Natural uranium metallic fuel elements: fabrication and operating experience

International Nuclear Information System (INIS)

Hammad, F.H.; Abou-Zahra, A.A.; Sharkawy, S.W.

1980-01-01

The main reactor types based on natural uranium metallic fuel element, particularly the early types, are reviewed in this report. The reactor types are: graphite moderated air cooled, graphite moderated gas cooled and heavy water moderated reactors. The design features, fabrication technology of these reactor fuel elements and the operating experience gained during reactor operation are described and discussed. The interrelation between operating experience, fuel design and fabrication was also discussed with emphasis on improving fuel performance. (author)

Fabrication of uranium carbide/beryllium carbide/graphite experimental-fuel-element specimens

International Nuclear Information System (INIS)

Muenzer, W.A.

1978-01-01

A method has been developed for fabricating uranium carbide/beryllium carbide/graphite fuel-element specimens for reactor-core-meltdown studies. The method involves milling and blending the raw materials and densifying the resulting blend by conventional graphite-die hot-pressing techniques. It can be used to fabricate specimens with good physical integrity and material dispersion, with densities of greater than 90% of the theoretical density, and with a uranium carbide particle size of less than 10 μm

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.

Monte Carlo calculations of fast effects in uranium graphite lattices

International Nuclear Information System (INIS)

Beardwood, J.E.; Tyror, J.G.

1962-12-01

Details are given of the results of a series of computations of fast neutron effects in natural uranium metal/graphite cells. The computations were performed using the Monte Carlo code SPEC. It is shown that neutron capture in U238 is conveniently discussed in terms of a capture escape probability ζ as well as the conventional probability p. The latter is associated with the slowing down flux and has the classical exponential dependence on fuel-to-moderator volume ratio whilst the former is identified with the component of neutron flux above 1/E. (author)

Thermodynamic Simulation of Equilibrium Composition of Reaction Products at Dehydration of a Technological Channel in a Uranium-Graphite Reactor

Science.gov (United States)

Pavliuk, A. O.; Zagumennov, V. S.; Kotlyarevskiy, S. G.; Bespala, E. V.

2018-01-01

The problems of accumulation of nuclear fuel spills in the graphite stack in the course of operation of uranium-graphite nuclear reactors are considered. The results of thermodynamic analysis of the processes in the graphite stack at dehydration of a technological channel, fuel element shell unsealing and migration of fission products, and activation of stable nuclides in structural elements of the reactor and actinides inside the graphite moderator are given. The main chemical reactions and compounds that are produced in these modes in the reactor channel during its operation and that may be hazardous after its shutdown and decommissioning are presented. Thermodynamic simulation of the equilibrium composition is performed using the specialized code TERRA. The results of thermodynamic simulation of the equilibrium composition in different cases of technological channel dehydration in the course of the reactor operation show that, if the temperature inside the active core of the nuclear reactor increases to the melting temperature of the fuel element, oxides and carbides of nuclear fuel are produced. The mathematical model of the nonstationary heat transfer in a graphite stack of a uranium-graphite reactor in the case of the technological channel dehydration is presented. The results of calculated temperature evolution at the center of the fuel element, the replaceable graphite element, the air gap, and in the surface layer of the block graphite are given. The numerical results show that, in the case of dehydration of the technological channel in the uranium-graphite reactor with metallic uranium, the main reaction product is uranium dioxide UO2 in the condensed phase. Low probability of production of pyrophoric uranium compounds (UH3) in the graphite stack is proven, which allows one to disassemble the graphite stack without the risk of spontaneous graphite ignition in the course of decommissioning of the uranium-graphite nuclear reactor.

Analysis of fuel cycles with natural uranium; Analiza gorivnih ciklusa sa prirodnim uranom

Energy Technology Data Exchange (ETDEWEB)

Stojanovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-05-15

A method was developed and a computer code was written for analysis of fuel cycles and it was applied for heavy water and graphite moderated power reactors. Among a variety of possibilities, three methods which enable best utilization of natural uranium and plutonium production were analyzed. Analysis has shown that reprocessing of irradiated uranium and plutonium utilization in the same or similar type of reactor could increase significantly utilization of natural uranium. Increase of burnup is limited exclusively by costs of reprocessing, plutonium extraction and fabrication of new fuel elements.

Purification and preparation of graphite oxide from natural graphite

Energy Technology Data Exchange (ETDEWEB)

Panatarani, C., E-mail: c.panatarani@phys.unpad.ac.id; Muthahhari, N.; Joni, I. Made [Instrumentation Systems and Functional Material Processing Laboratory, Department of Physics, Faculty of Mathematics and Natural Sciences, Universitas Padjadjaran, Padjadjaran University, Jl. Raya Bandung-Sumedang KM 21, Jatinangor, 45363, Jawa Barat (Indonesia); Rianto, Anton [Grafindo Nusantara Ltd., Belagio Mall Lantai 2, Unit 0 L3-19, Kawasan Mega Kuningan, Kav. B4 No.3, Jakarta Selatan (Indonesia)

2016-03-11

Graphite oxide has attracted much interest as a possible route for preparation of natural graphite in the large-scale production and manipulation of graphene as a material with extraordinary electronic properties. Graphite oxide was prepared by modified Hummers method from purified natural graphite sample from West Kalimantan. We demonstrated that natural graphite is well-purified by acid leaching method. The purified graphite was proceed for intercalating process by modifying Hummers method. The modification is on the reaction time and temperature of the intercalation process. The materials used in the intercalating process are H{sub 2}SO{sub 4} and KMNO{sub 4}. The purified natural graphite is analyzed by carbon content based on Loss on Ignition test. The thermo gravimetricanalysis and the Fouriertransform infrared spectroscopy are performed to investigate the oxidation results of the obtained GO which is indicated by the existence of functional groups. In addition, the X-ray diffraction and energy dispersive X-ray spectroscopy are also applied to characterize respectively for the crystal structure and elemental analysis. The results confirmed that natural graphite samples with 68% carbon content was purified into 97.68 % carbon content. While the intercalation process formed a formation of functional groups in the obtained GO. The results show that the temperature and reaction times have improved the efficiency of the oxidation process. It is concluded that these method could be considered as an important route for large-scale production of graphene.

Fine structure and spectral index measurements in natural uranium - graphite lattices; Mesures fines dans des reseaux a graphite

Energy Technology Data Exchange (ETDEWEB)

Cogne, F; Journet, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The experiments described in this report have been carried out for the most part in the critical facility MARIUS, and a few during the start up of the EDF-1 power reactor. The first part deals with the fine structure measurements made in various lattices and with their analysis. Integration over the neutron spectrum of the mono-kinetic disadvantage factor derived by the A.B.H method yields results in good agreement with the experiments. The second part deals with spectral indexes measurements (Pu/U, In/Mn) made at room temperature in MARIUS. Comparison are made of experiments with calculations using various thermalization models. Experiments carried out at higher temperatures in EDF-1 are also described. (authors) [French] Les mesures decrites dans ce rapport ont ete faites pour la plupart dans l'empilement critique MARIUS sur des reseaux a graphite-uranium naturel. Une premiere partie traite des mesures de structure fine faites dans differents reseaux et de leur interpretation. On montre en particulier qu'une integration sur le spectre d'un calcul monocinetique type A.B.H. rend bien compte des experiences. Dans une deuxieme partie, on donne les resultats de mesures d'indices de spectre Pu/U et In/Mn faites sur des reseaux froids a MARIUS et leur comparaison avec les differents modeles de calculs de thermalisation. On donne egalement les resultats de quelques mesures en temperature effectuees lors du demarrage du reacteur EDF-1. (auteurs)

Organic free radicals and micropores in solid graphitic carbonaceous matter at the Oklo natural fission reactors, Gabon

International Nuclear Information System (INIS)

Rigali, M.J.; Nagy, B.

1997-01-01

The presence, concentration, and distribution of organic free radicals as well as their association with specific surface areas and microporosities help characterize the evolution and behavior of the Oklo carbonaceous matter. Such information is necessary in order to evaluate uranium mineralization, liquid bitumen solidification, and radio nuclide containment at Oklo. In the Oklo ore deposits and natural fission reactors carbonaceous matter is often referred to as solid graphitic bitumen. The carbonaceous parts of the natural reactors may contain as much as 65.9% organic C by weight in heterogeneous distribution within the clay-rich matrix. The solid carbonaceous matter immobilized small uraninite crystals and some fission products enclosed in this uraninite and thereby facilitated radio nuclide containment in the reactors. Hence, the Oklo natural fission reactors are currently the subjects of detailed studies because they may be useful analogues to support performance assessment of radio nuclide containment at anthropogenic radioactive waste repository sites. Seven carbonaceous matter rich samples from the 1968 ± 50 Ma old natural fission reactors and the associated Oklo uranium ore deposit were studied by electron spin resonance (ESR) spectroscopy and by measurements of specific surface areas (BET method). Humic acid, fulvic acid, and fully crystalline graphite standards were also examined by ESR spectroscopy for comparison with the Oklo solid graphitic bitumens. With one exception, the ancient Oklo bitumens have higher organic free radical concentrations than the modem humic and fulvic acid samples. The presence of carbon free radicals in the graphite standard could not be determined due to the conductivity of this material. 72 refs., 7 figs., 1 tab

The experimental determination of the buckling in the bare heavy water natural uranium critical assembly 'RB'

Energy Technology Data Exchange (ETDEWEB)

Raisic, N M; Popovic, D D; Takac, S M; Djordjevic, M M [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

1960-03-15

The buckling in the bare heavy water natural uranium critical assembly was determined by measuring the thermal neutron flux distribution. The obtained value for the critical buckling at the temperature of 20 deg C is: B{sup 2} = (8.516 {+-} 0.02) m{sup -2}. The above error is a statistical one, obtained from several series of measurements. The possible systematic error was estimated as 0.1 m{sup -2}. (author)

The experimental determination of the buckling in the bare heavy water natural uranium critical assembly 'RB'

International Nuclear Information System (INIS)

Raisic, N.M.; Popovic, D.D.; Takac, S.M.; Djordjevic, M.M.

1960-01-01

The buckling in the bare heavy water natural uranium critical assembly was determined by measuring the thermal neutron flux distribution. The obtained value for the critical buckling at the temperature of 20 deg C is: B 2 = (8.516 ± 0.02) m -2 . The above error is a statistical one, obtained from several series of measurements. The possible systematic error was estimated as 0.1 m -2 . (author)

Diffusion of uranium in H-451 graphite at 900 to 14000C

International Nuclear Information System (INIS)

Tallent, O.K.; Wichner, R.P.; Towns, R.L.

1983-03-01

In this study, the diffusion of uranium (as a stand-in for plutonium) was investigated under conditions approximating those of the primary coolant loop in a High Temperature Gas-Cooled Reactor (HTGR). Profiles were obtained for uranium penetration in H-451 graphite (from the Great Lakes Carbon Company) at temperatures ranging from 900 to 1400 0 C. Diffusion coefficients are established for UO 2 and UC 2

Physics experiments in graphite lattices (1962); Experiences sur les reseaux a graphite (1962)

Energy Technology Data Exchange (ETDEWEB)

Bacher, P; Cogne, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1962-07-01

A review is made of the various experimental methods used to determine the physics of graphite, natural uranium lattices: integral lattice experiments; both absolute and differential, effective cross section measurements, both by activation methods and by analysis of irradiated fuels, fine structure measurements. A number of experimental results are also given. (authors) [French] On decrit les differentes methodes experimentales utilisees pour determiner les parametres physiques de reseaux a uranium-graphite. Il s'agit d'experiences globales: mesures absolues et relatives de laplaciens, mesures de sections efficaces effectives par activation et par analyses de combustibles irradies, mesures de structures fines. Un certain nombre de resultats experimentaux sont communiques. (auteurs)

Graphite Isotope Ratio Method Development Report: Irradiation Test Demonstration of Uranium as a Low Fluence Indicator

International Nuclear Information System (INIS)

Reid, B.D.; Gerlach, D.C.; Love, E.F.; McNeece, J.P.; Livingston, J.V.; Greenwood, L.R.; Petersen, S.L.; Morgan, W.C.

1999-01-01

This report describes an irradiation test designed to investigate the suitability of uranium as a graphite isotope ratio method (GIRM) low fluence indicator. GIRM is a demonstrated concept that gives a graphite-moderated reactor's lifetime production based on measuring changes in the isotopic ratio of elements known to exist in trace quantities within reactor-grade graphite. Appendix I of this report provides a tutorial on the GIRM concept

Effect of Graphite on the Properties of Natural Rubber

Directory of Open Access Journals (Sweden)

Auda jabber Braihi

2016-09-01

Full Text Available Natural rubber-graphite composites (0, 1, 2, 3, 4 pphr graphite were prepared on a laboratory two-roll mill. Swelling measurements were used to evaluate the impacts of graphite on the properties of natural rubber. Swelling results showed that the volume fraction of natural rubber in the swollen gel, the interaction parameter, and the cross-link density decreased by increasing graphite loadings, while the average molecular weight of natural rubber between cross-links increased. Vulcanization results showed that only scorch time parameter increased with increasing graphite loadings, while other parameters (Max. torque, Min. torque, cure rate and cure rate index decreased. Both thermal and AC conductivities increased.

31 CFR 540.317 - Uranium feed; natural uranium feed.

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Uranium feed; natural uranium feed... (Continued) OFFICE OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.317 Uranium feed; natural uranium feed. The...

31 CFR 540.309 - Natural uranium.

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Natural uranium. 540.309 Section 540... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.309 Natural uranium. The term natural uranium means uranium found in...

Buckling and reaction rate experiments in plutonium/uranium metal fuelled, graphite moderated lattices at temperatures up to 400 deg. C. Part I: Experimental techniques and results

Energy Technology Data Exchange (ETDEWEB)

Carter, D H; Clarke, W G; Gibson, M; Hobday, R; Hunt, C; Marshall, J; Puckett, B J; Symons, C R; Wass, T [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1964-07-15

This report presents experimental measurements of bucklings, flux fine structure and fission rate distributions in graphite moderated lattices fuelled with plutonium/uranium metal at temperatures up to 400 deg. C in the sub-critical assemblies SCORPIO I and SCORPIO II. The experimental techniques employed are described in some detail. The accuracy of the experimental measurements appears to be adequate for testing methods of calculation being developed for the calculation of reactivity and temperature coefficient of reactivity for power reactors containing plutonium and uranium. (author) 26 refs, 17 tabs, 17 figs

Strong demand for natural uranium

International Nuclear Information System (INIS)

Kalinowski, P.

1975-01-01

The Deutsches Atomforum and the task group 'fuel elements' of the Kerntechnische Gesellschaft had organized an international two-day symposium in Mainz on natural uranium supply which was attended by 250 experts from 20 countries. The four main themes were: Demand for natural uranium, uranium deposits and uranium production, attitude of the uranium producing countries, and energy policy of the industrial nations. (orig./AK) [de

Study of graphite reactivity worth on well-defined cores assembled on LR-0 reactor

International Nuclear Information System (INIS)

Košťál, Michal; Rypar, Vojtěch; Milčák, Ján; Juříček, Vlastimil; Losa, Evžen; Forget, Benoit; Harper, Sterling

2016-01-01

Highlights: • A light water critical facility for graphite reactivity worth measurements. • Comparison of calculated and measured k eff . • Effect of graphite description on k eff . - Abstract: Graphite is an often-used moderating material on the basis of its good moderating power and very low absorption cross section. This small absorption cross section permits the use of natural or low-enriched uranium in graphite moderated reactors. Graphite is now being considered as the moderator for Fluoride-salt-cooled High Temperature Reactors (FHR). The critical moderator level was measured for various graphite block configurations in an experimental dry assembly of the LR-0 reactor. Comparisons with experiments were performed between Monte Carlo simulation tools for which satisfactory agreement was obtained with the exception of some systematic discrepancies. The larger discrepancies were observed when using the ENDF/B-VII.0 library. To decrease the uncertainties, based on conservative assumptions, relative comparisons were done. The results provided by the different nuclear data libraries are within 3 sigma interval of experimental uncertainties. It has been determined that differences between the results of calculations are caused by variations in the (n,n), (n,n′), (n,g) reactions and also by various angular distributions, while the (n,g) cross section variations play only a minor role for these configurations.

Neutron study of fast neutron reactor systems by exponential experiments on Harmonie - Graphite program HUG-PHUG - Oxide program PHRIXOS - Uranium program UK

International Nuclear Information System (INIS)

Desprets, Alain.

1977-12-01

Exponential experiments allow to obtain the fundamental characteristics of a lattice (material buckling, reaction rate ratios) more economically than critical experiments. This report describes the experimental techniques and the methods of analysis used for this type of experiments. The results obtained with three programs performed with the source reactor HARMONIE are given: graphite-lattices program (3 U-fueled and 3 Pu-fueled lattices); oxide-fuel program (4 PuO 2 -UO 2 lattices); pure uranium program (one lattice). Some of these lattices were also studied in critical experiments. The coherence of the results obtained by the two types of experiments is established [fr

Dosage of boron traces in graphite, uranium and beryllium oxide; Dosage de traces de bore dans le graphite, l'uranium et l'oxyde de beryllium

Energy Technology Data Exchange (ETDEWEB)

Coursier, J [Ecole Nationale Superieure de Physique et Chimie Industrielles, 75 - Paris (France); Hure, J; Platzer, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

The problem of the dosage of the boron in the materials serving to the construction of nuclear reactors arises of the following way: to determine to about 0,1 ppm close to the quantities of boron of the order of tenth ppm. We have chosen the colorimetric analysis with curcumin as method of dosage. To reach the indicated contents, it is necessary to do a previous separation of the boron and the materials of basis, either by extraction of tetraphenylarsonium fluoborate in the case of the boron dosage in uranium and the beryllium oxide, either by the use of a cations exchanger resin of in the case of graphite. (M.B.) [French] Le probleme du dosage du bore dans les materiaux servant a la construction de reacteurs nucleaires se pose de la facon suivante: determiner a environ 0,1 ppm pres des quantites de bore de l'ordre de quelques dixiemes de ppm. Nous avons choisit la colorimetrie a la curcumine comme methode de dosage. Pour atteindre les teneurs indiquees, il est necessaire d'effectuer une separation prealable du bore et des materiaux de base, soit par extraction du fluoborate de tetraphenylarsonium dans le cas du dosage de bore dans l'uranium et l'oxyde de beryllium, soit par l'utilisation d'une resine echangeuse de cations dans le cas du graphite. (M.B.)

Experiments with HEU (93.14 wt.%) metal annuli with internal graphite cylinder

Energy Technology Data Exchange (ETDEWEB)

Liu, Xiaobo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wehmann, Udo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mihalczo, John T. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-04-01

A variety of critical experiments were constructed of enriched uranium metal (oralloy ) during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, only three experimental configurations are described here. They are internal graphite reflected metal uranium assemblies with three different diameter HEU annuli (15-9 inches, 15-7 inches and 13-7 inches). These experiments can be found in Reference 1 and in their associated logbook

On the search for uranium

International Nuclear Information System (INIS)

Forland, A.

1987-01-01

The research reactor JEEP, which was completed in 1951 at Institutt for atomenergi (IFA), Kjeller, Norway, became the first reactor in the world to be built outside the big-power states. Due to Norwegian production of heavy water, the reactor was constructed as a heavy water reactor using natural uranium as fuel. A graphite reflector surrounded the reactor tank. Both uranium and graphite had to be purchased abroad. Because of the Anglo-American monopoly of all sizable uranium sources in the Western part of the world, no uranium for the reactor was available on the free market. The present study analyses Norway's and IFA's foreign relations at the time of the reactor project, and focuses in particular on the choice of the future partner that IFA had to make in order to solve its uranium problem. Political considerations were among the factors behind the decision in 1951 to establish a joint Dutch-Norwegian atomic energy research institute

Recent developments concerning French fuel elements used in natural uranium - graphite - CO{sub 2} reactor systems; Developpements recents des elements combustibles francais de la filiere uranium naturel - graphite - CO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Salesse, M; Stohr, J A; Jeanpierre, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The policy followed in France for the development of fuel elements for reactors belonging to the Electricite de France has been to benefit as much as possible, for each new pile from the most recent technical progress by developing in each case a fuel element allowing the maximum power per channel. The two latest fuel elements thus studied by the French Atomic Energy Commission are of two different types: a tubular uranium element closed at both ends and cooled externally. (This type of element, chosen for the reactors EDF 2, EDF 3 and EDF 4 makes it possible to attain maximum specific powers of the order of 6 MW/metric ton.); an open tubular uranium element cooled both internally and externally, called an annular element which in being studied as a possibility for EDF 5. Such an element makes it possible to attain specific powers of over 12 MW/metric ton. The two types of element have the following common characteristics: - the can, for external cooling, has herning-bone type fins. This type of profile which has been vastly improved recently thereby increasing its thermal efficiency, has the important advantage of avoiding vibration of the element, but has posed problems of resistance to thermal cycling necessitating much research. - the fuel rods are placed inside graphite jackets, this limiting the vertical forces to which they are subjected and protecting them during charging and discharging. On the other hand, these elements present very different problems as for as the following points are concerned: - the characteristics required of the uranium tubes apart of course from a good dimensional stability during irradiation in the two cases are in the case of the closed tubes a very high resistance to external pressure, and in the case of the annular elements a low neutron absorption. Thus for each of these two cases it has been necessary to develop a suitable type of alloy. - a possible loosening of the can during thermal cycling, which is peculiar to the

Parametric study of the criticality of natural reactors

International Nuclear Information System (INIS)

Naudet, R.

1978-01-01

Conditions for the criticality of natural reactors are investigated from a general point of view; a parametric study is presented, which expresses the possibility of chain reactions as functions of five parameters: the age of the deposit, the ore's uranium content, the volume of high-grade ore, the neutron capture of the vein of ore and the amount of water associated with the uranium. It is demonstrated that although criticality could theoretically be attained for ages that are not in excess of 1000 to 1200 MA, conditions would have to be exceptionally favorable for it since the deposits are clearly much younger than those at Oklo. The study offers a much better appreciation of the probability for discovery of other natural fissionable reactors

The nucleation and growth of uranium on the basal plane of graphite studied by scanning tunneling microscopy

International Nuclear Information System (INIS)

Tench, R.J.

1992-11-01

For the first time, nanometer scale uranium clusters were created on the basal plane of highly oriented pyrolytic graphite by laser ablation under ultra-high vacuum conditions. The physical and chemical properties of these clusters were investigated by scanning tunneling microscopy (STM) as well as standard surface science techniques. Auger electron and X-ray photoelectron spectroscopies found the uranium deposit to be free of contamination and showed that no carbide had formed with the underlying graphite. Clusters with sizes ranging from 42 Angstrom 2 to 630 Angstrom 2 were observed upon initial room temperature deposition. Surface diffusion of uranium was observed after annealing the substrate above 800 K, as evidenced by the decreased number density and the increased size of the clusters. Preferential depletion of clusters on terraces near step edges as a result of annealing was observed. The activation energy for diffusion deduced from these measurements was found to be 15 Kcal/mole. Novel formation of ordered uranium thin films was observed for coverages greater than two monolayers after annealing above 900 K. These ordered films displayed islands with hexagonally faceted edges rising in uniform step heights characteristic of the unit cell of the P-phase of uranium. In addition, atomic resolution STM images of these ordered films indicated the formation of the β-phase of uranium. The chemical properties of these surfaces were investigated and it was shown that these uranium films had a reduced oxidation rate in air as compared to bulk metal and that STM imaging in air induced a polarity-dependent enhancement of the oxidation rate

Natural graphite demand and supply - Implications for electric vehicle battery requirements

Science.gov (United States)

Olson, Donald W.; Virta, Robert L.; Mahdavi, Mahbood; Sangine, Elizabeth S.; Fortier, Steven M.

2016-01-01

Electric vehicles have been promoted to reduce greenhouse gas emissions and lessen U.S. dependence on petroleum for transportation. Growth in U.S. sales of electric vehicles has been hindered by technical difficulties and the high cost of the lithium-ion batteries used to power many electric vehicles (more than 50% of the vehicle cost). Groundbreaking has begun for a lithium-ion battery factory in Nevada that, at capacity, could manufacture enough batteries to power 500,000 electric vehicles of various types and provide economies of scale to reduce the cost of batteries. Currently, primary synthetic graphite derived from petroleum coke is used in the anode of most lithium-ion batteries. An alternate may be the use of natural flake graphite, which would result in estimated graphite cost reductions of more than US$400 per vehicle at 2013 prices. Most natural flake graphite is sourced from China, the world's leading graphite producer. Sourcing natural flake graphite from deposits in North America could reduce raw material transportation costs and, given China's growing internal demand for flake graphite for its industries and ongoing environmental, labor, and mining issues, may ensure a more reliable and environmentally conscious supply of graphite. North America has flake graphite resources, and Canada is currently a producer, but most new mining projects in the United States require more than 10 yr to reach production, and demand could exceed supplies of flake graphite. Natural flake graphite may serve only to supplement synthetic graphite, at least for the short-term outlook.

Proserpine - plutonium 239 - Proserpine - uranium 235 - comparison of experimental results; Proserpine - plutonium 239 - proserpine - uranium 235 - comparaison de resultats experimentaux

Energy Technology Data Exchange (ETDEWEB)

Brunet, J P; Caizergues, R; Clouet D' Orval, Ch; Kremser, J; Moret-Bailly, J; Verriere, Ph [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The Proserpine homogeneous reactor is constituted by a tank, 25 cm dia, 30 cm high, surrounded by a composite reflector made of beryllium oxide and graphite. In this tank can be made critical plutonium or 90 per cent enriched uranium solutions, the fissile substances being in the form of a dissolved salt. In varying the concentration of the solution, critical masses were studied as a function of the level of the liquid in the tank. The minimum critical mass is 256 {+-} 2 grs for plutonium and 409 {+-} 3 grs for uranium 235. In the range of the critical concentrations which were studied, the neutronic properties of fissionable solutions of plutonium and enriched uranium were compared for identical geometries. (authors) [French] Proserpine est un reacteur homogene comportant une cuve de diametre 25 cm, de hauteur 30 cm, entouree d'un reflecteur composite d'oxyde de beryllium et de graphite. On y a rendu critiques des solutions de plutonium ou d'uranium enrichi a 90 pour cent, le produit fissile se trouvant sous la forme d'un sel dissous. En faisant varier la concentration de la solution, on a etudie les masses critiques en fonction de la hauteur du liquide dans la cuve. La masse- critique minimum est, pour le plutonium de 256 {+-} 2 g, pour l'uranium 235 de 409 {+-} 3 g. Dans la gamme des concentrations critiques etudiees, on a compare, dans des conditions de geometrie identique, les proprietes neutroniques des solutions fissiles de plutonium et d'uranium enrichi. (auteurs)

Organic tissues, graphite, and hydrocarbons in host rocks of the Rum Jungle Uranium Field, northern Australia

Science.gov (United States)

Foster, C.B.; Robbins, E.I.; Bone, Y.

1990-01-01

The Rum Jungle Uranium field consists of at least six early Proterozoic deposits that have been mined either for uranium and/or the associated base and precious metals. Organic matter in the host rocks of the Whites Formation and Coomalie Dolomite is now predominantly graphite, consistent with the metamorphic history of these rocks. For nine samples, the mean total organic carbon content is high (3.9 wt%) and ranged from 0.33 to 10.44 wt%. Palynological extracts from the host rocks include black, filamentous, stellate (Eoastrion-like), and spherical morphotypes, which are typical of early Proterozoic microbiota. The colour, abundance, and shapes of these morphotypes reflect the thermal history, organic richness, and probable lacustrine biofacies of the host rocks. Routine analysis of rock thin sections and of palynological residues shows that mineral grains in some of the host rocks are coated with graphitized organic matter. The grain coating is presumed to result from ultimate thermal degradation of a petroleum phase that existed prior to metamorphism. Hydrocarbons are, however, still present in fluid inclusions within carbonates of the Coomalie Dolomite and lower Whites Formation. The fluid inclusions fluoresce dull orange in blue-light excitation and their hydrocarbon content is confirmed by gas chromatography of whole-rock extracts. Preliminary analysis of the oil suggests that it is migrated, and because it has escaped graphitization through metamorphism it is probably not of early Proterozoic age. The presence of live oil is consistent with fluid inclusion data that suggest subsequent, low-temperature brine migration through the rocks. The present observations support earlier suggestions that organic matter in the host formations trapped uranium to form protore. Subsequent fluid migrations probably brought additional uranium and other metals to these formations, and the organic matter provided a reducing environment for entrapment. ?? 1990.

Synthesis of graphite intercalation compound of group VI metals and uranium hexafluorides

International Nuclear Information System (INIS)

Fukui, Toshihiro; Hagiwara, Rika; Ema, Keiko; Ito, Yasuhiko

1993-01-01

Systematic investigations were made on the synthesis of graphite intercalation compounds of group VI transition metals (W and Mo) and uranium hexafluorides. The reactions were performed by interacting liquid or gaseous metal hexafluorides with or without elemental fluorine at ambient temperature. The degree of intercalation of these metal fluorides depends on the formation enthalpy of fluorometallate anion from the original metal hexafluoride, as has been found for other intercalation reactions of metal fluorides. (author)

Criticality analysis in uranium enrichment plant

International Nuclear Information System (INIS)

Okamoto, Tsuyoshi; Kiyose, Ryohei

1977-01-01

In a large scale uranium enrichment plant, uranium inventory in cascade rooms is not very large in quantity, but the facilities dealing with the largest quantity of uranium in that process are the UF 6 gas supply system and the blending system for controlling the product concentration. When UF 6 spills out of these systems, the enriched uranium is accumulated, and the danger of criticality accident is feared. If a NaF trap is placed at the forestage of waste gas treatment system, plenty of UF 6 and HF are adsorbed together in the NaF trap. Thus, here is the necessity of checking the safety against criticality. Various assumptions were made to perform the computation surveying the criticality of the system composed of UF 6 and HF adsorbed on NaF traps with WIMS code (transport analysis). The minimum critical radius resulted in about 53 cm in case of 3.5% enriched fuel for light water reactors. The optimum volume ratio of fissile material in the double salt UF 6 .2NaF and NaF.HF is about 40 vol. %. While, criticality survey computation was also made for the annular NaF trap having the central cooling tube, and it was found that the effect of cooling tube radius did not decrease the multiplication factor up to the cooling tube radius of about 5 cm. (Wakatsuki, Y.)

Criticality of mixtures of plutonium and high enriched uranium

International Nuclear Information System (INIS)

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

2003-01-01

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

Graphite reactor physics; Physique des piles a graphite

Energy Technology Data Exchange (ETDEWEB)

Bacher, P; Cogne, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Noc, B [Electricite de France (EDF), 75 - Paris (France)

1964-07-01

The study of graphite-natural uranium power reactor physics, undertaken ten years ago when the Marcoule piles were built, has continued to keep in step with the development of this type of pile. From 1960 onwards the critical facility Marius has been available for a systematic study of the properties of lattices as a function of their pitch, of fuel geometry and of the diameter of cooling channels. This study has covered a very wide field: lattice pitch varying from 19 to 38 cm. uranium rods and tubes of cross-sections from 6 to 35 cm{sup 2}, channels with diameters between 70 and 140 mm. The lattice calculation methods could thus be checked and where necessary adapted. The running of the Marcoule piles and the experiments carried out on them during the last few years have supplied valuable information on the overall evolution of the neutronic properties of the fuel as a function of irradiation. More detailed experiments have also been performed in Marius with plutonium-containing fuels (irradiated or synthetic fuels), and will be undertaken at the beginning of 1965 at high temperature in the critical facility Cesar, which is just being completed at Cadarache. Spent fuel analyses complement these results and help in their interpretation. The thermalization and spectra theories developed in France can thus be verified over the whole valid temperature range. The efficiency of control rods as a function of their dimensions, the materials of which they are made and the lattices surrounding them has been measured in Marius, and the results compared with calculation on the one hand and with the measurements carried out in EDF 1 on the other. Studies on the control proper of graphite piles were concerned essentially with the risks of spatial instability and the means of detecting and controlling them, and with flux distortions caused by the control rods. (authors) [French] Entreprise il y a dix ans a l'occasion de la construction des piles de Marcoule, l'etude de la

Isotopic composition of uranium in U3O8 by neutron induced reactions utilizing thermal neutrons from critical facility and high resolution gamma-ray spectrometry

International Nuclear Information System (INIS)

Acharya, R.; Pujari, P.K.; Goel, Lokesh

2015-01-01

Uranium in oxide and metal forms is used as fuel material in nuclear power reactors. For chemical quality control, it is necessary to know the isotopic composition (IC) of uranium i.e., 235 U to 238 U atom ratio as well as 235 U atom % in addition to its total concentration. Uranium samples can be directly assayed by passive gamma ray spectrometry for obtaining IC by utilizing 185 keV (γ-ray abundance 57.2%) of 235 U and 1001 keV (γ-ray abundance 0.837%) of 234m Pa (decay product of 238 U). However, due to low abundance of 1001 keV, often it is not practiced to obtain IC by this method as it gives higher uncertainty even if higher mass of sample and counting time are used. IC of uranium can be determined using activity ratio of neutron induced fission product of 235 U to activation product of 238 U ( 239 Np). In the present work, authors have demonstrated methodologies for determination of IC of U as well as 235 U atom% in natural ( 235 U 0.715%) and low enriched uranium (LEU, 3-20 atom % of 235 U) samples of uranium oxide (U 3 O 8 ) by utilizing ratio of counts at 185 keV γ-ray or γ-rays of fission products with respect to 277 keV of 239 Np. Natural and enriched samples (about 25 mg) were neutron irradiated for 4 hours in graphite reflector position of AHWR Critical Facility (CF) using highly thermalized (>99.9% thermal component) neutron flux (∼10 7 cm -2 s -1 )

Uranium-thorium fuel cycle in a very high temperature hybrid system

International Nuclear Information System (INIS)

Hernandez, C.R.G.; Oliva, A.M.; Fajardo, L.G.; Garcia, J.A.R.; Curbelo, J.P.; Abadanes, A.

2011-01-01

Thorium is a potentially valuable energy source since it is about three to four times as abundant as Uranium. It is also a widely distributed natural resource readily accessible in many countries. Therefore, Thorium fuels can complement Uranium fuels and ensure long term sustainability of nuclear power. The main advantages of the use of a hybrid system formed by a Pebble Bed critical nuclear reactor and two Pebble Bed Accelerator Driven Systems (ADSs) using a Uranium-Thorium (U + Th) fuel cycle are shown in this paper. Once-through and two step U + Th fuel cycle was evaluated. With this goal, a preliminary conceptual design of a hybrid system formed by a Graphite Moderated Gas-Cooled Very High Temperature Reactor and two ADSs is proposed. The main parameters related to the neutronic behavior of the system in a deep burn scheme are optimized. The parameters that describe the nuclear fuel breeding and Minor Actinide stockpile are compared with those of a simple Uranium fuel cycle. (author)

Electrical properties of Egyptian natural graphite

International Nuclear Information System (INIS)

El-Shazly, O.; El-Wahidy, E.F.; Elanany, N.; Saad, N.A.

1992-06-01

The electrical properties of Egyptian natural graphite flakes, obtained from the graphite schists of Wadi Bent, Eastern Desert, were measured. The flakes were ground and compressed into pellets. The standard four probe dc method was used to measure the temperature dependence of the electric resistivity from room temperature down to 12 K. The transverse and longitudinal magnetoresistance were measured in the low magnetic field range at temperatures 300 K, 77 K and 12 K. The transverse magnetoresistance data was used to estimate the average mobility, assuming a simple two-band model. (author). 20 refs, 4 figs, 1 tab

Criticality safety concerns of uranium deposits in cascade equipment

International Nuclear Information System (INIS)

Plaster, M.J.

1996-01-01

The Paducah and Portsmouth Gaseous Diffusion Plants enrich uranium in the 235 U isotope by diffusing gaseous uranium hexafluoride (UF 6 ) through a porous barrier. The UF 6 gaseous diffusion cascade utilized several thousand open-quotes stagesclose quotes of barrier to produce highly enriched uranium (HEU). Historically, Portsmouth has enriched the Paducah Gaseous Diffusion Plant's product (typically 1.8 wt% 235 U) as well as natural enrichment feed stock up to 97 wt%. Due to the chemical reactivity of UF 6 , particularly with water, the formation of solid uranium deposits occur at a gaseous diffusion plant. Much of the equipment operates below atmospheric pressure, and deposits are formed when atmospheric air enters the cascade. Deposits may also be formed from UF 6 reactions with oil, UF 6 reactions with the metallic surfaces of equipment, and desublimation of UF 6 . The major deposits form as a result of moist air in leakage due to failure of compressor casing flanges, blow-off plates, seals, expansion joint convolutions, and instrument lines. This report describes criticality concerns and deposit disposition

Proserpine - plutonium 239 - Proserpine - uranium 235 - comparison of experimental results

International Nuclear Information System (INIS)

Brunet, J.P.; Caizergues, R.; Clouet D'Orval, Ch.; Kremser, J.; Moret-Bailly, J.; Verriere, Ph.

1964-01-01

The Proserpine homogeneous reactor is constituted by a tank, 25 cm dia, 30 cm high, surrounded by a composite reflector made of beryllium oxide and graphite. In this tank can be made critical plutonium or 90 per cent enriched uranium solutions, the fissile substances being in the form of a dissolved salt. In varying the concentration of the solution, critical masses were studied as a function of the level of the liquid in the tank. The minimum critical mass is 256 ± 2 grs for plutonium and 409 ± 3 grs for uranium 235. In the range of the critical concentrations which were studied, the neutronic properties of fissionable solutions of plutonium and enriched uranium were compared for identical geometries. (authors) [fr

Dietary intake and body content of natural uranium

International Nuclear Information System (INIS)

Anon.

1984-01-01

The members of the uranium series found in the body that arise primarily from dietary intake are 238 U, 234 U, 226 Ra and 210 Pb. Lead 210, the predominant series radionuclide in the body, decays to the alpha emitter 210 Po, while the others are alpha emitters themselves. While 210 Pb primarily enters the body through diet, inhalation must also be considered, especially in smokers. The primary site of deposition for these nuclides is the skeleton and the dose to bone is the critical factor. In this section, the average background, elevated natural and enhanced dietary intakes of the uranium series radionuclides are discussed. Human skeletal levels and consequent alpha doses are summarized

Dosage of boron traces in graphite, uranium and beryllium oxide

International Nuclear Information System (INIS)

Coursier, J.; Hure, J.; Platzer, R.

1955-01-01

The problem of the dosage of the boron in the materials serving to the construction of nuclear reactors arises of the following way: to determine to about 0,1 ppm close to the quantities of boron of the order of tenth ppm. We have chosen the colorimetric analysis with curcumin as method of dosage. To reach the indicated contents, it is necessary to do a previous separation of the boron and the materials of basis, either by extraction of tetraphenylarsonium fluoborate in the case of the boron dosage in uranium and the beryllium oxide, either by the use of a cations exchanger resin of in the case of graphite. (M.B.) [fr

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

International Nuclear Information System (INIS)

Elhardt, W.

1979-01-01

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

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.)

Uranium Enrichment Determination of the InSTEC Sub Critical Ensemble Fuel by Gamma Spectrometry

International Nuclear Information System (INIS)

Borrell Munnoz, Jose L.; LopezPino, Neivy; Diaz Rizo, Oscar; D'Alessandro Rodriguez, Katia; Padilla Cabal, Fatima; Arbelo Penna, Yunieski; Garcia Rios, Aczel R.; Quintas Munn, Ernesto L.; Casanova Diaz, Amaya O.

2009-01-01

Low background gamma spectrometry was applied to analyze the uranium enrichment of the nuclear fuel used in the InSTEC Sub Critical ensemble. The enrichment was calculated by two variants: an absolute method using the Monte Carlo method to simulated detector volumetric efficiency, and an iterative procedure without using standard sources. The results confirm that the nuclear fuel of the ensemble is natural uranium without any additional degree of enrichment. (author)

Minimum critical masses for uranium at the Portsmouth Gaseous Diffusion Plant

International Nuclear Information System (INIS)

Tayloe, R.W. Jr.; Davis, T.C.

1994-06-01

This report presents a tabulation of safe masses and minimum critical masses for uranium (U). These minimum critical mass and safe mass tables were obtained by interpolating between the values reported in the literature to obtain values as a function of enrichment within the 1.5 percent to 100 percent range. Equivalent mass values for uranium-235 (U 235 ), uranium hexafluoride (UF 6 ), and uranyl fluoride (UO 2 F 2 ) have been generated from the safe mass and minimum critical masses for uranium

Criticality of moderated and undermoderated low-enriched uranium oxide systems

International Nuclear Information System (INIS)

Goebel, G.R.

1980-06-01

Uranium oxide was enriched to 4.46 wt % 235 U compacted to a density of 4.68 g/cm 3 . The uranium oxide was packed into cubical aluminum cans and water added to the oxide until an H/U atomic ratio of 0.77 was achieved. A 5 x 5 x 5 array of uranium oxide cans for the experiments were used when no plastic moderator material was placed between cans. High enriched uranium drivers were used to achieve criticality. Criticality was achieved for smaller arrays without a driver when 24.5 mm plastic moderator material was placed between the cans. Twelve critical experiments are reported, six in each reflector

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Purification process of natural graphite as anode for Li-ion batteries: chemical versus thermal

Science.gov (United States)

Zaghib, K.; Song, X.; Guerfi, A.; Rioux, R.; Kinoshita, K.

The intercalation of Li ions in natural graphite that was purified by chemical and thermal processes was investigated. A new chemical process was developed that involved a mixed aqueous solution containing 30% H 2SO 4 and 30% NH xF y heated to 90 °C. The results of this process are compared to those obtained by heating the natural graphite from 1500 to 2400 °C in an inert environment (thermal process). The first-cycle coulombic efficiency of the purified natural graphite obtained by the chemical process is 91 and 84% after the thermal process at 2400 °C. Grinding the natural graphite before or after purification had no significant effect on electrochemical performance at low currents. However, grinding to a very small particle size before purification permitted optimization of the size distribution of the particles, which gives rise to a more homogenous electrode. The impurities in the graphite play a role as microabrasion agents during grinding which enhances its hardness and improves its mechanical properties. Grinding also modifies the particle morphology from a 2- to a 3-D structure (similar in shape to a potato). This potato-shaped natural graphite shows high reversible capacity at high current densities (about 90% at 1 C rate). Our analysis suggests that thermal processing is considerably more expensive than the chemical process to obtain purified natural graphite.

{sup 36}Cl and {sup 14}C behaviour in UNGG graphite during leaching experiments

Energy Technology Data Exchange (ETDEWEB)

Pichon, C.; Guy, C.; Comte, J. [Commissariat a l' Energie Atomique - C.E.A., Laboratoire d' Analyses Radiochimiques et Chimiques (L.A.R.C.) 13108 Saint Paul lez Durance (France)

2008-07-01

Graphite has been used as a moderator in Natural Uranium Graphite Gas reactors. Among the radionuclides, the long-lived activation product {sup 36}Cl and {sup 14}C, which are abundant in graphite after irradiation can be the main contributors to the dose during disposal. This paper deals with the first results obtained on irradiated graphite from French G2 reactor. Both leaching and diffusion experiments have been performed in order to understand and quantify the radionuclides behaviour. Chlorine leaching seems to be controlled by diffusion transport through graphite matrix. On the contrary {sup 14}C leaching is very low, probably because after irradiation, the remaining {sup 14}C was produced from {sup 13}C activation in the crystalline structure of graphite. (authors)

Projections on the future of the natural uranium industry

International Nuclear Information System (INIS)

Ishido, Akio

1995-01-01

This discussion looks at the future of the uranium industry and considers what type of procurement policy should be adopted. Viewing the future as an extension of the present, it is possible that supplies of natural uranium will begin to run short around 2015. However, natural uranium will have more resources available than petroleum. If rising uranium prices reinvigorate exploration and lead to the discovery of new uranium deposits, future shortages will be unlikely. Nonetheless, with structural changes expected in the world economy, the nature of natural uranium transactions will no doubt change, thereby increasing the present element of uncertainty that much more. At the same time, the oligopolistic situation created by today's major producers will intensify. Based on these projections, the author has reassessed Japan's past procurement policy of government exploration/development support combined with private-sector uranium purchasing and finds this shared risk approach to be the best. (author)

Study of the thermal drop at the uranium-can interface for fuel elements in gas-graphite reactors; Etude de la chute thermique au contact uranium-gaine pour des elements combustibles de reacteur de la filiere graphite-gaz

Energy Technology Data Exchange (ETDEWEB)

Faussat, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Levenes, G; Michel, M [Societe Industrielle de Combustible Nucleaire (France)

1964-07-01

The report reviews the tests now under way at the CEA, for determining the thermal contact resistance at the uranium-can interface for fuel elements used in gas-graphite type reactors. These are laboratory tests carried out with equipment based on the principle of a heat flow across a stack of test pieces having planar contact surfaces. The following points emerge from this work: - for a metallic uranium element canned in magnesium, of the type G-2 or EDF-2, a value of 0.2 deg C/W/cm{sup 2} seems reasonable for can temperatures of 400 deg C and above. - this value is independent of the micro-geometric state of the uranium surface in a range of roughness which easily includes those observed on tubes and rods produced industrially. - for the internal cans of elements cooled internally and externally, the value of the contact resistance for temperatures of under 400 deg C as a function of the stresses in the can has not yet been measured exactly. (authors) [French] Le rapport fait le point des essais actuellement en cours au CEA pour determiner la resistance thermique de contact uranium-gaine pour des reacteurs de la filiere graphite-gaz. Ces essais sont effectues en laboratoire sur des appareils bases sur le principe d'une circulation de flux de chaleur a travers un empilement d'eprouvettes dont les faces en contact sont planes. De l'etude, il ressort essentiellement que: - pour un element a uranium metallique et gaine de magnesium type G-2 ou EdF-2, on peut admettre la valeur de 0,2 deg C/W/cm{sup 2} pour des temperatures de gaines de 400 deg C et plus. - cette valeur ne depend pas de l'etat de surface microgeometrique de l'uranium pour un domaine de rugosites couvrant largement celles que l'on observe sur des tubes et barreaux fabriques en serie. - pour les gaines internes d'elements a refroidissement interne et externe la valeur de la resistance de contact reste a preciser pour les temperatures inferieures a 400 deg C, en fonction des contraintes existant dans les

Reference materials for nondestructive assay of special nuclear material. Volume 1. Uranium oxide plus graphite powder

International Nuclear Information System (INIS)

Sprinkle, J.K.; Likes, R.N.; Parker, J.L.; Smith, H.A.

1983-10-01

This manual describes the fabrication of reference materials for use in gamma-ray-based nondestructive assay of low-density uranium-bearing samples. The sample containers are 2-l bottles. The reference materials consist of small amounts of UO 2 spread throughout a graphite matrix. The 235 U content ranges from 0 to 100 g. The manual also describes the far-field assay procedure used with low-resolution detectors

Ecological considerations of natural and depleted uranium

International Nuclear Information System (INIS)

Hanson, W.C.

1980-01-01

Depleted 238 U is a major by-product of the nuclear fuel cycle for which increasing use is being made in counterweights, radiation shielding, and ordnance applications. This paper (1) summarizes the pertinent literature on natural and depleted uranium in the environment, (2) integrates results of a series of ecological studies conducted at Los Alamos Scientific Laboratory (LASL) in New Mexico where 70,000 kg of depleted and natural uranium has been expended to the environment over the past 34 years, and (3) synthesizes the information into an assessment of the ecological consequences of natural and depleted uranium released to the environment by various means. Results of studies of soil, plant, and animal communities exposed to this radiation and chemical environment over a third of a century provide a means of evaluating the behavior and effects of uranium in many contexts

The potential for criticality following disposal of uranium at low-level waste facilities: Uranium blended with soil

Energy Technology Data Exchange (ETDEWEB)

Toran, L.E.; Hopper, C.M.; Naney, M.T. [and others

1997-06-01

The purpose of this study was to evaluate whether or not fissile uranium in low-level-waste (LLW) facilities can be concentrated by hydrogeochemical processes to permit nuclear criticality. A team of experts in hydrology, geology, geochemistry, soil chemistry, and criticality safety was formed to develop achievable scenarios for hydrogeochemical increases in concentration of special nuclear material (SNM), and to use these scenarios to aid in evaluating the potential for nuclear criticality. The team`s approach was to perform simultaneous hydrogeochemical and nuclear criticality studies to (1) identify some achievable scenarios for uranium migration and concentration increase at LLW disposal facilities, (2) model groundwater transport and subsequent concentration increase via sorption or precipitation of uranium, and (3) evaluate the potential for nuclear criticality resulting from potential increases in uranium concentration over disposal limits. The analysis of SNM was restricted to {sup 235}U in the present scope of work. The outcome of the work indicates that criticality is possible given established regulatory limits on SNM disposal. However, a review based on actual disposal records of an existing site operation indicates that the potential for criticality is not a concern under current burial practices.

The potential for criticality following disposal of uranium at low-level waste facilities: Uranium blended with soil

International Nuclear Information System (INIS)

Toran, L.E.; Hopper, C.M.; Naney, M.T.

1997-06-01

The purpose of this study was to evaluate whether or not fissile uranium in low-level-waste (LLW) facilities can be concentrated by hydrogeochemical processes to permit nuclear criticality. A team of experts in hydrology, geology, geochemistry, soil chemistry, and criticality safety was formed to develop achievable scenarios for hydrogeochemical increases in concentration of special nuclear material (SNM), and to use these scenarios to aid in evaluating the potential for nuclear criticality. The team's approach was to perform simultaneous hydrogeochemical and nuclear criticality studies to (1) identify some achievable scenarios for uranium migration and concentration increase at LLW disposal facilities, (2) model groundwater transport and subsequent concentration increase via sorption or precipitation of uranium, and (3) evaluate the potential for nuclear criticality resulting from potential increases in uranium concentration over disposal limits. The analysis of SNM was restricted to 235 U in the present scope of work. The outcome of the work indicates that criticality is possible given established regulatory limits on SNM disposal. However, a review based on actual disposal records of an existing site operation indicates that the potential for criticality is not a concern under current burial practices

Criticality calculations for homogeneous mixtures of uranium and plutonium

International Nuclear Information System (INIS)

Spiegelberg, R. de S.H.

1981-05-01

Critical parameters were calculated using the one-dimensional multigroup transport theory. Calculations have been performed for water mixture of uranium metal and uranium oxides and plutonium nitrates to determine the dimensions of simple critical geometries (sphere and cylinder). The results of the calculations were plotted showing critical parameters (volume, radius or critical mass). The critical values obtained in Handbuch zur Kritikalitat were used to compare with critical parameters. A sensitivity study for the influences of mesh space size, multigroup structure and order of the S sub(n) approximation on the critical radius was carried out. The GAMTEC-II code was used to generate multigroup cross sections data. Critical radius were calculated using the one-dimensional multigroup transport code DTF-IV. (Author) [pt

SRTC criticality technical review: Nuclear Criticality Safety Evaluation 93-18 Uranium Solidification Facility's Waste Handling Facility

International Nuclear Information System (INIS)

Rathbun, R.

1993-01-01

Separate review of NMP-NCS-930058, open-quotes Nuclear Criticality Safety Evaluation 93-18 Uranium Solidification Facility's Waste Handling Facility (U), August 17, 1993,close quotes was requested of SRTC Applied Physics Group. The NCSE is a criticality assessment to determine waste container uranium limits in the Uranium Solidification Facility's Waste Handling Facility. The NCSE under review concludes that the NDA room remains in a critically safe configuration for all normal and single credible abnormal conditions. The ability to make this conclusion is highly dependent on array limitation and inclusion of physical barriers between 2x2x1 arrays of boxes containing materials contaminated with uranium. After a thorough review of the NCSE and independent calculations, this reviewer agrees with that conclusion

Synthesis of graphene oxide and reduced graphene oxide by needle platy natural vein graphite

Energy Technology Data Exchange (ETDEWEB)

Rathnayake, R.M.N.M. [National Institute of Fundamental Studies, Kandy (Sri Lanka); Graduate School of Engineering, Toyota Technological Institute, 2-12-1 Hisakata, Tempaku, Nagoya 468-8511 (Japan); Wijayasinghe, H.W.M.A.C., E-mail: athulawijaya@gmail.com [National Institute of Fundamental Studies, Kandy (Sri Lanka); Pitawala, H.M.T.G.A. [Department of Geology, University of Peradeniya, Peradeniya (Sri Lanka); Yoshimura, Masamichi; Huang, Hsin-Hui [Graduate School of Engineering, Toyota Technological Institute, 2-12-1 Hisakata, Tempaku, Nagoya 468-8511 (Japan)

2017-01-30

Highlights: • The high purity of this form of needle platy natural vein graphite is expected to synthesize GO and rGO readily and efficiently, as compared to the synthetic and less pure graphite raw materials. • Production of large-scale GO and rGO for industrial applications can be achieved by using this highly crystalline NPG vein graphite, and it adds value to the natural resources. • High quality, few-layer, and cost effective GO and rGO can achieve great results using this low cost, natural graphite. - Abstract: Among natural graphite varieties, needle platy vein graphite (NPG) has very high purity. Therefore, it is readily used to prepare graphene oxide (GO) and reduced graphene oxide (rGO). In this study, GO and rGO were prepared using chemical oxidation and reduction process, respectively. The synthesized materials were characterized by X-ray diffraction (XRD), atomic force microscopy (AFM), scanning electron microscopy (SEM), high-resolution transmission electron microscopy (HRTEM), X-ray photoelectron spectroscopy (XPS), and Fourier transform infrared (FTIR) spectroscopy. XRD studies confirmed the increase of the interlayer spacing of GO and rGO in between 3.35 to 8.66 A°. AFM studies showed the layer height of rGO to be 1.05 nm after the reduction process. TEM micrographs clearly illustrated that the prepared GO has more than 25 layers, while the rGO has only less than 15 layers. Furthermore, the effect of chemical oxidation and reduction processes on surface morphology of graphite were clearly observed in FESEM micrographs. The calculated R{sub O/C} of GO and rGO using XPS analysis are 5.37% and 1.77%, respectively. The present study revealed the successful and cost effective nature of the chemical oxidation, and the reduction processes for the production of GO and rGO out of natural vein graphite.

Synthesis of graphene oxide and reduced graphene oxide by needle platy natural vein graphite

International Nuclear Information System (INIS)

Rathnayake, R.M.N.M.; Wijayasinghe, H.W.M.A.C.; Pitawala, H.M.T.G.A.; Yoshimura, Masamichi; Huang, Hsin-Hui

2017-01-01

Highlights: • The high purity of this form of needle platy natural vein graphite is expected to synthesize GO and rGO readily and efficiently, as compared to the synthetic and less pure graphite raw materials. • Production of large-scale GO and rGO for industrial applications can be achieved by using this highly crystalline NPG vein graphite, and it adds value to the natural resources. • High quality, few-layer, and cost effective GO and rGO can achieve great results using this low cost, natural graphite. - Abstract: Among natural graphite varieties, needle platy vein graphite (NPG) has very high purity. Therefore, it is readily used to prepare graphene oxide (GO) and reduced graphene oxide (rGO). In this study, GO and rGO were prepared using chemical oxidation and reduction process, respectively. The synthesized materials were characterized by X-ray diffraction (XRD), atomic force microscopy (AFM), scanning electron microscopy (SEM), high-resolution transmission electron microscopy (HRTEM), X-ray photoelectron spectroscopy (XPS), and Fourier transform infrared (FTIR) spectroscopy. XRD studies confirmed the increase of the interlayer spacing of GO and rGO in between 3.35 to 8.66 A°. AFM studies showed the layer height of rGO to be 1.05 nm after the reduction process. TEM micrographs clearly illustrated that the prepared GO has more than 25 layers, while the rGO has only less than 15 layers. Furthermore, the effect of chemical oxidation and reduction processes on surface morphology of graphite were clearly observed in FESEM micrographs. The calculated R_O_/_C of GO and rGO using XPS analysis are 5.37% and 1.77%, respectively. The present study revealed the successful and cost effective nature of the chemical oxidation, and the reduction processes for the production of GO and rGO out of natural vein graphite.

Urea-assisted liquid-phase exfoliation of natural graphite into few-layer graphene

Science.gov (United States)

Hou, Dandan; Liu, Qinfu; Wang, Xianshuai; Qiao, Zhichuan; Wu, Yingke; Xu, Bohui; Ding, Shuli

2018-05-01

The mass production of graphene with high quality is desirable for its wide applications. Here, we demonstrated a facile method to exfoliate natural graphite into graphene in organic solvent by assisting of urea. The exfoliation of graphite may originate from the "molecular wedge" effect of urea, which can intercalate into the edge of natural graphite, thus facilitating the production of graphene dispersion with a high concentration up to 1.2 mg/mL. The obtained graphene is non-oxidized with negligible defects. Therefore, this approach has great promise in bulk production of graphene with superior quality for a variety of applications.

Depleted and natural uranium: chemistry and toxicological effects.

Science.gov (United States)

Craft, Elena; Abu-Qare, Aquel; Flaherty, Meghan; Garofolo, Melissa; Rincavage, Heather; Abou-Donia, Mohamed

2004-01-01

Depleted uranium (DU) is a by-product from the chemical enrichment of naturally occurring uranium. Natural uranium is comprised of three radioactive isotopes: (238)U, (235)U, and (234)U. This enrichment process reduces the radioactivity of DU to roughly 30% of that of natural uranium. Nonmilitary uses of DU include counterweights in airplanes, shields against radiation in medical radiotherapy units and transport of radioactive isotopes. DU has also been used during wartime in heavy tank armor, armor-piercing bullets, and missiles, due to its desirable chemical properties coupled with its decreased radioactivity. DU weapons are used unreservedly by the armed forces. Chemically and toxicologically, DU behaves similarly to natural uranium metal. Although the effects of DU on human health are not easily discerned, they may be produced by both its chemical and radiological properties. DU can be toxic to many bodily systems, as presented in this review. Most importantly, normal functioning of the kidney, brain, liver, and heart can be affected by DU exposure. Numerous other systems can also be affected by DU exposure, and these are also reviewed. Despite the prevalence of DU usage in many applications, limited data exist regarding the toxicological consequences on human health. This review focuses on the chemistry, pharmacokinetics, and toxicological effects of depleted and natural uranium on several systems in the mammalian body. A section on risk assessment concludes the review.

SRTC criticality safety technical review: Nuclear Criticality Safety Evaluation 93-04 enriched uranium receipt

International Nuclear Information System (INIS)

Rathbun, R.

1993-01-01

Review of NMP-NCS-930087, open-quotes Nuclear Criticality Safety Evaluation 93-04 Enriched Uranium Receipt (U), July 30, 1993, close quotes was requested of SRTC (Savannah River Technology Center) Applied Physics Group. The NCSE is a criticality assessment to determine the mass limit for Engineered Low Level Trench (ELLT) waste uranium burial. The intent is to bury uranium in pits that would be separated by a specified amount of undisturbed soil. The scope of the technical review, documented in this report, consisted of (1) an independent check of the methods and models employed, (2) independent HRXN/KENO-V.a calculations of alternate configurations, (3) application of ANSI/ANS 8.1, and (4) verification of WSRC Nuclear Criticality Safety Manual procedures. The NCSE under review concludes that a 500 gram limit per burial position is acceptable to ensure the burial site remains in a critically safe configuration for all normal and single credible abnormal conditions. This reviewer agrees with that conclusion

Laboratory simulation studies of uranium mobility in natural waters

International Nuclear Information System (INIS)

Giblin, A.M.; Swaine, D.J.; Batts, B.D.

1981-01-01

The effects of imposed variations of pH and Eh on aqueous uranium mobility at 25 0 C have been studied in three simulations of natural water systems. Constituents tested for their effect on uranium mobility were: (a) hydrous ferric oxide, to represent adsorptive solids which precipitate or dissolve in response to variations in pH and Eh; (b) kaolinite, representing minerals which, although modified by pH and Eh changes, are present as solids over the pH-Eh range of natural waters; and (c) carbonate, to represent a strong uranium-complexing species. Uranium mobility measurements from each simulation were regressed against pH and Eh within a range appropriate to natural waters. Hydrous ferric oxide and kaolinite each affected uranium mobility, but in separate pH-Eh domains. Aqueous carbonate increased mobility of uranium, and adsorption of UO 2 (CO 3 ) 3 4- caused colloidal dispersion of hydrous ferric oxide, possibly explaining the presence of 'hydrothermal hematite' in some uranium deposits. Enhanced uranium mobility observed in the pH-Eh domains of thermodynamically insoluble uranium oxides could be explained if the oxides were present as colloids. Uranium persisting as a mobile species, even after reduction, has implications for the near surface genesis of uranium ores. (author)

Buckling and reaction rate measurements in graphite moderated lattices fuelled with plutonium-uranium oxide clusters at temperatures up to 400 deg. C

International Nuclear Information System (INIS)

Carter, D.H.; Gibson, M.; King, D.C.; Marshall, J.; Puckett, B.J.; Richards, A.E.; Wass, T.; Wilson, D.J.

1965-07-01

The Report describes a series of experiments carried out in SCORPIO I and II on sub-critical graphite moderated lattices fuelled with 21-rod clusters of PuO 2 /UO 2 fuel. Three fuel batches with nominal plutonium: uranium ratios of 0.25%, 0.8% and 1.2% were investigated at temperatures between 20 deg. C and 400 deg. C. Because of the limited amounts of the three fuels, exponential measurements were made in 2-zone stacks, the outer regions of which were loaded with suitably matched 'reference fuel'. Fine structure distributions in the lattice cell were obtained with manganese and indium foils. Pu239/U235 fission ratios were determined both by fission chambers and by fission-product counting techniques. (author)

Nuclear criticality safety parameter evaluation for uranium metallic alloy

Energy Technology Data Exchange (ETDEWEB)

Sanchez, Andrea; Abe, Alfredo, E-mail: andreasdpz@hotmail.com, E-mail: abye@uol.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Energia Nuclear

2013-07-01

Nuclear criticality safety during fuel fabrication process, transport and storage of fissile and fissionable materials requires criticality safety analysis. Normally the analysis involves computer calculations and safety parameters determination. There are many different Criticality Safety Handbooks where such safety parameters for several different fissile mixtures are presented. The handbooks have been published to provide data and safety principles for the design, safety evaluation and licensing of operations, transport and storage of fissile and fissionable materials. The data often comprise not only critical values, but also subcritical limits and safe parameters obtained for specific conditions using criticality safety calculation codes such as SCALE system. Although many data are available for different fissile and fissionable materials, compounds, mixtures, different enrichment level, there are a lack of information regarding a uranium metal alloy, specifically UMo and UNbZr. Nowadays uranium metal alloy as fuel have been investigated under RERTR program as possible candidate to became a new fuel for research reactor due to high density. This work aim to evaluate a set of criticality safety parameters for uranium metal alloy using SCALE system and MCNP Monte Carlo code. (author)

Remediation of uranium mill tailings wastes in Australia: a critical review

International Nuclear Information System (INIS)

Mudd, G.M.

2000-01-01

Australia has been an active participant in the global uranium mining industry since its inception in the 1940s. By the late 1950s five major mining and milling projects were operating, several small mines supplied custom ores. All of these projects were closed by the early 1960s, except for Rum Jungle which continued under government subsidy. Most sites have had lasting Environmental impacts. The advances in nuclear power in the 1960s saw increasing demand for uranium and Australia again explored with remarkable success in the Northern Territory, South Australia and Western Australia. After several government inquiries in the 1970s, Ranger, Nabarlek and Olympic Dam were operating by the mid 1980s. The principal risks from uranium mill tailings wastes arise from their radioactive nature and often their chemical toxicities. A critical review of the rehabilitation of abandoned uranium mines and mill tailings as a comparison for current projects is presented. It is concluded that the management of uranium mill tailings wastes is a complex task, requiring a sound multi-disciplinary approach. The problems include groundwater contamination, erosion, radon emanation and gamma radiation. evidence to data from the remediation of old and modern sites does not demonstrate effective long-term closure and safety

Graphite

Science.gov (United States)

Robinson, Gilpin R.; Hammarstrom, Jane M.; Olson, Donald W.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

2017-12-19

Graphite is a form of pure carbon that normally occurs as black crystal flakes and masses. It has important properties, such as chemical inertness, thermal stability, high electrical conductivity, and lubricity (slipperiness) that make it suitable for many industrial applications, including electronics, lubricants, metallurgy, and steelmaking. For some of these uses, no suitable substitutes are available. Steelmaking and refractory applications in metallurgy use the largest amount of produced graphite; however, emerging technology uses in large-scale fuel cell, battery, and lightweight high-strength composite applications could substantially increase world demand for graphite.Graphite ores are classified as “amorphous” (microcrystalline), and “crystalline” (“flake” or “lump or chip”) based on the ore’s crystallinity, grain-size, and morphology. All graphite deposits mined today formed from metamorphism of carbonaceous sedimentary rocks, and the ore type is determined by the geologic setting. Thermally metamorphosed coal is the usual source of amorphous graphite. Disseminated crystalline flake graphite is mined from carbonaceous metamorphic rocks, and lump or chip graphite is mined from veins in high-grade metamorphic regions. Because graphite is chemically inert and nontoxic, the main environmental concerns associated with graphite mining are inhalation of fine-grained dusts, including silicate and sulfide mineral particles, and hydrocarbon vapors produced during the mining and processing of ore. Synthetic graphite is manufactured from hydrocarbon sources using high-temperature heat treatment, and it is more expensive to produce than natural graphite.Production of natural graphite is dominated by China, India, and Brazil, which export graphite worldwide. China provides approximately 67 percent of worldwide output of natural graphite, and, as the dominant exporter, has the ability to set world prices. China has significant graphite reserves, and

Performance enhancement of spherical natural graphite by phenol resin in lithium ion batteries

International Nuclear Information System (INIS)

Wu, Y.-S.; Wang, Y.-H.; Lee, Y.-H.

2006-01-01

The capacity of natural graphite in the lithium ion battery anode decays seriously. The phenol resin is used as a reaction material to modify the electrochemical performance of spherical graphite as the anode material in lithium ion batteries. Measuring the reversible capacity indicates change in the surface structure of spherical graphite. A dense layer of methyl groups was thus formed. Some structural imperfections are removed and the stability of the graphite structure is increased. Clearly, reducing the irreversible capacity is beneficial in controlling the uniformity of the spherical graphite surface structure

The market for natural uranium

International Nuclear Information System (INIS)

Bauder, P.

1981-01-01

The natural uranium market is characterized at present by its surplus. This is essentially due to a surplus on the production line. The uranium produced is no longer taken up by the market as it was up to the middle of 1979. The object of this contribution is therefore a survey on the present availability and demand situation, as well as to discuss market mechanisms and forecast the future market trend. (orig./IHO) [de

Aspects on optimization of natural uranium fuel utilization in heavy water reactors

International Nuclear Information System (INIS)

1978-08-01

This paper is dealing with a possibility to decrease the natural uranium consumption of CANDU PHWR using the once-through cycle. This possibility is based on the utilization of slightly enriched uranium. The optimal two-zone structure of a reactor using natural uranium is found out. The optimal criterium is the maximization of the burnup (equivalent to minimization of uranium requirements) with a constraint on power density radial uniformity factor. As regards the enriched uranium, the optimal enrichment and the two-zone structure of a reactor which minimizes the natural uranium requirement with constraints on uniformity factor and maximum burnup are established. Corresponding to a maximum burnup of 16,000 MWd/t and 1% enrichment, the natural uranium requirement is found to be 10% less than that of the natural uranium reactor

Criticality safety considerations for MSRE fuel drain tank uranium aggregation

International Nuclear Information System (INIS)

Hollenbach, D.F.; Hopper, C.M.

1997-01-01

This paper presents the results of a preliminary criticality safety study of some potential effects of uranium reduction and aggregation in the Molten Salt Reactor Experiment (MSRE) fuel drain tanks (FDTs) during salt removal operations. Since the salt was transferred to the FDTs in 1969, radiological and chemical reactions have been converting the uranium and fluorine in the salt to UF 6 and free fluorine. Significant amounts of uranium (at least 3 kg) and fluorine have migrated out of the FDTs and into the off-gas system (OGS) and the auxiliary charcoal bed (ACB). The loss of uranium and fluorine from the salt changes the chemical properties of the salt sufficiently to possibly allow the reduction of the UF 4 in the salt to uranium metal as the salt is remelted prior to removal. It has been postulated that up to 9 kg of the maximum 19.4 kg of uranium in one FDT could be reduced to metal and concentrated. This study shows that criticality becomes a concern when more than 5 kg of uranium concentrates to over 8 wt% of the salt in a favorable geometry

Special graphites; Graphites speciaux

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [French] Ameliorer les proprietes du graphite nucleaire pour empilements et ouvrir de nouveaux domaines d'application au graphite constituent une part importante de l'effort entrepris en commun par le Commissariat a l'Energie Atomique (CEA) et la compagnie PECHINEY. Des procedes nouveaux de fabrication de carbones et graphites speciaux ont ete mis au point: graphite forge, pyrocarbone, graphite de haute densite, agglomeration de poudres de graphite par craquage de gaz naturel, graphites impermeables. Les proprietes physiques de ces produits ainsi que leur reaction avec differents gaz oxydants sont decrites. Les premiers resultats d'irradiation sont aussi donnes. (auteurs)

Buckling and reaction rate measurements in graphite moderated lattices fuelled with plutonium-uranium oxide clusters at temperatures up to 400 deg. C

Energy Technology Data Exchange (ETDEWEB)

Carter, D H; Gibson, M; King, D C; Marshall, J; Puckett, B J; Richards, A E; Wass, T; Wilson, D J [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1965-07-15

The Report describes a series of experiments carried out in SCORPIO I and II on sub-critical graphite moderated lattices fuelled with 21-rod clusters of PuO{sub 2}/UO{sub 2} fuel. Three fuel batches with nominal plutonium: uranium ratios of 0.25%, 0.8% and 1.2% were investigated at temperatures between 20 deg. C and 400 deg. C. Because of the limited amounts of the three fuels, exponential measurements were made in 2-zone stacks, the outer regions of which were loaded with suitably matched 'reference fuel'. Fine structure distributions in the lattice cell were obtained with manganese and indium foils. Pu239/U235 fission ratios were determined both by fission chambers and by fission-product counting techniques. (author) 14 refs, 30 figs, 18 tabs

Alecto - results obtained with homogeneous critical experiments on plutonium 239, uranium 235 and uranium 233

International Nuclear Information System (INIS)

Bruna, J.G.; Brunet, J.P.; Caizegues, R.; Clouet d'Orval, Ch.; Kremser, J.; Tellier, H.; Verriere, Ph.

1965-01-01

In this report are given the results of the homogeneous critical experiments ALECTO, made on plutonium 239, uranium 235 and uranium 233. After a brief description of the equipment, the critical masses for cylinders of diameters varying from 25 to 42 cm, are given and compared with other values (foreign results, criticality guide). With respect to the specific conditions of neutron reflection in the ALECTO experiments the minimal values of critical masses are: Pu239 M c = 910 ± 10 g, U235 M c = 1180 ± 12 g and U233 M c = 960 ± 10 g. Experiments relating to cross sections and constants to be used on these materials are presented. Lastly, kinetic experiments allow to compare pulsed neutron methods to fluctuation methods [fr

Alternative repository criticality-control strategies for fissile uranium wastes

International Nuclear Information System (INIS)

Forsberg, C.W.

1998-01-01

Methods to prevent long term, disposal site nuclear criticality from fissile uranium isotopes in wastes were investigated. Long term refers to the time period after waste package (WP) failure and the subsequent loss of geometry and chemistry control within the WP. The preferred method of control was found to be the addition of sufficient depleted uranium to each WP so that the uranium enrichment is reduced to 235 U and 233 U in 238 U

Transmutation of uranium and thorium in the particle field of the Quinta sub-critical assembly

Science.gov (United States)

Hashemi-Nezhad, S. R.; Asquith, N. L.; Voronko, V. A.; Sotnikov, V. V.; Zhadan, Alina; Zhuk, I. V.; Potapenko, A.; Husak, Krystsina; Chilap, V.; Adam, J.; Baldin, A.; Berlev, A.; Furman, W.; Kadykov, M.; Khushvaktov, J.; Kudashkin, I.; Mar'in, I.; Paraipan, M.; Pronskih, V.; Solnyshkin, A.; Tyutyunnikov, S.

2018-03-01

The fission rates of natural uranium and thorium were measured in the particle field of Quinta, a 512 kg natural uranium target-blanket sub-critical assembly. The Quinta assembly was irradiated with deuterons of energy 4 GeV from the Nuclotron accelerator of the Joint Institute for Nuclear Research (JINR), Dubna, Russia. Fission rates of uranium and thorium were measured using Gamma spectroscopy and fission track techniques. The production rate of 239Np was also measured. The obtained experimental results were compared with Monte Carlo predictions using the MCNPX 2.7 code employing the physics and fission-evaporation models of INCL4-ABLA, CEM03.03 and LAQGSM03.03. Some of the neutronic characteristics of the Quinta are compared with the "Energy plus Transmutation (EpT)" subcritical assembly, which is composed of a lead target and natU blanket. This comparison clearly demonstrates the importance of target material, neutron moderator and reflector types on the performance of a spallation neutron driven subcritical system. As the dimensions of the Quinta are very close to those of an optimal multi-rod-uranium target, the experimental and Monte Carlo calculation results presented in this paper provide insights on the particle field within a uranium target as well as in Accelerator Driven Systems in general.

Analysis of Wigner energy release process in graphite stack of shut-down uranium-graphite reactor

OpenAIRE

Bespala, E. V.; Pavliuk, A. O.; Kotlyarevskiy, S. G.

2015-01-01

Data, which finding during thermal differential analysis of sampled irradiated graphite are presented. Results of computational modeling of Winger energy release process from irradiated graphite staking are demonstrated. It's shown, that spontaneous combustion of graphite possible only in adiabatic case.

Single nanocrystals of uranium dicarbide encapsulated in carbon

International Nuclear Information System (INIS)

Pasqualini, Enrique; Adelfang, Pablo

1996-01-01

The simultaneous volatilization of carbon and uranium compounds in an inert atmosphere produces the encapsulation of uranium dicarbide in graphitic material. These composite particles are of nanoscopic dimensions (smaller than 100 nanometers) and have the appearance of a black soot. Nanocapsules are as chemically inert as graphite and providers a more secure handling and processing of nuclear toxic materials. They preserve uranium dicarbide from environmental decomposition. The kernel of the nanoparticles is a single crystal of uranium dicarbide. The covers of the crystal are of graphitic structures of two types: parallel graphite layers and randomly oriented graphite crystallites. Both type of covers surround the UC 2 nanocrystal kernel. Density separation and measurements of specific area (BET technique) of purified soot were done. Particular adsorption and desorption kinetics of nitrogen monolayers indicate the presence of nanoparticles agglomeration. This is confirmed by direct observation by transmission electron microscopy (TEM). From density measurement and high resolution transmission electron microscopy (HRTEM) observation it can be inferred that there are gaps between kernel and cover and probably closed porosity inside clusters. Raman spectroscopy indicates only the presence of graphitic type carbon. Rietveld refinement of the nanoencapsulated material helps in the interpretation of X-Ray diffraction spectra. (author)

Single nanocrystals of uranium dicarbide encapsulated in carbon

Energy Technology Data Exchange (ETDEWEB)

Pasqualini, Enrique; Adelfang, Pablo [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Dept. de Combustibles Nucleares

1996-07-01

The simultaneous volatilization of carbon and uranium compounds in an inert atmosphere produces the encapsulation of uranium dicarbide in graphitic material. These composite particles are of nanoscopic dimensions (smaller than 100 nanometers) and have the appearance of a black soot. Nanocapsules are as chemically inert as graphite and providers a more secure handling and processing of nuclear toxic materials. They preserve uranium dicarbide from environmental decomposition. The kernel of the nanoparticles is a single crystal of uranium dicarbide. The covers of the crystal are of graphitic structures of two types: parallel graphite layers and randomly oriented graphite crystallites. Both type of covers surround the UC{sub 2} nanocrystal kernel. Density separation and measurements of specific area (BET technique) of purified soot were done. Particular adsorption and desorption kinetics of nitrogen monolayers indicate the presence of nanoparticles agglomeration. This is confirmed by direct observation by transmission electron microscopy (TEM). From density measurement and high resolution transmission electron microscopy (HRTEM) observation it can be inferred that there are gaps between kernel and cover and probably closed porosity inside clusters. Raman spectroscopy indicates only the presence of graphitic type carbon. Rietveld refinement of the nanoencapsulated material helps in the interpretation of X-Ray diffraction spectra. (author)

Untreated Natural Graphite as a Graphene Source for High-Performance Li-Ion Batteries

Directory of Open Access Journals (Sweden)

María Simón

2018-03-01

Full Text Available Graphene nanosheets (GNS are synthesized from untreated natural graphite (NG for use as electroactive materials in Li-ion batteries (LIBs, which avoids the pollution-generating steps of purifying graphite. Through a modified Hummer method and subsequent thermal exfoliation, graphitic oxide and graphene were synthesized and characterized structurally, morphologically and chemically. Untreated natural graphite samples contain 45–50% carbon by weight; the rest is composed of different elements such as aluminium, calcium, iron, silicon and oxygen, which are present as calcium carbonate and silicates of aluminium and iron. Our results confirm that in the GO and GNS synthesized, calcium is removed due to oxidation, though other impurities are maintained because they are not affected by the synthesis. Despite the remaining mineral phases, the energy storage capacity of GNS electrodes is very promising. In addition, an electrochemical comparison between GNS and NG demonstrated that the specific capacity in GNS is higher during the whole cycling process, 770 mA·g−1 at 100th cycle, which is twice that of graphite.

Sorption of natural uranium by algerian bentonite

International Nuclear Information System (INIS)

Megouda, N.; Kadi, H.; Hamla, M.S.; Brahimi, H.

2004-01-01

Full text.Batch sorption experiments have been used to assess the sorption behaviour of uranium onto natural and drilling bentonites. The operating parameters (pH, aolis-liquid ratio, particle size, time and initial uranium concentration) influenced the rate of adsorption. The distribution coefficient (Kd) range values at equilibrium time are 45.95-1079.26 ml/g and 32.81-463053 ml/g for the drilling and natural bentonites respectively. The equilibrium isotherms show that the data correlate with both Freundlich and Langmuir models

LiFePO4/polymer/natural graphite: low cost Li-ion batteries

International Nuclear Information System (INIS)

Zaghib, K.; Striebel, K.; Guerfi, A.; Shim, J.; Armand, M.; Gauthier, M

2004-01-01

The aging and performance of natural graphite/PEO-based gel electrolyte/LiFePO 4 cells are reported. The gel polymer electrolytes were produced by electron-beam irradiation and then soaked in a liquid electrolyte. The natural graphite anode in gel electrolyte containing LiBF4-EC/GBL exhibited high reversible capacity (345 mAh/g) and high coulombic efficiency (91%). The LiFePO 4 cathode in the same gel-polymer exhibited a reversible capacity of 160 mAh/g and 93% coulombic efficiency. Better performance was obtained at high-rate discharge with 6% carbon additive in the cathode, however the graphite anode performance suffers at high rate. The Li-ion gel polymer battery shows a capacity fade of 13% after 180 cycles and has poor performance at low temperature due to low diffusion of the lithium to the graphite in the GBL system. The LiFePO 4 /gel/Li system has an excellent rate capacity. LiFePO 4 cathode material is suitable for HEV application

Chemically reduced graphene contains inherent metallic impurities present in parent natural and synthetic graphite

Science.gov (United States)

Ambrosi, Adriano; Chua, Chun Kiang; Khezri, Bahareh; Sofer, Zdeněk; Webster, Richard D.; Pumera, Martin

2012-01-01

Graphene-related materials are in the forefront of nanomaterial research. One of the most common ways to prepare graphenes is to oxidize graphite (natural or synthetic) to graphite oxide and exfoliate it to graphene oxide with consequent chemical reduction to chemically reduced graphene. Here, we show that both natural and synthetic graphite contain a large amount of metallic impurities that persist in the samples of graphite oxide after the oxidative treatment, and chemically reduced graphene after the chemical reduction. We demonstrate that, despite a substantial elimination during the oxidative treatment of graphite samples, a significant amount of impurities associated to the chemically reduced graphene materials still remain and alter their electrochemical properties dramatically. We propose a method for the purification of graphenes based on thermal treatment at 1,000 °C in chlorine atmosphere to reduce the effect of such impurities on the electrochemical properties. Our findings have important implications on the whole field of graphene research. PMID:22826262

Chemically reduced graphene contains inherent metallic impurities present in parent natural and synthetic graphite.

Science.gov (United States)

Ambrosi, Adriano; Chua, Chun Kiang; Khezri, Bahareh; Sofer, Zdeněk; Webster, Richard D; Pumera, Martin

2012-08-07

Graphene-related materials are in the forefront of nanomaterial research. One of the most common ways to prepare graphenes is to oxidize graphite (natural or synthetic) to graphite oxide and exfoliate it to graphene oxide with consequent chemical reduction to chemically reduced graphene. Here, we show that both natural and synthetic graphite contain a large amount of metallic impurities that persist in the samples of graphite oxide after the oxidative treatment, and chemically reduced graphene after the chemical reduction. We demonstrate that, despite a substantial elimination during the oxidative treatment of graphite samples, a significant amount of impurities associated to the chemically reduced graphene materials still remain and alter their electrochemical properties dramatically. We propose a method for the purification of graphenes based on thermal treatment at 1,000 °C in chlorine atmosphere to reduce the effect of such impurities on the electrochemical properties. Our findings have important implications on the whole field of graphene research.

Long-term criticality control in radioactive waste disposal facilities using depleted uranium

International Nuclear Information System (INIS)

Forsberg, C.W.

1997-01-01

Plant photosynthesis has created a unique planetary-wide geochemistry - an oxidizing atmosphere with oxidizing surface waters on a planetary body with chemically reducing conditions near or at some distance below the surface. Uranium is four orders of magnitude more soluble under chemically oxidizing conditions than it is under chemically reducing conditions. Thus, uranium tends to leach from surface rock and disposal sites, move with groundwater, and concentrate where chemically reducing conditions appear. Earth's geochemistry concentrates uranium and can separate uranium from all other elements except oxygen, hydrogen (in water), and silicon (silicates, etc). Fissile isotopes include 235 U, 233 U, and many higher actinides that eventually decay to one of these two uranium isotopes. The potential for nuclear criticality exists if the precipitated uranium from disposal sites has a significant fissile enrichment, mass, and volume. The earth's geochemistry suggests that isotopic dilution of fissile materials in waste with 238 U is a preferred strategy to prevent long-term nuclear criticality in and beyond the boundaries of waste disposal facilities because the 238 U does not separate from the fissile uranium isotopes. Geological, laboratory, and theoretical data indicate that the potential for nuclear criticality can be minimized by diluting fissile materials with- 238 U to 1 wt % 235 U equivalent

Uranium Bio-accumulation and Cycling as revealed by Uranium Isotopes in Naturally Reduced Sediments from the Upper Colorado River Basin

Science.gov (United States)

Lefebvre, Pierre; Noël, Vincent; Jemison, Noah; Weaver, Karrie; Bargar, John; Maher, Kate

2016-04-01

Uranium (U) groundwater contamination following oxidized U(VI) releases from weathering of mine tailings is a major concern at numerous sites across the Upper Colorado River Basin (CRB), USA. Uranium(IV)-bearing solids accumulated within naturally reduced zones (NRZs) characterized by elevated organic carbon and iron sulfide compounds. Subsequent re-oxidation of U(IV)solid to U(VI)aqueous then controls the release to groundwater and surface water, resulting in plume persistence and raising public health concerns. Thus, understanding the extent of uranium oxidation and reduction within NRZs is critical for assessing the persistence of the groundwater contamination. In this study, we measured solid-phase uranium isotope fractionation (δ238/235U) of sedimentary core samples from four study sites (Shiprock, NM, Grand Junction, Rifle and Naturita, CO) using a multi-collector inductively coupled plasma mass spectrometer (MC-ICP-MS). We observe a strong correlation between U accumulation and the extent of isotopic fractionation, with Δ238U up to +1.8 ‰ between uranium-enriched and low concentration zones. The enrichment in the heavy isotopes within the NRZs appears to be especially important in the vadose zone, which is subject to variations in water table depth. According to previous studies, this isotopic signature is consistent with biotic reduction processes associated with metal-reducing bacteria. Positive correlations between the amount of iron sulfides and the accumulation of reduced uranium underline the importance of sulfate-reducing conditions for U(IV) retention. Furthermore, the positive fractionation associated with U reduction observed across all sites despite some variations in magnitude due to site characteristics, shows a regional trend across the Colorado River Basin. The maximum extent of 238U enrichment observed in the NRZ proximal to the water table further suggests that the redox cycling of uranium, with net release of U(VI) to the groundwater by

The light water natural uranium reactor

International Nuclear Information System (INIS)

Radkowsky, A.

A new type of light water seed blanket with the seed having 20% enrichment and the blanket a special combination of elements of natural uranium and thorium, relatively close packed, but sufficient spacing for heat transfer purpose is described. The blanket would deliver approximately half the total energy for about 10,000 MWDIT, so this type of core would be just as economical or better in uranium ore consumation as present cores. (author)

Hierarchically porous graphene in natural graphitic globules from silicate magmatic rocks

OpenAIRE

PONOMARCHUK V.A.; TITOV A.T.; MOROZ T.N.; PYRYAEV A.N.; PONOMARCHUK A.V.

2014-01-01

Naturally-occurring nanostructured graphites from silicate magmatic rocks, which are rare, were characterized using electron microscope and X-ray spectroscopy. This graphite consists of porous carbon, nanographite layers, microand nanotubes. The porous carbon is classified as macroporous matter with a small amount of mezopores. Evidence for the unusual properties of porous carbon are given: nanographite layers are created at the exposed surface of sample and the nanotubes occurs in the bulk o...

Structural characteristics of a graphite moderated critical assembly for a Zero Power reactor at IEA (Brazil)

International Nuclear Information System (INIS)

Almeida Ferreira, A.C. de; Hukai, R.Y.

1975-01-01

The structural characteristics of a graphite moderated core of a critical assembly to be installed in the Zero Power Reactor of IEA have been defined. These characteristics are the graphite block dimensions, the number and dimensions of the holes in the graphite, the pitch, the dimensions of the sticks of fuel and graphite to be inserted in the holes, and the mechanical reproducibility of the system. The composition of the fuel and moderator sticks were also defined. The main boundary conditions were the range of the relation C/U and C/TH used in commercial HTGR and the neutronics homogeneity

Electron transfer kinetics on natural crystals of MoS2 and graphite.

Science.gov (United States)

Velický, Matěj; Bissett, Mark A; Toth, Peter S; Patten, Hollie V; Worrall, Stephen D; Rodgers, Andrew N J; Hill, Ernie W; Kinloch, Ian A; Novoselov, Konstantin S; Georgiou, Thanasis; Britnell, Liam; Dryfe, Robert A W

2015-07-21

Here, we evaluate the electrochemical performance of sparsely studied natural crystals of molybdenite and graphite, which have increasingly been used for fabrication of next generation monolayer molybdenum disulphide and graphene energy storage devices. Heterogeneous electron transfer kinetics of several redox mediators, including Fe(CN)6(3-/4-), Ru(NH3)6(3+/2+) and IrCl6(2-/3-) are determined using voltammetry in a micro-droplet cell. The kinetics on both materials are studied as a function of surface defectiveness, surface ageing, applied potential and illumination. We find that the basal planes of both natural MoS2 and graphite show significant electroactivity, but a large decrease in electron transfer kinetics is observed on atmosphere-aged surfaces in comparison to in situ freshly cleaved surfaces of both materials. This is attributed to surface oxidation and adsorption of airborne contaminants at the surface exposed to an ambient environment. In contrast to semimetallic graphite, the electrode kinetics on semiconducting MoS2 are strongly dependent on the surface illumination and applied potential. Furthermore, while visibly present defects/cracks do not significantly affect the response of graphite, the kinetics on MoS2 systematically accelerate with small increase in disorder. These findings have direct implications for use of MoS2 and graphene/graphite as electrode materials in electrochemistry-related applications.

Comparisons of theoretical and experimental neutron spectra, 115In(n,n') and fission rates, in the centre of three spherical natural uranium and iron shell configurations, located at BR1

International Nuclear Information System (INIS)

De Leeuw-Gierts, G.; De Leeuw, S.; Gilliam, D.M.

1984-01-01

Three spherical configurations of iron and uranium shells have been studied. The configurations were a 1-cm thick natural uranium shell, a 1-cm thick natural uranium shell with an inner 7-cm thick iron shell and a 1-cm thick natural uranium shell with an inner iron shell of 14-cm thickness. For the measurements, the shells were located at the centre of a hollow cavity, 100-cm in diameter, in the vertical graphite thermal column of the BR1 reactor. The central neutron spectra were calculated by means of the DTF-IV code, using the 208-group KEDAK-3 library, and by means of the ANISN code, using the 171-group VITAMIN-C library. Central neutron spectra, measured by the proton-recoil and 6 Li(n,α)t spectrometry techniques, are compared to the theory between ∼ 100 keV and 5 MeV. Mean fission cross-sections of 240 Pu, 237 Np, 234 U, 235 U, 236 U and 238 U were deduced from the calculations. Their ratios with respect to 238 U are compared to measurements made with NBS dual fission chambers. (Auth.)

Criticality accident in uranium fuel processing plant. The estimation of the total number of fissions with related reactor physics parameters

International Nuclear Information System (INIS)

Nishina, Kojiro; Oyamatsu, Kazuhiro; Kondo, Shunsuke; Sekimoto, Hiroshi; Ishitani, Kazuki; Yamane, Yoshihiro; Miyoshi, Yoshinori

2000-01-01

This accident occurred when workers were pouring a uranium solution into a precipitation tank with handy operation against the established procedure and both the cylindrical diameter and the total mass exceeded the limited values. As a result, nuclear fission chain reactor in the solution reached not only a 'criticality' state continuing it independently but also an instantly forming criticality state exceed the criticality and increasing further nuclear fission number. The place occurring the accident at this time was not reactor but a place having not to form 'criticality' called by a processing process of uranium fuel. In such place, as because of relating to mechanism of chain reaction, it is required naturally for knowledge on the reactor physics, it is also necessary to understand chemical reaction in chemical process, and functions of tanks, valves and pumps mounted at the processes. For this purpose, some information on uranium concentration ratio, atomic density of nuclides largely affecting to chain reaction such as uranium, hydrogen, and so forth in the solution, shape, inner structure and size of container for the solution, and its temperature and total volume, were necessary for determining criticality volume of the accident uranium solution by using nuclear physics procedures. Here were described on estimation of energy emission in the JCO accident, estimation from analytical results on neutron and solution, calculation of various nuclear physics property estimation on the JCO precipitation tank at JAERI. (G.K.)

Lead determination in uranium mineralization soils by atomic absorption spectrometry with graphite oven

International Nuclear Information System (INIS)

Teixeira, Gleber Tacio

2001-01-01

The contamination of soils by lead has a great environmental importance due to its toxicity to vegetables, animals and humans. In general, the mobility of the lead is small due to its low solubility and strong adsorption in the soil. However, its solubility can be altered by several conditions (pH, redox potential and ionic stronger). Consequently, lead can migrate through the soil and can contaminate superficial and underground waters. The objective of this work was to determine the concentration of total lead in soil samples with uranium mineralization, in an area at Ipora/GO, having been evaluated as economically insuitable the extraction of that mineral. The radiogenic lead appears as a product of natural radioactive elements decay. In the decay series of uranium-238 we found the isotope lead-214 (half-life of 26,8 min), lead-210 (half-life of 22,3 min), and lead-206 that is stable. The sampling was done in profiles around north, south, east and west directions, starting from a reference point (FT), chosen by presenting the largest radiation of that place (4800 cps). A mass of 1 Kg of superficial soil was collected to each 20 m, in each profile, until 150 m of FT. Approximately, 1 g of dry soil, fraction 2 mm, was digested with a mixture of acids HNO 3 /HClO 4 2:1 (v/v), and the resulting solution was analyzed by atomic absorption. An atomic absorption spectrometer was used with graphite furnace, with deuterium arc to background correction and pyrolytic coated tube. Phosphoric acid was used as chemical modifier. The obtained results, using the standard additions method, presented a decrease of the lead concentration, in all profiles, when the distance of FT was increased. It was also made a radiometric screening in each sampling point. The lead concentration variate from 115,1 μg.g -1 in FT, to less than 40 μg.g -1 at 150 m of distance of FT ( 3 ) 2 was used. The method was applied to a certified sample, showing a good agreement between certified and

Long-term outlook for global natural uranium and uranium enrichment supply and demand situations after the impact of Fukushima Daiichi Nuclear Power Plant accident

International Nuclear Information System (INIS)

Matsuo, Yuhji; Murakami, Tomoko

2012-01-01

In this paper, the authors propose long-term projections of global nuclear power generation, uranium production, and uranium enrichment capacities by region, and estimate the trade flows of natural uranium and uranium enrichment activities in 2020 and 2035. In spite of the rapid nuclear power generation capacity growth expected especially in Asia, the natural uranium and uranium enrichment trade will not be tightened by 2020 due to the projected increase in both natural uranium production and uranium enrichment capacities, which may cause a drop in natural uranium and uranium enrichment prices. Thus, there is a great possibility that the current projects for capacity expansion will be delayed considerably. However, in the 'high-demand scenario', where nuclear expansion will be accelerated due to growing concerns about global warming and energy security issues, additional investments in uranium production and enrichment facilities will be needed by 2035. In Asia, the self-sufficiency ratio for both natural uranium supply and uranium enrichment activities will remain relatively low until 2035. However, the Herfindahl-Hirschman (HH) index of natural uranium and uranium enrichment activity trade to Asia will be lowered considerably up to 2035, indicating that nuclear capacity expansion can contribute to enhancing energy security in Asia. (author)

Immobilization of uranium from aqueous solutions by using natural diatomites

International Nuclear Information System (INIS)

Mokhambetbakr, Kh.E.; Burkitbaev, M.

2008-01-01

In this study, the adsorption of uranium on natural diatomite (as high abundant and low-cost material) obtained from Aktyubinsk (Kazakhstan) has been investigated. The main purpose of this work is the immobilization of uranium from liquid waste by using diatomites. The diatomites under study were subjected to treatment with various conditions. The first sample is the natural sample (D) Natural Diatomite, the second (D H CL) is purified 0,5 N HCl and the third is the Calcined Diatomite (D 9 00). The effects of concentration of uranium, contact time and type of diatomite treatment on the adsorption process were examined.

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

Criticality concerns in cleaning large uranium hexafluoride cylinders

International Nuclear Information System (INIS)

Sheaffer, M.K.; Keeton, S.C.; Lutz, H.F.

1995-06-01

Cleaning large cylinders used to transport low-enriched uranium hexafluoride (UF 6 ) presents several challenges to nuclear criticality safety. This paper presents a brief overview of the cleaning process, the criticality controls typically employed and their bases. Potential shortfalls in implementing these controls are highlighted, and a simple example to illustrate the difficulties in complying with the Double Contingency Principle is discussed. Finally, a summary of recommended criticality controls for large cylinder cleaning operations is presented

A preliminary definition of the parameters of an experimental natural - uranium, graphite - moderated, helium - cooled power reactor

International Nuclear Information System (INIS)

Baltazar, O.

1978-01-01

A preliminary study of the technical characteristic of an experiment at 32 MWe power with a natural uconium, graphite-moderated, helium cooled reactor is described. The national participation and the use of reactor as an instrument for the technological development of future high temperature gas cooled reactor is considered in the choice of the reactor type. Considerations about nuclear power plants components based in extensive bibliography about similar english GCR reactor is presented. The main thermal, neutronic an static characteristic and in core management of the nuclear fuel is stablished. A simplified scheme of the secondary system and its thermodynamic performance is determined. A scheme of parameters calculation of the reactor type is defined based in the present capacity of calculation developed by Coordenadoria de Engenharia Nuclear and Centro de Processamento de Dados, IEA, Brazil [pt

Obtention of uranium-molybdenum alloy ingots technique to avoid carbon contamination

Energy Technology Data Exchange (ETDEWEB)

Pedrosa, Tercio A.; Paula, Joao Bosco de; Reis, Sergio C.; Brina, Jose Giovanni M.; Faeda, Kelly Cristina M.; Ferraz, Wilmar B., E-mail: tap@cdtn.b, E-mail: jbp@cdtn.b, E-mail: jgmb@cdtn.b, E-mail: ferrazw@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2011-07-01

The replacement of high enriched uranium (U{sup 235} > 85 wt%) by low enriched uranium (U{sup 235} < 20wt%) nuclear fuels in research and test reactors is being implemented as an initiative of the Reduced Enrichment for Research and Test Reactors (RERTR) program, conceived in the USA since mid-70s, in order to avoid nuclear weapons proliferation. Such replacement implies in the use of compounds or alloys with higher uranium densities. Among the several uranium alloys investigated since then, U-Mo presents great application potential due to its physical properties and good behavior during irradiation, which makes it an important option as a nuclear fuel material for the Brazilian Multipurpose Reactor - RMB. The development of the plate-type nuclear fuel based on U-Mo alloy is being performed at the Nuclear Technology Development Centre (CDTN) and also at IPEN. The carbon contamination of the alloy is one of the great concerns during the melting process. It was observed that U-Mo alloy is more critical considering carbon contamination when using graphite crucibles. Alternative melting technique was implemented at CDTN in order to avoid carbon contamination from graphite crucible using Yttria stabilized ZrO{sub 2} crucibles. Ingots with low carbon content and good internal quality were obtained. (author)

Mineralogical study of uraniferous graphitic ore from Deogpyeong, Mogso and southern part of Daejeon area

Energy Technology Data Exchange (ETDEWEB)

Lee, D J; Nam, S K [Korean Inst. of Energy and Resources, Seoul (Republic of Korea)

1981-11-01

Uranium minerals of torbernite, metatorbernite, metatyuyamunite and autunite have been identified from the uraniferous ores in graphitic beds of Ogcheon Group in Deogpyeong, Mogso and southern part of Daejeon area. Polarizing and ore microscopic studies, and chemical and X-ray powder diffraction analyses were carried out on the uraniferous graphite and associated materials. Main component minerals of uraniferous samples are graphite and quartz. Minor minerals are calcite, muscovite, sericite, andalusite, barite, kaolinite, hyaline opal, uranium minerals, sulfides such as pyrite, chalcopyrite, zincblende, and pyrrhotite, limonite, zeolite minerals such as laumontite and heulandite. Metatyuyamunite, torbernite, metatorbernite and autunite generally occur together with secondary minerals such as kaolinite, hyaline opal, calcite and limonite. They were found along the minor fissures or on the surface. Secondary uranium minerals described above were formed by supergenetic origin from primary uranium mineral. Uraniferous phosphatic nodule from Deogpyeong area are mainly composed of graphite and fluorapatite. And minor minerals are barite, quartz, muscovite and pyrite. Autoradiograph from uraniferous nodule shows that uranium enrichment in outer part of nodules is much higher than in inner part. This feature coincides with chemical analyses data of this uraniferous nodule.

Critical experiments on low-enriched uranium oxide system with H/U=1.25

International Nuclear Information System (INIS)

Oh, I.; Rothe, R.E.; Tuck, G.

1982-01-01

Fifteen (15) critical experiments were performed on a horizontal split table machine using 4.48%-enriched sup(235)U uranium oxide(U 3 O 8 ). The oxide was compacted to a density of 4.68g/cm 3 and placed in 152 mm cubical aluminum cans. Water was added to achive an H/U of 1.25. Various arrays of oxide cans were distributed on each half of the split table, and the separation between halves reduced until criticality occurred. The critical table separation varied from 3.59 mm to 18.40 mm. Twelve (12) experiments required the addition of a high-enriched(-93 %sup(235)U) metal or solution driver to achieve criticality. These experiments were performed in a plastic, concrete, or thin steel reflector. Three additional experiments in the plastic reflector contained either 9.3-mm- or 24.3-mm-thick plastic moderator material between the oxide cans and did not require a driver to achieve criticality. Critical uranium driver masses ranged from 9.999 kg to 14.000 kg (solution driver), and from 25.378 kg to 29.278 kg (metal driver) for 5X5X5 arrays of uranium oxide cans. Always, one or four of these 125 cans had to be removed to make room for the drivers. Therefore, the uranium oxide masses used were 1823.8 kg and 1863.5 kg. For the moderated experiments, the uranium oxide mass ranged between 574.4 kg and 1210.0 kg. (Author)

Highly Enriched Uranium Metal Cylinders Surrounded by Various Reflector Materials

International Nuclear Information System (INIS)

Bernard Jones; J. Blair Briggs; Leland Monteirth

2007-01-01

A series of experiments was performed at Los Alamos Scientific Laboratory in 1958 to determine critical masses of cylinders of Oralloy (Oy) reflected by a number of materials. The experiments were all performed on the Comet Universal Critical Assembly Machine, and consisted of discs of highly enriched uranium (93.3 wt.% 235U) reflected by half-inch and one-inch-thick cylindrical shells of various reflector materials. The experiments were performed by members of Group N-2, particularly K. W. Gallup, G. E. Hansen, H. C. Paxton, and R. H. White. This experiment was intended to ascertain critical masses for criticality safety purposes, as well as to compare neutron transport cross sections to those obtained from danger coefficient measurements with the Topsy Oralloy-Tuballoy reflected and Godiva unreflected critical assemblies. The reflector materials examined in this series of experiments are as follows: magnesium, titanium, aluminum, graphite, mild steel, nickel, copper, cobalt, molybdenum, natural uranium, tungsten, beryllium, aluminum oxide, molybdenum carbide, and polythene (polyethylene). Also included are two special configurations of composite beryllium and iron reflectors. Analyses were performed in which uncertainty associated with six different parameters was evaluated; namely, extrapolation to the uranium critical mass, uranium density, 235U enrichment, reflector density, reflector thickness, and reflector impurities. In addition to the idealizations made by the experimenters (removal of the platen and diaphragm), two simplifications were also made to the benchmark models that resulted in a small bias and additional uncertainty. First of all, since impurities in core and reflector materials are only estimated, they are not included in the benchmark models. Secondly, the room, support structure, and other possible surrounding equipment were not included in the model. Bias values that result from these two simplifications were determined and associated

The measurements of critical mass with uranium fuel elements and thorium rods

International Nuclear Information System (INIS)

Yao Zhiquan; Chen Zhicheng; Yao Zewu; Ji Huaxiang; Bao Borong; Zhang Jiahua

1991-01-01

The critical experiments with uranium elements and Thorium rods have been performed in zero power reactor at Shanghai Institute of Nuclear Research. The critical masses have been measured in various U/Th ratios. The fuels are 3% 235 U-enriched uranium. The Thorium rods are made from power of ThF 4 . Ratios of calculated values to experimental values are nearly constant at 0.995

The use of graphite for the reduction of void reactivity in CANDU reactors

International Nuclear Information System (INIS)

Min, B.J.; Kim, B.G.; Sim, K-S.

1995-01-01

Coolant void reactivity can be reduced by using burnable poison in CANDU reactors. The use of graphite in the fuel bundle is introduced to reduce coolant void reactivity by adding an appropriate amount of burnable poison in the central rod. This study shows that sufficiently low void reactivity which in controllable by Reactor Regulating System (RRS) can be achieved by using graphite used fuel with slightly enriched uranium. Zero void reactivity can be also obtained by using graphite used fuel with a large central rod. A new fuel bundle with graphite rods can substantially reduce the void reactivity with less burnup penalty compared to previously proposed low void reactivity fuel with depleted uranium. (author)

Advanced Surface and Microstructural Characterization of Natural Graphite Anodes for Lithium Ion Batteries

Energy Technology Data Exchange (ETDEWEB)

Gallego, Nidia C [ORNL; Contescu, Cristian I [ORNL; Meyer III, Harry M [ORNL; Howe, Jane Y [ORNL; Meisner, Roberta Ann [ORNL; Payzant, E Andrew [ORNL; Lance, Michael J [ORNL; Yoon, Steve [A123 Systems, Inc.; Denlinger, Matthew [A123 Systems, Inc.; Wood III, David L [ORNL

2014-01-01

Natural graphite powders were subjected to a series of thermal treatments in order to improve the anode irreversible capacity loss (ICL) and capacity retention during long-term cycling of lithium ion batteries. A baseline thermal treatment in inert Ar or N2 atmosphere was compared to cases with a proprietary additive to the furnace gas environment. This additive substantially altered the surface chemistry of the natural graphite powders and resulted in significantly improved long-term cycling performance of the lithium ion batteries over the commercial natural graphite baseline. Different heat-treatment temperatures were investigated ranging from 950-2900 C with the intent of achieving the desired long-term cycling performance with as low of a maximum temperature and thermal budget as possible. A detailed summary of the characterization data is also presented, which includes X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), Raman spectroscopy, and temperature-programed desorption mass spectroscopy (TPD-MS). This characterization data was correlated to the observed capacity fade improvements over the course of long-term cycling at high charge-discharge rates in full lithium-ion coin cells. It is believed that the long-term performance improvements are a result of forming a more stable solid electrolyte interface (SEI) layer on the anode graphite surfaces, which is directly related to the surface chemistry modifications imparted by the proprietary gas environment during thermal treatment.

Special graphites

International Nuclear Information System (INIS)

Leveque, P.

1964-01-01

A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [fr

World nuclear fuel supply and demand prospects until 2030. Analysis of demand change factor of natural uranium and uranium separation work and its influence

International Nuclear Information System (INIS)

Murakami, Tomoko

2007-01-01

World nuclear power generation continues to spread gently until 2030 from the viewpoint of increase of the electricity demand around Asia, stable energy supply and anti-global warming measure, and the natural uranium demand is predicted to be increased from about 67 ktU in 2004 to 80-100 ktU in 2030. Steps of conversion/separation/reconversion/molding processing of the natural uranium are necessary for nuclear fuel, and the separation work of those is important because it needs high technology. There is a relation of the trade-off through the tale density (0.3% as a standard) between natural uranium and separation work demand. Therefore an analysis was performed of the influence on natural uranium and separation work demand by the change of the tale density and the influence on natural uranium supply and demand prospects by the recovery uranium use. In conclusion it was very likely that the supply and demand of separation work was tight at 0.2%-0.1% as for the cost of most suitable tale density which would appear earlier than natural uranium one and that the recovery uranium could become the backup of the natural uranium. (T. Tanaka)

Role of organic matter in the Proterozoic Oklo natural fission reactors, Gabon, Africa

International Nuclear Information System (INIS)

Nagy, B.; Rigali, M.J.; Gauthier-Lafaye, F.; Holliger, P.; Mossman, D.J.; Leventhal, J.S.

1993-01-01

Of the sixteen known Oklo and the Bangombe natural fission reactors (hydrothermally altered elastic sedimentary rocks that contain abundant uraninite and authigenic clay minerals), reactors 1 to 6 at Oklo contain only traces of organic matter, but the others are rich in organic substances. Reactors 7 to 9 are the subjects of this study. These organic-rich reactors may serve as time-tested analogues for anthropogenic nuclear-waste containment strategies. Organic matter helped to concentrate quantities of uranium sufficient to initiate the nuclear chain reactions. Liquid bitumen was generated from organic matter by hydrothermal reactions during nuclear criticality. The bitumen soon became a solid, consisting of polycyclic aromatic hydrocarbons and an intimate mixture of cryptocrystalline graphite, which enclosed and immobilized uraninite and the fission-generated isotopes entrapped in uraninite. This mechanism prevented major loss of uranium and fission products from the natural nuclear reactors for 1.2 b.y. 24 refs., 4 figs

The size distribution of dissolved uranium in natural waters

International Nuclear Information System (INIS)

Mann, D.K.; Wong, G.T.F.

1987-01-01

The size distribution of dissolved uranium in natural waters is poorly known. Some fraction of dissolved uranium is known to associate with organic matter which had a wide range of molecular weights. The presence of inorganic colloidal uranium has not been reported. Ultrafiltration has been used to quantify the size distribution of a number of elements, such as dissolved organic carbon, selenium, and some trace metals, in both the organic and/or the inorganic forms. The authors have applied this technique to dissolved uranium and the data are reported here

Critical and sub-critical experiments on U-BeO lattices

International Nuclear Information System (INIS)

Benoist, P.; Gourdon, Ch.; Martelly, J.; Sagot, M.; Wanner, G.

1958-01-01

Sub-critical experiments have allowed us to measure the material buckling of uranium natural oxide of beryllium lattices with a grid of 15 cm, and made up of uranium bars measuring 2.60 - 2.92 - 3.56 and 4.40 cm of diameter. A critical experiment has then been conducted with hollow 1.35 per cent enriched uranium bars. A study of U-BeO 18.03 cm grid lattices is at present being conducted. (author) [fr

Some equipment for graphite research in swimming pool reactors; Quelques dispositifs d'etude du graphite dans les piles piscines

Energy Technology Data Exchange (ETDEWEB)

Seguin, M; Arragon, Ph; Dupont, G; Gentil, J; Tanis, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1964-07-01

The irradiation devices described are used for research concerning reactors of the natural uranium type, moderated by graphite and cooled by carbon dioxide. The devices are generally designed for use in swimming pool reactors. The following points have been particularly studied: - maximum use of the irradiation volume, - use of the simplest technological solutions, - standardization of certain constituent parts. This standardization calls for precision machining and careful assembling; these requirements are also true when a relatively low irradiation temperature is required and the nuclear heating is pronounced. Finally, the design of these devices is suitable for the irradiation of other fissile or non-fissile materials. (authors) [French] Les dispositifs d'irradiation decrits servent aux etudes relatives a la filiere des reacteurs a uranium naturel, moderes au graphite et refroidis par le gaz carbonique. Ils sont generalement concus pour etre utilises dans des piles piscines. L'accent a ete mis sur: - l'utilisation au maximum du volume d'irradiation, - le recours aux solutions technologiques les plus simples, - la standardisation de certaines parties constitutives. Cette standardisation impose un usinage precis et un montage soigne, lesquels sont egalement necessaires lorsqu'on doit obtenir une temperature d'irradiation relativement basse alors que l'echauffement nucleaire est important. Enfin, la conception de ces dispositifs est valable pour irradier d'autres materiaux non fissiles ou fissiles. (auteurs)

A preliminary feasibility study on natural analogue in Korea

Energy Technology Data Exchange (ETDEWEB)

Kim, Chun Soo; Bae, Dae Seok; Kim, Kyung Su; Koh, Yong Kwon; Park, Byung Yun

2000-03-01

Preliminary study on the assessment of natural analogue study in Korea for the deep geological disposal of high-level radioactive waste was carried out. The project on natural analogue study in other countries are introduced. The uranium-bearing deposit in Okcheon belt are summarized, which reported to be uranium-bearing minerals in order to assess to feasibility for natural analogue study in Korea. Among the uranium-bearing deposits, the Deokpyeong area, reported to be the highest reservoir and grade, are selected as the study site, and the elementary investigation, including survey of radioactivity and geochemistry are carried out. According to the investigation of surface environment, the radioactivity and uranium content in the surface water and shallow groundwater does not show any anormal values. However, the radioactivity is expected to be increased in depth and the groundwater reacted with uranium-bearing graphite formation shows high unanium content, indicating the potential possibility for natural analogue study in Korea. In future, if more detail study are performed, the assessment of natural analogue study in Korea are expected.

Study on graphite samples for nuclear usage

International Nuclear Information System (INIS)

Suarez, J.C.M.; Silva Roseira, M. da

1994-01-01

Available as short communication only. The graphite, due to its properties (mechanical strength, thermal conductivity, high-temperature stability, machinability etc.) have many industrial applications, and consequently, an important strategic value. In the nuclear area, it has been used as moderator and reflector of neutrons in the fission process of uranium. The graphite can be produced from many types of carbonaceous materials by a variety of process dominated by the manufactures. This is the reason why there are in the world market a lot of graphite types with different physical and mechanical properties. The present investigation studies some physical characteristics of the graphite samples destined to use in a nuclear reactor. (author). 8 refs, 1 fig, 1 tab

Uranium nitride: a cubic antiferromagnet with anisotropic critical behavior

International Nuclear Information System (INIS)

Buyers, W.J.L.; Holden, T.M.; Svensson, E.C.; Lander, G.H.

1977-11-01

Highly anisotropic critical scattering associated with the transition at T/sub N/ = 49.5 K to the type-I antiferromagnetic structure has been observed in uranium nitride. The transverse susceptibility is found to be unobservably small. The longitudinal susceptibility diverges at T/sub N/ and its anisotropy shows that the spins within the (001) ferromagnetic sheets of the [001] domain are much more highly correlated than they are with the spins lying in adjacent (001) sheets. The correlation range within the sheets is much greater than that expected for a Heisenberg system with the same T/sub N/. The rod-like scattering extended along the spin and domain direction is reminiscent of two-dimensional behavior. The results are inconsistent with a simple localized model and may reflect the itinerant nature of the 5f electrons

Depleted uranium (DU) mobility in the natural environment

International Nuclear Information System (INIS)

Ragnarsdottir, K.V.

2002-01-01

In 1999 the Balkan's conflict lead NATO war planes to leave 10x10 3 kg of depleted uranium (DU) in the environment of Kosovo and neighbouring states (UNEP, 2001). DU behaves in the same manner in the environment as natural uranium and it can be traced with isotopic analysis due to the fact that DU has the isotopic composition of 0.2% 235 U and 99.8% 2 38 U as opposed to natural uranium which has 0.7% 2 35 U and 99.3% 2 38 U. DU is a waste product of the nuclear industry which enrich nuclear fuel by 2 35 U. Large stock piles of DU therefore exist in countries that produce nuclear energy and/or nuclear weapons. The DU is given to the weapons industry for free (or cheap) and has been a popular choice for armour penetrating arsenal due to the high density of uranium (19 g cm -3 ) and therefore its high penetrating power. Indeed the arsenal used in Kosovo consisted of DU penetrators that were shot from A-10 aeroplanes. They weigh roughly 300 g and have the shape of a fat 9 cm long pencil. (author)

A study on the formation of uranium carbide in an induction furnace

International Nuclear Information System (INIS)

Song, In Young; Lee, Yoon Sang; Kim, Eung Soo; Lee, Don Bae; Kim, Chang Kyu

2005-01-01

Uranium is a typical carbide-forming element. Three carbides, UC, U 2 C 3 and UC 2 , are formed in the uranium-carbon system. The most important of these as fuel is uranium monocarbide UC. It is well known that Uranium carbides can be obtained by three basic methods: 1) by reaction of uranium metal with carbon; 2) by reaction of uranium metal powder with gaseous hydrocarbons; 3) by reaction of uranium oxides with carbon. The use of uranium monocarbide, or materials based on it, has great prospects as fuel for nuclear reactors. It is quite possible that uranium dicarbide UC 2 may also acquire great importance as a fuel, particularly in dispersion fuel elements with graphite matrix. In the present study, uranium carbides are obtained by direct reaction of uranium metal with graphite in a high frequency induction furnace

10 CFR 40.66 - Requirements for advance notice of export shipments of natural uranium.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Requirements for advance notice of export shipments of natural uranium. 40.66 Section 40.66 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF SOURCE... natural uranium. (a) Each licensee authorized to export natural uranium, other than in the form of ore or...

Natural uranium

International Nuclear Information System (INIS)

Ammerich, Marc; Frot, Patricia; Gambini, Denis-Jean; Gauron, Christine; Moureaux, Patrick; Herbelet, Gilbert; Lahaye, Thierry; Pihet, Pascal; Rannou, Alain

2014-08-01

This sheet belongs to a collection which relates to the use of radionuclides essentially in unsealed sources. Its goal is to gather on a single document the most relevant information as well as the best prevention practices to be implemented. These sheets are made for the persons in charge of radiation protection: users, radioprotection-skill persons, labor physicians. Each sheet treats of: 1 - the radio-physical and biological properties; 2 - the main uses; 3 - the dosimetric parameters; 4 - the measurement; 5 - the protection means; 6 - the areas delimitation and monitoring; 7 - the personnel classification, training and monitoring; 8 - the effluents and wastes; 9 - the authorization and declaration administrative procedures; 10 - the transport; and 11 - the right conduct to adopt in case of incident or accident. This sheet deals specifically with natural uranium

Criticality safety aspects of K-25 Building uranium deposit removal

International Nuclear Information System (INIS)

Haire, M.J.; Jordan, W.C.; Ingram, J.C. III; Stinnet, E.C. Jr.

1995-01-01

The K-25 Building of the Oak Ridge Gaseous Diffusion Plant (now the K-25 Site) went into operation during World War II as the first large scale production plant to separate 235 U from uranium by the gaseous diffusion process. It operated successfully until 1964, when it was placed in a stand-by mode. The Department of Energy has initiated a decontamination and decommissioning program. The primary objective of the Deposit Removal (DR) Project is to improve the nuclear criticality safety of the K-25 Building by removing enriched uranium deposits from unfavorable-geometry process equipment to below minimum critical mass. The method utilized to accomplish this are detailed in this report

Criticality safety aspects of K-25 Building uranium deposit removal

Energy Technology Data Exchange (ETDEWEB)

Haire, M.J.; Jordan, W.C. [Oak Ridge National Lab., TN (United States); Ingram, J.C. III; Stinnet, E.C. Jr. [Oak Ridge K-25 Site, TN (United States)

1995-12-31

The K-25 Building of the Oak Ridge Gaseous Diffusion Plant (now the K-25 Site) went into operation during World War II as the first large scale production plant to separate {sup 235}U from uranium by the gaseous diffusion process. It operated successfully until 1964, when it was placed in a stand-by mode. The Department of Energy has initiated a decontamination and decommissioning program. The primary objective of the Deposit Removal (DR) Project is to improve the nuclear criticality safety of the K-25 Building by removing enriched uranium deposits from unfavorable-geometry process equipment to below minimum critical mass. The method utilized to accomplish this are detailed in this report.

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

Uranium absorption study pile; Empilement pour le controle de l'absorption de l'uranium

Energy Technology Data Exchange (ETDEWEB)

Raievski, V; Sautiez, B [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1959-07-01

The report describes a pile designed to measure the absorption of fuel slugs. The pile is of graphite and comprises a central section composed of uranium rods in a regular lattice. RaBe sources and BF{sub 3} counters are situated on either side of the center. A given uranium charge is compared with a specimen charge of about 560 kg, and the difference in absorption between the two noted. The sensitivity of the equipment will detect absorption variations of about a few ppm boron (10{sup -6} boron per gr. of uranium) or better. (author) [French] Nous decrivons un dispositif permettant de mesurer l'absorption des elements combustibles d'une pile. Ce dispositif est constitue par un empilement de graphite dont la region centrale est formee par un reseau regulier de barres d'uranium. Des sources de RaBe et des compteurs a BF{sub 3} sont places de part et d'autre de cette region. En comparant un chargement d'uranium a un chargement etalon d'environ 560 kg, on peut determiner la difference d'absorption entre ces deux chargements. La sensibilite permettrait de deceler une variation d'absorption de l'ordre du ppm de bore (10{sup -6} g de bore par gramme d'uranium) et peut-etre mieux. (auteur)

Minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems

Energy Technology Data Exchange (ETDEWEB)

Jordan, W.C.; Turner, J.C.

1992-12-01

A parametric calculational analysis has been performed in order to estimate the minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems. The analysis was performed using a version of the SCALE-4.0 code system and the 27-group ENDF/B-IV cross-section library. Water-moderated uranyl fluoride (UO{sub 2}F{sub 2} and H{sub 2}O) and hydrofluoric-acid-moderated uranium hexaflouride (UF{sub 6} and HF) systems were considered in the analysis over enrichments of 1.4 to 5 wt % {sup 235}U. Estimates of the minimum critical volume, minimum critical mass of uranium, and the minimum mass of moderator required for criticality are presented. There was significant disagreement between the values generated in this study when compared with a similar undocumented study performed in 1983 using ANISN and the Knight-modified Hansen-Roach cross sections. An investigation into the cause of the disagreement was made, and the results are presented.

Minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems

Energy Technology Data Exchange (ETDEWEB)

Jordan, W.C.; Turner, J.C.

1992-12-01

A parametric calculational analysis has been performed in order to estimate the minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems. The analysis was performed using a version of the SCALE-4.0 code system and the 27-group ENDF/B-IV cross-section library. Water-moderated uranyl fluoride (UO[sub 2]F[sub 2] and H[sub 2]O) and hydrofluoric-acid-moderated uranium hexaflouride (UF[sub 6] and HF) systems were considered in the analysis over enrichments of 1.4 to 5 wt % [sup 235]U. Estimates of the minimum critical volume, minimum critical mass of uranium, and the minimum mass of moderator required for criticality are presented. There was significant disagreement between the values generated in this study when compared with a similar undocumented study performed in 1983 using ANISN and the Knight-modified Hansen-Roach cross sections. An investigation into the cause of the disagreement was made, and the results are presented.

Distinction between natural and depleted uranium using instrumental neutron activation analysis

International Nuclear Information System (INIS)

Haddad, Kh.

2008-01-01

A convenient method to discriminate between natural and depleted uranium samples was developed in this work. Traces of natural and depleted uranium were irradiated separately and the ratios of 95 Zr/ 103 Ru, 239 Np/ 95 Zr, 239 Np/ 103 Ru were measured. The results show that these ratios can be used as indicators of the uranium isotopic composition of the sample. These ratios are independent of the secular equilibrium of the 238 U with its daughters in the sample and indicate the isotopic composition for trace amounts. Date and truffle samples has been analysed also using this method. Results show that the uranium content in this product was less than the detection limit.(author)

Research on calculation of mixing fraction for natural uranium equivalent fuel

International Nuclear Information System (INIS)

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

2013-01-01

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

Natural uranium toxicology - evaluation of internal contamination in man

International Nuclear Information System (INIS)

Chalabreysse, J.

1968-01-01

After reminding the physical and chemical properties of natural uranium which might affect its toxicology, a comprehensive investigation upon natural uranium metabolism and toxicity and after applying occupational exposure standards to this particular poison, it has been determined, from accident reports and human experience reported in the related literature, a series of formulae obtained by theoretical mathematical development giving principles for internal contamination monitoring and disclosure by determining uranium in the urine of occupationally exposed individuals. An assay is performed to determine individual internal contamination according to the various contamination cases. The outlined purposes, mainly practical, required some options and extrapolations. The proposed formula allows a preliminary approach and also to determine shortly a contamination extent or to discuss the systematical urinalysis results as compared with individual radio-toxicology monitoring professional standards. (author) [fr

Criticality Benchmark Analysis of Water-Reflected Uranium Oxyfluoride Slabs

International Nuclear Information System (INIS)

Marshall, Margaret A.; Bess, John D.

2009-01-01

A series of twelve experiments were conducted in the mid 1950's at the Oak Ridge National Laboratory Critical Experiments Facility to determine the critical conditions of a semi-infinite water-reflected slab of aqueous uranium oxyfluoride (UO2F2). A different slab thickness was used for each experiment. Results from the twelve experiment recorded in the laboratory notebook were published in Reference 1. Seven of the twelve experiments were determined to be acceptable benchmark experiments for the inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. This evaluation will not only be available to handbook users for the validation of computer codes and integral cross-section data, but also for the reevaluation of experimental data used in the ANSI/ANS-8.1 standard. This evaluation is important as part of the technical basis of the subcritical slab limits in ANSI/ANS-8.1. The original publication of the experimental results was used for the determination of bias and bias uncertainties for subcritical slab limits, as documented by Hugh Clark's paper 'Subcritical Limits for Uranium-235 Systems'.

The Potential for Criticality Following Disposal of Uranium at Low-Level-Waste Facilities. Containerized Disposal

International Nuclear Information System (INIS)

Colten-Bradley, V.A.; Hopper, C.M.; Parks, C.V.; Toran, L.E.

1999-01-01

The purpose of this study was to evaluate whether or not fissile uranium in low-level-waste (LLW) facilities can be concentrated by hydrogeochemical processes to permit nuclear criticality. A team of experts in hydrology, geology, geochemistry, soil chemistry, and criticality safety was formed to develop and test some reasonable scenarios for hydrogeochemical increases in concentration of special nuclear material (SNM) and to use these scenarios to aid in evaluating the potential for nuclear criticality. The team's approach was to perform simultaneous hydrogeochemical and nuclear criticality studies to (1) identify some possible scenarios for uranium migration and concentration increase at LLW disposal facilities, (2) model groundwater transport and subsequent concentration increase via precipitation of uranium, and (3) evaluate the potential for nuclear criticality resulting from potential increase in uranium concentration over disposal limits. The analysis of SNM was restricted to 235 U in the present scope of work. The work documented in this report indicates that the potential for a criticality safety concern to arise in an LLW facility is extremely remote, but not impossible. Theoretically, conditions that lead to a potential criticality safety concern might arise. However, study of the hydrogeochemical mechanisms, the associated time frames, and the factors required for an actual criticality event indicate that proper emplacement of the SNM at the site can eliminate practical concerns relative to the occurrence and possible consequences of a criticality event

Study on removal effect and mechanism of uranium by hydroxyapatite and natural apatite

International Nuclear Information System (INIS)

Zhang Xiaofeng; Chen Diyun; Tu Guoqing; Huang Xiaozhui

2014-01-01

By the static experiments, the effects of reaction time, pH value, initial concentration of uranium, dosage of apatite on adsorption of hydroxyapatite and natural apatite for uranium were studied respectively. The adsorption process was analyzed by thermodynamics and kinetics, and the adsorption mechanism was analyzed by infrared spectroscopy, X-ray diffraction and scanning electron microscope. The results of hydroxyapatite show that the removal capacity of uranium increases with the initial concentration of uranium, and the adsorption rate of hydroxyapatite on UO_2"2"+ reaches 85%, when the pH value is 4 to 5 and dosage of hydroxyapatite is 0.75 g. The results of natural apatite show that the removal capacity of uranium increases with the initial concentration of uranium, and the adsorption rate of natural apatite on UO_2"2"+ is up to 80%, when the pH value is 3 and dosage of hydroxyapatite is l.0 g. Similarly, at 120 minutes both of the removal reactions by hydroxyapatite and natural apatite substantially reach equilibrium. Moreover, both of the reactions by hydroxyapatite and natural apatite are in line with quasi secondary dynamics equation, and follow the Langmuir adsorption isotherm. Infrared spectra indicate that the removal of hydroxyapatite for uranium depends on the complexation of phosphate, which is almost the same as that of natural apatite. X-ray diffraction analysis shows that hydroxyapatite has the composition and structure of pure material, whereas the natural apatite is mainly composed of Ca_5H_2(PO_4)_3F and Ca_8H_2(PO_4)_6H_2O. In addition, scanning electron microscope demonstrates that hydroxyapatite has the appearance of spherical with a hole and the hole has a cavity containing a large amount of floc, while the surface becomes smooth and pores are closed after removal of uranium, which is due to the adsorption of UO_2"2"+ leading a link between molecules on hydroxyapatite surface. But for natural apatite, it depicts the angular mineral shape

CRITICALITY CONTROL DURING THE DISMANTLING OF A URANIUM CONVERSION PLANT

International Nuclear Information System (INIS)

LADURELLE, Laurent; LISBONNE, Pierre

2003-01-01

Within the Commissariat a l'Energie Atomique, in the Cadarache Research Center in southern France, the production at the Enriched Uranium Treatment Workshops started in 1965 and ended in 1995. The dismantling is in progress and will last until 2006. The decommissioning is planned in 2007. Since the authorized enrichment in 235U was 10% in some parts of the plant, and unlimited in others, the equipment and procedures were designed for criticality control during the operating period. Despite the best previous removing of the uranium in the inner parts of the equipment, evaluation of the mass of remaining fissile material by in site gamma spectrometry measurement shows that the safety of the ''clean up'' operations requires specific criticality control procedures, this mass being higher than the safe mass. The chosen method is therefore based on the mapping of fissile material in the contaminated parts of the equipment and on the respect of particular rules set for meeting the criticality control standards through mass control. The process equipment is partitioned in separated campaign, and for each campaign the equipment dismantling is conducted with a precise traceability of the pieces, from the equipment to the drum of waste, and the best final evaluation of the mass of fissile material in the drum. The first results show that the mass of uranium found in the dismantled equipment is less than the previous evaluation, and they enable us to confirm that the criticality was safely controlled during the operations. The mass of fissile material remaining in the equipment can be then carefully calculated, when it is lower than the minimal critical mass, and on the basis of a safety analysis, we will be free of any constraints regarding criticality control, this allowing to make procedures easier, and to speed up the operations

The measurement of natural uranium in urine by fluorometry

International Nuclear Information System (INIS)

Kramer, G.H.; Johnson, J.R.; Green, W.

1984-02-01

The fluorometric method of measuring natural uranium in urine that is currently used by the Bioassay Laboratory at Chalk River Nuclear Laboratories has been tested, optimized and documented. The method, which measures the fluorescence of uranium in a fused sodium fluoride pellet, has been shown to be quench independent and is routinely used to measure uranium concentrations in the range of 1 μg/L to 90 μg/L. The fluorimeter has a dynamic range of 0.2 μg/L to 200 μg/L

Studies on the development of special graphite for use in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Bhatia, G.; Aggarwal, R.K.; Saha, M.; Sengupta, P.R.; Mishra, A. [National Physical Lab., New Delhi (India). Carbon Technology Unit

2002-07-01

Special graphite is considered as a critical component of the present-day tokamaks wherein it acts as the armour material for plasma-facing components. This graphite is required to possess, besides other characteristics, high values of bulk density, bending strength and electrical and thermal conductivities and a low value of ash content. Since such graphite was not commercially available in the country, efforts to develop it were initiated at the National Physical Laboratory, New Delhi. The basic approach to develop this graphite was based on green coke method of making the high density graphite, wherein the green coke was modified by incorporating in it small amounts of conducting carbon materials, i.e. needle coke, synthetic graphite and natural graphite. The resulting graphites were characterized with respect to various physical characteristics, namely, green density, weight loss, volume shrinkage, linear shrinkage, bulk density, bending strength, Young's modulus and electrical resistivity, etc. The results are described and discussed in the present paper. 6 refs., 2 tabs.

Chapter 4: Exponential experiments on natural uranium graphite moderated systems. I: Performance and analysis

International Nuclear Information System (INIS)

Brown, G.; Moore, P.G.F.; Richmond, R.

1963-01-01

A description is given of the methods used on the first BICEP stack for the performance and analysis of graphite exponential experiments. These differ in many respects from the methods formerly employed at A.E.R.E. The accuracy of the measurements has been increased, and the time taken to carry out and analyse an experiment has been reduced by approximately a factor of four. The following have contributed to the experimental work reported in this paper: J. R. Dyson, W. M. Holderay, R. M. Turner, S. D. R. Summers. (author)

The challenge of uranium exploration

International Nuclear Information System (INIS)

Fountain, D.K.

1982-06-01

The first uranium discoveries at Beaverlodge were made using simple radiometric methods: hand-held geiger counters. Since then techniques of uranium exploration have evolved through airborne radiometric surveys, tracking glacial boulder trains to their origins, and electromagnetic surveys to detect graphite associated with buried uranium deposits. Simple radiometric surveys can cost around $1 000. per day, while testing for deposits at depths of over 400 meters will cost more than $60 000. per drill hole

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Experimental performance and results of the critical pebble bed facility KAHTER

Energy Technology Data Exchange (ETDEWEB)

Krings, F. J.; Drueke, V.; Kirch, N.; Neef, R. D.

1974-10-15

The paper provides a description and results of critical experiments performed in KAHTER fueled with pebbles containing coated particles of HEU/Th oxide with a ratio of uranium-to-thorium of 1.1:5. Core loadings with varying amounts of fuel and solid graphite pebbles were tested with fuel-to-graphite pebble ratios of 3:1, 1:1, and 1:3. Tests included criticality for various fuel loadings with all control rods removed, control rod worths for reflector-mounted control as single rods and in a bank and control worths for a central control rod, reaction rates by flux wire activations (Dy, Mn, In, Au, and U-235) and detector measurements (BF3 and fission chamber), simulated xenon stability testing using the motions of a Cf-252 source and Cd-absorber observed by an externally-located BF3 detector, and the reactivity worth of a Hf burnable absorber. For calculations of the room-temperature zero-power critical experiment, the values for nitrogen and hydrogen contents of the graphite were taken from previous experiments in CESAR.

Use of isotopic signature of radionuclides released from uranium mines and mills to discriminate low levels of environmental impact against natural background levels

International Nuclear Information System (INIS)

Zettwoog, P.; Lemaitre, N.; Bernhard, S.; Vauzelle, Y.

1997-01-01

In France, uranium ores have been exploited in rural areas with a low population density. The critical population group which is identified for radiological impact studies lives close to the uranium facilities, at distances from a few hundred metres to 1 kilometer. Within this range, the radioactivity of the environment is still detectable amongst the natural background. Mining companies manage surveillance networks according to strict specifications laid down by the government authorities. For active mining operations it is the exposure to daughter products of radon 222 that forms the bulk of the total effective dose. After closure of a mine and rehabilitation of the site, products such as uranium 238 and radium 226 in the water can become the major components of the total effective dose. Surveillance networks are built to allow direct measurement of the radon daughter products critical to the alpha energy and measurements of the water activities for uranium 238 and radium 226. Annual limits for the effective individual doses are given and determined by the authorities. The industry must manage effective annual doses of the order of 1 mSv. The natural exposure which is not part of the regulation is of the same order (1 mSv) as the exposure created by the industry. The results given are therefore lacking in clarity for the public. The Office de Protection contre les Rayonnements Ionisants (OPRI) and ALGADE are developing methods which allow differentiation between natural phenomena and man-made phenomena. It has been recognised that for a region where mining activities have taken place, the isotopic signature of uranium, radium and radon can clearly be recognised from the same product of natural origin. In the case of radon, for example, the industry produces only radon 222 while natural emanations are composed of radon 220 and radon 222. (author)

Influence of Uranium on Bacterial Communities: A Comparison of Natural Uranium-Rich Soils with Controls

Science.gov (United States)

Mondani, Laure; Benzerara, Karim; Carrière, Marie; Christen, Richard; Mamindy-Pajany, Yannick; Février, Laureline; Marmier, Nicolas; Achouak, Wafa; Nardoux, Pascal; Berthomieu, Catherine; Chapon, Virginie

2011-01-01

This study investigated the influence of uranium on the indigenous bacterial community structure in natural soils with high uranium content. Radioactive soil samples exhibiting 0.26% - 25.5% U in mass were analyzed and compared with nearby control soils containing trace uranium. EXAFS and XRD analyses of soils revealed the presence of U(VI) and uranium-phosphate mineral phases, identified as sabugalite and meta-autunite. A comparative analysis of bacterial community fingerprints using denaturing gradient gel electrophoresis (DGGE) revealed the presence of a complex population in both control and uranium-rich samples. However, bacterial communities inhabiting uraniferous soils exhibited specific fingerprints that were remarkably stable over time, in contrast to populations from nearby control samples. Representatives of Acidobacteria, Proteobacteria, and seven others phyla were detected in DGGE bands specific to uraniferous samples. In particular, sequences related to iron-reducing bacteria such as Geobacter and Geothrix were identified concomitantly with iron-oxidizing species such as Gallionella and Sideroxydans. All together, our results demonstrate that uranium exerts a permanent high pressure on soil bacterial communities and suggest the existence of a uranium redox cycle mediated by bacteria in the soil. PMID:21998695

Natural uranium and 226Ra in bottled potable waters of Argentina

International Nuclear Information System (INIS)

Bomben, Ana M.; Palacios, Miguel A.

2001-01-01

This paper presents the results obtained of the measurement of the natural uranium and 226 Ra concentrations carried out on 345 drinking water samples coming from different provinces of Argentina. The samples were collected from tap water systems and private wells. Six bottled mineral waters samples, selected from those most extensively consumed, were also analyzed. The natural uranium concentration was determined by a fluorimetric procedure and 226 Ra by the 222 Rn emanation technique and liquid scintillation counting. Values ranging from 0,03 to 50 μg L -1 of natural uranium and concentrations up to 22 mBq L -1 were found in the drinking water samples analyzed. Natural uranium concentrations from 0,04 to 3,8 μg L -1 and 226 Ra concentrations up to 2,4 mBq L -1 were measured in the bottled mineral waters samples. Based on the water intake rate and the measured concentrations of both radionuclides analyzed, an annual collective effective dose of 1,5 man Sv and an average committed effective dose of 0,5 μSv a -1 , were calculated for the City of Buenos Aires inhabitants. (author)

About the elaboration of pure uranium dicarbide

International Nuclear Information System (INIS)

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

1963-01-01

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

Control of criticality; Kawalan kegentingan

Energy Technology Data Exchange (ETDEWEB)

NONE

1988-12-31

The chapter briefly discussed the following subjects: basic and the principle of criticality, natural uranium, neutron utilization, criticality data for systems, criticality accidents, criticality control i.e. mass, volume and geometry control .

Uranium absorption study pile

International Nuclear Information System (INIS)

Raievski, V.; Sautiez, B.

1959-01-01

The report describes a pile designed to measure the absorption of fuel slugs. The pile is of graphite and comprises a central section composed of uranium rods in a regular lattice. RaBe sources and BF 3 counters are situated on either side of the center. A given uranium charge is compared with a specimen charge of about 560 kg, and the difference in absorption between the two noted. The sensitivity of the equipment will detect absorption variations of about a few ppm boron (10 -6 boron per gr. of uranium) or better. (author) [fr

The relationship of JNC and JCO in the uranium processing plant criticality accident

International Nuclear Information System (INIS)

Kanamori, Masashi; Yanagibashi, Katsumi; Okamoto, Naritoshi

2002-12-01

On September 30th 1999, the criticality accident occurred at JCO's uranium conversion building in Tokai. The accident occurred during reconversion from U 3 O 8 to uranium nitrate solution (UNH) with uranium enriched 18.8% and about 60 kgU. JCO contacted with JNC to supply UNH that is fuel material for the experimental fast breeder reactor 'JOYO'. JNC has contracted with JCO that had started nuclear fuel material processing business following a definite policy of Japanese government and developed SUMITOMO ADU PROCESS'. JNC made the first contract with JCO in 1985 and has made a contact every year. There had never been a problem in their products. JNC inspected products based on contract. JNC discharge our duty as customer inspecting products based on contract. As for safety control, JCO had taken licensing safety review and had been permitted to be 'a processing facility'. Therefore JNC understood that JCO produced following this license. 'The Uranium Processing Plant Criticality Accident Investigation' showed that JCO had been taking a different method from the permit and violating the license. However JNC had never been explained about that and JCO's operation procedures had never described about that. Therefore the Criticality Accident couldn't be avoided. This report describes the relationship of JNC and JCO in the uranium reconversion contract for JOYO, atomic development policy of Japanese government, process to the order and the contents of contract. (author)

Immobilization of uranium in contaminated soil by natural apatite addition

International Nuclear Information System (INIS)

Mrdakovic Popic, Jelena; Stojanovic, Mirjana; Milosevic, Sinisa; Iles, Deana; Zildzovic, Snezana

2007-01-01

Available in abstract form only. Full text of publication follows: The goal of this study was to evaluate the effectiveness of Serbian natural mineral apatite as soil additive for reducing the migration of uranium from contaminated sediments. In laboratory study we investigated the sorption properties of domestic apatite upon different experimental conditions, such as pH, adsorbent mass, reaction period, concentration of P 2 O 5 in apatite, solid/liquid ratio. In second part of study, we did the quantification of uranium in soil samples, taken from uranium mine site 'Kalna', by sequential extraction method. The same procedure was, also, used for uranium determination in contaminated soil samples after apatite addition, in order to determine the changes in U distribution in soil fraction. The obtained results showed the significant level of immobilization (96.7%) upon certain conditions. Increase of %P 2 O 5 in apatite and process of mechano-chemical activation led to increase of immobilization capacity from 17.50% till 91.64%. The best results for uranium binding were obtained at pH 5.5 and reaction period 60 days (98.04%) The sequential extraction showed the presence of uranium (48.2%) in potentially available soil fractions, but with the apatite addition uranium content in these fractions decreased (30.64%), what is considering environmental aspect significant fact. In situ immobilization of radionuclide using inexpensive sequestering agents, such as apatite, is very adequate for big contaminated areas of soil with low level of contamination. This investigation study on natural apatite from deposit 'Lisina' Serbia was the first one of this type in our country. Key words: apatite, uranium, immobilization, soil, contamination. (authors)

Characteristics of the natural uranium ingots developed in IPEN - CNEN/SP

International Nuclear Information System (INIS)

Soares, M.C.B.; Koshimizu, S.

1990-01-01

The natural uranium consists of two primary isotopes, the U sup(235) (0,7%) and the U sup(238) (99,3%). The isotopic separation carried out in order to obtain enriched uranium, generates a by-product called depleted uranium, which can be applied for industrial uses. The most singular property, from engineering standpoint, is its high density. When the density is the only important factor, the uranium has great advantage over other heavy metals related to economic and technical considerations. Among some applications of uranium are aircraft and missile counterweights, kinetics energy penetrators, radiation shielding, gyro rotors and oil-well sinker bars. The uranium ingot fabrication is done by direct reduction of UF, with magnesium, without remelting. The microstructure of as-cast uranium is, as in the other as-cast, formed by coarse and. (author)

Study on thermal neutron spectra in reactor moderators by time-of-flight method

International Nuclear Information System (INIS)

Akino, Fujiyoshi

1982-12-01

Prediction of thermal neutron spectra in a reactor core plays very important role in the neutronic design of the reactor for obtaining the accurate thermal group constants. It is well known that the neutron scattering properties of the moderator materials markedly influence the thermal neutron spectra. Therefore, 0 0 angular dependent thermal neutron spectra were measured by the time-of-flight method in the following moderator bulks 1) Graphite bulk poisoned with boron at the temperatures from 20 to 800 0 C, 2) Light water bulk poisoned with Cadmium and/or Indium, 3) Light water-natural uranium heterogeneous bulk. The measured results were compared with calculation utilizing Young-Koppel and Haywood scattering model for graphite and light water respectively. On the other hand, a variety of 20% enriched uranium loaded and graphite moderated cores consisting of the different lattice cell in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments related to Very High Temperature Reactor (VHTR). The experimental data were for the critical masses in 235 U, reactivity worths of experimental burnable poison rods, thorium rods, natural-uranium rods and experimental control rods and kinetic parameters. It is made clear from comparison between measurement and calculation that the accurate thermal group constants can be obtained by use of the Young-Koppel and Haywood neutron scattering models if heterogeneity of reactor core lattices is taken into account precisely. (author)

226Ra and natural uranium in egyptian bottled mineral waters

International Nuclear Information System (INIS)

Higgy, R.H.

2000-01-01

Concentration levels of 226 Ra and natural uranium have been analysed bottled mineral water commercially available in egypt. 226 Ra was determined by applying a chemical procedure in which Ra was coprecipitated with Ba as sulphate. The precipitate was then dissolved with EDTA and then measured by liquid scintillation system, after mixing with a scintillation cocktail. Natural uranium was determined by applying a chemical procedure for uranium extraction using MIBK and then measured using laser fluorimeter system. The concentration values obtained were compared with concentrations reported by other countries and with reference values accepted for drinking water. Based on the consumption rate and the measured concentrations, the collective committed effective doses were calculated. In addition, Ca, Mg and Na were measured using Icp system and compared with some worldwide values

The chemical industry of uranium in France; L'industrie chimique de l'uranium en France

Energy Technology Data Exchange (ETDEWEB)

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

1955-07-01

The actual CEA program is concerned with the construction of two large graphite reactors, each of those containing at least one hundred tons of uranium metal with nuclear purity. The uranium for these two reactors will be regularly supplied by new resources discovered in France and Madagascar in the last five years. The working and treatment of such ore have led to the creation of an important french industry of which the general outline and principle are described. The operated ores have got different natures and concentration, individual characteristics are described for the main ores.The most high-grade ore are transported to a central plant in Bouchet near Paris; the low-grade ore are concentrated by physical methods or chemical processes of which principles and economy are studied with constancy. The acid processes are the only used until now, although the carbonated alkaline processes has been studied in France. The next following steps after the acid process until the obtention of uranium rich concentrate are described. The purification steps of uranium compounds to nuclear purity material are described as well as the steps to elaborate metal of which the purity grade will be specify. Finally, the economic aspects of uranium production difficulty will be considered in relation with technical progresses which we can expect to achieve in the future. (M.P.)

Accumulation of uranium by filamentous green algae under natural environmental conditions

International Nuclear Information System (INIS)

Aleissa, K.A.; Shabana, El-Said K.; Al-Masoud, F.L.S.

2004-01-01

The capacity of algae to concentrate uranium under natural environmental conditions is measured by a-spectrometry. Spirogyra, a filamentous green fresh-water alga, has concentrated uranium from a surface concrete ponds with elevated uranium levels (140-1140 ppb). The concentration factors (CFs) ranged from 8.9-67 with an average value of 22. Cladophora spp, a filamentous green marine alga has concentrated uranium from the marine water with a concentration factor ranged from 220-280. The average concentration factor was 250. The factors affecting the sorption process are discussed in detail. (author)

Some equipment for graphite research in swimming pool reactors

International Nuclear Information System (INIS)

Seguin, M.; Arragon, Ph.; Dupont, G.; Gentil, J.; Tanis, G.

1964-01-01

The irradiation devices described are used for research concerning reactors of the natural uranium type, moderated by graphite and cooled by carbon dioxide. The devices are generally designed for use in swimming pool reactors. The following points have been particularly studied: - maximum use of the irradiation volume, - use of the simplest technological solutions, - standardization of certain constituent parts. This standardization calls for precision machining and careful assembling; these requirements are also true when a relatively low irradiation temperature is required and the nuclear heating is pronounced. Finally, the design of these devices is suitable for the irradiation of other fissile or non-fissile materials. (authors) [fr

Heat Transfer During Evaporation of Cesium From Graphite Surface in an Argon Environment

Directory of Open Access Journals (Sweden)

Bespala Evgeny

2016-01-01

Full Text Available The article focuses on discussion of problem of graphite radioactive waste formation and accumulation. It is shown that irradiated nuclear graphite being inalienable part of uranium-graphite reactor may contain fission and activation products. Much attention is given to the process of formation of radioactive cesium on the graphite element surface. It is described a process of plasma decontamination of irradiated graphite in inert argon atmosphere. Quasi-one mathematical model is offered, it describes heat transfer process in graphite-cesium-argon system. Article shows results of calculation of temperature field inside the unit cell. Authors determined the factors which influence on temperature change.

Paleo-channel deposition of natural uranium at a US Air Force landfill

International Nuclear Information System (INIS)

Young, Carl; Weismann, Joseph; Caputo, Daniel

2007-01-01

Available in abstract form only. Full text of publication follows: The US Air Force sought to identify the source of radionuclides that were detected in groundwater surrounding a closed solid waste landfill at the former Lowry Air Force Base in Denver, Colorado, USA. Gross alpha, gross beta, and uranium levels in groundwater were thought to exceed US drinking water standards and down-gradient concentrations exceeded up-gradient concentrations. Our study has concluded that the elevated radionuclide concentrations are due to naturally-occurring uranium in the regional watershed and that the uranium is being released from paleo-channel sediments beneath the site. Groundwater samples were collected from monitor wells, surface water and sediments over four consecutive quarters. A list of 23 radionuclides was developed for analysis based on historical landfill records. Concentrations of major ions and metals and standard geochemical parameters were analyzed. The only radionuclide found to be above regulatory standards was uranium. A search of regional records shows that uranium is abundant in the upstream drainage basin. Analysis of uranium isotopic ratios shows that the uranium has not been processed for enrichment nor is it depleted uranium. There is however slight enrichment in the U-234:U- 238 activity ratio, which is consistent with uranium that has undergone aqueous transport. Comparison of up-gradient versus down-gradient uranium concentrations in groundwater confirms that higher uranium concentrations are found in the down-gradient wells. The US drinking water standard of 30 μg/L for uranium was exceeded in some of the up-gradient wells and in most of the down-gradient wells. Several lines of evidence indicate that natural uranium occurring in streams has been preferentially deposited in paleo-channel sediments beneath the site, and that the paleo-channel deposits are causing the increased uranium concentrations in down-gradient groundwater compared to up

Electrochemically induced chemical sensor properties in graphite screen-printed electrodes: The case of a chemical sensor for uranium

International Nuclear Information System (INIS)

Kostaki, Vasiliki T.; Florou, Ageliki B.; Prodromidis, Mamas I.

2011-01-01

Highlights: → Electrochemical treatment endows analytical characteristics to SPEs. → A sensitive chemical sensor for uranium is described. → Performance is due to a synergy between electrochemical treatment and ink's solvents. → The amount of the solvent controls the achievable sensitivity. - Abstract: We report for the first time on the possibility to develop chemical sensors based on electrochemically treated, non-modified, graphite screen-printed electrodes (SPEs). The applied galvanostatic treatment (5 μA for 6 min in 0.1 M H 2 SO 4 ) is demonstrated to be effective for the development of chemical sensors for the determination of uranium in aqueous solutions. A detailed study of the effect of various parameters related to the fabrication of SPEs on the performance of the resulting sensors along with some diagnostic experiments on conventional graphite electrodes showed that the inducible analytical characteristics are due to a synergy between electrochemical treatment and ink's solvents. Indeed, the amount of the latter onto the printed working layer controls the achievable sensitivity. The preconcentration of the analyte was performed in an electroless mode in an aqueous solutions of U(VI), pH 4.6, and then, the accumulated species was reduced by means of a differential pulse voltammetry scan in 0.1 M H 3 BO 3 , pH 3. Under selected experimental conditions, a linear calibration curve over the range 5 x 10 -9 to 10 -7 M U(VI) was constructed. The 3σ limit of detection at a preconcentration time of 30 min, and the relative standard deviation of the method were 4.5 x 10 -9 M U(VI) and >12% (n = 5, 5 x 10 -8 M U(VI)), respectively. The effect of potential interferences was also examined.

Combination Carbon Nanotubes with Graphene Modified Natural Graphite and Its Electrochemical Performance

Directory of Open Access Journals (Sweden)

DENG Ling-feng

2017-04-01

Full Text Available The CNTs/rGO/NG composite lithiumion battery anode material was synthesized by thermal reducing, using graphene oxide (GO and carbon nanotubes (CNTs as precursors for a 5 ∶ 3 proportion. The morphology, structure, and electrochemical performance of the composite were characterized by scanning electron microscopy(SEM, X-ray diffractometry(XRD, Fourier transform infrared spectra (FTIR and electrochemical measurements. The results show that reduced graphene oxide and carbon nanotubes form a perfect three-dimensional network structure on the surface of natural graphite. CNTs/rGO/NG composite has good rate performance and cycle life,compared with pure natural graphite.The initial discharge capacity of designed anode is 479mAh/g at 0.1C, the reversible capacity up to 473mAh/g after 100 cycles,the capacity is still 439.5mAh/g, the capacity retention rate is 92%,and the capacity is 457, 433, 394mAh/g at 0.5, 1, 5C, respectively.

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

International Nuclear Information System (INIS)

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

1998-01-01

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

Experimental critical parameters of enriched uranium solution in annular tank geometries

International Nuclear Information System (INIS)

Rothe, R.E.

1996-04-01

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

Experimental critical parameters of enriched uranium solution in annular tank geometries

Energy Technology Data Exchange (ETDEWEB)

Rothe, R.E.

1996-04-01

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

Analytical strategies for uranium determination in natural water and industrial effluents samples

International Nuclear Information System (INIS)

Santos, Juracir Silva

2011-01-01

The work was developed under the project 993/2007 - 'Development of analytical strategies for uranium determination in environmental and industrial samples - Environmental monitoring in the Caetite city, Bahia, Brazil' and made possible through a partnership established between Universidade Federal da Bahia and the Comissao Nacional de Energia Nuclear. Strategies were developed to uranium determination in natural water and effluents of uranium mine. The first one was a critical evaluation of the determination of uranium by inductively coupled plasma optical emission spectrometry (ICP OES) performed using factorial and Doehlert designs involving the factors: acid concentration, radio frequency power and nebuliser gas flow rate. Five emission lines were simultaneously studied (namely: 367.007, 385.464, 385.957, 386.592 and 409.013 nm), in the presence of HN0 3 , H 3 C 2 00H or HCI. The determinations in HN0 3 medium were the most sensitive. Among the factors studied, the gas flow rate was the most significant for the five emission lines. Calcium caused interference in the emission intensity for some lines and iron did not interfere (at least up to 10 mg L -1 ) in the five lines studied. The presence of 13 other elements did not affect the emission intensity of uranium for the lines chosen. The optimized method, using the line at 385.957 nm, allows the determination of uranium with limit of quantification of 30 μg L -1 and precision expressed as RSD lower than 2.2% for uranium concentrations of either 500 and 1000 μg L -1 . In second one, a highly sensitive flow-based procedure for uranium determination in natural waters is described. A 100-cm optical path flow cell based on a liquid-core waveguide (LCW) was exploited to increase sensitivity of the arsenazo 111 method, aiming to achieve the limits established by environmental regulations. The flow system was designed with solenoid micro-pumps in order to improve mixing and minimize reagent consumption, as well as

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...

The Fracture Toughness of Nuclear Graphites Grades

Energy Technology Data Exchange (ETDEWEB)

Burchell, Timothy D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Erdman, III, Donald L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lowden, Rick R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunter, James A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hannel, Cara C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-04-01

New measurements of graphite mode I critical stress intensity factor, KIc (commonly referred to as the fracture toughness) and the mode II critical shear stress intensity, KIIc, are reported and compared with prior data for KIc and KIIc. The new data are for graphite grades PCEA, IG-110 and 2114. Variations of KIc and acoustic emission (AE) data with graphite texture are reported and discussed. The Codes and Standards applications of fracture toughness, KIc, data are also discussed. A specified minimum value for nuclear graphite KIc is recommended.

A comparative study of electrochemical performance of graphene sheets, expanded graphite and natural graphite as anode materials for lithium-ion batteries

International Nuclear Information System (INIS)

Bai, Li-Zhong; Zhao, Dong-Lin; Zhang, Tai-Ming; Xie, Wei-Gang; Zhang, Ji-Ming; Shen, Zeng-Min

2013-01-01

Highlights: • Graphene sheets (GSs), expanded graphite (EG) and natural graphite (NG) were comparatively investigated as anode materials for lithium-ion batteries. • The reversible capacity of GS electrode was almost twice that of EG electrode and three times that of NG electrode. • The first-cycle coulombic efficiency and capacity retention of NG were much bigger than those of GSs and EG. • GS and EG electrodes exhibited higher electrochemical activity and more favorable kinetic properties. -- Abstract: Three kinds of carbon materials, i.e., graphene sheets (GSs), expanded graphite (EG) and natural graphite (NG) were comparatively investigated as anode materials for lithium-ion batteries via scanning electron microscope, high-resolution transmission electron microscopy, X-ray diffraction, Fourier transform infrared spectroscopy, Raman spectroscopy and a variety of electrochemical testing techniques. The test results showed that the reversible capacities of GS electrode were 1130 and 636 mA h g −1 at the current densities of 0.2 and 1 mA cm −2 , respectively, which were almost twice those of EG electrode and three times those of NG electrode. The first-cycle coulombic efficiency and capacity retention of NG were much bigger than those of GSs and EG. The notable capacity fading observed in GSs and EG may be ascribed to the disorder-induced structure instability. The larger voltage hysteresis in GS and EG electrodes was not only related to the surface functional groups, but also to the active defects in GSs and EG, which results in greater hindrance and higher overvoltage during lithium extraction from electrode. The kinetics properties of GSs, EG and NG electrodes were compared by AC impedance measurements. GS and EG electrodes exhibited higher electrochemical activity and more favorable kinetic properties during charge and discharge process

An evaluation of the dissolution process of natural uranium ore as an analogue of nuclear fuel

International Nuclear Information System (INIS)

Stern, V.H.

1991-08-01

The assumption of congruent dissolution of uraninite as a mechanism for the dissolution behaviour of spent fuel was critically examined with regard to the fate of toxic radionuclides. The fission and daughter products of uranium are typically present in spent unreprocessed fuel rods in trace abundances. The principles of trace element geochemistry were applied in assessing the behaviour of these radionuclides during fluid/solid interactions. It is shown that the behaviour of radionuclides in trace abundances that reside in the crystal structure can be better predicted from the ionic properties of these nuclides rather than from assuming that they are controlled by the dissolution of uraninite. Geochemical evidence from natural uranium ore deposits (Athabasca Basin, Northern Territories of Australia, Oklo) suggests that in most cases the toxic radionuclides are released from uraninite in amounts that are independent of the solution behaviour of uranium oxide. Only those elements that have ionic and thus chemical properties similar to U 4+ , such as plutonium, americium, cadmium, neptunium and thorium can be satisfactorily modelled by the solution properties of uranium dioxide and then only if the environment is reducing. (84 refs., 7 tabs.)

Oxidation of naturally reduced uranium in aquifer sediments by dissolved oxygen and its potential significance to uranium plume persistence

Science.gov (United States)

Davis, J. A.; Smith, R. L.; Bohlke, J. K.; Jemison, N.; Xiang, H.; Repert, D. A.; Yuan, X.; Williams, K. H.

2015-12-01

The occurrence of naturally reduced zones is common in alluvial aquifers in the western U.S.A. due to the burial of woody debris in flood plains. Such reduced zones are usually heterogeneously dispersed in these aquifers and characterized by high concentrations of organic carbon, reduced mineral phases, and reduced forms of metals, including uranium(IV). The persistence of high concentrations of dissolved uranium(VI) at uranium-contaminated aquifers on the Colorado Plateau has been attributed to slow oxidation of insoluble uranium(IV) mineral phases found in association with these reducing zones, although there is little understanding of the relative importance of various potential oxidants. Four field experiments were conducted within an alluvial aquifer adjacent to the Colorado River near Rifle, CO, wherein groundwater associated with the naturally reduced zones was pumped into a gas-impermeable tank, mixed with a conservative tracer (Br-), bubbled with a gas phase composed of 97% O2 and 3% CO2, and then returned to the subsurface in the same well from which it was withdrawn. Within minutes of re-injection of the oxygenated groundwater, dissolved uranium(VI) concentrations increased from less than 1 μM to greater than 2.5 μM, demonstrating that oxygen can be an important oxidant for uranium in such field systems if supplied to the naturally reduced zones. Dissolved Fe(II) concentrations decreased to the detection limit, but increases in sulfate could not be detected due to high background concentrations. Changes in nitrogen species concentrations were variable. The results contrast with other laboratory and field results in which oxygen was introduced to systems containing high concentrations of mackinawite (FeS), rather than the more crystalline iron sulfides found in aged, naturally reduced zones. The flux of oxygen to the naturally reduced zones in the alluvial aquifers occurs mainly through interactions between groundwater and gas phases at the water table

Examination of uranium recovery technique from sea water using natural components for adsorbent

International Nuclear Information System (INIS)

Tanaka, Nobuyuki; Masaki, Hiroyuki; Shimizu, Takao; Tokiwai, Moriyasu

2010-01-01

In this study, we investigated the potency of natural components as adsorbent for uranium recovery from seawater. In addition, cost evaluation of uranium recovery from seawater using natural components for adsorbents was performed. Furthermore, new ideas on reservation system of adsorbents at sea area were proposed. Several poly-phenols were selected as adsorbent reagents, then they were adsorbed on the support such as cotton fiber by several methods as the followings; chemical syntheses, electrical beam irradiation, and traditional dyeing. As a result, the adsorbent made by traditional dyeing method using gallnut tannin as natural component, was showed high performance for uranium recovery from seawater on only the first. It was evaluated that traditional dyeing method had also advantage in the manufacturing cost, comparing with earlier method. Additionally, it was considered that reservation system of adsorbent at sea was able to be simplified compared with earlier system. Consequently, uranium recovery from sea water using natural components as adsorbent proposed in this study had a potency of practical use. (author)

Preconcentration of trace elements from high-purity thorium and uranium on Chelex-100 and determination by graphite furnace atomic absorption spectrometry with Zeeman-effect background correction

International Nuclear Information System (INIS)

Raje, Naina; Kayasth, Satish; Asari, T.P.S.; Gangadharan, S.

1994-01-01

Preconcentration of trace impurities from large-sized samples of uranium metal and thorium oxide using a small column of Chelex-100 followed by their determination using graphite furnace atomic absorption spectrometry (GFAAS) is reported. A 0.5-10-g amount of the sample (uranium metal or thorium oxide) was dissolved, complexed with ammonium carbonate and subjected to the ion-exchange procedure. The retained analytes were eluted with 2-4 M nitric acid and brought to a small volume for a final dilution to 10-25 ml for their determination using GFAAS. The validity of the separation procedure and recoveries at μg kg -1 levels was checked by standard addition; the recoveries were >95%

Migration of uranium daughter radionuclides in natural sediments

International Nuclear Information System (INIS)

Colley, S.; Thomson, J.

1991-01-01

An irregular concentration/depth profile of uranium in deep-sea turbidities, previously elucidated, has been exploited to obtain in-situ effective diffusion coefficients for the long-lived members of the 238 U natural series. The findings are relevant to the assessment of deep-sea sediments as potential repositories for high-level radioactive waste, because waste actinides decay through the same chains of daughter radionuclides as natural actinides. This work was part of the CEC Mirage project-Second phase, Natural analogues research area

Alecto - results obtained with homogeneous critical experiments on plutonium 239, uranium 235 and uranium 233; Alecto - resultats des experiences critiques homogenes realisees sur le plutonium 239, l'uranium 235 et l'uranium 233

Energy Technology Data Exchange (ETDEWEB)

Bruna, J G; Brunet, J P; Caizegues, R; Clouet d' Orval, Ch; Kremser, J; Tellier, H; Verriere, Ph [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-07-01

In this report are given the results of the homogeneous critical experiments ALECTO, made on plutonium 239, uranium 235 and uranium 233. After a brief description of the equipment, the critical masses for cylinders of diameters varying from 25 to 42 cm, are given and compared with other values (foreign results, criticality guide). With respect to the specific conditions of neutron reflection in the ALECTO experiments the minimal values of critical masses are: Pu239 M{sub c} = 910 {+-} 10 g, U235 M{sub c} = 1180 {+-} 12 g and U233 M{sub c} = 960 {+-} 10 g. Experiments relating to cross sections and constants to be used on these materials are presented. Lastly, kinetic experiments allow to compare pulsed neutron methods to fluctuation methods. [French] On presente dans ce rapport les resultats des experiences critiques homogenes ALECTO, effectuees sur le plutonium 239, l'uranium 235 et l'uranium 233. Apres avoir rappele la description des installations, on donne les masses critiques pour des cylindres de diametres variant entre 25 et 42 cm, qui sont comparees avec d'autres chiffres (resultats etrangers, guide de criticite). Dans les gammes des diametres etudies pour des cuves a fond plat reflechies lateralement, la valeur minimale des masses critiques est la suivante: Pu239 M{sub c} = 910 {+-} 10 g, U235 M{sub c} = 1180 {+-} 12 g et U233 M{sub c} 960 {+-} 10 g. Des experiences portant sur les sections efficaces et les constantes a utiliser sur ces milieux sont ensuite presentees. Enfin des experiences de cinetique permettent une comparaison entre la methode des neutrons pulses et la methode des fluctuations. (auteur)

Improved locations of reactivity devices in future CANDU reactors fuelled with natural uranium or enriched fuels

International Nuclear Information System (INIS)

Boczar, P.G.; Van Dyk, M.T.

1987-02-01

A new configuration of reactivity devices is proposed for future CANDU reactors which improves the core characteristics with enriched fuels, while still allowing the use of natural uranium fuel. Physics calculations for this new configuration are presented for four fuel types: natural uranium, mixed plutonium - uranium oxide (MOX) having a burnup of 21 MWd/kg, and slightly enriched uranium (SEU) having burnups of either 21 or 31 MWd/kg

Ion irradiation to simulate neutron irradiation in model graphites: Consequences for nuclear graphite

Science.gov (United States)

Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

2017-10-01

Due to its excellent moderator and reflector qualities, graphite was used in CO2-cooled nuclear reactors such as UNGG (Uranium Naturel-Graphite-Gaz). Neutron irradiation of graphite resulted in the production of 14C which is a key issue radionuclide for the management of the irradiated graphite waste. In order to elucidate the impact of neutron irradiation on 14C behavior, we carried out a systematic investigation of irradiation and its synergistic effects with temperature in Highly Oriented Pyrolitic Graphite (HOPG) model graphite used to simulate the coke grains of nuclear graphite. We used 13C implantation in order to simulate 14C displaced from its original structural site through recoil. The collision of the impinging neutrons with the graphite matrix carbon atoms induces mainly ballistic damage. However, a part of the recoil carbon atom energy is also transferred to the graphite lattice through electronic excitation. The effects of the different irradiation regimes in synergy with temperature were simulated using ion irradiation by varying Sn(nuclear)/Se(electronic) stopping power. Thus, the samples were irradiated with different ions of different energies. The structure modifications were followed by High Resolution Transmission Electron Microscopy (HRTEM) and Raman microspectrometry. The results show that temperature generally counteracts the disordering effects of irradiation but the achieved reordering level strongly depends on the initial structural state of the graphite matrix. Thus, extrapolating to reactor conditions, for an initially highly disordered structure, irradiation at reactor temperatures (200 - 500 °C) should induce almost no change of the initial structure. On the contrary, when the structure is initially less disordered, there should be a "zoning" of the reordering: In "cold" high flux irradiated zones where the ballistic damage is important, the structure should be poorly reordered; In "hot" low flux irradiated zones where the ballistic

Electrolysis of acidic sodium chloride solution with a graphite anode. I. Graphite electrode

NARCIS (Netherlands)

Janssen, L.J.J.; Hoogland, J.G.

1969-01-01

A graphite anode evolving Cl from a chloride soln. is slowly oxidized to CO and CO2. This oxidn. causes a change in the characteristics of the electrode in aging, comprising a change of the nature of the graphite surface and an increase of the surface area. It appears that a new graphite electrode

Problems of natural uranium supply

Energy Technology Data Exchange (ETDEWEB)

Huwyler, S [Eidgenoessisches Inst. fuer Reaktorforschung, Wuerenlingen (Switzerland)

1977-11-01

The estimated uranium reserves in the Western World and the forecast uranium requirement in this region make the supply of nuclear power stations appear guaranteed well beyond the turn of the century. At least in the next decade it will be possible to exploit the advantageous uranium reserves in low price category, provided that prospection activities are stepped up soon and production capacities are expanded in time which are not even fully utilized today. However, difficulties could arise earlier in those countries which have no uranium reserves of their own. There is an increasing tendency among uranium producing countries to link supplies of their uranium with restrictive conditions. This makes long term contractual uranium supply guarantees a most pressing matter for those countries which have no uranium of their own. Even if the delays in the addition of new nuclear power plants are likely to improve the supply situation in the next few years, supply shortages will have to be anticipated at least from the nineties onward, unless exploitation and dressing activities are expanded considerably and also low grade ores are included in the production. At the same time it appears that the use of plutonium fueled fast breeder reactors will be unavoidable in the nineties.

Estimated natural uranium requirements to the year 2000

International Nuclear Information System (INIS)

Bennett, L.L.

1981-01-01

The future requirements for natural uranium are mainly dependent on the future growth of nuclear energy generation and the types of reactors operated to provide that energy. These topics were examined extensively by the International Nuclear Fuel Cycle Evaluation (INFCE). The resulting projections of nuclear power plant capacity and estimated requirements for natural uranium, other nuclear raw materials and fuel cycle services were presented in the final report of INFCE. The projections from INFCE are the most recent results published by an international body, and can therefore be taken as the most authoritative estimates presently available. The INFCE results have been reviewed in the light of latest trends in national nuclear power capacity figures, and a sub-set of the INFCE results are used as the basis for the demand estimates presented in this article. The principal criteria involved in the selection of this sub-set are the nuclear power growth estimates and the reactor and fuel cycle strategies. These criteria are discussed in the following sections

The relationship between natural uranium and advanced fuel cycles in CANDU reactors

International Nuclear Information System (INIS)

Lane, A.D.; McDonnell, F.N.; Griffiths, J.

1988-11-01

CANDU is the most uranium-economic type of thermal power reactor, and is the only type used in Canada. CANDU reactors consume approximately 15% of Canadian uranium production and support a fuel service industry valued at ∼$250 M/a. In addition to their once-through, natural-uranium fuel cycle, CANDU reactors are capable of operating with slightly-enriched uranium (SEU), uranium-plutonium and thorium cycles, more efficiently than other reactors. Only SEU is economically attractive in Canada now, but the other cycles are of interest to countries without indigenous fuel resources. A program is underway to establish the fuel technologies necessary for the use of SEU and the other fuel cycles in CANDU reactors. 22 refs

Uranium concentrations in natural waters, South Park, Colorado

International Nuclear Information System (INIS)

Sharp, R.R. Jr.; Aamodt, P.L.

1976-08-01

During the summer of 1975, 464 water samples from 149 locations in South Park, Colorado, were taken for the Los Alamos Scientific Laboratory in order to test the field sampling and analytical methodologies proposed for the NURE Hydrogeochemical and Stream Sediment Reconnaissance for uranium in the Rocky Mountain states and Alaska. The study showed, in the South Park area, that the analytical results do not vary significantly between samples which were untreated, filtered and acidified, filtered only, or acidified only. Furthermore, the analytical methods of fluorometry and delayed-neutron counting, as developed at the LASL for the reconnaissance work, provide fast, adequately precise, and complementary procedures for analyzing a broad range of uranium in natural waters. The data generated using this methodology does appear to identify uraniferous areas, and when applied using sound geochemical, geological, and hydrological principles, should prove a valuable tool in reconnaissance surveying to delineate new districts or areas of interest for uranium exploration

Internal contamination by natural uranium: monitoring by analysis of urine of individuals exposed by occupational inhalation

International Nuclear Information System (INIS)

Ramalho, A.T.

1982-01-01

Urine samples from men working at Usina Santo Amaro (USAM - State of Sao Paulo), a monazite refinery, were analysed for uranium concentration, using fluorometric analysis and alpha spectrometry. All samples analysed presented uranium concentration below the lower limit of detection. Theoretical values were calculated for uranium concentration in urine samples from workers at the annual limit of intake (ALI) for inhalation of natural uranium, recommended in Publication 30 of the International Commission on Radiological Protection (ICRP, 1979). The two different methods used for analysis of natural uranium concentration in the urine samples were compared: fluorimetry and alpha spectrometry. (author)

Studies on yttrium oxide coatings for corrosion protection against molten uranium

International Nuclear Information System (INIS)

Chakravarthy, Y.; Bhandari, Subhankar; Pragatheeswaran; Thiyagarajan, T.K.; Ananthapadmanabhan, P.V.; Das, A.K.; Kumar, Jay; Kutty, T.R.G.

2012-01-01

Yttrium oxide is resistant to corrosion by molten uranium and its alloys. Yttrium oxide is recommended as a protective oxide layer on graphite and metal components used for melting and processing uranium and its alloys. This paper presents studies on the efficacy of plasma sprayed yttrium oxide coatings for barrier applications against molten uranium

The chemical industry of uranium in France

International Nuclear Information System (INIS)

Goldschmidt, B.

1955-01-01

The actual CEA program is concerned with the construction of two large graphite reactors, each of those containing at least one hundred tons of uranium metal with nuclear purity. The uranium for these two reactors will be regularly supplied by new resources discovered in France and Madagascar in the last five years. The working and treatment of such ore have led to the creation of an important french industry of which the general outline and principle are described. The operated ores have got different natures and concentration, individual characteristics are described for the main ores.The most high-grade ore are transported to a central plant in Bouchet near Paris; the low-grade ore are concentrated by physical methods or chemical processes of which principles and economy are studied with constancy. The acid processes are the only used until now, although the carbonated alkaline processes has been studied in France. The next following steps after the acid process until the obtention of uranium rich concentrate are described. The purification steps of uranium compounds to nuclear purity material are described as well as the steps to elaborate metal of which the purity grade will be specify. Finally, the economic aspects of uranium production difficulty will be considered in relation with technical progresses which we can expect to achieve in the future. (M.P.)

Uranium XAFS analysis of kidney from rats exposed to uranium.

Science.gov (United States)

Kitahara, Keisuke; Numako, Chiya; Terada, Yasuko; Nitta, Kiyohumi; Shimada, Yoshiya; Homma-Takeda, Shino

2017-03-01

The kidney is the critical target of uranium exposure because uranium accumulates in the proximal tubules and causes tubular damage, but the chemical nature of uranium in kidney, such as its chemical status in the toxic target site, is poorly understood. Micro-X-ray absorption fine-structure (µXAFS) analysis was used to examine renal thin sections of rats exposed to uranyl acetate. The U L III -edge X-ray absorption near-edge structure spectra of bulk renal specimens obtained at various toxicological phases were similar to that of uranyl acetate: their edge position did not shift compared with that of uranyl acetate (17.175 keV) although the peak widths for some kidney specimens were slightly narrowed. µXAFS measurements of spots of concentrated uranium in the micro-regions of the proximal tubules showed that the edge jump slightly shifted to lower energy. The results suggest that most uranium accumulated in kidney was uranium (VI) but a portion might have been biotransformed in rats exposed to uranyl acetate.

Irradiation-induced amorphization process in graphite

Energy Technology Data Exchange (ETDEWEB)

Abe, Hiroaki [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment

1996-04-01

Effects of the element process of irradiation damage on irradiation-induced amorphization processes of graphite was studied. High orientation thermal decomposed graphite was cut about 100 nm width and used as samples. The irradiation experiments are carried out under the conditions of electronic energy of 100-400 KeV, ion energy of 200-600 KeV, ionic species Xe, Ar, Ne, C and He and the irradiation temperature at from room temperature to 900 K. The critical dose ({phi}a) increases exponentially with increasing irradiation temperature. The displacement threshold energy of graphite on c-axis direction was 27 eV and {phi}a{sup e} = 0.5 dpa. dpa is the average number of displacement to atom. The critical dose of ion irradiation ({phi}a{sup i}) was 0.2 dpa at room temperature, and amorphous graphite was produced by less than half of dose of electronic irradiation. Amorphization of graphite depending upon temperature is discussed. (S.Y.)

49 CFR 173.403 - Definitions.

Science.gov (United States)

2010-10-01

... emitters and low toxicity alpha emitters or 0.04 Bq/cm2 for all other alpha emitters. Contamination exists... containers.” Graphite means, for the purposes of § 173.453, graphite with a boron equivalent content less... A2/g. Low toxicity alpha emitters means natural uranium; depleted uranium; natural thorium; uranium...

Uranium content of petroleum by fission track technique

International Nuclear Information System (INIS)

Paschoa, A.S.; Mafra, O.Y.; Oliveira, C.A.N. de; Pinto, L.R.

1981-03-01

The feasibility of the fission track registration technique to investigate the natural uranium concentration in petroleum is examined. The application of this technique to petroleum is briefly described and discussed critically. The results obtained so far indicate uranium concentrations in samples of Brazilian petroleum which are over the detect ion limit of fission track technique. (Author) [pt

ZPR-3 Assembly 11: A cylindrical sssembly of highly enriched uranium and depleted uranium with an average 235U enrichment of 12 atom % and a depleted uranium reflector

International Nuclear Information System (INIS)

Lell, R.M.; McKnight, R.D.; Tsiboulia, A.; Rozhikhin, Y.

2010-01-01

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was 235 U or 239 Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core 235 U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 8 in the Cross Section Evaluation Working Group (CSEWG) Benchmark

Measurements of natural uranium concentration in Caspian Sea and Persian Gulf water by laser fluorimetric method

International Nuclear Information System (INIS)

Garshasbi, H.; Karimi Diba, J.; Jahanbakhshian, M. H.; Asghari, S. K.; Heravi, G. H.

2005-01-01

Natural uranium exists in earth crust and seawater. The concentration of uranium might increase by human manipulation or geological changes. The aim of this study was to verify susceptibility of laser fluorimetry method to determine the uranium concentration in Caspian Sea and Persian Gulf water. Materials and Methods: Laser fluorimetric method was used to determine the uranium concentration in several samples prepared from Caspian Sea and Persian Gulf water. Biological and chemical substances were eliminated in samples for better evaluation of the method. Results: As the concentration of natural uranium in samples increases, the response of instrument (uranium analyzer) increases accordingly. The standard deviation also increased slightly and gradually. Conclusion: Results indicate that the laser fluorimetry method show a reliable and accurate response with uranium concentration up to 100 μg/L in samples after removal of biological and organic substances

Carbon nanoencapsulation of uranium dicarbide

International Nuclear Information System (INIS)

Pasqualini, E.

1996-01-01

Nanoparticles of uranium dicarbide encapsulated in carbon smaller than 100 nm have been obtained by chemical reactions at high temperature. Two types of nanocapsules were identified and characterized. The majority of them had small diffuse kernel surfaces, with dimensions between 5 and 15 nm, surrounded by thick spherical carbon cover. Others, in minor quantity and ranging from 15 to 40 nm, were polyhedrical and surrounded with several perfect graphite layers oriented parallel to their external surface. The nanocapsules are as chemically inert as graphite. (orig.)

Carbon nanoencapsulation of uranium dicarbide

Energy Technology Data Exchange (ETDEWEB)

Pasqualini, E. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Dept. Combustibles Nucleares; Adelfang, P. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Dept. Combustibles Nucleares; Regueiro, M.N. [EPM-Matformag, CNRS, Grenoble (France)

1996-07-01

Nanoparticles of uranium dicarbide encapsulated in carbon smaller than 100 nm have been obtained by chemical reactions at high temperature. Two types of nanocapsules were identified and characterized. The majority of them had small diffuse kernel surfaces, with dimensions between 5 and 15 nm, surrounded by thick spherical carbon cover. Others, in minor quantity and ranging from 15 to 40 nm, were polyhedrical and surrounded with several perfect graphite layers oriented parallel to their external surface. The nanocapsules are as chemically inert as graphite. (orig.).

Recent developments in graphite

International Nuclear Information System (INIS)

Cunningham, J.E.

1983-01-01

Overall, the HTGR graphite situation is in excellent shape. In both of the critical requirements, fuel blocks and support structures, adequate graphites are at hand and improved grades are sufficiently far along in truncation. In the aerospace field, GraphNOL N3M permits vehicle performance with confidence in trajectories unobtainable with any other existing material. For fusion energy applications, no other graphite can simultaneously withstand both extreme thermal shock and neutron damage. Hence, the material promises to create new markets as well as to offer a better candidate material for existing applications

Uranium prospecting program: memorandum of request United Nations Assistance Rotatory Fund for Naturals resources in Uranium Prospecting

International Nuclear Information System (INIS)

1976-01-01

The Uruguayan government required assistance to Unit Nations funds with the aim of studies the Natural resources in Uranium prospecting, their antecedent, actual and projected works, equipment and end considerations

Geochemical behaviour of natural uranium-series nuclides in geological formation

International Nuclear Information System (INIS)

Yamakawa, Minoru

1991-01-01

Recent research and investigation show that the Tono uranium deposit and its natural uranium-series nuclides have been preserved, without any significant changes like re-migration or reconcentration, throughout geological events such as upheaval-submergence, marine transgression-regression, and faulting which can readily change geological, hydrogeological, and geochemical conditions. This situation might have come about as a result of being kept in a geometrical closure system, with reducing and milk alkalic geochemical conditions, from the hydrogeological and geochemical point of view. (author)

MCNP5 CRITICALITY VALIDATION AND BIAS FOR INTERMEDIATE ENRICHED URANIUM SYSTEMS

International Nuclear Information System (INIS)

Finfrock, S.H.

2009-01-01

The purpose of this analysis is to validate the Monte Carlo N-Particle 5 (MCNP5) code Version 1.40 (LA-UR-03-1987, 2005) and its cross-section database for k-code calculations of intermediate enriched uranium systems on INTEL(reg s ign) processor based PC's running any version of the WINDOWS operating system. Configurations with intermediate enriched uranium were modeled with the moderator range of 39 (le) H/Fissile (le) 1438. See Table 2-1 for brief descriptions of selected cases and Table 3-1 for the range of applicability for this validation. A total of 167 input cases were evaluated including bare and reflected systems in a single body or arrays. The 167 cases were taken directly from the previous (Version 4C [Lan 2005]) validation database. Section 2.0 list data used to calculate k-effective (k eff ) for the 167 experimental criticality benchmark cases using the MCNP5 code v1.40 and its cross section database. Appendix B lists the MCNP cross-section database entries validated for use in evaluating the intermediate enriched uranium systems for criticality safety. The dimensions and atom densities for the intermediate enriched uranium experiments were taken from NEA/NSC/DOC(95)03, September 2005, which will be referred to as the benchmark handbook throughout the report. For these input values, the experimental benchmark k eff is approximately 1.0. The MCNP validation computer runs ran to an accuracy of approximately ± 0.001. For the cases where the reported benchmark k eff was not equal to 1.0000 the MCNP calculational results were normalized. The difference between the MCNP validation computer runs and the experimentally measured k eff is the MCNP5 v1.40 bias. The USLSTATS code (ORNL 1998) was utilized to perform the statistical analysis and generate an acceptable maximum k eff limit for calculations of the intermediate enriched uranium type systems.

ZPR-3 Assembly 11 : A cylindrical sssembly of highly enriched uranium and depleted uranium with an average {sup 235}U enrichment of 12 atom % and a depleted uranium reflector.

Energy Technology Data Exchange (ETDEWEB)

Lell, R. M.; McKnight, R. D.; Tsiboulia, A.; Rozhikhin, Y.; National Security; Inst. of Physics and Power Engineering

2010-09-30

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 8 in the Cross Section Evaluation

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.)

Uranium fixation by mineralization at the redox front

International Nuclear Information System (INIS)

Isobe, Hiroshi

1998-01-01

The behavior of actinide elements including uranium in geomedia is controlled by redox conditions. Under the oxidized conditions, uranium forms uranyl ion (UO 2 2+ ) and its complexes, and dissolves in ground water. Under the reduced conditions, U(IV) has much lower solubility than uranyl ion. In the Koongarra uranium deposit, Australia, lead-bearing uraninite, uranyl lead oxide and uranyl silicate minerals occur in the unweathered, primary ore zone, and uranyl phosphate minerals occur in the weathered, secondary ore zone. Between unweathered and weathered zones, the transition zone exists as a redox front. In the transition zone, graphite and sulfide minerals react as reducing agents for species dissolved in ground water. By SEM, spherical grains of uraninite were observed in veins with graphite. Pyrite had coffinite rim with crystals of uraninite. Calculation based on the ground water chemistry and hydrology at Koongarra shows that the uranium in the transition zone may be fixed from the ground water. In the Koongarra transition zone, recent mineralization of uranium by reduction takes place. Mineralization is much stronger fixation mechanism than adsorption on clay minerals. Pyrite in the buffer materials of possible radioactive waste repositories can fix radionuclides in oxidized ground water by mineralization with reducing reactions. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-09-15

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

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

Science.gov (United States)

Sprynskyy, Myroslav; Kovalchuk, Iryna; Buszewski, Bogusław

2010-09-15

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

Benchmarking of HEU mental annuli critical assemblies with internally reflected graphite cylinder

Directory of Open Access Journals (Sweden)

Xiaobo Liu

2017-01-01

Full Text Available Three experimental configurations of critical assemblies, performed in 1963 at the Oak Ridge Critical Experiment Facility, which are assembled using three different diameter HEU annuli (15-9 inches, 15-7 inches and 13-7 inches metal annuli with internally reflected graphite cylinder are evaluated and benchmarked. The experimental uncertainties which are 0.00057, 0.00058 and 0.00057 respectively, and biases to the benchmark models which are − 0.00286, − 0.00242 and − 0.00168 respectively, were determined, and the experimental benchmark keff results were obtained for both detailed and simplified models. The calculation results for both detailed and simplified models using MCNP6-1.0 and ENDF/B-VII.1 agree well to the benchmark experimental results within difference less than 0.2%. The benchmarking results were accepted for the inclusion of ICSBEP Handbook.

Uranium concentrations in natural waters, South Park, Colorado. [Part of National Uranium Resource Evaluation program

Energy Technology Data Exchange (ETDEWEB)

Sharp, R.R. Jr.; Aamodt, P.L.

1976-08-01

During the summer of 1975, 464 water samples from 149 locations in South Park, Colorado, were taken for the Los Alamos Scientific Laboratory in order to test the field sampling and analytical methodologies proposed for the NURE Hydrogeochemical and Stream Sediment Reconnaissance for uranium in the Rocky Mountain states and Alaska. The study showed, in the South Park area, that the analytical results do not vary significantly between samples which were untreated, filtered and acidified, filtered only, or acidified only. Furthermore, the analytical methods of fluorometry and delayed-neutron counting, as developed at the LASL for the reconnaissance work, provide fast, adequately precise, and complementary procedures for analyzing a broad range of uranium in natural waters. The data generated using this methodology does appear to identify uraniferous areas, and when applied using sound geochemical, geological, and hydrological principles, should prove a valuable tool in reconnaissance surveying to delineate new districts or areas of interest for uranium exploration.

Uranium redistribution under oxidizing conditions in Oklo natural reactor zone 2, Gabon

International Nuclear Information System (INIS)

Isobe, H.; Ohnuki, T.; Murakami, T.; Gauthier-Lafaye, F.

1995-01-01

This mineralogical study was completed to elucidate the relationships between uranium distribution and alteration products of the host rock of natural reactor zone clays just below the reactor core. Uraninite is preserved without any alteration in the reactor core. Uranium minerals are found to be present in the fractures in the reactor zone clays associated with iron-mineral veins, galena and Ti-bearing minerals. Uranium, for which the phases could not be identified, occurs in iron-mineral veins and the iron-mineral rim of pyrite grains in the reactor zone clays. Uranium is not associated with granular iron minerals occurring in the illite matrix of the reactor zone clays. The degree of crystallinity and uranium content of the three iron-bearing alteration products suggest that they formed under different conditions; the granular iron minerals, under alteration conditions where uranium was not mobilized while the iron-mineral veins and the iron-mineral rim of pyrite, under conditions in which uranium is mobilized after the formation of the granular iron minerals

From a critical assembly heavy water - natural uranium to the fast - thermal research reactor in the Institute Vinca

International Nuclear Information System (INIS)

Stefanovic, D.; Pesic, M.

1995-01-01

A part of the Institute in Vinca this monograph refers to is the thermal nuclear zero power reactor RB, with a heavy water moderator and variously enriched uranium fuel, that is, its present day version, the coupled fast-thermal system HERBE. A group of research workers, technicians, operators and skilled workmen in the workshop have worked continuously on it. Some of them have spent their whole working age at the reactor, and some a part of it. There is about a hundred and fifty internationally published papers, twenty master's and fourteen doctor's theses left behind them for the past thirty five years. This book is devoted to them. The first part of the text refers to the pioneering efforts on the reactor and fundamental research in reactor physics. The experimental reactor RB was designed and constructed at the time to operate with natural uranium and heavy water. Measurements are presented and the first results of reaching critical state, measurements of migration length of thermal neutrons and neutron multiplication factor in an infinite medium; also measurements of neutron flux density distribution and reactor parameter, and in the domain of safety, measurement of safety rods reactivity. Those were also the times when the known serious accident occurred with the uncontrolled rise of reactivity, which was especially minutely described in a publication of the International Atomic Energy Agency from Vienna. Later on, new fuel was acquired with 2 % enriched uranium. A series of experiments in reactor and neutron physics followed, with just the most interesting results of them presented here. In the period which followed, another type of fuel was available, with 80 % enriched uranium. New possibilities for work opened. Measurements with mixed lattices were performed, and the RA reactor lattices were simulated. After measurements mainly in the sphere of reactor and neutron physics, a need for investigations in the field of gamma and neutron radiation protection

Idaho National Engineering Laboratory materials in inventory natural and enriched uranium management and storage costs

International Nuclear Information System (INIS)

Nebeker, R.L.

1995-11-01

On July 13, 1994, the Office of Environmental Management (EM) was requested to develop a planning process that would result in management policies for dealing with nuclear materials in inventory. In response to this request, EM launched the Materials In Inventory (MIN) Initiative. A Headquarters Working Group was established to develop the broad policy framework for developing MIN management policies. MIN activities cover essentially all nuclear materials within the DOE complex, including such items as spent nuclear fuel, depleted uranium, plutonium, natural and enriched uranium, and other materials. In August 1995, a report discussing the natural and enriched uranium portion of the Initiative for the Idaho National Engineering Laboratory (INEL) was published. That report, 'Idaho National Engineering Laboratory Materials-in-Inventory, Natural and Enriched Uranium'.' identified MIN under the control of Lockheed Idaho Technologies Company at the INEL. Later, additional information related to the costs associated with the storage of MIN materials was requested to supplement this report. This report provides the cost information for storing, disposing, or consolidating the natural and enriched uranium portion of the MIN materials at the INEL. The information consists of eight specific tables which detail present management costs and estimated costs of future activities

Linearity assumption in soil-to-plant transfer factors of natural uranium and radium in Helianthus annuus L

International Nuclear Information System (INIS)

Rodriguez, P. Blanco; Tome, F. Vera; Fernandez, M. Perez; Lozano, J.C.

2006-01-01

The linearity assumption of the validation of soil-to-plant transfer factors of natural uranium and 226 Ra was tested using Helianthus annuus L. (sunflower) grown in a hydroponic medium. Transfer of natural uranium and 226 Ra was tested in both the aerial fraction of plants and in the overall seedlings (roots and shoots). The results show that the linearity assumption can be considered valid in the hydroponic growth of sunflowers for the radionuclides studied. The ability of sunflowers to translocate uranium and 226 Ra was also investigated, as well as the feasibility of using sunflower plants to remove uranium and radium from contaminated water, and by extension, their potential for phytoextraction. In this sense, the removal percentages obtained for natural uranium and 226 Ra were 24% and 42%, respectively. Practically all the uranium is accumulated in the roots. However, 86% of the 226 Ra activity concentration in roots was translocated to the aerial part

Linearity assumption in soil-to-plant transfer factors of natural uranium and radium in Helianthus annuus L

Energy Technology Data Exchange (ETDEWEB)

Rodriguez, P. Blanco [Departamento de Fisica, Facultad de Ciencias, Universidad de Extremadura, 06071 Badajoz (Spain); Tome, F. Vera [Departamento de Fisica, Facultad de Ciencias, Universidad de Extremadura, 06071 Badajoz (Spain)]. E-mail: fvt@unex.es; Fernandez, M. Perez [Area de Ecologia, Departamento de Fisica, Facultad de Ciencias, Universidad de Extremadura, 06071 Badajoz (Spain); Lozano, J.C. [Laboratorio de Radiactividad Ambiental, Facultad de Ciencias, Universidad de Salamanca, 37008 Salamanca (Spain)

2006-05-15

The linearity assumption of the validation of soil-to-plant transfer factors of natural uranium and {sup 226}Ra was tested using Helianthus annuus L. (sunflower) grown in a hydroponic medium. Transfer of natural uranium and {sup 226}Ra was tested in both the aerial fraction of plants and in the overall seedlings (roots and shoots). The results show that the linearity assumption can be considered valid in the hydroponic growth of sunflowers for the radionuclides studied. The ability of sunflowers to translocate uranium and {sup 226}Ra was also investigated, as well as the feasibility of using sunflower plants to remove uranium and radium from contaminated water, and by extension, their potential for phytoextraction. In this sense, the removal percentages obtained for natural uranium and {sup 226}Ra were 24% and 42%, respectively. Practically all the uranium is accumulated in the roots. However, 86% of the {sup 226}Ra activity concentration in roots was translocated to the aerial part.

Uranium half-lives: a critical review

International Nuclear Information System (INIS)

Holden, N.E.

1981-01-01

The experimental data are evaluated and values for the spontaneous fission half-life of 238 U and the total half-lives for 232 U, 233 U, 234 U, 235 U, 236 U, and 238 U are recommended. Also the variation of the isotopic abundance of 234 U in nature and the error involved in the assumption of secular equilibrium between 234 U and 238 U in the determination of the specific activity of natural uranium samples are discussed. The recommended half-life values and 95% confidence limits are: 238 U spontaneous fission: 8.09 +- 0.26 x 10 15 years; 232 U total: 69.8 +- 1.0 years; 233 U total: 1.592 +- 0.002 x 10 5 years; 234 U total: 2.454 +- 0.006 x 10 5 years; 235 U total: 7.037 +- 0.011 x 10 8 years; 236 U total: 2.342 +- 0.003 x 10 7 years 238 U total: 4.468 +- 0.005 x 10 9 years

Structure and Performance of Epoxy Resin Cladded Graphite Used as Anode

Science.gov (United States)

Zhou, Zhentao; Li, Haijun

This paper is concerning to prepare modified natural graphite which is low-cost and advanced materials used as lithium ion battery anode using the way of cladding natural graphite with epoxy resin. The results shows that the specific capacity and circular performance of the modified natural graphite, which is prepared in the range of 600°C and 1000°C, have been apparently improved compare with the not-modified natural graphite. The first reversible capacity of the modified natural graphite is 338mAh/g and maintain more than 330mAh/g after 20 charge/discharge circles.

Maximum permissible concentrations of uranium in air

CERN Document Server

Adams, N

1973-01-01

The retention of uranium by bone and kidney has been re-evaluated taking account of recently published data for a man who had been occupationally exposed to natural uranium aerosols and for adults who had ingested uranium at the normal dietary levels. For life-time occupational exposure to uranium aerosols the new retention functions yield a greater retention in bone and a smaller retention in kidney than the earlier ones, which were based on acute intakes of uranium by terminal patients. Hence bone replaces kidney as the critical organ. The (MPC) sub a for uranium 238 on radiological considerations using the current (1959) ICRP lung model for the new retention functions is slightly smaller than for earlier functions but the (MPC) sub a determined by chemical toxicity remains the most restrictive.

Elimination of natural uranium and 226Ra from contaminated waters by rhizofiltration using Helianthus annuus L

International Nuclear Information System (INIS)

Vera Tome, F.; Blanco Rodriguez, P.; Lozano, J.C.

2008-01-01

The elimination of natural uranium and 226 Ra from contaminated waters by rhizofiltration was tested using Helianthus annuus L. (sunflower) seedlings growing in a hydroponic medium. Different experiments were designed to determine the optimum age of the seedlings for the remediation process, and also to study the principal way in which the radionuclides are removed from the solution by the sunflower roots. In every trial a precipitate appeared which contained a major fraction of the natural uranium and 226 Ra. The results indicated that the seedlings themselves induced the formation of this precipitate. When four-week-old seedlings were exposed to contaminated water, a period of only 2 days was sufficient to remove the natural uranium and 226 Ra from the solution: about 50% of the natural uranium and 70% of the 226 Ra were fixed in the roots, and essentially the rest was found in the precipitate, with only very small percentages fixed in the shoots and left in solution

Natural analogue study of uranium deposits in Japan with special reference to the Tono uranium deposit

International Nuclear Information System (INIS)

Komuro, Kosei; Sasao, Eiji

2004-05-01

In order to verify the safety assessment for geological disposal system of high-level radioactive waste, it is necessary to evaluate properly the stability of the disposal system under natural hydrogeological environment over long period of time (ten to hundred thousands years). For the safety assessment for that in the Japanese Islands, many geological processes inherent in the tectonically active Island-Arc system should be also taken into consideration in addition to those in stable continental environment. However, it is difficult because some processes such as earthquake seem to be accidental and some are periodic or gradual over our life scale. The uranium deposits in Japan are subjected to many geological processes inherent in the tectonically active Island-Arc system. The studies on long-term preservation of uranium deposits in Japan from a natural analogue viewpoint would be expected to provide useful information for the assessment in the Japanese Islands over long period of time. In order to understand the behavior of radionuclides under natural hydrogeological environment in Japanese Islands over long period of time, the uranium deposits in Japan, especially of the Tono uranium deposit was investigated from a natural analogue viewpoint under the course of joint research program by University of Tsukuba and Japan Nuclear Cycle Development Institute. Important conclusions obtained in the present study are summarized as follows: The migration behavior of the radionuclides in the granite area is mainly controlled by the stability of original minerals in oxic condition, being due to poor reducing agents such as organic matter and sulfide minerals. In the case of hydrothermal alteration, yttrialite and fergusonite were decomposed and thorogummite was formed at the altered part, whereas zircon and allanite have not been significantly altered. In the case of weathering, autunite and torbernite were formed, probably due to the high phosphorus weathering

Measurement of critical mass for an assembly of bare uranium shells

International Nuclear Information System (INIS)

Myers, W.L.; Goulding, C.A.; Hollas, C.L.

1997-01-01

As part of the research into nuclear measurement techniques, a series of measurements was performed that have applications to criticality safety and nuclear material handling. The critical mass of a set of bare, enriched-uranium metal hemispherical shells, known as the Rocky Flats shells, was measured for an assembly having an inside radius of 2.347 cm. The critical mass value was extrapolated from a series of subcritical measurements using three different kinds of sources (AmBe, AmF, and 252 Cf) placed at the center of the shells. Two kinds of neutron detection configurations (a 1% efficiency and a 25% efficiency configuration) were used to make the measurements

Field technique for the measurement of uranium in natural waters

Energy Technology Data Exchange (ETDEWEB)

Robbins, J C [Scintrex Ltd., Concord, Ontario

1978-05-01

An analytical method suitable for field determination of trace levels of uranium in natural waters is described. Laser UV radiation causes persistent fluorescence of a uranyl complex. Electronic gating substantially rejects detection of short-lived natural organic matter fluorescence. Further work is required on effects of interferences in samples with complex matrices and interpretative aids such as concurrent conductivity and organic content measurements.

Aquifer restoration at in-situ leach uranium mines: evidence for natural restoration processes

International Nuclear Information System (INIS)

Deutsch, W.J.; Serne, R.J.; Bell, N.E.; Martin, W.J.

1983-04-01

Pacific Northwest Laboratory conducted experiments with aquifer sediments and leaching solution (lixiviant) from an in-situ leach uranium mine. The data from these laboratory experiments and information on the normal distribution of elements associated with roll-front uranium deposits provide evidence that natural processes can enhance restoration of aquifers affected by leach mining. Our experiments show that the concentration of uranium (U) in solution can decrease at least an order of magnitude (from 50 to less than 5 ppM U) due to reactions between the lixiviant and sediment, and that a uranium solid, possibly amorphous uranium dioxide, (UO 2 ), can limit the concentration of uranium in a solution in contact with reduced sediment. The concentrations of As, Se, and Mo in an oxidizing lixiviant should also decrease as a result of redox and precipitation reactions between the solution and sediment. The lixiviant concentrations of major anions (chloride and sulfate) other than carbonate were not affected by short-term (less than one week) contact with the aquifer sediments. This is also true of the total dissolved solids level of the solution. Consequently, we recommend that these solution parameters be used as indicators of an excursion of leaching solution from the leach field. Our experiments have shown that natural aquifer processes can affect the solution concentration of certain constituents. This effect should be considered when guidelines for aquifer restoration are established

Radioactive wastes of uranium mining and milling: Radiological consequences for human population and natural environment

International Nuclear Information System (INIS)

Sazykina, T.G.; Kryshev, I.I.

2002-01-01

The sources of wastes and levels of radioactive contamination are considered in the areas of uranium ore mining and milling. Assessments of doses to the population are made using the methodology of multiple sources and pathways of exposure, including calculations of inhalation dose and doses from consumption of contaminated agricultural and natural products, as well as external exposure from the radioactive cloud and soil. On the local (0-100 km) spatial scale, the dose from uranium mining and processing is, on average, about 0.7 man Sv (GWa) -1 . The most significant pathway of the population exposure is inhalation of radon. The impact of uranium ore mining and processing on natural flora and fauna is determined by specific characteristics of the production at uranium mining enterprises and has both radiation and non-radiation components. The estimates of external and internal exposures to the natural biota in the vicinity of hydro-metallurgical works and tailing dumps are presented. (author)

Benchmark critical experiments on low-enriched uranium oxide systems with H/U = 0.77

International Nuclear Information System (INIS)

Tuck, G.; Oh, I.

1979-08-01

Ten benchmark experiments were performed at the Critical Mass Laboratory at Rockwell International's Rocky Flats Plant, Golden, Colorado, for the US Nuclear Regulatory Commission. They provide accurate criticality data for low-enriched damp uranium oxide (U 3 O 8 ) systems. The core studied consisted of 152 mm cubical aluminum cans containing an average of 15,129 g of low-enriched (4.46% 235 U) uranium oxide compacted to a density of 4.68 g/cm 3 and with an H/U atomic ratio of 0.77. One hundred twenty five (125) of these cans were arranged in an approx. 770 mm cubical array. Since the oxide alone cannot be made critical in an array of this size, an enriched (approx. 93% 235 U) metal or solution driver was used to achieve criticality. Measurements are reported for systems having the least practical reflection and for systems reflected by approx. 254-mm-thick concrete or plastic. Under the three reflection conditions, the mass of the uranium metal driver ranged from 29.87 kg to 33.54 kg for an oxide core of 1864.6 kg. For an oxide core of 1824.9 kg, the weight of the high concentration (351.2 kg U/m 3 ) solution driver varied from 14.07 kg to 16.14 kg, and the weight of the low concentration (86.4 kg U/m 3 ) solution driver from 12.4 kg to 14.0 kg

Parametric analyses of planned flowing uranium hexafluoride critical experiments

Science.gov (United States)

Rodgers, R. J.; Latham, T. S.

1976-01-01

Analytical investigations were conducted to determine preliminary design and operating characteristics of flowing uranium hexafluoride (UF6) gaseous nuclear reactor experiments in which a hybrid core configuration comprised of UF6 gas and a region of solid fuel will be employed. The investigations are part of a planned program to perform a series of experiments of increasing performance, culminating in an approximately 5 MW fissioning uranium plasma experiment. A preliminary design is described for an argon buffer gas confined, UF6 flow loop system for future use in flowing critical experiments. Initial calculations to estimate the operating characteristics of the gaseous fissioning UF6 in a confined flow test at a pressure of 4 atm, indicate temperature increases of approximately 100 and 1000 K in the UF6 may be obtained for total test power levels of 100 kW and 1 MW for test times of 320 and 32 sec, respectively.

Obtention of nuclear grade graphite

International Nuclear Information System (INIS)

Ferreira, M.L.

1984-01-01

The impurity level of natural graphite found in some of the most important mines of the State of Minas Gerais - Brasil is determined. It is also concerned with the development and use of natural graphite in nuclear reactors. Standard methods for chemical and instrumentsal analysis such as Spectrografic Determination by Emission, Spectrografic Determination by X-Rays, Spectrografic Determination by Atomic Asorption, Photometric Determination, and also chemical and physical methods for separation of impurities as well standard method for Estimating the Thermal Neutron Absorption Cross Section of graphite were employed. Some aditionals methods of purification to the ordinary treatment such as the use of metanol and halogens are also described. (Author) [pt

Validation of the Monte Carlo Criticality Program KENO V. a for highly-enriched uranium systems

Energy Technology Data Exchange (ETDEWEB)

Knight, J.R.

1984-11-01

A series of calculations based on critical experiments have been performed using the KENO V.a Monte Carlo Criticality Program for the purpose of validating KENO V.a for use in evaluating Y-12 Plant criticality problems. The experiments were reflected and unreflected systems of single units and arrays containing highly enriched uranium metal or uranium compounds. Various geometrical shapes were used in the experiments. The SCALE control module CSAS25 with the 27-group ENDF/B-4 cross-section library was used to perform the calculations. Some of the experiments were also calculated using the 16-group Hansen-Roach Library. Results are presented in a series of tables and discussed. Results show that the criteria established for the safe application of the KENO IV program may also be used for KENO V.a results.

Validation of the Monte Carlo Criticality Program KENO V.a for highly-enriched uranium systems

International Nuclear Information System (INIS)

Knight, J.R.

1984-11-01

A series of calculations based on critical experiments have been performed using the KENO V.a Monte Carlo Criticality Program for the purpose of validating KENO V.a for use in evaluating Y-12 Plant criticality problems. The experiments were reflected and unreflected systems of single units and arrays containing highly enriched uranium metal or uranium compounds. Various geometrical shapes were used in the experiments. The SCALE control module CSAS25 with the 27-group ENDF/B-4 cross-section library was used to perform the calculations. Some of the experiments were also calculated using the 16-group Hansen-Roach Library. Results are presented in a series of tables and discussed. Results show that the criteria established for the safe application of the KENO IV program may also be used for KENO V.a results

Decommissioning: A critical component of the design for uranium tailings management facilities

International Nuclear Information System (INIS)

Clifton, W.A.; Barsi, R.G.; Misfeldt, G.A.

2000-01-01

Uranium was discovered in the Beaverlodge area of northern Saskatchewan in 1934 with the first major mill beginning operation in 1953. Little attention was paid to tailings quality or tailings management practices. With the onset of the modem uranium operations beginning in the late 1970's, it was repeatedly evident, that the public had significant concerns, particularly with respect to tailings management, that must be addressed if the developments were to be allowed to proceed. Primary considerations related to environmental protection, public safety and an assurance of the ongoing sustainable development of the region. Integrating the decommissioning of a mine/mill site into development planning from the very outset has proven to be a critical component that has contributed to the ongoing success of the Saskatchewan uranium operations. This paper will provide a case study of the evolution of the uranium tailings management technology utilized in Saskatchewan. It documents the evolution of tailings management processes and the characteristics of tailings produced by successive mines in northern Saskatchewan. It also discusses the evolution of technologies applied to management of uranium mill tailings and demonstrates how progressively increasing levels of environmental protection have been achieved during the last 47 years of uranium mill operation. The paper also shows that the planned and progressive decommissioning of an operational site is the key to: Minimizing environmental impacts; Satisfying public and regulatory concerns; Minimizing operational and decommissioning costs; Minimizing corporate liability; and Shifting public resistance to public support. (author)

Decommissioning: A critical component of the design for uranium tailings management facilities

International Nuclear Information System (INIS)

Clifton, A.W.; Barsi, R.G.; Misfeldt, G.A.

2002-01-01

Uranium was discovered in the Beaverlodge area of northern Saskatchewan in 1934 with the first major mill beginning operation in 1953. Little attention was paid to tailings quality or tailings management practices. With the onset of the modern uranium operations beginning in the late 1970's, it was repeatedly evident, that the public had significant concerns, particularly with respect to tailings management, that must be addressed if the developments were to be allowed to proceed. Primary considerations related to environmental protection, public safety and an assurance of the ongoing sustainable development of the region. Integrating the decommissioning of a mine/mill site into development planning from the very outset has proven to be a critical component that has contributed to the ongoing success of the Saskatchewan uranium operations. This paper will provide a case study of the evolution of the uranium tailings management technology utilized in Saskatchewan. It documents the evolution of tailings management processes and the characteristics of tailings produced by successive mines in northern Saskatchewan. It also discusses the evolution of technologies applied to management of uranium mill tailings and demonstrates how progressively increasing levels of environmental protection have been achieved during the last 47 years of uranium mill operation. The paper also shows that the planned and progressive decommissioning of an operational site is the key to: Minimizing environmental impacts; Satisfying public and regulatory concerns; Minimizing operational and decommissioning costs; Minimizing corporate liability; and Shifting public resistance to public support. (author)

SRTC criticality safety technical review: Nuclear criticality safety evaluation 94-02, uranium solidification facility pencil tank module spacing

International Nuclear Information System (INIS)

Rathbun, R.

1994-01-01

Review of NMP-NCS-94-0087, ''Nuclear Criticality Safety Evaluation 94-02: Uranium Solidification Facility Pencil Tank Module Spacing (U), April 18, 1994,'' was requested of the SRTC Applied Physics Group. The NCSE is a criticality assessment to show that the USF process module spacing, as given in Non-Conformance Report SHM-0045, remains safe for operation. The NCSE under review concludes that the module spacing as given in Non-Conformance Report SHM-0045 remains in a critically safe configuration for all normal and single credible abnormal conditions. After a thorough review of the NCSE, this reviewer agrees with that conclusion

Temperature Dependence of Uranium and Vanadium Adsorption on Amidoxime-Based Adsorbents in Natural Seawater

Energy Technology Data Exchange (ETDEWEB)

Kuo, Li-Jung [Marine Sciences Laboratory, Pacific Northwest National Laboratory, Sequim WA 98382 USA; Gill, Gary A. [Marine Sciences Laboratory, Pacific Northwest National Laboratory, Sequim WA 98382 USA; Tsouris, Costas [Oak Ridge National Laboratory, Oak Ridge TN 37831 USA; Rao, Linfeng [Chemical Sciences Division, Lawrence Berkeley National Laboratory, Berkeley CA 94720 USA; Pan, Horng-Bin [Department of Chemistry, University of Idaho, Moscow ID 83844 USA; Wai, Chien M. [Department of Chemistry, University of Idaho, Moscow ID 83844 USA; Janke, Christopher J. [Oak Ridge National Laboratory, Oak Ridge TN 37831 USA; Strivens, Jonathan E. [Marine Sciences Laboratory, Pacific Northwest National Laboratory, Sequim WA 98382 USA; Wood, Jordana R. [Marine Sciences Laboratory, Pacific Northwest National Laboratory, Sequim WA 98382 USA; Schlafer, Nicholas [Marine Sciences Laboratory, Pacific Northwest National Laboratory, Sequim WA 98382 USA; D' Alessandro, Evan K. [Rosensteil School of Marine and Atmospheric Chemistry, University of Miami, Miami FL 33149 USA

2018-01-16

The apparent enthalpy and entropy of the complexation of uranium (VI) and vanadium (V) with amidoxime ligands grafted onto polyethylene fiber was determined using time series measurements of adsorption capacities in natural seawater at three different temperatures. The complexation of uranium was highly endothermic, while the complexation of vanadium showed minimal temperature sensitivity. Amidoxime-based polymeric adsorbents exhibit significantly increased uranium adsorption capacities and selectivity in warmer waters.

Analysis of nuclear grade uranium oxides by atomic absorption spectrometry with electrothermal atomization

International Nuclear Information System (INIS)

Batistoni, D.A.; Erlijman, L.H.; Pazos, A.L.

1986-01-01

The application of atomic absorption spectrometry for the determination of five trace impurities in nuclear grade uranium oxides is described. The elements were separated from the uranium matrix by extraction chromatography and determined in 5.5 M nitric acid by electrothermal atomization using pyrolytic graphite coated tubes. Two elements, cadmium and chromium, with different volatility characteristics were employed to investigate the operating conditions. Drying and ashing conditions were studied for both elements. Ramp and constant potential (step) heating modes have also been studied and compared. Good reproducibility and a longer life of graphite tubes were obtained with ramp atomization. Detection limits (in micrograms per gram of uranium) were: Cd 0.01; Cr 0.1; Cu 0.4; Mn 0.04 and Ni 0.2. (author) [es

Simultaneous determination of Ra-226, natural uranium and natural thorium by gamma-ray spectrometry INa(Ti), in solid samples

International Nuclear Information System (INIS)

Salvador, S.; Navarro, T.; Alvarez, A.

1991-01-01

A method has been developed to determine activities of Ra-226, natural uranium and natural thorium by gamma-ray spectrometry. The measurement system has been calibrated using standards specially prepared at the laboratory. It is necessary to assume secular equilibrium in the samples, between Ra-226 and Th-232 and its daughters nuclides, and between U-238 and its immediate daughter Th-234, as the photo peaks measured are those of the daughters. The results obtained indicate that this method can of ter replace the radiochemical techniques used to measure activities in this type of sample. The method has been successfully used to determine these natural isotopes in samples from uranium mills. (Author) 9 refs

Cycle for fuel elements. Uranium production, programs for nuclear power stations and capital expenditure involved; Cycles de combustibles. Production d'uranium, programme de centrales electriques et effort financier correspondant

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

A number of different possible programs for nuclear power stations of various types are presented in this survey. These programs are established in relation to the use of uranium and thorium in amounts similar to those that shall probably be produced in France during the next fifteen years. As it is possible to draw plans for nuclear power stations in which several processes exist simultaneously, an unlimited number of variations being thinkable, this survey is limited to successive analysis of the results obtained by use of only one of each of the following three systems: - system natural uranium-graphite, - system natural uranium-heavy water, -system enriched uranium-pressurised light water. All schemes are considered as assemblages of these three simple systems. The effects of plutonium recycling are also considered for each system. The electric power installed and the capacity of stations situated up-stream and down-stream have been calculated by this method and an attempt has been made to establish the sum to be invested during the fifteen years necessary for the launching of the programs scheduled. A table of timing for the investments groups the results obtained. Considering the fact that French availabilities in capital shall not be unlimited during the coming years, this way of presenting the results seems to be interesting. (author)Fren. [French] L'etude presentee comporte l'examen d'un certain nombre d'hypotheses de programmes de centrales nucleaires de types differents. Ces programmes correspondent a l'utilisation de tonnages d'uranium et de thorium de l'ordre de grandeur de ceux qui seront probablement produits par la France dans les quinze prochaines annees. Comme il est possible de batir un programme de centrales nucleaires, comportant a la fois plusieurs filieres suivant des variantes en nombre infini, on s'est contente d'examiner successivement les resultats ous si on utilisait exclusivement l'une des trois filieres suivantes: - filiere uranium

Determination of 226Ra and natural uranium concentration in Botafogo river

International Nuclear Information System (INIS)

Nascimento, M.B. do; Amaral, R.S.; Khoury, H.J.; Andrade Lima, R. de

1990-01-01

In the Brazilian Northeast region at the coastal area from Pernambuco to Paraiba there is a 4 km wide strip deposit of phosphate rock. This phosphate is used to produce fertilizes by a factory located at the border of the Botafogo river, which cross this area. The phosphate is associated with uranium and no research has been conducted on the river radioactive contamination due the natural processes and to the fertizer factory the present investigation was undertaken to determine 226 Ra and natural uranium concentration in the river water, near the factory. Results show that the radionuclide concentration increases sharply in front of the place of the factory discharge and then decreases rapidly to the same levels found before the factory, 0,01 Bq/1. (author) [pt

Reactivity of Uranium and Ferrous Iron with Natural Iron Oxyhydroxides.

Science.gov (United States)

Stewart, Brandy D; Cismasu, A Cristina; Williams, Kenneth H; Peyton, Brent M; Nico, Peter S

2015-09-01

Determining key reaction pathways involving uranium and iron oxyhydroxides under oxic and anoxic conditions is essential for understanding uranium mobility as well as other iron oxyhydroxide mediated processes, particularly near redox boundaries where redox conditions change rapidly in time and space. Here we examine the reactivity of a ferrihydrite-rich sediment from a surface seep adjacent to a redox boundary at the Rifle, Colorado field site. Iron(II)-sediment incubation experiments indicate that the natural ferrihydrite fraction of the sediment is not susceptible to reductive transformation under conditions that trigger significant mineralogical transformations of synthetic ferrihydrite. No measurable Fe(II)-promoted transformation was observed when the Rifle sediment was exposed to 30 mM Fe(II) for up to 2 weeks. Incubation of the Rifle sediment with 3 mM Fe(II) and 0.2 mM U(VI) for 15 days shows no measurable incorporation of U(VI) into the mineral structure or reduction of U(VI) to U(IV). Results indicate a significantly decreased reactivity of naturally occurring Fe oxyhydroxides as compared to synthetic minerals, likely due to the association of impurities (e.g., Si, organic matter), with implications for the mobility and bioavailability of uranium and other associated species in field environments.

The French natural uranium industry in 1986

International Nuclear Information System (INIS)

Baron, Marcel

1987-01-01

France has relatively large uranium deposits. This led to the creation of an internationally significant uranium mining industry. The structure of this industry is explained. In 1985 world supply of uranium was greater than world demand leading to an increase in uranium stocks. However, as demand is expected to increase, the industry is undertaking extensive uranium exploration, mainly abroad. (UK)

Determination of uranium and thorium contents inside different materials using track detectors and mean critical angles

CERN Document Server

Misdaq, M A; Ktata, A; Merzouki, A; Youbi, N

1999-01-01

The critical angles of the CR-39 (theta sub c) and LR-115 type II (theta sub c ') solid state nuclear track detectors (SSNTD) for detecting alpha-particles emitted by the uranium and thorium series have been evaluated by calculating the corresponding ranges of the emitted alpha-particles in different material samples and in the SSNTD studied. The influence of the emitted alpha-particles initial and residual energies on the critical angles of the SSNTD studied has been investigated. The uranium and thorium contents of different geological samples have been evaluated by exploiting data obtained for the critical angles of the CR-39 and LR-115 type II solid state nuclear track detectors and measuring the corresponding densities of tracks.

The theory and uses of natural uranium isotopic variations in hydrology

International Nuclear Information System (INIS)

Osmond, J.K.; Cowart, J.B.

1976-01-01

The dissolved concentration of uranium and the relative abundance of two uranium isotopes, 234 U and 238 U, vary over a wide range of values in natural waters. The concentration is controlled mainly by the redox potential of the environment and by CO 2 . The mechanism of isotope fractionation is thought to be entrainment of 234 U in the aqueous phase either by selective leaching of the solid phase or by direct recoil of the daughter nuclide. Ion exchange techniques and alpha-spectrometry permit the measurement of uranium at concentrations as low as pp 10 11 and the isotopic ratio to a few per cent. In oxidizing conditions the uranium isotopes behave in a chemically stable conservative manner such that separate groundwater sources may have identifiably different characteristics and mixing volume calculations may be made. Other potential use of these isotopes include radiometric dating, tracing of hydrologic systems, ore prospecting and earthquake prediction. (author)

Brazing graphite to graphite

International Nuclear Information System (INIS)

Peterson, G.R.

1976-01-01

Graphite is joined to graphite by employing both fine molybdenum powder as the brazing material and an annealing step that together produce a virtually metal-free joint exhibiting properties similar to those found in the parent graphite. Molybdenum powder is placed between the faying surfaces of two graphite parts and melted to form molybdenum carbide. The joint area is thereafter subjected to an annealing operation which diffuses the carbide away from the joint and into the graphite parts. Graphite dissolved by the dispersed molybdenum carbide precipitates into the joint area, replacing the molybdenum carbide to provide a joint of graphite

Criticality safety for deactivation of the Rover dry headend process

International Nuclear Information System (INIS)

Henrikson, D.J.

1995-01-01

The Rover dry headend process combusted Rover graphite fuels in preparation for dissolution and solvent extraction for the recovery of 235 U. At the end of the Rover processing campaign, significant quantities of 235 U were left in the dry system. The Rover Dry Headend Process Deactivation Project goal is to remove the remaining uranium bearing material (UBM) from the dry system and then decontaminate the cells. Criticality safety issues associated with the Rover Deactivation Project have been influenced by project design refinement and schedule acceleration initiatives. The uranium ash composition used for calculations must envelope a wide range of material compositions, and yet result in cost effective final packaging and storage. Innovative thinking must be used to provide a timely safety authorization basis while the project design continues to be refined

Uranium enrichment plans

International Nuclear Information System (INIS)

Thomas, D.C.; Gagne, R.W.

1978-01-01

The following topics are covered: the status of the Government's existing uranium enrichment services contracts, natural uranium requirements based on the latest contract information, uncertainty in predicting natural uranium requirements based on uranium enrichment contracts, and domestic and foreign demand assumed in enrichment planning

Uranium mobility in the natural environment - evidence from sedimentary roll-front deposits

International Nuclear Information System (INIS)

Deutsch, W.J.; Serne, R.J.

1983-04-01

Roll-front deposits consist of naturally occurring ore-grade uranium in selected sandstone aquifers throughout the world. The geochemical environment of these roll-front deposits is analogous to the environment of a radioactive waste repository containing redox-sensitive elements during its post-thermal period. The ore deposits are formed by a combination of dissolution, complexation, sorption/precipitation, and mineral formation processes. The uranium, leached from the soil by percolating rainwater, complexes with dissolved carbonate and moves in the oxidizing ground water at very low concentration (parts per billion) levels. The uranium is extracted from the leaching solution by the chemical processes, over long periods of time, at the interfaces between oxidized and reduced sediments. The Eh of the ground water associated with the reduced sediments (Eh = -100 mv to +100 mv) is higher than the Eh expected for most waste repository environments (Eh = -100 mv to -300 mv); this suggests that uranium solids will not be very soluble in the repositories. Data from in-situ leach mining and restoration of roll-front uranium deposits also provide information on the potential mobility of the waste if oxidizing ground water should enter the repository. Uranium solids probably will be initially very soluble in carbonate ground water; however, as reducing conditions are re-estblished through water/rock interactions, the uranium will reprecipitate and the amount of uranium in solution will again equilibrate with the reduced uranium minerals

Nuclear criticality safety assessment of the Consolidated Edison Uranium-Solidification Program Facility

International Nuclear Information System (INIS)

Thomas, J.T.

1984-01-01

A nuclear criticality assessment of the Consolidated Edison Uranium-Solidification Program facility confirms that all operations involved in the process may be conducted with an acceptable margin of subcriticality. Normal operation presents no concern since subcriticality is maintained by design. Several recommendations are presented to prevent, or mitigate the consequences of, any abnormal events that might occur in the various portions of the process. These measures would also serve to reduce to a minimum the administrative controls required to prevent criticality

New insights into canted spiro carbon interstitial in graphite

Science.gov (United States)

EL-Barbary, A. A.

2017-12-01

The self-interstitial carbon is the key to radiation damage in graphite moderator nuclear reactor, so an understanding of its behavior is essential for plant safety and maximized reactor lifetime. The density functional theory is applied on four different graphite unit cells, starting from of 64 carbon atoms up to 256 carbon atoms, using AIMPRO code to obtain the energetic, athermal and mechanical properties of carbon interstitial in graphite. This study presents first principles calculations of the energy of formation that prove its high barrier to athermal diffusion (1.1 eV) and the consequent large critical shear stress (39 eV-50 eV) necessary to shear graphite planes in its presence. Also, for the first time, the gamma surface of graphite in two dimensions is calculated and found to yield the critical shear stress for perfect graphite. Finally, in contrast to the extensive literature describing the interstitial of carbon in graphite as spiro interstitial, in this work the ground state of interstitial carbon is found to be canted spiro interstitial.

Elimination of natural uranium and {sup 226}Ra from contaminated waters by rhizofiltration using Helianthus annuus L

Energy Technology Data Exchange (ETDEWEB)

Vera Tome, F. [Departamento de Fisica Aplicada, Facultad de Ciencias, Universidad de Extremadura, 06071 Badajoz (Spain)], E-mail: fvt@unex.es; Blanco Rodriguez, P. [Departamento de Fisica, Facultad de Ciencias, Universidad de Extremadura, 06071 Badajoz (Spain); Lozano, J.C. [Laboratorio de Radiactividad Ambiental, Facultad de Ciencias, Universidad de Salamanca, 37008 Salamanca (Spain)

2008-04-15

The elimination of natural uranium and {sup 226}Ra from contaminated waters by rhizofiltration was tested using Helianthus annuus L. (sunflower) seedlings growing in a hydroponic medium. Different experiments were designed to determine the optimum age of the seedlings for the remediation process, and also to study the principal way in which the radionuclides are removed from the solution by the sunflower roots. In every trial a precipitate appeared which contained a major fraction of the natural uranium and {sup 226}Ra. The results indicated that the seedlings themselves induced the formation of this precipitate. When four-week-old seedlings were exposed to contaminated water, a period of only 2 days was sufficient to remove the natural uranium and {sup 226}Ra from the solution: about 50% of the natural uranium and 70% of the {sup 226}Ra were fixed in the roots, and essentially the rest was found in the precipitate, with only very small percentages fixed in the shoots and left in solution.

Refinement of criticality and breeding parameters by means of experiments on a series of critical assemblies

International Nuclear Information System (INIS)

Golubev, V.I.; Dulin, V.A.; Kazanskij, Yu.A.; Mamontov, V.M.; Mozhaev, V.K.; Sidorov, G.I.

1980-01-01

A programme of measurements was performed on a number of critical assemblies with the aim of obtaining reliable experimental data under conditions approximating the simplest calculation model. To this end the neutron balance at the centres of the BFS-31, BFS-33, BFS-35, BFS-38, KBR-3 and KBR-7 critical assemblies was investigated. These assemblies contained central inserts made of uranium dioxide (BFS-33), natural uranium oxide and plutonium metal (BFS-31), natural uranium and plutonium metal (BFS-38), 90% enriched metallic uranium and stainless steel (KBR-3) and enriched uranium dioxide and nickel (KBR-7). The composition of the inserts was such that Ksub(infinite)=1. The K + values, the ratios of the reaction rates of the principal raw material and fissionable isotopes and the reactivity coefficients of a number of materials were measured in the inserts. The components of the breeding coefficient were measured at the centre of the BFS-39 critical assembly which simulates a power reactor (simplest composition with low- and high-enrichment zones and no control mechanism). The authors describe briefly the critical assemblies, the methods of measurement and calculation and methods of correcting for differences between the calculation model and the conditions under which the measurements were performed and compare the results of the experiments with the corresponding theoretical values obtained using various systems of group constants. In their latest versions, the group constants derived from different sets of integral experiments describe the experimental data much better than was previously possible. The deviations which occur in the predicted criticality and breeding parameters using different versions of the constants essentially reflect the difference in the results of the sets of integral experiments that were used for the group constants. (author)

Graphite oxidation and structural strength of graphite support column in VHTR

International Nuclear Information System (INIS)

Park, Byung Ha; No, Hee Cheno; Kim, Eung Soo; Oh, Chang H.

2009-01-01

The air-ingress event by a large pipe break is an important accident considered in design of very high-temperature gas-cooled reactors (VHTR). Core-collapse prediction is a main safety issue. Structural failure model are technically required. The objective of this study is to develop structural failure model for the supporting graphite material in the lower plenum of the GT-MHR (gas-turbine-modular high temperature reactor). Graphite support column is important for VHTR structural integrity. Graphite support columns are under the axial load. Critical strength of graphite column is related to slenderness ratio and bulk density. Through compression tests for fresh and oxidized graphite columns we show that compressive strength of IG-110 was 79.46 MPa. And, the buckling strength of IG-110 column was expressed by the empirical formula: σ 0 =σ straight-line - C L/r, σ straight-line =91.31 MPa, C=1.01. The results of uniform and non-uniform oxidation tests show that the strength degradation of oxidized graphite column is expressed in the following non-dimensional form: σ/σ 0 =exp(-kd), k=0.111. Also, from the results of the uniform oxidation test with a complicated-shape column, we found out that the above non-dimensional equation obtained from the uniform oxidation test is applicable to a uniform oxidation case with a complicated-shape column. (author)

Analysis of beryllium and depleted uranium: An overview of detection methods in aerosols and soils

International Nuclear Information System (INIS)

Camins, I.; Shinn, J.H.

1988-06-01

We conducted a survey of commercially available methods for analysis of beryllium and depleted uranium in aerosols and soils to find a reliable, cost-effective, and sufficiently precise method for researchers involved in environmental testing at the Yuma Proving Ground, Yuma, Arizona. Criteria used for evaluation include cost, method of analysis, specificity, sensitivity, reproducibility, applicability, and commercial availability. We found that atomic absorption spectrometry with graphite furnace meets these criteria for testing samples for beryllium. We found that this method can also be used to test samples for depleted uranium. However, atomic absorption with graphite furnace is not as sensitive a measurement method for depleted uranium as it is for beryllium, so we recommend that quality control of depleted uranium analysis be maintained by testing 10 of every 1000 samples by neutron activation analysis. We also evaluated 45 companies and institutions that provide analyses of beryllium and depleted uranium. 5 refs., 1 tab

Criticality Calculations for a Typical Nuclear Fuel Fabrication Plant with Low Enriched Uranium

International Nuclear Information System (INIS)

Elsayed, Hade; Nagy, Mohamed; Agamy, Said; Shaat, Mohmaed

2013-01-01

The operations with the fissile materials such as U 235 introduce the risk of a criticality accident that may be lethal to nearby personnel and can lead the facility to shutdown. Therefore, the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences. Sixty criticality accidents were occurred in the world. These are accidents divided into two categories, 22 accidents occurred in process facilities and 38 accidents occurred during critical experiments or operations with research reactor. About 21 criticality accidents including Japan Nuclear Fuel Conversion Co. (JCO) accident took place with fuel solution or slurry and only one accident occurred with metal fuel. In this study the nuclear criticality calculations have been performed for a typical nuclear fuel fabrication plant producing nuclear fuel elements for nuclear research reactors with low enriched uranium up to 20%. The calculations were performed for both normal and abnormal operation conditions. The effective multiplication factor (k eff ) during the nuclear fuel fabrication process (Uranium hexafluoride - Ammonium Diuranate conversion process) was determined. Several accident scenarios were postulated and the criticalities of these accidents were evaluated. The computer code MCNP-4B which based on Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations were performed for the cases of, change of moderator to fuel ratio, solution density and concentration of the solute in order to prevent or mitigate criticality accidents during the nuclear fuel fabrication process. The calculation results are analyzed and discussed

Nuclear graphite ageing and turnaround

International Nuclear Information System (INIS)

Marsden, B.J.; Hall, G.N.; Smart, J.

2001-01-01

Graphite moderated reactors are being operated in many countries including, the UK, Russia, Lithuania, Ukraine and Japan. Many of these reactors will operate well into the next century. New designs of High Temperature Graphite Moderated Reactors (HTRS) are being built in China and Japan. The design life of these graphite-moderated reactors is governed by the ageing of the graphite core due to fast neutron damage, and also, in the case of carbon dioxide cooled reactors by the rate of oxidation of the graphite. Nuclear graphites are polycrystalline in nature and it is the irradiation-induced damage to the individual graphite crystals that determines the material property changes with age. The life of a graphite component in a nuclear reactor can be related to the graphite irradiation induced dimensional changes. Graphites typically shrink with age, until a point is reached where the shrinkage stops and the graphite starts to swell. This change from shrinkage to swelling is known as ''turnaround''. It is well known that pre-oxidising graphite specimens caused ''turnaround'' to be delayed, thus extending the life of the graphite, and hence the life of the reactor. However, there was no satisfactory explanation of this behaviour. This paper presents a numerical crystal based model of dimensional change in graphite, which explains the delay in ''turnaround'' in the pre-oxidised specimens irradiated in a fast neutron flux, in terms of crystal accommodation and orientation and change in compliance due to radiolytic oxidation. (author)

Uranium enrichment plans

International Nuclear Information System (INIS)

Gagne, R.W.; Thomas, D.C.

1977-01-01

The status of existing uranium enrichment contracts in the US is reviewed and expected natural uranium requirements for existing domestic uranium enrichment contracts are evaluated. Uncertainty in natural uranium requirements associated with requirements-type and fixed-commitment type contracts is discussed along with implementation of variable tails assay

Enriched uranium recovery at Los Alamos

International Nuclear Information System (INIS)

Herrick, C.C.

1984-01-01

Graphite casting scrap, fuel elements and nongraphite combustibles are calcined to impure oxides. These materials along with zircaloy fuel elements and refractory solids are leach-dissolved separately in HF-HNO 3 acid to solubilize the contained enriched uranium. The resulting slurry is filtered and the clear filtrate (to which mineral acid solutions bearing enriched uranium may be added) are passed through solvent extraction. The solvent extraction product is filtered, precipitated with H 2 O 2 and the precipitate calcined to U 3 O 8 . Metal is made from U 3 O 8 by conversion to UO 2 , hydrofluorination and reduction to metal. Throughput is 150 to 900 kg uranium per year depending on the type of scrap

Uranium Biomineralization By Natural Microbial Phosphatase Activities in the Subsurface

Energy Technology Data Exchange (ETDEWEB)

Taillefert, Martial [Georgia Tech Research Corporation, Atlanta, GA (United States)

2015-04-01

This project investigated the geochemical and microbial processes associated with the biomineralization of radionuclides in subsurface soils. During this study, it was determined that microbial communities from the Oak Ridge Field Research subsurface are able to express phosphatase activities that hydrolyze exogenous organophosphate compounds and result in the non-reductive bioimmobilization of U(VI) phosphate minerals in both aerobic and anaerobic conditions. The changes of the microbial community structure associated with the biomineralization of U(VI) was determined to identify the main organisms involved in the biomineralization process, and the complete genome of two isolates was sequenced. In addition, it was determined that both phytate, the main source of natural organophosphate compounds in natural environments, and polyphosphate accumulated in cells could also be hydrolyzed by native microbial population to liberate enough orthophosphate and precipitate uranium phosphate minerals. Finally, the minerals produced during this process are stable in low pH conditions or environments where the production of dissolved inorganic carbon is moderate. These findings suggest that the biomineralization of U(VI) phosphate minerals is an attractive bioremediation strategy to uranium bioreduction in low pH uranium-contaminated environments. These efforts support the goals of the SBR long-term performance measure by providing key information on "biological processes influencing the form and mobility of DOE contaminants in the subsurface".

A comment on the metallogenic theory of exogenetic uranium ore deposits

International Nuclear Information System (INIS)

Liu Xiaodong; Yu Dagan

2010-01-01

The theory of exogenetic sandstone-type uranium followed the form process of construction in the early time, and discussed the uranium metallization by chemical enrichment during the phase of syn-deposition and diagenesis. Later, the epigenetic theory was put forward by emphasizing hydrodynamic influence on mineralization. The idea of uranium mineralization in open systems is a renovated metallogenic theory for uranium, which confirms the role of exogenesis playing in uranium mineralization. For open systems, this paper underlines that, as the most critical factors for uranium mineralization, both uranium sources and reduce agents should be open to form a dual-open system. Uranium ore deposits in the tectonic zone of eastern China formed in dual-open system, where uranium has been associated with coal, petroleum and natural gas in the sandstone sequence. (authors)

Critical analysis of world uranium resources

Science.gov (United States)

Hall, Susan; Coleman, Margaret

2013-01-01

The U.S. Department of Energy, Energy Information Administration (EIA) joined with the U.S. Department of the Interior, U.S. Geological Survey (USGS) to analyze the world uranium supply and demand balance. To evaluate short-term primary supply (0–15 years), the analysis focused on Reasonably Assured Resources (RAR), which are resources projected with a high degree of geologic assurance and considered to be economically feasible to mine. Such resources include uranium resources from mines currently in production as well as resources that are in the stages of feasibility or of being permitted. Sources of secondary supply for uranium, such as stockpiles and reprocessed fuel, were also examined. To evaluate long-term primary supply, estimates of uranium from unconventional and from undiscovered resources were analyzed. At 2010 rates of consumption, uranium resources identified in operating or developing mines would fuel the world nuclear fleet for about 30 years. However, projections currently predict an increase in uranium requirements tied to expansion of nuclear energy worldwide. Under a low-demand scenario, requirements through the period ending in 2035 are about 2.1 million tU. In the low demand case, uranium identified in existing and developing mines is adequate to supply requirements. However, whether or not these identified resources will be developed rapidly enough to provide an uninterrupted fuel supply to expanded nuclear facilities could not be determined. On the basis of a scenario of high demand through 2035, 2.6 million tU is required and identified resources in operating or developing mines is inadequate. Beyond 2035, when requirements could exceed resources in these developing properties, other sources will need to be developed from less well-assured resources, deposits not yet at the prefeasibility stage, resources that are currently subeconomic, secondary sources, undiscovered conventional resources, and unconventional uranium supplies. This

Recovery of uranium from seawater. 14. System arrangements for the recovery of uranium from seawater by spherical amidoxime chelating resins utilizing natural seawater motions

International Nuclear Information System (INIS)

Egawa, Hiroaki; Kabay, Nalan; Shuto, Taketomi; Jyo, Akinori

1993-01-01

In order to evaluate performances of lightly cross-linked highly porous amidoxime resins in uranium-adsorption systems utilizing natural seawater motions, uranium uptake by the resins from seawater was studied by different approaches, such as simulated sea current exposure tests, towing trials, and/or mooring trials. In general, the efficiency of uranium uptake became higher with a decrease in the thickness of packing layers, indicating important roles of fluidization of the resin particles. On the basis of these fundamental data, mooring tests in the natural sea current were designed and conducted. By mooring flat adsorption beds (base area 260 cm 2 , height 3.0 cm) packed with 780 ml of the resin for 40 h, promising uranium uptake as high as 44 mg/kg of resin (9.9 mg/l of resin) was achieved under sea conditions in which the velocity of sea currents and the vertical velocity of waves were 5.5-49.7 cm/s and 3.4-27 cm/s, respectively

Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

Energy Technology Data Exchange (ETDEWEB)

Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-07-19

The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

International Nuclear Information System (INIS)

Montierth, Leland M.

2016-01-01

The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

Structural analysis of polycrystalline (graphitized) materials

International Nuclear Information System (INIS)

Efremenko, M.M.; Kravchik, A.E.; Osmakov, A.S.

1993-01-01

Specific features of the structure of polycrystal carbon materials (CM), characterized by high enough degree of structural perfection and different genesis are analyzed. From the viewpoint of fine and supercrystallite structure analysis of the most characteristic groups of graphitized CM: artificial graphites, and natural graphites, as well, has been carried out. It is ascertained that in paracrystal CM a monolayer of hexagonally-bound carbon atoms is the basic element of the structure, and in graphitized CM - a microlayer. The importance of the evaluation of the degree of three-dimensional ordering of the microlayer is shown

Calculation of reactivity of control rods in graphite moderated reactors

International Nuclear Information System (INIS)

Nakata, H.

1978-01-01

A study about the method of calculation for the reactivity of control rods in graphite-moderated critical assemblies, is presented. The result of theoretical calculation, developed by super celles and Nordheim-Scalettar methods are compared with experimental results for the critical Assembly of General Atomic. The two methods are then applicable to reactivity calculation of the control rods of graphite moderated critical assemblies [pt

Uranium price reporting systems

International Nuclear Information System (INIS)

1987-09-01

This report describes the systems for uranium price reporting currently available to the uranium industry. The report restricts itself to prices for U 3 O 8 natural uranium concentrates. Most purchases of natural uranium by utilities, and sales by producers, are conducted in this form. The bulk of uranium in electricity generation is enriched before use, and is converted to uranium hexafluoride, UF 6 , prior to enrichment. Some uranium is traded as UF 6 or as enriched uranium, particularly in the 'secondary' market. Prices for UF 6 and enriched uranium are not considered directly in this report. However, where transactions in UF 6 influence the reported price of U 3 O 8 this influence is taken into account. Unless otherwise indicated, the terms uranium and natural uranium used here refer exclusively to U 3 O 8 . (author)

Uranium

International Nuclear Information System (INIS)

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

1992-01-01

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

Production of annular blanks for Mo-99 using natural uranium, LEU uranium, nickel and structural Al-3003 plates

International Nuclear Information System (INIS)

Lisboa, J.R.; Barrera, M.E.; Marin, J.

2010-01-01

The Tc-99m radioisotope for medical use is the one most used in nuclear medicine worldwide. In Chile the Tc-99m is applied in more than 90% of nuclear medicine studies. In order to supply the whole country with this radioisotope, in 2005-2007 the CCHEN developed its own production of Tc-99m generators from Mo-99 imported from Canada, which are prepared with the activity needed by the Chilean hospitals and clinics. As of 2007 Mo-99 was no longer imported, and since then the Tc-99m is produced only by neutron activation of the Mo. The present challenge is to produce Mo-99 by irradiating blanks that contain enriched uranium foils, with locally produced LEU. The annular blank consists of 2 concentric tubes of A1-3003 structural aluminum that, in an interior annular space, contain a LEU foil, covered on both sides by a nickel foil. This work presents the development of the production technology for annular blanks using natural uranium and U-325 enriched uranium. The structural components are made with A1-3003 aluminum alloy, the foils are 13 grams of uranium measuring 100 x 50 mm and 120-150 μ thick. The blank was assembled using a methodology to control, adapt and assemble the blank's different internal components. A foil of natural uranium and LEU uranium, and a nickel foil are included, used as a barrier for the escape of fission products. During the blank's expansion, for analysis alcohol as lubricant was used, allowing the expander to move smoothly through the inside of the blank. The blank was sealed by TIG welding with a pulsed AC current and a mixture of Ar-5% He gases. Two methods were used for the water tightness test; for high escape levels the temperature was used as a promoter of the ΔP provided by hot water and liquid nitrogen, for low escape levels high vacuum technology was used where the ΔP is provided by a high pressure helium atmosphere. The technology for the production of annular LEU blanks was achieved by applying innovations to technologies

Standard model for safety analysis report of hexafluoride power plants from natural uranium

International Nuclear Information System (INIS)

1983-01-01

The standard model for safety analysis report for hexafluoride production power plants from natural uranium is presented, showing the presentation form, the nature and the degree of detail, of the minimal information required by the Brazilian Nuclear Energy Commission - CNEN. (E.G.) [pt

Extrapolated experimental critical parameters of unreflected and steel-reflected massive enriched uranium metal spherical and hemispherical assemblies

International Nuclear Information System (INIS)

Rothe, R.E.

1997-12-01

Sixty-nine critical configurations of up to 186 kg of uranium are reported from very early experiments (1960s) performed at the Rocky Flats Critical Mass Laboratory near Denver, Colorado. Enriched (93%) uranium metal spherical and hemispherical configurations were studied. All were thick-walled shells except for two solid hemispheres. Experiments were essentially unreflected; or they included central and/or external regions of mild steel. No liquids were involved. Critical parameters are derived from extrapolations beyond subcritical data. Extrapolations, rather than more precise interpolations between slightly supercritical and slightly subcritical configurations, were necessary because experiments involved manually assembled configurations. Many extrapolations were quite long; but the general lack of curvature in the subcritical region lends credibility to their validity. In addition to delayed critical parameters, a procedure is offered which might permit the determination of prompt critical parameters as well for the same cases. This conjectured procedure is not based on any strong physical arguments

Case history of natural analogue research on sandstone type uranium occurrences, Japan

International Nuclear Information System (INIS)

Sakamaki, Y.; Kanai, Y.

1991-01-01

Previous fundamental studies on the ore genesis of uranium occurrences chiefly in Cenozoic sandstone formations in Japan, have been re-examined as the case history on natural analogue of radionuclides in high-level radioactive wastes (HLRW). Two principal mode of occurrences have been distinguished among Cenozoic uranium localities in Japan. In the Setouchi (Inland Sea) subregion, hot-spots are found in lacustrine to shallow sea facies of calm environment, corresponding to the first stage of formation of tectonic basins. As observed in Ningyo-toge and Tono area, stratabound ore bodies are generally arranged into paleo-channels. Another type of sporadic uranium indications are found within collapse basins in the 'Green-tuff' subregion, where intense volcanisms and block movements had been taken places throughout Middle miocene age. Well-developed fractures were to be favorable paths for uraniferous groundwater, as well as the suitable site for deposition of uranium. In both cases, the source material of uranium is granitic basement. Under oxidizing environment, uranium anomalies have been occasionally detected in surface- or fracture waters which passing through decomposed granite. In contrast to the behavior of uranium, one of the adequate analogues for mobile nuclides, thorium and REE are relatively immobile even under the same geologic and geochemical circumstances. In ore horizon, where reducing condition has still been kept, geochronological age of tetravalent uranium mineral is in concordance with the age of the host rock. Analysis of structural control shows that the principal factors for uranium concentration are the layout of redox front related to paleo-water tables. 234U/238U disequilibrium method has been proved to be the powerful tool for detecting mobility of uranium in the host rock throughout diagenesis and weathering process. The result of field and laboratory works on this is reported as an example. (author)

Critical and sub-critical experiments on U-BeO lattices; Experiences critiques et sous-critiques sur reseaux U-BeO

Energy Technology Data Exchange (ETDEWEB)

Benoist, P.; Gourdon, Ch.; Martelly, J.; Sagot, M.; Wanner, G. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Deniz, V.; Joshi, B.V.; Sahai, K. [Atomic Energy Establishment Trombay (India)

1958-07-01

Sub-critical experiments have allowed us to measure the material buckling of uranium natural oxide of beryllium lattices with a grid of 15 cm, and made up of uranium bars measuring 2.60 - 2.92 - 3.56 and 4.40 cm of diameter. A critical experiment has then been conducted with hollow 1.35 per cent enriched uranium bars. A study of U-BeO 18.03 cm grid lattices is at present being conducted. (author)Fren. [French] Nous avons mesure par des experiences sous-critiques le laplacien matiere de reseaux uranium naturel-oxyde de beryllium, dont la maille carree a un pas de 15 cm, realises avec des barreaux d'uranium de diametres 2,60 - 2,92 - 3,56 - 4,40 cm. Une experience critique a ete faite ensuite avec des barres creuses d'uranium enrichi a 1,35 pour cent; l'etude des reseaux U-BeO de pas 18,03 cm est actuellement en cours. (auteur)

Critical and sub-critical experiments on U-BeO lattices; Experiences critiques et sous-critiques sur reseaux U-BeO

Energy Technology Data Exchange (ETDEWEB)

Benoist, P; Gourdon, Ch; Martelly, J; Sagot, M; Wanner, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Deniz, V; Joshi, B V; Sahai, K [Atomic Energy Establishment Trombay (India)

1958-07-01

Sub-critical experiments have allowed us to measure the material buckling of uranium natural oxide of beryllium lattices with a grid of 15 cm, and made up of uranium bars measuring 2.60 - 2.92 - 3.56 and 4.40 cm of diameter. A critical experiment has then been conducted with hollow 1.35 per cent enriched uranium bars. A study of U-BeO 18.03 cm grid lattices is at present being conducted. (author)Fren. [French] Nous avons mesure par des experiences sous-critiques le laplacien matiere de reseaux uranium naturel-oxyde de beryllium, dont la maille carree a un pas de 15 cm, realises avec des barreaux d'uranium de diametres 2,60 - 2,92 - 3,56 - 4,40 cm. Une experience critique a ete faite ensuite avec des barres creuses d'uranium enrichi a 1,35 pour cent; l'etude des reseaux U-BeO de pas 18,03 cm est actuellement en cours. (auteur)

Criticality evaluations with moderators other than water for uranium metal fuels

International Nuclear Information System (INIS)

Toffer, H.; Tollefson, D.A.; Finfrock, S.H.

1986-01-01

Occasionally, nuclear criticality safety analyses of fissile material handling operations or transport situations require consideration of moderation other than water. Such moderators could be oils, plastics, wood, concrete, carbon, or even wet sand. All of these materials contain either hydrogen, carbon, or mixtures of the two elements as the principal moderators. Other elements as part of the compounds or mixtures contribute less to the neutron slowing down process and can possibly be significant parasitic neutron absorbers. Results of a series of calculations are presented illustrating the impact of various moderators on critical masses or critical parameters as a function of lattice pitch for different uranium metal fuel elements at low 235 U enrichments. Several nuclear criticality safety analyses performed at the Hanford N Reactor, operated by UNC Nuclear Industries for the US Department of Energy, have considered alternative moderators to assure that water moderation represented the most limiting case

Natural uranium equivalent fuel an innovative design for proven CANDU technology

Energy Technology Data Exchange (ETDEWEB)

Pineiro, F.; Ho, K.; Khaial, A.; Boubcher, M.; Cottrell, C.; Kuran, S., E-mail: fabricia.pineiro@candu.com [Candu Energy Inc., Mississauga, ON (Canada); Zhenhua, Z.; Zhiliang, M. [Third Qinshan Nuclear Power Company, Haiyan, Zhejiang (China)

2015-07-01

The high neutron economy, on-power refuelling capability and fuel bundle design simplicity in CANDU reactors allow for the efficient utilization of alternative fuels. Candu Energy Inc. (Candu), in collaboration with the Third Qinshan Nuclear Power Company (TQNPC), the China North Nuclear Fuel Corporation (CNNFC), and the Nuclear Power Institute of China (NPIC), has successfully developed an advanced fuel called Natural Uranium Equivalent (NUE). This innovative design consists of a mixture of recycled and depleted uranium, which can be implemented in existing CANDU stations thereby bringing waste products back into the energy stream, increasing fuel resources diversity and reducing fuel costs. (author)

Natural uranium equivalent fuel. An innovative design for proven CANDU technology

Energy Technology Data Exchange (ETDEWEB)

Pineiro, F.; Ho, K.; Khaial, A.; Boubcher, M.; Cottrell, C.; Kuran, S. [Candu Energy Inc., Mississauga, Ontario (Canada); Zhenhua, Z.; Zhiliang, M. [Third Qinshan Nuclear Power Co., Haiyan, Zhejiang (China)

2015-09-15

The high neutron economy, on-power refuelling capability and fuel bundle design simplicity in CANDU® reactors allow for the efficient utilization of alternative fuels. Candu Energy Inc. (Candu), in collaboration with the Third Qinshan Nuclear Power Company (TQNPC), the China North Nuclear Fuel Corporation (CNNFC), and the Nuclear Power Institute of China (NPIC), has successfully developed an advanced fuel called Natural Uranium Equivalent (NUE). This innovative design consists of a mixture of recycled and depleted uranium, which can be implemented in existing CANDU stations thereby bringing waste products back into the energy stream, increasing fuel resources diversity and reducing fuel costs. (author)

Draft critical mineral list—Summary of methodology and background information—U.S. Geological Survey technical input document in response to Secretarial Order No. 3359

Science.gov (United States)

Fortier, Steven M.; Nassar, Nedal T.; Lederer, Graham W.; Brainard, Jamie; Gambogi, Joseph; McCullough, Erin A.

2018-02-16

Pursuant to the Presidential Executive Order (EO) No. 13817, “A Federal Strategy to Ensure Secure and Reliable Supplies of Critical Minerals,” the Secretary of the Interior, in coordination with the Secretary of Defense, and in consultation with the heads of other relevant executive departments and agencies, was tasked with developing and submitting a draft list of minerals defined as “critical minerals” to the Federal Register within 60 days of the issue of the EO (December 20, 2017).Based on an analysis by the U.S. Geological Survey and other U.S. Government agencies, using multiple criteria, 35 minerals or mineral material groups have been identified that are currently (February 2018) considered critical. These include the following: aluminum (bauxite), antimony, arsenic, barite, beryllium, bismuth, cesium, chromium, cobalt, fluorspar, gallium, germanium, graphite (natural), hafnium, helium, indium, lithium, magnesium, manganese, niobium, platinum group metals, potash, rare earth elements group, rhenium, rubidium, scandium, strontium, tantalum, tellurium, tin, titanium, tungsten, uranium, vanadium, and zirconium. The categorization of minerals as critical may change during the course of the review process and is thus provisional.

Study by electronic microscopy of corrosion features of graphite after hot oxidation (air, 620 C)

International Nuclear Information System (INIS)

Jodon de Villeroche, Suzanne

1968-01-01

The author reports the study of corrosion features of graphite after hot oxidation in the air at 620 C. It is based on observations made by electronic microscopy. This study comes after another one dedicated to oxidation features obtained by hot corrosion of natural graphite, and aims at comparing pyrolytic graphite before and after irradiation in an atomic pile, and at performing tests on a graphite processed with ozone. After a recall of generalities about natural graphite and of some issues related to hot corrosion of natural graphite, the author presents some characteristics and features of irradiated and non-irradiated pyrolytic graphite. He reports the study of the oxidation of samples of pyrolytic graphite: production of thin lamellae, production of glaze-carbon replicates, oxidation of irradiated and of non-irradiated graphite, healing of irradiation defects, and oxidation of ozone-processed natural graphite [fr

Reactivity of hydrogen with uranium in the presence of Pt

International Nuclear Information System (INIS)

Balooch, M.; Siekhaus, W.J.

1997-07-01

The surface-reaction of di-hydrogen with uranium in the presence of Pt clusters has been studied using scanning tunneling microscopy (STM). Uranium was deposited on highly oriented pyrolytic graphite (HOPG) and annealed at temperatures up to 1200 degrees C to obtain atomically pyrolytic flat surfaces. Pt clusters were then formed using evaporation from a Pt source onto the surface and subsequent annealing. Hydrogen mainly attacked uranium in the vicinity of Pt clusters and formed hydride. The hydride formation probability is almost constant at 2.3x10 -4 over the range of exposures studied

Criticality Analysis of the U-H2O Subcritical Assembly Modified for Rand D of the High Temperature Reactor

International Nuclear Information System (INIS)

Syarip; Tri-Wulan-Tjiptono; Tegas-Sutondo

2000-01-01

A criticality analysis of the natural uranium - light water sub-criticalassembly available at the P3TM-BATAN Yogyakarta, converted into a naturaluranium - graphite system has been performed. The purpose of this study is toprovide the research facility on the basic static and kinetics studies forthe high temperature reactor (HTR) in which the HTR fuel system is underdevelopment at the P3TM. For the purpose of this study, a neutroniccalculation was performed using WIMSD/4 code, to determine the neutronmultiplication factor for various fuel configurations of the sub-criticalassemblies. The results show that the effective neutron multiplication factor(k ef ) for U-Be-H 2 O and U-Be-He systems are 1.0474 and 1.4666 respectively,while for the graphite moderated systems with coolants of H 2 O or He(U-C-H 2 O and U-C-He) systems, the corresponding k ef are 0.787 and 0.4211respectively. The results conclude that the modification of U-H 2 O toU-C-H 2 O system, in accordance with neutronic is quite feasible, safe, cheapand practical, and in addition, the treatment of H 2 O is relatively easy.(author)

Exposure pathways and health effects associated with chemical and radiological toxicity of natural uranium: a review.

Science.gov (United States)

Brugge, Doug; de Lemos, Jamie L; Oldmixon, Beth

2005-01-01

Natural uranium exposure derives from the mining, milling, and processing of uranium ore, as well as from ingestion of groundwater that is naturally contaminated with uranium. Ingestion and inhalation are the primary routes of entry into the body. Absorption of uranium from the lungs or digestive track is typically low but can vary depending on compound specific solubility. From the blood, two-thirds of the uranium is excreted in urine over the first 24 hours and up to 80% to 90% of uranium deposited in the bone leaves the body within 1.5 years. The primary health outcomes of concern documented with respect to uranium are renal, developmental, reproductive, diminished bone growth, and DNA damage. The reported health effects derive from experimental animal studies and human epidemiology. The Lowest Observed Adverse Effect Level (LOAEL) derived from animal studies is 50 microg/m3 for inhalation and 60 ug/kg body weight/day for ingestion. The current respiratory standard of the Occupational Safety and Health Administration (OSHA), 50 microg/m3, affords no margin of safety. Considering the safety factors for species and individual variation, the ingestion LOAEL corresponds to the daily consumption set by the World Health Organization Drinking Water Standard at 2 microg/L. Based on economic considerations, the United States Environmental Protection Agency maximum contaminant level is 30 microg/L. Further research is needed, with particular attention on the impact of uranium on indigenous populations, on routes of exposure in communities near uranium sites, on the combined exposures present at many uranium sites, on human developmental defects, and on health effects at or below established exposure standards.

Uranium, depleted uranium, biological effects; Uranium, uranium appauvri, effets biologiques

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

Physicists, chemists and biologists at the CEA are developing scientific programs on the properties and uses of ionizing radiation. Since the CEA was created in 1945, a great deal of research has been carried out on the properties of natural, enriched and depleted uranium in cooperation with university laboratories and CNRS. There is a great deal of available data about uranium; thousands of analyses have been published in international reviews over more than 40 years. This presentation on uranium is a very brief summary of all these studies. (author)

Cycle for fuel elements. Uranium production, programs for nuclear power stations and capital expenditure involved; Cycles de combustibles. Production d'uranium, programme de centrales electriques et effort financier correspondant

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

A number of different possible programs for nuclear power stations of various types are presented in this survey. These programs are established in relation to the use of uranium and thorium in amounts similar to those that shall probably be produced in France during the next fifteen years. As it is possible to draw plans for nuclear power stations in which several processes exist simultaneously, an unlimited number of variations being thinkable, this survey is limited to successive analysis of the results obtained by use of only one of each of the following three systems: - system natural uranium-graphite, - system natural uranium-heavy water, -system enriched uranium-pressurised light water. All schemes are considered as assemblages of these three simple systems. The effects of plutonium recycling are also considered for each system. The electric power installed and the capacity of stations situated up-stream and down-stream have been calculated by this method and an attempt has been made to establish the sum to be invested during the fifteen years necessary for the launching of the programs scheduled. A table of timing for the investments groups the results obtained. Considering the fact that French availabilities in capital shall not be unlimited during the coming years, this way of presenting the results seems to be interesting. (author)Fren. [French] L'etude presentee comporte l'examen d'un certain nombre d'hypotheses de programmes de centrales nucleaires de types differents. Ces programmes correspondent a l'utilisation de tonnages d'uranium et de thorium de l'ordre de grandeur de ceux qui seront probablement produits par la France dans les quinze prochaines annees. Comme il est possible de batir un programme de centrales nucleaires, comportant a la fois plusieurs filieres suivant des variantes en nombre infini, on s'est contente d'examiner successivement les resultats ous si on utilisait exclusivement l

Exfoliation approach for preparing high conductive reduced graphite oxide and its application in natural rubber composites

Energy Technology Data Exchange (ETDEWEB)

Wipatkrut, Pattharaporn [Department of Chemical Technology, Faculty of Science, Chulalongkorn University, Bangkok 10330 (Thailand); Poompradub, Sirilux, E-mail: sirilux.p@chula.ac.th [Department of Chemical Technology, Faculty of Science, Chulalongkorn University, Bangkok 10330 (Thailand); Center for Petroleum, Petrochemical and Advanced Material, Chulalongkorn University, Bangkok 10330 (Thailand)

2017-04-15

Highlights: • Graphite waste was exfoliated by oxidation and chemical and thermal reduction. • The obtained graphene-T was a single layer sheet with a high electrical conductivity. • Graphene-T incorporation at 5 phr improved the electrical conductivity of NR. • Graphene-T incorporation at 5–25 phr improved the mechanical properties of NR. - Abstract: High conductivity reduced graphite oxide (RGO) was prepared by exfoliation of graphite waste from the metal smelting industry. To improve the surface properties of the RGO, the graphite oxide obtained based on Hummers’ method was reduced by L-ascorbic acid to give RGOV, which was then subjected to thermal reduction to obtain RGOT. The residual oxygen-containing groups in RGOV were almost completely removed by the thermal reduction and the conjugated graphene networks were restored in RGOT. The effect of the RGOT content in natural rubber (NR) on the cure, electrical and mechanical properties of the NR-RGOT (NG) composites was evaluated. The electrical conductivity of NR was increased by the inclusion of RGOT at a percolation threshold of 5 phr, with an electrical conductivity of 8.71 × 10{sup −6} S/m. The mechanical properties, i.e., the modulus, tensile strength and hardness, of NG were comparable with those of conductive carbon black filled NR ones.

Cycle for fuel elements. Uranium production, programs for nuclear power stations and capital expenditure involved

International Nuclear Information System (INIS)

Andriot, J.; Gaussens, J.

1958-01-01

A number of different possible programs for nuclear power stations of various types are presented in this survey. These programs are established in relation to the use of uranium and thorium in amounts similar to those that shall probably be produced in France during the next fifteen years. As it is possible to draw plans for nuclear power stations in which several processes exist simultaneously, an unlimited number of variations being thinkable, this survey is limited to successive analysis of the results obtained by use of only one of each of the following three systems: - system natural uranium-graphite, - system natural uranium-heavy water, -system enriched uranium-pressurised light water. All schemes are considered as assemblages of these three simple systems. The effects of plutonium recycling are also considered for each system. The electric power installed and the capacity of stations situated up-stream and down-stream have been calculated by this method and an attempt has been made to establish the sum to be invested during the fifteen years necessary for the launching of the programs scheduled. A table of timing for the investments groups the results obtained. Considering the fact that French availabilities in capital shall not be unlimited during the coming years, this way of presenting the results seems to be interesting. (author) [fr

Market for natural uranium conversion. Commercial aspect

International Nuclear Information System (INIS)

Durret, L.F.

1986-01-01

The main activity of COMURHEX is the conversion into uranium hexafluoride of uranium concentrates from mines and owned by electricity producers. Capacities of the 5 uranium converters in the Western World are compared. About 50% of COMUREX turnover is exported. Evolution of the market and of stockpile are reviewed [fr

On the Deposition Equilibrium of Carbon Nanotubes or Graphite in the Reforming Processes of Lower Hydrocarbon Fuels

Directory of Open Access Journals (Sweden)

Zdzisław Jaworski

2017-11-01

Full Text Available The modeling of carbon deposition from C-H-O reformates has usually employed thermodynamic data for graphite, but has rarely employed such data for impure filamentous carbon. Therefore, electrochemical data for the literature on the chemical potential of two types of purified carbon nanotubes (CNTs are included in the study. Parameter values determining the thermodynamic equilibrium of the deposition of either graphite or CNTs are computed for dry and wet reformates from natural gas and liquefied petroleum gas. The calculation results are presented as the atomic oxygen-to-carbon ratio (O/C against temperature (200 to 100 °C for various pressures (1 to 30 bar. Areas of O/C for either carbon deposition or deposition-free are computed, and indicate the critical O/C values below which the deposition can occur. Only three types of deposited carbon were found in the studied equilibrium conditions: Graphite, multi-walled CNTs, and single-walled CNTs in bundles. The temperature regions of the appearance of the thermodynamically stable forms of solid carbon are numerically determined as being independent of pressure and the analyzed reactants. The modeling indicates a significant increase in the critical O/C for the deposition of CNTs against that for graphite. The highest rise in the critical O/C, of up to 290% at 30 bar, was found for the wet reforming process.

CRITICALITY ANALYSIS OF URANIUM STORAGE FACILITY WITH FORMATION RACKS

Directory of Open Access Journals (Sweden)

Sri Kuntjoro

2017-03-01

ANALISIS KRITIKALITAS DI FASILITAS PENYIMPANAN BAHAN URANIUM DENGAN FORMASI PENGATURAN RAK. Bahan uranium dibutuhkan untuk produksi bahan bakar reaktor penelitian dan radioisotop. Bahan uranium sebelum digunakan terlebih dahulu disimpan pada fasilitas penyimpanan. Salah satu prasyarat fasilitas penyimpanan bahan uranium adalah fasilitas tersebut harus dalam kondisi sub-kritis. Bila kondisi kritis terjadi mengakibatkan proses fissi pada bahan uranium tidak terkendali, sehingga akan menimbulkan suhu yang sangat tinggi. Tujuan dari penelitian ini adalah untuk menganalisa kondisi kritikalitas dari fasilitas penyimpanan bahan uranium yang berada di PT. INUKI (Persero untuk menjamin fasilitas tersebut dalam kondisi sub-kritis. Analisis kritikalitas dilakukan menggunakan program MCNP-5 untuk mengetahui tingkat kritikalitas dari tiga fasilitas penyimpanan bahan uranium untuk kondisi awal dan kondisi setelah ditambahkan rak penyimpanan. Untuk fasilitas penyimpanan 1 dan 2 dibuat tiga skenario pengaturan container pada rak penyimpanan, sedangkan pada fasilitas penyimpanan 3 dilakukan 1 skenario.  Hasil ini menunjukkan seluruh fasilitas penyimpanan pada kondisi awal dan setelah ditambah rak penyimpanan dalam kondisi sub-kritis (k-eff<1. Hasil tersebut selanjutnya dipergunakan sebagai dasar untuk menyusun manejemen pengelolaan bahan uranium. Selain itu juga digunakan sebagai dasar untuk pembuatan ijin dari badan pengawas (BAPETEN. Kata Kunci : kritikalitas, fasilitas penyimpanan berbahan uranium,  k-eff

Exposure of critical group of population to water radionuclides in area affected by uranium ore mining

Energy Technology Data Exchange (ETDEWEB)

Hladka, E; Zavadsky, M; Solnicka, H; Heroldova, J

1985-08-01

Waste waters from the uranium industry are decontaminated and then discharged into water courses. Inhabitants of the nearest village on the river form the critical group with regard to radiation burden. The critical radionuclides are Usub(nat), Ra 226, Pb 210 and Po 210 whose concentrations were determined in drinking water, in the water course and in plants watered with water from the river. From obtained data on the consumption of foods of own production and of water for drinking and cooking, a weighted sum was made of the intake of critical radionuclides per year on the conservative assumption that ingestion is the sole form of intake (permissible ingestion under Notice 59/72, Coll. of Laws). Under the said criteria the intake of radionuclides from water and foods of own production is for the critical population group 27 times less than the permissible intake for the population. Decontaminated waste waters from the operation of uranium industries contribute to the radiation burden of the population only negligibly. Radionuclides from the investigated sources represent a minute fraction of permissible intake.

Recycling of reprocessed uranium

International Nuclear Information System (INIS)

Randl, R.P.

1987-01-01

Since nuclear power was first exploited in the Federal Republic of Germany, the philosophy underlying the strategy of the nuclear fuel cycle has been to make optimum use of the resource potential of recovered uranium and plutonium within a closed fuel cycle. Apart from the weighty argument of reprocessing being an important step in the treatment and disposal of radioactive wastes, permitting their optimum ecological conditioning after the reprocessing step and subsequent storage underground, another argument that, no doubt, carried weight was the possibility of reducing the demand of power plants for natural uranium. In recent years, strategies of recycling have emerged for reprocessed uranium. If that energy potential, too, is to be exploited by thermal recycling, it is appropriate to choose a slightly different method of recycling from the one for plutonium. While the first generation of reprocessed uranium fuel recycled in the reactor cuts down natural uranium requirement by some 15%, the recycling of a second generation of reprocessed, once more enriched uranium fuel helps only to save a further three per cent of natural uranium. Uranium of the second generation already carries uranium-232 isotope, causing production disturbances, and uranium-236 isotope, causing disturbances of the neutron balance in the reactor, in such amounts as to make further fabrication of uranium fuel elements inexpedient, even after mixing with natural uranium feed. (orig./UA) [de

Starting up a programme of atomic piles using compressed gas

International Nuclear Information System (INIS)

Horowitz, J.; Yvon, J.

1959-01-01

1) An examination of the intellectual and material resources which have directed the French programme towards: a) the natural uranium and plutonium system, b) the use of compressed gas as heat transfer fluid (primary fluid). 2) The parts played in exploring the field by the pile EL2 and G1, EL2 a natural uranium, heavy water and compressed gas pile, G1 a natural uranium, graphite and atmospheric air pile. 3) Development of the neutronics of graphite piles: physical study of G1. 4) The examination of certain problem posed by centres equipped with natural uranium, graphite and compressed carbon dioxide piles: structure, special materials, fluid circuits, maximum efficiency. Economic aspects. 5) Aids to progress: a) piles for testing materials and for tests on canned fuel elements, b) laboratory and calculation facilities. 6) Possible new orientations of compressed gas piles: a) raising of the pressure, b) enriched fuel, c) higher temperatures, d) use of heavy water. (author) [fr

Determination of natural and depleted uranium in urine at the ppt level: an interlaboratory analytical exercise

International Nuclear Information System (INIS)

D'Agostino, P.A.; Ough, E.A.; Glover, S.E.; Vallerand, A.L.

2002-10-01

An analytical exercise was initiated in order to determine those analytical procedures with the capacity to measure uranium isotope ratios ( 238 U/ 235 U) in urine samples containing less that 1μ uranium /L urine. A host laboratory was tasked with the preparation of six sets (12 samples per set) of synthetic urine samples spiked with varying amounts of natural and depleted (0.2% 235 U) uranium. The sets of samples contained total uranium in the range 25 ng U/L urine to 770 ng U/L urine, with isotope ratios ( 238 U/ 235 U) from 137.9 (natural uranium) to 215 (∼50% depleted uranium). Sets of samples were shipped to five testing laboratories (four Canadian and one European) for total and isotopic assay. The techniques employed in the analyses included sector field inductively coupled plasma mass spectrometry (ICP-SF-MS), quadrupole inductively coupled plasma mass spectrometry (ICP-Q-MS), thermal ionization mass spectrometry (TIMS) and neutron activation analysis (NAA). Full results were obtained from three testing labs (ICP-SF-MS, ICP-Q-MS and TIMS). Their results, plus partial results from the NAA lab, have been included in this report. Total uranium and isotope ratio results obtained from ICP-SF-MS and ICP-Q-MS were in good agreement with the host lab values. Neutron activation analysis and TIMS reported total uranium concentrations that differed from the host lab. An incomplete set of isotopic ratios was obtained from the NAA lab with some results reporting enriched uranium (% 235 U > 0.7). Based on the reported results, the four analytical procedures were ranked: ICP-SF-MS (1), ICP-Q-MS (2), TIMS (3) and NAA (4). (author)

Analysis and exploitation of bacterial population from natural uranium-rich soils: selection of a model specie

International Nuclear Information System (INIS)

Mondani, L.

2010-01-01

It is well known that soils play a key role in controlling the mobility of toxic metals and this property is greatly influenced by indigenous bacterial communities. This study has been conducted on radioactive and controls soils, collected in natural uraniferous areas (Limousin). A physico-chemical and mineralogical analysis of soils samples was carried out.The structure of bacterial communities was estimated by Denaturing Gradient Gel Electrophoresis (DGGE). The community structure is remarkably more stable in the uranium-rich soils than in the control ones, indicating that uranium exerts a high selection from the soils was constructed and screened for uranium resistance in order to study bacteria-uranium interactions. Scanning electron microscopy revealed that a phylo-genetically diverse set of uranium-resistant species ware able to chelate uranium at the cell surface. (author) [fr

Collective modes in superconducting rhombohedral graphite

Energy Technology Data Exchange (ETDEWEB)

Kauppila, Ville [O.V. Lounasmaa Laboratory, Aalto University (Finland); Hyart, Timo; Heikkilae, Tero [University of Jyvaeskylae (Finland)

2015-07-01

Recently it was realized that coupling particles with a Dirac dispersion (such as electrons in graphene) can lead to a topologically protected state with flat band dispersion. Such a state could support superconductivity with unusually high critical temperatures. Perhaps the most promising way to realize such coupling in real materials is in the surface of rhombohedrally stacked graphite. We consider collective excitations (i.e. the Higgs modes) in surface superconducting rhombohedral graphite. We find two amplitude and two phase modes corresponding to the two surfaces of the graphite where the superconductivity lives. We calculate the dispersion of these modes. We also derive the Ginzburg-Landau theory for this material. We show that in superconducting rhombohedral graphite, the collective modes, unlike in conventional BCS superconductors, give a large contribution to thermodynamic properties of the material.

Critical management issues for the Uranium Mill Tailings Remedial Action (UMTRA) Project

International Nuclear Information System (INIS)

Themelis, J.G.; Krishnan, K.R.

1985-01-01

The Uranium Mill Tailings Radiation Control Act of 1978 (PL95-604) authorized the Secretary of Energy to enter into cooperative agreements with certain states and Indian Tribes to clean up 24 inactive uranium mill tailing sites and associated vicinity properties. The Uranium Mill Tailings Remedial Action (UMTRA) Project includes the three Federal agencies (EPA, DOE, and NRC), eleven state, Indian Tribes, and at least four major contractors. The UMTRA Project extends over a period of ten years. The standards for the Project require a design life of 1000 years with a minimum performance period of 200 years. This paper discusses the critical management issues in dealing with the UMTRA Project and identifies the development of solutions for many of those issues. The highlights to date are promulgation of EPA standards, continued support from Congress and participating states and Indian Tribes, significant leadership shown at all levels, establishment of credibility with the public, and continued motivation of the team. The challenge for tomorrow is making certain NRC will license the sites and maintaining the high level of coordination exhibited to date to assure Project completion on schedule

Uranium pollution in an estuary affected by pyrite acid mine drainage and releases of naturally occurring radioactive materials

International Nuclear Information System (INIS)

Villa, M.; Manjon, G.; Hurtado, S.; Garcia-Tenorio, R.

2011-01-01

Highlights: → Huelva estuary is affected by former phosphogypsum releases and pyrite acid mine drainage. → Time evolution of uranium concentration is analyzed after halting of NORM releases. → Two new contamination sources are preventing the complete uranium cleaning: (1) The leaching of phosphogypsum stacks located close to Tinto River. (2) Pyrite acid mine drainage. → High uranium concentrations are dissolved in water and precipitate subsequently. - Abstract: After the termination of phosphogypsum discharges to the Huelva estuary (SW Spain), a unique opportunity was presented to study the response of a contaminated environmental compartment after the cessation of its main source of pollution. The evolution over time of uranium concentrations in the estuary is presented to supply new insights into the decontamination of a scenario affected by Naturally Occurring Radioactive Material (NORM) discharges. The cleaning of uranium isotopes from the area has not taken place as rapidly as expected due to leaching from phosphogypsum stacks. An in-depth study using various techniques of analysis, including 234 U/ 238 U and 230 Th/ 232 Th ratios and the decreasing rates of the uranium concentration, enabled a second source of uranium contamination to be discovered. Increased uranium levels due to acid mine drainage from pyrite mines located in the Iberian Pyrite Belt (SW Spain) prevent complete uranium decontamination and, therefore, result in levels nearly twice those of natural background levels.

Distribution of natural uranium in groundwater around Kudankulam

International Nuclear Information System (INIS)

Selvi, B.S.; Vijayakumar, B.; Rana, B.K.; Ravi, P.M.

2016-01-01

A systematic study was carried out to estimate the uranium concentration in the ground water around Kudankulam in Southern Tamil Nadu. The uranium concentration in ground water varies from 0.2 to 6.6 μg/l, with a mean value of 2.0 μg/l. The Quantalase uranium analyzer was used to measure the uranium concentration. These groundwater samples were analyzed for the water quality parameters such as pH, conductance, total dissolved solids (TDS), salinity, chloride, and sulfate. An attempt has been made to correlate the uranium concentration with the water quality parameters. It is observed that conductance, TDS, salinity, chloride, and sulfate show positive correlation with uranium concentration. (author)

Natural uranium utilization without enrichment and reprocessing

Energy Technology Data Exchange (ETDEWEB)

Sekimoto, H.; Toshinsky, V.; Ryu, K. [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors

2001-07-01

Two types of fast reactor are investigated to utilize the natural uranium without enrichment and reprocessing in an equilibrium state. The first trial is SFPR. Its fuel-shuffling pattern is optimized. An obtained result gives its peak fuel burnup of 22,5%, power peaking factor of 1.5 and peak excess reactivity of 2,15%. The second trial is CANDLE burnup scheme, where distribution shapes of neutron flux and nuclide densities are constant but move in axial direction with a constant velocity. A feasible solution gives the speed of burning region of 4,1 cm/year, k{sub eff} of 1,02 and average spent fuel burnup of 41%. (author)

Starting up a programme of atomic piles using compressed gas; Le demarrage d'un programme de piles atomiques a gaz comprime

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

1) An examination of the intellectual and material resources which have directed the French programme towards: a) the natural uranium and plutonium system, b) the use of compressed gas as heat transfer fluid (primary fluid). 2) The parts played in exploring the field by the pile EL2 and G1, EL2 a natural uranium, heavy water and compressed gas pile, G1 a natural uranium, graphite and atmospheric air pile. 3) Development of the neutronics of graphite piles: physical study of G1. 4) The examination of certain problem posed by centres equipped with natural uranium, graphite and compressed carbon dioxide piles: structure, special materials, fluid circuits, maximum efficiency. Economic aspects. 5) Aids to progress: a) piles for testing materials and for tests on canned fuel elements, b) laboratory and calculation facilities. 6) Possible new orientations of compressed gas piles: a) raising of the pressure, b) enriched fuel, c) higher temperatures, d) use of heavy water. (author) [French] 1) Examen des ressources - intellectuelles et materielles - qui ont oriente le programme fran is vers: a) la voie de l'uranium naturel et du plutonium; b) l'emploi comme fluide pour le transfert de la chaleur (fluide primaire) d'un gaz comprime. 2) Le role d'exploration des piles EL2 et G1, EL2 pile a uranium naturel, eau lourde et gaz comprime, G1 pile a uranium naturel, graphite et air atmospherique. 3) Developpement de la neutronique des piles a graphite: l'etude physique de G1. 4) Examen de certains problemes poses par les centrales equipees de piles a uranium naturel, graphite et gaz carbonique comprime: structure, materiaux speciaux, circuits de fluides, optimisation. Aspects economiques. 5) Les auxiliaires du progres: a) piles pour essai de materiaux et pour essais de cartouches, b) moyens de laboratoire et moyens de calcul. 6) Orientations nouvelles possibles des piles a gaz comprime: a) elevation de la pression, b) combustible enrichi, c) temperatures elevees, d) emploi de l

Carbon-14 in neutron-irradiated graphite for graphite-moderated reactors. Joint research

Energy Technology Data Exchange (ETDEWEB)

Fujii, Kimio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Matsuo, Hideto [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokyo (Japan)

2002-12-01

The graphite moderated gas cooled reactor operated by the Japan Atomic Power Company was stopped its commercial operation on March 1998, and the decommissioning process has been started. Graphite material is often used as the moderator and the reflector materials in the core of the gas cooled reactor. During the operation, a long life nuclide of {sup 14}C is generated in the graphite by several transmutation reactions. Separation of {sup 14}C isotope and the development of the separation method have been recognized to be critical issues for the decommissioning of the reactor core. To understand the current methodologies for the carbon isotope separation, literature on the subject was surveyed. Also, those on the physical and chemical behavior of {sup 14}C were surveyed. This is because the larger part of the nuclides in the graphite is produced from {sup 14}N by (n,p) reaction, and the location of them in the material tends to be different from those of the other carbon atoms. This report summarizes the result of survey on the open literature about the behavior of {sup 14}C and the separation methods, including the list of the literature on these subjects. (author)

A graphite foam reinforced by graphite particles

Energy Technology Data Exchange (ETDEWEB)

Zhu, J.J.; Wang, X.Y.; Guo, L.F.; Wang, Y.M.; Wang, Y.P.; Yu, M.F.; Lau, K.T.T. [DongHua University, Shanghai (China). College of Material Science and Engineering

2007-11-15

Graphite foam was obtained after carbonization and graphitization of a pitch foam formed by the pyrolysis of coal tar based mesophase pitch mixed with graphite particles in a high pressure and temperature chamber. The graphite foam possessed high mechanical strength and exceptional thermal conductivity after adding the graphite particles. Experimental results showed that the thermal conductivity of modified graphite foam reached 110W/m K, and its compressive strength increased from 3.7 MPa to 12.5 MPa with the addition of 5 wt% graphite particles. Through the microscopic observation, it was also found that fewer micro-cracks were formed in the cell wall of the modified foam as compared with pure graphite foam. The graphitization degree of modified foam reached 84.9% and the ligament of graphite foam exhibited high alignment after carbonization at 1200{sup o}C for 3 h and graphitization at 3000{sup o}C for 10 min.

National Uranium Resource Evaluation: Spartanburg Quadrangle, South Carolina and North Carolina

International Nuclear Information System (INIS)

Schot, E.H.; Galipeau, J.M.

1980-11-01

The Spartanburg Quadrangle, South Carolina and North Carolina, was evaluated for uranium favorability using National Uranium Resource Evaluation criteria. The evaluation included the study and analysis of published and collected geologic, geophysical, and geochemical data from subsurface, surface, and aerial studies. Five environments are favorable for uranium deposits. The Triassic Wadesboro Basin has ground waters with anomalously high uranium concentrations and uranium-to-conductivity ratios. The Upper Cretaceous Tuscaloosa-Middendorf Formation is near a uranium source and has sediments favorable for uranium deposition. The contact-metamorphic aureoles associated with the Liberty Hill-Kershaw and Winnsboro-Rion plutonic complexes are close to uranium sources and contain the reductants (sulfides, graphite) necessary for precipitation. The East Fork area in the Charlotte Belt has ground waters with uranium concentrations 4 to 132 times the mean concentration reported for the surrounding Piedmont area. Unfavorable environments include the Catawba Granite, the area west of the Winnsboro-Rion complex, gold-quartz veins, the vermiculite district, and the Western Monazite Belt

Deuterium migration in nuclear graphite: consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste

International Nuclear Information System (INIS)

Le-Guillou, Mael

2014-01-01

In France, 23 000 t of irradiated graphite that will be generated by the decommissioning of the first generation Uranium Naturel-Graphite-Gaz (UNGG) nuclear reactors are waiting for a long term management solution. This work focuses on the behavior of tritium, which is one of the main contributors to the radiological inventory of graphite waste after reactor shutdown. In order to anticipate tritium release during dismantling or waste management, it is mandatory to collect data on its migration, location and inventory. Our study is based on the simulation of tritium by implantation of approximately 3 at. % of deuterium up to around 3 μm in a virgin nuclear graphite. This material was then annealed up to 300 h and 1300 C in inert atmosphere, UNGG coolant gas and humid gas, aiming to reproduce thermal conditions close to those encountered in reactor and during waste management operations. The deuterium profiles and spatial distribution were analyzed using the nuclear reaction 2 H( 3 He,p) 4 He. The main results evidence a thermal release of implanted deuterium occurring essentially through three regimes controlled by the detrapping of atomic deuterium located in superficial or interstitial sites. The extrapolation of our data to tritium suggests that its purely thermal release during reactor operations may have been lower than 30 % and would be located close to the graphite free surfaces. Consequently, most of the tritium inventory after reactor shutdown could be trapped deeply within the irradiated graphite structure. Decontamination of graphite waste should then require temperatures higher than 1300 C, and would be more efficient in dry inert gas than in humid gas. (author)

10 CFR 71.4 - Definitions.

Science.gov (United States)

2010-01-01

... § 71.15. Graphite means, for the purposes of §§ 71.15 and 71.22, graphite with a boron equivalent... exceed 2 × 10−3 A2/g. Low toxicity alpha emitters means natural uranium, depleted uranium, natural... microcurie/cm2) for beta and gamma and low toxicity alpha emitters, or 0.4 Bq/cm2 (10−5 microcurie/cm2) for...

Nuclear criticality safety controls for uranium deposits during D and D at the Oak Ridge Gaseous Diffusion Plant

International Nuclear Information System (INIS)

Haire, M.J.; Jordan, W.C.; Jollay, L.J. III; Dahl, T.L.

1997-01-01

The US Department of Energy (DOE) Deputy Assistant Secretary of Energy for Environmental Management has issued a challenge to complete DOE environmental cleanup within a decade. The response for Oak Ridge facilities is in accordance with the DOE ten-year plan which calls for completion of > 95% of environmental management work by the year 2006. This will result in a 99% risk reduction and in a significant savings in base line costs in waste management (legacy waste); remedial action (groundwater, soil, etc.); and decontamination and decommissioning (D and D). It is assumed that there will be long-term institutional control of cascade equipment, i.e., there will be no walk away from sites, and that there will be firm radioactivity release limits by 1999 for recycle metals. An integral part of these plants is the removal of uranium deposits which pose nuclear criticality safety concerns in the shut down of the Oak Ridge Gaseous Diffusion Plant. DOE has initiated the Nuclear Criticality Stabilization Program to improve nuclear criticality safety by removing the larger uranium deposits from unfavorable geometry equipment. Nondestructive assay (NDA) measurements have identified the location of these deposits. The objective of the K-25 Site Nuclear Criticality Stabilization Program is to remove and place uranium deposits into safe geometry storage containers to meet the double contingency principle. Each step of the removal process results in safer conditions where multiple controls are present. Upon completion of the Program, nuclear criticality risks will be greatly reduced

A plant taxonomic survey of the Uranium City region, Lake Athabasca north shore, emphasizing the naturally colonizing plants on uranium mine and mill wastes and other human-disturbed sites

International Nuclear Information System (INIS)

Harms, V.L.

1982-07-01

A goal of this study was to acquire more complete baseline data on the existing flora of the Uranium City region, both in natural and human-disturbed sites. Emphasis was given to determining which plant species were naturally revegetating various abandoned uranium mine and mill waste disposal areas, other human-disturbed sites, and ecologically analogous sites. Another goal was to document the occurrence and distribution in the study region of rare and possibly endangered species. A further objective was to suggest regionally-occurring species with potential value for revegetating uranium mine and mill waste sites. Field investigations were carried out in the Uranium City region during August, 1981. During this time 1412 plant collections were made; a total of 366 plant species - trees, shrubs, forbs, graminoids, lichens, and bryophytes were recorded. The report includes an annotated checklist of plant species of the Uranium City region and a reference index of plant taxa indicating species that have high revegetation potential

High temperature soldering of graphite

International Nuclear Information System (INIS)

Anikin, L.T.; Kravetskij, G.A.; Dergunova, V.S.

1977-01-01

The effect is studied of the brazing temperature on the strength of the brazed joint of graphite materials. In one case, iron and nickel are used as solder, and in another, molybdenum. The contact heating of the iron and nickel with the graphite has been studied in the temperature range of 1400-2400 ged C, and molybdenum, 2200-2600 deg C. The quality of the joints has been judged by the tensile strength at temperatures of 2500-2800 deg C and by the microstructure. An investigation into the kinetics of carbon dissolution in molten iron has shown that the failure of the graphite in contact with the iron melt is due to the incorporation of iron atoms in the interbase planes. The strength of a joint formed with the participation of the vapour-gas phase is 2.5 times higher than that of a joint obtained by graphite recrystallization through the carbon-containing metal melt. The critical temperatures are determined of graphite brazing with nickel, iron, and molybdenum interlayers, which sharply increase the strength of the brazed joint as a result of the formation of a vapour-gas phase and deposition of fine-crystal carbon

Analysis of fuel cycles with natural uranium, Phase I; Analiza gorivnih ciklusa sa prirodnim uranom, I faza

Energy Technology Data Exchange (ETDEWEB)

Stojadinovic, A; Zivkovic, Z; Raisic, N [Institute of Nuclear Sciences Boris Kidric, Laboratorija za fiziku i dinamiku reaktora, Vinca, Beograd (Serbia and Montenegro)

1964-12-15

This paper contains analyses of fuel cycles with natural uranium for the following cases: plutonium recycling is not done; recycling of plutonium and irradiated uranium with the condition of equal multiplication factor at the beginning of each cycle; and recycling of plutonium only.

Radon 226 and natural Uranium in potable waters to the Argentina Republic

International Nuclear Information System (INIS)

Bomben, A.M.; Palacios, M.A.

1998-01-01

157 samples were analyzed in the Buenos Aires City. Gathered in the domiciliary distribution net and private wells. The radon 226 concentration to determines for the radon 226 emanation technique and liquid scintilligraphy. The natural uranium concentration one carries out for fluorimetric methods

Electronic properties of graphite

International Nuclear Information System (INIS)

Schneider, J.

2010-10-01

In this thesis, low-temperature magneto-transport (T ∼ 10 mK) and the de Haas-van Alphen effect of both natural graphite and highly oriented pyrolytic graphite (HOPG) are examined. In the first part, low field magneto-transport up to B = 11 T is discussed. A Fourier analysis of the background removed signal shows that the electric transport in graphite is governed by two types of charge carriers, electrons and holes. Their phase and frequency values are in agreement with the predictions of the SWM-model. The SWM-model is confirmed by detailed band structure calculations using the magnetic field Hamiltonian of graphite. The movement of the Fermi at B > 2 T is calculated self-consistently assuming that the sum of the electron and hole concentrations is constant. The second part of the thesis deals with high field magneto-transport of natural graphite in the magnetic field range 0 ≤ B ≤ 28 T. Both spin splitting of magneto-transport features in tilted field configuration and the onset of the charge density wave (CDW) phase for different temperatures with the magnetic field applied normal to the sample plane are discussed. Concerning the Zeeman effect, the SWM calculations including the Fermi energy movement require a g-factor of g* equal to 2.5 ± 0.1 to reproduce the spin spilt features. The measurements of the charge density wave state confirm that its onset magnetic field can be described by a Bardeen-Cooper-Schrieffer (BCS)-type formula. The measurements of the de Haas-van Alphen effect are in agreement with the results of the magneto-transport measurements at low field. (author)

Study of the strength of the internal can for internally and externally cooled fuel elements intended for gas graphite reactors; Etude de la tenue de la gaine interne pour-element combustible a refroidissement interne et externe d'un reacteur graphite-gaz

Energy Technology Data Exchange (ETDEWEB)

Boudouresque, B; Courcon, P; Lestiboubois, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The cartridge of an internally and externally cooled annular fuel element used in gas-graphite reactors is made up of an uranium fuel tube, an external can and an internal can made of magnesium alloy. For the thermal exchange between the internal can and the fuel to be satisfactory, it is necessary for the can to stay in contact with the uranium under all temperature conditions. This report, based on a theoretical study, shows how the internal can fuel gap varies during the processes of canning, charging into the reactor and thermal cycling. The following parameters are considered: tube diameter, pressure of the heat carrying gas, gas entry temperature, plasticity of the can alloy. It is shown that for all operating conditions the internal can of a 77 x 95 element, planned for a gas-graphite reactor with a 40 kg/cm{sup 2} gas pressure, should remain in contact with the fuel. (authors) [French] La cartouche d'un element combustible annulaire, a refroidissement interne et externe pour reacteur graphite-gaz, est composee d'un tube combustible en uranium, d'une gaine externe et d'une gaine interne en alliage de magnesium. Pour que l'echange thermique entre la gaine interne et le combustible soit bon, il faut que la gaine reste appliquee sur l'uranium quel que soit le regime de temperature. Cette note a pour but de montrer comment, d'apres une etude theorique, le jeu combustible-gaine interne varie au cours des operations de gainage, de chargement dans le reacteur, et des cyclages thermiques. Les parametres suivants sont etudies: diametres de tube, pression du gaz caloporteur, temperature d'entree du gaz, plasticite de l'alliage de gaine. Il est montre que, quel que soit le regime de fonctionnement, la gaine interne d'un element 77 x 95, en projet pour un reacteur graphite-gaz sous pression de 40 kg/cm{sup 2}, doit rester appliquee sur le combustible. (auteurs)

Two criticisms of natural theology

Directory of Open Access Journals (Sweden)

Błażej Gębura

2014-01-01

Full Text Available The article aims at considering two general criticisms often formulated against the natural theology. First criticism is based on the thesis that the conclusions of the natural theology are not adequate with the religious beliefs of non-philosophers. It is widely known as opposition between God of Religion and God of Philosophers. One can find that argument in the writings of Blaise Pascal. I’m arguing for the thesis, that the natural theologian cannot fulfill the criteria given by the proponents of this argument. This is because the argument of the natural theology cannot contains the premises taken from the Revelation. If the argument of the natural theology would contain the premises taken from the Revelation, then it would be the argument of religion. But philosopher of religion (natural theologian can’t do this, if he wants to formulate an philosophical argument. The second criticism is based on the notion of a rational person. In the light of this argument, the natural theology is successful only, if every rational person will accept the conclusion “God exist”. I’m trying to show that there is no philosophical argument that can guarantee it’s acceptance by some rational persons. The acceptance of the conclusion of the argument of the natural theology is a matter of personal decision. There is no logical argument, which can “force” rational persons (rational subjects to accept it’s conclusion. But if this is true, the arguments for the existence of God are no worse than other philosophical arguments.

Lattice dynamical appraisal of the anisotropic Debye-Waller factors in graphite lattice

International Nuclear Information System (INIS)

Haridasan, T.M.; Sathyamurthy, G.

1989-12-01

The Debye-Waller factors in graphite for the atomic motions within the basal plane and also across the basal planes have been calculated using the various lattice dynamical models available to date and a critical comparison is made with the existing experimental data from X ray and neutron scattering studies. The present study reveals the need for further investigation on the nature of atomic motion across the basal planes. (author). 15 refs, 1 tab

Determination of natural uranium in urine (233U)

International Nuclear Information System (INIS)

Jeanmaire, L.; Jammet, H.

1959-01-01

A procedure for the quantitative analysis of uranium in urine is described. The residue obtained by mineralization is dissolved in diluted hydrochloric acid. Uranium is separated by fixation on a permutit 50 column, elution with 0,2 M oxalic acid and electrodeposition on nickel. Uranium is then measured by α counting. It is thus possible to detect less than 1 pico-curie of uranium in the sample. (author) [fr

An experimental stack for the control of uranium enrichment; Empilement pour le controle de l'enrichissement de l'uranium

Energy Technology Data Exchange (ETDEWEB)

Bailly du Bois, B; Raievski, V; Tretiakoff, O [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

An apparatus is described for controlling the enrichment of the uranium in a fuel element at all stages in its fabrication. The apparatus consists of a stack of graphite surrounded by cadmium, in which a thermal neutron density is maintained by a Ra-Be source placed in the stack. The fuel element to be controlled is placed in the stack at 16 cm from the cadmium wall and at 70 cm from the source. The fast neutrons produced in this fuel element are slowed down and detected by a BF{sub 3} counter placed in a second graphite stack adjacent to the first. The enrichment is obtained by comparing the counting rate measured with the element to be controlled, with that of a geometrically similar standard element, previously calibrated. (author) [French] On decrit un dispositif permettant de controler l'enrichissement de l'uranium contenu dans un element de combustible a tous les stades de sa fabrication. Le dispositif est constitue par un empilement de graphite entoure de cadnium dans lequel regne une densite de neutrons thermiques entretenus par une source de Ra-Be placee dans l'empilement. L'element de combustible a contr er est place dans cet empilement a 16 cm de la paroi de cadmium et a 70 cm de la source. Les neutrons rapides produits dans cet element de combustible sont ralentis et detectes par un compteur a BF{sub 3} place dans un deuxieme empilement de graphite contigu au premier. L'enrichissement est obtenu en comparant les taux de comptage mesures arec l'element a controler et un element standard geometriquement semblable, prealablement etalonne. (auteur)

Criticality safety of storage barrels for enriched uranium fresh fuel at the RB research reactor

International Nuclear Information System (INIS)

Pesic, M. P.

1997-01-01

Study on criticality safety of fresh low and high enriched uranium (LEU and HEU) fuel elements in the storage/transport barrels at the RB research reactor is carried out by using the well-known MCNP computer code. It is shown that studied arrays of tightly closed fuel barrels, each entirely loaded with 100 fresh (HEU or LEU) fuel slugs, are far away from criticality, even in cases of an unexpected flooding by light water.(author)

Laboratory studies on natural restoration of ground water after in-situ leach uranium mining

International Nuclear Information System (INIS)

Bell, N.E.; Deutsch, W.J.; Serne, R.J.

1983-05-01

When uranium is mined using in-situ leach techniques, the chemical quality of the ground water in the ore-zone aquifer is affected. This could lead to long-term degradation of the ground water if restoration techniques are not applied after the leaching is completed. Pacific Northwest Laboratory (PNL), is conducting an NRC-sponsored research project on natural restoration and induced-restoration techniques. Laboratory studies were designed to evaluate the ability of the natural system (ore-zone sediments and groundwater) to mitigate the effects of mining on aquifer chemistry. Using batch and flow-through column experiments [performed with lixiviant (leaching solution) and sediments from the reduced zone of an ore-zone aquifer], we found that the natural system can lower uranium and bicarbonate concentrations in solutions and reduce the lixiviant redox potential (Eh). The change in redox potential could cause some of the contaminants that were dissolved during the uranium leaching operation to precipitate, thereby lowering their solution concentration. The concentrations of other species such as calcium, potassium, and sulfate increased, possibly as a result of mineral dissolution and ion exchange. In this paper, we describe the experimentally determined mobility of contaminants after in-situ leach mining, and discuss the possible chemical process affecting mobility

Laboratory studies on natural restoration of ground water after in-situ leach uranium mining

Energy Technology Data Exchange (ETDEWEB)

Bell, N.E.; Deutsch, W.J.; Serne, R.J.

1983-05-01

When uranium is mined using in-situ leach techniques, the chemical quality of the ground water in the ore-zone aquifer is affected. This could lead to long-term degradation of the ground water if restoration techniques are not applied after the leaching is completed. Pacific Northwest Laboratory (PNL), is conducting an NRC-sponsored research project on natural restoration and induced-restoration techniques. Laboratory studies were designed to evaluate the ability of the natural system (ore-zone sediments and groundwater) to mitigate the effects of mining on aquifer chemistry. Using batch and flow-through column experiments (performed with lixiviant (leaching solution) and sediments from the reduced zone of an ore-zone aquifer), we found that the natural system can lower uranium and bicarbonate concentrations in solutions and reduce the lixiviant redox potential (Eh). The change in redox potential could cause some of the contaminants that were dissolved during the uranium leaching operation to precipitate, thereby lowering their solution concentration. The concentrations of other species such as calcium, potassium, and sulfate increased, possibly as a result of mineral dissolution and ion exchange. In this paper, we describe the experimentally determined mobility of contaminants after in-situ leach mining, and discuss the possible chemical process affecting mobility.

Validation of the Monte Carlo criticality program KENO IV and the Hansen-Roach sixteen-energy-group-cross sections for high-assay uranium systems

International Nuclear Information System (INIS)

Handley, G.R.; Masters, L.C.; Stachowiak, R.V.

1981-01-01

Validation of the Monte Carlo criticality code, KENO IV, and the Hansen-Roach sixteen-energy-group cross sections was accomplished by calculating the effective neutron multiplication constant, k/sub eff/, of 29 experimentally critical assemblies which had uranium enrichments of 92.6% or higher in the uranium-235 isotope. The experiments were chosen so that a large variety of geometries and of neutron energy spectra were covered. Problems, calculating the k/sub eff/ of systems with high-uranium-concentration uranyl nitrate solution that were minimally reflected or unreflected, resulted in the separate examination of five cases

Assay of uranium in fused salt cake generated at the natural uranium metal fuel fabrication plants by gamma-ray spectrometry

International Nuclear Information System (INIS)

Kalsi, P.C.; Bhanu, A.U.; Sahoo, S.; Iyer, R.H.

1986-01-01

A passive gamma-ray spectroscopic method is employed for the assay of uranium in fused salt cake, a scrap produced at the natural uranium metal fuel fabrication plants. The method makes use of NaI(TI) detector coupled with a multichannel analyser. The 1 MeV gamma-ray of 238 U was used for the calibration. The calibration curve was made by counting synthetic mixtures made of U 3 O 8 powder, the heat treatment salt and iron in the form of fine powder. The uranium content in these synthetic mixtures was kept in the range of 1-11 per cent. 23 lots of the fused salt cake taken from three different batches of the salt cake were then analysed by this method. The uranium content of fused salt cake was found to be in the range of 1.70-11.43 per cent. To compare the gamma spectrometric results with a completely independent method, chemical analysis of all the fused salt cakes were also carried out. The NDA results were found to agree within ± 17 per cent with the chemical analysis results. (author)

ZPR-3 Assembly 6F : A spherical assembly of highly enriched uranium, depleted uranium, aluminum and steel with an average {sup 235}U enrichment of 47 atom %.

Energy Technology Data Exchange (ETDEWEB)

Lell, R. M.; McKnight, R. D; Schaefer, R. W.; Nuclear Engineering Division

2010-09-30

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 6 consisted of six phases, A through F. In each phase a critical configuration was constructed to simulate a very simple shape such as a slab, cylinder or sphere that could be analyzed with the limited analytical tools available in the 1950s. In each case the configuration consisted of a core region of metal plates surrounded by a thick depleted uranium metal reflector. The average compositions of the core configurations were essentially identical in phases A - F. ZPR-3

Graphene-graphite oxide field-effect transistors.

Science.gov (United States)

Standley, Brian; Mendez, Anthony; Schmidgall, Emma; Bockrath, Marc

2012-03-14

Graphene's high mobility and two-dimensional nature make it an attractive material for field-effect transistors. Previous efforts in this area have used bulk gate dielectric materials such as SiO(2) or HfO(2). In contrast, we have studied the use of an ultrathin layered material, graphene's insulating analogue, graphite oxide. We have fabricated transistors comprising single or bilayer graphene channels, graphite oxide gate insulators, and metal top-gates. The graphite oxide layers show relatively minimal leakage at room temperature. The breakdown electric field of graphite oxide was found to be comparable to SiO(2), typically ~1-3 × 10(8) V/m, while its dielectric constant is slightly higher, κ ≈ 4.3. © 2012 American Chemical Society

Influence of the reduction-crucible material on the uranium properties

International Nuclear Information System (INIS)

Braga, F.J.C.; Bose, A.; Freitas, C.T. de

1979-01-01

The uranium obtained by UF 4 reduction using Mg in bombs coated with different materials such as alumina, blast furnace slag, Zirconia and graphite was studied. The reduction process involves a reaction that altains temperatures of the order of 1600 0 C at tightly closed enclosure environment. Assuming in this process that the only possible influencial agent on the reaction main product, i.e., metallic uranium is the own bomb coaling, different properties, mechanical-metallurgical and phase-transformation characteristics were examined and the influences of the coating materials were compared. The comparison of these properties was also studied in uranium refined by arc fusion. (Author) [pt

Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

International Nuclear Information System (INIS)

Monado, F.; Permana, S.

2013-01-01

Full-text: A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8 % HM. From the neutronic point of view, this design is in compliance with good performance. (author)

Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

International Nuclear Information System (INIS)

Monado, Fiber; Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik; Aziz, Ferhat; Sekimoto, Hiroshi

2014-01-01

A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance

LED vs laser fluorimetry: a comparative study for the determination of uranium in natural waters

International Nuclear Information System (INIS)

Shenoy, N.; Parab, H.; Sounderajan, S.; Kiran Kumar; Kumar, S.D.; Reddy, A.V.R.

2015-01-01

Measurement of uranium in water samples has acquired considerable importance ever since its occurrence in drinking water sources was reported. Among the various methods available for uranium quantification at ultra trace levels, laser fluorimetry (LF) method is the method of choice due to its simplicity, speed and high sensitivity compared to other analytical techniques. This technique is based on the measurement of fluorescence of uranium complexes in aqueous solution. Recently, laser source has been replaced by light emitting diode (LEDs) in the fluorimeter systems. In comparison to laser source, LED source is, cost effective, generates less heat and has extended lifetime. Herein, authors have presented a comparison of LED based fluorimeter (Quantalase, Indore, India) and laser fluorimeter (CAT, Indore, India) for the determination of uranium in natural waters

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

Radioactivity of uranium production cycle facilities in the Czech Republic compared to the natural environment

International Nuclear Information System (INIS)

Matolin, M.

2002-01-01

Forty-five years (1946-1990) of intensive uranium exploration and exploitation in the Czech Republic led to mining at 64 uranium deposits. These mining and milling activities left numerous accumulations of waste rock material in the landscape. The radioactivity of these man-made accumulations was measured and compared to the natural radiation environment. Waste rock dumps at the uranium deposits Pribram, Rozna, Jachymov, Straz-Hamr and deposits in the Zelezne Hory area show surface gamma dose rates mostly in the range of 200-1000 nGy/h, with a uranium concentration 10-100 ppm eU. An extremely high radioactivity of 3000-4200 nGy/h was detected at the extensive uranium processing tailings impoundments at Straz. Terrestrial gamma dose rate of regional geological units in the Czech Republic is in the range of 6-245 nGy/h. Reclamation and recultivation of dumps, control of their radioactivity and restriction of their accessibility are the major measures introduced to protect the public. (author)

Natural uranium fueled light water moderated breeding hybrid power reactors

International Nuclear Information System (INIS)

Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.; Levin, P.

The feasibility of fission-fusion hybrid reactors based on breeding light water thermal fission systems is investigated. The emphasis is on fuel-self-sufficient (FSS) hybrid power reactors that are fueled with natural uranium. Other LWHRs considered include FSS-LWHRs that are fueled with spent fuel from LWRs, and LWHRs which are to supplement LWRs to provide a tandem LWR-LWHR power economy that is fuel-self-sufficient

Smectite-zeolite envelope surrounding the Tsukiyoshi uranium deposit, central Japan. A natural analogue study

International Nuclear Information System (INIS)

Utada, Minoru

2003-01-01

The Tsukiyoshi uranium deposit in Gifu Prefecture is the largest one in Japan. It is embedded in lower part of the Mizunami Group of Miocene age. Relating to the existence of this uranium deposit, the constituent minerals in sediments were studied by XRD and SEM, using many drilling cores. The most abundant authigenic mineral is smectite. The amount of smectite increases generally from upper to lower horizons, and a highly smectitized zone is situated around the uranium deposit. Smectitization predominated in mafic glassy grains of sediments, which was probably formed in early burial diagenesis. Zeolites including clinoptilolite-heulandite, mordenite, analcime, chabazite and philipsite are secondly abundant authigenic minerals. They seem to have been formed at early to late diagenetic stages. Opaline silica is rather rare. Carbonate minerals, including calcite, dolomite, siderite and rhodocrosite are common. They may be formed by diagenesis as well. Gypsum and pyrite occur in upper horizons and lower horizons, respectively. In particular, a highly smectitized zone including pyrite probably played an important role for retarding the migration of uranium and as a result keeping the uranium deposit for past one million years. This smectite-zeolite envelope surrounding the Tsukiyoshi uranium deposit is regarded as a natural analogue of the buffer materials surrounding the high-level radioactive waste repository. (author)

On the nature of the phase transition in uranium dioxide

Science.gov (United States)

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

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

Search for other natural fission reactors

International Nuclear Information System (INIS)

Apt, K.E.; Balagna, J.P.; Bryant, E.A.; Cowan, G.A.; Daniels, W.R.; Vidale, R.J.

1977-01-01

Precambrian uranium ores have been surveyed for evidence of other natural fission reactors. The requirements for formation of a natural reactor direct investigations to uranium deposits with large, high-grade ore zones. Massive zones with volumes approximately greater than 1 m 3 and concentrations approximately greater than 20 percent uranium are likely places for a fossil reactor if they are approximately greater than 0.6 b.a. old and if they contained sufficient water but lacked neutron-absorbing impurities. While uranium deposits of northern Canada and northern Australia have received most attention, ore samples have been obtained from the following worldwide locations: the Shinkolobwe and Katanga regions of Zaire; Southwest Africa; Rio Grande do Norte, Brazil; the Jabiluka, Nabarlek, Koongarra, Ranger, and El Sharana ore bodies of the Northern Territory, Australia; the Beaverlodge, Maurice Bay, Key Lake, Cluff Lake, and Rabbit Lake ore bodies and the Great Bear Lake region, Canada. The ore samples were tested for isotopic variations in uranium, neodymium, samarium, and ruthenium which would indicate natural fission. Isotopic anomalies were not detected. Criticality was not achieved in these deposits because they did not have sufficient 235 U content (a function of age and total uranium content) and/or because they had significant impurities and insufficient moderation. A uranium mill monitoring technique has been considered where the ''yellowcake'' output from appropriate mills would be monitored for isotopic alterations indicative of the exhumation and processing of a natural reactor

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

Measurement of reactivity worths of burnable poison rods in enriched uranium graphite-moderated core simulated to high temperature gas cooled reactor

International Nuclear Information System (INIS)

Akino, Fujiyoshi; Takeuchi, Motoyoshi; Kitadate, Kenji; Yoshifuji, Hisashi; Kaneko, Yoshihiko

1980-11-01

As the core design for the Experimental Very High Temperature Gas Cooled Reactor progresses, evaluation of design precision has become increasingly important. For a high precision design, it is required to have adequate group constants based on accurate nuclear data, as well as calculation methods properly describing the physical behavior of neutrons. We, therefore, assembled a simulation core for VHTR, SHE-14, using a graphite-moderated 20%-enriched uranium Semi-Homogeneous Experimental Critical Facility (SHE), and obtained useful experimental data in evaluating the design precision. The VHTR is designed to accommodate burnable poison and control rods for reactivity compensation. Accordingly, the experimental burnable poison rods which are similar to those to be used in the experimental reactor were prepared, and their reactivity values were measured in the SHE-14 core. One to three rods of the above experimental burnable poison rods were inserted into the central column of the SHE-14 core, and the reactivity values were measured by the period and fuel rod substitution method. The results of the measurements have clearly shown that due to the self-shielding effect of B 4 C particles the reactivity value decreases with increasing particle diameter. For the particle diameter, the reactivity value is found to increase linearly with the logarithm of boron content. The measured values and those calculated are found to agree with each other within 5%. These results indicate that the reactivity of the burnable poison rod can be estimated fairly accurately by taking into account the self-shielding effect of B 4 C particles and the heterogeneity of the lattice cell. (author)

Hot pressing of U-UC cermets and stoichiometric uranium monocarbide; Preparation par frittage sous charge de cermets U-UC et de monocarbure stoechiometrique

Energy Technology Data Exchange (ETDEWEB)

Dubuisson, J; Houyvet, A; Le Boulbin, E; Lucas, R; Moranville, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

High density fuels, either in uranium monocarbide or in U-UC cermets have been prepared on laboratory-scale, by hot pressing of mixtures of uranium powder and graphite in suitable proportions. Uranium powder is prepared by calcium reduction of UO{sub 2} followed by an acetic leaching at low temperature. An adequate protection-treatment permits the manipulation of the powder in the open air. Uranium and Graphite powders are intimately mixed and then hot pressed in a double effect graphite die at a temperature of 900-1000 deg. C under a charge of 200 kg/cm{sup 2} during 3 hours. A special design of the die avoids the breaking of the graphite during the sintering. In this way, samples are prepared, the characteristics of which are: 1) {+-} 5 pour cent of homogeneity for a ratio height/diameter = 2. 2) almost theoretical density (98 pour cent) 3) low concentration of unreacted carbon (heat treatment of stoichiometric monocarbide can be useful for completion of reaction) 4) the micrographic examination shows: - a network of monocarbide surrounding uranium in the case of low concentration cermets (<2,5 per cent C) - two networks intimately mixed for high concentration cermets (<2,5 per cent C) - a fine grain structure for the monocarbide (10 u). 5) In every case, the X rays examinations show a fine grain structure without any orientation, and no UC{sub 2}. Some indications are given on the physical (thermal cycling, conductibility) and chemical properties (corrosion, reaction with cladding materials). (author)Fren. [French] Une methode de preparation de combustibles de haute densite, soit en monocarbure d'uranium, soit en cermets U-UC, a ete mise au point au laboratoire. Il s'agit du frittage sous charge de melanges de poudres d'uranium et de graphite en proportion convenable. La poudre d'uranium est elaboree par calciothermie de l'oxyde UO{sub 2} suivie d'un lavage acetique a basse temperature. Un traitement de protection adequat pe sa manipulation a l'air. Les poudres

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Initial conceptual design study of self-critical nuclear pumped laser systems

Science.gov (United States)

Rodgers, R. J.

1979-01-01

An analytical study of self-critical nuclear pumped laser system concepts was performed. Primary emphasis was placed on reactor concepts employing gaseous uranium hexafluoride (UF6) as the fissionable material. Relationships were developed between the key reactor design parameters including reactor power level, critical mass, neutron flux level, reactor size, operating pressure, and UF6 optical properties. The results were used to select a reference conceptual laser system configuration. In the reference configuration, the 3.2 m cubed lasing volume is surrounded by a graphite internal moderator and a region of heavy water. Results of neutronics calculations yield a critical mass of 4.9 U(235) in the form (235)UF6. The configuration appears capable of operating in a continuous steady-state mode. The average gas temperature in the core is 600 K and the UF6 partial pressure within the lasing volume is 0.34 atm.

Critical review of uranium resources and production capability to 2020

International Nuclear Information System (INIS)

1998-08-01

This report was prepared to assess the changing uranium supply and demand situation as well as the adequacy of uranium resources and the production capability to supply uranium concentrate to meet reactor demand through 2020. Uranium production has been meeting only 50 to 60 percent of the world requirements with the balance met from sale of excess inventory offered on the market at low prices. It is generally agreed by most specialists that the end of the excess inventory is approaching. With inventory no longer able to meet the production shortfall it is necessary to significantly expand uranium production to fill an increasing share of demand. Non-production supplies of uranium, such as the blending of highly enriched uranium (HEU) warheads to produce low enriched reactor fuel and reprocessing of spent fuel, are also expected to grow in importance as a fuel source. This analysis addresses three major concerns as follows: adequacy of resources to meet projected demand; adequacy of production capability to produce the uranium; and market prices to sustain production to fill demand. This analysis indicates uranium mine production to be the primary supply providing about 76 to 78 percent of cumulative needs through 2020. Alternative sources supplying the balance, in order of relative importance are: (1) low enriched uranium (LEU) blended from 500 tonnes of highly enriched uranium (HEU) Russian weapons, plus initial US Department of Energy (US DOE) stockpile sales (11 to 13%); (2) reprocessing of spent nuclear fuel (6%) and; (3) utility and Russian stockpiles. Further this report gives uranium production profiles by countries: CIS producers (Kazakhstan, Russian Federation, Ukraine, Uzbekistan) and other producers (Australia, Canada, China, Gabon, Mongolia, Namibia, Niger, South Africa, United States of America)

Standard test methods for arsenic in uranium hexafluoride

CERN Document Server

American Society for Testing and Materials. Philadelphia

2005-01-01

1.1 These test methods are applicable to the determination of total arsenic in uranium hexafluoride (UF6) by atomic absorption spectrometry. Two test methods are given: Test Method A—Arsine Generation-Atomic Absorption (Sections 5-10), and Test Method B—Graphite Furnace Atomic Absorption (Appendix X1). 1.2 The test methods are equivalent. The limit of detection for each test method is 0.1 μg As/g U when using a sample containing 0.5 to 1.0 g U. Test Method B does not have the complete collection details for precision and bias data thus the method appears as an appendix. 1.3 Test Method A covers the measurement of arsenic in uranyl fluoride (UO2F2) solutions by converting arsenic to arsine and measuring the arsine vapor by flame atomic absorption spectrometry. 1.4 Test Method B utilizes a solvent extraction to remove the uranium from the UO2F2 solution prior to measurement of the arsenic by graphite furnace atomic absorption spectrometry. 1.5 Both insoluble and soluble arsenic are measured when UF6 is...

Determination of natural uranium, thorium and radium isotopes in water and soil samples by alpha spectroscopy

Energy Technology Data Exchange (ETDEWEB)

Hao, Le Cong; Tao, Chau Van; Thong, Luong Van; Linh, Duong Mong [University of Science Ho Chi Minh City (Viet Nam). Faculty of Physics and Engineering Physics; Dong, Nguyen Van [University of Science Ho Chi Minh City (Viet Nam). Faculty of Chemistry

2011-08-15

In this study, a simple procedure for the determination of natural uranium, thorium and radium isotopes in water and soil samples by alpha spectroscopy is described. This procedure allows a sequential extraction polonium, uranium, thorium and radium radionuclides from the same sample in two to three days. It was tested and validated with the analysis of certified reference materials from the IAEA. (orig.)

Rapid determination of uranium in natural waters by fthermal emission mass spectrometry

International Nuclear Information System (INIS)

Ferguson, J.R.; Caylor, J.D.; Rogers, E.R.; Cole, S.H.

1977-03-01

A method has been developed to rapidly analyze natural water samples for part-per-trillion (ng/l) concentrations of uranium using a custom-built thermal-emission mass spectrometer. The filtered water sample is spiked with 233 U as an internal standard and extracted with a 2 percent solution of TOPO (trioctylphosphine oxide) in carbon tetrachloride. An aliquot of the organic phase is evaporated and the uranium in the residue extracted with aqueous ammonium carbonate. A 5j-μl aliquot is taken and dried on a flat uranium concentration of 3 ng/l will yield a count rate greater than three times the standard deviation, plus the mean of the background, and is defined as the lowest determinable concentration. The standard deviation of the method is 3 percent at accuracy of the method has been evaluated by comparing the results with a fluorescence procedure. There is very good agreement for water samples with uranium concentrations from 200 to 1000 ng/l. The mass spectrometer is a 6-in. -radius, 60-degree-sector instrument equipped for ion counting and having a vacuum system allowing rapid sample changing while maintaining a high source vacuum. A multiplexer and high-voltage s witch provide synchronized peak switching and scaler gating for monitoring three isotopes of uranium 238, 235, and 233. With this instrument, an analyst can achieve an analysis rate in excess of 50 samples per eight-hour shift

Spectrographic determination of impurities in uranium tetrafluoride

International Nuclear Information System (INIS)

Capdevila Perez, C.; Roca Adell, M.; Alvarez Gonzalez, F.

1967-01-01

A carrier distillation method for the determination of Ag, Al, As, B, Cd, Cr, Cu, Fe, Mg, Mn, Ni, Pb, and Si in uranium tetrafluoride was develop ped. the previous addition of 25% Y 2 3 prevents the excitation of uranium by conversion of the volatile UF 4 into U 3 0 8 during the arc discharge. NaCl or Ga 2 0 3 , containing Ge and V as internal standards, are used as carriers, and samples are arced in 10 Amp. d.c. arc in a graphite anode cup. 7 mm diameter, 10 mm deep, being the weight of charge 300 mg. (Author) 14 refs

The future of the uranium mining industry

International Nuclear Information System (INIS)

Capus, G.; Galaud, G.

1993-01-01

This paper presents the state of natural Uranium market today. In a first part, the author gives a brief history about nuclear programs history in Usa and Europe and describes natural Uranium demand and supply (Uranium mines, recycling, excessive civil stocks, military stocks using). In a second part, evolutions and futures of Uranium industry is studied: using of excessive stocks in Western Europe, using of military stocks, recycling of Uranium from spent fuels reprocessing, uranium deposits, future natural uranium market. 6 refs., 4 figs., 3 tabs., 3 photos

Once-through uranium thorium fuel cycle in CANDU reactors

International Nuclear Information System (INIS)

Ozdemir, S.; Cubukcu, E.

2000-01-01

In this study, the performance of the once-through uranium-thorium fuel cycle in CANDU reactors is investigated. (Th-U)O 2 is used as fuel in all fuel rod clusters where Th and U are mixed homogeneously. CANDU reactors have the advantage of being capable of employing various fuel cycle options because of its good neutron economy, continuous on line refueling ability and axial fuel replacement possibility. For lattice cell calculations transport code WIMS is used. WIMS cross-section library is modified to achieve precise lattice cell calculations. For various enrichments and Th-U mixtures, criticality, heavy element composition changes, diffusion coefficients and cross-sections are calculate. Reactor core is modeled by using the diffusion code CITATION. We conclude that an overall saving of 22% in natural uranium demand can be achieved with the use of Th cycle. However, slightly enriched U cycle still consumes less natural Uranium and is a lot less complicated. (author)

Low-energy electron observation of graphite and molybdenite crystals. Application to the study of graphite oxidation

International Nuclear Information System (INIS)

David, G.

1969-01-01

The LEED study of cleaved (0001) faces of crystals having a layered structure allowed to investigate flakes free of steps on graphite and molybdenite, to show twinning on natural graphite. By intensity measurements and computation in the case of a kinematical approximation it has been possible to determine an inner potential of 19 eV for graphite and to identify the direction of the Mo-S bond of the surface layer of molybdenite. The oxidation of graphite has been studied by observing changes, in symmetry of the diffraction patterns and by mass spectrometry of the gases evolved during the oxidation. No surface compounds have been detected and the carbon layers appeared to be peeled off one after the other. The oxidation took place at temperatures higher than 520 C under an oxygen pressure of 10 -5 torr. (author) [fr

Uranium, depleted uranium, biological effects

International Nuclear Information System (INIS)

2001-01-01

Physicists, chemists and biologists at the CEA are developing scientific programs on the properties and uses of ionizing radiation. Since the CEA was created in 1945, a great deal of research has been carried out on the properties of natural, enriched and depleted uranium in cooperation with university laboratories and CNRS. There is a great deal of available data about uranium; thousands of analyses have been published in international reviews over more than 40 years. This presentation on uranium is a very brief summary of all these studies. (author)

Management of UKAEA graphite liabilities

International Nuclear Information System (INIS)

Wise, M.

2001-01-01

development of these waste disposal strategies. UKAEA is liasing closely with the disposal organisations to ensure that all their concerns are addressed satisfactorily. This paper will describe: i. the varied nature, inventory and history of UKAEA's graphite liabilities; ii. the options that have been considered for the long-term storage and disposal of UKAEA's graphite liabilities; iii. the technical issues that have been considered in the development of these options; iv. the recent developments in the consideration of Wigner energy. The paper will summarise the current status and future plans of the UKAEA graphite waste packaging strategies, and describe the interactions with the disposal organisations. (author)

Highly conductive bridges between graphite spheres to improve the cycle performance of a graphite anode in lithium-ion batteries

Energy Technology Data Exchange (ETDEWEB)

Wang, Hongyu [IM and T Ltd., Advanced Research Center, Saga University, Yoga-machi 1341, Saga 840-0047 (Japan); Umeno, Tatsuo; Mizuma, Koutarou [Research Center, Mitsui Mining Co. Ltd., Hibiki-machi 1-3, Wakamatsu-ku, Kitakyushu 808-0021 (Japan); Yoshio, Masaki [Advanced Research Center, Saga University, Yoga-machi 1341, Saga 840-0047 (Japan)

2008-01-10

Spherical carbon-coated natural graphite (SCCNG) is a promising anode material for lithium-ion batteries, but the smooth surface of graphite spheres is difficult to wet with an aqueous binder solution, and lacks electrical contacts. As a result, the cycle performance of such a graphite anode material is not satisfactory. An effective method has been introduced to tightly connect adjacent SCCNG particles by a highly conductive binder, viz. acetylene black bridges. The effect of the conductive bridges on the cyclability of SCCNG electrode has been investigated. (author)

Spontaneous ignition of natural uranium in Tokai Research Establishment, Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

1989-01-01

At P.M. 7:23, May 30, 1989, a fire alarm functioned in the uranium enrichment laboratory building, and immediately investigation was carried out, as the result, smoke was confirmed in the nuclear fuel storage. In the nuclear fuel storage, there were five plastic bottles containing natural uranium chips, and smoke arose from three of them. Immediately fire fighting was carried out with powder extinguishers and others, the uranium chips which were regarded as the heat generating source were moved into stainless steel cans, and air was cut off with extinguishing sand, as the result, around P.M. 9:50, heat generation ceased. At present the detailed cause is being investigated, but it is considered that the uranium chips contained in plastic bottles reacted with air by some cause, and generated heat in the form of spontaneous ignition, as the result, the plastic bottles and the vinyl sheets placed under them smoked. The stack dust monitor in the uranium enrichment laboratory building showed the normal value, and there was not the effect to surrounding environment. The workers who did fire fighting with whole face masks were not affected by smoke. (K.I.)

Fuel Cycle Impacts of Uranium-Plutonium Co-extraction

International Nuclear Information System (INIS)

Taiwo, Temitope; Szakaly, Frank; Kim, Taek-Kyum; Hill, Robert

2008-01-01

A systematic investigation of the impacts of uranium and plutonium co-extraction during fuel separations on reactor performance and fuel cycle has been performed. Proliferation indicators, critical mass and radiation source levels of the separation products or fabricated fuel, were also evaluated. Using LWR-spent-uranium-based MOX fuel instead of natural-uranium-based fuel in a PWR MOX core requires a higher initial plutonium content (∼1%), and results in higher Np-237 content (factor of 5) in the spent fuel, and less consumption of Pu-238 (20%) and Am-241 (14%), indicating a reduction in the effective repository space utilization. Additionally, minor actinides continue to accumulate in the fuel cycle, and thus a separate solution is required for them. Differences were found to be quite smaller (∼0.4% in initial transuranics) between the equilibrium cycles of advanced fast reactor cores using spent and depleted uranium for make-up, in additional to transuranics. The critical masses of the co-extraction products were found to be higher than for weapons-grade plutonium (WG-Pu) and the decay heat and radiation sources of the materials (products) were also found to be generally higher than for WG-Pu in the transuranics content range of 10% to 100% in the heavy-metal. (authors)

Simultaneous determination of Ra-226, natural uranium and natural thorium by gamma-ray spectrometry INa(Ti), in solid samples.; Determinacion de U (Natural), Th (Natural) y Ra-226 en diversos materiales, mediante espectrometria con INa (TI)

Energy Technology Data Exchange (ETDEWEB)

Salvador, S.; Navarro, T.; Alvarez, A.

1991-07-01

A method has been developed to determine activities of Ra-226, natural uranium and natural thorium by gamma-ray spectrometry. The measurement system has been calibrated using standards specially prepared at the laboratory. It is necessary to assume secular equilibrium in the samples, between Ra-226 and Th-232 and its daughters nuclides, and between U-238 and its immediate daughter Th-234, as the photo peaks measured are those of the daughters. The results obtained indicate that this method can of ter replace the radiochemical techniques used to measure activities in this type of sample. The method has been successfully used to determine these natural isotopes in samples from uranium mills. (Author) 9 refs.

Sensibility test for uranium ores from Qianjiadian sandstone type uranium deposit

International Nuclear Information System (INIS)

Zhang Mingyu

2005-01-01

Sensibility tests for uranium ores from Qianjiadian sandstone type uranium deposit in Songliao Basin which is suitable to in-situ leach are carried out, including water sensibility, velocity sensibility, salt sensibility, acid sensibility and alkaline sensibility. The sensibility critical value of this ore is determined. Some references on mining process and technical parameter are provided for in-situ leaching of uranium. (authors)

Natural and depleted uranium in the topsoil of Qatar: Is it something to worry about?

International Nuclear Information System (INIS)

Shomar, Basem; Amr, Mohamed; Al-Saad, Khalid; Mohieldeen, Yasir

2013-01-01

Highlights: • Scientific studies on Uranium in the arid environment are almost absent. • Qatar is closed to Iraq and Iran where the two countries were exposed to long wars. • The paper introduces baseline study integrates chemistry, instrumentation and GIS mapping. • The study opens new horizons for similar studies on the field using similar approach. - Abstract: This study examines uranium in soils of Qatar to investigate whether there is any detectable traces of depleted uranium (DU). 409 soil samples were collected using a 10 km grid system throughout the State of Qatar. The U concentrations and isotopic compositions ( 235 U/ 238 U) were determined using an ICP-MS. The U concentrations range from 0.05 to 4.7 mg/kg and the 235 U/ 238 U isotopic signatures are in the range 0.007–0.008, i.e. comparable to the isotopic ratio in natural uranium (NU). The distribution of these concentrations in the topsoil were used to see correlations with locations of pollution point sources and environmentally hot areas associated with human activity: industrial estates, solid waste dumping sites, wastewater treatment plants, sea harbors, airports, and public transport network. New thematic maps were built using Geographic Information System (GIS) software. The results showed that there is no linkage between the occurrence, distribution, concentrations and isotopic ratios of U and these hotspots. More importantly, due to the low concentration of organic matter (OM) in soils of Qatar, very limited P-fertilization, the alkaline nature of soil (pH 8) and low Fe/Mn contents make soil uranium concentrations very low. The residential areas, including the capital Doha, had the lowest total concentrations of uranium and isotopic ratios of the country while the northern and western parts showed the highest values

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

Production of nanodiamonds by high-energy ion irradiation of graphite at room temperature

International Nuclear Information System (INIS)

Daulton, T.L.; Kirk, M.A.; Lewis, R.S.; Rehn, L.E.

2001-01-01

It has previously been shown that graphite can be transformed into diamond by MeV electron and ion irradiation at temperatures above approximately 600 deg. C. However, there exists geological evidence suggesting that carbonaceous materials can be transformed to diamond by irradiation at substantially lower temperatures. For example, submicron-size diamond aggregates have been found in uranium-rich, Precambrian carbonaceous deposits that never experienced high temperature or pressure. To test if diamonds can be formed at lower irradiation temperatures, sheets of fine-grain polycrystalline graphite were bombarded at 20 deg. C with 350±50 MeV Kr ions to fluences of 6x10 12 cm -2 using the Argonne tandem linear accelerator system (ATLAS). Ion-irradiated (and unirradiated control) graphite specimens were then subjected to acid dissolution treatments to remove untransformed graphite and isolate diamonds that were produced; these acid residues were subsequently characterized by high-resolution and analytical electron microscopy. The acid residue of the ion-irradiated graphite was found to contain nanodiamonds, demonstrating that ion irradiation of graphite at ambient temperature can produce diamond. The diamond yield under our irradiation conditions is low, ∼0.01 diamonds/ion. An important observation that emerges from comparing the present result with previous observations of diamond formation during irradiation is that nanodiamonds form under a surprisingly wide range of irradiation conditions. This propensity may be related to the very small difference in the graphite and diamond free-energies coupled with surface-energy considerations that may alter the relative stability of diamond and graphite at nanometer sizes

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-07-01

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

The biography of Uranium: from the Proto-solar cloud to the beginning of the oxygenic atmosphere

International Nuclear Information System (INIS)

Garzon Ruiperez, L.; Cavero Cavero, A.

2000-01-01

The geo-chemical properties of uranium and its materials have allowed us to consistently describe this element's characteristics in the evolution of matter from the proto-solar nebula to the formation and subsequent evolution of the Earth. The formation of the most primitive deposits is considered , and it is inferred that they were of a detrital nature. The ionizing radiations emitted by these deposits and the existence of critical episodes in them have been considered. The low concentration of O 2 until some 2.4 Ga ago was the reason why uranium deposits were not widespread and why their typology and the typology of their minerals were not very diversified. Uranium evolution, deposits, minerals, radiation, criticality. (Author)

The fluorimetry for control of internal contamination of exposed workers to natural and enriched uranium

International Nuclear Information System (INIS)

Gaburo, J.C.; Todo, A.S.; Sordi, G.M.A.A.

2000-01-01

This study is a part of bioassay program revision applied to the uranium processing plants at IPEN-CNEN/SP. The workers of these facilities handle both natural uranium and uranium compounds with different isotopic composition which could reach up to 20% in 235 U. The most commonly employed techniques for the determination of uranium in urine at IPEN are fluorimetry and alpha spectrometry with detection limit of 1.0 mgL-1. and 1,0 mBqL-1 , respectively. Based in advantages and disadvantages of each technique it is very important to identify the workers groups that should be submitted for these analysis. In this report a limiting value of uranium concentration in urine, mgL-1, obtained by fluorimetry is proposed. All the results greater than these limiting value indicate the necessity to carry out a additional measurement by alpha spectroscopy. The uranium mass that result in a pre-determined limit committed effective dose is function of isotopic composition. Consequently, the predicted value of the measured of urinary excretion is function of isotopic composition also and depends of absorption characteristics when inhaled and of the monitoring interval considered. In this report the uranium concentration values for reference levels and limits doses are determined. Based on these results the procedures to use the fluorimetry or both fluorimetry and alpha-spectrometry were adopted. (author)

Calculations on heavy-water moderated and cooled natural uranium fuelled power reactors

International Nuclear Information System (INIS)

Pinedo V, J.L.

1979-01-01

One of the codes that the Instituto Nacional de Investigaciones Nucleares (Mexico) has for the nuclear reactors design calculations is the LEOPARD code. This work studies the reliability of this code in reactors design calculations which component materials are the same of the heavy water moderated and cooled, natural uranium fuelled power reactors. (author)

Improving the Assessment of Internal Occupational Exposure to Natural Uranium from Urinalysis by Normalization to Creatinine

International Nuclear Information System (INIS)

Marko, R.; Kol, R.; Katorza, E.; German, U.; Balaish, Y.; Lorber, A.; Karpas, Z.

2002-01-01

The assessment of occupational internal exposure to natural uranium is normally carried out by combining Uranium Lung Detection (ULD) and urine analysis. The ULD is a direct measurement of the uranium content in lungs. The urine analysis measures the amount of uranium excreted from the body. The biokinetic models that are in use for dose assessments from urine analysis measurements are usually based on 24-hour urine collection. There are three traditional methods to collect urine samples: a) 24-hour collection - the subject is asked to collect all the urine excreted during a 24-hour period. b) Simulated 24-hour collection - the subject collects all the urine excreted during three consecutive 8-hour workdays. c) Spot samples - the subject gives a single urine sample at some time during work hours

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)

Recent developments in graphite. [Use in HTGR and aerospace

Energy Technology Data Exchange (ETDEWEB)

Cunningham, J.E.

1983-01-01

Overall, the HTGR graphite situation is in excellent shape. In both of the critical requirements, fuel blocks and support structures, adequate graphites are at hand and improved grades are sufficiently far along in truncation. In the aerospace field, GraphNOL N3M permits vehicle performance with confidence in trajectories unobtainable with any other existing material. For fusion energy applications, no other graphite can simultaneously withstand both extreme thermal shock and neutron damage. Hence, the material promises to create new markets as well as to offer a better candidate material for existing applications.

Geochemical behaviour of uranium in sedimentary formations: insights from a natural analogue study - 16340

International Nuclear Information System (INIS)

Noseck, Ulrich; Brasser, Thomas; Havlova, Vaclava; Cervinka, Radek; Suksi, Juhani

2009-01-01

Groundwater data from the natural analogue site Ruprechtov have been evaluated with special emphasis on the uranium behaviour in the so-called uranium-rich clay/lignite horizon. In this horizon in-situ Eh-values in the range of -160 to -280 mV seem to be determined by the SO 4 2- /HS - couple. Under these conditions U(IV) is expected to be the preferential redox state in solution. However, on-site measurements in groundwater from the clay/lignite horizon show only a fraction of about 20 % occurring in the reduced state U(IV). Thermodynamic calculations reveal that the high CO 2 partial pressure in the clay/lignite horizon can stabilise hexavalent uranium, which explains the occurrence of U(VI). The calculations also indicate that the low uranium concentrations in the range between 0.2 and 2.1 μg/l are controlled by amorphous UO 2 and/or the U(IV) phosphate mineral ningyoite. This confirms the findings from previous work that the uranium (IV) mineral phases are long-term stable under the reducing conditions in the clay/lignite horizon without any signatures for uranium mobilisation. It supports the current knowledge of the geological development of the site and is also another important indication for the long-term stability of the sedimentary system itself, namely of the reducing geochemical conditions in the near-surface (30 m to 60 m deep) clay/lignite horizon. Further work with respect to the impact of changes in redox conditions on the uranium speciation is on the way. (authors)

Metabolomics identifies a biological response to chronic low-dose natural uranium contamination in urine samples.

Science.gov (United States)

Grison, Stéphane; Favé, Gaëlle; Maillot, Matthieu; Manens, Line; Delissen, Olivia; Blanchardon, Eric; Banzet, Nathalie; Defoort, Catherine; Bott, Romain; Dublineau, Isabelle; Aigueperse, Jocelyne; Gourmelon, Patrick; Martin, Jean-Charles; Souidi, Maâmar

2013-01-01

Because uranium is a natural element present in the earth's crust, the population may be chronically exposed to low doses of it through drinking water. Additionally, the military and civil uses of uranium can also lead to environmental dispersion that can result in high or low doses of acute or chronic exposure. Recent experimental data suggest this might lead to relatively innocuous biological reactions. The aim of this study was to assess the biological changes in rats caused by ingestion of natural uranium in drinking water with a mean daily intake of 2.7 mg/kg for 9 months and to identify potential biomarkers related to such a contamination. Subsequently, we observed no pathology and standard clinical tests were unable to distinguish between treated and untreated animals. Conversely, LC-MS metabolomics identified urine as an appropriate biofluid for discriminating the experimental groups. Of the 1,376 features detected in urine, the most discriminant were metabolites involved in tryptophan, nicotinate, and nicotinamide metabolic pathways. In particular, N -methylnicotinamide, which was found at a level seven times higher in untreated than in contaminated rats, had the greatest discriminating power. These novel results establish a proof of principle for using metabolomics to address chronic low-dose uranium contamination. They open interesting perspectives for understanding the underlying biological mechanisms and designing a diagnostic test of exposure.

Method to Assess the Radionuclide Inventory of Irradiated Graphite from Gas-Cooled Reactors - 13072

Energy Technology Data Exchange (ETDEWEB)

Poncet, Bernard [EDF-CIDEN, 154 Avenue Thiers, CS 60018, F-69458 LYON cedex 06 (France)

2013-07-01

About 17,000 t of irradiated graphite waste will be produced from the decommissioning of the six French gas-cooled nuclear reactors. Determining the radionuclide (RN) content of this waste is of relevant importance for safety reasons and in order to determine the best way to manage them. For many reasons the impurity content that gave rise to the RNs in irradiated graphite by neutron activation during operation is not always well known and sometimes actually unknown. So, assessing the RN content by the use of traditional calculation activation, starting from assumed impurity content, leads to a false assessment. Moreover, radiochemical measurements exhibit very wide discrepancies especially on RN corresponding to precursor at the trace level such as natural chlorine corresponding to chlorine 36. This wide discrepancy is unavoidable and is due to very simple reasons. The level of impurity is very low because the uranium fuel used at that very moment was not enriched, so it was a necessity to have very pure nuclear grade graphite and the very low size of radiochemical sample is a simple technical constraint because device size used to get mineralization product for measurement purpose is limited. The assessment of a radionuclide inventory only based on few number of radiochemical measurements lead in most cases, to a gross over or under-estimation that is detrimental for graphite waste management. A method using an identification calculation-measurement process is proposed in order to assess a radiological inventory for disposal sizing purpose as precise as possible while guaranteeing its upper character. This method present a closer approach to the reality of the main phenomenon at the origin of RNs in a reactor, while also incorporating the secondary effects that can alter this result such as RN (or its precursor) release during reactor operation. (authors)

Preparation of graphite derivatives by selective reduction of graphite oxide and isocyanate functionalization

Energy Technology Data Exchange (ETDEWEB)

Santha Kumar, Arunjunai Raja Shankar [Materials Science Centre, Indian Institute of Technology, Kharagpur, 721302, West Bengal (India); Leibniz-Institut für Polymerforschung Dresden e.V., Hohe Straße 6, 01069, Dresden (Germany); Piana, Francesco [Leibniz-Institut für Polymerforschung Dresden e.V., Hohe Straße 6, 01069, Dresden (Germany); Organic Chemistry of Polymers, Technische Universität Dresden, 01062, Dresden (Germany); Mičušík, Matej [Polymer Institute, Slovak Academy of Sciences, Dúbravská cesta 9, 845 41, Bratislava (Slovakia); Pionteck, Jürgen, E-mail: pionteck@ipfdd.de [Leibniz-Institut für Polymerforschung Dresden e.V., Hohe Straße 6, 01069, Dresden (Germany); Banerjee, Susanta [Materials Science Centre, Indian Institute of Technology, Kharagpur, 721302, West Bengal (India); Voit, Brigitte [Leibniz-Institut für Polymerforschung Dresden e.V., Hohe Straße 6, 01069, Dresden (Germany); Organic Chemistry of Polymers, Technische Universität Dresden, 01062, Dresden (Germany)

2016-10-01

Heavily oxidized and ordered graphene nanoplatelets were produced from natural graphite by oxidation using a mixture of phosphoric acid, sulphuric acid, and potassium permanganate (Marcano's method). The atomic percentage of oxygen in the graphite oxide produced was more than 30% confirmed by XPS studies. The graphite oxide produced had intact basal planes and remains in a layered structure with interlayer distance of 0.8 nm, analyzed by WAXS. The graphite oxide was treated with 4,4′-methylenebis(phenyl isocyanate) (MDI) to produce grafted isocyanate functionalization. Introduction of these bulky functional groups widens the interlayer distance to 1.3 nm. In addition, two reduction methods, namely benzyl alcohol mediated reduction and thermal reduction were carried out on isocyanate modified and unmodified graphite oxides and compared to each other. The decrease in the oxygen content and the sp{sup 3} defect-repair were studied with XPS and RAMAN spectroscopy. Compared to the thermal reduction process, which is connected with large material loss, the benzyl alcohol mediated reduction process is highly effective in defect repair. This resulted in an increase of conductivity of at least 9 orders of magnitude compared to the graphite oxide. - Highlights: • Preparation of GO by Marcano's method results in defined interlayer spacing. • Treatment of GO with diisocyanate widens the interlayer spacing to 1.3 nm. • Chemical reduction of GO with benzyl alcohol is effective in defect repair. • Electrical conductivity increases by 9 orders of magnitude during chemical reduction. • The isocyanate functionalization is stable under chemical reducing conditions.

Bremsstrahlung doses from natural uranium ingots

International Nuclear Information System (INIS)

Anderson, J. L.; Hertel, N. E.

2005-01-01

In the past, some privately owned commercial facilities in the United States were involved in producing or processing radioactive materials used in the production of atomic weapons. Seven different geometrical objects, representative of the configurations of natural uranium metal potentially encountered by workers at these facilities, are modelled to determine gamma ray and Bremsstrahlung dose rates. The dose rates are calculated using the MCNP5 code and also by using the MICROSHIELD point-kernel code. Both gamma ray and Bremsstrahlung dose rates are calculated and combined to obtain a total dose rate. The two methods were found to be in good agreement despite differences in modelling assumptions and method differences. Computed total dose rates on the surface of these objects ranged from ∼51-84 μSv h -1 and 17-95 μSv h -1 using the MCNP5 and the MICROSHIELD modeling, respectively. The partitioning of the computed dose rates between gamma rays and Bremsstrahlung were the same order of magnitude for each object. (authors)

Bremsstrahlung doses from natural uranium ingots.

Science.gov (United States)

Anderson, Jeri L; Hertel, Nolan E

2005-01-01

In the past, some privately owned commercial facilities in the United States were involved in producing or processing radioactive materials used in the production of atomic weapons. Seven different geometrical objects, representative of the configurations of natural uranium metal potentially encountered by workers at these facilities, are modelled to determine gamma ray and bremsstrahlung dose rates. The dose rates are calculated using the MCNP5 code and also by using the MICROSHIELD point-kernel code. Both gamma ray and bremsstrahlung dose rates are calculated and combined to obtain a total dose rate. The two methods were found to be in good agreement despite differences in modelling assumptions and method differences. Computed total dose rates on the surface of these objects ranged from approximately 51-84 microSv h(-1) and 17-95 microSv h(-1) using the MCNP5 and the MICROSHIELD modeling, respectively. The partitioning of the computed dose rates between gamma rays and bremsstrahlung were the same order of magnitude for each object.

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

International Nuclear Information System (INIS)

Huang Chongyuan; Li Weicai; Zhang Yutai; Gao Xizhen

2000-01-01

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

Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

International Nuclear Information System (INIS)

Anfimov, S.S.

2000-01-01

As a result of decommissioning of water-cooled graphite-moderated reactors, a large amount of rad-waste in the form of graphite stack fragments is generated (on average 1500-2000 tons per reactor). That is why it is essentially important, although complex from the technical point of view, to develop advanced technologies based on up-to-date remotely-controlled systems for unmanned dismantling of the graphite stack containing highly-active long-lived radionuclides and for conditioning of irradiated graphite (IG) for the purposes of transportation and subsequent long term and ecologically safe storage either on NPP sites or in special-purpose geological repositories. The main characteristics critical for radiation and nuclear hazards of the graphite stack are as follows: the graphite stack is contaminated with nuclear fuel that has gotten there as a result of the accidents; the graphite mass is 992 tons, total activity -6?104 Ci (at the time of unit shutdown); the fuel mass in the reactor stack amounts to 100-140 kg, as estimated by IPPE and RDIPE, respectively; γ-radiation dose rate in the stack cells varies from 4 to 4300 R/h, with the prevailing values being in the range from 50 to 100 R/h. In this paper the traditional methods of rad-waste handling as bituminization technology, cementing technology are discussed. In terms of IG handling technology two lines were identified: long-term storage of conditioned IG and IG disposal by means of incineration. The specific cost of graphite immobilization in a radiation-resistant polymeric matrix amounts to -2600 USD per 1 t of graphite, whereas the specific cost of immobilization in slag-stone containers with an inorganic binder (cement) is -1400 USD per 1 t of graphite. On the other hand, volume of conditioned IG rad-waste subject for disposal, if obtained by means of the first technology, is 2-2.5 times less than the volume of rad-waste generated by means of the second technology. It can be concluded from the above that

Uranium tailings sampling manual

International Nuclear Information System (INIS)

Feenstra, S.; Reades, D.W.; Cherry, J.A.; Chambers, D.B.; Case, G.G.; Ibbotson, B.G.

1985-01-01

The purpose of this manual is to describe the requisite sampling procedures for the application of uniform high-quality standards to detailed geotechnical, hydrogeological, geochemical and air quality measurements at Canadian uranium tailings disposal sites. The selection and implementation of applicable sampling procedures for such measurements at uranium tailings disposal sites are complicated by two primary factors. Firstly, the physical and chemical nature of uranium mine tailings and effluent is considerably different from natural soil materials and natural waters. Consequently, many conventional methods for the collection and analysis of natural soils and waters are not directly applicable to tailings. Secondly, there is a wide range in the physical and chemical nature of uranium tailings. The composition of the ore, the milling process, the nature of tailings depositon, and effluent treatment vary considerably and are highly site-specific. Therefore, the definition and implementation of sampling programs for uranium tailings disposal sites require considerable evaluation, and often innovation, to ensure that appropriate sampling and analysis methods are used which provide the flexibility to take into account site-specific considerations. The following chapters describe the objective and scope of a sampling program, preliminary data collection, and the procedures for sampling of tailings solids, surface water and seepage, tailings pore-water, and wind-blown dust and radon

Critical review of uranium resources and production capability to 2020

International Nuclear Information System (INIS)

Underhill, D.H.

2002-01-01

Even with a modest forecast of nuclear power growth for the next 25 years, it is expected that the world uranium requirements will increase. This analysis indicates uranium mine production will continue to be the primary supply of requirements through 2020. Secondary supplies, such as low enriched uranium blended from highly enriched uranium, reprocessing of spent fuel would have to make-up the remaining balance, although the contribution of US and Russian strategic stockpiles is not well known at this time. (author)

Natural Radioactivity of Intrusive-Metamorphic and Sedimentary Rocks of the Balkan Mountain Range (Serbia, Stara Planina

Directory of Open Access Journals (Sweden)

Sanna Masod Abdulqader

2017-12-01

Full Text Available Stara Planina (also known as the Balkan mountain range is known for numerous occurrences and deposits of uranium and associated radionuclides. It is also famous for its geodiversity. The geologic framework is highly complex. The mountain is situated between the latitudes of 43° and 44° N and the longitudes from 22°16′ to 23°00′ E. Uranium exploration and radioactivity testing on Stara Planina began back in 1948. Uranium has also been mined in the zone of Kalna, within the Janja granite intrusive. The naturally radioactive geologic units of Stara Planina are presented in detail in this paper. The main sources of radioactivity on Stara Planina can be classified as: 1. Granitic endogenous—syngenetic–epigenetic deposits and occurrences; 2. Metamorphogenic—syngenetic; and 3. Sedimentary, including occurrences of uranium deposition and fluctuation caused by water in different types of sedimentary rocks formed in a continental setting, which could be classified under epigenetic types. The area of Stara Planina with increased radioactivity (higher than 200 cps, measured by airborne gamma spectrometry, is about 380 square kilometers. The highest values of measured radioactivity and uranium grade were obtained from a sample taken from the Mezdreja uranium mine tailing dump, where 226Ra measures 2600 ± 100 Bq/kg and the uranium grade is from 76.54 to 77.65 ppm U. The highest uranium (and lead concentration, among all samples, is measured in graphitic schist with high concentrations of organic (graphitic material from the Inovska Series—99.47 ppm U and 107.69 ppm Pb. Thorium related radioactivity is the highest in granite samples from the Janja granite in the vicinity of the Mezdreja granite mine and the Gabrovnica granite mine tailing dump, and it is the same—250 ± 10 Bq/kg for 232Th, while the thorium grade varies from 30.82 to 60.27 ppm Th. In gray siltstones with a small amount of organic material, the highest radioactivity is

Data feature World natural Uranium production 1992

International Nuclear Information System (INIS)

Anon.

1993-01-01

NUKEM estimates that world uranium production fell more than 13% last year, from 40,729 tonnes U [106 million lbs U308] in 1991 to 35,363 tonnes U [92 million lbs U308] in 1992. Production fell in both the Western World and non-Western World. How much of demand was met by production? World uranium production in 1992 amounted to about 65% of reactor consumption. That's assuming that reactor demand of the non-Western World has not changed much from the Uranium Institute's estimate for 1991. Civilian stockpiles are being drawn down on a massive scale while the world waits to see what will become of the military stockpiles that could soon enter the global supply picture

Influence of uranium and thorium in the natural radioactivity in shales from the middle Amazon river region

International Nuclear Information System (INIS)

Ferro, A.L.

1982-02-01

The feasibility of using the fission track registration technique in the determination of the uranium and thorium content in shales from the middle Amazon river region is studied. The above technique permits, through the determination of the uranium concentration, to establish a correlation between the uranium content and the organic matter present in the shale. In establishing the ratio between the fission tracks due to 238 U and 235 U, the sample was contaminated with natural uranium and analized, so that no modifications on the analysis conditions might change or distort the results. The experimental results were satisfactory and they may contribute to the study of the industrial exploration of these energy sources as well as to the analysis of problems related to environmental control. (Author) [pt

A comparative study between transport and criticality safety indexes for fissile uranium nuclearly pure

Energy Technology Data Exchange (ETDEWEB)

Moraes da Silva, T. de; Sordi, G.M.A.A. [Instituto de Pesquisas Energeticas e Nucleares, IPEN/CNEN (Brazil)]. e-mail: tmsilva@ipen.br

2006-07-01

The international and national standards determine that during the transport of radioactive materials the package to be sent should be identified by labels of risks specifying content, activity and the transport index. The result of the monitoring of the package to 1 meter identifies the transport index, TI, which represents the dose rate to 1 meter of this. The transport index is, by definition, a number that represents a gamma radiation that crosses the superficial layer the radioactive material of the package to 1 meter of distance. For the fissile radioactive material that is the one in which a neutron causes the division of the atom, the international standards specify criticality safety index CSI, which is related with the safe mass of the fissile element. In this work it was determined the respective safe mass for each considered enrichment for the compounds of uranium oxides UO{sub 2}, U{sub 3}O{sub 8} and U{sub 3}Si{sub 2}. In the study of CSI it was observed that the value 50 of the expression 50/N being N the number of packages be transported in subcriticality conditions it represents a fifth part of the safe mass of the element uranium or 9% of the smallest mass critical for a transport not under exclusive use. As conclusion of the accomplished study was observed that the transport index starting from 7% of enrichment doesn't present contribution and that criticality safety index is always greater than the transport index. Therefore what the standards demand to specify, the largest value between both indexes, was clearly identified in this study as being the criticality safety index. (Author)

From a critical assembly heavy water - natural uranium to the fast - thermal research reactor in the Institute Vinca; Od kriticnog sistema teska voda - prirodni uranium do brzo - termickog istrazivackog reaktora u Vinci

Energy Technology Data Exchange (ETDEWEB)

Stefanovic, D; Pesic, M [Vinca Institute of Nuclear Sciences, Beograd (Yugoslavia)

1995-07-01

A part of the Institute in Vinca this monograph refers to is the thermal nuclear zero power reactor RB, with a heavy water moderator and variously enriched uranium fuel, that is, its present day version, the coupled fast-thermal system HERBE. A group of research workers, technicians, operators and skilled workmen in the workshop have worked continuously on it. Some of them have spent their whole working age at the reactor, and some a part of it. There is about a hundred and fifty internationally published papers, twenty master's and fourteen doctor's theses left behind them for the past thirty five years. This book is devoted to them. The first part of the text refers to the pioneering efforts on the reactor and fundamental research in reactor physics. The experimental reactor RB was designed and constructed at the time to operate with natural uranium and heavy water. Measurements are presented and the first results of reaching critical state, measurements of migration length of thermal neutrons and neutron multiplication factor in an infinite medium; also measurements of neutron flux density distribution and reactor parameter, and in the domain of safety, measurement of safety rods reactivity. Those were also the times when the known serious accident occurred with the uncontrolled rise of reactivity, which was especially minutely described in a publication of the International Atomic Energy Agency from Vienna. Later on, new fuel was acquired with 2 % enriched uranium. A series of experiments in reactor and neutron physics followed, with just the most interesting results of them presented here. In the period which followed, another type of fuel was available, with 80 % enriched uranium. New possibilities for work opened. Measurements with mixed lattices were performed, and the RA reactor lattices were simulated. After measurements mainly in the sphere of reactor and neutron physics, a need for investigations in the field of gamma and neutron radiation protection

Method for producing dustless graphite spheres from waste graphite fines

Science.gov (United States)

Pappano, Peter J [Oak Ridge, TN; Rogers, Michael R [Clinton, TN

2012-05-08

A method for producing graphite spheres from graphite fines by charging a quantity of spherical media into a rotatable cylindrical overcoater, charging a quantity of graphite fines into the overcoater thereby forming a first mixture of spherical media and graphite fines, rotating the overcoater at a speed such that the first mixture climbs the wall of the overcoater before rolling back down to the bottom thereby forming a second mixture of spherical media, graphite fines, and graphite spheres, removing the second mixture from the overcoater, sieving the second mixture to separate graphite spheres, charging the first mixture back into the overcoater, charging an additional quantity of graphite fines into the overcoater, adjusting processing parameters like overcoater dimensions, graphite fines charge, overcoater rotation speed, overcoater angle of rotation, and overcoater time of rotation, before repeating the steps until graphite fines are converted to graphite spheres.

A study on criticality of coupled fast-thermal core HERBE at RB reactor

Energy Technology Data Exchange (ETDEWEB)

Pesic, M; Zavaljevski, M; Milosevic, M; Stefanovic, D; Nikolic, D; Avdic, S [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia); Popovic, D; Marinkovic, P [Faculty of Electrical Engineering, Beograd (Yugoslavia)

1991-07-01

The coupled fast-thermal core HERBE at the RB zero power heavy water reactor in Vinca was designed with the aim of improving the experimental possibilities in fast neutron fields. The requirements for minimum modifications in the RB construction and the use available fuel, restricted design flexibility of the coupled system. The following core is considered optimal in the light of the foregoing constraints: the central fast core of natural uranium is surrounded by a neutron filter zone (cadmium and natural uranium) and a converter zone (enriched uranium fuel, without moderator). The coupling region is heavy water. The thermal core in the form of the RB heavy water 80% enriched uranium lattice with 12 cm pitch. The criticality of the system is obtained by adjusting the moderator level. The critical heavy water levels were measured for normal reactor operation and some simulated accidental conditions. These data were analyzed by a computer code for the design of thermal and coupled fast-thermal reactor recently developed in IBK Nuclear Engineering Laboratory. Good agreement between the computations and experimental data was achieved. (author)

A study on criticality of coupled fast-thermal core HERBE at RB reactor

International Nuclear Information System (INIS)

Pesic, M.; Zavaljevski, M.; Milosevic, M.; Stefanovic, D.; Nikolic, D.; Avdic, S.; Popovic, D.; Marinkovic, P.

1991-01-01

The coupled fast-thermal core HERBE at the RB zero power heavy water reactor in Vinca was designed with the aim of improving the experimental possibilities in fast neutron fields. The requirements for minimum modifications in the RB construction and the use available fuel, restricted design flexibility of the coupled system. The following core is considered optimal in the light of the foregoing constraints: the central fast core of natural uranium is surrounded by a neutron filter zone (cadmium and natural uranium) and a converter zone (enriched uranium fuel, without moderator). The coupling region is heavy water. The thermal core in the form of the RB heavy water 80% enriched uranium lattice with 12 cm pitch. The criticality of the system is obtained by adjusting the moderator level. The critical heavy water levels were measured for normal reactor operation and some simulated accidental conditions. These data were analyzed by a computer code for the design of thermal and coupled fast-thermal reactor recently developed in IBK Nuclear Engineering Laboratory. Good agreement between the computations and experimental data was achieved. (author)

Kinetics of the chain uranium decay

International Nuclear Information System (INIS)

Zel'dovich, Ya.B.; Khariton, Yu.B.

1984-01-01

The development of chain reaction in uranium mass, when passing through the mass critical value, is studied. It is shown that thermal expanson is a powerful regulating factor, making the limit passing quite safe. The case of convergence of two uranium masses of precritical dimensions is considered in particular. The critical mass value is assessed and it is shown that its gradual increase above the critical value results in the reaction vibrational regime with the period inversely proportional to the square root of uranium supply rate. The role of delaying neutrons in the fission chain reaction development is pointed out

Measurement of fast assembly spectra using time-of-flight method

International Nuclear Information System (INIS)

Duquesne, Henry; Rotival, Michel; Schmitt, Andre; Allard, Christian; De Keyser, Albert; Hortsmann, Henri

1975-07-01

Measurement of neutron spectra made in fast subcritical assemblies HUG 3 and PHUG 3 (uranium-graphite and plutonium-graphite) utilizing time-of-flight techniques are described. The matrix were excited by the pulsed neutron source from the BCMN Linac beam impinging on a target of natural uranium. Details of the experimental procedure, safety studies, detector calibration and data reduction are given [fr

Natural radioactivity around a prospected uranium mining area in Finnish Lapland

International Nuclear Information System (INIS)

Rissanen, K.

1983-01-01

An environmental survey of natural radionuclides was carried out around the Pahtavuoma uranium occurrence site at Kittilae in Finnish Lapland. The aim of the survey was to determine the background levels of these nuclides in the terrestrial and aquatic ecosystems before changing the natural conditions by mining. All of the samples collected were analyzed for Ra-226 after radiochemical separation. Low Ra-226-content, < 0.02 - 1.9 Bg/kg d.w., was measured in locally produced foodstuffs, reindeer, elk and fish, cloudberry and blueberry; levels were 1.4 - 4.6 Bq/kg d.w. in cowberry. Contents of 0.3 - 5 Bq/kg were found in lichen, beard lichen, hay and fish bones, and higher concentrations in elk and reindeer bones (20 - 62 Bq/kg), aquatic plants Hippuris vulgaris (11 - 90 Bq/kg), and sediments (7 - 130 Bq/kg). The highest Ra-226 concentrations (110 - 3100 Bq/kg) were measured in aquatic mosses (Fontinalis sp). The Rn-222 and Ra-226-concentrations measured in surface and well waters were not higher than the average for Finland. Po-210 and Pb-210 determinations are in process. Dose rate and spectroscopic in situ measurements were performed as well. The results indicate lower environmental activity than the average for Lapland, except at the actual uranium mining site

Natural uranium toxicology - evaluation of internal contamination in man; Toxicologie de l'uranium naturel - essai d'evaluation de la contamination interne chez l'homme

Energy Technology Data Exchange (ETDEWEB)

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

1968-07-01

After reminding the physical and chemical properties of natural uranium which might affect its toxicology, a comprehensive investigation upon natural uranium metabolism and toxicity and after applying occupational exposure standards to this particular poison, it has been determined, from accident reports and human experience reported in the related literature, a series of formulae obtained by theoretical mathematical development giving principles for internal contamination monitoring and disclosure by determining uranium in the urine of occupationally exposed individuals. An assay is performed to determine individual internal contamination according to the various contamination cases. The outlined purposes, mainly practical, required some options and extrapolations. The proposed formula allows a preliminary approach and also to determine shortly a contamination extent or to discuss the systematical urinalysis results as compared with individual radio-toxicology monitoring professional standards. (author) [French] Apres le rappel des caracteristiques physiques et des proprietes chimiques de l'uranium naturel pouvant avoir une influence sur sa toxicologie, l'etude detaillee de son metabolisme et de sa toxicite, puis l'application des normes professionnelles d'exposition au cas particulier de ce toxique, il est etabli, a partir des comptes rendus d'accidents et de l'experimentation humaine rapportes dans la litterature, une serie de formules obtenues par developpement mathematique theorique qui posent les principes de la surveillance et de la mise en evidence de la contamination interne par la recherche et le dosage de l'uranium dans les urines d'individus professionnellement exposes. Un essai d'evaluation de la contamination interne individuelle suivant les differents cas de contamination est effectue. Le formulaire propose permet de faire une premiere approximation et d'apprecier rapidement l'importance d'une contamination ou bien d'interpreter les resultats d

Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

Energy Technology Data Exchange (ETDEWEB)

Luther, Erik; Rooyen, Isabella van; Leckie, Rafael; Papin, Pallas; Nelson, Andrew; Hunter, James

2015-03-01

In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabrication must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.

Application of 3D coupled code ATHLET-QUABOX/CUBBOX for RBMK-1000 transients after graphite block modernization

Energy Technology Data Exchange (ETDEWEB)

Samokhin, Aleksei [Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS), Moscow (Russian Federation); Zilly, Matias [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

2016-11-15

This work describes the application and the results of transient calculations for the RBMK-1000 with the coupled code system ATHLET 2.2A-QUABOX/CUBBOX which was developed in GRS. Within these studies the planned modernization of the graphite blocks of the RBMK-1000 reactor is taken into account. During the long-term operation of the uranium-graphite reactors RBMK-1000, a change of physical and mechanical properties of the reactor graphite blocks is observed due to the impact of radiation and temperature effects. These have led to a deformation of the reactor graphite columns and, as a result, a deformation of the control and protection system (CPS) and of fuel channels. Potentially, this deformation can lead to problems affecting the smooth movement of the control rods in the CPS channels and problems during the loading and unloading of fuel assemblies. The present paper analyzes two reactivity insertion transients, each taking into account three graphite removal scenarios. The presented work is directly connected with the modernization program of the RBMK- 1000 reactors and has an important contribution to the assessment of the safety-relevant parameters after the modification of the core graphite blocks.

Structural disorder of graphite and implications for graphite thermometry

Science.gov (United States)

Kirilova, Martina; Toy, Virginia; Rooney, Jeremy S.; Giorgetti, Carolina; Gordon, Keith C.; Collettini, Cristiano; Takeshita, Toru

2018-02-01

Graphitization, or the progressive maturation of carbonaceous material, is considered an irreversible process. Thus, the degree of graphite crystallinity, or its structural order, has been calibrated as an indicator of the peak metamorphic temperatures experienced by the host rocks. However, discrepancies between temperatures indicated by graphite crystallinity versus other thermometers have been documented in deformed rocks. To examine the possibility of mechanical modifications of graphite structure and the potential impacts on graphite thermometry, we performed laboratory deformation experiments. We sheared highly crystalline graphite powder at normal stresses of 5 and 25 megapascal (MPa) and aseismic velocities of 1, 10 and 100 µm s-1. The degree of structural order both in the starting and resulting materials was analyzed by Raman microspectroscopy. Our results demonstrate structural disorder of graphite, manifested as changes in the Raman spectra. Microstructural observations show that brittle processes caused the documented mechanical modifications of the aggregate graphite crystallinity. We conclude that the calibrated graphite thermometer is ambiguous in active tectonic settings.

Asymptomatic Intracorneal Graphite Deposits following Graphite Pencil Injury

OpenAIRE

Philip, Swetha Sara; John, Deepa; John, Sheeja Susan

2012-01-01

Reports of graphite pencil lead injuries to the eye are rare. Although graphite is considered to remain inert in the eye, it has been known to cause severe inflammation and damage to ocular structures. We report a case of a 12-year-old girl with intracorneal graphite foreign bodies following a graphite pencil injury.

Evaluation of activities of carbons in chemical equilibrium with uranium carbonitride

International Nuclear Information System (INIS)

Katsura, Masahiro; Hirota, Masayuki; Miyake, Masanobu; Hamada, Kazuo.

1992-01-01

A mixture of uranium sesquinitride and carbon was prepared by the reaction of UC of UC 2 with N 2 in the temperature range from 700 to 1400degC. When the mixture of uranium sesquinitride and carbon is kept at temperatures above 1200degC in the atmosphere of N 2 at low pressure, the state where uranium carbonitride (UC 1-x N x ) and carbon are present together in chemical equilibrium will be established. A thermodynamic analysis suggests that, in the equilibrium state, the composition of UC 1-x N x is determined by the chemical activity of carbon, a c , which is related to the chemical potential of the carbon, μ c , by the equation, μ c = μ c deg + RT 1n a c . Here μ c deg refers to graphite, which is usually taken as the standard state of carbon (a c = 1). Mixtures of U 2 N 3 and carbon with several degrees of graphitization were heat-treated at 1400degC, and the composition of UC 1-x N x in the reaction product was determined. From these experimental results and the thermodynamic analysis, values of the activity of the carbon coexisting with UC 1-x N x were estimated. (author)

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.

A Summary Report on Assembly 3 of FR0

International Nuclear Information System (INIS)

Andersson, T.L.; Brunfelter, B.; Cecchi, P.F.; Hellstrand, E.; Kockum, J.; Londen, S.O.; Tiren, L.I.

1966-06-01

The third core of the zero energy fast reactor FR0 consisted of 20 % enriched uranium diluted to 29 vol. % with graphite and had a volume of 50 litres. Like previous cores it was surrounded by a thick copper reflector. The report summarizes measurements of critical mass, control rod reactivities, fine structure flux variations and conversion ratio. In particular, effects associated with the heterogeneous arrangement of the uranium and graphite plates are examined

A Summary Report on Assembly 3 of FR0

Energy Technology Data Exchange (ETDEWEB)

Andersson, T L; Brunfelter, B; Cecchi, P F; Hellstrand, E; Kockum, J; Londen, S O; Tiren, L I

1966-06-15

The third core of the zero energy fast reactor FR0 consisted of 20 % enriched uranium diluted to 29 vol. % with graphite and had a volume of 50 litres. Like previous cores it was surrounded by a thick copper reflector. The report summarizes measurements of critical mass, control rod reactivities, fine structure flux variations and conversion ratio. In particular, effects associated with the heterogeneous arrangement of the uranium and graphite plates are examined.

Long term developments in irradiated natural uranium processing costs. Optimal size and siting of plants

International Nuclear Information System (INIS)

Thiriet, L.

1964-01-01

The aim of this paper is to help solve the problem of the selection of optimal sizes and sites for spent nuclear fuel processing plants associated with power capacity programmes already installed. Firstly, the structure of capital and running costs of irradiated natural uranium processing plants is studied, as well as the influence of plant sizes on these costs and structures. Shipping costs from the production site to the plant must also be added to processing costs. An attempt to reach a minimum cost for the production of a country or a group of countries must therefore take into account both the size and the location of the plants. The foreseeable shipping costs and their structure (freight, insurance, container cost and depreciation), for spent natural uranium are indicated. Secondly, for various annual spent fuel reprocessing programmes, the optimal sizes and locations of the plants are determined. The sensitivity of the results to the basic assumptions relative to processing costs, shipping costs, the starting up year of the plant programme and the length of period considered, is also tested. - this rather complex problem, of a combinative nature, is solved through dynamic programming methods. - It is shown that these methods can also be applied to the problem of selecting the optimal sizes and locations of processing plants for MTR type fuel elements, related to research reactor programmes, as well as to future plutonium element processing plants related to breeder reactors. Thirdly, the case where yearly extraction of the plutonium contained in the irradiated natural uranium is not compulsory is examined; some stockpiling of the fuel is then allowed some years, entailing delayed processing. The load factor of such plants is thus greatly improved with respect to that of plants where the annual plutonium demand is strictly satisfied. By including spent natural uranium stockpiling costs an optimal rhythm of introduction and optimal sizes for spent fuel

Status of the natural and enriched uranium market: the basic economical factor for the development of the fuel cycle

International Nuclear Information System (INIS)

Nochev, T.

1999-01-01

Status of the Natural and Enriched Uranium Market - the Basic. Economical Factor for the Development of the Fuel Cycle An overview of the status of the natural and enriched uranium market has been performed and it offers a possibility to estimate the changes and tendencies, the knowledge of which is needed in negotiations about the fresh fuel. The simplified financial analysis presented here demonstrates the economical profitability of the storage of the spent fuel making now the allocations for the future reprocessing

Uranium mineral - groundwater equilibrium at the Palmottu natural analogue study site, Finland

International Nuclear Information System (INIS)

Ahonen, L.; Ruskeeniemi, T.; Blomqvist, R.; Ervanne, H.; Jaakkola, T.

1993-01-01

The redox-potential, pH, chemical composition of fracture waters, and uraninite alteration associated with the Palmottu uranium mineralization (a natural analogue study site for radioactive waste disposal in southwestern Finland), have been studied. The data have been interpreted by means of thermodynamic calculations. The results indicate equilibrium between uraninite, ferric hydroxide and groundwater in the bedrock of the study site. Partially oxidized uraninite (UO 2 .33) and ferric hydroxide are in equilibrium with fresh, slightly acidic and oxidized water type, while primary uraninite is stable with deeper waters that have a higher pH and lower Eh. Measured Eh-pH values of groundwater cluster within a relatively narrow range indicating buffering by heterogenous redox-processes. A good consistency between measured Eh and analyzed uranium oxidation states was observed

Radioactivity and the French uranium bearing minerals

International Nuclear Information System (INIS)

Guiollard, P.Ch.; Boisson, J.M.; Leydet, J.C.; Meisser, N.

1998-01-01

This special issue of Regne Mineral journal is entirely devoted to the French uranium mining industry. It comprises 4 parts dealing with: the uranium mining industry in France (history, uranium rush, deposits, geologic setting, prosperity and recession, situation in 1998, ore processing); radioactivity and the uranium and its descendants (discovery, first French uranium bearing ores, discovery of radioactivity, radium and other uranium descendants, radium mines, uranium mines, atoms, elements and isotopes, uranium genesis, uranium decay, isotopes in an uranium ore, spontaneous fission, selective migration of radionuclides, radon in mines and houses, radioactivity units, radioprotection standards, new standards and controversies, natural and artificial radioactivity, hazards linked with the handling and collecting of uranium ores, conformability with radioprotection standards, radioactivity of natural uranium minerals); the French uranium bearing minerals (composition, crystal structure, reference, etymology, fluorescence). (J.S.)

Uranium industry in the USSR

International Nuclear Information System (INIS)

Nikipelov, B.V.; Chernov, A.G.

1990-01-01

A brief historical account of the Soviet production of natural and enriched uranium is given. The geological and geographical location of major uranium deposits are mentioned. The processing of natural ores including in-situ leaching (ISL) is also briefly described. Gas centrifuges play a large part in uranium enrichment. The role of Techsnabexport for the export of nuclear materials is explained

From USA operation experience of industrial uranium-graphite reactors

International Nuclear Information System (INIS)

Burdakov, N.S.

1996-01-01

The review on materials, presented by a group of the USA specialists at the seminar in Moscow on October 9-11, 1995 is considered. The above specialists shared their experience in operation of the Hanford industrial reactors, aimed at plutonium production for atomic bombs. The purpose of the above visit consisted in providing assistance to the Russian specialists by evaluation and modernization of operational conditions safety improvement of the RBMK type reactors. Special attention is paid to the behaviour of the graphite lining and channel tubes with an account of possible channel power interaction with the reactor structural units. The information on the experience of the Hanford reactor operation may be useful for specialists, operating the RBMK type reactors

The dose exposure of the environmental population by natural and released Ra-226 from an uranium mine prospect

International Nuclear Information System (INIS)

Schuettelkopf, H.; Kiefer, H.

1980-08-01

The concentration of natural Ra-226 in the environment of Baden-Baden was determined in samples of drinking water, surface water, sediments, soil, fish, milk and other food. The results partly are higher than normal caused by a local uranium deposit. Ra-226 releases with waste water from an uranium mine prospect are measured. The Ra-226 concentrations in creeks and sediments partly caused by this waste water were determined. Ra-226 concentrations in soil and plant samples collected on uranium ore dumps were partly much higher than in the environment. (orig.) [de

Natural radionuclides in the environment and problems of uranium mining

International Nuclear Information System (INIS)

Bowie, S.H.U.

1981-01-01

The subject is discussed under the headings: introduction (U-238, U-235, Th-232, K-40, and their decay products); distribution of radionuclides; α, β and γ radiation; uranium in rocks; uranium in soil and water; uranium mining (hazards of uranium and radon during mining and in tailings); assessment of risk. (U.K.)