and U(IV) in sediments. Although sediments were reduced, nitrate persisted in the groundwater. Nitrate concentrations decreased, 38 mg/L to 30 mg/L, with increasing concentrations of Fe(II) and U(IV). Dissolved U, primarily as U(VI), increased with depth, 30.3 μg/L to 302 μg/L. Analysis of sequentially extracted U(VI) and U(IV) revealed that virtually all U in sediments existed as U(IV). The presence of U(IV) is consistent with reduced Fe (Fe(II)) and low reduction potential. The increase in aqueous U concentrations with depth suggests active U cycling may occur at this site. Tetravalent U (U(IV)) phases are stable in reduced environments, however the input of an oxidant such as oxygen or nitrate into these systems would result in oxidation. Thus co-occurrence of nitrate suggests that nitrate could be used by bacteria as a U(IV) oxidant. Most probable number

  2. Processing of irradiated, enriched uranium fuels at the Savannah River Plant

    Energy Technology Data Exchange (ETDEWEB)

    Hyder, M L; Perkins, W C; Thompson, M C; Burney, G A; Russell, E R; Holcomb, H P; Landon, L F

    1979-04-01

    Uranium fuels containing /sup 235/U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction with dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of /sup 238/Pu is high enough to make its recovery desirable. Most of the /sup 238/Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, /sup 239/Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse.

  3. Processing of irradiated, enriched uranium fuels at the Savannah River Plant

    International Nuclear Information System (INIS)

    Hyder, M.L.; Perkins, W.C.; Thompson, M.C.; Burney, G.A.; Russell, E.R.; Holcomb, H.P.; Landon, L.F.

    1979-04-01

    Uranium fuels containing 235 U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction with dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of 238 Pu is high enough to make its recovery desirable. Most of the 238 Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, 239 Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse

  4. Neptunium separation in trace levels from uranium solutions by extraction chromatography

    International Nuclear Information System (INIS)

    Cotrim, M.B.; Matsuda, H.T.

    1994-01-01

    Neptunium and uranium behavior in extraction chromatograph system, aiming the separation of micro quantities of neptunium from uranyl solutions is described. Tri-n-octylamine (TOA), Tri-n-butylphosphate (TBP) as stationary phase, alumine, Voltalef UF-300 as support material were verified. The impregnation conditions as well as the best stationary phase/support material ratio were established. TBP/alumine, TBP/Voltalef and TOA/alumine system were selected to uranium and neptunium separation studies. (author) . 12 refs., 03 tabs., 03 figs

  5. Solubility of some phenolic compounds in aqueous alkali metal nitrate solutions from (293.15 to 318.15) K

    Energy Technology Data Exchange (ETDEWEB)

    Noubigh, Adel [Laboratoire de Physico-chimie des materiaux, IPEST, BP51, 2070 La MARSA (Tunisia)], E-mail: Adel.anoubigh@ipest.rnu.tn; Cherif, Mourad [IPEIEM, Universite de Tunis-El Manar, BP244. 2096. El Manar II (Tunisia); Provost, Elise [Laboratoire Chimie et procedes, ENSTA, 32 Rue de Boulevard Victor, 75739 Paris, Cedex 15 (France); Abderrabba, Manef [Laboratoire de Physico-chimie des materiaux, IPEST, BP51, 2070 La MARSA (Tunisia)

    2008-11-15

    This paper is continuation of the study concerning the solubility-temperature dependence data for some phenolic compounds (PhC), contained in olive mill wastewater (OMWW), in two nitrate salts (KNO{sub 3} and NaNO{sub 3}) aqueous solutions. The solubilities of PhC were determined in the temperature ranging from (293.15 to 318.15) K. It has been observed that the solubility, in aqueous nitrate solutions, increases with increasing temperature. Results showed that alkali metal nitrate has a salting-out effect on the solubility of PhC. The effect of the anion of the electrolyte on the solubility of PhC is observed by comparing these results with values reported in the previous papers for the effect of LiCl, NaCl and KCl. For each cation, the solubilites of the phenolic compounds are higher with nitrate anion than with chloride anion. Results were interpreted in terms of the salt hydration shells and the ability of the solute to form hydrogen-bond with water. The solubility data were accurately correlated by a semi empirical equation. The standard molar Gibbs free energies of transfer of PhC ({delta}{sub tr}G{sup 0}) from pure water to aqueous solutions of the nitrate salts have been calculated from the solubility data. The decrease in solubility is correlated to the positive {delta}{sub tr}G{sup 0} value which is mainly of enthalpic origin.

  6. Potentiality of uranium biosorption from nitric acid solutions using shrimp shells

    International Nuclear Information System (INIS)

    Ahmed, S.H.; El Sheikh, E.M.; Morsy, A.M.A.

    2014-01-01

    Biosorption has gained important credibility during recent years because of its good performance and low cost. This work is concerned with studying the potentiality of the chitin component of the shrimp shells for uranium biosorption from nitric acid liquid solutions. The structural characteristics of the working chitin have been determined via Fourier Transform Infrared Spectroscopy (FTIR). The surface morphology was examined using Scanning Electron Microscopy (SEM). The adsorption capacity of biomass was investigated experimentally. The influence of contact time, pH, metal ion concentration, solution volume to mass ratio and temperature were evaluated and the results were fitted using adsorption isotherm models. The kinetic of uranium biosorption was also investigated as well as biosorption thermodynamic. - Graphical abstract: Physicochemical process of biosorption is known to be promising technique due to the ease of operation and comparable low cost of biosorbant application. Chitin flakes extracted from shrimp shells show potentiality in uranium adsorption reached 7.48 mg uranium at the following conditions: 60 min contact time, pH 3.66, 50:1 V/m ration and at room temperature. The theoretical sorption capacity was 25.31 mg g −1 , according to Langmuir isotherm model. The rate of sorption follows pseudo second-order. The nature of biosorption process is spontaneous and exothermic. - Highlights: • This study shows the potentially for shrimp shell beads for uranium adsorption. • The max. biosorption was achieved at pH 3.6, contact time 60 min, S/L ratio 1:50. • Uranium sorption follows Langmuir isotherm with theoretical capacity of 25.31 g/kg. • The nature of sorption process of the sorbents is spontaneous and exothermic. • The rate of sorption follows pseudo second-order

  7. Criticality parameters for uranyl nitrate or plutonium nitrate systems in tributyl phosphate/kerosine and water

    International Nuclear Information System (INIS)

    Weber, W.

    1985-01-01

    This report presents the calculated values of smallest critical masses and volumina and neutron physical parameters for uranyl nitrate (3, 4, 5% U-235) or plutonium nitrate (5% Pu-240), each in a 30 per cent solution of tributyl phosphate (TBP)/kerosine. For the corresponding nitrate-water solutions, newly calculated results are presented together with a revised solution density model. A comparison of the data shows to what extent the criticality of nitrate-TBP/kerosine systems can be assessed on the basis of nitrate-water parameters, revealing that such data can be applied to uranyl nitrate/water systems, taking into account that the smallest critical mass of uranyl nitrate-TBP/kerosine systems, up to a 5 p.c. U-235 enrichment, is by 4.5 p.c. at the most smaller than that of UNH-water solutions. Plutonium nitrate (5% Pu-240) in the TBP/kerosine solution will have a smallest critical mass of up to 7 p.c. smaller, as compared with the water data. The suitability of the computing methods and cross-sections used is verified by recalculating experiments carried out to determine the lowest critical enrichment of uranyl nitrate. The calculated results are well in agreement with experimental data. The lowest critical enrichment is calculated to be 2.10 p.c. in the isotope U-235. (orig.) [de

  8. Determination of trace amounts of uranium in a reprocessing plant by solution spectrofluorimetry

    International Nuclear Information System (INIS)

    Mauchien, P.; Cauchetier, Ph.

    1983-01-01

    To establish inventory tables accurately and satisfy safeguards requirements - and also to ensure satisfactory operation of a reprocessing plant - it is essential to determine the uranium content of numerous solutions where the uranium is present only in trace quantities. For this purpose a method is proposed which relies on the fluorescence of uranyl solutions exposed to ultra-violet radiation. After a brief theoretical summary, the parameters which influence the measurements most strongly are enumerated: medium, temperature, nature of the matrix, and choice of wavelength of the incident radiation. It is then apparent that the measurement must be performed by internal calibration (using the proportional addition method) and that it is useful to obtain a fluorescence spectrum which enables us to verify the presence of uranium. The applications of this method at the La Hague plant are described, where it has been used since October 1981 by shift teams, notably to check the following points: the attack acid (before receiving the fuel), the foot of the first-cycle column, the carbonated solvent washing solution and the solutions of the effluent treatment unit. It is in fact used throughout the plant, even for checking uranium in PuO 2 oxide. The method makes it possible to avoid organic reagents such as pyridine and, in many cases, cumbersome effluent-generating separations. Determinations are possible - to give one example - by simple dilution in the fission-product concentrates and in plutonium solutions where the Pu/U ratio is as great as 1000/1. In pure solutions the detection limit with the equipment used at present is a few μg per litre in the measuring tank. In general, the accuracy is a few per cent. (author)

  9. Potentiality of uranium biosorption from nitric acid solutions using shrimp shells.

    Science.gov (United States)

    Ahmed, S H; El Sheikh, E M; Morsy, A M A

    2014-08-01

    Biosorption has gained important credibility during recent years because of its good performance and low cost. This work is concerned with studying the potentiality of the chitin component of the shrimp shells for uranium biosorption from nitric acid liquid solutions. The structural characteristics of the working chitin have been determined via Fourier Transform Infrared Spectroscopy (FTIR). The surface morphology was examined using Scanning Electron Microscopy (SEM). The adsorption capacity of biomass was investigated experimentally. The influence of contact time, pH, metal ion concentration, solution volume to mass ratio and temperature were evaluated and the results were fitted using adsorption isotherm models. The kinetic of uranium biosorption was also investigated as well as biosorption thermodynamic. Copyright © 2014 Elsevier Ltd. All rights reserved.

  10. A method for the gravimetric determination of plutonium in pure plutonium nitrate concentrate solution

    International Nuclear Information System (INIS)

    Mair, M.A.; Savage, D.J.

    1986-12-01

    Plutonium nitrate solution is treated with sulphuric acid before being heated and finally ignited. The stoichiometric plutonium dioxide so formed is weighed and hence the plutonium content is calculated. (author)

  11. Extraction of uranium from aqueous solution by phosphonic acid-imbedded polyurethane foam

    International Nuclear Information System (INIS)

    Katragadda, S.; Gesser, H.D.; Chow, A.

    1997-01-01

    Phenylphosphonic acid was imbedded into the matrix of the polyurethane foam during the fabrication process of the polymer. The extraction of uranium by phosphonic acid-imbedded polyurethane foam and blank polyurethane (i.e., foam without phosphonic acid functional groups) was investigated. Phosphonic acid-imbedded foam showed superior extractability of uranium from solutions with pH = 7.0 ± 1.5 over a wide range of temperatures. (author)

  12. Root uptake of uranium (6) in solution by a higher plant: speciation in hydroponic solution, bioavailability, micro-localisation and biological effects induced; Transfert racinaire de l'uranium (6) en solution chez une plante superieure: speciation en solution hydroponique, prise en charge par la plante, microlocalisation et effets biologiques induits

    Energy Technology Data Exchange (ETDEWEB)

    Laroche, L

    2005-01-15

    Uranium exists naturally in the environment, usually present in trace quantities. In soil solution and oxic conditions, uranium is present in the +VI oxidation state and forms a large number of inorganic and organic complexes. The exposure medium, an artificial soil solution, was designed in such a way as to control the uranium species in solution. The geochemical speciation code JCHESS was used to calculate the uranium aqueous species concentration and to define the domains of interest, each of them characterized by a limited number of dominant U species. These domains were defined as follows: pH 4.9 with uranyl ions as dominant species, pH 5.8 with hydroxyl complexes and pH 7 where carbonates play a major role. For each pH, short-duration (5 hours of exposure) well-defined laboratory experiments were carried out with Phaseolus vulgaris as plant model. The effect of competitive ions such as Ca{sup 2+} or the presence of ligands such as phosphate or citrate on root assimilation efficiency was explored. Results have shown that uranium transfer was not affected by the presence of calcium, phosphate or citrate (but was decreased of 60% with citrate (10 {mu}M) at pH 5.8) in our experimental conditions. Moreover, observation in Transmission Electronic Microscopy (TEM), equipped with an EDAX probe, have shown that uranium was associated with granules rich in phosphorus and that there were some chloroplast anomalies. Finally, the presence of uranium affects root CEC by reducing it and stimulates root elongation at low uranium concentrations (100 nM, 400 nM and 2 {mu}M at pHs 4.9, 5.8 and 7 respectively) and inhibits it at high uranium concentrations. (author)

  13. Ammonia volatilization from surface-applied nitrogen solution of urea and ammonium nitrate

    International Nuclear Information System (INIS)

    Trivellin, Paulo Cezar Ocheuze; Stefanutti, Ronaldo; Lima Filho, Oscar Fontvo de; Tziboy, Edgar Alfredo Tzi; Oliveira Junior, Jovo Alberto de; Bendassolli, Jose Albertino

    1996-08-01

    The urea is one of the fertilizers more utilized in modern agriculture. One of the problems in the urea utilization is the ammonium volatilization, resulting in low utilization of N-fertilizers by the plants.The objective of this study it was to evaluate and to compare in laboratories conditions , utilizing the 15 N technic the soil's ammonium lost by volatilization associated a superficial application of nitrogen corresponding doses like urea solution and urea and ammonium nitrates solution

  14. A new and simple gravimetric method for determination of uranium

    International Nuclear Information System (INIS)

    Saxena, A.K.

    1994-01-01

    A new and simple gravimetric method for determining uranium has been described. Using a known quantity of uranyl nitrate as the test solution, an alcoholic solution of 2-amino-2-methyl 1:3 propanediol (AMP) was added slowly. A yellow precipitate was obtained which was filtered through ashless filter paper, washed with alcohol, dried and ignited at 800 degC for 4h. It gave a black powder as a product which was shown by X-ray diffraction to be U 3 O 8 . The percentage error was found in the range -0.09 to +0.89. (author). 8 refs., 1 tab

  15. Developments in uranium solution mining in Australia

    International Nuclear Information System (INIS)

    Hunter, T.

    2001-01-01

    The last five years have seen rapid developments in uranium solution mining in Australia, with one deposit brought into production (Beverley, 1,000 tpa U 3 O 8 ) and another close to receiving development approval (Honeymoon, 500 expanding to 1,000 tpa U 3 O 8 proposed). The deposits were discovered during extensive exploration of the Frome Basin in South Australia in the early 1970s and were mothballed from 1983 to 1996 due to Government policies. Uranium mineralisation at Beverley, Honeymoon and other related prospects is hosted in unconsolidated coarse grained quartz sands which are sealed in buried palaeovalleys. Both projects have successfully trialled acid leaching methods and have confirmed high permeability and confinement of the target sands. At Beverley an ion exchange process has been adopted, whereas at Honeymoon solvent extraction has been trialled and is proposed for future production Australian production economics compare favourably with US counterparts and are likely to be within the lower quartile of world costs

  16. Determination of the uranium concentration in soil solutions by the fission track registration technique

    International Nuclear Information System (INIS)

    Fernandes, G.P.

    1980-02-01

    The fission tracks registration technique was used to determine the uranium concentration in soil solutions. The Makrofol KG, a synthetic plastic manufactured by Bayer, was used as a detector and the wet method was applied. From the calibration curves obtained, it was possible to determine uranium concentrations in soil solutions, from 90 to 320 μg U/l, with an error between 9.4% and 4.0%, respectively. The method was applied to a few soil samples from Pocos de Caldas, Minas Gerais in Brazil. The uranium concentrations in the sample and residues were also determined by other methods to compare the results obtained; only one sample showed deviation from the results obtained by the fission tracks method. And this discrepancy was explained in a reasonable way. It was shown that the fission tracks technique can be used with sucess for application in soil solutions. (Author) [pt

  17. Surface-enhanced Raman spectroscopy for uranium detection and analysis in environmental samples

    International Nuclear Information System (INIS)

    Ruan Chuanmin; Luo Wensui; Wang Wei; Gu Baohua

    2007-01-01

    Techniques for rapid screening of uranium in environmental samples are needed, and this study entails the development of surface-enhanced Raman scattering (SERS) for analyzing uranium in aqueous media with improved sensitivity and reproducibility. A new SERS substrate based on (aminomethyl)phosphonic acid (APA)-modified gold nanoparticles was found to give greater than three orders of magnitude SERS enhancement compared with unmodified bare gold nanoparticles. Intensities of uranyl band at about 830 cm -1 were proportional to the concentrations of uranium in solution, especially at relatively low concentrations ( -5 M). A detection limit of ∼8 x 10 -7 M was achieved with a good reproducibility since the measurement was performed directly in dispersed aqueous suspension. Without pretreatment, the technique was successfully employed for detecting uranium in a highly contaminated groundwater with a low pH, high dissolved salts (e.g., nitrate, sulfate, calcium and aluminum) and total organic carbon

  18. 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

  19. Influence of microstructure on stress corrosion cracking of mild steel in synthetic caustic-nitrate nuclear waste solution

    International Nuclear Information System (INIS)

    Sarafian, P.G.

    1975-12-01

    The influence of alloy microstructure on stress corrosion cracking of mild steel in caustic-nitrate synthetic nuclear waste solutions was studied. An evaluation was made of the effect of heat treatment on a representative material (ASTM A 516 Grade 70) used in the construction of high activity radioactive waste storage tanks at Savannah River Plant. Several different microstructures were tested for susceptibility to stress corrosion cracking. Precracked fracture specimens loaded in either constant load or constant crack opening displacement were exposed to a variety of caustic-nitrate and nitrate solutions. Results were correlated with the mechanical and corrosion properties of the microstructures. Crack velocity and crack arrest stress intensity were found to be related to the yield strength of the steel microstructures. Fractographic evidence indicated pH depletion and corrosive crack tip chemistry conditions even in highly caustic solutions. Experimental results were compatible with crack growth by a strain-assisted anodic dissolution mechanism; however, hydrogen embrittlement also was considered possible

  20. Removing ferric ions from concentrated acid leaching solution of an uranium ore by jarosite

    International Nuclear Information System (INIS)

    Song Huanbi; Hu Yezang

    1997-01-01

    The author expounds the fundamental rules of removing ferric ions by jarosite and presents results of removing ferric ions from concentrated acid curing-trickle leaching solution of an uranium ore. It turns out that the method can be applied to uranium hydrometallurgical process effectively

  1. U uranium. Suppl. Vol. D3

    International Nuclear Information System (INIS)

    Haug, H.O.

    1982-01-01

    This volume of the uranium series of the Gmelin handbook deals with the anion exchange of uranium. Compounds of the valence states of III, IV, V and VI of uranium in halide, nitrate, sulfate, phosphate, and carbonate media as well as in media containing organic complexing agents are treated. The literature cited covers the period from about 1947 to the end of 1980. (RB) [de

  2. Solidification of nitrate solutions with alkali-activated slag and slag–metakaolin cements

    International Nuclear Information System (INIS)

    Rakhimova, Nailia R.; Rakhimov, Ravil Z.; Osin, Yury N.; Naumkina, Natalia I.; Gubaidullina, Alfiya M.; Yakovlev, Grigory I.; Shaybadullina, Arina V.

    2015-01-01

    Highlights: • The effectiveness of an AASC matrix for NaNO 3 solution solidification is stated. • XRD, DTA-TG, and X-ray microtomography experiments were performed. • Crystallization of NaNO 3 reduces the shrinkage of hardened AASC-based waste forms. • Metakaolin shortens the setting time and increases the compressive strength of AASC. - Abstract: The solidification of nitrate solutions with alkali-activated slag (AASC) and slag–metakaolin cements (AASMC) and the resulting setting times, compressive strengths, dimensional stability, water resistance, hydration products, microstructures, and macroporous network structures were evaluated. The influences of the alkali activator concentration, mineral composition of metakaolin, ratio of slag to slag + metakaolin, and concentration of NaNO 3 on the cement performance were all evaluated in detail. The compressive strength of cemented nitrate solutions with AASC and AASMC aged for 28 days was from 13.4 to 42 MPa depending on the NaNO 3 concentration. X-ray diffractometer, differential thermal analyzer, and electron microscope analyses suggested that NaNO 3 crystallizes in cementitious matrices without reacting with the hydration products of AASC and AASMC. X-ray microtomography showed that the solidified NaNO 3 solution with a salt concentration of 700 g/l and AASC had a denser microstructure without shrinkage microcracks, a smaller macropore volume, and smaller macropore sizes than hardened AASC-based paste mixed with water

  3. Preliminary results from uranium/americium affinity studies under experimental conditions for cesium removal from NPP ''Kozloduy'' simulated wastes solutions

    International Nuclear Information System (INIS)

    Nikiforova, A.; Kinova, L.; Peneva, C.; Taskaeva, I.; Petrova, P.

    2005-01-01

    We use the approach described by Westinghouse Savannah River Company using ammonium molybdophosphate (AMP) to remove elevated concentrations of radioactive cesium to facilitate handling waste samples from NPP K ozloduy . Preliminary series of tests were carried out to determine the exact conditions for sufficient cesium removal from five simulated waste solutions with concentrations of compounds, whose complexing power complicates any subsequent processing. Simulated wastes solutions contain high concentrations of nitrates, borates, H 2 C 2 O 4 , ethylenediaminetetraacetate (EDTA) and Citric acid, according to the composition of the real waste from the NPP. On this basis a laboratory treatment protocol was created. This experiment is a preparation for the analysis of real waste samples. In this sense the results are preliminary. Unwanted removal of non-cesium radioactive species from simulated waste solutions was studied with gamma spectrometry with the aim to find a compromise between on the one hand the AMP effectiveness and on the other hand unwanted affinity to AMP of Uranium and Americium. Success for the treatment protocol is defined by proving minimal uptake of U and Am, while at the same time demonstrating good removal effectiveness through the use of AMP. Uptake of U and Am were determined as influenced by oxidizing agents at nitric acid concentrations, proposed by Savannah River National laboratory. It was found that AMP does not significantly remove U and Am when concentration of oxidizing agents is more than 0.1M for simulated waste solutions and for contact times inherent in laboratory treatment protocol. Uranium and Americium affinity under experimental conditions for cesium removal were evaluated from gamma spectrometric data. Results are given for the model experiment and an approach for the real waste analysis is chosen. Under our experimental conditions simulated wastes solutions showed minimal affinity to AMP when U and Am are most probably in

  4. Apparent and partial molal heat capacities of aqueous rare earth nitrate solutions at 250C

    International Nuclear Information System (INIS)

    Spedding, F.H.; Baker, J.L.; Walters, J.P.

    1979-01-01

    Specific heats of aqueous solutions of the trinitrates of La, Pr, Nd, Sm, Gd, Tb, Dy, Ho, Er, Tm, Yb, and Lu were measured from 0.1 m to saturation at 25 0 C. Apparent molal heat capacities, phi/sub cp/, were calculated for these solutions, and empirical polynomial equations were obtained which expressed phi/sub cp/ as a function of m/sup 1/2/ for each salt. The partial molal heat capacities of the solvent, anti C 1 /sub p/, and solute, anti C 2 /sub p/, were calculated from these equations. Unlike chloride and perchlorate data reported earlier, values of anti C 1 /sub p/ for nitrate solutions across the rare earth series did not show a two series effect. Instead, anti C 1 /sub p/ values at lower concentrations (0.5 and 1.0 m) appear correlated with reported first formation constants for rare earth-nitrate complexes. 31 references, 9 figures, 2 tables

  5. Fluorimetrich determination of uranium in mineral samples and phosphoric solutions

    International Nuclear Information System (INIS)

    Pupo Gonzales, I.; Cuevillas, J.; Estevez, J.

    1991-01-01

    In this paper an analytical technique for the determination of uranium in different cuban minerals and acid leaching liquors of phosphorites is proposed. The method used for solid samples includes their disolution and further dilution of the solution obtained. For liquid samples dilution is the only intermediate step used. A study of HNO3 concentration (pH) was made in fluorimetric measurements. The method was applied to the uranium analysis in two IAEA standard reference samples (phosphate base), in natural phosphorites and a synthetic one, in clays and phosphoric liquors. The results agree with those obtained by other methods

  6. Interaction of cadmium and indium nitrate mixture with sodium tungstate in aqueous solution

    Energy Technology Data Exchange (ETDEWEB)

    Belousova, E E; Krivobok, V I; Gruba, A I [Donetskij Gosudarstvennyj Univ. (Ukrainian SSR)

    1982-01-01

    The interaction of the mixture of cadmium and indium nitrates with sodium tungstate in aqueous solution is studied using the methods of ''residual concentrations'', pH potentiometry and conductometry. Independent of the ratio of components in the initial solution a mixture of coprecipitated normal tungstates of cadmium and indium is formed in the system. Heat treatment of the precipitates at 800 deg C for 50 hrs with subsequent hardening results in the formation of solid solutions on the basis of normal cadmium and indium tungstates.

  7. Processing of Sierra Albarrana uranium ores

    International Nuclear Information System (INIS)

    Gutierrez Jodra, L.; Perez Luina, A.; Perarnau, M.

    1960-01-01

    Uranium recovery by hydrometallurgy from brannerite, found in Hornachuelos (Cordoba) is described. It has been studied the acid and alkaline leaching and salt roasting, proving as more satisfactory the acid leaching. Besides the uranium solubilization by acid leaching, is described the further process to obtain pure uranyl nitrate. (Author)

  8. Literature review and preliminary analysis of inorganic ammonia pertinent to south Texas uranium in-situ leach

    International Nuclear Information System (INIS)

    Braswell, J.; Breland, M.; Chang, M.; Farley, J.; Hill, D.; Johnson, D.

    1978-01-01

    The purpose of this report is to review existing literature to aid in the determination of the potential impact of ammonia-containing lixiviants on uranium solution mining aquifers, perform studies based on the available literature, to identify potential ways to protect the groundwaters from ammonia contamination, and to propose further work where data are lacking or needed. The review of the literature includes an analysis and interpretation of the literature as it relates to the solution mining activities. Results focus on the range of geologic and hydrologic conditions representative of South Texas solution minig areas. Other pertinent data sources such as soils and agricultural literature are also reviewed and conclusions extrapolated to the solution mining situation. Specific tasks were: evaluate the potential of natural occurrence and influx of ammonia and/or nitrate species in confined aquifers typical of uranium solution mining sites; find available data on the sorption characteristics of ammonia and nitrates on pure and mixed minerals representative of South Texas geology in solution mining areas; determine applicable selectivity coefficients and kinetic data on sorption and desorption of ammonia on clay minerals; evaluate the potential for natural inorganic ammonia conversion by chemical or other mechanisms in typical solution mining aquifers; review available monitoring data from solution mining operations as it pertains to ammonia adsorption or migration; analyze and provide calculational bases for determining the predicted fate of ammonia under solution mining conditions; recommend continuation programs that focus on areas of uncertainty; provide comprehensive bibliography and abstracts of all pertinent articles

  9. Interpretation of criticality experiments on homogeneous solutions of plutonium and uranium; Interpretation des experiences de criticite sur des solutions homogenes de plutonium et d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Ithurralde, M F; Kremser, J; Leclerc, J; Lombard, Ch; Moreau, J; Robin, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    Criticality experiments on solutions of fissionable materials have been carried out in tanks of various geometries (cylinder, isolated annular cylinder, interacting annular cylinders); the reflexion conditions have also been varied (without reflection, semi-reflection and total reflexion by water). The range of the studied concentrations is rather large (18,8 to 104 gms/liter). The interpretation of these experiments has been undertaken in order to resolve the problems of the industrial use of homogeneous plutonium and uranium solutions. Several methods the fields of application of which are different have been used: diffusion method, transport method and Monte-Carlo method. (authors) [French] Des experiences critiques sur des solutions de matieres fissiles ont ete faites dans des cuves de diverses geometries (cylindre, cylindre annulaire isole, cylindre annulaire en interaction), les conditions de reflexion ont ete egalement variees (sans reflexion, semi reflexion et reflexion totale par l'eau). La gamme des concentrations etudiees est assez etendue (18,8 a 104 g/l ). L'interpretation de ces experiences a ete entreprise dans le but de pouvoir resoudre les problemes poses par l'emploi industriel de solutions homogenes de plutonium et d'uranium, plusieurs methodes dont les domaines d'application sont differents ont ete employees: methode de diffusion, methode de transport, methode de Monte-Carlo. (auteurs)

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

    Science.gov (United States)

    Novoselov, Ivan Yu.; Karengin, Alexander G.; Babaev, Renat G.

    2018-03-01

    The aim of this paper is to carry out thermodynamic simulation of mixed plutonium and uranium oxides compounds obtained after plasma treatment of plutonium and uranium nitrates and to determine optimal water-salt-organic mixture composition as well as conditions for their plasma treatment (temperature, air mass fraction). Authors conclude that it needs to complete the treatment of nitric solutions in form of water-salt-organic mixtures to guarantee energy saving obtainment of oxide compounds for mixed-oxide fuel and explain the choice of chemical composition of water-salt-organic mixture. It has been confirmed that temperature of 1200 °C is optimal to practice the process. Authors have demonstrated that condensed products after plasma treatment of water-salt-organic mixture contains targeted products (uranium and plutonium oxides) and gaseous products are environmental friendly. In conclusion basic operational modes for practicing the process are showed.

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

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

  12. Extraction of uranium with emulsion membrane process use tributylphosphate extractant

    International Nuclear Information System (INIS)

    Basuki, K.T.; Sudibyo, R.; Bambang EHB; Muhadi, A.W.

    1996-01-01

    To increase the effectiveness of extraction process with so for to occur, it was tried the extraction with a couple of extraction and stripping process. This couple process was called liquid membrane emulsion. As membrane was used mix surfactant (Span-80), tributylphosphate in kerosene, natrium carbonate, while as a feeder was uranium solution with 500 concentration ppm in 0.5 - 3 M nitrate acid. In this experiment the variable investigated were % surfactant (1 - 5 %), rotary speed for membrane making (2,500 - 10.000 rpm). The optimal condition result of experiment were 5 % surfactant, 3 M nitrate acid, rotary speed 10.000 rpm and (Kd eksU ) 57 %, and (Kd strippU ) 87 %, Kd eksU at liquid-liquid extraction is 44 %. (author)

  13. A thermal analysis - mass spectrometric study of the reactions in the binary nitrate system UO2(NO3)2.6H2O-Ba(NO3)2

    International Nuclear Information System (INIS)

    Kalekar, B.B.; Rajagopalan, K.V.; Ravindran, P.V.

    2008-01-01

    Solid state reaction between uranyl nitrate hexahydrate and barium nitrate at elevated temperatures has been studied. Barium nitrate does not react directly with uranyl nitrate hexahydrate (UNH). The decomposition product of the latter, UO 3 , however, reacts with molten barium nitrate to form nitrato complexes which decompose to form barium uranates. The decomposition of the nitrato complex is often attended by a change of oxidation state of uranium. Uranyl-nitrate-rich compositions form uranates with uranium in oxidation state V, whereas barium-nitrate-rich compositions form uranates with uranium in oxidation state VI. The tendency for hydrolysis of uranyl nitrate hexahydrate at elevated temperatures is lowered in presence of barium nitrate and, unlike in the case of dehydration of pure UNH, uranyl nitrate dihydrate is formed as dehydration intermediate. (author)

  14. Predictive geochemical modeling of uranium and other contaminants in laboratory columns in relatively oxidizing, carbonate-rich solutions

    International Nuclear Information System (INIS)

    Longmire, P.; Turney, W.R.; Mason, C.F.V.

    1994-01-01

    Carbonate heap leaching of uranium-contaminated soils and sediments represents a viable, cost-effective remediation technology. Column experiments have been conducted using 0.1, 0.25, and 0.5 M Na 2 CO 3 /NaHCO 3 solutions for leaching uranium from soils located adjacent to an incinerator at the Fernald Environmental Management Project (FEMP) site. Results from column experiments and geochemical modeling are used to quantitatively evaluate the effectiveness of heap leaching. Leach efficiencies of up to 72 wt.% of total uranium in CaO-agglomerated soil result from dissolution of uranium (U(VI)-dominated) minerals, formation of the soluble complex UO 2 (CO 3 ) 3 4- , and uranium desorption from clay minerals, ferric hydroxides, and humic acids. Parameters that control the extent of uranium extraction include pH, Eh, temperature, carbonate concentration, lixiviant-flow rate, pore-solution chemistry, solid phases, and soil texture

  15. Thermochemical nitrate reduction

    International Nuclear Information System (INIS)

    Cox, J.L.; Lilga, M.A.; Hallen, R.T.

    1992-09-01

    A series of preliminary experiments was conducted directed at thermochemically converting nitrate to nitrogen and water. Nitrates are a major constituent of the waste stored in the underground tanks on the Hanford Site, and the characteristics and effects of nitrate compounds on stabilization techniques must be considered before permanent disposal operations begin. For the thermochemical reduction experiments, six reducing agents (ammonia, formate, urea, glucose, methane, and hydrogen) were mixed separately with ∼3 wt% NO 3 - solutions in a buffered aqueous solution at high pH (13); ammonia and formate were also mixed at low pH (4). Reactions were conducted in an aqueous solution in a batch reactor at temperatures of 200 degrees C to 350 degrees C and pressures of 600 to 2800 psig. Both gas and liquid samples were analyzed. The specific components analyzed were nitrate, nitrite, nitrous oxide, nitrogen, and ammonia. Results of experimental runs showed the following order of nitrate reduction of the six reducing agents in basic solution: formate > glucose > urea > hydrogen > ammonia ∼ methane. Airnmonia was more effective under acidic conditions than basic conditions. Formate was also effective under acidic conditions. A more thorough, fundamental study appears warranted to provide additional data on the mechanism of nitrate reduction. Furthermore, an expanded data base and engineering feasibility study could be used to evaluate conversion conditions for promising reducing agents in more detail and identify new reducing agents with improved performance characteristics

  16. Study of accurate volume measurement system for plutonium nitrate solution

    Energy Technology Data Exchange (ETDEWEB)

    Hosoma, T. [Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan). Tokai Works

    1998-12-01

    It is important for effective safeguarding of nuclear materials to establish a technique for accurate volume measurement of plutonium nitrate solution in accountancy tank. The volume of the solution can be estimated by two differential pressures between three dip-tubes, in which the air is purged by an compressor. One of the differential pressure corresponds to the density of the solution, and another corresponds to the surface level of the solution in the tank. The measurement of the differential pressure contains many uncertain errors, such as precision of pressure transducer, fluctuation of back-pressure, generation of bubbles at the front of the dip-tubes, non-uniformity of temperature and density of the solution, pressure drop in the dip-tube, and so on. The various excess pressures at the volume measurement are discussed and corrected by a reasonable method. High precision-differential pressure measurement system is developed with a quartz oscillation type transducer which converts a differential pressure to a digital signal. The developed system is used for inspection by the government and IAEA. (M. Suetake)

  17. Recovery of uranium from alkaline ore (Tummalapalle) leach solution using novel precipitating method

    International Nuclear Information System (INIS)

    Biswas, Sujoy; Rupawate, V.H.; Hareendran, K.N.; Roy, S.B.; Chakravartty, J.K.

    2014-01-01

    The aim of present study is recovery of uranium from such ore leach solution containing 2 O 7 at pH ∼12.5. The average particle size of the MgU 2 O 7 particles was 20 micron and overall uranium recovery was 97%. The composition of final precipitate was characterized using XRD and surface morphology was studied using SEM

  18. Translocation of uranium from water to foodstuff while cooking

    International Nuclear Information System (INIS)

    Krishnapriya, K.C.; Baksi, Ananya; Chaudhari, Swathi; Gupta, Soujit Sen; Pradeep, T.

    2015-01-01

    Highlights: • Rice can efficiently uptake uranium from water contaminated with uranyl nitrate hexahydrate (UO 2 (NO 3 ) 2.6 H 2 O), while cooking. • Unusual uranium uptake to the extent of about 1000 ppm is observed when rice is cooked in highly concentrated uranium contaminated water (1240 ppm). • Nature of interaction of uranium with carbohydrates is probed using small monosaccharides like glucose and mannose. • Electrospray ionization mass spectrometry showed UO 2 2+ to be the most stable species in water in such solutions which can form complexes with sugars. • The species (UO 2 2+ ) is also observed in the case of water exposed to the common mineral, uranium oxide (UO 2 ) and similar type of complexation is observed with sugars. - Abstract: The present work report the unusual uranium uptake by foodstuff, especially those rich in carbohydrates like rice when they are cooked in water, contaminated with uranium. The major staple diet in South Asia, rice, was chosen to study its interaction with UO 2 2+ , the active uranium species in water, using inductively coupled plasma mass spectrometry. Highest uptake limit was checked by cooking rice at very high uranium concentration and it was found to be good scavenger of uranium. To gain insight into the mechanism of uptake, direct interaction of UO 2 2+ with monosaccharides was also studied, using electrospray ionization mass spectrometry taking mannose as a model. The studies have been done with dissolved uranium salt, uranyl nitrate hexahydrate (UO 2 (NO 3 ) 2 ·6H 2 O), as well as the leachate of a stable oxide of uranium, UO 2 (s), both of which exist as UO 2 2+ in water. Among the eight different rice varieties investigated, Karnataka Ponni showed the maximum uranium uptake whereas unpolished Basmati rice showed the minimum. Interaction with other foodstuffs (potato, carrot, peas, kidney beans and lentils) with and without NaCl affected the extent of chemical interaction but was not consistent with the

  19. Solid phase extraction of uranium and thorium on octadecyl bonded silica modified with Cyanex 302 from aqueous solutions

    International Nuclear Information System (INIS)

    Nilchi, A.; Shariati Dehaghan, T.; Rasouli Garmarodi, S.

    2013-01-01

    A simple and reliable method for rapid extraction and determination of uranium and thorium using octadecyl-bonded silica modified with Cyanex 302 is presented. Extraction efficiency and the influence of various parameters such as aqueous phase pH, flow rate of sample solution and amount of extractant has been investigated. The study showed that the extraction of uranium and thorium increase with increasing pH value and was found to be quantitative at pH 6; and the retention of ions was not affected significantly by the flow rate of sample solution. The extraction percent were found to be 89.55 and 86.27 % for uranium and thorium, respectively. The maximal capacity of the cartridges modified by 30 mg of Cyanex 302 was found to be 20 mg of uranium and thorium. The method was successfully applied to the extraction and determination of uranium and thorium in aqueous solutions. The percentage recovery of uranium and thorium in a number of natural as well as seawater samples of Iran were also investigated and found to be in the range of 85-95%. (author)

  20. 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.

  1. Systems of cerium(3) nitrate-dimethyl amine nitrate-water and cerium(3) nitrate-dimethyl amine nitrate-water

    International Nuclear Information System (INIS)

    Mininkov, N.E.; Zhuravlev, E.F.

    1976-01-01

    Solubility of solid phases in the systems cerium(3)nitrate-water-dimethyl amine nitrate and cerium(3)nitrate-water-dimethyl amine nitrate has been st ed by the method of isothermal sections at 25 and 50 deo. C. It has been shown that one anhydrous compound is formed in each system with a ratio of cerium(3) nitrate to amine nitrate 1:5. The compounds formed in the systems have been separated from the corresponding solutions and studied by microcrystalloscopic, X-ray phase, thermal and infrared spectroscopic methods. On the basis of spectroscopic studies the following formula has been assigned to the compound: [(CH 3 ) 2 NH 2 + ] 5 x[Ce(NO 3 ) 8 ]. The thermal analysis of the compound has shown that its melting point is 106 deg C. The solubility isotherms in the system Ce(NO 3 ) 3 -H 2 O-(C 2 H 5 ) 2 NHxHNO 3 consist of three branches which intersect in two eutonic points

  2. Laboratory scale vitrification of low-level radioactive nitrate salts and soils from the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Shaw, P.; Anderson, B.

    1993-07-01

    INEL has radiologically contaminated nitrate salt and soil waste stored above and below ground in Pad A and the Acid Pit at the Radioactive Waste Management Complex. Pad A contain uranium and transuranic contaminated potassium and sodium nitrate salts generated from dewatered waste solutions at the Rocky Flats Plant. The Acid Pit was used to dispose of liquids containing waste mineral acids, uranium, nitrate, chlorinated solvents, and some mercury. Ex situ vitrification is a high temperature destruction of nitrates and organics and immobilizes hazardous and radioactive metals. Laboratory scale melting of actual radionuclides containing INEL Pad A nitrate salts and Acid Pit soils was performed. The salt/soil/additive ratios were varied to determine the range of glass compositions (resulted from melting different wastes); maximize mass and volume reduction, durability, and immobilization of hazardous and radioactive metals; and minimize viscosity and offgas generation for wastes prevalent at INEL and other DOE sites. Some mixtures were spiked with additional hazardous and radioactive metals. Representative glasses were leach tested and showed none. Samples spiked with transuranic showed low nuclide leaching. Wasteforms were two to three times bulk densities of the salt and soil. Thermally co-processing soils and salts is an effective remediation method for destroying nitrate salts while stabilizing the radiological and hazardous metals they contain. The measured durability of these low-level waste glasses approached those of high-level waste glasses. Lab scale vitrification of actual INEL contaminated salts and soils was performed at General Atomics Laboratory as part of the INEL Waste Technology Development and Environmental Restoration within the Buried Waste Integrated Demonstration Program

  3. Surface-Enhanced Raman Spectroscopy for Uranium Detection and Analysis in Environmental Samples

    International Nuclear Information System (INIS)

    Ruan, Chuanmin; Luo, Wensui; Wang, Wei; Gu, Baohua

    2007-01-01

    Techniques for rapid screening of uranium in environmental samples are needed, and this study entails the development of surface-enhanced Raman scattering (SERS) for analyzing uranium in aqueous media with improved sensitivity and reproducibility. A new SERS substrate based on (aminomethyl)phosphonic acid (APA)-modified gold nanoparticles was found to give greater than three orders of magnitude SERS enhancement compared with unmodified bare gold nanoparticles. Intensities of uranyl band at about 830 cm-1 were proportional to the concentrations of uranium in solution, especially at relatively low concentrations (<10-5 M). A detection limit of ∼8 e10-7 M was achieved with a good reproducibility since the measurement was performed directly in dispersed aqueous suspension. Without pretreatment, the technique was successfully employed for the detection of uranium in a highly contaminated groundwater with a low pH, high dissolved salts (e.g., nitrate, sulfate, calcium and aluminum) and total organic carbon

  4. 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

  5. Simulation of transportation of low enriched uranium solutions

    International Nuclear Information System (INIS)

    Hope, E.P.; Ades, M.J.

    1996-01-01

    A simulation of the transportation by truck of low enriched uranium solutions has been completed for NEPA purposes at the Savannah River Site. The analysis involves three distinct source terms, and establishes the radiological risks of shipment to three possible destinations. Additionally, loading accidents were analyzed to determine the radiological consequences of mishaps during handling and delivery. Source terms were developed from laboratory measurements of chemical samples from low enriched uranium feed materials being stored at SRS facilities, and from manufacturer data on transport containers. The transportation simulations were accomplished over the INTERNET using the DOE TRANSNET system at Sandia National Laboratory. The HIGHWAY 3.3 code was used to analyze routing scenarios, and the RADTRAN 4 code was used to analyze incident free and accident risks of transporting radiological materials. Loading accidents were assessed using the Savannah River Site AXAIR89Q and RELEASE 2 codes

  6. Simultaneous determination of uranium and plutonium in dissolver solution of irradiated fuel, using ID-TIMS. IRP-11

    International Nuclear Information System (INIS)

    Shah, Raju; Sasi Bhushan, K.; Govindan, R.; Alamelu, D.; Khodade, P.S.; Aggarwal, S.K.

    2007-01-01

    A simple sample preparation and simultaneous analysis method to determine uranium and plutonium from dissolver solution, employing the technique of Isotope Dilution Mass spectrometry has been demonstrated. The method used, co-elusion of Uranium and Plutonium from anion exchanger column after initial elution of major part of uranium in 1:5 HNO 3 in order to reduce the initial U/Pu ratio from 1000 to about 100-200 in the co-eluted fraction. Due to the availability of variable multi-collector system, different Faraday cups were adjusted to collect the different ion intensities corresponding to the different masses, during the simultaneous analysis of Uranium and Plutonium, loaded on Re double filament assembly. 233 U and PR grade Plutonium were used as spikes to determine Uranium and Plutonium from dissolver solution of irradiated fuel from research reactor. The possibility of getting the isotopic composition of uranium from the simultaneous analysis of co-eluted purified fraction of U and Pu from spiked aliquots is also explained. (author)

  7. Enriched uranium processing with 7-1/2% TBP

    International Nuclear Information System (INIS)

    Orth, D.A.; Martin, W.H.; Pickett, C.E.

    1983-01-01

    The 7-1/2% TBP flowsheet gives adequate recovery of uranium and neptunium or plutonium, with reduced waste volume as compared to the prior aluminum-salted 3-1/2% TBP flowsheet. Decontamination from fission products is sensitive to numerous variables, including aluminum nitrate concentration in the feed, impeller speeds, and prior treatment of the fuel solution in head end operations. The impeller speed in the 1A bank also influences uranium losses as well as the fission product decontamination. The magnitudes of these effects suggest that stage efficiency is poor with this flowsheet in this mixer settler unit. The existing continuous solvent washers give evidence of low washing efficiency that limits permissible feed activity and that may be related to low contact time between the solvent and the carbonate wash solution. The most general conclusion is that satisfactory operation can be obtained with all projected domestic and foreign fuels under consideration for processing, by suitable adjustment of operating conditions. Also, possible flowsheet and equipment changes are known that could improve operations with these fuels further. 7 references

  8. 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.

  9. 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...

  10. Solvent-extraction and purification of uranium(VI) and molybdenum(VI) by tertiary amines from acid leach solutions

    International Nuclear Information System (INIS)

    La Gamma, Ana M.G.; Becquart, Elena T.; Chocron, Mauricio

    2008-01-01

    Considering international interest in the yellow-cake price, Argentina is seeking to exploit new uranium ore bodies and processing plants. A study of similar plants would suggest that solvent- extraction with Alamine 336 is considered the best method for the purification and concentration of uranium present in leaching solutions. In order to study the purification of these leach liquors, solvent-extraction tests under different conditions were performed with simulated solutions which containing molybdenum and molybdenum-uranium mixtures. Preliminary extraction tests carried out on mill acid-leaching liquors are also presented. (authors)

  11. 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)

  12. PROCESS OF RECOVERING URANIUM

    Science.gov (United States)

    Carter, J.M.; Larson, C.E.

    1958-10-01

    A process is presented for recovering uranium values from calutron deposits. The process consists in treating such deposits to produce an oxidlzed acidic solution containing uranium together with the following imparities: Cu, Fe, Cr, Ni, Mn, Zn. The uranium is recovered from such an impurity-bearing solution by adjusting the pH of the solution to the range 1.5 to 3.0 and then treating the solution with hydrogen peroxide. This results in the precipitation of uranium peroxide which is substantially free of the metal impurities in the solution. The peroxide precipitate is then separated from the solution, washed, and calcined to produce uranium trioxide.

  13. An automated solution enrichment system for uranium analysis

    International Nuclear Information System (INIS)

    Jones, S.A.; Sparks, R.; Sampson, T.; Parker, J.; Horley, E.; Kelly, T.

    1993-01-01

    An automated Solution Enrichment system (SES) for analysis of Uranium and U-235 isotopes in process samples has been developed through a joint effort between Los Alamos National Laboratory and Martin Marietta Energy systems, Portsmouth Gaseous Diffusion Plant. This device features an advanced robotics system which in conjuction with stabilized passive gamma-ray and X-ray fluorescence detectors provides for rapid, non-destructive analyses of process samples for improved special nuclear material accountability and process control

  14. Validation of the Monte Carlo criticality program KENO IV and the Hansen-Roach sixteen-energy-group-cross sections for high-assay uranium systems

    International Nuclear Information System (INIS)

    Handley, G.R.; Masters, L.C.; Stachowiak, R.V.

    1981-01-01

    Validation of the Monte Carlo criticality code, KENO IV, and the Hansen-Roach sixteen-energy-group cross sections was accomplished by calculating the effective neutron multiplication constant, k/sub eff/, of 29 experimentally critical assemblies which had uranium enrichments of 92.6% or higher in the uranium-235 isotope. The experiments were chosen so that a large variety of geometries and of neutron energy spectra were covered. Problems, calculating the k/sub eff/ of systems with high-uranium-concentration uranyl nitrate solution that were minimally reflected or unreflected, resulted in the separate examination of five cases

  15. The uptake of uranium from soil to vetiver grass (vetiver zizanioides (L.) nash)

    International Nuclear Information System (INIS)

    Luu Viet Hung; Bui Duy Cam; Dang Duc Nhan

    2012-01-01

    Uranium uptake of vetiver grass (Vetiveria zizanioides (L.) Nash) from Eutric Fluvisols (AK), Albic Acrisols (LP), Dystric Fluvisols (TT) and Ferralic Acrisols (TC) in northern Vietnam is assessed. The soils were mixed with aqueous solution of uranyl nitrate to make soils be contaminated with uranium at 0, 50, 100, 250 mg per kg before planting the grass. The efficiency of uranium uptake by the grass was assessed based on the soil-to-plant transfer factor (TF U , kg kg -1 ). It was found that the TF U values are dependent upon the soil properties. CEC facilitates the uptake and the increase soil pH could reduce the uptake and translocation of uranium in the plant. Organic matter content as well as ferrous and potassium inhibit the uranium uptake of the grass. It was revealed that the lower fertile soil the higher uranium uptake. The grass could tolerate to the high extent (up to 77%) of uranium in soils and could survive and grow well without fertilization. The translocation of uranium in root for all the soil types studies almost higher than that in its shoot. It seem that vetiver grass potentially be use for the purpose of phytoremediation of soils contaminated with uranium. (author)

  16. Synthesis of microspheres of triuranium octaoxide by simultaneous water and nitrate extraction from ascorbate-uranyl sols

    International Nuclear Information System (INIS)

    Brykala, M.; Deptula, A.; Rogowski, M.; Lada, W.; Olczak, T.; Wawszczak, D.; Smolinski, T.; Wojtowicz, P.; Modolo, G.

    2014-01-01

    A new method for synthesis of uranium oxide microspheres (diameter <100 μm) has been developed. It is a variant of our patented Complex Sol-Gel Process, which has been used to synthesize high-quality powders of a wide variety of complex oxides. Starting uranyl-nitrate-ascorbate sols were prepared by addition of ascorbic acid to uranyl nitrate hexahydrate solution and alkalizing by aqueous ammonium hydroxide and then emulsified in 2-ethylhexanol-1 containing 1v/o SPAN-80. Drops of emulsion were firstly gelled by extraction of water by the solvent. Destruction of the microspheres during thermal treatment, owing to highly reactive components in the gels, requires modification of the gelation step by Double Extraction Process-simultaneously extraction of water and nitrates using Primene JMT, which completely eliminates these problem. Final step was calcination in air of obtained microspheres of gels to triuranium octaoxide. (author)

  17. Recovery of uranium from uranium bearing black shale

    International Nuclear Information System (INIS)

    Das, Amrita; Yadav, Manoj; Singh, Ajay K.

    2016-01-01

    Black shale is the unconventional resource of uranium. Recovery of uranium from black shale has been carried out by the following steps: i) size reduction, ii) leaching of uranium in the aqueous medium, iii) fluoride ion removal, iv) solvent extraction of uranium from the aqueous leach solution, v) scrubbing of the loaded solvent after extraction to remove impurities as much as possible and vi) stripping of uranium from the loaded organic into the aqueous phase. Leaching of black shale has been carried out in hydrochloric acid. Free acidity of the leach solution has been determined by potentiometric titration method. Removal of fluoride ions has been done using sodium chloride. Solvent extraction has been carried out by both tributyl phosphate and alamine-336 as extractants. Scrubbing has been tried with oxalic acid and sulphuric acid. Stripping with sodium carbonate solution has been carried out. Overall recovery of uranium is 95%. (author)

  18. Recovery of uranium from the Syrian phosphate by solid-liquid method using alkaline solutions

    International Nuclear Information System (INIS)

    Shlewit, H.; Alibrahim, M.

    2007-01-01

    Uranium concentrations were analyzed in the Syrian phosphate deposits. Mean concentrations were found between 50 and 110 ppm. As a consequence, an average phosphate dressing of 22 kg/ha phosphate would charge the soil with 5-20 g/ha uranium when added as a mineral fertilizer. Fine grinding phosphate produced at the Syrian mines was used for uranium recovery by carbonate leaching. The formation of the soluble uranyl tricarbonate anion UO 2 (CO 3 ) 3 4- permits use of alkali solutions of sodium carbonate and sodium bicarbonate salts for the nearly selective dissolution of uranium from phosphate. Separation of iron, aluminum, titanium, etc., from the uranium during leaching was carried out. Formation of some small amounts of molybdates, vanadates, phosphates, aluminates, and some complexes metal was investigated. This process could be used before the manufacture of TSP fertilizer, and the final products would contain smaller uranium quantities. (author)

  19. Methods of Uranium Determination in solutions of Tributyl Phosphate and Kerosene; Metodos de determinacion de uranio en soluciones de fosfato de tributilo y queroseno

    Energy Technology Data Exchange (ETDEWEB)

    Petrement Eguiluz, J; Palomares Delgado, F

    1962-07-01

    A new analytical method for the determination of uranium in organic solutions of tributyl phosphate and kerosene is proposed. In this method the uranium is reectracted from the aqueous phase by reduction with cadmium in acid solution. The uranium can be determined in this solution by the usual methods. In case of very diluted solutions, a direct spectrophtometrical determination of uranium in the organic phase with dibenzoylmethane is proposed. (Author) 21 refs.

  20. Uranium hexafluoride and uranyl nitrate. Ionometric determination of bromine

    International Nuclear Information System (INIS)

    Anon.

    Bromine was determined in uranium hexafluoride. The method is suitable for determining 2 to 20 ppm with respect to uranium. Bromides are oxidized by potassium permanganate to give bromine which is extracted into carbon tetrachloride, reduced by ascorbic acid and determined by ionometry [fr

  1. Standard test method for determination of uranium or plutonium isotopic composition or concentration by the total evaporation method using a thermal ionization mass spectrometer

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This method describes the determination of the isotopic composition and/or the concentration of uranium and plutonium as nitrate solutions by the thermal ionization mass spectrometric (TIMS) total evaporation method. Purified uranium or plutonium nitrate solutions are loaded onto a degassed metal filament and placed in the mass spectrometer. Under computer control, ion currents are generated by heating of the filament(s). The ion beams are continually measured until the sample is exhausted. The measured ion currents are integrated over the course of the run, and normalized to a reference isotope ion current to yield isotopic ratios. 1.2 In principle, the total evaporation method should yield isotopic ratios that do not require mass bias correction. In practice, some samples may require this bias correction. When compared to the conventional TIMS method, the total evaporation method is approximately two times faster, improves precision from two to four fold, and utilizes smaller sample sizes. 1.3 The tot...

  2. Stoichiometry of the U3O8 phase formed during calcination of some uranium compounds

    International Nuclear Information System (INIS)

    El-Fekey, S.A.; Farah, M.Y.; Rofail, N.H.

    1981-01-01

    Although recent work has shown U 3 O 8 phase to be the decomposition product obtained after calcining uranyl nitrate, sulphate or ammonium uranate, neither the necessary conditions for obtaining stoichiometric U 3 O 8 nor the details of the reaction have been established. Presence of sulphate or nitrate ions during preparation greatly affects the O/U of the obtained oxides and the physico-chemical properties of uranium tetrafluoride prepared afterwards from it (1-3). The aim of the present investigation was to study the effect of calcination regimes on the stoichiometry of the U 3 O 8 phase produced by the thermal decomposition of uranyl nitrate, sulphate, and ammonium uranate, which was prepared by precipitation from nuclear-pure uranyl sulphate. Stoichiometry of the U 3 O 8 phase formed during calcination of ammonium uranate precipitated from nuclear pure uranyl nitrate solution was reported before (1)

  3. The role of solutions and uranium geochemistry in the geological formations of Jabal Abou-Rejmain-Northern Palmyrides

    International Nuclear Information System (INIS)

    Jubeli, Y.; Aissa, M.; Al-Hilall, M.

    1999-09-01

    The equivalent concentration of potassium, uranium, thorium and total radioactivity of various geological outcrops, along the major wadis of the southern and northern flanks of Jabal Abou-Rejmain, were determined. The radiometric anomalies found to be closely related to either phosphate beds of Companion age, or to spots of yellow secondary uranium minerals precipitated on very soft gray marly limestone rocks of Maestrichtian age. The present work indicates that phosphat deposition in the northern palmy rides seems to be widely distributed in the northern Patmyrides more than it was known earlier. The occurrences of secondary uranium mineralization coincide with the presence of evaporates, such as gypsum, which indicates that the precipitation occurs out of uranium enriched solution which were subjected to aridity and evaporation. The soft marly limestone rocks contains the highest percentage of clay fraction. This fraction is responsible for razing and absorbing the solutions, which were leached out of the Companion facies leading to precipitation of uranium minerals on the surface, as well as, within the pores of the rocks. However, transportation of these solutions, in the study area, seems not to extent for a long distance. (author

  4. Minimum critical values of uranyl and plutonium nitrate solutions calculated by various routes of the french criticality codes system CRISTAL using the new isopiestic nitrate density law

    International Nuclear Information System (INIS)

    Anno, Jacques; Rouyer, Veronique; Leclaire, Nicolas

    2003-01-01

    This paper provides for various cases of 235 U enrichment or Pu isotopic vectors, and different reflectors, new minimum critical values of uranyl nitrate and plutonium nitrate solutions (H + =0) obtained by the standard IRSN calculation route and the new isopiestic density laws. Comparisons are also made with other more accurate routes showing that the standard one's results are most often conservative and usable for criticality safety assessments. (author)

  5. Highly Efficient Interception and Precipitation of Uranium(VI) from Aqueous Solution by Iron-Electrocoagulation Combined with Cooperative Chelation by Organic Ligands.

    Science.gov (United States)

    Li, Peng; Zhun, Bao; Wang, Xuegang; Liao, PingPing; Wang, Guanghui; Wang, Lizhang; Guo, Yadan; Zhang, Weimin

    2017-12-19

    A new strategy combining iron-electrocoagulation and organic ligands (OGLs) cooperative chelation was proposed to screen and precipitate low concentrations (0-18.52 μmol/L) of uranium contaminant in aqueous solution. We hypothesized that OGLs with amino, hydroxyl, and carboxyl groups hydrophobically/hydrophilically would realize precuring of uranyl ion at pH electrocoagulation would achieve faster and more efficient uranium precipitation. Experimentally, the strategy demonstrated highly efficient uranium(VI) precipitation efficiency, especially with hydrophilic macromolecular OGLs. The uranium removal efficiency at optimized experimental condition reached 99.65%. The decrease of zeta potential and the lattice enwrapping between U-OGLs chelates and flocculation precursor were ascribed to the enhanced uranium precipitation activity. Uranium was precipitated as oxides of U(VI) or higher valences that were easily captured in aggregated micelles under low operation current potential. The actual uranium tailing wastewater was treated, and a satisfied uranium removal efficiency of 99.02% was discovered. After elution of the precipitated flocs, a concentrated uranium solution (up to 106.52 μmol/L) with very few other metallic impurities was obtained. Therefore, the proposed strategy could remove uranium and concentrate it concurrently. This work could provide new insights into the purification and recovery of uranium from aqueous solutions in a cost-effective and environmentally friendly process.

  6. Effect of diluent on extraction of uranyl nitrate from nitric acid solution by tri-n-octylamine

    International Nuclear Information System (INIS)

    Kojima, Takashi; Ukon, Toshiaki; Fukutomi, Hiroshi

    1979-01-01

    The distribution ratios in the extraction equilibriums of uranylnitrate from 3 M HNO 3 by tri-n-octylamine (TOA) nitrate salt in nitrobenzene, chlorobenzene, benzene, toluene, cyclohexane, nitrobenzene-benzene and benzene-cylohexane mixtures have been determined in varying the concentrations of uranyl nitrate and TOA nitrate salt. The extraction mechanisms have been discussed in detail based on the law of mass action. It has been concluded that the extractions of uranyl nitrate by TOA nitrate salt in nitrobenzene, 74% nitrobenzene-benzene and 49% nitrobenzene-benzene mixture are represented by the equation TOAHNO 3 (org) + UO 2 2+ (aq) + 2 NO 3 - (aq) = TOAHUO 2 (NO 3 - ) 3 (org), while the extractions of uranyl nitrate by TOA nitrate salt in chlorobenzene, benzene, toluene, cyclohexane, benzene-cyclohexane mixtures and 24% nitrobenzene-benzene mixture are represented by the equation 2 TOAHNO 3 (org) + UO 2 2+ (aq) + 2 NO 3 - (aq) = TOAHUO 2 (NO 3 ) 3 TOAHNO 3 (org). In the latter the extraction equilibrium constants increase in the order of chlorobenzene < benzene < toluene < cyclohexane and with decreasing of the volume fraction of benzene in benzene-cyclohexane mixtures. The effects of diluent have been discussed in detail on the basis of the Hildebrand-Scatchard theory of regular solutions. (author)

  7. 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

  8. Thermodynamics of uranium and nitric acid extraction from aqueous solution of TBP/diluent

    International Nuclear Information System (INIS)

    Souza Freitas, R.F. de.

    1982-06-01

    A thermodynamically consistent procedure for predicting distribution equilibria for uranyl nitrate and nitric acid between an aqueous solution and 30 vol % tributyl phosphate (TBP) in a hydrocarbon diluent is studied. Experimental work is developed in order to obtain equilibrium data for the system uranyl nitrate, nitric acid, water and 30 vol % TBP in n-dodecane, at 25 0 C and 40 0 C. The theoretical equilibrium data, obtained with the aid of a computer, are compared with the experimental results. (Author) [pt

  9. Standard test method for determination of bromine and chlorine in UF6 and uranyl nitrate by X-Ray fluorescence (XRF) spectroscopy

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2001-01-01

    1.1 This method covers the determination of bromine (Br) and chlorine (Cl) in uranium hexafluoride (UF6) and uranyl nitrate solution. The method as written covers the determination of bromine in UF6 over the concentration range of 0.2 to 8 μg/g, uranium basis. The chlorine in UF6 can be determined over the range of 4 to 160 μg/g, uranium basis. Higher concentrations may be covered by appropriate dilutions. The detection limit for Br is 0.2 μg/g uranium basis and for Cl is 4 μg/g uranium basis. 1.2 This standard may involve hazardous materials, operations and equipment. 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.

  10. Translocation of uranium from water to foodstuff while cooking

    Energy Technology Data Exchange (ETDEWEB)

    Krishnapriya, K.C.; Baksi, Ananya; Chaudhari, Swathi; Gupta, Soujit Sen; Pradeep, T.

    2015-10-30

    Highlights: • Rice can efficiently uptake uranium from water contaminated with uranyl nitrate hexahydrate (UO{sub 2}(NO{sub 3}) 2.6 H{sub 2}O), while cooking. • Unusual uranium uptake to the extent of about 1000 ppm is observed when rice is cooked in highly concentrated uranium contaminated water (1240 ppm). • Nature of interaction of uranium with carbohydrates is probed using small monosaccharides like glucose and mannose. • Electrospray ionization mass spectrometry showed UO{sub 2}{sup 2+} to be the most stable species in water in such solutions which can form complexes with sugars. • The species (UO{sub 2}{sup 2+}) is also observed in the case of water exposed to the common mineral, uranium oxide (UO{sub 2}) and similar type of complexation is observed with sugars. - Abstract: The present work report the unusual uranium uptake by foodstuff, especially those rich in carbohydrates like rice when they are cooked in water, contaminated with uranium. The major staple diet in South Asia, rice, was chosen to study its interaction with UO{sub 2}{sup 2+}, the active uranium species in water, using inductively coupled plasma mass spectrometry. Highest uptake limit was checked by cooking rice at very high uranium concentration and it was found to be good scavenger of uranium. To gain insight into the mechanism of uptake, direct interaction of UO{sub 2}{sup 2+} with monosaccharides was also studied, using electrospray ionization mass spectrometry taking mannose as a model. The studies have been done with dissolved uranium salt, uranyl nitrate hexahydrate (UO{sub 2}(NO{sub 3}){sub 2}·6H{sub 2}O), as well as the leachate of a stable oxide of uranium, UO{sub 2}(s), both of which exist as UO{sub 2}{sup 2+} in water. Among the eight different rice varieties investigated, Karnataka Ponni showed the maximum uranium uptake whereas unpolished Basmati rice showed the minimum. Interaction with other foodstuffs (potato, carrot, peas, kidney beans and lentils) with and

  11. Uranium determination in phosphatized materials by drop electrode polarography

    International Nuclear Information System (INIS)

    Sequeira, F.M.C. de; Abrao, A.

    1987-01-01

    An uranium determination procedure in phosphate rocks and crude phosphoric acid is outlined polarography is used. Uranium is previously separedted by extraction with tri-n-byte phosphate 10%-petroleum ether using aluminium nitrate as salting out agent. (M.L.J.) [pt

  12. Direct spectrophotometric analysis of low level Pu (III) in Pu(IV) nitrate solution

    International Nuclear Information System (INIS)

    Mageswaran, P.; Suresh Kumar, K.; Kumar, T.; Gayen, J.K.; Shreekumar, B.; Dey, P.K.

    2010-01-01

    Among the various methods demonstrated for the conversion of plutonium nitrate to its oxide, the oxalate precipitation process either as Pu (III) or Pu (IV) oxalate gained wide acceptance. Since uranous nitrate is the most successful partitioning agent used in the PUREX process for the separation of Pu from the bulk amount of U, the Pu (III) oxalate precipitation of the purified nitrate solution will not give required decontamination from U. Hence Pu IV oxalate precipitation process is a better option to achieve the end user's specified PuO 2 product. Prior to the precipitation process, ensuring of the Pu (IV) oxidation state is essential. Hence monitoring of the level of Pu oxidation state either Pu (III) or Pu (IV) in the feed solution plays a significant role to establish complete conversion of Pu (III). The method in vogue to estimate Pu(lV) content is extractive radiometry using Theonyl Trifluoro Acetone (TTA). As the the method warrants a sample preparation with respect to acidity, a precise measurement of Pu (IV) without affecting the Pu(III) level in the feed sample is difficult. Present study is focused on the exploration of direct spectrophotometry using an optic fiber probe of path length of 40mm to monitor the low level of Pu(III) after removing the bulk Pu(lV) which interfere in the Pu(III) absorption spectrum, using TTA-TBP synergistic mixture without changing the sample acidity

  13. Calibration of X-ray densitometers for the determination of uranium and plutonium concentrations in reprocessing input and product solutions

    International Nuclear Information System (INIS)

    Ottmar, H.; Eberle, H.; Michel-Piper, I.; Kuhn, E.; Johnson, E.

    1985-11-01

    In June 1985 a calibration exercise has been carried out, which included the calibration of the KfK K-Edge Densitometer for uranium assay in the uranium product solutions from reprocessing, and the calibration of the Hybrid K-Edge/K-XRF Instrument for the determination of total uranium and plutonium in reprocessing input solutions. The calibration measuremnts performed with the two X-ray densitometers are described and analyzed, and calibration constants are evaluated from the obtained results. (orig.)

  14. Study of uranium(VI) speciation in phosphoric acid solutions and of its recovery by solvent extraction

    International Nuclear Information System (INIS)

    Dartiguelongue, Adrien

    2014-01-01

    Because small amounts of uranium are present in phosphate rocks, wet phosphoric acids may contain up to 300 ppm of uranium(VI). Therefore, such acids are a cost-effective unconventional source of this metal. Its recovery is a challenge for metallurgical firms which must develop reliable and selective solvent extraction processes. Such processes need to know the chemical equilibria involved in the extraction process, the speciation of uranium and its thermodynamics in solution. These two last points have been investigated in this work. Firstly, the most probable species of uranium(VI) in phosphoric acid solutions have been selected thanks to a detailed review of the literature. Then, a thermodynamic model founded on an equation of state for electrolytes has been built according these hypotheses. It has been validated with speciation data coming from original ATR-IR spectroscopy measurements. Finally, the composition of the aqueous phosphoric acid solutions and the activity coefficients obtained have been combined with a chemical model of uranium(VI) extraction into an organic phase containing a synergistic mixture of bis(2-ethylhexyl)phosphoric acid (D2EHPA) and tri-n-octylphosphine oxide (TOPO) in order to represent the variation of the distribution coefficient of uranium(VI) with H 3 PO 4 concentration. This model had been previously developed at Chimie ParisTech at a given concentration of H 3 PO 4 (i.e., 5,3 mol/L), but in the present study we have tested its validity in an extended range of phosphoric acid concentrations (i.e., 1-7 mol/L) and improved it. (author)

  15. Sodium nitrate-cerium nitrate-water ternary system at 25 deg C

    International Nuclear Information System (INIS)

    Fedorenko, T.P.; Onishchenko, M.K.

    1978-01-01

    Solubility isotherm of sodium nitrate-cerium nitrate-water system at 25 deg C consists of three crystallization branches of initial salts and double compound of the composition 2NaNO 3 xCe(NO 3 ) 3 x2H 2 O. Sodium nitrate introduced in the solution strengthens complexing. Physico-chemical characteristics are in a good agreement with solubility curve

  16. Measurement of uranium concentration by molecular absorption spectrophotometry by means optical fibers; Medicion continua de concentracion de uranio por espectrofotometria de absorcion molecular mediante fibras opticas

    Energy Technology Data Exchange (ETDEWEB)

    Gauna, Alberto C.; Pascale, Ariel A. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Ezeiza. Agencia Minipost

    1996-07-01

    An on-line method for measuring the concentration of uranium in uranyl nitrate-nitric acid aqueous solutions is described. The method is based on molecular absorption spectrophotometry with transmission of light by means of optical fibers. It is ideally suited for control and processes development applications. (author)

  17. Plutonium-uranium separation in the Purex process using mixtures of hydroxylamine nitrate and ferrous sulfamate

    International Nuclear Information System (INIS)

    McKibben, J.M.; Chostner, D.F.; Orebaugh, E.G.

    1983-11-01

    Laboratory studies, followed by plant operation, established that a mixture of hydroxylamine nitrate (HAN) and ferrous sulfamate (FS) is superior to FS used alone as a reductant for plutonium in the Purex first cycle. FS usage has been reduced by about 70% (from 0.12 to 0.04M) compared to the pre-1978 period. This reduced the volume of neutralized waste due to FS by 194 liters/metric ton of uranium (MTU) processed. The new flowsheet also gives lower plutonium losses to waste and at least comparable fission product decontamination. To achieve satisfactory performance at this low concentration of FS, the acidity in the 1B mixer-settler was reduced by using a split-scrub - a low acid scrub in stage one and a higher acid scrub in stage three - to remove acid from the solvent exiting the 1A centrifugal contactor. 8 references, 14 figures, 1 table

  18. Uranium determination by spectrophotometry, in chloride solutions, using titanium (III) as reducer; Determinacao de uranio por espectrofotometria, em solucoes cloridricas, utilizando titanio (III) como redutor

    Energy Technology Data Exchange (ETDEWEB)

    Bastos, E T.R.; Bastos, M B.R.

    1986-08-01

    A simple method for determining uranium in uranium (VI) solutions with the presence of uranium (IV), iron (II), and titanium (IV) in chloridic solution is described. The method comprises in uranium (VI) reduction with titanium (III), acidity adjustment and uranium (IV) spectrophotometry in hydrochloric acid 2 M. (C.G.C.).

  19. Precipitation of uranium oxide by reduction in alkaline solution; Precipitation d'oxyde d'uranium par reduction en milieu alcalin

    Energy Technology Data Exchange (ETDEWEB)

    Pottier, P; Claus, J [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-07-01

    In the first part of the report the authors study the reaction mechanism for this reduction which makes it possible to precipitate a hydrated uranium oxide from alkaline uranyl carbonate solutions. The research into the effects of different variables on numerous cycles are then summarized. Optical, X-ray and thermogravimetric examinations then make it possible to predict the properties of this oxide. In the second part the authors carry out calculations for the continuous operation of single cells and cells in series. These calculations give the data required for the construction of 2 cells having capacities of 0.3 and 10 litres. Results obtained from the continuous operation of this latter cell lead to certain conclusions concerning the applicability of this method to the hydrometallurgy of uranium. (authors) [French] Dans une premiere partie, les auteurs etudient le mecanisme de reaction de cette reduction qui permet la precipitation d'un oxyde d'uranium hydrate dans les solutions d'uranyle-carbonates alcalins. Les etudes de diverses variables sur de nombreux cycles sont ensuite resumees. Puis des examens optiques, aux rayons X et par thermogravimetrie, permettent de proposer une hypothese sur les proprietes de l'oxyde obtenu. Dans la deuxieme partie, les auteurs developpent un calcul prevoyant la marche continue de cellules uniques et en cascades. De ces calculs on tire les elements permettant la realisation de deux cellules de 0,3 et 10 litres. Des resultats de marche continue sur cette derniere cellule, on peut conclure a l'applicabilite de cette methode a l'hydrometallurgie de l'uranium. (auteurs)

  20. Determination of uranium and plutonium in urine of people working with regenerated uranium

    International Nuclear Information System (INIS)

    Golutvina, M.M.; Ryzhova, E.A.

    1987-01-01

    Method of determining uranium and plutonium content in urine with their combined presence up to α-activity ratio Pu:U=1:100 is developed. The method is based on extraction chromatographic separation of nuclides using trimethyloctylammonium nitrate and their subsequent α-spectrometric determination. The coefficient of plutonium purification from uranium makes up 750. Chemical yield of Pu is 72±6%, U-76±8%. The method sensitivity is 0.2 decompositions per minute for a sample

  1. Determination of Ga in aqueous uranium solution by EDXRF

    International Nuclear Information System (INIS)

    Natarajan, V.; Purohit, P.J.; Goyal, Neelam; Seshagiri, T.K.; Godbole, S.V.; Manchanda, V.K.

    2009-01-01

    A method has been developed using EDXRF technique for the determination of gallium in aqueous solution using a set of solution standards in the concentration range 20-5000 μg/ml. When this method was applied to U containing solutions, the estimated values were found to be lower due to matrix effects. Hence the method was modified in order to determine gallium in the presence of uranium using lower tube current and another set of standards with U concentration at 100 mg/ml. The method was applicable for the estimation of Ga from 50 μg/ml to 5mg/ml (i.e.0.05-5% Ga in U). Three synthetic samples were analysed by the present methods in order to evaluate the method for its reliability and reproducibility. (author)

  2. Preparation of electrodeless discharge lamps for emission studies of uranium isotopes at trace level

    International Nuclear Information System (INIS)

    Bhowmick, G.K.; Verma, R.; Verma, M.K.; Raman, V.A.; Joshi, A.R.; Deo, M.N.; Gantayet, L.M.; Tiwari, A.K.; Ramakumar, K.L.; Kumar, Navin

    2010-01-01

    A simplified method for preparation of electrodeless discharge lamp for uranium isotopes with specific concerns for 232 U is described. Micro-gram quantities of solid uranium oxides and aqueous solution of uranium nitrate have been used as a starting material for in situ synthesis of uranium tetraiodide. High temperature iodination reaction is carried out in the presence of inert gas neon. By careful design, the preparation time and surface area of quartz reaction tubes have been reduced considerably. The latter decreases the level of contamination which has a direct bearing on the operational lifetime of the lamps. Incorporation of steps to purify the product from an unwanted material improved the stability of the lamps. The procedure provides a safe and convenient way of handling 232 U in particular but can be extended in general to any actinides having radioactivity similar to that of freshly separated 232 U. Characteristic emission of uranium isotopes have been recorded by Fourier Transform Spectrometer to show the satisfactory operation of the lamps as well as their usage for studying emission spectra of the specific isotope.

  3. Nitrate conversion and supercritical fluid extraction of UO{sub 2}-CeO{sub 2} solid solution prepared by an electrolytic reduction-coprecipitation method

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, L.Y. [Tsinghua Univ., Beijing (China). Inst. of Nuclear and New Energy Technology; China Institute of Atomic Energy, Beijing (China); Duan, W.H.; Wen, M.F.; Xu, J.M.; Zhu, Y.J. [Tsinghua Univ., Beijing (China). Inst. of Nuclear and New Energy Technology

    2014-04-01

    A low-waste technology for the reprocessing of spent nuclear fuel (SNF) has been developed recently, which involves the conversion of actinide and lanthanide oxides with liquid N{sub 2}O{sub 4} into their nitrates followed by supercritical fluid extraction of the nitrates. The possibility of the reprocessing of SNF from high-temperature gas-cooled reactors (HTGRs) with nitrate conversion and supercritical fluid extraction is a current area of research in China. Here, a UO{sub 2}-CeO{sub 2} solid solution was prepared as a surrogate for a UO{sub 2}-PuO{sub 2} solid solution, and the recovery of U and Ce from the UO{sub 2}-CeO{sub 2} solid solution with liquid N{sub 2}O{sub 4} and supercritical CO{sub 2} containing tri-n-butyl phosphate (TBP) was investigated. The UO{sub 2}-CeO{sub 2} solid solution prepared by electrolytic reduction-coprecipitation method had square plate microstructures. The solid solution after heat treatment was completely converted into nitrates with liquid N{sub 2}O{sub 4}. The XRD pattern of the nitrates was similar to that of UO{sub 2}(NO{sub 3}){sub 2} . 3H{sub 2}O. After 120 min of online extraction at 25 MPa and 50 , 99.98% of the U and 98.74% of the Ce were recovered from the nitrates with supercritical CO{sub 2} containing TBP. The results suggest a promising potential technology for the reprocessing of SNF from HTGRs. (orig.)

  4. Extraction with tributyl phosphate (TBP) from ferric nitrate solutions

    International Nuclear Information System (INIS)

    Kolarik, Z.; Grudpan, K.

    1985-01-01

    Ferric nitrate acts as a strong salting-out agent in the extraction of thorium(IV), uranyl, europium(III), samarium(III) and zirconium(IV) nitrates as well as of nitric acid with tributyl phosphate in dodecane. Nitric acid, if present in the extraction system together with large amounts of ferric nitrate, markedly suppresses the extraction of thorium(IV) and lanthanides(III) but significantly supports the extraction of zirconium(IV). Separation factors of different metal pairs are presented as functions of the concentrations of ferric nitrate and nitric acid

  5. Simultaneous removal and recovery of uranium from aqueous solution using TiO_2 photoelectrochemical reduction method

    International Nuclear Information System (INIS)

    Huichao He; Meirong Zong; Faqin Dong; Southwest University of Science and Technology, Sichuan; Pengpan Yang; Gaili Ke; Mingxue Liu; Xiaoqin Nie; Wei Ren; Liang Bian; Southwest University of Science and Technology, Sichuan; Chinese Academy of Sciences, Xinjiang

    2017-01-01

    U(VI)-containing wastewater has potential radiation hazard to the environment, but contains valuable uranium resource. Based on the reduction of U(VI) and the difference in solubility between U(VI) and U(IV), here we construct a TiO_2-based photoelectrochemical cell to remove U(VI) and recover uranium from aqueous solution. By irradiating TiO_2 photoanode at E = 0.45 V versus SCE, U(VI) can be simultaneously removed from aqueous solution and recovered as solid uranium compounds on a FTO glass cathode. Since ethanol can act as hole scavenger to protect the formed U(IV) and provide CO_2"−"· as reductant, ethanol adding improved the U(VI) reduction efficiency of TiO_2-based photoelectrochemical cell. (author)

  6. Role of electrodes in ambient electrolytic decomposition of hydroxylammonium nitrate (HAN) solutions

    OpenAIRE

    Koh, Kai Seng; Chin, Jitkai; Wahida Ku Chik, Tengku F.

    2013-01-01

    Decomposition of hydroxylammonium nitrate (HAN) solution with electrolytic decomposition method has attracted much attention in recent years due to its efficiencies and practicability. However, the phenomenon has not been well-studied till now. By utilizing mathematical model currently available, the effect of water content and power used for decomposition was studied. Experiment data shows that sacrificial material such as copper or aluminum outperforms inert electrodes in the decomposition ...

  7. Contribution to the study of uranyl salts in butyl phosphate solutions

    International Nuclear Information System (INIS)

    Coulon, A.

    1965-06-01

    A spectroscopic study in the normal infrared region and involving the following associations: tri-alkyl phosphates (tri-butyl, tri-ethyl, tri-methyl), uranyl salts (nitrate, chloride, acetate) has confirmed the existence of an interaction between the phosphoryl group and the uranium atom, as shown by a movement of absorption band for the valency P = 0 from ∼ 1270 cm -1 to ∼ 1180 cm -1 . A study of the preparation, analysis and spectroscopy of the solids obtained by the precipitation of uranyl salts by acid butyl phosphates has been carried out. By infrared spectrophotometry it has been shown that the tri-butyl and di-butyl phosphates are associated in non-polar diluents even before the uranium is introduced. The extraction of uranyl salts from acid aqueous solutions by a diluted mixture of tri-butyl and di-butyl phosphates proceeds by different mechanisms according to the nature of the ion (nitrate or chloride). (author) [fr

  8. Uranium speciation in plants

    International Nuclear Information System (INIS)

    Guenther, A.; Bernhard, G.; Geipel, G.; Reich, T.; Rossberg, A.; Nitsche, H.

    2003-01-01

    Detailed knowledge of the nature of uranium complexes formed after the uptake by plants is an essential prerequisite to describe the migration behavior of uranium in the environment. This study focuses on the determination of uranium speciation after uptake of uranium by lupine plants. For the first time, time-resolved laser-induced fluorescence spectroscopy and X-ray absorption spectroscopy were used to determine the chemical speciation of uranium in plants. Differences were detected between the uranium speciation in the initial solution (hydroponic solution and pore water of soil) and inside the lupine plants. The oxidation state of uranium did not change and remained hexavalent after it was taken up by the lupine plants. The chemical speciation of uranium was identical in the roots, shoot axis, and leaves and was independent of the uranium speciation in the uptake solution. The results indicate that the uranium is predominantly bound as uranyl(VI) phosphate to the phosphoryl groups. Dandelions and lamb's lettuce showed uranium speciation identical to lupine plants. (orig.)

  9. Solid state interaction studies on binary nitrate mixtures of uranyl nitrate hexahydrate and lanthanum nitrate hexahydrate at elevated temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Kalekar, Bhupesh; Raje, Naina, E-mail: nraje@barc.gov.in; Reddy, A.V.R.

    2017-02-15

    Interaction behavior of uranyl nitrate hexahydrate (UNH) and lanthanum nitrate hexahydrate (LaNH) have been investigated on the mixtures in different molar ratios of the two precursors and monitoring the reactions at elevated temperatures with thermoanalytical and X-ray diffraction measurement techniques. During the decomposition of equimolar mixture of LaNH and UNH, formation of lanthanum uranate (U{sub 0.5}La{sub 0.5})O{sub 2}, was seen by the temperature of 500 °C along with lanthanum oxide (La{sub 2}O{sub 3}) and uranium trioxide (UO{sub 3}). By the temperature of 700 °C, the formation of uranium sesquioxide (U{sub 3}O{sub 8}) was observed along with (U{sub 0.5}La{sub 0.5})O{sub 2} as end products in uranium rich mixtures. Lanthanum rich compositions decomposed by the temperature of 700 °C to give (U{sub 0.5}La{sub 0.5})O{sub 2} and La{sub 2}O{sub 3} as end products. - Highlights: • UO{sub 2}(NO{sub 3}){sub 2}.6H{sub 2}O and La(NO{sub 3}){sub 3}.6H{sub 2}O interact through their intermediates. • Formation of (U{sub 0.5}La{sub 0.5})O{sub 2} by 500 °C. • La deficient mixtures decompose to give U{sub 3}O{sub 8} and (U{sub 0.5}La{sub 0.5})O{sub 2} as end products. • La rich mixtures decompose to give La{sub 2}O{sub 3} and (U{sub 0.5}La{sub 0.5})O{sub 2} as end product.

  10. 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.)

  11. Decontamination and decommissioning of laboratory solutions enriched uranium (IR-01 b)

    International Nuclear Information System (INIS)

    Diaz Arocas, P. P.; Sama Colao, J.; Garcia Diaz, A.; Torre Rodriguez, J.; Martinez, A.; Argiles, E.; Garrido Delgado, C.

    2010-01-01

    Completed actions decontamination and decommissioning of the Laboratory of Enriched Uranium Solutions, attached to the Radioactivity lR-0l CIEMAT, was carried out final radiological control of the laboratory. From the documentation generated proceeded to request modification of the IR-01 installation by closing its laboratory IR-01 b.

  12. Insight into the electroreduction of nitrate ions at a copper electrode, in neutral solution, after determination of their diffusion coefficient by electrochemical impedance spectroscopy

    International Nuclear Information System (INIS)

    Aouina, Nizar; Cachet, Hubert; Debiemme-chouvy, Catherine; Tran, Thi Tuyet Mai

    2010-01-01

    The electrochemical reduction of nitrate ions at a copper electrode in an unbuffered neutral aqueous solution is studied. Using a two compartment electrochemical cell, three stationary cathodic waves, noted P1, P2 and P3, were evidenced by cyclic voltammetry at -0.9, -1.2 and -1.3 V/SCE, respectively. By comparing the electrochemical response of nitrate and nitrite containing solutions, P1 was attributed to the reduction of nitrate to nitrite. In order to assign P2 and P3 features by determining the number of electrons involved at the corresponding potential, rotating disk electrode experiments at various rotation speeds, combined with linear sweep voltammetry, were performed. Current data analysis at a given potential was carried out using Koutecky-Levich treatment taking into account water reduction. Confident values of the diffusion coefficient D of nitrate ions were assessed by electrochemical impedance spectroscopy for nitrate concentrations of 10 -3 , 10 -2 and 10 -1 M. For a nitrate concentration of 10 -2 M, D was found to be 1.31 x 10 -5 cm 2 s -1 allowing the number of electrons to be determined as 6 for P2 and 8 for P3, in accordance with nitrate reduction into hydroxylamine and ammonia, respectively. The formation of hydroxylamine was confirmed by the observation of its reoxidation at a Pt microelectrode present at the Cu electrode/nitrate solution interface.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  14. Distribution of uranium and radium radionuclides in the 'solid phase-interstitial soil solution' system and their migratory properties in ecosystems

    International Nuclear Information System (INIS)

    Sokolik, G.A.; Ovsyannikova, S.V.; Vojnikova, E.V.; Popenya, M.V.

    2008-01-01

    The background content of the main alpha-emitting radionuclides of uranium and radium in the soils of the south-east territory of the Republic of Belarus has been established. The reserve of migratory active species of uranium and radium in the soils has been determined using the data on the content of the radionuclides in the interstitial soil solutions, which are the most important chain of geochemical and biological migration of the chemical elements in ecosystems. The values of radionuclides distribution coefficients in the 'solid phase - interstitial solution of soil' system were estimated. It was shown that the migratory ability of uranium in the investigated soils is higher than that of radium. A direct correlation between the contents of organic components and uranium in the soil solution has been revealed. The used approach to the investigation of the uranium and radium behavior allows comparing their abilities to the migration in dependence of the soil medium peculiarities. (authors)

  15. Preparation of the pur uranium-metal

    International Nuclear Information System (INIS)

    Goldschmidt, B.; Vertes, P.

    1955-01-01

    A detailed description of the chemical processes used to prepare in the factory of Bouchet of the CEA (Seine-Et-Oise) pur metal uranium with either relatively rich ores, or extracts coming of physical or chemical treatment of poor ores. The nitric treatment of ores succeeds to the production of uranate of impure sodium carbonate. This last last product is dissolved in nitric acid and the uranyl nitrate is extracted by tributyl-phosphate diluted in an inert solvent. The uranyl nitrate pure is re-extracted and successively transformed in uranium peroxide, in orange oxide then in brown oxide which is transformed in fluoride by the anhydrous hydrofluoric acid. Uranate fluoride is then reduced in metal by the pure calcium with an yield superior to 99%. (authors) [fr

  16. Effective Uranium (VI) Sorption from Alkaline Solutions Using Bi-Functionalized Silica-Coated Magnetic Nanoparticles

    International Nuclear Information System (INIS)

    Chen, X.; He, L.; Liu, B.; Tang, Y.

    2015-01-01

    High temperature gas reactor is one of generation IV reactors that can adapt the future energy market, of which the preparation of fuel elements will produce a large amount of radioactive wastewater with uranium and high-level ammonia. Sorption treatment is one of the most important method to recover uranium from wastewater. However, there are few report on uranium sorbent that can directly be applied in wastewater with ammonia. Therefore, the development of a sorbent that can recover uranium in basic environment will greatly decrease the cost of fuel element production and the risk of radioactive pollution. In this work, ammonium-phosphonate-bifunctionalized silica-coated magnetic nanoparticles has been developed for effective sorption of uranium from alkaline media, which are not only advantaged in the uranium separation from liquid phase, but also with satisfactory adsorption rate, amount and reusability. The as-prepared sorbent is found to show a maximum uranium sorption capacity of 70.7 mg/g and a fast equilibrium time of 2 h at pH 9.5 under room temperature. Compared with the mono-functionalized (phosphonate alone and ammonium alone) particles, the combination of the bi-functionalized groups gives rise to an excellent ability to remove uranium from basic environment. The sorbent can be used as a promising solid phase candidate for highly-efficient removal of uranium from basic solution. (author)

  17. Role of electrodes in ambient electrolytic decomposition of hydroxylammonium nitrate (HAN solutions

    Directory of Open Access Journals (Sweden)

    Kai Seng Koh

    2013-09-01

    Full Text Available Decomposition of hydroxylammonium nitrate (HAN solution with electrolytic decomposition method has attracted much attention in recent years due to its efficiencies and practicability. However, the phenomenon has not been well-studied till now. By utilizing mathematical model currently available, the effect of water content and power used for decomposition was studied. Experiment data shows that sacrificial material such as copper or aluminum outperforms inert electrodes in the decomposition of HAN solution. In the case of using copper wire to electrolyse HAN solutions, approximately 10 seconds is required to reach 100 °C regardless of concentration of HAN. In term of power consumption, 100 W–300 W was found to be the range in which decomposition could be triggered effectively using copper wire as electrodes.

  18. Investigation of the effects of radiolytic-gas bubbles on the long-term operation of solution reactors for medical-isotope production

    Science.gov (United States)

    Souto Mantecon, Francisco Javier

    One of the most common and important medical radioisotopes is 99Mo, which is currently produced using the target irradiation technology in heterogeneous nuclear reactors. The medical isotope 99Mo can also be produced from uranium fission using aqueous homogeneous solution reactors. In solution reactors, 99Mo is generated directly in the fuel solution, resulting in potential advantages when compared with the target irradiation process in heterogeneous reactors, such as lower reactor power, less waste heat, and reduction by a factor of about 100 in the generation of spent fuel. The commercial production of medical isotopes in solution reactors requires steady-state operation at about 200 kW. At this power regime, the formation of radiolytic-gas bubbles creates a void volume in the fuel solution that introduces a negative coefficient of reactivity, resulting in power reduction and instabilities that may impede reactor operation for medical-isotope production. A model has been developed considering that reactivity effects are due to the increase in the fuel-solution temperature and the formation of radiolytic-gas bubbles. The model has been validated against experimental results from the Los Alamos National Laboratory uranyl fluoride Solution High-Energy Burst Assembly (SHEBA), and the SILENE uranyl nitrate solution reactor, commissioned at the Commissariat a l'Energie Atomique, in Valduc, France. The model shows the feasibility of solution reactors for the commercial production of medical isotopes and reveals some of the important parameters to consider in their design, including the fuel-solution type, 235U enrichment, uranium concentration, reactor vessel geometry, and neutron reflectors surrounding the reactor vessel. The work presented herein indicates that steady-state operation at 200 kW can be achieved with a solution reactor consisting of 120 L of uranyl nitrate solution enriched up to 20% with 235U and a uranium concentration of 145 kg/m3 in a graphite

  19. HPAT: A nondestructive analysis technique for plutonium and uranium solutions

    International Nuclear Information System (INIS)

    Aparo, M.; Mattia, B.; Zeppa, P.; Pagliai, V.; Frazzoli, F.V.

    1989-03-01

    Two experimental approaches for the nondestructive characterization of mixed solutions of plutonium and uranium, developed at BNEA - C.R.E. Casaccia, with the goal of measuring low plutonium concentration (<50 g/l) even in presence of high uranium content, are described in the following. Both methods are referred to as HPAT (Hybrid Passive-Active Technique) since they rely on the measurement of plutonium spontaneous emission in the LX-rays energy region as well as the transmission of KX photons from the fluorescence induced by a radioisotopic source on a suitable target. Experimental campaigns for the characterization of both techniques have been carried out at EUREX Plant Laboratories (C.R.E. Saluggia) and at Plutonium Plant Laboratories (C.R.E. Casaccia). Experimental results and theoretical value of the errors are reported. (author)

  20. Solvent extraction of uranyl nitrate (1962); Extraction du nitrate d'uranyle par solvant (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Talmont, X; Regnaut, P [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires; Durandet, J; Renault, Ph; Gladel, Y L [Institut Francais du Petrole (IFP), Dept. de Physico-Chimie Appliquee, 92 - Rueil-Malmaison (France)

    1962-07-01

    The transfer of uranyl nitrate from an aqueous acid phase into a solvent (dilute tributylphosphate) is attended by nitric acid transfer. In these conditions, equilibrium data cannot be represented by a plane curve, which complicates the determination of the efficiency of the extractor used, i.e. the calculation of either the number of theoretical plates or the number of transfer units equivalent to the apparatus. The authors are presenting a simple method for estimating the efficiency of a column, based upon either uranium or acid transfer. This method can be used when the profile of uranium and acid concentrations in a phase circulating in the equipment bas been previously determined. On another hand, it enables to study the variation of local efficiency, i.e. the efficiency of different sections of the column. (authors) [French] Le transfert du nitrate d'uranyle d'une phase aqueuse acide dans un solvant (phosphate de tributyle dilue) s'accompagne d'un transfert d'acide nitrique. Dans ces conditions, les donnees d'equilibre ne sont pas representees par une courbe plane, ce qui complique la determination de l'efficacite de l'extracteur utilise, c'est-a-dire le calcul, soit du nombre d'etages theoriques, soit du nombre d'unites de transfert auquel l'appareil est equivalent. Les auteurs presentent une methode simple d'evaluation de l'efficacite d'une colonne basee, soit sur le transfert d'uranium, soit sur celui d'acide. Cette methode est utilisable lorsqu'en a determine au prealable le profil des concentrations en uranium et en acide d'une phase circulant dans l'appareil. Elle permet, d'autre part, d'etudier la variation de l'efficacite locale, c'est-a-dire l'efficacite de differentes sections de la colonne. (auteurs)

  1. Recovery of valuable products from the raffinate of uranium and thorium pilot-plant

    International Nuclear Information System (INIS)

    Martins, E.A.J.

    1990-01-01

    IPEN-CNEN/SP has being very active in refining yellow cake to pure ammonium diuranate which is converted to uranium trioxide, uranium dioxide, uranium tetra-and hexa-fluoride in sequential way. The technology of the thorium purification and its conversion to nuclear grade products has been a practice since several years as well. For both elements the major waste to be worked is the raffinate from purification via TBP-varsol in pulsed columns. In this paper the actual processing technology is reviewed with special emphasis on the recovery of valuable products, mainly nitric acid, ammonium nitrate, uranium, thorium and rare earth elements. Ammonium nitrate from the precipitation of uranium diuranate is of good quality, being radioactivity and uranium-free, and recommended to be applied as fertilizer. In conclusion the main effort is to maximize the recycle and reuse of the above mentioned chemicals. (author)

  2. Extraction of plutonium and uranium from oxalate bearing solutions using phosphonic acid

    International Nuclear Information System (INIS)

    Godbole, A.G.; Mapara, P.M.; Swarup, Rajendra

    1995-01-01

    A feasibility study on the solvent extraction of plutonium and uranium from solutions containing oxalic and nitric acids using a phosphonic acid extractant (PC88A) was made to explore the possibility of recovering Pu from these solutions. Batch experiments on the extraction of Pu(IV) and U(VI) under different parameters were carried out using PC88A in dodecane. The results indicated that Pu could be extracted quantitatively by PC88A from these solutions. A good separation of Pu from U could be achieved at higher temperatures. (author). 6 refs., 3 tabs

  3. Calculated K-effectives using ENDF/B-V data for U + Pu solution critical experiments

    International Nuclear Information System (INIS)

    Primm, R.T. III; Mincey, J.F.

    1981-01-01

    Effective multiplication factors for 12 critical experiments have been calculated using multigroup cross sections derived from the ENDF/B-V library. All 12 experiments contained mixed plutonium and uranium nitrate solutions. The range of hydrogen-to-fissile plutonium atom ratios spanned by these experiments was 200 to 2200. A comparison with K-effectives calculated with ENDF/B-IV data is presented

  4. Simulation of effects of redox and precipitation on diffusion of uranium solution species in backfill

    International Nuclear Information System (INIS)

    Carnahan, C.L.

    1987-12-01

    This investigation addresses the problem of prediction of the rate of migration of redox-sensitive solution species within packing and backfill materials under conditions of variable oxidation potential. Effects of changes of oxidation potential and precipitation of stable uranium compounds during diffusion of uranium from a region of high oxidation potential into a region of low oxidation potential were simulated numerically. Questions of particular interest addressed in the investigation were the existence of a moving ''redox front'' and the influence of precipitation-dissolution processes on uranium migration. The simulations showed that no expanding redox fronts existed at any simulated time up to 3.2 x 10 5 years (10 13 s). In simulations where precipitation of stable solids was not allowed, variations of oxidation potential did not affect total uranium concentrations in solution. Concentration profiles could be predicted simply by diffusion of the (constant) source concentrations. In simulations where precipitation of stable solids was allowed, uraninite and calcium uranate accumulated at the source-transport domain interface, while coffinite penetrated further into the transport domain. Total uranium concentrations in regions of precipitation were determined by solubilities of the precipitated solids, and were six to seven orders of magnitude lower than those in the simulations without precipitation, throughout the domain of transport. 14 refs., 7 figs., 2 tabs

  5. Selective recovery of uranium from Ca-Mg uranates by chlorination

    Science.gov (United States)

    Pomiro, Federico J.; Gaviría, Juan P.; Quinteros, Raúl D.; Bohé, Ana E.

    2017-07-01

    A chlorination process is proposed for the uranium extraction and separation using Calciumsbnd Magnesium uranates such as starting reactants which were obtained by precipitation from uranyl nitrate solutions with calcium hydroxide. The study is based on thermodynamic and reaction analysis using chlorine gas as chlorination agent. The results showed that the chlorination reaction of Ca uranate is more feasible to occur than the Mg uranate. The products obtained after chlorination reactions were washed with deionized water to remove the chlorides produced and analyzed. The XRD patterns of the washed products indicated that the chlorination between 400 and 500 °C result in a single phase of calcium uranate (CaUO4) as reaction product. The formation of U3O8 and MgU3O10 was observed at temperatures between 600 °C and 700 °C for 8 hs. The optimal conditions to recover uranium were 3 l h-1 of chlorine and 10 hs of reaction at 700 °C being U3O8 the single uranium product obtained.

  6. The precipitation of double fluoride salts of uranium

    Energy Technology Data Exchange (ETDEWEB)

    Muir, C. W.A.

    1963-02-15

    Bench-scale kinetic tests were conducted to study the reduction and precipitation reactions involved in the preparation of ammonium uranous fluoride from high-purity uranyl nitrate solutions. Sulphur dioxide and formic acid were used to form the active reducing agent, nascent hyposulphite ion. The reduction was affected in the presence of ammonium fluoride, resulting in the precipitation of the highly insoluble double salt. It was found that uranium was precipituted at a constant rate throughout the progress of the reaction. It is postulated that the reducing agent was continuously regenerated, and that this reaction was rate controlling. As a result of this study, a reaction mechanism is proposed. (auth)

  7. Insight into the electroreduction of nitrate ions at a copper electrode, in neutral solution, after determination of their diffusion coefficient by electrochemical impedance spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Aouina, Nizar; Cachet, Hubert [Laboratoire Interfaces et Systemes Electrochimiques - UPR15 du CNRS, Universite Pierre et Marie Curie - Paris 6, 4, place Jussieu, F-75005 Paris (France); Debiemme-chouvy, Catherine, E-mail: catherine.debiemme-chouvy@upmc.f [Laboratoire Interfaces et Systemes Electrochimiques - UPR15 du CNRS, Universite Pierre et Marie Curie - Paris 6, 4, place Jussieu, F-75005 Paris (France); Tran, Thi Tuyet Mai [Laboratoire Interfaces et Systemes Electrochimiques - UPR15 du CNRS, Universite Pierre et Marie Curie - Paris 6, 4, place Jussieu, F-75005 Paris (France)

    2010-10-01

    The electrochemical reduction of nitrate ions at a copper electrode in an unbuffered neutral aqueous solution is studied. Using a two compartment electrochemical cell, three stationary cathodic waves, noted P1, P2 and P3, were evidenced by cyclic voltammetry at -0.9, -1.2 and -1.3 V/SCE, respectively. By comparing the electrochemical response of nitrate and nitrite containing solutions, P1 was attributed to the reduction of nitrate to nitrite. In order to assign P2 and P3 features by determining the number of electrons involved at the corresponding potential, rotating disk electrode experiments at various rotation speeds, combined with linear sweep voltammetry, were performed. Current data analysis at a given potential was carried out using Koutecky-Levich treatment taking into account water reduction. Confident values of the diffusion coefficient D of nitrate ions were assessed by electrochemical impedance spectroscopy for nitrate concentrations of 10{sup -3}, 10{sup -2} and 10{sup -1} M. For a nitrate concentration of 10{sup -2} M, D was found to be 1.31 x 10{sup -5} cm{sup 2} s{sup -1} allowing the number of electrons to be determined as 6 for P2 and 8 for P3, in accordance with nitrate reduction into hydroxylamine and ammonia, respectively. The formation of hydroxylamine was confirmed by the observation of its reoxidation at a Pt microelectrode present at the Cu electrode/nitrate solution interface.

  8. Development of FR fuel cycle in Japan (3) - Current state on unified technology of De-nitration conversion and granulation for the simplified pellet fuel production based on microwave heating

    International Nuclear Information System (INIS)

    Suzuki, M.; Ishii, K.; Yamamoto, T.; Kato, Y.; Kurita, T.; Yoshimoto, K.; Kihara, Y.; Namekawa, T.; Fujii, K. I.

    2008-01-01

    The major experimental results and current state on unified and simplified de-nitration conversion processing are explained. This technology was enabled by the nitrate solutions mixing method which is our original idea proposed in FS phase II project and transferred to present FaCT project where the mass production is the supreme subject. The major results obtained up to present, which are based on the two originalities, are as follows; (1) The mixing rate of the plutonium nitrate solution (PNS) into the uranium nitrate solution (UNS) was satisfactorily adjusted by the feeding of 0.5 litter step, (2) Excellent size uniformity around 10 μm was obtained in the MOX powders including 30 wt% of PUO 2 products, (3) In the microwave heating, a mild and stable boiling occurred at whole space in the evaporation dish and the size of bubbles gradually became smaller closing to the finishing stage, (4) The Can coefficient of raw MOX powders containing 30 wt% of PUO 2 was ranged 20-40 just before granulation, being rather difficult to pack them in a die, but after the granulation, the packing rate reached 100 %. (authors)

  9. Vein-type uranium deposits

    International Nuclear Information System (INIS)

    Rich, R.A.; Holland, H.D.; Petersen, U.

    1975-01-01

    A critical review is presented of published data bearing on the mineralogy, paragenesis, geochemistry, and origin of veiw-type uranium deposits. Its aim is to serve as a starting point for new research and as a basis for the development of new exploration strategies. During the formation of both vein and sandstone types of deposits uranium seems to have been dissolved by and transported in rather oxidized solutions, and deposited where these solutions encountered reducing agents such as carbon, sulfides, ferrous minerals and hydrocarbons. Granitic rocks abnormally enriched in uranium have apparently been the most common source for uranium in vein-type deposits. Oxidizing solutions have been derived either from the surface or from depth. Surface solutions saturated with atmospheric oxygen have frequently passed through red bed or clean sandstone conduits on their way to and from uranium source rocks. Deep solutions of non-surface origin have apparently become sufficiently oxidizing by passage through and equilibration with red beds. The common association of clean sandstones or red beds with uranium-rich granites in the vicinity of vein-type uranium deposits is probably not fortuitous, and areas where these rock types are found together are considered particularly favorable targets for uranium exploration

  10. Calculated k-effectives for light water reactor typical, U + Pu nitrate solution critical experiments

    International Nuclear Information System (INIS)

    Primm, R.T. III; Mincey, J.F.

    1982-01-01

    The Department of Energy's Consolidated Fuel Reprocessing Program has as a goal the design of nuclear fuel reprocessing equipment. In order to validate computer codes used for criticality analyses in the design of such equipment, k-effectives have been calculated for several U + Pu nitrate solution critical experiments. As of January 1981, descriptions of 45 unpoisoned, U + Pu solution experiments were available in the open literature. Twelve of these experiments were performed with solutions which have physical characteristics typical of dissolved, light water reactor fuel. This paper contains a discussion of these twelve experiments, a review of the calculational procedure used to determine k-effectives, and the results of the calculations

  11. Bioreduction and immobilization of uranium in situ: a case study at a USA Department of Energy radioactive waste site, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    Wu, Weimin; Carley, Jack M.; Watson, David B.; Gu, Baohua; Brooks, Scott C.; Kelly, Shelly D.; Kemner, Kenneth M.; Van Nostrand, Joy; Wu, Liyou; Zhou, Jizhong; Luo, Jian; Cardenas, Erick; Fields, Matthew Wayne; Marsh, Terence; Tiedje, James; Green, Stefan; Kostka, Joel; Kitanidis, Peter K.; Jardine, Philip; Criddle, Craig

    2011-01-01

    Bioremediation of uranium contaminated groundwater was tested by delivery of ethanol as an electron donor source to stimulate indigenous microbial bioactivity for reduction and immobilization of uranium in situ, followed by tests of stability of uranium sequestration in the bioreduced area via delivery of dissolved oxygen or nitrate at the US Department of energy's Integrated Field Research Challenge site located at Oak Ridge, Tennessee, USA. After long term treatment that spanned years, uranium in groundwater was reduced from 40-60 mg · L -1 to -1 , below the USA EPA standard for drinking water. The bioreduced uranium was stable under anaerobic or anoxic conditions, but addition of DO and nitrate to the bioreduced zone caused U remobilization. The change in the microbial community and functional microorganisms related to uranium reduction and oxidation were characterized. The delivery of ethanol as electron donor stimulated the activities of indigenous microorganisms for reduction of U(VI) to U(IV). Results indicated that the immobilized U could be partially remobilized by D0 and nitrate via microbial activity. An anoxic environmental condition without nitrate is essential to maintain the stability of bioreduced uranium.

  12. Denitrification in groundwater at uranium mill tailings sites

    International Nuclear Information System (INIS)

    Goering, Timothy J.; Groffman, Armando; Thomson, Bruce

    1992-01-01

    Nitrates are a major contaminant in groundwater at many Uranium Mill Tailings Remedial Action (UMTRA) sites. Microbial denitrification, the transformation of nitrate to nitrogen gas, may be occurring in groundwater at several UMTRA sites. Denitrification is a biologically mediated process whereby facultative anaerobes use nitrate for respiration under anaerobic conditions. Denitrifying bacteria are ubiquitous in soils, sediments, and water. Denitrification requires nitrate, organic carbon, oxygen-limiting conditions, and trace nutrients, especially phosphorus. The lack of organic carbon is the most common limiting factor for denitrification. Denitrification occurs under a limited range of temperature and pH. The uranium milling processes used at UMTRA sites provided a readily available source of carbon and nitrates for denitrifying bacteria. At the Maybell, Colorado, site, the denitrifying organisms Pseudomonas, Flavobacterium and Acinetobacter were identified in core samples of materials from beneath the tailings. In addition, microcosm experiments simulating aquifer conditions beneath the tailings pile showed an average 40 percent decrease in nitrate concentrations over 13 days. At the New Rifle, Colorado, site, aquifer conditions appear favorable for denitrification. Nitrate and organic carbon are readily available in the groundwater, and redox conditions beneath and downgradient of the tailings pile are relatively anoxic. Downgradient from the tailings, total nitrogen is being removed from the groundwater system at a greater rate than the geochemically conservative anion, chloride. This removal may be due to denitrification and adsorption of ammonium onto clay and silt particles. (author)

  13. Denitrification in groundwater at uranium mill tailings sites

    Energy Technology Data Exchange (ETDEWEB)

    Goering, Timothy J [Jacobs Engineering Group, Inc., Albuquerque, NM (United States); Groffman, Armando [Roy F. Weston, Inc., Albuquerque, NM (United States); Thomson, Bruce [University of New Mexico, Albuquerque, NM (United States)

    1992-07-01

    Nitrates are a major contaminant in groundwater at many Uranium Mill Tailings Remedial Action (UMTRA) sites. Microbial denitrification, the transformation of nitrate to nitrogen gas, may be occurring in groundwater at several UMTRA sites. Denitrification is a biologically mediated process whereby facultative anaerobes use nitrate for respiration under anaerobic conditions. Denitrifying bacteria are ubiquitous in soils, sediments, and water. Denitrification requires nitrate, organic carbon, oxygen-limiting conditions, and trace nutrients, especially phosphorus. The lack of organic carbon is the most common limiting factor for denitrification. Denitrification occurs under a limited range of temperature and pH. The uranium milling processes used at UMTRA sites provided a readily available source of carbon and nitrates for denitrifying bacteria. At the Maybell, Colorado, site, the denitrifying organisms Pseudomonas, Flavobacterium and Acinetobacter were identified in core samples of materials from beneath the tailings. In addition, microcosm experiments simulating aquifer conditions beneath the tailings pile showed an average 40 percent decrease in nitrate concentrations over 13 days. At the New Rifle, Colorado, site, aquifer conditions appear favorable for denitrification. Nitrate and organic carbon are readily available in the groundwater, and redox conditions beneath and downgradient of the tailings pile are relatively anoxic. Downgradient from the tailings, total nitrogen is being removed from the groundwater system at a greater rate than the geochemically conservative anion, chloride. This removal may be due to denitrification and adsorption of ammonium onto clay and silt particles. (author)

  14. Recovery of valuable products in liquid effluents from uranium and thorium pilot units

    International Nuclear Information System (INIS)

    Jardim, E.A.; Abrao, A.

    1988-01-01

    IPEN-CNEN/SP has being very active in refining yellowcake to pure ammonium diuranate which is converted to uranium trioxide, uranium dioxide, uranium tetra- and hexafluoride in a sequential way. The technology of the thorium purification and its conversion to nuclear grade products has been a practice since several years as well. For both elements the major waste to be worked is the refinate from the solvent extraction column where uranium and thorium are purified via TBP-varsol in pulsed columns. In this paper the actual processing technology is reviewed with special emphasis on the recovery of valuable products, mainly nitric acid and ammonium nitrate. Distilled nitric acid and the final sulfuric acid as residue are recycle. Ammonium nitrate from the precipitation of uranium diuranate is of good quality, being radioactivity and uranium-free, and recommended to be applied as fertilizer. In conclusion the main effort is to maximise the recycle and reuse of the abovementioned chemicals. (author) [pt

  15. Recovery of thorium along with uranium 233 from Thorex waste solution employing Chitosan

    International Nuclear Information System (INIS)

    Priya, S.; Reghuram, D.; Kumaraguru, K.; Vijayan, K.; Jambunathan, U.

    2003-01-01

    The low level waste solution, generated from Thorex process during the processing of U 233 , contains thorium along with traces of Th 228 and U 233 . Chitosan, a natural bio-polymer derived from Chitin, was earlier used to recover the uranium and americium. The studies were extended to find out its thorium sorption characteristics. Chitosan exhibited very good absorption of thorium (350 mg/g). Chitosan was equilibrated directly with the low level waste solution at different pH after adjusting its pH, for 60 minutes with a Chitosan to aqueous ratio of 1:100 and the raffinates were filtered and analysed. The results showed more than 99% of thorium and U 233 could be recovered by Chitosan between pH 4 and 5. Loaded thorium and uranium could be eluted from the Chitosan by 1M HNO 3 quantitatively. (author)

  16. Rapid removal of uranium from aqueous solutions using magnetic Fe3O4@SiO2 composite particles.

    Science.gov (United States)

    Fan, Fang-Li; Qin, Zhi; Bai, Jing; Rong, Wei-Dong; Fan, Fu-You; Tian, Wei; Wu, Xiao-Lei; Wang, Yang; Zhao, Liang

    2012-04-01

    Rapid removal of U(VI) from aqueous solutions was investigated using magnetic Fe(3)O(4)@SiO(2) composite particles as the novel adsorbent. Batch experiments were conducted to study the effects of initial pH, amount of adsorbent, shaking time and initial U(VI) concentrations on uranium sorption efficiency as well as the desorbing of U(VI). The sorption of uranium on Fe(3)O(4)@SiO(2) composite particles was pH-dependent, and the optimal pH was 6.0. In kinetics studies, the sorption equilibrium can be reached within 180 min, and the experimental data were well fitted by the pseudo-second-order model, and the equilibrium sorption capacities calculated by the model were almost the same as those determined by experiments. The Langmuir sorption isotherm model correlates well with the uranium sorption equilibrium data for the concentration range of 20-200 mg/L. The maximum uranium sorption capacity onto magnetic Fe(3)O(4)@SiO(2) composite particles was estimated to be about 52 mg/g at 25 °C. The highest values of uranium desorption (98%) was achieved using 0.01 M HCl as the desorbing agent. Fe(3)O(4)@SiO(2) composite particles showed a good selectivity for uranium from aqueous solution with other interfering cation ions. Present study suggested that magnetic Fe(3)O(4)@SiO(2) composite particles can be used as a potential adsorbent for sorption uranium and also provided a simple, fast separation method for removal of heavy metal ion from aqueous solution. Copyright © 2011 Elsevier Ltd. All rights reserved.

  17. Electrocatalytic reduction of nitrate at low concentration on coinage and transition-metal electrodes in acid solutions

    NARCIS (Netherlands)

    Dima, G.E.; Vooys, de A.C.A.; Koper, M.T.M.

    2003-01-01

    A comparative study was performed to determine the reactivity of nitrate ions at 0.1 M on eight different polycrystalline electrodes (platinum, palladium, rhodium, ruthenium, iridium, copper, silver and gold) in acidic solution using cyclic voltammetry (CV), chronoamperometry and differential

  18. Comparison of plasma generated nitrogen fertilizer to conventional fertilizers ammonium nitrate and sodium nitrate for pre-emergent and seedling growth

    Science.gov (United States)

    Andhavarapu, A.; King, W.; Lindsay, A.; Byrns, B.; Knappe, D.; Fonteno, W.; Shannon, S.

    2014-10-01

    Plasma source generated nitrogen fertilizer is compared to conventional nitrogen fertilizers in water for plant growth. Root, shoot sizes, and weights are used to examine differences between plant treatment groups. With a simple coaxial structure creating a large-volume atmospheric glow discharge, a 162 MHz generator drives the air plasma. The VHF plasma source emits a steady state glow; the high drive frequency is believed to inhibit the glow-to-arc transition for non-thermal discharge generation. To create the plasma activated water (PAW) solutions used for plant treatment, the discharge is held over distilled water until a 100 ppm nitrate aqueous concentration is achieved. The discharge is used to incorporate nitrogen species into aqueous solution, which is used to fertilize radishes, marigolds, and tomatoes. In a four week experiment, these plants are watered with four different solutions: tap water, dissolved ammonium nitrate DI water, dissolved sodium nitrate DI water, and PAW. Ammonium nitrate solution has the same amount of total nitrogen as PAW; sodium nitrate solution has the same amount of nitrate as PAW. T-tests are used to determine statistical significance in plant group growth differences. PAW fertilization chemical mechanisms are presented.

  19. Method of producing thin cellulose nitrate film

    International Nuclear Information System (INIS)

    Lupica, S.B.

    1975-01-01

    An improved method for forming a thin nitrocellulose film of reproducible thickness is described. The film is a cellulose nitrate film, 10 to 20 microns in thickness, cast from a solution of cellulose nitrate in tetrahydrofuran, said solution containing from 7 to 15 percent, by weight, of dioctyl phthalate, said cellulose nitrate having a nitrogen content of from 10 to 13 percent

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

    International Nuclear Information System (INIS)

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

    1991-09-01

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

  1. 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

  2. Method for determining microamounts of uranium in solutions from copper ores, by liquid-liquid extraction and spectrophotometry with arsenazo III

    International Nuclear Information System (INIS)

    Rodriguez, B.

    1972-01-01

    A spectrophotometric method is described for determining small amounts of uranium in aqueous solutions from copper ores. Uranium is quantitatively separated in a single extraction by a solution of tri-n-octylphosphine oxide in benzene, using ethylendiaminetetracetic acid and sodium fluoride as complexing agents, for improving the selectivity of the procedure. An aliquot of the extract is diluted with a hydrocolloidal solution of arsenazo III. Optical density is measured at 650 nm. (Author) 3 refs

  3. Uranyl nitrate-exposed rat alveolar macrophages cell death: Influence of superoxide anion and TNF α mediators

    International Nuclear Information System (INIS)

    Orona, N.S.; Tasat, D.R.

    2012-01-01

    Uranium compounds are widely used in the nuclear fuel cycle, military and many other diverse industrial processes. Health risks associated with uranium exposure include nephrotoxicity, cancer, respiratory, and immune disorders. Macrophages present in body tissues are the main cell type involved in the internalization of uranium particles. To better understand the pathological effects associated with depleted uranium (DU) inhalation, we examined the metabolic activity, phagocytosis, genotoxicity and inflammation on DU-exposed rat alveolar macrophages (12.5–200 μM). Stability and dissolution of DU could differ depending on the dissolvent and in turn alter its biological action. We dissolved DU in sodium bicarbonate (NaHCO 3 100 mM) and in what we consider a more physiological vehicle resembling human internal media: sodium chloride (NaCl 0.9%). We demonstrate that uranyl nitrate in NaCl solubilizes, enters the cell, and elicits its cytotoxic effect similarly to when it is diluted in NaHCO 3 . We show that irrespective of the dissolvent employed, uranyl nitrate impairs cell metabolism, and at low doses induces both phagocytosis and generation of superoxide anion (O 2 − ). At high doses it provokes the secretion of TNFα and through all the range of doses tested, apoptosis. We herein suggest that at DU low doses O 2 − may act as the principal mediator of DNA damage while at higher doses the signaling pathway mediated by O 2 − may be blocked, prevailing damage to DNA by the TNFα route. The study of macrophage functions after uranyl nitrate treatment could provide insights into the pathophysiology of uranium‐related diseases. -- Highlights: ► Uranyl nitrate effect on cultured macrophages is linked to the doses and independent of its solubility. ► At low doses uranyl nitrate induces generation of superoxide anion. ► At high doses uranyl nitrate provokes secretion of TNFα. ► Uranyl nitrate induces apoptosis through all the range of doses tested.

  4. Uranyl nitrate-exposed rat alveolar macrophages cell death: Influence of superoxide anion and TNF α mediators

    Energy Technology Data Exchange (ETDEWEB)

    Orona, N.S. [School of Science and Technology, National University of General Martín, Avda Gral Paz 5445 (1650) San Martín, Buenos Aires (Argentina); Tasat, D.R., E-mail: deborah.tasat@unsam.edu.ar [School of Science and Technology, National University of General Martín, Avda Gral Paz 5445 (1650) San Martín, Buenos Aires (Argentina); School of Dentistry, University of Buenos Aires, M. T. de Alvear 2142 (1122), Buenos Aires (Argentina)

    2012-06-15

    Uranium compounds are widely used in the nuclear fuel cycle, military and many other diverse industrial processes. Health risks associated with uranium exposure include nephrotoxicity, cancer, respiratory, and immune disorders. Macrophages present in body tissues are the main cell type involved in the internalization of uranium particles. To better understand the pathological effects associated with depleted uranium (DU) inhalation, we examined the metabolic activity, phagocytosis, genotoxicity and inflammation on DU-exposed rat alveolar macrophages (12.5–200 μM). Stability and dissolution of DU could differ depending on the dissolvent and in turn alter its biological action. We dissolved DU in sodium bicarbonate (NaHCO{sub 3} 100 mM) and in what we consider a more physiological vehicle resembling human internal media: sodium chloride (NaCl 0.9%). We demonstrate that uranyl nitrate in NaCl solubilizes, enters the cell, and elicits its cytotoxic effect similarly to when it is diluted in NaHCO{sub 3}. We show that irrespective of the dissolvent employed, uranyl nitrate impairs cell metabolism, and at low doses induces both phagocytosis and generation of superoxide anion (O{sub 2}{sup −}). At high doses it provokes the secretion of TNFα and through all the range of doses tested, apoptosis. We herein suggest that at DU low doses O{sub 2}{sup −} may act as the principal mediator of DNA damage while at higher doses the signaling pathway mediated by O{sub 2}{sup −} may be blocked, prevailing damage to DNA by the TNFα route. The study of macrophage functions after uranyl nitrate treatment could provide insights into the pathophysiology of uranium‐related diseases. -- Highlights: ► Uranyl nitrate effect on cultured macrophages is linked to the doses and independent of its solubility. ► At low doses uranyl nitrate induces generation of superoxide anion. ► At high doses uranyl nitrate provokes secretion of TNFα. ► Uranyl nitrate induces apoptosis through

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2000-01-01

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

  6. Uranium removal from acidic aqueous solutions by Saccharomyces cerevisiae, Debaryomyces hansenii, Kluyveromyces marxianus and Candida colliculosa

    International Nuclear Information System (INIS)

    Sarri, S.; Misaelides, P.; Papanikolaou, M.; Zamboulis, D.

    2009-01-01

    The sorption of uranium from acidic aqueous solutions (pH 4.5, C init = 10 to 1000 mg U/L) by Saccharomyces cerevisiae, Debaryomyces hansenii, Kluyveromyces marxianus and Candida colliculosa was investigated using a batch technique. The U-sorption onto Saccharomyces cerevisiae and Debaryomyces hansenii followed a Langmuir, while that onto Kluyveromyces marxianus and Candida colliculosa a Freundlich isotherm. The results demonstrated that all investigated biomasses could effectively remove uranium from acidic aqueous solutions. From all sorbents, Saccharomyces cerevisiae appeared to be the most effective with a maximum sorption capacity of 127.7 mg U/g dry biomass. (author)

  7. Influence of phosphates when uranium in solutions obtained by attacking Forez with sulfuric acid is precipitated by the action of lime; Influence des phosphates, lors de la precipitation par la chaux, de l'uranium contenu dans les solutions d'attaque sulfurique du Forez

    Energy Technology Data Exchange (ETDEWEB)

    Brebec, G

    1959-03-01

    Influence of phosphates when uranium in solutions obtained by attacking Forez with sulfuric acid is precipitated by the action of lime was studied. Most of the phosphates were eliminated in the form of ferric phosphates without noticeable losses of uranium: for this it is only necessary to add sufficient ferric sulfate to the solution to be treated so that [Po{sub 4}{sup 3-}]/[Fe{sup 3+}] {approx} 0,4. In these conditions, the preparation of a calcium concentrate rich in uranium takes place in two stages. The first is neutralization at pH 2,7 to 2,8 with elimination of phosphates, sulfates and iron; the second is precipitation of the concentrate at pH 6,5. (author) [French] Nous avons reussi a eliminer la majeure partie des phosphates sous forme de phosphates ferriques, sans pertes sensibles d'uranium. Pour cela, il suffit d'ajouter a la solution a traiter, du sulfate ferrique en quantite telle que: (Po{sub 4}{sup 3-}]/[Fe{sup 3+}] {approx} 0,4. Dans ces conditions, la preparation du concentre calcique, riche en uranium, s'effectue normalement en deux temps: 1) preneutralisation a pH 2,7-2,8: elimination des sulfates, phosphates et fer; 2) precipitation du concentre a pH 6,5. (auteur)

  8. Continuous denitration device by microwave heating

    International Nuclear Information System (INIS)

    Matsumaru, Ken-ichi; Sato, Hajime.

    1982-01-01

    Purpose: To continuously obtain powder of uranium dioxide, plutonium dioxide or a mixture of them respectively from the solution of uranyl nitrate, plutonium nitrate or a mixture of them effectively while maintaining a constant quality. Constitution: Plutonium nitrate or uranium nitrate solution is deposited on a rotational drum having a heater and dried into powderous products. The powderous products are scraped off by a blade, transferred to a belt conveyor, entered into a microwave heating furnace and heated by microwaves while stirring to obtain the powder of plutonium dioxide or uranium dioxide. The powderous products are scraped off by a scraper and collected in a receiving tank for denitration products, whereby the feeding solution can be denitrated continuously. (Horiuchi, T.)

  9. Electrochemical preparation of uranium and plutonium measuring probes for alpha spectroscopy from organic solutions

    International Nuclear Information System (INIS)

    Gruner, W.; Beutmann, A.

    1980-01-01

    A method for preparation of uranium and plutonium measuring probes for α-spectrometry is described. The method is based on electrodeposition from isopropanol and especially from ethanol and methanol solution. It was shown that a definite additions of a little amount of water lead to an increase of the deposition rate. It is possible to reach a 100% deposition in ethanol after an electrolysis time of 3 minutes for uranium and 30 minutes for plutonium with voltages of 150-200 V. (author)

  10. Extraction of nitric acid, uranyl nitrate, and bismuth nitrate from aqueous nitric acid solutions with CMPO

    International Nuclear Information System (INIS)

    Spencer, B.B.

    1995-08-01

    DOE sponsored development of the transuranium extraction (TRUEX) process for removing actinides from radioactive wastes. The solvent is a mixture of CMPO and TBP. Since the extraction characteristics of CMPO are not as well understood as those of TBP, the extraction of nitric acid, uranyl nitrate, and bismuth nitrate with CMPO (dissolved in n-dodecane) were studied. Results indicate that CMPO extracts nitric acid with a 1:1 stoichiometry; equilibrium constant is 2. 660±0.092 at 25 C, and extraction enthalpy is -5. 46±0.46 kcal/mol. Slope analysis indicates that uranyl nitrate extracts with a mixed equilibria of 1:1 and 2:1 stoichiometries in nearly equal proportion. Equil. constant of the 2: 1 extraction was 1.213 x 10 6 ±3.56 x 10 4 at 25 C; reaction enthalpy was -9.610±0.594 kcal/mol. Nitration complexation constant is 8.412±0.579, with an enthalpy of -10.72±1.87 kcal/mol. Bismuth nitrate also extracts with a mixed equilibria of (perhaps) 1:1 and 2:1 stoichiometries. A 2:1 extraction equilibrium and a nitrate complexation adequately model the data. Kinetics and enthalpies were also measured

  11. Extraction of nitric acid, uranyl nitrate, and bismuth nitrate from aqueous nitric acid solutions with CMPO

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, B.B.

    1995-08-01

    DOE sponsored development of the transuranium extraction (TRUEX) process for removing actinides from radioactive wastes. The solvent is a mixture of CMPO and TBP. Since the extraction characteristics of CMPO are not as well understood as those of TBP, the extraction of nitric acid, uranyl nitrate, and bismuth nitrate with CMPO (dissolved in n-dodecane) were studied. Results indicate that CMPO extracts nitric acid with a 1:1 stoichiometry; equilibrium constant is 2. 660{plus_minus}0.092 at 25 C, and extraction enthalpy is -5. 46{plus_minus}0.46 kcal/mol. Slope analysis indicates that uranyl nitrate extracts with a mixed equilibria of 1:1 and 2:1 stoichiometries in nearly equal proportion. Equil. constant of the 2: 1 extraction was 1.213 {times} 10{sup 6}{plus_minus}3.56 {times} 10{sup 4} at 25 C; reaction enthalpy was -9.610{plus_minus}0.594 kcal/mol. Nitration complexation constant is 8.412{plus_minus}0.579, with an enthalpy of -10.72{plus_minus}1.87 kcal/mol. Bismuth nitrate also extracts with a mixed equilibria of (perhaps) 1:1 and 2:1 stoichiometries. A 2:1 extraction equilibrium and a nitrate complexation adequately model the data. Kinetics and enthalpies were also measured.

  12. Method for the chemical reprocessing of irradiated nuclear fuels, in particular nuclear fuels containing uranium

    International Nuclear Information System (INIS)

    Koch, G.

    1976-01-01

    In the chemical processing of irradiated uranium-containing nuclear fuels which are hydrolyzed with aqueous nitric acid, a suggestion is made to use as quaternary ammonium nitrate trialkyl-methyl ammonium nitrates as extracting agent, in which the sum of C atoms is greater than 16. In the illustrated examples, tricaprylmethylammonium nitrate, trilaurylmethylammonium nitrate and tridecylmethylammonium nitrate are named. (HPH/LH) [de

  13. Potentiometric determination of free nitric-acid in trilaurylamine solutions containing plutonium nitrate; Dosage potentiometrique de l'acidite nitrique libre dans les solutions organiques de trilaurylamine

    Energy Technology Data Exchange (ETDEWEB)

    Perez, J J; Saey, J C [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1965-07-01

    A potentiometric method of determination of the free nitric acid in trilaurylamine solutions containing plutonium or thorium nitrates is described. The potentiometric titration is carried out in a mixture of benzene and 1,2-dichloro ethane with a standard solution of trilaurylamine as the titrant. When thorium nitrate is present the metal complex is not dissociated then the titration has a single end-point. In the case of plutonium nitrate the partial dissociation of the plutonium complex corresponds to a second point. The experimental error in duplicate analyses of 50 samples is about 1 per cent for free acid concentrations in the range of 0,03 to 0,1 N and plutonium concentrations between 1 to 5 g/l. (authors) [French] Une methode potentiometrique de dosage de l'acidite nitrique libre dans les solutions de trilaurylamine contenant un complexe de plutonium ou de thorium est decrite. La potentiometrie est effectuee en prenant comme base titrante la trilaurylamine et comme milieu de dilution un melange de benzene et de 1,2 dichloroethane. Dans le cas du thorium, le complexe organometallique n'est pas deplace et la courbe de titrage presente un seul point d'inflexion. Dans le cas du plutonium le complexe est partiellement dissocie ce qui correspond a un second saut de potentiel. La moyenne des erreurs experimentales sur 50 echantillons doses a ete d'environ {+-} 1 pour cent sur l'acide libre. Les solutions experimentees contenaient de 0,03 a 0,1 N en acide et de 1 a 5 g/l en plutonium. (auteurs)

  14. Engineering solutions to the long-term stabilization and isolation of uranium mill tailings in the United States

    International Nuclear Information System (INIS)

    Sanders, D.R.; Lommler, J.C.

    1995-01-01

    Engineering solutions to the safe and environmentally protective disposal and isolation of uranium mill tailings in the US include many factors. Cover design, materials selection, civil engineering, erosive forces, and cost effectiveness are only a few of those factors described in this paper. The systems approach to the engineering solutions employed in the US is described, with emphasis on the standards prescribed for the Uranium Mill Tailings Remedial Action Project. Stabilization and isolation of the tailings from humans and the environment are the primary goals of the US uranium mill tailings control standards. The performance of cover designs with respect to water infiltration, radon exhalation, geotechnical stability, erosion protection, human and animal intrusion prevention, and longevity are addressed. The need for and frequency of surveillance efforts to ensure continued disposal system performance are also assessed

  15. Production of thorium nitrate from uranothorianite ores

    International Nuclear Information System (INIS)

    Brodsky, M.; Sartorius, R.; Sousseuer, Y.

    1959-01-01

    The separation of thorium and uranium from uranothorianite ores, either by precipitation or solvent-extraction methods, are discussed, and an industrial process for the manufacture of thorium nitrate is described. Reprint of a paper published in 'Progress in Nuclear Energy' Series III, Vol. 2 - Process Chemistry, 1959, p. 68-76 [fr

  16. Kinetics of dissolution of thorium and uranium doped britholite ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Dacheux, N., E-mail: nicolas.dacheux@univ-montp2.f [Groupe de Radiochimie, Institut de Physique Nucleaire d' Orsay, Bat. 100, Universite Paris-Sud-11, 91406 Orsay (France); Institut de Chimie Separative de Marcoule, UMR 5257 (Universite Montpellier 2/CNRS/CEA/ENSCM), Bat. 426, Centre de Marcoule, BP 17171, 30207 Bagnols sur ceze cedex (France); Du Fou de Kerdaniel, E. [Groupe de Radiochimie, Institut de Physique Nucleaire d' Orsay, Bat. 100, Universite Paris-Sud-11, 91406 Orsay (France); Clavier, N. [Groupe de Radiochimie, Institut de Physique Nucleaire d' Orsay, Bat. 100, Universite Paris-Sud-11, 91406 Orsay (France); Institut de Chimie Separative de Marcoule, UMR 5257 (Universite Montpellier 2/CNRS/CEA/ENSCM), Bat. 426, Centre de Marcoule, BP 17171, 30207 Bagnols sur ceze cedex (France); Podor, R. [Institut de Chimie Separative de Marcoule, UMR 5257 (Universite Montpellier 2/CNRS/CEA/ENSCM), Bat. 426, Centre de Marcoule, BP 17171, 30207 Bagnols sur ceze cedex (France); Institut Jean Lamour - Departement CP2S - Equipe 206, Faculte des Sciences et Techniques - Nancy Universite, BP 70239, 54506 Vandoeuvre les Nancy cedex (France); Aupiais, J. [CEA DAM DIF, 91297 Arpajon (France); Szenknect, S. [Institut de Chimie Separative de Marcoule, UMR 5257 (Universite Montpellier 2/CNRS/CEA/ENSCM), Bat. 426, Centre de Marcoule, BP 17171, 30207 Bagnols sur ceze cedex (France)

    2010-09-01

    In the field of immobilization of actinides in phosphate-based ceramics, several thorium and uranium doped britholite samples were submitted to leaching tests. The normalized dissolution rates determined for several pH values, temperatures and acidic media from the calcium release range from 4.7 x 10{sup -2} g m{sup -2} d{sup -1} to 21.6 g m{sup -2} d{sup -1}. Their comparison with that determined for phosphorus, thorium and uranium revealed that the dissolution is clearly incongruent for all the conditions examined. Whatever the leaching solution considered, calcium and phosphorus elements were always released with higher R{sub L} values than the other elements (Nd, Th, U). Simultaneously, thorium was found to quickly precipitate as alteration product, leading to diffusion phenomena for uranium. For all the media considered, the uranium release is higher than that of thorium, probably due to its oxidation from tetravalent oxidation state to uranyl. Moreover, the evaluation of the partial order related to proton concentration and the apparent energy of activation suggest that the reaction of dissolution is probably controlled by surface chemical reactions occurring at the solid/liquid interface. Finally, comparative leaching tests performed in sulphuric acid solutions revealed a significant influence of such media on the chemical durability of the leached pellets, leading to higher normalized dissolution rates for all the elements considered. On the basis of the results of chemical speciation, this difference was mainly explained in the light of higher complexion constants by sulfate ions compared to nitrate, chloride and phosphate.

  17. Constant current coulometric method for the determination of uranium in active process solutions

    International Nuclear Information System (INIS)

    Chitnis, R.T.; Talnikar, S.G.; Paranjape, A.H.

    1980-01-01

    The determination of uranium in the range of 2.5-5 mg by constant current coulometry is described. The procedure is based on the modified version of the DAVIES - GRAY method, wherein uranium, after the reduction step, is oxidized by adding a known amount of potassium dichromate, and the excess of dichromate is determined by titration with Fe 2+ solution. Fe 2+ ions needed for the titration are generated in situ with 100% current efficiency by electrolytic reduction of Fe 3+ . The method is found to be accurate with a coefficient of variation better than 0.2%. (author)

  18. Deactivation and treatment of mine wastewaters and of aqueous solutions discharged in uranium ore treatment and processing

    International Nuclear Information System (INIS)

    Jilek, R.; Prochazka, H.; Fuska, J.; Nemec, P.; Katzer, J.

    1974-01-01

    A description is presented of decontamination and purification of mine wastewaters and water solutions discharged from uranium ore treatment and processing and of the simultaneous removal and concentration of uranium-radium daughters, mainly of 226 Ra and 210 Pb. These elements are incorporated in the mycelium of microorganisms, such as those of the Fungi imperfecti class or are sorbed on the mycelium surface. The mycelia are then mechanically separated from the decontaminated solution, e.g., by filtration, centrifugation or sedimentation. The mycelium may be cultivated in the purified solutions to which nutrients are added, such as carbon, nitrogen and phosphorus in concentrations necessary for the growth of the microorganisms used. The mycelium may be added to the purified solution either in the native or in the dried state. (B.S.)

  19. Oxidative destruction of ammonia for restoration of uranium solution mining sites

    International Nuclear Information System (INIS)

    Humenick, M.J.; Garwacka, K.

    1984-01-01

    A laboratory experimental research project was conducted to evaluate the use of chlorine for the oxidative destruction of residual ammonia that may remain in ground water after in-situ uranium solution mining operations. The work tested the idea of injecting high strength calcium hypochlorite solution into the mining zone to convert ammonia to nitrogen gas as a final cleanup process for ammonia removal from the ground water system. This paper details ammonia removal efficiency as a function of chlorine dose, reactant, and product material balances, and how the concept may be used as a final ground water restoration process

  20. Oxidative destruction of ammonia for restoration of uranium solution mining sites

    Energy Technology Data Exchange (ETDEWEB)

    Humenick, M.J.; Garwacka, K.

    1984-02-01

    A laboratory experimental research project was conducted to evaluate the use of chlorine for the oxidative destruction of residual ammonia that may remain in ground water after in-situ uranium solution mining operations. The work tested the idea of injecting high strength calcium hypochlorite solution into the mining zone to convert ammonia to nitrogen gas as a final cleanup process for ammonia removal from the ground water system. This paper details ammonia removal efficiency as a function of chlorine dose, reactant, and product material balances, and how the concept may be used as a final ground water restoration process.