Scale up issues involved with the ceramic waste form: ceramic-container interactions and ceramic cracking quantification

International Nuclear Information System (INIS)

Bateman, K. J.; DiSanto, T.; Goff, K. M.; Johnson, S. G.; O'Holleran, T.; Riley, W. P. Jr.

1999-01-01

Argonne National Laboratory is developing a process for the conditioning of spent nuclear fuel to prepare the material for final disposal. Two waste streams will result from the treatment process, a stainless steel based form and a ceramic based form. The ceramic waste form will be enclosed in a stainless steel container. In order to assess the performance of the ceramic waste form in a repository two factors must be examined, the surface area increases caused by waste form cracking and any ceramic/canister interactions that may release toxic material. The results indicate that the surface area increases are less than the High Level Waste glass and any toxic releases are below regulatory limits

Non-polluting treatment of uranium effluents from the alkaline digestion of an uranium ore containing sulfur

International Nuclear Information System (INIS)

Berger, Bernard.

1978-01-01

New non-polluting process for treating uranium effluents from the alkaline digestion of an uranium ore containing sulphur, which makes it possible (a) to extract and obtain relatively pure uranium and (b) to process the digestion liquor freed from the uranium and containing in an aqueous solution a mixture of alkaline carbonate and/or bicarbonate and sodium sulphate, consisting in the selective extraction of the sodium sulphate present and the recycling of the liquor free of SO 4 = ions, containing in solution the sole carbonates and/or bicarbonates involved, towards the digestion of the ore [fr

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

Immobilization of INEL low-level radioactive wastes in ceramic containment materials

International Nuclear Information System (INIS)

Seymour, W.C.; Kelsey, P.V.

1978-11-01

INEL low-level radioactive wastes have an overall chemical composition that lends itself to self-containment in a ceramic-based material. Fewer chemical additives would be needed to process the wastes than to process high-level wastes or use a mixture containment method. The resulting forms of waste material could include a basalt-type glass or glass ceramic and a ceramic-type brick. Expected leach resistance is discussed in relationshp to data found in the literature for these materials and appears encouraging. An overview of possible processing steps for the ceramic materials is presented

Thermal expansion of ceramic samples containing natural zeolite

Science.gov (United States)

Sunitrová, Ivana; Trník, Anton

2017-07-01

In this study the thermal expansion of ceramic samples made from natural zeolite is investigated. Samples are prepared from the two most commonly used materials in ceramic industry (kaolin and illite). The first material is Sedlec kaolin from Czech Republic, which contains more than 90 mass% of mineral kaolinite. The second one is an illitic clay from Tokaj area in Hungary, which contains about 80 mass% of mineral illite. Varying amount of the clay (0 % - 50 %) by a natural zeolite from Nižný Hrabovec (Slovak Republic), containing clinoptilolite as major mineral phase is replaced. The measurements are performed on cylindrical samples with a diameter 14 mm and a length about 35 mm by a horizontal push - rod dilatometer. Samples made from pure kaolin, illite and zeolite are also subjected to this analysis. The temperature regime consists from linear heating rate of 5 °C/min from 30 °C to 1100 °C. The results show that the relative shrinkage of ceramic samples increases with amount of zeolite in samples.

Purification method for calcium fluoride containing uranium

International Nuclear Information System (INIS)

Ogami, Takeshi

1998-01-01

Calcium fluoride (CaF 2 ) containing uranium is heated in an electrolytic bath having a cathode and an anode to form a molten salt, and the molten salt is electrolytically reduced to form metal uranium deposited on the surface of the cathode. The calcium fluoride molten salt separated by the deposition of generated metal uranium on the surface of the cathode is solidified by cooling. The solidified calcium fluoride is recovered. When metal uranium is deposited on the surface of the cathode by the electrolytic reduction of the molten salt, impurities such as plutonium and neptunium are also deposited on the surface of the anodes entrained by the metal uranium. Impurities having high vapor pressures such as americium and strontium are evaporated and removed from the molten salts. Then, nuclides such as uranium can thus be separated and recovered, and residual CaF 2 can be recovered in a state easily storable and reutilizable. (T.M.)

EXAFS and XANES analysis of plutonium and cerium edges from titanate ceramics for fissile materials disposal

International Nuclear Information System (INIS)

Fortner, J. A.; Kropf, A. J.; Bakel, A. J.; Hash, M. C.; Aase, S. B.; Buck, E. C.; Chamerlain, D. B.

1999-01-01

We report x-ray absorption near edge structure (XANES) and extended x-ray absorption fine structure (EXAFS) spectra from the plutonium L III edge and XANES from the cerium L II edge in prototype titanate ceramic hosts. The titanate ceramics studied are based upon the hafnium-pyrochlore and zirconolite mineral structures and will serve as an immobilization host for surplus fissile materials, containing as much as 10 weight % fissile plutonium and 20 weight % (natural or depleted) uranium. Three ceramic formulations were studied: one employed cerium as a ''surrogate'' element, replacing both plutonium and uranium in the ceramic matrix, another formulation contained plutonium in a ''baseline'' ceramic formulation, and a third contained plutonium in a formulation representing a high-impurity plutonium stream. The cerium XANES from the surrogate ceramic clearly indicates a mixed III-IV oxidation state for the cerium. In contrast, XANES analysis of the two plutonium-bearing ceramics shows that the plutonium is present almost entirely as Pu(IV) and occupies the calcium site in the zirconolite and pyrochlore phases. The plutonium EXAFS real-space structure shows a strong second-shell peak, clearly distinct from that of PuO 2 , with remarkably little difference in the plutonium crystal chemistry indicated between the baseline and high-impurity formulations

New ceramics containing dispersants for improved fracture toughness

Science.gov (United States)

Nevitt, M.V.; Aldred, A.T.; Chan, Sai-Kit

1985-07-01

The invention is a ceramic composition containing a new class of dispersant for hindering crack propagation by means of one or more energy-dissipative mechanisms. The composition is composed of a ceramic matrix with dispersed particles of a transformation-prone rare-earth niobate, tantalate or mixtures of these with each other and/or with a rare-earth vanadate. The dispersants, having a generic composition tRBO/sub 4/, where R is a rare-earth element, B if Nb or Ta and O is oxygen, are mixed in powder form with a powder of the matrix ceramic and sintered to produce a ceramic form or body. The crack-hindering mechanisms operates to provide improved performance over a wide range of temperature and operating conditions.

Determination of uranium in samples containing bulk aluminium

International Nuclear Information System (INIS)

Das, S.K.; Kannan, R.; Dhami, P.S.; Tripathi, S.C.; Gandhi, P.M.

2015-01-01

The determination of uranium is of great importance in PUREX process and need to be analyzed at different concentration ranges depending on the stage of reprocessing. Various techniques like volumetry, spectrophotometry, ICP-OES, fluorimetry, mass spectrometry etc. are used for the measurement of uranium in these samples. Fast and sensitive methods suitable for low level detection of uranium are desirable to cater the process needs. Microgram quantities of uranium are analyzed by spectrophotometric method using 2-(5- bromo-2-pyridylazo-5-diethylaminophenol) (Br-PADAP) as the complexing agent. But, the presence of some of the metal ions viz. Al, Pu, Zr etc. interferes in its analysis. Therefore, separation of uranium from such interfering metal ions is required prior to its analysis. This paper describes the analysis of uranium in samples containing aluminium as major matrix

Kinetics of dissolution of thorium and uranium doped britholite ceramics

Energy Technology Data Exchange (ETDEWEB)

Dacheux, N., E-mail: nicolas.dacheux@univ-montp2.f [Groupe de Radiochimie, Institut de Physique Nucleaire d' Orsay, Bat. 100, Universite Paris-Sud-11, 91406 Orsay (France); Institut de Chimie Separative de Marcoule, UMR 5257 (Universite Montpellier 2/CNRS/CEA/ENSCM), Bat. 426, Centre de Marcoule, BP 17171, 30207 Bagnols sur ceze cedex (France); Du Fou de Kerdaniel, E. [Groupe de Radiochimie, Institut de Physique Nucleaire d' Orsay, Bat. 100, Universite Paris-Sud-11, 91406 Orsay (France); Clavier, N. [Groupe de Radiochimie, Institut de Physique Nucleaire d' Orsay, Bat. 100, Universite Paris-Sud-11, 91406 Orsay (France); Institut de Chimie Separative de Marcoule, UMR 5257 (Universite Montpellier 2/CNRS/CEA/ENSCM), Bat. 426, Centre de Marcoule, BP 17171, 30207 Bagnols sur ceze cedex (France); Podor, R. [Institut de Chimie Separative de Marcoule, UMR 5257 (Universite Montpellier 2/CNRS/CEA/ENSCM), Bat. 426, Centre de Marcoule, BP 17171, 30207 Bagnols sur ceze cedex (France); Institut Jean Lamour - Departement CP2S - Equipe 206, Faculte des Sciences et Techniques - Nancy Universite, BP 70239, 54506 Vandoeuvre les Nancy cedex (France); Aupiais, J. [CEA DAM DIF, 91297 Arpajon (France); Szenknect, S. [Institut de Chimie Separative de Marcoule, UMR 5257 (Universite Montpellier 2/CNRS/CEA/ENSCM), Bat. 426, Centre de Marcoule, BP 17171, 30207 Bagnols sur ceze cedex (France)

2010-09-01

In the field of immobilization of actinides in phosphate-based ceramics, several thorium and uranium doped britholite samples were submitted to leaching tests. The normalized dissolution rates determined for several pH values, temperatures and acidic media from the calcium release range from 4.7 x 10{sup -2} g m{sup -2} d{sup -1} to 21.6 g m{sup -2} d{sup -1}. Their comparison with that determined for phosphorus, thorium and uranium revealed that the dissolution is clearly incongruent for all the conditions examined. Whatever the leaching solution considered, calcium and phosphorus elements were always released with higher R{sub L} values than the other elements (Nd, Th, U). Simultaneously, thorium was found to quickly precipitate as alteration product, leading to diffusion phenomena for uranium. For all the media considered, the uranium release is higher than that of thorium, probably due to its oxidation from tetravalent oxidation state to uranyl. Moreover, the evaluation of the partial order related to proton concentration and the apparent energy of activation suggest that the reaction of dissolution is probably controlled by surface chemical reactions occurring at the solid/liquid interface. Finally, comparative leaching tests performed in sulphuric acid solutions revealed a significant influence of such media on the chemical durability of the leached pellets, leading to higher normalized dissolution rates for all the elements considered. On the basis of the results of chemical speciation, this difference was mainly explained in the light of higher complexion constants by sulfate ions compared to nitrate, chloride and phosphate.

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

International Nuclear Information System (INIS)

Tsuruta, T.

2006-01-01

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

Criticality analyses of regions containing uranium in the earth history

International Nuclear Information System (INIS)

Ravnik, M.

2005-01-01

Investigations of necessary conditions for a self-sustained chain reaction in the Earth inner regions hypothetically containing uranium is presented for the time interval from Earth formation to present time. It is determined that criticality was theoretically possible up to 2.5 Ga before present if uranium concentrated in pure form. In the early geological history (4 Ga before present) the self-sustained criticality could occur even if uranium was diluted up to 1:20 by the average core material or 1:60 by the average mantle material. If other metallic materials of similar density as uranium (e.g., Au, W) or similar atomic weight (e.g., Th) concentrated from the primordial mixture in equal proportion as uranium, criticality was not possible for any period in Earth history provided that the basic material contained no light nuclides (H, C). Criticality in the Earth inner regions could have established only if uranium concentrated from the basic material more effectively than elements of similar density or atomic number. (orig.)

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

Recovery of uranium from sulphate solutions containing molybdenum

International Nuclear Information System (INIS)

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

1983-01-01

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

Treatment of uranium-containing effluent in the process of metallic uranium parts

International Nuclear Information System (INIS)

Yuan Guoqi

1993-01-01

The anion exchange method used in treatment of uranium-containing effluent in the process of metallic parts is the subject of the paper. The results of the experiments shows that the uranium concentration in created water remains is less than 10 μg/l when the waste water flowed through 10000 column volume. A small facility with column volume 150 litre was installed and 1500 m 3 of waste water can be cleaned per year. (1 tab.)

High temperature tribological properties of plasma-sprayed metallic coatings containing ceramic particles

International Nuclear Information System (INIS)

Dallaire, S.; Legoux, J.G.

1995-01-01

For sealing a moving metal component with a dense silica-based ceramic pre-heated at 800 C, coatings with a low coefficient of friction and moderate wear loss are required. As reported previously, plasma-sprayed coatings containing solid lubricants could reduce sliding wear in high-temperature applications. Plasma-sprayed metal-based coatings containing ceramic particles have been considered for high temperature sealing. Selected metal powders (NiCoCrAlY, CuNi, CuNiIn, Ag, Cu) and ceramic particles (boron nitride, Zeta-B ceramic) were agglomerated to form suitable spray powders. Plasma-sprayed composite coatings and reference materials were tested in a modified pin-on-disc apparatus in which the stationary disc consisted of a dense silica-based ceramic piece initially heated at 800 C and allowed to cool down during tests. The influence of single exposure and repeated contacts with a dense silica-based ceramic material pre-heated to 800 C on the coefficient of friction, wear loss and damage to the ceramic piece was evaluated. Being submitted to a single exposure at high temperature, coatings containing malleable metals such as indium, silver and copper performed well. The outstanding tribological characteristics of the copper-Zeta-B ceramic coating was attributed to the formation of a glazed layer on the surface of this coating which lasted over exposures to high temperature. This glazed layer, composed of fine oxidation products, provided a smooth and polished surface and helped maintaining the coefficient of friction low

Dense cermets containing fine grained ceramics and their manufacture

International Nuclear Information System (INIS)

King, H.L.

1986-01-01

This patent describes a method of producing a ceramic-metal composite (cermet) containing boride-oxide ceramic having components of a first metal boride and a second metal oxide, which ceramic is in mixture in the cermet with elemental metal of the second metal, wherein the cermet is produced by sintering a reaction mixture of the first metal oxide, boron oxide and the elemental second metal. The improvement consists of: combining for the reaction mixture; A. (a) first metal oxide; (b) boron oxide; (c) ceramic component in very finely divided form; and (d) elemental second metal in very finely divided form and in an amount of at least a 100 percent molar excess beyond that amount stoichiometrically required to produce the second metal oxide during sintering; and B. sintering the reaction mixture in inert gas atmosphere

The XPS study of the structure of uranium-containing ceramics

Directory of Open Access Journals (Sweden)

Teterin Anton Yu.

2010-01-01

Full Text Available The samples of the (Ca0.5GdU0.5Zr2O7 and (Ca0.5GdU0.5(ZrTiO7 ceramics with the fluorite and pyrochlore structures used as matrixes for the long-lived high-level radioactive waste disposal were studied with the X-ray photoelectron spectroscopy method. On the basis of the X-ray photoelectron spectroscopy parameters of the outer and core electrons from the binding energy range of 0-1250 eV the oxidation states of the included metal ions were determined, the quantitative elemental and ionic analysis was done, and the orderliness (monophaseness was evaluated. The obtained data agree with the X-ray diffraction and the scanning electron microscopy results.

Ceramics among Eurasian hunter-gatherers: 32 000 years of ceramic technology use and the perception of containment

Directory of Open Access Journals (Sweden)

Mihael Budja

2016-12-01

Full Text Available We present two parallel and 32 000 years long trajectories of episodic ceramic technology use in Eurasian pre-Neolithic hunter-gatherer societies. In eastern, Asian trajectory the pottery was produced from the beginning. Ceramic figurines mark the western, European trajectory. The western predates the eastern for about eleven millennia. While ceramic cones and figurines first appeared in Central Europe at c. 31 000 cal BC the earliest vessels in eastern Asia was dated at c. 20 000 cal BC. We discuss women’s agency, perception of containment, ‘cross-craft interactions’, and evolution of private property that that may influenced the inventions of ceramic (pyrotechnology.

Depleted uranium concrete container feasibility study

International Nuclear Information System (INIS)

Haelsig, R.T.

1994-09-01

The purpose of this report is to consider the feasibility of using containers constructed of depleted uranium aggregate concrete (DUCRETE) to store and transport radioactive materials. The method for this study was to review the advantages and disadvantages of DUCRETE containers considering design requirements for potential applications. The author found that DUCRETE is a promising material for onsite storage containers, provided DUCRETE vessels can be certified for one-way transport to disposal sites. The author also found that DUCRETE multipurpose spent nuclear fuel storage/transport packages are technically viable, provided altered temperature acceptance limits can be developed for DUCRETE

Ionic Liquids as templating agents in formation of uranium-containing nanomaterials

Science.gov (United States)

Visser, Ann E; Bridges, Nicholas J

2014-06-10

A method for forming nanoparticles containing uranium oxide is described. The method includes combining a uranium-containing feedstock with an ionic liquid to form a mixture and holding the mixture at an elevated temperature for a period of time to form the product nanoparticles. The method can be carried out at low temperatures, for instance less than about 300.degree. C.

Recovery of uranium from 30 vol % tributyl phosphate solvents containing dibutyl phosphate

International Nuclear Information System (INIS)

Mailen, J.C.; Tallent, O.K.

1986-01-01

A number of solid sorbents were tested for the removal of uranium and dibutyl phosphate (DBP) from 30% tributyl phosphate (TBP) solvent. The desired clean uranium product can be obtained either by removing the DBP, leaving the uranium in the solvent for subsequent stripping, or by removing the uranium, leaving the DBP in the solvent for subsequent treatment. The tests performed show that it is relatively easy to preferentially remove uranium from solvents containing uranium and DBP, but quite difficult to remove DBP preferentially. The current methods could be used by removing the uranium (as by a cation exchange resin) and then using either an anion exchange resin in the hydroxyl form or a conventional treatment with a basic solution to remove the DBP. Treatment of a solvent with a cation exchange resin could be useful for recovery of valuable metals from solvents containing DBP and might be used to remove cations before scrubbing a solvent with a basic solution to minimize emulsion formation. 6 refs., 9 figs

Bio-Corrosion Behavior of Ceramic Coatings Containing Hydroxyapatite on Mg-Zn-Ca Magnesium Alloy

Directory of Open Access Journals (Sweden)

Hong-Yan Ding

2018-04-01

Full Text Available Ceramic coatings containing hydroxyapatite (HA were fabricated on a biodegradable Mg66Zn29Ca5 magnesium alloy through micro-arc oxidation by adding HA particles into the electrolytes. The phase composition and surface morphology of the coatings were characterized by X-ray diffraction and scanning electron microscopy analyses, respectively. Electrochemical experiments and immersion tests were performed in Hank’s solution at 37 °C to measure the corrosion resistance of the coatings. Blood compatibility was evaluated by in vitro blood platelet adhesion tests and static water contact angle measurement. The results show that the typical ceramic coatings with a porous structure were prepared on the magnesium alloy surface with the main phases of MgO and MgSiO3 and a small amount of Mg3(PO42 and HA. The optimal surface morphology appeared at HA concentration of 0.4 g/L. The electrochemical experiments and immersion tests reveal a significant improvement in the corrosion resistance of the ceramic coatings containing HA compared with the coatings without HA or bare Mg66Zn29Ca5 magnesium alloy. The static water contact angle of the HA-containing ceramic coatings is 18.7°, which is lower than that of the coatings without HA (40.1°. The in vitro blood platelet adhesion tests indicate that the HA-containing ceramic coatings possess improved blood compatibility compared with the coatings without HA. Utilizing HA-containing ceramic coatings may be an effective way to improve the surface biocompatibility and corrosion resistance of magnesium alloys.

Effects of drop testing on scale model shipping containers shielded with depleted uranium

International Nuclear Information System (INIS)

Butler, T.A.

1980-02-01

Three scale model shipping containers shielded with depleted uranium were dropped onto an essentially unyielding surface from various heights to determine their margins to failure. This report presents the results of a thorough posttest examination of the models to check for basic structural integrity, shielding integrity, and deformations. Because of unexpected behavior exhibited by the depleted uranium shielding, several tests were performed to further characterize its mechanical properties. Based on results of the investigations, recommendations are made for improved container design and for applying the results to full-scale containers. Even though the specimens incorporated specific design features, the results of this study are generally applicable to any container design using depleted uranium

Ceramics as nuclear reactor fuels

International Nuclear Information System (INIS)

Reeve, K.D.

1975-01-01

Ceramics are widely accepted as nuclear reactor fuel materials, for both metal clad ceramic and all-ceramic fuel designs. Metal clad UO 2 is used commercially in large tonnages in five different power reactor designs. UO 2 pellets are made by familiar ceramic techniques but in a reactor they undergo complex thermal and chemical changes which must be thoroughly understood. Metal clad uranium-plutonium dioxide is used in present day fast breeder reactors, but may eventually be replaced by uranium-plutonium carbide or nitride. All-ceramic fuels, which are necessary for reactors operating above about 750 0 C, must incorporate one or more fission product retentive ceramic coatings. BeO-coated BeO matrix dispersion fuels and silicate glaze coated UO 2 -SiO 2 have been studied for specialised applications, but the only commercial high temperature fuel is based on graphite in which small fuel particles, each coated with vapour deposited carbon and silicon carbide, are dispersed. Ceramists have much to contribute to many aspects of fuel science and technology. (author)

Controlled thermolysis of uranium (alkoxy)siloxy complexes: a route to polymetallic complexes of Low-Valent uranium

Energy Technology Data Exchange (ETDEWEB)

Camp, Clement; Pecaut, Jacques; Mazzanti, Marinella [Laboratoire de Reconnaissance Ionique et Chimie de Coordination, SCIB, UMR-E3 CEA-UJF, INAC, CEA-Grenoble (France); Kefalidis, Christos E.; Maron, Laurent [LPCNO, CNRS et INSA, UPS, Universite de Toulouse (France)

2013-07-01

Decomposition into higher species: Intramolecular U{sup III}-mediated homolytic C-O bond cleavage in U{sup III} (alkoxy)siloxy complexes at low temperature and subsequent reduction with KC{sub 8} led to unprecedented polymetallic complexes containing siloxy, silanediolate, and silanetriolate ligands. Such compounds may be useful precursors to uranium ceramics relevant for catalysis and the storage of spent nuclear fuel. (Copyright copyright 2013 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

49 CFR 173.426 - Excepted packages for articles containing natural uranium or thorium.

Science.gov (United States)

2010-10-01

... outer surface of the uranium or thorium is enclosed in an inactive sheath made of metal or other durable... uranium or thorium. 173.426 Section 173.426 Transportation Other Regulations Relating to Transportation....426 Excepted packages for articles containing natural uranium or thorium. A manufactured article in...

Glass-Ceramic Waste Forms for Uranium and Plutonium Residues Wastes - 13164

International Nuclear Information System (INIS)

Stewart, Martin W.A.; Moricca, Sam A.; Zhang, Yingjie; Day, R. Arthur; Begg, Bruce D.; Scales, Charlie R.; Maddrell, Ewan R.; Hobbs, Jeff

2013-01-01

A program of work has been undertaken to treat plutonium-residues wastes at Sellafield. These have arisen from past fuel development work and are highly variable in both physical and chemical composition. The principal radiological elements present are U and Pu, with small amounts of Th. The waste packages contain Pu in amounts that are too low to be economically recycled as fuel and too high to be disposed of as lower level Pu contaminated material. NNL and ANSTO have developed full-ceramic and glass-ceramic waste forms in which hot-isostatic pressing is used as the consolidation step to safely immobilize the waste into a form suitable for long-term disposition. We discuss development work on the glass-ceramic developed for impure waste streams, in particular the effect of variations in the waste feed chemistry glass-ceramic. The waste chemistry was categorized into actinides, impurity cations, glass formers and anions. Variations of the relative amounts of these on the properties and chemistry of the waste form were investigated and the waste form was found to be largely unaffected by these changes. This work mainly discusses the initial trials with Th and U. Later trials with larger variations and work with Pu-doped samples further confirmed the flexibility of the glass-ceramic. (authors)

Uranium adsorption from the sulphuric acid leach liquor containing more chlorides with cation-exchange resin SL-406

International Nuclear Information System (INIS)

Hu Jun; Wang Zhaoguo; Chi Renqing; Niu Xuejun

1994-01-01

The feasibility of uranium adsorption was studied from the sulphuric acid leach liquor of a uranium ore containing more chlorides with cation-exchange resin SL-406. The influence of some factors on uranium adsorption was investigated. It was shown that the resin possesses better selectivity, stability and higher capacity. It can be effectively used to recovery uranium from leach liquors of uranium ores containing more chlorides

Purification process of uranium hexafluoride containing traces of plutonium fluoride and/or neptunium fluoride

International Nuclear Information System (INIS)

Aubert, J.; Bethuel, L.; Carles, M.

1983-01-01

In this process impure uranium hexafluoride is contacted with a metallic fluoride chosen in the group containing lead fluoride PbF 2 , uranium fluorides UFsub(4+x) (0 3 at a temperature such as plutonium and/or neptunium are reduced and pure uranium hexafluoride is recovered. Application is made to uranium hexafluoride purification in spent fuel reprocessing [fr

Chemical process for recovery of uranium values contained in phosphoric mineral lixivia

International Nuclear Information System (INIS)

Conceicao, E.L.H. da; Awwal, M.A.; Coelho, S. V.

1980-01-01

A recovery process of uranium values from phosporic mineral lixivia for obtaining uranio oxide concentrate adjusted to specifications of purity for its commercialization the process consists of the adjustment of electromotive force of lixiviem to suitable values for uranium extraction, extraction with organic solvent containing phosphoric acid ester and oxidant reextraction from this solvent with phosphoric acid solution, suggesting a new solvent extraction containing synergetic mixture of di-2-ethyl hexyl phosphoric acid and tri-octyl phosphine, leaching this solvent with water and re-extraction/precipitation with ammonium carbonate solution, resulting in the formation of uranyl tricarbonate and ammonium, that by drying and calcination gives the uranium oxide with purity degree for commercialization. (M.C.K.) [pt

Study of the quenching and subsequent return to room temperature of uranium-chromium, uranium-iron, and uranium-molybdenum alloys containing only small amounts of the alloying element

International Nuclear Information System (INIS)

Delaplace, J.

1960-09-01

By means of an apparatus which makes possible thermal pre-treatments in vacuo, quenching carried out in a high purity argon atmosphere, and simultaneous recording of time temperature cooling and thermal contraction curves, the author has examined the transformations which occur in uranium-chromium, uranium-iron and uranium-molybdenum alloys during their quenching and subsequent return to room temperature. For uranium-chromium and uranium-iron alloys, the temperature at which the γ → β transformation starts varies very little with the rate of cooling. For uranium-molybdenum alloys containing 2,8 atom per cent of Mo, this temperature is lowered by 120 deg. C for a cooling rate of 500 deg. C/mn. The temperature at which the β → α transformation starts is lowered by 170 deg. C for a cooling rate of 500 deg. C/mn in the case of uranium-chromium alloy containing 0,37 atom per cent of Cr. The temperature is little affected in the case of uranium-iron alloys. The addition of chromium or iron makes it possible to conserve the form β at ordinary temperatures after quenching from the β and γ regions. The β phase is particularly unstable and changes into needles of the α form even at room temperatures according to an autocatalytic transformation law similar to the austenitic-martensitic transformation law in the case of iron. The β phase obtained by quenching from the β phase region is more stable than that obtained by quenching from the γ region. Chromium is a more effective stabiliser of the β phase than is iron. Unfortunately it causes serious surface cracking. The β → α transformation in uranium-chromium alloys has been followed at room temperature by means of micro-cinematography. The author has not observed the direct γ → α transformation in uranium-molybdenum alloys containing 2,8 per cent of molybdenum even for cooling rates of up to 2000 deg. C/s. He has however observed the formation of several martensitic structures. (author) [fr

Magnetic properties of bioactive glass-ceramics containing nanocrystalline zinc ferrite

International Nuclear Information System (INIS)

Singh, Rajendra Kumar; Srinivasan, A.

2011-01-01

Glass-ceramics with finely dispersed zinc ferrite (ZnFe 2 O 4 ) nanocrystallites were obtained by heat treatment of x(ZnO,Fe 2 O 3 )(65-x)SiO 2 20(CaO,P 2 O 5 )15Na 2 O (6≤x≤21 mole%) glasses. X-ray diffraction patterns of the glass-ceramic samples revealed the presence of calcium sodium phosphate [NaCaPO 4 ] and zinc ferrite [ZnFe 2 O 4 ] as major crystalline phases. Zinc ferrite present in nanocrystalline form contributes to the magnetic properties of the glass-ceramic samples. Magnetic hysteresis cycles of the glass-ceramic samples were obtained with applied magnetic field sweeps of ±20 kOe and ±500 Oe, in order to evaluate the potential of these glass-ceramics for hyperthermia treatment of cancer. The evolution of magnetic properties in these samples, viz., from a partially paramagnetic to fully ferrimagnetic nature has been explored using magnetometry and X-ray diffraction studies. - Research highlights: → The glass-ceramics contain bone mineral and magnetic phases. → Calcium sodium phosphate and zinc ferrite nanocrystallites have been identified in all the sample. → With an increase in ZnO and Fe2O3 content, magnetic property of samples evolved from partially paramagnetic to fully ferrimagnetic nature. → Large magnetic hysteresis loops have been obtained for samples with high ZnO+Fe2O3 content.

Ionic flotation of uranium contained in industrial phosphoric acid

International Nuclear Information System (INIS)

Jdid; Blazy; Bessiere

1983-01-01

A new process for uranium recovery from industrial phosphoric acid at 30% of P 2 O 5 is applied by the ionic flotation process. Research is carried out on determination of the nature of ionic species of U in H 3 PO 4 5.5 M and the behavior of reagents from CECA Co. in very acid media. Reagents able to form complexes directly with uranium and stable in phosphoric acid selected are: potassium ethylene diamine tetra (methylene phosphonate) (INIPOL AD32) and sodium dialkyldiphosphonate (34S). Uranium IV, obtained by reduction of uranium VI with iron powder, is precipitated by these reagents. Flotation of the precipitate obtained with INIPOL AD 32 is realized by addition of hexylamino bis (methylene phosphonic acid). A recovery of 80 wt% is obtained. Flotation of the coprecipitate 34S-U(IV) is obtained without any other additions because 34S is a surfactant. Metal recovery is better than 90% and the coprecipitate contains more than 10% U. The process is fast precipitation 10 minutes and flotation 5 minutes and is efficient even at 60 0 C [fr

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.

Management of wastes containing radioactivity from mining and milling uranium ores in Northern Australia

International Nuclear Information System (INIS)

Costello, J.M.

1977-01-01

The procedures and controls to achieve safe management of wastes containing radioactivity during the mining and processing of uranium ores are mainly site-specific depending on the nature, location and distribution of the ore and gangue material. Waste rock and below-ore-grade material containing low levels of radioactivity require disposal at the mine site. In open-cut mining the material is generally stockpiled above ground, with revegetation and collection of run-off water. Some material may be used to backfill open cuts. Management of these wastes requires a thorough investigation of groundwater hydrology and surface soil characteristics to control dissipation of radioactive material. Dust containing radon and radioactive particulate is produced during ore milling, and dusts of ore concentrate are generated during calcination and packaging of the yellowcake product. These dusts are managed by ventilation and filtration systems; working conditions and discharges to atmosphere will be according to the Australian Code of Practice on Radiation Protection during Mining and Milling of Uranium Ores. The chemical waste stream from leaching and processing of the uranium ores contains most of the radioactivity resulting from radium and its decay products. Neutralized effluent is discharged into holding ponds for settling solids. The paper describes the nature of wastes containing radioactivity resulting from the mining and milling of uranium, and illustrates modern engineering practices and monitoring procedures to manage the wastes, as described in the Environmental Impact Statement produced by Ranger Uranium Mines Pty Ltd (RUM) for public hearings. (author)

Nickel container of highly-enriched uranium bodies and sodium

Science.gov (United States)

Zinn, Walter H.

1976-01-01

A fuel element comprises highly a enriched uranium bodies coated with a nonfissionable, corrosion resistant material. A plurality of these bodies are disposed in layers, with sodium filling the interstices therebetween. The entire assembly is enclosed in a fluid-tight container of nickel.

Nickel container of highly-enriched uranium bodies and sodium

International Nuclear Information System (INIS)

Zinn, W.H.

1976-01-01

A fuel element comprises highly enriched uranium bodies coated with a nonfissionable, corrosion resistant material. A plurality of these bodies are disposed in layers, with sodium filling the interstices therebetween. The entire assembly is enclosed in a fluid-tight container of nickel

NEUTRONICS STUDIES OF URANIUM-BASED FULLY CERAMIC MICRO-ENCAPSULATED FUEL FOR PWRs

Energy Technology Data Exchange (ETDEWEB)

George, Nathan M [ORNL; Maldonado, G Ivan [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL; Godfrey, Andrew T [ORNL

2012-01-01

This study evaluates the core neutronics and fuel cycle characteristics that result from employing uranium-based fully ceramic micro-encapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR bundle designs with FCM fuel have been developed, which by virtue of their TRISO particle based elements, are expected to safely reach higher fuel burnups while also increasing the tolerance to fuel failures. The SCALE 6.1 code package, developed and maintained at ORNL, was the primary software employed to model these designs. Analysis was performed using the SCALE double-heterogeneous (DH) fuel modeling capabilities. For cases evaluated with the NESTLE full-core three-dimensional nodal simulator, because the feature to perform DH lattice physics branches with the SCALE/TRITON sequence is not yet available, the Reactivity-Equivalent Physical Transformation (RPT) method was used as workaround to support the full core analyses. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a color-set array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In addition, a parametric study was performed by varying the various TRISO particle design features; such as kernel diameter, coating layer thicknesses, and packing fractions. Also, other features such as the selection of matrix material (SiC, Zirconium) and fuel rod dimensions were perturbed. After evaluating different uranium-based fuels, the higher physical density of uranium mononitride (UN) proved to be favorable, as the parametric studies showed that the FCM particle fuel design will need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. Neutronically, the FCM fuel designs evaluated maintain acceptable design features in the areas of fuel lifetime, temperature

76 FR 11275 - In the Matter of Certain Ceramic Capacitors and Products Containing Same; Notice of Commission...

Science.gov (United States)

2011-03-01

... INTERNATIONAL TRADE COMMISSION [Investigation No. 337-TA-692] In the Matter of Certain Ceramic Capacitors and Products Containing Same; Notice of Commission Determination To Review in Part A Final Initial... importation of certain ceramic capacitors and products containing the same by reason of infringement of...

Study on the treatment of waste waster containing uranium by organic modified vermiculite

International Nuclear Information System (INIS)

Liu Wenjuan; Zeng Yanhong

2012-01-01

The adsorption capability of uranium on organic modified Vermiculite was studied. The influence factors of the amount of adsorbent, initial pH, initial concentration of uranium and adsorption time have been investigated too. Through the orthogonal test, the primary factors of impacting the adsorption treatment can be obtained. Finally, the preliminary research and analysis on the principle adsorption of organic modified vermiculite test of uranium have been conducted. The results show that: Modifying Vermiculite by CTMAB makes Vermiculite adsorption capacity stronger when treating solution containing uranium. Combined flocculants with vermiculite to treat with low concentration of uranium solution has synergy, significantly enhancing its adsorption capacity. The impact factors of organic modified vermiculite's adsorption of uranium are adsorbent dosage, pH, initial concentration of uranium solution and adsorption time. The best adsorption pH is between 5∼6.5. (authors)

Uranium(VI) adsorption properties of a chelating resin containing polyamine-substituted methylphosphonic acid moiety

International Nuclear Information System (INIS)

Matsuda, Masaaki; Akiyoshi, Yoshirou

1991-01-01

Uranium(VI) adsorption and desorption properties of a chelating resin containing polyamine-substituted methylphosphonic acid moiety of 2.29 mmol/g-resin (APA) were examined. Uranium(VI) adsorption properties of several ion exchange resins and extractant agents which were known as excellent adsorbents for uranium(VI), were examined together for a comparison with those of APA. Uranium(VI) adsorption capacity of APA at the concentration of 100 mg·dm -3 -uranium(VI) in 100 g·dm -3 -H 2 SO 4 aq. soln., 190 g·dm -3 -H 3 PO 4 aq. soln. and uranium enriched sea water, was 0.2, 0.05 and 0.05 mmol·g -1 respectively. The adsorption capacity of APA for uranium(VI) in these solutions was larger than that of another adsorbents, except the adsorption of uranium(VI) in enriched sea water on ion exchange resin containing phosphoric acid moiety (adsorption capacity ; 0.2 mmol·g -1 ). Uranium(VI) adsorption rate on APA was high and the relation between treatment time (t : min) and uranium(VI) concentration (y : mg·dm -3 ) in 100 g·dm -3 H 2 SO 4 aq. soln. after treatment, was shown as following equation, y=20 0.048t+1.90 (0≤t≤30). The adsorbed uranium(VI) on APA was able to be eluted with a mixed aq. soln. of hydrogen peroxide and sodium hydroxide and also was able to be eluted with an aq. alkaline soln. dissolved reduction agents such as sodium sulfite and hydrazine. From these results, it was thought that uranium(VI) adsorbed on APA was eluted due to the reduction to uranium(VI) by these eluents. (author)

Ceramics for Molten Materials Containment, Transfer and Handling on the Lunar Surface

Science.gov (United States)

Standish, Evan; Stefanescu, Doru M.; Curreri, Peter A.

2009-01-01

As part of a project on Molten Materials Transfer and Handling on the Lunar Surface, molten materials containment samples of various ceramics were tested to determine their performance in contact with a melt of lunar regolith simulant. The test temperature was 1600 C with contact times ranging from 0 to 12 hours. Regolith simulant was pressed into cylinders with the approximate dimensions of 1.25 dia x 1.25cm height and then melted on ceramic substrates. The regolith-ceramic interface was examined after processing to determine the melt/ceramic interaction. It was found that the molten regolith wetted all oxide ceramics tested extremely well which resulted in chemical reaction between the materials in each case. Alumina substrates were identified which withstood contact at the operating temperature of a molten regolith electrolysis cell (1600 C) for eight hours with little interaction or deformation. This represents an improvement over alumina grades currently in use and will provide a lifetime adequate for electrolysis experiments lasting 24 hours or more. Two types of non-oxide ceramics were also tested. It was found that they interacted to a limited degree with the melt resulting in little corrosion. These ceramics, Sic and BN, were not wetted as well as the oxides by the melt, and so remain possible materials for molten regolith handling. Tests wing longer holding periods and larger volumes of regolith are necessary to determine the ultimate performance of the tested ceramics.

Plasma spraying of bioactive glass-ceramics containing bovine bone

Directory of Open Access Journals (Sweden)

Annamária Dobrádi

2017-06-01

Full Text Available Natural bone derived glass-ceramics are promising biomaterials for implants. However, due to their price and weak mechanical properties they are preferably applied as coatings on load bearing implants. This paper describes result obtained by plasma spraying of bioactive glass-ceramics containing natural bone onto selected implant materials, such as stainless steel, alumina, and titanium alloy. Adhesion of plasma sprayed coating was tested by computed X-ray tomography and SEM of cross sections. The results showed defect free interface between the coating and substrate, without cracks or gaps. Dissolution rate of the coating in simulated body fluid (SBF was readily controlled by the bone additives (phase composition, as well as microstructure. The SBF treatment of the plasma sprayed coating did not influence the boundary between the coating and substrate.

[Spectroscopic Research on Slag Nanocrystal Glass Ceramics Containing Rare Earth Elements].

Science.gov (United States)

Ouyang, Shun-li; Li, Bao-wei; Zhang, Xue-feng; Jia, Xiao-lin; Zhao, Ming; Deng, Lei-bo

2015-08-01

The research group prepared the high-performance slag nanocrystal glass ceramics by utilizing the valuable elements of the wastes in the Chinese Bayan Obo which are characterized by their symbiotic or associated existence. In this paper, inductively coupled plasma emission spectroscopy (ICP), X-ray diffraction (XRD), Raman spectroscopy (Raman) and scanning electron microscopy (SEM) are all used in the depth analysis for the composition and structure of the samples. The experiment results of ICP, XRD and SEM showed that the principal crystalline phase of the slag nanocrystal glass ceramics containing rare earth elements is diopside, its grain size ranges from 45 to 100 nm, the elements showed in the SEM scan are basically in consistent with the component analysis of ICP. Raman analysis indicated that its amorphous phase is a three-dimensional network structure composed by the structural unit of silicon-oxy tetrahedron with different non-bridging oxygen bonds. According to the further analysis, we found that the rare earth microelement has significant effect on the network structure. Compared the nanocrystal slag glass ceramic with the glass ceramics of similar ingredients, we found that generally, the Raman band wavenumber for the former is lower than the later. The composition difference between the glass ceramics and the slag nanocrystal with the similar ingredients mainly lies on the rare earth elements and other trace elements. Therefore, we think that the rare earth elements and other trace elements remains in the slag nanocrystal glass ceramics have a significant effect on the network structure of amorphous phase. The research method of this study provides an approach for the relationship among the composition, structure and performance of the glass ceramics.

Management of wastes containing radioactivity from mining and milling of uranium ores in Northern Australia

International Nuclear Information System (INIS)

Costello, J.M.

1977-01-01

The procedures and controls to achieve safe management of wastes containing radioactivity during the mining and processing of uranium ores are mainly site specific depending on the nature, location and distribution of the ore and gangue material. Waste rock and below-ore-grade material containing low levels of radioactivity require disposal at the mine site. In open cut mining the material is generally stockpiled above ground, with revegetation and collection of run-off water. Some material may be used to backfill open cuts. Management of these wastes requires a thorough investigation of ground water hydrology and surface soil characteristics to control dissipation of radioactive material. Dust containing radon and radioactive particulate is produced during ore milling, and dusts of ore concentrate are generated during calcination and packaging of the yellowcake product. These dusts are managed by ventilation and filtration systems, working conditions, and discharges to atmosphere will be according to the Australian Code of Practice on Radiation Protection during Mining and Milling of Uranium Ores. The chemical waste stream from leaching and processing of the uranium ores contains the majority of the radioactivity resulting from radium and its decay products. Neutralised effluent is discharged into holding ponds for settling of solids. This paper describes the nature of wastes containing radioactivity resulting from the mining and milling of uranium, and illustrates modern engineering practices and monitoring procedures to manage the wastes, as described in the Environmental Impact statement produced by Ranger Uranium Mines Proprietary Limited for public hearings

Process for the winning of a concentrate containing uranium and purified phosphoric acid, as well as the concentrate containing uranium and purified phosphoric acid obtained by this process

International Nuclear Information System (INIS)

1980-01-01

The uranium containing concentrate and purified phosphoric acid are obtained by treating wet phosphoric acid with an inorganic fluorine compound (ammonium fluoride) and an aliphatic ketone (acetone) in the presence of a reducing agent (finely divided iron). The ketone is added first and the formed uranium precipitate is separated from the solution. If the fluorine compound is added first, the yield is lowered by a factor of 2. (Th.P.)

Conditioning of uranium-containing technological radioactive waste

International Nuclear Information System (INIS)

Smodis, B.; Tavcar, G.; Stepisnik, M.; Pucelj, B.

2006-01-01

Conditioning of mostly liquid uranium containing technological radioactive waste emerging from the past research activities at the Jozef Stefan Institute is described. The waste was first thoroughly characterised, then the radionuclides present solidified by appropriate chemical treatment, and the final product separated and prepared for storage in compliance with the legislation. The activities were carried out within the recently renewed Hot Cells Facility of the Jozef Stefan Institute and the overall process resulted in substantial volume reduction of the waste initially present. (author)

Investigation of the fire at the Uranium Enrichment Laboratory. Analysis of samples and pressurization experiment/analysis of container

International Nuclear Information System (INIS)

Akabori, Mitsuo; Minato, Kazuo; Watanabe, Kazuo

1998-05-01

To investigate the cause of the fire at the Uranium Enrichment Laboratory of the Tokai Research Establishment on November 20, 1997, samples of uranium metal waste and scattered residues were analyzed. At the same time the container lid that had been blown off was closely inspected, and the pressurization effects of the container were tested and analyzed. It was found that 1) the uranium metal waste mainly consisted of uranium metal, carbides and oxides, whose relative amounts were dependent on the particle size, 2) the uranium metal waste hydrolyzed to produce combustible gases such as methane and hydrogen, and 3) the lid of the outer container could be blown off by an explosive rise of the inner pressure caused by combustion of inflammable gas mixture. (author)

Processing of uranium-containing coal

International Nuclear Information System (INIS)

Cordero Alvarez, M.

1987-01-01

A direct storage of uranium-bearing coal requires the processing of large amounts of raw materials while lacking guarantee of troublefree process cycles. With the example of an uranium-bearing bituminous coal from Stockheim, it was aimed at the production of an uranium ore concentrate by means of mechanical, thermal and chemical investigations. Above all, amorphous pitch blende was detected as a uranium mineralization which occurs homogeneously distributed in the grain size classes of the comminuted raw material with particle diameters of a few μm and, after the combustion, enriches in the field of finest grain of the axis. Heterogeneous and solid-state reactions in the thermal decarburization above 700deg C result in the development of hardly soluble uranium oxides and and calcium uranates as well as in enclosures in mineral glass. Thus, the pre-enrichment has to take place in a temperature range below 600deg C. By means of a sorting classification of the ash at ± 2.0 mm, it is possible to achieve an enrichment of up to factor 15 for a mineral of a mainly low carbonate content and, for a mineral of a rich carbonate content, up to the factor 4. The separation of the uranium from the concentrates produced is possible with a yield of 95% by means of leaching with sulphuric acid at a temperature of 20deg C. As far as their reproducibility was concerned, the laboratory tests were verified on a semi-industrial scale. A processing method is suggested on the basis of the data obtained. (orig.) [de

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)

10 CFR 34.67 - Records of leak testing of sealed sources and devices containing depleted uranium.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Records of leak testing of sealed sources and devices containing depleted uranium. 34.67 Section 34.67 Energy NUCLEAR REGULATORY COMMISSION LICENSES FOR INDUSTRIAL... Requirements § 34.67 Records of leak testing of sealed sources and devices containing depleted uranium. Each...

Application of acid dissolution and natural evaporation to wet cake containing uranium

International Nuclear Information System (INIS)

Kim, Kil J.; Kang, Il Sik; Shon, Jong S.; Hong, Kwon P.

2005-01-01

Chemical wastes containing small amounts of uranium cause environmental problems, if those wastes exceed the concentration of the EPA standard, 20 μg.. /L, and the concentrated sludge should be additionally dried and packaged into a drum, and categorized as a radioactive waste. Diphosil resin is developed to specifically remove actinides or multivalent metals. The immobilization technique is adopted to make a bead form of Diphosil by embedding into sodium alginate, and adsorption characteristics for uranium are reported for a simulated waste solution. In this study, acid dissolution is applied to dissolve uranium from the precipitates of sludge or the dewatered cake in the reduced volume of wastes solution, and removal characteristics of uranium is experimented. From the results, the most effective treatment method for the dissolved solution is suggested

Transport of uranium by supported liquid membrane containing bis(2-ethylhexyl) hydrogenphosphate and 1-octanol

International Nuclear Information System (INIS)

Akiba, Kenichi; Kanno, Takuji; Takahashi, Toshihiko.

1984-01-01

Carrier-mediated transport of uranium(VI) has been studied by means of liquid membranes impregnated in a microporous polymer. Liquid membranes containing bis(2-ethylhexyl) hydrogenphosphate (DEHPA) alone yielded inadequate stripping of uranium. The addition of 1-octanol to DEHPA solutions resulted in a decrease in extractability, and made it possible to control the distribution ratio of uranium. Uranium in the feed solution was sufficiently transported across the liquid membrane containing this DEHPA-1-octanol mixture into the product solution. The apparent rate constant (ksub(obs)) of transport increased slightly with an increase in carrier concentrations. Variations in acid concentrations of the feed solution (pH 2.5--3.2) and the product solution (0.1--1.0 M H 2 SO 4 ) had little effect on the transport rate. A large excess of uranium, more than the carrier content in the liquid membrane, was finally concentrated in the stripping acid. (author)

Measurement of Gross Alpha and Beta Emission Rates from Ceramic Tiles

International Nuclear Information System (INIS)

Wudthicharoonpun, Piyasak; Chankow, Nares

2007-08-01

Full text: Ceramic tiles normally used to cover floors and walls contain naturally occurring radioactive elements i.e. potassium-40, uranium, thorium and their daughters from raw materials. Thus, radioactivity was dependent upon source of raw materials and the amount used. The objective of this research was to measure gross alpha and beta emission rates to be used as a database for safety assessment and for selection of rooms to measure radioactive radon-222 gas

Synthesis and characterization of cerium containing iron phosphate based glass-ceramics

Science.gov (United States)

Deng, Yi; Liao, Qilong; Wang, Fu; Zhu, Hanzhen

2018-02-01

The structure and properties of xCeO2-(100-x)(40Fe2O3-60P2O5), where x = 0, 2, 4, 6 and 8 mol%, glass-ceramics prepared by melting and slow cooling method have been investigated by using X-ray diffraction (XRD), scanning electron microscope (SEM), Fourier transform infrared spectroscopy (FTIR), differential thermal analysis (DTA) and the Product Consistency Test (PCT). The results show that the 40Fe2O3-60P2O5 sample is homogeneously amorphous and the sample containing 2 mol% CeO2 has a small amount of FePO4 phase embedded. For the sample containing up to 4 mol% CeO2, monazite CePO4 and a small amount of FePO4 appear. Spectra analysis show that the structure networks of the glass-ceramics mainly consist of orthophosphate, along with pyrophosphate and a small amount of metaphosphate units. Moreover, the leaching rates of Fe and Ce are about 3.5 × 10-5 g m-2 d-1 and 5.0 × 10-5 g m-2 d-1 respectively after immersion in deionized water at 90 °C for 56 days, indicating their good chemical durability. The conclusions imply that the prepared method may be a promising process to immobilize nuclear waste into glass-ceramic matrix.

Selectivity of NF membrane for treatment of liquid waste containing uranium

International Nuclear Information System (INIS)

Oliveira, Elizabeth E.M.; Barbosa, Celina C.R.; Afonso, Julio C.

2013-01-01

The performance of two nanofiltration membranes were investigated for treatment of liquid waste containing uranium through two conditions permeation: permeation test and concentration test of the waste. In the permeation test solution permeated returned to the feed tank after collected samples each 3 hours. In the test of concentration the permeated was collected continuously until 90% reduction of the feed volume. The liquid waste ('carbonated water') was obtained during conversion of UF 6 to UO 2 in the cycle of nuclear fuel. This waste contains uranium concentration on average 7.0 mg L -1 , and not be eliminated to the environmental. The waste was permeated using a cross-flow membrane cell in the pressure of the 1.5 MPa. The selectivity of the membranes for separation of uranium was between 83% and 90% for both tests. In the concentration tests the waste was concentrated around for 5 times. The surface layer of the membranes was evaluated before and after the tests by infrared spectroscopy (ATR-FTIR), field emission microscopy (FESEM) and atomic force spectroscopy (AFM). The membrane separation process is a technique feasible to and very satisfactory for treatment the liquid waste. (author)

Single-source-precursor Synthesis and High-temperature Behavior of SiC Ceramics Containing Boron

Science.gov (United States)

Gui, Miaomiao; Fang, Yunhui; Yu, Zhaoju

2014-12-01

In this paper, a hyperbranched polyborocarbosilane (HPBCS) was prepared by a one-pot synthesis with Cl2Si(CH3)CH2Cl, Cl3SiCH2Cl and BCl3 as the starting materials. The obtained HPBCS was characterized by GPC, FT-IR and NMR, and was confirmed to have hyperbranched structures. The thermal property of the resulting HPBCS was investigated by TGA. The ceramic yield of the HPBCS is about 84% and that of the counterpart hyperbranched hydridopolycarbosilane is only 45%, indicating that the introduction of boron into the preceramic polymer significantly improved the ceramic yield. With the polymer-derived ceramic route, the final ceramics were annealed at 1800 °C in argon atmosphere for 2 h in order to characterize the microstructure and to evaluate the high-temperature behavior. The final ceramic microstructure was studied by XRD and SEM, indicating that the introduction of boron dramatically inhibits SiC crystallization. The boron-containing SiC ceramic shows excellent high-temperature behavior against decomposition and crystallization at 1800 °C.

Elaboration of new ceramic composites containing glass fibre production wastes

International Nuclear Information System (INIS)

Rozenstrauha, I.; Sosins, G.; Krage, L.; Sedmale, G.; Vaiciukyniene, D.

2013-01-01

Two main by-products or waste from the production of glass fibre are following: sewage sludge containing montmorillonite clay as sorbent material and ca 50 % of organic matter as well as waste glass from aluminium borosilicate glass fibre with relatively high softening temperature (> 600 degree centigrade). In order to elaborate different new ceramic products (porous or dense composites) the mentioned by-products and illitic clay from two different layers of Apriki deposit (Latvia) with illite content in clay fraction up to 80-90 % was used as a matrix. The raw materials were investigated by differential-thermal (DTA) and XRD analysis. Ternary compositions were prepared from mixtures of 15 - 35 wt % of sludge, 20 wt % of waste glass and 45 - 65 wt % of clay and the pressed green bodies were thermally treated in sintering temperature range from 1080 to 1120 degree centigrade in different treatment conditions. Materials produced in temperature range 1090 - 1100 degree centigrade with the most optimal properties - porosity 38 - 52 %, water absorption 39 -47 % and bulk density 1.35 - 1.67 g/cm 3 were selected for production of porous ceramics and materials showing porosity 0.35 - 1.1 %, water absorption 0.7 - 2.6 % and bulk density 2.1 - 2.3 g/cm 3 - for dense ceramic composites. Obtained results indicated that incorporation up to 25 wt % of sewage sludge is beneficial for production of both ceramic products and glass-ceramic composites according to the technological properties. Structural analysis of elaborated composite materials was performed by scanning electron microscopy(SEM). By X-ray diffraction analysis (XRD) the quartz, diopside and anorthite crystalline phases were detected. (Author)

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.

Early uranium mining in the United States

International Nuclear Information System (INIS)

Hahne, F.J.

1990-01-01

Uranium mining in the United States is closer to 100 years old than to the 200 years since the discovery of the element. Even then, for much of this time the rock was brought out of the ground for reasons other than its uranium content. The history of the US uranium industry is divided into five periods which follow roughly chronologically upon one another, although there is some overlap. The periods cover: uranium use in glass and ceramics; radium extraction; vanadium extraction; government uranium extraction and commercial extraction. (author)

Bioleaching of low grade uranium ore containing pyrite using A. ferrooxidans and A. thiooxidans

International Nuclear Information System (INIS)

Alexey Borisovich Umanskii; Anton Mihaylovich Klyushnikov

2013-01-01

A process of uranium extraction from ore containing 3.1 % pyrite by bacterial leaching was investigated in shaken flasks during 90 days. The highest uranium recovery amounting to 85.1 % was obtained using binary mixture of Acidithiobacillus ferrooxidans and Acidithiobacillus thiooxidans that was exceeding results obtained by traditional acid leaching technique up to 27 %. High uranium recovery was founded to be due to the high degree of pyrite dissolution that can be readily achieved by bacterial leaching (up to 98.0 %). (author)

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

Ion flotation of uranium contained in industrial phosphoric acid with collector recycling

International Nuclear Information System (INIS)

Jdid, E.; Blazy, P.; Bessiere, J.

1985-01-01

Uranium has been recovered from wet-process phosphoric acid (30% P 2 O 5 ) by ion flotation with an anionic organophosphorous collector. Recoveries greater than 90% were obtained even at temperatures of about 60 C, the uranium concentrate, which was collected in the froth as a precipitate, containing 7 to 10% U. Collector consumption without recycling of the surface-active reagent was about 12 kg/kg U. Much of the reagent, however, can be recovered for recycling by attack with sodium hydroxide on the floated phase after filtration. This enables a precipitate containing about 30% U to be produced and decreases collector consumption to about 3 kg/kg U. The results were obtained in laboratory-scale experiments on industrial wet-process acid. (author)

Recovery of fissile materials from plutonium residues, miscellaneous spent nuclear fuel, and uranium fissile wastes

International Nuclear Information System (INIS)

Forsberg, C.W.

1997-01-01

A new process is proposed that converts complex feeds containing fissile materials into a chemical form that allows the use of existing technologies (such as PUREX and ion exchange) to recover the fissile materials and convert the resultant wastes to glass. Potential feed materials include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel, and (3) uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, and organics. 14 refs., 4 figs

Neutronics Studies Of Uranium-Based Fully Ceramic Micro-Encapsulated Fuel For PWRs

International Nuclear Information System (INIS)

Maldonado, G. Ivan; Gehin, Jess C.

2012-01-01

This study evaluates the core neutronics and fuel cycle characteristics that result from employing uranium-based fully ceramic micro-encapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR bundle designs with FCM fuel have been developed, which by virtue of their TRISO particle based elements, are expected to safely reach higher fuel burnups while also increasing the tolerance to fuel failures. The SCALE 6.1 code package, developed and maintained at ORNL, was the primary software employed to model these designs. Analysis was performed using the SCALE double-heterogeneous (DH) fuel modeling capabilities. For cases evaluated with the NESTLE full-core three-dimensional nodal simulator, because the feature to perform DH lattice physics branches with the SCALE/TRITON sequence is not yet available, the Reactivity-Equivalent Physical Transformation (RPT) method was used as workaround to support the full core analyses. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a color-set array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In addition, a parametric study was performed by varying the various TRISO particle design features; such as kernel diameter, coating layer thicknesses, and packing fractions. Also, other features such as the selection of matrix material (SiC, Zirconium) and fuel rod dimensions were perturbed. After evaluating different uranium-based fuels, the higher physical density of uranium mononitride (UN) proved to be favorable, as the parametric studies showed that the FCM particle fuel design will need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. Neutronically, the FCM fuel designs evaluated maintain acceptable design features in the areas of fuel lifetime, temperature

Acid pressure leaching of a concentrate containing uranium, thorium and rare earth elements

International Nuclear Information System (INIS)

Lan Xinghua; Peng Ruqing.

1987-01-01

The acid pressure leaching of a concentrate containing rinkolite for recovering uranium, thorium and rare earth elements is described. The laboratory and the pilot plant test results are given. Under the optimum leaching conditions, the recovery of uranium, thorium and rare earth elements are 82.9%, 86.0% and 88.3% respectively. These results show that the acid pressure leaching process is a effective process for treating the concentrate

Characterization of PAH matrix with monazite stream containing uranium, gadolinium and iron

Energy Technology Data Exchange (ETDEWEB)

Pal, Sangita, E-mail: sangpal@barc.gov.in; Goswami, D. [Desalination Division, Bhabha Atomic Research Centre, Trombay, Mumbai-400 085 (India); Meena, Sher Singh [Solid State Physics Division, Bhabha Atomic Research Centre, Trombay, Mumbai-400 085 (India)

2016-05-23

Uranium (U) gadolinium (Gd) and iron (Fe) containing alkaline waste simulated effluent (relevant to alkaline effluent of monazite ore) has been treated with a novel amphoteric resin viz, Polyamidehydroxamate (PAH) containing amide and hydroxamic acid groups. The resin has been synthesized in an eco-friendly manner by polymerization nad conversion to functional groups characterized by FT-IR spectra and architectural overview by SEM. Coloration of the loaded matrix and de-coloration after extraction of uranium is the special characteristic of the matrix. Effluent streams have been analyzed by ICP-AES, U loaded PAH has been characterized by FT-IR, EXAFS, Gd and Fe by X-ray energy values of EDXRF at 6.053 KeVand 6.405 KeV respectively. The remarkable change has been observed in Mössbauer spectrum of Fe-loaded PAH samples.

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)

Ceramic heat exchanger

Science.gov (United States)

LaHaye, Paul G.; Rahman, Faress H.; Lebeau, Thomas P. E.; Severin, Barbara K.

1998-01-01

A tube containment system. The tube containment system does not significantly reduce heat transfer through the tube wall. The contained tube is internally pressurized, and is formed from a ceramic material having high strength, high thermal conductivity, and good thermal shock resistance. The tube containment system includes at least one ceramic fiber braid material disposed about the internally pressurized tube. The material is disposed about the tube in a predetermined axial spacing arrangement. The ceramic fiber braid is present in an amount sufficient to contain the tube if the tube becomes fractured. The tube containment system can also include a plurality of ceramic ring-shaped structures, in contact with the outer surface of the tube, and positioned between the tube and the ceramic fiber braid material, and/or at least one transducer positioned within tube for reducing the internal volume and, therefore, the energy of any shrapnel resulting from a tube fracture.

Standard test method for the analysis of refrigerant 114, plus other carbon-containing and fluorine-containing compounds in uranium hexafluoride via fourier-transform infrared (FTIR) spectroscopy

CERN Document Server

American Society for Testing and Materials. Philadelphia

2004-01-01

1.1 This test method covers determining the concentrations of refrigerant-114, other carbon-containing and fluorine-containing compounds, hydrocarbons, and partially or completely substituted halohydrocarbons that may be impurities in uranium hexafluoride. The two options are outlined for this test method. They are designated as Part A and Part B. 1.1.1 To provide instructions for performing Fourier-Transform Infrared (FTIR) spectroscopic analysis for the possible presence of Refrigerant-114 impurity in a gaseous sample of uranium hexafluoride, collected in a "2S" container or equivalent at room temperature. The all gas procedure applies to the analysis of possible Refrigerant-114 impurity in uranium hexafluoride, and to the gas manifold system used for FTIR applications. The pressure and temperatures must be controlled to maintain a gaseous sample. The concentration units are in mole percent. This is Part A. 1.2 Part B involves a high pressure liquid sample of uranium hexafluoride. This method can be appli...

Mechanical properties of concrete containing recycled concrete aggregate (RCA) and ceramic waste as coarse aggregate replacement

Science.gov (United States)

Khalid, Faisal Sheikh; Azmi, Nurul Bazilah; Sumandi, Khairul Azwa Syafiq Mohd; Mazenan, Puteri Natasya

2017-10-01

Many construction and development activities today consume large amounts of concrete. The amount of construction waste is also increasing because of the demolition process. Much of this waste can be recycled to produce new products and increase the sustainability of construction projects. As recyclable construction wastes, concrete and ceramic can replace the natural aggregate in concrete because of their hard and strong physical properties. This research used 25%, 35%, and 45% recycled concrete aggregate (RCA) and ceramic waste as coarse aggregate in producing concrete. Several tests, such as concrete cube compression and splitting tensile tests, were also performed to determine and compare the mechanical properties of the recycled concrete with those of the normal concrete that contains 100% natural aggregate. The concrete containing 35% RCA and 35% ceramic waste showed the best properties compared with the normal concrete.

The effect of sedimentation background of depression target stratum containing mineral in Erlian basin, Ulanqab to uranium mineralization type

International Nuclear Information System (INIS)

Kang Shihu; Jiao Yangquan; Men Hong; Kuang Wenzhan

2012-01-01

The ore bearing stratum in depression of Ulanqab contains target stratum of lower cretaceous Saihan formation, upper cretaceous Erlian formation, paleogene system etc. The uranium mineralization type which have found by now contains sandstone type, mudstone type and coal petrography. The genetic type of mineral deposit contains paleovalley-type, reformed type after superposition with sedimentation and diagenesis by sedimentation. Uranium mineralization of both the natural type and genetic type have close relationship with its ore bearing stratum. Different geological background forms different sedimentary system combination, and different sedimentary system combination forms different uranium mineralization type. (authors)

Hot pressing of uranium nitride and mixed uranium plutonium nitride

International Nuclear Information System (INIS)

Chang, J.Y.

1975-01-01

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

Recovery or removal of uranium contained in small quantity in waste water by tannic-group adsorbent

Energy Technology Data Exchange (ETDEWEB)

Komoto, Shigetoshi [Power Reactor and Nuclear Fuel Development Corp., Kamisaibara, Okayama (Japan). Ningyo Toge Works

1991-12-01

It was found that tannic compounds have a very strong affinity for uranium and thorium which are nuclear fuel materials, and the new uranium adsorbents composed mainly by tannic-group compounds were made. The solid-state refractory persimmon tannins in those compounds has the most superior capacity for uranium as high as 1.7 g of uranium on 1 g of the adsorbent. The tests adsorbing uranium on the adsorbent are carried out practically by using dam water of Ningyo-toge Works, PNC. Adsorption tests changed the pH or temperature of dam water, elution test, and adsorption-elution repeating test were performed, and it was found that uranium in dam water contained from ppb-level to ppm-level was recovered or removed with very excellent efficiency. (author).

Uranium's scientific history

International Nuclear Information System (INIS)

Goldschmidt, B.

1990-01-01

The bicentenary of the discovery of uranium coincides with the fiftieth anniversary of the discovery of fission, an event of worldwide significance and the last episode in the uranium -radium saga which is the main theme of this paper. Uranium was first identified by the German chemist Martin Klaproth in 1789. He extracted uranium oxide from the ore pitchblende which was a by-product of the silver mines at Joachimsthal in Bohemia. For over a century after its discovery, the main application for uranium derived from the vivid colours of its oxides and salts which are used in glazes for ceramics, and porcelain. In 1896, however, Becquerel discovered that uranium emitted ionizing radiation. The extraction by Pierre and Marie Curie of the more radioactive radium from uranium in the early years of the twentieth century and its application to the treatment of cancer shifted the chief interest to radium production. In the 1930s the discovery of the neutron and of artificial radioactivity stimulated research in a number of European laboratories which culminated in the demonstration of fission by Otto Frisch in January 1939. The new found use of uranium for the production of recoverable energy, and the creation of artificial radioelements in nuclear reactors, eliminated the radium industry. (author)

Uranium complex recycling method of purifying uranium liquors

International Nuclear Information System (INIS)

Elikan, L.; Lyon, W.L.; Sundar, P.S.

1976-01-01

Uranium is separated from contaminating cations in an aqueous liquor containing uranyl ions. The liquor is mixed with sufficient recycled uranium complex to raise the weight ratio of uranium to said cations preferably to at least about three. The liquor is then extracted with at least enough non-interfering, water-immiscible, organic solvent to theoretically extract about all of the uranium in the liquor. The organic solvent contains a reagent which reacts with the uranyl ions to form a complex soluble in the solvent. If the aqueous liquor is acidic, the organic solvent is then scrubbed with water. The organic solvent is stripped with a solution containing at least enough ammonium carbonate to precipitate the uranium complex. A portion of the uranium complex is recycled and the remainder can be collected and calcined to produce U 3 O 8 or UO 2

Preliminary study for treatment methodology establishment of liquid waste containing uranium in refining facility lagoon

Energy Technology Data Exchange (ETDEWEB)

Lee, Byung Jik; Lee, Kune Woo; Won, Hui Jun; Ahn, Byung Gil; Shim, Joon Bo

1999-12-01

The preliminary study which establishes the treatment methodology of the sludge waste containing uranium in the conversion facility lagoon was performed. The property of lagoon liquid waste such as the initial water content, the density including radiochemical analysis results were obtained using the samples taken from the lagoon. The objective of this study is to provide some basically needed materials for selection of the most proper lagoon waste treatment methodology by reviewing the effective processes and methods for minimizing the secondary waste resulting from the treatment and disposition of large amount of radioactive liquid waste according to the facility closing. The lagoon waste can be classified into two sorts, such as supernatant and precipitate. The supernatants contain uranium less than 5 ppm and their water content are about 35 percent. Therefore, supernatants are solutions composed of mainly salt components. However, the precipitates have lots of uranium compound contained in the coagulation matrix, and are formed as two kinds of crystalline structures. The most proper method minimizing the secondary waste would be direct drying and solidification of the supernatants and precipitates after separation of them by filtering. (author)

Method of sintering ceramic materials

Science.gov (United States)

Holcombe, Cressie E.; Dykes, Norman L.

1992-01-01

A method for sintering ceramic materials is described. A ceramic article is coated with layers of protective coatings such as boron nitride, graphite foil, and niobium. The coated ceramic article is embedded in a container containing refractory metal oxide granules and placed within a microwave oven. The ceramic article is heated by microwave energy to a temperature sufficient to sinter the ceramic article to form a densified ceramic article having a density equal to or greater than 90% of theoretical density.

Development of sorbers for the recovery of uranium from seawater. Part 2. The accumulation of uranium from seawater by resins containing amidoxime and imidoxime functional groups

International Nuclear Information System (INIS)

Astheimer, L.; Schenk, H.J.; Witte, E.G.; Schwochau, K.

1983-01-01

Hydroxylamine derivatives of cross-linked poly(acrylonitriles), so-called poly(acrylamidoxime) resins, are suitable for the accumulation of uranium from natural seawater of pH = 8.1 to 8.3. Depending on the method of manufacture, these sorbers yield excellent uranium loadings up to some thousand ppM which roughly equals the average uranium content of actually explored uranium ores. The rate of uranium uptake, which is 5 to 30 ppM/d at room temperature, increases with increasing temperature of seawater. Uranium can be eluted by 1 M HCl with an elution efficiency of more than 90%. Owing to a certain instability of the uranium binding groups in acid eluants, the uranium uptake decreases with increasing number of sorption-elution cycles. Hydroxylamine derivatives of poly(acrylonitrile) are shown to contain simultaneously at least two kinds of functional groups: open-chain amidoxime groups which are stable and cyclic imidoxime groups which are unstable in 1 M HCl. Experimental evidence is presented that the uptake of uranium from natural seawater is closely related to the presence of cyclic imidoxime configurations in the polyacrylic lattice. Polystyrene and poly(glycidylmethacrylate)-based amidoxime and imide dioxime resins are less effective in extracting uranium from natural seawater. 10 figures, 4 tables

Easy Debonding of Ceramic Brackets Bonded with a Light-Cured Orthodontic Adhesive Containing Microcapsules with a CO2 Laser.

Science.gov (United States)

Arima, Shiori; Namura, Yasuhiro; Tamura, Takahiko; Shimizu, Noriyoshi

2018-03-01

An easy debonding method for ceramic brackets using a light-cured Bis-GMA resin containing heat-expandable microcapsules and CO 2 laser was investigated. Ceramic brackets are used frequently in orthodontic treatment because of their desirable esthetic properties. However, the application of heavy force to ceramic brackets in debonding can fracture the tooth enamel and ceramic brackets, causing tooth pain. In total, 60 freshly extracted bovine permanent mandibular incisors were divided randomly into 10 groups of 6 specimens each, corresponding to the number of variables tested. Ceramic brackets were bonded to bovine permanent mandibular incisors using an orthodontic bonding agent containing heat-expandable microcapsules at different levels (0-30 wt%) and resin composite paste, and cured by a curing device. The bond strengths were measured before and after CO 2 laser irradiation, and the temperature increase in the pulp chamber in fresh human first premolars was also evaluated. With CO 2 laser irradiation for 5 sec to the bracket, the bond strength in the 25% microcapsule group decreased significantly, to ∼0.17-fold, compared with that of the no-laser group (p brackets, with less debonding time and enamel damage.

Development of high-density ceramic composites for ballistic applications

International Nuclear Information System (INIS)

Rupert, N.L.; Burkins, M.S.; Gooch, W.A.; Walz, M.J.; Levoy, N.F.; Washchilla, E.P.

1993-01-01

The application of ceramic composites for ballistic application has been generally developed with ceramics of low density, between 2.5 and 4.5 g/cm 2 . These materials have offered good performance in defeating small-caliber penetrators, but can suffer time-dependent degradation effects when thicker ceramic tiles are needed to defeat modem, longer, heavy metal penetrators that erode rather than break up. This paper addresses the ongoing development, fabrication procedures, analysis, and ballistic evaluation of thinner, denser ceramics for use in armor applications. Nuclear Metals Incorporated (NMI) developed a process for the manufacture of depleted uranium (DU) ceramics. Samples of the ceramics have been supplied to the US Army Research Laboratory (ARL) as part of an unfunded cooperative study agreement. The fabrication processes used, characterization of the ceramic, and a ballistic comparison between the DU-based ceramic with baseline Al 2 O 3 will be presented

49 CFR 173.477 - Approval of packagings containing greater than 0.1 kg of non-fissile or fissile-excepted uranium...

Science.gov (United States)

2010-10-01

... kg of non-fissile or fissile-excepted uranium hexafluoride. 173.477 Section 173.477 Transportation... non-fissile or fissile-excepted uranium hexafluoride. (a) Each offeror of a package containing more than 0.1 kg of uranium hexafluoride must maintain on file for at least one year after the latest...

Trace recovery of uranium and rare earth contained in phosphates by liquid-liquid extraction in sulfuric attack liquor

International Nuclear Information System (INIS)

Bousquet, F.; Foraison, D.; Leveque, A.; Sabot, J.L.

1980-06-01

Uranium and rare earths can be recovered in sedimentary phosphates during the wet processing of the ore by sulfuric acid giving raw phosphoric acid at 30 per cent of P 2 O 5 . Practically all the uranium contained and only part of rare earths are put into solution in this treatment. Separation of these elements in the phosphoric solution is obtained by liquid-liquid extraction with alkylphosphoric acids and especially with their mono and di esters. Partition isotherms are determined and counter-current tests are effected. Uranium and rare earths reextraction from these solvents can be simultaneous or separate with aqueous solutions alkaline or containing HF or by antisynergism. Pros and cons of each reextraction process are discussed. In conclusion HDEHP or OPPA are recommended because of availability, stability and hydrodynamic, OPPA less selective with rare earths allows the recovery with uranium of ceric earths, yttrium and yttric earths [fr

Method for converting uranium oxides to uranium metal

International Nuclear Information System (INIS)

Duerksen, W.K.

1988-01-01

A method for converting uranium oxide to uranium metal is described comprising the steps of heating uranium oxide in the presence of a reducing agent to a temperature sufficient to reduce the uranium oxide to uranium metal and form a heterogeneous mixture of a uranium metal product and oxide by-products, heating the mixture in a hydrogen atmosphere at a temperature sufficient to convert uranium metal in the mixture to uranium hydride, cooling the resulting uranium hydride-containing mixture to a temperature sufficient to produce a ferromagnetic transition in the uranium hydride, magnetically separating the cooled uranium hydride from the mixture, and thereafter heating the separated uranium hydride in an inert atmosphere to a temperature sufficient to convert the uranium hydride to uranium metal

Incentives for the use of depleted uranium alloys as transport cask containment structure

International Nuclear Information System (INIS)

McConnell, P.; Salzbrenner, R.; Wellman, G.W.; Sorenson, K.B.

1992-01-01

Radioactive material transport casks use either lead or depleted uranium (DU) as gamma-ray shielding material. Stainless steel is conventionally used for structural containment. If a DU alloy had sufficient properties to guarantee resistance to failure during both nominal use and accident conditions to serve the dual-role of shielding and containment, the use of other structure materials (i.e., stainless steel) could be reduced. (It is recognized that lead can play no structural role.) Significant reductions in cask weight and dimensions could then be achieved perhaps allowing an increase in payload. The mechanical response of depleted uranium has previously not been included in calculations intended to show that DU-shielded transport casks will maintain their containment function during all conditions. This paper describesa two-part study of depleted uranium alloys: First, the mechanical behavior of DU alloys was determined in order to extend the limited set of mechanical properties reported in the literature. The mechanical properties measured include the tensile behavior the impact energy. Fracture toughness testing was also performed to determine the sensitivity of DU alloys to brittle fracture. Fracture toughness is the inherent material property which quantifies the fracmm resistance of a material. Tensile strength and ductility are significant in terms of other failure modes, however, as win be discussed. These mechanical properties were then input into finite element calculations of cask response to loading conditions to quantify the potential for claiming structural credit for DU. (The term ''structural credit'' describes whether a material has adequate properties to allow it to assume a positive role in withstanding structural loadings.)

Radioactivity reference levels in ceramics tiles as building materials for different countries

International Nuclear Information System (INIS)

Ortiz, Josefina; Ballesteros, Luisa; Serradell, Vicente

2008-01-01

Measurements campaigns of ceramic tiles and raw materials used in them, shows that natural radionuclides of uranium ( 238 U) and thorium ( 232 Th) series, together with the radioactive isotope of potassium ( 40 K ), are presents. Uranium series contain radium, which decays to radon ( 222 Rn), an inert gas that can be released from materials and inhaled by individuals. Limits of 226 Ra concentrations are established by different countries in order to control Radon levels (200 Bq.m -3 in European Union). Potassium -40 and others gamma emitters of 226 Ra and 232 Th descendent, can cause an external dose. Therefore, with the purpose that individual doses due to building materials doesn't exceed a certain level recommendations or regulations have been established. A maximum value of 1 mSv.y -1 is recommended in European Union. In practice an easy way to avoid ceramic tiles provide doses to individuals over the reference level is to introduce an index, depending on activities concentrations of 226 Ra, 232 Th and 40 K, defined so that the dose limits due, exclusively, to building materials, will never be exceeded. These limits and indexes present differences between countries. In this paper indexes are compared and differences are discussed. (author)

INFLUENCE OF REOXIDATION ON SILICA-CONTAINING BARIUM TITANATE CERAMICS FOR PTCR THERMISTORS PREPARED BY TAPE CASTING

Directory of Open Access Journals (Sweden)

Jianqiao Liu

2016-03-01

Full Text Available Silica-containing barium-rich BaTiO₃ ceramics for thermistors with a positive temperature coefficient of resistance are prepared by a tape-casting technique. The ceramics are sintered in a reducing atmosphere at low temperatures of 1175-1225°C. The influences of reoxidation are investigated after the reduced ceramics are reoxidized in air at 700-900°C. An anomalous correlation is illustrated between room temperature resistivity and reoxidation temperature. The anomaly results from the ferroelectricity rebuilding mechanism, which includes the spontaneous polarization theory and the ferroelectricity degradation caused by oxygen vacancies. The acceptor-state densities are estimated from the temperature-dependent resistivity. A critical temperature of 750-800°C is concluded for the grain boundary reoxidation.

Enhanced lithium battery with polyethylene oxide-based electrolyte containing silane-Al2 O3 ceramic filler.

Science.gov (United States)

Zewde, Berhanu W; Admassie, Shimelis; Zimmermann, Jutta; Isfort, Christian Schulze; Scrosati, Bruno; Hassoun, Jusef

2013-08-01

A solid polymer electrolyte prepared by using a solvent-free, scalable technique is reported. The membrane is formed by low-energy ball milling followed by hot-pressing of dry powdered polyethylene oxide polymer, LiCF3 SO3 salt, and silane-treated Al2 O3 (Al2 O3 -ST) ceramic filler. The effects of the ceramic fillers on the properties of the ionically conducting solid electrolyte membrane are characterized by using electrochemical impedance spectroscopy, XRD, differential scanning calorimeter, SEM, and galvanostatic cycling in lithium cells with a LiFePO4 cathode. We demonstrate that the membrane containing Al2 O3 -ST ceramic filler performs well in terms of ionic conductivity, thermal properties, and lithium transference number. Furthermore, we show that the lithium cells, which use the new electrolyte together with the LiFePO4 electrode, operate within 65 and 90 °C with high efficiency and long cycle life. Hence, the Al2 O3 -ST ceramic can be efficiently used as a ceramic filler to enhance the performance of solid polymer electrolytes in lithium batteries. Copyright © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Chemical thermodynamics of uranium

International Nuclear Information System (INIS)

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

1992-01-01

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

The effect of dispersed materials on baro-membrane treatment of uranium-containing waters

International Nuclear Information System (INIS)

Kryvoruchko, Antonina P.; Atamanenkoa, Irina D.

2007-01-01

The paper investigated a treatment process of uranium-containing waters in a membrane reactor while using natural mineral kizelgur and synthetic sorbent SKN-1K with subsequent ultra- and nano-filtration separation of the mixture. The retention coefficient of U(VI) by membrane UPM-20 under conditions of quasi-stationary equilibrium reached the levels of 0.87-0.89 and 0.89-0.91, respectively, while using natural mineral kizelgur and synthetic sorbent SKN-1K. In the case of membrane OPMN-P and natural mineral kizelgur the retention coefficient of U(VI) was 0.990-0.991 and 0.993-0.996, respectively, while using natural mineral kizelgur and synthetic sorbent SKN-1K. Data regarding the state of water in membranes formed from natural mineral or synthetic sorbent on the surface of substrate membranes UPM-20 and OPMN-P made it possible to conclude that dispersed materials of different chemical nature affect the process of baro-membrane treatment of uranium-containing waters. (authors)

Experience in usage of T-108 titrimetric laboratory unit for precision analysis of uranium-containing materials

International Nuclear Information System (INIS)

Ryzhinskij, M.V.; Bronzov, P.A.

1989-01-01

Possibilities of the T-108 device of potentiometric titration for precise determination of uranium in various uranium-containing materials are studied, the results being presented. Principle flowsheet of the device and the sequence of analytic procedure of uranium potentiometric titration are considered. U 3 O 8 , UO 2 and UF 4 were used as materials to be analyzed, state standard samples of K 2 Cr 2 O 7 -SSS 2215-81 and U 3 O 8 SSS 2396-83P- as standard samples. It is shown that relative standard deviation during titration using the T-108 device is mainly determined by the error of determination of the final titration point potention and it must not exceed 0.002 for uranium titration considered. The conclusion is made that the variant of potentiometric titration of uranium with the use of the T-108 device is not inferior in its accuracy to gravimetry and surpasses it in productivity and possibility of automation. 4 refs.; 2 figs.; 2 tabs

Method for converting uranium oxides to uranium metal

Science.gov (United States)

Duerksen, Walter K.

1988-01-01

A process is described for converting scrap and waste uranium oxide to uranium metal. The uranium oxide is sequentially reduced with a suitable reducing agent to a mixture of uranium metal and oxide products. The uranium metal is then converted to uranium hydride and the uranium hydride-containing mixture is then cooled to a temperature less than -100.degree. C. in an inert liquid which renders the uranium hydride ferromagnetic. The uranium hydride is then magnetically separated from the cooled mixture. The separated uranium hydride is readily converted to uranium metal by heating in an inert atmosphere. This process is environmentally acceptable and eliminates the use of hydrogen fluoride as well as the explosive conditions encountered in the previously employed bomb-reduction processes utilized for converting uranium oxides to uranium metal.

Cyclic mechanical fatigue in ceramic-ceramic composites: an update

International Nuclear Information System (INIS)

Lewis, D. III

1983-01-01

Attention is given to cyclic mechanical fatigue effects in a number of ceramics and ceramic composites, including several monolithic ceramics in which significant residual stresses should be present as a result of thermal expansion mismatches and anisotropy. Fatigue is also noted in several BN-containing ceramic matrix-particulate composites and in SiC fiber-ceramic matrix composites. These results suggest that fatigue testing is imperative for ceramics and ceramic composites that are to be used in applications subject to cyclic loading. Fatigue process models are proposed which provide a rationale for fatigue effect observations, but do not as yet provide quantitative results. Fiber composite fatigue damage models indicate that design stresses in these materials may have to be maintained below the level at which fiber pullout occurs

Incentives for the use of depleted uranium alloys as transport cask containment structure

International Nuclear Information System (INIS)

McConnell, P.; Salzbrenner, R.; Wellman, G.W.; Sorenson, K.B.

1993-01-01

Radioactive material transport casks use either lead or depleted uranium (DU) as gamma-ray shielding material. Stainless steel is conventionally used for structural containment. If a DU alloy had sufficient properties to guarantee resistance to failure during both normal use and accident conditions to serve the dual-role of shielding and containment, the use of other structural materials (i.e., stainless steel) could be reduced. (It is recognized that lead can play no structural role.) Significant reductions in cask weight and dimensions could then be achieved perhaps allowing an increase in payload. The mechanical response of depleted uranium has previously not been included in calculations intended to show that DU-shielded transport casks will maintain their containment function during all conditions. This paper describes a two-part study of depleted uranium alloys: First, the mechanical behavior of DU alloys was determined in order to extend the limited set of mechanical properties reported in the literature (Eckelmeyer, 1991). The mechanical properties measured include the tensile behavior the impact energy. Fracture toughness testing was also performed to determine the sensitivity of DU alloys to brittle fracture. Fracture toughness is the inherent material property which quantifies the fracture resistance of a material. Tensile strength and ductility are significant in terms of other failure modes, however, as will be discussed. These mechanical properties were then input into finite element calculations of cask response to loading conditions to quantify the potential for claiming structural credit for DU. (The term 'structural credit' describes whether a material has adequate properties to allow it to assume a positive role in withstanding structural loadings.) (J.P.N.)

Separation of uranium and common impurities from solid analytical waste containing plutonium

International Nuclear Information System (INIS)

Pathak, Nimai; Kumar, Mithlesh; Thulasidas, S.K.; Hon, N.S.; Kulkarni, M.J.; Mhatre, Amol; Natarajan, V.

2014-07-01

The report describes separation of uranium (U) and common impurities from solid analytical waste containing plutonium (Pu). This will be useful in recovery of Pu from nuclear waste. This is an important activity of any nuclear program in view of the strategic importance of Pu. In Radiochemistry Division, the trace metal analysis of Pu bearing fuel materials such as PuO 2 , (U,Pu)O 2 and (U,Pu)C are being carried out using the DC arc-Carrier Distillation technique. During these analyses, solid analytical waste containing Pu and 241 Am is generated. This comprises of left-over of samples and prepared charges. The main constituents of this waste are uranium oxide, plutonium oxide and silver chloride used as carrier. This report describes the entire work carried out to separate gram quantities of Pu from large amounts of U and mg quantities of 241 Am and the effect of leaching of the waste with nitric acid as a function of batch size. The effect of leaching the solid analytical waste of (U,Pu)O 2 and AgCl with concentrated nitric acid for different time intervals was also studied. Later keeping the time constant, the effect of nitric acid molarity on the leaching of U and Pu was investigated. Four different lots of the waste having different amounts were subjected to multiple leaching with 8 M nitric acid, each for 15 minutes duration. In all the experiments the amount of Uranium, Plutonium and other impurities leached were determined using ICP as an excitation source. The results are discussed in this report. (author)

FY2015 ceramic fuels development annual highlights

Energy Technology Data Exchange (ETDEWEB)

Mcclellan, Kenneth James [Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)

2015-09-22

Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2015 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY15 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.

FY2016 Ceramic Fuels Development Annual Highlights

Energy Technology Data Exchange (ETDEWEB)

Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-01-24

Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2016 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY16 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.

Uranium induces oxidative stress in lung epithelial cells

International Nuclear Information System (INIS)

Periyakaruppan, Adaikkappan; Kumar, Felix; Sarkar, Shubhashish; Sharma, Chidananda S.; Ramesh, Govindarajan T.

2007-01-01

Uranium compounds are widely used in the nuclear fuel cycle, antitank weapons, tank armor, and also as a pigment to color ceramics and glass. Effective management of waste uranium compounds is necessary to prevent exposure to avoid adverse health effects on the population. Health risks associated with uranium exposure includes kidney disease and respiratory disorders. In addition, several published results have shown uranium or depleted uranium causes DNA damage, mutagenicity, cancer and neurological defects. In the current study, uranium toxicity was evaluated in rat lung epithelial cells. The study shows uranium induces significant oxidative stress in rat lung epithelial cells followed by concomitant decrease in the antioxidant potential of the cells. Treatment with uranium to rat lung epithelial cells also decreased cell proliferation after 72 h in culture. The decrease in cell proliferation was attributed to loss of total glutathione and superoxide dismutase in the presence of uranium. Thus the results indicate the ineffectiveness of antioxidant system's response to the oxidative stress induced by uranium in the cells. (orig.)

Study of the quenching and subsequent return to room temperature of uranium-chromium, uranium-iron, and uranium-molybdenum alloys containing only small amounts of the alloying element; Etude de la trempe et du revenu a la temperature ordinaire d'alliages uranium-chrome, uranium-fer et uranium-molybdene, a faible teneur en element d'alliage

Energy Technology Data Exchange (ETDEWEB)

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

1960-09-15

By means of an apparatus which makes possible thermal pre-treatments in vacuo, quenching carried out in a high purity argon atmosphere, and simultaneous recording of time temperature cooling and thermal contraction curves, the author has examined the transformations which occur in uranium-chromium, uranium-iron and uranium-molybdenum alloys during their quenching and subsequent return to room temperature. For uranium-chromium and uranium-iron alloys, the temperature at which the {gamma} {yields} {beta} transformation starts varies very little with the rate of cooling. For uranium-molybdenum alloys containing 2,8 atom per cent of Mo, this temperature is lowered by 120 deg. C for a cooling rate of 500 deg. C/mn. The temperature at which the {beta} {yields} {alpha} transformation starts is lowered by 170 deg. C for a cooling rate of 500 deg. C/mn in the case of uranium-chromium alloy containing 0,37 atom per cent of Cr. The temperature is little affected in the case of uranium-iron alloys. The addition of chromium or iron makes it possible to conserve the form {beta} at ordinary temperatures after quenching from the {beta} and {gamma} regions. The {beta} phase is particularly unstable and changes into needles of the {alpha} form even at room temperatures according to an autocatalytic transformation law similar to the austenitic-martensitic transformation law in the case of iron. The {beta} phase obtained by quenching from the {beta} phase region is more stable than that obtained by quenching from the {gamma} region. Chromium is a more effective stabiliser of the {beta} phase than is iron. Unfortunately it causes serious surface cracking. The {beta} {yields} {alpha} transformation in uranium-chromium alloys has been followed at room temperature by means of micro-cinematography. The author has not observed the direct {gamma} {yields} {alpha} transformation in uranium-molybdenum alloys containing 2,8 per cent of molybdenum even for cooling rates of up to 2000 deg. C

Formulation and synthesis by melting process of titanate enriched glass-ceramics and ceramics

International Nuclear Information System (INIS)

Advocat, T.; Fillet, C.; Lacombe, J.; Bonnetier, A.; McGlinn, P.

1999-01-01

The main objective of this work is to provide containment for the separated radionuclides in stable oxide phases with proven resistance to leaching and irradiation damage and in consequence to obtain a glass ceramic or a ceramic material using a vitrification process. Sphene glass ceramic, zirconolite glass ceramic and zirconolite enriched ceramic have been fabricated and characterized by XRD, SEM/EDX and DTA

Melt refining of uranium contaminated copper, nickel, and mild steel

International Nuclear Information System (INIS)

Ren Xinwen; Liu Wencang; Zhang Yuan

1993-01-01

This paper presents the experiment results on melt refining of uranium contaminated metallic discards such as copper, nickel, and mild steel. Based on recommended processes, uranium contents in ingots shall decrease below 1 ppm; metal recovery is higher than 96%; and slag production is below 5% in weight of the metal to be refined. The uranium in the slag is homogeneously distributed. The slag seems to be hard ceramics, insoluble in water, and can be directly disposed of after proper packaging

Analytical characterization of a loading resin containing chlorophosphonazo I and its application to the enrichment of trace uranium

International Nuclear Information System (INIS)

Tang Fulong; Mao Xueqin

1986-01-01

A loading resin containing chlorophosphonazo I was prepared. The analytical properties of this resin for uranium were studied by the batch and column methods. In case EDTA is used as a masking agent, this method can be successfully applied to the separation and enrichment of trace uranium in wastewater from mining. The uranium adsorbed can be eluted with 1.5N HCl, and determined using the arsenazo III at pH 2 by spectrophotometry. The result obtained agrees well with that of the conventional method

Abrasive wear behaviour of bio-active glass ceramics containing ...

Indian Academy of Sciences (India)

In this study, abrasive wear behaviour of bio-active glass ceramic materials produced with two different processes is studied. Hot pressing process and conventional casting and controlled crystallization process were used to produce bio-active ceramics. Fracture toughness of studied material was calculated by fracture ...

Uranium purchases report 1992

International Nuclear Information System (INIS)

1993-01-01

Data reported by domestic nuclear utility companies in their responses to the 1991 and 1992 ''Uranium Industry Annual Survey,'' Form EIA-858, Schedule B ''Uranium Marketing Activities,are provided in response to the requirements in the Energy Policy Act 1992. Data on utility uranium purchases and imports are shown on Table 1. Utility enrichment feed deliveries and secondary market acquisitions of uranium equivalent of US DOE separative work units are shown on Table 2. Appendix A contains a listing of firms that sold uranium to US utilities during 1992 under new domestic purchase contracts. Appendix B contains a similar listing of firms that sold uranium to US utilities during 1992 under new import purchase contracts. Appendix C contains an explanation of Form EIA-858 survey methodologies with emphasis on the processing of Schedule B data

Method of separation of uranium from contaminating ions in an aqueous feed liquid containing uranyl ions

International Nuclear Information System (INIS)

Sundar, P.S.; Elikan, L.; Lyon, W.L.

1975-01-01

A coupled cationic/anionic method for the separation of uranium from contaminated aqueous solutions which contain uranyl ions is proposed. The fluid is extracted using an organic solvent containing a reagent which, together with the uranyl ions, forms a soluble aggregate in that solvent. As an example, 0.1 - 1 Mol/l Di-2-ethyl-hexyl-phosphorous acid in kerosene is mentioned. The organic solvent is then treated with a sealing liquid (volume ratio 20 - 35). For separation, an aqueous carbonate solution or a sulfuric acid solution can be used; the most favorable pH-values and concentrations for both cases are mentioned. The U +4 -ion at the sulfuric acid separation is subsequently oxidized to the uranyl ion with air. In each case, an extraction with an amine follows; after that, the amine is separated using an ammonium-carbonate solution and the uranium aggregate is precipitated, for example as ammonium uranyl tricarbonate, and then further processed to uranium oxide. The solvents and fluids used are led back in closed circuit; a flow diagram is given. (UWI) [de

Overall viscoplastic behavior of non-irradiated porous nuclear ceramics

International Nuclear Information System (INIS)

Monerie, Yann; Gatt, Jean-Marie

2006-01-01

This paper deals with the overall behavior of nonlinear viscous and porous nuclear ceramics. Bi-viscous isotropic porous materials are considered: the matrix is subjected to two power-law viscosities with different exponents related to two stationary temperature-activated creeping mechanisms (scattering-creep and dislocation-creep), and this matrix contains a low porosity volume fraction. The overall behavior of these types of composite materials is obtained with the help of quadratic strain-rate potentials combined with experimental-based coupling function depending on stress and temperature. For each creeping mechanism, the hollow sphere model of [Michel, J.-C., Suquet, P., 1992. The constitutive law of nonlinear viscous and porous materials. Journal of the Mechanics and Physics of Solids 40, 783-812] is used. Mechanical parameters of the resulting model are identified and validated in the particular case of non-irradiated uranium dioxide nuclear ceramics. This model predicts, under pure thermo-mechanical loading, a variation of the material volume and a variation of the porosity volume fraction (the so-called densification or swelling). (authors)

Treatment of wastewater containing phenol using a tubular ceramic membrane bioreactor.

Science.gov (United States)

Ersu, C B; Ong, S K

2008-02-01

The performance of a membrane bioreactor (MBR) with a tubular ceramic membrane for phenol removal was evaluated under varying hydraulic retention times (HRT) and a fixed sludge residence time (SRT) of 30 days. The tubular ceramic membrane was operated with a mode of 15 minutes of filtration followed by 15 seconds of permeate backwashing at a flux of 250 l m(-2)hr(-1) along with an extended backwashing of 30 seconds every 3 hours of operation, which maintained the transmembrane pressure (TMP) below 100 kPa. Using a simulated municipal wastewater with varying phenol concentrations, the chemical oxygen demand (COD) and phenol removals observed were greater than 88% with excellent suspended solids (SS) removal of 100% at low phenol concentrations (approx. 100 mg l(-1) of phenol). Step increases in phenol concentration showed that inhibition was observed between 600 to 800 mg l(-1) of phenol with decreased sludge production rate, mixed liquor suspended solids (MLSS) concentration, and removal performance. The sludge volume index (SVI) of the biomass increased to about 450 ml g(-1) for a phenol input concentration of 800 mg l(-1). When the phenol concentration was decreased to 100 mg l(-1), the ceramic tubular MBR was found to recover rapidly indicating that the MBR is a robust system retaining most of the biomass. Experimental runs using wastewater containing phenol indicated that the MBR can be operated safely without upsets for concentrations up to 600 mg l(-1) of phenol at 2-4 hours HRT and 30 days SRT.

A uranium bed with ceramic body for tritium storage

Energy Technology Data Exchange (ETDEWEB)

Khapov, A.S.; Grishechkin, S.K.; Kiselev, V.G. [' All Russia Research Institute of Automatics' - FSUE VNIIA, Moscow (Russian Federation)

2015-03-15

It is widely recognized that ceramic coatings provide an attractive solution to lower tritium permeation in structural materials. Alumina based ceramic coatings have the highest permeation reduction factor for hydrogen. For this reason an attempt was made to apply crack-free low porous ceramics as a structural material of a bed body for tritium storage in a setup used for hydrogenating neutron tube targets at VNIIA. The present article introduces the design of the bed. This bed possesses essentially a lower hydrogen permeation factor than traditionally beds with stainless steel body. Bed heating in order to recover hydrogen from the bed is suggested to be implemented by high frequency induction means. Inductive heating allows decreasing the time necessary for tritium release from the bed as well as power consumption. Both of these factors mean less thermal power release into glove box where a setup for tritium handling is installed and thus causes fewer problems with pressure regulations inside the glove box. Inductive heating allows raising tritium sorbent material temperature up to melting point. The latter allows achieving nearly full tritium recovery.

Trace elements in ancient ceramics: Pt.4

International Nuclear Information System (INIS)

Li Huhou; Sun Yongjun; Zhang Xiangdong

1987-01-01

In the last period of Tong Dynasty, Jingdezhen began its production of ceramics. During the Song Dynasty, the ceramic industry greatly developed and produced fine white ware at Hutian. In the Yuan Dynastry, Hutian became the centre of production making the world famous blue and white wares. Here are reported results of analyses of ancient porcelians of Hutian in Jiangdezhen by reactor neutron activation analysis. The results show that the patterns of eight rare earth elements are apparently different for products in different periods, indicating that methods for producing ceramics or kinds of clay used were different. The contents of some other trace elements such as hafnium, tantalum, thorium and uranium show the same regularity in difference of composition also

Study of the quenching and subsequent return to room temperature of uranium-chromium, uranium-iron, and uranium-molybdenum alloys containing only small amounts of the alloying element; Etude de la trempe et du revenu a la temperature ordinaire d'alliages uranium-chrome, uranium-fer et uranium-molybdene, a faible teneur en element d'alliage

Energy Technology Data Exchange (ETDEWEB)

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

1960-09-15

By means of an apparatus which makes possible thermal pre-treatments in vacuo, quenching carried out in a high purity argon atmosphere, and simultaneous recording of time temperature cooling and thermal contraction curves, the author has examined the transformations which occur in uranium-chromium, uranium-iron and uranium-molybdenum alloys during their quenching and subsequent return to room temperature. For uranium-chromium and uranium-iron alloys, the temperature at which the {gamma} {yields} {beta} transformation starts varies very little with the rate of cooling. For uranium-molybdenum alloys containing 2,8 atom per cent of Mo, this temperature is lowered by 120 deg. C for a cooling rate of 500 deg. C/mn. The temperature at which the {beta} {yields} {alpha} transformation starts is lowered by 170 deg. C for a cooling rate of 500 deg. C/mn in the case of uranium-chromium alloy containing 0,37 atom per cent of Cr. The temperature is little affected in the case of uranium-iron alloys. The addition of chromium or iron makes it possible to conserve the form {beta} at ordinary temperatures after quenching from the {beta} and {gamma} regions. The {beta} phase is particularly unstable and changes into needles of the {alpha} form even at room temperatures according to an autocatalytic transformation law similar to the austenitic-martensitic transformation law in the case of iron. The {beta} phase obtained by quenching from the {beta} phase region is more stable than that obtained by quenching from the {gamma} region. Chromium is a more effective stabiliser of the {beta} phase than is iron. Unfortunately it causes serious surface cracking. The {beta} {yields} {alpha} transformation in uranium-chromium alloys has been followed at room temperature by means of micro-cinematography. The author has not observed the direct {gamma} {yields} {alpha} transformation in uranium-molybdenum alloys containing 2,8 per cent of molybdenum even for cooling rates of up to 2000 deg. C

Conversion of wood flour/SiO2/phenolic composite to porous SiC ceramic containing SiC whiskers

Directory of Open Access Journals (Sweden)

Li Zhong

2013-01-01

Full Text Available A novel wood flour/SiO2/phenolic composite was chosen to be converted into porous SiC ceramic containing SiC whiskers via carbothermal reduction. At 1550°C the composite is converted into porous SiC ceramic with pore diameters of 10~40μm, and consisting of β-SiC located at the position of former wood cell walls. β-SiC wire-like whiskers of less than 50 nm in diameter and several tens to over 100 μm in length form within the pores. The surface of the resulting ceramic is coated with β-SiC necklace-like whiskers with diameters of 1~2μm.

Radioactivity and associated radiation hazards in ceramic raw materials and end products.

Science.gov (United States)

Viruthagiri, G; Rajamannan, B; Suresh Jawahar, K

2013-12-01

Studies have been planned to obtain activity and associated radiation hazards in ceramic raw materials (quartz, feldspar, clay, zircon, kaolin, grog, alumina bauxite, baddeleyite, masse, dolomite and red mud) and end products (ceramic brick, glazed ceramic wall and floor tiles) as the activity concentrations of uranium, thorium and potassium vary from material to material. The primordial radionuclides in ceramic raw materials and end products are one of the sources of radiation hazard in dwellings made of these materials. By the determination of the activity level in these materials, the indoor radiological hazard to human health can be assessed. This is an important precautionary measure whenever the dose rate is found to be above the recommended limits. The aim of this work was to measure the activity concentration of (226)Ra, (232)Th and (40)K in ceramic raw materials and end products. The activity of these materials has been measured using a gamma-ray spectrometry, which contains an NaI(Tl) detector connected to multichannel analyser (MCA). Radium equivalent activity, alpha-gamma indices and radiation hazard indices associated with the natural radionuclides are calculated to assess the radiological aspects of the use of the ceramic end products as decorative or covering materials in construction sector. Results obtained were examined in the light of the relevant international legislation and guidance and compared with the results of similar studies reported in different countries. The results suggest that the use of ceramic end product samples examined in the construction of dwellings, workplace and industrial buildings is unlikely to give rise to any significant radiation exposure to the occupants.

Radioactivity and associated radiation hazards in ceramic raw materials and end products

International Nuclear Information System (INIS)

Viruthagiri, G.; Rajamannan, B.; Suresh Jawahar, K.

2013-01-01

Studies have been planned to obtain activity and associated radiation hazards in ceramic raw materials (quartz, feldspar, clay, zircon, kaolin, grog, alumina bauxite, baddeleyite, masse, dolomite and red mud) and end products (ceramic brick, glazed ceramic wall and floor tiles) as the activity concentrations of uranium, thorium and potassium vary from material to material. The primordial radionuclides in ceramic raw materials and end products are one of the sources of radiation hazard in dwellings made of these materials. By the determination of the activity level in these materials, the indoor radiological hazard to human health can be assessed. This is an important precautionary measure whenever the dose rate is found to be above the recommended limits. The aim of this work was to measure the activity concentration of 226 Ra, 232 Th and 40 K in ceramic raw materials and end products. The activity of these materials has been measured using a gamma-ray spectrometry, which contains an NaI(Tl) detector connected to multichannel analyser (MCA). Radium equivalent activity, alpha-gamma indices and radiation hazard indices associated with the natural radionuclides are calculated to assess the radiological aspects of the use of the ceramic end products as decorative or covering materials in construction sector. Results obtained were examined in the light of the relevant international legislation and guidance and compared with the results of similar studies reported in different countries. The results suggest that the use of ceramic end product samples examined in the construction of dwellings, workplace and industrial buildings is unlikely to give rise to any significant radiation exposure to the occupants. (authors)

Novel titanium dioxide ceramics containing bismuth and antimony

Directory of Open Access Journals (Sweden)

Zhenwei Li

2017-06-01

Full Text Available Here, we developed one kind of novel TiO2 ceramics with colossal dielectric constant by chemical modifications (Bi3+ and Sb5+, and discussed the physical origin for giant dielectric constant. Effects of Bi and/or Sb on their microstructure, dielectric properties as well as its frequency and temperature stability were studied in detail. It was found that their dielectric properties are strongly sensitive to (Bi,Sb contents, and colossal dielectric permittivity (CP (104∼105 together with low dielectric loss (∼5.7% can be obtained in a wide composition range. In addition, all the ceramics possessed good frequency (102∼106 Hz and temperature (−150–200 °C stability of dielectric properties. In addition, the defects caused by the Bi volatilization may be the reason for higher dielectric properties of (Bi0.5Sb0.5xTi1−xO2 ceramics with respect to (A0.5Sb0.5xTi1−xO2 (A = In, Pr, Dy, Sm, Gd, Yb, Ga, Al, Fe or Sc. According to the results of complex impedance and XPS, the electron-pinned defect-dipoles may be suitable to explain the CP phenomenon, and oxygen vacancies-induced by Bi3+&Sb5+ substitution for Ti4+ should be responsible for conduction mechanism. We believe that this profound investigation can benefit the development of TiO2 ceramics as a CP material.

Hydrothermal uranium deposits containing molybdenum and fluorite in the Marysvale volcanic field, west-central Utah

Science.gov (United States)

Cunningham, C.G.; Rasmussen, J.D.; Steven, T.A.; Rye, R.O.; Rowley, P.D.; Romberger, S.B.; Selverstone, J.

1998-01-01

Uranium deposits containing molybdenum and fluorite occur in the Central Mining Area, near Marysvale, Utah, and formed in an epithermal vein system that is part of a volcanic/hypabyssal complex. They represent a known, but uncommon, type of deposit; relative to other commonly described volcanic-related uranium deposits, they are young, well-exposed and well-documented. Hydrothermal uranium-bearing quartz and fluorite veins are exposed over a 300 m vertical range in the mines. Molybdenum, as jordisite (amorphous MoS2, together with fluorite and pyrite, increase with depth, and uranium decreases with depth. The veins cut 23-Ma quartz monzonite, 20-Ma granite, and 19-Ma rhyolite ash-flow tuff. The veins formed at 19-18 Ma in a 1 km2 area, above a cupola of a composite, recurrent, magma chamber at least 24 ?? 5 km across that fed a sequence of 21- to 14-Ma hypabyssal granitic stocks, rhyolite lava flows, ash-flow tuffs, and volcanic domes. Formation of the Central Mining Area began when the intrusion of a rhyolite stock, and related molybdenite-bearing, uranium-rich, glassy rhyolite dikes, lifted the fractured roof above the stock. A breccia pipe formed and relieved magmatic pressures, and as blocks of the fractured roof began to settle back in place, flat-lying, concave-downward, 'pull-apart' fractures were formed. Uranium-bearing, quartz and fluorite veins were deposited by a shallow hydrothermal system in the disarticulated carapace. The veins, which filled open spaces along the high-angle fault zones and flat-lying fractures, were deposited within 115 m of the ground surface above the concealed rhyolite stock. Hydrothermal fluids with temperatures near 200??C, ??18OH2O ~ -1.5, ?? -1.5, ??DH2O ~ -130, log fO2 about -47 to -50, and pH about 6 to 7, permeated the fractured rocks; these fluids were rich in fluorine, molybdenum, potassium, and hydrogen sulfide, and contained uranium as fluoride complexes. The hydrothermal fluids reacted with the wallrock resulting in

Evaluation of sol–gel based magnetic 45S5 bioglass and bioglass–ceramics containing iron oxide

International Nuclear Information System (INIS)

Shankhwar, Nisha; Srinivasan, A.

2016-01-01

Multicomponent oxide powders with nominal compositions of (45 − x)·SiO_2·24.5CaO·24.5Na_2O·6P_2O_5xFe_2O_3 (in wt.%) were prepared by a modified sol–gel procedure. X-ray diffraction (XRD) patterns and high resolution transmission electron microscope images of the sol–gel products show fully amorphous structure for Fe_2O_3 substitutions up to 2 wt.%. Sol–gel derived 43SiO_2·24.5CaO·24.5Na_2O·6P_2O_5·2Fe_2O_3 glass (or bioglass 45S5 with SiO_2 substituted with 2 wt.% Fe_2O_3), exhibited magnetic behavior with a coercive field of 21 Oe, hysteresis loop area of 33.25 erg/g and saturation magnetization of 0.66 emu/g at an applied field of 15 kOe at room temperature. XRD pattern of this glass annealed at 850 °C for 1 h revealed the formation of a glass–ceramic containing sodium calcium silicate and magnetite phases in nanocrystalline form. Temperature dependent magnetization and room temperature electron spin resonance data have been used to obtain information on the magnetic phase and distribution of iron ions in the sol–gel glass and glass–ceramic samples. Sol–gel derived glass and glass–ceramic exhibit in-vitro bioactivity by forming a hydroxyapatite surface layer under simulated physiological conditions and their bio-response is superior to their melt quenched bulk counterparts. This new form of magnetic bioglass and bioglass ceramics opens up new and more effective biomedical applications. - Highlights: • Bioglass 45S5 containing 2 wt.% Fe_2O_3 is prepared by sol–gel route. • Fully amorphous bioglass exhibits spontaneous magnetization. • Gel powders with more than 2 wt.% Fe_2O_3 formed glass–ceramics. • γ-Fe_2O_3 in bioglass transformed irreversibly to magnetite upon heat treatment. • In vitro bioactivity of sol–gel samples is superior to their bulk counterparts.

Management and Handling of Rejected Fuel of MTR Type and Process Effluents Contained Uranium at FEPI

International Nuclear Information System (INIS)

Ghaib Widodo; Bambang Herutomo

2007-01-01

Research Reactor Fuel Element Production Installation (FEPI) - Serpong has performed management and handling of all kinds of rejected fuel material during production (solids, liquids, and gases) and process effluents contained uranium. The methods that has been implemented are precipitation, absorption, evaporation, electrolysis, and electrodialysis. By these methods will finally be obtained forms of product which can be used directly as fuel material feed and solid/liquid radioactive waste that fulfil the requirements (uranium contents < 50 ppm) to be send to Radioactive Waste Management Installation. (author)

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

Method for Waterproofing Ceramic Materials

Science.gov (United States)

Cagliostro, Domenick E. (Inventor); Hsu, Ming-Ta S. (Inventor)

1998-01-01

Hygroscopic ceramic materials which are difficult to waterproof with a silane, substituted silane or silazane waterproofing agent, such as an alumina containing fibrous, flexible and porous, fibrous ceramic insulation used on a reentry space vehicle, are rendered easy to waterproof if the interior porous surface of the ceramic is first coated with a thin coating of silica. The silica coating is achieved by coating the interior surface of the ceramic with a silica precursor converting the precursor to silica either in-situ or by oxidative pyrolysis and then applying the waterproofing agent to the silica coated ceramic. The silica precursor comprises almost any suitable silicon containing material such as a silane, silicone, siloxane, silazane and the like applied by solution, vapor deposition and the like. If the waterproofing is removed by e.g., burning, the silica remains and the ceramic is easily rewaterproofed. An alumina containing TABI insulation which absorbs more that five times its weight of water, absorbs less than 10 wt. % water after being waterproofed according to the method of the invention.

High loading uranium plate

International Nuclear Information System (INIS)

Wiencek, T.C.; Domagala, R.F.; Thresh, H.R.

1990-01-01

Two embodiments of a high uranium fuel plate are disclosed which contain a meat comprising structured uranium compound confined between a pari of diffusion bonded ductile metal cladding plates uniformly covering the meat, the meat hiving a uniform high fuel loading comprising a content of uranium compound greater than about 45 Vol. % at a porosity not greater than about 10 Vol. %. In a first embodiment, the meat is a plurality of parallel wires of uranium compound. In a second embodiment, the meat is a dispersion compact containing uranium compound. The fuel plates are fabricated by a hot isostatic pressing process

Storage of plutonium and nuclear power plant actinide waste in the form of critical-mass-free ceramics containing neutron poisons

Energy Technology Data Exchange (ETDEWEB)

Nadykto, B.A. [RFNC-VNIIEF, Nizhni Novgorod Region (Russian Federation)

2001-07-01

The nuclear weapons production has resulted in accumulation of a large quantity of plutonium and uranium highly enriched with uranium-235 isotope (many tons). The work under ISTC Project 332B-97 treated the issues of safe plutonium storage through making critical-mass-free plutonium oxide compositions with neutron poisons. This completely excludes immediate utilization (without chemical reprocessing) of retained plutonium in nuclear devices. It is therewith possible to locate plutonium most compactly in the storage facility, which would allow reduction in required storage areas and costs. The issues of the surplus weapon-grade plutonium management and utilization have been comprehensively studied in the recent decade. The issues are treated in multiple scientific publications, conferences, and seminars. At the same time, issues of nuclear power engineering actinide waste storage are studied no less extensively. The general issues are material radioactivity and energy release and nuclear accident hazards due to critical mass generation. Plutonium accumulated in nuclear power plant spent fuel is more accessible than weapon-grade plutonium and can become of higher and higher interest with time as its activity reduces, including as material for nuclear devices. The urgency of plutonium management is presently related not only to accumulation of surplus weapon-grade plutonium, but also to the fact that it is high time to decide what has to be done regarding reactor plutonium. Presently, the possibility of actinide separation from NPP spent nuclear fuel and compact underground burial separately from other (mainly fragment) activity is being considered. Actinide and neutron poison base critical-mass-free ceramic materials (similar to plutonium ceramics) may be useful for this burial method. (author)

Storage of plutonium and nuclear power plant actinide waste in the form of critical-mass-free ceramics containing neutron poisons

International Nuclear Information System (INIS)

Nadykto, B.A.

2001-01-01

The nuclear weapons production has resulted in accumulation of a large quantity of plutonium and uranium highly enriched with uranium-235 isotope (many tons). The work under ISTC Project 332B-97 treated the issues of safe plutonium storage through making critical-mass-free plutonium oxide compositions with neutron poisons. This completely excludes immediate utilization (without chemical reprocessing) of retained plutonium in nuclear devices. It is therewith possible to locate plutonium most compactly in the storage facility, which would allow reduction in required storage areas and costs. The issues of the surplus weapon-grade plutonium management and utilization have been comprehensively studied in the recent decade. The issues are treated in multiple scientific publications, conferences, and seminars. At the same time, issues of nuclear power engineering actinide waste storage are studied no less extensively. The general issues are material radioactivity and energy release and nuclear accident hazards due to critical mass generation. Plutonium accumulated in nuclear power plant spent fuel is more accessible than weapon-grade plutonium and can become of higher and higher interest with time as its activity reduces, including as material for nuclear devices. The urgency of plutonium management is presently related not only to accumulation of surplus weapon-grade plutonium, but also to the fact that it is high time to decide what has to be done regarding reactor plutonium. Presently, the possibility of actinide separation from NPP spent nuclear fuel and compact underground burial separately from other (mainly fragment) activity is being considered. Actinide and neutron poison base critical-mass-free ceramic materials (similar to plutonium ceramics) may be useful for this burial method. (author)

Anthropogenic materials and products containing natural radionuclides. Pt. 2. Examination of radiation doses resulting from occupational exposure

International Nuclear Information System (INIS)

Reichelt, A.; Lehmann, K.H.

1993-11-01

The radiation doses are determined on the basis of dosimetric scanning of the materials and products and measurement of the ambient dose rates and inhaled doses at the place of work. For all places and conditions exmined, the average annual effective dose (ICRP) is of the order of 20mSv/annum. The substances and products examined are phosphate fertilizers. thoriated tungsten electrodes, or glass gas hoods, respectively, dental material containing uranium, and dental ceramics containing zirconium sands. The report also gives information on the occupational exposure in drinking-water conditioning plants. (Orig./DG) [de

Manufacturing of superconductive silver/ceramic composites

DEFF Research Database (Denmark)

Seifi, Behrouz; Bech, Jakob Ilsted; Eriksen, Morten

2000-01-01

Manufacturing of superconducting metal/ceramic composites is a rather new discipline within materials forming processes. High Temperature SuperConductors, HTSC, are manufactured applying the Oxide-Powder-In-Tube process, OPIT. A ceramic powder containing lead, calcium, bismuth, strontium, and cop......Manufacturing of superconducting metal/ceramic composites is a rather new discipline within materials forming processes. High Temperature SuperConductors, HTSC, are manufactured applying the Oxide-Powder-In-Tube process, OPIT. A ceramic powder containing lead, calcium, bismuth, strontium...

Controlled thermolysis of uranium (alkoxy)siloxy complexes. A route to polymetallic complexes of low-valent uranium

Energy Technology Data Exchange (ETDEWEB)

Camp, Clement; Pecaut, Jacques; Mazzanti, Marinella [CEA-Grenoble (France). Lab. de Reconnaissance Ionique et Chimie de Coordination; Kefalidis, Christos E.; Maron, Laurent [Toulouse Univ. (France). LPCNO, CNRS et INSA, UPS

2013-11-25

Decomposition into higher species: Intramolecular U{sup III}-mediated homolytic C-O bond cleavage in U{sup III} (alkoxy)siloxy complexes at low temperature and subsequent reduction with KC{sub 8} led to unprecedented polymetallic complexes containing siloxy, silanediolate, and silanetriolate ligands (see example: U green, Si yellow, K blue, O red). Such compounds may be useful precursors to uranium ceramics relevant for catalysis and the storage of spent nuclear fuel. [German] Zerfall in hoehere Spezies: Die intramolekulare U{sup III}-vermittelte homolytische C-O-Spaltung in U{sup III}-(Alkoxy)siloxy-Komplexen bei tiefer Temperatur mit nachfolgender Reduktion mit KC{sub 8} fuehrte zu ungewoehnlichen Polymetallkomplexen mit Siloxy-, Silandiolat- und Silantriolatliganden (siehe Beispiel: U gruen, Si gelb, K blau, O rot). Solche Verbindungen sind nuetzliche Vorstufen von Urankeramiken, die fuer die Katalyse und fuer die Speicherung verbrauchter Kernbrennstoffe wichtig sind.

Spectroscopic properties of Er3+ and Yb3+ co-doped glass ceramics containing SrF2 nanocrystals

International Nuclear Information System (INIS)

Qiao Xvsheng; Fan Xianping; Wang Minquan; Zhang Xianghua

2009-01-01

The spectroscopic properties of Er 3+ /Yb 3+ co-doped 50SiO 2 -10Al 2 O 3 -20ZnF 2 -20SrF 2 glass and glass ceramic containing SrF 2 nanocrystals were investigated. The formation of SrF 2 nanocrystals in the glass ceramic was confirmed by XRD. The oscillator strengths for several transitions of the Er 3+ ions in the glass ceramic have been obtained and the Judd-Ofelt parameters were then determined. The XRD result and Judd-Ofelt parameters suggested that Er 3+ and Yb 3+ ions had efficiently enriched in the SrF 2 nanocrystals in the glass ceramic. The lifetime of excited states has been used to reveal the surroundings of luminescent Er 3+ and Yb 3+ and energy transfer (ET) mechanism between Er 3+ and Yb 3+ . Much stronger upconversion luminescence and longer lifetime of the Er 3+ /Yb 3+ co-doped glass ceramic were observed in comparison with the Er 3+ /Yb 3+ co-doped glass, which could be ascribed to more efficient ET from Yb 3+ to Er 3+ due to the enrichment of Yb 3+ and Er 3+ and the shortening of the distance between lanthanide ions in the precipitated SrF 2 nanocrystals.

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

International Nuclear Information System (INIS)

Koch, G.

1975-01-01

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

Uranium recovering from slags generated in the metallic uranium by magnesiothermic reduction

International Nuclear Information System (INIS)

Fornarolo, F.; Carvalho, E.F. Urano de; Durazzo, M.; Riella, H.G.

2008-01-01

The Nuclear Fuel Center of IPEN/CNEN-SP has recent/y concluded a program for developing the fabrication technology of the nuclear fuel based on the U 3 Si 2 -Al dispersion, which is being used in the IEA-R1 research reactor. The uranium silicide (U 3 Si 2 ) fuel production starts with the uranium hexafluoride (UF 6 ) processing and uranium tetrafluoride (UF 4 ) precipitation. Then, the UF 4 is converted to metallic uranium by magnesiothermic reduction. The UF 4 reduction by magnesium generates MgF 2 slag containing considerable concentrations of uranium, which could reach 20 wt%. The uranium contained in that slag should be recovered and this work presents the results obtained in recovering the uranium from that slag. The uranium recovery is accomplished by acidic leaching of the calcined slag. The calcination transforms the metallic uranium in U 3 O 8 , promoting the pulverization of the pieces of metallic uranium and facilitating the leaching operation. As process variables, have been considered the nitric molar concentration, the acid excess regarding the stoichiometry and the leaching temperature. As result, the uranium recovery reached a 96% yield. (author)

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

Liquid membranes and process for uranium recovery therewith

International Nuclear Information System (INIS)

Frankenfeld, J.W.; Li, N.N.T.; Bruncati, R.L.

1981-01-01

A liquid membrane system consisting of water-in-oil type emulsions dispersed in water, which is capable of extracting uranium-containing ions from an aqueous feed solution containing uranium ions at a temperature in the range of 25 0 C to 80 0 C, is described. The emulsion comprises an aqueous interior phase surrounded by a surfactant-containing exterior phase. The exterior phase is immiscible with the interior phase and comprises a transfer agent capable of transporting selectively the desired uranium-containing ions and a solvent for the transfer agent. The interior phase comprises a reactant capable of removing uranium-containing ions from the transfer agent and capable of changing the valency of the uranium in uranium-containing ions to a second valency state and converting the uranium-containing ions into a nonpermeable form. (U.K.)

Extraction of uranium(VI) by emulsion liquid membrane containing 5,8-diethyl-7-hydroxy-6-dodecanone oxime

International Nuclear Information System (INIS)

Akiba, Kenichi; Takahashi, Toshihiko; Kanno, Takuji

1984-01-01

Extraction of uranium(VI) by a liquid surfactant membrane has been studied. The stability of water-in-oil (w/o) emulsion dispersed in the continuous aqueous phase increased with an increase in surfactant concentrations and in the fraction of the organic phase in emulsion globules. Uranium(VI) in dilute acid solutions was extracted into (w/o) emulsions containing 5,8-diethyl-7-hydroxy-6-dodecanone oxime (LIX 63) as a mobile carrier and its concentration decreased according to [U]sub(t)=[U]sub(o)exp(-ksub(obsd)t). The apparent rate constants (ksub(obsd)) increased with an increase in carrier concentrations and in external pH values, while they were slightly dependent on the stripping acid concentrations. Uranium was transported and concentrated into the internal aqueous droplets. The final concentration of uranium in the external aqueous phase dropped to about 10 -3 of its initial value. (author)

Spectroscopic study of local thermal effect in transparent glass ceramics containing nanoparticles

Institute of Scientific and Technical Information of China (English)

2007-01-01

Local thermal effect influencing the fluorescence of triply ionized rare earth ions doped in nanocrystals is studied with laser spectroscopy and theory of thermal transportation for transparent oxyfluoride glass ceramics containing nanocrystals. The result shows that the local temperature of the nanocrystals embedded in glass matrices is much higher than the environmental temperature of the sample. It is suggested that the temperature-dependent thermal energy induced by the light absorption must be considered when the theory of thermal transportation is applied to the study of local thermal effect.

Sulphuric Acid Resistant of Self Compacted Geopolymer Concrete Containing Slag and Ceramic Waste

Directory of Open Access Journals (Sweden)

Shafiq I.

2017-01-01

Full Text Available Malaysia is a one of the developing countries where the constructions of infrastructure is still ongoing, resulting in a high demand for concrete. In order to gain sustainability factors in the innovations for producing concrete, geopolymer concrete containing granulated blast-furnace slag and ceramics was selected as a cement replacement in concrete for this study. Since Malaysia had many ceramic productions and uses, the increment of the ceramic waste will also be high. Thus, a new idea to reuse this waste in construction materials have been tested by doing research on this waste. Furthermore, a previous research stated that Ordinary Portland Cement concrete has a lower durability compared to the geopolymer concrete. Geopolymer binders have been reported as being acid resistant and thus are a promising and alternative binder for sewer pipe manufacture. Lack of study regarding the durability of the geopolymer self-compacting concrete was also one of the problems. The waste will be undergoing a few processes in the laboratory in order to get it in the best form before undergoing the next process as a binder in geopolymer concrete. This research is very significant in order to apply the concept of sustainability in the construction field. In addition, the impact of this geopolymer binder is that it emits up to nine times less CO2 than Portland Cement.

RECOVERY OF URANIUM FROM ZIRCONIUM-URANIUM NUCLEAR FUELS

Science.gov (United States)

Gens, T.A.

1962-07-10

An improvement was made in a process of recovering uranium from a uranium-zirconium composition which was hydrochlorinated with gsseous hydrogen chloride at a temperature of from 350 to 800 deg C resulting in volatilization of the zirconium, as zirconium tetrachloride, and the formation of a uranium containing nitric acid insoluble residue. The improvement consists of reacting the nitric acid insoluble hydrochlorination residue with gaseous carbon tetrachloride at a temperature in the range 550 to 600 deg C, and thereafter recovering the resulting uranium chloride vapors. (AEC)

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

Glass-ceramic optical fiber containing Ba2TiSi2O8 nanocrystals for frequency conversion of lasers.

Science.gov (United States)

Fang, Zaijin; Xiao, Xusheng; Wang, Xin; Ma, Zhijun; Lewis, Elfed; Farrell, Gerald; Wang, Pengfei; Ren, Jing; Guo, Haitao; Qiu, Jianrong

2017-03-30

A glass-ceramic optical fiber containing Ba 2 TiSi 2 O 8 nanocrystals fabricated using a novel combination of the melt-in-tube method and successive heat treatment is reported for the first time. For the melt-in-tube method, fibers act as a precursor at the drawing temperature for which the cladding glass is softened while the core glass is melted. It is demonstrated experimentally that following heat treatment, Ba 2 TiSi 2 O 8 nanocrystals with diameters below 10 nm are evenly distributed throughout the fiber core. Comparing to the conventional rod-in-tube method, the melt-in-tube method is superior in terms of controllability of crystallization to allow for the fabrication of low loss glass-ceramic fibers. When irradiated using a 1030 nm femtosecond laser, an enhanced green emission at a wavelength of 515 nm is observed in the glass-ceramic fiber, which demonstrates second harmonic generation of a laser action in the fabricated glass-ceramic fibers. Therefore, this new glass-ceramic fiber not only provides a highly promising development for frequency conversion of lasers in all optical fiber based networks, but the melt-in-tube fabrication method also offers excellent opportunities for fabricating a wide range of novel glass-ceramic optical fibers for multiple future applications including fiber telecommunications and lasers.

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.

[Ceramic-on-ceramic bearings in total hip arthroplasty (THA)].

Science.gov (United States)

Sentürk, U; Perka, C

2015-04-01

The main reason for total hip arthroplasty (THA) revision is the wear-related aseptic loosening. Younger and active patients after total joint replacement create high demands, in particular, on the bearings. The progress, especially for alumina ceramic-on-ceramic bearings and mixed ceramics have solved many problems of the past and lead to good in vitro results. Modern ceramics (alumina or mixed ceramics containing alumina) are extremely hard, scratch-resistant, biocompatible, offer a low coefficient of friction, superior lubrication and have the lowest wear rates in comparison to all other bearings in THA. The disadvantage of ceramic is the risk of material failure, i.e., of ceramic fracture. The new generation of mixed ceramics (delta ceramic), has reduced the risk of head fractures to 0.03-0.05 %, but the risk for liner fractures remains unchanged at about 0.02 %. Assuming a non-impinging component implantation, ceramic-on-ceramic bearings have substantial advantages over all other bearings in THA. Due to the superior hardness, ceramic bearings produce less third body wear and are virtually impervious to damage from instruments during the implantation process. A specific complication for ceramic-on-ceramic bearings is "squeaking". The high rate of reported squeaking (0.45 to 10.7 %) highlights the importance of precise implant positioning and the stem and patient selection. With precise implant positioning this problem is rare with many implant designs and without clinical relevance. The improved tribology and the presumable resulting implant longevity make ceramic-on-ceramic the bearing of choice for young and active patients. Georg Thieme Verlag KG Stuttgart · New York.

Piezoelectric ceramic material, containing PbNb2O6, K2Nb2O6

International Nuclear Information System (INIS)

Fesenko, E.G.; Filip'ev, V.S.; Razumovskaya, O.N.; Cherner, Ya.E.; Rudkovskaya, L.M.; Zav'yalov, V.P.; Molchanova, R.A.; Kryshtop, V.G.; Panich, A.E.; Servuli, V.A.

1984-01-01

A new piezoelectric ceramic material including PbNb 2 O 6 , K 2 Nb 2 O 6 is prepared. Above the new material contains Nb 2 O 5 . The invention relates to piezotechnique. The principal advantage of this material for acoustic converters is high anisotropy of piezoelectric properties as well as high Curie temperature (T C =539-553 deg C). The composition containing 93.96 mole% PbNb 2 O 6 ; 2.48 mole% K 2 Nb 2 O 6 and 3.56 mole% Nb 2 O 5 has optimum content of parameters

Evaluation of plutonium, uranium, and thorium use in power reactor fuel cycles

International Nuclear Information System (INIS)

Kasten, P.R.; Homan, F.J.

1977-01-01

The increased cost of uranium and separative work has increased the attractiveness of plutonium use in both uranium and thorium fuel cycles in thermal reactors. A technology, fuel utilization, and economic evaluation is given for uranium and thorium fuel cycles in various reactor types, along with the use of plutonium and 238 U. Reactors considered are LWRs, HWRs, LWBRs, HTGRs, and FBRs. Key technology factors are fuel irradiation performance and associated physical property values. Key economic factors are unit costs for fuel fabrication and reprocessing, and for refabrication of recycle fuels; consistent cost estimates are utilized. In thermal reactors, the irradiation performance of ceramic fuels appears to be satisfactory. At present costs for uranium ore and separative work, recycle of plutonium with thorium rather than uranium is preferable from fuel utilization and economic viewpoints. Further, the unit recovery cost of plutonium is lower from LWR fuels than from natural-uranium HWR fuels; use of LWR product permits plutonium/thorium fueling to compete with uranium cycles. Converting uranium cycles to thorium cycles increases the energy which can be extracted from a given uranium resource. Thus, additional fuel utilization improvement can be obtained by fueling all thermal reactors with thorium, but this requires use of highly enriched uranium; use of 235 U with thorium is most economic in HTGRs followed by HWRs and then LWRs. Marked improvement in long-term fuel utilization can be obtained through high thorium loadings and short fuel cycle irradiations as in the LWBR, but this imposes significant economic penalties. Similar operating modes are possible in HWRs and HTGRs. In fast reactors, use of the plutonium-uranium cycle gives advantageous fuel resource utilization in both LMFBRs and GCFRs; use of the thorium cycle provides more negative core reactivity coefficients and more flexibility relative to use of recycle fuels containing uranium of less than 20

Hydrometallurgic treatment of a mineral containing uranium, vanadium and phosphorus

International Nuclear Information System (INIS)

Echenique, Patricia; Fruchtenicht, Fernando; Gil, Daniel; Vigo, Daniel; Bouza, Angel; Vert, Gabriela; Becquart, Elena

1987-01-01

A preliminary study of a mineral has been made towards the hydrometallurgy separation of uranium, vanadium and phosphorus. After the ore dressing, work on sulfuric acid with oxidation leaching has been made, to get the uranium, vanadium and phosphorus in solution. For the separation and purification of these elements, two alternative solvent extraction methods have been tested. One of them has been the extraction of uranium and vanadium and a selective stripping of both elements. The second one has been the selective extraction of uranium and vanadium at different aqueous solutions pH. In both methods, the same reagent has been used: di(2-ethylhexyl) phosphoric acid, kerosene as diluent with two different synergistic agents: TOPO (tri-n-octyl phosphine oxide) and TBP (tri-n-butyl phosphate). Batch studies have been made to determine the equilibrium isotherms for uranium and vanadium. A continuous countercurrent simulation method has been used to get the best phase ratio and to test different stripping agents. For the first method, an important loss of uranium and vanadium at the feed solution conditioning for the extraction step has been observed. For the second method, a good recovery of uranium has been reached, but there has been losses of vanadium in pH adjustment. Nevertheless, among these processes, the last seems to work better in this mineral hydrometallurgy. (Author) [es

Uranium health physics

International Nuclear Information System (INIS)

1980-01-01

This report contains the papers delivered at the Summer School on Uranium Health Physics held in Pretoria on the 14 and 15 April 1980. The following topics were discussed: uranium producton in South Africa; radiation physics; internal dosimetry and radiotoxicity of long-lived uranium isotopes; uranium monitoring; operational experience on uranium monitoring; dosimetry and radiotoxicity of inhaled radon daughters; occupational limits for inhalation of radon-222, radon-220 and their short-lived daughters; radon monitoring techniques; radon daughter dosimeters; operational experience on radon monitoring; and uranium mill tailings management

A new classification system for all-ceramic and ceramic-like restorative materials.

Science.gov (United States)

Gracis, Stefano; Thompson, Van P; Ferencz, Jonathan L; Silva, Nelson R F A; Bonfante, Estevam A

2015-01-01

Classification systems for all-ceramic materials are useful for communication and educational purposes and warrant continuous revisions and updates to incorporate new materials. This article proposes a classification system for ceramic and ceramic-like restorative materials in an attempt to systematize and include a new class of materials. This new classification system categorizes ceramic restorative materials into three families: (1) glass-matrix ceramics, (2) polycrystalline ceramics, and (3) resin-matrix ceramics. Subfamilies are described in each group along with their composition, allowing for newly developed materials to be placed into the already existing main families. The criteria used to differentiate ceramic materials are based on the phase or phases present in their chemical composition. Thus, an all-ceramic material is classified according to whether a glass-matrix phase is present (glass-matrix ceramics) or absent (polycrystalline ceramics) or whether the material contains an organic matrix highly filled with ceramic particles (resin-matrix ceramics). Also presented are the manufacturers' clinical indications for the different materials and an overview of the different fabrication methods and whether they are used as framework materials or monolithic solutions. Current developments in ceramic materials not yet available to the dental market are discussed.

Catalytic properties of mineral ion-exchangers previously used in uranium ore wastes treatment. I. Ammoniac octylation in gaseous phase using Y-type faujasite catalysts containing mainly uranium

International Nuclear Information System (INIS)

Azzouz, A.; Nibou, D.; Abbad, B.; Achache, M.

1990-06-01

Y-type faujasite, previously used in purifying aqueous wastes containing radioactive elements, are studied as catalysts in heterogeneous reactions such as octanol amination. This process consists in alkylating ammoniac NH 3 by octanol-1 in a flow reactor containing a fixed bed of pelleted catalysts. Acidic catalysts like Y-type faujasite impregnated with low concentration leaching solution containing between 10 and 1000 ppm of uranium show interesting activities and selectivities in yielding primary, secondary and tertiary amines. High octanol conversions were obtained reaching 60 mol.% for appreciable selectivities for producing amines of about 60-80 mol.%. Compared to the non-active fresh zeolite, the catalysts obtained present good activities essentially due to the presence of uranyl cations UO 2 ++ . The latter seems to play the main role in activating the zeolite by enhancing its Bronsted surface acidity. Another interest of this process consists in obtaining Trioctylamine (TOA), an important extraction agent in uranium hydrometallurgy or its derivatives (MOA and DOA) which are very used in chemical industry

Luminescent properties of Eu{sup 3+}-doped glass ceramics containing BaCl{sub 2} nanocrystals under NUV excitation for White LED

Energy Technology Data Exchange (ETDEWEB)

Jin, Han; Mo, Zhaojun, E-mail: mzjmzj163@163.com; Zhang, Xiaosong; Yuan, Linlin; Yan, Ming; Li, Lan, E-mail: lilan@tjut.edu.cn

2016-07-15

Eu{sup 3+} doped fluorozirconate glass ceramics containing BaCl{sub 2} nanocrystals were successfully fabricated by melt quenching method, and their structural and luminous properties were investigated. The existence of BaCl{sub 2} nanocrystals in the glass ceramics plays an important role on the improvement of luminescent properties. The emission intensity in glass ceramics was remarkably enhanced, which attributes to the phonon energy decrease by Eu{sup 3+} ions into BaCl{sub 2} nanocrystals. Meanwhile, the extended average fluorescence decay lifetime from 4.60 ms to 5.42 ms and the decreased Red/Orange ratio and spark splitting of {sup 7}F{sub 1} energy level also confirmed this view. Additionally, the excitation spectra showed that glass ceramics could be effectively excited by NUV light. The CIE chromaticity coordinates of glass ceramics (GC320) were calculated as (0.611, 0.371), which was close to the NTSC standard values for red (0.67, 0.33). The results suggested that the glass ceramics may be used as potential red phosphors under UV light excitation for white light-emitting diodes.

Ceramic UO2 powder production at Cameco Corporation

International Nuclear Information System (INIS)

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

1997-01-01

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

Effect of shape and size of amidoxime-group-containing adsorbent on the recovery of uranium from sea water

International Nuclear Information System (INIS)

Omichi, H.; Kataki, A.; Sugo, T.; Okamoto, J.; Katoh, S.; Sakane, K.; Sugasaka, K.; Itagaki, T.

1987-01-01

An amidoxime-group-containing adsorbent for the recovery of uranium from sea water was synthesized by radiation-induced graft polymerization of acrylonitrile onto polypropylene fiber of round and cross-shaped sections. The tensile strength and elongation of the synthesized adsorbent, both of which were one-half those of the raw material, were not affected by the shape of the fiber. The deterioration of the adsorption ability induced by immersing the adsorbent in HCl was negligible because of the short immersion time required for the desorption with HCl. The concentration factors for uranium and transition metals in 28 days were in the order of 10 5 , while those for alkali metals and alkaline earth metals were in the order 10 -1 -10 1 . The recovery of uranium with the cross-shaped adsorbent was superior to that of the round-shaped one. XMA line profiles show that the distribution of uranium is much restricted to the surface layer when compared with that of alkaline earth metals. Diminishing the diameter or increasing the surface area was effective for increasing the adsorption of uranium

Trends in uranium supply

Energy Technology Data Exchange (ETDEWEB)

Hansen, M [International Atomic Energy Agency, Division of Nuclear Power and Reactors, Nuclear Materials and Fuel Cycle Section, Vienna (Austria)

1976-07-01

Prior to the development of nuclear power, uranium ores were used to a very limited extent as a ceramic colouring agent, as a source of radium and in some places as a source of vanadium. Perhaps before that, because of the bright orange and yellow colours of its secondary ores, it was probably used as ceremonial paint by primitive man. After the discovery of nuclear fission a whole new industry emerged, complete with its problems of demand, resources and supply. Spurred by special incentives in the early years of this new nuclear industry, prospectors discovered over 20 000 occurrences of uranium in North America alone, and by 1959 total world production reached a peak of 34 000 tonnes uranium from mines in South Africa, Canada and United States. This rapid growth also led to new problems. As purchases for military purposes ended, government procurement contracts were not renewed, and the large reserves developed as a result of government purchase incentives, in combination with lack of substantial commercial market, resulted in an over-supply of uranium. Typically, an over-supply of uranium together with national stockpiling at low prices resulted in depression of prices to less than $5 per pound by 1971. Although forecasts made in the early 1970's increased confidence in the future of nuclear power, and consequently the demand for uranium, prices remained low until the end of 1973 when OPEC announced a very large increase in oil prices and quite naturally, prices for coal also rose substantially. The economics of nuclear fuel immediately improved and prices for uranium began to climb in 1974. But the world-wide impact of the OPEC decision also produced negative effects on the uranium industry. Uranium production costs rose dramatically, as did capital costs, and money for investment in new uranium ventures became more scarce and more expensive. However, the uranium supply picture today offers hope of satisfactory development in spite of the many problems to be

Trends in uranium supply

International Nuclear Information System (INIS)

Hansen, M.

1976-01-01

Prior to the development of nuclear power, uranium ores were used to a very limited extent as a ceramic colouring agent, as a source of radium and in some places as a source of vanadium. Perhaps before that, because of the bright orange and yellow colours of its secondary ores, it was probably used as ceremonial paint by primitive man. After the discovery of nuclear fission a whole new industry emerged, complete with its problems of demand, resources and supply. Spurred by special incentives in the early years of this new nuclear industry, prospectors discovered over 20 000 occurrences of uranium in North America alone, and by 1959 total world production reached a peak of 34 000 tonnes uranium from mines in South Africa, Canada and United States. This rapid growth also led to new problems. As purchases for military purposes ended, government procurement contracts were not renewed, and the large reserves developed as a result of government purchase incentives, in combination with lack of substantial commercial market, resulted in an over-supply of uranium. Typically, an over-supply of uranium together with national stockpiling at low prices resulted in depression of prices to less than $5 per pound by 1971. Although forecasts made in the early 1970's increased confidence in the future of nuclear power, and consequently the demand for uranium, prices remained low until the end of 1973 when OPEC announced a very large increase in oil prices and quite naturally, prices for coal also rose substantially. The economics of nuclear fuel immediately improved and prices for uranium began to climb in 1974. But the world-wide impact of the OPEC decision also produced negative effects on the uranium industry. Uranium production costs rose dramatically, as did capital costs, and money for investment in new uranium ventures became more scarce and more expensive. However, the uranium supply picture today offers hope of satisfactory development in spite of the many problems to be

Chapter 1. General information about uranium. 1.3. Uranium ores

International Nuclear Information System (INIS)

Khakimov, N.; Nazarov, Kh.M.; Mirsaidov, I.U.

2012-01-01

The uranium ores were described. It was found that uranium ores and natural mineral formations containing uranium and its compounds, can be found in concentrations that are technically possible for industrial utilization and which are economically profitable. It was defined that oxidation levels of uranium minerals have an impact on their reprocessing technology and behavior in hydrometallurgical re partition. It was found that the chemical composition of ores has a decisive importance during selection of their reprocessing method.

URANIUM DECONTAMINATION WITH RESPECT TO ZIRCONIUM

Science.gov (United States)

Vogler, S.; Beederman, M.

1961-05-01

A process is given for separating uranium values from a nitric acid aqueous solution containing uranyl values, zirconium values and tetravalent plutonium values. The process comprises contacting said solution with a substantially water-immiscible liquid organic solvent containing alkyl phosphate, separating an organic extract phase containing the uranium, zirconium, and tetravalent plutonium values from an aqueous raffinate, contacting said organic extract phase with an aqueous solution 2M to 7M in nitric acid and also containing an oxalate ion-containing substance, and separating a uranium- containing organic raffinate from aqueous zirconium- and plutonium-containing extract phase.

Containment and storage of uranium hexafluoride at US Department of Energy uranium enrichment plants

International Nuclear Information System (INIS)

Barlow, C.R.; Alderson, J.H.; Blue, S.C.; Boelens, R.A.; Conkel, M.E.; Dorning, R.E.; Ecklund, C.D.; Halicks, W.G.; Henson, H.M.; Newman, V.S.; Philpot, H.E.; Taylor, M.S.; Vournazos, J.P.; Pryor, W.A.; Ziehlke, K.T.

1992-07-01

Isotopically depleted UF 6 (uranium hexafluoride) accumulates at a rate five to ten times greater than the enriched product and is stored in steel vessels at the enrichment plant sites. There are approximately 55,000 large cylinders now in storage at Paducah, Kentucky; Portsmouth, Ohio; and Oak Ridge, Tennessee. Most of them contain a nominal 14 tons of depleted UF 6 . Some of these cylinders have been in the unprotected outdoor storage environment for periods approaching 40 years. Storage experience, supplemented by limited corrosion data, suggests a service life of about 70 years under optimum conditions for the 48-in. diameter, 5/16-in.-wall pressure vessels (100 psi working pressure), using a conservative industry-established 1/4-in.-wall thickness as the service limit. In the past few years, however, factors other than atmospheric corrosion have become apparent that adversely affect the serviceability of small numbers of the storage containers and that indicate the need for a managed program to ensure maintenance ofcontainment integrity for all the cylinders in storage. The program includes periodic visual inspections of cylinders and storage yards with documentation for comparison with other inspections, a group of corrosion test programs to permit cylinder life forecasts, and identification of (and scheduling for remedial action) situations in which defects, due to handling damage or accelerated corrosion, can seriously shorten the storage life or compromise the containment integrity of individual cylinders. The program also includes rupture testing to assess the effects of certain classes of damage on overall cylinder strength, aswell as ongoing reviews of specifications, procedures, practices, and inspection results to effect improvements in handling safety, containment integrity, and storage life

Uranium from phosphates to rabbit bones: Predicting dietary contribution to uranium deposition in animal bones

International Nuclear Information System (INIS)

Canella Avelar, A.; Motta Ferreira, W.; Menezes, M.

2014-01-01

Uranium is a hazardous element, both for radioactivity and metallotoxicity. Health implications of human overexposure to uranium are well documented: from reproduction impairment, liver and kidney diseases to some types of cancer. There are limited data in the modern literature concerning the levels of uranium in animal tissues and foods, as well the dietary daily intake of uranium is not fully known both for man and livestock. On the other hand, practically every phosphate and its products contain uranium in its structure. The average U content in agricultural phosphate may vary from 10 up to 390 ppm. In this particular feature, uranium can reach animal and man food chain by ingestion of feed and food grade phosphate containing U.

Adsorption equilibrium of uranium from seawater on chelating resin containing amide oxime group

International Nuclear Information System (INIS)

Hori, Takahiro; Saito, Kyoichi; Furusaki, Shintaro; Sugo, Takanobu; Okamoto, Jiro.

1987-01-01

Chelating resins containing amide oxime group were synthesized by radiation-induced graft polymerization. The amount of the amide oxime groups was controlled below about 0.1 mol per kg of base polymer. The adsorption equilibrium of uranium from seawater on this resin was investigated. It was suggested that two neighboring amide oxime groups on the grafted chain captured one uranyl ion, and that single amide oxime ligand had little capacity for the adsorption of uranium. The adsorption equilibrium was correlated by a Langmuir-type equation. The content of neighboring amide oxime groups was 0.406 x 10 -3 mol per kg of base polymer, which corresponded to 0.39 % of the total amount of amide oxime groups. The apparent stoichiometric stability constant for the complex of uranyl ion with the neighboring amide oxime groups in seawater was calculated to be 10 -21.7 . (author)

Influence of MgO containing strontium on the structure of ceramic film formed on grain oriented silicon steel surface

Directory of Open Access Journals (Sweden)

Daniela C. Leite Vasconcelos

1999-07-01

Full Text Available The oxide layer formed on the surface of a grain oriented silicon steel was characterized by SEM and EDS. 3% Si steel substrates were coated by two types of slurries: one formed by MgO and water and other formed by MgO, water and SrSO4. The ceramic films were evaluated by SEM, EDS and X-ray diffraction. Depth profiles of Fe, Si and Mg were obtained by GDS. The magnetic core losses (at 1.7 Tesla, 60 Hz of the coated steel samples were evaluated as well. The use of MgO containing strontium reduced the volume fraction of forsterite particles beneath the outermost ceramic layer. It was observed a reduced magnetic core loss with the use of the slurry with MgO containing strontium.

Recovering uranium from phosphates

Energy Technology Data Exchange (ETDEWEB)

Bergeret, M [Compagnie de Produits Chimiques et Electrometallurgiques Pechiney-Ugine Kuhlmann, 75 - Paris (France)

1981-06-01

Processes for the recovery of the uranium contained in phosphates have today become competitive with traditional methods of working uranium sources. These new possibilities will make it possible to meet more rapidly any increases in the demand for uranium: it takes ten years to start working a new uranium deposit, but only two years to build a recovery plant.

Criticality safety evaluation of a type B enriched uranium shipping container

International Nuclear Information System (INIS)

Hopper, C.M.

1978-01-01

The Oak Ridge Y-12 Plant Model DT-14 container was developed to replace and extend the enriched uranium shipping capabilities of the USA/5765/BF Vermiculite shipping container. This work was accomplished to comply with the DOE Immediate Action Directive Number 0529-02 for ''Phasing out the use of loose or bagged Vermiculite packaging material as a thermal shield and shock absorber in radioactive material packages''. The Model DT-14 is fabricated from a Specification 17H 30-gallon drum, cane fiberboard insulation, and a steel inner containment vessel (159 mm dia by 390 mm height). The single-package and array analyses are based upon results of the multigroup Monte Carlo criticality program, KENO, utilizing 16-energy-group Hansen-Roach, and Knight modified 238 U cross sections. The program and cross sections are considered well established on the basis of their success in calculating a large variety of critical experiments. Validation results show that a calculated neutron multiplication factor plus two standard deviations equal to 0.970 or greater must be considered critical, and all lower values may be considered safe

Ceramic component with reinforced protection against radiations

International Nuclear Information System (INIS)

Dubuisson, J.; Laville, H.; Le Gal, P.

1986-01-01

Ceramic components hardened against radiations are claimed (for example capacitors or ceramic substrates for semiconductors). They are prepared with a sintered ceramic containing a high proportion of heavy atoms (for instance barium titanate and a bismuth salt) provided with a glass layer containing a high proportion of light atoms. The two materials are joined by vitrification producing a diffusion zone at the interface [fr

Uranium purchases report 1993

International Nuclear Information System (INIS)

1994-01-01

Data reported by domestic nuclear utility companies in their responses to the 1991 through 1993 ''Uranium Industry Annual Survey,'' Form EIA-858, Schedule B,'' Uranium Marketing Activities,'' are provided in response to the requirements in the Energy Policy Act 1992. Appendix A contains an explanation of Form EIA-858 survey methodologies with emphasis on the processing of Schedule B data. Additional information published in this report not included in Uranium Purchases Report 1992, includes a new data table. Presented in Table 1 are US utility purchases of uranium and enrichment services by origin country. Also, this report contains additional purchase information covering average price and contract duration. Table 2 is an update of Table 1 and Table 3 is an update of Table 2 from the previous year's report. The report contains a glossary of terms

Uranium dioxide pellets

International Nuclear Information System (INIS)

Zawidzki, T.W.

1982-01-01

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

Corrosion of Ceramic Materials

Science.gov (United States)

Opila, Elizabeth J.; Jacobson, Nathan S.

1999-01-01

Non-oxide ceramics are promising materials for a range of high temperature applications. Selected current and future applications are listed. In all such applications, the ceramics are exposed to high temperature gases. Therefore it is critical to understand the response of these materials to their environment. The variables to be considered here include both the type of ceramic and the environment to which it is exposed. Non-oxide ceramics include borides, nitrides, and carbides. Most high temperature corrosion environments contain oxygen and hence the emphasis of this chapter will be on oxidation processes.

Uranium deposits in granitic rocks

International Nuclear Information System (INIS)

Nishimori, R.K.; Ragland, P.C.; Rogers, J.J.W.; Greenberg, J.K.

1977-01-01

This report is a review of published data bearing on the geology and origin of uranium deposits in granitic, pegmatitic and migmatitic rocks with the aim of assisting in the development of predictive criteria for the search for similar deposits in the U.S. Efforts were concentrated on the so-called ''porphyry'' uranium deposits. Two types of uranium deposits are primarily considered: deposits in pegmatites and alaskites in gneiss terrains, and disseminations of uranium in high-level granites. In Chapter 1 of this report, the general data on the distribution of uranium in igneous and metamorphic rocks are reviewed. Chapter 2 contains some comments on the classification of uranium deposits associated with igneous rocks and a summary of the main features of the geology of uranium deposits in granites. General concepts of the behavior of uranium in granites during crustal evolution are reviewed in Chapter 3. Also included is a discussion of the relationship of uranium mineralization in granites to the general evolution of mobile belts, plus the influence of magmatic and post-magmatic processes on the distribution of uranium in igneous rocks and related ore deposits. Chapter 4 relates the results of experimental studies on the crystallization of granites to some of the geologic features of uranium deposits in pegmatites and alaskites in high-grade metamorphic terrains. Potential or favorable areas for igneous uranium deposits in the U.S.A. are delineated in Chapter 5. Data on the geology of specific uranium deposits in granitic rocks are contained in Appendix 1. A compilation of igneous rock formations containing greater than 10 ppM uranium is included in Appendix 2. Appendix 3 is a report on the results of a visit to the Roessing area. Appendix 4 is a report on a field excursion to eastern Canada

On the use of dental ceramics as a possible second-line approach to accident irradiation dosimetry

International Nuclear Information System (INIS)

Davies, J.E.

1979-01-01

Recent development in dental ceramic production has resulted in natural or depleted uranium, used for over half a century to mimic the fluorescence of natural teeth, being substituted in such ceramics by non-radioactive fluorescent materials. This creates the possibility of using dental ceramics incorporating the latter as second-line dosimeters in cases of accidental irradiation. This pilot study shows the feasbility of such an approach using both thermally stimulated exoelectron and thermoluminescent techniques. In conclusion, it is considered that it would be of interest to continue this investigation of dental ceramic materials as second-line accident dosimeters

Crystallization kinetics and spectroscopic investigations on Tb3+ and Yb3+ codoped glass ceramics containing CaF2 nanocrystals

International Nuclear Information System (INIS)

Huang Lihui; Qin Guanshi; Arai, Yusuke; Jose, Rajan; Suzuki, Takenobu; Ohishi, Yasutake; Yamashita, Tatsuya; Akimoto, Yusuke

2007-01-01

Transparent Tb 3+ and Yb 3+ codoped oxyfluoride glass ceramics containing CaF 2 nanocrystals were prepared by melt quenching and subsequent heat treatment. Crystallization kinetics of CaF 2 nanocrystals was investigated by differential scanning calorimetric method. The average apparent activation energy E a of the crystallization was ∼498 kJ/mol. Moreover, the value of the Avrami exponent n was 1.01. These results suggest that the crystallization mechanism of CaF 2 is a diffusion controlled growth process of needles and plates of finite long dimensions. X-ray diffraction patterns and transmission electron microscopy image confirmed the CaF 2 nanocrystals in the glass ceramic. Ultraviolet (UV) and visible emission spectra of the as-made glass and the glass ceramic with an excitation of a 974 nm laser diode were recorded at room temperature. An intense UV emission at 381 nm was observed in the glass ceramic. The origin of the enhancement of the emission at 381 nm was investigated using spectroscopic technique and Judd-Ofelt analysis. The enhancement of the emission at 381 nm could be attributed to the change of the ligand field of Tb 3+ ions due to the incorporation of some Tb 3+ and Yb 3+ ions into CaF 2 nanocrystals in the glass ceramic

Method for preparing corrosion-resistant ceramic shapes

Science.gov (United States)

Arons, R.M.; Dusek, J.T.

1979-12-07

Ceramic shapes having impermeable tungsten coatings can be used for containing highly corrosive molten alloys and salts. The shapes are prepared by coating damp green ceramic shapes containing a small amount of yttria with a tungsten coating slip which has been adjusted to match the shrinkage rate of the green ceramic and which will fire to a theoretical density of at least 80% to provide an impermeable coating.

URANIUM BISMUTHIDE DISPERSION IN MOLTEN METAL

Science.gov (United States)

Teitel, R.J.

1959-10-27

The formation of intermetallic bismuth compounds of thorium or uranium dispersed in a liquid media containing bismuth and lead is described. A bismuthide of uranium dispersed in a liquid metal medium is formed by dissolving uranium in composition of lead and bismuth containing less than 80% lead and lowering the temperature of the composition to a temperature below the point at which the solubility of uranium is exceeded and above the melting point of the composition.

Automated uranium titration system

International Nuclear Information System (INIS)

Takahashi, M.; Kato, Y.

1983-01-01

An automated titration system based on the Davies-Gray method has been developed for accurate determination of uranium. The system consists of a potentiometric titrator with precise burettes, a sample changer, an electronic balance and a desk-top computer with a printer. Fifty-five titration vessels are loaded in the sample changer. The first three contain the standard solution for standardizing potassium dichromate titrant, and the next two and the last two contain the control samples for data quality assurance. The other forty-eight measurements are carried out for sixteen unknown samples. Sample solution containing about 100 mg uranium is taken in a titration vessel. At the pretreatment position, uranium (VI) is reduced to uranium (IV) by iron (II). After the valency adjustment, the vessel is transferred to the titration position. The rate of titrant addition is automatically controlled to be slower near the end-point. The last figure (0.01 mL) of the equivalent titrant volume for uranium is calculated from the potential change. The results obtained with this system on 100 mg uranium gave a precision of 0.2% (RSD,n=3) and an accuracy of better than 0.1%. Fifty-five titrations are accomplished in 10 hours. (author)

Uranium tailings bibliography

International Nuclear Information System (INIS)

Holoway, C.F.; Goldsmith, W.A.; Eldridge, V.M.

1975-12-01

A bibliography containing 1,212 references is presented with its focus on the general problem of reducing human exposure to the radionuclides contained in the tailings from the milling of uranium ore. The references are divided into seven broad categories: uranium tailings pile (problems and perspectives), standards and philosophy, etiology of radiation effects, internal dosimetry and metabolism, environmental transport, background sources of tailings radionuclides, and large-area decontamination

Polyphase ceramic and glass-ceramic forms for immobilizing ICPP high-level nuclear waste

International Nuclear Information System (INIS)

Harker, A.B.; Flintoff, J.F.

1984-01-01

Polyphase ceramic and glass-ceramic forms have been consolidated from simulated Idaho Chemical Processing Plant wastes by hot isostatic pressing calcined waste and chemical additives by 1000 0 C or less. The ceramic forms can contain over 70 wt% waste with densities ranging from 3.5 to 3.85 g/cm 3 , depending upon the formulation. Major phases are CaF 2 , CaZrTi 207 , CaTiO 3 , monoclinic ZrO 2 , and amorphous intergranular material. The relative fraction of the phases is a function of the chemical additives (TiO 2 , CaO, and SiO 2 ) and consolidation temperature. Zirconolite, the major actinide host, makes the ceramic forms extremely leach resistant for the actinide simulant U 238 . The amorphous phase controls the leach performance for Sr and Cs which is improved by the addition of SiO 2 . Glass-ceramic forms were also consolidated by HIP at waste loadings of 30 to 70 wt% with densities of 2.73 to 3.1 g/cm 3 using Exxon 127 borosilicate glass frit. The glass-ceramic forms contain crystalline CaF 2 , Al 203 , and ZrSi 04 (zircon) in a glass matrix. Natural mineral zircon is a stable host for 4+ valent actinides. 17 references, 3 figures, 5 tables

Health and environmental problems of using antiarmour munitions containing depleted uranium core

International Nuclear Information System (INIS)

Matousek, J.

2006-01-01

In the 1970s, core of depleted uranium commenced to be introduced into the breakthrough antitank munitions of various calibers and types in order to considerably enhance their effectiveness due to extremely high density in comparison with steel. The health and environmental threats of using this munitions and other weaponry where depleted uranium has been utilised as counterbalance stem from the pyrophoric character of uranium, burnt due to material deformation and friction when penetrating armour targets creating thus highly respirable aerosol of uranium oxides that are deposited in alveoli after being inhaled or in other tissues after being ingested. Composition and main properties of depleted uranium are presented. Chronic effects of deposited particles of uranium oxides are due to internal irradiation of sensitive organs at proceeding radioactive decay accompanied with alpha irradiation. Long-term internal irradiation by radionuclides producing alpha-rays leads to proved risk of increased incidence of carcinoma and leukaemia not to speak on chronic chemical toxicity of uranium, independent of its isotopic composition. Environmental impact of extensive use of munitions with depleted uranium in the recent armed conflicts is assessed. (authors)

Uranium hexafluoride: Handling procedures and container descriptions

International Nuclear Information System (INIS)

1987-09-01

The US Department of Energy (DOE) guidelines for packaging, measuring, and transferring uranium hexafluoride (UF 6 ) have been undergoing continual review and revision for several years to keep them in phase with developing agreements for the supply of enriched uranium. Initially, K-1323 ''A Brief Guide to UF 6 Handling,'' was issued in 1957. This was superceded by ORO-651, first issued in 1966, and reissued in 1967 to make editorial changes and to provide minor revisions in procedural information. In 1968 and 1972, Revisions 2 and 3, respectively, were issued as part of the continuing effort to present updated information. Revision 4 issued in 1977 included revisions to UF 6 cylinders, valves, and methods to use. Revision 5 adds information dealing with pigtails, overfilled cylinders, definitions and handling precautions, and cylinder heel reduction procedures. Weighing standards previously presented in ORO-671, Vol. 1 (Procedures for Handling and Analysis of UF 6 ) have also been included. This revision, therefore, supercedes ORO-671-1 as well as all prior issues of this report. These guidelines will normally apply in all transactions involving receipt or shipment of UF 6 by DOE, unless stipulated otherwise by contracts or agreements with DOE or by notices published in the Federal Register. Any questions or requests for additional information on the subject matter covered herein should be directed to the United States Department of Energy, P.O. Box E, Oak Ridge, Tennessee 37831, Attention: Director, Uranium Enrichment Operations Division. 33 figs., 12 tabs

Method for the chemical reprocessing of irradiated nuclear fuels, in particular nuclear fuels containing uranium

International Nuclear Information System (INIS)

Koch, G.

1976-01-01

In the chemical processing of irradiated uranium-containing nuclear fuels which are hydrolyzed with aqueous nitric acid, a suggestion is made to use as quaternary ammonium nitrate trialkyl-methyl ammonium nitrates as extracting agent, in which the sum of C atoms is greater than 16. In the illustrated examples, tricaprylmethylammonium nitrate, trilaurylmethylammonium nitrate and tridecylmethylammonium nitrate are named. (HPH/LH) [de

Determination of size and shape distributions of metal and ceramic powders

International Nuclear Information System (INIS)

Jovanovic, DI.

1961-01-01

For testing the size and shape distributions of metal and ceramic uranium oxide powders the following method for analysing the grain size of powders were developed and implemented: microscopic analysis and sedimentation method. A gravimetry absorption device was constructed for determining the specific surfaces of powders

Biomineral processing of high apatite containing low-grade indian uranium ore

International Nuclear Information System (INIS)

Abhilash; Mehta, K.D.; Pandey, B.D.; Ray, L.; Tamrakar, P.K.

2010-01-01

Microbial species isolated from source mine water, primarily an enriched culture of Acidithiobacillus ferrooxidans was employed for bio-leaching of uranium from a low-grade apatite rich uranium ore of Narwapahar Mines, India while varying pH, pulp density (PD), particle size, etc. The ore (0.047% U_3O_8), though of Singhbhum area (richest deposit of uranium ores in India), due to presence of some refractory minerals and high apatite (5%) causes a maximum 78% recovery through conventional processing. Bioleaching experiments were carried out by varying pH at 35"oC using 20%(w/v) PD and <76μm size particles resulting in 83.5% and 78% uranium bio-recovery at 1.7 and 2.0 pH in 40 days as against maximum recovery of 46% and 41% metal in control experiments respectively. Finer size (<45μm) ore fractions exhibited higher uranium dissolution (96%) in 40 days at 10% (w/v) pulp density (PD), 1.7 pH and 35"oC. On increasing the pulp density from 10% to 20% under the same conditions, the biorecovery of uranium fell down from 96% to 82%. The higher uranium dissolution during bioleaching at 1.7 pH with the fine size particles (<45μm) can be correlated with increase in redox potential from 598 mV to 708 mV and the corresponding variation of Fe(III) ion concentration in 40 days. (author)

Biomineral processing of high apatite containing low-grade indian uranium ore

Energy Technology Data Exchange (ETDEWEB)

Abhilash; Mehta, K.D.; Pandey, B.D., E-mail: biometnml@gmail.com [National Metallurgical Laboratory (CSIR), Jamshedpur (India); Ray, L. [Jadavpur Univ., FTBE Dept., Kolkata (India); Tamrakar, P.K. [Uranium Corp. of India Limited, CR& D Dept., Jaduguda (India)

2010-07-01

Microbial species isolated from source mine water, primarily an enriched culture of Acidithiobacillus ferrooxidans was employed for bio-leaching of uranium from a low-grade apatite rich uranium ore of Narwapahar Mines, India while varying pH, pulp density (PD), particle size, etc. The ore (0.047% U{sub 3}O{sub 8}), though of Singhbhum area (richest deposit of uranium ores in India), due to presence of some refractory minerals and high apatite (5%) causes a maximum 78% recovery through conventional processing. Bioleaching experiments were carried out by varying pH at 35{sup o}C using 20%(w/v) PD and <76μm size particles resulting in 83.5% and 78% uranium bio-recovery at 1.7 and 2.0 pH in 40 days as against maximum recovery of 46% and 41% metal in control experiments respectively. Finer size (<45μm) ore fractions exhibited higher uranium dissolution (96%) in 40 days at 10% (w/v) pulp density (PD), 1.7 pH and 35{sup o}C. On increasing the pulp density from 10% to 20% under the same conditions, the biorecovery of uranium fell down from 96% to 82%. The higher uranium dissolution during bioleaching at 1.7 pH with the fine size particles (<45μm) can be correlated with increase in redox potential from 598 mV to 708 mV and the corresponding variation of Fe(III) ion concentration in 40 days. (author)

A study on the recovery of TRU elements by a container-aided solid cathode

International Nuclear Information System (INIS)

Kwon, S.W.; Lee, J.H.; Woo, M.S.; Shim, J.B.; Kim, E.H.; Yoo, J.H.; Park, S.W.; Park, H.S.

2005-01-01

Pyroprocessing is a very prominent way for the recovery of the long-lived elements from the spent nuclear fuel. Electrorefining is a key technology of pyroprocessing and generally composed of two recovery steps - deposit of uranium onto a solid cathode and the recovery of TRU (TRansUranic) elements by a liquid cadmium cathode. The liquid cadmium cathode has some problems such as a cadmium volatilization problem, a low separation factor, and a complicates structure. In this study, CASC (Container-Aided Solid Cathode) was proposed as a candidate for replacing a liquid cadmium cathode and the deposition behavior of the cathode was examined during the electrorefining experiments. The CASC is a solid cathode surrounded with a porous ceramic container, where the container is used to capture the dripped deposit from the cathode. In the electrorefining experiment, the uranium used as a surrogate for the TRU elements, was effectively separated from cerium. The anode material and surface area were also investigated during electrolysis experiments for the more efficient electrorefining system. From the results of this study, it is concluded that the container-aided solid cathode can be a potential candidate for replacing a liquid cadmium cathode and the cathode should be developed further for the better electrolysis operation. (author)

Preparation of 147Pm ceramic source core

International Nuclear Information System (INIS)

Mielcarski, M.

1989-01-01

Preparation of ceramic pellets containing fixed promethium-147 is described. Incorporation rate of 147 Pm into the ceramic material was determined. The leachability and vaporization of promethium from the obtained ceramics was investigated. The ceramic pellets prepared by the described procedure, mounted in special holders, can be applied as point sources in beta backscatter thickness gauges. (author)

A mixed-valent uranium phosphonate framework containing U{sup IV}, U{sup V}, and U{sup VI}

Energy Technology Data Exchange (ETDEWEB)

Chen, Lanhua; Zheng, Tao; Wang, Yaxing; Diwu, Juan; Chai, Zhifang; Wang, Shuao [School for Radiological and interdisciplinary Sciences (RAD-X), Soochow University, Suzhou (China); Collaborative Innovation Center of Radiation Medicine, Jiangsu Higher Education Institutions, Suzhou (China); Bao, Songsong; Zheng, Limin [State Key Laboratory of Coordination Chemistry, School of Chemistry and Chemical Engineering, Collaborative Innovation Center of Advanced Microstructures, Nanjing University (China); Zhang, Linjuan; Wang, Jianqiang [Shanghai Institute of Applied Physics and, Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai (China); Liu, Hsin-Kuan [Department of Chemistry, National Central University, Jhongli (China); Albrecht-Schmitt, Thomas E. [Department of Chemistry and Biochemistry, Florida State University, Tallahassee, FL (United States)

2016-08-16

It is shown that U{sup V}O{sub 2}{sup +} ions can reside at U{sup VI}O{sub 2}{sup 2+} lattice sites during mild reduction and crystallization process under solvothermal conditions, yielding a complicated and rare mixed-valent uranium phosphonate compound that simultaneously contains U{sup IV}, U{sup V}, and U{sup VI}. The presence of uranium with three oxidation states was confirmed by various characterization techniques, including X-ray crystallography, X-ray photoelectron, electron paramagnetic resonance, FTIR, UV/Vis-NIR absorption, and synchrotron radiation X-ray absorption spectroscopy, and magnetism measurements. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Uranium hexafluoride: handling procedures and container criteria

International Nuclear Information System (INIS)

1977-04-01

The U.S. Energy Research and Development Administration's (ERDA) procedures for packaging, measuring, and transferring uranium hexafluoride (UF 6 ) have been undergoing continual review and revision for several years to keep them in phase with developing agreements for the supply of enriched uranium. This report, first issued in 1966, was reissued in 1967 to make editorial changes and to provide for minor revisions in procedural information. In 1968 and 1972, Revisions 2 and 3, respectively, were issued as part of the continuing effort to present updated information. This document, Revision 4, includes primarily revisions to UF 6 cylinders, valves, and methods of use. This revision supersedes all previous issues of this report. The procedures will normally apply in all transactions involving receipt or shipment of UF 6 by ERDA, unless stipulated otherwise by contracts or agreements with ERDA or by notices published in the Federal Register

Structural ceramics containing electric arc furnace dust.

Science.gov (United States)

Stathopoulos, V N; Papandreou, A; Kanellopoulou, D; Stournaras, C J

2013-11-15

In the present work the stabilization of electric arc furnace dust EAFD waste in structural clay ceramics was investigated. EAFD was collected over eleven production days. The collected waste was characterized for its chemical composition by Flame Atomic Absorption Spectroscopy. By powder XRD the crystal structure was studied while the fineness of the material was determined by a laser particle size analyzer. The environmental characterization was carried out by testing the dust according to EN12457 standard. Zn, Pb and Cd were leaching from the sample in significant amounts. The objective of this study is to investigate the stabilization properties of EAFD/clay ceramic structures and the potential of EAFD utilization into structural ceramics production (blocks). Mixtures of clay with 2.5% and 5% EAFD content were studied by TG/DTA, XRD, SEM, EN12457 standard leaching and mechanical properties as a function of firing temperature at 850, 900 and 950 °C. All laboratory facilities maintained 20 ± 1 °C. Consequently, a pilot-scale experiment was conducted with an addition of 2.5% and 5% EAFD to the extrusion mixture for the production of blocks. During blocks manufacturing, the firing step reached 950 °C in a tunnel kiln. Laboratory heating/cooling gradients were similar to pilot scale production firing. The as produced blocks were then subjected to quality control tests, i.e. dimensions according to EN772-17, water absorbance according to EN772-6, and compressive strength according to EN772-1 standard, in laboratory facilities certified under EN17025. The data obtained showed that the incorporation of EAFD resulted in an increase of mechanical strength. Moreover, leaching tests performed according to the Europeans standards on the EAFD-block samples showed that the quantities of heavy metals leached from crushed blocks were within the regulatory limits. Thus the EAFD-blocks can be regarded as material of no environmental concern. Copyright © 2013 Elsevier B

Controllable synthesis and tunable luminescence of glass ceramic containing Mn2+:ZnAl2O4 and Pr3+:YF3 nano-crystals

International Nuclear Information System (INIS)

Yu, Yunlong; Li, Xiaoyan

2016-01-01

Highlights: • Glass ceramic containing ZnAl 2 O 4 and YF 3 nano-crystals is fabricated. • Mn 2+ and Pr 3+ are selectively incorporated into ZnAl 2 O 4 and YF 3 , respectively. • The luminescence color can be tuned by adjusting the excitation wavelength. - Abstract: Glass ceramic containing spinel ZnAl 2 O 4 :Mn 2+ and orthorhombic YF 3 :Pr 3+ nano-crystals has been successfully prepared by a melt-quenching technique. X-ray diffraction and transmission electron microscopy demonstrated that two nano-phases, i.e. ZnAl 2 O 4 and YF 3 , were homogeneously distributed among the glass matrix. Importantly, the selective incorporation of Pr 3+ ions into the Y 3+ nine-fold coordinated sites of YF 3 and the segregation of Mn 2+ dopants in the Zn 2+ tetrahedral sites of ZnAl 2 O 4 were confirmed based on the excitation/emission spectra and the crystal field calculation. Under blue light excitation, both Pr 3+ and Mn 2+ in the glass ceramic can be simultaneously excited, and emit red and green luminescence, respectively, owing to the suppression of energy transfer between them. The luminescence color of the obtained glass ceramic can be easily tuned by adjusting the excitation wavelength. These results indicate the potential application of the glass ceramic as converting phosphor to generate white-light after coupling with the blue LED chip.

Uranium oxide recovering method

International Nuclear Information System (INIS)

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

1997-01-01

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

1967-06-15

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

Neutron activation analysis of rare earths in uranium containing rocks

International Nuclear Information System (INIS)

May, S.; Pinte, G.

1984-01-01

The determination of rare earths by activation analysis in uranium rocks is disturbed either by fission-produced rare earths, or by neptunium-239 originating from uranium-238. In order to eliminate these interferencies, the chemical separation of rare earths from uranium prior to activation should be performed. The chemical process is as follows: the rock sample is fused with sodium borate, then, after addition of hydrochloric acid, the resulting solution is passed through a Dowex 1x8 column. Uranium is retained on the resin, and rare earths and scandium are eluted. Aluminium is added as a carrier to the solution, and rare earths and scandium are coprecipitated with aluminium hydroxide. This precipitate is irradiated in the nuclear reactor. Gamma spectrometry is used for the determination of earth radionuclide. Activity measurements are performed in successive steps during one month. The following elements are determined: Pr, La, Sm, Nd, Yb, Lu, Ce, Tb, Eu and Sc. The chemical yield is measured by using scandium as an internal standard. (author)

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

III Advanced Ceramics and Applications Conference

CERN Document Server

Gadow, Rainer; Mitic, Vojislav; Obradovic, Nina

2016-01-01

This is the Proceedings of III Advanced Ceramics and Applications conference, held in Belgrade, Serbia in 2014. It contains 25 papers on various subjects regarding preparation, characterization and application of advanced ceramic materials.

Corrosion tests with uranium- and plutonium-loaded ceramic waste forms

International Nuclear Information System (INIS)

Morss, L. R.; Johnson, S. G.; Ebert, W. L.; DiSanto, T.; Frank, S. M.; Holly, J. L.; Kropf, A. J.; Mertz, C. J.; O'Holleran, T. P.; Richmann, M. K.; Sinkler, W.; Tsai, Y.; Warren, A. R.; Noy, M.

2003-01-01

Tests were conducted with ceramic waste form (CWF) materials that contained small amounts of uranium and plutonium to study their release behavior as the CWF corroded. Materials made using the hot isostatic press (HIP) and pressureless consolidation (PC) methods were examined and tested. Four different materials were made using the HIP method with two salts having different U:Pu mole ratios and two zeolite reagents having different residual water contents. Tests with the four HIP U,Pu-loaded CWF materials were conducted at 90 and 120 C, at CWF-to-water mass ratios of 1:10 and 1:20, and for durations between 7 and 365 days. Materials made using two PC processing conditions were also tested. Tests with the two PC U,Pu-loaded CWF materials were conducted at 90 and 120 C, at a CWF-to-water mass ratio of 1:10, and for durations between 7 and 182 days. The releases of matrix elements, U, and Pu in tests conducted under different test conditions and with different materials are compared to evaluate the effects of composition and processing conditions on the release behavior of U and Pu and the chemical durabilities of the different materials. The distributions of released elements among the fractions that were dissolved, in colloidal form in the solution, and fixed to test vessel walls were measured and compared. Characterization of Pu-bearing colloidal particles recovered from the test solutions using solids analysis techniques are also reported. The principal findings from this study are: (1) The release of U and Pu is about 10X less than the release of Si and 50X less than the release of B under all test conditions. This implies that U and Pu are in a phase that is less soluble than the sodalite and binder glass matrix. (2) Almost all of the plutonium that is released from U,Pu-loaded CWF is present either as colloidal-sized particles in the size range between 5 and 100 nm in the test solution (about 15% of the total) or becomes fixed on stainless steel test vessel

Development of a pharmaceutical form containing calixarene molecules for the treatment of intact or injured skin contaminated by uranium

International Nuclear Information System (INIS)

Spagnul, A.

2009-11-01

The first objective of this research thesis was to develop a formulation containing a tricarboxylic calixarene for cutaneous application for the local treatment of skin contamination by uranium. A second objective is to assess the efficiency of a calixarene nano-emulsion for such a treatment. In a first part, the author proposes an overview of risks associated with skin contamination by uranium, and of current treatments and treatments under development. In the second part, the author presents the oil-in-water-type nano-emulsion, reports an in vitro assessment of the decontamination efficiency of the calixarene nano-emulsion, reports an in vivo assessment of this efficiency (on pig ear skin explants contaminated by uranium), and presents the main publications and a patent request related to this research work

Oxidation studies of β-sialon ceramics containing amorphous and / or crystalline intergranular phases

International Nuclear Information System (INIS)

Persson, J.; Kall, P.O.; Jansson, K.; Nygren, M.

1992-01-01

β-sialon ceramics of equal overall compositions but containing amorphous, partly crystalline and almost completely crystalline intergranular phase(s) have been oxidized in oxygen at 1350 deg C for 20 hours. The obtained weight gain curves do not follow the parabolic rate law (ΔW/A 0 ) 2 = k p t + β. To the extent that crystallization occurs in the oxide scale during the oxidation experiment, the amorphous cross section area through which oxygen most easily diffuses will decrease with time. A brief description of this new rate law is given, and the obtained oxidation curves will be discussed within that framework. 4 refs., 2 tabs., 2 figs

Uranium resource technology, Seminar 3, 1980

International Nuclear Information System (INIS)

Morse, J.G.

1980-01-01

This conference proceedings contains 20 papers and 1 panel discussion on uranium mining and ore treatment, taking into account the environmental issues surrounding uranium supply. Topics discussed include: the US uranium resource base, the technology and economics of uranium recovery from phosphate resources, trends in preleash materials handling of sandstone uranium ores, groundwater restoration after in-situ uranium leaching, mitigation of the environmental impacts of open pit and underground uranium mining, remedial actions at inactive uranium mill tailings sites, environmental laws governing in-situ solution mining of uranium, and the economics of in-situ solution mining. 16 papers are indexed separately

Review of glass ceramic waste forms

International Nuclear Information System (INIS)

Rusin, J.M.

1981-01-01

Glass ceramics are being considered for the immobilization of nuclear wastes to obtain a waste form with improved properties relative to glasses. Improved impact resistance, decreased thermal expansion, and increased leach resistance are possible. In addition to improved properties, the spontaneous devitrification exhibited in some waste-containing glasses can be avoided by the controlled crystallization after melting in the glass-ceramic process. The majority of the glass-ceramic development for nuclear wastes has been conducted at the Hahn-Meitner Institute (HMI) in Germany. Two of their products, a celsian-based (BaAl 3 Si 2 O 8 ) and a fresnoite-based (Ba 2 TiSi 2 O 8 ) glass ceramic, have been studied at Pacific Northwest Laboratory (PNL). A basalt-based glass ceramic primarily containing diopsidic augite (CaMgSi 2 O 6 ) has been developed at PNL. This glass ceramic is of interest since it would be in near equilibrium with a basalt repository. Studies at the Power Reactor and Nuclear Fuel Development Corporation (PNC) in Japan have favored a glass-ceramic product based upon diopside (CaMgSi 2 O 6 ). Compositions, processing conditions, and product characterization of typical commercial and nuclear waste glass ceramics are discussed. In general, glass-ceramic waste forms can offer improved strength and decreased thermal expansion. Due to typcially large residual glass phases of up to 50%, there may be little improvement in leach resistance

The preparation of UO2 ceramic microspheres with an advanced process (TGU)

International Nuclear Information System (INIS)

Xu Zhichang; Tang Yaping; Zhang Fuhong

1994-04-01

The UO 2 ceramic microspheres are the most important materials in the spherical fuel elements for high temperature reactor (HTR). An advanced process for preparation of UO 2 ceramic microspheres has been developed at Institute of Nuclear Energy Technology, Tsinghua University. This process known as total gelation process of uranium (TGU), is based on the traditional sol-gel process, external gelation process and internal gelation process of uranium (EGU and IGU), and has been selected as the production process. The result of batch test is described. Accordance with the requirements of quality control (QC) and quality assurance (QA), the stabilization of operating parameters and product quality is tested., The results on batch test have shown that as well as all of the operated parameters are fixed, then the product quality can be stable as well as the product specification can be met. When the colloidal flow rate and the vibration frequency of nozzle are fixed, the kernel's size is also fixed. When the sintering temperature and time are fixed, the product density is also fixed. When the hydrogen atmosphere is used, the O/U ratio is very near to stoichiometry. The performance and structure of UO 2 ceramic microspheres are also given

Leaching behavior of glass ceramic nuclear waste forms

International Nuclear Information System (INIS)

Lokken, R.O.

1981-11-01

Glass ceramic waste forms have been investigated as alternatives to borosilicate glasses for the immobilization of high-level radioactive waste at Pacific Northwest Laboratory (PNL). Three glass ceramic systems were investigated, including basalt, celsian, and fresnoite, each containing 20 wt % simulated high-level waste calcine. Static leach tests were performed on seven glass ceramic materials and one parent glass (before recrystallization). Samples were leached at 90 0 C for 3 to 28 days in deionized water and silicate water. The results, expressed in normalized elemental mass loss, (g/m 2 ), show comparable releases from celsian and fresnoite glass ceramics. Basalt glass ceramics demonstrated the lowest normalized elemental losses with a nominal release less than 2 g/m 2 when leached in polypropylene containers. The releases from basalt glass ceramics when leached in silicate water were nearly identical with those in deionized water. The overall leachability of celsian and fresnoite glass ceramics was improved when silicate water was used as the leachant

Depleted uranium management alternatives

Energy Technology Data Exchange (ETDEWEB)

Hertzler, T.J.; Nishimoto, D.D.

1994-08-01

This report evaluates two management alternatives for Department of Energy depleted uranium: continued storage as uranium hexafluoride, and conversion to uranium metal and fabrication to shielding for spent nuclear fuel containers. The results will be used to compare the costs with other alternatives, such as disposal. Cost estimates for the continued storage alternative are based on a life-cycle of 27 years through the year 2020. Cost estimates for the recycle alternative are based on existing conversion process costs and Capital costs for fabricating the containers. Additionally, the recycle alternative accounts for costs associated with intermediate product resale and secondary waste disposal for materials generated during the conversion process.

Depleted uranium management alternatives

International Nuclear Information System (INIS)

Hertzler, T.J.; Nishimoto, D.D.

1994-08-01

This report evaluates two management alternatives for Department of Energy depleted uranium: continued storage as uranium hexafluoride, and conversion to uranium metal and fabrication to shielding for spent nuclear fuel containers. The results will be used to compare the costs with other alternatives, such as disposal. Cost estimates for the continued storage alternative are based on a life-cycle of 27 years through the year 2020. Cost estimates for the recycle alternative are based on existing conversion process costs and Capital costs for fabricating the containers. Additionally, the recycle alternative accounts for costs associated with intermediate product resale and secondary waste disposal for materials generated during the conversion process

Process and device for forming imprints on ceramic tubes

International Nuclear Information System (INIS)

1985-01-01

The purpose of the present invention is a process and a device for making imprints on ceramic tubes and these ceramic tubes with imprints. It is known that in uranium enrichment processes by gaseous diffusion, microporous tubes are used to made the diffuser units used for the application of this isotope enrichment process. It is known that these microporous tubes are generally made in two stages. In a first stage, a macroporous ceramic tube called a ''support'' is made. In a second stage, an internal microporous deposit is made which makes it possible to obtain a tube called a ''barrier'' finally having the required porosity to apply the gaseous diffusion enrichment process. The present invention involves the first stage of the manufacturing process of the barriers and, more precisely, a step in the manufacturing process of the supports that makes it possible to improve the efficiency of these barriers

Uranium prospecting; La prospection de l'uranium

Energy Technology Data Exchange (ETDEWEB)

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

1955-07-01

This report is an instruction book for uranium prospecting. It appeals to private prospecting. As prospecting is now a scientific and technical research, it cannot be done without preliminary studies. First of all, general prospecting methods are given with a recall of fundamental geologic data and some general principles which are common with all type of prospecting. The peculiarities of uranium prospecting are also presented and in particular the radioactivity property of uranium as well as the special aspect of uranium ores and the aspect of neighbouring ores. In a third part, a description of the different uranium ores is given and separated in two different categories: primary and secondary ores, according to the place of transformation, deep or near the crust surface respectively. In the first category, the primary ores include pitchblende, thorianite and rare uranium oxides as euxenite and fergusonite for example. In the second category, the secondary ores contain autunite and chalcolite for example. An exhaustive presentation of the geiger-Mueller counter is given with the presentation of its different components, its functioning and utilization and its maintenance. The radioactivity interpretation method is showed as well as the elaboration of a topographic map of the measured radioactivity. A brief presentation of other detection methods than geiger-Mueller counters is given: the measurement of fluorescence and a chemical test using the fluorescence properties of uranium salts. Finally, the main characteristics of uranium deposits are discussed. (M.P.)

Method for the recovery of uranium values from uranium tetrafluoride

International Nuclear Information System (INIS)

Kreuzmann, A.B.

1984-01-01

The invention comprises reacting particulate uranium tetrafluoride and alkaline earth metal oxide (e.g. CaO, MgO) in the presence of gaseous oxygen to effect formation of the corresponding alkaline earth metal uranate and alkaline earth metal fluoride. The product uranate is highly soluble in various acidic solutions whereas the product fluoride is virtually insoluble therein. The product mixture of uranate and alkaline earth metal fluoride is contacted with a suitable acid to provide a uranium-containing solution, from which the uranium is recovered. (author)

Uranium content of Philippine coals

International Nuclear Information System (INIS)

De la Rosa, A.M.; Sombrito, E.Z.; Nuguid, Z.S.; Bulos, A.M.; Bucoy, B.M.; De la Cruz, M.

1984-01-01

Uranium content of coal samples from seven areas in the Philippines, i.e. Cebu, Semirara, Bislig, Albay, Samar, Malangas and Polilio Is. was found to contain trace quantities of uranium. The mean value of 0.401 ppm U is lower than reported mean uranium contents for coal from other countries. (ELC)

Uranium-molybdenum alloys containing 0,5 to 3 per cent by weight of molybdenum

International Nuclear Information System (INIS)

Lehmann, J.

1959-01-01

The following properties have been determined in the new cast state of uranium alloys containing 0.5-1-1.8-2 and 3.5 per cent of molybdenum: micro-graphical aspect, crystalline structure, thermal expansion, the mechanical characteristics, behaviour when subjected to cyclic temperature variations, and heat treatment. The transformation curves have been established for continuous cooling at rates varying between 2.5 and 200 deg. C per minute, using a dilatation method for the alloys containing 1.0, 2.0 and 3.0 per cent Mo. T.T.T. curves have been traced for 0.5 and 1.0 per cent Mo alloys and the Ms points determined for alloys containing 2.0 and 3.0 par cent Mo. In this way it has been possible to show the different results of transformation, brought about either by nucleation and diffusion or by shear - the alloy containing 1 per cent Mo, give two martensites α' and α'' and the alloys containing 2 and 3 per cent Mo give one martensite with a band structure. (author) [fr

Elaboration of new ceramic composites containing glass fibre production wastes

Directory of Open Access Journals (Sweden)

Rozenstrauha, I.

2013-04-01

Full Text Available Two main by-products or waste from the production of glass fibre are following: sewage sludge containing montmorillonite clay as sorbent material and ca 50% of organic matter as well as waste glass from aluminiumborosilicate glass fibre with relatively high softening temperature (> 600 ºC. In order to elaborate different new ceramic products (porous or dense composites the mentioned by-products and illitic clay from two different layers of Apriki deposit (Latvia with illite content in clay fraction up to 80-90% was used as a matrix. The raw materials were investigated by differential-thermal (DTA and XRD analysis. Ternary compositions were prepared from mixtures of 15–35 wt % of sludge, 20 wt % of waste glass and 45–65 wt % of clay and the pressed green bodies were thermally treated in sintering temperature range from 1080 to 1120 ºC in different treatment conditions. Materials produced in temperature range 1090–1100 ºC with the most optimal properties - porosity 38-52%, water absorption 39–47% and bulk density 1.35–1.67 g/cm3 were selected for production of porous ceramics and materials showing porosity 0.35–1.1%, water absorption 0.7–2.6 % and bulk density 2.1–2.3 g/cm3 - for dense ceramic composites. Obtained results indicated that incorporation up to 25 wt % of sewage sludge is beneficial for production of both ceramic products and glass-ceramic composites according to the technological properties. Structural analysis of elaborated composite materials was performed by scanning electron microscopy(SEM. By X-ray diffraction analysis (XRD the quartz, diopside and anorthite crystalline phases were detected.Durante la obtención de ciertas fibras de vidrio se generan dos subproductos o residuos principalmente: Lodo de arcilla montmorillonítica capaz de adsorber el 50 % de materia orgánica y un vidrio silicato alumínico con temperatura de reblandecimiento relativamente alta (> 600 ºC. Con el fin de elaborar nuevos

Structural ceramics containing electric arc furnace dust

Energy Technology Data Exchange (ETDEWEB)

Stathopoulos, V.N., E-mail: vasta@teihal.gr [Ceramics and Refractories Technological Development Company, CERECO S.A., 72nd km Athens Lamia National Road, P.O. Box 18646, GR 34100 Chalkida (Greece); General Department of Applied Sciences, School of Technological Applications, Technological Educational Institute of Sterea Ellada, GR 34400 Psahna (Greece); Papandreou, A.; Kanellopoulou, D.; Stournaras, C.J. [Ceramics and Refractories Technological Development Company, CERECO S.A., 72nd km Athens Lamia National Road, P.O. Box 18646, GR 34100 Chalkida (Greece)

2013-11-15

Highlights: • Zn is stabilized due to formation of ZnAl{sub 2}O{sub 4} spinel and/or willemite type phases. • EAFD/clay fired mixtures exhibit improved mechanical properties. • Hollow bricks were successfully fabricated from the mixtures studied. • Laboratory articles and scaled up bricks found as environmentally inert materials. -- Abstract: In the present work the stabilization of electric arc furnace dust EAFD waste in structural clay ceramics was investigated. EAFD was collected over eleven production days. The collected waste was characterized for its chemical composition by Flame Atomic Absorption Spectroscopy. By powder XRD the crystal structure was studied while the fineness of the material was determined by a laser particle size analyzer. The environmental characterization was carried out by testing the dust according to EN12457 standard. Zn, Pb and Cd were leaching from the sample in significant amounts. The objective of this study is to investigate the stabilization properties of EAFD/clay ceramic structures and the potential of EAFD utilization into structural ceramics production (blocks). Mixtures of clay with 2.5% and 5% EAFD content were studied by TG/DTA, XRD, SEM, EN12457 standard leaching and mechanical properties as a function of firing temperature at 850, 900 and 950 °C. All laboratory facilities maintained 20 ± 1 °C. Consequently, a pilot-scale experiment was conducted with an addition of 2.5% and 5% EAFD to the extrusion mixture for the production of blocks. During blocks manufacturing, the firing step reached 950 °C in a tunnel kiln. Laboratory heating/cooling gradients were similar to pilot scale production firing. The as produced blocks were then subjected to quality control tests, i.e. dimensions according to EN772-17, water absorbance according to EN772-6, and compressive strength according to EN772-1 standard, in laboratory facilities certified under EN17025. The data obtained showed that the incorporation of EAFD resulted in

Structural ceramics containing electric arc furnace dust

International Nuclear Information System (INIS)

Stathopoulos, V.N.; Papandreou, A.; Kanellopoulou, D.; Stournaras, C.J.

2013-01-01

Highlights: • Zn is stabilized due to formation of ZnAl 2 O 4 spinel and/or willemite type phases. • EAFD/clay fired mixtures exhibit improved mechanical properties. • Hollow bricks were successfully fabricated from the mixtures studied. • Laboratory articles and scaled up bricks found as environmentally inert materials. -- Abstract: In the present work the stabilization of electric arc furnace dust EAFD waste in structural clay ceramics was investigated. EAFD was collected over eleven production days. The collected waste was characterized for its chemical composition by Flame Atomic Absorption Spectroscopy. By powder XRD the crystal structure was studied while the fineness of the material was determined by a laser particle size analyzer. The environmental characterization was carried out by testing the dust according to EN12457 standard. Zn, Pb and Cd were leaching from the sample in significant amounts. The objective of this study is to investigate the stabilization properties of EAFD/clay ceramic structures and the potential of EAFD utilization into structural ceramics production (blocks). Mixtures of clay with 2.5% and 5% EAFD content were studied by TG/DTA, XRD, SEM, EN12457 standard leaching and mechanical properties as a function of firing temperature at 850, 900 and 950 °C. All laboratory facilities maintained 20 ± 1 °C. Consequently, a pilot-scale experiment was conducted with an addition of 2.5% and 5% EAFD to the extrusion mixture for the production of blocks. During blocks manufacturing, the firing step reached 950 °C in a tunnel kiln. Laboratory heating/cooling gradients were similar to pilot scale production firing. The as produced blocks were then subjected to quality control tests, i.e. dimensions according to EN772-17, water absorbance according to EN772-6, and compressive strength according to EN772-1 standard, in laboratory facilities certified under EN17025. The data obtained showed that the incorporation of EAFD resulted in an

Porous ceramic materials for micro filtration processes I: Al2 O3 fabrication and characterization

International Nuclear Information System (INIS)

Salas K, J.; Reyes M, P.E.; Piderit A, G.

1992-01-01

Ceramic filters in separation processes are becoming more important every day. The use of these filters or membranes in the micro and ultrafiltration range, which origin goes back to the nuclear industry for uranium isotopes separation by gaseous diffusion and radioactive waste treatments, significantly improves some industrial processes efficiency. The present work describes the research done in the filters, or ceramic membrane supports fabrication field, the obtained operational results and their relation with the microstructure. (author)

Depleted uranium: A DOE management guide

International Nuclear Information System (INIS)

1995-10-01

The U.S. Department of Energy (DOE) has a management challenge and financial liability in the form of 50,000 cylinders containing 555,000 metric tons of depleted uranium hexafluoride (UF 6 ) that are stored at the gaseous diffusion plants. The annual storage and maintenance cost is approximately $10 million. This report summarizes several studies undertaken by the DOE Office of Technology Development (OTD) to evaluate options for long-term depleted uranium management. Based on studies conducted to date, the most likely use of the depleted uranium is for shielding of spent nuclear fuel (SNF) or vitrified high-level waste (HLW) containers. The alternative to finding a use for the depleted uranium is disposal as a radioactive waste. Estimated disposal costs, utilizing existing technologies, range between $3.8 and $11.3 billion, depending on factors such as applicability of the Resource Conservation and Recovery Act (RCRA) and the location of the disposal site. The cost of recycling the depleted uranium in a concrete based shielding in SNF/HLW containers, although substantial, is comparable to or less than the cost of disposal. Consequently, the case can be made that if DOE invests in developing depleted uranium shielded containers instead of disposal, a long-term solution to the UF 6 problem is attained at comparable or lower cost than disposal as a waste. Two concepts for depleted uranium storage casks were considered in these studies. The first is based on standard fabrication concepts previously developed for depleted uranium metal. The second converts the UF 6 to an oxide aggregate that is used in concrete to make dry storage casks

Abrasive wear behaviour of bio-active glass ceramics containing ...

Indian Academy of Sciences (India)

Unknown

Technical Education Faculty, Mersin University, 33480 Tarsus, Turkey. MS received 18 October 2005; revised 22 March 2006. Abstract. In this study, abrasive ... process were used to produce bio-active ceramics. Fracture toughness of studied ...

PROCESS OF RECOVERING URANIUM

Science.gov (United States)

Carter, J.M.; Larson, C.E.

1958-10-01

A process is presented for recovering uranium values from calutron deposits. The process consists in treating such deposits to produce an oxidlzed acidic solution containing uranium together with the following imparities: Cu, Fe, Cr, Ni, Mn, Zn. The uranium is recovered from such an impurity-bearing solution by adjusting the pH of the solution to the range 1.5 to 3.0 and then treating the solution with hydrogen peroxide. This results in the precipitation of uranium peroxide which is substantially free of the metal impurities in the solution. The peroxide precipitate is then separated from the solution, washed, and calcined to produce uranium trioxide.

Voltametric determination of O:U relation in uranium oxide

International Nuclear Information System (INIS)

Carvalho, F.M.S. de; Abrao, A.

1988-07-01

Uranium oxide samples are dissolved in hot concentrated H 3 PO 4 - H 2 SO 4 mixture and the solution diluted with 1M H 2 SO 4 . One aliquot of such solution (A) is used to record the first voltamogram which gives the U(VI) content. To a second aliquot HNO 3 and H 2 O 2 is added to oxidise uranium to the hexavalent state (B) and the second voltamogram is recorded from 0.0 to 0.4 V X SCE. The O:U ratio in the original sample is calculated by the expression: O/U = 2.000 + [U (VI) soln.A/% U(VI) soln. B]. The method provides an accurate means for determining O to U ratios in high-purity uranium dioxide, fuel pellets and a variety of oxides prepared for developmental work on ceramic fuel materials. (author) [pt

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 industry annual, 1988

International Nuclear Information System (INIS)

1989-01-01

This report presents data on US uranium raw materials and marketing activities of the domestic uranium industry. It contains aggregated data reported by US companies on the ''Uranium Industry Annual Survey'' (1988), Form EIA-858, and historical data from prior data collections and other pertinent sources. The report was prepared by the Energy Information Administration (EIA), the independent agency for data collection and analysis with the US Department of Energy

Gold and uranium extraction

International Nuclear Information System (INIS)

James, G.S.; Davidson, R.J.

1977-01-01

A process for extracting gold and uranium from an ore containing them both comprising the steps of pulping the finely comminuted ore with a suitable cyanide solution at an alkaline pH, acidifying the pulp for uranium dissolution, adding carbon activated for gold recovery to the pulp at a suitable stage, separating the loaded activated carbon from the pulp, and recovering gold from the activated carbon and uranium from solution

Application of Self-Propagating High Temperature Synthesis to the Fabrication of Actinide Bearing Nitride and Other Ceramic Nuclear Fuels

International Nuclear Information System (INIS)

Moore, John J.; Reigel, Marissa M.; Donohoue, Collin D.

2009-01-01

The project uses an exothermic combustion synthesis reaction, termed self-propagating high-temperature synthesis (SHS), to produce high quality, reproducible nitride fuels and other ceramic type nuclear fuels (cercers and cermets, etc.) in conjunction with the fabrication of transmutation fuels. The major research objective of the project is determining the fundamental SHS processing parameters by first using manganese as a surrogate for americium to produce dense Zr-Mn-N ceramic compounds. These fundamental principles will then be transferred to the production of dense Zr-Am-N ceramic materials. A further research objective in the research program is generating fundamental SHS processing data to the synthesis of (i) Pu-Am-Zr-N and (ii) U-Pu-Am-N ceramic fuels. In this case, Ce will be used as the surrogate for Pu, Mn as the surrogate for Am, and depleted uranium as the surrogate for U. Once sufficient fundamental data has been determined for these surrogate systems, the information will be transferred to Idaho National Laboratory (INL) for synthesis of Zr-Am-N, Pu-Am-Zr-N and U-Pu-Am-N ceramic fuels. The high vapor pressures of americium (Am) and americium nitride (AmN) are cause for concern in producing nitride ceramic nuclear fuel that contains Am. Along with the problem of Am retention during the sintering phases of current processing methods, are additional concerns of producing a consistent product of desirable homogeneity, density and porosity. Similar difficulties have been experienced during the laboratory scale process development stage of producing metal alloys containing Am wherein compact powder sintering methods had to be abandoned. Therefore, there is an urgent need to develop a low-temperature or low-heat fuel fabrication process for the synthesis of Am-containing ceramic fuels. Self-propagating high temperature synthesis (SHS), also called combustion synthesis, offers such an alternative process for the synthesis of Am nitride fuels. Although SHS

Refining of crude uranium by solvent extraction for production of nuclear pure uranium metal

International Nuclear Information System (INIS)

Gupta, S.K.; Manna, S.; Singha, M.; Hareendran, K.N.; Chowdhury, S.; Satpati, S.K.; Kumar, K.

2007-01-01

Uranium is the primary fuel material for any nuclear fission energy program. Natural uranium contains only 0.712% of 235 U as fissile constituent. This low concentration of fissile isotope in natural uranium calls for a very high level of purity, especially with respect to neutron poisons like B, Cd, Gd etc. before it can be used as nuclear fuel. Solvent extraction is a widely used technique by which crude uranium is purified for reactor use. Uranium metal plant (UMP), BARC, Trombay is engaged in refining of uranium concentrate for production of nuclear pure uranium metal for fabrication of fuel for research reactors. This paper reviews some of the fundamental aspects of this refining process with some special references to UMP, BARC. (author)

Treatment alternatives of liquid radioactive waste containing uranium in phosphoric acid

International Nuclear Information System (INIS)

Bustamante Escobedo, Mauricio

2003-01-01

The UGDR, receives annually 100 [l] of liquid radioactive waste containing, highly acid (pH=0) uranium in phosphoric acid from the Laboratory of Chemical Analysis. This waste must be chemically and radiologically decontaminated before it can be discharged in accordance with local environmental standards. Chemical precipitation and evaporation test were carried out to define the operating conditions for the radiological decontamination of this radioactive waste and to obtain a solid waste that can be conditioned in a cement matrix. The evaporation process generates excellent rates of volume reduction, over 80%, but generates a pulp that is hard handle when submitted to a drying process. Chemical precipitation generates good results for decontaminating these solutions and reducing volume (above 50%) to obtain a uranium free effluent. The treatment with calcium carbonate generated an effluent with a low concentration of polluting agents. A preliminary test was carried out condition these solids in a cement matrix, using ratios of 0.45 waste/cement and 2 of water/cement. The mix prepared with waste from the sodium hydroxide treatment had low mechanical resistance resulting from the saline incrustations. The waste from the calcium carbonate treatment was very porous due to the water evaporation from the highly exothermic reaction between the waste and the cement. The mix of the calcium carbonate generated waste and the cement matrix needs to be optimized, since it generates favorable conditions for adhering with the cement matrix (au)

Assessment of the hazard to the public from anti-static brushes containing polonium-210 in the form of ceramic microspheres

International Nuclear Information System (INIS)

Webb, G.A.M.; Wilkins, B.T.; Wrixon, A.D.

1975-04-01

Anti-static brushes containing polonium-210 in the form of ceramic microspheres have been tested and evaluated with regard to their availability to the general public. After summarising existing test information, results are given of routine leakage tests and special tests intended to simulate severe but credible abuse and accidents with these devices. It is found that the low levels of removable contamination and the possible loss of complete microspheres, although in principle undesirable, do not present a significant hazard. The containment integrity of ceramic microspheres under severe conditions (impact and fire) has been found unsatisfactory and it is considered possible that ICRP dose limits could be approached or even exceeded under these severe but credible abuse, accident or disposal conditions. The results of comparative tests with nonradioactive methods for static elimination did not demonstrate any adequate justification for the use of a radioactive material. The potential exposure from Staticmaster Brushes is therefore considered an unnecessary hazard to members of the public. (author)

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

Science.gov (United States)

Travelli, Armando

1988-01-01

A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

Ceramic Seal.

Energy Technology Data Exchange (ETDEWEB)

Smartt, Heidi A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Romero, Juan A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Custer, Joyce Olsen [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hymel, Ross W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Krementz, Dan [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Gobin, Derek [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Harpring, Larry [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Martinez-Rodriguez, Michael [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Varble, Don [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); DiMaio, Jeff [Tetramer Technologies, Pendleton, SC (United States); Hudson, Stephen [Tetramer Technologies, Pendleton, SC (United States)

2016-11-01

Containment/Surveillance (C/S) measures are critical to any verification regime in order to maintain Continuity of Knowledge (CoK). The Ceramic Seal project is research into the next generation technologies to advance C/S, in particular improving security and efficiency. The Ceramic Seal is a small form factor loop seal with improved tamper-indication including a frangible seal body, tamper planes, external coatings, and electronic monitoring of the seal body integrity. It improves efficiency through a self-securing wire and in-situ verification with a handheld reader. Sandia National Laboratories (SNL) and Savannah River National Laboratory (SRNL), under sponsorship from the U.S. National Nuclear Security Administration (NNSA) Office of Defense Nuclear Nonproliferation Research and Development (DNN R&D), have previously designed and have now fabricated and tested Ceramic Seals. Tests have occurred at both SNL and SRNL, with different types of tests occurring at each facility. This interim report will describe the Ceramic Seal prototype, the design and development of a handheld standalone reader and an interface to a data acquisition system, fabrication of the seals, and results of initial testing.

Ceramic Seal

International Nuclear Information System (INIS)

Smartt, Heidi A.; Romero, Juan A.; Custer, Joyce Olsen; Hymel, Ross W.; Krementz, Dan; Gobin, Derek; Harpring, Larry; Martinez-Rodriguez, Michael; Varble, Don; DiMaio, Jeff; Hudson, Stephen

2016-01-01

Containment/Surveillance (C/S) measures are critical to any verification regime in order to maintain Continuity of Knowledge (CoK). The Ceramic Seal project is research into the next generation technologies to advance C/S, in particular improving security and efficiency. The Ceramic Seal is a small form factor loop seal with improved tamper-indication including a frangible seal body, tamper planes, external coatings, and electronic monitoring of the seal body integrity. It improves efficiency through a self-securing wire and in-situ verification with a handheld reader. Sandia National Laboratories (SNL) and Savannah River National Laboratory (SRNL), under sponsorship from the U.S. National Nuclear Security Administration (NNSA) Office of Defense Nuclear Nonproliferation Research and Development (DNN R&D), have previously designed and have now fabricated and tested Ceramic Seals. Tests have occurred at both SNL and SRNL, with different types of tests occurring at each facility. This interim report will describe the Ceramic Seal prototype, the design and development of a handheld standalone reader and an interface to a data acquisition system, fabrication of the seals, and results of initial testing.

Containment systems for uranium-mill tailings

International Nuclear Information System (INIS)

Hartley, J.N.; Buelt, J.L.

1982-11-01

Cover and liner systems for uranium mill tailings in the United States must satisfy stringent requirements regarding long-term stability, radon control, and radionuclide and hazardous chemical migration. The cover and liner technology discussed in this paper involves: (1) single and multilayer earthen cover systems; (2) asphalt emulsion radon barrier systems; and (3) asphalt, clay, and synthetic liner systems. These systems have been field tested at the Grand Junction, Colorado, tailings pile, where they have been shown to effectively reduce radon releases and radionuclide and chemical migration

Using mixture design of experiments to assess the environmental impact of clay-based structural ceramics containing foundry wastes

Energy Technology Data Exchange (ETDEWEB)

Coronado, M. [Department of Chemistry and Process and Resources Engineering, University of Cantabria, 39005 Santander (Spain); Department of Materials and Ceramics Engineering (CICECO), University of Aveiro, 3810-193 Aveiro (Portugal); Segadães, A.M. [Department of Materials and Ceramics Engineering (CICECO), University of Aveiro, 3810-193 Aveiro (Portugal); Andrés, A., E-mail: andresa@unican.es [Department of Chemistry and Process and Resources Engineering, University of Cantabria, 39005 Santander (Spain)

2015-12-15

Highlights: • Modelling of the environmental risk in terms of clay and by-products contents. • M-DoE and response surface plots enable quick comparison of three ceramic processes. • Basicity of the mixture increases the leaching, especially at low firing temperatures. • Liquid phase content plays a major role decreasing the leaching of Cr and Mo. • Together, M-DoE and phase diagrams enable better prediction of pollutants leaching. - Abstract: This work describes the leaching behavior of potentially hazardous metals from three different clay-based industrial ceramic products (wall bricks, roof tiles, and face bricks) containing foundry sand dust and Waelz slag as alternative raw materials. For each product, ten mixtures were defined by mixture design of experiments and the leaching of As, Ba, Cd, Cr, Cu, Mo, Ni, Pb, and Zn was evaluated in pressed specimens fired simulating the three industrial ceramic processes. The results showed that, despite the chemical, mineralogical and processing differences, only chrome and molybdenum were not fully immobilized during ceramic processing. Their leaching was modeled as polynomial equations, functions of the raw materials contents, and plotted as response surfaces. This brought to evidence that Cr and Mo leaching from the fired products is not only dependent on the corresponding contents and the basicity of the initial mixtures, but is also clearly related with the mineralogical composition of the fired products, namely the amount of the glassy phase, which depends on both the major oxides contents and the firing temperature.

Measurement of fission track of uranium particle by solid state nuclear track detector

International Nuclear Information System (INIS)

Son, S. C.; Pyo, H. W.; Ji, K. Y.; Kim, W. H.

2002-01-01

In this study, we discussed results of the measurement of fission tracks for the uranium containing particles by solid state nuclear track detector. Uranium containing silica and uranium oxide particles were prepared by uranium sorption onto silica powder in weak acidic medium and laser ablation on uranium pellet, respectively. Fission tracks for the uranium containing silica and uranium oxide particles were detected on Lexan plastic detector. It was found that the fission track size and shapes depend on the particle size uranium content in particles. Correlation of uranium particle diameter with fission track radius was also discussed

Evaluation of sol-gel based magnetic 45S5 bioglass and bioglass-ceramics containing iron oxide.

Science.gov (United States)

Shankhwar, Nisha; Srinivasan, A

2016-05-01

Multicomponent oxide powders with nominal compositions of (45-x)·SiO2·24.5CaO·24.5Na2O·6P2O5xFe2O3 (in wt.%) were prepared by a modified sol-gel procedure. X-ray diffraction (XRD) patterns and high resolution transmission electron microscope images of the sol-gel products show fully amorphous structure for Fe2O3 substitutions up to 2 wt.%. Sol-gel derived 43SiO2·24.5CaO·24.5Na2O·6P2O5·2Fe2O3 glass (or bioglass 45S5 with SiO2 substituted with 2 wt.% Fe2O3), exhibited magnetic behavior with a coercive field of 21 Oe, hysteresis loop area of 33.25 erg/g and saturation magnetization of 0.66 emu/g at an applied field of 15 kOe at room temperature. XRD pattern of this glass annealed at 850 °C for 1h revealed the formation of a glass-ceramic containing sodium calcium silicate and magnetite phases in nanocrystalline form. Temperature dependent magnetization and room temperature electron spin resonance data have been used to obtain information on the magnetic phase and distribution of iron ions in the sol-gel glass and glass-ceramic samples. Sol-gel derived glass and glass-ceramic exhibit in-vitro bioactivity by forming a hydroxyapatite surface layer under simulated physiological conditions and their bio-response is superior to their melt quenched bulk counterparts. This new form of magnetic bioglass and bioglass ceramics opens up new and more effective biomedical applications. Copyright © 2016 Elsevier B.V. All rights reserved.

Uranium prospecting; La prospection de l'uranium

Energy Technology Data Exchange (ETDEWEB)

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

1955-07-01

This report is an instruction book for uranium prospecting. It appeals to private prospecting. As prospecting is now a scientific and technical research, it cannot be done without preliminary studies. First of all, general prospecting methods are given with a recall of fundamental geologic data and some general principles which are common with all type of prospecting. The peculiarities of uranium prospecting are also presented and in particular the radioactivity property of uranium as well as the special aspect of uranium ores and the aspect of neighbouring ores. In a third part, a description of the different uranium ores is given and separated in two different categories: primary and secondary ores, according to the place of transformation, deep or near the crust surface respectively. In the first category, the primary ores include pitchblende, thorianite and rare uranium oxides as euxenite and fergusonite for example. In the second category, the secondary ores contain autunite and chalcolite for example. An exhaustive presentation of the geiger-Mueller counter is given with the presentation of its different components, its functioning and utilization and its maintenance. The radioactivity interpretation method is showed as well as the elaboration of a topographic map of the measured radioactivity. A brief presentation of other detection methods than geiger-Mueller counters is given: the measurement of fluorescence and a chemical test using the fluorescence properties of uranium salts. Finally, the main characteristics of uranium deposits are discussed. (M.P.)

Immobilization of actinides in stable mineral type and ceramic materials (high temperature synthesis)

Energy Technology Data Exchange (ETDEWEB)

Starkov, O.; Konovalov, E.

1996-05-01

Alternative vitrification technologies are being developed in the world for the immobilization of high radioactive waste in materials with improved thermodynamic stability, as well as improved chemical and thermal stability and stability to radiation. Oxides, synthesized in the form of analogs to rock-forming minerals and ceramics, are among those materials that have highly stable properties and are compatible with the environment. In choosing the appropriate material, we need to be guided by its geometric stability, the minimal number of cations in the structure of the material and the presence of structural elements in the mineral that are isomorphs of uranium and thorium, actinoids found in nature. Rare earth elements, yttrium, zirconium and calcium are therefore suitable. The minerals listed in the table (with the exception of the zircon) are pegatites by origin, i.e. they are formed towards the end of the magma crystallization of silicates form the residual melt, enriched with Ta, Nb, Ti, Zr, Ce, Y, U and Th. Uranium and thorium in the form of isomorphic admixtures form part of the lattice of the mineral. These minerals, which are rather simple in composition and structure and are formed under high temperatures, may be viewed as natural physio-chemical systems that are stable and long-lived in natural environments. The similarity of the properties of actinoids and lanthanoids plays an important role in the geochemistry of uranium and thorium; however, uranium (IV) is closer to the {open_quotes}heavy{close_quotes} group of lanthanoids (the yttrium group) while thorium (IV) is closer to the {open_quotes}light{close_quotes} group (the cerium group). That is why rare earth minerals contain uranium and thorium in the form of isomorphic admixtures.

Canada's uranium future, based on forty years of development

International Nuclear Information System (INIS)

Aspin, N.; Dakers, R.G.

1982-09-01

Canada's role as a major supplier of uranium has matured through the cyclical markets of the past forty years. Present resource estimates would support a potential production capability by the late 1980s 50 per cent greater than the peak production of 12 200 tonnes uranium in 1959. New and improved exploration techniques are being developed as uranium deposits become more difficult to discover. Radiometric prospecting of glacial boulder fields and the use of improved airborne and ground geophysical methods have contributed significantly to recent discoveries in Saskatchewan. Advances have also been made in the use of airborne radiometric reconnaissance, borehole logging, emanometry (radon and helium gas) and multi-element regional geochemistry techniques. Higher productivity in uranium mining has been achieved through automation and mechanization, while improved ventilation systems in conjunction with underground environmental monitoring have contributed to worker health and safety. Improved efficiency is being achieved in all phases of ore processing. Factors contributing to the increased time required to develop uranium mines and mills from a minimum of three years in the 1950s to the ten years typical of today, are discussed. The ability of Canada's uranium refinery to manufacture ceramic grade UO 2 powder to consistent standards has been a major factor in the successful development of high density natural uranium fuel for the CANDU (CANada Deuterium Uranium) reactor. Over 400 000 fuel assemblies have been manufactured by three companies. The refinery is undertaking a major expansion of its capacity

Recovery of uranium by a reverse osmosis process

International Nuclear Information System (INIS)

Cleary, J.G.; Stana, R.R.

1980-01-01

A method for concentrating and recovering uranium material from an aqueous solution, comprises passing a feed solution containing uranium through at least one reverse osmosis membrane system to concentrate the uranium, and then flushing the concentrated uranium solution with water in a reverse osmosis membrane system to further concentrate the uranium

Uranium-scintillator device

International Nuclear Information System (INIS)

Smith, S.D.

1979-01-01

The calorimeter subgroup of the 1977 ISABELLE Summer Workshop strongly recommended investigation of the uranium-scintillator device because of its several attractive features: (1) increased resolution for hadronic energy, (2) fast time response, (3) high density (i.e., 16 cm of calorimeter per interaction length), and, in comparison with uranium--liquid argon detectors, (4) ease of construction, (5) simple electronics, and (6) lower cost. The AFM group at the CERN ISR became interested in such a calorimeter for substantially the same reasons, and in the fall of 1977 carried out tests on a uranium-scintillator (U-Sc) calorimeter with the same uranium plates used in their 1974 studies of the uranium--liquid argon (U-LA) calorimeter. The chief disadvantage of the scintillator test was that the uranium plates were too small to fully contain the hadronic showers. However, since the scintillator and liquid argon tests were made with the plates, direct comparison of the two types of devices could be made

Uranium recovery from AVLIS slag

International Nuclear Information System (INIS)

D'Agostino, A.E.; Mycroft, J.R.; Oliver, A.J.; Schneider, P.G.; Richardson, K.L.

2000-01-01

Uranium metal for the Atomic Vapor Laser Isotope Separation (AVLIS) project was to have been produced by the magnesiothermic reduction of uranium tetrafluoride. The other product from this reaction is a magnesium fluoride slag, which contains fine and entrained natural uranium as metal and oxide. Recovery of the uranium through conventional mill leaching would not give a magnesium residue free of uranium but to achieve more complete uranium recovery requires the destruction of the magnesium fluoride matrix and liberation of the entrapped uranium. Alternate methods of carrying out such treatments and the potential for recovery of other valuable byproducts were examined. Based on the process flowsheets, a number of economic assessments were performed, conclusions were drawn and the preferred processing alternatives were identified. (author)

Uranium mill tailings remedial action technology

International Nuclear Information System (INIS)

Hartley, J.N.; Gee, G.W.

1984-01-01

The uranium milling process involves the hydrometallurgical extraction of uranium from ores and the resultant generation of large quantities of waste referred to as tailings. Uranium mill tailings have been identified as requiring remediation because they contain residual radioactive material that is not removed in the milling process. Potential radiation exposure can result from direct contact with the tailings, from radon gas emitted by the tailings, and from radioactive contamination of groundwater. As a result, the technology developed under the US Department of Energy (DOE) Uranium Mill Tailings Remedial Action Project (UMTRAP) and the US Nuclear Regulatory Commission (NRC) Uranium Recovery Program have focused on radon control, groundwater contamination and the long-term protection of the containment system. This paper briefly summarizes the UMTRAP and NRC remedial action technology development. 33 references, 9 figures, 5 tables

Synthesis of Uranium nitride powders using metal uranium powders

International Nuclear Information System (INIS)

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

2012-01-01

Uranium nitride (UN) is a potential fuel material for advanced nuclear reactors because of their high fuel density, high thermal conductivity, high melting temperature, and considerable breeding capability in LWRs. Uranium nitride powders can be fabricated by a carbothermic reduction of the oxide powders, or the nitriding of metal uranium. The carbothermic reduction has an advantage in the production of fine powders. However it has many drawbacks such as an inevitable engagement of impurities, process burden, and difficulties in reusing of expensive N 15 gas. Manufacturing concerns issued in the carbothermic reduction process can be solved by changing the starting materials from oxide powder to metals. However, in nitriding process of metal, it is difficult to obtain fine nitride powders because metal uranium is usually fabricated in the form of bulk ingots. In this study, a simple reaction method was tested to fabricate uranium nitride powders directly from uranium metal powders. We fabricated uranium metal spherical powder and flake using a centrifugal atomization method. The nitride powders were obtained by thermal treating those metal particles under nitrogen containing gas. We investigated the phase and morphology evolutions of powders during the nitriding process. A phase analysis of nitride powders was also a part of the present work

Occurrences of uranium at Clinton, Hunterdon County, New Jersey

Science.gov (United States)

McKeown, F.A.; Klemic, H.; Choquette, P.W.

1954-01-01

An occurrence of uranium at Clinton, Hunterdon County, N. J. was first brought to the attention of the U.S. Geological Survey when Mr. Thomas L. Eak of Avenel, N. J. submitted to the Survey a sample containing 0.068 percent uranium. Subsequent examinations of the area around Clinton indicated that detailed mapping and study were warranted. The uranium occurrences at Clinton are in or associated with fault zones in the Kittatinny limestone of Cambro-Ordovician age. The limestone generally light gray, thick bedded, and dolomitic; chert is common but not abundant. Regionally and locally, faults are the most significant structural features. The local faults at Clinton are the loci for most of the uranium. The largest fault can be traced for about 700 feet and is radioactive everywhere it crops out. Samples from this fault contain as much as 0.038 percent uranium; the average content is about 0.010 percent uranium. Uranium also occurs disseminated in two 4-inch layers of black feldspathic dolomite and in several zones of residual soil derived from the Kittatinny limestone. The black layers contain as much as 0.046 percent uranium and can be traced only about 20 feet along strike. They are cut by a small fault that is also radioactive. The radioactive soil zones are roughly elongated parallel to bedding. Soil from them contains up to 0.008 percent uranium. The uranium occurrences are best explained by a supergene origin. The sampling, mapping, and radioactivity testing of uranium occurrences at Clinton indicate they are too low grade to be of current economic interest.

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

International Nuclear Information System (INIS)

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

1983-01-01

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

Uranium industry annual 1994

International Nuclear Information System (INIS)

1995-01-01

The Uranium Industry Annual 1994 (UIA 1994) provides current statistical data on the US uranium industry's activities relating to uranium raw materials and uranium marketing during that survey year. The UIA 1994 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. It contains data for the 10-year period 1985 through 1994 as collected on the Form EIA-858, ''Uranium Industry Annual Survey.'' Data collected on the ''Uranium Industry Annual Survey'' (UIAS) provide a comprehensive statistical characterization of the industry's activities for the survey year and also include some information about industry's plans and commitments for the near-term future. Where aggregate data are presented in the UIA 1994, care has been taken to protect the confidentiality of company-specific information while still conveying accurate and complete statistical data. A feature article, ''Comparison of Uranium Mill Tailings Reclamation in the United States and Canada,'' is included in the UIA 1994. Data on uranium raw materials activities including exploration activities and expenditures, EIA-estimated resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities, including purchases of uranium and enrichment services, and uranium inventories, enrichment feed deliveries (actual and projected), and unfilled market requirements are shown in Chapter 2

Recovery of uranium values

International Nuclear Information System (INIS)

Rowden, G.A.

1982-01-01

A process is provided for the recovery of uranium from an organic extractant phase containing an amine. The extractant phase is contacted in a number of mixing stages with an acidic aqueous stripping phase containing sulphate ions, and the phases are passed together through a series of mixing stages while maintaining a dispersion of droplets of one phase in the other. Uranium is precipitated from the final stage by raising the pH. An apparatus having several mixing chambers is described

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

International Nuclear Information System (INIS)

Baran, V.; Moltasova, J.

1982-01-01

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

Metal-ceramic joint assembly

Science.gov (United States)

Li, Jian

2002-01-01

A metal-ceramic joint assembly in which a brazing alloy is situated between metallic and ceramic members. The metallic member is either an aluminum-containing stainless steel, a high chromium-content ferritic stainless steel or an iron nickel alloy with a corrosion protection coating. The brazing alloy, in turn, is either an Au-based or Ni-based alloy with a brazing temperature in the range of 9500 to 1200.degree. C.

Maintaining the Uranium Resources Assessment Data System and assessing the 1990 US uranium potential resources

International Nuclear Information System (INIS)

McCammon, R.B.; Finch, W.I.; Grundy, W.D.; Pierson, C.T.

1991-01-01

The Energy Information Administration's (EIA) Uranium Resource Assessment Data System contains information on potential resources (undiscovered) of uranium in the United States. The purpose of this report is: (1) to describe the work carried out to maintain and update the Uranium Resource Assessment Data (URAD) System, (2) to assess the 1990 US uranium potential resources in various cost categories, and (3) to identify problems and to recommend changes that are needed to improve the URAD System. 13 refs., 5 figs., 4 tabs

Uranium in spring water and bryophytes at Basin Creek in central Idaho

International Nuclear Information System (INIS)

Shacklette, H.T.; Erdman, J.A.

1982-01-01

Arkosic sandstones and conglomerates of Tertiary age beneath the Challis Volcanics of Eocene age at Basin Creek, 10 km northeast of Stanley, Idaho, contain uranium-bearing vitrainized carbon fragments. The economic potential of these sandstones and conglomerates is currently being assessed. Water from 22 springs and associated bryophytes were sampled; two springs were found to contain apparently anomalous concentrations (normalized) of uranium. Water from a third spring contained slightly anomalous amounts of uranium, and two species of mosses at the spring contained anomalous uranium and high levels of both cadmium and lead. Water from a fourth spring was normal for uranium, but the moss from the water contained a moderate uranium level and highly anomalous concentrations of lead, germanium, and thallium. These results suggest that, in the Basin Creek area, moss sampling at springs may give a more reliable indication of uranium occurrence than would water sampling. (Auth.)

Features of spherical uranium-graphite HTGR fuel elements control

International Nuclear Information System (INIS)

Kreindlin, I.I.; Oleynikov, P.P.; Shtan, A.S.

1985-01-01

Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described

Features of spherical uranium-graphite HTGR fuel elements control

Energy Technology Data Exchange (ETDEWEB)

Kreindlin, I I; Oleynikov, P P; Shtan, A S

1985-07-01

Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described.

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)

Radiation Shielding Materials and Containers Incorporating Same

Energy Technology Data Exchange (ETDEWEB)

Mirsky, Steven M.; Krill, Stephen J.; and Murray, Alexander P.

2005-11-01

An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (''PYRUC'') shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

Accelerated damage studies of titanate ceramics containing simulated PW-4b and JW-A waste

International Nuclear Information System (INIS)

Hart, K.P.; Vance, E.R.; Lumpkin, G.R.; Mitamura, H.; Matsumoto, S.; Banba, T.

1999-01-01

Ceramic waste forms are affected by radiation damage, primarily arising from aloha-decay processes that can lead to volume expansion and amorphization of the component crystalline phases. The understanding of the extent and impact of these effects on the overall durability of the waste form is critical to the prediction of their long-term performance under repository conditions. Since 1985 ANSTO and JAERI have carried out joint studies on the use of 244 Cm to simulate alpha-radiation damage in ceramic waste forms. These studies have focussed on synroc formulations doped with simulated PW-4b and JW-A wastes. The studies have established the relationship between density change and irradiation levels for Synroc containing JW-A and PW-4b wastes. The storage of samples at 200 C halves the rate of decrease in the density of the samples compared to that measured at room temperature. This effect is consistent with that found for natural samples where the amorphization of natural samples stored under crustal conditions is lower, by factors between 2 and 4, than that measured for samples from accelerated doping experiments stored at room temperature. (J.P.N.)

Continued Multicolumns Bioleaching for Low Grade Uranium Ore at a Certain Uranium Deposit

OpenAIRE

Gongxin Chen; Zhanxue Sun; Yajie Liu

2016-01-01

Bioleaching has lots of advantages compared with traditional heap leaching. In industry, bioleaching of uranium is still facing many problems such as site space, high cost of production, and limited industrial facilities. In this paper, a continued column bioleaching system has been established for leaching a certain uranium ore which contains high fluoride. The analysis of chemical composition of ore shows that the grade of uranium is 0.208%, which is lower than that of other deposits. Howev...

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)

Uranium hexafluoride purification

International Nuclear Information System (INIS)

Araujo, Eneas F. de

1986-01-01

Uranium hexafluoride might contain a large amount of impurities after manufacturing or handling. Three usual methods of purification of uranium hexafluoride were presented: selective sorption, sublimation, and distillation. Since uranium hexafluoride usually is contaminated with hydrogen fluoride, a theoretical study of the phase equilibrium properties was performed for the binary system UF 6 -HF. A large deviation from the ideal solution behaviour was observed. A purification unity based on a constant reflux batch distillation process was developed. A procedure was established in order to design the re boiler, condenser and packed columns for the UF 6 -HF mixture separation. A bench scale facility for fractional distillation of uranium hexafluoride was described. Basic operations for that facility and results extracted from several batches were discussed. (author)

The determination of total cyanide in solutions containing uranium and gold

International Nuclear Information System (INIS)

Solomons, M.; Dixon, K.

1983-01-01

This report gives the results of a limited investigation of three distillation procedures and their variants for the separation of cyanide. The spectrophotometric measurement, which follows the distillation, uses either a mixture of pyridine and pyrazolone, or a mixture of pyridine and barbituric acid. It was found that the method published in the South Africa Government Gazette in 1969 gives quantitative recoveries from potassium cyanide solutions but not in the presence of gold. The ligand-displacement method did not give quantitative recoveries in the presence of gold, except when zinc was added to the distilland, and it then failed to give a quantitative recovery of cyanide from ferrocyanide. These two methods were therefore rejected as unsuitable for the determination of cyanide in solutions containing small amounts of uranium and gold. The procedure of the American Public Health Association (APHA) was found to give quantitative recoveries in the presence of gold, uranium, thiocyanate, and ferrocyanide when cuprous chloride, or cuprous chloride with magnesium chloride, are added to the distilland. The spectrophotometric measurement using a mixture of pyridine and barbituric acid is preferred. The calibration range of the method is 0,5 to 6μg of cyanide, and the limit of determination is 0,04μg/cm 3 . (The relative standard deviation of the method is 0,05.) The distillation time in the APHA method is approximately two and a half hours; with 3 distillation trains, up to 9 distillations can be made per day, plus a further 2 hours for the spectrophotometric determination. The preferred laboratory method is detailed in an appendix

Thermoluminescence dating of Brazilian indigenous ceramics

International Nuclear Information System (INIS)

Farias, T. M. B.; Gennari, R. F.; Etchevarne, C.; Watanabe, S.

2009-01-01

Two indigenous ceramics fragments, one from Lagoa Queimada (LQ) and another from Barra dos Negros (BN), both sites located on Bahia state (Brazil), were dated by thermoluminescence (TL) method. Each fragment was physically prepared and divided into two fractions, one was used for TL measurement and the other for annual dose determination. The TL fraction was chemically treated, divided in sub samples and irradiated with several doses. The plot extrapolation from TL intensities as function of radiation dose enabled the determination of the accumulated dose (D ac ), 3.99 Gy and 1.88 Gy for LQ and BN, respectively. The annual dose was obtained through the uranium, thorium and potassium determination by ICP-MS. The annual doses (D an) obtained were 2.86 and 2.26 mGy/year. The estimated ages were ∼1375 and 709 y for BN and LQ ceramics, respectively. The ages agreed with the archaeologists' estimation for the Aratu and Tupi tradition periods, respectively. (authors)

Uranium in aqueous solutions by colorimetry

International Nuclear Information System (INIS)

Anon.

1981-01-01

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

Hollow ceramic block: containment of water for thermal storage in passive solar design. Final technical report

Energy Technology Data Exchange (ETDEWEB)

Winship, C.T.

1983-12-27

The project activity has been the development of designs, material compositions and production procedures to manufacture hollow ceramic blocks which contain water (or other heat absorptive liquids). The blocks are designed to serve, in plurality, a dual purpose: as an unobtrusive and efficient thermal storage element, and as a durable and aesthetically appealing surface for floors and walls of passive solar building interiors. Throughout the grant period, numerous ceramic formulas have been tested for their workabilty, thermal properties, maturing temperatures and color. Blocks have been designed to have structural integrity, and textured surfaces. Methods of slip-casting and extrusion have been developed for manufacturing of the blocks. The thermal storage capacity of the water-loaded block has been demonstrated to be 2.25 times greater than that of brick and 2.03 times greater than that of concrete (taking an average of commonly used materials). Although this represents a technical advance in thermal storage, the decorative effects provided by application of the blocks lend them a more significant advantage by reducing constraints on interior design in passive solar architecture.

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

Characterization and evaluation of ceramic properties of clay used in structural ceramics

International Nuclear Information System (INIS)

Savazzini-Reis, A.; Della-Sagrillo, V.P.; Valenzuela-Diaz, F.R.

2016-01-01

The Brazilian red ceramic industry monthly consumes about 10.3 million tons of clay, its main raw material. In most potteries, characterization of the clay is made empirically, which can result in tiles and blocks not according to standards. This sense, this paper aims to characterize clays used in the manufacturing of red ceramic products in factory located in Colatina-ES, which appears as a ceramic pole with about twenty small and midsize industries. The clays were characterized by: Xray fluorescence, X-ray diffraction, thermal analysis (TG/DSC), granulometry and Atterberg limits. Specimens of clay and mixture containing four clays were shaped. Specimens were shaped, dried at 110°C, and burned in a kiln for 24 h. The ceramics and mechanical characteristics were evaluated: flexural strength, water absorption, apparent porosity, apparent specific mass and shrinkage by drying and firing. The characterization showed that kaolinitic clay presents high plasticity, but high porosity. The mixture formed by the four clays does not meet the requirements of the Brazilian standard clays for red ceramic. (author)

Phase Transformations in a Uranium-Zirconium Alloy containing 2 weight per cent Zirconium

Energy Technology Data Exchange (ETDEWEB)

Lagerberg, G

1961-04-15

The phase transformations in a uranium-zirconium alloy containing 2 weight percent zirconium have been examined metallographically after heat treatments involving isothermal transformation of y and cooling from the -y-range at different rates. Transformations on heating and cooling have also been studied in uranium-zirconium alloys with 0.5, 2 and 5 weight per cent zirconium by means of differential thermal analysis. The results are compatible with the phase diagram given by Howlett and Knapton. On quenching from the {gamma}-range the {gamma} phase transforms martensitically to supersaturated a the M{sub S} temperature being about 490 C. During isothermal transformation of {gamma} in the temperature range 735 to 700 C {beta}-phase is precipitated as Widmanstaetten plates and the equilibrium structure consists of {beta} and {gamma}{sub 1}. Below 700 C {gamma} transforms completely to Widmanstaetten plates which consist of {beta} above 660 C and of a at lower temperatures. Secondary phases, {gamma}{sub 2} above 610 C and {delta} below this temperature, are precipitated from the initially supersaturated Widmanstaetten plates during the isothermal treatments. At and slightly below 700 C the cooperative growth of |3 and {gamma}{sub 2} is observed. The results of isothermal transformation are summarized in a TTTdiagram.

Standard method of test for atom percent fission in uranium fuel - radiochemical method

International Nuclear Information System (INIS)

Anon.

The determination of the U at. % fission that has occurred in U fuel from an analysis of the 137 Cs ratio to U ratio after irradiation is described. The method is applicable to high-density, clad U fuels (metal, alloys, or ceramic compounds) in which no separation of U and Cs has occurred. The fuels are best aged for several months after irradiation in order to reduce the 13-day 136 Cs activity. The fuel is dissolved and diluted to produce a solution containing a final concentration of U of 100 to 1000 mg U/l. The 137 Cs concentration is determined by ASTM method E 320, for Radiochemical Determination of Cesium-137 in Nuclear Fuel Solutions, and the U concentration is determined by ASTM method E 267, for Determination of Uranium and Plutonium Concentrations and Isotopic Abundances, ASTM method E 318, for Colorimetric Determination of Uranium by Controlled-Potential Coulometry. Calculations are given for correcting the 137 Cs concentration for decay during and after irradiation. The accuracy of this method is limited, not only by the experimental errors with which the fission yield and the half-life of 137 Cs are known

Y-TZP ceramic processing from coprecipitated powders: a comparative study with three commercial dental ceramics.

Science.gov (United States)

Lazar, Dolores R R; Bottino, Marco C; Ozcan, Mutlu; Valandro, Luiz Felipe; Amaral, Regina; Ussui, Valter; Bressiani, Ana H A

2008-12-01

(1) To synthesize 3mol% yttria-stabilized zirconia (3Y-TZP) powders via coprecipitation route, (2) to obtain zirconia ceramic specimens, analyze surface characteristics, and mechanical properties, and (3) to compare the processed material with three reinforced dental ceramics. A coprecipitation route was used to synthesize a 3mol% yttria-stabilized zirconia ceramic processed by uniaxial compaction and pressureless sintering. Commercially available alumina or alumina/zirconia ceramics, namely Procera AllCeram (PA), In-Ceram Zirconia Block (CAZ) and In-Ceram Zirconia (IZ) were chosen for comparison. All specimens (6mmx5mmx5mm) were polished and ultrasonically cleaned. Qualitative phase analysis was performed by XRD and apparent densities were measured on the basis of Archimedes principle. Ceramics were also characterized using SEM, TEM and EDS. The hardness measurements were made employing Vickers hardness test. Fracture toughness (K(IC)) was calculated. Data were analyzed using one-way analysis of variance (ANOVA) and Tukey's test (alpha=0.05). ANOVA revealed that the Vickers hardness (pceramic materials composition. It was confirmed that the PA ceramic was constituted of a rhombohedral alumina matrix, so-called alpha-alumina. Both CAZ and IZ ceramics presented tetragonal zirconia and alpha-alumina mixture of phases. The SEM/EDS analysis confirmed the presence of aluminum in PA ceramic. In the IZ and CAZ ceramics aluminum, zirconium and cerium in grains involved by a second phase containing aluminum, silicon and lanthanum were identified. PA showed significantly higher mean Vickers hardness values (H(V)) (18.4+/-0.5GPa) compared to vitreous CAZ (10.3+/-0.2GPa) and IZ (10.6+/-0.4GPa) ceramics. Experimental Y-TZP showed significantly lower results than that of the other monophased ceramic (PA) (pceramics (pceramic processing conditions led to ceramics with mechanical properties comparable to commercially available reinforced ceramic materials.

Uranium industry annual 1998

International Nuclear Information System (INIS)

1999-01-01

The Uranium Industry Annual 1998 (UIA 1998) provides current statistical data on the US uranium industry's activities relating to uranium raw materials and uranium marketing. It contains data for the period 1989 through 2008 as collected on the Form EIA-858, ''Uranium Industry Annual Survey.'' Data provides a comprehensive statistical characterization of the industry's activities for the survey year and also include some information about industry's plans and commitments for the near-term future. Data on uranium raw materials activities for 1989 through 1998, including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment, are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2008, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, and uranium inventories, are shown in Chapter 2. The methodology used in the 1998 survey, including data edit and analysis, is described in Appendix A. The methodologies for estimation of resources and reserves are described in Appendix B. A list of respondents to the ''Uranium Industry Annual Survey'' is provided in Appendix C. The Form EIA-858 ''Uranium Industry Annual Survey'' is shown in Appendix D. For the readers convenience, metric versions of selected tables from Chapters 1 and 2 are presented in Appendix E along with the standard conversion factors used. A glossary of technical terms is at the end of the report. 24 figs., 56 tabs

Uranium in the rock fragments from Lunar soil

International Nuclear Information System (INIS)

Komarov, A.N.; Sergeev, S.A.

1983-01-01

Uranium content and distribution in Lunar rock fragments 0.4-0.9 mm in size from ''Lunar-16+ -20, -24'' stations were studied by the method of autoradiography. Uranium is almost absent in rock-forming minerals and is concentrated in some accessory mineral. Uranium content in microgabro fragments from ''Lunar-20 and -24'' equals (0.0n - n.0)16 -6 g/g. Variations are not related to fragment representation. Radiogra-- phies of fragments from Lunar soil showed the uranium distribution from uniform (in glasses) to extremely nonuniform in some holocrystalline rocks. It was pointed out, that uranium micro distributions in Lunar and Earth (effusive and magmatic) rocks have common features. In both cases rock-forming minerals don't contain appreciable uranium amount in the form of isomorphic admixture; uranium is highly concentrated in some accessory minerais. The difference lies in tne absence of hydroxyl -containing secondary minerals, which are enriched with uranium on Earth, in Lunar rocks. ''Film'' uranium micromineralization, which occurs in rocks of the Earth along the boundaries of mineral grains is absent in Lunar rocks as well

Uranium recovery from mine water

International Nuclear Information System (INIS)

Sarkar, K.M.

1984-01-01

In many plant trials it has been proven that very small amounts (10 to 20 ppm) of uranium dissolved in mine water can be effectively recovered by the use of ion exchange resins and this uranium recovery has many advantages. In this paper an economic analysis at different levels of uranium contamination and at different market prices of uranium are described. For this study an operating mine-mill complex with a sulphuric acid leach circuit, followed by solvent extraction (SX) process, is considered, where contaminated mine water is available in excess of process requirements. It is further assumed that the sulphuric acid eluant containing uranium would be mixed with the mill pregnant liquor stream that proceeds to the SX plant for final uranium recovery

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Geophysical methods in uranium mining

International Nuclear Information System (INIS)

Koehler, K.

1989-01-01

In uranium prospecting, exploration, milling, and mining there is an urgent need to have information on the concentration of uranium at all steps of handling uranium containing materials. To gain this information in an effective way modern geophysical methods have to be applied. Publications of the IAEA and NEA in this field are reviewed in order to characterize the state of the art of these methods. 55 refs

Novel processing of bioglass ceramics from silicone resins containing micro- and nano-sized oxide particle fillers.

Science.gov (United States)

Fiocco, L; Bernardo, E; Colombo, P; Cacciotti, I; Bianco, A; Bellucci, D; Sola, A; Cannillo, V

2014-08-01

Highly porous scaffolds with composition similar to those of 45S5 and 58S bioglasses were successfully produced by an innovative processing method based on preceramic polymers containing micro- and nano-sized fillers. Silica from the decomposition of the silicone resins reacted with the oxides deriving from the fillers, yielding glass ceramic components after heating at 1000°C. Despite the limited mechanical strength, the obtained samples possessed suitable porous architecture and promising biocompatibility and bioactivity characteristics, as testified by preliminary in vitro tests. © 2013 Wiley Periodicals, Inc.

Microstructure of SiC ceramics fabricated by pyrolysis of electron beam irradiated polycarbomethylsilane containing precursors

International Nuclear Information System (INIS)

Xu Yunshu; Tanaka, Shigeru

2003-01-01

A modified gel-casting method was developed to form the ceramics precursor matrix by using polycarbomehylsilane (PCMS) and SiC powder. The polymer precursor was mixed with SiC powder in toluene, and then the slurry samples were cast into designed shapes. The pre-ceramic samples were then irradiated by 2.0 MeV electron beam generated by a Cockcroft-Walton type accelerator in He gas flow to about 15 MGy. The cured samples were pyrolyzed and sintered into SiC ceramics at 1300degC in Ar gas. The modified gel-casting method leaves almost no internal stress in the pre-ceramic samples, and the electron beam curing not only diminished the amount of pyrolysis gaseous products but also enhanced the interface binding of the polymer converted SiC and the grains of SiC powder. Optical microscope, AFM and SEM detected no visible internal or surface cracks in the final SiC ceramics matrix. A maximum value of 122 MPa of flexural strength of the final SiC ceramics was achieved. (author)

Studies on the fluorination of tri uranium octa oxide to Uranium tetrafluoride

Energy Technology Data Exchange (ETDEWEB)

Rofail, N H; Elfekey, S A [Nuclear chemistry department, hot laboratories centre, atomic energy authority, Cairo, (Egypt)

1995-10-01

Uranium tetrafluoride suitable for both uranium metal and hexafluoride preparations, was prepared by fluorination of U{sub 3} O{sub 8} with C F{sub 2} Cl{sub 2}. It was found that the oct oxide must have certain physical and chemical specifications to satisfy the specifications needed for subsequent operations. X-ray diffraction analysis, infra red investigations and chemical analysis confirm that the obtained uranium tetrafluoride contains more than 97% of U F{sub 4} with tap density equals to 3.5 g/cc. 3 FIGS., 2 TABS.

Development of a pyro-partitioning process for long-lived radioactive nuclides. Process test for pretreatment of simulated high-level waste containing uranium

International Nuclear Information System (INIS)

Kurata, Masateru; Hijikata, Takatoshi; Kinoshita, Kensuke; Inoue, Tadashi

2000-01-01

A pyro-partitioning process developed at CRIEPI requires a pre-treatment process to convert high-level liquid waste to chloride. A combination process of denitration and chlorination has been developed for this purpose. Continuous process tests using simulated high-level waste were performed to certify the applicability of the process. Test results indicated a successful material balance sufficient for satisfying pyro-partitioning process criteria. In the present study, process tests using simulated high-level waste containing uranium were also carried out to prove that the pre-treatment process is feasible for uranium. The results indicated that uranium can be converted to chloride appropriate for the pyro-partitioning process. The material balance obtained from the tests is to be used to revise the process flow diagram. (author)

Recovery and removal of uranium by using plant wastes

International Nuclear Information System (INIS)

Nakajima, Akira; Sakaguchi, Takashi

1990-01-01

The uranium-adsorbing abilities of seven plant wastes were investigated. High abilities to adsorb uranium from non-saline water containing 10 mg dm -3 of uranium were observed with a number of plant wastes tested. However, with seawater supplemented with 10 mg dm -3 of uranium, similar results were found only with chestnut residues. When the plant wastes were immobilized with formaldehyde, their ability to adsorb uranium was increased. Uranium and copper ions were more readily adsorbed by all plant wastes tested than other metal ions from a solution containing a mixture of seven different heavy metals. The selective adsorption of heavy metal ions differs with different species of plant wastes. The immobilization of peanut inner skin, orange peel and grapefruit peel increased the selectivity for uranium. (author)

Synthesis, characterization and structural refinement of polycrystalline uranium substituted zirconolite

International Nuclear Information System (INIS)

Shrivastava, O.P.; Narendra Kumar; Sharma, I.B.

2005-01-01

Ceramic precursors of Zirconolite (CaZrTi 2 O 7 ) family have a remarkable property of substitution Zr 4+ cationic sites. This makes them potential material for nuclear waste management in 'synroc' technology. In order to simulate the mechanism of partial substitution of zirconium by tetravalent actinides, a solid phase of composition CaZr 0.95 U 0.5 Ti 2 O 7 has been synthesized through ceramic route by taking calculated quantities of oxides of Ca, Ti and nitrates of uranium and zirconium respectively. Solid state synthesis has been carried out by repeated pelletizing and sintering the finely powdered oxide mixture in a muffle furnace at 1050 degC. The polycrystalline solid phase has been characterized by its typical powder diffraction pattern. Step analysis data has been used for ab initio calculation of structural parameters. The uranium substituted zirconolite crystallizes in monoclinic symmetry with space group C2/c (15). The following unit cell parameters have been calculated: a =12.4883(15), b =7.2448(5), c 11.3973(10) and β = 100.615(9)0. The structure was refined to satisfactory completion. The Rp and Rwp are found to be 7.48% and 9.74% respectively. (author)

Uranium ore processing

International Nuclear Information System (INIS)

Ritcey, G.M.; Haque, K.E.; Lucas, B.H.; Skeaff, J.M.

1983-01-01

The authors have developed a complete method of recovering separately uranium, thorium and radium from impure solids such as ores, concentrates, calcines or tailings containing these metals. The technique involves leaching, in at least one stage. The impure solids in finely divided form with an aqueous leachant containing HCl and/or Cl 2 until acceptable amounts of uranium, thorium and radium are dissolved. Uranium is recovered from the solution by solvent extraction and precipitation. Thorium may also be recovered in the same manner. Radium may be recovered by at least one ion exchange, absorption and precipitation. This amount of iron in the solution must be controlled before the acid solution may be recycled for the leaching process. The calcine leached in the first step is prepared in a two stage roast in the presence of both Cl 2 and a metal sulfide. The first stage is at 350-450 0 and the second at 550-700 0

Production of uranium metal via electrolytic reduction of uranium oxide in molten LiCl and salt distillation

International Nuclear Information System (INIS)

Eun-Young Choi; Chan Yeon Won; Dae-Seung Kang; Sung-Wook Kim; Ju-Sun Cha; Sung-Jai Lee; Wooshin Park; Hun Suk Im; Jin-Mok Hur

2015-01-01

Recovery of metallic uranium has been achieved by electrolytic reduction of uranium oxide in a molten LiCl-Li 2 O electrolyte at 650 deg C, followed by the removal of the residual salt by vacuum distillation at 850 deg C. Four types of stainless steel mesh baskets, with various mesh sizes (325, 1,400 and 2,300 meshes) and either three or five ply layers, were used both as cathodes and to contain the reduced product in the distillation stage. The recovered uranium had a metal fraction greater than 98.8 % and contained no residual salt. (author)

METHOD OF APPLYING NICKEL COATINGS ON URANIUM

Science.gov (United States)

Gray, A.G.

1959-07-14

A method is presented for protectively coating uranium which comprises etching the uranium in an aqueous etching solution containing chloride ions, electroplating a coating of nickel on the etched uranium and heating the nickel plated uranium by immersion thereof in a molten bath composed of a material selected from the group consisting of sodium chloride, potassium chloride, lithium chloride, and mixtures thereof, maintained at a temperature of between 700 and 800 deg C, for a time sufficient to alloy the nickel and uranium and form an integral protective coating of corrosion-resistant uranium-nickel alloy.

Making Ceramic Cameras

Science.gov (United States)

Squibb, Matt

2009-01-01

This article describes how to make a clay camera. This idea of creating functional cameras from clay allows students to experience ceramics, photography, and painting all in one unit. (Contains 1 resource and 3 online resources.)

Inherently safe in situ uranium recovery

Science.gov (United States)

Krumhansl, James L; Brady, Patrick V

2014-04-29

An in situ recovery of uranium operation involves circulating reactive fluids through an underground uranium deposit. These fluids contain chemicals that dissolve the uranium ore. Uranium is recovered from the fluids after they are pumped back to the surface. Chemicals used to accomplish this include complexing agents that are organic, readily degradable, and/or have a predictable lifetime in an aquifer. Efficiency is increased through development of organic agents targeted to complexing tetravalent uranium rather than hexavalent uranium. The operation provides for in situ immobilization of some oxy-anion pollutants under oxidizing conditions as well as reducing conditions. The operation also artificially reestablishes reducing conditions on the aquifer after uranium recovery is completed. With the ability to have the impacted aquifer reliably remediated, the uranium recovery operation can be considered inherently safe.

Separation of chloride and fluoride from uranium compounds and their determination by ion selective electrodes

International Nuclear Information System (INIS)

Pires, M.A.F.; Abrao, A.

1982-01-01

Fluoride and chloride must be rigorously controlled in uranium compounds, especially in ceramic grade UO 2 . Their determination is very difficult without previous uranium separation, particularly when both are at a low concentration. A simple procedure is described for this separation using a strong cationic resin to retain the uranyl ion. Both anions are determined in the effluent solution. Uranium compounds of nuclear fuel cycle, especially ammonium diuranate, ammonium uranyl tricarbonate, sodium diuranate, uranium trioxide and dioxide and uranium peroxide are dissolved in nitric acid and the solutions are percolated through the resin column. Chloride and fluoride are determined in the effluent by selective electrodes, the detection limits being 0.02 μg F - /ml and 1.0 μg Cl - /ml. The dissolution of the sample, the acidity of the solution, the measurement conditions and the sensitivity of the method are discussed. (Author) [pt

Plutonium recovery from spent reactor fuel by uranium displacement

Science.gov (United States)

Ackerman, J.P.

1992-03-17

A process is described for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

Plutonium recovery from spent reactor fuel by uranium displacement

International Nuclear Information System (INIS)

Ackerman, J.P.

1992-01-01

A process is described for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished

Production of uranium peroxide

International Nuclear Information System (INIS)

Caropreso, F.E.; Kreuz, D.F.

1977-01-01

A process is claimed of recovering uranium values as uranium peroxide from an aqueous uranyl solution containing dissolved vanadium and sodium impurities by treating the uranyl solution with hydrogen peroxide in an amount sufficient to have an excess of at least 0.5 parts H 2 O 2 per part of vanadium (V 2 O 5 ) above the stoichiometric amount required to form the uranium peroxide, the hydrogen peroxide treatment is carried out in three sequential phases consisting of I, a precipitation phase in which the hydrogen peroxide is added to the uranyl solution to precipitate the uranium peroxide and the pH of the reaction medium maintained in the range of 2.5 to 5.5 for a period of from about 1 to 60 minutes after the hydrogen peroxide addition; II, a digestion phase in which the pH of the reaction medium is maintained in the range of 3.0 to 7.0 for a period of about 5 to 180 minutes and III, a final phase in which the pH of the reaction medium is maintained in the range of 4.0 to 7.0 for a period of about 1 to 60 minutes during which time the uranium peroxide is separated from the reaction solution containing the dissolved vanadium and sodium impurities. The excess hydrogen peroxide is maintained during the entire treatment up until the uranium peroxide is separated from the reaction medium

Uranium resources, production and demand 1993

International Nuclear Information System (INIS)

1994-10-01

This book is the Japanese edition of 'Uranium Resources, Production and Demand, 1993' published by OECD/NEA-IAEA in 1994. It contains data on uranium exploration activities, resources and production for about 50 countries. (K.I.)

Speciation of uranium in La{sub 2}Zr{sub 2}O{sub 7} pyrochlore by TRPLS

Energy Technology Data Exchange (ETDEWEB)

Mohapatra, M.; Rajeswari, B.; Hon, N. S.; Kadam, R. M., E-mail: rmkadam@barc.gov.in; Natarajan, V. [Radiochemistry Division, Bhabha Atomic Research Centre, Trombay, Mumbai-400085 (India)

2015-06-24

We discuss the speciation of uranium in lanthanum zirconate (La{sub 2}Zr{sub 2}O{sub 7} =LZO) pyrochlore ceramic prepared via a gel-combustion route. Uranium concentration in the pyrochlore was optimized to 2 mol%. XRD and SEM experiments were carried out to assess the phase and homogeneity of the prepared samples. Time resolved photoluminescence (TRPLS) investigations were carried out for understanding the species stabilized in the pyrochlore host. It was observed that, uranium exists as uranate ion (UO{sub 6}{sup 6−}) in the zirconate host where it replaces the ‘Zr’ ions at its regular site with surrounding defect centers created for charge compensation.

Method for preparation of uranium hydride

International Nuclear Information System (INIS)

Gorski, M.S.; Goncalves, Miriam; Mirage, A.; Lima, W. de.

1985-01-01

A method for preparation of Uranium Hydride starting from Hidrogen and Uranium is described. In the temperature range of 250 0 up to 350 0 C, and pressures above 10torr, Hydrogen reacts smoothly with Uranium turnings forming a fine black or dark gray powder (UH 3 ). Samples containing a significant amount of oxides show a delay before the reaction begging. (Author) [pt

Uranium industry annual 1990, September 1991

International Nuclear Information System (INIS)

1991-01-01

This report presents data on US uranium raw materials and uranium marketing activities of the domestic uranium industry including utilities with nuclear-powered electric generating plants. It contains aggregated data reported by US companies on the ''Uranium Industry Annual Survey'' (1990), Form EIA-858, and historical data from prior data collections and other pertinent sources. The report was prepared by the Energy Information Administration (EIA), the independent agency for data collection and analysis within the US Department of Energy. 19 figs., 47 tabs

PREFACE: 3rd International Congress on Ceramics (ICC3)

Science.gov (United States)

Niihara, Koichi; Ohji, Tatsuki; Sakka, Yoshio

2011-10-01

Early in 2005, the American Ceramic Society, the European Ceramic Society and the Ceramic Society of Japan announced a collaborative effort to provide leadership for the global ceramics community that would facilitate the use of ceramic and glass materials. That effort resulted in an agreement to organize a new biennial series of the International Congress on Ceramics, convened by the International Ceramic Federation (ICF). In order to share ideas and visions of the future for ceramic and glass materials, the 1st International Congress on Ceramics (ICC1) was held in Canada, 2006, under the organization of the American Ceramic Society, and the 2nd Congress (ICC2) was held in Italy, 2008, hosted by the European Ceramic Society. Organized by the Ceramic Society of Japan, the 3rd Congress (ICC3) was held in Osaka, Japan, 14-18 November 2010. Incorporating the 23rd Fall Meeting of the Ceramic Society of Japan and the 20th Iketani Conference, ICC3 was also co-organized by the Iketani Science and Technology Foundation, and was endorsed and supported by ICF, Asia-Oceania Ceramic Federation (AOCF) as well as many other organizations. Following the style of the previous two successful Congresses, the program was designed to advance ceramic and glass technologies to the next generation through discussion of the most recent advances and future perspectives, and to engage the worldwide ceramics community in a collective effort to expand the use of these materials in both conventional as well as new and exciting applications. ICC3 consisted of 22 voluntarily organized symposia in the most topical and essential themes of ceramic and glass materials, including Characterization, design and processing technologies Electro, magnetic and optical ceramics and devices Energy and environment related ceramics and systems Bio-ceramics and bio-technologies Ceramics for advanced industry and safety society Innovation in traditional ceramics It also contained the Plenary Session and the

Source identification of uranium-containing materials at mine legacy sites in Portugal.

Science.gov (United States)

Keatley, A C; Martin, P G; Hallam, K R; Payton, O D; Awbery, R; Carvalho, F P; Oliveira, J M; Silva, L; Malta, M; Scott, T B

2018-03-01

Whilst prior nuclear forensic studies have focused on identifying signatures to distinguish between different uranium deposit types, this paper focuses on providing a scientific basis for source identification of materials from different uranium mine sites within a single region, which can then be potentially used within nuclear forensics. A number of different tools, including gamma spectrometry, alpha spectrometry, mineralogy and major and minor elemental analysis, have been utilised to determine the provenance of uranium mineral samples collected at eight mine sites, located within three different uranium provinces, in Portugal. A radiation survey was initially conducted by foot and/or unmanned aerial vehicle at each site to assist sample collection. The results from each mine site were then compared to determine if individual mine sites could be distinguished based on characteristic elemental and isotopic signatures. Gamma and alpha spectrometry were used to differentiate between samples from different sites and also give an indication of past milling and mining activities. Ore samples from the different mine sites were found to be very similar in terms of gangue and uranium mineralogy. However, rarer minerals or specific impurity elements, such as calcium and copper, did permit some separation of the sites examined. In addition, classification rates using linear discriminant analysis were comparable to those in the literature. Crown Copyright © 2018. Published by Elsevier Ltd. All rights reserved.

Non-polluting treatment of alkaline uranium effluents contaning SO42- ions

International Nuclear Information System (INIS)

Berger, Bernard.

1978-01-01

New non-polluting process for treating uranium effluents from the alkaline digestion of an uranium ore containing sulfur, which makes it possible, on the one hand, to extract uranium and SO 4 2- contained in these effluents allowing the recycling of the sole alkaline carbonates and/or bicarbonates involved, towards the digestion of the ore and on the other hand the separation of the mixture uranium and SO 4 2- ions extracted simultaneously to obtain relatively pure uranium in oxide form [fr

Uranium occurences in calcrete and associated sediments in Western Australia

International Nuclear Information System (INIS)

Butt, C.R.M.; Horwitz, R.C.; Mann, A.W.

1977-10-01

The report is a compilation of data pertaining to the occurence and distribution of uranium mineralization in calcretes and associated sediments in Western Australia and contains brief descriptions of many of the calcrete-uranium occurences, including some of the most minor. Virtually all calcretes in the region are liable to contain traces of uranium mineralization, visible as coatings of carnotite. The locations of the uranium occurences are shown on a map which features the distribution of calcrete

Maintaining the uranium resources data system and assessing the 1991 US uranium potential resources

Energy Technology Data Exchange (ETDEWEB)

McCammon, R.B. (Geological Survey, Reston, VA (United States)); Finch, W.I.; Grundy, W.D.; Pierson, C.T. (Geological Survey, Denver, CO (United States))

1992-12-31

The Energy Information Administration's (EIA) Uranium Resource Assessment Data (URAD) System contains information on potential resources (undiscovered) of uranium in the United States. The purpose of this report is: (1) to describe the work carried out to maintain and update the URAD system; (2) to assess the 1991 U.S. uranium potential resources in various cost categories; and (3) to describe the progress that has been made to automate the generation of the assessment reports and their subsequent transmittal by diskette.

Lithium alkyl anions of uranium(IV) and uranium(V)

International Nuclear Information System (INIS)

Sigurdson, E.R.; Wilkinson, G.

1977-01-01

Organouranium compounds with six or eight uranium-to-carbon sigma-bonds have been synthesized for the first time. The interaction of uranium tetrachloride with lithium alkyls in diethyl ether leads to the isolation of unstable lithium alkyluranate(IV) compounds of stoicheiometry Li 2 UR 6 .8Et 2 0 (R = Me, CH 2 SiMe 3 . Ph, and o-Me 2 NCH 2 C 6 H 4 ). These lithium salts can also be obtained with other donor solvents, such as tetrahydrofuran or NNN'N'-tetramethylethylenediamine. From uranium pentaethoxide similar lithium salts of stoicheiometry Li 3 UR 8 .3 dioxan (R = Me, CH 2 CMe 3 , and CH 2 SiMe 3 ) can be obtained. The interaction of uranium(VI) hexaisopropoxide with lithium, magnesium, or aluminium alkyls does not give compounds containing U-C bonds, but green oils, e.g. U(OPrsup(i)) 6 (MgMe 2 ) 3 , that appear to be adducts in which the oxygen atom of the isopropoxide group bound to uranium is acting as a donor. I.r. and n.m.r. spectroscopy and analytical data for the new compounds are presented. (author)

Uranium accumulation by aquatic macrophyte, Pistia stratiotes

International Nuclear Information System (INIS)

Bhainsa, K.C.; D'Souza, S.F.

2012-01-01

Uranium accumulation by aquatic macrophyte, Pistia stratiotes from aqueous solution was investigated in laboratory condition. The objective was to evaluate the uranium accumulation potential and adopt the plant in uranium containing medium to improve its uptake capacity. The plant was found to tolerate and grow in the pH range of 3-7. Accumulation of uranium improved with increasing pH and the plant could remove 70% uranium from the medium (20 mg/L) within 24 hours of incubation at pH 5-6. Uptake of uranium on either side of this pH range decreased

COGEMA's UMF [Uranium Management Facility

International Nuclear Information System (INIS)

Lamorlette, G.; Bertrand, J.P.

1988-01-01

The French government-owned corporation, COGEMA, is responsible for the nuclear fuel cycle. This paper describes the activities at COGEMA's Pierrelatte facility, especially its Uranium Management Facility. UF6 handling and storage is described for natural, enriched, depleted, and reprocessed uranium. UF6 quality control specifications, sampling, and analysis (halocarbon and volatile fluorides, isotopic analysis, uranium assay, and impurities) are described. In addition, the paper discusses the filling and cleaning of containers and security at UMF

Carbon determination in uranium and its compounds

International Nuclear Information System (INIS)

Silva Queiroz, C.A. da; Abrao, A.

1982-01-01

Carbon content in uranium and its compounds, especially ceramic grade UO 2 , must be controlled rigorously. A method for the determination of carbon with the aid of commercial equipment which uses platinum as a catalyst for the oxidation of CO, and infrared cells for CO 2 measurement is described. The detection limit is 5μg C/g U and the determination range is 0.0005 to 5% C/U. The method is being used routinely to control the carbon content in nuclear fuel materials. (Author) [pt

Synthesis and luminescence properties of glass ceramics containing MSiO3:Eu2+ (M=Ca, Sr, Ba) phosphors for white LED

International Nuclear Information System (INIS)

Cui Zhiguang; Jia Guohua; Deng Degang; Hua Youjie; Zhao Shilong; Huang Lihui; Wang Huanping; Ma Hongping; Xu Shiqing

2012-01-01

Eu 2+ doped silicate glasses were prepared of the system 52SiO 2 -48MO: xEu 2+ (in molar ratio, M=Ca, Sr, Ba; x=1, 3, 5, 7, 9) by a high temperature melt-quenching method in a reducing atmosphere. Glass ceramics containing MSiO 3 :Eu 2+ (M=Ca, Sr, Ba) nano-phosphors were obtained after the heat treatment of the glass samples. The excitation, emission spectra and lifetime decay curves of 4f 6 5d 1 →4f 7 of Eu 2+ were measured and interpreted with respect to their crystal structures and multi-site occupations of divalent europium in the hosts. Their excitation bands mainly extend from 450 to 250 nm, which is adaptable to the main emission region of the UV LED chip. With UV light excitation, the Eu 2+ emission in CaSiO 3 , SrSiO 3 and BaSiO 3 shows blue, green and yellow colors centered at 440, 505 and 555 nm, respectively. The critical Eu 2+ concentration was studied and determined to be x=5 for both CaSiO 3 and SrSiO 3 and x=7 for BaSiO 3 phosphors. The results show that the Eu 2+ doped glass ceramic phosphors containing MSiO 3 (M=Ca, Sr, Ba) nano-crystals can be used as potential matrix materials for a high power white LED pumped by the UV LED chip. - Highlights: → Glass ceramic containing MSiO 3 :Eu 2+ (M=Ca, Sr, Ba) phosphors prepared. → Derived phosphors emit intensively blue, green and yellow colors. → Their luminescence properties and crystal structures have been investigated. → Concentration quenching effects observed and analyzed. → Potential application for UV chip exciting white LED evaluated.

Alloys of uranium and aluminium with low aluminium content; Alliages uranium-aluminium a faible teneur en aluminium

Energy Technology Data Exchange (ETDEWEB)

Cabane, G; Englander, M; Lehmann, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

Uranium, as obtained after spinning in phase {gamma}, presents an heterogeneous structure with large size grains. The anisotropic structure of the metal leads to an important buckling and surface distortion of the fuel slug which is incompatible with its tubular cladding for nuclear fuel uses. Different treatments have been made to obtain an isotropic structure presenting high thermal stability (laminating, hammering and spinning in phase {alpha}) without success. Alloys of uranium and aluminium with low aluminium content present important advantage in respect of non allied uranium. The introduction of aluminium in the form of intermetallic compound (UAl{sub 2}) gives a better resistance to thermal fatigue. Alloys obtained from raw casting present an improved buckling and surface distortion in respect of pure uranium. This improvement is obtained with uranium containing between 0,15 and 0,5 % of aluminium. An even more improvement in thermal stability is obtained by thermal treatments of these alloys. These new characteristics are explained by the fine dispersion of the UAl{sub 2} particles in uranium. The results after treatments obtained from an alloy slug containing 0,4 % of aluminium show no buckling or surface distortion and no elongation. (M.P.)

Fluorometric analysis for uranium in natural waters

International Nuclear Information System (INIS)

Waterbury, G.R.

1977-01-01

A fluorometric method is used for the routine determination of uranium at 0.2 to parts-per-billion (ppB) concentrations in natural surface waters. Duplicate 200-μl aliquots of the water samples are pipetted onto 0.4-g pellets of 98 percent NaF-2 percent LiF flux contained in platinum dishes. The pellets are dried under heat lamps and fused over special propane burners. The fused pellets are subjected to ultraviolet radiation and the fluorescence is measured in a fluorometer. The lower limit of detection is 0.2 ppB of uranium, and the precision is about 15 relative percent in the 0.2 to 10 ppB uranium concentration range. Two analysts determine uranium in 750 to 900 samples per week using this method. Samples containing solids or more than 19 ppB of uranium are analyzed by a delayed neutron counting method

U uranium. Suppl. Vol. D3

International Nuclear Information System (INIS)

Haug, H.O.

1982-01-01

This volume of the uranium series of the Gmelin handbook deals with the anion exchange of uranium. Compounds of the valence states of III, IV, V and VI of uranium in halide, nitrate, sulfate, phosphate, and carbonate media as well as in media containing organic complexing agents are treated. The literature cited covers the period from about 1947 to the end of 1980. (RB) [de

Solubility of airborne uranium samples from uranium processing plant

International Nuclear Information System (INIS)

Kravchik, T.; Oved, S.; Sarah, R.; Gonen, R.; Paz-Tal, O.; Pelled, O.; German, U.; Tshuva, A.

2005-01-01

Full text: During the production and machining processes of uranium metal, aerosols might be released to the air. Inhalation of these aerosols is the main route of internal exposure of workers. To assess the radiation dose from the intake of these uranium compounds it is necessary to know their absorption type, based on their dissolution rate in extracellular aqueous environment of lung fluid. The International Commission on Radiological Protection (ICRP) has assigned UF4 and U03 to absorption type M (blood absorption which contains a 10 % fraction with an absorption rate of 10 minutes and 90 % fraction with an absorption rate of 140 fays) and UO2 and U3O8 to absorption type S (blood absorption rate with a half-time of 7000 days) in the ICRP-66 model.The solubility classification of uranium compounds defined by the ICRP can serve as a general guidance. At specific workplaces, differences can be encountered, because of differences in compounds production process and the presence of additional compounds, with different solubility characteristics. According to ICRP recommendations, material-specific rates of absorption should be preferred to default parameters whenever specific experimental data exists. Solubility profiles of uranium aerosols were determined by performing in vitro chemical solubility tests on air samples taken from uranium production and machining facilities. The dissolution rate was determined over 100 days in a simultant solution of the extracellular airway lining fluid. The filter sample was immersed in a test vial holding 60 ml of simultant fluid, which was maintained at a 37 o C inside a thermostatic bath and at a physiological pH of 7.2-7.6. The test vials with the solution were shaken to simulate the conditions inside the extracellular aqueous environment of the lung as much as possible. The tests indicated that the uranium aerosols samples taken from the metal production and machining facilities at the Nuclear Research Center Negev (NRCN

Contribution to Yttria corrosion study by liquid uranium

International Nuclear Information System (INIS)

Tournier, C.

1995-02-01

We are studying liquid uranium and polycrystalline Yttria interactions under secondary vacuum. The type, morphology and thickness of interfacial reaction products between U and Y 2 O 3 are examined by optical and confocal microscopy, SEM, X ray diffraction, X analysis and XPS. The most important parameters are the stoechiometry and microstructure of the Yttria, the oxygen partial pressure of the furnace atmosphere, pO 2 , and the duration and temperature of experiments. In the thermodynamic modelization, we take into account exchanges at the ceramic/metal interface and exchanges between the molten metal and the furnace atmosphere. Liquid uranium reacts with Yttria to form UO 2 at the interface which gradually changes into a solid solution UO 2 -Y 2 O 3 . The total thickness of reaction products results from two opposing reactions: (i) oxidation of uranium by Yttria (low pO 2 ) or by the atmosphere (high pO 2 ), controlled by migration of oxygen vacancies at Yttria grain boundaries. (ii) deoxidation caused by the formation of volatile uranium monoxide. On the other hand, we observed a transition of the type ''non-wettability → wettability '' which occurs subsequent to an increase of the stoichiometric variation x in Y 2 O 3-x . (author). 69 refs., 76 figs., 30 tabs

Measurement of the radioactive concentration in consumer's goods containing natural uranium and thorium and evaluation of the exposure by their utilization

International Nuclear Information System (INIS)

Yoshida, Masahiro; Satou, Shigerou; Ohhata, Tsutomu; Watanabe, Masatoshi; Ohyama, Ryutaro; Furuya, Hirotaka; Endou, Akira

2005-01-01

A number of consumer's goods which contain natural uranium and thorium are circulated in the familiar living environment. Based on various kinds of information sources, 20 kinds of these consumer's goods were collected and their radioactive concentrations were measured by using ICP-MS and Ge semiconductor detector. As this result, it was found that the concentrations of uranium and thorium in the consumer's goods used at home and industries were below 34 Bq/g and below 270 Bq/g, respectively. Next, the concentrations of daughter nuclides were not so different from the ones of uranium or thorium, which showed that the secular radioactive equilibrium held between both concentrations. In addition, the radiation exposures for public consumer were evaluated when four kinds of typical consumer's goods frequently used in daily life are utilized. The results computed by MCNP-4C code were below 250 μSv/y. (author)

High-performance ceramics - state of the art and trends of development

International Nuclear Information System (INIS)

Gadow, R.; Keizer, K; Burggraaf, A.J.; Boch, P.; Chartier, T.; Thomann, H.

1989-01-01

This paper contains 4 lectures on the following topics: 1. fiber and whisker reinforced ceramics (R. Gadow), 2. ceramic membranes (K. Keizer, A.J. Burggraf), 3. ceramic processing techniques: The case of tape casting (P. Bach, T. Chartier), 4. ceramic superconductors (H. Thomann). Three contributions are separately analyzed for the ENERGIE database. (MM) [de

Stratigraphic implications of uranium deposits

International Nuclear Information System (INIS)

Langford, F.F.

1980-01-01

One of the most consistent characteristics of economic uranium deposits is their restricted stratigraphic distribution. Uraninite deposited with direct igneous affiliation contains thorium, whereas chemical precipitates in sedimentary rocks are characterized by thorium-free primary uranium minerals with vanadium and selenium. In marine sediments, these minerals form low-grade disseminations; but in terrestrial sediments, chiefly fluvial sandstones, the concentration of uranium varies widely, with the high-grade portions constituting ore. Pitchblende vein deposits not only exhibit the same chemical characteristics as the Colorado-type sandstone deposits, but they have a stratigraphically consistent position at unconformities covered by fluvial sandstones. If deposits in such diverse situations have critical features in common, they are likely to have had many features of their origin in common. Thus, vein deposits in Saskatchewan and Australia may have analogues in areas that contain Colorado-type sandstone deposits. In New Mexico, the presence of continental sandstones with peneconformable uranium deposits should also indicate good prospecting ground for unconformity-type vein deposits. All unconformities within the periods of continental deposition ranging from Permian to Cretaceous should have uranium potential. Some situations, such as the onlap of the Abo Formation onto Precambrian basement in the Zuni Mountains, may be directly comparable to Saskatchewan deposition. However, uranium occurrences in the upper part of the Entrada Sandstone suggest that unconformities underlain by sedimentary rocks may also be exploration targets

METHOD OF PROTECTIVELY COATING URANIUM

Science.gov (United States)

Eubank, L.D.; Boller, E.R.

1959-02-01

A method is described for protectively coating uranium with zine comprising cleaning the U for coating by pickling in concentrated HNO/sub 3/, dipping the cleaned U into a bath of molten zinc between 430 to 600 C and containing less than 0 01% each of Fe and Pb, and withdrawing and cooling to solidify the coating. The zinccoated uranium may be given a; econd coating with another metal niore resistant to the corrosive influences particularly concerned. A coating of Pb containing small proportions of Ag or Sn, or Al containing small proportions of Si may be applied over the zinc coatings by dipping in molten baths of these metals.

Plasma sprayed coatings on mild steel split moulds for uranium casting

International Nuclear Information System (INIS)

Sreekumar, K.P.; Padmanaban, P.V.A.; Venkatramani, N.; Singh, S.P.; Saha, D.P.; Date, V.G.

2002-01-01

High velocity high temperature plasma jets are used to deposit metals and ceramics on metallic substrates for oxidation and corrosion protection applications. Plasma sprayed ceramic coatings on metallic substrates are also used to prevent its reaction with molten metals. Metal-alumina duplex coatings on mild steel split moulds have been developed and successfully used for casting of uranium. Techno-economics of the coated moulds against the conventional graphite moulds are a major advantage. Mild steel moulds of 600 mm long and 75 mm in diameter have been plasma spray coated with alumina over a bond coat of molybdenum. In-plant tests showed an increase in number of castings per mould compared to the commonly used graphite moulds. (author)

The electronic conduction of glass and glass ceramics containing various transition metal oxides

International Nuclear Information System (INIS)

Yoshida, T.; Matsuno, Y.

1980-01-01

Nb 2 O 5 -V 2 O 5 -P 2 O 5 glasses containing only Group Va oxides have been investigated to elucidate their electronic conduction and structure, as compared with other glasses obtained by the addition of various transition metal oxides to vanadium phosphate. The P 2 O 5 introduction for Nb 2 O 5 in this glass with the same amount of V 2 O 5 increased the conductivity about two times. Glass ceramics having high conductivity increased by two orders of magnitude and the activation energy for conduction decreased from about 0.5 to 0.2 eV. The crystals were confirmed to be (V,Nb) 2 O 5 and Nb phosphate, one of which was highly conductive and developed a pillar-like shape with a length of more than 20 μm. (orig.)

Method for producing uranium atomic beam source

International Nuclear Information System (INIS)

Krikorian, O.H.

1976-01-01

A method is described for producing a beam of neutral uranium atoms by vaporizing uranium from a compound UM/sub x/ heated to produce U vapor from an M boat or from some other suitable refractory container such as a tungsten boat, where M is a metal whose vapor pressure is negligible compared with that of uranium at the vaporization temperature. The compound, for example, may be the uranium-rhenium compound, URe 2 . An evaporation rate in excess of about 10 times that of conventional uranium beam sources is produced

The Chemistry and Toxicology of Depleted Uranium

OpenAIRE

Sidney A. Katz

2014-01-01

Natural uranium is comprised of three radioactive isotopes: 238U, 235U, and 234U. Depleted uranium (DU) is a byproduct of the processes for the enrichment of the naturally occurring 235U isotope. The world wide stock pile contains some 1½ million tons of depleted uranium. Some of it has been used to dilute weapons grade uranium (~90% 235U) down to reactor grade uranium (~5% 235U), and some of it has been used for heavy tank armor and for the fabrication of armor-piercing bullets and missiles....

Jabiluka gold-uranium project

International Nuclear Information System (INIS)

1988-01-01

The Jabiluka gold-uranium deposit, 230km east of Darwin in the Alligator Rivers Region of the Northern Territory, was discovered by Pancontinental Mining Limited in 1971. Jabiluka, with reserves in excess of 200,000 tonnes of contained U 3 O 8 in two deposits 500 metres apart, is the world's largest high grade uranium deposit and also contains nearly 12 tonnes of gold. It is proposed that only the larger deposit, Jabiluka II will be mined - by underground extraction methods, and that 275,000 tonnes of ore per year will be mined and processed to produce 1,500 tonnes of U 3 O 8 and up to 30,000 oz of gold. The revenue from the uranium sales is estimated to be of the order of A$100 million per year at A$30/lb. By the end of 1982 all necessary mining and environmental approvals had been obtained and significant marketing progress made. With the Australian Labor Party winning Commonwealth Government in the 1983 election, Pancontinental's permission to seek sales contracts was withdrawn and development of the Jabiluka deposit ceased. Jabiluka remains undeveloped - awaiting a change in Australian Government policy on uranium. figs., maps

Zinc- and Y-group-bearing senaite from St Peters Dome, and new data on senaite from Dattas, Minas Gerais, Brazil.

Science.gov (United States)

Foord, E.E.; Sharp, W.N.; Adams, J.W.

1984-01-01

'Mineral Y', an unidentified phase described in association with murataite from a pegmatite in the Pikes Peak granite, El Paso County, Colorado (A.M. 59-172) is now found to be a senaite containing ZnO 7.05% and RE2O3 + Y2O3 5.24%, with the Zn and Y-group REE entering the (Ti,Fe,Mn) position. A Zn-bearing senaite from Dattas, Diamantina, Minas Gerais, has ZnO 7.7%.-R.A.H.

Interfacial characterization of ceramic core materials with veneering porcelain for all-ceramic bi-layered restorative systems.

Science.gov (United States)

Tagmatarchis, Alexander; Tripodakis, Aris-Petros; Filippatos, Gerasimos; Zinelis, Spiros; Eliades, George

2014-01-01

The aim of the study was to characterize the elemental distribution at the interface between all-ceramic core and veneering porcelain materials. Three groups of all-ceramic cores were selected: A) Glass-ceramics (Cergo, IPS Empress, IPS Empress 2, e-max Press, Finesse); B) Glass-infiltrated ceramics (Celay Alumina, Celay Zirconia) and C) Densely sintered ceramics (Cercon, Procera Alumina, ZirCAD, Noritake Zirconia). The cores were combined with compatible veneering porcelains and three flat square test specimens were produced for each system. The core-veneer interfaces were examined by scanning electron microscopy and energy dispersive x-ray microanalysis. The glass-ceramic systems showed interfacial zones reach in Si and O, with the presence of K, Ca, Al in core and Ca, Ce, Na, Mg or Al in veneer material, depending on the system tested. IPS Empress and IPS Empress 2 demonstrated distinct transitional phases at the core-veneer interface. In the glassinfiltrated systems, intermixing of core (Ce, La) with veneer (Na, Si) elements occurred, whereas an abrupt drop of the core-veneer elemental concentration was documented at the interfaces of all densely sintered ceramics. The results of the study provided no evidence of elemental interdiffusion at the core-veneer interfaces in densely sintered ceramics, which implies lack of primary chemical bonding. For the glass-containing systems (glassceramics and glass-infiltrated ceramics) interdiffusion of the glass-phase seems to play a critical role in establishing a primary bonding condition between ceramic core and veneering porcelain.

Management of depleted uranium

International Nuclear Information System (INIS)

2001-01-01

Large stocks of depleted uranium have arisen as a result of enrichment operations, especially in the United States and the Russian Federation. Countries with depleted uranium stocks are interested in assessing strategies for the use and management of depleted uranium. The choice of strategy depends on several factors, including government and business policy, alternative uses available, the economic value of the material, regulatory aspects and disposal options, and international market developments in the nuclear fuel cycle. This report presents the results of a depleted uranium study conducted by an expert group organised jointly by the OECD Nuclear Energy Agency and the International Atomic Energy Agency. It contains information on current inventories of depleted uranium, potential future arisings, long term management alternatives, peaceful use options and country programmes. In addition, it explores ideas for international collaboration and identifies key issues for governments and policy makers to consider. (authors)

U.S. uranium supply outlook

International Nuclear Information System (INIS)

Hogerton, J.F.

1977-01-01

The subject is analysed in the light of figures and forecasts contained in the following diagrams: forecasts of U.S. uranium production, 1977 to 1990; indicated relationship between annual U.S. uranium supply and demand, 1977 to 1986; presently indicated cumulative U.S. uranium supply/demand balance, 1977 to 1990; indicated cumulative U.S. supply/demand balance (shortage or surpluses) 1976 to 1990; presently indicated balance between outstanding U.S. utility procurement needs and uncommitted domestic supply capability 1977 to 1986; projected U.S. uranium requirements in relation to existing supply base and presently indicated additional domestic resource potential, 1977 to 2000. (U.K.)

Liquid membrane process for uranium recovery

International Nuclear Information System (INIS)

Valint, P.L. Jr.

1982-01-01

An improved liquid membrane emulsion extraction process for recovering uranium from a WPPA feed solution containing uranyl cations wherein said feed is contacted with a water-in-oil emulsion which extracts and captures the uranium in the interior aqueous phase thereof, wherein the improvement comprises the presence of an alkane diphosphonic acid uranium complexing agent in the interior phase of the emulsion. This improvement results in greater extraction efficiency

Physico-chemical and radiological characterization of uranium tailings from Tummalapalle uranium mining site

International Nuclear Information System (INIS)

Patra, A.C.; Sahoo, S.K.; Lenka, P.; Gupta, Anil; Jha, S.K.; Tripathi, R.M.; Molla, S.; Rana, B.K.

2018-01-01

Mining of uranium bearing minerals is essential for the extraction of uranium to meet the power requirements of India. Mining and milling activities produce large quantities of low active tailings, as wastes, which are contained in Tailings Ponds. The nature of tailings depends on the mineralogy of ore and host rock and their quantity depends on the configuration of the ore body and mining methods. The mobility of an element from these tailings depends on elemental concentration, pH, particle size, cation exchange capacity, bulk density and porosity of the tailings etc. This necessitates complete characterisation of the tailings. In this paper we aim to characterize the uranium mill tailings generated from Tummalapalle uranium mining facility in Kadappa district, Andhra Pradesh, India

Uranium in Canada: Billion-dollar industry

International Nuclear Information System (INIS)

Whillans, R.T.

1989-01-01

In 1988, Canada maintained its position as the world's leading producer and exporter of uranium; five primary uranium producers reported concentrate output containing 12,400 MT of uranium, or about one-third of Western production. Uranium shipments made by these producers in 1988 exceeded 13,200 MT, worth Canadian $1.1 billion. Because domestic requirements represent only 15% of current Canadian output, most of Canada's uranium production is available for export. Despite continued market uncertainty in 1988, Canada's uranium producers signed new sales contracts for some 14,000 MT, twice the 1987 level. About 90% of this new volume is with the US, now Canada's major uranium customer. The recent implementation of the Canada/US Free Trade agreement brings benefits to both countries; the uranium industries in each can now develop in an orderly, free market. Canada's uranium industry was restructured and consolidated in 1988 through merger and acquisition; three new uranium projects advanced significantly. Canada's new policy on nonresident ownership in the uranium mining sector, designed to encourage both Canadian and foreign investment, should greatly improve efforts to finance the development of recent Canadian uranium discoveries

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

Uranium. Resources, production and demand

International Nuclear Information System (INIS)

1997-01-01

The events characterising the world uranium market in the last several years illustrate the persistent uncertainly faced by uranium producers and consumers worldwide. With world nuclear capacity expanding and uranium production satisfying only about 60 per cent of demand, uranium stockpiles continue to be depleted at a high rate. The uncertainty related to the remaining levels of world uranium stockpiles and to the amount of surplus defence material that will be entering the market makes it difficult to determine when a closer balance between uranium supply and demand will be reached. Information in this report provides insights into changes expected in uranium supply and demand until well into the next century. The 'Red Book', jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency, is the foremost reference on uranium. This world report is based on official information from 59 countries and includes compilations of statistics on resources, exploration, production and demand as of 1 January 1997. It provides substantial new information from all of the major uranium producing centres in Africa, Australia, Eastern Europe, North America and the New Independent States, including the first-ever official reports on uranium production in Estonia, Mongolia, the Russian Federation and Uzbekistan. It also contains an international expert analysis of industry statistics and worldwide projections of nuclear energy growth, uranium requirements and uranium supply

Effect of added zinc on the properties of cobalt-containing ceramic pigments prepared from layered double hydroxides

International Nuclear Information System (INIS)

Perez-Bernal, M.E.; Ruano-Casero, R.J.; Rives, V.

2009-01-01

Layered double hydroxides (LDHs) with the hydrotalcite-type structure containing Co and Al, or Zn, Co and Al in the brucite-like layers and carbonate in the interlayer have been prepared by coprecipitation. The Zn/Co molar ratio was kept to 1 in all samples, while the divalent/trivalent molar ratio was varied from 2/1 to 1/2. The samples have been characterised by element chemical analysis, powder X-ray diffraction, differential thermal and thermogravimetric analysis, temperature-programmed reduction and FT-IR spectroscopy. A single hydrotalcite-like phase is formed for samples with molar ratio 2/1, which crystallinity decreases as the Al content is increased, developing small amounts of diaspore and dawsonite and probably an additional amorphous phase. Calcination at 1200 deg. C in air led to formation of spinels; a small amount of NaAlO 2 was observed in the Al-rich samples, which was removed by washing. The nature of the spinels formed (containing Co II , Co III , Al III and Zn II ) strongly depends on the cations molar ratio in the starting materials and the calcination treatment, leading to a partial oxidation of Co II species to Co III ones. Colour properties (L*a*b*) of the original and calcined solids have been measured. While the original samples show a pink colour (lighter for the series containing Zn), the calcined Co,Al samples show a dark blue colour and the Zn,Co,Al ones a green colour. Changes due to the different molar ratios within a given calcined series are less evident than between samples with the same composition in different series. These calcined materials could be usable as ceramic pigments. - Abstract: Mixed oxides from layered double hydroxides (LDHs) with the hydrotalcite-type structure containing Co and Al or Zn, Co and Al in the brucite-like layers are potential candidates for ceramic pigments with tunable colour properties. Display Omitted

Application of XRF and XRD in the study of ceramics and pottery

International Nuclear Information System (INIS)

Meor Yusoff Meor Sulaiman

2004-01-01

Ceramic artefacts are made from clay-based mineral and their elemental and mineral compositions tend to vary from one locality to another. The elemental and mineral composition data's besides able to verify the originality of the artifact also helps in regard to provenance, fabrication technology and also manufacturing technique. X-ray fluorescence XRF is a non-destructive technique to identify and quantify elements ranging from sodium (atomic number = 11 to uranium atomic number = 92). The paper also looks into recent advances of this technique in the study of ceramics and pottery. Microfocus XRF, besides able to do qualitative and quantitative elemental analysis, it also can perform accurate elemental mapping. Another aspect there is important in this study is the capability to do in-situ analysis. With the recent introduction of the peltiered-cooled silicon detector, in-situ analysis had become more easily available. X-ray diffraction (XRD) analysis on the other hand, helps to identify correctly the different mineral composition present in the ceramic artifact. This could also help in identifying the type of clay that is used in the manufacturing of these ceramic artifacts as well as its origin. Both x-ray techniques complement each other and are very important tool in the archaeological study of ceramic and pottery samples. (Author)

Extraction and desorption of accessible uranium

International Nuclear Information System (INIS)

Payne, T.

1987-01-01

The proportion of the uranium in natural ore samples which is in isotopic equilibrium with the uranium in the groundwater may be designated accessible uranium, and can be regarded as being in short-term exchange with the aqueous phase. Some of the natural uranium is secured in resistant crystalline minerals, and is described as inaccessible, because it may not be brought into solution unless the mineral is subjected to extreme chemical attack. It is not available for groundwater transport in the short term. An estimate of the proportion of accessible uranium is therefore useful when modeling radionuclide migration. The amount of accessible natural uranium is some uranium ore samples from the Ranger deposit has been determined by combining a sequential extraction with isotopic measurements of the extracted phases. The solid samples were crushed drill core form Ranger S1/146 which had previously been used for uranium adsorption experiments and therefore contained 236 U as well as natural uranium. This Section discusses how the uranium partitioning found with the sequential extraction procedure predicts the leaching behavior of these samples

Fluorometric determination of uranium in natural waters

International Nuclear Information System (INIS)

Hues, A.D.; Henicksman, A.L.; Ashley, W.H.; Romero, D.

1977-03-01

Duplicate 200-μl aliquots of the water samples, as received, are transferred by means of Eppendorf pipettors onto 0.4-g pellets of 2 percent LiF-98 percent NaF flux, contained in platinum dishes. The pellets are dried under heat lamps; then fused over special propane burners. The fused pellets are transferred to a Galvanek-Morrison fluorometer, where they are excited with ultraviolet radiation and the fluorescence is measured. The uranium is calculated by comparing the measured fluorescence with that of other pellets, carried through the same procedure, which contain aliquots of standard uranium solutions. The sensitivity of the method is about 0.2 ppB of uranium, and the precision is approximately 15 relative percent in the 0.2- to 10-ppB uranium concentration range

Long-term ecological effects of exposure to uranium

International Nuclear Information System (INIS)

Hanson, W.C.; Miera, F.R. Jr.

1976-03-01

The consequences of releasing natural and depleted uranium to terrestrial ecosystems during development and testing of depleted uranium munitions were investigated. At Eglin Air Force Base, Florida, soil at various distances from armor plate target butts struck by depleted uranium penetrators was sampled. The upper 5 cm of soil at the target bases contained an average of 800 ppM of depleted uranium, about 30 times as much as soil at 5- to 10-cm depth, indicating some vertical movement of depleted uranium. Samples collected beyond about 20 m from the targets showed near-background natural uranium levels, about 1.3 +- 0.3 μg/g or ppM. Two explosives-testing areas at the Los Alamos Scientific Laboratory (LASL) were selected because of their use history. E-F Site soil averaged 2400 ppM of uranium in the upper 5 cm and 1600 ppM at 5-10 cm. Lower Slobovia Site soil from two subplots averaged about 2.5 and 0.6 percent of the E-F Site concentrations. Important uranium concentration differences with depth and distance from detonation points were ascribed to the different explosive tests conducted in each area. E-F Site vegetation samples contained about 320 ppM of uranium in November 1974 and about 125 ppM in June 1975. Small mammals trapped in the study areas in November contained a maximum of 210 ppM of uranium in the gastrointestinal tract contents, 24 ppM in the pelt, and 4 ppM in the remaining carcass. In June, maximum concentrations were 110, 50, and 2 ppM in similar samples and 6 ppM in lungs. These data emphasized the importance of resuspension of respirable particles in the upper few millimeters of soil as a contamination mechanism for several components of the LASL ecosystem

Recovering uranium from phosphoric acid

International Nuclear Information System (INIS)

Anon.

1979-01-01

Wet-process phosphoric acid contains a significant amount of uranium. This uranium totals more than 1,500 tons/yr in current U.S. acid output--and projections put the uranium level at 8,000 tons/yr in the year 2000. Since the phosphoric acid is a major raw material for fertilizers, uranium finds its way into those products and is effectively lost as a resource, while adding to the amount of radioactive material that can contaminate the food chain. So, resource-conservation and environmental considerations both make recovery of the uranium from phosphoric acid desirable. This paper describes the newly developed process for recovering uranium from phosphoric acid by using solvent-extraction technique. After many extractants had been tested, the researchers eventually selected the combination of di (2-ethylhexyl) phosphoric acid (DEPA) and trioctylphosphine oxide (TOPO) as the most suitable. The flowscheme of the process is included

In situ leaching of uranium

International Nuclear Information System (INIS)

Martin, B.

1980-01-01

A process is described for the in-situ leaching of uranium-containing ores employing an acidic leach liquor containing peroxymonosulphuric acid. Preferably, additionally, sulphuric acid is present in the leach liquor. (author)

Synthesis and Structural Studies of Er3+ Containing Lead Cadmium Fluoroborate Glasses and Glass-Ceramics

Directory of Open Access Journals (Sweden)

Silva Maurício A.P.

2002-01-01

Full Text Available The vitreous domain was established in the PbF2-CdF2-B2O 3 system from melting and quenching experiments. Er3+ containing glasses were prepared and glass ceramics were obtained by selected heat-treatments. Lead fluoride was identified (beta-PbF2 as the crystalline phase. Structural studies were performed in some glassy and partially crystallized samples by means of X-ray Diffraction (XRD and Extended X-ray Absorption Fine Structure (EXAFS measurements. The role of Cd2+ and Pb2+ atoms on the glass network formation and also on the crystallization behavior was put forward by these techniques. After crystallization Er3+ atoms segregated in the crystal phase.

Determination of uranium distribution in the evaporation of simulated Savannah River Site waste

International Nuclear Information System (INIS)

Barnes, M.J.; Chandler, G.T.

1995-01-01

The results of an experimental program addressing the distribution of uranium in saltcake and supernate for two Savannah River Site waste compositions are presented. Successive batch evaporations were performed on simulated H-Area Modified Purex low-heat and post-aluminum dissolution wastes spiked with depleted uranium. Waste compositions and physical data were obtained for supernate and saltcake samples. For the H-Area Modified Purex low-heat waste, the product saltcake contained 42% of the total uranium from the original evaporator feed solution. However, precipitated solids only accounted for 10% of the original uranium mass; the interstitial liquid within the saltcake matrix contained the remainder of the uranium. In the case of the simulated post-aluminum dissolution waste; the product saltcake contained 68% of the total uranium from the original evaporator feed solution. Precipitated solids accounted for 52% of the original uranium mass; again, the interstitial liquid within the saltcake matrix contained the remainder of the uranium. An understanding of the distribution of uranium between supernatant liquid, saltcake, and sludge is required to develop a material balance for waste processing operations. This information is necessary to address nuclear criticality safety concerns

Mathematical simulation of the amplification of 1790-nm laser radiation in a nuclear-excited He - Ar plasma containing nanoclusters of uranium compounds

Science.gov (United States)

Kosarev, V. A.; Kuznetsova, E. E.

2014-02-01

The possibility of applying dusty active media in nuclearpumped lasers has been considered. The amplification of 1790-nm radiation in a nuclear-excited dusty He - Ar plasma is studied by mathematical simulation. The influence of nanoclusters on the component composition of the medium and the kinetics of the processes occurring in it is analysed using a specially developed kinetic model, including 72 components and more than 400 reactions. An analysis of the results indicates that amplification can in principle be implemented in an active laser He - Ar medium containing 10-nm nanoclusters of metallic uranium and uranium dioxide.

Thermodynamic properties of uranium in gallium–aluminium based alloys

International Nuclear Information System (INIS)

Volkovich, V.A.; Maltsev, D.S.; Yamshchikov, L.F.; Chukin, A.V.; Smolenski, V.V.; Novoselova, A.V.; Osipenko, A.G.

2015-01-01

Activity, activity coefficients and solubility of uranium was determined in gallium-aluminium alloys containing 1.6 (eutectic), 5 and 20 wt.% aluminium. Additionally, activity of uranium was determined in aluminium and Ga–Al alloys containing 0.014–20 wt.% Al. Experiments were performed up to 1073 K. Intermetallic compounds formed in the alloys were characterized by X-ray diffraction. Partial and excess thermodynamic functions of U in the studied alloys were calculated. - Highlights: • Thermodynamics of uranium is determined in Ga–Al alloys of various compositions. • Uranium in the mixed alloys interacts with both components, Ga and Al. • Interaction of U with Al increases with decreasing temperature. • Activity and solubility of uranium depend on Al content in Ga–Al alloys.

Thermodynamic properties of uranium in gallium–aluminium based alloys

Energy Technology Data Exchange (ETDEWEB)

Volkovich, V.A., E-mail: v.a.volkovich@urfu.ru [Department of Rare Metals and Nanomaterials, Institute of Physics and Technology, Ural Federal University, Ekaterinburg, 620002 (Russian Federation); Maltsev, D.S.; Yamshchikov, L.F. [Department of Rare Metals and Nanomaterials, Institute of Physics and Technology, Ural Federal University, Ekaterinburg, 620002 (Russian Federation); Chukin, A.V. [Department of Theoretical Physics and Applied Mathematics, Institute of Physics and Technology, Ural Federal University, Ekaterinburg, 620002 (Russian Federation); Smolenski, V.V.; Novoselova, A.V. [Institute of High-Temperature Electrochemistry UD RAS, Ekaterinburg, 620137 (Russian Federation); Osipenko, A.G. [JSC “State Scientific Centre - Research Institute of Atomic Reactors”, Dimitrovgrad, 433510 (Russian Federation)

2015-10-15

Activity, activity coefficients and solubility of uranium was determined in gallium-aluminium alloys containing 1.6 (eutectic), 5 and 20 wt.% aluminium. Additionally, activity of uranium was determined in aluminium and Ga–Al alloys containing 0.014–20 wt.% Al. Experiments were performed up to 1073 K. Intermetallic compounds formed in the alloys were characterized by X-ray diffraction. Partial and excess thermodynamic functions of U in the studied alloys were calculated. - Highlights: • Thermodynamics of uranium is determined in Ga–Al alloys of various compositions. • Uranium in the mixed alloys interacts with both components, Ga and Al. • Interaction of U with Al increases with decreasing temperature. • Activity and solubility of uranium depend on Al content in Ga–Al alloys.

Cordierite Glass-Ceramics for Dielectric Materials

International Nuclear Information System (INIS)

Siti Mazatul Azwa Saiyed Mohd Nurddin; Selamat, Malek; Ismail, Abdullah

2007-01-01

The objective of this project is to examine the potential of using Malaysian silica sand deposit as SiO2 raw material in producing cordierite glass-ceramics (2MgO-2Al2O3-5SiO2) for dielectric materials. Upgraded silica sands from Terengganu and ex-mining land in Perak were used in the test-works. The glass batch of the present work has a composition of 45.00% SiO2, 24.00% Al2O3, 15.00% MgO and 8.50% TiO2 as nucleation agent. From the differential thermal analysis results, the crystallization temperature was found to start around 900 deg. C. The glass samples were heat-treated at 900 deg. C and 1000 deg. C. The X-ray diffraction analysis (XRD) results showed glass-ceramics from Terengganu samples containing mainly cordierite and minor β-quartz crystals. However, glass-ceramics from ex-mining land samples contained mainly α-quartz and minor cordierite crystals. Glass-ceramics with different crystal phases exhibit different mechanical, dielectric and thermal properties. Based on the test works, both silica sand deposits, can be potentially used to produce dielectric material component

The precursors effects on biomimetic hydroxyapatite ceramic powders.

Science.gov (United States)

Yoruç, Afife Binnaz Hazar; Aydınoğlu, Aysu

2017-06-01

In this study, effects of the starting material on chemical, physical, and biological properties of biomimetic hydroxyapatite ceramic powders (BHA) were investigated. Characterization and chemical analysis of BHA powders were performed by using XRD, FT-IR, and ICP-AES. Microstructural features such as size and morphology of the resulting BHA powders were characterized by using BET, nano particle sizer, pycnometer, and SEM. Additionally, biological properties of the BHA ceramic powders were also investigated by using water-soluble tetrazolium salts test (WST-1). According to the chemical analysis of BHA ceramic powders, chemical structures of ceramics which are prepared under different conditions and by using different starting materials show differences. Ceramic powders which are produced at 80°C are mainly composed of hydroxyapatite, dental hydroxyapatite (contain Na and Mg elements in addition to Ca), and calcium phosphate sulfide. However, these structures are altered at high temperatures such as 900°C depending on the features of starting materials and form various calcium phosphate ceramics and/or their mixtures such as Na-Mg-hydroxyapatite, hydroxyapatite, Mg-Whitlockit, and chloroapatite. In vitro cytotoxicity studies showed that amorphous ceramics produced at 80°C and ceramics containing chloroapatite structure as main or secondary phases were found to be extremely cytotoxic. Furthermore, cell culture studies showed that highly crystalline pure hydroxyapatite structures were extremely cytotoxic due to their high crystallinity values. Consequently, the current study indicates that the selection of starting materials which can be used in the production of calcium phosphate ceramics is very important. It is possible to produce calcium phosphate ceramics which have sufficient biocompatibility at physiological pH values and by using appropriate starting materials. Copyright © 2017 Elsevier B.V. All rights reserved.

Hinkler Well - Centipede uranium deposits

International Nuclear Information System (INIS)

Crabb, D.; Dudley, R.; Mann, A.W.

1984-01-01

The Hinkler Well - Centipede deposits are near the northeastern margin of the Archean Yilgarn Block on a drainage system entering Lake Way. Basement rocks are granitoids and greenstones. The rocks are deeply weathered and overlain by alluvism. Granitoids, the probable uranium source, currently contain up to 25 ppm uranium, in spite of the weathering. The host calcrete body is 33 km long and 2 km wide. Uranium up to 1000 ppm occurs in carnotite over a 15 km by 2.5 km area. (author)

Multiscale Modeling of Ceramic Matrix Composites

Science.gov (United States)

Bednarcyk, Brett A.; Mital, Subodh K.; Pineda, Evan J.; Arnold, Steven M.

2015-01-01

Results of multiscale modeling simulations of the nonlinear response of SiC/SiC ceramic matrix composites are reported, wherein the microstructure of the ceramic matrix is captured. This micro scale architecture, which contains free Si material as well as the SiC ceramic, is responsible for residual stresses that play an important role in the subsequent thermo-mechanical behavior of the SiC/SiC composite. Using the novel Multiscale Generalized Method of Cells recursive micromechanics theory, the microstructure of the matrix, as well as the microstructure of the composite (fiber and matrix) can be captured.

Microstructures and luminescent properties of Ce-doped transparent mica glass-ceramics

International Nuclear Information System (INIS)

Taruta, Seiichi; Iwasaki, Yoshitomo; Nishikiori, Hiromasa; Yamakami, Tomohiko; Yamaguchi, Tomohiro; Kitajima, Kunio; Okada, Kiyoshi

2012-01-01

Highlights: ► Ce-doped transparent glass-ceramics and their parent glasses. ► TEM and STEM images for the microstructures. ► Each mica crystal did not contain Ce uniformly. ► Emission due to Ce 3+ ions in the glass phase and/or Ce 3+ ions in the mica crystals. - Abstract: Transparent mica glass-ceramics were prepared by heating parent glasses that had been doped with 0.5–15 mol% CeO 2 . During the melting and heat treatment, Ce 4+ ions in the specimens were reduced to Ce 3+ ions, and one or both of these ion species were then replaced with Li + ions in the interlayers of the separated mica crystals. However, scanning transmission electron microscope (STEM) and Z-contrast imaging revealed that the mica crystals did not contain the same amount of Ce. On excitation at 254 nm, the parent glasses and glass-ceramics emitted blue light, which originated from the 5d to 4f transition of the Ce 3+ ions. The emission of the glass-ceramic containing a smaller amount of Ce was attributed to the Ce 3+ ions in both the glass phase and the mica crystals, whereas that of the glass-ceramics containing a larger amount of Ce was caused mainly by Ce 3+ ions in the mica crystals. The dependence of the emission band of the parent glasses on the amount of Ce was a unique feature of the Ce-doped transparent mica glass-ceramics and was not observed in previous studies of Eu-doped parent glasses and mica glass-ceramics.

Exploration for uranium and other nuclear materials

International Nuclear Information System (INIS)

Hernandez, E.C.

1975-05-01

Prospecting and exploration for uranium and other nuclear minerals have one advantage over prospecting for other metals because of their inherent radioactivity. Radioactivity in the earth is not confined solely to these elements but also to radiations coming from cosmic rays and from fallouts from large-scale atomic and nuclear explosions. The primary uranium mineral is uranimite, however, concentrations of other uranium minerals may also lead to an economic deposit. Thorium is about three times more abundant than uranium in the earth's crust. Uranium is practically found in many types of geologic environment it being ubiquitous and very mobile. Uranium deposits are classified in a descriptive manner, owing to lack of basic information as to its origin. These classifications are peneconcordant, for deposits as conglomerates and sandstones, discordant for vein pegmatite and contact metamorphic deposits, concordant for deposits in shales and phosphate rocks, and miscellaneous for deposits in beach and placer sands containing mostly thorium minerals. The different exploration techniques and their associated instrumentations are discussed from a regional scale survey to a detailed survey. To date, only the Larap copper-molybdenum-magnetite deposit at the Paracale district, Camarines Norte in the Philippines, has been found to contain uranium as discrete uraninite grains in the ore mineral assemblage of the deposit

Uranium in Canada

International Nuclear Information System (INIS)

1985-09-01

In 1974 the Minister of Energy, Mines and Resources (EMR) established a Uranium Resource Appraisal Group (URAG) within EMR to audit annually Canada's uranium resources for the purpose of implementing the federal government's uranium export policy. A major objective of this policy was to ensure that Canadian uranium supplies would be sufficient to meet the needs of Canada's nuclear power program. As projections of installed nuclear power growth in Canada over the long term have been successively revised downwards (the concern about domestic security of supply is less relevant now than it was 10 years ago) and as Canadian uranium supply capabilities have expanded significantly. Canada has maintained its status as the western world's leading exporter of uranium and has become the world's leading producer. Domestic uranium resource estimates have increased to 551 000 tonnes U recoverable from mineable ore since URAG completed its last formal assessment (1982). In 1984, Canada's five primary uranium producers employed some 5800 people at their mining and milling operations, and produced concentrates containing some 11 170 tU. It is evident from URAG's 1984 assessment that Canada's known uranium resources, recoverable at uranium prices of $150/kg U or less, are more than sufficient to meet the 30-year fuelling requirements of those reactors that are either in opertaion now or committed or expected to be in-service by 1995. A substantial portion of Canada's identified uranium resources, recoverable within the same price range, is thus surplus to Canadian needs and available for export. Sales worth close to $1 billion annually are assured. Uranium exploration expenditures in Canada in 1983 and 1984 were an estimated $41 million and $35 million, respectively, down markedly from the $128 million reported for 1980. Exploration drilling and surface development drilling in 1983 and 1984 were reported to be 153 000 m and 197 000 m, respectively, some 85% of which was in

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

Plutonium and surrogate fission products in a composite ceramic waste form

International Nuclear Information System (INIS)

Esh, D. W.; Frank, S. M.; Goff, K. M.; Johnson, S. G.; Moschetti, T. L.; O'Holleran, T.

1999-01-01

Argonne National Laboratory is developing a ceramic waste form to immobilize salt containing fission products and transuranic elements. Preliminary results have been presented for ceramic waste forms containing surrogate fission products such as cesium and the lanthanides. In this work results from scanning electron microscopy/energy dispersive spectroscopy and x-ray diffraction are presented in greater detail for ceramic waste forms containing surrogate fission products. Additionally, results for waste forms containing plutonium and surrogate fission products are presented. Most of the surrogate fission products appear to be silicates or aluminosilicates whereas the plutonium is usually found in an oxide form. There is also evidence for the presence of plutonium within the sodalite phase although the chemical speciation of the plutonium is not known

Assessment of nonpoint source chemical loading potential to watersheds containing uranium waste dumps associated with uranium exploration and mining, Browns Hole, Utah

Science.gov (United States)

Marston, Thomas M.; Beisner, Kimberly R.; Naftz, David L.; Snyder, Terry

2012-01-01

During August of 2008, 35 solid-phase samples were collected from abandoned uranium waste dumps, undisturbed geologic background sites, and adjacent streambeds in Browns Hole in southeastern Utah. The objectives of this sampling program were (1) to assess impacts on human health due to exposure to radium, uranium, and thorium during recreational activities on and around uranium waste dumps on Bureau of Land Management lands; (2) to compare concentrations of trace elements associated with mine waste dumps to natural background concentrations; (3) to assess the nonpoint source chemical loading potential to ephemeral and perennial watersheds from uranium waste dumps; and (4) to assess contamination from waste dumps to the local perennial stream water in Muleshoe Creek. Uranium waste dump samples were collected using solid-phase sampling protocols. Solid samples were digested and analyzed for major and trace elements. Analytical values for radium and uranium in digested samples were compared to multiple soil screening levels developed from annual dosage calculations in accordance with the Comprehensive Environmental Response, Compensation, and Liability Act's minimum cleanup guidelines for uranium waste sites. Three occupancy durations for sites were considered: 4.6 days per year, 7.0 days per year, and 14.0 days per year. None of the sites exceeded the radium soil screening level of 96 picocuries per gram, corresponding to a 4.6 days per year exposure. Two sites exceeded the radium soil screening level of 66 picocuries per gram, corresponding to a 7.0 days per year exposure. Seven sites exceeded the radium soil screening level of 33 picocuries per gram, corresponding to a 14.0 days per year exposure. A perennial stream that flows next to the toe of a uranium waste dump was sampled, analyzed for major and trace elements, and compared with existing aquatic-life and drinking-water-quality standards. None of the water-quality standards were exceeded in the stream samples.

Behavior of radioactive elements (uranium and thorium) in Bayer process

International Nuclear Information System (INIS)

Sato, C.; Kazama, S.; Sakamoto, A.; Hirayanagi, K.

1986-01-01

It is essential that alumina used for manufacturing electronic devices should contain an extremely low level of alpha-radiation. The principal source of alpha-radiation in alumina is uranium, a minor source being thorium. Uranium in bauxite dissolves into the liquor in the digestion process and is fixed to the red mud as the desilication reaction progresses. A part of uranium remaining in the liquor precipitates together with aluminum hydroxide in the precipitation process. The uranium content of aluminum hydroxide becomes lower as the precipitation velocity per unit surface area of the seed becomes slower. Organic matters in the Bayer liquor has an extremely significant impact on the uranium content of aluminum hydroxide. Aluminum hydroxide free of uranium is obtainable from the liquor that does not contain organic matters

Radiation risk assessment of reprocessed uranium

International Nuclear Information System (INIS)

Cardenas, Hugo R.; Perez, Aldo E.; Luna, Manuel F.; Becerra, Fabian A.

1999-01-01

Reprocessed uranium contains 232 U, which is not found in nature, as well as 234 U which is present in higher proportion than in natural uranium. Both isotopes modify the radiological properties of the material. The paper evaluates the increase of the internal and external radiation risk on the base of experimental data and theoretical calculations. It also suggests measures to be taken in the production of fuel elements with slightly enriched uranium.The radiation risk of reprocessed uranium is directly proportional to the content of 232 U and 234 U as well as to the aging time of the material

Aftermath of Uranium Ore Processing on Floodplains: Lasting Effects of Uranium on Soil and Microbes

Science.gov (United States)

Tang, H.; Boye, K.; Bargar, J.; Fendorf, S. E.

2016-12-01

A former uranium ore processing site located between the Wind River and the Little Wind River near the city of Riverton, Wyoming, has generated a uranium plume in the groundwater within the floodplain. Uranium is toxic and poses a threat to human health. Thus, controlling and containing the spread of uranium will benefit the human population. The primary source of uranium was removed from the processing site, but a uranium plume still exists in the groundwater. Uranium in its reduced form is relatively insoluble in water and therefore is retained in organic rich, anoxic layers in the subsurface. However, with the aid of microbes uranium becomes soluble in water which could expose people and the environment to this toxin, if it enters the groundwater and ultimately the river. In order to better understand the mechanisms controlling uranium behavior in the floodplains, we examined sediments from three sediment cores (soil surface to aquifer). We determined the soil elemental concentrations and measured microbial activity through the use of several instruments (e.g. Elemental Analyzer, X-ray Fluorescence, MicroResp System). Through the data collected, we aim to obtain a better understanding of how the interaction of geochemical factors and microbial metabolism affect uranium mobility. This knowledge will inform models used to predict uranium behavior in response to land use or climate change in floodplain environments.

Bioassay for uranium mill tailings

International Nuclear Information System (INIS)

Tschaeche, A.N.

1986-01-01

Uranium mill tailings are composed of fine sand that contains, among other things, some uranium (U/sup 238/ primarily), and all of the uranium daughters starting with /sup 230/Th that are left behind after the usable uranium is removed in the milling process. Millions of pounds of tailings are and continue to be generated at uranium mills around the United States. Discrete uranium mill tailings piles exist near the mills. In addition, the tailings materials were used in communities situated near mill sites for such purposes as building materials, foundations for buildings, pipe runs, sand boxes, gardens, etc. The Uranium Mill Tailings Remedial Action Project (UMTRAP) is a U.S. Department of Energy Program designed with the intention of removing or stabilizing the mill tailings piles and the tailings used to communities so that individuals are not exposed above the EPA limits established for such tailings materials. This paper discusses the bioassay programs that are established for workers who remove tailings from the communities in which they are placed

Numerical modelling of evaporation in a ceramic layer in the tape casting process

DEFF Research Database (Denmark)

Jabbaribehnam, Mirmasoud; Jambhekar, V. A.; Hattel, Jesper Henri

2016-01-01

Evaporation of water from a ceramic layer is a key phenomenon in the drying process for the manufacturing of tape cast ceramics. This process contains mass, momentum and energy exchange between the porous medium and the free-flow region. In order to analyze such interaction processes, a Represent......Evaporation of water from a ceramic layer is a key phenomenon in the drying process for the manufacturing of tape cast ceramics. This process contains mass, momentum and energy exchange between the porous medium and the free-flow region. In order to analyze such interaction processes...

Uranium concentration in building materials used in the central region of Egypt

International Nuclear Information System (INIS)

Higgy, R.H.; El-Tahawy, M.S.; Ghods, A.

1997-01-01

Within a radiological survey of the building materials used in the urban dwellings in the central region of Egypt, the uranium concentration in 80 representative samples of raw and fabricated building materials are determined using laser fluorimetry technique. For 40 samples from the studied raw building materials of sand, gravel, gypsum, lime-stone, granite and marble the determined uranium concentration values range between 0.3 and 3.6 ppm for all these samples except for one type of granite having the corresponding value of 7.8 ppm. For 37 samples from studied fabricated building materials of normal cement, clay brick, sand brick, tiles and ceramic plates the determined uranium concentration values range from 0.5 to 3.4 ppm. The corresponding values for three types of iron cement are 3.1, 6.1 and 9.3 ppm. The radium-226 content (of the uranium-238 series) in the same samples was determined using high resolution gamma-ray spectrometers based on HP Ge-detectors. The data obtained by the two techniques are in good agreement for the majority of the studied samples. (author)

Fabrication of uranium-based ceramics using internal gelation for the conversion of trivalent actinides

International Nuclear Information System (INIS)

Daniels, Henrik

2012-01-01

Alternative to today's direct final waste disposal strategy of long-lived radionuclides, for example the minor actinides neptunium, americium, curium and californium, is their selective separation from the radioactive wastestream with subsequent transmutation by neutron irradiation. Hereby it is possible to obtain nuclides with a lower risk-potential concerning their radiotoxicity. 1 neutron irradiation can be carried out either with neutron sources or in the next generation of nuclear reactors. Before the treatment, the minor actinides need to be converted in a suitable chemical and physical form. Internal gelation offers a route through which amorphous gel-spheres can be obtained directly from a metal-salt solution. Due to the presence of different types of metal ions as well as changing pH-values in a stock solution, a complex hydrolysis behaviour of these elements before and during gelation occurs. Therefore, investigations with uranium and neodymium as a minor actinide surrogate were carried out. As a result of suitable gelation-parameters, uraniumneodymium gel-spheres were successfully synthesised. The spheres also stayed intact during the subsequent thermal treatment. Based upon these findings, uranium-plutonium and uranium-americium gels were successfully created. For theses systems, the determined parameters for the uraniumneodymium gelation could also be applied. Additionally, investigations to reduce the acidity of uranium-based stock solutions for internal gelation were carried out. The necessary amount of urea and hexamethylenetetramine to induce gelation could hereby be decreased. This lead to a general increase of the gel quality and made it possible to carry out uranium-americium gelation in the first place. To investigate the stability of urea and hexamethylenetetramine, solutions of these chemicals were irradiated with different radiation doses. These chemicals showed a high stability against radiolysis in aqueous solutions.

Precipitation of uranium peroxide from the leach liquor of uranium ores

International Nuclear Information System (INIS)

Gao Xizhen; Lin Sirong; Guo Erhua; Lu Shijie

1995-06-01

A chemical precipitation process of recovering uranium from the leach liquor of uranium ores was investigated. The process primarily includes the precipitation of iron with lime, the preprocessing of the slurry of iron hydroxides and the precipitation of uranium with H 2 O 2 . The leach liquor is neutralized by lime milk to pH 3.7 to precipitate the iron hydroxides which after flocculation and settle is separated out and preprocessed at 170 degree C in an autoclave. H 2 O 2 is then used to precipitate uranium in the leach liquor free of iron, and the pH of process for uranium precipitation adjusted by adding MgO slurry to 3.5. The barren solution can be used to wash the filter cakes of leach tailing. The precipitated slurry of iron hydroxides after being preprocessed is recycled to leaching processes for recovering uranium in it. This treatment can not only avoid the filtering of the slurry of iron hydroxides, but also prevent the iron precipitate from redissolving and consequently the increase of iron concentration in the leach liquor. The results of the investigation indicate that lime, H 2 O 2 and MgO are the main chemical reagents used to obtain the uranium peroxide product containing over 65% uranium from the leach liquor, and they also do not cause environmental pollution. In accordance with the uranium content in the liquor, the consumption of chemical reagent for H 2 O 2 (30%) and MgO are 0.95 kg/kgU and 0.169 kg/kgU, respectively. (1 fig., 8 tabs., 7 refs.)

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

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

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

Discussion on the interlayer oxidation and uranium metallogenesis in Qianjiadian uranium deposit, Songliao Basin

International Nuclear Information System (INIS)

Pang Yaqing; Chen Xiaolin; Fang Xiheng; Sun Ye

2010-01-01

Through systematic drill core observation, section contrast and analysis,it is proved that the ore-controlling interlayer oxidation zone of Qianjiadian uranium deposit is mainly composed by the red oxidized sandstone and locally distributed yellow and off-white sandstones. The red sandstone contains charcoal fragments, pyrite, ilmenite, siderite, which have been oxidized intensively, and it can be deduced that their original color was gray and became red due to the oxidization. The distribution of the oxidation zone is mainly controlled by the sedimentary facies,which also controll uranium metallization. The uranium orebodies mainly developed in the thinning or pinch parts of the red oxidation zone in section. On the plans, the uranium mineralization distributes near the front of the red interlayer oxidation zone. (authors)

Maintaining the uranium resources data system and assessing the 1991 US uranium potential resources. Final report

Energy Technology Data Exchange (ETDEWEB)

McCammon, R.B. [Geological Survey, Reston, VA (United States); Finch, W.I.; Grundy, W.D.; Pierson, C.T. [Geological Survey, Denver, CO (United States)

1992-12-31

The Energy Information Administration`s (EIA) Uranium Resource Assessment Data (URAD) System contains information on potential resources (undiscovered) of uranium in the United States. The purpose of this report is: (1) to describe the work carried out to maintain and update the URAD system; (2)to assess the 1991 U.S. uranium potential resources in various cost categories; and (3) to describe the progress that has been made to automate the generation of the assessment reports and their subsequent transmittal by diskette.

Geological and geochemical aspects of uranium deposits. A selected, annotated bibliography

Energy Technology Data Exchange (ETDEWEB)

Garland, P.A.; Thomas, J.M.; Brock, M.L.; Daniel, E.W. (comps.)

1980-06-01

A bibliography of 479 references encompassing the fields of uranium and thorium geochemistry and mineralogy, geology of uranium deposits, uranium mining, and uranium exploration techniques has been compiled by the Ecological Sciences Information Center of Oak Ridge National Laboratory. The bibliography was produced for the National Uranium Resource Evaluation Program, which is funded by the Grand Junction Office of the Department of Energy. The references contained in the bibliography have been divided into the following eight subject categories: (1) geology of deposits, (2) geochemistry, (3) genesis O deposits, (4) exploration, (5) mineralogy, (6) uranium industry, (7) reserves and resources, and (8) geology of potential uranium-bearing areas. All categories specifically refer to uranium and thorium; the last category contains basic geologic information concerning areas which the Grand Junction Office feels are particularly favorable for uranium deposition. The references are indexed by author, geographic location, quadrangle name, geoformational feature, taxonomic name, and keyword.

Geological and geochemical aspects of uranium deposits. A selected, annotated bibliography

International Nuclear Information System (INIS)

Garland, P.A.; Thomas, J.M.; Brock, M.L.; Daniel, E.W.

1980-06-01

A bibliography of 479 references encompassing the fields of uranium and thorium geochemistry and mineralogy, geology of uranium deposits, uranium mining, and uranium exploration techniques has been compiled by the Ecological Sciences Information Center of Oak Ridge National Laboratory. The bibliography was produced for the National Uranium Resource Evaluation Program, which is funded by the Grand Junction Office of the Department of Energy. The references contained in the bibliography have been divided into the following eight subject categories: (1) geology of deposits, (2) geochemistry, (3) genesis O deposits, (4) exploration, (5) mineralogy, (6) uranium industry, (7) reserves and resources, and (8) geology of potential uranium-bearing areas. All categories specifically refer to uranium and thorium; the last category contains basic geologic information concerning areas which the Grand Junction Office feels are particularly favorable for uranium deposition. The references are indexed by author, geographic location, quadrangle name, geoformational feature, taxonomic name, and keyword

Uranium supply/demand projections to 2030 in the OECD/NEA-IAEA ''Red Book''. Nuclear growth projections, global uranium exploration, uranium resources, uranium production and production capacity

International Nuclear Information System (INIS)

Vance, Robert

2009-01-01

World demand for electricity is expected to continue to grow rapidly over the next several decades to meet the needs of an increasing population and economic growth. The recognition by many governments that nuclear power can produce competitively priced, base load electricity that is essentially free of greenhouse gas emissions, combined with the role that nuclear can play in enhancing security of energy supplies, has increased the prospects for growth in nuclear generating capacity. Since the mid-1960s, with the co-operation of their member countries and states, the OECD Nuclear Energy Agency (NEA) and the International Atomic Energy Agency (IAEA) have jointly prepared periodic updates (currently every 2 years) on world uranium resources, production and demand. These updates have been published by the OECD/NEA in what is commonly known as the ''Red Book''. The 2007 edition replaces the 2005 edition and reflects information current as of 1 st January 2007. Uranium 2007: Resources, Production and Demand presents, in addition to updated resource figures, the results of a recent review of world uranium market fundamentals and provides a statistical profile of the world uranium industry. It contains official data provided by 40 countries (and one Country Report prepared by the IAEA Secretariat) on uranium exploration, resources, production and reactor-related requirements. Projections of nuclear generating capacity and reactor-related uranium requirements to 2030 as well as a discussion of long-term uranium supply and demand issues are also presented. (orig.)

Method of production of granulates of ceramic nuclear fuels

International Nuclear Information System (INIS)

Wilkinson, W.L.

1975-01-01

To obtain a classified granulate of ceramic nuclear fuels with narrow grain size spectrum, the nuclear fuel powder is made into a slurry in a non-aqueous solvent with a water content as low as possible (e.g. chlorated hydrocarbon), a binder added to it, and spray-dried. The dry granulate desired is already obtained by this working stage. Polybutyl methacrylate in dibutylphthalate is proposed as binder. An example in which uranium dioxide powder is slurried in trichloro-ethylene is described in detail. (UWI/LH) [de

Immobilization of preconditioned spent fuel from nuclear research reactors in a ceramic matrix

International Nuclear Information System (INIS)

Russo, Diego O.; Rodriguez, Diego S.; Heredia, Arturo D.; Sanfilippo, Miguel; Sterba, Mario E.; Mateos, Patricia

2002-01-01

The fuel elements from nuclear research reactors consist in a laminated sandwich of aluminum with a core of some uranium compound. To process this material its necessary to previously eliminate the aluminum covering the fuel, before the conditioning of the rest of the fuel in a stable matrix, in order to obtain an acceptable waste form for a subsequent disposition in a geological repository. Normally, mechanical and chemical methods are proposed for that purpose. One of the most developed techniques for immobilization of the radioactive elements above mentioned, is the vitrification. In this work we propose a method named CERUS (in Spanish Ceramizacion de Elementos Radiactivos con Uranio Sinterizado - Ceramization of radioactive elements with sintered uranium). This is a sinterization of the pre-treated fuel elements mixed with natural uranium oxide. The properties of the blocks obtained are adequate for final disposal in a deep geological reservoir. (author)

Uranium extraction in phosphoric acid

International Nuclear Information System (INIS)

Araujo Figueiredo, C. de

1984-01-01

Uranium is recovered from the phosphoric liquor produced from the concentrate obtained from phosphorus-uraniferous mineral from Itataia mines (CE, Brazil). The proposed process consists of two extraction cycles. In the first one, uranium is reduced to its tetravalent state and then extracted by dioctylpyrophosphoric acid, diluted in Kerosene. Re-extraction is carried out with concentrated phosphoric acid containing an oxidising agent to convert uranium to its hexavalent state. This extract (from the first cycle) is submitted to the second cycle where uranium is extracted with DEPA-TOPO (di-2-hexylphosphoric acid/tri-n-octyl phosphine oxide) in Kerosene. The extract is then washed and uranium is backextracted and precipitated as commercial concentrate. The organic phase is recovered. Results from discontinuous tests were satisfactory, enabling to establish operational conditions for the performance of a continuous test in a micro-pilot plant. (Author) [pt

Fatigue of dental ceramics.

Science.gov (United States)

Zhang, Yu; Sailer, Irena; Lawn, Brian R

2013-12-01

Clinical data on survival rates reveal that all-ceramic dental prostheses are susceptible to fracture from repetitive occlusal loading. The objective of this review is to examine the underlying mechanisms of fatigue in current and future dental ceramics. The nature of various fatigue modes is elucidated using fracture test data on ceramic layer specimens from the dental and biomechanics literature. Failure modes can change over a lifetime, depending on restoration geometry, loading conditions and material properties. Modes that operate in single-cycle loading may be dominated by alternative modes in multi-cycle loading. While post-mortem examination of failed prostheses can determine the sources of certain fractures, the evolution of these fractures en route to failure remains poorly understood. Whereas it is commonly held that loss of load-bearing capacity of dental ceramics in repetitive loading is attributable to chemically assisted 'slow crack growth' in the presence of water, we demonstrate the existence of more deleterious fatigue mechanisms, mechanical rather than chemical in nature. Neglecting to account for mechanical fatigue can lead to gross overestimates in predicted survival rates. Strategies for prolonging the clinical lifetimes of ceramic restorations are proposed based on a crack-containment philosophy. Copyright © 2013 Elsevier Ltd. All rights reserved.

Uranium isotopic signatures measured in samples of dirt collected at two former uranium facilities

International Nuclear Information System (INIS)

Meyers, L.A.; Stalcup, A.M.; LaMont, S.P.; Spitz, H.B.

2014-01-01

Nuclear forensics is a multidisciplinary science that uses a variety of analytical methods and tools to explore the physical, chemical, and isotopic characteristics of nuclear and radiological materials. These characteristics, when evaluated alone or in combination, become signatures that may reveal how and when the material was fabricated. The signatures contained in samples of dirt collected at two different uranium metal processing facilities in the United States were evaluated to determine uranium isotopic composition and compare results with processes that were conducted at these sites. One site refined uranium and fabricated uranium metal ingots for fuel and targets and the other site rolled hot forged uranium and other metals into dimensional rods. Unique signatures were found that are consistent with the activities and processes conducted at each facility and establish confidence in using these characteristics to reveal the provenance of other materials that exhibit similar signatures. (author)

Uranium deposits in Africa

International Nuclear Information System (INIS)

Wilpolt, R.H.; Simov, S.D.

1979-01-01

Africa is not only known for its spectacular diamond, gold, copper, chromium, platinum and phosphorus deposits but also for its uranium deposits. At least two uranium provinces can be distinguished - the southern, with the equatorial sub-province; and the south Saharan province. Uranium deposits are distributed either in cratons or in mobile belts, the first of sandstone and quartz-pebble conglomerate type, while those located in mobile belts are predominantly of vein and similar (disseminated) type. Uranium deposits occur within Precambrian rocks or in younger platform sediments, but close to the exposed Precambrian basement. The Proterozoic host rocks consist of sediments, metamorphics or granitoids. In contrast to Phanerozoic continental uranium-bearing sediments, those in the Precambrian are in marginal marine facies but they do contain organic material. The geology of Africa is briefly reviewed with the emphasis on those features which might control the distribution of uranium. The evolution of the African Platform is considered as a progressive reduction of its craton area which has been affected by three major Precambrian tectonic events. A short survey on the geology of known uranium deposits is made. However, some deposits and occurrences for which little published material is available are treated in more detail. (author)

Uranium chemistry: significant advances

International Nuclear Information System (INIS)

Mazzanti, M.

2011-01-01

The author reviews recent progress in uranium chemistry achieved in CEA laboratories. Like its neighbors in the Mendeleev chart uranium undergoes hydrolysis, oxidation and disproportionation reactions which make the chemistry of these species in water highly complex. The study of the chemistry of uranium in an anhydrous medium has led to correlate the structural and electronic differences observed in the interaction of uranium(III) and the lanthanides(III) with nitrogen or sulfur molecules and the effectiveness of these molecules in An(III)/Ln(III) separation via liquid-liquid extraction. Recent work on the redox reactivity of trivalent uranium U(III) in an organic medium with molecules such as water or an azide ion (N 3 - ) in stoichiometric quantities, led to extremely interesting uranium aggregates particular those involved in actinide migration in the environment or in aggregation problems in the fuel processing cycle. Another significant advance was the discovery of a compound containing the uranyl ion with a degree of oxidation (V) UO 2 + , obtained by oxidation of uranium(III). Recently chemists have succeeded in blocking the disproportionation reaction of uranyl(V) and in stabilizing polymetallic complexes of uranyl(V), opening the way to to a systematic study of the reactivity and the electronic and magnetic properties of uranyl(V) compounds. (A.C.)

Ceramic residue for producing cements, method for the production thereof, and cements containing same

OpenAIRE

Sánchez de Rojas, María Isabel; Frías, Moisés; Asensio, Eloy; Medina Martínez, César

2014-01-01

[EN] The invention relates to a ceramic residue produced from construction and demolition residues, as a puzzolanic component of cements. The invention also relates to a method for producing said ceramic residues and to another method of producing cements using said residues. This type of residue is collected in recycling plants, where it is managed. This invention facilitates a potential commercial launch.

Ageing of low-firing prehistoric ceramics in hydrothermal conditions

Directory of Open Access Journals (Sweden)

Petra Zemenová

2012-03-01

Full Text Available Remains of a prehistoric ceramic object, a moon-shaped idol from the Bronze Age found in archaeological site Zdiby near Prague in the Czech Republic, were studied especially in terms of the firing temperature. Archaeological ceramics was usually fired at temperatures below 1000 °C. It contained unstable non-crystalline products, residua after calcination of clay components of a ceramic material. These products as metakaolinite can undergo a reverse rehydration to a structure close to kaolinite. The aim of this work was to prove whether the identified kaolinite in archaeological ceramics is a product of rehydration. The model compound containing high amount of kaolinite was prepared in order to follow its changes during calcination and hydrothermal treatment. Archaeological ceramics and the model compound were treated by hydrothermal ageing and studied by XRF, XRD and IR analyses. It was proved that the presence of kaolinite in the border-parts of the archaeological object was not a product of rehydration, but that it originated from the raw materials.

Polymer derived non-oxide ceramics modified with late transition metals.

Science.gov (United States)

Zaheer, Muhammad; Schmalz, Thomas; Motz, Günter; Kempe, Rhett

2012-08-07

This tutorial review highlights the methods for the preparation of metal modified precursor derived ceramics (PDCs) and concentrates on the rare non-oxide systems enhanced with late transition metals. In addition to the main synthetic strategies for modified SiC and SiCN ceramics, an overview of the morphologies, structures and compositions of both, ceramic materials and metal (nano) particles, is presented. Potential magnetic and catalytic applications have been discussed for the so manufactured metal containing non-oxide ceramics.

Uranium isotopic composition and uranium concentration in special reference material SRM A (uranium in KCl/LiCl salt matrix)

International Nuclear Information System (INIS)

Graczyk, D.G.; Essling, A.M.; Sabau, C.S.; Smith, F.P.; Bowers, D.L.; Ackerman, J.P.

1997-07-01

To help assure that analysis data of known quality will be produced in support of demonstration programs at the Fuel Conditioning Facility at Argonne National Laboratory-West (Idaho Falls, ID), a special reference material has been prepared and characterized. Designated SRM A, the material consists of individual units of LiCl/KCl eutectic salt containing a nominal concentration of 2.5 wt. % enriched uranium. Analyses were performed at Argonne National Laboratory-East (Argonne, IL) to determine the uniformity of the material and to establish reference values for the uranium concentration and uranium isotopic composition. Ten units from a batch of approximately 190 units were analyzed by the mass spectrometric isotope dilution technique to determine their uranium concentration. These measurements provided a mean value of 2.5058 ± 0.0052 wt. % U, where the uncertainty includes estimated limits to both random and systematic errors that might have affected the measurements. Evidence was found of a small, apparently random, non-uniformity in uranium content of the individual SRM A units, which exhibits a standard deviation of 0.078% of the mean uranium concentration. Isotopic analysis of the uranium from three units, by means of thermal ionization mass spectrometry with a special, internal-standard procedure, indicated that the uranium isotopy is uniform among the pellets with a composition corresponding to 0.1115 ± 0.0006 wt. % 234 U, 19.8336 ± 0.0059 wt. % 235 U, 0.1337 ± 0.0006 wt. % 236 U, and 79.9171 ± 0.0057 wt. % 238 U

Containment of uranium in the proposed Egyptian geologic repository for radioactive waste using hydroxyapatite

International Nuclear Information System (INIS)

Moore, Robert Charles; Hasan, Ahmed Ali Mohamed; Headley, Thomas Jeffrey; Sanchez, Charles Anthony; Zhao, Hongting; Salas, Fred Manuel; Hasan, Mahmoud A.; Holt, Kathleen Caroline

2003-01-01

Currently, the Egyptian Atomic Energy Authority is designing a shallow-land disposal facility for low-level radioactive waste. To insure containment and prevent migration of radionuclides from the site, the use of a reactive backfill material is being considered. One material under consideration is hydroxyapatite, Ca 10 (PO 4 ) 6 (OH) 2 , which has a high affinity for the sorption of many radionuclides. Hydroxyapatite has many properties that make it an ideal material for use as a backfill including low water solubility (K sp > 10 -40 ), high stability under reducing and oxidizing conditions over a wide temperature range, availability, and low cost. However, there is often considerable variation in the properties of apatites depending on source and method of preparation. In this work, we characterized and compared a synthetic hydroxyapatite with hydroxyapatites prepared from cattle bone calcined at 500 C, 700 C, 900 C and 1100 C. The analysis indicated the synthetic hydroxyapatite was similar in morphology to 500 C prepared cattle hydroxyapatite. With increasing calcination temperature the crystallinity and crystal size of the hydroxyapatites increased and the BET surface area and carbonate concentration decreased. Batch sorption experiments were performed to determine the effectiveness of each material to sorb uranium. Sorption of U was strong regardless of apatite type indicating all apatite materials evaluated. Sixty day desorption experiments indicated desorption of uranium for each hydroxyapatite was negligible.

Containment of uranium in the proposed Egyptian geologic repository for radioactive waste using hydroxyapatite

International Nuclear Information System (INIS)

Moore, Robert Charles; Hasan, Ahmed Ali Mohamed; Headley, Thomas Jeffrey; Sanchez, Charles Anthony; Zhao, Hongting; Salas, Fred Manuel; Hasan, Mahmoud A.; Holt, Kathleen Caroline

2004-01-01

Currently, the Egyptian Atomic Energy Authority is designing a shallow-land disposal facility for low-level radioactive waste. To insure containment and prevent migration of radionuclides from the site, the use of a reactive backfill material is being considered. One material under consideration is hydroxyapatite, Ca 10 (PO 4 ) 6 (OH) 2 , which has a high affinity for the sorption of many radionuclides. Hydroxyapatite has many properties that make it an ideal material for use as a backfill including low water solubility (K sp >10 -40 ), high stability under reducing and oxidizing conditions over a wide temperature range, availability, and low cost. However, there is often considerable variation in the properties of apatites depending on source and method of preparation. In this work, we characterized and compared a synthetic hydroxyapatite with hydroxyapatites prepared from cattle bone calcined at 500 C, 700 C, 900 C and 1100 C. The analysis indicated the synthetic hydroxyapatite was similar in morphology to 500 C prepared cattle hydroxyapatite. With increasing calcination temperature the crystallinity and crystal size of the hydroxyapatites increased and the BET surface area and carbonate concentration decreased. Batch sorption experiments were performed to determine the effectiveness of each material to sorb uranium. Sorption of U was strong regardless of apatite type indicating all apatite materials evaluated. Sixty day desorption experiments indicated desorption of uranium for each hydroxyapatite was negligible

Uranium purchases report 1994

International Nuclear Information System (INIS)

1995-07-01

US utilities are required to report to the Secretary of Energy annually the country of origin and the seller of any uranium or enriched uranium purchased or imported into the US, as well as the country of origin and seller of any enrichment services purchased by the utility. This report compiles these data and also contains a glossary of terms and additional purchase information covering average price and contract duration. 3 tabs

Uranium and thorium phosphate based matrices; syntheses, characterizations and lixiviation

International Nuclear Information System (INIS)

Dacheux, N.

1995-03-01

In the framework of the search for a ceramic material usable in the radioactive waste storage, uranium and thorium phosphates have been investigated. Their experimental synthesis conditions have been entirely reviewed, they lead to the preparation of four new compounds: U(UO 2 )(PO 4 ) 2 , U 2 O(PO 4 ) 2 , UC1PO 4 ,H 2 O, and Th 4 (PO 4 ) 4 , U 2 O 3 P 2 O 7 and Th 3 (PO 4 ) 4 . Characterization by several techniques (X-rays and neutron powder diffractions, UV-Visible and Infra-red spectroscopies, XPS,...) were performed. The ab initio structure determination of U(UO 2 )(PO 4 ) 2 has been achieved by X-rays and refined by neutron diffractions. Through its physico-chemical analysis, we found that this compound was a new mixed valence uranium phosphate in which U 4+ and UO 2 2+ ions are ordered in pairs along parallel chains according to a new type of arrangement. Reaction mechanism, starting from UC1PO 4 , 4H 2 O and based on redox processes of uranium in solid state was set up. From two main matrices U(UO 2 )(PO 4 ) 2 and Th 4 (PO 4 ) 4 P 2 O 7 , solid solutions were studied. They consist of replacement of U(IV) by Th(IV) and reversely. The leaching tests on pure, loaded and doped matrices were performed in terms of storage time, pH of solutions, and determined by the use of solids labelled with 230 U or by the measurement of uranyl concentration by Laser-Induced Time-Resolved Spectrofluorometry. Average concentration of uranium in the liquid phase is around 10 -4 M to 10 -6 M. Taking into account the very low solubilities of the studied phosphate ceramics, we estimated their chemical performances promising as an answer to the important nuclear waste problem, if we compare them to the glasses used at the present time. (author). 47 figs., 23 tabs., 6 appendixes

Uranium and thorium based phosphate matrix: synthesis, characterizations and lixiviation

International Nuclear Information System (INIS)

Dacheux, N.

1995-03-01

In the framework of the search for a ceramic material usable in the radioactive waste storage, uranium and thorium phosphates have been investigated. Their experimental synthesis conditions have been entirely reviewed, they lead to the preparation of four new compounds: U(UO 2 )(PO 4 ) 2 , U 2 O(PO 4 ) 2 , UCIPO 4 , 4H 2 O, and Th 4 (PO 4 ) 4 P 2 O 7 . Experimental evidenced are advanced for non existent compounds such as: U 3 (PO 4 ) 4 , U 2 O 3 P 2 O 7 and Th 3 (PO 4 ) 4 . Characterization by several techniques (X-rays and neutron powder diffractions, UV-Visible and Infra-red spectroscopies, XPS,...) were performed. The ab initio structure determination of U(UO 2 )(PO 4 ) 2 has been achieved by X-rays and refined by neutron diffractions. Through its physico-chemical analysis, we found that this compound was a new mixed valence uranium phosphate in which U 4+ and UO 2 2+ ions are ordered in pairs along parallel chains according to a new type of arrangement. Reaction mechanism, starting from UCIPO 4 , 4H 2 O and based on redox processes of uranium in solid state was set up. From two main matrices U(UO 2 )(PO 4 ) 2 and Th 4 (PO 4 ) 4 P 2 O 7 , solid solutions were studied. They consist of replacement of U(IV) by Th(IV) and reversely. The leaching tests on pure, loaded and doped matrices were performed in terms of storage time, pH of solutions, and determined by the use of solids labelled with 230 U or by the measurement of uranyl concentration by Laser-Induced Time-Resolved Spectro-fluorimetry. Average concentration of uranium in the liquid phase is around 10 -4 M to 10 -6 M. Taking into account the very low solubilities of the studied phosphate ceramics, we estimated their chemical performances promising as an answer to the important nuclear waste problem, if we compare them to the glasses used at the present time. (author)

Encapsulation of sacrificial silicon containing particles for SH oxide ceramics via a boehmite precursor route

NARCIS (Netherlands)

Carabat, A.L.; Van der Zwaag, S.; Sloof, W.G.

2013-01-01

Easy crack propagation in oxide ceramic coatings limits their application in high temperature environment (e.g. such as engines and gas turbine components) [1]. In order to overcome this problem, incorporation of sacrificial particles into an oxide ceramic coating may be a viable option. Particles

Continued Multicolumns Bioleaching for Low Grade Uranium Ore at a Certain Uranium Deposit

Directory of Open Access Journals (Sweden)

Gongxin Chen

2016-01-01

Full Text Available Bioleaching has lots of advantages compared with traditional heap leaching. In industry, bioleaching of uranium is still facing many problems such as site space, high cost of production, and limited industrial facilities. In this paper, a continued column bioleaching system has been established for leaching a certain uranium ore which contains high fluoride. The analysis of chemical composition of ore shows that the grade of uranium is 0.208%, which is lower than that of other deposits. However, the fluoride content (1.8% of weight is greater than that of other deposits. This can be toxic for bacteria growth in bioleaching progress. In our continued multicolumns bioleaching experiment, the uranium recovery (89.5% of 4th column is greater than those of other columns in 120 days, as well as the acid consumption (33.6 g/kg. These results indicate that continued multicolumns bioleaching technology is suitable for leaching this type of ore. The uranium concentration of PLS can be effectively improved, where uranium recovery can be enhanced by the iron exchange system. Furthermore, this continued multicolumns bioleaching system can effectively utilize the remaining acid of PLS, which can reduce the sulfuric acid consumption. The cost of production of uranium can be reduced and this benefits the environment too.

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

International Nuclear Information System (INIS)

Worthington, R.E.; Magdics, A.

1987-01-01

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

National Uranium Resource Evaluation. Volume 1. Summary of the geology and uranium potential of Precambrian conglomerates in southeastern Wyoming

Energy Technology Data Exchange (ETDEWEB)

Karlstrom, K.E.; Houston, R.S.; Flurkey, A.J.; Coolidge, C.M.; Kratochvil, A.L.; Sever, C.K.

1981-02-01

A series of uranium-, thorium-, and gold-bearing conglomerates in Late Archean and Early Proterozoic metasedimentary rocks have been discovered in southern Wyoming. The mineral deposits were found by applying the time and strata bound model for the origin of uranium-bearing quartz-pebble conglomerates to favorable rock types within a geologic terrane known from prior regional mapping. No mineral deposits have been discovered that are of current (1981) economic interest, but preliminary resource estimates indicate that over 3418 tons of uranium and over 1996 tons of thorium are present in the Medicine Bow Mountains and that over 440 tons of uranium and 6350 tons of thorium are present in Sierra Madre. Sampling has been inadequate to determine gold resources. High grade uranium deposits have not been detected by work to date but local beds of uranium-bearing conglomerate contain as much as 1380 ppM uranium over a thickness of 0.65 meters. This project has involved geologic mapping at scales from 1/6000 to 1/50,000 detailed sampling, and the evaluation of 48 diamond drill holes, but the area is too large to fully establish the economic potential with the present information. This first volume summarizes the geologic setting and geologic and geochemical characteristics of the uranium-bearing conglomerates. Volume 2 contains supporting geochemical data, lithologic logs from 48 drill holes in Precambrian rocks, and drill site geologic maps and cross-sections from most of the holes. Volume 3 is a geostatistical resource estimate of uranium and thorium in quartz-pebble conglomerates.

National Uranium Resource Evaluation. Volume 1. Summary of the geology and uranium potential of Precambrian conglomerates in southeastern Wyoming

International Nuclear Information System (INIS)

Karlstrom, K.E.; Houston, R.S.; Flurkey, A.J.; Coolidge, C.M.; Kratochvil, A.L.; Sever, C.K.

1981-02-01

A series of uranium-, thorium-, and gold-bearing conglomerates in Late Archean and Early Proterozoic metasedimentary rocks have been discovered in southern Wyoming. The mineral deposits were found by applying the time and strata bound model for the origin of uranium-bearing quartz-pebble conglomerates to favorable rock types within a geologic terrane known from prior regional mapping. No mineral deposits have been discovered that are of current (1981) economic interest, but preliminary resource estimates indicate that over 3418 tons of uranium and over 1996 tons of thorium are present in the Medicine Bow Mountains and that over 440 tons of uranium and 6350 tons of thorium are present in Sierra Madre. Sampling has been inadequate to determine gold resources. High grade uranium deposits have not been detected by work to date but local beds of uranium-bearing conglomerate contain as much as 1380 ppM uranium over a thickness of 0.65 meters. This project has involved geologic mapping at scales from 1/6000 to 1/50,000 detailed sampling, and the evaluation of 48 diamond drill holes, but the area is too large to fully establish the economic potential with the present information. This first volume summarizes the geologic setting and geologic and geochemical characteristics of the uranium-bearing conglomerates. Volume 2 contains supporting geochemical data, lithologic logs from 48 drill holes in Precambrian rocks, and drill site geologic maps and cross-sections from most of the holes. Volume 3 is a geostatistical resource estimate of uranium and thorium in quartz-pebble conglomerates

Method of removing niobium from uranium-niobium alloy

International Nuclear Information System (INIS)

Pollock, E.N.; Schlier, D.S.; Shinopulos, G.

1992-01-01

This patent describes a method of removing niobium from a uranium-niobium alloy. It comprises dissolving the uranium-niobium alloy metal pieces in a first aqueous solution containing an acid selected from the group consisting of hydrochloric acid and sulfuric acid and fluoboric acid as a catalyst to provide a second aqueous solution, which includes uranium (U +4 ), acid radical ions, the acids insolubles including uranium oxides and niobium oxides; adding nitric acid to the insolubles to oxidize the niobium oxides to yield niobic acid and to complete the solubilization of any residual uranium; and separating the niobic acid from the nitric acid and solubilized uranium

Transformations of highly enriched uranium into metal or oxide

International Nuclear Information System (INIS)

Nollet, P.; Sarrat, P.

1964-01-01

The enriched uranium workshops in Cadarache have a double purpose on the one hand to convert uranium hexafluoride into metal or oxide, and on the other hand to recover the uranium contained in scrap materials produced in the different metallurgical transformations. The principles that have been adopted for the design and safety of these workshops are reported. The nuclear safety is based on the geometrical limitations of the processing vessels. To establish the processes and the technology of these workshops, many studies have been made since 1960, some of which have led to original achievements. The uranium hexafluoride of high isotopic enrichment is converted either by injection of the gas into ammonia or by an original process of direct hydrogen reduction to uranium tetrafluoride. The uranium contained m uranium-zirconium metal scrap can be recovered by combustion with hydrogen chloride followed treatment of the uranium chloride by fluorine in order to obtain the uranium in the hexafluoride state. Recovery of the uranium contained m various scrap materials is obtained by a conventional refining process combustion of metallic scrap, nitric acid dissolution of the oxide, solvent purification by tributyl phosphate, ammonium diuranate precipitation, calcining, reduction and hydro fluorination into uranium tetrafluoride, bomb reduction by calcium and slag treatment. Two separate workshops operate along these lines one takes care of the uranium with an isotopic enrichment of up to 3 p. 100, the other handles the high enrichments. The handling of each step of this process, bearing in mind the necessity for nuclear safety, has raised some special technological problems and has led to the conception of new apparatus, in particular the roasting furnace for metal turnings, the nitric acid dissolution unit, the continuous precipitator and ever safe filter and dryer for ammonium diuranate, the reduction and hydro fluorination furnace and the slag recovery apparatus These are

Australian uranium resources

International Nuclear Information System (INIS)

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

1987-01-01

Australia's uranium resources amount to 29% of the WOCA countries (world outside centrally-planned-economies areas) low-cost Reasonably Assured Resources and 28% of the WOCA countries low-cost Estimated Additional Resources. As at 1 January 1986, the Bureau of Mineral Resources estimated Australia's uranium resources as: (1) Cost range to US$80/kg U -Reasonably Assured Resources, 465 000 t U; Estimated Additional Resources, 256 000 t U; (2) Cost range US$80-130/kg U -Reasonably Assured Resources, 56 000 t U; Estimated Additional Resources, 127 000 t U. Most resources are contained in Proterozoic unconformity-related deposits in the Alligator Rivers uranium field in the Northern Territory (Jabiluka, Ranger, Koongarra, Nabarlek deposits) and the Proterozoic stratabound deposit at Olympic Dam on the Stuart Shelf in South Australia

World uranium resources

International Nuclear Information System (INIS)

Deffeyes, K.S.; MacGregor, I.D.

1980-01-01

To estimate the total resource availability of uranium, the authors' approach has been to ask whether the distribution of uranium in the earth's crust can be reasonably approximated by a bell-shaped log-normal curve. In addition they have asked whether the uranium deposits actually mined appear to be a portion of the high-grade tail, or ascending slope, of the distribution. This approach preserves what they feel are the two most important guiding principles of Hubbert's work, for petroleum, namely recognizing the geological framework that contains the deposits of interest and examining the industry's historical record of discovering those deposits. Their findings, published recently in the form of a book-length report prepared for the US Department of Energy, suggest that for uranium the crustal-distribution model and the mining-history model can be brought together in a consistent picture. In brief, they conclude that both sets of data can be described by a single log-normal curve, the smoothly ascending slope of which indicates approximately a 300-fold increase in the amount of uranium recoverable for each tenfold decrease in ore grade. This conclusion has important implications for the future availability of uranium. They hasten to add, however, that this is only an approximative argument; no rigorous statistical basis exists for expecting a log-normal distribution. They continue, pointing out the enormously complex range of geochemical behavior of uranium - and its wide variety of different binds of economic deposit. Their case study, supported by US mining records, indicates that the supply of uranium will not be a limiting factor in the development of nuclear power

77 FR 51579 - Application for a License To Export High-Enriched Uranium

Science.gov (United States)

2012-08-24

... NUCLEAR REGULATORY COMMISSION Application for a License To Export High-Enriched Uranium Pursuant.... Complex, July 30, 2012, August Uranium (93.35%). uranium-235 high-enriched 1, 2012, XSNM3726, 11006037. contained in 7.5 uranium in the kilograms uranium. form of broken metal to the Atomic Energy of Canada...

Effect of ingredients in waste water on property of ion exchange resin for uranium-contained waste water treatment

International Nuclear Information System (INIS)

Ren Junshu; Mu Tao; Zhang Wei; Yang Shengya

2008-01-01

The effect of ingredients in waste water on the property of ion exchange resin for uranium-contained waste water treatment was studied by the method of static ad- sorption combined with dynamic experiment. The experimental result shows that the efficiency or breackthrough volume of resin is reduced if there are other general anions, triethanolamine and oil in the solution. When the concentrations of CO 3 2- , HCO 3 - , SO 3 2- , Cl - in the solution are more than 0.24, 0.28, 0.23 and 0.09 mol/L, respectively, the concentrations of uranium in the outlet waste water will exceed 20 μg/L. The maximal allowable concentration of triethanolamine through the resin is no more than 250 mg/L. When the content of oil in the resin exceeds 1%(by quality), the breackthrough volume reduces by 16%, and when it exceeds 11%, the breackthrough volume almost loses at all. (authors)

International Uranium Resources Evaluation Project (IUREP) national favourability studies: Laos

International Nuclear Information System (INIS)

1977-11-01

Laos is a land locked country containing about 3.5 million people living primarily at a subsistence level. Geologically, the country contains a few places that may be marginally favourable for uranium deposits. A uranium potential in the upper half of Category 1 is assigned. (author)

Transparent ceramic lamp envelope materials

Energy Technology Data Exchange (ETDEWEB)

Wei, G C [OSRAM SYLVANIA, 71 Cherry Hill Drive, Beverly, MA 01915 (United States)

2005-09-07

Transparent ceramic materials with optical qualities comparable to single crystals of similar compositions have been developed in recent years, as a result of the improved understanding of powder-processing-fabrication- sintering-property inter-relationships. These high-temperature materials with a range of thermal and mechanical properties are candidate envelopes for focused-beam, short-arc lamps containing various fills operating at temperatures higher than quartz. This paper reviews the composition, structure and properties of transparent ceramic lamp envelope materials including sapphire, small-grained polycrystalline alumina, aluminium oxynitride, yttrium aluminate garnet, magnesium aluminate spinel and yttria-lanthana. A satisfactory thermal shock resistance is required for the ceramic tube to withstand the rapid heating and cooling cycles encountered in lamps. Thermophysical properties, along with the geometry, size and thickness of a transparent ceramic tube, are important parameters in the assessment of its resistance to fracture arising from thermal stresses in lamps during service. The corrosive nature of lamp-fill liquid and vapour at high temperatures requires that all lamp components be carefully chosen to meet the target life. The wide range of new transparent ceramics represents flexibility in pushing the limit of envelope materials for improved beamer lamps.

Naturally Occurring Radionuclides in Pottery, Ceramic and Glasswares Produced in Bangladesh

International Nuclear Information System (INIS)

Chowdhury, M.I.; Reaz, Rafia; Kamal, M.; Alam, M.N.; Mustafa, M.N.

2005-01-01

The concentrations of naturally occurring radionuclides were measured using gamma spectrometry in the finished products of pottery, glass, ceramic and tiles. Ceramic and pottery utensils, tiles, basin and glassware contained naturally occurring radionuclides. Pottery is produced from local clay materials, but ceramic, tiles, basin and glassware's are made from both local and imported raw materials. Radium and thorium radionuclides are concentrated during the making of pottery from the clay materials due to calcination. Radionuclides concentrated more in the highly calcined pottery products than the low calcined products. Glassware products contained very low quantities of radionuclides comparing with the ceramic and pottery products. Study on radioactivity in the pottery, ceramic and glassware products is important in the assessment of possible radiological hazards to human health. The knowledge is essential for the development of standards and guidelines for the use and management of these materials. (author)

Advantage of uranium contained in low grade dolomite ore

International Nuclear Information System (INIS)

Carneiro, A.L.M.

1988-01-01

The purpose of this work is to investigate a technological route to recover uranium from a lean mineral ore. The experimental work includes studies concerning calcination, carbonate leaching, settling, filtration and resin-ion-exchange. Experimental data confirm the technological feasibility of the proposed process and two different preliminary flowsheets of a pilot plant were suggested. (author) [pt

Interaction at interface between superconducting yttrium ceramics and copper or niobium

International Nuclear Information System (INIS)

Karpov, M.I.; Korzhov, V.P.; Medved', N.V.; Myshlyaeva, M.M.

1992-01-01

Light metallography, scanning electron microscopy and local energy dispersion analysis have been used to study the interaction of Y-ceramics with copper and niobium. Samples in the form of wire of two types were employed, that is, consisting of ceramic core YBaCuO and Cu shell or a ceramic core YBaCuO and bimetallic Cu/Nb shell. The interaction of the ceramics with the shell metal began already at 500 deg with the formation at the interafaces Cu-YBaCuO of oxide layers containing ceramic elements, and in the ceramic core - nonsuperconducting phases. A thin Al-layer placed between the ceramics and the shell appreciably decreased the reactability of the ceramics with respect to copper and niobium

Organic matter and containment of uranium and fissiogenic isotopes at the Oklo natural reactors

International Nuclear Information System (INIS)

Nagy, B.; Rigali, M.J.; Davis, D.W.; Parnell, J.

1991-01-01

Some of the Precambrian natural fission reactors at Oklo in Gabon contain abundant organic matter, part of which was liquefied at the time of criticality and subsequently converted to a graphitic solid. The liquid organic matter helps to reduce U(VI) to U(IV) from aqueous solutions, resulting in the precipitation of uraninite. It is known that in the prevailing reactor environments, precipitated uraninite grains incorporated fission products. We report here observations which show that these uraninite crystals were held immobile within the re-solidified, graphitic bituminous organics at Oklo thus enhanced radionuclide containment. Uraninite encased in solid graphitic matter in the organic-rich reactor zones lost virtually no fissiogenic lanthanide isotopes. The first major episode of uranium and lead migration was caused by the intrusion of a swarm of adjacent dolerite dykes about 1,100 Myr after the reactors went critical. Our results from Oklo imply that the use of organic, hydrophobic solids such as graphitic bitumen as a means of immobilizing radionuclides in pre-treated nuclear waste warrants further investigation. (author)

Facility for continuous CVD coating of ceramic fibers

International Nuclear Information System (INIS)

Moore, A.W.

1992-01-01

The development of new and improved ceramic fibers has spurred the development and application of ceramic composites with improved strength, strength/weight ratio, toughness, and durability at increasingly high temperatures. For many systems, the ceramic fibers can be used without modification because their properties are adequate for the chosen application. However, in order to take maximum advantage of the fiber properties, it is often necessary to coat the ceramic fibers with materials of different composition and properties. Examples include (1) boron nitride coatings on a ceramic fiber, such as Nicalon silicon carbide, to prevent reaction with the ceramic matrix during fabrication and to enhance fiber pullout and increase toughness when the ceramic composite is subjected to stress; (2) boron nitride coatings on ceramic yarns, such as Nicalon for use as thermal insulation panels in an aerodynamic environment, to reduce abrasion of the Nicalon and to inhibit the oxidation of free carbon contained within the Nicalon; and (3) ceramic coatings on carbon yarns and carbon-carbon composites to permit use of these high-strength, high-temperature materials in oxidizing environments at very high temperatures. This paper describes a pilot-plant-sized CVD facility for continuous coating of ceramic fibers and some of the results obtained so far with this equipment

Epidemiological study of workers at risk of internal exposure to uranium

International Nuclear Information System (INIS)

Guseva Canu, I.

2008-09-01

This work is a pilot-study among nuclear fuel cycle workers potentially exposed to alpha radiation. Internal exposure from inhalation of uranium compounds during uranium conversion and enrichment operations was estimated at the AREVA NC Pierrelatte plant. A plant specific semi-quantitative job exposure matrix (JEM) was elaborated for 2709 workers employed at this plant between 1960 and 2006. The JEM has permitted to estimate the exposure to uranium and 16 other categories of pollutants and to calculate individual cumulative exposure score. Numerous correlations were detected between uranium compounds exposure and exposure to other pollutants, such as asbestos, ceramic refractive fibers, TCE and so on. 1968-2005 mortality follow-up showed an increasing risk of mortality from pleural cancer, rectal cancer and lymphoma on the basis of national mortality rates. Analyses of association between cancer mortality and uranium exposure suggested an increase in mortality due to lung cancer among workers exposed to slowly soluble uranium compounds derived from natural and reprocessed uranium. However these results are not statistically significant and based on a small number of observed deaths. These results are concordant with previously reported results from other cohorts of workers potentially exposed to uranium. Experimental studies of biokinetic and action mechanism of slowly soluble uranium oxides bear the biological plausibility of the observed results. Influence of bias was reduced by taking into account of possible confounding including co-exposure to other carcinogenic pollutants and tobacco consumption in the study. Nevertheless, at this stage statistical power of analyses is too limited to obtain more conclusive results. This pilot study shows the interest and feasibility of an epidemiological investigation among workers at risk of internal exposure to uranium and other alpha emitters at the national level. It demonstrates the importance of exposure assessment for

Preparation and characterization of uranium alkoxides through oxidation of uranium metal

International Nuclear Information System (INIS)

Gordon, P.L.; Sauer, N.N.; Burns, C.J.; Watkin, J.G.; Van Der Sluys, W.G.

1993-01-01

Currently the authors are investigating the preparation of halide-containing uranium alkoxides by simultaneous halogen and alcohol oxidation of uranium metal. They recently reported the formation of U 2 I 4 (O-i-Pr) 4 (HO-i-Pr) 2 which upon addition of excess isopropanol forms UI 2 (O-i-Pr) 2 (HO-i-Pr) 2 . They report further characterization and reactivity for this monomeric species. Attempts to prepare similar complexes are being made using chlorine gas in the presence of other alcohols. They describe this ongoing research

Ceramic Nanocomposites from Tailor-Made Preceramic Polymers

Directory of Open Access Journals (Sweden)

Gabriela Mera

2015-04-01

Full Text Available The present Review addresses current developments related to polymer-derived ceramic nanocomposites (PDC-NCs. Different classes of preceramic polymers are briefly introduced and their conversion into ceramic materials with adjustable phase compositions and microstructures is presented. Emphasis is set on discussing the intimate relationship between the chemistry and structural architecture of the precursor and the structural features and properties of the resulting ceramic nanocomposites. Various structural and functional properties of silicon-containing ceramic nanocomposites as well as different preparative strategies to achieve nano-scaled PDC-NC-based ordered structures are highlighted, based on selected ceramic nanocomposite systems. Furthermore, prospective applications of the PDC-NCs such as high-temperature stable materials for thermal protection systems, membranes for hot gas separation purposes, materials for heterogeneous catalysis, nano-confinement materials for hydrogen storage applications as well as anode materials for secondary ion batteries are introduced and discussed in detail.

Ceramic Nanocomposites from Tailor-Made Preceramic Polymers.

Science.gov (United States)

Mera, Gabriela; Gallei, Markus; Bernard, Samuel; Ionescu, Emanuel

2015-04-01

The present Review addresses current developments related to polymer-derived ceramic nanocomposites (PDC-NCs). Different classes of preceramic polymers are briefly introduced and their conversion into ceramic materials with adjustable phase compositions and microstructures is presented. Emphasis is set on discussing the intimate relationship between the chemistry and structural architecture of the precursor and the structural features and properties of the resulting ceramic nanocomposites. Various structural and functional properties of silicon-containing ceramic nanocomposites as well as different preparative strategies to achieve nano-scaled PDC-NC-based ordered structures are highlighted, based on selected ceramic nanocomposite systems. Furthermore, prospective applications of the PDC-NCs such as high-temperature stable materials for thermal protection systems, membranes for hot gas separation purposes, materials for heterogeneous catalysis, nano-confinement materials for hydrogen storage applications as well as anode materials for secondary ion batteries are introduced and discussed in detail.

Ceramic Nanocomposites from Tailor-Made Preceramic Polymers

Science.gov (United States)

Mera, Gabriela; Gallei, Markus; Bernard, Samuel; Ionescu, Emanuel

2015-01-01

The present Review addresses current developments related to polymer-derived ceramic nanocomposites (PDC-NCs). Different classes of preceramic polymers are briefly introduced and their conversion into ceramic materials with adjustable phase compositions and microstructures is presented. Emphasis is set on discussing the intimate relationship between the chemistry and structural architecture of the precursor and the structural features and properties of the resulting ceramic nanocomposites. Various structural and functional properties of silicon-containing ceramic nanocomposites as well as different preparative strategies to achieve nano-scaled PDC-NC-based ordered structures are highlighted, based on selected ceramic nanocomposite systems. Furthermore, prospective applications of the PDC-NCs such as high-temperature stable materials for thermal protection systems, membranes for hot gas separation purposes, materials for heterogeneous catalysis, nano-confinement materials for hydrogen storage applications as well as anode materials for secondary ion batteries are introduced and discussed in detail. PMID:28347023

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

International Nuclear Information System (INIS)

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

1991-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1991-01-01

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

History of the use of uranium

International Nuclear Information System (INIS)

Miettinen, J.K.

1989-01-01

Uranium was found 200 years ago, though the first use for it - in colouring glass yellow, orange or green - was only found 40 years later. When its radioactivity was discovered in 1896, interest in research into uranium increased and for a brief period it was used for improving the ductility of steel. The isolation of radium from uranium ore in 1904 caused a boom for uranium mining for radium. It found use in healing skin cancer, for various 'health' preparations like radon-containing water, and for making self-luminous paints. The discovery of fission 50 years ago increased the use of uranium into large industrial-scale applications. For fission weapons highly enriched U-235 and Pu-239 were needed. Today the main use is for uranium enriched to about 3 per cent U-235 for light water power reactors. Other important uses are for submarines, icebreakers and satellites

Decontamination of uranium-contaminated waste oil using supercritical fluid and nitric acid

International Nuclear Information System (INIS)

Sung, J.; Kim, J.; Lee, Y.; Seol, J.; Ryu, J.; Park, K.

2011-01-01

The waste oil used in nuclear fuel processing is contaminated with uranium because of its contact with materials or environments containing uranium. Under current law, waste oil that has been contaminated with uranium is very difficult to dispose of at a radioactive waste disposal site. To dispose of the uranium-contaminated waste oil, the uranium was separated from the contaminated waste oil. Supercritical R-22 is an excellent solvent for extracting clean oil from uranium-contaminated waste oil. The critical temperature of R-22 is 96.15 deg. C and the critical pressure is 49.9 bar. In this study, a process to remove uranium from the uranium-contaminated waste oil using supercritical R-22 was developed. The waste oil has a small amount of additives containing N, S or P, such as amines, dithiocarbamates and dialkyldithiophosphates. It seems that these organic additives form uranium-combined compounds. For this reason, dissolution of uranium from the uranium-combined compounds using nitric acid was needed. The efficiency of the removal of uranium from the uranium-contaminated waste oil using supercritical R-22 extraction and nitric acid treatment was determined. (authors)

Sorption of Uranium Ions from Their Aqueous Solution by Resins Containing Nanomagnetite Particles

Directory of Open Access Journals (Sweden)

Mahmoud O. Abd El-Magied

2016-01-01

Full Text Available Magnetic amine resins composed of nanomagnetite (Fe3O4 core and glycidyl methacrylate (GMA/N,N′-methylenebisacrylamide (MBA shell were prepared by suspension polymerization of glycidyl methacrylate with N,N′-methylenebisacrylamide in the presence of nanomagnetite particles and immobilized with different amine ligands. These resins showed good magnetic properties and could be easily retrieved from their suspensions using an external magnetic field. Adsorption behaviors of uranium ions on the prepared resins were studied. Maximum sorption capacities of uranium ions on R-1 and R-2 were found to be 92 and 158 mg/g. Uranium was extracted successfully from three granite samples collected from Gabal Gattar pluton, North Eastern Desert, Egypt. The studied resins showed good durability and regeneration using HNO3.

Uranium resources in New Mexico

International Nuclear Information System (INIS)

McLemore, V.T.; Chenoweth, W.L.

1989-01-01

For nearly three decades (1951-1980), the Grants uranium district in northwestern New Mexico produced more uranium than any other district in the world. The most important host rocks containing economic uranium deposits in New Mexico are sandstones within the Jurassic Morrison Formation. Approximately 334,506,000 lb of U 3 O 8 were produced from this unit from 1948 through 1987, accounting for 38% of the total uranium production from the US. All of the economic reserves and most of the resources in New Mexico occur in the Morrison Formation. Uranium deposits also occur in sandstones of Paleozoic, Triassic, Cretaceous, Tertiary, and Quaternary formations; however, only 468,680 lb of U 3 O 8 or 0.14% of the total production from New Mexico have been produced from these deposits. Some of these deposits may have a high resource potential. In contrast, almost 6.7 million lb of U 3 O 8 have been produced from uranium deposits in the Todilto Limestone of the Wanakah Formation (Jurassic), but potential for finding additional economic uranium deposits in the near future is low. Other uranium deposits in New Mexico include those in other sedimentary rocks, vein-type uranium deposits, and disseminated magmatic, pegmatitic, and contact metasomatic uranium deposits in igneous and metamorphic rocks. Production from these deposits have been insignificant (less than 0.08% of the total production from New Mexico), but there could be potential for medium to high-grade, medium-sized uranium deposits in some areas. Total uranium production from New Mexico from 1948 to 1987 amounts to approximately 341,808,000 lb of U 3 O 8 . New Mexico has significant uranium reserves and resources. Future development of these deposits will depend upon an increase in price for uranium and lowering of production costs, perhaps by in-situ leaching techniques

Uranium and plutonium in marine sediments

International Nuclear Information System (INIS)

Ordonez R, E.; Almazan T, M. G.; Ruiz F, A. C.

2011-11-01

The marine sediments contain uranium concentrations that are considered normal, since the seawater contains dissolved natural uranium that is deposited in the bed sea in form of sediments by physical-chemistry and bio-genetics processes. Since the natural uranium is constituted of several isotopes, the analysis of the isotopic relationship 234 U/ 238 U are an indicator of the oceanic activity that goes accumulating slowly leaving a historical registration of the marine events through the profile of the marine soil. But the uranium is not the only radioelement present in the marine sediments. In the most superficial strata the presence of the 239+140 Pu has been detected that it is an alpha emitter and that recently it has been detected with more frequency in some coasts of the world. The Mexican coast has not been the exception to this phenomenon and in this work the presence of 239-140 Pu is shown in the more superficial layers of an exploring coming from the Gulf of Tehuantepec. (Author)

Contribution to Yttria corrosion study by liquid uranium; Contribution a l`etude de la corrosion de l`yttria par l`uranium liquide

Energy Technology Data Exchange (ETDEWEB)

Tournier, C

1995-02-01

We are studying liquid uranium and polycrystalline Yttria interactions under secondary vacuum. The type, morphology and thickness of interfacial reaction products between U and Y{sub 2}O{sub 3} are examined by optical and confocal microscopy, SEM, X ray diffraction, X analysis and XPS. The most important parameters are the stoechiometry and microstructure of the Yttria, the oxygen partial pressure of the furnace atmosphere, pO{sub 2}, and the duration and temperature of experiments. In the thermodynamic modelization, we take into account exchanges at the ceramic/metal interface and exchanges between the molten metal and the furnace atmosphere. Liquid uranium reacts with Yttria to form UO{sub 2} at the interface which gradually changes into a solid solution UO{sub 2}-Y{sub 2}O{sub 3}. The total thickness of reaction products results from two opposing reactions: (i) oxidation of uranium by Yttria (low pO{sub 2}) or by the atmosphere (high pO{sub 2}), controlled by migration of oxygen vacancies at Yttria grain boundaries. (ii) deoxidation caused by the formation of volatile uranium monoxide. On the other hand, we observed a transition of the type ``non-wettability {yields} wettability `` which occurs subsequent to an increase of the stoichiometric variation x in Y{sub 2}O{sub 3-x}. (author). 69 refs., 76 figs., 30 tabs.

International Uranium Resources Evaluation Project (IUREP) national favourability studies: Belgium

International Nuclear Information System (INIS)

1977-12-01

Uranium occurrences and resources - To date the uranium identified in Belgium is limited to a number of occurrences and none of these have as yet proved significant from a reserve or resource viewpoint. The main uranium occurrences ares (1) In the Upper Cambrian graphite schists corresponding to the culm of Sweden small zones are found (30 - 50 cm thick) with an average of 20 ppm uranium. (2) Near Vise at the base of the Carboniferous the Visean formation is discordantly superimposed on the Permian (Frasnian) and overlain by shales and phyllites. Solution pockets at the boundary contain phosphatic lenses that contain uranium values of up to 200 ppm. Autunite and Torbernite are the main uranium minerals associated with a number of complex phosphatic minerals. Within the Chalk (Maestrichtien) of the Mons basin, that is mainly in the Ciply - St. Symphorien and Baudow district. Here is found enrichment of uranium up to 140 ppm over large areas related to phosphatic chalk. The thickness of the zone varies from a few to 20 metres. However, as the P 2 O 5 content is not high enough for the deposits to be exploited at present for phosphate there is little possibility of the uranium being concentrated at high enough levels to be exploited for itself alone. (4) Near to Vielsalm (in the Stavelot Massif) are some thin quartz veins containing small amounts of copper and uranium minerals (Torbornite). Values of up to 70 ppm are recorded. (5) A number of low uranium values are recorded associated with phosphatic nodules and zones in the Lower Pleistocene and Tertiary