Uranium Dioxide Powder Flow ability Improvement Using Sol-Gel

International Nuclear Information System (INIS)

Juanda, D.; Sambodo Daru, G.

1998-01-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Sintering uranium oxide using a preheating step

International Nuclear Information System (INIS)

Jensen, N.J.; Nivas, Y.; Packard, D.R.

1977-01-01

Compacted pellets of uranium oxide or uranium oxide with one or more additives are heated in a kiln in a process having a preheating step, a sintering step, a reduction step, and a cooling step in a controlled atmosphere. The process is practiced to give a range of temperature and atmosphere conditions for obtaining optimum fluoride removal from the compacted pellets along with optimum sintering in a single process. The preheating step of this process is conducted in a temperature range of about 600 0 to about 900 0 C and the pellets are held for at least twenty min, and preferably about 60 min, in an atmosphere having a composition in the range of about 10 to about 75 vol % hydrogen with the balance being carbon dioxide. The sintering step is conducted at a temperature in the range of about 900 0 C to 1500 0 C in the presence of an atmosphere having a composition in the range of about 0.5 to about 90 vol % hydrogen with the balance being carbon dioxide. The reduction step reduces the oxygen to metal ratio of the pellets to a range of about 1.98 to 2.10:1 and this is accomplished by gradually cooling the pellets for about 30 to about 120 min from the temperature of the sintering step to about 1100 0 C in an atmosphere of about 10 to 90 vol % hydrogen with the balance being carbon dioxide. Thereafter the pellets are cooled to about 100 0 C under a protective atmosphere, and in one preferred practice the same atmosphere used in the reduction step is used in the cooling step. The preheating, sintering and reduction steps may also be conducted with their respective atmospheres having an initial additional component of water vapor and the water vapor can comprise up to about 20 vol %

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-01-24

Flash sintering is a novel type of field assisted sintering that uses an electric field and current to provide densification of materials on very short time scales. The potential for field assisted sintering techniques to be used in producing nuclear fuel is gaining recognition due to the potential economic benefits and improvements in material properties. The flash sintering behavior has so far been linked to applied and material parameters, but the underlying mechanisms active during flash sintering have yet to be identified. This report summarizes the efforts to investigate flash sintering of uranium dioxide using dilatometer studies at Los Alamos National Laboratory and two separate sets of in-situ studies at Brookhaven National Laboratory’s NSLS-II XPD-1 beamline. The purpose of the dilatometer studies was to understand individual parameter (applied and material) effects on the flash behavior and the purpose of the in-situ studies was to better understand the mechanisms active during flash sintering. As far as applied parameters, it was found that stoichiometry, or oxygen-to-metal ratio, has a significant effect on the flash behavior (time to flash and speed of flash). Composite systems were found to have degraded sintering behavior relative to pure UO₂. The critical field studies are complete for UO_2.00 and will be analyzed against an existing model for comparison. The in-situ studies showed that the strength of the field and current are directly related to the sample temperature, with temperature-driven phase changes occurring at high values. The existence of an ‘incubation time’ has been questioned, due to a continuous change in lattice parameter values from the moment that the field is applied. Some results from the in-situ experiments, which should provide evidence regarding ion migration, are still being analyzed. Some preliminary conclusions can be made from these results with regard to using field assisted sintering to

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

International Nuclear Information System (INIS)

Landspersky, H.; Tympl, M.

1984-01-01

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

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

International Nuclear Information System (INIS)

Landspersky, H.; Urbanek, V.

1979-01-01

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

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

International Nuclear Information System (INIS)

Sampath, S.; Chawla, K.L.

1975-01-01

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

Kinetics of sintering of uranium dioxide

International Nuclear Information System (INIS)

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

1978-01-01

The kinetics of sintering of UO 2 powders derived from ADU route and calcined at different temperatures was studied. The activation energy for sintering was found to depend on the calcination temperature, the density chosen and the sintering temperature range. The motive force for sintering is the excess free energy in the particle system. This exists in the powder compact in the form of surface energy and the excess lattice energy due to defects. The defects which can be eliminated at the operating temperature are responsible for the mobility and hence sintering. This concept of the motive force for sintering has been used to explain the difference in the activation energies observed in the present study. This would also explain phenomena such as attainment of limiting density, presence of optimum sintering temperature and the influence of calcination treatments on the sintering behaviour of powders. (author)

Uranium dioxide and beryllium oxide enhanced thermal conductivity nuclear fuel development

International Nuclear Information System (INIS)

Andrade, Antonio Santos; Ferreira, Ricardo Alberto Neto

2007-01-01

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

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

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

International Nuclear Information System (INIS)

Porneuf, A.

1960-05-01

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

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.)

Current state of the Uranium dioxide sintering theory

International Nuclear Information System (INIS)

Baranov, V.; Devyatko, Y.; Tenishev, A.; Khlunov, A.; Khomyakov, O.

2011-01-01

The basic approaches to the description of the ceramics sintering phenomenon are considered. It is established that diffusive sintering models incorrectly describe an intermediate stage of this process. The physical model of sintering, considering the substance plastic flow of pressing under the influence of internal stress forces and capillary forces, as the basic mechanism defining the shrinkage of sintering oxide nuclear fuel, is offered. (authors)

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

International Nuclear Information System (INIS)

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

1977-01-01

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

Preparation, sintering and leaching of optimized uranium thorium dioxides

International Nuclear Information System (INIS)

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

2009-01-01

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

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

CERN Document Server

American Society for Testing and Materials. Philadelphia

2001-01-01

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

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.

A METHOD OF PREPARING URANIUM DIOXIDE

Science.gov (United States)

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

1963-12-17

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

The cohesive energy of uranium dioxide and thorium dioxide

International Nuclear Information System (INIS)

Childs, B.G.

1958-08-01

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

Thermodynamic and transport properties of uranium dioxide and related phases

International Nuclear Information System (INIS)

1965-01-01

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

Production of uranium dioxide

International Nuclear Information System (INIS)

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

1977-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1960-05-15

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

Sintering of uranium oxide of high specific surface area

International Nuclear Information System (INIS)

Bel, Alain; Francois, Bernard; Delmas, Roger; Caillat, Roger

1959-01-01

The extent to which a uranium oxide powder deriving from ammonium uranate or uranium peroxide lends itself to the sintering process depends largely on its specific surface area. When this is greater than 5 m 2 / g there is an optimum temperature for sintering in hydrogen. This temperature becomes less as the specific area of the powder is greater. Reprint of a paper published in Comptes rendus des seances de l'Academie des Sciences, t. 249, p. 1045-1047, sitting of 21 September 1959 [fr

Predictor of regulation of uranium dioxide powder pressing process

International Nuclear Information System (INIS)

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

2007-01-01

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

Immobilization of Uranium Silicide in Sintered Iron-Phosphate Glass

International Nuclear Information System (INIS)

Mateos, Patricia; Russo, Diego; Rodriguez, Diego; Heredia, A; Sanfilippo, M.; Sterba, Mario

2003-01-01

This work is a continuation of a previous one performed in vitrification of uranium silicide in borosilicate and iron-silicate glasses, by sintering.We present the results obtained with an iron-phosphate glass developed at our laboratory and we compare this results with those obtained with the above mentioned glasses. The main objective was to develop a method as simple as possible, so as to get a monolithic glass block with the appropriate properties to be disposed in a deep geological repository.The thermal transformation of the uranium silicide was characterized by DTA/TG analysis and X-ray diffraction.We determined the evolution of the crystalline phases and the change in weight.Calcined uranium silicide was mixed with natural U 3 O 8 , the amount of U 3 O 8 was calculated to simulate an isotopic dilution of 4%.This material was mixed with powdered iron-phosphate glass (in wt.%: 64,9 P 2 O 5 ; 22,7 Fe 2 O 3 ; 8,1 Al 2 O 3 ; 4,3 Na 2 O) in different proportions (in wt%): 7%, 10% y 15%.The powders were pressed and sintered at temperatures between 585 y 670 °C. Samples of the sintered pellet were prepared for the lixiviation tests (MCC-1P: monolithic samples; deionised water; 90° C; 7, 14 and 28 days).The samples showed a quite good durability (0,6 g.m -2 .day -1 ), similar to borosilicate glasses.The microstructure of the glass samples showed that the uranium particles are much better integrated to the glass matrix in the iron-phosphate glasses than in the borosilicate or iron-silicate glasses.We can conclude that the sintered product obtained could be a good alternative for the immobilization of nuclear wastes with high content of uranium, as the ones arising from the conditioning of research reactors spent fuels

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1960-05-15

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

Uranium dioxide electrolysis

Science.gov (United States)

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

2009-12-29

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

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

International Nuclear Information System (INIS)

2008-01-01

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

Sintered nuclear fuel and method of preparing same

International Nuclear Information System (INIS)

Abate-Daga, G.; Amato, I.

1975-01-01

A description is given of a method of preparing a nuclear fuel containing a consumable nuclear poison uniformly distributed therein in the form of coated micro-spheres of between 10 and 2,000 microns diameter, consisting in preparing sintered micro-spheres of the consumable poison, covering those micro-spheres with a protective coating and incorporating the coated micro-spheres into uranium dioxide powder, followed by sintering

Uranium migration in spark plasma sintered W/UO2 CERMETS

Science.gov (United States)

Tucker, Dennis S.; Wu, Yaqiao; Burns, Jatuporn

2018-03-01

W/UO2 CERMET samples were sintered in a Spark Plasma Sintering (SPS) furnace at various temperature under vacuum and pressure. High Resolution Transmission Electron Microscopy (HRTEM) with Energy Dispersive Spectroscopy (EDS) was performed on the samples to determine interface structures and uranium diffusion from the UO2 particles into the tungsten matrix. Local Electrode Atom Probe (LEAP) was also performed to determine stoichiometry of the UO2 particles. It was seen that uranium diffused approximately 10-15 nm into the tungsten matrix. This is explained in terms of production of oxygen vacancies and Fick's law of diffusion.

Influence of various manufacturing parameters on some characteristics of UO2 powders and their sintering behaviour

International Nuclear Information System (INIS)

Mintz, M.H.; Vaknin, Sh.; Kremener, A.; Hadari, Z.

1977-02-01

Various parameters in the process of manufacturing uranium dioxide are examined and their influence on the characteristics and sintering behaviour of the powders obtained established. In addition some correlations between the powder aggregates microstructure and their adhesion properties and sintering behaviour are indicated. Shrinkage during the sintering process is also discussed

Behaviour of uranium dioxide in liquid nitrogen tetraoxide

International Nuclear Information System (INIS)

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

1983-01-01

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

Internal friction in uranium dioxide

International Nuclear Information System (INIS)

Paulin Filho, Pedro Iris

1979-01-01

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

Sintering of uranium dioxide obtained by continuous precipitation of AUC

International Nuclear Information System (INIS)

Amaya, C.D.; Sterba, M.E.; Russo, D.O.

1993-01-01

The Nuclear Materials Division in Bariloche Atomic Center evaluates the ceramic behaviour of UO 2 powders obtained from continuously precipitated and reduced AUC (Ammonium Uranyl Tri Carbonate). An analysis is made of powder characteristics (particle morphology and size distribution and specific area) on behaviour of UO 2 during sintering (compaction, sintering, pore and grain microstructure, etc.). 1 ref

Low density, variation in sintered density and high nitrogen in uranium dioxide

International Nuclear Information System (INIS)

Balakrishna, Palanki; Murty, B.N.; Anuradha, M.; Nageshwara Rao, P.; Jayaraj, R.N.; Ganguly, C.

2000-01-01

Low sintered density and density variation in sintered UO 2 were found to have been caused by non uniformity in the granule feed characteristics to the compacting press. The nitrogen impurity content of sintered UO 2 was found to be sintering furnace related and associated with low sintered density pellets. The problems of low density, variation in sintered density and high nitrogen could be solved by the replacement of the prevailing four punch precompaction by a single punch process; by the introduction of a vibro-sieve for the separation of fine particles from the press feed granules; by innovation in the powder feed shoe design for simultaneous and uniform dispensing of powder in all the die holes; by increasing the final compaction pressure and by modifying the gas flows and preheat temperature in the sintering furnace. (author)

Investigation of transformation of uranium hexafluoride into dioxide

International Nuclear Information System (INIS)

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

1982-01-01

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

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

International Nuclear Information System (INIS)

Schrire, D.I.

1983-12-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Sintered ceramics having controlled density and porosity

International Nuclear Information System (INIS)

Brassfield, H.C.; DeHollander, W.R.; Nivas, Y.

1980-01-01

A new method was developed for sintering ceramic uranium dioxide powders, in which ammonium oxalate is admixed with the powder prior to being pressed into a cylindrical green body, so that the end-point density of the final nuclear-reactor fuel product can be controlled. When the green body is heated, the ammonium oxalate decomposes and leaves discrete porosity in the sintered body, which corresponds to the ammonium oxalate regions in the green body. Thus the end-point density of the sintered body is a function of the amount of ammonium oxalate added. The final density of the sintered product is about 90-97% of the theoretical. The addition of ammonium oxalate also allows control of the pore size and distribution throughout the fuel. The process leaves substantially no impurities in the sintered strucuture. (DN)

Contribution to the study of the sintering of uranium oxide; Contribution a l'etude du frittage de l'oxyde d'uranium

Energy Technology Data Exchange (ETDEWEB)

Bel, A; Carteret, Y [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The sintering ofnium oxide has been considered and the following factors have been particularly taken in consideration: - the particle size and the particles in shape of the initial powder, - the specific area of the initial powder, - the chemical composition of the oxide, - and the medium in which the sintering was carried out. A method of sintering uranium oxide on semi-industrial scale is presented. (author)Fren. [French] On xamine l'influence de differents facteurs sur le frittage de l'oxyde d'uranium. Sont particulierement prises en consideration: - la taille et la forme des grains de la poudre initiale, - la surface specifique de la poudre initiale, - la composition chimique de l'oxyde, - ainsi que la nature de l'atmosphere durant le frittage. D'autre part, une technique de frittage de l'oxyde d'uranium a l'echelle semi-industrielle est presentee. (auteur)

Yellow cake to ceramic uranium dioxide

International Nuclear Information System (INIS)

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

1983-01-01

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

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

International Nuclear Information System (INIS)

Ponelis, A.A.

1989-01-01

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

Thermal properties of nonstoichiometry uranium dioxide

Science.gov (United States)

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

2016-04-01

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

Sintering with a chemical reaction as applied to uranium monocarbide

International Nuclear Information System (INIS)

Accary, A.; Caillat, R.

1960-01-01

The present paper provides a survey of different investigations whose aim was the preparation and fabrication of uranium monocarbide for nuclear use. If a chemical reaction takes place in the sample during the sintering operation, it may be expected that the atom rearrangements involved in this reaction should favour the sintering process and thereby lower the temperature needed to yield a body of a given density. With this hypothesis in mind, the following methods have been studied: - Sintering of U-C mixtures; - Sintering of UO 2 -C mixtures; - Hot pressing of U-C mixtures; - Extrusion of U-C mixtures. To generalize our result, it could be said that a chemical reaction does not lead to high densification, if one depends on a simple contact between discrete particles. On the contrary, a chemical reaction can help sintering if, as our hot pressing experiments shows, the densification can be achieved prior to the reaction. (author) [fr

Synthesis, sintering properties and thermal conductivity of uranium carbonitrides

International Nuclear Information System (INIS)

Wolters, R.A.M.

1978-01-01

An introduction to the applications and chemistry of uranium carbonitrides is given including the potential use as a nuclear fuel. The powder synthesis of UC, UN and mixtures of UC and UN by a cyclic process is described. The correlation between the composition ratio UN/(UC+UN) in the final product and the parameters of the process is only determined qualitatively. Batch synthesis of a powder does not lead to an increase of the content of metallic impurities and oxygen. The impurity level is determined by that of the starting uranium metal and the thermal conductivity of the sintered compacts of uranium carbonitrides are determined via the measurement of the thermal diffusivity at 1100-1700 K. (Auth.)

Uranium tetrafluoride production via dioxide by wet process

International Nuclear Information System (INIS)

Aquino, A.R. de.

1988-01-01

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

Study of automatic boat loading unit and horizontal sintering process of uranium dioxide pellet

International Nuclear Information System (INIS)

He Zhongjing; Chen Yu; Yao Dengfeng; Wang Youliang; Shu Binhua; Wu Genjiu

2014-01-01

Sintering process is a key process for the manufacture of nuclear fuel UO_2 pellet. In our factory, the continuous high temperature sintering furnace is used for sintering process. During the sintering of green pellets, the furnace, the boat and the accumulation way can influence the quality of the final product. In this text, on the basis of early process research, The automatic loading boat Unit and horizontal sintering process is studied successively. The results show that the physical and chemical properties of the products manufactured by automatic loading boat unit and horizontal sintering process can meet the technique requirements completely, and this system is reliable and continuous. (authors)

Investigation of control conditions of uranium dioxide pellets sinterability through microspheres

International Nuclear Information System (INIS)

Assis, Gino de.

1996-01-01

Promotion or inhibition of ceramic powders sinterability, the decisive question in ceramic processing is approached in this dissertation. Each high density microsphere has been considered as a solid inclusion in a low density microspheres matrix, generating big pores. Such pores make it difficult for the pellets density due the fact that they are difficult to be eliminated. A master mixture, allowing the pellet densification in the projected range has been reached. Batches of microspheres have been observed sometimes with high apparent density and sometimes with low apparent density. This apparent density variation was attributed to changing the oxygen partial pressure during calcination under air atmosphere. It is evident that the control of the apparent density of the microspheres needs a further research in order to adjust the sinterability of the microspheres on the desired level.It was demonstrated that the produced microspheres do not have impurities levels that can promote its sinterability or avoid their use in nuclear area

Modifier free supercritical fluid extraction of uranium from sintered UO2, soil and ore samples

International Nuclear Information System (INIS)

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

2011-01-01

Direct extraction of uranium from different samples viz. sintered UO 2 , soil and ores was carried out by modifier free supercritical fluid using tri-n-butyl phosphate-nitric acid (TBP-HNO 3 ) adduct as extractant. These studies showed that pre-equilibration with more concentrated nitric acid helps in better dissolution and extraction of uranium from sintered UO 2 samples. Modifier free supercritical fluid extraction appears attractive with respect to minimization of secondary wastes. This method resulted 80-100% extraction of uranium from different soil/ore samples. The results were confirmed by performing neutron activation analysis of original (before extraction) and residue (after extraction) samples. (author)

Improvement of cesium retention in uranium dioxide by additional phases

International Nuclear Information System (INIS)

Gamaury Dubois, S.

1995-01-01

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

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)

Metallographic preparation of sintered oxides, carbides and nitrides of uranium and plutonium

International Nuclear Information System (INIS)

Martin, A.; Arles, L.

1967-12-01

We describe the methods of polishing, attack and coloring used at the section of plutonium base ceramics studies. These methods have stood the test of experience on the uranium and plutonium carbides, nitrides and carbonitrides as well on the mixed uranium and plutonium oxides. These methods have been particularly adapted to fit to the low dense and sintered samples [fr

Thermal conductivity of uranium dioxide

International Nuclear Information System (INIS)

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

1993-01-01

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

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

International Nuclear Information System (INIS)

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

1984-01-01

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

Density determination of sintered ceramic nuclear fuel materials

International Nuclear Information System (INIS)

Landspersky, H.; Medek, J.

1980-01-01

The feasibility was tested of using solids for pycnometric determination of the density of uranium dioxide-based sintered ceramic fuel materials manufactured by the sol-gel method in the shape of spherical particles of 0.7 to 1.0 mm in size and of particles smaller than 200 μm. For fine particles, this is the only usable method of determining their density which is a very important parameter of the fine fraction when it is employed for the manufacture of fuel elements by vibration compacting. The method consists in compacting a mixture of pycnometric material and dispersed particles of uranium dioxide, determining the size and weight of the compact, and in calculating the density of the material measured from the weight of the oxide sample in the mixture. (author)

Immobilization of chlorine dioxide modified cells for uranium absorption

International Nuclear Information System (INIS)

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

2014-01-01

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

Uranium dioxide calcining apparatus

International Nuclear Information System (INIS)

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

1978-01-01

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

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

International Nuclear Information System (INIS)

Peres, V.

1994-06-01

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

Sintering of uranium dioxide pellets (UO2) in an oxidizing atmosphere (C O2)

International Nuclear Information System (INIS)

Santos, G.R.T.

1992-01-01

This work consists in the study of the sintering process of U O 2 pellets in an oxidizing atmosphere. Sintering tests were performed in an CO 2 atmosphere and the influence of temperature and time on the pellets density and microstructure were verified. The results obtained were compared to those from the conventional sintering process and its efficiency was confirmed. (author)

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

International Nuclear Information System (INIS)

1994-01-01

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

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

International Nuclear Information System (INIS)

Leitao Junior, C.B.

1992-01-01

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

Method of manufacturing sintered nuclear fuel

International Nuclear Information System (INIS)

Watarumi, Kazutoshi.

1984-01-01

Purpose: To obtain composite pellets with an improved strength. Method: A core mainly composed of fuel materials is previously prepared, embedded into the central portion of a pellet, silted therearound with cladding material, and then pressmolded and sintered. For instance, a rugby-ball like core body with the maximum outer diameter of 6 mm and the height of 6 mm is made by compressive molding with uranium dioxide powder, then coating material comprising the same powder incorporated with 0.1 % by weight of SiC fibers is filled around the core body, which is molded into a composite pellet by means of pressing and then sintered at 1600 0 C, to obtain a sintered pellet of 93.5 % theoretical density. As the result of the compression test for the pellet, it showed a strength greater by 15 % than that of the similar mono-layer pellet. (Kamimura, M.)

Uranium tetracyclopentadienyl interaction with carbon oxide and dioxide

International Nuclear Information System (INIS)

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

1983-01-01

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

The preparation of uranium tetrafluoride from dioxide by aqueous way

International Nuclear Information System (INIS)

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

1990-01-01

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

Contribution to the study of uranium dioxide aqueous corrosion mechanisms

International Nuclear Information System (INIS)

Gallien, J.-P.

1994-01-01

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

Study of Physical modifications induced by chromium doping of uranium dioxide

International Nuclear Information System (INIS)

Fraczkiewicz, M.

2010-01-01

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

Dissolution experiments of unirradiated uranium dioxide pellets

International Nuclear Information System (INIS)

Ollila, K.

1985-01-01

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

Manufacture of uranium dioxide powder

International Nuclear Information System (INIS)

Becker, M.

1976-01-01

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

Fluorophotometric determination of uranium: an automated sintering furnace and factors affecting precision

International Nuclear Information System (INIS)

Strain, J.E.

1978-07-01

The fusion furnace consists of four individually controlled, slotted-tube furnaces that automatically dry, sinter and anneal the fluoride or carbonate pellet used in the fluorometric determination of uranium. The furnace operates in air and prepares approximately 90 pellets per hour for fluorometric measurement. The factors that were thought to affect the precision of the method were investigated. The two factors that seem to be the most influential are (1) the manner in which the sample is loaded onto the pellet; and (2) the surface characteristics of the platinum dish in which the pellet is sintered and measured fluorometrically

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Study of non stoichiometric uranium dioxide samples (UO2)

International Nuclear Information System (INIS)

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

1999-01-01

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

Sintering with a chemical reaction as applied to uranium monocarbide; Frittage-reaction dans le cas du monocarbure d'uranium

Energy Technology Data Exchange (ETDEWEB)

Accary, A; Caillat, R [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

The present paper provides a survey of different investigations whose aim was the preparation and fabrication of uranium monocarbide for nuclear use. If a chemical reaction takes place in the sample during the sintering operation, it may be expected that the atom rearrangements involved in this reaction should favour the sintering process and thereby lower the temperature needed to yield a body of a given density. With this hypothesis in mind, the following methods have been studied: - Sintering of U-C mixtures; - Sintering of UO{sub 2}-C mixtures; - Hot pressing of U-C mixtures; - Extrusion of U-C mixtures. To generalize our result, it could be said that a chemical reaction does not lead to high densification, if one depends on a simple contact between discrete particles. On the contrary, a chemical reaction can help sintering if, as our hot pressing experiments shows, the densification can be achieved prior to the reaction. (author) [French] Le present article resume les etudes faites pour le compte du Commissariat a l'Energie Atomique dans le but de preparer du monocarbure d'uranium pour usage nucleaire. Si, en meme temps que l'on fritte une poudre, celle-ci est le siege d'une reaction chimique, on peut s'attendre a ce que le rearrangement atomique d'une reaction chimique favorise le frittage et, ainsi abaisse la temperature de travail necessaire pour obtenir une densite donnee. Nous avons etudie les methodes suivantes: - frittage des melanges U-C; - frittage des melanges UO{sub 2}-C; - frittage sous charge des melanges U-C; - filage des melanges U-C. Nos resultats montrent qu'une reaction chimique en cours de frittage ne conduit pas a un produit de haute densite si on opere sur un melange de poudres. Par contre, elle permet d'atteindre de hautes densites si la densification peut etre obtenue avant la reaction chimique. (auteur)

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)

Nuclear fuel elements

International Nuclear Information System (INIS)

Obara, Hiroshi.

1981-01-01

Purpose: To suppress iodine release thereby prevent stress corrosion cracks in fuel cans by dispersing ferrous oxide at the outer periphery of sintered uranium dioxide pellets filled and sealed within zirconium alloy fuel cans of fuel elements. Constitution: Sintered uranium dioxide pellets to be filled and sealed within a zirconium alloy fuel can are prepared either by mixing ferric oxide powder in uranium dioxide powder, sintering and then reducing at low temperature or by mixing iron powder in uranium dioxide powder, sintering and then oxidizing at low temperature. In this way, ferrous oxide is dispersed on the outer periphery of the sintered uranium dioxide pellets to convert corrosive fission products iodine into iron iodide, whereby the iodine release is suppressed and the stress corrosion cracks can be prevented in the fuel can. (Moriyama, K.)

Improved ionic model of liquid uranium dioxide

NARCIS (Netherlands)

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

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

Pyrochemical reduction of uranium dioxide and plutonium dioxide by lithium metal

International Nuclear Information System (INIS)

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

2002-01-01

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

Certification of a uranium dioxide reference material for chemical analyses

International Nuclear Information System (INIS)

Le Duigou, Y.

1984-01-01

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

Determination of gas residues in uranium dioxide pellets

International Nuclear Information System (INIS)

Riella, H.G.

1978-01-01

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

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

Science.gov (United States)

Grisaffe, Salvatore J.; Caves, Robert M.

1964-01-01

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

Determination of the stoichiometric ratio uranium dioxide samples

International Nuclear Information System (INIS)

Moura, Sergio Carvalho

1999-01-01

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

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

CERN Document Server

American Society for Testing and Materials. Philadelphia

1999-01-01

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

Uranium dioxide calcining apparatus and method

International Nuclear Information System (INIS)

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

1978-01-01

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

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

International Nuclear Information System (INIS)

Danroc, J.

1982-09-01

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

Grain growth in uranium nitride prepared by spark plasma sintering

Science.gov (United States)

Johnson, Kyle D.; Lopes, Denise Adorno

2018-05-01

Uranium mononitride (UN) has long been considered a potential high density, high performance fuel candidate for light water reactor (LWR) and fast reactor (FR) applications. However, deployability of this fuel has been limited by the notable resistance to sintering and subsequent difficulty in producing a desirable microstructure, the high costs associated with 15N enrichment, as well as the known proclivity to oxidation and interaction with steam. In this study, the stimulation of grain growth in UN pellets sintered using SPS has been investigated. The results reveal that by using SPS and controlling temperature, time, and holding pressure, grain growth can be stimulated and controlled to produce a material featuring both a desired porosity and grain size, at least within the range of interest for nuclear fuel candidates. Grain sizes up to 31 μm were obtained using temperatures of 1650 °C and hold times of 15 min. Evaluation by EBSD reveal grain rotation and coalescence as the dominant mechanism in grain growth, which is suppressed by the application of higher external pressure. Moreover, complete closure of the porosity of the material was observed at relative densities of 96% TD, resulting in a material with sufficient porosity to accommodate LWR burnup. These results indicate that a method exists for the economic fabrication of an 15N-bearing uranium mononitride fuel with favorable microstructural characteristics compatible with use in a light water-cooled nuclear reactor.

Method of preparing uranium nitride or uranium carbonitride bodies

International Nuclear Information System (INIS)

Wilhelm, H.A.; McClusky, J.K.

1976-01-01

Sintered uranium nitride or uranium carbonitride bodies having a controlled final carbon-to-uranium ratio are prepared, in an essentially continuous process, from U 3 O 8 and carbon by varying the weight ratio of carbon to U 3 O 8 in the feed mixture, which is compressed into a green body and sintered in a continuous heating process under various controlled atmospheric conditions to prepare the sintered bodies. 6 claims, no drawings

SEM hot stage sintering of UO2

International Nuclear Information System (INIS)

Miller, D.J.

1976-06-01

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

Procedure for the obtainment of ammonium uranyl-tricarbonate suitable for the preparation of sinterable UO2

International Nuclear Information System (INIS)

Anasco, Roberto; Amendolara, M.M.; De La Fuente, M.; Gonzalez, A.G.; La Gamma de Batistoni, A.M.; Garcia, E.

1980-01-01

Experiments carried out to obtain Ammonium Uranyl-Tricarbonate (AUC) of nuclear purity and with the appropriate physical characteristics to serve as an intermediate stage for the obtainment of sinterable Uranium Dioxide are described. AUC was obtained by precipitation with gaseous ammonium and carbon dioxide from aqueous solutions re-circulation, controlling, in both cases, the flow of the reactive gases, the pH and the temperature. The analyzed working conditions are described, giving also the results from the distribution of the particle size and morphology of the crystals. (M.E.L.) [es

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

International Nuclear Information System (INIS)

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

1979-01-01

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

Design of a Uranium Dioxide Spheroidization System

Science.gov (United States)

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

2013-01-01

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

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

International Nuclear Information System (INIS)

Tolk, A.; Lingerak, W.A.

1970-09-01

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

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

International Nuclear Information System (INIS)

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

1981-01-01

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

Guideline tests on the corrosion of sintered uranium oxide by lead and sodium about 450 deg. C

International Nuclear Information System (INIS)

Portnoff, A.; Pointud, R.

1958-05-01

Within the frame of the investigation of behaviour of a fuel element (such as UO 2 ) under irradiation, the authors report the study of the physical-chemical action of the coolant at different temperatures on the body to be irradiated. Thus, sintered uranium oxide has been submitted to corrosion by lead and by sodium during 250 hours at temperatures between 400 and 500 C. The physical characteristics of the UO 2 powder and of different sintered UO 2 pellets produced from this powder under different sintering processes are indicated, as well as the results of a spectrographic analysis of the sintered UO 2 . Corrosion devices, treatments and obtained results are reported for corrosion by lead and by sodium. In the latter case, extraction processes are discussed (using butyl alcohol, or melting in vaseline oil)

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

International Nuclear Information System (INIS)

Santon, J.P.

1964-09-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

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)

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

International Nuclear Information System (INIS)

Fraczkiewicz, M.; Baldinozzi, G.

2008-01-01

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

Process for the preparation of uranium dioxide

International Nuclear Information System (INIS)

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

1977-01-01

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

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

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

International Nuclear Information System (INIS)

Manojlovic, Lj.

1962-12-01

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

SULPHUR DIOXIDE LEACHING OF URANIUM CONTAINING MATERIAL

Science.gov (United States)

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

1958-12-01

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Advances in heterogeneous autocatalytic reactions applied to uranium dissolution - 5317

International Nuclear Information System (INIS)

Marc, P.; Magnaldo, A.; Godard, J.; Schaer, E.

2015-01-01

Dissolution and the solubilization of the chemical elements is a milestone of the head-end of hydrometallurgical processes. When dissolving spent nuclear fuels, additional constraints are added due to the permanent need to strictly control and limit the hold-up. Thus the need for kinetic modeling concerning the dissolution of spent nuclear fuels in nitric acid. This study aims at better understanding the chemical and physical-chemical phenomena of uranium dioxide dissolution reactions in nitric medium. It has been documented that the nitric acid attack of sintering-manufactured uranium dioxide solids occurs through preferential attack sites. This non uniform attack leads to the development of cracks in the solids. Optical microscopy observations show that in some cases, the development of these cracks can lead to the solid cleavage. In this case, we show that the dissolution of the detached fragments is much slower than the time required for the complete cleavage of the solid. These points motivated the measurements of dissolution kinetics using optical microscopy and image processing. A comparison of the measured kinetics with the diffusion kinetics by the mean of the external resistance fraction allows discriminating between measured kinetics corresponding to the chemical reaction or mass-transport limitation. This capability to measure, for the very first time, the 'true' chemical kinetics of the reaction has enabled the confirmation of the highly autocatalytic nature of the reaction, and first evaluation of the constants of the chemical reactions kinetic laws. These data are fundamental to set the kinetic parameters of the chemical reactions in a future model of the dissolution of uranium dioxide sintered pellets. (authors)

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

Science.gov (United States)

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

2018-03-01

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

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

Directory of Open Access Journals (Sweden)

Marc Philippe

2018-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Evaluation of Hydrothermally Synthesized Uranium Dioxide for Novel Semiconductor Applications

Science.gov (United States)

2016-08-29

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

The creation of a uranium oxide industry, from the laboratory stage to a pilot plant (1961)

International Nuclear Information System (INIS)

Caillat, R.; Delange, M.; Sauteron, J.

1961-01-01

The qualities of uranium oxide, in particular its good in-pile characteristics and its resistance to corrosion by the usual heat-exchange fluids, have led to this material being chose at the present time as a nuclear fuel in many power reactors, either planned or under construction. A great effort has been made these last few years in France in studying processes for transforming powdered uranium oxide into a dense material with satisfactory behaviour in a neutron flux. The laboratories at Saclay have studied the physico-chemical features of the phenomena accompanying the calcination of uranium peroxide or ammonium uranate to give uranium trioxide, and the subsequent reduction of the latter to dioxide as well as the sintering of the powders obtained. This work has made it possible on one hand to prepare powder of known specific surface area, and on the other to show the overriding influence of this factor, all other things being equal, on the behaviour of powders during sintering in a hydrogen atmosphere. The work has led to defining two methods for sintering stoichiometric uranium oxide of high density. The technological study of the preparation of the powder and its industrial production are carried out at the plant of Le Bouchet which produces at the moment powders of known characteristics suitable for sintering in hydrogen at 1650 deg. C without prior grinding. The industrial sintering is carried out by the Compagnie industrielle des Combustibles Atomiques Frittes who has set up a pilot plant having a capacity of 25 metric tons/year, for the Commissariat l'Energie Atomique and has been operating this plant since May 1958. This plant is presented by a film entitled 'uranium oxide'. (author) [fr

Polarographic determination of uranium dioxide stoichiometry

International Nuclear Information System (INIS)

Viguie, J.; Uny, G.

1966-10-01

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

Fabrication and testing of the sintered ceramic UO2 fuel - I - III, Part III - testing of sintered uranium dioxide properties dependent on the fabrication procedure

International Nuclear Information System (INIS)

Novakovic, M.; Ristic, M.M.

1961-12-01

The objective of this task was testing the influence of some parameters on the properties of sintered UO 2 . The influence of parameters tested were as follows: adhesives; pressure in the pressing procedure; temperature of sintering of the UO 2 powder. Other parameters were chosen according to the theoretical study. Sintering was done in argon atmosphere. Characterization of the UO 2 powder was performed meaning determining the needed chemical, physical and physico-chemical properties. Some new methods were developed within this task: SET method for measuring the specific surfaces, DTA, TGA, high-temperature torsion

Methods for oxygen/uranium ratio determination in substoichiometric uranium dioxide

International Nuclear Information System (INIS)

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

1994-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Gamaury Dubois, S

1995-09-19

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

Process for continuous production of metallic uranium and uranium alloys

Science.gov (United States)

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

1995-01-01

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

Process for continuous production of metallic uranium and uranium alloys

Science.gov (United States)

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

1995-06-06

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

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

Science.gov (United States)

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

2018-02-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-09-28

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

Determination of carbon chlorine and fluorine in uranium dioxide

International Nuclear Information System (INIS)

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

1983-01-01

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

Immobilization of Uranium Silicides in Sintered Glass

International Nuclear Information System (INIS)

Mateos, P.; Russo, D.O.; Heredia, A.D.; Sanfilippo, M.

2003-01-01

High activity nuclear spent fuels vitrification by fusion is a well known technology which has industrial scale in France, England, Japan, EEUU. Borosilicates glasses are used in this process.Sintered glasses are an alternative to the immobilization task in which there is also a wide experience around the world.The available technics are: cold pressing and sintering , hot-pressing and hot isostatic pressing.This work compares Borosilicates and Iron silicates sintered glasses behaviour when different ammounts of nuclear simulated waste is added

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

Energy Technology Data Exchange (ETDEWEB)

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

1961-12-15

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

Surface Characterization and Electrochemical Oxidation of Metal Doped Uranium Dioxide

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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

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

International Nuclear Information System (INIS)

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

1988-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Hingant, N

2008-12-15

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

Process for producing uranium oxide rich compositions from uranium hexafluoride

International Nuclear Information System (INIS)

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

1978-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-09-15

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

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

International Nuclear Information System (INIS)

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

1977-08-01

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

Fluorination reaction uranium dioxide by fluorine

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

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

1989-01-01

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

Note on measurement of thermal conductivity of sintered uranium dioxide; Note relative a la mesure de la conductivite thermique du bioxyde d'uranium fritte

Energy Technology Data Exchange (ETDEWEB)

Englander, M

1951-06-01

Thermal conductivity of sintered UO{sub 2} was determined by measuring the quantity of heat having passed in unit time through a plate of given dimensions when a certain temperature difference was being maintained at the faces of the plate. Specimens, about 10 and 40 mm thick and about 65 mm in diameter, were heated electrically, the temperature of both faces being measured by means of iron-constantan thermocouples. The accuracy of the device in its present shape is not high, the relative error being {approx} 15%. The thermal conductivity of sintered UO{sub 2} in the temperature range 20 to 250 deg. C was found to be about 9 x 10{sup -3} cgs units. (author)

Boron nitride coated uranium dioxide and uranium dioxide-gadolinium oxide fuels

Energy Technology Data Exchange (ETDEWEB)

Gunduz, G [Department of Chemical Engineering, Middle East Technical Univ., Ankara (Turkey); Uslu, I; Tore, C; Tanker, E [Turkiye Atom Enerjisi Kurumu, Ankara (Turkey)

1997-08-01

Pure Urania and Urania-gadolinia (5 and 10%) fuels were produced by sol-gel technique. The sintered fuel pellets were then coated with boron nitride (BN). This is achieved through chemical vapor deposition (CVD) using boron trichloride and ammonia. The coated samples were sintered at 1600 K. The analyses under scanning electron microscope (SEM) showed a variety of BN structures, mainly platelike and rodlike structures were observed. Burnup calculations by using WIMSD4 showed that BN coated and gadolinia containing fuels have larger burnups than other fuels. The calculations were repeated at different pitch distances. The change of the radius of the fuel pellet or the moderator/fuel ratio showed that BN coated fuel gives the highest burnups at the present design values of a PWR. Key words: burnable absorber, boron nitride, gadolinia, CVT, nuclear fuel. (author). 32 refs, 14 figs.

Boron nitride coated uranium dioxide and uranium dioxide-gadolinium oxide fuels

International Nuclear Information System (INIS)

Gunduz, G.; Uslu, I.; Tore, C.; Tanker, E.

1997-01-01

Pure Urania and Urania-gadolinia (5 and 10%) fuels were produced by sol-gel technique. The sintered fuel pellets were then coated with boron nitride (BN). This is achieved through chemical vapor deposition (CVD) using boron trichloride and ammonia. The coated samples were sintered at 1600 K. The analyses under scanning electron microscope (SEM) showed a variety of BN structures, mainly platelike and rodlike structures were observed. Burnup calculations by using WIMSD4 showed that BN coated and gadolinia containing fuels have larger burnups than other fuels. The calculations were repeated at different pitch distances. The change of the radius of the fuel pellet or the moderator/fuel ratio showed that BN coated fuel gives the highest burnups at the present design values of a PWR. Key words: burnable absorber, boron nitride, gadolinia, CVT, nuclear fuel. (author). 32 refs, 14 figs

Evaluation of uranium dioxide thermal conductivity using molecular dynamics simulations

International Nuclear Information System (INIS)

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

2014-01-01

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

Sintering of dioxide pellets in an oxidizing atmosphere (CO2)

International Nuclear Information System (INIS)

Santos, G.R.T.

1992-01-01

This work consists in the study of the sintering process of U O 2 pellets in an oxidizing atmosphere. Sintering tests were performed in an CO 2 atmosphere and the influence of temperature and time on the pellets density and microstructure were verified. The results obtained were compared to those from the conventional sintering process and its efficiency was confirmed. (author)

Production of sized particles of uranium oxides and uranium oxyfluorides

International Nuclear Information System (INIS)

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

1976-01-01

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

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

International Nuclear Information System (INIS)

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

1977-01-01

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

On the nature of the phase transition in uranium dioxide

Science.gov (United States)

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

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

Compaction and sintering of nickel powder used encapsulation of irradiation targets

Energy Technology Data Exchange (ETDEWEB)

Miyano, Rosana S.L.; Guimaraes, Raquel R.F.L.; Rossi, Jesualdo L., E-mail: rosatac@gmail.com, E-mail: raquel.lucchesi@icloud.com, E-mail: jelrossi@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (CCTM/IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Ciencia e Tecnologia de Materiais; Wendhausen, Paulo A.P.; Evangelista, Leandro L., E-mail: paulo.wendhausen@ufsc.br, E-mail: leandro.materiais@gmail.com [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil). Laboratorio de Materiais

2015-07-01

The objective of this study was to develop an alternative way to produce targets for irradiation containing uranium, for the pair of {sup 99}Mo production {sup 99m}Tc radionuclide. These targets were obtained by powder metallurgy, the compact serving as means for encapsulation a uranium cylinder to be irradiated. The targets were compacted in an axial hydraulic press applying different pressures up to 800 MPa. The sintering temperature was 600 °C in hydrogen atmosphere and it was used two sintering cycles, one for 4 h and the for 4 h plus 8 h time. The nickel powder was of high purity, that in order to provide the sealing of the fissile content within the compacted. The bulk density of compacted was evaluated by the method geometric. The porosity was measured by mercury porosimetry technique. The microstructure was investigated by optical microscopy. The results obtained with sintering powders involving confirm the feasibility of achieving a casing for uranium targets. (author)

Compaction and sintering of nickel powder used encapsulation of irradiation targets

International Nuclear Information System (INIS)

Miyano, Rosana S.L.; Guimaraes, Raquel R.F.L.; Rossi, Jesualdo L.; Wendhausen, Paulo A.P.; Evangelista, Leandro L.

2015-01-01

The objective of this study was to develop an alternative way to produce targets for irradiation containing uranium, for the pair of 99 Mo production 99m Tc radionuclide. These targets were obtained by powder metallurgy, the compact serving as means for encapsulation a uranium cylinder to be irradiated. The targets were compacted in an axial hydraulic press applying different pressures up to 800 MPa. The sintering temperature was 600 °C in hydrogen atmosphere and it was used two sintering cycles, one for 4 h and the for 4 h plus 8 h time. The nickel powder was of high purity, that in order to provide the sealing of the fissile content within the compacted. The bulk density of compacted was evaluated by the method geometric. The porosity was measured by mercury porosimetry technique. The microstructure was investigated by optical microscopy. The results obtained with sintering powders involving confirm the feasibility of achieving a casing for uranium targets. (author)

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

International Nuclear Information System (INIS)

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

1987-07-01

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

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

International Nuclear Information System (INIS)

Baque, P.; Leclercq, D.

1964-01-01

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

XAS characterisation of xenon bubbles in uranium dioxide

Energy Technology Data Exchange (ETDEWEB)

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

2008-06-15

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

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

CERN Document Server

American Society for Testing and Materials. Philadelphia

2000-01-01

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

The creation of a uranium oxide industry, from the laboratory stage to a pilot plant (1961); Creation d'une industrie de l'oxyde d'uranium du laboratoire a l'usine pilote (1961)

Energy Technology Data Exchange (ETDEWEB)

Caillat, R; Delange, M; Sauteron, J [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires; Hauser, R [Compagnie Industrielle des Combustibles atomiques frittes (France)

1961-07-01

The qualities of uranium oxide, in particular its good in-pile characteristics and its resistance to corrosion by the usual heat-exchange fluids, have led to this material being chose at the present time as a nuclear fuel in many power reactors, either planned or under construction. A great effort has been made these last few years in France in studying processes for transforming powdered uranium oxide into a dense material with satisfactory behaviour in a neutron flux. The laboratories at Saclay have studied the physico-chemical features of the phenomena accompanying the calcination of uranium peroxide or ammonium uranate to give uranium trioxide, and the subsequent reduction of the latter to dioxide as well as the sintering of the powders obtained. This work has made it possible on one hand to prepare powder of known specific surface area, and on the other to show the overriding influence of this factor, all other things being equal, on the behaviour of powders during sintering in a hydrogen atmosphere. The work has led to defining two methods for sintering stoichiometric uranium oxide of high density. The technological study of the preparation of the powder and its industrial production are carried out at the plant of Le Bouchet which produces at the moment powders of known characteristics suitable for sintering in hydrogen at 1650 deg. C without prior grinding. The industrial sintering is carried out by the Compagnie industrielle des Combustibles Atomiques Frittes who has set up a pilot plant having a capacity of 25 metric tons/year, for the Commissariat l'Energie Atomique and has been operating this plant since May 1958. This plant is presented by a film entitled 'uranium oxide'. (author) [French] Les qualites de l'oxyde d'uranium, en particulier son bon comportement en pile et sa resistance a la corrosion par les fluides caloporteurs habituels, font choisir aujourd'hui ce materiau comme combustible de nombreux reacteurs de puissance en construction ou en

Guideline tests on the corrosion of sintered uranium oxide by lead and sodium about 450 deg. C; Essais d'orientation sur la corrosion de l'oxyde d'uranium fritte par le plomb et le sodium aux environs de 450 deg. C

Energy Technology Data Exchange (ETDEWEB)

Portnoff, A.; Pointud, R.

1958-05-15

Within the frame of the investigation of behaviour of a fuel element (such as UO{sub 2}) under irradiation, the authors report the study of the physical-chemical action of the coolant at different temperatures on the body to be irradiated. Thus, sintered uranium oxide has been submitted to corrosion by lead and by sodium during 250 hours at temperatures between 400 and 500 C. The physical characteristics of the UO{sub 2} powder and of different sintered UO{sub 2} pellets produced from this powder under different sintering processes are indicated, as well as the results of a spectrographic analysis of the sintered UO{sub 2}. Corrosion devices, treatments and obtained results are reported for corrosion by lead and by sodium. In the latter case, extraction processes are discussed (using butyl alcohol, or melting in vaseline oil)

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

Science.gov (United States)

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

2011-11-01

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

Study of the oxidation risks during the sintering of uranium dioxide, and characterization of the excess oxygen; Etude du risque d'oxydation lors du frittage du bioxyde d'uranium et caracterisation de l'oxygene excedentaire

Energy Technology Data Exchange (ETDEWEB)

Conte, M; Brandela, M

1966-05-01

During sintering in reducing atmospheres, UO{sub 2} pellets can be oxidized by gaseous impurities. The effects of temperature cycles, the partial pressure of O{sub 2} and the flow rate of the gas over the pellets were investigated. In these atmospheres, the O{sub 2} partial pressure during sintering is low at high temperatures, as a consequence of the dissociation rate of the combined water, but below 1000 deg C, it can be high enough to result in a noticeable oxidation of the surface of the pellets during cooling. The crystalline phases which can occur have been identified and two methods of detection have been proposed: a micrographic examination after chemical etching and radiocrystallography. (authors) [French] Lors du frittage industriel du bioxyde d'uranium en atmosphere reductrice (hydrogene ou ammoniac dissocie) la presence d'impuretes oxydantes dans l'atmosphere peut provoquer l'oxydation des pastilles d'UO{sub 2}; les auteurs ont etudie les phenomenes en faisant varier le cycle de temperature, la pression partielle d'oxygene introduit dans l'hydrogene, la vitesse de passage du gaz sur les pastilles. Dans les atmospheres considerees la pression partielle d'oxygene au-dessus de l'UO{sub 2} en cours de frittage est faible a temperature elevee car elle resulte de la dissociation de l'eau formee, mais a t < 1000 degrees C elle, peut etre assez importante pour provoquer une oxydation notable de la surface des pastilles lors du refroidissement. Les phases cristallines susceptibles d'etre formees ont ete reperees et deux methodes de detection proposees: la micrographie apres attaque chimique specifique et la radiocristallographie. (auteurs)

Discrimination symbol applying method for sintered nuclear fuel product

International Nuclear Information System (INIS)

Ishizaki, Jin

1998-01-01

The present invention provides a symbol applying method for applying discrimination information such as an enrichment degree on the end face of a sintered nuclear product. Namely, discrimination symbols of information of powders are applied by a sintering aid to the end face of a molded member formed by molding nuclear fuel powders under pressure. Then, the molded product is sintered. The sintering aid comprises aluminum oxide, a mixture of aluminum oxide and silicon dioxide, aluminum hydride or aluminum stearate alone or in admixture. As an applying means of the sintering aid, discrimination symbols of information of powders are drawn by an isostearic acid on the end face of the molded product, and the sintering aid is sprayed thereto, or the sintering aid is applied directly, or the sintering aid is suspended in isostearic acid, and the suspension is applied with a brush. As a result, visible discrimination information can be applied to the sintered member easily. (N.H.)

CALCIUM OXIDE SINTERING IN ATMOSPHERES CONTAINING WATER AND CARBON DIOXIDE

Science.gov (United States)

The paper gives results of measurements of the effects of water vapor and CO2 on the sintering rate of nascent CaO, as a function of partial pressure and temperature using CaO prepared by rapid decomposition of CaCO3 and CA(OH)2. Each gas strongly catalyzed the sintering process ...

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

International Nuclear Information System (INIS)

Desrues, R.; Paidassi, J.

1965-01-01

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

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

International Nuclear Information System (INIS)

Delwaulle, Celine

2011-01-01

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

The creation of a uranium oxide industry, from the laboratory stage to a pilot plant (1961); Creation d'une industrie de l'oxyde d'uranium du laboratoire a l'usine pilote (1961)

Energy Technology Data Exchange (ETDEWEB)

Caillat, R.; Delange, M.; Sauteron, J. [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires; Hauser, R. [Compagnie Industrielle des Combustibles atomiques frittes (France)

1961-07-01

The qualities of uranium oxide, in particular its good in-pile characteristics and its resistance to corrosion by the usual heat-exchange fluids, have led to this material being chose at the present time as a nuclear fuel in many power reactors, either planned or under construction. A great effort has been made these last few years in France in studying processes for transforming powdered uranium oxide into a dense material with satisfactory behaviour in a neutron flux. The laboratories at Saclay have studied the physico-chemical features of the phenomena accompanying the calcination of uranium peroxide or ammonium uranate to give uranium trioxide, and the subsequent reduction of the latter to dioxide as well as the sintering of the powders obtained. This work has made it possible on one hand to prepare powder of known specific surface area, and on the other to show the overriding influence of this factor, all other things being equal, on the behaviour of powders during sintering in a hydrogen atmosphere. The work has led to defining two methods for sintering stoichiometric uranium oxide of high density. The technological study of the preparation of the powder and its industrial production are carried out at the plant of Le Bouchet which produces at the moment powders of known characteristics suitable for sintering in hydrogen at 1650 deg. C without prior grinding. The industrial sintering is carried out by the Compagnie industrielle des Combustibles Atomiques Frittes who has set up a pilot plant having a capacity of 25 metric tons/year, for the Commissariat l'Energie Atomique and has been operating this plant since May 1958. This plant is presented by a film entitled 'uranium oxide'. (author) [French] Les qualites de l'oxyde d'uranium, en particulier son bon comportement en pile et sa resistance a la corrosion par les fluides caloporteurs habituels, font choisir aujourd'hui ce materiau comme combustible de nombreux reacteurs de

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

International Nuclear Information System (INIS)

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

1990-07-01

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

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)

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

International Nuclear Information System (INIS)

Liu Husheng; Li Jun

1997-01-01

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

Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

Science.gov (United States)

Mcdonald, G. E.

1973-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

International Nuclear Information System (INIS)

Schlutig, S.

2001-03-01

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

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

Directory of Open Access Journals (Sweden)

Dąbrowski Ludwik

2014-08-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

DEFF Research Database (Denmark)

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

2012-01-01

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

International Nuclear Information System (INIS)

Dherbey, Francine

2000-01-01

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

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

International Nuclear Information System (INIS)

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

1991-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Electrical impedance studies of uranium oxide

International Nuclear Information System (INIS)

Hampton, R.N.

1986-11-01

The thesis presents data on the electrical properties of uranium oxide at temperatures from 1700K to 4.2K, and pressures between 25 K bar and 70 K bar. The impedance data were analysed using the technique of complex plane representation to establish the conductivity and dielectric constant of uranium dioxide. The thermophysical data were compared with previously reported experimental and theoretical work on uranium dioxide and other fluorite structured oxides. (U.K.)

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-06-18

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

Uranium-bearing wastes and their radon emanation

International Nuclear Information System (INIS)

Sasaki, Tomozo; Imamura, Mitsutaka; Gunji, Yasuyoshi

2007-01-01

There are no data available with regard to radon emanation coefficients for uranium-bearing wastes; such data are needed for the assessment of radiation exposure from radon that will be generated in the distant future as one uranium progeny at shallow land disposal sites for uranium-bearing wastes. There are many kinds of uranium-bearing wastes. However, it is not necessary to measure the radon emanation coefficients for all of them for two reasons. First, the radon emanation coefficients for uranium-bearing wastes contaminated by dissolved uranium are determined by the uranium chemical form, the manner of uranium deposition on the waste matrix, and the size of the particles which constitute the waste matrix. Therefore, only a few representative measurements are sufficient for such uranium-bearing wastes. Second, it is possible to make theoretical calculations of radon emanation coefficients for uranium-bearing wastes contaminated by UO 2 particles before sintering. In the present study, simulated uranium-bearing wastes contaminated by dissolved uranium were prepared, their radon emanation coefficients were measured and radon emanation coefficients were calculated theoretically for uranium-bearing wastes contaminated by UO 2 particles before sintering. The obtained radon emanation coefficients are distributed at higher values than those for ubiquitous soils and rocks in the natural environment. Therefore, it is not correct to just compare uranium concentrations among uranium-bearing wastes, ubiquitous soils and rocks in terms of radiation exposure. The radon emanation coefficients obtained in the present study have to be employed together with the uranium concentration in uranium-bearing wastes in order to achieve proper assessment of radiation exposure. (author)

Spark plasma sintering and porosity studies of uranium nitride

Energy Technology Data Exchange (ETDEWEB)

Johnson, Kyle D., E-mail: kylej@kth.se; Wallenius, Janne; Jolkkonen, Mikael; Claisse, Antoine

2016-05-15

In this study, a number of samples of UN sintered by the SPS method have been fabricated, and highly pure samples ranging in density from 68% to 99.8%TD – corresponding to an absolute density of 14.25 g/cm{sup 3} out of a theoretical density of 14.28 g/cm{sup 3} – have been fabricated. By careful adjustment of the sintering parameters of temperature and applied pressure, the production of pellets of specific porosity may now be achieved between these ranges. The pore closure behaviour of the material has also been documented and compared to previous studies of similar materials, which demonstrates that full pore closure using these methods occurs near 97.5% of relative density. - Highlights: • UN pellets are fabricated over a wide array of densities using the SPS method. • The sintereing parameters necessary to produce pellets over a wide array of density space are charted. • Pellets of extremely high density (99.9% of TD, absolute density of 14.25 g/cm{sup 3}) are fabricated. • Full-closure of the porosity in this material is obtained at around 2.5% of total porosity.

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-09-15

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

Nuclear fuel rod with burnable plate and pellet-clad interaction fix

International Nuclear Information System (INIS)

Boyle, R.F.

1987-01-01

This patent describes a nuclear fuel rod comprising a metallic tubular cladding containing nuclear fuel pellets, the pellets containing enriched uranium-235. The improvement described here comprises: ceramic wafers, each wafter comprising a sintered mixture of gadolinium oxide and uranium dioxide, the uranium oxide having no more uranium-235 than is present in natural uranium dioxide. Each of the wafers is axially disposed between a major portion of adjacent the nuclear fuel pellets, whereby the wafers freeze out volatile fission products produced by the nuclear fuel and prevent interaction of the fission products with the metallic tubing cladding

Study of the low temperature oxidation of uranium powders and its application to the sintering of uranium oxide powders; Etude de l'oxydation des poudres dtranium a basse temperature et son application au frittage de poudres d'uranium oxyde

Energy Technology Data Exchange (ETDEWEB)

Conte-Albert, M. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-06-01

The uranium oxygen reaction has been studied with a view to obtaining U-UO{sub 2} samples containing about 20 per cent by weight of UO{sub 2} starting from spherical grain uranium powder (36 {mu} < {phi} < 50 {mu}). The techniques used are micrography, thermogravimetry, sintering under pressure, radio-crystallography. At 170 deg. C in air or argon + oxygen mixtures, the uranium oxide formed is always UO{sub 2} and it is uniformly distributed around the initial uranium spheres. These mixed powders can easily be sintered under pressure in the {gamma}-phase. The density of the samples obtained is 85 to 90 per cent of the theoretical density. The influence of UO{sub 2} on the properties of uranium has been shown by the use of dilatometry and thermal cycling in the {alpha} phase. The temperatures at which the phase changes {alpha} {r_reversible} {beta} and {beta} {r_reversible} {gamma} occur are lowered, the remnant expansion is decreased. High density samples resist well to thermal cycling; the characteristic defects of uranium: high distortion, wrinkled surface, have almost disappeared. Heat treatments in a secondary vacuum at 1050 deg. C cause crystallization of UO{sub 2} in a geometrical form and the appearance of a phase of the F.C.C. crystalline type having the composition U{sub W}C{sub X}O{sub Y}N{sub Z}. This phase causes a new decrease in the {alpha} {r_reversible} {beta}, {beta} {r_reversible} {gamma} transformation temperatures for the uranium. After ten dilatometric cycles the remanent expansion of the sample is about 0.5 per cent. The resistance to thermal cycling of a low density sample which has been heat-treated is similar to that of a high density sample which has not undergone a heat treatment. (author) [French] La reaction uranium-oxygene a ete etudiee pour permettre l'obtention d'echantillons U-UO{sub 2} a 20 pour cent en poids environ d'UO{sub 2}, a partir de billes d'uranium pulverulent (36 {mu} < {phi} < 50 {mu}). Les

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

Energy Technology Data Exchange (ETDEWEB)

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

1962-12-15

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

Alumina-zirconium ceramics synthesis by selective laser sintering/melting

International Nuclear Information System (INIS)

Shishkovsky, I.; Yadroitsev, I.; Bertrand, Ph.; Smurov, I.

2007-01-01

In the present paper, porous refractory ceramics synthesized by selective laser sintering/melting from a mixture of zirconium dioxide, aluminum and/or alumina powders are subjected to optical metallography and X-ray analysis to study their microstructure and phase composition depending on the laser processing parameters. It is shown that high-speed laser sintering in air yields ceramics with dense structure and a uniform distribution of the stabilizing phases. The obtained ceramic-matrix composites may be used as thermal and electrical insulators and wear resistant coating in solid oxide fuel cells, crucibles, heating elements, medical tools. The possibility to reinforce refractory ceramics by laser synthesis is shown on the example of tetragonal dioxide of zirconium with hardened micro-inclusion of Al 2 O 3 . By applying finely dispersed Y 2 O 3 powder inclusions, the type of the ceramic structure is significantly changed

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-06-15

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

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

International Nuclear Information System (INIS)

Hannerz, K.

1976-01-01

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

Compacted and Sintered Microstructure Depending on Uranium Powder Size in Zr-U Metallic Fuel

Energy Technology Data Exchange (ETDEWEB)

Yun, Chang Gun; Jun, Hyun-Joon; Ju, Jung Hwan; Lee, Ho Jin; Lee, Chong-Tak; Kim, Hyung Lae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-03-15

In case of the uranium (U) and zirconium (Zr) powders which have been utilized for the production of a metallic fuel in the various nuclear applications, the homogenous distribution of U powders in the Zr-U pellet has influenced significantly on the nuclear fuel performance. The inhomogeneity in a powder process was changed by various intricate factors, e.g. powder size, shape, distribution and so on. Particularly, the U inhomogeneity in the Zr-U pellets occurs by segregation derived from the great gaps of densities between Zr and U during compaction of the mixed powders. In this study, the relationship between powder size and homogeneity was investigated by using the different-sized U powders. The microstructure in Zr-U pellets reveals more homogeneity when the weight ration of Zr and U powders are close to 1. In addition, homogeneous pellets which were produced by fine U powders have higher density because the homogeneity affects the alloying reaction during sintering and the densification behavior of pore induced by powder size.

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Xenon Defects in Uranium Dioxide From First Principles and Interatomic Potentials

Science.gov (United States)

Thompson, Alexander

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

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

International Nuclear Information System (INIS)

Wlodarski, R.

1977-01-01

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

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

International Nuclear Information System (INIS)

Dorado, B.

2010-09-01

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

Nuclear fuel recycling system

International Nuclear Information System (INIS)

Lee, H.R.; Koch, A.K.; Krawczyk, A.

1981-01-01

A process is provided for recycling sintered uranium dioxide fuel pellets rejected during fuel manufacture and the swarf from pellet grinding. The scrap material is prepared mechanically by crushing and milling as a high solids content slurry, using scrap sintered UO 2 pellets as the grinding medium under an inert atmosophere

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

International Nuclear Information System (INIS)

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

1985-01-01

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

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

Contribution to the study of the textures of uranium rods prepared by sintering-extrusion, and their consequences on the thermal cycling behaviour; Contribution a l'etude des textures de barreaux d'uranium mis en forme par frittage-extrusion et leurs consequences sur le comportement au cyclage thermique

Energy Technology Data Exchange (ETDEWEB)

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

1964-11-01

Uranium rods prepared by sintering-extrusion in the {alpha} or {beta} phase (at various extrusion ratios) using slightly oxidised powders have been subjected to the thermal cycling test. At the same time, dilatometric and X-ray techniques have made it possible to determine the structures produced in these materials during their fabrication. A relationship is then proposed linking the texture to the increase in length on thermal cycling. 1. Two types of rods have been studied: Sintered-extruded in the {beta} phase: low density (88 per cent theoretical density), large grain-size and no preferential texture. Sintered-extruded in the {alpha} phase: high density (96 per cent theoretical density), fine grain with pronounced preferential texture. 2. After 1000 thermal cycles between 20 and 550 C, the increases in length are the following: 2 per cent for a uranium sintered-extruded in the {beta} phase (with surface cracking). between 14 and 56 per cent according to the extrusion ratio for on uranium sintered-extruded in the {alpha} phase (with no surface effects). 3. In the case of rods sintered-extruded in the {alpha} phase, determination of the pole figure using the Schulz reflection method showed the existence of two preferential orientations parallel to the direction of extrusion: one close to [100], the other close to [110]. By dilatometry it was then possible to measure quantitatively the proportion of each constituent in the overall texture and to show that an increase in the percentage of [100] occurs with increasing amounts of cold-working. 4. Finally, by comparing 2 and 3 it can be seen that the increases in length due to thermal cycling are connected to the percentage amounts of each component. It seems that the increases in length diminish as the percentage of [100] increases. On the other hand the behaviour of materials containing large amount of [110] is still far from clear. (author) [French] Des barreaux d'uranium realises par frittage-extrusion en phase

Production of sintered porous metal fluoride pellets

Science.gov (United States)

Anderson, L.W.; Stephenson, M.J.

1973-12-25

Porous pellets characterized by a moderately reactive crust and a softer core of higher reactivity are produced by forming agglomerates containing a metal fluoride powder and a selected amount ofwater. The metal fluoride is selected to be sinterable and essentially non-reactive with gaseous fluorinating agents. The agglomerates are contacted with a gaseous fluorinating agent under controlled conditions whereby the heat generated by localized reaction of the agent and water is limited to values effccting bonding by localized sintering. Porous pellets composed of cryolite (Na/sub 3/AlF/sub 6/) can be used to selectively remove trace quantities of niobium pentafluoride from a feed gas consisting predominantly of uranium hexafluoride. (Official Gazette)

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-04-01

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

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

International Nuclear Information System (INIS)

Baker, C.

1977-05-01

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

Micro-hardness of non-irradiated uranium dioxide

International Nuclear Information System (INIS)

Kim, Sung-Sik; Takagi, Osamu; Obata, Naomi; Kirihara, Tomoo.

1983-01-01

In order to obtain the optimum conditions for micro-hardness measurements of sintered UO 2 , two kinds of hardness tests (Vickers and Knoop) were examined with non-irradiated UO 2 of 2.5 and 5 μm in grain size. The hardness values were obtained as a function of the applied load in the load range of 25 -- 1,000 g. In the Vickers test, cracks were generated around the periphery of an indentation even at lower load of 50 g, which means the Vickers hardness is not suitable for UO 2 specimens. In the Knoop test, three stages of load dependence were observed for sintered pellet as well as for a single crystal by Bates. Load dependence of Knoop hardness and crack formation were discussed. In the range of applied load around 70 -- 100 g there were plateau region where hardness values were nearly unchanged and did not contain any cracks in the indentation. The plateau region represents a hardness of a specimen. From a comparison between the hardness values of 2.5 μm and those of 5 μm UO 2 , it was approved that the degree of sintering controls the hardness in the plateau region. (author)

Strain fields and line energies of dislocations in uranium dioxide

International Nuclear Information System (INIS)

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

2010-01-01

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

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

International Nuclear Information System (INIS)

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

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

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

International Nuclear Information System (INIS)

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

1981-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1966-10-01

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

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.

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

Synthesis and sintering of UN-UO{sub 2} fuel composites

Energy Technology Data Exchange (ETDEWEB)

Jaques, Brian J., E-mail: BrianJaques@BoiseState.edu [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States); Watkins, Jennifer; Croteau, Joseph R.; Alanko, Gordon A. [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States); Tyburska-Püschel, Beata [Department of Engineering Physics, University of Wisconsin–Madison, 1500 Engineering Dr., Madison, WI 53706 (United States); Meyer, Mitch [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Xu, Peng; Lahoda, Edward J. [Westinghouse Electric Company LLC, Pittsburgh, PA 15235 (United States); Butt, Darryl P., E-mail: DarrylButt@BoiseState.edu [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States)

2015-11-15

The design and development of an economical, accident tolerant fuel (ATF) for use in the current light water reactor (LWR) fleet is highly desirable for the future of nuclear power. Uranium mononitride has been identified as an alternative fuel with higher uranium density and thermal conductivity when compared to the benchmark, UO{sub 2}, which could also provide significant economic benefits. However, UN by itself reacts with water at reactor operating temperatures. In order to reduce its reactivity, the addition of UO{sub 2} to UN has been suggested. In order to avoid carbon impurities, UN was synthesized from elemental uranium using a hydride-dehydride-nitride thermal synthesis route prior to mixing with up to 10 wt% UO{sub 2} in a planetary ball mill. UN and UN – UO{sub 2} composite pellets were sintered in Ar – (0–1 at%) N{sub 2} to study the effects of nitrogen concentration on the evolved phases and microstructure. UN and UN-UO{sub 2} composite pellets were also sintered in Ar – 100 ppm N{sub 2} to assess the effects of temperature (1700–2000 °C) on the final grain morphology and phase concentration.

Process for in-situ leaching of uranium

International Nuclear Information System (INIS)

Espenscheid, W.F.; Yan, F.Y.

1983-01-01

The present invention relates to the recovery of uranium from subterranean ore deposits, and more particularly to an in-situ leaching operation employing an aqueous solution of sulfuric acid and carbon dioxide as the lixiviant. Uranium is solubilized in the lixiviant as it traverses the subterranean uranium deposit. The lixiviant is subsequently recovered and treated to remove the uranium

Automated fluorometer for uranium analysis

International Nuclear Information System (INIS)

McElhaney, R.J.; Caylor, J.D.; Cole, S.H.; Futrell, T.L.; Giles, V.M.

1978-03-01

An utomated fluorometer has proven to be a valuable analytical tool for analyzing natural waters for the Uranium Resource Evaluation (URE) project. Uranium is isolated from potential quenching ions and concentrated by extraction with tri-n-octylphosphine oxide (TOPO) in Varsol. A portion of the extract is placed on a sodium fluoride pellet which is then dried, sintered, and cooled. Sixteen samples can be analyzed in about 1.5 hours. The lower reporting limit has been set at 0.20 micrograms per liter

Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

Energy Technology Data Exchange (ETDEWEB)

McDeavitt, Sean M

2011-04-29

outlining the beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A—MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled “Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled “Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors” Appendix B—External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, “Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, “Uranium Powder Production Using a Hydride-Dehydride Process,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C—Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys” presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis

Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

International Nuclear Information System (INIS)

McDeavitt, Sean M.

2011-01-01

beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A - MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled 'Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications' A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled 'Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications' A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled 'Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors' Appendix B - External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, 'Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,' Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, 'Uranium Powder Production Using a Hydride-Dehydride Process,' Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C - Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled 'Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys' presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis by William Sames, Research Fellow

Preparation of uranium tetrafluoride

International Nuclear Information System (INIS)

Wirths, G.

1981-01-01

Uranium dioxide is converted to uranium tetrafluoride under stoichiometric excess of hydrogen fluoride. The water formed in the process and the unreacted hydrogen fluoride are cooled and the condensate fractionally distilled into water and approx. 40% hydrofluoric acid. The hydrofluoric acid and water-free hydrogen fluoride are fed back into the process. (WI) [de

Development of empirical relation for isotope of uranium in enriched uranium matrix

International Nuclear Information System (INIS)

Srivastava, S.K.; Vidyasagar, D.; Jha, S.K.; Tripathi, R.M.

2018-01-01

Uranium enriched in 235 U is required in commercial light water reactors to produce a controlled nuclear reaction. Enrichment allows the 235 U isotopes to be increased from 0.71% to a range between 2% to 5% depending upon requirement. The enriched uranium in the form of sintered UO 2 pellet is used for any commercially operating boiling light water reactors. The enriched uranium fuel bundle surface swipes sample is being analysed to assess the tramp uranium as a quality control parameter. It is known that the 234 U isotope also enriched along with 235 U isotope in conventional gaseous diffusion enrichment process. The information about enrichment percentage of 234 U helps to characterize isotopic properties of enriched uranium. A few reports provide the empirical equation and graphs for finding out the specific activity, activity percentage, activity ratio of 234 U isotopes for enriched uranium. Most of them have not provided the reference for the data used and their source. An attempt has been made to model the relationship between 234 U and 235 U as a function of uranium enrichment at low level

A spectroscopic study of uranium species formed in chloride melts

International Nuclear Information System (INIS)

Volkovich, Vladimir A.; Bhatt, Anand I.; May, Iain; Griffiths, Trevor R.; Thied, Robert C.

2002-01-01

The chlorination of uranium metal or uranium oxides in chloride melts offers an acceptable process for the head-end of pyrochemical reprocessing of spent nuclear fuels. The reactions of uranium metal and ceramic uranium dioxide with chlorine and with hydrogen chloride were studied in the alkali metal chloride melts, NaCl-KCl at 973K, NaCl-CsCl between 873 and 923K and LiCl-KCl at 873K. The uranium species formed therein were characterized from their electronic absorption spectra measured in situ. The kinetic parameters of the reactions depend on melt composition, temperature and chlorinating agent used. The reaction of uranium dioxide with oxygen in the presence of alkali metal chlorides results in the formation of alkali metal uranates. A spectroscopic study, between 723 and 973K, on their formation and their solutions was undertaken in LiCl, LiCl-KCl eutectic and NaCl-CsCl eutectic melts. The dissolution of uranium dioxide in LiCl-KCl eutectic at 923K containing added aluminium trichloride in the presence of oxygen has also been investigated. In this case, the reaction leads to the formation of uranyl chloride species. (author)

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

International Nuclear Information System (INIS)

Shi, Lei

2016-01-01

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

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

International Nuclear Information System (INIS)

Harmon, H.D.

1975-01-01

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

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

PRODUCTION OF URANIUM METAL BY CARBON REDUCTION

Science.gov (United States)

Holden, R.B.; Powers, R.M.; Blaber, O.J.

1959-09-22

The preparation of uranium metal by the carbon reduction of an oxide of uranium is described. In a preferred embodiment of the invention a charge composed of carbon and uranium oxide is heated to a solid mass after which it is further heated under vacuum to a temperature of about 2000 deg C to produce a fused uranium metal. Slowly ccoling the fused mass produces a dendritic structure of uranium carbide in uranium metal. Reacting the solidified charge with deionized water hydrolyzes the uranium carbide to finely divide uranium dioxide which can be separated from the coarser uranium metal by ordinary filtration methods.

Production of nuclear ceramic fuel for nuclear power plants at 'Ulba metallurgical plant' OSC

International Nuclear Information System (INIS)

Khadeev, V.G.

2000-01-01

The paper describes the flow-sheet of production of uranium dioxide powders and nuclear ceramic fuel pellets of them existing at the facility. 'UMP' OSC applies ADU extraction process of UO2 powders production. An indisputable success of the process is the possibility of use of the wide range of raw materials. Uranium hexafluoride, uranium oxides, uranium metal, uranium tetrafluoride, uranyl salts, uranium ore concentrates, all possible types of uranium-containing materials the processing of which by routine methods is difficult (ashes, scraps, etc.) are used as the raw materials. In addition, a reprocessed nuclear fuel can be used for fuel production. The quality of uranium dioxide powder produced does not depend on the type of uranium raw material used. High selectivity of extraction refining makes possible to obtain material with rather low impurities content that meets practically all specifications for uranium dioxide known to us. Ceramic and process features of uranium dioxide powders, namely, specific surface, bulk density, grain size and sinterability make possible to produce nuclear ceramic fuel with specified features. Quality of uranium dioxide powders produced by 'UMP' OSC was highly rated by General Electric company that is one of the leading companies from fuel manufactures in the USA market . It has certified 'UMP' OSC as its supplier. Currently, our company makes great efforts on establishing production of uranium dioxide powders with natural isotopes content for production of fuel for CANDU reactors. Trial lots of such powders are under tests at some companies manufacturing fuel for this type reactors in Canada, USA and Corea

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.

Uranium conversion

International Nuclear Information System (INIS)

Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina

2006-03-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

International Nuclear Information System (INIS)

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

1978-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Martin, G

2007-06-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-06-15

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

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

International Nuclear Information System (INIS)

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

1988-04-01

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

Fabrication and testing of ceramic UO2 fuel - I-III. Part I

International Nuclear Information System (INIS)

Novakovic, M.

1961-12-01

The task described consists of the following: fabrication of UO 2 with different granulation from uranyl nitrate by ammonia diuranate; determination of size and shape distributions of metal and ceramic powders; fabrication of sintered pressed samples UO 2 ; investigating the properties of sintered uranium dioxide dependent on the fabrication process; producing a vibrator for compacting UO 2 powder. This volume includes reports on the first two tasks

Feasibility study of the dissolution rates of uranium ore dust, uranium concentrates and uranium compounds in simulated lung fluid

International Nuclear Information System (INIS)

Robertson, R.

1986-01-01

A flow-through apparatus has been devised to study the dissolution in simulated lung fluid of aerosol materials associated with the Canadian uranium industry. The apparatus has been experimentally applied over 16 day extraction periods to approximately 2g samples of < 38um and 53-75um particle-size fractions of both Elliot Lake and Mid-Western uranium ores. The extraction of uranium-238 was in the range 24-60% for these samples. The corresponding range for radium-226 was 8-26%. Thorium-230, lead-210, polonium-210, and thorium-232 were not significantly extracted. It was incidentally found that the elemental composition of the ores studied varies significantly with particle size, the radionuclide-containing minerals and several extractable stable elements being concentrated in the smaller size fraction. Samples of the refined compounds uranium dioxide and uranium trioxide were submitted to similar 16 day extraction experiments. Approximately 0.5% of the uranium was extracted from a 0.258g sample of unsintered (fluid bed) uranium dioxide of particle size < 38um. The corresponding figure for a 0.292g sample of uranium trioxide was 97%. Two aerosol samples on filters were also studied. Of the 88ug uranium initially measured on stage 2 of a cascade impactor sample collected from the yellow cake packing area of an Elliot Lake mill, essentially 100% was extracted over a 16 day period. The corresponding figure for an open face filter sample collected in a fuel fabrication plant and initially measured at 288ug uranium was approximately 3%. Recommendations are made with regard to further work of a research nature which would be useful in this area. Recommendations are also made on sampling methods, analytical methods and extraction conditions for various aerosols of interest which are to be studied in a work of broader scope designed to yield meaningful data in connection with lung dosimetry calculations

boron nitride coating of uranium dioxide and uranium dioxide-gadolinium oxide fuels by chemical precipitation method

International Nuclear Information System (INIS)

Uslu, I.; Tanker, E.; Guenduez, G.

1997-01-01

In this research pure urania and urania-gadolinia (5 and 10 %) fuels were coated with boron nitride (BN). This is achieved through chemical vapor deposition (CVD) using boron tricloride BCl 3 ) and ammonia (NH 3 ) at 600 C.Boron tricloride and ammonia are carried to tubular furnace using hydrogen as carrier gas. The coated samples were sintered at 1600 K. The properties of the coated samples were observed using BET surface area analysis, infrared spectra (IR), X-Ray Diffraction and Scanning Electron Microscope (SEM) techniques

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

International Nuclear Information System (INIS)

Vivant-Duguay, Christelle

1998-01-01

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

About the elaboration of pure uranium dicarbide

International Nuclear Information System (INIS)

Besson, J.; Blum, P.; Guinet, Ph.; Spitz, J.

1963-01-01

In order to develop methods for the elaboration of as pure as possible uranium dicarbide, the authors report the study of different elaboration processes based on the reaction between uranium and carbon, or between uranium and hydrocarbon, or between uranium oxide and carbon. They finally choose a method which comprises an arc-induced fusion of a mixture of uranium dioxide and carbon. The fusion process is described. The influence of thermal treatments is discussed as well as the graphite electrode carburization

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

Energy Technology Data Exchange (ETDEWEB)

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

1983-01-01

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

Fabrication of lanthanum-doped thorium dioxide by high-energy ball milling and spark plasma sintering

Energy Technology Data Exchange (ETDEWEB)

Scott, Spencer M.; Yao, Tiankai [Department of Mechanical, Aerospace, and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY 12180 (United States); Lu, Fengyuan [Department of Mechanical & Industrial Engineering, Louisiana State University, Baton Rouge, LA 70803 (United States); Xin, Guoqing; Zhu, Weiguang [Department of Mechanical, Aerospace, and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY 12180 (United States); Lian, Jie, E-mail: lianj@rpi.edu [Department of Mechanical, Aerospace, and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY 12180 (United States)

2017-03-15

Abstract: High-energy ball milling was used to synthesize Th{sub 1-x}La{sub x}O{sub 2-0.5x} (x = 0.09, 0.23) solid solutions, as well as improve the sinterability of ThO{sub 2} powders. Dense La-doped ThO{sub 2} pellets with theoretical density above 94% were consolidated by spark plasma sintering at temperatures above 1400 °C for 20 min, and the densification behavior and the non-equilibrium effects on phase and structure were investigated. A lattice contraction of the SPS-densified pellets occurred with increasing ball milling duration, and a secondary phase with increased La-content was observed in La-doped pellets. A dependence on the La-content and sintering duration for the onset of localized phase segregation has been proposed. The effects of high-energy ball milling, La-content, and phase formation on the thermal diffusivity were also studied for La-doped ThO{sub 2} pellets by laser flash measurement. Increasing La-content and high energy ball milling time decreases thermal diffusivity; while the sintering peak temperature and holding time beyond 1600 °C dramatically altered the temperature dependence of the thermal diffusivity beyond 600 °C. - Highlights: • Lanthanum incorporation into ThO{sub 2} by high energy ball milling and rapid consolidation by spark plasma sintering. • Elucidation of phase behavior of the La-doped ThO{sub 2} and the contributions of La incorporation and SPS sintering conditions. • Investigation of the effects of La incorporation and high energy ball milling on the thermal behavior of La-doped ThO{sub 2}.

Models for the adsorption of uranium on titanium dioxide

International Nuclear Information System (INIS)

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

1980-01-01

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

Titanium di-oxide films using a less hygroscopic colloidal precursor

Energy Technology Data Exchange (ETDEWEB)

Vandana,, E-mail: vandana1@nplindia.org; Batra, Neha; Kumar, Praveen; Sharma, Pooja; Singh, P.K., E-mail: pksingh@nplindia.org

2014-04-01

We report the study of titanium dioxide films (TiO{sub 2}) using titanium di-isopropoxyl di-2ethyl hexanoate Ti(OC{sub 3}H{sub 7}){sub 2} (C{sub 7}H{sub 15}COO){sub 2} colloidal precursor. This compound is less hygroscopic in nature and easy to use with processes like spin or dip coating. Thin films of TiO{sub 2} are made on silicon substrates and their structural and optical properties are studied. The effect of Ti content in the precursor, sintering temperature and its duration on film thickness and refractive index are investigated. Refractive index shows an increasing trend with the rise in the sintering temperature but remains unchanged with the time. The film thickness decreases with both sintering temperature and time and increases with Ti content in the precursor. Reflectivity measurements show marked reduction in the reflection losses compared to bare silicon surface wherein the film thickness is altered by spin speed. XRD results show anatase phase in the samples sintered at lower temperature (<680 °C), however, a mix of anatase, brookite and rutile phases is seen above this temperature. In the samples sintered above 1100 °C, rutile phase is dominant. These results are supported by the X-ray photoelectron spectroscopy. Atomic force microscopy reveals larger grain size at higher sintering temperature. The titanium dioxide films of desirable thickness and refractive index could be used as an antireflection coating on solar cells. - Highlights: • TiO{sub 2} films are made using titanium di-isopropoxyl di-2ethyl hexanoate precursor. • Effect of Ti content in the precursor, sintering temperature and time is studied. • Refractive index (μ) increases with sintering temperature but is independent of time. • Films of desired thickness and μ could be used as an antireflection coating. • XRD results show that rutile phase dominates in samples sintered above 1100 °C.

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

Science.gov (United States)

Cavender, Daniel

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

The Mexican mesozoic uranium province: its distribution and metallogeny

International Nuclear Information System (INIS)

Bazan B, S.

1981-01-01

The distribution of uranium scattered in sedimentary terrains of the continental jurassic such as those found in the Tlaxiaco-Guerrero Basin encourage the outlook for uncovering extensive new deposits of strato-bound uranium belonging to the Mexican mesozoic in other structurally similar intercratonic basins. Stratographic and paleographic structural references define the simultaneous evolution of five sedimentary basins during the Mexican geotechtonic cycle: 1. the Tlaxiaco-Guerrero basin, 2. the Huayacocotla basin, 3. the Gulf of Sabinas basin, 4. the Chihuahua basin and 5. the Sonora basin. From the various lithostratographic formations in them we favourably infer the presence of intermountainous mesozoic concentrations of uranium sediments leached from crystalline precambric packets and from nevadian plutonites and volcanic rocks. During the metallogeny process described under the techtonic evolution of the Mexican structural belt, the presence is established of extensive terciary hydrothermal uranium deposits in the districts of Aldama, Chihuahua; Coneto-El Rodeo, Durango; Vizarron de Montes, Queretaro; Tlaucingo, Puebla; Los Amoles, Sonora; El Picacho, Sonora; Amalia Margarita, Coahuila; etc., scattered in sandstones and sinters of the continental mesozoic and shifted during the postorogenic phase of the Mexican geotectonic cycle. The extensive mesozoic province defined within the Mexican territory favourable to large deposits of uranium, scattered and strato-bound in triassic, jurassic and cretaceous sandstone and sinters, could resolve future demands for energetics within a modified philosophy and resourceful policy of regional mining. (author)

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

Fabrication and testing of ceramic UO{sub 2} fuel - I-III. Part I; Izrada i ispitivanje keramickog goriva na bazi UO{sub 2}- I-III, I Deo

Energy Technology Data Exchange (ETDEWEB)

Novakovic, M [Institute of Nuclear Sciences Boris Kidric, Laboratorija za termotehniku reaktora, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

The task described consists of the following: fabrication of UO{sub 2} with different granulation from uranyl nitrate by ammonia diuranate; determination of size and shape distributions of metal and ceramic powders; fabrication of sintered pressed samples UO{sub 2}; investigating the properties of sintered uranium dioxide dependent on the fabrication process; producing a vibrator for compacting UO{sub 2} powder. This volume includes reports on the first two tasks.

Microwave combustion and sintering without isostatic pressure

International Nuclear Information System (INIS)

Ebadian, M.A.

1998-01-01

In recent years interest has grown rapidly in the application of microwave energy to the processing of ceramics, composites, polymers, and other materials. Advances in the understanding of microwave/materials interactions will facilitate the production of new ceramic materials with superior mechanical properties. One application of particular interest is the use of microwave energy for the mobilization of uranium for subsequent redeposition. Phase III (FY98) will focus on the microwave assisted chemical vapor infiltration tests for mobilization and redeposition of radioactive species in the mixed sludge waste. Uranium hexachloride and uranium (IV) borohydride are volatile compounds for which the chemical vapor infiltration procedure might be developed for the separation of uranium. Microwave heating characterized by an inverse temperature profile within a preformed ceramic matrix will be utilized for CVI using a carrier gas. Matrix deposition is expected to commence from the inside of the sample where the highest temperature is present. The preform matrix materials, which include aluminosilicate based ceramics and silicon carbide based ceramics, are all amenable to extreme volume reduction, densification, and vitrification. Important parameters of microwave sintering such as frequency, power requirement, soaking temperature, and holding time will be investigated to optimize process conditions for the volatilization of uranyl species using a reactive carrier gas in a microwave chamber

Fracture toughness of yttria-stabilized zirconia sintered in conventional and microwave ovens.

Science.gov (United States)

Marinis, Aristotelis; Aquilino, Steven A; Lund, Peter S; Gratton, David G; Stanford, Clark M; Diaz-Arnold, Ana M; Qian, Fang

2013-03-01

The fabrication of zirconium dioxide (ZrO2) dental prosthetic substructures requires an extended sintering process (8 to 10 hours) in a conventional oven. Microwave sintering is a shorter process (2 hours) than conventional sintering. The purpose of this study was to compare the fracture toughness of 3 mol % Y2O3-stabilized ZrO2 sintered in a conventional or microwave oven. Partially sintered ZrO2 specimens from 3 manufacturers, KaVo, Lava 3M, and Crystal HS were milled (KaVo Everest engine) and randomly divided into 2 groups: conventional sintering and microwave sintering (n=16 per group). The specimens were sintered according to the manufacturers' recommendations and stored in artificial saliva for 10 days. Fracture toughness was determined by using a 4-point bend test, and load to fracture was recorded. Mean fracture toughness for each material was calculated. A 2-way ANOVA followed by the Tukey HDS post hoc test was used to assess the significance of sintering and material effects on fracture toughness, including an interaction between the 2 factors (α=.05). The 2-way ANOVA suggested a significant main effect for ZrO2 manufacturer (P.05). The main effect of the sintering process (Conventional [5.30 MPa·m(1/2) ±1.00] or Microwave [5.36 MPa·m(1/2) ±0.92]) was not significant (P=.76), and there was no interaction between sintering and ZrO2 manufacturer (P=.91). Based on the results of this study, no statistically significant difference was observed in the fracture toughness of ZrO2 sintered in microwave or conventional ovens. Copyright © 2013 The Editorial Council of the Journal of Prosthetic Dentistry. Published by Mosby, Inc. All rights reserved.

Preparation of UO_2 Fine Particle by Hydrolysis of Uranium(IV) Alkoxide

OpenAIRE

Satoh, Isamu; Takahashi, Mitsuyuki; Miura, Shigeyuki

1997-01-01

Fine particles of uranium(IV) dioxides were obtained by hydrolysis of uranium(IV) ethoxide which was synthesized by reacting uranium tetrachloride with sodium ethoxide. The monodispersed submicrometer particles were confirmed by SEM observation.

Uranium conversion; Urankonvertering

Energy Technology Data Exchange (ETDEWEB)

Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina [Swedish Defence Research Agency (FOI), Stockholm (Sweden)

2006-03-15

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

Fabrication of chamfered uranium-plutonium mixed carbide pellets

International Nuclear Information System (INIS)

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

1985-10-01

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

Synthesis and preservation of graphene-supported uranium dioxide nanocrystals

Energy Technology Data Exchange (ETDEWEB)

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

2016-07-15

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

Micromechanical approach of behavior of uranium dioxide nuclear fuel

International Nuclear Information System (INIS)

Soulacroix, Julian

2014-01-01

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

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

International Nuclear Information System (INIS)

Webb, Jonathan A.; Charit, Indrajit

2011-01-01

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

A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

International Nuclear Information System (INIS)

Vasudevamurthy, G.; Radecka, A.; Massey, C.

2015-01-01

High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

Energy Technology Data Exchange (ETDEWEB)

Vasudevamurthy, G.; Radecka, A.; Massey, C. [Virginia Commonwealth Univ., Richmond, VA (United States). High Temperature Materials Lab.

2015-07-01

High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

Contribution to the study of the sintering mechanisms of uranium powders in the {alpha}, {beta}, and {gamma} phases; Contribution a l'etude des mecanismes de frittage de poudre d'uranium en phases {alpha}, {beta}, et {gamma}

Energy Technology Data Exchange (ETDEWEB)

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

1964-06-01

This study of the sintering mechanisms of uranium powders prepared by calci-thermy has been effected using continuous dilatometric measurements of the shrinkage of samples previously compressed at room temperature in purified argon gas. The tests carried out in the {alpha}, {beta} and {gamma} phases have led to the observation that the first step of the sintering appears to be governed by a volume self-diffusion mechanism; the activation heat values found for the sintering mechanisms are close to those deduced during studies of volume self-diffusion using the direct radio-tracer method. Furthermore it has been possible to show that in the {gamma} domain a second sintering mechanism occurs involving much longer sintering times; the heats of activation are much lower and this appears to indicate that there occurs a mechanism involving pore elimination by grain boundary diffusion of the vacancies. Furthermore, the dilatometric tests have shown the simultaneous influence of two important parameters in this work: grain boundaries and the diffusion coefficients. In the second part of the report are given results concerning the examination of sintered samples by various methods with a view to elucidating their structure and some of their physical properties. In this way it has been possible, by carrying out metallographic examinations after etching by ionic bombardment, to determine the changes in the porosity of the three phases {alpha}, {beta} and {gamma}, as well as the structure and the nature of the inclusions in each sample. Density and porosity measurements have also been carried out. The variations in these two sets of results make it possible to confirm the preceding dilatometric end micro-graphic examinations. Finally a detailed dilatometric study of the samples sintered in the {gamma} phase has shown the effect of oxide layers, associated with the existence of porosity, on the amplitudes and temperatures of the allotropic transformations, these latter being

The development of the production process for the thorium/uranium dicarbide fuel kernels for the first charge of the Dragon Reactor

International Nuclear Information System (INIS)

Burnett, R.C.; Hankart, L.J.; Horsley, G.W.

1965-05-01

The development of methods of producing spheroidal sintered porous kernels of hyperstoichiometric thorium/uranium dicarbide solid solution from thorium/uranium monocarbide/carbon and thoria/urania/carbon powder mixes is described. The work has involved study of (i) Methods of preparing green kernels from UC/Th/C powder mixes using the rotary sieve technique. (ii) Methods of producing green kernels from UO2/Th02/C powder mixes using the planetary mill technique. (iii) The conversion by appropriate heat treatment of green kernels produced by both routes to sintered porous kernels of thorium/uranium carbide. (iv) The efficiency of the processes. (author)

Technological aspects of UO2 sintering at low temperature

International Nuclear Information System (INIS)

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

1999-01-01

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

Surface studies on uranium monocarbide using XPS and SIMS

International Nuclear Information System (INIS)

Asuvathraman, R.

1995-01-01

The air-exposed surfaces of sintered and arc-melted UC samples were examined by XPS and SIMS. XPS results indicate that the surface is covered with a very thin layer of UO 2 mixed with free carbon, which would have formed along with the oxide during the reaction between UC and oxygen or moisture. From the SIMS depth profile of oxygen, the thickness of the oxide layer is found to be approximately 10 nm. The SIMS oxygen images of the surface as a function of etching time reveal that the surface of UC consists of a top layer of adsorbed moisture/oxygen; this contamination layer is followed by a layer containing uranium oxide, uranium hydroxide and free carbon and then grain boundary oxide and finally bulk UC. The behaviour of sintered and arc-melted samples is similar. ((orig.))

Sintering and densification; new techniques: sinter forging

International Nuclear Information System (INIS)

Winnubst, A.J.A.

1998-01-01

In this chapter pressure assisted sintering methods will be described. Attention will mainly be paid to sinter forging as a die-wall free uniaxial pressure sintering technique, where large creep strains are possible. Sinter forging is an effective tool to reduce sintering temperature and time and to obtain a nearly theoretically dense ceramic. In this way grain size in tetragonal zirconia ceramics can be reduced down to 100 nm. Another important phenomenon is the reduction of the number density and size of cracks and flaws resulting in higher strength and improved reliability, which is of utmost importance for engineering ceramics. The creep deformation during sinter forging causes a rearrangement of the grains resulting in a reduction of interatomic spaces between grains, while grain boundary (glassy) phases can be removed. The toughness and in some cases the wear resistance is enhanced after sinter forging as a result of the grain-boundary-morphology improvement. (orig.)

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

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

Science.gov (United States)

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

2018-06-01

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

Spent fuel from nuclear research reactors immobilized in sintered glass

International Nuclear Information System (INIS)

Mateos, P.; Russo, D.O.; Rodriguez, D.; Heredia, A.; Sanfilippo, M.; Sterba, M.

2002-01-01

Different kinds of glasses, borosilicates, Iron borosilicates and Iron phosphates, were tested in order to determine its capability to immobilize calcined uranium silicide in a sintering process. Iron phosphate glass developed in our laboratory showed the best results in SEM analysis. Also its gravimetric leaching rate is less than 0.45 g.m -2 .day -1 for 7 and 10% loading which is lower than any previously studied for us. (author)

Kinetic study of uranium carburization by different carbonated gases

International Nuclear Information System (INIS)

Feron, Guy

1963-01-01

The kinetic study of the reaction U + CO 2 and U + CO has been performed by a thermogravimetric method on a spherical uranium powder, in temperature ranges respectively from 460 to 690 deg. C and from 570 to 850 deg. C. The reaction with carbon dioxide leads to uranium dioxide. A carbon deposition takes place at the same time. The global reactions is the result of two reactions: U + 2 CO 2 → UO 2 + 2 CO U + CO 2 → UO 2 + C The reaction with carbon monoxide leads to a mixture of dioxide UO 2 , dicarbide UC 2 and free carbon. The main reaction can be written. U + CO → 1/2 UO 2 + 1/2 UC 2 The free carbon results of the disproportionation of the carbon monoxide. A remarkable separation of the two phases UO 2 and UC 2 can be observed. A mechanism accounting for the phenomenon has been proposed. The two reactions U + CO 2 and U + CO begin with a long germination period, after which, the reaction velocity seems to be limited in both cases by the ionic diffusion of oxygen through the uranium dioxide. (author) [fr

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

Science.gov (United States)

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

1973-01-01

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

High temperature behavior of metallic inclusions in uranium dioxide

International Nuclear Information System (INIS)

Yang, R.L.

1980-08-01

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

Gasification reactivity and ash sintering behaviour of biomass feedstocks

Energy Technology Data Exchange (ETDEWEB)

Moilanen, A.; Nasrullah, M.

2011-12-15

Char gasification reactivity and ash sintering properties of forestry biomass feedstocks selected for large-scale gasification process was characterised. The study was divided into two parts: (1) Internal variation of the reactivity and the ash sintering of feedstocks. (2) Measurement of kinetic parameters of char gasification reactions to be used in the modelling of a gasifier. The tests were carried out in gases relevant to pressurized oxygen gasification, i.e. steam and carbon dioxide, as well as their mixtures with the product gases H{sub 2} and CO. The work was based on experimental measurements using pressurized thermobalance. In the tests, the temperatures were below 1000 deg C, and the pressure range was between 1 and 20 bar. In the first part, it was tested the effect of growing location, storage, plant parts and debarking method. The following biomass types were tested: spruce bark, pine bark, aspen bark, birch bark, forestry residue, bark feedstock mixture, stump chips and hemp. Thick pine bark had the lowest reactivity (instantaneous reaction rate 14%/min) and hemp the highest (250%/min); all other biomasses laid between these values. There was practically no difference in the reactivities among the spruce barks collected from the different locations. For pine bark, the differences were greater, but they were probably due to the thickness of the bark rather than to the growth location. For the spruce barks, the instantaneous reaction rate measured at 90% fuel conversion was 100%/min, for pine barks it varied between 14 and 75%/min. During storage, quite large local differences in reactivity seem to develop. Stump had significantly lower reactivity compared with the others. No clear difference in the reactivity was observed between barks obtained with the wet and dry debarking, but, the sintering of the ash was more enhanced for the bark from dry debarking. Char gasification rate could not be modelled in the gas mixture of H{sub 2}O + CO{sub 2} + H{sub 2

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

International Nuclear Information System (INIS)

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

1981-01-01

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

Oxidation and crystal field effects in uranium

Energy Technology Data Exchange (ETDEWEB)

Tobin, J. G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Booth, C. H. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Shuh, D. K. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); van der Laan, G. [Diamond Light Source, Didcot (United Kingdom); Sokaras, D. [Stanford Synchrotron Radiation Lightsource, Stanford, CA (United States); Weng, T. -C. [Stanford Synchrotron Radiation Lightsource, Stanford, CA (United States); Yu, S. W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bagus, P. S. [Univ. of North Texas, Denton, TX (United States); Tyliszczak, T. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Nordlund, D. [Stanford Synchrotron Radiation Lightsource, Stanford, CA (United States)

2015-07-06

An extensive investigation of oxidation in uranium has been pursued. This includes the utilization of soft x-ray absorption spectroscopy, hard x-ray absorption near-edge structure, resonant (hard) x-ray emission spectroscopy, cluster calculations, and a branching ratio analysis founded on atomic theory. The samples utilized were uranium dioxide (UO₂), uranium trioxide (UO₃), and uranium tetrafluoride (UF₄). As a result, a discussion of the role of non-spherical perturbations, i.e., crystal or ligand field effects, will be presented.

Thorium dioxide: properties and nuclear applications

International Nuclear Information System (INIS)

Belle, J.; Berman, R.M.

1984-01-01

This is the sixth book on reactor materials published under sponsorship of the Naval Reactors Office of the United States Department of Energy, formerly the United States Atomic Energy Commission. This book presents a comprehensive compilation of the most significant properties of thorium dioxide, much like the book Uranium Dioxide: Properties and Nuclear Applications presented information on the fuel material used in the Shippingport Pressurized Water Reactor core

Thorium dioxide: properties and nuclear applications

Energy Technology Data Exchange (ETDEWEB)

Belle, J.; Berman, R.M. (eds.)

1984-01-01

This is the sixth book on reactor materials published under sponsorship of the Naval Reactors Office of the United States Department of Energy, formerly the United States Atomic Energy Commission. This book presents a comprehensive compilation of the most significant properties of thorium dioxide, much like the book Uranium Dioxide: Properties and Nuclear Applications presented information on the fuel material used in the Shippingport Pressurized Water Reactor core.

Dry uranium tetrafluoride process preparation using the uranium hexafluoride reconversion process effluents

International Nuclear Information System (INIS)

Silva Neto, Joao Batista da

2008-01-01

It is a well known fact that the use of uranium tetrafluoride allows flexibility in the production of uranium suicide and uranium oxide fuel. To its obtention there are two conventional routes, the one which reduces uranium from the UF 6 hydrolysis solution with stannous chloride, and the hydro fluorination of a solid uranium dioxide. In this work we are introducing a third and a dry way route, mainly utilized to the recovery of uranium from the liquid effluents generated in the uranium hexafluoride reconversion process, at IPEN/CNEN-SP. Working in the liquid phase, this route comprises the recuperation of ammonium fluoride by NH 4 HF 2 precipitation. Working with the solid residues, the crystallized bifluoride is added to the solid UO 2 , which comes from the U mini plates recovery, also to its conversion in a solid state reaction, to obtain UF 4 . That returns to the process of metallic uranium production unity to the U 3 Si 2 obtention. This fuel is considered in IPEN CNEN/SP as the high density fuel phase for IEA-R1m reactor, which will replace the former low density U 3 Si 2 -Al fuel. (author)

In situ leaching process for recording uranium values

International Nuclear Information System (INIS)

McKnight, W.M.; Timmins, T.H.; Sherry, H.S.

1977-01-01

A method of recovering uranium values from a subterranean deposit comprising: injecting an alkaline carbonate lixiviant into said deposit; flowing said alkaline carbonate lixiviant through said deposit to dissolve said uranium values into said lixiviant; producing said lixiviant and said dissolved uranium values from said deposit; flowing said lixiviant and said dissolved uranium values through an adsorption material to adsorp said uranium values from said lixiviant; eluting said adsorption material with an eluant of ammonium carbonate to desorb said uranium values from said adsorption material into said eluate in a concentration greater than in said lixiviant; heating said eluate and said desorbed uranium values to vaporize off ammonia and carbon dioxide therefrom, thereby causing uranium values to crystallize from the eluate; and recovering said solid uranium values

Protection of uranium by metallic coatings

International Nuclear Information System (INIS)

Baque, P.; Koch, P.; Dominget, R.; Darras, R.

1968-01-01

A study is made of the possibilities of inhibiting or limiting, by means of protective metallic coatings, the oxidation of uranium by carbon dioxide at high temperature. In general, surface films containing intermetallic compounds or solid solutions of uranium with aluminium, zirconium, copper, niobium, nickel or chromium are formed, according to the techniques employed which are described here. The processes most to be recommended are those of direct diffusion starting from a thin sheet or tube, of vacuum deposition, or of immersion in a molten bath of suitable composition. The conditions for preparing these coatings have been optimized as a function of the protective effect obtained in carbon dioxide at 450 or at 500 C. Only the aluminium and zirconium based coatings are really satisfactory since they can lead to a reduction by a factor of 5 to 10 in the oxidation rate of uranium in the conditions considered; they make it possible in particular to avoid or to reduce to a very large extent the liberation of powdered oxide. Furthermore, the coatings produced generally give the uranium good protection against atmospheric corrosion. (author) [fr

Synthesis of uranium metal using laser-initiated reduction of uranium tetrafluoride by calcium metal

International Nuclear Information System (INIS)

West, M.H.; Martinez, M.M.; Nielsen, J.B.; Court, D.C.; Appert, Q.D.

1995-09-01

Uranium metal has numerous uses in conventional weapons (armor penetrators) and nuclear weapons. It also has application to nuclear reactor designs utilizing metallic fuels--for example, the former Integral Fast Reactor program at Argonne National Laboratory. Uranium metal also has promise as a material of construction for spent-nuclear-fuel storage casks. A new avenue for the production of uranium metal is presented that offers several advantages over existing technology. A carbon dioxide (CO 2 ) laser is used to initiate the reaction between uranium tetrafluoride (UF 4 ) and calcium metal. The new method does not require induction heating of a closed system (a pressure vessel) nor does it utilize iodine (I 2 ) as a chemical booster. The results of five reductions of UF 4 , spanning 100 to 200 g of uranium, are evaluated, and suggestions are made for future work in this area

Green strength of zirconium sponge and uranium dioxide powder compacts

International Nuclear Information System (INIS)

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

2008-01-01

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

METHOD OF PRODUCING URANIUM METAL BY ELECTROLYSIS

Science.gov (United States)

Piper, R.D.

1962-09-01

A process is given for making uranium metal from oxidic material by electrolytic deposition on the cathode. The oxidic material admixed with two moles of carbon per one mole of uranium dioxide forms the anode, and the electrolyte is a mixture of from 40 to 75% of calcium fluoride or barium fluoride, 15 to 45% of uranium tetrafluoride, and from 10 to 20% of lithium fluoride or magnesium fluoride; the temperature of the electrolyte is between 1150 and 1175 deg C. (AEC)

Obtaining of U-2.5Zr7.5Nb and U-3Zr-9Nb alloys by sintering process

International Nuclear Information System (INIS)

Mazzeu, Thiago de Oliveira; Paula, Joao Bosco de; Ferraz, Wilmar Barbosa; Santos, Ana Maria Matildes dos; Brina, Jose Giovanni Mascarenhas

2011-01-01

The development of metallic fuels with low enrichment to be used in research and test reactors, as well in the future pressurized water reactors, focuses on the search for uranium alloys of high density. Alloying elements such as Zr, Nb and Mo are added to uranium to improve fuel performance in reactors. In this context, the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) in Belo Horizonte is developing the U-2.5Zr-7.5Nb and U- 3Zr-9Nb (weight %) alloys by the innovative process of sintering that utilizes raw materials in the form of powders. The powders were pressed at 400MPa and then sintered under a vacuum of about 5 x 10-6 Torr at temperatures ranging from 1050 deg to 1300 deg C. The densities of the alloys were measured geometrically and by hydrostatic method using water. The microstructures of the pellets were observed by scanning electron microscopy (SEM) and the elements of alloying were identified by energy dispersive X-ray spectroscopy (SEM/EDS) analysis. The obtained results showed a small increasing density with rising sintering temperature. The highest density achieved was approximately 80% of theoretical density. It was also qualitatively observed that the superficial oxidation of the pellets increased with increasing sintering temperature thus avoiding the fusion of the alloys at higher temperatures. (author)

Solid state processing of massive uranium mononitride, using uranium and uranium higher nitride powders as starting materials (1962); Preparation a l'etat solide de mononitrure d'uranium massif a partir de poudres d'uranium et de nitrures superieurs d'uranium (1962)

Energy Technology Data Exchange (ETDEWEB)

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

1960-12-15

The mechanism and the optimum conditions for preparing uranium mononitride have been studied. The results have been used for hot pressing (250 kg/cm{sup 2}, 1000 deg. C, under vacuum) a mixture of powders of uranium and uranium higher nitrides. The products obtained have been identified by X-ray measurements and may be - at will and depending upon the stoichiometry - either UN, or a cermet a U{sub {alpha}}-UN. As revealed by the curved shape of grain boundaries, the sinters obtained here do not easily evolve towards physico-chemical equilibrium when submitted to heat treatment. This behaviour is quite different from the one observed with uranium monocarbide prepared by a similar method. This fact may be ascribed to the insolubility in the matrix UN of particles of UO{sub 2} being present as impurities. The density, hardness and thermal conductivity of these products are higher than those measured on uranium nitride or cermets U-UN obtained by other methods. (author) [French] Apres une etude prealable du mecanisme et des conditions optimales de nitruration de l'uranium, on a montre qu'il est possible de preparer par frittage sous charge (250 kg/cm{sup 2}, 1000 deg. C sous vide) d'un melange de poudres d'uranium et de nitrures superieurs d'uranium, un produit qui a ete identifie par diffraction de rayons X. On peut ainsi obtenir a volonte, soit le monocarbure UN, soit un cermet U{sub {alpha}}-UN dans le cas de compositions sous-stoechiometriques. Au contraire du monocarbure d'uranium prepare dans des conditions analogues, les produits obtenus ici, soumis a un traitement thermique, n'evoluent pas facilement vers un etat d'equilibre physico-chimique caracterise par l'existence de joints de grains rectilignes. On attribue ce phenomene a l'insolubilite de l'impurete UO{sub 2} dans UN. La densite, la durete, la conductibilite thermique de ces produits se revelent superieures a celles des nitrures d'uranium ou des cermets U-UN obtenus par les autres methodes. (auteur)

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

Energy Technology Data Exchange (ETDEWEB)

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

1965-07-01

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

Morphological analysis and modelling of sintering and of sintered materials

International Nuclear Information System (INIS)

Jernot, Jean-Paul

1982-01-01

This research thesis addresses the study of solid phase sintering of metallic powders, and aims at describing as precisely as possible the different involved matter transport mechanisms, first by using a thermodynamic approach to sintering. Sintering diagrams are also used to determine prevailing mechanisms. The microstructure of sintered materials has been studied by using image quantitative analysis, thus by using a morphological approach to sintering. Morphological parameters allow, on the one hand, the evolution of powders during sintering to be followed, and, on the other hand, sintered products to be correctly characterised. Moreover, the author reports the study of the evolution of some physical properties of sintered materials with respect to their microstructure parameters. This leads to the development of a modelling of the behaviour of these materials [fr

Contribution to the study of sputtering and damage of uranium dioxide by fast heavy ions; Contribution a l'etude de la pulverisation et de l'endommagement du dioxyde d'uranium par les ions lourds rapides

Energy Technology Data Exchange (ETDEWEB)

Schlutig, S

2001-03-01

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

Measurement of uranium dioxide thermophysical properties by the laser flash method

International Nuclear Information System (INIS)

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

2009-01-01

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

A thermal modelling of displacement cascades in uranium dioxide

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

The behaviour of uranium metal in hydrogen atmospheres

International Nuclear Information System (INIS)

Allen, G.C.; Stevens, J.C.H.

1988-01-01

The reaction between commercial H 2 and uranium metal leads to the formation of UO 2 due to traces of water vapour or oxygen. When extremely pure H 2 is used uranium hydride may be formed but, even with 99.9999% H 2 , uranium dioxide forms preferentially. The present work identifies the presence of UH 3 in the X-ray photoelectron spectrum of a uranium sample which has been exposed to ca. 10 10 L† H 2 at ca. 200 0 C. This spectrum indicates that the hydride possesses a high degree of covalency, since the oxidation state of uranium in UH 3 appears to be ca. 1.4. (author)

Electrochemical preparation of new uranium oxide phases

International Nuclear Information System (INIS)

Smolenskij, V.V.; Lyalyushkin, N.V.; Bove, A.L.; Komarov, V.K.; Kapshukov, I.I.

1992-01-01

Behaviour of uranium ions in oxidation states 3+ and 4+ in molten chlorides of alkali metals in the temperature range of 700-900 degC in the atmosphere of an inert gas was studied by the method of cyclic voltametry. It is shown that as a result of introduction of crystal uranium dioxide into the salt melt formation of uranium oxide ions of the composition UO + and UO 2+ occurs, the ions participating in electrode reactions and bringing about formation of the following uranium oxides on the cathode: UO and, presumably, U 3 O 4 . Oxides UO and U 3 O 4 are thermodynamically unstable at low temperatures and decompose into uranium oxide of the composition UO 2-x , where x varies from 0 to 0.05, and metal uranium

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Fabrication and testing of the sintered ceramic UO{sub 2} fuel - I - III, Part III - testing of sintered uranium dioxide properties dependent on the fabrication procedure; Izrada i ispitivanje keramickog goriva na bazi UO{sub 2}- I-III, III Deo - Ispitivanje osobina sinterovanog urandioksida u zavisnosti od procesa dobijanja

Energy Technology Data Exchange (ETDEWEB)

Novakovic, M; Ristic, M M [Institute of Nuclear Sciences Boris Kidric, Laboratorija za termotehniku reaktora, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

The objective of this task was testing the influence of some parameters on the properties of sintered UO{sub 2}. The influence of parameters tested were as follows: adhesives; pressure in the pressing procedure; temperature of sintering of the UO{sub 2} powder. Other parameters were chosen according to the theoretical study. Sintering was done in argon atmosphere. Characterization of the UO{sub 2} powder was performed meaning determining the needed chemical, physical and physico-chemical properties. Some new methods were developed within this task: SET method for measuring the specific surfaces, DTA, TGA, high-temperature torsion.

Etching of uranium dioxide in nitrogen trifluoride RF plasma glow discharge

Science.gov (United States)

Veilleux, John Mark

1999-10-01

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

Production and analysis of ultradispersed uranium oxide powders

Science.gov (United States)

Zajogin, A. P.; Komyak, A. I.; Umreiko, D. S.; Umreiko, S. D.

2010-05-01

Spectroscopic studies are made of the laser plasma formed near the surface of a porous body containing nanoquantities of uranium compounds which is irradiated by two successive laser pulses. The feasibility of using laser chemical methods for obtaining nanoclusters of uranium oxide particles in the volume of a porous body and the simultaneous possibility of determining the uranium content with good sensitivity are demonstrated. The thermochemical and spectral characteristics of the analogs of their compounds with chlorine are determined and studied. The possibility of producing uranium dioxides under ordinary conditions and their analysis in the reaction products is demonstrated.

Simulation of uranium oxides reduction kinetics by hydrogen. Reactivities of germination and growth

International Nuclear Information System (INIS)

Brun, C.

1997-01-01

The aim of this work is to simulate the reduction by hydrogen of the tri-uranium octo-oxide U 3 O 8 (obtained by uranium trioxide calcination) into uranium dioxide. The kinetics curves have been obtained by thermal gravimetric analysis, the hydrogen and steam pressures being defined. The geometrical modeling which has allowed to explain the trend of the kinetics curves and of the velocity curves is an anisotropic germination-growth modeling. The powder is supposed to be formed of spherical grains with the same radius. The germs of the new UO 2 phase appear at the surface of the U 3 O 8 grains with a specific germination frequency. The growth reactivity is anisotropic and is very large in the tangential direction to the grains surface. Then, the uranium dioxide growths inside the grain and the limiting step is the grain surface. The variations of the growth reactivity and of the germination specific frequency in terms of the gases partial pressures and of the temperature have been explained by two different mechanisms. The limiting step of the growth mechanism is the desorption of water in the uranium dioxide surface. Concerning the germination mechanism the limiting step is a water desorption too but in the tri-uranium octo-oxide surface. The same geometrical modeling and the same germination and growth mechanisms have been applied to the reduction of a tri-uranium octo-oxide obtained by calcination of hydrated uranium trioxide. The values of the germination specific frequency of this solid are nevertheless weaker than those of the solid obtained by direct calcination of the uranium trioxide. (O.M.)

Influence of sintering atmospheres on the aluminium sintering characteristics

International Nuclear Information System (INIS)

Mintzer, S.; Bermudez Belkys, S.

1993-01-01

This paper describes the aluminium powder (Al) cool compacted (at 95% from theoretical density) which was sintered at 903 K during 4 hours at different atmospheres; oxidizing (air), inert Argon (Ar), Nitrogen (N) and high vacuum. The results obtained show: a) porosity measurements; greater porosity when sintering in Ar and air. b) Metallographic and Scanning observations: many fine pores (< 1 μm) and pore lines distributed at random, at air sintering and greater pores distributed preferentially near the surface, in Ar and N atmospheres. c) Dimensional changes: tendency to contraction of the samples at N and vacuum sintering and expansion in Ar or air. d) Mechanical properties: greater strength and fluence stresses at air and N sintering. The analysis of the results is performed considering sintering modes in presence of an oxide layer and dropped inert gases. (Author)

Boric oxide or boric acid sintering aid for sintering ceramics

International Nuclear Information System (INIS)

Lawler, H.A.

1979-01-01

The invention described relates to the use of liquid sintering aid in processes involving sintering of ceramic materials to produce dense, hard articles having industrial uses. Although the invention is specifically discussed in regard to compositions containing silicon carbide as the ceramic material, other sinterable carbides, for example, titanium carbide, may be utilized as the ceramic material. A liquid sintering aid for densifying ceramic material is selected from solutions of H 3 BO 3 , B 2 O 3 and mixtures of these solutions. In sintering ceramic articles, e.g. silicon carbide, a shaped green body is formed from a particulate ceramic material and a resin binder, and the green body is baked at a temperature of 500 to 1000 0 C to form a porous body. The liquid sintering aid of B 2 O 3 and/or H 3 BO 3 is then dispersed through the porous body and the treated body is sintered at a temperature of 1900 to 2200 0 C to produce the sintered ceramic article. (U.K.)

Standard test method for determination of impurities in plutonium: acid dissolution, ion exchange matrix separation, and inductively coupled plasma-atomic emission spectroscopic (ICP/AES) analysis

CERN Document Server

American Society for Testing and Materials. Philadelphia

2003-01-01

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

Study of reducing pyrohydrolysis of uranyl fluoride into uranium dioxide

International Nuclear Information System (INIS)

Favre, P.

1977-06-01

The dry process studied in this paper for the preparation of UO 2 (for sintering) from UF 6 presents the following advantages on other wet or dry processes (fluidized beds or discontinuous processes): it is completely continuous, one chemical reactor only is required for the successive reactions hydrolysis, pyrolysis and reduction, it is possible to obtain various densities after sintering and particularly high densities. Safety, environmental, economical and technical aspects are also improved. Pyrohydrolysis and reduction reactions of UO 2 F 2 into UO 2 are studied because kinetics are not well known although they have been used for several years. Reaction temperature and pressure are examined for optimization. Influence of the gaseous mixture hydrogen and inert gas on reaction inhibition could lead to rate, in particular nitrogen flow should be reduced. Operation and product quality should be both improved. 68 refs [fr

Chemical reactions during sintering of Fe-Cr-Mn-Si-Ni-Mo-C-steels with special reference to processing in semi-closed containers

Directory of Open Access Journals (Sweden)

Cias A.

2015-01-01

Full Text Available Sintering of Cr, Mn and Si bearing steels has recently attracted both experimental and theoretical attention and processing in semiclosed containers has been reproposed. This paper brings together relevant thermodynamic data and considers the kinetics of some relevant chemical reactions. These involve iron and carbon, water vapour, carbon monoxide and dioxide, hydrogen and nitrogen of the sintering atmospheres and the alloying elements Cr, Mn, Mo and Si. The paper concludes by presenting mechanical properties data for three steels sintered in local microatmosphere with nitrogen, hydrogen, nitrogen-5% hydrogen and air as the furnace gas.

Physical chemistry and modelling of the sintering of actinide oxides

International Nuclear Information System (INIS)

Lechelle, Jacques

2013-01-01

This report gives a synthesis of the work I have carried out or to which I have numerically contributed to from 1996 up to 2012 in the Department of Plutonium Uranium and minor Actinides in Cadarache CEA Center. Their main goal is the study and the modeling of the sintering process of nuclear fuels which is the unifying thread of this document. Both in order to take into account the physical and chemical features of the actinide bearing oxide material and in order to combine the different transport phenomena leading to sintering, a sub-granular scale model is under development. Extension to a varying chemical composition as well as exchanges with the gaseous phase are foreseen. A simulation on a larger scale (pellet scale) is ongoing in the framework of a PhD thesis. Validation means have been tested with (U,Pu)O 2 material on the scale of the pellet (Small Angle Neutron Diffusion), on the scale of powder granules (X-Ray High Resolution Micro-Tomography) and with CeO 2 at the 'Institut de Chimie Separative' in Marcoule on a single crystal scale (Environmental Scanning Electron Microscope). The required microstructure homogeneity for nuclear fuels has led to a campaign of experimental studies about the role of Cr 2 O 3 as a sintering aid. Whole of these studies improve our understanding of fuel sintering and hence leads to an improved mastering of this process. (author) [fr

Kinetic study of the reaction of uranium with various carbon-containing gases

International Nuclear Information System (INIS)

Feron, G.

1963-09-01

The kinetic study of the reaction U + CO 2 and U + CO has been performed by a thermogravimetric method on a spherical uranium powder, in temperature ranges respectively from 460 to 690 deg. C and from 570 to 850 deg. C. The reaction with carbon dioxide leads to uranium dioxide. A carbon deposition takes place at the same time. The global reactions is the result of two reactions: U + 2 CO 2 → UO 2 + 2 CO U + CO 2 → UO 2 + C The reaction with carbon monoxide leads to a mixture of dioxide UO 2 , dicarbide UC 2 and free carbon. The main reaction can be written. U + CO → 1/2 UO 2 + 1/2 UC 2 The free carbon results of the disproportionation of the carbon monoxide. A remarkable separation of the two phases UO 2 and UC 2 can be observed. A mechanism accounting for the phenomenon has been proposed. The two reactions U + CO 2 and U + CO begin with a long germination period, after which, the reaction velocity seems to be limited in both cases by the ionic diffusion of oxygen through the uranium dioxide. (author) [fr

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

International Nuclear Information System (INIS)

Lewis, M.B.

1980-01-01

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

Evolution of microstructure of U-Mo alloys in as cast and sintered forms

International Nuclear Information System (INIS)

Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Kamath, H.S.; Dey, G.K.

2009-01-01

Over the years U 3 Si 2 compound dispersed in aluminium matrix has been successfully used as potential Low Enriched Uranium (LEU 235 ) base dispersion fuel in new research and test reactors and also for converting High Enriched Uranium (HEU > 85% U 235 ) cores to LEU in most of the existing research and test reactors. The maximum density achievable with U 3 Si 2 -AI dispersion fuel is around 4.8 g U cm -3 . To achieve a uranium density of 8.0 to 9.0 g U cm -3 in dispersion fuel with aluminium as matrix material, it is required to use γ-stabilized uranium metal powders. At Metallic Fuels Division, R and D efforts are on to develop these high density uranium alloys. Molybdenum plays a crucial role in metastabilising the γ-phase of uranium at room temperature which is very much evident when we see the microstructures of different U-Mo alloys with varying molybdenum concentration as solute atom. The paper describes the role of molybdenum in imparting metastability in U-Mo alloys from their microstructures in as cast and sintered forms. The paper also covers the role of tailored microstructure in U-Mo alloy for the purpose of hydriding and dehydriding treatment to generate alloy powders. (author)

Simulation of uranium oxides reduction kinetics by hydrogen. Reactivities of germination and growth; Modelisation de la cinetique de reduction d`oxydes d`uranium par l`hydrogene. Reactivites de germination et de croissance

Energy Technology Data Exchange (ETDEWEB)

Brun, C

1997-12-04

The aim of this work is to simulate the reduction by hydrogen of the tri-uranium octo-oxide U{sub 3}O{sub 8} (obtained by uranium trioxide calcination) into uranium dioxide. The kinetics curves have been obtained by thermal gravimetric analysis, the hydrogen and steam pressures being defined. The geometrical modeling which has allowed to explain the trend of the kinetics curves and of the velocity curves is an anisotropic germination-growth modeling. The powder is supposed to be formed of spherical grains with the same radius. The germs of the new UO{sub 2} phase appear at the surface of the U{sub 3}O{sub 8} grains with a specific germination frequency. The growth reactivity is anisotropic and is very large in the tangential direction to the grains surface. Then, the uranium dioxide growths inside the grain and the limiting step is the grain surface. The variations of the growth reactivity and of the germination specific frequency in terms of the gases partial pressures and of the temperature have been explained by two different mechanisms. The limiting step of the growth mechanism is the desorption of water in the uranium dioxide surface. Concerning the germination mechanism the limiting step is a water desorption too but in the tri-uranium octo-oxide surface. The same geometrical modeling and the same germination and growth mechanisms have been applied to the reduction of a tri-uranium octo-oxide obtained by calcination of hydrated uranium trioxide. The values of the germination specific frequency of this solid are nevertheless weaker than those of the solid obtained by direct calcination of the uranium trioxide. (O.M.) 45 refs.

Production of cerium dioxide microspheres by an internal gelation sol–gel method

Energy Technology Data Exchange (ETDEWEB)

Katalenich, Jeffrey A.

2017-03-27

An internal gelation sol-gel technique was used to prepare cerium dioxide microspheres with uniform diameters near 100 µm. In this process, chilled aqueous solutions containing cerium, hexamethylenetetramine (HMTA), and urea are transformed into a solid gel by heat addition and are subsequently washed, dried, and sintered to produce pure cerium dioxide. Cerous nitrate and ceric ammonium nitrate solutions were compared for their usefulness in microsphere production. Gelation experiments were performed with both cerous nitrate and ceric ammonium nitrate to determine desirable concentrations of cerium, HMTA, and urea in feed solutions as well as the necessary quantity of ammonium hydroxide added to cerium solutions. Analysis of the pH before and after sample gelation was found to provide a quantitative metric for optimal parameter selection along with subjective evaluations of gel qualities. The time necessary for chilled solutions to gel upon inserting into a hot water bath was determined for samples with a variety of parameters and also used to determine desirable formulations for microsphere production. A technique for choosing the optimal mixture of ceric ammonium nitrate, HMTA, and urea was determined using gelation experiments and used to produce microspheres by dispersion of the feed solution into heated silicone oil. Gelled spheres were washed to remove excess reactants and reaction products before being dried and sintered. X-ray diffraction of air-dried microspheres, sintered microspheres, and commercial CeO₂ powders indicated that air-dried and sintered spheres were pure CeO₂.

Uranium kernel formation via internal gelation

International Nuclear Information System (INIS)

Hunt, R.D.; Collins, J.L.

2004-01-01

In the 1970s and 1980s, U.S. Department of Energy (DOE) conducted numerous studies on the fabrication of nuclear fuel particles using the internal gelation process. These amorphous kernels were prone to flaking or breaking when gases tried to escape from the kernels during calcination and sintering. These earlier kernels would not meet today's proposed specifications for reactor fuel. In the interim, the internal gelation process has been used to create hydrous metal oxide microspheres for the treatment of nuclear waste. With the renewed interest in advanced nuclear fuel by the DOE, the lessons learned from the nuclear waste studies were recently applied to the fabrication of uranium kernels, which will become tri-isotropic (TRISO) fuel particles. These process improvements included equipment modifications, small changes to the feed formulations, and a new temperature profile for the calcination and sintering. The modifications to the laboratory-scale equipment and its operation as well as small changes to the feed composition increased the product yield from 60% to 80%-99%. The new kernels were substantially less glassy, and no evidence of flaking was found. Finally, key process parameters were identified, and their effects on the uranium microspheres and kernels are discussed. (orig.)

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

Science.gov (United States)

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

2018-03-01

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

Treatment of uranium turning with the controllable oxidizing process

International Nuclear Information System (INIS)

Shen Bingyi; Zhang Yonggang; Zhen Huikuan

1989-02-01

The concept, procedure and safety measures of the controllable oxidizing for uranium turning is described. The feasibility study on technological process has been made. The process provided several advantages such as: simplicity of operation, no pollution environment, safety, high efficiency and low energy consumption. The process can yield nuclear pure uranium dioxide under making no use of a great number of chemical reagent. It may supply raw material for fluoration and provide a simply method of treatment for safe store of uranium turning

Chemical treatment of ammonium fluoride solution in uranium reconversion plant

International Nuclear Information System (INIS)

Carvalho Frajndlich, E.U. de.

1992-01-01

A chemical procedure is described for the treatment of the filtrate, produced from the transformation of uranium hexafluoride (U F 6 ) into ammonium uranyl carbonate (AUC). This filtrate is an intermediate product in the U F 6 to uranium dioxide (U O 2 ) reconversion process. The described procedure recovers uranium as ammonium peroxide fluoro uranate (APOFU) by precipitation with hydrogen peroxide (H 2 O 2 ), and as later step, its calcium fluoride (CaF 2 ) co-precipitation. The recovered uranium is recycled to the AUC production plant. (author)

Pilot production of 325 kg of uranium carbide

International Nuclear Information System (INIS)

Clozet, C.; Dessus, J.; Devillard, J.; Guibert, M.; Morlot, G.

1969-01-01

This report describes the pilot fabrication of uranium carbide rods to be mounted in bundles and assayed in two channels of the EL 4 reactor. The fabrication process includes: - elaboration of uranium carbide granules by carbothermic reduction of uranium dioxide; - electron bombardment melting and continuous casting of the granules; - machining of the raw ingots into rods of the required dimensions; finally, the rods will be piled-up to make the fuel elements. Both qualitative and quantitative results of this pilot production chain are presented and discussed. (authors) [fr

Romania, producer and consumer of nuclear fuel

International Nuclear Information System (INIS)

Iuhas, Tiberius

1998-01-01

A historical sketch of the activity of Romanian Rare Metals Enterprises is presented stressing the valorization of rare metals like: - radioactive metals, uranium and thorium; - dispersed rare metals, molybdenum, monazite; - heavy and refractory metals, titanium and zirconium; rare earths, lanthanides and yttrics. The beginning and developing of research in the nuclear field is in closed relation to the existence on the domestic territory of important uranium ores the mining of which begun early in 1954. The exploitation of Baita-Bihor orebody was followed by that at Ciudanovita, Natra and Dobrei ores in Caras-Severin county. Concomitantly with the ore mining, geological research was developed covering vast areas of country's surface and using advanced investigation tools suitable for increasing depths. The next step in the nuclear fuel program was made by building a uranium concentrate (as ammonium or sodium diuranate) plant. Two purification units for processing the uranium concentrate to sintered uranium dioxide powder were completed and commissioned at Feldioara in 1986. The quality of the uranium dioxide product meets the quality standards requirements for CANDU type nuclear fuel as certified in 1994. Currently, part of the fuel load of Cernavoda reactor is fuel element clusters produced by Nuclear Fuel Plant at Pitesti of sintered powder processed at Feldioara. A list of strategic objectives of the Uranium National Company is presented among which: - maintaining the uranium mining and milling activities in close relation with the fuel requirements of Cernavoda NPP; continuing geological research in promising zones, to find new uranium orebodies, easy to mill cost effectively; decreasing the environmental impact in the geological research areas, in mining and transport affected areas and in the processing plants. The fuel demand of current operation of Cernavoda NPP Unit 1 as well as of future Unit 2 after commissioning are and will be satisfied by the

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

International Nuclear Information System (INIS)

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

1981-05-01

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

Depleted uranium oxides as spent-nuclear-fuel waste-package fill materials

International Nuclear Information System (INIS)

Forsberg, C.W.

1997-01-01

Depleted uranium dioxide fill inside the waste package creates the potential for significant improvements in package performance based on uranium geochemistry, reduces the potential for criticality in a repository, and consumes DU inventory. As a new concept, significant uncertainties exist: fill properties, impacts on package design, post- closure performance

Preparation of uranium-plutonium mixed nitride pellets with high purity

International Nuclear Information System (INIS)

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

1992-01-01

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

Effects of sintering temperature on the mechanical properties of sintered NdFeB permanent magnets prepared by spark plasma sintering

International Nuclear Information System (INIS)

Wang, G.P.; Liu, W.Q.; Huang, Y.L.; Ma, S.C.; Zhong, Z.C.

2014-01-01

Sintered NdFeB-based permanent magnets were fabricated by spark plasma sintering (SPS) and a conventional method to investigate the mechanical and magnetic properties. The experimental results showed that sintered NdFeB magnet prepared by the spark plasma sintering (SPS NdFeB) possesses a better mechanical properties compared to the conventionally sintered one, of which the maximum value of bending strength and Vickers hardness was 402.3 MPa and 778.1 MPa, respectively. The effects of sintering temperature on bending strength and Vickers hardness were investigated. It was shown that the bending strength firstly increases to the maximum value and then decreases with the increase of sintering temperature in a certain range. The investigations of microstructures and mechanical properties indicated that the unique sintering mechanism in the SPS process is responsible for the improvement of mechanical properties of SPS NdFeB. Furthermore, the relations between the mechanical properties and relevant microstructure have been analyzed based on the experimental fact. - Highlights: • Studied the sintering temperature effect on strengthening mechanism of NdFeB magnet firstly. • It showed that sintering temperature may effectively affect the mechanical properties. • The maximum bending strength and Vickers hardness was 402.3 MPa and 778.1 MPa, respectively

Effects of sintering temperature on the mechanical properties of sintered NdFeB permanent magnets prepared by spark plasma sintering

Energy Technology Data Exchange (ETDEWEB)

Wang, G.P., E-mail: wgp@jxnu.edu.cn [College of Physics and Communication Electronics, Jiangxi Normal University, Nanchang 330022 (China); Liu, W.Q. [Key Laboratory of Advanced Functional Materials Science and Engineering, Ministry of Education, Beijing University of Technology, Beijing 100022 (China); Huang, Y.L.; Ma, S.C.; Zhong, Z.C. [School of Materials Science and Engineering, Nanchang Hangkong University, Nanchang 330063 (China)

2014-01-15

Sintered NdFeB-based permanent magnets were fabricated by spark plasma sintering (SPS) and a conventional method to investigate the mechanical and magnetic properties. The experimental results showed that sintered NdFeB magnet prepared by the spark plasma sintering (SPS NdFeB) possesses a better mechanical properties compared to the conventionally sintered one, of which the maximum value of bending strength and Vickers hardness was 402.3 MPa and 778.1 MPa, respectively. The effects of sintering temperature on bending strength and Vickers hardness were investigated. It was shown that the bending strength firstly increases to the maximum value and then decreases with the increase of sintering temperature in a certain range. The investigations of microstructures and mechanical properties indicated that the unique sintering mechanism in the SPS process is responsible for the improvement of mechanical properties of SPS NdFeB. Furthermore, the relations between the mechanical properties and relevant microstructure have been analyzed based on the experimental fact. - Highlights: • Studied the sintering temperature effect on strengthening mechanism of NdFeB magnet firstly. • It showed that sintering temperature may effectively affect the mechanical properties. • The maximum bending strength and Vickers hardness was 402.3 MPa and 778.1 MPa, respectively.

Fugitive binder for nuclear fuel materials

International Nuclear Information System (INIS)

Gallivan, T.J.

1980-01-01

A compound consisting of ammonium cations and carbonate, bicarbonate, or carbamate anions, or a mixture of such compounds, is useful as a binder for uranium dioxide fuel pellets for which it is desired to maintain a certain degree of porosity, uniformity of pore size, a lack of interconnections between the pores, and the shape or configuration of the base material particles in the final article after sintering. Upon heating, these binders decompose into gases and leave substantially no impurities. A process for sintering green nuclear fuel pellets using these binders is provided. (LL)

Study of uranium dioxide pellets by micro-acoustic techniques

International Nuclear Information System (INIS)

Roque, V.

1999-01-01

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

Gravimetric determination of uranium in SALE samples

International Nuclear Information System (INIS)

Anon.

1981-01-01

As a participant in the Safeguards Analytical Laboratory Evaluation (SALE) program, the Analytical Chemistry Laboratory at General Atomic routinely assays uranium dioxide and uranyl nitrate SALE samples for uranium content. Gravimetric methods are relatively easy and inexpensive to apply when the samples for uranium content. Gravimetric methods are relatively easy and inexpensive to apply when the samples are free from substantial amounts of metallic impurities. Clearly the gravimetric procedure alone is not specific for uranium and must be enhanced by the use of impurity corrections. Emission spectrography is used routinely as the technique of choice for making such corrections. In cases where it is essential to assay specifically for uranium, the modified Davies-Gray titration using a weighed titrant method is applied. In this paper some essential features of these gravimetric and titrimetric procedures are discussed

Photoluminescence and hydrogen gas-sensing properties of titanium dioxide nanostructures synthesized by hydrothermal treatments

CSIR Research Space (South Africa)

Sikhwivhilu, LM

2012-03-01

Full Text Available Titanium dioxide (TiO2) nanostructures were synthesized by microwave-assisted and conventionally heated hydrothermal treatment of TiO2 powder. The tubular structures were converted to a rodlike shape by sintering the samples at various temperatures...

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

Energy Technology Data Exchange (ETDEWEB)

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

1975-01-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

Bartscher, W.

1982-01-01

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

Sinterability and microstructure evolution during sintering of ferrous powder mixtures

Directory of Open Access Journals (Sweden)

Kétner Bendo Demétrio

2013-01-01

Full Text Available The present work is focused on ferrous powder metallurgy and presents some results of a development of a suitable masteralloy for use as an additive to iron powder for the production of sintered steels. The masteralloy was produced by melting a powder mixture containing approximately Fe + 20% Ni + 20% Mn + 20% Si + 1% C (wt%, in order to obtain a cast billet that was converted into fine powder by crushing and milling. It was observed presence of SiC in the masteralloy after melting that is undesirable in the alloy. Si element should be introduced by using ferrosilicon. Sintered alloys with distinct contents of alloying elements were prepared by mixing the masteralloy powder to plain iron powder. Samples were produced by die compaction of the powder mixtures and sintering at 1200 °C in a differential dilatometer in order to record their linear dimensional behaviour during heating up and isothermal sintering, aiming at studying the sinterability of the compacts. Microstructure development during sintering was studied by SEM, XRD and microprobe analyses.

Calculation of the energy of stacking faults in uranium dioxide

International Nuclear Information System (INIS)

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

1976-01-01

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

Sintering of nonstoichiometric UO2

International Nuclear Information System (INIS)

Susnik, D.; Holc, J.

1983-01-01

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

Titanium Powder Sintering in a Graphite Furnace and Mechanical Properties of Sintered Parts

Directory of Open Access Journals (Sweden)

Changzhou Yu

2017-02-01

Full Text Available Recent accreditation of titanium powder products for commercial aircraft applications marks a milestone in titanium powder metallurgy. Currently, powder metallurgical titanium production primarily relies on vacuum sintering. This work reported on the feasibility of powder sintering in a non-vacuum furnace and the tensile properties of the as-sintered Ti. Specifically, we investigated atmospheric sintering of commercially pure (C.P. titanium in a graphite furnace backfilled with argon and studied the effects of common contaminants (C, O, N on sintering densification of titanium. It is found that on the surface of the as-sintered titanium, a severely contaminated porous scale was formed and identified as titanium oxycarbonitride. Despite the porous surface, the sintered density in the sample interiors increased with increasing sintering temperature and holding time. Tensile specimens cut from different positions within a large sintered cylinder reveal different tensile properties, strongly dependent on the impurity level mainly carbon and oxygen. Depending on where the specimen is taken from the sintered compact, ultimate tensile strength varied from 300 to 580 MPa. An average tensile elongation of 5% to 7% was observed. Largely depending on the interstitial contents, the fracture modes from typical brittle intergranular fracture to typical ductile fracture.

Fabrication of Cerium Oxide and Uranium Oxide Microspheres for Space Nuclear Power Applications

Energy Technology Data Exchange (ETDEWEB)

Jeffrey A. Katalenich; Michael R. Hartman; Robert C. O' Brien

2013-02-01

Cerium oxide and uranium oxide microspheres are being produced via an internal gelation sol-gel method to investigate alternative fabrication routes for space nuclear fuels. Depleted uranium and non-radioactive cerium are being utilized as surrogates for plutonium-238 (Pu-238) used in radioisotope thermoelectric generators and for enriched uranium required by nuclear thermal rockets. While current methods used to produce Pu-238 fuels at Los Alamos National Laboratory (LANL) involve the generation of fine powders that pose a respiratory hazard and have a propensity to contaminate glove boxes, the sol-gel route allows for the generation of oxide microsphere fuels through an aqueous route. The sol-gel method does not generate fine powders and may require fewer processing steps than the LANL method with less operator handling. High-quality cerium dioxide microspheres have been fabricated in the desired size range and equipment is being prepared to establish a uranium dioxide microsphere production capability.

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

International Nuclear Information System (INIS)

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

2013-01-01

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

The Influence of Sintering Temperature of Reactive Sintered (Ti, MoC-Ni Cermets

Directory of Open Access Journals (Sweden)

Marek Jõeleht

2015-09-01

Full Text Available Titanium-molybdenum carbide nickel cermets ((Ti, MoC-Ni were produced using high energy milling and reactive sintering process. Compared to conventional TiC-NiMo cermet sintering the parameters for reactive sintered cermets vary since additional processes are present such as carbide synthesis. Therefore, it is essential to acquire information about the suitable sintering regime for reactive sintered cermets. One of the key parameters is the final sintering temperature when the liquid binder Ni forms the final matrix and vacancies inside the material are removed. The influence of the final sintering temperature is analyzed by scanning electron microscopy. Mechanical properties of the material are characterized by transverse rupture strength, hardness and fracture toughness.DOI: http://dx.doi.org/10.5755/j01.ms.21.3.7179

Thermal stabilization of uranium mill tailings

International Nuclear Information System (INIS)

Dreesen, D.R.; Williams, J.M.; Cokal, E.J.

1981-01-01

The sintering of tailings at high temperatures (1200 0 C) has shown promise as a conditioning approach that greatly reduces the 222 Rn emanation of uranium mill tailings. The structure of thermally stabilized tailings has been appreciably altered producing a material that will have minimal management requirements and will be applicable to on-site processing and disposal. The mineralogy of untreated tailings is presented to define the structure of the original materials. Quartz predominates in most tailings samples; however, appreciable quantities of gypsum, clay, illite, or albites are found in some tailings. Samples from the Durango and Shiprock sites have plagioclase-type aluminosilicates and non-aluminum silicates as major components. The iron-rich vanadium tailings from the Salt Lake City site contain appreciable quantities of α-hematite and chloroapatite. The reduction in radon emanation power and changes in mineralogy as a function of sintering temperature are presented

Fabrication of Uranium Oxycarbide Kernels for HTR Fuel

International Nuclear Information System (INIS)

Barnes, Charles; Richardson, Clay; Nagley, Scott; Hunn, John; Shaber, Eric

2010-01-01

Babcock and Wilcox (B and W) has been producing high quality uranium oxycarbide (UCO) kernels for Advanced Gas Reactor (AGR) fuel tests at the Idaho National Laboratory. In 2005, 350-(micro)m, 19.7% 235U-enriched UCO kernels were produced for the AGR-1 test fuel. Following coating of these kernels and forming the coated-particles into compacts, this fuel was irradiated in the Advanced Test Reactor (ATR) from December 2006 until November 2009. B and W produced 425-(micro)m, 14% enriched UCO kernels in 2008, and these kernels were used to produce fuel for the AGR-2 experiment that was inserted in ATR in 2010. B and W also produced 500-(micro)m, 9.6% enriched UO2 kernels for the AGR-2 experiments. Kernels of the same size and enrichment as AGR-1 were also produced for the AGR-3/4 experiment. In addition to fabricating enriched UCO and UO2 kernels, B and W has produced more than 100 kg of natural uranium UCO kernels which are being used in coating development tests. Successive lots of kernels have demonstrated consistent high quality and also allowed for fabrication process improvements. Improvements in kernel forming were made subsequent to AGR-1 kernel production. Following fabrication of AGR-2 kernels, incremental increases in sintering furnace charge size have been demonstrated. Recently small scale sintering tests using a small development furnace equipped with a residual gas analyzer (RGA) has increased understanding of how kernel sintering parameters affect sintered kernel properties. The steps taken to increase throughput and process knowledge have reduced kernel production costs. Studies have been performed of additional modifications toward the goal of increasing capacity of the current fabrication line to use for production of first core fuel for the Next Generation Nuclear Plant (NGNP) and providing a basis for the design of a full scale fuel fabrication facility.

Ecological aspects of air pollution from an iron-sintering plant at Wawa, Ontario

Energy Technology Data Exchange (ETDEWEB)

Gordon, A G; Gorham, E

1963-01-01

At Wawa, in northern Ontario, vegetation has been damaged severely by sulphur dioxide pollution from an iron-sintering plant. Damage is mainly restricted to a narrow strip northeast from the sinter plant, since southwest winds are strongly predominant. It is traceable from the air for at least 20 miles in this direction and is estimated as severe within 11 miles and very severe within 5 miles. Within about 10 miles NE. From the sinter plant ground flora variety declines markedly, from about 20-40 species per 40 square meter quadrant beyond this distance to 0-1 species within 2 miles of the pollution source. At the same time sulphate in lake and pond waters increases greatly, from normal levels of about 0.2-0.3 milliequivalents per liter to more than 0.5 meq/l within 11 miles NE and up to 2 meq/l within 2 miles NE from the sinter plant. Waters within about 5 miles NE are strongly acid (pH 3.2-3.8), but are not low in calcium. Soluble sulphate in the surface soil rises sharply within about 4 miles NE from the pollution source, where, also, soil erosion is very pronounced, though traceable farther out. The phanerogms most tolerant of air pollution are Polygonum cilinode and Sambucus pubens, which are infrequent in the normal forest vegetation. In quadrant studies along a northeast transect, seedlings of Pinus strobus were not observed within 30 miles from the sinter plant, while those of Picea glauca, P. mariana, and Populus tremuliodes were not recorded within 15 miles.

ELECTROCHEMICAL STUDIES OF URANIUM METAL CORROSION MECHANISM AND KINETICS IN WATER

International Nuclear Information System (INIS)

Boudanova, Natalya; Maslennikov, Alexander; Peretroukhine, Vladimir F.; Delegard, Calvin H.

2006-01-01

During long-term underwater storage of low burn-up uranium metal fuel, a corrosion product sludge forms containing uranium metal grains, uranium dioxide, uranates and, in some cases, uranium peroxide. Literature data on the corrosion of non-irradiated uranium metal and its alloys do not allow unequivocal prediction of the paragenesis of irradiated uranium in water. The goal of the present work conducted under the program 'CORROSION OF IRRADIATED URANIUM ALLOYS FUEL IN WATER' is to study the corrosion of uranium and uranium alloys and the paragenesis of the corrosion products during long-term underwater storage of uranium alloy fuel irradiated at the Hanford Site. The elucidation of the physico-chemical nature of the corrosion of irradiated uranium alloys in comparison with non-irradiated uranium metal and its alloys is one of the most important aspects of this work. Electrochemical methods are being used to study uranium metal corrosion mechanism and kinetics. The present part of work aims to examine and revise, where appropriate, the understanding of uranium metal corrosion mechanism and kinetics in water

Thermal diffusivity and conductivity of thorium- uranium mixed oxides

Science.gov (United States)

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

2018-03-01

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

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

International Nuclear Information System (INIS)

Fang Chao; Cao Jianzhu; Sun Lifeng

2011-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Direct dissolution and supercritical fluid extraction of uranium from UO2 powder, granule, green pellet and sintered pellet

International Nuclear Information System (INIS)

Rao, Ankita; Kumar, Pradeep; Ramakumar, K.L.

2009-01-01

In the present work, direct dissolution and extraction of UO 2 from the solid rejects various stages of fuel fabrication viz. powder granules green pellet and, sintered pellet has been studied. Powder and granules could be easily dissolved in TBP-HNO 3 complex at 50 deg C., whereas in case of green and sintered pellets at elevated temperature at raised to 80 deg C in TBP-HNO 3 complex. With supercritical (SC) CO 2 alone the efficiency was ∼70%. But with SC CO 2 +2.5% TBP, the efficiency was ∼95% for powder and granules, and ∼60% for green and sintered pellets. Nearly complete extraction (∼99%) was achievable for SC CO 2 + 2.5 % TTA in all cases. The method has distinct advantage of elimination of acid usage and minimization of liquid waste generation. (author)

Activation of Chalcogens and Chalcogenides at Reactive Uranium Centers

OpenAIRE

Franke, Sebastian

2015-01-01

The coordination chemistry of uranium has experienced a tremendous recent increase of interest within the last three decades, likely due to the fact that complexes of trivalent uranium can effectively engage activation and functionalization of small molecules, such as carbon monoxide (CO), carbon dioxide (CO2), dinitrogen (N2), or dioxygen (O2). Many small molecules are of great biochemical and industrial relevance, but their thermodynamical stability requires high pressures and temperatures...

X-ray photoelectron spectroscopy of the uranium/oxygen system: Part 13

International Nuclear Information System (INIS)

Allen, G.C.; Stevens, J.C.H.

1987-02-01

The reaction between commercial H 2 and uranium metal leads to the formation of UO 2 due to traces of water vapour or oxygen. When extremely pure H 2 is used uranium hydride may be formed but, even with 99.9999% H 2 , uranium dioxide forms preferentially. The present work identifies the presence of UH 3 in the X-ray photoelectron spectrum of a uranium sample which has been exposed to ∼ 5 mbar H 2 at ∼ 200 0 C for 1 hour. This spectrum indicates that the hydride possesses a high degree of covalency, since the oxidation state of uranium in UH 3 appears to be ∼ 1.4. (U.K.)

Carbon dioxide, the feedstock for using renewable energy

Science.gov (United States)

Hashimoto, K.; Kumagai, N.; Izumiya, K.; Kato, Z.

2011-03-01

Extrapolation of world energy consumption between 1990 and 2007 to the future reveals the complete exhaustion of petroleum, natural gas, uranium and coal reserves on Earth in 2040, 2044, 2049 and 2054, respectively. We are proposing global carbon dioxide recycling to use renewable energy so that all people in the whole world can survive. The electricity will be generated by solar cell in deserts and used to produce hydrogen by seawater electrolysis at t nearby desert coasts. Hydrogen, for which no infrastructures of transportation and combustion exist, will be converted to methane at desert coasts by the reaction with carbon dioxide captured by energy consumers. Among systems in global carbon dioxide recycling, seawater electrolysis and carbon dioxide methanation have not been performed industrially. We created energy-saving cathodes for hydrogen production and anodes for oxygen evolution without chlorine formation in seawater electrolysis, and ideal catalysts for methane formation by the reaction of carbon dioxide with hydrogen. Prototype plant and industrial scale pilot plant have been built.

Decontamination of radioactive clothing using microemulsion in carbon dioxide

Energy Technology Data Exchange (ETDEWEB)

Yoo, Jaeryong; Jang, Jina; Park, Kwangheon; Kim, Hongdoo; Kim, Hakwon [Kyunghee Univ., Seoul (Korea, Republic of); Yim, Sanghak; Yoon, Weonseob [Ulchin Nuclear Power Site, Ulchin (Korea, Republic of)

2006-07-01

Nuclear power is intrinsically a clean energy source due to its high energy density and low generation of waste. However, as the nuclear industry grows, a variety of radioactive wastes are increased gradually. Major subjects include contaminated components, tools, equipment, containers and facilities as well as nuclear waste such as uranium scrap and radioactive clothing. The radioactive waste can be classified by its creation. There are Trans-Uranium Nuclides (TRU), Fission Products (FP) and corrosion products. Nuclear decontamination has become an important issue in the nuclear industry. The conventional methods have some problems such as the production of secondary wastes and the use of toxic solvents. We need to develop a new method of decontamination and suggest a use of microemulsion in carbon dioxide to overcome these disadvantages. The microemulsion is the clear solution that contains the water, surfactant and carbon dioxide. The surfactant surrounded the droplet into carbon dioxide and this state is thermodynamically stable. That is, the microemulsion has a structure similar to that of a conventional water-based surfactant system. Generally, the size of droplet is about 5 {approx} 10nm. The microemulsion is able to decontaminate radioactive waste so that the polar substance is removed by water and the non-polar substance is removed by carbon dioxide. After the decontamination process, the microemulsion is separated easily to surfactant and water by decreasing the pressure under the cloud point. This way, only radioactive wastes are left in the system. Cleaned carbon dioxide is then collected and reused. Thus, there are no secondary wastes. Carbon dioxide is considered an alternative process medium. This is because it is non-toxic, non-flammable, inexpensive and easy to handle. Additionally, the tunable properties of carbon dioxide through pressure and temperature control are versatile for use in extracting organic materials. In this paper, we examine the

Decontamination of radioactive clothing using microemulsion in carbon dioxide

International Nuclear Information System (INIS)

Yoo, Jaeryong; Jang, Jina; Park, Kwangheon; Kim, Hongdoo; Kim, Hakwon; Yim, Sanghak; Yoon, Weonseob

2006-01-01

Nuclear power is intrinsically a clean energy source due to its high energy density and low generation of waste. However, as the nuclear industry grows, a variety of radioactive wastes are increased gradually. Major subjects include contaminated components, tools, equipment, containers and facilities as well as nuclear waste such as uranium scrap and radioactive clothing. The radioactive waste can be classified by its creation. There are Trans-Uranium Nuclides (TRU), Fission Products (FP) and corrosion products. Nuclear decontamination has become an important issue in the nuclear industry. The conventional methods have some problems such as the production of secondary wastes and the use of toxic solvents. We need to develop a new method of decontamination and suggest a use of microemulsion in carbon dioxide to overcome these disadvantages. The microemulsion is the clear solution that contains the water, surfactant and carbon dioxide. The surfactant surrounded the droplet into carbon dioxide and this state is thermodynamically stable. That is, the microemulsion has a structure similar to that of a conventional water-based surfactant system. Generally, the size of droplet is about 5 ∼ 10nm. The microemulsion is able to decontaminate radioactive waste so that the polar substance is removed by water and the non-polar substance is removed by carbon dioxide. After the decontamination process, the microemulsion is separated easily to surfactant and water by decreasing the pressure under the cloud point. This way, only radioactive wastes are left in the system. Cleaned carbon dioxide is then collected and reused. Thus, there are no secondary wastes. Carbon dioxide is considered an alternative process medium. This is because it is non-toxic, non-flammable, inexpensive and easy to handle. Additionally, the tunable properties of carbon dioxide through pressure and temperature control are versatile for use in extracting organic materials. In this paper, we examine the

Synthesis and Optimization of the Sintering Kinetics of Actinide Nitrides

International Nuclear Information System (INIS)

Butt, Drryl P.; Jaques, Brian

2009-01-01

Research conducted for this NERI project has advanced the understanding and feasibility of nitride nuclear fuel processing. In order to perform this research, necessary laboratory infrastructure was developed; including basic facilities and experimental equipment. Notable accomplishments from this project include: the synthesis of uranium, dysprosium, and cerium nitrides using a novel, low-cost mechanical method at room temperature; the synthesis of phase pure UN, DyN, and CeN using thermal methods; and the sintering of UN and (U x , Dy 1-x )N (0.7 (le) X (le) 1) pellets from phase pure powder that was synthesized in the Advanced Materials Laboratory at Boise State University.

Synthesis and Optimization of the Sintering Kinetics of Actinide Nitrides

Energy Technology Data Exchange (ETDEWEB)

Drryl P. Butt; Brian Jaques

2009-03-31

Research conducted for this NERI project has advanced the understanding and feasibility of nitride nuclear fuel processing. In order to perform this research, necessary laboratory infrastructure was developed; including basic facilities and experimental equipment. Notable accomplishments from this project include: the synthesis of uranium, dysprosium, and cerium nitrides using a novel, low-cost mechanical method at room temperature; the synthesis of phase pure UN, DyN, and CeN using thermal methods; and the sintering of UN and (Ux, Dy1-x)N (0.7 ≤ X ≤ 1) pellets from phase pure powder that was synthesized in the Advanced Materials Laboratory at Boise State University.

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

Energy Technology Data Exchange (ETDEWEB)

Sauter, F

2001-02-01

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

Development of Novel Porous Sorbents for Extraction of Uranium from Seawater

Energy Technology Data Exchange (ETDEWEB)

Lin, Wenbin [Univ. of Chicago, IL (United States)

2017-05-25

Climate disruption is one of the greatest crises the global community faces in the 21st century. Alarming increases in CO₂, NO, SO₂ and particulate matter levels will have catastrophic consequences on the environment, food supplies, and human health if no action is taken to lessen their worldwide prevalence. Nuclear energy remains the only mature technology capable of continuous base-load power generation with ultralow carbon dioxide, nitric oxide, and sulfur dioxide emissions. Over the lifetime of the technology, nuclear energy outputs less than 1.5% the carbon dioxide emissions per gigawatt hour relative to coal—about as much as onshore wind power.1 However, in order for nuclear energy to be considered a viable option in the future, a stable supply of uranium must be secured. Current estimates suggest there is less than 100 years’ worth of uranium left in terrestrial ores (6.3 million tons) if current consumption levels remain unchanged.2 It is likely, however, that demand for nuclear fuel will rise as a direct consequence of the ratification of global climate accords. The oceans, containing approximately 4.5 billion tons of uranium (U) at a uniform concentration of ~3 ppb, represent a virtually limitless supply of this resource.3 Development of technologies to recover uranium from seawater would greatly improve the U resource availability, providing a U price ceiling for the current generation and sustaining the nuclear fuel supply for future generations. Several methods have been previously evaluated for uranium sequestration including solvent extraction, ion exchange, flotation, biomass collection, and adsorption; however, none have been found to be suitable for reasons including cost effectiveness, long term stability, and selectivity.4,5 While polymer beads and fibers have been functionalized with amidoxime functional groups to afford U adsorption capacities as high as 1.5 g U/kg,6 further discoveries are needed to make uranium

Development of Novel Porous Sorbents for Extraction of Uranium from Seawater

International Nuclear Information System (INIS)

Lin, Wenbin

2017-01-01

Climate disruption is one of the greatest crises the global community faces in the 21st century. Alarming increases in CO_2, NO, SO_2 and particulate matter levels will have catastrophic consequences on the environment, food supplies, and human health if no action is taken to lessen their worldwide prevalence. Nuclear energy remains the only mature technology capable of continuous base-load power generation with ultralow carbon dioxide, nitric oxide, and sulfur dioxide emissions. Over the lifetime of the technology, nuclear energy outputs less than 1.5% the carbon dioxide emissions per gigawatt hour relative to coal-about as much as onshore wind power.1 However, in order for nuclear energy to be considered a viable option in the future, a stable supply of uranium must be secured. Current estimates suggest there is less than 100 years' worth of uranium left in terrestrial ores (6.3 million tons) if current consumption levels remain unchanged.2 It is likely, however, that demand for nuclear fuel will rise as a direct consequence of the ratification of global climate accords. The oceans, containing approximately 4.5 billion tons of uranium (U) at a uniform concentration of ~3 ppb, represent a virtually limitless supply of this resource.3 Development of technologies to recover uranium from seawater would greatly improve the U resource availability, providing a U price ceiling for the current generation and sustaining the nuclear fuel supply for future generations. Several methods have been previously evaluated for uranium sequestration including solvent extraction, ion exchange, flotation, biomass collection, and adsorption; however, none have been found to be suitable for reasons including cost effectiveness, long term stability, and selectivity.4,5 While polymer beads and fibers have been functionalized with amidoxime functional groups to afford U adsorption capacities as high as 1.5 g U/kg,6 further discoveries are needed to make uranium extraction from seawater

Low Temperature Two-Steps Sintering (LTTSS) - an innovative method for consolidating porous UO2 pellets

International Nuclear Information System (INIS)

Sanjay Kumar, D.; Ananthasivan, K.; Senapati, Abhiram; Venkata Krishnan, R.

2015-01-01

Metallic uranium and its alloys are an important fuel for fast reactors. Presently, metallic uranium is being prepared using expensive fluoro-metallothermic process. Recent reports suggest that metal oxide could be reduced to metal using a novel electrochemical de-oxidation method and this could serve as attractive alternate for expensive metallothermic process. In view of which, a research program is being pursued in our Centre to develop an optimum process parameter for the scaled up preparation of metallic uranium efficiently. One of the important process parameter is the size, nature and distribution of porosity in the urania pellet. Essentially the ceramic form of the urania should encompass interconnected porosity that would allow percolation of melts into the UO 2 . However, the matrix density of the pellet should be high to ensure that it possesses good handling strength and is electrically conducting. Hence preparation of high dense porous UO 2 pellets was required. In this study, we report the preparation of porous UO 2 pellets possessing a very high matrix density by using the citrate gel-combustion method. The 'as-prepared' powders were consolidated at various compaction pressures as such and these pellets were sintered in 8 mol %Ar+H 2 gas with a flow rate of 250 mL/min at 1073 K for 30 min followed by soaking at 1473 K for 4 h with heating rate of 5 K min -1 in a molybdenum furnace. X-ray diffraction studies revealed that these pellets contained UO 2 . The morphological analysis sintered pellets was carried out by using Scanning Electron Microscope (M/s. Philips model XL 30, Netherlands). All these pellets were gold coated

Optimization of dissolution process parameters for uranium ore concentrate powders

Energy Technology Data Exchange (ETDEWEB)

Misra, M.; Reddy, D.M.; Reddy, A.L.V.; Tiwari, S.K.; Venkataswamy, J.; Setty, D.S.; Sheela, S.; Saibaba, N. [Nuclear Fuel Complex, Hyderabad (India)

2013-07-01

Nuclear fuel complex processes Uranium Ore Concentrate (UOC) for producing uranium dioxide powder required for the fabrication of fuel assemblies for Pressurized Heavy Water Reactor (PHWR)s in India. UOC is dissolved in nitric acid and further purified by solvent extraction process for producing nuclear grade UO{sub 2} powder. Dissolution of UOC in nitric acid involves complex nitric oxide based reactions, since it is in the form of Uranium octa oxide (U{sub 3}O{sub 8}) or Uranium Dioxide (UO{sub 2}). The process kinetics of UOC dissolution is largely influenced by parameters like concentration and flow rate of nitric acid, temperature and air flow rate and found to have effect on recovery of nitric oxide as nitric acid. The plant scale dissolution of 2 MT batch in a single reactor is studied and observed excellent recovery of oxides of nitrogen (NO{sub x}) as nitric acid. The dissolution process is automated by PLC based Supervisory Control and Data Acquisition (SCADA) system for accurate control of process parameters and successfully dissolved around 200 Metric Tons of UOC. The paper covers complex chemistry involved in UOC dissolution process and also SCADA system. The solid and liquid reactions were studied along with multiple stoichiometry of nitrous oxide generated. (author)

Science of sintering

International Nuclear Information System (INIS)

Kuczynski, G.

1977-01-01

Although the methods of integration of materials by sintering, have been used since the early history of humanity, the actual understanding of the process involved came only in the last three decades. As in the most human endeavors, the art preceded theory. The comprehension of the elementary processes occuring during sintering comes from the studies of model system. Although the elementary processes occuring during sintering are today quite well understood, the problem of shrinkage of a powder compact which was at the origin of Sintering Science is still far from solved. This is due to the complexity of the internal geometry of the compacts. The recent attempts to apply statistics to this problem, seem to offer some promise

A study on some properties of sintered stainless steel powder compacts with sintering conditions

International Nuclear Information System (INIS)

Lee, Bang Sik; Kim, Kwan Hyu; Lee, Doh Jae; Choi, Dap Chun

1986-01-01

Sintered specimens for the mechanical and corrosion tests were prepared from 316L, 410L and 434L stainless steel powder compacts with green densities in the range of 6.2∼7.0g/cm 3 . The experimental variables studied were green density, sintering atmosphere, temperature and time, type of lubricant used and cooling rate after sintering operation. Mechanical properties of green compacts and sintered specimens were evaluated. The corrosion tests were performed by potentiodynamic anodic polarization technique. Mechanical properties were very sensitive to the sintering atmosphere; sintering in dissociated ammonia resulted in the strengthing but embrittlement of sintered 316L, 410L and 434L strainless steel powder compacts. Their corrosion resistance was also decreased. The tensile strength was increased with increases in sintering time and temperature while the decreases in the yield strength were observed. The tensile properties of green compacts were shown to closely related to the green density. Addition of 1% acrawax as a lubricant was appeared to be most effective for the improvement of green strength. (Author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-11-01

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

Recent developments in the field of refractory fuels

International Nuclear Information System (INIS)

Accary, A.; Delmas, R.

1964-01-01

The main part of the work carried out in the field of ceramic fuels by the Commissariat a l'Energie Atomique during recent years, has been in the direction of uranium dioxide and the uranium-carbon alloys. Uranium dioxide is being studied with the aim of using it as a fuel in the first core of EL-4, in which an integrated thermal conductivity of 29 W/cm is expected at the hottest point for a surface temperature of about 750 C. We concentrated on developing a process for preparing a dioxide powder of suitable characteristics and for sintering this powder, and on evaluating the main properties of the material obtained in the light of the conditions under which they will be used - micro-structural aspect and pore distribution, - mechanical and thermal behaviour in cylindrical form, - control of excess oxygen in the industrial products, - behaviour of the gaseous fission products at high temperatures after or during the course of irradiation. Our aim in the case of uranium carbides has been to determine the conditions of industrial manufacturing of a suitable fuel with a composition close to This has led us to undertake a number of fundamental investigations into - the domain of existence of non-stoichiometric UC, - the influence of elements of O and N on the properties of the UC in which they are dissolved, - the compatibility of uranium-carbon alloys with the different metallic or ceramic materials used for the sheath, - the corrosion of uranium-carbon alloys by H 2 O and CO 2 , - methods of preparing high purity samples, - in-pile irradiation devices for the investigation of these materials in the region of possible operating temperatures. In parallel with these fundamental investigations, we have attempted to define a procedure for the fabrication of uranium-carbon alloys of composition very close to UC which would, in its industrial application, lead to better results than the existent methods from the technical or economical points of view (i.e. sintering or arc

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

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

International Nuclear Information System (INIS)

Knight, Travis; Anghaie, Samim

2002-01-01

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

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

International Nuclear Information System (INIS)

Michel, A.

2011-01-01

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Studies on supercritical fluid extraction of uranium from sodium diuranate

International Nuclear Information System (INIS)

Prabhat, Parimal; Vithal, G.K.; Rao, Ankita; Kumar, Pradeep; Tomar, B.S.

2014-01-01

Crude sodium diuranate (SDU) produced from phosphoric acid by solvent extraction process with di-2-ethyl hexyl phosphoric acid (D2EHPA) and tri-n-butyl phosphate(TBP) contains iron and other rare earth impurities along with uranium. For further use of this uranium for fuel fabrication and its subsequent use in nuclear reactors, it has to be purified up to nuclear grade ammonium diuranate (ADU) specifications. Conventionally crude SDU is being purified by dissolving it in nitric acid followed by solvent extraction process using TBP in diluent. Use of large amount of acid and organic solvents for industrial processes is an environmental concern. Nowadays there are efforts to minimize use of acid and organic solvents in industrial processes. Supercritical Fluid Extraction (SFE) of uranium from different matrices (solid as well as liquid) has been reported by several authors in recent years. Near complete extraction of uranium from UO 2 (powder, green pellet and sintered pellet) using TBP-HNO 3 adduct by SFE has been reported. We attempted to explore possibility to purify crude SDU to nuclear grade by SFE of uranium from crude SDU matrix and study the effect of different operational parameters, mode of extraction and complexation

Uranium doping and neutron irradiation of Bi-2223 superconduction tapes for improved critical current density

International Nuclear Information System (INIS)

Moss, S.D.; Wang, W.G.; Dou, S.X.; Weinstein, R.

1998-01-01

It is demonstrated that a combination of neutron irradiation with uranium doping introduce fission tracks through a Bi-2223 tape which act as effective pinning centres, leading to a substantial increase in critical current. Preliminary data suggests that the combination of uranium doping and neutron irradiation produces improved flux pinning in Bi-2223 tapes over neutron irradiation alone. Before irradiation, SEM, DTA and XRD analyses were performed on the tapes. Both before and after irradiation the trapped maximum magnetic flux was measured at 77K. Before neutron irradiation, uranium doping has no effect on critical current. Preliminary SEM data suggested that the uranium is homogeneously distributed throughout the oxide core of the tape. The presence of 2212 and other secondary phases in the doped tapes suggest further refinement of the sintering procedure is necessary. (authors)

Development of uranium industry in Romania

International Nuclear Information System (INIS)

Iuhas, Tiberiu

2000-01-01

The management of the uranium resources is performed in Romania by the National Uranium Company. The tasks to be done are: 1. management and protection of rare and radioactive metal ores in the exploitation areas; 2. mining, preparation, refining and trading the radioactive ores, as well as reprocessing the uranium stock from the uranium concentrate in the national reserve; 3. performing geologic and technologic studies in the exploitation areas; 4. performing studies and projects concerning the maintenance of the present facilities and unearthing new ores; 5. building industrial facilities; 6. carrying out technological transport; 7. importation-exportation operations; 8. performing micro-production activity in experimental research units; 9. personnel training; 10. medical assistance for the personnel; 11. environment protection. The company is organized as follows: 1.three branches for uranium ore mining, located at Suceava, Bihor and Banat; 2. one branch for geologic survey, located at Magurele; 3. one branch for uranium ore preparation and concentration and for refining uranium concentrates, located at Feldioara; 4. One group for mine conservation, closure and ecology, located at Bucuresti. The final product, sintered powder of UO 2 produced at Feldioara plant, was tested in 1994 by the Canadian partner and met successfully the required standards. The Feldioara plant was certified as supplier of raw material for CANDU nuclear fuel production and as such, Romania is the only authorized producer of CANDU nuclear fuel in Europe and the second in the world, after Canada. Maintaining the uranium production in Romania is justified by the existence of uranium ore resources, the declining of natural gas resources, lower costs per kWh for electric nuclear power as compared to fossil-fuel power production, the possibility for Romania to become an important supplier of CANDU nuclear fuel, the low environmental impact and high costs for total shutdown of activity, high

Assessment of the meteorological data and atmospheric dispersion estimates in the Ranger 1 Uranium Mining Environmental Impact Statement

International Nuclear Information System (INIS)

Clark, G.H.

1977-03-01

Wind records from Jabiru, Northern Territory, Australia have been re-analysed to give atmospheric dispersion estimates of sulphur dioxide and radioactive contaminants associated with a proposed uranium mining and milling operation. Revisions in the plume rise equations have led to lower annual average sulphur dioxide air concentrations than those presented in the Ranger 1 Uranium Mining Environmental Impact Statement. Likewise, the short term peak air concentrations of sulphur dioxide were all within the United States Environment Protection Agency air quality standards. Even though the radon gas inventory was revised upwards, predicted concentrations were only slightly higher than those in the RUMEIS. An attempt was made at a first estimate of the uranium dust source term caused by wind suspension from stockpiled ore and waste rock. In a preliminary analysis using a 'surface depletion' model, it was estimated that uranium dust air concentrations would be decreased by about an order of magnitude when dry deposition was included in the atmospheric dispersion model. Integrating over all sources, radionuclides and meteorological conditions, the annual radiation dose to members of the public in the Regional Centre is estimated to be a maximum of 5 per cent of the recommended annual limits. (author)

Kinetics of UO2 sintering

International Nuclear Information System (INIS)

Ristic, M.M.

1962-01-01

Detailed conclusions related to the UO 2 sintering can be drawn from investigating the kinetics of the sintering process. This report gives an thorough analysis of the the data concerned with sintering available in the literature taking into account the Jander and Arrhenius laws. This analysis completes the study of influence of the O/U ratio and the atmosphere on the sintering. Results presented are fundamentals of future theoretical and experimental work related to characterisation of the UO 2 sintering process

Model for the behaviour of thorium and uranium fuels at pelletization

International Nuclear Information System (INIS)

Ferreira Neto, Ricardo Alberto

2000-11-01

In this work, a model for the behaviour of thorium-uranium-mixed oxide microspheres in the pelletizing process is presented. This model was developed in a program whose objective was to demonstrate the viability of producing fissile material through the utilization of thorium in pressurized water reactors. This is important because it allows the saving of the strategic uranium reserves, and makes it possible the nuclear utilization of the large brazilian thorium reserves. The objective was to develop a model for optimizing physical properties of the microspheres, such as density, fracture strength and specific surface, so as to produce fuel pellets with microstructure, density, open porosity and impurity content, in accordance with the fuel specification. And, therefore, to adjust the sol-gel processing parameters in order to obtain these properties, and produce pellets with an optimized microstructure, adequate to a stable behaviour under irradiation. The model made it clear that to achieve this objective, it is necessary to produce microspheres with density and specific surface as small as possible. By changing the sol-gel processing parameters, microspheres with the desired properties were produced, and the model was experimentally verified by manufacturing fuel pellets with optimized microstructures, density, open porosity and impurity content, meeting the specifications for this new nuclear fuel for pressurized water reactors. Furthermore it was possible to obtain mathematical expressions that enables to calculate from the microspheres properties and the utilized compaction pressure, the sinter density that will be obtained in the sintered pellet and the necessary compaction pressure to reach the sintered density specified for the fuel. (author)

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

International Nuclear Information System (INIS)

Pages, Simon

2014-01-01

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

Reaction of uranium and plutonium carbides with nitrogen; Reaction avec l'azote des carbures d'uranium et de plutonium

Energy Technology Data Exchange (ETDEWEB)

Lorenzelli, R; Martin, A; Schickel, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1966-03-01

Uranium and plutonium carbides react with nitrogen during the grinding process preceding the final sintering. The reaction occurs even in argon atmospheres containing a few percent of residual nitrogen. The resulting contamination is responsible for the appearance of an equivalent quantity of higher carbide in the sintered products; nitrogen remains quantitatively in the monocarbide phase. UC can be transformed completely into nitride under a nitrogen pressure, at a temperature as low as 400 C. The reaction is more sluggish with PuC. The following reactions take places: UC + 0,8 N{sub 2} {yields}> UN{sub 1.60} + C and PuC + 0,5 N{sub 2} {yields} PuN + C. (authors) [French] Les carbures d'uranium et de plutonium reagissent avec l'azote au cours du broyage qui precede le frittage final. Cette reaction est sensible meme sous des atmospheres d'argon ne contenant que quelques pour cent d'azote. Cette contamination se traduit sur les produits frittes par l'apparition d'une quantite equivalente de carbure superieur, l'azote restant fixe quantitativement dans la phase monocarbure. On peut transformer entierement UC en nitrure par action de l'azote sous pression des 400 C. La reaction est plus difficile avec PuC. Les reactions sont les suivantes: UC + 0,8 N{sub 2} {yields} UN{sub 1.60} + C et PuC + 0,5 N{sub 2} {yields} PuN + C.

Procedure for the conversion of a metal oxide powder to a fine grained ceramic material

International Nuclear Information System (INIS)

Ferrell, L.J.

1978-01-01

A procedure for sintering metal oxides is described which gives a product with significantly smaller grain size and better grain size distribution than previous processes. The procedure is presented as applied to aluminium oxide, but it is also stated to be applicable to uranium dioxide. A pellet density of within 1/2 percent of the theoretical maximum can be obtained. No grinding or surface treatment of the pellets is necessary. (JIW)

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

Energy Technology Data Exchange (ETDEWEB)

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

1966-07-01

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

Master Sintering Surface: A practical approach to its construction and utilization for Spark Plasma Sintering prediction

Directory of Open Access Journals (Sweden)

Pouchly V.

2012-01-01

Full Text Available The sintering is a complex thermally activated process, thus any prediction of sintering behaviour is very welcome not only for industrial purposes. Presented paper shows the possibility of densification prediction based on concept of Master Sintering Surface (MSS for pressure assisted Spark Plasma Sintering (SPS. User friendly software for evaluation of the MSS is presented. The concept was used for densification prediction of alumina ceramics sintered by SPS.

Alternative sintering methods compared to conventional thermal sintering for inkjet printed silver nanoparticle ink

NARCIS (Netherlands)

Niittynen, J.; Abbel, R.; Mäntysalo, M.; Perelaer, J.; Schubert, U.S.; Lupo, D.

2014-01-01

In this contribution several alternative sintering methods are compared to traditional thermal sintering as high temperature and long process time of thermal sintering are increasing the costs of inkjet-printing and preventing the use of this technology in large scale manufacturing. Alternative

Flash sintering of ceramic materials

Science.gov (United States)

Dancer, C. E. J.

2016-10-01

During flash sintering, ceramic materials can sinter to high density in a matter of seconds while subjected to electric field and elevated temperature. This process, which occurs at lower furnace temperatures and in shorter times than both conventional ceramic sintering and field-assisted methods such as spark plasma sintering, has the potential to radically reduce the power consumption required for the densification of ceramic materials. This paper reviews the experimental work on flash sintering methods carried out to date, and compares the properties of the materials obtained to those produced by conventional sintering. The flash sintering process is described for oxides of zirconium, yttrium, aluminium, tin, zinc, and titanium; silicon and boron carbide, zirconium diboride, materials for solid oxide fuel applications, ferroelectric materials, and composite materials. While experimental observations have been made on a wide range of materials, understanding of the underlying mechanisms responsible for the onset and latter stages of flash sintering is still elusive. Elements of the proposed theories to explain the observed behaviour include extensive Joule heating throughout the material causing thermal runaway, arrested by the current limitation in the power supply, and the formation of defect avalanches which rapidly and dramatically increase the sample conductivity. Undoubtedly, the flash sintering process is affected by the electric field strength, furnace temperature and current density limit, but also by microstructural features such as the presence of second phase particles or dopants and the particle size in the starting material. While further experimental work and modelling is still required to attain a full understanding capable of predicting the success of the flash sintering process in different materials, the technique non-etheless holds great potential for exceptional control of the ceramic sintering process.

Effects of sintering atmosphere and initial particle size on sintering of gadolinia-doped ceria

International Nuclear Information System (INIS)

Batista, Rafael Morgado

2014-01-01

The effects of the sintering atmosphere and initial particle size on the sintering of ceria containing 10 mol% gadolinia (GdO 1.5 ) were systematically investigated. The main physical parameter was the specific surface area of the initial powders. Nanometric powders with three different specific surface areas were utilized, 210 m 2 /g, 36,2 m 2 /g e 7,4 m 2 /g. The influence on the densification, and micro structural evolution were evaluated. The starting sintering temperature was verified to decrease with increasing on the specific surface area of raw powders. The densification was accelerated for the materials with smaller particle size. Sintering paths for crystallite growth were obtained. Master sintering curves for gadolinium-doped ceria were constructed for all initial powders. A computational program was developed for this purpose. The results for apparent activation energy showed noticeable dependence with specific surface area. In this work, the apparent activation energy for densification increased with the initial particle size of powders. The evolution of the particle size distributions on non isothermal sintering was investigated by WPPM method. It was verified that the grain growth controlling mechanism on gadolinia doped ceria is the pore drag for initial stage and beginning of intermediate stage. The effects of the sintering atmosphere on the stoichiometry deviation of ceria, densification, microstructure evolution, and electrical conductivity were analyzed. Inert, oxidizing, and reducing atmospheres were utilized on this work. Deviations on ceria stoichiometry were verified on the bulk materials. The deviation verified was dependent of the specific surface area and sintering atmosphere. Higher reduction potential atmospheres increase Ce 3+ bulk concentration after sintering. Accelerated grain growth and lower electrical conductivities were verified when reduction reactions are significantly present on sintering. (author)

Simulation of the interaction between uranium dioxide and zircaloy

International Nuclear Information System (INIS)

Denis, A.; Garcia, E.A.

1984-01-01

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

Micromechanical simulation of Uranium dioxide polycrystalline aggregate behaviour under irradiation

International Nuclear Information System (INIS)

Pacull, J.

2011-02-01

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

Production of pure sintered alumina

International Nuclear Information System (INIS)

Rocha, J.C. da; Huebner, H.W.

1982-01-01

With the aim of optimizing the sintering parameters, the strength of a large number of alumina samples was determined which were produced under widely varying sintering conditions and with different amounts of MgO content. The strength as a function of sintering time or temperature was found to go through a maximum. With increasing time, this maximum is shifted to lower temperatures, and with decreasing temperature to longer times. Data pairs of sintering times and temperatures which yeld the strength maximum were determined. The value of the strength at the maximum remains unchanged. The strength is high (= 400 MN/m 2 , at a grain size of 3 um and a porosity of 2 per cent) and comparable to foreign aluminas produced for commercial purposes, or even higher. The increase in the sintering time from 1 h to 16 h permits a reduction of the sintering temperature from 1600 to 1450 0 C without losing strength. The practical importance of this fact for a production of sintered alumina on a large scale is emphasized. (Author) [pt

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

In situ leach method for recovering uranium and related values

International Nuclear Information System (INIS)

Yan, T.Y.

1981-01-01

A process is provided for in-situ leaching of uranium from a calcium-containing clay which does not result in contamination of the clay formation by any cations not already present. A lixiviant is prepared by dissolving carbon dioxide into water having essentially the same cationic composition as that of the formation connate water. The solution is injected along with an oxidant, for example oxygen, into the formation. Calcium that has become dissolved in the lixiviant must be removed to control the pH, preferably by the addition of lime in a calcium precipitator. After calcium removal the lixiviant is filtered to remove suspended solids and is passed through an ion exchange resin or other uranium extraction means. The barren solution goes to a mix tank where carbon dioxide is added, and the fresh lixiviant is injected along with additional oxidant into the formation

Sintering Theory and Practice

Science.gov (United States)

German, Randall M.

1996-01-01

Although sintering is an essential process in the manufacture of ceramics and certain metals, as well as several other industrial operations, until now, no single book has treated both the background theory and the practical application of this complex and often delicate procedure. In Sintering Theory and Practice, leading researcher and materials engineer Randall M. German presents a comprehensive treatment of this subject that will be of great use to manufacturers and scientists alike. This practical guide to sintering considers the fact that while the bonding process improves strength and other engineering properties of the compacted material, inappropriate methods of control may lead to cracking, distortion, and other defects. It provides a working knowledge of sintering, and shows how to avoid problems while accounting for variables such as particle size, maximum temperature, time at that temperature, and other problems that may cause changes in processing. The book describes the fundamental atomic events that govern the transformation from particles to solid, covers all forms of the sintering process, and provides a summary of many actual production cycles. Building from the ground up, it begins with definitions and progresses to measurement techniques, easing the transition, especially for students, into advanced topics such as single-phase solid-state sintering, microstructure changes, the complications of mixed particles, and pressure-assisted sintering. German draws on some six thousand references to provide a coherent and lucid treatment of the subject, making scientific principles and practical applications accessible to both students and professionals. In the process, he also points out and avoids the pitfalls found in various competing theories, concepts, and mathematical disputes within the field. A unique opportunity to discover what sintering is all about--both in theory and in practice What is sintering? We see the end product of this thermal

The industrial application of a uranium dioxide electrode

International Nuclear Information System (INIS)

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

1975-01-01

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

Preparation of Uranium Dioxide by Electrochemical Reduction in Ammonium Carbonate Solutions and Subsequent Precipitation; Preparation de bioxyde d'uranium par reduction electrochimique dans des solutions de carbonate d'ammonium et precipitation; Prigotovlenie dvuokisi urana metodom ehlektrokhimicheskogo vosstanovleniya v rastvore karbonata ammoniya s posleduyushchim osazhdeniem; Preparacion de dioxido de uranio por reduccion electroquimica en soluciones de carbonato amonico u precipitacion subsiguiente

Energy Technology Data Exchange (ETDEWEB)

Pravdic, V.; Branica, M.; Pucar, Z. [Department of Physical Chemistry, Rudjer Boskovic Institute, Zagreb, Yugoslavia (Serbia)

1963-11-15

Experiments in a small scale electrolysis cell on cathodic reduction of uranium (VI) to uranium (IV) show the possibility of an efficient way to obtain uranium (IV) in carbonate solutions. From this solution uranium (IV) hydrous oxide precipitates by merely raising the temperature. To obtain larger quantities of material needed for technological testing, a scale-up of the process was attempted. An electrolysis cell of hard PVC (polyvinylchloride) was constructed with a mercury pool cathode of approximately 2.5 dm{sup 2} and platinum anodes. The catholyte was separated from the anolyte by cationexchange membranes. The catholyte was circulated between two 50-1 reservoirs and streamed toward the vigorously stirred mercury cathode. The working potential of mercury was controlled against an Ag/AgCl/KC1 (sat.) reference electrode, the potential being held constant at -1.5 V. The current efficiency is approximately 90%; the power consumed for the reduction process is about 0.8 kWh/kg of uranium dioxide. After the electrolysis was completed the precipitation was initiated only by heating the deeply green clear solution up to 70 deg. C in a separate all-glass vessel of 60-1 volume. From 50, 1 of the catholyte solution 1 kg of a centrifuged product (containing about 20% of water) was obtained. The coulometric analysis of the oxygen-uranium ratio always gave results in the range of 2.04 to 2.09. By the procedure described uranium (IV) hydrous oxide is selectively precipitated, and the oxygen-uranium ratio in the precipitate was found to be independent of the degree of completion of the reduction. The product was identified as the alpha phase of uranium dioxide by the X-ray powder diffraction. Experiments in sintering and characterization of uranium dioxide thus obtained for the ceramic nuclear fuel requirements are under way. (author) [French] Des experiences faites dans une petite cellule d'electrolyse sur la reduction cathodique d'uranium (VI) en uranium (IV) montrent qu

Fabrication of Cr-doped UO2 Fuel Pellet using Liquid Phase Sintering

International Nuclear Information System (INIS)

Kim, Dong Joo; Yang, Jae Ho; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Oh, Jang Soo; Koo, Yang Hyun

2013-01-01

An enhancement of the thermal conductivity of a pellet can be obtained by the addition of a higher thermal conductive material in the pellet. In addition, the resistance to the PCI can be increased through a plasticity increase of the pellet. Thermal conductivity of ceramic materials is generally lower than that of metallic materials. The thermal conductivity of uranium oxide which is a typical ceramic material is low as well. The steep temperature gradient in the fuel pellet results from the low thermal conductivity. Therefore, the thermal conductivity improvement of a nuclear fuel pellet can enhance the fuel performance in various aspects. The lower centerline temperature of a fuel pellet affects the enhancement of fuel safety as well as fuel pellet integrity during nuclear reactor operation. Besides, the nuclear reactor power can be uprated due to the higher safety margin. So, many researches to enhance the thermal conductivity of nuclear fuel pellet have been performed in various ways. To improve the thermal conductivity of UO 2 pellet, an appropriate arrangement of the high thermal conductive material in UO 2 matrix is one of the various methods. We intended to control a placement of chromium as the high thermal conductive material. The metallic chromium and chromium oxide were arranged in a grain boundary of UO 2 using a liquid phase sintering method. The liquid phase sintering of Cr-doped UO 2 pellet could be adjusted using a control of an oxygen potential in sintering atmosphere

Canada's nuclear fuel industry: An overview. Background paper

International Nuclear Information System (INIS)

Nixon, A.

1993-11-01

Canada was among the first countries to mine and process uranium-bearing ores. Such ores contain trace amounts of radium, which was in great demand for medical treatment and for use by research laboratories in the early part of the century. For the last half century, the same basic processes have been used to extract uranium from its ores and convert it to a form suitable for use in nuclear reactors. The process described here is that currently in use in Canada. Mining can take a variety of forms, from open-pit to deep, hard-rock. Mining is typically the most costly step in the process, particularly for lower-grade ores. The ore is crushed and ground in the mill to the consistency of fine sand from which the uranium is extracted chemically to produce the impure concentrate known as yellowcake. In the next step, the impure uranium concentrate is chemically refined into highly purified, nuclear-grade, uranium trioxide (UO 3 ). Uranium trioxide is then converted, in two separate chemical processes, into uranium dioxide (UO 2 ) which is destined for domestic consumption and uranium hexafluoride (UF 6 ) which is exported. In Canada, fabrication is the final step of the fuel production process. Uranium dioxide powder is compressed and sintered into very dense ceramic pellets which are then sealed in zirconium tubes and assembled into fuel bundles for Candu reactors. This background paper will review the Canadian nuclear fuels industry. 1 fig

Kinetic study of the reaction of uranium with various carbon-containing gases; Etude cinetique de la reaction sur l'uranium de differents gaz carbones

Energy Technology Data Exchange (ETDEWEB)

Feron, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1963-09-15

The kinetic study of the reaction U + CO{sub 2} and U + CO has been performed by a thermogravimetric method on a spherical uranium powder, in temperature ranges respectively from 460 to 690 deg. C and from 570 to 850 deg. C. The reaction with carbon dioxide leads to uranium dioxide. A carbon deposition takes place at the same time. The global reactions is the result of two reactions: U + 2 CO{sub 2} {yields} UO{sub 2} + 2 CO U + CO{sub 2} {yields} UO{sub 2} + C The reaction with carbon monoxide leads to a mixture of dioxide UO{sub 2}, dicarbide UC{sub 2} and free carbon. The main reaction can be written. U + CO {yields} 1/2 UO{sub 2} + 1/2 UC{sub 2} The free carbon results of the disproportionation of the carbon monoxide. A remarkable separation of the two phases UO{sub 2} and UC{sub 2} can be observed. A mechanism accounting for the phenomenon has been proposed. The two reactions U + CO{sub 2} and U + CO begin with a long germination period, after which, the reaction velocity seems to be limited in both cases by the ionic diffusion of oxygen through the uranium dioxide. (author) [French] L'etude cinetique des reactions U sol + CO{sub 2} gaz et U sol + CO gaz a ete effectuee par thermogravirnetrie sur une poudre d'uranium a grains spheriques, les domaines de temperature etudies s'etendant respectivement de 460 a 690 deg. C et de 570 a 850 deg. C. L'action du dioxyde de carbone conduit au dioxyde d'uranium UO{sub 2}; il se produit en meme temps un depot de carbone. La reaction globale resulte des deux reactions: U + 2 CO{sub 2} {yields} UO{sub 2} + 2 CO U + CO{sub 2} {yields} UO{sub 2} + C Le mono-oxyde de carbone conduit a un melange de dioxyde UO{sub 2}, de dicarbure UC{sub 2} et de carbone libre. La reaction principale s'ecrit: U + CO {yields} 1/2 UO{sub 2} + 1/2 UC{sub 2} Le carbone libre provient de la dismutation du mono-oxyde de carbone. On observe une separation remarquable des deux phases UO{sub 2} et UC{sub 2}; un mecanisme rendant compte de ce phenomene a

A novel method for the preparation of uranium metal, oxide and carbide via electrolytic amalgamation

International Nuclear Information System (INIS)

Wang, L.C.; Lee, H.C.; Lee, T.S.; Lai, W.C.; Chang, C.T.

1978-01-01

A solid uranium amalgam was prepared electrolytically using a two-compartment cell separated with an ion exchange membrane for the purpose of regulating pH value within a narrowly restricted region of 2 to 3. The mercury cathode was kept at -1.8V vs SCE during electrolysis. The thereby obtained amalgam containing as high as 1.9gm U/ml Hg is easily converted into uranium metal by heating in vacuo above 1300 0 C. Uranium dioxide and uranium monocarbide could be easily obtained at relatively low temperature by reacting the amalgam with water vapor and methane. (author)

Strain-enhanced sintering of iron powders

Energy Technology Data Exchange (ETDEWEB)

Amador, D.R.; Torralba, J.M. [Universidad Carlos III de Madrid, Departamento de Ciencias de Materiales e Ingenieria Metalurgica, Leganes, Madrid (Spain); Monge, M.A.; Pareja, R. [Universidad Carlos III de Madrid, Departamento de Fisica, Madrid (Spain)

2005-02-01

Sintering of ball-milled and un-milled Fe powders has been investigated using dilatometry, X-ray, density, and positron annihilation techniques. A considerable sintering enhancement is found in milled powders showing apparent activation energies that range between 0.44 and 0.80 eV/at. The positron annihilation results, combined with the evolution of the shrinkage rate with sintering temperature, indicate generation of lattice defects during the sintering process of milled and un-milled powders. The sintering enhancement is attributed to pipe diffusion along the core of moving dislocations in the presence of the vacancy excess produced by plastic deformation. Positron annihilation results do not reveal the presence of sintering-induced defects in un-milled powders sintered above 1200 K, the apparent activation energy being in good agreement with that for grain-boundary diffusion in {gamma}-Fe. (orig.)

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

International Nuclear Information System (INIS)

Garnier, J.E.; Begej, S.

1979-04-01

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

A review of the rates of reaction of unirradiated uranium in gaseous atmospheres

International Nuclear Information System (INIS)

Pearce, R.J.

1989-10-01

The review collates available quantitative rate data for the reaction of unirradiated uranium in dry and moist air, steam and carbon dioxide based atmospheres at temperatures ranging from room temperature to above the melting point of uranium. Reactions in nitrogen and carbon monoxide are also considered. The aim of the review is to provide a compilation of base data for the hazard analysis of fault conditions relating to Magnox fuel. (author)

The composition and character of oxycarbide phase in uranium metal

International Nuclear Information System (INIS)

Liu Kezhao; Lai Xinchun; Yu Yong; Ni Ranfu

1999-08-01

The oxide layer of uranium metal formed by vacuum heating were examined with X-ray photoelectron spectroscopy (XPS) and Auger Electron Spectroscopy (AES). XPS results indicated that the air-exposed surface of the oxide layer were mainly consisted of UO 2 and free carbon. After the air-exposed surface were removed by low energy argon ion sputtering, C1s spectra shifted from 284.8 eV to 281.8 eV, indicating the existence of carbide phase. AES results of C(KVV) Auger transitions confirmed this result. Resolved and fitted using a combination of Gaussian and Lorentzian peak shape, U4f 7/2 spectra showed that three uranium chemical states existed in the layer, there were uranium dioxide, uranium carbide (or oxycarbide, UC x O 1-x ) and uranium metal phase. Calculated the AES data by relatively sensitive factor, the composition of oxycarbide was given as UC 0.41+-0.04 O 0.62+-0.01

Optimization of process parameters in precipitation for consistent quality UO2 powder production

International Nuclear Information System (INIS)

Tiwari, S.K.; Reddy, A.L.V.; Venkataswamy, J.; Misra, M.; Setty, D.S.; Sheela, S.; Saibaba, N.

2013-01-01

Nuclear reactor grade natural uranium dioxide powder is being produced through precipitation route, which is further processed before converting into sintered pellets used in the fabrication of PHWR fuel assemblies of 220 and 540 MWe type reactors. The process of precipitating Uranyl Nitrate Pure Solution (UNPS) is an important step in the UO 2 powder production line, where in soluble uranium is transformed into solid form of Ammonium Uranate (AU), which in turn reflects and decides the powder characteristics. Precipitation of UNPS with vapour ammonia is being carried out in semi batch process and process parameters like ammonia flow rate, temperature, concentration of UNPS and free acidity of UNPS are very critical and decides the UO 2 powder quality. Variation in these critical parameters influences powder characteristics, which in turn influences the sinterability of UO 2 powder. In order to get consistent powder quality and sinterability the critical parameter like ammonia flow rate during precipitation is studied, optimized and validated. The critical process parameters are controlled through PLC based automated on-line data acquisition systems for achieving consistent powder quality with increased recovery and production. The present paper covers optimization of process parameters and powder characteristics. (author)

Studies involving direct heating of uranium and plutonium oxides by microwaves

Energy Technology Data Exchange (ETDEWEB)

Mallik, G K; Malav, R K; Karande, A P; Bhargava, V K; Kamath, H S [Bhabha Atomic Research Centre, Tarapur (India). Advanced Fuel Fabrication Facility

1997-08-01

Nuclear fuel fabrication and recovery of nuclear materials from scraps generated during fabrication involve heating steps like dewaxing, sintering, roasting of scraps, calcination, etc. The dielectric properties of uranium and plutonium oxides place them in the category of materials which are susceptible to absorption of microwaves. The studies were carried out to explore the microwave heating technique for these steps required in nuclear fuel fabrication and scrap recovery laboratories. (author). 1 ref.

Solid-state sintering of tungsten heavy alloys

International Nuclear Information System (INIS)

Gurwell, W.E.

1994-10-01

Solid-state sintering is a technologically important step in the fabrication of tungsten heavy alloys. This work addresses practical variables affecting the sinterability: powder particle size, powder mixing, and sintering temperature and time. Compositions containing 1 to 10 micrometer (μM) tungsten (W) powders can be fully densified at temperatures near the matrix solidus. Blending with an intensifier bar provided good dispersion of elemental powders and good as-sintered mechanical properties under adequate sintering conditions. Additional ball milling increases powder bulk density which primarily benefits mold and die filling. Although fine, 1 μm W powder blends have high sinterability, higher as-sintered ductilities are reached in shorter sintering times with coarser, 5 μm W powder blends; 10μm W powder blends promise the highest as-sintered ductilities due to their coarse microstructural W

Present state and problems of uranium fuel fabrication businesses

International Nuclear Information System (INIS)

Yuki, Akio

1981-01-01

The businesses of uranium fuel fabrication converting uranium hexafluoride to uranium dioxide powder and forming fuel assemblies are the field of most advanced industrialization among nuclear fuel cycle industries in Japan. At present, five plants of four companies engage in this business, and their yearly sales exceeded 20 billion yen. All companies are planning the augmentation of installation capacity to meet the growth of nuclear power generation. The companies of uranium fuel fabrication make the nuclear fuel of the specifications specified by reactor manufacturers as the subcontractors. In addition to initially loaded fuel, the fuel for replacement is required, therefore the demand of uranium fuel is relatively stable. As for the safety of enriched uranium flowing through the farbicating processes, the prevention of inhaling uranium powder by workers and the precaution against criticality are necessary. Also the safeguard measures are imposed so as not to convert enriched uranium to other purposes than peacefull ones. The strict quality control and many times of inspections are carried out to insure the soundness of nuclear fuel. The growth of the business of uranium fuel fabrication and the regulation of the businesses by laws are described. As the problems for the future, the reduction of fabrication cost, the promotion of research and development and others are pointed out. (Kako, I.)

Microwave Combustion and Sintering Without Isostatic Pressure. Topical Report August 1, 1995 - October 30, 1996

International Nuclear Information System (INIS)

Ebadian, M.A.; Monroe, N.D.H.

1998-01-01

This investigation involves a study of the influence of key processing parameters on the heating of materials using microwave energy. Selective and localized heating characteristics of microwaves will be utilized in the sintering of ceramics without hydrostatic pressure. In addition, combustion synthesis will be studied for the production of powders, carbides, and nitrides by combining two or more solids or a solid and a gas to form new materials. The insight gained from the interaction of microwaves with various materials will be utilized in the mobilization and subsequent redeposition of uranium

Microwave sintering of hydroxyapatite-based composites

International Nuclear Information System (INIS)

Fang, Y.; Roy, D.M.; Cheng, J.; Roy, R.; Agrawal, D.K.

1993-01-01

Composites of hydroxyapatite/partially stabilized zirconia (HAp/PSZ) and hydroxyapatite/silicon carbide whiskers (HAp/SiC) were sintered at 1100-1200 degrees C by microwave at 2.45 GHz. Characterization of the sintered composites was carried out by density, microstructure, phase composition, and fracture toughness measurements. The results show that although not yet fully densified, a much higher sintered density in the HAp/PSZ composite was achieved by microwave sintering than by conventional sintering at the same temperature. A relative density of 93% was achieved by 20 min. microwave processing at 1200 degrees C. Comparatively, 2 h conventional sintering of the same material at 1200 degrees C led to only 75.5% relative density. K IC of this microwave sintered HAp/PSZ of 93% density was found to be 3.88 MPa√m, which is 250% of the value for pure HAp of the same density. A further increase in K IC could be expected if full or nearly full densification was achieved. Sintering of PSZ particles in the HAp/PSZ composite was also observed in the microwave processed sample. Microwave sintering of HAp/SiC was not successful in the current study due to the oxidation of SiC in air at high temperature. 8 refs., 4 figs., 1 tab

Inversion defects in MgAl2O4 elaborated by pressureless sintering, pressureless sintering plus hot isostatic pressing, and spark plasma sintering

International Nuclear Information System (INIS)

Mussi, A.; Granger, G. Bernard; Addad, A.; Benameur, N.; Beclin, F.; Bataille, A.

2009-01-01

The distribution of inversion defects of Al was investigated in dense magnesium-aluminate spinel elaborated by pressureless sintering, pressureless sintering plus hot isostatic pressing, and spark plasma sintering. This study was conducted by energy electron loss spectroscopy analyses and more particularly by energy loss near edge structure investigations of the Al-L 2,3 edge. Several aspects are discussed with the purpose of understanding why charged defects dispersal reveals a special configuration.

Obtention of uranium tetrafluoride from effluents generated in the hexafluoride conversion process

International Nuclear Information System (INIS)

Silva Neto, J.B.; Urano de Carvalho, E.F.; Durazzo, M.; Riella, H.G.

2009-01-01

Full text: The uranium silicide (U3Si2) fuel is produced from uranium hexafluoride (UF6) as the primary raw material. The uranium tetrafluoride (UF4) and metallic uranium are the two subsequent steps. There are two conventional routes for UF4 production: the first one reduces the uranium from the UF6 hydrolysis solution by adding stannous chloride (SnCl2). The second one is based on the hydrofluorination of solid uranium dioxide (UO2) produced from the ammonium uranyl carbonate (AUC). This work introduces a third route, a dry way route which utilizes the recovering of uranium from liquid effluents generated in the uranium hexafluoride reconversion process adopted at IPEN/CNEN-SP. Working in the liquid phase, this route comprises the recovery of ammonium fluoride by NH4HF2 precipitation. The crystallized bifluoride is added to the solid UO2 to get UF4, which returns to the metallic uranium production process and, finally, to the U3Si2 powder production. The UF4 produced by this new route was chemically and physically characterized and will be able to be used as raw material for metallic uranium production by magnesiothermic reduction. (author)

Behaviour of magnesium and two magnesium alloys heated in a carbon dioxide flow

International Nuclear Information System (INIS)

Boussion, M.-L.; Darras, R.; Leclercq, D.

1959-01-01

Magnesium is a particularly attractive material for sheathing uranium fuel elements in nuclear reactors in order to avoid uranium hot temperature oxidation by the cooling fluid. As this cooling fluid will be carbon dioxide at the (future) Marcoule plants, a thorough study of magnesium and magnesium alloys behaviour when heated by carbon dioxide at a 400 C temperature, have been completed. Tests on three materials (Mg, Mg-Zr and Mg-Zr-Zn) have been performed with CO 2 up to a temperature of 550 C, at atmospheric pressure in the presence of a certain amount of oxygen and nitrogen (in order to study the influence of these impurities), and at a pressure of 15 kg / cm 2 . Oxidation results are detailed. Reprint of a paper published in 'Revue de Metallurgie', LVI, n. 1, 1959, p. 61-67

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

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

International Nuclear Information System (INIS)

Zhou Wei; Liu Kezhao; Yang Jiangrong; Xiao Hong

2009-01-01

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

Literature information applicable to the reaction of uranium oxides with chlorine to prepare uranium tetrachloride

International Nuclear Information System (INIS)

Haas, P.A.

1992-02-01

The reaction of uranium oxides and chlorine to prepare anhydrous uranium tetrachloride (UCl 4 ) are important to more economical preparation of uranium metal. The most practical reactions require carbon or carbon monoxide (CO) to give CO or carbon dioxide (CO 2 ) as waste gases. The chemistry of U-O-Cl compounds is very complex with valances of 3, 4, 5, and 6 and with stable oxychlorides. Literature was reviewed to collect thermochemical data, phase equilibrium information, and results of experimental studies. Calculations using thermodynamic data can identify the probable reactions, but the results are uncertain. All the U-O-Cl compounds have large free energies of formation and the calculations give uncertain small differences of large numbers. The phase diagram for UCl 4 -UO 2 shows a reaction to form uranium oxychloride (UOCl 2 ) that has a good solubility in molten UCl 4 . This appears more favorable to good rates of reaction than reaction of solids and gases. There is limited information on U-O-Cl salt properties. Information on the preparation of titanium, zirconium, silicon, and thorium tetrachlorides (TiCl 4 , ZrCl 4 , SiCl 4 , ThCl 4 ) by reaction of oxides with chlorine (Cl 2 ) and carbon has application to the preparation of UCl 4

Magnesium bicarbonate as an in situ uranium lixiviant

International Nuclear Information System (INIS)

Sibert, J.W.

1984-01-01

In the subsurface solution mining of mineral values, especially uranium, in situ, magnesium bicarbonate leaching solution is used instead of sodium, potassium and ammonium carbonate and bicarbonates. The magnesium bicarbonate solution is formed by combining carbon dioxide with magnesium oxide and water. The magnesium bicarbonate lixivant has four major advantages over prior art sodium, potassium and ammonium bicarbonates

Comparison of Ti(C,N)-based cermets processed by hot-pressing sintering and conventional pressureless sintering

International Nuclear Information System (INIS)

Xu, Qingzhong; Ai, Xing; Zhao, Jun; Qin, Weizhen; Wang, Yintao; Gong, Feng

2015-01-01

Highlights: • The HP sintered Ti(C,N)-based cermets exhibit high hardness with fine grain size. • The PLS sintered cermets possess high mechanical properties with low porosity. • The applied pressure can rearrange particles and contribute to grain refinement. • The heating rate can greatly affect the solid and liquid phase sintering of cermets. - Abstract: A suitable sintering method is important to obtain the Ti(C,N)-based cermets with superior properties. In this paper, Ti(C,N)-based cermets were fabricated by hot-pressing sintering (HP) and conventional pressureless sintering (PLS) technology, respectively, to investigate the influence of different sintering methods on the microstructure and mechanical properties of cermets materials. The microstructure, fracture morphology, indention cracks and phase composition were observed and detected using scanning electron microscope (SEM), energy dispersive spectroscopy (EDS) and X-ray diffraction (XRD). The transverse rupture strength (TRS), Vickers hardness (HV) and fracture toughness (K IC ) were also measured. The results reveal that all of the Ti(C,N)-based cermets exhibit core–rim microstructures with black cores, white cores and grey rims embedded into metal binder phases. The grain size of the samples fabricated by HP is much finer and the structure is more compact than those fabricated by PLS, while there exist pores in the HP sintered samples. The sintering process has no influence on the phase composition of cermets, but affects the phase content and crystallinity. The samples fabricated by PLS present higher transverse rupture strength, fracture toughness and density than samples fabricated by HP. However, the HP sintered samples possess a higher hardness

The uranium industry of South Africa

International Nuclear Information System (INIS)

McLean, C.S.

1994-01-01

This paper was originally published in 1954 and is reproduced in this centenary issue of the journal of the South African Institute of Mining and Metallurgy. South Africa's economy was (and is) based on mining. The early history of the uranium mining industry (until 1954) is discussed in detail, together with its status and economy. The first quantitative assessment of the uranium potential of the Witwatersrand goldfield was made in 1945 when it was reported that South Africa had one of the largest low-grade uranium fields in the world. The first metallurgical plants brought considerable benefit to the area. The process of uranium extraction was basically similar to that employed in the recovery of gold. It could be divided into the same three main headings: agitation, filtration and precipitation. It was predicted that the program, in full swing, would possibly consume as much as 20,000 tons of manganese ore a month, as the extraction process requires dioxide. It was for this reason that manganese recovery plants have been incorporated in the process. Other materials that were to be used in large quantities were lime, limestone, animal glue and water. Considering the increasing importance of uranium in the economy of the country, the question of secrecy was becoming a problem. At that time the demand for South African uranium was guaranteed by a ten-year agreement with the British and American authorities. 3 figs

Plutonium oxides and uranium and plutonium mixed oxides. Carbon determination

International Nuclear Information System (INIS)

Anon.

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

Gas chromatographic method fr determination of carbon in metallic uranium

International Nuclear Information System (INIS)

Nikol'skij, V.A.; Markov, V.K.; Evseeva, T.I.; Cherstvenkova, E.P.

1983-01-01

Gas chromatographic device to determine carbon in metal uranium is developed. Burnout unite, permitting to load in the burnout tube simultaneously quite a few (up to 20) weight amounts of materials to be burned is a characteristic feature of the device. As a result amendments for control experiment and determination limit are decreased. The time of a single determination is also reduced. Conditions of carbon burn out from metal uranium are studied and temperature and time of complete extraction of carbon in the form of dioxide from weight amount into gaseous phase are established

Production of sintered alumina from powder; optimization of the sinterized parameters for the maximum mechanical resistence

International Nuclear Information System (INIS)

Rocha, J.C. da.

1981-02-01

Pure, sinterized alumina and the optimization of the parameters of sinterization in order to obtain the highest mechanical resistence are discussed. Test materials are sinterized from a fine powder of pure alumina (Al 2 O 3 ), α phase, at different temperatures and times, in air. The microstructures are analysed concerning porosity and grain size. Depending on the temperature or the time of sinterization, there is a maximum for the mechanical resistence. (A.R.H.) [pt

Development of DU-AGG (Depleted Uranium Aggregate)

Energy Technology Data Exchange (ETDEWEB)

Lessing, P.A.

1995-09-01

Depleted uranium oxide (UO{sub 2} or U0{sub 3}) powder was mixed with fine mineral additives, pressed, and heated to about 1,250{degree}C. The additives were chemically constituted to result in an iron-enriched basalt (IEB). Melting and wetting of the IEB phase caused the urania powder compact to densify (sinter) via a liquid phase sintering mechanism. An inorganic lubricant was found to aid in green-forming of the body. Sintering was successful in oxidizing (air), inert (argon), or reducing (dry hydrogen containing) atmospheres. The use of ground U0{sub 3} powders (93 vol %) followed by sintering in a dry hydrogen-containing atmosphere significantly increased the density of samples (bulk density of 8.40 g/cm{sup 3} and apparent density of 9.48 g/cm{sup 3}, open porosity of 11.43%). An improvement in the microstructure (reduction in open porosity) was achieved when the vol % of U0{sub 3} was decreased to 80%. The bulk density increased to 8.59 g/cm{sup 3}, the apparent density decreased slightly to 8.82 g/cm{sup 3} (due to increase of low density IEB content), while the open porosity decreased to an excellent number of 2.78%. A representative sample derived from 80 vol % U0{sub 3} showed that most pores were closed pores and that, overall, the sample achieved the excellent relative density value of 94.1% of the estimated theoretical density (composite of U0{sub 2} and IEB). It is expected that ground powders of U0{sub 3} could be successfully used to mass produce lowcost aggregate using the green-forming technique of briquetting.

Magnesium and uranium ignition in different gaseous atmospheres

International Nuclear Information System (INIS)

Darras, R.; Baque, P.; Leclercq, D.

1960-01-01

Magnesium, uranium and some of their alloys burning temperatures have been systematically determined in an air or carbon dioxide atmosphere, either dry or wet. Two different ways of heating have been used: either continuously rising up the temperature, or heating to and then maintaining a constant temperature. The results are clearly different in the two cases. Besides, if moisture has little effect on the magnesium burning temperatures in air, it does lower them by about 130-140 deg. C in CO 2 . The differences of sight between the burning of magnesium and uranium have been noticed; this leads to distinguish between an 'ignition' and an 'inflammation'. (author) [fr

New method for conversion of uranium hexafluoride to uranium dioxide

International Nuclear Information System (INIS)

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

1987-01-01

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

Synthesis of Uranium-based Microspheres for Transmutation of Minor Actinides

International Nuclear Information System (INIS)

Daniels, Henrik; Neumeier, Stefan; Modolo, Giuseppe

2010-01-01

Utilisation of the internal gelation process is a promising perspective for the fabrication of advanced nuclear fuels containing minor actinides (MA). The formulation of appropriate precursor solutions for this process is an important step towards a working process as the chemistry of uranium-MA systems is quite complex. In this work, actinide surrogates were utilised for basic research on their influence on the system. The ceramics obtained through thermal treatment of the gels were characterised to optimise the calcination and sintering process. (authors)

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

International Nuclear Information System (INIS)

Elhardt, W.

1979-01-01

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

Recycling of mill scale in sintering process

Directory of Open Access Journals (Sweden)

El-Hussiny N.A.

2011-01-01

Full Text Available This investigation deals with the effect of replacing some amount of Baharia high barite iron ore concentrate by mill scale waste which was characterized by high iron oxide content on the parameters of the sintering process., and investigation the effect of different amount of coke breeze added on sintering process parameters when using 5% mill scale waste with 95% iron ore concentrate. The results of this work show that, replacement of iron ore concentrate with mill scale increases the amount of ready made sinter, sinter strength and productivity of the sinter machine and productivity at blast furnace yard. Also, the increase of coke breeze leads to an increase the ready made sinter and productivity of the sintering machine at blast furnace yard. The productivity of the sintering machine after 5% decreased slightly due to the decrease of vertical velocity.

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)

Sintering of beryllium oxide

International Nuclear Information System (INIS)

Caillat, R.; Pointud, R.

1955-01-01

This study had for origin to find a process permitting to manufacture bricks of beryllium oxide of pure nuclear grade, with a density as elevated as possible and with standardized shape. The sintering under load was the technique kept for the manufacture of the bricks. Because of the important toxicity of the beryllium oxide, the general features for the preliminary study of the sintering, have been determined while using alumina. The obtained results will be able to act as general indication for ulterior studies with sintering under load. (M.B.) [fr

Sintering studies on iron-carbon-copper compacts

Directory of Open Access Journals (Sweden)

Perianayagam Philomen-D-Anand Raj

2016-01-01

Full Text Available Sintered Iron-Carbon-Copper parts are among the most widely used powder metallurgy product in automobile. In this paper, studies have been carried out to find out the sintering characteristics of iron-carbon-copper compacts when sintered in nitrogen atmosphere. The effects of various processing parameters on the sintering characteristics were studied. The various processing parameters considered were compaction pressure, green density and sintering temperature. The sintering characteristics determined were sintered density, porosity, dimensional change, micro hardness and radial crush strength. The results obtained have been discussed on the basis of micro structural observations. The characteristics of SEM fractography were also used to determine the mechanism of fracture. The fracture energy is strongly dependent on density of the compact.

Laser sintering of copper nanoparticles

International Nuclear Information System (INIS)

Zenou, Michael; Saar, Amir; Ermak, Oleg; Kotler, Zvi

2014-01-01

Copper nanoparticle (NP) inks serve as an attractive potential replacement to silver NP inks in functional printing applications. However their tendency to rapidly oxidize has so far limited their wider use. In this work we have studied the conditions for laser sintering of Cu-NP inks in ambient conditions while avoiding oxidation. We have determined the regime for stable, low-resistivity copper (< ×3 bulk resistivity value) generation in terms of laser irradiance and exposure duration and have indicated the limits on fast processing. The role of pre-drying conditions on sintering outcome has also been studied. A method, based on spectral reflectivity measurements, was used for non-contact monitoring of the sintering process evolution. It also indicates preferred spectral regions for sintering. Finally, we illustrated how selective laser sintering can generate high-quality, fine line (<5 µm wide) and dense copper circuits. (paper)

New materials through a variety of sintering methods

Science.gov (United States)

Jaworska, L.; Cyboroń, J.; Cygan, S.; Laszkiewicz-Łukasik, J.; Podsiadło, M.; Novak, P.; Holovenko, Y.

2018-03-01

New sintering techniques make it possible to obtain materials with special properties that are impossible to obtain by conventional sintering techniques. This issue is especially important for ceramic materials for application under extreme conditions. Following the tendency to limit critical materials in manufacturing processes, the use of W, Si, B, Co, Cr should be limited, also. One of the cheapest and widely available materials is aluminum oxide, which shows differences in phase composition, grain size, hardness, strain and fracture toughness of the same type of powder, sintered via various methods. In this paper the alumina was sintered using the conventional free sintering process, microwave sintering, Spark Plasma Sintering (SPS), high pressure-high temperature method (HP-HT) and High Pressure Spark Plasma Sintering (HP SPS). Phase composition analysis, by X-ray diffraction of the alumina materials sintered using various methods, was carried out. For the conventional sintering method, compacts are composed of α-Al2O3 and θ-Al2O3. For compacts sintered using SPS, microwave and HP-HT methods, χ-Al2O3 and γ-Al2O3 phases were additionally present. Mechanical and physical properties of the obtained materials were compared between the methods of sintering. On the basis of images from scanning electron microscope quantitative analysis was performed to determine the degree of grain growth of alumina after sintering.

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

Ternary carbide uranium fuels for advanced reactor design applications

International Nuclear Information System (INIS)

Knight, Travis; Anghaie, Samim

1999-01-01

Solid-solution mixed uranium/refractory metal carbides such as the pseudo-ternary carbide, (U, Zr, Nb)C, hold significant promise for advanced reactor design applications because of their high thermal conductivity and high melting point (typically greater than 3200 K). Additionally, because of their thermochemical stability in a hot-hydrogen environment, pseudo-ternary carbides have been investigated for potential space nuclear power and propulsion applications. However, their stability with regard to sodium and improved resistance to attack by water over uranium carbide portends their usefulness as a fuel for advanced terrestrial reactors. An investigation into processing techniques was conducted in order to produce a series of (U, Zr, Nb)C samples for characterization and testing. Samples with densities ranging from 91% to 95% of theoretical density were produced by cold pressing and sintering the mixed constituent carbides at temperatures as high as 2650 K. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-15

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

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

International Nuclear Information System (INIS)

Yao, Aining; Chu, Taiwei

2016-01-01

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

Pressureless sintering of whisker-toughened ceramic composites

Science.gov (United States)

Tiegs, T.N.

1993-05-04

A pressureless sintering method is disclosed for use in the production of whisker-toughened ceramic composites wherein the sintered density of composites containing up to about 20 vol. % SiC whiskers is improved by reducing the average aspect ratio of the whiskers to from about 10 to about 20. Sintering aids further improve the density, permitting the production of composites containing 20 vol. % SiC with sintered densities of 94% or better of theoretical density by a pressureless sintering method.

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)

Influence of sintering temperature on mechanical properties of spark plasma sintered pre-alloyed Ti-6Al-4 V powder

Energy Technology Data Exchange (ETDEWEB)

Muthuchamy, A.; Patel, Paridh; Rajadurai, M. [VIT Univ., Vellore, Tamil Nadu (India); Chaurisiya, Jitendar K. [NIT, Suratkal (India); Annamalai, A. Raja [VIT Univ., Vellore, Tamil Nadu (India). Centre for Innovative Manufacturing Research

2018-04-01

Spark plasma sintering provides faster heating that can create fully, or near fully, dense samples without significant grain growth. In this study, pre-alloyed Ti-6Al-4 V powder compact samples produced through field assisted sintering in a spark plasma sintering machine are compared as a function of consolidation temperature. The effect of sintering temperature on the densification mechanism, microstructural evolution and mechanical properties of spark plasma sintered Ti-6Al-4 V alloy compacts was investigated in detail. The compact, sintered at 1100 C, exhibited near net density, highest hardness and strength as compared to the other compacts processed at a temperature lower than 1100 C.

Laser sintering of metal powders on top of sintered layers under multiple-line laser scanning